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Sample records for babcock and wilcox lpr reactor

  1. Babcock and Wilcox advanced PWR development

    International Nuclear Information System (INIS)

    Kulynych, G.E.; Lemon, J.E.

    1986-01-01

    The Babcock and Wilcox 600 MWe PWR design is discussed. Main features of the new B-600 design are improvements in reactor system configuration, glandless coolant pumps, safety features, core design and steam generators

  2. History of research reactor fuel fabrication at Babcock and Wilcox

    International Nuclear Information System (INIS)

    Freim, James B.

    1983-01-01

    B and W Research Reactor Fuel Element facility at Lynchburg, Virginia now produces national laboratory and university fuel assemblies. The Company's 201000 square foot facility is devoted entirely to supplying research fuel and related products. B and W re-entered the research reactor fuel market in 1981

  3. Standard technical specifications for Babcock and Wilcox pressurized water reactors

    International Nuclear Information System (INIS)

    1978-06-01

    The Standard Technical Specification (STS) has been structured for the broadest possible use on B and W NSSS plants currently being reviewed for an Operating License. Two separate and discrete containment specification sections are provided for each of the following containment types: Atmospheric, and Dual. Optional specifications are provided for those features and systems which may be included in individual plant designs but are not generic in their scope of application. Alternate specifications are provided in a limited number of cases to cover situations where alternate specification requirements are necessary on a generic basis because of design differences. The format of the STS addresses the categories required by 10 CFR 50 and consists of six sections covering the areas of: Definitions, Safety Limits and Limiting Safety System Settings, Limiting Conditions for Operation, Surveillance Requirements, Design Features, and Administrative Controls

  4. Standard technical specifications for Babcock and Wilcox pressurized water reactors

    International Nuclear Information System (INIS)

    Virgilio, M.

    1979-07-01

    This Standard Technical Specification (STS) has been structured for the broadest possible use on B and W NSSS plants currently being reviewed for an Operating License. Two separate and discrete containment specification sections are provided for each of the following containment types: Atmospheric and Dual. Optional specifications are provided for those features and systems which may be included in individual plant designs but are not generic in their scope of application. Alternate specifications are provided in a limited number of cases to cover situations where alternate specification requirements are necessary on a generic basis because of design differences. This revision of STS does not typically include requirements which may be added or revised as a result of the NRC staff's further review of the Three Mile Island incident

  5. Standard Technical Specifications, Babcock and Wilcox Plants

    International Nuclear Information System (INIS)

    1992-09-01

    This NUREG contains improved Standard Technical Specifications (STS) for Babcock and Wilcox (B ampersand W) plants and documents the positions of the Nuclear Regulatory Commission (NRC) based on the B ampersand W Owners Group's proposed STS. This document is the result of extensive technical meetings and discussions among the NRC staff, the Nuclear Steam Supply System (NSSS) Owners Groups, the NSSS vendors, and the Nuclear Management and Resources Council. The improved STS were developed based on the criteria in the interim Commission Policy Statement on Technical Specification Improvements for Nuclear Power Reactors, dated February 6, 1987. The improved STS will be used as the basis for developing improved plant-specific technical specifications by individual nuclear power plant licensees. This volume contains sections 3.4--3.9 which cover: Reactor coolant systems, emergency core cooling systems, containment systems, plant systems, electrical power systems, refueling operations

  6. Standard Technical Specifications, Babcock and Wilcox plants

    International Nuclear Information System (INIS)

    1992-09-01

    This NUREG contains improved Standard Technical Specifications (STS) for Babcock and Wilcox (B ampersand W) Plants and documents the positions of the Nuclear Regulatory Commission based on the B ampersand W Owners Group's proposed STS. This document is the result of extensive technical meetings and discussions among the NRC staff, the Nuclear Steam Supply System (NSSS) Owners Groups, the NSSS vendors, and the Nuclear Management and Resources Council (NUMARC). The unproved STS were developed based on the, criteria in the interim Commission Policy Statement on Technical Specification Improvements for Nuclear Power Reactors, dated February 6, 1987. The improved STS will be used as the basis for individual nuclear power plant licensees to develop proved plant-specific technical specifications. This report contains three volumes. This document, Volume 1 contains the Specifications for all chapters and sections of the improved STS

  7. Uranium silicide activities at Babcock and Wilcox

    International Nuclear Information System (INIS)

    Noel, W.W.; Freim, J.B.

    1983-01-01

    Babcock and Wilcox, Naval Nuclear Fuel Division (NNFD) in conjunction with Argonne National Laboratory (ANL) is actively involved in the Reduced Enrichment Research Test Reactor (RERTR) Program to produce low enriched fuel elements for research reactors. B and W and ANL have undertaken a joint effort in which NNFD will fabricate two low enriched uranium (LEU), Oak Ridge Reactor (ORR) elements with uranium silicide fuel furnished by ANL. These elements are being fabricated for irradiation testing at Oak Ridge National Laboratory (ORNL). Concurrently with this program, NNFD is developing and implementing the uranium silicide and uranium aluminide fuel fabrication technology. NNFD is fabricating the uranium silicide ORR elements in a two-phase program, Development and Production. To summarize: 1. Full size fuel plates can be made with U 3 SiAl but the fabricator must prevent oxidation of the compact prior to hot roll bonding; 2. Providing the ANL U 3 Si x irradiation results are successful, NNFD plans to provide two ORR elements during February 1983; 3. NNFD is developing and implementing U 3 Si x and UAI x fuel fabrication technology to be operational in 1983; 4. NNFD can supply U 3 O 8 high enriched uranium (HEU) or low enriched uranium (LEU) research reactor elements; 5. NNFD is capable of providing high quality, cost competitive LEU or HEU research reactor elements to meet the needs of the customer

  8. Production of leu high density fuels at Babcock and Wilcox

    International Nuclear Information System (INIS)

    Freim, J.B.

    1983-01-01

    A large number of fuel elements of all types are produced for both international and domestic customers by Nuclear Fuel Division of Babcock and Wilcox. A brief history of the division, included previous and present research reactor fuel element fabrication experience is discussed. The manufacturing facilities are briefly described. The fabrication of LEU fuels and economic analysis of the production are included. (A.J.)

  9. Status of LEU programs at Babcock and Wilcox

    International Nuclear Information System (INIS)

    McCormick, G.L.

    1995-01-01

    The primary focus of Babcock and Wilcox's (B and W) Research and Test Reactor Fuel Element Facility's (RTRFE) most recent activities is to continue to improve its successful LEU fuel element production process. This is being done by expanding its R and D efforts (expenditures for CFY 92 are twice that of CFY 91) and applying statistical process control to its production processes. B and W's total commitment to quality and integrity has led to the successful fabrication of silicide production elements for five (5) reactors and development/qualification elements for four (4) other reactors. The results of B and W's recent production and development efforts are highlighted in this report. (author)

  10. RELAP5/MOD2 assessment at Babcock and Wilcox

    International Nuclear Information System (INIS)

    Nithianandan, C.K.; Shah, N.H.; Schomaker, R.J.; Turk, C.

    1986-01-01

    Babcock and Wilcox (B and W) has been working with the code developers at EG and G Idaho, Inc. and the NRC assessing the RELAP5/MOD2 computer code by simulating selected separate effects tests. The purpose of this B and W Owners Group-sponsored assessment was to evaluate RELAP5/MOD2 for use in design calculations for the MIST and OTIS integral system tests and in predicting pressurized water reactor (PWR) transients. B and W evaluated various versions of the code and made recommendations to improve code performance. As a result, the currently released version (Cycle 36.1) has been improved considerably over earlier versions. However, further refinements to some of the constitutive models may still be needed to further improve specific predictive capabilities of RELAP5/MOD2

  11. Babcock and Wilcox Canada steam generators past, present and future

    International Nuclear Information System (INIS)

    Smith, J.C.

    1998-01-01

    The steam generators in all of the domestic CANDU Plants, and most of the foreign CANDU plants, were supplied by Babcock and Wilcox Canada, either on their own or in co-operation with local manufacturers. More than 200 steam generators have been supplied. In addition, Babcock and Wilcox Canada has taken the technology which evolved out of the CANDU steam generators and has adapted the technology to supply of replacement steam generators for PWR's. There is enough history and operating experience, plus laboratory experience, to point to the future directions which will be taken in steam generator design. This paper documents the steam generators which have been supplied, the experience in operation and maintenance, what has worked and not worked, and how the design, materials, and operating and maintenance philosophy have evolved. The paper also looks at future requirements in the market, and the continuing research and product development going on at Babcock and Wilcox to address the future steam generator requirements. (author)

  12. Standard technical specifications: Babcock and Wilcox Plants. Revision 1

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-04-01

    This report documents the results of the combined effort of the NRC and the industry to produce improved Standard Technical Specifications (STS), Revision 1 for Babcock & Wilcox Plants. The changes reflected in Revision 1 resulted from the experience gained from license amendment applications to convert to these improved STS or to adopt partial improvements to existing technical specifications. This NUREG is the result of extensive public technical meetings and discussions between the Nuclear Regulatory Commission (NRC) staff and various nuclear power plant licensees, Nuclear Steam Supply System (NSSS) Owners Groups, NSSS vendors, and the Nuclear Energy Institute (NEI). The improved STS were developed based on the criteria in the Final Commission Policy Statement on Technical Specifications Improvements for Nuclear Power Reactors, dated July 22, 1993. The improved STS will be used as the basis for individual nuclear power plant licensees to develop improved plant-specific technical specifications. This report contains three volumes. Volume 1 contains the Specifications for all chapters and sections of the improved STS. Volume 2 contains the Bases for Chapters 2.0 and 3.0, and Sections 3.1--3.3 of the improved STS. Volume 3 contains the Bases for Sections 3.4--3.9 of the improved STS.

  13. Standard technical specifications: Babcock and Wilcox Plants. Revision 1

    International Nuclear Information System (INIS)

    1995-04-01

    This report documents the results of the combined effort of the NRC and the industry to produce improved Standard Technical Specifications (STS), Revision 1 for Babcock ampersand Wilcox Plants. The changes reflected in Revision 1 resulted from the experience gained from license amendment applications to convert to these improved STS or to adopt partial improvements to existing technical specifications. This NUREG is the result of extensive public technical meetings and discussions between the Nuclear Regulatory Commission (NRC) staff and various nuclear power plant licensees, Nuclear Steam Supply System (NSSS) Owners Groups, NSSS vendors, and the Nuclear Energy Institute (NEI). The improved STS were developed based on the criteria in the Final Commission Policy Statement on Technical Specifications Improvements for Nuclear Power Reactors, dated July 22, 1993. The improved STS will be used as the basis for individual nuclear power plant licensees to develop improved plant-specific technical specifications. This report contains three volumes. Volume 1 contains the Specifications for all chapters and sections of the improved STS. Volume 2 contains the Bases for Chapters 2.0 and 3.0, and Sections 3.1--3.3 of the improved STS. Volume 3 contains the Bases for Sections 3.4--3.9 of the improved STS

  14. 75 FR 50009 - Babcock & Wilcox Nuclear Operations Group, Inc.; Establishment of Atomic Safety and Licensing Board

    Science.gov (United States)

    2010-08-16

    ... COMMISSION Babcock & Wilcox Nuclear Operations Group, Inc.; Establishment of Atomic Safety and Licensing... Safety and Licensing Board (Board) is being established to preside over the following proceeding: Babcock & Wilcox Nuclear Operations Group, Inc. (Lynchburg, VA Facility). This proceeding concerns an Order...

  15. Results of a neutron flux perturbation experiment with Babcock and Wilcox Owners Group surveillance capsules

    International Nuclear Information System (INIS)

    Snidow, N.L.; Hassler, L.A.

    1989-01-01

    The Babcock and Wilcox Owners Group (B and WOG) Flux Perturbation Experiment in the Oak Ridge National Laboratory Poolside Facility simulated the thermal shield, downcomer, pressure vessel, and cavity region of a B and W-designed 177-fuel assembly reactor by an arrangement of steel slabs and a void box. Two simulated surveillance capsules located in the downcomer were irradiated as part of the NRC-sponsored Light Water Reactor Pressure Vessel Surveillance Dosimetry Improvement Pregram. The capsules contained extensive dosimetry provided B and W and the Hanford Engineering Development Laboratory (HEDL). Dosimeters were also located outside of the capsules in the downcomer region. Flux distributions were calculated throughout the test configuration using the two-dimensional DOT 4.3 transport theory code. The calculated and measured data are compared in this paper

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

    International Nuclear Information System (INIS)

    1987-11-01

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

  17. Assessment of ISLOCA risk: Methodology and application to a Babcock and Wilcox nuclear power plant

    International Nuclear Information System (INIS)

    Galyean, W.J.; Gertman, D.I.

    1992-04-01

    This report presents information essential to understanding the risk associated with inter-system loss-of-coolant accidents (ISLOCAs). The methodology developed and presented in the report provides a state-of-the-art method for identifying and evaluating plant-specific hardware design, human performance issues, and accident consequence factors to relevant to the prediction of the ISLOCA risk. This ISLOCA methodology was developed and then applied to a Babcock and Wilcox (B ampersand W) nuclear power plants. The results from this application are described in detail. For this particular B ampersand W reference plant, the assessment indicated that the probability of a severe ISLOCA is approximately 2.2E-06/reactor-year. This document Volume 3 provides appendices A--H of the report. Topics are: Historical experience related to ISLOCA events; component failure rates; reference B ampersand W plant system descriptions; reference B ampersand W plant ISLOCA event trees; Human reliability analysis for the B ampersand W ISLOCA probabilistic risk assessment; thermal hydraulic calculations; bounding core uncovery time calculations; and system rupture probability

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

    International Nuclear Information System (INIS)

    Thoma, J.O.

    1985-01-01

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

  19. Superconducting performance of CEBAF/Cornell prototype cavities fabricated by Babcock and Wilcox

    International Nuclear Information System (INIS)

    Bensiek, W.; Dateo, J.; Hager, J.; Pruitt, W.; Williams, P.; Padamsee, H.

    1987-01-01

    Babcock and Wilcox (B and W) is participating in the development of an industrial production capability for CEBAF superconducting rf accelerator cavities. Five-cell elliptical cavities of the Cornell design (operating frequency 1500 MHz) have been fabricated at B and W and tested at the Cornell Laboratory of Nuclear Studies (LNS). Performance specifications (accelerating field of 5 MeV/m at a residual quality factor of 3 x 10 9 ) have been exceeded by comfortable margins in the first two prototypes. A comparison between the performance of cavities fabricated from niobium of different purities is presented

  20. Summary description of the Babcock and Wilcox integrated nuclear design system

    International Nuclear Information System (INIS)

    Wittkopf, W.A.

    1976-03-01

    The Babcock and Wilcox integrated nuclear design system is divided into three broad areas: basic nuclear data processing, applications data processing, and nuclear design calculations. In basic nuclear data processing, basic nuclear data are collected, evaluated, and processed into a specified fine-energy mesh multigroup data file called a Master Library. In applications data processing, data for selected materials are retrieved from the Master Library and processed into an optimally structured, multigroup Production Library. Using these data and input descriptions of cells or regions, neutron spectra are generated and few-group constants are computed and fitted as a function of fuel burnup, initial enrichment, temperature, etc. In nuclear design calculations, few-group cross-section fits and descriptions of each core region and core geometry are input to a diffusion-depletion program or a nodal program that computes core reactivity, core power distribution, control rod worth, fuel cycle studies, core operating limitations, etc

  1. Health hazard evaluation report No. HETA-81-003-980, Babcock and Wilcox Co. , Milwaukee, Wisconsin

    Energy Technology Data Exchange (ETDEWEB)

    Zey, J.N.; Ahrenholz, S.; Klemme, J.C.

    1981-10-01

    On October 1, 1980, the National Institute for Occupational Safety and Health (NIOSH) received a request from the International Brotherhood of Boilermakers Union, Local 1849, for a Health Hazard Evaluation of the Babcock and Wilcox Co., Tubular Products Division, Milwaukee, Wisconsin. The request involved the potential for employee exposure to biocides, dispersant and anti-scaling agents as they are added to four separate circulating water systems which cool four annealing furnaces, two reheat furnaces and one air compressor. NIOSH conducted a combined environmental and medical survey at the Milwaukee facility on November 19-20, 1980. While conducting a walk-through survey on November 19, 1980, NIOSH observed that furnace operators working near cooling systems were potentially exposed to cooling system chemicals. The furnace operators were included in employee monitoring on November 20, 1980. All concentrations obtained were below current environmental criteria. Medical interview data suggested that workers may have been exposed to potentially hazardous levels of DMF in the past.

  2. Seismic risk analysis for the Babcock and Wilcox facility, Leechburg, Pennsylvania

    International Nuclear Information System (INIS)

    1977-01-01

    The results of a detailed seismic risk analysis of the Babcock and Wilcox Plutonium Fuel Fabrication facility at Leechburg, Pennsylvania are presented. This report focuses on earthquakes; the other natural hazards, being addressed in separate reports, are severe weather (strong winds and tornados) and floods. The calculational method used is based on Cornell's work (1968); it has been previously applied to safety evaluations of major projects. The historical seismic record was established after a review of available literature, consultation with operators of local seismic arrays and examination of appropriate seismic data bases. Because of the aseismicity of the region around the site, an analysis different from the conventional closest approach in a tectonic province was adapted. Earthquakes as far from the site as 1,000 km were included, as were the possibility of earthquakes at the site. In addition, various uncertainties in the input were explicitly considered in the analysis. The results of the risk analysis, which include a Bayesian estimate of the uncertainties, are presented, expressed as return period accelerations. The best estimate curve indicates that the Babcock and Wilcox facility will experience 0.05 g every 220 years and 0.10 g every 1400 years. The bounding curves roughly represent the one standard deviation confidence limits about the best estimate, reflecting the uncertainty in certain of the input. Detailed examination of the results show that the accelerations are very insensitive to the details of the source region geometries or the historical earthquake statistics in each region and that each of the source regions contributes almost equally to the cumulative risk at the site. If required for structural analysis, acceleration response spectra for the site can be constructed by scaling the mean response spectrum for alluvium in WASH 1255 by these peak accelerations

  3. Babcock and Wilcox Owners' Group program: Trip reduction and transient response improvement

    International Nuclear Information System (INIS)

    O'Connor, W.T.; Mercado, A.L.; Ganthner, R.W.

    1989-01-01

    In 1985, the average trip frequency for the industry was 4.3 trips per plant per year while Babcock ampersand Wilcox (B ampersand W)-designed plants had 4.5 trips. In early 1986, the B ampersand W Owners' Group (B ampersand WOG) established goals to reduce trip frequency and improve posttrip transient response. Through the recommendations of the B ampersand WOG Trip Reduction and Transient Response Improvement Program (TR/TRIP) and other utility initiatives, the trip frequency for the B ampersand WOG plants has been on a progressive downward trend and has been consistently below the industry average since 1986. The successful results in trip reduction for the B ampersand WOG plants are shown. The B ampersand WOG has implemented several programs that have resulted in fewer trips per plant. This success can be attributed to the following: (1) a comprehensive program to evaluate each trip and transient for root-cause determination, define corrective actions, share information, and peer reviews; (2) a broad program to review systems and components that contribute to trips and transients, identify specific recommendations to correct deficiencies, utility commitment to implementation, conduct internal monitoring and indirectly exert peer pressure; (3) an awareness of the goals at all levels in the organization coupled with strong executive-level involvement; and (4) timely implementation of recommendations

  4. BABCOCK & WILCOX CYCLONE VITRIFICATION TECHNOLOGY FOR CONTAMINATED SOIL

    Science.gov (United States)

    The Babcock & Wilcox 6 million Btu/hr pilot cyclone furnace was successfully used in a 2-yr Superfund Innovative Technology Evaluation (SITE) Emerging Technology project to melt and vitrify an EPA Synthetic Soil Matrix (SSM) spiked with 7,000 ppm lead, 1,000 ppm cadmium, and 1,5...

  5. Generic evaluation of small break loss-of-coolant accident behavior in Babcock and Wilcox designed 177-FA operating plants

    International Nuclear Information System (INIS)

    1980-01-01

    Slow system depressurization resulting from small break loss-of-coolant accidents (LOCAs) in the reactor coolant system have not, until recently, received detailed analytical study comparable to that devoted to large breaks. Following the TMI-2 accident, the staff had a series of meetings with Babcock and Wilcox (B and W) and the B and W licensees. The staff requested that B and W and the licensees: (1) systematically evaluate plant response for small break loss-of-coolant accidents; (2) address each of the concerns documented in the Michelson report; (3) validate the computer codes used against the TMI-2 accident; (4) extend the break spectrum analysis to very small breaks, giving special consideration to failure of pressurizer valves to close; (5) analyze degraded conditions where AFW is not available; (6) prepare design changes aimed at reducing the probability of loss-of-coolant accidents produced by the failure of a PORV to close; and (7) develop revised emergency procedures for small breaks. This report describes the review of the generic analyses performed by B and W based on the requests stated above

  6. Standard technical specifications: Babcock and Wilcox plants. Volume 3, Revision 1: Bases (Sections 3.4--3.9)

    International Nuclear Information System (INIS)

    1995-04-01

    This report documents the results of the combined effort of the NRC and the industry to produce improved Standard Technical Specifications (STS), Revision 1 for Babcock and Wilcox Plants. The changes reflected in Revision 1 resulted from the experience gained from license amendment applications to convert to these improved STS or to adopt partial improvements to existing technical specifications. This NUREG is the result of extensive public technical meetings and discussions between the Nuclear Regulatory Commission (NRC) staff and various nuclear power plant licensees, Nuclear Steam Supply System (NSSS) Owners Groups, NSSS vendors, and the Nuclear Energy Institute (NEI). The improved STS were developed based on the criteria in the Final Commission Policy Statement on Technical Specifications Improvements for Nuclear Power Reactors, dated July 22, 1993. The improved STS will be used as the basis for individual nuclear power plant licensees to develop improved plant-specific technical specifications. This report contains three volumes. Volume 1 contains the Specifications for all chapters and sections of the improved STS. Volume 2 contains the Bases for Chapters 2.0 and 3.0, and Sections 3.1--3.3 of the improved STS. Volume 3 contains the Bases for Sections 3.4--3.9 of the improved STS

  7. Standard technical specifications - Babcock and Wilcox Plants: Bases (Sections 2.0-3.3). Volume 2, Revision 1

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-04-01

    This NUREG contains the improved Standard Technical Specifications (STS) for Babcock and Wilcox (B&W) plants. Revision 1 incorporates the cumulative changes to Revision 0, which was published in September 1992. The changes reflected in Revision 1 resulted from the experience gained from license amendment applications to convert to these improved STS or to adopt partial improvements to existing technical specifications. This NUREG is the result of extensive public technical meetings and discussions between the Nuclear Regulatory Commission (NRC) staff and various nuclear power plant licensees, Nuclear Steam Supply System (NSSS) Owners Groups, specifically the B&W Owners Group (BWOG), NSSS vendors, and the Nuclear Energy Institute (NEI). The improved STS were developed based on the criteria in the Final Commission Policy Statement on Technical Specifications Improvements for Nuclear Power Reactors, dated July 22, 1993 (58 FR 39132). Licensees are encouraged to upgrade their technical specifications consistent with those criteria and conforming, to the extent practical and consistent with the licensing basis for the facility, to Revision 1 to the improved STS. The Commission continues to place the highest priority on requests for complete conversions to the improved STS. Licensees adopting portions of the improved STS to existing technical specifications should adopt all related requirements, as applicable, to achieve a high degree of standardization and consistency.

  8. The water treatment in the dual-purpose nuclear plants of Babcock and Wilcox with straight pipes

    International Nuclear Information System (INIS)

    Martynova, O.I.

    1978-01-01

    A report is given on water processing and water chemistry in the dual-purpose nuclear power plants (as compared to the single-purpose nuclear power plants) of Babcock and Wilcox, with flow steam generators with straight pipes. The most important materials, especially regarding their corrosion resistance, and the water composition during 'hot' start-up of the Okonie-I power plant, the quality factors of the feedwater, the water quality factors of the steam generator with fast start-up and the experience with numerous corrosion-caused defects in steam generator pipes are dealt with from the aspect of optimum water processing and successful continuous operation. (HK) [de

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

    International Nuclear Information System (INIS)

    Thoma, J.O.

    1984-05-01

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

  10. Calculation of particulate dispersion in a design-basis tornadic storm from the Babcock and Wilcox Plant, Leechburg, Pennsylvania

    International Nuclear Information System (INIS)

    Pepper, D.W.

    1978-03-01

    A three-dimensional numerical model is used to calculate ground-level air concentration and deposition (due to precipitation scavenging) after a hypothetical tornado strike at the Babcock and Wilcox Plutonium Fabrication Facility at Leechburg, Pennsylvania. Plutonium particles lss than 20 μm in diameter are assumed to be lifted into the tornadic storm cell by the vortex. The rotational characteristics of the tornadic storm are embedded within the larger mesoscale flow of the storm system. The design-basis translational wind values are based on probabilities associated with existing records of tornado strikes in the vicinity of the plant site. Turbulence exchange coefficients are based on empirical values deduced from experimental data in severe storms and from theoretical assumptions obtained from the literature. The quasi-Lagrangian method of moments is used to model the transport of concentration within a grid cell volume

  11. McMunn, et al. v. Babcock and Wilcox Power Generation Group, Inc., et al.: The long road to dismissal

    International Nuclear Information System (INIS)

    Berger, Marjorie

    2016-01-01

    McMunn, et al. v Babcock and Wilcox Power Generation Group, Inc., et al. was one of 17 related public liability actions filed between 2010 and 2015 by individuals living and/or working in the vicinity of two former fuel fabrication facilities who alleged that releases of radioactive materials from those facilities contaminated the air, soil, surface water and groundwater in the surrounding communities, causing them personal injury and property damage. The plaintiffs in all 17 cases claimed they had contracted various cancers and their property was contaminated with uranium. Plaintiffs brought their claims pursuant to the Price-Anderson Amendments Act (PAA) and the Atomic Energy Act of 1954, as amended (AEA), and also asserted related state law claims of negligence, negligence per se, strict liability, civil conspiracy, and wrongful death and survival. The defendants, Babcock and Wilcox Power Generation Group, Inc., B and W Technical Services, Inc. and Atlantic Richfield Company (ARCO), were unrelated companies who, at different times, owned and operated those facilities. The PAA, which became law on 2 September 1957, is a federal statute that governs claims for personal injury and property damage 'arising from the activities of NRC licensees and DOE contractors'. These claims are defined in the PAA as public liability actions. In order to prevail in a public liability action, plaintiffs must establish through expert evidence that the defendants released radiation into the environment in excess of the limits then permitted by federal regulations and that the plaintiffs were exposed to those releases. They must also establish that their respective exposures to radionuclides were capable of causing their illnesses and that the doses of radiation they received did in fact cause their illnesses

  12. Babcock and Wilcox Safety Anaysis Report (B-SAR-205). Volume 1

    International Nuclear Information System (INIS)

    1976-01-01

    The design of the BW-205 standard reactor with a plant output of 1295 and 1200 MW(e) is described. The reactor is arranged in two closed coolant loops connected in parallel to the reactor vessel, and is controlled by a coordinated combination of chemical shim and mechanical control rods. The coolant serves as a neutron moderator, reflector, and solvent for the soluble boron used in chemical shim reactivity control. The fuel elements consist of slightly enriched UO 2 pellets enclosed in zircaloy tubes

  13. Compact Process Development at Babcock & Wilcox

    Energy Technology Data Exchange (ETDEWEB)

    Eric Shaber; Jeffrey Phillips

    2012-03-01

    Multiple process approaches have been used historically to manufacture cylindrical nuclear fuel compacts. Scale-up of fuel compacting was required for the Next Generation Nuclear Plant (NGNP) project to achieve an economically viable automated production process capable of providing a minimum of 10 compacts/minute with high production yields. In addition, the scale-up effort was required to achieve matrix density equivalent to baseline historical production processes, and allow compacting at fuel packing fractions up to 46% by volume. The scale-up approach of jet milling, fluid-bed overcoating, and hot-press compacting adopted in the U.S. Advanced Gas Reactor (AGR) Fuel Development Program involves significant paradigm shifts to capitalize on distinct advantages in simplicity, yield, and elimination of mixed waste. A series of compaction trials have been completed to optimize compaction conditions of time, temperature, and forming pressure using natural uranium oxycarbide (NUCO) fuel at packing fractions exceeding 46% by volume. Results from these trials are included. The scale-up effort is nearing completion with the process installed and operable using nuclear fuel materials. Final process testing is in progress to certify the process for manufacture of qualification test fuel compacts in 2012.

  14. Babcock and Wilcox BR-100 100-ton rail/barge spent fuel shipping cask

    Energy Technology Data Exchange (ETDEWEB)

    None

    1990-02-01

    This Preliminary Design Report (PDR) provides a detailed description of the design, analyses, and testing programs for the BR-100 cask. The BR-100 is a Type B(U) cask designed for transport by rail or barge. This report presents the preliminary analyses and tests which have been performed for the BR-100 and outlines the confirmatory analyses and tests which will be performed.

  15. Babcock and Wilcox BR-100 100-ton rail/barge spent fuel shipping cask

    International Nuclear Information System (INIS)

    1990-02-01

    This Preliminary Design Report (PDR) provides a detailed description of the design, analyses, and testing programs for the BR-100 cask. The BR-100 is a Type B(U) cask designed for transport by rail or barge. This report presents the preliminary analyses and tests which have been performed for the BR-100 and outlines the confirmatory analyses and tests which will be performed

  16. Environmental consequences of postulated plutonium releases from the Babcock and Wilcox Plant, Leechburg, Pennsylvania, as a result of severe natural phenomena

    International Nuclear Information System (INIS)

    McPherson, R.B.; Watson, E.C.

    1979-03-01

    Potential environmental consequences in terms of radiation dose to people are presented for postulated plutonium releases accidently caused by severe natural phenomena at the Babcock and Wilcox plant, Leechburg, Pennsylvania. The severe natural phenomena considered are earthquakes, tornadoes, high straight-line winds, and floods. Maximum plutonium deposition values are given for significant locations around the site. All important potential exposure pathways are examined. The most likely 50-year committed dose equivalents are given for the maximum-exposed individual and the population within a 50-mile radius of the plant. The maximum plutonium deposition values most likely to occur at the site boundary are also given. The most likely calculated 50-year collective committed dose equivalents are all much lower than the collective dose equivalent expected from 50 years of exposure to natural background radiation and medical x-rays. The most likely maximum residual plutonium contamination estimated to be deposited at the site boundary following Earthquake No. 3, the 110-mph and 130-mph winds, and the 130 mph tornado are above the Environmental Protection Agency's (EPA) proposed guideline for plutonium in the general environment of 0.2 μCi/m 2

  17. Preliminary design report: Babcock and Wilcox BR-100 100-ton rail/barge spent fuel shipping cask

    International Nuclear Information System (INIS)

    1990-02-01

    The purpose of this document is to provide information on burnup credit as applied to the preliminary design of the BR-100 shipping cask. There is a brief description of the preliminary basket design and the features used to maintain a critically safe system. Following the basket description is a discussion of various criticality analyses used to evaluate burnup credit. The results from these analyses are then reviewed in the perspective of fuel burnups expected to be shipped to either the final repository or a Monitored Retrievable Storage (MRS) facility. The hurdles to employing burnup credit in the certification of any cask are then outlines and reviewed. the last section gives conclusions reached as to burnup credit for the BR-100 cask, based on our analyses and experience. All information in this study refers to the cask configured to transport PWR fuel. Boiling Water Reactor (BWR) fuel satisfies the criticality requirements so that burnup credit is not needed. All calculations generated in the preparation of this report were based upon the preliminary design which will be optimized during the final design. 8 refs., 19 figs., 16 tabs

  18. An Analysis of the Corporate Merger between the Babcock & Wilcox Co. and J. Ray Mcdermott & Co., Inc.

    Science.gov (United States)

    1980-09-01

    United States’ attempt to lessen their dependence on imported oil. The world’s tallest and heaviest steel platform, the Cognac platform, was recently...field of the Gulf in 1981. This jacket’s height will be second only to Cognac . 2. Middle East The Middle East’s operations have been active, but not

  19. Babcock and Wilcox revisions to CONTEMPT, computer program for predicting containment pressure-temperature response to a loss-of-coolant accident

    International Nuclear Information System (INIS)

    Hsii, Y.H.

    1975-01-01

    The CONTEMPT computer program predicts the pressure-temperature response of a single-volume reactor building to a loss-of-coolant accident. The analytical model used for the program is described. CONTEMPT assumes that the loss-of-coolant accident can be separated into two phases; the primary system blowdown and reactor building pressurization. The results of the blowdown analysis serve as the boundary conditions and are input to the CONTEMPT program. Thus, the containment model is only concerned with the pressure and temperature in the reactor building and the temperature distribution through the reactor building structures. The program also calculates building leakage and the effects of engineered safety features such as reactor building sprays, decay heat coolers, sump coolers, etc. 11 references. (U.S.)

  20. In core measurement and monitoring of reactor (neutron) radiation field

    International Nuclear Information System (INIS)

    Erben, O.

    1985-01-01

    A survey is presented of in core radiation detectors. The principles are described of activation detectors, fission chambers, self-powered neutron detectors and thermal sensors. Systems of in core measurement for WWER nuclear power plants, nuclear reactors of power plants operated by KWU, Babcock and Wilcox, Combustion Engineering and FRAMATOME are described. (E.S.)

  1. Report of the Bulletins and Orders Task Force. Volume II. Appendices

    International Nuclear Information System (INIS)

    1980-01-01

    Appendices include: Office of Inspection and Enforcement bulletins; NRR status report on feedwater transients in BWR plants; orders on Babcock and Wilcox Company plants; letters lifting orders; letters issuing auxiliary feedwater system requirements; letter to licensees of all operating reactors, dated October 30, 1979 concerning short-term lessons learned requirements; and letters approving guidelines for preparation of small-break LOCA operating procedures

  2. Potentiometric surfaces and water-level trends in the Cockfield (upper Claiborne) aquifer in southern Arkansas and the Wilcox (lower Wilcox) aquifer of northeastern and southern Arkansas, 2012

    Science.gov (United States)

    Rodgers, Kirk D.

    2015-01-01

    The Cockfield aquifer, located in southern Arkansas, is composed of Eocene-age sand beds found near the base of the Cockfield Formation of Claiborne Group. The Wilcox aquifer, located in northeastern and southern Arkansas, is composed of Paleocene-age sand beds found in the middle to lower part of the Wilcox Group. The Cockfield and Wilcox aquifers are primary sources of groundwater. In 2010, withdrawals from the Cockfield aquifer in Arkansas totaled 19.2 million gallons per day (Mgal/d), and withdrawals from the Wilcox aquifer totaled 36.5 Mgal/d.

  3. The preliminary processing and analysis of LPR Channel-2B data from Chang'E-3

    Science.gov (United States)

    Zhao, Na; Zhu, PeiMin; Yang, KeSi; Yuan, YueFeng; Guo, ShiLi

    2014-12-01

    The Lunar Penetrating Radar (LPR) carried by Chang'E-3 has imaged the shallow subsurface of the landing site at the northern Mare Imbrium. The antenna B of the Channel-2 onboard the LPR (LPR Channel-2B) has collected more than 2000 traces of usable raw data. Because of the low resolution and noise of the raw data, only a few shallow geological structures are visible. To improve the resolution and the signal-to-noise ratio of the LPR data, we processed the LPR data including amplitude compensation, filtering, and deconvolution processes. The processing results reveal that the data processing in this study not only improves the signal-to-noise ratio of the LPR Channel-2B data but also makes the geological structures vivid. The processing results will lay the foundation for the subsequent geological interpretation and physical property inversion of lunar materials.

  4. Reactor protection system design using micro-computers

    International Nuclear Information System (INIS)

    Fairbrother, D.B.

    1976-01-01

    Reactor protection systems for nuclear power plants have traditionally been built using analog hardware. This hardware works quite well for single parameter trip functions; however, optimum protection against DNBR and KW/ft limits requires more complex trip functions than can easily be handled with analog hardware. For this reason, Babcock and Wilcox has introduced a Reactor Protection System, called the RPS-II, that utilizes a micro-computer to handle the more complex trip functions. The paper describes the design of the RPS-II and the operation of the micro-computer within the Reactor Protection System

  5. Reactor protection system design using micro-computers

    International Nuclear Information System (INIS)

    Fairbrother, D.B.

    1977-01-01

    Reactor Protection Systems for Nuclear Power Plants have traditionally been built using analog hardware. This hardware works quite well for single parameter trip functions; however, optimum protection against DNBR and KW/ft limits requires more complex trip functions than can easily be handled with analog hardware. For this reason, Babcock and Wilcox has introduced a Reactor Protection System, called the RPS-II, that utilizes a micro-computer to handle the more complex trip functions. This paper describes the design of the RPS-II and the operation of the micro-computer within the Reactor Protection System

  6. Accident at the Three Mile Island Nuclear Powerplant. Part 1. Oversight hearings before a task force of the Subcommittee on Energy and the Environment of the Committee on Interior and Insular Affairs, House of Representatives, Ninety-Sixth Congress

    International Nuclear Information System (INIS)

    Anon.

    1979-01-01

    The Committee on Interior and Insular Affairs conducted an informal review of the accident beginning on March 28, 1979 at the Three Mile Island Nuclear Power Plant. Officials of the Nuclear Regulatory Commission, plant operating personnel employed by General Public Utilities, and representatives of the reactor manufacturer, Babcock and Wilcox Company, related their activities during the accident and their analyses of the sequence of events

  7. Howard Wilcox Haggard and the Institutionalization of Modern Alcohol Studies.

    Science.gov (United States)

    Allred, Nicholas; Bejarano, William; Ward, Judit

    2017-03-01

    This biographical sketch and accompanying bibliography provide a new look at Howard Wilcox Haggard, M.D., Ph.D., whose career highlights the consolidation of alcohol studies as a field in twentieth-century America. The article relies in large part on the works of Haggard assembled for the bibliography project, supplemented by published and unpublished documents and records from collections at Rutgers University. Haggard began his career in respiratory physiology, influenced by his work on chemical weapons for the Army during the First World War. As his reputation grew, he moved into anesthesiology and supplemented his research with textbooks and popular science bestsellers. Haggard moved into the burgeoning field of alcohol studies after the repeal of National Prohibition and, in 1940, became the inaugural editor and president of the corporation of the Quarterly Journal of Studies on Alcohol, now the Journal of Studies on Alcohol and Drugs. Under the aegis of the Yale Laboratory for Applied Physiology, he also assembled and oversaw what would become the Center of Alcohol Studies. Haggard died in 1959, his legacy established as a central figure in the 20th-century transformation of alcohol studies in the United States. A prolific researcher with a talent for tapping into the public zeitgeist, Haggard helped provide the institutional infrastructure, academic credibility, and broad audience that made the renaissance of alcohol studies in post-Prohibition America possible.

  8. Selection, training, qualification and licensing of Three Mile Island reactor operating personnel

    International Nuclear Information System (INIS)

    Eytchison, R.M.

    1980-01-01

    The various programs which were intended to staff Three Mile Island with competent, trained operators and supervisors are reviewed. The analysis includes a review of the regulations concerning operator training and licensing, and describes how the requirements were implemented by the NRC, Metropolitan Edison Company, and Babcock and Wilcox Company. Finally the programs conducted by these three organisations are evaluated. (U.K.)

  9. Defects in the peripheral taste structure and function in the MRL/lpr mouse model of autoimmune disease.

    Directory of Open Access Journals (Sweden)

    Agnes Kim

    Full Text Available While our understanding of the molecular and cellular aspects of taste reception and signaling continues to improve, the aberrations in these processes that lead to taste dysfunction remain largely unexplored. Abnormalities in taste can develop in a variety of diseases, including infections and autoimmune disorders. In this study, we used a mouse model of autoimmune disease to investigate the underlying mechanisms of taste disorders. MRL/MpJ-Fas(lpr/J (MRL/lpr mice develop a systemic autoimmunity with phenotypic similarities to human systemic lupus erythematosus and Sjögren's syndrome. Our results show that the taste tissues of MRL/lpr mice exhibit characteristics of inflammation, including infiltration of T lymphocytes and elevated levels of some inflammatory cytokines. Histological studies reveal that the taste buds of MRL/lpr mice are smaller than those of wild-type congenic control (MRL/+/+ mice. 5-Bromo-2'-deoxyuridine (BrdU pulse-chase experiments show that fewer BrdU-labeled cells enter the taste buds of MRL/lpr mice, suggesting an inhibition of taste cell renewal. Real-time RT-PCR analyses show that mRNA levels of several type II taste cell markers are lower in MRL/lpr mice. Immunohistochemical analyses confirm a significant reduction in the number of gustducin-positive taste receptor cells in the taste buds of MRL/lpr mice. Furthermore, MRL/lpr mice exhibit reduced gustatory nerve responses to the bitter compound quinine and the sweet compound saccharin and reduced behavioral responses to bitter, sweet, and umami taste substances compared with controls. In contrast, their responses to salty and sour compounds are comparable to those of control mice in both nerve recording and behavioral experiments. Together, our results suggest that type II taste receptor cells, which are essential for bitter, sweet, and umami taste reception and signaling, are selectively affected in MRL/lpr mice, a model for autoimmune disease with chronic

  10. Hydrogeologic characteristics and water levels of Wilcox aquifer in southwestern and northeastern Arkansas

    Science.gov (United States)

    Pugh, Aaron L.; Schrader, Tony P.

    2009-01-01

    The Wilcox Group of Eocene and Paleocene age is located throughout most of southern and eastern Arkansas. The Wilcox Group in southern Arkansas is undifferentiated, while in northeastern Arkansas, the Wilcox Group is subdivided into three units: Flour Island, Fort Pillow Sand, and Old Breastworks Formation. The Wilcox Group crops out in southwestern Arkansas in discontinuous, 1 to 3 mi wide bands. In northeastern Arkansas, the Wilcox Group crops out along a narrow, discontinuous, band along the western edge of Crowleys Ridge. The Wilcox aquifer provides sources of groundwater in southwestern and northeastern Arkansas. In 2005, reported withdrawals from the Wilcox aquifer in Arkansas totaled 27.0 million gallons per day, most of which came from the northeastern area. Major withdrawals from the aquifer were for public supplies with lesser but locally important withdrawals for commercial, domestic, and industrial uses. A study was conducted by the U.S. Geological Survey in cooperation with the Arkansas Natural Resources Commission and the Arkansas Geological Survey to determine the water levels associated with the Wilcox aquifer in southwestern and northeastern Arkansas. During February 2009, 58 water-level measurements were made in wells completed in the Wilcox aquifer. The results from this study and previous studies are presented as potentiometric-surface maps, water-level difference maps, and long-term hydrographs. The direction of groundwater flow in the southwestern area is affected by two potentiometric-surface mounds, one in the north and the other in the southwest, and a cone of depression in the center. The direction of water flowing off of the northern mound of water is generally to the south and east with some to the north. The direction of water flowing off of the southwestern mound is generally to the south and east. The direction of water flowing into the cone of depression is generally from the north, west, and south. The direction of groundwater flow

  11. Spent fuel working group report on inventory and storage of the Department's spent nuclear fuel and other reactor irradiated nuclear materials and their environmental, safety and health vulnerabilities

    International Nuclear Information System (INIS)

    1993-11-01

    In a memo dated 19 August 1993, Secretary O'Leary assigned the Office of Environment, Safety and Health the primary responsibility to identify, characterize, and assess the safety, health, and environmental vulnerabilities of the DOE's existing storage conditions and facilities for the storage of irradiated reactor fuel and other reactor irradiated nuclear materials. This volume is divided into three major sections. Section 1 contains the Working Group Assessment Team reports on the following facilities: Hanford Site, INEL, SRS, Oak Ridge Site, West Valley Site, LANL, BNL, Sandia, General Atomics (San Diego), Babcock ampersand Wilcox (Lynchburg Technology Center), and ANL. Section 2 contains the Vulnerability Development Forms from most of these sites. Section 3 contains the documents used by the Working Group in implementing this initiative

  12. What is LPR (Laryngopharyngeal Reflux)

    Science.gov (United States)

    ... press registration and more below. Please direct any interview requests or policy questions to our media and public relations staff at newsroom@entnet.org . Insight into the diagnosis, prevention, and treatment of laryngopharyngeal reflux (LPR) What ...

  13. Susceptible cytotoxicity to ultraviolet B light in fibroblasts and keratinocytes cultured from autoimmune-prone MRL/Mp-lpr/lpr mice

    Energy Technology Data Exchange (ETDEWEB)

    Furukawa, F.; Lyon, M.B.; Norris, D.A. (Univ. of Colorado School of Medicine, Denver (USA))

    1989-09-01

    The MRL/Mp-lpr/lpr (MRL/l) mouse is an autoimmune model of spontaneous lupus erythematosus (LE), in addition to lupus nephritis. In order to better understand the mechanisms of photosensitivity in LE, in vitro photocytotoxicity was examined by using fibroblasts and keratinocytes cultured from MRL/l mice, control MRL/Mp- +/+ (MRL/n) mice, and normal BALB/c mice. A colorimetric 3-(4,5-dimethylthiazol-2-yl)-2,5-diphenyl tetrazolium bromide assay and the acridine orange/ethidium bromide assay were used for determination of cytotoxicity. Fibroblasts cultured from newborn MRL/l mice showed higher susceptibility to single ultraviolet light B (UVB) light irradiation at a dose of 100-500 mJ than those from MRL/n, F1 hybrid of (MRL/l x MRL/n mice), and BALB/c mice. However, the susceptibility to UVB was not observed in young (1-month-old) and adult (4-month-old) MRL/l mice. UVA light irradiation was not cytotoxic. Keratinocytes cultured from MRL mice showed lower cytotoxicity to UVB irradiation than fibroblasts cultured. However, keratinocytes from newborn MRL/l mice showed higher cytotoxicity to 50 mJ UVB irradiation than cells from MRL/n mice. Syngeneic or allogeneic sera augmented UVB-induced cytotoxicity of fibroblasts cultured. UVB irradiation of spleen cells induced no significant difference of cytotoxicity between MRL/l and MRL/n mice. Based on the results of F1 hybrid of (MRL/l x MRL/n) mice, the susceptibility seemed to be associated with autoimmune traits and to be regulated by genetical background.

  14. Evaluation of B&W UO2/ThO2 VIII experimental core: criticality and thermal disadvantage factor analysis

    Energy Technology Data Exchange (ETDEWEB)

    Carlo Parisi; Emanuele Negrenti

    2017-02-01

    In the framework of the OECD/NEA International Reactor Physics Experiment (IRPHE) Project, an evaluation of core VIII of the Babcock & Wilcox (B&W) Spectral Shift Control Reactor (SSCR) critical experiment program was performed. The SSCR concept, moderated and cooled by a variable mixture of heavy and light water, envisaged changing of the thermal neutron spectrum during the operation to encourage breeding and to sustain the core criticality. Core VIII contained 2188 fuel rods with 93% enriched UO2-ThO2 fuel in a moderator mixture of heavy and light water. The criticality experiment and measurements of the thermal disadvantage factor were evaluated.

  15. xLPR - a probabilistic approach to piping integrity analysis

    International Nuclear Information System (INIS)

    Harrington, C.; Rudland, D.; Fyfitch, S.

    2015-01-01

    The xLPR Code is a probabilistic fracture mechanics (PFM) computational tool that can be used to quantitatively determine a best-estimate probability of failure with well characterized uncertainties for reactor coolant system components, beginning with the piping systems and including the effects of relevant active degradation mechanisms. The initial application planned for xLPR is somewhat narrowly focused on validating LBB (leak-before-break) compliance in PWSCC-susceptible systems such as coolant systems of PWRs. The xLPR code incorporates a set of deterministic models that represent the full range of physical phenomena necessary to evaluate both fatigue and PWSCC degradation modes from crack initiation through failure. These models are each implemented in a modular form and linked together by a probabilistic framework that contains the logic for xLPR execution, exercises the individual modules as required, and performs necessary administrative and bookkeeping functions. The completion of the first production version of the xLPR code in a fully documented, releasable condition is presently planned for spring 2015

  16. Description of project for pretreatment and storage of wastes of L.P.R. (Radiochemical Processes Laboratory)

    International Nuclear Information System (INIS)

    Doval, J.C.F.; Mehlich, A.M.; Quilici, D.F.

    1987-01-01

    The aim of the project is to allow the start up and operation of LPR (Radiochemical Processes Laboratory) as part of the intended activities in the plant. In this paper, the pretreatment and storage of liquid wastes generated at the LPR are described. The pretreatment section will be set up inside the shielded cells already existent in the LPR, where a previous concentration through the evaporation of liquid wastes will take place. The storage section has to be constructed on purpose in order to temporarily store the concentrates. The cells of transference and preconditioning of solid wastes are also described. These cells will be mounted inside the building, allowing the handling of radioactive solids generated as effluents during the reprocessing plan. In the description, the use of non conventional materials for the boiler making and the construction of cells is specially mentioned. (Author)

  17. xLPR Scenario Analysis Report.

    Energy Technology Data Exchange (ETDEWEB)

    Eckert-Gallup, Aubrey Celia [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Lewis, John R. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Brooks, Dusty Marie [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Martin, Nevin [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Hund, Lauren [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Clark, Andrew Jordan [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Mariner, Paul [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2017-03-01

    This report describes the methods, results, and conclusions of the analysis of 11 scenarios defined to exercise various options available in the xLPR (Extremely Low Probability of Rupture) Version 2 .0 code. The scope of the scenario analysis is three - fold: (i) exercise the various options and components comprising xLPR v2.0 and defining each scenario; (ii) develop and exercise methods for analyzing and interpreting xLPR v2.0 outputs ; and (iii) exercise the various sampling options available in xLPR v2.0. The simulation workflow template developed during the course of this effort helps to form a basis for the application of the xLPR code to problems with similar inputs and probabilistic requirements and address in a systematic manner the three points covered by the scope.

  18. Using FARIS [Fuel Assembly Repair and Inspection Station] for assembly clean-up and debris removal

    International Nuclear Information System (INIS)

    Tucker, J.S.; Sapyta, J.J.

    1990-01-01

    Because fuel inspection and repair tasks are commonly done on the critical path during plant refuelling outages, they must be completed quickly and efficiently with minimal costs. To fulfil these demands, the Babcock and Wilcox Fuel Company has designed a Fuel Assembly Repair and Inspection Station (FARIS) for fuel assembly clean-up and debris removal in Pressurized Water Reactors. The system is portable and can also be used for carrying out visual inspections on fuel assemblies, spacer grid repair, fuel rod oxide thickness measurements and for fuel rod water channel inspections. (author)

  19. Pressurized-water-reactor station blackout

    International Nuclear Information System (INIS)

    Dobbe, C.A.

    1983-01-01

    The purpose of the Severe Accident Sequence Analysis (SASA) Program was to investigate accident scenarios beyond the design basis. The primary objective of SASA was to analyze nuclear plant transients that could lead to partial or total core melt and evaluate potential mitigating actions. The following summarizes the pressurized water reactor (PWR) SASA effort at the Idaho National Engineering Laboratory (INEL). The INEL is presently evaluating Unresolved Safety Issue A-44 - Station Blackout from initiation of the transient to core uncovery. The balance of the analysis from core uncovery until fission product release is being performed at Sandia National Laboratory (SNL). The current analyses involve the Bellefonte Nuclear Steam Supply System (NSSS), a Babcock and Wilcox (B and W) 205 Fuel Assembly (205-FA) raised loop design to be operated by the Tennessee Valley Authority

  20. Nuclear plant-aging research on reactor protection systems

    International Nuclear Information System (INIS)

    Meyer, L.C.

    1988-01-01

    This report presents the rsults of a review of the Reactor Trip System (RTS) and the Engineered Safety Feature Actuating System (ESFAS) operating experiences reported in Licensee Event Reports (LER)s, the Nuclear Power Experience data base, Nuclear Plant Reliability Data System, and plant maintenance records. Our purpose is to evaluate the potential significance of aging, including cycling, trips, and testing as contributors to degradation of the RTS and ESFAS. Tables are presented that show the percentage of events for RTS and ESFAS classified by cause, components, and subcomponents for each of the Nuclear Steam Supply System vendors. A representative Babcock and Wilcox plant was selected for detailed study. The US Nuclear Regulatory Commission's Nuclear Plant Aging Research guidelines were followed in performing the detailed study that identified materials susceptible to aging, stressors, environmental factors, and failure modes for the RTS and ESFAS as generic instrumentation and control systems. Functional indicators of degradation are listed, testing requirements evaluated, and regulatory issues discussed

  1. Parameters and structure of lunar regolith in Chang'E-3 landing area from lunar penetrating radar (LPR) data

    Science.gov (United States)

    Dong, Zehua; Fang, Guangyou; Ji, Yicai; Gao, Yunze; Wu, Chao; Zhang, Xiaojuan

    2017-01-01

    Chang'E-3 (CE-3) landed in the northwest Mare Imbrium, a region that has not been explored before. Yutu rover that released by CE-3 lander carried the first lunar surface penetrating radar (LPR) for exploring lunar regolith thickness and subsurface shallow geological structures. In this paper, based on the LPR data and the Panoramic Camera (PC) data, we first calculate the lunar surface regolith parameters in CE-3 landing area including its permittivity, density, conductivity and FeO + TiO2 content. LPR data provides a higher spatial resolution and more accuracy for the lunar regolith parameters comparing to other remote sensing techniques, such as orbit radar sounder and microwave sensing or earth-based powerful radar. We also derived the regolith thickness and its weathered rate with much better accuracy in the landing area. The results indicate that the regolith growth rate is much faster than previous estimation, the regolith parameters are not uniform even in such a small study area and the thickness and growth rate of lunar regolith here are different from other areas in Mare Imbrium. We infer that the main reason should be geological deformation that caused by multiple impacts of meteorites in different sizes.

  2. Selective elimination of non-lpr lymphoid cells in mice undergoing lpr-mediated graft-vs-host disease

    International Nuclear Information System (INIS)

    Perkins, D.L.; Michaelson, J.; Glaser, R.M.; Marshak-Rothstein, A.

    1987-01-01

    The transfer of lpr BM stem cells into lethally irradiated non-lpr recipients (including the congenic MRL/+ differing only at the lpr locus) causes GVHD characterized by a wasting syndrome. In this study we investigated the interaction between the autoimmune (lpr) and normal (A-Thy) B, T, and RBC cell lineages in two types of radiation chimeras: MRL/lpr plus A-Thy----(MRL/lpr X A-Thy)F1 and MRL/+ plus A-Thy----(MRL/lpr X A-Thy)F1. Analysis of B cell repopulation by competitive RIA of serum Igh-1 allotype showed that both the MRL and the A-Thy donor cells initially engrafted. However, by 2 to 4 mo post-transplantation the normal A-Thy allotype was barely detectable (reduced greater than 2 orders of magnitude), whereas the autoimmune MRL/lpr allotype persisted at normal levels. Similarly, investigation of the donor origin of peripheral blood T cells by two-color flow cytometry showed that by 8 mo post-transplantation normal A-Thy T cells had been eliminated and only MRL/lpr T cells were present in the circulation. In contrast, erythrocytes from both the MRL/lpr and A-Thy donor strains successfully engrafted the F1 recipients and persisted until the termination of the study. Control chimeras transplanted with a mixture of MRL/+ plus A-Thy BM were stably engrafted with both donor strains in both the erythroid and lymphoid populations. Additional experiments in which either B6/lpr or MRL/lpr (and B6/+ or MRL/+ control) BM cells were transferred into (MRL/lpr X B6/+)F1 and (MRL/lpr X B6/lpr)F1 recipients demonstrated that the development of GVHD was not simply due to increased alloreactivity by the lpr donor cells. In these chimeras only the recipients heterozygous (but not homozygous) for the lpr gene developed lpr-GVHD, although both types of recipients had identical genotypes except at the lpr locus

  3. In-vessel inspection before head removal: TMI II, Phase III (tooling and systems design and verification)

    International Nuclear Information System (INIS)

    Carter, G.S.; Ryan, R.F.; Pieleck, A.W.; Bibb, H.Q.

    1982-09-01

    Under EG and G contract K-9003 to General Public Utilities Corporation, a Task Order was assigned to Babcock and Wilcox to develop and provide equipment to facilitate early assessment of core damage in the Three Mile Island Unit 2 reactor vessel head. Described is the work performed, the equipment developed, and the tests conducted with this equipment on various mockups used to simulate the constraints inside and outside the reactor vessel that affect the performance of the inspection. The tooling developed provides several methods of removing a few control rod drive leadscrews from the reactor, thereby providing paths into which cameras and lights may be inserted to permit video viewing of many potentially damaged areas in the reactor vessel. The tools, equipment, and cameras demonstrated that these tasks could be accomplished

  4. Spent fuel working group report on inventory and storage of the Department's spent nuclear fuel and other reactor irradiated nuclear materials and their environmental, safety and health vulnerabilities

    International Nuclear Information System (INIS)

    1993-11-01

    Each Site Team, consisting of M ampersand O contractor and Operations Office personnel, performed data collection and identified ES ampersand H concerns relative to RINM storage by preparing responses to the detailed question set for each storage facility at the site. These responses formed the basis for the Site Team reports. These reports are contained in this volume and are from the following facilities: Hanford Site, Idaho National Engineering Laboratory Site, Savannah River Site, Oak Ridge Site, West Valley Demonstration Project Site, Los Alamos National Laboratory, Brookhaven National Laboratory, Sandia National Laboratories, General Atomics, San Diego, Babcock ampersand Wilcox, Lynchburg Technical Center, Argonne National Laboratory - East, Naval Reactors Facilities, Rocky Flats Critical Mass Laboratory, EG ampersand G Mound Applied Technologies, Ohio, Lawrence Berkeley Laboratory, and Battelle Columbus Laboratory. This volume also contains information received from the sites that were not visited. These sites include the Naval Reactor Facility at the INEL, EG ampersand G Mound Applied Technologies, The Catholic University of America, Rocky Flats Site, Lawrence Livermore National Laboratory, Stanford Linear Accelerator Laboratory, Energy Technology Engineering Center, and Lawrence Berkeley Laboratory. Information received through the Chicago Operations Office for University Reactors, Massachusetts Institute of Technology, and Battelle Columbus Laboratory is also included. Materials contained in this volume consist of information, data and site documents. They are unedited

  5. U.S. Nuclear Regulatory Commission Extremely Low Probability of Rupture pilot study : xLPR framework model user's guide.

    Energy Technology Data Exchange (ETDEWEB)

    Kalinich, Donald A.; Sallaberry, Cedric M.; Mattie, Patrick D.

    2010-12-01

    For the U.S. Nuclear Regulatory Commission (NRC) Extremely Low Probability of Rupture (xLPR) pilot study, Sandia National Laboratories (SNL) was tasked to develop and evaluate a probabilistic framework using a commercial software package for Version 1.0 of the xLPR Code. Version 1.0 of the xLPR code is focused assessing the probability of rupture due to primary water stress corrosion cracking in dissimilar metal welds in pressurizer surge nozzles. Future versions of this framework will expand the capabilities to other cracking mechanisms, and other piping systems for both pressurized water reactors and boiling water reactors. The goal of the pilot study project is to plan the xLPR framework transition from Version 1.0 to Version 2.0; hence the initial Version 1.0 framework and code development will be used to define the requirements for Version 2.0. The software documented in this report has been developed and tested solely for this purpose. This framework and demonstration problem will be used to evaluate the commercial software's capabilities and applicability for use in creating the final version of the xLPR framework. This report details the design, system requirements, and the steps necessary to use the commercial-code based xLPR framework developed by SNL.

  6. Development, analysis, and evaluation of a commercial software framework for the study of Extremely Low Probability of Rupture (xLPR) events at nuclear power plants.

    Energy Technology Data Exchange (ETDEWEB)

    Kalinich, Donald A.; Helton, Jon Craig; Sallaberry, Cedric M.; Mattie, Patrick D.

    2010-12-01

    Sandia National Laboratories (SNL) participated in a Pilot Study to examine the process and requirements to create a software system to assess the extremely low probability of pipe rupture (xLPR) in nuclear power plants. This project was tasked to develop a prototype xLPR model leveraging existing fracture mechanics models and codes coupled with a commercial software framework to determine the framework, model, and architecture requirements appropriate for building a modular-based code. The xLPR pilot study was conducted to demonstrate the feasibility of the proposed developmental process and framework for a probabilistic code to address degradation mechanisms in piping system safety assessments. The pilot study includes a demonstration problem to assess the probability of rupture of DM pressurizer surge nozzle welds degraded by primary water stress-corrosion cracking (PWSCC). The pilot study was designed to define and develop the framework and model; then construct a prototype software system based on the proposed model. The second phase of the project will be a longer term program and code development effort focusing on the generic, primary piping integrity issues (xLPR code). The results and recommendations presented in this report will be used to help the U.S. Nuclear Regulatory Commission (NRC) define the requirements for the longer term program.

  7. U.S. regulatory requirements for nuclear plant license renewal: The B and W Owners Group License Renewal Program

    International Nuclear Information System (INIS)

    Staudinger, Deborah K.

    2004-01-01

    This paper discusses the current U.S. Regulatory Requirements for License Renewal and describes the Babcock and Wilcox Owners Group (B and WOG) Generic License Renewal Program (GLRP). The B and W owners, recognizing the need to obtain the maximum life for their nuclear generating units, embarked on a program to renew the licenses of the seven reactors in accordance with the requirements of the Atomic Energy Act of 1954 and further defined by Title 10 of the Code of Federal Regulation Part 54 (10 CFR 54). These reactors, owned by five separate utilities, are Pressurized Water Reactors (PWR) ranging in net rated capacity from approximately 800 to 900 MW. The plants, predominately constructed in the 70s, have USNRC Operating Licenses that expire between 2013 to 2017. (author)

  8. A reexamination of the North American Crepis agamic complex and comparison with the findings of Babcock and Stebbins' classic biosystematic monograph.

    Science.gov (United States)

    Sears, Christopher J; Whitton, Jeannette

    2016-07-01

    Babcock and Stebbins coined the term agamic complex in their 1938 monograph of the North American Crepis agamic complex. Despite the historical role that this complex holds in the evolutionary literature, it has not been reexamined in over 75 years. We present a thorough reevaluation of the complex to test hypotheses proposed by Babcock and Stebbins about its origins and spread, the relationships of diploids, and the nature and origins of polyploids. We used flow cytometry to infer ploidy of roughly 600 samples spanning the morphological and taxonomic diversity of the complex and a phylogenetic analysis of plastid DNA variation to infer maternal relationships among diploids and to infer maternal origins of polyploids. We identified populations of all seven recognized diploids plus one new lineage. Phylogenetic analysis of plastid DNA variation in diploids revealed a well-resolved, but moderately supported phylogeny, with evidence for monophyly of the North America Crepis agamic complex and no evidence of widespread homoploid hybridization. Polyploids showed evidence of multiple origins and a pattern of frequent local co-occurrence consistent with repeated colonization of suitable sites. Our findings agree broadly with the distribution and variation of ploidy within and among species described by Babcock and Stebbins. One key difference is finding support for monophyly of North American species, and refuting their hypothesis of polyphyly. Our results provide an explicit phylogenetic framework for further study of this classic agamic complex. © 2016 Botanical Society of America.

  9. Program to determine in-reactor performance of B and W fuels. Cladding creep collapse

    International Nuclear Information System (INIS)

    Eckert, A.F.J.; Wilson, H.W.; Yoon, K.E.

    1975-01-01

    An analytical procedure is described along with the experimental data developed by Babcock and Wilcox (B and W) to determine the minimum time to collapse for B and W fuel cladding under operating conditions. The resultant computer code is referred to as CROV. The analytical procedure was developed from the engineering principles of shell theory. The experimental data consisted of two types. One type of data was employed to develop empirical creep constants applicable to B and W fuel cladding. The other type of data provided an experimental comparison for the ovalities and collapse times predicted by CROV. 9 references. (U.S.)

  10. Role of Fas and Treg cells in fracture healing as characterized in the fas-deficient (lpr) mouse model of lupus.

    Science.gov (United States)

    Al-Sebaei, Maisa O; Daukss, Dana M; Belkina, Anna C; Kakar, Sanjeev; Wigner, Nathan A; Cusher, Daniel; Graves, Dana; Einhorn, Thomas; Morgan, Elise; Gerstenfeld, Louis C

    2014-06-01

    Previous studies showed that loss of tumor necrosis factor α (TNFα) signaling delayed fracture healing by delaying chondrocyte apoptosis and cartilage resorption. Mechanistic studies showed that TNFα induced Fas expression within chondrocytes; however, the degree to which chondrocyte apoptosis is mediated by TNFα alone or dependent on the induction of Fas is unclear. This question was addressed by assessing fracture healing in Fas-deficient B6.MRL/Fas(lpr) /J mice. Loss of Fas delayed cartilage resorption but also lowered bone fraction in the calluses. The reduced bone fraction was related to elevated rates of coupled bone turnover in the B6.MRL/Fas(lpr) /J calluses, as evidenced by higher osteoclast numbers and increased osteogenesis. Analysis of the apoptotic marker caspase 3 showed fewer positive chondrocytes and osteoclasts in calluses of B6.MRL/Fas(lpr) /J mice. To determine if an active autoimmune state contributed to increased bone turnover, the levels of activated T cells and Treg cells were assessed. B6.MRL/Fas(lpr) /J mice had elevated Treg cells in both spleens and bones of B6.MRL/Fas(lpr) /J but decreased percentage of activated T cells in bone tissues. Fracture led to ∼30% to 60% systemic increase in Treg cells in both wild-type and B6.MRL/Fas(lpr) /J bone tissues during the period of cartilage formation and resorption but either decreased (wild type) or left unchanged (B6.MRL/Fas(lpr) /J) the numbers of activated T cells in bone. These results show that an active autoimmune state is inhibited during the period of cartilage resorption and suggest that iTreg cells play a functional role in this process. These data show that loss of Fas activity specifically in chondrocytes prolonged the life span of chondrocytes and that Fas synergized with TNFα signaling to mediate chondrocyte apoptosis. Conversely, loss of Fas systemically led to increased osteoclast numbers during later periods of fracture healing and increased osteogenesis. These findings

  11. Geologic assessment of undiscovered conventional oil and gas resources in the Lower Paleogene Midway and Wilcox Groups, and the Carrizo Sand of the Claiborne Group, of the Northern Gulf coast region

    Science.gov (United States)

    Warwick, Peter D.

    2017-09-27

    The U.S. Geological Survey (USGS) recently conducted an assessment of the undiscovered, technically recoverable oil and gas potential of Tertiary strata underlying the onshore areas and State waters of the northern Gulf of Mexico coastal region. The assessment was based on a number of geologic elements including an evaluation of hydrocarbon source rocks, suitable reservoir rocks, and hydrocarbon traps in an Upper Jurassic-Cretaceous-Tertiary Composite Total Petroleum System defined for the region by the USGS. Five conventional assessment units (AUs) were defined for the Midway (Paleocene) and Wilcox (Paleocene-Eocene) Groups, and the Carrizo Sand of the Claiborne Group (Eocene) interval including: (1) the Wilcox Stable Shelf Oil and Gas AU; (2) the Wilcox Expanded Fault Zone Gas and Oil AU; (3) the Wilcox-Lobo Slide Block Gas AU; (4) the Wilcox Slope and Basin Floor Gas AU; and (5) the Wilcox Mississippi Embayment AU (not quantitatively assessed).The USGS assessment of undiscovered oil and gas resources for the Midway-Wilcox-Carrizo interval resulted in estimated mean values of 110 million barrels of oil (MMBO), 36.9 trillion cubic feet of gas (TCFG), and 639 million barrels of natural gas liquids (MMBNGL) in the four assessed units. The undiscovered oil resources are almost evenly divided between fluvial-deltaic sandstone reservoirs within the Wilcox Stable Shelf (54 MMBO) AU and deltaic sandstone reservoirs of the Wilcox Expanded Fault Zone (52 MMBO) AU. Greater than 70 percent of the undiscovered gas and 66 percent of the natural gas liquids (NGL) are estimated to be in deep (13,000 to 30,000 feet), untested distal deltaic and slope sandstone reservoirs within the Wilcox Slope and Basin Floor Gas AU.

  12. Coal gasification systems engineering and analysis. Appendix G: Commercial design and technology evaluation

    Science.gov (United States)

    1980-01-01

    A technology evaluation of five coal gasifier systems (Koppers-Totzek, Texaco, Babcock and Wilcox, Lurgi and BGC/Lurgi) and procedures and criteria for evaluating competitive commercial coal gasification designs is presented. The technology evaluation is based upon the plant designs and cost estimates developed by the BDM-Mittelhauser team.

  13. An extended conventional fuel cycle for the B and W mPowerTM small modular nuclear reactor

    International Nuclear Information System (INIS)

    Scarangella, M. J.

    2012-01-01

    The B and W mPower TM reactor is a small pressurized water reactor (PWR) with an integral once-through steam generator and a thermal output of about 500 MW; it is intended to replace aging fossil power plants of similar output. The core is composed of 69 reduced-height PWR assemblies with the familiar 17 x 17 fuel rod array. The Babcock and Wilcox Company (B and W) is offering a core loading and cycle management plan for a four-year cycle based on its presumed attractiveness to potential customers. This option is a once-through fuel cycle in which the entire core is discharged and replaced after four years. In addition, a conventional fuel utilization strategy, employing a periodic partial reload and shuffle, was developed as an alternative to the four-year once-through fuel cycle. This study, which was performed using the Studsvik core design code suite, is a typical multi-cycle projection analysis of the type performed by most fuel management organizations such as fuel vendors and utilities. In the industry, the results of such projections are used by the financial arms of these organizations to assist in making long-term decisions. In the case of the B and W mPower reactor, this analysis demonstrates flexibility for customers who consider the once-through fuel cycle unacceptable from a fuel utilization standpoint. As expected, when compared to the once-through concept, reloads of the B and W mPower reactor will achieve higher batch average discharge exposure, will have adequate shut-down margin, and will have a relatively flat hot excess reactivity trend at the expense of slightly increased peaking. (authors)

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

    International Nuclear Information System (INIS)

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

    1976-08-01

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

  15. Application of industrial robots in tubesheet cladding and tube to tubesheet welding

    International Nuclear Information System (INIS)

    Berbakov, P.J.

    1984-01-01

    This paper deals with the implementation of industrial robots in two areas of fabrication of nuclear power generation components at The Babcock and Wilcox Company facility in Barberton, Ohio. The applications described are robotic cladding of tubesheets, and tube-to-tubesheet Welding in nuclear steam generators

  16. Effects of Fetal and Neonatal Murine Peripheral Blood Mononuclear Cells Infusion on MicroRNA-145 Expression in Renal Vascular Smooth Muscle Cells in MRL/lpr Mice.

    Science.gov (United States)

    Wen, C; Liu, X Y; Wan, W Q; Yi, Z W

    2015-10-01

    For patients with refractory systemic lupus erythematosus, current medications are insufficient to control their condition, and new treatments are necessary. We aimed to evaluate the therapeutic effect of fetal and neonatal murine peripheral blood (FNPB) mononuclear cells and their impact on microRNA-145 (miR-145) in renal vascular smooth muscle cells (VSMCs) of MRL/lpr lupus-prone mice. MRL/lpr mice aged 20 weeks were randomized to 3 groups of 15 (control group, radiation group, infusion group). The renal tissues were subjected to pathological examination. In situ hybridization assay was applied to measure miR-145 expression in renal vessels of MRL/lpr mice. The infusion group had significantly better results for pathological renal tissue lesions than either the control or radiation group. In MRL/lpr mice, there was positive expression of miR-145 in renal VSMCs, although the expression of miR-145 was not discernible in renal vascular intima and adventitia. The miR-145 expression in renal VSMCs in the infusion group was significantly higher than in the control or radiation group, and higher in the radiation group than in the control group; however, the difference was not statistically significant. The increased expression of miR-145 in renal VSMCs might be one of the mechanisms supporting FNPB as a therapy for lupus nephritis; it also suggests that the miR-145 in renal vessels might be a new target for treatment of lupus nephritis. Copyright © 2015 Elsevier Inc. All rights reserved.

  17. Paleocene Wilcox cross-shelf channel-belt history and shelf-margin growth: Key to Gulf of Mexico sediment delivery

    Science.gov (United States)

    Zhang, Jinyu; Steel, Ronald; Ambrose, William

    2017-12-01

    Shelf margins prograde and aggrade by the incremental addition of deltaic sediments supplied from river channel belts and by stored shoreline sediment. This paper documents the shelf-edge trajectory and coeval channel belts for a segment of Paleocene Lower Wilcox Group in the northern Gulf of Mexico based on 400 wireline logs and 300 m of whole cores. By quantitatively analyzing these data and comparing them with global databases, we demonstrate how varying sediment supply impacted the Wilcox shelf-margin growth and deep-water sediment dispersal under greenhouse eustatic conditions. The coastal plain to marine topset and uppermost continental slope succession of the Lower Wilcox shelf-margin sediment prism is divided into eighteen high-frequency ( 300 ky duration) stratigraphic sequences, and further grouped into 5 sequence sets (labeled as A-E from bottom to top). Sequence Set A is dominantly muddy slope deposits. The shelf edge of Sequence Sets B and C prograded rapidly (> 10 km/Ma) and aggraded modestly ( 80 m/Ma) characterizes Sequence Sets D and E, which is associated with smaller (9-10 m thick on average) and isolated channel belts. This stratigraphic trend is likely due to an upward decreasing sediment supply indicated by the shelf-edge progradation rate and channel size, as well as an upward increasing shelf accommodation indicated by the shelf-edge aggradation rate. The rapid shelf-edge progradation and large rivers in Sequence Sets B and C confirm earlier suggestions that it was the early phase of Lower Wilcox dispersal that brought the largest deep-water sediment volumes into the Gulf of Mexico. Key factors in this Lower Wilcox stratigraphic trend are likely to have been a very high initial sediment flux to the Gulf because of the high initial release of sediment from Laramide catchments to the north and northwest, possibly aided by modest eustatic sea-level fall on the Texas shelf, which is suggested by the early, flat shelf-edge trajectory, high

  18. Lessons from American-German nuclear power plant construction. Quality, safety and costs of an attempt to integrate American and German nuclear power plant technology

    International Nuclear Information System (INIS)

    Buchwald, K.

    1979-05-01

    The 1300 MW nuclear power plant at Muelheim-Kaerlich has been under construction since the beginning of 1975. It is being equipped with a pressurised water reactor which has been adapted to the German client's requirements and German licensing practice, based on a license held by Babcock and Wilcox USA (B and W). The problems which have arisen in making this adaptation are the result of different requirements in the USA and the Federal Republic of Germany which make it very difficult to integrate the two technologies. Full integration will almost certainly be impossible, but integration to the widest possible extent is important because it might mean both greater safety and reduced costs. In this article it is intended to show where the problems of integration lie and how they might perhaps be overcome. (author)

  19. Loss of P2X7 receptor plasma membrane expression and function in pathogenic B220+ double-negative T lymphocytes of autoimmune MRL/lpr mice.

    Directory of Open Access Journals (Sweden)

    Sylvain M Le Gall

    Full Text Available Lupus is a chronic inflammatory autoimmune disease influenced by multiple genetic loci including Fas Ligand (FasL and P2X7 receptor (P2X7R. The Fas/Fas Ligand apoptotic pathway is critical for immune homeostasis and peripheral tolerance. Normal effector T lymphocytes up-regulate the transmembrane tyrosine phosphatase B220 before undergoing apoptosis. Fas-deficient MRL/lpr mice (lpr mutation exhibit lupus and lymphoproliferative syndromes due to the massive accumulation of B220(+ CD4(-CD8(- (DN T lymphocytes. The precise ontogeny of B220(+ DN T cells is unknown. B220(+ DN T lymphocytes could be derived from effector CD4(+ and CD8(+ T lymphocytes, which have not undergone activation-induced cell death due to inactivation of Fas, or from a special cell lineage. P2X7R is an extracellular ATP-gated cell membrane receptor involved in the release of proinflammatory cytokines and TNFR1/Fas-independent cell death. P2X7R also regulate early signaling events involved in T-cell activation. We show herein that MRL/lpr mice carry a P2X7R allele, which confers a high sensitivity to ATP. However, during aging, the MRL/lpr T-cell population exhibits a drastically reduced sensitivity to ATP- or NAD-mediated stimulation of P2X7R, which parallels the increase in B220(+ DN T-cell numbers in lymphoid organs. Importantly, we found that this B220(+ DN T-cell subpopulation has a defect in P2X7R-mediated responses. The few B220(+ T cells observed in normal MRL(+/+ and C57BL/6 mice are also resistant to ATP or NAD treatment. Unexpectedly, while P2X7R mRNA and proteins are present inside of B220(+ T cells, P2X7R are undetectable on the plasma membrane of these T cells. Our results prompt the conclusion that cell surface expression of B220 strongly correlates with the negative regulation of the P2X7R pathway in T cells.

  20. Development and testing of a diagnostic system for intelligen distributed control at EBR-2

    International Nuclear Information System (INIS)

    Edwards, R.M.; Ruhl, D.W.; Klevans, E.H.; Robinson, G.E.

    1990-01-01

    A diagnostic system is under development for demonstration of Intelligent Distributed Control at the Experimental Breeder Reactor (EBR--II). In the first phase of the project a diagnostic system is being developed for the EBR-II steam plant based on the DISYS expert systems approach. Current testing uses recorded plant data and data from simulated plant faults. The dynamical simulation of the EBR-II steam plant uses the Babcock and Wilcox (B ampersand W) Modular Modeling System (MMS). At EBR-II the diagnostic system operates in the UNIX workstation and receives live plant data from the plant Data Acquisition System (DAS). Future work will seek implementation of the steam plant diagnostic in a distributed manner using UNIX based computers and Bailey microprocessor-based control system. 10 refs., 6 figs

  1. Flexible Architecture System & Topology License Plate Recognition (Fast LPR) and Concept of Operations in Thailand

    Science.gov (United States)

    2008-09-01

    operating in jungle areas north of Thai- Malaysian border, the successes of communist forces in Vietnam and Laos, and other regional unrest and protest...FARC being in control of these areas, “it began to traffic in drugs.”68 FARC’s revenue in the beginning was based on taxation on the drug loads as

  2. Obituary: Horace Welcome Babcock, 1912-2003

    Science.gov (United States)

    Vaughan, Arthur Harris

    2003-12-01

    Horace Welcome Babcock died in Santa Barbara, California on 29 August 2003, fifteen days short of his ninety-first birthday. An acclaimed authority on solar and stellar magnetism and the originator of ingenious advances in astronomical instrumentation in his earlier career, he served as Director of Mount Wilson and Palomar (later Hale) Observatories from 1964 until his retirement in 1978. The founding of the Carnegie Institution of Washington's Las Campanas Observatory in Chile was the culmination of his directorship. Horace was born in Pasadena California on 13 September 1912, the only child of Harold Delos Babcock and Mary G. Henderson. His father, an electrical engineer and physicist by training, had been hired by George Ellery Hale to work at the recently founded Mount Wilson Solar Observatory beginning in 1909. Thus Horace spent much of his boyhood on Mount Wilson in the company of astronomers. Horace developed an early interest in astronomy, worked as a volunteer solar observer at Mount Wilson and published his first paper in 1932, with his father. He was fascinated by fine mechanisms and by optical and electrical instruments. After graduating from Caltech with a degree in structural engineering in 1934, he earned his PhD in astronomy at Lick Observatory in 1938. His dissertation provided the first measurement of the rotational velocity curve and a derivation of the mass-to-luminosity ratio for M31; this work is still cited in reviews of the study of ``dark matter." Horace served as a research assistant at Lick Observatory (1938 39) and an Instructor at the University of Chicago's McDonald and Yerkes Observatories (1939--41) under Otto Struve. He undertook radar-related wartime electronics work at the MIT Radiation Laboratory (1941 42) and then worked on aircraft rocket launchers as part of the Caltech Rocket Project (1942 45). This project brought him into contact with Ira S. Bowen, head of the project's Photographic Division. Impressed with his knowledge of

  3. xLPR Sim Editor 1.0 User's Guide

    Energy Technology Data Exchange (ETDEWEB)

    Mariner, Paul

    2017-03-01

    The United States Nuclear Regulatory Commission in cooperation with the Electric Power Research Institute contracted Sandia National Laboratories to develop the framework of a probabilistic fracture mechanics assessment code called xLPR ( Extremely Low Probability of Rupture) Version 2.0 . The purpose of xLPR is to evaluate degradation mechanisms in piping systems at nuclear power plants and to predict the probability of rupture. This report is a user's guide for xLPR Sim Editor 1.0 , a graphical user interface for creating and editing the xLPR Version 2.0 input file and for creating, editing, and using the xLPR Version 2.0 database files . The xLPR Sim Editor, provides a user - friendly way for users to change simulation options and input values, s elect input datasets from xLPR data bases, identify inputs needed for a simulation, and create and modify an input file for xLPR.

  4. Babcock Redux: An Amendment of Babcock's Schematic of the Sun's Magnetic Cycle

    Science.gov (United States)

    Moore, Ronald L.; Cirtain, Jonathan W.; Sterling, Alphonse C.

    2017-08-01

    We amend Babcock's original scenario for the global dynamo process that sustains the Sun's 22-year magnetic cycle. The amended scenario fits post-Babcock observed features of the magnetic activity cycle and convection zone, and is based on ideas of Spruit & Roberts (1983, Nature, 304, 401) about magnetic flux tubes in the convection zone. A sequence of four schematic cartoons lays out the proposed evolution of the global configuration of the magnetic field above, in, and at the bottom of the convection zone through sunspot Cycle 23 and into Cycle 24. Three key elements of the amended scenario are: (1) as the net following-polarity magnetic field from the sunspot-region Ω-loop fields of an ongoing sunspot cycle is swept poleward to cancel and replace the opposite-polarity polar-cap field from the previous sunspot cycle, it remains connected to the ongoing sunspot cycle's toroidal source-field band at the bottom of the convection zone; (2) topological pumping by the convection zone's free convection keeps the horizontal extent of the poleward-migrating following-polarity field pushed to the bottom, forcing it to gradually cancel and replace old horizontal field below it that connects the ongoing-cycle source-field band to the previous-cycle polar-cap field; (3) in each polar hemisphere, by continually shearing the poloidal component of the settling new horizontal field, the latitudinal differential rotation low in the convection zone generates the next-cycle source-field band poleward of the ongoing-cycle band. The amended scenario is a more-plausible version of Babcock's scenario, and its viability can be explored by appropriate kinematic flux-transport solar-dynamo simulations. A paper giving a full description of our dynamo scenario is posted on arXiv (http://arxiv.org/abs/1606.05371).This work was funded by the Heliophysics Division of NASA's Science Mission Directorate through the Living With a Star Targeted Research and Technology Program and the Hinode

  5. Reactor analysis support package (RASP). Volume 7. PWR set-point methodology. Final report

    International Nuclear Information System (INIS)

    Temple, S.M.; Robbins, T.R.

    1986-09-01

    This report provides an overview of the basis and methodology requirements for determining Pressurized Water Reactor (PWR) technical specifications related setpoints and focuses on development of the methodology for a reload core. Additionally, the report documents the implementation and typical methods of analysis used by PWR vendors during the 1970's to develop Protection System Trip Limits (or Limiting Safety System Settings) and Limiting Conditions for Operation. The descriptions of the typical setpoint methodologies are provided for Nuclear Steam Supply Systems as designed and supplied by Babcock and Wilcox, Combustion Engineering, and Westinghouse. The description of the methods of analysis includes the discussion of the computer codes used in the setpoint methodology. Next, the report addresses the treatment of calculational and measurement uncertainties based on the extent to which such information was available for each of the three types of PWR. Finally, the major features of the setpoint methodologies are compared, and the principal effects of each particular methodology on plant operation are summarized for each of the three types of PWR

  6. Taxonomic revision of Plyomydas Wilcox & Papavero, 1971 with the description of two new species and its transfer to Mydinae (Insecta: Diptera: Mydidae

    Directory of Open Access Journals (Sweden)

    Stephanie Castillo

    Full Text Available ABSTRACT The monotypic Neotropical Mydidae genus Plyomydas Wilcox & Papavero, 1971, to date confined to coastal Peru, is reviewed. Two new species, Plyomydas adelphe sp. nov. and Plyomydas phalaros sp. nov., are described from mid-elevational western Argentina, which extends the distribution of the genus considerably. Distribution, occurrence in biodiversity hotspots sensu Conservation International, and seasonal incidence are discussed. Descriptions/re-descriptions, photographs, illustrations, and identification keys are provided and made openly accessible in data depositories to support future studies of the included taxa. Plyomydas is transferred from the Leptomydinae to the Mydinae: Messiasiini based on the absence of acanthophorite spines on abdominal tergite 10 in females and the presence of vein M3 + M4 terminating in the costal vein C. Leptomydinae is therefore restricted to the Northern Hemisphere with the exception of Hessemydas Kondratieff, Carr & Irwin, 2005 known from Madagascar. Messiasia notospila (Wiedemann, 1828 is compared to Plyomydas species.

  7. Four critical facilities: their capabilities and programs

    International Nuclear Information System (INIS)

    Whitesides, G.E.

    1980-01-01

    Information is presented on the critical experiments facilities at Babcock and Wilcox, Lynchburg, Virginia; at Battelle Pacific Northwest Laboratory in Hanford, Washington; at Rockwell-International in Rocky Flats, Colorado; and at Los Alamos Scientific Laboratory in New Mexico. It is noted that the critical mass facilities which still exist in this country represent a bare minimum for maintaining a measurement program sufficient for meeting data requirements

  8. Reactor physics and reactor computations

    International Nuclear Information System (INIS)

    Ronen, Y.; Elias, E.

    1994-01-01

    Mathematical methods and computer calculations for nuclear and thermonuclear reactor kinetics, reactor physics, neutron transport theory, core lattice parameters, waste treatment by transmutation, breeding, nuclear and thermonuclear fuels are the main interests of the conference

  9. Hot Isostatic Press Can Optimization for Aluminum Cladding of U-10Mo Reactor Fuel Plates: FY12 Final Report and FY13 Update

    Energy Technology Data Exchange (ETDEWEB)

    Clarke, Kester D. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Crapps, Justin M. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Scott, Jeffrey E. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Aikin, Beverly [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Vargas, Victor D. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Dvornak, Matthew J. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Duffield, Andrew N. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Weinberg, Richard Y. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Alexander, David J. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Montalvo, Joel D. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Hudson, Richard W. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Mihaila, Bogdan [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Liu, Cheng [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Lovato, Manuel L. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Dombrowski, David E. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2013-08-26

    Currently, the proposed processing path for low enriched uranium – 10 wt. pct. molybdenum alloy (LEU-10Mo) monolithic fuel plates for high power research and test reactors includes hot isostatic pressing (HIP) to bond the aluminum cladding that encapsulates the fuel foil. Initial HIP experiments were performed at Idaho National Laboratory (INL) on approximately ¼ scale “mini” fuel plate samples using a HIP can design intended for these smaller experimental trials. These experiments showed that, with the addition of a co-rolled zirconium diffusion barrier on the LEU-10Mo alloy fuel foil, the HIP bonding process is a viable method for producing monolithic fuel plates. Further experimental trials at Los Alamos National Laboratory (LANL) effectively scaled-up the “mini” can design to produce full-size fuel prototypic plates. This report summarizes current efforts at LANL to produce a HIP can design that is further optimized for higher volume production runs. The production-optimized HIP can design goals were determined by LANL and Babcock & Wilcox (B&W) to include maintaining or improving the quality of the fuel plates produced with the baseline scaled-up mini can design, while minimizing material usage, improving dimensional stability, easing assembly and disassembly, eliminating machining, and significantly reducing welding. The initial small-scale experiments described in this report show that a formed-can approach can achieve the goals described above. Future work includes scaling the formed-can approach to full-size fuel plates, and current progress toward this goal is also summarized here.

  10. Paleocene coal deposits of the Wilcox group, central Texas

    Science.gov (United States)

    Hook, Robert W.; Warwick, Peter D.; SanFilipo, John R.; Schultz, Adam C.; Nichols, Douglas J.; Swanson, Sharon M.; Warwick, Peter D.; Karlsen, Alexander K.; Merrill, Matthew D.; Valentine, Brett J.

    2011-01-01

    Coal deposits in the Wilcox Group of central Texas have been regarded as the richest coal resources in the Gulf Coastal Plain. Although minable coal beds appear to be less numerous and generally higher in sulfur content (1 percent average, as-received basis; table 1) than Wilcox coal deposits in the Northeast Texas and Louisiana Sabine assessment areas (0.5 and 0.6 percent sulfur, respectively; table 1), net coal thickness in coal zones in central Texas is up to 32 ft thick and more persistent along strike (up to 15 mi) at or near the surface than coals of any other Gulf Coast assessment area. The rank of the coal beds in central Texas is generally lignite (table 1), but some coal ranks as great as subbituminous C have been reported (Mukhopadhyay, 1989). The outcrop of the Wilcox Group in central Texas strikes northeast, extends for approximately 140 mi between the Trinity and Colorado Rivers, and covers parts of Bastrop, Falls, Freestone, Lee, Leon, Limestone, Milam, Navarro, Robertson, and Williamson Counties (Figure 1). Three formations, in ascending order, the Hooper, Simsboro, and Calvert Bluff, are recognized in central Texas (Figure 2). The Wilcox Group is underlain conformably by the Midway Group, a mudstone-dominated marine sequence, and is overlain and scoured locally by the Carrizo Sand, a fluvial unit at the base of the Claiborne Group.

  11. Sunlight triggers cutaneous lupus through a CSF-1-dependent mechanism in MRL-Fas(lpr) mice.

    Science.gov (United States)

    Menke, Julia; Hsu, Mei-Yu; Byrne, Katelyn T; Lucas, Julie A; Rabacal, Whitney A; Croker, Byron P; Zong, Xiao-Hua; Stanley, E Richard; Kelley, Vicki R

    2008-11-15

    Sunlight (UVB) triggers cutaneous lupus erythematosus (CLE) and systemic lupus through an unknown mechanism. We tested the hypothesis that UVB triggers CLE through a CSF-1-dependent, macrophage (Mø)-mediated mechanism in MRL-Fas(lpr) mice. By constructing mutant MRL-Fas(lpr) strains expressing varying levels of CSF-1 (high, intermediate, none), and use of an ex vivo gene transfer to deliver CSF-1 intradermally, we determined that CSF-1 induces CLE in lupus-susceptible MRL-Fas(lpr) mice, but not in lupus-resistant BALB/c mice. UVB incites an increase in Møs, apoptosis in the skin, and CLE in MRL-Fas(lpr), but not in CSF-1-deficient MRL-Fas(lpr) mice. Furthermore, UVB did not induce CLE in BALB/c mice. Probing further, UVB stimulates CSF-1 expression by keratinocytes leading to recruitment and activation of Møs that, in turn, release mediators, which induce apoptosis in keratinocytes. Thus, sunlight triggers a CSF-1-dependent, Mø-mediated destructive inflammation in the skin leading to CLE in lupus-susceptible MRL-Fas(lpr) but not lupus-resistant BALB/c mice. Taken together, CSF-1 is envisioned as the match and lupus susceptibility as the tinder leading to CLE.

  12. CNSS plant concept, capital cost, and multi-unit station economics

    International Nuclear Information System (INIS)

    1984-07-01

    United Engineers and Constructors (UE and C) and the Babcock and Wilcox Company (B and W) have performed several studies over the last eight years related to small integral pressurized water reactors. These reactors include the 365 MWt (100 MWe) Consolidated Nuclear Steam Generator (CNSG) and the 1200 MWt Consolidated Nuclear Steam System (CNSS). The studies, mostly performed under contract to the Oak Ridge National Laboratory, have led to a 1250 MWt (400 MWe) Consolidated Nuclear Steam System (CNSS) plant concept, with unique design and cost features. This report contains an update of earlier studies of the CNSS reactor and balance-of-plant concept design, capital costs, and multi-unit plant economics incorporating recent design developments, improvements, and post-TMI-2 upgrades. The economic evaluation compares the total system economic impact of a phased, three stage 400 MWe CNSS implementation program, i.e., a three-unit station, to the installation of a single 1200 MWe Pressurized Water Reactor (PWR) into a typical USA utility system

  13. Additions to the flora of the Wilcox group

    Science.gov (United States)

    Berry, Edward Wilber

    1923-01-01

    A rather full account of the extensive flora contained in the lower Eocene strata of the Mississippi embayment which are referred to the Wilcox group was published in 1916. At that time it was not possible to obtain sections of the numerous specimens of petrified wood that had been collected from these beds. These woods have since been sectioned and studied, and it seems eminently desirable to place the results of this study on record, for although much of the material had suffered greatly from decay before silicification, some of it is fairly well preserved and shows, among other results, that conifers were individually much more plentiful during Wilcox time than would be inferred from the almost total absence of their foliage in the very large collections of remains of this class that have been studied.

  14. Reclamation and reuse of LEU silicide fuel from manufacturing scrap

    International Nuclear Information System (INIS)

    Gale, G.R.; Pace, B.W.; Evans, R.S.

    2004-01-01

    In order to provide an understanding of the organization which is the sole supplier of United States plate type research and test reactor fuel and LEU core conversions, a brief description of the structure and history is presented. Babcock and Wilcox (B and W) is a part of McDermott International, Inc. which is a large diversified corporation employing over 20,000 people primarily in engineering and construction for the off-shore oil and power generation industries throughout the world. B and W provides many energy related products requiring precision machining and high quality systems. This is accomplished by using state-of-the-art equipment, technology and highly skilled people. The RTRFE group within B and W has the ability to produce various complexly shaped fuel elements with a wide variety of fuels and enrichments. B and W RTRFE has fabricated over 200,000 plates since 1981 and gained the diversified experience necessary to satisfy many customer requirements. This accomplishment was possible with the support of McDermott International and all of its resources. B and W has always had a commitment to high quality and integrity. This is apparent by the success and longevity (125 years) of the company. A lower cost to convert cores to LEU provides direct support to RERTR and demonstrates Babcock and Wilcox's commitment to the program. As a supporter of RERTR reactor conversion from HEU to LEU, B and W has contributed a significant amount of R and D money to improve the silicide fuel process which ultimately lowers the LEU core costs. In the most recent R and D project, B and W is constructing a LEU silicide reclamation facility to re-use the unirradiated fuel scrap generated from the production process. Remanufacturing use of this fuel completes the fuel cycle and provides a contribution to LEU cores by reducing scrap inventory and handling costs, lowering initial purchase of fuel due to increasing the process yields, and lowering the replacement costs. This

  15. Feeder replacement tooling and processes

    International Nuclear Information System (INIS)

    Mallozzi, R.; Goslin, R.; Pink, D.; Askari, A.

    2008-01-01

    Primary heat transport system feeder integrity has become a concern at some CANDU nuclear plants as a result of thinning caused by flow accelerated corrosion (FAC). Feeder inspections are indicating that life-limiting wall thinning can occur in the region between the Grayloc hub weld and second elbow of some outlet feeders. In some cases it has become necessary to replace thinned sections of affected feeders to restore feeder integrity to planned end of life. Atomic Energy of Canada Limited (AECL) and Babcock and Wilcox Canada Ltd. (B and W) have developed a new capability for replacement of single feeders at any location on the reactor face without impacting or interrupting operation of neighbouring feeders. This new capability consists of deploying trained crews with specialized tools and procedures for feeder replacements during planned outages. As may be expected, performing single feeder replacement in the congested working environment of an operational CANDU reactor face involves overcoming many challenges with respect to access to feeders, available clearances for tooling, and tooling operation and performance. This paper describes some of the challenges encountered during single feeder replacements and actions being taken by AECL and B and W to promote continuous improvement of feeder replacement tooling and processes and ensure well-executed outages. (author)

  16. Fate of injected CO2 in the Wilcox Group, Louisiana, Gulf Coast Basin: Chemical and isotopic tracers of microbial–brine–rock–CO2 interactions

    Science.gov (United States)

    Shelton, Jenna L.; McIntosh, Jennifer C.; Warwick, Peter D.; Lee Zhi Yi, Amelia

    2014-01-01

    The “2800’ sandstone” of the Olla oil field is an oil and gas-producing reservoir in a coal-bearing interval of the Paleocene–Eocene Wilcox Group in north-central Louisiana, USA. In the 1980s, this producing unit was flooded with CO2 in an enhanced oil recovery (EOR) project, leaving ∼30% of the injected CO2 in the 2800’ sandstone post-injection. This study utilizes isotopic and geochemical tracers from co-produced natural gas, oil and brine to determine the fate of the injected CO2, including the possibility of enhanced microbial conversion of CO2 to CH4 via methanogenesis. Stable carbon isotopes of CO2, CH4 and DIC, together with mol% CO2 show that 4 out of 17 wells sampled in the 2800’ sandstone are still producing injected CO2. The dominant fate of the injected CO2appears to be dissolution in formation fluids and gas-phase trapping. There is some isotopic and geochemical evidence for enhanced microbial methanogenesis in 2 samples; however, the CO2 spread unevenly throughout the reservoir, and thus cannot explain the elevated indicators for methanogenesis observed across the entire field. Vertical migration out of the target 2800’ sandstone reservoir is also apparent in 3 samples located stratigraphically above the target sand. Reservoirs comparable to the 2800’ sandstone, located along a 90-km transect, were also sampled to investigate regional trends in gas composition, brine chemistry and microbial activity. Microbial methane, likely sourced from biodegradation of organic substrates within the formation, was found in all oil fields sampled, while indicators of methanogenesis (e.g. high alkalinity, δ13C-CO2 and δ13C-DIC values) and oxidation of propane were greatest in the Olla Field, likely due to its more ideal environmental conditions (i.e. suitable range of pH, temperature, salinity, sulfate and iron concentrations).

  17. The Lipocalin LPR-1 Cooperates with LIN-3/EGF Signaling To Maintain Narrow Tube Integrity in Caenorhabditis elegans

    Science.gov (United States)

    Pu, Pu; Stone, Craig E.; Burdick, Joshua T.; Murray, John I.; Sundaram, Meera V.

    2017-01-01

    Lipocalins are secreted cup-shaped glycoproteins that bind sterols, fatty acids, and other lipophilic molecules. Lipocalins have been implicated in a wide array of processes related to lipophilic cargo transport, sequestration, and signaling, and several are used as biomarkers for human disease, but the functions of most lipocalins remain poorly understood. Here we show that the Caenorhabditis elegans lipocalin LPR-1 is required to maintain apical membrane integrity and a continuous lumen in two narrow unicellular tubes, the excretory duct and pore, during a period of rapid lumen elongation. LPR-1 fusion protein is expressed by the duct and pore and accumulates both intracellularly and in apical extracellular compartments, but it can also function cell nonautonomously when provided from outside of the excretory system. lpr-1 mutant defects can be rescued by increased signaling through the epidermal growth factor (EGF)-Ras-extracellular signal regulated kinase (ERK) pathway, which promotes the more elongated duct vs. less elongated pore tube fate. Spatial and temporal rescue experiments indicate that Ras signaling acts within the duct and pore tubes during or prior to cell fate determination to bypass the requirement for LPR-1. lpr-1 mutations did not disrupt LIN-3/EGF-dependent duct-fate specification, prevent functioning of any specific LIN-3/EGF isoform, or alter LET-23/EGFR localization, and reduced signaling did not phenocopy or enhance lpr-1 mutant defects. These data suggest that LPR-1 protects lumen integrity through a LIN-3/EGF-independent mechanism, but that increased signaling upregulates some target(s) that can compensate for lpr-1 absence. PMID:28040739

  18. TRAC-PF1/MOD1 calculations and data comparisons for MIST [Multi-Loop Integral System Test] small-break loss-of-coolant accidents with scaled 10 cm2 and 50 cm2 breaks

    International Nuclear Information System (INIS)

    Steiner, J.L.; Siebe, D.A.; Boyack, B.E.

    1987-01-01

    Los Alamos National Laboratory is a participant in the Integral System Test (IST) program initiated in June 1983 for the purpose of providing integral system test data on specific issues/phenomena relevant to post-small-break loss-of-coolant accidents (SBLOCAs), loss of feedwater and other transients in Babcock and Wilcox (B and W) plant designs. The Multi-Loop Integral System Test (MIST) facility is the largest single component in the IST program. MIST is a 2 x 4 (2 hot legs and steam generators, 4 cold legs and reactor-coolant pumps) representation of lowered-loop reactor systems of the B and W design. It is a full-height, full-pressure facility with 1/817 power and volume scaling. Two other experimental facilities are included in the IST program: test loops at the University of Maryland, College Park, and at Stanford Research Institute. The objective of the IST tests is to generate high-quality experimental data to be used for assessing thermal-hydraulic safety computer codes. Efforts are underway at Los Alamos to assess TRAC-PF1/MOD1 against data from each of the IST facilities. Calculations and data comparisons for TRAC-PF1/MOD1 assessment have been completed for two transients run in the MIST facility. These are the MIST nominal test. Test 3109AA, a scaled 10 cm 2 SBLOCA and Test 320201, a scaled 50 cm 2 SBLOCA. Only MIST assessment results are presented in this paper

  19. SPACE-R Thermionic Space Nuclear Power System: Design and Technology Demonstration Program. Semiannual technical progress report for period ending March 1993

    Energy Technology Data Exchange (ETDEWEB)

    1993-05-01

    This Semiannual Technical Progress Report summarizes the technical progress and accomplishments for the Thermionic Space Nuclear Power System (TI-SNPS) Design and Technology Demonstration Program of the Prime Contractor, Space Power Incorporated (SPI), its subcontractors and supporting National Laboratories during the first half of the Government Fiscal Year (GFY) 1993. SPI`s subcontractors and supporting National Laboratories include: Babcock & Wilcox for the reactor core and externals; Space Systems/Loral for the spacecraft integration; Thermocore for the radiator heat pipes and the heat exchanger; INERTEK of CIS for the TFE, core elements and nuclear tests; Argonne National Laboratories for nuclear safety, physics and control verification; and Oak Ridge National laboratories for materials testing. Parametric trade studies are near completion. However, technical input from INERTEK has yet to be provided to determine some of the baseline design configurations. The INERTEK subcontract is expected to be initiated soon. The Point Design task has been initiated. The thermionic fuel element (TFE) is undergoing several design iterations. The reactor core vessel analysis and design has also been started.

  20. Acceptance Test Data for BWXT Coated Particle Batch 93164A Defective IPyC Fraction and Pyrocarbon Anisotropy

    Energy Technology Data Exchange (ETDEWEB)

    Helmreich, Grant W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Hunn, John D. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Skitt, Darren J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Dyer, John A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2017-02-01

    Coated particle fuel batch J52O-16-93164 was produced by Babcock and Wilcox Technologies (BWXT) for possible selection as fuel for the Advanced Gas Reactor Fuel Development and Qualification (AGR) Program’s AGR-5/6/7 irradiation test in the Idaho National Laboratory (INL) Advanced Test Reactor (ATR), or may be used as demonstration production-scale coated particle fuel for other experiments. The tristructural-isotropic (TRISO) coatings were deposited in a 150-mm-diameter production-scale fluidizedbed chemical vapor deposition (CVD) furnace onto 425-μm-nominal-diameter spherical kernels from BWXT lot J52L-16-69316. Each kernel contained a mixture of 15.5%-enriched uranium carbide and uranium oxide (UCO) and was coated with four consecutive CVD layers: a ~50% dense carbon buffer layer with 100-μm-nominal thickness, a dense inner pyrolytic carbon (IPyC) layer with 40-μm-nominal thickness, a silicon carbide (SiC) layer with 35-μm-nominal thickness, and a dense outer pyrolytic carbon (OPyC) layer with 40-μm-nominal thickness. The TRISO-coated particle batch was sieved to upgrade the particles by removing over-sized and under-sized material, and the upgraded batch was designated by appending the letter A to the end of the batch number (i.e., 93164A).

  1. Rolling Process Modeling Report. Finite-Element Model Validation and Parametric Study on various Rolling Process parameters

    Energy Technology Data Exchange (ETDEWEB)

    Soulami, Ayoub [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Lavender, Curt A. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Paxton, Dean M. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Burkes, Douglas [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2015-06-15

    Pacific Northwest National Laboratory (PNNL) has been investigating manufacturing processes for the uranium-10% molybdenum alloy plate-type fuel for high-performance research reactors in the United States. This work supports the U.S. Department of Energy National Nuclear Security Administration’s Office of Material Management and Minimization Reactor Conversion Program. This report documents modeling results of PNNL’s efforts to perform finite-element simulations to predict roll-separating forces for various rolling mill geometries for PNNL, Babcock & Wilcox Co., Y-12 National Security Complex, Los Alamos National Laboratory, and Idaho National Laboratory. The model developed and presented in a previous report has been subjected to further validation study using new sets of experimental data generated from a rolling mill at PNNL. Simulation results of both hot rolling and cold rolling of uranium-10% molybdenum coupons have been compared with experimental results. The model was used to predict roll-separating forces at different temperatures and reductions for five rolling mills within the National Nuclear Security Administration Fuel Fabrication Capability project. This report also presents initial results of a finite-element model microstructure-based approach to study the surface roughness at the interface between zirconium and uranium-10% molybdenum.

  2. Acceptance Test Data for Candidate AGR-5/6/7 TRISO Particle Batches BWXT Coater Batches 93165 93172 Defective IPyC Fraction and Pyrocarbon Anisotropy

    Energy Technology Data Exchange (ETDEWEB)

    Helmreich, Grant W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Hunn, John D. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Skitt, Darren J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Dyer, John A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Schumacher, Austin T. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2017-03-01

    Coated particle fuel batches J52O-16-93165, 93166, 93168, 93169, 93170, and 93172 were produced by Babcock and Wilcox Technologies (BWXT) for possible selection as fuel for the Advanced Gas Reactor Fuel Development and Qualification (AGR) Program’s AGR-5/6/7 irradiation test in the Idaho National Laboratory (INL) Advanced Test Reactor (ATR). Some of these batches may alternately be used as demonstration coated particle fuel for other experiments. Each batch was coated in a 150-mm-diameter production-scale fluidized-bed chemical vapor deposition (CVD) furnace. Tristructural isotropic (TRISO) coatings were deposited on 425-μm-nominal-diameter spherical kernels from BWXT lot J52R-16-69317 containing a mixture of 15.5%-enriched uranium carbide and uranium oxide (UCO). The TRISO coatings consisted of four consecutive CVD layers: a ~50% dense carbon buffer layer with 100-μm-nominal thickness, a dense inner pyrolytic carbon (IPyC) layer with 40-μm-nominal thickness, a silicon carbide (SiC) layer with 35-μm-nominal thickness, and a dense outer pyrolytic carbon (OPyC) layer with 40-μmnominal thickness. The TRISO-coated particle batches were sieved to upgrade the particles by removing over-sized and under-sized material, and the upgraded batches were designated by appending the letter A to the end of the batch number (e.g., 93165A).

  3. The association of 5-HTR2A-1438A/G, COMTVal158Met, MAOA-LPR, DATVNTR and 5-HTTVNTR gene polymorphisms and antisocial personality disorder in male heroin-dependent Chinese subjects.

    Science.gov (United States)

    Yang, Mei; Kavi, Vasish; Wang, Wenfu; Wu, Zhimei; Hao, Wei

    2012-03-30

    To explore the association between the 5-HTR2A-1438A/G, COMTVal158Met, MAOA-LPR, DATVNTR and 5-HTTVNTR polymorphisms with comorbidity of antisocial personality disorder in male heroin-dependent patients. In case control study, we compared the polymorphic distributions of 5-HTR2A-1438A/G, COMTVal158Met, MAOA-LPR, DATVNTR and 5-HTTVNTR in 588 male heroin-dependent patients (including 311 patients with antisocial personality disorder and 277 patients without antisocial personality disorder) and 194 normal males by genotypes, alleles, and interaction between genes. Between male heroin-dependent patients with antisocial personality disorder and normal males, and between male heroin-dependent patients with and without antisocial personality disorder, the distributions of 5-HTTVNTR polymorphic genotypes and alleles were in statistical significance. Individuals carrying 10R allele were in higher risk of the comorbidity of antisocial personality disorder and heroin dependence. By MDR analyses, the interaction between 5-HTTVNTR and DATVNTR was close to statistical significance in predicting the risk of antisocial personality disorder in male heroin dependent patients. In male heroin dependent patients, individuals carrying 5-HTTVNTR 10R allele or/and DATVNTR 9R allele were in higher risks of co-occurring antisocial personality disorder, while individuals with 5-HTTVNTR 12R/12R and DATVNTR 10R/10R genotypes together were in lower risks of antisocial personality disorder. 5-HTTVNTR, and the interaction between 5-HTTVNTR and DATVNTR may be associated with the comorbidity of antisocial personality disorder in male heroin-dependent patients. Copyright © 2011 Elsevier Inc. All rights reserved.

  4. Qualification of the B and W Mark B fuel assembly for high burnup. First semi-annual progress report, July-December 1978

    International Nuclear Information System (INIS)

    Coleman, T.A.; Coppola, E.J.; Doss, P.L.; Uotinen, V.O.; Davis, H.H.

    1979-08-01

    Five Babcock and Wilcox standard Mark B (15 x 15) fuel assemblies are being irradiated in Duke Power Company's Oconee Unit 1 reactor under a research and development program sponsored by the U.S. Department of Energy. Valuable experimental data on fuel performance characteristics at burnups of > 40,000 MWd/mtU will be obtained from these assemblies. This information, at a duty approximately 20% greater than that achieved by typical discharged assemblies, will be used to qualify standard Mark B fuel assemblies for extended burnups. Efforts during this period included fuel cycle design and reload licensing of Oconee 1 for cycle 5, in which the assemblies are being irradiated, and nondestructive examination of the assemblies during the refueling outage between cycles 4 and 5. The Oconee 1 cycle 5 startup tests proceeded in a routine manner, and the reactor has operated with a 92% capacity factor since completion of power escalation testing on November 10, 1978. Irradiation of the fuel assemblies is currently in progress

  5. Revisiting the Integrated Pressurized Thermal Shock Studies of an Aging Pressurized Water Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Bryson, J.W.; Dickson, T.L.; Malik, S.N.M.; Simonen, F.A.

    1999-08-01

    The Integrated Pressurized Thermal Shock (IPTS) studies were a series of studies performed in the early-mid 1980s as part of an NRC-organized comprehensive research project to confirm the technical bases for the pressurized thermal shock (PTS) rule, and to aid in the development of guidance for licensee plant-specific analyses. The research project consisted of PTS pilot analyses for three PWRs: Oconee Unit 1, designed by Babcock and Wilcox; Calvert Cliffs Unit 1, designed by Combustion Engineering; and H.B. Robinson Unit 2, designed by Westinghouse. The primary objectives of the IPTS studies were (1) to provide for each of the three plants an estimate of the probability of a crack propagating through the wall of a reactor pressure vessel (RPV) due to PTS; (2) to determine the dominant overcooling sequences, plant features, and operator actions and the uncertainty in the plant risk due to PTS; and (3) to evaluate the effectiveness of potential corrective actions. The NRC is currently evaluating the possibility of revising current PTS regulatory guidance. Technical bases must be developed to support any revisions. In the years since the results of IPTS studies were published, the fracture mechanics model, the embrittlement database, embrittlement correlation, inputs for flaw distributions, and the probabilistic fracture mechanics (PFM) computer code have been refined. An ongoing effort is underway to determine the impact of these fracture-technology refinements on the conditional probabilities of vessel failure calculated in the IPTS Studies. This paper discusses the results of these analyses performed for one of these plants.

  6. D and DR Reactors

    Data.gov (United States)

    Federal Laboratory Consortium — The world's second full-scale nuclear reactor was the D Reactor at Hanford which was built in the early 1940's and went operational in December of 1944.D Reactor ran...

  7. The reactor accident in the Three Mile Island-2 reactor plant. (Harrisburg, USA)

    International Nuclear Information System (INIS)

    Anon.

    1979-01-01

    The presentation of the accident development is based on the bulletin of the Atomic Industrial Forum (AIF) from the 6th of April, 1979. In addition, there are some short pieces of information, No. 14 and 15 of the association for reactor security as well as written and verbal information of the firms Brown, Boveri and Co/Babcock-Brown Boveri Reaktor GmbH (BBC/BBR). (orig.) [de

  8. A THREE-DIMENSIONAL BABCOCK-LEIGHTON SOLAR DYNAMO MODEL

    International Nuclear Information System (INIS)

    Miesch, Mark S.; Dikpati, Mausumi

    2014-01-01

    We present a three-dimensional (3D) kinematic solar dynamo model in which poloidal field is generated by the emergence and dispersal of tilted sunspot pairs (more generally bipolar magnetic regions, or BMRs). The axisymmetric component of this model functions similarly to previous 2.5 dimensional (2.5D, axisymmetric) Babcock-Leighton (BL) dynamo models that employ a double-ring prescription for poloidal field generation but we generalize this prescription into a 3D flux emergence algorithm that places BMRs on the surface in response to the dynamo-generated toroidal field. In this way, the model can be regarded as a unification of BL dynamo models (2.5D in radius/latitude) and surface flux transport models (2.5D in latitude/longitude) into a more self-consistent framework that builds on the successes of each while capturing the full 3D structure of the evolving magnetic field. The model reproduces some basic features of the solar cycle including an 11 yr periodicity, equatorward migration of toroidal flux in the deep convection zone, and poleward propagation of poloidal flux at the surface. The poleward-propagating surface flux originates as trailing flux in BMRs, migrates poleward in multiple non-axisymmetric streams (made axisymmetric by differential rotation and turbulent diffusion), and eventually reverses the polar field, thus sustaining the dynamo. In this Letter we briefly describe the model, initial results, and future plans

  9. The Therapeutic Effects of the Chinese Herbal Medicine, Lang Chuang Fang Granule, on Lupus-Prone MRL/lpr Mice

    Directory of Open Access Journals (Sweden)

    Kai-Peng Huang

    2016-01-01

    Full Text Available Systemic lupus erythematosus (SLE is a chronic autoimmune disease that leads to severe multiorgan damage. Lang Chuang Fang (LCF is a Chinese herbal medicine that is clinically prescribed for treating SLE. In this study, we examined the therapeutic effects of LCF granule on lupus-prone MRL/lpr mice. Female mice were randomly separated into six groups, and LCF treatment groups received LCF granule at the dosage of 0.97 g/kg/d, 1.95 g/kg/d, and 3.90 g/kg/d, respectively. Here, we found that, compared to the MRL/lpr mice, both the spleen coefficient and thymus coefficient were reduced in the LCF granule-treated mice. There was a marked downregulation in CRP and anti-dsDNA autoantibody and an evident upregulation of CH50 in LCF granule-treated mice. LCF granule treatment also obviously reduced the proteinuria, BUN, and SCr levels in MRL/lpr mice at the dosage of 0.97 g/kg/d, 1.95 g/kg/d, and 3.90 g/kg/d, indicating that LCF granule alleviated the renal injury of MRL/lpr mice. Furthermore, LCF granule decreased p65 NF-κB levels and increased Sirt1 and Nrf2 levels in the kidney tissues of MRL/lpr mice, which might elucidate the beneficial effects of LCF on lupus nephritis. In conclusion, this study demonstrates that LCF granule has therapeutic effects on lupus-prone MRL/lpr mice.

  10. GPU v. B and W lawsuit review and its effect on TMI-1 (Docket 50-289)

    International Nuclear Information System (INIS)

    1983-09-01

    This report documents a review by the Nuclear Regulatory Commission (NRC) staff of the General Public Utilities Corporation, et al. v. the Babcock and Wilcox Company, et al. (GPU v. B and W) lawsuit record to assess whether any of the staff's previous conclusions or their principal bases presented at the Three Mile Island Unit 1 (TMI-1) restart hearing, supporting restart of TMI-1, should be amended in light of the information contained in the lawsuit record. Details of the lawsuit record are provided in the appendices contained in Volume II of this report

  11. GPU v. B and W lawsuit review and its effect on TMI-1 (Docket 50-289)

    Energy Technology Data Exchange (ETDEWEB)

    1983-09-01

    This report documents a review by the Nuclear Regulatory Commission (NRC) staff of the General Public Utilities Corporation, et al. v. the Babcock and Wilcox Company, et al. (GPU v. B and W) lawsuit record to assess whether any of the staff's previous conclusions or their principal bases presented at the Three Mile Island Unit 1 (TMI-1) restart hearing, supporting restart of TMI-1, should be amended in light of the information contained in the lawsuit record. Details of the lawsuit record are provided in the appendices contained in Volume II of this report.

  12. Green County Nuclear Power Plant. License application

    International Nuclear Information System (INIS)

    1975-07-01

    The Green County reactor, a PWR to be supplied by Babcock and Wilcox, will be a baseload generating facility planned to provide for mass transit and other public agency electrical needs. The plant is scheduled for completion by 1983 and will have a generating capacity of about 1200 MW(e). (FS)

  13. Clinical efficacy of buprenorphine to minimize distress in MRL/lpr mice

    Science.gov (United States)

    Swenson, Julie; Olgun, Selen; Radjavi, Ali; Kaur, Taranjit; Reilly, Christopher M.

    2007-01-01

    MRL/MpJ-Faslpr (MRL/lpr) mice are an accepted animal model to study human systemic lupus erythematosus. We tested if a commonly used analgesic (buprenorphine hydrochloride) would reduce pain and distress in these mice without impacting the progression of autoimmune disease. Female MRL/lpr mice were randomly separated into four groups. Experimental groups received cyclophosphamide (25 mg/kg i.p. weekly), buprenorphine (0.09 mg/kg/mouse/day via drinking water), or cyclophosphamide + buprenorphine from 11 - 21 weeks of age. Controls received no treatments. Mice were monitored daily by a licensed veterinarian (blinded observer) and assigned a score weekly on parameters associated with pain and distress as well as progression of disease. Proteinuria was measured weekly, and serum anti-dsDNA antibody levels were determined at 11, 15, and 18 weeks of age. At 21 weeks of age, the animals were euthanized and the kidneys and spleens were removed for evaluation. Regardless of the parameter observed, buprenorphine did not significantly decrease distress when compared to the controls. Buprenorphine did not alter the progression of autoimmune disease, based on characteristics of splenic architecture and splenocyte cell profiles, development of lymphadenopathy, or kidney histology as compared to controls. This study indicates that buprenorphine at this dose and route of administration was ineffective in reducing distress associated with disease progression in the MRL/lpr strain. More studies are needed to determine if, at a different dose or route, buprenorphine would be useful as adjunctive therapy in reducing distress in MRL/lpr mice. PMID:17490635

  14. [Therapeutic effect of total glucosides of paeony on lupus nephritis in MRL/lpr mice].

    Science.gov (United States)

    Ding, Zhao-Xia; Yang, Shao-Feng; Wu, Qi-Fu; Lu, Ying; Chen, Yu-Yao; Nie, Xiao-Li; Jie, Hong-Yu; Qi, Jing-Min; Wang, Fan-Sheng

    2011-04-01

    To observe the therapeutic effect of total glucosides of paeony (TGP) on lupus nephritis (LN) in MRL/lpr mice. MRL/lpr mice with lupus nephritis were randomized into model group and TGP group. The urinary protein content was detected using Coomassie brilliant blue, and the serum levels of IgG anti-double-stranded DNA (dsDNA) antibodies and antinuclear antibodies (ANA) were measured by enzyme-linked immunosorbent assay (ELISA). The changes in the renal pathology were examined microscopically, and the spleen and thymus were weighed to calculate the spleen and thymus indexes. At 15 and 30 days after TGP administration, the urinary protein content in the TGP group was significantly lower than that in the model group (PTGP treatment significantly lowered the serum levels of anti-dsDNA antibodies and ANA and the weight and index of spleen (PTGP treatment, the urinary protein content and the levels of anti-dsDNA antibodies and ANA decreased significantly at 15 and 30 days after TGP administration (PTGP administration, the urinary protein content was significantly lowered in the TGP group as compared to that at 15 days (PTGP can reduce urinary protein content and serum levels of anti-dsDNA antibodies and ANA, and lessen renal pathology in MRL/lpr mice with lupus nephritis, suggesting its therapeutic effect on lupus nephritis.

  15. Naja naja atra Venom Protects against Manifestations of Systemic Lupus Erythematosus in MRL/lpr Mice

    Directory of Open Access Journals (Sweden)

    Jiali Zhu

    2014-01-01

    Full Text Available Systemic lupus erythematosus (SLE is an autoimmune disease and effective therapy for this pathology is currently unavailable. We previously reported that oral administration of Naja naja atra venom (NNAV had anti-inflammatory and immune regulatory actions. We speculated that NNAV may have therapeutic effects in MRL/lpr SLE mice. Twelve-week-old MRL/lpr mice received oral administration of NNAV (20, 40, and 80 μg/kg or Tripterygium wilfordii polyglycosidium (10 mg/kg daily for 16 weeks. The effects of NNAV on SLE manifestations, including skin erythema, proteinuria, and anxiety-like behaviors, were assessed with visual inspection and Multistix 8 SG strips and open field test, respectively. The pathology of spleen and kidney was examined with H&E staining. The changes in autoimmune antibodies and cytokines were determined with ELISA kits. The results showed that NNAV protected against the manifestation of SLE, including skin erythema and proteinuria. In addition, although no apparent histological change was found in liver and heart in MRL/lpr SLE mice, NNAV reduced the levels of glutamate pyruvate transaminase and creatine kinase. Furthermore, NNAV increased serum C3 and reduced concentrations of circulating globulin, anti-dsDNA antibody, and inflammatory cytokines IL-6 and TNF-α. NNAV also reduced lymphadenopathy and renal injury. These results suggest that NNAV may have therapeutic values in the treatment of SLE by inhibiting autoimmune responses.

  16. Naja naja atra Venom Protects against Manifestations of Systemic Lupus Erythematosus in MRL/lpr Mice.

    Science.gov (United States)

    Zhu, Jiali; Cui, Kui; Kou, Jianqun; Wang, Shuzhi; Xu, Yinli; Ding, Zhihui; Han, Rong; Qin, Zhenghong

    2014-01-01

    Systemic lupus erythematosus (SLE) is an autoimmune disease and effective therapy for this pathology is currently unavailable. We previously reported that oral administration of Naja naja atra venom (NNAV) had anti-inflammatory and immune regulatory actions. We speculated that NNAV may have therapeutic effects in MRL/lpr SLE mice. Twelve-week-old MRL/lpr mice received oral administration of NNAV (20, 40, and 80 μg/kg) or Tripterygium wilfordii polyglycosidium (10 mg/kg) daily for 16 weeks. The effects of NNAV on SLE manifestations, including skin erythema, proteinuria, and anxiety-like behaviors, were assessed with visual inspection and Multistix 8 SG strips and open field test, respectively. The pathology of spleen and kidney was examined with H&E staining. The changes in autoimmune antibodies and cytokines were determined with ELISA kits. The results showed that NNAV protected against the manifestation of SLE, including skin erythema and proteinuria. In addition, although no apparent histological change was found in liver and heart in MRL/lpr SLE mice, NNAV reduced the levels of glutamate pyruvate transaminase and creatine kinase. Furthermore, NNAV increased serum C3 and reduced concentrations of circulating globulin, anti-dsDNA antibody, and inflammatory cytokines IL-6 and TNF-α. NNAV also reduced lymphadenopathy and renal injury. These results suggest that NNAV may have therapeutic values in the treatment of SLE by inhibiting autoimmune responses.

  17. The Role of Magnetic Buoyancy in a Babcock-Leighton Type Solar ...

    Indian Academy of Sciences (India)

    tribpo

    J. Astrophys. Astr. (2000) 21, 381-385. The Role of Magnetic Buoyancy in a Babcock-Leighton. Type Solar Dynamo. Dibyendu Nandy* & Arnab Rai Choudhuri, ... model of the solar dynamo—which draws inspiration from the Babcock- .... are still of rather exploratory nature, since none of the authors have succeeded yet.

  18. B and W model boiler tests: effect of temperature on IGA rate. Initial and post-1878 operating conditions of the model boilers

    International Nuclear Information System (INIS)

    Anon.

    1986-01-01

    The Babcock and Wilcox (B and W) model boiler operated with 10 ppm weekly injections of NaOH for 41,900 hours (4.8 years). The model boiler operating conditions are given. Tube No. 24 failed by caustic intergranular attack/stress corrosion cracking (IGA/SCC) at the steam-water zone. IGA defect depths on tube 24 is compared at different locations, which also have different temperature conditions. The specific locations are: steam/water zone, drilled baffle plate, and lower tube sheet crevice. In all locations caustic will concentrate (although to different concentration levels). Nevertheless, an effect of temperature on IGA rate can be estimated. The degree of attack relative to the location and environment is shown. SEM fractographs illustrate the completely intergranular failure of Tube 24. A summary of the estimated results is presented. These results show the estimated IGA rate as a function of primary/secondary temperature and estimated caustic concentration. Details of the failure analysis of the model boiler can be found in the final report Destructive Examination of Babcock and Wilcox's Model Boiler for Intergranular Attack (IGA) on Tubes, EPRI S302-6, J.L.; Barna and L.W. Sarver

  19. White Paper on Data Repository Reorganization Proposal for the xLPR Project

    Energy Technology Data Exchange (ETDEWEB)

    Klasky, Hilda B [ORNL; Williams, Paul T [ORNL; Bass, Bennett Richard [ORNL

    2012-09-01

    As the xLPR project moves along, it is important to properly manage the knowledge generated by the different groups. We focus specifically on the knowledge and communications written in files, including general documents, source code and executable files. Data generated through the project are different in nature and, for this reason, need to be treated differently. To that end, ORNL put in place a series of tools that facilitate proper storage and management of project data, document and code changes, group collaboration, knowledge transfer, transparency, accountability and auditability. This paper describes the approaches/tools that we recommend for moving the project forward on knowledge management.

  20. Role of Fas and Treg Cells in Fracture Healing as Characterized in the Fas-Deficient (lpr) Mouse Model of Lupus†

    Science.gov (United States)

    Al-Sebaei, Maisa O; Daukss, Dana M; Belkina, Anna C; Kakar, Sanjeev; Wigner, Nathan A; Cusher, Daniel; Graves, Dana; Einhorn, Thomas; Morgan, Elise; Gerstenfeld, Louis C

    2014-01-01

    Previous studies showed that loss of tumor necrosis factor α (TNFα) signaling delayed fracture healing by delaying chondrocyte apoptosis and cartilage resorption. Mechanistic studies showed that TNFα induced Fas expression within chondrocytes; however, the degree to which chondrocyte apoptosis is mediated by TNFα alone or dependent on the induction of Fas is unclear. This question was addressed by assessing fracture healing in Fas-deficient B6.MRL/Faslpr/J mice. Loss of Fas delayed cartilage resorption but also lowered bone fraction in the calluses. The reduced bone fraction was related to elevated rates of coupled bone turnover in the B6.MRL/Faslpr/J calluses, as evidenced by higher osteoclast numbers and increased osteogenesis. Analysis of the apoptotic marker caspase 3 showed fewer positive chondrocytes and osteoclasts in calluses of B6.MRL/Faslpr/J mice. To determine if an active autoimmune state contributed to increased bone turnover, the levels of activated T cells and Treg cells were assessed. B6.MRL/Faslpr/J mice had elevated Treg cells in both spleens and bones of B6.MRL/Faslpr/J but decreased percentage of activated T cells in bone tissues. Fracture led to ∼30% to 60% systemic increase in Treg cells in both wild-type and B6.MRL/Faslpr/J bone tissues during the period of cartilage formation and resorption but either decreased (wild type) or left unchanged (B6.MRL/Faslpr/J) the numbers of activated T cells in bone. These results show that an active autoimmune state is inhibited during the period of cartilage resorption and suggest that iTreg cells play a functional role in this process. These data show that loss of Fas activity specifically in chondrocytes prolonged the life span of chondrocytes and that Fas synergized with TNFα signaling to mediate chondrocyte apoptosis. Conversely, loss of Fas systemically led to increased osteoclast numbers during later periods of fracture healing and increased osteogenesis. These findings suggest that retention

  1. Prolactin Levels Correlate with Abnormal B Cell Maturation in MRL and MRL/lpr Mouse Models of Systemic Lupus Erythematosus-Like Disease

    Directory of Open Access Journals (Sweden)

    Maria Victoria Legorreta-Haquet

    2013-01-01

    Full Text Available Prolactin (PRL plays an important role in modulating the immune response. In B cells, PRL enhances antibody production, including antibodies with self-specificity. In this study, our aims were to determine the level of PRL receptor expression during bone-marrow B-cell development and to assess whether the presence of high PRL serum concentrations influences absolute numbers of developing populations and disease outcome in lupus-prone murine models. We observed that the PRL-receptor is expressed in early bone-marrow B-cell; the expression in lupus-prone mice, which had the highest level of expression in pro-B cells and immature cells, differed from that in wild-type mice. These expression levels did not significantly change in response to hyperprolactinemia; however, populations of pro-B and immature cells from lupus-prone strains showed a decrease in the absolute numbers of cells with high PRL-receptor expression in response to PRL. Because immature self-reactive B cells are constantly being eliminated, we assessed the expression of survival factor BIRC5, which is more highly expressed in both pro-B and immature B-cells in response to PRL and correlates with the onset of disease. These results identify an important role of PRL in the early stages of the B-cell maturation process: PRL may promote the survival of self-reactive clones.

  2. Prolactin Levels Correlate with Abnormal B Cell Maturation in MRL and MRL/lpr Mouse Models of Systemic Lupus Erythematosus-Like Disease

    Science.gov (United States)

    Legorreta-Haquet, Maria Victoria; Flores-Fernández, Rocio; Blanco-Favela, Francisco; Fuentes-Pananá, Ezequiel M; Chávez-Sánchez, Luis; Hernández-González, Rafael; Tesoro-Cruz, Emiliano; Arriaga-Pizano, Lourdes; Chávez-Rueda, Adriana Karina

    2013-01-01

    Prolactin (PRL) plays an important role in modulating the immune response. In B cells, PRL enhances antibody production, including antibodies with self-specificity. In this study, our aims were to determine the level of PRL receptor expression during bone-marrow B-cell development and to assess whether the presence of high PRL serum concentrations influences absolute numbers of developing populations and disease outcome in lupus-prone murine models. We observed that the PRL-receptor is expressed in early bone-marrow B-cell; the expression in lupus-prone mice, which had the highest level of expression in pro-B cells and immature cells, differed from that in wild-type mice. These expression levels did not significantly change in response to hyperprolactinemia; however, populations of pro-B and immature cells from lupus-prone strains showed a decrease in the absolute numbers of cells with high PRL-receptor expression in response to PRL. Because immature self-reactive B cells are constantly being eliminated, we assessed the expression of survival factor BIRC5, which is more highly expressed in both pro-B and immature B-cells in response to PRL and correlates with the onset of disease. These results identify an important role of PRL in the early stages of the B-cell maturation process: PRL may promote the survival of self-reactive clones. PMID:24454471

  3. A nuclear power unit with a Babcock type steam generating system-analysis of the break-down in the Three Mile Island power plant

    International Nuclear Information System (INIS)

    Werner, A.

    1980-01-01

    Installations of the primary and the secondary circuits and basic automatic control and protection systems for a nuclear power unit with Babcock type vertical, once-through steam generator are described. On this background the course of the break-down in the Three Mile Island power plant at Harrisburg is presented and analysed. (author)

  4. Environmental Assessment: Geothermal Energy Geopressure Subprogram. Gulf Coast Well Drilling and Testing Activity (Frio, Wilcox, and Tuscaloosa Formations, Texas and Louisiana)

    Energy Technology Data Exchange (ETDEWEB)

    None

    1981-09-01

    The Department of Energy (DOE) has initiated a program to evaluate the feasibility of developing the geothermal-geopressured energy resources of the Louisiana-Texas Gulf Coast. As part of this effort, DOE is contracting for the drilling of design wells to define the nature and extent of the geopressure resource. At each of several sites, one deep well (4000-6400 m) will be drilled and flow tested. One or more shallow wells will also be drilled to dispose of geopressured brines. Each site will require about 2 ha (5 acres) of land. Construction and initial flow testing will take approximately one year. If initial flow testing is successful, a continuous one-year duration flow test will take place at a rate of up to 6400 m{sup 3} (40,000 bbl) per day. Extensive tests will be conducted on the physical and chemical composition of the fluids, on their temperature and flow rate, on fluid disposal techniques, and on the reliability and performance of equipment. Each project will require a maximum of three years to complete drilling, testing, and site restoration.

  5. NTRE extended life feasibility assessment

    Science.gov (United States)

    Results of a feasibility analysis of a long life, reusable nuclear thermal rocket engine are presented in text and graph form. Two engine/reactor concepts are addressed: the Particle Bed Reactor (PBR) design and the Commonwealth of Independent States (CIS) concept. Engine design, integration, reliability, and safety are addressed by various members of the NTRE team from Aerojet Propulsion Division, Energopool (Russia), and Babcock & Wilcox.

  6. Wilcox sandstone reservoirs in the deep subsurface along the Texas Gulf Coast: their potential for production of geopressured geothermal energy. Report of Investigations No. 117

    Energy Technology Data Exchange (ETDEWEB)

    Debout, D.G.; Weise, B.R.; Gregory, A.R.; Edwards, M.B.

    1982-01-01

    Regional studies of the lower Eocene Wilcox Group in Texas were conducted to assess the potential for producing heat energy and solution methane from geopressured fluids in the deep-subsurface growth-faulted zone. However, in addition to assembling the necessary data for the geopressured geothermal project, this study has provided regional information of significance to exploration for other resources such as lignite, uranium, oil, and gas. Because the focus of this study was on the geopressured section, emphasis was placed on correlating and mapping those sandstones and shales occurring deeper than about 10,000 ft. The Wilcox and Midway Groups comprise the oldest thick sandstone/shale sequence of the Tertiary of the Gulf Coast. The Wilcox crops out in a band 10 to 20 mi wide located 100 to 200 mi inland from the present-day coastline. The Wilcox sandstones and shales in the outcrop and updip shallow subsurface were deposited primarily in fluvial environments; downdip in the deep subsurface, on the other hand, the Wilcox sediments were deposited in large deltaic systems, some of which were reworked into barrier-bar and strandplain systems. Growth faults developed within the deltaic systems, where they prograded basinward beyond the older, stable Lower Cretaceous shelf margin onto the less stable basinal muds. Continued displacement along these faults during burial resulted in: (1) entrapment of pore fluids within isolated sandstone and shale sequences, and (2) buildup of pore pressure greater than hydrostatic pressure and development of geopressure.

  7. Chemical evolution of groundwater in the Wilcox aquifer of the northern Gulf Coastal Plain, USA

    Science.gov (United States)

    Haile, Estifanos; Fryar, Alan E.

    2017-12-01

    The Wilcox aquifer is a major groundwater resource in the northern Gulf Coastal Plain (lower Mississippi Valley) of the USA, yet the processes controlling water chemistry in this clastic aquifer have received relatively little attention. The current study combines analyses of solutes and stable isotopes in groundwater, petrography of core samples, and geochemical modeling to identify plausible reactions along a regional flow path ˜300 km long. The hydrochemical facies evolves from Ca-HCO3 upgradient to Na-HCO3 downgradient, with a sequential zonation of terminal electron-accepting processes from Fe(III) reduction through SO4 2- reduction to methanogenesis. In particular, decreasing SO4 2- and increasing δ34S of SO4 2- along the flow path, as well as observations of authigenic pyrite in core samples, provide evidence of SO4 2- reduction. Values of δ13C in groundwater suggest that dissolved inorganic carbon is contributed both by oxidation of sedimentary organic matter and calcite dissolution. Inverse modeling identified multiple plausible sets of reactions between sampled wells, which typically involved cation exchange, pyrite precipitation, CH2O oxidation, and dissolution of amorphous Fe(OH)3, calcite, or siderite. These reactions are consistent with processes identified in previous studies of Atlantic Coastal Plain aquifers. Contrasts in groundwater chemistry between the Wilcox and the underlying McNairy and overlying Claiborne aquifers indicate that confining units are relatively effective in limiting cross-formational flow, but localized cross-formational mixing could occur via fault zones. Consequently, increased pumping in the vicinity of fault zones could facilitate upward movement of saline water into the Wilcox.

  8. Reactor

    International Nuclear Information System (INIS)

    Toyama, Masahiro; Kasai, Shigeo.

    1978-01-01

    Purpose: To provide a lmfbr type reactor wherein effusion of coolants through a loop contact portion is reduced even when fuel assemblies float up, and misloading of reactor core constituting elements is prevented thereby improving the reactor safety. Constitution: The reactor core constituents are secured in the reactor by utilizing the differential pressure between the high-pressure cooling chamber and low-pressure cooling chamber. A resistance port is formed at the upper part of a connecting pipe, and which is connect the low-pressure cooling chamber and the lower surface of the reactor core constituent. This resistance part is formed such that the internal sectional area of the connecting pipe is made larger stepwise toward the upper part, and the cylinder is formed larger so that it profiles the inner surface of the connecting pipe. (Aizawa, K.)

  9. Reactor

    International Nuclear Information System (INIS)

    Ikeda, Masaomi; Kashimura, Kazuo; Inoue, Kazuyuki; Nishioka, Kazuya.

    1979-01-01

    Purpose: To facilitate the construction of a reactor containment building, whereby the inspections of the outer wall of a reactor container after the completion of the construction of the reactor building can be easily carried out. Constitution: In a reactor accommodated in a container encircled by a building wall, a space is provided between the container and the building wall encircling the container, and a metal wall is provided in the space so that it is fitted in the building wall in an attachable or detatchable manner. (Aizawa, K.)

  10. Nuclear Reactors and Technology

    Energy Technology Data Exchange (ETDEWEB)

    Cason, D.L.; Hicks, S.C. [eds.

    1992-01-01

    This publication Nuclear Reactors and Technology (NRT) announces on a monthly basis the current worldwide information available from the open literature on nuclear reactors and technology, including all aspects of power reactors, components and accessories, fuel elements, control systems, and materials. This publication contains the abstracts of DOE reports, journal articles, conference papers, patents, theses, and monographs added to the Energy Science and Technology Database during the past month. Also included are US information obtained through acquisition programs or interagency agreements and international information obtained through the International Energy Agency`s Energy Technology Data Exchange or government-to-government agreements. The digests in NRT and other citations to information on nuclear reactors back to 1948 are available for online searching and retrieval on the Energy Science and Technology Database and Nuclear Science Abstracts (NSA) database. Current information, added daily to the Energy Science and Technology Database, is available to DOE and its contractors through the DOE Integrated Technical Information System. Customized profiles can be developed to provide current information to meet each user`s needs.

  11. Reactor instrumentation and control

    International Nuclear Information System (INIS)

    Wach, D.; Beraha, D.

    1980-01-01

    The methods for measuring radiation are shortly reviewed. The instrumentation for neutron flux measurement is classified into out-of-core and in-core instrumentation. The out-of-core instrumentation monitors the operational range from the subcritical reactor to full power. This large range is covered by several measurement channels which derive their signals from counter tubes and ionization chambers. The in-core instrumentation provides more detailed information on the power distribution in the core. The self-powered neutron detectors and the aeroball system in PWR reactors are discussed. Temperature and pressure measurement devices are briefly discussed. The different methods for leak detection are described. In concluding the plant instrumentation part some new monitoring systems and analysis methods are presented: early failure detection methods by noise analysis, acoustic monitoring and vibration monitoring. The presentation of the control starts from an qualitative assessment of the reactor dynamics. The chosen control strategy leads to the definition of the part-load diagram, which provides the set-points for the different control systems. The tasks and the functions of these control systems are described. In additiion to the control, a number of limiting systems is employed to keep the reactor in a safe operating region. Finally, an outlook is given on future developments in control, concerning mainly the increased application of process computers. (orig./RW)

  12. Wilcox 1:100000 Quad Hydrography DLGs

    Data.gov (United States)

    Earth Data Analysis Center, University of New Mexico — Digital line graph (DLG) data are digital representations of cartographic information. DLG's of map features are converted to digital form from maps and related...

  13. Results and Analysis of the Infrastructure Request for Information (DE-SOL-0008318)

    Energy Technology Data Exchange (ETDEWEB)

    Heidrich, Brenden John [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-07-01

    The Department of Energy (DOE) Office of Nuclear Energy (NE) released a request for information (RFI) (DE-SOL-0008318) for “University, National Laboratory, Industry and International Input on Potential Office of Nuclear Energy Infrastructure Investments” on April 13, 2015. DOE-NE solicited information on five specific types of capabilities as well as any others suggested by the community. The RFI proposal period closed on June 19, 2015. From the 26 responses, 34 individual proposals were extracted. Eighteen were associated with a DOE national laboratory, including Argonne National Laboratory (ANL), Brookhaven National Laboratory (BNL), Idaho National Laboratory (INL), Los Alamos National Laboratory (LANL), Pacific Northwest National Laboratory (PNNL) and Sandia National Laboratory (SNL). Oak Ridge National Laboratory (ORNL) was referenced in a proposal as a proposed capability location, although the proposal did not originate with ORNL. Five US universities submitted proposals (Massachusetts Institute of Technology, Pennsylvania State University, Rensselaer Polytechnic Institute, University of Houston and the University of Michigan). Three industrial/commercial institutions submitted proposals (AREVA NP, Babcock and Wilcox (B&W) and the Electric Power Research Institute (EPRI)). Eight major themes emerged from the submissions as areas needing additional capability or support for existing capabilities. Two submissions supported multiple areas. The major themes are: Advanced Manufacturing (AM), High Performance Computing (HPC), Ion Irradiation with X-Ray Diagnostics (IIX), Ion Irradiation with TEM Visualization (IIT), Radiochemistry Laboratories (RCL), Test Reactors, Neutron Sources and Critical Facilities (RX) , Sample Preparation and Post-Irradiation Examination (PIE) and Thermal-Hydraulics Test Facilities (THF).

  14. Nuclear reactors: physics and materials

    Energy Technology Data Exchange (ETDEWEB)

    Yadigaroglu, G

    2005-07-01

    In the form of a tutorial addressed to non-specialists, the article provides an introduction to nuclear reactor technology and more specifically to Light Water Reactors (LWR); it also shows where materials and chemistry problems are encountered in reactor technology. The basics of reactor physics are reviewed, as well as the various strategies in reactor design and the corresponding choices of materials (fuel, coolant, structural materials, etc.). A brief description of the various types of commercial power reactors follows. The design of LWRs is discussed in greater detail; the properties of light water as coolant and moderator are put in perspective. The physicochemical and metallurgical properties of the materials impose thermal limits that determine the performance and the maximum power a reactor can deliver. (author)

  15. Remission of systemic lupus erythematosus disease activity with regulatory cytokine interleukin (IL)-35 in Murphy Roths Large (MRL)/lpr mice.

    Science.gov (United States)

    Cai, Z; Wong, C K; Dong, J; Chu, M; Jiao, D; Kam, N W; Lam, C W K; Tam, L S

    2015-08-01

    The immunological mechanisms mediated by regulatory cytokine interleukin (IL)-35 are unclear in systemic lupus erythematosus (SLE). We investigated the frequency of CD4(+) CD25(+) forkhead box protein 3 (FoxP3)(+) regulatory T (Treg ) and IL-10(+) regulatory B (Breg ) cells and related immunoregulatory mechanisms in a female Murphy Roths Large (MRL)/lpr mouse model of spontaneous lupus-like disease, with or without IL-35 treatment. A remission of histopathology characteristics of lupus flare and nephritis was observed in the MRL/lpr mice upon IL-35 treatment. Accordingly, IL-35 and IL-35 receptor subunits (gp130 and IL-12Rβ2) and cytokines of MRL/lpr and BALB/c mice (normal controls) were measured. The increased anti-inflammatory cytokines and decreased proinflammatory cytokines were possibly associated with the restoration of Treg and Breg frequency in MRL/lpr mice with IL-35 treatment, compared to phosphate-buffered saline (PBS) treatment. mRNA expressions of Treg -related FoxP3, IL-35 subunit (p35 and EBI3) and soluble IL-35 receptor subunit (gp130 and IL12Rβ2) in splenic cells were up-regulated significantly in IL-35-treated mice. Compared with the PBS treatment group, IL-35-treated MRL/lpr mice showed an up-regulation of Treg -related genes and the activation of IL-35-related intracellular Janus kinase/signal transducer and activator of transcription signal pathways, thereby indicating the immunoregulatory role of IL-35 in SLE. These in vivo findings may provide a biochemical basis for further investigation of the regulatory mechanisms of IL-35 for the treatment of autoimmune-mediated inflammation. © 2015 British Society for Immunology.

  16. The multi-interlock and check of logical system for 5 MW low power reactor automatic rod

    International Nuclear Information System (INIS)

    Li Guangjian; Zhao Zengqiao

    1992-01-01

    The safety and reliability of the logical system for 5 MW LPR automatic rod are improved, because of using multi-interlock and manual check on line. The design character and function of the logical system are introduced

  17. Research in nuclear reactor theory and experimental reactors

    International Nuclear Information System (INIS)

    Pop-Jordanov, J.

    1978-01-01

    The paper is devoted to the possibilities of using experimental reactors for scientific research in nuclear power with a stress on problems in nuclear reactor theory. The stationary and nonstationary neutron fields, burnup prediction and analyses as well as fuel element development and the corresponding role of test-reactors were dealt with. It was shown that the investigations in nuclear reactor theory in Yugoslavia were developing continuously and in a useful interaction with experiments on research reactors. The needs for continuing the work on fundamental problems in neutron transport theory and on improving the calculation methods for thermal power reactors, together with the improvement of performances of existing research systems, were pointed out. A new quality in scientific work could be obtained dealing with the problems connected to a possible introduction of test-reactors, and fast systems later on. It was also pleaded for the corresponding orientations in fundamental sciences. (author) [sr

  18. Nuclear reactor kinetics and control

    International Nuclear Information System (INIS)

    Lewins, J.

    1978-01-01

    A consistent, integrated account of modern developments in the study of nuclear reactor kinetics and the problem of their efficient and safe control. It aims to prepare the student for advanced study and research or practical work in the field. Special features include treatments of noise theory, reliability theory and safety related studies. It covers all aspects of the operation and control of nuclear reactors, power and research and is complete in providing physical data methods of calculation and solution including questions of equipment reliability. The work uses illustrations of the main types of reactors in use in the UK, USA and Europe. Each chapter contains problems and worked examples suitable for course work and study. The subject is covered in chapters, entitled: introductory review; neutron and precursor equations; elementary solutions at low power; linear reactor process dynamics with feedback; power reactor control systems; fluctuations and reactor noise; safety and reliability; nonlinear systems (safety and control); analogue computing. (author)

  19. Supply strategy for SMR deployment

    International Nuclear Information System (INIS)

    Coccagna, A.F.

    2013-01-01

    This document provides a description of Babcock and Wilcox's deployment strategy for the mPower™ Small Modular Reactor from the perspective of Supply Chain and Manufacturing. A desirable future state of readiness is described as one which leverages and revitalizes an existing supply chain and manufacturing infrastructure, as well as leveraging an existing workforce of engineering, construction, and project management employees. B and W's mPower™ SMR value proposition offers many desired design and operating advantages to the SMR market. (author)

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

    International Nuclear Information System (INIS)

    Powell, J.R.

    1977-01-01

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

  1. FFTF and CRBRP reactor vessels

    International Nuclear Information System (INIS)

    Morgan, R.E.

    1977-01-01

    The Fast Flux Test Facility (FFTF) reactor vessel and the Clinch River Breeder Reactor Plant (CRBRP) reactor vessel each serve to enclose a fast spectrum reactor core, contain the sodium coolant, and provide support and positioning for the closure head and internal structure. Each vessel is located in its reactor cavity and is protected by a guard vessel which would ensure continued decay heat removal capability should a major system leak develop. Although the two plants have significantly different thermal power ratings, 400 megawatts for FFTF and 975 megawatts for CRBRP, the two reactor vessels are comparable in size, the CRBRP vessel being approximately 28% longer than the FFTF vessel. The FFTF vessel diameter was controlled by the space required for the three individual In-Vessel Handling Machines and Instrument Trees. Utilization of the triple rotating plug scheme for CRBRP refueling enables packaging of the larger CRBRP core in a vessel the same diameter as the FFTF vessel

  2. Coal geology of the Paleocene-Eocene Calvert Bluff Formation (Wilcox Group) and the Eocene Manning Formation (Jackson Group) in east-central Texas; field trip guidebook for the Society for Organic Petrology, Twelfth Annual Meeting, The Woodlands, Texas, August 30, 1995

    Science.gov (United States)

    Warwick, Peter D.; Crowley, Sharon S.

    1995-01-01

    The Jackson and Wilcox Groups of eastern Texas (fig. 1) are the major lignite producing intervals in the Gulf Region. Within these groups, the major lignite-producing formations are the Paleocene-Eocene Calvert Bluff Formation (Wilcox) and the Eocene Manning Formation (Jackson). According to the Keystone Coal Industry Manual (Maclean Hunter Publishing Company, 1994), the Gulf Coast basin produces about 57 million short tons of lignite annually. The state of Texas ranks number 6 in coal production in the United States. Most of the lignite is used for electric power generation in mine-mouth power plant facilities. In recent years, particular interest has been given to lignite quality and the distribution and concentration of about a dozen trace elements that have been identified as potential hazardous air pollutants (HAPs) by the 1990 Clean Air Act Amendments. As pointed out by Oman and Finkelman (1994), Gulf Coast lignite deposits have elevated concentrations of many of the HAPs elements (Be, Cd, Co, Cr, Hg, Mn, Se, U) on a as-received gm/mmBtu basis when compared to other United States coal deposits used for fuel in thermo-electric power plants. Although regulations have not yet been established for acceptable emissions of the HAPs elements during coal burning, considerable research effort has been given to the characterization of these elements in coal feed stocks. The general purpose of the present field trip and of the accompanying collection of papers is to investigate how various aspects of east Texas lignite geology might collectively influence the quality of the lignite fuel. We hope that this collection of papers will help future researchers understand the complex, multifaceted interrelations of coal geology, petrology, palynology and coal quality, and that this introduction to the geology of the lignite deposits of east Texas might serve as a stimulus for new ideas to be applied to other coal basins in the U.S. and abroad.

  3. Inorganic membranes and catalytic reactors

    OpenAIRE

    Rangel, Maria do Carmo

    1997-01-01

    Membrane reactors are reviewed with emphasis in their applications in catalysis field. The basic principles of these systems are presented as well as a historical development. The several kinds of catalytic membranes and their preparations are discussed including the problems, needs and challenges to be solved in order to use these reactors in commercial processes. Some applications of inorganic membrane reactors are also shown. It was concluded that these systems have a great potential for i...

  4. Nuclear power reactors: reactor safety and military and civil defence

    International Nuclear Information System (INIS)

    Hvinden, T.

    1976-01-01

    The formation of fission products and plutonium in reactors is briefly described, followed by a short general discussion of reactor safety. The interaction of reactor safety and radioactive release considerations with military and civil defence is thereafter discussed. Reactors and other nuclear plants are factors which must be taken into account in the defence of the district around the site, and as potential targets of both conventional and guerilla attacks and sabotage, requiring special defence. The radiological hazards arising from serious damage to a power reactor by conventional weapons are briefly discussed, and the benefits of underground siting evaluated. Finally the author discusses the significance of the IAEA safeguards work as a preventive factor. (JIW)

  5. In service inspection of the reactor pressure vessel coolant and moderator nozzles at Atucha 1. 1998/1999 outages

    International Nuclear Information System (INIS)

    Antonaccio, Carlos; Conde, Alberto; Fittipaldi, Andres H.; Maniotti, Jorge; Moliterno, Gabriel E.

    2000-01-01

    During the August 1998 and the August 1999 Atucha 1 outages, two areas were inspected on the Reactor Pressure Vessel: the nozzle inner radii and the nozzle shell welds on all 3 moderator nozzles and all 4 main coolant nozzles. The inspections themselves were carried out by Mitsui Babcock Energy Limited from Scotland. The coordination, maintenance assistant and mounting of the manipulator devices over the nozzles were carried out by NASA personnel. Although it was not the first time the nozzle shell welds were inspected, due to the technologies advances in the ultrasonic field and in the inspection manipulators (magnetic ones), it was possible to inspect more volume than in previous inspections. In the other hand, it was the first time NASA was able to inspect the inner radii. In this last case the mayor problems to inspect them were the nozzles geometry and the small space available to install manipulators. The result of the inspections were: 1) There were no reportable indications at any of the inner radii inspected; 2) The inspection of nozzle to shell welds in main-coolant nozzles R3 and R4 detected flaws (one in each nozzle) which were reported as exceeding the dimensions specified as the acceptance level under Table IWB 3512-1, Section XI of the ASME code. Subsequent analysis requested by NASA and performed by Mitsui Babcock, demonstrated that the flaws were over dimensioned and could be explained as due to 'point' flaws. The analysis was based on theoretical mathematic model and experimental trials. Therefore their dimension were under the acceptance level of the ASME XI code. Although the Mitsui Babcock analysis, and at the same time it was in progress, it was assumed that the flaws were as they were originally presented (exceeding the acceptance level). NASA asked SIEMENS/KWU, the designer of the plant, to perform the fracture assessment according to ASME XI App. A. The assessment shows that the expected crack growth is negligibly small and the safety

  6. Comparison between TRU burning reactors and commercial fast reactor

    International Nuclear Information System (INIS)

    Fujimura, Koji; Sanda, Toshio; Ogawa, Takashi

    2001-03-01

    Research and development for stabilizing or shortening the radioactive wastes including in spent nuclear fuel are widely conducted in view point of reducing the environmental impact. Especially it is effective way to irradiate and transmute long-lived TRU by fast reactors. Two types of loading way were previously proposed. The former is loading relatively small amount of TRU in all commercial fast reactors and the latter is loading large amount of TRU in a few TRU burning reactors. This study has been intended to contribute to the feasibility studies on commercialized fast reactor cycle system. The transmutation and nuclear characteristics of TRU burning reactors were evaluated and compared with those of conventional transmutation system using commercial type fast reactor based upon the investigation of technical information about TRU burning reactors. Major results are summarized as follows. (1) Investigation of technical information about TRU burning reactors. Based on published reports and papers, technical information about TRU burning reactor concepts transmutation system using convectional commercial type fast reactors were investigated. Transmutation and nuclear characteristics or R and D issue were investigated based on these results. Homogeneously loading of about 5 wt% MAs on core fuels in the conventional commercial type fast reactor may not cause significant impact on the nuclear core characteristics. Transmutation of MAs being produced in about five fast reactors generating the same output is feasible. The helium cooled MA burning fast reactor core concept propose by JAERI attains criticality using particle type nitride fuels which contain more than 60 wt% MA. This reactor could transmute MAs being produced in more than ten 1000 MWe-LWRs. Ultra-long life core concepts attaining more than 30 years operation without refueling by utilizing MA's nuclear characteristics as burnable absorber and fertile nuclides were proposed. Those were pointed out that

  7. The breeder reactor and Europe

    International Nuclear Information System (INIS)

    Daglish, J.

    1979-01-01

    A report is given of a conference on the breeder reactor and Europe held in Lucerne, Switzerland from 14 - 17 October 1979 sponsored by the Swiss Association for Atomic Energy and the Association of European Atomic Forums. The underlying theme of the conference was the question that if nuclear power is to play a major role in meeting world energy needs in the long term, thermal reactors must in time be complemented with more advanced reactor systems that conserve uranium resources which are huge but not unlimited. This is not questioned; disagreement begins with discussion of the desirability of the breeder, and how fast and how far the introduction of such reactors should go. Aspects considered at the conference which are especially dealt with in this review are; why breed, commercial aspects, alternatives to the LMFBR, how to build a fast reactor, the breeder programmes in Europe, Britain, the Soviet Union, Japan and the United States. (U.K.)

  8. Ageing implementation and refurbishment development at the IEA-R1 nuclear research reactor: a 15 years experience

    International Nuclear Information System (INIS)

    Cardenas, Jose Patricio N.; Ricci Filho, Walter; Carvalho, Marcos R. de; Berretta, Jose Roberto; Marra Neto, Adolfo

    2011-01-01

    IPEN (Instituto de Pesquisas Energeticas e Nucleares) is a nuclear research center established into the Secretary of Science and Technology from the government of the state of Sao Paulo, and administered both technically and financially by Comissao Nacional de Energia Nuclear (CNEN), a federal government organization under the Ministry of Science and Technology. The institute is located inside the campus of the University of Sao Paulo, Sao Paulo city, Brazil. One of major nuclear facilities at IPEN is the IEA-R1 nuclear research reactor. It is the unique Brazilian research reactor with substantial power level suitable for application with research in physics, chemistry, biology and engineering, as well as radioisotope production for medical and other applications. Designed and built by Babcok-Wilcox, in accordance with technical specifications established by the Brazilian Nuclear Energy Commission, and financed by the US Atoms for Peace Program, it is a swimming pool type reactor, moderated and cooled by light water and uses graphite and beryllium as reflector elements. The first criticality was achieved on September 16, 1957 and the reactor is currently operating at 4.0 MW on a 64h per week cycle. Since 1996, an IEA-R1 reactor ageing study was established at the Research Reactor Center (CRPq) related with general deterioration of components belonging to some operational systems, as cooling towers from secondary cooling system, piping and pumps, sample irradiation devices, radiation monitoring system, fuel elements, rod drive mechanisms, nuclear and process instrumentation and safety operational system. Although basic structures are almost the same as the original design, several improvements and modifications in components, systems and structures had been made along reactor life. This work aims to show the development of the ageing program in the IEA-R1 reactor and the upgrading (modernization) that was carried out, concerning several equipment and system in the

  9. Ageing implementation and refurbishment development at the IEA-R1 nuclear research reactor: a 15 years experience

    Energy Technology Data Exchange (ETDEWEB)

    Cardenas, Jose Patricio N.; Ricci Filho, Walter; Carvalho, Marcos R. de; Berretta, Jose Roberto; Marra Neto, Adolfo, E-mail: ahiru@ipen.b, E-mail: wricci@ipen.b, E-mail: carvalho@ipen.b, E-mail: jrretta@ipen.b, E-mail: amneto@ipen.b [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2011-07-01

    IPEN (Instituto de Pesquisas Energeticas e Nucleares) is a nuclear research center established into the Secretary of Science and Technology from the government of the state of Sao Paulo, and administered both technically and financially by Comissao Nacional de Energia Nuclear (CNEN), a federal government organization under the Ministry of Science and Technology. The institute is located inside the campus of the University of Sao Paulo, Sao Paulo city, Brazil. One of major nuclear facilities at IPEN is the IEA-R1 nuclear research reactor. It is the unique Brazilian research reactor with substantial power level suitable for application with research in physics, chemistry, biology and engineering, as well as radioisotope production for medical and other applications. Designed and built by Babcok-Wilcox, in accordance with technical specifications established by the Brazilian Nuclear Energy Commission, and financed by the US Atoms for Peace Program, it is a swimming pool type reactor, moderated and cooled by light water and uses graphite and beryllium as reflector elements. The first criticality was achieved on September 16, 1957 and the reactor is currently operating at 4.0 MW on a 64h per week cycle. Since 1996, an IEA-R1 reactor ageing study was established at the Research Reactor Center (CRPq) related with general deterioration of components belonging to some operational systems, as cooling towers from secondary cooling system, piping and pumps, sample irradiation devices, radiation monitoring system, fuel elements, rod drive mechanisms, nuclear and process instrumentation and safety operational system. Although basic structures are almost the same as the original design, several improvements and modifications in components, systems and structures had been made along reactor life. This work aims to show the development of the ageing program in the IEA-R1 reactor and the upgrading (modernization) that was carried out, concerning several equipment and system in the

  10. Fuel element production at BWX technologies

    International Nuclear Information System (INIS)

    Pace, Brett

    1997-01-01

    Effective July 1, 1997, the Government Group portion of the Babcock and Wilcox company was incorporated separately to become BWX Technologies, Inc. (BWXT) a wholly-owned subsidiary of the Babcock and Wilcox Company. The names of the divisions and other business units of the former Babcock and Wilcox Government Group (Advanced Systems Operations, Naval Nuclear Fuel Division, and Nuclear Equipment Division) will remain unchanged, but they are now known as divisions or business units of BWXT. The management of all units and their reporting relationships will likewise remain unchanged. (author)

  11. Windows Calorimeter Control (WinCal) system configuration control board (SCCB) operating procedure

    International Nuclear Information System (INIS)

    Westsik, G.A.

    1997-01-01

    This document describes the operating procedure for the System Configuration Control Board (SCCB) performed in support of the Windows Calorimeter Control (WinCal) system. This board will consist of representatives from Babcock and Wilcox Hanford Company Babcock and Wilcox Protec, Inc.; and Lockheed Martin Services, Inc. In accordance with agreements for the joint use of the Babcock and Wilcox Hanford Company calorimeters located in the Hanford Site Plutonium Finishing Plant (PFP) Nondestructive Assay Laboratory, concurrence regarding changes to the WinCal system will be obtained from the International Atomic Energy Agency (IAEA). Further, changes to the WinCal software will be communicated to Los Alamos National Laboratory

  12. Radiation protection and reactor safety

    International Nuclear Information System (INIS)

    1990-03-01

    The Chernobyl reactor accident caused bewilderment, fear and anxiety among the population. How safe are reactors? Which precautions to protect lives and health have been taken? These questions are posed particularly in the areas of radiation protection, reactor safety, supply and waste management of nuclear power plants and other nuclear installations. For all these areas the present report contains an analysis of facts; it informs about political measures during the 11th legislative period of the German Bundestag, and shows prospects of future developments. (orig.) [de

  13. 4SC-101, a novel small molecule dihydroorotate dehydrogenase inhibitor, suppresses systemic lupus erythematosus in MRL-(Fas)lpr mice.

    Science.gov (United States)

    Kulkarni, Onkar P; Sayyed, Sufyan G; Kantner, Claudia; Ryu, Mi; Schnurr, Max; Sárdy, Miklós; Leban, Johann; Jankowsky, Ruediger; Ammendola, Aldo; Doblhofer, Robert; Anders, Hans-Joachim

    2010-06-01

    Immunosuppressive treatments of systemic lupus (SLE) remain associated with significant toxicities; hence, compounds with better toxicity profiles are needed. Dihydroorotate dehydrogenase (DHODH) inhibition with leflunomide has proven to be effective in autoimmune diseases including SLE, but leflunomide can cause a variety of side effects. We hypothesized that 4SC-101, a novel DHODH inhibitor with a more favorable toxicity profile, would be as effective as high-dose cyclophosphamide (CYC) in controlling experimental SLE of female MRL(Fas)lpr mice. Daily oral gavage of 30, 100, and 300 mg/kg 4SC-101 from 12 to 22 weeks of age was compared with either vehicle or CYC treatment (30 mg/kg/week, i.p.) in terms of efficacy and toxicity. Three hundred milligrams per kilogram 4SC-101 was as effective as CYC in depleting spleen autoreactive T cells, B cells, and plasma cells as well as the respective DNA and RNA serum autoantibodies. This was associated with a comparable amelioration of the renal, dermal, and pulmonary SLE manifestations of MRL(Fas)lpr mice. However, even the highest dose of 4SC-101 had no effect on bone marrow neutrophil counts, which were significantly reduced in CYC-treated mice. Together, the novel DHODH inhibitor 4SC-101 is as effective as high dose CYC in controlling SLE without causing myelosuppression. Hence, DHODH inhibition with 4SC-101 might be suitable to treat active SLE with fewer side effects than CYC.

  14. Fuel assembly stress and deflection analysis for loss-of-coolant accident and seismic excitation

    International Nuclear Information System (INIS)

    DeMars, R.V.; Steinke, R.R.

    1975-01-01

    Babcock and Wilcox has evaluated the capability of the fuel assemblies to withstand the effects of a loss-of-coolant accident (LOCA) blowdown, the operational basis earthquake (OBE) and design basis earthquake (DBE), and the simultaneous occurrence of the DBE and LOCA. This method of analysis is applicable to all of B and W's nuclear steam system contracts that specify the skirt-supported pressure vessel. Loads during the saturated and subcooled phases of blowdown following a loss-of-coolant accident were calculated. The maximum loads on the fuel assemblies were found to be below allowable limits, and the maximum deflections of the fuel assemblies were found to be less than those that could prevent the insertion of control rods or the flow of coolant through the core. (U.S.)

  15. Fuel Fabrication and Nuclear Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Karpius, Peter Joseph [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2017-02-02

    The uranium from the enrichment plant is still in the form of UF6. UF6 is not suitable for use in a reactor due to its highly corrosive chemistry as well as its phase diagram. UF6 is converted into UO2 fuel pellets, which are in turn placed in fuel rods and assemblies. Reactor designs are variable in moderators, coolants, fuel, performance etc.The dream of energy ‘too-cheap to meter’ is no more, and now the nuclear power industry is pushing ahead with advanced reactor designs.

  16. Fuel Fabrication and Nuclear Reactors

    International Nuclear Information System (INIS)

    Karpius, Peter Joseph

    2017-01-01

    The uranium from the enrichment plant is still in the form of UF 6 . UF 6 is not suitable for use in a reactor due to its highly corrosive chemistry as well as its phase diagram. UF 6 is converted into UO 2 fuel pellets, which are in turn placed in fuel rods and assemblies. Reactor designs are variable in moderators, coolants, fuel, performance etc.The dream of energy ''too-cheap to meter'' is no more, and now the nuclear power industry is pushing ahead with advanced reactor designs.

  17. RA reactor operation and maintenance

    International Nuclear Information System (INIS)

    Zecevic, V.

    1963-02-01

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

  18. Steam generator waterlancing at Darlington NGS (system development and field application)

    International Nuclear Information System (INIS)

    Seppala, D.; Malaugh, J.; Kiisel, E.; Kamler, F.

    1996-01-01

    From the initial steam generator (SG) inspections at Darlington Nuclear Generating Station (DNGS), the authors know that the sludge accumulations on the secondary side tubesheets have been minimal. DNGS is a fairly new station but the experience at the older Ontario Hydro plants have shown that significant accumulations will happen. A pro-active strategy has been adopted for maintaining SGs that will minimize corrosion product accumulation and the potential for component degradation. During the four year planned Unit maintenance outages, SGs will be inspected and waterlanced using a waterlance system designed and built by Babcock and Wilcox International. This automated state-of-the-art system also allows fully recorded inspections of the tubesheet/first half-lattice supports. Some of the key elements covered include results of the initial field application (May, 1995), system development and design, system qualification, cleaning performance, and lessons learned for future outages

  19. Galectin-9 ameliorates clinical severity of MRL/lpr lupus-prone mice by inducing plasma cell apoptosis independently of Tim-3.

    Directory of Open Access Journals (Sweden)

    Masahiro Moritoki

    Full Text Available Galectin-9 ameliorates various murine autoimmune disease models by regulating T cells and macrophages, although it is not known what role it may have in B cells. The present experiment shows that galectin-9 ameliorates a variety of clinical symptoms, such as proteinuria, arthritis, and hematocrit in MRL/lpr lupus-prone mice. As previously reported, galectin-9 reduces the frequency of Th1, Th17, and activated CD8(+ T cells. Although anti-dsDNA antibody was increased in MRL/lpr lupus-prone mice, galectin-9 suppressed anti-dsDNA antibody production, at least partly, by decreasing the number of plasma cells. Galectin-9 seemed to decrease the number of plasma cells by inducing plasma cell apoptosis, and not by suppressing BAFF production. Although about 20% of CD19(-/low CD138(+ plasma cells expressed Tim-3 in MRL/lpr lupus-prone mice, Tim-3 may not be directly involved in the galectin-9-induced apoptosis, because anti-Tim-3 blocking antibody did not block galectin-9-induced apoptosis. This is the first report of plasma cell apoptosis being induced by galectin-9. Collectively, it is likely that galectin-9 attenuates the clinical severity of MRL lupus-prone mice by regulating T cell function and inducing plasma cell apoptosis.

  20. Model tests of a once-through steam generator for land-blocker assessment and THEDA code verification. Final report

    International Nuclear Information System (INIS)

    Carter, H.R.; Childerson, M.T.; Moskal, T.E.

    1983-06-01

    The Babcock and Wilcox Company (B and W) operating Once-Through Steam Generators (OTSGs) have experienced leaking tubes in a region adjacent to the untubed inspection lane. The tube leaks have been attributed to an environmentally-assisted fatigue mechanism with moisture transported up the inspection lane being a major factor in the tube-failure process. B and W has developed a hardware modification (lane blockers) to mitigate the detrimental effects of inspection lane moisture. A 30-tube Laboratory Once-through Steam Generator (Designated OTSGC) was designed, fabricated, and tested. Tests were performed with and without five flat-plate lane blockers installed on tube-support plates (TSPs) 10, 11, 12, 13, and 14. The test results were utilized to determine the effectiveness of lane blockers for eliminating moisture transport to the upper tubesheet in the inspection lanes and to benchmark the predictive capabilities of a three-dimensional steam-generator computer code, THEDA

  1. Research reactors: design, safety requirements and applications

    International Nuclear Information System (INIS)

    Hassan, Abobaker Mohammed Rahmtalla

    2014-09-01

    There are two types of reactors: research reactors or power reactors. The difference between the research reactor and energy reactor is that the research reactor has working temperature and fuel less than the power reactor. The research reactors cooling uses light or heavy water and also research reactors need reflector of graphite or beryllium to reduce the loss of neutrons from the reactor core. Research reactors are used for research training as well as testing of materials and the production of radioisotopes for medical uses and for industrial application. The difference is also that the research reactor smaller in terms of capacity than that of power plant. Research reactors produce radioactive isotopes are not used for energy production, the power plant generates electrical energy. In the world there are more than 284 reactor research in 56 countries, operates as source of neutron for scientific research. Among the incidents related to nuclear reactors leak radiation partial reactor which took place in three mile island nuclear near pennsylvania in 1979, due to result of the loss of control of the fission reaction, which led to the explosion emitting hug amounts of radiation. However, there was control of radiation inside the building, and so no occurred then, another accident that lead to radiation leakage similar in nuclear power plant Chernobyl in Russia in 1986, has led to deaths of 4000 people and exposing hundreds of thousands to radiation, and can continue to be effect of harmful radiation to affect future generations. (author)

  2. Fast Reactors and Nuclear Nonproliferation

    International Nuclear Information System (INIS)

    Avrorina, E.N.; Chebeskovb, A.N.

    2013-01-01

    Conclusion remarks: 1. Fast reactor start-up with U-Pu fuel: – dependent on thermal reactors, – no needs in U enrichment, – needs in SNF reprocessing, – Pu is a little suitable for NED, – practically impossible gun-type NED, – difficulties for implosion-type NED: necessary tests, advanced technologies, etc. – Pu in blankets is similar to WPu by isotopic composition, – Use of blanket for production isotopes (e.g. 233 U), – Combined reprocessing of SNF: altogether blanket and core, – Blanket elimination: decrease in Pu production – No pure Pu separation. 2. Fast reactor start-up with U fuel: - Needs in both U enrichment and SNF reprocessing, - Independent of thermal reactors, - Good Pu bred in the core let alone blankets, - NED of simple gun-type design, - Increase of needs in SWU, - Increased demands in U supply. 3. Fast reactors for export: - Uranium shortage, - To replace thermal reactors in future, - No blankets (depends on the country, though), - Fuel supply and SNF take back, - International centers for rendering services of NFC. Time has come to remove from FRs and their NFC the label unfairly identifying them as the most dangerous installations of nuclear power from the standpoint of being a proliferation problem

  3. Training and Certification of Research Reactor Personnel

    International Nuclear Information System (INIS)

    Zarina Masood

    2011-01-01

    The safe operation of a research reactor requires that reactor personnel be fully trained and certified by the relevant authorities. Reactor operators at PUSPATI TRIGA Reactor underwent extensive training and are certified, ever since the reactor first started its operation in 1982. With the emphasis on enhancing reactor safety in recent years, reactor operator training and certification have also evolved. This paper discusses the changes that have to be implemented and the challenges encountered in developing a new training programme to be in line with the national standards. (author)

  4. Pressurized Water Reactors (PWR) and Boiling Water Reactors (BWR) are compared

    International Nuclear Information System (INIS)

    Greneche, D.

    2014-01-01

    This article compares the 2 types of light water reactors that are used to produce electricity: the Pressurized Water Reactor (PWR) and the Boiling Water Reactor (BWR). Historically the BWR concept was developed after the PWR concept. Today 80% of light water reactors operating in the world are of PWR-type. This comparison is comprehensive and detailed. First the main technical features are reviewed and compared: reactor architecture, core and fuel design, reactivity control, reactor vessel, cooling systems and reactor containment. Secondly, various aspects concerning reactor operations like reactor control, fuel management, maintenance, inspections, radiation protection, waste generation and reactor reliability are presented and compared for both reactors. As for the issue of safety, it is highlighted that the accidental situations are too different for the 2 reactors to be compared. The main features of reactor safety are explained for both reactors

  5. PARR-2: reactor description and experiments

    International Nuclear Information System (INIS)

    Wyne, M.F.; Meghji, J.H.

    1990-12-01

    PARR-2 is a miniature neutron source reactor (MNSR) research reactor has been designed at the rate of 27 kW. Reactor assembly comprises of peaking characteristics with a self limiting flux. In this report reactor description with its assembly and instrumentation control system has been explained. The reactor engineering and physics experiments which can be performed on this reactor are explained in this report. PARR-2 is fueled with HEU fuel pins which are about 90% enriched in U-235. Specific requirements for the safety of the reactor, its building and the personnel, normal instrumentation as required in an industrial environment is sufficient. (A.B.)

  6. Nuclear heating reactor, an advanced and passive reactor

    International Nuclear Information System (INIS)

    Wang Dazhong; Zheng Wenxiang

    1994-01-01

    The nuclear heating reactor (NHR) is designed with a number of the advanced and innovative features, including integrated arrangement, natural circulation, self-pressurized performance, dual vessel structure, hydraulic control rod drive and passive safety systems. Being an advanced and passive reactor, the NHR can serve as a clean, safe and economic energy source. This paper describes the development status, main design and safety features of the NHR. 3 refs., 2 tabs., 5 figs

  7. Solar Cycle Variability Induced by Tilt Angle Scatter in a Babcock-Leighton Solar Dynamo Model

    Science.gov (United States)

    Karak, Bidya Binay; Miesch, Mark

    2017-09-01

    We present results from a three-dimensional Babcock-Leighton (BL) dynamo model that is sustained by the emergence and dispersal of bipolar magnetic regions (BMRs). On average, each BMR has a systematic tilt given by Joy’s law. Randomness and nonlinearity in the BMR emergence of our model produce variable magnetic cycles. However, when we allow for a random scatter in the tilt angle to mimic the observed departures from Joy’s law, we find more variability in the magnetic cycles. We find that the observed standard deviation in Joy’s law of {σ }δ =15^\\circ produces a variability comparable to the observed solar cycle variability of ˜32%, as quantified by the sunspot number maxima between 1755 and 2008. We also find that tilt angle scatter can promote grand minima and grand maxima. The time spent in grand minima for {σ }δ =15^\\circ is somewhat less than that inferred for the Sun from cosmogenic isotopes (about 9% compared to 17%). However, when we double the tilt scatter to {σ }δ =30^\\circ , the simulation statistics are comparable to the Sun (˜18% of the time in grand minima and ˜10% in grand maxima). Though the BL mechanism is the only source of poloidal field, we find that our simulations always maintain magnetic cycles even at large fluctuations in the tilt angle. We also demonstrate that tilt quenching is a viable and efficient mechanism for dynamo saturation; a suppression of the tilt by only 1°-2° is sufficient to limit the dynamo growth. Thus, any potential observational signatures of tilt quenching in the Sun may be subtle.

  8. Nuclear Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hogerton, John

    1964-01-01

    This pamphlet describes how reactors work; discusses reactor design; describes research, teaching, and materials testing reactors; production reactors; reactors for electric power generation; reactors for supply heat; reactors for propulsion; reactors for space; reactor safety; and reactors of tomorrow. The appendix discusses characteristics of U.S. civilian power reactor concepts and lists some of the U.S. reactor power projects, with location, type, capacity, owner, and startup date.

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

    International Nuclear Information System (INIS)

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

    1990-01-01

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

  10. Research reactor standards and their impact on the TRIGA reactor community

    International Nuclear Information System (INIS)

    Richards, W.J.

    1980-01-01

    The American Nuclear Society has established a standards committee devoted to writing standards for research reactors. This committee was formed in 1971 and has since that time written over 15 standards that cover all aspects of research reactor operation. The committee has representation from virtually every group concerned with research reactors and their operation. This organization includes University reactors, National laboratory reactors, Nuclear Regulatory commission, Department of Energy and private nuclear companies and insurers. Since its beginning the committee has developed standards in the following areas: Standard for the development of technical specifications for research reactors; Quality control for plate-type uranium-aluminium fuel elements; Records and reports for research reactors; Selection and training of personnel for research reactors; Review of experiments for research reactors; Research reactor site evaluation; Quality assurance program requirements for research reactors; Decommissioning of research reactors; Radiological control at research reactor facilities; Design objectives for and monitoring of systems controlling research reactor effluents; Physical security for research reactor facilities; Criteria for the reactor safety systems of research reactors; Emergency planning for research reactors; Fire protection program requirements for research reactors; Standard for administrative controls for research reactors. Besides writing the above standards, the committee is very active in using communications with the nuclear regulatory commission on proposed rules or positions which will affect the research reactor community

  11. Nuclear reactor PBMR and cogeneration; Reactor nuclear PBMR y cogeneracion

    Energy Technology Data Exchange (ETDEWEB)

    Ramirez S, J. R.; Alonso V, G., E-mail: ramon.ramirez@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2013-10-15

    In recent years the nuclear reactor designs for the electricity generation have increased their costs, so that at the moment costs are managed of around the 5000 US D for installed kw, reason for which a big nuclear plant requires of investments of the order of billions of dollars, the designed reactors as modular of low power seek to lighten the initial investment of a big reactor dividing the power in parts and dividing in modules the components to lower the production costs, this way it can begin to build a module and finished this to build other, differing the long term investment, getting less risk therefore in the investment. On the other hand the reactors of low power can be very useful in regions where is difficult to have access to the electric net being able to take advantage of the thermal energy of the reactor to feed other processes like the water desalination or the vapor generation for the processes industry like the petrochemical, or even more the possible hydrogen production to be used as fuel. In this work the possibility to generate vapor of high quality for the petrochemical industry is described using a spheres bed reactor of high temperature. (Author)

  12. Propose Reactor Control and Monitoring System for RTP

    International Nuclear Information System (INIS)

    Mohd Sabri Minhat; Izhar Abu Hussin; Mohd Idris Taib; Mohd Khairulezwan Abdul Manan; Nurfarhana Ayuni Joha

    2011-01-01

    Reactor control and monitoring system is a one of the important features used in reactor. The control and monitoring must come together to provide safety, excellent performance and reliable in nuclear reactor technology application. Objectives of this technical paper are to design and propose reactor control system and reactor monitoring system in Research Reactor (RTP) for Reactor Upgrading Project. (author)

  13. Research reactor education and training

    International Nuclear Information System (INIS)

    Gless, B.; Chanteux, P.

    2003-01-01

    CORYS T.E.S.S. and TECHNICATOME present in this document some of the questions that can be rightfully raised concerning education and training of nuclear facilities' staffs. At first, some answers illustrate the tackled generic topics: importance of training, building of a training program, usable tools for training purposes. Afterwards, this paper deals more specifically with research reactors as an actual training tool. The pedagogical advantages they can bring are illustrated through an example consisting in the description of the AZUR facility training capabilities followed by the detailed experiences CORYS T.E.S.S. and TECHNICATOME have both gathered and keeps on gaining using research reactors for training means. The experience shows that this incomparable training material is not necessarily reserved to huge companies or organisations' numerous personnel. It offers enough flexibility to be adapted to the specific needs of a thinner audience. Thus research reactor staffs can also take advantages of this training method. (author)

  14. Research Reactors Types and Utilization

    International Nuclear Information System (INIS)

    Nasr, Nahla

    2008-01-01

    A nuclear reactor, in gross terms, is a device in which nuclear chain reactions are initiated, controlled, and sustained at a steady rate. The nuclei of fuel heavy atoms (mostly 235 U or 239 Pu), when struck by a slow neutron, may split into two or more smaller nuclei as fission products,releasing energy and neutrons in a process called nuclear fission. These newly-born fast neutrons then undergo several successive collisions with relatively low atomic mass material, the moderator, to become thermalized or slow. Normal water, heavy water, graphite and beryllium are typical moderators. These neutrons then trigger further fissions, and so on. When this nuclear chain reaction is controlled, the energy released can be used to heat water, produce steam and drive a turbine that generates electricity. The fission process, and hence the energy release, are controlled by the insertion (or extraction) of control rods through the reactor. These rods are strongly neutron absorbents, and thus only enough neutrons to sustain the chain reaction are left in the core. The energy released, mostly in the form of heat, should be continuously removed, to protect the core from damage. The most significant use of nuclear reactors is as an energy source for the generation of electrical power and for power in some military ships. This is usually accomplished by methods that involve using heat from the nuclear reaction to power steam turbines. Research reactors are used for radioisotope production and for beam experiments with free neutrons. Historically, the first use of nuclear reactors was the production of weapons grade plutonium for nuclear weapons. Currently all commercial nuclear reactors are based on nuclear fission. Fusion power is an experimental technology based on nuclear fusion instead of fission.

  15. Research reactor modernization and refurbishment

    International Nuclear Information System (INIS)

    2009-08-01

    Many recent, high profile research reactor unplanned shutdowns can be directly linked to different challenges which have evolved over time. The concept of ageing management is certainly nothing new to nuclear facilities, however, these events are highlighting the direct impact unplanned shutdowns at research reactors have on various stakeholders who depend on research reactor goods and services. Provided the demand for these goods and services remains strong, large capital projects are anticipated to continue in order to sustain future operation of many research reactors. It is within this context that the IAEA organized a Technical Workshop to launch a broader Agency activity on research reactor modernization and refurbishment (M and R). The workshop was hosted by the operating organization of the HOR Research Reactor in Delft, the Netherlands, in October 2006. Forty participants from twenty-three countries participated in the meeting: with representation from Africa, Asia Pacific, Eastern Europe, North America, South America and Western Europe. The specific objectives of this workshop were to present facility reports on completed, existing and planned M and R projects, including the project objectives, scope and main characteristics; and to specifically report on: - the project impact (planned or actual) on the primary and key supporting motivation for the M and R project; - the project impact (planned or actual) on the design basis, safety, and/or regulatory-related reports; - the project impact (planned or actual) on facility utilization; - significant lessons learned during or following the completion of M and R work. Contributions from this workshop were reviewed by experts during a consultancy meeting held in Vienna in December 2007. The experts selected final contributions for inclusion in this report. Requests were also distributed to some authors for additional detail as well as new authors for known projects not submitted during the initial 2006 workshop

  16. Update on reactors and reactor instruments in Asia

    Science.gov (United States)

    Rao, K. R.

    1991-10-01

    The 1980s have seen the commissioning of several medium flux (∼10 14 neutrons/cm 2s) research reactors in Asia. The reactors are based on indigenous design and development in India and China. At Dhruva reactor (India), a variety of neutron spectrometers have been established that have provided useful data related to the structure of high- Tc materials, phonon density of states, magnetic moment distributions and micellar aggregation during the last couple of years. Polarised neutron analysis, neutron interferometry and neutron spin echo methods are some of the new techniques under development. The spectrometers and associated automaton, detectors and neutron guides have all been indigenously developed. This paper summarises the developments and on-going activities in Bangladesh, China, India, Indonesia, Korea, Malaysia, the Philippines and Thailand.

  17. Update on reactors and reactor instruments in Asia

    International Nuclear Information System (INIS)

    Rao, K.R.

    1991-01-01

    The 1980s have seen the commissioning of several medium flux (∝10 14 neutrons/cm 2 s) research reactors in Asia. The reactors are based on indigenous design and development in India and China. At Dhruva reactor (India), a variety of neutron spectrometers have been established that have provided useful data related to the structure of high-Tc materials, phonon density of states, magnetic moment distributions and micellar aggregation during the last couple of years. Polarised neutron analysis, neutron interferometry and neutron spin echo methods are some of the new techniques under development. The spectrometers and associated automaton, detectors and neutron guides have all been indigenously developed. This paper summarises the developments and on-going activities in Bangladesh, China, India, Indonesia, Korea, Malaysia, the Philippines and Thailand. (orig.)

  18. Nuclear Reactors and Technology; (USA)

    Energy Technology Data Exchange (ETDEWEB)

    Cason, D.L.; Hicks, S.C. (eds.)

    1991-01-01

    Nuclear Reactors and Technology (NRT) announces on a monthly basis the current worldwide information available from the open literature on nuclear reactors and technology, including all aspects of power reactors, components and accessories, fuel elements, control systems, and materials. This publication contains the abstracts of DOE reports, journal articles, conference papers, patents, theses, and monographs added to the Energy Science and Technology Database (EDB) during the past month. Also included are US information obtained through acquisition programs or interagency agreements and international information obtained through the International Energy Agency's Energy Technology Data Exchange or government-to-government agreements. The digests in NRT and other citations to information on nuclear reactors back to 1948 are available for online searching and retrieval on EDB and Nuclear Science Abstracts (NSA) database. Current information, added daily to EDB, is available to DOE and its contractors through the DOE integrated Technical Information System. Customized profiles can be developed to provide current information to meet each user's needs.

  19. Nuclear reactors and fuel cycle

    International Nuclear Information System (INIS)

    2014-01-01

    The Nuclear Fuel Center (CCN) of IPEN produces nuclear fuel for the continuous operation of the IEA-R1 research reactor of IPEN. The serial production started in 1988, when the first nuclear fuel element was delivered for IEA-R1. In 2011, CCN proudly presents the 100 th nuclear fuel element produced. Besides routine production, development of new technologies is also a permanent concern at CCN. In 2005, U 3 O 8 were replaced by U 3 Si 2 -based fuels, and the research of U Mo is currently under investigation. Additionally, the Brazilian Multipurpose Research Reactor (RMB), whose project will rely on the CCN for supplying fuel and uranium targets. Evolving from an annual production from 10 to 70 nuclear fuel elements, plus a thousand uranium targets, is a huge and challenging task. To accomplish it, a new and modern Nuclear Fuel Factory is being concluded, and it will provide not only structure for scaling up, but also a safer and greener production. The Nuclear Engineering Center has shown, along several years, expertise in the field of nuclear, energy systems and correlated areas. Due to the experience obtained during decades in research and technological development at Brazilian Nuclear Program, personnel has been trained and started to actively participate in design of the main system that will compose the Brazilian Multipurpose Reactor (RMB) which will make Brazil self-sufficient in production of radiopharmaceuticals. The institution has participated in the monitoring and technical support concerning the safety, licensing and modernization of the research reactors IPEN/MB-01 and IEA-R1. Along the last two decades, numerous specialized services of engineering for the Brazilian nuclear power plants Angra 1 and Angra 2 have been carried out. The contribution in service, research, training, and teaching in addition to the development of many related technologies applied to nuclear engineering and correlated areas enable the institution to fulfill its mission that is

  20. Design and construction of multi research reactor

    International Nuclear Information System (INIS)

    1985-05-01

    This is the report about design and construction of multi research reactor, which introduces the purpose and necessity of the project, business contents, plan of progress of project and budget for the project. There are three appendixes about status of research reactor in other country, a characteristic of research reactor, three charts about evaluation, process and budget for the multi research reactor and three drawings for the project.

  1. X-ray Analysis of Defects and Anomalies in AGR-5/6/7 TRISO Particles

    Energy Technology Data Exchange (ETDEWEB)

    Helmreich, Grant W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Hunn, John D. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Skitt, Darren J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Dyer, John A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Schumacher, Austin T. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2017-06-01

    Coated particle fuel batches J52O-16-93164, 93165, 93166, 93168, 93169, 93170, and 93172 were produced by Babcock and Wilcox Technologies (BWXT) for possible selection as fuel for the Advanced Gas Reactor Fuel Development and Qualification (AGR) Program’s AGR-5/6/7 irradiation test in the Idaho National Laboratory (INL) Advanced Test Reactor (ATR), or may be used for other tests. Each batch was coated in a 150-mm-diameter production-scale fluidized-bed chemical vapor deposition (CVD) furnace. Tristructural isotropic (TRISO) coatings were deposited on 425-μm-nominal-diameter spherical kernels from BWXT lot J52R-16-69317 containing a mixture of 15.4%-enriched uranium carbide and uranium oxide (UCO), with the exception of Batch 93164, which used similar kernels from BWXT lot J52L-16-69316. The TRISO-coatings consisted of a ~50% dense carbon buffer layer with 100-μmnominal thickness, a dense inner pyrolytic carbon (IPyC) layer with 40-μm-nominal thickness, a silicon carbide (SiC) layer with 35-μm-nominal thickness, and a dense outer pyrolytic carbon (OPyC) layer with 40-μm-nominal thickness. Each coated particle batch was sieved to upgrade the particles by removing over-sized and under-sized material, and the upgraded batch was designated by appending the letter A to the end of the batch number (e.g., 93164A). Secondary upgrading by sieving was performed on the upgraded batches to remove specific anomalies identified during analysis for Defective IPyC, and the upgraded batches were designated by appending the letter B to the end of the batch number (e.g., 93165B). Following this secondary upgrading, coated particle composite J52R-16-98005 was produced by BWXT as fuel for the AGR Program’s AGR-5/6/7 irradiation test in the INL ATR. This composite was comprised of coated particle fuel batches J52O-16-93165B, 93168B, 93169B, and 93170B.

  2. Acceptance Test Data for BWXT Coated Particle Batches 93172B and 93173B—Defective IPyC and Pyrocarbon Anisotropy

    Energy Technology Data Exchange (ETDEWEB)

    Hunn, John D. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Helmreich, Grant W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Dyer, John A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Schumacher, Austin T. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Skitt, Darren J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2017-08-01

    Coated particle batches J52O-16-93172B and J52O-16-93173B were produced by Babcock and Wilcox Technologies (BWXT) as part of the production campaign for the Advanced Gas Reactor Fuel Development and Qualification (AGR) Program’s AGR-5/6/7 irradiation test in the Idaho National Laboratory (INL) Advanced Test Reactor (ATR), but were not used in the final fuel composite. However, these batches may be used as demonstration production-scale coated particle fuel for other experiments. Each batch was coated in a 150-mm-diameter production-scale fluidized-bed chemical vapor deposition (CVD) furnace. Tristructural isotropic (TRISO) coatings were deposited on 425-μm-nominal-diameter spherical kernels from BWXT lot J52R-16-69317 containing a mixture of 15.5%-enriched uranium carbide and uranium oxide (UCO). The TRISO coatings consisted of four consecutive CVD layers: a ~50% dense carbon buffer layer with 100-μm-nominal thickness, a dense inner pyrolytic carbon (IPyC) layer with 40-μm-nominal thickness, a silicon carbide (SiC) layer with 35-μm-nominal thickness, and a dense outer pyrolytic carbon (OPyC) layer with 40-μm-nominal thickness. The TRISO-coated particle batches were sieved to upgrade the particles by removing over-sized and under-sized material, and the upgraded batches were designated by appending the letter A to the end of the batch number (e.g., 93172A). Secondary upgrading by sieving was performed on the A-designated batches to remove particles with missing or very-thin buffer layers that were identified during previous analysis of the individual batches for defective IPyC, as reported in the acceptance test data report for the AGR-5/6/7 production batches [Hunn et al. 2017b]. The additionally-upgraded batches were designated by appending the letter B to the end of the batch number (e.g., 93172B).

  3. In-core instrumentation and reactor assessment

    International Nuclear Information System (INIS)

    Anon.

    1984-01-01

    Information on the conditions in the reactor core is essential for the safe and economic operation of nuclear reactors. This book reviews the important aspects of measurement and interpretation of reactor core parameters. Contributions of industry and research laboratories on the state of the art cover measurement methods, core performance evaluation, and operating experience

  4. Reactor protection and shut-down system

    International Nuclear Information System (INIS)

    Klar

    1980-01-01

    The reactor protection system being a part of the reactor safety system. The requirements on the reactor protection system are: high safety with regard to signal processing, high availability, self-reporting of faults etc. The functional sections of the reactor protection system are the analog section, the logic section and the generating of output signals. Description of the operation characteristics and of the extension of function. (orig.)

  5. Safety of research reactors (Design and Operation)

    International Nuclear Information System (INIS)

    Dirar, H. M.

    2012-06-01

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

  6. Reactor

    International Nuclear Information System (INIS)

    Fujibayashi, Toru.

    1976-01-01

    Object: To provide a boiling water reactor which can enhance a quake resisting strength and flatten power distribution. Structure: At least more than four fuel bundles, in which a plurality of fuel rods are arranged in lattice fashion which upper and lower portions are supported by tie-plates, are bundled and then covered by a square channel box. The control rod is movably arranged within a space formed by adjoining channel boxes. A spacer of trapezoidal section is disposed in the central portion on the side of the channel box over substantially full length in height direction, and a neutron instrumented tube is disposed in the central portion inside the channel box. Thus, where a horizontal load is exerted due to earthquake or the like, the spacers come into contact with each other to support the channel box and prevent it from abnormal vibrations. (Furukawa, Y.)

  7. Acceptance Test Data for the AGR-5/6/7 Irradiation Test Fuel Composite Defective IPyC Fraction and Pyrocarbon Anisotropy

    Energy Technology Data Exchange (ETDEWEB)

    Helmreich, Grant W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Hunn, John D. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Skitt, Darren J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Dyer, John A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Schumacher, Austin T. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2017-05-01

    Coated particle composite J52R-16-98005 was produced by Babcock and Wilcox Technologies (BWXT) as fuel for the Advanced Gas Reactor Fuel Development and Qualification (AGR) Program’s AGR-5/6/7 irradiation test in the Idaho National Laboratory (INL) Advanced Test Reactor (ATR). This composite was comprised of four coated particle fuel batches J52O-16-93165B (26%), 93168B (26%), 93169B (24%), and 93170B (24%), chosen based on the Quality Control (QC) data acquired for each individual candidate AGR-5/6/7 batch. Each batch was coated in a 150-mm-diameter production-scale fluidized-bed chemical vapor deposition (CVD) furnace. Tristructural isotropic (TRISO) coatings were deposited on 425-μm-nominal-diameter spherical kernels from BWXT Lot J52R-16-69317 containing a mixture of 15.5%-enriched uranium carbide and uranium oxide (UCO). The TRISO coatings consisted of four consecutive CVD layers: a ~50% dense carbon buffer layer with 100-μm-nominal thickness, a dense inner pyrolytic carbon (IPyC) layer with 40-μm-nominal thickness, a silicon carbide (SiC) layer with 35-μm-nominal thickness, and a dense outer pyrolytic carbon (OPyC) layer with 40-μm-nominal thickness. The TRISO-coated particle batches were sieved to upgrade the particles by removing over-sized and under-sized material, and the upgraded batches were designated by appending the letter A to the end of the batch number (e.g., 93165A). Secondary upgrading by sieving was performed on the A-designated batches to remove particles with missing or very-thin buffer layers that were identified during previous analysis of the individual batches for defective IPyC, as reported in the acceptance test data report for the AGR-5/6/7 production batches [Hunn et al. 2017]. The additionally-upgraded batches were designated by appending the letter B to the end of the batch number (e.g., 93165B).

  8. Reactor

    International Nuclear Information System (INIS)

    Evans, R.M.

    1976-01-01

    Disclosed is a neutronic reactor having a moderator, coolant tubes traversing the moderator from an inlet end to an outlet end, bodies of material fissionable by neutrons of thermal energy disposed within the coolant tubes, and means for circulating water through said coolant tubes characterized by the improved construction wherein the coolant tubes are constructed of aluminum having an outer diameter of 1.729 inches and a wall thickness of 0.059 inch, and the means for circulating a liquid coolant through the tubes includes a source of water at a pressure of approximately 350 pounds per square inch connected to the inlet end of the tubes, and said construction including a pressure reducing orifice disposed at the inlet ends of the tubes reducing the pressure of the water by approximately 150 pounds per square inch. 1 claim, 16 figures

  9. Reactor physics and reactor strategy investigations into the fissionable material economy of the thorium and uranium cycle in fast breeder reactors and high temperature reactors

    International Nuclear Information System (INIS)

    Schikorr, W.M.

    In this work the properties governing the fissionable material economy of the uranium and thorium cycles are investigated for the advanced reactor types currently under development - the fast breeder reactor (FBR) and the high temperature reactor (HTR) - from the point of view of the optimum utilization of the available nuclear fuel reserves and the continuance of supply of these reserves. For this purpose, the two reactor types are first of all considered individually and are subsequently discussed as a complementary overall system

  10. The United States Advanced Reactor Technologies Research and Development Program

    International Nuclear Information System (INIS)

    O’Connor, Thomas J.

    2014-01-01

    The following aspects are addressed: • Nuclear energy mission; • Reactor research development and deployment (RD&D) programs: - Light Water Reactor Sustainability Program; - Small Modular Reactor Licensing Technical Support; - Advanced Reactor Technologies (ART)

  11. Method and apparatus for stopping nuclear reactor

    International Nuclear Information System (INIS)

    Sakurai, Mikio.

    1974-01-01

    Object: To safely attain shut-down of a nuclear reactor even when control rods are not inserted into the core of the reactor and the shut-down of the reactor is incomplete. Structure: After operating the control rods in accordance with a scramble signal, the signal from an output detector is discriminated by an output discriminator, and a passage for a liquid poison is opened to allow the liquid poison to be poured from a liquid poison container through the passage into the core of the reactor when the output of the reactor exceeds the predetermined value. (Kamimura, M.)

  12. Advanced reactor concepts and safety

    International Nuclear Information System (INIS)

    Lipsett, J.J.

    1988-06-01

    The need for some consistency in the terms used to describe the evolution of methods for ensuring the safety of nuclear reactors has been identified by the IAEA. This is timely since there appears to be a danger that the precision of many valuable words is being diluted and that a new jargon may appear that will confuse rather than aid the communication of important but possibly diverse philosophies and concepts. Among the difficulties faced by the nuclear industry is promoting and gaining a widespread understanding of the risks actually posed by nuclear reactors. In view of the importance of communication to both the public and to the technical community generally, the starting point for the definition of terms must be with dictionary meanings and common technical usage. The nuclear engineering community should use such words in conformance with the whole technical world. This paper addresses many of the issues suggested in the invitation to meet and also poses some additional issues for consideration. Some examples are the role of the operator in either enhancing or degrading safety and how the meaning or interpretation of the word 'safety' can be expected to change during the next few decades. It is advantageous to use criteria against which technologies and ongoing operating performance can be judged provided that the criteria are generic and not specific to particular reactor concepts. Some thoughts are offered on the need to frame the criteria carefully so that innovative solutions and concepts are fostered, not stifled

  13. Reactor power control method and device

    International Nuclear Information System (INIS)

    Fushimi, Atsushi; Ishii, Yoshihiko; Miyamoto, Yoshiyuki; Ishii, Kazuhiko; Kiyoharu, Norihiko; Aizawa, Yuko.

    1997-01-01

    The present invention provides a method and a device suitable to rise the temperature and increase the pressure of the reactor to an aimed pressure in accordance with an aimed value for a reactor water temperature changing rate in the course of rising temperature and increasing pressure of the reactor upon start up of a BWR type power plant. Namely, neutron fluxes in the reactor and the temperature of reactor water are detected respectively. The maximum value among the detected values for the neutron fluxes is detected. The reactor water temperature changing rate is calculated based on the detected values of the reactor water temperature, from which the maximum value of the reactor water temperature changing rate is detected. An aimed value for the neutron flux is calculated in accordance with both detected maximum values and the aimed value of the reactor water temperature changing rate. The position of control rods is adjusted in accordance with the aimed value for the calculated neutron flux. Then, an aimed value for the neutron flux for realizing the aimed value for the reactor water temperature changing rate can be obtained accurately with no influence of the sensitivity of the detected values of the neutron fluxes and the time delay of the reactor water temperature changing rate. (I.S.)

  14. The profile of tuberculosis infection at the Babcock University ...

    African Journals Online (AJOL)

    2016-01-23

    Poor. Countries: Challenges and Opportunities. Clinical Microbiology Reviews 2011;24:314-350. 12. Ansari NA, Kombe AH, Kenyon. TA.Pathology and causes of death in a group of. 128 predominantly HIV-positive patients in.

  15. 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

  16. The profile of tuberculosis infection at the Babcock University ...

    African Journals Online (AJOL)

    Background: Tuberculosis is the leading cause of death from any single pathogen and it has consistently continued to be a major public health challenge globally. Data show that Nigeria ranks tenth among the 22 high tuberculosis burden countries. Aim: This study intends to describe the profile of tuberculosis infections in ...

  17. Geogenic organic contaminants in the low-rank coal-bearing Carrizo-Wilcox aquifer of East Texas, USA

    Science.gov (United States)

    Chakraborty, Jayeeta; Varonka, Matthew S.; Orem, William H.; Finkelman, Robert B.; Manton, William

    2017-01-01

    The organic composition of groundwater along the Carrizo-Wilcox aquifer in East Texas (USA), sampled from rural wells in May and September 2015, was examined as part of a larger study of the potential health and environmental effects of organic compounds derived from low-rank coals. The quality of water from the low-rank coal-bearing Carrizo-Wilcox aquifer is a potential environmental concern and no detailed studies of the organic compounds in this aquifer have been published. Organic compounds identified in the water samples included: aliphatics and their fatty acid derivatives, phenols, biphenyls, N-, O-, and S-containing heterocyclic compounds, polycyclic aromatic hydrocarbons (PAHs), aromatic amines, and phthalates. Many of the identified organic compounds (aliphatics, phenols, heterocyclic compounds, PAHs) are geogenic and originated from groundwater leaching of young and unmetamorphosed low-rank coals. Estimated concentrations of individual compounds ranged from about 3.9 to 0.01 μg/L. In many rural areas in East Texas, coal strata provide aquifers for drinking water wells. Organic compounds observed in groundwater are likely to be present in drinking water supplied from wells that penetrate the coal. Some of the organic compounds identified in the water samples are potentially toxic to humans, but at the estimated levels in these samples, the compounds are unlikely to cause acute health problems. The human health effects of low-level chronic exposure to coal-derived organic compounds in drinking water in East Texas are currently unknown, and continuing studies will evaluate possible toxicity.

  18. Reactor core and control rod assembly in FBR type reactor

    International Nuclear Information System (INIS)

    Fujimura, Koji; Kawashima, Katsuyuki; Itooka, Satoshi.

    1993-01-01

    Fuel assemblies and control rod assemblies are attached respectively to reactor core support plates each in a cantilever fashion. Intermediate spacer pads are disposed to the lateral side of a wrapper tube just above the fuel rod region. Intermediate space pads are disposed to the lateral side of a control rod guide tube just above a fuel rod region. The thickness of the intermediate spacer pad for the control rod assembly is made smaller than the thickness of the intermediate spacer pad for the fuel assembly. This can prevent contact between intermediate spacer pads of the control guide tube and the fuel assembly even if the temperature of coolants is elevated to thermally expand the intermediate spacer pad, by which the radial displacement amount of the reactor core region along the direction of the height of the control guide tube is reduced substantially to zero. Accordingly, contribution of the control rod assembly to the radial expansion reactivity can be reduced to zero or negative level, by which the effect of the negative radial expansion reactivity of the reactor is increased to improve the safety upon thermal transient stage, for example, loss of coolant flow rate accident. (I.N.)

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

    International Nuclear Information System (INIS)

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

    1992-01-01

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

  20. Reactor calculations and nuclear information

    International Nuclear Information System (INIS)

    Lang, D.W.

    1977-12-01

    The relationship of sets of nuclear parameters and the macroscopic reactor quantities that can be calculated from them is examined. The framework of the study is similar to that of Usachev and Bobkov. The analysis is generalised and some properties required by common sense are demonstrated. The form of calculation permits revision of the parameter set. It is argued that any discrepancy between a calculation and measurement of a macroscopic quantity is more useful when applied directly to prediction of other macroscopic quantities than to revision of the parameter set. The mathematical technique outlined is seen to describe common engineering practice. (Author)

  1. Computerized reactor monitor and control for research reactors

    International Nuclear Information System (INIS)

    Buerger, L.; Vegh, E.

    1981-09-01

    The computerized process control system developed in the Central Research Institute for Physics, Budapest, Hungary, is described together with its special applications at research reactors. The nuclear power of the Hungarian research reactor is controlled by this computerized system, too, while in Lybia many interesting reactor-hpysical calculations are built into the computerized monitor system. (author)

  2. Nuclear reactors and disarmament

    International Nuclear Information System (INIS)

    Almagro, J.C.; Estrada Oyuela, M.E.; Garcia Moritan, R.

    1987-01-01

    From a brief analysis of the perspectives of nuclear weapons arsenals reduction, a rational use of the energetic potential of the ogives and the authentic destruction of its warlike power is proposed. The fissionable material conversion contained in the nuclear fuel ogives for peaceful uses should be part of the disarmament agreements. This paper pretends to give an approximate idea on the resources re assignation implicancies. (Author)

  3. U.S. and foreign breeder reactors

    International Nuclear Information System (INIS)

    Hill, E.H.

    1977-01-01

    The running battle between Congress and the Administration over the Clinch River Breeder Reactor Plant (CRBRP) Project has provoked an increased interest in domestic and foreign breeder reactor programs. Perhaps an understanding of the history of breeders here and abroad will serve to place the CRBRP in perspective and allow some analysis of how the U.S. appears on the global canvas. Breeder reactor technology has, for the most part, settled down to concentration on the liquid metal fast breeder reactor (LMFBR). This is the result of 32 years of experience with reactors employing a fast neutron flux and even longer experience with liquid metal coolants. However, a number of U.S. utilities are sponsoring a gas cooled fast reactor program as an alternative technology to the LMFBR. This development program is supported by the U.S. Department of Energy

  4. Education and Training on ISIS Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Foulon, F.; Badeau, G.; Lescop, B.; Wohleber, X. [French Atomic Energy and Alternative Energies Commission, Paris (France)

    2013-07-01

    In the frame of academic and vocational programs the National Institute for Nuclear Science and Technology uses the ISIS research reactor as a major tool to ensure a practical and comprehensive understanding of the nuclear reactor physics, principles and operation. A large set of training courses have been developed on ISIS, optimising both the content of the courses and the pedagogical approach. Programs with duration ranging from 3 hours (introduction to reactor operation) to 24 hours (full program for the future operators of research reactors) are carried out on ISIS reactor. The reactor is operated about 350 hours/year for education and training, about 40 % of the courses being carried out in English. Thus, every year about 400 trainees attend training courses on ISIS reactor. We present here the ISIS research reactor and the practical courses that have been developed on ISIS reactor. Emphasis is given to the pedagogical method which is used to focus on the operational and safety aspects, both in normal and incidental operation. We will present the curricula of the academic and vocational courses in which the practical courses are integrated, the courses being targeted to a wide public, including operators of research reactors, engineers involved in the design and operation of nuclear reactors as well as staff of the regulatory body. We address the very positive impact of the courses on the development of the competences and skills of participants. Finally, we describe the Internet Reactor Laboratories (IRL) that are under development and will consist in broadcasting the training courses via internet to remote facilities or institutions.

  5. Education and Training on ISIS Research Reactor

    International Nuclear Information System (INIS)

    Foulon, F.; Badeau, G.; Lescop, B.; Wohleber, X.

    2013-01-01

    In the frame of academic and vocational programs the National Institute for Nuclear Science and Technology uses the ISIS research reactor as a major tool to ensure a practical and comprehensive understanding of the nuclear reactor physics, principles and operation. A large set of training courses have been developed on ISIS, optimising both the content of the courses and the pedagogical approach. Programs with duration ranging from 3 hours (introduction to reactor operation) to 24 hours (full program for the future operators of research reactors) are carried out on ISIS reactor. The reactor is operated about 350 hours/year for education and training, about 40 % of the courses being carried out in English. Thus, every year about 400 trainees attend training courses on ISIS reactor. We present here the ISIS research reactor and the practical courses that have been developed on ISIS reactor. Emphasis is given to the pedagogical method which is used to focus on the operational and safety aspects, both in normal and incidental operation. We will present the curricula of the academic and vocational courses in which the practical courses are integrated, the courses being targeted to a wide public, including operators of research reactors, engineers involved in the design and operation of nuclear reactors as well as staff of the regulatory body. We address the very positive impact of the courses on the development of the competences and skills of participants. Finally, we describe the Internet Reactor Laboratories (IRL) that are under development and will consist in broadcasting the training courses via internet to remote facilities or institutions

  6. Nuclear reactors design study and parameters calculation

    International Nuclear Information System (INIS)

    Morcos, H.N.

    2002-01-01

    the nuclear design a reactor core needs to determine a set of system parameters which will lead to safe, reliable and economical reactor operation at the rated power level over the desired core lifetime. the principal tools used in this task consist of a number of models of neutron behavior in the reactor that are implemented by a multiplicity of computer programs or codes used to simulate the nuclear behavior of the reactor core. the study of the interaction of the core power distributions with the time-dependent production or depletion of nuclei in the core is known as depletion or burn up analysis the main objective of the present thesis is to study the fuel depletion analysis under different reactor operating regimes and their influence on the build up of actinides and fission products (F P). therefore, one can estimate the optimum reactor-operating regime at which the accumulation of certain actinide isotope can reach maximum

  7. Advanced nuclear reactor types and technologies

    Energy Technology Data Exchange (ETDEWEB)

    Ignatiev, V. [ed.; Feinberg, O.; Morozov, A. [Russian Research Centre `Kurchatov Institute`, Moscow (Russian Federation); Devell, L. [Studsvik Eco and Safety AB, Nykoeping (Sweden)

    1995-07-01

    The document is a comprehensive world-wide catalogue of concepts and designs of advanced fission reactor types and fuel cycle technologies. Two parts have been prepared: Part 1 Reactors for Power Production and Part 2 Heating and Other Reactor Applications. Part 3, which will cover advanced waste management technology, reprocessing and disposal for different nuclear fission options is planned for compilation during 1995. The catalogue was prepared according to a special format which briefly presents the project title, technical approach, development status, application of the technology, reactor type, power output, and organization which developed these designs. Part 1 and 2 cover water cooled reactors, liquid metal fast reactors, gas-cooled reactors and molten salt reactors. Subcritical accelerator-driven systems are also considered. Various reactor applications as power production, heat generation, ship propulsion, space power sources and transmutation of such waste are included. Each project is described within a few pages with the main features of an actual design using a table with main technical data and figure as well as references for additional information. Each chapter starts with an introduction which briefly describes main trends and approaches in this field. Explanations of terms and abbreviations are provided in a glossary.

  8. Advanced nuclear reactor types and technologies

    International Nuclear Information System (INIS)

    Ignatiev, V.; Devell, L.

    1995-01-01

    The document is a comprehensive world-wide catalogue of concepts and designs of advanced fission reactor types and fuel cycle technologies. Two parts have been prepared: Part 1 Reactors for Power Production and Part 2 Heating and Other Reactor Applications. Part 3, which will cover advanced waste management technology, reprocessing and disposal for different nuclear fission options is planned for compilation during 1995. The catalogue was prepared according to a special format which briefly presents the project title, technical approach, development status, application of the technology, reactor type, power output, and organization which developed these designs. Part 1 and 2 cover water cooled reactors, liquid metal fast reactors, gas-cooled reactors and molten salt reactors. Subcritical accelerator-driven systems are also considered. Various reactor applications as power production, heat generation, ship propulsion, space power sources and transmutation of such waste are included. Each project is described within a few pages with the main features of an actual design using a table with main technical data and figure as well as references for additional information. Each chapter starts with an introduction which briefly describes main trends and approaches in this field. Explanations of terms and abbreviations are provided in a glossary

  9. Space reactor fuels performance and development issues

    International Nuclear Information System (INIS)

    Wewerka, E.M.

    1984-01-01

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

  10. Reactor safety research program. A description of current and planned reactor safety research sponsored by the Nuclear Regulatory Commission's Division of Reactor Safety Research

    International Nuclear Information System (INIS)

    1975-06-01

    The reactor safety research program, sponsored by the Nuclear Regulatory Commission's Division of Reactor Safety Research, is described in terms of its program objectives, current status, and future plans. Elements of safety research work applicable to water reactors, fast reactors, and gas cooled reactors are presented together with brief descriptions of current and planned test facilities. (U.S.)

  11. Physics and safety of advanced research reactors

    International Nuclear Information System (INIS)

    Boening, K.; Hardt, P. von der

    1987-01-01

    Advanced research reactor concepts are presently being developed in order to meet the neutron-based research needs of the nineties. Among these research reactors, which are characterized by an average power density of 1-10 MW per liter, highest priority is now generally given to the 'beam tube reactors'. These provide very high values of the thermal neutron flux (10 14 -10 16 cm -2 s -1 ) in a large volume outside of the reactor core, which can be used for sample irradiations and, in particular, for neutron scattering experiments. The paper first discusses the 'inverse flux trap concept' and the main physical aspects of the design and optimization of beam tube reactors. After that two examples of advanced research reactor projects are described which may be considered as two opposite extremes with respect to the physical optimization principle just mentioned. The present situation concerning cross section libraries and neutronic computer codes is more or less satisfactory. The safety analyses of advanced research reactors can largely be updated from those of current new designs, partially taking advantage of the immense volume of work done for power reactors. The paper indicates a few areas where generic problems for advanced research reactor safety are to be solved. (orig.)

  12. Design and construction of reactor containment systems of the prototype fast breeder reactor MONJU

    International Nuclear Information System (INIS)

    Ikeda, Makinori; Kawata, Koji; Sato, Masaki; Ito, Masashi; Hayashi, Kazutoshi; Kunishima, Shigeru.

    1991-01-01

    The MONJU reactor containment systems consist of a reactor containment vessel, reactor cavity walls and cell liners. The reactor containment vessel is strengthened by ring stiffeners for earthquake stresses. To verify its earthquake-resistant strength, vibration and buckling tests were carried out by using 1/19 scale models. The reactor cavity walls, which form biological shield and support the reactor vessel, are constructed of steel plate frames filled with concrete. The cell liner consists of liner plates and thermal insulation to moderate the effects of sodium spills, and forms a gastight cell to maintain a nitrogen atmosphere. (author)

  13. Structures and Materials of Reactor Internals for PWR in Korea

    Energy Technology Data Exchange (ETDEWEB)

    Kim, D. W.; Kim, W. S.; Kwon, S. C.; Kwon, J. H.; Kim, Y. S.; Kim, H. P.; Yoo, C. S.; Lee, S. R.; Jung, M. K.; Hwang, S. S

    2007-10-15

    Nuclear reactor types in Korea are PWR type reactor (Westinghouse, Combustion Engineering, Farmatome type) and CANDU type reactor. Structures and Materials for reactor internal of PWR type were investigated. Reactor internal was composed of lower core support structure, upper core support assembly, incore instrumentation support structure. Lower core support structure of these structures is the most important. The major material for the reactor internal is type 304 and 316 stainless steel and radial support clevis bolts are made of Inconel. The main damage mechanism for reactor internal was IASCC and the effect of IASCC on reactor internal was investigated. The accident for reactor internal was also investigate.

  14. The properties of the lunar regolith at Chang'e-3 landing site: A study based on LPR data

    Science.gov (United States)

    Feng, J.; Su, Y.; Xing, S.; Ding, C.; Li, C.

    2015-12-01

    In situ sampling from surface is difficult in the exploration of planets and sometimes radar sensing is a better choice. The properties of the surface material such as permittivity, density and depth can be obtained by a surface penetrating radar. The Chang'e-3 (CE-3) landed in the northern Mare Imbrium and a Lunar Penetrating Radar (LPR) is carried on the Yutu rover to detect the shallow structure of the lunar crust and the properties of the lunar regolith, which will give us a close look at the lunar subsurface. We process the radar data in a way which consist two steps: the regular preprocessing step and migration step. The preprocessing part includes zero time correction, de-wow, gain compensation, DC removal, geometric positioning. Then we combine all radar data obtained at the time the rover was moving, and use FIR filter to reduce the noise in the radar image with a pass band frequency range 200MHz-600MHz. A normal radar image is obtained after the preprocessing step. Using a nonlinear least squares fitting method, we fit the most hyperbolas in the radar image which are caused by the buried objects or rocks in the regolith and estimate the EM wave propagation velocity and the permittivity of the regolith. For there is a fixed mathematical relationship between dielectric constant and density, the density profile of the lunar regolith is also calculated. It seems that the permittivity and density at the landing site is larger than we thought before. Finally with a model of variable velocities, we apply the Kirchhoff migration method widely used in the seismology to transform the the unfocused space-time LPR image to a focused one showing the object's (most are stones) true location and size. From the migrated image, we find that the regolith depth in the landing site is smaller than previous study and the stone content rises rapidly with depth. Our study suggests that the landing site is a young region and the reworked history of the surface is short, which is

  15. Analysis of accidents at the LPR (Radiochemical Processes Laboratory)

    International Nuclear Information System (INIS)

    Kaufmann, F.; Boutet, L.I.

    1987-01-01

    Accidents are defined as not planned events that may result in the emission of significative quantities of radioactive materials to the environment. The pilot plant has been specifically designed to prevent this type of accidents but there still exists the possibility that one or more accidents can be produced during the plant life. In a first phase, the emission of radionuclides to the environment were evaluated for 13 credible accidents. In a second phase, by means of the calculation program SEDA, specially adapted to this purpose, the critical doses of critical group were calculated for each accident. Due to the small capacity of the pilot plant and the long cooling period of treated fuel, it is concluded that the radiological consequences for the external environment are of very small magnitude. In this way, without need of developing complex fault- or event-trees, it is shown that any of the accidents falls into the non acceptable zone of Farmer diagram. (Author)

  16. Monitoring and control of anaerobic reactors

    DEFF Research Database (Denmark)

    Pind, Peter Frode; Angelidaki, Irini; Ahring, Birgitte Kiær

    2003-01-01

    The current status in monitoring and control of anaerobic reactors is reviewed. The influence of reactor design and waste composition on the possible monitoring and control schemes is examined. After defining the overall control structure, and possible control objectives, the possible process...

  17. Research reactors and alternative devices for research

    International Nuclear Information System (INIS)

    1985-01-01

    This report includes papers on research reactors and alternatives to the research reactors - radioisotopic neutron sources, cyclotrons, D-T neutron generators and small accelerators, used for radioisotope production, neutron activation analysis, material science, applied and basic research using neutron beams. A separate abstract was prepared for each of the 7 papers

  18. Exporting apocalypse: CANDU reactors and nuclear proliferation

    International Nuclear Information System (INIS)

    McKay, Paul.

    The author believes that the peaceful use of nuclear technology leads inevitably to the production of nuclear weapons, and that CANDU reactors are being bought by countries that are likely to build bombs. He states that exports of reactors and nuclear materials cannot be defended and must be stopped

  19. Fast reactor fuel design and development

    International Nuclear Information System (INIS)

    Bishop, J.F.W.; Chamberlain, A.; Holmes, J.A.G.

    1977-01-01

    Fuel design parameters for oxide and carbide fast reactor fuels are reviewed in the context of minimising the total uranium demands for a combined thermal and fast reactor system. The major physical phenomena conditioning fast reactor fuel design, with a target of high burn-up, good breeding and reliable operation, are characterised. These include neutron induced void swelling, irradiation creep, pin failure modes, sub-assembly structural behaviour, behaviour of defect fuel, behaviour of alternative fuel forms. The salient considerations in the commercial scale fabrication and reprocessing of the fuels are reviewed, leading to the delineation of possible routes for the manufacture and reprocessing of Commercial Reactor fuel. From the desiderata and restraints arising from Surveys, Performance and Manufacture, the problems posed to the Designer are considered, and a narrow range of design alternatives is proposed. The paper concludes with a consideration of the development areas and the conceptual problems for fast reactors associated with those areas

  20. TRIGA reactor owners' seminar. Papers and abstracts

    International Nuclear Information System (INIS)

    1970-01-01

    The TRIGA Reactor Owners' Conference was planned with the aim of bringing together a group of persons interested in the ownership and operation of TRIGA reactors in the hope that an interchange of viewpoints, information, and experience would prove of mutual benefit

  1. Scanning tunneling microscope assembly, reactor, and system

    Science.gov (United States)

    Tao, Feng; Salmeron, Miquel; Somorjai, Gabor A

    2014-11-18

    An embodiment of a scanning tunneling microscope (STM) reactor includes a pressure vessel, an STM assembly, and three spring coupling objects. The pressure vessel includes a sealable port, an interior, and an exterior. An embodiment of an STM system includes a vacuum chamber, an STM reactor, and three springs. The three springs couple the STM reactor to the vacuum chamber and are operable to suspend the scanning tunneling microscope reactor within the interior of the vacuum chamber during operation of the STM reactor. An embodiment of an STM assembly includes a coarse displacement arrangement, a piezoelectric fine displacement scanning tube coupled to the coarse displacement arrangement, and a receiver. The piezoelectric fine displacement scanning tube is coupled to the coarse displacement arrangement. The receiver is coupled to the piezoelectric scanning tube and is operable to receive a tip holder, and the tip holder is operable to receive a tip.

  2. Operational and reliability experience with reactor instrumentation

    International Nuclear Information System (INIS)

    Dixon, F.; Gow, R.S.

    1978-01-01

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

  3. ISIS Training Reactor: A Reactor Dedicated to Education and Training for Students and Professionals

    International Nuclear Information System (INIS)

    Foulon, F.

    2014-01-01

    Conclusion: • INSTN strategy: complete theoretical courses by practical courses on the ISIS research reactor. • Training courses integrated both in Academic degree programs and continuing education. • 27 hours of training courses have been developed focusing on the practical and safety aspects of reactor operation. • The Education and Training activity became the main activity of ISIS reactor: 400 trainees/year; 360 hours/year; 40% in English. • Remote access to the Training courses: Internet Reactor Laboratory under development to be started from 2014 to broadcast training courses from ISIS reactor to guest institutions

  4. New reactors concepts and scenarios

    International Nuclear Information System (INIS)

    Gandini, A.

    2001-01-01

    In recent years an increasing interest is observed with respect to subcritical, accelerator driven systems (ADS), for their possible role in perspective future nuclear energy scenarios, as actinide (Pu and MA) incinerators, and/or claimed energy plants with potential enhanced safety characteristics. Important research programs are devoted to the various related fields of research. Extensive studies on the ADS behavior under incidental conditions are in particular made, for verifying their claimed advantage, under the safety point of view, with respect to the corresponding critical reactors. Corresponding medium and long range scenarios are being studied to cope with a number of concerns associated with the safety (power excursions. residual heat risk), as well as with the fuel flow (criticality accidents, fuel diversion, radiological risk, proliferation). In the present work we shall try to review current lines of research in this field, and comment on possible scenarios so far envisaged. (author)

  5. Possibilities of TWR and long life reactor

    International Nuclear Information System (INIS)

    Sekimoto, Hiroshi; Shimazu, Yoichiro; Handa, Norihiko

    2010-01-01

    Bill Gates identified the need to switch to zero-emission energy and clarified investing in Terra Power developing the TWR (Traveling Wave Reactor) in February 2010. He also visited Toshiba developing small reactor 4S (Super Safe Small and Simple). In Japan design studies of the TWR have been conducted on the CANDLE reactor without refueling and the 4S long life reactor with maintenance free. In this feature article, the state of R and D on the TWR in Japan and IAEA's activities on small reactors without online refueling were reviewed in addition to articles on impacts of Bill Gates' investment in the TWR and state of the TWR development from an interview with John Gilleland of Terra Power. (T. Tanaka)

  6. TREAT Reactor Control and Protection System

    International Nuclear Information System (INIS)

    Lipinski, W.C.; Brookshier, W.K.; Burrows, D.R.; Lenkszus, F.R.; McDowell, W.P.

    1985-01-01

    The main control algorithm of the Transient Reactor Test Facility (TREAT) Automatic Reactor Control System (ARCS) resides in Read Only Memory (ROM) and only experiment specific parameters are input via keyboard entry. Prior to executing an experiment, the software and hardware of the control computer is tested by a closed loop real-time simulation. Two computers with parallel processing are used for the reactor simulation and another computer is used for simulation of the control rod system. A monitor computer, used as a redundant diverse reactor protection channel, uses more conservative setpoints and reduces challenges to the Reactor Trip System (RTS). The RTS consists of triplicated hardwired channels with one out of three logic. The RTS is automatically tested by a digital Dedicated Microprocessor Tester (DMT) prior to the execution of an experiment. 6 refs., 5 figs., 1 tab

  7. RA research reactor - potentials and prospective

    International Nuclear Information System (INIS)

    Sotic, O.

    1984-01-01

    Since December 1959, the RA reactor was operated successfully, except for a few shorter periods needed for maintenance and a four longer shutdown periods caused by decrease in the heavy water quality. Accordingly, reconstruction of some reactor systems was started at the beginning of this decad, as well as increase of its experimental potential which would enable its efficient reliable operation in the future period. Reconstruction is concerned with emergency core cooling system, special ventilation system, and modernization of the reactor instrumentation. Improvement of the experimental potential is related to modifications of the neutron scattering instruments. Development of methods for isotope production is described as well. Design of the reactor experimental loop with external cooling system will be of significant importance in improvement of reactor potential in the future

  8. K-East and K-West Reactors

    Data.gov (United States)

    Federal Laboratory Consortium — Hanford's "sister reactors", the K-East and the K-West Reactors, were built side-by-side in the early 1950's. The two reactors went operational within four months of...

  9. Power Nuclear Reactors: technology and innovation for development in future

    International Nuclear Information System (INIS)

    Suarez Antola, R.

    2009-01-01

    The conference is about some historicals task of the fission technology as well as many types of Nuclear Reactors. Enrichment of fuel, wastes, research reactors and power reactors, a brief advertisment about Uruguay electric siystem and power generation, energetic worldwide, proliferation, safety reactors, incidents, accidents, Three-Mile Island accident, Chernobil accident, damages, risks, classification and description of Power reactors steam generation, nuclear reactor cooling systems, future view

  10. The program of reactors and nuclear power plants

    International Nuclear Information System (INIS)

    Calabrese, Carlos R.

    2001-01-01

    Into de framework of the program of research reactors and nuclear power plants, the operating Argentine reactors are described. The uses of the research reactors in Argentina are summarized. The reactors installed by Argentina in other countries (Peru, Algeria, Egypt) are briefly described. The CAREM project for the design and construction of an innovator small power reactor (27 MWe) is also described in some detail. The next biennial research and development program for reactor is briefly outlined

  11. Thermal and flow design of helium-cooled reactors

    International Nuclear Information System (INIS)

    Melese, G.; Katz, R.

    1984-01-01

    This book continues the American Nuclear Society's series of monographs on nuclear science and technology. Chapters of the book include information on the first-generation gas-cooled reactors; HTGR reactor developments; reactor core heat transfer; mechanical problems related to the primary coolant circuit; HTGR design bases; core thermal design; gas turbines; process heat HTGR reactors; GCFR reactor thermal hydraulics; and gas cooling of fusion reactors

  12. Nuclear power reactors

    International Nuclear Information System (INIS)

    1982-11-01

    After an introduction and general explanation of nuclear power the following reactor types are described: magnox thermal reactor; advanced gas-cooled reactor (AGR); pressurised water reactor (PWR); fast reactors (sodium cooled); boiling water reactor (BWR); CANDU thermal reactor; steam generating heavy water reactor (SGHWR); high temperature reactor (HTR); Leningrad (RMBK) type water-cooled graphite moderated reactor. (U.K.)

  13. Ceramic oxygen transport membrane array reactor and reforming method

    Energy Technology Data Exchange (ETDEWEB)

    Kelly, Sean M.; Christie, Gervase Maxwell; Robinson, Charles; Wilson, Jamie R.; Gonzalez, Javier E.; Doraswami, Uttam R.

    2016-11-08

    The invention relates to a commercially viable modular ceramic oxygen transport membrane reforming reactor configured using repeating assemblies of oxygen transport membrane tubes and catalytic reforming reactors.

  14. REACTOR: an expert system for diagnosis and treatment of nuclear reactor accidents

    International Nuclear Information System (INIS)

    Nelson, W.R.

    1982-01-01

    REACTOR is an expert system under development at EG and G Idaho, Inc., that will assist operators in the diagnosis and treatment of nuclear reactor accidents. This paper covers the background of the nuclear industry and why expert system technology may prove valuable in the reactor control room. Some of the basic features of the REACTOR system are discussed, and future plans for validation and evaluation of REACTOR are presented. The concept of using both event-oriented and function-oriented strategies for accident diagnosis is discussed. The response tree concept for representing expert knowledge is also introduced

  15. MELCOR analyses of severe accident scenarios in Oconee, a B ampersand W PWR plant

    International Nuclear Information System (INIS)

    Madni, I.K.; Nimnual, S.; Foulds, R.

    1993-01-01

    This paper presents the results and insights gained from MELCOR analyses of two severe accident scenarios, a Loss of Coolant Accident (LOCA) and a Station Blackout (TMLB) in Oconee, a Babcock ampersand Wilcox (B ampersand W) designed PWR with a large dry containment, and comparisons with Source Term Code Package (STCP) calculations of the same sequences. Results include predicted timing of key events, thermal-hydraulic response in the reactor coolant system and containment, and environmental releases of fission products. The paper also explores the impact of varying concrete type, vessel failure temperature, and break location on the accident progression, containment pressurization, and environmental releases of radionuclides

  16. Advanced Demonstration and Test Reactor Options Study

    Energy Technology Data Exchange (ETDEWEB)

    Petti, David Andrew [Idaho National Lab. (INL), Idaho Falls, ID (United States); Hill, R. [Argonne National Lab. (ANL), Argonne, IL (United States); Gehin, J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (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); Heidet, F. [Argonne National Lab. (ANL), Argonne, IL (United States); Kinsey, J. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Grandy, Christopher [Argonne National Lab. (ANL), Argonne, IL (United States); Qualls, A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Brown, Nicholas [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Powers, J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Hoffman, E. [Argonne National Lab. (ANL), Argonne, IL (United States); Croson, D. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2017-01-01

    Global efforts to address climate change will require large-scale decarbonization of energy production in the United States and elsewhere. Nuclear power already provides 20% of electricity production in the United States (U.S.) and is increasing in countries undergoing rapid growth around the world. Because reliable, grid-stabilizing, low emission electricity generation, energy security, and energy resource diversity will be increasingly valued, nuclear power’s share of electricity production has a potential to grow. In addition, there are non electricity applications (e.g., process heat, desalination, hydrogen production) that could be better served by advanced nuclear systems. Thus, the timely development, demonstration, and commercialization of advanced nuclear reactors could diversify the nuclear technologies available and offer attractive technology options to expand the impact of nuclear energy for electricity generation and non-electricity missions. The purpose of this planning study is to provide transparent and defensible technology options for a test and/or demonstration reactor(s) to be built to support public policy, innovation and long term commercialization within the context of the Department of Energy’s (DOE’s) broader commitment to pursuing an “all of the above” clean energy strategy and associated time lines. This planning study includes identification of the key features and timing needed for advanced test or demonstration reactors to support research, development, and technology demonstration leading to the commercialization of power plants built upon these advanced reactor platforms. This planning study is consistent with the Congressional language contained within the fiscal year 2015 appropriation that directed the DOE to conduct a planning study to evaluate “advanced reactor technology options, capabilities, and requirements within the context of national needs and public policy to support innovation in nuclear energy

  17. Different types of power reactors and provenness

    International Nuclear Information System (INIS)

    Goodman, E.I.

    1977-01-01

    The lecture guides the potential buyer in the selection of a reactor type. Recommended criteria regarding provenness, licensability, and contractual arrangements are defined and discussed. Tabular data summarizing operating experience and commercial availability of units are presented and discussed. The status of small and medium power reactors which are of interest to many developing countries is presented. It is stressed that each prospective buyer will have to establish his own criteria based on specific conditions which will be applied to reactor selection. In all cases it will be found that selection, either pre-selection of bidders or final selection of supplier, will be a fairly complex evaluation. (orig.) [de

  18. GPU is in great jeopardy: PUC report details GPU's deteriorating financial position

    International Nuclear Information System (INIS)

    Anon.

    1980-01-01

    The General Public Utilities Corporation (GPU) faces bankruptcy or reorganization without Federal financial help and rate relief for the cleanup at Three Mile Island, but neither the Administration nor the Pennsylvania Public Utilities Commission is inclined to help. Bankruptcy will leave GPU's customers without power and will leave Unit 2 contaminated, making it unlikely that the courts will permit dissolution. The Nuclear Regulatory Commission's permission to restore Unit 1 could make financial recovery possible. Its reluctance to do so and its use of the psychological stress factor can be interpreted as applying a double standard to Babcock and Wilcox reactors

  19. Mechanical core coupling and reactors stability

    International Nuclear Information System (INIS)

    Suarez Antola, R.

    2006-01-01

    Structural parts of nuclear reactors are complex mechanical systems, able to vibrate with a set of proper frequencies when suitably excited. Cyclical variations in the strain state of the materials, including density perturbations, are produced. This periodic changes may affect reactor reactivity. But a variation in reactivity affects reactor thermal power, thus modifying the temperature field of the abovementiones materials. If the variation in temperature fields is fast enough, thermal-mechanical coupling may produce fast variations in strain states, and this, at its turn, modifies the reactivity, and so on. This coupling between mechanical vibrations of the structure and the materials of the core, with power oscillations of the reactor, not only may not be excluded a priori, but it seems that it has been present in some stage of the incidents or accidents that happened during the development of nuclear reactor technology. The purpose of the present communication is: (a) To review and generalize some mathematical models that were proposed in order to describe thermal-mechanical coupling in nuclear reactors. (b) To discuss some conditions in which significant instabilities could arise, including large amplitude power oscillations coupled with mechanical vibrations whose amplitudes are too small to be excluded by conventional criteria of mechanical design. Enough Certain aspects of thr physical safety of nuclear power reactors, that are objected by people that opposes to the renaissance of nucleoelectric generation, are discussed in the framework of the mathematical model proposed in this paper [es

  20. University Research Reactors: Issues and Challenges

    International Nuclear Information System (INIS)

    Bernard, John A.; Hu Linwen

    2000-01-01

    University research reactors are underutilized and, as a result, are being decommissioned. The reason for the lack of utilization is shown to be a chronic inability to generate sufficient funds to procure and maintain state-of-the-art instrumentation for prospective researchers. The role of these reactors in nuclear science/engineering education is explored and the rationale for their continued operation is presented. It is argued that base financial support for both reactor operations and the technical support staff needed to interface with experimenters is necessary if these research facilities are not to be irretrievably lost from the educational infrastructure of the United States

  1. High-temperature and breeder reactors - economic nuclear reactors of the future

    International Nuclear Information System (INIS)

    Djalilzadeh, A.M.

    1977-01-01

    The thesis begins with a review of the theory of nuclear fission and sections on the basic technology of nuclear reactors and the development of the first generation of gas-cooled reactors applied to electricity generation. It then deals in some detail with currently available and suggested types of high temperature reactor and with some related subsidiary issues such as the coupling of different reactor systems and various schemes for combining nuclear reactors with chemical processes (hydrogenation, hydrogen production, etc.), going on to discuss breeder reactors and their application. Further sections deal with questions of cost, comparison of nuclear with coal- and oil-fired stations, system analysis of reactor systems and the effect of nuclear generation on electricity supply. (C.J.O.G.)

  2. Reactor and method for production of nanostructures

    Science.gov (United States)

    Sunkara, Mahendra Kumar; Kim, Jeong H.; Kumar, Vivekanand

    2017-04-25

    A reactor and method for production of nanostructures, including metal oxide nanowires or nanoparticles, are provided. The reactor includes a regulated metal powder delivery system in communication with a dielectric tube; a plasma-forming gas inlet, whereby a plasma-forming gas is delivered substantially longitudinally into the dielectric tube; a sheath gas inlet, whereby a sheath gas is delivered into the dielectric tube; and a microwave energy generator coupled to the dielectric tube, whereby microwave energy is delivered into a plasma-forming gas. The method for producing nanostructures includes providing a reactor to form nanostructures and collecting the formed nanostructures, optionally from a filter located downstream of the dielectric tube.

  3. Some questions and answers concerning fast reactors

    International Nuclear Information System (INIS)

    Marshall, W.

    1980-01-01

    The theme of the lecture is the place of the fast reactor in an evolving nuclear programme. The whole question of plutonium is first considered, ie its method of production and the ways in which it can be used in the fast reactor fuel cycle. Whether fast reactors are necessary is then discussed. Their safety is examined with particular attention to those design features which are most criticised ie high volumetric power density of the core, and the use of liquid sodium as coolant. Attention is then paid to environmental and safeguard aspects. (U.K.)

  4. Nuclear reactor philosophy and criteria

    International Nuclear Information System (INIS)

    Atchison, R.J.

    1979-07-01

    Nuclear power plant safety criteria and principles developed in Canada are directed towards minimizing the chance of failure of the fuel and preventing or reducing to an acceptably low level the escape of fission products should fuel failure occur. Safety criteria and practices are set forth in the Reactor Siting Guide, which is based upon the concept of defence in depth. The Guide specifies that design and construction shall follow the best applicable code, standard or practice; the total of all serious process system failures shall not exceed one in three years; special safety systems are to be physically and functionally separate from process systems and each other; and safety systems shall be testable, with unavailability less than 10 - 3 . Doses to the most exposed member of the public due to normal operation, serious process failures, and dual failures are specified. Licensees are also required to consider the effects of extreme conditions due to airplane crashes, explosions, turbine disintegration, pipe burst, and natural disasters. Safety requirements are changing as nuclear power plant designs evolve and in response to social and economic pressures

  5. University of Florida potato variety trials spotlight: 'Peter Wilcox'

    Science.gov (United States)

    'Peter Wilcox’ is a fresh market potato variety selected from progeny of a cross between B0810-1 and B0918-5, and tested under the pedigree B1816-5 by K.G. Haynes. It was jointly released by United States Department of Agriculture, North Carolina Agricultural Research Service, Agricultural Experimen...

  6. Fully integrated analysis of reactor kinetics, thermalhydraulics and the reactor control system in the MAPLE-X10 research reactor

    International Nuclear Information System (INIS)

    Shim, S.Y.; Carlson, P.A.; Baxter, D.K.

    1992-01-01

    A prototype research reactor, designated MAPLE-X10 (Multipurpose Applied Physics Lattice Experimental - X 10MW), is currently being built at AECL's Chalk River Laboratories. The CATHENA (Canadian Algorithm for Thermalhydraulic Network Analysis) two-fluid code was used in the safety analysis of the reactor to determine the adequacy of core cooling during postulated reactivity and loss-of-forced-flow transients. The system responses to a postulated transient are predicted including the feedback between reactor kinetics, thermalhydrauilcs and the reactor control systems. This paper describes the MAPLE-X10 reactor and the modelling methodology used. Sample simulations of postulated loss-of-heat-sink and loss-of-regulation transients are presented. (author)

  7. Reactor Simulator Integration and Testing

    Science.gov (United States)

    Schoenfield, M. P.; Webster, K. L.; Pearson, J. B.

    2013-01-01

    As part of the Nuclear Systems Office Fission Surface Power Technology Demonstration Unit (TDU) project, a reactor simulator (RxSim) test loop was designed and built to perform integrated testing of the TDU components. In particular, the objectives of RxSim testing were to verify the operation of the core simulator, the instrumentation and control system, and the ground support gas and vacuum test equipment. In addition, it was decided to include a thermal test of a cold trap purification design and a pump performance test at pump voltages up to 150 V because the targeted mass flow rate of 1.75 kg/s was not obtained in the RxSim at the originally constrained voltage of 120 V. This Technical Memorandum summarizes RxSim testing. The gas and vacuum ground support test equipment performed effectively in NaK fill, loop pressurization, and NaK drain operations. The instrumentation and control system effectively controlled loop temperature and flow rates or pump voltage to targeted settings. The cold trap design was able to obtain the targeted cold temperature of 480 K. An outlet temperature of 636 K was obtained, which was lower than the predicted 750 K but 156 K higher than the cold temperature, indicating the design provided some heat regeneration. The annular linear induction pump tested was able to produce a maximum flow rate of 1.53 kg/s at 800 K when operated at 150 V and 53 Hz.

  8. DIGITAL FLOOD INSURANCE RATE MAP DATABASE, Wilcox COUNTY, AL

    Data.gov (United States)

    Federal Emergency Management Agency, Department of Homeland Security — The Digital Flood Insurance Rate Map (DFIRM) Database depicts flood risk information and supporting data used to develop the risk data. The primary risk...

  9. OrthoImagery Submission for Wilcox County, GA

    Data.gov (United States)

    Federal Emergency Management Agency, Department of Homeland Security — NAIP imagery is available for distribution within 60 days of the end of a flying season and is intended to provide current information of agricultural conditions in...

  10. Nuclear reactor safety: physics and engineering aspects

    International Nuclear Information System (INIS)

    Kinchin, G.H.

    1982-01-01

    In order to carry out the sort of probabilistic analysis referred to by Farmer (Contemp. Phys.; 22:349(1981)), it is necessary to have a good understanding of the processes involved in both normal and accident conditions in a nuclear reactor. Some of these processes, for a variety of different reactor systems, are considered in sections dealing with the neutron chain reaction, the removal of heat from the reactor, material problems, reliability of protective systems and a number of specific topics of particular interest from the point of view of physics or engineering. (author)

  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. Reactor accident analysis and evaluation

    International Nuclear Information System (INIS)

    Chang, J.W.

    1983-01-01

    Reactor Management Division of Korea Advanced Energy Research Institute has, so far, adopted, modified and developed quite a number of large programs for nuclear core analysis. During the course of this work, it was found necessary to employ some standard subroutines for handling data, input procedures, core memory management and search files. Many programs share lots of common subroutines and/or functions with other programs. Above all, some of them are in lack of transmittal. During the installation of big codes for CYBER computer, it has drawn our keen attention that many elementary subroutines are heavily machine-dependent and that their conversion is extremely difficult. After having collected and modified the subroutines to fit in different codes, it was finally named KINEP (KAERI Improved Nuclear Environmental Package). KINEP has been proved to be convenient even for smaller programs for general purpose. The KINEP includes about one hundred subroutines to facilitate data handling, operator communications, storage allocation, decimal input, file maintence and scratch I/O. (Author)

  13. Current status and prospects of research reactors

    International Nuclear Information System (INIS)

    Gabaraev, A.B.; Cherepnin, Yu.S.; Tretyakov, I.T.; Khmelshikov, V.V.; Dollezhal, N.A.

    2009-01-01

    Full text: The first nuclear research reactors (RR) appeared in the 1940s. Their initial purpose was to provide knowledge of the main processes associated with neutron-induced nuclear reactions. Later, the rang of problems addressed expanded substantially. Besides fundamental research in the properties of matter, such reactors are successfully used for dealing with problems in the fields of materials science, nuclear engineering, medicine, isotope production, education, etc. Over the whole period of RR fleet growth, more than six hundred nuclear research facilities were built in 70 countries of the world. As of the end of 2008, the number of Russian research reactors in service was about 20% of the globally operating RR fleet. This paper discusses the current status of the world's RR fleet and describes the capabilities of the experimental reactor facilities existing in Russia. In the 21st century, research reactors will remain in demand to solve scientific and technological problems for innovative development of society. The emerging renaissance of nuclear power, the expanding RR uses for production of isotopes and other applications, the increase in the number of countries willing to use nuclear technologies in energy production, industry and science - all contribute to a rebirth of interest in research reactors. One of the ways to improve the experimental capabilities lies in radical upgrading of the reactor facilities with qualitative changes in the main neutronic characteristics of the core. The associated design approaches are illustrated with the example of the IBR-2M reactor at the JNRI in Dubna. The imperative need restricting the spread of nuclear threat leads us to give up using highly enriched uranium in most research reactors. Development of RR fuel with reduced enrichment in uranium has been one of the priority objectives of NIKIET for many years. This paper presents the latest results obtained along these lines, as applied to pool-type research

  14. Cooperation in reactor design evaluation and licensing

    Energy Technology Data Exchange (ETDEWEB)

    Kaufer, B.; Wasylyk, A. [World Nuclear Association, London (United Kingdom)

    2014-07-01

    In January 2007 the World Nuclear Association (WNA) established the Cooperation in Reactor Design Evaluation and Licensing (CORDEL) Working Group with the aim of stimulating a dialogue between the nuclear industry (including reactor vendors, operators and utilities) and nuclear regulators (national and international organisations) on the benefits and means of achieving a worldwide convergence of reactor safety standards for reactor designs. From the time of its inception to the present, CORDEL has evolved from a group of experts discussing how to achieve international standardisation in nuclear safety design to an established and recognised working group dedicated to analysing and forging common understandings in key areas as input to major decisions on nuclear energy policy. This paper will review the general directions and activities CORDEL plans to undertake during the next five-year period, including its general strategy, activities, priorities and interactions with its customers in order to meet its objectives. (author)

  15. Monitoring and control of anaerobic reactors

    DEFF Research Database (Denmark)

    Pind, Peter Frode; Angelidaki, Irini; Ahring, Birgitte Kiær

    2003-01-01

    The current status in monitoring and control of anaerobic reactors is reviewed. The influence of reactor design and waste composition on the possible monitoring and control schemes is examined. After defining the overall control structure, and possible control objectives, the possible process mea...... control approaches that have been used are comprehensively described. These include simple and adaptive controllers, as well as more recent developments such as fuzzy controllers, knowledge-based controllers and controllers based on neural networks....

  16. Space reactors - past, present, and future

    International Nuclear Information System (INIS)

    Buden, D.; Angelo, J.

    1983-01-01

    In the 1990s and beyond, advanced-design nuclear reactors could represent the prime source of both space power and propulsion. Many sophisticated military and civilian space missions of the future will require first kilowatt and then megawatt levels of power. This paper reviews key technology developments that accompanied past US space nuclear power development efforts, describes on-going programs, and then explores reactor technologies that will satisfy megawatt power level needs and beyond

  17. Laryngopharyngeal Reflux: Diagnosis, Treatment, and Latest Research

    Directory of Open Access Journals (Sweden)

    Campagnolo, Andrea Maria

    2014-01-01

    Full Text Available Introduction Laryngopharyngeal reflux (LPR is a highly prevalent disease and commonly encountered in the otolaryngologist's office. Objective To review the literature on the diagnosis and treatment of LPR. Data Synthesis LPR is associated with symptoms of laryngeal irritation such as throat clearing, coughing, and hoarseness. The main diagnostic methods currently used are laryngoscopy and pH monitoring. The most common laryngoscopic signs are redness and swelling of the throat. However, these findings are not specific of LPR and may be related to other causes or can even be found in healthy individuals. Furthermore, the role of pH monitoring in the diagnosis of LPR is controversial. A therapeutic trial with proton pump inhibitors (PPIs has been suggested to be cost-effective and useful for the diagnosis of LPR. However, the recommendations of PPI therapy for patients with a suspicion of LPR are based on the results of uncontrolled studies, and high placebo response rates suggest a much more complex and multifactorial pathophysiology of LPR than simple acid reflux. Molecular studies have tried to identify biomarkers of reflux such as interleukins, carbonic anhydrase, E-cadherin, and mucin. Conclusion Laryngoscopy and pH monitoring have failed as reliable tests for the diagnosis of LPR. Empirical therapy with PPIs is widely accepted as a diagnostic test and for the treatment of LPR. However, further research is needed to develop a definitive diagnostic test for LPR.

  18. Safety and environmental aspects of fusion reactors

    International Nuclear Information System (INIS)

    Kilic, H.; Jensen, B.

    1982-01-01

    This paper deals with those problems concerning safety and environmental aspects of the future fusion reactors (e.g. fuel cycle, magnetic failure, after heat disturbances, radioactive waste and magnetic field)

  19. Reactor Neutrino Experiments: Present and Future

    Science.gov (United States)

    Wen, L. J.; Cao, J.; Wang, Y. F.

    2017-10-01

    Reactor neutrinos have been an important tool for both discovery and precision measurement in the history of neutrino studies. Since the first generation of reactor neutrino experiments in the 1950s, the detector technology has advanced greatly. New ideas, new knowledge, and modern software have also enhanced the power of the experiments. The current reactor neutrino experiments, Daya Bay, Double Chooz, and RENO, have led neutrino physics into the precision era. In this article, we review these developments and advances, address the key issues in designing a state-of-the-art reactor neutrino experiment, and explain how the challenging requirements of determining the neutrino mass hierarchy with the next-generation experiment JUNO could be realized in the near future.

  20. Small and medium power reactors 1987

    International Nuclear Information System (INIS)

    1987-12-01

    This TECDOC follows the publication of TECDOC-347 Small and Medium Power Reactors Project Initiation Study - Phase I published in 1985 and TECDOC-376 Small and Medium Power Reactors 1985 published in 1986. It is mainly intended for decision makers in Developing Member States interested in embarking on a nuclear power programme. It consists of two parts: 1) Guidelines for the Introduction of Small and Medium Power Reactors in Developing Countries. These Guidelines were established during the Advisory Group Meeting held in Vienna from 11 to 15 May 1987. Their purpose is to review key aspects relating to the introduction of Small and Medium Power Reactors in developing countries; 2) Up-dated Information on SMPR Concepts Contributed by Supplier Industries. According to the recommendations of the Second Technical Committee Meeting on SMPRs held in Vienna in March 1985, this part contains the up-dated information formerly published in Annex I of the above mentioned TECDOC-347. Figs

  1. Research reactors: a tool for science and medicine

    International Nuclear Information System (INIS)

    Ordonez, Juan

    2001-01-01

    The types and uses of research reactors are reviewed. After an analysis of the world situation, the demand of new research reactors of about 20 MW is foreseen. The experience and competitiveness of INVAP S.E. as designer and constructor of research reactors is outlined and the general specifications of the reactors designed by INVAP for Egypt and Australia are given

  2. Review of current and proposed reactor upgrades

    International Nuclear Information System (INIS)

    Moon, R.M.

    1985-01-01

    In an effort to foresee the future health of neutron scattering, a survey of plans to upgrade reactors and associated experimental facilities was undertaken. The results indicate that we are now entering a period characterized by a substantial reinvestment in reactor sources and expansion in the number of neutron scattering instruments. For the group of institutions participating in this survey there will be a total investment in improved sources and experimental facilities of $500 M to $1,000 M over the next decade. This investment will result in a 30 to 40% increase in the total power of research reactors and an increase of 30 to 50% in the number of neutron scattering instruments. It is therefore reasonable to anticipate an approximate doubling in the number of reactor neutrons incident on samples in the mid 90s compared to the present

  3. 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

  4. Blankets for fusion reactors : materials and neutronics

    International Nuclear Information System (INIS)

    Carvalho, S.H. de.

    1980-03-01

    The studies about Fusion Reactors have lead to several problems for which there is no general agreement about the best solution. Nevertheless, several points seem to be well defined, at least for the first generation of reactors. The fuel, for example, should be a mixture of deuterium and tritium. Therefore, the reactor should be able to generate the tritium to be burned and also to transform kinetic energy of the fusion neutrons into heat in a process similar to the fission reactors. The best materials for the composition of the blanket were first selected and then the neutronics for the proposed system was developed. The neutron flux in the blanket was calculated using the discrete ordinates transport code, ANISN. All the nuclides cross sections came from the DLC-28/CTR library, that processed the ENDF/B data, using the SUPERTOG Program. (Author) [pt

  5. Small reactors and the 'second nuclear era'

    International Nuclear Information System (INIS)

    Egan, J.R.

    1984-01-01

    Predictions of the nuclear industry's demise are premature and distort both history and politics. The industry is reemerging in a form commensurate with the priorities of those people and nations controlling the global forces of production. The current lull in plant orders is due primarily to the world recession and to factors related specifically to reactor size. Traditional economies of scale for nuclear plants have been greatly exaggerated. Reactor vendors and governments in Great Britain, France, West Germany, Japan, the United States, Sweden, Canada, and the Soviet Union are developing small reactors for both domestic applications and export to the Third World. The prefabricated, factory-assembled plants under 500 MWe may alleviate many of the existing socioeconomic constraints on nuclear manufacturing, construction, and operation. In the industrialized world, small reactors could furnish a qualitatively new energy option for utilities. But developing nations hold the largest potential market for small reactors due to the modest size of their electrical systems. These units could double or triple the market potential for nuclear power in this century. Small reactors will both qualitatively and quantitatively change the nature of nuclear technology transfers, offering unique advantages and problems vis-a-vis conventional arrangements. (author)

  6. The fast reactor and energy supply

    International Nuclear Information System (INIS)

    1979-01-01

    The progress made with fast reactor development in many countries is summarised showing that the aim is to provide to the nation concerned an ability to instal fast reactor power stations at the end of this century or early in the next one. Accepting the importance of fast reactors as a potential independent source of energy, problems concerning economics, industrial capability, technical factors, public acceptibility and in particular plutonium management, are discussed. It is concluded that although fast reactors have reached a comparatively advanced stage of development, a number of factors make it likely that their introduction for electricity generation will be a gradual process. Nevertheless it is necessary to complete demonstration and development phases in good time. (U.K.)

  7. Gas-cooled reactors and their applications

    International Nuclear Information System (INIS)

    1987-10-01

    The purpose of the meeting was to review and discuss the current status and recent progress made in the technology and design of gas-cooled reactors and their application for electricity generation, process steam and process heat production. The meeting was attended by more than 200 participants from 25 countries and International Organizations presenting 34 papers. The technical part of the meeting was subdivided into 7 sessions: A. Overview of the Status of Gas-Cooled Reactors and Their Prospects (2 papers); B. Experience with Gas-Cooled Reactors (5 papers); C. Description of Current GCR Plant Designs (10 papers); D. Safety Aspects (4 papers); E. Gas-Cooled Reactor Applications (3 papers); F. Gas-Cooled Reactor Technology (6 papers); G. User's Perspectives on Gas-Cooled Reactors (4 papers). At the end of the meeting a round table discussion was organized in order to summarize the meeting and to make recommendations for future activities. A separate abstract was prepared for each of the 34 presentations of this meeting. Refs, figs and tabs

  8. MOX in reactors: present and future

    International Nuclear Information System (INIS)

    Arslan, Marc; Gros, Jean Pierre; Niquille, Aurelie; Marincic, Alexis

    2010-01-01

    In Europe, MOX fuel has been supplied by AREVA for more than 30 years, to 36 reactors: 21 in France, 10 in Germany, 3 in Switzerland, 2 in Belgium. For the present and future, recycling is compulsory in the frame of sustainable development of nuclear energy. By 2030 the overall volume of used fuel will reach about 400 000 t worldwide. Their plutonium and uranium content represents a huge resource of energy to recycle. That is the reason why, the European Utilities issued an EUR (European Utilities Requirement) demanding new builds reactors to be able of using MOX Fuel Assemblies in up to 50 % of the core. AREVA GEN3+ reactors, like EPR TM or ATMEA TM designed with MHI partnership, are designed to answer any utility need of MOX recycling. The example of the EPR TM reactor operated with 100 % MOX core optimized for MOX recycling will be presented. A standard EPR TM can be operated with 100 % MOX core using an advanced homogeneous MOX (single Pu content) with highly improved performances (burn-up and Cycle length). The adaptations needed and the main operating and safety reactor features will be presented. AREVA offers the utilities throughout the world, fuel supply (UO 2 , ERU, MOX), and reactors designed with all the needed capability for recycling. For each country and each utility, an adapted global solution, competitive and non proliferant can be proposed. (authors)

  9. The energy gap and the fast reactor

    International Nuclear Information System (INIS)

    Hill, J.

    1977-01-01

    The background to the development of fast reactors is summarized. In Britain, the results of the many experiments performed, the operation of the Dounreay Fast Reactor for the past 18 years and the first year's operation of the larger Prototype Fast Reactor have all been very encouraging, in that they demonstrated that the performance corresponded well with predictions, breeding is possible, and the system is exceptionally stable in operation. The next step in fast reactor engineering is to build a full-scale fast reactor power station. There would seem to be little reason to expect more trouble than could reasonably be expected in constructing any large project of this general nature. However, from an engineering point of view continuity of experience is required. If a decision to build a commercial fast reactor were taken today there would be a 14-year gap between strating this and the start of the Prototype Fast Reactor. This is already much too long. From an environmental standpoint we have to demonstrate that we can manufacture and reprocess fast reacctor fuel for a substantial programme in a way that does not lead to pollution of the environment, and that plutonium-containing fuel can be transported in the quantities required in safety and in a way that does not attract terrorists or require a private army to ensure its security. Finally, we have to find a way to allow many countries to obtain the energy they need from fast reactors, without leading to the proliferation of nuclear weapons or weapons capability. (author)

  10. Future fuel cycle and reactor strategies

    International Nuclear Information System (INIS)

    Meneley, D.A.

    1999-01-01

    Within the framework of the 1997 IAEA Symposium 'Future Fuel Cycle and Reactor Strategies Adjusting to New Realities', Working Group No.3 produced a Key Issues paper addressing the title of the symposium. The scope of the Key Issues paper included those factors that are expected to remain or become important in the time period from 2015 to 2050, considering all facets of nuclear energy utilization from ore extraction to final disposal of waste products. The paper addressed the factors influencing the choice of reactor and fuel cycle. It then addressed the quantitatively largest category of reactor types expected to be important during the period; that is, thermal reactors burning uranium and plutonium fuel. The fast reactor then was discussed both as a stand-alone technology and as might be used in combination with thermal reactors. Thorium fuel use was discussed briefly. The present paper includes of a digest of the Key Issues Paper. Some comparisons arc made between the directions suggested in that paper and those indicated by the Abstracts of this Technical Committee Meeting- Recommendations are made for work which might be undertaken in the short and medium time frames, to ensure that fuel cycle technologies and processes established by the year 2050 will support the continuation of nuclear energy applications in the long term. (author)

  11. Actinide recycling for reactor waste mass and radiotoxicity reduction

    International Nuclear Information System (INIS)

    Renard, A.; Maldague, T.; Pilate, S.; Journet, J.; Rome, M.; Harislur, A.; Vergnes, J.

    1994-01-01

    The long-term radiotoxicity of nuclear waste from a Light Water Reactor fuel is analyzed; it can be reduced by multiple recycling of actinides in fast reactors. The capabilities of a first recycling in the light water reactor itself are evaluated with regard to implications on reactor physics and core management. Two main options are compared with their penalties and efficiency

  12. Operation and utilizations of Dalat nuclear research reactor

    International Nuclear Information System (INIS)

    Hien, P.Z.

    1988-01-01

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

  13. History and evolution of the breeder reactor

    International Nuclear Information System (INIS)

    Carle, R.

    1989-01-01

    The concept of the breeder reactor is almost as old as the idea of the nuclear reactor itself. From the very first years following the discovery of nuclear fission, scientists and technicians tried to turn mankind's eternal dream into reality; that is, enjoy an abundant source of energy without using up our raw material reserves. Nuclear energy offered several solutions to realize this dream. One of them, fusion, seemed out of our grasp in the near future. But fission of 235 U was possible, and the Manhattan Project soon furnished ample proof of this theory. However, everyone working in this field was conscious of the fact that thermal neutron reactors make very inefficient use of the energy potential contained in natural uranium. The solution was to use in a core sufficiently rich in fissile matter, the excess neutrons to convert the 238 U, so poorly used by other types of reactors, into fissile 239 Pu. Regeneration, or 'breeding' of fuel, can multiply the energy drawn from a ton of uranium by a factor of 50 to 100. This would enable us to ward off the specter of an energy shortage and the rapid depletion of uranium mines. As early as 1945 in Los Alamos, Enrico Fermi stated: 'The country which first develops a breeder reactor will have a great competitive edge in atomic energy.' The development of the breeder reactor in the USA and around the world is discussed

  14. TREATMENT OF METHANOLIC WASTEWATER BY ANAEROBIC DOWN-FLOW HANGING SPONGE (ANDHS) REACTOR AND UASB REACTOR

    Science.gov (United States)

    Sumino, Haruhiko; Wada, Keiji; Syutsubo, Kazuaki; Yamaguchi, Takashi; Harada, Hideki; Ohashi, Akiyoshi

    Anaerobic down-flow hanging sponge (AnDHS) reactor and UASB reactor were operated at 30℃ for over 400 days in order to investigate the process performance and the sludge characteristics of treating methanolic wastewater (2 gCOD/L). The settings OLR of AnDHS reactor and of UASB reactor were 5.0 -10.0 kgCOD/m3/d and 5.0 kgCOD/m3/d. The average of the COD removal demonstrated by both reactors were over 90% throughout the experiment. From the results of methane producing activities and the PCR-DGGE method, most methanol was directly converted to methane in both reactors. The conversion was carried out by different methanogens: one closely related to Methanomethylovorans hollandica in the AnDHS retainted sludge and the other closely related to Methanosarcinaceae and Metanosarciales in the UASB retainted sludge.

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

    International Nuclear Information System (INIS)

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

    1986-01-01

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

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

    International Nuclear Information System (INIS)

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

    1987-01-01

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

  17. Review of advanced reactor transient analysis capabilities and applications for Savannah River Plant reactors

    International Nuclear Information System (INIS)

    Buckner, M.R.; Hostetler, D.E.; Anderson, M.M.; Dodds, H.L.

    1977-01-01

    GRASS is a three-dimensional, coupled neutronic and engineering code for analysis of the radioisotope production reactors at the Savannah River Plant. The capabilities of GRASS are reviewed with emphasis on recent additions to model accident conditions involving the transport of molten fuel material and to accurately characterize neutronic and engineering feedback. The general application of GRASS to the Savannah River reactors is discussed, and results are presented for the analyses of severla reactor transient calculations

  18. Future view of total energy system and reactor engineering and reactor physics

    International Nuclear Information System (INIS)

    Ozawa, T.

    1974-01-01

    This paper outlines the present status of fission reactors and fusion reactors. The conversion ratio of light water reactors is 0.5, and the efficiency is 32% because of relatively low temperature. Both pressurized water reactors and boiling water reactors are technically well developed, their performances are well known, and the fuel cycle is well developed, so that both reactors have monopolized power reactor market. But the reprocessing of spent fuel and the treatment of their hazards are inevitable, and the construction and enlargement of reprocessing facilities are indispensable. In LMFBR's tight sealing is easy because they are non-pressurized, and the efficiency is 41%. But liquid sodium is strongly activated and recirculated, so that chemical obstruction due to the breakage of recirculating pumps, pipings, and heat exchangers may occur, and the hazard of plutonium is large. Regarding controlled thermo-nuclear fusion reactors, because Lawson criterion must be satisfied, two methods of plasma confinement are now experimented. One is the plasma confinement by strong magnetic field of 50 KG to 100 KG, and the other is the confinement by the implosion method with high-power laser beam. The latter has much more uncertainties than the former, but recently both methods have made much progress. (Tai, I)

  19. RELAP/SCDAPSIM Reactor System Simulator Development and Training for University and Reactor Applications

    International Nuclear Information System (INIS)

    Hohorst, J.K.; Allison, C.M.

    2010-01-01

    The RELAP/SCDAPSIM code, designed to predict the behaviour of reactor systems during normal and accident conditions, is being developed as part of an international nuclear technology development program called SDTP (SCDAP Development and Training Program). SDTP involves more than 60 organizations in 28 countries. One of the important applications of the code is for simulator training of university faculty and students, reactor analysts, and reactor operations and technical support staff. Examples of RELAP/SCDAPSIM-based system thermal hydraulic and severe accident simulator packages include the SAFSIM simulator developed by NECSA for the SAFARI research reactor in South Africa, university-developed simulators at the University of Mexico and Shanghai Jiao Tong University in China, and commercial VISA and RELSIM packages used for analyst and reactor operations staff training. This paper will briefly describe the different packages/facilities. (authors)

  20. Anti-high Mobility Group Box 1 Antibody Ameliorates Albuminuria in MRL/lpr Lupus-Prone Mice

    Directory of Open Access Journals (Sweden)

    Haruki Watanabe

    2017-09-01

    Full Text Available We evaluated the efficacy of a neutralizing anti-high mobility group box 1 (HMGB1 monoclonal antibody in MRL/lpr lupus-prone mice. The anti-HMGB1 monoclonal antibody (5 mg/kg weight or class-matched control immunoglobulin G2a (IgG2a was administered intravenously twice a week for 4–15 weeks. Urine albumin was monitored, and histological evaluation of the kidneys was conducted at 16 weeks. Lymphadenopathies were evaluated by 1-(2′-deoxy-2′-[18F]fluoro-β-D-arabinofuranosylcytosine ([18F]FAC positron emission tomography/computed tomography (PET/CT at 12 weeks. Following 4-week treatment, [18F]FAC-PET/CT showed similar accumulation in cervical and axillary lymph nodes at 12 weeks of age. However, anti-HMGB1 monoclonal antibody sufficiently inhibited the increase in albuminuria compared to an isotype control following 15-week treatment. Complement deposition was also improved; however, there were no significant differences in IgG deposition and renal pathological scores between the two groups. Anti-double-stranded DNA (dsDNA antibody titers and cytokine and chemokine levels were also unaltered. Although there were no significant differences in glomerular macrophage infiltration, neutrophil infiltration was significantly decreased by the anti-HMGB1 monoclonal antibody. Antagonizing HMGB1 treatment suppressed HMGB1 translocation from nuclei in the kidney and suppressed neutrophil extracellular traps. The anti-HMGB1 monoclonal antibody demonstrated therapeutic potential against albuminuria in lupus nephritis by inhibiting neutrophil recruitment and neutrophil extracellular traps.

  1. Status of advanced technology and design for water cooled reactors: Heavy water reactors

    International Nuclear Information System (INIS)

    1989-07-01

    In 1987 the IAEA established the International Working Group on Advanced Technologies for Water-Cooled Reactors (IWGATWR). Within the framework of the IWGATWR the IAEA Technical Report on Status of Advanced Technology and Design for Water Cooled Reactors, Part I: Light Water Reactors and Part II: Heavy Water Reactors, has been undertaken to document the major current activities and trends of technological improvement and development for future water reactors. Part I of the report dealing with Light Water Reactors (LWRs) was published in 1988 (IAEA-TECDOC-479). Part II of the report covers Heavy Water Reactors (HWRs) and has now been prepared. This report is based largely upon submissions from Member States. It has been supplemented by material from the presentations at the IAEA Technical Committee and Workshop on Progress in Heavy Water Reactor Design and Technology held in Montreal, Canada, December 6-9, 1988. It is hoped that this part of the report, containing the status of advanced heavy water reactor technology up to 1988 and ongoing development programmes will aid in disseminating information to Member States and in stimulating international cooperation. Refs, figs and tabs

  2. Reactor core and passive safety systems descriptions of a next generation pressure tube reactor - mechanical aspects

    Energy Technology Data Exchange (ETDEWEB)

    Yetisir, M.; Gaudet, M.; Rhodes, D.; Hamilton, H.; Pencer, J. [Atomic Energy of Canada Limited, Chalk River, ON (Canada)

    2014-07-01

    Canada has been developing a channel-type supercritical water-cooled nuclear reactor concept, often called the Canadian SCWR. The objective of this reactor concept is to meet the technology goals of the Generation IV International Forum (GIF) for the next generation nuclear reactor development, which include enhanced safety features (inherent safe operation and deploying passive safety features), improved resource utilization, sustainable fuel cycle, and greater proliferation resistance than Generation III nuclear reactors. The Canadian SCWR core concept consists of a high-pressure inlet plenum, a separate low-pressure heavy water moderator contained in a calandria vessel, and 336 pressure tubes surrounded by the moderator. The reactor uses supercritical water as a coolant, and a direct steam power cycle to generate electricity. The reactor concept incorporates advanced safety features such as passive core cooling, long-term decay heat rejection to the environment and fuel melt prevention via passive moderator cooling. These features significantly reduce core damage frequency relative to existing nuclear reactors. This paper presents a description of the design concepts for the Canadian SCWR core, reactor building layout and the plant layout. Passive safety concepts are also described that address containment and core cooling following a loss-of coolant accident, as well as long term reactor heat removal at station blackout conditions. (author)

  3. Homogeneous SLOWPOKE reactors for replacing SLOWPOKE-2 research reactors and the production of radioisotopes

    Energy Technology Data Exchange (ETDEWEB)

    Bonin, H.W., E-mail: bonin-h@rmc.ca [Royal Military College of Canada, Kingston, Ontario (Canada); Hilborn, J.W. [Canadian Nuclear Laboratories, Chalk River, Ontario (Canada); Carlin, G.E. [Ontario Power Generation, Toronto, Ontario (Canada); Gagnon, R.; Busatta, P. [Canadian Forces (Canada)

    2014-07-01

    Inspired from the inherently safe SLOWPOKE-2 research reactor, the Homogeneous SLOWPOKE reactor was conceived with a double goal: replacing the heterogeneous SLOWPOKE-2 reactors when they reach end-of-core life to continue their missions of neutron activation analysis and neutron radiography at universities, and to produce radioisotopes such as {sup 99}Mo for medical applications. A homogeneous reactor core allows a much simpler extraction of radioisotopes (such as {sup 99}Mo) for applications in industry and nuclear medicine. The 20 kW Homogeneous SLOWPOKE reactor was modelled using both the deterministic WIMS-AECL and the probabilistic MCNP 5 reactor simulation codes. The homogeneous fuel mixture was a dilute aqueous solution of Uranyl Sulfate (UO{sub 2}SO{sub 4}) with 994.2 g of {sup 235}U (enrichment at 20%) providing an excess reactivity at operating temperature (40 {sup o}C) of 3.8 mk for a molality determined as 1.46 mol kg{sup -1} for a Zircaloy-2 reactor vessel. Because this reactor is intended to replace the core of SLOWPOKE-2 reactors, the Homogeneous SLOWPOKE reactor core had a height about twice its diameter. The reactor could be controlled by mechanical absorber rods in the beryllium reflector, chemical control in the core, or a combination of both. The safety of the Homogeneous SLOWPOKE reactor was analysed for both normal operation and transient conditions. Thermal-hydraulics calculations used COMSOL Multiphysics and the results showed that natural convection was sufficient to ensure adequate reactor cooling in all situations. The most severe transient simulated resulted from a 5.87 mk step positive reactivity insertion to the reactor in operation at critical and at steady state at 20 {sup o}C. Peak temperature and power were determined as 83 {sup o}C and 546 kW, respectively, reached 5.1 s after the reactivity insertion. However, the power fell rapidly to values below 20 kW some 35 s after the peak and remained below that value thereafter. Both the

  4. Homogeneous SLOWPOKE reactors for replacing SLOWPOKE-2 research reactors and the production of radioisotopes

    International Nuclear Information System (INIS)

    Bonin, H.W.; Hilborn, J.W.; Carlin, G.E.; Gagnon, R.; Busatta, P.

    2014-01-01

    Inspired from the inherently safe SLOWPOKE-2 research reactor, the Homogeneous SLOWPOKE reactor was conceived with a double goal: replacing the heterogeneous SLOWPOKE-2 reactors when they reach end-of-core life to continue their missions of neutron activation analysis and neutron radiography at universities, and to produce radioisotopes such as 99 Mo for medical applications. A homogeneous reactor core allows a much simpler extraction of radioisotopes (such as 99 Mo) for applications in industry and nuclear medicine. The 20 kW Homogeneous SLOWPOKE reactor was modelled using both the deterministic WIMS-AECL and the probabilistic MCNP 5 reactor simulation codes. The homogeneous fuel mixture was a dilute aqueous solution of Uranyl Sulfate (UO 2 SO 4 ) with 994.2 g of 235 U (enrichment at 20%) providing an excess reactivity at operating temperature (40 o C) of 3.8 mk for a molality determined as 1.46 mol kg -1 for a Zircaloy-2 reactor vessel. Because this reactor is intended to replace the core of SLOWPOKE-2 reactors, the Homogeneous SLOWPOKE reactor core had a height about twice its diameter. The reactor could be controlled by mechanical absorber rods in the beryllium reflector, chemical control in the core, or a combination of both. The safety of the Homogeneous SLOWPOKE reactor was analysed for both normal operation and transient conditions. Thermal-hydraulics calculations used COMSOL Multiphysics and the results showed that natural convection was sufficient to ensure adequate reactor cooling in all situations. The most severe transient simulated resulted from a 5.87 mk step positive reactivity insertion to the reactor in operation at critical and at steady state at 20 o C. Peak temperature and power were determined as 83 o C and 546 kW, respectively, reached 5.1 s after the reactivity insertion. However, the power fell rapidly to values below 20 kW some 35 s after the peak and remained below that value thereafter. Both the temperature and void coefficients are

  5. Advanced Demonstration and Test Reactor Options Study

    International Nuclear Information System (INIS)

    Petti, David Andrew; Hill, R.; Gehin, J.; Gougar, Hans David; Strydom, Gerhard; Heidet, F.; Kinsey, J.; Grandy, Christopher; Qualls, A.; Brown, Nicholas; Powers, J.; Hoffman, E.; Croson, D.

    2017-01-01

    Global efforts to address climate change will require large-scale decarbonization of energy production in the United States and elsewhere. Nuclear power already provides 20% of electricity production in the United States (U.S.) and is increasing in countries undergoing rapid growth around the world. Because reliable, grid-stabilizing, low emission electricity generation, energy security, and energy resource diversity will be increasingly valued, nuclear power's share of electricity production has a potential to grow. In addition, there are non electricity applications (e.g., process heat, desalination, hydrogen production) that could be better served by advanced nuclear systems. Thus, the timely development, demonstration, and commercialization of advanced nuclear reactors could diversify the nuclear technologies available and offer attractive technology options to expand the impact of nuclear energy for electricity generation and non-electricity missions. The purpose of this planning study is to provide transparent and defensible technology options for a test and/or demonstration reactor(s) to be built to support public policy, innovation and long term commercialization within the context of the Department of Energy's (DOE's) broader commitment to pursuing an 'all of the above' clean energy strategy and associated time lines. This planning study includes identification of the key features and timing needed for advanced test or demonstration reactors to support research, development, and technology demonstration leading to the commercialization of power plants built upon these advanced reactor platforms. This planning study is consistent with the Congressional language contained within the fiscal year 2015 appropriation that directed the DOE to conduct a planning study to evaluate 'advanced reactor technology options, capabilities, and requirements within the context of national needs and public policy to support innovation in nuclear

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

    International Nuclear Information System (INIS)

    Jin, Kyungho; Heo, Gyunyoung; Park, Jaekwan

    2014-01-01

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

  7. Open-ended fusion devices and reactors

    International Nuclear Information System (INIS)

    Kawabe, T.; Nariai, H.

    1983-01-01

    Conceptual design studies on fusion reactors based upon open-ended confinement schemes, such as the tandem mirror and rf plugged cusp, have been carried out in Japan. These studies may be classified into two categories: near-term devices (Fusion Engineering Test Facility), and long-term fusion power recators. In the first category, a two-component cusp neutron source was proposed. In the second category, the GAMMA-R, a tandem-mirror power reactor, and the RFC-R, an axisymetric mirror and cusp, reactor studies are being conducted at the University of Tsukuba and the Institute of Plasma Physics. Mirror Fusion Engineering Facility parameters and a schematic are shown. The GAMMA-R central-cell design schematic is also shown

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

    Science.gov (United States)

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

    2016-01-01

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

  9. H Reactor

    Data.gov (United States)

    Federal Laboratory Consortium — The H Reactor was the first reactor to be built at Hanford after World War II.It became operational in October of 1949, and represented the fourth nuclear reactor on...

  10. Advanced reactors and future energy market needs

    International Nuclear Information System (INIS)

    Paillere, Henri; )

    2017-01-01

    Based on the results of a very well-attended international workshop on 'Advanced Reactor Systems and Future Energy Market Needs' that took place in April 2017, the NEA has embarked on a two-year study with the objective of analysing evolving energy market needs and requirements, as well as examining how well reactor technologies under development today will fit into tomorrow's low-carbon world. The NEA Expert Group on Advanced Reactor Systems and Future Energy Market Needs (ARFEM) held its first meeting on 5-6 July 2017 with experts from Canada, France, Italy, Japan, Korea, Poland, Romania, Russia and the United Kingdom. The outcome of the study will provide much needed insight into how well nuclear can fulfil its role as a key low-carbon technology, and help identify challenges related to new operational, regulatory or market requirements

  11. A comparison of radioactive waste from first generation fusion reactors and fast fission reactors with actinide recycling

    International Nuclear Information System (INIS)

    Koch, M.; Kazimi, M.S.

    1991-04-01

    Limitations of the fission fuel resources will presumably mandate the replacement of thermal fission reactors by fast fission reactors that operate on a self-sufficient closed fuel cycle. This replacement might take place within the next one hundred years, so the direct competitors of fusion reactors will be fission reactors of the latter rather than the former type. Also, fast fission reactors, in contrast to thermal fission reactors, have the potential for transmuting long-lived actinides into short-lived fission products. The associated reduction of the long-term activation of radioactive waste due to actinides makes the comparison of radioactive waste from fast fission reactors to that from fusion reactors more rewarding than the comparison of radioactive waste from thermal fission reactors to that from fusion reactors. Radioactive waste from an experimental and a commercial fast fission reactor and an experimental and a commercial fusion reactor has been characterized. The fast fission reactors chosen for this study were the Experimental Breeder Reactor 2 and the Integral Fast Reactor. The fusion reactors chosen for this study were the International Thermonuclear Experimental Reactor and a Reduced Activation Ferrite Helium Tokamak. The comparison of radioactive waste parameters shows that radioactive waste from the experimental fast fission reactor may be less hazardous than that from the experimental fusion reactor. Inclusion of the actinides would reverse this conclusion only in the long-term. Radioactive waste from the commercial fusion reactor may always be less hazardous than that from the commercial fast fission reactor, irrespective of the inclusion or exclusion of the actinides. The fusion waste would even be far less hazardous, if advanced structural materials, like silicon carbide or vanadium alloy, were employed

  12. A comparison of radioactive waste from first generation fusion reactors and fast fission reactors with actinide recycling

    Energy Technology Data Exchange (ETDEWEB)

    Koch, M.; Kazimi, M.S.

    1991-04-01

    Limitations of the fission fuel resources will presumably mandate the replacement of thermal fission reactors by fast fission reactors that operate on a self-sufficient closed fuel cycle. This replacement might take place within the next one hundred years, so the direct competitors of fusion reactors will be fission reactors of the latter rather than the former type. Also, fast fission reactors, in contrast to thermal fission reactors, have the potential for transmuting long-lived actinides into short-lived fission products. The associated reduction of the long-term activation of radioactive waste due to actinides makes the comparison of radioactive waste from fast fission reactors to that from fusion reactors more rewarding than the comparison of radioactive waste from thermal fission reactors to that from fusion reactors. Radioactive waste from an experimental and a commercial fast fission reactor and an experimental and a commercial fusion reactor has been characterized. The fast fission reactors chosen for this study were the Experimental Breeder Reactor 2 and the Integral Fast Reactor. The fusion reactors chosen for this study were the International Thermonuclear Experimental Reactor and a Reduced Activation Ferrite Helium Tokamak. The comparison of radioactive waste parameters shows that radioactive waste from the experimental fast fission reactor may be less hazardous than that from the experimental fusion reactor. Inclusion of the actinides would reverse this conclusion only in the long-term. Radioactive waste from the commercial fusion reactor may always be less hazardous than that from the commercial fast fission reactor, irrespective of the inclusion or exclusion of the actinides. The fusion waste would even be far less hazardous, if advanced structural materials, like silicon carbide or vanadium alloy, were employed.

  13. Reactors

    International Nuclear Information System (INIS)

    Onuki, Koji; Sasanuma, Katsumi.

    1980-01-01

    Purpose: To make it possible to correctly measure the flow rate and temperatures of the coolants flowing through fuel assemblies. Constitution: One or more holes are formed at the side surface of the guide tube of a control rod driving mechanism thereby to reduce the flow path resistance within the guide tube of the control rod driving mechanism and to prevent the outlet coolant of the control rod guide tube from flowing into the guide tube of the mechanism as it is and also from flowing into ambient rectifying lattice guide tubes, so that the quantities and temperatures of the coolants flowing through respective fuel assemblies can be measured correctly. (Kamimura, M.)

  14. Tandem mirror and tokamak reactor maintainability comparison

    International Nuclear Information System (INIS)

    Zahn, H.S.

    1981-01-01

    The analysis proceeds through estimates of downtime and resources required for selected maintenance actions and optimization of the replacement fraction, availability and cost of electricity. Scheduled downtime estimates and availability goals provide a basis for determining allowable forced outage downtimes. These analyses have been conducted with the assumption of redundancy wherever feasible but without the impact of maintenance equipment outages. Annual maintenance cost estimates and availabilities for both reactors are found to be approximately equal. However, the tandem mirror reactor capital costs are higher. Reduction of these costs appears feasible with the trend of current design studies toward smaller and more accessible machines

  15. Nuclear reactor kinetics and plant control

    CERN Document Server

    Oka, Yoshiaki

    2013-01-01

    Understanding time-dependent behaviors of nuclear reactors and the methods of their control is essential to the operation and safety of nuclear power plants. This book provides graduate students, researchers, and engineers in nuclear engineering comprehensive information on both the fundamental theory of nuclear reactor kinetics and control and the state-of-the-art practice in actual plants, as well as the idea of how to bridge the two. The first part focuses on understanding fundamental nuclear kinetics. It introduces delayed neutrons, fission chain reactions, point kinetics theory, reactivit

  16. Nuclear reactor plants and control systems therefor

    International Nuclear Information System (INIS)

    de Boer, G.A.; de Hex, M.

    1976-01-01

    A nuclear reactor plant is described comprising at least two hydraulically separated but thermally interconnected heat conveying circuits, of which one is the reactor circuit filled with a non-water medium and the other one is the water-steam-circuit equipped with a steam generator, a feed water conduit controlled by a valve and a steam turbine, and a control system mainly influenced by the pressure drop caused in said feed water conduit and its control valve and having a value of at least 10 bars at full load

  17. The Traveling Wave Reactor: Design and Development

    Directory of Open Access Journals (Sweden)

    John Gilleland

    2016-03-01

    Full Text Available The traveling wave reactor (TWR is a once-through reactor that uses in situ breeding to greatly reduce the need for enrichment and reprocessing. Breeding converts incoming subcritical reload fuel into new critical fuel, allowing a breed-burn wave to propagate. The concept works on the basis that breed-burn waves and the fuel move relative to one another. Thus either the fuel or the waves may move relative to the stationary observer. The most practical embodiments of the TWR involve moving the fuel while keeping the nuclear reactions in one place−sometimes referred to as the standing wave reactor (SWR. TWRs can operate with uranium reload fuels including totally depleted uranium, natural uranium, and low-enriched fuel (e.g., 5.5% 235U and below, which ordinarily would not be critical in a fast spectrum. Spent light water reactor (LWR fuel may also serve as TWR reload fuel. In each of these cases, very efficient fuel usage and significant reduction of waste volumes are achieved without the need for reprocessing. The ultimate advantages of the TWR are realized when the reload fuel is depleted uranium, where after the startup period, no enrichment facilities are needed to sustain the first reactor and a chain of successor reactors. TerraPower's conceptual and engineering design and associated technology development activities have been underway since late 2006, with over 50 institutions working in a highly coordinated effort to place the first unit in operation by 2026. This paper summarizes the TWR technology: its development program, its progress, and an analysis of its social and economic benefits.

  18. Status of advanced technology and design for water cooled reactors: Light water reactors

    International Nuclear Information System (INIS)

    1988-10-01

    Water reactors represent a high level of performance and safety. They are mature technology and they will undoubtedly continue to be the main stream of nuclear power. There are substantial technological development programmes in Member States for further improving the technology and for the development of new concepts in water reactors. Therefore the establishment of an international forum for the exchange of information and stimulation of international co-operation in this field has emerged. In 1987 the IAEA established the International Working Group on Advanced Technologies for Water-Cooled Reactors (IWGATWR). Within the framework of IWGATWR the IAEA Technical Report on Status of Advanced Technology and Design for Water Cooled Reactors, Part I: Light Water Reactors and Part II: Heavy Water Reactors has been undertaken to document the major current activities and different trends of technological improvements and developments for future water reactors. Part I of the report dealing with LWRs has now been prepared and is based mainly on submissions from Member States. It is hoped that this part of the report, containing the status of advanced light water reactor design and technology of the year 1987 and early 1988 will be useful for disseminating information to Agency Member States and for stimulating international cooperation in this subject area. 93 refs, figs and tabs

  19. Anaerobic granular sludge and biofilm reactors

    DEFF Research Database (Denmark)

    Skiadas, Ioannis V.; Gavala, Hariklia N.; Schmidt, Jens Ejbye

    2003-01-01

    The long retention time of the active biomass in the high-rate anaerobic digesters is the key factor for the successful application of the high rate anaerobic wastewater treatment. The long solids retention time is achieved due to the specific reactor configuration and it is enhanced by the immob......The long retention time of the active biomass in the high-rate anaerobic digesters is the key factor for the successful application of the high rate anaerobic wastewater treatment. The long solids retention time is achieved due to the specific reactor configuration and it is enhanced...... by the immobilization of the biomass, which forms static biofilms, particle-supported biofilms, or granules depending on the reactor's operational conditions. The advantages of the high-rate anaerobic digestion over the conventional aerobic wastewater treatment methods has created a clear trend for the change...

  20. Decommissioning and decontrolling the R1-reactor

    International Nuclear Information System (INIS)

    Bergman, C.; Holmberg, B.T.

    1985-01-01

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

  1. Distributed computing and nuclear reactor analysis

    International Nuclear Information System (INIS)

    Brown, F.B.; Derstine, K.L.; Blomquist, R.N.

    1994-01-01

    Large-scale scientific and engineering calculations for nuclear reactor analysis can now be carried out effectively in a distributed computing environment, at costs far lower than for traditional mainframes. The distributed computing environment must include support for traditional system services, such as a queuing system for batch work, reliable filesystem backups, and parallel processing capabilities for large jobs. All ANL computer codes for reactor analysis have been adapted successfully to a distributed system based on workstations and X-terminals. Distributed parallel processing has been demonstrated to be effective for long-running Monte Carlo calculations

  2. On the research activities in reactor and neutron physics using the first egyptian research reactor

    International Nuclear Information System (INIS)

    Hassan, A.M.

    2000-01-01

    A review on the most important research activities in reactor and neutron physics using the first Egyptian Research Reactor (ET-RR-1) is given. An out look on: neutron cross-sections, neutron flux, neutron capture gamma-ray spectroscopy, neutron activation analysis, neutron diffraction and radiation shielding experiments, is presented

  3. Reactivity estimation for subcritical and critical reactors

    International Nuclear Information System (INIS)

    Benhaim A; Bellino P; Gomez A

    2012-01-01

    We developed a digital reactimeter that works in both current and pulse mode. This reactimeter will allow to estimate the reactivity of the reactor at any state. We st obtained for the measurements taken in the experimental reactor RA-1 the reactivity around the critical state without a neutron source. Measurements were made using simultaneously a compensated ionization chamber and a 3He proportional counter. The results were compared with the ones obtained from the digital reactimeter of reference with matching results within the experimental errors (author)

  4. Advanced Computational Thermal Fluid Physics (CTFP) and Its Assessment for Light Water Reactors and Supercritical Reactors

    International Nuclear Information System (INIS)

    D.M. McEligot; K. G. Condie; G. E. McCreery; H. M. McIlroy; R. J. Pink; L.E. Hochreiter; J.D. Jackson; R.H. Pletcher; B.L. Smith; P. Vukoslavcevic; J.M. Wallace; J.Y. Yoo; J.S. Lee; S.T. Ro; S.O. Park

    2005-01-01

    Background: The ultimate goal of the study is the improvement of predictive methods for safety analyses and design of Generation IV reactor systems such as supercritical water reactors (SCWR) for higher efficiency, improved performance and operation, design simplification, enhanced safety and reduced waste and cost. The objective of this Korean/US/laboratory/university collaboration of coupled fundamental computational and experimental studies is to develop the supporting knowledge needed for improved predictive techniques for use in the technology development of Generation IV reactor concepts and their passive safety systems. The present study emphasizes SCWR concepts in the Generation IV program

  5. Advanced Computational Thermal Fluid Physics (CTFP) and Its Assessment for Light Water Reactors and Supercritical Reactors

    Energy Technology Data Exchange (ETDEWEB)

    D.M. McEligot; K. G. Condie; G. E. McCreery; H. M. McIlroy; R. J. Pink; L.E. Hochreiter; J.D. Jackson; R.H. Pletcher; B.L. Smith; P. Vukoslavcevic; J.M. Wallace; J.Y. Yoo; J.S. Lee; S.T. Ro; S.O. Park

    2005-10-01

    Background: The ultimate goal of the study is the improvement of predictive methods for safety analyses and design of Generation IV reactor systems such as supercritical water reactors (SCWR) for higher efficiency, improved performance and operation, design simplification, enhanced safety and reduced waste and cost. The objective of this Korean / US / laboratory / university collaboration of coupled fundamental computational and experimental studies is to develop the supporting knowledge needed for improved predictive techniques for use in the technology development of Generation IV reactor concepts and their passive safety systems. The present study emphasizes SCWR concepts in the Generation IV program.

  6. Decommissioning and dismantling reactors and managing waste

    International Nuclear Information System (INIS)

    Bensoussan, E.; Reicher-Fournel, N.

    2005-01-01

    In the early forties/fifties, a number of countries launched the first developments in the field of nuclear power. Some of them now have large numbers of nuclear facilities and nuclear power plants which have met, and continue to meet, the objectives for which they were designed and built. Other plants, including nuclear fuel production and enrichment plants, experimental reactors or research reactors, will have to be dismantled and demolished in the near future. These activities are handled differently in different countries as a function of specific energy policies, advanced development plants, current financial resources, the availability of qualified engineers and specialized industries able to handle projects of this kind, as well as other factors. All dismantling and demolition projects serve the purpose of returning the respective sites to green-field conditions. (orig.)

  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. Design guide for category II reactors light and heavy water cooled reactors

    International Nuclear Information System (INIS)

    Brynda, W.J.; Lobner, P.R.; Powell, R.W.; Straker, E.A.

    1978-05-01

    The Department of Energy (DOE), in the ERDA Manual, requires that all DOE-owned reactors be sited, designed, constructed, modified, operated, maintained, and decommissioned in a manner that gives adequate consideration to health and safety factors. Specific guidance pertinent to the safety of DOE-owned reactors is found in Chapter 0540 of the ERDA Manual. The purpose of this Design Guide is to provide additional guidance to aid the DOE facility contractor in meeting the requirement that the siting, design, construction, modification operation, maintainance, and decommissioning of DOW-owned reactors be in accordance with generally uniform standards, guide and codes which are comparable to those applied to similar reactors licensed by the Nuclear Regulatory Commission (NRC). This Design Guide deals principally with the design and functional requirements of Category II reactor structure, components, and systems

  9. Seclazone Reactor Modeling And Experimental Validation

    Energy Technology Data Exchange (ETDEWEB)

    Osinga, T. [ETH-Zuerich (Switzerland); Olalde, G. [CNRS Odeillo (France); Steinfeld, A. [PSI and ETHZ (Switzerland)

    2005-03-01

    A numerical model is formulated for the SOLZINC solar chemical reactor for the production of Zn by carbothermal reduction of ZnO. The model involves solving, by the finite-volume technique, a 1D unsteady state energy equation that couples heat transfer to the chemical kinetics for a shrinking packed bed exposed to thermal radiation. Validation is accomplished by comparison with experimentally measured temperature profiles and Zn production rates as a function of time, obtained for a 5-kW solar reactor tested at PSI's solar furnace. (author)

  10. Reactor pressure vessel aging and countermeasures

    International Nuclear Information System (INIS)

    Leitz, C.

    1987-01-01

    The considerable aging effect on reactor pressure vessels is the effect of irradiation on material properties in the core beltline region. Modern LWRs in the Federal Republic of Germany are designed such that irradiation effects are very low. Countermeasures applicable separately or in combination for plants with higher than normally expected irradiation effects are described in three steps: first, examinations and calculations to extend the formal reactor lifetime by reducing over-conservative margins; second, changes in core design to reduce future irradiation effects; third, a procedure to recover irradiation effect on material properties already sustained. (orig./HP) [de

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

    International Nuclear Information System (INIS)

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

    1981-12-01

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

  12. Hydrogen and water reactor safety: proceedings

    International Nuclear Information System (INIS)

    1982-01-01

    Separate abstracts were prepared for papers presented in the following areas of interest: 1) hydrogen research programs; 2) hydrogen behavior during light water reactor accidents; 3) combustible gas generation; 4) hydrogen transport and mixing; 5) combustion modeling and experiments; 6) accelerated flames and detonations; 7) combustion mitigation and control; and 8) equipment survivability

  13. Hydrogen and water reactor safety: proceedings

    Energy Technology Data Exchange (ETDEWEB)

    1982-01-01

    Separate abstracts were prepared for papers presented in the following areas of interest: 1) hydrogen research programs; 2) hydrogen behavior during light water reactor accidents; 3) combustible gas generation; 4) hydrogen transport and mixing; 5) combustion modeling and experiments; 6) accelerated flames and detonations; 7) combustion mitigation and control; and 8) equipment survivability.

  14. Present and future oscillation experiments at reactors

    International Nuclear Information System (INIS)

    Mikaehlyan, L.A.

    2001-01-01

    A report is presented on recent progress and developments (since the NANP'99 Conference) in the current and future long baseline (∼100 - 800 km) oscillation experiments at reactors. These experiments, under certain assumptions, can fully reconstruct the internal mass structure of the electron neutrino and provide a laboratory test of solar and atmospheric neutrino problems

  15. Reactor costs and maintenance, with reference to the Culham Mark II conceptual tokamak reactor design

    International Nuclear Information System (INIS)

    Hancox, R.; Mitchell, J.T.D.

    1977-01-01

    Published designs of tokamak reactors have proposed conceptual solutions for most of the technological problems encountered. Two areas which remain uncertain, however, are the capital cost of the reactor and the practicability of reactor maintenance. A cost estimate for the Culham Conceptual Tokamak Reactor (Mk I) is presented. The capital cost of a power station incorporating this reactor would be significantly higher than that of an equivalent fast breeder fission power station, mainly because of the low power density of the fusion reactor which affects both the reactor and building costs. To reduce the fusion station capital costs a new conceptual design is proposed (Mk II) which incorporates a shaped plasma cross-section to give a higher plasma pressure ratio, βsub(t) approximately 0.1. Since the higher power density implies more severe radiation damage of the blanket structure, the question of reactor maintenance assumes greater importance. With the proposed scheme for regular replacement of the blanket, a fusion power station availability around 0.9 should be achievable. (author)

  16. Current Abstracts Nuclear Reactors and Technology

    Energy Technology Data Exchange (ETDEWEB)

    Bales, J.D.; Hicks, S.C. [eds.

    1993-01-01

    This publication Nuclear Reactors and Technology (NRT) announces on a monthly basis the current worldwide information available from the open literature on nuclear reactors and technology, including all aspects of power reactors, components and accessories, fuel elements, control systems, and materials. This publication contains the abstracts of DOE reports, journal articles, conference papers, patents, theses, and monographs added to the Energy Science and Technology Database during the past month. Also included are US information obtained through acquisition programs or interagency agreements and international information obtained through acquisition programs or interagency agreements and international information obtained through the International Energy Agency`s Energy Technology Data Exchange or government-to-government agreements. The digests in NRT and other citations to information on nuclear reactors back to 1948 are available for online searching and retrieval on the Energy Science and Technology Database and Nuclear Science Abstracts (NSA) database. Current information, added daily to the Energy Science and Technology Database, is available to DOE and its contractors through the DOE Integrated Technical Information System. Customized profiles can be developed to provide current information to meet each user`s needs.

  17. Power reactor noise studies and applications

    Energy Technology Data Exchange (ETDEWEB)

    Arzhanov, V

    2002-03-01

    The present thesis deals with the neutron noise arising in power reactor systems. Generally, it can be divided into two major parts: first, neutron noise diagnostics, or more specifically, novel methods and algorithms to monitor nuclear industrial reactors; and second, contributions to neutron noise theory as applied to power reactor systems. Neutron noise diagnostics is presented by two topics. The first one is a theoretical study on the possibility to use a newly proposed current-flux (C/F) detector in Pressurised Water Reactors (PWR) for the localisation of anomalies. The second topic concerns various methods to detect guide tube impacting in Boiling Water Reactors (BWR). The significance of these problems comes from the operational experience. The thesis describes a novel method to localise vibrating control rods in a PWR by using only one C/F detector. Another novel method, based on wavelet analysis, is put forward to detect impacting guide tubes in a BWR. Neutron noise theory is developed for both Accelerator Driven Systems (ADS) and traditional reactors. By design the accelerator-driven systems would operate in a subcritical mode with a strong external source. This calls for a revision of many concepts and methods that have been developed for traditional reactors and also it poses a number of new problems. As for the latter, the thesis investigates the space-dependent neutron noise caused by a fluctuating source. It is shown that the frequency-dependent spatial behaviour exhibits some new properties that are different from those known in traditional critical systems. On the other hand, various reactor physics approximations (point kinetic, adiabatic etc.) have not been defined yet for the subcritical systems. In this respect the thesis presents a systematic formulation of the above mentioned approximations as well as investigations of their properties. Another important problem in neutron noise theory is the treatment of moving boundaries. In this case one

  18. Power reactor noise studies and applications

    International Nuclear Information System (INIS)

    Arzhanov, V.

    2002-03-01

    The present thesis deals with the neutron noise arising in power reactor systems. Generally, it can be divided into two major parts: first, neutron noise diagnostics, or more specifically, novel methods and algorithms to monitor nuclear industrial reactors; and second, contributions to neutron noise theory as applied to power reactor systems. Neutron noise diagnostics is presented by two topics. The first one is a theoretical study on the possibility to use a newly proposed current-flux (C/F) detector in Pressurised Water Reactors (PWR) for the localisation of anomalies. The second topic concerns various methods to detect guide tube impacting in Boiling Water Reactors (BWR). The significance of these problems comes from the operational experience. The thesis describes a novel method to localise vibrating control rods in a PWR by using only one C/F detector. Another novel method, based on wavelet analysis, is put forward to detect impacting guide tubes in a BWR. Neutron noise theory is developed for both Accelerator Driven Systems (ADS) and traditional reactors. By design the accelerator-driven systems would operate in a subcritical mode with a strong external source. This calls for a revision of many concepts and methods that have been developed for traditional reactors and also it poses a number of new problems. As for the latter, the thesis investigates the space-dependent neutron noise caused by a fluctuating source. It is shown that the frequency-dependent spatial behaviour exhibits some new properties that are different from those known in traditional critical systems. On the other hand, various reactor physics approximations (point kinetic, adiabatic etc.) have not been defined yet for the subcritical systems. In this respect the thesis presents a systematic formulation of the above mentioned approximations as well as investigations of their properties. Another important problem in neutron noise theory is the treatment of moving boundaries. In this case one

  19. Reactor power system deployment and startup

    Science.gov (United States)

    Wetch, J. R.; Nelin, C. J.; Britt, E. J.; Klein, G.

    1985-01-01

    This paper addresses issues that should receive further examination in the near-term as concept selection for development of a U.S. space reactor power system is approached. The issues include: the economics, practicality and system reliability associated with transfer of nuclear spacecraft from low earth shuttle orbits to operational orbits, via chemical propulsion versus nuclear electric propulsion; possible astronaut supervised reactor and nuclear electric propulsion startup in low altitude Shuttle orbit; potential deployment methods for nuclear powered spacecraft from Shuttle; the general public safety of low altitude startup and nuclear safe and disposal orbits; the question of preferred reactor power level; and the question of frozen versus molten alkali metal coolant during launch and deployment. These issues must be considered now because they impact the SP-100 concept selection, power level selection, weight and size limits, use of deployable radiators, reliability requirements, and economics, as well as the degree of need for and the urgency of developing space reactor power systems.

  20. Present status and future prospect of research reactors

    International Nuclear Information System (INIS)

    Takemi, Hirokatsu

    1996-01-01

    The present status of research reactors more than MW class reactor in JAERI and the Kyoto University and the small reactors in the Musashi Institute of Technology, the Rikkyo University, the Tokyo University, the Kinki University and other countries are explained in the paper. The present status of researches are reported by the topics in each field. The future researches of the beam reactor and the irradiation reactor are reviewed. On various kinds of use of research reactor and demands of neutron field of a high order, new type research reactors under investigation are explained. Recently, the reactors are used in many fields such as the basic science: the basic physics, the material science, the nuclear physics, and the nuclear chemistry and the applied science; the earth and environmental science, the biology and the medical science. (S.Y.)

  1. Monitoring and control of anaerobic reactors

    DEFF Research Database (Denmark)

    Pind, Peter Frode; Angelidaki, Irini; Ahring, Birgitte Kiær

    2003-01-01

    measurements are reviewed in detail. In the sequel, possible manipulated variables, such as the hydraulic retention time, the organic loading rate, the sludge retention time, temperature, pH and alkalinity are evaluated with respect to the two main reactor types: high-rate and low-rate. Finally, the different...

  2. Instrumentation and control strategies for an integral pressurized water reactor

    Directory of Open Access Journals (Sweden)

    Belle R. Upadhyaya

    2015-03-01

    Full Text Available Several vendors have recently been actively pursuing the development of integral pressurized water reactors (iPWRs that range in power levels from small to large reactors. Integral reactors have the features of minimum vessel penetrations, passive heat removal after reactor shutdown, and modular construction that allow fast plant integration and a secure fuel cycle. The features of an integral reactor limit the options for placing control and safety system instruments. The development of instrumentation and control (I&C strategies for a large 1,000 MWe iPWR is described. Reactor system modeling—which includes reactor core dynamics, primary heat exchanger, and the steam flashing drum—is an important part of I&C development and validation, and thereby consolidates the overall implementation for a large iPWR. The results of simulation models, control development, and instrumentation features illustrate the systematic approach that is applicable to integral light water reactors.

  3. Results and recommendations from the reactor chemistry and corrosion tasks of the reactor materials program

    Energy Technology Data Exchange (ETDEWEB)

    Baumann, E.W.; Ondrejcin, R.S.

    1990-11-01

    Within the general context of extended service life, the Reactor Materials Program was initiated in 1984. This comprehensive program addressed material performance in SRS reactor tanks and the primary coolant or Process Water System (PWS) piping. Three of the eleven tasks concerned moderator quality and corrosion mitigation. Definition and control of the stainless steel aqueous environment is a key factor in corrosion mitigation. The Reactor Materials Program systematically investigated the SRS environment and its effect on crack initiation and propagation in stainless steel, with the objective of improving this environment. The purpose of this report is to summarize the contributions of Tasks 6, 7 and 10 of the Reactor Materials Program to the understanding and control of moderator quality and its relationship to mitigation of stress corrosion cracking.

  4. Results and recommendations from the reactor chemistry and corrosion tasks of the reactor materials program

    International Nuclear Information System (INIS)

    Baumann, E.W.; Ondrejcin, R.S.

    1990-11-01

    Within the general context of extended service life, the Reactor Materials Program was initiated in 1984. This comprehensive program addressed material performance in SRS reactor tanks and the primary coolant or Process Water System (PWS) piping. Three of the eleven tasks concerned moderator quality and corrosion mitigation. Definition and control of the stainless steel aqueous environment is a key factor in corrosion mitigation. The Reactor Materials Program systematically investigated the SRS environment and its effect on crack initiation and propagation in stainless steel, with the objective of improving this environment. The purpose of this report is to summarize the contributions of Tasks 6, 7 and 10 of the Reactor Materials Program to the understanding and control of moderator quality and its relationship to mitigation of stress corrosion cracking

  5. Research reactor de-fueling and fuel shipment

    International Nuclear Information System (INIS)

    Ice, R.D.; Jawdeh, E.; Strydom, J.

    1998-01-01

    Planning for the Georgia Institute of Technology Research Reactor operations during the 1996 Summer Olympic Games began in early 1995. Before any details could be outlined, several preliminary administrative decisions had to be agreed upon by state, city, and university officials. The two major administrative decisions involving the reactor were (1) the security level and requirements and (2) the fuel status of the reactor. The Georgia Tech Research Reactor (GTRR) was a heavy-water moderated and cooled reactor, fueled with high-enriched uranium. The reactor was first licensed in 1964 with an engineered lifetime of thirty years. The reactor was intended for use in research applications and as a teaching facility for nuclear engineering students and reactor operators. Approximately one year prior to the olympics, the Georgia Tech administration decided that the GTRR fuel would be removed. In addition, a heightened, beyond regulatory requirements, security system was to be implemented. This report describes the scheduling, operations, and procedures

  6. Prometheus Reactor I and C Software Development Methodology, for Action

    International Nuclear Information System (INIS)

    T. Hamilton

    2005-01-01

    The purpose of this letter is to submit the Reactor Instrumentation and Control (I and C) software life cycle, development methodology, and programming language selections and rationale for project Prometheus to NR for approval. This letter also provides the draft Reactor I and C Software Development Process Manual and Reactor Module Software Development Plan to NR for information

  7. Temperature and Doppler Coefficients of Various Space Nuclear Reactors

    Science.gov (United States)

    Mughabghab, Said F.; Ludewig, Hans; Schmidt, Eldon

    1994-07-01

    Temperature and Doppler feedback effects for a Particle Bed Reactor (PBR) designed to operate as a propulsion reactor are investigated. Several moderator types and compositions fuel enrichments and reactor sizes are considered in this study. From this study it could be concluded that a PBR can be configured which has a negative prompt feedback, zero coolant worth, and a small positive to zero moderator worth. This reactor would put the lowest demands on the control system.

  8. Membrane assisted fluidized bed reactors: Potentials and hurdles

    NARCIS (Netherlands)

    Deshmukh, S.A.R.K.; Heinrich, S.; Mörl, L.; van Sint Annaland, M.; Kuipers, J.A.M.

    2007-01-01

    Recent advances in the development of more stable membranes with increased permeance have significantly enhanced the possibilities for integrating membranes into catalytic reactors in order to achieve a major increase in reactor performance by process integration and process intensification. Several

  9. Gas cooled reactor experience and programs in France

    International Nuclear Information System (INIS)

    Rastoin, J.; Brisbois, J.

    1978-01-01

    After discussing the state of development of natural uranium graphite-gas cooled reactors in France, the current program focused on electricity generating high temperature reactors and the future program based on heat generating applications are presented

  10. Revision of the second basic plans of power reactor development in Power Reactor and Nuclear Fuel Development Corporation

    International Nuclear Information System (INIS)

    1978-01-01

    Revision of the second basic plans concerning power reactor development in PNC (Power Reactor and Nuclear Fuel Development Corporation) is presented. (1) Fast breeder reactors: As for the experimental fast breeder reactor, after reaching the criticality, the power is raised to 50 MW thermal output within fiscal 1978. The prototype fast breeder reactor is intended for the electric output of 200 MW -- 300 MW, using mixed plutonium/uranium oxide fuel. Along the above lines, research and development will be carried out on reactor physics, sodium technology, machinery and parts, nuclear fuel, etc. (2) Advanced thermal reactor: The prototype advanced thermal reactor, with initial fuel primarily of slightly enriched uranium and heavy water moderation and boiling water cooling, of 165 MW electric output, is brought to its normal operation by the end of fiscal 1978. Along the above lines, research and development will be carried out on reactor physics, machinery and parts, nuclear fuel, etc. (Mori, K

  11. Report of scientific results 1976. Section nuclear chemistry and reactor

    International Nuclear Information System (INIS)

    1976-01-01

    The report of the section Nuclear Chemistry and Reactor presents the results of R and D in the fields of neutron scattering, radiation damage in solids, reactor chemistry, trace elements research in biomedicine, geochemistry, reactor operation, radioisotope production, and gives a survey of publications and lectures. (HK) [de

  12. Inherently safe reactors and a second nuclear era.

    Science.gov (United States)

    Weinberg, A M; Spiewak, I

    1984-06-29

    The Swedish PIUS reactor and the German-American small modular high-temperature gas-cooled reactor are inherently safe-that is, their safety relies not upon intervention of humans or of electromechanical devices but on immutable principles of physics and chemistry. A second nuclear era may require commercialization and deployment of such inherently safe reactors, even though existing light-water reactors appear to be as safe as other well-accepted sources of central electricity, particularly hydroelectric dams.

  13. Reactor D and D at Argonne National Laboratory - lessons learned

    International Nuclear Information System (INIS)

    Fellhauer, C. R.

    1998-01-01

    This paper focuses on the lessons learned during the decontamination and decommissioning (D and D) of two reactors at Argonne National Laboratory-East (ANL-E). The Experimental Boiling Water Reactor (EBWR) was a 100 MW(t), 5 MSV(e) proof-of-concept facility. The Janus Reactor was a 200 kW(t) reactor located at the Biological Irradiation Facility and was used to study the effects of neutron radiation on animals

  14. Development of supercritical water reactors in Russia and abroad

    International Nuclear Information System (INIS)

    Glebov, A.P.; Klushin, A.V.

    2014-01-01

    The results of Russian and foreign studies on the water-cooled high critical parameters reactors are analyzed. Developments on this subject are conducted in more than 15 countries. The advantages of WWER- SCP and characteristics of experimental reactor of WWER-SCP-30 are discussed. It is noted that priority task is to develop a reactor with thermal neutron spectrum with a subsequent transition to the reactor with a fast neutron spectrum [ru

  15. Plastic reactor suitable for high pressure and supercritical fluid electrochemistry

    OpenAIRE

    Branch, Jack; Alibouri, Mehrdad; Cook, David A.; Richardson, Peter; Bartlett, Philip N.; Matefi-Tempfli, Maria; Matefi-Tempfli, Stefan; Bampton, Mark; Cookson, Tamsin; Connell, Phil; Smith, David

    2017-01-01

    The paper describes a reactor suitable for high pressure, particularly supercritical fluid, electrochemistry and electrodeposition at pressures up to 30 MPa at 115°C. The reactor incorporates two key, new design concepts; a plastic reactor vessel and the use of o-ring sealed brittle electrodes. These two innovations widen what can be achieved with supercritical fluid electrodeposition. The suitability of the reactor for electroanalytical experiments is demonstrated by studies of the voltammet...

  16. Operation and Utilizations of Dalat Nuclear Research Reactor

    International Nuclear Information System (INIS)

    Hien, P.Z.

    1988-01-01

    The reconstructed Dalat nuclear research reactor was commissioned in March 1984 and up to September 1988 more than 6200 hours of operation at nominal power have been recorded. The major utilizations of the reactor include radioisotope production, activation analysis, nuclear data research and training. A brief review of the utilization of the reactor is presented. Some aspects of reactor safety are also discussed. (author) 2 figs.; 5 refs.; 1 tab

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

    International Nuclear Information System (INIS)

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

    1988-01-01

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

  18. The role of research reactor and its future

    International Nuclear Information System (INIS)

    Nakagome, Yoshihiro

    2005-01-01

    About a half century passed since the start of operation of research reactors. Many research reactors were stopped their operation or decommissioned. With the practical use of nuclear energy, the meaning of research reactor has been buried in oblivion in the developed countries. Furthermore, under the nuclear weapons nonproliferation policy, the use of high enriched uranium fuel in research reactors is obliged to change to the use of low enriched uranium fuel. In such severe situation, this paper refers to the role of the research reactor once more through the operation experience of university-owned research reactor KUR (Kyoto University Reactor, Japan) and describes that research reactor is indispensable for the preparation to the second coming nuclear age. (author)

  19. Reactor coolant pump seal issues and their applicability to new reactor designs

    International Nuclear Information System (INIS)

    Ruger, C.J.; Higgins, J.C.

    1993-01-01

    Reactor Coolant Pumps (RCPs) of various types are used to circulate the primary coolant through the reactor in most reactor designs. RCPs generally contain mechanical seals to limit the leakage of pressurized reactor coolant along the pump drive shaft into the containment. The relatively large number of RCP seal and seal auxiliary system failures experienced at U.S. operating plants during the 1970's and early 1980's raised concerns from the U.S. Nuclear Regulatory Commission (NRC) that gross failures may lead to reactor core uncovery and subsequent core damage. Some seal failure events resulted in a loss of primary coolant to the containment at flow rates greater than the normal makeup capacity of Pressurized Water Reactor (PWR) plants. This is an example of RCP seal failures resulting in a small Loss of Coolant Accident (LOCA). This paper discusses observed and potential causes of RCP seal failure and the recommendations for limiting the likelihood of a seal induced small LOCA. Issues arising out of the research supporting these recommendations and subsequent public comments by the utility industry on them, serve as lessons learned, which are applicable to the design of new reactor plants

  20. Status of Dalat research reactor and progress of new reactor plan in Vietnam

    International Nuclear Information System (INIS)

    Dien, Nguyen Nhi; Vien, Luong Ba

    2005-01-01

    The Dalat Nuclear Research Reactor (DNRR) is a 500-kW pool-type reactor loaded with the Soviet WWR-M2 Fuel Assemblies (FA), moderated and cooled by light water. The reactor was reconstructed from the USA 250-kW TRIGA Mark-II reactor built in early 1960s. The first criticality of the renovated reactor was achieved on 1 st November 1983, and then on 20 March 1984 the reactor was officially inaugurated and its activities restarted. During the last twenty years, the DNRR has played an important role as a large national research facility to implement researches and applications, and its utilization has been broadened in various fields of human life. However, due to the limitation of the neutron flux and power level, the out-of date design of the experimental facilities and the ageing of the reactor facilities, it cannot meet the increasing user's demands even in the existing utilization areas. In addition, the utilization demands of the Research Reactor (RR) will be increased along with the development of the nation's economy growth. In this aspect, it is necessary to have in Vietnam a new high performance multipurpose RR with a sufficient neutron flux and power level. According to the last draft of a national strategy for atomic energy development submitted to the Government for consideration and approval, it is expected that a new high power RR would be put into operation before 2020. The operation and utilization status of the DNRR is presented and some preliminary results of the national research project on new reactor plan for Vietnam are discussed in this paper

  1. Safety philosophy and safety technology of the Soviet RBMK reactors

    International Nuclear Information System (INIS)

    Zuend, H.; Jarvis, A.S.; Haennis, H.P.; Tikal, J.

    1986-01-01

    Safety requirements and control in USSR are outlined. Safety criteria and practical application in the case of the RBMK type reactor Chernobyl-4 are discussed. An overview of the Chernobyl-4 reactor accident including its causes is given. Measures to improve the safety of RBMK reactors are described

  2. Evolution patterns and family relations in G-S reactors

    NARCIS (Netherlands)

    van Swaaij, Willibrordus Petrus Maria; van der Ham, Aloysius G.J.; Kronberg, Alexandre E.

    2002-01-01

    Reactor selection strategies for gas–solid (G–S) heterogeneously catalysed processes can be based on the requirements of the desired process and the properties of the reactions and catalysts involved. Ultimately a reactor selection will nearly always be grounded on existing or emerging reactor types

  3. Scaling of reactor cavity wall loads and stresses

    International Nuclear Information System (INIS)

    Bohachevsky, I.O.

    1977-11-01

    Scalings of reactor cavity wall loads and stresses are determined by deriving an analytic expression in terms of relevant parameters for each loading induced in the reactor cavity walls by fuel pellet microexplosion and by deriving associated expressions relating resulting stresses to shell thicknesses. Also identified are problems that require additional investigations to obtain satisfactory explicit stress estimates for the reactor cavity walls

  4. QUARTERLY PROGRESS REPORT JANUARY, FEBRUARY, MARCH, 1968 REACTOR FUELS AND MATERIALS DEVELOPMENT PROGRAMS FOR FUELS AND MATERIALS BRANCH OF USAEC DIVISION OF REACTOR DEVELOPMENT AND TECHNOLOGY

    Energy Technology Data Exchange (ETDEWEB)

    Cadwell, J. J.; de Halas, D. R.; Nightingale, R. E.; Worlton, D. C.

    1968-06-01

    Progress is reported in these areas: nuclear graphite; fuel development for gas-cooled reactors; HTGR graphite studies; nuclear ceramics; fast-reactor nitrides research; non-destructive testing; metallic fuels; basic swelling studies; ATR gas and water loop operation and maintenance; reactor fuels and materials; fast reactor dosimetry and damage analysis; and irradiation damage to reactor metals.

  5. The digital reactor protection system for the instrumentation and control of reactor TRIGA PUSPATI (RTP)

    International Nuclear Information System (INIS)

    Nurfarhana Ayuni Joha; Izhar Abu Hussin; Mohd Idris Taib; Zareen Khan Abdul Jalil Khan

    2010-01-01

    Reactor Protection System (RPS) is important for Reactor Instrumentation and Control System. The RPS comprises all redundant electrical devices and circuitry involved in the generation of those initiating signals associated to the trip protective function. The instrumentation system for the RPS provides automatic protection signals against unsafe and improper reactor operation. The physical separation is provided for all of the redundant instrumentation systems to preserve redundancy. The safety protection systems using circuits composed of analog instruments and relays with relay contacts is difficult to realize from various reasons. Therefore, an application of digital technology can be said a logical conclusion also in the light of its functional superiority. (author)

  6. JANUS reactor d and d project

    International Nuclear Information System (INIS)

    Fellhauer, C. R.

    1998-01-01

    Argonne National Laboratory (ANL-E) has recently completed the decontamination and decommissioning (D and D) of the JANUS Reactor Facility located in Building 202. The 200 KW reactor operated from August 1963 to March 1992. The facility was used to study the effects of both high and low doses of fission neutrons in animals. There were two exposure rooms on opposite sides of the reactor and the reactor was therefore named after the two-faced Roman god. The High Dose Room was capable of specimen exposure at a dose rate of 3,600 rads per hour. During calendar year 1996 a detailed characterization of the facility was performed by ANL-E Health Physics personnel. ANL-E Analytical Services performed the required sample analysis. An Auditable Safety Analysis and an Environmental Assessment were completed. D and D plans, procedures and procurement documents were prepared and approved. A D and D subcontractor was selected and a firm, fixed price contract awarded for the field work and final survey effort. The D and D subcontractor was mobilized to ANL-E in January 1997. Electrical isolation of all reactor equipment and control panels was accomplished and the equipment removed. A total of 207,230 pounds (94,082 Kg) of lead shielding was removed, surveyed and sampled, and free-released for recycle. All primary and secondary piping was removed, size reduced and packaged for disposal or recycled as appropriate. The reactor vessel was removed, sized reduced and packaged as radioactive waste in April. The activated graphite block reflector was removed next, followed by the bioshield concrete and steel. All of this material was packaged as low level waste. Total low level radioactive waste generation was 4002.1 cubic feet (113.3 cubic meters). Mixed waste generation was 538 cubic feet (15.2 cubic meters). The Final Release Survey was completed in September. The project field work was completed in 38 weeks without any lost-time accidents, personnel contaminations or unplanned

  7. Safety in the utilization and modification of research reactors

    International Nuclear Information System (INIS)

    1994-01-01

    This Safety Guide presents guidelines, approved by international consensus, for the safe utilization and modification of research reactors. While the Guide is most applicable to existing reactors, it is also recommended for use by organizations planning to put a new reactor into operation. 1 fig

  8. Development and application of reactor noise diagnostics

    Energy Technology Data Exchange (ETDEWEB)

    Karlsson, Joakim K.H

    1999-04-01

    A number of problems in reactor noise diagnostics have been investigated within the framework of the present thesis. The six papers presented cover three relatively different areas, namely the use of analytical calculations of the neutron noise in simple reactor models, some aspects of boiling water reactor (BWR) stability and diagnostics of core barrel motion in pressurized water reactors (PWRs). The noise induced by small vibrations of a strong absorber has been the subject of several previous investigations. For a conventional {delta}-function source model, the equations can not be linearized in the traditional manner. Thus, a new source model, which is called the {epsilon}/d model, was developed. The correct solution has been derived in the {epsilon}/d model for both 1-D and 2-D reactor models. Recently, several reactor diagnostic problems have occurred which include a control rod partially inserted into the reactor core. In order to study such problems, we have developed an analytically solvable, axially non-homogenous, 2-D reactor model. This model has also been used to study the noise induced by a rod maneuvering experiment. Comparisons of the noise with the results of different reactor kinetic approximations have yielded information on the validity of the approximations in this relatively realistic model. In case of an instability event in a BWR, the noise may consist of one or several co-existing modes of oscillation and besides the fundamental mode, a regional first azimuthal mode has been observed in e.g. the Swedish BWR Ringhals-1. In order to determine the different stability characteristics of the different modes separately, it is important to be able to decompose the noise into its mode constituents. A separation method based on factorisation of the flux has been attempted previously, but without success. The reason for the failure of the factorisation method is the presence of the local component of the noise and its axial correlation properties. In

  9. Development and application of reactor noise diagnostics

    International Nuclear Information System (INIS)

    Karlsson, Joakim K.H.

    1999-04-01

    A number of problems in reactor noise diagnostics have been investigated within the framework of the present thesis. The six papers presented cover three relatively different areas, namely the use of analytical calculations of the neutron noise in simple reactor models, some aspects of boiling water reactor (BWR) stability and diagnostics of core barrel motion in pressurized water reactors (PWRs). The noise induced by small vibrations of a strong absorber has been the subject of several previous investigations. For a conventional δ-function source model, the equations can not be linearized in the traditional manner. Thus, a new source model, which is called the ε/d model, was developed. The correct solution has been derived in the ε/d model for both 1-D and 2-D reactor models. Recently, several reactor diagnostic problems have occurred which include a control rod partially inserted into the reactor core. In order to study such problems, we have developed an analytically solvable, axially non-homogenous, 2-D reactor model. This model has also been used to study the noise induced by a rod maneuvering experiment. Comparisons of the noise with the results of different reactor kinetic approximations have yielded information on the validity of the approximations in this relatively realistic model. In case of an instability event in a BWR, the noise may consist of one or several co-existing modes of oscillation and besides the fundamental mode, a regional first azimuthal mode has been observed in e.g. the Swedish BWR Ringhals-1. In order to determine the different stability characteristics of the different modes separately, it is important to be able to decompose the noise into its mode constituents. A separation method based on factorisation of the flux has been attempted previously, but without success. The reason for the failure of the factorisation method is the presence of the local component of the noise and its axial correlation properties. In the paper

  10. Reactor structure and superconducting magnet system of ITER

    International Nuclear Information System (INIS)

    Tada, Eisuke; Yoshida, Kiyoshi; Shibanuma, Kiyoshi; Okuno, Kiyoshi; Tsuji, Hiroshi; Shimamoto, Susumu

    1993-01-01

    Fusion Experimental Reactors are one of the major steps toward realization of the fusion energy and the key objective are to demonstrate the scientific and technological feasibility prior to the Demo Fusion Reactor. ITER (International Thermonuclear Experimental Reactor) is one of experimental reactors and the conceptual design has been completed by the united efforts of USA, USSR, EC and Japan. In parallel with the conceptual design, key technology development in various areas has being conducted. This paper describes the overall design concepts and the latest technological achievements of the ITER reactor structure and superconducting magnet system. (author)

  11. Reactor and turbine building layout of the high performance light water reactor

    International Nuclear Information System (INIS)

    Bittermann, D.

    2010-01-01

    Based on the information generated within the European funded project ''High Per-formance Light Water Reactor Phase 2'', a general plant layout has been developed. The central building is the reactor building, in which the containment and safety sys-tems are located. The reactor building is with app. 90.000 m 3 considerably smaller compared to other BWR buildings, thus providing a huge potential for cost savings. The turbine building with app 250,000 m 3 is of approximately the same size like for existing BWRs. (orig.)

  12. Novelties in design and construction of the advanced reactors

    International Nuclear Information System (INIS)

    Acosta Ezcurra, T.; Garcia Rodriguez, B.M.

    1996-01-01

    The advanced pressurized water reactors (APWR), advanced boiling water reactors (ABWR), advanced liquid metal reactors (ALMR), and modular high temperature gas-cooled reactors (MHTGR), as well as heavy water reactors (AHWR), are analyzed taking into account those characteristics which make them less complex, but safer than their current homologous ones. This fact simplifies their construction which reduces completion periods and costs, increasing safety and protection of the plants. It is demonstrated how the accumulated operational experience allows to find more standardized designs with some enhancement in the material and component technology and thus achieve also a better use of computerized systems

  13. Research on nuclear energy in the fields of fuel cycle, PWR reactors and LMFBR reactors

    International Nuclear Information System (INIS)

    Barre, B.; Camarcat, N.

    1995-01-01

    In this article we present the CEA research programs to improve the safety of the next generation of reactors, to manage the Plutonium and the wastes of the fuel cycle end and to ameliorate the competitiveness. 6 refs

  14. The Global Outlook for Small Reactors: Opportunities, Challenges and Implementation

    International Nuclear Information System (INIS)

    Hughes, A.

    2012-01-01

    The fascinating topic of small nuclear is becoming more prevalent on the nuclear agenda. The discussions are generally focused within the country of technical origin. In this presentation 'The global outlook for small reactors' Rolls-Royce along with energy business analysts Douglas-Westwood present their shared views on the global opportunities for Small Reactor deployment in the context of the wider energy market. The presentation will: provide a compressive overview of trends and dynamics relating to Small Reactors in the context of the current world energy market, identify specific Small Reactor opportunities and areas of interest, address the challenges and potential solutions for Small Reactor deployment and operation.(author).

  15. Pakistan research reactor and its utilization

    International Nuclear Information System (INIS)

    Iqbal Hussain Qureshi; Naeem Ahmad Khan.

    1983-01-01

    The 5 MW enriched uranium fuelled, light water moderated and cooled Pakistan Research reactor became critical on 21st December, 1965 and was taken to full power on 22nd June, 1966. Since then is has been operated for about 23000 hours till 30th June, 1983 without any major break down. It has been used for the studies of neutron cross-sections, nuclear structure, fission physics, structure of material, radiation damage in crystals and semiconductors, studies of geological, biological and environmental samples by neutron activation techniques, radioisotope production, neutron radiography and for training of scientists, engineers and technicians. In the paper we have described briefly the facility of Pakistan Research Reactor and the major work carried around it during the last decade. (author)

  16. Status of Fast Reactor Research and Technology Development

    International Nuclear Information System (INIS)

    2012-01-01

    In 1985, the International Atomic Energy Agency (IAEA) published a report titled 'Status of Liquid Metal Cooled Fast Breeder Reactors' (Technical Reports Series No. 246). The report was a general review of the status of fast reactor development at that time, covering some aspects of design and operation and reviewing experience from the earliest days. It summarized the programmes and plans in all countries which were pursuing the development of fast reactors. In 1999, the IAEA published a follow-up report titled 'Status of Liquid Metal Cooled Fast Reactor Technology' (IAEA-TECDOC-1083), necessitated by the substantial advances in fast reactor technology development and changes in the economic and regulatory environment which took place during the period of 1985-1998. Chief among these were the demonstration of reliable operation by several prototypes and experimental reactors, the reliable operation of fuel at a high burnup and the launch of new fast reactor programmes by some additional Member States. In 2006, the Technical Working Group on Fast Reactors (TWG-FR) identified the need to update its past publications and recommended the preparation of a new status report on fast reactor technology. The present status report intends to provide comprehensive and detailed information on the technology of fast neutron reactors. The focus is on practical issues that are useful to engineers, scientists, managers, university students and professors, on the following topics: experience in construction, operation and decommissioning; various areas of research and development; engineering; safety; and national strategies and public acceptance of fast reactors.

  17. Status of Fast Reactor Research and Technology Development

    International Nuclear Information System (INIS)

    2013-01-01

    In 1985, the International Atomic Energy Agency (IAEA) published a report titled 'Status of Liquid Metal Cooled Fast Breeder Reactors' (Technical Reports Series No. 246). The report was a general review of the status of fast reactor development at that time, covering some aspects of design and operation and reviewing experience from the earliest days. It summarized the programmes and plans in all countries which were pursuing the development of fast reactors. In 1999, the IAEA published a follow-up report titled 'Status of Liquid Metal Cooled Fast Reactor Technology' (IAEA-TECDOC-1083), necessitated by the substantial advances in fast reactor technology development and changes in the economic and regulatory environment which took place during the period of 1985-1998. Chief among these were the demonstration of reliable operation by several prototypes and experimental reactors, the reliable operation of fuel at a high burnup and the launch of new fast reactor programmes by some additional Member States. In 2006, the Technical Working Group on Fast Reactors (TWG-FR) identified the need to update its past publications and recommended the preparation of a new status report on fast reactor technology. The present status report intends to provide comprehensive and detailed information on the technology of fast neutron reactors. The focus is on practical issues that are useful to engineers, scientists, managers, university students and professors, on the following topics: experience in construction, operation and decommissioning; various areas of research and development; engineering; safety; and national strategies and public acceptance of fast reactors.

  18. Aging considerations for PWR [pressurized water reactor] control rod drive mechanisms and reactor internals

    International Nuclear Information System (INIS)

    Ware, A.G.

    1988-01-01

    This paper describes age-related degradation mechanisms affecting life extension of pressurized water reactor control rod drive mechanisms and reactor internals. The major sources of age-related degradation for control rod drive mechanisms are thermal transients such as plant heatups and cooldowns, latchings and unlatchings, long-term aging effects on electrical insulation, and the high temperature corrosive environment. Flow induced loads, the high-temperature corrosive environment, radiation exposure, and high tensile stresses in bolts all contribute to aging related degradation of reactor internals. Another problem has been wear and fretting of instrument guide tubes. The paper also discusses age-related failures that have occurred to date in pressurized water reactors

  19. Trends and developments in magnetic confinement fusion reactor concepts

    International Nuclear Information System (INIS)

    Baker, C.C.; Carlson, G.A.; Krakowski, R.A.

    1981-01-01

    An overview is presented of recent design trends and developments in reactor concepts for magnetic confinement fusion. The paper emphasizes the engineering and technology considerations of commercial fusion reactor concepts. Emphasis is placed on reactors that operate on the deuterium/tritium/lithium fuel cycle. Recent developments in tokamak, mirror, and Elmo Bumpy Torus reactor concepts are described, as well as a survey of recent developments on a wide variety of alternate magnetic fusion reactor concepts. The paper emphasizes recent developments of these concepts within the last two to three years

  20. Space reactors - past, present, and future

    International Nuclear Information System (INIS)

    Buden, D.; Angelo, J.A.

    1983-01-01

    The successful test flights of the Space Shuttle mark the start of a new era--an era of routine manned access into cislunar space. Human technical development at the start of the next Millenium will be highlighted by the creation of Man's extraterrestrial civilization with off-planet expansion of the human resource base. In the 1990s and beyond, advanced-design nuclear reactors could represent the prime source of both space power and propulsion. Many sophisticated military and civilian space missions of the future will require first kilowatt and then megawatt levels of power. This paper reviews key technology developments that accompanied past US space nuclear power development efforts, describes on-going programs, and then explores reactor technologies that will satisfy megawatt power level needs and beyond

  1. The present status and the prospect of China research reactors

    International Nuclear Information System (INIS)

    Yongmao, Z.; Yizheng, C.

    1990-01-01

    A total of 100 reactor operation years' experience of research reactors has now been obtained in China. The type and principal parameters of China research reactors and their operating status are briefly introduced in this paper. Chinese research reactors have been playing an important role in nuclear power and nuclear weapon development, industrial and agricultural production, medicine, basic and applied science research and environmental protection, etc. The utilization scale, benefits and achievements will be given. There is a good safety record in the operation of these reactors. A general safety review is discussed. The important incidents and accidents happening during a hundred reactor operating years are described and analyzed. China has the capability of developing any type of research reactor. The prospective projects are briefly introduced

  2. Development of Reactor Protection System (RPS) in Reactor Digital Instrumentation and Control System (ReDICS)

    International Nuclear Information System (INIS)

    Mohd Khairulezwan Abdul Manan; Mohd Sabri Minhat; Ridzuan Abdul Mutalib

    2013-01-01

    RTP Research Reactor are in the process upgraded from analogue control console system to a digital control console system . Upgrade process requires a statistical study to improve safety during reactor operation. RPS was developed to meet the needs of operational safety and at the same time comply with the guidelines set by the IAEA. RPS is in analog and hardware with industry standard interfaced with digital DAC (Data Acquisition and Control) and OWS (Operator Work Station). (author)

  3. Neutronics and mass transport in a chemical reactor associated with controlled thermonuclear fusion reactor

    International Nuclear Information System (INIS)

    Dang, V.D.; Steinberg, M.; Lazareth, O.W.; Powell, J.R.

    1976-05-01

    The formation of ozone from oxygen and the dissociation carbon dioxide to carbon monoxide and oxygen is studied in a gamma-neutron chemical process blanket associated with a controlled thermonuclear reactor. Materials used for reactor tube wall will affect the efficiency of the energy absorption by the reactants and consequently the yield of reaction products. Three kinds of materials, aluminum, stainless steel and fiber (Al 2 O 3 )-aluminium are investigated for the tube wall material in the study

  4. Reactor coolant system and containment aqueous chemistry

    International Nuclear Information System (INIS)

    Torgerson, D.F.

    1986-01-01

    Fission products released from fuel during reactor accidents can be subject to a variety of environments that will affect their ultimate behavior. In the reactor coolant system (RCS), for example, neutral or reducing steam conditions, radiation, and surfaces could all have an effect on fission product retention and chemistry. Furthermore, if water is encountered in the RCS, the high temperature aqueous chemistry of fission products must be assessed to determine the quantity and chemical form of fission products released to the containment building. In the containment building, aqueous chemistry will determine the longer-term release of volatile fission products to the containment atmosphere. Over the past few years, the principles of physical chemistry have been rigorously applied to the various chemical conditions described above. This paper reviews the current state of knowledge and discusses the future directions of chemistry research relating to the behavior of fission products in the RCS and containment

  5. Optical inspections of research reactor tanks and tank components

    International Nuclear Information System (INIS)

    Boeck, H.; Hammer, J.

    1988-01-01

    By the end of 1987 worldwide there were 326 research reactors in operation, 276 of them operating more than 10 years, and 195 of them operating more than 20 years. The majority of these reactors are swimming-pool type or tank type reactors using aluminium as structural material. Although aluminium has prooven its excellent properties for reactor application in primary system, it is however subjected to various types of corrosion if it gets into contact with other materials such as mild steel in the presence of destilled water. This paper describes various methods of research reactor tank inspections, maintenance and repair possibilities. 9 figs. (Author)

  6. Computational mathematics and physics of fusion reactors.

    Science.gov (United States)

    Garabedian, Paul R

    2003-11-25

    Theory has contributed significantly to recent advances in magnetic fusion research. New configurations have been found for a stellarator experiment by computational methods. Solutions of a free-boundary problem are applied to study the performance of the plasma and look for islands in the magnetic surfaces. Mathematical analysis and numerical calculations have been used to study equilibrium, stability, and transport of optimized fusion reactors.

  7. Reactor building assembly and method of operation

    International Nuclear Information System (INIS)

    Fennern, L.E.; Caraway, H.A.; Hsu, Li C.

    1993-01-01

    A reactor building assembly is described comprising: a reactor pressure vessel containing a reactor core for generating heat in the form of steam; a containment vessel enclosing said pressure vessel; a first enclosure surrounding said containment vessel and spaced laterally therefrom to define a first chamber there between, and having a top and a bottom; a second enclosure surrounding said first enclosure and spaced laterally therefrom to define a second chamber there between, and having a top and a bottom; a building inlet for receiving into said second chamber fresh air from outside said second enclosure; a building outlet for discharging stale air from said first chamber; a transfer duct disposed through said first enclosure selectively joining in flow communication said first and second chambers; said building inlet being disposed at said second enclosure top, said building outlet being disposed at said first enclosure top, and said transfer duct being disposed adjacent said first enclosure bottom for allowing said fresh air to flow downwardly by gravity through said second chamber and through said transfer duct into said first chamber for cooling said first chamber, said stale air flowing upwardly by natural buoyancy for discharger from said first chamber through said building outlet; an exhaust stack disposed above said building outlet and in flow communication therewith for channeling upwardly said stale air from said first chamber for discharge into the surrounding environs; and a passive first driving means for increasing flow of said stale air from said building outlet comprising: an isolation pool containing isolation water; an isolation condenser disposed in said isolation pool, and joined in flow communication with said reactor pressure vessel for receiving primary steam therefrom, said primary steam being cooled in said isolation condenser for heating said isolation water to generate secondary steam

  8. Construction of Research Reactors for Gen 3 and Gen 4 Reactors Development

    International Nuclear Information System (INIS)

    Behar, Christophe

    2014-01-01

    Christophe Behar, Director of the Nuclear Energy Division at CEA, detailed the different kind of research reactors and the issues in term of investment, use, side application such as the medical isotopes production

  9. Nuclear reactor insulation and preheat system

    International Nuclear Information System (INIS)

    Wampole, N.C.

    1978-01-01

    An insulation and preheat system is disclosed for preselected components of a fluid cooled nuclear reactor. A gas tight barrier or compartment of thermal insulation surrounds the selected components and includes devices to heat the internal atmosphere of the ocmpartment. An external surface of the compartment of enclosure is cooled, such as by a circulating fluid. The heating devices provide for preheating of the components, as well as maintenance of a temperature sufficient to ensure that the reactor coolant fluid will not solidify during shutdown. The external cooling limits the heat transferred to other plant structures, such as supporting concrete and steel. The barrier is spaced far enough from the surrounded components so as to allow access for remote or manual inspection, maintenance, and repair

  10. Safe operation and maintenance of research reactor

    Energy Technology Data Exchange (ETDEWEB)

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

    1999-10-01

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

  11. The experimental reactor Osiris and the nuclear fuel technology for the P.W.R. reactors

    International Nuclear Information System (INIS)

    Lestiboudois, G.; Contenson, G. de; Genthon, J.P.; Molvault, M.; Roche, M.

    1977-01-01

    The possibility of employing research reactors to study and to improve the nuclear fuel of the power reactors is presented. Measurements of temperature, pressure, stresses, thermal balance, gamma spectrometry and neutron radiography, allow the study of fuel densification, the influence of the initial filling pressure on the fission gas release and the gadolinium efficiency evolution. The solutions of the problems of failed element detection, power increase, remote handling, are presented [fr

  12. LBB application in the US operating and advanced reactors

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-04-01

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

  13. Present and possible utilization of PUSPATI reactor

    International Nuclear Information System (INIS)

    Gui Ah Auu.

    1983-01-01

    The utilization of PUSPATI TRIGA Mark II Reactor (PTR) has increased reasonably well since its commissioning last year. PTR was used mainly for training of operators, neutron flux measurements and neutron activation analysis. However, the present utilization data indicates that further increase in PTR utilization to include teaching and the usage of the beam ports is desirable. Some possible areas of PTR applications in the future in relevance to our needs are also described in this paper. (author)

  14. Light-water reactor research and development

    International Nuclear Information System (INIS)

    1985-05-01

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

  15. On disruption of reactor core of the Chernobylsk-4 reactor (retrospective analysis of experiments and facts)

    International Nuclear Information System (INIS)

    Platonov, P.A.

    2007-01-01

    Fragments of graphite blocks from the damaged Chernobyl NPP, unit 4 are studied, the results are analyzed. The temperature of the graphite blocks at the moment of accident release from the reactor is evaluated. Results of studying the fragments of fuel channel and fuel dispersion are considered. The fuel heat content at the moment of the explosion is evaluated and some conclusions are made about the character of the reactor core destruction [ru

  16. Operation and maintenance of 1MW PUSPATI TRIGA reactor

    International Nuclear Information System (INIS)

    Adnan Bokhari; Mohammad Suhaimi Kassim

    2006-01-01

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

  17. MELCOR development for existing and advanced reactors

    International Nuclear Information System (INIS)

    Summers, R.M.

    1993-01-01

    Recent efforts in MELCOR development to address previously identified deficiencies have resulted in release of MELCOR 1.8.2, a much-improved version of the code. Major new models have been implemented for direct containment heating, ice condensers, debris quenching, lower plenum debris behavior, core materials interactions' and radial relocation of debris. Significant improvements have also been made in the modeling of interfacial momentum exchange and in the modeling of fission product release, condensation/evaporation, and aerosol behavior. Efforts are underway to address two-phase hydrodynamics difficulties, to improve modeling of water condensation on structures and fine-scale natural circulation within the reactor vessel, and to implement CORCON-Mod3. Improvements are also being made to MELCOR's capability to handle new features of the advanced light water reactor designs, including drainage of water films on connected heat structures, heat transfer from the external surface of the reactor vessel to a flooded cavity, and creep rupture failure of the lower head. Additional development needs in other areas are discussed

  18. Radiation Effects in Fission and Fusion Reactors

    Science.gov (United States)

    Odette, G. Robert; Wirth, Brian D.

    Since the prediction of "Wigner disease" [1] and the subsequent observation of anisotropic growth of the graphite used in the Chicago Pile, the effects of radiation on materials has been an important technological concern. The broad field of radiation effects impacts many critical advanced technologies, ranging from semiconductor processing to severe materials degradation in nuclear reactor environments. Radiation effects also occur in many natural environments, ranging from deep space to inside the Earth's crust. As selected examples that involve many basic phenomena that cross-cut and illustrate the broader impacts of radiation exposure on materials, this article focuses on modeling microstructural changes in iron-based ferritic alloys under high-energy neutron irradiation relevant to light water fission reactor pressure vessels. We also touch briefly on radiation effects in structural alloys for fusion reactor first wall and blanket structures; in this case the focus is on modeling the evolution of self-interstitial atom clusters and dislocation loops. Note, since even the narrower topic of structural materials for nuclear energy applications encompass a vast literature dating from 1942, the references included in this article are primarily limited to these two narrower subjects. Thus, the references cited here are presented as examples, rather than comprehensive bibliographies. However, the interested reader is referred to proceedings of continuing symposia series that have been sponsored by several organizations, several monographs [2-4] and key journals (e.g., Journal of Nuclear Materials, Radiation Effects and Defects in Solids).

  19. Communication and computer technologies for teaching physics in nuclear reactors

    International Nuclear Information System (INIS)

    Murua, C; Chautemps, A; Odetto, J; Keil, W; Trivino, S; Rossi, F; Perez Lucero, A

    2012-01-01

    In order to train personnel inn order to train personnel in Embalse Nuclear Power Plant, and provided that such training given primarily on the location of such a facility, we designed a pedagogical strategy that combined the use of conventional resources with new information technologies. Since the Nuclear Reactor RA-0 is an ideal tool for teaching Reactor Physics, priority was the use of it, both locally remotely. The teaching strategy is based on four pillar: -Lectures on the Power Plant (using a virtual classroom to support); -Remote monitoring of Ra-0 Nuclear Reactor parameters while operating (RA0REMOTO); -Use, through the Internet, of the Ra-0 Nuclear Reactor Simulator (RA0SIMUL); -Made in the Nuclear Reactor RA-0 of Reactor Physics practical. The work emphasizes RA0REMOTO and RA0SIMUL systems. The RA0REMOTO system is an appendix of the Electronic Data Acquisition System (SEAD) of the Nuclear Reactor RA-0. This system acquires signals from Reactor instrumentation and sends them to a server running the software that 'publish' the reactor parameters on the internet. Students may, during the lectures, monitor any parameter of the reactor while it operates, which allows teachers to compare theory with reality. RA0SIMUL is a simulator on the RA-0, which allows students to 'operate' a reactor analyzing the underlying physics concepts (author)

  20. Joining structure between diaphragms for reactor container and foundations for reactor

    International Nuclear Information System (INIS)

    Furukawa, Hideyasu; Maki, Yasufumi; Honma, Keiji.

    1990-01-01

    In the conventional joining structure of a diaphragm floor for a reactor container, it is necessary to join lower end steel reinforcements in the radial direction to thread couplers welded to a rim plate of foundations for the reactor main body in view of the application of steel reinforcements. However, since there are many perforations for pipelines in the vicinity thereof, this method involves a drawback. According to the method of the present invention, the rim plate recessed at the top end of the foundations for the reactor main body are removed and steel reinforcements are fixed by means of a fixing plate in the diaphragm floor, and loads caused in the joined portions are transmitted by studs disposed on a horizontal plate recessed at the top of the foundations for the reactor main body. Fixed loads of equipments and pipelines, as well as downward pressing loads upon accidents are directly transmitted as a surface pressure to the foundations for the reactor main body, while radial and tangential loads upon accidents or earthquakes are transmitted by way of the studs from the diaphragm floor to the foundations for the reactor main body and, accordingly, the amount of contraction materials can be reduced and the number of steps such as of operation can be saved. (N.H.)

  1. Request for Naval Reactors Comment on Proposed PROMETHEUS Space Flight Nuclear Reactor High Tier Reactor Safety Requirements and for Naval Reactors Approval to Transmit These Requirements to Jet Propulsion Laboratory

    International Nuclear Information System (INIS)

    D. Kokkinos

    2005-01-01

    The purpose of this letter is to request Naval Reactors comments on the nuclear reactor high tier requirements for the PROMETHEUS space flight reactor design, pre-launch operations, launch, ascent, operation, and disposal, and to request Naval Reactors approval to transmit these requirements to Jet Propulsion Laboratory to ensure consistency between the reactor safety requirements and the spacecraft safety requirements. The proposed PROMETHEUS nuclear reactor high tier safety requirements are consistent with the long standing safety culture of the Naval Reactors Program and its commitment to protecting the health and safety of the public and the environment. In addition, the philosophy on which these requirements are based is consistent with the Nuclear Safety Policy Working Group recommendations on space nuclear propulsion safety (Reference 1), DOE Nuclear Safety Criteria and Specifications for Space Nuclear Reactors (Reference 2), the Nuclear Space Power Safety and Facility Guidelines Study of the Applied Physics Laboratory

  2. Segmentation and packaging reactor vessels internals

    International Nuclear Information System (INIS)

    Boucau, Joseph

    2014-01-01

    Document available in abstract form only, full text follows: With more than 25 years of experience in the development of reactor vessel internals and reactor vessel segmentation and packaging technology, Westinghouse has accumulated significant know-how in the reactor dismantling market. The primary challenges of a segmentation and packaging project are to separate the highly activated materials from the less-activated materials and package them into appropriate containers for disposal. Since disposal cost is a key factor, it is important to plan and optimize waste segmentation and packaging. The choice of the optimum cutting technology is also important for a successful project implementation and depends on some specific constraints. Detailed 3-D modeling is the basis for tooling design and provides invaluable support in determining the optimum strategy for component cutting and disposal in waste containers, taking account of the radiological and packaging constraints. The usual method is to start at the end of the process, by evaluating handling of the containers, the waste disposal requirements, what type and size of containers are available for the different disposal options, and working backwards to select a cutting method and finally the cut geometry required. The 3-D models can include intelligent data such as weight, center of gravity, curie content, etc, for each segmented piece, which is very useful when comparing various cutting, handling and packaging options. The detailed 3-D analyses and thorough characterization assessment can draw the attention to material potentially subject to clearance, either directly or after certain period of decay, to allow recycling and further disposal cost reduction. Westinghouse has developed a variety of special cutting and handling tools, support fixtures, service bridges, water filtration systems, video-monitoring systems and customized rigging, all of which are required for a successful reactor vessel internals

  3. The TRIGA reactor Frankfurt FRF 2, construction and present status

    International Nuclear Information System (INIS)

    Rossberg, D.; Wolf, G.; Bass, F.

    1980-01-01

    The reactor FR2 was planned to replace the old FR1 reactor at the Frankfurt University, which was shutdown due to a failure. The FR2 is a tank reactor; the core is divided into 2 parts. The 1 MW operational core was calculated to consist of about 70 fuel elements. The grid plates contain 110 fuel element positions, surplus positions are occupied by dummy elements filled with water. The core has 5 control rods, 4 of which have fuel element followers. The reactor plant including all protection facilities was completed in 1977. The fuel elements were delivered in 1978. The reactor has been ready for critical experiments, but the permission for the start-up was not granted. In 1980 was decided that the reactor was not going in operation. Preparations are being made to dismantle the reactor components and to re-use them elsewhere

  4. Development Directions For CANDU and Slowpoke Reactors

    International Nuclear Information System (INIS)

    Brooks, Gordon L.

    1990-01-01

    This paper provides a broader-based discussion of overall development directions foreseen for CANDU reactors, particularly those which have further evolved sine the earlier paper. The paper then discusses development directions for the Slowpokes Energy System which is a small nuclear heat source intended to meet local heating needs for building complexes and municipal heating systems. In evolving a sound development direction, it is necessary, firstly, to address the question of requirements, viz., what are the requirements which future nuclear power plants must satisfy if they are to be successful? Today, some in the nuclear industry believe that the most important of such requirements relates to the need for 'safer' reactors. Indeed, some proponents of this view would seem to suggest that if only we could develop such 'safer' reactors, suddenly all of our problem s with public acceptance would disappear and utilities would form long lines waiting to purchase such marvellous machines. I do not share such a simplistic view nor, indeed, do many of my colleagues in the international nuclear power industry

  5. Research reactors - an overview

    International Nuclear Information System (INIS)

    West, C.D.

    1997-01-01

    A broad overview of different types of research and type reactors is provided in this paper. Reactor designs and operating conditions are briefly described for four reactors. The reactor types described include swimming pool reactors, the High Flux Isotope Reactor, the Mark I TRIGA reactor, and the Advanced Neutron Source reactor. Emphasis in the descriptions is placed on safety-related features of the reactors. 7 refs., 7 figs., 2 tabs

  6. Reactor Physics

    Energy Technology Data Exchange (ETDEWEB)

    Ait Abderrahim, A

    2002-04-01

    SCK-CEN's Reactor Physics and MYRRHA Department offers expertise in various areas of reactor physics, in particular in neutron and gamma calculations, reactor dosimetry, reactor operation and control, reactor code benchmarking and reactor safety calculations. This expertise is applied in the Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 materials testing reactor dosimetry, and the preparation and interpretation of irradiation experiments by means of neutron and gamma calculations. The activities of the Fuzzy Logic and Intelligent Technologies in Nuclear Science programme cover several domains outside the department. Progress and achievements in these topical areas in 2001 are summarised.

  7. Reactor Physics

    Energy Technology Data Exchange (ETDEWEB)

    Ait Abderrahim, A

    2001-04-01

    The Reactor Physics and MYRRHA Department of SCK-CEN offers expertise in various areas of reactor physics, in particular in neutronics calculations, reactor dosimetry, reactor operation, reactor safety and control and non-destructive analysis of reactor fuel. This expertise is applied in the Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 reactor dosimetry, and the preparation and interpretation of irradiation experiments by means of neutron and gamma calculations. The activities of the Fuzzy Logic and Intelligent Technologies in Nuclear Science programme cover several domains outside the department. Progress and achievements in these topical areas in 2000 are summarised.

  8. Reactor Physics

    International Nuclear Information System (INIS)

    Ait Abderrahim, A.

    2001-01-01

    The Reactor Physics and MYRRHA Department of SCK-CEN offers expertise in various areas of reactor physics, in particular in neutronics calculations, reactor dosimetry, reactor operation, reactor safety and control and non-destructive analysis of reactor fuel. This expertise is applied in the Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 reactor dosimetry, and the preparation and interpretation of irradiation experiments by means of neutron and gamma calculations. The activities of the Fuzzy Logic and Intelligent Technologies in Nuclear Science programme cover several domains outside the department. Progress and achievements in these topical areas in 2000 are summarised

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

    International Nuclear Information System (INIS)

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

    2002-01-01

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

  10. Noise and vibration analysis system

    International Nuclear Information System (INIS)

    Johnsen, J.R.; Williams, R.L.

    1985-01-01

    The analysis of noise and vibration data from an operating nuclear plant can provide valuable information that can identify and characterize abnormal conditions. Existing plant monitoring equipment, such as loose parts monitoring systems (LPMS) and neutron flux detectors, may be capable of gathering noise data, but may lack the analytical capability to extract useful meanings hidden in the noise. By analyzing neutron noise signals, the structural motion and integrity of core components can be assessed. Computer analysis makes trending of frequency spectra within a fuel cycle and from one cycle to another a practical means of core internals monitoring. The Babcock and Wilcox Noise and Vibration Analysis System (NVAS) is a powerful, compact system that can automatically perform complex data analysis. The system can acquire, process, and store data, then produce report-quality plots of the important parameter. Software to perform neutron noise analysis and loose parts analysis operates on the same hardware package. Since the system is compact, inexpensive, and easy to operate, it allows utilities to perform more frequency analyses without incurring high costs and provides immediate results

  11. Research reactor design and utilization: the Korean experience

    International Nuclear Information System (INIS)

    JUN BYUNG-JIN; PARK CHEOL; KIM HARK-RHO

    2008-01-01

    HANARO is the first high power research reactor in Korea designed by Korean technology with specific requirements for its utilization. Since the construction and operation of multiple research reactors would have been almost impossible when considering the circumstances at the time of its initiation, we designed the HANARO for as many purposes as possible to satisfy future national demands for the 21st century. Installation of a majority of the experimental instruments and facilities, to fully utilize the reactor, followed its commissioning and initial operation. Some of them are still being developed and installed, and the reactor is capable of accommodating more instruments in the future. Under a limited national condition, one should proceed in a step-by-step manner but pursue a high performance of a reactor and its instruments. Existence of a high performance reactor is the starting point to attract users, and thus establishing a system for a nation-wide utilization of the reactor is indispensable. Now the system selects the required instruments with priorities. As the investment of the government accumulates and the user society grows, a request for a self reliance has become stronger. For a gradual approach, all the reactor systems should be carefully designed by considering future additional installations. A provision should also be made at the reactor design stage for the self reliance of a reactor operation for its long term survival. (author)

  12. Concept and designs of new-generation fast reactors

    International Nuclear Information System (INIS)

    Mitenkov, F.M.

    1993-01-01

    This article discusses the general safety requirements and characteristics for future nuclear power plants. It examines various designs - loop, block, and integrated layouts for reactors. Specifically, the article focuses an integrated design for sodium-cooled fast reactors noting that the BN-600 reactor has operated accident-free over the past 12 years. An obvious advantage of this scheme is that the coolant of the primary loop is localized in one volume (in a vessel), there are no short connections and large-diameter pipes, which of course sharply reduces the probability in coolant leaks. With an integrated scheme the problem of embrittlement of the reactor vessel by neutron irradiation is obviated. The neutron fluence for the vessels of the AST-500 and VPBER-600 reactors, built with an integrated scheme, is less than 10 17 cm -2 . Such a fluence does not cause any appreciable change in the mechanical properties of the vessel steel. The integrated layout of the reactor makes it possible to build a containment vessel. In this case it is possible to eliminate the danger of the reactor core drying out and thus cooling of the reactor in emergency situations can be simplified substantially. In an integrated layout, however, access is more difficult to the equipment inside the reactor, thus limiting or complicating maintenance work. The integrated layout, therefore, requires the use of highly reliable equipment built according to designs that have been proven in operation and have been passed representative service-life tests under laboratory conditions. The integrated layout considerably increases the mass and size characteristics of the reactor. New solutions thus are needed for the organization of work on reactor fabrication and assembly. In the case of the BN-600 and Superphenix reactors the welding of the reactor vessels and the assembly work were done on the building site

  13. Reactor operation

    CERN Document Server

    Shaw, J

    2013-01-01

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

  14. Development of technologies for nuclear reactors of small and medium sized

    International Nuclear Information System (INIS)

    2011-08-01

    This meeting include: countries presentations, themes and objectives of the training course, reactor types, design, EPR, APR1400, A P 1000, A PWR, ATMEA 1, VVER-1000, A PWR, ATMEA 1, VVER 1200, Boiling Water Reactor, A BWR, A BWR -II, ESBUR, Ke ren, AREVA, Heavy Water Reactor, Candu 6, Acr-1000, HWR, Bw, Iris, CAREM NuCcale, Smart, KLT-HOS, Westinghouse small modular Reactor, Gas Cooled Reactors, PBMR, React ores enfriados con metales liquidos, Hs, Prism,Terra Power, Hyper ion, appliance's no electric as de energia, Generation IV Reactors,VHTR, Gas Fast Reactor, Sodium Fast Reactor, Molten salt Reactor, Lfr, Water Cooled Reactor, Technology Assessment Process, Fukushima accident.

  15. Westinghouse small modular reactor design and application

    International Nuclear Information System (INIS)

    Blinn, R.; Godfrey, M.

    2012-01-01

    The AP1000 is currently under construction in both China and the US with the first one scheduled to come on line in late 2013. Nuclear power is a proven, safe, plentiful and clean source of power generation, and Westinghouse Electric Company, the pioneer and global leader in nuclear plant design and construction, is ready with the AP1000™ pressurized water reactor (PWR). The AP1000, based on the proven performance of Westinghouse-designed PWRs, is an advanced 1154 MWe nuclear power plant that uses the forces of nature and simplicity of design to enhance plant safety and operations and reduce construction costs.

  16. Westinghouse small modular reactor design and application

    Energy Technology Data Exchange (ETDEWEB)

    Blinn, R.; Godfrey, M. [Westinghouse Electric Company, Cranberry Township, Pennsilvania (United States)

    2012-07-01

    The AP1000 is currently under construction in both China and the US with the first one scheduled to come on line in late 2013. Nuclear power is a proven, safe, plentiful and clean source of power generation, and Westinghouse Electric Company, the pioneer and global leader in nuclear plant design and construction, is ready with the AP1000™ pressurized water reactor (PWR). The AP1000, based on the proven performance of Westinghouse-designed PWRs, is an advanced 1154 MWe nuclear power plant that uses the forces of nature and simplicity of design to enhance plant safety and operations and reduce construction costs.

  17. Startup and commissioning of pressurized water reactors

    International Nuclear Information System (INIS)

    Albert, L.J.; Gilbert, C.F.

    1983-05-01

    A critical phase of plant development is the test, startup, and commissioning period. The effort expended prior to commissioning has a definite effect on the reliability and continuing availability of the plant during its life. This paper describes a test, startup, and commissioning program for a pressurized water reactor (PWR) plant. This program commences with the completion of construction and continues through the turnover of equipment/systems to the owner's startup/ commissioning group. The paper addresses the organization of the test/startup group, planning and scheduling, test procedures and initial testing, staffing and certification of the test group, training of operators, and turnover to the owner

  18. Systemization of Design and Analysis Technology for Advanced Reactor

    International Nuclear Information System (INIS)

    Kim, Keung Koo; Lee, J.; Zee, S. K.

    2009-01-01

    The present study is performed to establish the base for the license application of the original technology by systemization and enhancement of the technology that is indispensable for the design and analysis of the advanced reactors including integral reactors. Technical reports and topical reports are prepared for this purpose on some important design/analysis methodology; design and analysis computer programs, structural integrity evaluation of main components and structures, digital I and C systems and man-machine interface design. PPS design concept is complemented reflecting typical safety analysis results. And test plans and requirements are developed for the verification of the advanced reactor technology. Moreover, studies are performed to draw up plans to apply to current or advanced power reactors the original technologies or base technologies such as patents, computer programs, test results, design concepts of the systems and components of the advanced reactors. Finally, pending issues are studied of the advanced reactors to improve the economics and technology realization

  19. Reactor and process design in sustainable energy technology

    CERN Document Server

    Shi, Fan

    2014-01-01

    Reactor Process Design in Sustainable Energy Technology compiles and explains current developments in reactor and process design in sustainable energy technologies, including optimization and scale-up methodologies and numerical methods. Sustainable energy technologies that require more efficient means of converting and utilizing energy can help provide for burgeoning global energy demand while reducing anthropogenic carbon dioxide emissions associated with energy production. The book, contributed by an international team of academic and industry experts in the field, brings numerous reactor design cases to readers based on their valuable experience from lab R&D scale to industry levels. It is the first to emphasize reactor engineering in sustainable energy technology discussing design. It provides comprehensive tools and information to help engineers and energy professionals learn, design, and specify chemical reactors and processes confidently. Emphasis on reactor engineering in sustainable energy techn...

  20. History, Development and Future of TRIGA Research Reactors

    International Nuclear Information System (INIS)

    2016-01-01

    Due to its particular fuel design and resulting enhanced inherent safety features, TRIGA reactors (Training, Research, Isotopes, General Atomics) constitute a ‘class of their own’ among the large variety of research reactors built world-wide. This publication summarizes in a single document the information on the past and present of TRIGA research reactors and presents an outlook in view of potential issues to be solved by TRIGA operating organizations in the near future. It covers the historical development and basic TRIGA characteristics, followed by utilization, fuel conversion and ageing management of TRIGA research reactors. It continues with issues and challenges, introduction to the global TRIGA research reactor network and concludes with future perspectives. The publication is complemented with a CD-ROM to illustrate the historical developments of TRIGA research reactors through individual facility examples and experiences

  1. Experimental reactors in the European Community and their utilization

    International Nuclear Information System (INIS)

    Ehringer, H.; Lecoq, J.P.

    1976-01-01

    Research and test reactors which in the first years of the peaceful use of nuclear energy had to found the basis for building and operation of commercial nuclear power plants, having achieved their aim, have faded into the background of the report. They still play an important role, however, for the further development of today's power reactor generation and for the development of progressing reactor lines as well as for fuel and material irradiation, for isotope production and, last but not least, for research and training. At the moment, over 100 test reactors are being operated in the widest sense in the European Community. In the present survey, their purpose and charge are dealt with particular consideration to the more important materials test reactors and to the programme reactors. (orig./LH) [de

  2. The role of nuclear reactors in space exploration and development

    International Nuclear Information System (INIS)

    Lipinski, R.J.

    2000-01-01

    The United States has launched more than 20 radioisotopic thermoelectric generators (RTGs) into space over the past 30 yr but has launched only one nuclear reactor, and that was in 1965. Russia has launched more than 30 reactors. The RTGs use the heat of alpha decay of 238 Pu for power and typically generate 235 U; typical designs are for 100 to 1000 kW of electricity. The only US space reactor launch (SNAP-10A) was a demonstration mission. One reason for the lack of space reactor use by the United States was the lack of space missions that required high power. But, another was the assumed negative publicity that would accompany a reactor launch. The net result is that all space reactor programs after 1970 were terminated before an operating space reactor could be developed, and they are now many years from recovering the ability to build them. Two major near-term needs for space reactors are the human exploration of Mars and advanced missions to and beyond the orbit of Jupiter. To help obtain public acceptance of space reactors, one must correct some of the misconceptions concerning space reactors and convey the following facts to the public and to decision makers: Space reactors are 1000 times smaller in power and size than a commercial power reactor. A space reactor at launch is only as radioactive as a pile of dirt 60 m (200 ft) across. A space reactor contains no plutonium at launch. It does not become significantly radioactive until it is turned on, and it will be engineered so that no launch accident can turn it on, even if that means fueling it after launch. The reactor will not be turned on until it is in a high stable orbit or even on an earth-escape trajectory for some missions. The benefits of space reactors are that they give humanity a stairway to the planets and perhaps the stars. They open a new frontier for their children and their grandchildren. They pave the way for all life on earth to move out into the solar system. At one time, humans built

  3. TMI-2 and reactor safety

    International Nuclear Information System (INIS)

    Kato, W.Y.

    1982-01-01

    The accident on March 28, 1979 at the Three Mile Island Unit 2 (TMI-2) Nuclear Power Station located near Harrisburg, Pennsylvania has had a major impact upon the design, construction, operation, regulation and safety research of nuclear power plants. It also has had an impact on utility and regulatory organizations. This paper presents a review of the TMI-2 accident scenario, its consequences and the conclusions of various investigating groups. An overview of the nuclear licensing process, computer codes used in safety analysis, safety research, and current issues is provided, using as a basis the TMI-2 accident. Some of the current safety issues which are included in the proposed NRC rulemaking hearings on such subjects as safety goals, siting criteria, emergency plans, and degraded or damaged core analysis are discussed. (author)

  4. Rationalization and future planning for AECL's research reactor capability

    International Nuclear Information System (INIS)

    Slater, J.B.

    1990-01-01

    AECL's research reactor capability has played a crucial role in the development of Canada's nuclear program. All essential concepts for the CANDU reactors were developed and tested in the NRX and NRU reactors, and in parallel, important contributions to basic physics were made. The technical feasibility of advanced fuel cycles and of the organic-cooled option for CANDU reactors were also demonstrated in the two reactors and the WR-1 reactor. In addition, an important and growing radio-isotope production industry was established and marketed on a world-wide basis. In 1984, however, it was recognized that a review and rationalization of the research reactor capability was required. The commercial success of the CANDU reactor system had reduced the scope and size of the required development program. Limited research and development funding and competition from other research facilities and programs, required that the scope be reduced to a support basis essential to maintain strategic capability. Currently, AECL, is part-way through this rationalization program and completion should be attained during 1992/93 when the MAPLE reactor is operational and decisions on NRX decommissioning will be made. A companion paper describes some of the unique operational and maintenance problems which have resulted from this program and the solutions which have been developed. Future planning must recognize the age of the NRU reactor (currently 32 years) and the need to plan for eventual replacement. Strategy is being developed and supporting studies include a full technical assessment of the NRU reactor and the required age-related upgrading program, evaluation of the performance characteristics and costs of potential future replacement reactors, particularly the advanced MAPLE concept, and opportunities for international co-operation in developing mutually supportive research programs

  5. The reactor and the production of isotopes

    International Nuclear Information System (INIS)

    Hevesy, G. de

    1962-01-01

    The construction of the cyclotron immensely advanced the availability of radioactive tracers, a few of which even today can be produced only with the aid of this device. But even this great advance was overshadowed by the fabulous production of isotopes by the reactors. Isotopes of almost any element and of almost unlimited activity became available. It now became possible to apply H 3 - discovered already in the 'thirties by Rutherford and Oliphant - and C 14 , and these were used in thousands of investigations

  6. Reactor numerical simulation and hydraulic test research

    International Nuclear Information System (INIS)

    Yang, L. S.

    2009-01-01

    In recent years, the computer hardware was improved on the numerical simulation on flow field in the reactor. In our laboratory, we usually use the Pro/e or UG commercial software. After completed topology geometry, ICEM-CFD is used to get mesh for computation. Exact geometrical similarity is maintained between the main flow paths of the model and the prototype, with the exception of the core simulation design of the fuel assemblies. The drive line system is composed of drive mechanism, guide bush assembly, fuel assembly and control rod assembly, and fitted with the rod level indicator and drive mechanism power device

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

    International Nuclear Information System (INIS)

    Wei Li; He Xuhong; Zhao Bingquan

    2003-01-01

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

  8. Present status of research reactor and future prospects

    International Nuclear Information System (INIS)

    Nakajima, Ken

    2013-01-01

    Research reactors have been playing an important role in the research and development of the various fields, such as physics, chemistry, biology, engineering, agriculture, medicine, etc. as well as human resource development. However, the most of them are older than 40 years, and the ageing management is an important issue. In Japan, only two research reactors are operational after the Great East Japan Earthquake in 2011. JAEA's reactors suffered from the quake and they are under inspections. Kyoto University Research Reactor, one of the operational reactors, has been widely used for research and human resource development, and the additional safety measures against the station blackout were installed. Besides the affect of the quake, the disposal or treatment of spent fuel becomes an inevitable problem for research reactors. The way of spent fuel disposal or treatment should be determined with the nation-wide and/or international coalition. (author)

  9. New or improved computational methods and advanced reactor design

    International Nuclear Information System (INIS)

    Nakagawa, Masayuki; Takeda, Toshikazu; Ushio, Tadashi

    1997-01-01

    Nuclear computational method has been studied continuously up to date, as a fundamental technology supporting the nuclear development. At present, research on computational method according to new theory and the calculating method thought to be difficult to practise are also continued actively to find new development due to splendid improvement of features of computer. In Japan, many light water type reactors are now in operations, new computational methods are induced for nuclear design, and a lot of efforts are concentrated for intending to more improvement of economics and safety. In this paper, some new research results on the nuclear computational methods and their application to nuclear design of the reactor were described for introducing recent trend of the nuclear design of the reactor. 1) Advancement of the computational method, 2) Reactor core design and management of the light water reactor, and 3) Nuclear design of the fast reactor. (G.K.)

  10. Reactor container and method of constructing the same

    International Nuclear Information System (INIS)

    Ichiki, Tadaharu; Muramatsu, Yutaka.

    1976-01-01

    Purpose: To reduce the construction term when constructing a reactor container and also alleviate the temperature requirements for the shielding wall by concurrently forwarding the work of constructing the reactor container and the shielding wall. Constitution: When constructing a reactor container, a shielding wall is constructed from lower layers on the outer side of the reactor container via a gap therefrom concurrently with the work of building the steel container. After the construction of the reactor container is completed, the reactor container is subjected to a pressure resistance test, leakage test, welding inspection and so forth. After the various tests and inspections are completed, concrete blocks as a protective material are stacked on the inner side of a steel frame mounted on the inner wall of the shielding wall, and they are secured to each other and also to the steel frame by concrete block holding members. (Moriyama, K.)

  11. Media and Australia's replacement reactor project

    International Nuclear Information System (INIS)

    Keenan, Pamela

    2001-01-01

    In September 1997, the Commonwealth Government of Australia announced a proposal to build a replacement nuclear research reactor at Lucas Heights in Sydney. Extensive public consultation, parliamentary debate and independent reports were prepared to ensure that the new facility would meet strict international requirements, national safety and environmental standards, and performance specifications servicing the needs of Australia - for decades to come. On 6 June 2000, Argentine company INVAP SE was announced as the preferred tenderer. In July 2000 contracts were signed between INVAP and the Australian Nuclear Science and Technology Organisation for the construction the replacement reactor, due to be completed in 2005. In order to retain a strong local presence, INVAP undertook a joint venture with two of Australia's foremost heavy construction businesses. Briefly the new research reactor will be a replacement for the ageing Australian Reactor (HIFAR). Nuclear science and technology, in Australia, is no stranger to media controversy and misinformation. Understandably the announcement of a preferred tenderer followed by the signing of contracts, attracted significant national and international media attention. However in the minds of the media, the issue is far from resolved and is now a constant 'news story' in the Australian media. Baseless media stories have made claims that the project will cost double the original estimates; question the credibility of the contractors; and raise issues of international security. The project is currently linked with Australia's requirements for long term nuclear waste management and there has been an attempt to bring national Indigenous People's issues into play. Some of these issues have been profiled in the press internationally. So, just to set the record straight and give you an appropriate impression of what's 'really happening' I would like to highlight a few issues, how ANSTO dealt with these, and what was finally reported

  12. Reactor safeguards

    CERN Document Server

    Russell, Charles R

    1962-01-01

    Reactor Safeguards provides information for all who are interested in the subject of reactor safeguards. Much of the material is descriptive although some sections are written for the engineer or physicist directly concerned with hazards analysis or site selection problems. The book opens with an introductory chapter on radiation hazards, the construction of nuclear reactors, safety issues, and the operation of nuclear reactors. This is followed by separate chapters that discuss radioactive materials, reactor kinetics, control and safety systems, containment, safety features for water reactor

  13. International Reactor Innovative and Secure (IRIS) summary

    International Nuclear Information System (INIS)

    Carelli, Mario D.

    2001-01-01

    The IRIS (International Reactor Innovative and Secure) reactor is described in the first part of the presentation. IRIS is a light water cooled reactor with an integral configuration, where steam generators, pumps and pressurizer are inside the reactor vessel. Partially funded by the DOE NERI program, IRIS is being developed by an international consortium of 16 organizations from seven countries. A key IRIS characteristic is its 'safety by design' approach which strives to eliminate, by design, as many accidents as possible rather than coping with their consequences. Initial returns are very positive; out of the eight Class IV accidents considered in the AP600 only one remains as a Class IV in IRIS, and at much reduced probability. Small-to-medium LOCAs have minimal consequences as the core remains safely under water for days, without the need for safety injection or water makeup. In spite of its novelty IRIS is firmly grounded on proven LWR technology and therefore a prototype is not needed to assure design certification. Rather, very extensive scaled tests will be performed to investigate the performance of in-vessel components such as steam generators and pumps, both individually and as interactive systems. Accident sequences will also be simulated and tested to prove IRIS safety by design claims. The first core fuel is less than 5% enriched and the fuel assembly is very similar to existing PWR assemblies, so there is no licensing challenge regarding the fuel. Because of the safety by design approach, yielding simplifications In design and accident management (e.g., IRIS does not have an emergency core cooling system), some accident scenarios are eliminated and others have lesser consequences. Thus, simplification and streamlining of the regulatory process might be possible. Risk informed regulation will be coupled with safety by design to show lower accident and damage probabilities. This could lead to a relaxation of siting regulatory requirements. It is

  14. BWR [boiling-water reactor] and PWR [pressurized-water reactor] off-normal event descriptions

    International Nuclear Information System (INIS)

    1987-11-01

    This document chronicles a total of 87 reactor event descriptions for use by operator licensing examiners in the construction of simulator scenarios. Events are organized into four categories: (1) boiling-water reactor abnormal events; (2) boiling-water reactor emergency events; (3) pressurized-water reactor abnormal events; and (4) pressurized-water reactor emergency events. Each event described includes a cover sheet and a progression of operator actions flow chart. The cover sheet contains the following general information: initial plant state, sequence initiator, important plant parameters, major plant systems affected, tolerance ranges, final plant state, and competencies tested. The progression of operator actions flow chart depicts, in a flow chart manner, the representative sequence(s) of expected immediate and subsequent candidate actions, including communications, that can be observed during the event. These descriptions are intended to provide examiners with a reliable, performance-based source of information from which to design simulator scenarios that will provide a valid test of the candidates' ability to safely and competently perform all licensed duties and responsibilities

  15. Reactor physics innovations of the advanced CANDU reactor core: adaptable and efficient

    International Nuclear Information System (INIS)

    Chan, P.S.W.; Hopwood, J.M.; Bonechi, M.

    2003-01-01

    The Advanced CANDU Reactor (ACR) is designed to have a benign, operator-friendly core physics characteristic, including a slightly negative coolant-void reactivity and a moderately negative power coefficient. The discharge fuel burnup is about three times that of natural uranium fuel in current CANDU reactors. Key features of the reactor physics innovations in the ACR core include the use of H 2 O coolant, slightly enriched uranium (SEU) fuel, and D 2 O moderator in a reduced lattice pitch. These innovations result in substantial improvements in economics, as well as significant enhancements in reactor performance and waste reduction over the current reactor design. The ACR can be readily adapted to different power outputs by increasing or decreasing the number of fuel channels, while maintaining identical fuel and fuel-channel characteristics. The flexibility provided by on-power refuelling and simple fuel bundle design enables the ACR to easily adapt to the use of plutonium and thorium fuel cycles. No major modifications to the basic ACR design are required because the benign neutronic characteristics of the SEU fuel cycle are also inherent in these advanced fuel cycles. (author)

  16. Reactor power automatically controlling method and device for BWR type reactor

    International Nuclear Information System (INIS)

    Murata, Akira; Miyamoto, Yoshiyuki; Tanigawa, Naoshi.

    1997-01-01

    For an automatic control for a reactor power, when a deviation exceeds a predetermined value, the aimed value is kept at a predetermined value, and when the deviation is decreased to less than the predetermined value, the aimed value is increased from the predetermined value again. Alternatively, when a reactor power variation coefficient is decreased to less than a predetermine value, an aimed value is maintained at a predetermined value, and when the variation coefficient exceeds the predetermined value, the aimed value is increased. When the reactor power variation coefficient exceeds a first determined value, an aimed value is increased to a predetermined variation coefficient, and when the variation coefficient is decreased to less than the first determined value and also when the deviation between the aimed value and an actual reactor power exceeds a second determined value, the aimed value is maintained at a constant value. When the deviation is increased or when the reactor power variation coefficient is decreased, since the aimed value is maintained at predetermined value without increasing the aimed value, the deviation is not increased excessively thereby enabling to avoid excessive overshoot. (N.H.)

  17. Research and development of a super fast reactor (12). Considerations for the reactor characteristics

    International Nuclear Information System (INIS)

    Goto, Shoji; Ishiwatari, Yuki; Oka, Yoshiaki

    2008-01-01

    A research program aimed at developing the Super Fast Reactor (Super FR) has been entrusted by the Ministry of Education, Culture, Sports, Science and Technology (MEXT) of Japan since December 2005. It includes the following three projects. (A) Development of the Super Fast Reactor concept. (B)Thermal-hydraulic experiments. (C) Materials development. Tokyo Electric Power Company (TEPCO) has joined this program and works on part (A) together with the University of Tokyo. From the utility's viewpoint, it is important to consider the most desirable characteristics for Super FR to have. Four issues were identified in project (A), (1) Fuel design, (2) Reactor core design, (3) Safety, and (4) Plant characteristics of Super FR. This report describes the desired characteristics of Super FR with respect to item (1) fuel design and item (2) Reactor core design, as compared with a boiling water reactor (BWR) plant. The other two issues will be discussed in this project, and will also be considered in the design process of Super FR. (author)

  18. A friendly Maple module for one and two group reactor model

    International Nuclear Information System (INIS)

    Baptista, Camila O.; Pavan, Guilherme A.; Braga, Kelmo L.; Silva, Marcelo V.; Pereira, P.G.S.; Werner, Rodrigo; Antunes, Valdir; Vellozo, Sergio O.

    2015-01-01

    The well known two energy groups core reactor design model is revisited. A simple and friendly Maple module was built to cover the steps calculations of a plate reactor in five situations: 1. one group bare reactor, 2. two groups bare reactor, 3. one group reflected reactor, 4. 1-1/2 groups reflected reactor and 5. two groups reflected reactor. The results show the convergent path of critical size, as it should be. (author)

  19. Research reactors for power reactor fuel and materials testing - Studsvik's experience

    International Nuclear Information System (INIS)

    Grounes, M.

    1998-01-01

    Presently Studsvik's R2 test reactor is used for BWR and PWR fuel irradiations at constant power and under transient power conditions. Furthermore tests are performed with defective LWR fuel rods. Tests are also performed on different types of LWR cladding materials and structural materials including post-irradiation testing of materials irradiated at different temperatures and, in some cases, in different water chemistries and on fusion reactor materials. In the past, tests have also been performed on HTGR fuel and FBR fuel and materials under appropriate coolant, temperature and pressure conditions. Fuel tests under development include extremely fast power ramps simulating some reactivity initiated accidents and stored energy (enthalpy) measurements. Materials tests under development include different types of in-pile tests including tests in the INCA (In-Core Autoclave) facility .The present and future demands on the test reactor fuel in all these cases are discussed. (author)

  20. DOE fundamentals handbook: Nuclear physics and reactor theory

    International Nuclear Information System (INIS)

    1993-01-01

    The Nuclear Physics and Reactor Theory Handbook was developed to assist nuclear facility operating contractors in providing operators, maintenance personnel, and the technical staff with the necessary fundamentals training to ensure a basic understanding of nuclear physics and reactor theory. The handbook includes information on atomic and nuclear physics; neutron characteristics; reactor theory and nuclear parameters; and the theory of reactor operation. This information will provide personnel with a foundation for understanding the scientific principles that are associated with various DOE nuclear facility operations and maintenance

  1. 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

  2. International Seminar on Gasification 2009 - Biomass Gasification, Gas Clean-up and Gas Treatment

    Energy Technology Data Exchange (ETDEWEB)

    2009-10-15

    During the seminar international and national experts gave presentations concerning Biomass gasification, Gas cleaning and gas treatment; and Strategy and policy issues. The presentations give an overview of the current status and what to be expected in terms of development, industrial interest and commercialization of different biomass gasification routes. The following PPT presentations are reproduced in the report: Black Liquor Gasification (Chemrec AB.); Gasification and Alternative Feedstocks for the Production of Synfuels and 2nd Generation Biofuels (Lurgi GmbH); Commercial Scale BtL Production on the Verge of Becoming Reality (Choren Industries GmbH.); Up-draft Biomass Gasification (Babcock and Wilcox Voelund A/S); Heterogeneous Biomass Residues and the Catalytic Synthesis of Alcohols (Enerkem); Status of the GoBiGas-project (Goeteborg Energi AB.); On-going Gasification Activities in Spain (University of Zaragoza,); Biomass Gasification Research in Italy (University of Perugia.); RDandD Needs and Recommendations for the Commercialization of High-efficient Bio-SNG (Energy Research Centre of the Netherlands.); Cleaning and Usage of Product Gas from Biomass Steam Gasification (Vienna University of Technology); Biomass Gasification and Catalytic Tar Cracking Process Development (Research Triangle Institute); Syngas Cleaning with Catalytic Tar Reforming (Franhofer UMSICHT); Biomass Gas Cleaning and Utilization - The Topsoee Perspective (Haldor Topsoee A/S); OLGA Tar Removal Technology (Dahlman); Bio-SNG - Strategy and Activities within E.ON (E.ON Ruhrgas AG); Strategy and Gasification Activities within Sweden (Swedish Energy Agency); 20 TWh/year Biomethane (Swedish Gas Association)

  3. Retrospective Reactor Dosimetry for Neutron Fluence, Helium, and Boron Measurements

    International Nuclear Information System (INIS)

    Greenwood, Lawrence R.; Oliver, Brian M.

    2003-01-01

    Neutron fluences can be measured and radiation damage parameters determined by analyzing the neutron reaction products in very small samples removed from components of an operating power research reactor. This process, known as retrospective reactor dosimetry, provides precise neutron exposure parameters for establishing or validating calculations of neutron fluences, helium generation, and radiation damage to reactor materials. Correlation of the neutron fluence and helium data helps to establish and validate models of radiation damage and helium production that are needed to address important issues such as irradiation assisted stress corrosion cracking, void swelling, and weld repair of cracks. Results are presented for samples recently obtained from several operating reactors

  4. Tritium release from lithium silicate and lithium aluminate, in-reactor and out-of-reactor

    Energy Technology Data Exchange (ETDEWEB)

    Johnson, A.B. Jr.

    1976-09-01

    Studies were conducted to determine the generation and evolution of tritium and helium in lithium aluminate (LiAlO/sub 2/) and lithium silicate (Li/sub 2/SiO/sub 3/) by the reaction: Li/sup 6/ + n ..-->.. /sup 4/He + T. Targets were irradiated 4.4 days in the K-West Reactor snout facility. (Silicate GVR* approximately 2.0 cc/cc; aluminate GVR approximately 1.4 cc/cc.) Gas release in-reactor was determined by post-irradiation drilling experiments on aluminum ampoules containing silicate and aluminate targets. In-reactor tritium release (at approximately 100/sup 0/C) was found to decrease linearly with increasing target density. Tritium released in-reactor was primarily in the noncondensible form (HT and T/sub 2/), while in laboratory extractions (300-1300/sup 0/C), the tritium appeared primarily in the condensible form (HTO and T/sub 2/O). Concentrations of HT (and presumably HTO) were relatively high, indicating moisture pickup in canning operations or by inleakage of moisture after the capsule was welded. Impurities in extracted gases included H/sub 2/O, CO/sub 2/, CO, O/sub 2/, H/sub 2/, NO, SO/sub 2/, SiF/sub 4/ and traces of hydrocarbons.

  5. Tritium release from lithium silicate and lithium aluminate, in-reactor and out-of-reactor

    International Nuclear Information System (INIS)

    Johnson, A.B. Jr.

    1976-09-01

    Studies were conducted to determine the generation and evolution of tritium and helium in lithium aluminate (LiAlO 2 ) and lithium silicate (Li 2 SiO 3 ) by the reaction: Li 6 + n → 4 He + T. Targets were irradiated 4.4 days in the K-West Reactor snout facility. (Silicate GVR* approximately 2.0 cc/cc; aluminate GVR approximately 1.4 cc/cc.) Gas release in-reactor was determined by post-irradiation drilling experiments on aluminum ampoules containing silicate and aluminate targets. In-reactor tritium release (at approximately 100 0 C) was found to decrease linearly with increasing target density. Tritium released in-reactor was primarily in the noncondensible form (HT and T 2 ), while in laboratory extractions (300-1300 0 C), the tritium appeared primarily in the condensible form (HTO and T 2 O). Concentrations of HT (and presumably HTO) were relatively high, indicating moisture pickup in canning operations or by inleakage of moisture after the capsule was welded. Impurities in extracted gases included H 2 O, CO 2 , CO, O 2 , H 2 , NO, SO 2 , SiF 4 and traces of hydrocarbons

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

    International Nuclear Information System (INIS)

    Ho, V.; Apostolakis, G.

    1988-01-01

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

  7. Fixed-bed Reactor Dynamics and Control - A Review

    DEFF Research Database (Denmark)

    Jørgensen, S. B.

    1986-01-01

    The industrial diversity of fixed bed reactors offers a challenging and relevant set of control problems. These intricate problems arise due to the rather complex dynamics of fixed bed reactors and to the complexity of actual reactor configurations. Many of these control problems are nonlinear...... and multi-variable. During the last decade fixed bed reactor control strategies have been proposed and investigated experimentally. This paper reviews research on these complex control problems with an emphasis upon solutions which have been demon-strated to work in the laboratory and hold promise...

  8. Analysis and evaluation of the Dual Fluid Reactor concept

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Xiang

    2017-06-27

    The Dual Fluid Reactor is a molten salt fast reactor developed by IFK in Berlin based on the Gen-IV Molten-Salt Reactor concept and the Liquid-Metal Cooled Reactor. The design aims to combine these two concepts to improve these two concepts. The Dissertation focuses on the concept and performs diverse calculations and estimations on the subjects of neutron physics, depletion and thermal-hydraulic behaviors to validate the new features of the concept. Based on the results it is concluded that this concept is feasible to its desired purpose and with great potential.

  9. An analysis of reactor pit pressurization and forces applied on reactor vessel

    International Nuclear Information System (INIS)

    Wang Rongzhong; Li Feng

    1997-12-01

    The pressure and temperature transients with the time of the reactor pit during LOCA have been analyzed by using Catem computer code for Qinshan-2 nuclear power plant. The force and bending moment on the inlet and outlet nozzles of the reactor vessel also have been calculated by using Wformom code. Qinshan-2 NPP is a two-loop nuclear power plant. The cold water of the accumulators are directly injected into the downcomer of reactor vessel. Injection line of accumulators is located at the same level with the inlet and outlet nozzles. These geometry characteristics have been taken into account in the circumferential vessel pit nodding using five volumes around the vessel. The assumptions used in the analysis and calculation results have been presented. Many sensitive calculations have been performed for different break size and circumferential nodding

  10. Conceptual designs of tokamak reactor and R D

    International Nuclear Information System (INIS)

    Fukai, Yuzo; Yamato, Harumi; Sawada, Yoshio

    1983-01-01

    The conceptual design of both FER (Fusion Experimental Reactor) and R-project is now under way as the new step of JT-60. From the engineering viewpoint, these reactors, requiring D-T operation, have the challenge, such as the handling of tritium and components irradiated by neutron bombardment. Toshiba's design team is participating to these projects in order to realize the reactor and plant concept coping with the above objectives. This paper represents the conceptual design contributions of the FER and R-project as well as R D technology which are now under development, such as tritium handling app aratus, reactor materials, etc. (author)

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

    International Nuclear Information System (INIS)

    Montes Ponce de Leon, J.

    1974-01-01

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

  12. 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)

  13. Tank Vapor Sampling and Analysis Data Package for Tank 241-Z-361 Sampled 09/22/1999 and 09/271999 During Sludge Core Removal

    Energy Technology Data Exchange (ETDEWEB)

    VISWANATH, R.S.

    1999-12-29

    This data package presents sampling data and analytical results from the September 22 and 27, 1999, headspace vapor sampling of Hanford Site Tank 241-2-361 during sludge core removal. The Lockheed Martin Hanford Corporation (LMHC) sampling team collected the samples and Waste Management Laboratory (WML) analyzed the samples in accordance with the requirements specified in the 241-2361 Sludge Characterization Sampling and Analysis Plan, (SAP), HNF-4371, Rev. 1, (Babcock and Wilcox Hanford Corporation, 1999). Six SUMMA{trademark} canister samples were collected on each day (1 ambient field blank and 5 tank vapor samples collected when each core segment was removed). The samples were radiologically released on September 28 and October 4, 1999, and received at the laboratory on September 29 and October 6, 1999. Target analytes were not detected at concentrations greater than their notification limits as specified in the SAP. Analytical results for the target analytes and tentatively identified compounds (TICs) are presented in Section 2.2.2 starting on page 2B-7. Three compounds identified for analysis in the SAP were analyzed as TICs. The discussion of this modification is presented in Section 2.2.1.2.

  14. Tank Vapor Sampling and Analysis Data Package for Tank 241-Z-361 Sampled 09/22/1999 and 09/27/1999, During Sludge Core Removal

    International Nuclear Information System (INIS)

    VISWANATH, R.S.

    1999-01-01

    This data package presents sampling data and analytical results from the September 22 and 27, 1999, headspace vapor sampling of Hanford Site Tank 241-2-361 during sludge core removal. The Lockheed Martin Hanford Corporation (LMHC) sampling team collected the samples and Waste Management Laboratory (WML) analyzed the samples in accordance with the requirements specified in the 241-2361 Sludge Characterization Sampling and Analysis Plan, (SAP), HNF-4371/Rev. 1, (Babcock and Wilcox Hanford Corporation, 1999). Six SUMMA(trademark) canister samples were collected on each day (1 ambient field blank and 5 tank vapor samples collected when each core segment was removed). The samples were radiologically released on September 28 and October 4, 1999, and received at the laboratory on September 29 and October 6, 1999. Target analytes were not detected at concentrations greater than their notification limits as specified in the SAP. Analytical results for the target analytes and tentatively identified compounds (TICs) are presented in Section 2.2.2 starting on page 2B-7. Three compounds identified for analysis in the SAP were analyzed as TICs. The discussion of this modification is presented in Section 2.2.1.2

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

    International Nuclear Information System (INIS)

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

    1994-01-01

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

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

    International Nuclear Information System (INIS)

    Matejka, K.

    1998-01-01

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

  17. Decommissioning and re-utilization of the Musashi Reactor

    International Nuclear Information System (INIS)

    Tomio Tanzawa; Nobukazu Iijima; Norikazu Horiuchi; Tadashi Yoshida; Tetsuo Matsumoto; Naoto Hagura; Ryouhei Kamiya

    2008-01-01

    The Musashi Institute of Technology Research Reactor (the Musashi Reactor) is a TRIGA-? with maximum thermal power of 100 kW. The decommissioning was decided in May, 2003. The reactor facility is now under decommissioning. The phased decommissioning was selected. Phase 1 consists of permanent shutdown of the reactor and stopping the operational functions, and transportation of the spent nuclear fuels. After completion of the transportation, the reactor facility is characterized as the storage of low level radioactive materials. This is phase 2. Activities of phase 1 were completed and the facility is now under phase 2. Activities of phase 3 consist of dismantling the reactor tank and the shielding, and delivering the radioactive waste to a waste disposal facility. The phase 3 will be started on condition that the undertaking of the waste disposal for research reactors will be established. On the other hand, re-utilization of the facility has being studied, and 'realistic' reactor simulator was turned out by utilizing the reactor installations such as control rod drive and operation console. (authors)

  18. Reactivity requirements and safety systems for heavy water reactors

    International Nuclear Information System (INIS)

    Kati, S.L.; Rustagi, R.S.

    1977-01-01

    The natural uranium fuelled pressurised heavy water reactors are currently being installed in India. In the design of nuclear reactors, adequate attention has to be given to the safety systems. In recent years, several design modifications having bearing on safety, in the reactor processes, protective and containment systems have been made. These have resulted either from new trends in safety and reliability standards or as a result of feed-back from operating reactors of this type. The significant areas of modifications that have been introduced in the design of Indian PHWR's are: sophisticated theoretical modelling of reactor accidents, reactivity control, two independent fast acting systems, full double containment and improved post-accident depressurisation and building clean-up. This paper brings out the evolution of design of safety systems for heavy water reactors. A short review of safety systems which have been used in different heavy water reactors, of varying sizes, has been made. In particular, the safety systems selected for the latest 235 MWe twin reactor unit station in Narora, in Northern India, have been discussed in detail. Research and Development efforts made in this connection are discussed. The experience of design and operation of the systems in Rajasthan and Kalpakkam reactors has also been outlined

  19. Zero energy reactor RB technical characteristics and experimental possibilities

    International Nuclear Information System (INIS)

    Jovanovic, S.; Takac, S.; Raisic, N.; Lolic, B.; Markovic, H.

    1963-04-01

    The zero energy reactor RB was constructed in 1958 in accordance with the nuclear reactor development programme of the Boris Kidric Institute of Nuclear Sciences. The reactor was in operation until the middle of 1959 when the heavy water, serving as the moderator, was transported to the high flux reactor RA, built at the same time at the Boris Kidric Institute. Owing to the fact that the purchase of new quantities of heavy water was planned for 1961 it was decided to reconstruct the RB reactor in order to improve the safety of the system and to obtain better flexibility in performing the experiments. New control, safety and radiation monitoring systems were constructed. Some changes were also made on the reactor tank, water circulation system and the water level monitoring equipment. The reconstruction was completed in 1961. and the heavy water was delivered early in 1962. The reconstructed reactor was critical for the first time in summer 1962, and from that time was in continuous operation. This report presents an outline of the design and construction characteristics of the reactor. The main intention is to inform potential users of the reactor about experimental possibilities, advantages and disadvantages of such a critical facility

  20. Chernobyl: recovery operations and the entombment of Reactor 4

    International Nuclear Information System (INIS)

    Dalziel, S.P.C.

    1988-01-01

    The immediate actions taken following the accident at the Chernobyl-number 4 reactor in April 1986 are described. These included actions to put out the fires, initial medical aid and the dropping of sand, lead, dolomite and boron onto the reactor from helicopters. Following this the chamber below the damaged reactor core was filled with concrete to prevent any further explosions or meltdown. The reactor was subsequently entombed in steel and concrete. The evacuation of the surrounding area is also mentioned. (U.K.)

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

    International Nuclear Information System (INIS)

    Krull, W.

    1981-01-01

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

  2. Contributions of research Reactors in science and technology

    International Nuclear Information System (INIS)

    Butt, N.M.; Bashir, J.

    1992-12-01

    In the present paper, after defining a research reactor, its basic constituents, types of reactors, their distribution in the world, some typical examples of their uses are given. Particular emphasis in placed on the contribution of PARR-I (Pakistan Research Reactor-I), the 5 MW Swimming Pool Research reactor which first became critical at the Pakistan Institute of Nuclear Science and Technology (PINSTECH) in Dec. 1965 and attained its full power in June 1966. This is still the major research facility at PINSTECH for research and development. (author)

  3. Plastic reactor suitable for high pressure and supercritical fluid electrochemistry

    DEFF Research Database (Denmark)

    Branch, Jack; Alibouri, Mehrdad; Cook, David A.

    2017-01-01

    The paper describes a reactor suitable for high pressure, particularly supercritical fluid, electrochemistry and electrodeposition at pressures up to 30 MPa at 115◦C. The reactor incorporates two key, new design concepts; a plastic reactor vessel and the use of o-ring sealed brittle electrodes....... These two innovations widen what can be achieved with supercritical fluid electrodeposition. The suitability of the reactor for electroanalytical experiments is demonstrated by studies of the voltammetry of decamethylferrocene in supercritical difluromethane and for electrodeposition is demonstrated...

  4. 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.

  5. Nuclear reactor internals construction and failed fuel rod detection system

    International Nuclear Information System (INIS)

    Frisch, E.; Andrews, H.N.

    1976-01-01

    A system is provided for determining during operation of a nuclear reactor having fluid pressure operated control rod mechanisms the exact location of a fuel assembly with a defective fuel rod. The construction of the reactor internals is simplified in a manner to facilitate the testing for defective fuel rods and the reduce the cost of producing the upper internals of the reactor. 13 claims, 10 drawing figures

  6. Risk-assessment techniques and the reactor licensing process

    International Nuclear Information System (INIS)

    Levine, S.

    1979-01-01

    A brief description of the Reactor Safety Study (WASH-1400), concentrating on the engineering aspects of the contribution to reactor accident risks is followed by some comments on how we have applied the insights and techniques developed in this study to prepare a program to improve the safety of nuclear power plants. Some new work we are just beginning on the application of risk-assessment techniques to stablize the reactor licensing process is also discussed

  7. Chernobyl and the safety of nuclear reactors in OECD countries

    International Nuclear Information System (INIS)

    1987-01-01

    This report assesses the possible bearing of the Chernobyl accident on the safety of nuclear reactors in OECD countries. It discusses analyses of the accident performed in several countries as well as improvements to the safety of RBMK reactors announced by the USSR. Several remaining questions are identified. The report compares RBMK safety features with those of commercial reactors in OECD countries and evaluates a number of issues raised by the Chernobyl accident

  8. BR2 reactor: medical and industrial applications

    International Nuclear Information System (INIS)

    Ponsard, B.

    2005-01-01

    The radioisotopes are produced for various applications in the nuclear medicine (diagnostic, therapy, palliation of metastatic bone pain), industry (radiography of welds, ...), agriculture (radiotracers, ...) and basic research. Due to the availability of high neutron fluxes (thermal neutron flux up to 10 15 n/cm 2 .s), the BR2 reactor is considered as a major facility through its contribution for a continuous supply of products such 99 Mo ( 99 mTc), 131 I, 133 Xe, 192 Ir, 186 Re, 153 Sm, 90 Y, 32 P, 188 W ( 188 Re), 203 Hg, 82 Br, 41 Ar, 125 I, 177 Lu, 89 Sr, 60 Co, 169 Yb, 147 Nd, and others. Neutron Transmutation Doped (NTD) silicon is produced for the semiconductor industry in the SIDONIE (Silicon Doping by Neutron Irradiation Experiment) facility, which is designed to continuously rotate and traverse the silicon through the neutron flux. These combined movements produce exceptional dopant homogeneity in batches of silicon measuring 4 and 5-inches in diameter by up to 750 mm in length. The main objectives of work performed were to provide a reliable and qualitative supply of radioisotopes and NTD-silicon to the customers in accordance with a quality system that has been certified to the requirements of the EN ISO 9001: 2000. This new Quality System Certificate has been obtained in November 2003 for the Production of radioisotopes for medical and industrial applications and the Production of Neutron Transmutation Doped (NTD) Silicon in the BR2 reactor

  9. ANAEROBIC DIGESTION AND THE DENITRIFICATION IN UASB REACTOR

    Directory of Open Access Journals (Sweden)

    José Tavares de Sousa

    2008-01-01

    Full Text Available The environmental conditions in Brazil have been contributing to the development of anaerobic systems in the treatment of wastewaters, especially UASB - Upflow Anaerobic Sludge Blanket reactors. The classic biological process for removal of nutrients uses three reactors - Bardenpho System, therefore, this work intends an alternative system, where the anaerobic digestion and the denitrification happen in the same reactor reducing the number of reactors for two. The experimental system was constituted by two units: first one was a nitrification reactor with 35 L volume and 15 d of sludge age. This system was fed with raw sanitary waste. Second unit was an UASB, with 7.8 L and 6 h of hydraulic detention time, fed with ¾ of effluent nitrification reactor and ¼ of raw sanitary waste. This work had as objective to evaluate the performance of the UASB reactor. In terms of removal efficiency, of bath COD and nitrogen, it was verified that the anaerobic digestion process was not affected. The removal efficiency of organic material expressed in COD was 71%, performance already expected for a reactor of this type. It was also observed that the denitrification process happened; the removal nitrate efficiency was 90%. Therefore, the denitrification process in reactor UASB is viable.

  10. Predictions and measurements of isothermal airflow in a model once-through steam generator

    Energy Technology Data Exchange (ETDEWEB)

    Carter, H R; Promey, G J; Rush, G C

    1982-11-01

    Once-Through Steam Generators (OTSGs) are used in the Nuclear Steam Supply Systems marketed by The Babcock and Wilcox Company (B and W). To analytically predict the three-dimensional, steady-state thermohydraulic conditions in the OTSG, B and W has developed a proprietary code THEDA-1 and is working in cooperation with EPRI to develop an improved version, THEDA-2. Confident application of THEDA requires experimental verification to demonstrate that the code can accurately describe the thermohydraulic conditions in geometries characteristic of the OTSG. The first step in the THEDA verification process is the subject of this report. A full-scale, partial-section model of two OTSG spans was constructed and tested using isothermal air as the working fluid. Model local velocities and pressure profiles were measured and compared to THEDA prediction for five model configurations. Over 3000 velocity measurements were taken and the results were compared to THEDA predictions. Agreement between measured and predicted velocity data was generally better than +-12.5%.

  11. Uranium utilization of light water cooled reactors and fast breeders

    International Nuclear Information System (INIS)

    Stojadinovic, Timm

    1991-08-01

    The better uranium utilization of fast breeder reactors as compared with water cooled reactors is one argument in favour of the breeder introduction. This report tries to quantify this difference. It gives a generally valid formalism for the uranium utilization as a function of the fuel burnup, the conversion rate, fuel cycle losses and the fuel enrichment. On the basis of realistic assumptions, the ratio between the utilizations of breeder reactors to that of light water cooled reactors (LWR) amounts to 180 for the open LWR cycle and 100 in case of plutonium recycling in LWRs

  12. Extended life for the Borssele reactor. Risks and cost

    International Nuclear Information System (INIS)

    Van Gelder, J.W.; De Rijk, P.

    2005-01-01

    In the last few years several decisions were taken by the Dutch government with respect to the service life of the nuclear power reactor Borssele. The last decision was taken in 2003 to close the reactor in the year 2013. However, the discussion started again while it appears, according to the Dutch Minister of State for the Environment, there is not enough political and legal basis to enforce closure of the reactor in 2013. In this report insight is given into the risks and societal cost to continue the operation of the Borssele reactor after 2013 [nl

  13. Shutdown channels and fitted interlocks in atomic reactors

    International Nuclear Information System (INIS)

    Furet, J.; Landauer, C.

    1968-01-01

    This catalogue consists of tables (one per reactor) giving the following information: number and type of detectors, range of the shutdown channels, nature of the associated electronics, thresholds setting off the alarms, fitted interlocks. These cards have been drawn up with a view to an examination of the reactors safety by the 'Reactor Safety Sub-Commission', they take into account the latest decisions. The reactors involved in this review are: Azur, Cabri, Castor-Pollux, Cesar-Marius-2, Edf-2, EL3, EL4, Eole, G1, G2-G3, Harmonie, Isis, Masurca, Melusine, Minerve, Osiris, Pegase, Peggy, PAT, Rapsodie, SENA, Siloe, Siloette, Triton-Nereide, and Ulysse. (authors) [fr

  14. Antineutrino and gamma emission from the OSIRIS research reactor

    Directory of Open Access Journals (Sweden)

    Giot Lydie

    2017-01-01

    Full Text Available For the first time, the summation method has been coupled with a complete reactor model, in order to predict the antineutrino emission of a research reactor. This work, discussed in the first part of this paper, allows us to predict the low energy part of the antineutrino spectrum, evidencing the important contribution of actinides to the antineutrino emission. Experimental conditions at short distance from research reactors are challenging, because the reactor itself produces huge gamma background that induce accidental and correlated backgrounds in an antineutrino target. The understanding of this background is of utmost importance and triggered the second part of the work presented here.

  15. Development of an MMS/PC based real time simulation of the B and W NSS plant for advanced control system design

    International Nuclear Information System (INIS)

    Bartells, P.S.; Brownell, R.B.

    1990-01-01

    The development of this personal-computer-based simulation of the Babcock and Wilcox nuclear steam system (NSS) was prompted in part by the need for a real-time analysis tool to be used in evaluating advanced control concepts for the NSS. NSS control is currently accomplished via conventional analog systems that are becoming increasingly obsolete. With the widespread use of digital micro-processor-based control systems for fossil power and other applications, the B and W Owners Group Advanced Control System Task Force is developing a next-generation control system for upgrading existing B and W power plants. To take advantage of the digital control technology, it is desirable to have a flexible, cost-effective, and portable control analysis tool available that can simulate various postulated control strategies and algorithms and couple these with simulated plant responses in real time to determine overall effectiveness. To develop the desired capability, B and W has incorporated the simulation methodology of the Modular Modeling System (MMS) and the knowledge gained during development of a similar Department of Energy-funded project. The MMS-based NSS model was developed and then modified to increase execution speed, ported to an IBM Personal System 2 (Model 80) and interfaced with user-friendly graphics. The user can develop alternative control strategies and readily interface them with the NSS model for real-time display and evaluation. The paper addresses the key considerations and programming techniques used to accomplish the resulting simulation

  16. Reactor Neutrinos

    OpenAIRE

    Kim, Soo-Bong; Lasserre, Thierry; Wang, Yifang

    2013-01-01

    We review the status and the results of reactor neutrino experiments. Short-baseline experiments have provided the measurement of the reactor neutrino spectrum, and their interest has been recently revived by the discovery of the reactor antineutrino anomaly, a discrepancy between the reactor neutrino flux state of the art prediction and the measurements at baselines shorter than one kilometer. Middle and long-baseline oscillation experiments at Daya Bay, Double Chooz, and RENO provided very ...

  17. Reactor abnormality diagnosis device and its method

    International Nuclear Information System (INIS)

    Honma, Hitoshi; Hirayama, Tatsuya.

    1992-01-01

    The present invention rapidly detects leakage of primary coolants due to rupture of heat transfer pipes of a steam generator in a PWR type reactor to diagnose the operation state of the reactor. That is, a radiation detector is disposed to a secondary main steam pipeline for supplying steams generated from the steam generator to a turbine. The radiation detector detects a dose rate or a counting rate continuously. The measured data are transferred to an calculation and processing system and compared with the standard of normal values to diagnose the presence of leaks. Alternatively, radiation detectors are disposed at the upstream and the downstream of the secondary system main steam pipeline respectively. The signals from each of the radiation detectors are processed by the calculation and processing system as the change with lapse of time. As a result, the scale of the ruptured portion of the heat transfer pipe in the steam generator is diagnosed based on the value of radioactivity concentration in the main steams. (I.S.)

  18. RA Research reactor Annual report 1982 - Part 1, Operation, maintenance and utilization of the RA reactor

    International Nuclear Information System (INIS)

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

    1982-12-01

    Reactor test operation started in September 1981 at 2 MW power with 80% enriched fuel continued during 1982 according to the previous plan. The initial reactor core was made of 44 fuel channel each containing 10 fuel slugs. The first half of 1982 was used for the needed measurements and analysis of operating parameters and functioning of reactor systems and equipment under operating conditions. Program concerned with the testing operation at higher power levels was started in the second half of this year. It was found that the inherent excess reactivity and control rod worths ensure safe operation according to the IAEA safety standards. Excess reactivity is high enough to enable higher power level of 4.7 MW during 4 monthly cycles each lasting 15-20 days. Favourable conditions for cooling exist for the initial core configuration. Effects of poisoning at startup on the reactivity and power density distribution were measured as well as initial spatial distribution of the neutron flux which was 3,9 10 13 cm -2 s -1 at 2 MW power. Modification of the calibration coefficient in the system for automated power level control was determined. All the results show that all the safety criteria and limitations concerned with fuel utilization are fulfilled if reactor power would be 4.7 MW. Additional testing operation at 3, 4, and 4.7 MW power levels will be needed after obtaining the licence for operating at nominal power. Transition from the initial core with 44 fuel channels to the equilibrium lattice configuration with 72 fuel channels each containing 10 fuel slugs, would be done gradually. Reactor was not operated in September because of the secondary coolant pipes were exchanged between Danube and the horizontal sedimentary. Control and maintenance of the reactor equipment was done regularly and efficiently dependent on the availability of the spare parts. Difficulties in maintenance of the reactor instrumentation were caused by unavailability of the outdated spare parts

  19. Research Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Martens, Frederick H. [Argonne National Laboratory; Jacobson, Norman H.

    1968-09-01

    This booklet discusses research reactors - reactors designed to provide a source of neutrons and/or gamma radiation for research, or to aid in the investigation of the effects of radiation on any type of material.

  20. Nuclear reactors

    International Nuclear Information System (INIS)

    Middleton, J.E.

    1977-01-01

    Reference is made to water cooled reactors and in particular to the cooling system of steam generating heavy water reactors (SGHWR). A two-coolant circuit is described for the latter. Full constructural details are given. (U.K.)

  1. Space reactor system and subsystem investigations: assessment of technology issues for the reactor and shield subsystem. SP-100 Program

    International Nuclear Information System (INIS)

    Atkins, D.F.; Lillie, A.F.

    1983-01-01

    As part of Rockwell's effort on the SP-100 Program, preliminary assessment has been completed of current nuclear technology as it relates to candidate reactor/shield subsystems for the SP-100 Program. The scope of the assessment was confined to the nuclear package (to the reactor and shield subsystems). The nine generic reactor subsystems presented in Rockwell's Subsystem Technology Assessment Report, ESG-DOE-13398, were addressed for the assessment

  2. Kinetics and Product Selectivity (Yield) of Second Order Competitive Consecutive Reactions in Fed-Batch Reactor and Plug Flow Reactor

    OpenAIRE

    Selvamony, Subash Chandra Bose

    2013-01-01

    This literature compares the performance of second order competitive consecutive reaction in Fed-Batch Reactor with that in continuous Plug Flow Reactor. In a kinetic sense, this simulation study aims to develop a case for continuous Plug Flow Reactor in pharmaceutical, fine chemical, and related other chemical industries. MATLAB is used to find solutions for the differential equations. The simulation results show that, for certain cases of nonelementary scenario, product selectivity is highe...

  3. BOILING REACTORS

    Science.gov (United States)

    Untermyer, S.

    1962-04-10

    A boiling reactor having a reactivity which is reduced by an increase in the volume of vaporized coolant therein is described. In this system unvaporized liquid coolant is extracted from the reactor, heat is extracted therefrom, and it is returned to the reactor as sub-cooled liquid coolant. This reduces a portion of the coolant which includes vaporized coolant within the core assembly thereby enhancing the power output of the assembly and rendering the reactor substantially self-regulating. (AEC)

  4. Research and training reactors in the German Democratic Republic - basic data and ultilization

    International Nuclear Information System (INIS)

    Weibrecht, R.; Ziegenbein, D.; Knorr, J.; Meyer, K.

    1988-01-01

    A description of the following five research and training reactors in operation is given: (1) RFR, a light water cooled and moderated reactor of the Soviet type WWR-S, 10 MW in power; (2) RRR, a zero-power reactor of Argonaut type; (3) RAKE, a zero-power reactor of tank type; (4) AKR, a zero-power reactor of the homogeneous type; (5) ZLFR, a zero-power reactor of tank type. Utilization of the reactors is outlined for problems of reactor physics and neutronic research, for production of radioactive isotopes and of neutron-doped silicon, of activation analysis, and for education and training. (author)

  5. Reactor coolant and associated systems in nuclear power plants

    International Nuclear Information System (INIS)

    1986-01-01

    This Safety Guide outlines the design requirements for the reactor coolant and associated systems (RCAS) and the features required in order to achieve their safety functions. It covers design considerations for various reactor types and encompasses the safety aspects of the functions of the RCAS both during normal operation and following postulated initiating events, and to some extent also for decommissioning

  6. Reactor vessel

    NARCIS (Netherlands)

    Makkee, M.; Kapteijn, F.; Moulijn, J.A.

    1999-01-01

    A reactor vessel (1) comprises a reactor body (2) through which channels (3) are provided whose surface comprises longitudinal inwardly directed parts (4) and is provided with a catalyst (6), as well as buffer bodies (8, 12) connected to the channels (3) on both sides of the reactor body (2) and

  7. Feasibility study for fast reactor and related fuel cycle. Preliminary studies in 1998

    International Nuclear Information System (INIS)

    Hayafune, Hiroki; Enuma, Yasuhiro; Kubota, Kenichi; Yoshida, Masashi; Uno, Osamu; Ishikawa, Hiroyasu; Kobayashi, Jun; Umetsu, Youichiro; Ichimiya, Masakazu

    1999-10-01

    Prior to the feasibility study for fast reactors (FRs) starting from the 1999 fiscal year, planned in the medium and long-term program of JNC, preliminarily studies were performed on 'FR systems except sodium cooled MOX fueled reactors'. Small scale or module type reactors, heavy metal (Pb or Pb-Bi) cooled reactors, gas cooled reactors, light water cooled reactors, and molten salt reactors were studied on the basis of literature. They were evaluated from the viewpoint of the technical possibility (the structure integrity, earthquake resistance, safety, productivity, operability, maintenance repair, difficulty of the development), the long-term targets (market competitiveness as an energy system, utilization of uranium resources, reduction of radioactive waste, security of the non-proliferation), and developmental risk. As the result, the following concepts should be studied for future commercialized FRs. Small scale and module type reactor: Middle-sized reactor with an excellent economical efficiency. Small power reactor with a multipurpose design concept. Gas cooled reactor: CO2 gas cooled reactor, He gas cooled reactor. Heavy metal cooled reactor: Russian type lead cooled reactor. Light water cooled reactor: Light water cooled high converter reactor and super critical pressure light water cooled reactor. Molten salt reactor: Trichloride molten salt reactor which matches the U-Pu cycle. (author)

  8. Analysis of vertical flow during ambient and pumped conditions in four monitoring wells at the Pantex Plant, Carson County, Texas, July-September 2008

    Science.gov (United States)

    Stanton, Gregory P.; Thomas, Jonathan V.; Stoval, Jeffery

    2009-01-01

    The Pantex Plant is a U.S. Department of Energy/National Nuclear Security Administration (USDOE/NNSA)-owned, contractor-operated facility managed by Babcock & Wilcox Technical Services Pantex, LLC (B&W Pantex) in Carson County, Texas, approximately 17 miles northeast of Amarillo. The U.S. Geological Survey, in cooperation with B&W Pantex through the USDOE/NNSA, made a series of flowmeter measurements and collected other borehole geophysical logs during July–September 2008 to analyze vertical flow in screened intervals of four selected monitoring wells (PTX01–1012, PTX06–1044, PTX06–1056, and PTX06–1068) at the Pantex Plant. Hydraulic properties (transmissivity values) of the section of High Plains (Ogallala) aquifer penetrated by the wells also were computed. Geophysical data were collected under ambient and pumped flow conditions in the four monitoring wells. Unusually large drawdowns occurred at two monitoring wells (PTX06–1044 and PTX06–1056) while the wells were pumped at relatively low rates. A decision was made to redevelop those wells, and logs were run again after redevelopment in the two monitoring wells.

  9. Cascading pressure reactor and method for solar-thermochemical reactions

    Science.gov (United States)

    Ermanoski, Ivan

    2017-11-14

    Reactors and methods for solar thermochemical reactions are disclosed. The reactors and methods include a cascade of reduction chambers at successively lower pressures that leads to over an order of magnitude pressure decrease compared to a single-chambered design. The resulting efficiency gains are substantial, and represent an important step toward practical and efficient solar fuel production on a large scale.

  10. Reactor power reduction system and method

    International Nuclear Information System (INIS)

    1980-01-01

    An improved control method for maintaining the operation of a nuclear reactor system in response to an event which requires an immediate but less than complete power reduction called an accelerated power reduction, by rapidly inserting into said reactor core a portion of said regulating rods selected from said predetermined regulation sequence. (author)

  11. Safety and licensing for small and medium power reactors

    International Nuclear Information System (INIS)

    Trauger, D.B.

    1987-01-01

    Proposed new concepts for small and medium power reactors differ substantially from traditional Light Water Reactors (LWRs). Although designers have a large base of experience in safety and licensing, much of it is not relevant to new concepts. It can be a disadvantage if regulators apply LWR rules directly. A fresh start is appropriate. The extensive interactions between industry, regulators, and the public complicates but may enhance safety. It is basic to recognize the features that distinguish nuclear energy safety from that for other industries. These features include: nuclear reactivity, fission product radiation, and radioactive decay heat. Small and medium power reactors offer potential advantages over LWRs, particularly for reactivity and decay heat

  12. Extension of the technical scope of the Paris and Vienna Conventions: fusion reactors and reactors in means of transport

    International Nuclear Information System (INIS)

    Reye, S.

    1993-01-01

    This paper examines the possibility of extending the technical scope of the Vienna and Paris Conventions to two types of nuclear installation presently excluded. Industrial use of fusion reactors is not expected for several decades, but the present revision of the liability regime provides a useful opportunity to ensure in advance that future industrial reactors will be covered, as well as covering risks arising from existing research reactors. Inclusion of nuclear reactors comprised in means of transport (in practice, in ships) in the liability regime would have certain advantages, but given their almost exclusively military use, such a proposal would be politically controversial. 18 refs

  13. Correction | Sofola | International Journal of Medicine and ...

    African Journals Online (AJOL)

    Authors' names should also appear as: “Shobowale E.O1*, Elikwu C.J1,Olusanya A.O2, Abiodun O3”. The affiliations remain the same: “'1Department of Medical Microbiology and Parasitology, Babcock University, Ogun State. 2Pathcare Laboratories, Lagos. 3Department of Community Health, Babcock University, Ogun ...

  14. Problems and prospects of small and medium power reactors

    International Nuclear Information System (INIS)

    Matin, A.

    1977-01-01

    Prior to 1973 it was generally believed that small and medium power reactors (SMPRs) had a potentially large market and only their high capital costs prevented their large scale commercial application. In December, 1973, crude oil price rose from US $2.50 per barrel to more than US $11 per barrel. This changed the economic position of SMPRs so much so that even 100-200 MWe nuclear reactors were considered economic compared to oil-fired plants. A Market Survey by the International Atomic Energy Agency in 1974 showed that the potential market for reactors ranging from 150 to 400 MWe during 1980-1990 amounted to 140 units with a total installed capacity of 38,000 MWe. This potential market did not, however, generate the desired interest among the reactor manufacturers. So far only three manufacturers based in Europe have shown interest in SMPRs and at present small reactors are being built commercially only in India. Among developing countries, Bangladesh, Jamaica and Kuwait are seriously looking for reactors in sizes of 100-200 MWe. The paper analyses the historic background of SMPRs and problems related to their commercial application and suggests the following actions: i) The British 100 MWe SGHWR is considered proven and suitable for small grids and hence deserves financial support by British/International Financing Agencies. ii) Any re-engineered or slightly re-designed version of operating small light water reactors will find wider acceptability than available new adaptions of marine reactors. Manufacturers of operating small LWRs may be encouraged through international financial assistance to make such designs commercially available. iii) Small CANDU reactors may be suitable for most developing countries and need technical and economic support from Canada for their export. iv) The Agency must continue their effort more vigorously for making SMPRs commercially available to small developing countries

  15. Experience, status and future of the computerized reactor instrumentation at the TRIGA reactor Vienna

    International Nuclear Information System (INIS)

    Frankl, M.; Boeck, H.; Katrik, P.; Schachner, H.

    1997-01-01

    The paper describes the 33 years old history of the instrumentation of the TRIGA reactor Vienna and focuses on the present computerized instrumentation installed in 1992. The experience of three years of operation is discussed and some of the failures are analyzed. Potential future problems both with soft- and hardware as well as with spare part supplies are analyzed. (author). 6 figs

  16. Development of core design and analyses technology for integral reactor

    International Nuclear Information System (INIS)

    Zee, Sung Quun; Lee, C. C.; Kim, K. Y.

    2002-03-01

    In general, small and medium-sized integral reactors adopt new technology such as passive and inherent safety concepts to minimize the necessity of power source and operator actions, and to provide the automatic measures to cope with any accidents. Specifically, such reactors are often designed with a lower core power density and with soluble boron free concept for system simplification. Those reactors require ultra long cycle operation for higher economical efficiency. This cycle length requirement is one of the important factors in the design of burnable absorbers as well as assurance of shutdown margin. Hence, both computer code system and design methodology based on the today's design technology for the current commercial reactor cores require intensive improvement for the small and medium-sized soluble boron free reactors. New database is also required for the development of this type of reactor core. Under these technical requirements, conceptual design of small integral reactor SMART has been performed since July 1997, and recently completed under the long term nuclear R and D program. Thus, the final objectives of this work is design and development of an integral reactor core and development of necessary indigenous design technology. To reach the goal of the 2nd stage R and D program for basic design of SMART, design bases and requirements adequate for ultra long cycle and soluble boron free concept are established. These bases and requirements are satisfied by the core loading pattern. Based on the core loading pattern, nuclear, and thermal and hydraulic characteristics are analyzed. Also included are fuel performance analysis and development of a core protection and monitoring system that is adequate for the soluble boron free core of an integral reactor. Core shielding design analysis is accomplished, too. Moreover, full scope interface data are produced for reactor safety and performance analyses and other design activities. Nuclear, thermal and

  17. Sodium fast reactor safety and licensing research plan. Volume II.

    Energy Technology Data Exchange (ETDEWEB)

    Ludewig, H. (Brokhaven National Laboratory, Upton, NY); Powers, D. A.; Hewson, John C.; LaChance, Jeffrey L.; Wright, A. (Argonne National Laboratory, Argonne, IL); Phillips, J.; Zeyen, R. (Institute for Energy Petten, Saint-Paul-lez-Durance, France); Clement, B. (IRSN/DPAM.SEMIC Bt 702, Saint-Paul-lez-Durance, France); Garner, Frank (Radiation Effects Consulting, Richland, WA); Walters, Leon (Advanced Reactor Concepts, Los Alamos, NM); Wright, Steve; Ott, Larry J. (Oak Ridge National Laboratory, Oak Ridge, TN); Suo-Anttila, Ahti Jorma; Denning, Richard (Ohio State University, Columbus, OH); Ohshima, Hiroyuki (Japan Atomic Energy Agency, Ibaraki, Japan); Ohno, S. (Japan Atomic Energy Agency, Ibaraki, Japan); Miyhara, S. (Japan Atomic Energy Agency, Ibaraki, Japan); Yacout, Abdellatif (Argonne National Laboratory, Argonne, IL); Farmer, M. (Argonne National Laboratory, Argonne, IL); Wade, D. (Argonne National Laboratory, Argonne, IL); Grandy, C. (Argonne National Laboratory, Argonne, IL); Schmidt, R.; Cahalen, J. (Argonne National Laboratory, Argonne, IL); Olivier, Tara Jean; Budnitz, R. (Lawrence Berkeley National Laboratory, Berkeley, CA); Tobita, Yoshiharu (Japan Atomic Energy Agency, Ibaraki, Japan); Serre, Frederic (Centre d' %C3%94etudes nucl%C3%94eaires de Cadarache, Cea, France); Natesan, Ken (Argonne National Laboratory, Argonne, IL); Carbajo, Juan J. (Oak Ridge National Laboratory, Oak Ridge, TN); Jeong, Hae-Yong (Korea Atomic Energy Research Institute, Daejeon, Korea); Wigeland, Roald (Idaho National Laboratory, Idaho Falls, ID); Corradini, Michael (University of Wisconsin-Madison, Madison, WI); Thomas, Justin (Argonne National Laboratory, Argonne, IL); Wei, Tom (Argonne National Laboratory, Argonne, IL); Sofu, Tanju (Argonne National Laboratory, Argonne, IL); Flanagan, George F. (Oak Ridge National Laboratory, Oak Ridge, TN); Bari, R. (Brokhaven National Laboratory, Upton, NY); Porter D. (Idaho National Laboratory, Idaho Falls, ID); Lambert, J. (Argonne National Laboratory, Argonne, IL); Hayes, S. (Idaho National Laboratory, Idaho Falls, ID); Sackett, J. (Idaho National Laboratory, Idaho Falls, ID); Denman, Matthew R.

    2012-05-01

    Expert panels comprised of subject matter experts identified at the U.S. National Laboratories (SNL, ANL, INL, ORNL, LBL, and BNL), universities (University of Wisconsin and Ohio State University), international agencies (IRSN, CEA, JAEA, KAERI, and JRC-IE) and private consultation companies (Radiation Effects Consulting) were assembled to perform a gap analysis for sodium fast reactor licensing. Expert-opinion elicitation was performed to qualitatively assess the current state of sodium fast reactor technologies. Five independent gap analyses were performed resulting in the following topical reports: (1) Accident Initiators and Sequences (i.e., Initiators/Sequences Technology Gap Analysis), (2) Sodium Technology Phenomena (i.e., Advanced Burner Reactor Sodium Technology Gap Analysis), (3) Fuels and Materials (i.e., Sodium Fast Reactor Fuels and Materials: Research Needs), (4) Source Term Characterization (i.e., Advanced Sodium Fast Reactor Accident Source Terms: Research Needs), and (5) Computer Codes and Models (i.e., Sodium Fast Reactor Gaps Analysis of Computer Codes and Models for Accident Analysis and Reactor Safety). Volume II of the Sodium Research Plan consolidates the five gap analysis reports produced by each expert panel, wherein the importance of the identified phenomena and necessities of further experimental research and code development were addressed. The findings from these five reports comprised the basis for the analysis in Sodium Fast Reactor Research Plan Volume I.

  18. Sodium fast reactor safety and licensing research plan - Volume II

    International Nuclear Information System (INIS)

    Ludewig, H.; Powers, D.A.; Hewson, John C.; LaChance, Jeffrey L.; Wright, A.; Phillips, J.; Zeyen, R.; Clement, B.; Garner, Frank; Walters, Leon; Wright, Steve; Ott, Larry J.; Suo-Anttila, Ahti Jorma; Denning, Richard; Ohshima, Hiroyuki; Ohno, S.; Miyhara, S.; Yacout, Abdellatif; Farmer, M.; Wade, D.; Grandy, C.; Schmidt, R.; Cahalen, J.; Olivier, Tara Jean; Budnitz, R.; Tobita, Yoshiharu; Serre, Frederic; Natesan, Ken; Carbajo, Juan J.; Jeong, Hae-Yong; Wigeland, Roald; Corradini, Michael; Thomas, Justin; Wei, Tom; Sofu, Tanju; Flanagan, George F.; Bari, R.; Porter D.

    2012-01-01

    Expert panels comprised of subject matter experts identified at the U.S. National Laboratories (SNL, ANL, INL, ORNL, LBL, and BNL), universities (University of Wisconsin and Ohio State University), international agencies (IRSN, CEA, JAEA, KAERI, and JRC-IE) and private consultation companies (Radiation Effects Consulting) were assembled to perform a gap analysis for sodium fast reactor licensing. Expert-opinion elicitation was performed to qualitatively assess the current state of sodium fast reactor technologies. Five independent gap analyses were performed resulting in the following topical reports: (1) Accident Initiators and Sequences (i.e., Initiators/Sequences Technology Gap Analysis), (2) Sodium Technology Phenomena (i.e., Advanced Burner Reactor Sodium Technology Gap Analysis), (3) Fuels and Materials (i.e., Sodium Fast Reactor Fuels and Materials: Research Needs), (4) Source Term Characterization (i.e., Advanced Sodium Fast Reactor Accident Source Terms: Research Needs), and (5) Computer Codes and Models (i.e., Sodium Fast Reactor Gaps Analysis of Computer Codes and Models for Accident Analysis and Reactor Safety). Volume II of the Sodium Research Plan consolidates the five gap analysis reports produced by each expert panel, wherein the importance of the identified phenomena and necessities of further experimental research and code development were addressed. The findings from these five reports comprised the basis for the analysis in Sodium Fast Reactor Research Plan Volume I.

  19. Current tendencies and perspectives of development research reactors of Russia

    International Nuclear Information System (INIS)

    Gabaraev, B.A.; Kchmelschikov, V.V.

    2004-01-01

    Full text: During more than fifty years many Research Reactors were constructed under Russian projects, and that is a considerable contribution to the world reactor building. The designs of Research Reactors, constructed under Russian projects, appeared to be so successful, that permitted to raise capacity and widen the range of their application. The majority of Russian Research Reactors being middle-aged are far from having their designed resources exhausted and are kept on the intensive run still. In 2000 'Strategy of nuclear power development in Russia in the first half of XXI century' was elaborated and approved. The national nuclear power requirements and possible ways of its development determined in this document demanded to analyze the state of the research reactors base. The analysis results are presented in this report. The main conclusion consists in the following statement: on the one hand quantity and experimental potentialities of domestic Research Reactors are sufficient for the solution of reactor materials science tasks, and on the other hand the reconstruction and modernization appears to be the most preferable way of research reactors development for the near-term outlook. At present time the modernization and reconstruction works and works on extension of operational life of high-powered multipurpose MIR-M1, SM-3, IRV-1M, BOR-60, IVV-2M and others are conducted. There is support for the development of Research Reactors, intended for carrying out the fundamental investigations on the neutron beams. Toward this end the Government of Russia gives financial and professional support with a view to complete the reactor PIK construction in PINPh and the reactor IBR-2 modernization in JINR. In future prospect Research Reactors branch in Russia is to acquire the following trends: - limited number of existent scientific centers, based on the construction sites, with high flux materials testing research reactors, equipped with experimental facilities

  20. Small reactors in the Canadian context: opportunities and challenges

    Energy Technology Data Exchange (ETDEWEB)

    Walker, R.S. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada)

    2013-07-01

    This presentation discusses the opportunities and challenges for small reactors in Canada. It concludes by suggesting that the success of small reactors in Canada will depend on a number of factors including private sector investment, access to international markets, stable, equitable and adaptable regulatory regime, public trust and technology.

  1. Oak Ridge Research reactor shutdown maintenance and surveillance

    International Nuclear Information System (INIS)

    Coleman, G.H.; Laughlin, D.L.

    1991-05-01

    The Department of Energy ordered the Oak Ridge Research Reactor to be placed in permanent shutdown on July 14, 1987. The paper outlines routine maintenance activities and surveillance tests performed April through September, 1990, on the reactor instrumentation and controls, process system, and the gaseous waste filter system. Preparations are being made to transfer the facility to the Remedial Action Program. 6 tabs

  2. Preliminary conceptual design and analysis on KALIMER reactor structures

    International Nuclear Information System (INIS)

    Kim, Jong Bum

    1996-10-01

    The objectives of this study are to perform preliminary conceptual design and structural analyses for KALIMER (Korea Advanced Liquid Metal Reactor) reactor structures to assess the design feasibility and to identify detailed analysis requirements. KALIMER thermal hydraulic system analysis results and neutronic analysis results are not available at present, only-limited preliminary structural analyses have been performed with the assumptions on the thermal loads. The responses of reactor vessel and reactor internal structures were based on the temperature difference of core inlet and outlet and on engineering judgments. Thermal stresses from the assumed temperatures were calculated using ANSYS code through parametric finite element heat transfer and elastic stress analyses. While, based on the results of preliminary conceptual design and structural analyses, the ASME Code limits for the reactor structures were satisfied for the pressure boundary, the needs for inelastic analyses were indicated for evaluation of design adequacy of the support barrel and the thermal liner. To reduce thermal striping effects in the bottom are of UIS due to up-flowing sodium form reactor core, installation of Inconel-718 liner to the bottom area was proposed, and to mitigate thermal shock loads, additional stainless steel liner was also suggested. The design feasibilities of these were validated through simplified preliminary analyses. In conceptual design phase, the implementation of these results will be made for the design of the reactor structures and the reactor internal structures in conjunction with the thermal hydraulic, neutronic, and seismic analyses results. 4 tabs., 24 figs., 4 refs. (Author)

  3. Dynamics and Control of Chemical Reactors-Selectively Surveyed

    DEFF Research Database (Denmark)

    Jørgensen, S. B.; Jensen, N.

    1989-01-01

    in industry, many reactor control problems are still left unsolved or only partly solved using open loop strategies where disturbance rejection and model inaccuracies have to be handled through manual reactor control and feedback control of raw material preprocessing and product purification operations...

  4. Operation monitoring and protection method for nuclear reactor

    International Nuclear Information System (INIS)

    Tochihara, Hiroshi.

    1995-01-01

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

  5. SRAC: JAERI thermal reactor standard code system for reactor design and analysis

    International Nuclear Information System (INIS)

    Tsuchihashi, Keichiro; Takano, Hideki; Horikami, Kunihiko; Ishiguro, Yukio; Kaneko, Kunio; Hara, Toshiharu.

    1983-01-01

    The SRAC (Standard Reactor Analysis Code) is a code system for nuclear reactor analysis and design. It is composed of neutron cross section libraries and auxiliary processing codes, neutron spectrum routines, a variety of transport, 1-, 2- and 3-D diffusion routines, dynamic parameters and cell burn-up routines. By making the best use of the individual code function in the SRAC system, the user can select either the exact method for an accurate estimate of reactor characteristics or the economical method aiming at a shorter computer time, depending on the purpose of study. The user can select cell or core calculation; fixed source or eigenvalue problem; transport (collision probability or Sn) theory or diffusion theory. Moreover, smearing and collapsing of macroscopic cross sections are separately done by the user's selection. And a special attention is paid for double heterogeneity. Various techniques are employed to access the data storage and to optimize the internal data transfer. Benchmark calculations using the SRAC system have been made extensively for the Keff values of various types of critical assemblies (light water, heavy water and graphite moderated systems, and fast reactor systems). The calculated results show good prediction for the experimental Keff values. (author)

  6. Status of the advanced boiling water reactor and simplified boiling water reactor

    International Nuclear Information System (INIS)

    Smith, P.F.

    1992-01-01

    This paper reports that the excess of U.S. electrical generating capacity which has existed for the past 15 years is coming to an end as we enter the 1990s. Environmental and energy security issues associated with fossil fuels are kindling renewed interest in the nuclear option. The importance of these issues are underscored by the National Energy Strategy (NES) which calls for actions which are designed to ensure that the nuclear power option is available to utilities. Utilities, utility associations, and nuclear suppliers, under the leadership of the Nuclear Power Oversight Committee (NPOC), have jointly developed a 14 point strategic plan aimed at establishing a predictable regulatory environment, standardized and pre-licensed Advanced Light Water Reactor (ALWR) nuclear plants, resolving the long-term waste management issue, and other enabling conditions. GE is participating in this national effort and GE's family of advanced nuclear power plants feature two new reactor designs, developed on a common technology base, aimed at providing a new generation of nuclear plants to provide safe, clean, economical electricity to the world's utilities in the 1990s and beyond. Together, the large-size (1300 MWe) Advanced Boiling Water Reactor (ABWR) and the small-size (600 MWe) Simplified Boiling Water Reactor (SBWR) are innovative, near-term candidates for expanding electrical generating capacity in the U.S. and worldwide. Both possess the features necessary to do so safely, reliably, and economically

  7. The training and research reactor of the Zittau Technical College

    International Nuclear Information System (INIS)

    Ackermann, G.; Hampel, R.; Konschak, K.

    1979-01-01

    The light-water moderated training and research reactor of the Zittau Technical College, which has been put into operation 1 July 1979, is described. Having a power of 10 MW, it is provided for education of students and advanced training of nuclear power plant staff members. High inherent nuclear safety and economy of operation are achieved by appropriate design of the reactor core and the use of fresh fuel elements provided for the 10-MW research reactor at the Rossendorf Central Institute for Nucleear Research for one year on a loan basis. Further characteristics of the reactor are easy accessibility of the core interior for in-core studies, sufficient external experimental channels, and a control and protection system meeting the requirements of teaching operation. The installed technological and dosimetric devices not only ensure reliable operation of the reactor, but also extend the potentialities of experimental work and education that is reported in detail. The principles on which the training programs are based are explained in the light of some examples. The training reactor is assumed to serve for providing basic knowledge about processes in nuclear power stations with pressurized water reactors. Where the behaviour of a nuclear power station cannot sufficiently be demonstrated by the training reactor, a reasonable completion of practical training at special simulation models and experimental facilities of the Technical College and at the nuclear power plant simulator of the Rheinsberg nuclear power plant school has been conceived. (author)

  8. Recommended safety objectives, principles and requirements for mini-reactors

    International Nuclear Information System (INIS)

    1991-05-01

    Canadian and international publications containing objectives, principles and requirements for the safety of nuclear facilities in general and nuclear power plants in particular have been reviewed for their relevance to mini-reactors. Most of the individual recommendations, sometimes with minor wording changes, are applicable to mini-reactors. However, some prescriptive requirements for the shutdown, emergency core cooling and containment systems of power reactors are considered inappropriate for mini-reactors. The Advisory Committee on Nuclear Safety favours a generally non-prescriptive approach whereby the applicant for a mini-reactor license is free to propose any means of satisfying the fundamental objectives, but must convince the regulatory agency to that effect. To do so, a probabilistic safety assessment (PSA) would be the favoured procedure. A generic PSA for all mini-reactors of the same design would be acceptable. Notwithstanding this non-prescriptive approach, the ACNS considers that it would be prudent to require the existence of at least one independent shutdown system and two physically independent locations from which the reactor can be shut down and the shutdown condition monitored, and to require provision for an assumed loss of integrity of the primary cooling system's boundary unless convincing arguments to the contrary are presented. The ACNS endorses in general the objectives and fundamental principles proposed by the interorganizational Small Reactor Criteria working group, and intends to review and comment on the documents on specific applications to be issued by that working group

  9. Safe design and operation of fluidized-bed reactors: Choice between reactor models

    NARCIS (Netherlands)

    Westerink, E.J.; Westerterp, K.R.

    1990-01-01

    For three different catalytic fluidized bed reactor models, two models presented by Werther and a model presented by van Deemter, the region of safe and unique operation for a chosen reaction system was investigated. Three reaction systems were used: the oxidation of benzene to maleic anhydride, the

  10. Reactor - and accelerator-based filtered beams

    International Nuclear Information System (INIS)

    Mill, A.J.; Harvey, J.R.

    1980-01-01

    The neutrons produced in high flux nuclear reactors and in accelerator, induced fission and spallation reactions, represent the most intense sources of neutrons available for research. However, the neutrons from these sources are not monoenergetic, covering the broad range extending from 10 -3 eV up to 10 7 eV or so. In order to make quantitative measurements of the effects of neutrons and their dependence on neutron energy it is desirable to have mono-energetic neutron sources. The paper describes briefly methods of obtaining mono-energetic neutrons and different methods of filtration. This is followed by more detailed discussion of neutron window filters and a summary of the filtered beam facilities using this technique. The review concludes with a discussion of the main applications of filtered beams and their present and future importance

  11. A Joint Report on PSA for New and Advanced Reactors

    International Nuclear Information System (INIS)

    2013-01-01

    This report addresses the application of Probabilistic Safety Assessment (PSA) to new and advanced nuclear reactors. As far as advanced reactors are concerned, the objectives were to characterize the ability of current PSA technology to address key questions regarding the development, acceptance and licensing of advanced reactor designs, to characterize the potential value of advanced PSA methods and tools for application to advanced reactors, and to develop recommendations for any needed developments regarding PSA for these reactors. As far as the design and commissioning of new nuclear power plants is concerned, the objectives were to identify and characterize current practices regarding the role of PSA, to identify key technical issues regarding PSA, lessons learned and issues requiring further work; to develop recommendations regarding the use of PSA, and to identify future international cooperative work on the identified issues. In order to reach these objectives, questionnaires had been sent to participating countries and organisations

  12. Treatment of spent fuels from research reactors and reactor development programs in Germany

    International Nuclear Information System (INIS)

    Closs, K.D.

    1999-01-01

    Quite a great number of different types of spent fuel from research reactors and development programs exists in Germany. The general policy is to send back to the USA as long as possible fuel from MTRs and TRIGAs of USA origin. An option is reprocessing in Great Britain or France. This option is pursued as long as reprocessing and reuse of the recovered material is economically justifiable. For those fuels which cannot be returned to the USA or which will not be reprocessed, a domestic back-up solution of spent fuel management has been developed in Germany, compatible with the management of spent fuel from power reactors. It consists in dry storage in special casks and, later on, direct disposal. Preliminary results from experimental R and D investigations with research reactor fuel and experience from LWR fuel lead to the conclusion that the direct disposal option even for research reactor fuel or exotic fuel does not impose major technical difficulties for the German waste management and disposal concept. (author)

  13. Stop valve with automatic control and locking for nuclear reactors

    International Nuclear Information System (INIS)

    Chung, D.K.

    1980-01-01

    This invention generally concerns an automatic control and locking stop valve. Specifically it relates to the use of such a valve in a nuclear reactor of the type containing absorber elements supported by a fluid and intended for stopping the reactor in complete safety [fr

  14. Industrial and commercial applications for a Triga reactor

    International Nuclear Information System (INIS)

    Green, D.

    1986-01-01

    The Physics and Radioisotope Services Group of ICI operates a Triga Reactor in support of a commercial, Industrial Radioisotope Technology Service. The technical and commercial development of this business is discussed in the context of operating a Triga Reactor in an Industrial Environment. (author)

  15. Management and storage of nuclear fuel from Belgian research reactors

    International Nuclear Information System (INIS)

    Gubel, P.

    1996-01-01

    Experiences and problems with the storage of irradiated fuel at research reactors in Belgium are described. In particular, interim storage problems exist for spent fuel elements at the BR2 and the shut down BR3 reactors in Mol. (author). 1 ref

  16. Hydrogen production from high temperature electrolysis and fusion reactor

    International Nuclear Information System (INIS)

    Dang, V.D.; Steinberg, J.F.; Issacs, H.S.; Lazareth, O.; Powell, J.R.; Salzano, F.J.

    1978-01-01

    Production of hydrogen from high temperature electrolysis of steam coupled with a fusion reactor is studied. The process includes three major components: the fusion reactor, the high temperature electrolyzer and the power conversion cycle each of which is discussed in the paper. Detailed process design and analysis of the system is examined. A parametric study on the effect of process efficiency is presented

  17. Fast reactor physics at CEA: present studies and future prospects

    International Nuclear Information System (INIS)

    Hammer, P.

    1980-09-01

    This paper aims at giving a general survey of the fast reactor core physics and shielding studies wich are in progress at CEA (1979-1983) in order to solve the neutronic problems related to: - core design optimization, - reactor operation and fuel management, - safety, for the development of fast commercial breeders in France after the SUPER-PHENIX 1 construction is achieved

  18. Potential market and characteristics of low-temperature reactors

    International Nuclear Information System (INIS)

    Lerouge, B.

    1975-01-01

    The low-temperature (100 to 200 deg C) heat market for industrial applications and district heating is very important. Two main studies have been developed: a swimming pool reactor delivering water at 110 deg C and a prestressed concrete vessel reactor delivering water at 200 deg C [fr

  19. The accident of Chernobylsk-4 reactor and its consequences

    International Nuclear Information System (INIS)

    1986-01-01

    This report deals with the particulars of the accident as communicated by the Soviet delegation at an IAEA meeting by the and of August 1986. It was stated that the consequences emanated from the inherent instability of the design of the reactor, the deviation from the safety rules by the operators and the lack of a sight reactor containment. (G.B.)

  20. Thermal hydraulic reactor safety analyses and experiments

    International Nuclear Information System (INIS)

    Holmstroem, H.; Eerikaeinen, L.; Kervinen, T.; Kilpi, K.; Mattila, L.; Miettinen, J.; Yrjoelae, V.

    1989-04-01

    The report introduces the results of the thermal hydraulic reactor safety research performed in the Nuclear Engineering Laboratory of the Technical Research Centre of Finland (VTT) during the years 1972-1987. Also practical applications i.e. analyses for the safety authorities and power companies are presented. The emphasis is on description of the state-of-the-art know how. The report describes VTT's most important computer codes, both those of foreign origin and those developed at VTT, and their assessment work, VTT's own experimental research, as well as international experimental projects and other forms of cooperation VTT has participated in. Appendix 8 contains a comprehensive list of the most important publications and technical reports produced. They present the content and results of the research in detail.(orig.)