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Sample records for babcock and wilcox lpr reactor

  1. Transient response of Babcock and Wilcox-designed reactors

    International Nuclear Information System (INIS)

    On February 26, 1980, the Crystal River Unit No. 3 Nuclear Generating Plant, designed by the Babcock and Wilcox Company (B and W), experienced an incident involving a malfunction in an instrumentation and control system power supply. Faced with the Crystal River Unit 3 incident and the apparently high frequency of such near similar types of transients in other B and W designed plants, a special Task Force was established within the Office of Nuclear Reactor Regulation to provide an assessment of the apparent sensitivity of the B and W designed plants to such transients and the consequences of malfunctions and failures of the integrated control system and non-nuclear instrumentation. This report provides an assessment of these issues

  2. Probabilistic analysis for the Babcock and Wilcox advanced light water reactor

    International Nuclear Information System (INIS)

    The Babcock and Wilcox (B and W) Advanced Light Water Reactor (ALWR) design employs design features that will provide enhanced safety, reliability, and design margin over the current generation of commercial nuclear power plants. This paper presents a probabilistic analysis performed to provide early feedback to the designers to enhance the reliability of these systems. Feedback from the probabilistic analysis was used to improve the system design by incorporating the insights gained. The calculated core melt frequency for the ALWR design was better than the design targets since most of the features that dominate the risk profile in conventional pressurized water reactors (PWRs) were eliminated in the redesign for the ALWR

  3. Experimental simulation of a small-scale Babcock and Wilcox reactor model: Final report

    International Nuclear Information System (INIS)

    This report documents the design, the instrumentation system, the data-acquisition system, and the testing of a small-scale, low pressure model of the cooling systems of a Babcock and Wilcox pressurized water reactor. This work is part of a larger program to address some of the safety issues in the B and W design. This test data are stored on data tapes; they are available to qualified requesters through EPRI. The primary use of these data is expected to be code verification and comparisons with results from other test facilities in the program

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

    International Nuclear Information System (INIS)

    The Standard Technical Specifications for Babcock and Wilcox Pressurized Water Reactors (BandW-STS) is a generic document prepared by the U.S. NRC for use in the licensing process. The BandW-STS provide applicants with model specifications to be used in formulation plant-specific technical specifications required by 10 CFR Part 50, Section 50.36, which set forth the specific characteristics of the facility and the conditions for its operation that are required to provide adequate protection to the health and safety of the public. This document is revised periodically to reflect current licensing requirements

  5. Babcock and Wilcox experiments interpretation

    International Nuclear Information System (INIS)

    This work consists of enlargement of critical experiments used for the pin power distribution calculation validation by the EDF industrial methodology. The Babcock and Wilcox critical experiments have measured the pin by pin radial power distribution in UO2 assemblies with and without glass or gadolinium absorber rods. The EDF scheme calculates two energy groups collapsed and homogenized neutron cross-sections and diffusion coefficients for the different pins types present in the mock-up, using the APOLLO2 cell code, based on the Pij collision probability modelling, fed with the 99 energy groups CEA93V6 data base library. These cross sections are then corrected by the HERMES transport-diffusion equivalence and used as entry data by the COCCINELLE core calculation code using finite difference method with one mesh for each calculation cell. The comparison between measured and calculated pin power values has confirmed the very satisfactory accuracy level of EDF industrial scheme for the treatment of assemblies without and with gadolinium pins. It exists a margin of improvement: the future calculation methodology currently under development will have the benefit of more accurate transport calculations for generating the two groups cross-sections used by the core diffusion code. (authors)

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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

  8. Standard Technical Specifications, Babcock and Wilcox Plants

    International Nuclear Information System (INIS)

    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

  9. Numerical simulation of natural circulation in a geometry simulating a Babcock and Wilcox type nuclear reactor

    International Nuclear Information System (INIS)

    In this paper, the authors present the results of numerical calculations for natural circulation in the facility called Once-Through Integral System (OTIS) Test Facility simulating a Babcock and Wilcox type nuclear reactor. The OTIS test facility was constructed to represent the main features of a Babcok and Wilcox raised loop plant. The computer code adopted for the study is RETRAN-02. A small break LOCA is simulated, and a number of important physical variables are calculated and compared with test data. These variables are temperature, pressure, void fraction, mass flow rate and liquid level in the steam generator secondary side. The analysis conducted indicates that the RETRAN-02 calculated response agrees reasonably well with the measured system response. Figure 1 shows cold leg fluid temperature during a two-phase natural circulation transient. Complex phenomena such as flow oscillations due to void generation are calculated well with RETRAN-02. Hot and cold fluid mixing near the HPI injection port is also well represented using RETRAN-02. The results do indicate, however, the need to account for piping heat losses to accurately represent the detailed phenomena occurring in the hot leg

  10. LEU silicide programs at Babcock and Wilcox

    International Nuclear Information System (INIS)

    The low enriched silicide development project at Babcock and Wilcox has matured into a production operation that has resulted in the completion of fuel elements for three research reactors; ORR, R-2 Studsvik and SAPHIR. Characteristic anomalies of silicide fuel which make the fabrication of fuel plates and elements more difficult than UAlx, have either been avoided, eliminated or significantly improved. One such anomaly is the reaction between uranium silicide fuel and aluminum matrix material. A detailed analysis was performed to characterize the extent of this reaction. Data suggests that a solid state diffusion of aluminum atoms into the uranium silicide lattice results in the formation of several intermediate Al-Si-U phases before forming a stable UAl4 phase

  11. Thermal-hydraulic research plan for Babcock and Wilcox plants

    International Nuclear Information System (INIS)

    This document presents a plan for thermal-hydraulic research for Babcock and Wilcox designed reactor systems. It describes the technical issues, regulatory needs, and the research necessary to address these needs. The plan also discusses the relationship between current and proposed research, and provides a tentative schedule to complete the required work

  12. Standard Technical Specifications, Babcock and Wilcox plants

    International Nuclear Information System (INIS)

    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 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 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. This document 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--3.10 of the improved STS

  13. Status of LEU programs at Babcock and Wilcox

    International Nuclear Information System (INIS)

    Within the Low Enriched Programs being conducted at Babcock and Wilcox the primary effort has been to establish, from past LEU development work and current production technology, an efficient production process that maintains product quality for both LEU UAlx and U3Si2 elements. This effort has allowed the Babcock and Wilcox Company to successfully complete a second LEU production contract for the 2-MW Ford Nuclear Reactor at the University of Michigan. Current U3Si2 contracts which include Standard and Control Elements for the Oak Ridge Reactor, SAPHIR Elements for the Swiss Federal Institute for Reactor Research, silicide (U3Si2) powder for the Danish Riso National Laboratory and Elements for Sweden's R2 Reactor at Studsvik are being manufactured under the same guidelines of quality and efficiency improvements. The transition from developmental work to a production process for powder fabrication; compacting; plate and element fabrication along with inspection methods are highlighted within this report. (author)

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

    International Nuclear Information System (INIS)

    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.)

  15. Shutdown decay heat removal analysis of a Babcock and Wilcox pressurized water reactor: Case study

    International Nuclear Information System (INIS)

    This is one of six case studies for USI A-45 Decay Heat Removal (DHR) Requirements. The purpose of this study is to identify any potential vulnerabilities in the DHR systems of a typical Babcock and Wilcox PWR, to suggest possible modifications to improve the DHR capability, and to assess the value and impact of the most promising alternatives to the existing DHR systems. The systems analysis considered small LOCAs and transient internal initiating events, and seismic, fire, extreme wind, internal and external flood, and lightning external events. A full-scale systems analysis was performed with detailed fault trees and event trees including support system dependencies. The system analysis results were extrapolated into release categories using applicable past PRA phenomenological results and improved containment failure mode probabilities. Public consequences were estimated using site specific CRAC2 calculations. The Value-Impact (VI) analysis of possible alternatives considered both onsite and offsite impacts arriving at several risk measures such as averted population dose out to a 50-mile radius and dollars per person rem averted. Uncertainties in the VI analysis are discussed and the issues of feed and bleed and secondary blowdown are analyzed

  16. APPLICATIONS ANALYSIS REPORT: BABCOCK AND WILCOX CYCLONE FURNACE

    Science.gov (United States)

    This document is an evaluation of the performance of the Babcock & Wilcox (B&W) Cyclone Furnace Vitrification Technology and its applicability as a treatment technique for soils contaminated with heavy metals, radionuclides, and organics. oth the technical and economic aspects of...

  17. TECHNOLOGY EVALUATION REPORT: BABCOCK AND WILCOX CYCLONE FURNACE VITRIFICATION TECHNOLOGY

    Science.gov (United States)

    The Babcock & Wilcox (B&W) Cyclone Furnace Vitrification Technology is a treatment process for contaminated soils. he process was evaluated to determine its ability to destroy semivolatile organics and to isolate metals and simulated radionuclides into a non-leachable slag materi...

  18. Assessment of thermal aging embrittlement of cast austenitic stainless steel components in the Babcock and Wilcox -designed PWR reactor internals

    International Nuclear Information System (INIS)

    The currently operating Babcock and Wilcox (BW) designed pressurized water reactors (PWRs) were constructed during the late sixties and seventies. Some of the reactor internals components were fabricated from cast austenitic stainless steel (CASS). The selection of CASS for the internals components was made to expedite the construction schedule by reducing machining and allowing production in large quantities. Since then, test data have shown that some CASS materials are susceptible to thermal aging embrittlement at PWR operating temperatures and its effect on functionality is of concern. Recently, the US nuclear power industry has developed inspection and evaluation guidelines (MRP-227, Rev.0) for managing aging degradation in PWR reactor internals for both the current and extended license periods. The MRP-227, Rev.0 guidelines recommend additional inspections for certain internals components including CASS components in BW PWRs due to thermal aging embrittlement concerns. The thermal aging embrittlement susceptibility for CASS can be assessed by the casting method and ferrite content if sufficient information in the original fabrication records is available. AREVA NP has performed a fabrication records search to identify several CASS components in the BW PWR internals and reviewed the archived fabrication records. A database has been assembled as a result of this records search. Based on the fabrication records, the ferrite content is determined using Hull's equivalent factors. Grade CF8 castings (without molybdenum) have been found to not be susceptible to thermal aging embrittlement. However, thermal aging embrittlement is a potential concern for Grade CF3M castings (containing 2 to 3% molybdenum). As a result of this assessment, several CASS components in the BW PWRs are concluded to not be susceptible to thermal aging embrittlement. The findings provide the basis for the removal of these CASS components from the additional inspection requirements in MRP-227

  19. Babcock and Wilcox comes up with a recipe for longevity

    International Nuclear Information System (INIS)

    With many nuclear power plants ten to twelve years old, there is a growing awareness of the desirability of extending their useful life. Babcock and Wilcox recently released details of its life extension strategy. The five-step approach to plant life extension is outlined. (U.K.)

  20. Status of LEU programs at Babcock and Wilcox

    International Nuclear Information System (INIS)

    The primary focus of Babcock and Wilcox's Research and Test Reactor Fuel Element Facility (B and W-RTRFE) is to continuously improve its fabrication and inspection processes in order to provide the highest quality product available. Beginning with fuel powder production and progressing through final element inspection, all operations are continuously reviewed for potential improvement. In addition, B and W provides significant corporate R and D funding to further test and improve critical operations, inspections, and equipment. This total commitment to quality and integrity has led to B and W's success as a premier fabricator of plate fuel assemblies. The results of these recent production and development activities are highlighted in this report. (author)

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

    International Nuclear Information System (INIS)

    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)

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

    Energy Technology Data Exchange (ETDEWEB)

    Smith, J.C. [Babcock and Wilcox Canada, Cambridge, Ontario (Canada)

    1998-07-01

    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)

  3. Quality assurance and inspection techniques in use at Babcock and Wilcox

    International Nuclear Information System (INIS)

    When Babcock and Wilcox reentered the aluminum research reactor fuel element business, most of the equipment, processes and techniques were provided by our customers. Some of the equipment, both manufacturing and inspection, has been in use since the beginning of the program and dates back some 20 years. Babcock and Wilcox has applied the expertise and technology gained from our naval fuel program in several areas of research, fabrication and inspection to update this equipment. Areas of improvement are capacitance non-contact gaging, min-clad gage evaluation and the future of real time x-ray systems. With production and inspection costs rising, Babcock and Wilcox has also initiated alternative possibilities for inspecting components at lower costs and increased precision

  4. Babcock and Wilcox plate fabrication experience with uranium silicide spherical fuel

    International Nuclear Information System (INIS)

    This report is written to present the fuel fabrication experience of Babcock and Wilcox using atomized spherical uranium silicide powder. The intent is to demonstrate the ability to fabricate fuel plates using spherical powder and to provide useful information proceeding into the next phase of work using this type of fuel. The limited quantity of resources- spherical powder and time, did not allow for much process optimizing in this work scope. However, the information contained within provides optimism for the future of spherical uranium silicide fuel plate fabrication at Babcock and Wilcox.The success of assembling fuel elements with spherical powder will enable Babcock and Wilcox to reduce overall costs to its customers while still maintaining our reputation for providing high quality research and test reactor products. (author)

  5. Babcock and Wilcox assessment of the Pratt and Whitney XNR2000

    Science.gov (United States)

    Westerman, Kurt O.; Scoles, Stephen W.; Jensen, R. R.; Rodes, J. R.; Ales, M. W.

    1993-01-01

    Babcock & Wilcox performed four subtasks related to the assessment of the Pratt & Whitney XNR2000 nuclear reactor as follows: (1) cermet fuel element fabricability assessment; (2) mechanical design review of the reactor system; (3) neutronic analysis review; and (4) safety assessment. The results of the mechanical and physics reviews have been integrated into the reactor design. The results of the fuel and safety assessments are presented.

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

    International Nuclear Information System (INIS)

    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

  7. Nuclear criticality safety for drums at Babcock and Wilcox

    International Nuclear Information System (INIS)

    The Babcock and Wilcox Company (B ampersand W) operates a nuclear fuel facility in Lynchburg, Virginia, processing uranium over the full range of possible enrichments (depleted to 97.65 wt% 235U). Nuclear fuel is produced for defense programs and for various research and test reactors worldwide. The facility has a uranium recovery operation that handles scrap produced at B ampersand W as well as scrap from other U.S. Department of Energy sites. B ampersand W also has a down-blending operation that is currently completing the down-blending of the fully enriched Project Sapphire Uranium to commercial-grade fuel (4 Wt% 235U). The facility generates approximately two hundred 55-gal drums of radioactive waste each month. Just a few years ago the number of waste drums on-site stood at ∼5000; however, through an aggressive waste reduction program, this number has been reduced to ∼2000. B ampersand W strives to avoid storing uranium scrap in 55-gal drums; however, there are approximately sixty-four 55-gal drums of scrap on-site. Scrap is that material from which the uranium is recovered because of financial, contractual, or regulatory considerations; waste is that material destined for disposal. Whether waste or scrap, nuclear criticality safety is of paramount concern in the handling, processing, and storing of uranium-bearing drums at B ampersand W. Any shipment complies with applicable U.S. Nuclear Regulatory Commission and U.S. Department of Transportation regulations

  8. Compact Process Development at Babcock and Wilcox

    International Nuclear Information System (INIS)

    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.

  9. Babcock and Wilcox STACSL integration options in ACSL

    International Nuclear Information System (INIS)

    The pertinent features of the Babcock and Wilcox ordinary differential equation solver STACSL which has been implemented in the ACSL Advanced Continuous Simulation Language are described. STACSL solves systems which have either a dense or a sparse Jacobian matrix. Root-finding techniques are incorporated in STACSL to efficiently solve problems with derivative discontinuities or other special events which must be detected and processed. Extensive diagnostics are also included in STACSL to assist in developing and debugging complex models. Each of these features is described and illustrated

  10. LWRWIMS analysis of Babcock and Wilcox LWR fuel storage experiments

    International Nuclear Information System (INIS)

    The report describes very briefly an analysis of a series of critical experiments made by Babcock and Wilcox to study the relative importance on fuel storage reactivity of assembly spacing and various types of absorber. LWRWIMS in its standard design mode of calculation was used for the analysis. The results demonstrate that even the simplest options in LWRWIMS produce eigenvalues which are a very useful check of the Monte Carlo calculations normally made for criticality clearances. An appendix examines some of the eigenvalue trends in more detail. (author)

  11. Safety Evaluation Report related to Babcock and Wilcox Owners Group Plant Reassessment Program

    International Nuclear Information System (INIS)

    Supplement 1 to the ''Safety Evaluation Report (SER) Related to the Babcock and Wilcox Owners Group (BWOG's) Plant Reassessment Program'' has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission (NRC). This supplement contains the NRC staff's evaluation of the BWOG reassessment of the integrated control system/non-nuclear instrumentation system, the emergency feedwater initiation and control system, reactor trip initiating events, several additional open items identified in the SER, and BWOG comments on the SER

  12. Design, operating and maintenance experience of Babcock and Wilcox nuclear steam generators

    International Nuclear Information System (INIS)

    Babcock and Wilcox (B and W) has designed and manufactured nuclear steam generators since the beginnings of the nuclear era in the 1950's. This paper describes how the B and W recirculating steam generator design evolved, the operating and maintenance history of the design, and the evolution of design and manufacturing methods into replacement steam generators for non-B and W reactors. (author)

  13. A probabilistic evaluation of the safety of Babcock and Wilcox nuclear reactor power plants with emphasis on historically observed operational events

    International Nuclear Information System (INIS)

    This report summarizes a study performed by Brookhaven National Laboratory for the Office of Nuclear Reactor Regulation, Division of Engineering and System Technology (A/D for Systems), US Nuclear Regulatory Commission. This study was requested by the NRC to assist their staff in assessing the risk significance of features of the Babcock and Wilcox (B and W) reactor plant design in the light of recent operational events. This study focuses on a critical review of submissions from the B and W Owners Group (BWOG) and as an independent assessment of the risk significance of ''Category C'' events at each operating B and W reactor. Category C events are those in which system conditions reach limits which require significant safety system and timely operator response to mitigate. A precursor study for each of the major B and W historical Category C events also was carried out. In addition, selected PRAs for B and W reactor plants and plants with other pressurized water reactor (PWR) designs were reviewed to appraise their handling of Category C events, thereby establishing a comparison between the risk profiles of B and W reactor plants and those of other PWR designs. The effectiveness of BWOG recommendations set forth in Appendix J of the BWOG SPIP (Safety and Performance Improvement Program) report (BAW-1919) also was evaluated. 49 refs., 21 figs., 52 tabs

  14. Babcock and Wilcox version of PDQ07: user's manual

    International Nuclear Information System (INIS)

    The Babcock and Wilcox version of PDQ07 solves the neutron diffusion depletion problem in one, two, and three dimensions and in up to five lethargy groups. Adjoint and boundary value calculations may also be performed. Geometries available are rectangular, cylindrical, spherical, and hexagonal. Special capabilities of the code include thermal-hydraulic feedback with subcooled boiling effects, boron iteration, rod bank placement, automatic partial rod movement, and flux synthesis. Time-independent group diffusion equations are solved by Gaussian elimination in one dimension, single-line cyclic Chebyshev semi-iterative technique in two dimensions, and a modified block Gauss-Siedel in three dimensions. Diffusion coefficients, macroscopic data, and depletion use a modified HARMONY system. Thermal feedback effects use an iterative approach based on relative power density in the core. Flux synthesis uses two-dimensional trial functions to solve three-dimensional problems

  15. Overview of Babcock and Wilcox involvement in the RERTR program

    International Nuclear Information System (INIS)

    The Nuclear Fuel Division (NFD) of the Babcock and Wilcox (BandW) Company is fully committed to the goals/objectives of the RERTR program. In support of this program, the NFD has fabricated and shipped two full size ORR elements of U3Si2. In addition, developmental work has been done with U3SiAl. This paper provides an overview of this manufacturing experience, discusses the facility modifications both for LEU and increased capacity, and briefly reviews manufacturing changes for LEU fuels. Overall, the fabrication of the ORR silicide elements proceeded smoothly. To better improve the efficiency, additional information is being gathered on crushing schedules, blending times, and dies. (author)

  16. Evaluation of operational safety at Babcock and Wilcox plants

    International Nuclear Information System (INIS)

    A methodology was developed to assess the operational performance of nuclear power plants through an integration of thermal-hydraulic, human factors, and risk analysis techniques. This methodology was applied to evaluate the effectiveness of plant systems and operator actions in lessening the severity of selected transients for Babcock and Wilcox (B and W) plants. Comparisons were also performed to assess differences in operational performance capabilities and limitations between selected Combustion Engineering, Westinghouse, and B and W plants. For the selected B and W plant, the results show the probability that an operating crew would not respond within the times available (the non-response probability) is estimated to be relatively small for the three transients studied. Results also show a strong correlation between operator performance and the influence of eight performance shaping factors (PSFs). Comparison of results from the Oconee, Calvert Cliffs, and H. B. Robinson plants indicate that the likelihood operators would take the proper actions to return Oconee to a safe stable state is not judged to be significantly different than the likelihood operators at Calvert Cliffs or H. B. Robinson would recover their plants for the transients investigated. The validity of this conclusion depends on the assumption that the performance shaping factors at all three plants are essentially equivalent. Recommendations are made that influence performance shaping factors positively and thereby influence operator performance positively

  17. Babcock ampersand Wilcox experience with alloy A-286 reactor vessel internal bolting

    International Nuclear Information System (INIS)

    Multiple reactor vessel internal bolt failures were discovered during the 1981 and 1982 in service inspections performed at three PWR nuclear power plants. All the failures were limited to bolts that fastened the lower portion of the reactor vessel internal thermal shield to the lower grid assembly. Subsequent examinations during 1982, 1983 and 1984 revealed bolt failures at four additional plants. These failures included bolts that fastened the core barrel to the core support shield and lower grid assembly. Additional failures were also discovered in the bolts used to join the surveillance specimen holder tube to the thermal shield. All the affected fasteners were fabricated from Alloy A-286 (ASTM A453 Grade 660) material. Alloy A-286 is a high strength precipitation hardened austenitic stainless steel containing a nominal Cr and Ni content of 15% and 25%, respectively. As a result of these bolt failures, the Babcock ampersand Wilcox Co., under the direction of the B ampersand W Owners Group, performed extensive evaluations of Alloy A-286 reactor vessel internal fasteners. The principal conclusions obtained from this investigation are given below. 1. Internals bolting failures have been observed at nominal peak calculated stress levels of greater than or equal to 690 MPa (100 ksi). The number of failures generally increases with increasing stress. Variations in this correlation are postulated to be the result of scatter in the calculated peak stress data. 2. A variety of material conditions including the use of highly cold worked barstock in the fabrication of some of the bolts, degree of annealing and hot forging may have contributed to the bolt failures. 3. No specific upset environmental conditions were found that could be judged to be a leading cause of the bolt failures. 4 refs., 2 tabs

  18. An assessment of RELAP5/MOD2 applicability to loss-of-feedwater transient analysis in a Babcock and Wilcox reactor plant

    International Nuclear Information System (INIS)

    The applicability and scaling capability of RELAP5/MOD2 when applied to a Babcock and Wilcox (B and W) loss-of-feedwater transient is assessed using a code applicability methodology. A loss-of-feedwater test with a feed-and-bleed recovery was selected from the once-through integral system (OTIS) test data as a reference transient. Nondimensional comparisons are made between code assessment calculations and code applications calculations using computer code models scaled according to scaling criteria derived from the work of Ishii and others. The results indicate that RELAP5/MOD2 can scale the phenomena observed in the experiment and that the code is applicable for transients for which phenomena are within this envelope. The results also demonstrate the usefulness of the code applicability methodology for interpreting and verifying code calculations. 21 refs., 59 figs., 12 tabs

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    This document presents information essential to understanding the risk associated with inter-system loss-of-coolant accidents (ISLOCAs). The methodology developed and presented in this document provides a state-of-the-art method for identifying and evaluating plant-specific hardware designs, human performance issues, and accident consequence factors 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 plant. 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

  1. TECHNOLOGY DEMONSTRATION SUMMARY. BABCOCK AND WILCOX CYCLONE FURNACE VITRIFICATION TECHNOLOGY (EPA/540/SR-92/017)

    Science.gov (United States)

    A Superfund Innovative Technology Evaluation (SITE) Demonstration of the Babcock & Wilcox Cyclone Furnace Vitrification Technology was conducted in November 1991. This Demonstration occurred at the Babcock & Wilcox (B&W) Alliance Research Center (ARC) in Alliance, OH. The B&W cyc...

  2. Radioactive waste shipments to Hanford retrievable storage from Babcock and Wilcox, Leechburg, Pennsylvania

    International Nuclear Information System (INIS)

    This report characterizes, as far as possible, the solid radioactive wastes generated by Babcock and Wilcox's Park Township Plutonium Facility near Leechburg, Pennsylvania that were sent to retrievable storage at the Hanford Site. Solid waste as defined in this document is any containerized or self-contained material that has been declared waste. The objective is a description of characteristics of solid wastes that are or will be managed by the Restoration and Upgrades Program; gaseous or liquid effluents are discussed only at a summary level This characterization is of particular interest in the planning of transuranic (TRU) waste retrieval operations, including the Waste Receiving and Processing (WRAP) Facility, because Babcock and Wilcox generated greater than 2.5 percent of the total volume of TRU waste currently stored at the Hanford Site

  3. Technology evaluation report: Babcock and Wilcox Cyclone Furnace Vitrification technology. Volume 2

    International Nuclear Information System (INIS)

    The Babcock and Wilcox (B and W) Cyclone Furnace Vitrification Technology is a treatment process for contaminated soils. The process was evaluated to determine its ability to destroy semivolatile organics and to isolate metals and simulated radionuclides into a non-leachable slag material. The feed material for the system was a prepared synthetic soil matrix (SSM) that was spiked with two organic compounds and six metals. This volume contains the appendices

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

    International Nuclear Information System (INIS)

    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

  5. Description of the Babcock and Wilcox owners group cavity dosimetry benchmark experiment

    International Nuclear Information System (INIS)

    The Babcock and Wilcox Owners Group (B and WOG) Cavity Dosimetry Benchmark experiment is the first step in the B and WOG program to develop measurement-based methodology for use in monitoring vessel fluence in the post-Reactor Vessel Surveillance Program timeframe. Ex-vessel dosimetry has been chosen as the vehicle to provide fluence measurements for use in this measurement-based methodology. (Fluence is measured indirectly by first measuring a relatable quantity and then applying the known correspondence between the measured quantity and the fluence, (e.g., Cs137 activity of a fission foil or tracks on an SSTR). The results of the In-Out Experiment will be used in refining the analytical models and benchmarking the final methodology. The experiment will provide neutron and gamma fluence measurements, at points both inside and outside the reactor vessel, through the use of numerous fluence measuring devices. Four different categories of ex-vessel monitors have been specified. The in-vessel fluence will be measured using an unirradiated, standard B and W reactor vessel surveillance capsule that will be installed in a spare holder tube at the same azimuthal position as the main ex-vessel dosimetry stringer. This paper presents a detailed description of the experiment

  6. Technology evaluation report: Babcock and Wilcox Cyclone Furnace Vitrification technology. Volume 1

    International Nuclear Information System (INIS)

    The project consists of an analysis of the Babcock and Wilcox (B and W) Cyclone Furnace Vitrification process. The SITE Demonstration took place at the B and W Research and Development Division in Alliance, Ohio. The vitrification process was performed on a synthetic soil matrix (SSM) that was spiked with known concentrations of semivolatile organic compounds, metals, and simulated radionuclides. The Demonstration effort was directed at obtaining information on the performance and cost of the process for use at other sites. Documentation will consist of two reports. This Technology Evaluation Report (TER) is contained in two volumes and describes the field activities and laboratory results

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

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

    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 109) 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

  9. Evaluation of operational safety at Babcock and Wilcox Plants: Volume 1, Results overview

    International Nuclear Information System (INIS)

    A methodology was developed to assess the operational performance of nuclear power plants through an integration of thermal-hydraulic and human factors analysis techniques together with inputs from information used in the assessment of risk. This methodology was applied to evaluate the extent to which plant systems and/or operator actions are effective in lessening the severity of selected transients for Babcock and Wilcox (B and W) plants. Comparisons were also performed to assess differences in operational performance capabilities and limitations between selected Combustion Engineering, Westinghouse, and B and W plants. Detailed results from the methodology application are presented in two volumes. This report Volume 1, presents an overview of the results with emphasis on the systems and operator performance. Volume 2 presents detailed results from thermal-hydraulic calculations. 22 refs., 9 figs., 16 tabs

  10. High density LEU [low enriched uranium] fuel development at Babcock and Wilcox

    International Nuclear Information System (INIS)

    An aggressive pursuit of developing a high-density LEU fuel process has been undertaken over the past six years at the Babcock and Wilcox Co. A major effort has been devoted to the U3Si2 fuel development. Today B and W feels confident that their current U3Si2 manufacturing process is comparable to existing U3O8 and UAlx fuel technologies. A continued effort will be maintained within the U3Si2 product line to provide the highest product quality and to increased process efficiencies. Investigations into other high density LEU fuel development such as U(x)Si(y) alloys will only be secondary considerations. (Author)

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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

  13. Aging assessment of the Combustion Engineering and Babcock and Wilcox control rod drives

    International Nuclear Information System (INIS)

    The effects of aging upon the Babcock ampersand Wilcox (B ampersand W) and Combustion Engineering (CE) control rod drive systems have been evaluated. For this study, the CRD system boundary included the control rod assemblies, guide tubes, control rod drive mechanism, control system components, rod position indication components, and cooling system. Detailed operation experience data for 1980 to 1990 was evaluated to identify the predominant failure modes, causes, and effects. The results of this evaluation, along with an assessment of component material and operating environment, lead to the conclusion that both the B ampersand W and CE CRD systems are susceptible to age degradation. Failures of the CRD system have resulted in significant plant effects including power reductions, plant shutdowns, scrams, and ESF actuations. Information on current plant system inspection and maintenance practices were obtained from two B ampersand W plants, and four CE plants through an industry survey. The results of this survey indicate that some plants have modified the system, replaced components, and established preventive maintenance programs, some of which effectively address the aging issue, while others do not. The potential application of some advanced monitoring inspection techniques are discussed

  14. An aerial radiological survey of the Babcock and Wilcox Nuclear Facilities and surrounding area, Lynchburg, Virginia

    International Nuclear Information System (INIS)

    An aerial radiological survey was conducted from July 18 through July 25, 1988, over a 41-square-kilometer (16-square-mile) area surrounding the Babcock and Wilcox nuclear facilities located near Lynchburg, Virginia. The survey was conducted at a nominal altitude of 61 meters (200 feet) with line spacings of 91 meters (300 feet). A contour map of the terrestrial gamma exposure rate extrapolated to 1 meter above ground level (AGL) was prepared and overlaid on an aerial photograph. The terrestrial exposure rates varied from 8 to 12 microroentgens per hour (μR/h). A search of the data for man-made radiation sources revealed the presence of three areas of high count rates in the survey area. Spectra accumulated over the main plant showed the presence of cobalt-60 (60Co) and cesium-137 (137Cs). A second area near the main plant indicated the presence of uranium-235 (235U). Protactinium-234m (234mPa) and 60Co Were detected over a building to the east of the main plant. Soil samples and pressurized ion chamber measurements were obtained at four locations within the survey boundaries in support of the aerial data

  15. IE Information Notice No. 86-04: Transient due to loss of power to integrated control system at a pressurized water reactor designed by Babcock and Wilcox

    International Nuclear Information System (INIS)

    On December 26, 1985, Rancho Seco was operating on automatic control at a constant power level of 710 MWe (76% of licensed power). At 4:14 a.m., power to the integrated control system (ICS) was lost. The annunciator alarm for ''Loss of ICS or Fan Power'' sounded. As designed, ICS demand signals went to midscale. The main feedwater valves closed to 50%, and the atmospheric dump valves, turbine bypass valves, and one set of auxiliary feedwater valves opened to 50%. The main feedwater pump speed was reduced to minimum. Low discharge pressure at the main feedwater pump caused the motor-driven auxiliary feedwater pump to start automatically. The net decrease in feedwater flow caused the reactor to trip on high reactor coolant system (RCS) pressure. After the reactor trip, the above ICS valves remained at 50% (i.e., could not be operated from the control room) causing excessive cooling of the RCS which was exacerbated by autostarting of the dual-drive auxiliary feedwater pump. During the 26 minutes required to restore ICS power, operators acted to minimize the resulting transient. However, difficulties were experienced with manipulation of valves, operation of pumps, and control of various liquid levels, pressures, and temperatures. RCS pressure decreased to a minimum of 1,064 psig at 4:21 a.m. At 4:40 a.m., the lowest RCS temperature (386 F) during the cooling transient was reached. RCS pressure at that time was 1,413 psig. Eventually, a senior reactor operator discovered that switches which supplied power to the ICS dc power supplies were in the off position and set them to the on position

  16. 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...

  17. DEMONSTRATION BULLETIN: CYCLONE FURNACE SOIL VITRI- FICATION TECHNOLOGY - BABCOCK & WILCOX

    Science.gov (United States)

    Babcock and Wilcox's (B&W) cyclone furnace is an innovative thermal technology which may offer advantages in treating soils containing organics, heavy metals, and/or radionuclide contaminants. The furnace used in the SITE demonstration was a 4- to 6-million Btu/hr pilot system....

  18. SITE EMERGING TECHNOLOGIES PROJECT: BABCOCK & WILCOX CYCLONE VITRIFICATION

    Science.gov (United States)

    The Babcock & Wilcox 6 million Btu/hr pilot cyclone furnace was successfully used in a 2-year 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 1,500 ppm chromium. n advantage of vitrificatio...

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    This NUREG contains the improved Standard Technical Specifications (STS) for Babcock and Wilcox (B ampersand 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 ampersand 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

  2. Steam generator channel head dose rates at Babcock ampersand Wilcox reactors

    International Nuclear Information System (INIS)

    This report describes the results of a program to collect historical OTSG dose rate data from the five (5) B ampersand W operating plants. Data is presented for Arkansas Nuclear One, Crystal River, Davis-Besse, Oconee, and Three Mile Island. Also included are dose reduction measures employed at each site

  3. Nuclear criticality safety at Babcock ampersand Wilcox Company

    International Nuclear Information System (INIS)

    The Babcock ampersand Wilcox Company (B ampersand W) operates a nuclear fuel production plant in Virginia. It is a privately owned facility licensed by the U.S. Nuclear Regulatory Commission (NRC). The NRC maintains a resident inspector on-site. The plant produces highly enriched fuel for both certain defense programs and the various U.S. research and test reactors. The plant also produces nuclear fuel at an intermediate enrichment (20 wt%) for research and test reactors in the United States and overseas. B ampersand W operates a highly enriched uranium recovery operation for its scrap and as a service to various U.S. Department of Energy sites. B ampersand W's downblending operations are designed to produce low-enriched fuel (5 wt%); the company is currently under contract to clean up and downblend Sapphire material. Operations within the facility include ceramic (oxides, silicide, and carbides), foundry (metal), chemical (nitrates, ADUN, etc.), and mechanical assembly with extensive laboratory and quality assurance operations. Also located on-site is a hot cell facility for the examination of irradiated fuel. This report discusses B ampersand W's license renewal considerations

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

    International Nuclear Information System (INIS)

    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

  5. Modeling operator actions during a small break loss-of-coolant accident in a Babcock and Wilcox nuclear power plant

    International Nuclear Information System (INIS)

    A small break loss-of-accident (SBLOCA) in a typical Babcock and Wilcox (B ampersand W) nuclear power plant was modeled using RELAP5/MOD3. This work was performed as part of the United States Regulatory Commission's (USNRC) Code, Scaling, Applicability and Uncertainty (CSAU) study. The break was initiated by severing one high pressure injection (HPI) line at the cold leg. Thus, the small break was further aggravated by reduced HPI flow. Comparisons between scoping runs with minimal operator action, and full operator action, clearly showed that the operator plays a key role in recovering the plant. Operator actions were modeled based on the emergency operating procedures (EOPs) and the Technical Bases Document for the EOPs. The sequence of operator actions modeled here is only one of several possibilities. Different sequences of operator actions are possible for a given accident because of the subjective decisions the operator must make when determining the status of the plant, hence, which branch of the EOP to follow. To assess the credibility of the modeled operator actions, these actions and results of the simulated accident scenario were presented to operator examiners who are familiar with B ampersand W nuclear power plants. They agreed that, in general, the modeled operator actions conform to the requirements set forth in the EOPs and are therefore plausible. This paper presents the method for modeling the operator actions and discusses the simulated accident scenario from the viewpoint of operator actions

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

    International Nuclear Information System (INIS)

    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 UO2 pellets enclosed in zircaloy tubes

  7. Effects of natural phenomena on the Babcock and Wilcox Co. Plutonium Fabrication Plant at the Parks Township site, Leechburg, Pennsylvania. Docket No. 70-364

    International Nuclear Information System (INIS)

    The proposed action is to issue a renewal to the full-term Special Nuclear Material License No. SNM-414 (Docket No. 70-364) authorizing the Nuclear Material Division of the Babcock and Wilcox Company (BandW) to operate nuclear-fuel-fabrication facilities located in Leechburg, Pennsylvania. The plutonium fuel facility is presently being used to fabricate fuel for the fast test reactor under construction at the Hanford Reservation near Richland, Washington. Implicit in Sections 70.22 and 70.23 of 10CFR70 is a requirement that existing plutonium fabrication plants be examined with the objective of improving, to the extent practicable, their abilities to withstand adverse natural phenomena without loss of capability to protect the public. In accordance with these regulations, an analysis was initiated of the effects of natural phenomena on the BandW Plutonium Fabrication Plant. Following completion of the analysis, a condensation was prepared of the effects of natural phenomena on the facility

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

    International Nuclear Information System (INIS)

    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

  9. 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.

  10. LOCA pipe break criteria for the design of Babcock and Wilcox nuclear steam systems

    International Nuclear Information System (INIS)

    The document describes the criteria applied by B and W to determine design basis break locations, types of breaks, and break sizes in the primary piping system. Appendixes are provided in support of the basic assumptions made in the development of the criteria

  11. 75 FR 50009 - Babcock & Wilcox Nuclear Operations Group, Inc.; Establishment of Atomic Safety and Licensing Board

    Science.gov (United States)

    2010-08-16

    ... Board Pursuant to delegation by the Commission dated December 29, 1972 (37 FR 28710), and the Commission...., on February 23, 2010. Pursuant to a Request for Hearing published in the Federal Register (74 FR 75... (72 FR 49139). Issued at Rockville, Maryland, this 6th day of August 2010. E. Roy Hawkens,...

  12. Scaled experiments for support of code modeling of main steam-line break phenomena in a B ampersand W [Babcock and Wilcox]-type once-through steam generator

    International Nuclear Information System (INIS)

    This paper describes and aspect of ongoing research to provide information on the performance of once-through steam generators (OTSGs) commonly used in the Babcock and Wilcox (B ampersand W) nuclear steam supply system. This program is funded by the U.S. Nuclear Regulatory Commission and is being conducted in part at the U.S. Department of Energy's Idaho National Engineering Laboratory. The objectives of the program are to develop an experimental data base that can be used to assess existing models and to develop improved models for characterizing the behavior of an OTSG during various off-normal operating conditions and accident scenarios. The models are then implemented in the nuclear reactor safety codes RELAP5 and TRAC

  13. Benchmarking of flowtran with Mark-22 mockup flow excursion test data from Babcock ampersand Wilcox

    International Nuclear Information System (INIS)

    Version 16.2 of the FLOWTRAN code with a Savannah River Site (SRS) working criterion (St=0.00455) for the onset of significant void (OSV) was benchmarked against power and flow excursion data derived from tests at the Babcock ampersand Wilcox Alliance Research Center test facility. This document presents analyses which show that FLOWTRAN accurately predicts the mockup test assembly thermal-hydraulic behavior during the steady state and LOCA transient conditions, and that FLOWTRAN with a Savannah River Site (SRS) working limits criterion (St=0.00455) conservatively predicts the OFI power

  14. 75 FR 35846 - In the Matter of Babcock & Wilcox Nuclear Operations Group, Inc., Lynchburg, VA; Order Imposing...

    Science.gov (United States)

    2010-06-23

    ... under 10 CFR 2.315(c), must be filed in accordance with the NRC E-Filing rule (72 FR 49139, August 28... COMMISSION In the Matter of Babcock & Wilcox Nuclear Operations Group, Inc., Lynchburg, VA; Order Imposing Civil Monetary Penalty I Babcock & Wilcox Nuclear Operations Group, Inc., (Licensee) is the holder...

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

    International Nuclear Information System (INIS)

    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

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

    Energy Technology Data Exchange (ETDEWEB)

    None

    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.

  17. GERD (Gastroesophageal Reflux) and LPR (Laryngopharyngeal Reflux)

    Science.gov (United States)

    ... Meeting Calendar Find an ENT Doctor Near You GERD and LPR GERD and LPR Patient Health Information ... relations staff at newsroom@entnet.org . What is GERD? Gastroesophageal reflux disease, often referred to as GERD, ...

  18. 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)

    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.)

  19. Replacement steam generators for pressurized water reactors

    International Nuclear Information System (INIS)

    Babcock and Wilcox Canada has developed an Advanced Series steam generator for PWR Systems. This design incorporates all of the features that have contributed to the successful CANDU steam generator performance. This paper presents an overview of the design features and how the overall design relates to the requirements of a PWR reactor system

  20. Reactor internals design/analysis for normal, upset, and faulted conditions

    International Nuclear Information System (INIS)

    The analytical procedures used by Babcock and Wilcox to demonstrate the structural integrity of the 205-FA reactor internals are described. Analytical results are presented and compared to ASME Code allowable limits for Normal, Upset, and Faulted conditions. The particular faulted condition considered is a simultaneous loss-of-coolant accident and safe shutdown earthquake. The operating basis earthquake is addressed as an Upset condition

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

    International Nuclear Information System (INIS)

    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. 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

  3. The interactive effect of MAOA-LPR genotype and childhood physical neglect on aggressive behaviors in Italian male prisoners

    OpenAIRE

    Gorodetsky, Elena; Bevilacqua, Laura; Carli, Vladimir; Sarchiapone, Marco; Roy, Alec; Goldman, David; Enoch, Mary-Anne

    2014-01-01

    Aggressive disorders are moderately heritable; therefore, identification of genetic influences is important. The X-linked MAOA gene, encoding the MAOA enzyme, has a functional 30bp repeat polymorphism in the promoter region (MAOA-LPR) that has been shown to influence aggression. Childhood trauma is a known risk factor for numerous psychopathologies in adulthood including aggressive behaviors. We investigated the interactive effect of MAOA-LPR genotype and a history of childhood trauma in pred...

  4. Babcock-Leighton solar dynamo: the role of downward pumping and the equatorward propagation of activity

    OpenAIRE

    Karak, Bidya Binay; Cameron, Robert

    2016-01-01

    The key elements of the Babcock-Leighton dynamo are the generation of poloidal field through the decay of tilted bipolar active regions and the generation of toroidal field through the observed differential rotation. There are two classes of Babcock-Leighton models: flux transport dynamos where an equatorward flow at the bottom of the convection zone (CZ) causes the equatorial propagation of the butterfly wings, and dynamo waves where the radial shear and the $\\alpha$ effect act in conjunctio...

  5. 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.

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

    International Nuclear Information System (INIS)

    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)

  7. Mycobacterial Metabolic Syndrome: LprG and Rv1410 Regulate Triacylglyceride Levels, Growth Rate and Virulence in Mycobacterium tuberculosis.

    Directory of Open Access Journals (Sweden)

    Amanda J Martinot

    2016-01-01

    Full Text Available Mycobacterium tuberculosis (Mtb mutants lacking rv1411c, which encodes the lipoprotein LprG, and rv1410c, which encodes a putative efflux pump, are dramatically attenuated for growth in mice. Here we show that loss of LprG-Rv1410 in Mtb leads to intracellular triacylglyceride (TAG accumulation, and overexpression of the locus increases the levels of TAG in the culture medium, demonstrating a role of this locus in TAG transport. LprG binds TAG within a large hydrophobic cleft and is sufficient to transfer TAG from donor to acceptor membranes. Further, LprG-Rv1410 is critical for broadly regulating bacterial growth and metabolism in vitro during carbon restriction and in vivo during infection of mice. The growth defect in mice is due to disrupted bacterial metabolism and occurs independently of key immune regulators. The in vivo essentiality of this locus suggests that this export system and other regulators of metabolism should be considered as targets for novel therapeutics.

  8. Mycobacterial Metabolic Syndrome: LprG and Rv1410 Regulate Triacylglyceride Levels, Growth Rate and Virulence in Mycobacterium tuberculosis.

    Science.gov (United States)

    Martinot, Amanda J; Farrow, Mary; Bai, Lu; Layre, Emilie; Cheng, Tan-Yun; Tsai, Jennifer H; Iqbal, Jahangir; Annand, John W; Sullivan, Zuri A; Hussain, M Mahmood; Sacchettini, James; Moody, D Branch; Seeliger, Jessica C; Rubin, Eric J

    2016-01-01

    Mycobacterium tuberculosis (Mtb) mutants lacking rv1411c, which encodes the lipoprotein LprG, and rv1410c, which encodes a putative efflux pump, are dramatically attenuated for growth in mice. Here we show that loss of LprG-Rv1410 in Mtb leads to intracellular triacylglyceride (TAG) accumulation, and overexpression of the locus increases the levels of TAG in the culture medium, demonstrating a role of this locus in TAG transport. LprG binds TAG within a large hydrophobic cleft and is sufficient to transfer TAG from donor to acceptor membranes. Further, LprG-Rv1410 is critical for broadly regulating bacterial growth and metabolism in vitro during carbon restriction and in vivo during infection of mice. The growth defect in mice is due to disrupted bacterial metabolism and occurs independently of key immune regulators. The in vivo essentiality of this locus suggests that this export system and other regulators of metabolism should be considered as targets for novel therapeutics. PMID:26751071

  9. 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)

    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

  10. A functional polymorphism in the MAOA gene promoter (MAOA-LPR) predicts central dopamine function and body mass index.

    Science.gov (United States)

    Ducci, F; Newman, T K; Funt, S; Brown, G L; Virkkunen, M; Goldman, D

    2006-09-01

    Variation in brain monoaminergic activity is heritable and modulates risk of alcoholism and other addictions, as well as food intake and energy expenditure. Monoamine oxidase A deaminates the monoamine neurotransmitters serotonin, dopamine (DA), and noradrenaline. The monoamine oxidase A (MAOA) gene (Xp11.5) contains a length polymorphism in its promoter region (MAOA-LPR) that putatively affects transcriptional efficiency. Our goals were to test (1) whether MAOA-LPR contributes to interindividual variation in monoamine activity, assessed using levels of cerebrospinal fluid (CSF) monoamine metabolites; and (2) whether MAOA-LPR genotype influences alcoholism and/or body mass index (BMI). Male, unrelated criminal alcoholics (N=278) and controls (N=227) were collected from a homogeneous Finnish source population. CSF concentration of 5-hydroxyindoleacetic acid (5-HIAA), homovanillic acid (HVA), and 3-methoxy-4-hydroxyphenylglycol (MHPG) were available from 208 participants. Single allele, hemizygous genotypes were grouped according to inferred effect of the MAOA alleles on transcriptional activity. MAOA-LPR genotypes had a significant effect on CSF HVA concentration (P=0.01) but explained only 3% of the total variance. There was a detectable but nonsignificant genotype effect on 5-HIAA and no effects on MHPG. Specifically, the genotype conferring high MAOA activity was associated with lower HVA levels in both alcoholics and controls, a finding that persisted after accounting for the potential confounds of alcoholism, BMI, height, and smoking. MAOA-LPR genotype predicted BMI (P<0.005), with the high-activity genotype being associated with lower BMI. MAOA-LPR genotypes were not associated with alcoholism or related psychiatric phenotypes in this data set. Our results suggest that MAOA-LPR allelic variation modulates DA turnover in the CNS, but does so in a manner contrary to our prior expectation that alleles conferring high activity would predict higher HVA levels in

  11. Defect in negative selection in lpr donor-derived T cells differentiating in non-lpr host thymus

    International Nuclear Information System (INIS)

    Transplantation of bone marrow cells of lpr/lpr mice into irradiated normal mice fails to develop massive lymphadenopathy or autoimmunity but causes severe graft-vs.-host-like syndrome. To elucidate an abnormality of lpr/lpr bone marrow-derived T cells, we transplanted bone marrow cells of Mlsb lpr/lpr mice into H-2-compatible Mlsa non-lpr mice. Although lpr/lpr T cell precursors repopulated the host thymus as well as +/+ cells, a proportion of CD4+CD8+ cells decreased, and that of both CD4- and CD8- single-positive cells increased compared with those of +/+ recipients. Notably, in MRL/lpr----AKR and C3H/lpr----AKR chimeras, CD4 single-positive thymocytes contained an increased number of V beta 6+ cells in spite of potentially deleting alleles of Mlsa, whereas V beta 6+ mature T cells were deleted in the MRL/+ ----AKR and C3H/+ ----AKR chimeras. There was no difference between MRL/+ ----AKR and MRL/lpr----AKR chimeras in their proportion of V beta 3+ cells because both host and donor strain lack the deleting alleles. Interleukin 2 receptor expression of mature T cells, in the thymus and lymph node, was obviously higher in the MRL/lpr----AKR chimeras, in particular in the forbidden V beta 6+ subset. Moreover, lpr donor-derived peripheral T cells showed vigorous anti-CD3 response. These results indicate that lpr-derived T cells escape not only tolerance-related clonal deletion but also some induction of unresponsiveness in the non-lpr thymus

  12. Research reactor and fuel development/production facility decommissioning technology and experience

    International Nuclear Information System (INIS)

    This paper discusses the technology and experience gained in a series of reactor and fuels development facility decommissioning programs carried out by Babcock and Wilcox (B and W) at its US Nuclear Regulatory Commission (NRC)-licensed sites in Lynchburg, Virginia. Areas of generic application to future projects are particularly emphasized. The projects included one test and one research reactor, four low-power critical experiment facilities, and two buildings that housed plutonium/uranium fuels development laboratories. These projects were comprehensive; they included developing the decommissioning and quality assurance plans, interfacing with the NRC, performing the actual decontamination/dismantling work, performing predecontamination and final radiological surveys, and volume reducing, packaging, certifying, classifying and shipping the radioactive waste for disposal

  13. Development of the steam generator by Babcock Atlantic and Stein Industries, for the super Phenix Project

    International Nuclear Information System (INIS)

    The development program of steam generators studied by Babcock Atlantic and Stein Industries Companies, jointly with CEA and EDF, for the Super Phenix 1200 MWe Fast Breeder Power Plant is presented. The main characteristics of both sodium heated steam generators are emphasized and the experimental studies related to their key features are reported

  14. Reactor protection system

    International Nuclear Information System (INIS)

    The report describes the reactor protection system (RPS-II) designed for use on Babcock and Wilcox 145-, later 177-, and 205-fuel assembly pressurized water reactors. In this system, relays in the trip logic have been replaced by solid state devices. A calculating module for the low DNBR, pump status, and offset trip functions has replaced the overpower trip (based on flow and imbalance), the power/RC pump trip, and the variable low-pressure trip. Included is a description of the changes from the present Oconee-type reactor protection system (RPS-I), a functional and hardware description of the calculating module, a description of the software programmed in the calculating module, and a discussion of the qualification program conducted to ensure that the degree of protection provided by RPS-II is not less than that provided by previously licensed systems supplied by B and W

  15. Studying the possibilities of the IRT-2000 Sophia research reactor reconstruction into a low-power reactor

    International Nuclear Information System (INIS)

    The work is aimed at substantiating the possibility of the IRT-2000-Sofia research reactor reconstruction into the low-power reactor (LPR). The IRT-2000 reactor was shutdown on July 13, 1989 due to the safety-related causes. The following aspects of this reconstruction are discussed: namely, the LPR possibilities; the possibility of utilizing the IRT-2000 reactor site, fuel, buildings and equipment; auxiliary technological systems and mechanical equipment; measures for providing the LPR safety

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

    International Nuclear Information System (INIS)

    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.)

  17. Low enriched aluminide and silicide fuel element technology at B and W (USA)

    International Nuclear Information System (INIS)

    Babcock and Wilcox is fabricating full size fuel elements with low enriched uranium silicide and uranium aluminide. BandW also provides high enrichred U3O8 and UA Lsub(x) for United States Research Reactors, and Test Research and Training Reactors (TRTR). BandW and Argonne National Laboratry (ANL) are actively involved in the Reduced Enrichment Research and Test Reactor (RERTR) Program and have undertaken a joint effort in which BandW is fabricating two Oak Ridge Reactor (ORR ) elements with uranium silicide fuel. During plate development, fuel plates were fabricated with compacts containing U3SiAl and U3Si2 fuel. (author)

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

    International Nuclear Information System (INIS)

    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

  19. Preliminary study of uranium favorability of the Wilcox and Claiborne Groups (Eocene) in Texas

    Energy Technology Data Exchange (ETDEWEB)

    Wilbert, W.P.; Templain, C.J.

    1978-01-01

    Rocks of the Wilcox and Claiborne Groups crop out in the Texas Gulf Coastal Plain and are represented by a series of sands and shales which reflect oscillation of the strandline. The Wilcox Group (lower Eocene), usually undifferentiated in Texas, consists of very fine sands and clays and abundant lignite. The Claiborne Group (middle Eocene) comprises, in ascending order, Carrizo Sand, Reklaw Formation (clay), Queen City Sand, Weches Formation (clay), Sparta Sand, Cook Mountain Formation (clay), and Yegua Formation (sand). Fluvial systems of the Wilcox and Claiborne Groups exist in east Texas and trend perpendicular to the present coastline. In central Texas, sand bodies are parallel to the present coastline and are strand-plain, barrier-bar systems. Since the time of deposition of the Queen City Sand, a significant fluvial sand buildup occurred in the area of the present Rio Grande embayment where the marine clays pinch out. Known occurrences of mineral matter in the Wilcox and Claiborne (up to the Yegua) are limited to lignite (particularly in the Wilcox), cannel coal in the upper Claiborne, and hydrocarbons throughout. No uranium mineralization is known, and no uranium is likely to be discovered in the Claiborne and Wilcox. Approximately 50 surface samples and many gamma-ray logs showed no significant anomalies. The sands are very good potential host rocks, but no uranium source was discovered. During deposition of the Wilcox and Claiborne Groups, there was no volcanism to serve as a source of uranium (as with the prolific occurrences in the younger rocks of south Texas); also, Precambrian crystalline rocks in the Llano uplift were not exposed.

  20. Babcock-Leighton solar dynamo: the role of downward pumping and the equatorward propagation of activity

    CERN Document Server

    Karak, Bidya Binay

    2016-01-01

    The key elements of the Babcock-Leighton dynamo are the generation of poloidal field through the decay of tilted bipolar active regions and the generation of toroidal field through the observed differential rotation. There are two classes of Babcock-Leighton models: flux transport dynamos where an equatorward flow at the bottom of the convection zone (CZ) is responsible for the equatorial propagation of the butterfly wings, and dynamo waves where the radial gradient of differential rotation and the $\\alpha$ effect act in conjunction to produce the equatorial propagation. Here we investigate the role of downward magnetic pumping near the surface using a kinematic Babcock-Leighton model. We find that the pumping causes the poloidal field to become predominately radial in the near-surface shear layer. This allows the negative radial shear in the near-surface layer to effectively act on the radial field to produce a toroidal field. Consequently, we observe a clear equatorward migration of the toroidal field at lo...

  1. White Paper on the Use of Team Calendars with the JIRA Issue Tracking System and Confluence Collaboration Tools for the xLPR Project

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-09-01

    ORNL was tasked by xLPR project management to propose a team calendar for use within the xLPR consortium. Among various options that were considered, the approach judged by ORNL to best fit the needs of the xLPR project is presented in this document. The Atlassian Team Calendars plug-in used with the Confluence collaboration tool was recommended for several reasons, including the advantage that it provides for a tight integration between Confluence (found at https://xlpr.ornl.gov/wiki ) and xLPR s JIRA issue tracking system (found at https://xlpr.ornl.gov/jira ). This document is divided into two parts. The first part (Sections 1-6) consists of the white paper, which highlights some of the ways that Team Calendars can improve com mun ication between xLPR project managers, group leads, and team members when JIRA is applied for both issue tracking and change-management activities. Specific points emphasized herein are as follows: The Team Calendar application greatly enhances the added value that the JIRA and Confluence tools bring to the xLPR Project. The Team Calendar can improve com mun ication between xLPR project managers, group leads, and team members when JIRA is applied for both issue tracking and change-management activities. The Team Calendar works across different email tools such as Outlook 2011, Outlook 2010, Outlook 2007, Google Calendars and Mac s iCalendar to name a few. xLPR users can now access the wiki Confluence (with embedded Team Calendars) directly from JIRA without having to re-validate their login. The second part consists of an Annex (Section 7), which describes how users can subscribe to Team Calendars from different calendar applications. Specific instructions are given in the Annex that describe how to Import xLPR Team Calendar to Outlook Version Office 2010 Import xLPR Team Calendar to Outlook Version Office 2007 Subscribe to Team Calendar from Google Calendar The reader is directed to Section 4 for instructions on adding events to the

  2. 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)

    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

  3. Magnetic flux transport and the sun's dipole moment - New twists to the Babcock-Leighton model

    Science.gov (United States)

    Wang, Y.-M.; Sheeley, N. R., Jr.

    1991-01-01

    The mechanisms that give rise to the sun's large-scale poloidal magnetic field are explored in the framework of the Babcock-Leighton (BL) model. It is shown that there are in general two quite distinct contributions to the generation of the 'alpha effect': the first is associated with the axial tilts of the bipolar magnetic regions as they erupt at the surface, while the second arises through the interaction between diffusion and flow as the magnetic flux is dispersed over the surface. The general relationship between flux transport and the BL dynamo is discussed.

  4. WILCOX COUNTY, ALABAMA--A STUDY OF SOCIAL, ECONOMIC, AND EDUCATIONAL BANKRUPTCY. REPORT OF AN INVESTIGATION.

    Science.gov (United States)

    BROADUS, JAMES; AND OTHERS

    THE REQUEST FOR THIS INVESTIGATION BY THE SPECIAL COMMITTEE OF THE NATIONAL EDUCATION ASSOCIATION COMMISSION ON PROFESSIONAL RIGHTS AND RESPONSIBILITIES RESULTED FROM THE FIRING OF NINE NEGRO TEACHERS IN WILCOX COUNTY. THE STUDY ITSELF IS MORE INCLUSIVE, INCORPORATING THE FINDINGS AND CONCLUSIONS OF SEPARATE STUDIES IN POVERTY, SCHOOL FINANCE,…

  5. Wilcox Group Apparent Thickness, Gulf Coast (wlcxthkg)

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — The Apparent Wilcox Group thickness maps are contoured from location and top information derived from the Petroleum Information (PI) Wells database. The Wilcox...

  6. 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.

  7. Research reactor and fuel development facility decommissioning experience and technology

    International Nuclear Information System (INIS)

    This paper discusses the technology and experience gained in research reactor and fuels development facility decommissioning programs carried out by Babcock and Wilcox (B and W) at one of its NRC-licensed sites in Lynchburg, VA. The projects included two buildings that housed plutonium/uranium fuels development laboratories, four low-power critical experiment facilities, and two (megawatt-level) research reactors. This paper concentrates on the experiences with the plutonium/uranium fuels development laboratories and critical experiment facilities. These were comprehensive projects that included: developing the decommissioning and quality assurance plans; interfacing with the U.S. Nuclear Regulatory Commission, performing the actual decontamination/dismantling work; performing decontamination and final radiological surveys; and volume reducing, packaging, certifying, classifying, and shipping the radioactive waste for disposal. This broad experience has involved handling radioactive contamination from the following sources: low- and high-enriched U-235 fuel; depleted uranium; mixed oxide fuel (Pu/UO); thorium fuel; U Al alloy fuel; and fission activation products (beta-gamma). Areas of application to future projects are highlighted in this paper

  8. Qualification of the B and W Mark B fuel assembly for high burnup. Third semi-annual progress report, July-December 1979

    International Nuclear Information System (INIS)

    Five Babcock and Wilcox-designed Mark B (15 x 15) pressurized water reactor fuel assemblies were irradiated to extended burnups in Duke Power Company's Oconee Unit 1 reactor. An assembly average burnup of 40,000 MWd/mtU, which is about 29% greater than previous discharge burnups at Oconee 1, was attained. The nondestructive examination of these five assemblies, which have been irradiated for four fuel cycles, was begun. Data obtained included fuel assembly and fuel dimensions, water channel spacings, fuel rod surface deposit samples, and holddown spring preload forces. Visual examination of the assemblies indicated that good fuel performance was maintained through four cycles of irradiation

  9. Resetting the epigenetic histone code in the MRL-lpr/lpr mouse model of lupus by histone deacetylase inhibition.

    Science.gov (United States)

    Garcia, Benjamin A; Busby, Scott A; Shabanowitz, Jeffrey; Hunt, Donald F; Mishra, Nilamadhab

    2005-01-01

    The baseline level of gene expression varies between healthy controls and systemic lupus erythematosus (SLE) patients, and among SLE patients themselves. These variations may explain the different clinical manifestations and severity of disease observed in SLE. Epigenetic mechanisms, which involve DNA and histone modifications, are predictably associated with distinct transcriptional states. To understand the interplay between various histone modifications, including acetylation and methylation, and lupus disease, we performed differential expression histone modification analysis in splenocytes from the MRL-lpr/lpr mouse model of lupus. Using stable isotope labeling in combination with mass spectrometry, we found global site-specific hypermethylation (except H3 K4 methylation) and hypoacetylation in histone H3 and H4 MRL-lpr/lpr mice compared to control MRL/MPJ mice. Moreover, we have identified novel histone modifications such as H3 K18 methylation, H4 K31 methylation, and H4 K31 acetylation that are differentially expressed in MRL-lpr/lpr mice compared to controls. Finally, in vivo administration of the histone deacetylase inhibitor trichostatin A (TSA) corrected the site-specific hypoacetylation states on H3 and H4 in MRL-lpr/lpr mice with improvement of disease phenotype. Thus, this study is the first to establish the association between aberrant histone codes and pathogenesis of autoimmune disease SLE. These aberrant post-translational histone modifications can therefore be reset with histone deacetylase inhibition in vivo. PMID:16335948

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

    International Nuclear Information System (INIS)

    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.

  11. 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.

  12. Deletion of microRNA-155 reduces autoantibody responses and alleviates lupus-like disease in the Fas(lpr) mouse.

    Science.gov (United States)

    Thai, To-Ha; Patterson, Heide Christine; Pham, Duc-Hung; Kis-Toth, Katalin; Kaminski, Denise A; Tsokos, George C

    2013-12-10

    MicroRNA-155 (miR-155) regulates antibody responses and subsequent B-cell effector functions to exogenous antigens. However, the role of miR-155 in systemic autoimmunity is not known. Using the death receptor deficient (Fas(lpr)) lupus-prone mouse, we show here that ablation of miR-155 reduced autoantibody responses accompanied by a decrease in serum IgG but not IgM anti-dsDNA antibodies and a reduction of kidney inflammation. MiR-155 deletion in Fas(lpr) B cells restored the reduced SH2 domain-containing inositol 5'-phosphatase 1 to normal levels. In addition, coaggregation of the Fc γ receptor IIB with the B-cell receptor in miR-155(-/-)-Fas(lpr) B cells resulted in decreased ERK activation, proliferation, and production of switched antibodies compared with miR-155 sufficient Fas(lpr) B cells. Thus, by controlling the levels of SH2 domain-containing inositol 5'-phosphatase 1, miR-155 in part maintains an activation threshold that allows B cells to respond to antigens. PMID:24282294

  13. Piperlongumine alleviates lupus nephritis in MRL-Fas(lpr) mice by regulating the frequency of Th17 and regulatory T cells.

    Science.gov (United States)

    Yao, Lan; Chen, Hai-ping; Ma, Qing

    2014-09-01

    Recent data have shown that piperlongumine (PL), an important component of Piper longum fruits, is known to possess anti-inflammatory and vascular-protective activities. This study aimed to examine the therapeutic effects and underlying mechanisms of PL on lupus-prone MRL-Fas(lpr) mice. Female MRL-Fas(lpr) mice were intraperitoneally treated with PL (2.4 mg kg(-1) d(-1)) for 10 weeks, and the proteinuria level was biweekly monitored. After the mice were euthanized, serum biochemical parameters and renal damage were determined. Splenocytes of MRL-Fas(lpr) mice were isolated for in vitro study. Treatment of the mice with PL significantly attenuated the progression of proteinuria and glomerulonephritis. The improvement was accompanied by decreased serum levels of nephritogenic anti-dsDNA antibodies, IL-6, IL-17, IL-23 and TNF-α. Treatment of the mice with PL suppressed the frequency of Th17 cells and increased the regulatory T cells (Tregs). In vitro, the levels of IL-6, IL-17, IL-23 and TNF-α were significantly decreased in the cultures of splenocytes from PL-treated mice compared with those from vehicle-treated mice. In addition, PL treatment impeded activation of the JAK/STAT3 signaling in splenocytes. Of great important, the survival of MRL-Fas(lpr) mice were improved by PL treatment. In summary, PL effectively ameliorates lupus syndrome in MRL-Fas(lpr) mice by suppressing the pathogenic Th17 cells and increasing the Tregs as well as inhibiting activation of the JAK/STAT3 signaling pathway. This study sheds new light on the immune-modulatory role of PL. PMID:24837470

  14. Technical and economic studies of small reactors for supply of electricity and steam

    International Nuclear Information System (INIS)

    Several years ago conventional opinion held that nuclear power plants must be very large to be competitive with fossil fuels. This situation has changed markedly in most countries within recent years, as oil and gas supplies have become more scarce and costly. Studies have been carried out for several nuclear steam supply systems in the small and intermediate size range. Detail studies are reported of the Consolidated Nuclear Steam Generator (CNSG), a 365 MW(th) pressurized water reactor being developed by Babcock and Wilcox, as applied to industrial energy needs. Both conventional and barge-mounted nuclear steam supply systems are considered. Conceptual studies have been started of pressurized and boiling water reactors in the range of 1000 MW(th), which are envisioned for utility operation for supply of electric power and steam. Design studies of a 500 MW(th) high temperature reactor are also reported. The small reactors are expected to have higher unit costs than the large commercial plants, but to have compensating advantages in higher plant availability, shorter construction schedule, and greater siting flexibility. Studies are also reported of power cycle parameters and cost allocations for extraction of steam from steam turbine plants. This steam could be used for industrial energy, district heating, or desalination

  15. miR-155 Deficiency Ameliorates Autoimmune Inflammation of Systemic Lupus Erythematosus by Targeting S1pr1 in Faslpr/lpr Mice.

    Science.gov (United States)

    Xin, Qian; Li, Jiangxia; Dang, Jie; Bian, Xianli; Shan, Shan; Yuan, Jupeng; Qian, Yanyan; Liu, Zhaojian; Liu, Guangyi; Yuan, Qianqian; Liu, Na; Ma, Xiaochun; Gao, Fei; Gong, Yaoqin; Liu, Qiji

    2015-06-01

    MicroRNA-155 (miR-155) was previously found involved in the development of systemic lupus erythematosus (SLE) and other autoimmune diseases and the inflammatory response; however, the detailed mechanism of miR-155 in SLE is not fully understood. To explore the in vivo role of miR-155 in the pathogenesis of SLE, miR-155-deficient Fas(lpr/lpr) (miR-155(-/-)Fas(lpr/lpr)) mice were obtained by crossing miR-155(-/-) and Fas(lpr/lpr) mice. Clinical SLE features such as glomerulonephritis, autoantibody levels, and immune system cell populations were compared between miR-155(-/-)Fas(lpr/lpr) and Fas(lpr/lpr) mice. Microarray analysis, RT-PCR, Western blot, and luciferase reporter gene assay were used to identify the target gene of miR-155. miR-155(-/-)Fas(lpr/lpr) mice showed milder SLE clinical features than did Fas(lpr/lpr)mice. As compared with Fas(lpr/lpr) mice, miR-155(-/-)Fas(lpr/lpr) mice showed less deposition of total IgA, IgM, and IgG and less infiltration of inflammatory cells in the kidney. Moreover, the serum levels of IL-4 and IL-17a, secreted by Th2 and Th17 cells, were lower in miR-155(-/-)Fas(lpr/lpr) than Fas(lpr/lpr) mice; the CD4(+)/CD8(+) T cell ratio was restored in miR-155(-/-)Fas(lpr/lpr) mice as well. Sphingosine-1-phosphate receptor 1 (S1PR1) was found as a new target gene of miR-155 by in vitro and in vivo studies; its expression was decreased in SLE patients and Fas(lpr/lpr) mice. miR-155(-/-)Fas(lpr/lpr) mice are resistant to the development of SLE by the regulation of the target gene S1pr1. miR-155 might be a new target for therapeutic intervention in SLE. PMID:25911753

  16. Thorium base fuels reprocessing at the L.P.R. (Radiochemical Processes Laboratory) experimental plant

    International Nuclear Information System (INIS)

    The availability of the LPR (Radiochemical Processes Laboratory) plant offers the possibility to demonstrate and create the necessary technological basis for thorium fuels reprocessing. To this purpose, the solvents extraction technique is used, employing TBP (at 30%) as solvent. The process is named THOREX, a one-cycle acid, which permits an adequate separation of Th232 and U233 components and fission products. For thorium oxide elements dissolution, the 'chopp-leach' process (installed at LPR) is used, employing a NO3 H 13N, 0.05M FH and 0.1M Al (NO3)3, as solvent. To adapt the pilot plant to the flow-sheet requirements proposed, minor modifications must be carried out in the interconnection of the existing decanting mixers. The input of the plant has been calculated by Origin Code modified for irradiations in reactors of the HWR type. (Author)

  17. Nuclear plant-aging research on reactor protection systems

    International Nuclear Information System (INIS)

    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

  18. Babcock-Leighton solar dynamo: the role of downward pumping and the equatorward propagation of activity

    Science.gov (United States)

    Karak, Bidya Binay; Cameron, Robert

    2016-05-01

    We investigate the role of downward magnetic pumping near the surface using a kinematic Babcock-Leighton model. We find that the pumping causes the poloidal field to become predominately radial in the near-surface shear layer. This allows the negative radial shear in the near-surface layer to effectively act on the radial field to produce a toroidal field. Consequently, we observe a clear equatorward migration of the toroidal field at low latitudes even when there is no meridional flow in the deep CZ. We show a case where the period of a dynamo wave solution is approximately 11 years. Flux transport models are also shown with periods close to 11 years. Both the dynamo wave and flux transport dynamo are thus able to reproduce some of the observed features of solar cycle. The main difference between the two types of dynamo is the value of $\\alpha$ required to produce dynamo action. In both types of dynamo, the surface meridional flow helps to advect and build the polar field in high latitudes, while in flux transport dynamo the equatorward flow near the bottom of CZ advects toroidal field to cause the equatorward migration in butterfly wings and this advection makes the dynamo easier by transporting strong toroidal field to low latitudes where $\\alpha$ effect works. Another conclusion of our study is that the magnetic pumping suppresses the diffusion of fields through the photospheric surface which helps to achieve the 11-year dynamo cycle at a moderately larger value of magnetic diffusivity than has previously been used.

  19. Scientific Opinion on the safety assessment of the process LPR based on EREMA Advanced and Colortronic SSP ® technology used to recycle post-consumer PET into food contact materials

    Directory of Open Access Journals (Sweden)

    EFSA Panel on Food Contact Materials, Enzymes, Flavourings and Processing Aids (CEF

    2014-02-01

    Full Text Available This scientific opinion of the EFSA Panel on Food Contact Materials, Enzymes, Flavourings and Processing Aids deals with the safety assessment of the recycling process LPR (EU register No RECYC061 which is based on the EREMA advanced and Colortronic SSP ® technologies. The input to the process is hot caustic washed and dried PET flakes originating from collected post-consumer PET bottles and containing no more than 5 % of PET from non-food consumer applications. In this process, washed and dried PET flakes are heated successively in two continuous reactors under vacuum before being extruded into pellets. After extrusion they are crystallised and solid state polymerized. Having examined the results of the challenge test provided, the Panel concluded that the four steps, the decontamination in two continuous reactors, extrusion, crystallisation and solid state polymerization are the critical steps that determine the decontamination efficiency of the process. The operating parameters to control the performance of these critical steps are temperature, pressure, gas flow and residence time. Under these conditions, it was demonstrated that the recycling process is able to ensure that the level of migration of potential unknown contaminants into food is below the modelled migration of 0.1 μg/kg food derived from exposure scenario for infants and 0.15 μg/kg food derived from the exposure scenario for toddlers. The Panel concluded that recycled PET obtained from LPR process is not of safety concern when used to manufacture articles intended for food contact materials applications in compliance with the conditions as specified in the conclusion of the opinion.

  20. Preliminary Design Concept for a Reactor-internal CRDM

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jae Seon; Kim, Jong Wook; Kim, Tae Wan; Choi, Suhn; Kim, Keung Koo [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-10-15

    A rod ejection accident may cause severer result in SMRs because SMRs have relatively high control rod reactivity worth compared with commercial nuclear reactors. Because this accident would be perfectly excluded by adopting a reactor-internal CRDM (Control Rod Drive Mechanism), many SMRs accept this concept. The first concept was provided by JAERI with the MRX reactor which uses an electric motor with a ball screw driveline. Babcock and Wilcox introduced the concept in an mPower reactor that adopts an electric motor with a roller screw driveline and hydraulic system, and Westinghouse Electric Co. proposes an internal Control Rod Drive in its SMR with an electric motor with a latch mechanism. In addition, several other applications have been reported thus far. The reactor-internal CRDM concept is now widely adopted in many SMR designs, and this concept may also be applied in an evolutionary reactor development. So the preliminary study is conducted based on the SMART CRDM design. A preliminary design concept for a reactor-internal CRDM was proposed and evaluated through an electromagnetic analysis. It was found that there is an optimum design for the motor housing, and the results may contribute to the realization a reactor-internal CRDM for an evolutionary reactor development. More detailed analysis results will be reported later.

  1. On the Meaning of Formative Measurement and How It Differs from Reflective Measurement: Comment on Howell, Breivik, and Wilcox (2007)

    Science.gov (United States)

    Bagozzi, Richard P.

    2007-01-01

    D. Howell, E. Breivik, and J. B. Wilcox (2007) have presented an important and interesting analysis of formative measurement and have recommended that researchers abandon such an approach in favor of reflective measurement. The author agrees with their recommendations but disagrees with some of the bases for their conclusions. He suggests that…

  2. Apparent Depth to the Wilcox Group, Gulf Coast (wlcxdpthg)

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — The depth to top of the Wilcox Group is contoured from location and top information derived from the Petroleum Information (PI) Wells database. The depth to Wilcox...

  3. Interpretational Confounding Is Due to Misspecification, Not to Type of Indicator: Comment on Howell, Breivik, and Wilcox (2007)

    Science.gov (United States)

    Bollen, Kenneth A.

    2007-01-01

    R. D. Howell, E. Breivik, and J. B. Wilcox (2007) have argued that causal (formative) indicators are inherently subject to interpretational confounding. That is, they have argued that using causal (formative) indicators leads the empirical meaning of a latent variable to be other than that assigned to it by a researcher. Their critique of causal…

  4. A Babcock-Leighton solar dynamo model with multi-cellular meridional circulation in advection- and diffusion-dominated regimes

    CERN Document Server

    Belucz, Bernadett; Forgacs-Dajka, Emese

    2015-01-01

    Babcock-Leighton type solar dynamo models with single-celled meridional circulation are successful in reproducing many solar cycle features. Recent observations and theoretical models of meridional circulation do not indicate a single-celled flow pattern. We examine the role of complex multi-cellular circulation patterns in a Babcock-Leighton solar dynamo in advection- and diffusion-dominated regimes. We show from simulations that presence of a weak, second, high-latitude reverse cell speeds up the cycle and slightly enhances the poleward branch in butterfly diagram, whereas the presence of a second cell in depth reverses the tilt of butterfly wing to an anti-solar type. A butterfly diagram constructed from middle of convection zone yields a solar-like pattern, but this may be difficult to realize in the Sun because of magnetic buoyancy effects. Each of the above cases behaves similarly in higher and lower magnetic diffusivity regimes. However, our dynamo with a meridional circulation containing four cells in...

  5. Reports of the Technical Assessment Task Force on selection, training, qualification, and licensing of Three Mile Island reactor operating personnel; technical assessment of operating, abnormal, and emergency procedures; control room design and performance

    International Nuclear Information System (INIS)

    As a part of the effort to identify and evaluate the possible causes of the Three Mile Island accident, an analysis of operator training, qualification, licensing, selection, and manning was conducted by the staff. The study included review of documents, interviews, and depositions at Three Mile Island, Babcock and Wilcox, and the Nuclear Regulatory Commission (NRC) during June, July, and August 1979. Analysis of the information obtained was conducted almost exclusively by the writer. This paper examines the roles of the actors involved in training and it reviews the various programs which were intended to staff Three Mile Island with sufficient numbers of competent, trained operators and supervisors. The analysis includes a review of the regulations concerning operator training and licensing; describes how the requirements were implemented by the NRC, Metropolitan Edison Company (Met Ed), and Babcock and Wilcox Company (B and W), and then evaluates the programs conducted by these three organizations

  6. 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)

    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

  7. The New US Public-Private Partnership to License and DeploySmall Modular Reactors, With Focus on The B and W mPower Reactor

    International Nuclear Information System (INIS)

    On December 16, 2011, The US Congress and the President approved new Fiscal Year 2012 funding for a Government - Industry cost shared program called 'Small Modular Reactor (SMR) Licensing Technical Support'. The new legislation appropriates $67 million in 2012 to provide licensing and first-of-a-kind engineering support for small modular reactor designs that can be deployed expeditiously. The legislation requires the Department of Energy to consider applications utilizing any small modular reactor technology. Competitive solicitations are likely to begin shortly and two or three SM R designs will be selected for U S Government support. Such support will likely accelerate deployment and operation of at least one such design by 2020. The Congressional language states that the Government portion of the program is expected to total $452 million over five years One of the candidates for this competition is the B and W mPower reactor being developed by Generation mPower, a company recently formed by the Babcock and Wilcox Company and Bechtel Power Corporation. This presentation will summarize the main features of this design, and explain why it meets the requirements for the Government program, and will be fully developed, licensed and deployed in the US within the next 8 years. Importantly, this design has many features that favor its introduction and use in smaller countries with critical needs for future electric generation capacity, with arid conditions that may require air cooled condensers, and with potential need for a desalination component of the new energy source. The relatively small capacity of the modules (e.g. 320 MWe for an initial two unit plant) will require much lower initial capital investment, as compared to the very large investment of $4 to $6 billion required for the newer 1100 to 1400 MWe plants now being constructed in China, France, Finland, Korea, the US, and the United Arab Emirates

  8. Electrical system regulations of the IEA-R1 reactor

    International Nuclear Information System (INIS)

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

  9. Organic petrology and coalbed gas content, Wilcox Group (Paleocene-Eocene), northern Louisiana

    Science.gov (United States)

    Hackley, P.C.; Warwick, P.D.; Breland, F.C., Jr.

    2007-01-01

    Wilcox Group (Paleocene-Eocene) coal and carbonaceous shale samples collected from four coalbed methane test wells in northern Louisiana were characterized through an integrated analytical program. Organic petrographic analyses, gas desorption and adsorption isotherm measurements, and proximate-ultimate analyses were conducted to provide insight into conditions of peat deposition and the relationships between coal composition, rank, and coalbed gas storage characteristics. The results of petrographic analyses indicate that woody precursor materials were more abundant in stratigraphically higher coal zones in one of the CBM wells, consistent with progradation of a deltaic depositional system (Holly Springs delta complex) into the Gulf of Mexico during the Paleocene-Eocene. Comparison of petrographic analyses with gas desorption measurements suggests that there is not a direct relationship between coal type (sensu maceral composition) and coalbed gas storage. Moisture, as a function of coal rank (lignite-subbituminous A), exhibits an inverse relationship with measured gas content. This result may be due to higher moisture content competing for adsorption space with coalbed gas in shallower, lower rank samples. Shallower ( 600??m) coal samples containing less moisture range from under- to oversaturated with respect to their CH4 adsorption capacity.

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

    International Nuclear Information System (INIS)

    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)

  11. Spontaneous elaboration of transforming growth factor beta suppresses host defense against bacterial infection in autoimmune MRL/lpr mice

    OpenAIRE

    1994-01-01

    Infection with gram-negative and gram-positive bacteria remains a leading cause of death in patients with systemic lupus erythematosis (SLE), even in the absence of immunosuppressive therapy. To elucidate the mechanisms that underly the increased risk of infection observed in patients with systemic autoimmunity, we have investigated host defense against bacterial infection in a murine model of autoimmunity, the MRL/Mp-lpr/lpr (MRL/lpr) mouse. Our previous study implicated transforming growth ...

  12. The simulation of thermohydraulic phenomena in a pressurized water reactor primary loop

    International Nuclear Information System (INIS)

    Several important fluid flow and heat transfer phenomena essential to nuclear power reactor safety were investigated. Scaling and modeling laws for pressurized water reactors are reviewed and a new scaling approach focusing on the overall loop behavior is presented. Scaling criteria for one- and two-phase natural circulation are developed, as well as a simplified model describing the first phase of a small break loss of coolant accident. Reactor vessel vent valve effects are included in the analysis of steady one-phase natural circulation flow. Two new dimensionless numbers, which uniquely describe one-phase flow in natural circulation loops, were deduced and are discussed. A scaled model of the primary loop of a typical Babcock and Wilcox reactor was designed, built, and tested. The particular prototype modeled was the TMI unit 2 reactor. The electrically heated, stainless steel model operates at a maximum pressure of 300 psig and has a maximum heat input of 188 kW. The model is about 4 times smaller in height than the prototype reactor, with a nominal volume scale of 1:500. Experiments were conducted establishing subcooled natural circulation in the model loop. Both steady flow and power transients were investigated

  13. Comparative Investigation of River Water Quality by OWQI, NSFWQI and Wilcox Indexes (Case study: the Talar River – IRAN

    Directory of Open Access Journals (Sweden)

    Darvishi Gholamreza

    2016-03-01

    Full Text Available Rivers are considered as one of the main resources of water supply for various applications such as agricultural, drinking and industrial purposes. Also, these resources are used as a place for discharge of sewages, industrial wastewater and agricultural drainage. Regarding the fact that each river has a certain capacity for acceptance of pollutants, nowadays qualitative and environmental investigations of these resources are proposed. In this study, qualitative investigation of the Talar river was done according to Oregon Water Quality Index (OWQI, National Sanitation Foundation Water Quality Index (NSFWQI and Wilcox indicators during 2011–2012 years at upstream, midstream and downstream of the river in two periods of wet and dry seasons. According to the results of OWQI, all of the values at 3 stations and both periods are placed at very bad quality category and the water is not acceptable for drinking purposes. According to NSFWQI, the best condition was related to the upstream station at wet season period (58, medium quality and the worst condition was related to the downstream in wet season period (46, very bad quality. Also the results of Wilcox showed that in both periods of wet season and dry season, the water quality is getting better from upstream station to the downstream station, and according to the index classification, the downstream water quality has shown good quality and it is suitable for agriculture.

  14. 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.

  15. Analysis of local perfusion rate (LPR) and local glucose transport rate (LGTR) in brain and heart in man by means of C-11-methyl-D-glucose (CMG) and dynamic positron emission tomography (dPET)

    International Nuclear Information System (INIS)

    A method has been developed to measure simultaneously the LPR and LGTR. CMG is used as an indicator. The transaxial distribution of activity in organism is registered with dPET. On the basis of a mathematical model, the LPR and LGTR can be calculated in terms of parameters of the time activity curves registered over different brain or heart regions and over the sup. long. sinus (brain) or the ventricular cavity (heart) (blood activity). The method was used in 10 normal subjects and 20 patients with ischemic brain or heart disease. The values of LGTR range from 0.43 to 0.6 μmol/min g in normal cortex and from 0.09 to 0.12 μmol/min g in normal white matter. The LPR was 0.9-098 ml/min g for the cortex and 0.3-0.4 ml/min g for the white matter. In patients with stroke the ischemic defects appeared to be larger in CMG scans than in CT. The results obtained in a patient with left homonymous hemianopia, caused by infarctions in the distribution area of RMCA, and in a patient with TIA, demonstrate that the inactivation of morphologically intact, cerebral cortex, observed in stroke patients, may be caused by undercutting of cortical fiber tracts as well as by the impairment of the glucose transport systems in the inactivated area. In myocardial studies the LPR in normal left myocardium was 0.68 ml/min g (subendocardium 0.74 ml/min g; subepicardiuim 0.65 ml/min g). In patients with old myocardial infarction, the infarcted areas could be easily recognized as accumulation defects. The results obtained in a patient with narrowing of the RCA indicate that repeated exposure of myocardial tissue to transient ischemia produces an irreversible damage of the glucose transport system. We conclude from the data that for diagnostic evaluation of ultimate brain or heart damage simultaneous quantitative assessment of both LPR and LGTR is of basic importance. (Author)

  16. Statistical study of conductivity probe output signals in a high-pressure and -temperature test facility

    International Nuclear Information System (INIS)

    A scaled test facility was designed to evaluate the thermal-hydraulics conditions in the reactor coolant system and steam generator of a model of a Babcock and Wilcox pressurized water reactor (PWR) during the natural circulation phases of a small-break loss-of-coolant accident. The test facility, referred to as the Once-Through Integral System (OTIS), was equipped with ∼ 250 instruments, including 36 conductivity probes to measure the thermal-hydraulics response of the system during the transient tests. The purpose of this study is to present statistical characteristics of the conductivity probe output signals. Autocorrelation and cross-correlation analyses of signals produced by spatially separated probes were computed using long and conditional sampling techniques. The cross-correlation signal analysis of conductivity probes revealed some information about the flow patterns in the hot leg and U-bend pipe of the PWR

  17. 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. PMID:18981160

  18. Interview with Professor Mark Wilcox.

    Science.gov (United States)

    Wilcox, Mark

    2016-08-01

    Mark Wilcox speaks to Georgia Patey, Commissioning Editor: Professor Mark Wilcox is a Consultant Microbiologist and Head of Microbiology at the Leeds Teaching Hospitals (Leeds, UK), the Professor of Medical Microbiology at the University of Leeds (Leeds, UK), and is the Lead on Clostridium difficile and the Head of the UK C. difficile Reference Laboratory for Public Health England (PHE). He was the Director of Infection Prevention (4 years), Infection Control Doctor (8 years) and Clinical Director of Pathology (6 years) at the Leeds Teaching Hospitals. He is Chair of PHE's Rapid Review Panel (reviews utility of infection prevention and control products for National Health Service), Deputy Chair of the UK Department of Health's Antimicrobial Resistance and Healthcare Associated Infection Committee and a member of PHE's HCAI/AR Programme Board. He is a member of UK/European/US working groups on C. difficile infection. He has provided clinical advice as part of the FDA/EMA submissions for the approval of multiple novel antimicrobial agents. He heads a healthcare-associated infection research team at University of Leeds, comprising approximately 30 doctors, scientists and nurses; projects include multiple aspects of C. difficile infection, diagnostics, antimicrobial resistance and the clinical development of new antimicrobial agents. He has authored more than 400 publications, and is the coeditor of Antimicrobial Chemotherapy (5th/6th/7th Editions, 15 December 2007). PMID:27494150

  19. Peter Wilcox: A new purple-skin, yellow flesh fresh market potato cultivar

    Science.gov (United States)

    Peter Wilcox is a new, medium-maturing, purple-skin, yellow-flesh potato cultivar for fresh market. Peter Wilcox also produces light-colored chips, although it is being released primarily as a fresh market potato because of its skin and flesh colors. Tubers of Peter Wilcox are attractive, smooth, wi...

  20. Use of the modular modeling system in severe transient analysis of Penn State advanced light water reactor

    International Nuclear Information System (INIS)

    The Department of Nuclear Engineering of the Pennsylvania State University has designed and developed, with Department of Energy support, an inherently safe pressurized light water reactor concept. The Penn State University Advanced Light Water Reactor (PSU ALWR) incorporates various passive and active ultra-safe features, such as continuous online injection and letdown for pressure control, a raised-loop reactor primary system for enhanced natural circulation, a dedicated primary reservoir (the atmospheric tank) for enhanced thermal hydraulic control, and a secondary shutdown turbine. Because of the conceptual design basis of the project, the dynamic system modeling was to be performed using a code with a high degree of flexibility. For these reasons, the modeling was performed with the Modular Modeling System (MMS) developed by Babcock and Wilcox for EPRI. The demonstrative transient chosen for the PSU ALWR was a turbine trip and reactor scram, concurrent with total station blackout. This transient demonstrates the utility of the pressure control system, the shutdown turbine generator, and the enhanced natural circulation of the PSU ALWR. However, the low flow rates, low pressure drops, and large derivative states encountered in such a transient pose special problems for the modeler and for MMS. The use of Extended Range MMS, the development of the necessary system controls, and certain local modifications to the MMS itself are described below, along with the final results of the prescribed transient

  1. Experience of Deutsche Babcock AG with the maintenance and rehabilitation of existing power plants

    Energy Technology Data Exchange (ETDEWEB)

    Horstmann, H.; Frank, R.

    1997-12-31

    With regard to the exponentially increasing power demand in a lot of countries the rehabilitation of existing power plants has become more and more important during the last years. From the economic viewpoint, it is necessary to develop a tailor made rehabilitation program for the individual power plant by defining the relevant measures. The main benefits of power plant rehabilitation are shown in relation to the necessary technical measures and with regard to their economic effect. 15 figs.

  2. Altitude of the water table in the alluvial and Wilcox aquifers in the vicinity of Richland and Tehuacana creeks and the Trinity River, Texas, December 1979

    Science.gov (United States)

    Garza, Sergio

    1980-01-01

    This map shows the altitude of the water table in the alluvial and Wilcox aquifers in the vicinity of Richland and Tehuacana Creeks and the Trinity River, Tex., in December 1979. The water-table contours were constructed on the basis of water-level control derived from an inventory of shallow wells in the area, topographic maps, and field locations of numerous small springs and seeps. (USGS)

  3. Cutting Edge: IL-23 Receptor Deficiency Prevents the Development of Lupus Nephritis in C57BL/6–lpr/lpr Mice

    OpenAIRE

    Kyttaris, Vasileios C.; Zhang, Zheng; Kuchroo, Vijay K.; Oukka, Mohamed; Tsokos, George C.

    2010-01-01

    IL-17–producing T cells infiltrate kidneys of patients with lupus nephritis, and IL-23–treated lymph node cells from lupus-prone mice may transfer disease to Rag1-deficient mice. In this study, we show that IL-23R–deficient lupus-prone C57BL/6–lpr/lpr mice display decreased numbers of CD3+CD4−CD8− cells and IL-17A–producing cells in the lymph nodes and produce less anti-DNA Abs. In addition, clinical and pathology measures of lupus nephritis are abrogated. The presented experiments document t...

  4. Scientific Opinion on the safety assessment of the process LPR based on EREMA Advanced and Colortronic SSP ® technology used to recycle post-consumer PET into food contact materials

    OpenAIRE

    EFSA Panel on Food Contact Materials, Enzymes, Flavourings and Processing Aids (CEF)

    2014-01-01

    This scientific opinion of the EFSA Panel on Food Contact Materials, Enzymes, Flavourings and Processing Aids deals with the safety assessment of the recycling process LPR (EU register No RECYC061) which is based on the EREMA advanced and Colortronic SSP ® technologies. The input to the process is hot caustic washed and dried PET flakes originating from collected post-consumer PET bottles and containing no more than 5 % of PET from non-food consumer applications. In this process, washed and d...

  5. 3D Babcock-Leighton Solar Dynamo Models

    Science.gov (United States)

    Miesch, Mark S.; Hazra, Gopal; Karak, Bidya Binay; Teweldebirhan, Kinfe; Upton, Lisa

    2016-05-01

    We present results from the new STABLE (Surface flux Transport and Babcock Leighton) Dynamo Model. STABLE is a 3D Babcock-Leighton/Flux Transport dynamo model in which the source of poloidal field is the explicit emergence, distortion, and dispersal of bipolar magnetic regions (BMRs). In this talk I will discuss initial results with axisymmetric flow fields, focusing on the operation of the model, the general features of the cyclic solutions, and the challenge of achieving supercritical dynamo solutions using only the Babcock-Leighton source term. Then I will present dynamo simulations that include 3D convective flow fields based on the observed velocity power spectrum inferred from photospheric Dopplergrams. I'll use these simulations to assess how the explicit transport and amplification of fields by surface convection influences the operation of the dynamo. I will also discuss the role of surface magnetic fields in regulating the subsurface toroidal flux budget.

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

    International Nuclear Information System (INIS)

    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

  7. Study of Channel Morphology and Infill Lithology in the Wilcox Group Central Louisiana Using Seismic Attribute Analysis

    Science.gov (United States)

    Chen, Feng

    The fluvial and deltaic Wilcox Group is a major target for hydrocarbon and coal exploration in northern and central Louisiana. However, the characterization and delineation of fluvial systems is a difficult task due to the variability and complexity of fluvial systems and their internal heterogeneities. Seismic geomorphology is studied by recognizing paleogeographic features in seismic stratal slices, which are seismic images of paleo-depositional surfaces. Seismic attributes, which are extracted along seismic stratal slices, can reveal information that is not readily apparent in raw seismic data. The existence and distribution of fluvial channels are recognized by the channel geomorphology in seismic attributes displayed on stratal slices. The lithologies in the channels are indicated by those seismic attributes that are directly related to the physical properties of rocks. Selected attributes utilized herein include similarity, spectral decomposition, sweetness, relative acoustic impedance, root mean square (RMS) amplitude, and curvature. Co-rendering and Red/Green/Blue (RGB) display techniques are also included to better illuminate the channel geometry and lithology distribution. Hydrocarbons may exist in the channel sand-bodies, but are not explicitly identified herein. Future drilling plans for oil and gas exploration may benefit from the identification of the channels and the lithologies that fill them.

  8. International Reactor Physics Experiment Evaluation (IRPhE) Project. IRPhE Handbook - 2015 edition

    International Nuclear Information System (INIS)

    performed at 50 reactor facilities. To be published as approved benchmarks the experiments must be evaluated against agreed technical criteria and reviewed by the IRPhE Technical Review Group. A total of 139 of the 143 evaluations are published as approved benchmarks. The remaining four evaluations are published as draft documents only. New to the handbook are benchmark specifications for selected measurements from the Babcock and Wilcox (B and W) Spectral Shift Reactor Lattice Experiment that was performed to study the nuclear properties of rod lattices moderated by D2O-H2O mixtures. The International Handbook of Evaluated Reactor Physics Benchmark Experiments was prepared by a working party comprised of experienced reactor physics personnel from Argentina, Belgium, Brazil, Canada, P.R. of China, Czech Republic, France, Germany, Hungary, Italy, Japan, Republic of Korea, Russian Federation, Serbia, Slovenia, South Africa, Sweden, Switzerland, United Kingdom, and the United States of America

  9. Lipoprotein LprI of Mycobacterium tuberculosis Acts as a Lysozyme Inhibitor.

    Science.gov (United States)

    Sethi, Deepti; Mahajan, Sahil; Singh, Chaahat; Lama, Amrita; Hade, Mangesh Dattu; Gupta, Pawan; Dikshit, Kanak L

    2016-02-01

    Mycobacterium tuberculosis executes numerous defense strategies for the successful establishment of infection under a diverse array of challenges inside the host. One such strategy that has been delineated in this study is the abrogation of lytic activity of lysozyme by a novel glycosylated and surface-localized lipoprotein, LprI, which is exclusively present in M. tuberculosis complex. The lprI gene co-transcribes with the glbN gene (encoding hemoglobin (HbN)) and both are synchronously up-regulated in M. tuberculosis during macrophage infection. Recombinant LprI, expressed in Escherichia coli, exhibited strong binding (Kd ≤ 2 nm) with lysozyme and abrogated its lytic activity completely, thereby conferring protection to fluorescein-labeled Micrococcus lysodeikticus from lysozyme-mediated hydrolysis. Expression of the lprI gene in Mycobacterium smegmatis (8-10-fold) protected its growth from lysozyme inhibition in vitro and enhanced its phagocytosis and survival during intracellular infection of peritoneal and monocyte-derived macrophages, known to secrete lysozyme, and in the presence of exogenously added lysozyme in secondary cell lines where lysozyme levels are low. In contrast, the presence of HbN enhanced phagocytosis and intracellular survival of M. smegmatis only in the absence of lysozyme but not under lysozyme stress. Interestingly, co-expression of the glbN-lprI gene pair elevated the invasion and survival of M. smegmatis 2-3-fold in secondary cell lines in the presence of lysozyme in comparison with isogenic cells expressing these genes individually. Thus, specific advantage against macrophage-generated lysozyme, conferred by the combination of LprI-HbN during invasion of M. tuberculosis, may have vital implications on the pathogenesis of tuberculosis. PMID:26589796

  10. Virulence phenotypes of Legionella pneumophila associated with noncoding RNA lpr0035.

    Science.gov (United States)

    Jayakumar, Deepak; Early, Julie V; Steinman, Howard M

    2012-12-01

    The Philadelphia-1 strain of Legionella pneumophila, the causative organism of Legionnaires' disease, contains a recently discovered noncoding RNA, lpr0035. lpr0035 straddles the 5' chromosomal junction of a 45-kbp mobile genetic element, pLP45, which can exist as an episome or integrated in the bacterial chromosome. A 121-bp deletion was introduced in strain JR32, a Philadelphia-1 derivative. The deletion inactivated lpr0035, removed the 49-bp direct repeat at the 5' junction of pLP45, and locked pLP45 in the chromosome. Intracellular multiplication of the deletion mutant was decreased by nearly 3 orders of magnitude in Acanthamoeba castellanii amoebae and nearly 2 orders of magnitude in J774 mouse macrophages. Entry of the deletion mutant into amoebae and macrophages was decreased by >70%. The level of entry in both hosts was restored to that in strain JR32 by plasmid copies of two open reading frames immediately downstream of the 5' junction and plasmid lpr0035 driven by its endogenous promoter. When induced from a tac promoter, plasmid lpr0035 completely reversed the intracellular multiplication defect in macrophages but was without effect in amoebae. These data are the first evidence of a role for noncoding RNA lpr0035, which has homologs in six other Legionella genomes, in entry of L. pneumophila into amoebae and macrophages and in host-specific intracellular multiplication. The data also demonstrate that deletion of a direct-repeat sequence restricts the mobility of pLP45 and is a means of studying the role of pLP45 mobility in Legionella virulence phenotypes. PMID:22966048

  11. 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

  12. Use of the modular modeling system in the design of the Penn State advanced light water reactor

    International Nuclear Information System (INIS)

    The Modular Modeling System (MMS), developed by Babcock and Wilcox under a contract from the Electric Power Research Institute (EPRI), is a computer code designed for the simulation of nuclear and fossil power plants. MMS uses preprogrammed modules to present specific power plant components and allows for the interconnection of these modules in a wide variety of configurations to model present and future plant configurations. MMS requires the use of a simulation language to translate and execute the plant model. The Advanced Continuous Simulation Language (ASCL), a general purpose simulation language by Mitchell and Gauthier, was used in conjunction with MMS for the Advanced Light Water Reactor (ALWR) studies at the Pennsylvania State University (PSU). For the past year, the Nuclear Engineering Department at PSU, under a contract from the Department of Energy (DOE), has been involved in the conceptual design and evaluation of a reconfigured Ultra-Safe ALWR. The underlying design philosophy was that the large amounts of energy stored in a reactor at shutdown could be used in such a way as to ensure safe plant shutdown, even if all AC power to the plant is lost. A secondary shutdown turbine was employed to recover energy to power the initial cooldown of the plant until natural circulation can develop and dissipate the remaining decay heat in the core. Primary system pressure is no longer controlled using a conventional pressurizer. Instead a modified let-down injection system connected to an inside containment atmospheric tank controls pressure

  13. Use of the modular modeling system in the design of the Penn State Advanced Light Water Reactor

    International Nuclear Information System (INIS)

    This study involves the design and subsequent transient analysis of the Penn State Advanced Light Water Reactor (PSU ALWR). The performance of the PSU ALWR is evaluated during small step changes in power and a turbine trip from full power without scram. The Modular Modeling System (MMS), developed by Babcock and Wilcox under a contract from the Electric Power Research Institute (EPRI), is a computer code designed for the simulation of nuclear and fossil power plants. MMS uses preprogrammed modules to represent specific power plant components such as pipes, pumps, steam generators, and a nuclear reactor. These components can then be connected in any manner the user desires providing certain simple interconnection rules are followed. In this study, MMS is used to develop computer models of both the PSU ALWR and a conventional PWR operating at the same power level. These models are then subjected to the transients mentioned above to evaluate the ability of the letdown-injection system to maintain primary system pressure. The transient response of the PSU ALWR and conventional PWR MMS models were compared to each other and whenever possible to actual plant transient data. 14 refs., 29 figs., 5 tabs

  14. Reactor physics and reactor computations

    International Nuclear Information System (INIS)

    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

  15. Babcock Redux: An Ammendment of Babcock's Schematic of the Sun's Magnetic Cycle

    CERN Document Server

    Moore, Ronald L; Sterling, Alphonse C

    2016-01-01

    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 and Roberts (1983) 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 field from the sunspot-region omega-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 t...

  16. FABRICATION PROCESS AND PRODUCT QUALITY IMPROVEMENTS IN ADVANCED GAS REACTOR UCO KERNELS

    Energy Technology Data Exchange (ETDEWEB)

    Charles M Barnes

    2008-09-01

    A major element of the Advanced Gas Reactor (AGR) program is developing fuel fabrication processes to produce high quality uranium-containing kernels, TRISO-coated particles and fuel compacts needed for planned irradiation tests. The goals of the AGR program also include developing the fabrication technology to mass produce this fuel at low cost. Kernels for the first AGR test (“AGR-1) consisted of uranium oxycarbide (UCO) microspheres that werre produced by an internal gelation process followed by high temperature steps tot convert the UO3 + C “green” microspheres to first UO2 + C and then UO2 + UCx. The high temperature steps also densified the kernels. Babcock and Wilcox (B&W) fabricated UCO kernels for the AGR-1 irradiation experiment, which went into the Advance Test Reactor (ATR) at Idaho National Laboratory in December 2006. An evaluation of the kernel process following AGR-1 kernel production led to several recommendations to improve the fabrication process. These recommendations included testing alternative methods of dispersing carbon during broth preparation, evaluating the method of broth mixing, optimizing the broth chemistry, optimizing sintering conditions, and demonstrating fabrication of larger diameter UCO kernels needed for the second AGR irradiation test. Based on these recommendations and requirements, a test program was defined and performed. Certain portions of the test program were performed by Oak Ridge National Laboratory (ORNL), while tests at larger scale were performed by B&W. The tests at B&W have demonstrated improvements in both kernel properties and process operation. Changes in the form of carbon black used and the method of mixing the carbon prior to forming kernels led to improvements in the phase distribution in the sintered kernels, greater consistency in kernel properties, a reduction in forming run time, and simplifications to the forming process. Process parameter variation tests in both forming and sintering steps led

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

    International Nuclear Information System (INIS)

    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

  18. Coal gasification systems engineering and analysis. Appendix D: Cost and economic studies

    Science.gov (United States)

    1980-01-01

    The detailed cost estimate documentation for the designs prepared in this study are presented. The include: (1) Koppers-Totzek, (2) Texaco (3) Babcock and Wilcox, (4) BGC-Lurgi, and (5) Lurgi. The alternate product cost estimates include: (1) Koppers-Totzek and Texaco single product facilities (methane, methanol, gasoline, hydrogen), (2) Kopers-Totzek SNG and MBG, (3) Kopers-Totzek and Texaco SNG and MBG, and (4) Lurgi-methane and Lurgi-methane and methanol.

  19. Nondestructive analysis at B and W's uranium conversion plant

    International Nuclear Information System (INIS)

    Containers and processing lines bearing high and low enriched uranium are routinely analyzed by nondestructive assay. Measurement systems used at Babcock and Wilcox's nuclear fuels plant in Apollo, Pennsylvania include the segmented gamma scanner (SGS) and the stabilized assay meter (SAM-II). These systems have been calibrated and used for a variety of tasks including uranium holdup measurements prior to decommissioning, in situ filter analysis and assay of calcined waste. 2 refs

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

    International Nuclear Information System (INIS)

    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

  1. Ethology of Omniablautus nigronotum (Wilcox) (Diptera: Asilidae) in Wyoming

    Science.gov (United States)

    In southwest Wyoming, Omniablautus nigronotum (Wilcox), hunted primarily from the surface of the sandy substrate in a greasewood community. Prey, captured in flight, represented four insect orders with Diptera and Hymenoptera predominating. Courtship consisted of the male approaching the female from...

  2. Azalea's Worst Nightmare: The Strawberry Rootworm, Paria fargariae Wilcox

    Science.gov (United States)

    The strawberry rootworm (SRW), Paria fargariae Wilcox, is an emergent pest of azaleas in commercial production nurseries in the southeastern US. Larvae feed on roots but do minimal damage. Adults feed at night and make small holes in the foliage. Severe damage has been reported in many nurseries, es...

  3. Babcock-pocket guide energy- and environmental engineering in the plant technology. Refractory construction, heat- and sound insulation, coatings. 4. ed.; Babcock-Taschenbuch Energie- und Umwelttechnik im Anlagenbau. Feuerfestbau, Waerme- und Schallschutz, Beschichtungen

    Energy Technology Data Exchange (ETDEWEB)

    Fuchs, W.E.

    2003-07-01

    Main topics of the pocket guide: constructions for refractories; thermal insulation for pipes, boilers and ceramic components; sound protection and measures on buildings; corrosion protection by coatings; ventilation in power plants; fire prevention in components and general fundamentals as data of technical, physical and chemical data of important materials. (GL)

  4. 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.

  5. Development of ground water from the Carrizo sand and Wilcox group in Dimmit, Zavala, Maverick, Frio, Atacosa, Median, Bexar, Live Oak, McMullen, La Salle, and Webb Counties, Texas

    Science.gov (United States)

    Moulder, E.A.

    1957-01-01

    The development of ground water for irrigation from the Carrizo sand south and southwest of San Antonio, Tex., has increased rapidly during the past few years. Declining pumping water levels in irrigation wells, caused by increased withdrawals, have caused considerable concern among the residents of the area. In response, the Nueces River Conservation and Reclamation District entered into a cooperative agreement with the Texas Board of Water Engineers and the United States Geological Survey to determine the extent of development and the rate of withdrawal that has cause the decline. All wells that discharged more than 150 gallons per minute for extended periods of time in 1955 from either the Carrizo sand or sands of the Wilcox group were studied and are shown on [late 1. Estimates were made of the total withdrawals by county and are given in table 2. Similar estimates of withdrawals in some of the counties for the irrigation years 1929-30, 1938-39, 1944-45, and 1947-48 are presented for comparison in table 3. Although the Carrizo sand is the principal source of ground water pumped in the area, estimate of withdrawals of water from the Wilcox were included in this inventory because (1) the formation appears to be hydraulically connected to the Carrizo sand, (2) the quality of water generally is good in the outcrop area of the Wilcox, and (3) appreciable withdrawals are being made from the Wilcox for irrigation in a few areas. The investigation covered an area of about 7,500 square miles and included all or parts of the following counties: Dimmit, Zavala, Maverick, Frio, Atascosa, Medina, Bexar, Live Oak, McMullen, La Salle, and Webb (fig. 1).

  6. Repeated 0.5 Gy gamma-ray irradiation attenuates autoimmune disease in MRL-lpr/lpr mice with up-regulation of regulatory T cells

    International Nuclear Information System (INIS)

    Complete text of publication follows. MRL-lpr/lpr mice present a single gene mutation on the Fas (CD95) gene that leads to reduced signaling for apoptosis. With aging, these mice spontaneously develop autoimmune disease and are used as a model of systemic lupus erythematosus. We previously reported attenuation of autoimmune disease in MRL-lpr/lpr mice by repeated γ-ray irradiation (0.5 Gy each time). In this study, we investigated the mechanisms of this attenuation focusing the highly activated CD3+CD4-CD8-B220+ T cells, which are characteristically involved in autoimmune pathology in these mice. We measured the weight of the spleen and the population of CD3+CD4-CD8-B220+ T cells. Splenomegaly and increase in percentage of CD3+CD4-CD8-B220+ T cells, which occur with aging in non-irradiated mice, were suppressed in irradiated mice. To investigate the function of CD3+CD4-CD8-B220+ T cells, we isolated these cells from splenocytes by magnetic cell sorting. Isolated CD3+CD4-CD8-B220+ T cells were more resistant to irradiation-induced cell death than isolated CD4+ T cells. Although high proliferation rate and IL-6 production were observed in isolated CD3+CD4-CD8-B220+ T cells, the proliferation rate and IL-6 production were lower in the cells isolated from the irradiated mice. Moreover, the production of autoantibodies (anti-collagen antibody and anti-single strand DNA antibody) was also lowered by irradiation. These results indicate that activation of CD3+CD4-CD8-B220+ T cells and progression of pathology would be suppressed by repeated 0.5 Gy γ-ray irradiation. To uncover the mechanism of the immune suppression, we analyzed population of regulatory T cells (CD4+CD25+Foxp3+), which suppress activated T cells and excessive autoimmune responses. Intriguingly, significant increase of the percentage of regulatory T cells was observed in irradiated mice. In conclusion, we found that repeated 0.5 Gy γ-ray irradiation suppresses proliferation rate of CD3+CD4-CD8-B220+ T

  7. On applicability of plate and shell heat exchangers for steam generation in naval PWR

    Energy Technology Data Exchange (ETDEWEB)

    Freire, Luciano Ondir, E-mail: luciano.ondir@gmail.com; Andrade, Delvonei Alves de, E-mail: delvonei@ipen.br

    2014-12-15

    Highlights: • Given emissions restrictions, nuclear propulsion may be an alternative. • Plate and shell heat exchangers (PSHE) are a mature technology on market. • PSHE are compact and could be used as steam generators. • Preliminary calculations to obtain a PWR for a large container ship are performed. • Results suggest PSHE improve overall compactness and cost. - Abstract: The pressure on reduction of gas emissions is going to raise the price of fossil fuels and an alternative to fossil fuels is nuclear energy. Naval reactors have some differences from stationary PWR because they have limitations on volume and weight, requiring compact solutions. On the other hand, a source of problems in naval reactors across history is the steam generation function. In order to reduce nuclear containment footprint, it is desirable to employ integral designs, which, however, poses complications and design constraints for recirculation type steam generators, being interesting to employ once through steam generators, whose historic at Babcock and Wilcox is better than recirculation steam generators. Plate and shell heat exchangers are a mature technology made available by many suppliers which allows heat exchange at high temperature and pressure. This work investigates the feasibility of the use of an array of welded plate heat exchangers of a material approved by ASME for pressure barrier (Ti-3Al-2.5V) in a hypothetical naval reactor. It was found it is feasible from thermal-hydraulic point of view and presents advantages over other steam generator designs.

  8. On applicability of plate and shell heat exchangers for steam generation in naval PWR

    International Nuclear Information System (INIS)

    Highlights: • Given emissions restrictions, nuclear propulsion may be an alternative. • Plate and shell heat exchangers (PSHE) are a mature technology on market. • PSHE are compact and could be used as steam generators. • Preliminary calculations to obtain a PWR for a large container ship are performed. • Results suggest PSHE improve overall compactness and cost. - Abstract: The pressure on reduction of gas emissions is going to raise the price of fossil fuels and an alternative to fossil fuels is nuclear energy. Naval reactors have some differences from stationary PWR because they have limitations on volume and weight, requiring compact solutions. On the other hand, a source of problems in naval reactors across history is the steam generation function. In order to reduce nuclear containment footprint, it is desirable to employ integral designs, which, however, poses complications and design constraints for recirculation type steam generators, being interesting to employ once through steam generators, whose historic at Babcock and Wilcox is better than recirculation steam generators. Plate and shell heat exchangers are a mature technology made available by many suppliers which allows heat exchange at high temperature and pressure. This work investigates the feasibility of the use of an array of welded plate heat exchangers of a material approved by ASME for pressure barrier (Ti-3Al-2.5V) in a hypothetical naval reactor. It was found it is feasible from thermal-hydraulic point of view and presents advantages over other steam generator designs

  9. 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.

  10. 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.

  11. Replacement steam generators for Calvert Cliffs, Oconee and future replacement design

    International Nuclear Information System (INIS)

    After the completion of steam generators presently being fabricated, a total of forty replacement steam generators will have been built for fourteen reactor units located at ten reactor sites. This represents approximately $1 billion of manufacture excluding installation costs. Replacement steam generator work began with the initiation of the Millstone 2 steam generator replacement program for Northeast Utilities in 1989. Manufacture is presently underway on replacement recirculating steam generators for Calvert Cliffs Units 1 and 2 plants of Constellation Nuclear (OEM Combustion Engineering) and the once-through steam generators for the Oconee 1, 2 and 3 plants of Duke Power (OEM Babcock and Wilcox). These two sites are the first and second respectively to have applied for and received approval for a life extension of 20 years beyond their original operating license. The application and granting of these license extensions reflects a major change in the nuclear industry over the recent past. The attitude to nuclear power has changed from a relatively defensive strategy to a much more optimistic agenda of utility reorganization, purchase of well performing older plants, replacement of aging components, plant refurbishment, and upgrades and applications for license extension. Possible new plants are also being considered. The paper discusses specific features, attributes, performance and operating experience with replacement steam generators (RSGs) both in service and under construction. Industry issues and design features applicable to future replacement steam generators are also reviewed. (author)

  12. The erection and commissioning of the Muelheim-Kaerlich nuclear power plant

    International Nuclear Information System (INIS)

    The contract for the Muelheim-Kaerlich nuclear power plant, equipped with a 1300 MWe pressurized water reactor, was awarded in 1973. The erection phase of the plant had been characterized in the mid-seventies by an aggravation of circumstances in connection with the nuclear energy controversy in the Federal Republic of Germany, the tightening of stipulations regarding safety philosophy, regulations and documentation, and by the consequences of the TMI accident in 1979. These led to considerable additional difficulties and delays. The commissioning phase on the other hand proceeded smoothly and speedily without major disturbances. The Muelheim-Kaerlich Nuclear Power Plant has some major technical features distinguishing it from other pressurized water reactor plants built in the Federal Republic of Germany. Its nuclear steam system is based on a license from the Babcock and Wilcox Company, USA, but it was adapted to German rules and regulations. The Muelheim-Kaerlich power plant is the first of this type and size built and put into operation. Its main technical features are described and, after a brief survey of the erection phase, the results of the start-up operations are discussed. (orig.)

  13. Characteristics of spent fuel, high-level waste, and other radioactive wastes which may require long-term isolation: Appendix 2A, Physical descriptions of LWR [Light-Water Reactor] fuel assemblies

    International Nuclear Information System (INIS)

    This appendix includes a four-page Physical Description report for each assembly type identified from the current data. Where available, a drawing of an assembly follows the appropriate Physical Description report. If no drawing is available for an assembly, a cross-reference to a similar assembly is provided if possible. For Advanced Nuclear Fuels, Babcock and Wilcox, Combustion Engineering, and Westinghouse assemblies, information was obtained via subcontracts with these fuel vendors. Data for some assembly types are not available. For such assemblies, the information shown in this report was obtained from the open literature and by inference from reload fuels made by other vendors. Efforts to obtain additional information are continuing. Individual Physical Description reports can be generated interactively through the menu-driven LWR Assemblies Data Base system. These reports can be viewed on the screen or directed to a printer. Special reports and compilations of specific data items can be produced on request

  14. 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)

    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 cm2 SBLOCA and Test 320201, a scaled 50 cm2 SBLOCA. Only MIST assessment results are presented in this paper

  15. 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.

  16. 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.

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

    International Nuclear Information System (INIS)

    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.

  18. 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

    2016-01-01

    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).

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

    International Nuclear Information System (INIS)

    Highlights: • 1980s enhanced oil recovery attempt investigated using isotope geochemistry in north-central Louisiana, USA. • CO2 injection was not the primary cause for increased microbial methanogenesis previously seen in the Olla Field. • Injected CO2 did not migrate uniformly within the reservoir. • Ideal geochemical conditions likely promoted methanogenesis in the Olla Field. - Abstract: 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 CO2 appears 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

  20. Correction to Wilcox et al. (2016).

    Science.gov (United States)

    2016-05-01

    Reports an error in "How being busy can increase motivation and reduce task completion time" by Keith Wilcox, Juliano Laran, Andrew T. Stephen and Peter P. Zubcsek (Journal of Personality and Social Psychology, 2016[Mar], Vol 110[3], 371-384). In the article, the affiliation of the author Andrew T. Stephen was incorrectly listed in the byline and the author note. The author is affiliated with the University of Oxford. The author note paragraph "Andrew T. Stephen is now at the University of Oxford" should have been omitted. All versions of this article have been corrected. (The following abstract of the original article appeared in record 2016-11945-002.) This research tests the hypothesis that being busy increases motivation and reduces the time it takes to complete tasks for which people miss a deadline. This effect occurs because busy people tend to perceive that they are using their time effectively, which mitigates the sense of failure people have when they miss a task deadline. Studies 1 and 2 show that when people are busy, they are more motivated to complete a task after missing a deadline than those who are not busy, and that the perception that one is using time effectively mediates this effect. Studies 3 and 4 show that this process makes busy people more likely to complete real tasks than people who are not busy. Study 5 uses data from over half a million tasks submitted by thousands of users of a task management software application to show that busy people take less time to complete a task after they miss a deadline for completing it. The findings delineate the conditions under which being busy can mitigate the negative effects of missing a deadline and reduce the time it takes to complete tasks. (PsycINFO Database Record PMID:27176772

  1. Locating leaking fuel rods in light water reactors

    International Nuclear Information System (INIS)

    Several techniques have been developed to perform the rod-by-rod leakage discrimination tests on nuclear fuel elements that rod replacement requires, including visual, vibrational analysis, eddy current and ultrasonic techniques. The ultrasonic technique has proved to have the most potential. It is the only system that in the field has provided a reliable, unambiguous indication of which fuel pins have leaked and which are intact without moving any fuel rods in the assembly. The through-transmission system is shown to be reliable and has been successfully used in many countries. It depends however, on specialised personnel to operate it and interpret the data. A new system, Echo-330, has been developed by Babcock and Wilcox. This is fully automated, and uses a multiple probe system with computerized control and data evaluation. The probe design is illustrated and typical output data shown. The time needed to locate leaking fuel rods is considerably reduced. (U.K.)

  2. Comparative Investigation of River Water Quality by OWQI, NSFWQI and Wilcox Indexes (Case study: the Talar River – IRAN)

    OpenAIRE

    Darvishi Gholamreza; Kootenaei Farshad Golbabaei; Ramezani Maedeh; Lotfi Eissa; Asgharnia Hosseinali

    2016-01-01

    Rivers are considered as one of the main resources of water supply for various applications such as agricultural, drinking and industrial purposes. Also, these resources are used as a place for discharge of sewages, industrial wastewater and agricultural drainage. Regarding the fact that each river has a certain capacity for acceptance of pollutants, nowadays qualitative and environmental investigations of these resources are proposed. In this study, qualitative investigation of the Talar riv...

  3. A benzenediamine derivate FC-99 attenuates lupus nephritis in MRL/lpr mice via inhibiting myeloid dendritic cell-secreted BAFF.

    Science.gov (United States)

    Ji, Jianjian; Xu, Jingjing; Li, Fanlin; Li, Xiaojing; Gong, Wei; Song, Yuxian; Dou, Huan; Hou, Yayi

    2016-05-01

    Myeloid dendritic cells (DCs) can produce B-cell-activating factor (BAFF) that modulates survival and differentiation of B cells and plays a pivotal role in the pathogenesis of systemic lupus erythematosus (SLE). Toll-like receptor 4 (TLR4) signaling has important functions in the process of BAFF production. Our previous study showed that a benzenediamine derivate FC-99 possesses anti-inflammation activity and directly interacts with interleukin-1 receptor-associated kinase 4 (IRAK4), which was a pivotal molecule in TLR4 signaling. In this study, we demonstrated that FC-99 attenuated lupus nephritis in the MRL/lpr mice. FC-99 also decreased the levels of total immunoglobulin G (IgG), total IgG2a and IgM in sera, as well as the activation of B cells in the spleens of MRL/lpr mice. Moreover, FC-99 inhibited abnormal activation of myeloid DCs in spleens and reduced the levels of BAFF in sera, spleens, and kidneys of MRL/lpr mice. Furthermore, upon TLR4 stimulation with lipopolysaccharide in vitro, FC-99 inhibited IRAK4 phosphorylation, as well as the activation and BAFF production in murine bone marrow-derived DCs. These data indicate that FC-99 attenuates lupus nephritis in MRL/lpr mice via inhibiting DC-secreted BAFF, suggesting that FC-99 may be a potential therapeutic candidate for the treatment of SLE. PMID:27121231

  4. Prevalence of latent eosinophilia among occupational gardeners at Babcock University, Nigeria

    Institute of Scientific and Technical Information of China (English)

    Ayodele Olushola Ilesanmi; Ginnikachi Jennifer Ekwe; Rosemary Isioma Ilesanmi; Damilola Temitope Ogundele; Jacob Kehinde Akintunde; Oluwasogo Adewole Olalubi

    2016-01-01

    Objective: To determine the level of eosinophils present in the blood and sputum samples, presumably as a result of continual occupational exposure to allergens while on duty, as gardeners at Babcock University, Nigeria. Methods: Haemocytometer and Olympus microscope were utilized to estimate eosino-phils population in 44 blood samples and 21 sputum samples respectively. Results: Relationship between the occurrence of eosinophil in blood and the exposure period among Babcock University gardeners had a positive correlation (r = + 0.08, t=4.55, P Conclusions: The nature and the gardening activities are not a risk factor that signifi-cantly affect eosinophil level but duration of exposure to allergens. However, all safety precautionary kits and wears should be enforced and embraced by the concerned occu-pational gardeners so as to avert and subvert its pre-disposing deleterious effect on them.

  5. Materials Ageing in Light Water Reactors - Handbook of Destructive Assays

    International Nuclear Information System (INIS)

    From the end of the 60's, LWRs have extensively been used in many countries around the world for electricity production. As in many other industrial facilities, some components failures have occurred during operation. This Handbook captures the results of some typical destructive examinations that have been carried out to understand and furthermore mitigate these failures. This Handbook is specific to PWRs from Western design, typically reactors supplied by Westinghouse Babcock and Wilcox, Combustion Engineering, Framatome (now Areva) and Mitsubishi Heavy Industries. However, some information on the history of BWRs failures along with some examples are also provided. This Handbook mainly addresses NSSS components destructive examinations as numerous typical balance of plant components failures have already been well addressed in Pierre Mousset's book written at the end of the 80's (Pierre Mousset, L'Expertise metallurgique appliquee aux centrales thermiques, electricite de France editions Kirk 1990). Furthermore, the materials addressed here are metallic materials, with a focus on the materials for which at least one example of destructive examination is presented in the Handbook. The treatment of fuel (and by extension the zirconium alloys) are out of the scope of this Handbook. The Handbook is organized by chapters. Following the introduction, the second chapter presents some LWRs basics. The third chapter gives some insights into the relevant failure mechanisms. Next, some properties of the materials having experienced field failures are presented. Regarding materials characteristics, the reference (Materials Handbook for Nuclear Plant Pressure Boundary Applications (2008). EPRI, Palo Alto, CA: 2008. 1016550) has provided valuable information. The destructive examinations results are sorted by the main field material issues such as: - Ni alloys PWSCC; - SG tubes issues; -Cold work SSSCC; -SS IASCC; - SS SCC in polluted environment or in occluded areas

  6. Converting the Audience: A Conversation with Agnes Wilcox

    Science.gov (United States)

    Becker, Becky

    2006-01-01

    This article presents a conversation with Agnes Wilcox, Executive Director of Prison Performing Arts in St. Louis, Missouri, about Prison Performing Arts. Although the average person might balk at the notion of interacting with prison inmates, finding it intimidating, worrisome, or self-sacrificial, for Wilcox, Prison Performing Arts is a…

  7. HYDRAULICS, WILCOX COUNTY, ALABAMA, USA

    Data.gov (United States)

    Federal Emergency Management Agency, Department of Homeland Security — Recent developments in digital terrain and geospatial database management technology make it possible to protect this investment for existing and future projects to...

  8. HYDROLOGY, WILCOX COUNTY, ALABAMA USA

    Data.gov (United States)

    Federal Emergency Management Agency, Department of Homeland Security — Hydrology data include spatial datasets and data tables necessary for documenting the hydrologic procedures for estimating flood discharges for a flood insurance...

  9. TERRAIN, WILCOX COUNTY, ALABAMA USA

    Data.gov (United States)

    Federal Emergency Management Agency, Department of Homeland Security — Terrain data, as defined in FEMA Guidelines and Specifications, Appendix N: Data Capture Standards, describes the digital topographic data that was used to create...

  10. FLOODPLAIN, WILCOX COUNTY, ALABAMA USA

    Data.gov (United States)

    Federal Emergency Management Agency, Department of Homeland Security — The Floodplain Mapping/Redelineation study deliverables depict and quantify the flood risks for the study area. The primary risk classifications used are the...

  11. The effect of aging upon CE and B and W control rod drives

    International Nuclear Information System (INIS)

    The effect of aging upon the Babcock and Wilcox and Combustion Engineering control rod drive systems has been evaluated as part of the US Nuclear Regulatory Commission Nuclear Plant Aging Research program. Operating experience data for the 1980-1990 time period was reviewed to identify predominant failure modes, causes, and effects. These results, in conjunction with an assessment of component materials and operating environments, conclude that both systems are susceptible to age degradation. System failures have resulted in significant plant effects, including power reductions, plant shutdowns, scrams, and engineered safety feature actuation. Current industry inspection and maintenance practices were assessed. Some of these practices effectively address aging, while others do not

  12. Comparing dynamic responses of recirculating and once-through steam generators for next generation LWRs

    International Nuclear Information System (INIS)

    In this paper two types of steam generators are under consideration for next-generation (pressurized) light water reactors: a recirculating type and a once-through type. The steady-state and dynamic characteristics of these steam generators were compared to facilitate optimization of a particular reactor system design. To compare, the dynamic responses of the two types, as indicated by the feedwater flow, steam generator level, steam flow, steam pressure, steam enthalpy, primary-side pressure and cold-leg temperature, were assessed using Babcock and Wilcox's Modular Modeling System. The once-through steam generator showed a tremendous flexibility to produce superheated steam under diverse conditions (i.e., constant or variable steam throttle pressure and constant or variable average primary temperature) with excellent speed and accuracy in following the load demand. Since the primary and steam sides are closely coupled with the feedwater, the pressurizer should be sized liberally to lessen the sensitivity of the primary response to feedwater upsets and the reliability of the feedwater train should be enhanced. In contrast, the recirculating steam generator must be operated with variable steam throttle pressure and variable primary average temperature, and the speed and accuracy of following the load demand are not as good. While the recirculation provides an effective cushion for the primary and steam sides from feedwater upsets, it also amplifies the level response caused by upsets in steam pressure and feedwater temperature affecting the level controllability and moisture separation performance. The recirculating steam generator should be designed to incorporate features to improve level controllability by constant-inventory control strategy. Also to survive a reactor-coolant pump trip, the design with one reactor-coolant pump per loop should be considered

  13. A 3D Babcock-Leighton Solar Dynamo Model

    CERN Document Server

    Miesch, Mark S

    2014-01-01

    We present a 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 2D 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 (2D in radius/latitude) and surface flux transport models (2D in latitude/longitude) into a more self-consistent framework that captures 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-p...

  14. Comment on a Wilcox Test Statistic for Comparing Means When Variances Are Unequal.

    Science.gov (United States)

    Hsiung, Tung-Hsing; And Others

    1994-01-01

    The alternative proposed by Wilcox (1989) to the James second-order statistic for comparing population means when variances are heterogeneous can sometimes be invalid. The degree to which the procedure is invalid depends on differences in sample size, the expected values of the observations, and population variances. (SLD)

  15. Growing Readers: Wendy Wilcox--West Bloomfield Township Public Library, MI

    Science.gov (United States)

    Library Journal, 2005

    2005-01-01

    In 2001 youth services librarian Wendy Wilcox begged her boss for the chance to make West Bloomfield Township Public Library (WBTPL) one of 20 demonstration sites for the Public Library Association (PLA)/Association for Library Service to Children initiative Every Child Ready To Read. While all participating libraries teach parents and caregivers…

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

    International Nuclear Information System (INIS)

    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

  17. 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.

  18. Factory acceptance of the compressor skids at Samifi-Babcock. All pictures show the second stage compressor skid.

    CERN Multimedia

    G. Perinic

    2001-01-01

    Most recent pictures taken during the factory acceptance of the compressor skids at Samifi-Babcock. All pictures show the second stage compressor skid. Picture two was taken during the leak tests and shows all the pockets around flanges and valves.

  19. Green County Nuclear Power Plant. License application

    International Nuclear Information System (INIS)

    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)

  20. Reactor containment and reactor safety in the United States

    International Nuclear Information System (INIS)

    The reactor safety systems of two reactors are studied aiming at the reactor containment integrity. The first is a BWR type reactor and is called Peachbottom 2, and the second is a PWR type reactor, and is called surry. (E.G.)

  1. The CAREM reactor and present currents in reactor design

    International Nuclear Information System (INIS)

    INVAP has been working on the CAREM project since 1983. It concerns a very low power reactor for electrical energy generation. The design of the reactor and the basic criteria used were described in 1984. Since then, a series of designs have been presented for reactors which are similar to CAREM regarding the solutions presented to reduce the chance of major nuclear accidents. These designs have been grouped under different names: Advanced Reactors, Second Generation Reactors, Inherently Safe Reactors, or even, Revolutionary Reactors. Every reactor fabrication firm has, at least, one project which can be placed in this category. Presently, there are two main currents of Reactor Design; Evolutionary and Revolutionary. The present work discusses characteristics of these two types of reactors, some revolutionary designs and common criteria to both types. After, these criteria are compared with CAREM reactor design. (Author)

  2. Buoyancy-induced time delays in Babcock-Leighton flux-transport dynamo models

    Science.gov (United States)

    Jouve, L.; Proctor, M. R. E.; Lesur, G.

    2010-09-01

    Context. The Sun is a magnetic star whose cyclic activity is thought to be linked to internal dynamo mechanisms. A combination of numerical modelling with various levels of complexity is an efficient and accurate tool to investigate such intricate dynamical processes. Aims: We investigate the role of the magnetic buoyancy process in 2D Babcock-Leighton dynamo models, by modelling more accurately the surface source term for poloidal field. Methods: To do so, we reintroduce in mean-field models the results of full 3D MHD calculations of the non-linear evolution of a rising flux tube in a convective shell. More specifically, the Babcock-Leighton source term is modified to take into account the delay introduced by the rise time of the toroidal structures from the base of the convection zone to the solar surface. Results: We find that the time delays introduced in the equations produce large temporal modulation of the cycle amplitude even when strong and thus rapidly rising flux tubes are considered. Aperiodic modulations of the solar cycle appear after a sequence of period doubling bifurcations typical of non-linear systems. The strong effects introduced even by small delays is found to be due to the dependence of the delays on the magnetic field strength at the base of the convection zone, the modulation being much less when time delays remain constant. We do not find any significant influence on the cycle period except when the delays are made artificially strong. Conclusions: A possible new origin of the solar cycle variability is here revealed. This modulated activity and the resulting butterfly diagram are then more compatible with observations than what the standard Babcock-Leighton model produces.

  3. Replacement of shutdown cooling system and repair of reactor pressure vessel nozzle welds at NPP Forsmark unit 1 and unit 2

    International Nuclear Information System (INIS)

    The Forsmark Nuclear Power Plant is located about 150 km north of Stockholm. The plant consists of three units with boiling water reactors. Unit 1 and Unit 2 were put into operation in 1981 and 1982, respectively. Both of these units are identical each having a capacity of 970 MW. Unit 3 was completed in 1985 and has a capacity of 1160 MW. In November 1998 Babcock Noell Nuclear was awarded the contract to replace the pipe-work of the two-sectioned Shutdown Cooling System 321 from the nozzles at the reactor pressure vessel to 10 meters outside the containment. Moreover, the inner and outer isolation valves including the penetrations had to be replaced. Finally, the repair of the RPV (reactor pressure vessel) connecting welds of the System 415 (Feed Water) and System 323 (Emergency Cooling) was to be performed. The work was carried out by a Babcock Noell Nuclear team integrating Swedish companies during the outages May/June 2000 in Forsmark 2 and August/September 2000 in Forsmark 1. In the Forsmark Nuclear Power Plant, Units 1 and 2, 19 RPV nozzle connections were improved successfully. All relevant start-up deadlines could be kept. All new tools and manipulators met the stringent project requirements. The mockup qualification of the equipment and the special personnel training performed in advance proved that such challenging work can be managed despite limited preparation time and planned effectively in order to recognize and avoid possible risks. (authors)

  4. Welding and reactor safety

    International Nuclear Information System (INIS)

    The high safety requirements which must be demanded of the quality of the welded joints in reactor technique have so far not been fulfilled in all cases. The errors occuring have caused considerable loss of availability and high material costs. They were not, however, so serious that one need have feared any immediate danger to the personnel or to the environment. The safety devices of reactor plants were only called upon in a few cases and to these they responded perfectly. The intensive efforts to complete and improve the specifications are to contribute to that in future, the reactor plants can be counted even more so as one of the safest technical plants ever. (orig./LH)

  5. Material and fabrication of the HTTR reactor pressure vessel

    International Nuclear Information System (INIS)

    The High Temperature Engineering Test Reactor (HTTR) is under construction at Oarai Research Establishment, Japan Atomic Energy Research Institute (JAERI) and planned to be critical in October 1997. Fabrication of the HTTR reactor pressure vessel (RPV) at Kure Works, Babcock-Hitachi K.K. took about two years, and the RPV was transported to the Oarai site in August 1994. Pressure test of the primary and secondary cooling system including the RPV was performed successfully in March 1996. Because temperature of the HTTR RPV becomes about 400 deg. C at normal operation, 2 1/4 Cr-1 Mo steel is chosen for it. Fluence of the RPV is calculated to be less than 1 X 1017 n/cm2 (E>l MeV), and so irradiation embrittlement, is presumed to be negligible, but temper embrittlement is not. For the purpose of reducing embrittlement, content of some elements is limited on 2 1/4 Cr-1 Mo steel for the HTTR RPV using embrittlement parameters: J-factor and X-bar. In this paper design and structure of the HTTR RPV is briefly reviewed first. Fabrication procedure of the RPV and its special feature is shown. Material data on 2 1/4 Cr-1 Mo steel manufactured for the RPV, especially the embrittlement parameters J-factor and X-bar, and nil-ductility transition temperatures TNDT by drop weight tests, are shown, and increase in the transition temperature is estimated based on data available in literature. Technology of the HTTR RPV is applicable to RPVs of future commercial High Temperature Gas-cooled Reactors (HTGRs). (author)

  6. The influence of solar wind on extratropical cyclones – Part 1: Wilcox effect revisited

    Directory of Open Access Journals (Sweden)

    M. Rybanský

    2009-01-01

    Full Text Available A sun-weather correlation, namely the link between solar magnetic sector boundary passage (SBP by the Earth and upper-level tropospheric vorticity area index (VAI, that was found by Wilcox et al. (1974 and shown to be statistically significant by Hines and Halevy (1977 is revisited. A minimum in the VAI one day after SBP followed by an increase a few days later was observed. Using the ECMWF ERA-40 re-analysis dataset for the original period from 1963 to 1973 and extending it to 2002, we have verified what has become known as the "Wilcox effect" for the Northern as well as the Southern Hemisphere winters. The effect persists through years of high and low volcanic aerosol loading except for the Northern Hemisphere at 500 mb, when the VAI minimum is weak during the low aerosol years after 1973, particularly for sector boundaries associated with south-to-north reversals of the interplanetary magnetic field (IMF BZ component. The "disappearance" of the Wilcox effect was found previously by Tinsley et al. (1994 who suggested that enhanced stratospheric volcanic aerosols and changes in air-earth current density are necessary conditions for the effect. The present results indicate that the Wilcox effect does not require high aerosol loading to be detected. The results are corroborated by a correlation with coronal holes where the fast solar wind originates. Ground-based measurements of the green coronal emission line (Fe XIV, 530.3 nm are used in the superposed epoch analysis keyed by the times of sector boundary passage to show a one-to-one correspondence between the mean VAI variations and coronal holes. The VAI is modulated by high-speed solar wind streams with a delay of 1–2 days. The Fourier spectra of VAI time series show peaks at periods similar to those found in the solar corona and solar wind time series. In the modulation of VAI by solar wind the IMF BZ seems to control the phase of the Wilcox effect and the depth of the VAI minimum. The

  7. 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.

  8. VVER and RBMK reactors

    International Nuclear Information System (INIS)

    The safety of VVER and RBMK reactors has been discussed a lot after Chernobyl accident. Some improvements have been performed since that especially in RBMK-reactors and extensive programmes for backfitting have been planned and are partly underway. There are two different sizes of VVER reactors, 440 MW and 1000 MW. The design bases and designs itself vary inside the family of two size classes depending on the age of the plant. The oldest VVER-440 is called model 230 and the newest model 213. The oldest VVER-1000 units (two units) are prototypes that have some unique, nonfavorable features. The next stage of VVER-1000 developement (three units) is model V-302 and the remaining 15 plants in operation are model V-320, but even within this latest model there are some differences. The design bases and designs vary also inside the family of the RBMK reactors exactly the same way as in VVERs. The most important design bases of nuclear power plants designed in the former Soviet Union is presented in this paper. Also some safety advantages and disadvantages of these NPPs are discussed. (au). (5 figs.)

  9. BODYFIT-2PE-HEM: LWR core thermal-hydraulic code using boundary-fitted coordinates and two-phase homogeneous equilibrium model. Volume 3: validation and applications

    International Nuclear Information System (INIS)

    The BODYFIT-2PE-HEM code was used to simulate several Pressurized Water Reactor (PWR) and Boiling Water Reactor (BWR) types of experiments to validate its applicability and accuracy. Five simulations are reported in this volume. The first comparison was between the closed form analytical solution and the BODYFIT calculation of 3-D flows in an inifinite square array of circular tubes. Both the velocity profiles along symmetry lines and Nusselt numbers as a function of the entrance distance were given in the report. The second simulation was on the Columbia University 4 x 4 rod bundle experiment with a power skew of 2 to 1. The calculated mass flow rates and qualities for both hot and cold subchannels at the exit of the rod bundle were compared with the experimental isokinetic measurements. The third simulation was on the Babcock and Wilcox 4 x 6 rod bundle experiments with a power skew of 1.5 to 1. Again, the calculated mass flow rates and qualities for both hot and cold subchannels at the exit of the rod bundle were compared with the experimental isokinetic measurements. The fourth simulation was on the Westinghouse 4 x 5 rod bundle critical heat flux experiments and transient pressure drop tests. In this simulation, the critical heat fluxes calculated by the code with several CHF correlations were compared with the experimental measurements. Furthermore, the pressure drops, as a function of time, were compared with the experimental values for the flow rundown transients. The fifth simulation was on the GE 3 x 3 CHF experiments. Many operating conditions with different inlet temperatures, inlet velocities, and system pressures were used in the experiments. Code calculations were based on the Biasi correlation and the Columbia University correlation. Comparisons between calcuations and measurements show good agreements, demonstrating the validity and accuracy of the BODYFIT-2PE-HEM code. 14 refs., 36 figs., 11 tabs

  10. A Three-Dimensional Babcock-Leighton Solar Dynamo Model: Initial Results with Axisymmetric Flows

    CERN Document Server

    Miesch, Mark S

    2015-01-01

    The main objective of this paper is to introduce the STABLE (Surface flux Transport And Babcock-LEighton) solar dynamo model. STABLE is a 3D Babcock-Leighton/Flux Transport dynamo model in which the source of poloidal field is the explicit emergence, distortion, and dispersal of bipolar magnetic regions (BMRs). Here we describe the STABLE model in more detail than we have previously and we verify it by reproducing a 2D mean-field benchmark. We also present some representative dynamo simulations, focusing on the special case of kinematic magnetic induction and axisymmetric flow fields. Not all solutions are supercritical; it can be a challenge for the BL mechanism to sustain the dynamo when the turbulent diffusion near the surface is $\\geq 10^{12}$ cm$^2$ s$^{-1}$. However, if BMRs are sufficiently large, deep, and numerous, then sustained, cyclic, dynamo solutions can be found that exhibit solar-like features. Furthermore, we find that the shearing of radial magnetic flux by the surface differential rotation ...

  11. The profile of tuberculosis infection at the Babcock University Teaching Hospital

    Directory of Open Access Journals (Sweden)

    Shobowale E.O

    2016-01-01

    Full Text Available 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 Babcock University Teaching Hospital. Methods: This was a retrospective cross sectional study of patients presenting to the Tuberculosis Laboratory of Babcock University Teaching Hospital. Results: Patients presenting to BUTH were 2.29 times more likely to have a positive AFB sputum smear result when compared to samples from Primary Health Care Centers – P = 0.05, χ 2 = 3.83, O.R = 2.29, R.R = 1.17, CI = 1.0 – 5.34. Patients presenting to BUTH were more likely to be HIV positive when compared to those from PHC’s p = 0.00, χ 2 = 24.74, df = 2. Conclusion: The burden of tuberculosis is still high in our environment and challenges in its rapid and accurate diagnosis still remain. In order to strengthen tuberculosis control, attention needs to be placed on rapid diagnosis and prompt treatment.

  12. Buoyancy-induced time delays in Babcock-Leighton flux-transport dynamo models

    CERN Document Server

    Jouve, L; Lesur, G

    2010-01-01

    The Sun is a magnetic star whose cyclic activity is thought to be linked to internal dynamo mechanisms. A combination of numerical modelling with various levels of complexity is an efficient and accurate tool to investigate such intricate dynamical processes. We investigate the role of the magnetic buoyancy process in 2D Babcock-Leighton dynamo models, by modelling more accurately the surface source term for poloidal field. Methods. To do so, we reintroduce in mean-field models the results of full 3D MHD calculations of the non-linear evolution of a rising flux tube in a convective shell. More specifically, the Babcock-Leighton source term is modified to take into account the delay introduced by the rise time of the toroidal structures from the base of the convection zone to the solar surface. We find that the time delays introduced in the equations produce large temporal modulation of the cycle amplitude even when strong and thus rapidly rising flux tubes are considered. Aperiodic modulations of the solar cy...

  13. Effect of diameter and geometry on two-phase flow regimes and carry-over in a model pwr hot leg

    International Nuclear Information System (INIS)

    This paper describes a series of tests investigating two-phase flow characterization and carry-over in a transparent model of a Babcock and Wilcox (BandW) pressurized water reactor (PWR) hot leg geometry. This work was performed, in part, to support the interpretation of results from the Once-Through Integral System (OTIS) and Multiloop Integral Test (MIST) facilities. Test conditions were selected to cover a wide range of gas and liquid superficial velocities expected to occur in a prototypical reactor geometry during a small break loss of coolant accident (SBLOCA). Tests at high gas superficial velocities were also performed for comparison with semi-analytical predictions. Tests were conducted in a test rig with 30.5-cm (12-inch) diameter pipe. Results include average void fraction, amount of water carry-over through the U-bend and a description of the two-phase flow phenomena. Results of these tests indicate that slug flow is not observed in large diameter pipes. Instead, as the air flow rate is increased, the flow regime progresses from bubbly to churn-type flow with the presence of large bubbles (approximately 15-cm diameter). The results also indicate that flow regimes and collapsed liquid level are more strongly dependent on air superficial velocity than the water superficial velocity and that the amount of water carry-over for a given air flow rate is a strong function of collapsed water level. Furthermore, the results show that similar thresholds for breakdown in natural circulation flow exist between small and large diameter pipes for gas and liquid superficial velocities expected in a SBLOCA

  14. Reactor physics and economic aspects of the CANDU reactor system

    International Nuclear Information System (INIS)

    A history of the development of the CANDU system is given along with a fairly detailed description of the 600 MW(e) CANDU reactor. Reactor physics calculation methods are described, as well as comparisons between calculated reactor physics parameters and those measured in research and power reactors. An examination of the economics of CANDU in the Ontario Hydro system and a comparison between fossil fuelled and light water reactors is presented. Some physics, economics and resources aspects are given for both low enriched uranium and thorium-fuelled CANDU reactors. Finally the RβD program in Advanced Fuel Cycles is briefly described

  15. Modern research reactors in the world and RA research reactor

    International Nuclear Information System (INIS)

    This paper covers the following topics: fundamentals of research reactors, thermal neutron flux density, classification of research reactors in the world, properties of research reactors of higher power in the world according to IAEA data for 1995, their application, and trend of development, experimental feasibility and status of RA reactor. Trend of research reactors development in the world (after 1980) is directed towards increasing the neutron production quality factor, i.e. ratio between thermal neutron flux density and reactor power, which is achieved by designing compact reactor cores. With the aim of renewal of RA reactor (without analysis of reactor components and staff aging, possibility of restart and commercialization), according to the analysis in this paper, it can be concluded: there is very few reactors under construction in the world, all the important countries in Europe have research reactors; RA reactor is not very interesting for development of reactor physics; nowadays RA reactor is in the group of reactors which are 30-40 years old; its inventories of fuel and heavy water are enough for about 20 years of operation; it has achieved high quality factor of neutron production with low and highly enriched fuel; core transfer from low highly enriched to low enriched fuel should be carefully studies from operation, experimental and economical point of view; it is necessary to use the advantages of RA reactor (minimum investment): volume of the core and reflector which enables availability of neutron flux for the users (numerous experimental loops), fuel in shape of slugs enabling efficient fuel management and flexible neutron flux distribution in the core in the reflector, reactor operation should be directed towards commercial applications. Bibliography of more than 140 relevant papers used is included in this paper

  16. Safety systems of heavy water reactors and small power reactors

    International Nuclear Information System (INIS)

    After introductional descriptions of heavy water reactors and natural circulation boiling water reactors the safety philosophy and safety systems like ECCS, residual heat removal, protection systems etc., are described. (RW)

  17. The IRIS network site at the Wilcox Solar Observatory

    Science.gov (United States)

    Hoeksema, J. T.; Scherrer, P. H.

    1991-01-01

    The site for the International Research on the Interior of the Sun (IRIS) instrument housed at the Wilcox Solar Observatory at Stanford University (near San Francisco, USA) is described together with the instrument operation procedure. The IRIS instrument, which measures global oscillations of the sun, operates continuously every clear day since it was installed in August 1987.

  18. The effect of salmeterol and salbutamol on mediator release and skin responses in immediate and late phase allergic cutaneous reactions

    DEFF Research Database (Denmark)

    Petersen, Lars Jelstrup; Skov, P S

    1999-01-01

    clinical and biochemical EAR and LPR in human skin. METHODS: Measurement of wheal and flare reactions to allergen, codeine, and histamine, and LPR (induration) to allergen. Assessment of histamine and prostaglandin D2 (PGD2) release by microdialysis technique in EAR, and measurement of mediators in LPR by...

  19. Reactor regulating and protection system for a light water reactor

    International Nuclear Information System (INIS)

    Microprocessor based systems are developed for reactor regulation and protection of LWR. A triple modular redundancy approach is followed for the design of this system. This system is functionally partitioned into two sub-systems - Reactor Regulating System (RRS) and Reactor Trip Logic System (RTLS). RRS controls the reactor power as per demand and RTLS generates the reactor trip on abnormal process conditions. This paper describes the details of RRS and RTLS system architecture and fault tolerant and fail-safe features used in the system design. (author)

  20. Fast reactors and nonproliferation

    International Nuclear Information System (INIS)

    1.Three aspects of nonproliferation relevant to nuclear power are: Pu buildup in NPP spent fuel cooling ponds (∼ 104 t in case of consumption of ∼ 107 t cheap uranium). Danger of illegal radiochemical extraction of Pu for weapons production; Pu extraction from NPP fuel at the plants available in nuclear countries, its burning along with weapon-grade Pu in NPP reactors or in special-purpose burners; increased hazard of nuclear weapons sprawl with breeders and closed fuel cycle technology spreading all over the world. 2.The latter is one of major obstacles to creation of large-scale nuclear power. 3.Nuclear power of the first stage using 235 U will be able to meet the demands of certain fuel-deficient countries and regions, replacing ∼ 5-10% of conventional fuels in the global consumption for a number of decades. 4.Fast reactors of the first generation and the currently employed fuel technology are far from exhausting their potential for solving economic problems and meeting the challenges of safety, radioactive waste and nonproliferation. Development of large-scale nuclear power will become an option accepted by society for solving energy problems in the following century, provided a breeder technology is elaborated and demonstrated in the next 15-20 years, which would comply with the totality of the following requirement: full internal Pu breeding deterministic elimination of severe accidents involving fuel damage and high radioactivity releases: fast runaway, loss of coolant, fires, steam and hydrogen explosions, etc.; reaching a balance between radioactive wastes disposed of and uranium mined in terms of radiation hazard; technology of closed fuel cycle preventing its use for Pu extraction and permitting physical protection from fuel thefts;economic competitiveness of nuclear power for most of countries and regions, i.e. primarily the cost of NPPs with fat reactors is to be below the cost of modern LWR plants, etc

  1. 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.

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

    International Nuclear Information System (INIS)

    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-He3 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

  3. Safety review, assessment and inspection on research reactors, experimental reactors, nuclear heating reactors and critical facilities

    International Nuclear Information System (INIS)

    In 1998, the NNSA organized to complete the nuclear safety review on the test loop in-reactor operation of the High-flux Engineering Experimental Reactor (HFEER) and the re-operation of the China Pulsed Reactor and the Uranium-water Criticality Facility. The NNSA conducted the nuclear safety review on the CP application of the China Experimental Fast Reactor (CEFR) and the siting of China Advanced Research Reactor (CARR), and carried out the construction supervision on HTR-10, and dealt with the event about the technological tube breakage of HWRR and other events

  4. An integrated maintenance strategy for the Babcock 10E Coal Mill

    Energy Technology Data Exchange (ETDEWEB)

    MacIntyre, J. [University of Sunderland (United Kingdom). Centre for Adaptive Systems; Stansfield, D.; Allot, P.; Harris, M. [National Power plc (United Kingdom)

    1998-07-01

    Coal-fired power station around the world have many common features, including similar types of auxiliary plant. One example of such a common area is the coal milling plant. This paper describes how an integrated approach to maintenance of the Babcock 10E Coal Mill has been developed at National Power`s Blyth `B` Station on the North East coast of England. The paper gives details of the types of mechanical problems experienced with the plant, and the various engineering, maintenance, and monitoring strategies which have been integrated into a comprehensive and effective maintenance strategy for this plant. The paper also gives detailed examples of the application of these techniques, the results obtained from them, and goes on to show how this integrated approach has reaped substantial rewards for the Station in terms of availability, reliability and profitability. (author)

  5. Alternative approaches to fusion. [reactor design and reactor physics for Tokamak fusion reactors

    Science.gov (United States)

    Roth, R. J.

    1976-01-01

    The limitations of the Tokamak fusion reactor concept are discussed and various other fusion reactor concepts are considered that employ the containment of thermonuclear plasmas by magnetic fields (i.e., stellarators). Progress made in the containment of plasmas in toroidal devices is reported. Reactor design concepts are illustrated. The possibility of using fusion reactors as a power source in interplanetary space travel and electric power plants is briefly examined.

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

    Indian Academy of Sciences (India)

    Dibyendu Nandy; Arnab Rai Choudhuri

    2000-09-01

    We study the effects of incorporating magnetic buoyancy in a model of the solar dynamo—which draws inspiration from the Babcock-Leighton idea of surface processes generating the poloidal field. We present our main results here.

  7. 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).

  8. Fast reactors and nuclear nonproliferation

    International Nuclear Information System (INIS)

    Problems are discussed with regard to nuclear fuel cycle resistance in fast reactors to nuclear proliferation risk due to the potential for use in military programs of the knowledge, technologies and materials gained from peaceful nuclear power applications. Advantages are addressed for fast reactors in the creation of a more reliable mode of nonproliferation in the closed nuclear fuel cycle in comparison with the existing fully open and partially closed fuel cycles of thermal reactors. Advantages and shortcomings are also discussed from the point of view of nonproliferation from the start with fast reactors using plutonium of thermal reactor spent fuel and enriched uranium fuel to the gradual transition using their own plutonium as fuel. (author)

  9. Computerized reactor monitor and control for nuclear reactors

    International Nuclear Information System (INIS)

    The analysis of a computerized process control system developed by Transelektro-KFKI-Videoton (Hangary) for a twenty-year-old research reactor in Budapest and or a new one in Tajura (Libya) is given. The paper describes the computer hardware (R-10) and the implemented software (PROCESS-24K) as well as their applications at nuclear reactors. The computer program provides for man-machine communication, data acquisition and processing, trend and alarm analysis, the control of the reactor power, reactor physical calculations and additional operational functions. The reliability and the possible further development of the computerized systems which are suitable for application at reactors of different design are also discussed. (Sz.J.)

  10. 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

  11. RA reactor operation and maintenance

    International Nuclear Information System (INIS)

    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

  12. Reactor containment research and development

    Energy Technology Data Exchange (ETDEWEB)

    Weil, N. A.

    1963-06-15

    An outline is given of containment concepts, sources and release rates of energy, responses of containment structures, effects of projectiles, and leakage rates of radioisotopes, with particular regard to major reactor accidents. (T.F.H.)

  13. Safety review, assessment and inspection on research reactors, experimental reactors, nuclear heating reactors and critical facilities

    International Nuclear Information System (INIS)

    The NNSA organized mainly in 1999 to complete the verification loop in core of the high flux experimental reactor with the 2000 kW fuel elements, the re-starting of China Pulsed Reactor, review and assessment on nuclear safety for the restarting of the Uranium-water critical Facility and treat the fracture event with the fuel tubes in the HWRR

  14. Supply strategy for SMR deployment

    International Nuclear Information System (INIS)

    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)

  15. Lupus eritematoso sistémico en ratones MRL lpr/lpm y knockouts del receptor de quimioquina CCR2

    OpenAIRE

    Camarasa Lillo, Natalia

    2009-01-01

    INTRODUCCIÓN El lupus eritematoso sistémico es una enfermedad autoinmune cuya principal manifestación y debut de la enfermedad es la glomerulonefritis mediada por complejos inmunes. Los ratones MRL/MpJ-Fas lpr/J (MRL/lpr) llevan una mutación en el gen Fas de la apoptosis que da lugar a una proliferación de linfocitos autoreactivos y son considerados un modelo de ratón que reproduce muy bien la enfermedad lúpica en el humano, con linfadenopatía asociada a proliferación aberrante de células T,...

  16. Nuclear reactor PBMR and cogeneration

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

    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

  18. Research reactors: design, safety requirements and applications

    International Nuclear Information System (INIS)

    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)

  19. Wilcox 1:100000 Quad Transportation 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...

  20. 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...

  1. Training and Certification of Research Reactor Personnel

    International Nuclear Information System (INIS)

    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)

  2. PARR-2: reactor description and experiments

    International Nuclear Information System (INIS)

    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.)

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

    International Nuclear Information System (INIS)

    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)

  4. The need to address the larger universe of HEU-fueled reactors, including: Critical assemblies, pulsed reactors and propulsion reactors

    International Nuclear Information System (INIS)

    The RERTR program has focused on ending shipments of HEU fuel to research reactors. Highest priority has been given to reactors with steady thermal powers ≥ 1 megawatt. Since the cores of critical assemblies and pulsed reactors can contain huge amounts of HEU, they should be a second focus. Also, since many aging and specialized HEU-fuelled reactors may no longer be needed, more emphasis should be given to initiatives that could assist in their shutdown and decommissioning, including providing access to regional reactors with superior facilities. HEU-fuelled ship-propulsion reactors should also be addressed. Russia's civilian icebreaker reactors are of particular interest because their fuel design is considered less sensitive than that of naval reactor fuel. Moreover, Russia's KLT-40 icebreaker reactor is being adapted for a floating nuclear power plant and LEU icebreaker fuel could be used for converting Russian research reactors such as PIK and SM-3, that operate at power-reactor temperatures. (author)

  5. 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.

  6. Reactor core and fueling method

    International Nuclear Information System (INIS)

    When MOX fuel assemblies are used in a portion of fuel assembly of a BWR type nuclear reactor, neutron spectra are hardened due to the difference of the nuclear property between uranium and plutonium. As a result, the reactivity controllability of burnable poisons such as gadolinia is lowered, and the multiplication factor of the MOX fuel assembly at the initial stage of burning is increased greater than that of an uranium fuel assembly, to reduce thermal margin and reactor shutdown margin. Then, in the present invention, fresh fuel assemblies containing plutonium are disposed in a first region at the second layer from the outermost circumference of the reactor core and in a second region in adjacent with a control cell. Since the MOX fuel assemblies with increasing reactivity are disposed in the first and the second regions of small neutron importance, the power at the periphery of the reactor core and the circumference of the control cell can be kept substantially constant throughout the operation period. Further, satisfactory reactor operation can be kept without causing excess distortion of power distribution. (N.H.)

  7. 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)

  8. Nuclear reactors and fuel cycle

    International Nuclear Information System (INIS)

    The Center for Nuclear Engineering has shown expertise in the field of nuclear and energy systems ad correlated areas. Due to the experience obtained over decades in research and technological development at Brazilian Nuclear Program personnel has been trained and started to actively participate in the design of the main system that will compose the Brazilian Multipurpose Reactor (RMB) which will make Brazil self-sufficient in the 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. The Nuclear Fuel Center is responsible for the production of the nuclear fuel necessary for the continuous operation of the IEA-R1 research reactor. Development of new fuel technologies is also a permanent concern

  9. A comparison of three self-tuning control algorithms developed for the Bristol-Babcock controller

    Energy Technology Data Exchange (ETDEWEB)

    Tapp, P.A.

    1992-04-01

    A brief overview of adaptive control methods relating to the design of self-tuning proportional-integral-derivative (PID) controllers is given. The methods discussed include gain scheduling, self-tuning, auto-tuning, and model-reference adaptive control systems. Several process identification and parameter adjustment methods are discussed. Characteristics of the two most common types of self-tuning controllers implemented by industry (i.e., pattern recognition and process identification) are summarized. The substance of the work is a comparison of three self-tuning proportional-plus-integral (STPI) control algorithms developed to work in conjunction with the Bristol-Babcock PID control module. The STPI control algorithms are based on closed-loop cycling theory, pattern recognition theory, and model-based theory. A brief theory of operation of these three STPI control algorithms is given. Details of the process simulations developed to test the STPI algorithms are given, including an integrating process, a first-order system, a second-order system, a system with initial inverse response, and a system with variable time constant and delay. The STPI algorithms` performance with regard to both setpoint changes and load disturbances is evaluated, and their robustness is compared. The dynamic effects of process deadtime and noise are also considered. Finally, the limitations of each of the STPI algorithms is discussed, some conclusions are drawn from the performance comparisons, and a few recommendations are made. 6 refs.

  10. A comparison of three self-tuning control algorithms developed for the Bristol-Babcock controller

    Energy Technology Data Exchange (ETDEWEB)

    Tapp, P.A.

    1992-04-01

    A brief overview of adaptive control methods relating to the design of self-tuning proportional-integral-derivative (PID) controllers is given. The methods discussed include gain scheduling, self-tuning, auto-tuning, and model-reference adaptive control systems. Several process identification and parameter adjustment methods are discussed. Characteristics of the two most common types of self-tuning controllers implemented by industry (i.e., pattern recognition and process identification) are summarized. The substance of the work is a comparison of three self-tuning proportional-plus-integral (STPI) control algorithms developed to work in conjunction with the Bristol-Babcock PID control module. The STPI control algorithms are based on closed-loop cycling theory, pattern recognition theory, and model-based theory. A brief theory of operation of these three STPI control algorithms is given. Details of the process simulations developed to test the STPI algorithms are given, including an integrating process, a first-order system, a second-order system, a system with initial inverse response, and a system with variable time constant and delay. The STPI algorithms' performance with regard to both setpoint changes and load disturbances is evaluated, and their robustness is compared. The dynamic effects of process deadtime and noise are also considered. Finally, the limitations of each of the STPI algorithms is discussed, some conclusions are drawn from the performance comparisons, and a few recommendations are made. 6 refs.

  11. A comparison of three self-tuning control algorithms developed for the Bristol-Babcock controller

    International Nuclear Information System (INIS)

    A brief overview of adaptive control methods relating to the design of self-tuning proportional-integral-derivative (PID) controllers is given. The methods discussed include gain scheduling, self-tuning, auto-tuning, and model-reference adaptive control systems. Several process identification and parameter adjustment methods are discussed. Characteristics of the two most common types of self-tuning controllers implemented by industry (i.e., pattern recognition and process identification) are summarized. The substance of the work is a comparison of three self-tuning proportional-plus-integral (STPI) control algorithms developed to work in conjunction with the Bristol-Babcock PID control module. The STPI control algorithms are based on closed-loop cycling theory, pattern recognition theory, and model-based theory. A brief theory of operation of these three STPI control algorithms is given. Details of the process simulations developed to test the STPI algorithms are given, including an integrating process, a first-order system, a second-order system, a system with initial inverse response, and a system with variable time constant and delay. The STPI algorithms' performance with regard to both setpoint changes and load disturbances is evaluated, and their robustness is compared. The dynamic effects of process deadtime and noise are also considered. Finally, the limitations of each of the STPI algorithms is discussed, some conclusions are drawn from the performance comparisons, and a few recommendations are made. 6 refs

  12. Research reactor education and training

    International Nuclear Information System (INIS)

    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)

  13. Research Reactors Types and Utilization

    International Nuclear Information System (INIS)

    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 235U or 239Pu), 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.

  14. Research reactor modernization and refurbishment

    International Nuclear Information System (INIS)

    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

  15. An example of Ensemble Kalman Filter data assimilation in a Babcock-Leighton solar dynamo model

    Science.gov (United States)

    Dikpati, Mausumi; Anderson, Jeffrey L.

    2016-05-01

    Atmospheric and oceanic prediction models have been greatly advanced over the past 40 years by using modern data assimilation techniques. Application of similar techniques in solar models started about 7 years ago. However, acceptance of such techniques by the solar community has been slow to develop. In order to make accurate predictions of solar activity as well as reconstruction of certain model parameters that cannot be directly measured, it will be essential to implement sophisticated data assimilation techniques as used by atmospheric and oceanic models. We will present here an example of parameter reconstruction, namely the time variation in meridional flow-speed, done by assimilating data into a Babcock-Leighton solar dynamo model in the framework of NCAR's Data Assimilation Research Testbed (NCAR-DART). By performing many 'Observing System Simulation Experiments' (OSSEs) we find that an optimally good reconstruction in time series of meridional circulation can be obtained by using 16 ensemble members and assimilating one magnetic observation with less than 40 percent observational error. However, the RMS error in reconstruction reduces with increase in ensemble size, increase in number of observations and decrease in observational error. We also find that assimilation of magnetic field observations taken from low-to-mid latitudes at the surface compared to any other locations produces the best reconstruction. We will close by showing that assimilation cycle of 15 days is optimal; generally a longer assimilation cycle deteriorates the results, but the Dynamo DART system needs a minimum time to develop the dynamics.

  16. 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.

  17. Propose Reactor Control and Monitoring System for RTP

    International Nuclear Information System (INIS)

    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)

  18. Disturbance analysis and surveillance system scoping and feasibility system. Final report

    International Nuclear Information System (INIS)

    This report summarizes the results of a disturbance analysis and surveillance system (DASS) scoping and feasibility study conducted by The Babcock and Wilcox Company, Burns and Roe, Incorporated, General Physics Corporation, and Duke Power Company for Sandia Laboratories and the US Department of Energy. The report addresses selection of DASS goals and functions, development of a design concept for a DASS based on monitoring the nuclear plant subsystem functions and states against predetermined targets, and creation of engineering procedures for the design and implementation of a DASS. The validity of the procedures is evaluated based on application to a subset of the DASS functions. It is concluded that the DASS design concept is a feasible, systematic, and modular approach to plant disturbance identification

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

    International Nuclear Information System (INIS)

    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.)

  20. 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.

  1. Nuclear reactors and fuel cycle

    International Nuclear Information System (INIS)

    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 100th nuclear fuel element produced. Besides routine production, development of new technologies is also a permanent concern at CCN. In 2005, U3O8 were replaced by U3Si2-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 to

  2. Coal gasification systems engineering and analysis. Volume 1: Executive summary

    Science.gov (United States)

    1980-01-01

    Feasibility analyses and systems engineering studies for a 20,000 tons per day medium Btu (MBG) coal gasification plant to be built by TVA in Northern Alabama were conducted. Major objectives were as follows: (1) provide design and cost data to support the selection of a gasifier technology and other major plant design parameters, (2) provide design and cost data to support alternate product evaluation, (3) prepare a technology development plan to address areas of high technical risk, and (4) develop schedules, PERT charts, and a work breakdown structure to aid in preliminary project planning. Volume one contains a summary of gasification system characterizations. Five gasification technologies were selected for evaluation: Koppers-Totzek, Texaco, Lurgi Dry Ash, Slagging Lurgi, and Babcock and Wilcox. A summary of the trade studies and cost sensitivity analysis is included.

  3. Review of the use and state of development of the various reactor types

    International Nuclear Information System (INIS)

    The report gives a review of the reactor types being of importance from today's point of view for use as stationary power reactors. These are heavy water reactors, light water reactors (pressurized water reactor, Soviet pressurized water reactor, Soviet light-water-graphite reactors, boiling water reactors), gas-cooled reactors (gas-graphite reactors, high temperature reactors), and fast breeder reactors. (HJ)

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

    International Nuclear Information System (INIS)

    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

  5. Modeling and kinetics research of IGR reactor

    International Nuclear Information System (INIS)

    The effort addresses issues related to modeling and studying of IGR reactor dynamic behavior; an example of IGR reactor kinetics model realization and study results in time and frequency domains are given. (author)

  6. The SAPPHIRE and 50 MT projects at BWXT, Lynchburg, VA

    International Nuclear Information System (INIS)

    When the SAPPHIRE project for the down-blending of HEU material of Khazak origin was initiated in 1996 at BWX Technologies (BWXT) formally Babcock and Wilcox in Lynchburg, VA and the Agency was requested to apply its specially designed safeguards measures to the process with a view to provide assurance to the international community that down-blending had actually taken place as stipulated in the USA-Khazak agreement a learning process was initiated from this effort culminating in the current 50 MT downblending process at the same facility with BWXT, the USA Authorities, and the Agency as partners in this technologically advanced enterprise aimed at the downgrading of a substantial quantity of weapons grade material. In the present paper an overview is provided of the road leading to an effective, and mutually agreeable safeguards approach for carrying out verifications in the sensitive environment of a facility devoted to HEU uranium processing. (author)

  7. The SAPPHIRE and 50 MT projects at BWXT, Lynchburg, VA

    International Nuclear Information System (INIS)

    Full text: When the SAPPHIRE project for the down-blending of HEU material of Khazak origin was initiated in 1996 at BWX Technologies (BWXT) formally Babcock and Wilcox in Lynchburg, VA and the Agency was requested to apply its specially designed safeguards measures to the process with a view to provide assurance to the international community that down-blending had actually taken place as stipulated in the USA-Khazak agreement a learning process was initiated from this effort culminating in the current 50 MT downblending process at the same facility with BWXT, the USA Authorities, and the Agency as partners in this technologically advanced enterprise aimed at the downgrading of a substantial quantity of weapons grade material. In the present paper an overview is provided of the road leading to an effective, and mutually agreeable safeguards approach for carrying out verifications in the sensitive environment of a facility devoted to HEU uranium processing. (author)

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

    International Nuclear Information System (INIS)

    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

  9. Fabrication and closure development of nuclear waste containers for storage at the Yucca Mountain, Nevada repository

    International Nuclear Information System (INIS)

    US Congress and the President have determined that the Yucca Mountain site in Nevada is to be characterized to determine its suitability for construction of the first US high-level nuclear waste repository. Work in connection with this site is carried out within the Yucca Mountain Project (YMP). Lawrence Livermore National Laboratory (LLNL) has the responsibility for designing, developing, and projecting the performance of the waste package for the permanent storage of high-level nuclear waste. Babcock ampersand Wilcox (B ampersand W) is involved with the YMP as a subcontractor to LLNL. B ampersand W's role is to recommend and demonstrate a method for fabricating the metallic waste container and a method for performing the final closure of the container after it has been filled with waste. Various fabrication and closure methods are under consideration for the production of containers. This paper presents progress to date in identifying and evaluating the candidate manufacturing processes. 2 refs., 1 fig., 7 tabs

  10. Reactors

    International Nuclear Information System (INIS)

    Purpose: To provide a spray cooling structure wherein the steam phase in a bwr reactor vessel can sufficiently be cooled and the upper cap and flanges in the vessel can be cooled rapidly which kept from direct contaction with cold water. Constitution: An apertured shielding is provided in parallel spaced apart from the inner wall surface at the upper portion of a reactor vessel equipped with a spray nozzle, and the lower end of the shielding and the inner wall of the vessel are closed to each other so as to store the cooling water. Upon spray cooling, cooling water jetting out from the nozzle cools the vapor phase in the vessel and then hits against the shielding. Then the cooling water mostly falls as it is, while partially enters through the apertures to the back of the shielding plate, abuts against stoppers and falls down. The stoppers are formed in an inverted L shape so that the spray water may not in direct contaction with the inner wall of the vessel. (Horiuchi, T.)

  11. Intermediate leak protection/automatic shutdown for B and W helical coil steam generator

    International Nuclear Information System (INIS)

    The report summarizes a follow-on study to the multi-tiered Intermediate Leak/Automatic Shutdown System report. It makes the automatic shutdown system specific to the Babcock and Wilcox (B and W) helical coil steam generator and to the Large Development LMFBR Plant. Threshold leak criteria specific to this steam generator design are developed, and performance predictions are presented for a multi-tier intermediate leak, automatic shutdown system applied to this unit. Preliminary performance predictions for application to the helical coil steam generator were given in the referenced report; for the most part, these predictions have been confirmed. The importance of including a cover gas hydrogen meter in this unit is demonstrated by calculation of a response time one-fifth that of an in-sodium meter at hot standby and refueling conditions

  12. The effect of aging upon CE and B ampersand W control rod drives

    International Nuclear Information System (INIS)

    The effect of aging upon the Babcock ampersand Wilcox (B ampersand W) and Combustion Engineering (CE) Control Rod Drive (CRD) systems has been evaluated as part of the USNRC Nuclear Plant Aging Research (NPAR) program. Operating experience data for the 1980--1990 time period was reviewed to identify predominant failure modes, causes, and effects. These results, in conjunction with an assessment of component materials and operating environment, conclude that both systems are susceptible to age degradation. System failures have resulted in significant plant effects, including power reductions, plant shutdowns, scrams, and Engineered Safety Feature (ESF) actuation. Current industry inspection and maintenance practices were assessed. Some of these practices effectively address aging, while others do not

  13. The pulsed reactor and its application

    International Nuclear Information System (INIS)

    The situation of the first pulsed reactor in China is briefly described. The pulsed reactor has a large prompt negative temperature coefficient of reactivity provided by combination of the uranium-zirconium hydride fuel and the moderator. Therefore, its most outstanding features are 'inherent safety' and fairly high pulsed-power capacity. The pulsed reactor is now extensively used in science and technology

  14. Safety of research reactors (Design and Operation)

    International Nuclear Information System (INIS)

    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)

  15. Simulation analysis of static and dynamic characteristics of once-through steam generator in concentric annuli tube

    Institute of Scientific and Technical Information of China (English)

    ZHANG Wei; BIAN Xin-qian; XIA Guo-qing

    2006-01-01

    The once-through steam generator (OTSG) in concentric annuli tube is a new type of steam generator which applies double side to transfer heat. The heat flux between the water of centric tube, outside annuli tube and that of annulus channel is assumed to be equal, and then the steam generator's model is built by lumped parameters with moving boundary. In the basis of the built model, static and dynamic characteristics are analyzed.The static characteristics are proved by experiment results in a 19-tube once-through steam generator of Babcock & Wilcox. The characteristics that the lengths of three regions (subcooled region, nucleate boiling region, superheat region) change with power can be explained by theory analysis. The dynamic characteristics accord with the heat and hydraulics and the results of analysis according to the mechanism.

  16. Safety review, assessment and inspection on research reactors, experimental reactors and nuclear heating reactors

    International Nuclear Information System (INIS)

    The NNSA and its regional office step further strengthened the regulation on the safety of in-service research reactors in 1996. A lot of work has been done on the supervision of safe in rectifying the review and assessment of modified items, the review of operational documents, the treatment of accidents, the establishment of the system for operational experience feedback, daily and routine inspection on nuclear safety. The internal management of the operating organization on nuclear safety was further strengthened, nuclear safety culture was further enhanced, the promotion in nuclear safety and the safety situation for in-service research reactors were improved

  17. 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)

    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

  18. Paradoxical effects of all-trans-retinoic acid on lupus-like disease in the MRL/lpr mouse model.

    Directory of Open Access Journals (Sweden)

    Xiaofeng Liao

    Full Text Available Roles of all-trans-retinoic acid (tRA, a metabolite of vitamin A (VA, in both tolerogenic and immunogenic responses are documented. However, how tRA affects the development of systemic autoimmunity is poorly understood. Here we demonstrate that tRA have paradoxical effects on the development of autoimmune lupus in the MRL/lpr mouse model. We administered, orally, tRA or VA mixed with 10% of tRA (referred to as VARA to female mice starting from 6 weeks of age. At this age, the mice do not exhibit overt clinical signs of lupus. However, the immunogenic environment preceding disease onset has been established as evidenced by an increase of total IgM/IgG in the plasma and expansion of lymphocytes and dendritic cells in secondary lymphoid organs. After 8 weeks of tRA, but not VARA treatment, significantly higher pathological scores in the skin, brain and lung were observed. These were accompanied by a marked increase in B-cell responses that included autoantibody production and enhanced expression of plasma cell-promoting cytokines. Paradoxically, the number of lymphocytes in the mesenteric lymph node decreased with tRA that led to significantly reduced lymphadenopathy. In addition, tRA differentially affected renal pathology, increasing leukocyte infiltration of renal tubulointerstitium while restoring the size of glomeruli in the kidney cortex. In contrast, minimal induction of inflammation with tRA in the absence of an immunogenic environment in the control mice was observed. Altogether, our results suggest that under a predisposed immunogenic environment in autoimmune lupus, tRA may decrease inflammation in some organs while generating more severe disease in others.

  19. Process regime variability across growth faults in the Paleogene Lower Wilcox Guadalupe Delta, South Texas Gulf Coast

    Science.gov (United States)

    Olariu, Mariana I.; Ambrose, William A.

    2016-07-01

    The Wilcox Group in Texas is a 3000 m thick unit of clastic sediments deposited along the Gulf of Mexico coast during early Paleogene. This study integrates core facies analysis with subsurface well-log correlation to document the sedimentology and stratigraphy of the Lower Wilcox Guadalupe Delta. Core descriptions indicate a transition from wave- and tidally-influenced to wave-dominated deposition. Upward-coarsening facies successions contain current ripples, organic matter, low trace fossil abundance and low diversity, which suggest deposition in a fluvial prodelta to delta front environment. Heterolithic stratification with lenticular, wavy and flaser bedding indicate tidal influence. Pervasively bioturbated sandy mudstones and muddy sandstones with Cruziana ichnofacies and structureless sandstones with Ophiomorpha record deposition in wave-influenced deltas. Tidal channels truncate delta front deposits and display gradational upward-fining facies successions with basal lags and sandy tabular cross-beds passing into heterolithic tidal flats and biologically homogenized mudstones. Growth faults within the lower Wilcox control expanded thickness of sedimentary units (up to 4 times) on the downdip sides of faults. Increased local accommodation due to fault subsidence favors a stronger wave regime on the outer shelf due to unrestricted fetch and water depth. As the shoreline advances during deltaic progradation, successively more sediment is deposited in the downthrown depocenters and reworked along shore by wave processes, resulting in a thick sedimentary unit characterized by repeated stacking of shoreface sequences. Thick and laterally continuous clean sandstone successions in the downthrown compartments represent attractive hydrocarbon reservoirs. As a consequence of the wave dominance and increased accommodation, thick (tens of meters) sandstone-bodies with increased homogeneity and vertical permeability within the stacked shoreface successions are created.

  20. Advanced reactor concepts and safety

    International Nuclear Information System (INIS)

    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

  1. Thermal hydraulic R and D of Chinese advanced reactors

    International Nuclear Information System (INIS)

    The Chinese government sponsors a program of research, development, and demonstration related to advanced reactors, both small modular reactors and larger systems. These advanced reactors encompass innovative reactor concepts, such as CAP1400 - Chinese large advanced passive pressurized water reactor, Hualong one - Chinese large advanced active and passive pressurized water reactor, ACP100 - Chinese small modular reactor, SCWR- R and D of super critical water-cooled reactor in China, CLEAR - Chinese lead-cooled fast reactor, TMSR - Chinese Thorium molten-salt reactor. The thermal hydraulic R and D of those reactors are summarised. (J.P.N.)

  2. Three-dimensional ground-water modeling in depositional systems, Wilcox Group, Oakwood salt dome area, east Texas

    International Nuclear Information System (INIS)

    The data base includes not only measurements of hydraulic head and hydraulic conductivity but also lithofacies maps constructed in a previous study of Wilcox depositional systems. The Carrizo aquifer is a fairly homogeneous sand sheet overlying the much thicker Wilcox Group, a multiple-aquifer system composed primarily of fluvial channel-fill sand bodies distributed among lower permeability interchannel sands and muds. The interconnectedness of the channel-fill sands, which have predictable values of hydraulic conductivity, strongly influences the rate and direction of ground-water flow. Lateral interconnectedness may depend largely on frequency distributions of channel-fill sands (that is, sand percent). Vertical interconnectedness is apparently poor owing to the horizontal stratification of sand and mud. Simulating observed pressure-depth trends by manipulating values of equivalent vertical hydraulic conductivity (K/sub v/') demonstrates that the ratio of vertical to horizontal conductivity (K/sub v/'/K/sub h/') is very low (about 10-3 to 10-4). Locally high values of K/sub v/' could result in locally rapid vertical flow, which could in turn be detected using pressure-depth residence times of 103 to 104 years in channel-fill facies and 105 to 106 years in interchannel facies. Because Oakwood Dome is apparently surrounded by interchannel facies as a result of syndepositional dome growth, the dome may be essentially isolated from circulating Wilcox ground water. A possible exception is where channel-fill facies appear to touch or come close to the northeast flank, coinciding with a brackish-water plume that apparently results from dissolution of salt of cap rock. The northeast orientation of the plume appears to be caused by sand-body distribution and interconnection. 38 references

  3. Physics and kinetics of TRIGA reactor

    International Nuclear Information System (INIS)

    This training module is written as an introduction to reactor physics for reactor operators. It assumes the reader has a basic, fundamental knowledge of physics, materials and mathematics. The objective is to provide enough reactor theory knowledge to safely operate a typical research reactor. At this level, it does not necessarily provide enough information to evaluate the safety aspects of experiment or non-standard operation reviews. The material provides a survey of basic reactor physics and kinetics of TRIGA type reactors. Subjects such as the multiplication factor, reactivity, temperature coefficients, poisoning, delayed neutrons and criticality are discussed in such a manner that even someone not familiar with reactor physics and kinetics can easily follow. A minimum of equations are used and several tables and graphs illustrate the text. (author)

  4. Fast reactors and advanced light water reactors for sustainable development

    International Nuclear Information System (INIS)

    Complete text of publication follows: The importance of nuclear energy, as a realistic option to solve the issues of the depletion of energy resources and the global environment, has been re-acknowledged worldwide. In response to this international movement, the papers compiling the most recent findings in the fields of fast reactors (FR) and advanced light water reactors (LWR) were gathered and published in this special issue. This special issue compiles six articles, most of which are very meticulously performed studies of the multi year development of design and assessment methods for large sodium-cooled FRs (SFRs), and two are related to the fuel cycle options that are leading to a greater understanding on the efficient utilization of energy resources. The Japanese sodium-cooled fast reactor (JSFR) is addressed in two manuscripts. H. Yamano et al. reviewed the current design which adopts a number of innovative technologies in order to achieve economic competitiveness, enhanced reliability, and safety. Their safety assessments of both design basis accidents and severe accidents indicate that the devised JSFR satisfies well their risk target. T. Takeda et al. discussed the improvement of the modeling accuracy for the detailed calculation of JSFR's features in three areas: neutronics, fuel materials, and thermal hydraulics. The verification studies which partly use the measured data from the prototype FBR Monju are also described. Two of these manuscripts deal with those aspects of advanced design of SFR that have hitherto not been explored in great depth. The paper by G. Palmiotti et al. explored the possibility of using the sensitivity methodologies in the reactor physics field. A review of the methods used is provided, and several examples illustrate the success of the methodology in reactor physics. A new application as the improvement of nuclear basic parameters using integral experiments is also described. F. Baque et al. reviewed the evolution of the in

  5. Prospects for small and medium power reactors

    International Nuclear Information System (INIS)

    A searching examination of the present status of nuclear power technology and economics was made in 64 papers presented to the Conference on Small and Medium Power Reactors held by the IAEA in Vienna during the week 5 - 9 September 1960. The IAEA Conference concentrated on small and medium power reactors because these are the sizes of primary interest to less-developed countries around the world. The Conference brought forward information on a wide range of subjects related to power reactors, including power costs, summaries of national programs, applications in less-developed countries, process heat reactors, reactor safety, results of experience in the actual construction and operation of power reactors and technical appraisals of various reactor types

  6. Reactors and physics education

    International Nuclear Information System (INIS)

    This paper discussed some ideas for using neutrons in physics education, including experiments which demonstrate diffraction and optical refraction, divergence imaging, Zeeman splitting, polarization, Larmor precession, and neutron spin-echo. (author)

  7. The automatic control design and simulation of reactor control system in small modular reactor

    International Nuclear Information System (INIS)

    In China, the development and application goals of Small Modular Reactor (SMR) aim at electricity generation area, heat supply area, and seawater desalination area, etc. The main technical features of the SMR are as follows: integrated pressurized water reactor, reactor coolant pump and reactor pressure vessel connected by short pipe, steam generator sets in reactor pressure vessel, control rod drive mechanism (CRDM), pressure vessel, reactor internals, Once-Through Steam Generator (OTSG), and canned motor pump are all mature technology. Based on the characteristic of the reactor and OTSG, the automatic control design of the SMR is discussed in this paper, and the simulation results are presented to illustrate the control scheme. (author)

  8. Nuclear reactors and disarmament

    International Nuclear Information System (INIS)

    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)

  9. Refurbishment and activities at Tajoura reactor

    Energy Technology Data Exchange (ETDEWEB)

    Abutweirat, F.; Abusta, M. [Renewable Energies and Water Desalination Research Centre, Basic and Applied Research Dept., Tajoura (Libyan Arab Jamahiriya)

    2007-07-01

    The Tajoura Research Reactor was built in the late seventies by the former Soviet Union for Libya. The Tajoura Research Reactor is a 10 MW light water cooled and moderated, beryllium reflected, pool type reactor. Its design facilitates the production of radioisotopes and the performance of material testing experiments. The reactor is provided with a critical assembly that is an exact mockup of the reactor core to test and study neutron transport in the different core configurations. The utilization of the reactor suffered the most due to the hardship which had confronted Libya during the years 1985 - 2000. During that time the utilization was limited to the use of the reactor as an educational tool for university students, for training reactor operators and for capacity building in the field of radiation safety, radiation chemistry, isotope production and neutron activation analysis. Both the Critical Assembly and the reactor were recently converted from the high enrichment uranium (HEU) fuel (Type IRT-2M) to low enrichment (LEU) fuel (Type IRT-4M). The refurbishment of the control and safety systems of the reactor and the critical assembly is due to start in a near future.

  10. Refurbishment and activities at Tajoura reactor

    International Nuclear Information System (INIS)

    The Tajoura Research Reactor was built in the late seventies by the former Soviet Union for Libya. The Tajoura Research Reactor is a 10 MW light water cooled and moderated, beryllium reflected, pool type reactor. Its design facilitates the production of radioisotopes and the performance of material testing experiments. The reactor is provided with a critical assembly that is an exact mockup of the reactor core to test and study neutron transport in the different core configurations. The utilization of the reactor suffered the most due to the hardship which had confronted Libya during the years 1985 - 2000. During that time the utilization was limited to the use of the reactor as an educational tool for university students, for training reactor operators and for capacity building in the field of radiation safety, radiation chemistry, isotope production and neutron activation analysis. Both the Critical Assembly and the reactor were recently converted from the high enrichment uranium (HEU) fuel (Type IRT-2M) to low enrichment (LEU) fuel (Type IRT-4M). The refurbishment of the control and safety systems of the reactor and the critical assembly is due to start in a near future

  11. 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 JPL

    Energy Technology Data Exchange (ETDEWEB)

    D. Kokkinos

    2005-04-28

    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.

  12. RA nuclear reactor - revitalisation, renewal and applications

    International Nuclear Information System (INIS)

    This book is meant to give professional support in solving the problem of RA reactor, its revitalisation and renewal, as a special help for decision makers. Facts in favor of restarting RA reactor are prevailing. This report is made of six parts. First part includes an overview of basic properties of research reactors in the world and a discussion concerning their future development. RA reactor parameters are analyzed both with low enriched and highly enriched fuel and it has been concluded that the aim of RA reactor renewal should be to obtaining as high as possible thermal neutron flux density. The second part deals with possible applications of RA reactor in fundamental and applied research programs, commercial applications and its role in education and training programs. The third part discusses application of RA reactor as a source of thermal neutrons for fundamental and applied sciences, especially in the condensed matter physics and development of new materials. The role of RA reactor in development of radiation protection systems is emphasised in part four. Some possible commercial applications of Ra reactor are described in part five: isotope production, and their different applications. Part six deals with education and training of staff, with special accent on scientific international cooperation. Basic conclusions of this material meant for decision makers are: restarting RA reactor is the most reasonable and activities related to its revitalisation and renewal should be continued; this program should include solving the problems of education and training of the staff for reactor operation, improvement and different applications; renewal program should include renewal of the experimental devices as a condition of reactor efficient application immediately after its startup

  13. 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.

  14. Education and Training on ISIS Research Reactor

    International Nuclear Information System (INIS)

    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

  15. FBR and RBR particle bed space reactors

    International Nuclear Information System (INIS)

    Compact, high-performance nuclear reactor designs based on High-Temperature Gas Reactors (HTGRs) particulate fuel are investigated. The large surface area available with the small-diameter (approx. 500 microns) particulate fuel allows very high power densities (MW's/liter), small temperature differences between fuel and coolant (approx. 100K), high coolant-outlet temperatures (1500 to 30000K, depending on design), and fast reactor startup (approx. 2 to 3 seconds). Two reactor concepts are developed - the Fixed Bed Reactor (FBR), where the fuel particles are packed into a thin annular bed between two porous cylindrical drums, and the Rotating Bed Reactor (RBR), where the fuel particles are held inside a cold rotating (typically approx. 500 rpm) porous cylindrical drum. The FBR can operate steady-state in the closed-cycle He-cooled mode or in the open-cycle H2-cooled mode. The RBR will operate only in the open-cycle H2-cooled mode

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

    International Nuclear Information System (INIS)

    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

  17. 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.

  18. Advanced nuclear reactor types and technologies

    International Nuclear Information System (INIS)

    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

  19. Westinghouse's small and medium reactor portfolio

    International Nuclear Information System (INIS)

    Full text: Westinghouse has been a pioneer in the civil nuclear power industry. The first commercial nuclear reactor was a Westinghouse reactor in Shippingport, PA of the United States of America. The company was founded in 1886 by the inventor and entrepreneur George Westinghouse. Today, Westinghouse Electric Company is a nuclear technology company and 60% of the electricity produced from nuclear power in the world is based on Westinghouse technology. Westinghouse is working with partners worldwide to build its 1100 MWe advanced passive PWR. It is also developing small and medium size reactors to fill market niches and for what is known as Generation IV reactors. These reactors (<700 MWe) are suitable where there exists one or more of the following limitations: grid, financing, site etc. IRIS is one such reactor that utilizes a simplified, integral configuration. This integral, advanced PWR at 335 MWe locates major components inside the reactor pressure vessel to eliminate system piping and other components. IRIS is being developed by an international development team that includes ten countries and twenty four organizations. PMBR (Pebble Bed Modular Reactor) is a Generator IV high temperature gas reactor that supports co-generation operation. The PBMR design is being developed through a partnership between Westinghouse Electric Co. and PBMR (Pty) Ltd of the Republic of South Africa. The PBMR design is sized at 200 MWt and 80 MWe to support a broad range of process steam applications. Furthermore, the PBMR achieves inherent safety levels through the use of innovative TRISO fuel. In addition, Westinghouse parent company, Toshiba Corporation, is developing the 4S sodium fast reactor which is a 10-50 MWe reactor that is ideal for isolated areas of small power demand. The conference presentation will include specific product features and the development status of the small and medium reactors in the Westinghouse portfolio

  20. Space reactor fuels performance and development issues

    International Nuclear Information System (INIS)

    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

  1. Space reactor fuels performance and development issues

    International Nuclear Information System (INIS)

    Three compact reactor concepts are now under consideration by the U.S. 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

  2. Inertial fusion reactors and magnetic fields

    International Nuclear Information System (INIS)

    The application of magnetic fields of simple configurations and modest strengths to direct target debris ions out of cavities can alleviate recognized shortcomings of several classes of inertial confinement fusion (ICF) reactors. Complex fringes of the strong magnetic fields of heavy-ion fusion (HIF) focusing magnets may intrude into reactor cavities and significantly affect the trajectories of target debris ions. The results of an assessment of potential benefits from the use of magnetic fields in ICF reactors and of potential problems with focusing-magnet fields in HIF reactors conducted to set priorities for continuing studies are reported. Computational tools are described and some preliminary results are presented

  3. The need to address the larger universe of HEU-fueled reactors, including critical assemblies, pulsed reactors and propulsion reactors

    International Nuclear Information System (INIS)

    Full text: The RERTR program has focused thus far primarily on ending shipments of HEU fuel to research reactors. This has resulted in giving highest priority to reactors with steady thermal powers of 1 megawatt or more, because they require regular refuelling. Critical facilities and pulsed reactors can also of serious concern, because some of them contain very large amounts of barely-irradiated HEU and plutonium. They could be costly to convert - and conversion to LEU may be impractical for fast-neutron critical assemblies. An assessment should be carried out first, therefore, as to which are still needed. Critical assemblies are required today primarily to benchmark Monte Carlo neutron-transport codes. Perhaps the world nuclear community could share a few instead of each reactor-design institute having its own. There is also a whole universe of HEU-fuelled pressurized-water reactors used to power submarines and other types of nuclear-powered ships. These reactors collectively require much more HEU fuel each year than research reactors. The risk of HEU diversion from their fuel cycles is not zero but it is difficult for outsiders to discuss conversion because of the fuel designs are classified. This makes the conversion of Russia's civilian icebreaker reactors of particular interest because issues of classified fuel design are less problematic and these reactors load annually fuel containing about 400 kg of U-235. Another reason for interest in developing LEU fuel for these reactors is that the KLT-40 icebreaker reactor is being adapted for a floating nuclear power plant. Finally, the research-reactor community is, in any case, faced with developing fuels that can operate at power-reactor-fuel temperatures because there are a few high-powered research reactors that operate in this temperature range. (author)

  4. The Shippingport Pressurized Water Reactor and Light Water Breeder Reactor

    International Nuclear Information System (INIS)

    This report discusses the Shippingport Atomic Power Station, located in Shippingport, Pennsylvania, which was the first large-scale nuclear power plant in the United States and the first plant of such size in the world operated solely to produce electric power. A program was started in 1953 at the Bettis Laboratory to confirm the practical application of nuclear power for large-scale electric power generation. It led to the development of zirconium alloy (Zircaloy) clad fuel element containing bulk actinide oxide ceramics (UO2, ThO2, ThO2 -- UO2, ZrO2 -- UO2) as nuclear reactor fuels. The program provided much of the technology being used for design and operation of the commercial, central-station nuclear power plants now in use. The Shippingport Pressurized Water Reactor (PWR) began initial power operation on December 18, 1957, and was a reliable electric power producer until February 1974. In 1965, subsequent to the successful operation of the Shippingport PWR (UO2, ZrO2 -- UO2 fuels), the Bettis Laboratory undertook a research and development program to design and build a Light Water Breeder Reactor (LWBR) core for operation in the Shippingport Station. Thorium was the fertile fuel in the LWBR core and was the base oxide for ThO2 and ThO2 -- UO2 fuel pellets. The LWBR core was installed in the pressure vessel of the original Shippingport PWR as its last core before decommissioning. The LWBR core started operation in the Shippingport Station in the autumn of 1977 and finished routine power operation on October 1, 1982. Successful LWBR power operation to over 160% of design lifetime demonstrated the performance capability of the core for both base-load and swing-load operation. Postirradiation examinations confirmed breeding and successful performance of the fuel system

  5. Safety review and assessment and inspection on research reactors, experimental reactors, nuclear heating reactors and critical facilities

    International Nuclear Information System (INIS)

    More operational events were occurred at various research reactors in 1995. The NNSA and its regional offices conducted careful investigation and strict regulation. In order to analyze comprehensively the safety situation of inservice research reactors and find same countermeasures the NNSA convened a meeting of the safety regulation on research reactors and a meeting for change experience of the safety regulation on research reactors that were participated in by the operating organizations in 1995. A lot of work has been done in the respects of propagation of regulations on nuclear safety, education of nuclear safety culture, the investigation and treatment of operational events, the reexamine of operation documents, the implementation of rectifying items on nuclear safety, the daily inspection and routine inspection on nuclear safety and the studying on the extending service life of research reactors etc

  6. New reactors concepts and scenarios

    International Nuclear Information System (INIS)

    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)

  7. Research reactors and alternative devices for research

    International Nuclear Information System (INIS)

    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

  8. Exporting apocalypse: CANDU reactors and nuclear proliferation

    International Nuclear Information System (INIS)

    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

  9. Reactor Physics and the Nuclear Fuel Cycle

    Directory of Open Access Journals (Sweden)

    Md Minhaj Ahmed

    2013-11-01

    Full Text Available Questions regarding the feasibility of fusion power are examined, taking into account fuel cycles and breeding reactions, energy balance and reactor conditions, approaches to fusion, magnetic confinement, magneto hydro dynamic instabilities, micro instabilities, and the main technological problems which have to be solved. Basic processes and balances in fusion reactors are considered along with some aspects of the neutronics in fusion reactors, the physics of neutral beam heating, plasma heating by relativistic electrons, radiofrequency heating of fusion plasmas, adiabatic compression and ignition of fusion reactors, dynamics and control of fusion reactors, and aspects of thermal efficiency and waste heat. Attention is also given to fission-fusion hybrid systems, inertial-confinement fusion systems, the radiological aspects of fusion reactors, design considerations of fusion reactors, and a comparative study of the approaches to fusion power. The nuclear fuel cycle, also called nuclear fuel chain, is the progression of nuclear fuel through a series of differing stages. It consists of steps in the front end, which are the preparation of the fuel, steps in the service period in which the fuel is used during reactor operation, and steps in the back end, which are necessary to safely manage, contain, and either reprocess or dispose of spent nuclear fuel. If spent fuel is not reprocessed, the fuel cycle is referred to as an open fuel cycle (or a once-through fuel cycle; if the spent fuel is reprocessed, it is referred to as a closed fuel cycle..

  10. Roles of interferon-gamma and interleukin-4 in murine lupus.

    OpenAIRE

    Peng, S L; Moslehi, J; Craft, J.

    1997-01-01

    The systemic autoimmune syndrome of MRL/Mp-lpr/lpr (MRL/lpr) mice consists of severe pan-isotype hypergammaglobulinemia, autoantibody production, lymphadenopathy, and immune complex-associated end-organ disease. Its pathogenesis has been largely attributed to helper alphabeta T cells that may require critical cytokines to propagate pathogenic autoantibody production. To investigate the roles of prototypical Th1 and Th2 cytokines in the pathogenesis of murine lupus, IFN-gamma -/- and IL-4 -/- ...

  11. TRIGA reactor owners' seminar. Papers and abstracts

    International Nuclear Information System (INIS)

    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

  12. Comparison of Inflammatory Events during Developing Immunoglobulin E-Mediated Late-Phase Reactions and Delayed-Hypersensitivity Reactions

    OpenAIRE

    Zweiman, Burton; Moskovitz, Anne R.; von Allmen, Carolyn

    1998-01-01

    To compare cellular and mediator responses in early developing late-phase skin reactions (LPR) and delayed-hypersensitivity (DH) reactions in the same subjects, responses in skin chambers overlying sites of challenge with pollen antigen and Candida albicans antigens were compared in six humans with demonstrated prominent LPR and DH responses. Histamine levels in overlying chamber fluids at 1 h were much higher at LPR than at DH sites (P = 0.002). After the next 4 h, leukocyte exudation was hi...

  13. Operational and reliability experience with reactor instrumentation

    International Nuclear Information System (INIS)

    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)

  14. Scanning tunneling microscope assembly, reactor, and system

    Energy Technology Data Exchange (ETDEWEB)

    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.

  15. Maintenance and material aspects of DREAM reactor

    International Nuclear Information System (INIS)

    A concept of a commercial fusion power reactors (Fusion Power: 5.5 GW, electric output: 2.7 GW) having high environmental safety, high thermal efficiency and high availability has been studied in JAERI. The gross reactor configuration was designed to achieve good maintainability, high performance breeding blanket, high efficient power generation system and little radwastes. Design was based on the use of low activation structural material (SiC/SiC composites) and helium as a coolant. In this paper, maintenance and material aspects of DREAM reactor design is discussed. The concluding remarks are as follows. (1) The difficulty of development of maintenance tool is alleviated by sector replacement and the radiation dose environment less than 10 Gy/h in a reactor chamber. (2) Design requirement and present status of SiC/SiC composites was investigated. (3) The SiC/SiC composite development program is planned to satisfy the requirements of DREAM reactor

  16. Research reactor and its application in Thailand

    International Nuclear Information System (INIS)

    The first Thai Research Reactor (TRR-1) was established in 1961. TRR-1 had been operated with power of 1 MW from 1962 to 1975 and was shut down for modification during 1975 to 1977. The Thai Research Reactor1/modification1 (TRR-1/M1) is a multipurpose reactor with nominal power of 2 MW. Since 1977 TRR-1/M1 has been operated and utilized for various applications such as neutron activation analysis, radioisotope production, gem irradiation, neutron radiography and research works. To expand and promote the utilization of research reactor, the new 10 MW Research Reactor will be established in the Ongkarak Nuclear Research Center (ONRC) project and the project will be finished in the near future. (author)

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

    International Nuclear Information System (INIS)

    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

  18. TREAT Reactor Control and Protection System

    International Nuclear Information System (INIS)

    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

  19. Problems and experience of research reactor decommissioning

    International Nuclear Information System (INIS)

    According to the IAEA research reactor database there are about 300 research reactors worldwide. At present above 30% of them have lifetime more than 35 years, 60% - more then 25 years. After the Chernobyl accident significant efforts have been made by many countries to modernize old research reactors aiming, first of all, at ensuring of its safe operation. However, a large number of aging research reactor will be facing shutdown in the near future. Before developing the design and planning of the works it is necessary to define the concept of the reactor decommissioning. It is defined by the time of the beginning of dismantling works after the reactor shutdown and the finite state of the reactor site.The concept of the reactor decommissioning provides 3 variants in a general case: reactor conservation, or partial dismantling, or complete dismantling to 'green field' state. Specialists of three International institutions (European Commission, IAEA and the Nuclear Energy Agency/Organization for Economic Cooperation and Development) have developed a detailed plan of all actions and operations on nuclear power plants decommissioning in the framework of a joint project for cost assessment. For the reactor decontamination the following main constructions, equipment and devices are necessary: temporary storage facility for the spent fuel; general site-dismantling equipment including manipulators and 'hot' cells; facilities for 'active' equipment, personnel, tooling and washing decontamination; equipment for concentration of liquid and compactness of solid radioactive waste; temporary storage facility for radioactive waste; instrumentation and radiometric devices including , α,β,γ-spectrometers; transportable containers and other means for transportation of fuel and radioactive materials

  20. Reactor power measuring method and device therefor

    International Nuclear Information System (INIS)

    The present invention concerns measurement of a BWR type reactor power and provides a method of and a device for ensuring accuracy of calibration of sensitivity of neutron detectors and measurement of reactor power even if γ-ray thermometers are failed. Namely, the output signals of the γ-ray thermometers are compared with previously determined judging values to detect failures. The reactor power is measured based on the signals of neutron detectors calibrated by integral thermometers except for neutron detectors calibrated by γ-ray thermometers detected as failed. Calibration for sensitivity of neutron detectors as objects of γ-ray thermometers detected as failed is preferably prohibited. Accuracy of measurement of the reactor power can be ensured by the method described above. If axial power distribution of the reactor core is measured while eliminating the signals of γ-ray thermometers detected as failed, accuracy of the measurement of axial power distribution can be ensured. (N.H.)

  1. RA research reactor - potentials and prospective

    International Nuclear Information System (INIS)

    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

  2. The program of reactors and nuclear power plants

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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

  4. Effective utilization and management of research reactors

    International Nuclear Information System (INIS)

    The problem of utilizing a research reactor effectively is closely related to its management and therefore should not be considered separately. Too often, attention has been focused on specific techniques and methods rather than on the overall programme of utilization, with the result that skills and equipment have been acquired without any active continuing programme of applications and services. The seminar reported here provided a forum for reactor managers, users, and operators to discuss their experience. At the invitation of the Government of Malaysia, it was held at the Asia Pacific Development Centre, Kuala Lumpur, from 7 to 11 November 1983. It was attended by about 50 participants from 19 Member States; it is hoped that a report on the seminar, including papers presented, can be published and thus reach a wider audience. Thirty-one lectures and contributions were presented at a total of seven sessions: Research reactor management; Radiation exposure and safety; Research reactor utilization (two sessions); PUSPATI Research Reactor Project Development; Core conversion to low-enriched uranium, and safeguards; Research reactor technology. In addition, a panel discussed the causes and resolutions of the under-utilization of research reactors

  5. Nuclear reactor philosophy and criteria

    International Nuclear Information System (INIS)

    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

  6. 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.

  7. 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...... 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...

  8. Reactor accident analysis and evaluation

    International Nuclear Information System (INIS)

    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)

  9. Mechanical core coupling and reactors stability

    International Nuclear Information System (INIS)

    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

  10. Nuclear reactors for research and radioisotope production in Argentina

    International Nuclear Information System (INIS)

    In Argentina, the construction, operation, and use of research and radioisotope production reactors is and has been an important method of personnel preparation for the nuclear power program. Moreover, it is a very suitable means for technology transfer to countries developing their own nuclear programs. At present, the following research reactors are in operation in Argentina: Argentine Reactor 0 (RA-0); Argentine Reactor 1 (RA-1); Argentine Reactor 2 (RA-2); Argentine Reactor 3 (RA-3); Argentine Reactor 4 (RA-4). The Argentine Reactor 6 (RA-6), under construction, should reach criticality in 1981

  11. Evaluating the use and limitations of the Danish National Hospital Register in registry-based research using an example of multiple sclerosis

    DEFF Research Database (Denmark)

    Mason, K; Thygesen, Lau Caspar; Stenager, Egon; Brønnum-Hansen, Henrik; Koch-Henriksen, Nils

    2012-01-01

    BACKGROUND: The Danish National Patient Register, Landspatientregistret (LPR), is a register of all hospital discharges and outpatient treatments in Denmark. AIMS: It is increasingly used in research so it is important to understand to what extent this can be used as an accurate source of...... epidemiological MS research, in particular incidence studies. The study also found that the completeness of the LPR could be increased to 92.8% by including LPR records from other departments in addition, but this reduced the validity of the LPR to 95.1%. However, these results cannot uncritically be applied to...

  12. Monitoring and control of anaerobic reactors

    DEFF Research Database (Denmark)

    Pind, Peter Frode; Angelidaki, Irini; Ahring, Birgitte Kiær; Stamatelatou, K.; Lyberatos, G.

    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...... 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....

  13. Monitoring and control of anaerobic reactors

    DEFF Research Database (Denmark)

    Pind, Peter Frode; Angelidaki, Irini; Ahring, Birgitte Kiær; Stamatelatou, K.; Lyberatos, G.

    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...... 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....

  14. Space reactors - past, present, and future

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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.)

  16. The CANDUR Reactor - The Practical Path to RU and TH use in Nuclear Reactors

    International Nuclear Information System (INIS)

    The CANDU heavy water reactor has unrivalled flexibility for using a variety of fuels, such as Natural Uranium (NU), Low Enriched Uranium (LEU), Recycled Uranium (RU), Mixed Oxide (MOX), and Thorium (Th). Recently, this unique CANDU reactor feature attracted considerable attention due to favourable commercial, environmental and strategic needs. This paper summarizes the solid progress over the last three years and outlines CANDU Energy Incorporated's (CEI) multi-stage vision of utilizing various fuels in currently operational and new build CANDU reactors. In CEI's fuel-cycle vision, CANDU reactors will operate in conjunction with other reactor types and use advanced fuels to produce more energy and ensure the most efficient and least costly method of utilizing Light Water Reactor (LWR) used fuel. With this vision and the tandem goal of systematic adoption of Thorium based fuels, CANDU reactors will be a strong technology partner in ensuring the availability of long-term stable resources for nuclear power plants

  17. Safety and environmental aspects of fusion reactors

    International Nuclear Information System (INIS)

    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)

  18. Current status and prospects of research reactors

    International Nuclear Information System (INIS)

    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

  19. Current status and prospects of research reactors

    International Nuclear Information System (INIS)

    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 range 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 for 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

  20. Organization and management of operation of the research reactor MARIA

    International Nuclear Information System (INIS)

    The MARIA research reactor belongs to the Institute of Atomic Energy. The MARIA research reactor operation provides basing on the Atomic Law code and requirements of the State Nuclear Safety. Main task of the operation Department is the current MARIA reactor operation and relevant technological systems. The Head of the Reactor bears the direct responsibility for nuclear safety and radiological protection of the reactor plant. Service of reactor operation is accomplished by the Shift Groups. The cooperation with the reactor users is based on the principles defined by the Regulations of MARIA Reactor Operation. In the abnormal and emergency states the procedure is determined by 'Schedule of emergency procedure for the MARIA reactor plant'. Reactor has got valid and actual documents which are compulsory to all the persons being involved in operation and usage of reactor. (author)

  1. Intrinsically Safe and Economical Reactor (ISER)

    International Nuclear Information System (INIS)

    The Intrinsically Safe and Economical Reactor (ISER) is designed based on the principle of a process inherent ultimate safe reactor, PIUS, a so-called inherently safe reactor (ISR). ISER has been developed joingly by the members of the Kanagawa Institute of Technology, the University of Tokyo, the Japan Atomic Energy Research Institute (JAERI) and several industrial firms in Japan. This paper describes the requirements for the next generation of power reactor, the safety design philosphy of ISR and ISER, the controllability of ISER and the results of analyses of some of the design-based accidents (DBA) of ISER, namely station blackout, accidents in which the pressurizer relief valve becomes jammed and stuck in open position and tube breaks in the steam generator. It is concluded that the ISER can ensure a wide range of contraollabitily and fuel integrity for all the analysed DBAs. (orig.)

  2. SVBR-75: a reactor module for renewal of WWER-440 decommissioning reactors - safety and economic aspects

    International Nuclear Information System (INIS)

    IPPE has been developing jointly with EDB 'Gidropress' a reactor called SVBR-75 (in English spelling LBFR-75: Lead-Bismuth Fast Reactor), an innovative heavy metal reactor with transparent safety characteristics. The SVBR-75 Reactor Module is designed for the steam production instead of WWER-440 reactors to be decommissioned. At the renewal of the NPP unit the reactor vessel is remained at the same place, and steam generators are replaced by the new modules for SVBR-75. The main targets are: passively safe behaviour, no pressurisation of the reactor containment (building) under any accident conditions, reduction of plant capital costs and the construction schedule by means of the modular concept and the compact layout. The SVBR-75 design, described in general terms in this paper, is based upon the safety concept proposed by IPPE and on proven technology from both LMR and HMR for submarines. (author)

  3. MIT research reactor. Power uprate and utilization

    International Nuclear Information System (INIS)

    The MIT Research Reactor (MITR) is a university research reactor located on MIT campus. and has a long history in supporting research and education. Recent accomplishments include a 20% power rate to 6 MW and expanding advanced materials fuel testing program. Another important ongoing initiative is the conversion to high density low enrichment uranium (LEU) monolithic U-Mo fuel, which will consist of a new fuel element design and power increase to 7 MW. (author)

  4. RA reactor operation and maintenance in 1999, Part 1

    International Nuclear Information System (INIS)

    Activities at the RA reactor in 1999 were defined according to the needs of maintaining the reactor components and systems according to the existing funding. Basic activities during the past year were related to the maintenance of the reactor devices which must be in constant operation (special and regular ventilation power supply system, radioactivity and contamination control system, internal transportation system), reactor security system, and other systems that are useful independent of the future status of the reactor. (secondary cooling system, hot cells). maintenance of the reactor building was done on a limited scale due to lack of financial support. Possible solutions for the future status of the RA reactor discussed in this report are: renewal of reactor components for the reactor restart, conservation of the reactor (temporary shutdown) or permanent reactor shutdown

  5. RA reactor operation and maintenance in 1998, Part 1

    International Nuclear Information System (INIS)

    Activities at the RA reactor in 1998 were defined according to the needs of maintaining the reactor components and systems according to the existing funding. Basic activities during the past year were related to the maintenance of the reactor devices which must be in constant operation (special and regular ventilation power supply system, radioactivity and contamination control system, internal transportation system), reactor security system, and other systems that are useful independent of the future status of the reactor. (secondary cooling system, hot cells). maintenance of the reactor building was done on a limited scale due to lack of financial support. Possible solutions for the future status of the RA reactor discussed in this report are: renewal of reactor components for the reactor restart, conservation of the reactor (temporary shutdown) or permanent reactor shutdown

  6. Small and medium power reactors 1987

    International Nuclear Information System (INIS)

    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

  7. Blankets for fusion reactors : materials and neutronics

    International Nuclear Information System (INIS)

    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)

  8. Small reactors and the 'second nuclear era'

    International Nuclear Information System (INIS)

    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)

  9. MODELING AND SIMULATION OF AMMONIA SYNTHESIS REACTOR

    Directory of Open Access Journals (Sweden)

    Madjid Kakavand

    2006-06-01

    Full Text Available In this paper an industrial ammonia synthesis reactor has been modeled. The reactor under study is of horizontal type. This reactor which is under the license of Kellogg Company is equipped with three axial flow catalyst beds and an internal heat exchanger in accompany with a cooling flow. The achieved modeling is one dimensional and non-homogenous. Considering the sever effect of internal heat exchanger on reactor operation, it has been simulated by calculation of film heat transfer coefficients in its tube and shell and then, taking into account the shell thermal resistance and fouling coefficient, obtaining the overall heat transfer coefficient. So in the developed software, the heat transfer coefficient is first calculated using the conditions of the input flow to the exchanger and then the input flows to the first and second beds are calculated. The differential equations have been solved using Rung Kutta 4 method and the results have been compared with the available industrial data. Finally the capability of the developed software for industrial application has been investigated by changing the reactor operation conditions and studying their effects on reactor output.

  10. Gas-cooled reactors and their applications

    International Nuclear Information System (INIS)

    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

  11. Research reactors: a tool for science and medicine

    International Nuclear Information System (INIS)

    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

  12. Future fuel cycle and reactor strategies

    International Nuclear Information System (INIS)

    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)

  13. TITAN program and direct cycle fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kato, Yasuyoshi; Yoshizawa, Yoshio; Nitawaki, Takeshi [Research Laboratory for Nuclear Reactors, Tokyo Institute of Technology, Tokyo (Japan)

    2000-07-01

    In December 1999, the Research Laboratory for Nuclear Reactors of the Tokyo Institute of Technology (TIT) started a new program for the development of advanced nuclear reactors with small and medium size. TITAN is the acronym for the program. A novel concept of a carbon dioxide cooled direct cycle fast reactor with a Rankin cycle has been proposed as the advanced nuclear reactors and evaluated for an alternative option to liquid metal cooled fast reactors (LMFRs). The use of carbon dioxide as coolant eliminates major safety related problems of sodium cooled fast reactors: positive sodium void reactivity, hazardous reaction between sodium and water or air. The decay heat is passively removed by allocating a storage tank of liquidized carbon dioxide between the regenerator and the condenser, and by introducing naturally the carbon dioxide vaporized from the tank into the core in the event of the depressurization accident. The direct cycle results in considerable simplification of the heat transport system owing to the absence of intermediate cooling and water-steam loops comparing with the LMFRs. The thermal efficiency of the direct cycle is evaluated as 34.3 %, which is slightly higher than those in the current BWRs and PWRs. (author)

  14. RA Reactor operation and maintenance (I-IX), Part IV, Task 3.08/04, Refurbishment of the RA reactor

    International Nuclear Information System (INIS)

    This volume contains reports describing maintenance and repair work of the RA reactor instrumentation, equipment of the reactor dosimetry control system, and equipment for regulation and control systems

  15. History and evolution of the breeder reactor

    International Nuclear Information System (INIS)

    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 235U 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 238U, so poorly used by other types of reactors, into fissile 239Pu. 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

  16. The effect of heavy water reactors and liquid fuel reactors on the long-term development of nuclear energy

    International Nuclear Information System (INIS)

    The effects of the rates at which various combinations of power reactor types are installed on the long-range (to the year 2040) uranium and plutonium inventory requirements are examined. Consideration is given to light water reactors, fast breeder reactors, high temperature gas-cooled reactors, heavy water reactors, and thermal breeder reactors, in various combinations, and assuming alternatively a 3% and a 5% growth in energy demand

  17. Comparisons of prediction methods for peak cladding temperature and effective thermal conductivity in spent fuel assemblies of transportation/storage casks

    International Nuclear Information System (INIS)

    Highlights: • Peak cladding temperature (PCT) of spent fuel were evaluated by various methods. • The methods are Wooton–Epstein correlation, two-region model, and CFD. • Temperature difference between two-region and CFD ranges from −0.2 to 9 K. • CFD could be used to calculate PCT because of over-predicting PCT of two-region. • Application using CFD was conducted for spent fuel assembly used in Republic of Korea. - Abstract: When spent fuel assemblies from the reactor of nuclear power plants (NPPs) are transported or stored, the assemblies are exposed to a variety of environments that can affect the peak cladding temperature. There are three models to calculate the peak cladding temperature of spent fuel assemblies in a cask: Manteufel and Todreas’s two-region model, Bahney Lotz’s effective thermal conductivity model, and Wooton–Epstein correlation. The peak cladding temperatures of Babcock and Wilcox (B and W) 15 × 15 PWR spent fuel assembly under helium backfill gas were evaluated by using two-dimensional CFD simulation and compared with two models (Wooton–Epstein correlation, two-region model). The peak cladding temperature difference between the two-region model and CFD simulation ranges from −0.2 K to 9 K. Two-region model over-predicts the measured peak cladding temperature that performs in a spent fuel dry storage cask. Therefore the simulation could be used to calculate peak cladding temperature of spent fuel assemblies. Application using CFD simulation was conducted to investigate the peak cladding temperature and effective thermal conductivity of spent fuel assembly used in Korea NPPs: 16 × 16 (CE type) and 17 × 17 (WH type) PWR spent fuel assembly. CFD simulation results are similar to each other, and the difference of temperature drop between the three arrays occurs slightly in all basket wall temperatures. The effective thermal conductivity calculated from the 16 × 16 PWR spent fuel assembly results was more conservative

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    International Nuclear Information System (INIS)

    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

  20. Operation and utilizations of Dalat nuclear research reactor

    International Nuclear Information System (INIS)

    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)

  1. The construction, installation and commissioning of the PUSPATI TRIGA reactor

    International Nuclear Information System (INIS)

    A TRIGA Mark II research reactor has been installed at the Tun Ismail Atomic Research Centre (PUSPATI), Selangor, Malaysia. The reactor was commissioned in July 1982. With the commissioning of the reactor, a new era in the development of nuclear science and technology in Malaysia has just begun. This report describes the construction, installation and commissioning of the reactor. (author)

  2. Argentinean integrated small reactor design and scale economy analysis of integrated reactor

    International Nuclear Information System (INIS)

    This paper describes the design of CAREM, which is Argentinean integrated small reactor project and the scale economy analysis results of integrated reactor. CAREM project consists on the development, design and construction of a small nuclear power plant. CAREM is an advanced reactor conceived with new generation design solutions and standing on the large experience accumulated in the safe operation of Light Water Reactors. The CAREM is an indirect cycle reactor with some distinctive and characteristic features that greatly simplify the reactor and also contribute to a highly level of safety: integrated primary cooling system, self pressurized, primary cooling by natural circulation and safety system relying on passive features. For a fully doupled economic evaluation of integrated reactors done by IREP (Integrated Reactor Evaluation Program) code transferred to IAEA, CAREM have been used as a reference point. The results shows that integrated reactors become competitive with power larger than 200MWe with Argentinean cheapest electricity option. Due to reactor pressure vessel construction limit, low pressure drop steam generator are used to reach power output of 200MWe for natural circulation. For forced circulation, 300MWe can be achieved. (author)

  3. Development and demonstration of an advanced extended-burnup fuel-assembly design incorporating urania-gadolinia. Second semi-annual progress report, October 1981-March 1982

    Energy Technology Data Exchange (ETDEWEB)

    Newman, L W; Rombough, C T; Thornton, T A

    1982-08-01

    The Babcock and Wilcox Company, Duke Power Company, and the US Department of Energy are participating in an extended-burnup program for pressurized water reactors that will demonstrate an advanced fuel assembly design. This advanced fuel assembly will use a UO/sub 2/-Gd/sub 2/O/sub 3/ burnable-poison fuel mixture along with other state-of-the-art fuel performance and uranium utilization-enhancing design features that include annular pellets, annealed guide tubes, Zircaloy intermediate grids, and removable upper end fittings. Comparisons of the thermal properties of UO/sub 2/-Gd/sub 2/O/sub 3/ specimens containing 2.98, 5.66, and 8.50 wt % Gd/sub 2/O/sub 3/ with UO/sub 2/ specimens showed that thermal conductivity is the only thermal parameter significantly affected by the addition of Gd/sub 2/O/sub 3/. The milling steps used to prepare UO/sub 2/-Gd/sub 2/O/sub 3/ powder result in a powder that is more active than standard UO/sub 2/ powder. As a result, UO/sub 2/-Gd/sub 2/O/sub 3/ fuel has shown more variability than UO/sub 2/ fuel in as-sintered theoretical density and densification behavior. However, a poreforming material, added to the UO/sub 2/-Gd/sub 2/O/sub 3/ powder mixture before sintering, can be used to achieve the desired density. Measured results from critical experiments were compared with predicted data and confirmed the accuracy of the standard two-group diffusion theory model for predicting global and discrete UO/sub 2/-Gd/sub 2/O/sub 3/ effects when cross-section input is appropriately adjusted. The preliminary first two fuel cycles for lead test assemblies of the advanced design were developed. Irradiation of the lead test assemblies is scheduled to begin in 1983 in Duke Power Company's Oconee Unit 1. An intercalibrated movable incore detector system will be used to monitor the performance of the test assemblies during irradiation.

  4. Sodium fast neutron reactors. Status and perspective of development

    International Nuclear Information System (INIS)

    This report reveals data on development history of domestic fast neutron reactors cooled with sodium (BN reactors). It also shows BN reactors' unique role in expanding source of nuclear power raw materials and in solving ecological problems relating to radioactive wastes. There is brief information on characteristics and operation experience of research reactors BR-10, BOR-60, pilot-industrial reactors BN-350 and BN-600. As well there is data on BN-800 reactor designing that obtained a license for building. There are considered BN reactor peculiarities in regard of safety and design decisions on safety provision at the level meeting standard document requirements. BN reactor technical and economic indices and the ways of their improvement are evaluated. There is brief information on alternative perspective technologies of fast reactors, in particular regarding 'BREST-300' reactor cooled with lead coolant

  5. A coupled $2\\times2$D Babcock-Leighton solar dynamo model. II. Reference dynamo solutions

    CERN Document Server

    Lemerle, Alexandre

    2016-01-01

    In this paper we complete the presentation of a new hybrid $2\\times2$D flux transport dynamo (FTD) model of the solar cycle based on the Babcock-Leighton mechanism of poloidal magnetic field regeneration via the surface decay of bipolar magnetic regions (BMRs). This hybrid model is constructed by allowing the surface flux transport (SFT) simulation described in Lemerle et al. 2015 to provide the poloidal source term to an axisymmetric FTD simulation defined in a meridional plane, which in turn generates the BMRs required by the SFT. A key aspect of this coupling is the definition of an emergence function describing the probability of BMR emergence as a function of the spatial distribution of the internal axisymmetric magnetic field. We use a genetic algorithm to calibrate this function, together with other model parameters, against observed cycle 21 emergence data. We present a reference dynamo solution reproducing many solar cycle characteristics, including good hemispheric coupling, phase relationship betwe...

  6. Technology, safety and costs of decommissioning nuclear reactors at multiple-reactor stations

    International Nuclear Information System (INIS)

    Safety and cost information is developed for the conceptual decommissioning of large (1175-MWe) pressurized water reactors (PWR) and large (1155-MWe) boiling water reactors (BWR) at multiple-reactor stations. Three decommissioning alternatives are studied: DECON (immediate decontamination), SAFSTOR (safe storage followed by deferred decontamination), and ENTOMB (entombment). Safety and costs of decommissioning are estimated by determining the impact of probable features of multiple-reactor-station operation that are considered to be unavailable at a single-reactor station, and applying these estimated impacts to the decommissioning costs and radiation doses estimated in previous PWR and BWR decommissioning studies. The multiple-reactor-station features analyzed are: the use of interim onsite nuclear waste storage with later removal to an offsite waste disposal facility, the use of permanent onsite nuclear waste disposal, the dedication of the site to nuclear power generation, and the provision of centralized services

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

    International Nuclear Information System (INIS)

    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)

  8. 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.

  9. Styrene-maleic anhydride copolymerization in a recycle tubular reactor: reactor stability and product quality

    OpenAIRE

    Belkhiria, Sahbi; Meyer, Thierry; Renken, Albert

    1994-01-01

    A tubular recycle reactor was developed to ensure good homogeneity of concn. and temp. in the copolymn. of styrene and maleic anhydride. The compn. of the copolymer obtained is in good agreements with predicted values and the uniformity of compn. was measured for the entire mol.-wt. distribution. The characterization of the reactor (both hydrodynamic and stability) and the quality of the resulting polymer are presented herein. The limits of use of this reactor for the styrene-maleic anhydride...

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

    International Nuclear Information System (INIS)

    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

  11. Operating experience of BN-600 fast neutron reactor and BN-800 reactor design

    International Nuclear Information System (INIS)

    Full text: Experience gained in Russia (USSR) in R and D work in the area of sodium cooled fast reactors in the period of 1950-1970s has been used in the design of NPP with the BN-600 reactor. Since its start-up in 1980, BN-600 reactor has demonstrated operating characteristics, which are unique for this nuclear technology. Average load factor value for 23 years of operation is near 74%, its values in 2002 and 2003 being respectively 77.35% and 75.7%. Release of inert radioactive gases is within 0.3% of reference value, while average collective dose rate of personnel is about 0.3 man. Sv per year. In the course of operation of NPP with the BN-600 reactor, effectiveness of steam generator protection system was demonstrated in 12 cases of small and large water-into-sodium leaks. Besides, unique experience was gained in confining either radioactive and non-radioactive sodium fires in case of sodium leaks from the circuits. Radioactive sodium leak from the primary auxiliary circuit occurred in December 1994 is a typical example. Total amount of sodium released from the circuit was about 1000 kg, and protection system was capable of confining sodium ignition nucleation site and limiting radioactivity release to the atmosphere by 10 Ci value. This release has had almost zero effect on radiological conditions of the NPP controlled area. BN-800 reactor design is the next stage of development of sodium cooled fast reactor technology. Fourth power unit with the BN-800 reactor is now under construction on Beloyarskaya NPP site. Innovative design approaches have been used in the BN-800 reactor in order to further improve safety of fast reactors with sodium coolant. Among these innovations are as follows: Additional 'passive' safety system using three absorber rods hydraulically suspended by the sodium flow; Passive decay heat removal system using sodium-air heat exchangers; Device for collection and retaining of the core debris in case of its disruption under conditions of

  12. Reactor modeling and physicochemical properties characterization for a polyethylene fluidized bed reactor

    OpenAIRE

    F. A. N. Fernandes; L. M. F. LONA BATISTA

    1999-01-01

    A new steady state model for the fluidized bed reactor and a physicochemical characterization model were developed for the simulation of polyethylene production using gas-phase technology. The association of these models was done in order to predict the characteristics of the polymer produced in the fluidized bed reactor (molecular weight, polydispersity, melt index, and other characteristics) throughout the reactor and also to predict the growth of the polymer particle.

  13. Reactor modeling and physicochemical properties characterization for a polyethylene fluidized bed reactor

    Directory of Open Access Journals (Sweden)

    FERNANDES F. A. N.

    1999-01-01

    Full Text Available A new steady state model for the fluidized bed reactor and a physicochemical characterization model were developed for the simulation of polyethylene production using gas-phase technology. The association of these models was done in order to predict the characteristics of the polymer produced in the fluidized bed reactor (molecular weight, polydispersity, melt index, and other characteristics throughout the reactor and also to predict the growth of the polymer particle.

  14. Reactor and fuel assembly design for improved fuel utilization in liquid moderated thermal reactors

    International Nuclear Information System (INIS)

    An improved reactor and fuel assembly design is disclosed wherein a light water reactor is initially run with undermoderated fuel assemblies to take advantage of increased conversion ratio, and after a suitable period of operation, the neutron spectrum for the undermoderated assemblies is shifted to lower energies to increase reactivity by withdrawing a number of fuel rods from the assemblies. The increased reactivity allows for continued operation of the modified assembly, and the fuel rods which are removed are used to construct similar assemblies which are also capable of continued operation. The improved reactor and fuel assembly design results in improved fuel utilization and neutron economy and reduced control requirements for the reactor

  15. Optimally moderated nuclear fission reactor and fuel source therefor

    Science.gov (United States)

    Ougouag, Abderrafi M.; Terry, William K.; Gougar, Hans D.

    2008-07-22

    An improved nuclear fission reactor of the continuous fueling type involves determining an asymptotic equilibrium state for the nuclear fission reactor and providing the reactor with a moderator-to-fuel ratio that is optimally moderated for the asymptotic equilibrium state of the nuclear fission reactor; the fuel-to-moderator ratio allowing the nuclear fission reactor to be substantially continuously operated in an optimally moderated state.

  16. Conceptual design of nuclear fusion power reactor DREAM. Reactor structures and remote maintenance

    International Nuclear Information System (INIS)

    Nuclear fusion reactors are required to be able to compete another energy sources in economy, reliability, safety and environmental integrity for commercial use. In the DREAM (DRastically EAsy Maintenance) reactor, a very low activated material of SiC/SiC composite has been introduced for the structural material, a reactor configuration for very easy maintenance and the helium gas of a high temperature for the cooling system, and hence DREAM has been proven to be very attractively as the commercial power reactor due to the high availability and efficiency of the plant and minimization of radioactive wastes. (author)

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

    International Nuclear Information System (INIS)

    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

  18. 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...... 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...... 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....

  19. Nuclear reaction data and nuclear reactors

    International Nuclear Information System (INIS)

    These two volumes contain the lecture notes of the workshop 'Nuclear Reaction Data and Nuclear Reactors: Physics, Design and Safety', which was held at the Abdus Salam ICTP in the Spring of 2000. The workshop consisted of five weeks of lecture courses followed by practical computer exercises on nuclear data treatment and design of nuclear power systems. The spectrum of topics is wide enough to timely cover the state-of-the-art and the perspectives of this broad field. The first two weeks were devoted to nuclear reaction models and nuclear data evaluation. Nuclear data processing for applications to reactor calculations was the subject of the third week. On the last two weeks reactor physics and on-going projects in nuclear power generation, waste disposal and safety were presented

  20. Reactor Physics Modeling Of Spent Nuclear Research Reactor Fuel For SNM Attribution And Nuclear Forensics

    International Nuclear Information System (INIS)

    Nuclear research reactors are the least safeguarded type of reactor; in some cases this may be attributed to low risk and in most cases it is due to difficulty from dynamic operation. Research reactors vary greatly in size, fuel type, enrichment, power and burnup providing a significant challenge to any standardized safeguard system. If a whole fuel assembly was interdicted, based on geometry and other traditional forensics work, one could identify the material's origin fairly accurately. If the material has been dispersed or reprocessed, in-depth reactor physics models may be used to help with the identification. Should there be a need to attribute research reactor fuel material, the Savannah River National Laboratory would perform radiochemical analysis of samples of the material as well as other non-destructive measurements. In depth reactor physics modeling would then be performed to compare to these measured results in an attempt to associate the measured results with various reactor parameters. Several reactor physics codes are being used and considered for this purpose, including: MONTEBURNS/ORIGEN/MCNP5, CINDER/MCNPX and WIMS. In attempt to identify reactor characteristics, such as time since shutdown, burnup, or power, various isotopes are used. Complexities arise when the inherent assumptions embedded in different reactor physics codes handle the isotopes differently and may quantify them to different levels of accuracy. A technical approach to modeling spent research reactor fuel begins at the assembly level upon acquiring detailed information of the reactor to be modeled. A single assembly is run using periodic boundary conditions to simulate an infinite lattice which may be repeatedly burned to produce input fuel isotopic vectors of various burnups for a core level model. A core level model will then be constructed using the assembly level results as inputs for the specific fuel shuffling pattern in an attempt to establish an equilibrium cycle. The

  1. Hydraulic characteristics of the N Reactor core and reactor cooling system

    International Nuclear Information System (INIS)

    In conjunction with the NUSAR program, the need was recognized for well substantiated pressure drop correlations for the N Reactor core to support in-depth safety analysis consistent with currently-available technology. Additionally, it was considered desirable to reconfirm the hydraulic characteristics of the reactor coolant system in the light of improved understanding of the hydraulic features of the current reactor fuel loading. The report summarizes the results of laboratory tests and analysis accomplished to meet the above objectives

  2. Fusion-Fission hybrid reactors and nonproliferation

    International Nuclear Information System (INIS)

    New options for the development of the nuclear energy economy which might become available by a successful development of fusion-breeders or fusion-fission hybrid power reactors, identified and their nonproliferative attributes are discussed. The more promising proliferation-resistance ettributes identified include: (1) Justification for a significant delay in the initiation of fuel processing, (2) Denaturing the plutonium with 238Pu before its use in power reactors of any kind, and (3) Making practical the development of denatured uranium fuel cycles and, in particular, denaturing the uranium with 232U. Fuel resource utilization, time-table and economic considerations associated with the use of fusion-breeders are also discussed. It is concluded that hybrid reactors may enable developing a nuclear energy economy which is more proliferation resistant than possible otherwise, whileat the same time, assuring high utilization of t he uranium and thorium resources in an economically acceptable way. (author)

  3. Design features of BREST reactors and experimental work to advance the concept of BREST reactors

    International Nuclear Information System (INIS)

    Principle design features of BREST-300 (300 MWe) and BREST-1200 (1200 MWe) lead.cooled fast reactors are presented in this paper. Several experimental works have been performed or under way in order to justify lead-cooled reactor design concepts. BREST reactor designs of different outputs have been developed using the same principles. In conjunction with the increased output and the implement of inherent safety concept, a number of new solutions, which may be applied to the BREST-300 reactor design too, have been considered in the BREST-1200 reactor design. The new design features adopted in the BREST-1200 reactor design include: pool-type reactor design not requiring metal vessel, hence, not limiting reactor power; new handling system allowing to reduce central hall and building dimensions as a whole; emergency cooling system using field pipes, immersed directly in lead, which may be used to cool down reactor under normal conditions; by--pass line incorporated in coolant loop allowing to refuse the actively actuating valve initiated in pumps shut down. (author)

  4. Reactor Neutrino Experiments: $\\theta_{13}$ and Beyond

    CERN Document Server

    Qian, X

    2014-01-01

    We review the current-generation short-baseline reactor neutrino experiments that have firmly established the third neutrino mixing angle $\\theta_{13}$ to be non-zero. The relative large value of $\\theta_{13}$ (around 9$^\\circ$) has opened many new and exciting opportunities for future neutrino experiments. Daya Bay experiment with the first measurement of $\\Delta m^2_{ee}$ is aiming for a precision measurement of this atmospheric mass-squared splitting with a comparable precision as $\\Delta m^2_{\\mu\\mu}$ from accelerator muon neutrino experiments. JUNO, a next-generation reactor neutrino experiment, is targeting to determine the neutrino mass hierarchy with medium baselines ($\\sim$50 km). Beside these {\\color{black} opportunities enabled by the large $\\theta_{13}$}, the current-generation (Daya Bay, Double Chooz, and RENO) and the next-generation (JUNO, RENO-50, and PROSPECT) reactor experiments, with their unprecedented statistics, are also leading the precision era of the 3-flavor neutrino oscillation phys...

  5. 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

  6. Bowing and interaction of fast reactor subassemblies

    International Nuclear Information System (INIS)

    Deformations of the subassembly structural components, in particular the bowing of the hexagonal wrapper which encloses the pin bundles, due to stainless steel swelling as a result of fast neutron irradiation give rise to operational and safety problems especially in large breeder reactors where the neutron flux is much larger than in smaller reactors. The restraint on bowing induces heavy restraint loads and high stresses in the wrapper, which tend to limit the target burn-up of the fast reactor fuels. Therefore, a realistic analysis has to include the phenomenon of creep to determine the extent to which the stresses in the wrapper would be relaxed due to both thermally induced and irradiation induced creep. Apart from this, determination of deformations of the subassemblies in the core due to the interaction among them is also necessary. (author)

  7. Innovative reactor core: potentialities and design

    International Nuclear Information System (INIS)

    Gen IV nuclear reactors are considered a very attractive answer for the demand of energy. Because public acceptance they have to fulfil very clearly the requirement of sustainable development. In this sense a reactor concept, having by itself a rather no significant interaction with the environment both on the front and back end ('adiabatic concept'), is vital. This goal in mind, a new way of designing such a core has to be assumed. The starting point must be the 'zero impact'. Therefore the core will be designed having as basic constraints: a) fed with only natural or depleted Uranium, and b) discharges only fission products. Meantime its potentiality as a net burner of Minor Actinide has to be carefully estimated. This activity, referred to the ELSY reactor, shows how to design such an 'adiabatic' core and states its reasonable capability of burning MA legacy in the order of 25-50 kg/GWey. (authors)

  8. Geoneutrinos and reactor antineutrinos at SNO+

    CERN Document Server

    Baldoncini, M; Wipperfurth, S A; Fiorentini, G; Mantovani, F; McDonough, W F; Ricci, B

    2016-01-01

    In the heart of the Creighton Mine near Sudbury (Canada), the SNO+ detector is foreseen to observe almost in equal proportion electron antineutrinos produced by U and Th in the Earth and by nuclear reactors. SNO+ will be the first long baseline experiment to measure a reactor signal dominated by CANDU cores ($\\sim$55\\% of the total reactor signal), which generally burn natural uranium. Approximately 18\\% of the total geoneutrino signal is generated by the U and Th present in the rocks of the Huronian Supergroup-Sudbury Basin: the 60\\% uncertainty on the signal produced by this lithologic unit plays a crucial role on the discrimination power on the mantle signal as well as on the geoneutrino spectral shape reconstruction, which can in principle provide a direct measurement of the Th/U ratio in the Earth.

  9. Reactor group constants and benchmark test

    International Nuclear Information System (INIS)

    The evaluated nuclear data files such as JENDL, ENDF/B-VI and JEF-2 are validated by analyzing critical mock-up experiments for various type reactors and assessing applicability for nuclear characteristics such as criticality, reaction rates, reactivities, etc. This is called Benchmark Testing. In the nuclear calculations, the diffusion and transport codes use the group constant library which is generated by processing the nuclear data files. In this paper, the calculation methods of the reactor group constants and benchmark test are described. Finally, a new group constants scheme is proposed. (author)

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

    International Nuclear Information System (INIS)

    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

  11. Oklo natural reactors: geological and geochemical conditions

    International Nuclear Information System (INIS)

    Published as well as unpublished material on the Oklo natural reactors in Gabon was evaluated with regard to the long-term aspects of nuclear waste disposal. Even though the vast data base available at present can provide only a site specific description of the phenomenon, already this material gives relevant information on plutonium retention, metamictization, fission product release, hydrogeochemical stability and migration of fission products. Generalized conclusions applicable to other nuclear waste repository would require the quantitative reconstruction of t s coupled thermo-hydrologic-chemical processes. This could be achieved by studying the deviations in the 2H/1H and 18O/16O ratios of minerals at Oklo. A further generalization of the findings from Oklo could be realized by examining the newly-discovered reactor zone 10, which was active under very different thermal conditions than the other reactors. 205 refs

  12. Pakistan research reactor-1 and its upgradation

    International Nuclear Information System (INIS)

    In this article the author describes the procedure of renovation and upgradation of a swimming pool type Pakistan Research Reactor-1 (PARR-1) installed at PINSTECH. The reactor originally designed for a thermal power of 5 MW using highly enriched uranium as has been upgraded 10 MW with low enriched uranium as fuel. All the required safety precaution has been also modified with the new requirements. The cooling system of PARR-1 was modified to meet the requirements of upgraded power of 10 MW. In order to ensure safety for upgraded PARR-1 and to bring the reactor the current safety standards, some additional safety systems have been provided. An emergency core cooling system ECCS has been installed to remove core decay heat in case of loss of coolant accident (LOCA). (A.B.)

  13. Distributed computing and nuclear reactor analysis

    International Nuclear Information System (INIS)

    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

  14. Technology, Safety and Costs of Decommissioning Nuclear Reactors At Multiple-Reactor Stations

    Energy Technology Data Exchange (ETDEWEB)

    Wittenbrock, N. G.

    1982-01-01

    Safety and cost information is developed for the conceptual decommissioning of large (1175-MWe) pressurized water reactors (PWRs) and large (1155-MWe) boiling water reactors {BWRs) at multiple-reactor stations. Three decommissioning alternatives are studied: DECON (immediate decontamination), SAFSTOR (safe storage followed by deferred decontamination), and ENTOMB (entombment). Safety and costs of decommissioning are estimated by determining the impact of probable features of multiple-reactor-station operation that are considered to be unavailable at a single-reactor station, and applying these estimated impacts to the decommissioning costs and radiation doses estimated in previous PWR and BWR decommissioning studies. The multiple-reactor-station features analyzed are: the use of interim onsite nuclear waste storage with later removal to an offsite nuclear waste disposal facility, the use of permanent onsite nuclear waste disposal, the dedication of the site to nuclear power generation, and the provision of centralized services. Five scenarios for decommissioning reactors at a multiple-reactor station are investigated. The number of reactors on a site is assumed to be either four or ten; nuclear waste disposal is varied between immediate offsite disposal, interim onsite storage, and immediate onsite disposal. It is assumed that the decommissioned reactors are not replaced in one scenario but are replaced in the other scenarios. Centralized service facilities are provided in two scenarios but are not provided in the other three. Decommissioning of a PWR or a BWR at a multiple-reactor station probably will be less costly and result in lower radiation doses than decommissioning an identical reactor at a single-reactor station. Regardless of whether the light water reactor being decommissioned is at a single- or multiple-reactor station: • the estimated occupational radiation dose for decommissioning an LWR is lowest for SAFSTOR and highest for DECON • the estimated

  15. International Experience with Fast Reactor Operation and Maintenance

    International Nuclear Information System (INIS)

    This paper reviews the most important lessons learned from operation of the world’s sodium cooled fast reactors, both test reactors and power producing reactors, which represent nearly 400 reactor-years of cumulative operating experience. The first reactor in the world to produce electricity was a fast reactor, the Experimental Breeder Reactor I, in December 1951. International experience with fast reactor technology exists in France, Germany, India, Japan, the Russian Federation, the United Kingdom and the United States of America. The operating experience with these reactors has been mixed; early problems were associated with fuel cladding, steam generators, fuel handling and sodium leakage. Excellent experience has been gained, however, that demonstrates the robust nature of the technology, the potential for exceedingly safe designs, ease of maintenance, ease of operation and the ability to effectively manage waste from spent fuel. It is a mature technology. (author)

  16. 400-MWe consolidated nuclear steam system (CNSS). Conceptual design. Executive summary

    International Nuclear Information System (INIS)

    There are a number of small nuclear unit concepts under active study. These include the Process Inherent Ultimate Safety (PIUS) unit and a smaller version of the High-Temperature Gas-Cooled Reactor (HTGR). This study has focused on the Consolidated Nuclear Steam System (CNSS) plant concept. Studies performed by The Babcock and Wilcox Company (B and W) and United Engineers and Constructors (UE and C) starting in 1974 have led to a 400 MW PWR CNSS plant concept of compact design. Recent economic studies for the CNSS plant show that it offers economic advantages for electric power generation in certain situations. This executive summary presents the results of these studies

  17. A proposed paradigm for solar cycle dynamics mediated via turbulent pumping of magnetic flux in Babcock-Leighton type solar dynamos

    CERN Document Server

    Hazra, Soumitra

    2016-01-01

    At present, Babcock-Leighton flux transport solar dynamo models appear as the most promising model for explaining diverse observational aspects of the sunspot cycle. The success of these flux transport dynamo models is largely dependent upon a single-cell meridional circulation with a deep equatorward component at the base of the Sun's convection zone. However, recent observations suggest that the meridional flow may in fact be very shallow (confined to the top 10 % of the Sun) and more complex than previously thought. Taken together these observations raise serious concerns on the validity of the flux transport paradigm. By accounting for the turbulent pumping of magnetic flux as evidenced in magnetohydrodynamic simulations of solar convection, we demonstrate that flux transport dynamo models can generate solar-like magnetic cycles even if the meridional flow is shallow. Solar-like periodic reversals is recovered even when meridional circulation is altogether absent, however, in this case the solar surface m...

  18. G2 and G3 reactors design

    International Nuclear Information System (INIS)

    The 'FRANCE ATOME' Manufacturers Party has been entrusted with the G2 and G3 reactors engineering by the french A.E.C., for the first-five-year french project. Although these reactors are essentially plutonium generators, everyone has been linked with a power station which is supposed to supply with 40 MW, 'Electricite de France' has taken the liability upon itself. The reactor core includes most of G1 reactor parts (central gap excluded): horizontal channels, graphite parallelepipedic bricks stacking, steel thermal shield. The cooling is provided with CO2 under a 15 atmospheres pressure. This pressure is kept steady in a press-stressed concrete packing-case which is a cylinder horizontally shaped. Steel strips tightened encircle the concrete cylinder; itself protected by sole-plates. The cylinder bottom has brought about unusual problems which have been solved by the choice of an hemispheric shape. Packing-case tightness is provided by a 30 mm iron-plate connected with the inner wall of concrete. One of the reactor's special characteristics is the possibility of loading and unloading while operating. On loading side, barrel locks, each weighting 50 tons, allow new cans, at a pressure of 15 atmospheres, to pass. The cans process almost in a steady way through the channel, and finally drop down through bent spouts, then through spiral toboggans into a new lock. The cooling CO2 flow is provided with 3 turbo-bellows, these are actuated by average pressure-steam, obtained from exchangers. Every reactor supplies 4 exchangers which have been very difficult to build and to set up. The secondary cycle is standard and contains 3 stages (pressure 10,3: 2 and 0,5 kg/cm2). Steam can be condensed in the event of a group turbo-generator stopping, with no modification for the normal operating conditions of the reactor. Auxiliary circuits have to assure the continuous purifying of cooling CO2, its storage and drain. 49 boron carbide rods are used to control the operating power

  19. BN-600 and BN-350 reactors

    International Nuclear Information System (INIS)

    The nuclear power plant (NPP) BN-600 has been operating since 1980 as the Beloyarsk-3 power plant. The NPP construction cost was ∼ 312 million Rubles [1980] [approximately 620 million US$ (1980)]. The planned budget was exceeded by less than 5%. First criticality was reached on 26 February 1980. The basic result of the physical startup in March 1980 (213 low (21%) enrichment fuel subassemblies (FSAs), 143 high (33%) enrichment FSAs and 13 permanent reactivity compensators) showed that the measured physical characteristics of the reactor were correspondent with the design values. Measurement of sodium flow through each FSA was carried out two times: before and after the power startup of the reactor

  20. The renaissance of fast sodium reactors 2007 assessment: situation and contributions from the Phenix experimental reactor

    Energy Technology Data Exchange (ETDEWEB)

    Guidez, J. [Phenix Plant (France)

    2007-07-01

    The first nuclear reactor to produce electrical current was the fast sodium/potassium reactor EBR-1 in Idaho (Usa). Following this pioneering experience, France, Germany, Great Britain, Usa, Japan, Russia and India launched construction of fast sodium reactors. In the post Chernobyl years, waves of protest against nuclear power grew and swelled, leading to a strong overall slowdown for this reactor type. The SNR-300 project in Germany never started up, and was shut down. In Great Britain, PFR was definitely shut down, operation of MONJU in Japan and BN-800 project in Russia were frozen, FFTF in the United States shut down, and finally the SPX-1 project in France was also stopped. When PHENIX started back up in 2003, there were only three other research reactors operating worldwide: FBTR in India, BOR-60 in Russia and JOYO in Japan, and one power reactor BN-600 in Russia. The Generation-IV initiative was the opportunity for global thinking about reactors for the future, referred to as fourth generation reactors. Six reactor designs were selected, including the fast sodium reactor. However, after several years, most of the countries have officially announced or confirmed that the fast sodium reactor is their priority reference design. In France, within the scope of the law of 28 June 2006, the country has announced and confirmed the decision to build a prototype scheduled for operation in 2020. These and other plans are all sustained in a very practical manner by the ongoing production in the field. PHENIX has been operating since 2003, contributing to the development of future systems and demonstrating the fast reactors ability to burn waste. Following the excellent results obtained by the BN-600, Russia has re-launched the BN-800 project. China is currently in the process of building a 65 MW research reactor, scheduled for divergence in 2009. In Japan, work is underway on MONJU for divergence in 2008. In India, a 1200 MW (thermal) power reactor is under

  1. The renaissance of fast sodium reactors 2007 assessment: situation and contributions from the Phenix experimental reactor

    International Nuclear Information System (INIS)

    The first nuclear reactor to produce electrical current was the fast sodium/potassium reactor EBR-1 in Idaho (Usa). Following this pioneering experience, France, Germany, Great Britain, Usa, Japan, Russia and India launched construction of fast sodium reactors. In the post Chernobyl years, waves of protest against nuclear power grew and swelled, leading to a strong overall slowdown for this reactor type. The SNR-300 project in Germany never started up, and was shut down. In Great Britain, PFR was definitely shut down, operation of MONJU in Japan and BN-800 project in Russia were frozen, FFTF in the United States shut down, and finally the SPX-1 project in France was also stopped. When PHENIX started back up in 2003, there were only three other research reactors operating worldwide: FBTR in India, BOR-60 in Russia and JOYO in Japan, and one power reactor BN-600 in Russia. The Generation-IV initiative was the opportunity for global thinking about reactors for the future, referred to as fourth generation reactors. Six reactor designs were selected, including the fast sodium reactor. However, after several years, most of the countries have officially announced or confirmed that the fast sodium reactor is their priority reference design. In France, within the scope of the law of 28 June 2006, the country has announced and confirmed the decision to build a prototype scheduled for operation in 2020. These and other plans are all sustained in a very practical manner by the ongoing production in the field. PHENIX has been operating since 2003, contributing to the development of future systems and demonstrating the fast reactors ability to burn waste. Following the excellent results obtained by the BN-600, Russia has re-launched the BN-800 project. China is currently in the process of building a 65 MW research reactor, scheduled for divergence in 2009. In Japan, work is underway on MONJU for divergence in 2008. In India, a 1200 MW (thermal) power reactor is under

  2. 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.

  3. Hydrogen and water reactor safety: proceedings

    International Nuclear Information System (INIS)

    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

  4. 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.

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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

  7. Reactor Physics and Core Design Issues: India

    International Nuclear Information System (INIS)

    In ADS, since reactor control system is not required to maintain criticality, it is possible to increase burnup i.e. to extract more energy from a given mass of fuel till such time that the keff of the system falls to a value below which it is no more economical to maintain the fission power merely by increasing accelerator current. An interesting argument in support of ADS-based thorium utilization emanates from possibility of starting such reactor without a seed fissile species. On the basis to these possibilities, development of calculation codes and some investigative simulations of ADS operation with them were carried out, which are presented in the following section

  8. 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)

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

    International Nuclear Information System (INIS)

    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

  10. 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.

  11. Exploitation questions regarding channel type reactors: water graphite channel reactors (operation, reconstruction, advantages and disadvantages)

    International Nuclear Information System (INIS)

    An overview of up-grade of the RBMK-type reactors is given. I this paper the core design and core monitoring, pressure boundary integrity, RBMK basic design and safety improvements emergency core cooling system (ECCS) as well as reactor cavity overpressure protection system (RCOPS) are discussed

  12. Fast Reactor and ADS development in China

    International Nuclear Information System (INIS)

    Conclusion: • The Fukushima accident influence China deeply. “The 12th five years plan and 2020 perspective goal of nuclear safety and radioactive pollution prevention” has been approved which means the nuclear may restart in the near future. • A demonstration fast reactor is under design. • More and more research works will be executed on CEFR

  13. Reactor antineutrino fluxes - status and challenges

    CERN Document Server

    Huber, Patrick

    2016-01-01

    In this contribution we describe the current understanding of reactor antineutrino fluxes and point out some recent developments. This is not intended to be a complete review of this vast topic but merely a selection of observations and remarks, which despite their incompleteness, will highlight the status and the challenges of this field.

  14. Effect of interphase friction on void fraction prediction

    International Nuclear Information System (INIS)

    The RELAP5/MOD1 transient thermal-hydraulic system analysis code has been developed to evaluate pressurized water reactor behavior during accident conditions. Benchmark cases have shown RELAP5/MOD1 to differ from the measured void distribution in certain relevant tests. In an attempt to achieve better agreement, the Wallis annular interphase drag correlation was replaced by the Bharathan-Wallis correlation in the Babcock and Wilcox version of RELAP5/MOD1 (REDBL5). The simulation of two Christensen's subcooled boiling tests using REDBL5 shows better agreement between the measured and calculated void distributions

  15. Power reactor noise studies and applications

    International Nuclear Information System (INIS)

    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

  16. 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

  17. Status and problems of fusion reactor development.

    Science.gov (United States)

    Schumacher, U

    2001-03-01

    Thermonuclear fusion of deuterium and tritium constitutes an enormous potential for a safe, environmentally compatible and sustainable energy supply. The fuel source is practically inexhaustible. Further, the safety prospects of a fusion reactor are quite favourable due to the inherently self-limiting fusion process, the limited radiologic toxicity and the passive cooling property. Among a small number of approaches, the concept of toroidal magnetic confinement of fusion plasmas has achieved most impressive scientific and technical progress towards energy release by thermonuclear burn of deuterium-tritium fuels. The status of thermonuclear fusion research activity world-wide is reviewed and present solutions to the complicated physical and technological problems are presented. These problems comprise plasma heating, confinement and exhaust of energy and particles, plasma stability, alpha particle heating, fusion reactor materials, reactor safety and environmental compatibility. The results and the high scientific level of this international research activity provide a sound basis for the realisation of the International Thermonuclear Experimental Reactor (ITER), whose goal is to demonstrate the scientific and technological feasibility of a fusion energy source for peaceful purposes. PMID:11402837

  18. Solution structure of loperamide and -cyclodextrin inclusion complexes using NMR spectroscopy

    Indian Academy of Sciences (India)

    Santosh Kumar Upadhyay; Syed Mashhood Ali

    2009-07-01

    Loperamide (LPR) is a synthetic, poorly water soluble, peripherally acting opiate agonist drug used for the treatment of diarrhea. Major challenges in formulating this drug for clinical applications include solubility enhancement and improved stability in biological systems. Cyclodextrins (CDs) are chiral, truncated cone shaped; cyclic oligosaccharides that can encapsulate a variety of poorly water soluble drug molecules into inclusion complexes, thereby increasing their stability and solubility. 1H NMR spectroscopic studies showed the inclusion complexation between -CD and LPR, based on the upfield shift changes in the -CD cavity protons (H-3' and H-5') and downfield shift changes in the guest (LPR) protons. 2D COSY spectral data was used for assignment of -CD as well as LPR protons and 2D ROESY spectral data to know the inclusion of LPR inside the -CD cavity. The 1 : 1 stoichiometry and overall association constant (a) were determined by using Scott’s plot method to be 68.805 M-1. 2D ROESY spectral data suggest that the inclusion of aromatic rings of LPR in -CD cavity can be from narrower as well as the wider rim side and the six possible 1 : 1 LPR : -CD inclusion complexes have been proposed. Thus, we anticipate that complexation of LPR with -CD would increase its solubility and stability in biological system.

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

    International Nuclear Information System (INIS)

    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

  1. Design guide for category II reactors light and heavy water cooled reactors. [US DOE

    Energy Technology Data Exchange (ETDEWEB)

    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.

  2. Dose rate in the reactor room and environment during maintenance in fusion reactors

    International Nuclear Information System (INIS)

    According to the International Thermonuclear Experimental Reactor (ITER) conceptual design activity, after reactor shutdown, damaged segments are pulled up from the reactor and hung from the reactor room ceiling by a remote handling device. The dose rate in the reactor room and the environment is estimated for this situation, and the following results are obtained. First, the dose rate in the room is > 108 μSv/h. Since this dose rate is 107 times greater than the biological radiation shielding design limit of 25 μSv/h, workers cannot enter the room. Second, lenses and optical fiber composed of glass that is radiation resistant up to 106 Gy would be damaged after <100 h near the segment, and devices using semiconductors could not work after several hours or so in the aforementioned dose-rate conditions. Third, during suspension of one blanket segment from the ceiling, the dose rate in the site boundary can be reduced by one order by a 23-cm-thicker reactor building roof. To reduce dose rate in public exposure to a value that is less than one-tenth of the public exposure radiation shielding design limit of 100 μSv/yr, the distance of the site boundary from the reactor must be greater than 200 m for a reactor building with a 160-cm-thick concrete roof. 9 refs., 6 figs., 2 tabs

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

    International Nuclear Information System (INIS)

    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

  4. Goals and requirements for advanced reactor concepts

    International Nuclear Information System (INIS)

    Economic problems and public concerns about safety have lead to a reassessment of current nuclear power plant designs and the development of improved designs or new reactor concepts to better meet the needs of United States utilities. This paper presents a set of goals and requirements, developed by the Idaho National Engineering Laboratory (INEL), to provide a means for evaluating the relative merits of alternate advanced reactor concepts. This set of requirements and goals is intended to be independent of any particular reactor concept, and is predicated on the assumption that nuclear power cannot become viable option until the public is favorable to the use of nuclear power for electric power generation in the United States. Under this assumption, the top level requirements defined for new reactor concepts are (1) public acceptability, (2) acceptable investment risk, (3) competitive life cycle costs, and (4) early deployment. Each of these requirements is supported by several related lower level requirements and design goals that are necessary or desirable to meet the top level requirements

  5. Advanced test reactor. Testing capabilities and plans

    International Nuclear Information System (INIS)

    The Advanced Test Reactor (ATR), at the Idaho National Laboratory (INL), is one of the world's premier test reactors for providing the capability for studying the effects of intense neutron and gamma radiation on reactor materials and fuels. The physical configuration of the ATR, a 4-leaf clover shape, allows the reactor to be operated at different power levels in the corner 'lobes' to allow for different testing conditions for multiple simultaneous experiments. The combination of high flux (maximum thermal neutron fluxes of 1E15 neutrons per square centimeter per second and maximum fast [E>1.0 MeV] neutron fluxes of 5E14 neutrons per square centimeter per second) and large test volumes (up to 122 cm long and 12.7 cm diameter) provide unique testing opportunities. For future research, some ATR modifications and enhancements are currently planned. In 2007 the US Department of Energy designated the ATR as a National Scientific User Facility (NSUF) to facilitate greater access to the ATR for material testing research by a broader user community. This paper provides more details on some of the ATR capabilities, key design features, experiments, and plants for the NSUF. (author)

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

    International Nuclear Information System (INIS)

    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

  7. Sandwich reactor lattices and Bloch's theorem

    International Nuclear Information System (INIS)

    The study of the neutron flux distribution in repetitive sandwiches of reactor material leads to results analogous to the 1-dimensional form of Bloch's theorem for the electronic structure in crystals. This principle makes it possible to perform analytically accurate homogenisations of sandwich lattices The method has been extended to cover multi group diffusion and transport theory. (author)

  8. Advanced reactor design and safety objectives - The heavy water reactor perspective

    International Nuclear Information System (INIS)

    This paper provides a summary of the major requirements for future nuclear reactors from CANDU operating station owners based on the various studies and plans prepared. Most of the specific technical requirements for Advanced Heavy Water reactor Systems are based on systematic reviews of current operating CANDU stations to identify opportunities for generic improvements in reliability, operability, maintainability and to address emerging licensing or safety issues. Hence these requirements represent those for the evolutionary development of the Advanced Heavy Water Reactor systems factoring in the considerable operating experience of the CANDU stations. This evolutionary approach to the development of advanced heavy water reactors will be consistent with a philosophy of minimizing the risk to future reactor owners whose requirements are for a reliable, low cost unit

  9. Irradiation fuel for high converter reactor and neutron irradiation method for high converter reactor

    International Nuclear Information System (INIS)

    The present invention provides fuels capable of simulating irradiation of fuels for a high conversion reactor and an irradiation method of fuels for the high conversion reactor. Namely, fuel pins to be irradiated are disposed to the central portion of a fuel assembly. The fuels of the fuel assembly surrounding the central portion comprise driver fuels which generate neutral energy spectra higher than neutron energy spectra generated by other fuel assemblies. With such a constitution, fuels capable of simulating the state of a reactor core of a high conversion reactor can be provided by a portion of a reactor core in a thermal neutron reactor. In addition, the driver fuels comprise plutonium mixed oxide fuels. With such a constitution, the plutonium enrichment degree can be increased at the periphery of the driver fuels. Neutrons of the high neutron energy spectra generated by the driver fuels at high plutonium enrichment degree are irradiated to the fuels to be irradiated. As a result, fuels capable of simulating the state of a reactor core of a high converter reactor can be provided. (I.S.)

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

    International Nuclear Information System (INIS)

    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

  11. Research reactor de-fueling and fuel shipment

    International Nuclear Information System (INIS)

    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

  12. Reactor inspection and maintenance machine senses and homes in on reactor end fittings

    International Nuclear Information System (INIS)

    The Universal Delivery Machine (UDM) is a new CANDU reactor maintenance tool that allows safe, timely, and cost-effective inspection and maintenance of fuel channels. The UDM must align precisely with reactor end-fittings in order to clamp onto fuel channels without applying excessive force. This alignment process is called fine homing. This paper describes the UDM instrumentation and control design features used in the fine homing process. (author)

  13. Papers on reactor physics for operators and unit managers

    International Nuclear Information System (INIS)

    The monograph contains papers submitted to the Dukovany nuclear power plant personnel with the aim of improving professional knowledge of reactor operators and unit managers and helping them in their preparation for state examinations. It presents an easy to understand explanation of phenomena unit control room personnel actually encounter. The following topics are covered: radioactivity, nuclear reactions, nuclear fission and the fate of neutrons in the reactor; delayed neutrons; reactor period, reactivity; subcriticality and transition to criticality; heat generation in the reactor; reactivity coefficients; reactivity effects during the fuel cycle; reactivity compensation during power changes; reactor response to reactivity changes; xenon poisoning, samarium poisoning; residual power; unit start-up after refuelling; unit power rise to the minimal controllable level following emergency shutdown; shutdown concentrations; reactor control and safety system; scram rod drop; neutron sensors in the reactor; monitoring system inside the reactor; 3rd unit computer; ''operator's ten commandments''. (P.A.). 36 figs., 2 tabs., 6 refs

  14. Nuclear Reactor Safeguards and Monitoring with Antineutrino Detectors

    OpenAIRE

    Bernstein, Adam; Wang, Yifang; Gratta, Giorgio; West, Todd

    2001-01-01

    Cubic-meter-sized antineutrino detectors can be used to non-intrusively, robustly and automatically monitor and safeguard a wide variety of nuclear reactor types, including power reactors, research reactors, and plutonium production reactors. Since the antineutrino spectra and relative yields of fissioning isotopes depend on the isotopic composition of the core, changes in composition can be observed without ever directly accessing the core itself. Information from a modest-sized antineutrino...

  15. Basic conceptions for reactor pressure vessel manipulators and their evaluation

    International Nuclear Information System (INIS)

    The study deals with application fields and basic design conceptions of manipulators in reactor pressure vessels as well as their evaluation. It is shown that manipulators supported at the reactor flange have essential advantages

  16. Safety and Behavior of Nuclear Reactors After Shutdown

    International Nuclear Information System (INIS)

    the performance of the nuclear reactors after shutdown is studied. The decay power and xenon poisoning buildup are calculated using a developed mathematical models. Xenon buildup after reactor shutdown is evaluated and discussed from the safety point of view. The calculations are applied for a typical pressurized water reactor (PWR) and for the Egyptian research reactor (ETRR-1). The calculated results for ETRR-1 are compared with the available experimental data

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

    International Nuclear Information System (INIS)

    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

  18. JANUS reactor d and d project

    International Nuclear Information System (INIS)

    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

  19. Advanced CANDU reactor, evolution and innovation

    International Nuclear Information System (INIS)

    Atomic Energy of Canada Limited (AECL) has developed the ACRTM (Advanced CANDU(1) ReactorTM) to meet today's market challenges. It is a light water tube type pressurized water reactor and is the latest evolution of CANDU technology. The design was launched to be cost-competitive with other generating sources, while building on the unique safety and operational advantages of the CANDU design. The ACR is an evolutionary design that retains the proven CANDU features delivered at Qinshan Phase III, while incorporating a set of innovative features and proven state-of-the-art technologies that have emerged from AECL's ongoing Research, Development and Demonstration programs. This approach ensures that key design parameters are well supported by existing reactor experience and R and D. The result is a design that delivers a new threshold in safety, performance and economics while retaining ample design margin. AECL has developed the enabling technologies and components for the ACR design, and has applied them to two plant sizes, ACR-700 and ACR-1000. The ACR integrates hallmark characteristics of traditional CANDU plants (e.g. horizontal pressure tubes, on power fuelling, automated reactor control systems, and dual independent shutdown systems), new innovations (e.g. state-of-the- art control room, extensive use of modular construction techniques, smaller reactor core, enriched uranium fuel), and certain PWR features (e.g. light water coolant, negative void reactivity). The ACR is designed for a high capacity factor and low operation and maintenance costs. It fully exploits the construction techniques that contributed to the impressive schedule accomplishments at Qinshan Phase III and therefore features a very short construction schedule, 40 months construction schedule (First Concrete to Fuel Loading ) for the first unit with improvements to 36 months for later units. The ACR is a true Gen-III plus product with a broad application. It has been proven to be an ideal

  20. Chemistry and technology of Molten Salt Reactors - history and perspectives

    International Nuclear Information System (INIS)

    Molten Salt Reactors represent one of promising future nuclear reactor concept included also in the Generation IV reactors family. This reactor type is distinguished by an extraordinarily close connection between the reactor physics and chemical technology, which is given by the specific features of the chemical form of fuel, representing by molten fluoride salt and circulating through the reactor core and also by the requirements of continuous 'on-line' reprocessing of the spent fuel. The history of Molten Salt Reactors reaches the period of fifties and sixties, when the first experimental Molten Salt Reactors were constructed and tested in ORNL (US). Several molten salt techniques dedicated to fresh molten salt fuel processing and spent fuel reprocessing were studied and developed in those days. Today, after nearly thirty years of discontinuance, a renewed interest in the Molten Salt Reactor technology is observed. Current experimental R and D activities in the area of Molten Salt Reactor technology are realized by a relatively small number of research institutions mainly in the EU, Russia and USA. The main effort is directed primarily to the development of separation processes suitable for the molten salt fuel processing and reprocessing technology. The techniques under development are molten salt/liquid metal extraction processes, electrochemical separation processes from the molten salt media, fused salt volatilization techniques and gas extraction from the molten salt medium

  1. Operation and Utilizations of Dalat Nuclear Research Reactor

    International Nuclear Information System (INIS)

    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

  2. Development and application of reactor noise diagnostics

    International Nuclear Information System (INIS)

    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

  3. 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

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  7. Research reactors for the social safety and prosperous neutron use

    International Nuclear Information System (INIS)

    The present status of nuclear reactors in Japan and the world was briefly described in this report. Aiming to construct a background of stable future society dependent on nuclear energy, the necessity to establish an organization for research reactors in Japan was pointed out. There are a total of 468 reactors in the world, but only 248 of them are running at present and most of them are superannuated. In Japan, 15 research reactors are running and 8 of them are under collaborative utilization, but not a few of them have various problems. In the education of atomic energy, a reactor is dispensable for understanding its working principle through practice learning. Furthermore, a research reactor has important roles for development of power reactor in addition to various basic studies such as activation analysis, fission track, biological irradiation, neutron scattering, etc. Application of a reactor has been also progressing in industrial and medical fields. However, operation of the reactors has become more and more difficult in Japan because of a large running cost and a lack of residential consensus for nuclear reactor. Here, the author proposed an establishment of organization of research reactor in order to promote utilization of a reactor in the field of education, rearing of professionals and science and engineering. (M.N.)

  8. R- AND P- REACTOR BUILDING IN-SITU DECOMISSIONING VISUALIZATION

    Energy Technology Data Exchange (ETDEWEB)

    Bobbitt, J.; Vrettos, N.; Howard, M.

    2010-06-15

    During the early 1950s, five production reactor facilities were built at the Savannah River Site. These facilities were built to produce materials to support the building of the nation's nuclear weapons stockpile in response to the Cold War. R-Reactor and P-Reactor were the first two facilities completed in 1953 and 1954.

  9. HEXNOD and HEXMED, nodal reactor codes for the design of high converter reactors

    International Nuclear Information System (INIS)

    The purpose of the paper is to describe the nodal reactor codes HEXNOD and HEXMED which are used for high converter reactors at SIEMENS/KWU. HEXNOD is a three-dimensional nodal diffusion and transport theory program for reactors with hexagonal-z geometry. The code is based on advanced hexagonal nodal methods for solving global diffusion theory and nodal transport problems for fast reactors and light water high converter reactors (LWHCR). The methods have a high accuracy and efficiency for 3D fast reactor benchmark problems. The present paper gives similar benchmark results for epithermal and thermal reactor systems. The numerical results for HTGR and thermal reactor benchmark problems of the Soviet PWR types WWER-440 and WWER-1000 show that HEXNOD, with one single node per assembly, is as accurate as conventional finite difference computer codes using a mesh of 96 triangles per hexagon. The computing times of HEXNOD are extremely small and allow to perform 3D calculations routinely with reliable results at very small costs. On the basis of HEXNOD the hexagonal nodal reactor burnup program HEXMED has been developed for LWHCR design calculations, allowing 4 to 10 energy groups. HEXMED is a microscopic burnup program. Based on a library of microscopic cross sections the nuclide depletion equations for multi-branched nuclide chains are solved nodewise with the local assembly spectrum. The main operational features of HEXMED are: flexible control of time stepping; refueling operator for the removal, shuffling and loading of fresh fuel assemblies; movement of control rods in bank configuration; criticality search with control banks in discrete axial steps; criticality search with soluble boron; computation of the axial variation of moderator temperatures and densities. (author). 19 refs, 2 figs, 5 tabs

  10. Reactor Coolant Pump seal issues and their applicability to new reactor designs

    International Nuclear Information System (INIS)

    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 US operating plants during the 1970's and early 1980's raised concerns from the US 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

  11. Reactor Coolant Pump seal issues and their applicability to new reactor designs

    Energy Technology Data Exchange (ETDEWEB)

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

    1993-11-01

    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 US operating plants during the 1970`s and early 1980`s raised concerns from the US 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.

  12. Sodium fast reactors (SFRs) and recyclers

    International Nuclear Information System (INIS)

    This presentation is about Sodium Fast Reactor (SFRs) and Recyclers. Their pursuit has been going on in the United States (U.S.) since 1941 and that development work could help support the penetration of SFRs into the current nuclear power market in three forms: 1. A breeding SFR to increase the supply of fissile material. It will not happen for many decades because of increased uranium (U) resources, nuclear market ability to absorb increased U prices, and/or switch to a Thorium (Th) fuel cycle (under development in India) until the anticipated stringent regulations for breeding SFRs are defined and tested. 2. An economic SFR capable of competing with the Advance Light Water Reactor (ALWR) expected to produce electricity in the near future. The Generation IV (Gen IV) program is pursuing that goal under conceptual studies in South Korea (1) and, particularly under the demonstration Japan Sodium Fast Reactor (JSFR) (2) forecasted to start up by 2025 followed by the deployment of commercial JSFRs before 2050. 3. To use the pyro-processing and electro refining methodology developed under the Integral Fast Reactor (IFR) (3) to separate the Light Water Reactor (LWR) spent nuclear fuel (SNF) Transuranics (TRUs) and to burn them in SFRs referred to as Advanced Burner Reactors (ABR). That innovative approach can significantly increase the capacity of geological repositories for disposition of LWR SNF. That last form of SFR is needed urgently to cope with the continued increase in U.S. inventories of recyclable fissile and fertile materials and, particularly, with the projected growth in LWR SNF. According to a recent Electrical Power Research Institute (EPRI) study (4) to reduce CO2 emissions, the U.S. nuclear generated electricity will increase by 64 Gigawatt electrical (GWe) by 2030. While it is realized that additional long term interim storage can alleviate this need, it is not a long term solution because it will have to be followed eventually by final disposal or

  13. 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.

  14. Fast reactor technology innovation and visualization

    International Nuclear Information System (INIS)

    Innovations in safety, operations, and maintenance for improving the availability, reliability, and capital cost of the sodium fast reactor are described. Concerning safety these innovations deal with on-line limiting safety settings, inherent core protection, detection of subassembly coolant mis-allocation. Concerning reactor operations these innovations deal with advanced energy conversion, adapting non-base load nuclear plants and on-line diagnostics. Other innovations concern inspection, servicing, refueling. The development of these innovations rely on visualization technology for their use and for demonstration of improvements achievable. A visualization platform for running these innovations and the nuclear plant thermal-hydraulic, structure, and process codes that underlie them are described. The platform hardware consists of a large-scale tiled display and a haptic hand-controller and in the future will grow to include a high-speed network and multiple graphics-client systems

  15. The Australian nuclear reactor, HIFAR, its past present and future

    International Nuclear Information System (INIS)

    The role of the AAEC's reactor, HIFAR has changed from support for the development of an indigenous power reactor to radioisotope producer and neutron beam source. An account is given of the last twenty-four years' operating experience. A brief review of modernisation programs for reactors of the DIDO class is followed by details of the current HIFAR refurbishing program

  16. Safety in the utilization and modification of research reactors

    International Nuclear Information System (INIS)

    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

  17. Pakistan research reactor and its utilization

    International Nuclear Information System (INIS)

    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)

  18. Small and medium power reactors 1985

    International Nuclear Information System (INIS)

    This report is intended for designers and planners concerned with Small and Medium Power Reactors. It provides a record of the presentations during the meetings held on this subject at the Agency's General Conference in September 1985. This information should be useful as it indicates the principal findings and main conclusions and recommendations resulting from these meetings. A separate abstract was prepared for each of the 10 presentations in this report

  19. Nuclear reactor and associated fuel element

    International Nuclear Information System (INIS)

    Nuclear reactor with a high instantaneous negative reactivity temperature coefficient, comprising a vessel containing a certain quantity of water serving as coolant and moderator, a reactor core immersed in this water and comprising a series of fuel assemblies. Each fuel element contains a solid homogeneous mixture of zirconium hydride, uranium and erbium, in which the uranium constitutes 20 to 50% of the mixture by weight, the zirconium hydride 70 to 50% by weight and the erbium 0.5 to 1.5% by weight, the uranium present in the mixture being not more than 20% of U-235, the remainder being mostly U-238. The ratio of hydrogen/zirconium atom numbers is between 1.5/1 and 1.7/1 and the erbium is evenly distributed in the entire uranium-zirconium hydride mixture

  20. U.S. ALMR sodium cooled reactor design and performance

    International Nuclear Information System (INIS)

    The U.S. Advanced Liquid Metal Reactor (ALMR) has innovative features for reliable operation, accident prevention and severe accident accommodation. The relatively small size of the reactor allows factory fabrication. The reactor shutdown system, inherent reactivity control and the passive decay heat removal systems result in extremely high scram and shutdown reliability. The robust primary boundary and lower internal structures allow accommodation of a hypothetical core disruptive accident and retention of a whole core molten fuel-clad mixture within the reactor. The reactor's low operating temperatures and seismic isolation result in large structural margins and high seismic capabilities. (author)

  1. Titer-plate formatted continuous flow thermal reactors: Design and performance of a nanoliter reactor

    OpenAIRE

    Chen, Pin-Chuan; Park, Daniel S.; You, Byoung-Hee; Kim, Namwon; Park, Taehyun; Steven A Soper; Nikitopoulos, Dimitris E.; Murphy, Michael C.

    2010-01-01

    Arrays of continuous flow thermal reactors were designed, configured, and fabricated in a 96-device (12 × 8) titer-plate format with overall dimensions of 120 mm × 96 mm, with each reactor confined to a 8 mm × 8 mm footprint. To demonstrate the potential, individual 20-cycle (740 nL) and 25-cycle (990 nL) reactors were used to perform the continuous flow polymerase chain reaction (CFPCR) for amplification of DNA fragments of different lengths. Since thermal isolation of the required temperatu...

  2. Dismantling id the reactor pressure vessel insulation and dissecting of the MZFR reactor pressure vessel

    International Nuclear Information System (INIS)

    The MZFR reactor was decommissioned in 1984. The authors describe the dismantling of the reactor pressure vessel insulation that consists of asbestos containing mineral fiber wool. The appropriate remote handling and cutting tools had to be adapted with respect to the restrained space in the containment. The dismantling of the reactor pressure vessel has been completed, the dissected parts have been packaged into 200 containers for the final repository Konrad. During the total project time no reportable events and no damage to persons occurred.

  3. Experience and prospects for developing research reactors of different types

    International Nuclear Information System (INIS)

    NIKIET has a 60-year experience in the development of research reactors. Altogether, there have been more than 25 NIKIET-designed plants of different types built in Russia and 20 more in other countries, including pool-type water-cooled and water moderated research reactors, tank-type and pressure-tube research reactors, pressurized high-flux, heavy-water, pulsed and other research reactors. Most of the research reactors were designed as multipurpose plants for operation at research centers in a broad range of applications. Besides, unique research reactors were developed for specific application fields. Apart from the experience in the development of research reactor designs and the participation in the reactor construction, a unique amount of knowledge has been gained on the operation of research reactors. This makes it possible to use highly reliable technical solutions in the designs of new research reactors to ensure increased safety, greater economic efficiency and maintainability of the reactor systems. A multipurpose pool-type research reactor of a new generation is planned to be built at the Center for Nuclear Energy Science & Technology (CNEST) in the Socialist Republic of Vietnam to be used to support a spectrum of research activities, training of skilled personnel for Vietnam nuclear industry and efficient production of isotopes. It is exactly the applications a research reactor is designed for that defines the reactor type, design and capacity, and the selection of fuel and components subject to all requirements of industry regulations. The design of the new research reactor has a great potential in terms of upgrading and installation of extra experimental devices. (author)

  4. Selection and challenges for LFR reactor materials

    International Nuclear Information System (INIS)

    Nuclear energy using Fast GenIV reactors can fulfil future demands concerning CO2 free, base load capability and sustainability. One of the most promising coolants especially due to its high thermal inertia is liquid lead (Pb). Since several years researches all over the world investigate this coolant and its impact on the reactor design and by that on the materials to be selected. The LEADER project, a follow up of ELSY, aims to design a prototypical demonstrator ALFRED and to continue with several design related aspects of the ELFR reactor. For a demonstrator the criteria of material selection are somewhat different to a commercial type like the ELFR. Material selection for ELFR of course considers all the aspects relevant for ALFRED plus the targeted burn up and the expected total dpa related damage especially of the fuel pins. In the past compatibility of structural material (steels like 316L, T91 and 15-15Ti (1.4970)) that can be employed for Pb cooled fast nuclear reactors were investigated in several EU projects like EUROTRANS and worldwide. Solubility of steel alloying elements like Ni, Fe, Cr is the driving force for the reduced corrosion resistance in contact with Pb. In-situ oxidation is the acknowledged measure to protect steels in Pb up to certain temperatures that are material dependent. Based on experiments and the derived temperature limits the average core outlet temperatures of ALFRED and the ELFR are set to 480 C. The most challenging conditions with respect to temperature are at the fuel assembly and the heat exchangers. For both, thin stable oxide scales with negligible reduction in heat transfer are the requested protection method. This presentation will give an overview on the selected materials for ALFRED and ELFR considering, beside pure compatibility, the influence of mechanical interaction like creep and fretting. (orig.)

  5. Computational mathematics and physics of fusion reactors

    OpenAIRE

    Garabedian, Paul R.

    2003-01-01

    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.

  6. Reactor alarm system development and application issues

    International Nuclear Information System (INIS)

    The new hardware and software technologies, and the need in research reactors for assistance systems in operation and maintenance, have given an appropriate background to develop a computer based system named ''Reactor Alarm System'' (RAS). RAS is a software package, user oriented, with emphasis on production, experiments and maintenance goals. It is designed to run on distributed systems conformed with microcomputers under QNX operating system. RAS main features are: a) Alarm Panel Display; b) Alarm Page; c) Alarm Masking and Inhibition; d) Alarms Color and Attributes; e) Condition Classification; and f) Arrangement Presentation. RAS design allows it to be installed as a part of a computer based Supervision and Control System in new installations or retrofit existing reactor instrumentation systems. The analysis of human factors during development stage and successive user feedback from different applications, brought out several RAS improvements: a) Multiple-copy alarm summaries; b) Improved alarm handling; c) Extended dictionary; and d) Enhanced hardware availability. It has proved successful in providing new capabilities for operators, and also has shown the continuous increase of user-demands, reflecting the expectations placed today on computer-based systems. (author). 6 figs, 1 tabs

  7. Experimental neutronic science and instrumentation: from hybrid reactors to fourth generation reactors

    International Nuclear Information System (INIS)

    After an overview of his academic career and scientific and research activities, the author proposes a rather detailed synthesis and overview of his scientific activities in the fields of cross sections and Doppler effect (development and validation of a code), on the MUSE-4 hybrid reactor (experiments, static and dynamic measurements), on the TRADE hybrid reactor (experimental means, sub-critical reactivity measurement), on the RACE hybrid reactor (experimental results, modelling and interpretation), and on neutron detection (design and modelling of fission chamber, on-line measurement of the fast flow). The next part gives an overview of some research programs (neutron monitoring in sodium-cool fast reactors, research and development on fission chambers, improvement of effective delayed neutron measurements)

  8. Safe operation and maintenance of research reactor

    International Nuclear Information System (INIS)

    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 U3O8- 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)

  9. 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)

  10. Novelties in design and construction of the advanced reactors

    International Nuclear Information System (INIS)

    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

  11. A review of calculation methods for fast and intermediate reactors

    International Nuclear Information System (INIS)

    This paper discusses the development of methods for calculating intermediate and fast reactors. It deals with various approaches to the problems of physical calculation. The calculation of resonance effects is discussed. Consideration is given to multi-group systems of fundamental and conjugate equations, various applications of perturbation theory to the problems of physical reactor calculation, and numerical methods of solving fundamental and conjugate reactor equations, which approximate the method of spherical harmonics. The paper describes an application of the response method to the solution of critical-mass problems, and methods of calculating reactors with hydrogeneous moderators. The fundamental features of an effective one-group reactor model are described. (author)

  12. International Conference on Physics and Technology of Reactors and Applications

    International Nuclear Information System (INIS)

    Full text : The international conference on physics and technology of reactors is organized by the Moroccan Association for Nuclear enggineering and Reactor Technology (GMTR) with the collaboration of the Centre for Energy and Nuclear Sciences and Techniques (CNESTEN) and under the auspices of the ministry of Energy, Mining, Water and Environment. The programme of the PHYTRA2 conference covers a wide variety of topics. The conference is organised in one plenary session, eight oral technical sessions and one poster session. The oral and poster technical sessions covers the usual topics of nuclear engineering including one session on research reactors utilisation and computational methods for research reactors

  13. Safety analysis of RA reactor operation, I-II, Part I - RA reactor technical and operation characteristics

    International Nuclear Information System (INIS)

    RA research reactor is a thermal, heavy water moderated system with graphite reflector having nominal power 6.5 MW. The 2% enriched metal uranium fuel in the reactor core produces mean thermal neutron flux of 2.9 1013 neutrons/cm2 s, and maximum neutron flux 5.5 1013 neutrons/cm2 s. main components of the reactor described in this report are: rector core, reflector, biological shield, heavy water cooling system, ordinary water cooling system, helium system, reactor control system, reactor safety system, dosimetry system, power supply system, and fuel transport system. Detailed reactor properties and engineering drawings of all the system are part of this volume

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

    International Nuclear Information System (INIS)

    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 m3 considerably smaller compared to other BWR buildings, thus providing a huge potential for cost savings. The turbine building with app 250,000 m3 is of approximately the same size like for existing BWRs. (orig.)

  15. Nuclear reactor dismantling method and device

    International Nuclear Information System (INIS)

    The reactor dismantling device according to the present invention comprises an elevator lift extending from the lower portion in a biological-shielding walls of the reactor to an operation floor thereabove and a scaffolding for cutting operation vertically disposed at the periphery thereof. The scaffolding is rotated around a cutting mast by remote control to displace a cutting position for a drilling device. Then, pieces of the biological-shielding walls are cut out and automatically transported from the inside of the biological-shielding walls to an operation chamber by a recovery device, a truck and an elevator lift. This makes the dismantling operation highly efficient, to shorten the term of works. Further, operators' exposure dose can be mitigated, thereby enabling to improve safety of the dismantling operation. (T.M.)

  16. Shielding design for research and education reactor

    International Nuclear Information System (INIS)

    For the purpose of education and research at the University, 20-KW powered SLOWPOKE-2 research reactor has been chosen as a prototype reactor. In order to study the safety characteristics of the reactor, exposure rate has been estimated at the pool boundary. Reactor core as a radiation source is assumed to be cylindrical volume source. Thus point kernel integration method can be applied to determine the exposure rate. For the sake of simplicity, calculation was done only for the prompt fission gamma rays and fission product gamma rays. As a result, the maximum exposure rate at the pool boundary was estimated to be 18R/min at the same height of the center of the core. In order to examine the accuracy for the point kernel integration method, two shielding experiments were carried out: one for the water tank only and the other for with concrete blocks outside the water tank. Water tank was made of wood pieces which is 13.4cm wide, 1.5cm thick and 2.15m long. Thus the water tank has the total dimension of 1 m radius and 2.1 m height. The experiment was carried out for the radiation source of 0.968 mCi Co-60 at the center of the water tank and the penetrated gamma rays were measured at 5 different detector positions. For the measurement and analysis of the responses, NaI(T1) 3''x3'' detector and 256 channel multichannel analyzer was utilized. To convert pulse height distribution to the exposure rate, Moriuchi conversion factor was adopted. Data from the calculations by point kernel method were well agreed within 10% band with the data from the the experiments. (Author)

  17. Present and possible utilization of PUSPATI reactor

    International Nuclear Information System (INIS)

    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)

  18. Perspective of nuclear energy and advanced reactors

    International Nuclear Information System (INIS)

    Future nuclear energy growth will be the result of the contributions of every single plant being constructed or projected at present as it is connected to the grid. As per IAEA, there exists presently 34 nuclear power plants under construction 81 with the necessary permits and funding and 223 proposed, which are plants seriously pursuing permits and financing. This means that in a few decades the number of nuclear power plants in operation will have doubled. This growth rate is characterised by the incorporation of new countries to the nuclear club and the gradually increasing importance of Asian countries. During this expansive phase, generation III and III+designs are or will be used. These designs incorporate the experience from operating plants, and introduce innovations on rationalization design efficiency and safety, with emphasis on passive safety features. In a posterior phase, generation IV designs, presently under development, will be employed. Generation IV consists of several types of reactors (fast reactors, very high temperature reactors, etc), which will improve further sustain ability, economy, safety and reliability concepts. The described situation seems to lead to a renaissance of the nuclear energy to levels hardly thinkable several years ago. (Author)

  19. Design and manufacture for the reactor internals

    International Nuclear Information System (INIS)

    The reactor internals are main equipment in the reactor of NPP. In order to ensure the safe and reliable operation of NPP, a lot of test researches and calculation analyses on the internal structure design, stress analyses, thermal hydraulics as well as technological trial-manufacture etc. have been performed, so as to make it perfect and to improve it constantly. Among them special structures such as alignment orientation, reactor driving line and high pressure high temperature seal etc. are designed. A complete working diagram and methods for design by analysis are developed. In the researches, intersect disciplines on flow-solid coupling, irradiation-heat-stress coupling, shock-vibration-elastic-plastic dynamics, similar theory of model test etc. are concerned with, and a series of technological difficulties in manufacture such as thin plate machining, precision boring, precision welding and driving line aligning etc. are overcome. These experiences can be used as reference for the design and manufacture of new NPP. (17 refs., 5 figs., 1 tab.)

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

    International Nuclear Information System (INIS)

    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).

  1. Regulation concerning installation and operation of reactors for power generation

    International Nuclear Information System (INIS)

    The regulations applying to reactors for power generation mentioned in the law for the regulations of nuclear source materials, nuclear fuel materials and reactors. Covered are the following: definitions of terms, application for the permission to install reactors, application for the permission to alter installations, reactor operation plans, keeping of various records, limitations on ingress and ingress in radiation controlled areas, measures concerning radiation exposure doses, operation of reactors, on-site transport, storage of nuclear fuel materials and radioactive waste, security regulations, steps taken during times of danger, making of various reports, and so on. (Mori, K.)

  2. Development of Education and Training Programs Using ISIS Research Reactor

    International Nuclear Information System (INIS)

    As a part of the French Alternative Energies and Atomic Energy Commission (CEA), the National Institute for Nuclear Science and Technology (INSTN) carries out various education and training programs on nuclear reactor theory and operation. These programs take advantage of the use of an extensive range of training tools that includes software applications, simulators, as well as the use of research reactors. After a presentation of ISIS reactor, we present the training courses that have been developed on ISIS reactor and their use in education and training programs developed by INSTN. We report on how the training courses carried out on ISIS research reactor ensure a practical and comprehensive understanding of the reactor principle and operation, bringing tremendous benefit to the trainees. We also discuss the future development of education and training programs using the ISIS research reactor as a very powerful tool for the development of the human resources needed by the nuclear industry and the nuclear programs. (author)

  3. Status of Fast Reactor Research and Technology Development

    International Nuclear Information System (INIS)

    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.

  4. Fabrication, testing, and qualification of reactor graphites

    International Nuclear Information System (INIS)

    The work performed under the HBK project for development and testing of reactor graphites could have recourse to results and experience already gained in Great Britain, in the F.R.G., the USA, and the Netherlands. The specific problems to be tackled by the HBK project activities result from the particularly exacting requirements with regard to behaviour under irradiation that are to be met by the graphite reflector for the THTR follower plant. From a great number of candidate graphites, selected for testing and evaluation, the extensive irradiation experiments revealed a variety of graphites best suited to the various tasks in mind, as defined by the operational conditions. The tests examined radiation-induced changes of linear dimension, E-module, thermal expansion, and heat conductivity, as well as radiation-induced creep and corrosion in reactor graphites under specified normal and under accident conditions. The work performed also includes tests for defining design criteria for reactor graphite components. The goals have been achieved, but further work will be necessary, as new requirements are taking shape in the course of current THTR follower plant development. (orig.)

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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)

  7. 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 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...... of the role of the anaerobic digestion in the wastewater treatment plants from a pre-treatment method to the main biological treatment method. The application of staged high-rate anaerobic digesters has shown the larger potential among the recent developments in this direction. The most common high...

  8. Optical inspections of research reactor tanks and tank components

    International Nuclear Information System (INIS)

    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)

  9. Reactor technology comparison method using modeling and decision making technique

    International Nuclear Information System (INIS)

    To make wise choices about the future of nuclear power, we need improved knowledge of the safety, safeguards, and security features of both existing and new nuclear energy plants. Understanding the potential advantages and disadvantages of nuclear energy is critical for those stake holders and decision-makers facing Korea's energy challenges. This report provides an overview of the evolution of nuclear reactor technology and discusses development and deployment of reactor technology in future Korea. Following reactor technology options will be considered within this report: 1) Active Reactor (GEN Π : OPR 1000) 2) Advanced Reactor (GEN ΠI : APR 1400) 3) Passive Reactor (GEN ΠI+ : AP 1000 ) 4) Small Modular Reactor (GEN ΙV )

  10. DCS Terrain for Wilcox County GA MapMod08

    Data.gov (United States)

    Federal Emergency Management Agency, Department of Homeland Security — Terrain data, as defined in FEMA Guidelines and Specifications, Appendix M: Data Capture Standards, describes the digital topographic data that was used to create...

  11. 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...

  12. DIGITAL FLOOD INSURANCE RATE MAP DATABASE, WILCOX COUNTY, GEORGIA, USA

    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;...

  13. 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...

  14. Optimized Design and Discussion on Middle and Large CANDLE Reactors

    OpenAIRE

    Xiaoming Chai; Yong Zhang; Mingyu Yan

    2012-01-01

    CANDLE (Constant Axial shape of Neutron flux, nuclide number densities and power shape During Life of Energy producing reactor) reactors have been intensively researched in the last decades [1–6]. Research shows that this kind of reactor is highly economical, safe and efficiently saves resources, thus extending large scal...

  15. Tritium chemistry in fission and fusion reactors

    International Nuclear Information System (INIS)

    We are interested in the behaviour of tritium inside the solids where it is generated both in the case of fission nuclear reactor fuel elements, and in that of blankets of future fusion reactor. In the first case it is desirable to be able to predict whether tritium will be found in the hulls or in the uranium oxide, and under what chemical form, in order to take appropriate steps for it's removal in reprocessing plants. In fusion reactors breeding large amounts of tritium and burning it in the plasma should be accomplished in as short a cycle as possible in order to limit inventories that are associated with huge activities. Mastering the chemistry of every step is therefore essential. Amounts generated are not of the same order of magnitude in the two cases studied. Ternary fissions produce about 66 1013Bq (18 000 Ci) per year of tritium in a 1000 MWe fission generator, i.e., about 1.8 1010Bq (0.5 Ci) per day per ton of fuel

  16. Burn up calculations for the Iranian miniature reactor: A reliable and safe research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Faghihi, F. [Department of Nuclear Engineering, School of Engineering, Shiraz University, Shiraz 71345 (Iran, Islamic Republic of); Research Center for Radiation Protection, Shiraz University, Shiraz (Iran, Islamic Republic of)], E-mail: faghihif@shirazu.ac.ir; Mirvakili, S.M. [Department of Nuclear Engineering, School of Engineering, Shiraz University, Shiraz 71345 (Iran, Islamic Republic of)

    2009-06-15

    Presenting neutronic calculations pertaining to the Iranian miniature research reactor is the main goal of this article. This is a key to maintaining safe and reliable core operation. The following reactor core neutronic parameters were calculated: clean cold core excess reactivity ({rho}{sub ex}), control rod and shim worth, shut down margin (SDM), neutron flux distribution of the reactor core components, and reactivity feedback coefficients. Calculations for the fuel burnup and radionuclide inventory of the Iranian miniature neutron source reactor (MNSR), after 13 years of operational time, are carried out. Moreover, the amount of uranium burnup and produced plutonium, the concentrations and activities of the most important fission products, the actinide radionuclides accumulated, and the total radioactivity of the core are estimated. Flux distribution for both water and fuel temperature increases are calculated and changes of the central control rod position are investigated as well. Standard neutronic simulation codes WIMS-D4 and CITATION are employed for these studies. The input model was validated by the experimental data according to the final safety analysis report (FSAR) of the reactor. The total activity of the MNSR core is calculated including all radionuclides at the end of the core life and it is found to be equal to 1.3 x 10{sup 3}Ci. Our investigation shows that the reactor is operating under safe and reliable conditions.

  17. Burn up calculations for the Iranian miniature reactor: A reliable and safe research reactor

    International Nuclear Information System (INIS)

    Presenting neutronic calculations pertaining to the Iranian miniature research reactor is the main goal of this article. This is a key to maintaining safe and reliable core operation. The following reactor core neutronic parameters were calculated: clean cold core excess reactivity (ρex), control rod and shim worth, shut down margin (SDM), neutron flux distribution of the reactor core components, and reactivity feedback coefficients. Calculations for the fuel burnup and radionuclide inventory of the Iranian miniature neutron source reactor (MNSR), after 13 years of operational time, are carried out. Moreover, the amount of uranium burnup and produced plutonium, the concentrations and activities of the most important fission products, the actinide radionuclides accumulated, and the total radioactivity of the core are estimated. Flux distribution for both water and fuel temperature increases are calculated and changes of the central control rod position are investigated as well. Standard neutronic simulation codes WIMS-D4 and CITATION are employed for these studies. The input model was validated by the experimental data according to the final safety analysis report (FSAR) of the reactor. The total activity of the MNSR core is calculated including all radionuclides at the end of the core life and it is found to be equal to 1.3 x 103Ci. Our investigation shows that the reactor is operating under safe and reliable conditions.

  18. RA reactor operation and maintenance in 2000, Part 1

    International Nuclear Information System (INIS)

    Activities related to revitalisation of the RA reactor started in 1986 were fulfilled except the exchange of the complete reactor instrumentation. Since 1992, due to economic and political reasons, RA reactor is in a difficult situation. The old RA reactor instrumentation was dismantled. Decision about the future status of the reactor should be made because the aging of all the components is becoming dramatic. Control and maintenance of the reactor components was done regularly and efficiently. The most important activity and investment in 1998 was improvement of conditions for spent fuel storage in the existing pools at the RA reactor. Russian company ENTEK and IAEA are involved in this activity which was initiated 1997. Fuel inspection by the IAEA safeguards inspectors was done on a monthly basis

  19. 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.

  20. INVAP Experience in the Design and Construction of Research Reactors. (Research Reactors in and from Argentina)

    International Nuclear Information System (INIS)

    Full Text: Argentina has a long tradition in the design and construction of Research Reactors. The first research reactor in Argentina, RA-1, was built by CNEA (Argentina Atomic Energy Commission) in 1958, using drawings lent by USA. RA-2, RA-3, RA-4 and RA-0 followed through. In 1976, a career degree in Nuclear Engineering was started by CNEA and the University of Cuyo in Bariloche. It was decided that there would be a university type reactor to assist with the training of the students. INVAP, a recently created company, was assigned the task of building the reactor in accordance with the engineering developed by CNEA. The RA-6 was a very successful project, which allowed INVAP to build the knowledge for participating in RR projects abroad. Since 1982, INVAP has built research reactors in Algeria, Egypt, Argentina and Australia and had a large participation in the RRs CNEA built in Peru. INVAP has also designed several other RR for different clients, which were not subsequently built. This paper explores this history, giving details of the RR projects in which INVAP has been involved through the years. (author)