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  1. 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

  2. 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

  3. 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

  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. 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

  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. 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

  8. 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

  9. 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

  10. 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)

  11. 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)

  12. 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

  13. 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

  14. 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

  15. 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.)

  16. 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

  17. 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

  18. 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

  19. 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)

  20. 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...

  1. 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...

  2. 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

  3. 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.)

  4. 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)

  5. 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)

  6. 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)

  7. 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

  8. 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)

  9. 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.

  10. 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

  11. 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.

  12. 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

  13. 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

  14. 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)

  15. 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

  16. 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

  17. 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)

  18. 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

  19. 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

  20. 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

  1. 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

  2. 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

  3. 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

  4. 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...

  5. 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

  6. 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

  7. 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

  8. 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

  9. 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

  10. 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)

  11. 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

  12. 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

  13. 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)

  14. 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

  15. 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

  16. 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

  17. 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

  18. 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...

  19. 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....

  20. 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...

  1. 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

  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. 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

  6. 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

  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. 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.)

  18. 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

  19. 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

  20. 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

  1. 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...

  2. American Nuclear Society standards for TRIGA reactors and their use

    International Nuclear Information System (INIS)

    The American Nuclear Society established a committee (ANS-15) with the expressed charter to develop standards for research reactors. These standards were to cover all aspects of research reactor operations, maintenance and administration. Standards have been written in every area of research reactor operations that the research reactor community has deemed important. One of the uppermost goals of the Standards Committee work is to produce standards that provide guidance and help to the research reactor community in a timely manner. To make the standards meaningful requires a great deal of cooperation between all segments of the reactor community. The research reactors - whether they are private, university or government owned - have a mission to perform. At the same time, the regulatory agencies also have a mission to perform, and with a spirit of mutual respect and cooperation, both can accomplish their goals. In the last five years this spirit has been present, and a number of very good standards have resulted. These standards should be a part of every research reactor library. In particular ANS-15.16 and ANS-15.1 have been endorsed by the regulatory agencies and are being used to evaluate submittals

  3. 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)

  4. 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

  5. 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

  6. 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

  7. 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

  8. ASTM Standards for Reactor Dosimetry and Pressure Vessel Surveillance

    International Nuclear Information System (INIS)

    The ASTM standards provide guidance and instruction on how to field and interpret reactor dosimetry. They provide a roadmap towards understanding the current ''state-of-the-art'' in reactor dosimetry, as reflected by the technical community. The consensus basis to the ASTM standards assures the user of an unbiased presentation of technical procedures and interpretations of the measurements. Some insight into the types of standards and the way in which they are organized can assist one in using them in an expeditious manner. Two example are presented to help orient new users to the breadth and interrelationship between the ASTM nuclear metrology standards. One example involves the testing of a new ''widget'' to verify the radiation hardness. The second example involves quantifying the radiation damage at a pressure vessel critical weld location through surveillance dosimetry and calculation

  9. 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.)

  10. 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)

  11. 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.

  12. 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...

  13. Neutron standard cross sections in reactor physics - Need and status

    International Nuclear Information System (INIS)

    The design and improvement of nuclear reactors require detailed neutronics calculations. These calculations depend on comprehensive libraries of evaluated nuclear cross sections. Most of the cross sections that form the data base for these evaluations have been measured relative to neutron cross-section standards. The use of these standards can often simplify the measurement process by eliminating the need for a direct measurement of the neutron fluence. The standards are not known perfectly, however; thus the accuracy of a cross-section measurement is limited by the uncertainty in the standard cross section relative to which it is measured. Improvements in a standard cause all cross sections measured relative to that standard to be improved. This is the reason for the emphasis on improving the neutron cross-section standards. The continual process of measurement and evaluation has led to improvements in the accuracy and range of applicability of the standards. Though these improvements have been substantial, this process must continue in order to obtain the high-quality standards needed by the user community

  14. 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

  15. Monitoring system for accuracy and reliability characteristics of standard temperature measurements in WWER-440 reactors

    International Nuclear Information System (INIS)

    Description of a monitoring system for accuracy and reliability characteristics of standard temperature measurements in WWER-440 reactors and benefits obtained from its use are shown in the presentation. As standard reactor temperature measurement, coolant temperature measurement at fuel assembly outlets and in loops, entered into the In-Reactor Control System , are considered. Such systems have been implemented at two V-230 reactors and are under implementation at other four V-213 reactors. (Authors)

  16. 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.

  17. 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,…

  18. SRAC: JAERI thermal reactor standard code system for reactor design and analysis

    International Nuclear Information System (INIS)

    The SRAC (Standard Reactor Analysis Code) is a code system for nuclear reactor analysis and design. It is composed of neutron cross section libraries and auxiliary processing codes, neutron spectrum routines, a variety of transport, 1-, 2- and 3-D diffusion routines, dynamic parameters and cell burn-up routines. By making the best use of the individual code function in the SRAC system, the user can select either the exact method for an accurate estimate of reactor characteristics or the economical method aiming at a shorter computer time, depending on the purpose of study. The user can select cell or core calculation; fixed source or eigenvalue problem; transport (collision probability or Sn) theory or diffusion theory. Moreover, smearing and collapsing of macroscopic cross sections are separately done by the user's selection. And a special attention is paid for double heterogeneity. Various techniques are employed to access the data storage and to optimize the internal data transfer. Benchmark calculations using the SRAC system have been made extensively for the Keff values of various types of critical assemblies (light water, heavy water and graphite moderated systems, and fast reactor systems). The calculated results show good prediction for the experimental Keff values. (author)

  19. 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...

  20. 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

  1. Standards for reactor accident cases

    International Nuclear Information System (INIS)

    The Committee on Standards for reactor accident cases in the Netherlands published its recommendations to the Minister of Health. Maximum permissible quantities of radiation and radionuclide intake are presented for adults and children as well as pregnant women. Exposure limit standards for the whole body as well as specific organs and other parts are given. Also considered is the contamination of cattle and cows' milk. The standards are compared with those of the ICRP and the English Medical Research Council

  2. 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

  3. 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

  4. 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

  5. Standard interface files and procedures for reactor physics codes. Version IV

    International Nuclear Information System (INIS)

    Standards, procedures, and recommendations of the Committee on Computer Code Coordination for promoting the exchange of reactor physics codes are updated to Version IV status. Standards and procedures covering general programming, program structure, standard interface files, and file management and handling subroutines are included

  6. IAEA Workshop (Training Course) on Codes and Standards for Sodium Cooled Fast Reactors. Working Material

    International Nuclear Information System (INIS)

    The training course consisted of lectures and Q&A sessions. The lectures dealt with the history of the development of Design Codes and Standards for Sodium Cooled Fast Reactors (SFRs) in the respective country, the detailed description of the current design Codes and Standards for SFRs and their application to ongoing Fast Reactor design projects, as well as the ongoing development work and plans for the future in this area. Annex 1 contains the detailed Workshop program

  7. Choosing a standard reactor: International competition and domestic politics in Chinese nuclear policy

    International Nuclear Information System (INIS)

    China has ambitious plans to expand its nuclear power capacity. One of the policy goals that high-level policymakers have desired is to base the nuclear program on a standardized reactor design. However, this has not materialized so far. By examining its nuclear reactor choices for individual projects, we argue that China’s policymaking process has been greatly influenced by international competition and domestic politics. Multiple international nuclear vendors are intent upon maintaining their respective niches in the expanding Chinese reactor market, and they have used various forms of economic and political pressure to achieve their objectives. On the other hand, China’s policymaking process is fragmented and the shifting power balances among powerful domestic actors do not allow a fixed path to be followed. Further, because of the high costs and potential profits involved, nuclear reactor choices in China have been driven not just by technical considerations but also by foreign and trade policy objectives. All of these make it unlikely that China will standardize the reactor type it constructs in the near future. -- Highlights: ► China’s nuclear power policymaking has been fragmented and without central control. ► Multiple domestic actors have pursued independent agendas. ► International nuclear vendors have intensely competed for Chinese reactor contracts. ► Economic, political and foreign policy goals have driven reactor contract decisions. ► China is unlikely to construct only a standardized reactor design.

  8. Conversion and standardization of university reactor fuels using low-enrichment uranium: Plans and schedules

    Energy Technology Data Exchange (ETDEWEB)

    Young, H.H.; Brown, K.R.; Matos, J.E.

    1986-01-01

    The highly-enriched uranium (HEU) fuel used in twenty United States university reactors can be viewed as contributing to the risk of theft or diversion of weapons-useable material. To minimize this risk, the US Nuclear Regulatory Commission issued its final rule on ''Limiting the Use of Highly Enriched Uranium in Domestically Licensed Research and Test Reactors,'' in February 1986. This paper describes the plans and schedules developed by the US Department of Energy to coordinate an orderly transition from HEU to LEU fuel in most of these reactors. An important element in the planning process has been the desire to standardize the LEU fuels used in US university reactors and to enhance the performance and utilization of a number of these reactors. The program is estimated to cost about $10 million and to last about five years.

  9. 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

  10. New revisions of reactor physics standards

    International Nuclear Information System (INIS)

    This paper presents an overview of two of the most basic reactor physics standards, the newly revised “Steady State Neutronics Methods for Power Reactor Analysis” and its companion standard, “Nuclear Data for Reactor Design”. These two popular state-of-the-art standards provide important guidance for developing the necessary input data needed to calculate reactor lattice and core parameters such as reaction rates spatial distributions, reactivity and flux distributions in power reactors for all currently used reactor types, from fast to thermal reactors. The standards provide guidance for the selection of cross section data and libraries, the development of nuclear data sets suitable for specific applications, energy group structures and group collapsing. Key elements in the complex sequence of power reactor calculations are outlined. The effects of simplifications and approximations made in the treatment of the neutronic and geometric models and the biases and uncertainties resulting from such simplifications and assumptions are discussed. In the newly-revised standard on neutronics methods, clear distinction is made between the two important tools for assessing the reliability of the results of the calculations, verification and validation. To provide an auditable path in the verification and validation processes, the standard requires detailed documentation including methods used, selection of calculation models and experimental data and results of higher order calculations. (author)

  11. 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.

  12. 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.

  13. 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…

  14. Assessement of Codes and Standards Applicable to a Hydrogen Production Plant Coupled to a Nuclear Reactor

    Energy Technology Data Exchange (ETDEWEB)

    M. J. Russell

    2006-06-01

    This is an assessment of codes and standards applicable to a hydrogen production plant to be coupled to a nuclear reactor. The result of the assessment is a list of codes and standards that are expected to be applicable to the plant during its design and construction.

  15. 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...

  16. 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…

  17. High Temperature Gas Reactors: Assessment of Applicable Codes and Standards

    Energy Technology Data Exchange (ETDEWEB)

    McDowell, Bruce K.; Nickolaus, James R.; Mitchell, Mark R.; Swearingen, Gary L.; Pugh, Ray

    2011-10-31

    Current interest expressed by industry in HTGR plants, particularly modular plants with power up to about 600 MW(e) per unit, has prompted NRC to task PNNL with assessing the currently available literature related to codes and standards applicable to HTGR plants, the operating history of past and present HTGR plants, and with evaluating the proposed designs of RPV and associated piping for future plants. Considering these topics in the order they are arranged in the text, first the operational histories of five shut-down and two currently operating HTGR plants are reviewed, leading the authors to conclude that while small, simple prototype HTGR plants operated reliably, some of the larger plants, particularly Fort St. Vrain, had poor availability. Safety and radiological performance of these plants has been considerably better than LWR plants. Petroleum processing plants provide some applicable experience with materials similar to those proposed for HTGR piping and vessels. At least one currently operating plant - HTR-10 - has performed and documented a leak before break analysis that appears to be applicable to proposed future US HTGR designs. Current codes and standards cover some HTGR materials, but not all materials are covered to the high temperatures envisioned for HTGR use. Codes and standards, particularly ASME Codes, are under development for proposed future US HTGR designs. A 'roadmap' document has been prepared for ASME Code development; a new subsection to section III of the ASME Code, ASME BPVC III-5, is scheduled to be published in October 2011. The question of terminology for the cross-duct structure between the RPV and power conversion vessel is discussed, considering the differences in regulatory requirements that apply depending on whether this structure is designated as a 'vessel' or as a 'pipe'. We conclude that designing this component as a 'pipe' is the more appropriate choice, but that the ASME BPVC

  18. Compiled reports on the applicability of selected codes and standards to advanced reactors

    International Nuclear Information System (INIS)

    The following papers were prepared for the Office of Nuclear Regulatory Research of the U.S. Nuclear Regulatory Commission under contract DE-AC06-76RLO-1830 NRC FIN L2207. This project, Applicability of Codes and Standards to Advance Reactors, reviewed selected mechanical and electrical codes and standards to determine their applicability to the construction, qualification, and testing of advanced reactors and to develop recommendations as to where it might be useful and practical to revise them to suit the (design certification) needs of the NRC

  19. Standard review plan for the review and evaluation of emergency plans for research and test reactors

    International Nuclear Information System (INIS)

    This document provides a Standard Review Plan to assure that complete and uniform reviews are made of research and test reactor radiological emergency plans. The report is organized under ten planning standards which correspond to the guidance criteria in American National Standard ANSI/ANS 15.16 - 1982 as endorsed by Revision 1 to Regulatory Guide 2.6. The applicability of the items under each planning standard is indicated by subdivisions of the steady-state thermal power levels at which the reactors are licensed to operate. Standard emergency classes and example action levels for research and test reactors which should initiate these classes are given in an Appendix. The content of the emergency plan is as follows: the emergency plan addresses the necessary provisions for coping with radiological emergencies. Activation of the emergency plan is in response to the emergency action levels. In addition to addressing those severe emergencies that will fall within one of the standard emergency classes, the plan also discusses the necessary provisions to deal with radiological emergencies of lesser severity that can occur within the operations boundary. The emergency plan allows for emergency personnel to deviate from actions described in the plan for unusual or unanticipated conditions

  20. Assessment of United States industry structural codes and standards for application to advanced nuclear power reactors: Appendices. Volume 2

    International Nuclear Information System (INIS)

    Throughout its history, the USNRC has remained committed to the use of industry consensus standards for the design, construction, and licensing of commercial nuclear power facilities. The existing industry standards are based on the current class of light water reactors and as such may not adequately address design and construction features of the next generation of Advanced Light Water Reactors and other types of Advanced Reactors. As part of their on-going commitment to industry standards, the USNRC commissioned this study to evaluate US industry structural standards for application to Advanced Light Water Reactors and Advanced Reactors. The initial review effort included (1) the review and study of the relevant reactor design basis documentation for eight Advanced Light Water Reactors and Advanced Reactor Designs, (2) the review of the USNRCs design requirements for advanced reactors, (3) the review of the latest revisions of the relevant industry consensus structural standards, and (4) the identification of the need for changes to these standards. The results of these studies were used to develop recommended changes to industry consensus structural standards which will be used in the construction of Advanced Light Water Reactors and Advanced Reactors. Over seventy sets of proposed standard changes were recommended and the need for the development of four new structural standards was identified. In addition to the recommended standard changes, several other sets of information and data were extracted for use by USNRC in other on-going programs. This information included (1) detailed observations on the response of structures and distribution system supports to the recent Northridge, California (1994) and Kobe, Japan (1995) earthquakes, (2) comparison of versions of certain standards cited in the standard review plan to the most current versions, and (3) comparison of the seismic and wind design basis for all the subject reactor designs

  1. Assessment of United States industry structural codes and standards for application to advanced nuclear power reactors: Final report. Volume 1

    International Nuclear Information System (INIS)

    Throughout its history, the USNRC has remained committed to the use of industry consensus standards for the design, construction, and licensing of commercial nuclear power facilities. The existing industry standards are based on the current class of light water reactors and as such may not adequately address design and construction features of the next generation of Advanced Light Water Reactors and other types of Advanced Reactors. As part of their on-going commitment to industry standards, the USNRC commissioned this study to evaluate US industry structural standards for application to Advanced Light Water Reactors and Advanced Reactors. The initial review effort included: (1) the review and study of the relevant reactor design basis documentation for eight Advanced Light Water Reactors and Advanced Reactor Designs, (2) the review of the USNRCs design requirements for advanced reactors, (3) the review of the latest revisions of the relevant industry consensus structural standards, and (4) the identification of the need for changes to these standards. The results of these studies were used to develop recommended changes to industry consensus structural standards which will be used in the construction of Advanced Light Water Reactors and Advanced Reactors. Over seventy sets of proposed standard changes were recommended and the need for the development of four new structural standards was identified. In addition to the recommended standard changes, several other sets of information and data were extracted for use by USNRC in other ongoing programs. This information included: (1) detailed observations on the response of structures and distribution system supports to the recent Northridge, California (1994) and Kobe, Japan (1995) earthquakes, (2) comparison of versions of certain standards cited in the standard review plan to the most current versions, and (3) comparison of the seismic and wind design basis for all the subject reactor designs

  2. 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...

  3. Nuclear reactors' construction costs: The role of lead-time, standardization and technological progress

    International Nuclear Information System (INIS)

    This paper provides the first comparative analysis of nuclear reactor construction costs in France and the United States. Studying the cost of nuclear power has often been a challenge, owing to the lack of reliable data sources and heterogeneity between countries, as well as the long time horizon which requires controlling for input prices and structural changes. We build a simultaneous system of equations for overnight costs and construction time (lead-time) to control for endogeneity, using expected demand variation as an instrument. We argue that benefits from nuclear reactor program standardization can arise through short term coordination gains, when the diversity of nuclear reactors' technologies under construction is low, or through long term benefits from learning spillovers from past reactor construction experience, if those spillovers are limited to similar reactors. We find that overnight construction costs benefit directly from learning spillovers but that these spillovers are only significant for nuclear models built by the same Architect-Engineer (A- E). In addition, we show that the standardization of nuclear reactors under construction has an indirect and positive effect on construction costs through a reduction in lead-time, the latter being one of the main drivers of construction costs. Conversely, we also explore the possibility of learning by searching and find that, contrary to other energy technologies, innovation leads to construction costs increases. (authors)

  4. Impact of proposed research reactor standards on reactor operation

    International Nuclear Information System (INIS)

    A Standards Committee on Operation of Research Reactors, (ANS-15), sponsored by the American Nuclear Society, was organized in June 1971. Its purpose is to develop, prepare, and maintain standards for the design, construction, operation, maintenance, and decommissioning of nuclear reactors intended for research and training. Of the 15 original members, six were directly associated with operating TRIGA facilities. This committee developed a standard for the Development of Technical Specifications for Research Reactors (ANS-15.1), the revised draft of which was submitted to ANSI for review in May of 1973. The Committee then identified 10 other critical areas for standards development. Nine of these, along with ANS-15.1, are of direct interest to TRIGA owners and operators. The Committee was divided into subcommittees to work on these areas. These nine areas involve proposed standards for research reactors concerning: 1. Records and Reports (ANS-15.3) 2. Selection and Training of Personnel (ANS-15.4) 3. Effluent Monitoring (ANS-15.5) 4. Review of Experiments (ANS-15.6) 5. Siting (ANS-15.7) 6. Quality Assurance Program Guidance and Requirements (ANS-15.8) 7. Restrictions on Radioactive Effluents (ANS-15.9) 8. Decommissioning (ANS-15.10) 9. Radiological Control and Safety (ANS-15.11). The present status of each of these standards will be presented, along with their potential impact on TRIGA reactor operation. (author)

  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. Nuclear reactors' construction costs: The role of lead-time, standardization and technological progress

    International Nuclear Information System (INIS)

    This paper provides an econometric analysis of nuclear reactor construction costs in France and the United States based on overnight costs data. We build a simultaneous system of equations for overnight costs and construction time (lead-time) to control for endogeneity, using change in expected electricity demand as instrument. We argue that the construction of nuclear reactors can benefit from standardization gains through two channels. First, short term coordination benefits can arise when the diversity of nuclear reactors' designs under construction is low. Second, long term benefits can occur due to learning spillovers from past constructions of similar reactors. We find that construction costs benefit directly from learning spillovers but that these spillovers are only significant for nuclear models built by the same Architect–Engineer. In addition, we show that the standardization of nuclear reactors under construction has an indirect and positive effect on construction costs through a reduction in lead-time, the latter being one of the main drivers of construction costs. Conversely, we also explore the possibility of learning by searching and find that, contrary to other energy technologies, innovation leads to construction costs increases. -- Highlights: •This paper analyses the determinants of nuclear reactors construction costs and lead-time. •We study short term (coordination gains) and long term (learning by doing) benefits of standardization in France and the US. •Results show that standardization of nuclear programs is a key factor for reducing construction costs. •We also suggest that technological progress has contributed to construction costs escalation

  7. The IAEA Safety Standards and Small and Medium Reactor Design Features

    International Nuclear Information System (INIS)

    The renewed interest in nuclear energy worldwide prompted the development of small and medium size reactors (SMR). These reactors are more suitable for smaller grids as well as they promise increased safety, possibly better economy, and they offer flexibility in applications (including non-power), siting and fuel cycle options. The IAEA safety standards, especially the Fundamental Safety Principles (IAEA Safety Standards series No. SF-1) provide excellent principles for design of nuclear power systems, these principles however, as such, are very generic. The IAEA safety standard Safety of Nuclear Power Plants: Design (SSR-2/1) provides specific requirements for the design of nuclear power plants, yet these requirements were developed based on best practices related to the design of current generation of reactors. On other hand, the SMRs currently under design incorporate features that are innovative but with which there is no operational experience, therefore the current standards and design requirement might not be directly applicable to the new SMR designs. This paper examines four most advanced in their development designs (NuScale, mPower, SMART, CAREM) with respect to design safety requirements of the IAEA. These designs were selected because all of them are pressurized water reactors and among SMRs they are closest relatives to the current designs currently Being offered to the market. Therefore, the existing safety standards should be potentially applicable. Since these SMR designs are under development and most of the detailed information is not available this paper utilizes only information that is public domain. This limits the depth of the evaluation and mainly principal technical requirements and some general design requirements are addressed. Focus is on issues such as defence in depth, design principles (single failure criterion, fail safe design, etc.), design extension conditions, external hazards. (author)

  8. Standard Guide for Benchmark Testing of Reactor Dosimetry in Standard and Reference Neutron Fields

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2010-01-01

    Return to Contents page 1.1 This guide covers facilities and procedures for benchmarking neutron measurements and calculations. Particular sections of the guide discuss: the use of well-characterized benchmark neutron fields to calibrate integral neutron sensors; the use of certified-neutron-fluence standards to calibrate radiometric counting equipment or to determine interlaboratory measurement consistency; development of special benchmark fields to test neutron transport calculations; use of well-known fission spectra to benchmark spectrum-averaged cross sections; and the use of benchmarked data and calculations to determine the uncertainties in derived neutron dosimetry results. 1.2 The values stated in SI units are to be regarded as standard. No other units of measurement are included in this standard.

  9. 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

  10. 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

  11. 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.

  12. Obituary: Horace Welcome Babcock, 1912-2003

    Science.gov (United States)

    Vaughan, Arthur Harris

    2003-12-01

    sunspot cycles. Until about 1957 this work had been done at the Hale Solar Laboratory on Holladay Road in Pasadena. Improved models of the magnetograph developed by Robert F. Howard, in collaboration with Horace, went into operation in the 150-foot solar tower telescope at Mount Wilson in 1959 and later, and similar instruments are now employed at many other solar observatories. In 1961 Horace proposed an explanation of the Sun's 22-year magnetic cycle that contained many of the features still embodied in contemporary theoretical models of the phenomenon. The advance in our understanding of solar and stellar magnetism brought forth by Horace Babcock is a worthy sequel to the pioneering efforts initiated by George E. Hale early in the twentieth century. Faced with the growing obsolescence of the Carnegie Institution of Washington's facilities at Mount Wilson along with the competition from Caltech's 200-inch telescope, the Carnegie Trustees in 1963 adopted the idea of founding a major observatory in the Southern Hemisphere as its master plan for modernizing the astronomical facilities of the Institution. Upon becoming Director of the Mount Wilson and Palomar Observatories in 1964, Horace Babcock embraced the job of carrying out this plan, although it meant giving up his own science. Beginning in 1963, and with his usual ingenuity, Horace developed apparatus for measuring astronomical ``seeing." In collaboration with John Irwin and others, he carried out site surveys in Chile, Australia and New Zealand with the aim of selecting the best available location for the anticipated array of large telescopes. Some five years of exploration led, in 1968, to the selection and purchase of a 276 square-kilometer tract on Cerro Las Campanas in north central Chile as the site for the new observatory. Babcock and Irwin had first climbed to its summit, on foot, in October 1966. The team Horace assembled to build the observatory and its infrastructure proved equal to the high standards he

  13. Standardization of specifications and inspection procedures for LEU plate-type research reactor fuels

    International Nuclear Information System (INIS)

    With the transition to high density uranium LEU fuel, fabrication costs of research reactor fuel elements have a tendency to increase because of two reasons. First, the amount of the powder of the uranium compound required increases by more than a factor of five. Second, fabrication requirements are in many cases nearer the fabrication limits. Therefore, it is important that measures be undertaken to eliminate or reduce unnecessary requirements in the specification or inspection procedures of research reactor fuel elements utilizing LEU. An additional stimulus for standardizing specifications and inspection procedures at this time is provided by the fact that most LEU conversions will occur within a short time span, and that nearly all of them will require preparation of new specifications and inspection procedures. In this sense, the LEU conversions offer an opportunity for improving the rationality and efficiency of the fuel fabrication and inspection processes. This report focuses on the standardization of specifications and inspection processes of high uranium density LEU fuels for research reactors. However, in many cases the results can also be extended directly to other research reactor fuels. 15 refs, 1 fig., 3 tabs

  14. High flux materials testing reactor HFR Petten. Characteristics of facilities and standard irradiation devices

    International Nuclear Information System (INIS)

    For the materials testing reactor HFR some characteristic information is presented. Besides the nuclear data for the experiment positions short descriptions are given of the most important standard facilities for material irradiation and radionuclide production. One paragraph deals with the experimental set-ups for solid state and nuclear structure investigations. The information in this report refers to a core type, which is operational since March 1977. The numerical data compiled have been up-dated to June 1978

  15. Selective methods for the maintainability and standardization of the engineering of a research reactor

    International Nuclear Information System (INIS)

    the same function in each specialty. These diversities bring about conflicts and confusion between the maintenance and operation crew, besides modifying dangerously the fail rate and thus the overall reliability of the reactor. The maintainability is the capacity of being maintained an equipment/system has, serving as a design parameter. A system must be designed in a way in which it is maintained without a great investment of time and with low costs, minimum environmental impact and the least resources possible. Standardization is the action of normalizing the engineering of all systems/equipments of the reactor from its design, in all the disciplines, (mechanical, electrical, electronic, chemical, etc.) taking into consideration the facility of its maintenance and conserving or increasing the reliability of the system. The intention of this Program of Maintainability and Standardization in Research Reactors is based on procedures and calculations to improve the reliability of the equipments/systems according to pre-established criterion. (author)

  16. 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)

  17. 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

  18. Development of standards and investigation of safety examination items for advancement of safety regulation of fast breeder reactor

    International Nuclear Information System (INIS)

    The purposes of this study are to prepare the fuel technical standard and the structure and materials standard of fast breeder reactors (FBRs), and to develop the requirements in a reactor establishment permission. The objects of this study are mainly the Monju high performance core and a demonstration FBR. In JFY 2012, the following results were obtained. As for the fuel technical standard, the fuel technical standard adapting the examination of integrity of the FBR fuels was prepared based on the information and data obtained in this study. As for the structure and material standard, the investigation of the revised parts of the standard was carried out. And as for the examination of the safety requirements, safety evaluation items for the future FBR plant and the fission products to be considered in a reactor establishment permission were investigated and examined. (author)

  19. Catalogue and classification of technical safety standards, rules and regulations for nuclear power reactors and nuclear fuel cycle facilities

    International Nuclear Information System (INIS)

    The present report is an up-dated version of the report 'Catalogue and Classification of Technical Safety Rules for Light-water Reactors and Reprocessing Plants' edited under code No EUR 5362e, August 1975. Like the first version of the report, it constitutes a catalogue and classification of standards, rules and regulations on land-based nuclear power reactors and fuel cycle facilities. The reasons for the classification system used are given and discussed

  20. Standard- and extended-burnup PWR [pressurized-water reactor] and BWR [boiling-water reactor] reactor models for the ORIGEN2 computer code

    International Nuclear Information System (INIS)

    The purpose of this report is to describe an updated set of reactor models for pressurized-water reactors (PWRs) and boiling-water reactors (BWRs) operating on uranium fuel cycles and the methods used to generate the information for these models. Since new fuel cycle schemes and reactor core designs are introduced from time to time by reactor manufacturers and fuel vendors, an effort has been made to update these reactor models periodically and to expand the data bases used by the ORIGEN2 computer code. In addition, more sophisticated computational techniques than previously available were used to calculate the resulting reactor model cross-section libraries. The PWR models were based on a Westinghouse design, while the BWR models were based on a General Electric BWR/6 design. The specific reactor types considered in this report are as follows (see Glossary for the definition of these and other terms): (1) PWR-US, (2) PWR-UE, (3) BWR-US, (4) BWR-USO, and (5) BWR-UE. Each reactor model includes a unique data library that may be used to simulate the buildup and deletion of isotopes in nuclear materials using the ORIGEN2 computer code. 33 refs., 44 tabs

  1. Brief overview of American Nuclear Society's research reactor standards

    International Nuclear Information System (INIS)

    The American Nuclear Society (ANS) established the research reactor standards group in 1968. The standards group, known as ANS-15, was established for the purpose of developing, preparing, and maintaining standards for the design, construction, operation, maintenance, and decommissioning of nuclear reactors intended for research and training

  2. Standard review plan for the review and evaluation of emergency plans for research and test reactors. Technical report

    International Nuclear Information System (INIS)

    This document provides a Standard Review Plan for the guidance of the NRC staff to assure that complete and uniform reviews are made of research and test reactor emergency plans. The report is organized under ten planning standards which correspond to the guidance criteria in Draft II of ANSI/ANS 15.16 as endorsed by Revision 1 to Regulatory Guide 2.6. The applicability of the items under each planning standard is indicated by subdivisions of the steady state thermal power levels at which the reactors are licensed to operate. Standard emergency classes and example action levels for research and test reactors which should initiate these classes are given in an Appendix

  3. 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.

  4. Standard mirror fusion reactor design study

    International Nuclear Information System (INIS)

    This report covers the work of the Magnetic Fusion Energy Division's reactor study group during FY 1976 on the standard mirror reactor. The ''standard'' mirror reactor is characterized as a steady state, neutral beam sustained, D-T fusioning plasma confined by a Yin-Yang magnetic mirror field. The physics parameters are obtained from the same physics model that explains the 2XIIB experiment. The model assumes that the drift cyclotron loss cone mode occurs on the boundary of the plasma, and that it is stabilized by warm plasma with negligible energy investment. The result of the study was a workable mirror fusion power plant, steady-state blanket removal made relatively simple by open-ended geometry, and no impurity problem due to the positive plasma potential. The Q (fusion power/injected beam power) turns out to be only 1.1 because of loss out the ends from Coulomb collisions, i.e., classical losses. This low Q resulted in 77% of the gross electrical power being used to power the injectors, thereby causing the net power cost to be high. The low Q stimulated an intensive search for Q-enhancement concepts, resulting in the LLL reactor design effort turning to the field reversal mirror and the tandem mirror, each having Q of order 5

  5. Reactor Section standard analytical methods. Part 1

    Energy Technology Data Exchange (ETDEWEB)

    Sowden, D.

    1954-07-01

    the Standard Analytical Methods manual was prepared for the purpose of consolidating and standardizing all current analytical methods and procedures used in the Reactor Section for routine chemical analyses. All procedures are established in accordance with accepted practice and the general analytical methods specified by the Engineering Department. These procedures are specifically adapted to the requirements of the water treatment process and related operations. The methods included in this manual are organized alphabetically within the following five sections which correspond to the various phases of the analytical control program in which these analyses are to be used: water analyses, essential material analyses, cotton plug analyses boiler water analyses, and miscellaneous control analyses.

  6. 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.

  7. 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

  8. Standardization of advanced light water reactors and progress on achieving utility requirements

    International Nuclear Information System (INIS)

    This paper reports that for a number of years, the U.S. utilities had led an industry-wide effort to establish a technical foundation for the design of the next generation of light water reactors in the United States. Since 1985, this utility initiative has been effected through a major technical program managed by the Electric Power Research Institute (EPRI); the U.S. Advanced Light Water Reactor (ALWR) Program. In addition to the U.S. utility leadership and sponsorship, the ALWR Program also has the participation and sponsorship of a number of international utility companies and close cooperation with the U.S. Department of Energy (DOE). The NPOC Strategic Plan for Building New Nuclear Plants creates a framework within which new standardized nuclear plants may be built. The Strategic Plan is an expression of the nuclear energy industry's serious intent to create the necessary conditions for new plant construction and operation. The industry has assembled a comprehensive, integrated list of actions that must be taken before new plants will be built and assigns responsibility for managing the various issues and sets time-tables and milestones against which we must measure progress

  9. Guidelines for estimating present and forecasting future population distributions surrounding power reactor sites. (Draft of a standard)

    International Nuclear Information System (INIS)

    This report presents demographic procedures for estimating and forecasting population totals and distributions within a 50-mile radius of a proposed power plant site. Procedures such as those described are needed to comply with the license application procedure for a proposed power reactor site. The report includes a draft for a prospective American National Standard to estimate present and forecast future population distributions surrounding proposed reactor sites

  10. 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

  11. Self absorption and geometric correction factors for reactor off-gas samples relative to NBS standards

    International Nuclear Information System (INIS)

    Although they can be counted in identical bottles using identical counting systems, real gas samples differ from the NBS solution standards (e.g., mock reactor off-gas) in two respects--geometry and self absorption. Because both detector and source are real and finite, the simple ''narrow beam'' linear attenuation coefficient approximations currently used in the industry are quite inadequate for correction. Accordingly, the well-tested, complete-analog program, BIM 130, was used to compute the fraction of photons, and the photon energy spectra, reaching typical detectors used in the industry. Using this method, it was possible to correct the given NBS standard activity to its effective activity relative to a gas sample in an identical bottle. Factors were much closer to unity than predictions based on ''narrow beam'' linear attenuation coefficient approximations. At 80 keV, for example, such approximations gave 0.76, whereas the factor proved to be 1.03 for a 3'' x 3'' NaI(Tl) crystal and a 3 cm distance. Results are presented for various gamma energies of interest from 80 keV to 1,830 keV, and for the commonly used industrial distances of 3, 10, and 30 cm from the bottom of the sample bottle to the top of the detector container. Complete spectra for photons entering the detectors, as well as factors derived from these for typical NaI(Tl) and Ge(Li) detector resolutions, are given

  12. 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...

  13. Standard Test Method for Application and Analysis of Helium Accumulation Fluence Monitors for Reactor Vessel Surveillance, E706 (IIIC)

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2007-01-01

    1.1 This test method describes the concept and use of helium accumulation for neutron fluence dosimetry for reactor vessel surveillance. Although this test method is directed toward applications in vessel surveillance, the concepts and techniques are equally applicable to the general field of neutron dosimetry. The various applications of this test method for reactor vessel surveillance are as follows: 1.1.1 Helium accumulation fluence monitor (HAFM) capsules, 1.1.2 Unencapsulated, or cadmium or gadolinium covered, radiometric monitors (RM) and HAFM wires for helium analysis, 1.1.3 Charpy test block samples for helium accumulation, and 1.1.4 Reactor vessel (RV) wall samples for helium accumulation. This standard does not purport to address all of the safety concerns, if any, associated with its use. It is the responsibility of the user of this standard to establish appropriate safety and health practices and determine the applicability of regulatory limitations prior to use.

  14. 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

  15. IAEA role in nuclear reactor safety standardization

    International Nuclear Information System (INIS)

    In 1981 the electricity generation by nuclear power plants all over the world reached 8% of total production. It can be expected that at the turn of century up to 25% of electric power will be provided by means of nuclear fuel burning. In connection with NPP total number growth, their attraction to large population centres, widening of the atomic energy application areas, the importance of nuclearreactor safety problems can only increase. The safety measures have usually the structure of sequential barriers: for accident preventing, for protection from accidents, for accident localization. NPP safety is a complex problem having scientific, engineering, juridical, social and political aspects. Since these problems have an international importance, IAEA should actively work on their solving. Practically all the topics of nuclear power development and nuclear reactor s;fety lie within the activity area of the Department of nuclear power and safety, its sections: of nuclear safety, nuclear power, nuclear fuel cycle. In 1974 a decision was made in IAEA about initiation of work on development of an international nuclear safety standards system (NUSS Programme). These activities are divided into five major branches: a government organization for nuclear safety regulations; site selection for NPP; NPP desing; operation, start of operation and decommissioning; quality provision for NPP. The report presents a list of documents, comprising the NUSS Programme. The complection of all the works within the scope of the Programme is planned for 1985. After 1985 the start of development of fast neutron reactor and fuel cycle enterprise safety standards is planned

  16. 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.

  17. 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)

  18. Integrated lid unit for a nuclear reactor of standard construction

    International Nuclear Information System (INIS)

    This is an integrated lid unit for a nuclear reactor of standard construction, where many components and sub-groups of the upper reactor structure are collected into one unit, which is lifted in one lifting operation from the reactor containment vessel. The integrated lid unit includes, in particular, the pressure vessel lid, a cooling jacket, the control rod drive mechanisms, a catch plate, a lifting device, a winch and a cable connection plate. (orig.)

  19. The Advanced Candu reactor annunciation system - Compliance with IEC standard and US NRC guidelines

    International Nuclear Information System (INIS)

    Annunciation is a key plant information system that alerts Operations staff to important changes in plant processes and systems. Operational experience at nuclear stations worldwide has shown that many annunciation implementations do not provide the support needed by Operations staff in all plant situations. To address utility needs for annunciation improvement in Candu plants, AECL in partnership with Canadian Candu utilities, undertook an annunciation improvement program in the early nineties. The outcome of the research and engineering development program was the development and simulator validation of alarm processing, display, and information presentation techniques that provide practical and effective solutions to key operational deficiencies with earlier annunciation implementations. The improved annunciation capabilities consist of a series of detection, information processing and presentation functions called the Candu Annunciation Message List System (CAMLS). The CAMLS concepts embody alarm processing, presentation and interaction techniques, and strategies and methods for annunciation system configuration to ensure improved annunciation support for all plant situations, especially in upset situations where the alarm generation rate is high. The Advanced Candu Reactor (ACR) project will employ the CAMLS annunciation concepts as the basis for primary annunciation implementations. The primary annunciation systems will be implemented from CAMLS applications hosted on AECL Advanced Control Centre Information System (ACCIS) computing technology. The ACR project has also chosen to implement main control room annunciation aspects in conformance with the following international standard and regulatory review guide for control room annunciation practice: International Electrotechnical Commission (IEC) 62241 - Main Control Room, Alarm Function and Presentation (International standard) US NRC NUREG-0700 - Human-System Interface Design Review Guidelines, Section 4

  20. 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.

  1. 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

  2. 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.

  3. International standardization of nuclear reactor designs - the way forward

    International Nuclear Information System (INIS)

    The concept of 'International Standardization of Nuclear Reactor Designs' means that vendors could build their designs in every country without having to adapt it specifically to national safety requirements. Such standardization would have two main effects. It would greatly facilitate nuclear new build worldwide by giving greater efficiency and certainty to the national licensing procedures; by taking into account the fact that vendors, and nowadays also utilities, are active across borders; by helping developing countries to establish their nuclear new build programmes; and by reducing the strain on human resources on both the regulators' and the industry's side. The second valuable effect of standardization would be to further enhance safety by improving the exchange of construction and operating experience among a number of reactors belonging to fleets of the same design. The World Nuclear Association's CORDEL (Cooperation in Reactor Design Evaluation and Licensing) Group has developed a concept for implementation of international standardization of reactor designs. It has defined a number of steps to be taken by industry. At the same time, possibilities offered by national and international regulatory mechanisms would have to be fully made use of, and some changes in regulatory frameworks might be necessary. Some steps especially towards greater cooperation of regulators have already been taken; however, much still remains to be done. The concept of deploying standardized reactor designs across a number of countries supposes an alignment and, if possible, harmonization of national safety standards; a streamlining of national licensing procedures, making them more efficient and predictable; and the willingness of national regulators to take into account licensing done in other countries. In the end, this should lead to a mutual acceptance of design approvals or, in a more distant future, even to a multinational design approval process. All in all, the concept

  4. Computer-assisted reactor NAA of geological and other reference materials, using the ksub(o)-standardization method

    International Nuclear Information System (INIS)

    USGS BCR-1 and G-2, NBS 1633a Coal Fly-Ash and a 7-element synthetic standard for biological materials were analyzed by reactor NAA, using the ksub(o)-standardization method. The analyses were performed independently in the analytical laboratories of the Institute for Nuclear Sciences (INW), Gent, and the Central Research Institute for Physics (KFKI), Budapest. This procedure allowed not only a comparison with the specified data or with other published values, but enabled a check of the consistency of our own results obtained in largely different experimental conditions. As concluded, the ksub(o)-standardization method combines general versatility (with respect to irradiation and counting conditions) with good accuracy, while the experimental work remains as simple as possible. Since the ksub(o) method is a computer-oriented technique, a FORTRAN IV program was designed and applied on a VAX 11/780 machine. (author)

  5. Development and implementation of k0-INAA standardization at 10 MW Pakistan research reactor-1

    International Nuclear Information System (INIS)

    The k0-method has been developed for 10 MW Pakistan Research Reactor (PARR-1). It involved the full energy peak efficiency calibration of HPGe detector for different counting geometries and characterization of neutron flux at three irradiation channels. Neutron flux was characterized for thermal to epithermal flux ratio, epithermal flux shape factor, modified spectral index. Westcott's g-factor and fast flux. The method was validated by analyzing IAEA-SL1 (lake sediment) and NIST-SRM-1572 (citrus leaves). All calculations were performed in Excel. The results revealed most of the elements with good accuracy. (orig.)

  6. Standard Technical Specifications for Combustion Engineering Pressurized Water Reactors

    International Nuclear Information System (INIS)

    The Standard Technical Specifications for Combustion Engineering Pressurized Water Reactors (CE-STS) is a generic document prepared by the US NRC for use in the licensing process of current Combustion Engineering Pressurized Water Reactors. The CE-STS sets forth the limits, operating conditions, and other requirements applicable to nuclear reactor facility operation as set forth by Section 50.36 of 10 CFR 50 for the protection of the health and safety of the public. The document is revised periodically to reflect current licensing requirements

  7. Standard Technical Specifications for Westinghouse pressurized water reactors

    International Nuclear Information System (INIS)

    The Standard Technical Specifications for Westinghouse Pressurized Water Reactors (W-STS) is a generic document prepared by the U.S. NRC for use in the licensing process of current Westinghouse Pressurized Water Reactors. The W-STS sets forth the Limits, Operating Conditions and other requirements applicable to nuclear reactor facility operation as set forth in by Section 50.36 of 10 CFR Part 50 for the protection of the health and safety of the public. This document is revised periodically to reflect current licensing requirements

  8. Reference design for the standard mirror hybrid reactor

    Energy Technology Data Exchange (ETDEWEB)

    Bender, D.J.; Fink, J.H.; Galloway, T.R.; Kastenberg, W.E.; Lee, J.D.; Devoto, R.S.; Neef, W.S. Jr.; Schultz, K.R.; Culver, D.W.; Rao, S.B.; Rao, S.R.

    1978-05-22

    This report describes the results of a two-year study by Lawrence Livermore Laboratory and General Atomic Co. to develop a conceptual design for the standard (minimum-B) mirror hybrid reactor. The reactor parameters have been chosen to minimize the cost of producing nuclear fuel (/sup 239/Pu) for consumption in fission power reactors (light water reactors). The deuterium-tritium plasma produces approximately 400 MW of fusion power with a plasma Q of 0.64. The fast-fission blanket, which is fueled with depleted uranium and lithium, generates sufficient tritium to run the reactor, has a blanket energy multiplication of M = 10.4, and has a net fissile breeding ratio of Pu/n = 1.51. The reactor has a net electrical output of 600 MWe, a fissile production of 2000 kg of plutonium per year (at a capacity factor of 0.74), and a net plant efficiency of 0.18. The plasma-containment field is generated by a Yin-Yang magnet using NbTi superconductor, and the neutral beam system uses positive-ion acceleration with beam direct conversion. The spherical blanket is based on gas-cooled fast reactor technology. The fusion components, blanket, and primary heat-transfer loop components are all contained within a prestressed-concrete reactor vessel, which provides magnet restraint and supports the primary heat-transfer loop and the blanket.

  9. Reference design for the standard mirror hybrid reactor

    International Nuclear Information System (INIS)

    This report describes the results of a two-year study by Lawrence Livermore Laboratory and General Atomic Co. to develop a conceptual design for the standard (minimum-B) mirror hybrid reactor. The reactor parameters have been chosen to minimize the cost of producing nuclear fuel (239Pu) for consumption in fission power reactors (light water reactors). The deuterium-tritium plasma produces approximately 400 MW of fusion power with a plasma Q of 0.64. The fast-fission blanket, which is fueled with depleted uranium and lithium, generates sufficient tritium to run the reactor, has a blanket energy multiplication of M = 10.4, and has a net fissile breeding ratio of Pu/n = 1.51. The reactor has a net electrical output of 600 MWe, a fissile production of 2000 kg of plutonium per year (at a capacity factor of 0.74), and a net plant efficiency of 0.18. The plasma-containment field is generated by a Yin-Yang magnet using NbTi superconductor, and the neutral beam system uses positive-ion acceleration with beam direct conversion. The spherical blanket is based on gas-cooled fast reactor technology. The fusion components, blanket, and primary heat-transfer loop components are all contained within a prestressed-concrete reactor vessel, which provides magnet restraint and supports the primary heat-transfer loop and the blanket

  10. Efficient Implementation of Volterra Filters for De-interlacing TV Images - Snell and Wilcox

    DEFF Research Database (Denmark)

    Budd, Chris; Gravesen, Jens; Willson, Eddie

    A standard TV image is transmitted as a series of horizontal lines. To reduce band-width effects in transmission, half of a picture is transmitted in each frame, i.e., information is only given about pictures on alternate lines, a process called interlacing. A difficulty with this process is that...

  11. 78 FR 59981 - Proposed Revision to Physical Security-Standard Design Certification and Operating Reactors

    Science.gov (United States)

    2013-09-30

    ..., incorporate licensing experience from previous design certification application reviews, and to inform... From the Federal Register Online via the Government Publishing Office NUCLEAR REGULATORY... AGENCY: Nuclear Regulatory Commission. ACTION: Standard review plan-draft section revision; request...

  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. Outlines of revised regulation standards for experimental research reactors

    International Nuclear Information System (INIS)

    In response to the accident of TEPCO Fukushima Daiichi Nuclear Power Station, the government took actions through the revision of regulatory standards as well as the complete separation of regulation administrative department from promotion administrative department. The Nuclear and Industrial Safety Agency of the Ministry of Economy, Trade and Industry, which has been in charge of the regulations of commercial reactors, and the Office of Nuclear Regulations of the Ministry of Education, Culture, Sports, Science and Technology, which has been in charge of the regulations of reactors for experiment and research, were separated from both ministries, and integrated into the Nuclear Regulation Authority, which was newly established as the affiliated agency of the Ministry of the Environment. As for the revision of regulations and standards, the Nuclear Safety Commission was dismantled, and regulation enacting authority was given to the new Nuclear Regulation Authority, and the regulations that stipulated new regulatory standards were enacted. This paper outlines the contents of regulations related mainly to the reactors for experiment and research, and explains the following: (1) retroactive application of the new regulatory standards to existing reactor facilities, (2) examinations at the Nuclear Regulatory Agency, (3) procedures to confirm the compliance to the new standards, (4) seismic design classification, and (5) importance classification of safety function. (A.O.)

  15. 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

  16. The application and development of k0-standardization method of neutron activation analysis at Dalat research reactor

    International Nuclear Information System (INIS)

    In recent years the k0-NAA method has been applied and developed at the 500 kW Dalat research reactor, which includes the establishment of a PC database of k0-NAA-related nuclear parameters, e.g., radionuclide produced, half-lives, k0-factors, Q0, E-barr Eγ, etc; the access to the database is able by a k0-NAA software or by manual; the detection efficiency calibration of gamma spectrometers used in k0-NAA, the determination of reactor neutron spectrum parameters such as α and f factors and neutron fluxes in the irradiation channels, and the validation of the developed k0-NAA procedure by analysing some SRMs, namely Coal Fly Ash (NIST-1633b), Bovine Liver (NIST-1577b) and IAEA-Soil7. The analytical results showed the deviations between experimental and certified values were mostly less than 15% with most Z-scores lower than 2. The k0-NAA procedure established at the Dalat research reactor has been regarded as a reliable standardization method of NAA and as available for practical applications, in particularly for airborne particulate and crude oil samples. (author)

  17. Safety-evaluation report related to the license renewal and power increase for the National Bureau of Standards Reactor (Docket No. 50-184)

    International Nuclear Information System (INIS)

    This Safety Evaluation Report for the application filed by the National Bureau of Standards (NBS) for an increase in power from 10 MWt to 20 MWt and for a renewal of the Operating License TR-5 to continue to operate the test reactor has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. The facility is located in Gaithersburg, Maryland, on the site of the National Bureau of Standards, which is a bureau of the Department of Commerce. The staff concludes that the NBS reactor can operate at the 20 MWt power level without endangering the health and safety of the public

  18. Fusion reactor design studies: standard accounts for cost estimates

    International Nuclear Information System (INIS)

    The fusion reactor design studies--standard accounts for cost estimates provides a common format from which to assess the economic character of magnetically confined fusion reactor design concepts. The format will aid designers in the preparation of design concept costs estimates and also provide policymakers with a tool to assist in appraising which design concept may be economically promising. The format sets forth a categorization and accounting procedure to be used when estimating fusion reactor busbar energy cost that can be easily and consistently applied. Reasons for developing the procedure, explanations of the procedure, justifications for assumptions made in the procedure, and the applicability of the procedure are described in this document. Adherence to the format when evaluating prospective fusion reactor design concepts will result in the identification of the more promising design concepts thus enabling the fusion power alternatives with better economic potential to be quickly and efficiently developed

  19. Guidance of clearance related standards in reactor facilities

    International Nuclear Information System (INIS)

    The reactor regulation law was amended in May 2005 to provide for the clearance system. The regulatory body confirmed radioactivity concentration of material and allowed the material be classified as 'material cleared from regulatory control'. The procedure of confirmation consisted of two steps; 1) methodology and 2) results, of measurement and assessment. Clearance related standards were issued as ordinance of Ministry and NISA information notice, which specified radioactive nuclides, radioactivity concentration, evaluation unit, how to decide radioactivity concentration, radiation measurement equipment and control of object material. This report was guidance of clearance related standards in reactor facilities with explanatory notes. By applying the clearance system, material of insignificant radiation level, which was part of concrete or metals generated from decommissioning of reactor facilities, could be released from regulatory control of radioactive waste as clearance material, and recycle of such material could result in reduction of radioactive waste amount. (T. Tanaka)

  20. 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...

  1. 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

  2. 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

  3. Exit reactor Thetis/Ghent (1967-2003). A recollection of its significant contribution to NAA and its leading role in the development of the k0-standardization

    International Nuclear Information System (INIS)

    After 36 years of operation, reactor Thetis at the Institute for Nuclear Sciences (Ghent University) was decommissioned in December 2003. On this occasion, a survey is presented of the characteristics and features of Thetis, which were opening the way to its significant contribution to NAA and its leading role in the development of the k0-standardization. A summary is given, including a few specific examples, of fundamental analytical developments and practical applications based on irradiations in the reactor Thetis. (author)

  4. 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.

  5. Evaluation and standardization of neutron activation analysis according to the K0 method in the RP-10 reactor

    International Nuclear Information System (INIS)

    It has been characterized and standardized an irradiation of the RP-10 Research Nuclear Reactor for use of the K0 method of neutron activation analysis using the Hoegdahl convention; also it has been evaluate the behaviour of such method in regard to the accuracy and precision of the results obtained in the quantitative multi elemental analysis of several certified materials of reference. In order to prove that the analytical method is totally under statistical control, it has been used the Heydorn method. It has been verified that the method is exact, precise and reliable to determine the aluminium, antimuonium, arsenic, bromine, calcium, chloride, copper, magnesium, manganese, sodium, titanium, vanadium, zinc and other elements. Also, they are discussed, in regard to the use of K0 constants, the different formalisms employed to calculate the integral of the reaction rate by nucleus in the activation. (author). 58 refs., 18 tabs., 6 figs

  6. Final Environmental Statement related to license renewal and power increase for the National Bureau of Standards Reactor: Docket No. 50-184

    International Nuclear Information System (INIS)

    This Final Environmental Statement contains an assessment of the environmental impact associated with renewal of Operating License No. TR-5 for the National Bureau of Standards (NBS) reactor for a period of 20 years at a power level of 20 MW. This reactor is located on the 576-acre NBS site near Gaithersburg in Montgomery County, Maryland, about 20 mi northwest of the center of Washington, DC. The reactor is a high-flux heavy-water-moderated, cooled and reflected test reactor, which first went critical on December 7, 1967. Though the reactor was originally designed for 20-MW operation, it has been operating for 14 years at a maximum authorized power level to 10 MW. Program demand is now great enough to warrant operation at a power level of 20 MW. No additional major changes to the physical plant are required to operate at 20 MW

  7. 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.

  8. Safety evaluation report related to the license renewal and power increase for the National Bureau of Standards reactor (Docket No. 50-184)

    International Nuclear Information System (INIS)

    Supplement 1 to the Safety Evaluation Report (SER) related to the renewal of the operating license and for a power increase (10 MWt to 20 MWt) for the research reactor at the National Bureau of Standards (NBS) facility has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. This supplement reports on the review of the licensee's emergency plan, which had not been reviewed at the time the Safety Evaluation Report (NUREG-1007) was published, and the review of the NBS application by the Advisory Committee on Reactor Safeguards, which was completed subsequent to the publication of the SER

  9. 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

  10. General principles of nuclear reactor instrumentation (International Electrotechnical Commission Standard Publication 60231:1967)

    International Nuclear Information System (INIS)

    This Recommendation given guidance on the provision of reactor instrumentation and recommends standards of good practice. The main body of the Recommendation is of general application and aspects applicable only to particular types of reactors are included in Appendices. Items of instrumentation are included only where they have a direct bearing on the over-all safety and effective control of the reactor

  11. Radiation streaming analysis in the korean standard nuclear power plant reactor cavity

    International Nuclear Information System (INIS)

    Radiation shield plugs are provided in the reactor cavity of the KSNP (Korean Standard Nuclear Power Plant) to assure an acceptable low level of radiation streaming up through the reactor cavity. An analysis of radiation streaming in the reactor cavity was performed to evaluate the effectiveness of the reactor cavity shield plugs in meeting their design goals. The MCNP code was used for this analysis with DORT leakage flux at the reactor vessel outer surface. Based on the results of this analysis, the upper and lower shield plugs are important design features to reduce dose rates at the reactor vessel flange level and at the operating floor

  12. 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

  13. 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...

  14. 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...

  15. 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)

  16. Non-Power Reactor Operator Licensing Examiner Standards. Revision 1

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-06-01

    The Non-Power Reactor Operator Licensing Examiner Standards provide policy and guidance to NRC examiners and establish the procedures and practices for examining and licensing of applicants for NRC operator licenses pursuant to Part 55 of Title 10 of the Code of Federal Regulations (10 CFR 55). They are intended to assist NRC examiners and facility licensees to understand the examination process better and to provide for equitable and consistent administration of examinations to all applicants by NRC examiners. These standards are not a substitute for the operator licensing regulations and are subject to revision or other internal operator examination licensing policy changes. As appropriate, these standards will be revised periodically to accommodate comments and reflect new information or experience.

  17. Non-Power Reactor Operator Licensing Examiner Standards. Revision 1

    International Nuclear Information System (INIS)

    The Non-Power Reactor Operator Licensing Examiner Standards provide policy and guidance to NRC examiners and establish the procedures and practices for examining and licensing of applicants for NRC operator licenses pursuant to Part 55 of Title 10 of the Code of Federal Regulations (10 CFR 55). They are intended to assist NRC examiners and facility licensees to understand the examination process better and to provide for equitable and consistent administration of examinations to all applicants by NRC examiners. These standards are not a substitute for the operator licensing regulations and are subject to revision or other internal operator examination licensing policy changes. As appropriate, these standards will be revised periodically to accommodate comments and reflect new information or experience

  18. Non-Power Reactor Operator Licensing Examiner Standards

    International Nuclear Information System (INIS)

    The Non-Power Reactor Operator Licensing Examiner Standards provide policy and guidance to NRC examiners and establish the procedures and practices for examining and licensing of applicants for NRC operator licenses pursuant to Part 55 of Title 10 of the Code of Federal Regulations (10 CFR Part 55). They are intended to assist NRC examiners and facility licensees to understand the examination process better and to provide for equitable and consistent administration of examinations to all applicants by NRC examiners. These standards are not a substitute for the operator licensing regulations and are subject to revision or other internal operator examination licensing policy changes. As appropriate, this standard will be revised periodically to accommodate comments and reflect new information or experience

  19. On the path to ordering standardized advanced light water reactors

    International Nuclear Information System (INIS)

    The international Advanced Light Water Reactor (ALWR) program is specifying, designing, and certifying the next generation of nuclear power plants. Begun in the mid-1980's, the program is on track to permit ordering and construction of families of standardized plants at the start of the twenty-first century. ALWRs will be constructed only if they are economically competitive with alternative forms of electricity generation and are recognized as acceptable and favorable by the public, prospective owners, and investors. This paper first gives an overview of the major building blocks ensuring safe, reliable, and economic designs and the status of those designs. Next it lays out the path the industry has charted toward adopting the ALWR option and indicates the status of three key steps -- design certification, utility requirements, and first-of-a-kind engineering. Lastly, the paper focuses on one of the most important building blocks for ensuring economic viability -- life-cycle standardization. Among the topics are the definition and scope of standardization; its advantages and disadvantages; design team standardization plans that describe the desired or optimum degree of standardization and the processes used to achieve it; and the need for an agreement among all plant owners and operators for implementing and sustaining standardization in families of ALWRs. 10 refs., 5 figs

  20. Analysis of standard and FLIP fuel mixed loading patterns in TRIGA Mark-III reactor

    International Nuclear Information System (INIS)

    Concanavalin-A, lectin extracted from Korean native plant ''Banha'' (Pinellia Termata) and their conjugation products with tyrosine and 5-iodo-6-aminouracil were labelled by radioactive iodine-125, and these labelled compounds were subjected to a hemagglutination test in vitro. The labelling procedures were presented with respect to labelling yield (80-90%) and the trends on the conjugation reaction of lectin with various amino acids were discussed. The results of the agglutination test using 6-aminouracil and tyrosine conjugated lectin labelled with 125I indicated that the white blood cell portion shows high radioactivity comparing that of red blood cell portions. Thus the labelled Banha lectin could show a quantitative profile for the agglutination test and its binding of the white blood cell portion was substantial enough to identify such case as a leukemia. (Author)

  1. Production of 232U from irradiation of standard and thorium-based fuels in PWR reactors

    International Nuclear Information System (INIS)

    The production of small quantities of 232U can induce radiation protection issues in the back end of the fuel cycle, particularly for thorium-based fuels. This is due to its relatively short half life (69 years) and the emission of a high energy gamma ray of 2.6 MeV at the end of its decay chain. With the depletion code MURE, we determine the different reactions pathways, and their proportions, leading to the synthesis of 232U in UO2 and (Th,Pu)O2 fuels irradiated in a PWR. Moreover, the impact, on the 232U production, of cycle times such as time separating the fabrication of the fuel and its irradiation as well as influence of the fissile content has been investigated for UO2 fuel. The impact of the thorium ore provenance and of the plutonium quality has been studied for the (Th,Pu)O2 case. (author)

  2. 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).

  3. 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.

  4. 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

  5. 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...

  6. Standard technical specifications for Westinghouse pressurized water reactors (revision issued Fall 1981). Technical report

    International Nuclear Information System (INIS)

    The Standard Technical Specifications for Westinghouse Pressurized Water Reactors (W-STS) is a generic document prepared by the U.S. NRC for use in the licensing process of current Westinghouse Pressurized Water Reactors. The W-STS sets forth the Limits, Operating Conditions and other requirements applicable to nuclear reactor facility operation as set forth in Section 50.36 of 10 CFR Part 50 for the protection of the health and safety of the public

  7. Standard Master Matrix for Light-Water Reactor Pressure Vessel Surveillance Standards, E706(0)

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2002-01-01

    1.1 This master matrix standard describes a series of standard practices, guides, and methods for the prediction of neutron-induced changes in light-water reactor (LWR) pressure vessel (PV) and support structure steels throughout a pressure vessel's service life (Fig. 1). Some of these are existing ASTM standards, some are ASTM standards that have been modified, and some are proposed ASTM standards. General requirements of content and consistency are discussed in Section 6 . More detailed writers' and users' information, justification, and specific requirements for the nine practices, ten guides, and three methods are provided in Sections 3-5. Referenced documents are discussed in Section 2. The summary-type information that is provided in Sections 3 and 4 is essential for establishing proper understanding and communications between the writers and users of this set of matrix standards. It was extracted from the referenced documents, Section 2 and references (1-106) for use by individual writers and users. 1...

  8. Method of realization and exploitation of monitoring system for accuracy and reliability characteristics of standard temperature measurements in WWER-440 reactors at NPP Bohunice V1

    International Nuclear Information System (INIS)

    The sequences in development of computer equipment s and the sequences in development of measurement tools and procedures are listed in submitted presentation - from start-up the power plant in operation until present days. Present status of integration of a monitoring system for accuracy and reliability characteristics of standard temperature measurements in WWER-440 reactors in NPP V1 is presented here. The ways of data acquisition, storing of results and their evaluation are described in this presentation. In conclusion some practical possibilities of using a a monitoring system for accuracy and reliability are listed. (Authors)

  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. Validation of standardized computer analyses for licensing evaluation/TRITON two-dimensional and three-dimensional models for light water reactor fuel

    International Nuclear Information System (INIS)

    The isotopic depletion capabilities of the new Standardized Computer Analyses for Licensing Evaluation control module TRITON, coupled with ORIGEN-S, were evaluated using spent fuel assays from several commercial light water reactors with both standard and mixed-oxide fuel assemblies. Calculations were performed using the functional modules NEWT and KENO-VI. NEWT is a two-dimensional, arbitrary-geometry, discrete-ordinates transport code, and KENO-VI is a three-dimensional Monte Carlo transport code capable of handling complex three-dimensional geometries. To validate the codes and data used in depletion calculations, numerical predictions were compared with experimental measurements for a total of 29 samples taken from the Calvert Cliffs, Obrigheim, and San Onofre pressurized water reactors and the Gundremmingen boiling water reactor. Similar comparisons have previously been performed at the Oak Ridge National Laboratory for the one-dimensional SAS2H control module. The SAS2H, TRITON/KENO-VI, and TRITON/NEWT results were compared for corresponding samples. All analyses showed that TRITON/KENO-VI and TRITON/NEWT produced typically similar or better results than SAS2H. The calculations performed in this validation study demonstrate that the depletion capabilities of TRITON accurately model spent fuel depletion and decay. (authors)

  11. Experience of the standardization of the vibratory condition pipe line when working the reactor on powers

    International Nuclear Information System (INIS)

    Analysis of the experience of the standardization of the vibratory condition pipe line and considered approaches of the motivation of the normative requirements is organized in article to vibratory load on pipe lines when working the reactor on powers

  12. Utilization of Research Reactors in Standard Reference Material Certification

    Energy Technology Data Exchange (ETDEWEB)

    Capannesi, G.; Rosada, A. [UTFISST-CATNUC, ENEA, R.C.-Casaccia, via Anguillarese 301, 00060 Rome (Italy); Avino, P. [DIPIA, INAIL (ex-ISPESL), via Urbana 167, 00184 Rome (Italy)

    2011-07-01

    The certification issue of Standard Reference Materials is one of the most complex analytical problems and runs over different research fields. International organization, e.g. NIST, BCR etc., organize continuously systematic intercomparison campaigns among worldwide laboratories using different analytical techniques. Samples are irradiated in nuclear research reactors and analyzed by Instrumental Neutron Activation Analysis, a technique strongly involved in this field for its significant analytical properties. This paper shows a study on Zircaloy-4. The importance of accurate measurements of minor constituents, i.d. Cr, Fe, Hf and Sn, regards its characteristics of corrosion resistance and mechanical properties. The samples were irradiated in the rotating rack of the TRIGA Mark II reactor of the R.C.-Casaccia (ENEA). The gamma spectrometry measurements were performed after 30 and 90 days of decay by means of HPGe detector. The results obtained by interlaboratory intercomparison can highlight an excellent precision for Cr, Hf and Sn, and a good precision for Fe. The reliability of the technique is confirmed by Hf determination, since the INAA is one of the few analytical techniques measuring and delivering accurate and homogeneous data. (author)

  13. Development and implementation of k{sub 0}-INAA standardization at 10 MW Pakistan research reactor-1

    Energy Technology Data Exchange (ETDEWEB)

    Wasim, M.; Arif, M.; Zaidi, J.H.; Anwar, Y. [Chemistry Div., Pakistan Inst. of Nuclear Science and Technology, Islamabad (Pakistan)

    2009-07-01

    The k{sub 0}-method has been developed for 10 MW Pakistan Research Reactor (PARR-1). It involved the full energy peak efficiency calibration of HPGe detector for different counting geometries and characterization of neutron flux at three irradiation channels. Neutron flux was characterized for thermal to epithermal flux ratio, epithermal flux shape factor, modified spectral index. Westcott's g-factor and fast flux. The method was validated by analyzing IAEA-SL1 (lake sediment) and NIST-SRM-1572 (citrus leaves). All calculations were performed in Excel. The results revealed most of the elements with good accuracy. (orig.)

  14. Standard Test Method for Application and Analysis of Solid State Track Recorder (SSTR) Monitors for Reactor Surveillance, E706(IIIB)

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2003-01-01

    1.1 This test method describes the use of solid-state track recorders (SSTRs) for neutron dosimetry in light-water reactor (LWR) applications. These applications extend from low neutron fluence to high neutron fluence, including high power pressure vessel surveillance and test reactor irradiations as well as low power benchmark field measurement. (1) This test method replaces Method E 418. This test method is more detailed and special attention is given to the use of state-of-the-art manual and automated track counting methods to attain high absolute accuracies. In-situ dosimetry in actual high fluence-high temperature LWR applications is emphasized. 1.2 This test method includes SSTR analysis by both manual and automated methods. To attain a desired accuracy, the track scanning method selected places limits on the allowable track density. Typically good results are obtained in the range of 5 to 800 000 tracks/cm2 and accurate results at higher track densities have been demonstrated for some cases. (2) Trac...

  15. NDE standards for materials of fast reactors

    International Nuclear Information System (INIS)

    Primary objective of this paper is to bring out the salient features of the specifications followed for the procurement of various materials such as 316LN plates, tube sheets and 9Cr-1Mo tube sheets, 2.25Cr-1Mo dished ends and chrome-moly tubes and the difficulties encountered in procurement. 4 figs

  16. 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)

  17. 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)

  18. 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.)

  19. 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

  20. 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

  1. 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.

  2. 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.

  3. 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.

  4. 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).

  5. 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

  6. 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

  7. 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.)

  8. Comparison between MAAP and ECART predictions of radionuclide transport throughout a French standard PWR reactor coolant system

    International Nuclear Information System (INIS)

    In the framework of a collaboration agreement between EDF and ENEL, the MAAP (Modular Accident Analysis Program) and ECART (ENEL Code for Analysis of radionuclide Transport) predictions about the fission product retention inside the reactor cooling system of a French PWR 1300 MW during a small Loss of Coolant Accident were compared. The volatile fission products CsI, CsOH, TeO2 and the structural materials, all of them released early by the core, are more retained in MAAP than in ECART. On the other hand, the non-volatile fission products, released later, are more retained in ECART than in MAAP, because MAAP does not take into account diffusion-phoresis: in fact, this deposition phenomenon is very significant when the molten core vaporizes the water of the vessel lower plenum. Centrifugal deposition in bends, that can be modeled only with ECART, slightly increases the whole retention in the circuit if it is accounted for. (authors). 18 refs., figs., tabs

  9. Standard Technical Specifications for General Electric Boiling Water Reactors (BWR/5)

    International Nuclear Information System (INIS)

    The Standard Technical Specifications for General Electric Boiling Water Reactors (GE-STS) is a generic document prepared by the US NRC for use in the licensing process of current General Electric Boiling Water Reactors. The GE-STS sets forth the limits, operating conditions, and other requirements applicable to nuclear reactor facility operation as set forth by Section 50.36 of 10 CFR Part 50 for the protection of the health and safety of the public. The document is revised periodically to reflect current licensing requirements

  10. 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.

  11. 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...

  12. Draft of standard for graphite core components in high temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    For the design of the graphite components in the High Temperature Engineering Test Reactor (HTTR), the graphite structural design code for the HTTR etc. were applied. However, general standard systems for the High Temperature Gas-cooled Reactor (HTGR) have not been established yet. The authors had studied on the technical issues which is necessary for the establishment of a general standard system for the graphite components in the HTGR. The results of the study were documented and discussed at a 'Special committee on research on preparation for codes for graphite components in HTGR' at Atomic Energy Society of Japan (AESJ). As a result, 'Draft of Standard for Graphite Core Components in High Temperature Gas-cooled Reactor.' was established. In the draft standard, the graphite components are classified three categories (A, B and C) in the standpoints of safety functions and possibility of replacement. For the components in the each class, design standard, material and product standards, and in-service inspection and maintenance standard are determined. As an appendix of the design standard, the graphical expressions of material property data of 1G-110 graphite as a function of fast neutron fluence are expressed. The graphical expressions were determined through the interpolation and extrapolation of the irradiated data. (author)

  13. ANS shielding standards for light-water reactors

    International Nuclear Information System (INIS)

    The purpose of the American Nuclear Society Standards Subcommittee, ANS-6, Radiation Protection and Shielding, is to develop standards for radiation protection and shield design, to provide shielding information to other standards-writing groups, and to develop standard reference shielding data and test problems. A total of seven published ANS-6 standards are now current. Additional projects of the subcommittee, now composed of nine working groups, include: standard reference data for multigroup cross sections, gamma-ray absorption coefficients and buildup factors, additional benchwork problems for shielding problems and energy spectrum unfolding, power plant zoning design for normal and accident conditions, process radiation monitors, and design for postaccident radiological conditions

  14. Standard Guide for Benchmark Testing of Light Water Reactor Calculations

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2010-01-01

    1.1 This guide covers general approaches for benchmarking neutron transport calculations in light water reactor systems. A companion guide (Guide E2005) covers use of benchmark fields for testing neutron transport calculations and cross sections in well controlled environments. This guide covers experimental benchmarking of neutron fluence calculations (or calculations of other exposure parameters such as dpa) in more complex geometries relevant to reactor surveillance. Particular sections of the guide discuss: the use of well-characterized benchmark neutron fields to provide an indication of the accuracy of the calculational methods and nuclear data when applied to typical cases; and the use of plant specific measurements to indicate bias in individual plant calculations. Use of these two benchmark techniques will serve to limit plant-specific calculational uncertainty, and, when combined with analytical uncertainty estimates for the calculations, will provide uncertainty estimates for reactor fluences with ...

  15. Inconsistencies of neutron flux parameters for k0 standardization in neutron activation analysis determined with the use of Au+Zr and Au+Mo+Cr monitor sets at the LVR-15 reactor in Rez

    International Nuclear Information System (INIS)

    Multipurpose research reactors require monitoring of neutron flux parameters (α, f, Fc or Φth ) during every irradiation of samples when k0 standardization in neutron activation analysis is to be used, because the above parameters may change quite frequently and unpredictably. We tested two monitor sets consisting of Au+Zr and Au+Mo+Cr for determination of the neutron flux parameters in k0-NAA at the LVR-15 reactor in Rez. Based on analysis of several reference materials, we found the use of the Au+Zr monitor set superior over the Au+Mo+Cr monitor set in the specific conditions of the LVR-15 reactor. (author)

  16. Benchmark calculations by the thermal reactor standard nuclear design code system SRAC

    International Nuclear Information System (INIS)

    This report summarizes the present status of the thermal reactor standard nuclear design code system SRAC developed by the nuclear design working group of the JAERI thermal reactor standard code committee which was started on July 1978. Descriptions are given at first on the brief introduction and the process of development of the code system SRAC, and then, the several benchmark tests performed to evaluate the performance of the code system. The results show the good predictions of the experimental keff values of the critical facilities; TCA for LWR, JMTRC for JAERI MTR, DCA for the Japanese Advanced Thermal Reactor and SHE for VHTR. A trial to the IAEA benchmark calculations on the Reduction of uranium Enrichment of Research and Test Reactors yields satisfactory agreements with the results of ANL. Another test to evaluate the fast group constants was also attempted by tracing the fast reactor benchmark problems which have been used to evaluate nuclear data file in the FBR reactor physics field. (author)

  17. 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.

  18. 76 FR 14437 - Economic Simplified Boiling Water Reactor Standard Design: GE Hitachi Nuclear Energy; Issuance of...

    Science.gov (United States)

    2011-03-16

    ... From the Federal Register Online via the Government Publishing Office ] NUCLEAR REGULATORY COMMISSION Economic Simplified Boiling Water Reactor Standard Design: GE Hitachi Nuclear Energy; Issuance of... GE Hitachi Nuclear Energy (GEH) for the economic simplified boiling water reactor (ESBWR)...

  19. Neutron flux characterization of the Moroccan Triga Mark II research reactor and validation of the k0 standardization method of NAA using k0-IAEA program

    International Nuclear Information System (INIS)

    The aim of this work was to implement and to validate the k0 standardization method in neutron activation analysis (k0-NAA) at the Moroccan TRIGA Mark II research reactor. This technique was used in order to determine, the calibration of several HPGe detectors and calibration of neutron flux parameters in the typical irradiation channels [rotary specimen rack (RSR) and the pneumatic tube system (PTS) facilities]. Calibrations and calculations of k0-NAA results were carried out using the k0-IAEA program. The two parameters of neutron flux in the selected irradiation channels used for elemental concentration calculation, f (thermal-to-epithermal ratio) and α (deviation from the 1/E distribution), have been determined as well in the PTS as in the RSR facilities using the zirconium bare triple method. Results obtained for f and α in two irradiation channels show that f parameter determined in this way is different in the RSR and the PTS facilities. This can be explained by the fact that the RSR channel is situated in a graphite reflector and is relatively far from the reactor core, while the PTS is in the core. Five reference materials of different origin obtained from USGS (basalt BE-N, bauxite BX-N, biotite mica-Fe, granite GS-N) and IAEA (Soil-7) were used to evaluate the validity of this method in our laboratory by analyzing the elemental concentrations with respect to the certified values. In general, good agreement was obtained between results of this work and values in certificates of the individual reference materials, thus proving the accuracy of our results and successful implementation of the method for analysis of real samples. (author)

  20. 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

  1. An inspection standard of fuel for the high temperature engineering test reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kobayashi, Fumiaki; Shiozawa, Shusaku; Sawa, Kazuhiro; Sato, Sadao (Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment); Hayashi, Kimio; Fukuda, Kosaku; Kaneko, Mitsunobu; Sato, Tsutomu.

    1992-06-01

    The High Temperature Engineering Test Reactor (HTTR) uses the fuel comprising coated fuel particles. A general inspection standard for the coated particle fuel, however, has not been established in Japan. Therefore, it has been necessary to prescribe the inspection standard of the fuel for HTTR. Under these circumstances, a fuel inspection standard of HTTR has been established under cooperation of fuel specialists both inside and outside of JAERI on referring to the inspection methods adopted in USA, Germany and Japan for HTGR fuels. Since a large number of coated fuel particle samples is needed to inspect the HTTR fuel, the sampling inspection standard has also been established considering the inspection efficiency. This report presents the inspection and the sampling standards together with an explanation of these standards. These standards will be applied to the HTTR fuel acceptance tests. (author).

  2. An inspection standard of fuel for the high temperature engineering test reactor

    International Nuclear Information System (INIS)

    The High Temperature Engineering Test Reactor (HTTR) uses the fuel comprising coated fuel particles. A general inspection standard for the coated particle fuel, however, has not been established in Japan. Therefore, it has been necessary to prescribe the inspection standard of the fuel for HTTR. Under these circumstances, a fuel inspection standard of HTTR has been established under cooperation of fuel specialists both inside and outside of JAERI on referring to the inspection methods adopted in USA, Germany and Japan for HTGR fuels. Since a large number of coated fuel particle samples is needed to inspect the HTTR fuel, the sampling inspection standard has also been established considering the inspection efficiency. This report presents the inspection and the sampling standards together with an explanation of these standards. These standards will be applied to the HTTR fuel acceptance tests. (author)

  3. 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…

  4. 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...

  5. 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...

  6. 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...

  7. 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

  8. 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

  9. 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...

  10. Examination of fast-reactor fuels and FBR analytical quality-assurance standards and methods. Progress report, October 1-December 31, 1980

    International Nuclear Information System (INIS)

    This project is directed toward the examination and comparison of the effects of neutron irradiation on Liquid Metal Fast Breeder Reactor (LMFBR) Program fuel materials. Unirradiated and irradiated materials will be examined as requested by the Reference Fuels System Branch of the Division of Reactor Research and Technology (DRRT). Capabilities have been established and are being expanded for providing conventional preirradiation and postirradiation examinations. Nondestructive tests will be conducted in a hot-cell facility specifically modified for examining irradiated prototype fuel pins at a rate commensurate with schedules established by DRRT

  11. 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

  12. 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)

  13. 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…

  14. 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

  15. 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.

  16. 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)

  17. 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.

  18. 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)

  19. Guidelines for preparing and reviewing applications for the licensing of non-power reactors: Standard review plan and acceptance criteria. NUREG - 1537, Part 2

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-02-01

    NUREG - 1537, Part 2 gives guidance on the conduct of licensing action reviews to NRC staff who review non-power reactor licensing applications. These licensing actions include construction permits and initial operating licenses, license renewals, amendments, conversions from highly enriched uranium to low-enriched uranium, decommissioning, and license termination.

  20. Guidelines for preparing and reviewing applications for the licensing of non-power reactors: Standard review plan and acceptance criteria. NUREG - 1537, Part 2

    International Nuclear Information System (INIS)

    NUREG - 1537, Part 2 gives guidance on the conduct of licensing action reviews to NRC staff who review non-power reactor licensing applications. These licensing actions include construction permits and initial operating licenses, license renewals, amendments, conversions from highly enriched uranium to low-enriched uranium, decommissioning, and license termination

  1. Modelling and research of hypothetical non-standard situations of new type nuclear reactors. Research report for 2012

    International Nuclear Information System (INIS)

    The present research report assesses and summarizes the activity for the period from March to December 2012 and is related to the physical and thermophysical properties of the CORIUM. In the first stage of the four-year project, the database of thermophysical data was created. The data define the metals and their oxides contained in CORIUM in the situation of a severe accident of a VVER nuclear power The data were used to predict the parameter setting on the COMETA facility, especially of the high-frequency generator. A system for melt sampling was designed and procedure for sampling from an ingot (the final solidified product of the melting process) was developer for special analysis and irradiation experiments. Several experiments with radioactive material were performed using the COMETA facility, with a precise composition of the batches. The samples taken during the melting and samples from ingots were analyzed by the SEM EDX and X-ray techniques. In parallel, irradiation experiments of non-radioactive samples were performed, and the estimated activities of samples after irradiation were carried out with the aim was to determine the mechanical properties of materials after exposure to radiation. (author)

  2. 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.)

  3. Shifts of neutrino oscillation parameters in reactor antineutrino experiments with non-standard interactions

    CERN Document Server

    Li, Yu-Feng

    2014-01-01

    We discuss reactor antineutrino oscillations with non-standard interactions (NSIs) at the neutrino production and detection processes. The neutrino oscillation probability is calculated with a parametrization of the NSI parameters by splitting them into the averages and differences of the production and detection processes respectively. The average parts induce constant shifts of the neutrino mixing angles from their true values, and the difference parts can generate the energy (and baseline) dependent corrections to the initial mass-squared differences. We stress that only the shifts of mass-squared differences are measurable in reactor antineutrino experiments. Taking Jiangmen Underground Neutrino Observatory (JUNO) as an example, we analyze how NSIs influence the standard neutrino measurements and to what extent we can constrain the NSI parameters.

  4. Shifts of neutrino oscillation parameters in reactor antineutrino experiments with non-standard interactions

    International Nuclear Information System (INIS)

    We discuss reactor antineutrino oscillations with non-standard interactions (NSIs) at the neutrino production and detection processes. The neutrino oscillation probability is calculated with a parametrization of the NSI parameters by splitting them into the averages and differences of the production and detection processes respectively. The average parts induce constant shifts of the neutrino mixing angles from their true values, and the difference parts can generate the energy (and baseline) dependent corrections to the initial mass-squared differences. We stress that only the shifts of mass-squared differences are measurable in reactor antineutrino experiments. Taking Jiangmen Underground Neutrino Observatory (JUNO) as an example, we analyze how NSIs influence the standard neutrino measurements and to what extent we can constrain the NSI parameters

  5. Operator licensing examination standards for power reactors. Interim revision 8

    International Nuclear Information System (INIS)

    These examination standards are intended to assist NRC examiners and facility licensees to better understand the processes associated with initial and requalification examinations. The standards also ensure the equitable and consistent administration of examinations for all applicants. These standards are for guidance purposes and are not a substitute for the operator licensing regulations (i.e., 10 CFR Part 55), and they are subject to revision or other changes in internal operator licensing policy. This interim revision permits facility licensees to prepare their initial operator licensing examinations on a voluntary basis pending an amendment to 10 CFR Part 55 that will require facility participation. The NRC intends to solicit comments on this revision during the rulemaking process and to issue a final Revision 8 in conjunction with the final rule

  6. Three-dimensional flow field analysis of the standard fuel assembly for China advanced research reactor

    International Nuclear Information System (INIS)

    Numerical simulation of the flow field of the standard fuel assembly in China Advanced Research Reactor is carried out by using computational fluid dynamics software CFX4.4 and CFX5.5. The flow distribution and pressure difference of different coolant channels in the standard fuel assembly at rated operating condition are reached. Based on the computational pressure drop results of different flow rate, the resistance characteristic curve is given and compared with experimental results. The two results are in good agreement. (authors)

  7. Standardization of the SLOWPOKE-2 reactor in Jamaica for routine NAA

    International Nuclear Information System (INIS)

    The International Centre for Environmental and Nuclear Sciences (ICENS) has been involved in conducting multipurpose geochemical surveys, the results of which were published in 'A Geochemical Atlas of Jamaican Soils'. The primary analytical tool for these studies was neutron activation analysis (NAA) using the SLOWPOKE-2 reactor at the Centre. The neutron flux of the SLOWPOKE-2 reactor is extremely stable, thus allowing a semi-absolute method for quantitative NAA. This has several advantages, but requires preparation and measurement of the single- or multi-element standards for each gamma-spectroscopy system (GSS). The NAA laboratory at ICENS operates three GSSs. The primary ('master') GSS was standardized using single element standards for over 50 elements, naturally occurring in most geological and biological materials. The standardization of the secondary GSS's was achieved by transferring of the elemental sensitivities of the master GSS using an instrumentation-free standardization approach. Implementation of this methodology and its utilization in the routine analytical work is described. (author)

  8. 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)

  9. 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)

  10. Corrosion experiment on non standard austenitic steel A1, in reactor coolant water

    International Nuclear Information System (INIS)

    Experimental corrosion studies on non standard austenitic SS, A1, have been carried out. The samples were immersed in reactor coolant water medium with pH variation of 5.95, 6.0, 6.1, and 6.31. The experiments were carried out using a type of M-273 EG&G potentiostate/galvanometer test instrument. The post-corrosion samples' microstructure were analyzed with the aid of EDS (energy dispersive spectroscopy) equipped SEM instrument to detect the presence of any viable corrosion products. For further verification x-ray diffraction method was also used to detect any possible emerging corrosion products type on the samples' surfaces. Experimental results confirm that non standard austenitic SS immersed in reactor coolant water corrosion medium with a variation of concentration experience very little or almost no corrosion, and that according to the so-called Fontana's criteria these test-materials turn out to have an excellent resistance toward reactor coolant water corrosion medium. This is also evidenced by the very low corrosion rate value measured in this study. EDS study and X-ray diffraction results indicate that the possible ensuing corrosion by products are chrome oxides and iron oxides. (author)

  11. Standard technical specifications for General Electric boiling water reactors

    International Nuclear Information System (INIS)

    This Standard Technical Specification (STS) has been structured for the broadest possible use on General Electric plants currently being reviewed for an Operating License. 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. This revision of the GE-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

  12. Standard Guide for Application of Neutron Transport Methods for Reactor Vessel Surveillance, E706 (IID)

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2011-01-01

    1.1 Need for Neutronics Calculations—An accurate calculation of the neutron fluence and fluence rate at several locations is essential for the analysis of integral dosimetry measurements and for predicting irradiation damage exposure parameter values in the pressure vessel. Exposure parameter values may be obtained directly from calculations or indirectly from calculations that are adjusted with dosimetry measurements; Guide E944 and Practice E853 define appropriate computational procedures. 1.2 Methodology—Neutronics calculations for application to reactor vessel surveillance encompass three essential areas: (1) validation of methods by comparison of calculations with dosimetry measurements in a benchmark experiment, (2) determination of the neutron source distribution in the reactor core, and (3) calculation of neutron fluence rate at the surveillance position and in the pressure vessel. 1.3 This standard does not purport to address all of the safety concerns, if any, associated with its use. It is th...

  13. 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

  14. 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)

  15. 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)

  16. 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

  17. Implementation of k0-standardization method of the INAA at ETRR-2 research reactor

    International Nuclear Information System (INIS)

    The k0-method of INAA standardization has been implemented using the irradiation facilities of the fast pneumatic rabbit and some selected manually loaded irradiation positions, which designated for short and long irradiation, respectively, at Egypt second research reactor. The neutron flux parameters (f and α) in each site have been determined using Zr-Au sets as neutron flux monitors. The reference materials coal NIST 1632c and IAEA-Soil 7 were analyzed for data validation and good agreement between the experimental values and the certified values was obtained. (author)

  18. 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.

  19. 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.)

  20. Standard technical specifications for Westinghouse pressurized water reactors

    International Nuclear Information System (INIS)

    This Standard Technical Specification (STS) has been structured for the broadest possible use on Westinghouse plants currently being reviewed for an Operating License. Accordingly, the document contains specifications applicable to plants with (1) either 3 or 4 loops and (2) with and without loop stop valves. In addition, four separate and discrete containment specification sections are provided for each of the following containment types: Atmospheric, Ice Condenser, Sub-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 the 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

  1. 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

  2. Standard technical specifications for combustion engineering pressurized water reactors

    International Nuclear Information System (INIS)

    This Standard Technical Specification (STS) has been structured for the broadest possible use on Combustion Engineering 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

  3. 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 ...

  4. 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.

  5. Study on Kο-standardization method of neutron activation analysis in HANARO research reactor

    International Nuclear Information System (INIS)

    The neutron activation analysis based on kο-standardization method (kο-NAA) has been known as one of the remarkable developments of NAA in recent with advantages of experimental simplicity, high accuracy, excellent flexibility with respect to irradiation and counting conditions, and suitability for computerization. This work has carried out to consider the application of kο-NAA on NAA 1 of HANARO research reactor such as the calibration of gamma-ray spectrometry and irradiation facility. α and f-values related to the reactor neutron spectrum parameters as main factor of irradiation quality are measured and some SRMs are analyzed to compare the accuracy of the measured results. The analytical results with the deviation mostly less than ±15% and all Z-score lower 1.5 in comparison between the experiment value and that value certified of some SRMs has shown a promising capability in application of kο-NAA on the HANARO research reactor

  6. 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...

  7. 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

  8. 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

  9. Application of NAA standardization methods using a low power research reactor

    International Nuclear Information System (INIS)

    Two widely used neutron activation analysis (NAA) standardization methods (relative and k0) have been validated at the Ghana Research Reactor-1 (GHARR-1) Centre using environmental and biological standard reference materials (SRMs). The samples were IAEA Soil-7 as an environmental sample, and NIST Orchard Leaves 1571 as a biological sample. The qualitative and quantitative analyses were done using a high resolution Canberra N-type high purity germanium (HPGe) detector. The accuracy and precision were evaluated for the elements analysed. The concentrations of most of the elements were found to be within 10% of the certified values. Precision was calculated from six replicate measurements and was found to be within 15%. (author)

  10. 76 FR 23630 - Office of New Reactors; Proposed Revision 2 to Standard Review Plan, Section 1.0 on Introduction...

    Science.gov (United States)

    2011-04-27

    ... COMMISSION Office of New Reactors; Proposed Revision 2 to Standard Review Plan, Section 1.0 on Introduction...), Section 1.0, ``Introduction and Interfaces'' (Agencywide Documents Access and Management System (ADAMS...: Cindy Bladey, Chief, Rules, Announcements, and Directives Branch (RADB), Office of Administration,...

  11. 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

  12. 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)

  13. Comparison of Standard Light Water Reactor Cross-Section Libraries using the United States Nuclear Regulatory Commission Pressurized Water Reactor Standard Core Loading Benchmark Problem

    Science.gov (United States)

    Arzu Alpan, F.; Kulesza, Joel A.

    2016-02-01

    This paper compares contemporary and historical light water reactor shielding and pressure vessel dosimetry cross-section libraries for a pressurized water reactor calculational benchmark problem with a standard out-in core loading. The calculational benchmark problem was developed at Brookhaven National Laboratory by the request of the U. S. Nuclear Regulatory Commission and used the Oak Ridge National Laboratory two-dimensional discrete ordinates code DORT and the BUGLE-93 cross-section library for the calculations. In this paper, a Westinghouse three-dimensional discrete ordinates code with parallel processing, the RAPTOR-M3G code was used. A variety of cross section libraries were used with RAPTOR-M3G including the BUGLE-93, BUGLE-96, and BUGLE-B7 cross-section libraries developed at Oak Ridge National Laboratory, and the broad-group ALPAN-VII.0 cross-section library developed at Westinghouse. In comparing the calculation-to-calculation reaction rates using the BUGLE-93 cross-section library at the thermal shield, pressure vessel, and cavity capsules, for eleven dosimetry reaction rates, a maximum relative difference of 5% was observed, with the exception of 65Cu(n,2n) in the pressure vessel capsule that had a 90% relative difference with respect to the reference results. It is thought that the 65Cu(n,2n) reaction rate reported in the reference for the pressure vessel capsule is not correct. In considering the libraries developed after BUGLE-93, a maximum relative difference of 12% was observed in reaction rates, with respect to the reference results, for 237Np(n,f) in the cavity capsule using BUGLE-B7.

  14. 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.

  15. Towards standardization of the dissemination measures and tritium solubility in materials of fusion reactors; Hacia la estandarizacion de las medidas de difusion y solubilidad de tritio en materiales de reactores de fusion

    Energy Technology Data Exchange (ETDEWEB)

    Alberto, G.; Penalva, I.; Aranburu, I.; Sarrionandia-Ibarra, A.; Legarda, F.; Martinez, P. M.; Sedano, L.; Moral, N.

    2011-07-01

    The standardization of the measurements of hydrogen isotope interaction with different materials is a challenge and goal of fusion technology programs worldwide. For decades the programs have promoted the need for a reference laboratory for measurements of hydrogen transport to the evolution of fusion technology, but that goal is still pending, in contrast to the situation in other goals I+D.

  16. 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)

  17. Review of ANS-19 standards and activities

    International Nuclear Information System (INIS)

    The standards subcommittee ANS-19, 'Physics of Reactor Design' sponsored by the ANS Reactor Physics Division, has been in existence since 1972. Five standards have been produced, two data review groups have been established, and a working group has reviewed reactor physics terms for inclusion in the N 1.1 American National Standard Glossary of Terms in Nuclear Science and Technology. Much to the disappointment of those of us involved in their development, these standards have for the most part been uniformly ignored by all segments of the nuclear industry. Nevertheless, there is still great interest in revising the standards to make them more useful. It is hoped that the changes now under consideration as well as possible new standards areas of physics inputs to safety and transient calculations and operational power reactor measurements for license compliance will help to bring this about. There is a general feeling that within the large and diverse nuclear industry there is real benefit to be gained both technically and in public perception by the introduction of increasing standardization of good engineering practices. Such standardization to have maximum benefit must come from the technical experts within the industry, rather than being imposed by the regulators. (orig.)

  18. 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

  19. 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

  20. 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.

  1. 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

  2. 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

  3. Study of reactor characteristics for the adaptation of the monoelement standard method in activation analysis. Application to impurity determination in silicon

    International Nuclear Information System (INIS)

    Nuclear reactions by irradiation in a nuclear reactor are reviewed. Quantitative analysis by comparison with multielement standard is treated. Comparison methods using a monoelement standard, easier to use they require a previous study of neutronic characteristics of irradiation channels of reactors (thermal and epithermal flux ratio and eventual deviations fo epithermal neutron energy spectra from the 1/E relationship). Then analysis fo silicon polycrystals by both methods is studied and interfering reactions during irradiation are examined

  4. Performance testing of irradiation facility rabbit system pneumatic reactor RSG-GAS using standard reference material

    International Nuclear Information System (INIS)

    The irradiation facility function test of rabbit system pneumatic (RS-5) has been done using standard reference material SRM 1633 Coal Fly ash through the sending station. Long irradiation of about 4-5 seconds. The results of qualitative analysis showed that the dominant elements listed in the certificate can be detected are Al, Ca, Mg, Si, Na, Ti, V, Mn. But only an element of Mn and Na which has a relative refractive values below 10%. And the other elements have a value relative refractive index 25% - 60%. The significant difference of value was not influenced by the position of irradiation in the reactor facility but due to the influence of the time difference between the sample and the standard count, and the half-life nuclide itself. Overall it can be said that the performance of the irradiation facility pneumatic rabbit system is good, but needs to be tested again by using different standard reference materials, in order to obtain the test results of analysis that can be trusted. (author)

  5. 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)

  6. Reliability study: digital reactor protection system of Korean standard nuclear power plant

    Energy Technology Data Exchange (ETDEWEB)

    Kang, H. G.; Jang, S. C.; Eom, H. S.; Jeong, H. S

    2003-02-01

    Digital safety-critical systems which are now installed in Korean Standard Nuclear Power Plants (KSNPP) would be quantitatively evaluated in order to prove the safety. In this study, we quantify the safety of the digital reactor protection system in KSNPPs using PSA technology. This study also includes the detailed investigation of the target system operation. The Fault Tree (FT) models were constructed for 15 reactor trip parameters. For digital parts, because the operation data for the same type PWR was unavailable, we used the data provided by vendors. On the other hand, for the conventional analog/mechanical parts, we used experience data presented in KAERI/TR-2164/2002.The result of quantification shows that the system unavailability varies from 4.36E-5 to 8.96E-4 according to the trip parameter. Main contributor to the difference from the conventional analysis would be the difference in human failure probability estimation. Generally, the system unavailability depends on several important factors: Human failure probability, software failure probability, watchdog timer coverage, and common cause failure estimation.

  7. 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.

  8. 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

  9. 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.

  10. 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.

  11. 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)

  12. 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).

  13. 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)

  14. Standardization and the European Standards Organisations

    Directory of Open Access Journals (Sweden)

    Marta Orviska

    2014-01-01

    Full Text Available Standardization is a relatively neglected aspect of the EU regulatory process and yet it is fundamental to that process and arguably has recently been the key vehicle in making the single market an economic reality. Yet the key standardization bodies in the EU, the ESOs, are scarcely known to the public and seldom discussed in the literature. In this article we redress this imbalance, arguing that standardization and integration are closely related concepts. We also argue that the ESOs have developed a degree of autonomy in expanding the boundaries of standardization and even in developing their own links with the rest of the world. Recent proposals put forward by the European Commission can be seen as an attempt to reduce that autonomy. These proposals emphasize the speed of, and stakeholder involvement in, standards production, which we further suggest are somewhat conflicting aims.

  15. Standards and standpoints

    DEFF Research Database (Denmark)

    Nissen, Morten

    2016-01-01

    This article argues that critique is a necessary component in any study of standards, just as it is implied in the concept of standard itself. From this follows the relevance of reflexively situating our research in relation to the cultural-historical development of standards and standardization....... The argument takes off from two different conceptualizations of standards in the literature. On the one hand, standards as immanent to practices (the “Neo-Aristotelian” approach), and on the other hand, standards as imposed to regulate practices (the “neo-pragmatist” and “governmentality” approaches......). It is suggested that this opposition can be superseded by articulating the former alternative, not as an essentialism of “practice,” but as the reflexive assumption of standpoint. Some intricacies of the articulation of standpoint are then discussed, concluding in a proposed dialectics of standard...

  16. 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

  17. 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.)

  18. Evidence Standards and Litigation

    DEFF Research Database (Denmark)

    Guerra, Alice; Luppi, Barbara; Parisi, Francesco

    aspect of the legal system: the evidence standard. We recast the conventional rent-seeking model to consider how alternative evidence standards affect litigation choices. We analyze the interrelation between different evidence standards, the effectiveness of the parties’ efforts, and the merits of the...... case. We study how these factors jointly affect the parties’ litigation expenditures and the selection of cases brought to the courts. The evidence standard has different effects on different types of cases, reducing litigation for high-merit cases when standards are set low and increasing litigation...... for low-merit cases when standards are set high. This provides a valuable key for understanding the sorting effect of evidence standards and their role as a policy instrument in civil litigation....

  19. 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.)

  20. 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

  1. 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

  2. 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.)

  3. 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

  4. 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)

  5. Quality evaluation of the k0-standardized neutron activation analysis on Dalat research reactor

    International Nuclear Information System (INIS)

    Laboratory for neutron activation analysis (NAA) at the 500 kW Dalat Research Reactor( DRR) has been accredited following ISO/IEC 17025: 2005 (TCVN VILAS-519). Successful introduction of the k0-based NAA using Ko-Dalat software written in house at DRR has allowed to extend its applications in petroleum, archaeology and environment besides other traditional fields, i.e. geology, biomedicine, industry and materials. This study aimed to assess the quality of k0-NAA by analyzing a number of standard reference materials: SMELS, NIST-1547, NIST-2711a, IAEA-Soil-7 and IAEA-V-10. The laboratory has also participated in proficiency testing schemes organized by IAEA and FNCA. External and internal quality assessment revealed that the k0-NAA using Ko-Dalat software established at DRR has met the requirements of multi-element analysis in the intended applications. About 42 elements: Al, As, Au, Ba, Br, Ca, Ce, Cl, Co, Cr, Cs, Cu, Dy, Eu, Fe, Hf, I, In, K, La, Mg, Mn, Mo, Na, Nd, Pr, Rb, Sb, Sc, Se, Sm, Sr, Ta, Tb, Th, Ti, U, V, W, Yb, Zn and Zr, were determined in the above mentioned materials. The results were evaluated and reported in this paper. (author)

  6. 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

  7. Design and fabrication of HTTR reactor pressure vessel

    International Nuclear Information System (INIS)

    The High Temperature Engineering Test Reactor (HTTR) is under construction at the Oarai Research Establishment, JAERI and planned to be critical at the end of 1997. The HTTR is a High Temperature Gas-cooled Reactors (HTGRs) with thermal output of 30MW, inlet coolant temperature of 395degC, and outlet coolant temperature of 850degC at rated operation and 950degC at high temperature test operation. 2.25Cr-1Mo steel is chosen for the reactor pressure vessel of the HTTR because its temperature reaches about 400degC at normal operation. 2.25Cr-1Mo steel has higher creep rupture strength than Mn-Mo steel used for the reactor pressure vessels of Light Water Reactors (LWRs). For the components of the HTTR reactor pressure vessel subjected to low temperatures where creep deformation is negligible, a design guideline based on Japanese structural design standard for LWRs 'Technical standards for LWR power plant components-Ministry of International Trade and Industry Standard No.501' is utilized. On the other hand, design of the components for high temperature application, where creep behavior dominates, is conducted under newly determined high temperature structural design guideline and design material data. The fabrication of the HTTR reactor pressure vessel took about 23 months. It was installed in a reactor containment vessel in August, 1994. After core components had been installed in the reactor pressure vessel, pressure test of the primary and secondary cooling system including the reactor pressure vessel was performed and successfully ended in March, 1996. This paper reports issues of the HTTR reactor pressure vessel such as structure, material, stress analysis, fabrication, examination and testing. (author)

  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. 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)

  10. 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

  11. 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)

  12. 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...

  13. 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...

  14. 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.)

  15. CFD Simulation of Hydrodynamic Characteristics in Stirred Reactors Equipped with Standard Rushton or 45°-Upward PBT Impeller

    Institute of Scientific and Technical Information of China (English)

    未作君; 徐世民; 元英进; 许松林

    2003-01-01

    The hydrodynamic characteristics generated by the standard Rushton or 45°-upward pitched-blade-turbine (PBT) impellers in a baffled reactor are numerically simulated for different off-bottom clearances (C= 1/3H and 1/2H) and agitator speeds (100, 150, 200, 250 and 300r·min-1) by using FLUENT code (Version 5.4). The results are compared with the experimental and simulated data in the published papers and good agreement is observed. The shapes of the profile of mean velocities seem independent to the speed of agitators under the experimental conditions (100-300r·min-1).

  16. 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)

  17. 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.

  18. Standard Practice for Design of Surveillance Programs for Light-Water Moderated Nuclear Power Reactor Vessels

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2010-01-01

    1.1 This practice covers procedures for designing a surveillance program for monitoring the radiation-induced changes in the mechanical properties of ferritic materials in light-water moderated nuclear power reactor vessels. This practice includes the minimum requirements for the design of a surveillance program, selection of vessel material to be included, and the initial schedule for evaluation of materials. 1.2 This practice was developed for all light-water moderated nuclear power reactor vessels for which the predicted maximum fast neutron fluence (E > 1 MeV) at the end of license (EOL) exceeds 1 × 1021 neutrons/m2 (1 × 1017 n/cm2) at the inside surface of the reactor vessel. 1.3 This practice applies only to the planning and design of surveillance programs for reactor vessels designed and built after the effective date of this practice. Previous versions of Practice E185 apply to earlier reactor vessels. 1.4 This practice does not provide specific procedures for monitoring the radiation induced cha...

  19. 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.)

  20. The French Fast Reactor Programme: Innovations in Support of Higher Standards

    International Nuclear Information System (INIS)

    Deriving from the feedback of experience of former sodium cooled fast reactors, very high levels of requirements have been set for the ASTRID reactor (Advanced Sodium Technological Reactor for Industrial Demonstration). Innovations are needed to further enhance safety, reduce capital cost and improve efficiency, reliability and operability, making the Generation IV SFR an attractive option for electricity production. This puts great pressure not only on the R&D and design teams at CEA but also on the CEA’s R&D and industrial partners. During the first phase of the ASTRID conceptual design (2010–2012), promising innovative options have been identified. They will be further developed in the next phases of ASTRID design studies, keeping a strong interaction between design and R&D. (author)

  1. 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

  2. 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.

  3. 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.

  4. 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)

  5. 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

  6. Standard Guide for In-Service Annealing of Light-Water Moderated Nuclear Reactor Vessels

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2003-01-01

    1.1 This guide covers the general procedures to be considered for conducting an in-service thermal anneal of a light-water moderated nuclear reactor vessel and demonstrating the effectiveness of the procedure. The purpose of this in-service annealing (heat treatment) is to improve the mechanical properties, especially fracture toughness, of the reactor vessel materials previously degraded by neutron embrittlement. The improvement in mechanical properties generally is assessed using Charpy V-notch impact test results, or alternatively, fracture toughness test results or inferred toughness property changes from tensile, hardness, indentation, or other miniature specimen testing (1). 1.2 This guide is designed to accommodate the variable response of reactor-vessel materials in post-irradiation annealing at various temperatures and different time periods. Certain inherent limiting factors must be considered in developing an annealing procedure. These factors include system-design limitations; physical constrain...

  7. 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.

  8. 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.

  9. 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

  10. A comprehensive study on the thermal movement of the reactor coolant system for Korea Standard Nuclear Power Plant

    International Nuclear Information System (INIS)

    The specific thermal pattern of each component of the reactor coolant system (RCS) is a basic design parameter for the establishment of the RCS arrangement and support system. Development of an effective method to analyze the global behavior of thermal movement of the RCS is required for design process. In this paper a fundamental but reliable and comprehensive method of obtaining thermal movement of the Korea Standard Nuclear Power Plant (KSNPP) is presented. Experimental results are included to illustrate the validity of the analyses

  11. 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)

  12. 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.

  13. 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

  14. 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.

  15. 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.

  16. 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)

  17. Recent activities on nuclear codes and standards

    International Nuclear Information System (INIS)

    The technical codes and standards relating to the nuclear power stations in Japan are prepared by shapes of laws (ministerial ordinances and bulletins) issued by the government and obliged to comply with by 'the Law concerning the Regulations of Nuclear Material Substances, Nuclear Fuel Substances and Nuclear Reactors' and 'the Electricity Business Act' and of guides defined by the Nuclear Safety Commission, and further some private standards have been issued at a shape of complement of these laws and guides by receiving national recommendation. On the other hand, in the fields of electricity and heat facilities except atomic energy, simplification and feature stipulation of the national technical codes and standards was recently carried out, by which a system usable for the private standards in and out of Japan were prepared through approval of the private Japan Electrotechnical Standards and Codes Committee (JESC). As the nuclear field was now excepted from simultaneous transfer to the private standard and the standard application system, it is expected in future to realize similar transfer if possible and preparation of the private standards is now being advanced. Here were introduced on present state on technical codes and standards relating to the nuclear power generation facilities and recent trends on their private standardization. (G.K.)

  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. Qualification of the reactor physics toolset for the design and analysis of the advanced CANDU reactor

    International Nuclear Information System (INIS)

    The qualification of reactor physics toolset for Advanced CANDU Reactor (ACR) applications is described in this paper. The qualification process follows AECL standard code validation methodology. The ACR nuclear design incorporates certain features that challenge the physics code-suite capabilities. The physics codes were first assessed, and development work required to meet these challenges was undertaken. A Validation Matrix Document was prepared to identify the physics phenomena that could arise during postulated accident events, and specify the experimental data required for code validation. Key issues related to physics modelling and code validation are also discussed. (author)

  20. Standards and quality

    CERN Document Server

    El-Tawil, Anwar

    2015-01-01

    The book brings together a number of subjects of prime importance for any practicing engineer and, students of engineering. The book explains the concepts and functions of voluntary standards, mandatory technical regulations, conformity assessment (testing and measurement of products), certification, quality and quality management systems as well as other management systems such as environmental, social responsibility and food safety management systems.The book also gives a comprehensive description of the role of metrology systems that underpin conformity assessment. A description is given of typical national systems of standards, quality and metrology and how they relate directly or through regional structures to international systems. The book also covers the relation between standards and trade and explains the context and stipulations of the Technical Barriers to Trade Agreement of the World Trade Organization (WTO).

  1. Formation Rate of Natural Gas Hydrate - Reactor Experiments and Models

    OpenAIRE

    Mork, Marit

    2002-01-01

    The rate of methane hydrate and natural gas hydrate formation was measured in a 9.5 litre stirred tank reactor of standard design. The experiments were performed to better understand the performance and scale-up of a reactor for continuous production of natural gas hydrates. The hydrate formation rate was measured at steady-state conditions at pressures between 70 and 90 bar and temperatures between 7 and 15 °C. Between 44 and 56 % of the gas continuously supplied to the reactor was converted...

  2. 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.)

  3. 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.

  4. 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

  5. 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.

  6. 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)

  7. 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

  8. Technical modifications and management innovations in exporting nuclear reactor projects

    International Nuclear Information System (INIS)

    As a main channel for the foreign economic cooperation of China nuclear industry, China Zhongyuan Engineering Corporation (CZEC) has been constantly engaged in technical modifications and management innovations in its exporting nuclear reactor projects. In the implementation of heavy water research reactor contract in Algeria, CZEC had established a complete and adequate design standards system in compliance with the international standards, and made significant modifications to the reference reactor in the aspects of reactor power and reactor safety, solved quite some technical issues which-affected the reactor technical performance. The modifications and improvements enabled the technical parameters, safety features, reactor multipurpose application to attain to the advanced level in the world. In the 300 MWe PWR NPPs in Pakistan, safety features had been updated in line with upgrading regulatory requisites. The design philosophy and technology application demonstrated CZEC' s creation and innovation on basis of constant safety enhancement of nuclear power projects. Efforts had also been made by CZEC' s creation and innovation on basis of constant safety enhancement of nuclear power projects. Efforts had also been made by CZEC in promoting China made equipment items and components exportation. (authors)

  9. Recent experience and new developments in reactor pressure vessel manufacture

    International Nuclear Information System (INIS)

    In this paper, Framatome's recent experience and new developments in the manufacture of pressurized water reactor (PWR) reactor pressure vessel (RPV)s is described to show the very high standards of quality achieved to meet the most stringent requirements. Outstanding new developments include: qualification and utilisation of thick forged shell rings made from large hollow ingots; fully automatic submerged arc narrow gap welding; electroslag stainless steel cladding process; nozzle buttering by automatic hotwire TIG process

  10. 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)

  11. A probabilistic safety assessment of the standard French 900MWe pressurized water reactor. Main report

    International Nuclear Information System (INIS)

    To situate the probabilistic safety assessment of standardized 900 MWe units made by the Institute for Nuclear Safety and Protection (IPSN), it is necessary to consider the importance and possible utilization of a study of this type. At the present time, the safety of nuclear installations essentially depends on the application of the defence in-depth approach. The design arrangements adopted are justified by the operating organization on the basis of deterministic studies of a limited number of conventional situations with corresponding safety margins. These conventional situations are grouped in categories by frequency, it being accepted that the greater the consequences the lesser the frequency must be. However in the framework of the analysis performed under the control of the French safety authority, the importance was rapidly recognized of setting an overall reference objective. By 1977, on the occasion of appraisal of the fundamental safety options of the standardized 1300 MWe units, the Central Service for the Safety of Nuclear Installations (SCSIN) set the following global probabilistic objective: 'Generally speaking, the design of installations including a pressurized water nuclear reactor must be such that the global probability of the nuclear unit being the origin of unacceptable consequences does not exceed 10-6 per year...' Probabilistic analyses making reference to this global objective gradually began to supplement the deterministic approach, both for examining external hazards to be considered in the design basis and for examining the possible need for additional means of countering the failure of doubled systems in application of the deterministic single-failure criterion. A new step has been taken in France by carrying out two level 1 probabilistic safety assessments (calculation of the annual probability of core meltdown), one for the 900 MWe series by the IPSN and the other for the 1300 MWe series by Electricite de France. The objective of

  12. Refurbishment and Modernisation of PUSPATI TRIGA Reactor and Lessons Learnt

    International Nuclear Information System (INIS)

    The PUSPATI TRIGA Reactor first became critical in June 1982, and has been in operation since then. Over the years, several of the reactor systems, structures and components (SSCs) experience ageing and obsolescence problems and had to be refurbished, replaced or modernised. Initially refurbishment or replacements were carried out with SSCs of equivalent quality or capability. Subsequently SSCs were replaced with higher specification to allow for future upgrading of the reactor. Features of new SSCs should include all features of SSCs to be replaced and consider human machine interface to avoid any incidents. Lessons learnt over the years have been applied to the reactor control console modernisation project. In this project the involvement of our personnel during the design, fabrication and testing stages will enable us to have the capability to solve any associated problems with minimal vendor involvement. The close cooperation between regulators of Malaysia and vendor country was also beneficial to ensure that the project meet international safety standards

  13. Application of coupled code technique to a safety analysis of a standard MTR research reactor

    International Nuclear Information System (INIS)

    Accident analyses in nuclear research reactors have been performed, up to now, using simple computational tools based on conservative physical models. These codes, developed to focus on specific phenomena in the reactor, were widely used for licensing purposes. Nowadays, the advances in computer technology make it possible to switch to a new generation of computational tools that provides more realistic description of the phenomena occurring in a nuclear research reactor. Recent International Atomic Energy Agency (IAEA) activities have emphasized the maturity in using Best Estimate (BE) Codes in the analysis of accidents in research reactors. Indeed, some assessments have already been performed using BE thermal-hydraulic system codes such as RELAP5/Mod3. The challenge today is oriented to the application of coupled code techniques for research reactors safety analyses. Within the framework of the current study, a Three-Dimensional Neutron Kinetics Thermal-Hydraulic Model (3D-NKTH) based on coupled PARCS and RELAP5/Mod3.3 codes has been developed for the IAEA High Enriched Uranium (HEU) benchmark core. The results of the steady state calculations are sketched by comparison to tabulated results issued from the IAEA TECDOC 643. These data were obtained using conventional diffusion codes as well as Monte Carlo codes. On the other hand, the transient analysis was assessed with conventional coupled point kinetics-thermal-hydraulic channel codes such as RELAP5 stand alone, RETRAC-PC, and PARET codes. Through this study, the applicability of the coupled code technique is emphasized with an outline of some remaining challenges.

  14. Application of coupled code technique to a safety analysis of a standard MTR research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hamidouche, Tewfik [Division de l' Environnement, de la Surete et des Dechets Radioactifs, Centre de Recherche Nucleaire d' Alger (CRNA), Alger (Algeria); Laboratoire de Mecanique des Fluides Theorique et Appliquee, Faculte de Physique, Universite Des Sciences et de la Technologie Houari Boumediene, (USTHB), Bab-Ezzouar, Alger (Algeria)], E-mail: t.hamidouche@crna.dz; Bousbia-Salah, Anis [Dipartimento di Ingegneria Meccanica, Nucleari e della Produzione-Facolta di Ingegneria, Universita di Pisa, Pisa (Italy)], E-mail: b.salah@ing.unipi.it; Si-Ahmed, El Khider [Laboratoire de Mecanique des Fluides Theorique et Appliquee, Faculte de Physique, Universite Des Sciences et de la Technologie Houari Boumediene, (USTHB), Bab-Ezzouar, Alger (Algeria)], E-mail: esi-ahmed@usthb.dz; Mokeddem, Mohamed Yazid [Division de la Physique et des Applications Nucleaires, Centre de Recherche Nucleaire de Draria (CRND) (Algeria); D' Auria, Franscesco [Dipartimento di Ingegneria Meccanica, Nucleari e della Produzione-Facolta di Ingegneria, Universita di Pisa, Pisa (Italy)

    2009-10-15

    Accident analyses in nuclear research reactors have been performed, up to now, using simple computational tools based on conservative physical models. These codes, developed to focus on specific phenomena in the reactor, were widely used for licensing purposes. Nowadays, the advances in computer technology make it possible to switch to a new generation of computational tools that provides more realistic description of the phenomena occurring in a nuclear research reactor. Recent International Atomic Energy Agency (IAEA) activities have emphasized the maturity in using Best Estimate (BE) Codes in the analysis of accidents in research reactors. Indeed, some assessments have already been performed using BE thermal-hydraulic system codes such as RELAP5/Mod3. The challenge today is oriented to the application of coupled code techniques for research reactors safety analyses. Within the framework of the current study, a Three-Dimensional Neutron Kinetics Thermal-Hydraulic Model (3D-NKTH) based on coupled PARCS and RELAP5/Mod3.3 codes has been developed for the IAEA High Enriched Uranium (HEU) benchmark core. The results of the steady state calculations are sketched by comparison to tabulated results issued from the IAEA TECDOC 643. These data were obtained using conventional diffusion codes as well as Monte Carlo codes. On the other hand, the transient analysis was assessed with conventional coupled point kinetics-thermal-hydraulic channel codes such as RELAP5 stand alone, RETRAC-PC, and PARET codes. Through this study, the applicability of the coupled code technique is emphasized with an outline of some remaining challenges.

  15. 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)

  16. 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)

  17. Learning and nuclear safety: New reactors and US regulation

    International Nuclear Information System (INIS)

    Gathering and analyzing data from operating reactors has become part of government and industry programs to improve performance in plants already on line and to inform development of future reactors. In the United States, however, early development and certain other factors combined to encourage a bias in learning. Regulation and learning from operational data intersect in ways that limit participation, data collection, and positive response to findings. Past learning has shown the advantage of simpler more standard designs with passive or inherent safety features. However, even designs incorporating these past lessons are apt to face tough regulatory tests and much criticism as operating experience is gathered. Only the operational success of new standardized reactors is apt to help rationalize regulation. (orig.)

  18. 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

  19. 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

  20. 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.)

  1. 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

  2. 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

  3. 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

  4. 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)

  5. The French Fast Reactor Program - Innovations in Support to Higher Standards

    International Nuclear Information System (INIS)

    • From the experience of ASTRID first phase of conceptual design studies (2010-2012), two remarks can be made: → Higher requirements in safety and operability lead to higher costs that cannot be fully recovered by advances in technology. This puts additional pressure on the next phases of the design to optimize the design and to keep the costs to the minimum. → There is a clear link between the level of safety that can be achieved and the maturity of the technology, i.e. the experience accumulated in R&D, design, construction, operation and decommissioning of past reactors. In the field of fast neutron reactors, this gives a strong advantage to the sodium technology, because strengths and weaknesses are well mastered. • Meeting the high requirements set for ASTRID and serving R&D needs of innovative options will require increased industrial and international collaboration

  6. 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.

  7. Design of standardized WWER-1000 reactor power plant allowing industrialization of production

    International Nuclear Information System (INIS)

    The improvement consists in the siting of the individual units which allows streamlined construction, the assembly of power units and improved quality of construction work. To protect it against vibrations the reactor building is designed as a symmetric box-shaped reinforced concrete structure. The heaviest equipment is placed in the lowest parts, which increases stability and facilitates the solution of the problem of the interaction of foundations and structure proper. The cylindrical part of the sealed envelope of the reactor part may be assembled of large units up to 100 t in weight and another design of the envelope copula allows the assembly of basic equipment to be started 3 to 4 months earlier. These and other improvements make it possible to shorten construction time by 35 to 40%, to reduce material consumption and to increase productivity. (E.S.)

  8. 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

  9. Islam, Standards, and Technoscience

    DEFF Research Database (Denmark)

    Fischer, Johan

    Halal (literally, "permissible" or "lawful") production, trade, and standards have become essential to state-regulated Islam and to companies in contemporary Malaysia and Singapore, giving these two countries a special position in the rapidly expanding global market for halal products: in these n......Halal (literally, "permissible" or "lawful") production, trade, and standards have become essential to state-regulated Islam and to companies in contemporary Malaysia and Singapore, giving these two countries a special position in the rapidly expanding global market for halal products......: in these nations state bodies certify halal products as well as spaces (shops, factories, and restaurants) and work processes, and so consumers can find state halal-certified products from Malaysia and Singapore in shops around the world. Building on ethnographic material from Malaysia, Singapore, and Europe......, this book provides an exploration of the role of halal production, trade, and standards. Fischer explains how the global markets for halal comprise divergent zones in which Islam, markets, regulatory institutions, and technoscience interact and diverge. Focusing on the "bigger institutional picture...

  10. AGR-5/6/7 LEUCO Kernel Fabrication Readiness Review

    Energy Technology Data Exchange (ETDEWEB)

    Marshall, Douglas W. [Idaho National Lab. (INL), Idaho Falls, ID (United States). Fuel Design and Development; Bailey, Kirk W. [Idaho National Lab. (INL), Idaho Falls, ID (United States). ART Quality Assurance Engineer

    2015-02-01

    In preparation for forming low-enriched uranium carbide/oxide (LEUCO) fuel kernels for the Advanced Gas Reactor (AGR) fuel development and qualification program, Idaho National Laboratory conducted an operational readiness review of the Babcock & Wilcox Nuclear Operations Group – Lynchburg (B&W NOG-L) procedures, processes, and equipment from January 14 – January 16, 2015. The readiness review focused on requirements taken from the American Society Mechanical Engineers (ASME) Nuclear Quality Assurance Standard (NQA-1-2008, 1a-2009), a recent occurrence at the B&W NOG-L facility related to preparation of acid-deficient uranyl nitrate solution (ADUN), and a relook at concerns noted in a previous review. Topic areas open for the review were communicated to B&W NOG-L in advance of the on-site visit to facilitate the collection of objective evidences attesting to the state of readiness.

  11. Specific Considerations and Milestones for a Research Reactor Project

    International Nuclear Information System (INIS)

    One of the IAEA's statutory objectives is to 'seek to accelerate and enlarge the contribution of atomic energy to peace, health and prosperity throughout the world'. One way this objective is achieved is through the publication of a range of technical series. Two of these are the IAEA Nuclear Energy Series and the IAEA Safety Standards Series. According to Article III.A.6 of the IAEA Statute, the safety standards establish 'standards of safety for protection of health and minimization of danger to life and property.' The safety standards include the Safety Fundamentals, Safety Requirements and Safety Guides. These standards are written primarily in a regulatory style, and are binding on the IAEA for its own programmes. The principal users are the regulatory bodies in Member States and other national authorities. The IAEA Nuclear Energy Series comprises reports designed to encourage and assist R and D on, and application of, nuclear energy for peaceful uses. This includes practical examples to be used by owners and operators of utilities in Member States, implementing organizations, academia, and government officials, among others. This information is presented in guides, reports on technology status and advances, and best practices for peaceful uses of nuclear energy based on inputs from international experts. The IAEA Nuclear Energy Series complements the IAEA Safety Standards Series. Services provided by research reactors can impact every aspect of social and community development. Research reactors can contribute to a country's scientific and educational resources, raise living standards through improved health care and industrial and agricultural productivity, or pave the way to the utilization of nuclear energy. However, the decision to construct a new research reactor requires national recognition of international responsibilities and the implementation of essential policy and technical infrastructure. If appropriately conceived, managed and organized, a

  12. Standard model and beyond

    International Nuclear Information System (INIS)

    The SU(3)/sub c/ circle crossSU(2)/sub L/circle crossU(1)/sub Y/ gauge theory of ineractions among quarks and leptons is briefly described, and some recent notable successes of the theory are mentioned. Some shortcomings in our ability to apply the theory are noted, and the incompleteness of the standard model is exhibited. Experimental hints that Nature may be richer in structure than the minimal theory are discussed. 23 references

  13. 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

  14. 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.

  15. Frequency Standards and Metrology

    Science.gov (United States)

    Maleki, Lute

    2009-04-01

    Preface / Lute Maleki -- Symposium history / Jacques Vanier -- Symposium photos -- pt. I. Fundamental physics. Variation of fundamental constants from the big bang to atomic clocks: theory and observations (Invited) / V. V. Flambaum and J. C. Berengut. Alpha-dot or not: comparison of two single atom optical clocks (Invited) / T. Rosenband ... [et al.]. Variation of the fine-structure constant and laser cooling of atomic dysprosium (Invited) / N. A. Leefer ... [et al.]. Measurement of short range forces using cold atoms (Invited) / F. Pereira Dos Santos ... [et al.]. Atom interferometry experiments in fundamental physics (Invited) / S. W. Chiow ... [et al.]. Space science applications of frequency standards and metrology (Invited) / M. Tinto -- pt. II. Frequency & metrology. Quantum metrology with lattice-confined ultracold Sr atoms (Invited) / A. D. Ludlow ... [et al.]. LNE-SYRTE clock ensemble: new [symbol]Rb hyperfine frequency measurement - spectroscopy of [symbol]Hg optical clock transition (Invited) / M. Petersen ... [et al.]. Precise measurements of S-wave scattering phase shifts with a juggling atomic clock (Invited) / S. Gensemer ... [et al.]. Absolute frequency measurement of the [symbol] clock transition (Invited) / M. Chwalla ... [et al.]. The semiclassical stochastic-field/atom interaction problem (Invited) / J. Camparo. Phase and frequency noise metrology (Invited) / E. Rubiola ... [et al.]. Optical spectroscopy of atomic hydrogen for an improved determination of the Rydberg constant / J. L. Flowers ... [et al.] -- pt. III. Clock applications in space. Recent progress on the ACES mission (Invited) / L. Cacciapuoti and C. Salomon. The SAGAS mission (Invited) / P. Wolf. Small mercury microwave ion clock for navigation and radioScience (Invited) / J. D. Prestage ... [et al.]. Astro-comb: revolutionizing precision spectroscopy in astrophysics (Invited) / C. E. Kramer ... [et al.]. High frequency very long baseline interferometry: frequency standards and

  16. 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

  17. 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.)

  18. LVR-15 reactor performance and transformation to low enriched fuel

    International Nuclear Information System (INIS)

    Experimental research reactor LVR-15 situated in Nuclear Research Institute Rez, plc. has been utilized since 1957. The present reactor nominal power is 10 MW. Standard reactor cycle is 21 days and the reactor operates 8 - 10 cycles per year. The State Office for Nuclear Safety newly licensed the reactor till 2014. The reactor is of multi-purpose use. The basic research is carried out using horizontal neutron beams and one of them is used for development and application of the boron neutron captures therapy for brain tumors. Mostly material testing of PWR and BWR specimens is performed in high-pressure loops and irradiation rigs operated in the reactor. Several vertical channels serve for production of neutron transmutation doped silicon and isotopes production for medical purposes. Reactor was originally designed with EK-10 fuel type and consequently reconstructed for HEU with the 80% 235U enrichment in 1987-89. In period 1987-1989 the reactor undertook the second reconstruction to enhance its experimental and commercial utilizing. Later, the fuel enrichment was changed to 36 % with the use of the IRT-2M fuel type. The neutron-physical characteristics of the reactor core and fuel cycle design and analyses are carried out using the WIMS-D4m, NODER and OMEGA programs. The codes were used for a preliminary evaluation of essential changes of main neutronic characteristics of the LVR-15 core with the prospective conversion to low enriched fuel (LEU). Three types of FA-s has been assessed: 1) currently used IRT-2M(36%), 2) IRT-3M(20%), 3) IRT-4M(20%) and results are presented in the article. (author)

  19. 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

  20. Considerations and Infrastructure Milestones for a Research Reactor Project

    International Nuclear Information System (INIS)

    Establishment of a research reactor is a major project requiring careful planning, preparation, implementation, and investment in time and human resources. The implementation of such a project requires establishment of sustainable infrastructures, including legal and regulatory, safety, technical, and economic. This paper discusses the scope of these infrastructures and the major stages in their development; starting with a robust pre-project justification for the research reactor and moving through three milestones in the establishment of the infrastructure itself. The paper discusses also the main elements of the feasibility study for a new research reactor project and specific safety and technical considerations in different phases of the project as well as the major activities to be performed along with the project phases, including progressive involvement of the main organizations in the project, and application of the IAEA Code of Conduct on the Safety of Research Reactors and IAEA Safety Standards. (author)

  1. 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

  2. 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

  3. On the classification of structures, systems and components of nuclear research and test reactors

    International Nuclear Information System (INIS)

    The classification of structures, systems and components of nuclear reactors is a relevant issue related to their design because it is directly associated with their safety functions. There is an important statement regarding quality standards and records that says Structures, systems, and components important to safety shall be designed, fabricated, erected, and tested to quality standards commensurate with the importance of the safety functions to be performed. The definition of the codes, standards and technical requirements applied to the nuclear reactor design, fabrication, inspection and tests may be seen as the main result from this statement. There are well established guides to classify structures, systems and components for nuclear power reactors such as the Pressurized Water Reactors but one can not say the same for nuclear research and test reactors. The nuclear reactors safety functions are those required to the safe reactor operation, the safe reactor shutdown and continued safe conditions, the response to anticipated transients, the response to potential accidents and the control of radioactive material. So, it is proposed in this paper an approach to develop the classification of structures, systems and components of these reactors based on their intended safety functions in order to define the applicable set of codes, standards and technical requirements. (author)

  4. The reference neutron field - a standard neutron source for neutron measurements at the research reactor IRT-2000 in Sofia

    International Nuclear Information System (INIS)

    A reference neutron field (RFN) is used as a standard neutron source (SNS) that is influenced by the changes in the reactor core due to recharging or other causes. A whole range of measurements is carried out in a full scope, to specify its characteristics precisely. The SNS comprises: 1) the RNF certificated to the neutron energy spectrum, its location in the reactor field, being a reference measure of the differential energy distribution in the neutron flux; 2) exposure monitoring tools (detectors revealing the certified physical characteristics); 3) functional measurement apparatus (revealing the spectral characteristics). The following basic metrological characteristics are given: differential neutron energy spectrum, described by F(E) [1/cm2.s.MeV], normalized by 1 in the range 3-19 MeV and the measurement error; the conventional neutron flux density and its error. The methodology of measuring the neutron flux integral density comprises the following six steps: 1) assessment of the influence of the changes in the core configuration on the stability of the RNF (estimated in six energy ranges); 2) demonstration of RNF application in reactor physics studies; 3) irradiation of two sets of activation detectors (Au, Sc and Au, Sc, S in Al and Cd shields); 4) measurement of the detector activities by calibrated gamma- and beta- spectrometric apparatus; 5) determination of the neutron field characteristics at a certain point of the RNF by the method of activating ratios; 6) the result accuracy assessment and probabilistic error limits determination with 95% upper bound frequency. The RNF neutron energy range have been measured 6 times for a period of two years. 6 refs., 8 figs. (M.A.)

  5. Equipment and piping for nuclear power plants, test and research reactors, and nuclear installations

    International Nuclear Information System (INIS)

    The standard concerns the primary and secondary circuits as well as the safety and protection equipment in nuclear power plants with PWR or LWGR type reactors. Rules for design, manufacturing, erection, operation, and maintenance of the reactors, steam generators, vessels, pumps and housings, and pressure pipes are provided

  6. Radiological shielding of low power compact reactor: calculation and design

    International Nuclear Information System (INIS)

    The development of compact reactors becoming a technology that offers great projection and innumerable use possibilities, both in electricity generation and in propulsion.One of the requirements for the operation of this type of reactor is that it must include a radiological shield that will allow for different types of configurations and that, may be moved with the reactor if it needs to be transported.The nucleus of a reactor emits radiation, mainly neutrons and gamma rays in the heat of power, and gamma radiation during the radioactive decay of fission products.This radiation must be restrained in both conditions of operation to avoid it affecting workers or the public.The combination of different materials and properties in layers results in better performance in the form of a decrease in radiation, hence causing the dosage outside the reactor, whether in operation or shut down, to fall within the allowed limits.The calculations and design of radiological shields is therefore of paramount importance in reactor design.The choice of material and the design of the shield have a strong impact on the cost and the load capacity, the latter being one of the characteristics to optimize.The imposed condition of design is that the reactor can be transported together with the decay shield in a standard container of 40 foot

  7. Design Features and Operating Experience of Experimental Fast Reactors

    International Nuclear Information System (INIS)

    One of the IAEA's statutory objectives is to 'seek to accelerate and enlarge the contribution of atomic energy to peace, health and prosperity throughout the world'. One way this objective is achieved is through the publication of a range of technical series. Two of these are the IAEA Nuclear Energy Series and the IAEA Safety Standards Series. According to Article III.A.6 of the IAEA Statute, the safety standards establish 'standards of safety for protection of health and minimization of danger to life and property'. The safety standards include the Safety Fundamentals, Safety Requirements and Safety Guides. These standards are written primarily in a regulatory style, and are binding on the IAEA for its own programmes. The principal users are the regulatory bodies in Member States and other national authorities. The IAEA Nuclear Energy Series comprises reports designed to encourage and assist R and D on, and application of, nuclear energy for peaceful uses. This includes practical examples to be used by owners and operators of utilities in Member States, implementing organizations, academia, and government officials, among others. This information is presented in guides, reports on technology status and advances, and best practices for peaceful uses of nuclear energy based on inputs from international experts. The IAEA Nuclear Energy Series complements the IAEA Safety Standards Series. The IAEA has begun an initiative to help coordinate Member State efforts in the field of fast neutron nuclear reactors. This initiative is primarily targeted at the preservation of knowledge in the areas of design, construction and operation, for both experimental and power fast reactors. The ultimate goal of this activity is to establish a comprehensive, international inventory of fast reactor data and knowledge, which will be an essential resource for the future development and deployment of fast reactor technology. In this project, carried out within the framework of the

  8. 75 FR 68009 - Office of New Reactors; Notice of Availability of the Final Staff Guidance Standard Review Plan...

    Science.gov (United States)

    2010-11-04

    ... COMMISSION Office of New Reactors; Notice of Availability of the Final Staff Guidance Standard Review Plan Section 13.6.2, Revision 1 on Physical Security--Design Certification AGENCY: Nuclear Regulatory Commission (NRC). ACTION: Notice of Availability. SUMMARY: The NRC is issuing its Final Revision 1 to...

  9. Oconee 1, cycle 5 design report

    International Nuclear Information System (INIS)

    The Oconee 1, cycle 5 fuel cycle was designed to irradiate five fuel assemblies to a burnup of approximately 40,000 MWd/mtU. The fuel cycle design was performed using standard Babcock and Wilcox calculational techniques for nuclear fuel cycles. All applicable design criteria were satisfied. Valuable experimental data on the performance characteristics of high-burnup fuel assemblies will be obtained from these assemblies in subsequent post-irradiation examinations

  10. 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)

  11. Simulation of primary to secondary break in a VVER-type reactor: Results of the IAEA's Third Standard Problem Exercise

    International Nuclear Information System (INIS)

    Seeking to enlarge the experimental data base for code assessment, the International Atomic Energy Agency (IAEA), in collaboration with the Central Research Institute for Physics of the Hungarian Academy of Sciences, has organized the Third Standard Problem Exercise (SPE-3) involving the simulation of a break from primary to secondary in the steam generator of the PMK-NVH experimental facility. The facility is a scaled-down model of a VVER-type reactor, and the experiment addresses the possibility of a break developing in the steam generator collector of the actual plant. This paper presents a brief description of the experimental facility and the experiment. Results of a comparison of pretest and posttest calculations performed by some of the 24 participants in the exercise are also presented. The complete report of the exercise has been published as an IAEA technical document

  12. 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)

  13. Designs characteristics, and development of fast reactors for utilization of thorium

    International Nuclear Information System (INIS)

    Fast breeder reactors will be necessary in the next century in order to meet increasing demands for electricity resulting from industrialization and general improvement of standards of living. A scheme for a smooth development of liquid metal fast breeder reactors in Brazil is proposed and designs and characteristics of required reactors are discussed. Emphasis is placed on utilization of thorium that is abundant in the country, on reactor safety in order to promote public acceptance and smoothness of the development. The initial step is the construction of a 5 MW experimental reactor in order to acquire basic experiences and technologies. The second step is the construction of a series of small power reactors designed with particular emphasis on safety and ease of operation. In the final phase when fast breeder reactors are to play a central role in electricity generation, large power reactors that utilize both uranium and thorium fuel cycles will be built to establish a practically permanent power system. (Author)

  14. Developing research reactor coalitions and centres of excellence

    International Nuclear Information System (INIS)

    could otherwise result in a coordinated approach to market development, building upon strengths of various facilities. Moreover, belief that the markets for research reactor products and services are a 'zero-sum' game, with market gains by one research reactor coming at the expense of another facility, result in a general lack of openness within the research reactor community. et there is evidence to suggest that the market for research reactor services is supply limited, rather than demand limited. A number of factors limit the ability of research reactors to expand their user base and to generate new sources of revenue: - Many potential customers do not know how, or where, to contact the research reactor community, and have only limited knowledge or awareness of the range of research reactor services, equipment and locations available. - The standards of quality control and quality assurance between research reactors are not uniform, impede business development, and may result in a lack of confidence in service reliability. As a consequence, customers need to conduct due diligence for each facility to be used, reducing the enthusiasm and financial rationale for developing additional sources of supply. - Transport of radionuclides is becoming increasingly difficult, with examples of shipments held in customs, prevented from leaving the country of origin or from entering the customer destination, and requires specific expertise and experience to manage this issue. In order to address the complex of issues related to sustainability, security, and non-proliferation aspects of research reactors, and to promote international and regional cooperation, the IAEA is initiating the Research Reactor Coalitions and Centres of Excellence initiative. This activity is supported by a two-year grant from the Nuclear Threat Initiative, Inc. (NTI), and by a 2007-2008 IAEA Technical Cooperation Project, 'Enhancement of the Sustainability of Research Reactors and their Safe Operation

  15. Emergency Management Standards and Schools

    Science.gov (United States)

    National Clearinghouse for Educational Facilities, 2009

    2009-01-01

    This publication discusses emergency management standards for school use and lists standards recommended by FEMA's National Incident Management System (NIMS). Schools are encouraged to review these standards carefully and to adopt, where applicable, those that meet their needs. The lists of standards, resources, and references contained herein…

  16. Standard for prevention of gas entrainment phenomena in fast reactors. (2) Proposal of gas entrainment evaluation method

    International Nuclear Information System (INIS)

    For the stable operation of fast breeder reactors (FBRs), the occurrences of gas entrainment (GE) phenomena should be suppressed below an allowance level. Therefore, a reliable evaluation method for the GE phenomena is necessary to determine the operating conditions of FBRs. However, such a method has not yet been established, especially for the vortex-type GE. In this paper, the authors propose a GE evaluation method in which free surface vortices are identified from velocity fields by using the second invariant of the velocity gradient tensor, and GE evaluation parameters, e.g., gas core length, are calculated by using the Burgers vortex model. In addition, the standard method for the prevention of three kinds of vortex-type GE is shown by considering experimental data, evaluation results obtained by the proposed method and comparison results. Finally, it is confirmed that the onset conditions of the vortex-type GE can be evaluated by the proposed method. (author)

  17. 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)

  18. Reliability Analysis of I and C Architecture of Research Reactors Using Bayesian Networks

    International Nuclear Information System (INIS)

    The objective of this research project is to identify a configuration of architecture which gives highest availability with maintaining low cost of manufacturing. In this regard, two configurations of a single channel of RPS are formulated in the current article and BN models were constructed. Bayesian network analysis was performed to find the reliability features. This is a continuation of study towards the standardization of I and C architecture for low and medium power research reactors. This research is the continuation of study to analyze the reliability of single channel of Reactor Protection System (RPS) using Bayesian networks. The focus of research was on the development of architecture for low power research reactors. What level of reliability is sufficient for protection, safety and control systems in case of low power research reactors? There should be a level which should satisfy all the regulatory requirements as well as operational demands with optimized cost of construction. Scholars, researchers and material investigators from educational and research institutes are demanding for construction of more research reactors. In order to meet this demand and construct more units, it is necessary to do more research in various areas. The research is also needed to make a standardization of research reactor I and C architectures on the same lines of commercial power plants. The research reactors are categorized into two broad categories, Low power research reactors and medium to high power research reactors. According to IAEA TECDOC-1234, Research reactors with 0.250-2.0 MW power rating or 2.5-10 Χ 1011 n/cm2.s. flux are termed low power reactor whereas research reactors ranging from 2-10 MW power rating or 0.1-10 Χ 1013 n/cm2.s. are considered as Medium to High power research reactors. Some other standards (IAEA NP-T-5.1) define multipurpose research reactor ranging from power few hundred KW to 10 MW as low power research reactor

  19. 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.

  20. 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

  1. 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...

  2. Recycle Strategies for Fast Reactors and Related Fuel Cycle Technologies

    International Nuclear Information System (INIS)

    Fast reactors and related fuel cycle (hereafter referred to as 'fast reactor cycle') technologies have the potential to contribute to long term energy security owing to their effective use of uranium and plutonium resources, and to a reduction in the heat generation and potential toxicity of high level radioactive wastes by burning long lived minor actinides recovered from spent fuel from light water reactors and fast reactors. Further, it is likely that fast reactor cycle technologies can play a certain role in non-proliferation as addressed in the Global Nuclear Energy Partnership. With these features, the research and development towards their commercialization has been promoted vigorously and globally as a future vision of nuclear energy. The introduction of fast reactor cycle systems will be carried out independently in each country according to its national conditions and nuclear energy policy. It should then be considered important to have a globally common consensus relating to safety philosophy, concepts of proliferation resistance, transuranic element burnup and recycling and so on. For the development and utilization of fast reactor cycle systems, while respecting each country's concept, it is essential to organize the technologies and concepts which countires should have in common globally and build a framework to make them standardized. The use of existing frameworks such as the Generation IV International Forum and the International Project on Innovative Nuclear Reactors and Fuel Cycles is considered effective to achieving this. Furthermore, a vigorous promotion such as international cooperative developments enables the formation of international consensus on major technologies for the fast reactor cycle as well as the saving of resources by infrastructure sharing. (author)

  3. Standards for bullets and casings

    OpenAIRE

    J.F Song; T.V Vorburger; Clary, R.; E Whitenton; Ma, L.; S Ballou

    2002-01-01

    The National Institute of Standards and Technology is developing reference standards through its Office of Law Enforcement Standards with funding provided by the National Institute of Justice. The standard reference materials are used by crime laboratories to verify that results obtained when using their protocols and methodologies meet legal requirements and that equipment is operating properly. The NIST Reference Materials 8240/8250 standard bullets and casings is an example of materials th...

  4. 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.

  5. Status of the advanced boiling water reactor and simplified boiling water reactor

    International Nuclear Information System (INIS)

    This paper reports that the excess of U.S. electrical generating capacity which has existed for the past 15 years is coming to an end as we enter the 1990s. Environmental and energy security issues associated with fossil fuels are kindling renewed interest in the nuclear option. The importance of these issues are underscored by the National Energy Strategy (NES) which calls for actions which are designed to ensure that the nuclear power option is available to utilities. Utilities, utility associations, and nuclear suppliers, under the leadership of the Nuclear Power Oversight Committee (NPOC), have jointly developed a 14 point strategic plan aimed at establishing a predictable regulatory environment, standardized and pre-licensed Advanced Light Water Reactor (ALWR) nuclear plants, resolving the long-term waste management issue, and other enabling conditions. GE is participating in this national effort and GE's family of advanced nuclear power plants feature two new reactor designs, developed on a common technology base, aimed at providing a new generation of nuclear plants to provide safe, clean, economical electricity to the world's utilities in the 1990s and beyond. Together, the large-size (1300 MWe) Advanced Boiling Water Reactor (ABWR) and the small-size (600 MWe) Simplified Boiling Water Reactor (SBWR) are innovative, near-term candidates for expanding electrical generating capacity in the U.S. and worldwide. Both possess the features necessary to do so safely, reliably, and economically

  6. Standard curves and formulae for neutron kinetics calculations

    International Nuclear Information System (INIS)

    The response of the neutron kinetic equations to a wide range of step and ramp additions of reactivity has been evaluated on the PACE 231R analogue computer for two fuels, U235 and Pu239, with a full range of neutron lifetimes. The results are presented in the form of standard curves which may be readily used to assess the 'zero-energy' performance of a reactor at the early stages of a reactor concept. Appendices contain the derivation of several useful expressions associated with neutron kinetics calculations and demonstrate the use of the curves to estimate reactor behaviour during shut-down following trip action. (author)

  7. RA Research reactor Annual report 1982 - Part 1, Operation, maintenance and utilization of the RA reactor

    International Nuclear Information System (INIS)

    Reactor test operation started in September 1981 at 2 MW power with 80% enriched fuel continued during 1982 according to the previous plan. The initial reactor core was made of 44 fuel channel each containing 10 fuel slugs. The first half of 1982 was used for the needed measurements and analysis of operating parameters and functioning of reactor systems and equipment under operating conditions. Program concerned with the testing operation at higher power levels was started in the second half of this year. It was found that the inherent excess reactivity and control rod worths ensure safe operation according to the IAEA safety standards. Excess reactivity is high enough to enable higher power level of 4.7 MW during 4 monthly cycles each lasting 15-20 days. Favourable conditions for cooling exist for the initial core configuration. Effects of poisoning at startup on the reactivity and power density distribution were measured as well as initial spatial distribution of the neutron flux which was 3,9 1013 cm-2 s-1 at 2 MW power. Modification of the calibration coefficient in the system for automated power level control was determined. All the results show that all the safety criteria and limitations concerned with fuel utilization are fulfilled if reactor power would be 4.7 MW. Additional testing operation at 3, 4, and 4.7 MW power levels will be needed after obtaining the licence for operating at nominal power. Transition from the initial core with 44 fuel channels to the equilibrium lattice configuration with 72 fuel channels each containing 10 fuel slugs, would be done gradually. Reactor was not operated in September because of the secondary coolant pipes were exchanged between Danube and the horizontal sedimentary. Control and maintenance of the reactor equipment was done regularly and efficiently dependent on the availability of the spare parts. Difficulties in maintenance of the reactor instrumentation were caused by unavailability of the outdated spare parts

  8. 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

  9. 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.

  10. 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

  11. 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

  12. 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

  13. 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.

  14. 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

  15. 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

  16. 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

  17. 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

  18. Nuclear reactor safety and Federal regulation

    International Nuclear Information System (INIS)

    Public confidence in nuclear reactors requires that technical people translate complex safety information into a form that the public can understand well enough to make a judgment. An overall picture is drawn of the major areas of concern: (1) risks and safety measures, (2) government regulation, (3) licensing, (4) plant operation, (5) safety experience, and (6) quality assurance. Although the possibilities of a reactor core melting through the concrete containment barrier are slight, rigorous safety efforts are required. Government regulation and technical developments have developed concurrently so that the high standards set for government facilities can be carried over to commercial efforts. There are two stages in the licensing procedure: a construction permit and an operating license. Reviews of the proposed site, design, emergency cooling systems are all held, followed by a public hearing. Inspection and backfitting of new safety equipment are required in operating plants. The 60 plants now in operation have a good performance record, but good management for quality assurance increases safety and efficiency factors

  19. 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

  20. Standards and measurement units

    International Nuclear Information System (INIS)

    Today, the measurement is a permanent component of our environment, of our training and of the life of our companies. However, any decision taken from a measurement requires a proper interpretation of this measurement. In particular, it is necessary to know the reference system used and the uncertainty linked with the measurement. This book, published by the French national office of metrology (BNM), the official organization in charge of the implementation of the national metrology policy, must be considered as the basic document for any people from the industry, the teaching and the research world concerned by measurement problems. This book is devoted to the definition of the basic measurement units, to the units derived from the international unit system, and to the realization of the specific instrumentation necessary for the materialization of these units. This instrumentation of very high technical quality is based on multiple physical principles. It presents also the national reference standards. Beside the basic units of time and frequencies, of mechanics, of electricity and magnetism, of temperature, of radiometry and photometry and of mass quantity, one chapter is devoted to ionizing radiations and to their derived units: the Becquerel (Bq), the Gray (Gy) and the Sievert (Sv). (J.S.)

  1. ESDIS Standards Office (ESO): Requirements, Standards and Practices

    Science.gov (United States)

    Mitchell, Andrew E.; Mcinerney, Mark Allen; Enloe, Yonsok K.; Conover, Helen T.; Doyle, Allan

    2016-01-01

    The ESDIS Standards Office assists the ESDIS Project in formulating standards policy for NASA Earth Science Data Systems (ESDS), coordinates standards activities within ESDIS, and provides technical expertise and assistance with standards related tasks within the NASA Earth Science Data System Working Groups (ESDSWG). This poster summarizes information found on the earthdata.nasa.gov site that describes the ESO.

  2. 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)

  3. 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

  4. Recycle strategies for fast reactors and related fuel cycle technologies

    International Nuclear Information System (INIS)

    countries and one international organization from both technology holders and technology users countries, has been improving INPRO methodology and developing institutions for introducing innovative nuclear systems including fast reactor cycle in the future. GNEP, a voluntary international partnership, aims to expand clean, sustainable nuclear power worldwide in a safe and secure manner while responsibly managing nuclear waste and reducing proliferation risks, and one of its objectives is to develop, demonstrate, and in due course deploy advanced fast reactors that consume transuranic elements recovered from nuclear spent fuel. Such multilateral cooperation has now been promoted and received achievements in each framework. If these frameworks of multilateral cooperation can be unified to a manageable level in the future and further developed, it will be expected that fast reactor cycle technologies can be available on a global scale. 4. Conclusion For the development and utilization of fast reactor cycle systems, while respecting each country's concept, it is essential to organize the technologies and concepts which should be in common globally and build a framework to make it standardizing. Utilization of existing frameworks such as GIF and INPRO are considered as effective to realize it. Furthermore, a vigorous promotion such as international cooperative developments enables formation of international consensus on major technologies for fast reactor cycle as well as saving of resources by infrastructure sharing. Japan, as a non-nuclear-weapon nation, considers that we should play a great role in studying and creating a system where fast reactor cycle technologies can be used in peace and we would like to actively contribute to the international community. Finally, it is firmly confirmed that Japan will make a great effort to enable the global contribution as one of the few nations who have both experimental and prototype fast

  5. Media and Australia's replacement reactor project

    International Nuclear Information System (INIS)

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

  6. Mirror fusion reactors

    International Nuclear Information System (INIS)

    Conceptual design studies were made of fusion reactors based on the three current mirror-confinement concepts: the standard mirror, the tandem mirror, and the field-reversed mirror. Recent studies of the standard mirror have emphasized its potential as a fusion-fission hybrid reactor, designed to produce fuel for fission reactors. We have designed a large commercial hybrid and a small pilot-plant hybrid based on standard mirror confinement. Tandem mirror designs include a commercial 1000-MWe fusion power plant and a nearer term tandem mirror hybrid. Field-reversed mirror designs include a multicell commercial reactor producing 75 MWe and a single-cell pilot plant

  7. Reactor technology assessment and selection utilizing systems engineering approach

    International Nuclear Information System (INIS)

    Full-text: The first Nuclear power plant (NPP) deployment in a country is a complex process that needs to consider technical, economic and financial aspects along with other aspects like public acceptance. Increased interest in the deployment of new NPPs, both among newcomer countries and those with expanding programs, necessitates the selection of reactor technology among commercially available technologies. This paper reviews the Systems Decision Process (SDP) of Systems Engineering and applies it in selecting the most appropriate reactor technology for the deployment in Malaysia. The integrated qualitative and quantitative analyses employed in the SDP are explored to perform reactor technology assessment and to select the most feasible technology whose design has also to comply with the IAEA standard requirements and other relevant requirements that have been established in this study. A quick Malaysian case study result suggests that the country reside with PWR (pressurized water reactor) technologies with more detailed study to be performed in the future for the selection of the most appropriate reactor technology for Malaysia. The demonstrated technology assessment also proposes an alternative method to systematically and quantitatively select the most appropriate reactor technology. (author)

  8. The DALAT nuclear research reactor operation and conversion status

    International Nuclear Information System (INIS)

    This paper presents operation and conversion status of the DALAT Nuclear Research Reactor (DNRR). The DNRR is a pool type research reactor which was reconstructed from the 250 kW TRIGA-MARK II reactor. The core is loaded with Soviet-designed standard type WWR-M2 fuel assemblies with 36% enrichment. The reconstructed reactor reached its initial criticality in November 1983 and attained its nominal power of 500 kW in February 1984. The DNRR is operated mainly in continuous runs of 100 hours, once every 4 weeks, for radioisotope production, neutron activation analyses, training and research purposes. The remaining time between two continuous runs, is devoted to maintenance activities and to short runs. Until now 4 fuel reloading were executed. The reactor control and instrumentation system was upgraded in 1994. And now the reactor control system is being replaced by new one, the replacement will be fulfilled in March 2007. The study on fuel conversion has been done on the basis of a new LEU of 19.75% with UO2-Al dispersion fuel meat instead of the current HEU of 36% with aluminium-uranium alloy. The results of the study show that operation time of mixed core by inserting 36 LEU fuel assemblies lasts much longer than by inserting 36 HEU fuel assemblies (14.5 instead of 10.5 years). Neutron flux performances at irradiation positions are not significantly changed. Now we are working for realizing fuel conversion of the DNRR

  9. The studies for application of Ko-standardization method of neutron activation analysis on Dalat reactor

    International Nuclear Information System (INIS)

    The studies for application of the Ko-NAA method have been carried out in order to aim at creating a concrete procedure of the method, which is able to apply for practical samples. The results of the calibration of neutron spectrum parameters at cell 7-1, 13-2 and rotary rack on Dalat reactor, the calibration of detection efficiency of gamma spectrometers, the application and development of computer software for processing of gamma-ray spectra and for calculating of related problems (i.e. detector efficiency, neutron spectrum parameters, elemental concentration, etc.) and the quality control/assessment (QC/QA) with the analysis of the reference material IAEA-Soil7 have been presented as well as the application of the ko-NAA for practical samples in geological, archaeological, bio-medical and environmental objects has also been shown. (author)

  10. Standards and (self)implosion

    DEFF Research Database (Denmark)

    Brøgger, Katja; Staunæs, Dorthe

    2015-01-01

    of standards often tends to conceptualize the travelling of standards as contagious processes resulting in epidemic spreads. In this article, the abstract metaphor of epidemic spread is replaced by an analytical configuration of a new mode of educational governance in which orchestrating webs of incentives......Standards is an interesting phenomenon in discussions on emotional circulation and subjectivities. Because standards do travel. And they are usually on a mission: to standardize policies or products and consequently also those who are administered by them or consume them. Literature on the spread...... and anticipations is a major driver. New international standards is propelled by material-affective infrastructures and the embodied interpretations that, educational agents and organizations make of them. The article displays how standards makes organizations and selves implode and how the impact of standardizing...

  11. 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

  12. Labour Standards and International Trade

    OpenAIRE

    Krisztina Kis-Katos; Günther G. Schulze

    2002-01-01

    Can a case be made for the imposition of international minimum labour standards? And if so, on what grounds? The authors systematically present the existing theoretical and empirical arguments for and against introducing minimum labour standards on the international level, and discuss whether trade sanctions are the instrument of choice to improve labour standards around the world.

  13. 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)

  14. Lexicon management and standard formats

    CERN Document Server

    Laporte, Eric

    2005-01-01

    International standards for lexicon formats are in preparation. To a certain extent, the proposed formats converge with prior results of standardization projects. However, their adequacy for (i) lexicon management and (ii) lexicon-driven applications have been little debated in the past, nor are they as a part of the present standardization effort. We examine these issues. IGM has developed XML formats compatible with the emerging international standards, and we report experimental results on large-coverage lexica.

  15. Regulations for RA reactor operation

    International Nuclear Information System (INIS)

    Regulations for RA reactor operation are written in accordance with the legal regulations defined by the Law about radiation protection and related legal acts, as well as technical standards according to the IAEA recommendations. The contents of this book include: fundamental data about the reactor; legal regulations for reactor operation; organizational scheme for reactor operation; general and detailed instructions for operation, behaviour in the reactor building, performing experiments; operating rules for operation under steady state and accidental conditions

  16. 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

  17. 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

  18. Offsite dose calculation manual guidance: Standard radiological effluent controls for boiling water reactors

    International Nuclear Information System (INIS)

    This report contains guidance which may be voluntarily used by licensees who choose to implement the provision of Generic Letter 89-- 01, which allows Radiological Effluent Technical Specifications (RETS) to be removed from the main body of the Technical Specifications and placed in the Offsite Dose Calculation Manual (ODCM). Guidance is provided for Standard Effluent Controls definitions, Controls for effluent monitoring instrumentation, Controls for effluent releases, Controls for radiological environmental monitoring, and the basis for Controls. Guidance on the formulation of RETS has been available in draft form for a number of years; the current effort simply recasts those RETS into Standard Radiological Effluent Controls for application to the ODCM. 11 tabs

  19. Surveillance dosimetry of the French 900-MW pressurized water reactors (PWRs): Results, uncertainties, and reactor series effects

    International Nuclear Information System (INIS)

    Experimental neutron characterization (or dosimetry) of surveillance capsule irradiation is a main step in the surveillance program of irradiation effects on pressurized water reactor (PWR) vessels. Recently in France, results were obtained with 50 capsules removed from 29 900-MW (electric) reactors and analyzed using a standard method implementing, among others, the TRIPOLI Code to determine the perturbations caused by structural parts (stiffeners) and the neutron spectra in the capsules with their variance-covariance matrices. The method is briefly explained, emphasizing some choices that were made concerning the uncertainty evaluations. Then the results of neutron fluxes and fluences are discussed, and the mean measured value for each reactor series is compared with the computed one. In France, PWRs in operation are highly standardized. There are only 2 models of 900-MW reactors, called CPO and CPY. The 50 capsules examined were taken from these 2 series of reactors and can be arranged into 6 classes only. Some classes have as many as 18 elements, as, for example the U capsule class, irradiated at a 20 degree position in the CPY series. In each class, the observed standard deviation and estimated uncertainty are discussed. Because some classes have a strong statistical weight, the averaged results are highly significant. These reactors are constructed and operated very close to a standard model in power, flux, and dosimetry measurements. Therefore, results are weakly scattered. They confirm the calculations and indicate that the uncertainty the authors find for capsule fluence is reasonable. The best is only 6.1% (2σ). Knowledge of the vessel fluence (and of its uncertainty, which is being evaluated) is an essential factor for plant life anticipation

  20. 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..

  1. The BR2-material testing reactor and its major contribution to the reactor material, fuel and safety research

    International Nuclear Information System (INIS)

    The BR2 was shutdown at the end of June 1995 for a programme of extensive refurbishment after more than 30 years utilization. The beryllium matrix was replaced and the aluminum vessel inspected and requalified for the envisaged 15 years life extension. Other aspects of the refurbishment programme were aimed at reliability and availability of the installations, safety of operation and compliance with modem safety standards. The reactor was restarted in April 1997. This paper deals with aspects of this refurbishment in general as well as the ongoing experimental projects in the areas of reactor material, fuel behaviour and safety research. (author)

  2. 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

  3. Minimum quality standards and exports

    OpenAIRE

    Birg, Laura; Voßwinkel, Jan S.

    2015-01-01

    This paper studies the interaction of a minimum quality standard and exports in a vertical product differentiation model when firms sell global products. If exante quality of foreign firms is lower (higher) than the quality of exporting firms, a mild minimum quality standard in the home market hinders (supports) exports. The minimum quality standard increases quality in both markets. A welfare maximizing minimum quality standard is always lower under trade than under autarky. A Minimum qualit...

  4. 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)

  5. 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.

  6. Process and Control Design for a Novel Chemical Heat Exchange Reactor

    OpenAIRE

    Haugwitz, Staffan; Hagander, Per; Norén, Tommy

    2006-01-01

    A new chemical reactor, the Open Plate Reactor (OPR), is being developed by Alfa Laval AB. It has a very flexible configuration with distributed inlet ports, cooling zones and internal sensors. This gives the OPR improved control capabilities compared to standard chemical reactors in addition to better heat transfer capacity. In this paper, we address the relationship between the process design, the number of actuators used and how to use these actuators in feedback contro...

  7. Review and comparison of WWER and LWR Codes and Standards

    International Nuclear Information System (INIS)

    The results of work on a collaborative project on comparison of Codes and Standards used for safety related components of the WWER and LWR type reactors is presented. This work was performed on behalf of the European Commission, Working Group Codes and Standards and considers areas such as rules, criteria and provisions, failure mechanisms , derivation and understanding behind the fatigue curves, piping, materials and aging, manufacturing and ISI. WWERs are essentially designed and constructed using the Russian PNAE Code together with special provisions in a few countries (e.g. Czech Republic) from national standards. The LWR Codes have a strong dependence on the ASME Code. Also within Western Europe other codes are used including RCC-M, KTA and British Standards. A comparison of procedures used in all these codes and standards have been made to investigate the potential for equivalencies between the codes and any grounds for future cooperation between eastern and western experts in this field. (author)

  8. 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

  9. Economic viability of innovative nuclear reactor and fuel cycle technologies

    International Nuclear Information System (INIS)

    need to incorporate such changes of electricity market; This may suggest that small, modular-type reactor could be more advantageous than large scale, conventional reactor, especially in a low-growth, small grid market. This is especially true for low-growth and small grid market. A model cash flow analysis suggests that given the low (or uncertain) growth market, modular reactors have high economic advantage, while large scale reactor can enjoy scale-merit in faster growth market: Given high growth and large grid market in Asia, large reactor design should not be excluded from advanced reactor designs. It is important to note that for fast-growing or large grid market large reactor may be more advantageous than small reactor. It is, therefore, very important to keep the large scale designs in advanced reactor programs; Uncertainty infuel cycle (back end) costs should be minimized. This may be a unique issue for Japan and for other Asian market where back end of fuel cycle program is not well developed. Institutional mechanism can help to reduce such uncertainty in fuel cycle costs, but reactor and fuel cycle design should also aim to minimize the uncertainty; Breeding capability and/or fuel efficiency criteria are not the highest priority at present, but could become important factor in high growth scenario, and after the latter half of century. Based on the global resource availability and growth potential of nuclear power, it can be concluded that breeding or recycling capability are not the highest priority at present for next generation of advanced nuclear reactor. In general, it is desirable to have a standardized reactor design all over the world, so that production scale merit can be maximized. However, it is also important to recognize that market condition and need may vary and thus criteria for reactor design may also vary. Given the high risk of development of advanced reactor designs for future generation, therefore, it is critically important to keep

  10. 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)

  11. 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

  12. Developing research reactor coalitions and centres of excellence

    International Nuclear Information System (INIS)

    The IAEA, in line with its statute and mandatory responsibilities to support its member states in the promotion of peaceful uses of nuclear energy in concert with global nuclear non-proliferation, nuclear material security, and threat reduction objectives is well positioned to provide support for regional and international cooperation involving the research reactor community. The IAEA is pleased to announce an initiative to form one or more coalitions of research reactor operators and stakeholders to improve the sustainability of research reactors through improved market analysis and strategic/business planning, joint marketing of services, increased contacts with prospective customers and enhanced public information. Such coalition(s) will also be designed to promulgate high standards of nuclear material security, safety, quality control/assurance and to conform with global non-proliferation trends. (authors)

  13. 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

  14. A Note on Standard Deviation and Standard Error

    Science.gov (United States)

    Hassani, Hossein; Ghodsi, Mansoureh; Howell, Gareth

    2010-01-01

    Many students confuse the standard deviation and standard error of the mean and are unsure which, if either, to use in presenting data. In this article, we endeavour to address these questions and cover some related ambiguities about these quantities.

  15. 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

  16. 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.)

  17. 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

  18. Globalization of ASME Nuclear Codes and Standards

    International Nuclear Information System (INIS)

    With the globalization of the nuclear industry, it is clear that the reactor suppliers are based in many countries around the world (such as United States, France, Japan, Canada, South Korea, South Africa) and they will be marketing their reactors to many countries around the world (such as US, China, South Korea, France, Canada, Finland, Taiwan). They will also be fabricating their components in many different countries around the world. With this situation, it is clear that the requirements of ASME Nuclear Codes and Standards need to be adjusted to accommodate the regulations, fabricating processes, and technology of various countries around the world. It is also very important for the American Society of Mechanical Engineers (ASME) to be able to assure that products meeting the applicable ASME Code requirements will provide the same level of safety and quality assurance as those products currently fabricated under the ASME accreditation process. To do this, many countries are in the process of establishing or changing their regulations, and it is important for ASME to interface with the appropriate organizations in those countries, in order to ensure there is effective use of ASME Codes and standards around the world. (authors)

  19. Standards and Standard Setting and the Post School Curriculum

    Science.gov (United States)

    Young, Michael

    2014-01-01

    This paper is concerned with the role of standards and standard setting in shaping the expansion of post school education in highly unequal society. It draws on an account of the debates and policies on standards in the UK from the 1980's to today and the wider lessons that can be learned from them. It argues that relying on any type of…

  20. Antineutrino emission and gamma background characteristics from a thermal research reactor

    CERN Document Server

    Bui, V M; Fallot, M; Communeau, V; Cormon, S; Estienne, M; Lenoir, M; Peuvrel, N; Shiba, T; Cucoanes, A S; Elnimr, M; Martino, J; Onillon, A; Porta, A; Pronost, G; Remoto, A; Thiolliere, N; Yermia, F; Zakari-Issoufou, A -A

    2016-01-01

    The detailed understanding of the antineutrino emission from research reactors is mandatory for any high sensitivity experiments either for fundamental or applied neutrino physics, as well as a good control of the gamma and neutron backgrounds induced by the reactor operation. In this article, the antineutrino emission associated to a thermal research reactor: the OSIRIS reactor located in Saclay, France, is computed in a first part. The calculation is performed with the summation method, which sums all the contributions of the beta decay branches of the fission products, coupled for the first time with a complete core model of the OSIRIS reactor core. The MCNP Utility for Reactor Evolution code was used, allowing to take into account the contributions of all beta decayers in-core. This calculation is representative of the isotopic contributions to the antineutrino flux which can be found at research reactors with a standard 19.75\\% enrichment in $^{235}$U. In addition, the required off-equilibrium correction...

  1. Standard and non-standard weak interactions

    International Nuclear Information System (INIS)

    This work consists of independent chapters, all deal with weak interactions. The first chapter deals with left-right symmetric theories. Two main versions of these theories are discussed and compared. In addition, the K - K-bar mixing term is analysed: it has been known for several years now that in a left-right symmetric model there are new contributions to the mixing of kaons. We show that in the most appealing left-right symmetric model - the new contributions add up constructively. Consequently, we may derive reliable bounds on the mass of the right-handed gauge boson and the average mass of the (unavoidable) physical Higgs scalars. We also show that the new contributions are proportional to a new CP violating phase. While all previous treatments of the K - K-bar system were limited to the minimal model, we are able to show that our results hold also in the general case of nonminimal models. The second chapter deals with the possibility that W and Z are composite. Three experimental tests are discussed: (i) Universality -if W is composite then its coupling to the fermions is expected to deviate from universality. Since such deviations were not yet seen -we derive a lower bound on the compositeness scale. (ii) Possible enhancement of the reaction p-bar+p→Z0+γ+any - we show that if Z0 is composite then the cross section for the above process might be considerably enhanced and this effect can be measured at CERN and Fermilab.(iii) The eeγ events of the 1983 run in CERN - we show that in contradiction to suggestions made in several papers, these events may not be explained by a composite-Z decaying through a scalar. In the last chapter we discuss the quark mixing angles

  2. Hydropower Reservoir Operation using Standard Operating and Standard Hedging Policies

    OpenAIRE

    T.R. Neelakantan; K. Sasireka

    2013-01-01

    Standard operating policy and hedging policies are commonly used for reservoir operation for municipal or irrigation water supply. Application of these policies to hydropower reservoir operation is complex. In this paper, new standard operating policies and standard hedging policy are proposed for hydropower reservoir operation. The newly proposed policies were applied to the operation of Indira Sagar reservoir in India and demonstrated.

  3. Compact Reactor

    International Nuclear Information System (INIS)

    Weyl's Gauge Principle of 1929 has been used to establish Weyl's Quantum Principle (WQP) that requires that the Weyl scale factor should be unity. It has been shown that the WQP requires the following: quantum mechanics must be used to determine system states; the electrostatic potential must be non-singular and quantified; interactions between particles with different electric charges (i.e. electron and proton) do not obey Newton's Third Law at sub-nuclear separations, and nuclear particles may be much different than expected using the standard model. The above WQP requirements lead to a potential fusion reactor wherein deuterium nuclei are preferentially fused into helium nuclei. Because the deuterium nuclei are preferentially fused into helium nuclei at temperatures and energies lower than specified by the standard model there is no harmful radiation as a byproduct of this fusion process. Therefore, a reactor using this reaction does not need any shielding to contain such radiation. The energy released from each reaction and the absence of shielding makes the deuterium-plus-deuterium-to-helium (DDH) reactor very compact when compared to other reactors, both fission and fusion types. Moreover, the potential energy output per reactor weight and the absence of harmful radiation makes the DDH reactor an ideal candidate for space power. The logic is summarized by which the WQP requires the above conditions that make the prediction of DDH possible. The details of the DDH reaction will be presented along with the specifics of why the DDH reactor may be made to cause two deuterium nuclei to preferentially fuse to a helium nucleus. The presentation will also indicate the calculations needed to predict the reactor temperature as a function of fuel loading, reactor size, and desired output and will include the progress achieved to date

  4. Offsite dose calculation manual guidance: Standard radiological effluent controls for pressurized water reactors

    International Nuclear Information System (INIS)

    This report contains guidance which may be voluntarily used by licensees who choose to implement the provision of Generic Letter 89-01, which allows Radiological Effect Technical Specifications (RETS) to be removed from the main body of the Technical Specifications and placed in the Offsite Dose Calculation Manual (ODCM). Guidance is provided for Standard Effluent Controls definitions, Controls for effluent monitoring instrumentation, Controls for effluent releases, Controls for radiological environmental monitoring, and the basis for Controls. Guidance on the formulation of RETS has been available in draft from (NUREG-0471 and -0473) for a number of years; the current effort simply recasts those RETS into Standard Radiological Effluent Controls for application to the ODCM. Also included for completeness are: (1) radiological environmental monitoring program guidance previously which had been available as a Branch Technical Position (Rev. 1, November 1979); (2) existing ODCM guidance; and (3) a reproduction of generic Letter 89-01

  5. 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

  6. 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

  7. 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

  8. Status and possibility of fuel and structural materials experimental irradiation in BN-600 reactor. Stages of BN-600 reactor core development

    International Nuclear Information System (INIS)

    The results of the irradiation of standard and experimental fuel subassemblies (SA) in BN-600 reactor are presented. The prospects of further tests on experimental SAs and on standard SAs up to 12% h.a. burnup and damage doses ≥ 90 dpa are also analyzed. (author)

  9. 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.

  10. Standards for bullets and casings

    Directory of Open Access Journals (Sweden)

    J.F Song

    2002-11-01

    The Office of Law Enforcement Standards (OLES at the National Institute of Standards and Technology (NIST manages research in many different disciplines of forensic science. One of these projects supports the National Integrated Ballistics Information Network (NIBIN. NIST digitized six bullet signatures from samples provided by the Bureau of Alcohol, Tobacco, and Firearms (ATF and the Federal Bureau of Investigation (FBI. Using these signatures as a virtual standard, NIST’s Instrument Shop manufactured 20 reference materials (RM 8240 standard bullets using a numerically-controlled diamond turning machine. Test results show high reproducibility of the bullet signatures on standard bullets. NIST has also developed a new parameter for bullet signature comparisons, using autocorrelation functions, and proposed a diagram for tracing local ballistics measurements to the National Laboratory Center of the ATF and to the FBI. Using an electro-forming process, NIST has manufactured prototype standard casings and test results show high reproducibility for the casing signatures.

  11. 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...

  12. Halden Reactor Project activities, achievements and international collaboration

    International Nuclear Information System (INIS)

    This paper concentrates on the Halden Project research programme related to fuel testing. An overview of ongoing tests on WWER fuel performance is also included. The ongoing and planned experiments containing WWER-related fuels and materials - Irradiation of Standard and Modified WWER Fuel (IFA-503) and Corrosion Testing of Different Cladding Alloys (IFA-638) - are presented. The future experiments involving WWER fuel and cladding types foreseen in of the Halden Reactor Project programme are given

  13. Project to replace the control and protection system at the WWR-K research reactor

    International Nuclear Information System (INIS)

    Replacement of reactor control and protection system (CPS) is a part of the wider program of the reactor modifications related to its conversion from HEU to LEU fuel, supported by the US Department of Energy, the Nuclear Threat Initiative and the Kazakhstan government. The program includes development of a new LEU fuel assembly design and re-configuration of the reactor core, with change in number and positions of the control rods, which requires modification of the CPS. Furthermore, replacement of the existing instrumentation, some of which is 40 years old, will improve reactor safety, bring the CPS up to current international standards, and provide an upgraded control interface

  14. 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

  15. Thermohydraulic design and safety analysis of research reactors

    International Nuclear Information System (INIS)

    This contribution presents briefly the trend of thermal hydraulic design and safety analysis of medium and high flux research reactors. This field of deterministic safety analysis is being considered by the IAEA in the framework of coordinated research project (CRP) initiated in 2002 on the Assessment of Analytical Tools for Different Research Reactor Types. The objective of this project is to establish a forum of international experts in order to integrate the activities for improvement and verification of selected computer codes that can be considered as reference tools in the safety analysis of research reactors, similar to that by power reactors. This undertaken supports the international ambition in improving the safety features and standards of research reactors, which can be useful for countries with long experience on RR and very helpful for countries having research reactors with low neutron flux and may looking for to extend them or build other reactors with higher neutron flux. In this regard the methodological approach on modification, verification and application of advanced computer codes for the safety analysis of research reactors is presented. In this regard the methodological approach on modification, verification and application of advanced computer codes for the safety analysis of research reactors is presented. The presented paper deals with the prediction of a semi empirical correlation for the first design limit regarding the onset of flow instability. Using the experimental data by considering the verification results for the thermalhydraulic and safety analysis code ATHLET a simple correlation for the Onset of Flow Instability (OFI) for medium and high flux reactors has been suggested. This correlation predicts by a given maximum heat flux of the hot channel the amount of inlet flow velocity at which an onset of instability is expected. It presents a simple procedure to estimate the minimum allowed flow velocity at which the fuel element

  16. Inconsistencies of neutron flux parameters for k(0) standardization in neutron activation analysis determined with the use of Au+Zr and Au+Mo+Cr monitor sets at the LVR-15 reactor in Rez

    Czech Academy of Sciences Publication Activity Database

    Kubešová, Marie; Kučera, Jan

    2012-01-01

    Roč. 293, č. 2 (2012), s. 665-674. ISSN 0236-5731 R&D Projects: GA ČR GA202/09/0363 Institutional support: RVO:61389005 Keywords : Neutron activation analysis * K(0) standardization * Neutron flux parameters Subject RIV: BG - Nuclear, Atomic and Molecular Physics, Colliders Impact factor: 1.467, year: 2012

  17. Radiological Calibration and Standards Facility

    Data.gov (United States)

    Federal Laboratory Consortium — PNNL maintains a state-of-the-art Radiological Calibration and Standards Laboratory on the Hanford Site at Richland, Washington. Laboratory staff provide expertise...

  18. 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

  19. Determination of silicon in Japanese iron reference standard materials by reactor fast neutron activation analysis combined with a simple pre-concentration

    International Nuclear Information System (INIS)

    A reactor fast neutron activation analysis was used in combination with a simple pre-concentration procedure for determining silicon in some iron reference standard materials of Japan Iron and Steel Federation. The samples were dissolved with aqua regia and digested with perchloric or sulfuric acid. The precipitated silica was collected on a filter paper and irradiated in a cadmium case with reactor fast neutrons. Silicon can be determined in tool steel SKD6, low-alloy steel Nos 2 and 4 and silico-manganese samples by a present method measuring 1,273.4 keV γ-rays from 6.63-minute 29Al produced by 29Si(n,p)29Al reaction. (author)

  20. Core labour standards and exports

    OpenAIRE

    Jean-Marc Siroën

    2012-01-01

    (english) Core labour standards defined by the ILO in 1998 are universal, but applied very differently across countries. Compliance is much higher in high income countries. However, the causality between improved labour standards and economic growth remains a controversial issue. Export-led growth strategies might encourage developing countries to curb the process of standards improvement. In this way, they can raise the volume of their unskilled labour endowments (child and/or forced labour)...