WorldWideScience

Sample records for babcock and wilcox test reactor

  1. History of Research Reactor Fuel Fabrication at Babcock and Wilcox

    International Nuclear Information System (INIS)

    Freim, J.B.

    1983-01-01

    1982 was a year of tremendous growth for Babcock and Wilcox and its Research Reactor Fuel Facility. The Division has progressed from essentially being a non-competitor to a position where we are growing in strength. This paper describes some of the general aspects of past history and where B and W is now

  2. Transient response of Babcock and Wilcox-designed reactors

    International Nuclear Information System (INIS)

    1980-05-01

    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

  3. Babcock and Wilcox advanced PWR development

    International Nuclear Information System (INIS)

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

    1986-01-01

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

  4. Babcock and Wilcox model for predicting in-reactor densification

    International Nuclear Information System (INIS)

    Buescher, B.J.; Pegram, J.W.

    1977-07-01

    The B and W densification model is based on a correlation between in-reactor densification and a thermal resintering test. The densification model has been found to predict in-reactor densification with a remarkable degree of accuracy for fuel pellets operated at heat rates above 5 kW/ft and with considerable conservatism for pellelts operating at heat rates below 5 kW/ft

  5. Babcock and Wilcox model for predicting in-reactor densification

    International Nuclear Information System (INIS)

    Buescher, B.J.; Pegram, J.W.

    1975-06-01

    The B and W fuel densification model is used to describe the extent and kinetics of in-reactor densification in B and W production fuel. The model and approach are qualified against an extensive data base available through B and W's participation in the EEI Fuel Densification Program. Out-of-reactor resintering tests on representative pellets from each batch of fuel are used to provide input parameters to the B and W densification model. The B and W densification model predicts in-reactor densification very accurately for pellets operated at heat rates above 5 kW/ft and with considerable conservation for pellets operated at heat rates less than 5 kW/ft. This model represents a technically rigorous and conservative basis for predicting the extent and kinetics of in-reactor densification. 9 references. (U.S.)

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

    International Nuclear Information System (INIS)

    Freim, James B.

    1983-01-01

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

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

    International Nuclear Information System (INIS)

    Virgilio, M.J.

    1980-10-01

    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

  8. Uranium silicide activities at Babcock and Wilcox

    International Nuclear Information System (INIS)

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

    1983-01-01

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

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

    International Nuclear Information System (INIS)

    1978-06-01

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

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

    International Nuclear Information System (INIS)

    Virgilio, M.

    1979-07-01

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

  11. Standard Technical Specifications, Babcock and Wilcox Plants

    International Nuclear Information System (INIS)

    1992-09-01

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

  12. Standard Technical Specifications, Babcock and Wilcox plants

    International Nuclear Information System (INIS)

    1992-09-01

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

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

    International Nuclear Information System (INIS)

    Lee, R.Y.

    1988-02-01

    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

  14. Standard Technical Specifications, Babcock and Wilcox plants

    International Nuclear Information System (INIS)

    1992-09-01

    This NUREG contains improved Standard Technical Specifications (STS) for Babcock and Wilcox (B ampersand W) Plants and documents the positions of the Nuclear Regulatory Commission based on the B ampersand W Owners Group's proposed STS. This document is the result of extensive technical meetings and discussions among the NRC staff, the Nuclear Steam Supply System (NSSS) Owners Groups, the NSSS vendors, and the Nuclear Management and Resources Council (NUMARC). The 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

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

    International Nuclear Information System (INIS)

    Freim, J.B.

    1983-01-01

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

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

    International Nuclear Information System (INIS)

    Todd, Lawrence E.; Pace, Brett W.

    1996-01-01

    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)

  17. RELAP5/MOD2 assessment at Babcock and Wilcox

    International Nuclear Information System (INIS)

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

    1986-01-01

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

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

    International Nuclear Information System (INIS)

    Cramond, W.R.; Ericson, D.M. Jr.; Sanders, G.A.

    1987-03-01

    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

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

    International Nuclear Information System (INIS)

    Smith, J.C.

    1998-01-01

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

  20. 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)

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

    International Nuclear Information System (INIS)

    1995-04-01

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

  2. Effect of flow leakage on the benchmarking of FLOWTRAN with Mark-22 mockup flow excursion test data from Babcock and Wilcox

    International Nuclear Information System (INIS)

    Chen, Kuo-Fu.

    1992-10-01

    This report presents a revised analysis of the Babcock and Wilcox (B and W) downflow flow excursion tests that accounts for leakage between flow channels in the test assembly. Leak rates were estimated by comparing results from the downflow tests with those for upflow tests conducted using an identical assembly with some minor modifications. The upflow test assembly did not contain leaks. This revised analyses shows that FLOWTRAN with the SRS working criterion conservatively predicts onset of flow instability without using a local peaking factor to model heat transfer variations near the ribs

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

    International Nuclear Information System (INIS)

    Hassan, H.H.; Wittkopf, W.A.; Mullan, W.H.

    1977-01-01

    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

  4. Evaluation of operational safety at Babcock and Wilcox Plants: Volume 2, Thermal-hydraulic results

    International Nuclear Information System (INIS)

    Wheatley, P.D.; Davis, C.B.; Callow, R.A.; Fletcher, C.D.; Dobbe, C.A.; Beelman, R.J.

    1987-11-01

    The Nuclear Regulatory Commission has initiated a research program to develop a methodology to assess the operational performance of Babcock and Wilcox plants and to apply this methodology on a trial basis. The methodology developed for analyzing Babcock and Wilcox plants integrated methods used in both thermal-hydraulics and human factors and compared results with information used in the assessment of risk. The integrated methodology involved an evaluation of a selected plant for each pressurized water reactor vendor during a limited number of transients. A plant was selected to represent each vendor, and three transients were identified for analysis. The plants were Oconee Unit 1 for Babcock and Wilcox, H.B. Robinson Unit 2 for Westinghouse, and Calvert Cliffs Unit 1 for Combustion Engineering. The three transients were a complete loss of all feedwater, a small-break loss-of-coolant accident, and a steam-generator overfill with auxiliary feedwater. Included in the integrated methodology was an assessment of the thermal-hydraulic behavior, including event timing, of the plants during the three transients. Thermal-hydraulic results are presented in this volume (Volume 2) of the report. 26 refs., 30 figs., 7 tabs

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

    International Nuclear Information System (INIS)

    Duncan, D.R.

    1994-01-01

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

    International Nuclear Information System (INIS)

    1987-11-01

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

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

    International Nuclear Information System (INIS)

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

    1992-04-01

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

  8. LEU fuel powder technology at Babcock and Wilcox (USA)

    International Nuclear Information System (INIS)

    Bogacik, K.E.

    1984-01-01

    This paper traces BandW involvement in HEU fuel manufacturing to the current work directed at LEU reactor technology. Past work at BandW in areas such as alloying, fuel handling and core manufacturing has been of significant benefit to the current LEU fuel processing requirements. Recent investigations and process developments for production of LEU aluminide and silicide fuels are discussed. Techniques for alloying by vacuum are melting, followed by comminution methods after alloying, are presented for both the LEU aluminide and silicide fuel powders. Powder processing discussions include compacting techniques used by BandW for these alloys. This overview of BandW's LEU i nvolvement provides details of specific modifications and process developments in powdered fuels. Product attributes such as powder chemistry, size, and other physical properties of each LEU fuel are presented. (author)

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

    International Nuclear Information System (INIS)

    Thoma, J.O.

    1985-01-01

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

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

    International Nuclear Information System (INIS)

    Wittkopf, W.A.

    1976-03-01

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

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

    International Nuclear Information System (INIS)

    1977-01-01

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

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

    International Nuclear Information System (INIS)

    Guss, P.P.

    1993-04-01

    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 ( 60 Co) and cesium-137 ( 137 Cs). A second area near the main plant indicated the presence of uranium-235 ( 235 U). Protactinium-234m ( 234m Pa) and 60 Co 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

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

    International Nuclear Information System (INIS)

    1980-01-01

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

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

    International Nuclear Information System (INIS)

    1995-04-01

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

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

    International Nuclear Information System (INIS)

    1995-04-01

    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

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

    International Nuclear Information System (INIS)

    Martynova, O.I.

    1978-01-01

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

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

    International Nuclear Information System (INIS)

    Thoma, J.O.

    1984-05-01

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

  18. 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)

    1979-03-01

    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

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

    International Nuclear Information System (INIS)

    Pepper, D.W.

    1978-03-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Pepper, D.W.

    1978-03-01

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

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

    International Nuclear Information System (INIS)

    Berger, Marjorie

    2016-01-01

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

  2. Babcock & Wilcox technologies for power plant stack emissions control

    Energy Technology Data Exchange (ETDEWEB)

    Polster, M.; Nolan, P.S.; Batyko, R.J. [Babcock & Wilcox, Barberton, OH (United States)

    1994-12-31

    The current status of sulfur dioxide control in power plants is reviewed with particular emphasis on proven, commercial technologies. This paper begins with a detailed review of Babcock & Wilcox commercial wet flue gas desulfurization (FGD) systems. This is followed by a brief discussion of B&W dry FGD technologies, as well as recent full-scale and pilot-scale demonstration projects which focus on lower capital cost alternatives to conventional FGD systems. A comparison of the economics of several of these processes is also presented. Finally, technology selections resulting from recent acid rain legislation in various countries are reviewed.

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

    International Nuclear Information System (INIS)

    Ghan, L.S.; Ortiz, M.G.

    1991-01-01

    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

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

    International Nuclear Information System (INIS)

    1976-01-01

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

  5. Compact Process Development at Babcock & Wilcox

    Energy Technology Data Exchange (ETDEWEB)

    Eric Shaber; Jeffrey Phillips

    2012-03-01

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

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

    International Nuclear Information System (INIS)

    1990-02-01

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

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

    International Nuclear Information System (INIS)

    1990-02-01

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

  8. Estimate of airborne release of plutonium from Babcock and Wilcox plant as a result of severe wind hazard and earthquake

    International Nuclear Information System (INIS)

    Mishima, J.; Schwendiman, L.C.; Ayer, J.E.

    1978-10-01

    As part of an interdisciplinary study to evaluate the potential radiological consequences of wind hazard and earthquake upon existing commercial mixed oxide fuel fabrication plants, the potential mass airborne releases of plutonium (source terms) from such events are estimated. The estimated souce terms are based upon the fraction of enclosures damaged to three levels of severity (crush, puncture penetrate, and loss of external filter, in order of decreasing severity), called damage ratio, and the airborne release if all enclosures suffered that level of damage. The discussion of damage scenarios and source terms is divided into wind hazard and earthquake scenarios in order of increasing severity. The largest airborne releases from the building were for cases involving the catastrophic collapse of the roof over the major production areas--wind hazard at 110 mph and earthquakes with peak ground accelerations of 0.20 to 0.29 g. Wind hazards at higher air velocities and earthquakes with higher ground acceleration do not result in significantly greater source terms. The source terms were calculated as additional mass of respirable particles released with time up to 4 days; and, under these assumptions, approximately 98% of the mass of material of concern is made airborne from 2 h to 4 days after the event. The overall building source terms from the damage scenarios evaluated are shown in a table. The contribution of individual areas to the overall building source term is presented in order of increasing severity for wind hazard and earthquake

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

    Science.gov (United States)

    1980-09-01

    06. 1 21 Kenai Corp............ 39 21 Globuarin.......... 4.0 21 Za4pata ................ 0.8 22 IKea Corp............. 47.3 22 Tea Intl...months, to find the causes and discuss ways to clean up the worst nuclear accident in history . But the impact that T.M.I. will have on the economy will be

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

    International Nuclear Information System (INIS)

    Hsii, Y.H.

    1976-06-01

    The CONTEMPT computer program predicts the pressure-temperature response of a single-volume reactor building to a loss-of-coolant accident. The report describes the analytical model used for the program. 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 user is required to input the description of the discharge of coolant, the boiling of residual water by reactor decay heat, the superheating of steam passing through the core, and metal-water reactions. The reactor building is separated into liquid and vapor regions. Each region is in thermal equilibrium itself, but the two may not be in thermal equilibrium; the liquid and gaseous regions may have different temperatures. The reactor building is represented as consisting of several heat-conducting structures whose thermal behavior can be described by the one-dimensional multi-region heat conduction equation. 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. Babcock and Wilcox revisions to CONTEMPT, computer program for predicting containment pressure-temperature response to a loss-of-coolant accident

    International Nuclear Information System (INIS)

    Hsii, Y.H.

    1975-01-01

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

  12. Babcock experience of automated ultrasonic non-destructive testing of PWR pressure vessels during manufacture

    International Nuclear Information System (INIS)

    Dikstra, B.J.; Farley, J.M.; Scruton, G.

    1990-01-01

    Major developments in ultrasonic techniques, equipment and systems for automated inspection have lead, over a period of about ten years, to the regular application of sophisticated computer-controlled systems during the manufacture of nuclear reactor pressure vessels. Ten years ago the use of procedures defined in a code such as ASME XI might have been considered sufficient, but it is now necessary, as was demonstrated by the results of the UKAEA defect detection trials and the PISC II trials, to apply more comprehensive arrays of probes and higher test sensitivities. The ultrasonic techniques selected are demonstrated to be adequate by modelling or test-block exercises, the automated systems applied are subject to stringent quality assurance testing, and very rigorous inspection procedures are used in conjunction with a high degree of automation to ensure reproducibility of inspection quality. The state-of-the-art in automated ultrasonic testing of pressure vessels by Babcock is described. Current developments by the company, including automated flaw recognition, integrated modelling of inspection capability, and the use of electronically scanned variable-angle probes are reviewed. Examples quoted include the automated ultrasonic inspections of the Sizewell B pressurized water reactor vessel. (author)

  13. 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, 2007). The E-Filing process requires participants to submit and serve all adjudicatory documents over... below. To comply with the procedural requirements of E-Filing, at least 10 days prior to the filing...

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

    International Nuclear Information System (INIS)

    Burke, F.R.

    1977-06-01

    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

  15. Multiloop integral system test (MIST)

    International Nuclear Information System (INIS)

    Gloudemans, J.R.

    1989-07-01

    The multiloop integral system test (MIST) was part of a multiphase program started in 1983 to address small-break loss-of-coolant accidents (SBLOCAs) specific to Babcock and Wilcox-designed plants. MIST was sponsored by the US Nuclear Regulatory Commission, the Babcock and Wilcox Owners Group, the Electric Power Research Institute, and Babcock and Wilcox. The unique features of the Babcock and Wilcox design, specifically the hot leg U-bends and steam generators, prevented the use of existing integral system data or existing integral system facilities to address the thermal-hydraulic SBLOCA questions. MIST and two other supporting facilities were specifically designed and constructed for this program, and an existing facility -- the once-through integral system (OTIS) -- was also used. Data from MIST and the other facilities will be used to benchmark the adequacy of system codes, such as RELAP5 and TRAC, for predicting abnormal plant transients. The individual tests are described in detail in Volumes 2 through 8 and Volume 11, and are summarized in Volume 1. Inter-group comparisons are addressed in this document, Volume 9. These comparisons are grouped as follows: mapping versus SBLOCA transients, SBLOCA, pump effects, and the effects of noncondensible gases. Appendix A provides an index and description of the microfiched plots for each test, which are enclosed with the corresponding Volumes 2 through 8. 147 figs., 5 tabs

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

    International Nuclear Information System (INIS)

    1980-01-01

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

  17. Multiloop Integral System Test (MIST): MIST Facility Functional Specification

    International Nuclear Information System (INIS)

    Habib, T.F.; Koksal, C.G.; Moskal, T.E.; Rush, G.C.; Gloudemans, J.R.

    1991-04-01

    The Multiloop Integral System Test (MIST) is part of a multiphase program started in 1983 to address small-break loss-of-coolant accidents (SBLOCAs) specific to Babcock and Wilcox designed plants. MIST is sponsored by the US Nuclear Regulatory Commission, the Babcock ampersand Wilcox Owners Group, the Electric Power Research Institute, and Babcock and Wilcox. The unique features of the Babcock and Wilcox design, specifically the hot leg U-bends and steam generators, prevented the use of existing integral system data or existing integral facilities to address the thermal-hydraulic SBLOCA questions. MIST was specifically designed and constructed for this program, and an existing facility -- the Once Through Integral System (OTIS) -- was also used. Data from MIST and OTIS are used to benchmark the adequacy of system codes, such as RELAP5 and TRAC, for predicting abnormal plant transients. The MIST Functional Specification documents as-built design features, dimensions, instrumentation, and test approach. It also presents the scaling basis for the facility and serves to define the scope of work for the facility design and construction. 13 refs., 112 figs., 38 tabs

  18. Multiloop integral system test (MIST): Final report, Inter-group comparisons

    Energy Technology Data Exchange (ETDEWEB)

    Gloudemans, J.R.

    1989-07-01

    The multiloop integral system test (MIST) was part of a multiphase program started in 1983 to address small-break loss-of-coolant accidents (SBLOCAs) specific to Babcock and Wilcox-designed plants. MIST was sponsored by the US Nuclear Regulatory Commission, the Babcock and Wilcox Owners Group, the Electric Power Research Institute, and Babcock and Wilcox. The unique features of the Babcock and Wilcox design, specifically the hot leg U-bends and steam generators, prevented the use of existing integral system data or existing integral system facilities to address the thermal-hydraulic SBLOCA questions. MIST and two other supporting facilities were specifically designed and constructed for this program, and an existing facility -- the once-through integral system (OTIS) -- was also used. Data from MIST and the other facilities will be used to benchmark the adequacy of system codes, such as RELAP5 and TRAC, for predicting abnormal plant transients. The individual tests are described in detail in Volumes 2 through 8 and Volume 11, and are summarized in Volume 1. Inter-group comparisons are addressed in this document, Volume 9. These comparisons are grouped as follows: mapping versus SBLOCA transients, SBLOCA, pump effects, and the effects of noncondensible gases. Appendix A provides an index and description of the microfiched plots for each test, which are enclosed with the corresponding Volumes 2 through 8. 147 figs., 5 tabs.

  19. Research reactors and materials testing

    International Nuclear Information System (INIS)

    Vidal, H.

    1986-01-01

    Research reactors can be classified in three main groups according to the moderator which is used. Their technical characteristics are given and the three most recent research and materials testing reactors are described: OSIRIS, ORPHEE and the high-flux reactor of Grenoble. The utilization of research reactors is reviewed in four fields of activity: training, fundamental or applied research and production (eg. radioisotopes) [fr

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

    International Nuclear Information System (INIS)

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

    1990-01-01

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

  1. Performance of self-powered neutron detectors in pressurized water reactors

    International Nuclear Information System (INIS)

    Warren, H.D.; Bozarch, D.P.

    1977-01-01

    A typical Babcock and Wilcox pressurized water reactor (PWR) contains 364 rhodium self-powered neutron detectors (SPNDs) and 52 background detectors. The detectors are inserted into the reactor core in 52 dry, multidetector assemblies. Each assembly contains seven SPNDs and one background detector. By mid-1977, eight B and W PWRs, each fitted with SPNDs, were in operation. Many of the SPNDs have operated successfully for more than four years. This paper describes the operational performance of the SPNDs and special tests conducted to improve that performance. Topics included are (1) insulation performance versus neutron dose to the SPND, (2) background signals in the leadwire region of the SPND, and (3) depletion of the SPND emitter versus absorbed neutron dose

  2. Nuclear plant-aging research on reactor protection systems

    International Nuclear Information System (INIS)

    Meyer, L.C.

    1988-01-01

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

  3. Reactor protection system design using micro-computers

    International Nuclear Information System (INIS)

    Fairbrother, D.B.

    1976-01-01

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

  4. Reactor protection system design using micro-computers

    International Nuclear Information System (INIS)

    Fairbrother, D.B.

    1977-01-01

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

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

    International Nuclear Information System (INIS)

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

    1982-09-01

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

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

    International Nuclear Information System (INIS)

    Anon.

    1979-01-01

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

  7. TRAC-PF1 code verification with data from the OTIS test facility

    International Nuclear Information System (INIS)

    Childerson, M.T.; Fujita, R.K.

    1985-01-01

    A computer code (TRAC-PF1/MOD1) developed for predicting transient thermal and hydraulic integral nuclear steam supply system (NSSS) response was benchmarked. Post-small break loss-of-coolant accident (LOCA) data from a scaled, experimental facility, designated the One-Through Integral System (OTIS), were obtained for the Babcock and Wilcox NSSS and compared to TRAC predictions. The OTIS tests provided a challenging small break LOCA data set for TRAC verification. The major phases of a small break LOCA observed in the OTIS tests included pressurizer draining and loop saturation, intermittent reactor coolant system circulation, boiler-condenser mode, and the initial stages of refill. The TRAC code was successful in predicting OTIS loop conditions (system pressures and temperatures) after modification of the steam generator model. In particular, the code predicted both pool and auxiliary-feedwater initiated boiler-condenser mode heat transfer

  8. TRAC-PF1 code verification with data from the OTIS test facility

    International Nuclear Information System (INIS)

    Childerson, M.T.; Fujits, R.K.

    1985-01-01

    A computer code (TRAC-PFI/MODI; denoted as TRAC) developed for predicting transient thermal and hydraulic integral nuclear steam supply system (NSSS) response was benchmarked. Post-small break loss-of-coolant accident (LOCA) data from a scaled, experimental facility, designated the Once-Through Integral Systems (OTIS), were obtained for the Babcock and Wilcox NSSS and compared to TRAC predictions. The OTIS tests provided a challenging small break LOCA data set for TRAC verification. The major phases of a small break LOCA observed in the OTIS tests included pressurizer draining and saturation, intermittent reactor coolant system circulation, boiler-condenser mode and the initial stages of refill. The TRAC code was successful in predicting OTIS loop conditions (system pressures and temperatures) after modification of the steam generator model. In particular, the code predicted both pool- and auxiliary- feedwater initiated boiler-condenser mode heat transfer

  9. Test plan for glass melter system technologies for vitrification of hign-sodium content low-level radioactive liquid waste, Project No. RDD-43288

    International Nuclear Information System (INIS)

    Higley, B.A.

    1995-01-01

    This document provides a test plan for the conduct of combustion fired cyclone vitrification testing by a vendor in support of the Hanford Tank Waste Remediation System, Low-Level Waste Vitrification Program. The vendor providing this test plan and conducting the work detailed within it is the Babcock ampersand Wilcox Company Alliance Research Center in Alliance, Ohio. This vendor is one of seven selected for glass melter testing

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    Eytchison, R.M.

    1980-01-01

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

  12. 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.

  13. Reactor protection system

    International Nuclear Information System (INIS)

    Fairbrother, D.B.; Lesniak, L.M.; Orgera, E.G.

    1977-10-01

    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

  14. RELAP5 analyses of overcooling transients in a pressurized water reactor

    International Nuclear Information System (INIS)

    Bolander, M.A.; Fletcher, C.D.; Ogden, D.M.; Stitt, B.D.; Waterman, M.E.

    1983-01-01

    In support of the Pressurized Thermal Shock Integration Study sponsored by the United States Nuclear Regulatory Commission, the Idaho National Engineering Laboratory has performed analyses of overcooling transients using the RELAP5/MOD1.5 computer code. These analyses were performed for Oconee Plants 1 and 3, which are pressurized water reactors of Babcock and Wilcox lowered-loop design. Results of the RELAP5 analyses are presented, including a comparison with plant data. The capabilities and limitations of the RELAP5/MOD1.5 computer code in analyzing integral plant transients are examined. These analyses require detailed thermal-hydraulic and control system computer models

  15. Reactor protection system. Revision 1

    International Nuclear Information System (INIS)

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

    1975-04-01

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

  16. Containment for small pressurized water reactors

    International Nuclear Information System (INIS)

    Siler, W.C.; Marda, R.S.; Smith, W.R.

    1977-01-01

    Babcock and Wilcox Company has prepared studies under ERDA contract of small and intermediate size (313, 365 and 1200 MWt) PWR reactor plants, for industrial cogeneration or electric power generation. Studies and experience with nuclear plants in this size range indicate unfavorable economics. To offset this disadvantage, modular characteristics of an integral reactor and close-coupled vapor suppression containment have been exploited to shorten construction schedules and reduce construction costs. The resulting compact reactor/containment complex is illustrated. Economic studies to date indicate that the containment design and the innovative construction techniques developed to shorten erection schedules have been important factors in reducing estimated project costs, thus potentially making such smaller plants competetive with competing energy sources

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

    International Nuclear Information System (INIS)

    1993-11-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-02-01

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

  19. Commissioning of the new heat exchanger for the research nuclear reactor IEA-R1

    Energy Technology Data Exchange (ETDEWEB)

    Castro, Alfredo Jose Alvim de; Cassiano, Douglas Alves; Umbehaun, Pedro Ernesto; Carvalho, Marcos Rodrigues de; Frajndlich, Roberto [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)]. E-mails: ajcastro@ipen.br; docass@gmail.com.br; umbehaun@ipen.br; carvalho@ipen.br; frajndli@ipen.br

    2008-07-01

    The Research Reactor IEA-R1 placed at IPEN/CNEN-SP is of the swimming pool type, light water moderated and with graphite reflectors, and was build and designed by Babcock and Wilcox Co. Start up operation was in September the 16{sup th}, 1957, being the first criticality for South Hemisphere. Although designed to operate at 5 MW, the IEA-R1 was operated until 2001 with 2 MW and was suitable for use in basic and applied research as well as the production of medical radioisotopes, industry and natural sciences applications. Due to a recent demand increase on radioisotopes in Brazil for medical diagnoses and therapies applications, IPEN /CNEN updated the IEA-R1 power to 5 MW and to work at continuous operation regime. Studies on the Ageing Management for the Research Reactor IEA-R1 were conducted according to IAEA procedures. As result of these studies critical components within the Ageing Management Program were identified. Also were made recommendations on the implementation of test scheduling and standardization procedures to organize data and documents. One of the main results was the need of monitoring the two heat exchangers, the two primary circuit pumps and the data acquisition system. During monitoring procedures, issues were observed on the IEA-R1 operation at 5 MW mainly due to the ageing of the Babcox and Wilcox TCA heat exchanger, and excessive vibrations at high flow rates on CBC's TCB heat exchanger. So, from 2005 on, it was decided to work with 3,5 MW and provide a new IESA heat exchanger with 5 MW capacity, to substitute the TCA heat exchanger. This work presents results on the commissioning of the new heat exchanger and compares against the values calculated in the IESA project. The results show that the IEA-R1 Reactor can be operated more safety and continuously at 5 MW with the new IESA heat exchanger. (author)

  20. RELAP5 analyses and support of Oconee-1 PTS studies

    International Nuclear Information System (INIS)

    Charlton, T.R.

    1983-01-01

    The integrity of a reactor vessel during a severe overcooling transient with primary system pressurization is a current safety concern and has been identified as an Unresolved Safety Issue(USI) A-49 by the US Nuclear Regulatory Commission (NRC). Resolution of USI A-49, denoted as Pressurized Thermal Shock (PTS), is being examined by the US NRC sponsored PTS integration study. In support of this study, the Idaho National Engineering Laboratory (INEL) has performed RELAP5/MOD1.5 thermal-hydraulic analyses of selected overcooling transients. These transient analyses were performed for the Oconee-1 pressurized water reactor (PWR), which is Babcock and Wilcox designed nuclear steam supply system

  1. Advanced Demonstration and Test Reactor Options Study

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-01-01

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

  2. Advanced Demonstration and Test Reactor Options Study

    International Nuclear Information System (INIS)

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

    2017-01-01

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

  3. High conversion ratio plutonium recycle in pressurized water reactors

    International Nuclear Information System (INIS)

    Edlund, M.C.

    1975-01-01

    The use of Pu light water reactors in such a way as to minimise the depletion of Pu needed for future use, and therefore to reduce projected demands for U ore and U enrichment is envisaged. Fuel utilisation in PWRs could be improved by tightly-packed fuel rod lattices with conversion ratios of 0.8 to 0.9 compared with ratios of about 0.5 in Pu recycle designs using fuel to water volume ratios of currently operating PWRs. A conceptual design for the Babcock and Wilcox Company reactors now in operation is presented and for illustrative purposes thermalhydraulic design considerations and the reactor physics are described. Principle considerations in the mechanical design of the fuel assemblies are the effect of hydraulic forces, thermal expansion, and fission gas release. The impact of high conversion ratio plutionium recycle in separative work and natural U requirements for PWRs likely to be in operation by 1985 are examined. (U.K.)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-03-01

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

  5. Reactor group constants and benchmark test

    Energy Technology Data Exchange (ETDEWEB)

    Takano, Hideki [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2001-08-01

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

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

    International Nuclear Information System (INIS)

    Grover, S.B.

    2004-01-01

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

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

    International Nuclear Information System (INIS)

    S. Blaine Grover

    2004-01-01

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

  8. Correlations between power and test reactor data bases

    International Nuclear Information System (INIS)

    Guthrie, G.L.; Simonen, E.P.

    1989-02-01

    Differences between power reactor and test reactor data bases have been evaluated. Charpy shift data has been assembled from specimens irradiated in both high-flux test reactors and low-flux power reactors. Preliminary tests for the existence of a bias between test and power reactor data bases indicate a possible bias between the weld data bases. The bias is nonconservative for power predictive purposes, using test reactor data. The lesser shift for test reactor data compared to power reactor data is interpreted primarily in terms of greater point defect recombination for test reactor fluxes compared to power reactor fluxes. The possibility of greater thermal aging effects during lower damage rates is also discussed. 15 refs., 5 figs., 2 tabs

  9. Advanced test reactor testing experience-past, present and future

    International Nuclear Information System (INIS)

    Marshall, Frances M.

    2006-01-01

    The Advanced Test Reactor (ATR), at the Idaho National Laboratory (INL), is one of the world's premier test reactors for providing the capability for studying the effects of intense neutron and gamma radiation on reactor materials and fuels. The physical configuration of the ATR, a 4-leaf clover shape, allows the reactor to be operated at different power levels in the corner 'lobes' to allow for different testing conditions for multiple simultaneous experiments. The combination of high flux (maximum thermal neutron fluxes of 1E15 neutrons per square centimeter per second and maximum fast [E>1.0 MeV] neutron fluxes of 5E14 neutrons per square centimeter per second) and large test volumes (up to 122 cm long and 12.7 cm diameter) provide unique testing opportunities. The current experiments in the ATR are for a variety of test sponsors - US government, foreign governments, private researchers, and commercial companies needing neutron irradiation services. There are three basic types of test configurations in the ATR. The simplest configuration is the sealed static capsule, which places the capsule in direct contact with the primary coolant. The next level of experiment complexity is an instrumented lead experiment, which allows for active control of experiment conditions during the irradiation. The most complex experiment is the pressurized water loop, in which the test sample can be subjected to the exact environment of a pressurized water reactor. For future research, some ATR modifications and enhancements are currently planned. This paper provides more details on some of the ATR capabilities, key design features, experiments, and future plans

  10. 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…

  11. Pressurized-water-reactor station blackout

    International Nuclear Information System (INIS)

    Dobbe, C.A.

    1983-01-01

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

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

    International Nuclear Information System (INIS)

    Spiewak, I.; Klepper, O.H.; Fuller, L.C.

    1977-02-01

    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

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

    International Nuclear Information System (INIS)

    Spiewak, I.; Klepper, O.H.; Fuller, L.C.

    1977-01-01

    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 of several nuclear steam supply systems in the small and intermediate size range. Detail studies are reported of the Consolidated Nuclear Steam Generator (CNSG), a 313 MW(t) 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(t), which are envisioned for utility operation for supply of electric power and steam. Design studies of a 500 MW(t) 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

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

    International Nuclear Information System (INIS)

    Spiewak, I.; Klepper, O.H.; Fuller, L.C.

    1977-01-01

    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 of several nuclear steam supply systems in the small and intermediate size range. Detail studies are reported of the Consolidated Nuclear Steam Generator (CNSG), a 313MW(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 1000MW(th), which are envisioned for utility operation for supply of electric power and steam. Design studies of a 500MW(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. (author)

  15. 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…

  16. Automated reactor protection testing saves time and avoids errors

    International Nuclear Information System (INIS)

    Raimondo, E.

    1990-01-01

    When the Pressurized Water Reactor units in the French 900MWe series were designed, the instrumentation and control systems were equipped for manual periodic testing. Manual reactor protection system testing has since been successfully replaced by an automatic system, which is also applicable to other instrumentation testing. A study on the complete automation of process instrumentation testing has been carried out. (author)

  17. TREAT [Transient Reactor Test Facility] reactor control rod scram system simulations and testing

    International Nuclear Information System (INIS)

    Solbrig, C.W.; Stevens, W.W.

    1990-01-01

    Air cylinders moving heavy components (100 to 300 lbs) at high speeds (above 300 in/sec) present a formidable end-cushion-shock problem. With no speed control, the moving components can reach over 600 in/sec if the air cylinder has a 5 ft stroke. This paper presents an overview of a successful upgrade modification to an existing reactor control rod drive design using a computer model to simulate the modified system performance for system design analysis. This design uses a high speed air cylinder to rapidly insert control rods (278 lb moved 5 ft in less than 300 msec) to scram an air-cooled test reactor. Included is information about the computer models developed to simulate high-speed air cylinder operation and a unique new speed control and end cushion design. A patent application is pending with the US Patent ampersand Trade Mark Office for this system (DOE case number S-68,622). The evolution of the design, from computer simulations thru operational testing in a test stand (simulating in-reactor operating conditions) to installation and use in the reactor, is also described. 6 figs

  18. Preliminary Design Concept for a Reactor-internal CRDM

    International Nuclear Information System (INIS)

    Lee, Jae Seon; Kim, Jong Wook; Kim, Tae Wan; Choi, Suhn; Kim, Keung Koo

    2013-01-01

    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

  19. Emergency operating procedures guidelines for pressurized water reactors - a progress report

    International Nuclear Information System (INIS)

    Lyon, W.C.

    1984-01-01

    Emergency Operating Procedures (EOPs) contain the instructions the operator will follow to control a nuclear plant whenever a condition exists that potentially jeopardizes the fuel cladding, the reactor coolant system (RCS) pressure boundary, or the containment. The EOPs are prepared from guidelines which contain the major operator instructions that will be in the EOPs. Guidelines have been prepared by owners' groups having Babcock and Wilcox (BandW), Combustion Engineering (CE), General Electric (GE), and Westinghouse (W) plants. These guidelines cover many aspects of full power operation. Future effort is anticipated to complete coverage of transient events, including severe accidents, all power conditions, and shutdown. This paper describes the philosophy which has guided NRC technical review of guidelines, progress achieved in providing comprehensive coverage of emergency conditions for PWRs, and anticipated future technical activities

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

    International Nuclear Information System (INIS)

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

    1983-06-01

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

  1. The combined use of test reactor experiments and power reactor tests for the development of PCI-resistant fuel

    International Nuclear Information System (INIS)

    Junkrans, S.; Vesterlund, G.; Vaernild, O.

    1980-01-01

    The theme of this paper is that for development of PCI-resistant fuel acceptable from the commercial and licensing aspects, extensive and time-consuming work is needed both in a test reactor and in power reactors. The test reactor is necessary for ramp testing to power levels not allowed in power reactors and with the aim of generating fuel failures. It is also used for other special irradiation experiments. The access to power reactors is necessary to generate information on performance in a real LWR core and to incubate at a reasonable cost the large amount of rods required for test reactor ramping. Selected results from the ASEA-ATOM work are used to support these conclusions. (author)

  2. Reactor numerical simulation and hydraulic test research

    International Nuclear Information System (INIS)

    Yang, L. S.

    2009-01-01

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

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

    International Nuclear Information System (INIS)

    Grounes, M.

    1998-01-01

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

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

    International Nuclear Information System (INIS)

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

    2004-01-01

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

  5. Safety re-assessment of AECL test and research reactors

    International Nuclear Information System (INIS)

    Winfield, D.J.

    1990-01-01

    Atomic Energy of Canada Limited currently has four operating engineering test/research reactors of various sizes and ages; a new isotope-production reactor Maple-X10, under construction at Chalk River Nuclear Laboratories (CRNL), and a heating demonstration reactor, SDR, undergoing high-power commissioning at Whiteshell Nuclear Research Establishment (WNRE). The company is also performing design studies of small reactors for hot water and electricity production. The older reactors are ZED-2, PTR, NRX, and NRU; these range in age from 42 years (NRX) to 29 years (ZED-2). Since 1984, limited-scope safety re-assessments have been underway on three of these reactors (ZED-2, NRX AND NRU). ZED-2 and PTR are operated by the Reactor Physics Branch; all other reactors are operated by the respective site Reactor Operations Branches. For the older reactors the original safety reports produced were entirely deterministic in nature and based on the design-basis accident concept. The limited scope safety re-assessments for these older reactors, carried out over the past 5 years, have comprised both quantitative probabilistic safety-assessment techniques, such as event tree and fault analysis, and/or qualitative techniques, such as failure mode and effect analysis. The technique used for an individual assessment was dependent upon the specific scope required. This paper discusses the types of analyses carried out, specific insights/recommendations resulting from the analysis, and the plan for future analysis. In addition, during the last four years safety assessments have been carried out on the new isotope-, heat-, and electricity-producing reactors, as part of the safety design review, commissioning and licensing activities

  6. Development and testing of control rod drives for ship reactors

    International Nuclear Information System (INIS)

    Bruelheide, K.; Mundt, D.; Peters, C.-H.; Manthey, H.-J.

    1978-01-01

    The following paper deals with the development and testings of a new control rod drive design for marine reactors. Starting from the good operating experience with the advanced pressurized water reactor (FDR) of the NS OTTO HAHN a control rod drive system with an hermetically sealed drive principle was developed. A prototype control rod drive system was put through extensive tests and developed ready for standard production at the 'Gesellschaft fuer Kernenergieverwertung in Schiffbau und Schiffahrt'

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

    International Nuclear Information System (INIS)

    1993-11-01

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

  8. Electrical system regulations of the IEA-R1 reactor

    International Nuclear Information System (INIS)

    Mello, Jose Roberto de; Madi Filho, Tufic

    2013-01-01

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

  9. Current and prospective fuel test programmes in the MIR reactor

    Energy Technology Data Exchange (ETDEWEB)

    Izhutov, A.L.; Burukin, A.V.; Iljenko, S.A.; Ovchinnikov, V.A.; Shulimov, V.N.; Smirnov, V.P. [State Scientific Centre of Russia Research Institute of Atomic Reactors, Ulyanovsk region (Russian Federation)

    2007-07-01

    MIR reactor is a heterogeneous thermal reactor with a moderator and a reflector made of metal beryllium, it has a channel-type design and is placed in a water pool. MIR reactor is mainly designed for testing fragments of fuel elements and fuel assemblies (FA) of different nuclear power reactor types under normal (stationary and transient) operating conditions as well as emergency situations. At present six test loop facilities are being operated (2 PWR loops, 2 BWR loops and 2 steam coolant loops). The majority of current fuel tests is conducted for improving and upgrading the Russian PWR fuel, these tests involve issues such as: -) long term tests of short-size rods with different modifications of cladding materials and fuel pellets; -) further irradiation of power plant re-fabricated and full-size fuel rods up to achieving 80 MW*d/kg U; -) experiments with leaking fuel rods at different burnups and under transient conditions; -) continuation of the RAMP type experiments at high burnup of fuel; and -) in-pile tests with simulation of LOCA and RIA type accidents. Testing of the LEU (low enrichment uranium) research reactor fuel is conducted within the framework of the RERTR programme. Upgrading of the gas cooled and steam cooled loop facilities is scheduled for testing the HTGR fuel and sub-critical water-cooled reactor, correspondingly. The present paper describes the major programs of the WWER high burn-up fuel behavior study in the MIR reactor, capabilities of the applied techniques and some results of the performed irradiation tests. (authors)

  10. Rise-to-power test in High Temperature Engineering Test Reactor. Test progress and summary of test results up to 30 MW of reactor thermal power

    International Nuclear Information System (INIS)

    Nakagawa, Shigeaki; Fujimoto, Nozomu; Shimakawa, Satoshi

    2002-08-01

    The High Temperature Engineering Test Reactor (HTTR) is a graphite moderated and gas cooled reactor with the thermal power of 30 MW and the reactor outlet coolant temperature of 850degC/950degC. Rise-to-power test in the HTTR was performed from April 23rd to June 6th in 2000 as phase 1 test up to 10 MW in the rated operation mode, from January 29th to March 1st in 2001 as phase 2 test up to 20 MW in the rated operation mode and from April 14th to June 8th in 2001 as phase 3 test up to 20 MW in the high temperature test the mechanism of the reactor outlet coolant temperature becomes 850degC at 30 MW in the rated operation mode and 950degC in the high temperature test operation mode. Phase 4 rise-to-power test to achieve the thermal reactor power of 30 MW started on October 23rd in 2001. On December 7th in 2001 it was confirmed that the thermal reactor power and the reactor outlet coolant temperature reached to 30 MW and 850degC respectively in the single loaded operation mode in which only the primary pressurized water cooler is operating. Phase 4 test was performed until March 6th in 2002. JAERI (Japan Atomic Energy Research Institute) obtained the certificate of the pre-operation test from MEXT (Ministry of Education Culture Sports Science and Technology) after all the pre-operation tests by MEXT were passed successfully with the reactor transient test at an abnormal event as a final pre-operation test. From the test results of the rise-up-power test up to 30 MW in the rated operation mode, performance of the reactor and cooling system were confirmed, and it was also confirmed that an operation of reactor facility can be performed safely. Some problems to be solved were found through the tests. By solving them, the reactor operation with the reactor outlet coolant temperature of 950degC will be achievable. (author)

  11. Simulating Neutronic Core Parameters in a Research and Test Reactor

    International Nuclear Information System (INIS)

    Selim, H.K.; Amin, E.A.; Koutb, M.E.

    2011-01-01

    The present study proposes an Artificial Neural Network (ANN) modeling technique that predicts the control rods positions in a nuclear research reactor. The neutron, flux in the core of the reactor is used as the training data for the neural network model. The data used to train and validate the network are obtained by modeling the reactor core with the neutronic calculation code: CITVAP. The type of the network used in this study is the feed forward multilayer neural network with the backpropagation algorithm. The results show that the proposed ANN has good generalization capability to estimate the control rods positions knowing neutron flux for a research and test reactor. This method can be used to predict critical control rods positions to be used for reactor operation after reload

  12. Permeated defect detecting test method and device in reactor

    International Nuclear Information System (INIS)

    Sakurai, Yoshishige.

    1996-01-01

    The present invention provides a method of and a device capable of performing a test for entire inner surfaces of the reactor upon periodical inspection of a BWR type reactor while sufficiently taking countermeasures for radiation rays into consideration. Namely, the present invention comprises following steps. (1) A provisional step for taking a shroud head of a reactor core shroud and incore structural components above and below the shroud out of the reactor, discharging reactor water and water tightly closing openings such as reactor wall perforation holes, (2) a pretreatment step for washing exposed inner surfaces of the reactor and peeling deteriorated materials, (3) a first drying step for drying portions washed and peeled in the step (2), (4) a permeation step for applying a permeation liquid of a defect detecting medium on the exposed inner surfaces of the reactor, (5) a permeation liquid removing step for removing the an excess permeation liquid in the step (4), (6) a second drying step for drying corresponding portions after performing the step (5), and (7) a flaw detecting step for optically observing the corresponding portions after performing the step (6) and detecting flaws. (I.S.)

  13. Remote inspection manipulators for AGR II: Babcock Power's interstitial manipulator

    International Nuclear Information System (INIS)

    Whyley, S.R.

    1985-01-01

    The interstitial manipulator has been designed and built by Babcock Power for the remote visual inspection of AGR II reactors at Heysham and Torness. Its five drives are operated from a console local to the manipulator on the pile cap, or from a similar console located remotely. The need to operate from an interstitial ISI standpipe has restricted the size of the components entering the reactor, and this has consequently provided the major design constraint. A detailed structural assessment of the manipulator was carried out to demonstrate the ability to operate with payloads in excess of the largest camera weight of 13.6 kg. The manipulator finite element model was also used to determine static deflections, and, as a consequence, has provided data from which the control system is able to predict accurately the camera's position. Other computer aided design techniques have enabled the step by step sequences of manipulator deployment, in the restricted space available, to be successfully demonstrated. (author)

  14. Safety requirements, facility user needs, and reactor concepts for a new Broad Application Test Reactor

    International Nuclear Information System (INIS)

    Ryskamp, J.M.; Liebenthal, J.L.; Denison, A.B.; Fletcher, C.D.

    1992-07-01

    This report describes the EG ampersand G Laboratory Directed Research and Development Program (LDRD) Broad Application Test Reactor (BATR) Project that was conducted in fiscal year 1991. The scope of this project was divided into three phases: a project process definition phase, a requirements development phase, and a preconceptual reactor design and evaluation phase. Multidisciplinary teams of experts conducted each phase. This report presents the need for a new test reactor, the project process definition, a set of current and projected regulatory compliance and safety requirements, a set of facility user needs for a broad range of projected testing missions, and descriptions of reactor concepts capable of meeting these requirements. This information can be applied to strategic planning to provide the Department of Energy with management options

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

    International Nuclear Information System (INIS)

    Staudinger, Deborah K.

    2004-01-01

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

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

    Science.gov (United States)

    Warwick, Peter D.

    2017-09-27

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

  17. Design and testing of reactors for 735 kV

    Energy Technology Data Exchange (ETDEWEB)

    Erb, W; Kraaij, D J

    1965-11-01

    The design and testing of five large, single phase shunt reactors rated either 110 or 55 MVAR, supplied for the 735 kV system of the Quebec Hydro Electric Commission which came into operation in the autumn of 1965 are described. As these reactors are permanently connected to the transmission lines, their losses must be considered as being continuously present and must be determined exactly. In addition to the use of a new bridge method, the losses were also measured calorimetrically for the purpose of comparison, the agreement between the two tests being remarkably good. The impulse tests with full wave and chopped wave are subsequently described.

  18. Reduced enrichment for research and test reactors: Proceedings

    International Nuclear Information System (INIS)

    1993-07-01

    The 15th annual Reduced Enrichment for Research and Test Reactors (RERTR) international meeting was organized by Ris oe National Laboratory in cooperation with the International Atomic Energy Agency and Argonne National Laboratory. The topics of the meeting were the following: National Programs, Fuel Fabrication, Licensing Aspects, States of Conversion, Fuel Testing, and Fuel Cycle. Individual papers have been cataloged separately

  19. Reduced enrichment for research and test reactors: Proceedings

    Energy Technology Data Exchange (ETDEWEB)

    1993-07-01

    The 15th annual Reduced Enrichment for Research and Test Reactors (RERTR) international meeting was organized by Ris{o} National Laboratory in cooperation with the International Atomic Energy Agency and Argonne National Laboratory. The topics of the meeting were the following: National Programs, Fuel Fabrication, Licensing Aspects, States of Conversion, Fuel Testing, and Fuel Cycle. Individual papers have been cataloged separately.

  20. Design and testing of integrated circuits for reactor protection channels

    International Nuclear Information System (INIS)

    Battle, R.E.; Vandermolen, R.I.; Jagadish, U.; Swail, B.K.; Naser, J.

    1995-01-01

    Custom and semicustom application-specific integrated circuit design and testing methods are investigated for use in research and commercial nuclear reactor safety systems. The Electric Power Research Institute and Oak Ridge National Laboratory are working together through a cooperative research and development agreement to apply modern technology to a nuclear reactor protection system. The purpose of this project is to demonstrate to the nuclear industry an alternative approach for new or upgrade reactor protection and safety system signal processing and voting logic. Motivation for this project stems from (1) the difficulty of proving that software-based protection systems are adequately reliable, (2) the obsolescence of the original equipment, and (3) the improved performance of digital processing. A demonstration model for protection system of PWR reactor has been designed and built

  1. 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.

  2. Chinese nuclear heating test reactor and demonstration plant

    International Nuclear Information System (INIS)

    Wang Dazhong; Ma Changwen; Dong Duo; Lin Jiagui

    1992-01-01

    In this report the importance of nuclear district heating is discussed. From the viewpoint of environmental protection, uses of energy resources and transport, the development of nuclear heating in China is necessary. The development program of district nuclear heating in China is given in the report. At the time being, commissioning of the 5 MW Test Heating Reactor is going on. A 200 MWt Demonstration Plant will be built. In this report, the main characteristics of these reactors are given. It shows this type of reactor has a high inherent safety. Further the report points out that for this type of reactor the stability is very important. Some experimental results of the driving facility are included in the report. (orig.)

  3. 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).

  4. Research and Test Reactor Fuel Elements (RTRFE)

    International Nuclear Information System (INIS)

    Pace, Brett W.; Marinak, Edward A.

    1999-01-01

    BWX Technologies Inc. (BWXT) has experienced several production improvements over the past year. The homogeneity yields in 4.8 gU/cc U 3 Si 2 plates have increased over last year's already high yields. Through teamwork and innovative manufacturing techniques, maintaining high quality surface finishes on plates and elements is becoming easier and less expensive. Currently, BWXT is designing a fabrication development plan to reach a fuel loading of 9 gU/cc within 2 - 4 years. This development will involve a step approach requested by ANL to produce plates using U-8Mo at a loading of 6 gU/cc first and qualify the fuel at those levels. In achieving the goal of a very high-density fuel loading of 9 gU/cc, BWXT is considering employing several new, state of the art, ultrasonic testing techniques for fuel core evaluation. (author)

  5. Improving the proliferation resistance of research and test reactors

    International Nuclear Information System (INIS)

    Lewis, R.A.

    1978-01-01

    Elimination, or substantial reduction, of the trade in unirradiated highly-enriched fuel elements for research and test reactors would significantly reduce the proliferation risk associated with the current potential for diversion of these materials. To this end, it is the long-term goal of U.S. policy to fuel all new and existing research and test reactors with uranium of less-than-20% enrichment (but substantially greater than natural) excepting, perhaps, only a small number of high-power, high-performance, reactors. The U.S. development program for enrichment reduction in research and test reactor designs currently using 90-93% enriched uranium is based on the practical criterion that enrichment reduction should not cause significant flux performance (flux per unit power) or burnup performance degradation relative to the unmodified reactor design. To first order, this implies the requirement that the 235 U loading in the reduced-enrichment fuel elements be the same as the 235 U loading in the 90-93% enriched fuel elements. This can be accomplished by substitution of higher uranium density fuel technology for currently-used fuel technology in the fuel meat volume of the current fuel element design and/or by increasing the usable fuel meat volume. For research and test reactors of power greater than 5-10 megawatts, fuel technology does not currently exist that would permit enrichment reductions to below 20% utilizing this criterion. A program is now beginning in the U.S. to develop the necessary fuel technology. Currently-proven fuel technology is capable, however, of accommodating enrichment reductions to the 30-45% range (from 90-93%) for many reactors in the 5-50MW range. Accordingly the U.S. is proposing to convert existing reactors (and new designs) in the 5-50MW range from the use of highly-enriched fuel to the use of 30-45% enriched fuel, and reactors of less that about 5MW to less-than-20% enrichment, wherever this can be done without significant

  6. Present status and future perspective of research and test reactors in JAERI

    International Nuclear Information System (INIS)

    Baba, Osamu; Kaieda, Keisuke

    1999-01-01

    Since 1957, Japan Atomic Energy Research Institute (JAERI) has constructed several research and test reactors to fulfil a major role in the study of nuclear energy and fundamental research. At present, four reactors, the Japan Research Reactor No. 3 and No. 4 (JRR-3M and JRR-4 respectively), the Japan Materials Testing Reactor (JMTR) and the Nuclear Safety Research Reactor (NSRR), are in operation, and a new High Temperature Engineering Test Reactor (HTTR) has reached first criticality and is waiting for the power-up test. This paper introduce these reactors and describe their present operational status. The recent tendency of utilization and future perspectives are also reported. (author)

  7. Present status and future perspective of research and test reactors in JAERI

    Energy Technology Data Exchange (ETDEWEB)

    Baba, Osamu [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment; Kaieda, Keisuke

    1999-08-01

    Since 1957, Japan Atomic Energy Research Institute (JAERI) has constructed several research and test reactors to fulfil a major role in the study of nuclear energy and fundamental research. At present, four reactors, the Japan Research Reactor No. 3 and No. 4 (JRR-3M and JRR-4 respectively), the Japan Materials Testing Reactor (JMTR) and the Nuclear Safety Research Reactor (NSRR), are in operation, and a new High Temperature Engineering Test Reactor (HTTR) has reached first criticality and is waiting for the power-up test. This paper introduce these reactors and describe their present operational status. The recent tendency of utilization and future perspectives are also reported. (author)

  8. Reactor protection system with automatic self-testing and diagnostic

    International Nuclear Information System (INIS)

    Gaubatz, D.C.

    1996-01-01

    A reactor protection system is disclosed having four divisions, with quad redundant sensors for each scram parameter providing input to four independent microprocessor-based electronic chassis. Each electronic chassis acquires the scram parameter data from its own sensor, digitizes the information, and then transmits the sensor reading to the other three electronic chassis via optical fibers. To increase system availability and reduce false scrams, the reactor protection system employs two levels of voting on a need for reactor scram. The electronic chassis perform software divisional data processing, vote 2/3 with spare based upon information from all four sensors, and send the divisional scram signals to the hardware logic panel, which performs a 2/4 division vote on whether or not to initiate a reactor scram. Each chassis makes a divisional scram decision based on data from all sensors. Automatic detection and discrimination against failed sensors allows the reactor protection system to automatically enter a known state when sensor failures occur. Cross communication of sensor readings allows comparison of four theoretically ''identical'' values. This permits identification of sensor errors such as drift or malfunction. A diagnostic request for service is issued for errant sensor data. Automated self test and diagnostic monitoring, sensor input through output relay logic, virtually eliminate the need for manual surveillance testing. This provides an ability for each division to cross-check all divisions and to sense failures of the hardware logic. 16 figs

  9. B and W owners group scram reduction efforts

    International Nuclear Information System (INIS)

    Rose, S.T.

    1985-01-01

    Reducing the frequency of reactor scrams is an important and highly visible task. Scram frequency is one indicator of how consistently a unit is operated within desired bounds and hence is a key performance indicator. To be successful and efficient in this undertaking, diligent effort by the individual utilities should be complimented by collective action to resolve generic problems. The Babcock and Wilcox (B and W) Owners Group is committed to improving the productivity of its operating units and, in particular, to reducing the number of scrams experienced at the units. Toward this goal, the owners group has undertaken several initiatives to improve the reliability and performance of systems that historically have reactor trips. This paper describes those efforts and how they were identified

  10. Scheduling and recording of reactor maintenance and testing by computer

    International Nuclear Information System (INIS)

    Gray, P.L.

    1975-01-01

    The use of a computer program, Maintenance Information and Control (MIAC), at the Savannah River Laboratory (SRL) assists a small operating staff in maintaining three research reactors and a subcritical facility. The program schedules and defines preventive maintenance, schedules required periodic tests, logs repair and cost information, specifies custodial and service responsibilities, and provides equipment maintenance history, all with a minimum of record-keeping

  11. Design and testing of integrated circuits for reactor protection channels

    International Nuclear Information System (INIS)

    Battle, R.E.; Vandermolen, R.I.; Jagadish, U.; Swail, B.K.; Naser, J.; Rana, I.

    1995-01-01

    Custom and semicustom application-specific integrated circuit design and testing methods are investigated for use in research and commercial nuclear reactor safety systems. The Electric Power Research Institute and Oak Ridge National Laboratory are working together through a cooperative research and development agreement to apply modern technology to a nuclear reactor protection system. Purpose of this project is to demonstrate to the nuclear industry an alternative approach for new or upgrade reactor protection and safety system signal processing and voting logic. Motivation for this project stems from (1) the difficulty of proving that software-based protection systems are adequately reliable, (2) the obsolescence of the original equipment, and (3) the improved performance of digital processing

  12. Education and training by utilizing irradiation test reactor simulator

    International Nuclear Information System (INIS)

    Eguchi, Shohei; Koike, Sumio; Takemoto, Noriyuki; Tanimoto, Masataka; Kusunoki, Tsuyoshi

    2016-01-01

    The Japan Atomic Energy Agency, at its Japan Materials Testing Reactor (JMTR), completed an irradiation test reactor simulator in May 2012. This simulator simulates the operation, irradiation test, abnormal transient change during operation, and accident progress events, etc., and is able to perform operation training on reactor and irradiation equipment corresponding to the above simulations. This simulator is composed of a reactor control panel, process control panel, irradiation equipment control panel, instructor control panel, large display panel, and compute server. The completed simulator has been utilized in the education and training of JMTR operators for the purpose of the safe and stable operation of JMTR and the achievement of high operation rate after resuming operation. For the education and training, an education and training curriculum has been prepared for use in not only operation procedures at the time of normal operation, but also learning of fast and accurate response in case of accident events. In addition, this simulator is also being used in operation training for the purpose of contributing to the cultivation of human resources for atomic power in and out of Japan. (A.O.)

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

    International Nuclear Information System (INIS)

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

    1976-08-01

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

  14. Situation of test and research reactors' spent fuels

    International Nuclear Information System (INIS)

    Shimizu, Kenichi; Uchiyama, Junzo; Sato, Hiroshi

    1996-01-01

    The U.S. DOE decided a renewal Off-Site Fuel Policy for stopping to spread a highly enriched uranium which was originally enriched at the U.S., the policy declared that to receive all HEU spent fuels from Test and Research reactors in all the world. In Japan, under bilateral agreement of cooperation between the government of the United States and the government of Japan concerning peaceful uses of nuclear energy, the highly enriched uranium of Test and Research Reactors' fuels was purchased from the U.S. and the fuels had been manufactured in Japan, America, Germany and France. On the other hand, a former president of the U.S. J. Carter proposed that to convert the fuels from HEU to LEU concerning a nonproliferation of nuclear materials in 1978, and Japan absolutely supported this policy. Under this condition, the U.S. stopped to receive the spent fuels from the other countries concerning legal action to the Off-Site Fuels Policy. As a result, the spent fuels are increasing, and to cross to each reactor's storage capacity, and if this policy start, a faced crisis of Test and Research Reactors will be avoided. (author)

  15. Safety analysis calculations for research and test reactors

    Energy Technology Data Exchange (ETDEWEB)

    Chen, S Y; MacDonald, R; MacFarlane, D [Argonne National Laboratory, Argonne, IL (United States)

    1983-08-01

    The goal of the RERTR (Reduced Enrichment in Research and Test Reactor) Program at ANL is to provide technical means for conversion of research and test reactors from HEU (High-Enrichment Uranium) to LEU (Low-Enrichment Uranium) fuels. In exploring the feasibility of conversion, safety considerations are a prime concern; therefore, safety analyses must be performed for reactors undergoing the conversion. This requires thorough knowledge of the important safety parameters for different types of reactors for both HEU and LEU fuel. Appropriate computer codes are needed to predict transient reactor behavior under postulated accident conditions. In this discussion, safety issues for the two general types of reactors i.e., the plate-type (MTR-type) reactor and the rod-type (TRIGA-type) reactor, resulting from the changes associated with LEU vs. HEU fuels, are explored. The plate-type fuels are typically uranium aluminide (UAl{sub x}) compounds dispersed in aluminum and clad with aluminum. Moderation is provided by the water coolant. Self shut-down reactivity coefficients with EU fuel are entirely a result of coolant heating, whereas with LEU fuel there is an additional shut down contribution provided by the direct heating of the fuel due to the Doppler coefficient. In contrast, the rod-type (TRIGA) fuels are mixtures of zirconium hydride, uranium, and erbium. This fuel mixture is formed into rods ( {approx} 1 cm diameter) and clad with stainless steel or Incoloy. In the TRIGA fuel the self-shutdown reactivity is more complex, depending on heating of the fuel rather than the coolant. The two most important mechanisms in providing this feedback are: spectral hardening due to neutron interaction with the ZrH moderator as it is heated and Doppler broadening of resonances in erbium and U-238. Since these phenomena result directly from heating of the fuel, and do not depend on heat transfer to the moderator/coolant, the coefficients are prompt acting. Results of transient

  16. Reduced enrichment for research and test reactors: Proceedings

    International Nuclear Information System (INIS)

    1988-05-01

    The international effort to develop new research reactor fuel materials and designs based on the use of low-enriched uranium, instead of highly-enriched uranium, has made much progress during the eight years since its inception. To foster direct communication and exchange of ideas among the specialist in this area, the Reduced Enrichment Research and Test Reactor (RERTR) Program, at the Argonne National Laboratory, sponsored this meeting as the ninth of a series which began in 1978. All previous meetings of this series are listed on the facing page. The focus of this meeting was on the LEU fuel demonstration which was in progress at the Oak Ridge Research (ORR) reactor, not far from where the meeting was held. The visit to the ORR, where a silicide LEU fuel with 4.8 g A/cm 3 was by then in routine use, illustrated how far work has progressed

  17. Reduced enrichment for research and test reactors: Proceedings

    Energy Technology Data Exchange (ETDEWEB)

    1988-05-01

    The international effort to develop new research reactor fuel materials and designs based on the use of low-enriched uranium, instead of highly-enriched uranium, has made much progress during the eight years since its inception. To foster direct communication and exchange of ideas among the specialist in this area, the Reduced Enrichment Research and Test Reactor (RERTR) Program, at the Argonne National Laboratory, sponsored this meeting as the ninth of a series which began in 1978. All previous meetings of this series are listed on the facing page. The focus of this meeting was on the LEU fuel demonstration which was in progress at the Oak Ridge Research (ORR) reactor, not far from where the meeting was held. The visit to the ORR, where a silicide LEU fuel with 4.8 g A/cm/sup 3/ was by then in routine use, illustrated how far work has progressed.

  18. Decontamination and Decommissioning of the Tokamak Fusion Test Reactor

    International Nuclear Information System (INIS)

    Perry, E.; Chrzanowski, J.; Rule, K.; Viola, M.; Williams, M.; Strykowsky, R.

    1999-01-01

    The Tokamak Fusion Test Reactor (TFTR) is a one-of-a-kind, tritium-fueled fusion research reactor that ceased operation in April 1997. The Decontamination and Decommissioning (D and D) of the TFTR is scheduled to occur over a period of three years beginning in October 1999. This is not a typical Department of Energy D and D Project where a facility is isolated and cleaned up by ''bulldozing'' all facility and hardware systems to a greenfield condition. The mission of TFTR D and D is to: (a) surgically remove items which can be re-used within the DOE complex, (b) remove tritium contaminated and activated systems for disposal, (c) clear the test cell of hardware for future reuse, (d) reclassify the D-site complex as a non-nuclear facility as defined in DOE Order 420.1 (Facility Safety) and (e) provide data on the D and D of a large magnetic fusion facility. The 100 cubic meter volume of the donut-shaped reactor makes it the second largest fusion reactor in the world. The record-breaking deuterium-tritium experiments performed on TFTR resulted in contaminating the vacuum vessel with tritium and activating the materials with 14 Mev neutrons. The total tritium content within the vessel is in excess of 7,000 Curies while dose rates approach 75 mRem/hr. These radiological hazards along with the size and shape of the Tokamak present a unique and challenging task for dismantling

  19. Advanced In-pile Instrumentation for Material and Test Reactors

    International Nuclear Information System (INIS)

    Rempe, J.L.; Knudson, D.L.; Daw, J.E.; Unruh, T.C.; Chase, B.M.; Davis, K.L.; Palmer, A.J.; Schley, R.S.

    2013-06-01

    The US Department of Energy sponsors the Advanced Test Reactor (ATR) National Scientific User Facility (NSUF) program to promote U.S. research in nuclear science and technology. By attracting new research users - universities, laboratories, and industry - the ATR NSUF facilitates basic and applied nuclear research and development, advancing U.S. energy security needs. A key component of the ATR NSUF effort is to design, develop, and deploy new in-pile instrumentation techniques that are capable of providing real-time measurements of key parameters during irradiation. This paper describes the strategy developed by the Idaho National Laboratory (INL) for identifying instrumentation needed for ATR irradiation tests and the program initiated to obtain these sensors. New sensors developed from this effort are identified; and the progress of other development efforts is summarized. As reported in this paper, INL staff is currently involved in several tasks to deploy real-time length and flux detection sensors, and efforts have been initiated to develop a crack growth test rig. Tasks evaluating 'advanced' technologies, such as fiber-optics based length detection and ultrasonic thermometers are also underway. In addition, specialized sensors for real-time detection of temperature and thermal conductivity are not only being provided to NSUF reactors, but are also being provided to several international test reactors. (authors)

  20. Fuels for research and test reactors, status review: July 1982

    International Nuclear Information System (INIS)

    Stahl, D.

    1982-12-01

    A thorough review is provided on nuclear fuels for steady-state thermal research and test reactors. The review was conducted to provide a documented data base in support of recent advances in research and test reactor fuel development, manufacture, and demonstration in response to current US policy on availability of enriched uranium. The review covers current fabrication practice, fabrication development efforts, irradiation performance, and properties affecting fuel utilization, including thermal conductivity, specific heat, density, thermal expansion, corrosion, phase stability, mechanical properties, and fission-product release. The emphasis is on US activities, but major work in Europe and elsewhere is included. The standard fuel types discussed are the U-Al alloy, UZrH/sub x/, and UO 2 rod fuels. Among new fuels, those given major emphasis include H 3 Si-Al dispersion and UO 2 caramel plate fuels

  1. Mechanical behaviour of the reactor vessel support of a pressurized water reactor: tests and analysis

    International Nuclear Information System (INIS)

    Bolvin, M.; L'huby, Y.; Quillico, J.J.; Humbert, J.M.; Thomas, J.P.; Hugenschmitt, R.

    1985-08-01

    The PWR reactor vessel is supported by a steel ring laying on the reactor pit. This support has to ensure a good behaviour of the vessel in the event of accidental conditions (earthquake and pipe rupture). A new evolution of the evaluation methods of the applied forces has shown a significant increase in the design loads used until now. In order to take into account these new forces, we carried out a test on a representative mock-up of the vessel support (scale 1/6). This test was performed by CEA, EDF and FRAMATOME. Several static equivalent forces were applied on the experimental mock-up. Displacements and strains were simultaneously recorded. The results of the test have enabled to justify the design of the pit and the ring, to show up a wide safety margin until the collapse of the structures and to check our hypothesis about the transmission of the forces between the ring and the pit

  2. Manufacturing and material properties of forgings for reactor pressure vessel of high temperature engineering test reactor

    International Nuclear Information System (INIS)

    Sato, I.; Suzuki, K.

    1994-01-01

    For the reactor pressure vessel (RPV) of high temperature engineering test reactor (HTTR) which has been developed by Japan Atomic Energy Research Institute (JAERI), 2 1/4Cr-1Mo steel is used first in the world. Material confirmation test has been carried out to demonstrate good applicability of forged low Si 2 1/4Cr-1Mo steel to the RPV of HTTR. Recently, JSW has succeeded in the manufacturing of large size ring forgings and large size forged cover dome integrated with nozzles for stand pipe for the RPV. This paper describes the results of the material confirmation test as well as the manufacturing and material properties of the large forged cover dome integrated with nozzles for stand pipe. (orig.)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-05-01

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

  4. Test reactor risk assessment methodology

    International Nuclear Information System (INIS)

    Jennings, R.H.; Rawlins, J.K.; Stewart, M.E.

    1976-04-01

    A methodology has been developed for the identification of accident initiating events and the fault modeling of systems, including common mode identification, as these methods are applied in overall test reactor risk assessment. The methods are exemplified by a determination of risks to a loss of primary coolant flow in the Engineering Test Reactor

  5. An extended conventional fuel cycle for the B and W mPower{sup TM} small modular nuclear reactor

    Energy Technology Data Exchange (ETDEWEB)

    Scarangella, M. J. [Babcock and Wilcox Company, 109 Ramsey Place, Lynchburg, VA 24502 (United States)

    2012-07-01

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

  6. Instrument accuracy in reactor vessel inventory tracking systems

    International Nuclear Information System (INIS)

    Anderson, J.L.; Anderson, R.L.; Morelock, T.C.; Hauang, T.L.; Phillips, L.E.

    1986-01-01

    Instrumentation needs for detection of inadequate core cooling. Studies of the Three Mile Island accident identified the need for additional instrumentation to detect inadequate core cooling (ICC) in nuclear power plants. Industry studies by plant owners and reactor vendors supported the conclusion that improvements were needed to help operators diagnose the approach to or existence of ICC as well as to provide more complete information for operator control of safety injection flow to minimize the consequences of such an accident. In 1980, the US Nuclear Regulatory Commission (NRC) required further studies by the industry and described ICC instrumentation design requirements that included human factors and environmental considerations. On December 10, 1982, NRC issued to Babcock and Wilcox (B and W) licensees orders for Modification of License and transmitted to pressurized water reactor licensees Generic Letter 82-28 to inform them of the revised NRC requirements. The instrumentation requirements include upgraded subcooling margin monitors (SMM), upgraded core exit thermocouples (CET), and installation of a reactor coolant inventory tracking system. NRC Regulatory Guide 1.97, which covers accident monitoring instrumentation, was revised (Rev. 3) to be consistent with the requirements of item II.F.2 of NUREG-0737

  7. The Advanced Test Reactor Irradiation Facilities and Capabilities

    International Nuclear Information System (INIS)

    S. Blaine Grover; Raymond V. Furstenau

    2007-01-01

    The Advanced Test Reactor (ATR) is one of the world's premiere test reactors for performing long term, high flux, and/or large volume irradiation test programs. The ATR is a very versatile facility with a wide variety of experimental test capabilities for providing the environment needed in an irradiation experiment. These different capabilities include passive sealed capsule experiments, instrumented and/or temperature-controlled experiments, and pressurized water loop experiment facilities. The ATR has enhanced capabilities in experiment monitoring and control systems for instrumented and/or temperature controlled experiments. The control systems utilize feedback from thermocouples in the experiment to provide a custom blended flowing inert gas mixture to control the temperature in the experiments. Monitoring systems have also been utilized on the exhaust gas lines from the experiment to monitor different parameters, such as fission gases for fuel experiments, during irradiation. ATR's unique control system provides axial flux profiles in the experiments, unperturbed by axially positioned control components, throughout each reactor operating cycle and over the duration of test programs requiring many years of irradiation. The ATR irradiation positions vary in diameter from 1.6 cm (0.625 inches) to 12.7 cm (5.0 inches) over an active core length of 122 cm (48.0 inches). Thermal and fast neutron fluxes can be adjusted radially across the core depending on the needs of individual test programs. This paper will discuss the different irradiation capabilities available and the cost/benefit issues related to each capability. Examples of different experiments will also be discussed to demonstrate the use of the capabilities and facilities at ATR for performing irradiation experiments

  8. Test reactors in the world

    International Nuclear Information System (INIS)

    Corella, M.R.; Gomez Alonso, M.

    1983-01-01

    INFCE work on research reactor core conversion from HEU to LEU, attracted a raising interest on this type of nuclear reactors. In this context, the present work shows a compilation of worldwide research and test nuclear reactors, now in operation, under construction, or planned, as well as decommissioned reactors (tables A to F). Brief descriptions of these reactors are included in tables G to L. In table M a summary view of reactors with power level between 10 and 30 MWt is shown. Attention is focused on that power range, as it has been considered in very preliminar studies for a new research reactor. Almost all data have been obtained from current available bibliography. (author)

  9. Implosion and staging systems for a Scyllac Fusion Test Reactor

    International Nuclear Information System (INIS)

    Gribble, R.F.; Linford, R.K.; Thomassen, K.I.

    1976-01-01

    The implosion heating and adiabatic compression processes will be separated in future theta pinch devices. The circuit to achieve the fast implosion heating and power crowbar (staging) for the Scyllac Fusion Test Reactor is described here. The plasma is very tightly coupled to the circuit and presents a varying inductive load. Computer-aided circuit designs which achieve a programmed magnetic field waveform are described. The field approximates a two-step waveform, on-off-on, which is ideal for achieving the large initial plasma radius needed for stability. The components for the circuits have been developed and are being tested in experiments at Los Alamos

  10. Implosion and staging systems for a Scyllac fusion test reactor

    International Nuclear Information System (INIS)

    Gribble, R.F.; Linford, R.K.; Thomassen, K.I.

    1975-01-01

    The implosion heating and adiabatic compression processes will be separated in future theta pinch devices. The circuit to achieve the fast implosion heating and power crowbar (staging) for the Scyllac Fusion Test Reactor is described here. The plasma is very tightly coupled to the circuit and presents a varying inductive load. Computer-aided circuit designs which achieve a programmed magnetic field waveform are described. The field approximates a two-step waveform, on-off-on, which is ideal for achieving the large initial plasma radius needed for stability. The components for the circuits have been developed and are being tested in experiments at Los Alamos. (auth)

  11. Simulator for materials testing reactors

    International Nuclear Information System (INIS)

    Takemoto, Noriyuki; Sugaya, Naoto; Ohtsuka, Kaoru; Hanakawa, Hiroki; Onuma, Yuichi; Hosokawa, Jinsaku; Hori, Naohiko; Kaminaga, Masanori; Tamura, Kazuo; Hotta, Kohji; Ishitsuka, Tatsuo

    2013-06-01

    A real-time simulator for both reactor and irradiation facilities of a materials testing reactor, “Simulator of Materials Testing Reactors”, was developed for understanding reactor behavior and operational training in order to utilize it for nuclear human resource development and to promote partnership with developing countries which have a plan to introduce nuclear power plant. The simulator is designed based on the JMTR (Japan Materials Testing Reactor), and it simulates operation, irradiation tests and various kinds of anticipated operational transients and accident conditions caused by the reactor and irradiation facilities. The development of the simulator was sponsored by the Japanese government as one of the specialized projects of advanced research infrastructure in order to promote basic as well as applied researches. This report summarizes the simulation components, hardware specification and operation procedure of the simulator. (author)

  12. Present status and future perspectives of research and test reactor in Japan

    International Nuclear Information System (INIS)

    Kaneko, Yoshihiko; Kaieda, Keisuke

    2000-01-01

    Since 1957, Japan Atomic Energy Research Institute (JAERI) has constructed several research and test reactors to fulfill a major role in the study of nuclear energy and fundamental research. At present four reactors, the Japan Research Reactor No. 3 and No. 4 (JRR-3M and JRR-4 respectively), the Japan Materials Testing Reactor (JMTR) and the Nuclear Safety Research Reactor (NSRR) are in operation, and a new High Temperature Engineering Test Reactor (HTTR) has recently reached first criticality and now in the power up test. In 1966, the Kyoto University built the Kyoto University Reactor (KUR) and started its operation for joint use program of the Japanese universities. This paper introduces these reactors and describes their present operational status and also efforts for aging management. The recent tendency of utilization and future perspectives is also reported. (author)

  13. Present status and future perspectives of research and test reactor in Japan

    Energy Technology Data Exchange (ETDEWEB)

    Kaneko, Yoshihiko [Atomic Energy Research Laboratory, Musashi Institute of Technology, Kawasaki, Kanagawa (Japan); Kaieda, Keisuke [Department of Research Reactor, Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan)

    2000-10-01

    Since 1957, Japan Atomic Energy Research Institute (JAERI) has constructed several research and test reactors to fulfill a major role in the study of nuclear energy and fundamental research. At present four reactors, the Japan Research Reactor No. 3 and No. 4 (JRR-3M and JRR-4 respectively), the Japan Materials Testing Reactor (JMTR) and the Nuclear Safety Research Reactor (NSRR) are in operation, and a new High Temperature Engineering Test Reactor (HTTR) has recently reached first criticality and now in the power up test. In 1966, the Kyoto University built the Kyoto University Reactor (KUR) and started its operation for joint use program of the Japanese universities. This paper introduces these reactors and describes their present operational status and also efforts for aging management. The recent tendency of utilization and future perspectives is also reported. (author)

  14. Operation, test, research and development of the High Temperature Engineering Test Reactor (HTTR). FY2013

    International Nuclear Information System (INIS)

    2014-12-01

    The High Temperature Engineering Test Reactor (HTTR), a graphite-moderated and helium gas-cooled reactor with 30MW of thermal power, constructed at the Oarai Research and Development Center of the Japan Atomic Energy Agency (JAEA) is the first high-temperature gas-cooled reactor (HTGR) in Japan. The HTTR was attained at the full power operation of 30MW in December 2001 and achieved the 950degC of outlet coolant temperature at the outside the reactor pressure vessel in June 2004. To establish and upgrade basic technologies for HTGRs, we have obtained demonstration test data necessary for several R and Ds, and accumulated operation and maintenance experience of HTGRs throughout the HTTR's operation such as rated power operations, safety demonstration tests and long-term high temperature operations, and so on. In fiscal year 2013, we started to prepare the application document of reactor installation license for the HTTR to prove conformity with the new research reactor's safety regulatory requirements taken effect from December 2013. We had been making effort to restart the HTTR which was stopped since the 2011 when the Pacific coast of Tohoku Earthquake (2011.3.11) occurred. This report summarizes activities and results of HTTR operation, maintenance, and several R and Ds, which were carried out in the fiscal year 2013. (author)

  15. Operation, test, research and development of the High Temperature Engineering Test Reactor (HTTR). FY2014

    International Nuclear Information System (INIS)

    2016-02-01

    The High Temperature Engineering Test Reactor (HTTR), a graphite-moderated and helium gas-cooled reactor with 30 MW of thermal power, constructed at the Oarai Research and Development Center of the Japan Atomic Energy Agency is the first high-temperature gas-cooled reactor (HTGR) in Japan. The HTTR was attained at the full power operation of 30 MW in December 2001 and achieved the 950degC of coolant outlet temperature at outside of the reactor pressure vessel in June 2004. To establish and upgrade basic technologies for HTGRs, we have obtained demonstration test data necessary for several R and Ds, and accumulated operation and maintenance experience of HTGRs throughout the HTTR's operation such as rated power operations, safety demonstration tests and long-term high temperature operations, and so on. In fiscal year 2014, we started to apply the application document of reactor installation license for the HTTR to prove conformity with the new research reactor's safety regulatory requirements taken effect from December 2013. We had been making effort to restart the HTTR which was stopped since the 2011 by the Pacific coast of Tohoku Earthquake. This report summarizes activities and results of HTTR operation, maintenance, and several R and Ds, which were carried out in the fiscal year 2014. (author)

  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

    2007-06-01

    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 (formative) indicators rests on several claims: (a) A latent variable exists apart from the model when there are effect (reflective) indicators but not when there are causal (formative) indicators, (b) causal (formative) indicators need not have the same consequences, (c) causal (formative) indicators are inherently subject to interpretational confounding, and (d) a researcher cannot detect interpretational confounding when using causal (formative) indicators. This article shows that each claim is false. Rather, interpretational confounding is more a problem of structural misspecification of a model combined with an underidentified model that leaves these misspecifications undetected. Interpretational confounding does not occur if the model is correctly specified whether a researcher has causal (formative) or effect (reflective) indicators. It is the validity of a model not the type of indicator that determines the potential for interpretational confounding. Copyright 2007 APA, all rights reserved.

  17. Non-process instrumentation surveillance and test reduction

    International Nuclear Information System (INIS)

    Ferrell, R.; LeDonne, V.; Donat, T.; Thomson, I.; Sarlitto, M.

    1993-12-01

    Analysis of operating experience, instrument failure modes, and degraded instrument performance has led to a reduction in Technical Specification surveillance and test requirements for nuclear power plant process instrumentation. These changes have resulted in lower plant operations and maintenance (O ampersand M) labor costs. This report explores the possibility of realizing similar savings by reducing requirements for non-process instrumentation. The project team reviewed generic Technical Specifications for the four major US nuclear steam supply system (NSSS) vendors (Westinghouse, General Electric, Combustion Engineering, and Babcock ampersand Wilcox) to identify nonprocess instrumentation for which surveillance/test requirements could be reduced. The team surveyed 10 utilities to identify specific non-process instrumentation at their plants for which requirements could be reduced. The team evaluated utility analytic approaches used to justify changes in surveillance/test requirements for process equipment to determine their applicability to non-process instrumentation. The report presents a prioritized list of non-process instrumentation systems suitable for surveillance/test requirements reduction. The top three systems in the list are vibration monitors, leak detection monitors, and chemistry monitors. In general, most non-process instrumentation governed by Technical Specification requirements are candidates for requirements reduction. If statistical requirements are somewhat relaxed, the analytic approaches previously used to reduce requirements for process instrumentation can be applied to non-process instrumentation. The report identifies as viable the technical approaches developed and successfully used by Southern California Edison, Arizona Public Service, and Boston Edison

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

    Science.gov (United States)

    Zhang, Jinyu; Steel, Ronald; Ambrose, William

    2017-12-01

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

  19. Reduced enrichment for research and test reactors: Proceedings

    Energy Technology Data Exchange (ETDEWEB)

    1993-08-01

    November 9--10, 1978, marked the first of what has become an annual event--the International Meeting on Reduced Enrichment for Research and Test Reactors (RERTR). The meeting brought together for the first time many people who became major program participants in later years. This first meeting emphasized fuel development, and it established the basis for all later meetings. Believing that the proceedings of this first meeting are important as a historical record of the beginning of the international RERTR effort. This report provides presentations and discussions of this original meeting. Individual papers have been cataloged separately.

  20. Reduced enrichment for research and test reactors: Proceedings

    International Nuclear Information System (INIS)

    1993-08-01

    November 9--10, 1978, marked the first of what has become an annual event--the International Meeting on Reduced Enrichment for Research and Test Reactors (RERTR). The meeting brought together for the first time many people who became major program participants in later years. This first meeting emphasized fuel development, and it established the basis for all later meetings. Believing that the proceedings of this first meeting are important as a historical record of the beginning of the international RERTR effort. This report provides presentations and discussions of this original meeting. Individual papers have been cataloged separately

  1. Broad-Application Test Reactor

    International Nuclear Information System (INIS)

    Motloch, C.G.

    1992-05-01

    This report is about a new, safe, and operationally efficient DOE reactor of nuclear research and testing proposed for the early to mid- 21st Century. Dubbed the Broad-Application Test Reactor (BATR), the proposed facility incorporates a multiple-application, multiple-mission design to support DOE programs such as naval reactors and space power and propulsion, as well as research in medical, science, isotope, and electronics arenas. DOE research reactors are aging, and implementing major replacement projects requires long lead times. Primary design drivers include safety, low risk, minimum operation cost, mission flexibility, waste minimization, and long life. Scientists and engineers at the Idaho National Engineering Laboratory are evaluating possible fuel forms, structural materials, reactor geometries, coolants, and moderators

  2. High flux testing reactor Petten. Replacement of the reactor vessel and connected components. Overall report

    International Nuclear Information System (INIS)

    Chrysochoides, N.G.; Cundy, M.R.; Von der Hardt, P.; Husmann, K.; Swanenburg de Veye, R.J.; Tas, A.

    1985-01-01

    The project of replacing the HFR originated in 1974 when results of several research programmes confirmed severe neutron embrittlement of aluminium alloys suggesting a limited life of the existing facility. This report contains the detailed chronology of events concerning preparation and execution of the replacement. After a 14 months' outage the reactor resumed routine operation on 14th February, 1985. At the end of several years of planning and preparation the reconstruction proceded in the following steps: unloading of the old core, decay of short-lived radioactivity in December 1983, removal of the old tank and of its peripheral equipment in January-February 1984, segmentation and waste disposal of the removed components in March-April, decontamination of the pools, bottom penetration overhauling in May-June, installation of the new tank and other new components in July-September, testing and commissioning, including minor modifications in October-December, and, trials runs and start-up preparation in January-February 1985. The new HFR Petten features increased and improved experimental facilities. Among others the obsolete thermal columns was replaced by two high flux beam tubes. Moreover the new plant has been designed for future increases of reactor power and neutron fluxes. For the next three to four years the reactor has to cope with a large irradiation programme, claiming its capacity to nearly 100%

  3. Tokamak engineering test reactor

    International Nuclear Information System (INIS)

    Conn, R.W.; Jassby, D.L.

    1975-07-01

    The design criteria for a tokamak engineering test reactor can be met by operating in the two-component mode with reacting ion beams, together with a new blanket-shield design based on internal neutron spectrum shaping. A conceptual reactor design achieving a neutron wall loading of about 1 MW/m 2 is presented. The tokamak has a major radius of 3.05 m, the plasma cross-section is noncircular with a 2:1 elongation, and the plasma radius in the midplane is 55 cm. The total wall area is 149 m 2 . The plasma conditions are T/sub e/ approximately T/sub i/ approximately 5 keV, and ntau approximately 8 x 10 12 cm -3 s. The plasma temperature is maintained by injection of 177 MW of 200-keV neutral deuterium beams; the resulting deuterons undergo fusion reactions with the triton-target ions. The D-shaped toroidal field coils are extended out to large major radius (7.0 m), so that the blanket-shield test modules on the outer portion of the torus can be easily removed. The TF coils are superconducting, using a cryogenically stable TiNb design that permits a field at the coil of 80 kG and an axial field of 38 kG. The blanket-shield design for the inner portion of the torus nearest the machine center line utilizes a neutron spectral shifter so that the first structural wall behind the spectral shifter zone can withstand radiation damage for the reactor lifetime. The energy attenuation in this inner blanket is 8 x 10 -6 . If necessary, a tritium breeding ratio of 0.8 can be achieved using liquid lithium cooling in the []outer blanket only. The overall power consumption of the reactor is about 340 MW(e). A neutron wall loading greater than 1 MW/m 2 can be achieved by increasing the maximum magnetic field or the plasma elongation. (auth)

  4. The RERTR [Reduced Enrichment Research and Test Reactor] program:

    International Nuclear Information System (INIS)

    Travelli, A.

    1987-01-01

    The progress of the Reduced Enrichment Research and Test Reactor (RERTR) program is described. After a brief summary of the results which the RERTR program, in collaboration with its many international partners, had achieved by the end of 1986, the activities, results and new developments which ocurred in 1987 are reviewed. Irradiation of the second miniplate series, concentrating on U 3 Si 2 -Al and U 3 Si-Al fuels was completed and postirradiation examinations were performed on many of its miniplates. The whole-core ORR demonstration with U 3 Si 2 -Al fuel at 4.8 g U/cm 3 was completed at the end of March with excellent results and with 29 elements estimated to have reached at least 40 % average burnup. Good progress was made in the area of LEU usage for the production of fission 99 Mo, and in the coordination of safety evaluations related to LEU conversions of U.S. university reactors. Planned activities include testing and demonstrating advanced fuels intended to allow use of reduced enrichment uranium in very-high-performance reactors. Two candidate fuels are U 3 Si-Al with 19.75 % enrichment and U 3 Si 2 -Al with 45 % enrichment. Demonstration of these fuels will include irradiation of full-size elements and, possibly, a full-core demonstration. Achievement of the final program goals is still projected for 1990. This progress could not have been possible without the close international cooperation which has existed from the beginning, and which is essential to the ultimate success of the RERTR program. (Author)

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

    International Nuclear Information System (INIS)

    Buchwald, K.

    1979-05-01

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

  6. Operation, test, research and development of the high temperature engineering test reactor (HTTR). FY2003

    International Nuclear Information System (INIS)

    2005-03-01

    The High Temperature Engineering Test Reactor (HTTR) constructed at the Oarai Research Establishment of The Japan Atomic Energy Research Institute (JAERI) is the first high-temperature gas-cooled reactor (HTGR) in Japan, which is a graphite-moderated and helium gas-cooled reactor with 30MW of thermal power. Coolant of helium-gas circulates under the pressure of about 4Mpa, and the reactor inlet and outlet temperature are 395degC and 950degC (maximum), respectively coated particle fuel is used as fuel, and the HTTR core is composed of graphite prismatic blocks. The full power operation of 30MW was attained in December, 2001, and then JAERI received the commissioning license for the HTTR in March, 2002. Since 2002, we have been carrying out rated power operation, safety demonstration tests and several R and Ds, etc., and conducted the high-temperature test operation of 950degC in April, 2004. This report summarizes activities and test results on HTTR operation and maintenance as well as safety demonstration tests and several R and Ds, which were carried out in the fiscal year of 2003 before the high temperature test operation of 950degC. (author)

  7. FASTER Test Reactor Preconceptual Design Report

    Energy Technology Data Exchange (ETDEWEB)

    Grandy, C. [Argonne National Lab. (ANL), Argonne, IL (United States); Belch, H. [Argonne National Lab. (ANL), Argonne, IL (United States); Brunett, A. J. [Argonne National Lab. (ANL), Argonne, IL (United States); Heidet, F. [Argonne National Lab. (ANL), Argonne, IL (United States); Hill, R. [Argonne National Lab. (ANL), Argonne, IL (United States); Hoffman, E. [Argonne National Lab. (ANL), Argonne, IL (United States); Jin, E. [Argonne National Lab. (ANL), Argonne, IL (United States); Mohamed, W. [Argonne National Lab. (ANL), Argonne, IL (United States); Moisseytsev, A. [Argonne National Lab. (ANL), Argonne, IL (United States); Passerini, S. [Argonne National Lab. (ANL), Argonne, IL (United States); Sienicki, J. [Argonne National Lab. (ANL), Argonne, IL (United States); Sumner, T. [Argonne National Lab. (ANL), Argonne, IL (United States); Vilim, R. [Argonne National Lab. (ANL), Argonne, IL (United States); Hayes, S. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2016-03-31

    The FASTER test reactor plant is a sodium-cooled fast spectrum test reactor that provides high levels of fast and thermal neutron flux for scientific research and development. The 120MWe FASTER reactor plant has a superheated steam power conversion system which provides electrical power to a local grid allowing for recovery of operating costs for the reactor plant.

  8. Decontamination and decommissioning the Tokamak Fusion Test Reactor

    International Nuclear Information System (INIS)

    Walton, G.R.; Perry, E.D.; Commander, J.C.; Spampinato, P.T.

    1994-01-01

    The Tokamak Fusion Test Reactor (TFTR) is scheduled to complete its end-of-life deuterium-tritium (D-T) experiments in September 1994. The D-T operation will result in the TFTR machine structure becoming activated, and plasma facing and vacuum components will be contaminated with tritium. The resulting machine activation levels after a two year cooldown period will allow hands on dismantling for external structures, but require remote dismantling for the vacuum vessel. The primary objective of the Decontamination and Decommissioning (D ampersand D) Project is to provide a facility for construction of a new Department of Energy (DOE) experimental fusion reactor by March 1998. The project schedule calls for a two year shutdown period when tritium decontamination of the vacuum vessel, neutral beam injectors and other components will occur. Shutdown will be followed by an 18 month period of D ampersand D operations. The technical objectives of the project are to: safely dismantle and remove components from the test cell complex; package disassembled components in accordance with applicable regulations; ship packages to a DOE approved disposal or material recycling site; and develop expertise using remote disassembly techniques on a large scale fusion facility. This paper discusses the D ampersand D objectives, the facility to be decommissioned, and the technical plan that will be implemented

  9. Nuclear start-up, testing and core management of the Fast Test Reactor (FTR)

    International Nuclear Information System (INIS)

    Bennett, R.A.; Daughtry, J.W.; Harris, R.A.; Jones, D.H.; Nelson, J.V.; Rawlins, J.A.; Rothrock, R.B.; Sevenich, R.A.; Zimmerman, B.D.

    1980-01-01

    Plans for the nuclear start-up, low and high power physics testing, and core management of the Fast Test Reactor (FTR) are described. Owing to the arrangement of the fuel-handling system, which permits continuous instrument lead access to experiments during refuelling, it is most efficient to load the reactor in an asymmetric fashion, filling one-third core sectors at a time. The core neutron level will be monitored during this process using both in-core and ex-core detectors. A variety of physics tests are planned following the core loading. Because of the experimental purpose of the reactor, these tests will include a comprehensive characterization programme involving both active and passive neutron and gamma measurements. Following start-up tests, the FTR will be operated as a fast neutron irradiation facility, to test a wide variety of fast reactor core components and materials. Nuclear analyses will be made prior to each irradiation cycle to confirm that the planned arrangement of standard and experimental components satisfies all safety and operational constraints, and that all experiments are located so as to achieve their desired irradiation environment. (author)

  10. Evaluation of scaling concepts for integral system test facilities

    International Nuclear Information System (INIS)

    Condie, K.G.; Larson, T.K.; Davis, C.B.

    1987-01-01

    A study was conducted by EG and G Idaho, Inc., to identify and technically evaluate potential concepts which will allow the U.S. Nuclear Regulatory Commission to maintain the capability to conduct future integral, thermal-hydraulic facility experiments of interest to light water reactor safety. This paper summarizes the methodology used in the study and presents a rankings for each facility concept relative to its ability to simulate phenomena identified as important in selected reactor transients in Babcock and Wilcox and Westinghouse large pressurized water reactors. Established scaling methodologies are used to develop potential concepts for scaled integral thermal-hydraulic experiment facilities. Concepts selected included: full height, full pressure water; reduced height, reduced pressure water; reduced height, full pressure water; one-tenth linear, full pressure water; and reduced height, full scaled pressure Freon. Results from this study suggest that a facility capable of operating at typical reactor operating conditions will scale most phenomena reasonably well. Local heat transfer phenomena is best scaled by the full height facility, while the reduced height facilities provide better scaling where multi-dimensional phenomena are considered important. Although many phenomena in facilities using Freon or water at nontypical pressure will scale reasonably well, those phenomena which are heavily dependent on quality can be distorted. Furthermore, relation of data produced in facilities operating with nontypical fluids or at nontypical pressures to large plants will be a difficult and time-consuming process

  11. Similarity Analysis for Reactor Flow Distribution Test and Its Validation

    Energy Technology Data Exchange (ETDEWEB)

    Hong, Soon Joon; Ha, Jung Hui [Heungdeok IT Valley, Yongin (Korea, Republic of); Lee, Taehoo; Han, Ji Woong [KAERI, Daejeon (Korea, Republic of)

    2015-05-15

    The newly derived dimensionless groups are slightly different from Hetsroni's. Reynolds number, relative wall roughness, and Euler don't appear, instead, friction factor appears newly. In order to conserve friction factor Reynolds number and relative wall roughness should be conserved. Since the effect of Reynolds number in high range is small, and since the scaled model is far smaller than prototype the conservation of friction factor is easily obtained by making the model wall just smooth. It is much easier to implement the test design than Hetsroni's because the Reynolds number and relative wall roughness do not appear explicitly. In case that there is no free surface within the interested domain of the reactor, the gravity is of second importance, and in this case the pressure drops should be compensated for in order to compare them between prototype and model. The gravity head compensated pressure drop is directly same to the measured value by a differential pressure transmitter. In order to conserve the gravity effect Froude number should be conserved. In pool type SFR (Sodium Cooled Fast Reactor) there exists liquid level difference, and if the level difference is desired to be conserved, the Froude number should be conserved. Euler number, which represents pressure terms in momentum equation, should be well conserved according to Hetsroni's approach. It is not a wrong statement, but it should be noted that Euler number is NOT an independent variable BUT a dependent variable according to Hong et al. It means that if all the geometrical similarity and the dimensionless numbers are conserved, Euler number is automatically conserved. So Euler number need not be considered in case that the perfect geometrical similarity is kept. However, even in case that the geometrical similarity is not conserved, it possible to conserved the velocity field similarity by just conserve Euler number. It gives tolerance to the engineer who designs the test

  12. Development and verification test of integral reactor major components

    Energy Technology Data Exchange (ETDEWEB)

    Kim, J. I.; Kim, Y. W.; Kim, J. H. and others

    1999-03-01

    The conceptual designs for SG, MCP, CEDM to be installed in the integral reactor SMART were developed. Three-dimensional CAD models for the major components were developed to visualize the design concepts. Once-through helical steam generator was conceptually designed for SMART. Canned motor pump was adopted in the conceptual design of MCP. Linear pulse motor type and ballscrew type CEDM, which have fine control capabilities were studied for adoption in SMART. In parallel with the structural design, the electro-magnetic design was performed for the sizing motors and electro-magnet. Prototypes for the CEDM and MCP sub-assemblies were developed and tested to verify the performance. The impeller design procedure and the computer program to analyze the dynamic characteristics of MCP rotor shaft were developed. The design concepts of SG, MCP, CEDM were also invetigated for the fabricability.

  13. Development and verification test of integral reactor major components

    International Nuclear Information System (INIS)

    Kim, J. I.; Kim, Y. W.; Kim, J. H. and others

    1999-03-01

    The conceptual designs for SG, MCP, CEDM to be installed in the integral reactor SMART were developed. Three-dimensional CAD models for the major components were developed to visualize the design concepts. Once-through helical steam generator was conceptually designed for SMART. Canned motor pump was adopted in the conceptual design of MCP. Linear pulse motor type and ballscrew type CEDM, which have fine control capabilities were studied for adoption in SMART. In parallel with the structural design, the electro-magnetic design was performed for the sizing motors and electro-magnet. Prototypes for the CEDM and MCP sub-assemblies were developed and tested to verify the performance. The impeller design procedure and the computer program to analyze the dynamic characteristics of MCP rotor shaft were developed. The design concepts of SG, MCP, CEDM were also invetigated for the fabricability

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

    African Journals Online (AJOL)

    2016-01-23

    Jan 23, 2016 ... International Journal of Medicine and Biomedical Research. Volume 5 Issue 1 January – April 2016 www.ijmbr.com. © Shobowale et al.; licensee Michael Joanna Publications ... Patients presenting to BUTH were more likely to be HIV positive .... 18. 10.6. BUTH –Babcock University Teaching Hospital, PHC ...

  15. Obituary: Horace Welcome Babcock, 1912-2003

    Science.gov (United States)

    Vaughan, Arthur Harris

    2003-12-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-08-01

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

  17. Operation, test, research and development of the high temperature engineering test reactor (HTTR). (FY2005)

    International Nuclear Information System (INIS)

    2007-03-01

    The High Temperature Engineering Test Reactor (HTTR) constructed at the Oarai Research and Development Center of the Japan Atomic Energy Agency (JAEA) is the first high-temperature gas-cooled reactor (HTGR) in Japan, which is a graphite-moderated and helium gas-cooled reactor with 30 MW of thermal power. The full power operation of 30 MW was attained in December, 2001, and then JAERI (JAEA) received the commissioning license for the HTTR in March, 2002. Since 2002, we have been carrying out rated power operation, safety demonstration tests and several R and Ds, etc., and conducted the high-temperature test operation of 950degC in April, 2004. In fiscal 2005 year, periodical inspection and overhaul of reactivity control system were conducted, and safety demonstration tests were promoted. This report summarizes activities and test results on HTTR operation and maintenance as well as safety demonstration tests and several R and Ds, which were carried out in the fiscal year of 2005. (author)

  18. Performance Test for Neutron Detector and Associated System using Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Seongwoo; Park, Sung Jae; Cho, Man Soon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Oh, Se Hyun [USERS, Daejeon (Korea, Republic of); Shin, Ho Cheol [KHNP CRI, Daejeon (Korea, Republic of)

    2016-10-15

    SPND (Self-Powered Neutron Detector) has been developed to extend its lifespan. ENFMS (Ex-Core Flux Monitoring System) of pressurized water reactor has been also improved. After the development and improvement, their performance must be verified under the neutron irradiation environment. We used a research reactor for the performance verification of neutron detector and associated system because the research reactor can meet the neutron flux level of commercial nuclear reactor. In this paper, we report the performance verification method and result for the SPND and ENFMS using the research reactor. The performance tests for the SPND and ENFMS were conducted using UCI TRIGA reactor. The test environment of commercial reactor’s neutron flux level must be required. However, it is difficult to perform the test in the commercial rector due to the constraint of time and space. The research reactor can be good alternative neutron source for the test of neutron detectors and associated system.

  19. Study for improvement of performance of the test and research reactors

    Energy Technology Data Exchange (ETDEWEB)

    Sakurai, Fumio [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1999-03-01

    Current utilization needs for the test and research reactors become more advanced and diversified along with the advance of nuclear science and technology. Besides, the requested safety for the research and test reactors grows strictly every year as well as a case of the power reactors. Under this circumstance, every effort to improve reactor performance including its safety is necessary to be sustained for allowing more effective utilization of the test and research reactors as experimental apparatus for advanced researches. In this study, the following three themes i.e., JMTR high-performance fuel element, evaluation method of fast neutron irradiation dose in the JMTR, evaluation method of performance of siphon break valve as core covering system for water-cooled test and research reactors, were investigated respectively from the views of improvement of core performance as a neutron source, utilization performance as an experimental apparatus, and safety as a reactor plant. (author)

  20. Review of Transient Fuel Test Results at Sandia National Laboratories and the Potential for Future Fast Reactor Fuel Transient Testing in the Annular Core Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Wright, Steven A.; Pickard, Paul S.; Parma, Edward J.; Vernon, Milton E.; Kelly, John; Tikare, Veena [Sandia National Laboratories, Org 6872 MS-1146, PO Box 5800 Albuquerque, New Mexico 87185 (United States)

    2009-06-15

    Reactor driven transient tests of fast reactor fuels may be required to support the development and certification of new fuels for Fast Reactors. The results of the transient fuel tests will likely be needed to support licensing and to provide validation data to support the safety case for a variety of proposed fast fuel types and reactors. In general reactor driven transient tests are used to identify basic phenomenology during reactor transients and to determine the fuel performance limits and margins to failure during design basis accidents such as loss of flow, loss of heat sink, and reactivity insertion accidents. This paper provides a summary description of the previous Sandia Fuel Disruption and Transient Axial Relocation tests that were performed in the Annular Core Research Reactor (ACRR) for the U.S. Nuclear Regulatory Commission almost 25 years ago. These tests consisted of a number of capsule tests and flowing gas tests that used fission heating to disrupt fresh and irradiated MOX fuel. The behavior of the fuel disruption, the generation of aerosols and the melting and relocation of fuel and cladding was recorded on high speed cinematography. This paper will present videos of the fuel disruption that was observed in these tests which reveal stark differences in fuel behavior between fresh and irradiated fuel. Even though these tests were performed over 25 years ago, their results are still relevant to today's reactor designs. These types of transient tests are again being considered by the Advanced Fuel Cycle Initiative to support the Global Nuclear Energy Partnership because of the need to perform tests on metal fuels and transuranic fuels. Because the Annular Core Research Reactor is the only transient test facility available within the US, a brief summary of Sandia's continued capability to perform these tests in the ACRR will also be provided. (authors)

  1. Development and testing of the EDF-2 reactor fuel element

    International Nuclear Information System (INIS)

    Delpeyroux, P.

    1964-01-01

    This technical report reviews the work which has been necessary for defining the EDF-2 fuel element. After giving briefly the EDF-2 reactor characteristics and the preliminary choice of parameters which made it possible to draw up a draft plan for the fuel element, the authors consider the research proper: - Uranium studies: tests on the passage into the β phase of an internal crown of a tube, bending of the tube under the effect of a localized force, welding of the end-pellets and testing for leaks. The resistance of the tube to crushing and of the pellets to yielding under the external pressure have been studied in detail in another CEA report. - Can studies: conditions of production and leak proof testing of the can, resistance of the fins to creep due to the effect of the gas flow. - Studies of the extremities of the element: creep under compression and welding of the plugs to the can. - Cartridge studies: determination of the characteristics of the can fuel fixing grooves and of the canning conditions, verification of the resistance of the fuel element to thermal cycling, determination of the temperature drop at the can-fuel interface dealt with in more detail in another CEA report. - Studies of the whole assembly: this work which concerns the graphite jacket, the support and the cartridge vibrations has been carried out by the Mechanical and Thermal Study Service (Mechanics Section). In this field the Fuel Element Study Section has investigated the behaviour of the centering devices in a gas current. The outcome of this research is the defining of the plan of the element the production process and the production specifications. The validity of ail these out-of-pile tests will be confirmed by the in-pile tests already under way and by irradiation of the elements in the EDF-2 reactor itself. In conclusion the programme is given for improving the fuel element and for defining the fuel element for the second charge. (authors) [fr

  2. FASTER test reactor preconceptual design report summary

    Energy Technology Data Exchange (ETDEWEB)

    Grandy, C. [Argonne National Lab. (ANL), Argonne, IL (United States); Belch, H. [Argonne National Lab. (ANL), Argonne, IL (United States); Brunett, A. [Argonne National Lab. (ANL), Argonne, IL (United States); Heidet, F. [Argonne National Lab. (ANL), Argonne, IL (United States); Hill, R. [Argonne National Lab. (ANL), Argonne, IL (United States); Hoffman, E. [Argonne National Lab. (ANL), Argonne, IL (United States); Jin, E. [Argonne National Lab. (ANL), Argonne, IL (United States); Mohamed, W. [Argonne National Lab. (ANL), Argonne, IL (United States); Moisseytsev, A. [Argonne National Lab. (ANL), Argonne, IL (United States); Passerini, S. [Argonne National Lab. (ANL), Argonne, IL (United States); Sienicki, J. [Argonne National Lab. (ANL), Argonne, IL (United States); Sumner, T. [Argonne National Lab. (ANL), Argonne, IL (United States); Vilim, R. [Argonne National Lab. (ANL), Argonne, IL (United States); Hayes, Steven [Argonne National Lab. (ANL), Argonne, IL (United States)

    2016-02-29

    The FASTER reactor plant is a sodium-cooled fast spectrum test reactor that provides high levels of fast and thermal neutron flux for scientific research and development. The 120MWe FASTER reactor plant has a superheated steam power conversion system which provides electrical power to a local grid allowing for recovery of operating costs for the reactor plant.

  3. Establishing a safety and licensing basis for generation IV advanced reactors. License by test

    International Nuclear Information System (INIS)

    Kadak, Andrew C.

    2001-01-01

    The license by test approach to licensing is a novel method of licensing reactors. It provides an opportunity to deal with innovative non-water reactors in a direct way on a time scale that could permit early certification based on tests of a demonstration reactor. The uncertainties in the design and significant contributors to risk would be identified in the PRA during the design. Deterministic analysis computer codes could be tested on a real reactor. Scaling effects and associated uncertainties would be minimized. License by test is an approach that has sufficient merit to be developed and tested

  4. Irradiation capsule for testing magnetic fusion reactor first-wall materials at 60 and 2000C

    International Nuclear Information System (INIS)

    Conlin, J.A.

    1985-08-01

    A new type of irradiation capsule has been designed, and a prototype has been tested in the Oak Ridge Research Reactor (ORR) for low-temperature irradiation of Magnetic Fusion Reactor first-wall materials. The capsule meets the requirements of the joint US/Japanese collaborative fusion reactor materials irradiation program for the irradiation of first-wall fusion reactor materials at 60 and 200 0 C. The design description and results of the prototype capsule performance are presented

  5. Review of inservice inspection and nondestructive examination practices at DOE Category A test and research reactors

    International Nuclear Information System (INIS)

    Anderson, M.T.; Aldrich, D.A.

    1990-09-01

    In-service inspection (ISI) programs are used at commercial nuclear power plants for monitoring the pressure boundary integrity of various systems and components to ensure their continued safe operation. The Department of Energy (DOE) operates several test and research reactors. This report represents an evaluation of the ISI and nondestructive examination (NDE) practices at five DOE Category A (> 20 MW thermal) reactors as compared, where applicable, to the current ISI activities of commercial nuclear power facilities. The purpose of an inservice inspection (ISI) program is to establish regular surveillance of safety-related components to ensure their safe and reliable operation. The integrity of materials comprising these components is generally monitored by means of periodic nondestructive examinations (NDE), which, if appropriately performed, provide methods for identifying degradation that could render components unable to perform their intended safety functions. The reactors evaluated during this review were the Experimental Breeder Reactor 2 and the Fast Flux Test Facility (liquid-metal cooled plants), the Advanced Test Reactor and the High Flux Isotopes Reactor (light-water cooled reactors), and the High Flux Beam Reactor (a heavy-water cooled facility). Although these facilities are extremely diverse in design and operation, they all have less stored energy, smaller inventories of radionuclides, and generally, more remote locations than commercial reactors. However, all DOE test and research facilities contain components similar to those of commercial reactors for which continued integrity is important to maintain plant safety. 10 refs., 6 tabs

  6. A review of experiments and results from the transient reactor test (TREAT) facility

    International Nuclear Information System (INIS)

    Deitrich, L. W.

    1998-01-01

    The TREAT Facility was designed and built in the late 1950s at Argonne National Laboratory to provide a transient reactor for safety experiments on samples of reactor fuels. It first operated in 1959. Throughout its history, experiments conducted in TREAT have been important in establishing the behavior of a wide variety of reactor fuel elements under conditions predicted to occur in reactor accidents ranging from mild off normal transients to hypothetical core disruptive accidents. For much of its history, TREAT was used primarily to test liquid-metal reactor fuel elements, initially for the Experimental Breeder Reactor-II (EBR-II), then for the Fast Flux Test Facility (FFTF), the Clinch River Breeder Reactor Plant (CRBRP), the British Prototype Fast Reactor (PFR), and finally, for the Integral Fast Reactor (IFR). Both oxide and metal elements were tested in dry capsules and in flowing sodium loops. The data obtained were instrumental in establishing the behavior of the fuel under off-normal and accident conditions, a necessary part of the safety analysis of the various reactors. In addition, TREAT was used to test light-water reactor (LWR) elements in a steam environment to obtain fission-product release data under meltdown conditions. Studies are now under way on applications of TREAT to testing of the behavior of high-burnup LWR elements under reactivity-initiated accident (RIA) conditions using a high-pressure water loop

  7. Reactor transients tests for SNR fuel elements in HFR reactor

    International Nuclear Information System (INIS)

    Plitz, H.

    1989-01-01

    In HFR reactor, fuel pins of LMFBR reactors are putted in irradiation specimen capsules cooled with sodium for reactor transients tests. These irradiation capsules are instrumented and the experiences realized until this day give results on: - Fuel pins subjected at a continual variation of power - melting fuel - axial differential elongation of fuel pins

  8. Irradiation Facilities at the Advanced Test Reactor

    International Nuclear Information System (INIS)

    S. Blaine Grover

    2005-01-01

    The Advanced Test Reactor (ATR) is the third generation and largest test reactor built in the Reactor Technology Complex (RTC) (formerly known as the Test Reactor Area), located at the Idaho National Laboratory (INL), to study the effects of intense neutron and gamma radiation on reactor materials and fuels. The RTC was established in the early 1950s with the development of the Materials Testing Reactor (MTR), which operated until 1970. The second major reactor was the Engineering Test Reactor (ETR), which operated from 1957 to 1981, and finally the ATR, which began operation in 1967 and will continue operation well into the future. These reactors have produced a significant portion of the world's data on materials response to reactor environments. The wide range of experiment facilities in the ATR and the unique ability to vary the neutron flux in different areas of the core allow numerous experiment conditions to co-exist during the same reactor operating cycle. Simple experiments may involve a non-instrumented capsule containing test specimens with no real-time monitoring or control capabilities. More sophisticated testing facilities include inert gas temperature control systems and pressurized water loops that have continuous chemistry, pressure, temperature, and flow control as well as numerous test specimen monitoring capabilities. There are also apparatus that allow for the simulation of reactor transients on test specimens

  9. Material and geometry options and performance characteristics for a test reactor

    International Nuclear Information System (INIS)

    Jahshan, S.N.; Fletcher, C.D.; Terry, W.K.

    1993-01-01

    For the past 3 yr, an Idaho National Engineering Laboratory (INEL) design team has studied design options for a new test reactor to provide continued testing services after several aging test reactors in the United States are decommissioned. This new reactor, the Broad Application Test Reactor (BATR), would also fill other currently unmet needs, such as medical isotope production and space reactor component testing. Consideration of user needs, safety requirements, developmental uncertainties, and other factors led to the selection of an evolutionary design with plate fuel and several independently cooled test loops. The fuel would be cooled by light water, but most neutron moderation would come from heavy water or beryllium. The BATR design was tentatively scaled to the Advanced Test Reactor (ATR), an existing reactor at INEL: The power output of BATR is 250 MW(thermal), and the active core heights is 1 m. For safety in loss-of-flow events, the coolant flows upward through the core. The BATR design has one large test loop (with a test space diameter of 15.0 cm) along the central axis of the core and six smaller test loops (with test space diameters of 8.0 cm) centered at 6-deg azimuthal intervals on a 24.71-cm-diam circle around the central core axis

  10. Conversion of research and test reactors to low enriched uranium fuel: technical overview and program status

    International Nuclear Information System (INIS)

    Roglans-Ribas, J.

    2008-01-01

    Many of the nuclear research and test reactors worldwide operate with high enriched uranium fuel. In response to worries over the potential use of HEU from research reactors in nuclear weapons, the U.S Department of Energy (DOE) initiated a program - the Reduced Enrichment for Research and Test Reactors (RERTR) - in 1978 to develop the technology necessary to reduce the use of HEU fuel by converting research reactors to low enriched uranium (LEU) fuel. The Reactor Conversion program is currently under the DOE's National Nuclear Security Administration's Global Threat Reduction Initiative (GTRI). 55 of the 129 reactors included in the scope have been already converted to LEU fuel or have shutdown prior to conversion. The major technical activities of the Conversion Program include: (1) the development of advanced LEU fuels; (2) conversion analysis and conversion support; and (3) technology development for the production of Molybdenum-99 (Mo 99 ) with LEU targets. The paper provides an overview of the status of the program, the technical challenges and accomplishments, and the role of international collaborations in the accomplishment of the Conversion Program objectives. Nuclear research and test reactors worldwide have been in operation for over 60 years. Many of these facilities operate with high enriched uranium fuel. In response to increased worries over the potential use of HEU from research reactors in the manufacturing of nuclear weapons, the U.S Department of Energy (DOE) initiated a program - the Reduced Enrichment for Research and Test Reactors (RERTR) - in 1978 to develop the technology necessary to reduce the use of HEU fuel in research reactors by converting them to low enriched uranium (LEU) fuel. The reactor conversion program was initially focused on U.S.-supplied reactors, but in the early 1990s it expanded and began to collaborate with Russian institutes with the objective of converting Russian supplied reactors to the use of LEU fuel.

  11. Testing of HTR UO{sub 2} TRISO fuels in AVR and in material test reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kania, Michael J., E-mail: MichaelJKania@googlemail.com [Retired from Lockheed Martin Corp, 20 Beach Road, Averill Park, NY 12018 (United States); Nabielek, Heinz, E-mail: heinznabielek@me.com [Retired from Research Center Jülich, Monschauerstrasse 61, 52355 Düren (Germany); Verfondern, Karl [Research Center Juelich,Research Center Jülich, Institute of Energy and Climate Research, 52425 Jülich (Germany); Allelein, Hans-Josef [Research Center Juelich,Research Center Jülich, Institute of Energy and Climate Research, 52425 Jülich (Germany); RWTH Aachen, 52072 Aachen (Germany)

    2013-10-15

    The German High Temperature Reactor Fuel Development Program successfully developed, licensed and manufactured many thousands of spherical fuel elements that were used to power the experimental AVR reactor and the commercial THTR reactor. In the 1970s, this program extended the performance envelope of HTR fuels by developing and qualifying the TRISO-coated particle system. Irradiation testing in real-time AVR tests and accelerated MTR tests demonstrated the superior manufacturing process of this fuel and its irradiation performance. In the 1980s, another program direction change was made to a low enriched UO{sub 2} TRISO-coated particle system coupled with high-quality manufacturing specifications designed to meet new HTR plant design needs. These needs included requirements for inherent safety under normal operation and accident conditions. Again, the German fuel development program met and exceeded these challenges by manufacturing and qualifying the low-enriched UO{sub 2} TRISO-fuel system for HTR systems with steam generation, gas-turbine systems and very high temperature process heat applications. Fuel elements were manufactured in production scale facilities that contained near defect free UO{sub 2} TRISO coated particles, homogeneously distributed within a graphite matrix with very low levels of uranium contamination. Good irradiation performance for these elements was demonstrated under normal operating conditions to 12% FIMA and under accident conditions not exceeding 1600 °C.

  12. Advanced Test Reactor probabilistic risk assessment methodology and results summary

    International Nuclear Information System (INIS)

    Eide, S.A.; Atkinson, S.A.; Thatcher, T.A.

    1992-01-01

    The Advanced Test Reactor (ATR) probabilistic risk assessment (PRA) Level 1 report documents a comprehensive and state-of-the-art study to establish and reduce the risk associated with operation of the ATR, expressed as a mean frequency of fuel damage. The ATR Level 1 PRA effort is unique and outstanding because of its consistent and state-of-the-art treatment of all facets of the risk study, its comprehensive and cost-effective risk reduction effort while the risk baseline was being established, and its thorough and comprehensive documentation. The PRA includes many improvements to the state-of-the-art, including the following: establishment of a comprehensive generic data base for component failures, treatment of initiating event frequencies given significant plant improvements in recent years, performance of efficient identification and screening of fire and flood events using code-assisted vital area analysis, identification and treatment of significant seismic-fire-flood-wind interactions, and modeling of large loss-of-coolant accidents (LOCAs) and experiment loop ruptures leading to direct damage of the ATR core. 18 refs

  13. The roles of EBR-II and TREAT [Transient Reactor Test] in establishing liquid metal reactor safety

    International Nuclear Information System (INIS)

    Sackett, J.I.; Lehto, W.K.; Solbrig, C.W.

    1990-01-01

    This paper examines the role of the Experimental Breeder Reactor II (EBR-II) and Transient Reactor Test (TREAT) facilities in contributing to the understanding and resolution of key safety issues in liquid metal reactor safety during the decade of the 80's. Fuels and materials testing has been carried out to address questions on fuels behavior during steady-state and upset conditions. In addition, EBR-II has conducted plant tests to demonstrate passive response to ATWS events and to develop control and diagnostic strategies for safe operation of advanced LMRs. TREAT and EBR-II complement each other and between them provide a transient testing capability that covers the whole range of concerns during overpower conditions. EBR-II, with use of the special Automatic Control Rod Drive System, can generate power change rates that overlap the lower end of the TREAT capability. 21 refs

  14. Cavity temperature and flow characteristics in a gas-core test reactor

    Science.gov (United States)

    Putre, H. A.

    1973-01-01

    A test reactor concept for conducting basic studies on a fissioning uranium plasma and for testing various gas-core reactor concepts is analyzed. The test reactor consists of a conventional fuel-element region surrounding a 61-cm-(2-ft-) diameter cavity region which contains the plasma experiment. The fuel elements provide the neutron flux for the cavity region. The design operating conditions include 60-MW reactor power, 2.7-MW cavity power, 200-atm cavity pressure, and an average uranium plasma temperature of 15,000 K. The analytical results are given for cavity radiant heat transfer, hydrogen transpiration cooling, and uranium wire or powder injection.

  15. Reduced enrichment for research and test reactors. Proceedings

    International Nuclear Information System (INIS)

    Thamm, G.; Brandt, M.

    1991-01-01

    The 12th meeting was attended by 113 participants coming from 21 countries and from EURATOM and IAEA.42 reports were presented orally within 10 sessions dealing with 5 main topics: 1) programs(5); 2) fuels(12); 3) reactor conversions(17); 5) high performance neutron sources(4); 5) others(4). (HP)

  16. Benchmark tests for fast and thermal reactor applications

    International Nuclear Information System (INIS)

    Seki, Yuji

    1984-01-01

    Integral tests of JENDL-2 library for fast and thermal reactor applications are reviewed including relevant analyses of JUPITER experiments. Criticality and core center characteristics were tested with one-dimensional models for a total of 27 fast critical assemblies. More sofisticated problems such as reaction rate distributions, control rod worths and sodium void reactivities were tested using two-dimensional models for MOZART and ZPPR-3 assemblies. Main observations from the fast core benchmark tests are as follows. 1) The criticality is well predicted; the average C/E value is 0.999+-0.008 for uranium cores and 0.997+-0.005 for plutonium cores. 2) The calculation underpredicts the reaction rate ratio 239 Pusub(fis)/ 235 Usub(fis) by 3% and overpredicts 238 Usub(cap)/ 239 Pusub(fis) by 6%. The results are consistent with those of JUPITER analyses. 3) The reaction rate distributions in the cores of prototype size are well predicted within +-3%. In larger JUPITER cores, however, the C/E value increases with the radial distance from the core center up to 6% at the outer core edge. 4) The prediction of control rod worths is satisfactory; C/E values are within the range from 0.92 to 0.97 with no apparent dependence on 10 B enrichment and the number of control rods inserted. Spatial dependence of C/E is also observed in the JUPITER cores. 5) The sodium void reactivity is overpredicted by 30% to 50% to the positive side. 1) The criticality is well predicted, as is the same in the fast core tests; the average C/E is 0.997+-0.003. 2) The calculation overpredicts 238 Usub(fis)/ 235 Usub(fis) by 3% to 6%, which shows the same tendency as in the small and medium size fast assemblies. The 238 Usub(cap)/ 235 Usub(fis) ratio is well predicted in the thermal cores. The calculated reaction rate ratios of 232 Th deviate from the measurements by 10% to 15%. (author)

  17. Real time simulator for material testing reactor

    Energy Technology Data Exchange (ETDEWEB)

    Takemoto, Noriyuki; Imaizumi, Tomomi; Izumo, Hironobu; Hori, Naohiko; Suzuki, Masahide [Japan Atomic Energy Agency, Oarai Research and Development Center, Oarai, Ibaraki (Japan); Ishitsuka, Tatsuo; Tamura, Kazuo [ITOCHU Techno-Solutions Corp., Tokyo (Japan)

    2012-03-15

    Japan Atomic Energy Agency (JAEA) is now developing a real time simulator for a material testing reactor based on Japan Materials Testing Reactor (JMTR). The simulator treats reactor core system, primary and secondary cooling system, electricity system and irradiation facility systems. Possible simulations are normal reactor operation, unusual transient operation and accidental operation. The developed simulator also contains tool to revise/add facility in it for the future development. (author)

  18. Real time simulator for material testing reactor

    International Nuclear Information System (INIS)

    Takemoto, Noriyuki; Imaizumi, Tomomi; Izumo, Hironobu; Hori, Naohiko; Suzuki, Masahide; Ishitsuka, Tatsuo; Tamura, Kazuo

    2012-01-01

    Japan Atomic Energy Agency (JAEA) is now developing a real time simulator for a material testing reactor based on Japan Materials Testing Reactor (JMTR). The simulator treats reactor core system, primary and secondary cooling system, electricity system and irradiation facility systems. Possible simulations are normal reactor operation, unusual transient operation and accidental operation. The developed simulator also contains tool to revise/add facility in it for the future development. (author)

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

    International Nuclear Information System (INIS)

    Mattar Neto, Miguel

    2009-01-01

    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)

  20. Safety analysis calculations for research and test reactors

    International Nuclear Information System (INIS)

    Chen, S.Y.; MacDonald, R.; MacFarlane, D.

    1983-01-01

    Safety issues for the two general types of reactors, i.e., the plate-type (MTR-type) reactor and the rod-type (TRIGA-type) reactor, resulting from the changes associated with LEU vs HEU fuels, are explored. The plate-type fuels are typically uranium aluminide (UAl/sub x/) compounds dispersed in aluminum and clad with aluminum. Moderation is provided by the water coolant. Self shut-down reactivity coefficients with HEU fuel are entirely a result of coolant heating, whereas with LEU fuel there is an additional shut down contribution provided by the direct heating of the fuel due to the Doppler coefficient. In contrast, the rod-type (TRIGA) fuels are mixtures of zirconium hydride, uranium, and erbium. This fuel mixture is formed into rods (approx. 1 cm diameter) and clad with stainless steel or Incoloy. In the TRIGA fuel the self-shutdown reactivity is more complex, depending on heating of the fuel rather than the coolant. Results of transient calculations performed with existing computer codes, most suited for each type of reactor, are presented

  1. Refurbish research and test reactors corresponding to global age of nuclear energy

    International Nuclear Information System (INIS)

    Mishima, Kaichiro; Oyama, Yukio; Okamoto, Koji; Yamana, Hajime; Yamaguchi, Akira

    2011-01-01

    This special article featured arguments for refurbishment of research and test reactors corresponding to global age of nuclear energy, based on the report: 'Investigation of research facilities necessary for future joint usage' issued by the special committee of Atomic Energy Society of Japan (AESJ) in September 2010. It consisted of six papers titled as 'Introduction-establishment of AESJ special committee for investigation', 'State of research and test reactors in Japan', 'State of overseas research and test reactors', 'Needs analysis for research and test reactors', 'Proposal of AESJ special committee' and 'Summary and future issues'. In order to develop human resources and promote research and development needed in global age of nuclear energy, research and test reactors would be refurbished as an Asian regional center of excellence. (T. Tanaka)

  2. PITR: Princeton Ignition Test Reactor

    International Nuclear Information System (INIS)

    1978-12-01

    The principal objectives of the PITR - Princeton Ignition Test Reactor - are to demonstrate the attainment of thermonuclear ignition in deuterium-tritium, and to develop optimal start-up techniques for plasma heating and current induction, in order to determine the most favorable means of reducing the size and cost of tokamak power reactors. This report describes the status of the plasma and engineering design features of the PITR. The PITR geometry is chosen to provide the highest MHD-stable values of beta in a D-shaped plasma, as well as ease of access for remote handling and neutral-beam injection

  3. MTR (Materials Testing Reactors) cores fuel management. Application of a low enrichment reactor for the equilibrium and transitory core calculation

    International Nuclear Information System (INIS)

    Relloso, J.M.

    1990-01-01

    This work describes a methodology to define the equilibrium core and a MTR (Materials Testing Reactors) type reactor's fuel management upon multiple boundary conditions, such as: end cycle and permitted maximum reactivities, burn-up extraction and maximun number of movements by rechange. The methodology proposed allows to determine the best options through conceptual relations, prior to a detailed calculation with the core code, reducing the test number with these codes and minimizing in this way CPU cost. The way to better systematized search of transient cores from the first one to the equilibrium one is presented. (Author) [es

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

    International Nuclear Information System (INIS)

    Bess, John D.; Gullifor, Jim

    2015-03-01

    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 D 2 O-H 2 O 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

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

    International Nuclear Information System (INIS)

    Smith, K.A.

    1988-12-01

    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

  6. Ground test facility for nuclear testing of space reactor subsystems

    International Nuclear Information System (INIS)

    Quapp, W.J.; Watts, K.D.

    1985-01-01

    Two major reactor facilities at the INEL have been identified as easily adaptable for supporting the nuclear testing of the SP-100 reactor subsystem. They are the Engineering Test Reactor (ETR) and the Loss of Fluid Test Reactor (LOFT). In addition, there are machine shops, analytical laboratories, hot cells, and the supporting services (fire protection, safety, security, medical, waste management, etc.) necessary to conducting a nuclear test program. This paper presents the conceptual approach for modifying these reactor facilities for the ground engineering test facility for the SP-100 nuclear subsystem. 4 figs

  7. Application of non-destructive testing and in-service inspections to research reactors and preparation of ISI programme and manual for WWR-C research reactors

    International Nuclear Information System (INIS)

    Khattab, M.

    1996-01-01

    The present report gives a review on the results of application of non-destructive testing and in-service inspections to WWR-C reactors in different countries. The major problems related to reactor safety and the procedure of inspection techniques are investigated to collect the experience gained from this type of reactors. Exchangeable experience in solving common problems in similar reactors play an important role in the effectiveness of their rehabilitation programmes. 9 figs., 4 tabs

  8. Proceedings of the 1984 international meeting on Reduced Enrichment for Research and Test Reactors. Base technology

    International Nuclear Information System (INIS)

    1985-07-01

    More than 40 papers were presented at this RERTR Meeting during the following sessions: Status of RERTR programs and licensing procedures; LEU fuel element development; fuel fabrication and testing; economics; mixed reactor cores; and applications, i.e. neutronics and thermal hydraulics design of upgraded reactors, with new LEU fuel, fuel cycle studies, feasibility and safety analyses

  9. Proceedings of the 1984 international meeting on Reduced Enrichment for Research and Test Reactors. Base technology

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1985-07-01

    More than 40 papers were presented at this RERTR Meeting during the following sessions: Status of RERTR programs and licensing procedures; LEU fuel element development; fuel fabrication and testing; economics; mixed reactor cores; and applications, i.e. neutronics and thermal hydraulics design of upgraded reactors, with new LEU fuel, fuel cycle studies, feasibility and safety analyses.

  10. Reactor operator screening test experiences

    International Nuclear Information System (INIS)

    O'Brien, W.J.; Penkala, J.L.; Witzig, W.F.

    1976-01-01

    When it became apparent to Duquesne Light Company of Pittsburgh, Pennsylvania, that the throughput of their candidate selection-Phase I training-reactor operator certification sequence was something short of acceptable, the utility decided to ask consultants to make recommendations with respect to candidate selection procedures. The recommendation implemented was to create a Nuclear Training Test that would predict the success of a candidate in completing Phase I training and subsequently qualify for reactor operator certification. The mechanics involved in developing and calibrating the Nuclear Training Test are described. An arbitration decision that resulted when a number of International Brotherhood of Electrical Workers union employees filed a grievance alleging that the selection examination was unfair, invalid, not job related, inappropriate, and discriminatorily evaluated is also discussed. The arbitration decision favored the use of the Nuclear Training Test

  11. Use of Reactor Pressure Vessel Surveillance Materials for Extended Life Evaluations Using Power and Test Reactor Irradiations

    International Nuclear Information System (INIS)

    Server, W.L.; Nanstad, R.K.; Odette, G.R.

    2012-01-01

    The most important component in assuring safety of the nuclear power plant is the reactor pressure (RPV). Surveillance programs have been designed to cover the licensed life of operating nuclear RPVs. The original surveillance programs were designed when the licensed life was 40 years. More than one-half of the operating nuclear plants in the USA have an extended license out to 60 years, and there are plans to continue to operate many plants out to 80 years. Therefore, the surveillance programs have had to be adjusted or enhanced to generate key data for 60 years, and now consideration must be given for 80 or more years. To generate the necessary data to assure safe operation out to these extended license lives, test reactor irradiations have been initiated with key RPV and model alloy steels, which include several steels irradiated in the current power reactor surveillance programs out to relatively high fluence levels. These data are crucial in understanding the radiation embrittlement mechanisms and to enable extrapolation of the irradiation effects on mechanical properties for these extended time periods. This paper describes the potential radiation embrittlement mechanisms and effects when assessing much longer operating times and higher neutron fluence levels. Potential methods for adjusting higher neutron flux test reactor data for use in predicting power reactor vessel conditions are discussed. (author)

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

    Science.gov (United States)

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

    2017-08-01

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

  13. The role and use of materials-testing reactors in France

    International Nuclear Information System (INIS)

    Colomez, Gerard; Mas, Pierre

    1981-01-01

    The authors outline the role played by polyvalent materials-testing reactors in France - in the area of primary and applied research - in neutronic irradiation production and the acquisition and diffusion of nuclear know-how. They then go on to describe the fields of application of these reactors [fr

  14. Design and safety consideration in the High-Temperature Engineering Test Reactor (HTTR)

    International Nuclear Information System (INIS)

    Saito, Shinzo; Tanaka, Toshiuki; Sudo, Yukio; Baba, Osamu; Shiozawa, Shusaku; Okubo, Minoru

    1990-01-01

    The budget for construction of the High-Temperature Engineering Test Reactor (HTTR) was recently committed by the Government in Japan. The HTTR is a test reactor with thermal output of 30 MW and reactor outlet coolant temperature of 950 deg. C at high temperature test operation. The HTTR plant uses a pin-in-block design core and will be used as an experience leading to high temperature applications. Several major important safety considerations are adopted in the design of the HTTR. These are as follows: 1) A coated particle fuel must not be failed during a normal reactor operation and an anticipated operational occurrence; 2) Two independent and diverse reactor shut-down systems are provided in order to shut down the reactor safely and reliably in any condition; 3) Back-up reactor cooling systems which are safety ones are provided in order to remove residual heat of reactor in any condition; 4) Multiple barriers and countermeasures are provided to contain fission products such as a containment, pressure gradient between the primary and secondary cooling circuit and so on, though coated particle fuels contain fission products with high reliability; 5) The functions of materials used in the primary cooling circuit are separated to be pressure-resisting and heat-resisting in order to resolve material problems and maintain high reliability. The detailed design of the HTTR was completed with extensive accumulation of material data and component tests. (author)

  15. Homogeneous fast reactor benchmark testing of CENDL-2 and ENDF/B-6

    International Nuclear Information System (INIS)

    Liu Guisheng

    1995-11-01

    How to choose correct weighting spectrum has been studied to produce multigroup constants for fast reactor benchmark calculations. A correct weighting option makes us obtain satisfying results of K eff and central reaction rate ratios for nine fast reactor benchmark testing of CENDL-2 and ENDF/B-6. (author). 8 refs, 2 figs, 4 tabs

  16. Homogeneous fast reactor benchmark testing of CENDL-2 and ENDF/B-6

    International Nuclear Information System (INIS)

    Liu Guisheng

    1995-01-01

    How to choose correct weighting spectrum has been studied to produce multigroup constants for fast reactor benchmark calculations. A correct weighting option makes us obtain satisfying results of K eff and central reaction rate ratios for nine fast reactor benchmark testings of CENDL-2 and ENDF/B-6. (4 tabs., 2 figs.)

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

    Directory of Open Access Journals (Sweden)

    Stephanie Castillo

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

  18. the JHR Material Testing Reactor

    International Nuclear Information System (INIS)

    Roure, C.; Cornu, B.; Berthet, B.; Simon, E.; Estre, N.; Guimbal, P.; Kinnunen, P.; Kotiluoto, P.

    2013-06-01

    The Jules Horowitz Reactor (JHR) is a European experimental reactor under construction in CEA Cadarache. It will be dedicated to material and fuel irradiation tests, and to medical isotopes production. Non-Destructive nuclear Examinations systems (NDE) will be implemented in pools to analyse the irradiated fuel or tested material in their supporting experimental irradiation devices extracted from the core or its immediate periphery. The Nuclear Measurement Laboratory (NML) of CEA Cadarache is working in collaboration with VTT (Technical Research Centre in Finland) in designing and developing NDE systems implementing gamma-ray spectroscopy and high energy X-ray imaging of the sample and irradiation device. CEA is also designing a neutron radiography system for which NML is working on the detection system. Design studies are performed with Monte Carlo transport codes and specific simulation tools developed by the NML for Xray and neutron imaging. (authors)

  19. Four critical facilities: their capabilities and programs

    International Nuclear Information System (INIS)

    Whitesides, G.E.

    1980-01-01

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

  20. Test reactor: basic to U.S. breeder reactor development

    International Nuclear Information System (INIS)

    Miller, B.J.; Harness, A.J.

    1975-01-01

    Long-range energy planning in the U. S. includes development of a national commercial breeder reactor program. U. S. development of the LMFBR is following a conservative sequence of extensive technology development through use of test reactors and demonstration plants prior to construction of commercial plants. Because materials and fuel technology development is considered the first vital step in this sequence, initial U. S. efforts have been directed to the design and construction of a unique test reactor. The Fast Flux Test Facility, FFTF, is a 400 MW(t) reactor with driver fuel locations, open test locations, and closed loops for higher risk experiments. The FFTF will provide a prototypic LMFBR core environment with sufficient instrumentation for detailed core environmental characterization and a testing capability substituted for breeder capability. The unique comprehensive fuel and materials testing capability of the FFTF will be key to achieving long-range objectives of increased power density, improved breeding gain and shorter doubling times. (auth)

  1. In-reactor testing of self-powered neutron detectors and miniature fission chambers

    International Nuclear Information System (INIS)

    Duchene, J.; LeMeur, R.; Verdant, R.

    1975-01-01

    The CEA has tested a variety of ''slow'' self-powered neutron detectors with rhodium, silver and vanadium emitters. Currently there are 120 vanadium detectors in the EL4 heavy water reactor. In addition, ''fast'' detectors with cobalt emitters have been tested at Saclay and 50 of these are in reactor. Other studies are concerned with 6 mm diameter miniature fission chambers. Two fast response chambers with argon-nitrogen filling gas became slow during irradiation, but operated to 600 deg C. An argon filled chamber of 4.7 mm diameter, for traversing in core system in pressurized water reactor, has shown satisfactory test results. (author)

  2. Development of research reactor simulator and its application to dynamic test-bed

    International Nuclear Information System (INIS)

    Kwon, Kee-Choon; Baang, Dane; Park, Jae-Chang; Lee, Seung-Wook; Bae, Sung Won

    2014-01-01

    We developed a real-time simulator for 'High-flux Advanced Neutron Application ReactOr (HANARO), and the Jordan Research and Training Reactor (JRTR). The main purpose of this simulator is operator training, but we modified this simulator into a dynamic test-bed (DTB) to test the functions and dynamic control performance of reactor regulating system (RRS) in HANARO or JRTR before installation. The simulator hardware consists of a host computer, 6 operator stations, a network switch, and a large display panel. The software includes a mathematical model that implements plant dynamics in real-time, an instructor station module that manages user instructions, and a human machine interface module. The developed research reactor simulators are installed in the Korea Atomic Energy Research Institute nuclear training center for reactor operator training. To use the simulator as a dynamic test-bed, the reactor regulating system modeling software of the simulator was replaced by actual RRS cabinet, and was interfaced using a hard-wired and network-based interface. RRS cabinet generates control signals for reactor power control based on the various feedback signals from DTB, and the DTB runs plant dynamics based on the RRS control signals. Thus the Hardware-In-the-Loop Simulation between RRS and the emulated plant (DTB) has been implemented and tested in this configuration. The test result shows that the developed DTB and actual RRS cabinet works together simultaneously resulting in quite good dynamic control performances. (author)

  3. Reactor benchmarks and integral data testing and feedback into ENDF/B-VI

    International Nuclear Information System (INIS)

    McKnight, R.D.; Williams, M.L.

    1992-01-01

    The role of integral data testing and its feedback into the ENDF/B evaluated nuclear data files are reviewed. The use of the CSEWG reactor benchmarks in the data testing process is discussed and selected results based on ENDF/B Version VI data are presented. Finally, recommendations are given to improve the implementation in future integral data testing of ENDF/B

  4. Requirements, needs, and concepts for a new broad-application test reactor

    International Nuclear Information System (INIS)

    Ryskamp, J.M.; Fletcher, C.D.; Denison, A.B.; Liebenthal, J.L.

    1992-01-01

    For a variety of reasons, including (a) the increasing demands of the 1990s regulatory environment, (b) limited existing test capactiy and capability to satisfy projected future testing missions, and (c) an expected increasing need for nuclear information to support development of advanced reactors, there is a need for requirements and preliminary concepts for a new broad-application test reactor (BATR). These requirements must include consideration not only for a broad range of projected testing missions but also for current and projected regulatory compliance and safety requirements. The requirements will form the basis for development and assessment of preconceptual reactor designs and lead to the identification of key technologies to support the government's long-term strategic and programmatic planning. This paper outlines the need for a new BATR and suggests a few preliminary reactor concepts that can meet that need

  5. REACTOR FUEL ELEMENTS TESTING CONTAINER

    Science.gov (United States)

    Whitham, G.K.; Smith, R.R.

    1963-01-15

    This patent shows a method for detecting leaks in jacketed fuel elements. The element is placed in a sealed tank within a nuclear reactor, and, while the reactor operates, the element is sparged with gas. The gas is then led outside the reactor and monitored for radioactive Xe or Kr. (AEC)

  6. Reactor recirculation pump test loop

    International Nuclear Information System (INIS)

    Taka, Shusei; Kato, Hiroyuki

    1979-01-01

    A test loop for a reactor primary loop recirculation pumps (PLR pumps) has been constructed at Ebara's Haneda Plant in preparation for production of PLR pumps under license from Byron Jackson Pump Division of Borg-Warner Corporation. This loop can simulate operating conditions for test PLR pumps with 130 per cent of the capacity of pumps for a 1100 MWe BWR plant. A main loop, primary cooling system, water demineralizer, secondary cooling system, instrumentation and control equipment and an electric power supply system make up the test loop. This article describes the test loop itself and test results of two PLR pumps for Fukushima No. 2 N.P.S. Unit 1 and one main circulation pump for HAZ Demonstration Test Facility. (author)

  7. Thermal and fast reactor benchmark testing of ENDF/B-6.4

    International Nuclear Information System (INIS)

    Liu Guisheng

    1999-01-01

    The benchmark testing for B-6.4 was done with the same benchmark experiments and calculating method as for B-6.2. The effective multiplication factors k eff , central reaction rate ratios of fast assemblies and lattice cell reaction rate ratios of thermal lattice cell assemblies were calculated and compared with testing results of B-6.2 and CENDL-2. It is obvious that 238 U data files are most important for the calculations of large fast reactors and lattice thermal reactors. However, 238 U data in the new version of ENDF/B-6 have not been renewed. Only data of 235 U, 27 Al, 14 N and 2 D have been renewed in ENDF/B-6.4. Therefor, it will be shown that the thermal reactor benchmark testing results are remarkably improved and the fast reactor benchmark testing results are not improved

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

    Science.gov (United States)

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

    2011-01-01

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

  9. 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).

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

    International Nuclear Information System (INIS)

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

    1986-09-01

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

  11. Plutonium-containing aerosols found within containment enclosures in industrial mixed-oxide reactor fuel fabrication

    International Nuclear Information System (INIS)

    Newton, G.J.; Yeh, H.C.; Stanley, J.A.

    1977-01-01

    Mixed oxide (PuO 2 and UO 2 ) nuclear reactor fuel pellets are fabricated within safety enclosures at Babcock and Wilcox's Park Township site near Apollo, PA. Forty-two sample runs of plutonium-containing aerosols were taken from within glove boxes during routine industrial operations. A small, seven-stage cascade impactor and the Lovelace Aerosol Particle Separator (LAPS) were used to determine aerodynamic size distribution and gross alpha aerosol concentration. Powder comminution and blending produced aerosols with lognormal size distributions characterized by activity median aerodynamic diameters (AMAD) of 1.89 +- 0.33 μm, sigma/sub g/ = 1.62 +- 0.09 and a gross alpha aerosol concentration range of 0.1 to 150 nCi/l. Slug pressing and grinding produced aerosols of AMAD = 3.08 +- 0.1 μm, sigma/sub g/ = 1.53 +- 0.01 and AMAD = 2.26 +- 0.16 μm, sigma/sub g/ = 1.68 +- 0.20, respectively. Gross alpha aerosol concentrations ranged from 3.4 to 450 nCi/l. Centerless grinding produced similar-sized aerosols but the gross alpha concentration ranged from 220 to 1690 nCi/l. In vitro solubility studies on selected LAPS samples in a lung fluid simulant indicate that plutonium mixed-oxide aerosols are more soluble than laboratory-produced plutonium aerosols

  12. Research and developments on nondestructive testing in fabrications of fast breeder reactor structural components in Japan

    International Nuclear Information System (INIS)

    Ebata, M.; Ooka, K.; Miyoshi, S.; Senda, T.

    1985-01-01

    Research and developments (R and D) have been conducted on the nondestructive testing techniques necessary for the construction of fast breeder reactor (FBR). Radiographic tests have been made on tube-tube plate welds and small-diameter tube welds, etc. Ultrasonic tests have been conducted on austenitic stainless steel welds. In the penetrant tests and magnetic particle tests, the investigations have been performed on the effects of various test factors on the test results

  13. Proceedings of the international meeting on reduced enrichment for research and test reactors

    International Nuclear Information System (INIS)

    Tsuchihashi, Keichiro

    1984-05-01

    The purpose of the Meeting was to exchange and discuss the most up-to-date information on the progress of various programs related to research and test reactor core conversion from high enriched uranium to lower enriched uranium. The papers presented during the Meeting were divided into 9 sessions and one round able discussion which concluded the Meeting. The Sessions were: Program, Fuel Development, Fuel Fabrication, Irradiation testing, Safety Analysis, Special Reactor Conversion, Reactor Design, Critical Experiments, and Reprocessing and Spent Fuel Storage. Thus, topics of this Meeting were of a very wide range that was expected to result in information exchange valuable for all the participants in the RERTR program

  14. Proceedings of the international meeting on reduced enrichment for research and test reactors

    Energy Technology Data Exchange (ETDEWEB)

    Tsuchihashi, Keichiro [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1984-05-01

    The purpose of the Meeting was to exchange and discuss the most up-to-date information on the progress of various programs related to research and test reactor core conversion from high enriched uranium to lower enriched uranium. The papers presented during the Meeting were divided into 9 sessions and one round able discussion which concluded the Meeting. The Sessions were: Program, Fuel Development, Fuel Fabrication, Irradiation testing, Safety Analysis, Special Reactor Conversion, Reactor Design, Critical Experiments, and Reprocessing and Spent Fuel Storage. Thus, topics of this Meeting were of a very wide range that was expected to result in information exchange valuable for all the participants in the RERTR program.

  15. Preconceptual design and analysis of a solid-breeder blanket test in an existing fission reactor

    International Nuclear Information System (INIS)

    Deis, G.A.; Hsu, P.Y.; Watts, K.D.

    1983-01-01

    Preconceptual design and analysis have been performed to examine the capabilities of a proposed fission-based test of a water-cooled Li 2 O blanket concept. The mechanical configuration of the test piece is designed to simulate a unit cell of a breeder-outside-tube concept. This test piece will be placed in a fission test reactor, which provides an environment similar to that in a fusion reactor. The neutron/gamma flux from the reactor produces prototypical power density, tritium production rates, and operating temperatures and stresses. Steady-state tritium recovery from the test piece can be attained in short-duration (5-to-6-day) tests. The capabilities of this test indicate that fission-based testing can provide important near-term engineering information to support the development of fusion technology

  16. Proceedings of the international meeting on development, fabrication and application of reduced enrichment fuels for research and test reactors

    International Nuclear Information System (INIS)

    1983-08-01

    Separate abstracts were prepared for each of the papers presented in the following areas: (1) Reduced Enrichment Fuels for Research and Test Reactors (RERTR) Program Status; (2) Fuel Development; (3) Fuel Demonstrations; (4) General Topics; and (5) Specific Reactor Applications

  17. New facilities in Japan materials testing reactor for irradiation test of fusion reactor components

    International Nuclear Information System (INIS)

    Kawamura, H.; Sagawa, H.; Ishitsuka, E.; Sakamoto, N.; Niiho, T.

    1996-01-01

    The testing and evaluation of fusion reactor components, i.e. blanket, plasma facing components (divertor, etc.) and vacuum vessel with neutron irradiation is required for the design of fusion reactor components. Therefore, four new test facilities were developed in the Japan Materials Testing Reactor: an in-pile functional testing facility, a neutron multiplication test facility, an electron beam facility, and a re-weldability facility. The paper describes these facilities

  18. Testing and qualification of Control and Safety Rod and its drive mechanism of Fast Breeder Reactor

    International Nuclear Information System (INIS)

    Rajan Babu, V.; Veerasamy, R.; Patri, Sudheer; Ignatius Sundar Raj, S.; Kumar Krovvidi, S.C.S.P.; Dash, S.K.; Meikandamurthy, C.; Rajan, K.K.; Puthiyavinayagam, P.; Chellapandi, P.; Vaidyanathan, G.; Chetal, S.C.

    2010-01-01

    Prototype Fast Breeder Reactor (PFBR) has two independent fast acting diverse shutdown systems. The absorber rod of the first system is called Control and Safety Rod (CSR). CSR and its Drive Mechanism (CSRDM) are used for reactor control and for safe shutdown of the reactor by scram action. In view of the safety role, the qualification of CSRDM is one of the important requirements. CSR and CSRDM were qualified in two stages by extensive testing. In the first stage, the critical subassemblies of the mechanism, such as scram release electromagnet, hydraulic dashpot and dynamic seals and CSR subassembly, were tested and qualified individually simulating the operating conditions of the reactor. Experiments were also carried out on sodium vapour deposition in the annular gaps between the stationary and mobile parts of the mechanism. In the second stage, full-scale CSRDM and CSR were subjected to all the integrated functional tests in air, hot argon and subsequently in sodium simulating the operating conditions of the reactor and finally subjected to endurance tests. Since the damage occurring in CSRDM and CSR is mainly due to fatigue cycles during scram actions, the number of test cycles was decided based on the guidelines given in ASME, Section III, Div. 1. The results show that the performance of CSRDM and CSR is satisfactory. Subsequent to the testing in sodium, the assemblies having contact with liquid sodium/sodium vapour were cleaned using CO 2 process and the total cleaning process has been established, so that the mechanism can be reused in sodium. The various stages of qualification programmes have raised the confidence level on the performance of the system as a whole for the intended and reliable operation in the reactor.

  19. Reactor building pressure proof test (PPT) and leak rate test (LRT) of Qinshan phase III (CANDU) project

    International Nuclear Information System (INIS)

    Gu Jun; Shi Jinqi; Fan Fuping

    2004-12-01

    As the first reactor building (R/B) without stainless steel liner in china, TQNPC studied the containment characteristics, such as strong concrete absorb/release air effect, poor containment penetration. etc. And carefully prepared test scheme and emergency response, creatively introduced the instrument air self-supply system in reactor building, developed the special measurement and analysis system for PPT and LRT, organized work under high-pressure on large-scale in the test. Finally got the containment leak rate result and the test-cost-time value is the best in all same type tests. (authors)

  20. Reduced Enrichment for Research and Test Reactors. Proceedings of the XIV international meeting

    Energy Technology Data Exchange (ETDEWEB)

    Suripto, Asmedi; Hastowo, Hudi; Hersubeno, J B [eds.

    1995-07-01

    Apart from the progress of the Reduced Enrichment Research and Test Reactor (RERTR) Program the national programs of Indonesia, Japan and China were presented. The major events, findings, and activities of 1991 are reviewed with a brief summary of the results which the RERTR Program had achieved by the end of 1990 in collaboration with its many international partners. The RERTR program, has concentrated its efforts on technology transfer and implementation activities consistent with the guidance received from the Department of Energy at the end of 1990. A number of presentations were devoted to development of new fuel uranium silicide fuel elements, fuel irradiation testing and reactor core conversions from highly enriched (HEU) to slightly enriched uranium (LEU). Calculations and measurements of converted reactor core parameters were shown related to safety test and analysis. Fuel cycle issue were discussed as well. One should note that a significant number of papers were devoted to Indonesian GA SIWABESSY reactor core conversion and related topics.

  1. Reduced Enrichment for Research and Test Reactors. Proceedings of the XIV international meeting

    International Nuclear Information System (INIS)

    Suripto, Asmedi; Hastowo, Hudi; Hersubeno, J.B.

    1995-01-01

    Apart from the progress of the Reduced Enrichment Research and Test Reactor (RERTR) Program the national programs of Indonesia, Japan and China were presented. The major events, findings, and activities of 1991 are reviewed with a brief summary of the results which the RERTR Program had achieved by the end of 1990 in collaboration with its many international partners. The RERTR program, has concentrated its efforts on technology transfer and implementation activities consistent with the guidance received from the Department of Energy at the end of 1990. A number of presentations were devoted to development of new fuel uranium silicide fuel elements, fuel irradiation testing and reactor core conversions from highly enriched (HEU) to slightly enriched uranium (LEU). Calculations and measurements of converted reactor core parameters were shown related to safety test and analysis. Fuel cycle issue were discussed as well. One should note that a significant number of papers were devoted to Indonesian GA SIWABESSY reactor core conversion and related topics

  2. Irradiation facilitates at the advanced test reactor

    International Nuclear Information System (INIS)

    Grover, Blaine S.

    2006-01-01

    The Advanced Test Reactor (ATR) is the third generation and largest test reactor built in the Reactor Technology Complex (RTC - formerly known as the Test Reactor Area), located at the Idaho National Laboratory (INL), to study the effects of intense neutron and gamma radiation on reactor materials and fuels. The RTC was established in the early 1950's with the development of the Materials Testing Reactor (MTR), which operated until 1970. The second major reactor was the Engineering Test Reactor (ETR), which operated from 1957 to 1981, and finally the ATR, which began operation in 1967 and will continue operation well into the future. These reactors have produced a significant portion of the world's data on materials response to reactor environments. The wide range of experiment facilities in the ATR and the unique ability to vary the neutron flux in different areas of the core allow numerous experiment conditions to co-exist during the same reactor operating cycle. Simple experiments may involve a non-instrumented capsule containing test specimens with no real-time monitoring or control capabilities. More sophisticated testing facilities include inert gas temperature control systems and pressurized water loops that have continuous chemistry, pressure, temperature, and flow control as well as numerous test specimen monitoring capabilities. There are also apparatus that allow for the simulation of reactor transients on test specimens. The paper has the following contents: ATR description and capabilities; ATR operations, quality and safety requirements; Static capsule experiments; Lead experiments; Irradiation test vehicle; In-pile loop experiments; Gas test loop; Future testing; Support facilities at RTC; Conclusions. To summarize, the ATR has a long history in fuel and material irradiations, and will be fulfilling a critical role in the future fuel and material testing necessary to develop the next generation reactor systems and advanced fuel cycles. The

  3. Status of the RERTR [Reduced Enrichment Research and Test Reactor] program in Argentina

    International Nuclear Information System (INIS)

    Giorsetti, D.R.

    1987-01-01

    The Argentine Atomic Energy Commission started in 1978 the Reduced Enrichment Research and Test Reactors in the field of reactor engineering; engineering, development and manufacturing of fuel elements and research reactors operators. This program was initiated with the conviction that it would contribute to the international efforts to reduce risks of nuclear weapons proliferation owing to an uncontrolled use of highly enriched uranium. It was intended to convert RA-3 reactor to make possible its operation with low enriched fuel (LEU), instead of high enriched fuel (HEU) and to develop manufacturing techniques for said LEU. Afterwards, this program was adapted to assist other countries in reactors conversion, development of the corresponding fuel elements and supply of fuel elements to other countries. (Author)

  4. Oxidation of aluminum alloy cladding for research and test reactor fuel

    Science.gov (United States)

    Kim, Yeon Soo; Hofman, G. L.; Robinson, A. B.; Snelgrove, J. L.; Hanan, N.

    2008-08-01

    The oxide thicknesses on aluminum alloy cladding were measured for the test plates from irradiation tests RERTR-6 and 7A in the ATR (advanced test reactor). The measured thicknesses were substantially lower than those of test plates with similar power from other reactors available in the literature. The main reason is believed to be due to the lower pH (pH 5.1-5.3) of the primary coolant water in the ATR than in the other reactors (pH 5.9-6.5) for which we have data. An empirical model for oxide film thickness predictions on aluminum alloy used as fuel cladding in the test reactors was developed as a function of irradiation time, temperature, surface heat flux, pH, and coolant flow rate. The applicable ranges of pH and coolant flow rates cover most research and test reactors. The predictions by the new model are in good agreement with the in-pile test data available in the literature as well as with the RERTR test data measured in the ATR.

  5. Oxidation of aluminum alloy cladding for research and test reactor fuel

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Yeon Soo [Argonne National Laboratory, Nuclear Engineering, 9700 South Cass Avenue, Argonne, IL 60439 (United States)], E-mail: yskim@anl.gov; Hofman, G.L. [Argonne National Laboratory, Nuclear Engineering, 9700 South Cass Avenue, Argonne, IL 60439 (United States); Robinson, A.B. [Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-6188 (United States); Snelgrove, J.L.; Hanan, N. [Argonne National Laboratory, Nuclear Engineering, 9700 South Cass Avenue, Argonne, IL 60439 (United States)

    2008-08-31

    The oxide thicknesses on aluminum alloy cladding were measured for the test plates from irradiation tests RERTR-6 and 7A in the ATR (advanced test reactor). The measured thicknesses were substantially lower than those of test plates with similar power from other reactors available in the literature. The main reason is believed to be due to the lower pH (pH 5.1-5.3) of the primary coolant water in the ATR than in the other reactors (pH 5.9-6.5) for which we have data. An empirical model for oxide film thickness predictions on aluminum alloy used as fuel cladding in the test reactors was developed as a function of irradiation time, temperature, surface heat flux, pH, and coolant flow rate. The applicable ranges of pH and coolant flow rates cover most research and test reactors. The predictions by the new model are in good agreement with the in-pile test data available in the literature as well as with the RERTR test data measured in the ATR.

  6. Reactor physics tests and benchmark analyses of STACY

    International Nuclear Information System (INIS)

    Miyoshi, Yoshinori; Umano, Takuya

    1996-01-01

    The Static Experiment Critical Facility, STACY in the Nuclear Fuel Cycle Safety Engineering Research Facility, NUCEF is a solution type critical facility to accumulate fundamental criticality data on uranyl nitrate solution, plutonium nitrate solution and their mixture. A series of critical experiments have been performed for 10 wt% enriched uranyl nitrate solution using a cylindrical core tank. In these experiments, systematic data of the critical height, differential reactivity of the fuel solution, kinetic parameter and reactor power were measured with changing the uranium concentration of the fuel solution from 313 gU/l to 225 gU/l. Critical data through the first series of experiments for the basic core are reported in this paper for evaluating the accuracy of the criticality safety calculation codes. Benchmark calculations of the neutron multiplication factor k eff for the critical condition were made using a neutron transport code TWOTRAN in the SRAC system and a continuous energy Monte Carlo code MCNP 4A with a Japanese evaluated nuclear data library, JENDL 3.2. (J.P.N.)

  7. The U.S. reduced enrichment research and test reactor (RERTR) program

    International Nuclear Information System (INIS)

    Travelli, A.

    1993-01-01

    Research and test reactors are widely deployed to study the irradiation behavior of materials of interest in nuclear engineering, to produce radioisotopes for medicine, industry, and agriculture, and as a basic research and teaching tool. In order to maximize neutron flux per unit power and/or to minimize capital costs and fuel cycle costs, most of these reactors were de- signed to utilize uranium with very high enrichment (in the 70% to 95% range). Research reactor fuels with such high uranium enrichment cause a potential risk of nuclear weapons proliferation. Over 140 research and test reactors of significant power (between 10 kW and 250 MW) are in operation with very highly enriched uranium in more than 35 countries, with total power in excess of 1,700 MW. The overall annual fuel requirement of these reactors corresponds to approximately 1,200 kg of 235 U. This highly strategic material is normally exported from the United States, converted to metal form, shipped to a fuel fabricator, and then shipped to the reactor site in finished fuel element form. At the reactor site the fuel is first stored, then irradiated, stored again, and eventually shipped back to the United States for reprocessing. The whole cycle takes approximately four years to complete, bringing the total required fuel inventory to approximately 5,000 kg of 235 U. The resulting international trade in highly-enriched uranium may involve several countries in the process of refueling a single reactor and creates a considerable concern that the highly-enriched uranium may be diverted for non-peaceful purposes while in fabrication, transport, or storage, particularly when it is in the unirradiated form. The proliferation resistance of nuclear fuels used in research and test reactors can be considerably improved by reducing their uranium enrichment to a value less than 20%, but significantly greater than natural to avoid excessive plutonium production

  8. Analyses and testing of model prestressed concrete reactor vessels with built-in planes of weakness

    International Nuclear Information System (INIS)

    Dawson, P.; Paton, A.A.; Fleischer, C.C.

    1990-01-01

    This paper describes the design, construction, analyses and testing of two small scale, single cavity prestressed concrete reactor vessel models, one without planes of weakness and one with planes of weakness immediately behind the cavity liner. This work was carried out to extend a previous study which had suggested the likely feasibility of constructing regions of prestressed concrete reactor vessels and biological shields, which become activated, using easily removable blocks, separated by a suitable membrane. The paper describes the results obtained and concludes that the planes of weakness concept could offer a means of facilitating the dismantling of activated regions of prestressed concrete reactor vessels, biological shields and similar types of structure. (author)

  9. Reliability and testing considerations in the design of nuclear reactor filtration systems

    International Nuclear Information System (INIS)

    O'Nan, A.; Williams, R.P.; Goldsmith, J.M.

    1975-01-01

    The high performance standards set by USAEC-DRL Regulatory Guides for nuclear reactor filtration systems pose difficult problems for on-site leakage tests. These problems are compounded by the crowded conditions inside reactor structures, and by the fact that, until recently, little consideration has been given by system designers to the needs of testing. Techniques for coping with testing problems on existing systems, and suggestions for improving the testability of future systems, are given. Test crew safety considations are discussed, and a pair of easily portable contaminant generators is described. (U.S.)

  10. Installation and testing of the ERANOS computer code for fast reactor calculations

    International Nuclear Information System (INIS)

    Gren, Milan

    2010-12-01

    The French ERANOS computer code was acquired and tested by solving benchmark problems. Five problems were calculated: 1D XZ Model, 1D RZ Model, 3D HEX SNR 300 reactor, 2S HEX and 3D HEX VVER 440 reactor. The multi-group diffuse approximation was used. The multiplication coefficients were compared within the first problem, neutron flux density in the calculation points was obtained within the second problem, and powers in the various reactor areas and in the assemblies were calculated within the remaining problems. (P.A.)

  11. Fabrication and Testing of a Modular Micro-Pocket Fission Detector Instrumentation System for Test Nuclear Reactors

    Science.gov (United States)

    Reichenberger, Michael A.; Nichols, Daniel M.; Stevenson, Sarah R.; Swope, Tanner M.; Hilger, Caden W.; Roberts, Jeremy A.; Unruh, Troy C.; McGregor, Douglas S.

    2018-01-01

    Advancements in nuclear reactor core modeling and computational capability have encouraged further development of in-core neutron sensors. Measurement of the neutron-flux distribution within the reactor core provides a more complete understanding of the operating conditions in the reactor than typical ex-core sensors. Micro-Pocket Fission Detectors have been developed and tested previously but have been limited to single-node operation and have utilized highly specialized designs. The development of a widely deployable, multi-node Micro-Pocket Fission Detector assembly will enhance nuclear research capabilities. A modular, four-node Micro-Pocket Fission Detector array was designed, fabricated, and tested at Kansas State University. The array was constructed from materials that do not significantly perturb the neutron flux in the reactor core. All four sensor nodes were equally spaced axially in the array to span the fuel-region of the reactor core. The array was filled with neon gas, serving as an ionization medium in the small cavities of the Micro-Pocket Fission Detectors. The modular design of the instrument facilitates the testing and deployment of numerous sensor arrays. The unified design drastically improved device ruggedness and simplified construction from previous designs. Five 8-mm penetrations in the upper grid plate of the Kansas State University TRIGA Mk. II research nuclear reactor were utilized to deploy the array between fuel elements in the core. The Micro-Pocket Fission Detector array was coupled to an electronic support system which has been specially developed to support pulse-mode operation. The Micro-Pocket Fission Detector array composed of four sensors was used to monitor local neutron flux at a constant reactor power of 100 kWth at different axial locations simultaneously. The array was positioned at five different radial locations within the core to emulate the deployment of multiple arrays and develop a 2-dimensional measurement of

  12. The SPHINX reactor for engineering tests

    International Nuclear Information System (INIS)

    Adamov, E.O.; Artamkin, K.N.; Bovin, A.P.; Bulkin, Y.M.; Kartashev, E.F.; Korneev, A.A.; Stenbok, I.A.; Terekhov, A.S.; Khmel'Shehikov, V.V.; Cherkashov, Y.M.

    1990-01-01

    A research reactor known as SPHINX is under development in the USSR. The reactor will be used mainly to carry out tests on mock-up power reactor fuel assemblies under close-to-normal parameters in experimental loop channels installed in the core and reflector of the reactor, as well as to test samples of structural materials in ampoule and loop channels. The SPHINX reactor is a channel-type reactor with light-water coolant and moderator. Maximum achievable neutron flux density in the experimental channels (cell composition 50% Fe, 50% H 2 O) is 1.1 X 10 15 neutrons/cm 2 · s for fast neutrons (E > 0.1 MeV) and 1.7 X 10 15 for thermal neutrons at a reactor power of 200 MW. The design concepts used represent a further development of the technical features which have met with approval in the MR and MIR channel-type engineering test reactors currently in use in the USSR. The 'in-pond channel' construction makes the facility flexible and eases the carrying out of experimental work while keeping discharges of radioactivity into the environment to a low level. The reactor and all associated buildings and constructions conform to modern radiation safety and environmental protection requirements

  13. Handbook of materials testing reactors and ancillary hot laboratories in the European Community

    International Nuclear Information System (INIS)

    1977-01-01

    The purpose of this Handbook is to make available to those interested in 'in-pile' irradiation experiments important data on Materials Testing Reactors in operation in the European Community. Only thermal reactors having a power output of more than 5 MW(th) are taken into consideration. In particular, detailed technical information is given on the experimental irradiation facilities of the reactors, their specialized irradiation devices (loops and instrumented capsules), and the associated hot cell facilities for post-irradiation examination of samples

  14. The feature of high flux engineering test reactor and its role in nuclear power development

    International Nuclear Information System (INIS)

    Lu Guangquan

    1987-01-01

    The High Flux Engineering Test Reactor (HFETR) designed and built by Chinese own efforts reached to its initial criticality on Dec. 27, 1979, and then achieved high power operation on Dec. 16, 1980. Until Nov. 11. 1986, the reactor had been operated for thirteen cycles. The paper presents briefly main feature of HFETR and its utilization during past years. The paper also deals with its role in nuclear power development. Finally, author gives his opinion on comprehensive utilization of HFETR. (author)

  15. Irradiation experiments and materials testing capabilities in High Flux Reactor in Petten

    International Nuclear Information System (INIS)

    Luzginova, N.; Blagoeva, D.; Hegeman, H.; Van der Laan, J.

    2011-01-01

    The text of publication follows: The High Flux Reactor (HFR) in Petten is a powerful multi-purpose research and materials testing reactor operating for about 280 Full Power Days per year. In combination with hot cells facilities, HFR provides irradiation and post-irradiation examination services requested by nuclear energy research and development programs, as well as by industry and research organizations. Using a variety of the custom developed irradiation devices and a large experience in executing irradiation experiments, the HFR is suitable for fuel, materials and components testing for different reactor types. Irradiation experiments carried out at the HFR are mainly focused on the understanding of the irradiation effects on materials; and providing databases for irradiation behavior of materials to feed into safety cases. The irradiation experiments and materials testing at the HFR include the following issues. First, materials irradiation to support the nuclear plant life extensions, for instance, characterization of the reactor pressure vessel stainless steel claddings to insure structural integrity of the vessel, as well as irradiation of the weld material coupons to neutron fluence levels that are representative for Light Water Reactors (LWR) internals applications. Secondly, development and qualification of the structural materials for next generation nuclear fission reactors as well as thermo-nuclear fusion machines. The main areas of interest are in both conventional stainless steel and advanced reduced activation steels and special alloys such as Ni-base alloys. For instance safety-relevant aspects of High Temperature Reactors (HTR) such as the integrity of fuel and structural materials with increasing neutron fluence at typical HTR operating conditions has been recently assessed. Thirdly, support of the fuel safety through several fuel irradiation experiments including testing of pre-irradiated LWR fuel rods containing UO 2 or MOX fuel. Fourthly

  16. Reactor building integrity testing: A novel approach at Gentilly 2 - principles and methodology

    International Nuclear Information System (INIS)

    Collins, N.; Lafreniere, P.

    1991-01-01

    In 1987, Hydro-Quebec embarked on an ambitious development program to provide the Gentilly 2 nuclear power station with an effective, yet practical reactor building Integrity Test. The Gentilly 2 Integrity Test employs an innovative approach based on the reference volume concept. It is identified as the Temperature Compensation Method (TCM) System. This configuration has been demonstrated at both high and low test pressure and has achieved extraordinary precision in the leak rate measurement. The Gentilly 2 design allows the Integrity Test to be performed at a nominal 3 kPa(g) test pressure during an (11) hour period with the reactor at full power. The reactor building Pressure Test by comparison, is typically performed at high pressure 124 kPa(g)) in a 7 day window during an annual outage. The Integrity Test was developed with the goal of demonstrating containment availability. Specifically it was purported to detect a leak or hole in the 'bottled-up' reactor building greater in magnitude than an equivalent pipe of 25 mm diameter. However it is considered feasible that the high precision of the Gentilly 2 TCM System Integrity Test and a stable reactor building leak characteristic will constitute sufficient grounds for the reduction of the Pressure Test frequency. It is noted that only the TCM System has, to this date, allowed a relevant determination of the reactor building leak rate at a nominal test pressure of 3 kPa(g). Classical method tests at low pressure have lead to inconclusive results due to the high lack of precision

  17. Test facility for auxiliary cooling system (ACS) of fast breeder reactor for Power Reactor and Nuclear Fuel Development Corporation (PNC)

    International Nuclear Information System (INIS)

    1983-01-01

    In preparation of constructing ''Monju'', a prototype fast breeder reactor, PNC has been pushing forward its research and development projects and the ACS was constructed under these projects. The auxiliary cooling system is an important engineered safety feature, and is used for safe removal of heat from the reactor at the shutdown. The ACS serves as a means of testing and assessing the auxiliary cooling system for the ''Monju'' and is designed and manufactured to have one fifth capacity of the Monju. The air heat exchanger and the ACS system was designed to withstand higher temperature range of the conventional design code (MITI-501), and finned tubes were applied for effective heat removal. Preheating system was designed to heat up the whole system over 200 0 C within 20 hours to prevent sodium from freezing. Basic performance of ACS was verified satisfactorily by a series of performance tests, such as start up test, flow rate measurement and preheating test before delivery. The experience from designing and construction of ACS and data obtained by these tests will be very instructive for designing and construction of the ''Monju''. (author)

  18. Warm Water Oxidation Verification - Scoping and Stirred Reactor Tests

    Energy Technology Data Exchange (ETDEWEB)

    Braley, Jenifer C.; Sinkov, Sergey I.; Delegard, Calvin H.; Schmidt, Andrew J.

    2011-06-15

    Scoping tests to evaluate the effects of agitation and pH adjustment on simulant sludge agglomeration and uranium metal oxidation at {approx}95 C were performed under Test Instructions(a,b) and as per sections 5.1 and 5.2 of this Test Plan prepared by AREVA. (c) The thermal testing occurred during the week of October 4-9, 2010. The results are reported here. For this testing, two uranium-containing simulant sludge types were evaluated: (1) a full uranium-containing K West (KW) container sludge simulant consisting of nine predominant sludge components; (2) a 50:50 uranium-mole basis mixture of uraninite [U(IV)] and metaschoepite [U(VI)]. This scoping study was conducted in support of the Sludge Treatment Project (STP) Phase 2 technology evaluation for the treatment and packaging of K-Basin sludge. The STP is managed by CH2M Hill Plateau Remediation Company (CHPRC) for the U.S. Department of Energy. Warm water ({approx}95 C) oxidation of sludge, followed by immobilization, has been proposed by AREVA and is one of the alternative flowsheets being considered to convert uranium metal to UO{sub 2} and eliminate H{sub 2} generation during final sludge disposition. Preliminary assessments of warm water oxidation have been conducted, and several issues have been identified that can best be evaluated through laboratory testing. The scoping evaluation documented here was specifically focused on the issue of the potential formation of high strength sludge agglomerates at the proposed 95 C process operating temperature. Prior hydrothermal tests conducted at 185 C produced significant physiochemical changes to genuine sludge, including the formation of monolithic concretions/agglomerates that exhibited shear strengths in excess of 100 kPa (Delegard et al. 2007).

  19. Warm Water Oxidation Verification - Scoping and Stirred Reactor Tests

    International Nuclear Information System (INIS)

    Braley, Jenifer C.; Sinkov, Sergey I.; Delegard, Calvin H.; Schmidt, Andrew J.

    2011-01-01

    Scoping tests to evaluate the effects of agitation and pH adjustment on simulant sludge agglomeration and uranium metal oxidation at ∼95 C were performed under Test Instructions(a,b) and as per sections 5.1 and 5.2 of this Test Plan prepared by AREVA. (c) The thermal testing occurred during the week of October 4-9, 2010. The results are reported here. For this testing, two uranium-containing simulant sludge types were evaluated: (1) a full uranium-containing K West (KW) container sludge simulant consisting of nine predominant sludge components; (2) a 50:50 uranium-mole basis mixture of uraninite (U(IV)) and metaschoepite (U(VI)). This scoping study was conducted in support of the Sludge Treatment Project (STP) Phase 2 technology evaluation for the treatment and packaging of K-Basin sludge. The STP is managed by CH2M Hill Plateau Remediation Company (CHPRC) for the U.S. Department of Energy. Warm water (∼95 C) oxidation of sludge, followed by immobilization, has been proposed by AREVA and is one of the alternative flowsheets being considered to convert uranium metal to UO 2 and eliminate H 2 generation during final sludge disposition. Preliminary assessments of warm water oxidation have been conducted, and several issues have been identified that can best be evaluated through laboratory testing. The scoping evaluation documented here was specifically focused on the issue of the potential formation of high strength sludge agglomerates at the proposed 95 C process operating temperature. Prior hydrothermal tests conducted at 185 C produced significant physiochemical changes to genuine sludge, including the formation of monolithic concretions/agglomerates that exhibited shear strengths in excess of 100 kPa (Delegard et al. 2007).

  20. Strain measurement in and analysis for hydraulic test of CPR1000 reactor pressure vessel

    International Nuclear Information System (INIS)

    Zhou Dan; Zhuang Dongzhen

    2013-01-01

    The strain measurement in hydraulic test of CPR1000 reactor pressure vessel performed in Dongfang Heavy Machinery Co., Ltd. is introduced. The detail test scheme and method was introduced and the measurement results of strain and stress was given. Meanwhile the finite element analysis was performed for the pressure vessel, which was generally matched with the measurement results. The reliability of strain measurement was verified and the high strength margin of vessel was shown, which would give a good reference value for the follow-up hydraulic tests and strength analysis of reactor pressure vessel. (authors)

  1. Further development of remote testing of submerged bolts and screws in reactors

    International Nuclear Information System (INIS)

    Mohr, F.; Schirner, G.; Meier, R.; Wiesinger, W.

    2007-01-01

    Since the eighties, intelligeNDT has been carrying out ultrasonic tests of bolts in reactor containments and pressure vessels both in Germany and abroad. The ultrasonic equipment used belonged to the SAPHIR/SAPHIRplus line. The recording and online evaluation software was adapted to the test requirements and optimized for high test rates and quality-assured documentation. As test manipulator, the ''SUSI'' submarine by AREVA NP was used with good results. (orig.)

  2. NRC review of passive reactor design certification testing programs: Overview, progress, and regulatory perspective

    Energy Technology Data Exchange (ETDEWEB)

    Levin, A.E.

    1995-09-01

    New reactor designs, employing passive safety systems, are currently under development by reactor vendors for certification under the U.S. Nuclear Regulatory Commission`s (NRC`s) design certification rule. The vendors have established testing programs to support the certification of the passive designs, to meet regulatory requirements for demonstration of passive safety system performance. The NRC has, therefore, developed a process for the review of the vendors` testing programs and for incorporation of the results of those reviews into the safety evaluations for the passive plants. This paper discusses progress in the test program reviews, and also addresses unique regulatory aspects of those reviews.

  3. Reliability test for reactor internals rejuvenation technology

    International Nuclear Information System (INIS)

    Uchiyama, Junichi

    1998-01-01

    41 transparencies were presented on the subject of 'Reliability test for reactor internals rejuvenation technology'. The items presented give an introduction on the management of plant life in Japan and introduce the Nuclear Power Engineering Corporation (NUPEC). The question of what reliability tests for rejuvenation of reactor internals are is discussed in some detail and an outline of each test is given. Altogether six methods to rejuvenate reactor internals are presented, two of which have already been applied to actual plants. The presentation was supported by many detailed drawings and images

  4. Design and fabrication of irradiation testing capsule for research reactor materials

    International Nuclear Information System (INIS)

    Yang, Seong Woo; Kim, Bong Goo; Park, Seung Jae; Cho, Man Soon; Choo, Kee Nam; Oh, Jong Myeong; Choi, Myeong Hwan; Lee, Byung Chul; Kang, Suk Hoon; Kim, Dae Jong; Chun, Young Bum; Kim, Tae Kyu

    2012-01-01

    Recently, the demand of research reactors is increasing because there are many ageing research reactors in the world. Also, the production of radioisotope related with the medical purpose is very important. Korea Atomic Energy Research Institute (KAERI) is designing and licensing for Jordan Research and Training Reactor (JRTR) and new type research reactor for export which will be constructed in Amman, Jordan and Busan, Korea, respectively. Thus, It is expected that more research reactors will be designed and constructed by KAERI. To design the research reactor, the irradiation performance and behavior of core structure material are necessary. However, the irradiation behavior of these materials is not yet investigated. Therefore, the irradiation performance must be verified by irradiation test. 11M 20K and 11M 21K irradiation capsules were designed and fabricated to conduct the irradiation test for some candidate core materials, Zircaloy 4, beryllium, and graphite, at HANARO. In this paper, the design and fabrication features of 11M 20K and 11M 21K were discussed

  5. Design and fabrication of irradiation testing capsule for research reactor materials

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Seong Woo; Kim, Bong Goo; Park, Seung Jae; Cho, Man Soon; Choo, Kee Nam; Oh, Jong Myeong; Choi, Myeong Hwan; Lee, Byung Chul; Kang, Suk Hoon; Kim, Dae Jong; Chun, Young Bum; Kim, Tae Kyu [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2012-10-15

    Recently, the demand of research reactors is increasing because there are many ageing research reactors in the world. Also, the production of radioisotope related with the medical purpose is very important. Korea Atomic Energy Research Institute (KAERI) is designing and licensing for Jordan Research and Training Reactor (JRTR) and new type research reactor for export which will be constructed in Amman, Jordan and Busan, Korea, respectively. Thus, It is expected that more research reactors will be designed and constructed by KAERI. To design the research reactor, the irradiation performance and behavior of core structure material are necessary. However, the irradiation behavior of these materials is not yet investigated. Therefore, the irradiation performance must be verified by irradiation test. 11M 20K and 11M 21K irradiation capsules were designed and fabricated to conduct the irradiation test for some candidate core materials, Zircaloy 4, beryllium, and graphite, at HANARO. In this paper, the design and fabrication features of 11M 20K and 11M 21K were discussed.

  6. The Jules Horowitz Reactor - A new High Performance European Material Testing Reactor open to International Users Present Status and Objectives

    International Nuclear Information System (INIS)

    Iracane, Daniel; Bignan, Gilles; Lindbaeck, Jan-Erik; Blomgren, Jan

    2010-01-01

    The development of sustainable nuclear energy requires R and D on fuel and material behaviour under irradiation with a high level of performance in order to meet the needs and challenges for the benefit of industry, research and public bodies. These stakes require a sustainable and secured access to an up-to-date high performance Material Testing Reactor. Following a broad survey within the European Research Area, the international community agreed that the need for Material Test Reactors in support of nuclear power plant safety and operation will continue in the context of sustainable nuclear energy. The Jules Horowitz Reactor project (JHR) copes with this context. JHR is designed as a user facility addressing the needs of the international community. This means: - flexibility with irradiation loops able to reproduce a large variation in operation conditions of different power reactor technologies, - high flux capacity to address Generations II, III, and IV needs. JHR is designed, built and operated as an international user facility because: - Given the maturity and globalization of the industry, domestic tools have no more the required level of economic and technical efficiency. Meanwhile, countries with nuclear energy need an access to high performance irradiation experimental capabilities to support technical skill and guarantee the competitiveness and safety of nuclear energy. - Many research items related to safety or public policy (waste management, etc.) require international cooperation to share costs and benefits of resulting consensus. JHR design is optimised for offering high performance material and fuel irradiation capability for the coming decades. This project is driven and funded by an international consortium gathering vendors, utilities and public stakeholders. This consortium has been set up in March 2007 when the construction began. The construction is in progress and the start of operation is scheduled for 2014. The JHR is a research

  7. The Jules Horowitz reactor, a new high performance European material testing reactor open to international users: present status and objectives

    International Nuclear Information System (INIS)

    Iracane, D.; Bignan, G.

    2010-01-01

    The development of nuclear power as a sustainable and competitive energy source will continue to require research and development of fuel and material behaviour under irradiation. This necessitates a high performance material testing reactor (MTR). Facing the obsolescence of most of the existing MTR in Europe, France decided a few years ago the construction of the RJH (Jules Horowitz reactor). RJH is designed, built and will be operated as an international user facility. A first set of experimental hosting devices is being designed. For instance, there are the in-core CALIPSO Nak integrated loop for material studies and other loops for fuel studies under nominal or off-normal or accidental conditions. The RJH international program will focus on the following subjects: -) fuel reliability, assessed through power ramps tests and post-irradiation examination; -) Loss of coolant tests done out-of-pile in a first phase and in-pile in a possible second phase; and -) source term tests addressing fission products release. The paper reports also the point of view of VATTENFALL (a Swedish power utility), as a potential European RJH user. (A.C.)

  8. The reactor core configuration and important systems related to physics tests of Daya Bay NPP

    International Nuclear Information System (INIS)

    Tao Shaoping

    1995-06-01

    A brief introduction to reactor core configuration and important systems related to physics tests of Daya Bay NPP is given. These systems involve the reactor core system (COR), the full length rod control system (RGL), the in-core instrumentation system (RIC), the out-of-core nuclear instrumentation system (RPN), and the LOCA surveillance system (LSS), the centralized data processing system (KIT) and the test data acquisition system (KDO). In addition, that the adjustment and evaluation of boron concentration related to other systems, for example the reactor coolant system (RCP), the chemical and volume control system (RCV), the reactor boron and water makeup system (REA), the nuclear sampling system (REN) and the reactor control system (RRC), etc. is also described. Analysis of these systems helps not only to familiarize their functions and acquires a deepen understanding for the principle procedure, points for attention and technical key of the core physics tests, but also to further analyze the test results. (3 refs., 11 figs., 1 tab.)

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

    Energy Technology Data Exchange (ETDEWEB)

    1984-07-01

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

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

    International Nuclear Information System (INIS)

    1984-07-01

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

  11. Abbreviated sampling and analysis plan for planning decontamination and decommissioning at Test Reactor Area (TRA) facilities

    International Nuclear Information System (INIS)

    1994-10-01

    The objective is to sample and analyze for the presence of gamma emitting isotopes and hazardous constituents within certain areas of the Test Reactor Area (TRA), prior to D and D activities. The TRA is composed of three major reactor facilities and three smaller reactors built in support of programs studying the performance of reactor materials and components under high neutron flux conditions. The Materials Testing Reactor (MTR) and Engineering Test Reactor (ETR) facilities are currently pending D/D. Work consists of pre-D and D sampling of designated TRA (primarily ETR) process areas. This report addresses only a limited subset of the samples which will eventually be required to characterize MTR and ETR and plan their D and D. Sampling which is addressed in this document is intended to support planned D and D work which is funded at the present time. Biased samples, based on process knowledge and plant configuration, are to be performed. The multiple process areas which may be potentially sampled will be initially characterized by obtaining data for upstream source areas which, based on facility configuration, would affect downstream and as yet unsampled, process areas. Sampling and analysis will be conducted to determine the level of gamma emitting isotopes and hazardous constituents present in designated areas within buildings TRA-612, 642, 643, 644, 645, 647, 648, 663; and in the soils surrounding Facility TRA-611. These data will be used to plan the D and D and help determine disposition of material by D and D personnel. Both MTR and ETR facilities will eventually be decommissioned by total dismantlement so that the area can be restored to its original condition

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

    International Nuclear Information System (INIS)

    Anon.

    1986-01-01

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

  13. Use of research and test reactors for SPD development and calibration

    International Nuclear Information System (INIS)

    LaFontaine, M.W.R.

    2011-01-01

    Prior to using a research or test reactor for performance studies or calibration of self powered detectors, it is first necessary to fully characterize the reactor environment in the region to be utilized. This presentation details Characterization Experiments performed to quantify research/test reactor core/site parameters as they would apply for use with SPD applications. Methods will be described to: Determine the Westcott parameter, r (T n /T o ) , for the region of interest; Characterize the neutron energy spectrum in terms of the cadmium absorption cut-off, i.e., consider neutrons of energy 5kT 0.13 eV to be epithermal neutrons; Determine T n , the effective neutron temperature, in the region of interest; Determine the gamma flux in the region of interest; and, Establish SPD calibration standard detectors.

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

    International Nuclear Information System (INIS)

    Ansari, S.A.

    1990-01-01

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

  15. Basic research on high-uranium density fuels for research and test reactors

    International Nuclear Information System (INIS)

    Ugajin, M.; Itoh, A.; Akabori, M.

    1992-01-01

    High-uranium density fuels, uranium silicides (U 3 Si 2 , U 3 Si) and U 6 Me-type uranium alloys (Me = Fe, Mn, Ni), were prepared and examined metallurgically as low-enriched uranium (LEU) fuels for research and test reactors. Miniature aluminum-dispersion plate-type fuel (miniplate) and aluminum-clad disk-type fuel specimens were fabricated and subjected to the neutron irradiation in JMTR (Japan Materials Testing Reactor). Fuel-aluminum compatibility tests were conducted to elucidate the extent of reaction and to identify reaction products. The relative stability of the fuels in an aluminum matrix was established at 350degC or above. Experiments were also performed to predict the chemical form of the solid fission-products in the uranium silicide (U 3 Si 2 ) simulating a high burnup anticipated for reactor service. (author)

  16. Design of Seismic Test Rig for Control Rod Drive Mechanism of Jordan Research and Training Reactor

    International Nuclear Information System (INIS)

    Sun, Jongoh; Kim, Gyeongho; Yoo, Yeonsik; Cho, Yeonggarp; Kim, Jong In

    2014-01-01

    The reactor assembly is submerged in a reactor pool filled with water and its reactivity is controlled by locations of four control absorber rods(CARs) inside the reactor assembly. Each CAR is driven by a stepping motor installed at the top of the reactor pool and they are connected to each other by a tie rod and an electromagnet. The CARs scram the reactor by de-energizing the electromagnet in the event of a safe shutdown earthquake(SSE). Therefore, the safety function of the control rod drive mechanism(CRDM) which consists of a drive assembly, tie rod and CARs is to drop the CAR into the core within an appropriate time in case of the SSE. As well known, the operability for complex equipment such as the CRDM during an earthquake is very hard to be demonstrated by analysis and should be verified through tests. One of them simulates the reactor assembly and the guide tube of the CAR, and the other one does the pool wall where the drive assembly is installed. In this paper, design of the latter test rig and how the test is performed are presented. Initial design of the seismic test rig and excitation table had its first natural frequency at 16.3Hz and could not represent the environment where the CRDM was installed. Therefore, experimental modal analyses were performed and an FE model for the test rig and table was obtained and tuned based on the experimental results. Using the FE model, the design of the test rig and table was modified in order to have higher natural frequency than the cutoff frequency. The goal was achieved by changing its center of gravity and the stiffness of its sliding bearings

  17. Design of Seismic Test Rig for Control Rod Drive Mechanism of Jordan Research and Training Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Sun, Jongoh; Kim, Gyeongho; Yoo, Yeonsik; Cho, Yeonggarp; Kim, Jong In [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    The reactor assembly is submerged in a reactor pool filled with water and its reactivity is controlled by locations of four control absorber rods(CARs) inside the reactor assembly. Each CAR is driven by a stepping motor installed at the top of the reactor pool and they are connected to each other by a tie rod and an electromagnet. The CARs scram the reactor by de-energizing the electromagnet in the event of a safe shutdown earthquake(SSE). Therefore, the safety function of the control rod drive mechanism(CRDM) which consists of a drive assembly, tie rod and CARs is to drop the CAR into the core within an appropriate time in case of the SSE. As well known, the operability for complex equipment such as the CRDM during an earthquake is very hard to be demonstrated by analysis and should be verified through tests. One of them simulates the reactor assembly and the guide tube of the CAR, and the other one does the pool wall where the drive assembly is installed. In this paper, design of the latter test rig and how the test is performed are presented. Initial design of the seismic test rig and excitation table had its first natural frequency at 16.3Hz and could not represent the environment where the CRDM was installed. Therefore, experimental modal analyses were performed and an FE model for the test rig and table was obtained and tuned based on the experimental results. Using the FE model, the design of the test rig and table was modified in order to have higher natural frequency than the cutoff frequency. The goal was achieved by changing its center of gravity and the stiffness of its sliding bearings.

  18. Subcritical Measurements Research Program for Fresh and Spent Materials Test Reactor Fuels

    International Nuclear Information System (INIS)

    Blanchard, A.

    1999-01-01

    'A series of subcritical noise measurements were performed on fresh and spent University of Missouri Research Reactor fuel assemblies. These experimental measurements were performed for the purposes of providing benchmark quality data for validating transport theory computer codes and nuclear cross-section data used to perform criticality safety analyses for highly enriched, uranium-aluminum Material Test Reactor fuel assemblies. A mechanical test rig was designed and built to hold up to four fuel assemblies and neutron detectors in a subcritical array. The rig provided researchers with the ability to evaluate the reactivity effects of variable fuel/detector spacing, fuel rotation, and insertion of metal reflector plates into the lattice.'

  19. An overview of the fuels and materials testing programme at the OECD Halden Reactor Project

    Energy Technology Data Exchange (ETDEWEB)

    Wiesenack, W [Institutt for Energiteknikk, Halden (Norway). OECD Halden Reaktor Projekt

    1997-08-01

    The fuels and materials testing programme of the OECD Halden Reactor Project is aimed at investigations of fuel and cladding properties at high burnup, water chemistry effects and in-core materials ageing problems. For the execution of this programme, different types of irradiation rigs and experimental facilities providing typical power reactors conditions are available. Data are obtained from in-core sensors developed at the Halden Project; these are shortly described. An overview of the current test programme and the scope of the following years are briefly presented. (author). 5 refs, 3 figs.

  20. The BR2 materials testing reactor. Past, ongoing and under-study upgradings

    Energy Technology Data Exchange (ETDEWEB)

    Baugnet, J M; Roedt, Ch de; Gubel, P; Koonen, E [Centre d' Etude de I' Energie Nucleaire, Studiecentrum voor Kernenergie, C.E.N./S.C.K., Mol (Belgium)

    1990-05-01

    The BR2 reactor (Mol, Belgium) is a high-flux materials testing reactor. The fuel is 93% {sup 235}U enriched uranium. The nominal power ranges from 60 to 100 MW. The main features of the design are the following: 1) maximum neutron flux, thermal: 1.2 x 10{sup 15} n/cm{sup 2} s; fast (E > 0.1 MeV) : 8.4 x 10{sup 14} n /cm{sup 2} s; 2) great flexibility of utilization: the core configuration and operation mode can be adapted to the experimental loading; 3) neutron spectrum tailoring; 4) availability of five 200 mm diameter channels besides the standard channels (84 mm diameter); 5) access to the top and bottom covers of the reactor authorizing the irradiation of loops. The reactor is used to study the behaviour of fuel elements and structural materials intended for future nuclear power stations of several types (fission and fusion). Irradiations are carried out in connection with performance tests up to very high burn-up or neutron fluence as well as for safety experiments, power cycling experiments, and generally speaking, tests under off-normal conditions. Irradiations for nuclear transmutation (production of high specific activity radio-isotopes and transplutonium elements), neutron-radiography, use of beam tubes for physics studies, and gamma irradiations are also carried out. The BR2 is used in support of Belgian programs, at the request of utilities, industry and universities and in the framework of international agreements. The paper reviews the past and ongoing upgrading and enhancement of reactor capabilities as well as those under study or consideration, namely with regard to: reactor equipment, fuel elements, irradiation facilities, reactor operation conditions and long-term strategy. (author)

  1. Model tests and numerical analysis on restoring force characteristics of reactor buildings

    International Nuclear Information System (INIS)

    Uchiyama, Y.; Suzuki, S.; Akino, K.

    1987-01-01

    Seismic shear walls of nuclear reactor buildings are composed of cylindrical, truncated cone-shape, box-shape, irregular polygonal walls or its combination and they are generally heavily reinforced concrete (RC) walls. So the elasto-plastic behaviors of those RC structures in ultimate regions have many unsolved and may be considered as especially important factors for explaining nonlinear response of nuclear reactor buildings. Following these research demands, the authors have prepared a nonlinear F.E.M. code called ''SANREF'' and made an extensive study for the restoring force characteristics of the inner concrete structures (I/C) of a PWR-type containment vessel and the principal seismic shear walls of a BWR-type reactor building by some series of reduced model tests and simulation analysis for the tests results. The detailed objectives of this study can be summarized as follows: (1) Examine the effectiveness of the configurations of shear walls, reinforcement ratios, shear span ratios (M/Qd) and vertical axial stress by ''partial model test'' which simulates some independent shear walls of the PWR-type and BWR-type reactor buildings. (2) Obtain fundamental data of restoring force characteristics of the complex shaped RC structures by ''composite model test'' which models are composed of the partial model test specimens. (3) Verify the applicability of analytical methods and constitutive modelings in SANREF code for complex shaped RC structures through nonlinear simulation analysis for the composite model test

  2. Advanced Test Reactor National Scientific User Facility

    International Nuclear Information System (INIS)

    Marshall, Frances M.; Benson, Jeff; Thelen, Mary Catherine

    2011-01-01

    The Advanced Test Reactor (ATR), at the Idaho National Laboratory (INL), is a large test reactor for providing the capability for studying the effects of intense neutron and gamma radiation on reactor materials and fuels. The ATR is a pressurized, light-water, high flux test reactor with a maximum operating power of 250 MWth. The INL also has several hot cells and other laboratories in which irradiated material can be examined to study material irradiation effects. In 2007 the US Department of Energy (DOE) designated the ATR as a National Scientific User Facility (NSUF) to facilitate greater access to the ATR and the associated INL laboratories for material testing research by a broader user community. This paper highlights the ATR NSUF research program and the associated educational initiatives.

  3. Advanced Test Reactor National Scientific User Facility

    Energy Technology Data Exchange (ETDEWEB)

    Frances M. Marshall; Jeff Benson; Mary Catherine Thelen

    2011-08-01

    The Advanced Test Reactor (ATR), at the Idaho National Laboratory (INL), is a large test reactor for providing the capability for studying the effects of intense neutron and gamma radiation on reactor materials and fuels. The ATR is a pressurized, light-water, high flux test reactor with a maximum operating power of 250 MWth. The INL also has several hot cells and other laboratories in which irradiated material can be examined to study material irradiation effects. In 2007 the US Department of Energy (DOE) designated the ATR as a National Scientific User Facility (NSUF) to facilitate greater access to the ATR and the associated INL laboratories for material testing research by a broader user community. This paper highlights the ATR NSUF research program and the associated educational initiatives.

  4. Development of Research Reactor Simulator and Its Application to Dynamic Test-bed

    International Nuclear Information System (INIS)

    Kwon, Kee Choon; Park, Jae Chang; Lee, Seung Wook; Bang, Dane; Bae, Sung Won

    2014-01-01

    We developed HANARO and the Jordan Research and Training Reactor (JRTR) real-time simulator for operating staff training. The main purpose of this simulator is operator training, but we modified this simulator as a dynamic test-bed to test the reactor regulating system in HANARO or JRTR before installation. The simulator configuration is divided into hardware and software. The simulator hardware consists of a host computer, 6 operator stations, a network switch, and a large display panel. The simulator software is divided into three major parts: a mathematical modeling module, which executes the plant dynamic modeling program in real-time, an instructor station module that manages user instructions, and a human machine interface (HMI) module. The developed research reactors are installed in the Korea Atomic Energy Research Institute nuclear training center for reactor operator training. To use the simulator as a dynamic test-bed, the reactor regulating system modeling software of the simulator was replaced by a hardware controller and the simulator and target controller were interfaced with a hard-wired and network-based interface

  5. Development of Research Reactor Simulator and Its Application to Dynamic Test-bed

    Energy Technology Data Exchange (ETDEWEB)

    Kwon, Kee Choon; Park, Jae Chang; Lee, Seung Wook; Bang, Dane; Bae, Sung Won [KAERI, Daejeon (Korea, Republic of)

    2014-08-15

    We developed HANARO and the Jordan Research and Training Reactor (JRTR) real-time simulator for operating staff training. The main purpose of this simulator is operator training, but we modified this simulator as a dynamic test-bed to test the reactor regulating system in HANARO or JRTR before installation. The simulator configuration is divided into hardware and software. The simulator hardware consists of a host computer, 6 operator stations, a network switch, and a large display panel. The simulator software is divided into three major parts: a mathematical modeling module, which executes the plant dynamic modeling program in real-time, an instructor station module that manages user instructions, and a human machine interface (HMI) module. The developed research reactors are installed in the Korea Atomic Energy Research Institute nuclear training center for reactor operator training. To use the simulator as a dynamic test-bed, the reactor regulating system modeling software of the simulator was replaced by a hardware controller and the simulator and target controller were interfaced with a hard-wired and network-based interface.

  6. Forced vibration tests and simulation analyses of a nuclear reactor building. Part 2: simulation analyses

    International Nuclear Information System (INIS)

    Kuno, M.; Nakagawa, S.; Momma, T.; Naito, Y.; Niwa, M.; Motohashi, S.

    1995-01-01

    Forced vibration tests of a BWR-type reactor building. Hamaoka Unit 4, were performed. Valuable data on the dynamic characteristics of the soil-structure interaction system were obtained through the tests. Simulation analyses of the fundamental dynamic characteristics of the soil-structure system were conducted, using a basic lumped mass soil-structure model (lattice model), and strong correlation with the measured data was obtained. Furthermore, detailed simulation models were employed to investigate the effects of simultaneously induced vertical response and response of the adjacent turbine building on the lateral response of the reactor building. (author). 4 refs., 11 figs

  7. Fabrication, inspection, and test plan for the Advanced Test Reactor (ATR) Mixed-Oxide (MOX) fuel irradiation project

    International Nuclear Information System (INIS)

    Wachs, G.W.

    1997-11-01

    The Department of Energy (DOE) Fissile Materials Disposition Materials Disposition Program (FMDP) has announced that reactor irradiation of MOX fuel is one of the preferred alternatives for disposal of surplus weapons-usable plutonium (Pu). MOX fuel has been utilized domestically in test reactors and on an experimental basis in a number of Commercial Light Water Reactors (CLWRs). Most of this experience has been with Pu derived from spent low enriched uranium (LEU) fuel, known as reactor grade (RG) Pu. The MOX fuel test will be irradiated in the ATR to provide preliminary data to demonstrate that the unique properties of surplus weapons-derived or weapons-grade (WG) plutonium (Pu) do not compromise the applicability of this MOX experience base. In addition, the test will contribute experience with irradiation of gallium-containing fuel to the data base required for resolution of generic CLWR fuel design issues (ORNL/MD/LTR-76). This Fabrication, Inspection, and Test Plan (FITP) is a level 2 document as defined in the FMDP LWR MOX Fuel Irradiation Test Project Plan (ORNL/MD/LTR-78)

  8. Final safety and hazards analysis for the Battelle LOCA simulation tests in the NRU reactor

    International Nuclear Information System (INIS)

    Axford, D.J.; Martin, I.C.; McAuley, S.J.

    1981-04-01

    This is the final safety and hazards report for the proposed Battelle LOCA simulation tests in NRU. A brief description of equipment test design and operating procedure precedes a safety analysis and hazards review of the project. The hazards review addresses potential equipment failures as well as potential for a metal/water reaction and evaluates the consequences. The operation of the tests as proposed does not present an unacceptable risk to the NRU Reactor, CRNL personnel or members of the public. (author)

  9. Scyllac fusion test reactor design

    International Nuclear Information System (INIS)

    Dudziak, D.J.; Gerstl, S.A.; Houck, D.L.; Jalbert, R.A.; Krakowski, R.A.; Linford, R.K.; McDonald, T.E.; Rogers, J.D.; Thomassen, K.I.

    1975-01-01

    A general design of the system is given. The implosion heating and compression systems (METS) are described. Tritium handling, shielding and activation of the reactor, and safety and environmental aspects are discussed

  10. TIBER II/ETR [Engineering Test Reactor] nuclear shielding and optional tritium breeding system: An overview

    International Nuclear Information System (INIS)

    Lee, J.D.; Sawan, M.

    1987-01-01

    TIBER II, the Tokamak Ignition/Burn Experimental Reactor II, is a design concept developed as the US candidate for an International Engineering Test Reactor (ETR). An important objective of this design is to minimize cost by minimizing major radius while providing a wall loading greater than 1.0 MW/m2 and a total fluence greater than 3.0 MWY/m2 needed for blanket module testing. The shielding required for the superconducting TF coils is an important element in setting TIBER II's 3.0m major radius. 6 refs., 1 fig., 1 tab

  11. Operation, test, research and development of the high temperature engineering test reactor (HTTR). FY1999-2001

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2003-05-01

    The HTTR (High Temperature Engineering Test Reactor) with the thermal power of 30 MW and the reactor outlet coolant temperature of 850/950 degC is the first high temperature gas-cooled reactor (HTGR) in Japan, which uses coated fuel particle, graphite for core components, and helium gas for primary coolant. The HTTR, which locates at the south-west area of 50,000 m{sup 2} in the Oarai Research Establishment, had been constructed since 1991 before accomplishing the first criticality on November 10, 1998. Rise to power tests of the HTTR started in September, 1999 and the rated thermal power of 30 MW and the reactor outlet coolant temperature of 850 degC was attained in December 2001. JAERI received the certificate of pre-operation test, that is, the commissioning license for the HTTR in March 2002. This report summarizes operation, tests, maintenance, radiation control, and construction of components and facilities for the HTTR as well as R and Ds on HTGRs from FY1999 to 2001. (author)

  12. Comparison of nuclear irradiation parameters of fusion breeder materials in high flux fission test reactors and a fusion power demonstration reactor

    International Nuclear Information System (INIS)

    Fischer, U.; Herring, S.; Hogenbirk, A.; Leichtle, D.; Nagao, Y.; Pijlgroms, B.J.; Ying, A.

    2000-01-01

    Nuclear irradiation parameters relevant to displacement damage and burn-up of the breeder materials Li 2 O, Li 4 SiO 4 and Li 2 TiO 3 have been evaluated and compared for a fusion power demonstration reactor and the high flux fission test reactor (HFR), Petten, the advanced test reactor (ATR, INEL) and the Japanese material test reactor (JMTR, JAERI). Based on detailed nuclear reactor calculations with the MCNP Monte Carlo code and binary collision approximation (BCA) computer simulations of the displacement damage in the polyatomic lattices with MARLOWE, it has been investigated how well the considered HFRs can meet the requirements for a fusion power reactor relevant irradiation. It is shown that a breeder material irradiation in these fission test reactors is well suited in this regard when the neutron spectrum is well tailored and the 6 Li-enrichment is properly chosen. Requirements for the relevant nuclear irradiation parameters such as the displacement damage accumulation, the lithium burn-up and the damage production function W(T) can be met when taking into account these prerequisites. Irradiation times in the order of 2-3 full power years are necessary for the HFR to achieve the peak values of the considered fusion power Demo reactor blanket with regard to the burn-up and, at the same time, the dpa accumulation

  13. Evaluating and planning the radioactive waste options for dismantling the Tokamak Fusion Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Rule, K.; Scott, J.; Larson, S. [Princeton Plasma Physics Lab., NJ (United States)] [and others

    1995-12-31

    The Tokamak Fusion Test Reactor (TFTR) is a one-of-a kind tritium fusion research reactor, and is planned to be decommissioned within the next several years. This is the largest fusion reactor in the world and as a result of deuterium-tritum reactions is tritium contaminated and activated from 14 Mev neutrons. This presents many unusual challenges when dismantling, packaging and disposing its components and ancillary systems. Special containers are being designed to accommodate the vacuum vessel, neutral beams, and tritium delivery and processing systems. A team of experienced professionals performed a detailed field study to evaluate the requirements and appropriate methods for packaging the radioactive materials. This team focused on several current and innovative methods for waste minimization that provides the oppurtunmost cost effective manner to package and dispose of the waste. This study also produces a functional time-phased schedule which conjoins the waste volume, weight, costs and container requirements with the detailed project activity schedule for the entire project scope. This study and project will be the first demonstration of the decommissioning of a tritium fusion test reactor. The radioactive waste disposal aspects of this project are instrumental in demonstrating the viability of a fusion power reactor with regard to its environmental impact and ultimate success.

  14. On exposure of workers in nuclear reactor facilities for test and in nuclear reactor facilities in research and development stage in fiscal 1988

    International Nuclear Information System (INIS)

    1989-01-01

    The Law for Regulation on Nuclear Reactor requires the operators of nuclear reactors that the exposure dose of workers engaged in work for nuclear reactors should not exceed the limits specified in official notices that are issued based on the Law. The present article summarizes the contents of the Report on Radiation Management in 1988 submitted by the operators of nuclear reactor facilities for test and those of nuclear reactor facilities in research and development stage based on the Law, and the Report on Management of Exposure Dose of Workers submitted by them based on administrative notices. The reports demonstrate that the exposure of workers was below the permissible exposure dose in 1988 in all nuclear reactor facilities. The article presents data on the distribution of exposure dose among workers in all facilities with a nuclear reactor for test, and data on personal exposure of employees and non-employees and overall exposure of all workers in the facilities of Japan Atomic Energy Research Institute and Power Reactor and Nuclear Fuel Development Corporation. (N.K.)

  15. Further Development of Crack Growth Detection Techniques for US Test and Research Reactors

    International Nuclear Information System (INIS)

    Kohse, Gordon; Carpenter, David M.; Ostrovsky, Yakov; Joseph Palmer, A.; Teysseyre, Sebastien P.; Davis, Kurt L.; Rempe, Joy L.

    2015-01-01

    One of the key issues facing Light Water Reactors (LWRs) in extending lifetimes beyond 60 years is characterizing the combined effect of irradiation and water chemistry on material degradation and failure. Irradiation Assisted Stress Corrosion Cracking (IASCC), in which a crack propagates in a susceptible material under stress in an aggressive environment, is a mechanism of particular concern. Full understanding of IASCC depends on real time crack growth data acquired under relevant irradiation conditions. Techniques to measure crack growth in actively loaded samples under irradiation have been developed outside the US - at the Halden Boiling Water Reactor, for example. Several types of IASCC tests have also been deployed at the MITR, including passively loaded crack growth measurements and actively loaded slow strain rate tests. However, there is not currently a facility available in the US to measure crack growth on actively loaded, pre-cracked specimens in LWR irradiation environments. A joint program between the Idaho National Laboratory (INL) and the Massachusetts Institute of Technology (MIT) Nuclear Reactor Laboratory (NRL) is currently underway to develop and demonstrate such a capability for US test and research reactors. Based on the Halden design, the samples will be loaded using miniature high pressure bellows and a compact loading mechanism, with crack length measured in real time using the switched Direct Current Potential Drop (DCPD) method. The basic design and initial mechanical testing of the load system and implementation of the DCPD method have been previously reported. This paper presents the results of initial autoclave testing at INL and the adaptation of the design for use in the high pressure, high temperature water loop at the MITR 6 MW research reactor, where an initial demonstration is planned in mid-2015. Materials considerations for the high pressure bellows are addressed. Design modifications to the loading mechanism required by the

  16. Material test reactor fuel research at the BR2 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Dyck, Steven Van; Koonen, Edgar; Berghe, Sven van den [Institute for Nuclear Materials Science, SCK-CEN, Boeretang, Mol (Belgium)

    2012-03-15

    The construction of new, high performance material test reactor or the conversion of such reactors' core from high enriched uranium (HEU) to low enriched uranium (LEU) based fuel requires several fuel qualification steps. For the conversion of high performance reactors, high density dispersion or monolithic fuel types are being developed. The Uranium-Molybdenum fuel system has been selected as reference system for the qualification of LEU fuels. For reactors with lower performance characteristics, or as medium enriched fuel for high performance reactors, uranium silicide dispersion fuel is applied. However, on the longer term, the U-Mo based fuel types may offer a more efficient fuel alternative and-or an easier back-end solution with respect to the silicide based fuels. At the BR2 reactor of the Belgian nuclear research center, SCK-CEN in Mol, several types of fuel testing opportunities are present to contribute to such qualification process. A generic validation test for a selected fuel system is the irradiation of flat plates with representative dimensions for a fuel element. By flexible positioning and core loading, bounding irradiation conditions for fuel elements can be performed in a standard device in the BR2. For fuel element designs with curved plates, the element fabrication method compatibility of the fuel type can be addressed by incorporating a set of prototype fuel plates in a mixed driver fuel element of the BR2 reactor. These generic types of tests are performed directly in the primary coolant flow conditions of the BR2 reactor. The experiment control and interpretation is supported by detailed neutronic and thermal-hydraulic modeling of the experiments. Finally, the BR2 reactor offers the flexibility for irradiation of full size prototype fuel elements, as 200mm diameter irradiation channels are available. These channels allow the accommodation of various types of prototype fuel elements, eventually using a dedicated cooling loop to provide the

  17. Possibilities for power reactor structural material and fuel testing in reactor RA; Mogucnosti reaktora RA za testiranje konstrukcionih materijala i goriva energetskih reaktora

    Energy Technology Data Exchange (ETDEWEB)

    Martinc, R; Lazarevic, Dj; Stefanovic, D; Cupac, S; Pesic, M [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1978-05-15

    Nuclear reactor RA at Vinca has been designed as a high flux general purpose research reactor. Among other it was intended to play a role of material testing reactor. A scope of activities of Material Laboratory and Reactor RA Department of Boris Kidric Institute is presented in this report. Reactor RA capacity for reactor structural material and fuel irradiation is also described. The increase of RA reactor irradiation capacity is based on the improvement of VISA type fuel channel for fast neutron irradiations, as well as on the general neutron flux increase, due to introduction of highly enriched uranium fuel into reactor core and the advanced in-core fuel management. The irradiation capacities described allow for the reactor material and fuel testing to the considerable extent. Istrazivacki reaktor RA u Vinci je projektovan kao visokofluksni istrazivacki reaktor opste namene. Pored ostalog, on je namenjen i za testiranje reaktorskih konstrukcionih materijala i goriva. U radu je dat pregled aktivnosti Laboratorije za materijale IBK i reaktora RA na tom podrucju, kao i opis povecanih mogucnosti reaktora RA za ozracivanje reaktorskih materijala i goriva u cilju njihovog testiranja. Povecanje mogucnosti reaktora RA zasniva se na usavrsavanju specijalnog gorivnog kanala tipa VISA (za ozracivanje materijala brzim neutronima), kao i na opstem povecanju neutronskog fluksa na osnovu uvodjenja i nacina koriscenja visokoobogacenog uranskog goriva u reaktoru RA. Opisane mogucnosti reaktora RA dozvoljavaju u znatnoj meri ispitivanje konstrukcionih materijala i goriva energetskih reaktora.

  18. The operating experience and incident analysis for High Flux Engineering Test Reactor

    International Nuclear Information System (INIS)

    Zhao Guang

    1999-01-01

    The paper describes the incidents analysis for High Flux Engineering test reactor (HFETR) and introduces operating experience. Some suggestion have been made to reduce the incidents of HFETR. It is necessary to adopt new improvements which enhance the safety and reliability of operation. (author)

  19. Opening address at the international meeting on reduced enrichment for research and test reactors

    International Nuclear Information System (INIS)

    Kazuo Sato

    1984-01-01

    The purpose of the Meeting was to exchange and discuss the most up-to-date information on the progress of variuos programs related to research and test reactor core conversion from high enriched uranium to lower enriched uranium. More detailed status of the RERTR program in Japan, as the host country is covered in this presentation

  20. Mechanical properties of reactor pressure vessel steels studied by static and dynamic torsion tests

    International Nuclear Information System (INIS)

    Munier, A.; Maamouri, M.; Schaller, R.; Mercier, O.

    1993-01-01

    Internal friction measurements and torsional plastic deformation tests have been performed in reactor pressure vessel steels (unirradiated, irradiated and irradiated/annealed specimens). The results of these experiments have been interpreted with help of transmission electron microscopy observations (conventional and in situ). It is shown how the interactions between screw dislocations and obstacles (Peierls valleys, impurities and precipitates) could explain the low temperature hardening and the irradiation embrittlement of ferritic steels. In addition, it appears that the nondestructive internal friction technique could be used advantageously to follow the evolution of the material properties under irradiation, as for instance the irradiation embrittlement of the reactor pressure vessel steels. (orig.)

  1. Technical Bases to Consider for Performance and Demonstration Testing of Space Fission Reactors

    International Nuclear Information System (INIS)

    Hixson, Laurie L.; Houts, Michael G.; Clement, Steven D.

    2004-01-01

    Performance and demonstration testing are critical to the success of a space fission reactor program. However, the type and extent to which testing of space reactors should be performed has been a point of discussion within the industry for many years. With regard to full power ground nuclear tests, questions such as 'Do the benefits outweigh the risks? Are there equivalent alternatives? Can a test facility be constructed (or modified) in a reasonable amount of time? Will the test article accurately represent the flight system? Are the costs too restrictive?' have been debated for decades. There are obvious benefits of full power ground nuclear testing such as obtaining systems integrated reliability data on a full-scale, complete end-to-end system. But these benefits come at some programmatic risk. In addition, this type of testing does not address safety related issues. This paper will discuss and assess these and other technical considerations essential in deciding which type of performance and demonstration testing to conduct on space fission reactor systems. (authors)

  2. Proceedings of the 1988 International Meeting on Reduced Enrichment for Research and Test Reactors

    Energy Technology Data Exchange (ETDEWEB)

    1993-07-01

    The international effort to develop and implement new research reactor fuels utilizing low-enriched uranium, instead of highly- enriched uranium, continues to make solid progress. This effort is the cornerstone of a widely shared policy aimed at reducing, and possibly eliminating, international traffic in highly-enriched uranium and the nuclear weapon proliferation concerns associated with this traffic. To foster direct communication and exchange of ideas among the specialists in this area, the Reduced Enrichment Research and Test Reactor (RERTR) Program, at Argonne National Laboratory, sponsored this meeting as the eleventh of a series which began 1978. Individual papers presented at the meeting have been cataloged separately.

  3. Proceedings of the 1988 International Meeting on Reduced Enrichment for Research and Test Reactors

    International Nuclear Information System (INIS)

    1993-07-01

    The international effort to develop and implement new research reactor fuels utilizing low-enriched uranium, instead of highly- enriched uranium, continues to make solid progress. This effort is the cornerstone of a widely shared policy aimed at reducing, and possibly eliminating, international traffic in highly-enriched uranium and the nuclear weapon proliferation concerns associated with this traffic. To foster direct communication and exchange of ideas among the specialists in this area, the Reduced Enrichment Research and Test Reactor (RERTR) Program, at Argonne National Laboratory, sponsored this meeting as the eleventh of a series which began 1978. Individual papers presented at the meeting have been cataloged separately

  4. A simulated test of physical starting and reactor physics on zero power facility of PWR

    International Nuclear Information System (INIS)

    Yao Zewu; Ji Huaxiang; Chen Zhicheng; Yao Zhiquan; Chen Chen; Li Yuwen

    1995-01-01

    The core neutron economics has been verified through experiments conducted at a zero power reactor with baffles of various thickness. A simulated test of physical starting of Qinshan PWR has been introduced. The feasibility and safety of the programme are verified. The research provides a valuable foundation for developing physical starting programme

  5. Construction and tests of a gamma device for experimental measurements of burnup of nuclear reactor fuel

    International Nuclear Information System (INIS)

    Brandao Junior, F.A.

    1982-01-01

    The gamma-scanning method is an important tool for the measurement of burnup of nuclear reactor fuel. The adequate knowledge of burnup allows for a better inventory of 'sensitive' fissile materials, better fuel management and provides insight on fuel behaviour and safety margins. This paper is related to the description, construction and operation of a first gamma scanning device, tested by irradiation of prototype PWR fuel pins, 14 cm long, in a Triga Mark-I reactor at very low power. Despite the limitations imposed by the low burnup, the experiment permitted a good checking of the main physical concepts and devices involved in the method. (Author) [pt

  6. Development and Testing of a High Capacity Plasma Chemical Reactor in the Ukraine

    Energy Technology Data Exchange (ETDEWEB)

    Reilly, Raymond W.

    2012-07-30

    This project, Development and Testing of a High Capacity Plasma Chemical Reactor in the Ukraine was established at the Kharkiv Institute of Physics and Technology (KIPT). The associated CRADA was established with Campbell Applied Physics (CAP) located in El Dorado Hills, California. This project extends an earlier project involving both CAP and KIPT conducted under a separate CRADA. The initial project developed the basic Plasma Chemical Reactor (PCR) for generation of ozone gas. This project built upon the technology developed in the first project, greatly enhancing the output of the PCR while also improving reliability and system control.

  7. Scram and nonlinear reactor system seismic analysis for the Fast Flux Test Facility

    International Nuclear Information System (INIS)

    Morrone, A.

    1975-01-01

    A description is given of the analysis and results for the Fast Flux Test Facility (FFTF) reactor system which was analyzed for both scram times and seismic responses such as bending moments and impact forces. The reactor system was represented with a one-dimensional nonlinear mathematical model with two degrees of freedom per node. The results give time history plots of various seismic responses and plots of scram times as a function of control rod travel distance for the most critical scram initiation times. The total scram time considering the effects of the earthquake was still acceptable but about 4 times longer than that calculated without the earthquake. (U.S.)

  8. Design and Test of Advanced Thermal Simulators for an Alkali Metal-Cooled Reactor Simulator

    Science.gov (United States)

    Garber, Anne E.; Dickens, Ricky E.

    2011-01-01

    The Early Flight Fission Test Facility (EFF-TF) at NASA Marshall Space Flight Center (MSFC) has as one of its primary missions the development and testing of fission reactor simulators for space applications. A key component in these simulated reactors is the thermal simulator, designed to closely mimic the form and function of a nuclear fuel pin using electric heating. Continuing effort has been made to design simple, robust, inexpensive thermal simulators that closely match the steady-state and transient performance of a nuclear fuel pin. A series of these simulators have been designed, developed, fabricated and tested individually and in a number of simulated reactor systems at the EFF-TF. The purpose of the thermal simulators developed under the Fission Surface Power (FSP) task is to ensure that non-nuclear testing can be performed at sufficiently high fidelity to allow a cost-effective qualification and acceptance strategy to be used. Prototype thermal simulator design is founded on the baseline Fission Surface Power reactor design. Recent efforts have been focused on the design, fabrication and test of a prototype thermal simulator appropriate for use in the Technology Demonstration Unit (TDU). While designing the thermal simulators described in this paper, effort were made to improve the axial power profile matching of the thermal simulators. Simultaneously, a search was conducted for graphite materials with higher resistivities than had been employed in the past. The combination of these two efforts resulted in the creation of thermal simulators with power capacities of 2300-3300 W per unit. Six of these elements were installed in a simulated core and tested in the alkali metal-cooled Fission Surface Power Primary Test Circuit (FSP-PTC) at a variety of liquid metal flow rates and temperatures. This paper documents the design of the thermal simulators, test program, and test results.

  9. Dynamic Response Testing in an Electrically Heated Reactor Test Facility

    Science.gov (United States)

    Bragg-Sitton, Shannon M.; Morton, T. J.

    2006-01-01

    Non-nuclear testing can be a valuable tool in development of a space nuclear power or propulsion system. In a non-nuclear test bed, electric heaters are used to simulate the heat from nuclear fuel. Standard testing allows one to fully assess thermal, heat transfer, and stress related attributes of a given system, but fails to demonstrate the dynamic response that would be present in an integrated, fueled reactor system. The integration of thermal hydraulic hardware tests with simulated neutronic response provides a bridge between electrically heated testing and full nuclear testing. By implementing a neutronic response model to simulate the dynamic response that would be expected in a fueled reactor system, one can better understand system integration issues, characterize integrated system response times and response characteristics, and assess potential design improvements at a relatively small fiscal investment. Initial system dynamic response testing was demonstrated on the integrated SAFE-100a heat pipe cooled, electrically heated reactor and heat exchanger hardware, utilizing a one-group solution to the point kinetics equations to simulate the expected neutronic response of the system (Bragg-Sitton, 2005). The current paper applies the same testing methodology to a direct drive gas cooled reactor system, demonstrating the applicability of the testing methodology to any reactor type and demonstrating the variation in system response characteristics in different reactor concepts. In each testing application, core power transients were controlled by a point kinetics model with reactivity feedback based on core average temperature; the neutron generation time and the temperature feedback coefficient are provided as model inputs. Although both system designs utilize a fast spectrum reactor, the method of cooling the reactor differs significantly, leading to a variable system response that can be demonstrated and assessed in a non-nuclear test facility.

  10. Modeling and simulation of loss of the ultimate heat sink in a typical material testing reactor

    International Nuclear Information System (INIS)

    El-Khatib, Hisham; El-Morshedy, Salah El-Din; Higazy, Maher G.; El-Shazly, Karam

    2013-01-01

    Highlights: ► A thermal–hydraulic model has been developed to simulate loss of the ultimate heat sink in MTR. ► The model involves three coupled sub-models for core, heat exchanger and cooling tower. ► The model is validated against PARET for steady-state and verified by operation data for transients. ► The model is used to simulate the behavior of the reactor under a loss of the ultimate heat sink. ► The model results are analyzed and discussed. -- Abstract: A thermal–hydraulic model has been developed to simulate loss of the ultimate heat sink in a typical material testing reactor (MTR). The model involves three interactively coupled sub-models for reactor core, heat exchanger and cooling tower. The model is validated against PARET code for steady-state operation and verified by the reactor operation records for transients. Then, the model is used to simulate the thermal–hydraulic behavior of the reactor under a loss of the ultimate heat sink event. The simulation is performed for two operation regimes: regime I representing 11 MW power and three cooling tower cells operated, and regime II representing 22 MW power and six cooling tower cells operated. In regime I, the simulation is performed for 1, 2 and 3 cooling tower cells failed while in regime II, it is performed for 1, 2, 3, 4, 5 and 6 cooling tower cells failed. The simulation is performed under protected conditions where the safety action called power reduction is triggered by reactor protection system to decrease the reactor power by 20% when the coolant inlet temperature to the core reaches 43 °C and scram is triggered if the core inlet temperature reaches 44 °C. The model results are analyzed and discussed.

  11. Review of Savannah River Site K Reactor inservice inspection and testing restart program

    International Nuclear Information System (INIS)

    Anderson, M.T.; Hartley, R.S.; Kido, C.

    1992-09-01

    Inservice inspection (ISI) and inservice testing (IST) programs are used at commercial nuclear power plants to monitor the pressure boundary integrity and operability of components in important safety-related systems. The Department of Energy (DOE) - Office of Defense Programs (DP) operates a Category A (> 20 MW thermal) production reactor at the Savannah River Site (SRS). This report represents an evaluation of the ISI and IST practices proposed for restart of SRS K Reactor as compared, where applicable, to current ISI/IST activities of commercial nuclear power facilities

  12. Tokamaks with high-performance resistive magnets: advanced test reactors and prospects for commercial applications

    International Nuclear Information System (INIS)

    Bromberg, L.; Cohn, D.R.; Williams, J.E.C.; Becker, H.; Leclaire, R.; Yang, T.

    1981-10-01

    Scoping studies have been made of tokamak reactors with high performance resistive magnets which maximize advantages gained from high field operation and reduced shielding requirements, and minimize resistive power requirements. High field operation can provide very high values of fusion power density and n tau/sub e/ while the resistive power losses can be kept relatively small. Relatively high values of Q' = Fusion Power/Magnet Resistive Power can be obtained. The use of high field also facilitates operation in the DD-DT advanced fuel mode. The general engineering and operational features of machines with high performance magnets are discussed. Illustrative parameters are given for advanced test reactors and for possible commercial reactors. Commercial applications that are discussed are the production of fissile fuel, electricity generation with and without fissioning blankets and synthetic fuel production

  13. Proceedings of the 1990 International Meeting on Reduced Enrichment for Research and Test Reactors

    Energy Technology Data Exchange (ETDEWEB)

    1993-07-01

    The global effort to reduce, and possibly, eliminate the international traffic in highly-enriched uranium caused by its use in research reactors requires extensive cooperation and free exchange of information among all participants. To foster this free exchange of information, the Reduced Enrichment Research and Test Reactor (RERTR) Program, at Argonne National Laboratory, sponsored this meeting as the thirteenth of a series which began in 1978. The common effort brought together, past, a large number of specialists from many countries. On hundred twenty-three participants from 26 countries, including scientists, reactor operators, and personnel from commercial fuel suppliers, research centers, and government organizations, convened in Newport, Rhode Island to discuss their results, their activities, and their plans relative to converting research reactors to low-enriched fuels. As more and more reactors convert to the use of low-enriched uranium, the emphasis of our effort has begun to shift from research and development to tasks more directly related to implementation of the new fuels and technologies that have been developed, and to refinements of those fuels and technologies. It is appropriate, for this reason, that the emphasis of this meeting was placed on safety and on conversion experiences. This individual papers in this report have been cataloged separately.

  14. Proceedings of the 1990 International Meeting on Reduced Enrichment for Research and Test Reactors

    International Nuclear Information System (INIS)

    1993-07-01

    The global effort to reduce, and possibly, eliminate the international traffic in highly-enriched uranium caused by its use in research reactors requires extensive cooperation and free exchange of information among all participants. To foster this free exchange of information, the Reduced Enrichment Research and Test Reactor (RERTR) Program, at Argonne National Laboratory, sponsored this meeting as the thirteenth of a series which began in 1978. The common effort brought together, past, a large number of specialists from many countries. On hundred twenty-three participants from 26 countries, including scientists, reactor operators, and personnel from commercial fuel suppliers, research centers, and government organizations, convened in Newport, Rhode Island to discuss their results, their activities, and their plans relative to converting research reactors to low-enriched fuels. As more and more reactors convert to the use of low-enriched uranium, the emphasis of our effort has begun to shift from research and development to tasks more directly related to implementation of the new fuels and technologies that have been developed, and to refinements of those fuels and technologies. It is appropriate, for this reason, that the emphasis of this meeting was placed on safety and on conversion experiences. This individual papers in this report have been cataloged separately

  15. Standard practice for analysis and interpretation of physics dosimetry results for test reactors

    International Nuclear Information System (INIS)

    Anon.

    1984-01-01

    This practice describes the methodology summarized in Annex Al to be used in the analysis and interpretation of physics-dosimetry results from test reactors. This practice relies on, and ties together, the application of several supporting ASTM standard practices, guides, and methods that are in various stages of completion (see Fig. 1). Support subject areas that are discussed include reactor physics calculations, dosimeter selection and analysis, exposure units, and neutron spectrum adjustment methods. This practice is directed towards the development and application of physics-dosimetrymetallurgical data obtained from test reactor irradiation experiments that are performed in support of the operation, licensing, and regulation of LWR nuclear power plants. It specifically addresses the physics-dosimetry aspects of the problem. Procedures related to the analysis, interpretation, and application of both test and power reactor physics-dosimetry-metallurgy results are addressed in Practice E 853, Practice E 560, Matrix E 706(IE), Practice E 185, Matrix E 706(IG), Guide E 900, and Method E 646

  16. French safety authority projects in the field of research and test reactors

    International Nuclear Information System (INIS)

    Saint Raymond, P.; Duthe, M.; Abou Yehia, H.

    2001-01-01

    This paper gives an outline of some actions initiated by the French safety authority in the field of research and test reactors. An important action concerns the definition of the authorisation criteria for the implementation of experiments in these reactors. In particular, it is necessary to define clearly in which conditions an experiment may be authorised internally by the operating organisation or needs a formal approval by the safety authority. The practice related to the systematic safety reassessment of old facilities and the regulatory provisions associated with the decommissioning are presented after a discussion on the ageing issues. (author)

  17. Monitoring and Analysis of In-Pile Phenomena in Advanced Test Reactor using Acoustic Telemetry

    International Nuclear Information System (INIS)

    Agarwal, Vivek; Smith, James A.; Jewell, James Keith

    2015-01-01

    The interior of a nuclear reactor presents a particularly harsh and challenging environment for both sensors and telemetry due to high temperatures and high fluxes of energetic and ionizing particles among the radioactive decay products. A number of research programs are developing acoustic-based sensing approach to take advantage of the acoustic transmission properties of reactor cores. Idaho National Laboratory has installed vibroacoustic receivers on and around the Advanced Test Reactor (ATR) containment vessel to take advantage of acoustically telemetered sensors such as thermoacoustic (TAC) transducers. The installation represents the first step in developing an acoustic telemetry infrastructure. This paper presents the theory of TAC, application of installed vibroacoustic receivers in monitoring the in-pile phenomena inside the ATR, and preliminary data processing results.

  18. Monitoring and Analysis of In-Pile Phenomena in Advanced Test Reactor using Acoustic Telemetry

    Energy Technology Data Exchange (ETDEWEB)

    Agarwal, Vivek [Idaho National Lab. (INL), Idaho Falls, ID (United States). Dept. of Human Factors, Controls, and Statistics; Smith, James A. [Idaho National Lab. (INL), Idaho Falls, ID (United States). Dept. of Fuel Performance and Design; Jewell, James Keith [Idaho National Lab. (INL), Idaho Falls, ID (United States). Dept. of Fuel Performance and Design

    2015-02-01

    The interior of a nuclear reactor presents a particularly harsh and challenging environment for both sensors and telemetry due to high temperatures and high fluxes of energetic and ionizing particles among the radioactive decay products. A number of research programs are developing acoustic-based sensing approach to take advantage of the acoustic transmission properties of reactor cores. Idaho National Laboratory has installed vibroacoustic receivers on and around the Advanced Test Reactor (ATR) containment vessel to take advantage of acoustically telemetered sensors such as thermoacoustic (TAC) transducers. The installation represents the first step in developing an acoustic telemetry infrastructure. This paper presents the theory of TAC, application of installed vibroacoustic receivers in monitoring the in-pile phenomena inside the ATR, and preliminary data processing results.

  19. Vibration test on KMRR reactor structure and primary cooling system piping

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Seung Hoh; Kim, Tae Ryong; Park, Jin Hoh; Park, Jin Suk; Ryoo, Jung Soo [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1994-10-01

    Most equipments, piping systems and reactor structures in nuclear power plants are subjected to flow induced vibration due to high temperature and high pressure coolant flowing inside or outside of the equipments, systems and structures. Because the flow induced vibration sometimes causes significant damage to reactor structures and piping systems, it is important and necessary to evaluate the vibration effect on them and to prove their structural integrity. Korea Multipurpose Research Reactor (KMRR) being constructed by KAERI is 30 MWt pool type research reactor. Since its main structures and piping systems were designed and manufactured in accordance with the standards and guidelines for commercial nuclear power plant, it was decided to evaluate their vibratory response in accordance with the standards and guidelines for commercial NPP. The objective of this vibration test is the assessment of vibration levels of KMRR reactor structure and primary cooling piping system for their structural integrity under the steady-state or transient operating condition. 38 figs, 14 tabs, 2 refs. (Author).

  20. Vibration test on KMRR reactor structure and primary cooling system piping

    International Nuclear Information System (INIS)

    Chung, Seung Hoh; Kim, Tae Ryong; Park, Jin Hoh; Park, Jin Suk; Ryoo, Jung Soo

    1994-10-01

    Most equipments, piping systems and reactor structures in nuclear power plants are subjected to flow induced vibration due to high temperature and high pressure coolant flowing inside or outside of the equipments, systems and structures. Because the flow induced vibration sometimes causes significant damage to reactor structures and piping systems, it is important and necessary to evaluate the vibration effect on them and to prove their structural integrity. Korea Multipurpose Research Reactor (KMRR) being constructed by KAERI is 30 MWt pool type research reactor. Since its main structures and piping systems were designed and manufactured in accordance with the standards and guidelines for commercial nuclear power plant, it was decided to evaluate their vibratory response in accordance with the standards and guidelines for commercial NPP. The objective of this vibration test is the assessment of vibration levels of KMRR reactor structure and primary cooling piping system for their structural integrity under the steady-state or transient operating condition. 38 figs, 14 tabs, 2 refs. (Author)

  1. U.S. uranium supply to the research and test reactor community

    International Nuclear Information System (INIS)

    Parker, Elaine M.

    2002-01-01

    From the 1950s through the early 1990s, the U.S. Department of Energy (DOE) was the primary supplier of low enriched uranium (LEU) and highly enriched uranium (HEU) to research and test reactors worldwide. The formerly called Y-12 Plant in Oak Ridge, Tennessee, was put into operational stand down in 1994 due to inadequate safety documentation. This paper will discuss the re-start of the Y-12 Plant and its current capabilities. Additionally, the paper will address recent changes within the DOE, with the creation of the National Nuclear Security Administration (NNSA). It will show how the change to NNSA and an organizational re-alignment has improved efficiencies. NNSA is committed to operate its sales program so that it is complementary to, and in support of, the Reduced Enrichment for Research and Test Reactors (RERTR) and Foreign Research Reactor Spent Nuclear Fuel (FRR SNF) Return Programs. The NNSA is committed to provide an assurance of competitively-priced, high-quality uranium supply to the research and test reactor community under long-term contracts. This paper will discuss some of NNSA's recent successes in long-term contracting and meeting deliveries. (author)

  2. Technology, safety, and costs of decommissioning reference nuclear research and test reactors. Main report

    International Nuclear Information System (INIS)

    Konzek, G.J.; Ludwick, J.D.; Kennedy, W.E. Jr.; Smith, R.I.

    1982-03-01

    Safety and Cost Information is developed for the conceptual decommissioning of two representative licensed nuclear research and test reactors. Three decommissioning alternatives are studied to obtain comparisons between costs (in 1981 dollars), occupational radiation doses, potential radiation dose to the public, and other safety impacts. The alternatives considered are: DECON (immediate decontamination), SAFSTOR (safe storage followed by deferred decontamination), and ENTOMB (entombment). The study results are presented in two volumes. Volume 1 (Main Report) contains the results in summary form

  3. The advanced test reactor strategic evaluation program

    International Nuclear Information System (INIS)

    Buescher, B.J.

    1989-01-01

    Since the Chernobly accident, the safety of test reactors and irradiation facilities has been critically evaluated from the public's point of view. A systematic evaluation of all safety, environmental, and operational issues must be made in an integrated manner to prioritize actions to maximize benefits while minimizing costs. Such a proactive program has been initiated at the Advanced Test Reactor (ATR). This program, called the Strategic Evaluation Program (STEP), is being conducted for the ATR to provide integrated safety and operational reviews of the reactor against the standards applied to licensed commercial power reactors. This has taken into consideration the lessons learned by the US Nuclear Regulatory Commission (NRC) in its Systematic Evaluation Program (SEP) and the follow-on effort known as the Integrated Safety Assessment Program (ISAP). The SEP was initiated by the NRC to review the designs of older operating nuclear power plants to confirm and document their safety. The ATR STEP objectives are discussed

  4. Thermal simulations and tests in the development of a helmet transport spent fuel elements Research Reactor

    International Nuclear Information System (INIS)

    Saliba, R.; Quintana, F.; Márquez Turiello, R.; Furnari, J.C.; Pimenta Mourão, R.

    2013-01-01

    A packaging for the transport of irradiated fuel from research reactors was designed by a group of researchers to improve the capability in the management of spent fuel elements from the reactors operated in the region. Two half-scale models for MTR fuel were constructed and tested so far and a third one for both MTR and TRIGA fuels will be constructed and tested next. Four test campaigns have been carried out, covering both normal and hypothetical accident conditions of transportation. The thermal test is part of the requirements for the qualification of transportation packages for nuclear reactors spent fuel elements. In this paper both the numerical modelling and experimental thermal tests performed are presented and discussed. The cask is briefly described as well as the finite element model developed and the main adopted hypotheses for the thermal phenomena. The results of both numerical runs and experimental tests are discussed as a tool to validate the thermal modelling. The impact limiters, attached to the cask for protection, were not modelled. (author) [es

  5. Automated systems help prevent operator error during [reactor] I and C [instrumentation and control] testing

    International Nuclear Information System (INIS)

    Courcoux, R.

    1989-01-01

    On a nuclear steam supply system, even a minor failure can involve actuation of the whole reactor protection system (RPS). To reduce the likelihood of human error leading to unwanted trips during the maintenance of instrumentation and control systems, Framatome has been developing and installing various automated testing systems. Such automated systems are particularly helpful when periodic tests with a potential for RPS actuation have to be carried out, or when the test is on the critical path for the refuelling outage. The Sensitive Channel Programme described is an example of the sort of work that has been done. (author)

  6. On exposure management of workers in nuclear reactor facilities for test and in nuclear reactor facilities in research and development stage in fiscal 1993

    International Nuclear Information System (INIS)

    1994-01-01

    The Law of Regulation on Nuclear Reactor requires the operators of nuclear reactors that the exposure dose of workers engaged in work for nuclear reactors should not exceed the limits specified in official notices that are issued based on the Law. The present article summarizes the contents of the Report on Radiation Management in 1993 submitted by the operators of nuclear reactor facilities for test and those of nuclear reactor facilities in research and development stage based on the Law, and the Report on Management of Exposure Dose of Workers submitted by them based on administrative notices. The reports demonstrate that the the exposure of workers was below the permissible exposure dose in 1993 in all nuclear reactor facilities. The article presents data on the distribution of exposure dose among workers in all facilities with a nuclear reactor for test, and data on personal exposure of employees and non-employees and overall exposure of all workers in the facilities of JAERI and PNC. (J.P.N.)

  7. Concerning control of radiation exposure to workers in nuclear reactor facilities for testing and nuclear reactor facilities in research and development phase (fiscal 1987)

    International Nuclear Information System (INIS)

    1988-01-01

    A nuclear reactor operator is required by the Nuclear Reactor Control Law to ensure that the radiation dose to workers engaged in the operations of his nuclear reactor is controlled below the permissible exposure doses that are specified in notifications issued based on the Law. The present note briefly summarizes the data given in the Reports on Radiation Control, which have been submitted according to the Nuclear Reactor Control Law by the operators of nuclear reactor facilities for testing and those in the research and development phase, and the Reports on Control of Radiation Exposure to Workers submitted in accordance with the applicable administrative notices. According to these reports, the measured exposure to workers in 1987 were below the above-mentioned permissible exposure doses in all these nuclear facilities. The 1986 and 1987 measurements of radiation exposure dose to workers in nuclear reactor facilities for testing are tabulated. The measurements cover dose distribution among the facilities' personnel and workers of contractors. They also cover the total exposure dose for all workers in each of four plants operated under the Japan Atomic Energy Research Institute and the Power Reactor and Nuclear Fuel Development Corporation. (N.K.)

  8. A premature demise for RERTR [Reduced Enrichment for Research and Test Reactors programme]?

    International Nuclear Information System (INIS)

    Rydell, R.J.

    1990-01-01

    A common commitment from France, Belgium, Germany and the US to eliminate highly enriched uranium from their research reactors is needed to help guard against this material falling into the wrong hands. In the US, an essential part of this commitment would be rekindling the weakened Reduced Enrichment for Research and Test Reactors programme (RERTR). This is an American initiative to develop low-enrichment uranium fuel for research reactors that have previously required weapons-usable material. Underway since 1978 at Argonne National Laboratory, RERTR has achieved some impressive results: the development of higher density, low enriched fuels that are suitable for use at over 90% of the world's research reactors; a net reduction of US exports of highly enriched uranium (HEU) from the annual 700kg levels in the late 1970s to a 1990 level of just over 100kg; the encouragement of international scientific co-operation aimed at developing new fuels and facilitating the conversion of existing reactors to these fuels. However, in recent years, the US commitment to RERTR has been declining -budgets have fallen and advanced fuel development work has terminated. (author)

  9. Startup testing of Romania dual-core test reactor

    International Nuclear Information System (INIS)

    Whittemore, W.L.

    1980-01-01

    Late in 1979 both the Annular Core Pulsed Reactor (ACPR) and the 14-MW steady-state reactor (SSR) were loaded to critical. The fuel loading in both was then carried to completion and low-power testing was conducted. Early in 1980 both reactors successfully underwent high-power testing. The ACPR was operated for several hours at 500 kW and underwent pulse tests culminating in pulses with reactivity insertions of $4.60, peak power levels of about 20,000 MW, energy releases of 100 MW-sec, and peak measured fuel temperatures of 830 deg. C. The SSR was operated in several modes, both with natural convection and forced cooling with one or more pumps. The reactor successfully completed a 120-hr full-power test. Subsequent fuel element inspections confirmed that the fuel has performed without fuel damage or distortion. (author)

  10. Reliability tests for reactor internals replacement technology

    International Nuclear Information System (INIS)

    Fujimaki, K.; Uchiyama, J.; Ohtsubo, T.

    2000-01-01

    Structural damage due to aging degradation of LWR reactor internals has been reported in several nuclear plants. NUPEC has started a project to test the reliability of the technology for replacing reactor internals, which was directed at preventive maintenance before damage and repair after damage for the aging degradation. The project has been funded by the Ministry of International Trade and Industry (MITI) of Japan since 1995, and it follows the policy of a report that the MITI has formally issued in April 1996 summarizing the countermeasures to be considered for aging nuclear plants and equipment. This paper gives an outline of the whole test plans and the test results for the BWR reactor internals replacement methods; core shroud, ICM housing, and CRD Housing and stub tube. The test results have shown that the methods were reliable and the structural integrity was appropriate based on the evaluation. (author)

  11. Design, Testing and Modeling of the Direct Reactor Auxiliary Cooling System for AHTRs

    Energy Technology Data Exchange (ETDEWEB)

    Lv, Quiping [The Ohio State Univ., Columbus, OH (United States); Sun, Xiaodong [The Ohio State Univ., Columbus, OH (United States); Chtistensen, Richard [The Ohio State Univ., Columbus, OH (United States); Blue, Thomas [The Ohio State Univ., Columbus, OH (United States); Yoder, Graydon [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Wilson, Dane [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-05-08

    The principal objective of this research is to test and model the heat transfer performance and reliability of the Direct Reactor Auxiliary Cooling System (DRACS) for AHTRs. In addition, component testing of fluidic diodes is to be performed to examine the performance and viability of several existing fluidic diode designs. An extensive database related to the thermal performance of the heat exchangers involved will be obtained, which will be used to benchmark a computer code for the DRACS design and to evaluate and improve, if needed, existing heat transfer models of interest. The database will also be valuable for assessing the viability of the DRACS concept and benchmarking any related computer codes in the future. The experience of making a liquid fluoride salt test facility available, with lessons learned, will greatly benefit the development of the Fluoride Salt-cooled High-temperature Reactor (FHR) and eventually the AHTR programs.

  12. Design, Testing and Modeling of the Direct Reactor Auxiliary Cooling System for AHTRs

    International Nuclear Information System (INIS)

    Lv, Quiping; Sun, Xiaodong; Chtistensen, Richard; Blue, Thomas; Yoder, Graydon; Wilson, Dane

    2015-01-01

    The principal objective of this research is to test and model the heat transfer performance and reliability of the Direct Reactor Auxiliary Cooling System (DRACS) for AHTRs. In addition, component testing of fluidic diodes is to be performed to examine the performance and viability of several existing fluidic diode designs. An extensive database related to the thermal performance of the heat exchangers involved will be obtained, which will be used to benchmark a computer code for the DRACS design and to evaluate and improve, if needed, existing heat transfer models of interest. The database will also be valuable for assessing the viability of the DRACS concept and benchmarking any related computer codes in the future. The experience of making a liquid fluoride salt test facility available, with lessons learned, will greatly benefit the development of the Fluoride Salt-cooled High-temperature Reactor (FHR) and eventually the AHTR programs.

  13. Fabrication and testing of uranium nitride fuel for space power reactors

    Science.gov (United States)

    Matthews, R. B.; Chidester, K. M.; Hoth, C. W.; Mason, R. E.; Petty, R. L.

    1988-02-01

    Uranium nitride fuel was selected for previous space power reactors because of its attractive thermal and physical properties; however, all UN fabrication and testing activities were terminated over ten years ago. An accelerated irradiation test, SP-1, was designed to demonstrate the irradiation performance of Nb-1 Zr clad UN fuel pins for the SP-100 program. A carbothermic-reduction/nitriding process was developed to synthesize UN powders. These powders were fabricated into fuel pellets by conventional cold-pressing and sintering. The pellets were loaded into Nb-1 Zr cladding tubes, irradiated in a fast-test reactor, and destructively examined after 0.8 at% burnup. Preliminary postirradiation examination (PIE) results show that the fuel pins behaved as designed. Fuel swelling, fission-gas release, and microstructural data are presented, and suggestions to enhance the reliability of UN fuel pins are discussed.

  14. The US Advanced Liquid Metal Reactor and the Fast Flux Test Facility Phase IIA passive safety tests

    International Nuclear Information System (INIS)

    Shen, P.K.; Harris, R.A.; Campbell, L.R.; Dautel, W.A.; Dubberley, A.E.; Gluekler, E.L.

    1992-07-01

    This report discusses the safety approach of the Advanced Liquid Metal reactor program, sponsored by the US Department of Energy, which relies upon passive reactor responses to off-normal condition to limit power and temperature excursions to levels that allow safety margins. Gas expansion modules (GEM) have included in the design to provide negative reactivity to enhance these margins in the extremely unlikely event that pumping power is lost and the highly reliable scram system fails to operate. The feasibility and beneficial features of these devices were first demonstrated in the core of the Fast Flux Test Facility (FFTF) in 1986. Preapplication safety evaluations by the US Nuclear Regulatory Commission have identified areas that must be addressed if these devices are to be relied on. One of these areas is the response of the reactor when it is critical and the pumps are turned on, resulting in positive reactivity being added to the core. Tests to examine such transients have been performed as part of the continuing FFTF program to confirm the passive safety characteristics of liquid metal reactors (LMR). The primary tests consisted of starting the main coolant pumps, which forced sodium coolant into the GEMS, decreasing neutron leakage and adding positive reactivity. The resulting transients were shown to be benign and easily mitigated by the reactivity feedbacks inherent in the FFTF and all LMRs. Steady-state auxiliary tests of the GEM and feedback reactivity worths accurately predicted the transient results. The auxiliary GEM worth tests also demonstrated that the worth can be determined at a subcritical state, which allows for a verification of the GEM's availability prior to ascending to power

  15. Plutonium recycle test reactor characterization activities and results

    International Nuclear Information System (INIS)

    Cornwell, B.C.

    1997-01-01

    Report contains results of PRTR core and associated structures characterization performed in January and February of 1997. Radiation survey data are presented, along with recommendations for stabilization activities before transitioning to a decontamination and decommissioning function. Recommendations are also made about handling the waste generated by the stabilization activities, and actions suggested by the Decontamination and Decommissioning organization

  16. Reduced enrichment for research and test reactors: proceedings

    International Nuclear Information System (INIS)

    1985-07-01

    Separate abstracts are presented for each of the papers included in the data base concerning RERTR programs and licensing; fuel development; plate-type fuel fabrication; fuel demonstration; economics; mixed cores; and applications

  17. Advanced Instrumentation for Transient Reactor Testing

    Energy Technology Data Exchange (ETDEWEB)

    Corradini, Michael L.; Anderson, Mark; Imel, George; Blue, Tom; Roberts, Jeremy; Davis, Kurt

    2018-01-31

    Transient testing involves placing fuel or material into the core of specialized materials test reactors that are capable of simulating a range of design basis accidents, including reactivity insertion accidents, that require the reactor produce short bursts of intense highpower neutron flux and gamma radiation. Testing fuel behavior in a prototypic neutron environment under high-power, accident-simulation conditions is a key step in licensing nuclear fuels for use in existing and future nuclear power plants. Transient testing of nuclear fuels is needed to develop and prove the safety basis for advanced reactors and fuels. In addition, modern fuel development and design increasingly relies on modeling and simulation efforts that must be informed and validated using specially designed material performance separate effects studies. These studies will require experimental facilities that are able to support variable scale, highly instrumented tests providing data that have appropriate spatial and temporal resolution. Finally, there are efforts now underway to develop advanced light water reactor (LWR) fuels with enhanced performance and accident tolerance. These advanced reactor designs will also require new fuel types. These new fuels need to be tested in a controlled environment in order to learn how they respond to accident conditions. For these applications, transient reactor testing is needed to help design fuels with improved performance. In order to maximize the value of transient testing, there is a need for in-situ transient realtime imaging technology (e.g., the neutron detection and imaging system like the hodoscope) to see fuel motion during rapid transient excursions with a higher degree of spatial and temporal resolution and accuracy. There also exists a need for new small, compact local sensors and instrumentation that are capable of collecting data during transients (e.g., local displacements, temperatures, thermal conductivity, neutron flux, etc.).

  18. Procedures and techniques for the management of experimental fuels from research and test reactors. Proceedings of an advisory group meeting

    International Nuclear Information System (INIS)

    1999-04-01

    Almost all countries that have undertaken fuel development programs for power, research or military reactors have experimental and exotic fuels, either stored at the original research reactors where they have been tested or at some away-from-reactor storage facility. These spent fuel liabilities cannot follow the standard treatment recognized for modern power reactor fuels. They include experimental and exotic fuels ranging from liquids to coated spheres and in configurations ranging from full test assemblies to post irradiation examination specimens set in resin. This document contains an overview of the extent of the problem of managing experimental and exotic fuels from research and test reactors and an expert evaluation of the overall situation in countries which participated in the meeting

  19. Heat Transfer Behaviour and Thermohydraulics Code Testing for Supercritical Water Cooled Reactors (SCWRs)

    International Nuclear Information System (INIS)

    2014-08-01

    The supercritical water cooled reactor (SCWR) is an innovative water cooled reactor concept which uses water pressurized above its thermodynamic critical pressure as the reactor coolant. This concept offers high thermal efficiencies and a simplified reactor system, and is hence expected to help to improve economic competitiveness. Various kinds of SCWR concepts have been developed, with varying combinations of reactor type (pressure vessel or pressure tube) and core spectrum (thermal, fast or mixed). There is great interest in both developing and developed countries in the research and development (R&D) and conceptual design of SCWRs. Considering the high interest shown in a number of Member States, the IAEA established in 2008 the Coordinated Research Project (CRP) on Heat Transfer Behaviour and Thermo-hydraulics Code Testing for SCWRs. The aim was to foster international collaboration in the R&D of SCWRs in support of Member States’ efforts and under the auspices of the IAEA Nuclear Energy Department’s Technical Working Groups on Advanced Technologies for Light Water Reactors (TWG-LWR) and Heavy Water Reactors (TWG-HWR). The two key objectives of the CRP were to establish accurate databases on the thermohydraulics of supercritical pressure fluids and to test analysis methods for SCWR thermohydraulic behaviour to identify code development needs. In total, 16 institutes from nine Member States and two international organizations were involved in the CRP. The thermohydraulics phenomena investigated in the CRP included heat transfer and pressure loss characteristics of supercritical pressure fluids, development of new heat transfer prediction methods, critical flow during depressurization from supercritical conditions, flow stability and natural circulation in supercritical pressure systems. Two code testing benchmark exercises were performed for steady state heat transfer and flow stability in a heated channel. The CRP was completed with the planned outputs in

  20. Heat Transfer Behaviour and Thermohydraulics Code Testing for Supercritical Water Cooled Reactors (SCWRs)

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2014-08-15

    The supercritical water cooled reactor (SCWR) is an innovative water cooled reactor concept which uses water pressurized above its thermodynamic critical pressure as the reactor coolant. This concept offers high thermal efficiencies and a simplified reactor system, and is hence expected to help to improve economic competitiveness. Various kinds of SCWR concepts have been developed, with varying combinations of reactor type (pressure vessel or pressure tube) and core spectrum (thermal, fast or mixed). There is great interest in both developing and developed countries in the research and development (R&D) and conceptual design of SCWRs. Considering the high interest shown in a number of Member States, the IAEA established in 2008 the Coordinated Research Project (CRP) on Heat Transfer Behaviour and Thermo-hydraulics Code Testing for SCWRs. The aim was to foster international collaboration in the R&D of SCWRs in support of Member States’ efforts and under the auspices of the IAEA Nuclear Energy Department’s Technical Working Groups on Advanced Technologies for Light Water Reactors (TWG-LWR) and Heavy Water Reactors (TWG-HWR). The two key objectives of the CRP were to establish accurate databases on the thermohydraulics of supercritical pressure fluids and to test analysis methods for SCWR thermohydraulic behaviour to identify code development needs. In total, 16 institutes from nine Member States and two international organizations were involved in the CRP. The thermohydraulics phenomena investigated in the CRP included heat transfer and pressure loss characteristics of supercritical pressure fluids, development of new heat transfer prediction methods, critical flow during depressurization from supercritical conditions, flow stability and natural circulation in supercritical pressure systems. Two code testing benchmark exercises were performed for steady state heat transfer and flow stability in a heated channel. The CRP was completed with the planned outputs in

  1. Storage and management of fuel from fast breeder test reactor and KAlpakkam MINI

    International Nuclear Information System (INIS)

    Sodhi, B.S.; Rao, M.S.; Natarajan, R.

    1999-01-01

    Two Research Reactors, FBTR (Fast Breeder Test Reactor) and KAMINI (KAlpakkam MINI) are in operation at Kalpakkam, India. FBTR is a 40 MWt reactor. It is the first reactor to use mixed carbide (70% PuC-30% UC) as driver fuel. Special precautions are needed to fabricate pellets in glove boxes under inert atmosphere to take into account the possibility of criticality, radiation, pyrophoricity and toxicity of PuC. FBTR has been operating with small core up to 12 MWt power. The initial limit was 250 W/cm, linear heat rating and 25,000 MWd/t peak burnup. This limit was increased to 320 W/cm and 50,000 MWd/t respectively after rigorous analysis. At present the core has reached 40,000 MWd/t without any pin failure. After 25,000 MWd/t burnup one fuel subassembly (SA) was removed and PEE was carried out. The results were as expected by the analysis. In FBTR, fuel is stored in a container filled with argon and the container is cooled by forced circulation of air (during storage). Closing the fuel cycle is important for the breeder programme. Therefore, efforts have been made to set up a reprocessing plant. It uses the well proven purex process. The irradiated fuel is sheared in a single pin chopper and dissolved in an electrochemical dissolver. The resulting solution after adjusting the valency of Pu to IVth state is processed in the solvent extraction plant using 30% Tri-n-Butyl phosphate/n-dodecane as solvent. KAMINI is 30 kWt neutron source reactor which uses light water as moderator and coolant and has as a fuel U-233 aluminium alloy. Uranium-233 has been indigenously recovered from thorium irradiated in CIRUS reactor at Trombay. KAMINI was made critical on October 1996. It is housed in a vault below one of the hot cells in the Radiometallurgy laboratories of IGCAR. This reactor is planned to be used for neutron radiography of fuel elements and neutron activation analysis. It is available for use by research institutions and universities also. This paper describes the

  2. Operating the Advanced Test Reactor in today's economic and regulatory environment

    International Nuclear Information System (INIS)

    Furstenau, R.V.; Patrick, M.E.; Mecham, D.C.

    1999-01-01

    The Advanced Test Reactor (ATR), located at the Idaho National Engineering and Environmental Laboratory, is the US Department of Energy's largest and most versatile test reactor. Base programs at ATR are planned well into the 21st century. The ATR and support facilities along with an overview of current programs will be reviewed, but the main focus of the presentation will be on the impact that today's economic and regulatory concerns have had on the operation of this test reactor. Today's economic and regulatory concerns have demanded more work be completed at lower cost while increasing the margin of safety. By the beginning of the 1990 s, federal budgets for research generally and particularly for nuclear research had decreased dramatically. Many national needs continued to require testing in the ATR; but demanded lower cost, increased efficiency, improved performance, and an increased margin of safety. At the same time budgets were decreasing, there was an increase in regulatory compliance activity. The new standards imposed higher margins of safety. The new era of greater openness and higher safety standards complemented research demands to work safer, smarter and more efficiently. Several changes were made at the ATR to meet the demands of the sponsors and public. Such changes included some workforce reductions, securing additional program sponsors, upgrading some facilities, dismantling other facilities, and implementing new safety programs. (author)

  3. Reactor physics and reactor computations

    International Nuclear Information System (INIS)

    Ronen, Y.; Elias, E.

    1994-01-01

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

  4. Present status and prospects of high-temperature engineering test reactor (HTTR) program

    International Nuclear Information System (INIS)

    Tanaka, Toshiyuki; Baba, Osamu; Shiozawa, Shusaku; Okubo, Minoru; Tobioka, Toshiaki

    1995-01-01

    It is essentially important in Japan, which has limited amount of natural resources, to make efforts to obtain more reliable and stable energy supply by extended use of nuclear energy including high temperature heat from nuclear reactors. Hence, efforts are to be continuously devoted to establish and upgrade High Temperature Gas-cooled Reactor (HTGR) technologies and to make much of research resources accumulated so far. It is also expected that making basic researches at high temperature using HTGR will contribute to innovative basic research in future. Then, the construction of High Temperature engineering Test Reactor (HTTR), which is an HTGR with a maximum helium coolant temperature of 950degC at the reactor outlet, was decided by the Japanese Atomic Energy Commission (JAEC) in 1987 and is now under way by the Japan Atomic Energy Research Institute (JAERI). The construction of the HTTR started in March 1991, with first criticality in 1998 to be followed after commissioning testing. At present the HTTR reactor building and its containment vessel have been nearly completed and its main components, such as a reactor pressure vessel, an intermediate heat exchanger, hot gas pipings and core support structures, have been manufactured at their factories and delivered to the Oarai Research Establishment of the JAERI for their installation in the middle of 1994. Fuel fabrication will be started as well. The project is intended to establish and upgrade the technology basis necessary for HTGR developments. The IAEA Coordinated Research Programme on Design and Evaluation of Heat Utilization Systems for the HTTR, such as steam reforming of methane and thermochemical water splitting for hydrogen production, was launched successfully in January 1994. Some heat utilization system is planned to be connected to the HTTR and demonstrated at the former stage of the second core. At present, steam-reforming of methane is the first candidate. The JAERI also plans to conduct material

  5. Shock and vibration tests of uranium mononitride fuel pellets for a space power nuclear reactor

    Science.gov (United States)

    Adams, D. W.

    1972-01-01

    Shock and vibration tests were conducted on cylindrically shaped, depleted, uranium mononitride (UN) fuel pellets. The structural capabilities of the pellets were determined under exposure to shock and vibration loading which a nuclear reactor may encounter during launching into space. Various combinations of diametral and axial clearances between the pellets and their enclosing structures were tested. The results of these tests indicate that for present fabrication of UN pellets, a diametral clearance of 0.254 millimeter and an axial clearance of 0.025 millimeter are tolerable when subjected to launch-induced loads.

  6. Application of advanced irradiation analysis methods to light water reactor pressure vessel test and surveillance programs

    International Nuclear Information System (INIS)

    Odette, R.; Dudey, N.; McElroy, W.; Wullaert, R.; Fabry, A.

    1977-01-01

    Inaccurate characterization and inappropriate application of neutron irradiation exposure variables contribute a substantial amount of uncertainty to embrittlement analysis of light water reactor pressure vessels. Damage analysis involves characterization of the irradiation environment (dosimetry), correlation of test and surveillance metallurgical and dosimetry data, and projection of such data to service conditions. Errors in available test and surveillance dosimetry data are estimated to contribute a factor of approximately 2 to the data scatter. Non-physical (empirical) correlation procedures and the need to extrapolate to the vessel may add further error. Substantial reductions in these uncertainties in future programs can be obtained from a more complete application of available damage analysis tools which have been developed for the fast reactor program. An approach to reducing embrittlement analysis errors is described, and specific examples of potential applications are given. The approach is based on damage analysis techniques validated and calibrated in benchmark environments

  7. LARGE-SCALE MECURY CONTROL TECHNOLOGY TESTING FOR LIGNITE-FIRED UTILITIES-OXIDATION SYSTEMS FOR WET FGD

    Energy Technology Data Exchange (ETDEWEB)

    Michael J. Holmes; Steven A. Benson; Jeffrey S. Thompson

    2004-03-01

    The Energy & Environmental Research Center (EERC) is conducting a consortium-based effort directed toward resolving the mercury (Hg) control issues facing the lignite industry. Specifically, the EERC team--the EERC, EPRI, URS, ADA-ES, Babcock & Wilcox, the North Dakota Industrial Commission, SaskPower, and the Mercury Task Force, which includes Basin Electric Power Cooperative, Otter Tail Power Company, Great River Energy, Texas Utilities (TXU), Montana-Dakota Utilities Co., Minnkota Power Cooperative, BNI Coal Ltd., Dakota Westmoreland Corporation, and the North American Coal Company--has undertaken a project to significantly and cost-effectively oxidize elemental mercury in lignite combustion gases, followed by capture in a wet scrubber. This approach will be applicable to virtually every lignite utility in the United States and Canada and potentially impact subbituminous utilities. The oxidation process is proven at the pilot-scale and in short-term full-scale tests. Additional optimization is continuing on oxidation technologies, and this project focuses on longer-term full-scale testing. The lignite industry has been proactive in advancing the understanding of and identifying control options for Hg in lignite combustion flue gases. Approximately 1 year ago, the EERC and EPRI began a series of Hg-related discussions with the Mercury Task Force as well as utilities firing Texas and Saskatchewan lignites. This project is one of three being undertaken by the consortium to perform large-scale Hg control technology testing to address the specific needs and challenges to be met in controlling Hg from lignite-fired power plants. This project involves Hg oxidation upstream of a system equipped with an electrostatic precipitator (ESP) followed by wet flue gas desulfurization (FGD). The team involved in conducting the technical aspects of the project includes the EERC, Babcock & Wilcox, URS, and ADA-ES. The host sites include Minnkota Power Cooperative Milton R. Young

  8. Reliability tests for reactor internals rejuvenation technology

    International Nuclear Information System (INIS)

    Fujimaki, Katsumi; Hitoki, Yoichi; Otsubo, Toru; Uchiyama, Junichi

    1998-01-01

    Structural damage due to aging degradation of LWR reactor internals has been reported in several nuclear plants. NUPEC has started a project to test the reliability of the technology for rejuvenating reactor internals which has been funded by the Ministry of International Trade and Industry (MITI) of Japan since 1995. The project follows the policy of a report that the MITI has formally issued in April 1996 summarizing the countermeasures to be considered for aging nuclear plants and equipment. This paper gives an outline of the test plans and results which are directed at preventive maintenance before damage and repair after damage for reactor internals aging degradation. The test results for the replacement methods of ICM housing and BWR core shroud have shown that the methods were reliable and the structural integrity was appropriate based on the evaluation. (author)

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

    International Nuclear Information System (INIS)

    1983-10-01

    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

  10. Pre-service proof pressure and leak rate tests for the Qinshan CANDU project reactor buildings

    International Nuclear Information System (INIS)

    Petrunik, K.J.; Khan, A.; Ricciuti, R.; Ivanov, A.; Chen, S.

    2003-01-01

    The Qinshan CANDU Project Reactor Buildings (Units 1 and 2) have been successfully tested for the Pre-Service Proof Pressure and Integrated Leak Rate Tests. The Unit 1 tests took place from May 3 to May 9, 2002 and from May 22 to May 25, 2002, and the Unit 2 tests took place from January 21 to January 27, 2003. This paper discusses the significant steps taken at minimum cost on the Qinshan CANDU Project, which has resulted in a) very good leak rate (0.21%) for Unit 1 and excellent leak rate (0.130%) for Unit 2; b) continuous monitoring of the structural behaviour during the Proof Pressure Test, thus eliminating any repeat of the structural test due to lack of data; and c) significant schedule reduction achieved for these tests in Unit 2. (author)

  11. Heat resistant/radiation resistant cable and incore structure test device for FBR type reactor

    International Nuclear Information System (INIS)

    Tanimoto, Hajime; Shiono, Takeo; Sato, Yoshimi; Ito, Kazumi; Sudo, Shigeaki; Saito, Shin-ichi; Mitsui, Hisayasu.

    1995-01-01

    A heat resistant/radiation resistant coaxial cable of the present invention comprises an insulation layer, an outer conductor and a protection cover in this order on an inner conductor, in which the insulation layer comprises thermoplastic polyimide. In the same manner, a heat resistant/radiation resistant power cable has an insulation layer comprising thermoplastic polyimide on a conductor, and is provided with a protection cover comprising braid of alamide fibers at the outer circumference of the insulation layer. An incore structure test device for an FBR type reactor comprises the heat resistant/radiation resistant coaxial cable and/or the power cable. The thermoplastic polyimide can be extrusion molded, and has excellent radiation resistant by the extrusion, as well as has high dielectric withstand voltage, good flexibility and electric characteristics at high temperature. The incore structure test device for the FBR type reactor of the present invention comprising such a cable has excellent reliability and durability. (T.M.)

  12. Model test on interaction of reactor building and soil. Part 1

    International Nuclear Information System (INIS)

    Iguchi, M.; Akino, K.; Kiva, Y.

    1989-01-01

    Theoretical and experimental studies on the effects of dynamic interaction between structures and soil have been carried out in recent years. Most of the dynamic tests, however, have been conducted using comparatively small-scale models. In order to evaluate the effects of soil-structure interaction for rigid structure such as reactor building, a series of tests, including forced vibration test and earthquake observations, was carried out. Large-scale models constructed on an actual soil were used. These tests included forced vibration tests on individual foundations, on foundations with superstructures, on cross interaction through the soil between adjacent structures. Tests on the embedded effects of foundation, on artificial ground-shaking, on large amplitude excitation, and aging effects in soil properties were performed. This paper describes the results of forced vibration tests and analyses of cross interaction through the soil between adjacent structures

  13. First TREAT [Transient Reactor Test Facility] transient overpower tests on U-Pu-Zr fuel: M5 and M6

    International Nuclear Information System (INIS)

    Robinson, W.R.; Bauer, T.H.; Wright, A.E.; Rhodes, E.A.; Stanford, G.S.; Klickman, A.E.

    1987-01-01

    Transient Reactor Test Facility (TREAT) tests M5 and M6 were the first transient overpower (TOP) tests of the margin to cladding breach and prefailure elongation of metallic U-Pu-Zr ternary fuel, the reference fuel of the Integral Fast Reactor concept. Similar tests on U-Fs fueled EBR-II driver pins were previously performed and reported [1,2]. Results from these earlier tests indicated a margin to failure of about 4 times nominal power and significant axial elongation prior to failure, a feature that was very pronounced at low burnups. While these two fuel types are similar in many respects, the ternary alloy exhibits a much more complex physical structure and is typically irradiated at much higher temperatures. Thus, a prime motivation for performing M5 and M6 was to compare the safety related fuel performance characteristics of U-Fs and U-Pu-Zr. This report described conditions, results, and conclusions of testing of these fuel types

  14. Vortex Diode Analysis and Testing for Fluoride Salt-Cooled High-Temperature Reactors

    International Nuclear Information System (INIS)

    Yoder, Graydon L. Jr.; Elkassabgi, Yousri M.; De Leon, Gerardo I.; Fetterly, Caitlin N.; Ramos, Jorge A.; Cunningham, Richard Burns

    2012-01-01

    Fluidic diodes are presently being considered for use in several fluoride salt-cooled high-temperature reactor designs. A fluidic diode is a passive device that acts as a leaky check valve. These devices are installed in emergency heat removal systems that are designed to passively remove reactor decay heat using natural circulation. The direct reactor auxiliary cooling system (DRACS) uses DRACS salt-to-salt heat exchangers (DHXs) that operate in a path parallel to the core flow. Because of this geometry, under normal operating conditions some flow bypasses the core and flows through the DHX. A flow diode, operating in reverse direction, is-used to minimize this flow when the primary coolant pumps are in operation, while allowing forward flow through the DHX under natural circulation conditions. The DRACSs reject the core decay heat to the environment under loss-of-flow accident conditions and as such are a reactor safety feature. Fluidic diodes have not previously been used in an operating reactor system, and therefore their characteristics must be quantified to ensure successful operation. This report parametrically examines multiple design parameters of a vortex-type fluidic diode to determine the size of diode needed to reject a particular amount of decay heat. Additional calculations were performed to size a scaled diode that could be tested in the Oak Ridge National Laboratory Liquid Salt Flow Loop. These parametric studies have shown that a 152.4 mm diode could be used as a test article in that facility. A design for this diode is developed, and changes to the loop that will be necessary to test the diode are discussed. Initial testing of a scaled flow diode has been carried out in a water loop. The 150 mm diode design discussed above was modified to improve performance, and the final design tested was a 171.45 mm diameter vortex diode. The results of this testing indicate that diodicities of about 20 can be obtained for diodes of this size. Experimental

  15. Field test and evaluation of the passive neutron coincidence collar for prototype fast reactor fuel subassemblies

    International Nuclear Information System (INIS)

    Menlove, H.O.; Keddar, A.

    1982-08-01

    The passive neutron Coincidence Collar, which was developed for the verification of plutonium content in fast reactor fuel subassemblies, has been field tested using Prototype Fast Reactor fuel. For passive applications, the system measures the 240 Pu-effective mass from the spontaneous fission rate, and in addition, a self-interrogation technique is used to determine the fissile content in the subassembly. Both the passive and active modes were evaluated at the Windscale Works in the United Kingdom. The results of the tests gave a standard deviation 0.75% for the passive count and 3 to 7% for the active measurement for a 1000-s counting time. The unit will be used in the future for the verification of plutonium in fresh fuel assemblies

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

    International Nuclear Information System (INIS)

    Roettger, H.; Hardt, P. von der; Tas, A.; Voorbraak, W.P.

    1981-01-01

    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 January 1981

  17. Advancing nuclear technology and research. The advanced test reactor national scientific user facility

    Energy Technology Data Exchange (ETDEWEB)

    Benson, Jeff B; Marshall, Frances M [Idaho National Laboratory, Idaho Falls, ID (United States); Allen, Todd R [Univ. of Wisconsin, Madison, WI (United States)

    2012-03-15

    The Advanced Test Reactor (ATR), at the Idaho National Laboratory (INL), is one of the world's premier test reactors for providing the capability for studying the effects of intense neutron and gamma radiation on reactor materials and fuels. The INL also has several hot cells and other laboratories in which irradiated material can be examined to study material radiation effects. In 2007 the US Department of Energy (DOE) designated the ATR as a National Scientific User Facility (NSUF) to facilitate greater access to the ATR and the associated INL laboratories for material testing research. The mission of the ATR NSUF is to provide access to world-class facilities, thereby facilitating the advancement of nuclear science and technology. Cost free access to the ATR, INL post irradiation examination facilities, and partner facilities is granted based on technical merit to U.S. university-led experiment teams conducting non-proprietary research. Proposals are selected via independent technical peer review and relevance to United States Department of Energy. To increase overall research capability, ATR NSUF seeks to form strategic partnerships with university facilities that add significant nuclear research capability to the ATR NSUF and are accessible to all ATR NSUF users. (author)

  18. Experience in handling core subassemblies in sodium cooled reactor KNK and test rigs

    International Nuclear Information System (INIS)

    Althaus; Jansing; Kesseler; Kirchner; Menck

    1974-01-01

    Compared with a water cooled reactor plant a sodium cooled reactor plant presents a number of problems which result from the specific nature of sodium. These problems that must be faced during all handling operations are mainly: 1. The rapid reaction of sodium in air requires handling to be done only under cover gas. 2. The temperature of all sodium-wetted components is to be kept above the melting point of sodium. 3. Poor draining of removed reactor components due to the high surface tension of sodium and the associated danger of dripping radioactive sodium may produce radiation or contamination problems. 4. Sodium is not transparent. The sum of these and further influences dictate that the general handling usually is carried out without visual means, though a method is under development in the USA to use ultrasonic for under sodium 'viewing'. These limitations to sodium component handling are applicable to all sodium reactor plants, several of which are discussed in this report. After the description of the handling systems of the KNK plant now operating at Karlsruhe, the experience with the SNR test rig and finally the handling systems for SNR 300 and SNR 2 are discussed

  19. Design and development of a zoom lens objective for the fast breeder test reactor periscope

    International Nuclear Information System (INIS)

    Das, N.C.; Udupa, D.V.; Shukla, R.P.

    2003-10-01

    A three lens optically compensated zoom lens useful for the 5 meter long periscope in the Fast Breeder Test Reactor (FBTR) has been designed, fabricated and tested. The zoom lens fabricated using radiation resistant glasses has a zoom ratio of 2.5 with a focal length range of l00 mm to 250 mm. The zoom lens objective has been designed for viewing the objects kept at a distance in the range of 1.5 m to 3 m from the objective lens. It is found that the zoom lens objective can be used for resolving objects with a linear resolution of 0.2 mm inside the reactor when viewed with an eye piece of focal length 50 mm. (author)

  20. Flow inversion and natural convection in a MTR (Materials Testing Reactor)

    International Nuclear Information System (INIS)

    Gimenez, M.O.; Clausse, A.

    1990-01-01

    The thermohydraulic evolution of a refrigerating channel of the MTR (Materials Testing Reactors) RA-6 reactor's core, at the Bariloche Atomic Center, has been studied during the transient caused by the primary system's pump decommissioning. This transient constitutes one of the reactor's operating power boundaries due to the maximum temperature permissible in fuel plates. The problem regarding the thermohydraulic code altered for the rectangular geometry calculation characteristic of the MTR design is analyzed. (Author) [es

  1. Feeder replacement tooling and processes

    International Nuclear Information System (INIS)

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

    2008-01-01

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

  2. 'Experience with decommissioning of research and test reactors at Argonne National Laboratory'

    International Nuclear Information System (INIS)

    Bhattacharyya, S.K.; Yule, T.J.; Fellhauer, C.R.; Boing, L.E.

    2002-01-01

    A large number of research reactors around the world have reached the end of their useful operational life. Many of these are kept in a controlled storage mode awaiting decontamination and decommissioning (D and D). At Argonne National Laboratory located near Chicago in the United States of America, significant experience has been gained in the D and D of research and test reactors. These experiences span the entire range of activities in D and D - from planning and characterization of the facilities to the eventual disposition of all waste. A multifaceted D nd D program has been in progress at the Argonne National Laboratory - East site for nearly a decade. The program consists of three elements: - D and D of nuclear facilities on the site that have reached the end of their useful life; - Development and demonstrations of technologies that help in safe and cost effective D and D; - Presentation of training courses in D and D practices. Nuclear reactor facilities have been constructed and operated at the ANL-E site since the earliest days of nuclear power. As a result, a number of these early reactors reached end-of-life long before reactors on other sites and were ready for D and D earlier. They presented an excellent set of test beds on which D and D practices and technologies could be demonstrated in environments that were similar to commercial reactors, but considerably less hazardous. As shown, four reactor facilities, plutonium contaminated glove boxes and hot cells, a cyclotron facility and assorted other nuclear related facilities have been decommissioned in this program. The overall cost of the program has been modest relative to the cost of comparable projects undertaken both in the U.S. and abroad. The safety record throughout the program was excellent. Complementing the actual operations, a set of D and D technologies are being developed. These include robotic methods of tool handling and operation, chemical and laser decontamination techniques, sensors

  3. Hanford low-level waste process chemistry testing data package

    International Nuclear Information System (INIS)

    Smith, H.D.; Tracey, E.M.; Darab, J.G.; Smith, P.A.

    1996-03-01

    Recently, the Tri-Party Agreement (TPA) among the State of Washington Department of Ecology, U.S. Department of Energy (DOE) and the US Environmental Protection Agency (EPA) for the cleanup of the Hanford Site was renegotiated. The revised agreement specifies vitrification as the encapsulation technology for low level waste (LLW). A demonstration, testing, and evaluation program underway at Westinghouse Hanford Company to identify the best overall melter-system technology available for vitrification of Hanford Site LLW to meet the TPA milestones. Phase I is a open-quotes proof of principleclose quotes test to demonstrate that a melter system can process a simulated highly alkaline, high nitrate/nitrite content aqueous LLW feed into a glass product of consistent quality. Seven melter vendors were selected for the Phase I evaluation: joule-heated melters from GTS Duratek, Incorporated (GDI); Envitco, Incorporated (EVI); Penberthy Electomelt, Incorporated (PEI); and Vectra Technologies, Incorporated (VTI); a gas-fired cyclone burner from Babcock ampersand Wilcox (BCW); a plasma torch-fired, cupola furnace from Westinghouse Science and Technology Center (WSTC); and an electric arc furnace with top-entering vertical carbon electrodes from the U.S. Bureau of Mines (USBM)

  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

    2014-01-01

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

  5. The RERTR [Reduced Enrichment Research and Test Reactor] program: A progress report

    International Nuclear Information System (INIS)

    Travelli, A.

    1986-11-01

    The progress of the Reduced Enrichment Research and Test Reactor (RERTR) Program is described. After a brief summary of the results which the RERTR Program, in collaboration with its many international partners, had achieved by the end of 1985, the activities, results, and new developments which occurred in 1986 are reviewed. The second miniplate series, concentrating on U 3 Si 2 -Al and U 3 Si-Al fuels, was expanded and its irradiation continued. Postirradiation examinations of several of these miniplates and of six previously irradiated U 3 Si 2 -Al full-size elements were completed with excellent results. The whole-core ORR demonstration with U 3 Si 2 -Al fuel at 4.8 g U/cm 3 is well under way and due for completion before the end of 1987. DOE removed an important barrier to conversions by announcing that the new LEU fuels will be accepted for reprocessing. New DOE prices for enrichment and reprocessing services were calculated to have minimal effect on HEU reactors, and to reduce by about 8 to 10% the total fuel cycle costs of LEU reactors. New program activities include preliminary feasibility studies of LEU use in DOE reactors, evaluation of the feasibility to use LEU targets for the production of fission-product 99 Mo, and responsibility for coordinating safety evaluations related to LEU conversions of US university reactors, as required by NRC. Achievement of the final program goals is projected for 1990. This progress could not have been achieved without close international cooperation, whose continuation and intensification are essential to the achievement of the ultimate goals of the RERTR Program

  6. Hydraulic and hydrodynamic tests for design evaluation of research reactors fuel elements

    International Nuclear Information System (INIS)

    Kulichevsky, R.; Martin Ghiselli, A.; Fiori, J.; Yedros, P.

    2002-01-01

    During the design steps of research reactors fuel elements some tests are usually necessary to verify its design, i.e.: its hydraulic characteristics, dynamical response and structural integrity. The hydraulic tests are developed in order to know the pressure drops characteristics of different parts or elements of the prototype and of the whole fuel element. Also, some tests are carried out to obtain the velocity distribution of the coolant water across different prototype's sections. The hydrodynamic tests scopes are the assessment of the dynamical characteristics of the fuel elements and their components and its dynamical response considering the forces generated by the coolant flowing water at different flow rate conditions. Endurance tests are also necessary to qualify the structural design of the FE prototypes and their corresponding clamp tools, verifying the whole system structural integrity and wear processes influences. To carry out these tests a special test facility is needed to obtain a proper representation of the hydraulic and geometric boundary conditions of the fuel element. In some cases changes on the fuel element prototype or dummy are necessary to assure that the data results are representative of the case under study. Different kind of sensors are mounted on the test section and also on the fuel element itself when necessary. Some examples of the instrumentation used are strain gauges, displacement transducers, absolute and differential pressure transducers, pitot tubes, etc. The obtained data are, for example, plates' vibration amplitudes and frequencies, whole bundle displacement characterization, pressure drops and flow velocity measurements. The Experimental Low Pressure Loop is a hydraulic loop located at CNEA's Constituyentes Atomic Center and is the test facility where different kind of tests are performed in order to support and evaluate the design of research reactor fuel elements. A brief description of the facility, and examples of

  7. NRC review of passive reactor design certification testing programs: Overview and regulatory perspective

    International Nuclear Information System (INIS)

    Levin, A.E.

    1993-01-01

    Reactor vendors are developing new designs for future deployment, including open-quotes passiveclose quotes light water reactors (LWRs), such as General Electric's (G.E.'s) simplified boiling water reactor (SBWR) and Westinghouse's AP600, which depend primarily on inherent processes, such as national convection and gravity feed, for safety injection and emergency core cooling. The U.S. Nuclear Regulatory Commission (NRC) has implemented a new process, certification of standardized reactor designs, for licensing these Plants. Part 52 of Title 10 of the Code of Federal Regulations (10CFR52) contains the requirements that vendors must meet for design certification. One important section, 10CFR52.47, reads open-quotes Certification of a standard design which . . . utilizes simplified, inherent, passive, or other innovative means to accomplish its safety functions will be granted only if: (1) The performance of each safety feature of the design has been demonstrated through either analysis, appropriate test programs, experience, or a combination thereof; (2) Interdependent effects among the safety features have been found acceptable by analysis, appropriate test programs, experience, or a combination thereof; and (3) Sufficient data exist on the safety features of the design to assess the analytical tools used for safety analyses. . . . The vendors have initiated programs to test innovative features of their designs and to develop data bases needed to validate their analytical codes, as required by the design certification rule. Accordingly, the NRC is reviewing and evaluating the vendors programs to ensure that they address adequately key issues concerning safety system performance. This paper provides an overview of the NRC's review process and regulatory perspective

  8. Industrial Hygiene Concerns during the Decontamination and Decommissioning of the Tokamak Fusion Test Reactor

    International Nuclear Information System (INIS)

    M.E. Lumia; C.A. Gentile

    2002-01-01

    A significant industrial hygiene concern during the Decontamination and Decommissioning (D and D) of the Tokamak Fusion Test Reactor (TFTR) was the oxidation of the lead bricks' surface, which were utilized for radiation shielding. This presented both airborne exposure and surface contamination issues for the workers in the field removing this material. This paper will detail the various protection and control methods tested and implemented to protect the workers, including those technologies deployed to decontaminate the work surfaces. In addition, those techniques employed to recycle the lead for additional use at the site will be discussed

  9. Industrial Hygiene Concerns during the Decontamination and Decommissioning of the Tokamak Fusion Test Reactor

    CERN Document Server

    Lumia, M E

    2002-01-01

    A significant industrial hygiene concern during the Decontamination and Decommissioning (D and D) of the Tokamak Fusion Test Reactor (TFTR) was the oxidation of the lead bricks' surface, which were utilized for radiation shielding. This presented both airborne exposure and surface contamination issues for the workers in the field removing this material. This paper will detail the various protection and control methods tested and implemented to protect the workers, including those technologies deployed to decontaminate the work surfaces. In addition, those techniques employed to recycle the lead for additional use at the site will be discussed.

  10. Technology, safety, and costs of decommissioning reference nuclear research and test reactors. Appendices

    International Nuclear Information System (INIS)

    Konzek, G.J.; Ludwick, J.D.; Kennedy, W.E. Jr.; Smith, R.I.

    1982-03-01

    Safety and Cost Information is developed for the conceptual decommissioning of two representative licensed nuclear research and test reactors. Three decommissioning alternatives are studied to obtain comparisons between costs (in 1981 dollars), occupational radiation doses, potential radiation dose to the public, and other safety impacts. The alternatives considered are: DECON (immediate decontamination), SAFSTOR (safe storage followed by deferred decontamination), and EMTOMB (entombment). The study results are presented in two volumes. Volume 2 (Appendices) contains the detailed data that support the results given in Volume 1, including unit-component data

  11. Technology, safety, and costs of decommissioning reference nuclear research and test reactors. Appendices

    Energy Technology Data Exchange (ETDEWEB)

    Konzek, G.J.; Ludwick, J.D.; Kennedy, W.E. Jr.; Smith, R.I.

    1982-03-01

    Safety and Cost Information is developed for the conceptual decommissioning of two representative licensed nuclear research and test reactors. Three decommissioning alternatives are studied to obtain comparisons between costs (in 1981 dollars), occupational radiation doses, potential radiation dose to the public, and other safety impacts. The alternatives considered are: DECON (immediate decontamination), SAFSTOR (safe storage followed by deferred decontamination), and EMTOMB (entombment). The study results are presented in two volumes. Volume 2 (Appendices) contains the detailed data that support the results given in Volume 1, including unit-component data.

  12. Testing of plain and fibrous concrete single cavity prestressed concrete reactor vessel models

    International Nuclear Information System (INIS)

    Oland, C.B.

    1985-01-01

    Two single-cavity prestressed concrete reactor vessel (PCRV) models were fabricated and tested to failure to demonstrate the structural response and ultimate pressure capacity of models cast from high-strength concretes. Concretes with design compressive strengths in excess of 70 MPa (10,000 psi) were developed for this investigation. One model was cast from plain concrete and failed in shear at the head region. The second model was cast from fiber reinforced concrete and failed by rupturing the circumferential prestressing at the sidewall of the structure. The tests also demonstrated the capabilities of the liner system to maintain a leak-tight pressure boundary. 3 refs., 4 figs

  13. Fission reactors and materials

    International Nuclear Information System (INIS)

    Frost, B.R.T.

    1981-12-01

    The American-designed boiling water reactor and pressurized water reactor dominate the designs currently in use and under construction worldwide. As in all energy systems, materials problems have appeared during service; these include stress-corrosion of stainless steel pipes and heat exchangers and questions regarding crack behavior in pressure vessels. To obtain the maximum potential energy from our limited uranium supplies is is essential to develop the fast breeder reactor. The materials in these reactors are subjected to higher temperatures and neutron fluxes but lower pressures than in the water reactors. The performance required of the fuel elements is more arduous in the breeder than in water reactors. Extensive materials programs are in progress in test reactors and in large test rigs to ensure that materials will be available to meet these conditions

  14. Los Alamos PWR feed-and-bleed studies summary results and conclusions

    International Nuclear Information System (INIS)

    Boyack, B.E.; Henninger, R.J.; Lime, J.F.

    1985-01-01

    The adequacy of shutdown decay heat removal in pressurized water reactors (PWRs) is currently under investigation by the Nuclear Regulatory Commission. One part of this effort is review of feed-and-bleed procedures that could be used if the normal cooling mode through the steam generators was unavailable. Feed-and-bleed cooling is effected by manually activating the high-pressure injection (HPI) system and opening the power-operated relief valves (PORVs) to release the core decay energy. The feasibility of the feed-and-bleed concept as a diverse mode of heat removal has been evaluated at the Los Alamos National Laboratory. The TRAC-PF1 code has been used to predict the expected performances of the Oconee-1 and Calvert Cliffs-1 reactors of Babcock and Wilcox and Combustion Engineering, respectively, and the Zion-1 and H.B. Robinson-2 plants of Westinghouse. Feed and bleed was successfully applied in each of the four plants studied, provided it was initiated no later than the time of loss-of-secondary heat sink

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-06-01

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

  16. The development and testing of reduced enrichment fuels for Canadian research reactors

    International Nuclear Information System (INIS)

    Wood, J.C.; Foo, M.T.; Berthiaume, L.C.

    1983-01-01

    Fuel rods of uranium silicide dispersed in aluminum and clad in aluminum have been developed and tested in the laboratory and in-reactor. The properties of the dispersion fuel materials proved satisfactory with regard to thermal conductivity, aqueous corrosion resistance, strength and ductility, and thermal stability below 473 K. A vacancy condensation model is proposed to account for the thermally-induced swelling that occurs above 473 K by virtue of the chemical reactions that occur between the dispersed silicide fuel particles and the aluminum matrix. The in-reactor fuel core swelling was less than % after irradiation at high powers 76-131 kW/m) to a high terminal burnup (79.2 at% of U-235 atoms). (author)

  17. Ultracold neutron source at the PULSTAR reactor: Engineering design and cryogenic testing

    Energy Technology Data Exchange (ETDEWEB)

    Korobkina, E., E-mail: ekorobk@ncsu.edu [Department of Nuclear Engineering, North Carolina State University, 2500 Stinson Drive, Box 7909, Raleigh, NC 27695 (United States); Medlin, G. [Department of Physics, North Carolina State University, 2401 Stinson Drive, Box 8202, Raleigh, NC 27695 (United States); Triangle Universities Nuclear Laboratory, 116 Science Drive, Box 90308, Durham, NC 27708 (United States); Wehring, B.; Hawari, A.I. [Department of Nuclear Engineering, North Carolina State University, 2500 Stinson Drive, Box 7909, Raleigh, NC 27695 (United States); Huffman, P.R.; Young, A.R. [Department of Physics, North Carolina State University, 2401 Stinson Drive, Box 8202, Raleigh, NC 27695 (United States); Triangle Universities Nuclear Laboratory, 116 Science Drive, Box 90308, Durham, NC 27708 (United States); Beaumont, B. [Department of Physics, North Carolina State University, 2401 Stinson Drive, Box 8202, Raleigh, NC 27695 (United States); Palmquist, G. [Department of Physics, North Carolina State University, 2401 Stinson Drive, Box 8202, Raleigh, NC 27695 (United States); Triangle Universities Nuclear Laboratory, 116 Science Drive, Box 90308, Durham, NC 27708 (United States)

    2014-12-11

    Construction is completed and commissioning is in progress for an ultracold neutron (UCN) source at the PULSTAR reactor on the campus of North Carolina State University. The source utilizes two stages of neutron moderation, one in heavy water at room temperature and the other in solid methane at ∼40K, followed by a converter stage, solid deuterium at 5 K, that allows a single down scattering of cold neutrons to provide UCN. The UCN source rolls into the thermal column enclosure of the PULSTAR reactor, where neutrons will be delivered from a bare face of the reactor core by streaming through a graphite-lined assembly. The source infrastructure, i.e., graphite-lined assembly, heavy-water system, gas handling system, and helium liquefier cooling system, has been tested and all systems operate as predicted. The research program being considered for the PULSTAR UCN source includes the physics of UCN production, fundamental particle physics, and material surface studies of nanolayers containing hydrogen. In the present paper we report details of the engineering and cryogenic design of the facility as well as results of critical commissioning tests without neutrons.

  18. Nondestructive testing of PWR type fuel rods by eddy currents and metrology in the OSIRIS reactor pool

    International Nuclear Information System (INIS)

    Faure, M.; Marchand, L.

    1985-02-01

    The Saclay Reactor Department has developed a nondestructive test bench, now installed above channel 1 of the OSIRIS reactor. As part of investigations into the dynamics of PWR fuel degradation, a number of fuel rods underwent metrological and eddy current inspection, after irradiation [fr

  19. The Tokamak Fusion Test Reactor D-T modifications and operations

    International Nuclear Information System (INIS)

    1992-01-01

    This Environmental Assessment (EA) was prepared in accordance with the National Environmental Policy Act (NEPA) of 1969, as amended, in support of the Department of Energy's proposal for the Tokamak Fusion Test Reactor (TFTR) D-T program. The objective of the proposed D-T program is to take the initial step in studying the effects of alpha particle heating and transport in a magnetic fusion device. These studies would enable the successful completion of the original TFTR program objectives, and would support the research and development needs of the Burning Plasma Experiment, BPX (formerly the Compact Ignition Tokamak (CIT)) and International Thermonuclear Experimental Reactor (ITER) in the areas of alpha particle physics, tritium retention, alpha particle diagnostic development, and tritium handling

  20. Summary of Thermocouple Performance During Advanced Gas Reactor Fuel Irradiation Experiments in the Advanced Test Reactor and Out-of-Pile Thermocouple Testing in Support of Such Experiments

    Energy Technology Data Exchange (ETDEWEB)

    A. J. Palmer; DC Haggard; J. W. Herter; M. Scervini; W. D. Swank; D. L. Knudson; R. S. Cherry

    2011-07-01

    High temperature gas reactor experiments create unique challenges for thermocouple based temperature measurements. As a result of the interaction with neutrons, the thermoelements of the thermocouples undergo transmutation, which produces a time dependent change in composition and, as a consequence, a time dependent drift of the thermocouple signal. This drift is particularly severe for high temperature platinum-rhodium thermocouples (Types S, R, and B); and tungsten-rhenium thermocouples (Types C and W). For lower temperature applications, previous experiences with type K thermocouples in nuclear reactors have shown that they are affected by neutron irradiation only to a limited extent. Similarly type N thermocouples are expected to be only slightly affected by neutron fluxes. Currently the use of these Nickel based thermocouples is limited when the temperature exceeds 1000°C due to drift related to phenomena other than nuclear irradiation. High rates of open-circuit failure are also typical. Over the past ten years, three long-term Advanced Gas Reactor (AGR) experiments have been conducted with measured temperatures ranging from 700oC – 1200oC. A variety of standard Type N and specialty thermocouple designs have been used in these experiments with mixed results. A brief summary of thermocouple performance in these experiments is provided. Most recently, out of pile testing has been conducted on a variety of Type N thermocouple designs at the following (nominal) temperatures and durations: 1150oC and 1200oC for 2000 hours at each temperature, followed by 200 hours at 1250oC, and 200 hours at 1300oC. The standard Type N design utilizes high purity crushed MgO insulation and an Inconel 600 sheath. Several variations on the standard Type N design were tested, including Haynes 214 alloy sheath, spinel (MgAl2O4) insulation instead of MgO, a customized sheath developed at the University of Cambridge, and finally a loose assembly thermocouple with hard fired alumina

  1. Summary of thermocouple performance during advanced gas reactor fuel irradiation experiments in the advanced test reactor and out-of-pile thermocouple testing in support of such experiments

    Energy Technology Data Exchange (ETDEWEB)

    Palmer, A. J.; Haggard, DC; Herter, J. W.; Swank, W. D.; Knudson, D. L.; Cherry, R. S. [Idaho National Laboratory, P.O. Box 1625, MS 4112, Idaho Falls, ID, (United States); Scervini, M. [University of Cambridge, Department of Material Science and Metallurgy, 27 Charles Babbage Road, CB3 0FS, Cambridge, (United Kingdom)

    2015-07-01

    High temperature gas reactor experiments create unique challenges for thermocouple-based temperature measurements. As a result of the interaction with neutrons, the thermoelements of the thermocouples undergo transmutation, which produces a time-dependent change in composition and, as a consequence, a time-dependent drift of the thermocouple signal. This drift is particularly severe for high temperature platinum-rhodium thermocouples (Types S, R, and B) and tungsten-rhenium thermocouples (Type C). For lower temperature applications, previous experiences with Type K thermocouples in nuclear reactors have shown that they are affected by neutron irradiation only to a limited extent. Similarly, Type N thermocouples are expected to be only slightly affected by neutron fluence. Currently, the use of these nickel-based thermocouples is limited when the temperature exceeds 1000 deg. C due to drift related to phenomena other than nuclear irradiation. High rates of open-circuit failure are also typical. Over the past 10 years, three long-term Advanced Gas Reactor experiments have been conducted with measured temperatures ranging from 700 deg. C - 1200 deg. C. A variety of standard Type N and specialty thermocouple designs have been used in these experiments with mixed results. A brief summary of thermocouple performance in these experiments is provided. Most recently, out-of-pile testing has been conducted on a variety of Type N thermocouple designs at the following (nominal) temperatures and durations: 1150 deg. C and 1200 deg. C for 2,000 hours at each temperature, followed by 200 hours at 1250 deg. C and 200 hours at 1300 deg. C. The standard Type N design utilizes high purity, crushed MgO insulation and an Inconel 600 sheath. Several variations on the standard Type N design were tested, including a Haynes 214 alloy sheath, spinel (MgAl{sub 2}O{sub 4}) insulation instead of MgO, a customized sheath developed at the University of Cambridge, and finally a loose assembly

  2. Basic Characteristics of Human Erroneous Actions during Test and Maintenance Activities Leading to Unplanned Reactor Trips

    International Nuclear Information System (INIS)

    Kim, Jae Whan; Park, Jin Kyun

    2010-01-01

    Test and maintenance (T and M) activities of nuclear power plants are essential for sustaining the safety of a power plant and maintaining the reliability of plant systems and components. However, the potential of human errors during T and M activities has also the potential to induce unplanned reactor trips or power derate or making safety-related systems unavailable. According to the major incident/accident reports of nuclear power plants in Korea, contribution of human errors takes up about 20% of the total events. The previous study presents that most of human-related unplanned reactor trip events during normal power operation are associated with T and M activities (63%), which are comprised of plant maintenance activities such as a 'periodic preventive maintenance (PPM)', a 'planned maintenance (PM)' and a 'corrective maintenance (CM)'. This means that T and M activities should be a major subject for reducing the frequency of human-related unplanned reactor trips. This paper aims to introduce basic characteristics of human erroneous actions involved in the test and maintenance-induced unplanned reactor trip events that have occurred between 1986 and 2006 in Korean nuclear power plants. The basic characteristics are described by dividing human erroneous actions into planning-based errors and execution-based errors. For the events associated with planning failures, they are, firstly, classified according to existence of the work procedure and then described for what aspects of the procedure or work plan have deficiency or problem. On the other hand, for the events associated with execution failures, they are described from the aspect of external error modes

  3. Monitoring and Control Research Using a University Reactor and SBWR Test-Loop

    International Nuclear Information System (INIS)

    Edwards, Robert M.

    2003-01-01

    The existing hybrid simulation capability of the Penn State Breazeale nuclear reactor was expanded to conduct research for monitoring, operations and control. Hybrid simulation in this context refers to the use of the physical time response of the research reactor as an input signal to a real-time simulation of power-reactor thermal-hydraulics which in-turn provides a feedback signal to the reactor through positioning of an experimental changeable reactivity device. An ECRD is an aluminum tube containing an absorber material that is positioned in the central themble of the reactor kinetics were used to expand the hybrid reactor simulation (HRS) capability to include out-of-phase stability characteristics observed in operating BWRs

  4. Dynamic analysis of the PEC fast reactor vessel: On-site tests and mathematical models

    International Nuclear Information System (INIS)

    Zola, M.; Martelli, A.; Masoni, P.; Scandola, G.

    1988-01-01

    This paper presents the main features and results of the on-site dynamic tests and the related numerical analyses carried out for the PEC reactor vessel. The purpose is to provide an example of on-site testing of large components, stressing the problems encountered during the experiments, as well as in the processing phase of the test results and for the comparisons between calculations and measurements. Tests, performed by ISMES on behalf of ENEA, allowed the dynamic response of the empty vessel to be measured, thus providing data for the verification of the numerical models of the vessel supporting structure adopted in the PEC reactor-block seismic analysis. An axisymmetric model of the vessel, implemented in the NOVAX code, had been developed in the framework of the detailed numerical studies performed by NOVATOME (again on behalf of ENEA), to check the beam schematization with fluid added mass model adopted by ANSALDO in SAP-IV and ANSYS for the reactor-block design calculations. Furthermore, a numerical model, describing vessel supporting structure in detail, was also developed by ANSALDO and implemented in the SAP-IV code. The test conditions were analysed by use of these and the design models. Comparisons between calculations and measurements showed particularly good agreement with regard to first natural frequency of the vessel and rocking stiffness of the vessel supporting structure, i.e. those parameters on which vessel seismic amplification mainly depends: this demonstrated the adequacy of the design analysis to correctly calculate the seismic motion at the PEC core diagrid. (author). 5 refs, 23 figs, 4 tabs

  5. Dynamic analysis of the PEC fast reactor vessel: on-site tests and mathematical models

    International Nuclear Information System (INIS)

    Zola, Maurizio; Martelli, Alessandro; Maresca, Giuseppe; Masoni, Paolo; Scandola, Giani; Descleves, Pierre

    1988-01-01

    This paper presents the main features and results of the on-site dynamic tests and the related numerical analysis carried out for the PEC reactor vessel. The purpose is to provide an example of on-site testing of large components, stressing the problems encountered during the experiments, as well as in the processing phase of the test results and for the comparisons between calculations and measurements. Tests, performed by ISMES on behalf of ENEA, allowed the dynamic response of the empty vessel to be measured, thus providing data for the verification of the numerical models of the vessel supporting structure adopted in the PEC reactor-block seismic analysis. An axisymmetric model of the vessel, implemented in the vessel, implemented in the NOVAK code, had been developed in the framework of the detailed numerical studies performed by NOVATOME (again on behalf of ENEA), to check the beam schematization with fluid added mass model adopted by ANSALDO in SAP-IV and ANSYS for the reactor-block design calculations. Furthermore, a numerical model, describing vessel supporting structure in detail, was also developed by ANSALDO and implemented in the SAP-IV code. The test conditions were analysed by use of these and the design models. Comparisons between calculations and measurements showed particularly good agreement with regard to first natural frequency of the vessel and rocking stiffness of the vessel supporting structure, i.e. those parameters on which vessel seismic amplification mainly depends: this demonstrated the adequacy of the design analysis to correctly calculate the seismic motion at the PEC core diagrid. (author)

  6. Modeling and Experimental Tests on the Hydraulically Driven Control Rod option for IRIS Reactor

    International Nuclear Information System (INIS)

    Cammi, Antonio; Ricotti, Marco E.; Vitulo, Alessia

    2004-01-01

    The adoption of Internal Control Rod Drive Mechanisms (ICRDMs) represents a valuable alternative to classical, external CRDMs based on electro-magnetic devices, as adopted in current PWRs. The advantages on the safety features of the reactor are apparent: inherent elimination of the Rod Ejection accidents and of possible concerns about the vessel head penetrations. A further positive feedback on the design is the reduction of the primary system overall dimensions. Within the frame of the ICRDM concepts, the Hydraulically Driven Control Rod solution is investigated as a possible option for the IRIS integral reactor. After a brief comparison of the solutions currently proposed for integral reactors, the configuration of the Hydraulic Control Rod device for IRIS, made up by an external movable piston and an internal fixed cylinder, is described. A description of the whole control system is reported as well. Particular attention is devoted to the Control Rod profile characterization, performed by means of a Computational Fluid Dynamics (CFD) analysis. The investigation of the system behavior has been carried out, including the dynamic equilibrium and its stability properties, the withdrawal and insertion step movement and the sensitivity study on command time periods. A suitable dynamic model has been set up for the mentioned purposes: the models corresponding to the various Control Rod system devices have been written in an Object-Oriented language (Modelica), thus allowing an easy implementation of such a system into the simulator for the whole reactor. Finally, a preliminary low pressure, low temperature, reduced length experimental facility has been built. Tests on HDCR stability and operational transients have been performed. The results are compared with the dynamic system model and CFD simulation model, showing good agreement between simulations and experimental data. During these preliminary tests, the control system performed correctly, allowing stable dynamic

  7. Impact of closed Brayton cycle test results on gas cooled reactor operation and safety

    International Nuclear Information System (INIS)

    Wright, St.A.; Pickard, P.S.

    2007-01-01

    This report summarizes the measurements and model predictions for a series of tests supported by the U.S. Department of Energy that were performed using the recently constructed Sandia Brayton Loop (SBL-30). From the test results we have developed steady-state power operating curves, controls methodologies, and transient data for normal and off-normal behavior, such as loss of load events, and for decay heat removal conditions after shutdown. These tests and models show that because the turbomachinery operates off of the temperature difference (between the heat source and the heat sink), that the turbomachinery can continue to operate (off of sensible heat) for long periods of time without auxiliary power. For our test hardware, operations up to one hour have been observed. This effect can provide significant operations and safety benefits for nuclear reactors that are coupled to a Brayton cycles because the operating turbomachinery continues to provide cooling to the reactor. These capabilities mean that the decay-heat removal can be accommodated by properly managing the electrical power produced by the generator/alternator. In some conditions, it may even be possible to produce sufficient power to continue operating auxiliary systems including the waste heat circulatory system. In addition, the Brayton plant impacts the consequences of off-normal and accident events including loss of load and loss of on-site power. We have observed that for a loss of load or a loss of on-site power event, with a reactor scram, the transient consists initially of a turbomachinery speed increase to a new stable operating point. Because the turbomachinery is still spinning, the reactor is still being cooled provided the ultimate heat sink remains available. These highly desirable operational characteristics were observed in the Sandia Brayton loop. This type of behavior is also predicted by our models. Ultimately, these results provide the designers the opportunity to design gas

  8. High Flux Materials Testing Reactor (HFR), Petten

    International Nuclear Information System (INIS)

    1975-09-01

    After conversion to burnable poison fuel elements, the High Flux Materials Testing Reactor (HFR) Petten (Netherlands), operated through 1974 for 280 days at 45 MW. Equipment for irradiation experiments has been replaced and extended. The average annual occupation by experiments was 55% as compared to 38% in 1973. Work continued on thirty irradiation projects and ten development activities

  9. The RERTR [Reduced Enrichment Research and Test Reactor] Program: Progress and plans

    International Nuclear Information System (INIS)

    Travelli, A.

    1987-01-01

    The progress of the Reduced Enrichment Research and Test Reactor (RERTR) Program is described. After a brief summary of the results which the RERTR Program, in collaboration with its many international partners, had achieved by the end of 1986, the activities, results, and new developments which occurred in 1987 are reviewed. Irradiation of the second miniplate series, concentrating on U 3 Si 2 -Al and U 3 Si-Al fuels, was completed and postirradiation examinations were performed on many of its miniplates. The whole-core ORR demonstration with U 3 Si 2 -Al fuel at 4.8 g U/cm 3 was completed at the end of March with excellent results and with 29 elements estimated to have reached at least 40% average burnup. Good progress was made in the area of LEU usage for the production of fission 99 Mo, and in the coordination of safety evaluations related to LEU conversions of US university reactors. Planned activities include testing and demonstrating advanced fuels intended to allow use of reduced enrichment uranium in very-high-performance reactors. Two candidate fuels are U 3 Si-Al with 19.75% enrichment and U 3 Si 2 -Al with 45% enrichment. Demonstration of these fuels will include irradiation of full-size elements and, possibly, a full-core demonstration. Achievement of the final program goals is still projected for 1990. This progress could not have been possible without the close international cooperation which has existed from the beginning, and which is essential to the ultimate success of the RERTR Program

  10. Testing and analyses of a high temperature duct for gas-cooled reactors

    International Nuclear Information System (INIS)

    Black, W.E.; Roberge, A.; Felten, P.; Bastien, R.

    1979-01-01

    A 0.6 scale model of a steam cycle gas-cooled reactor high temperature duct was tested in a closed loop helium facility. The object of the test series was to determine: 1) the thermal effects of gas permeation within the thermal barrier, 2) the plastic deformation of the metallic components, and 3) the thermal performance of the fibrous insulation. A series of tests was performed with thermal cyclings from 100 0 C to 760 0 C at 50 atmospheres until the system thermal performance had stabilized hence enabling predictions for the reactor life. Additional tests were made to assess permeation by deliberately simulating sealing weld failures thereby allowing gas flow by-pass within the primary thermal barrier. After 100 cycles the entire primary structure was found to have performed without structural failure. Due to high pressures exerted by the insulation on the cover plates and a design oversight, the thin seal sheets were unable to expand in an anticipated manner. Local buckling resulted. Pre and post test metallurgical analyses were conducted on the Hastelloy-X structures and reference specimens. The results gave evidence of aging in the form of noticeable changes in room temperature tensile and reduction in area parameters. The Hastelloy-X welds exhibited greater changes in properties due to thermal aging. The antifriction coating (Cr 3 C 2 ) performed well without spallation or excessive wear. (orig.)

  11. The past, present, and future of test and research reactor physics

    International Nuclear Information System (INIS)

    Ryskamp, J.M.

    1992-01-01

    Reactor physics calculations have been performed on research reactors since the first one was built 50 yr ago under the University of Chicago stadium. Since then, reactor physics calculations have evolved from Fermi-age theory calculations performed with slide rules to three-dimensional, continuous-energy, coupled neutron-photon Monte Carlo computations performed with supercomputers and workstations. Such enormous progress in reactor physics leads us to believe that the next 50 year will be just as exciting. This paper reviews this transition from the past to the future

  12. Present status of Japan materials testing reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hori, Naohiko; Kaminaga, Masanori; Kusunoki, Tsuyoshi; Ishihara, Masahiro; Niimi, Motoji; Komori, Yoshihiro; Suzuki, Masahide; Kawamura, Hiroshi [Japan Atomic Energy Agency, Oarai Research and Development Center, Oarai, Ibaraki (Japan)

    2012-03-15

    The Japan Materials Testing Reactor (JMTR) in Japan Atomic Energy Agency (JAEA) is a light water cooled tank type reactor with first criticality in March 1968. Owing to the connection between the JMTR and hot laboratory by a canal, easy re-irradiation tests can be conducted with safe and quick transportation of irradiated samples. The JMTR has been applied to fuel/material irradiation examinations for LWRs, HTGR, fusion reactor and RI production. However, the JMTR operation was once stopped in August 2006, and check and review on the reoperation had been conducted by internal as well as external committees. As a result of the discussion, the JMTR reoperation was determined, and refurbishment works started from the beginning of JFY 2007. The refurbishment works have finished in March 2011 taking four years from JFY 2007. Unfortunately, at the end of the JFY 2010 on March 11, the Great-Eastern-Japan-Earthquake occurred, and functional tests before the JMTR restart, such as cooling system, reactor control system and so on, were delayed by the earthquake. Moreover, a detail inspection found some damages such as slight deformation of the truss structure at the roof of the JMTR reactor building. Consequently, the restart of the JMTR will be delayed from June to next October, 2012. Now, the safety evaluation after the earthquake disaster is being carried out aiming at the restart of the JMTR. The renewed JMTR will be started from JFY 2012 and operated for a period of about 20 years until around JFY 2030. The usability improvement of the JMTR, e.g. higher reactor availability, shortening turnaround time to get irradiation results, attractive irradiation cost, business confidence, is also discussed with users as the preparations for re-operation. (author)

  13. Present status of Japan materials testing reactor

    International Nuclear Information System (INIS)

    Hori, Naohiko; Kaminaga, Masanori; Kusunoki, Tsuyoshi; Ishihara, Masahiro; Niimi, Motoji; Komori, Yoshihiro; Suzuki, Masahide; Kawamura, Hiroshi

    2012-01-01

    The Japan Materials Testing Reactor (JMTR) in Japan Atomic Energy Agency (JAEA) is a light water cooled tank type reactor with first criticality in March 1968. Owing to the connection between the JMTR and hot laboratory by a canal, easy re-irradiation tests can be conducted with safe and quick transportation of irradiated samples. The JMTR has been applied to fuel/material irradiation examinations for LWRs, HTGR, fusion reactor and RI production. However, the JMTR operation was once stopped in August 2006, and check and review on the reoperation had been conducted by internal as well as external committees. As a result of the discussion, the JMTR reoperation was determined, and refurbishment works started from the beginning of JFY 2007. The refurbishment works have finished in March 2011 taking four years from JFY 2007. Unfortunately, at the end of the JFY 2010 on March 11, the Great-Eastern-Japan-Earthquake occurred, and functional tests before the JMTR restart, such as cooling system, reactor control system and so on, were delayed by the earthquake. Moreover, a detail inspection found some damages such as slight deformation of the truss structure at the roof of the JMTR reactor building. Consequently, the restart of the JMTR will be delayed from June to next October, 2012. Now, the safety evaluation after the earthquake disaster is being carried out aiming at the restart of the JMTR. The renewed JMTR will be started from JFY 2012 and operated for a period of about 20 years until around JFY 2030. The usability improvement of the JMTR, e.g. higher reactor availability, shortening turnaround time to get irradiation results, attractive irradiation cost, business confidence, is also discussed with users as the preparations for re-operation. (author)

  14. Safety report content and development for test loop facility on MARIA reactor

    International Nuclear Information System (INIS)

    Konechko, A.; Shumskij, A.M.; Mikul'ahin, V.E.

    1982-01-01

    A 600 kW test loop facility for investigatin.o safety problems is realized on MARIA reactor in Poland together with USSR organizations. Safety reports have been developed in two steps at the designstage. The 1st report being essentially a preliminary safety analysis was developed within the scope of the feasibility study. At the engineering design stage the preliminary test loop facility safety report had been prepared considering measures excluding the possibility of the MARIA reactor damage. The test loop facility safety report is fulfilled for normal, transient and emergency operation regimes. Separate safety basing for each group of experiments will be prepared. The report presents the test loop facility safety criteria coordinated by the nuclear safety comission. They contains the preliminary reports on the test loop facility safety. At the final stage of construction and at thecommitioning stage the start-up safety report will be developed which after required correction and adding up the putting into operation data will turn into operation safety report [ru

  15. Nuclear fuels for material test reactors

    International Nuclear Information System (INIS)

    Ramanathan, L.V.; Durazzo, M.; Freitas, C.T. de

    1982-01-01

    Experimental results related do the development of nuclear fuels for reactors cooled and moderated by water have been presented cylindrical and plate type fuels have been described in which the core consists of U compouns dispersed in an Al matrix and is clad with aluminium. Fabrication details involving rollmilling, swaging or hot pressing have been described. Corrosion and irradiation test results are also discussed. The performance of the different types of fuels indicates that it is possible to locally fabricate fuel plates with U 3 O 8 +Al cores (20% enriched U) for use in operating Brazilian research reactors. (Author) [pt

  16. The behavior of fission products during nuclear rocket reactor tests

    International Nuclear Information System (INIS)

    Bokor, P.C.; Kirk, W.L.; Bohl, R.J.

    1991-01-01

    Fission product release from nuclear rocket propulsion reactor fuel is an important consideration for nuclear rocket development and application. Fission product data from the last six reactors of the Rover program are collected in this paper to provide as basis for addressing development and testing issues. Fission product loss from the fuel will depend on fuel composition and reactor design and operating parameters. During ground testing, fission products can be contained downstream of the reactor. The last Rover reactor tested, the Nuclear Furnance, was mated to an effluent clean-up system that was effective in preventing the discharge of fission products into the atmosphere

  17. Utilization of fusion neutrons in the tokamak fusion test reactor for blanket performance testing and other nuclear engineering experiments

    International Nuclear Information System (INIS)

    Caldwell, C.S.; Pettus, W.G.; Schmotzer, J.K.; Welfare, F.; Womack, R.

    1979-01-01

    In addition to developing a set of reacting-plasma/blanket-neutronics benchmark data, the TFTR fusion application experiments would provide operational experience with fast-neutron dosimetry and the remote handling of blanket modules in a tokamak reactor environment; neutron streaming and hot-spot information invaluable for the optimal design of penetrations in future fusion reactors; and the identification of the most damage-resistant insulators for a variety of fusion-reactor components

  18. Characterization and testing of materials for nuclear reactors. Proceedings of a technical meeting

    International Nuclear Information System (INIS)

    2007-03-01

    Nuclear techniques in general and neutrons based methods in particular have played and will continue to play an important role in research in materials science and technology. Today the world is looking at nuclear fission and nuclear fusion as the main sources of energy supply for the future. Research reactors have played a key role in the development of nuclear technology. A materials development programme will thus play a major role in the design and development of new nuclear power plants, for the extension of the life of operating reactors as well as for fusion reactors. Against this background, the IAEA had organized a Technical Meeting on Development, Characterization and Testing of Materials - With Special Reference to the Energy Sector under the activity on specific applications of research reactors. The meeting was held in Vienna, May 29- June 2, 2006. There was also participation by experts in techniques, complementary to neutrons. The participants for the technical meeting were experts in the utilization of nuclear techniques namely the high flux and medium flux research reactors, fusion research and positron annihilation. They presented the design, development and utilization of the facilities at their respective centres for materials characterization with main focus on materials for nuclear energy, both fission and fusion. In core irradiation of materials, development of instrument for residual stress measurement in large and / or irradiated specimen, neutron radiography for inspection of irradiated fuel, work on oxide dispersion strengthened (ODS) steels and SiC composites, relevant to future power systems were cited as application of nuclear techniques in fission reactors. The use of neutron scattering for helium bubbles in steel, application of positron annihilation to study helium bubbles in Cu, Ti-stabilized stainless steel and voidswelling studies etc. show that these techniques have an important role in the development of materials for energy

  19. Testing and analyses of a high temperature thermal barrier for gas-cooled reactors

    International Nuclear Information System (INIS)

    Black, W.E.; Betts, W.S.; Felten, P.

    1979-01-01

    A full size, multi-panel section of a thermal barrier system was fabricated from a nickel-base superalloy and a combination of fibrous blanket insulation materials for specific application in a steam cycle gas-cooled nuclear reactor. The 2.4 m square array was representative of the sidewall of the lower core outlet plenum and included coverplates, attachments, seals, and a simulated water-cooled liner. Testing was conducted in a reactor grade, helium-filled chamber at 816 0 C for 100 hours, which established a normal (baseline) condition; 982 0 C for 10 hours, which satisfied an emergency condition; 1093 0 C for 1 hour, which simulated a faulted condition; and 1260 0 C, which was a non-design condition test to demonstrate the temperature overshoot capability of the system. Post-test examination indicated: (1) an acceptable performance by the anti-friction chromium carbide (Cr 3 C 2 ) coating; (2) no significant galling between non-coated surfaces; (3) no distortion of attachment fixtures; (4) predictable coverplate deflection during the design conditions testing (normal, emergency, and faulted); and (5) considerable plastic deformation resulting from the near-incipient melting temperature. (orig.)

  20. IMPROVED COMPUTATIONAL NEUTRONICS METHODS AND VALIDATION PROTOCOLS FOR THE ADVANCED TEST REACTOR

    Energy Technology Data Exchange (ETDEWEB)

    David W. Nigg; Joseph W. Nielsen; Benjamin M. Chase; Ronnie K. Murray; Kevin A. Steuhm

    2012-04-01

    The Idaho National Laboratory (INL) is in the process of modernizing the various reactor physics modeling and simulation tools used to support operation and safety assurance of the Advanced Test Reactor (ATR). Key accomplishments so far have encompassed both computational as well as experimental work. A new suite of stochastic and deterministic transport theory based reactor physics codes and their supporting nuclear data libraries (HELIOS, KENO6/SCALE, NEWT/SCALE, ATTILA, and an extended implementation of MCNP5) has been installed at the INL. Corresponding models of the ATR and ATRC are now operational with all five codes, demonstrating the basic feasibility of the new code packages for their intended purpose. Of particular importance, a set of as-run core depletion HELIOS calculations for all ATR cycles since August 2009 was successfully completed during 2011. This demonstration supported a decision late in the year to proceed with the phased incorporation of the HELIOS methodology into the ATR fuel cycle management process beginning in 2012. On the experimental side of the project, new hardware was fabricated, measurement protocols were finalized, and the first four of six planned physics code validation experiments based on neutron activation spectrometry were conducted at the ATRC facility. Data analysis for the first three experiments, focused on characterization of the neutron spectrum in one of the ATR flux traps, has been completed. The six experiments will ultimately form the basis for a flexible, easily-repeatable ATR physics code validation protocol that is consistent with applicable ASTM standards.

  1. Organic geochemistry and petrology of subsurface Paleocene-Eocene Wilcox and Claiborne Group coal beds, Zavala County, Maverick Basin, Texas, USA

    Science.gov (United States)

    Hackley, Paul C.; Warwick, Peter D.; Hook, Robert W.; Alimi, Hossein; Mastalerz, Maria; Swanson, Sharon M.

    2012-01-01

    Coal samples from a coalbed methane exploration well in northern Zavala County, Maverick Basin, Texas, were characterized through an integrated analytical program. The well was drilled in February, 2006 and shut in after coal core desorption indicated negligible gas content. Cuttings samples from two levels in the Eocene Claiborne Group were evaluated by way of petrographic techniques and Rock–Eval pyrolysis. Core samples from the Paleocene–Eocene Indio Formation (Wilcox Group) were characterized via proximate–ultimate analysis in addition to petrography and pyrolysis. Two Indio Formation coal samples were selected for detailed evaluation via gas chromatography, and Fourier transform infrared (FTIR) and 13C CPMAS NMR spectroscopy. Samples are subbituminous rank as determined from multiple thermal maturity parameters. Elevated rank (relative to similar age coal beds elsewhere in the Gulf Coast Basin) in the study area is interpreted to be a result of stratigraphic and/or structural thickening related to Laramide compression and construction of the Sierra Madre Oriental to the southwest. Vitrinite reflectance data, along with extant data, suggest the presence of an erosional unconformity or change in regional heat flow between the Cretaceous and Tertiary sections and erosion of up to >5 km over the Cretaceous. The presence of liptinite-rich coals in the Claiborne at the well site may indicate moderately persistent or recurring coal-forming paleoenvironments, interpreted as perennially submerged peat in shallow ephemeral lakes with herbaceous and/or flotant vegetation. However, significant continuity of individual Eocene coal beds in the subsurface is not suggested. Indio Formation coal samples contain abundant telovitrinite interpreted to be preserved from arborescent, above-ground woody vegetation that developed during the middle portion of mire development in forested swamps. Other petrographic criteria suggest enhanced biological, chemical and physical

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-01-26

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

  3. Comparison of diffusion and transport theory analysis with experimental results in fast breeder test reactor

    International Nuclear Information System (INIS)

    Sathyabama, N.; Mohanakrishnan, P.; Lee, S.M.

    1994-01-01

    A systematic analysis has been performed by 3 dimensional diffusion and transport methods to calculate the measured control rod worths and subassembly wise power distribution in fast breeder test reactor. Geometry corrections (rectangular to hexagonal and diffusion to transport corrections are estimated for multiplication factors and control rod worths. Calculated control rod worths by diffusion and transport theory are nearly the same and 10% above measured values. Power distribution in the core periphery is over predicted (15%) by diffusion theory. But, this over prediction reduces to 8% by use of the S N method. (authors). 9 refs., 4 tabs., 3 fig

  4. Packaging and transport case of test fuel assembly irradiated in the Creys-Malville reactor

    International Nuclear Information System (INIS)

    Geffroy, J.; Vivien, J.; Pouard, M.; Dujardin, G.N.; Veron, B.; Michoux, H.

    1986-06-01

    Some irradiated fuel assemblies from the fast neutron Creys Malville reactor will be sent to hot laboratories to follow fuel behavior. These test assemblies will be examined after a limited cooling time and transport is realized at high residual power (about 10kW) and cladding temperature should not rise over 500deg C. The fuel assemblies are not dismantled and transported into sodium. The assembly is placed into a case containing sodium plugged and put into a packaging. Dimensioning, thermal behavior, radiation protection and containment are examined [fr

  5. Analyzing and comparing the dynamic response of test reactor main workshop

    International Nuclear Information System (INIS)

    Wang Jiachun; Fu Jiyang; Cai Laizhong

    2001-01-01

    Analyzing soil-structure interaction is an important section in anti-seismic design and analysis of nuclear engineering. The factors that influence on the response of nuclear structures include the properties of earthquake, soil and structures. So the soil-structure interaction in the non-rock foundation is different from that in the surface free field. And the interaction must be considered under the anti-seismic design standard of test reactors. The FLUSH program and SASSI2000 are applied to dynamic analysis. Moreover, comparing the obtained data and diagrams draws some conclusions

  6. Nuclear electronic equipment for control and monitoring panel. Procedure guide for on-site tests of nuclear reactor instruments

    International Nuclear Information System (INIS)

    1975-10-01

    By the use of a procedure for on-site testing of nuclear reactor instruments it should be possible to judge their ability to guarantee the reactor safety and availability at the moment of divergence or during operation. Such a procedure must therefore be created as a work implement for the quick and reliable installation of electronic devices necessary for nuclear reactor control and supervision. A standard document is proposed for this purpose, allowing a ''test programme'' to be set up before the equipment is installed on the site [fr

  7. Progress of the RERTR [Reduced Enrichment Research and Test Reactor] Program in 1989

    International Nuclear Information System (INIS)

    Travelli, A.

    1989-01-01

    The progress of the Reduced Enrichment Research and Test Reactor (RERTR) Program is described. After a brief summary of the results which the RERTR Program, in collaboration with its many international partners, had achieved by the end of 1988, the major events, findings, and activities of 1989 are reviewed. The scope of the RERTR Program activities was curtailed, in 1989, by an unexpected legislative restriction which limited the ability of the Arms Control and Disarmament Agency to adequately fund the program. Nevertheless, the thrust of the major planned program activities was maintained, and meaningful results were obtained in several areas of great significance for future work. 15 refs., 12 figs

  8. Calorific energy deposited by gamma radiations in a test reactor. Calorimetric measurements and calculations

    International Nuclear Information System (INIS)

    Mecheri, K.-F.

    1977-01-01

    The purpose of this work was to determine the calorific energy deposited by gamma radiations in the experimental devices irradiated in the test reactors of the Grenoble Nuclear Study Centre. A theoretical study briefly recalls to mind the various sorts of nuclear reactions that occur in a reactor, from the special angle of their ability to deposit calorific energy in the materials. A special study with the help of a graphite calorimeter made it possible to show the possible effect of the various parameters intervening in this energy absorption: the nature of the materials, their geometry, the spectrum of the incident gamma rays and the fact that the variation of this spectrum is due to the position of the measuring point with respect to the reactor core or to the presence of structures around the measuring instrument. The results of the calculations made with the help of the Mercury IV and ANISN codes are compared with those of the determinations in order to ascertain that very are adapted to the forecasts of energy deposition in the various materials. The conclusion was reached that in order to calculate with accuracy the depositifs of gamma energy in the experimental devices, it is necessary either to introduce the build-up calculation for the low energy photons, in the Mercury IV calculation code or to associate the DOT code to the ANISN calculation code [fr

  9. Development of an aging evaluation and life extension program for the Advanced Test Reactor

    International Nuclear Information System (INIS)

    Dwight, J.E. Jr.

    1988-01-01

    A life extension program has been developed for the US Department of Energy's Advanced Test Reactor. The program is an adaptation of life extension pilot programs at the Surry Unit 1 and Monticello generating stations and is being completed in three phases. In Phase 1, the critical plant components were identified. In Phase 2, existing lifetime analyses and support data for the critical components were reviewed. The results from the review give a preliminary indication that an overall plant lifetime in excess of forty years is feasible. In Phase 3, now in progress, detailed evaluations for component life extensions are being performed. 2 refs., 2 figs., 1 tab

  10. INEL test reactor facility alarms: descriptions, technical specifications, and modification procedure

    International Nuclear Information System (INIS)

    Potash, L.M.; Boone, M.P.

    1980-04-01

    This report identifies standards, procedures, and practices which will affect any attempt to integrate or introduce human engineering principles into nuclear power plant alarm systems. Additional information concerning type of signal used, expected reaction, type of sensor, etc., is presented because of its relevance to future work on alarm system integration. The INEL test reactors were studied. Interviews were conducted with operators, designers, and management personnel. Additional information was obtained from available documentation. Only fire-alarm systems, and to a lesser extent, criticality alarms, have detailed industry-wide standards. One general standard has been written for control-room annunciators

  11. Main boiler feed pump for fast breeder test reactor. Failure analysis and remedial measures

    International Nuclear Information System (INIS)

    Iyer, M.A.K.; Chande, S.K.; Raghuvir, A.D.; Baskar, S.; Kale, R.D.

    1994-01-01

    A small capacity ten stage 670 kw feed water pump is used for supplying feed water at a temperature of 190 deg C to a once through steam generator in the Fast Breeder Test Reactor at Kalpakkam. During preparatory heating up stage to commission the steam generator the pump suffered a severe loss of suction which resulted in failure of hydrostatic journal bearings and extensive damage to pump internals. This paper discusses the detailed mechanism of loss of suction, details of damage to the pump and various modifications carried out to prevent recurrence of the problem. (author). 4 refs., 3 figs., 2 tabs

  12. Integrity assessment of research reactor fuel cladding and material testing using eddy current inspection

    International Nuclear Information System (INIS)

    Alencar, Donizete Anderson de

    2004-01-01

    A methodology to perform the integrity assessment of research reactors nuclear fuels cladding, such as those installed in IPR-Rl (TRIGA) and IEA-R1 (MTR), using nondestructive electromagnetic inspection (eddy current) is presented. This methodology is constituted by: the development of calibration reference standards, specific for each type of fuel; the development of special test probes; the recommendations for the inspection equipment calibration; the construction of voltage based evaluation curves and the inspection procedures developed for the characterization of detected flaws. The test probes development, specially those designed for the inspection of MTR fuels cladding, which present access difficulties due to the narrow gap between fuel plates (2,89 mm for IEAR-R1), constituted a challenge that demanded the introduction of unusual materials and constructive techniques. The operational performance of the developed resources, as well as the special operative characteristics of the test probes, such as their immunity to adjacent fuel plates interference and electrical resistivity changes of the fuels meat are experimentally demonstrated. The practical applicability of the developed methodology is verified in non radioactive environment, using a dummy MTR fuel element model, similar to an IEA-R1 reactor fuel element, produced and installed in IPEN, Sao Paulo. The efficacy of the proposed methodology was verified by the achieved results. (author)

  13. Proposal of world network on material testing reactors

    International Nuclear Information System (INIS)

    Takemoto, Noriyuki; Izumo, Hironobu; Hori, Naohiko; Ishitsuka, Etsuo; Ishihara, Masahiro

    2011-01-01

    Establishment of an international cooperation system of worldwide testing reactor network (world network) is proposed in order to achieve efficient facility utilization and provide high quality irradiation data by role sharing of irradiation tests with materials testing reactors in the world. As for the first step, mutual understanding among materials testing reactors is thought to be necessary. From this point, an international symposium on materials testing reactors (ISMTR) was held to construct the world network from 2008, and a common understanding of world network has begun to be shared. (author)

  14. Experimental tests and calculation methods for missile crashing effects on a reactor containment

    International Nuclear Information System (INIS)

    Goldstein, S.; Berriaud, C.; Labrot, R.

    1975-01-01

    In the analysis of missile crashing on a reactor containment there are two main effects to be taken into account: the overall behaviour of the building; the local perforation. The overall behaviour of the building is easily calculated when the applied force as a function of time is known. Two calculation examples are presented. The local perforation is a much more difficult problem and experimental work is necessary. The report presents a series of perforation tests of concrete plates by cylindrical missiles with a flat nose. The aim of these tests is to extrapolate for the lower speeds the existing experimental correlations and to check the calculation methods. The calculations are made with the PASTEL code (Finite elements, implicit integration), with elastoplasticity of the reinforcing steel bars and the concrete. Various plastification and fracturation laws are tested. (Auth.)

  15. Experimental tests and calculation methods for missile crashing effects on a reactor containment

    International Nuclear Information System (INIS)

    Goldstein, S.; Berriaud, C.

    1975-01-01

    In the analysis of missile crashing on a reactor containment there are two main effects to be taken into account: the overall behavior of the building; the local perforation. The overall behavior of the building is easily calculated when the applied force as a function of time is known. Two calculation examples are presented. The local perforation is a much more difficult problem and experimental work is necessary. The report presents a series of perforation tests of concrete plates by cylindrical missiles with a flat nose. The aim of these tests is to extrapolate for the lower speeds the existing experimental correlations (Petry, HN-NDRC, BRL...) and to check the calculation methods. The calculations are made with the PASTEL Code (Finite elements, implicit integration), with elastoplasticity of the reinforcing steel bars and the concrete. Various plastification and fracturation laws will be tested

  16. Review of WWER fuel and material tests in the Halden reactor

    International Nuclear Information System (INIS)

    Volkov, B.; Kolstad, E.

    2006-01-01

    A review of the tests with WWER fuels and materials conducted in HBWR over the years of cooperation with Russia is presented. The first test with old generation WWER-440 fuel and PWR specification fuel was carried out from 1995 to 1998. Some differences between these fuels regarding irradiation induced densification and pellet design as well as similar fuel thermal behaviour, swelling and FGR were revealed during the test. The data from this test are reviewed and compared with PIE recently performed to confirm the in-pile measurements. The second test was started in March 1999 with the main objective to study different modified WWER fuels also in comparison with PWR fuel. The results indicated that all these modified WWER fuels exhibit improved densification properties relative to earlier tested fuel. In-pile data on fuel densification have been analysed with respect to as fabricated fuel microstructure and can be used for verification of fuel behaviour models. Corrosion and creep tests in the Halden reactor encompass WWER cladding alloys and some results are given. Prospective WWER fuel and material tests foreseen within the frame of the joint program of OECD HRP are also presented. (authors)

  17. JENDL-3.3 thermal reactor benchmark test

    International Nuclear Information System (INIS)

    Akie, Hiroshi

    2001-01-01

    Integral tests of JENDL-3.2 nuclear data library have been carried out by Reactor Integral Test WG of Japanese Nuclear Data Committee. The most important problem in the thermal reactor benchmark testing was the overestimation of the multiplication factor of the U fueled cores. With several revisions of the data of 235 U and the other nuclides, JENDL-3.3 data library gives a good estimation of multiplication factors both for U and Pu fueled thermal reactors. (author)

  18. Automated testing of reactor protection instrumentation made easy

    International Nuclear Information System (INIS)

    Iborra, A.; De Marcos, F.; Pastor, J.A.; Alvarez, B.; Jimenez, A.; Mesa, E.; Alsonso, L.; Regidor, J.J.

    1997-01-01

    Maintenance and testing of reactor protection systems is an important cause of unplanned reactor trips. Automated testing is the answer because it minimises test times and reduces human error. The GAMA I system, developed and implemented at Vandellos II in Spain, has the added advantage that it uses visual programming, which means that changing the software does not need specialist programming skills. (author)

  19. Nuclear reactor pressure vessel surveillance capsule examinations. Application of American Society for Testing and Materials Standards

    International Nuclear Information System (INIS)

    Perrin, J.S.

    1978-01-01

    A series of pressure vessel surveillance capsules is installed in each commercial nuclear power plant in the United States. A capsule typically contains neutron dose meters, thermal monitors, tensile specimens, and Charpy V-notch impact specimens. In order to determine property changes of the pressure vessel resulting from irradiation, surveillance capsules are periodically removed during the life of a reactor and examined. There are numerous standards, regulations, and codes governing US pressure vessel surveillance capsule programmes. These are put out by the US Nuclear Regulatory Commission, the Boiler and Pressure Vessel Committee of the American Society of Mechanical Engineers, and the American Society for Testing and Materials (ASTM). A majority of the pertinent ASTM standards are under the jurisdiction of ASTM Committee E-10 on Nuclear Applications and Measurements of Radiation Effects. The standards, regulations, and codes pertaining to pressure vessel surveillance play an important role in ensuring reliability of the nuclear pressure vessels. ASTM E 185-73 is the Standard Recommended Practice for Surveillance Tests for Nuclear Reactors. This standard recommends procedures for both the irradiation and subsequent testing of surveillance capsules. ASTM E 185-73 references many additional specialized ASTM standards to be followed in specific areas of a surveillance capsule examination. A key element of surveillance capsule programmes is the Charpy V-notch impact test, used to define curves of fracture behaviour over a range of temperatures. The data from these tests are used to define the adjusted reference temperature used in determining pressure-temperature operating curves for a nuclear power plant. (author)

  20. Unusual occurrences in fast breeder test reactor

    International Nuclear Information System (INIS)

    Kapoor, R.P.; Srinivasan, G.; Ellappan, T.R.; Ramalingam, P.V.; Vasudevan, A.T.; Iyer, M.A.K.; Lee, S.M.; Bhoje, S.B.

    2000-01-01

    Fast Breeder Test Reactor (FBTR) is a 40 MWt/13.2 MWe sodium cooled mixed carbide fuelled reactor. Its main aim is to generate experience in the design, construction and operation of fast reactors including sodium systems and to serve as an irradiation facility for the development of fuel and structural materials for future fast reactors. It achieved first criticality in Oct 85 with Mark I core (70% PuC - 30% UC). Steam generator was put in service in Jan 93 and power was raised to 10.5 MWt in Dec 93. Turbine generator was synchronised to the grid in Jul 97. The indigenously developed mixed carbide fuel has achieved a burnup of 44,000 MW-d/t max at a linear heat rating of 320 W/cm max without any fuel clad failure. The commissioning and operation of sodium systems and components have been smooth and performance of major components, viz., sodium pumps, intermediate heat exchangers and once through sodium heated steam generators (SG) have been excellent. There have been three minor incidents of Na/NaK leaks during the past 14 years, which are described in the paper. There have been no incident of a tube leak in SG. However, three incidents of water leaks from water / steam headers have been detailed. The plant has encountered some unusual occurrences, which were critically analysed and remedial measures, in terms of system and procedural modifications, incorporated to prevent recurrence. This paper describes unusual occurrences of fuel handling incident of May 1987, main boiler feed pump seizure in Apr 1992, reactivity transients in Nov 1994 and Apr 1995, and malfunctioning of the core cover plate mechanism in Jul 1995. These incidents have resulted in long plant shutdowns. During the course of investigation, various theoretical and experimental studies were carried out for better understanding of the phenomena and several inspection techniques and tools were developed resulting in enriching the technology of sodium cooled reactors. FBTR has 36 neutronic and process

  1. Creep-Fatigue Damage Evaluation of a Model Reactor Vessel and Reactor Internals of Sodium Test Facility according to ASME-NH and RCC-MRx Codes

    International Nuclear Information System (INIS)

    Lim, Dong-Won; Lee, Hyeong-Yeon; Eoh, Jae-Hyuk; Son, Seok-Kwon; Kim, Jong-Bum; Jeong, Ji-Young

    2016-01-01

    The objective of the STELLA-2 is to support the specific design approval for PGSFR by synthetic reviews of key safety issues and code validations through the integral effect tests. Due to its high temperature operation in SFRs (and in a testing facility) up to 550 °C, thermally induced creep-fatigue damage is very likely in components including a reactor vessel, reactor internals (interior structures), heat exchangers, pipelines, etc. In this study, structural integrity of the components such as reactor vessel and internals in STELLA-2 has been evaluated against creep-fatigue failures at a concept-design step. As 2D analysis yields far conservative results, a realistic 3D simulation is performed by a commercial software. A design integrity guarding against a creep-fatigue damage failure operating at high temperature was evaluated for the reactor vessel with its internal structure of the STELLA-2. Both the high temperature design codes were used for the evaluation, and results were compared. All the results showed the vessel as a whole is safely designed at the given operating conditions, while the ASME-NH gives a conservative evaluation

  2. Creep-Fatigue Damage Evaluation of a Model Reactor Vessel and Reactor Internals of Sodium Test Facility according to ASME-NH and RCC-MRx Codes

    Energy Technology Data Exchange (ETDEWEB)

    Lim, Dong-Won; Lee, Hyeong-Yeon; Eoh, Jae-Hyuk; Son, Seok-Kwon; Kim, Jong-Bum; Jeong, Ji-Young [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    The objective of the STELLA-2 is to support the specific design approval for PGSFR by synthetic reviews of key safety issues and code validations through the integral effect tests. Due to its high temperature operation in SFRs (and in a testing facility) up to 550 °C, thermally induced creep-fatigue damage is very likely in components including a reactor vessel, reactor internals (interior structures), heat exchangers, pipelines, etc. In this study, structural integrity of the components such as reactor vessel and internals in STELLA-2 has been evaluated against creep-fatigue failures at a concept-design step. As 2D analysis yields far conservative results, a realistic 3D simulation is performed by a commercial software. A design integrity guarding against a creep-fatigue damage failure operating at high temperature was evaluated for the reactor vessel with its internal structure of the STELLA-2. Both the high temperature design codes were used for the evaluation, and results were compared. All the results showed the vessel as a whole is safely designed at the given operating conditions, while the ASME-NH gives a conservative evaluation.

  3. Testing and analyses of a high temperature duct for gas-cooled reactors

    International Nuclear Information System (INIS)

    Black, W.E.; Roberge, A.; Felten, P.; Bastien, D.

    1979-01-01

    A 0.6 scale model of a steam cycle gas-cooled reactor high temperature duct was tested in a closed loop helium facility. The object of the test series was to determine: 1) the thermal effects of gas permeation within the thermal barrier, 2) the plastic deformation of the metallic components, and 3) the thermal performance of the fibrous insulation. A series of tests was performed with thermal cyclings from 100 0 C to 760 0 C at 50 atmospheres until the system thermal performance had stabilized hence enabling predictions for the reactor life. Additional tests were made to assess permeation by deliberately simulating sealing weld failures thereby allowing gas flow by-pass within the primary thermal barrier. After 100 cycles the entire primary structure was found to have performed without structural failure. Due to high pressures exerted by the insulation on the cover plates and a design oversight, the thin seal sheets were unable to expand in an anticipated manner. Local buckling resulted. The insulation retained an acceptable degree of resiliency. However, some fiber damage was observed within both the high and low temperature insulation blankets. A thermal analysis was conducted to correlate the hot duct heat transfer results with those obtained from the analytical techniques used for the HTGR design using a computer thermal model representative of the duct and test setup. The thermal performance of the insulation, the temperature gradient through the structural components, the heating load to the cooling system and the permeation flow effect on heat transfer were verified. Exellent correlation between the experimental data and the analytical techniques were obtained

  4. Laser-Ultrasonic Testing and its Applications to Nuclear Reactor Internals

    Science.gov (United States)

    Ochiai, M.; Miura, T.; Yamamoto, S.

    2008-02-01

    A new nondestructive testing technique for surface-breaking microcracks in nuclear reactor components based on laser-ultrasonics is developed. Surface acoustic wave generated by Q-switched Nd:YAG laser and detected by frequency-stabilized long pulse laser coupled with confocal Fabry-Perot interferometer is used to detect and size the cracks. A frequency-domain signal processing is developed to realize accurate sizing capability. The laser-ultrasonic testing allows the detection of surface-breaking microcrack having a depth of less than 0.1 mm, and the measurement of their depth with an accuracy of 0.2 mm when the depth exceeds 0.5 mm including stress corrosion cracking. The laser-ultrasonic testing system combined with laser peening system, which is another laser-based maintenance technology to improve surface stress, for inner surface of small diameter tube is developed. The generation laser in the laser-ultrasonic testing system can be identical to the laser source of the laser peening. As an example operation of the system, the system firstly works as the laser-ultrasonic testing mode and tests the inner surface of the tube. If no cracks are detected, the system then changes its work mode to the laser peening and improves surface stress to prevent crack initiation. The first nuclear industrial application of the laser-ultrasonic testing system combined with the laser peening was completed in Japanese nuclear power plant in December 2004.

  5. Proceedings of the 1994 international meeting on reduced enrichment for research and test reactors

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-08-01

    This meeting brought together participants in the international effort to minimize and eventually eliminate the use of highly enriched uranium in civilian nuclear programs. Papers cover the following topics: National programs; fuel cycle; nuclear fuels; analyses; advanced reactors; and reactor conversions. Selected papers have been indexed separately for inclusion to the Energy Science and Technology Database.

  6. Proceedings of the 1994 international meeting on reduced enrichment for research and test reactors

    International Nuclear Information System (INIS)

    1997-08-01

    This meeting brought together participants in the international effort to minimize and eventually eliminate the use of highly enriched uranium in civilian nuclear programs. Papers cover the following topics: National programs; fuel cycle; nuclear fuels; analyses; advanced reactors; and reactor conversions. Selected papers have been indexed separately for inclusion to the Energy Science and Technology Database

  7. Naval Reactors Prime Contractor Team (NRPCT) Experiences and Considerations With Irradiation Test Performance in an International Environment

    International Nuclear Information System (INIS)

    MH Lane

    2006-01-01

    This letter forwards a compilation of knowledge gained regarding international interactions and issues associated with Project Prometheus. The following topics are discussed herein: (1) Assessment of international fast reactor capability and availability; (2) Japanese fast reactor (JOYO) contracting strategy; (3) NRPCT/Program Office international contract follow; (4) Completion of the Japan Atomic Energy Agency (JAEA)/Pacific Northwest National Laboratory (PNNL) contract for manufacture of reactor test components; (5) US/Japanese Departmental interactions and required Treaties and Agreements; and (6) Non-technical details--interactions and considerations

  8. Naval Reactors Prime Contractor Team (NRPCT) Experiences and Considerations With Irradiation Test Performance in an International Environment

    Energy Technology Data Exchange (ETDEWEB)

    MH Lane

    2006-02-15

    This letter forwards a compilation of knowledge gained regarding international interactions and issues associated with Project Prometheus. The following topics are discussed herein: (1) Assessment of international fast reactor capability and availability; (2) Japanese fast reactor (JOYO) contracting strategy; (3) NRPCT/Program Office international contract follow; (4) Completion of the Japan Atomic Energy Agency (JAEA)/Pacific Northwest National Laboratory (PNNL) contract for manufacture of reactor test components; (5) US/Japanese Departmental interactions and required Treaties and Agreements; and (6) Non-technical details--interactions and considerations.

  9. Advanced burner test reactor preconceptual design report.

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Y. I.; Finck, P. J.; Grandy, C.; Cahalan, J.; Deitrich, L.; Dunn, F.; Fallin, D.; Farmer, M.; Fanning, T.; Kim, T.; Krajtl, L.; Lomperski, S.; Moisseytsev, A.; Momozaki, Y.; Sienicki, J.; Park, Y.; Tang, Y.; Reed, C.; Tzanos, C; Wiedmeyer, S.; Yang, W.; Chikazawa, Y.; JAEA

    2008-12-16

    The goals of the Global Nuclear Energy Partnership (GNEP) are to expand the use of nuclear energy to meet increasing global energy demand, to address nuclear waste management concerns and to promote non-proliferation. Implementation of the GNEP requires development and demonstration of three major technologies: (1) Light water reactor (LWR) spent fuel separations technologies that will recover transuranics to be recycled for fuel but not separate plutonium from other transuranics, thereby providing proliferation-resistance; (2) Advanced Burner Reactors (ABRs) based on a fast spectrum that transmute the recycled transuranics to produce energy while also reducing the long term radiotoxicity and decay heat loading in the repository; and (3) Fast reactor fuel recycling technologies to recover and refabricate the transuranics for repeated recycling in the fast reactor system. The primary mission of the ABR Program is to demonstrate the transmutation of transuranics recovered from the LWR spent fuel, and hence the benefits of the fuel cycle closure to nuclear waste management. The transmutation, or burning of the transuranics is accomplished by fissioning and this is most effectively done in a fast spectrum. In the thermal spectrum of commercial LWRs, some transuranics capture neutrons and become even heavier transuranics rather than being fissioned. Even with repeated recycling, only about 30% can be transmuted, which is an intrinsic limitation of all thermal spectrum reactors. Only in a fast spectrum can all transuranics be effectively fissioned to eliminate their long-term radiotoxicity and decay heat. The Advanced Burner Test Reactor (ABTR) is the first step in demonstrating the transmutation technologies. It directly supports development of a prototype full-scale Advanced Burner Reactor, which would be followed by commercial deployment of ABRs. The primary objectives of the ABTR are: (1) To demonstrate reactor-based transmutation of transuranics as part of an

  10. Instrumentation to Enhance Advanced Test Reactor Irradiations

    Energy Technology Data Exchange (ETDEWEB)

    J. L. Rempe; D. L. Knudson; K. G. Condie; J. E. Daw; S. C. Taylor

    2009-09-01

    The Department of Energy (DOE) designated the Advanced Test Reactor (ATR) as a National Scientific User Facility (NSUF) in April 2007 to support U.S. leadership in nuclear science and technology. By attracting new research users - universities, laboratories, and industry - the ATR will support basic and applied nuclear research and development, further advancing the nation's energy security needs. A key component of the ATR NSUF effort is to prove new in-pile instrumentation techniques that are capable of providing real-time measurements of key parameters during irradiation. To address this need, an assessment of instrumentation available and under-development at other test reactors has been completed. Based on this review, recommendations are made with respect to what instrumentation is needed at the ATR and a strategy has been developed for obtaining these sensors. Progress toward implementing this strategy is reported in this document. It is anticipated that this report will be updated on an annual basis.

  11. Instrumentation to Enhance Advanced Test Reactor Irradiations

    International Nuclear Information System (INIS)

    Rempe, J.L.; Knudson, D.L.; Condie, K.G.; Daw, J.E.; Taylor, S.C.

    2009-01-01

    The Department of Energy (DOE) designated the Advanced Test Reactor (ATR) as a National Scientific User Facility (NSUF) in April 2007 to support U.S. leadership in nuclear science and technology. By attracting new research users - universities, laboratories, and industry - the ATR will support basic and applied nuclear research and development, further advancing the nation's energy security needs. A key component of the ATR NSUF effort is to prove new in-pile instrumentation techniques that are capable of providing real-time measurements of key parameters during irradiation. To address this need, an assessment of instrumentation available and under-development at other test reactors has been completed. Based on this review, recommendations are made with respect to what instrumentation is needed at the ATR and a strategy has been developed for obtaining these sensors. Progress toward implementing this strategy is reported in this document. It is anticipated that this report will be updated on an annual basis.

  12. The Great Southwest of the Fred Harvey Company and the Santa Fe Railway, edited by Marta Weigle and Barbara A. Babcock. The Heard Museum, Pheonix (printed by The University of Arizona Press, Tucson, for The Heard Museum, 1996

    Directory of Open Access Journals (Sweden)

    Douglas R. Givens

    1997-11-01

    Full Text Available One of the more colorful eras in American Southwestern archaelogy is reflected in The Great Southwest of the Fred Harvey Company and the Santa Fe Railway. Marta Weigle and Barbara A. Babcock, editors of the volume, have done a superb job weaving in early Southwestern archaeological activities with the role of the Fred Harvey Company and the Santa Fe Railway in bring the American Southwest to those "east of the Mississippi River". Many early Southwestern archaeologists made their way throughout the Southwest on the Santa Fe Railway while the "outposts of civilization" that the Fred Harvey Company provided in many railroad stations served as a " bit of home" to the traveler. This book describes the collaboration of both Fred Harvey and the Santa Fe Railroad tourism in the American Southwest and provides an excellent look into the Native American artists and their comumnities which were transformed on a massive scale by the Fred Harvey Company as it bought, sold, and popularized Native American art. Also part of the volume is an excellent discussion of the network of major museums that hold art collections which were purchased through the Harvey Company's Indian Department.

  13. Evaluation of deformation and fracture characteristics of nuclear reactor materials using ball indentation test technique

    International Nuclear Information System (INIS)

    Byun, T. S.; Hong, J. H.; Lee, B. S.; Park, D. G.; Kim, J. H.; Oh, Y. J.; Yoon, J. H.; Chi, S. H.; Kuk, I. H.; Kwon, D. I.; Lee, J. H.

    1998-05-01

    The present report describes the automated ball indentation test techniques and the results of their applications. The ball indentation test technique is an innovative method for evaluating the key mechanical properties from the indentation load-depth data. In the 1st chapter, the existing technique for evaluating basic deformation (tensile) properties is described in detail, and also the application result of the technique is presented. The through-thickness variations of mechanical properties in SA 508 C1.3 reactor pressure vessel steels were measured using an automated ball indentation (ABI) technique. In the 2nd chapter, a method under development, which is similar to that in the 1st chapter, is new method is based on the theoretical solutions rather than experimental relationships. The result of the application showed that the stress-strain curves of various metals were successfully determined with the method. In the 3rd chapter, a new theoretical model was proposed to estimate the fracture toughness of ferritic steels in the transition temperature region. The key concept of the model is that the indention energy to a critical load is related to the fracture energy of the material. The theory was applied to the reactor pressure vessel (RPV) base and weld metals. (author). 24 refs., 3 tabs., 6 figs

  14. Annual report of the Neutron Irradiation and Testing Reactor Center. FY2007. April 1, 2007 - March 31, 2008

    International Nuclear Information System (INIS)

    2009-03-01

    The Japan Materials Testing Reactor (JMTR), achieving first criticality in March 1968, has been used in testing the durability and integrity of reactor fuels and components, basic nuclear research, the production of radioisotopes (RIs), and other purposes. The JMTR, however, stopped in August 2006 after its 165th operation cycle, and is currently under going partial renewal of reactor facilities and installation of new irradiation Facilities, geared toward being restarted in 2011. In addition, to cope with the strong requests from users to improve usability of the JMTR, efforts are being made to increase reactor operation efficiency, shorten the turnaround time for obtaining results, and other necessary tasks for JMTR to commence reoperation. The present report summarizes the activities carried out in 2007 for the refurbishment and restart of JMTR. (author)

  15. Safety of power transformers, power supplies, reactors and similar products - Part 1: General requirements and tests

    CERN Document Server

    International Electrotechnical Commission. Geneva

    1998-01-01

    This International Standard deals with safety aspects of power transformers, power supplies, reactors and similar products such as electrical, thermal and mechanical safety. This standard covers the following types of dry-type transformers, power supplies, including switch mode power supplies, and reactors, the windings of which may be encapsulated or non-encapsulated. It has the status of a group safety publication in accordance with IEC Guide 104.

  16. Review on conformance of JMTR reactor facility to safety design examination guides for water-cooled reactors for test and research

    International Nuclear Information System (INIS)

    Ide, Hiroshi; Naka, Michihiro; Sakuta, Yoshiyuki; Hori, Naohiko; Matsui, Yoshinori; Miyazawa, Masataka

    2009-03-01

    The safety design examination guides for water-cooled reactors for test and research are formulated as fundamental judgements on the basic design validity for licensing from a viewpoint of the safety. Taking the refurbishment opportunity of the JMTR, the conformance of the JMTR reactor facility to current safety design examination guides was reviewed with licensing documents, annexes and related documents. As a result, it was found that licensing documents fully satisfied the requirements of the current guides. Moreover, it was found that the JMTR reactor facility itself also satisfied the guides requirements as well as the safety performance, since the facility with safety function such as structure, systems, devices had been installed based on the licensing documents under the permission by the regulation authority. Important devices for safety have been produced under authorization of regulating authority. Therefore, it was confirmed that the licensing was conformed to guides, and that the JMTR has enough performance. (author)

  17. Vibration tests and analyses of the reactor building model on a small scale

    International Nuclear Information System (INIS)

    Tsuchiya, Hideo; Tanaka, Mitsuru; Ogihara, Yukio; Moriyama, Ken-ichi; Nakayama, Masaaki

    1985-01-01

    The purpose of this paper is to describe the vibration tests and the simulation analyses of the reactor building model on a small scale. The model vibration tests were performed to investigate the vibrational characteristics of the combined super-structure and to verify the computor code based on Dr. H. Tajimi's Thin Layered Element Theory, using the uniaxial shaking table (60 cm x 60 cm). The specimens consist of ground model, three structural model (prestressed concrete containment vessel, inner concrete structure, and enclosure building), a combined structural model and a combined structure-soil interaction model. These models are made of silicon-rubber, and they have a scale of 1:600. Harmonic step by step excitation of 40 gals was performed to investigate the vibrational characteristics for each structural model. The responses of the specimen to harmonic excitation were measured by optical displacement meters, and analyzed by a real time spectrum analyzer. The resonance and phase lag curves of the specimens to the shaking table were obtained respectively. As for the tests of a combined structure-soil interaction model, three predominant frequencies were observed in the resonance curves. These values were in good agreement with the analytical transfer function curves on the computer code. From the vibration tests and the simulation analyses, the silicon-rubber model test is useful for the fundamental study of structural problems. The computer code based on the Thin Element Theory can simulate well the test results. (Kobozono, M.)

  18. Hot cell works and related irradiation tests in fission reactor for development of new materials for nuclear application

    International Nuclear Information System (INIS)

    Shikama, Tatsuo

    1999-01-01

    Present status of research works in Oarai Branch, Institute for Materials Research, Tohoku University, utilizing Japan Materials Testing Reactor and related hot cells will be described.Topics are mainly related with nuclear materials studies, excluding fissile materials, which is mainly aiming for development of materials for advanced nuclear systems such as a nuclear fusion reactor. Conflict between traditional and routined procedures and new demands will be described and future perspective is discussed. (author)

  19. Investigation on cause of malfunction of wide range monitor (WRM) in high temperature engineering test reactor (HTTR). Sample tests and destructive tests

    International Nuclear Information System (INIS)

    Shinohara, Masanori; Saito, Kenji; Haga, Hiroyuki; Sasaki, Shinji; Katsuyama, Kozo; Motegi, Toshihiro; Takada, Kiyoshi; Higashimura, Keisuke; Fujii, Junichi; Ukai, Takayuki; Moriguchi, Yusuke

    2012-11-01

    An event, in which one of WRMs were disabled to detect the neutron flux in the reactor core, occurred during the period of reactor shut down of HTTR in March, 2010. The actual life time of WRM was unexpectedly shorter than the past developed life time. Investigation of the cause of the outage of WRM toward the recovery of the life time up to the past developed life is one of the issues to develop the technology basis of High Temperature Gas cooled Reactor (HTGR). Then, two experimental investigations were carried out to reveal the cause of the malfunction by specifying the damaged part causing the event in the WRM. One is an experiment using a mock-up sample test which strength degradation at assembly process and heat cycle to specify the damaged part in the WRM. The other is a destructive test in Fuels Monitoring Facility (FMF) to specify the damaged part in the WRM. This report summarized the results of the destructive test and the experimental investigation using the mock-up to reveal the cause of malfunction of WRM. (author)

  20. Design and fabrication report on capsule (11M 19K for out of pile test) for irradiation testing of research reactor materials at HANARO

    Energy Technology Data Exchange (ETDEWEB)

    Kim, B.G.; Yang, S.W.; Park, S.J.; Shim, K.T.; Choo, K.N.; Oh, J.M.; Lee, B.C.; Choi, M.H.; Kim, D.J.; Kim, J.M.; Kang, S.H.; Chun, Y.B.; Kim, T.K.; Jeong, Y.H.

    2012-05-15

    As a part of the research reactor development project with a plate type fuel, the irradiation tests of graphite (Gr), beryllium (Be), and zircaloy 4 materials using the capsule have been investigating to obtain the mechanical characteristics such as an irradiation growth, hardness, swelling and tensile strength at the temperature below 100 .deg. C and the 30 MW reactor power. Then, A capsule to be able to irradiate materials(graphite, Be, zircaloy 4) under 100 .deg. C at the HANARO was designed and fabricated. After performing out of pile testing in single channel test loop by using the capsule, the final design of the capsules to be irradiated in CT and IR2 test hole of HANARO was approved, and 2 sets of capsule were fabricated. These capsules will be loaded in CT and IR2 test hole of HANARO, and be started the irradiation from the end of June, 2012. After performing the irradiation testing of 2 sets of capsule, PIE (Post Irradiation Examination) on irradiated specimens (Gr, Be, and zircaloy 4) will be carry out in IMEF (Irradiated Material Examination Facility). So, the irradiation testing will be contributed to obtain the characteristic data induced neutron irradiation on Gr, Be, and zircaloy 4. And then, it is convinced that these data will be also contributed to obtain the license for JRTR (Jordan Research and Training Reactor) and new research reactor in Korea, and export research reactors.

  1. Imperfection detection probability at ultrasonic testing of reactor vessels

    International Nuclear Information System (INIS)

    Kazinczy, F. de; Koernvik, L.Aa.

    1980-02-01

    The report is a lecture given at a symposium organized by the Swedish nuclear power inspectorate on February 1980. Equipments, calibration and testing procedures are reported. The estimation of defect detection probability for ultrasonic tests and the reliability of literature data are discussed. Practical testing of reactor vessels and welded joints are described. Swedish test procedures are compared with other countries. Series of test data for welded joints of the OKG-2 reactor are presented. Future recommendations for testing procedures are made. (GBn)

  2. Design of Fire/Gas Penetration Seals and fire exposure tests for Tokamak Fusion Test Reactor experimental areas

    International Nuclear Information System (INIS)

    Cavalluzzo, S.

    1983-01-01

    A Fire/Gas Penetration Seal is required in every penetration through the walls and ceilings into the Test Cell housing the Tokamak Fusion Test Reactor (TFTR), as well as other adjacent areas to protect the TFTR from fire damage. The penetrations are used for field coil lead stems, diagnostics systems, utilities, cables, trays, mechanical devices, electrical conduits, vacuum liner, air conditioning ducts, water pipes, and gas pipes. The function of the Fire/Gas Penetration Seals is to prevent the passage of fire and products of combustion through penetrations for a period of time up to three hours and remain structurally intact during fire exposure. The Penetration Seal must withstand, without rupture, a fire hose water stream directed at the hot surface. There are over 3000 penetrations ranging in size from several square inches to 100 square feet, and classified into 90 different types. The material used to construct the Fire/Gas Penetration Seals consist of a single and a two-component room temperature vulcanizing (RTV) silicone rubber compound. Miscellaneous materials such as alumina silica refractory fibers in board, blanket and fiber forms are also used in the construction and assembly of the Seals. This paper describes some of the penetration seals and the test procedures used to perform the three-hour fire exposure tests to demonstrate the adequacy of the seals

  3. Conceptual design of the integral test loop (I): Reactor coolant system and secondary system

    Energy Technology Data Exchange (ETDEWEB)

    Song, Chul Hwa; Lee, Seong Je; Kwon, Tae Soon; Moon, Sang Ki [Korea Atomic Energy Research Institute, Taejon (Korea)

    1999-10-01

    This report describes the conceptual design of the primary coolant system and the secondary system of the Integral Test Loop (ITL) which simulates overall thermal hydraulic phenomena of the primary system of a nuclear power plant during postulated accidents or transients. The design basis for the primary coolant system and secondary system is as follows ; Reference plant: Korean Standard Nuclear Plant (KSNP), Height ratio : 1/1, Volume ratio : 1/200, Power scale : Max. 15% of the scaled nominal power, Temperature, Pressure : Real plant conditions. The primary coolant system includes a reactor vessel, which contains a core simulator, a steam generator, a reactor coolant pump simulator, a pressurizer and piping, which consists of two hot legs, four cold legs and four intermediate legs. The secondary system consists of s steam discharge system, a feedwater supply system and a steam condensing system. This conceptual design report describes general configuration of the reference plant, and major function and operation of each system of the plant. Also described is the design philosophy of each component and system of the ITL, and specified are the design criteria and technical specifications of each component and system of the ITL in the report. 17 refs., 43 figs., 51 tabs. (Author)

  4. Definition and Analysis of Heavy Water Reactor Benchmarks for Testing New Wims-D Libraries

    International Nuclear Information System (INIS)

    Leszczynski, Francisco

    2000-01-01

    This work is part of the IAEA-WIMS Library Update Project (WLUP). A group of heavy water reactor benchmarks have been selected for testing new WIMS-D libraries, including calculations with WIMSD5B program and the analysis of results.These benchmarks cover a wide variety of reactors and conditions, from fresh fuels to high burnup, and from natural to enriched uranium.Besides, each benchmark includes variations in lattice pitch and in coolants (normally heavy water and void).Multiplication factors with critical experimental bucklings and other parameters are calculated and compared with experimental reference values.The WIMS libraries used for the calculations were generated with basic data from JEF-2.2 Rev.3 (JEF) and ENDF/B-VI iNReleaseln 5 (E6) Results obtained with WIMS-86 (W86) library, included with WIMSD5B package, from Windfrith, UK with adjusted data, are included also, for showing the improvements obtained with the new -not adjusted- libraries.The calculations with WIMSD5B were made with two methods (input program options): PIJ (two-dimension collision probability method) and DSN (one-dimension Sn method, with homogenization of materials by ring).The general conclusions are: the library based on JEF data and the DSN meted give the best results, that in average are acceptable

  5. Local flow distribution analysis inside the reactor pools of KALIMER-600 and PDRC performance test facility

    International Nuclear Information System (INIS)

    Jeong, Ji Hwan; Hwang, Seong Won; Choi, Kyeong Sik

    2010-05-01

    In the study, 3-dimensional thermal hydraulic analysis was carried out focusing on the thermal hydraulic behavior inside the reactor pools for both KALIMER-600 and one-fifth scale-down test facility. STAR-CD, one of the commercial CFD codes, was used to analyze 3-dimensional incompressible steady-state thermal hydraulic behavior in both designs of KALIMER-600 and the scale-down test facility. In the KALIMER-600 CFD analysis, the pressure drops in the core and IHX gave a good agreement within 1% error range. It was found that the porous media model was appropriate to analyze the pressure distribution inside reactor core and IHX. Also, a validation analysis showed the pressure drop through the porous media under the condition of 80% flow rate and thermal power was calculated 64% less than in 100% condition giving a physically reasonable analytic result. Since the temperatures in the hot-side pool and cold-side pool were estimated to be very close to 540 and 390 .deg. C specified on the design values respectively, the CFD models of heat source and sink was confirmed. Through the study, the methodology of 3-dimensional CFD analysis about KALIMER-600 has been established and proven. Performed with the methodology, the analysis data such as flow velocity, temperature and pressure distribution were compared by normalizing those data for the actual sized modeling and scale-down modeling. As a result, the characteristics of thermal hydraulic behavior were almost identical for the actual sized modeling and scale-down modeling and the similarity scaling law used in the design of the sodium test facility by KAERI was found to be correct

  6. Results of assembly test of HTTR reactor internals

    International Nuclear Information System (INIS)

    Maruyama, S.; Saikusa, A.; Shiozawa, S.; Tsuji, N.; Miki, T.

    1996-01-01

    The assembly test of the HTTR actual reactor internals had been carried out at the works, prior to their installation in the actual reactor pressure vessel(RPV) at the construction site. The assembly test consists of several items such as examining fabricating precision of each component and alignment of piled-up structures, measuring circumferential coolant velocity profile in the passage between the simulated RPV and the reactor internals as well as under the support plates, measuring by-pass flow rate through gaps between the reactor internals, and measuring the binding force of the core restraint mechanism. Results of the test showed good performance of the HTTR reactor internals. Installation of the reactor internals in the actual RPV was started at the construction site of HTTR in April, 1995. In the installation process, main items of the assembly test at the works were repeated to investigate the reproducibility of installation. (author). 5 refs, 11 figs

  7. 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.

  8. Improvement of transport-corrected scattering stability and performance using a Jacobi inscatter algorithm for 2D-MOC

    International Nuclear Information System (INIS)

    Stimpson, Shane; Collins, Benjamin; Kochunas, Brendan

    2017-01-01

    The MPACT code, being developed collaboratively by the University of Michigan and Oak Ridge National Laboratory, is the primary deterministic neutron transport solver being deployed within the Virtual Environment for Reactor Applications (VERA) as part of the Consortium for Advanced Simulation of Light Water Reactors (CASL). In many applications of the MPACT code, transport-corrected scattering has proven to be an obstacle in terms of stability, and considerable effort has been made to try to resolve the convergence issues that arise from it. Most of the convergence problems seem related to the transport-corrected cross sections, particularly when used in the 2D method of characteristics (MOC) solver, which is the focus of this work. Here in this paper, the stability and performance of the 2-D MOC solver in MPACT is evaluated for two iteration schemes: Gauss-Seidel and Jacobi. With the Gauss-Seidel approach, as the MOC solver loops over groups, it uses the flux solution from the previous group to construct the inscatter source for the next group. Alternatively, the Jacobi approach uses only the fluxes from the previous outer iteration to determine the inscatter source for each group. Consequently for the Jacobi iteration, the loop over groups can be moved from the outermost loop-as is the case with the Gauss-Seidel sweeper-to the innermost loop, allowing for a substantial increase in efficiency by minimizing the overhead of retrieving segment, region, and surface index information from the ray tracing data. Several test problems are assessed: (1) Babcock & Wilcox 1810 Core I, (2) Dimple S01A-Sq, (3) VERA Progression Problem 5a, and (4) VERA Problem 2a. The Jacobi iteration exhibits better stability than Gauss-Seidel, allowing for converged solutions to be obtained over a much wider range of iteration control parameters. Additionally, the MOC solve time with the Jacobi approach is roughly 2.0-2.5× faster per sweep. While the performance and stability of the Jacobi

  9. Benchmark tests of JENDL-3.2 for thermal and fast reactors

    International Nuclear Information System (INIS)

    Takano, Hideki

    1995-01-01

    Benchmark calculations for a variety of thermal and fast reactors have been performed by using the newly evaluated JENDL-3 Version-2 (JENDL-3.2) file. In the thermal reactor calculations for the uranium and plutonium fueled cores of TRX and TCA, the k eff and lattice parameters were well predicted. The fast reactor calculations for ZPPR-9 and FCA assemblies showed that the k eff , reactivity worth of Doppler, sodium void and control rod, and reaction rate distribution were in a very good agreement with the experiments. (author)

  10. Preliminary Feasibility, Design, and Hazard Analysis of a Boiling Water Test Loop Within the Idaho National Laboratory Advanced Test Reactor National Scientific User Facility

    International Nuclear Information System (INIS)

    Gerstner, Douglas M.

    2009-01-01

    The Advanced Test Reactor (ATR) is a pressurized light-water reactor with a design thermal power of 250 MW. The principal function of the ATR is to provide a high neutron flux for testing reactor fuels and other materials. The ATR and its support facilities are located at the Idaho National Laboratory (INL). A Boiling Water Test Loop (BWTL) is being designed for one of the irradiation test positions within the. The objective of the new loop will be to simulate boiling water reactor (BWR) conditions to support clad corrosion and related reactor material testing. Further it will accommodate power ramping tests of candidate high burn-up fuels and fuel pins/rods for the commercial BWR utilities. The BWTL will be much like the pressurized water loops already in service in 5 of the 9 'flux traps' (region of enhanced neutron flux) in the ATR. The loop coolant will be isolated from the primary coolant system so that the loop's temperature, pressure, flow rate, and water chemistry can be independently controlled. This paper presents the proposed general design of the in-core and auxiliary BWTL systems; the preliminary results of the neutronics and thermal hydraulics analyses; and the preliminary hazard analysis for safe normal and transient BWTL and ATR operation

  11. Solar magnetic field - 1976 through 1985: an atlas of photospheric magnetic field observations and computed coronal magnetic fields from the John M. Wilcox Solar Observatory at Stanford, 1976-1985

    International Nuclear Information System (INIS)

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

    1986-01-01

    Daily magnetogram observations of the large-scale photospheric magnetic field have been made at the John M. Wilcox Solar Observatory at Stanford since May of 1976. These measurements provide a homogeneous record of the changing solar field through most of Solar Cycle 21. Using the photospheric data, the configuration of the coronal and heliospheric fields can be calculated using a Potential Field -- Source Surface model. This provides a 3-dimensional picture of the heliospheric field-evolution during the solar cycle. In this report the authors present the complete set of synoptic charts of the measured photospheric magnetic field, the computed field at the source surface, and the coefficients of the multipole expansion of the coronal field. The general underlying structure of the solar and heliospheric fields, which determine the environment for solar - terrestrial relations and provide the context within which solar-activity-related events occur, can be approximated from these data

  12. Accelerating the design and testing of LEU fuel assemblies for conversion of Russian-designed research reactors outside Russia

    International Nuclear Information System (INIS)

    Matos, J.E

    2003-01-01

    This paper identifies proposed geometries and loading specifications of LEU tube-type and pin-type test assemblies that would be suitable for accelerating the conversion of Russian-designed research reactors outside of Russia if these fuels are manufactured, qualified by irradiation testing, and made commercially available in Russia. (author)

  13. Decommissioning of the Tokamak Fusion Test Reactor

    International Nuclear Information System (INIS)

    Perry, E.; Chrzanowski, J.; Gentile, C.; Parsells, R.; Rule, K.; Strykowsky, R.; Viola, M.

    2003-01-01

    The Tokamak Fusion Test Reactor (TFTR) at the Princeton Plasma Physics Laboratory was operated from 1982 until 1997. The last several years included operations with mixtures of deuterium and tritium. In September 2002, the three year Decontamination and Decommissioning (D and D) Project for TFTR was successfully completed. The need to deal with tritium contamination as well as activated materials led to the adaptation of many techniques from the maintenance work during TFTR operations to the D and D effort. In addition, techniques from the decommissioning of fission reactors were adapted to the D and D of TFTR and several new technologies, most notably the development of a diamond wire cutting process for complex metal structures, were developed. These techniques, along with a project management system that closely linked the field crews to the engineering staff who developed the techniques and procedures via a Work Control Center, resulted in a project that was completed safely, on time, and well below budget

  14. Decontamination and decommissioning of the MTR [Materials Testing Reactor]-603 HB-2 cubicle

    International Nuclear Information System (INIS)

    Smith, D.L.

    1987-10-01

    This paper describes the decontamination and decommissioning (D and D) of the MTR-603 HB-2 cubicle located at the Idaho National Engineering Laboratory (INEL). The HB-2 cubicle became radioactively contaminated during out-of-pile circulating water loop experiments conducted in the Materials Testing Reactor in the 1950s and 1960s. This paper describes work performed to accomplish the D and D objectives of reducing the high radiation fields caused by contamination inside the cubicle, preventing future contamination spread, and making about 1400 ft 2 of floor space available for reuse. Decommissioning of the HB-2 cubicle consisted of total dismantlement of the cubicle and its contents and was performed without disrupting ongoing laboratory work being conducted in areas surrounding the HB-2 cubicle. 3 refs., 7 figs., 4 tabs

  15. Proceedings of the international meeting on research and test reactor core conversions from HEU to LEU fuels

    International Nuclear Information System (INIS)

    1983-09-01

    Separate abstracts have been prepared for each paper presented in the following areas of interest: (1) fuel development; (2) post-irradiation examinations; (3) reprocessing; (4) thermite reaction; (5) fuel fabrication; (6) element tests; (7) core tests; (8) criticals; (9) shipping; and (10) reactors and methods

  16. Testing FLUKA on neutron activation of Si and Ge at nuclear research reactor using gamma spectroscopy

    Science.gov (United States)

    Bazo, J.; Rojas, J. M.; Best, S.; Bruna, R.; Endress, E.; Mendoza, P.; Poma, V.; Gago, A. M.

    2018-03-01

    Samples of two characteristic semiconductor sensor materials, silicon and germanium, have been irradiated with neutrons produced at the RP-10 Nuclear Research Reactor at 4.5 MW. Their radionuclides photon spectra have been measured with high resolution gamma spectroscopy, quantifying four radioisotopes (28Al, 29Al for Si and 75Ge and 77Ge for Ge). We have compared the radionuclides production and their emission spectrum data with Monte Carlo simulation results from FLUKA. Thus we have tested FLUKA's low energy neutron library (ENDF/B-VIIR) and decay photon scoring with respect to the activation of these semiconductors. We conclude that FLUKA is capable of predicting relative photon peak amplitudes, with gamma intensities greater than 1%, of produced radionuclides with an average uncertainty of 13%. This work allows us to estimate the corresponding systematic error on neutron activation simulation studies of these sensor materials.

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

    International Nuclear Information System (INIS)

    Grove, E.; Gunther, W.

    1991-01-01

    Though mechanically different, the control rod drive (CRD) systems used at both Combustion Engineering (CE) and Babcock and Wilcox (B and W) plants position the control rod assemblies (CRA) in the core in response to automatic or manual reactivity control signals. Both systems are also designed to provide a rapid insertion of the CRAs upon a loss of AC power. The CRD system consists of the actual drive mechanisms, power and control, rod position indication, and cooling system components. This aging evaluation included the individual absorber rods, and the fuel assembly and upper internal guide tubes, since failure of these components could preclude the insertion of the control assemblies. Aging and environmental degradation have resulted in system and component failures. Many of these failures caused dropped or slipped rods which adversely affected plant operations by resulting in power reductions, scrams, and safety system actuation. No CRD system failure has ever resulted in the inability to shut down a reactor. However, unplanned, automatic trips challenge the operation of the plants safety systems. Consequently, their occurrence represents a potentially significant increase in plant risk. System and component failures have resulted in four Information Notices during the past decade

  18. Developing the MAPLE materials test reactor concept

    International Nuclear Information System (INIS)

    Lee, A.G.; Lidstone, R.F.; Donnelly, J.V.

    1992-05-01

    MAPLE-MTR is a new multipurpose research facility being planned by AECL Research as a possible replacement for the 35-year-old NRU reactor. In developing the MAPLE-MTR concept, AECL is starting from the recent design and licensing experience with the MAPLE-X10 reactor. By starting from technology developed to support the MAPLE-X10 design and adapting it to produce a concept that satisfies the requirements of fuel channel materials testing and fuel irradiation programs, AECL expects to minimize the need for major advances in nuclear technology (e.g., fuel, heat transfer). Formulation of the MAPLE-MTR concept is at an early stage. This report describes the irradiation requirements of the research areas, how these needs are translated into design criteria for the project and elements of the preliminary design concept

  19. Tensile and impact testing of an HFBR [High Flux Beam Reactor] control rod follower

    International Nuclear Information System (INIS)

    Czajkowski, C.J.; Schuster, M.H.; Roberts, T.C.; Milian, L.W.

    1989-08-01

    The Materials Technology Group of the Department of Nuclear Energy (DNE) at Brookhaven National Laboratory (BNL) undertook a program to machine and test specimens from a control rod follower from the High Flux Beam Reactor (HFBR). Tensile and Charpy impact specimens were machined and tested from non-irradiated aluminum alloys in addition to irradiated 6061-T6 from the HFBR. The tensile test results on irradiated material showed a two-fold increase in tensile strength to a maximum of 100.6 ksi. The impact resistance of the irradiated material showed a six-fold decrease in values (3 in-lb average) compared to similar non-irradiated material. Fracture toughness (K I ) specimens were tested on an unirradiated compositionally and dimensionally similar (to HFBR follower) 6061 T-6 material with K max values of 24.8 ± 1.0 Ksi√in (average) being obtained. The report concludes that the specimens produced during the program yielded reproducible and believable results and that proper quality assurance was provided throughout the program. 9 figs., 6 tabs

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

    International Nuclear Information System (INIS)

    Bates, E.F.; Grimes, B.K.; Ramos, S.L.

    1982-05-01

    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

  1. Inspection qualification programme for VVER reactors and review of round robin test results

    International Nuclear Information System (INIS)

    Horacek, L.; Zdarek, J.

    1998-01-01

    Experience obtained, especially from in-service inspections of VVER 440-type reactor pressure vessels and from the Czech round test trials with international participation of ultrasonic teams, has highlighted the need for an in-service inspection qualification programme in the Czech Republic focused on NDT procedures, equipment and personnel. Recently, several national and international regional projects included in the PHARE programme (projects 4.1.2/93 and 1.02/94), briefly described, have been initiated. These projects are to cover step by step the programme of the in-service inspection qualification in view of technical justification as well as of practical assessment-performance demonstration-for all the main VVER-type primary circuit components. (orig.)

  2. Performance tests for integral reactor nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Sohn, Dong-Seong; Yim, Jeong-Sik; Lee, Chong-Tak; Kim, Han-Soo; Koo, Yang-Hyun; Lee, Byung-Ho; Cheon, Jin-Sik; Oh, Je-Yong

    2006-02-15

    An integral type reactor SMART plans to utilize metallic Zr-U fuel which is Zr-based alloy with 34{approx}38 wt% U. In order to verify the technologies for the design and manufacturing of the fuel and get a license, performance tests were carried out. Experimental Fuel Assembly (EFA) manufactured in KAERI is being successfully irradiated in the MIR reactor of RIAR from September 4 2004, and it has achieved burnup of 0.21 g/cc as of January 25 2006. Thermal properties of irradiated Zr-U fuel were measured. Up to the phase transformation temperature, thermal diffusivity increased linearly in proportion to temperature. However its dependence on the burnup was not significant. RIA tests with 4 unirradiated Zr-U fuel rods were performed in Kurchatov Institute to establish a safety criterion. In the case of the un-irradiated Zr-U fuel, the energy deposition during the control rod ejection accident should be less than 172 cal/g to prevent the failure accompanying fuel fragmentation and dispersal. Finally the irradiation tests of fuel rods have been performed at HANARO. The HITE-2 test was successfully completed up to a burnup of 0.31 g/cc. The HITE-3 test began in February 2004 and will be continued up to a target burnup of 0.6 g/cc.

  3. Long- and short-term trends in vessel conditioning of TFTR [Tokamak Fusion Test Reactor

    International Nuclear Information System (INIS)

    LaMarche, P.H.; Dylla, H.F.; Bell, M.G.

    1986-10-01

    We have investigated trends in the conditioning of the Tokamak Fusion Test Reactor (TFTR) vacuum vessel during the May 1984 to April 1985 run period. The initial conditioning of the vessel, consisting of glow discharge cleaning (GDC) and pulse discharge cleaning (PDC) in concert with a 150 0 C vessel bakeout, is necessary to assure plasma operation after atmospheric venting. A long-term conditioning process, ascribed to limiter conditioning, effectively improves operational conditions during the course of the run. Over several thousand high power plasma discharges, the improvement was documented by using standard parameter (fiducial) plasma discharges. Several techniques demonstrated short-term improvements in vessel conditioning during this time period, including: Cr gettering and programming the plasma position relative to the limiter contact area

  4. The fabrication and performance of Canadian silicide dispersion fuel for test reactors

    International Nuclear Information System (INIS)

    Sears, D.F.; Wood, J.C.; Berthiaume, L.C.; Herbert, L.N.; Schaefer, J.D.

    1985-01-01

    Fuel fabrication effort is now concentrated on the commissioning of large-scale process equipment, defining product specifications, developing a quality assurance plan, and setting up a mini-computer material accountancy system. In the irradiation testing program, full-size NRU assemblies containing 20% enriched silicide dispersion fuel have been Irradiated successfully to burnups in the range 65-80 atomic percent. Irradiations have also been conducted on mini-elements having 1.2 mm diameter holes In their mid-sections, some drilled before irradiation and others after irradiation to 22-83 atomic percent burnup. Uranium was lost to the coolant in direct proportion to the surface area of exposed core material. Pre-irradiation in the intact condition appeared to reduce in-reactor corrosion. Fuel cores developed for the NRU reactor are dimensionally very stable, swelling by only 6-8% at the very high burnup of 93 atomic percent. Two important factors contributing to this good performance are cylindrical clad restraint and coarse silicide particles. Thermal ramping tests were conducted on irradiated silicide aspersion fuels. Small segments of fuel cores released 85 Kr starting at about 520 deg. C and peaking at about 680 deg C. After a holding period of 1 hour at 720 deg. C a secondary 85 Kr peak occurred during cooling (at about 330 deg. C) probably due to thermal contraction cracking. Whole mini-elements irradiated to 93 atomic percent burnup were also ramped thermally, with encouraging results. After about 0.25 h at 530 deg. C the aluminum cladding developed very localized small blisters, some with penetrating pin-hole cracks preventing gross pillowing or ballooning. (author)

  5. The fabrication and performance of Canadian silicide dispersion fuel for test reactors

    Energy Technology Data Exchange (ETDEWEB)

    Sears, D F; Wood, J C; Berthiaume, L C; Herbert, L N; Schaefer, J D

    1985-07-01

    Fuel fabrication effort is now concentrated on the commissioning of large-scale process equipment, defining product specifications, developing a quality assurance plan, and setting up a mini-computer material accountancy system. In the irradiation testing program, full-size NRU assemblies containing 20% enriched silicide dispersion fuel have been Irradiated successfully to burnups in the range 65-80 atomic percent. Irradiations have also been conducted on mini-elements having 1.2 mm diameter holes In their mid-sections, some drilled before irradiation and others after irradiation to 22-83 atomic percent burnup. Uranium was lost to the coolant in direct proportion to the surface area of exposed core material. Pre-irradiation in the intact condition appeared to reduce in-reactor corrosion. Fuel cores developed for the NRU reactor are dimensionally very stable, swelling by only 6-8% at the very high burnup of 93 atomic percent. Two important factors contributing to this good performance are cylindrical clad restraint and coarse silicide particles. Thermal ramping tests were conducted on irradiated silicide aspersion fuels. Small segments of fuel cores released {sup 85}Kr starting at about 520 deg. C and peaking at about 680 deg C. After a holding period of 1 hour at 720 deg. C a secondary {sup 85}Kr peak occurred during cooling (at about 330 deg. C) probably due to thermal contraction cracking. Whole mini-elements irradiated to 93 atomic percent burnup were also ramped thermally, with encouraging results. After about 0.25 h at 530 deg. C the aluminum cladding developed very localized small blisters, some with penetrating pin-hole cracks preventing gross pillowing or ballooning. (author)

  6. Engineering and technology in the deconstruction of nuclear materials production facilities

    International Nuclear Information System (INIS)

    Kingsley, R.S.; Reynolds, W.E.; Heffner, D.C.

    1996-01-01

    Technology and equipment exist to support nuclear facility deactivation, decontamination, and decommissioning. In reality, this statement is not surprising because the nuclear industry has been decontaminating and decommissioning production plants for decades as new generations of production technology were introduced. Since the 1950s, the Babcock and Wilcox Company (B ampersand W) has operated a number of nuclear materials processing facilities to manufacture nuclear fuel for the commercial power industry and the U.S. Navy. These manufacturing facilities included a mixed oxide (PuO 2 -UO 2 ) nuclear fuel manufacturing plant, low- and high-enriched uranium (HEU/LEU) chemical and fuel plants, and fuel assembly plants. In addition, B ampersand W designed and build a major nuclear research center in Lynchburg, Virginia, to support these nuclear fuel manufacturing activities and to conduct nuclear power research. These nuclear research facilities included two research reactors, a hot-cell complex for nuclear materials research, four critical experiment facilities, and a plutonium fuels research and development facility. This article describes the B ampersand W deactivation, decomtanimation, and decommisioning program

  7. Advanced Test Reactor probabilistic risk assessment

    International Nuclear Information System (INIS)

    Atkinson, S.A.; Eide, S.A.; Khericha, S.T.; Thatcher, T.A.

    1993-01-01

    This report discusses Level 1 probabilistic risk assessment (PRA) incorporating a full-scope external events analysis which has been completed for the Advanced Test Reactor (ATR) located at the Idaho National Engineering Laboratory

  8. Thermal Hydraulic Tests for Reactor Core Safety

    Energy Technology Data Exchange (ETDEWEB)

    Moon, S. K.; Baek, W. P.; Chun, S. Y. (and others)

    2007-06-15

    The main objectives of the present project are to resolve the current issues of reactor core thermal hydraulics, to develop an advanced measurement and analytical techniques, and to perform reactor core safety verification tests. 6x6 reflood experiments, various heat transfer experiments using Freon, and experiments on the spacer grids effects on the post-dryout are carried out using spacer grids developed in Korea in order to resolve the current issues of the reactor core thermal hydraulics. In order to develop a reflood heat transfer model, the detailed reflood phenomena are visualized and measured using round tube and 2x2 rod bundle. A detailed turbulent mixing ph