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

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

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

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

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

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

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

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

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

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

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

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

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

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.

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

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.)

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

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

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

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

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

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

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

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

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.)

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

xLPR - a probabilistic approach to piping integrity analysis

International Nuclear Information System (INIS)

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

2015-01-01

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

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

Autoreactive T cells in MRL/Mpr-lpr/lpr mice. Characterization of the lymphokines produced and analysis of antigen-presenting cells required

International Nuclear Information System (INIS)

Weston, K.M.; Ju, S.T.; Lu, C.Y.; Sy, M.S.

1988-01-01

Lymph node cells from 4-wk-old MRL/Mp-lpr/lpr mice, but not from MRL/Mp-+/+ mice, when cultured in vitro for 5 to 7 days, will spontaneously proliferate and produce IL-2. We examined the expression of several cell surface Ag on lymph node cells from MRL/Mp-lpr/lpr mice before and after in vitro culture. There is an increase in the expression of Thy-1, L3T4, IL-2R, T cell activating protein, T cell receptor, and T3 complex on the surface of cultured cells. Cultured cells produced IL-3, IFN-gamma, and small but detectable amounts of IL-1 in addition to IL-2. Gamma irradiation of APC from young MRL/Mp-lpr/lpr mice or treatment of APC with a mAb (J11D) and C, completely abrogated their stimulatory capacity. These experiments suggest that B cells are the predominant APC responsible in the activation of autoreactive T cells in MRL/Mp-lpr/lpr mice. Lymph node cells from C57BL/6-lpr/lpr or C3H-lpr/lpr mice were unable to spontaneously proliferate or produce IL-2. Lymph node cells from (MRL/Mp-lpr/lpr x C57BL/6-lpr/lpr) F1 mice or (C3H-lpr/lpr x MRL/Mp-lpr/lpr) F1 mice did proliferate and produced IL-2 after in vitro culture. Using T cells from these F1 animals and APC from each parental haplotype, we found that APC from MRL/Mp-lpr/lpr mice induced more proliferation and greater amounts of IL-2, when compared to APC from F1 animals. APC from C57BL6-lpr/lpr mice or C3H-lpr/lpr were unable to induce spontaneous proliferation and IL-2 production. Therefore, B cells from MRL/Mp-lpr/lpr mice appear to possess unique features that enable them to activate autoreactive T cells more effectively than B cells from other mice bearing the lpr/lpr gene

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

International Nuclear Information System (INIS)

Matsumoto, K.; Yoshikai, Y.; Asano, T.; Himeno, K.; Iwasaki, A.; Nomoto, K.

1991-01-01

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

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

International Nuclear Information System (INIS)

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

1987-01-01

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

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

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.

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)

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

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

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

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.)

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

International Nuclear Information System (INIS)

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

2010-01-01

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

Fuel element production at BWX technologies

International Nuclear Information System (INIS)

Pace, Brett

1997-01-01

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

xLPR Scenario Analysis Report.

Energy Technology Data Exchange (ETDEWEB)

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

2017-03-01

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

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

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

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

Green County Nuclear Power Plant. License application

International Nuclear Information System (INIS)

1975-07-01

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

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

NTRE extended life feasibility assessment

Science.gov (United States)

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2010-12-01

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

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.)

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)

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

International Nuclear Information System (INIS)

Westsik, G.A.

1997-01-01

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

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

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

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

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

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

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)

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

FK506: therapeutic effects on lupus dermatoses in autoimmune-prone MRL/Mp-lpr/lpr mice.

Science.gov (United States)

Furukawa, F; Imamura, S; Takigawa, M

1995-01-01

The effects of FK506, a new immunosuppressive agent, on the development of lupus dermatoses were investigated in the autoimmune-prone MRL/Mp-lpr/lpr (MRL/lpr) mouse, which is an animal model for the spontaneous development of skin lesions similar to those of human lupus erythematosus (LE). FK506 reduced the incidence of skin lesions, lupus nephritis, the titre of serum anti-double-stranded DNA antibodies and the massive T cell proliferation. The incidence and magnitude of IgG deposition at the dermoepidermal junction were not changed. These results suggest that FK506 is a promising immunosuppressive agent for the control of autoimmune skin diseases.

xLPR Sim Editor 1.0 User's Guide

Energy Technology Data Exchange (ETDEWEB)

Mariner, Paul

2017-03-01

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

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

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.

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

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

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

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

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

International Nuclear Information System (INIS)

Almagro, J.C.; Dupetit, G.A.; Deandreis, R.A.

1987-01-01

The availability of the LPR (Radiochemical Processes Laboratory) plant offers the possibility to demonstrate and create the necessary technological basis for thorium fuels reprocessing. To this purpose, the solvents extraction technique is used, employing TBP (at 30%) as solvent. The process is named THOREX, a one-cycle acid, which permits an adequate separation of Th 232 and U 233 components and fission products. For thorium oxide elements dissolution, the 'chopp-leach' process (installed at LPR) is used, employing a NO 3 H 13N, 0.05M FH and 0.1M Al (NO 3 ) 3 , as solvent. To adapt the pilot plant to the flow-sheet requirements proposed, minor modifications must be carried out in the interconnection of the existing decanting mixers. The input of the plant has been calculated by Origin Code modified for irradiations in reactors of the HWR type. (Author)

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

Science.gov (United States)

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

2014-12-01

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

Supply strategy for SMR deployment

International Nuclear Information System (INIS)

Coccagna, A.F.

2013-01-01

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

Integrating LPR with CCTV systems: problems and solutions

Science.gov (United States)

Bissessar, David; Gorodnichy, Dmitry O.

2011-06-01

A new generation of high-resolution surveillance cameras makes it possible to apply video processing and recognition techniques on live video feeds for the purpose of automatically detecting and identifying objects and events of interest. This paper addresses a particular application of detecting and identifying vehicles passing through a checkpoint. This application is of interest to border services agencies and is also related to many other applications. With many commercial automated License Plate Recognition (LPR) systems available on the market, some of which are available as a plug-in for surveillance systems, this application still poses many unresolved technological challenges, the main two of which are: i) multiple and often noisy license plate readings generated for the same vehicle, and ii) failure to detect a vehicle or license plate altogether when the license plate is occluded or not visible. This paper presents a solution to both of these problems. A data fusion technique based on the Levenshtein distance is used to resolve the first problem. An integration of a commercial LPR system with the in-house built Video Analytic Platform is used to solve the latter. The developed solution has been tested in field environments and has been shown to yield a substantial improvement over standard off-the-shelf LPR systems.

Assessment of nonbackfittable concepts for improving uranium utilization in LWRs

International Nuclear Information System (INIS)

Newman, D.F.; Goldsmith, S.; Fleischman, R.M.

1980-01-01

Recent efforts to improve uranium utilization in light water reactors (LWRs) have involved backfittable changes to fuel or operations. The Advanced Reactor Design Study sponsored by the US Department of Energy identified and evaluated nonbackfittable LWR concepts to provide a basis for selecting and demonstrating specific improvements that have good implementation potential. Because the application of nonbackfittable concepts necessitates modifications to contemporary reactor designs, it was apparent that the most qualified organizations to assess implementation potential would be LWR designers/vendors. Accordingly, Babcock and Wilcox, Combustion Engineering, and General Electric were the principal participants in selecting, assessing, and evaluating the nonbackfittable concepts included in this study. The results of the industrial assessments of nonbackfittable LWR concepts are shown

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

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

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)

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

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

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)

Development of steam generator manufacturing technology

International Nuclear Information System (INIS)

Grant, J.A.

1979-01-01

In 1968 Babcock and Wilcox (Operations) Ltd., received an order from the CEGB to design, manufacture, install and commission 16 Steam Generators for 2 x 660 Mw (e) Advanced Gas Cooled Reactor Power Station at Hartlepool. This order was followed in 1970 by a similar order for the Heysham Power Station. The design and manufacture of the Steam Generators represented a major advance in technology and the paper discusses the methods by which a manufacturing facility was developed, by the Production Division of Babcock, to produce components to a quality, complexity and accuracy unique in the U.K. commercial boilermaking industry. The discussion includes a brief design background, a description of the Steam Generators and a view of the Production Division background. This is followed by a description of the organisation of the technological development and a consideration of the results. (author)

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)

Joint pathology and behavioral performance in autoimmune MRL-lpr Mice.

Science.gov (United States)

Sakić, B; Szechtman, H; Stead, R H; Denburg, J A

1996-09-01

Young autoimmune MRL-lpr mice perform more poorly than age-matched controls in tests of exploration, spatial learning, and emotional reactivity. Impaired behavioral performance coincides temporally with hyperproduction of autoantibodies, infiltration of lymphoid cells into the brain, and mild arthritic-like changes in hind paws. Although CNS mechanisms have been suggested to mediate behavioral deficits, it was not clear whether mild joint pathology significantly affected behavioral performance. Previously we observed that 11-week-old MRL-lpr mice showed a trend for disturbed performance when crossing a narrow beam. The first aim of the present study was to test the significance of this trend by increasing the sample size and, second, to examine the possibility that arthritis-like changes interfere with performance in brief locomotor tasks. For the purpose of the second goal, 18-week-old mice that differ widely in severity of joint disease were selectively taken from the population and tested in beam walking and swimming tasks. It was expected that the severity of joint inflammation would be positively correlated with the degree of locomotor impairment. The larger sample size revealed that young MRL-lpr mice perform significantly more poorly than controls on the beam-walking test, as evidenced by more foot slips and longer traversing time. However, significant correlation between joint pathology scores and measures of locomotion could not be detected. The lack of such relationship suggests that mild joint pathology does not significantly contribute to impaired performance in young, autoimmune MRL-lpr mice tested in short behavioral tasks.

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

400-MWe consolidated nuclear steam system (CNSS). 1255 MWt CNSS design/cost update

International Nuclear Information System (INIS)

1984-07-01

Since 1976 Babcock and Wilcox (B and W) has been extensively involved in the development of a medium-sized (1255 MWt/400 MWe) reactor. Under the sponsorship of the U.S. Department of Energy (DOE) and through a contract with Oak Ridge National Laboratories (ORNL), B and W investigated the feasibility of the concept for utility power generation and cogenerated process heat. The potential benefits of the design, called the Consolidated Nuclear Steam System (CNSS), were also identified. This study provides an update of the CNSS design and cost reflecting current regulatory requirements and operating reactor experience. The study was funded by DOE through ORNL and was performed by B and W and UE and C

Prototypical consolidation demonstration project - Final fuel recommendation report

International Nuclear Information System (INIS)

Piscitella, R.R.; Paskey, W.R.

1987-01-01

The Prototypical Consolidation Demonstration (PCD) Project will, in its final phase, conduct a demonstration of the equipment's ability to consolidate actual spent commercial fuel. Since budget and schedule limitations do not allow this demonstration to include all types of fuel assemblies, a selection process was utilized to identify the fuel types that would represent predominate fuel inventories and that would demonstrate the equipment's abilities. The Pressurized Water Reactor (PWR) fuel assemblies that were suggested for use in the PCD Project Hot Demonstration were Babcock and Wilcox (B and W) 15 x 15's, and Westinghouse (WE) 15 x 15's. The Boiling Water Reactor (BWR) fuel suggested was the General Electric (GE) 8 x 8

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

Directory of Open Access Journals (Sweden)

Agnes Kim

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

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

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.

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

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

Science.gov (United States)

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

2008-11-15

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

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

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

An expert system approach for safety diagnosis

International Nuclear Information System (INIS)

Erdmann, R.C.; Sun, B.K.H.

1988-01-01

An expert system was developed with the intent to provide real-time information about an accident to an operator who is in the process of diagnosing and bringing that accident under control. Explicit use was made of probabilistic risk analysis techniques and plant accident response information in constructing this system. The expert system developed contains 70 logic rules and provides contextual messages during simulated accident sequences and logic sequence information on the entire sequence in graphical form for accident diagnosis. The present analysis focuses on integrated control system-related transients with Babcock and Wilcox-type reactors. While the system developed here is limited in extent and was built for a composite reactor, it demonstrates that an expert system may enhance the operator's capability in the control room

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.)

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)

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.

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.

Improvement of Diagnostic Flow Chart in Severe Accident Management Guidance for Nuclear Power Plants in Korea

Energy Technology Data Exchange (ETDEWEB)

Jang, So Won; Lee, Su Won [FNC Technology Co., Yongin (Korea, Republic of); Oh, Hae Cheol [KHNP CRI, Daejeon (Korea, Republic of)

2016-10-15

Development of generic SAMG in the United States was undertaken by WOG, Combustion Engineering Owner’s Group (CEOG), and Babcock and Wilcox Owner’s Group (BWOG) in the 1990s. They made efforts to develop generic SAMG specific to the individual plant designs to satisfy the regulatory concerns for severe accident managements. Recently, the SAMG of the Pressurized Water Reactor Owner's Group (PWROG) which is applicable to Babcock and Wilcox (B and W), Combustion Engineering (CE), and Westinghouse Pressurizer Water Reactor (PWR) Nuclear Steam Supply System (NSSS) designs was developed in February 2016 by incorporating the best features from the previous PWR generic SAMG. However, the structure of the PWROG SAMG is also based on the WOG SAMG. The purpose of this study is to provide improvements of the Korean SAMG by comparing between the Korean SAMG and the PWROG SAMG to reflect the state of the art and trends. The improved DFC for the Korean SAMG is proposed by comparing between the Korean SAMG and the PWROG SAMG. If the improved DFC is adopted, the level to identify the severity of the plant condition is diversified, the order of priority of the strategies is changed, the entry condition into the RCS injection strategy is changed, LTMG-01and SAEG-01 come under the DFC as parts of the step, and the step whether CCI is occurred or not is added in DFC. In order to increase the effectiveness of the Korean SAMG and to reflect the state of the art and trends, it is appropriate to apply the suggestions contained in this paper to the Korean SAMG.

Field characterization of plutonium aerosols in mixed-oxide fuel fabrication

International Nuclear Information System (INIS)

Newton, G.J.; Teague, S.V.; Yeh, H.C.

1976-01-01

Nuclear reactor fuel pellets of PuO 2 and UO 2 are fabricated within safety enclosures at Babcock and Wilcox's Parker Township Site near Apolla, Pa. Nineteen sample runs were taken from within glove boxes of aerosols formed during powder comminution and blending. Eight sampling runs were also taken of a centerless grinding operation 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 concentrations. The potential toxicity of inhaled plutonium originating in the nuclear fuel cycle following accidental releases of these aerosols and possible inhalation by industrial workers is considered

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

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

International Nuclear Information System (INIS)

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

1987-01-01

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

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

On-line monitoring of main coolant pump seals

International Nuclear Information System (INIS)

Stevens, D.M.; Spencer, J.W.; Morris, D.J.; Glass, S.W.; Sommerfield, G.A.; Harrison, D.

1984-06-01

The Babcock and Wilcox Company has developed and implemented a Reactor Coolant Pump Monitoring and Diagnostic System (RCPM and DS). The system has been installed at Toledo Edison Company's Davis-Besse Nuclear Power Station Unit 1. The RCPM and PS continuously monitors a number of indicators of pump performance and notifies the plant operator of out-of-tolerance conditions or pump performance trending toward out-of-tolerance conditions. Pump seal parameters being monitored include pump internal pressures, temperatures, and flow rates. Rotordynamic performanvce and plant operating conditions are also measured with a variety of dynamic sensors. This paper describes the implementation of the system and the results of on-line monitoring of four RC pumps

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

Energy Technology Data Exchange (ETDEWEB)

Klasky, Hilda B. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Williams, Paul T. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Bass, Bennett Richard [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2012-09-01

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

FIELD EVALUATION OF LOW-EMISSION COAL BURNER TECHNOLOGY ON UTILITY BOILERS VOLUME III. FIELD EVALUATIONS

Science.gov (United States)

The report gives results of field tests conducted to determine the emission characteristics of a Babcock and Wilcox Circular burner and Dual Register burner (DRB). The field tests were performed at two utility boilers, generally comparable in design and size except for the burner...

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)

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).

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 ...

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

Post-implementation review of inadequate core cooling instrumentation

International Nuclear Information System (INIS)

Anderson, J.L.; Anderson, R.L.; Hagen, E.W.; Morelock, T.C.; Huang, T.L.; Phillips, L.E.

1988-01-01

Studies of Three Mile Island (TMI) 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 and 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 (BandW) licensees' orders for Modification of License and transmitted to all pressurized water reactor (PWR) licensees Generic Letter 82-28 to inform them of the revised NRC requirements. The instrumentation requirements for detection of ICC include upgraded subcooling margin monitors (SMMs), upgraded core exit thermocouples (CETs), and installation of a reactor coolant inventory tracking system (RCITS)

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)

Teaming up to improve reliability

International Nuclear Information System (INIS)

Malone, E.A.; Ayres, D.J.; Lear, R.C. van

1989-01-01

Responding to increasingly stringent regulatory requirements, Babcock and Wilcox has teamed up with three specialist companies to provide services for nuclear utilities aiming to improve the performance of their valves and actuators. The services, which are outlined here, include inspection, repair, overhaul and valve and actuator reliability programmes. (author)

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

Science.gov (United States)

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

2014-06-01

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

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

International Nuclear Information System (INIS)

Mitsutoshi Tsukimoto; Fumitoshi Tago; Hiroko Nakatsukasa; Shuji Kojima

2007-01-01

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

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

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

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

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

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

International Nuclear Information System (INIS)

Anon.

1979-01-01

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

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

Good Engineering + Poor Communication = Three Mile Island.

Science.gov (United States)

Mathes, J. C.

The accident at the Three Mile Island nuclear power plant resulted from a communication failure. Following an incident at an Ohio plant a year and a half earlier, B. M. Dunn, manager of Emergency Core Cooling Systems Analysis at Babcock and Wilcox (engineers), wrote a memorandum making specific recommendations on written instructions for nuclear…

Vitrification of low-level radioactive waste in a slagging combustor

International Nuclear Information System (INIS)

Holmes, M.J.; Downs, W.; Higley, B.A.

1995-07-01

The suitability of a Babcock ampersand Wilcox cyclone furnace to vitrify a low-level radioactive liquid waste was evaluated. The feed stream contained a mixture of simulated radioactive liquid waste and glass formers. The U.S. Department of Energy is testing technologies to vitrify over 60,000,000 gallons of this waste at the Hanford site. The tests reported here demonstrated the technical feasibility of Babcock ampersand Wilcox's cyclone vitrification technology to produce a glass for near surface disposal. Glass was produced over a period of 24-hours at a rate of 100 to 150 lb/hr. Based on glass analyses performed by an independent laboratory, all of the glass samples had leachabilities at least as low as those of the laboratory glass that the recipe was based upon. This paper presents the results of this demonstration, and includes descriptions of feed preparation, glass properties, system operation, and flue gas composition. The paper also provides discussions on key technical issues required to match cyclone furnace vitrification technology to this U.S. Department of Energy Hanford site application

Nuclear Regulatory Commission issuances, January 1995. Volume 41, Number 1

Energy Technology Data Exchange (ETDEWEB)

NONE

1995-01-01

This book contains issuances of the Atomic Safety and Licensing Boards for January 1995. The issuances include Babcock and Wilcox Company materials license; Hydro Resources, Inc. application for uranium mining; low-level waste storage in Utah; communication of emerging and existing generic, technical issues with PWR owners groups; and radioactive waste management by Sierra Nuclear Corporation.

Nuclear Regulatory Commission issuances, January 1995. Volume 41, Number 1

International Nuclear Information System (INIS)

1995-01-01

This book contains issuances of the Atomic Safety and Licensing Boards for January 1995. The issuances include Babcock and Wilcox Company materials license; Hydro Resources, Inc. application for uranium mining; low-level waste storage in Utah; communication of emerging and existing generic, technical issues with PWR owners groups; and radioactive waste management by Sierra Nuclear Corporation

Severe accident management; the approach in the USA. Applications of US methods in Europe. Other approaches in Europe

International Nuclear Information System (INIS)

Vayssier, G.

1999-01-01

In this lecture severe accident management, applications of US methods in Europe are presented. Author deals with historical perspective, US industry position to core melt accidents, method of Westinghouse owners group, method of Combustion Engineering owners group, method of Babcock and Wilcox Owners group, interaction with/inspection by the USNRC and with assessment of US SAMG methods

Nondestructive examination of Oconee 1 fuel assemblies after three cycles of irradiation

International Nuclear Information System (INIS)

Pyecha, T.D.; Davis, H.H.; Mayer, J.T.; Guthrie, B.A. III; Larson, J.G.

1979-09-01

The Babcock and Wilcox Company (B and W) in conjunction with Duke Power Company is participating in a Department of Energy sponsored research and development program to qualify current design pressurized water reactor (PWR) fuel assemblies for extended burnup (>40,000 MWd/mtU). The information obtained from this program will provide a basis for future design improvements in PWR fuel assemblies culminating in an extended burnup assembly having a nominal operating limit of approximately 50,000 MWd/mtU. An extension of the current assembly design to higher burnups will result in the following benefits: (1) lower uranium ore requirements, (2) greater fuel cycle efficiency, (3) reduction in spent fuel storage requirements, and (4) increased flexibility in tailoring fuel batch sizes to better accommodate the varying energy requirements of the utilities

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.

Handbook of International alloy Compositions and Designations. Volume II. Superalloys

Science.gov (United States)

1978-12-01

Aubert et Duval, 92200 Neuilly-sur-Seine, Paris, France C-0180 August Dreckshage Eisen und Stahl, 4800 Bielefeld, West Germany C-0021 Avesta Jernverks...Aktiebolag, 774 01 Avesta , Sweden B C-0087 Babcock & Wilcox, Beaver Falls, Pennsylvania 15010 C-0126 Bergische Stahl-lndustrie, (BSD 5630 Remscheid...C-0033) C-0021 Avesta Jernverks Aktiebolag, 774 01 Avesta , Sweden C-0022 Howmet Turbine Components Corporation, (Div. of PUG), Muskegon, Michigan

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

Directory of Open Access Journals (Sweden)

M. Rybanský

2009-01-01

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

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

Science.gov (United States)

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

2011-04-01

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

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

Science.gov (United States)

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

2017-01-01

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

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

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

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

TRAC analysis of steam-generator overfill transients for TMI-1

International Nuclear Information System (INIS)

Bassett, B.

1983-01-01

A reactor safety issue concerning the overfilling of once-through steam generators leading to combined primary/secondary blowdown has been raised recently. A series of six calculations, performed with the LWR best-estimate code, TRAC-PD2, on a Babcock and Wilcox Plant (TMI-1), was performed to investigate this safety issue. The base calculation assumed runaway main feedwater to one steam generator causing it to overfill and to break the main steam line. Four additional calculations build onto the base case with combinations of a pump-seal failure, a steam-generator tube rupture, and the pilot-operated relief valve not reseating. A sixth calculation involved only the rupture of a single steam-generator tube. The results of these analyses indicate that for the transients investigated, the emergency cooling system provided an adequate make-up coolant flow to mitigate the accidents

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

Energy Technology Data Exchange (ETDEWEB)

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

2015-02-01

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

New steam generators slated for nuclear units

International Nuclear Information System (INIS)

Anon.

1992-01-01

This article is a brief discussion of Duke Power's plans to replace steam generators at its McGuire and Catawba nuclear units. A letter of intent to purchase (from Babcock and Wilcox) the 12 Westinghouse steam generators has been signed, but no constructor has been selected at this time. This action is brought about by the failures of more than 3000 tubes in these units

Ginsenoside Rh1 Improves the Effect of Dexamethasone on Autoantibodies Production and Lymphoproliferation in MRL/lpr Mice

Directory of Open Access Journals (Sweden)

Yinglu Feng

2015-01-01

Full Text Available Ginsenoside Rh1 is able to upregulate glucocorticoid receptor (GR level, suggesting Rh1 may improve glucocorticoid efficacy in hormone-dependent diseases. Therefore, we investigated whether Rh1 could enhance the effect of dexamethasone (Dex in the treatment of MRL/lpr mice. MRL/lpr mice were treated with vehicle, Dex, Rh1, or Dex + Rh1 for 4 weeks. Dex significantly reduced the proteinuria and anti-dsDNA and anti-ANA autoantibodies. The levels of proteinuria and anti-dsDNA and anti-ANA autoantibodies were further decreased in Dex + Rh1 group. Dex, Rh1, or Dex + Rh1 did not alter the proportion of CD4+ splenic lymphocytes, whereas the proportion of CD8+ splenic lymphocytes was significantly increased in Dex and Dex + Rh1 groups. Dex + Rh1 significantly decreased the ratio of CD4+/CD8+ splenic lymphocytes compared with control. Con A-induced CD4+ splenic lymphocytes proliferation was increased in Dex-treated mice and was inhibited in Dex + Rh1-treated mice. Th1 cytokine IFN-γ mRNA was suppressed and Th2 cytokine IL-4 mRNA was increased by Dex. The effect of Dex on IFN-γ and IL-4 mRNA was enhanced by Rh1. In conclusion, our data suggest that Rh1 may enhance the effect of Dex in the treatment of MRL/lpr mice through regulating CD4+ T cells activation and Th1/Th2 balance.

Supercompactor force effectiveness as related to dry active waste volume reduction

International Nuclear Information System (INIS)

Williams, P.C.; Phillips, W.S.

1986-01-01

The first U.S. permanently installed supercompactor is now in operation at the Babcock and Wilcox volume reduction center, Parks Township, Pennsylvania. Tests with various DAW (dry active waste) material have been conducted, recording press force versus drum height as one means of estimating volume reduction capability of this machine at various compaction forces. The results of these tests, as well as other factors, are presented herein

MANTIS helps collection of piping erosion/corrosion data

International Nuclear Information System (INIS)

Fairbrother, D.B.

1988-01-01

One of the challenges posed by erosion/corrosion inspection of piping in nuclear power plants is handling the large amounts of data generated. Babcock and Wilcox's MANTIS (Modular Automated Non-destructive Thickness Inspection System) uses a small battery-powered, waterproof, dustproof computer to log thickness readings. The readings are obtained from a flaw detector with an analog output signal that is proportional in thickness. (author)

TRAC analyses of severe overcooling transients for the Oconee-1 PWR

Energy Technology Data Exchange (ETDEWEB)

Ireland, J R [comp.

1985-05-01

This report describes the results of several Transient Reactor Analysis Code (TRAC)-PF1 calculations of overcooling transients in a Babcock and Wilcox lowered-loop, pressurized water reactor (Oconee-1). The purpose of this study is to provide detailed input on thermal-hydraulic data to Oak Ridge National Laboratory for pressurized thermal-shock analyses. The transient calculations performed were plant specific in that details of the primary system, the secondary system, and the plant-integrated control system of Oconee-1 were included in the TRAC input model. The results of the calculations indicate that the turbine-bypass valve failure transient was the most severe in terms of resulting in relatively cold liquid temperatures in the downcomer region of the vessel. The power-operated relief valve loss-of-coolant accident transient was the least severe in terms of downcomer liquid temperatures because of vent-valve fluid mixing and near-saturated conditions in the primary system. It is recommended that future calculations consider a wider range of operator actions to cover the spectra of overcooling transient sequences more completely. 6 refs., 287 figs., 32 tabs.

TRAC analyses of severe overcooling transients for the Oconee-1 PWR

International Nuclear Information System (INIS)

Ireland, J.R.

1985-05-01

This report describes the results of several Transient Reactor Analysis Code (TRAC)-PF1 calculations of overcooling transients in a Babcock and Wilcox lowered-loop, pressurized water reactor (Oconee-1). The purpose of this study is to provide detailed input on thermal-hydraulic data to Oak Ridge National Laboratory for pressurized thermal-shock analyses. The transient calculations performed were plant specific in that details of the primary system, the secondary system, and the plant-integrated control system of Oconee-1 were included in the TRAC input model. The results of the calculations indicate that the turbine-bypass valve failure transient was the most severe in terms of resulting in relatively cold liquid temperatures in the downcomer region of the vessel. The power-operated relief valve loss-of-coolant accident transient was the least severe in terms of downcomer liquid temperatures because of vent-valve fluid mixing and near-saturated conditions in the primary system. It is recommended that future calculations consider a wider range of operator actions to cover the spectra of overcooling transient sequences more completely. 6 refs., 287 figs., 32 tabs

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

Energy Technology Data Exchange (ETDEWEB)

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

2010-12-01

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

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

OpenAIRE

Camarasa Lillo, Natalia

2009-01-01

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

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

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

Science.gov (United States)

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

2016-05-01

Myeloid dendritic cells (DCs) can produce B-cell-activating factor (BAFF) that modulates survival and differentiation of B cells and plays a pivotal role in the pathogenesis of systemic lupus erythematosus (SLE). Toll-like receptor 4 (TLR4) signaling has important functions in the process of BAFF production. Our previous study showed that a benzenediamine derivate FC-99 possesses anti-inflammation activity and directly interacts with interleukin-1 receptor-associated kinase 4 (IRAK4), which was a pivotal molecule in TLR4 signaling. In this study, we demonstrated that FC-99 attenuated lupus nephritis in the MRL/lpr mice. FC-99 also decreased the levels of total immunoglobulin G (IgG), total IgG2a and IgM in sera, as well as the activation of B cells in the spleens of MRL/lpr mice. Moreover, FC-99 inhibited abnormal activation of myeloid DCs in spleens and reduced the levels of BAFF in sera, spleens, and kidneys of MRL/lpr mice. Furthermore, upon TLR4 stimulation with lipopolysaccharide in vitro, FC-99 inhibited IRAK4 phosphorylation, as well as the activation and BAFF production in murine bone marrow-derived DCs. These data indicate that FC-99 attenuates lupus nephritis in MRL/lpr mice via inhibiting DC-secreted BAFF, suggesting that FC-99 may be a potential therapeutic candidate for the treatment of SLE. © The Author 2016. Published by ABBS Editorial Office in association with Oxford University Press on behalf of the Institute of Biochemistry and Cell Biology, Shanghai Institutes for Biological Sciences, Chinese Academy of Sciences.

Davis-Besse uncertainty study

International Nuclear Information System (INIS)

Davis, C.B.

1987-08-01

The uncertainties of calculations of loss-of-feedwater transients at Davis-Besse Unit 1 were determined to address concerns of the US Nuclear Regulatory Commission relative to the effectiveness of feed and bleed cooling. Davis-Besse Unit 1 is a pressurized water reactor of the raised-loop Babcock and Wilcox design. A detailed, quality-assured RELAP5/MOD2 model of Davis-Besse was developed at the Idaho National Engineering Laboratory. The model was used to perform an analysis of the loss-of-feedwater transient that occurred at Davis-Besse on June 9, 1985. A loss-of-feedwater transient followed by feed and bleed cooling was also calculated. The evaluation of uncertainty was based on the comparisons of calculations and data, comparisons of different calculations of the same transient, sensitivity calculations, and the propagation of the estimated uncertainty in initial and boundary conditions to the final calculated results

Recent trends of SMR developments in USA

International Nuclear Information System (INIS)

Kang, Han Ok

2011-11-01

Recently several nuclear reactor vendors from the United States officially announced the development and construction plans for the Small and Modular Reactors (SMR). NuScale Power completed a preliminary plant design and intends to submit a design certification application to the NRC next year. Babcock and Wilcox is developing the mPower reactor, which has the capacity to provide 125 MWe to 750 MWe or more for a 4.5-year operating cycle without refueling, with its scalable and modular design. Recently Westinghouse has officially introduced new 200 MWe Small Modular Reactor (SMR) for the US Department of Energy's demonstration program. The SMART reactor is an advanced integral type reactor with a power of 330MWt. A passive residual heat removal system (PRHRS) is also installed to prevent over-heating and over-pressurization of the primary system during accidental conditions. Safety injection system for the SBLOCA and containment spray system for the containment cooling and precipitation of radioactive material are based on the active systems of conventional nuclear plant. Reliance on the active systems for the safety function lowers public acceptance and makes the SMART less competitive with other SMRs developed in the United States. Two approaches are suggested to solve this problem. The first approach is the passive safety system on the basis of the steel safeguard vessel and the second is on the basis of concrete containment

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

International Nuclear Information System (INIS)

1983-09-01

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

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

Energy Technology Data Exchange (ETDEWEB)

1983-09-01

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

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)

Development of Krsko Severe Accident Management Guidance (SAMG)

International Nuclear Information System (INIS)

Cizel, F.

1999-01-01

In this lecture development of severe accident management guidances for Krsko NPP are described. Author deals with the history of severe accident management and implementation of issues (validation, review of E-plan and other aspects SAMG implementation guidance). Methods of Westinghouse owners group, of Combustion Engineering owners group, of Babcock and Wilcox owners group, of the BWR owners group, as well as application of US SAMG methodology in Europe and elsewhere are reviewed

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

NRC's object-oriented simulator instructor station

International Nuclear Information System (INIS)

Griffin, J.I.; Griffin, J.P.

1995-06-01

As part of a comprehensive simulator upgrade program, the simulator computer systems associated with the Nuclear Regulatory Commission's (NRC) nuclear power plant simulators were replaced. Because the original instructor stations for two of the simulators were dependent on the original computer equipment, it was necessary to develop and implement new instructor stations. This report describes the Macintosh-based Instructor Stations developed by NRC engineers for the General Electric (GE) and Babcock and Wilcox (B and W) simulators

Uncertainty analysis of minimum vessel liquid inventory during a small-break LOCA in a B ampersand W Plant: An application of the CSAU methodology using the RELAP5/MOD3 computer code

International Nuclear Information System (INIS)

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

1992-12-01

The Nuclear Regulatory Commission (NRC) revised the emergency core cooling system licensing rule to allow the use of best estimate computer codes, provided the uncertainty of the calculations are quantified and used in the licensing and regulation process. The NRC developed a generic methodology called Code Scaling, Applicability, and Uncertainty (CSAU) to evaluate best estimate code uncertainties. The objective of this work was to adapt and demonstrate the CSAU methodology for a small-break loss-of-coolant accident (SBLOCA) in a Pressurized Water Reactor of Babcock ampersand Wilcox Company lowered loop design using RELAP5/MOD3 as the simulation tool. The CSAU methodology was successfully demonstrated for the new set of variants defined in this project (scenario, plant design, code). However, the robustness of the reactor design to this SBLOCA scenario limits the applicability of the specific results to other plants or scenarios. Several aspects of the code were not exercised because the conditions of the transient never reached enough severity. The plant operator proved to be a determining factor in the course of the transient scenario, and steps were taken to include the operator in the model, simulation, and analyses

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

Evaluation of safety implications of control systems in LWR nuclear power plants

International Nuclear Information System (INIS)

Szukiewicz, A.J.

1989-06-01

An in-depth evaluation was performed on non-safety-related control systems (see Section 1) that are typically used during normal plant operation on four nuclear steam supply system plants: a General Electric Company boiling-water reactor, a Westinghouse 3-loop pressurized-water reactor (PWR), a Babcock ampersand Wilcox Co. (B ampersand W) once-through steam generator PWR, and a Combustion Engineering PWR design. A study was also conducted to determine the generic applicability of the results to the class of plants represented by the specific plants analyzed. Generic conclusions were then developed. Steam generator and reactor vessel overfill events and reactor vessel overcooling events were identified as major classes of events having the potential to be more severe than previously analyzed. Specific substasks of this issue were to study these events to determine the need for preventive and/or mitigating design measures. This report describes the technical studies performed by the laboratories, the NRC staff assessment of the results, the generic applicability of the evaluations, and the technical findings resulting from these studies. This final report contains the staff's responses to, and resolution of, the public comments that were solicited and received before September 16,1988, in response to the draft reports issued for public comment on May 27, 1988. 39 refs, 1 fig., 7 tabs

Concept selection for advanced low-emission coal fired boiler

Energy Technology Data Exchange (ETDEWEB)

Gorrell, R.L. [Babcock and Wilcox Co., Barberton, OH (United States); Rodgers, L.W.; Farthing, G.A. [Babcock and Wilcox Co., Alliance, OH (United States)

1993-12-31

The Babcock & Wilcox Company (B&W), under contract to the US Department of Energy (DOE) with subcontract to Physical Sciences, Inc. (PSIT), the Massachusetts Institute of Technology (MIT) and United Engineers and Constructors (UE&C) has begun development of an advanced low-emission boiler system (LEBS). The initial phase of this multi-phase program required a thorough review and assessment of potential advanced technologies and techniques for control of combustion and flue gas emissions. Results of this assessment are presented in this paper.

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

A 600 MWe advanced PWR for the 1990's

International Nuclear Information System (INIS)

Lemon, J.E.; Malloy, J.D.; Allen, R.E.

1987-01-01

The Babcock and Wilcox Company (B and W) and United Engineers and Constructors (UE and C) have prepared a conceptual design of an advanced 600 MWe Presurized Water Nuclear Power Plant. This design utilizes the large body of design and operating experience on PWRs in the U.S. and abroad and incorporates improvements emphasizing simplicity, safety, licensability, ease of construction, operability, reliability and maintainability. Cost and schedule estimates based on U.S. utility experience indicate that this plant design should be competitive with alternate options

The effect of hydroxychloroquine on lupus erythematosus-like skin lesions in MRL/lpr mice.

Science.gov (United States)

Shimomatsu, Tatsuya; Kanazawa, Nobuo; Mikita, Naoya; Nakatani, Yumi; Li, Hong-Jin; Inaba, Yutaka; Ikeda, Takaharu; Kondo, Toshikazu; Furukawa, Fukumi

2016-09-01

To evaluate the effect and safety of hydroxychloroquine (HCQ) on lupus erythematosus (LE)-like skin lesions in the MRL/lpr mouse, a model for systemic LE (SLE). We divided the MRL/lpr mice into three groups that were given: (1) drinking water, (2) HCQ at a dose of 4 mg/kg/d, or (3) HCQ at a dose of 40 mg/kg/d. The HCQ was administered to examine the effect and safety of HCQ on skin lesions and the number of infiltrating cells including mast cells in the dermis. Six of 13 mice in the group given drinking water, 3 of 11 mice in the group administered low-dose HCQ (4 mg/kg/d), and 1 of 10 mice in the group administered high-dose HCQ (40 mg/kg/d) presented the skin lesions. The average number of mast cells was 81, 50, and 12 (magnification, ×100), the mortality rate was 24%, 8%, and 9% and the mean body weight gain was 4.6 g, 8.0 g and 5.1 g, respectively. HCQ was demonstrated to decrease the appearance of LE-like lesions and the number of mast cells in the dermis. Furthermore, there were no obvious systemic adverse effects. This study provides evidence that suggests benefits in human patients.

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).

CONTEMPT: computer program for predicting containment pressure-temperature response to a loss-of-coolant accident

International Nuclear Information System (INIS)

Hsii, Y.H.

1978-04-01

The CONTEMPT code is used by Babcock and Wilcox for containment analysis following a postulated loss of coolant accident. An additional model is described which is used for the calculation of long term post reflood mass and energy releases to the containment that is used for the containment design basis LOCA calculations. These calculations maximize the rate of energy flow to the containment. The mass and energy data are given to the containment designer for use in calculating the containment building design pressure and temperature and in sizing containment heat removal equipment

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…

Nuclear Regulatory Commission issuances, April 1995. Volume 41, Number 4

International Nuclear Information System (INIS)

1995-04-01

This book contains issuances of the Nuclear Regulatory Commission and of the Atomic Safety and Licensing Boards, and an issuance of the Director's decision. The issuances concern a petition filed by Dr. James E Bauer seeking interlocutory Commission review of the Atomic Safety and Licensing Board's order imposing several restrictions on Dr. Bauer; a denial of an Interveners' Petition for Review addressing the application of Babcock and Wilcox for a renewal of its Special Nuclear Materials License; granting a motion for a protective order, by Sequoyah Fuel Corporation and General Atomics, limiting the use of the protected information to those individuals participating in the litigation and for the purposes of the litigation only; granting a Petitioner's petition for leave to intervene and request for a hearing concerning Georgia Institute of Technology (Georgia Tech Research Reactor) renewal of a facility license; and a denial of a petition filed by Mr. Ted Dougherty requesting a shutdown of the San Onofre Nuclear Generating Station based on concerns regarding the vulnerability of the plant to earthquakes and defensibility of the plant to a terrorist threat

White Paper on Data Repository Reorganization Proposal for the xLPR Project

Energy Technology Data Exchange (ETDEWEB)

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

2012-09-01

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

Measurements on spent-fuel assemblies at Arkansas Nuclear One using the Fork system. Final report, January 1995

International Nuclear Information System (INIS)

Ewing, R.I.; Bronowski, D.R.; Bosler, G.E.; Siebelist, R.; Priore, J.; Hansford, C.H.; Sullivan, S.

1997-03-01

The Fork measurement system has been used to examine spent-fuel assemblies at the two reactors of Arkansas Nuclear One, operated by Entergy Operations, Inc. The Unit 1 reactor is a Babcock and Wilcox (B and W) design, and the Unit 2 reactor is a Combustion Engineering (CE) design. The neutron and gamma-ray emissions from individual spent-fuel assemblies were measured in the storage pools by raising each assembly pathway out of the storage rack and performing a measurement near the center of the assembly. The overall accuracy of the measurements after corrections is about 2%. Thirty-four assemblies were examined at Unit 1, and forty-one assemblies at Unit 2. The average deviation of the burnup measurements from the calibration was 3.0% at Unit 1 and 3.5% at Unit 2, indicating 2 to 3% random variation among the reactor records. There was no indication of clearly anomalous assemblies. Axial Scans of the variation in neutron and gamma ray emission were obtained by collecting data at several locations along the length of three assemblies at Unit 2. Two of these assemblies were nonstandard in that each contained a small neutron source. The sources were detected by the axial scans. The test program was a cooperative effort involving Sandia National Laboratories, Los Alamos National Laboratory, Entergy Operations, Inc., the Electric Power Research Institute, and the Office of Civilian Radioactive Waste Management of the US Department of Energy

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

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…

Drug-like analogues of the parasitic worm-derived immunomodulator ES-62 are therapeutic in the MRL/Lpr model of systemic lupus erythematosus

Science.gov (United States)

Rodgers, D T; Pineda, M A; Suckling, C J; Harnett, W

2015-01-01

Introduction ES-62, a phosphorylcholine (PC)-containing immunomodulator secreted by the parasitic worm Acanthocheilonema viteae, protects against nephritis in the MRL/Lpr mouse model of systemic lupus erythematosus (SLE). However, ES-62 is not suitable for development as a therapy and thus we have designed drug-like small molecule analogues (SMAs) based around its active PC-moiety. To provide proof of concept that ES-62-based SMAs exhibit therapeutic potential in SLE, we have investigated the capacity of two SMAs to protect against nephritis when administered to MRL/Lpr mice after onset of kidney damage. Methods SMAs 11a and 12b were evaluated for their ability to suppress antinuclear antibody (ANA) generation and consequent kidney pathology in MRL/Lpr mice when administered after the onset of proteinuria. Results SMAs 11a and 12b suppressed development of ANA and proteinuria. Protection reflected downregulation of MyD88 expression by kidney cells and this was associated with reduced production of IL-6, a cytokine that exhibits promise as a therapeutic target for this condition. Conclusions SMAs 11a and 12b provide proof of principle that synthetic compounds based on the safe immunomodulatory mechanisms of parasitic worms can exhibit therapeutic potential as a novel class of drugs for SLE, a disease for which current therapies remain inadequate. PMID:26085597

Determination of fuel assembly vibrational modes through analysis of incore detector noise

International Nuclear Information System (INIS)

Johnson, R.S.

1986-01-01

In order to better characterize fuel assembly vibration at Duke Power Company's Oconee Nuclear Station, incore noise data were acquired an analyzed from prompt responding detectors incorporated in the Oconee 2, Cycle 7 core. Duke Power Company began actively pursuing an inhouse Neutron Noise Analysis program for routine surveillance of reactor internals vibration in 1979. Noise data has since been acquired and analyzed for twelve cycles of operation for the three Oconee units. Duke Power's Oconee Unit 2 is a Babcock and Wilcoxs pressurized water reactor with a rate thermal power of 2568MW. For Oconee 2, Cycle 7 operation, two test assemblies, each employing a string of seven axially-spaced, prompt responding hafnium detectors, were included in the final core design. Incore detector noise data were obtained during Cycle 7 at approximately 281 and 430 effective full power days (EFPD). In addition to the incore test detector signals, noise signals from the upper and lower chambers of the four excore power range detectors were recorded to aid in the analysis. The comparison of RMS signal levels for each incore detector and the phase relationships between detector locations within two test assemblies identified the first four fuel assembly bending modes associated with fixed end conditions

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).

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

International Nuclear Information System (INIS)

Vyska, K.; Freundlieb, C.; Hoeck, A.; Becker, V.; Schmidt, A.; Feinendegen, L.E.; Kloster, G.; Stoecklin, G.; Heiss, W.D.

1982-01-01

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

Fuel quality and its effect on the design of power boilers in the USA

Energy Technology Data Exchange (ETDEWEB)

Kotler, V.R.

1984-05-01

Statistical data, taken from Power, Proceedings of the American Power Conference and others, on developments since the 1950s in boiler design caused by the increasing use of lower quality fuel (subbituminous and lignite coals) are presented. The effect of pollution regulations in the USA on boiler design is discussed. The results of a 16 year study by the TVA on the decrease in coal quality fired in its boilers and its effect on boiler efficiency are presented. Methods of transport are surveyed. Descriptions and characteristics of several modern boilers designed by Babcock and Wilcox, Combustion Engineering, Foster-Wheeler and Riley Stoker are given. 13 references.

Effect of boric acid on intergranular corrosion in tube support plate crevices

International Nuclear Information System (INIS)

Brunet, J.P.; Campan, J.L.

1993-10-01

Intergranular attack on steam generator tubing is one important phenomenon involved in availability of Pressurized Water Reactors. Boric acid appears to be a possible candidate for inhibiting the corrosion process. The program performed in Cadarache was supposed to give statistical informations on the boric acid effect. It was based on a large number of samples initially attacked during a program performed by BABCOCK ampersand WILCOX. These samples were sleeved onto Alloy 690 tubes, in order to prevent premature cracking. Unfortunately it was not possible to find chemical conditions able to produce significant additional corrosion; we postulated mainly due to a drastic reduction of the thermal flux resulting from the increase of the tube wall thickness under the tube support plates (TSP). The tests demonstrate that such sleeve could be a possible remedy of the corrosion when introduced under the TSP. The tests show indications of a possible beneficial effect of the boric acid, a large variability of the heats sensitivity to the IGA and a predominant effect of Na 2 CO 3 on IGA production

Pre-service baseline inspection using x-probe of Oconee replacement steam generators

International Nuclear Information System (INIS)

Addario, M.; Shipp, P.; Davis, K.; Fogal, C.

2003-01-01

The eddy current method has been the industry standard for inspecting steam generator tubing for many years and the level of sophistication of coil technology has continued to evolve during that time. State of the art array probe systems now employ multiple sensitivity zones in the probe to better detect and characterize defects in an efficient manner. Owners and regulators of nuclear power plants are interested in the most effective and efficient inspection possible. The ultimate goal has been to meet or exceed new and existing regulatory and design requirements by maximizing the quantity and quality of eddy current data collected while minimizing both the time needed to perform the inspection and the radiation exposure. The X-Probe is an example of this new eddy current array technology. Qualified to detect all types of known defects in steam generator tubing, the technology is comprised of a system of probe, data acquisition instrumentation, computer and human interface software. Recently, Duke Power, along with Babcock and Wilcox Canada and the system developer R/D Tech, collaborated to implement this technology in a first of a kind full scale pre-service inspection of replacement steam generators for Duke Power's Oconee nuclear generating station at Babcock and Wilcox Canada's Cambridge plant. The discussion in this paper will briefly describe the X-Probe technology, describe the system required to perform the inspection, present the general results of the inspection and finally draw some comparative benefit conclusions for both pre-service and in-service applications. (author)

Boosting INPO's programs with supplier know-how

International Nuclear Information System (INIS)

Matson, B.

1984-01-01

Nuclear industry suppliers, such as Stone and Webster and Babcock and Wilcox, are key participants in the Institute of Nuclear Power Operators (INPO) effort to promote technical information exchanges throughout the industry and to contribute to INPO training programs. Suppliers have 21 employees on loan to the Institute, although steps must be taken to protect proprietary information and to avoid breaking antitrust laws. Among the NUCLEAR NETWORK programs using supplier expertise is the Significant Event Evaluation and Information Network (SEE-IN), which reviews important reports involving their products. Suppliers benefit by observing evaluation teams at construction sites and plant start-ups

Nuclear snow job: utilities struggle to avoid credibility meltdown

International Nuclear Information System (INIS)

Feeney, A.

1979-01-01

Electric utilities and the US nuclear industry are criticized for giving a public relations response to safety concerns following the accident at Three Mile Island and for passing the cost of the response on to utility customers. The large amounts of money budgeted by Babcock and Wilcox, Bechtel Power Corporation, the Atomic Industrial Forum, Electric Power Research Institute, Nuclear Energy Women, and others to minimize the incident and counter anti-nuclear arguments with media events are seen here by Feeney as evidence that the utilities and nuclear industry are afraid of political consequences as their technical credibility vanishes

Refractory experience in circulating fluidized bed combustors, Task 7

Energy Technology Data Exchange (ETDEWEB)

Vincent, R.Q.

1989-11-01

This report describes the results of an investigation into the status of the design and selection of refractory materials for coal-fueled circulating fluidized-bed combustors. The survey concentrated on operating units in the United States manufactured by six different boiler vendors: Babcock and Wilcox, Combustion Engineering, Foster Wheeler, Keeler Dorr-Oliver, Pyropower, and Riley Stoker. Information was obtained from the boiler vendors, refractory suppliers and installers, and the owners/operators of over forty units. This work is in support of DOE's Clean Coal Technology program, which includes circulating fluidized-bed technology as one of the selected concepts being evaluated.

Refractory experience in circulating fluidized bed combustors, Task 7. Final report

Energy Technology Data Exchange (ETDEWEB)

Vincent, R.Q.

1989-11-01

This report describes the results of an investigation into the status of the design and selection of refractory materials for coal-fueled circulating fluidized-bed combustors. The survey concentrated on operating units in the United States manufactured by six different boiler vendors: Babcock and Wilcox, Combustion Engineering, Foster Wheeler, Keeler Dorr-Oliver, Pyropower, and Riley Stoker. Information was obtained from the boiler vendors, refractory suppliers and installers, and the owners/operators of over forty units. This work is in support of DOE`s Clean Coal Technology program, which includes circulating fluidized-bed technology as one of the selected concepts being evaluated.

Partnering for Canada's nuclear future

International Nuclear Information System (INIS)

Koenderman, P.P.

1997-01-01

''Partnering'' is an evolving relationship that could lead to a partnership or joint ownership. Babcock and Wilcox (BW) has used a variety of forms of contracting and partnering to develop global strategy for the supply of its products, both fossil and nuclear steam generating equipment. A strategic mix of consortia, strategic alliances and joint ventures has provided the impetus for BW to lead in worldwide market share in many categories, including the supply of nuclear replacement steam generators to the USA since 1992. The implication is that continuing cooperation with BW will benefit the Canadian nuclear industry. 6 refs., 12 figs

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

International Nuclear Information System (INIS)

Grove, E.; Gunther, W.

1991-01-01

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

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

International Nuclear Information System (INIS)

Dowling, E.F.; Benedict, B.J.; Snidow, N.L.

1981-05-01

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

The Heuristic Model Based on LPR in the Context of Material Conversion

Directory of Open Access Journals (Sweden)

Wilk-Kołodziejczyk D.

2017-09-01

Full Text Available High complexity of the physical and chemical processes occurring in liquid metal is the reason why it is so difficult, impossible even sometimes, to make analytical models of these phenomena. In this situation, the use of heuristic models based on the experimental data and experience of technicians is fully justified since, in an approximate manner at least, they allow predicting the mechanical properties of the metal manufactured under given process conditions. The study presents a methodology applicable in the design of a heuristic model based on the formalism of the logic of plausible reasoning (LPR. The problem under consideration consists in finding a technological variant of the process that will give the desired product parameters while minimizing the cost of production. The conducted tests have shown the effectiveness of the proposed approach.

Disposition of surplus highly enriched uranium: Draft environmental impact statement

International Nuclear Information System (INIS)

1995-10-01

This document assesses the environmental impacts at four potential sites that may result from alternatives for the disposition of United States-origin weapons-usable highly enriched uranium (HEU) that has been or may be declared surplus to national defense or defense-related program needs. In addition to the no action alternative, it assesses four alternatives that would eliminate the weapons-usability of HEU by blending it with depleted uranium, natural uranium, or low-enriched uranium (LEU) to create low-enriched uranium, either as commercial reactor fuel feedstock or as low-level radioactive waste. The potential blending sites are DOE's Y-12 Plant at Oak Ridge Reservation in Oak Ridge, Tennessee; DOE's Savannah River Site in Aiken, South Carolina; the Babcock ampersand Wilcox Naval Nuclear Fuel Division Facility in Lynchburg, Virginia; and the Nuclear Fuel Services Fuel Fabrication Plant in Erwin, Tennessee. Evaluations of impacts on site infrastructure, water resources, air quality and noise, socioeconomic resources, waste management, public and occupational health, and environmental justice for the potential blending sites are included in the assessment. The intersite transportation of nuclear and hazardous materials is also assessed. The preferred alternative is to blend down surplus HEU to LEU for maximum commercial use as reactor fuel feed which would likely be done at a combination of DOE and commercial sites

Analysis of zircaloy oxide thickness data from PWRs

International Nuclear Information System (INIS)

Sheppard, K.D.; Speyer, D.M.; Chan, Y.Y.; Frankl, I.; Strasser, A.A.

1990-02-01

Prior EPRI funded research (Project 1250-1) resulted in a set of Zircaloy waterside corrosion models. These models were based principally on KWU reactor data. The objective of this study was to evaluate the ability of the KWU corrosion models to predict available domestic USA data for all domestic PWR vendors in order to further validate the models and to provide a consistent basis to judge the corrosion data of the domestic plants. A methodology for analyzing the large amount of data was developed and implemented in a single channel model. This model includes the capability, by a method described herein, of accounting for open core related effects (crossflow) and the effect of the immediately adjacent fuel rods, guide tubes, etc., on the coolant conditions around the fuel rods that were measured for oxide thickness. Data from the Arkansas Unit number-sign 2 (ANO-2) Combustion Engineering (C-E), Oconee Units 1 and 2 built by Babcock ampersand Wilcox (B ampersand W), and the Trojan reactor built by Westinghouse (W) were used in this study. The corrosion models previously developed, and the present single channel model methodology, were able to predict the corrosion data quite well. The maximum corrosion thickness was on the order of 20 to 40 microns in all plants studied. 13 refs., 11 figs., 5 tabs

Nuclear Regulatory Commission issuances, April 1995. Volume 41, Number 4

Energy Technology Data Exchange (ETDEWEB)

NONE

1995-04-01

This book contains issuances of the Nuclear Regulatory Commission and of the Atomic Safety and Licensing Boards, and an issuance of the Director`s decision. The issuances concern a petition filed by Dr. James E Bauer seeking interlocutory Commission review of the Atomic Safety and Licensing Board`s order imposing several restrictions on Dr. Bauer; a denial of an Interveners` Petition for Review addressing the application of Babcock and Wilcox for a renewal of its Special Nuclear Materials License; granting a motion for a protective order, by Sequoyah Fuel Corporation and General Atomics, limiting the use of the protected information to those individuals participating in the litigation and for the purposes of the litigation only; granting a Petitioner`s petition for leave to intervene and request for a hearing concerning Georgia Institute of Technology (Georgia Tech Research Reactor) renewal of a facility license; and a denial of a petition filed by Mr. Ted Dougherty requesting a shutdown of the San Onofre Nuclear Generating Station based on concerns regarding the vulnerability of the plant to earthquakes and defensibility of the plant to a terrorist threat.

Technical notes for the conceptual design for an atmospheric fluidized-bed direct combustion power generating plant. [570 MWe plant

Energy Technology Data Exchange (ETDEWEB)

None

1978-04-01

The design, arrangement, thermodynamics, and economics of a 592 MW(e) (nominal gross) electric power generating plant equipped with a Babcock and Wilcox Company (B and W) atmospheric fluidized bed (AFB) boiler are described. Information is included on capital and operating costs, process systems, electrical systems, control and instrumentation, and environmental systems. This document represents a portion of an overall report describing the conceptual designs of two atmospheric fluidized bed boilers and balance of plants for the generation of electric power and the analysis and comparison of these conceptual designs to a conventional pulverized coal-fired electric power generation plant equipped with a wet limestone flue gas desulfurization system.

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

International Nuclear Information System (INIS)

Werner, A.

1980-01-01

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

Application of a B ampersand W developed computer aided pictorial process planning system to CQMS for manufacturing process control

International Nuclear Information System (INIS)

Johanson, D.C.; VandeBogart, J.E.

1992-01-01

Babcock ampersand Wilcox (B ampersand W) will utilize its internally developed Computer Aided Pictorial Process Planning or CAPPP (pronounced open-quotes cap cubedclose quotes) system to create a paperless manufacturing environment for the Collider Quadruple Magnets (CQM). The CAPPP system consists of networked personal computer hardware and software used to: (1) generate and maintain the documents necessary for product fabrication, (2) communicate the information contained in these documents to the production floor, and (3) obtain quality assurance and manufacturing feedback information from the production floor. The purpose of this paper is to describe the various components of the CAPPP system and explain their applicability to product fabrication, specifically quality assurance functions

Applications of natural analogue studies to Yucca Mountain as a potential high level radioactive waste repository

International Nuclear Information System (INIS)

1995-02-01

The 5-member group convened in Las Vegas, Nov. 11-13, 1991, to clarify the extent to which studies of natural analogues can assist the Yucca Mountain site characterization (SC) project. This document is to provide guidance and recommendations to DOE for the implementation of natural analogue studies in the SC program. Performance assessment, integrity of engineered barriers, and communication to the public and the scientific community are stressed. The reference design being developed by Babcock ampersand Wilcox Fuel Company are reviewed. Guidelines for selecting natural analogues are given. Quality assurance is discussed. Recommendations are given for developing an effective natural analogue program within the SC program

Programmable Automated Welding System (PAWS)

Science.gov (United States)

Kline, Martin D.

1994-01-01

An ambitious project to develop an advanced, automated welding system is being funded as part of the Navy Joining Center with Babcock & Wilcox as the prime integrator. This program, the Programmable Automated Welding System (PAWS), involves the integration of both planning and real-time control activities. Planning functions include the development of a graphical decision support system within a standard, portable environment. Real-time control functions include the development of a modular, intelligent, real-time control system and the integration of a number of welding process sensors. This paper presents each of these components of the PAWS and discusses how they can be utilized to automate the welding operation.

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

International Nuclear Information System (INIS)

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

1975-01-01

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

A THREE-DIMENSIONAL BABCOCK-LEIGHTON SOLAR DYNAMO MODEL

International Nuclear Information System (INIS)

Miesch, Mark S.; Dikpati, Mausumi

2014-01-01

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

Environmental Impact Appraisal for renewal of special nuclear material license No. SNM-42 (Docket No. 70-27)

International Nuclear Information System (INIS)

1984-03-01

The Babcock and Wilcox Company (B and W) Naval Nuclear Fuel Division (NNFD) facility near Lynchburg, Virginia, produces fuel assemblies and complete fuel modules for reactors used in the US Navy nuclear propulsion program and fuel components for university and other research reactors; and processes scrap material to recover the enriched uranium content. No significant modifications of the production procedures for the US Navy nuclear fuel fabrication have been made since the previous environmental assessment, and none are anticipated during the five-year license renewal period being considered. In 1982 the fabrication of fuel assemblies for university and other research reactors was begun. This environmental assessment provides a review of the past five years of operation and an analysis of future impacts, including the effect of plant changes. The proposed action is the renewal of the license necessary for B and W to continue the existing fuel fabrication operations. Principal operations in the fabrication facility include the processing of highly enriched uranium (> 90% 235 U) into fuel elements and assembling the elements into complete reactor cores for shipment and eventual installation in US Navy facilities. The principal environmental impacts of current operation of the NNFD result from release of radioactive gases to the atmosphere and of radioactively contaminated liquids to the adjacent James River. The actual gaseous and liquid pollutants released during normal operation of the plant have been monitored and documented. The principal subjects addressed in this environmental assessment include water use, pollutant controls, environmental monitoring, and environmental impact of operation and accidents. Other site factors and plant operations necessary for this assessment are described, and aspects of insignificant impacts are identified. 10 figures, 36 tables

Synergy for a Strong Future FY 2008

International Nuclear Information System (INIS)

Devore, L.; Chrzanowski, P.

2008-01-01

Lawrence Livermore National Security, LLC is committed to delivering the best combination of scientific research, technology development, business management, and safe, secure operations in support of Lawrence Livermore National Laboratory's critical national security mission. LLNS was formed specifically to manage LLNL for the Department of Energy's National Nuclear Security Administration. LLNS consists of a team of five organizations renowned for their expertise and accomplishments throughout the U.S. nuclear weapons complex and beyond - Bechtel National, University of California, Babcock and Wilcox, Washington Division of URS Corporation, and Battelle. Bechtel is the nation's largest engineering and construction firm and a leader in project management. The University of California is the world's largest public research institution. Babcock and Wilcox and the Washington Division of URS Corporation are top nuclear facilities contractors and between them manage four of DOE's five safest sites. Battelle is a global leader in science and technology development and commercialization. The LLNS Board of Governors provides oversight for the management of the Laboratory and holds the Director and LLNS President responsible for the Laboratory's performance. The Board has seven standing committees that assist in assessing Laboratory performance and monitoring risks and internal controls. Through the Board of Governors, the Laboratory can reach back to LLNS partner organizations to help ensure that it fulfills its national security mission with excellence in scientific research, technology development, business management, and safe, secure operations. LLNS assumed management of LLNL on October 1, 2007. This report highlights LLNS accomplishments in FY2008, its first year as the Laboratory's managing contractor. It is clear that LLNS and the Laboratory have exploited numerous synergies inherent in their relationship - for example, science and engineering, mission and

Nuclear Law Bulletin No. 98. Volume 2016/2

International Nuclear Information System (INIS)

Wetherall, Anthony C.; Soedersten, Anna; Berger, Marjorie; Paez, M.R.; Touitou-Durand, F.; Pelzer, N.; Adomaityte, U.; Majerus, P.; Nowacki, T.; Pospisil, M.; Skraban, A.; Noelliste, N.E.; Popov, A.; Drillat, C.; Reynaers Kini, E.

2016-01-01

The Nuclear Law Bulletin is a unique international publication for both professionals and academics in the field of nuclear law. It provides readers with authoritative and comprehensive information on nuclear law developments. Published free online twice a year in both English and French, it features topical articles written by renowned legal experts, covers legislative developments worldwide and reports on relevant case law, bilateral and international agreements as well as regulatory activities of international organisations. Feature articles in this issue include 'Strengthening the international legal framework for nuclear security: Better sooner rather than later'; 'Brexit, Euratom and nuclear proliferation'; and 'McMunn et al. v Babcock and Wilcox Power Generation Group, Inc., et al.: The long road to dismissal'

Continuous fiber ceramic composite. Phase I final report, April 1992--April 1993

Energy Technology Data Exchange (ETDEWEB)

Goettler, R.W.

1995-04-01

Babcock and Wilcox assembled a team to develop the Continuous Fiber Ceramic Composite (CFCC) processing technology, identify the industrial applications, generate design and life prediction software, and to begin the necessary steps leading to full commercialization of CFCC components. Following is a summary of Phase I activities on this program. B&W has selected an oxide-oxide composite system for development and optimization in this program. This selection was based on the results of exposure tests in combustion environments typical of the boiler and heat exchanger applications. Aluminum oxide fiber is the selected reinforcement, and both aluminum oxide and zirconium oxide matrices were selected, based on their superior resistance to chemical attack in hostile industrial service.

RELAP5/MOD2 code assessment

International Nuclear Information System (INIS)

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

1985-01-01

Babcock and Wilcox (B and W) has been working with the code developers at EG and G and the US Nuclear Regulatory Commission in assessing the RELAP5/MOD2 computer code for the past year by simulating selected separate-effects tests. The purpose of this assessment has been to evaluate the code for use in MIST (Ref. 2) and OTIS integral system tests simulations and in the prediction of pressurized water reactor 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 the predictive capability of RELAP5/MOD2. The following versions of the code were evaluated. (1) RELAP/MOD2/Cycle 22 - first released version; (2) YELAP5/Cycle 32 - EG and G test version of RELAP5/MOD2/Cycle 32; (3) RELAP5/MOD2/Cycle 36 - frozen cycle for international code assessment; (4) updates to cycle 36 based on recommendations developed by B and W during the simulation of a Massachusetts Institute of Technology (MIT) pressurizer test; and (5) cycle 36.1 updates received from EG and G

RELAP5/MOD2 code assessment

Energy Technology Data Exchange (ETDEWEB)

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

1985-11-01

Babcock and Wilcox (B and W) has been working with the code developers at EG and G and the US Nuclear Regulatory Commission in assessing the RELAP5/MOD2 computer code for the past year by simulating selected separate-effects tests. The purpose of this assessment has been to evaluate the code for use in MIST (Ref. 2) and OTIS integral system tests simulations and in the prediction of pressurized water reactor 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 the predictive capability of RELAP5/MOD2. The following versions of the code were evaluated. (1) RELAP/MOD2/Cycle 22 - first released version; (2) YELAP5/Cycle 32 - EG and G test version of RELAP5/MOD2/Cycle 32; (3) RELAP5/MOD2/Cycle 36 - frozen cycle for international code assessment; (4) updates to cycle 36 based on recommendations developed by B and W during the simulation of a Massachusetts Institute of Technology (MIT) pressurizer test; and (5) cycle 36.1 updates received from EG and G.

Automatic characterization of loose parts impact damage risk parameters

International Nuclear Information System (INIS)

Glass, S.W.; Phillips, J.M.

1985-01-01

Loose parts caught in the high-velocity flows of the reactor coolant fluid strike against nuclear steam supply system (NSSS) components and can cause significant damage. Loose parts monitor systems (LPMS) have been available for years to detect metal-to-metal impacts. Once detected, however, an assessment of the damage risk potential for leaving the part in the system versus shutting it down and removing the part must be made. The principal parameters used in the damage risk assessment are time delays between the first and subsequent sensor indications (used to assess the impact location) and a correlation between the waveform and the impact energy of the part (how hard the part impacted). These parameters are not well suited to simple automatic techniques. The task has historically been performed by loose parts diagnostic experts who base much of their evaluation on experience and subjective interpretation of impact data waveforms. Three of the principal goals in developing the Babcock and Wilcox (B and W) LPMS-III were (a) to develop an accurate automatic assessment for the time delays, (b) to develop an automatic estimate of the impact energy, and (c) to present the data in a meaningful manner to the operator

An analytical model for prediction of two-phase (noncondensable) flow pump performance

International Nuclear Information System (INIS)

Furuya, O.

1985-01-01

During operational transients or a hypothetical LOCA (loss of coolant accident) condition, the recirculating coolant of PWR (pressurized water reactor) may flash into steam due to a loss of line pressure. Under such two-phase flow conditions, it is well known that the recirculation pump becomes unable to generate the same head as that of the single-phase flow case. Similar situations also exist in oil well submersible pumps where a fair amount of gas is contained in oil. Based on the one dimensional control volume method, an analytical method has been developed to determine the performance of pumps operating under two-phase flow conditions. The analytical method has incorporated pump geometry, void fraction, flow slippage and flow regime into the basic formula, but neglected the compressibility and condensation effects. During the course of model development, it has been found that the head degradation is mainly caused by higher acceleration on liquid phase and deceleration on gas phase than in the case of single-phase flows. The numerical results for head degradations and torques obtained with the model favorably compared with the air/water two-phase flow test data of Babcock and Wilcox (1/3 scale) and Creare (1/20 scale) pumps

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

Science.gov (United States)

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

2017-01-01

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

HEU to LEU fuel conversion. Final report

International Nuclear Information System (INIS)

Mulder, R.U.

1994-10-01

The Nuclear Regulatory Commission issued a ruling, effective March 27, 1986, that all U.S. non-power reactors convert from HEU fuel to LEU fuel. A Reduced Enrichment for Research and Test Reactors Program was conducted by the Department of Energy at Argonne National Laboratory to coordinate the development of the high density LEU fuel and assist in the development of Safety Analysis Reports for the smaller non-power reactors. Several meetings were held at Argonne in 1987 with the non-power reactor community to discuss the conversion and to set up a conversion schedule for university reactors. EG ampersand G at Idaho was assigned the coordination of the fuel element redesigns. The fuel elements were manufactured by the Babcock ampersand Wilcox Company in Lynchburg, Virginia. The University of Virginia was awarded a grant by the DOE Idaho Operations Office in 1988 to perform safety analysis studies for the LEU conversion for its 2 MW UVAR and 100 Watt CAVALIER reactors. The University subsequently decided to shut down the CAVALIER reactor. A preliminary SAR on the UVAR, along with Technical Specification changes, was submitted to the NRC in November, 1990. An updated SAR was approved by the NRC in January, 1991. In September, 1992, representatives from the fuel manufacturer (B ampersand W) and the fuel designer (EG ampersand G, Idaho) came to the UVAR facility to observe trial fittings of new 22 plate LEU mock fuel elements. B ampersand W fabricated two non-fuel bearing elements, a regular 22 plate element and a control rod element. The elements were checked against the drawings and test fitted in the UVAR grid plate. The dimensions were acceptable and the elements fit in the grid plate with no problems. The staff made several suggestions for minor construction changes to the end pieces on the elements, which were incorporated into the final design of the actual fuel elements. Selected papers are indexed separately for inclusion in the Energy Science and Technology

HEU to LEU fuel conversion. Final report

Energy Technology Data Exchange (ETDEWEB)

Mulder, R.U.

1994-10-01

The Nuclear Regulatory Commission issued a ruling, effective March 27, 1986, that all U.S. non-power reactors convert from HEU fuel to LEU fuel. A Reduced Enrichment for Research and Test Reactors Program was conducted by the Department of Energy at Argonne National Laboratory to coordinate the development of the high density LEU fuel and assist in the development of Safety Analysis Reports for the smaller non-power reactors. Several meetings were held at Argonne in 1987 with the non-power reactor community to discuss the conversion and to set up a conversion schedule for university reactors. EG&G at Idaho was assigned the coordination of the fuel element redesigns. The fuel elements were manufactured by the Babcock & Wilcox Company in Lynchburg, Virginia. The University of Virginia was awarded a grant by the DOE Idaho Operations Office in 1988 to perform safety analysis studies for the LEU conversion for its 2 MW UVAR and 100 Watt CAVALIER reactors. The University subsequently decided to shut down the CAVALIER reactor. A preliminary SAR on the UVAR, along with Technical Specification changes, was submitted to the NRC in November, 1990. An updated SAR was approved by the NRC in January, 1991. In September, 1992, representatives from the fuel manufacturer (B&W) and the fuel designer (EG&G, Idaho) came to the UVAR facility to observe trial fittings of new 22 plate LEU mock fuel elements. B&W fabricated two non-fuel bearing elements, a regular 22 plate element and a control rod element. The elements were checked against the drawings and test fitted in the UVAR grid plate. The dimensions were acceptable and the elements fit in the grid plate with no problems. The staff made several suggestions for minor construction changes to the end pieces on the elements, which were incorporated into the final design of the actual fuel elements. Selected papers are indexed separately for inclusion in the Energy Science and Technology Database.

Analysis of the ATR fuel element swaging process

International Nuclear Information System (INIS)

Richins, W.D.; Miller, G.K.

1995-12-01

This report documents a detailed evaluation of the swaging process used to connect fuel plates to side plates in Advanced Test Reactor (ATR) fuel elements. The swaging is a mechanical process that begins with fitting a fuel plate into grooves in the side plates. Once a fuel plate is positioned, a lip on each of two side plate grooves is pressed into the fuel plate using swaging wheels to form the joints. Each connection must have a specified strength (measured in terms, of a pullout force capacity) to assure that these joints do not fail during reactor operation. The purpose of this study is to analyze the swaging process and associated procedural controls, and to provide recommendations to assure that the manufacturing process produces swaged connections that meet the minimum strength requirement. The current fuel element manufacturer, Babcock and Wilcox (B ampersand W) of Lynchburg, Virginia, follows established procedures that include quality inspections and process controls in swaging these connections. The procedures have been approved by Lockheed Martin Idaho Technologies and are designed to assure repeatability of the process and structural integrity of each joint. Prior to July 1994, ATR fuel elements were placed in the Hydraulic Test Facility (HTF) at the Idaho National Engineering Laboratory (AGNAIL), Test Reactor Area (TRA) for application of Boehmite (an aluminum oxide) film and for checking structural integrity before placement of the elements into the ATR. The results presented in this report demonstrate that the pullout strength of the swaged connections is assured by the current manufacturing process (with several recommended enhancements) without the need for- testing each element in the HTF

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

Apollo decommissioning project, Apollo, Pennsylvania. Final technical report

International Nuclear Information System (INIS)

1997-01-01

In November, 1991 Babcock and Wilcox (B and W) received a grant to partially fund the decommissioning of the former Apollo Nuclear Fuel Facility. The decommissioning was performed in accordance with a Nuclear Regulatory Commission (NRC) approved decommissioning plan. This report summarizes the decommissioning of the Apollo Nuclear Fuel Facility and the radiological surveying of the site to demonstrate that these decommissioning activities were effective in reducing residual activity well below NRC's criteria for release for unrestricted use. The Apollo Nuclear Fuel Facility was utilized by the Nuclear Materials and Equipment Corporation (NUMEC) and B and W for nuclear research and production under Atomic Energy Commission and Department of Energy (DOE) contracts during 20 plus years of nuclear fuel manufacturing operations

Status of SOFCo SOFC technology development

Energy Technology Data Exchange (ETDEWEB)

Privette, R.; Perna, M.A.; Kneidel, K. [SOFCo, Alliance, OH (United States)] [and others

1996-12-31

SOFCo, a Babcock & Wilcox/Ceramatec Research & Development Limited Partnership, is a collaborative research and development venture to develop technologies related to planar, solid-oxide fuel cells (SOFCs). SOFCo has successfully demonstrated a kW-class, solid-oxide fuel cell module operating on pipeline natural gas. The SOFC system design integrates the air preheater and the fuel processor with the fuel cell stacks into a compact test unit; this is the platform for multi-kW modules. The cells, made of tape-cast zirconia electrolyte and conventional electrode materials, exhibit excel lent stability in single-cell tests approaching 40,000 hours of operation. Stack tests using 10-cm and 15-cm cells with ceramic interconnects also show good performance and stability in tests for many thousands of hours.

Life extension program initiation at Davis-Besse nuclear power station

International Nuclear Information System (INIS)

Staudinger, Deborah K.

1991-01-01

Davis-Besse is a 900 MW Babcock and Wilcox designed plant located in Northwest Ohio. Effective December 31, 1990, the construction period was recovered making the current license expiration 2007. The economic effects of this extension reduced the depreciation expense for 1990 by $9,790,000 and increased earnings per share by $.04. This positive impact has resulted in an evaluation of pursuing license renewal for Davis-Besse in accordance with the proposed rule on license renewal (10CFR54 'Requirements for renewal of operating licenses for nuclear power plants'). This paper reviews preliminary efforts to evaluate these actions and summarizes strategies planned to ensure continued operation of Davis-Besse remains a viable option for base load generation for Toledo Edison. (author)

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

International Nuclear Information System (INIS)

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

2000-01-01

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

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

International Nuclear Information System (INIS)

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

1996-01-01

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

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

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

International Nuclear Information System (INIS)

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

2011-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2011-07-01

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

An investigation of differences between measured and calculated bucklings of a series of light water and heavy water moderated experimental cores

International Nuclear Information System (INIS)

Figgins, A.J.G.

1966-02-01

A series of light water and light and heavy water moderated exponential and critical experiments performed by the Babcock and Wilcox Company were analysed using the METHUSELAH programme and it was found that the calculated and measured critical bucklings differed significantly. The effect was most marked as the temperature of the moderator was raised in the light water cores where it amounted to 10 m -2 for a 200 deg. C rise above room temperature. Of this discrepancy 3 m -2 , at the most, could be explained as being caused by the experimental cores not being large enough to have a central asymptotic region, leaving an unexplained difference of 7 m -2 . It is suggested that the only region in which METHUSELAH could be usefully modified to improve this agreement is in the calculation of the resonance escape probability. The last section of the report compares the calculated and measured results obtained at room temperatures. (author)

Remote controlled stud bolt handling device for reactor pressure vessel

International Nuclear Information System (INIS)

Shindo, Takenori; Shigehiro, Katsuya; Ito, Morio; Okada, Kenji

1988-01-01

In nuclear power stations, at the time of regular inspection, the works of opening and fixing the upper covers of reactor pressure vessels are carried out for inspecting the inside of reactor pressure vessels and exchanging fuel rods. These upper covers are fastened with many stud bolts, therefore, the works of opening and fixing require a large amount of labor, and are done under the restricted condition of wearing protective clothings and masks. Babcock Hitachi K.K. has completed the development of a remotely controlled automatic bolt tightenig device for this purpose, therefore, its outline is reported. The conventional method of these works and the problems in it are described. The design of the new device aimed at the parallel execution of cleaning screw threads, loosening and tightening nuts, and taking off and putting on nuts and washers, thus contributing to the shortening of regular inspection period, the reduction of the radiation exposure of workers, and the decrease of the number of workers. The function, reliability and endurance of the new device were confirmed by the verifying test using a device made for trial. The device is composed of a stand, a rail and four stations each with a cleaning unit, a stud tensioner and a nut handling unit. (K.I.)

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

International Nuclear Information System (INIS)

Thiele, R.; Horn, B.; Coates, C.W.; Stainback, J.R.

2001-01-01

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

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).

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...

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

International Nuclear Information System (INIS)

Russell, E.W.; Nelson, T.A.; Domian, H.A.; LaCount, D.F.; Robitz, E.S.; Stein, K.O.

1989-04-01

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

Fabrication and closure development of corrosion resistant containers for Nevada's Yucca Mountain high-level nuclear waste repository

International Nuclear Information System (INIS)

Russell, E.W.; Nelson, T.A.; Domian, H.A.; LaCount, D.F.; Robitz, E.S.; Stein, K.O.

1989-11-01

US Congress and the President have determined that the Yucca Mountain site in Nevada is to be characterized to determine its suitability for construction of the first US high-level nuclear waste repository. Work in connection with this site is carried out within the Yucca Mountain Project (YMP). Lawrence Livermore National Laboratory (LLNL) has the responsibility for designing, developing, and projecting the performance of the waste package for the permanent storage of high-level nuclear waste. Babcock ampersand Wilcox (B ampersand W) is involved with the YMP as a subcontractor to LLNL. B ampersand W's role is to recommend and demonstrate a method for fabricating the metallic waste container and a method for performing the final closure of the container after it has been filled with waste. Various fabrication and closure methods are under consideration for the production of containers. This paper presents progress to date in identifying and evaluating the candidate manufacturing processes. 2 refs., 2 figs., 4 tabs

TACO: fuel pin performance analysis

International Nuclear Information System (INIS)

Stoudt, R.H.; Buchanan, D.T.; Buescher, B.J.; Losh, L.L.; Wilson, H.W.; Henningson, P.J.

1977-08-01

The thermal performance of fuel in an LWR during its operational lifetime must be described for LOCA analysis as well as for other safety analyses. The determination of stored energy in the LOCA analysis, for example, requires a conservative fuel pin thermal performance model that is capable of calculating fuel and cladding behavior, including the gap conductance between the fuel and cladding, as a function of burnup. The determination of parameters that affect the fuel and cladding performance, such as fuel densification, fission gas release, cladding dimensional changes, fuel relocation, and thermal expansion, should be accounted for in the model. Babcock and Wilcox (B and W) has submitted a topical report, BAW-10087P, December 1975, which describes their thermal performance model TACO. A summary of the elements that comprise the TACO model and an evaluation are presented

Dual fuel gradients in uranium silicide plates

Energy Technology Data Exchange (ETDEWEB)

Pace, B.W. [Babock and Wilcox, Lynchburg, VA (United States)

1997-08-01

Babcock & Wilcox has been able to achieve dual gradient plates with good repeatability in small lots of U{sub 3}Si{sub 2} plates. Improvements in homogeneity and other processing parameters and techniques have allowed the development of contoured fuel within the cladding. The most difficult obstacles to overcome have been the ability to evaluate the bidirectional fuel loadings in comparison to the perfect loading model and the different methods of instilling the gradients in the early compact stage. The overriding conclusion is that to control the contour of the fuel, a known relationship between the compact, the frames and final core gradient must exist. Therefore, further development in the creation and control of dual gradients in fuel plates will involve arriving at a plausible gradient requirement and building the correct model between the compact configuration and the final contoured loading requirements.

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

Science.gov (United States)

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

2015-12-01

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

Remerschen nuclear power station with BBR pressurized water reactor

International Nuclear Information System (INIS)

Hoffmann, J.P.

1975-01-01

On the basis of many decades of successful cooperation in the electricity supply sector with the German RWE utility, the Grand Duchy of Luxemburg and RWE jointly founded Societe Luxembourgeoise d'Energie Nucleaire S.A. (SENU) in 1974 in which each of the partners holds a fifty percent interest. SENU is responsible for planning, building and operating this nuclear power station. Following an international invitation for bids on the delivery and turnkey construction of a nuclear power station, the consortium of the German companies of Brown, Boveri and Cie. AG (BBC), Babcock - Brown Boveri Reaktor GmbH (BBR) and Hochtief AG (HT) received a letter of intent for the purchase of a 1,300 MW nuclear power station equipped with a pressurized water reactor. The 1,300 MW station of Remerschen will be largely identical with the Muelheim-Kaerlich plant under construction by the same consortium near Coblence on the River Rhine since early 1975. According to present scheduling, the Remerschen nuclear power station could start operation in 1981. (orig.) [de

AGE RELATED DEGRADATION OF STEAM GENERATOR INTERNALS BASED ON INDUSTRY RESPONSES TO GENERIC LETTER 97-06

International Nuclear Information System (INIS)

SUBUDHI, M.; SULLIVAN, JR. E.J.

2002-01-01

THIS PAPER PRESENTS THE RESULTS OF AN AGING ASSESSMENT OF THE NUCLEAR POWER INDUSTRY RESPONSES TO NRC GENERIC LETTER 97-06 ON THE DEGRADATION OF STEAM GENERATOR INTERNALS EXPERIENCED AT ELECTRICITE DE FRANCE (EDF) PLANTS IN FRANCE AND AT A UNITED STATES PRESSURIZED WATER REACTOR (PWR). WESTINGHOUSE (W), COMBUSTION ENGINEERING (CE), AND BABCOCK AND WILCOX (BW) STEAM GENERATOR MODELS, CURRENTLY IN SERVICE AT U.S. NUCLEAR POWER PLANTS, POTENTIALLY COULD EXPERIENCE DEGRADATION SIMILAR TO THATFOUND AT EDF PLANTS AND THE U.S. PLANT. THE STEAM GENERATORS IN MANY OF THE U.S. PWRS HAVE BEEN REPLACED WITH STEAM GENERATORS WITH STEAM GENERATORS WITH IMPROVED DESIGNS AND MATERIALS. THESE REPLACEMENT STEAM GENERATORS HAVE BEEN MANUFACTURED IN THE U.S. AND ABROAD. DURING THIS ASSESSMENT, EACH OF THE THREE OWNERS GROUPS (W,CE, AND BW) IDENTIFIED FOR ITS STEAM GENERATOR, MODELS ALL THE POTENTIAL INTERNAL COMPONENTS THAT ARE VULNERABLE TO DEGRADATION WHILE IN SERVICE. EACH OWNERS GROUPDEVELOPED INSPEC TION AND MONITORING GUIDANCE AND RECOMMENDATIONS FOR ITS PARTICULAR STEAM GENERATOR MODELS. THE NUCLEAR ENERGY INSTITUTE INCORPORATED IN NEI 97-06 STEAM GENERATOR PROGRAM GUIDELINES, A REQUIREMENT TO MONITOR SECONDARY SIDE STEAM GENERATOR COMPONENTS IF THEIR FAILURE COULD PREVENT THE STEAM GENERATOR FROM FULFILLING ITS INTENDED SAFETY-RELATED FUNCTION. LICENSEES INDICATED THAT THEY IMPLEMENTED OR PLANNED TO IMPLEMENT, AS APPROPRIATE FOR THEIR STEAM GENERATORS, THEIR OWNERS GROUPRECOMMENDATIONS TO ADDRESS THE LONG-TERM EFFECTS OF THE POTENTIAL DEGRADATION MECHANISMS ASSOCIATED WITH THE STEAM GENERATOR INTERNALS

Safety of Ikata Nuclear Power Station from the accident of Three Mile Island

International Nuclear Information System (INIS)

Nonaka, Hiroshi

1979-01-01

The leak of radioactive substances occurred on March 28, 1979, in the No. 2 plant of Three Mile Island Nuclear Power Station, and this accident must be put to use to prevent similar accidents and to secure safety hereafter in the nuclear power stations being operated in Japan. In the TMI accident, too many problems concerning the operation management seemed to exist in a series of events. In this paper, a few matters related to the TMI accident among the aspects of the operation management in Ikata Nuclear Power Station are reported. As the problems of operation management, it is considered that the operation of the TMI plant was continued as the exit valve of auxiliary feed line was closed, that it took long time to close the root valve for a pressurizer relief valve manually, and that the ECCS was stopped manually. In TMI, the abnormal phenomenon of losing main feed water has occurred 6 times since the attainment of criticality in March, 1978, and the opening and sticking of pressurizer relief valves occurred at least twice in about 150 times of their actuation in the nuclear reactors designed by Babcock and Wilcox Co. In Ikata Nuclear Power Station, these problems are detected early and the suitable measures are taken immediately, therefore it never happens to continue the operation as the problems are left as they are. It is not conceivable that similar troubles occur many times. (Kako, I.)

Noise and vibration analysis system

International Nuclear Information System (INIS)

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

1985-01-01

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

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

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

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

Science.gov (United States)

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

2012-03-30

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

Brayton Point coal conversion project (NEPCO)

Energy Technology Data Exchange (ETDEWEB)

Sullivan, W.F. Jr.

1982-05-01

The New England Power Company (NEPCO) recently converted Brayton Point Power Station Units 1, 2, and 3 from oil to coal. The coal conversion project is the largest coal conversion project in the nation to date. Stone and Webster Engineering Corporation (SWEC) was hired as the engineer/constructor for the project. Units 1 and 2 are 250-MW Combustion Engineering boilers, and Unit 3 is a 650-MW Babcock and Wilcox boiler. All three units were originally designed to burn pulverized coal but were converted to oil during the years of low oil prices. Studies performed by NEPCO and SWEC indicated that the areas discussed in the following paragraphs required upgrading before the units could efficiently burn coal and meet Federal and State environmental requirements. All units have been converted and are operating. This paper discusses design modifications required to burn coal, startup, and initial operating problems, and solutions.

Commercialization of the modular modeling system

International Nuclear Information System (INIS)

Bartells, P.S.

1987-01-01

The Modular Modeling System (MMS) was developed by the Electric Power Research Institute (EPRI) to provide an efficient, economical, and user-friendly computer code for use in the analysis of the dynamic performance of nuclear and fossil power plants. The MMS has been under development since 1978 and is now ready for general release to the industry. To most effectively transfer the MSS technology on a worldwide basis, EPRI has licensed the Babcock and Wilcox Company (B and W) to act as the MMS code coordinator and commercial licensing agent. B and W is responsible for promoting the use of MMS and assisting users in the effective utilization of this new analysis tool. The history behind the commercialization of the MMS and the mechanism by which the MMS technology will be effectively transferred to provide for a self-sustaining and technically current analysis tool are outlined

On-line chemistry monitoring for the secondary side

International Nuclear Information System (INIS)

Anon.

1990-01-01

Babcock and Wilcox (B and W) has developed a computerized water chemistry data acquisition and management system for nuclear plant secondary coolant systems. The Integrated Water Chemistry Monitoring System (IWCMS) provides on-line monitoring of conditions and rapid trend analysis of sampled data. So far it has been installed at GPU Three Mile Island unit 1 and at Toledo Edison Davis-Besse. The IWCMS meets the following utility needs for monitoring power plant chemistry: control of chemistry conditions to minimize corrosion and extend component/system life; continuous analysis of data from on-line detectors and grab samples; expediting of transient recovery actions with trend, alarm and evaluation capability; provision for rapid sharing of useful operational chemistry information; concentration of attention on evaluation instead of data manipulation. The system is composed of three functional parts: data acquisition hardware; PC-based computer system and customised system software. (author)

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).

Advanced Emissions Control Development Program

Energy Technology Data Exchange (ETDEWEB)

A.P.Evans; K.E. Redinger; M.J. Holmes

1998-04-01

The objective of the Advanced Emissions Control Development Program (AECDP) is to develop practical, cost-effective strategies for reducing the emissions of air toxics from coal-fired boilers. Ideally, the project aim is to effectively control air toxic emissions through the use of conventional flue gas cleanup equipment such as electrostatic precipitators (ESPS), fabric filters (baghouse), and wet flue gas desulfurization. Development work to date has concentrated on the capture of mercury, other trace metals, fine particulate and hydrogen chloride. Following the construction and evaluation of a representative air toxics test facility in Phase I, Phase II focused on the evaluation of mercury and several other air toxics emissions. The AECDP is jointly funded by the United States Department of Energy's Federal Energy Technology Center (DOE), the Ohio Coal Development Office within the Ohio Department of Development (oCDO), and Babcock& Wilcox-a McDermott company (B&W).

A two-parameter preliminary optimization study for a fluidized-bed boiler through a comprehensive mathematical simulator

Energy Technology Data Exchange (ETDEWEB)

Rabi, Jose A.; Souza-Santos, Marcio L. de [Universidade Estadual de Campinas, SP (Brazil). Faculdade de Engenharia Mecanica. Dept. de Energia]. E-mails: jrabi@fem.unicamp.br; dss@fem.unicamp.br

2000-07-01

Modeling and simulation of fluidized-bed equipment have demonstrated their importance as a tool for design and optimization of industrial equipment. Accordingly, this work carries on an optimization study of a fluidized-bed boiler with the aid of a comprehensive mathematical simulator. The configuration data of the boiler are based on a particular Babcock and Wilcox Co. (USA) test unit. Due to their importance, the number of tubes in the bed section and the air excess are chosen as the parameters upon which the optimization study is based. On their turn, the fixed-carbon conversion factor and the boiler efficiency are chosen as two distinct optimization objectives. The results from both preliminary searches are compared. The present work is intended to be just a study on possible routes for future optimization of larger boilers. Nonetheless, the present discussion might give some insight on the equipment behavior. (author)

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

International Nuclear Information System (INIS)

Tapp, P.A.

1992-04-01

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

Preliminary assessment of the benefits of derating a cask for increasing age/burnup capability

International Nuclear Information System (INIS)

Broadhead, B.L.; Parks, C.V.; Joy, D.S.; Tang, J.S.

1992-01-01

This paper discusses a study performed to determine the extent to which the age/burnup capability of the Babcock and Wilcox BR-100 rail cask could be extended by reducing the number of fuel assemblies. Only the shielding effects of derating are accounted for in this study. Separate analyses will be necessary to address the enhanced heat loads due to increased burnup or decreased age. The criterion used to assess the derating was the calculated dose 2 m from the rail car. The reference calculations were based on the 70% design of the BR-100 cask with 21 PWR fuel assemblies. Seven different basket/assembly loading configurations were investigated. The results indicate that both an alternative 18-assembly basket configuration and a 17-assembly/4-empty-hole configuration for the 21-element basket offer substantial gains over the fully loaded reference 21-element basket configuration

Timing analysis of PWR fuel pin failures

International Nuclear Information System (INIS)

Jones, K.R.; Wade, N.L.; Katsma, K.R.; Siefken, L.J.; Straka, M.

1992-09-01

Research has been conducted to develop and demonstrate a methodology for calculation of the time interval between receipt of the containment isolation signals and the first fuel pin failure for loss-of-coolant accidents (LOCAs). Demonstration calculations were performed for a Babcock and Wilcox (B ampersand W) design (Oconee) and a Westinghouse (W) four-loop design (Seabrook). Sensitivity studies were performed to assess the impacts of fuel pin bumup, axial peaking factor, break size, emergency core cooling system availability, and main coolant pump trip on these times. The analysis was performed using the following codes: FRAPCON-2, for the calculation of steady-state fuel behavior; SCDAP/RELAP5/MOD3 and TRACPF1/MOD1, for the calculation of the transient thermal-hydraulic conditions in the reactor system; and FRAP-T6, for the calculation of transient fuel behavior. In addition to the calculation of fuel pin failure timing, this analysis provides a comparison of the predicted results of SCDAP/RELAP5/MOD3 and TRAC-PFL/MOD1 for large-break LOCA analysis. Using SCDAP/RELAP5/MOD3 thermal-hydraulic data, the shortest time intervals calculated between initiation of containment isolation and fuel pin failure are 10.4 seconds and 19.1 seconds for the B ampersand W and W plants, respectively. Using data generated by TRAC-PF1/MOD1, the shortest intervals are 10.3 seconds and 29.1 seconds for the B ampersand W and W plants, respectively. These intervals are for a double-ended, offset-shear, cold leg break, using the technical specification maximum peaking factor and applied to fuel with maximum design bumup. Using peaking factors commensurate widi actual bumups would result in longer intervals for both reactor designs. This document also contains appendices A through J of this report

An evaluation of the defence in depth concept in the light of the Three Mile Island accident

International Nuclear Information System (INIS)

Fuchs, H.; Gibbs, K.P.

1981-01-01

One of the fundamental bases of reactor safety has been the 'defence in depth' concept. Valuable lessons for plants outside the USA can be learned by studying both the strengths and weaknesses of all the various lines of defence in the light of the TMI findings. These lessons can be summarized as follows: Quality: lack of quality led to the initiation of the event through frequent trips, one of which eventually led to the sticking open of the PORV. It also led to wrong diagnosis of the fault. Multiple failures during the many hours the incident lasted distracted the operator and made a careful diagnosis of its causes and progress impossible. Reactor protection system: the Babcock and Wilcox PWR, with its once-through steam generators (OTSG), responds more rapidly than other LWRs to some operating transients. This does not necessarily result in a less safe system provided that the reactor protection system is designed with the fast response in mind, e.g. by providing anticipatory trips. This emphasizes the importance of a systems approach to design rather than too much reliance on codes and standards, which may not recognize important plant differences. Safety systems: For similar reasons the low water content of the OTSG could have been compensated for by a higher reliability in the emergency feed system. In US practice the operator plays an important role in aligning safety systems, whereas in Europe there is more extensive automation. The latter system leads to an analysis of a wider spectrum of faults to ensure that the automatics respond correctly in all circumstances. Containment: The TMI incident showed that the containment can be very valuable for situations where far more core damage and hydrogen production occurs than is predicted for design basis accidents. Continuing containment integrity depends on core cooling and on the correct operation or isolation of many other systems, whose role in such accidents has not always been recognized

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.

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

International Nuclear Information System (INIS)

Li Guangjian; Zhao Zengqiao

1992-01-01

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

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).

Development and commissioning of a links type inspection manipulator

International Nuclear Information System (INIS)

Whitcomb, H.M.; Ford, G.T.

1988-01-01

In order to access a number of tortuous inspection routes in the Heysham II/Torness reactors, the Central Electricity Generating Board (CEGB) Remote Inspection Group (RIG) have developed a 'links' type manipulator in conjunction with Babcock Power Ltd. This manipulator operates from the reactor pile cap and allows placement of a television camera within the pressure vessel up to 4m laterally from a standpipe of only 140mm in diameter. The design features of this novel manipulator are described together with the solutions developed to overcome particular difficulties experienced during commissioning. (author)

Laboratory evaluation of alloy X-750 clevis bolts removed from D.C. cook unit 1

International Nuclear Information System (INIS)

Hyres, J.; Xu, H.; Kalchik, K.; Thompson, G.

2015-01-01

This paper summarizes the results of the laboratory evaluation performed by Babcock and Wilcox on Alloy X-750 clevis bolts removed from the Lower Radial Support System (LRSS) at D.C. Cook Unit 1. A total of 29 clevis bolts - 16 broken and 13 intact - were provided for laboratory analysis and testing to document the extent of degradation, evaluate the integrity of the intact bolts, and identify the bolt degradation/failure mechanism(s). The laboratory work scope included visual and stereo-visual examinations of all bolts. Based on the results of these examinations, four bolts - two broken and two intact - were selected for more detailed analysis/testing, including Scanning Electron Microscopy (SEM), Energy Dispersive Spectroscopy (EDS), optical metallography, microhardness, chemical analysis by Inductively Coupled Plasma-Mass Spectroscopy (ICP-MS), Rockwell hardness testing, and tensile testing. The laboratory data indicated the bolts failed by intergranular stress corrosion cracking (IGSCC). There was no evidence that the bolts failed due to fatigue cracking or mechanical overload. (authors)

SGN-Reseau Eurisys participates to the Hanford military site rehabilitation

International Nuclear Information System (INIS)

Anon.

1996-01-01

Numatec Hanford Corporation, a subsidiary company of SGN-Reseau Eurisys and Cogema, gained with Fluor Daniel the contract for the rehabilitation of the old military nuclear centre of Hanford (Washington, USA). This contract of 5 years represents 5 billions of US dollars with 300 millions of dollars for the French part. This short paper gives a general description of the Hanford installations and of the partners involved in the contract: Fluor Daniel consortium, Lockheed Martin, Babcock and Wilcox, Duke Engineering and Services, Rust Federal Services, Numatec Hanford Corporation (NHC), SGN-Eurisys Services Corporation (SESC). The schedule comprises: the stabilisation of the residual plutonium in all installations before December 1999, the removal of muds and debris from the K storage pool of irradiated fuels before June 2000, the draining and cleaning of the high activity storage tanks before December 2001 and the general decontamination of the installations up to the year 2005. (J.S.)

The use of burnup credit for spent fuel cask design

International Nuclear Information System (INIS)

Lake, W.H.

1993-01-01

A new generation of high capacity spent fuel transport casks is being developed by the U.S. Department of Energy (DOE) as part of the Federal Waste Management System (FWMS). Burnup credit, which recognizes the reduced reactivity of spent fuel is being used for these casks. Two cask designs being developed for DOE by Babcock and Wilcox and General Atomics use burnup credit. The cask designs must be certified by the U.S. Nuclear Regulatory Commission (NRC) if they are to be used in the FWMS. Certification of these casks by the NRC would not require any change in the NRC's transport regulations, and would be consistent with past practices. Furthermore, use of burnup credit casks appears to be consistent with current International Atomic Energy Agency (IAEA) rules and regulations. To support NRC certification, DOE has identified the technical issues related to burnup credit, and embarked on a development program to resolve them. (J.P.N.)

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.

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.

Laryngopharyngeal Reflux: Diagnosis, Treatment, and Latest Research

Directory of Open Access Journals (Sweden)

Campagnolo, Andrea Maria

2014-01-01

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

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

International Nuclear Information System (INIS)

VISWANATH, R.S.

1999-01-01

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

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

DEFF Research Database (Denmark)

Petersen, Lars Jelstrup; Skov, P S

1999-01-01

on clinical and biochemical EAR and LPR in human skin. METHODS: Measurement of wheal and flare reactions to allergen, codeine, and histamine, and LPR (induration) to allergen. Assessment of histamine and prostaglandin D2 (PGD2) release by microdialysis technique in EAR, and measurement of mediators in LPR......, myeloperoxidase, or eosinophil cationic protein in LPR. CONCLUSIONS: Salmeterol and salbutamol inhibited allergen-induced skin responses, and reduced mediator release in EAR but not LPR. In general, the anti-inflammatory effects of salmeterol did not differ from those induced by salbutamol....

Upgrading the electrical system of the IEA-R1 reactor to avoid triggering event of accidents

International Nuclear Information System (INIS)

Mello, Jose Roberto de; Madi Filho, Tufic

2015-01-01

The IEA-R1 research reactor at the Institute of Energy and Nuclear Research (IPEN) is a research reactor open pool type, built and designed by the American firm 'Babcox and Wilcox', having as coolant and moderator demineralized light water and Beryllium and graphite, as reflectors. The power supply system is designed to meet the electricity demand required by the loads of the reactor (Security systems and systems not related to security) in different situations the plant can meet, such as during startup, normal operation at power, shutdown, maintenance, exchange of fuel elements and accident situations. Studies have been done on possible accident initiating events and deterministic techniques were applied to assess the consequences of such incidents. Thus, the methods used to identify and select the accident initiating events, the methods of analysis of accidents, including sequence of events, transient analysis and radiological consequences, have been described. Finally, acceptance criteria of radiological doses are described. Only a brief summary of the item concerning loss of electrical power will be presented. The loss of normal electrical power at the IEA-R1 reactor is very common. In the case of Electric External Power Loss, at the IEA-R1 reactor building, there may be different sequences of events, as described below. When the supply of external energy in the IEA-R1 facility fails, the Electrical Distribution Vital System, consisting of 4 (four) generators type 'UPS', starts operation, immediately and it will continue supplying power to the reactor control table, core cooling system and other security systems. To contribute to security, in the electric power failure, starts to operate the Emergency Cooling System (SRE). SRE has the function of removing residual heat from the core to prevent the melting of fuel elements in the event of loss of refrigerant to the core. Adding to the generators with batteries group system, new auxiliary

Significant role of Fas ligand-binding but defective Fas receptor (CD95) in lymph node hyperplasia composed of abnormal double-negative T cells

Science.gov (United States)

Matsuzawa, Akio; Shimizu, Motomu; Takeda, Yasutaka; Nagase, Hisashi; Sayama, Kazutoshi; Kimura, Mikio

2002-01-01

The functional differences between two mutations of the Fas (CD95) locus, Faslpr (lpr) and Faslprcg (lprcg), were investigated using bone marrow (BM) transplantation on the C3H mouse background. Both lpr/lpr and lprcg/lprcg BM transferred caused lymph node (LN) hyperplasia in lpr/+ and lprcg/+ recipients, although it was clearly smaller than that in lpr/lpr and lprcg/lprcg recipients of lpr/lpr and lprcg/lprcg BM. In addition, both BM induced significantly larger LN hyperplasia in lprcg/+ than lpr/+ recipients. Appearance of CD4− CD8−[double negative (DN)] T cells in the periphery is the most consistent phenotype of Fas mutations. Importantly, the proportion of DN T cells was higher in larger LN hyperplasia in the order of lpr/+, lprcg/+ and lpr/lpr or lprcg/lprcg recipients. On the other hand, both lpr/lpr and lprcg/lprcg BM transferred into wild-type (+/+) mice caused marked LN atrophy. The former, but not the latter, induced wasting syndrome. Faslg1d (gld)-homozygous lpr/lpr BM transferred into +/+ mice elicited LN hyperplasia of the same extent as that in lpr/lpr mice transferred with lpr/lpr BM, but not wasting syndrome. Taken together with the fact that DN T cells massively express Fas ligand (FasL), this study implied that FasL overexpressed on DN cells may be involved in the accumulation of DN T cells in LN, LN atrophy and wasting syndrome, and that lprcg Fas, which can bind to Fas ligand but not transduce apoptosis signal into cells, may modulate these pathological conditions by interfering with the binding of FasL to Fas. PMID:12153509

Evaluating the use and limitations of the Danish National Patient Register in register-based research using an example of multiple sclerosis.

Science.gov (United States)

Mason, K; Thygesen, L C; Stenager, E; Brønnum-Hansen, H; Koch-Henriksen, N

2012-03-01

The Danish National Patient Register, Landspatientregistret (LPR), is a register of all hospital discharges and outpatient treatments in Denmark. It is increasingly used in research so it is important to understand to what extent this can be used as an accurate source of information. Virtually all patients in Denmark with multiple sclerosis (MS) are reported to the Combined MS Registry (DMSR), so this was used as the standard which the LPR was compared against. All residents of Denmark are assigned a unique Civil Register (CPR) number; this was used to compare data between registers. The LPR completeness was estimated by the proportion of cases from the DMSR that could be retrieved from the LPR. The LPR validity was estimated by the proportion of cases, listed in the LPR and DMSR, in whom the MS diagnosis could be confirmed as definite/probable/possible by the DMSR. We found that 86.9% of those who were DMSR listed with an approved MS diagnosis were also listed in the LPR with a MS diagnosis. The diagnosis was valid in 96.3% of patients listed in the LPR when compared against the DMSR. The low completeness reduces the usefulness of the LPR in epidemiological MS research, in particular incidence studies. The study also found that the completeness of the LPR could be increased to 92.8% by including LPR records from other departments in addition, but this reduced the validity of the LPR to 95.1%. However, these results cannot uncritically be applied to registration of other diseases in the LPR. © 2011 John Wiley & Sons A/S.

Fabrication of Uranium Oxycarbide Kernels for HTR Fuel

International Nuclear Information System (INIS)

Barnes, Charles; Richardson, Clay; Nagley, Scott; Hunn, John; Shaber, Eric

2010-01-01

Babcock and Wilcox (B and W) has been producing high quality uranium oxycarbide (UCO) kernels for Advanced Gas Reactor (AGR) fuel tests at the Idaho National Laboratory. In 2005, 350-(micro)m, 19.7% 235U-enriched UCO kernels were produced for the AGR-1 test fuel. Following coating of these kernels and forming the coated-particles into compacts, this fuel was irradiated in the Advanced Test Reactor (ATR) from December 2006 until November 2009. B and W produced 425-(micro)m, 14% enriched UCO kernels in 2008, and these kernels were used to produce fuel for the AGR-2 experiment that was inserted in ATR in 2010. B and W also produced 500-(micro)m, 9.6% enriched UO2 kernels for the AGR-2 experiments. Kernels of the same size and enrichment as AGR-1 were also produced for the AGR-3/4 experiment. In addition to fabricating enriched UCO and UO2 kernels, B and W has produced more than 100 kg of natural uranium UCO kernels which are being used in coating development tests. Successive lots of kernels have demonstrated consistent high quality and also allowed for fabrication process improvements. Improvements in kernel forming were made subsequent to AGR-1 kernel production. Following fabrication of AGR-2 kernels, incremental increases in sintering furnace charge size have been demonstrated. Recently small scale sintering tests using a small development furnace equipped with a residual gas analyzer (RGA) has increased understanding of how kernel sintering parameters affect sintered kernel properties. The steps taken to increase throughput and process knowledge have reduced kernel production costs. Studies have been performed of additional modifications toward the goal of increasing capacity of the current fabrication line to use for production of first core fuel for the Next Generation Nuclear Plant (NGNP) and providing a basis for the design of a full scale fuel fabrication facility.

A fuzzy-based reliability approach to evaluate basic events of fault tree analysis for nuclear power plant probabilistic safety assessment

International Nuclear Information System (INIS)

Purba, Julwan Hendry

2014-01-01

Highlights: • We propose a fuzzy-based reliability approach to evaluate basic event reliabilities. • It implements the concepts of failure possibilities and fuzzy sets. • Experts evaluate basic event failure possibilities using qualitative words. • Triangular fuzzy numbers mathematically represent qualitative failure possibilities. • It is a very good alternative for conventional reliability approach. - Abstract: Fault tree analysis has been widely utilized as a tool for nuclear power plant probabilistic safety assessment. This analysis can be completed only if all basic events of the system fault tree have their quantitative failure rates or failure probabilities. However, it is difficult to obtain those failure data due to insufficient data, environment changing or new components. This study proposes a fuzzy-based reliability approach to evaluate basic events of system fault trees whose failure precise probability distributions of their lifetime to failures are not available. It applies the concept of failure possibilities to qualitatively evaluate basic events and the concept of fuzzy sets to quantitatively represent the corresponding failure possibilities. To demonstrate the feasibility and the effectiveness of the proposed approach, the actual basic event failure probabilities collected from the operational experiences of the David–Besse design of the Babcock and Wilcox reactor protection system fault tree are used to benchmark the failure probabilities generated by the proposed approach. The results confirm that the proposed fuzzy-based reliability approach arises as a suitable alternative for the conventional probabilistic reliability approach when basic events do not have the corresponding quantitative historical failure data for determining their reliability characteristics. Hence, it overcomes the limitation of the conventional fault tree analysis for nuclear power plant probabilistic safety assessment

Analysis of vertical flow during ambient and pumped conditions in four monitoring wells at the Pantex Plant, Carson County, Texas, July-September 2008

Science.gov (United States)

Stanton, Gregory P.; Thomas, Jonathan V.; Stoval, Jeffery

2009-01-01

The Pantex Plant is a U.S. Department of Energy/National Nuclear Security Administration (USDOE/NNSA)-owned, contractor-operated facility managed by Babcock & Wilcox Technical Services Pantex, LLC (B&W Pantex) in Carson County, Texas, approximately 17 miles northeast of Amarillo. The U.S. Geological Survey, in cooperation with B&W Pantex through the USDOE/NNSA, made a series of flowmeter measurements and collected other borehole geophysical logs during July–September 2008 to analyze vertical flow in screened intervals of four selected monitoring wells (PTX01–1012, PTX06–1044, PTX06–1056, and PTX06–1068) at the Pantex Plant. Hydraulic properties (transmissivity values) of the section of High Plains (Ogallala) aquifer penetrated by the wells also were computed. Geophysical data were collected under ambient and pumped flow conditions in the four monitoring wells. Unusually large drawdowns occurred at two monitoring wells (PTX06–1044 and PTX06–1056) while the wells were pumped at relatively low rates. A decision was made to redevelop those wells, and logs were run again after redevelopment in the two monitoring wells.

Heavy-Section Steel Irradiation Program

International Nuclear Information System (INIS)

Anon.

1991-01-01

In FY1990 the Heavy-Section Steel Irradiation (HSSI) Program was arranged into 8 tasks: (1) program management, (2) K Ic curve shift in high-copper welds, (3) K Ia curve shift in high-copper welds, (4) irradiation effects on cladding, (5) K Ic and K Ia curve shifts in low upper-shelf (LUS) welds, (6) irradiation effects in a commercial LUS weld, (7) microstructural analysis of irradiation effects, and (8) in-service aged material evaluations. Of particular interest are the efforts in FY1990 concerning the shifts in fracture toughness and crack arrest toughness in high-copper welds, the unirradiated examination of a LUS weld from the Midland reactor, and the continued investigation into the causes of accelerated low-temperature embrittlement recently observed in RPV support steels. In the Fifth and Sixth Irradiation Series, designed to examine the shifts and possible changes in shape in the ASME K Ic and K Ia curves for two irradiated high-copper welds, it was seen that both the lower bound and mean fracture toughness shifts were greater than those of the associated Charpy-impact energies, whereas the shifts in crack arrest toughness were comparable. The irradiation-shifted fracture toughness data fell slightly below the appropriately indexed ASME K Ic curve even when it was shifted according to Revision 2 of Regulatory Guide 1.99 including its margins. The beltline weld, which was removed from the Midland reactor, fabricated by Babcock and Wilcox, Co. using Linde 80 flux, is being examined in the Tenth Irradiation Series to establish the effects of irradiation on a commercial LUS weld. A wide variation in the unirradiated fracture properties of the Midland weld were measured with values of RT NDT ranging from -22 to 54F through its thickness. In addition, a wide range of copper content from 0.21 to 0.45 wt % was found, compared to the 0.42 wt % previously reported

Balance-of-plant outage availability study. Phase I. Extension report. [PWR

Energy Technology Data Exchange (ETDEWEB)

Thomasson, F.R.

1978-09-01

After completion of the Phase 1 Refueling Outage Availability Study, Babcock and Wilcox and the U.S. Department of Energy entered into a supplemental agreement to perform a balance-of-plant maintenance, inspection, and test study with the cooperation of Duke Power Company and Arkansas Power and Light Company. The objectives were (1) to expand the Phase 1 data base, including balance-of-plant activities, to reduce outage time and increase plant availability and (2) to conduct an onsite review of plant maintenance, practices to complement the utility efforts in reducing outage time and increasing on-line operational time. Data were obtained from (1) observations during the 1977 refueling outage at Oconee 3, (2) review of maintenance practices during the Arkansas Nuclear One, Unit 1, operational cycle in 1977, and (3) selected observations of the 1978 refueling outage at ANO-1. Accumulated data were then reviewed and analyzed to produce a list of improvement recommendations for Oconee 3 and ANO-1 that can be generically applied to plants of similar design and construction.

Balance-of-plant outage availability study. Phase I. Extension report

International Nuclear Information System (INIS)

Thomasson, F.R.

1978-09-01

After completion of the Phase 1 Refueling Outage Availability Study, Babcock and Wilcox and the U.S. Department of Energy entered into a supplemental agreement to perform a balance-of-plant maintenance, inspection, and test study with the cooperation of Duke Power Company and Arkansas Power and Light Company. The objectives were (1) to expand the Phase 1 data base, including balance-of-plant activities, to reduce outage time and increase plant availability and (2) to conduct an onsite review of plant maintenance, practices to complement the utility efforts in reducing outage time and increasing on-line operational time. Data were obtained from (1) observations during the 1977 refueling outage at Oconee 3, (2) review of maintenance practices during the Arkansas Nuclear One, Unit 1, operational cycle in 1977, and (3) selected observations of the 1978 refueling outage at ANO-1. Accumulated data were then reviewed and analyzed to produce a list of improvement recommendations for Oconee 3 and ANO-1 that can be generically applied to plants of similar design and construction

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

Science.gov (United States)

Karak, Bidya Binay; Miesch, Mark

2017-09-01

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

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

Quality of life, patient satisfaction, and disease burden in patients with gastroesophageal reflux disease with or without laryngopharyngeal reflux symptoms.

Science.gov (United States)

Gong, Eun Jeong; Choi, Kee Don; Jung, Hye-Kyung; Youn, Young Hoon; Min, Byung-Hoon; Song, Kyung Ho; Huh, Kyu Chan

2017-07-01

Patients with gastroesophageal reflux disease (GERD) have decreased health-related quality of life (HRQL). The quality of life in patients with laryngopharyngeal reflux (LPR) symptoms is also significantly impaired. However, the impact of LPR symptoms on HRQL in GERD patients has not been studied. A nationwide, random-sample, and face-to-face survey of 300 Korean patients with GERD was conducted from January to March 2013. Gastroesophageal reflux symptoms were assessed using the Rome III questionnaire, LPR symptoms using the reflux symptom index, and HRQL using the EuroQol five dimensions (EQ-5D) questionnaire. A structured questionnaire on patient satisfaction, sickness-related absences, and health-related work productivity was also used. Among the 300 patients with GERD, 150 had LPR symptoms. The mean EQ-5D index was lower in patients with GERD and LPR symptoms than in those without LPR (0.88 vs 0.91, P = 0.002). A linear regression model showed that the severity of LPR symptoms was related to decreased HRQL and was independent of age, marital status, body mass index, or household income. The overall satisfaction rate regarding treatment was lower in patients with GERD and LPR (40.0% vs 69.1%, P = 0.040). GERD patients with LPR symptoms reported greater sickness-related absent hours per week (0.36 vs 0.02 h, P = 0.016) and greater percentages of overall work impairment than those without LPR (31.1% vs 20.8%, P Gastroesophageal reflux disease patients with LPR symptoms have a poorer HRQL, a lower satisfaction rate, and a greater disease burden than those without LPR. © 2017 Journal of Gastroenterology and Hepatology Foundation and John Wiley & Sons Australia, Ltd.

Advanced Emissions Control Development Program: Mercury Control

International Nuclear Information System (INIS)

Evans, A.P.; Redinger, K.W.; Holmes, M.J.

1997-07-01

McDermott Technology, Inc. (a subsidiary of Babcock ampersand Wilcox) is conducting the Advanced Emissions Control Development Project (AECDP) which is aimed at the development of practical, cost-effective strategies for reducing the emissions of hazardous air pollutants (HAPS) from coal-fired electric utility plants. The need for such controls may arise as the US Environmental Protection Agency (EPA) proceeds with implementation of requirements set forth in the Clean Air Act Amendments (CAAA's) of 1990. Promulgation of air toxics emissions regulations for electric utility plants could dramatically impact utilities burning coal, their industrial and residential customers, and the coal industry. AECDP project work will supply the information needed by utilities to respond to potential HAPs regulations in a timely, cost-effective, enviromnentally-sound manner which supports the continued use of the Nation's abundant reserves of coal, such as those in the State of Ohio. The development work is being carried out using the 10 MW Clean Environment Development Facility wherein air toxics emissions control strategies can be developed under controlled conditions. The specific objectives of the project are to (1) measure and understand production and partitioning of air toxics species for a variety of coals, (2) optimize the air toxics removal performance of conventional flue gas cleanup systems, (3) develop advanced air toxics emissions control concepts, (4) develop and validate air toxics emissions measurement and monitoring techniques, and (5) establish a comprehensive, self-consistent air toxics data library. This project is supported by the Department of Energy, the Ohio Coal Development Office within the Ohio Department of Development and Babcock ampersand Wilcox. A comprehensive assessment of HAP emissions from coal-fired electric utility boilers sponsored by the Department of Energy and the Electric Power Research Institute concluded that with the exception of

FUMAC-84. A hybrid PCI analytical tool

International Nuclear Information System (INIS)

Matheson, J.E.; Walton, L.A.

1984-01-01

''FUMAC-84'', a new computer code currently under development at Babcock and Wilcox, will be used to analyze PCMI in light water reactor fuel rods. This is a hybrid code in the sense that the pellet behaviour is predicted from deterministic models which incorporate the large data base being generated by the international fuel performance programs (OVERRAMP, SUPER-RAMP, NFIR, etc.), while the cladding is modelled using finite elements. The fuel cracking and relocation model developed for FUMAC is semi-empirical and includes data up to 35 GWd/mtU and linear heat rates ranging from 100 to 700 W/Cm. With this model the onset of cladding ridging has been accurately predicted for steady-state operation. Transient behaviour of the pellet is still under investigation and the model is being enhanced to include these effects. The cladding model integrates the mechanical damage over a power history by solving the finite element assumed displacement problem in a quasistatic manner. Early work on FUMAC-84 has been directed at the development and benchmarking of the interim code. The purpose of the interim code is to provide a vehicle to proof out the deterministic pellet models which have been developed. To date the cracking model and the relocation model have been benchmarked. The thermal model for the pellet was developed by fitting data from several Halden experiments. The ability to accurately predict cladding ridging behaviour has been used to test how well the pellet swelling, densification and compliance models work in conjunction with fuel cladding material models. Reasonable results have been achieved for the steady-state cases while difficulty has been encountered in trying to reproduce transient results. Current work includes an effort to improve the ability of the models to handle transients well. (author)

SOx-NOx-Rox Box{trademark} flue gas clean-up demonstration. Final report

Energy Technology Data Exchange (ETDEWEB)

NONE

1995-09-01

Babcock and Wilcox`s (B and W) SOx-NOx-Rox Box{trademark} process effectively removes SOx, NOx and particulate (Rox) from flue gas generated from coal-fired boilers in a single unit operation, a high temperature baghouse. The SNRB technology utilizes dry sorbent injection upstream of the baghouse for removal of SOx and ammonia injection upstream of a zeolitic selective catalytic reduction (SCR) catalyst incorporated in the baghouse to reduce NOx emissions. Because the SOx and NOx removal processes require operation at elevated gas temperatures (800--900 F) for high removal efficiency, high-temperature fabric filter bags are used in the baghouse. The SNRB technology evolved from the bench and laboratory pilot scale to be successfully demonstrated at the 5-MWe field scale. This report represents the completion of Milestone M14 as specified in the Work Plan. B and W tested the SNRB pollution control system at a 5-MWe demonstration facility at Ohio Edison`s R.E. Burger Plant located near Shadyside, Ohio. The design and operation were influenced by the results from laboratory pilot testing at B and W`s Alliance Research Center. The intent was to demonstrate the commercial feasibility of the SNRB process. The SNRB facility treated a 30,000 ACFM flue gas slipstream from Boiler No. 8. Operation of the facility began in May 1992 and was completed in May 1993. About 2,300 hours of high-temperature operation were achieved. The main emissions control performance goals of: greater than 70% SO{sub 2} removal using a calcium-based sorbent; greater than 90% NOx removal with minimal ammonia slip; and particulate emissions in compliance with the New Source Performance Standards (NSPS) of 0.03 lb/million Btu were exceeded simultaneously in the demonstration program when the facility was operated at optimal conditions. Testing also showed significant reductions in emissions of some hazardous air pollutants.

Pepsin and bile acids in saliva in patients with laryngopharyngeal reflux - a prospective comparative study.

Science.gov (United States)

Sereg-Bahar, M; Jerin, A; Jansa, R; Stabuc, B; Hocevar-Boltezar, I

2015-06-01

Laryngopharyngeal reflux (LPR) and biliary duodenogastric reflux can cause damage to the laryngeal mucosa and voice disorders. The aim of this study was to find out whether levels of pepsin and bile acids in the saliva can serve as diagnostic markers of LPR. A prospective comparative study. Twenty-eight patients with LPR proven via high-resolution manometry and combined multichannel intraluminal impedance and 24-h pH monitoring and 48 healthy controls without symptoms of LPR were included in the study. In the patients with LPR symptoms, oesophagogastroscopy with oesophageal biopsy was performed. The levels of total pepsin, active pepsin, bile acids and the pH of the saliva were determined in all participants and compared between the groups. Reflux symptom index (RSI) and reflux finding score (RFS) were also obtained and compared. The groups differed significantly in RSI (P = 0.00), RFS (P = 0.00), the levels of bile acids (P = 0.005) and total pepsin in saliva (P = 0.023). The levels of total pepsin and bile acids were about three times higher in the patients with LPR than in the healthy controls. There was a significant correlation between the RSI and RFS score and the level of total pepsin and bile acids in the saliva. Histopathological examination of the oesophageal biopsy taken 5 cm above the lower oesophageal sphincter confirmed reflux in almost 93% of patients with symptoms. The study results show that the levels of total pepsin and bile acids in saliva are significantly higher in patients with LPR than in the controls, thus suggesting this as a useful tool in the diagnosis of LPR and particularly biliary LPR. © 2014 John Wiley & Sons Ltd.

Engineering and innovative erection concept for the containment liner for an EPR trademark

International Nuclear Information System (INIS)

Goehring, Rainer

2010-01-01

The determination of the optimal design of the containment liner considering amount of material, manufacturing and erection was the challenge for the engineering team of Babcock Noell GmbH. The construction costs of a Nuclear Power Plant are also impacted by the erection time of components. Therefore, it is necessary to optimize the erection times and consequently essentially to shorten. One possibility has been demonstrated by Babcock Noell GmbH with the erection concept of a containment liner. The liner is preassembled on the pre-assembly place in rings of up to 12 meter height, with the nominal diameter of 47 meter and a weight of approximately up to 200 tonnes. These rings, as well as the containment cup and the dome, are lifted with a heavy load crane into the reactor building. Several load cases for normal operation and accidental conditions as well as severe accidents have been analysed. The special load cases in the vicinity of penetrations and anchor plates have been calculated. The results of theses analyses have been considered in the actual design of the liner. This is the topic of the speech '' Computational Concept for the Containment Liner for EPR trademark '', given by Mr. Franz Nagelstutz, Abt. Planning, Babcock Noell GmbH und Mr. Nils Anders, Babcock Noell GmbH, Abt. Planning. With this concept, the construction activities on the inner containment wall are only five times disrupted by the welding and coating of the circumferential weld (approx. 25 calendar days). In comparison with the common known erection concept of welding of liner shells in situ, at least 20 weeks are saved on the schedule. An integrated concept from planning, manufacturing and erection of this large component has been implemented. It could be demonstrated that within the given time frame, with the required quality and within the required tolerances the containment liner for the Nuclear Power Plant can be delivered to the Client. (orig.)

Science & Technology Review October/November 2016

Energy Technology Data Exchange (ETDEWEB)

Vogt, R. L. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Meissner, C. N. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Kotta, P. R. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

2016-11-11

At Lawrence Livermore National Laboratory, we focus on science and technology research to ensure our nation’s security. We also apply that expertise to solve other important national problems in energy, bioscience, and the environment. Science & Technology Review is published eight times a year to communicate, to a broad audience, the Laboratory’s scientific and technological accomplishments in fulfilling its primary missions. The publication’s goal is to help readers understand these accomplishments and appreciate their value to the individual citizen, the nation, and the world. The Laboratory is operated by Lawrence Livermore National Security, LLC (LLNS), for the Department of Energy’s National Nuclear Security Administration. LLNS is a partnership involving Bechtel National, University of California, Babcock & Wilcox, Washington Division of URS Corporation, and Battelle in affiliation with Texas A&M University. More information about LLNS is available online at www.llnsllc.com. Please address any correspondence (including name and address changes) to S&TR, Mail Stop L-664, Lawrence Livermore National Laboratory, P.O. Box 808, Livermore, California 94551, or telephone (925) 423-3893. Our e-mail address is str-mail@llnl.gov. S&TR is available on the Web at str.llnl.gov.

Coal reburning technology for cyclone boilers

International Nuclear Information System (INIS)

Yagiela, A.S.; Maringo, G.J.; Newell, R.J.; Farzan, H.

1990-01-01

Babcock and Wilcox has obtained encouraging results from engineering feasibility and pilot-scale proof-of-concept studies of coal reburning for cyclone boiler NO x control. Accordingly, B and W completed negotiations for a clean coal cooperative agreement with the Department of Energy to demonstrate coal reburning technology for cyclone boilers. The host site for the demonstration is the Wisconsin Power and Light (WP and L) Company's 100MWe Nelson Dewey Station. Reburning involves the injection of a supplemental fuel (natural gas, oil, or coal) into the main furnace to produce locally reducing stoichiometric conditions which convert the NO x produced therein to molecular nitrogen, thereby reducing overall NO x emissions. There are currently no commercially-demonstrated combustion modification techniques for cyclone boilers which reduce NO x emissions. The emerging reburning technology offers cyclone boiler operators a promising alternative to expensive flue gas cleanup techniques for NO x emission reduction. This paper reviews baseline testing results at the Nelson Dewey Station and pilot-scale results simulating Nelson Dewey operation using pulverized coal (PC) as the reburning fuel. Outcomes of the model studies as well as the full-scale demonstration preliminary design are discussed

Lipocalins Are Required for Apical Extracellular Matrix Organization and Remodeling in Caenorhabditis elegans.

Science.gov (United States)

Forman-Rubinsky, Rachel; Cohen, Jennifer D; Sundaram, Meera V

2017-10-01

A lipid and glycoprotein-rich apical extracellular matrix (aECM) or glycocalyx lines exposed membranes in the body, and is particularly important to protect narrow tube integrity. Lipocalins ("fat cups") are small, secreted, cup-shaped proteins that bind and transport lipophilic cargo and are often found in luminal or aECM compartments such as mammalian plasma, urine, or tear film. Although some lipocalins can bind known aECM lipids and/or matrix metalloproteinases, it is not known if and how lipocalins affect aECM structure due to challenges in visualizing the aECM in most systems. Here we show that two Caenorhabditis elegans lipocalins, LPR-1 and LPR-3, have distinct functions in the precuticular glycocalyx of developing external epithelia. LPR-1 moves freely through luminal compartments, while LPR-3 stably localizes to a central layer of the membrane-anchored glycocalyx, adjacent to the transient zona pellucida domain protein LET-653 Like LET-653 and other C. elegans glycocalyx components, these lipocalins are required to maintain the patency of the narrow excretory duct tube, and also affect multiple aspects of later cuticle organization. lpr-1 mutants cannot maintain a continuous excretory duct apical domain and have misshapen cuticle ridges (alae) and abnormal patterns of cuticular surface lipid staining. lpr-3 mutants cannot maintain a passable excretory duct lumen, properly degrade the eggshell, or shed old cuticle during molting, and they lack cuticle barrier function. Based on these phenotypes, we infer that both LPR-1 and LPR-3 are required to build a properly organized aECM, while LPR-3 additionally is needed for aECM clearance and remodeling. The C. elegans glycocalyx provides a powerful system, amenable to both genetic analysis and live imaging, for investigating how lipocalins and lipids affect aECM structure. Copyright © 2017 by the Genetics Society of America.

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)

Comparison between gradient-dependent hydraulic conductivities of roots using the root pressure probe: the role of pressure propagations and implications for the relative roles of parallel radial pathways.

Science.gov (United States)

Bramley, Helen; Turner, Neil C; Turner, David W; Tyerman, Stephen D

2007-07-01

Hydrostatic pressure relaxations with the root pressure probe are commonly used for measuring the hydraulic conductivity (Lp(r)) of roots. We compared the Lp(r) of roots from species with different root hydraulic properties (Lupinus angustifolius L. 'Merrit', Lupinus luteus L. 'Wodjil', Triticum aestivum L. 'Kulin' and Zea mays L. 'Pacific DK 477') using pressure relaxations, a pressure clamp and osmotic gradients to induce water flow across the root. Only the pressure clamp measures water flow under steady-state conditions. Lp(r) determined by pressure relaxations was two- to threefold greater than Lp(r) from pressure clamps and was independent of the direction of water flow. Lp(r) (pressure clamp) was two- to fourfold higher than for Lp(r) (osmotic) for all species except Triticum aestivum where Lp(r) (pressure clamp) and Lp(r) (osmotic) were not significantly different. A novel technique was developed to measure the propagation of pressure through roots to investigate the cause of the differences in Lp(r). Root segments were connected between two pressure probes so that when root pressure (P(r)) was manipulated by one probe, the other probe recorded changes in P(r). Pressure relaxations did not induce the expected kinetics in pressure in the probe at the other end of the root when axial hydraulic conductance, and probe and root capacitances were accounted for. An electric circuit model of the root was constructed that included an additional capacitance in the root loaded by a series of resistances. This accounted for the double exponential kinetics for intact roots in pressure relaxation experiments as well as the reduced response observed with the double probe experiments. Although there were potential errors with all the techniques, we considered that the measurement of Lp(r) using the pressure clamp was the most unambiguous for small pressure changes, and provided that sufficient time was allowed for pressure propagation through the root. The differences in

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

Directory of Open Access Journals (Sweden)

Xiaofeng Liao

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

Reactor theory and power reactors. 1. Calculational methods for reactors. 2. Reactor kinetics

International Nuclear Information System (INIS)

Henry, A.F.

1980-01-01

Various methods for calculation of neutron flux in power reactors are discussed. Some mathematical models used to describe transients in nuclear reactors and techniques for the reactor kinetics' relevant equations solution are also presented

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

Reactor core and initially loaded reactor core of nuclear reactor

International Nuclear Information System (INIS)

Koyama, Jun-ichi; Aoyama, Motoo.

1989-01-01

In BWR type reactors, improvement for the reactor shutdown margin is an important characteristic condition togehter with power distribution flattening . However, in the reactor core at high burnup degree, the reactor shutdown margin is different depending on the radial position of the reactor core. That is , the reactor shutdown margin is smaller in the outer peripheral region than in the central region of the reactor core. In view of the above, the reactor core is divided radially into a central region and as outer region. The amount of fissionable material of first fuel assemblies newly loaded in the outer region is made less than the amount of the fissionable material of second fuel assemblies newly loaded in the central region, to thereby improve the reactor shutdown margin in the outer region. Further, the ratio between the amount of the fissionable material in the upper region and that of the fissionable material in the lower portion of the first fuel assemblies is made smaller than the ratio between the amount of the fissionable material in the upper region and that of the fissionable material in the lower region of the second fuel assemblies, to thereby obtain a sufficient thermal margin in the central region. (K.M.)

Reactor science and technology: operation and control of reactors

International Nuclear Information System (INIS)

Qiu Junlong

1994-01-01

This article is a collection of short reports on reactor operation and research in China in 1991. The operation of and research activities linked with the Heavy Water Research Reactor, Swimming Pool Reactor and Miniature Neutron Source Reactor are briefly surveyed. A number of papers then follow on the developing strategies in Chinese fast breeder reactor technology including the conceptual design of an experimental fast reactor (FFR), theoretical studies of FFR thermo-hydraulics and a design for an immersed sodium flowmeter. Reactor physics studies cover a range of topics including several related to work on zero power reactors. The section on reactor safety analysis is concerned largely with the assessment of established, and the presentation of new, computer codes for use in PWR safety calculations. Experimental and theoretical studies of fuels and reactor materials for FBRs, PWRs, BWRs and fusion reactors are described. A final miscellaneous section covers Mo-Tc isotope production in the swimming pool reactor, convective heat transfer in tubes and diffusion of tritium through plastic/aluminium composite films and Li 2 SiO 3 . (UK)

Correlation of obstructive sleep apnoea and laryngopharyngeal reflux: phmetry study.

Science.gov (United States)

Elhennawi, D M; Ahmed, M R; Abou-Halawa, A S

2016-12-01

To study the correlation of obstructive sleep apnoea (OSA) and laryngopharyngeal reflux (LPR). A descriptive study. Suez Canal University Hospital, Ismailia, Egypt. 62 patients with polysomnography confirmed OSA. Patients were evaluated with ambulatory 24-h double channel pH monitoring. Mean reflux symptom index in the study group was 9 ± 5.5, and it was > 13 in all patients with severe OSA. Signs of LPR reflux were present in 34 (55%) patients. Abnormal reflux was detected in the distal oesophagus in 41 patients (66%) and in the proximal oesophagus in 21 patients (34%). Patients with severe OSA had significantly higher nocturnal LPR reflux episodes compared to patients with mild disease (P .05). LPR is common in patients with OSA. Patients with severe OSA have significantly higher nocturnal LPR. This should be considered when treating this group of patients. © 2016 John Wiley & Sons Ltd.

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

DEFF Research Database (Denmark)

Mason, K; Thygesen, Lau Caspar; Stenager, Egon

2012-01-01

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

A container for storage and disposal of low-level waste

International Nuclear Information System (INIS)

Fish, R.L.; Butler, B.D.

1989-01-01

A unique concept for corrosion-resistant containers for storing and disposing of low-level radioactive, mixed and toxic wastes has been developed. The strength and low cost of carbon steel has been combined with the corrosion and abrasion resistance of a proprietary combination of polymers to provide an inexpensive alternative to currently available waste containers. The initial development effort has focused on a 55-gallon container, the B and W ECOSAFE-55 tm . However, Babcock and Wilcox (B and W) can develop a family of ECOSAFE waste containers using this technology to accommodate user-preferred configurations and volumes. The containers will be capable of accepting a wide range of low-level radioactive (LLRW) and industrial waste forms. Basic engineering design analyses and functional tests were performed to show compliance of the container with transportation functional requirements. These tests and analyses, along with chemical resistance tests, qualify the container for use in storing a wide range of radioactive and chemical wastes. For the container to be licensed for use as a high-integrity container in shallow land, low-level radioactive waste burial facilities, the Nuclear Regulatory Commission requires certain tests and analyses to demonstrate that container gross physical properties and identity can be maintained for 300 years. This paper describes the container concept in generic terms and provides information on the initial, ECOSAFE-55 container design, testing and engineering analysis efforts

Closing the loops between plant design and operator-An automatic logging system

International Nuclear Information System (INIS)

Tally, C.

1985-01-01

The close relationship between plant owner and NSSS designer frequently ceases after the plant is through startup testing. Thus, there is no continuous feedback between the operations staff and the designer. As a result, there is no assurance that the plant is being operated within the design envelope defined by the NSSS component stress reports. The link between plant operation and the plant design basis is vital to ensure that the plant can be safely operated for its full licensed life. This link is also a key to extending the life of the plant since the fatigue history of critical components is an important element of any justification for extended component life. An allowable Operating Transient Cycles Program established by Duke Power and Babcock and Wilcox successfully closed the operator-designer loop at the Oconee Nuclear Station. This paper describes that program, some of its conclusions, and also describes the next logical step in its development...automation of the transient logging process. A transient monitoring program must satisfy many requirements ranging from sensing the onset of a transient or slow power maneuver to recording sufficient data to provide for human checking of all computerized conclusions and results. Although not yet available to the industry, this type of program will ultimately be a virtual necessity for all nuclear stations

Polarization ratio property and material classification method in passive millimeter wave polarimetric imaging

Science.gov (United States)

Cheng, Yayun; Qi, Bo; Liu, Siyuan; Hu, Fei; Gui, Liangqi; Peng, Xiaohui

2016-10-01

Polarimetric measurements can provide additional information as compared to unpolarized ones. In this paper, linear polarization ratio (LPR) is created to be a feature discriminator. The LPR properties of several materials are investigated using Fresnel theory. The theoretical results show that LPR is sensitive to the material type (metal or dielectric). Then a linear polarization ratio-based (LPR-based) method is presented to distinguish between metal and dielectric materials. In order to apply this method to practical applications, the optimal range of incident angle have been discussed. The typical outdoor experiments including various objects such as aluminum plate, grass, concrete, soil and wood, have been conducted to validate the presented classification method.

The CAREM reactor and present currents in reactor design

International Nuclear Information System (INIS)

Ordonez, J.P.

1990-01-01

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

Characterization of a Mycobacterium avium subsp. avium Operon Associated with Virulence and Drug Detoxification

Directory of Open Access Journals (Sweden)

Mariana Noelia Viale

2014-01-01

Full Text Available The lprG-p55 operon of Mycobacterium tuberculosis and Mycobacterium bovis is involved in the transport of toxic compounds. P55 is an efflux pump that provides resistance to several drugs, while LprG is a lipoprotein that modulates the host's immune response against mycobacteria. The knockout mutation of this operon severely reduces the replication of both mycobacterial species during infection in mice and increases susceptibility to toxic compounds. In order to gain insight into the function of LprG in the Mycobacterium avium complex, in this study, we assayed the effect of the deletion of lprG gene in the D4ER strain of Mycobacterium avium subsp. avium. The replacement of lprG gene with a hygromycin cassette caused a polar effect on the expression of p55. Also, a twofold decrease in ethidium bromide susceptibility was observed and the resistance to the antibiotics rifampicin, amikacin, linezolid, and rifabutin was impaired in the mutant strain. In addition, the mutation decreased the virulence of the bacteria in macrophages in vitro and in a mice model in vivo. These findings clearly indicate that functional LprG and P55 are necessary for the correct transport of toxic compounds and for the survival of MAA in vitro and in vivo.

NLRP3 and ASC suppress lupus-like autoimmunity by driving the immunosuppressive effects of TGF-β receptor signalling.

Science.gov (United States)

Lech, Maciej; Lorenz, Georg; Kulkarni, Onkar P; Grosser, Marian O O; Stigrot, Nora; Darisipudi, Murthy N; Günthner, Roman; Wintergerst, Maximilian W M; Anz, David; Susanti, Heni Eka; Anders, Hans-Joachim

2015-12-01

The NLRP3/ASC inflammasome drives host defence and autoinflammatory disorders by activating caspase-1 to trigger the secretion of mature interleukin (IL)-1β/IL-18, but its potential role in autoimmunity is speculative. We generated and phenotyped Nlrp3-deficient, Asc-deficient, Il-1r-deficient and Il-18-deficient C57BL/6-lpr/lpr mice, the latter being a mild model of spontaneous lupus-like autoimmunity. While lack of IL-1R or IL-18 did not affect the C57BL/6-lpr/lpr phenotype, lack of NLRP3 or ASC triggered massive lymphoproliferation, lung T cell infiltrates and severe proliferative lupus nephritis within 6 months, which were all absent in age-matched C57BL/6-lpr/lpr controls. Lack of NLRP3 or ASC increased dendritic cell and macrophage activation, the expression of numerous proinflammatory mediators, lymphocyte necrosis and the expansion of most T cell and B cell subsets. In contrast, plasma cells and autoantibody production were hardly affected. This unexpected immunosuppressive effect of NLRP3 and ASC may relate to their known role in SMAD2/3 phosphorylation during tumour growth factor (TGF)-β receptor signalling, for example, Nlrp3-deficiency and Asc-deficiency significantly suppressed the expression of numerous TGF-β target genes in C57BL/6-lpr/lpr mice and partially recapitulated the known autoimmune phenotype of Tgf-β1-deficient mice. These data identify a novel non-canonical immunoregulatory function of NLRP3 and ASC in autoimmunity. Published by the BMJ Publishing Group Limited. For permission to use (where not already granted under a licence) please go to http://www.bmj.com/company/products-services/rights-and-licensing/

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

Dual regulation of root hydraulic conductivity and plasma membrane aquaporins by plant nitrate accumulation and high-affinity nitrate transporter NRT2.1.

Science.gov (United States)

Li, Guowei; Tillard, Pascal; Gojon, Alain; Maurel, Christophe

2016-04-01

The water status and mineral nutrition of plants critically determine their growth and development. Nitrate (NO3(-)), the primary nitrogen source of higher plants, is known to impact the water transport capacity of roots (root hydraulic conductivity, Lpr). To explore the effects and mode of action of NO3(-) on Lpr, we used an extended set of NO3(-) transport (nrt1.1, nrt1.2, nrt1.5 and nrt2.1), signaling (nrt1.1 and nrt2.1) and metabolism (nia) mutants in Arabidopsis, grown under various NO3(-) conditions. First, a strong positive relationship between Lpr and NO3(-) accumulation, in shoots rather than in roots, was revealed. Secondly, a specific 30% reduction of Lpr in nrt2.1 plants unraveled a major role for the high-affinity NO3(-) transporter NRT2.1 in increasing Lpr These results indicate that NO3(-)signaling rather than nitrogen assimilation products governs Lpr in Arabidopsis. Quantitative real-time reverse transcription-PCR and enzyme-linked immunosorbent assays (ELISAs) were used to investigate the effects of NO3(-) availability on plasma membrane aquaporin (plasma membrane intrinsic protein; PIP) expression. Whereas PIP regulation mostly occurs at the post-translational level in wild-type plants, a regulation of PIPs at both the transcriptional and translational levels was uncovered in nrt2.1 plants. In conclusion, this work reveals that control of Arabidopsis Lpr and PIP functions by NO3(-) involves novel shoot to root signaling and NRT2.1-dependent functions. © The Author 2016. Published by Oxford University Press on behalf of Japanese Society of Plant Physiologists. All rights reserved. For permissions, please email: journals.permissions@oup.com.

Reactor containment and reactor safety in the United States

International Nuclear Information System (INIS)

Kouts, H.

1986-01-01

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

Comparison between TRU burning reactors and commercial fast reactor

International Nuclear Information System (INIS)

Fujimura, Koji; Sanda, Toshio; Ogawa, Takashi

2001-03-01

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

Reactor as furnace and reactor as lamp

International Nuclear Information System (INIS)

Goldanskii, V.I.

1992-01-01

There are presented general characteristics of the following ways of transforming of nuclear energy released in reactors into chemical : ordinary way (i.e. trough the heat, mechanical energy and electricity); chemonuclear synthesis ; use of high-temperature fuel elements (reactor as furnace); use of the mixed nγ-radiation of reactors; use of the radiation loops; radiation - photochemical synthesis (reactor as lamp). Advantage and disadvantages of all above variants are compared. The yield of the primary product of fixation of nitrogen (nitric oxide NO) in reactor with the high-temperature (above ca. 1900degC) fuel elements (reactor-furnace) can exceed W ∼ 200 kg per gram of burned uranium. For the latter variant (reactor-lamp) the yield of chemical products can reach W ∼ 60 kg. per gram of uranium. Such values of W are close to or even strongly exceed the yields of chemical products for other abovementioned variants and - what is particularly important - are not connected to the necessity of archscrupulous removal of radioactive contamination of products. (author)

Neutron behavior, reactor control, and reactor heat transfer. Volume four

International Nuclear Information System (INIS)

Anon.

1986-01-01

Volume four covers neutron behavior (neutron absorption, how big are nuclei, neutron slowing down, neutron losses, the self-sustaining reactor), reactor control (what is controlled in a reactor, controlling neutron population, is it easy to control a reactor, range of reactor control, what happens when the fuel burns up, controlling a PWR, controlling a BWR, inherent safety of reactors), and reactor heat transfer (heat generation in a nuclear reactor, how is heat removed from a reactor core, heat transfer rate, heat transfer properties of the reactor coolant)

The Processing of Lunar Penetrating Radar Channel-2B Data from Chang'E-3

Science.gov (United States)

Zhu, P.; Zhao, N.; Yang, K.; Yuan, Y.; Guo, S.

2014-12-01

The Lunar Penetrating Radar (LPR) carried by Chang'E-3 has imaged the shallow subsurface of the landing site at the northern Mare Imbrium. The antenna B of the Channel-2 onboard the LPR (LPR Channel-2B) has collected more than 20000 traces of raw data. The raw LPR data could not be directly used for geological interpretation because of the operation mode of the LPR, noise and fast attenuation of radar wave. This study focuses data preprocessing and processing methods to obtain higher quality data. A section of usable LPR data of over 2000 traces is gained after the preprocessing of selecting, splicing, removing delay time and fine-correcting to raw data, but only a few shallow geological structures are visible. To further improve the resolution and the signal-to-noise ratio of the LPR data, we have processed the LPR data including amplitude compensation, filtering, and deconvolution processes based on electromagnetic wave theory. The processing results reveal that (1) the spherical spreading compensation and auto gain control enhance the amplitude of reflection echoes from deeper strata and make the geological structures more obvious, (2) the spiking deconvolution applied to narrow reflection events down makes it possible to identify thin layers with 30% improved resolution, and (3) the band-pass filtering removes the multiple reflections and, consequently, improves the signal-to-noise ratio of LPR data. The processing results will lay the foundation for the subsequent geological interpretation and physical property inversion of lunar materials.

Research reactor standards and their impact on the TRIGA reactor community

International Nuclear Information System (INIS)

Richards, W.J.

1980-01-01

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

Effect of component aging on PWR control rod drive systems

International Nuclear Information System (INIS)

Grove, E.; Gunther, W.; Sullivan, K.

1992-01-01

An aging assessment of PWR control rod drive (CRD) systems has been completed as part of the US NRC Nuclear Plant Aging Research (NPAR) Program. The design, construction, maintenance, and operation of the Babcock ampersand Wilcox (B ampersand W), Combustion Engineering (CE), and Westinghouse (W) systems were evaluated to determine the potential for degradation as each system ages. Operating experience data were evaluated to identify the predominant failure modes, causes, and effects. This, coupled with an assessment of the materials of construction and operating environment, demonstrate that each design is subject to degradation, which if left unchecked, could affect its safety function as the plant ages. An industry survey, conducted with the assistance of EPRI and NUMARC, identified current CRD system maintenance and inspection practices. The results of this survey indicate that some plants have performed system modifications, replaced components, or augmented existing preventive maintenance practices in response to system aging. The survey results also supported the operating experience data, which concluded that the timely replacement of degraded components, prior to failure, was not always possible using existing condition monitoring techniques. The recommendations presented in this study also include a discussion of more advanced monitoring techniques, which provide trendable results capable of detecting aging

Demonstration of coal reburning for cyclone boiler NO{sub x} control. Appendix, Book 1

Energy Technology Data Exchange (ETDEWEB)

1994-06-01

Based on the industry need for a pilot-scale cyclone boiler simulator, Babcock Wilcox (B&W) designed, fabricated, and installed such a facility at its Alliance Research Center (ARC) in 1985. The project involved conversion of an existing pulverized coal-fired facility to be cyclone-firing capable. Additionally, convective section tube banks were installed in the upper furnace in order to simulate a typical boiler convection pass. The small boiler simulator (SBS) is designed to simulate most fireside aspects of full-size utility boilers such as combustion and flue gas emissions characteristics, fireside deposition, etc. Prior to the design of the pilot-scale cyclone boiler simulator, the various cyclone boiler types were reviewed in order to identify the inherent cyclone boiler design characteristics which are applicable to the majority of these boilers. The cyclone boiler characteristics that were reviewed include NO{sub x} emissions, furnace exit gas temperature (FEGT) carbon loss, and total furnace residence time. Previous pilot-scale cyclone-fired furnace experience identified the following concerns: (1) Operability of a small cyclone furnace (e.g., continuous slag tapping capability). (2) The optimum cyclone(s) configuration for the pilot-scale unit. (3) Compatibility of NO{sub x} levels, carbon burnout, cyclone ash carryover to the convection pass, cyclone temperature, furnace residence time, and FEGT.

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

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

International Nuclear Information System (INIS)

Greneche, D.

2014-01-01

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

Computerized reactor monitor and control for nuclear reactors

International Nuclear Information System (INIS)

Buerger, L.

1982-01-01

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

Rikkunshito improves globus sensation in patients with proton-pump inhibitor-refractory laryngopharyngeal reflux.

Science.gov (United States)

Tokashiki, Ryoji; Okamoto, Isaku; Funato, Nobutoshi; Suzuki, Mamoru

2013-08-21

To investigate the effect of rikkunshito on laryngopharyngeal reflux (LPR) symptoms and gastric emptying in patients with proton-pump inhibitor (PPI)-refractory LPR. In total, 22 patients with LPR were enrolled. Following a 2-wk treatment with PPI monotherapy, PPI-refractory LPR patients were randomly divided into two treatment groups (rikkunshito alone or rikkunshito plus the PPI, lansoprazole). LPR symptoms were assessed using a visual analog scale (VAS) score, gastrointestinal symptoms were assessed using the gastrointestinal symptom rating scale (GSRS), and gastric emptying was assessed using the radio-opaque marker method prior to and 4 wk following treatments. The 4-wk treatment with rikkunshito alone and with rikkunshito plus the PPI significantly decreased the globus sensation VAS scores. The VAS score for sore throat was significantly decreased following treatment with rikkunshito plus PPI but not by rikkunshito alone. Neither treatment significantly changed the GSRS scores. Rikkunshito improved delayed gastric emptying. We found a significant positive correlation between improvements in globus sensation and in gastric emptying (r² = 0.4582, P sensation in patients with PPI-refractory LPR, in part, because of stimulation of gastric emptying. Thus, rikkunshito is an effective treatment for PPI-refractory LPR.

Reactor technology: power conversion systems and reactor operation and maintenance

International Nuclear Information System (INIS)

Powell, J.R.

1977-01-01

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

Anatomy of reflux: a growing health problem affecting structures of the head and neck.

Science.gov (United States)

Lipan, Michael J; Reidenberg, Joy S; Laitman, Jeffrey T

2006-11-01

Gastroesophageal reflux disease (GERD) and laryngopharyngeal reflux (LPR) are sibling diseases that are a modern-day plague. Millions of Americans suffer from their sequelae, ranging from subtle annoyances to life-threatening illnesses such as asthma, sleep apnea, and cancer. Indeed, the recognized prevalence of GERD alone has increased threefold throughout the 1990s. Knowledge of the precise etiologies for GERD and LPR is becoming essential for proper treatment. This review focuses on the anatomical, physiological, neurobiological, and cellular aspects of these diseases. By definition, gastroesophageal reflux (GER) is the passage of gastric contents into the esophagus; when excessive and damaging to the esophageal mucosa, GERD results. Reflux that advances to the laryngopharynx and, subsequently, to other regions of the head and neck such as the larynx, oral cavity, nasopharynx, nasal cavity, paranasal sinuses, and even middle ear results in LPR. While GERD has long been identified as a source of esophageal disease, LPR has only recently been implicated in causing head and neck problems. Recent research has identified four anatomical/physiological "barriers" that serve as guardians to prevent the cranial incursion of reflux: the gastroesophageal junction, esophageal motor function and acid clearance, the upper esophageal sphincter, and pharyngeal and laryngeal mucosal resistance. Sequential failure of all four barriers is necessary to produce LPR. While it has become apparent that GER must precede both GERD and LPR, the head and neck distribution of the latter clearly separates these diseases as distinct entities warranting specialized focus and treatment.

Nuclear reactor, reactor core thereof, and device for constituting the reactor

International Nuclear Information System (INIS)

Takiyama, Masashi.

1994-01-01

A reactor core is constituted by charging coolants (light water) in a reactor pressure vessel and distributing fuel assemblies, reflecting material sealing pipes, moderator (heavy water and helium gas) sealing pipes, and gas sealing pipes therein. A fuel guide tube is surrounded by a cap and the gap therebetween is made hollow and filled with coolant steams. The cap is supported by a baffle plate. The moderator sealing pipe is disposed in a flow channel of coolants in adjacent with the cap. The position of the moderator sealing tube in the reactor core is controlled by water stream from a hydraulic pump with a guide tube extending below the baffle plate being as a guide. Then, the position of the moderator sealing tube is varied to conduct power control, burnup degree compensation, and reactor shut down. With such procedures, moderator cooling facility is no more necessary to simplify the structure. Further, heat generated from the moderator is transferred to the coolants thereby improving heat efficiency of the reactor. (I.N.)

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

Energy Technology Data Exchange (ETDEWEB)

2009-10-15

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

Effects of a Diet-Based Weight-Reducing Program with Probiotic Supplementation on Satiety Efficiency, Eating Behaviour Traits, and Psychosocial Behaviours in Obese Individuals

Directory of Open Access Journals (Sweden)

Marina Sanchez

2017-03-01

Full Text Available This study evaluated the impact of probiotic supplementation (Lactobacillus rhamnosus CGMCC1.3724 (LPR on appetite sensations and eating behaviors in the context of a weight-reducing program. Obese men (n = 45 and women (n = 60 participated in a double-blind, randomized, placebo-controlled trial that included a 12-week weight loss period (Phase 1 based on moderate energy restriction, followed by 12 weeks of weight maintenance (Phase 2. During the two phases of the program, each subject consumed two capsules per day of either a placebo or a LPR formulation (10 mg of LPR equivalent to 1.6 108 CFU/capsule, 210 mg of oligofructose, and 90 mg of inulin. The LPR supplementation increased weight loss in women that was associated with a greater increase in the fasting desire to eat (p = 0.03. On the other hand, satiety efficiency (satiety quotient for desire to eat at lunch increased (p = 0.02, whereas disinhibition (p = 0.05 and hunger (p = 0.02 scores decreased more in the LPR-treated women, when compared with the female control group. Additionally, the LPR female group displayed a more pronounced decrease in food craving (p = 0.05, and a decrease in the Beck Depression Inventory score (p = 0.05 that was significantly different from the change noted in the placebo group (p = 0.02, as well as a higher score in the Body Esteem Scale questionnaire (p = 0.06. In men, significant benefits of LPR on fasting fullness and cognitive restraint were also observed. Taken together, these observations lend support to the hypothesis that the gut-brain axis may impact appetite control and related behaviors in obesity management.

Three Mile Island Unit 1 Main Steam Line Break Three-Dimensional Neutronics/Thermal-Hydraulics Analysis: Application of Different Coupled Codes

International Nuclear Information System (INIS)

D'Auria, Francesco; Moreno, Jose Luis Gago; Galassi, Giorgio Maria; Grgic, Davor; Spadoni, Antonino

2003-01-01

A comprehensive analysis of the double ended main steam line break (MSLB) accident assumed to occur in the Babcock and Wilcox Three Mile Island Unit 1 (TMI-1) has been carried out at the Dipartimento di Ingegneria Meccanica, Nucleare e della Produzione of the University of Pisa, Italy, in cooperation with the University of Zagreb, Croatia. The overall activity has been completed within the framework of the participation in the Organization for Economic Cooperation and Development-Committee on the Safety of Nuclear Installations-Nuclear Science Committee pressurized water reactor MSLB benchmark.Thermal-hydraulic system codes (various versions of Relap5), three-dimensional (3-D) neutronics codes (Parcs, Quabbox, and Nestle), and one subchannel code (Cobra) have been adopted for the analysis. Results from the following codes (or code versions) are assumed as reference:1. Relap5/mod3.2.2, beta version, coupled with the 3-D neutron kinetics Parcs code parallel virtual machine (PVM) coupling2. Relap5/mod3.2.2, gamma version, coupled with the 3-D neutron kinetics Quabbox code (direct coupling)3. Relap5/3D code coupled with the 3-D neutron kinetics Nestle code.The influence of PVM and of direct coupling is also discussed.Boundary and initial conditions of the system, including those relevant to the fuel status, have been supplied by Pennsylvania State University in cooperation with GPU Nuclear Corporation (the utility, owner of TMI) and the U.S. Nuclear Regulatory Commission. The comparison among the results obtained by adopting the same thermal-hydraulic nodalization and the coupled code version is discussed in this paper.The capability of the control rods to recover the accident has been demonstrated in all the cases as well as the capability of all the codes to predict the time evolution of the assigned transient. However, one stuck control rod caused some 'recriticality' or 'return to power' whose magnitude is largely affected by boundary and initial conditions

Nuclear power reactors: reactor safety and military and civil defence

International Nuclear Information System (INIS)

Hvinden, T.

1976-01-01

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

Drosophila lipophorin receptors mediate the uptake of neutral lipids in oocytes and imaginal disc cells by an endocytosis-independent mechanism.

Directory of Open Access Journals (Sweden)

Esmeralda Parra-Peralbo

2011-02-01

Full Text Available Lipids are constantly shuttled through the body to redistribute energy and metabolites between sites of absorption, storage, and catabolism in a complex homeostatic equilibrium. In Drosophila, lipids are transported through the hemolymph in the form of lipoprotein particles, known as lipophorins. The mechanisms by which cells interact with circulating lipophorins and acquire their lipidic cargo are poorly understood. We have found that lipophorin receptor 1 and 2 (lpr1 and lpr2, two partially redundant genes belonging to the Low Density Lipoprotein Receptor (LDLR family, are essential for the efficient uptake and accumulation of neutral lipids by oocytes and cells of the imaginal discs. Females lacking the lpr2 gene lay eggs with low lipid content and have reduced fertility, revealing a central role for lpr2 in mediating Drosophila vitellogenesis. lpr1 and lpr2 are transcribed into multiple isoforms. Interestingly, only a subset of these isoforms containing a particular LDLR type A module mediate neutral lipid uptake. Expression of these isoforms induces the extracellular stabilization of lipophorins. Furthermore, our data indicate that endocytosis of the lipophorin receptors is not required to mediate the uptake of neutral lipids. These findings suggest a model where lipophorin receptors promote the extracellular lipolysis of lipophorins. This model is reminiscent of the lipolytic processing of triglyceride-rich lipoproteins that occurs at the mammalian capillary endothelium, suggesting an ancient role for LDLR-like proteins in this process.

Development of an MMS/PC based real time simulation of the B and W NSS plant for advanced control system design

International Nuclear Information System (INIS)

Bartells, P.S.; Brownell, R.B.

1990-01-01

The development of this personal-computer-based simulation of the Babcock and Wilcox nuclear steam system (NSS) was prompted in part by the need for a real-time analysis tool to be used in evaluating advanced control concepts for the NSS. NSS control is currently accomplished via conventional analog systems that are becoming increasingly obsolete. With the widespread use of digital micro-processor-based control systems for fossil power and other applications, the B and W Owners Group Advanced Control System Task Force is developing a next-generation control system for upgrading existing B and W power plants. To take advantage of the digital control technology, it is desirable to have a flexible, cost-effective, and portable control analysis tool available that can simulate various postulated control strategies and algorithms and couple these with simulated plant responses in real time to determine overall effectiveness. To develop the desired capability, B and W has incorporated the simulation methodology of the Modular Modeling System (MMS) and the knowledge gained during development of a similar Department of Energy-funded project. The MMS-based NSS model was developed and then modified to increase execution speed, ported to an IBM Personal System 2 (Model 80) and interfaced with user-friendly graphics. The user can develop alternative control strategies and readily interface them with the NSS model for real-time display and evaluation. The paper addresses the key considerations and programming techniques used to accomplish the resulting simulation

Reactor core in FBR type reactor

International Nuclear Information System (INIS)

Masumi, Ryoji; Kawashima, Katsuyuki; Kurihara, Kunitoshi.

1989-01-01

In a reactor core in FBR type reactors, a portion of homogenous fuels constituting the homogenous reactor core is replaced with multi-region fuels in which the enrichment degree of fissile materials is lower nearer to the axial center. This enables to condition the composition such that a reactor core having neutron flux distribution either of a homogenous reactor core or a heterogenous reactor core has substantially identical reactivity. Accordingly, in the transfer from the homogenous reactor core to the axially heterogenous reactor core, the average reactivity in the reactor core is substantially equal in each of the cycles. Further, by replacing a portion of the homogenous fuels with a multi-region fuels, thereby increasing the heat generation near the axial center, it is possiable to reduce the linear power output in the regions above and below thereof and, in addition, to improve the thermal margin in the reactor core. (T.M.)

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

International Nuclear Information System (INIS)

Schikorr, W.M.

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

The nuclear reactor strategy between fast breeder reactors and advanced pressurized water reactors

International Nuclear Information System (INIS)

Seifritz, W.

1983-01-01

A nuclear reactor strategy between fast breeder reactors (FBRs) and advanced pressurized water reactors (APWRs) is being studied. The principal idea of this strategy is that the discharged plutonium from light water reactors (LWRs) provides the inventories of the FBRs and the high-converter APWRs, whereby the LWRs are installed according to the derivative of a logistical S curve. Special emphasis is given to the dynamics of reaching an asymptotic symbiosis between FBRs and APWRs. The main conclusion is that if a symbiotic APWR-FBR family with an asymptotic total power level in the terawatt range is to exist in about half a century from now, we need a large number of FBRs already in an early phase

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

International Nuclear Information System (INIS)

Djalilzadeh, A.M.

1977-01-01

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

Computerized reactor monitor and control for research reactors

International Nuclear Information System (INIS)

Buerger, L.; Vegh, E.

1981-09-01

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

Chinese herbal medicine and prednisone increase proportion of ...

African Journals Online (AJOL)

Objective: This study investigated the effects of Chinese herbal medicine and prednisone onCD4+FoxP3+ T cells (Tregs) and Th17 cells in the MRL/lpr mouse model of systemic lupus erythematosus. Methods: MRL/lpr mice were treated with herbal medicine (yin-nourishing and heat-clearing therapy), prednisone, and a ...

Antiallergic activity of unripe Citrus hassaku fruits extract and its flavanone glycosides on chemical substance-induced dermatitis in mice.

Science.gov (United States)

Itoh, Kimihisa; Masuda, Megumi; Naruto, Shunsuke; Murata, Kazuya; Matsuda, Hideaki

2009-10-01

Oral administration of a 50% ethanolic extract (CH-ext) obtained from unripe Citrus hassaku fruits collected in July exhibited a potent dose-dependent inhibition of IgE (immunoglobulin E)-mediated triphasic cutaneous reaction at 1 h [immediate phase response (IPR)], 24 h [late phase response (LPR)] and 8 days [very late phase response (vLPR)] after dinitrofluorobenzene challenge in mice. Naringin, a major flavanone glycoside component of CH-ext, showed a potent dose-dependent inhibition against IPR, LPR and vLPR. Neohesperidin, another major glycoside component of CH-ext, showed an inhibition against vLPR. The effect of CH-ext on type IV allergic reaction was examined by determining inhibitory activity against ear swelling in mice by using the picryl chloride-induced contact dermatitis (PC-CD) model. Oral administration (p.o.) of CH-ext and subcutaneous administration (s.c.) of prednisolone inhibited ear swelling during the induction phase of PC-CD. The inhibitory activities of combinations of CH-ext (p.o.) and prednisolone (s.c.) against PC-CD in mice were more potent than those of CH-ext alone and prednisolone alone, without enhancing the adverse effects. Other combinations of prednisolone (s.c.) and flavanone glycoside (p.o.) components of CH-ext, i.e. naringin and neohesperidin, exerted similar synergistic effects.

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

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

Science.gov (United States)

Warwick, Peter D.; Crowley, Sharon S.

1995-01-01

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

Nuclear reactor physics course for reactor operators

International Nuclear Information System (INIS)

Baeten, P.

2006-01-01

The education and training of nuclear reactor operators is important to guarantee the safe operation of present and future nuclear reactors. Therefore, a course on basic 'Nuclear reactor physics' in the initial and continuous training of reactor operators has proven to be indispensable. In most countries, such training also results from the direct request from the safety authorities to assure the high level of competence of the staff in nuclear reactors. The aim of the basic course on 'Nuclear Reactor Physics for reactor operators' is to provide the reactor operators with a basic understanding of the main concepts relevant to nuclear reactors. Seen the education level of the participants, mathematical derivations are simplified and reduced to a minimum, but not completely eliminated

Nuclear reactors

International Nuclear Information System (INIS)

Barre, Bertrand

2015-10-01

After some remarks on the nuclear fuel, on the chain reaction control, on fuel loading and unloading, this article proposes descriptions of the design, principles and operations of different types of nuclear reactors as well as comments on their presence and use in different countries: pressurized water reactors (design of the primary and secondary circuits, volume and chemistry control, backup injection circuits), boiling water reactors, heavy water reactors, graphite and boiling water reactors, graphite-gas reactors, fast breeder reactors, and fourth generation reactors (definition, fast breeding). For these last ones, six concepts are presented: sodium-cooled fast reactor, lead-cooled fast reactor, gas-cooled fast reactor, high temperature gas-cooled reactor, supercritical water-cooled reactor, and molten salt reactor