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

Uranium silicide activities at Babcock and Wilcox

International Nuclear Information System (INIS)

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

1983-01-01

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

Standard technical specifications: Babcock and Wilcox Plants. Revision 1

International Nuclear Information System (INIS)

1995-04-01

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

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

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

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

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

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

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

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

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

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

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

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.

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

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

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)

Inspirations from Dupont Safety Management System

Institute of Scientific and Technical Information of China (English)

Ma Yong

2009-01-01

@@ Dupont,with its 200 years of safety management experience,tells us:all safety accidents can be prevented. Dupont has a history of more than 200 years,the concept of "safety is priority"has never changed.Dupont is just another word for safety.

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

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

DuPont Chemical Vapor Technical Report

International Nuclear Information System (INIS)

MOORE, T.L.

2003-01-01

DuPont Safety Resources was tasked with reviewing the current chemical vapor control practices and providing preventive recommendations on best commercial techniques to control worker exposures. The increased focus of the tank closure project to meet the 2024 Tri-Party Agreement (TPA) milestones has surfaced concerns among some CH2MHill employees and other interested parties. CH2MHill is committed to providing a safe working environment for employees and desires to safely manage the tank farm operations using appropriate control measures. To address worker concerns, CH2MHill has chartered a ''Chemical Vapors Project'' to integrate the activities of multiple CH2MHill project teams, and solicit the expertise of external resources, including an independent Industrial Hygiene expert panel, a communications consultant, and DuPont Safety Resources. Over a three-month time period, DuPont worked with CH2MHill ESH and Q, Industrial Hygiene, Engineering, and the independent expert panel to perform the assessment. The process included overview presentations, formal interviews, informal discussions, documentation review, and literature review. DuPont Safety Resources concluded that it is highly unlikely that workers in the tank farms are exposed to chemicals above established standards. Additionally, the conventional and radiological chemistry is understood, the inherent chemical hazards are known, and the risk associated with chemical vapor exposure is properly managed. The assessment highlighted management's commitment to addressing chemical vapor hazards and controlling the associated risks. Additionally, we found the Industrial Hygiene staff to be technically competent and well motivated. The tank characterization data resides in a comprehensive database containing the tank chemical compositions and relevant airborne concentrations

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)

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

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

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

The DuPont Conference: Implications for the Chemical Technology Curriculum

Science.gov (United States)

Kenkel, John; Rutledge, Sue; Kelter, Paul B.

1998-05-01

Southeast Community College (SCC) hosted the first DuPont Conference for Chemical Technology Education at its Lincoln, Nebraska campus October 4-6, 1997. The conference brought together fourteen practicing chemists and chemistry technicians and five college and university faculty members for the express purpose of suggesting new laboratory activities that would help relate the real world of work to the education of chemical laboratory technicians in community colleges. Participants included seven men and seven women from DuPont, Procter & Gamble, Eastman Chemical, Eastman Kodak, Dow Chemical, Air Products and Chemicals, Monsanto, Union Carbide, the Nebraska Agriculture Laboratory, and the University of Nebraska Biological Process Development Facility, Department of Food Science. The conference, sponsored by the E. I. DuPont DeNemours & Company through a grant awarded to SCC in June 1997, was intended to help further the goals of the two major projects underway at SCC, funded by the National Science Foundation's Advanced Technological Education Program. These projects, dubbed "Assignment: Chemical Technology I and II", or ACT-I and ACT-II, are curriculum and materials development projects. The invited scientists had between 2 and 32 years of experience that ranged from bench work to management levels. Many are or have been active on the national scene as members and officers of the American Chemical Society's Division of Chemical Technicians and the ACS Committee on Technician Activities.

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

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

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.

Comparative study on DuPont analysis and DEA models for measuring stock performance using financial ratio

Science.gov (United States)

Arsad, Roslah; Shaari, Siti Nabilah Mohd; Isa, Zaidi

2017-11-01

Determining stock performance using financial ratio is challenging for many investors and researchers. Financial ratio can indicate the strengths and weaknesses of a company's stock performance. There are five categories of financial ratios namely liquidity, efficiency, leverage, profitability and market ratios. It is important to interpret the ratio correctly for proper financial decision making. The purpose of this study is to compare the performance of listed companies in Bursa Malaysia using Data Envelopment Analysis (DEA) and DuPont analysis Models. The study is conducted in 2015 involving 116 consumer products companies listed in Bursa Malaysia. The estimation method of Data Envelopment Analysis computes the efficiency scores and ranks the companies accordingly. The Alirezaee and Afsharian's method of analysis based Charnes, Cooper and Rhodes (CCR) where Constant Return to Scale (CRS) is employed. The DuPont analysis is a traditional tool for measuring the operating performance of companies. In this study, DuPont analysis is used to evaluate three different aspects such as profitability, efficiency of assets utilization and financial leverage. Return on Equity (ROE) is also calculated in DuPont analysis. This study finds that both analysis models provide different rankings of the selected samples. Hypothesis testing based on Pearson's correlation, indicates that there is no correlation between rankings produced by DEA and DuPont analysis. The DEA ranking model proposed by Alirezaee and Asharian is unstable. The method cannot provide complete ranking because the values of Balance Index is equal and zero.

Evaluating Bank Profitability in Ghana: A five step Du-Pont Model Approach

Directory of Open Access Journals (Sweden)

Baah Aye Kusi

2015-09-01

Full Text Available We investigate bank profitability in Ghana using periods before, during and after the globe financial crises with the five step du-pont model for the first time.We adapt the variable of the five step du-pont model to explain bank profitability with a panel data of twenty-five banks in Ghana from 2006 to 2012. To ensure meaningful generalization robust errors fixed and random effects models are used.Our empirical results suggests that bank operating activities (operating profit margin, bank efficiency (asset turnover, bank leverage (asset to equity and financing cost (interest burden  were positive and significant determinants of bank profitability (ROE during the period of study implying that bank in Ghana can boost return to equity holders through the above mentioned variables. We further report that the five step du-pont model better explains the total variation (94% in bank profitability in Ghana as compared to earlier findings suggesting that bank specific variables are keen in explaining ROE in banks in Ghana.We cited no empirical study that has employed five step du-pont model making our study unique and different from earlier studies as we assert that bank specific variables are core to explaining bank profitability.                

Weyerhaeuser Export Facility at DuPont. Volume I.

Science.gov (United States)

1982-05-01

exists on the DuPont site, in which the most common trees are lodgepole pine ( Pinus contorta) and western hemlock (Tsuga heterophtlla). Bog birch (Betula...Weyerhaeuser production facilities and storage at the proposed export facility. Tropolone methyl ethers have also been mentioned as potential contam

DuPont IsoTherming clean fuel technology

Energy Technology Data Exchange (ETDEWEB)

Levinski, E. [E.I. DuPont Co., Wilmington, DE (United States)

2009-07-01

This poster described a hydroprocessing technology that DuPont has acquired from Process Dynamics, Inc. The IsoTherming clean fuel technology significantly reduces sulphur in motor fuels. The technology provides petroleum refiners the solution for meeting ultra low sulphur diesel requirements, at much lower costs than conventional technologies. IsoTherming hydroprocessing operates in a kinetically limited mode, with no mass transfer limitation. Hydrogen is delivered to the reactor in the liquid phase as soluble hydrogen, allowing for much higher space velocities than conventional hydrotreating reactors. Treated diesel is recycled back to the inlet of the reactor, generating less heat and more hydrogen into the reactor. The process results in a more isothermal reactor operation that allows for better yields, fewer light ends and greater catalyst life. The technology reduces coking, because the process provides enough hydrogen in the solution when cracking reactions take place. As a result, the process yields longer catalyst life. Other advantages for refiners include lower total investment; reduced equipment delivery lead times; reduced maintenance and operating costs; and configuration flexibility. tabs., figs.

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

Investigating the effect of adjusted DuPont ratio and its components on investor's decisions in short and long term

Directory of Open Access Journals (Sweden)

Parvaneh Khaleghi Kasbi

2014-03-01

Full Text Available This paper investigates the effect of adjusted DuPont ratio and its components on investors’ decisions in short and long term. The primary objective of this study is to find the effect of adjusted DuPont ratio and its components on herding behavior of investors in one and several year period. Hence, 85 corporations as the member of Tehran stock exchange over the period 2006-2011 are selected. In order to recognize the herding, by market index consideration, the herded β and in order to hypothesis validity SPSS software and multivariable linear regression have been used. As the results of this study indicate, the adjusted DuPont ratio and its components have more effect on investors’ decisions in short term but in long the period, the effect of this ratio on herding investors’ behavior are reduced. Furthermore, from the two components of adjusted DuPont ratio, profit margin has more effect on investor's decisions.

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...

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 ...

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)

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

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

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.)

Marion duPont Scott Equine Medical Center uses innovative lameness treatment

OpenAIRE

Lee, Kate

2009-01-01

Virginia Tech's Marion duPont Scott Equine Medical Center is now offering an equine lameness therapy that prevents further degeneration of the affected joint and offers a longer-lasting benefit than traditional steroid treatment.

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

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

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

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)

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)

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

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

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

Analysis of DuPont and Kodak duplicating films and chemistries in a Fultron spray processor

Science.gov (United States)

Weinstein, M. S.

1972-01-01

A test program was conducted with duPont duplicating film type SR 112 and SCOLOR developer and Kodak duplicating film types 2430, 2422, and FE 2628 (SO-467) and MX-641 developer to determine sensitometric and image quality characteristics of these materials when used with a fultron spray processor. The test results show that the SCOLOR developer foams excessively in the fultron processor when used with or without the addition of an antifoaming agent. The Kodak type FE 2628 film with MX-641 chemistry had the longest linear Log E range at a 1.0 gamma. Sensitometric curves and granularity traces for all film process combinations tested are included.

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

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

Marion duPont Scott Equine Medical Center offers new treatment for lameness

OpenAIRE

Musick, Marjorie

2006-01-01

The Virginia-Maryland Regional College of Veterinary Medicine's Marion duPont Scott Equine Medical Center has begun offering a new therapy for treating lameness associated with osteoarthritis and cartilage damage in horses, a problem that affects all segments of the equine industry.

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

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

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

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.

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

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)

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)

The DuPont, Monsanto, General Electric open-quotes Lasagnaclose quotes Remediation Project - joint R ampersand D and financing

International Nuclear Information System (INIS)

Palmer, P.A.

1995-01-01

The Lasagna project is the first of what we expect will be several large cooperative projects between industry consortia and government to develop improved remediation technologies. In 1992, Monsanto Company began contacting other major corporations to see if they were experiencing similar difficulties in applying cost-effective, or even workable technologies for industrial site remediation. Both General Electric and DuPont were early participants in the effort to develop a meeting with the EPA to discuss technical problems faced in cleanup, research needs, and ways to accelerate development of more cost-effective techniques. This paper provides some background on how this cooperative process came to reality, what the Lasagna process is and how the cooperative arrangements and financing are structured

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

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

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.

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.

A Decomposition of Hospital Profitability: An Application of DuPont Analysis to the US Market.

Science.gov (United States)

Turner, Jason; Broom, Kevin; Elliott, Michael; Lee, Jen-Fu

2015-01-01

This paper evaluates the drivers of profitability for a large sample of U.S. hospitals. Following a methodology frequently used by financial analysts, we use a DuPont analysis as a framework to evaluate the quality of earnings. By decomposing returns on equity (ROE) into profit margin, total asset turnover, and capital structure, the DuPont analysis reveals what drives overall profitability. Profit margin, the efficiency with which services are rendered (total asset turnover), and capital structure is calculated for 3,255 U.S. hospitals between 2007 and 2012 using data from the Centers for Medicare & Medicaid Services' Healthcare Cost Report Information System (CMS Form 2552). The sample is then stratified by ownership, size, system affiliation, teaching status, critical access designation, and urban or non-urban location. Those hospital characteristics and interaction terms are then regressed (OLS) against the ROE and the respective DuPont components. Sensitivity to regression methodology is also investigated using a seemingly unrelated regression. When the sample is stratified by hospital characteristics, the results indicate investor-owned hospitals have higher profit margins, higher efficiency, and are substantially more leveraged. Hospitals in systems are found to have higher ROE, margins, and efficiency but are associated with less leverage. In addition, a number of important and significant interactions between teaching status, ownership, location, critical access designation, and inclusion in a system are documented. Many of the significant relationships, most notably not-for-profit ownership, lose significance or are predominately associated with one interaction effect when interaction terms are introduced as explanatory variables. Results are not sensitive to the alternative methodology. The results of the DuPont analysis suggest that although there appears to be convergence in the behavior of NFP and IO hospitals, significant financial differences remain

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.

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

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

Application of DuPont photopolymer films to automotive holographic display

Science.gov (United States)

Nakazawa, Norihito; Ono, Motoshi; Takeuchi, Shoichi; Sakurai, Hiromi; Hirano, Masahiro

1998-03-01

Automotive holographic head-up display (HUD) systems employing DuPont holographic photopolymer films are presented. Holographic materials for automotive application are exposed to severe environmental conditions and are required high performance. This paper describes the improvement of DuPont photopolymer films for the automotive use, critical technical issues such as optical design, external color and stray light. The holographic HUD combiner embedded in a windshield of an automobile has peculiar problems called external color. Diffraction light from holographic combiner makes its external color tone stimulative. We have introduced RGB three color recording and color simulation in order to improve the external color. A moderate external color tone was realized by the optimization in terms of wavelengths and diffraction efficiencies of the combiner hologram. The stray light called flare arises from a reflection by glass surface of windshield. We have developed two techniques to avoid the flare. First is a diffuser type trap beam guard hologram which reduces the intensity of the flare. Second is the optimization of the design of hologram so that the incident direction of flare is lower than the horizon line. As an example of automotive display a stand-alone type holographic HUD system attached on the dashboard of an automobile is demonstrated, which provides useful driving information such as route guidance. The display has a very simple optical system that consists of only a holographic combiner and a vacuum fluorescent display. Its thin body is only 35 mm high and does not obstruct driver's view. The display gives high contrast and wide image.

Evaluating Banking Profit Performance in Ghana during and post Profit Decline: A five Step Du-Pont Approach

Directory of Open Access Journals (Sweden)

Baah Aye Kusi

2015-11-01

Full Text Available In this study we aimed at three objectives. First, identify and rank banks based on a composite score comprising of all five du-pont variables. Second, we identify variables in the five step du-pont set up that are most likely to influence bank ROE during and post profit declining periods. And third, we estimate a model to capture the variables that drive bank ROE during and post profit declining periods. We first establish from our rankings that, foreign banks in Ghana performed better during profit declining periods while the local banks performed better in post profit decline periods using the top ten banks as a benchmark in both periods. Employing Pearson correlation coefficients matrix, we recognized that operating profit margin, asset turnover and leverage were most likely to influence bank ROE in both time periods. We further employ OLS regression and find that bank ROE was impacted by operating profit margin and leverage during profit declining periods and post profit decline while tax effect added up in post profit declining periods.

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

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

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.

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

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)

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

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)

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)

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.

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

Soil salinity data from Bayou Dupont and flanking marshes, New Orleans, LA, 2015-09-16 to 2016-03-30 (NCEI Accession 0151633)

Data.gov (United States)

National Oceanic and Atmospheric Administration, Department of Commerce — The project restored both structural and habitat functions of Bayou Dupont and flanking marshes. The project created and nourished marsh and restored a ridge on the...

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

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

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

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

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

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…

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...

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

Health assessment for E. I. Dupont Newport Plant Landfill, Newport, Delaware, Region 3. CERCLIS No. DED980555122. Preliminary report

International Nuclear Information System (INIS)

1988-01-01

The E.I. Dupont Newport Plant Landfill, a seven acre site located adjacent to the Dupont Pigment Plant in Newport, Delaware, was used to bury paint pigments (heavy metals and chlorinated solvents) and radioactive materials from 1902 to 1975. The site was closed in 1975, the surface covered, graded and vegetated, and groundwater monitoring wells installed on- and off-site. Heavy metals (cadmium, barium, lead, zinc), and chlorinated solvents (trichloroethylene, tetrachloroethylene) have been measured in groundwater on-site. Radiation levels appear to be close to background. The site is considered to be of potential public health concern because of the risk to human health caused by the possibility of exposure to hazardous substances via groundwater. More information is needed concerning levels of contamination in private wells and the Christina River, the direction of groundwater flow, the size and extent of the groundwater contamination plume, the location of the private wells and the public water supply wells, and the composition of the population down gradient of the site

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…

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.

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

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)

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)

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

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.

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

Decontamination of process equipment using recyclable chelating solvent

International Nuclear Information System (INIS)

Jevec, J.; Lenore, C.; Ulbricht, S.

1995-01-01

The Department of Energy (DOE) is now faced with the task of meeting decontamination and decommissioning obligations at numerous facilities by the year 2019. Due to the tremendous volume of material involved, innovative decontamination technologies are being sought that can reduce the volumes of contaminated waste materials and secondary wastes requiring disposal. With sufficient decontamination, some of the material from DOE facilities could be released as scrap into the commercial sector for recycle, thereby reducing the volume of radioactive waste requiring disposal. Although recycling may initially prove to be more costly than current disposal practices, rapidly increasing disposal costs are expected to make recycling more and more cost effective. Additionally, recycling is now perceived as the ethical choice in a world where the consequences of replacing resources and throwing away reusable materials are impacting the well-being of the environment. Current approaches to the decontamination of metals most often involve one of four basic process types: (1) chemical, (2) manual and mechanical, (3) electrochemical, and (4) ultrasonic. open-quotes Hardclose quotes chemical decontamination solutions, capable of achieving decontamination factors (Df's) of 50 to 100, generally involve reagent concentrations in excess of 5%, tend to physically degrade the surface treated, and generate relatively large volumes of secondary waste. open-quotes Softclose quotes chemical decontamination solutions, capable of achieving Df's of 5 to 10, normally consist of reagents at concentrations of 0.1 to 1%, generally leave treated surfaces in a usable condition, and generate relatively low secondary waste volumes. Under contract to the Department of Energy, the Babcock ampersand Wilcox Company is developing a chemical decontamination process using chelating agents to remove uranium compounds and other actinide species from process equipment

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.

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

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

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

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)

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

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

Le contrôle de l’activité législative de la nation en 1789 : l’opinion de Dupont de Nemours

Directory of Open Access Journals (Sweden)

Anthony Mergey

2014-06-01

Full Text Available On the eve of the French Revolution, Dupont de Nemours became infatuated with constitutional issues that had hitherto been the exclusive domain of some of his fellow physiocrats, in particular Le Mercier de La Rivière. Endorsing the ideas of the time, he was convinced of the necessity of creating a new political regime that would take the features of a moderate monarchy. The prince would thus be submitted to constitutional rules, sovereignty transferred to the nation and the legislative power exercised by its representatives. However, even if sharing authority between the nation and the king was then claimed, the representatives of the nation should not exercise their prerogatives outside a clearly established, defined constitutional framework. Partly consistent with the original physiocratic theories, Dupont de Nemours identified constitutional norms that should be promoted and protected from abuses of the legislature. With this aim in mind, he didn’t plan to resort to a jurisdictional authority, because abuses of Ancien Régime parliaments were still live memories. He decided, in an original way, to entrust the heavy burden of controlling legislative activity to political instances, namely, successively, the people and the monarch. Quite naturally, these suggestions, expressed a few months apart, exposed various forms of control. The examination carried out by the nation appeared as a prerequisite and tended to verify the constitutionality of the law, while the one carried out by the monarch through his veto, as holder of the executive power, was more of a control of opportunity devoid of binding. Although the omnipotence of the revolutionary legicentrism seemed incompatible with such a system of verification of the norm, it remains that Dupont de Nemours was part of a current which considerably developed during the XVIIIth century and which found a certain resonance in the decade following the Revolution.

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

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.)

Process waste assessment for the Radiography Laboratory

International Nuclear Information System (INIS)

Phillips, N.M.

1994-07-01

This Process Waste Assessment was conducted to evaluate the Radiography Laboratory, located in Building 923. It documents the processes, identifies the hazardous chemical waste streams generated by these processes, recommends possible ways to minimize waste, and serves as a reference for future assessments of this facility. The Radiography Laboratory provides film radiography or radioscopy (electronic imaging) of weapon and nonweapon components. The Radiography Laboratory has six x-ray machines and one gamma ray source. It also has several other sealed beta- and gamma-ray isotope sources of low microcurie (μCi) activity. The photochemical processes generate most of the Radiography Laboratory's routinely generated hazardous waste, and most of that is generated by the DuPont film processor. Because the DuPont film processor generates the most photochemical waste, it was selected for an estimated material balance

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

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

Marketingová strategie u vybraného produktu na příkladu společnosti DuPont

OpenAIRE

Wojnar, Michal

2009-01-01

The bachelor thesis "Marketing Strategy of a Chosen Product on The Example of DuPont" is aimed at analyzing the marketing strategy and marketing mix of the Elvaloy material which is used in paving segment. The main applied methodology is called situation analysis through which the author compares individual key markets, reviews internal and external conditions of the company and studies the competition. The main objective of the thesis is also to present diminishing differences between indust...

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

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

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)

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

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

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'

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

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

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

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)

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.

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

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

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

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)

Overview of an energy management process

International Nuclear Information System (INIS)

Chantraine, P.

2004-01-01

Invista is a global and vertically integrated fiber, resin and intermediates business which belonged to Dupont but is now a subsidiary of Koch Industries. A background of Invista and its former relationship with Dupont was presented. This presentation was based on goals and work done as Dupont Canada Inc., up to the end of 2003. Details of Invista's approach to climate change in Canada were provided along with the company's relationship with Natural Resources Canada. The historical position of Dupont Canada was reviewed in detail, including their commitment to voluntary approach; participation in the national process; their goal of 85 per cent reduction in greenhouse gas (GHG) by 2000; 93 per cent reduction in nitrous oxide emissions; energy efficiency goals; and continuing growth of the company. An outline and mission of the energy management team established in 1974 was presented, with details of the 1974 oil shortage, stabilization in the 1980s through to rises in electricity prices in the 1990s and concerns over climate change in recent years. Details of the team's operational procedures were presented. Results were presented in graph form and include: total energy use from 1972 to 2003 as well as cumulative energy conservation projects and resulting energy savings. Examples of activities and projects were provided, including details of energy performance contracting. It was concluded that in order to conserve energy, top management support was necessary, as well as passion and dedication in both leaders and teams. A broad scope for creativity in finding solutions within evolving constraints was also important, as was the nurturing of capability, capacity and recognition for results achieved. tabs., figs

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.

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

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.)

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

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

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

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

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)

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

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)

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

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

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

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

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

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)

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

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

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)

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

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

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.)

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

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

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)

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.)

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)

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.)

Causes of death of thoroughbred racehorses at Octavio Dupont Veterinary Hospital, Brazilian Jockey Club, Rio de Janeiro

Directory of Open Access Journals (Sweden)

Marsel C. Pereira

2012-03-01

Full Text Available There are few studies that approach the epidemiology of deaths in racehorses in a broad manner. The majority focus on a specific affection or procedure. Brazil does not have a program instituted for the monitoring of deaths of horses. By means of a descriptive study in association with a multivariate analysis method, an epidemiologic profile was determined for deaths related to musculoskeletal (MS, gastrointestinal (GI, respiratory (RES systems, neurologic origin (NEU and sudden death (SD for the years of 2002 to 2008, at the Octavio Dupont Veterinary Hospital-Rio de Janeiro (ODVH. Males comprised the majority of deaths and that deaths were related to, decreasing order, MS>GI>SD>NEU>RES, with respect to general mortality rate per large group of determined causes (TSPMr. The majority of deaths registered included horses aged four to five years (ID4-ID5. We observed the following correspondence relations: (3-year period = SM - ID>5 - SD; ID>5 - GI; ID4-5 - MS; SF - ID5 - GI; SF - ID5. The present study points out the importance and necessity of epidemiologic studies of lesions in horses, based on diagnosis for the recognition of predisposing factors and prevention.

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.

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

THE IMPACT OF THE CONVERSION OF HISTORICAL COST TO CORRECTED CURRENT COST ON RETURN ON ASSETS AND EQUITY: ANALYSIS FROM DUPONT SYSTEM.

Directory of Open Access Journals (Sweden)

Roberto Silva da Penha

2013-12-01

Full Text Available Historical cost is the most used form of accounting valuation, but it does not reflect the effect of currency in time. The Theory of Corrected Current Cost, albeit with limitations, may show evidence of real and relevant accounting information of the entity. In this context, a question can be formulated: What is the impact of the conversion of historical cost to corrected current cost in return on assets and equity? In order to answer this question, the DuPont system was applied in a model company as a tool for analysis. The results showed that the return of assets at historical cost and corrected current cost were 12.11% and 3.14%, respectively, a decrease of 74.07%, and the result of the return on equity, the historical cost and corrected current cost were 20.59% and 4.96%, respectively, decreased 75.91%. Based on these variations it is observed that the impacts of the conversion of the corrected current cost were significant,confirming the importance of the adoption of this practice, whenever possible.

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.

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

Tritium processing using metal hydrides

International Nuclear Information System (INIS)

Mallett, M.W.

1986-01-01

E.I. duPont de Nemours and Company is commissioned by the US Department of Energy to operate the Savannah River Plant and Laboratory. The primary purpose of the plant is to produce radioactive materials for national defense. In keeping with current technology, new processes for the production of tritium are being developed. Three main objectives of this new technology are to ease the processing of, ease the storage of, and to reduce the operating costs of the tritium production facility. Research has indicated that the use of metal hydrides offers a viable solution towards satisfying these objectives. The Hydrogen and Fuels Technology Division has the responsibility to conduct research in support of the tritium production process. Metal hydride technology and its use in the storage and transportation of hydrogen will be reviewed

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)

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.

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

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

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

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)

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

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.

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)

Dynawave froth scrubbing technology

International Nuclear Information System (INIS)

McLean, J.E.

1989-01-01

The DynaWave family of scrubbers was developed and patented by DuPont in the 1970's and is used extensively within the company. Because the technology was treated as confidential, information about the scrubbers was not published outside of DuPont. In 1986 Monsanto Enviro-Chem was asked to install a Reverse Jet scrubber at DuPont's Burnside spent acid recovery pant to replace an aging spray humidifying tower. Enviro-Chem was impressed with the technology and entered into a licensing agreement with DuPont in late 1987. This paper addresses the application of DynaWave froth scrubbing technology to metallurgical plant process and effluent gas streams. Froth scrubber design principles an sampling results from pilot plant trails on fluid bed roaster off-gas are presented

Uranium hexaflouride freezer/sublimer process simulator/trainer

International Nuclear Information System (INIS)

Carnal, C.L.; Belcher, J.D.; Tapp, P.A.; Ruppel, F.R.; Wells, J.C.

1991-01-01

This paper describes a software and hardware simulation of a freezer/sublimer unit used in gaseous diffusion processing of uranium hexafluoride (UF 6 ). The objective of the project was to build a plant simulator that reads control signals and produces plant signals to mimic the behavior of an actual plant. The model is based on physical principles and process data. Advanced Continuous Simulation Language (ACSL) was used to develop the model. Once the simulation was validated with actual plant process data, the ACSL model was translated into Advanced Communication and Control Oriented Language (ACCOL). A Bristol Babcock Distributed Process Controller (DPC) Model 3330 was the hardware platform used to host the ACCOL model and process the real world signals. The DPC will be used as a surrogate plant to debug control system hardware/software and to train operators to use the new distributed control system without disturbing the process. 2 refs., 4 figs

Polymerization and processing of organic polymers in a magnetic field

Energy Technology Data Exchange (ETDEWEB)

Douglas, E.P. [Los Alamos National Laboratory, NM (United States)

1995-05-01

The use of magnetic fields to affect the structure and properties of polymeric materials remains an area of great promise. Liquid crystalline polymers have been actively studied over the past 20 years for use in high performance structural applications. In particular, highly oriented fibers can exhibit remarkable increases in strength to weight performance compared to conventional materials. For example, the fibers marketed by DuPont under the tradename Kevlar are 20 times stronger than steel on an equivalent weight basis. However, larger bulk parts do not exhibit the same increases in strength due to a lack of orientation of the polymer molecules. Magnetic field processing of polymers remains an attractive solution to this problem.

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

E.I. duPont de Nemours & Co., Process Section Manufacturing Division, November 1956

Energy Technology Data Exchange (ETDEWEB)

NONE

1956-12-15

This document presents monthly progress at the Savannah River Plant in the 100, 200, 300, and 400 Areas. In the 100 Area (reactors), the power ascension program, the productivity improvement program and general progress is given. Tritium processing, 200-F increased capacity, and waste disposal are reported for the solvent extraction plants in the 200 Area. The increased productivity program is reported for the 300 Area. Dana plant activities and 400-D progress are reported in the 400 area report (heavy water).

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

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.

The Defense Waste Processing Facility, from vision to reality

International Nuclear Information System (INIS)

Randall, C.T.

2000-01-01

When the Savannah River Plant began operation in the early 1950's producing nuclear materials for the National defense, liquid, highly radioactive waste was generated as a by-product. Since that time the waste has been stored in large, carbon steel tanks that are buried underground. In 1960 one of the tanks developed a leak, and before recovery measures could be taken, about 25-gallons of radioactive salt solution had overflowed the secondary liner and seeped into the soil surrounding the tank. Significant improvements to the tanks were made, but constant surveillance was still required. Thus, the opinion began forming that storage of the mobile, highly radioactive waste in tanks was not a responsible long-term practice. So in the late 1960's the Savannah River Laboratory began research to find a suitable long-term solution to the waste disposal problem. Several alternative waste forms were evaluated, and in 1972 the first Savannah River waste was vitrified on a laboratory scale. By the mid-1970's, the DuPont Company, prime contractor at the Savannah River Plant, began to develop a vision of constructing America's first vitrification plant to immobilize the high level radioactive waste in borosilicate glass. This vision was later championed by DuPont in the form of a vitrification plant called the Defense Waste Processing Facility (DWPF). Today, the DWPF processes Savannah River High Level Waste sludge turning it into a solid, durable waste form of borosilicate glass. The DWPF is the world's largest vitrification facility. It was brought to reality through over 25-years of research and 13-years of careful construction, tests, and reviews at a cost of approximately $3 billion dollars

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

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.

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

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

A DOUBLE-RING ALGORITHM FOR MODELING SOLAR ACTIVE REGIONS: UNIFYING KINEMATIC DYNAMO MODELS AND SURFACE FLUX-TRANSPORT SIMULATIONS

International Nuclear Information System (INIS)

Munoz-Jaramillo, Andres; Martens, Petrus C. H.; Nandy, Dibyendu; Yeates, Anthony R.

2010-01-01

The emergence of tilted bipolar active regions (ARs) and the dispersal of their flux, mediated via processes such as diffusion, differential rotation, and meridional circulation, is believed to be responsible for the reversal of the Sun's polar field. This process (commonly known as the Babcock-Leighton mechanism) is usually modeled as a near-surface, spatially distributed α-effect in kinematic mean-field dynamo models. However, this formulation leads to a relationship between polar field strength and meridional flow speed which is opposite to that suggested by physical insight and predicted by surface flux-transport simulations. With this in mind, we present an improved double-ring algorithm for modeling the Babcock-Leighton mechanism based on AR eruption, within the framework of an axisymmetric dynamo model. Using surface flux-transport simulations, we first show that an axisymmetric formulation-which is usually invoked in kinematic dynamo models-can reasonably approximate the surface flux dynamics. Finally, we demonstrate that our treatment of the Babcock-Leighton mechanism through double-ring eruption leads to an inverse relationship between polar field strength and meridional flow speed as expected, reconciling the discrepancy between surface flux-transport simulations and kinematic dynamo models.

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

The origin and distribution of HAPs elements in relation to maceral composition of the A1 lignite bed (Paleocene, Calvert Bluff Formation, Wilcox Group), Calvert mine area, east-central Texas

Science.gov (United States)

Crowley, Sharon S.; Warwick, Peter D.; Ruppert, Leslie F.; Pontolillo, James

1997-01-01

The origin and distribution of twelve potentially Hazardous Air Pollutants (HAPs; As, Be, Cd, Cr, Co, Hg, Mn, Ni, Pb, Sb, Se, and U) identified in the 1990 Clean Air Act Amendments were examined in relation to the maceral composition of the A1 bed (Paleocene, Calvert Bluff Formation, Wilcox Group) of the Calvert mine in east-central Texas. The 3.2 m-thick A1 bed was divided into nine incremental channel samples (7 lignite samples and 2 shaley coal samples) on the basis of megascopic characteristics. Results indicate that As, Cd, Cr, Ni, Pb, Sb, and U are strongly correlated with ash yield and are enriched in the shaley coal samples. We infer that these elements are associated with inorganic constituents in the coal bed and may be derived from a penecontemporaneous stream channel located several kilometers southeast of the mining block. Of the HAPs elements studied, Mn and Hg are the most poorly correlated to ash yield. We infer an organic association for Mn; Hg may be associated with pyrite. The rest of the trace elements (Be, Co, and Se) are weakly correlated with ash yield. Further analytical work is necessary to determine the mode of occurrence for these elements. Overall, concentrations of the HAPs elements are generally similar to or less than those reported in previous studies of lignites of the Wilcox Group, east-central region, Texas. Petrographic analysis indicates the following ranges in composition for the seven lignite samples: liptinites (5–8%), huminites (88–95%), and inertinites (trace amounts to 7%). Samples from the middle portion of the A1 bed contain abundant crypto-eugelinite compared to the rest of the samples; this relationship suggests that the degradation of plant material was an important process during the development of the peat mire. With the exception of Hg and Mn, relatively low levels of the HAPs elements studied are found in the samples containing abundant crypto-eugelinite. We infer that the peat-forming environment for this

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

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

Scientific Opinion related to a notification from DuPont Nutrition Biosciences Aps on behenic acid from mustard seeds to be used in the manufacturing of certain emulsifiers pursuant to Article 21(2) of Regulation (EU) No 1169/2011 – for permanent exemption from labelling

DEFF Research Database (Denmark)

Sjödin, Anders Mikael

2016-01-01

Following a request from the European Commission, the Panel on Dietetic Products, Nutrition and Allergies (NDA) was asked to deliver a scientific opinion related to a notification from DuPont Nutrition Biosciences Aps on behenic acid from mustard seeds to be used in the manufacturing of certain...... and peptides may not be carried over into behenic acid after the two distillation steps reported in the manufacturing process in amounts beyond 1 mg/kg. The Panel notes that the maximum amount of mustard protein that could be consumed from emulsifiers manufactured from behenic acid (E 470a, E 471 and E 477......) on a single occasion by an adult under the proposed conditions of use would be around 0.00119 mg, which is about 1,000 times lower than the protein doses reported to trigger allergic reactions in mustard-allergic individuals (around 1 mg). On the basis of the data presented, the Panel concludes that oral...

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

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).

End of project report on degradation processes in hydrogen fuel cells.

Science.gov (United States)

2008-01-01

Proton exchange membrane (PEM) fuel cells are one of the most popular types of fuel cells. They operate similarly to others with the electrolyte material inbetween the electrodes being a patented polymer called Nafion, made by DuPont. This polyelec...

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).

Oxy-gasoline torch. Innovative technology summary report

International Nuclear Information System (INIS)

1998-12-01

Under the deactivation and decommissioning (D and D) Implementation Plan of the US Department of Energy's (DOE) Fernald Environmental Management Project (FEMP), non-recyclable process components and debris that are removed from buildings undergoing D and D are disposed of in an on-site disposal facility (OSDF). Critical to the design and operation of the FEMP's OSDF are provisions to protect against subsidence of the OSDF's cap. Subsidence of the cap could occur if void spaces within the OSDF were to collapse under the overburden of debris and the OSDF cap. Subsidence may create significant depressions in the OSDF's cap in which rainwater could collect and eventually seep into the OSDF. To minimize voids in the FEMP's OSDF, large metallic components are cut into smaller segments that can be arranged more compactly when placed in the OSDF. Component segmentation using an oxy-acetylene cutting torch was the baseline approach used by the FEMP's D and D contractor on Plant 1, Babcock and Wilcox (B and W) Services, Inc., for the dismantlement and size-reduction of large metal components. Although this technology has performed satisfactorily, improvements are sought in the areas of productivity, airborne contamination, safety, and cost. This demonstration investigated the feasibility of using an oxy-gasoline torch as an alternative to the baseline oxy-acetylene torch for segmenting D and D components. This report provides a comparative analysis of the cost and performance of the baseline oxy-acetylene torch currently used by B and W Services, Inc., and the innovative oxy-gasoline torch

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

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

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

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

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

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

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.

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)

O IMPACTO DA CONVERSÃO DO CUSTO HISTÓRICO AO CUSTO CORRENTE CORRIGIDO NO RETORNO SOBRE ATIVOS E PATRIMÔNIO LÍQUIDO: ANÁLISE A PARTIR DO SISTEMA DUPONT.

Directory of Open Access Journals (Sweden)

Roberto Silva da Penha

2013-07-01

Full Text Available Historical cost is the most used form of accounting valuation, but it does not reflect the effect of currency in time. The Theory of Corrected Current Cost, albeit with limitations, may show evidence of real and relevant accounting information of the entity. In this context, a question can be formulated: What is the impact of the conversion of historical cost to corrected current cost in return on assets and equity? In order to answer this question, the DuPont system was applied in a model company as a tool for analysis. The results showed that the return of assets at historical cost and corrected current cost were 12.11% and 3.14%, respectively, a decrease of 74.07%, and the result of the return on equity, the historical cost and corrected current cost were 20.59% and 4.96%, respectively, decreased 75.91%. Based on these variations it is observed that the impacts of the conversion of the corrected current cost were significant, confirming the importance of the adoption of this practice, whenever possible.

FutureGen 2.0 Oxy-Coal Combustion Carbon Capture Plant Pre-FEED Design and Cost

Energy Technology Data Exchange (ETDEWEB)

Flanigan, Tom; Pybus, Craig; Roy, Sonya; Lockwood, Frederick; McDonald, Denny; Maclnnis, Jim

2011-09-30

This report summarizes the results of the Pre-Front End Engineering Design (pre-FEED) phase of a proposed advanced oxy-combustion power generation plant to repower the existing 200 MWe Unit 4 at Ameren Energy Resources’ (AER) Meredosia Power Plant. AER has formed an alliance with Air Liquide Process and Construction, Inc. (ALPC) and Babcock & Wilcox Power Generation Group (B&W PGG) for the design, construction, and testing of the facility, and has contracted with URS Corporation (URS) for preliminary design and Owner’s engineering services. The Project employs oxy-combustion technology – combustion of coal with nearly pure oxygen and recycled flue gas (instead of air) – to capture approximately 90% of the flue gas CO2 for transport and sequestration by another Project. Plant capacity and configuration has been developed based on the B&W PGG-ALPC cool recycle process firing high-sulfur bituminous coal fuel, assuming baseload plant operation to maximize existing steam turbine capability, with limited consideration for plant redundancy and performance optimization in order to keep plant costs as low as practical. Activities and preliminary results from the pre-FEED phase addressed in this report include the following: Overall plant thermal performance; Equipment sizing and system configuration; Plant operation and control philosophy; Plant emissions and effluents; CO₂ production and recovery characteristics; Project cost estimate and economic evaluation; Integrated project engineering and construction schedule; Project risk and opportunity assessment; Development of Project permitting strategy and requirements During the Phase 2 of the Project, additional design details will be developed and the Phase 1 work products updated to support actual construction and operation of the facility in Phase 3. Additional information will be provided early in Phase 2 to support Ameren-Environmental in finalizing the appropriate permitting strategies and permit

Radioactive Demonstration of Caustic Recovery from Low-Level Alkaline Nuclear Waste by an Electrochemical Separation Process

International Nuclear Information System (INIS)

Hobbs, D.T.

1998-04-01

Bench-scale radioactive tests successfully demonstrated an electrochemical process for the recovery of sodium hydroxide (caustic) from Decontaminated Salt Solution produced from the In-Tank Precipitation and Effluent Treatment Processes at the Savannah River Site (SRS). This testing evaluated two membranes: an organic-based membrane, Nafion Type 350, manufactured by E. I. duPont de Nemours ampersand Company, Inc. (DuPont) and an inorganic-based membrane, NAS D, being developed by Ceramatec. Both membranes successfully separated caustic from radioactive SRS waste.Key findings of the testing indicate the following attributes and disadvantages of each membrane. The commercially-available Nafion membrane proved highly conductive. Thus, the electrochemical cell can operate at high current density minimizing the number of cells at the desired volumetric processing rate. Testing indicated cesium transported across the Nafion membrane into the caustic product. Therefore, the caustic product will contain low-levels of radioactive cesium due to the presence of 134,137 Cs in the waste feed. To meet customer requirements, a post treatment stage may prove necessary to remove radioactive cesium resulting in increased overall process costs and decreased cost savings. In contrast to the Nafion membrane, the NAS D membrane demonstrated the production of caustic with much lower levels of gamma radioactivity ( 137 Cs activity was < 51 dpm/g). Therefore, the caustic product could possibly release for onsite/offsite use without further treatment. The NAS D membrane remains in the development stage and does not exist as a commercial product. Operating costs and long-term membrane durability remain unknown.Caustic recovery has been successfully demonstrated in a bench-scale, 2-compartment electrochemical reactor operated for brief periods of time with simulated and radioactive waste solutions and two different types of membranes. The next phase of testing should be directed at (1

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

75 FR 51845 - Chrysler Group, LLC Manufacturing Division St. Louis North Plant Including On-Site Leased Workers...

Science.gov (United States)

2010-08-23

... produce performance coating solutions for vehicles. The company reports that workers leased from DuPont Performance Coatings, a wholly-owned subsidiary of E.I. DuPont de Nemours Company, OEM, were employed on-site..., C R Associates, Syncreon, Robinson Solutions and Dupont Performance Coatings Fenton, MO; Amended...

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.

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

Cost estimate of high-level radioactive waste containers for the Yucca Mountain Site Characterization Project

Energy Technology Data Exchange (ETDEWEB)

Russell, E.W.; Clarke, W. [Lawrence Livermore National Lab., CA (United States); Domian, H.A. [Babcock and Wilcox Co., Lynchburg, VA (United States); Madson, A.A. [Kaiser Engineers California Corp., Oakland, CA (United States)

1991-08-01

This report summarizes the bottoms-up cost estimates for fabrication of high-level radioactive waste disposal containers based on the Site Characterization Plan Conceptual Design (SCP-CD). These estimates were acquired by Babcock and Wilcox (B&S) under sub-contract to Lawrence Livermore National Laboratory (LLNL) for the Yucca Mountain Site Characterization Project (YMP). The estimates were obtained for two leading container candidate materials (Alloy 825 and CDA 715), and from other three vendors who were selected from a list of twenty solicited. Three types of container designs were analyzed that represent containers for spent fuel, and for vitrified high-level waste (HLW). The container internal structures were assumed to be AISI-304 stainless steel in all cases, with an annual production rate of 750 containers. Subjective techniques were used for estimating QA/QC costs based on vendor experience and the specifications derived for the LLNL-YMP Quality Assurance program. In addition, an independent QA/QC analysis is reported which was prepared by Kasier Engineering. Based on the cost estimates developed, LLNL recommends that values of $825K and $62K be used for the 1991 TSLCC for the spent fuel and HLW containers, respectively. These numbers represent the most conservative among the three vendors, and are for the high-nickel anstenitic steel (Alloy 825). 6 refs., 7 figs.

Cost estimate of high-level radioactive waste containers for the Yucca Mountain Site Characterization Project

International Nuclear Information System (INIS)

Russell, E.W.; Clarke, W.; Domian, H.A.; Madson, A.A.

1991-08-01

This report summarizes the bottoms-up cost estimates for fabrication of high-level radioactive waste disposal containers based on the Site Characterization Plan Conceptual Design (SCP-CD). These estimates were acquired by Babcock and Wilcox (B ampersand S) under sub-contract to Lawrence Livermore National Laboratory (LLNL) for the Yucca Mountain Site Characterization Project (YMP). The estimates were obtained for two leading container candidate materials (Alloy 825 and CDA 715), and from other three vendors who were selected from a list of twenty solicited. Three types of container designs were analyzed that represent containers for spent fuel, and for vitrified high-level waste (HLW). The container internal structures were assumed to be AISI-304 stainless steel in all cases, with an annual production rate of 750 containers. Subjective techniques were used for estimating QA/QC costs based on vendor experience and the specifications derived for the LLNL-YMP Quality Assurance program. In addition, an independent QA/QC analysis is reported which was prepared by Kasier Engineering. Based on the cost estimates developed, LLNL recommends that values of $825K and $62K be used for the 1991 TSLCC for the spent fuel and HLW containers, respectively. These numbers represent the most conservative among the three vendors, and are for the high-nickel anstenitic steel (Alloy 825). 6 refs., 7 figs

Subsystem selection for advanced low emission boiler system

Energy Technology Data Exchange (ETDEWEB)

Rodgers, L.W.; Farthing, G.A. [Babcock & Wilcox, Alliance, OH (United States). Research and Development Div.; Gorrell, R.L. [Babcock & Wilcox, Barberton, OH (United States). Fossil Power Div.

1993-12-31

In 1992 the Pittsburgh Energy Technology Center (PETC) initiated a new program called Combustion 2000. The purpose of the program was to address the design issues facing new and replacement coal-fired power plants. The work presented in this paper was conducted under the low-emission boiler system (LEBS) portion of the program. LEBS major goals are: NO{sub x} - No more than 0.20 lbs per million Btu of fuel input firing bituminous coal; SO{sub x} -- no more than 0.2 lbs of SO{sub 2} per million Btu firing coal with at least 3 lbs of sulfur per million btu; Particulate -- no more than 0.015 lbs per million Btu of fuel input; Waste and Air Toxics -- reduced; and Plant Efficiency -- no less than 38%. Other objectives include reducing waste generation, producing usable by-products, improving ash disposability, and increasing plant thermal efficiency while keeping the cost of electricity comparable to a state-of-the-art plant. The Babcock and Wilcox Company has completed the first year of work toward the development of an advanced low-emission boiler system (LEBS). The results of this work have led to a preliminary engineering design and a plan to address remaining technical uncertainties. This was accomplished by conducting a thorough technical assessment and performing a concept selection analysis. A summary of the results of this work is presented in this paper.

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

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

Water and power installation for Lanzarote (Canary Islands) Spain

Energy Technology Data Exchange (ETDEWEB)

Krebs, F.

1977-03-01

Water producing and electrical generating facilities had to be increased on the Spanish Canary Island of Lanzarote for it to develop beyond its 1972 status. The Ministry of Public Works, prepared specifications for a turnkey installation of two 2500 M/sup 3//D desalting plants and a 5 MW (net) generating system. In 1973 through international competition Babcock and Wilcox, Espanola won a contract to build the installation. In December 1975 the construction of the project was completed. In early January 1976 the first line was producing 103% of the rated production at a thermal efficiency of 120% of the rated value. The second line was modified to operate at a reduced maximum brine temperature and commissioned. This plant was brought to full production which was 115% of the rated production in 2/sup 1///sub 2/ days. Problems and delays during the commissioning, were minimal. A 5 days performance test of the installation was run. During this test the water production was 111/sup 1///sub 2/% of the rated value at a specific energy consumption of 93% of the rated value while the maximum brine temperature was maintained 6/sup 0/C less than the design value. The high performances were achieved because the plants were well designed and built and properly commissioned. During the first 4/sup 1///sub 2/ months that the owner operated the installation it provided water and power as demanded. During that time the average plant production was 107% of the rated capacity.

Differential solubilities of CO[2] and methane and the potential for coupling in TFA

NARCIS (Netherlands)

Golombok, M.

2003-01-01

We wish to add to the recent insightful comments of Wilcox et al. [1] regarding the thermodynamics of light alkane carboxylation. This process is particularly of industrial interest when it comes to methane which is vented at oil production sites where there are significant quantities of

In-house validation study of the DuPont Qualicon BAX system Q7 instrument with the BAX system PCR Assay for Salmonella (modification of AOAC Official Method 2003.09 and AOAC Research Institute Performance-Tested Method 100201).

Science.gov (United States)

Tice, George; Andaloro, Bridget; White, H Kirk; Bolton, Lance; Wang, Siqun; Davis, Eugene; Wallace, Morgan

2009-01-01

In 2006, DuPont Qualicon introduced the BAX system Q7 instrument for use with its assays. To demonstrate the equivalence of the new and old instruments, a validation study was conducted using the BAX system PCR Assay for Salmonella, AOAC Official Method 2003.09, on three food types. The foods were simultaneously analyzed with the BAX system Q7 instrument and either the U.S. Food and Drug Administration Bacteriological Analytical Manual or the U.S. Department of Agriculture-Food Safety and Inspection Service Microbiology Laboratory Guidebook reference method for detecting Salmonella. Comparable performance between the BAX system and the reference methods was observed. Of the 75 paired samples analyzed, 39 samples were positive by both the BAX system and reference methods, and 36 samples were negative by both the BAX system and reference methods, demonstrating 100% correlation. Inclusivity and exclusivity for the BAX system Q7 instrument were also established by testing 50 Salmonella strains and 20 non-Salmonella isolates. All Salmonella strains returned positive results, and all non-Salmonella isolates returned a negative response.

Quantifying Contaminant Mass for the Feasibility Study of the DuPont Chambers Works FUSRAP Site - 13510

Energy Technology Data Exchange (ETDEWEB)

Young, Carl; Rahman, Mahmudur; Johnson, Ann; Owe, Stephan [Cabrera Services Inc., 1106 N. Charles St., Ste. 300, Baltimore, MD 21201 (United States)

2013-07-01

The U.S. Army Corps of Engineers (USACE) - Philadelphia District is conducting an environmental restoration at the DuPont Chambers Works in Deepwater, New Jersey under the Formerly Utilized Sites Remedial Action Program (FUSRAP). Discrete locations are contaminated with natural uranium, thorium-230 and radium-226. The USACE is proposing a preferred remedial alternative consisting of excavation and offsite disposal to address soil contamination followed by monitored natural attenuation to address residual groundwater contamination. Methods were developed to quantify the error associated with contaminant volume estimates and use mass balance calculations of the uranium plume to estimate the removal efficiency of the proposed alternative. During the remedial investigation, the USACE collected approximately 500 soil samples at various depths. As the first step of contaminant mass estimation, soil analytical data was segmented into several depth intervals. Second, using contouring software, analytical data for each depth interval was contoured to determine lateral extent of contamination. Six different contouring algorithms were used to generate alternative interpretations of the lateral extent of the soil contamination. Finally, geographical information system software was used to produce a three dimensional model in order to present both lateral and vertical extent of the soil contamination and to estimate the volume of impacted soil for each depth interval. The average soil volume from all six contouring methods was used to determine the estimated volume of impacted soil. This method also allowed an estimate of a standard deviation of the waste volume estimate. It was determined that the margin of error for the method was plus or minus 17% of the waste volume, which is within the acceptable construction contingency for cost estimation. USACE collected approximately 190 groundwater samples from 40 monitor wells. It is expected that excavation and disposal of

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).

77 FR 284 - Change in Bank Control Notices; Acquisitions of Shares of a Bank or Bank Holding Company

Science.gov (United States)

2012-01-04

... Affairs Officer) 90 Hennepin Avenue, Minneapolis, Minnesota 55480-0291: 1. Gladys E. Youse Trust and.... Additionally, The Gladys E. Youse Trust, and its trustee, Madolyn Y. Babcock, propose to join the Babcock...

Development of High Temperature Superconducting Josephson Junction Device Technology

National Research Council Canada - National Science Library

Myers, Kirsten

1998-01-01

The DuPont program was successful in generating useful knowledge about thallium cuprate materials, photoresist reflow processing, and radiant heater technology though it did not lead to a new junction technology...

The Great Southwest of the Fred Harvey Company and the Santa Fe Railway, edited by Marta Weigle and Barbara A. Babcock. The Heard Museum, Pheonix (printed by The University of Arizona Press, Tucson, for The Heard Museum, 1996

Directory of Open Access Journals (Sweden)

Douglas R. Givens

1997-11-01

Full Text Available One of the more colorful eras in American Southwestern archaelogy is reflected in The Great Southwest of the Fred Harvey Company and the Santa Fe Railway. Marta Weigle and Barbara A. Babcock, editors of the volume, have done a superb job weaving in early Southwestern archaeological activities with the role of the Fred Harvey Company and the Santa Fe Railway in bring the American Southwest to those "east of the Mississippi River". Many early Southwestern archaeologists made their way throughout the Southwest on the Santa Fe Railway while the "outposts of civilization" that the Fred Harvey Company provided in many railroad stations served as a " bit of home" to the traveler. This book describes the collaboration of both Fred Harvey and the Santa Fe Railroad tourism in the American Southwest and provides an excellent look into the Native American artists and their comumnities which were transformed on a massive scale by the Fred Harvey Company as it bought, sold, and popularized Native American art. Also part of the volume is an excellent discussion of the network of major museums that hold art collections which were purchased through the Harvey Company's Indian Department.

Qualification of a Vitrified High Level Waste Product to Support Used Nuclear Fuel Recycling in the US

International Nuclear Information System (INIS)

Murray, P.; Bailly, F.; Strachan, D.; Senentz, G.; Veyer, C.

2009-01-01

As part of the Department of Energy (DOE) Global Nuclear Energy Partnership (GNEP), AREVA formed the International Nuclear Recycling Alliance (INRA) consisting of recognized world-leading companies in the area of used nuclear fuel (UNF) recycling,. The INRA team, consisting of AREVA, Mitsubishi Heavy Industries (MHI), Japan Nuclear Fuel Ltd (JNFL), Batelle Memorial Institute (BMI), URS Washington Division and Babcock and Wilcox (B and W), prepared a pre-conceptual design for an upgradable engineering-scale recycling plant with a nominal through put of 800 tHM/y. The pre-conceptual design of this leading-edge facility was based upon the extensive experience of the INRA team in recycling plant design and real world 'lessons learned' from actually building, commissioning, and operating recycling facilities in both France and Japan. The conceptual flowsheet, based upon the COEX TM separations process, separates the useful products for recycling into new fuel and sentences all the remaining fission products and minor actinides (MA) to the high level waste, (HLW) for vitrification. The proposed vitrified waste product will be similar to that currently produced in recycling plants in France. This wasteform has been qualified in France by conducting extensive studies and demonstrations. In the US, the qualification of vitrified glass products has been conducted by the US National Laboratories for the Defence Waste Processing Facility (DWPF), the West Valley Demonstration Plant (WVDP), and the Waste Treatment Plant (WTP). The vitrified waste product produced by recycling is sufficiently different from these current waste forms to warrant additional trials and studies. In this paper we review the differences in the vitrified waste forms previously qualified in the US with that produced from recycling of UNF in France. The lessons learned from qualifying a vitrified waste form in Europe is compared to the current US process for vitrified waste qualification including waste

NASA Trapezoidal Wing Simulation Using Stress-w and One- and Two-Equation Turbulence Models

Science.gov (United States)

Rodio, J. J.; Xiao, X; Hassan, H. A.; Rumsey, C. L.

2014-01-01

The Wilcox 2006 stress-omega model (also referred to as WilcoxRSM-w2006) has been implemented in the NASA Langley code CFL3D and used to study a variety of 2-D and 3-D configurations. It predicted a variety of basic cases reasonably well, including secondary flow in a supersonic rectangular duct. One- and two-equation turbulence models that employ the Boussinesq constitutive relation were unable to predict this secondary flow accurately because it is driven by normal turbulent stress differences. For the NASA trapezoidal wing at high angles of attack, the WilcoxRSM-w2006 model predicted lower maximum lift than experiment, similar to results of a two-equation model.

Investigation of a high pressure oxy-coal process

Energy Technology Data Exchange (ETDEWEB)

Renz, U. [RWTH Aachen Univ. (Germany). Inst. of Heat and Mass Transfer

2013-07-01

A study was conducted to investigate the feasibility of an oxy-coal process, which is pressurized to a combustion pressure of 80 bar. At that pressure the water-vapor can be separated economically from the CO{sub 2}/H{sub 2}O flue gases, either by nucleate condensation or by condensation on cooled surfaces in condenser heat exchangers at a temperature of about 300 C. The heat of condensation can be recaptured to preheat the boiler feed water. So the number of economizers is drastically reduced compared to a conventional steam cycle. Another interesting feature of the high pressure oxy-coal process is the fact, that low rank coal with high moisture content can be fired. Such a process at a pressure of about 80 bar is currently investigated by Babcock, USA, as the ThermoEnergy Integrated Power System (TIPS) and will be analyzed in the present paper. A known disadvantage of the oxy-coal processes is the large recirculating flue gas stream to control the combustion temperature, and which need large pipes and heavy recirculation fans. This disadvantage could be avoided if instead of flue gas a part of the condensed water from the condenser heat exchangers is recirculated. Within the present study both types of processes have been simulated and for an electric power output of about 220 MW. Furthermore, results of CFD simulations of a pressurized 250 MW combustor with a single swirl burner and flue gas recirculation will be presented.

NEW OPTICAL SENSOR SUITE FOR ULTRAHIGH TEMPERATURE FOSSIL FUEL APPLICATIONS

Energy Technology Data Exchange (ETDEWEB)

Russell G. May; Tony Peng; Tom Flynn

2004-12-01

Accomplishments during the Phase I of a program to develop and demonstrate technology for the instrumentation of advanced powerplants are described. Engineers from Prime Research, LC and Babcock and Wilcox Research Center collaborated to generate a list of potential applications for robust photonic sensors in existing and future boiler plants. From that list, three applications were identified as primary candidates for initial development and demonstration of high-temperature sensors in an ultrasupercritical power plant. A matrix of potential fiber optic sensor approaches was derived, and a data set of specifications for high-temperature optical fiber was produced. Several fiber optic sensor configurations, including interferometric (extrinsic and intrinsic Fabry-Perot interferometer), gratings (fiber Bragg gratings and long period gratings), and microbend sensors, were evaluated in the laboratory. In addition, progress was made in the development of materials and methods to apply high-temperature optical claddings to sapphire fibers, in order to improve their optical waveguiding properties so that they can be used in the design and fabrication of high-temperature sensors. Through refinements in the processing steps, the quality of the interface between core and cladding of the fibers was improved, which is expected to reduce scattering and attenuation in the fibers. Numerical aperture measurements of both clad and unclad sapphire fibers were obtained and used to estimate the reduction in mode volume afforded by the cladding. High-temperature sensors based on sapphire fibers were also investigated. The fabrication of an intrinsic Fabry-Perot cavity within sapphire fibers was attempted by the bulk diffusion of magnesium oxide into short localized segments of longer sapphire fibers. Fourier analysis of the fringes that resulted when the treated fiber was interrogated by a swept laser spectrometer suggested that an intrinsic cavity had been formed in the fiber. Also

Quality assurance program plan for SNF characterization support project

International Nuclear Information System (INIS)

Tanke, J.M.

1997-01-01

This Quality Assurance Program Plan (QAPP) provides information on how the Quality Assurance Program is implemented for the Spent Nuclear Fuel Characterization Support Project. This QAPP has been developed specifically for the Spent Nuclear Fuel Characterization Support Project, per Letter of Instruction (LOI) from Duke Engineering and Services Company, letter No. DESH-9655870, dated Nov. 22, 1996. It applies to those items and tasks which affect the completion of activities identified in the work breakdown structure of the Project Management Plan (PMP) and LOI. These activities include installation of sectioning equipment and furnace, surface and subsurface examinations, sectioning for metallography, and element drying and conditioning testing, as well as project related operations within the 327 facility as it relates to the specific activities of this project. General facility activities are covered in other appropriate QA-PPS. In addition, this QAPP supports the related quality assurance activities addressed in CM-2-14, Hazardous Material Packaging and Shipping,1261 and HSRCM-1, Hanford Site Radiological Control Manual. The 327 Building is currently transitioning from being a Pacific Northwest National Laboratory (PNNL) managed facility to a Babcock and Wilcox Hanford Company (BVMC) managed facility. During this transition process existing procedures and documents will be utilized until replaced by BVMC procedures and documents. These documents conform to the requirements found in PNL-MA-70, Quality Assurance Manual and PNL-MA-8 1, Hazardous Materials Shipping Manual. The Quality Assurance Program Index (QAPI) contained in Table 1 provides a matrix which shows how project activities relate to IO CFR 830.120 and 5700.6C criteria. Quality Assurance program requirements will be addressed separate from the requirements specified in this document. Other Hanford Site organizations/companies may be utilized in support of this project and the subject organizations are

Component Test Facility (Comtest) Phase 1 Engineering For 760°C (1400°F) Advanced Ultrasupercritical (A-USC) Steam Generator Development

Energy Technology Data Exchange (ETDEWEB)

Weitzel, Paul [Babcock & Wilcox Power Generation Group, Inc., Barberton, OH (United States)

2016-05-13

The Babcock & Wilcox Company (B&W) performed a Pre-Front End Engineering Design (Pre-FEED) of an A-USC steam superheater for a proposed component test program achieving 760°C (1400°F) steam temperature. This would lead to follow-on work in a Phase 2 and Phase 3 that would involve detail design, manufacturing, construction and operation of the ComTest. Phase 1 results have provided the engineering data necessary for proceeding to the next phase of ComTest. The steam generator superheater would subsequently supply the steam to an A-USC prototype intermediate pressure steam turbine. The ComTest program is important in that it will place functioning A-USC components in operation and in coordinated boiler and turbine service. It is also important to introduce the power plant operation and maintenance personnel to the level of skills required and provide the first background experience with hands-on training. The project will provide a means to exercise the complete supply chain events required in order to practice and perfect the process for A-USC power plant design, supply, manufacture, construction, commissioning, operation and maintenance. Representative participants will then be able to transfer knowledge and recommendations to the industry. ComTest is conceived in the manner of using a separate standalone plant facility that will not jeopardize the host facility or suffer from conflicting requirements in the host plant’s mission that could sacrifice the nickel alloy components and not achieve the testing goals. ComTest will utilize smaller quantities of the expensive materials and reduce the risk in the first operational practice for A-USC technology in the United States. Components at suitable scale in ComTest provide more assurance before putting them into practice in the full size A-USC demonstration plant.

Genetic diversity in 14 tomato (Lycopersicon esculentum Mill ...

African Journals Online (AJOL)

Administrator

2011-06-08

Jun 8, 2011 ... 1Department of Biosciences and Biotechnology, Babcock University, Ilishan Remo, Ogun State,PMB 21244, Ikeja,. Nigeria. 2Department of Agriculture and Industrial Technology, Babcock University, Ogun State, Nigeria. 3Department of Botany, University of Ibadan, Ibadan, Oyo State, Nigeria. Accepted 6 ...

Calculation of fuel pin failure timing under LOCA conditions

International Nuclear Information System (INIS)

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

1991-10-01

The objective of this research was 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) 4-loop design (Seabrook). Sensitivity studies were performed to assess the impacts of fuel pin burnup, axial peaking factor, break size, emergency core cooling system (ECCS) availability, and main coolant pump trip on these items. The analysis was performed using a four-code approach, comprised of FRAPCON-2, SCDAP/RELAP5/MOD3, TRAC-PF1/MOD1, and FRAP-T6. In addition to the calculation of timing results, this analysis provided a comparison of the capabilities of SCDAP/RELAP5/MOD3 with TRAC-PF1/MOD1 for large-break LOCA analysis. This paper discusses the methodology employed and the code development efforts required to implement the methodology. The shortest time intervals calculated between initiation of containment isolation and fuel pin failure were 11.4 s and 19.1 for the B ampersand W and W plants, respectively. The FRAP-T6 fuel pin failure times calculated using thermal-hydraulic data generated by SCDAP/RELAP5/MOD3 were more conservative than those calculated using data generated by TRAC-PF1/MOD1. 18 refs., 7 figs., 4 tabs

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

Microbiologically influenced corrosion in condenser water boxes at Crystal River-3

International Nuclear Information System (INIS)

Hayner, G.O.; Pope, D.H.; Crane, B.E.

1988-01-01

During the spring of 1986, several welds in the lower half of the condenser inlet water boxes at Crystal River-3 (CR-3) were found to be seeping seawater. The leakage produced red-brown and black-green colored deposits on the outside surface of the water boxes. The welds in affected areas were not uniformly attacked, and the severity of attack varied between water boxes; however, there were instances of attack on each type of pressure-retaining weld in the affected regions. Weld seepage was also seen on the outside of the inlet piping to the water boxes. A few small pin holes were seen in the base metal of the water boxes not associated with welds. In this paper the authors report the results of examinations performed at both the CR-3 site and at The Babcock and Wilcox Company Lynchburt Research Center (LRC). The inside of a water box and the exterior of the condenser inlet piping were visually inspected at the Cr-3 site. Nodules inside the water box were probed and examined. Parts of nodules were collected and microscopically examined for bacteria. Two corrosion-deposit samples removed from condenser instrument piping and the condenser inlet piping were chemically analyzed at the LRC. Four pipe samples removed from the condenser instrument piping were destructively examined at the LRC. This work included visual inspection, metallographic, scanning electron microscopy (SEM), and energy dispersive spectroscopy (EDS) examinations performed on selected locations of the piping samples

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

Once-through integral system (OTIS): Final report

International Nuclear Information System (INIS)

Gloudemans, J.R.

1986-09-01

A scaled experimental facility, designated the once-through integral system (OTIS), was used to acquire post-small break loss-of-coolant accident (SBLOCA) data for benchmarking system codes. OTIS was also used to investigate the application of the Abnormal Transient Operating Guidelines (ATOG) used in the Babcock and Wilcox (B and W) designed nuclear steam supply system (NSSS) during the course of an SBLOCA. OTIS was a single-loop facility with a plant to model power scale factor of 1686. OTIS maintained the key elevations, approximate component volumes, and loop flow resistances, and simulated the major component phenomena of a B and W raised-loop nuclear plant. A test matrix consisting of 15 tests divided into four categories was performed. The largest group contained 10 tests and was defined to parametrically obtain an extensive set of plant-typical experimental data for code benchmarking. Parameters such as leak size, leak location, and high-pressure injection (HPI) shut-off head were individually varied. The remaining categories were specified to study the impact of the ATOGs (2 tests), to note the effect of guard heater operation on observed phenomena (2 tests), and to provide a data set for comparison with previous test experience (1 test). A summary of the test results and a detailed discussion of Test 220100 is presented. Test 220100 was the nominal or reference test for the parametric studies. This test was performed with a scaled 10-cm 2 leak located in the cold leg suction piping

A comparison of floor response spectra techniques

International Nuclear Information System (INIS)

Yan, M.J.; Galford, J.E.

1983-01-01

Floor response spectra (FRS) conventionally have been generated using a time-history method. Babcock and Wilcox has developed a new technique, the Fast Floor Response Spectra (FFRS) method, in which dynamic analyses are done entirely in the frequency domain. This paper compares the two techniques and demonstrates that the FFRS method complies with the 'equivalency' and 'conservatism' requirements of the US NRC's Standard Review Plan. The upper end of a once-through steam generator in the B and W 205 nuclear steam supply system (NSSS) was used to demonstrate that the FFRS method is equivalent to the time-history technique. The two techniques were compared with respect to frequency content and magnitude of response for a given point on the structure. First, the specified forcing function was described in terms of an acceleration time history and an acceleration spectra enveloping that time history. The time-history forcing function was then used in a direct transient analysis to determine the response at the specified point on the NSSS. The resultant response was subsequently converted to a floor response spectra for that point. To show that the FFRS method gave equivalent and conservative results, the FFRS technique was used to determine the modal response directly from the spectral description of the forcing function. The FFRS- and time-history-generated data agreed to within 13 (worst case on conservative side) of each other with the former cutting analytical costs by 99%. (orig./HP)

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.)

Development of compaction technique for spent fuel skeletons

Energy Technology Data Exchange (ETDEWEB)

Yoon, Ji Sup; Kim, Young Hwan; Jung, Jae Hoo [Korea Atomic Energy Research Institute, Taejon (Korea)

1999-12-01

To increase the utilization of uranium resources contained in the spent fuel, the spent fuel is reused. For this, the spent fuel is dismantled or spent fuel rod is extracted from the spent fuel assembly. When the rod is extracted, the remaining components of spent fuel assembly, so called a NFBC(Non-Fuel Bearing Components), should be compacted for the final disposal. To this end, several companies developed the NFBC compactors. German company, named as GNS has developed the direct compression devices of the NFBCs for the rod consolidation and installed it at the PKA(2) of pilot conditioning plant. B and W (Babcock and Wilcox) in USA adopted cutting method rather than the compression method and developed the special cutting devices of NFBC which can be applied underwater environment. In this study the characteristics of these two methods was investigated, in terms of fabrication cost of devices, maintainability in a high radioactive environment, required power and work volume for operation. Also, the optimal power source is selected by comparing the maximum power versus the work volume for operation. In addition to these, the reduction ratio of the bulk volume is obtained while varying the cutting length of the NFBC through a series of experiments. Based on the results of analysis and experiments, the cutting method after compression is selected as an optimal volume reduction method and its design specification is obtained. 8 refs., 62 figs., 32 tabs. (Author)

Advanced non-linear flow-induced vibration and fretting-wear analysis capabilities

Energy Technology Data Exchange (ETDEWEB)

Toorani, M.; Pan, L.; Li, R.; Idvorian, N. [Babcock and Wilcox Canada Ltd., Cambridge, Ontario (Canada); Vincent, B.

2009-07-01

Fretting wear is a potentially significant degradation mechanism in nuclear steam generators and other shell and tube heat transfer equipment as well. This paper presents an overview of the recently developed code FIVDYNA which is used for the non-linear flow-induced vibration and fretting wear analysis for operating steam generators (OTSG and RSG) and shell-and-tube heat exchangers. FIVDYNA is a non-linear time-history Flow-Induced Vibration (FIV) analysis computer program that has been developed by Babcock and Wilcox Canada to advance the understanding of tube vibration and tube to tube-support interaction. In addition to the dynamic fluid induced forces the program takes into account other tube static forces due to axial and lateral tube preload and thermal interaction loads. The program is capable of predicting the location where the fretting wear is most likely to occur and its magnitude taking into account the support geometry including gaps. FIVDYNA uses the general purpose finite element computer code ABAQUS as its solver. Using ABAQUS gives the user the flexibility to add additional forces to the tube ranging from tube preloads and the support offsets to thermal loads. The forces currently being modeled in FIVDYNA are the random turbulence, steady drag force, fluid-elastic forces, support offset and pre-strain force (axial loads). This program models the vibration of tubes and calculates the structural dynamic characteristics, and interaction forces between the tube and the tube supports. These interaction forces are then used to calculate the work rate at the support and eventually the predicted depth of wear scar on the tube. A very good agreement is found with experiments and also other computer codes. (author)

Mechanistic insights into the hydrocyanation reaction

NARCIS (Netherlands)

Bini, L.

2009-01-01

The hydrocyanation of an alkene is a catalytic carbon-carbon bond formation reaction and the obtained nitriles can be converted into a variety of valuable products. The investigation of this reaction has mainly focused on the DuPont adiponitrile (AdN) process. This process is so far the only example

Field to fuel: developing sustainable biorefineries.

Science.gov (United States)

Jenkins, Robin; Alles, Carina

2011-06-01

Life-cycle assessment (LCA) can be used as a scientific decision support technique to quantify the environmental implications of various biorefinery process, feedstock, and integration options. The goal of DuPont's integrated corn biorefinery (ICBR) project, a cost-share project with the United States Department of Energy, was to demonstrate the feasibility of a cellulosic ethanol biorefinery concept. DuPont used LCA to guide research and development to the most sustainable cellulosic ethanol biorefinery design in its ICBR project and will continue to apply LCA in support of its ongoing effort with joint venture partners. Cellulosic ethanol is a biofuel which has the potential to provide a sustainable solution to the nation's growing concerns around energy supply and climate change. A successful biorefinery begins with sustainable removal of biomass from the field. Michigan State University (MSU) used LCA to estimate the environmental performance of corn grain, corn stover, and the corn cob portion of the stover, grown under various farming practices for several corn growing locations in the United States Corn Belt. In order to benchmark the future technology options for producing cellulosic ethanol with existing technologies, LCA results for fossil energy consumption and greenhouse gas (GHG) emissions are compared to alternative ethanol processes and conventional gasoline. Preliminary results show that the DuPont ICBR outperforms gasoline and other ethanol technologies in the life-cycle impact categories considered here.

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.

76 FR 14393 - Notice of Receipt of Requests for Amendments To Delete Uses in Certain Pesticide Registrations

Science.gov (United States)

2011-03-16

... Aldicarb Aldicarb Use on Coffee. Pesticide. 352-604 DuPont Tanos Famoxadone & Cymoxanil.. Leaf Petioles...: Manager, U.S.-- Registration, DuPont Crop Protection, 1007 Market Street, Wilmington, DE 19898-0001. 8536...

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)

76 FR 10584 - Notice of Receipt of Several Pesticide Petitions Filed for Residues of Pesticide Chemicals in or...

Science.gov (United States)

2011-02-25

... dimethylcyclopropanecarboxylate and its inactive R-isomers, in or on avocado; papaya; star apple; black sapote; mango; sapodilla... 0F7763. (EPA-HQ-OPP-2010-0888). E. I. DuPont de Nemours and Company, DuPont Crop Protection, 1700 Market...

Assessing aquifer storage and recovery feasibility in the Gulf Coastal Plains of Texas

Directory of Open Access Journals (Sweden)

W. Benjamin Smith

2017-12-01

Full Text Available Study region: The Gulf Coast and Carrizo-Wilcox aquifer systems in the Gulf Coastal Plains of Texas. Study focus: Aquifer storage and recovery is a water storage alternative that is underutilized in Texas, a state with both long periods of drought and high intensity storms. Future water storage plans in Texas almost exclusively rely on surface reservoirs, subject to high evaporative losses. This study seeks to identify sites where aquifer storage and recovery (ASR may be successful, especially in recovery of injected waters, by analyzing publicly-available hydrogeologic data. Transmissivity, hydraulic gradient, well density, depth to aquifer, and depth to groundwater are used in a GIS-based index to determine feasibility of implementing an ASR system in the Gulf Coast and Carrizo-Wilcox aquifer systems. New hydrological insights for the region: Large regions of the central and northern Gulf Coast and the central and southern Carrizo-Wilcox aquifer systems are expected to be hydrologically feasible regions for ASR. Corpus Christi, Victoria, San Antonio, Bryan, and College Station are identified as possible cities where ASR would be a useful water storage strategy. Keywords: Aquifer storage and recovery (ASR, GIS, Gulf coast, Carrizo-Wilcox, Managed aquifer recharge (MAR

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

Science.gov (United States)

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

2014-01-01

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

Dupont Analysis of An Information Technology Enabled Competitive Advantage

OpenAIRE

Nustini, Yuni

2009-01-01

The transformation of business caused by e-business and e-commerce applications of the internet and related technologies demonstrates that information systems and information technologies are essential ingredients for business survival and success. The most often cited benefit of IT are integrating business process, increasing efficiency, sales, productivity, and competitiveness. This study is designed for determining where managers considering investment in information technology (IT) projec...

Dupont Analysis of an Information Technology Enabled Competitive Advantage

OpenAIRE

Nustini, Yuni

2003-01-01

The transformation of business caused by e-business and e-commerce applications of the internet and related technologies demonstrates that information systems and information technologies are essential ingredients for business survival and success. The most often cited benefit of IT are integrating business process, increasing efficiency, sales, productivity, and competitiveness. This study is designed for determining where managers considering investment in information technology (IT) projec...

A Methodology for a Sustainable CO2 Capture and Utilization Network

DEFF Research Database (Denmark)

Frauzem, Rebecca; Fjellerup, Kasper; Gani, Rafiqul

2015-01-01

hydrogenation highlights the application. This case study illustrates the utility of the utilization network and elements of the methodology being developed. In addition, the conversion process is linked with carbon capture to evaluate the overall sustainability. Finally, the production of the other raw...... of Climate Change. New York: Cambridge University Press, 2007. [2] J. Wilcox, Carbon Capture. New York: Springer, 2012....

Achievement report for fiscal 1982 on Sunshine Program. Research and development of coal gasification (Feasibility study of fluidized gasification process by pressurized hydrogenation); 1982 nendo sekitan gas ka no kenkyu kaihatsu seika hokokusho. Kaatsu suiten ryudo gas ka process no feasibility study

Energy Technology Data Exchange (ETDEWEB)

NONE

1983-01-01

This report covers the results of a feasibility study conducted for the 'Research and development of high-calorie gas production technology (Fluidized gasification process by pressurized hydrogenation)' entrusted to the Mifuji Ironworks during the period 1974-1977 and then entrusted to the Babcock-Hitachi K.K. in the period 1978-present. The hydrogenation gasification process is characterized in that hydrogen serves as the gasification agent for the generation of a methane-rich high-calorie gas. It is preferred that the gasification temperature be relatively low and the pressure high for a hydrogenation gasification reaction. Under the Sunshine Program, development efforts are under way to produce a high-calorie gas, clean and its combustion easy to control, under a pressure of 30kg/cm{sup 2}G, more or less. In this pressurized gasification process, it is important to acquire high-concentration hydrogen for hydrogenation gasification simultaneously with the hydrogenation gasification reaction. In order to achieve the goal, a study is made of a new separated reaction type wet gasification furnace capable of generating high-concentration hydrogen efficiently and economically, and research is under way for the development of such a furnace. The results of element studies and of this feasibility study, and future tasks are compiled into this report. (NEDO)

Steam generator maintenance and life management at Embalse Nuclear Station

International Nuclear Information System (INIS)

Sainz, R.; Diaz, G.; Sveruga, H.; Ramakrishnan, T.K.; Azeez, S.

2004-01-01

The Embalse Nuclear Station has four steam generators (SGs) with inverted vertical U tubes manufactured by Babcock and Wilcox Canada (B and W). These are main components, both from the operative point of view as the heat transfer from the Primary Heat Transport System (PHTS) to the Secondary System, and from the point of view of safety, as they are the part of the PHTS and its radioactive inventory pressure barrier. In addition, they are one of the most important cost-related elements for potential life extensions. Maintenance and inspections are carried out in order to maintain a high availability of the SGs, as they have had a positive impact on the operational availability of the plant, and to reduce the tube failure probabilities, thus minimizing the amount of radioactive effluents and taking care of the condition of the main components in order to enable the plant life management and the planning of the plant life extension. The most relevant maintenance activities performed have been the inspections performed on 100% of the tubes every 3 years. the mechanical cleaning of the inside of the tubes, the sludge removal from the secondary side tubesheet, the divider plate replacement, and the inspection of internals of the secondary side.Thanks to the latter and to the eddy current inspections, the degradation in the U-bend supports was detected early and every effort is being made to repair them shortly. Besides, a life management program has been started covering the entire plant starting with this important component. The Embalse Nuclear Station's SGs show a low percentage of plugged tubes compared to other stations in similar conditions, but they must be monitored continually and systematically if a life extension is intended. (author)

Design and performance of BWC replacement steam generators for PWR systems

International Nuclear Information System (INIS)

Klarner, R.; Steinmoeller, F.; Millman, J.; Schneider, W.

1998-01-01

In recent years, Babcock and Wilcox Canada (BWC) has provided a number of PWR Replacement Steam Generators (RSGS) to replace units that had experienced extensive Alloy 600 tube degradation. BWC RSG units are in operation at Northeast Utilities' Millstone Unit 2, Rochester Gas and Electric's Ginna Station, Duke Energy's Catawba Unit 1, McGuire Unit 1 and 2, Florida Power and Light's St. Lucie Unit 1 and Commonwealth Edison's Byron 1 Station. Extensive start-up performance characteristics have been obtained for Millstone 2, Ginna, McGuire 1, and Catawba 1 RSGS. The Millstone 2, Ginna and Catawba 1 RSGs have also undergone extensive inspections following their first cycle of operation. The design and start-up performance characteristics of these RSGs are presented. The BWC Replacement Steam generators were designed to fit the existing envelope of pressure boundary dimensions to ensure licensability and integration into the Nuclear Steam Supply System. The RSGs were provided with a tube bundle of Alloy 690TT tubing, sized to match or exceed the original steam generator (OSG) thermal performance including provision for the reduced thermal conductivity of Alloy 690 relative to Alloy 600. The RSG tube bundle configurations provide a higher circulation design relative to the OSG, and feature corrosion resistant lattice grid and U-bend tube supports which provide effective anti-vibration support. The tube bundle supports accommodate relatively unobstructed flow and allow unrestrained structural interactions during thermal transients. Efficient steam separators assure low moisture carryover as well as high circulation. Performance measurements obtained during start-up verify that the BWC RSGs meet or exceed the specified thermal and moisture carryover performance requirements. RSG water level stability results at nor-mal operation and during plant transients have been excellent. Visual and ECT inspections have confirmed minimal deposition and 100% tube integrity following

Factorial Analysis of Profitability

OpenAIRE

Georgeta VINTILA; Ilie GHEORGHE; Ioana Mihaela POCAN; Madalina Gabriela ANGHEL

2012-01-01

The DuPont analysis system is based on decomposing the profitability ratio in factors of influence. This paper describes the factorial analysis of profitability based on the DuPont system. Significant importance is given to the impact on various indicators on the shares value and profitability.

Low NO{sub x} burner modifications to front-fired pulverized coal boilers

Energy Technology Data Exchange (ETDEWEB)

Broderick, R G; Wagner, M

1998-07-01

Madison Gas and Electric Blount Street Station Units 8 and 9 are Babcock and Wilcox pulverized coal fired and natural gas fired boilers. These boilers were build in the late 1950's and early 1960's with each boiler rated at 425,000 lb./hr of steam producing 50 MW of electricity. The boilers are rated at 9,500 F at 1,350 psig. Each unit is equipped with one Ljungstroem air heater and two B and W EL pulverizers. These units burn subbituminous coal with higher heating value of 10,950 Btu/LB on an as-received basis. The nitrogen content is approximately 1.23% with 15% moisture. In order to comply with the new Clean Air Act Madison Gas and Electric needs to reduce NO{sub x} on these units to less than .5 LB/mmBtu. Baseline NO{sub x} emissions on these units range between .8--.9 lb./mmBtu. LOIs average approximately 8%. Madison Gas and Electric contracted with RJM Corporation to modify the existing burners to achieve this objective. These modifications consisted of adding patented circumferentially and radially staged flame stabilizers, modifying the coal pipe, and replacing the coal impeller with a circumferentially staged coal spreader. RJM Corporation utilized computational fluid dynamics modeling in order to design the equipment to modify these burners. The equipment was installed during the March 1997 outage and start-up and optimization was conducted in April 1997. Final performance results and economic data will be included in the final paper.

Monte Carlo code criticality benchmark comparisons for waste packaging

International Nuclear Information System (INIS)

Alesso, H.P.; Annese, C.E.; Buck, R.M.; Pearson, J.S.; Lloyd, W.R.

1992-07-01

COG is a new point-wise Monte Carlo code being developed and tested at Lawrence Livermore National Laboratory (LLNL). It solves the Boltzmann equation for the transport of neutrons and photons. The objective of this paper is to report on COG results for criticality benchmark experiments both on a Cray mainframe and on a HP 9000 workstation. COG has been recently ported to workstations to improve its accessibility to a wider community of users. COG has some similarities to a number of other computer codes used in the shielding and criticality community. The recently introduced high performance reduced instruction set (RISC) UNIX workstations provide computational power that approach mainframes at a fraction of the cost. A version of COG is currently being developed for the Hewlett Packard 9000/730 computer with a UNIX operating system. Subsequent porting operations will move COG to SUN, DEC, and IBM workstations. In addition, a CAD system for preparation of the geometry input for COG is being developed. In July 1977, Babcock ampersand Wilcox Co. (B ampersand W) was awarded a contract to conduct a series of critical experiments that simulated close-packed storage of LWR-type fuel. These experiments provided data for benchmarking and validating calculational methods used in predicting K-effective of nuclear fuel storage in close-packed, neutron poisoned arrays. Low enriched UO2 fuel pins in water-moderated lattices in fuel storage represent a challenging criticality calculation for Monte Carlo codes particularly when the fuel pins extend out of the water. COG and KENO calculational results of these criticality benchmark experiments are presented

Retrospective Geospatial Modeling of PM10 Exposures from Open Burning at Joint Base Balad, Iraq

Science.gov (United States)

2011-03-01

Gullett & Linak, 2007), and complex munitions (Wilcox, Entezam, Molenaar , & Shreeve, 1996). Documentation on emissions from the open burning of typical...of a Petrochemical Refinery: a Cross- sectional Study. Environmental Health , 8 (45). 194 Wilcox, J., Entezam, B., Molenaar , B., & Shreeve, T

Organic geochemistry and petrology of subsurface Paleocene-Eocene Wilcox and Claiborne Group coal beds, Zavala County, Maverick Basin, Texas, USA

Science.gov (United States)

Hackley, Paul C.; Warwick, Peter D.; Hook, Robert W.; Alimi, Hossein; Mastalerz, Maria; Swanson, Sharon M.

2012-01-01

Coal samples from a coalbed methane exploration well in northern Zavala County, Maverick Basin, Texas, were characterized through an integrated analytical program. The well was drilled in February, 2006 and shut in after coal core desorption indicated negligible gas content. Cuttings samples from two levels in the Eocene Claiborne Group were evaluated by way of petrographic techniques and Rock–Eval pyrolysis. Core samples from the Paleocene–Eocene Indio Formation (Wilcox Group) were characterized via proximate–ultimate analysis in addition to petrography and pyrolysis. Two Indio Formation coal samples were selected for detailed evaluation via gas chromatography, and Fourier transform infrared (FTIR) and 13C CPMAS NMR spectroscopy. Samples are subbituminous rank as determined from multiple thermal maturity parameters. Elevated rank (relative to similar age coal beds elsewhere in the Gulf Coast Basin) in the study area is interpreted to be a result of stratigraphic and/or structural thickening related to Laramide compression and construction of the Sierra Madre Oriental to the southwest. Vitrinite reflectance data, along with extant data, suggest the presence of an erosional unconformity or change in regional heat flow between the Cretaceous and Tertiary sections and erosion of up to >5 km over the Cretaceous. The presence of liptinite-rich coals in the Claiborne at the well site may indicate moderately persistent or recurring coal-forming paleoenvironments, interpreted as perennially submerged peat in shallow ephemeral lakes with herbaceous and/or flotant vegetation. However, significant continuity of individual Eocene coal beds in the subsurface is not suggested. Indio Formation coal samples contain abundant telovitrinite interpreted to be preserved from arborescent, above-ground woody vegetation that developed during the middle portion of mire development in forested swamps. Other petrographic criteria suggest enhanced biological, chemical and physical

NEW SOLID FUELS FROM COAL AND BIOMASS WASTE; FINAL

International Nuclear Information System (INIS)

Hamid Farzan

2001-01-01

Under DOE sponsorship, McDermott Technology, Inc. (MTI), Babcock and Wilcox Company (B and W), and Minergy Corporation developed and evaluated a sludge derived fuel (SDF) made from sewage sludge. Our approach is to dry and agglomerate the sludge, combine it with a fluxing agent, if necessary, and co-fire the resulting fuel with coal in a cyclone boiler to recover the energy and to vitrify mineral matter into a non-leachable product. This product can then be used in the construction industry. A literature search showed that there is significant variability of the sludge fuel properties from a given wastewater plant (seasonal and/or day-to-day changes) or from different wastewater plants. A large sewage sludge sample (30 tons) from a municipal wastewater treatment facility was collected, dried, pelletized and successfully co-fired with coal in a cyclone-equipped pilot. Several sludge particle size distributions were tested. Finer sludge particle size distributions, similar to the standard B and W size distribution for sub-bituminous coal, showed the best combustion and slagging performance. Up to 74.6% and 78.9% sludge was successfully co-fired with pulverized coal and with natural gas, respectively. An economic evaluation on a 25-MW power plant showed the viability of co-firing the optimum SDF in a power generation application. The return on equity was 22 to 31%, adequate to attract investors and allow a full-scale project to proceed. Additional market research and engineering will be required to verify the economic assumptions. Areas to focus on are: plant detail design and detail capital cost estimates, market research into possible project locations, sludge availability at the proposed project locations, market research into electric energy sales and renewable energy sales opportunities at the proposed project location. As a result of this program, wastes that are currently not being used and considered an environmental problem will be processed into a renewable

Medline Plus

Full Text Available ... Retaining (CR) Knee System (Wilcox Medical Center, Lihue, HI, 3/06/2014) Knee Replacement for Women with ... Retaining (CR) Knee System (Wilcox Medical Center, Lihue, HI, 3/06/2014) Knee Replacement for Women with ...

Biogenic origin of coalbed gas in the northern Gulf of Mexico Coastal Plain, U.S.A.

Energy Technology Data Exchange (ETDEWEB)

Warwick, Peter D.; Hackley, Paul C. [U.S. Geological Survey, 956 National Center, Reston, VA 20192 (United States); Breland, F. Clayton Jr. [Louisiana Department of Natural Resources, 617 North 3rd Street, Baton Rouge, LA 70802 (United States)

2008-10-02

New coal-gas exploration and production in northern Louisiana and south-central Mississippi, Gulf of Mexico Basin, is focused on the Wilcox Group (Paleocene-Eocene), where the depth to targeted subbituminous C to high volatile C bituminous coal beds ranges from 300 to 1680 m, and individual coal beds have a maximum thickness of about 6 m. Total gas content (generally excluding residual gas) of the coal beds ranges from less than 0.37 cm{sup 3}/g (as-analyzed or raw basis; 1.2 cm{sup 3}/g, dry, ash free basis, daf) at depths less than 400 m, to greater than 7.3 cm{sup 3}/g (as-analyzed basis; 8.76 cm{sup 3}/g, daf) in deeper (> 1,500 m) parts of the basin. About 20 Wilcox coal-gas wells in northern Louisiana produce from 200 to 6485 m{sup 3} of gas/day and cumulative gas production from these wells is approximately 25 million m{sup 3} (as of December, 2006). U.S. Geological Survey assessment of undiscovered, technically recoverable gas resources in the Gulf of Mexico Coastal Plain, including northern and south-central Mississippi, indicates that coal beds of the Wilcox Group contain an estimated mean total 109.3 million m{sup 3} (3.86 trillion ft{sup 3}) of producible natural gas. To determine the origin of the Wilcox Group coal gases in northern Louisiana, samples of gas, water, and oil were collected from Wilcox coal and sandstone reservoirs and from under- and overlying Late Cretaceous and Eocene carbonate and sandstone reservoirs. Isotopic data from Wilcox coal-gas samples have an average {delta}{sup 13}C{sub CH4} value of - 62.6 permille VPDB (relative to Vienna Peedee Belemnite) and an average {delta}D{sub CH4} value of - 199.9 permille VSMOW (relative to Vienna Standard Mean Ocean Water). Values of {delta}{sup 13}C{sub CO2} range from - 25.4 to 3.42 permille VPDB. Produced Wilcox saline water collected from oil, conventional gas, and coalbed gas wells have {delta}D{sub H2O} values that range from - 27.3 to - 18.0 permille VSMOW. These data suggest that the

Validation of the detection of STEC (O26, O45, O103, O111, O121, O145 and O157 in raw burgers using real-time PCR (BAX® System Q7, DuPont in “WET POOLS”

Directory of Open Access Journals (Sweden)

Paula Mussio

2014-12-01

Full Text Available Shiga toxin-producing Escherichia coli (STEC O157:H7 and no-O157:H7 have been identified as emerging foodborne pathogens responsible for an increasing number of outbreaks worldwide. Many foods have been associated to these outbreaks, mainly undercooked beef burgers. Due to the increasing production and consumption of burgers in our country, it is important to have a rapid technique to identify and isolate the most important STEC strains in foods matrixes.The objective of these investigation was to assess the sensitivity, specificity and limit detection for the screening of the most prevalent STEC in frozen raw beef burgers, using the "STEC Screening stx/eae " kit to detect the stx/eae genes and the "Panel 1 STEC E. coli O26, O111, O121" and "Panel 2 STEC E. coli O45, O103, O145" and "E. coli O157: H7 MP” (DuPont kits for the detection of the different serogroups. The use of composed samples (wet pools, the recovery of each strain from the positives samples on selective media after specified inmunoconcentration and the detection of stx/eae virulence genes in isolates by PCR were also evaluated. We validated a technique that allows the detection of the mentioned STEC strains with limits of detection between 1-5 CFU in 65 grams of raw frozen hamburgers, using wet pools. 

Coupled hydrology and biogeochemistry of Paleocene–Eocene coal beds, northern Gulf of Mexico

Science.gov (United States)

McIntosh, Jennifer C.; Warwick, Peter D.; Martini, Anna M.; Osborn, Stephen G.

2010-01-01

Thirty-six formation waters, gas, and microbial samples were collected and analyzed from natural gas and oil wells producing from the Paleocene to Eocene Wilcox Group coal beds and adjacent sandstones in north-central Louisiana, USA, to investigate the role hydrology plays on the generation and distribution of microbial methane. Major ion chemistry and Cl−Br relations of Wilcox Group formation waters suggest mixing of freshwater with halite-derived brines. High alkalinities (up to 47.8 meq/L), no detectable SO4, and elevated δ13C values of dissolved inorganic carbon (up to 20.5‰ Vienna Peedee belemnite [VPDB]) and CO2 (up to 17.67‰ VPDB) in the Wilcox Group coals and adjacent sandstones indicate the dominance of microbial methanogenesis. The δ13C and δD values of CH4, and carbon isotope fractionation of CO2 and CH4, suggest CO2 reduction is the major methanogenic pathway. Geochemical indicators for methanogenesis drop off significantly at chloride concentrations above ∼1.7 mol/L, suggesting that high salinities inhibit microbial activity at depths greater than ∼1.6 km. Formation waters in the Wilcox Group contain up to 1.6% modern carbon (A14C) to at least 1690 m depth; the covariance of δD values of co-produced H2O and CH4 indicate that the microbial methane was generated in situ with these Late Pleistocene or younger waters. The most enriched carbon isotope values for dissolved inorganic carbon (DIC) and CO2, and highest alkalinities, were detected in Wilcox Group sandstone reservoirs that were CO2 flooded in the 1980s for enhanced oil recovery, leading to the intriguing hypothesis that CO2 sequestration may actually enhance methanogenesis in organic-rich formations.

76 FR 69746 - Eunice Kennedy Shriver National Institute of Child Health & Human Development; Notice of Closed...

Science.gov (United States)

2011-11-09

...) of the Federal Advisory Committee Act, as amended (5 U.S.C. App.), notice is hereby given of the..., 2011. Time: 8 a.m. to 5 p.m. Agenda: To review and evaluate grant applications. Place: The Dupont Hotel.... Agenda: To review and evaluate grant applications. Place: The Dupont Hotel, 1500 New Hampshire Avenue NW...

Application of photopolymer holograms for the anticounterfeit protection of banknotes

Science.gov (United States)

Markova, Nina V.; Yamnikov, Leonid S.; Turkina, Elena S.; Semenov, Erick G.; Bulygin, Theodor V.; Levin, Gennady G.

1998-04-01

The article analyses holograms counterfeiting techniques and presents grounds for the choice of the optimum type of holograms best protected against counterfeiting -- volume holograms. There is a description of these holograms registration process on DuPont OmniDex photopolymer film and the analysis of the perspective of their use during banknotes production.

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

DEFF Research Database (Denmark)

Budd, Chris; Gravesen, Jens; Willson, Eddie

1998-01-01

A standard TV image is transmitted as a series of horizontal lines. To reduce band-width effects in transmission, half of a picture is transmitted in each frame, i.e., information is only given about pictures on alternate lines, a process called interlacing. A difficulty with this process is that...... is that other images (for example computer images) do not have alternate lines omitted. Thus it is desirable to be able to interpolate an existing TV images to obtain information on the images between alternate lines; this is the de-interlacing process.......A standard TV image is transmitted as a series of horizontal lines. To reduce band-width effects in transmission, half of a picture is transmitted in each frame, i.e., information is only given about pictures on alternate lines, a process called interlacing. A difficulty with this process...

75 FR 29656 - Amendment of Class E Airspace; Mountain View, AR

Science.gov (United States)

2010-05-27

...-1181; Airspace Docket No. 09-ASW-36] Amendment of Class E Airspace; Mountain View, AR AGENCY: Federal... Mountain View, AR. Decommissioning of the Wilcox non-directional beacon (NDB) at Mountain View Wilcox Memorial Field Airport, Mountain View, AR, has made this action necessary to enhance the safety and...

75 FR 12163 - Class E Airspace; Mountain View, AR

Science.gov (United States)

2010-03-15

...-1181; Airspace Docket No. 09-ASW-36] Class E Airspace; Mountain View, AR AGENCY: Federal Aviation... Class E airspace at Mountain View, AR. Decommissioning of the Wilcox non-directional beacon (NDB) at... View, AR. Airspace reconfiguration is necessary due to the decommissioning of the Wilcox NDB and the...

Enhancing Carbon Reactivity in Mercury Control in Lignite-Fired Systems

Energy Technology Data Exchange (ETDEWEB)

Chad Wocken; Michael Holmes; John Pavlish; Jeffrey Thompson; Katie Brandt; Brandon Pavlish; Dennis Laudal; Kevin Galbreath; Michelle Olderbak

2008-06-30

This project was awarded through the U.S. Department of Energy (DOE) National Energy Technology Laboratory Program Solicitation DE-PS26-03NT41718-01. The Energy & Environmental Research Center (EERC) led a consortium-based effort to resolve mercury (Hg) control issues facing the lignite industry. The EERC team-the Electric Power Research Institute (EPRI); the URS Corporation; the Babcock & Wilcox Company; ADA-ES; Apogee; Basin Electric Power Cooperative; Otter Tail Power Company; Great River Energy; Texas Utilities; Montana-Dakota Utilities Co.; Minnkota Power Cooperative, Inc.; BNI Coal Ltd.; Dakota Westmoreland Corporation; the North American Coal Corporation; SaskPower; and the North Dakota Industrial Commission-demonstrated technologies that substantially enhanced the effectiveness of carbon sorbents to remove Hg from western fuel combustion gases and achieve a high level ({ge} 55% Hg removal) of cost-effective control. The results of this effort are applicable to virtually all utilities burning lignite and subbituminous coals in the United States and Canada. The enhancement processes were previously proven in pilot-scale and limited full-scale tests. Additional optimization testing continues on these enhancements. These four units included three lignite-fired units: Leland Olds Station Unit 1 (LOS1) and Stanton Station Unit 10 (SS10) near Stanton and Antelope Valley Station Unit 1 (AVS1) near Beulah and a subbituminous Powder River Basin (PRB)-fired unit: Stanton Station Unit 1 (SS1). This project was one of three conducted by the consortium under the DOE mercury program to systematically test Hg control technologies available for utilities burning lignite. The overall objective of the three projects was to field-test and verify options that may be applied cost-effectively by the lignite industry to reduce Hg emissions. The EERC, URS, and other team members tested sorbent injection technologies for plants equipped with electrostatic precipitators (ESPs) and

Scaling up ecohydrological processes: role of surface water flow in water-limited landscapes

CSIR Research Space (South Africa)

Popp, A

2009-11-01

Full Text Available –30% of global drylands [Winand Staring Centre, 1991; 71Reid, 2005; Zika and Erb, 2009] and 30% of drylands in JOURNAL OF GEOPHYSICAL RESEARCH, VOL. 114, XXXXXX, doi:10.1029/2008JG000910, 2009 Click Here for Full Article 1Potsdam Institute for Climate.... Copyright 2009 by the American Geophysical Union. 0148-0227/09/2008JG000910$09.00 XXXXXX 1 of 10 72 Sub-Sahara Africa [Zika and Erb, 2009] are already 73 degraded. Maintaining so-called ‘resource conserving’ dry 74 lands [Wilcox et al., 2003] will have...

Translating DWPF design criteria into an engineered facility design

International Nuclear Information System (INIS)

Kemp, J.B.

1986-01-01

The Defense Waste Processing Facility (DWPF) takes radioactive defense waste sludge and the radioactive nuclides, cesium and strontium, from the salt solution, and incorporates them in borosilicate glass in stainless steel canisters, for subsequent disposal in a deep geologic repository. The facility was designed by Bechtel National, Inc. under a subcontract from E.I. DuPont de Nemurs and Co., the prime contractor for the Department of Energy, for the design, construction and commissioning of the plant. The design criteria were specified by the DuPont Company, based upon their extensive experience as designer, and operator since the early 1950's, of the existing Savannah River Plant facilities. Some of the design criteria imposed unusual or new requirements on the detailed design of the facilities. This paper describes some of these criteria, encompassing several engineering disciplines, and discusses the solutions and designs which were developed for the DWPF

Biogenic origin of coalbed gas in the northern Gulf of Mexico Coastal Plain, U.S.A

Science.gov (United States)

Warwick, Peter D.; Breland, F. Clayton; Hackley, Paul C.

2008-01-01

New coal-gas exploration and production in northern Louisiana and south-central Mississippi, Gulf of Mexico Basin, is focused on the Wilcox Group (Paleocene–Eocene), where the depth to targeted subbituminous C to high volatile C bituminous coal beds ranges from 300 to 1680 m, and individual coal beds have a maximum thickness of about 6 m. Total gas content (generally excluding residual gas) of the coal beds ranges from less than 0.37 cm3/g (as-analyzed or raw basis; 1.2 cm3/g, dry, ash free basis, daf) at depths less than 400 m, to greater than 7.3 cm3/g (as-analyzed basis; 8.76 cm3/g, daf) in deeper (> 1,500 m) parts of the basin. About 20 Wilcox coal-gas wells in northern Louisiana produce from 200 to 6485 m3 of gas/day and cumulative gas production from these wells is approximately 25 million m3 (as of December, 2006). U.S. Geological Survey assessment of undiscovered, technically recoverable gas resources in the Gulf of Mexico Coastal Plain, including northern and south-central Mississippi, indicates that coal beds of the Wilcox Group contain an estimated mean total 109.3 million m3 (3.86 trillion ft3) of producible natural gas.To determine the origin of the Wilcox Group coal gases in northern Louisiana, samples of gas, water, and oil were collected from Wilcox coal and sandstone reservoirs and from under- and overlying Late Cretaceous and Eocene carbonate and sandstone reservoirs. Isotopic data from Wilcox coal-gas samples have an average δ13CCH4 value of − 62.6‰ VPDB (relative to Vienna Peedee Belemnite) and an average δDCH4 value of − 199.9‰ VSMOW (relative to Vienna Standard Mean Ocean Water). Values of δ13CCO2 range from − 25.4 to 3.42‰ VPDB. Produced Wilcox saline water collected from oil, conventional gas, and coalbed gas wells have δDH2O values that range from − 27.3 to − 18.0‰ VSMOW. These data suggest that the coal gases primarily are generated in saline formation water by bacterial reduction of CO2

Meniscus effect in microgravity materials processing

Science.gov (United States)

1998-01-01

While the microgravity environment of orbit eliminates a number of effects that impede the formation of materials on Earth, the change can also cause new, unwanted effects. A mysterious phenomenon, known as detached solidification, apparently stems from a small hydrostatic force that turns out to be pervasive. The contact of the solid with the ampoule transfers stress to the growing crystal and causing unwanted dislocations and twins. William Wilcox and Liya Regel of Clarkson University theorize that the melt is in contact with the ampoule wall, while the solid is not, and the melt and solid are cornected by a meniscus. Their work is sponsored by NASA's Office of Biological and Physical Researcxh, and builds on earlier work by Dr. David Larson of the State University of New York at Stony Brook.

Evaluation of a high-temperature burner-duct-recuperator system

Science.gov (United States)

1990-07-01

The U.S. Department of Energy's (DOE) Office of Industrial Technologies (OIT) sponsors research and development (R and D) to improve the energy efficiency of American industry and to provide for fuel flexibility. OIT has funded a multiyear R and D project by the Babcock and Wilcox Company (B and W) to design, fabricate, field test, and evaluate a high-temperature burner-duct-recuperator (HTBDR) system. This ceramic-based recuperator system recovers waste heat from the corrosive, high-temperature (2170 F) flue gas stream of a steel soaking pit to preheat combustion air to as high as 1700 F. The preheated air is supplied to a high-temperature burner. The B and W R and D program, which is now complete, involved several activities, including selecting and evaluating ceramic materials, designing the system, and developing and evaluating the prototype. In addition, a full-scale unit was tested at a B and W steel soaking pit. The full-scale system consisted of a modular single-stage ceramic recuperator, a conventional two-pass metallic recuperator, a high-temperature burner, fans, insulated ducting, and associated controls and instrumentation. The metallic recuperator preheated combustion air to about 750 F before it passed to the ceramic module. This technical case study describes the DOE/B and W recuperator project and highlights the field tests of the full-scale recuperator system. The document makes results of field tests and data analysis available to other researchers and private industry. It discusses project status, summarizes field tests, and reviews the potential effects the technology will have on energy use and system economics.

Large-Scale Mercury Control Technology Testing for Lignite-Fired Utilities - Oxidation Systems for Wet FGD

Energy Technology Data Exchange (ETDEWEB)

Steven A. Benson; Michael J. Holmes; Donald P. McCollor; Jill M. Mackenzie; Charlene R. Crocker; Lingbu Kong; Kevin C. Galbreath

2007-03-31

Mercury (Hg) control technologies were evaluated at Minnkota Power Cooperative's Milton R. Young (MRY) Station Unit 2, a 450-MW lignite-fired cyclone unit near Center, North Dakota, and TXU Energy's Monticello Steam Electric Station (MoSES) Unit 3, a 793-MW lignite--Powder River Basin (PRB) subbituminous coal-fired unit near Mt. Pleasant, Texas. A cold-side electrostatic precipitator (ESP) and wet flue gas desulfurization (FGD) scrubber are used at MRY and MoSES for controlling particulate and sulfur dioxide (SO{sub 2}) emissions, respectively. Several approaches for significantly and cost-effectively oxidizing elemental mercury (Hg{sup 0}) in lignite combustion flue gases, followed by capture in an ESP and/or FGD scrubber were evaluated. The project team involved in performing the technical aspects of the project included Babcock & Wilcox, the Energy & Environmental Research Center (EERC), the Electric Power Research Institute, and URS Corporation. Calcium bromide (CaBr{sub 2}), calcium chloride (CaCl{sub 2}), magnesium chloride (MgCl{sub 2}), and a proprietary sorbent enhancement additive (SEA), hereafter referred to as SEA2, were added to the lignite feeds to enhance Hg capture in the ESP and/or wet FGD. In addition, powdered activated carbon (PAC) was injected upstream of the ESP at MRY Unit 2. The work involved establishing Hg concentrations and removal rates across existing ESP and FGD units, determining costs associated with a given Hg removal efficiency, quantifying the balance-of-plant impacts of the control technologies, and facilitating technology commercialization. The primary project goal was to achieve ESP-FGD Hg removal efficiencies of {ge}55% at MRY and MoSES for about a month.

Knowledge, Attitude, and Healthcare-Seeking Behavior Towards ...

African Journals Online (AJOL)

Dysmenorrhea among Female Students of a Private University in ... Maternal and Child Health, School of Nursing, Babcock University, Ilishan Remo, Nigeria ..... affect perception of pain among individuals. .... Appropriate counseling and.

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

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).

Fertility and pregnancy: an epidemiologic perspective

National Research Council Canada - National Science Library

Wilcox, Allen J

2010-01-01

.... Weaving together history, biology, obstetrics, pediatrics, demography, infectious diseases, molecular genetics, and evolutionary biology, Allen Wilcox brings a fresh coherence to the epidemiologic...

The Effects of Mass Media Advertising on U.S. Army Recruiting.

Science.gov (United States)

1982-03-16

the analysis of the product, markets desired to be entered, mo- tives of the advertiser , messages to be presented, media to r -------- 12 be used to... market its products. Sales is a major DuPont advertising objective, but not the only objective. DuPont used selected panels of individuals to expose to its...center, objectives could be formed and operationalized for each advertising program. Overall re- search strategy could be planned, and specific

Latest design of gate valves

Energy Technology Data Exchange (ETDEWEB)

Kurzhofer, U.; Stolte, J.; Weyand, M.

1996-12-01

Babcock Sempell, one of the most important valve manufacturers in Europe, has delivered valves for the nuclear power industry since the beginning of the peaceful application of nuclear power in the 1960s. The latest innovation by Babcock Sempell is a gate valve that meets all recent technical requirements of the nuclear power technology. At the moment in the United States, Germany, Sweden, and many other countries, motor-operated gate and globe valves are judged very critically. Besides the absolute control of the so-called {open_quotes}trip failure,{close_quotes} the integrity of all valve parts submitted to operational forces must be maintained. In case of failure of the limit and torque switches, all valve designs have been tested with respect to the quality of guidance of the gate. The guidances (i.e., guides) shall avoid a tilting of the gate during the closing procedure. The gate valve newly designed by Babcock Sempell fulfills all these characteristic criteria. In addition, the valve has cobalt-free seat hardfacing, the suitability of which has been proven by friction tests as well as full-scale blowdown tests at the GAP of Siemens in Karlstein, West Germany. Babcock Sempell was to deliver more than 30 gate valves of this type for 5 Swedish nuclear power stations by autumn 1995. In the presentation, the author will report on the testing performed, qualifications, and sizing criteria which led to the new technical design.

Design and performance of BWC replacement steam generators for PWR systems

Energy Technology Data Exchange (ETDEWEB)

Klarner, R.; Steinmoeller, F.; Millman, J.; Schneider, W. [Babcock and Wilcox Canada, Cambridge, Ontario (Canada)

1998-07-01

In recent years, Babcock and Wilcox Canada (BWC) has provided a number of PWR Replacement Steam Generators (RSGS) to replace units that had experienced extensive Alloy 600 tube degradation. BWC RSG units are in operation at Northeast Utilities' Millstone Unit 2, Rochester Gas and Electric's Ginna Station, Duke Energy's Catawba Unit 1, McGuire Unit 1 and 2, Florida Power and Light's St. Lucie Unit 1 and Commonwealth Edison's Byron 1 Station. Extensive start-up performance characteristics have been obtained for Millstone 2, Ginna, McGuire 1, and Catawba 1 RSGS. The Millstone 2, Ginna and Catawba 1 RSGs have also undergone extensive inspections following their first cycle of operation. The design and start-up performance characteristics of these RSGs are presented. The BWC Replacement Steam generators were designed to fit the existing envelope of pressure boundary dimensions to ensure licensability and integration into the Nuclear Steam Supply System. The RSGs were provided with a tube bundle of Alloy 690TT tubing, sized to match or exceed the original steam generator (OSG) thermal performance including provision for the reduced thermal conductivity of Alloy 690 relative to Alloy 600. The RSG tube bundle configurations provide a higher circulation design relative to the OSG, and feature corrosion resistant lattice grid and U-bend tube supports which provide effective anti-vibration support. The tube bundle supports accommodate relatively unobstructed flow and allow unrestrained structural interactions during thermal transients. Efficient steam separators assure low moisture carryover as well as high circulation. Performance measurements obtained during start-up verify that the BWC RSGs meet or exceed the specified thermal and moisture carryover performance requirements. RSG water level stability results at nor-mal operation and during plant transients have been excellent. Visual and ECT inspections have confirmed minimal deposition and 100

South Louisiana Enhanced Oil Recovery/Sequestration R&D Project Small Scale Field Tests of Geologic Reservoir Classes for Geologic Storage

Energy Technology Data Exchange (ETDEWEB)

Hite, Roger [Blackhorse Energy LLC, Houston, TX (United States)

2016-10-01

The project site is located in Livingston Parish, Louisiana, approximately 26 miles due east of Baton Rouge. This project proposed to evaluate an early Eocene-aged Wilcox oil reservoir for permanent storage of CO₂. Blackhorse Energy, LLC planned to conduct a parallel CO₂ oil recovery project in the First Wilcox Sand. The primary focus of this project was to examine and prove the suitability of South Louisiana geologic formations for large-scale geologic sequestration of CO₂ in association with enhanced oil recovery applications. This was to be accomplished through the focused demonstration of small-scale, permanent storage of CO₂ in the First Wilcox Sand. The project was terminated at the request of Blackhorse Energy LLC on October 22, 2014.

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.)

Super-compactor and grouting. Efficient and safe treatment of nuclear waste

International Nuclear Information System (INIS)

Li, Hongyou; Starke, Holger; Muetzel, Wolfgang; Winter, Marc

2014-01-01

The conditioning and volume reduction of nuclear waste are increasingly important factors throughout the world. Efficient and safe treatment of nuclear waste therefore plays a decisive role. Babcock Noell designed, manufactured and supplied a complete waste treatment facility for conditioning of the solid radioactive waste of a nuclear power plant to China. This facility consists of a Sorting Station, a Super-Compactor, a Grouting Unit with Capping Device and other auxiliary equipment which is described in more detail in the following article. This article gives an overview of the efficient and safe treatment of nuclear waste. Babcock Noell is a subsidiary of the Bilfinger Power Systems and has 40 years of experience in the field of design, engineering, construction, static and dynamic calculations, manufacturing, installation, commissioning, as well as in the service and operation of a wide variety of nuclear components and facilities worldwide.

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.

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

Field test of a post-closure radiation monitor

International Nuclear Information System (INIS)

Reed, S.; Christy, C.E.; Heath, R.E.

1995-01-01

The DOE is conducting remedial actions at many sites contaminated with radioactive materials. After closure of these sites, long-term subsurface monitoring is typically required by law. This monitoring is generally labor intensive and expensive using conventional sampling and analysis techniques. The U.S. Department of Energy's Morgantown Energy Technology Center (METC) has contracted with Babcock and Wilcox to develop a Long-Term Post-Closure Radiation Monitoring System (LPRMS) to reduce these monitoring costs. A prototype LPRMS probe was built, and B ampersand W and FERMCO field tested this monitoring probe at the Fernald Environmental Management Project in the fall of 1994 with funding from the DOE's Office of Technology Development (EM-50) through METC. The system was used to measure soil and water with known uranium contamination levels, both in drums and in situ at depths up to 3 meters. For comparison purposes, measurements were also performed using a more conventional survey probe with a sodium iodide scintillator directly butt-coupled to detection electronics. This paper presents a description and the results of the field tests. The results were used to characterize the lower detection limits, precision and bias of the system, which allowed the DOE to judge the monitoring system's ability to meet its long-term post-closure radiation monitoring needs. Based on the test results, the monitoring system has been redesigned for fabrication and testing in a potential Phase III of this program. If the DOE feels that this system can meet its needs and chooses to continue into Phase III of this program, this redesigned full scale prototype system will be built and tested for a period of approximately a year. Such a system can be used at a variety of radioactively contaminated sites

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)

SALTSTONE VAULT CLASSIFICATION SAMPLES MODULAR CAUSTIC SIDE SOLVENT EXTRACTION UNIT/ACTINIDE REMOVAL PROCESS WASTE STREAM APRIL 2011

Energy Technology Data Exchange (ETDEWEB)

Eibling, R.

2011-09-28

Savannah River National Laboratory (SRNL) was asked to prepare saltstone from samples of Tank 50H obtained by SRNL on April 5, 2011 (Tank 50H sampling occurred on April 4, 2011) during 2QCY11 to determine the non-hazardous nature of the grout and for additional vault classification analyses. The samples were cured and shipped to Babcock & Wilcox Technical Services Group-Radioisotope and Analytical Chemistry Laboratory (B&W TSG-RACL) to perform the Toxic Characteristic Leaching Procedure (TCLP) and subsequent extract analysis on saltstone samples for the analytes required for the quarterly analysis saltstone sample. In addition to the eight toxic metals - arsenic, barium, cadmium, chromium, mercury, lead, selenium and silver - analytes included the underlying hazardous constituents (UHC) antimony, beryllium, nickel, and thallium which could not be eliminated from analysis by process knowledge. Additional inorganic species determined by B&W TSG-RACL include aluminum, boron, chloride, cobalt, copper, fluoride, iron, lithium, manganese, molybdenum, nitrate/nitrite as Nitrogen, strontium, sulfate, uranium, and zinc and the following radionuclides: gross alpha, gross beta/gamma, 3H, 60Co, 90Sr, 99Tc, 106Ru, 106Rh, 125Sb, 137Cs, 137mBa, 154Eu, 238Pu, 239/240Pu, 241Pu, 241Am, 242Cm, and 243/244Cm. B&W TSG-RACL provided subsamples to GEL Laboratories, LLC for analysis for the VOCs benzene, toluene, and 1-butanol. GEL also determines phenol (total) and the following radionuclides: 147Pm, 226Ra and 228Ra. Preparation of the 2QCY11 saltstone samples for the quarterly analysis and for vault classification purposes and the subsequent TCLP analyses of these samples showed that: (1) The saltstone waste form disposed of in the Saltstone Disposal Facility in 2QCY11 was not characteristically hazardous for toxicity. (2) The concentrations of the eight RCRA metals and UHCs identified as possible in the saltstone waste form were present at levels below the UTS. (3) Most of the

77 FR 67363 - Notice of Suspension and Commencement of Proposed Debarment Proceedings; Schools and Libraries...

Science.gov (United States)

2012-11-09

... of Federal Funds, May 11, 2011, at http://www.justice.gov/usao/ias/news/2011/Babcock%20-%20plea%20... District; $21,789.40 payable to the Lena-Winslow School District; $17,933.80 payable to North Boone School...

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).

Avaliação de diferentes tipos de atmosferas modificadas na vida útil de carambolas minimamente processadas Evaluation of different types of modified atmosphere in cold storage of star fruits minimally processed

Directory of Open Access Journals (Sweden)

Leandro Camargo Neves

2006-12-01

qualidade durante o AR. Assim, esse tratamento proporcionou adequado controle microbiológico e manutenção das características de qualidade por 18 dias para as carambolas minimamente processadas.This work aimed to evaluate the potential of cold storage of star fruits cv. Golden Star, minimally processed, by using different types of packing, for the analysis of the modified passive atmosphere. The fruits harvested in nature-green stage showed green-yellowish coloration, soluble solids (SS an average of 6,8 ºBrix and an average mass of 185g . Before the treatments were set, the fruits had been selected, cleaned in solution of NaOCl at 10 mg. L-1, cooled for 12 hours at 15 ± 0,5 °C, cut transversally, and cleaned in solution of NaOCl at 10 mg. L-1, for 3 minutes. After, the pieces in form of stars had been displayed in rigid polystyrene trays, with capacity for 250g, and coated with the following materials: T1: perforated film plastic of low density polyethylene (LDPE, with 0,006 mm; T2: polyolephinic plastic film with anti-fog of DuPont® (AGF, with 0,015mm; T3: polyolephinic plastic film of Dupont® (HF, with 0,015 mm; T4: LDPE plastic film, with 0,060 mm; T5: LDPE plastic film, with 0,080 mm; T6: polypropylene plastic film (PP, with 0,022 mm and T7: rigid polyethylene terephtalat tray (PET, with capacity for 500 mL, with cover of the same material. The treatments had been conditioned in refrigerating chamber at 12 ± 0,5 °C and 90 ± 3% of R.U., for 18 days. At the end, it was not observed significant difference in the tested treatments in the analysis of soluble solids (SS and titratable acidity (TA. The biggest microbiological activity and the greater pH had been detected in the fruits packed in LDPE films with 6 µm, AGF and HF with 15µm from Dupont. However, the fruits conditioned in packing PET had shown the biggest contents of ascorbic acid. In the same way, only in PET packing was obtained an efficient atmospheric modification in the point of maintaining the

Fraud and Fraudulent Practices in Nigeria Banking

African Journals Online (AJOL)

FIRST LADY

Owolabi, S. A. - Accounting Department, Babcock University, Nigeria. E-mail: ... Abstract. The paper reviews the various forms of fraudulent practice their impact and ... legislations targeted at reducing fraud and introducing some forms of control .... money market and treasury, risk assets, information technology, financial.

Multidrug and vancomycin resistance among clinical isolates of ...

African Journals Online (AJOL)

2017-09-03

Sep 3, 2017 ... Babcock University, Ilisan-Remo, Ogun State, Nigeria ... 2017 Olufunmiso et al; licensee African Health Sciences. This is an Open ..... Cole AM, Tahk S, Oren A, Yoshioka D, Kim YH, Park .... Part 2: Cambridge University press,.

Healy Clean Coal Project: A DOE Assessment

Energy Technology Data Exchange (ETDEWEB)

National Energy Technology Laboratory

2003-09-01

The goal of the U.S. Department of Energy's (DOE) Clean Coal Technology (CCT) Program is to provide the energy marketplace with advanced, more efficient, and environmentally responsible coal utilization options by conducting demonstrations of new technologies. These demonstration projects are intended to establish the commercial feasibility of promising advanced coal technologies that have been developed to a level at which they are ready for demonstration testing under commercial conditions. This document serves as a DOE post-project assessment (PPA) of the Healy Clean Coal Project (HCCP), selected under Round III of the CCT Program, and described in a Report to Congress (U.S. Department of Energy, 1991). The desire to demonstrate an innovative power plant that integrates an advanced slagging combustor, a heat recovery system, and both high- and low-temperature emissions control processes prompted the Alaska Industrial Development and Export Authority (AIDEA) to submit a proposal for this project. In April 1991, AIDEA entered into a cooperative agreement with DOE to conduct this project. Other team members included Golden Valley Electric Association (GVEA), host and operator; Usibelli Coal Mine, Inc., coal supplier; TRW, Inc., Space & Technology Division, combustor technology provider; Stone & Webster Engineering Corp. (S&W), engineer; Babcock & Wilcox Company (which acquired the assets of Joy Environmental Technologies, Inc.), supplier of the spray dryer absorber technology; and Steigers Corporation, provider of environmental and permitting support. Foster Wheeler Energy Corporation supplied the boiler. GVEA provided oversight of the design and provided operators during demonstration testing. The project was sited adjacent to GVEA's Healy Unit No. 1 in Healy, Alaska. The objective of this CCT project was to demonstrate the ability of the TRW Clean Coal Combustion System to operate on a blend of run-of-mine (ROM) coal and waste coal, while meeting strict

A comparative study of EVA with and without thermal history for different lamination process parameters

Energy Technology Data Exchange (ETDEWEB)

Drabczyk, Kazimierz, E-mail: kazimierz.drabczyk@wp.pl [Institute of Metallurgy and Materials Science, 25 Reymonta St., 30-059 Krakow (Poland); Panek, Piotr [Institute of Metallurgy and Materials Science, 25 Reymonta St., 30-059 Krakow (Poland)

2012-09-01

In more than 80% of the worldwide photovoltaic (PV) modules, mostly very fragile and 200 {mu}m thick, crystalline silicon solar cells are encapsulated into ethylene-vinyl acetate (EVA) foils, which bond the module components together, provide physical protection, electrical insulation and a barrier for moisture ingress. The understanding of what can happen with EVA during its transport, storage and lamination process is necessary to optimize the quality of the PV module for its long exposure to outdoor weather conditions. Achieving a proper cross-link density of over 70%, it is essential to overcome the cold flow of EVA and to make the module durable. In this work, the feasibility of the use of differential scanning calorimetry (DSC) compared with the solvent extraction (SE) method by toluene were evaluated in order to provide structural information on the EVA curing kinetics and the cross-link density. DSC tests were performed on a DTA DuPont1600 tester. The temperature range for the test was from -50 Degree-Sign C to 200 Degree-Sign C, with the heating rate of 10 Degree-Sign C/min, and the endothermic and exothermic peaks were evaluated. Toluene solvent extractions were performed on the same set of samples that were analyzed by DSC. The measured cross-link density shows a direct dependence on the pre-lamination conditions of EVA, which is in good agreement with the data obtained with the DSC method.

A comparative study of EVA with and without thermal history for different lamination process parameters

International Nuclear Information System (INIS)

Drabczyk, Kazimierz; Panek, Piotr

2012-01-01

In more than 80% of the worldwide photovoltaic (PV) modules, mostly very fragile and 200 μm thick, crystalline silicon solar cells are encapsulated into ethylene-vinyl acetate (EVA) foils, which bond the module components together, provide physical protection, electrical insulation and a barrier for moisture ingress. The understanding of what can happen with EVA during its transport, storage and lamination process is necessary to optimize the quality of the PV module for its long exposure to outdoor weather conditions. Achieving a proper cross-link density of over 70%, it is essential to overcome the cold flow of EVA and to make the module durable. In this work, the feasibility of the use of differential scanning calorimetry (DSC) compared with the solvent extraction (SE) method by toluene were evaluated in order to provide structural information on the EVA curing kinetics and the cross-link density. DSC tests were performed on a DTA DuPont1600 tester. The temperature range for the test was from −50 °C to 200 °C, with the heating rate of 10 °C/min, and the endothermic and exothermic peaks were evaluated. Toluene solvent extractions were performed on the same set of samples that were analyzed by DSC. The measured cross-link density shows a direct dependence on the pre-lamination conditions of EVA, which is in good agreement with the data obtained with the DSC method.

Containment technology and the success of a joint initiative

International Nuclear Information System (INIS)

Campbell, H.J. Jr.

1997-01-01

This paper describes the efforts of the DuPont Company and DOE on containment of pollution. A test program at the Groundwater Remediation Field Laboratory National Test Site at Dover Air Force Base is discussed

77 FR 51739 - Approval and Promulgation of Implementation Plans; Tennessee; Regional Haze State Implementation...

Science.gov (United States)

2012-08-27

... Tennessee: (1) Aluminum Company of America (Alcoa)--South Plant; (2) DuPont White Pigment and Mineral... reductions than BART (i.e., dispersion modeling is not required to evaluate the differences in visibility...

Incineration with energy recovery

Energy Technology Data Exchange (ETDEWEB)

Mahoney, T.G.

1986-02-01

Motherwell Bridge Tacol Ltd. operate a 'Licence Agreement' with Deutsche Babcock Anlagen of Krefeld, West Germany, for the construction of Municipal Refuse Incineration plant and Industrial Waste plant with or without the incorporation of waste heat recovery equipment. The construction in the UK of a number of large incineration plants incorporating the roller grate incinerator unit is discussed. The historical background, combustion process, capacity, grate details, refuse analysis and use as fuel, heat recovery and costs are outlined.

Technology transfer for reduced environmental pollution in the former USSR - the role of Western companies

Energy Technology Data Exchange (ETDEWEB)

Hill, M.R.

1995-12-01

The paper is divided into two main sections, the first presenting information on the volumes of Western exports of power engineering products to the former USSR, and the second describing the technology transfer experiences of a sample of Western power generation, fuels, and power engineering companies. The information relating to those companies was obtained from published information and press reports, and visits for discussions with senior executives during 1994 and 1995. Companies involved in technology transfer and investment activities described in the report are: IVO International Ltd., CRE Group Ltd., Asea Brown Boveri, Siemens Kraftwerkunion, GEC Alsthorn Rolly Royce Industrial Power Group, Babcock Energy, Deutsche Babcock AG, Alsthom Boilers, and GE Corporation. 35 refs., 5 tabs.

STEM Equality and Diversity Toolkit

Science.gov (United States)

Collins, Jill

2011-01-01

In 2008, the Centre for Science Education at Sheffield Hallam University teamed up with VT Enterprise (now Babcock International) in their submission of a successful bid to deliver the national STEM (Science, Technology, Engineering and Maths) Subject Choice and Careers Project. An integral part of the bid was the promotion of equality and…

Establishing a conversion factor between electrical conductivity and ...

African Journals Online (AJOL)

2015-07-04

Jul 4, 2015 ... In large areas of South Africa, acid mine drainage (AMD) from gold, coal ..... salinity (Scofield and Wilcox, 1931; Rainwater and Thatcher,. 1960; McNeil .... sites belonging to different river systems, but varied little for individual ...

Magnetically Enhanced Solid-Liquid Separation

Science.gov (United States)

Rey, C. M.; Keller, K.; Fuchs, B.

2005-07-01

DuPont is developing an entirely new method of solid-liquid filtration involving the use of magnetic fields and magnetic field gradients. The new hybrid process, entitled Magnetically Enhanced Solid-Liquid Separation (MESLS), is designed to improve the de-watering kinetics and reduce the residual moisture content of solid particulates mechanically separated from liquid slurries. Gravitation, pressure, temperature, centrifugation, and fluid dynamics have dictated traditional solid-liquid separation for the past 50 years. The introduction of an external field (i.e. the magnetic field) offers the promise to manipulate particle behavior in an entirely new manner, which leads to increased process efficiency. Traditional solid-liquid separation typically consists of two primary steps. The first is a mechanical step in which the solid particulate is separated from the liquid using e.g. gas pressure through a filter membrane, centrifugation, etc. The second step is a thermal drying process, which is required due to imperfect mechanical separation. The thermal drying process is over 100-200 times less energy efficient than the mechanical step. Since enormous volumes of materials are processed each year, more efficient mechanical solid-liquid separations can be leveraged into dramatic reductions in overall energy consumption by reducing downstream drying requirements have a tremendous impact on energy consumption. Using DuPont's MESLS process, initial test results showed four very important effects of the magnetic field on the solid-liquid filtration process: 1) reduction of the time to reach gas breakthrough, 2) less loss of solid into the filtrate, 3) reduction of the (solids) residual moisture content, and 4) acceleration of the de-watering kinetics. These test results and their potential impact on future commercial solid-liquid filtration is discussed. New applications can be found in mining, chemical and bioprocesses.

A preliminary report on the artesian water supply of Memphis, Tennessee

Science.gov (United States)

Wells, F.G.

1932-01-01

Memphis is located in the part of the Gulf Coastal Plain known as the Mississippi embayment. It is underlain by unconsolidated sand and clay formations of Tertiary and Cretaceous age. The Wilcox group, of Tertiary age, and the Ripley formation, of Cretaceous age, are excellent aquifers, and all the water consumed in Memphis is derived from them. The maximum pumpage from the Wilcox group was reached about 1920; in that year an estimated average of 37,575,000 gallons a day was pumped. In 1928 the average daily pumpage from the Wilcox group was about 33,984,000 gallons, and in addition to this the Memphis Artesian Water Department pumped an average of 4,616,000 gallons a day from the Ripley formation. The static level at Memphis varies with the pumpage and the stage of the Mississippi River. The original static level was about 235 feet above mean sea level. In 1928 the average static level at the Auction Avenue plant was 202 feet above mean sea level, which was about 33 feet lower than the original level. The yield is therefore about a million gallons a day for each foot of drawdown. The drawdown is not excessive, and additional pumpage can be developed without undue lowering of head. The water from both the Wilcox group and the Ripley formation is fairly soft and has a moderately low content of dissolved mineral matter. The iron content is sufficiently high to be objectionable, but the iron is easily removed by aeration followed by either settling or filtration for removal of sediment.

ORF Alignment: NC_003063 [GENIUS II[Archive

Lifescience Database Archive (English)

Full Text Available in C58, Dupont) ... Length = 112 ... Query: 205 IARSIVWIRKHFRETCSVEAGAEVAGMSRSTFHAHFKAVTNMSPLEFRTQLRLQEA...RRLM 264 ... IARSIVWIRKHFRETCSVEAGAEVAGMSRSTFHAHFKAVTNMSPLEFRTQLRLQEARRLM Sbjct: 1 ... IARSIVWIRKHFRETCSVEAGAEVAGMSRSTFHAHFKAVTNMSPLEFRTQLRLQEARRLM 60 ...

X-Ray Characterization of Resistor/Dielectric Material for Low Temperature Co-Fired Ceramic Packages

International Nuclear Information System (INIS)

DIMOS, DUANE B.; KOTULA, PAUL G.; RODRIGUEZ, MARK A.; YANG, PIN

1999-01-01

High temperature XRD has been employed to monitor the devitrification of Dupont 951 low temperature co-fired ceramic (LTCC) and Dupont E84005 resistor ink. The LTCC underwent devitrification to an anorthite phase in the range of 835-875 C with activation energy of 180 kJ/mol as calculated from kinetic data. The resistor paste underwent devitrification in the 835-875 C range forming monoclinic and hexagonal celcian phases plus a phase believed to be a zinc-silicate. RuO(sub 2) appeared to be stable within this devitrified resistor matrix. X-ray radiography of a co-fired circuit indicated good structural/chemical compatibility between the resistor and LTCC

COST-EFFECTIVE CONTROL OF NOx WITH INTEGRATED ULTRA LOW-NOx BURNERS AND SNCR

International Nuclear Information System (INIS)

Hamid Farzan

2001-01-01

Coal-fired electric utilities are facing a serious challenge with regards to curbing their NO(sub x) emissions. At issue are the NO(sub x) contributions to the acid rain, ground level ozone, and particulate matter formation. Substantial NO(sub x) control requirements could be imposed under the proposed Ozone Transport Rule, National Ambient Air Quality Standards, and New Source Performance Standards. McDermott Technology, Inc. (MTI), Babcock and Wilcox (B and W), and Fuel Tech are teaming to provide an integrated solution for NO(sub x) control. The system will be comprised of an ultra low-NO(sub x) pulverized coal (PC) burner technology plus a urea-based, selective non-catalytic reduction (SNCR) system. This system will be capable of meeting a target emission limit of 0.15 lb NO(sub x)/10(sup 6) Btu and target ammonia (NH3) slip level targeted below 5 ppmV for commercial units. Our approach combines the best available combustion and post-combustion NO(sub x) control technologies. More specifically, B and W's DRB-4Z TM ultra low-NO(sub x) PC burner technology will be combined with Fuel Tech's NO(sub x)OUT (SNCR) and NO(sub x)OUT Cascade (SNCR/SCR hybrid) systems and jointly evaluated and optimized in a state-of-the-art test facility at MTI. Although the NO(sub x)OUT Cascade (SNCR/SCR hybrid) system will not be tested directly in this program, its potential application for situations that require greater NO(sub x) reductions will be inferred from other measurements (i.e., SNCR NO(sub x) removal efficiency plus projected NO(sub x) reduction by the catalyst based on controlled ammonia slip). Our analysis shows that the integrated ultra low-NO(sub x) burner and SNCR system has the lowest cost when the burner emissions are 0.25 lb NO(sub x)/10(sup 6) Btu or less. At burner NO(sub x) emission level of 0.20 lb NO(sub x)/10(sup 6) Btu, the levelized cost per ton of NO(sub x) removed is 52% lower than the SCR cost

ЗОИГЛФЛГ СЕЕЛЖЗРЦГОЛХ ССН

African Journals Online (AJOL)

USER

2015-02-17

Feb 17, 2015 ... Address: Department of Agricultural Economics & Extension, Babcock University, Ilishan-. Remo ... This study examined determinants of farm productivity among fluted pumpkin ..... Asian Vegetable Research Development.

Natural Phenomena Hazards Design Criteria and Other Characterization Information for the MFFF at SRS

Energy Technology Data Exchange (ETDEWEB)

Wyatt, D.E.

2000-12-01

This report is a comprehensive complication applicable to the general Savannah River Site area, developed by both the original contractor, the DuPont Company, and by the current plant operator, Westinghouse Savannah River Company, over the full plant lifetime period (1950 - 2000).

Comparative Assessment of Copper-Coated Kapton: Analysis of Microclad from Several Manufacturing Lots

Energy Technology Data Exchange (ETDEWEB)

Martz, Joseph Christopher [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Spearing, Dane Robert [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Labouriau, Andrea [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Judge, Elizabeth [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Kelly, Daniel [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Dirmyer, Matthew R. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Milenski, Helen Marie [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Patterson, Brian M. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Sandoval, Cynthia Wathen [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Usov, Igor Olegovich [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Beaux, Miles Frank II [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Henderson, Kevin C. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Torres, Joseph Angelo [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Edwards, Stephanie Lynn [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Vodnik, Douglas R. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Keller, Jennie [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Mccabe, Rodney James [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Livescu, Veronica [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Cowan, Joseph Sarno [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Aragonez, Robert J. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Tokash, Justin Charles [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Lawrence, Samantha Kay [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Leon Brito, Neliza [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2017-04-03

Microclad is a composite material consisting of a thin copper coating applied on a single side over a Kapton substrate. Kapton is the commercial designator for polyimide supplied by DuPont. Microclad is a key material in detonator manufacture and function. Detonators which utilize Microclad function when a large current applied through a thin bridge etched into the copper produces a plasma, accelerating a Kapton flyer into an explosive (PETN) pellet. The geometry and properties of the Microclad are a critical element of this process.

77 FR 3774 - Formations of, Acquisitions by, and Mergers of Bank Holding Companies

Science.gov (United States)

2012-01-25

..., Dupont, Indiana. River Valley Financial Bank, Madison, Indiana, the existing thrift subsidiary of... FEDERAL RESERVE SYSTEM Formations of, Acquisitions by, and Mergers of Bank Holding Companies The companies listed in this notice have applied to the Board for approval, pursuant to the Bank Holding Company...

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