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Stress analyses of flat plates with attached nozzles. Vol. 3. Experimental stress analyses of a flat plate with two closely spaced nozzles of equal diameter attached

Description: The complete test results for a flat plate with two closely spaced nozzles attached are presented. Test loadings were 1:1, 1:2, and 2:1 biaxial planar tension loadings on the plate, axial thrust loadings applied separately to the nozzles, and bending moment loadings applied to the nozzles both within and normal to the plane of symmetry containing the nozzle axes. The test plate was 36 x 36 x 0.375 in., and the attached nozzles had outer diameters of 2.625 in. and wall thicknesses of 0.250 in. T… more
Date: December 1, 1975
Creator: Bryson, J. W. & Swinson, W. F.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

FLANGE: a computer program for the analysis of flanged joints with ring- type gaskets

Description: The computer program FLANGE was written to calculate not only the stresses due to moment loads on the flange ring but also stresses due to internal pressure; stresses due to a temperature difference between the hub and ring; and stresses due to the variations in bolt load that result from pressure, hub-ring temperature gradient, and/or bolt-ring temperature difference. The program FLANGE is applicable to tapered-hub, straight, and blind flanges. The analysis method is based on the differential … more
Date: January 1, 1976
Creator: Rodabaugh, E. C.; O'Hara, F. M., Jr. & Moore, S. E.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

Stress-assisted, microbial-induced corrosion of stainless steel primary piping and other aging issues at the Omega West Reactor

Description: After the discovery of cooling system leak of about 284 liters per twenty-four (24) hour period, an investigation determined that the 76.2-cm diameter, 33.5-m long stainless-steel (304) OWR delay line was losing water at the same nominal rate. An excavation effort revealed that a circumferential crack, approximately 0.0025 cm in width, extended around the bottom half of the delay line. In addition, other evidence of what appeared to be microcracking and pitting that originated at random nucleat… more
Date: July 1, 1995
Creator: Andrade, A.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
open access

Assessment of Field Experience Related to Pressurized Water Reactor Primary System Leaks

Description: This paper presents our assessment of field experience related to pressurized water reactor (PWR) primary system leaks in terms of their number and rates, how aging affects frequency of leak events, the safety significance of such leaks, industry efforts to reduce leaks, and effectiveness of current leak detection systems. We have reviewed the licensee event reports to identify the events that took place during 1985 to the third quarter of 1996, and reviewed related technical literature and vis… more
Date: February 1, 1999
Creator: Shah, Vikram N.; Ware, Arthur G.; Atwood, Cory L.; Sattison, Martin B.; Hartley, R. Scott & Hsu, Chuck
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
open access

Adaptive robust control of the EBR-II reactor

Description: Simulation results are presented for an adaptive H{sub {infinity}} controller, a fixed H{sub {infinity}} controller, and a classical controller. The controllers are applied to a simulation of the Experimental Breeder Reactor II primary system. The controllers are tested for the best robustness and performance by step-changing the demanded reactor power and by varying the combined uncertainty in initial reactor power and control rod worth. The adaptive H{sub {infinity}} controller shows the fast… more
Date: May 1996
Creator: Power, M. A. & Edwards, R. M.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
open access

Review of ASME code criteria for control of primary loads on nuclear piping system branch connections and recommendations for additional development work

Description: This report collects and uses available data to reexamine the criteria for controlling primary loads in nuclear piping branch connections as expressed in Section III of the ASME Boiler and Pressure Vessel Code. In particular, the primary load stress indices given in NB-3650 and NB-3683 are reexamined. The report concludes that the present usage of the stress indices in the criteria equations should be continued. However, the complex treatment of combined branch and run moments is not supported … more
Date: November 1, 1993
Creator: Rodabaugh, E. C.; Gwaltney, R. C. & Moore, S. E.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

Reactor primary coolant system pipe rupture study. Progress report No. 29, January--June 1973

Description: The pipe rupture study is designed to extend the understanding of failure-causing mechanisms and to provide improved capability for evaluating reactor piping systems to minimize the probability of failures. Following a detailed review to determine the effort most needed to improve nuclear system piping (Phase I), analytical and experimental efforts (Phase II) were started in 1965. The recent accomplishments of a broad program in (a) fatigue studies through both bench-scale and large-component t… more
Date: August 1, 1973
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

Influence of coolant pH on corrosion of 6061 aluminum under reactor heat transfer conditions

Description: To support the design of the Advanced Neutron Source (ANS), an experimental program was conducted wherein aluminum alloy specimens were exposed at high heat fluxes to high-velocity aqueous coolants in a corrosion test loop. The aluminum alloys selected for exposure were candidate fuel cladding materials, and the loop system was constructed to emulate the primary coolant system for the proposed ANS reactor. One major result of this program has been the generation of an experimental database defi… more
Date: October 1, 1995
Creator: Pawel, S. J.; Felde, D. K. & Pawel, R. E.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

Coupled reactor physics and coolant dynamics of heavy liquid metal coolant systems.

Description: Cooling of advanced nuclear designs with heavy liquid metals such as lead or lead-bismuth eutectic offers the potential for plant simplifications and higher operating efficiencies compared to previously considered liquid metal coolants such as sodium or NaK. Such applications would however also introduce additional safety concerns and design challenges, therefore necessitating a verifiable computational tool for transient design-basis analysis of heavy liquid metal coolant (HLMC) systems. This … more
Date: July 15, 1999
Creator: Cahalan, J. E.; Dunn, F. E. & Taiwo, T. A.
Item Type: Refine your search to only Article
Partner: UNT Libraries Government Documents Department
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