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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. The nozzles were located in the center of the plate with their centers 3.0 in. apart and were considered to be free of weld distortions and irregularities in the junction region. 6 references. (auth)
Date: December 1, 1975
Creator: Bryson, J.W. & Swinson, W.F.
Partner: UNT Libraries Government Documents Department

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 nucleated sites around the pipe were also found. Results of destructive analysis and nondestructive testing allowed Los Alamos staff to conclude that the direct cause for the main crack and other pitting resulted from stress-assisted, microbial-induced corrosion of the stainless steel primary piping. The results also indicated that microbial action from bacteria that are normally present in earth can be extremely harmful to stainless- steel piping under certain conditions. Other potential problems that could have also eventually led to a permanent shutdown of the OWR were discussed. These problems, although never encountered nor associated with the current shutdown, were identified in aging studies and are associated with: (1) the water-cooled, bismuth gamma-ray shield and, (2) the aluminum thermal column head seal that prevents reactor vessel water from entering into the graphite-filled thermal column.
Date: July 1, 1995
Creator: Andrade, A.
Partner: UNT Libraries Government Documents Department

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 equations for thin plates and shells. The stresses due to moment loading calculated by the two methods are essentially identical for identical boundary conditions. A description of the general model of flanges used in the theoretical development of the computer code is provided. The actual mathematical expressions for calculating stresses and displacements due to moment and pressure loads are derived. Example calculations, listings, and flowcharts of the program and its subroutines are included as appendices. (auth)
Date: January 1, 1976
Creator: Rodabaugh, E.C.; O'Hara, F.M. Jr. & Moore, S.E.
Partner: UNT Libraries Government Documents Department

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 fastest settling time, fastest rise time and smallest peak overshoot when compared to the fixed H{sub {infinity}} and classical controllers. This makes for a superior and more robust controller.
Date: May 1996
Creator: Power, M. A. & Edwards, R. M.
Partner: UNT Libraries Government Documents Department

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 visited PWR plants to analyze these events. Our assessment shows that USNRC licensees have taken effective actions to reduce the number of leak events. One main reason for this decreasing trend was the elimination or reportable leakages from valve stem packing after 1991. Our review of leak events related to vibratory fatigue reveals a statistically significant decreasing trend with age (years of operation), but not in calendar time. Our assessment of worldwide data on leakage caused by thermal fatigue cracking is that the fatigue of aging piping is a safety significant issue. Our review of leak events has identified several susceptible sites in piping having high safety significance; but the inspection of some of these sites is not required by the ASME Code. These sites may be included in the risk-informed inspection programs.
Date: February 1, 1999
Creator: Shah, Vikram N.; Ware, Arthur G.; Atwood, Cory L.; Sattison, Martin B.; Hartley, R. Scott & Hsu, Chuck
Partner: UNT Libraries Government Documents Department

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 by available information. Therefore, it is recommended that this combined loading evaluation procedure be replaced for primary loads by the separate leg evaluation procedure specified in NC/ND-3653.3(c) and NC/ND-3653.3(d). No recommendation is made for fatigue or secondary load evaluations for Class 1 piping. Further work should be done on the development of better criteria for treatment of combined branch and run moment effects.
Date: November 1, 1993
Creator: Rodabaugh, E. C.; Gwaltney, R. C. & Moore, S. E.
Partner: UNT Libraries Government Documents Department

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 defining oxide film growth on 6061 aluminum alloy cladding. Additionally, a data correlation was developed from the database to permit the prediction of film growth for any reasonable thermal-hydraulic excursion. This capability was utilized effectively during the conceptual design stages of the reactor. During the course of this research, it became clear that the kinetics of film growth on the aluminum alloy specimens were sensitively dependent on the chemistry of the aqueous coolant and that relatively small deviations from the intended pH 5 operational level resulted in unexpectedly large changes in the corrosion behavior. Examination of the kinetic influences and the details of the film morphology suggested that a mechanism involving mass transport from other parts of the test loop was involved. Such a mechanism would also be expected to be active in the operating reactor. This report emphasizes the results of experiments that best illustrate the influence of the nonthermal-hydraulic parameters on film growth and presents data to show that comparatively small variations in pH near 5.0 invoke a sensitive response. Simply, for operation in the temperature and heat flux range appropriate for the ANS studies, coolant pH levels from 4.5 to 4.9 produced significantly less film growth than those from pH 5.1 to 6. A mechanism for this behavior based on the concept of treating the entire loop as an active corrosion ...
Date: October 1, 1995
Creator: Pawel, S.J.; Felde, D.K. & Pawel, R.E.
Partner: UNT Libraries Government Documents Department

Deterministic and probabilistic evaluations for uncertainty in pipe fracture parameters in leak-before-break and in-service flaw evaluations

Description: This report presents new results from deterministic and probabilistic analyses to evaluate the significance of a number of technical aspects that may affect LBB or in-service flaw evaluations. The following summarizes the objectives and results from both the deterministic and probabilistic studies. The reasons for including each technical aspect being evaluated are given first. Then a table is given that summarizes the relative significance of each technical aspect. In most cases there are both deterministic and probabilistic results. The deterministic analyses were conducted independently of the probabilistic analysis, which offered the opportunity to validate conclusions from each of these studies.
Date: June 1, 1996
Creator: Ghadiali, N.; Wilkowski, G.; Rahman, S. & Choi, Y.H.
Partner: UNT Libraries Government Documents Department

Sensitivity of SRP LOCA power limit to break size and location

Description: SRP reactors are low pressure, heavy water reactors with six external process water loops that drive the coolant into an upper plenum and then downward through the assemblies. Assembly LOCA power limits are currently set in these reactors to prevent Ledinegg flow instability (FI) in any assembly flow channel. These limits are based on a postulated break area and location. This study determined the sensitivity of the power limit to the break area and location.
Date: 1989-06~
Creator: White, A. M.; Pevey, R. E. & Smith, F. G.
Partner: UNT Libraries Government Documents Department

TRAC analysis of design basis events for the accelerator production of tritium target/blanket

Description: A two-loop primary cooling system with a residual heat removal system was designed to mitigate the heat generated in the tungsten neutron source rods inside the rungs of the ladders and the shell of the rungs. The Transient Reactor Analysis Code (TRAC) was used to analyze the thermal-hydraulic behavior of the primary cooling system during a pump coastdown transient; a cold-leg, large-break loss-of-coolant accident (LBLOCA); a hot-leg LBLOCA; and a target downcomer LBLOCA. The TRAC analysis results showed that the heat generated in the tungsten neutron source rods can be mitigated by the primary cooling system for the pump coastdown transient and all the LBLOCAs except the target downcomer LBLOCA. For the target downcomer LBLOCA, a cavity flood system is required to fill the cavity with water at a level above the large fixed headers.
Date: August 1, 1997
Creator: Lin, J.C. & Elson, J.
Partner: UNT Libraries Government Documents Department

FY 1995 progress report on the ANS thermal-hydraulic test loop operation and results

Description: The Thermal-Hydraulic Test Loop (THTL) is an experimental facility constructed to support the development of the Advanced Neutron Source Reactor (ANSR) at Oak Ridge National Laboratory (ORNL). The THTL facility was designed and built to provide known thermal-hydraulic (T/H) conditions for a simulated full-length coolant subchannel of the ANS reactor core, thus facilitating experimental determination of FE and CHF thermal limits under expected ANSR T/H conditions. Special consideration was given to allow operation of the system in a stiff mode (constant flow) and in a soft mode (constant pressure drop) for proper implementation of true FE and DNB experiments. The facility is also designed to examine other T/H phenomena, including onset of incipient boiling (IB), single-phase heat transfer coefficients and friction factors, and two-phase heat transfer and pressure drop characteristics. Tests will also be conducted that are representative of decay heat levels at both high pressure and low pressure as well as other quasi-equilibrium conditions encountered during transient scenarios. A total of 22 FE tests and 2 CHF tests were performed during FY 1994 and FY 1995 with water flowing vertically upward. Comparison of these data as well as extensive data from other investigators led to a proposed modification to the Saha and Zuber correlation for onset of significant void (OSV), applied to FE prediction. The modification takes into account a demonstrated dependence of the OSV or FE thermal limits on subcooling levels, especially in the low subcooling regime.
Date: July 1, 1997
Creator: Siman-Tov, M.; Felde, D.K.; Farquharson, G.; McDuffee, J.L.; McFee, M.T.; Ruggles, A.E. et al.
Partner: UNT Libraries Government Documents Department