15 Matching Results

Search Results

Advanced search parameters have been applied.

Prestressed Concrete Reactor Vessel Model 1

Description: From introduction: The goal of engineers associated with nuclear power plants is the achievement of safe plants with low generating costs.One possible means of lowering costs is to increase the power generating capability for a single generating unit. To accomplish this, the sizes of nuclear reactors have been increased.
Date: October 25, 1966
Partner: UNT Libraries Government Documents Department

Analysis of Proposed Gamma-Ray Detection System for the Monitoring of Core Water Inventory in a Pressurized Water Reactor

Description: An initial study has been performed of the feasibility of employing an axial array of gamma detectors located outside the pressure vessel to monitor the coolant in a PWR. A one-dimensional transport analysis model is developed for the LOFT research reactor and for a mock-PWR geometry. The gamma detector response to coolant voiding in the core and down-comer has been determined for both geometries. The effects of various conditions (for example, time after shutdown, materials in the transport path, and the relative void fraction in different water regions) on the detector response are studied. The calculational results have been validated by a favorable comparison with LOFT experimental data. Within the limitations and approximations considered in the analysis, the results indicate that the gamma-ray detection scheme is able to unambiguously respond to changes in the coolant inventory within any vessel water region.
Date: December 1987
Creator: Markoff, Diane Melanie
Partner: UNT Libraries Government Documents Department

DYNAPCON: A Computer Code for Dynamic Analysis of Prestressed Concrete Structures

Description: A finite element computer code for the transient analysis of prestressed concrete reactor vessels (PCRVs) for LMFBR containment is described. The method assumes rotational symmetry of the structure. Time integration is by an explicit method. The quasistatic prestressing operation of the PCRV model is performed by a dynamic relaxation technique. The material model accounts for the crushing and tensile cracking in arbitrary direction in concrete and the elastic-plastic behavior of reinforcing steel. The variation of the concrete tensile cracking and compressive crushing limits with strain rate is taken into account. Relative slip is permitted between the concrete and tendons. Several example solutions are presented and compared with experimental results. These sample problems range from simply supported beams to small scale models of PCRV's. It is shown that the analytical methods correlate quite well with experimental results, although in the vicinity of the failure load the response of the models tend to be quite sensitive to input parameters.
Date: September 1982
Creator: Marchertas, A. H.
Partner: UNT Libraries Government Documents Department

The Rolling of Gold Sheet for LAPRE 2 Pressure Liner

Description: From abstract: "This report discusses a special handling procedure devised to prevent contamination of high-purity gold while processing the metal from cast billet to final sheet. Gold of 99.99% purity was cast into rectangular billets and rolled into sheet whose final dimensions were 48 by 12-1/2 by 0.015 inch...As a final step, the sheet was annealed dead-soft in preparation for subsequent forming operations."
Date: July 1958
Creator: Keil, R. W.; Hanks, G. S. & Taub, J. M.
Partner: UNT Libraries Government Documents Department

Design of the Argonne Low Power Reactor (ALPR)

Description: Report containing "[a] description (...) of the design of a prototype "packaged" nuclear power plant. The purpose of the plant is to alleviate fuel oil logistics and storage problems posed by remote auxiliary DEW Line radar stations north of the Arctic Circle. The ALPR (redesignated SL-1) is a 3 Mwt, heterogeneous, highly enriched uranium- fueled, natural-circulation boiling water reactor, cooled and moderated with light water. Steam at 300 psig, dry and saturated (421 deg F) is passed directly from the reactor to a conventional turbine-generator to produce electric power (300 kw nominal) and space-heating (400 kw) requirements consistent with rigid mechanical and structural specifications prescribed by the military, and dictated by the extreme geophysics prevailing at the ultimate site. The over-all design criteria emphasize: simplicity and reliability of operation and maintenance, with minimum supervision; minimum on-site construction; maximum use of standard components; limited water supply; utilization of local gravel for biological shielding; transportability by air lift; and nominal 3-year fuel operating lifetime per core loading." (p. 15)
Date: May 1961
Creator: Hamer, E. E.; Grant, N. R.; Hooker, H. H.; Jorgensen, G. L.; Kann, W. J.; Lipinski, W. C. et al.
Partner: UNT Libraries Government Documents Department

Design-Development and Operation of the Experimental Boiling-Water Reactor (EBWR) Facility, 1955--1967

Description: The Experimental Boiling-Water Reactor (EBWR) was designed, built, and operated to provide experience and engineering data that would demonstrate the feasibility of the direct-cycle, boiling-water reactor and be applicable to improved, larger nuclear power stations; and was based on information obtained in the first test boiling-water reactors, the BORAX series. EBWR initially produced 20 MW(t), 5 MW(e); later modified and upgraded, as described and illustrated, it was operated at up to 100 MW(t). The facility fulfilled its primary mission - demonstrating the practicality of the direct-boiling concept - and, in fact, was the prototype of some of the first commercial plants and of reactor programs in some other countries. After successful completion of the Water-Cooled Reactor Program, EBWR was utilized in the joint Argonne-Hanford Plutonium Recycle Program to develop data for the utilization of plutonium as a fuel in light-water thermal systems. Final shutdown of the EBWR facility followed the termination of the latter program.
Date: November 1990
Creator: Boing, L. E.; Wimunc, E. A. & Whittington, G. A.
Partner: UNT Libraries Government Documents Department

Comparison of REXCO Code Predictions with SRI SM-2 Experimental Results

Description: This report deals with the REXCO-code predictions of the SRI SM-2 test. Two calculations were performed with the REXCO-HEP code: one used the pressure history of the core detonation products as input and the other the pressure-volume relations of the detonation products as input. The other inputs of the computer analysis are the vessel and the core-barrel dimensions and boundary conditions, the constitutive equations of the vessel and the core barrel materials, and the equation of state for the coolant. The REXCO-predicted well deformations, pressure loadings, and dynamic strain histories at various gauge positions are compared with the experimental data. Results of the comparisons are discussed.
Date: August 1978
Creator: Chang, Y. W. & Gvildys, J.
Partner: UNT Libraries Government Documents Department

An Evaluation of Alternative Reactor Vessel Cutting Technologies for the Experimental Boiling Water Reactor at Argonne National Laboratory

Description: Metal cutting techniques that can be used to segment the reactor pressure vessel of the Experimental Boiling Water Reactor (EBWR) at Argonne National Laboratory (ANL) have been evaluated by Nuclear Energy Services. Twelve cutting technologies are described in terms of their ability to perform the required task, their performance characteristics, environmental and radiological impacts, and cost and schedule considerations. Specific recommendations regarding which technology should ultimately be used by ANL are included. The selection of a cutting method was the responsibility of the decommissioning staff at ANL, who included a relative weighting of the parameters described in this document in their evaluation process.
Date: December 1989
Creator: Boing, L. E.; Henley, D. R.; Manion, William J. & Gordon, J. W.
Partner: UNT Libraries Government Documents Department