Search Results

Advanced search parameters have been applied.
open access

Assessment of light water reactor fuel damage during a reactivity initiated accident

Description: This paper presents an assessment of LWR fuel damage during a reactivity initiated accident and comments on the adequacy of the present USNRC design requirements. Results from early SPERT tests are reviewed and compared with results from recent computer simulations and PBF tests. A progression of fuel rod and cladding damage events is presented. High strain rate deformation of relatively cool irradiated cladding early in the transient may result in fracture at a radial average peak fuel enthalp… more
Date: January 1, 1980
Creator: MacDonald, P. E.; Seiffert, S. L.; Martinson, Z. R.; McCardell, R. K.; Owen, D. E. & Fukuda, S. K.
Partner: UNT Libraries Government Documents Department
open access

Status Report on the Argonne Advanced Research Reactor

Description: The interim design and development status is reported. The scope of the work was limited to conceptual design studies supported by critical experiments plus heat transfer and hydraulic tests. Design criteria, facility and site, reactor, core geometry and composition, fuel elements, reflector, core and reflector support structure, reactor vessel, control and instruments, primary coolant systems, secondary coolant system, auxiliary systems, experimental facilities, building layout and constructio… more
Date: November 1, 1961
Creator: Lennox, D. H.; Barts, E. W.; Batch, R. V.; Beyer, F. C.; Jorgensen, G. L.; Kelber, C. N. et al.
Partner: UNT Libraries Government Documents Department
open access

GAS-COOLED REACTOR PROGRAM QUARTERLY PROGRESS REPORT FOR PERIOD ENDING SEPTEMBER 30, 1961

Description: Progress is reported on investigations in support of the Experimental Gas-Cooled Reactor, the Pebble-Bed Reactor Experiment, Advanced reactor design and development, test facilities, components, and materials. Topics covered include EGCR physics, EGCR performance analyses, structural investigations, EGCR component and materials development and testing, EGCR experimental facilities, PBRE physics and design studies, fueled-graphite investigations, clad fuel development, design studies of advanced… more
Date: February 1, 1962
Partner: UNT Libraries Government Documents Department
open access

Severe accident core heatup transients in modular high temperature gas-cooled reactors without operating Reactor Cavity Cooling Systems

Description: The ultimate decay heat removal system for the current Modular High Temperature Gas-Cooled reactors is a completely passive natural convection air cooling loop. This paper considers an extremely remote accident scenario, where even this passive system fails, and heat rejection is only via a layer of thermal insulation to the reactor silo structure and the surrounding soil. The results show that even in this case the peak fuel temperatures remain well within safe limits. However, vessel and conc… more
Date: January 1, 1988
Creator: Kroeger, P. G.
Partner: UNT Libraries Government Documents Department
open access

Examination of Uranium-2 w/o Zirconium Experimental Fuel Slugs Irradiated in EBR-I. Final Report-Program 6.1.11

Description: Six groups of U-2 wt% Zr fuel slugs were irradiated in the first core of the EBR-I to burnups of 0.080 to 0.189 at.% at calculated temperatures of 307 to 353 deg C. Two groups of cast specimens were found to be more dimensionally stable than four groups of wrought slugs. Of the wrought slungs, the as quenched group showed less tendency to grow than the three groups which had some annealing after quenching. Specimens at burnups of about 0.189 at.% and at 383 deg C showed the onset of swelling as… more
Date: February 1, 1962
Creator: Murphy, W. F.; Klank, A. C. & Paine, S. H.
Partner: UNT Libraries Government Documents Department
open access

Chemical Technology Division Unit Operations Section Monthly Progress Report, September 1961

Description: Nine samples of ThO/sub 2/ -UO/sub 2/ prepared as part of the solgel process development studies showed no consistent effects from small variations in several process parameters. The reaction of methane and copper oxide was studied. Engineering studies of the continuous dissolution of simulated U-Zr-Sn fuels in 6.5 M NH4F, 0.6-1.0 M NH/sub 4/NO/sub 3/, 0.1 M H/sub 2/O/sub 2/ were continued in modified 6-in.-dia. equipment. A total of 1642 kg of U from NaK bonded SRE Core I fuel rods wad dejacke… more
Date: April 1, 1962
Creator: Whatley, M. E.; Haas, P. A.; Horton, R. W.; Ryon, A. D.; Suddath, J. C. & Watson, C. D.
Partner: UNT Libraries Government Documents Department
open access

Startup and Initial Testing of SM-1 Core II With Special Components

Description: The loading operation for SM-1 Core II is described. Results of startup physics measurements (Test A-300 (Series) and fission product iodine monitoring in the primary coolant are given. The SM-1 Core II initial loading progressed satisfactorily, fulfilling the predictions of the zero power experiment performed at the Alco Criticality Facility. The initial cold clean five rod bank position was 6.53 in.; the initial hot, no xenon, five rod bank position was 9.62 in.; the initial hot, equilibrium … more
Date: February 28, 1962
Creator: Moote, F. G. & Schrader, E. W.
Partner: UNT Libraries Government Documents Department
open access

DEVELOPMENT OF A PROCESS FOR SODIUM BONDING OF EBR-II FUEL AND BLANKET ELEMENTS

Description: Procedures for assembling EBR-II fuel elements with annular sodium bonds between the uranium rods and the stainless steel claddings are outlined. The results of several meltdown and uranium-settling experiments are given. Bonding experiments were performed: furnace bonding, submerged canning, ultrasonic bonding, centrifuging, pressure pulsing, and vibratory bonding. Vibratory bonding was chosen for the production of the first EBR-II core. (D.L.C.)
Date: July 1, 1961
Creator: Sowa, E.S. & Kimont, E.L.
Partner: UNT Libraries Government Documents Department
open access

Fuel performance annual report for 1989

Description: This annual report, the twelfth in a series, provides a brief description of fuel performance during 1989 in commercial nuclear power plants and an indication of trends. Brief summaries of fuel design changes, fuel surveillance programs, fuel operating experience, fuel problems, high-burnup fuel experience, and items of general significance are provided. References to more detailed information and related US Nuclear Regulatory Commission evaluations are included.
Date: June 1, 1992
Creator: Bailey, W.J.; Berting, F.M. (Pacific Northwest Lab., Richland, WA (United States)) & Wu, S. (Nuclear Regulatory Commission, Washington, DC (United States). Div. of Systems Technology)
Partner: UNT Libraries Government Documents Department
open access

Supporting Analysis for Thermal Suitability of Fuel Elements for SM-1A Core I Loading

Description: A recommended SM-1A Core I loading chart was derived from available, metallurgically acceptable elements at the SM-1A and SM-1 sites. The derivation was based on local thermal and hydraulic considerations of minimum elementto- element coolant channel clearances. These clearances were determined from field inspection measurements of outer fuel plate spacing, as modified by analytical calculations of plate ripple growth during exposure to reactor operating thermal stresses. (auth)
Date: January 10, 1962
Creator: Brondel, J. O.
Partner: UNT Libraries Government Documents Department
open access

Chemical Technology Division, Unit Operations Section Monthly Progress Report, June 1961

Description: An interfacial viseometer was built for use in an interfacial phenomena study. Installation of a 6-in.-ID foam separation column system was completed. The dispersiondrying-sintering characteristics of six low-nitrate batches of thoria sol material were studied. The average effective porosity of the CuO pellets used for reactor helium purification was determined to be 0.0545 for H/ sub 2/ transport and 0.0526 for CO transport. In continuous Zirflex dissolution studies, no H/sub 2/O/sub 2/ decomp… more
Date: January 23, 1962
Creator: Whatley, M. E.; Haas, P. A.; Horton, R. W.; Ryon, A. D.; Suddath, J. C. & Watson, C. D.
Partner: UNT Libraries Government Documents Department
open access

BRAZING OF CERAMICS. Progress Report

Description: Brazing alloys such as 48 Ti-48 Zr-4 Be (wt%) and 49 Ti-49 Cu-2 Be (wt%) have been found to readily flow on oxide and graphite ceramics. Two demonstrati on fuel element assemblies were fabricated to illustrate the usefulness of these procedures for nuclear applications. One of these assemblies contained graphite tubes and end caps which were brazed to a molybdenum hanger. The second demonstration fuel element was composed of a compartmented aluminum oxide plate to which aluminum oxide cover pla… more
Date: November 1, 1962
Creator: Fox, C.W.
Partner: UNT Libraries Government Documents Department
open access

Calculations of thermal-reactor spent-fuel nuclide inventories and comparisons with measurements

Description: Comparisons with integral measurements have demonstrated the accuracy of CINDER codes and libraries in calculating aggregate fission-product properties, including neutron absorption, decay power, and decay spectra. CINDER calculations have, alternatively, been used to supplement measured integral data describing fission-product decay power and decay spectra. Because of the incorporation of the extensive actinide library and the use of ENDF/B-V data, it is desirable to compare the inventory of i… more
Date: January 1, 1982
Creator: Wilson, W. B.; LaBauve, R. J. & England, T. R.
Partner: UNT Libraries Government Documents Department
open access

Coolant mixing in LMFBR rod bundles and outlet plenum mixing transients. Progress report, December 1, 1979-February 29, 1980

Description: Information is presented concerning bundle geometry with wrapped and bare rods; LMFBR outlet plenum flow mixing; and theoretical determination of local temperature fields in LMFBR fuel rod bundles.
Date: January 1, 1980
Creator: Todreas, N.E.; Golay, M.W. & Wolf, L.
Partner: UNT Libraries Government Documents Department
open access

NUCLEAR SUPERHEAT PROJECT TENTH QUARTERLY PROGRESS REPORT, OCTOBER-DECEMBER 1961

Description: Results in the Nuclear Superheat Project are summarized. Topics covered include: conceptual design and program evaluation, fuel technology, materials development, experimental physics, coolant chemistry, heat transfer, mechanical development, SADE and E-SADE, and mixed spectrum superheat study. (M.C.G.)
Date: October 31, 1962
Creator: Pennington, R.T.
Partner: UNT Libraries Government Documents Department
open access

Analysis of fission product behavior in the Saclay Spitfire Loop Test SSL-1. [HTGR]

Description: The behavior of the fission metal cesium and the fission gases krypton and xenon in the Saclay Spitfire Loop SSL-1 test has been compared to that predicted using General Atomic reference data and computer code models. This is the first in a series of analyses planned in order to provide quantitative validation of HTGR fission product design methods. In this analysis, the first attempt to rigorously verify fission product design methods, the FIPERQ code was used to model the diffusion of cesium … more
Date: February 1, 1978
Creator: Jensen, D. D.; Haire, M. J. & Ballagny, A.
Partner: UNT Libraries Government Documents Department
Back to Top of Screen