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in subjects: "Fuel Assemblies"

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SCORE-EVET: a computer code for the multidimensional transient thermal-hydraulic analysis of nuclear fuel rod arrays. [BWR; PWR]

The SCORE-EVET code was developed to study multidimensional transient fluid flow in nuclear reactor fuel rod arrays. The conservation equations used were derived by volume averaging the transient compressible three-dimensional local continuum equations in Cartesian coordinates. No assumptions associated with subchannel flow have been incorporated into the derivation of the conservation equations. In addition to the three-dimensional fluid flow equations, the SCORE-EVET code ocntains: (a) a one-… more

Date: February 1, 1978

Creator: Benedetti, R. L.; Lords, L. V. & Kiser, D. M.

Item Type: Report

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

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.

Item Type: Report

Partner: UNT Libraries Government Documents Department

open access

Design configuration of GCFR core assemblies

The current design configurations of the core assemblies for the gas-cooled fast reactor (GCFR) demonstration plant reactor core conceptual design are described. Primary emphasis is placed upon the design innovations that have been incorporated in the design of the core assemblies since the establishment of the initial design of an upflow GCFR core. A major feature of the design configurations is that they are prototypical of core assemblies for use in commercial plants; a larger number of the … more

Date: May 1, 1980

Creator: LaBar, M.P.; Lee, G.E. & Meyer, R.J.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Development of an extended-burnup Mark B design. First semi-annual progress report, July-December 1978. Report BAW-1532-1. [PWR]

The primary objective of this program is to develop and demonstrate an improved PWR fuel assembly design capable of batch average burnups of 45,000-50,000 MWd/mtU. To accomplish this, a number of technical areas must be investigated to verify acceptable extended-burnup fuel performance. This report is the first semi-annual progress report for the program, and it describes work performed during the July-December 1978 time period. Efforts during this period included the definition of a preliminar… more

Date: October 1, 1979

Item Type: Report

Partner: UNT Libraries Government Documents Department

open access

Methods and techniques of NDA (nondestructive assay)

Nondestructive assay (NDA) refers to techniques and instruments developed to measure nuclear materials in the many forms in which they occur throughout the fuel cycle. These techniques were first developed to support nuclear safeguards inspections and nuclear material accountability; they are also used extensively for process and quality control. Most accountability measurements are based on analytical chemistry and require that a sample be drawn and analyzed destructively. Destructive analysis… more

Date: January 1, 1988

Creator: Reilly, T. D.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Analysis of the thorium axial blanket experiments in the PROTEUS reactor

An extensive program of reactor physics experiments in GCFR fuel pin lattices has been completed recently at the PROTEUS critical facility located at EIR laboratory in Switzerland. The PROTEUS reactor consists of a central test zone surrounded by a uranium buffer and thermal driver region. The test lattices included a PuO/sub 2//UO/sub 2/ fuel region with internal and axial blankets of UO/sub 2/, ThO/sub 2/, and thorium metal. Detailed analysis of the thorium-bearing lattices has been performed… more

Date: January 1, 1980

Creator: White, J.R.; Ingersoll, D.T. & Schmocker, U.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Design of a full scale model fuel assembly for full power production reactor flow excursion experiments

A novel full scale production reactor fuel assembly model was designed and built to study thermal-hydraulic effects of postulated Savannah River Site (SRS) nuclear reactor accidents. The electrically heated model was constructed to simulate the unique annular concentric tube geometry of fuel assemblies in SRS nuclear production reactors. Several major design challenges were overcome in order to produce the prototypic geometry and thermal-hydraulic conditions. The two concentric heater tubes (to… more

Date: January 1, 1990

Creator: Nash, C. A. (Westinghouse Savannah River Co., Aiken, SC (United States)); Blake, J. E. & Rush, G. C. (Babcock and Wilcox Co., Alliance, OH (United States))

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Turbulent interchange in triangular array bare rod bundles

Bulk mixing coefficients were measured for single plane water flow in a simulated rod bundle with a pitch to diameter ratio of 1.10. A tracer technique employing Rhodamine B as the tracer and measuring fluorescence was used. Isokinetic sampling was achieved by using a pressure balance method. The results were corrected for both entrance effects and diversion crossflows. The results showed a change in Reynolds number behavior as the laminar sublayer began to ''choke'' the turbulent mixing. This,… more

Date: July 1, 1977

Creator: Kelly, J.M. & Todreas, N.

Item Type: Report

Partner: UNT Libraries Government Documents Department

open access

RODCON: a finite difference heat conduction computer code in cylindrical coordinates

RODCON, a finite difference computer code, was developed to calculate the internal temperature distribution of the fuel rod simulator (FRS) for the Core Flow Test Loop (CFTL). RODCON solves the implicit, time-dependent forward-differencing heat transfer equation in 2-dimensional (Rtheta) cylindrical coordinates at an axial plane with user specified radial material zones and surface conditions at the FRS periphery. Symmetry of the boundary conditions of coolant bulk temperatures and film coeffic… more

Date: September 16, 1980

Creator: Conklin, J. C.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

GCFR core thermal-hydralic design

The approach for developing the thermal-hydraulic core assembly designs for the gas-cooled fast reactor (GCFR) is reviewed, and key considerations for improving the core performance at all power and flow conditions are discussed. It is shown how the thermal-hydraulic core assembly designs evolve from evaluations of plant size, material limitations, safety criteria, and structural performance considerations.

Date: May 1, 1980

Creator: Schleuter, G.; Baxi, C.B. & Bennett, F.O.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

US GCFR demonstration plant design

A general description of the US GCFR demonstration plant conceptual design is given to provide a context for more detailed papers to follow. The parameters selected for use in the design are presented and the basis for parameter selection is discussed. Nuclear steam supply system (NSSS) and balance of plant (BOP) component arrangements and systems are briefly discussed.

Date: May 1, 1980

Creator: Hunt, P. S. & Snyder, H. J.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Hydrodynamic behavior of a bare rod bundle. [LMFBR]

The temperature distribution within the rod bundle of a nuclear reactor is of major importance in nuclear reactor design. However temperature information presupposes knowledge of the hydrodynamic behavior of the coolant which is the most difficult part of the problem due to complexity of the turbulence phenomena. In the present work a 2-equation turbulence model--a strong candidate for analyzing actual three dimensional turbulent flows--has been used to predict fully developed flow of infinite … more

Date: June 1, 1977

Creator: Bartzis, J.G. & Todreas, N.E.

Item Type: Report

Partner: UNT Libraries Government Documents Department

open access

Test PCM-5 rod bowing and bow direction reversal. [PWR]

Test PCM-5 was the first bundle test in the PCM Test Series being conducted by the Thermal Fuels Behavior Program of EG and G Idaho, Inc. as part of the Nuclear Regulatory Commission's Fuel Behavior Program. The experiment was performed in the Power Burst Facility (PBF) reactor at the Idaho National Engineering Laboratory. The bundle consisted of nine previously unirradiated PWR-type fuel rods, arranged in a 3 x 3 array within a square cross section flow shroud, with rod-to-rod spacing typical … more

Date: January 1, 1980

Creator: Kerwin, D. K.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Time series analysis of reactor thermocouple data. [LMFBR]

Time-series analysis techniques are applied to nuclear reactor thermocouple data to investigate coolant temperatures measured within the fueled test assembly. The coolant temperature distribution within a fuel assembly affects the length of time a fuel assembly may be operated in a power reactor and, therefore, is an important economic consideration in the design of reactor fuel systems. Frequency-domain signal conditioning techniques were used to reveal the smoothly varying thermocouple signal… more

Date: April 10, 1980

Creator: Devary, J.L.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Natural convection heat transfer analysis of ATR fuel elements

Natural convection air cooling of the Advanced Test Reactor (ATR) fuel assemblies is analyzed to determine the level of decay heat that can be removed without exceeding the melting temperature of the fuel. The study was conducted to assist in the level 2 PRA analysis of a hypothetical ATR water canal draining accident. The heat transfer process is characterized by a very low Rayleigh number (Ra {approx} 10{sup {minus}5}) and a high temperature ratio. Since neither data nor analytical models wer… more

Date: May 1, 1992

Creator: Langerman, M.A.

Item Type: Report

Partner: UNT Libraries Government Documents Department

open access

Low enrichment fuel conversion for Iowa State University

Work during the reported period was centered primarily in preparation for receiving the LEU fuel and the shipping of the HEU fuel. This included development of procedures and tools for the disassembly process. During the period we held many practice sessions applying these tools and practices to a dummy fuel assembly. The LEU fuel was received on April 10, 1991 and the reactor was shut down on May 3, 1991 for refueling. The twelve HEU fuel assemblies in the UTR-10 reactor core were removed and … more

Date: August 1, 1991

Creator: Rohach, A.F. & Hendrickson, R.A.

Item Type: Report

Partner: UNT Libraries Government Documents Department

open access

Disposal of spent fuel

Based on preliminary analyses, spent fuel assemblies are an acceptable form for waste disposal. The following studies appear necessary to bring our knowledge of spent fuel as a final disposal form to a level comparable with that of the solidified wastes from reprocessing: 1. A complete systems analysis is needed of spent fuel disposition from reactor discharge to final isolation in a repository. 2. Since it appears desirable to encase the spent fuel assembly in a metal canister, candidate mater… more

Date: January 1, 1978

Creator: Blomeke, J. O.; Ferguson, D. E. & Croff, A. G.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Investigation of stainless steel clad fuel rod failures and fuel performance in the Connecticut Yankee Reactor. Final report

Significant levels of fuel rod failures were observed in the batch 8 fuel assemblies of the Connecticut Yankee reactor. Failure of 304 stainless steel cladding in a PWR environment was not expected. Therefore a detailed poolside and hot cell examination program was conducted to determine the cause of failure and identify differences between batch 8 fuel and previous batches which had operated without failures. Hot cell work conducted consisted of detailed nondestructive and destructive examinat… more

Date: November 1, 1981

Creator: Pasupathi, V. & Klingensmith, R. W.

Item Type: Report

Partner: UNT Libraries Government Documents Department

open access

Fabrication development of full-sized components for GCFR core assemblies

This paper presents the status of the development of full-sized components for gas-cooled fast reactor (GCFR) core assemblies. Methods for ribbing of the fuel rod cladding, fabrication of grid spacers of two different designs, drawing of assembly flow ducts, and fabrication of fission gas collection manifolds by several methods are discussed.

Date: May 1980

Creator: Lindgren, J. R.; Flynn, P. W. & Foster, L. C.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

FLOWTRAN-TF code description

FLOWTRAN-TF is a two-component (air-water), two-phase thermal-hydraulics code designed for performing accident analyses of SRS reactor fuel assemblies during the Emergency Cooling System (ECS) phase of a Double Ended Guillotine Break (DEGB) Loss of Coolant Accident (LOCA). This report provides a brief description of the physical models in the version of FLOWTRAN-TF used to compute the Recommended K-Reactor Restart ECS Power Limit. This document is viewed as an interim report and should ultimate… more

Date: December 1, 1990

Creator: Flach, G.P. (ed.)

Item Type: Report

Partner: UNT Libraries Government Documents Department

open access

Irradiated fuel inspection in a storage pond with no fuel movement and an uncollimated detector

It is assumed that the assemblies are not to be moved during the inspection in order to reduce the time required. The task is simplified further when the detector is well collimated. If the collimation is insufficient to respond to only one assembly from among an array of fuel assemblies, then an unfolding process is needed. A formulation of the detector's behavior is given followed by an unfolding process to yield the detector's response to each assembly as if it were isolated. Some simulated … more

Date: July 1, 1980

Creator: Rinard, P. M.

Item Type: Report

Partner: UNT Libraries Government Documents Department

open access

Historical overview of domestic spent fuel shipments: Update

This report presents available historic data on most commercial and research reactor spent fuel shipments in the United States from 1964 through 1989. Data include sources of the spent fuel shipped, types of shipping casks used, number of fuel assemblies shipped, and number of shipments made. This report also addresses the shipment of spent research reactor fuel. These shipments have not been documented as well as commercial power reactor spent fuel shipment activity. Available data indicate th… more

Date: July 1, 1991

Item Type: Report

Partner: UNT Libraries Government Documents Department

open access

Analysis of fuel relocation for the NRC/PNL Halden assemblies IFA-431, IFA-432, and IFA-513

The effects of the thermally-induced cracking and subsequent relocation of UO/sub 2/ fuel pellets on the thermal and mechanical behavior of light-water reactor fuel rods during irradiation are quantified in this report. Data from the Nuclear Regulatory Commission/Pacific Northwest Laboratory Halden experiments on instrumented fuel assemblies (IFA) IFA-431, IFA-432, and IFA-513 are analyzed. Beginning-of-life in-reactor measurements of fuel center temperatures, linear heat ratings, and cladding … more

Date: April 1, 1980

Creator: Williford, R. E.; Mohr, C. L.; Lanning, D. D.; Cunningham, M. E.; Rausch, W. N. & Bradley, E. R.

Item Type: Report

Partner: UNT Libraries Government Documents Department

open access

GCFR development status report

This report describes the major design features of the gas-cooled fast breeder reactor being developed in the United States principally at General Atomic Company. The report gives the general design strategy and highlights the design features of the reactor core and the nuclear steam supply components. It describes the design results on plant safety and licensing.

Date: May 1, 1980

Item Type: Report

Partner: UNT Libraries Government Documents Department