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in subjects: "Fuel Channels"

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Numerical accuracy of linear triangular finite elements in modeling multi-holed structures

A study has been performed to quantify the accuracy of linear triangular finite elements for modeling temperature and stress fields in structures with multiple holes. The purpose of the study was to evaluate the use of these elements for the analysis of HTGR fuel blocks, which may contain up to 325 holes. Since an accurate full scale analysis was not feasible with existing methods, a representative small scale benchmark problem containing only seven holes was selected. The finite element codes … more

Date: June 1, 1980

Creator: Sullivan, R.M. & Griffen, J.E.

Item Type: Report

Partner: UNT Libraries Government Documents Department

open access

The impact of duct-to-duct interaction on the hex duct dilation

Dilation of the hex duct is an important factor in the operational lifetime of fuel subassemblies in liquid metal fast reactors. It is caused primarily by the irradiation-enhanced creep and void swelling of the hex duct material. Excessive dilation may jeopardize subassembly removal from the core or cause a subassembly storage problem where the grid size of the storage basket is limited. Dilation of the hex duct in Experimental Breeder Reactor II (EBR-II) limits useful lifetime because of these… more

Date: January 1, 1992

Creator: Lee, M. J.; Chang, L. K.; Lahm, C. E. & Porter, D. L.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Global stiffness of hex-can assembly in a uniform force field. [LMFBR]

Two approximate constitutive equations are derived that can be used to represent the global stiffness of a hexagonal cross-section duct in a uniform force field. The first equation uses a single coefficient that can be determined from Poisson's ratio for the material and the duct geometry. This equation is useful for isothermal applications. The second equation can be used to account for temperature-varying material properties and requires that two coefficients be determined from Poisson's rati… more

Date: July 1, 1980

Creator: Ju, F.D. & Bennett, J.

Item Type: Report

Partner: UNT Libraries Government Documents Department

open access

Nuclear reactor fuel-assembly duct-tube-to-handling-socket attachment system. [LMFBR]

A reusable system is described for removably attaching the upper end of a nuclear reactor duct tube to the lower end of a nuclear reactor fuel assembly handling socket. A transition ring, fixed to the duct tube's upper end, has an interior-threaded section with a first locking hole segment. An adaptor ring, fixed to the handling socket's lower end has an outside-threaded section with a second locking hole segment. The inside and outside threaded sections match and can be joined so that the firs… more

Date: March 5, 1981

Item Type: Patent

Partner: UNT Libraries Government Documents Department

open access

DILATE: a 2-d structural program for the dilation response of hexagonal ducts

An analytical method is introduced for determining the dilation of hexagonal ducts in fast reactors. The method, which is valid for temperatures where creep is linearly dependent on stress, was implemented in a fast-running computer called DILATE. A bench-mark program is presented, which shows the results of the DILATE program in close agreement with the results of the finite element program MARC-CDC. User instructions for the DILATE program are described in detail and a listing of the program … more

Date: February 1, 1980

Creator: Chan, D. P.

Item Type: Report

Partner: UNT Libraries Government Documents Department

open access

Core support block thermal mixing test analysis report

The extent of gas mixing and pressure drop within the core support block was experimentally investigated for various geometric and height configurations. These tests were conducted by the Experimental Engineering Branch of General Atomic Company. As a result of this investigation, the core support block thermal mixing and pressure drop has been quantified. Thermal mixing and the temperature sensor accuracy can be substantially improved at the cost of higher pressure drop. A 70-degree miter angl… more

Date: May 1, 1979

Creator: Chin, E.

Item Type: Report

Partner: UNT Libraries Government Documents Department

open access

Added Mass and Damping Coefficients for Hexagonal Tube Arrays

An analytical investigation of the fluid coupling effects from an array of hexagonal cylindrical ducts undergoing harmonic oscillations is presented. A closed form solution for the velocity and pressure is obtained under a thin gap approximation for the case of moderate frequencies. From this solution, the usual viscous and inertial fluid coupling coefficients are easily obtained. These analytically derived coefficients indicate a strong dependence upon gap spacing and oscillating Reynolds numb… more

Date: August 1, 1979

Creator: Wilson, D. E.

Item Type: Report

Partner: UNT Libraries Government Documents Department

open access

Calculation of the reactivity feedback due to core-assembly bowing in LMFBRs

The nonuniformity of the temperature distribution in an LMFBR leads to differential thermal expansion of the walls of an assembly hexcan. These thermal expansion differentials cause the hexcan to distort or bow. Consequentially, the assembly experiences a spatial displacement, which results in a change in reactivity for the core. A computational model to calculate the reactivity feedback due to material displacements induced by assembly bowing effects has been developed.

Date: January 1, 1983

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

SAS3A analysis of natural convection boiling behavior in the Sodium Boiling Test Facility

An analysis of natural convection boiling behavior in the Sodium Boiling Test (SBT) Facility has been performed using the SAS3A computer code. The predictions from this analysis indicate that stable boiling can be achieved for extensive periods of time for channel powers less than 1.4 kW and indicate intermittent dryout at higher powers up to at least 1.7 kW. The results of this anaysis are in reasonable agreement with the SBT Facility test results.

Date: January 1, 1979

Creator: Klein, G A

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Use of cermet fueled nuclear reactors for direct nuclear propulsion

There has been a renewal of interest in Direct Nuclear Propulsion (DNP) because of the Air Force Forecast II recommendation for the development of the technology. Several nuclear concepts have been proposed to meet the Direct Nuclear Propulsion challenge. In this paper we will present results of an initial study of the potential of a cermet fueled nuclear system in providing the desired DNP capabilities and featuring a set of unique safety characteristics. The concept of cermet fuel for DNP app… more

Date: July 1, 1988

Creator: Bhattacharyya, S. K.; Carlson, L. W.; Kuczen, K. D.; Hanan, N. A.; Palmer, R. G.; Von Hoomissen, J. et al.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Evaluation and selection of hot channel (peaking) factors for research reactor applications

A proposed method for selecting and applying hot channel factors is presented along with some justification for these selections. The method is illustrated by example, and the sensitivity to some of the choices is examined. The uncertainty in the heat transfer coefficient is a major contributor to the reduction in thermal-hydraulic safety margins. The uncertainty introduced by the heterogeneity in the fuel is another important contributor and an area where more information may be useful in redu… more

Date: January 1, 1987

Creator: Woodruff, W.L.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Flow excursion experiments with a production reactor assembly mockup

A series of power ramp and loss-of-coolant accidents were simulated with an electrically heated mockup of a Savannah River Site production reactor assembly. The one-to-one scale mockup had full multichannel annular geometry in its heated section in addition to prototypical inlet and outlet endfitting hardware. Power levels causing void generation and flow instability in the water coolant flowing through the mockup were found under different transient and quasi-steady state test conditions. A re… more

Date: January 1, 1990

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

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

A parametric analysis of decay ratio calculations in a boiling water reactor model

The results of an investigation of the effects of several parameters on the reactivity instability of a Boiling Water Reactor (BWR) calculational model are summarized. Calculations were performed for a typical BWR operated at low flow conditions, where reactivity instabilities are more likely to occur. The parameters investigated include the axial power shape (characterized by two separate parameters), the core pressure, and operating flow. All calculations were performed using the LAPUR code w… more

Date: January 1, 1989

Creator: Blakeman, E.D. & March-Leuba, J.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Coolant mixing in LMFBR rod bundles and outlet plenum mixing transients. Progress report, March 1, 1980-May 31, 1980

Experimental and theoretical work is reported on four tasks: bundle geometry (wrapped and bare rods), subchannel geometry (bare rods), LMFBR outlet plenum flow mixing, and theoretical local temperature files in LMFBR fuel rod bundles. (DLC)

Date: January 1, 1980

Creator: Todreas, N.E.; Golay, M.W. & Wolf, L.

Item Type: Report

Partner: UNT Libraries Government Documents Department

open access

Critical heat flux experiments with a local hot patch in an internally heated annulus

Critical heat flux experiments were conducted for upflow of water in a vertical 84 inch annular flow channel, 0.303 inch heated I.D. and 0.500 inch unheated O.D. Test data were obtained at pressures from 1200 to 2000 psia, mass velocities from 0.25 x 10/sup 6/ to 2.8 x 10/sup 6/ lb/hr-ft/sup 2/ and inlet temperatures ranging from 200 to 600/sup 0/F. Three different test sections were employed with (1) axially uniform heat flux over the 84 inch length to serve as a no-hot-patch data base, (2) ax… more

Date: February 1, 1979

Creator: Mortimore, E.P. & Beus, S.G.

Item Type: Report

Partner: UNT Libraries Government Documents Department

open access

Algebraic stress model for axial flow in a bare rod-bundle

The problem of predicting transport properties for momentum and heat across the boundaries of interconnected channels has been the subject of many investigations. In the particular case of axial flow through rod-bundles, transport coefficients for channel faces aligned with rod centers are known to be considerably higher than those calculated by simple isotropic theories. And yet, it was been found that secondary flows play only a minor role in this overall transport, being turbulence highly en… more

Date: January 1, 1987

Creator: de Lemos, M.J.S.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Analysis of IFR driver fuel hot channel factors

Thermal-hydraulic uncertainty factors for Integral Fast Reactor (IFR) driver fuels have been determined based primarily on the database obtained from the predecessor fuels used in the IFR prototype, Experimental Breeder Reactor II. The uncertainty factors were applied to the channel factors (HCFs) analyses to obtain separate overall HCFs for fuel and cladding for steady-state analyses. A ``semistatistical horizontal method`` was used in the HCFs analyses. The uncertainty factor of the fuel ther… more

Date: March 1, 1994

Creator: Ku, J. Y.; Chang, L. K. & Mohr, D.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Effect of power on demand surface and flow instability for annular geometry fuel elements with upflow of coolant

Analytical studies of the effects of power on flow instability in parallel channels with upward flow of coolant, have predicted that the Ledinegg flow instability, encountered as flow is decreased for typical operating power levels, would not be experienced at low power levels. For a system in which the flow of coolant is upward, the increased buoyancy enhances flow in the channel, so that as the void increases, the overall pressure loss decreases. Under this condition, flow instability does no… more

Date: May 1, 1993

Creator: Smith, A. C.; Blake, J. E; Childerson, M. T.; Ohrn, T. R. & Privette, R. M.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Comparison of the PLTEMP code flow instability predictions with measurements made with electrically heated channels for the advanced test reactor.

When the University of Missouri Research Reactor (MURR) was designed in the 1960s the potential for fuel element burnout by a phenomenon referred to at that time as 'autocatalytic vapor binding' was of serious concern. This type of burnout was observed to occur at power levels considerably lower than those that were known to cause critical heat flux. The conversion of the MURR from HEU fuel to LEU fuel will probably require significant design changes, such as changes in coolant channel thicknes… more

Date: June 9, 2011

Creator: Feldman, E. (Nuclear Engineering Division)

Item Type: Report

Partner: UNT Libraries Government Documents Department

open access

The impact of duct-to-duct interaction on the hex duct dilation

Date: July 1, 1992

Creator: Lee, M. J.; Chang, L. K.; Lahm, C. E. & Porter, D. L.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Fatigue tests of dowel-socket systems. [HTGR]

A test program was conducted to determine the fatigue behavior of LHTGR fuel element dowel/socket systems. Two dowel/socket systems, namely, a four-dowel system and a five-dowel system, were tested to failure under shear loads applied through a fatigue test apparatus to simulate repetitive loading during a seismic event.

Date: June 15, 1976

Creator: Chiang, D. D.

Item Type: Report

Partner: UNT Libraries Government Documents Department

open access

Feasibility Study of Supercritical Light Water Cooled Reactors for Electric Power Production, Nuclear Energy Research Initiative Project 2001-001, Westinghouse Electric Co. Grant Number: DE-FG07-02SF22533, Final Report

The supercritical water-cooled reactor (SCWR) is one of the six reactor technologies selected for research and development under the Generation IV program. SCWRs are promising advanced nuclear systems because of their high thermal efficiency (i.e., about 45% versus about 33% efficiency for current Light Water Reactors [LWRs]) and considerable plant simplification. SCWRs are basically LWRs operating at higher pressure and temperatures with a direct once-through cycle. Operation above the critica… more

Date: January 1, 2005

Creator: MacDonald, Philip E.

Item Type: Report

Partner: UNT Libraries Government Documents Department

open access

Columbia University flow instability experimental program: Volume 12. Single annulus transient test program data tables: Part 1

The single annulus test program was designed to investigate the onset of flow instability in an annular geometry similar to the fuel assemblies used in the Savannah River Site production reactors. Data files were transmitted from Columbia University to Savannah River Site in a DOS compatible format. This report provides a hardcopy version of the electronic media data files.

Date: September 1993

Creator: Coutts, D. A.

Item Type: Report

Partner: UNT Libraries Government Documents Department

open access

Physical characteristics of non-fuel assembly reactor components

The primary objective of this report is to enhance the utility of the Characteristics Data Base (CDB). This has been accomplished by providing a pictorial representation of the principal non-fuel assembly (NFA) components along with a tabular summary of key information about each type of component. This report is intended for use as an adjunct to the CDB. Toward this end, the report may be used either as a complement to the detailed descriptions in the CDB, or as a stand-alone document that act… more

Date: September 1994

Creator: Hawkes, E. C.

Item Type: Report

Partner: UNT Libraries Government Documents Department