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in subjects: "Reactor 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

Annular Flow Distribution test

This report documents the Babcock and Wilcox (B W) Annular Flow Distribution testing for the Savannah River Laboratory (SRL). The objective of the Annular Flow Distribution Test Program is to characterize the flow distribution between annular coolant channels for the Mark-22 fuel assembly with the bottom fitting insert (BFI) in place. Flow rate measurements for each annular channel were obtained by establishing hydraulic similarity'' between an instrumented fuel assembly with the BFI removed an… more

Date: December 1, 1990

Creator: Kielpinski, A.L. (ed.) (Westinghouse Savannah River Co., Aiken, SC (United States)); Childerson, M.T.; Knoll, K.E.; Manolescu, M.I. & Reed, M.J. (Babcock and Wilcox Co., Alliance, OH (United States). Research Center)

Item Type: Report

Partner: UNT Libraries Government Documents Department

open access

Experimental facilities

We have completed an engineering feasibility study of a major modification of the HFIR facility and are now beginning a similar study of an entirely new facility. The design of the reactor itself is common to both options. In this paper, a general description of the modified HFIR is presented with some indications of the additional facilities that might be available in an entirely new facility.

Date: January 1, 1984

Creator: Moon, R.M.

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

Operation of a nuclear test gage at low multiplications

The Nuclear Test Gage (NTG) at the Savannah River Plant is a subcritical multiplying facility (low k) with H/sub 2/O moderator and 2.54-cm-diameter fuel slugs of 5 wt percent /sup 235/U in aluminum alloy at a 4.285-cm triangular pitch. The core of the facility is 61-cm long with a normal diameter of 27 cm. The NTG is used for quality control of reactor components, such as /sup 235/U-Al fuel tubes, Li--Al target tubes, control and safety rods, and miscellaneous special irradiation elements. A co… more

Date: January 1, 1977

Creator: Baumann, N.P.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Performance of conjugate gradient-like algorithms in transient two-phase subchannel analysis

A transient, drift-flux subchannel analysis code (SWIRL) has been created for the development and evaluation of algorithms for the solution of weakly three-dimensional fluid flow problems. Spatial discretization on a staggered grid, semi-implicit temporal discretization, and algebraic reduction of the conservation equations of mass, energy, and momentum result in nonsymmetric block-tridiagonal linear systems of equations that must be solved for the pressure distribution at each time step of a t… more

Date: January 1, 1991

Creator: Turner, J.A. (Los Alamos National Lab., NM (USA)) & Doster, M.J. (North Carolina State Univ., Raleigh, NC (USA). Dept. of Nuclear Engineering)

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Heat transfer and pressure drop in an annular channel with downflow

The onset of a flow instability (OFI) determines the minimum flow rate for cooling in the flow channels of a nuclear fuel assembly. A test facility was constructed with full-scale models (length and diameter) of annular flow channels incorporating many instruments to measure heat transfer and pressure drop with downflow in the annulus. Tests were performed both with and without axial centering ribs at prototypical values of pressure, flow rate and uniform wall heat flux. The axial ribs have the… more

Date: January 1, 1992

Creator: Dolan, F.X.; Crowley, C.J. (Creare, Inc., Hanover, NH (United States)) & Qureshi, Z.H. (Westinghouse Savannah River Co., Aiken, SC (United States))

Item Type: Article

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

Design inputs document: Boiling behavior during flow instability

The coolant flow in a nuclear reactor core under normal operating conditions is kept as a subcooled liquid. This coolant is evenly distributed throughout the multiple flow channels with a uniform pressure profile across each coolant flow channel. If the coolant flow is reduced, the flow through individual channels will also decrease. A decrease in coolant flow will result in higher coolant temperatures if the heat flux is not reduced. When flow is significantly decreased, localized boiling may … more

Date: January 1, 1991

Creator: Coutts, D. A.

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

Critical heat flux for free convection boiling in thin rectangular channels

A review of the experimental data on free convection boiling critical heat flux (CHF) in vertical rectangular channels reveals three mechanisms of burnout. They are the pool boiling limit, the circulation limit, and the flooding limit associated with a transition in flow regime from churn to annular flow. The dominance of a particular mechanism depends on the dimensions of the channel. Analytical models were developed for each free convection boiling limit. Limited agreement with data is observ… more

Date: January 1, 1991

Creator: Cheng, Lap Y. & Tichler, P.R.

Item Type: Article

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

PADLOC: a one-dimensional computer program for calculating coolant and plateout fission-product concentrations. Part 2

The behavior of some of the prominent fission products along their convection pathways is dominated by the interaction of other species with them. This gave rise to the development of a plateout code capable of analyzing coupled species effects. The single species plateout computer program PADLOC is described in Part I of this report. The present Part II is concerned with the extension of PADLOC to MULTI*PADLOK, a multiple species version of PADLOC. MULTI*PADLOC is designed to analyze the time … more

Date: September 1, 1981

Creator: Hudritsch, W.W.

Item Type: Report

Partner: UNT Libraries Government Documents Department

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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

Counter-current flow limited CHF in thin rectangular channels

An analytical expression for counter-current-flow-limitation (CCFL) was used to predict critical heat flux (CHF) for downward flow in thin vertical rectangular channels which are prototypes of coolant channels in test and research nuclear reactors. Top flooding is the mechanism for counter-current flow limited CHF. The CCFL correlation also was used to determine the circulation and flooding-limited CHF. Good agreements were observed between the period the model predictions and data on the CHF f… more

Date: January 1, 1990

Creator: Cheng, L. Y.

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

Experience with non-fuel-bearing components in LWR (light-water reactor) fuel systems

Many non-fuel-bearing components are so closely associated with the spent fuel assemblies that their integrity and behavior must be taken into consideration with the fuel assemblies, when handling spent fuel of planning waste management activities. Presented herein is some of the experience that has been gained over the past two decades from non-fuel-bearing components in light-water reactors (LWRs), both pressurized-water reactors (PWRs) and boiling-water reactors (BWRs). Among the most import… more

Date: December 1, 1990

Creator: Bailey, W.J. & Berting, F.M.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Modeling of thermal hydraulic behavior and fission product releases in degraded cores

When core material reaches melting conditions severe degradation of the core geometry occurs. Data available on the core behavior in a severely degraded state suggest that extensive blockage of the flow channels would occur. If a sufficient bypass is available for the gas flow, such as in the LOFT LP-FP-2 test, severe retardation of the hydrogen and fission product sources from the degraded channel is suggested from the available data. This phenomena is expected to occur in an LWR core and shou… more

Date: January 1, 1988

Creator: Sharon, A.; Ellison, P.G.; Henry, R.E.; Kenton, M.A. & Hammersley, R.J. (Fauske and Associates, Inc., Burr Ridge, IL (USA))

Item Type: Article

Partner: UNT Libraries Government Documents Department