D. S. Rowe

发表

Validation of the assert subchannel code : prediction of critical heat flux in standard and nonstandard CANDU bundle geometries

M. B. Carver, J. C. Kiteley, R. Q.-N. Zhou, 1995 .

COBRA-IV-I: an interim version of COBRA for thermal-hydraulic analysis of rod bundle nuclear fuel elements and cores

R. J. Cena, C. L. Wheeler, D. S. Rowe, 1976 .

Numerical model for nuclear core thermal-hydraulic analysis

D. S. Rowe, 1973 .

COBRA IIIC: digital computer program for steady state and transient thermal-hydraulic analysis of rod bundle nuclear fuel elements

D. S. Rowe, 1973 .

COBRA-II: A DIGITAL COMPUTER PROGRAM FOR THERMAL-HYDRAULIC SUBCHANNEL ANALYSIS OF ROD BUNDLE NUCLEAR FUEL ELEMENTS.

D. S. Rowe, 1970 .

COBRA4I. Rod-Bundle and Core Thermal-Hydraulics

C. L. Wheeler, D. S. Rowe, C. W. Stewart, 1976 .

Subchannel code development and data analysis. Progress report, July 1-- December 31, 1974

D. S. Rowe, 1974 .

Transient thermal-hydraulic experiments: N-Reactor inlet connector break

W. L. Thorne, J. K. Anderson, D. S. Rowe, 1964 .

Experimental study of axial and crossflow velocity in a 7-pin wire wrapped bundle

C. L. Wheeler, D. S. Rowe, J. D. Smith, 1974 .

Experimental study of flow and pressure in rod bundle subchannels containing blockages

C. L. Wheeler, D. S. Rowe, D. E. Fitzsimmons, 1973 .

VIPRE-01. a thermal-hydraulic analysis code for reactor cores. Volume 1. Mathematical modeling. [PWR; BWR]

K. L. Basehore, J. M. Kelly, D. S. Rowe, 1983 .

EXPERIMENTAL STUDY OF MIXING BETWEEN ROD-BUNDLE FUEL-ELEMENT FLOW CHANNELS DURING BOILING.

C. W. Angle, D. S. Rowe, 1967 .

Computational analysis of two-phase flow in horizontal bundles

M. B. Carver, D. S. Rowe, A. Tapucu, 1984 .

Preliminary analysis of K-Reactor downcomer approach section pressures during temperature transients

D. S. Rowe, 1965 .

Advanced continuous fluid Eulerian computation scheme for flows with large density gradients. [PWR]

D. S. Rowe, C. W. Stewart, C. W. Stewart, 1976 .

Two-phase turbulent mixing and buoyancy drift in rod bundles

D. S. Rowe, L. Carlucci, N. Hammouda, 2004 .

Implications concerning rod bundle crossflow mixing based on measurements of turbulent flow structure

D. S. Rowe, B. Johnson, J. Knudsen, 1974 .

Turbulent flow in a model nuclear fuel rod bundle containing partial flow blockages

J. M. Bates, D. S. Rowe, A. M. Sutey, 1979 .

CROSSFLOW MIXING BETWEEN PARALLEL FLOW CHANNELS DURING BOILING. PART II. MEASUREMENT OF FLOW AND ENTHALPY IN TWO PARALLEL CHANNELS.

C. W. Angle, D. S. Rowe, 1967 .

An Analytical Model for Transient Gas Flow in Nuclear Fuel Rods

R. N. Oehlberg, D. S. Rowe, 1981 .

CROSS-FLOW MIXING BETWEEN PARALLEL FLOW CHANNELS DURING BOILING. PART I. COBRA: COMPUTER PROGRAM FOR COOLANT BOILING IN ROD ARRAYS.

D. S. Rowe, 1967 .

THERMAL HYDRAULIC SUBCHANNEL MODEL BASED ON VOID-DRIFT

D. S. Rowe, R. E. Collingham, R. B. Macduff, 1990 .

Design of a translating field-reversed configuration reactor

D. S. Rowe, G. Vlases, 1986 .

Simulation of the distribution of flow and phases in vertical and horizontal bundles using the ASSERT-4 subchannel code☆

M. B. Carver, J. C. Kiteley, D. S. Rowe, 1990 .

Effects of sleeve blockages on axial velocity and intensity of turbulence in an unheated 7 x 7 rod bundle. [PWR]

J. M. Bates, D. S. Rowe, A. M. Sutey, 1976 .

VIPRE-01: a thermal-hydraulic code for reactor cores. Volume 1: mathematical modeling (Revision 2)

D. S. Rowe, K. L. Basehore, D. S. Rowe, 1985 .