ANPP CODE DEVELOPMENT PROGRAM PRESSURIZED WATER TASK QUARTERLY PROGRESS REPORT NUMBER ONE. Period Covered : August 1 through October 31, 1959

Technical Report · Sun Nov 01 04:00:00 EST 1959

OSTI ID:4207928

Olsen, T M; Welshans, L; Eicheldinger, C

The objective of the program is the development of a nuclear physical model and its calculation code for pressurized-water-cooled and -moderated reactors to predict accurately reactivity, rod positions, and power distributions at operating temperatures and reactor life from a nuclear standpoint. The BENIKO and RSP physical models for calculating buckling and reflector savings coupling coefficients for synthesis were revised, extended, and combined for improved accuracy, flexibility, and convenience under the new rame, BRAVE. Program C was modified to eliminate the assumption of the validity of using a bare core slowing-down spectrum and to correct for inelastic scattering in stainless steel, and renamed Program M. Program F was modified to calculate adjoint fluxes, to include extra lattice intervals, and to include the option of spherical geometry, and was renamed Program T. Characteristics and current status of 15 machine codes of interest are presented. The details of automating the BRAVE, M and T calculations were developed. The resulting leakage of these three programs is called SYNFAR, for " Synthesized Flux and Reactivity Calculation." The Nuclear FORTRAN Library Tape System was revised for increased efficiency and ease of operation. A FORTRAN subroutine, called CHANGE, for making efficient changes in a few of the input data of a case without reading in all of the data. was written and checked out. A FORTRAN subroutine for changing the tape unit numbers used in an IBM-704 FORTRAN program, without reassembly, was written and checked out. A scheme was developed for reading basic cross-section data from experimental curves using Benson-Lehner equipment and generating a basic crosssection data tape from this information. A FORTRAN program was outlined and started for generating a multigroup cross-section production data tape from the basic data tape. A FORTRAN program for applying the Breit-Wigner formula to fill in gaps in the experimental data was written and check out was started. Use of the Benson-Lehner equipment to read and convert the basic curvilinear data to punched card form was stanted for U/sup 235/. Critical experiments were performed on two cores. In the first core, critical buckling measurements were performed using gold, thorium, and dysprosium foils and the fuel plate scanning technique. Analytical buckling and reflector savings were compared with the experimental values. The thorium data agreed within five per cent; poor correlation was obtained using the other materials and techniques. Experiments performed on the second core were: multiplication runs to criticality; control rod calibration; fuel and boron worth determinations; and three-axes buckling measurements. Plans for the homogeneous element substitution experiments were advanced. Two types of homogeneous elements are expected to be utilized: one in which all materials in a reference element are homogenized; and a second type in which a single material, (boron or uranium), is homogenized. (See also MND-E- 2119). (auth)

• Research Organization: Martin Co. Nuclear Div., Baltimore
• NSA Number: NSA-14-010177
• OSTI ID: 4207928
• Report Number(s): MND-C-2200
• Country of Publication: United States
• Language: English

Related Subjects

ANALOG SYSTEMS
BUCKLING
COMPUTERS
CONFIGURATION
DISTRIBUTION
LIFETIME
MATHEMATICS
POWER
PRESSURE
REACTIVITY
REACTOR CORE
REACTOR TECHNOLOGY
REACTORS
REFLECTORS
URANIUM 235
WATER COOLANT
WATER MODERATOR