RSICC CODE PACKAGE PSR-422
1. NAME AND TITLE
RELAP3B/MOD110: Reactor System Transient Code.
Brookhaven National Laboratory, Upton, New York through the Energy Science and Technology Software Center, Oak Ridge, Tennessee.
3. CODING LANGUAGE AND COMPUTER
FORTRAN IV; CDC 7600 (P00422C760000).
4. NATURE OF PROBLEM SOLVED
RELAP3B describes the behavior of water-cooled nuclear reactors during postulated accidents or power transients, such as large reactivity excursions, coolant losses or pump failures. The program calculates flows, mass and energy inventories, pressures, temperatures, and steam qualities along with variables associated with reactor power, reactor heat transfer, or control systems. Its versatility allows one to describe simple hydraulic systems as well as complex reactor systems.
5. METHOD OF SOLUTION
RELAP3B obtains a time-dependent thermal and hydraulic description of a rector by integrating a set of differential equations subject to certain algebraic relationships. These equations and relationships are presented in the referenced document.
6. RESTRICTIONS OR LIMITATIONS
RELAP3B allows a maximum of 75 volumes connected by a maximum of 100 junctions, with no restrictions as to the order of these connections. However, these maxima may be increased to far greater limits on a larger computer.
7. TYPICAL RUNNING TIME
NESC executed the sample problem in less than 1 second of CP time on a CDC7600.
8. COMPUTER HARDWARE REQUIREMENTS
153,000 (octal) words of central memory and 154,000 (octal) words of large core memory are required on a CDC 7600.
9. COMPUTER SOFTWARE REQUIREMENTS
Thermal Reactor Safety Division Staff, "User's Manual for RELAP3B-MOD110, A Reactor System Transient Code," BNL-NUREG-22011 (December 1977, revised July 1980).
11. CONTENTS OF CODE PACKAGE
Included in the package are the referenced document and one 3.5" diskette containing a self-extracting compressed DOS file which includes a readme file, Fortran IV source file, and an input file.
12. DATE OF ABSTRACT
KEYWORDS: FLUID DYNAMICS; HEAT TRANSFER; LOCA; REACTOR SAFETY; THERMAL HYDRAULICS