RSIC CODE PACKAGE PSR-537
1. NAME AND TITLE
TRISTAN-IJS: Steady-State Axial Temperature and Flow Velocity in Triga Channel.
Jozef Stefan Institute, Reactor Physics Division and Agencija za radioaktivne odpadke, Ljubljana, Slovenia, through the OECD Nuclear Energy Agency Data Bank, Issy‑les-Moulineaux, France.
3. CODING LANGUAGE AND COMPUTER
Fortran 77; Pentium Personal Computers (P00537IBMPC00).
4. NATURE OF PROBLEM SOLVED
TRISTAN-IJS is a computer program for calculating steady-state axial temperature distribution and flow velocity through a vertical coolant channel in low power TRIGA reactor core, cooled by natural circulation. It is designed for steady-state thermohydraulic analysis of TRIGA research reactors operating at a low power level of 1-2 MW.
5. METHOD OF SOLUTION
Stable operating point is calculated by iterations according to Bernoulli equation.
6. RESTRICTIONS OR LIMITATIONS
7. TYPICAL RUNNING TIME
Less than 1 minute in the sample case on Windows PC.
8. COMPUTER HARDWARE REQUIREMENTS
The included executable was tested on Intel Windows personal computers under the Windows XP SP2 and Vista operating systems.
9. COMPUTER SOFTWARE REQUIREMENTS
At RSICC, TRISTAN-IJS was compiled on a Pentium 4 running Windows XP SP2 with the Intel 9.0 ifort compiler for 32-bit Windows applications. This executable and the Fortran source code are included in the package.
RSICC obtained TRISTAN-IJS through the NEADB, where it is distributed as IAEA1337/01. The code was originally compiled under Windows NT using Lahey F77L-EM/32 Version 5.20, which is incompatible with WindowsXP. RSICC compiled the source and replaced the original executable with the Intel 9.0 executable.
10.a included in documentation:
I. Mele and B. Zefran, “TRISTAN, A Computer Program for Calculating Natural Convection Flow Parameters in TRIGA Core, Program Manual,” IJS-DP-6548 (December 1992).
10.b background references:
Varnostono porocilo za reaktor TRIGA Mark II v Podgorici, Revizija 3, IJS-DP-5832, Junij 1992
I. Mele: Izracun porazdelitve moci in temperature v reaktorju TRIGA med tranzientom, Magisrsko delo, Univerza v Mariboru, Maj 1991
Safety Analysis Report for the 3000 kW forced-Flow TRIGA Mark II Reactor for the Bangladesh Atomic Energy Commission, GA E-117-990 July 1981
Physics Handbook, American Institute of Physics Handbook, McGraw-Hill Book Co., 1982
M. Opresnik, M. Opara: Termodinamicne tabele in diagrami, Fakulteta za strojnistvo, Ljubljana 1974
L.S. Tong, J. Weisman: Thermal Analysis for Pressurized Water Reactors, American Nuclear Society, 1970
M.M. El-Wakil: Nuclear Heat Transport, An Intext Publisher, London, 1971
11. CONTENTS OF CODE PACKAGE
The package is transmitted on a CD that includes the referenced document cited in 10.a above and a WinZIP file which contains the Fortran source files, PC executables, data files, and test case input and outputs.
12. DATE OF ABSTRACT
KEYWORDS: HEAT TRANSFER; REACTOR PHYSICS; THERMAL HYDRAULICS