THYDE-B1/MOD2: Computer Code for the Analysis of Small-Break Loss-of-Coolant Accident of Boiling Water Reactors.
Department of Nuclear Safety Evaluation Japan Atomic Energy Research Institute, Tokai-mura, Naka-gun, Ibaraki-Ken, Japan, through the OECD NEA Data Bank, Issy-les-Moulineaux, France.
FACOM M-200; FORTRAN77 (P00553FM20000). (NEADB ID: NEA-0778/003).
THYDE-B1/MOD2 is a computer code to analyze thermo-hydraulic transients of the reactor cooling system of a BWR, mainly during a small-break loss-of coolant accident (SB-LOCA) with special emphasis on the behavior of pressure and mixture level in the pressure vessel. THYDE-B1/MOD2 uses the steam table subroutines of RELAP-4, so the same steam table dataset as used in RELAP-4 must be supplied.
The coolant behavior in THYDE-B1/MOD2 is simulated with a volume-and-junction method based on the assumptions of thermal equilibrium and homogeneous conditions for two-phase flow. A characteristic feature of this code is a three-region representation of the state of the coolant in a control volume in which three regions (i.e., sub cooled liquid, saturated mixture and saturated steam) are allowed to exist. The regions are separated by moving boundaries tracked by mass and energy balances in each region. The pressure vessel is represented by two volumes with three regions—one for the inside of the shroud and the other for the outside. Other portions of the system are treated with the homogeneous model. This method, although it seems to be very simple, has been verified to be adequate for cases of BWR SB-LOCAs in which the hydraulic transient is relatively slow and the cooling of the core strongly depends on the mixture level behavior in the vessel. In order to simulate the reactor system behavior, THYDE-B1/MOD2 is provided with analytical models for reactor kinetics, heat generation and conduction in fuel rods and structures, heat transfer between coolant and solid surfaces, coolant injection systems, breaks and discharge systems, jet pumps, recirculation pumps, etc.
Because this code has been developed and verified for relatively slow transients of BWR, users should be careful to apply this code to any rapid transients such as an early phase of a large break LOCA.
The running time depends on the number of time steps and the complexity of the nodalization of the system to be analyzed. A rough estimate of CPU time requirements is about 6ms/volume.step on FACOM M-200. As a BWR is usually modeled by 4 to 6 volumes with time step size of 3 to 10 milliseconds, the total CPU time for a transient of 500 seconds is about 0.5 to 2 hours.
The code was developed on a FACOM M-200 and was tested on an IBM 3090. The code was developed on mainframe computers and has not been ported to Unix or Windows operating systems. Storage requirement for the test case on an IBM 3090 computer is 988K bytes.
A Fortran compiler is required. No executables are included in the distribution. The results of a THYDE-B1/MOD2 calculation were plotted by the SPLPACK-1 program generally designed for transient analysis codes and used the CALCOMP plotting system. The NEADB first released this package in 1989; it was not modified when RSICC released it in 2009.
10.a Included in the document
K. Muramatsu and M. Akimoto: THYDE-B1/MOD1: A Computer Code for Analysis of Small-Break Loss-of-Coolant Accidents of Boiling Water Reactors, JAERI-M 82-126, Japan Atomic Energy Research Institute report (August 1982).
Library Functions and Subroutines Used in THYDE-B1 and SPLPLOT-1.
10.b Background information
K. Muramatsu et al., Users Manual for SPLPACK-1 A Program Package for Plotting and Editing of Experimental and Analytical Data of Various Transient Systems, JAERI-M 83-166 (1983).
K. Muramatsu et al., Analysis of ROSA-III Small-Break LOCA Experiment RUN 804 by THYDE-B1 Computer Code, JAERI-M 9413 (1981).
T. Simizu, Verification of LOCA/ECCS Analysis Code ALARM-B2 and THYDE-B1 by Comparison with RELAP4/MOD6/U4/J3, JAERI-M 82-094 (1982).
THYDE-B1/MOD2 SPLPLOT-1: Plotting Output
The package contains the documents cited in Section 10.a and one self-extracting compressed DOS file on a single CD. This file includes Fortran and assembler source code, steam table data and sample problem input and output.
KEYWORDS: LOCA; HEAT TRANSFER; THERMAL HYDRAULICS; NUCLEAR SAFETY