CSNI1029: REWET, PWR LOCA accidents experiments

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CSNI1029 REWET. (Abstract last modified 01-JAN-1989)

REWET, PWR LOCA accidents experiments

Abstract Table of Contents:

1. NAME OF EXPERIMENT - REWET.

2. COMPUTER FOR WHICH PROGRAM IS DESIGNED AND OTHER MACHINE VERSION PACKAGES AVAILABLE -

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Explanation of the status codes.

Program-name Package-ID Status REWET CSNI1029/01 Report

Machines used: Package-ID Orig.Computer Test Computer CSNI1029/01 Many Computers

3. DESCRIPTION OF TEST FACILITY - The REWET-II facility was designed for the investigation of the reflooding phase of a LOCA. The main design principle is the accurate simulation of the rod bundle geometry and the primary system elevations. This is necessary in order to have the correct flow channels and hydrostatic pressures for the reflooding process. The reactor vessel is simulated by a stainless steel U-tube construction consisting of downcomer, lower plenum, core and upper plenum. The primary loops contain a pipe simulating the broken loop and a connection line between the upper plenum and the downcomer simulating five intact loops. The containment is simulated by a pressure vessel (not in scale). Steam generators and primary pumps are simulated with flow resistances. The ECC-water can be injected to the downcomer and/or to the upper plenum by a pump or from an accumulator.
All the elevation in the reactor vessel simulator are scaled to 1:1 (except the reactor upper head). The scale of the volumes and flow areas is 1:2333 referring to the number of the fuel rod simulators in the facility and the fuel rods in the reference reactor.
The rod bundle is either in a hexagonal shroud, the inside distance of the opposite walls is 54.3 mm and the wall thickness 2 mm, or in a round shroud, the inner diameter is 66.0 mm and the wall thickness 2 mm.
The simulation of the fuel-rod bundle consists of 19 indirectly electrically heated simulator rods. The heating coils are inside stainless steel claddings in magnesium oxide insulation. The heated length, the outer diameter and the lattice pitch of the fuel-rod simulators as well as the number (= 10) and construction of the rod bundle spacers are the same as in the reference reactor. The upper ends of the rods are attached to the upper tie plate.

4. DESCRIPTION OF TEST ¿
Pressurized water reactors in use in Finland reactors have certain unique features which make them different from most other PWR designs. The 6 horizontal steam generators differ from the commonly used vertical steam generators. The fuel rods are arranged hexagonally, and the fuel bundles of 127 rods are in BWR-like channels. The lower plenum volume is large because of the fuel followers of the control rods. In particular, the ECC water is injected directly into the upper plenum and the downcomer of the reactor vessel.
The main measurements in the experiments are coolant and rod cladding temperatures with thermocouples at different radial and axial locations. System pressure, pressure difference along the test section (core simulator), coolant flow rates and heating power are also measured. The data acquisition system consists of measurement and control processor, digital voltmeter and desk-top computer. The maximum speed to scan all 96 data channels during an experiment is once per second. With this speed it takes 10 minutes to collect a maximum of 58,800 readouts.
Three different fuel rod simulator bundles have been used in experiments, two bundles with nine-step chopped cosine power distribution and one with uniform linear power.
The simulator rods have been compared with the VVER-440 fuel rods by performing reflood calculations using the FLOOD4 computer code developed at the INEL, USA.

6. PHENOMENA TESTED - Improve the understanding of the basic phenomena of the accident situations and to provide experimental data for the development and verification of the LOCA and SBLOCA codes aimed for analyzing pressurized water reactors in use in Finland.

9. STATUS
CSNI1029/01: 01-JAN-1989 Report Only

10. REFERENCES -
CSNI1029/01:
- Timo Kervinen, Heikki Purhonen and Timo Haapalehto:
REWET-II and REWET-III facilities for PWR LOCA experiments
Research note 929 (January 1989)

11. TEST DESIGNATION - PWR LOCA accidents

12. PROGRAMMING LANGUAGE -CSNI1029/01:

15. ESTABLISHMENT -
VTT
Nuclear Engineering Laboratory
Espoo, Finland

16. MATERIAL AVAILABLE -CSNI1029/01:
Documentation in electronic format

17. CATEGORIES -

- Y. Integral Experiments Data, Databases, Benchmarks

Keywords: LOSS-OF-COOLANT ACCIDENT, PWR REACTORS