Validation of the thermohydraulic code ATHLET
E. Krepper and F. Schäfer
1. Participation in the external verification group of the ATHLET code
An important component of nuclear safety research is the analysis of abnormal transients and accident scenarios in nuclear power plants (NPP). Such analyses are carried out with complex thermohydraulic computer codes, like ATHLET. ATHLET is a one-dimensional nodal thermohydraulic computer code, developed by GRS (Gesellschaft für Anlagen- und Reaktorsicherheit) for the analysis of anticipated and abnormal plant transients, small and intermediate leaks as well as large breaks in light water reactors. For the calculations the primary and secondary coolant system of a NPP is described by a set of one-dimensional control volumes. In each control volume the conservation equations for energy, mass and momentum are applied. Additionally different physical models (drift-flux model, discharge model, ...) have to be used to solve the resulting equation system. For the application to real NPPs the code must be validated through the comparison of calculated results with experimental data. The experimental data are obtained from various test facilities, modelling the primary and secondary circuit of the real plant.
1.1 Post-test analyses of the experiments 5.2c and 9.3 at the integral test facility BETHSY
In the framework of the external validation of the thermohydraulic code ATHLET Mod 1.1 Cycle D, which has been developed by the GRS, post test analyses of two experiments were done, which were performed at the french integral test facility BETHSY.
The facility is a 1:100 scaled thermohydraulic model of a 900 MW (el.) pressurized water reactor (FRAMATOME). The test facility is designed to investigate various accident scenarios and to provide an experimental data base for code validation and for the verification of accident management measures. The three identical loops enable the simulation of asymmetric loop behavior. In the test facility, the primary circuit is modeled with a volume scaling ratio of 1:100 retaining the original heights. Each primary loop is equipped with a vertical steam generator.
BETHSY test facility
The BETHSY experiment 5.2c investigates the accident procedures in case of a total loss of feedwater at the steam generator secondary side. In such an accident the emergency cooling of the reactor core with primary bleed and feed, the behaviour of the steam generators in case of dry out and the long time behaviour of the test facility are special subjects of interest. During the experiment the high pressure injection system, the hydroaccumulators and the low pressure injection system were available.
During test 9.3 the consequences of a steam generator U-tube rupture with failure of the high pressure injection and of the auxiliary feedwater supply were investigated. As accident management measures, the depressurization of the secondary sides, first of the two intact steam generators, then of the damaged steam generator and finally the primary depressurization by opening of the pressurizer valve were performed.
More about the ATHLET-calculations:
1.2 Post-test analyses of experiments at the integral test facility CCTF
The CCTF test facility is a 1:25 scaled model of a 1000 MW (el.) pressurized water reactor and is located in Japan. The test facility has a full hight core section, four primary loops and two steam generators. The facility is mainly designed to provide detailed information about two-phase flow phenomena in the reactor core, downcomer an upper plenum during the refill and reflood phases of a hypothetical "large break loss of coolant accident" (LB-LOCA).
CCTF test facility
Two experiments at the CCTF test facility were analyzed with the code ATHLET. The CCTF tests C2-19 and C2-4 are characterized by a double-end break in the cold leg and different modes of emergency cooling injection into the hot and cold legs. In both tests the development of the quench front depends on the radial power distribution in the core. In ATHLET the three-dimensional behaviour of the quench front can be simulated by a two-channel representation of the reactor core.
More about the ATHLET-calculations:
2. ATHLET calculations of several experiments at test facilities for VVER-type reactors
Test facilities - overview:
- PACTEL (VVER-440, Lappeenranta, Finland)
- PMK-2 (VVER-440, Budapest, Hungary)
- ISB-VVER (VVER-1000, Moscow, Russia)
The PACTEL test facility was constructed to investigate accident scenarios with small and intermediate leaks and also to study the natural circulation behaviour of the primary circuit. The facility models the six loops of the real NPP by three symmetric loops with a volume scaling ratio of 1:305. The Research Center Rossendorf was a participant at the International Standard Problem ISP-33, a natural circulation experiment with stepwise reduced primary coolant inventory.
The PMK-2 test facility is a full-pressure, volume-scaled model of the Paks NPP. The facility was constructed by the KFKI Atomic Energy Research Institute Budapest and is mainly designed to investigate processes following small and medium size breaks in the primary circuit and to study the natural circulation behaviour of VVER-440 type reactors. The 6 loops of the NPP are modelled by a single active loop with a scaling ratio of 1:2070. The experiments calculated by ATHLET are a 1% Cold Leg Break, a 1% Cold Leg Break with additionally accident management measures (primary bleed) and a Surge Line Break with full and with reduced high pressure and accumulator injection.
Currently ISB is the only operating test facility for russian type VVER-1000 reactors. The primary circuit of the NPP consists of four loops, which are modelled by two loops (1+3) in the ISB facility with a volume scaling ratio of 1:3000. Several experiments were calculated with ATHLET, like a small break in the upper plenum with locked rotor of all circulation pumps (1st Russian Standard Problem) and a natural circulation experiment with stepwise reduced mass inventory.
More about ATHLET-calculations for VVER-type reactors:
3. Theoretical and experimental investigations of natural circulation phenomena in VVER-type reactors
In co-operation with the KFKI Atomic Energy Research Institute Budapest a series of LOCA experiments were performed at the PMK-2 test facility:
- 1% Cold Leg Break
- 1% Cold Leg Break with additionally accident management measures (primary bleed)
- Surge Line Break
- Surge Line Break with reduced high pressure and accumulator injection
The thermohydraulic computer code ATHLET was used for calculations of these experiments.
PMK-2 test facility
LOCA experiments are characterized by a more or less rapid primary pressure decrease in the early phase of the transient. After pump coast down natural circulation becomes the dominant decay heat removal mechanism. A few hundred seconds after leak initiation boiling in the reactor core leads to formation of two-phase flow conditions in the primary circuit.
At these conditions different types of two-phase flow instabilities can appear. The instabilities play an important role in the behavior of the primary circuit, because the instabilities can disturb the decay heat removal from the reactor core.The appearance of such instabilities strongly depends on the thermohydraulic and geometrical conditions in the loop.
More about natural circulation phenomena in VVER-type reactors: