The SuperCritical Water Loop (SCWL)

The SCWL (SuperCritical Water Loop) was designed by CVŘ under the framework of SUSEN project. The main goals are to study nuclear cladding materials and neutron irradiation in supercritical conditions. Furthermore, it gives possibility to investigate some phenomenon encountered in SCW systems (i.e. corrosion layer creation, cracking, initiation, heat transfer deterioration DHT). The loop (Figure 1) is intended to be inserted into the LVR-15 research reactor found in the CVŘ platform.

SCWL was designed to reach a maximum temperature of 600º C in the sample area and a pressure of up to 25MPa. The basic technological scheme and flow of the AC (active channel) is shown in Figure 2. The active channel consists of a pressure envelope, a lower chamber, a heat exchanger, an electric heater, a sample section, a bypass, two wolfram targets, and an inlet and outlet cooling nozzles (Figure 3). The AC will be installed in the reactor in an aluminum reciever in order to physically separate the pressure envelope from the reactor’s coolant and reduce the thermal losses of the AC. The pressure envelope consists of a stainless steel tube with a diameter of 61 mm, a thickness of 10 mm and a length of 4662 mm. The material used is steel 08Ch18N10T.

The flow passage is divided into two upward and downward streams. At the end of the first upward stream, the flow is divided into the bypass and into the cold part of the recuperator by a ratio of 65% and 35%. Therefore, the in-pile AC can be redesigned to meet the requirements of various experiments, like, for example LWR testing conditions.

For the amendment to the LWR-15 the AC was modeled using ATHLET 3.1A code in order to prove its safe operation under both normal and accident conditions that could lead into a potential damage to the reactor. The code was developed by GRS and has integrated for the supercritical water regime several state of the art heat transfer correlations. Out of which, two were selected and certified by SUJB (State Office for Nuclear Safety) thermo-hydraulic commission before performing the analyses as part of an amendment to LVR-15 Final Safety Report.

For accident conditions Loss of Flow Accident (LOFA) and Loss of Coolant Accident were selected assessing in particular the pressure envelope and reciever temperature profile. In both normal operation and accidental conditions the results showed that the parameters are in accordance with the acceptance criteria.


Alis Musa, CV REZ