Storage drum operating experience

The SPX fuel handling equipment was designed for a maximum residual power of about 28 kW per subassembly (Figs 17–19). In the initial design of fuel handling operations, the fuel storage drum played a double role:

— It ensured the transfer of all new and spent fuel subassemblies during reactor loading and unloading operations;

— It ensured the storage in sodium of spent fuel subassemblies to allow the decay of their residual power from 28 to 7.5 kW before cleaning and storage in the on-site water pool.

FIG. 17a

FIG. 17b

FIG. 17a,b. SPX new and spent fuel handling systems.

FIG. 18

FIG. 19

FIGs 18, 19. SPX storage drum and leak location.

The drum played the role of a storage buffer between the reactor and the water storage pool; it allowed reducing loading and unloading time to a minimum, and thus the duration of reactor outages. The storage drum is part of the fuel handling line.

To facilitate handling of the fuel between the storage positions and to reach the required capacity of fuel assemblies, two-level storage was provided; the maximum residual power from fuel assemblies stored was 3 MW, which was removed through two redundant sodium loops (Figs 17 and 18).

The storage drum consists of a 9 m diameter and 13 m high vessel made of 15 D3 ferritic steel with a low molybdenum alloy. It contained about 800 m³ (700 tons) of sodium at a nominal temperature of 200°C. This sodium was cooled or heated by sodium circuits circulating in a bundle of tubes fixed on the inner side of the vessel. The storage drum is equipped with a safety vessel. Both vessels were separated from each other by a 150 mm gap, and were located in a 2 m thick concrete pit.

In the initial plant project, it was envisaged to remove the fuel subassemblies to a reprocessing plant after a period of one year in the sodium-filled storage drum. It was also planned to renew 50% of the core after a half-cycle of 320 EFPD. The storage drum could receive 409 subassemblies with a decay heat of up to 28 kW. Removal of the subassemblies was carried out when the decay heat was less than 7.5 kW.

3.2.3.2. The incident

On 8 March 1987 a leak was detected in the space between the storage drum vessels by the leak detection system. Several methods were implemented to confirm the leak, and assess its importance. The balances for the levels of the storage drum and the storage tank showed that there were 20 m³ of sodium between the vessels.

The leak was confirmed and the first investigations to locate the leak and size of the crack started. This work took place between 27 August and 9 September 1987. Infrared thermograph, xenon and helium detection in the inter-vessel space were used. The localization of the leak had been carried out during the fall of the sodium level as a result of emptying the storage drum in August 1987. The leak was found to be along one of the lower support-plates for the storage drum cooling circuit (Fig. 19).

At the time of the leak, the storage drum contained only virtually radiation-free subassemblies whose decay heat was negligible. The most serious risk, although very improbable, was, in the short term, a leak on the safety vessel made of the same material as the storage drum vessel, not knowing the cause of the leak. The unloading operations of the storage drum were finished on 31 July 1987. The fuel and fertile sub-assemblies were then, after approval, put back in the reactor, and the slightly irradiated dummy subassemblies were at this stage stored in special containers constructed for the purpose.

After emptying the sodium completely, and when the storage drum had cooled down, investigations consisting in particular of X-rays through the two drums showed up that other plates were affected by the same type of not-yet-penetrating fault. It became obvious that local repair would be impossible, and that it was necessary to anticipate human access into the drum. This meant an air-fill of the main drum and its cleaning.

3.2.3.3. Causes of the accident

The leak was caused by a horizontal crack around 60 cm long on the lower angle welding bead which secured a plate. After the drum sodium drainage and the first investigations, identical faults to those observed on the plate at the origin of the leak were found on similar plates. Numerous other observations made in-situ showed that cracks of similar type, but

nevertheless less severe, existed not only in the vicinity of the plate-supports but also on the structure welds of the drum.

The reuse of the initial sodium drum after repair proved to be impossible and it was necessary to define replacement. When the drum was removed, it was found out that long (several meters) cracks had also formed in the constituting weld beads of the main vessel.

Considerable effort was expended on researching into the cause of the fault, in order to determine the follow-up necessary (reuse or not of the other facilities in 15 D3 ferritic carbon steel grade), and to assess possible risk on other structures or components.

The destructive examination samples taken at the beginning of 1988 showed that cracking was very probably due to:

— The existence of start sites (micro-cracking) in zones of high hardness;

— Embrittlement by hydrogen, and

— The cracks developed disruptive zones under the influence of residual welding stresses close to the elastic limit of the material.

3.2.3.4. Dispositions chosen for installation repair

After weighing up a considerable number of solutions to solve the problem posed by the occurrence of the leak, it appeared from October 1987 that the choices were between two categories of solutions:

— The first solution consisted of reconstructing the two geometrically very similar drums, but in austenitic steel;

— The second solution consisted of abandoning the decay storage function of the drum (this function being therefore assured in the reactor itself), and maintaining only the function of subassembly transfer.

It was the last solution that was finally chosen, in March 1988, particularly because it allowed personnel to resume more rapidly operation in normal conditions. The new device was called the “fuel transfer station” (PTC). The implementation of this solution required dismantling the fuel storage drum. This was started by the conversion of the residual sodium into sodium carbonate through a controlled additional of water vapor and carbon dioxide gas.

This operation (began in August 1988) allowed placing the vessel in contact with atmosphere and were performed with pressure suits. So as to ensure compatibility of site-work with plant operation, a drum work zone (zone de travail barilla, ZTB) was created to isolate the repair area from the rest of the reactor building, not only from the point of view of ventilation but also detection, protection, fire risk, health physics, and handling operations. In July 1989, the dismantling was over.

The schedule for the realization of the PTC shows that it will be available at the beginning of 1992: the in-situ assembling of the new stainless steel chamber started in December 1989.

Reactor operation was not possible without a fuel subassembly discharge route during the PTC construction. Consequently a special flask was designed to transfer the subassemblies directly from the reactor to the fuel cave.

In 1982 the necessity for fast reactors in France was less acute and therefore there was less need for a dedicated reprocessing centre. This led NERSA to decide on the on-site

construction of temporary storage for several spent cores. This was referred to as the fuel storage pool building (APEC). Then in 1988, when repair of the storage drum turned out to be impossible, NERSA chose to eliminate it altogether and replace it with a gas-filled transfer chamber

.

This modification, which was made possible through the existence of the fuel storage pool building, in turn led to modification of the management mode of the core which was based on frequency one. The core was renewed entirely after a cycle of 2 to 3 years (640 EFPD).

Replacing the subassemblies at a later stage required 7 to 8 months delay, including an initial period of 2 months, for decay of the first subassemblies to a level of 7.5 kW.

The APEC and the PTC make up two links of the handling line. Construction of the APEC covered the period 1984 to 1989, and that of the PTC lasted two years (1990/91). The PTC was a facility that implements a fuel removal process with no intermediate storage between the reactor core and the water filled pool. This facility was very different from the ferritic steel fuel storage drum.

The subassemblies were placed by the transfer machine in a sodium-filled container which is carried by the pot in the A-frame. In the argon-filled fuel transfer station under argon, the container carrying the subassembly is placed on a pivot arm which transfers it to the handling line. The pivot arm has a third position to receive the new subassemblies to be loaded.

After washing, the subassemblies are placed in a transfer shuttle which can receive three subassemblies. The APEC offers storage for approximately 1 700 subassemblies between the pool with a capacity of about 1 400 subassemblies and a hall which can house about 300 subassemblies in casks under gas atmosphere. Unloading capacity of the handling line is about four to five subassemblies per day.