Equipment tests

A special attention was given to verify the correctness of the design principles used in the BN-600 and the functional qualities of the equipment. The primary circuit hydraulics tests were performed on a 1:6.8 scale reactor model with operating pumps (Fig. 4).

FIG. 4. Hydraulic model of the BN-600 reactor.

6 For more detailed information see: LEIPUNSKII, A.I. et al, The BN-600 fast reactor; BUDOV, V.M., et al., A NPP BN-600 - the plant for the near future; Reports presented at the Nuclex-75, Basel, 1975.

The data produced in this way enabled checks and changes to be made with the results of calculation studies, on such basic issues as:

─ The distribution of the coolant flow rate through the heat exchangers in various reactor operating modes with one loop not operating;

─ The variation in the coolant level in the reactor vessel and the pump stand pipe under various reactor operating modes;

─ The absence of any mechanical entrainment of gas by the coolant from the reactor vessel into the circuit, etc.

On a separate rig a study was performed of the coolant distribution through assemblies in the pressure headers. The influence on the flow rate through the assembly of the position of the holes in the assembly spikes was determined. Tests and final adjustments to the refuelling mechanisms were performed on sodium rigs with circulation of hot sodium. These tests particularly envisaged the performance of a number of studies in artificially created

“accident” situations, such as failure of the control system, complete loss of power to mechanisms, high oxide content in the sodium, etc. Comprehensive checking of the refuelling system together with the control system was performed on a control assembly rig which comprised: metal structures simulating the vessel and reactor transfer cell, rotating plugs, central column with control and safety system drivers, ramps, mechanisms for refuelling and transfer of assemblies, dummy subassemblies, a simulated inlet plenum and electrical equipment and movement systems. On the rig, a number of subassembly refuelling operations were carried out, both in normal mode and in conditions simulating of all types of events. The latter included:

— Loading and unloading of deformed subassembly with a 10 mm bend, with subassembly rotated through 60, 120 and 180°;

— Loading and unloading of subassembly, with misalignment of axis of refuelling mechanism relative to subassembly head axis by 10 mm;

— Checking possibility of disconnection and removal of refuelling machine from jammed subassembly, raised to a height of 500 mm.

Under all these conditions the refuelling mechanisms were found to function satisfactorily.

The refuelling mechanisms, elevators and fuel transfer mechanisms in the cells have been operated without any disturbances; above 40 reactor reloads have been carried out since the power unit BN-600 was put into operation. The total operating age in terms of double strokes of the in-reactor refuelling mechanism amounts to 36 100, for the elevators to 9 060 and for the ex-reactor fuel transfer mechanism to 29 200. These exceed the respective design values.

Based on results of audits of the refuelling mechanisms, their operating life has been extended.

4.1.2.1. Sodium rig for testing of drivers and control and safety system rods

Experimental final design work and full life tests were performed on the control and safety rod drive and their rods using a special sodium rig at a temperature of 550-580°C (Fig. 5).

The control and safety drive prototypes with absorber rods worked about 16 000 h on the rig, during which time the following operations were carried out:

— Automatic power regulation drive mechanism - 7 500 raising and lowering;

— Safety rod drive mechanism - 650 drops in fast scram mode and 600 raising and lowering in slow emergency protection mode;

— Shim rod drive mechanism - 2 600 raising and lowering.

These lengthy tests confirmed the reliability of all the control and safety rod drive mechanisms and the conformity between the main working characteristics and the design characteristics.

FIG. 5. Sodium rig for testing of drives, control and safety system rods for the BN-600.

4.1.2.2. Intermediate heat exchanger (IHX)

The IHX shell-and-tube sodium/sodium heat exchanger has 4 974 OD 16×1.4 mm tubes with expansion bends. In order to minimize the temperature difference between the bundle and the central tube, a double-tube arrangement was introduced. The annular gap is gas-filled. The perforated plate at the below part of the IHX was used to reduce the temperature difference between central tube and bundle. The distribution of the coolant flow rate through the heat exchangers in various reactor operating modes, including modes with one loop disconnected were performed on a 1:6.8 scale reactor unit model (Fig. 6). The study of IHX thermal hydraulics was performed on the mock-up tube bundles with the limited number of tube rows (one-, three-row bundle).

1, 2-secondary Na inlet/outlet; 3-shielding block; 4,9-primary Na inlet/outlet; 5-seeling element;

6-central downcomer tube; 7,8-heat exchange tube; 10-lattice FIG. 6. BN-600 intermediate heat exchanger.

The low cycle fatigue and fracture investigations done on tube-tubesheet joint IHX nozzle model and IHX to IHX support shell joint. For the tube-tubesheet joint, two models with 19 and 37 tubes respectively were studied. The test results for heating rate 50-70 K/h hold period of 3 h at 550°C and cooling rate with of 5-6.5 K/s had not indicated any fatigue damage.

The tests on nozzle model with the heat rate 60-90 K/h hold at 580-600°C for 3 h and cooling in time 30-40 s up to 400°C revealed no crack initiations in the all surfaces of 2 test models.

The calculated peak to peak stream values was 0.62% which can permit up to 3 000 cycles.

The last aspect was on the creep fatigue damage on IHX support shell joint. The calculations showed that the margin on number of load cycles is less than the permissible value. Hence tests were conducted on the flat model, with 12Cr18 Ni steel.

The test temperature was 490°C. The calculation and experiments’ comparison showed a good coincide. Accordingly crack initiation in the weld bead zone was observed after

1 100 cycles by acoustic emission method for the clearance of 0.5 mm between the shells. If this width is increased to 2 mm, cracks were seen only after 2 000 cycles. All six IHXs are operated since 1980 without any faults or troubles.

4.1.2.3. The pumps

The pumps underwent various model and full scale tests on rigs, where there was the possibility of simulating widely (bearing) varying operating conditions were carried out to guarantee the required high level of operational reliability. When the pumps were being designed, much thought was given to the hydraulic side, using computers.

Models of the pumps were created, and these were subjected to various hydraulic tests with water. Some of the pump units (shaft seal, bearings) were tested on special rigs, which enabled design changes to be made to them promptly and various designs to be tried out.

Besides the usual testing of the functional properties of the pumps and their drives in all possible operating conditions, the full scale water tests also included checking of the pumps under extreme conditions, e.g. during operation under cavitation conditions, with non-regulation start-up, with delivery overload, etc. Rigs for testing pumps with hot sodium enabled the hydraulic and electro-mechanical characteristics of the pumps and their parts to be checked under various thermal conditions. Successful long term life tests on pumps confirmed the correctness of the designs selected for them.

The strength of the highly loaded reactor vessel parts was tested using models. In particular, a 1:10 scale model of the roof slab was made. The basic principle of the modeling work was equivalence of the stress level of the model to that of the “real-life” version. On a rig simulating a real pipe with normal pump, studies were performed of the vibration and vibration-strength characteristics.

Construction work and the manufacture and assembly of equipment for the BN-600 began in 1969. The reactor plant building was made ready for equipment assembly in 1973. Assembly of the main equipment commenced in 1974. By mid 1975 welding of the reactor vessel was practically complete. While the vessel was being assembled, the secondary circuit pumps and the component handling and storage equipment were also being put together. Assembly work was completed in 1979. The power station with BN-600 reactor was commissioned in April 1980.