A beyond the design basis loss-of-coolant accident (LOCA) for the floating nuclear power plant (FNPP) with two water-cooled integral ABV reactors (Fig. 1) has been considered. Primary coolant natural circulation is ensured in the ABV reactor. The multi-sectional once-through steam generator is built inside the reactor pressure vessel. A noncondensible driven pressurizer is connected to the reactor by a pipeline. The design characteristics of the ABV reactor plant are presented in Table 1.

The emergency core cooling system of the FNPP includes two active (high and low pressure) subsystems designed to make up the primary coolant losses even at the guillotine break of any primary pipeline. Besides the ECCS, the active residual heat removal (RHR) system with emergency feedwater pumps (EFP) or the passive RHR system with pressurized accumulators can be used to provide emergency heat removal through the steam generator.

All primary system pipelines, including the pressurizer surge line, have 15 mm dia. orifices to reduce the leak flow rate during a LOCA.

The FNPP accident localization system includes the following leak-tight compartments (Fig. 2):

• Reactor containment shell.

• Pressure suppression pool (PSP) compartment.

• Auxiliary equipment compartment (not shown on Fig. 2).

Characteristic	Value
Reactor thermal power, MW	38
Primary coolant
pressure, MPa	15.4
core inlet/outlet temperature, C	245/327
natural circulation mass flow rate, kg/s	85.3
inventory, kg (m3)	5500 (7.6)
Secondary coolant
steam pressure, MPa	3.14
steam temperature, C	290
feed water temperature, C	106
feed water mass flow rate, kg/s	14.7
Pressurizer
type	gas
total volume, m3	3.0
water inventory, m3	2.4

Safety membranes are installed:

(1) in the connection between the containment shell and the pressure suppression pool compartment,

(2) in the connection between PSP and auxiliary equipment compartments (not shown on Fig.2), so that all plant compartments are separated at normal operation. The rupture of the first safety membrane occurs when the containment shell absolute pressure exceeds 0.2 MPa. After that, the mixture of water, steam, and noncondensible gases starts to flow through the gap around the metal-water shielding tank (Fig. 2) to the pressure suppression pool. In this way, a major portion of the steam mixture condenses on the outer surface of shielding tank and the remaining steam condenses in the pressure suppression pool. Cooled noncondensible gases accumulate in the free volume of the PSP compartment. The rupture of the second safety membrane occurs when the absolute pressure inside this compartment exceeds 0.3 MPa.

The above-mentioned metal-water shielding tank surrounds the reactor and contains around 26 tons of water. Under normal operation this water is cooled by special heat exchangers.

The following specific features of the ABV reactor plant design can be highlighted as the factors of high importance to the LOCA progression:

— elimination of large break LOCA thanks to exclusion of large-diameter primary pipelines;

Control rod drive mechanism

Closure head assembly

Reactor pressure vessel

Steam generator modules

Reactor Core

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Fig. 1 ABV REACTOR

Table 2 Containment system design characteristics Plant premises Parameter Containment shell PSP compartment Auxiliary equipment compartment Initial free volume, m3 172 40 70 Water inventory, m3 — 25 — Total heat exchange surface of compartment walls and equipment, m2 228 85 22 Heat exchange surface of the walls cooled with external air, m2 128 38 22 Total mass of compartment walls and equipment, kg 35600 13300 20600 Maximum permissible pressure in compartments, MPa 0.6 0.6 0.6

— employment of the outflow restrictors to decrease the primary coolant loss during a LOCA;

— upper location of the penetrations of the primary pipelines through the reactor pressure vessel;

— large water inventory above the core;

— favourable conditions for the emergency heat removal through the steam generator thanks to the primary system integral design;

— nitrogen presence in the primary system, including nitrogen dissolved in the primary coolant during reactor plant operation in transient regimes. The analysis of this factor is one of the main objectives of the study.

Accident Scenario

The postulated LOCA scenario is based on the following assumptions and events:

1. The accident is initiated by the double-ended break of the pressurizer surge line near the reactor pressure vessel.

2. Both trains of the ECCS fail to operate.

3. Additional failure of the shielding tank cooling system is supposed.

4. Full power operation of the reactor plant is assumed before the start of the accident.

5. Reactor scram occurs when the reactor pressure reduces to 12.5 MPa.

6. The EFP starts to operate after reactor shutdown and provides around 15 m3/h of feedwater at a temperature of 40 C.

7. The total amount of nitrogen dissolved in the primary water inside the reactor vessel is assumed to be 9.4 kg. The assessment of initial nitrogen concentration in the reactor coolant was based on the Henry law.

8. No operator action is performed to mitigate the accident.