3. Accidents

1. Postulation of Events

For the purposes of safety analysis, events that have lower possibility of occurrence than "abnormal transients in operation” but, if they should occur, could have a much more serious consequence on the plant and the general public in the vicinity of the plant are selected as "accidents.” For PWR plants the following 13 events of four categories are analyzed.

© Loss of coolant or significant change in the state of core cooling

© 4 Loss of coolant

© -2 Loss of coolant flow

© -3 Locked rotor of reactor coolant pump

© 4 Break of main feedwater piping

© -5 Break of main steam piping

@ Excessive reactivity insert or significant change of reactor power

© -6 Control rod ejection

© Excessive release of radioactive materials into the environment

@-7 Rupture in the radioactive waste gas handling facility

© -8 Steam generator tube rupture

© -9 Fuel assembly drop

© 40 Loss of coolant

© -11 Control rod ejection

@ Excessive change in pressure, atmosphere, etc. inside the reactor containment vessel.

@ -12 Loss of reactor coolant

© -13 Generation of flammable gases

The criteria for “accidents" are shown in Table 7.1.1. In this case, unlike that of “abnormal transients during operation,” the events selected involve core damage and release of radioactive materials outside the facility, and therefore acceptance criteria based on the premises (e.g. possibility of cooling of reactor core and effective dose equivalent of the general public in the vicinity of the facility) are adopted.

“Loss of reactor coolant” and “steam generator tube rupture“ are explained as examples of safety evaluation of “accidents" for 3-loop plants. These events are also selected for a “major accident" and a “hypothetical accident" to evaluate siting pertinence of the NPP.

2. Loss of Reactor Coolant

In the postulated event, a crack or a break, etc., in the reactor coolant piping or in its connected piping up to the first isolation valve, is assumed to occur and cause loss of reactor coolant from the primary coolant system. The progress of the event varies depending on the size of the break of the piping. Here as the worst case, an instantaneous complete severance of a main piping of the reactor coolant system is postulated.

If nothing is done against the accident, the coolant water flows out from the primary coolant system, and the reactor core is heated up and “burns dry” which eventually leads to significant damage of fuel rods.

In reality, however, selection of materials, construction and inspection are stringently performed, and hence the probability of coolant piping break is practically incredible. Furthermore, multiple countermeasures are provided for loss of coolant accident. They are as follows.

1. Provision of monitoring devices for early detection of minor leaks of coolant

2. Provision of an accumulator tank in each coolant loop that automatically injects boric acid water into the reactor when the pressure of the reactor system drops below the accumulator tank pressure due to the release of reactor coolant.

3. The emergency reactor core cooling system is actuated by the following signals and the boric acid water in the refueling water storage tank is injected into the reactor by the pumps: •Coincident of low reactor pressure and low

pressurizer water level

•Low-low reactor pressure

•High pressure of reactor containment vessel

4. If the pressure of the reactor system decreases and many voids are formed in the reactor core, reactivity is suppressed and the reactor is led to the shutdown condition. Furthermore, either the “low reactor pressure" signal or the “actuation of emergency reactor core cooling system" signal automatically trips the reactor.

The results indicated below are with the assumption of loss of external power and failure of one train of the low pressure injection system to make the analysis conservative.

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downward again (@ in Figure 8.3.1).

3. About 16s after the break, the reactor pressure decreases below the accumulator tank pressure, and boric acid water is automatically injected from the tank to the primary coolant system.

4. After the break, heat generation by nuclear fission ceases due to the voids generated in the core. But stored energy in fuel pellets continues to be released and generation of decay heat continues.

On the other hand, as indicated in (2) above, the temperature of the fuel cladding rapidly rises as a result of stagnation of flow in the reactor core, but starts to decrease as coolant discharge from the break opening proceeds and flow in the reactor core resumes. However, with discharge of the reactor coolant progressing further, the coolant flowing through the core gradually decreases and the temperature of the fuel cladding rises again.

Core Flow (t/s) Core Pressure (MPa (gage))

"5

£

to s o u g

&

(Z)

Time (s)

Figure 8.3.1 Loss of reactor coolant (large break)

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