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Steam line break accident description

The main steam line break accident is initiated by a full size break of one of the main steam line at outlet of one steam generator.
The accident is supposed to occur at end of cycle with the reactor at hot zero power condition.
All the control rods are then fully inserted, except for the most reactive one, which is supposed to remain stuck at its position at nominal operating condition.

Nuclear design must insure that, at accident time, the core be subcritical by at least 2% (rc= -.02) (typical, example value).
At end of cycle, the boron concentration boc is close to zero, so that the moderator density reactivity coefficient is strongly positive (typically .35e-3 /kg/m³), so that the rapid fall of moderator temperature caused by the liquid discharge out of the faulted sg's risks to use up the subcriticalty margin and bring the core again at power.
The safety margins must be verified in that adverse situation

The accident is particularly of concern for several reason:
the main process for checking the accident is the counter-reactivity developed by the intrinsic core reactivity feedback's as core returns to power.
boron injected by the safety injection system triggered at the accident onset appears to comes too late for checking the reactivity rise.
SG's abnormal working conditions: in the faulted sg, modeling the heat exchange, as the tubes start to be uncovered by saturated water, is difficult because the weak heat capacity of dry steam causes rapid steam temperature increase and the uncoupled heat exchange model fails.
Modeling stream discharge through the break is not obvious.
Normally, the unaffected sg's will be isolated at accident's inception by automatic closure of the fast isolating valves, and these sg ceases to behave as a natural recirculation sg because the secondary water becomes warmer than the primary water.
Modeling the large temperature and flow unbalance effects through the vessel downcomer, bottom, core entrance and exit is difficult.
In the core itself, the neutronic-hydraulic interactions will cause rapid spatial power radial and axial redistributions which challenge the core calculation model. cf for a discussion of how these effects are attempted to be accounted for by the additional neutronic tabulations.
In case primary water shrinks too rapidly before being reheated by the possible return to power, the pressu becomes voided and boiling can occur in the primary system.
Boiling will generally start in the vessel dome, which tends to keep its initial temperature.
Boiling in the primary system mitigates the pressure decrease but challenges the calculation if core is still at power.
When pressu refills after complete voiding of the faulted sg, its behavior departs totally from normal operation.

In order to tackle the problem, we will, as usual, try uncoupling the effects in order to identify the critical ones and adopt a bounding analysis approach when modeling difficulties are encountered.

We will start analyzing the accident for the highly improbable case that the isolating valves of the intact sg's fail to operate, so that all sg loose their water through the break.
In that case we will, however, not retain the, equally highly improbable, stuck rod assumption.

Because of the paramount importance of spatial redistribution effects in the core, it is not relevant to try analyzing the transient with a point kinetics representation of the core.
Furthermore, in the situations where boiling appears in the core, the program execution may crash because of the strong pressure-reactivity coupling oscillations. Therefore, all the applications need to be carried out with a fine mesh representation of the core.

In case of failure of closing the rapid isolation valves, the minor loops unbalances caused by the location of the pressu on loop1 hot leg and by longer discharge paths of the intact sg, will be neglected and the loops and core representation can be simplified: l9= 1, xwl= 3 for a 3-loop plant and osplit= f.

This will allow gaining a first feeling of the accident trend and on the importance of parameters such as core initial power, effective fraction of delayed neutrons, neutronic reactivity feedback's, steam discharge rate, sg feed water flow, sg and pressu initial water inventory,...
Furthermore this will provide envelope case for the normally protected MSLB.
Next, the classical unsymmetrical, single faulted sg, accident event will be investigated with a 2 loop representation: loop1 feeding the faulted sg and loop2 duplicated, unsymmetrical core with osplit= t, effects of isolating valves closure time, discharge rate, flow mixing parameters, ... can thus be examined.

Because of the simplification of the various input data, it is obvious that the findings gained from our first preliminary investigations must be confirmed whith more realistic and more complete data sets.

For some cases at least, it will be observed that, at most dangerous conditions, the core and system are quasi stationary, which makes it feasible checking the state point by means of core and sg stationary calculations with normal 3-d core and sg design tools.