What Is the Big Rip?

Pressurized water reactor nuclear power plant large break loss of water accident (LBLOCA) refers to the reactor coolant system loss of reactor coolant caused by a major rupture of the main pipeline of the reactor coolant system. The limit case of the design basis large-scale water loss accident is the case where both ends of the cold pipe section are broken and completely staggered.

Major breach

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Pressurized water reactor nuclear power plant large break loss of water accident (LBLOCA) refers to the reactor coolant system loss of reactor coolant caused by a major rupture of the main pipeline of the reactor coolant system. The limit case of the design basis large-scale water loss accident is the case where both ends of the cold pipe section are broken and completely staggered.

The large-scale loss of water accident is the limit of the rapid loss of coolant in a type of accident in which the reactor coolant capacity is reduced. The harm is very large, mainly manifested in:

(1) At the beginning of the accident, the sudden loss of pressure of the coolant outside the breach will form a strong shock wave in the primary circuit system. This shock propagates in the system at the speed of sound waves, which may damage the core structure. In addition, the violent spraying of the coolant will cause the pipeline to sway and damage the facilities in the containment.

(2) The core cooling capacity is greatly reduced, which may cause damage to the fuel element.

(3) High-temperature and high-pressure coolant is sprayed into the containment, which causes the pressure and temperature of the gas in the containment to rise, endangering the integrity of the containment.

(4) Zirconium cladding of fuel element

The accident process is divided into four phases:

(1) The spraying phase, at which time the coolant is sprayed out of the reactor vessel in large quantities;

(2) In the refilling phase, at this time, the emergency core cooling water starts to be injected into the reactor pressure vessel but the water level does not exceed the bottom of the core;

(3) Re-submergence stage, when the water level rises to a sufficient height to cool the core;

(4) In the long-term core cooling stage, the core is completely submerged, and the low-pressure safety injection system is put into use to remove the decay heat.

The details of each stage are as follows:

(1) Discharge stage

The discharge is performed in two stages, namely, underheated discharge and saturated discharge. During the initial underheated blasting phase, the pressure of the primary circuit quickly drops to the corresponding steam (saturation) pressure of the current water temperature. At this stage, pressure loss is accompanied by the propagation of pressure waves; therefore, a safety analysis must demonstrate that the mechanical loads caused by these pressure waves do not cause system damage to a level that poses a danger to the public.

In the longer saturated spraying phase, steam bubbles are generated in the coolant, and a mixture of water and steam is sprayed from the breach; this phase will generally last 15 to 20 seconds until the system pressure is substantially equal to the pressure in the containment vessel. . This two-phase flow is equivalent to the flow of steam and water mixture in the tube, which is called "plugging" flow. It will last for a period of time, in which the steam-water mixture is ejected at the maximum speed of sound, and the fuel rod is cooled to a certain extent when the steam-water mixture flows, so the cladding temperature decreases for a period of time. Then, as the enthalpy of the liquid increases, the critical heat flux density drops below the maximum heat flux density; the heat transfer coefficient decreases significantly, and the cladding temperature rises accordingly.

When the saturated discharge is carried out, the liquid phase content in the coolant is continuously reduced until the remainder becomes a foamy mixture of water and steam. The level of the foam dropped, leaving the upper part of the core exposed to dryness. At this time, the core can only dissipate heat by radiating to surrounding structures. Then the cladding temperature may rise to the extent that some hot fuel rods are damaged due to overheating of the cladding. At this time, the thermal energy generated by the zirconium water reaction also contributed to the further increase in temperature.

(2) Refilling stage

During the discharge process, the pressure drop signal of the PWR primary circuit will trigger ECCS action. The injection tank will inject boron-containing water into the reactor pressure vessel through the unbroken cold end or directly through the descending section; this will provide part of the cooling means for the fuel, but still a large amount of water will become steam-water mixture in the initial stage Mouth squirting. In the cold section fracture accident, it is desirable that the injected water descends along the descending section and rises through the core, but this movement is blocked by the liquid flow sprayed in the opposite direction. Due to the high resistance, the "ECC water bypass loss" phenomenon is likely to occur (see Figure 4.3-1), that is, the injected water flows along the annular space of the descending section and is ejected from the breach on the pipe. Part of the boron water began to focus on the bottom of the reactor pressure vessel and was refilled with water. If external power is not lost, the low-pressure safety injection subsystem of ECCS will begin to inject boron-containing water into the reactor pressure vessel. However, if the external power supply has also been lost, and the start-up emergency generator always lags behind for a period of time, the low-pressure safety injection system can only start to operate at the end of the refilling phase. In either case, the fuel must Not enough cooling; only the convection of the soda-water mixture can carry some heat.

(3) Re-submersion stage

When the water level rises to the lower end of the fuel rod, the submergence process begins. When the cooling water rises again, it comes into contact with the red-hot fuel cladding; as a result, a vapor film is formed on the surface of the latter, which makes it difficult for the cooling water to penetrate. Then the working condition of pool membrane boiling occurred. However, the temperature at the bottom of the core quickly dropped to such an extent that cold water could penetrate the vapor film, and nucleate boiling occurred. Then the temperature of this part of the cladding dropped rapidly. After the pressurized water reactor had a pipeline fracture accident, its injection tank was completely emptied about one minute later, but the supplementary water obtained from other sources of ECCS could completely submerge the core in about two minutes.

A problem in the re-submersion process is that the pressure of leaking steam tends to prevent the injection of re-submerged water. This phenomenon is called "steam plugging", which significantly reduces the rate of rise in the core water level. As a result of the interaction of the inertia of water and the compressibility of the remaining steam in the core, flow oscillations may occur, which should be taken into account in estimating the time required for resubmergence.

(4) Long-term core cooling stage

After the core is completely submerged, the low-pressure safety injection system will continue to take water from the refueling water tank and inject it into the reactor pressure vessel to maintain the core cooling. The containment ground pit takes water and achieves long-term core cooling through low-pressure injection and recirculation conditions.