Drywell pressure

Selected safety/relief valves are associated with the automatic depressurization of the primary system under assumed LOCA conditions. These valves have two independent logic channels powered from different power sources, either of which can initiate depressurization. Valves open automatically and remain open imtil the pressiue falls to a preset closure pressure. These valves open automatically upon signals of high drywell pressure and low reactor water level and confirmation of one LPCI function of the RHR system or LPCS system running. Initiation signals need not be simultaneous. The valves remain open until the primary system pressure is reduced to a point where the LPCI function of the RHR system and/or the LPCS system can adequately cool the core. The initiation of automatic depressurization is delayed from 90 to 120 s to allow the operator to terminate the initiation should the HPCS system initiation and acceptable reactor vessel level have been confirmed. 

The operation of the ECCS network is automatically activated by the reactor protection system upon redundant signals that are indicating low reactor vessel water level or high drywell pressure or a combination of indicators showing low reactor vessel water level and high drywell pressure. 

High pressure in the dryrvell Abnormal drywell pressure trips the reactor, initiates the automatic depressurization function, the HPCS system, the LPCS system, and the RHR system. 

The RPV control EOF is designed to maintain adequate core cooling, shutdown of the reactor, and decrease RPV temperature to cold shutdown conditions. Entry into this procedure is required at an RPV water level below (low level scram set point), RPV pressure above (high pressure scram set point), drywell pressure above (high pressure scram set point), and SCRAM condition and reactor power above a certain value (average power range monitor downscale trip) or reactor power cannot be determined. 

The PC control EOF is designed to provide a barrier to the uncontrolled release of fission products, contain and condense steam discharged through the safety relief valves and primary cooling system breaks, shield personnel from radiation emitted by the reactor, and provide a protected environment for key equipment important to safety. Entry into this procedure is required at a suppression pool temperature above the limiting condition for operation (LCO), a drywell temperature above LCO, a containment temperature above LCO, a drywell pressure above the high pressure scram set point, a suppression pool water level above maximum level LCO, a suppression pool water level below minimum level LCO, and an SC hydrogen concentration above the alarm set point. 

Pressure high (level 1) Pressure low (level 1) MSIV closure (90%) ECCS start-up Drywell pressure high Reactor period short (10 s)... 

Main coolant flow rate low (level 3) Drywell pressure high... 

When the critical pressure is higher than the drywell pressure, the mass flux is determined regardless of the critical pressure. 

The system pressure, defined as the upper plenum pressure, is calculated by adding the pressure loss to the drywell pressure. The pressure drop consists of friction, nozzle, acceleration, and head loss in the suppression pool. The depth of the suppression pool and the submergence of the quencher will be determined from the containment design. It is presently assumed as 2.0 m, which means that the head loss is 0.02 MPa. The sensitivity of this parameter is investigated in Sect. 6.7. 

The pressure-suppression type primary containment is a cylindrical concrete structure with an embedded steel liner type with two major compartments a drywell and wetwell. The lower part of wetwell volume is filled with water that works as the condensation pool, and the upper part is a gas compression chamber. The blow-down pipes from the safety relief valves are routed to the pressure suppression pool. 

The pressure suppression function is supported by a four train containment spray system that is continuously in service, with one train supplying spray water from the condensation pool to the gas compression chamber in accident situations the system will start operation at full capacity. Spraying is also possible for the upper drywell - after rerouting, on operator action. The drywell spray is generally initiated only in the event of "small" LOG As to "depressurize" the containment. 

A containment is provided that completely encloses the reactor systems, drywell, and suppression chambers. The containment employs the pressure suppression concept. 

The PCCS loops are driven by the pressure difference created between the containment drywell and the suppression pool during a LOCA and require no sensing, control, logic or power actuated devices for operation. Together with the pressure suppression containment system, the three PCCS condensers limit containment pressure to less than its design pressure for at least 72 hours after a LOCA without make-up to the IC/PCC pool. 

The SPWR adopts a water-filled CV. The RPV is covered with a water-tight shell. A mirror insulator of laminated thin stainless steel plates is mounted inside the shell. Heat loss is estimated to be below 1 MW. The space between the water-tight shell and the RPV corresponds to a drywell of the suppression type CV of conventional BWRs. To allow for a pipe rupture in this space, the shell is equipped with pressure relief valves. The advantages of the water-filled CV are the compactness of the reactor plant and ease of application of a passive decay heat removal system. 

A pressure relief function is used to control large pressure transients. This system will operate safety/relief valves following closure of the main steam isolation valves or the sudden closure of the turbine admission or stop valves and failure of the turbine bypass system to relieve the excess pressure. For this fimction, the safety/relief valves discharge steam from the steam lines inside the drywell to the suppression chamber. Each safety/ relief valve is operated from its own overpressure signal for the relief fimction, and by direct spring action for the safety function. 

During the evolution of the BWRs, three m or types of contaiiunents were built Mark 1 (page 3-16), Mark 11, and the Mark 111 (page 3-18). Unlike the Mark HI, that consists of a primary containment and a dry well, the Mark 1 and Mark n designs consist of a drywell and a wetwell (suppression pool). All three containment designs use the principle of pressure suppression for LOCAs. The primary containment is designed to condense steam and to contain fission products released from a LOCA so that offsite radiation doses specified are not exceeded and to provide a heat sink and water source for certain safety related equipment. 

The Mark 111 primary containment consists of drywell which is a cylindrical, reinforced concrete structure with a removable head and suppression pool. The drywell and wetwell are connected via the weir wall and the horizontal vents. The suppression pool contains a large volume of water for rapidly condensing steam directed to it. A leak-tight, cylindrical, steel containment vessel surrounds the drywell and the suppression pool to prevent gaseous and particulate fission products from escaping to the enviromnent. The containment system has water spray system at top of the drywell that activates when the containment pressure increases from a set point. 

The steam which in a loss-of-coolant accident is released from the primary system may lead to a pressure increase inside the containment and to a pressure difference between the drywell and the condensation chamber. As a consequence, a steam-air mixture is transported to the pressure suppression pool where the steam is condensed. Simultaneously, fission products which might be carried with the steam are retained in the water volume of the pool, thus efficiently reducing airborne radioactivity. 

If the pressure in the outer containment exceeds the pressure in the drywell, then vacuum breakers open to equalize the pressure. Long-term containment heat removal can be accomplished with suppression pool cooling or by containment sprays (with appropriate circulation of the water through heat exchangers) in the outer containment. 

Generation of pressures and shock waves that can fail the drywell floor of a BWR Mark II containment or the drywell wall of a Mark III containment. 

The Mark-I liner attack problem is well known and it will not be elaborated here. Voy briefly, it is concerned with the possibility that the molten corium released from the reactor vessel (in an unmitigated severe accident in a BWR with a Mark-I pressure siq>ptession containment) will come in contact and cause a breach in die containment liner. This liner is the containment pressure boundary, and such a breach would constitute an "early containment failure. The situation is illustrated in Figure 1. The important phenomenology is sketched in Figure 2, which also notes the k terminology enq>loyed in such discussions. The object is to detomine the likelihood of such a liner breach (conditional on die occurrence of an unmitigated severe accident), and especially to consider tte effect of flooding (by wat ) of the drywell floor. 

One of the important issues in the analysis of acddait consequences in BWRs and with MARK-I containments is concerned with the possibility of drywell liner failure as a result of attack by hofcoiium spreading over the drywell floor. The probability of failure strongly depends on the composition of corium released from the reactor pressure vessel [Theofanous et aL, 1990]. 

The objective of the present work was to perform numerical analyses in which the effect of heat transfer from a degraded partially molten reactor core towards the upper intemals (upper shroud, standpipes, steam separator dryer) in a BWR reactor pressure vessel (RPV) would be accounted for, and the resultant failure and melting of these structures, on e composition of corium released to flie drywell following RPV failure. The calculations for this study were performed using the APRIL.MOD3 computer code, and the Peach Bottom BWR was used a reference power plant with MARK-I containment... 

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