Pressurized water reactors auxiliary systems

The nuclear steam supply system (NSSS) of the QP300 consists of a pressurized water reactor, reactor coolant system (RCS) and associated auxiliary systems. The NSSS has retained the general design features of current PWR plant design. 

The primary system, the reactor coolant pressure boundary, the safety systems and the high-pressure components of the reactor auxiliary systems are enclosed in the primary containment, a cylindrical concrete structure with an embedded steel liner. The primary containment is a pressure-suppression type with two major compartments a drywell and wet well. The drywell includes the volume that surrounds the reactor pressure vessel and the second shutdown system rooms. A partition-floor and cylindrical wall separate the drywell from the wet well. The lower part of the wet well is filled with water that acts as the condensation pool and the upper part is a gas compression chamber. 

In this work, a Motor-Operated Valve (MOV) of the Auxiliary Feed Water System (AFWS) has been selected based on several arguments. First, the basic event representing MOV fails to remain open is one of the most important contributors to the CDF based on the standard PSA available. Second, equipment aging, preventive maintenance and surveillance requirements are meaningful for this MOV. This basic event is modelled as a standby-related failure. This valve is normally open and its function is to control the flow from AFWS until Steam Generators on the secondary of a typical Pressurized Water Reactor (PWR) NPP. 

The primary circuit includes the equipment of the main circulation circuit, pressurization system and also auxiliary systems connected to the reactor. The fi ee volume above the reactor water level is used as steam-gas pressurizer for which gas is supplied to the pressurizer before reactor startup. The primary circuit operates in a non-boiling mode. 

The RB is designed for a relative pressure of 0,5 bar which conservatively covers accidental pressure increase due to the evaporation of the water of the Reactor Pool. Two RB sizes are foreseen a single module size and a three modules size with common fuel element and component handling facilities. Apart from this, each module has its own dedicated auxiliary systems, of which those which process primary fluids are located in segregated areas. A separate access to each potentially contaminated area is provided from the common areas in the upper portion of the reactor building. 

The HPCS system can operate independently of normal auxiliary AC power, plant service air, or the emergency cooling water system. Operation of the system is automatically initiated from independent redundant signals indicating low reactor vessel water level or high pressure in the primary containment. The system also provides for remote-manual startup, operation, and shutdown. A testable check valve in the discharge line prevents backflow from the reactor pressure vessel when the reactor vessel pressure exceeds the HPCS system pressure such as may occur during initial activation of the system. A low flow bypass system is placed into operation until pump head exceeds the nuclear system pressure and permits flow into the reactor vessel. 

As reactor heavy water coolant and the moderator heavy water are kept at nearly the same pressure, thin-walled tubes were sufficient to separate the fluids. The fuel channel tubes can thus be categorized as reactor internals. Also, the two systems use the same auxiliary systems to maintain water quality. 

Flow proceeds from the lower plenum, through the core. The steam and water are separated the steam is then dried and passed to the turbine. Other flow (see above) returns to the recirculation system. Feedwater is introduced to the annulus between the core shroud and reactor vessel (Fig. 4). The recirculation system piping is a primary pressure boundary for the high-pressure, high-temperature reactor coolant. Type 304 stainless steel was selected for recirculation system piping and numerous other auxiliary systems (such as the reactor water cleanup system, residual heat removal system, core spray, and other emergency core cooling systems) for its corrosion resistance and adequate mechanical properties. Failures of weld heat affected zones... 

The main and auxiliary cooling systems are based on natural circulation of water coolant. The containment vessel (CV) is water-filled, preventing activity release to the environment and acting as a radiation shield. The control rod drive mechanism (CRDM) is in-vessel type, with no penetrations in the reactor pressure vessel (RPV). No chemical and volume control system is used during reactor power operation. The PSRD has a passive reactor shutdown system. 

The major auxiliary systems of SMART consist of a component cooling system (CCS), purification system and make-up system. The function of the CCS is to remove heat generated in the main coolant pumps (MCPs), control element drive mechanisms (CEDMs), pressurizer (PZR), and the internal shielding tank. Feedwater supplied from the condensate pump of the turbo-generator is used as the coolant to remove heat. The purification system purifies the primary coolant and controls water chemistry to provide reliable and safe operation of the reactor core and all equipment in any mode of operation. The make-up system fills and makes-up the primary coolant in case of a primary system leak and supplies water to the compensating tanks for the PRHRS it consists of two independent trains, each with one positive displacement makeup pump, a makeup tank, and piping and valves. 

In addition to this opening failure, the failed closure of the motor-operated valves also became a problem after a valve in the auxiliary feed-water system in the US Catawba 2 plant didn t succeed in closing completely against an elevated differential pressure (14 March 1988). The reactor was shut down and no consequences ensued except for the overfilling of a steam generator. It was determined that the cause had been an underestimate of the friction coefficient between discs and seats of the valve by the valve fabricator. 

The capacities of the emergency core cooling systems suffice to provide water under all postulated pipe break conditions. This statement is also valid assuming that only two of the four redundant subsystems are operable. The postulated loss-of-coolant conditions include a hypothetical 80 cm leak at the bottom of the reactor vessel In this context, it can be noted that the capacity of the low pressure coolant injection punq)S has been reduced for BWR 90, following comprehensive core cooling analyses. As a secondary effect, it has been possible to simplify the auxiliary power supply systems. 

The HGP, owned by the Supply System, received steam via the steam piping system from the N Reactor. The HGP consists of two 430-MW (electrical) low-pressure turbine generator systems with associated auxiliary equipment normally found in a steam power station. The HGP is operated by the Supply System. The HGP condensers and auxiliary cooling systems were supplied by raw water pumped from the Columbia River and discharged back to the river approximately 90 m (300 ft) upstream from the N Reactor raw water intake structure. 

During reactor operation, the secondary system is operated as a steam condensate system to transfer heat removed from the primary loop to the circulating raw water system. Heat transfer from the primary loop results in boiling on the secondary side of the ten main heat exchangers. This steam then flows to the main 46 steam header atop the 109 Building. Sufficient steam to drive the six drive turbines and to supply auxiliary areas uses flows to the turbine supply header the remainder is condensed in the dump condensers, completing the heat transfer to the circulating raw water system. Secondary coolant pressures, flow rates and Inventory are directly controlled coolant temperatures are dependent variables. 

The reactor trips due to low pressure in the pressurizer. It is assumed that the emergency core coolant system does not operate during the transient. It is also assumed that there is not auxiliary feed water to the steam generators and no steam dump to the condenser. Only the accumulators participate in the sequence, since they are passive elements their discharge is only produced when vessel breaches, as, until this moment, the primary pressure is over the setpoint of the accumulator valves. Under these conditions the transient leads to core damage and melt-down. 

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