Generation III/III+ PWR Designs


This section provides a description of the AP1000 reactor features, particularly the features relating to the containment design, based on the information from publicly available documents [28, 29]. The AP1000 is a two-loop PWR built and improved upon the established of technology of major components (steam generators, fuels, pressurizer, reactor vessels, etc.) used in current Westinghouse-designed plants with proven, reliable operating experience over the past 35 years. The AP1000 design with rated power at 1117 MWe includes advanced passive safety features and extensive plant simplifications to enhance the safety, construction, operation, and maintenance of the plant. The safety systems use natural circulations, gravity flow, pressurized gas, and convective flow, but do not rely on active components such as pumps, fans, and diesel generators as were seen in most current operating reactors. The AP1000 standard design was certified by the NRC in 2005. Subsequently, the Westinghouse applied an amendment to the AP1000 to revise but not limited to the shield building for utilizing advanced modular structural designs for better performance against the effect of commercial aircraft impact and for taking advantage of advanced construction technologies. The AP1000 amendment was approved by the NRC in 2011.

Some of the high-level design characteristics of the AP1000 are highlighted below: [1]

  • • Predicted core damage frequency of 2.4E - 07/yr is well below the 1E - 05/yr requirement and frequency of significant release of 1.95E - 08/yr is well below the 1E - 06/yr requirement
  • • Standard design is applicable to anticipated US and international sites
  • • Occupational radiation exposure expected to be below 0.7 man-Sv/yr (70 man-rem/yr)
  • • Core is designed for a 18-month fuel cycle
  • • Refueling outages can be conducted in 17 days or less
  • • Plant design life of 60 years without replacement of the reactor vessel
  • • Overall plant availability greater than 93%, including forced and planned outages; the goal for unplanned reactor trips is less than one per year
  • • Leak-before-break on primary lines > 6 in. and on main steamlines
  • • Seismic based on 0.3 g ground acceleration
  • • Security enhanced with all safe shutdown equipment located in safety reinforced concrete Nuclear Island buildings
  • • In-vessel retention of core debris following core melt which significantly reduces the uncertainty in the assessment of containment failure and radioactive release to the environment due to ex-vessel severe accident phenomena
  • • No reactor pressure vessel penetrations below the top of the core. This eliminates the possibility of a loss of coolant accident by leakage from the reactor vessel, which could lead to core uncover
  • • Containment design pressure is 0.407 MPa and containment pressure capacity at severe accident conditions is estimated about 0.9 MPa

The AP1000 uses the nuclear island to support Seismic Category I structures including: the containment vessel, the shield building and the auxiliary building. The nuclear island structures are designed to withstand the effects of postulated internal events, as well as the effects of natural phenomenon hazards such as winds, tornadoes, hurricanes, and earthquakes. The containment vessel is a freestanding cylindrical steel shell structure with a wall thickness of 4 mm (1.75 in.), and a diameter of 40 m (130 ft.), and is surrounded by the shield building (Fig. 9.7).


The principle systems located inside the containment include the reactor coolant system, the passive core cooling system (PCCS), and the reactor coolant purification portion of the chemical and volume control system. Both the containment vessel and the shield building are an integral part of the PCCS.

The AP1000 containment contains a 4.9-m (16 ft) diameter main equipment hatch and a personnel airlock at the operating deck level, and a 4.9-m (16 ft) diameter maintenance hatch and a personnel airlock at grade level. Because of their locations, these large hatches significantly enhance accessibility to the containment during outages and, consequently, reduce the potential for congestion at the containment entrances. These containment hatches, located at the two different levels, allow activities occurring above the operating deck to be unaffected by activities occurring below the operating deck.

The containment vessel provides protection against the uncontrolled release of radioactive fission materials to the environment. It is designed with a leakage rate of 0.1 wt.% per day of the containment air mass which is present at the start of a design basis accident and the resulting containment isolation.

As shown in Fig. 9.7, the shield building surrounds the primary steel containment vessel and provides the natural convective annulus for containment cooling. During normal operations, the shield building, in conjunction with the internal structures of the containment vessel, provides the required shielding for the reactor coolant system and all the other radioactive systems and components housed in the containment. During accident conditions, the shield building provides an additional barrier for radioactive airborne materials that may be dispersed in the containment as well as radioactive particles in the water distributed throughout the containment.

The PCCS air baffle is located in the upper annulus area of the shield building. The function of the PCCS air baffle is to provide a pathway for natural circulation of cooling air in the event that a design basis accident results in a large release of energy into the containment. In this event the outer surface of the containment vessel transfers heat to the air between the baffle and the containment shell. This heated and thus, lower density air flows up through the air baffle to the air diffuser and cooler and higher density air is drawn into the shield building through the air inlet in the upper part of the shield building.

Another function of the shield building is to protect the containment building from external events. The shield building protects the containment vessel and the reactor coolant system from the effects of tornadoes and tornado produced missiles.

The shield building design in the original design certification is a cylindrical, reinforced concrete structure. It has a conical roof which supports the PCCS water storage tank and air diffuser. The PCCS is also referred to as PCS (passive cooling system). During the amendment application, the Westinghouse revised the shield building to substantially improve the structural performance of the shield building against the effect of potential commercial aircraft impact. The new design incorporates a steel-plate composite (SC) walls for the upper portion of the shielding building, which are connected to the traditional reinforced concrete (RC) walls in both the roof and the lower portion of the shielding building (Fig. 9.8).

Although the applications of SC constructions have seen in other countries, it is the first time that such a non-conventional design feature is proposed for nuclear structures in the United States. The design for traditional RC structures typically meets the provisions of ACI codes (such as ACI-318 for building designs, ACI-349 for nuclear structures) to ensure the adequate quality for the design. However, the design codes had not addressed the conventional designs such as SC constructions. During the amendment application review, the NRC staff raised a number of technical concerns pertaining to the adequate design and testing with respect to the SC constructions. Subsequently, the Westinghouse and its consultants made substantial design improvements to the design SC and its connections to RC sections, and also performed a number of testing to confirm adequate structural performance of the SC design [30-32]. To address the apparent lacking of code requirements, the American Institute of Steel Construction (AISC) formed a sub-committee to develop an appendix to AISC N690 focusing on SC walls.

A significant technological advance of the AP1000 over the previous PWRs is that the defense-in-depth of the plant relies on completely passive safety features without the need for either pumps or AC powers. These safety features include: (1) the passive core cooling system (PXS), (2) the passive containment cooling system (PCS), (3) the main control room emergency habitability system (VES), (4) in-vessel retention of molten core melt.

FIG. 9.8


Safety injection and depressurization — The PXS (Fig. 9.9) is the ECCS to protect the plant against RCS leaks and ruptures of various sizes and locations, and is designed to perform safety injection, reactor depressurization, and passive residual heat removal (PRHR). The PXS performs safety injections to cool the core relying on water from three passive sources: the core makeup tanks (CMT), the accumulators, and the in-containment refueling water storage tank (IRWST), which all located within the primary containment, as compared to conventional PWRs which locate these tanks outside of the containments.

The PXS injection sources are directly connected to two nozzles on the reactor vessel through dedicated direct vessel injection (DVI) lines so that no injection flow can be spilled for the main reactor coolant pip breaks. The high pressure safety injection is accomplished by the two CMTs which effectively replace the high pressure injection systems in conventional PWRs. The CMTs consist of large volume stainless steel tanks with inlet lines that connect the RCS cold legs to the top of the CMTs and outlet lines connecting to the DVI line. The DVIs are connected to the reactor vessel downcomer. Each CMT is filled with bo- rated water. The CMT inlet valve is normally open and therefore the CMT is normally at primary system pressure. The CMT outlet line is normally closed, preventing natural circulation during normal operation. When the reactor is tripped, the CMT outlet valve is open, establishing a natural circulation path. The cold borated water is injected in the reactor by gravity flow and hot primary fluid flows upward into the top of the CMT.

The intermediate pressure injection is done in a manner similar to current PWRs through accumulators. The PXS uses accumulators which are large spherical tanks approximately three-quarters filled with cold

FIG. 9.9


borated water and pressurized to 4.83 MPa (700 psig) with nitrogen. A pair of check valves prevents injection flow during normal operation conditions. When system pressure drops below the accumulator pressure, the check valves open allowing cold borated water flow into the reactor downcomer via DVI line.

The low pressure injection is supplied by gravity flow of coolant makeup from the IRWST which is located at a height above the RCS loops. The IRWST is a very large concrete pool (Fig. 9.9) filled with borated water at atmospheric pressure and the injection occurs only when the RCS is depressurized to near atmospheric pressure. The AP1000 automatic depressurization system (ADS) has four stages of valves sufficient for the controlled reduction of primary system pressure. The three stages consist of two trains of valves connected to the top of the pressurizer. The ADS 1-3 valves discharge primary system steam into a sparger line than vents into the IRWST. The steam is condensed by direct connect with the highly subcooled borated water in the IRWST. The fourth stage of the ADS consists of two large valves attached to ADS lines on each hot leg. The ADS-4 valves open on low CMT liquid level and bring the primary system pressure to the containment conditions. The ADS-4 valves vent directly into the containment airspace.

The IRWST also serves as the heat sink for the PRHR heat exchanger (PRHR HX) which is submerged in the IRWST. The PRHR HX removes the decay heat and protects the plant against transients that upset the normal steam generator feedwater and steam systems. The PRHR HX satisfies the safety criteria for loss of feedwater, feedwater line breaks and steam line breaks. The PRHR HX and the passive containment cooling system provide indefinite decay heat removal capability without operator actions.

Passive containment cooling — The PCS provides the safety-related ultimate heat sink for decay heat and protect the containment from overheating and exceeding the design pressure. As shown in Fig. 9.7, the PCS remove decay heat through convective natural circulation. The steel containment vessel provides the heat transfer surface that removes heat from inside the containment into the annulus space between the containment and the shield building. Outside air is pulled in through orifices near the top of the shield building cylindrical walls and circulates around the baffle and through the annulus to remove the heat from steel containment surface, and then discharges the heated air upward out of the shield building through the central opening in the shield building roof. In addition, the PCS gravity drain water tank located on the top of the shielding building drains by gravity to cover the outer surface of the primary steel containment vessel.

Main control room emergency habitability system — The VES provides fresh air, cooling, and pressurization to the main control room (MCR) following a plant accident. Operation of the VES is automatically initiated upon receipt of a high MCR radiation signal, which isolates the normal control room ventilation path and initiates pressurization. Following system actuation, all functions are completely passive. The VES air supply is contained in a set of compressed air storage tanks. The VES also maintains the MCR at a slight positive pressure, to minimize the infiltration of airborne contaminants from the surrounding areas.

In-vessel retention (IVR) of molten core melt — The AP1000 manages the severe accident molten core debris cooling utilizing the IVR strategy. During postulated severe accidents, the accident management strategy is to flood the reactor cavity by IRWST water to submerge the reactor vessel. The water cools the external surface of the vessel and prevents molten debris in the lower head from failing the vessel wall and relocating into the containment. Retaining the debris in the reactor vessel protects the containment integrity by preventing ex-vessel severe accident phenomena, such as ex-vessel steam explosion and core-concrete interaction, which have large uncertainties with respect to containment integrity.

The passive AP1000 design is uniquely suited to in-vessel retention because it contains features that promote external cooling of the reactor vessel. Figure 9.10 provides a schematic of the AP1000 reactor vessel, vessel cavity, vessel insulation and vents configuration that promotes IVR of molten core debris.

  • [1] Net electrical power of at least 1117 MWe; and a thermal power of 3415 MWt • Rated performance is achieved with up to 10% of the steam generator tubes plugged and witha maximum hot leg temperature of 321°C (610°F) • Core design is robust with at least a 15% operating margin on core power parameters • Short lead time (5 years from owner’s commitment to commercial operation) and construction schedule (3 years) • No plant prototype is needed since proven power generating system components are used • Major safety systems are passive; they require no operator action for 72 hours after an accident and maintain core and containment cooling for a protracted time without ac power
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