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Structural Aspects of Standardized Designs

The structural designs for Generation III/III+ reactors typically utilize the concept of a power block or nuclear island which is a large reinforced concrete common basemat shared by the containment and other safety-related structures to provide vital safety functions for the operation of the facility. A salient feature of the NI construction is to reduce the failures induced by the foundation differential settlement, especially for appurtenances and commodities that are connected between buildings. The nuclear island (NI) structures are designed to meet 10 CFR Part 50, Appendix A, General Design Criteria (GDC), including:

  • • GDC 1, “Quality Standards and Records,” as it relates to the structures being designed, fabricated, erected, and tested to quality standards commensurate with the importance of the safety function to be performed.
  • • GDC 2, “Design Bases for Protection against Natural Phenomena,” as it relates to the design of the safety related structures being able to withstand the most severe natural phenomena such winds, tornadoes, floods, and earthquakes and the appropriate combination of all loads.
  • • GDC 4, “Environmental and Dynamic Effects Design Bases,” as it relates to the structures being appropriately protected against dynamic effects including the effects of missiles, pipe whipping, and discharging fluids due to blowdown loads associated with the loss-of-coolant accident (LOCA).
  • • GDC 5, “Sharing of Structures, Systems, and Components,” as it relates to safety related structures not being shared among nuclear power units, unless it can be shown that such sharing will not significantly impair their ability to perform their safety functions.
  • • GDC 16, “Containment Design,” as it relates to the containment being design with the capability to act as a leak-tight membrane to prevent the uncontrolled release of radioactive effluents to the environment.
  • • GDC 50, “Containment Design Bases,” as it relates to the containment being designed with sufficient margin of safety to accommodate appropriate design loads.
  • • GDC 53, “Provisions for Containment Testing and Inspection,” as it relates to the containment being designed to permit appropriate in-service inspections (ISI).

In addition, 10 CFR 50.44, “Combustible Gas Control for Nuclear Power Reactors,” requires that the design demonstrates structural integrity of the containment against loads associated with combustible gas generation. Finally, the design of NI structures must satisfy the quality assurance criteria of 10 CFR Appendix B, “Quality Assurance Criteria for Nuclear Power Plants and Fuel Processing Plants.”

These regulatory requirements for the design of safety related nuclear structures are typically met by following the NRC Standard Review Plan (SRP) [9], Section 3.8. The SRP provides an acceptable method for the structural design to comply with NRC regulations. Although the SRP is not a requirement, but according to

10 CFR 52.47 (9), applications based on methods that deviate from the SRP acceptance criteria need to demonstrate that the proposed alternative provides an acceptable method of complying with the NRC’s regulations that underlie the corresponding SRP acceptance criteria. Therefore, the SRP criteria provide the framework for an acceptable standard to meet the regulations in the designs for nuclear power plant facilities.

The acceptance criteria for design of containments and other Category I structures and foundations are provided in SRP Section 3.8 to address the following technical aspects:

  • • Level of details in structural description
  • • Applicable codes, standards, and specifications
  • • Loads and load combinations
  • • Design and analysis procedures
  • • Structural acceptance criteria
  • • Materials, quality control, and special construction techniques
  • • Testing and in-service surveillance requirements
  • • ITAAC, technical specifications, and site interfaces

Containment can be of concrete or steel constructions. Concrete containments are further divided into reinforced concrete and prestressed concrete constructions. For prestressed containments, the current operating fleet of reactors in the United States typically utilized greased tendon technology except for H.B. Robinson steam electric plant which used grouted tendons for the containment in the vertical direction. For Generation III/III+ designs, an increasing number of PWR containments around the world including the U.S. EPR design have relied on grouted tendon deployments to withstand design basis LOCA loads, as the grouted tendons tend to provide stronger resistance to corrosions and better post tensioned performance than greased tendons.

Because of the dual function of the containment: (1) to house and to protect from design-basis hazards, safety-related structures, mechanical and electrical components, and distribution systems associated with a reactor coolant system; and (2) to serve a primary function as an engineered safeguard to contain the postulated radiological consequences of a LOCA accident, the containment structure is typically designed for a design pressure as high as 4 atmospheres (420 kPa, 60 psi) and a design temperature of 170°C (350°F) for a short duration design basis LOCA, and 377°C (650°F) for local hot spots with the concurrent occurrence of a 10-4 year or 10,000 year return period earthquake with design-basis mean peak ground surface accelerations that range from 0.1 to 0.75 g (1.0 g equals 980 gals). Concrete containments are also required to resist the effects of tornadoes with maximum wind speeds ranging from 386 km/h (kph) [240 miles/hours (mph)] to 579 kph (360 mph) concurrent with differential pressure drops and design basis tornado missiles. Hurricane wind effects are typically bounded by the design basis tornadoes; however, a recent study [10] indicated that for a number of coastal areas of the United States, the hurricane wind could exceed the tornado wind. Therefore, the standard designs are also required to evaluate the design basis hurricane wind effect in the design of concrete containments as well as other Category I structures [11]. Additionally, generation III/ III+ reactors have also incorporated additional protections such as a shield structure to ensure that the containment structures can withstand the impact of a commercial jetliner crash as required by 10 CFR 50.150.

Since the design requirements to contain pressure and temperature effects are more stringent (except for earthquake-induced membrane shear and possibly membrane tension in a concrete wall segment) than those required to protect the safety-related components, concrete containment structures tend to follow pressure vessel design code (ASME) more closely than they do with building design practice (i.e., ACI 318). Concrete containments are designed to the requirements of the ASME B&PV Code, Section III, Division 2, Subsection CC, “Code for Concrete Reactor Vessels and Containments,” also known as ACI Standard 359-01, American Concrete Institute [12]; while steel containments including containment penetration steel assemblies are designed to the ASME Code, Section III, Division 1, Subsection NE, “Class MC Components” [13]. Internal structures of the containment, which are typically comprised of concrete walls and slabs providing support to the reactor systems, are generally subject to differential pressure loads during the blowdown of the reactor coolant system. Since they are not required to remain leak-tight, the ACI-349 Code [14] is normally used for their designs. Internal structures could be supported on 2- or 3-ft-thick deformed-bar, reinforced-concrete slabs locate above the containment base mat liner or anchored via penetrations through the containment base liner directly to the containment base mat.

The current fleet of operating reactors were designed and constructed by customizing the design to the specific site conditions pursuant to provisions of 10 CFR Part 50. In this approach, the design for structures, systems, components (SSCs) considers site-specific natural phenomenon hazards such as: earthquake, tornado, hurricane and flooding, based on the physical characteristics of the site including geology, seismology, meteorology and hydrology of the site. For sites located in low seismicity zones, the customized design may not be controlled by seismic loads; rather, the design may be governed by other hazards such as tornadoes or hurricanes if the plant is located near a coast. And vise versa.

Standard designs, on the other hand, are generic in nature and are generally intended for multiple site conditions. Therefore, standard designs generally do not aim at addressing any particular site; rather, they focus the design to satisfy a broad range of site characteristics, so that any site can incorporate the standard design by reference without major modifications to the design. The site characteristics for a standard design are typically established by defining a generic site. In 10 CFR Part 52, the generic site is established by identifying associated site parameters. According to Part 52.47 (a) (1), a DC must include the site parameters postulated for the standard design, and an analysis and evaluation of the design in terms of those site parameters. Since the generic site dictates the extent of sites to which the standard design can apply, the scope of the site parameters should adequately cover the range of site conditions intended for future constructions of the standard design. SRP Section 2.0 provides guidance for developing site parameters for a standard design.

As discussed previously, a customized design needs to consider natural phenomena hazards such as: earthquakes, tornadoes, hurricanes, and flooding, which are most severe based on historic data for the particular site, and to ensure that the structures can withstand these hazards without loss of the capability to perform their safety functions. In fact, one can apply the performance-based approach such as ASCE 43 [15] for the seismic design by establishing a performance goal or objective in terms of annual probability of exceeding a design limit state. For nuclear structures, the design limit state typically corresponds to the onset of significant inelastic deformation or essentially elastic limit state, which means that the structure’s stress or strain state are within elastic limits globally while allowing limited inelastic excursions in certain confined regions of geometric discontinuities where localized high stress concentrations are expected. The same design principles can also apply to the standard designs. However, instead of considering the most severe of site-specific natural phenomenon hazards as is the case for a customized design, a standard design typically considers the most severe of natural phenomenon hazards for a wide range of sites or regions. For instance, because of the consideration for potential future sites, a standard design may need to include in the design the most severe snow loads, tornado loads, hurricanes and seismic loads across a region such as central and eastern United States. The design is performed to the envelope of all these loads. Therefore, the final design tends to be much more conservative than any individual sites targeted for the design. This would ensure the standardization of the plant with substantial benefits in enhanced safety as well as the use of modular construction techniques, streamlined process for constructions, procurements of systems and components, etc.

Lastly, a discussion is worthwhile on aspects of the constructions to ensure that as-constructed plant conforms to the design. The structural design is completed in a DC application with respect to the so called “critical sections.” These critical section designs are typically presented in terms of design drawings with sufficient details. To ensure that the critical sections and other structural elements which the critical sections represent are constructed in a manner that is consistent with the design drawings in the FSAR. 10 CFR Part 52 institutes the ITAAC verification process which will allow the NRC to verify through ITAACs established for the critical sections that the as-built structures are consistent with the DC design before loading the fuels into the reactor. Any deviations or departures from the standard design are controlled through rigorous regulatory processes in 10 CFR Part 52 to ensure the proposed changes to the design will enhance safety and result in an efficient and effective process for the constructions.

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