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CODE FOR CONCRETE REACTOR VESSELS AND CONTAINMENTS
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14.1 INTRODUCTION
This Chapter describes the bases and provisions of the Code for
Concrete Reactor Vessels and Containments. After a short description
of the provisions for Concrete Reactor Vessels, the Chapter
describes the concrete containment general environment, types of
existing containments, future containment configurations, and background
development including the regulatory bases of concrete containment
construction code requirements. The description covers
sequentially the following topics: Introduction, Concrete Reactor
Vessels, Concrete Reactor Containments, Types of Containments,
Future Containments, Regulatory Bases for the Code Development,
Background Development of the Code, Reinforced Concrete
Containment Behavior, Containment Design Analysis and Related
Testing, Code Design Requirements, Fabrication and Construction,
Construction Testing and Examination, Containment Structural
Integrity Testing, Containment Overpressure Protection, Stamping
and Reports, Containment Structure and Aircraft Impact,
Containment and Severe Accident Considerations, Other Relevant
Information, Summary and Conclusion.
The previous editions of this Chapter were developed by John
D. Stevenson. The basic format of this chapter is kept the same as
in the previous editions. The updates and additional information
relating to the regulatory bases for the code requirements, future
containments and considerations for future revisions of the Code
included in this update are based on contributions from Hansraj
Ashar, Barry Scott, and Joseph Artuso.
14.1.1 Concrete Reactor Vessels
Concrete reactor vessel construction requirements are contained
in Subsection CB of Section III, Division 2 of the
American Society of Mechanical Engineers (ASME) Boiler and
Pressure Vessel (B&PV) Code. Fort St. Vrain is the only concrete
reactor vessel built in the United States. The Subsection CB
requirements were developed in parallel with the design and construction
of Fort St. Vrain. The reactor is no longer in operation.
Because there are no plans to construct concrete nuclear reactor
vessels in the foreseeable future, the current text of Subsection
CB, though still being published in the periodic Code editions, is
no longer being actively maintained by the ASME Boiler and
Pressure Vessel Code committees.
14.1.2 Concrete Reactor Containments
Concrete containment construction requirements are contained
in Subsection CC of Section III, Division 2 of the ASME Boiler
and Pressure Vessel Code.
In general there are two types of barriers intended to resist
release of radioactive material at nuclear facilities: confinements
and containments. A confinement is a barrier intended to keep
unpressured, usually solid radiological material from being
released from its intended location and is not generally required to
be leak-tight. In many instances it is operated at a slight negative
pressure (i.e., in. water gauge: 0.018 psi, 122 Pa) to ensure that
any confinement leakage during normal operation would occur
only into the confinement space. Confinement construction is not
addressed by the ASME Boiler and Pressure Vessel Code.
Containment as considered herein is often referred to as secondary
containment and acts as a final barrier to radiological
releases to the environment. The primary containment is typically
considered fuel element cladding in a water-cooled and moderated
nuclear reactor. Secondary containment acts as a barrier to the
release of radioactive fluids under pressure and is designed to be
leak tight under pressure. In general, large nuclear power, watercooled
reactors require containment vessels or structures surrounding
a high pressure and temperature reactor coolant system.
Confinements are often used as barriers to release of unpressurized
or solid spent nuclear fuel and high-level waste storage and
processing facilities, and also have been used as barriers for
release of reactor coolant in gas-cooled nuclear power plants.
Containment construction1 criteria, which include design loads,
load combinations and acceptable behavior applicable to concrete
nuclear containments, developed originally as a unique combination
of mechanical and civil structural engineering procedures. They are
composed of procedures considered in design and analysis of boiler
12
CODE FOR CONCRETE REACTOR
VESSELS AND CONTAINMENTS
Hansraj Ashar, Barry Scott, Joseph F. Artuso and
John D. Stevenson
1 Typical design and layout characteristics for both BWR and PWR systems are shown in Table 14.1. Construction as defined by the ASME Code includes
the prescription of administrative, documentation, material selection and material qualification, design analysis, fabrication, erection, examination, startup
testing, structural integrity testing, overpressure protection, and Code-stamping requirements.
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2 • Chapter 14
and pressure vessel components developed by the ASME Boiler
and Pressure Vessel Code Committee, and those procedures used
in design of conventional concrete building structures as developed
by the American Concrete Institute (ACI) 318 Code
Committee. This situation naturally follows from an understanding
that such containments perform a dual function: (1) to be a
building structure used to house, and to protect from design-basis
hazards, nuclear 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 loss-of-coolant accident in the nuclear steam supply
system.
As a result of such design basis accidents, the containment
structure in the existing greater than 1,000 M Wt light-water reactor
nuclear power plants may use internal stream vapor design
pressures as high as 4 atmospheres (60 psi, 420 KPa) and a design
temperature of 350F (177C) for a short duration design basis,
loss-of-coolant accident, and 650F (377C) 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.
Normal long-term operation temperatures in the containment
concrete are limited generally to 150F (66C) and in local areas
(such as near penetrations) to 200F (93C).
Concrete containments in the United States are also required to
resist the effects of tornadoes with maximum wind speeds ranging
from 240 to 360 mph concurrent with differential pressure drops
and design basis tornado missiles.
Normally, the design requirements to contain pressure and temperature
effects are more severe (except for earthquake-induced
membrane shear and possibly membrane tension in a concrete
wall segment) than those required to protect the safety-related
components from extreme natural hazard effects, such as earthquake,
tornado, and flooding, and human-induced design-basis loads
such as blast, small airplane crash, and external missiles. Hence,
concrete containment structures tend to follow current pressure
vessel design more closely than they do building design practice.
This section is a brief description of the construction requirements,
techniques, and procedures developed by the joint ACI
359 and ASME B&PV Code Section III, Division 2, Subsection
CC committees for concrete containments. They bridge the gap
between steel pressure component construction developed by the
ASME and nuclear safety–related concrete building structures
developed by the ACI. Inherent in this discussion is the noting of
differences between working stress design (WSD) normally used
by ASME and ultimate strength design (USD) used by ACI
codes, as well as the use of load factors instead of allowable
stresses to provide necessary design margins. Understanding these
differences in philosophy between the two technical societies and
the codes and standards they have developed is important to the
understanding of the content and application of the Joint ACIASME
Concrete Containment Code for construction of nuclear
containments.
TABLE 14.1 RANGE OF WATER REACTOR CONCRETE CONTAINMENT DESIGN AND LAYOUT PARAMETERS
Percentage above
Design Maximum Enclosure Height to Base Material Cylinder Wall Dome
Pressure, Calculated Volume Springline Diameter Thickness, Thickness, Thickness,
Type psi Pressure 100 ft.3 100 ft 100 ft ft. ft. ft.
PWR I-S [Note (1)] 45–60 10 to 20 1.2–2.4 1.2–1.5 1.05–1.20 8–10 3.0 2.0
PWR I-L [Note (2)] 47–75 10 to 20 2.25–3.3 1.2–2.0 1.24–1.5 8–12 3.0–4.5 2.0–2.5
PWR II-S 45 10 3.0 1.5 1.3 — — —
PWR III-L [Note (3)] 12–15 10 1.2 1.2 1.15–1.2 8.0 4.5 2.0
BWR Mark I [Note (4)] — 10 to 20 — N/A — — N/A N/A
BWR Mark II — 10 to 25 — N/A — — N/A N/A
BWR Mark III 23 10 to 20 1.5 1.2 1.2 3.0 3.0 2.5
General Notes:
(a) PWR I-SI: small prestressed concrete containment; see Fig. 14.1.
(b) PWR I and II, S and I: small and large deformed-bar concrete containment; see Fig. 14.2.
© PWR I-L: large prestressed concrete containment; see Fig. 14.3.
(d) PWR III-L: large deformed bar concrete containment; see Fig. 14.4.
(e) PWR III-L: large hybrid steel and concrete containment; see Fig. 14.5.
(f) BWR Mark I: steel containment; see Fig. 14.6.
(g) BWR Mark I: deformed-bar concrete containment; see Fig. 14.7.
(h) BWR Mark II: deformed-bar concrete containment; see Fig. 14.8.
(i) BWR Mark III: deformed-bar concrete containment; see Fig. 14.9.
(j) BWR Mark III: hybrid concrete and steel containment; see Fig. 14.10.
Notes:
(1) The small PWR deformed-bar reinforced-concrete containment is similar in appearance to a large deformed-bar reinforced-concrete
containment, except smaller in size.
(2) The large deformed-bar reinforced-concrete containment, except for the elimination of the tendon access gallery, is similar in appearance
to the Fig. 14.3 PWR I-L reinforced, prestressed concrete containment; small 600 MWe 25%; large 1000 MWe 25%.
(3) Only two all-concrete ice condensers were constructed, D.C. Cook Units 1 and 2.
(4) Only one Mark I concrete containment was constructed, Brunswick Station.
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COMPANION GUIDE TO THE ASME BOILER & PRESSURE VESSEL CODE • 3
14.1.2 1 Subsection CC requirements are too stringent for confinement
structure design and construction. Reactor Buildings of
Mark I, Mark II containments may be defined as low leakage confinement
buildings. They are designed using the requirements of
ACI 349. In the definition of “Defense in Depth,” fuel element
cladding and RC loop (including RPV) are the primary barrier to
the release of radioactive materials. However, they are not defined
as containments. Hence the containment, as envisioned in
Subsection CC, implies primary containments.
Table 14.1.1 has been added in this update to show parameters
for the containments of standardized designs (proposed to be used
in the future reactors), e.g. ABWR, ESBWR, and EPR.
14.1.3 Types of Containment Systems
In the United States there are two basic types of commercial
nuclear power plants, both of which consist of water-cooled and
moderated reactors. The Boiling Water Reactor (BWR) in the
United States is a product of the General Electric Co. It typically
operates at a design pressure of 1,100 psi and 600F with the
direct cycle of coolant from the reactor coolant system through
the electricity-generating steam turbine. Pressurized Water
Reactor (PWR) systems in the United States have been provided
by the Combustion Engineering Co., the Babcock Wilcox Co.,
and the Westinghouse Electric Co. These types of reactor coolant
systems have a subcooled design pressure of 2,500 psi and 650F
with an intermediate steam generator heat exchanger between the
reactor coolant system (primary) and the steam system used to
drive the electricity-generating turbine system (secondary).
Commercial nuclear power plant containments are designed to
effectively contain without leakage the total primary coolant
system inventory released into the containment volume. Because
of the larger reactor coolant inventory at rated power in a BWRtype
commercial power reactor system, the containment for
BWRs employs a water-based pressure suppression system,
which includes a dry well that surrounds the reactor coolant system.
In the event of a loss-ofcoolant accident, the resultant steam
and air in the dry well blow down into a wet-well water pool
where the steam contained in the reactor coolant system condenses,
thereby significantly reducing the design pressure of the dryand
wet-well containment system.
Essentially two different sizes, 600 MWe and 1,000 MWE
25%, are used in both BWR and PWR plant design. Three distinct
types of containment structures, Mark I, Mark II, and Mark III,
are used for BWR pressure-suppression–type nuclear power plant
containments. Three different types of containment structures are
used for PWRs in the United States: dry, reduced pressure, and
pressure-suppression ice containments.
The Mark I steel containment system was in general use
between 1964 and 1972. The Mark I type of containment for a
BWR-type reactor consists mostly of a steel dry-well (light bulb)
and wet-well (torus) design as shown in Fig. 14.6 with a 4 in. gap
between the steel light bulb and concrete shield structure. The
single exception to the Mark I steel containment design was the
deformed bar–reinforced concrete design used in the Brunswick
Nuclear Power Station shown in Fig. 14.7. This containment type
for a BWR reactor system was followed by the Mark II, which
TABLE 14.1.1 CONCRETE CONTAINMENT DESIGN PARAMETERS
Design % Margin Basemat Wall Dome
Press. over Calc. Volume Diam Thickness Thickness Thickness
Type kPa P 103 m3 m m m m Remarks
Pressurized Water Reactors
PWR-Dry 310–433 10–20 51–93 35–46 2–3.5 1.0–1.4 0.6–0.8 Reinf &
Prestr Fig. 14.1
PWR-Ice 80–84 10 34 36 2.2 1.4 0.6 Reinforced
Fig. 14.2
EPR-Planned 427 — 80 47 — 1.3 1 PC-Containm
RC-Outer
Shield Fig. 14.6
Boiling Water Reactors
BWR- Mk I 420 10–20 4.5 DW Vary1 2–3 Below 1–2 Steel Head Reinforced
7 Torus DW Conc. Fig. 14.3
Base
BWR-MkII 380 10–25 7.1 DW Vary2 2–3 1.2–2 Steel Head RC & PC
7.5–8 Fig. 14.4
Wetwell
BWR- Mk III 207-DW 10–20 39.6 DW 37 1 1 .8 Reinforced
103-WW 11 WW Fig. 14.5
ABWR-Planned 310 15 7.35 DW 29 1.6 2 Steel Head Reinforced
5.96 WW Fig. 14.7
ESBWR-Planned 310 15 7.2 DW 18 UDW 5.1 2 Steel Head Reinforced
9.9 WW 5.6 LDW Fig. 14.8
The parameters shown are approximate and meant to be for general information
Explanation: DW – Drywell, WW – Wetwell, Vary1 – Inverted Light Bulb, Vary2 – Truncated Conical Drywell & Cylindrical Wetwell,
UDW – Upper Drywell includes Suppression Pools, LDW – Lower Drywell around Reactor Pedestal
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4 • Chapter 14
consists mostly of a concrete base mat, a concrete cylindrical section
housing the wet well, and a concrete conical section housing
the dry well, as shown in Fig. 14.8. This containment type design
for BWRs was popular in the United States from 1972 to 1974.
The most recent BWR containment structural design in the United
States after 1974 is the Mark III containment, which consists of
concrete deformed bar type as shown in Fig. 14.9 or a hybrid
deformed bar-reinforced concrete-slab base mat and a deformed
bar-reinforced concrete cylinder reinforcing a steel liner vapor
barrier in the wet-well region and a steel cylinder and dome above
the wet well region of the containment, as shown in Fig. 14.10.
It should be noted that there have been no new containment
designs employed in the United States since 1976. Other countries,
however, have constructed BWR designs since that time; in Japan,
its advanced BWR designs have used a modified version of the
Mark II type of containment.
Because of its relatively smaller reactor coolant system inventory,
most PWR-type reactors have generally employed the socalled
dry containment designed to simply contain the inventory
of the reactor coolant system. However, pressure suppression systems
of two types have been employed in some cases. The Stone
Webster Corp. designed several containments with reduced internal
air pressure typically in the range of 5–10 psig, which had the
effect of significantly reducing the design pressure in the containment
subjected to a design basis loss-of-coolant accident.
Most dry and reduced-pressure PWR concrete containment volumes
vary from 1.5 to 3.2 million ft3 depending on the volume of
the reactor coolant system size of the unit and the containment
design pressure. They consist of a reinforced and deformed-bar
concrete base mat typically 10 ft thick; either a reinforcedconcrete
4 ft thick deformed bar or a 3 ft thick prestressed
cylinder; and a 2–2 ft thick hemispherically and elliptically
shaped dome as shown in Figs. 14.1, 14.2, and 14.3. Two different
sizes of containments were used: one enclosing a 600 MWe
25% and 1,000 MWe 25%. Table 14.1 gives typical design
parameters for containments built in the United States.
In 1968 Westinghouse Electric Co. introduced a pressure-suppression–
type containment called the ice condenser. For the most
part the ice condenser containments are a hybrid concrete and
steel mixture having a reinforced-concrete, deformed-bar base
mat with steel cylinders and hemispherical domes, as shown in
Fig. 14.5. The single exception was the deformed-bar, reinforcedconcrete
ice condenser containment used for the D.C. Cook
power plant, as shown in Fig. 14.4.
The containment design pressures applied to most concrete containments
are usually statically applied as a function of the several
seconds it takes pressure to build up from the postulated initial discharge
of the reactor coolant system into containment. Exceptions
to this static application of load for containment design is the area
of the BWR wet wall that is exposed to pool swell, oscillationcondensation,
jet impingement, and chugging phenomena associated
with the dry-well blow down into the wet well, which are
dynamic events. These localized dynamic loads in BWR Mark
III–type containments have generally required the use of reinforced
concrete in the region of the wet well rather than steel
shells. The localized differential pressure loads across interior barriers
of the ice condenser and ice component doors impact, as well
as steam and air flow through the ice beds, are also dynamic in