U.S. patent application number 17/192784 was filed with the patent office on 2022-02-17 for containment structure and arrangement for nuclear reactor.
The applicant listed for this patent is TerraPower, LLC. Invention is credited to Brian C Johnson.
Application Number | 20220051815 17/192784 |
Document ID | / |
Family ID | 1000005840641 |
Filed Date | 2022-02-17 |
United States Patent
Application |
20220051815 |
Kind Code |
A1 |
Johnson; Brian C |
February 17, 2022 |
CONTAINMENT STRUCTURE AND ARRANGEMENT FOR NUCLEAR REACTOR
Abstract
A safety system for a nuclear reactor includes a first
containment structure and a second containment structure. The
double containment configuration is designed and configured to meet
all design basis accidents and beyond design basis events with
independent redundancy. The remaining systems that control
reactivity, decay heat removal, and fission product retention may
be categorized and designed as business systems, structures, and
components, and can therefore be designed and licensed according to
an appropriate quality grade for business systems.
Inventors: |
Johnson; Brian C; (Issaquah,
WA) |
|
Applicant: |
Name |
City |
State |
Country |
Type |
TerraPower, LLC |
Bellevue |
WA |
US |
|
|
Family ID: |
1000005840641 |
Appl. No.: |
17/192784 |
Filed: |
March 4, 2021 |
Related U.S. Patent Documents
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Application
Number |
Filing Date |
Patent Number |
|
|
63066778 |
Aug 17, 2020 |
|
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Current U.S.
Class: |
1/1 |
Current CPC
Class: |
G21C 13/10 20130101;
G21C 13/028 20130101; G21C 15/18 20130101; G21C 13/093
20130101 |
International
Class: |
G21C 13/10 20060101
G21C013/10; G21C 15/18 20060101 G21C015/18; G21C 13/028 20060101
G21C013/028; G21C 13/093 20060101 G21C013/093 |
Claims
1. A nuclear reactor, comprising: a nuclear reactor core; a reactor
vessel, the nuclear reactor core within the reactor vessel; a
reactivity control system that is categorized as a business system;
a decay heat removal system that is categorized as a business
system; a fission product retention system that is categorized as a
business system; a first containment structure surrounding the
reactor vessel, the first containment structure categorized as a
first safety system; and a second containment structure surrounding
the first containment structure, the second containment structure
categorized as a second safety system; wherein the first
containment structure and second containment structure are
sufficient to meet all design basis accidents and the second
containment structure provides redundancy to the first containment
structure.
2. The nuclear reactor of claim 1, wherein safety-related equipment
associated with the nuclear reactor consists essentially of the
first containment structure and the second containment
structure.
3. The nuclear reactor of claim 1, wherein the decay heat removal
system is not categorized as safety related equipment.
4. The nuclear reactor of claim 1, wherein the first containment
structure comprises an air-tight steel structure surrounded by
concrete.
5. The nuclear reactor of claim 1, wherein the second containment
structure comprises reinforced concrete.
6. The nuclear reactor of claim 5, wherein the second containment
structure comprises steel-reinforce concrete.
7. The nuclear reactor of claim 1, wherein the first containment
structure defines a first volume and the second containment
structure defines a second volume greater than the first
volume.
8. The nuclear reactor of claim 7, wherein a ratio of the second
volume to the first volume is greater than 10.
9. The nuclear reactor of claim 7, wherein a ratio of the second
volume to the first volume is greater than 20.
10. The nuclear reactor of claim 7, wherein a ratio of the second
volume to the first volume is greater than 50.
11. A safety system for a nuclear reactor consisting essentially of
a first containment structure surrounding a nuclear reactor vessel,
and a second containment structure surrounding the first
containment structure.
12. The safety system as in claim 11, wherein the first containment
structure comprises reinforced concrete.
13. The safety system as in claim 11, wherein the first containment
structure comprises a sealed steel structure.
14. The safety system as in claim 13, wherein the first containment
structure comprises an airlock through the first containment
structure to provide access to an interior portion of the first
containment structure.
15. The safety system as in claim 11, wherein the second
containment structure comprises reinforced concrete.
16. The safety system as in claim 15, wherein the second
containment structure comprises steel reinforced concrete.
17. The safety system as in claim 11, wherein the first containment
structure and the second containment structure are decoupled from
one another.
18. The safety system as in claim 11, wherein the first containment
structure and the second containment structure are designed to
eliminate any public safety consequence of a design basis
accident.
19. The safety system as in claim 11, wherein the first containment
structure defines a first volume and the second containment
structure defines a second volume greater than the first
volume.
20. The safety system as in claim 19, wherein a ratio of the second
volume to the first volume is greater than 10.
Description
CROSS-REFERENCE TO RELATED APPLICATIONS
[0001] This application claims the benefit of U.S. Provisional
Patent Application No. 63/066,778, filed Aug. 17, 2020, entitled
"MODULAR MANUFACTURE, DELIVERY, AND ASSEMBLY OF NUCLEAR REACTOR,"
the contents of which is incorporated herein by reference in its
entirety.
BACKGROUND
[0002] According to the United States Nuclear Regulatory
Commission, a containment structure is a gas-tight shell or other
enclosure around a nuclear reactor to confine fission products that
otherwise might be released to the atmosphere in the event of an
accident. Such enclosures are usually dome-shaped and made of
steel-reinforced concrete.
[0003] The containment structure must meet certain regulatory
guidelines and is usually the last line of defense in the event of
a design basis accident. Other safety systems usually include fuel
cladding, the reactor vessel, and the coolant system, among others.
These and other safety systems must be designed and constructed to
deal with design basis accidents and must pass regulatory licensing
requirements. These systems therefore are often complex, robust,
engineered with safety factors to withstand any of the numerous
design basis accidents. As a result, the engineering, construction,
and licensing of these safety-related system is often an arduous,
time and capital-intensive process. The safety systems associated
with a nuclear reactor are some of the primary drivers of
construction cost, construction time, and regulatory licensing
impediments.
[0004] It would be a significant advantage to simplify the systems,
construction times, and regulatory licensing requirements. These,
and other benefits, will become readily apparent from the following
description and attendant figures.
SUMMARY
[0005] According to some embodiments, the safety grade systems for
a nuclear reactor consist essentially of a first containment
structure and a second containment structure. For example, a
nuclear reactor may include a nuclear reactor core; a reactor
vessel, the nuclear reactor core within the reactor vessel; a
reactivity control system that is categorized as a business system;
a decay heat removal system that is categorized as a business
system; a fission product retention system that is categorized as a
business system; a first containment structure surrounding the
reactor vessel, the first containment structure categorized as a
first safety system; and a second containment structure surrounding
the first containment structure, the second containment structure
categorized as a second safety system; wherein the first
containment structure and second containment structure are
sufficient to meet all design basis accidents and the second
containment structure provides redundancy to the first containment
structure. As used herein, a system "categorized" as a business
system is a system that is designed, constructed, and licensed as a
business system and does not include a safety grade system. Safety
grade systems have particular regulations regarding their design,
construction, importance, and required redundancy. On the other
hand, business systems have much lower requirements in terms of
design, construction, importance, and redundancy.
[0006] In some cases, the safety-related equipment associated with
the nuclear reactor consists essentially of the first containment
structure and the second containment structure.
[0007] For example, in some embodiments, the decay heat removal
system is not categorized as safety related equipment. The first
containment structure may include an air-tight steel structure
surrounded by concrete. The second containment structure may
include reinforced concrete. In some instances, the second
containment structure is formed of steel-reinforce concrete.
[0008] According to some embodiments, the first containment
structure defines a first volume and the second containment
structure defines a second volume greater than the first volume. In
some cases, a ratio of the second volume to the first volume is
greater than 10, or greater than 20, or greater than 50, or greater
than 100.
[0009] According to some embodiments, a safety system for a nuclear
reactor consists essentially of a first containment structure
surrounding a nuclear reactor vessel, and a second containment
structure surrounding the first containment structure.
[0010] The first containment structure may be formed of reinforced
concrete. In some instances, the first containment structure may
include a sealed steel structure. The first containment structure
may include an airlock through the first containment structure to
provide access to an interior portion of the first containment
structure.
[0011] In some cases, the second containment structure comprises
reinforced concrete, and may include steel reinforced concrete.
[0012] According to some embodiments, the first containment
structure and the second containment structure are decoupled from
one another.
[0013] In some cases, the first containment structure and the
second containment structure are designed to eliminate any public
safety consequence of a design basis accident.
[0014] The first containment structure may define a first volume
and the second containment structure may define a second volume
greater than the first volume. In some cases, a ratio of the second
volume to the first volume is greater than 10, 20, 30, 40, 50, 80,
100 or more.
BRIEF DESCRIPTION OF THE DRAWINGS
[0015] FIG. 1 is a schematic representation of a containment
structure for a light water reactor ("LWR");
[0016] FIG. 2 is a categorized list of sample systems used with a
nuclear reactor, in accordance with some embodiments;
[0017] FIG. 3A shows example systems and functions that are safety
related, in accordance with some embodiments; and
[0018] FIG. 3B shows an example safety system for meeting design
basis accidents, in accordance with some embodiments.
DETAILED DESCRIPTION
[0019] This disclosure generally relates to containment structures
for nuclear reactors and a strategy for mitigating design basis
accidents. In some respects, the containment structures and
arrangements described herein significantly reduce the time and
cost of engineering, constructing, and licensing a nuclear reactor
as the containments described herein can efficiently withstand any
design basis accident ("DBA") and beyond design basis events
("BDBEs") with a large safety factor.
[0020] In the United States, the general design criteria is
governed by federal law and outlines the basic design criteria for
the containment structure including isolating lines penetrating the
containment wall. The containment building is generally an airtight
structure enclosing the nuclear reactor and is sealed from the
outside atmosphere. The containment building is typically built to
withstand the impact of a fully loaded passenger airliner without
breaching the structure.
[0021] The requirements for the containment structure are largely
dependent upon the size and type of reactor, the generation of the
reactor, and other specific needs of the nuclear plant. In typical
reactor installations, suppression systems are critical to safety
analysis and affect the design of the containment structure.
[0022] There is typically mandatory testing of the containment
structure and isolation systems, which provide for redundant
containment in the even of a design basis accident. In addition,
local leakage rate tests are performed regularly to identify
possible leakage points in an accident and to fix leakage paths. In
many cases, a nuclear plant operator is required to prove
satisfactory containment integrity before a restart following each
shutdown event.
[0023] In addition to containment structure, which in many cases,
is a last line of defense to a design basis accident, there are
numerous additional safety systems that need to be designed and
constructed to withstand and/or deal with a design basis accident.
For example, depending on the type of nuclear reactor, in the event
of an accident, safety systems are designed to shut down the
reactor, maintain it in a shutdown condition, and prevent the
release of radioactive material.
[0024] Examples of safety systems include control rods within the
core; a reactor protection system ("RPS"), emergency core cooling
systems ("ECCS"), decay heat removal systems, sodium-water reaction
protection systems (SWRPS), emergency electrical systems, standby
gas treatment system ("SGTS"), containment systems, and ventilation
systems. Of course, depending on the type of reactor, additional or
fewer systems may be required for regulatory licensing and the
above list is provided as representative. Generally, control rods
act as neutron absorbers and can be inserted into the core to
reduce neutron flux and terminate the critical nuclear reaction.
The reactor protection system is configured to terminate the
nuclear reaction by initiating a scram event, usually by inserting
negative reactivity mass into the core, which may be control rods.
The ECCS are designed to safely shut down a nuclear reactor in the
even of an accident and may include additional systems such as
depressurization systems, coolant injections systems, isolation
systems, and containment spray systems.
[0025] The emergency electrical system may include diesel
generators, batteries, grid power, or some other form of electrical
power so the safety systems can function as intended in the event
of an accident. The SGTS filters and pumps air from a secondary
containment and maintains a negative pressure within the secondary
containment to prevent the release of radioactive material. The
ventilation systems may be configured to remove radioactivity from
the air, thus protecting the control room and plant operators from
the effects of radioactivity.
[0026] In general, structures, systems, and components ("SSCs") are
classified as part of a defense in depth approach in the life cycle
of a nuclear plant. There is a graded approach to safety that
mandates that system having higher safety importance must be higher
quality, more robust and able to withstand failures, and more
resistant to hazards. The safety class has a direct impact on the
requirements for design, qualification, quality assurance, fault
tolerance, system architecture, and layout/location within the
nuclear island.
[0027] Many of the safety systems associated with a nuclear reactor
have a high safety significance, and therefore must be designed,
constructed, and licensed to very high quality standards to ensure
that even in the event of a design basis failure, there is minimal
risk of harm to the public or environment. As would be expected, in
many cases, the cost and time involved to design, construct, and
license a safety system may be somewhat tied to the safety
classification of the system or component.
[0028] According to the International Atomic Energy Agency
("IAEA"), systems are broadly divided into categories that perform
functions important to safety and those performing functions that
are not important to safety. Those systems important to safety are
those items where malfunction or failure could lead to radiation
exposure of site personnel or members of the public. The systems
important to safety include control of reactivity, removal of
residual heat, and confinement of radioactive materials.
[0029] The safety related systems are further categorized into
multiple classifications depending on their function and safety
importance, and in many cases, a tiered classification system for
safety related equipment includes 3 tiers. While there is not
currently an international harmonization of safety categorization,
the concepts described herein will suffice for any classification
system in local jurisdictions.
[0030] In order to provide redundancy, a primary means of
preventing accidents and mitigating the consequences of such
accidents is the application of defense in depth that provides for
diverse backup systems that are independent and redundant. This
ensures that no single safety layer, no matter how robust, is
exclusively relied upon to compensate for potential human or
mechanical failures.
[0031] With reference to FIG. 1, a typical containment building 100
for an LWR is illustrated. The containment building 100 is
typically formed of steel, concrete, and/or steel reinforced
concrete. The containment building 100 is designed to prevent the
uncontrolled release of radioactive material to the environment. In
some cases, the containment building is shaped to contain a
pressure increase within the containment building, such as by a
loss of coolant ("LOC") accident, and for this reason, is typically
shaped to be hemispherical, cylindrical, or a combination (e.g. a
domed cylinder).
[0032] In many cases, the current state of the art containment
structure includes a steel shell 102 surrounded by reinforced
concrete 104 that surrounds the nuclear reactor vessel and core
106.
[0033] With reference to FIG. 2, a high-level generic designation
of plant equipment is illustrated showing various categories of
nuclear plant equipment. At a high level, Plant Equipment 200 can
be broken down into categories that fit into either Safety Items
202 or Business Items 204. From a regulatory standpoint, Safety
Items 202 are those SSC's that promote or ensure the safe operation
of the nuclear reactor and prevent public harm. Other systems that
support the functioning of the reactor on a day to day basis and
are not specifically directed to safety can be categorized as
Business Items 204.
[0034] The Safety Items 202 can be further broken down into SSCs
that are Safety Related 206 versus those that are specific Safety
Systems 208. The Safety Systems 208 include systems such as
Protection Systems 210, Safety Actuation Systems 212, and Safety
Support Systems 214, among others. The SSCs that fit into any of
the Safety Item 202 category or subcategories generally must be
constructed to withstand and mitigate DBAs.
[0035] It should be appreciated that there are numerous SSCs,
including all the redundant systems that fall within the safety
items 202 classifications and therefore require adherence to
stringent licensing requirements. Because of the difficulties in
adhering to the stringent licensing requirements which were
mandated as a result of decades of experience with LWRs, it has
become difficult to apply the historical prescriptive methods to
more advanced reactor designs. The licensing requirements are not
necessarily directly applicable to next generation reactor designs
with their inherent safety features, and thus, many regulatory
authorities have either had to provide exemptions from some of the
requirements or deny licensing to more advanced reactor
designs.
[0036] As a result, in the United Sates, the NRC completed a
Licensing Modernization Project which culminated in a new approach
to licensing non-LWR reactor technologies. The new guidance reduces
the regulatory uncertainty within the industry and streamlines the
advanced reactor design and licensing process.
[0037] The finalized approach focuses on a technology-inclusive,
risk-informed, performance-based review process (rather than the
prior prescriptive based licensing approach) and is tailored to the
unique aspects of each advanced reactor design to provide a clear
and consistent review of its safety case. In short, the guidance
focuses on identifying licensing basis events; categorizing and
establishing performance criteria for SSCs, and evaluating the
safety margins of advanced reactor designs.
[0038] Even given the opportunities for increased regulatory
certainty, there are still significant obstacles to the
engineering, design, and licensing of nuclear reactor SSCs to meet
the licensing performance criteria. For example, it is possible to
design many, or even most, of the systems that were historically
categorized within the Safety Items 202 category or subcategories,
as Business Items 204 and therefore design those SSCs to a lower
threshold of design standard. By proactively dealing with all
contemplatable DBA's with other systems, many of the traditional
safety systems and their redundant systems can be eliminated, or
designed to a lesser standard, while still meeting all the
licensing requirement for DBAs and BDBEs.
[0039] With a clear performance-based licensing approach, there
arise opportunities to meet the performance-based criteria in an
efficient and cost-effective manner. For example, while the
fundamental safety functions continue to focus on reactivity
control, decay heat removal, and fission product retention, only
those systems and functions selected by the designer for responding
to DBAs and some high consequence BDBEs are properly categorized as
safety-related. While many advanced reactor designers are
accustomed to past licensing regulations and continue with robust
design of safety related SSCs, containment may not typically be
identified as a safety system necessary to meet DBA goals.
[0040] FIG. 3A illustrates a typical case of safety-related systems
that include a reactor vessel 302, a direct reactor auxiliary
cooling system ("DRACS") 304, and numerous SSCs 306a, 306b, 306n
located within or adjacent to the reactor vessel 302. The
containment building 308 is typically identified as not necessary
to meet DBA goals and is therefore not an identified safety system.
The non-safety related systems are shown in dashed outline while
the safety related systems are shown in solid line. As can be
imagined, there are numerous safety-related SSCs that must be
deigned to robust licensing standards.
[0041] However, with the paradigm shift to reactor licensing
requirements being technology-inclusive, risk-informed, and
performance-based, the licensing requirements now rely on
quantitative risk metrics to evaluate the risk significance of
events which leads to the formulation of performance targets on the
capability and reliability of SSCs to prevent and mitigate
accidents. This aligns the design and licensing efforts with the
safety objectives while providing greater safety margins.
[0042] As shown in FIG. 3B, according to some embodiments, the
containment can be identified as safety-related and can be designed
to meet all of the DBA goals and BDBE goals. That is, the
containment can be designed to meet all the performance targets to
prevent and mitigate accidents. Under some licensing regimes, SSCs
must be designed with the expectation of fission product release to
the containment structure. Therefore, providing a robust
containment structure and identifying the containment structure as
safety-related, it can be designed to meet the DBA and BDBE
conditions. Furthermore, according to some embodiments, two
containment structures can be identified as safety related and
thereby provide a redundant backup to all of the SSCs, which may
not be required to have a safety classification. The reactor vessel
and core 302 may continue to include a heat removal system, such as
a DRACS 404, but it may no longer need to be identified as
safety-related equipment. Similarly, equipment for reactivity
control, decay heat removal, and fission product retention 406a,
406b, 406c . . . 406n, may continue to be provided, but may no
longer be identified as safety-related.
[0043] According to the Licensing Modernization Project,
Anticipated Operational Occurrences ("AOOs") encompass anticipated
event sequences expected to occur one or more times during the life
of a nuclear power plant, which may include one or more reactor
modules. Event sequences with mean frequencies of
1.times.10.sup.-2/plant-year and greater are classified as AOOs.
AOOs take into account the expected response of all SSCs within the
plant, regardless of safety classification. Design Basis Events
("DBEs") encompass infrequent event sequences that are not expected
to occur in the life of a nuclear power plant, which may include
one or more reactor modules, but are less likely than AOOs. Even
sequences with mean frequencies of 1.times.10.sup.-4/plant-year to
1.times.10.sup.-2/plant-year are classified as DBEs. DBEs take into
account the expected response of all SSCs within the plant
regardless of safety classification. Beyond Design Basis Events
("BDBEs") are rare event sequences that are not expected to occur
in the life of a nuclear power plant, which may include one or more
reactor modules, and are less likely than a DBE. Event sequences
with mean frequencies of 5.times.10.sup.-7/plant-year to
1.times.10-4/plant-year are classified as BDBEs. BDBEs take into
account the expected response of all SSCs within the plant
regardless of safety classification.
[0044] According to some embodiments, a first containment structure
408 and a second containment structure 410 can be appropriately
designed as a double containment configuration to mitigate all
AOOs, DBAs, and BDBEs, resulting in an acceptable potential
accident consequence, which in nearly all cases, results in zero
public consequences. All of the dose requirements can be met with
two containment barriers which allow the remaining equipment to not
be safety classified, but rather, classified as business items for
plant product retention and normal reactor operation.
[0045] With this methodology, the categorization of the SSCs
reduces to two categories: (1) safety grade equipment, and (2)
business grade equipment, with the majority of the plant SSCs
fitting within the business grade equipment designation. According
to some embodiments, the first containment structure 408 and the
second containment structure 410 are the primary safety grade
systems. In some instances, the first containment structure 408 and
the second containment structure 410 are the only safety grade
systems, and are configured to perform both radionuclide retention
and allow sufficient heat transfer to the environment to inhibit
continuous heat build-up due to the decay heat load. In some
embodiments, there may be additional safety grade SSC's that help
to manage DBAs or BDBEs.
[0046] FIG. 3B illustrates a double containment structure
configuration in which a primary containment 408 structure
surrounds the nuclear reactor and a secondary containment structure
410 encompasses the primary containment structure. The safety grade
equipment can include a primary containment structure 408 and a
secondary containment structure 410. In some cases, the safety
grade equipment consists essentially of the primary containment
structure 408 and the secondary containment structure 410. The
primary containment structure 408 surrounds the nuclear reactor and
attached structures. The secondary containment structure 410 can be
constructed to encompass the primary containment structure 408. The
double containment configuration can be designed to exceed the
safety and licensing requirements for all DBAs and BDBEs, and
therefore, can be used to meet the licensing performance-based
criteria. The primary and secondary containment structures 408, 410
can be decoupled from one another such that an incident affecting
one structure is not transmitted to the other. Consequently, the
primary and secondary containment structures 408, 410 may provide
for decoupled and redundant safety systems to meet all DBA's and
BDBE's.
[0047] As indicated by the solid lines, the primary and secondary
containment structures 408, 410 are indicated as safety-related
equipment, and the remainder of the equipment, shown by dotted
lines which includes the reactor vessel 302, DRACS 404, and SSCs
406a . . . 406n within the reactor vessel, are not considered
safety related equipment, and can therefore be designed and
constructed to business grade equipment standards.
[0048] Of course, the non-safety related equipment may continue to
be needed for the reliable operation of the reactor, and the
paradigm shift is that the equipment necessary for the reliable
operation of the reactor is no longer relied upon to ensure public
safety. Of course, the reactor may continue to be designed to
control reactivity, reliably shut down and remove decay heat.
[0049] The primary and secondary containment structures may be
constructed similarly, or have different construction materials,
thicknesses, and characteristics. The primary and secondary
containment structures may be designed based upon deterministic and
probabilistic inputs that help drive design decisions. As an
example, the primary containment structure may be designed to
predominantly protect against internal hazards while the secondary
containment structure may be designed to primarily protect against
external hazards. In either case, postulated event sequences can be
used to set design criteria for the primary and secondary
containment structures to meet performance objectives. In other
words, the double containment structures can be designed to meet
any postulated event sequence consequences within the prescribed
dose limits.
[0050] In some cases, the containment structures may be formed of
any suitable steel, concrete, and may include fiber reinforced
concrete, steel reinforced concrete, geopolymer concrete, or other
suitable materials. One or more of the containment structures may
alternatively or additionally be formed of steel, and may
incorporate steel into a concrete matrix, or may be a steel-lined
concrete structure. In many cases, the primary and/or secondary
containment structures are sealed from the atmosphere. In some
cases, the secondary containment structure includes a sealed steel
structure surrounded by a missile shield, which may be formed of
any suitable material, such as concrete. The steel structure may be
isolated from the missile shield, or may be coupled to it. Where
the secondary containment structure may be configured to address
potential external hazards, the primary containment structure may
be configured to mitigate internal hazards and may be constructed
differently than the secondary containment structure. For example,
a hardened prestressed concrete building may be used as the outer
containment, while the inner containment may be a relatively thin
metal structure that is compatible with the coolant to ensure the
core remains covered assuming a failure of the primary coolant
system. In some cases, the inner containment may be a metal
structure. In some cases, the metal structure may have a wall
thickness averaging between 1 inch and 6 inches, or between 2
inches, and 4 inches.
[0051] In some embodiments, the primary containment structure may
be sealed and only permit access through an airlock to inhibit the
egress of radioactive material. The primary and or secondary
containment structure may have any suitable thickness, such as up
to 3 feet, or 4 feet, or 5 feet, or a greater thickness. In some
cases, the primary containment structure is a metal structure
configured to cover the core should the primary coolant system
fail. In some cases, the secondary containment structure is a
hardened structure and provides a volume larger than the primary
containment structure. According to some embodiments, the primary
containment structure defines a first volume and the secondary
containment structure defines a second volume, larger than the
first volume. A ratio of the second volume to the first volume may
be on the order of 1.5, 2, 3, 4, or 5 or more. In some embodiments,
the ratio of the second volume to the first volume is equal to or
greater than about 10, 20, 50, 80, or 100, or more. In some cases,
the difference in volume provides for separation between the
primary and secondary containment structures and provides a
significant volume for gas expansion should the first containment
structure fail by pressure rupture. As an example, a primary
containment structure may have an internal first volume on the
order of about 2,000 m.sup.3, and the second containment structure
may have a second volume on the order of about 100,000 m.sup.3.
[0052] In some cases, the primary containment structure may be
formed as a cylinder, and in some cases may have one or more
hemispherical ends. In some cases, the primary containment
structure may be spherical. In some embodiments, the secondary
containment structure may be generally rectangular, prismatic, or
any other building shape. In some cases, the secondary containment
structure may appear to be a normal building in terms of shape,
aspect ratio, building materials and the like. For example, the
rector hall can be configured as the second containment structure
and the reactor hall can be designed and built to safety grade
standards to provide a fully redundant safety system to the primary
containment structure.
[0053] In some cases, the secondary containment structure may be
fabricated as a metal building having a larger volume than a
primary containment building. The secondary containment building
may provide a low integral leakage rate and be configured to
contain any radionuclides from being released to the environment,
and the primary containment structure may be configured as a
hardened shield to protect the reactor from external hazards.
[0054] In some cases, the primary containment structure and
secondary containment structures are formed of similar, or the
same, materials and may have generally the same shape and
construction techniques, with a primary difference being the volume
of the secondary containment structure sized to completely
encapsulate the primary containment structure to provide a
redundant and decoupled safety system.
[0055] The result is a nuclear reactor licensing process that is
very efficient because each SSC is not needed to be designed or
evaluated for a safety case, but rather, the containment structures
can meet every safety case for all DBAs and BDBEs. As a further
result, even in the worst-case scenarios, there is no potential
harm to the public because the containment structures are designed
to mitigate any possible event sequence and avoid any public
accident consequence.
[0056] In some cases, residual decay heat can be handled by a DRACS
unit or additionally or alternatively, be handled between the
primary containment and secondary containment based on thermal
inertia and normal flow paths. In some embodiments, the secondary
containment may include a dedicated decay heat removal system that
is separate from any heat removal systems of the primary
containment structure. Another benefit of the proposed arrangement
is that a DRACS system is no longer needed as a primary safety
system, although may still be provided as a non-safety grade
system. Similarly, a SCRAM system is no longer needed as a primary
safety system. These systems may ultimately be provided, but they
are not necessary as safety systems and thereby do not need to be
designed or constructed to meet safety grade requirements.
[0057] In many cases, the primary containment and secondary
containment structures are independent from one another, thereby
providing full redundancy and depth in defense protection from any
postulated DBA or BDBE. The secondary containment 410, by its
nature of encompassing the primary containment structure 408, will
have a substantial volume, larger than the volume of the primary
containment structure, and can accept pressure conditions in the
event of a primary containment failure and pressure spikes.
[0058] The disclosure sets forth example embodiments and, as such,
is not intended to limit the scope of embodiments of the disclosure
and the appended claims in any way. Embodiments have been described
above with the aid of functional building blocks illustrating the
implementation of specified functions and relationships thereof.
The boundaries of these functional building blocks have been
arbitrarily defined herein for the convenience of the description.
Alternate boundaries can be defined to the extent that the
specified functions and relationships thereof are appropriately
performed.
[0059] The foregoing description of specific embodiments will so
fully reveal the general nature of embodiments of the disclosure
that others can, by applying knowledge of those of ordinary skill
in the art, readily modify and/or adapt for various applications
such specific embodiments, without undue experimentation, without
departing from the general concept of embodiments of the
disclosure. Therefore, such adaptation and modifications are
intended to be within the meaning and range of equivalents of the
disclosed embodiments, based on the teaching and guidance presented
herein. The phraseology or terminology herein is for the purpose of
description and not of limitation, such that the terminology or
phraseology of the specification is to be interpreted by persons of
ordinary skill in the relevant art in light of the teachings and
guidance presented herein.
[0060] The breadth and scope of embodiments of the disclosure
should not be limited by any of the above-described example
embodiments, but should be defined only in accordance with the
following claims and their equivalents.
[0061] Conditional language, such as, among others, "can," "could,"
"might," or "may," unless specifically stated otherwise, or
otherwise understood within the context as used, is generally
intended to convey that certain implementations could include,
while other implementations do not include, certain features,
elements, and/or operations. Thus, such conditional language
generally is not intended to imply that features, elements, and/or
operations are in any way required for one or more implementations
or that one or more implementations necessarily include logic for
deciding, with or without user input or prompting, whether these
features, elements, and/or operations are included or are to be
performed in any particular implementation.
[0062] The specification and annexed drawings disclose examples of
systems, apparatus, devices, and techniques that may provide
control and optimization of separation equipment. It is, of course,
not possible to describe every conceivable combination of elements
and/or methods for purposes of describing the various features of
the disclosure, but those of ordinary skill in the art recognize
that many further combinations and permutations of the disclosed
features are possible. Accordingly, various modifications may be
made to the disclosure without departing from the scope or spirit
thereof. Further, other embodiments of the disclosure may be
apparent from consideration of the specification and annexed
drawings, and practice of disclosed embodiments as presented
herein. Examples put forward in the specification and annexed
drawings should be considered, in all respects, as illustrative and
not restrictive. Although specific terms are employed herein, they
are used in a generic and descriptive sense only, and not used for
purposes of limitation.
[0063] Those skilled in the art will appreciate that, in some
implementations, the functionality provided by the processes,
systems, and arrangements discussed above may be provided in
alternative ways. The various methods, configurations, and
arrangements as illustrated in the figures and described herein
represent example implementations. From the foregoing, it will be
appreciated that, although specific implementations have been
described herein for purposes of illustration, various
modifications may be made without deviating from the spirit and
scope of the appended claims and the elements recited therein. In
addition, while certain aspects are presented below in certain
claim forms, the inventors contemplate the various aspects in any
available claim form. For example, while only some aspects may
currently be recited as being embodied in a particular
configuration, other aspects may likewise be so embodied. Various
modifications and changes may be made as would be obvious to a
person skilled in the art having the benefit of this disclosure. It
is intended to embrace all such modifications and changes and,
accordingly, the above description is to be regarded in an
illustrative rather than a restrictive sense.
* * * * *