US20220051815A1

US20220051815A1 - Containment structure and arrangement for nuclear reactor

Info

Publication number: US20220051815A1
Application number: US17/192,784
Authority: US
Inventors: Brian C Johnson
Original assignee: TerraPower LLC
Current assignee: TerraPower LLC
Priority date: 2020-08-17
Filing date: 2021-03-04
Publication date: 2022-02-17
Also published as: JP2023538343A; EP4197000A1; US20230323695A1; EP4197002A1; EP4197003A1; CN115917672A; KR20230049060A; CA3188367A1; KR20230050266A; US11746550B2; JP2023538174A; JP2023538175A; WO2022039798A1; US20240309665A1; US20220051811A1; JP7731379B2; KR20230049092A; WO2022039784A1; CA3177785A1; US12416175B2

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.

Description

CROSS-REFERENCE TO RELATED APPLICATIONS

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

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.
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.
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

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.
In some cases, the safety-related equipment associated with the nuclear reactor consists essentially of the first containment structure and the second containment structure.
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.
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.
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.
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.
In some cases, the second containment structure comprises reinforced concrete, and may include steel reinforced concrete.
According to some embodiments, the first containment structure and the second containment structure are decoupled from one another.
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.
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

FIG. 1 is a schematic representation of a containment structure for a light water reactor (“LWR”);

FIG. 2 is a categorized list of sample systems used with a nuclear reactor, in accordance with some embodiments;

FIG. 3A shows example systems and functions that are safety related, in accordance with some embodiments; and

FIG. 3B shows an example safety system for meeting design basis accidents, in accordance with some embodiments.

DETAILED DESCRIPTION

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.
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.
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.
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.
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.
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.
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.
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.
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.
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.
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.
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.
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).
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.
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.
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.
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.
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.
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.
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.
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.
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 306 a, 306 b, 306 n 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.
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.
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 406 a, 406 b, 406 c . . . 406 n, may continue to be provided, but may no longer be identified as safety-related.
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 1×10⁻²/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 1×10⁻⁴/plant-year to 1×10⁻²/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 5×10⁻⁷/plant-year to 1×10−4/plant-year are classified as BDBEs. BDBEs take into account the expected response of all SSCs within the plant regardless of safety classification.
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.
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.
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.
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 406 a . . . 406 n within the reactor vessel, are not considered safety related equipment, and can therefore be designed and constructed to business grade equipment standards.
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.
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.
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.
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 m³, and the second containment structure may have a second volume on the order of about 100,000 m³.
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.
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.
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.
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.
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.
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.
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.
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.
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.
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.
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.
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.

Claims

What is claimed is:

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.