WO2020239531A1

WO2020239531A1 - Containment for a pwr nuclear power plant

Info

Publication number: WO2020239531A1
Application number: PCT/EP2020/063934
Authority: WO
Inventors: Philip ELLINGHAM
Original assignee: Rolls-Royce Plc
Priority date: 2019-05-24
Filing date: 2020-05-19
Publication date: 2020-12-03
Also published as: GB201907315D0

Abstract

A containment which in use forms at least part of a PWR nuclear power plant in which a primary coolant circuit circulates pressurised water between a reactor pressure vessel containing fuel assemblies, one or more steam generators and a pressurizer. The containment includes: (i) one or more power plant components selected from the one or more steam generators and the pressurizer, (ii) a pressure-containing housing which contains the component(s) and which defines a release space whereby, in the event of a loss-of-coolant accident releasing the pressurised water from the component(s), the released water and steam formed therefrom is received by and contained within the housing, the housing having a floor beneath the component(s) and sidewalls extending upwards from the floor to a roof above the component(s), and (iii) a support structure for the component(s), wherein the support structure is attached to the sidewalls of the housing and provides bearing surfaces on which are located corresponding bearing surfaces of the component(s). The containment is configured such that the component(s) is elevated above the floor of the housing by the support structure, and a portion of the vertical load of the weight of the component(s) is transmitted into the sidewalls of the housing via the bearing surfaces of the component(s) and the support structure.

Description

TITLE:

CONTAINMENT FOR A PWR NUCLEAR POWER PLANT

Field of Disclosure

The present disclosure relates to a containment which in use forms part of a pressurised water reactor nuclear power plant.

Background

Nuclear power plants convert heat energy from the nuclear decay of fissile material contained in fuel assemblies into electrical energy. Pressurised water reactor (PWR) nuclear power plants have a primary coolant circuit which typically connects the following pressurised components: a reactor pressure vessel (RPV) containing the fuel assemblies; one or more steam generators; and one or more pressurizers. Coolant pumps in the primary circuit circulate pressurised water through pipework between these components. The RPV houses the nuclear reactor which heats the water in the primary circuit. The steam generator functions as a heat exchanger between the primary circuit and a secondary system where steam is generated to power turbines. The pressurizers maintain pressure of around 155 bar in the primary circuit.

These components of a PWR plant are conventionally housed in an airtight containment building, which has to be able to contain and retain any primary coolant water, and steam formed therefrom, released during a loss-of-coolant accident from the primary coolant circuit or any of the components. The containment building also has to be able to contain and retain any accidental release of radioactivity, such as may be associated with a loss-of- coolant accident.

The steam generators and pressurizers are conventionally supported using vertical structural supports which attach near the bottom of the steam generator/pressurizer. Conventionally, a single vertical structural support may extend from the bottom centre of a steam generator or pressurizer. Alternatively, a plurality of vertical structural supports may connect elsewhere on the domed lower bottom portion of a steam generator or pressurizer and extend vertically downwards to the containment floor. Thermal expansion of the pipes from the reactor vessel to the steam generator or pressurizer creates a force that pushes the steam generator or pressurizer away from the reactor vessel. This movement is conventionally accommodated by flex of the vertical supports, or rotation at the joints between the vertical support and the vessel, and between the vertical support and a concrete floor of the containment building. By connecting a central vertical structural support to the base of the steam generator or pressurizer and not around the circumference, thermal expansion of the steam generator or pressurizer may also accommodated for.

However, because these supports must be present beneath the steam generator/pressurizer to function, other components, such as primary coolant circuit pumps, are typically located laterally of the steam generator/pressurizer, e.g. between the RPV and the steam

generators. This increases the distance of the steam generators from the centre of the containment building and increases the necessary size of the building. This in turn increases the cost and build time of the power plant.

Summary

According to a first aspect there is provided a containment which in use forms at least part of a PWR nuclear power plant in which a primary coolant circuit circulates pressurised water between a reactor pressure vessel containing fuel assemblies, one or more steam generators and a pressurizer, the containment including: (i) one or more power plant components selected from the one or more steam generators and the pressurizer, (ii) a pressure-containing housing which contains the component(s) and which defines a release space whereby, in the event of a loss-of-coolant accident releasing the pressurised water from the component(s), the released water and steam formed therefrom is received by and contained within the housing, the housing having a floor beneath the component(s) and sidewalls extending upwards from the floor to a roof above the component(s), and (iii) a support structure for the component(s), wherein the support structure is attached at least to the sidewall of the housing and provides bearing surfaces on which are located

corresponding bearing surfaces of the component(s);

wherein the containment is configured such that the component(s) is elevated above the floor of the housing by the support structure, and at least a portion of the vertical load of the weight of the component(s) is transmitted into the sidewall of the housing via the bearing surfaces of the component(s) and the support structure.

The support structure thus frees up a volume beneath the component where, e.g. a coolant pump of the primary coolant circuit and/or pipework of the primary coolant circuit can be located. This allow the steam generators and the pressurizer to be brought closer to the RPV, and in turn allows the containment to be smaller, saving time and money in plant construction.

According to a second aspect there is provided a PWR nuclear power plant in which a primary coolant circuit circulates pressurised water between a reactor pressure vessel containing fuel assemblies, one or more steam generators and a pressurizer, the plant having one or more of the containments of the first aspect for the one or more steam generators and the pressurizer.

The pressure-containing housing of the first aspect may contain just one of the one or more steam generators and the pressurizer. In this case, according to a third aspect, there is provided a PWR nuclear power plant in which a primary coolant circuit circulates pressurised water between a reactor pressure vessel containing fuel assemblies, one or more steam generators and a pressurizer, the plant having respective of the containments of the first aspect for the one or more steam generators and the pressurizer.

Optional features of the present disclosure will now be set out. These are applicable singly or in any combination with any aspect of the present disclosure.

The bearing surfaces of the component(s) and the support structure may be configured to accommodate relative sliding movement therebetween. This can be caused by thermal expansion and contraction of the component(s) and/or by lateral movement of the component(s) relative to the RPV.

The bearing surfaces of the component may be positioned on two opposing lateral sides of the component, wherein a line between the opposing lateral sides is perpendicular to a line from the centre of the component to the centre of the reactor pressure vessel when viewed from above.

The support structure and bearing surfaces may allow tilting of the component, e.g. as thermal expansion of the pipework pushes on the bottom portion of the component, the frame and bearing surface may allow some rotation of about the bearing contact point.

The support structure may extend partially or completely around the circumference of the component. Alternatively, the support structure may extend on at least two opposing lateral sides of the containment. In some embodiments, the support surface may comprise two elongate bearing surfaces aligned parallel to the line between the component and the reactor pressure vessel when viewed from above.

The support structure may be spaced from the component to allow the component to expand or contract from thermal expansion. The bearing surfaces may be oversized with enough overlap to maintain a bearing surface connection as the component expands and contracts. The bearing surfaces may overlap a distance that corresponds to the maximum expansion in the radial direction from the centre of the component, from cold to peak running temperature.

The bearing surface of the component may comprise a smooth planar bearing surface or a roller bearing. The bearing surface of the support structure may comprise a smooth planar surface or a roller bearing.

The interface between the bearing surfaces of the support frame and the component bearing surface may be lubricated. The lubrication may comprise a low friction coating, a dry lubricant or an oil-based lubricant.

The support structure may be a frame anchored in the sidewalls of the housing. For example, the frame may be a steel frame. The frame may be embedded or otherwise integrated into the, typically concrete, sidewalls of the housing.

The support structure may be entirely supported by the sidewalls of the housing. The vertical load of substantially the entire weight of the component may be transmitted into the sidewalls of the housing via the bearing surfaces of the component. Alternatively, the support structure may be at least partially supported by the sidewalls of the housing. A portion of the vertical load of the weight of the component may be transmitted into the sidewalls of the housing via the bearing surfaces of the component. A second portion of the load may be transmitted in to a second support structure. The second support structure positioned laterally of the component i.e. not beneath the component. A second support structure may comprise any an internal wall positioned within the housing. This may be, for example, a wall of a containment pool, a support structure supporting a working floor level or a dedicated support structure to support the support structure.

The bearing surfaces of the component(s) may be provided by one or more flanges externally projecting from an outer containment skin of the or each component. One or more mechanical snubbers may extend between the housing and the component(s) to provide seismic restraint to the component(s) while accommodating normal relative movement between the housing and the component(s). Conveniently, these snubbers may be located above the support structure.

The or each component may have one or more hatches allowing access thereinto, the hatches being located below the bearing surfaces of that component.

The containment may further include pipework of the primary coolant circuit, the pipework extending downwards from the or each component into a volume beneath that component produced by its elevation above the floor of the housing. The containment may further include a coolant pump of the primary coolant circuit, the pump being located in the volume beneath that component.

The pressure-containing housing may contain just one of the one or more steam generators and the pressurizer.

The present invention may comprise or be comprised as part of a nuclear reactor power plant (referred to herein as a nuclear reactor). In particular, the present invention may relate to a Pressurized water reactor. The nuclear reactor power plant may have a power output between 250 and 600 MW or between 300 and 550 MW.

The nuclear reactor power plant may be a modular reactor. A modular reactor may be considered as a reactor comprised of a number of modules that are manufactured off site (e.g. in a factory) and then the modules are assembled into a nuclear reactor power plant on site by connecting the modules together. Any of the primary, secondary and/or tertiary circuits may be formed in a modular construction.

The nuclear reactor of the present disclosure may comprise a primary circuit comprising a reactor pressure vessel; one or more steam generators and one or more pressurizer. The primary circuit circulates a medium (e.g. water) through the reactor pressure vessel to extract heat generated by nuclear fission in the core, the heat is then to delivered to the steam generators and transferred to the secondary circuit. The primary circuit may comprise between one and six steam generators; or between two and four steam generators; or may comprise three steam generators; or a range of any of the aforesaid numerical values. The primary circuit may comprise one; two; or more than two pressurizers. The primary circuit may comprise a circuit extending from the reactor pressure vessel to each of the steam generators, the circuits may carry hot medium to the steam generator from the reactor pressure vessel, and carry cooled medium from the steam generators back to the reactor pressure vessel. The medium may be circulated by one or more pumps. In some

embodiments, the primary circuit may comprise one or two pumps per steam generator in the primary circuit.

In some embodiments, the medium circulated in the primary circuit may comprise water. In some embodiments, the medium may comprise a neutron absorbing substance added to the medium (e.g., boron, gadolinium). In some embodiments the pressure in the primary circuit may be at least 50, 80 100 or 150 bar during full power operations, and pressure may reach 80, 100, 150 or 180 bar during full power operations. In some embodiments, where water is the medium of the primary circuit, the heated water temperature of water leaving the reactor pressure vessel may be between 540 and 670 K, or between 560 and 650 K, or between 580 and 630 K during full power operations. In some embodiments, where water is the medium of the primary circuit, the cooled water temperature of water returning to the reactor pressure vessel may be between 510 and 600k, or between 530 and 580 K during full power operations.

The nuclear reactor of the present disclosure may comprise a secondary circuit comprising circulating loops of water which extract heat from the primary circuit in the steam generators to convert water to steam to drive turbines. In embodiments, the secondary loop may comprise one or two high pressure turbines and one or two low pressure turbines.

The secondary circuit may comprise a heat exchanger to condense steam to water as it is returned to the steam generator. The heat exchanger may be connected to a tertiary loop which may comprise a large body of water to act as a heat sink.

The reactor vessel may comprise a steel pressure vessel, the pressure vessel may be from 5 to 15 m high, or from 9.5 to 11.5 m high and the diameter may be between 2 and 7 m, or between 3 and 6 m, or between 4 to 5 m. The pressure vessel may comprise a reactor body and a reactor head positioned vertically above the reactor body. The reactor head may be connected to the reactor body by a series of studs that pass through a flange on the reactor head and a corresponding flange on the reactor body.

The reactor head may comprise an integrated head assembly in which a number of elements of the reactor structure may be consolidated into a single element. Included among the consolidated elements are a pressure vessel head, a cooling shroud, control rod drive mechanisms, a missile shield, a lifting rig, a hoist assembly, and a cable tray assembly.

The nuclear core may be comprised of a number of fuel assemblies, with the fuel assemblies containing fuel rods. The fuel rods may be formed of pellets of fissile material. The fuel assemblies may also include space for control rods. For example, the fuel assembly may provide a housing for a 17 x 17 grid of rods i.e. 289 total spaces. Of these 289 total spaces, 24 may be reserved for the control rods for the reactor, each of which may be formed of 24 control rodlets connected to a main arm, and one may be reserved for an instrumentation tube. The control rods are movable in and out of the core to provide control of the fission process undergone by the fuel, by absorbing neutrons released during nuclear fission. The reactor core may comprise between 100 - 300 fuel assemblies. Fully inserting the control rods may typically lead to a subcritical state in which the reactor is shutdown. Up to 100% of fuel assemblies in the reactor core may contain control rods.

Movement of the control rod may be moved by a control rod drive mechanism. The control rod drive mechanism may command and power actuators to lower and raise the control rods in and out of the fuel assembly, and to hold the position of the control rods relative to the core. The control rod drive mechanism rods may be able to rapidly insert the control rods to quickly shut down (i.e. scram) the reactor.

The primary circuit may be housed within a containment structure to retain steam from the primary circuit in the event of an accident. The containment may be between 15 and 60 m in diameter, or between 30 and 50 m in diameter. The containment structure may be formed from steel or concrete, or concrete lined with steel. The containment may house one or more lifting devices (e.g. a polar crane). The lifting device may be housed in the top of the containment above the reactor pressure vessel. The containment may contain within or support exterior to, a water tank for emergency cooling of the reactor. The containment may contain equipment and facilities to allow for refuelling of the reactor, for the storage of fuel assemblies and transportation of fuel assemblies between the inside and outside of the containment.

The power plant may contain one or more civil structures to protect reactor elements from external hazards (e.g. missile strike) and natural hazards (e.g. tsunami). The civil structures may be made from steel, or concrete, or a combination of both. Brief Description of the Drawings

Embodiments will now be described by way of example only, with reference to the Figures, in which:

Figure 1 is a schematic diagram of a PWR nuclear power plant; and

Figure 2 shows schematically the bottom half of a silo for a steam generator of the plant of Figure 1.

Detailed Description

Figure 1 is a schematic diagram of a PWR nuclear power plant 10. An RPV 12 containing fuel assemblies is centrally located in the plant. Clustered around the RPV are three steam generators 14 connected to the RPV by pipework 16 of the pressurised water, primary coolant circuit. Coolant pumps 18 circulate pressurised water around the primary coolant circuit, taking heated water from the RPV to the steam generators, and cooled water from the steam generators to the RPV.

A pressurizer 20 maintains the water pressure in the primary coolant circuit at about 155 bar.

In the steam generators 14, heat exchangers transfer heat from the pressurised water to feed water circulating in pipework 22 of a secondary coolant circuit, thereby producing steam which is used to drive turbines which in turn drive an electricity-generator. The steam is then condensed before returning to the steam generators.

In one arrangement of the plant 10, each of the RPV 12, steam generators 14 and pressurizer 20 is contained in a respective pressure-containing silo. This makes each silo significantly smaller and easier to fabricate than a conventional containment building for the whole plant. Figure 2 shows schematically the bottom half of the silo 24 for one of the steam generators 14. The silo has a concrete floor and vertically extending sidewalls, which extend up to a domed roof (not shown). The concrete may be of reinforced construction and/or may have a steel inner liner. The interior volume of the silo defines a cylindrical release space which, in the event of a loss-of-coolant accident releasing the pressurised water from the steam generator, contains the released water and steam formed therefrom within the silo.

A flange 26 is welded onto and projects outwards from the containment skin of the steam generator 14 to provide bearing surfaces at a bottom side thereof. Anchored to the sidewalls of the silo 24 is a steel support frame 28 which provides corresponding bearing surfaces at an upper side thereof. Conveniently, the support frame can be integrated and embedded into the concrete of the silo. The steam generator is positioned in the silo such that the bearing surfaces of the steam generator are located on the corresponding bearing surfaces of the support frame. In this way, the vertical load of substantially the entire weight of the steam generator is transmitted into the sidewalls of the silo via these bearing surfaces.

Moreover, the bearing surfaces can accommodate relative sliding movement therebetween caused by thermal expansion and contraction of the steam generator, as well as by lateral movement of the steam generator relative to the RPV 12 due to thermal expansion and contraction of the pipework 16 of the of the primary coolant circuit.

There is therefore no need for any weight-bearing structure beneath the steam generator 14, which can be elevated above the floor of the silo 24. The pipework 16 of the primary coolant circuit can thus enter the steam generator from below, and a pump 18 for the circuit can also be located in the space created in the bottom of the silo beneath the steam generator.

These adjustments allow the steam generator to be brought closer to the RPV 12.

Similar considerations apply to the other steam generators 14, the pressurizer 20 and their respective silos, with a result that the overall footprint of the plant can be reduced, producing savings in construction cost and time.

Conveniently, the flange 26 can be located above any access hatch(es) 30 into the steam generator/pressurizer 14, 20. The hatch is then not obscured by the flange, and can be accessed via the space below the respective component.

One or more mechanical snubbers (not shown) can extend between the sidewalls of the silo 24 and the respective component 14, 20 to provide seismic restraint to the component while accommodating normal relative movement between the housing and the component(s). Conveniently, these snubbers may be located above the flange 26 to avoid impinging on any of the space created below.

Although descried above in respect of an arrangement having individual silos for each of the main components of the plant, aspects and embodiments may be applied to other arrangements, such as for example ones in which a single pressure-tight containment building houses all of the RPV 12, steam generators 14 and pressurizer 20. In this case, the support frames for the different components may be interconnected to strengthen the overall support structure.

Aspects and embodiments of the containment and PWR nuclear power plant may be applicable to a small modular nuclear reactor. A small modular nuclear reactor is a nuclear power plant designed such that a significant number of power plant components can be manufactured and preassembled in a factory. The components are then connected on site. The small modular reactor concept avoids the need for onsite manufacturing or complex assembly of components on site. This saves build time and construction costs.

The component of the present invention may be manufactured off site. The component may be manufactured with the corresponding bearing surfaces incorporated into the component. Further, the support structure may also be manufactured off site and optionally connected to the component off site or shipped with the component.

It will be understood that the invention is not limited to the embodiments above-described and various modifications and improvements can be made without departing from the concepts described herein. Except where mutually exclusive, any of the features may be employed separately or in combination with any other features and the disclosure extends to and includes all combinations and sub-combinations of one or more features described herein.

Claims

CLAIMS We claim:

1. A containment which in use forms at least part of a PWR nuclear power plant (10) in which a primary coolant circuit circulates pressurised water between a reactor pressure vessel (12) containing fuel assemblies, one or more steam generators (14) and a pressurizer (20), the containment including: (i) one or more power plant components selected from the one or more steam generators and the pressurizer, (ii) a pressure-containing housing (24) which contains the component(s) and which defines a release space whereby, in the event of a loss-of-coolant accident releasing the pressurised water from the component(s), the released water and steam formed therefrom is received by and contained within the housing, the housing having a floor beneath the component(s) and sidewalls extending upwards from the floor to a roof above the component(s), and (iii) a support structure (28) for the component(s), wherein the support structure is attached to the sidewall of the housing and provides bearing surfaces on which are located corresponding bearing surfaces of the component(s);

wherein the containment is configured such that the component(s) is elevated above the floor of the housing by the support structure (28), and a portion of the vertical load of the weight of the component(s) is transmitted into the sidewalls of the housing (24) via the bearing surfaces of the component(s) and the support structure.

2. The containment of claim 1 , wherein the bearing surfaces of the component(s) and the support structure (28) are configured to accommodate relative sliding movement therebetween.

3. The containment of claim 1 , wherein the support structure (28) is a frame anchored in the sidewall of the housing.

4. The containment of any one of the previous claims wherein the bearing surfaces of the component(s) are provided by one or more flanges (26) externally projecting from an outer containment skin of the or each component.

5. The containment of any one of the previous claims further including one or more mechanical snubbers which extend between the housing and the component(s) to provide seismic restraint to the component(s) while accommodating normal relative movement between the housing and the component(s).

6. The containment of any one of the previous claims wherein the or each component has one or more hatches (30) allowing access thereinto, the hatches being located below the bearing surfaces of that component.

7. The containment of any one of the previous claims further including pipework (16) of the primary coolant circuit, the pipework extending downwards from the or each component into a volume beneath that component produced by its elevation above the floor of the housing.

8. The containment of claim 7 further including a coolant pump (18) of the primary coolant circuit, the pump being located in the volume beneath that component.

9. The containment of any one of the previous claims wherein the pressure-containing housing contains just one of the one or more steam generators (14) and the pressurizer (20).

10. The containment of any one of the previous claims wherein the support structure is entirely supported by the sidewalls and the vertical load of substantially the entire weight of the component(s) is transmitted into the sidewalls of the housing via the bearing surfaces of the component(s) and the support structure.

11. A PWR nuclear power plant in which a primary coolant circuit circulates pressurised water between a reactor pressure vessel containing fuel assemblies, one or more steam generators and a pressurizer, the plant having one or more of the containments of any one of claims 1 to 8 for the one or more steam generators and the pressurizer.

12. A PWR nuclear power plant (10) in which a primary coolant circuit circulates pressurised water between a reactor pressure vessel (12) containing fuel assemblies, one or more steam generators (14) and a pressurizer (20), the plant having respective of the containments of claim 9 for the one or more steam generators and the pressurizer.