HK1197491B - Submerged containment vessel for a nuclear reactor - Google Patents

Description

Submerged containment vessel for nuclear reactor

The present application is a divisional application of invention patent application No. 200880118524.3 entitled "submerged containment vessel for nuclear reactor" filed on 6.11.2008.

Technical Field

The present invention relates to a system for removing decay heat from a nuclear reactor.

Background

In order to obtain cheap and reliable energy sources, some nuclear reactors have been designed with the aim of passive operation. In these passive systems, laws of physics are applied to ensure that safe operation of the nuclear reactor is maintained during normal operation, or even in emergency situations, for at least some predetermined period of time without operator intervention or supervision. One goal of the passive operating system is to minimize the number of motors, pumps, or other electrical or mechanical devices that are conventionally relied upon to operate a nuclear reactor.

The multipurpose Small Light Water Reactor (Multi-Application Small Light Water Reactor) project, implemented under the support of the Idaho National Engineering and environmental Laboratory (Idaho National Engineering and environmental Laboratory), NEXANT and Oregon State University nuclear energy Engineering (nuclear Engineering Department of organic State University), seeks to develop a safe and economical natural Light Water Reactor. FIG. 1 illustrates a nuclear reactor design 20 from this project.

The nuclear reactor design 20 includes a reactor core 6 surrounded by a reactor vessel 2. Water 10 in the reactor vessel 2 surrounds the reactor core 6. The reactor core 6 is also placed in a shroud (shroud)22, the shroud 22 surrounding the reactor core 6 around the sides of the reactor core 6. As the water 10 is heated by the reactor core 6 due to a fission event, the water 10 is directed from the shield 22 upwardly into an annular space (annuus) 23 above the reactor core 6 and out a riser 24. This causes further water 10 to be drawn into the shield 22 to be correspondingly heated by the reactor core 6, drawing more water 10 into the shield 22. The water 10 discharged from the riser 24 is cooled and guided toward the outside of the reactor vessel 2, and then returned to the bottom of the reactor vessel 2 by natural circulation. As the water 10 is heated, pressurized steam 11 is generated in the reactor vessel 2.

The heat exchanger 35 circulates feedwater and steam in the auxiliary cooling system 30 to produce electricity with the turbine 32 and the generator 34. The feed water flows through the heat exchanger 35 and becomes superheated steam (superheated steam). The auxiliary cooling system 30 includes a condenser 36 and feed pump 38. The steam and feedwater in the auxiliary cooling system 30 are isolated from the water 10 in the reactor vessel 2 so that they must not mix or come into direct contact with each other.

The reactor vessel 2 is surrounded by a containment vessel (containment vessel) 4. The containment vessel 4 is placed in a pool of water 16. The pool 16 and containment vessel 4 are located below the ground 28 in the reactor bay 26. Containment vessel 4 does not allow any water or steam from reactor vessel 2 to escape into pool 16 or the surrounding environment. In an emergency situation, steam 11 is vented from the reactor vessel 2 through the steam valve 8 into the upper half 14 of the containment vessel 4, and the water 10 is flashed as it is released through a submerged discharge valve (blowdown valve)18 located in the suppression pool 12. The suppression pool 12 includes subcooled water. Thus, by releasing steam 11 to containment vessel 4 through steam valve 8, and water 10 to containment vessel 4 through vent valve 18, the overpressure of reactor vessel 2 is reduced. The release rate of steam 11 and water 10 to containment vessel 4 varies depending on the pressure within reactor vessel 2. Decay heat is removed from the nuclear reactor core 6 by a combination of condensation of the steam 11 and energy transfer of the water 10 to the suppression pool 12.

In the event of loss of coolant or pipeline rupture in the containment vessel 4, the water in the suppression pool 12 provides pressure suppression as well as liquid makeup (liquid makeup) performance. However, this also means that the electrical and mechanical components within containment vessel 4 are often subjected to corrosive environments, which cause reliability problems. When placed in a wet or humid environment, the insulation surrounding the reactor vessel 2 loses part of its insulating properties and may need to be replaced periodically. Expensive and exotic (exotic) materials may be used for the insulation of the reactor vessel. In addition, electrical and mechanical components must be maintained, monitored and inspected to ensure their continued operational reliability.

The present invention addresses these and other problems.

Drawings

FIG. 1 illustrates a nuclear power system as known in the art.

FIG. 2 illustrates a novel power module assembly including an internally dry containment.

FIG. 3 illustrates the power module assembly of FIG. 2 during an emergency operation.

FIG. 4 illustrates an exemplary condensation rate at which water vapor is released into the containment vessel.

FIG. 5 illustrates an exemplary pressure fluctuation in the containment vessel under an overpressure event.

FIG. 6 illustrates an alternative embodiment of a power module assembly including a containment vessel having cooling fins.

FIG. 7 illustrates an embodiment of a power module assembly including multiple containment regions.

FIG. 8 illustrates a novel method of cooling a power module assembly.

Disclosure of Invention

A power module assembly is disclosed herein as including a reactor core immersed in a coolant, and a reactor vessel containing the coolant and the reactor core. An internally dry containment vessel is submerged in the liquid and substantially surrounds the reactor vessel in a gaseous environment. During an overpressure event, the reactor vessel is configured to release the coolant into the containment vessel and remove decay heat of the reactor core by condensation of the coolant on the interior walls of the containment vessel.

A nuclear reactor module is disclosed herein as including a containment vessel designed to inhibit the release of liquids, and a reactor vessel mounted inside the containment vessel, wherein the exterior surface of the reactor vessel is exposed to sub-atmospheric pressure. The nuclear reactor module also includes a reactor core submerged in the liquid, and a steam outlet (steam vent) connected to the reactor vessel, wherein the steam outlet is configured to vent steam into the containment vessel when the reactor core becomes superheated.

Disclosed herein is a method of cooling a nuclear reactor, wherein the method comprises: in the event of a high pressure event within a reactor vessel, the nuclear reactor is emergency shutdown, and coolant is released to a containment region (containment region) located between the containment and the reactor vessel. The containment region surrounds the reactor vessel and is substantially dry prior to the high pressure event. The method further comprises the following steps: condensing the coolant on an inner wall of the containment vessel; transferring decay heat to a liquid medium surrounding the containment vessel; and maintaining the pressure in said containment region within design limits by condensation of the coolant on the inner wall.

The present invention will be more readily understood from the following detailed description of a preferred embodiment thereof, taken with reference to the accompanying drawings.

Detailed Description

Conventional nuclear facilities are expensive to license and build, with large upfront investment costs and delayed return on profit. In addition to energy cost considerations, efficiency requirements, and reliability issues, nuclear reactor design today must also take into account issues of nuclear spread, terrorist activity, and enhanced awareness of environmental management.

Developing countries that could otherwise benefit greatly from nuclear power often employ other energy sources, such as coal, gas or hydroelectric generators that produce significant pollution or have other harmful environmental effects. These developing countries may not have the technical or natural resources that enable them to build nuclear power plants. Countries that have developed nuclear power may be hesitant to introduce these technologies into developing countries for fear of loss of control of nuclear materials or technologies.

Passive safety nuclear power systems help address some of the concerns discussed above. Further system improvements and innovative designs are expected to open up a new era of nuclear power as the main energy source of global feasibility.

FIG. 2 illustrates a novel power module assembly 50 that includes an internally dry containment vessel 54. The containment vessel 54 is cylindrical in shape and has spherical upper and lower ends. The entire power module assembly 50 may be submerged in the water basin 16 which serves as an active heat sink. The containment vessel 54 may be welded or otherwise sealed from the environment such that liquids and gases cannot escape from the power module assembly 50 or enter the power module assembly 50. The containment vessel 54 may be bottom supported, top supported, or supported about its center. Supporting the containment vessel 54 at the top may facilitate maintenance and removal of the power module assembly 50 from the pool of water 16.

In one embodiment, the containment vessel 54 is suspended in the pool of water 16 by one or more mounting connections 80. The fitting connection 80 may be attached to an upper portion of the containment vessel 54. The mounting connection 80 may be a rigid or flexible member that helps position the containment vessel 54 substantially in the center of the pool of water 16. During seismic activity, such as an earthquake, the pool of water 16 acts as a protective cushion (cushion) around the containment vessel 54 to avoid damage that may result if the containment vessel 54 comes into contact with the reactor bay 26. Flexible mounting connections, such as chains or cables, attached to the walls of the reactor bay 26 may reduce the amount of vibration or stress that may otherwise be transmitted to the containment vessel 54. In one embodiment, a flexible tie-down connection is attached to the bottom of the containment vessel 54 to reduce rocking or lateral movement. The power module assembly 50 may be arranged to float in the basin 16 to minimize support requirements and provide shock resistance. A support pedestal may be provided at the bottom of the containment vessel 54 to support the power module assembly 50 in an upright position.

A reactor vessel 52 is located or mounted inside a containment vessel 54. The inner surfaces of the reactor vessel 52 may be exposed to a humid environment including the coolant 100 or a liquid such as water, and the outer surfaces may be exposed to a dry environment such as air. The reactor vessel 52 may be made of stainless steel or carbon steel, may include cladding, and may be supported within the containment vessel 54.

The power module system 50 may be sized so that it can be transported on a rail car. For example, the containment vessel 54 may be configured to have a diameter of about 4.3 meters and a height (length) of about 17.7 meters. By completely sealing the containment vessel 54, access to the reactor core 6 may be restricted. Any unauthorized access or intervention may be supervised. Furthermore, the underground profile of the nuclear power system makes it less visible and more easily hidden. The water sump 16 may be covered by a protective cover (not shown) to further isolate the power module assembly 50 from external threat or airborne (airborne) objects, such as aircraft or missiles.

Refueling of the reactor core 6 may be performed by: the entire power module assembly 50 is transported, for example, by rail car or sea, and the power module assembly 50 is replaced with a new or refurbished power module assembly having a new supply of fuel rods. Refueling and maintenance operations may be performed by opening (underfilling) flanges (flanges) or cutting the vessel at a height where the cylindrical section is above the reactor core 6. Refuelling may be performed every 2-10 years or more, depending on the type of fuel and the specifications of the system.

The containment vessel 54 encloses the reactor core 6 and, in some cases, cools the reactor core 6. The containment vessel 54 is relatively small, has high strength, and can withstand six or seven times the pressure of conventional containment vessel designs due in part to its smaller overall size. Even if there is a disruption of the primary cooling system of the power module assembly 50, no nuclear fission products are released to the environment. Decay heat may be removed from the power module assembly 50 in an emergency situation.

The reactor core 6 is shown submerged or immersed in a primary coolant 100, such as water. The reactor vessel 52 contains the coolant 100 and the reactor core 6. A shroud 22 surrounds the reactor core 6 around its sides and serves to direct the coolant 100 upwardly through the annular space 23 and out of the riser 24 in the upper half of the reactor vessel 52 as a result of the natural circulation of the coolant 100. In one embodiment, the reactor vessel 52 has a diameter of about 2.7 meters and an overall height (length) of about 13.7 meters. The reactor vessel 52 may comprise a predominantly cylindrical shape with spherical upper and lower ends. The reactor vessel 52 is typically at an operating pressure and temperature. The containment vessel 54 is internally dry and may operate at atmospheric pressure with a wall temperature at or near the temperature of the pool of water 16.

The containment vessel 54 surrounds the reactor vessel 52 primarily in a dry or gaseous environment, which is considered the containment region 44. The containment region 44 may be filled with air. The containment vessel 54 includes an inner surface 55 or wall adjacent to the containment region 44. The containment region 44 may include one or more gases instead of or in addition to air. In one embodiment, the containment region is maintained at a sub-atmospheric pressure condition, such as a partial vacuum condition. The gas or gases in the containment vessel may be removed such that the reactor vessel 52 is within the fully or partially evacuated containment region 44.

During normal operation, thermal energy from a fission event in the reactor core 6 causes the coolant 100 to be heated. As the coolant 100 heats up, it becomes less dense and tends to rise through the annular space 23 and out the riser 24. As the coolant 100 cools, it becomes relatively denser than the heated coolant, and circulates around the outside of the annulus 23, descends to the bottom of the reactor vessel 52, and ascends through the shroud 22 to be heated again by the reactor core 6. This natural circulation causes the coolant 100 to circulate through the reactor core 6, transferring heat to an auxiliary cooling system, such as the auxiliary cooling system 30 in FIG. 1, to generate electricity.

This natural circulation can be enhanced by providing a two-phase condition (two-phase condition) of the coolant 100 in the riser 24. In one embodiment, gas is injected into or near the reactor core 6 to create or enhance the two-phase operating condition and increase the flow rate of the coolant 100 through the riser 24. Although the discharging (voiding) of the reactor core 6 may produce negative induction of reactivity, steady state conditions that follow non-discharging conditions may result in positive induction of reactivity. In one embodiment, the reactivity is also controlled by managing a combination of control rod insertion rate and temperature-sensitive control rod travel (taps).

Fig. 3 illustrates the power module assembly 50 of fig. 2 during an emergency operation. The emergency operation may include, for example, a response to overheating of the reactor core 6 or an overpressure event of the reactor vessel 52. During emergency operation, the reactor vessel 6 may be configured to release the coolant 100 into the containment region 44 of the otherwise dry containment vessel 54. The decay heat of the reactor core 6 may be removed by the coolant 100 condensing on the inner surface 55 of the containment vessel 54. Although the containment vessel 54 may be submerged in the pool of water 16, the inner surface 55 of the containment vessel 54 may be completely dry prior to an emergency operation or an over-pressurization event. For example, the suppression pool 12 in FIG. 1 is not present in the containment vessel 54 during normal operation.

A flow limiter 58 or steam vent may be provided on the reactor vessel 52 to vent the coolant 100 into the containment vessel 54 during emergency operation. The coolant 100 may be released into the containment vessel 54 as a gas or vapor 41, such as steam. The flow restrictor 58 may be directly connected or mounted to the outer wall of the reactor vessel 52 without any intervening structures (conduits) such as pipes or connectors. In one embodiment, the flow restrictor 58 is welded directly to the reactor vessel 52 to minimize the possibility of any leaks or structural failures. The flow restrictor 58 may be a Venturi (Venturi) flow valve sized to release the coolant 100 to the containment vessel 54 at a controlled rate. In one embodiment, the coolant 100 is released only as steam or water vapor from the reactor vessel 52. The condensation of the water vapor 41 may reduce the pressure within the containment vessel 54 at about the same rate as the vented water vapor 41 adds pressure to the containment vessel 54. In one embodiment, the flow restrictor 58 is configured to release approximately 5 megawatts of heat contained in the water vapor 41.

The coolant 100 released into the containment vessel 54 in the form of steam 41 condenses on the inner surface 55 of the containment vessel 54 as a liquid such as water. As the water vapor 41 is converted to the liquid coolant 100, the condensation of the water vapor 41 causes a pressure drop within the containment vessel 54. A sufficiently large amount of heat may be removed from the power module assembly 50 by condensation of the water vapor 41 on the inner surface 55 of the containment to enable management of the removal of decay heat from the reactor core 6. In one embodiment, no liquid coolant 100 is released from the reactor vessel 52, even during emergency operation. The condensed coolant 100 descends to the bottom of the containment vessel 54 and collects as a pool of liquid. As more and more water vapor 41 condenses on the inner surface 55, the level of the coolant 100 at the bottom of the containment vessel 54 gradually rises. The heat stored in the steam 41 is transferred through the walls of the containment vessel 54 to the water sump 16, which is the ultimate heat sink. The heat stored in the coolant 100 located at the bottom of the containment vessel 54 is transferred to the inner surface 55 by liquid convection and conductive heat transfer.

The heat removed from the steam or water vapor 41 may be transferred to the relatively cool inner surface 55 by condensation on the inner wall of the cold containment vessel 54 and by natural convection from the hot coolant to the inner surface 55. Heat may be transferred to the pool of water 16 by conduction through the containment wall and by natural convection on the outer surface of the containment vessel 54. After the reactor core 6 becomes overheated and during emergency operation, the coolant 100 remains confined within the power module assembly 50. The heat transferred to the water sump 16 may provide sufficient passive decay heat removal for three or more days without any operator intervention.

The containment vessel 54 may be designed to withstand the maximum pressure that would be incurred given the instantaneous release of high pressure fluid from the reactor vessel 52 to the containment vessel 54. The pressure within the containment vessel 54 may be designed to equalize with the pressure within the reactor vessel 52 such that the vent flow caused by the pressure differential ceases. Over time, the amount of pressure within the containment vessel 54 may be equalized with the amount of pressure within the reactor vessel 52, resulting in a coolant level 100A within the reactor vessel 52 and a coolant level 100B within the containment vessel 54 as shown in FIG. 3. Since the coolant temperature within the reactor vessel 52 is higher compared to the temperature within the containment vessel 54, the coolant level 100B is shown elevated relative to the coolant level 100A. Fig. 3 shows that the coolant levels 100A and 100B may be balanced such that the coolant level 100A in the reactor vessel 52 remains above the top of the reactor core 6 such that the reactor core 6 is always covered by the coolant 100.

The flow valve 57 may be configured to allow the coolant 100 to flow back from the containment vessel 54 to the reactor vessel 52 once a steady state condition of the coolant levels 100A, 100B is reached. The coolant 100 that is allowed to reenter the reactor vessel 52 through the flow valve 57 replenishes the coolant 100 that is bled off through the flow restrictor 58 as water vapor 41. The flow of coolant 100 through the flow valve 57 may be achieved by natural circulation of a passive system resulting from different water densities caused by temperature differences in the containers 52, 54. No mechanical or electrical pump or motor is required. In one embodiment, the flow valve 57 restricts the flow of the coolant 100 in a single direction, from the containment vessel 54 to the reactor vessel 52.

When the reactor core 6 becomes superheated, the flow limiter 58 or steam outlet is configured to bleed the coolant 100, e.g., in the form of steam or steam 41, into the containment vessel 54 at a rate that maintains an approximately constant pressure within the containment vessel 54 during steady state conditions. In one embodiment, the containment vessel 54 experiences an initial pressure spike (spike) before reaching a steady state condition. By controlling the rate of pressure build-up within the containment vessel 54, the thickness of the containment vessel wall may be designed to have less material strength due to the lower, controlled pressure therein. Reducing the wall thickness may reduce the shipping weight of the power module assembly 50 and reduce manufacturing and shipping costs.

Although achieving or maintaining a full or true vacuum may be commercially or technically impractical, a partial vacuum may be created within the containment vessel 54. Thus, references herein to a vacuum are to be understood as a partial or complete vacuum. In one embodiment, the containment region 44 is maintained under vacuum pressure with significantly reduced convective and conductive heat transfer through containment gases (containment gases). By substantially removing gas from the containment region 44, such as by maintaining a vacuum within the containment vessel 54, the initial rate of condensation of the water vapor 41 on the inner surface 55 is increased. The increase in the rate of condensation increases the rate of heat transfer through the containment vessel 54.

The vacuum within the containment region 44 acts as a type of thermal insulation during normal operation, thereby retaining heat and energy within the reactor vessel 52 that can continue to be utilized. For this reason, less insulating material may be used in the design of the reactor vessel 52. In one embodiment, a reflective insulator is used in place of or in addition to conventional thermal insulator. The reflective insulation may be disposed on one or both of the reactor vessel 52 or the containment vessel 54. The reflective insulator can resist water loss (water damp) better than conventional thermal insulators. Additionally, the reflective insulation does not prevent as much heat transfer from the reactor vessel 52 during emergency conditions as conventional thermal insulation. Thus, the combination of vacuum and reflective insulation provides thermal insulation during normal operation and facilitates heat transfer away from the reactor core 6 during emergency conditions.

In the event of a loss of vacuum within the containment region 44, the introduced gas or liquid provides another passive safety cooling mechanism to transfer heat between the reactor vessel 52 and the containment vessel 54 through natural circulation. For example, by reducing or eliminating conventional thermal insulation, a more efficient heat transfer from the reactor vessel 52 during emergency operation may be created due to condensed liquid coolant 100 collecting in the bottom of the containment vessel 54. Heat can be transferred from the reactor vessel 52 to the containment vessel 54 through the liquid coolant 100.

Additionally, removing air and other gases from the containment region 44 may reduce or entirely eliminate the need for any hydrogen recombiners (hydrogen recombiners) that are typically used to reduce flammable gas mixtures that may otherwise be developed. During emergency operation, steam may chemically react with the fuel rods to produce a high level of hydrogen. When hydrogen is mixed with air or oxygen, a flammable mixture may be produced. By removing a substantial portion of the air or oxygen from the containment vessel 54, the amount of hydrogen and oxygen that is allowed to mix is minimized or eliminated. In one embodiment, any air or other gas remaining in the containment region 44 is removed or vented when an emergency condition is detected.

FIG. 4 illustrates an exemplary condensation rate of the coolant 100 released into the containment vessel 54. As previously described, the coolant 100 may be vented as steam or water vapor 41 that condenses on the inner surface 55 of the containment vessel 54. The flow limiter 58 controls the rate of release of the coolant 100 in the form of steam 41 to the containment vessel 54 so that the rate of increase of the coolant level 100B in the containment vessel 54 may be determined or managed. According to the graph in FIG. 4, about 110 inches of coolant 100 may collect in the bottom of the containment vessel 54 after 9500 seconds or about 2 hours and 38 minutes. Of course, this rate of increase of the coolant height 100B will depend on the size of the reactor vessel 52 and containment vessel 54, in addition to the design of the flow limiter 58.

In one embodiment, once the pressures in the reactor vessel 52 and the containment vessel 54 equalize or reach a steady state, the rate of increase of the coolant height 100B flattens out at a near constant value. The flow of the coolant 100 through the flow valve 57 in FIG. 3 to the reactor vessel 52 may remove approximately the same amount of the coolant 100 that condenses to a liquid on the inner surface 55 of the containment vessel 54.

A flow limiter 58 connected to the reactor vessel 52 may vent the water vapor 41 at a rate that maintains an approximately constant pressure within the containment vessel 54 during steady state conditions. FIG. 5 illustrates an exemplary pressure fluctuation within the containment vessel 54 during an overpressure event. In one embodiment, the pressure within the containment vessel 54 may be at or near atmospheric pressure prior to the overpressure event. In another embodiment, the pressure within the containment vessel is maintained as a vacuum. The containment vessel 54 may then be subjected to a pressure spike that increases the pressure to some predetermined upper limit.

In one embodiment, the upper pressure value is about 300 pounds per inch 2 (psia). Once the pressure reaches the upper limit value, the flow limiter 58 may close or inhibit further release of the coolant 100 as steam 41 into the containment vessel 54. The pressure within the containment vessel 54 then decreases as the water vapor 41 condenses into a liquid. The pressure may be allowed to decrease to some predetermined lower limit. In one embodiment, the lower limit pressure value is less than 150 lbs/inch². Once the pressure reaches the lower limit value, the flow limiter 58 may open or otherwise allow additional coolant 100 to be released to the containment vessel 54. The pressure in the containment vessel 54 thenUp until the upper limit value is reached again, continuing the pressurization and depressurization cycle while removing decay heat from the reactor core 6. Thus, the pressure within the containment vessel 54 may be maintained between the upper and lower limits.

The steam nozzle flow area of the flow limiter 58 may be calculated from measurements or estimates of the rate of condensation of steam within the containment vessel 54, the rate of energy removal from the containment vessel 54, and the rate of heating of the pool 16 of FIG. 3. In one embodiment, the rate of change of the height of the liquid in the containment vessel 54 may be about.0074 inches/second. According to the law of conservation of mass, the mass flow rate at which vapor condenses into a liquid can be determined according to the following equation:

(1)dM_L/dt＝ρ_LA_C(dL/dt)_C＝m

the rate of heat transfer to the inner surface 55 of the containment vessel 54 may be given according to the following equation:

(2)q＝mh_fg

the heating rate for the basin 16 can be determined using the following equation:

(3)MC_PdT/dt＝q

assuming that the cooling pool mass, the cooling pool water specific heat at constant pressure, and the heat input are constant, the time required to heat the water pool 16 can be obtained according to the following equation in combination with equation (2):

(4)Δt＝MC_PΔT/q

in one embodiment, the upper temperature of the water basin 16 is set below boiling point, such as 200 degrees (Fahrenheit). Finally, the equation for steam choke (choke flow) can be given by the following equation:

(5)m＝C_dA[Kg_cρ_gP]^(1/2)

wherein C is_dIs an emission coefficient of about 0.95, and wherein

K＝γ[2/(γ+1)]^{(γ+1)/(γ-1)}

During the first 100 second steam discharge event, the initial 6% decay heat may be experienced by the main cooling system, however this will be flat to about 2% or 3% under steady state conditions. Releasing pressure to the containment vessel 54 may result in approximately 3% of the decay heat being transferred from the reactor vessel 52, which includes a significant amount of decay heat released during the steady state phase. This is accomplished by the passive emergency feedwater and decay heat removal system described herein, without the need for a pre-existing water source or suppression pool located within the containment vessel 54.

FIG. 6 illustrates an alternative embodiment of the power module assembly 60, the power module assembly 60 including a containment vessel 64 having fins 65 to increase the cooling surface area. The heat sink 65 may be attached to the outer wall of the containment vessel 64 to remove the decay heat of the reactor core during emergency operation. During normal operation of the power module assembly 60, the inside of the containment vessel 64 remains dry while the reactor vessel 62 contains the reactor core as well as the coolant. In one embodiment, the containment vessel 64 is under reduced pressure or vacuum during normal operating conditions. The coolant may be a liquid or a gas. During emergency operation, such as an overpressure in the reactor vessel 62, coolant is released from the flow restrictor 68 to the containment vessel 64. The coolant circulates within the containment vessel 64 and releases heat to the walls of the containment vessel 64. This heat is then removed from the containment vessel by convection or conduction to the surrounding heat sink 66.

The hot trap 66 may be a fluid such as water or gas. In one embodiment, the heat sink is comprised of earth (e.g., stone, soil, or other solid material) that completely surrounds containment vessel 66. Fins 65 may be attached to containment vessel 64 and provide additional surface area for transferring decay heat to heat sink 66. The fins 65 may surround the containment vessel 64. In one embodiment, the fins 65 are oriented in a horizontal plane. The heat traps 66 may be contained in a containment structure 61, such as concrete. A cover 63, which may also be made of concrete, may completely enclose the power module assembly 60 and the heat sink 66. The containment structure 61 and cover 63 may be used to protect against foreign projectiles and may also act as a biological shield.

FIG. 7 illustrates an embodiment of a power module assembly 70 including multiple containment regions 71, 72. The containment region may be divided into a first containment region 71 and a second containment region 72. The first containment region 71 may be located in an upper portion of the containment vessel 74 and the second containment region 72 may be located in a lower portion of the containment vessel 74. The first containment region 71 may be maintained at atmospheric pressure while the second containment region 72 may be maintained below atmospheric pressure.

One or more valves 75 may be disposed between the first and second containment regions 71, 72. The valve 75 may be operated to release pressure in the event of an emergency condition. In one embodiment, the valve 75 operates to transfer liquid coolant that condenses within the first containment region 71 such that it is collected within the second containment region 72. In one embodiment, conventional thermal insulation 76 is included within the first containment region 71, while reflective insulation 78 is included within the second containment region 72. Any number of containment regions may be provided, some or all of which may be maintained in a vacuum state.

FIG. 8 illustrates a novel method of cooling a power system, such as the power module assembly 50 of FIG. 3. In operation 810, the power module assembly 50 is emergency stopped in the event of a high pressure event indicated in a reactor vessel, such as the reactor vessel 52 in fig. 3.

At operation 820, coolant is released to a containment region, such as the containment region 44 in FIG. 3, located between a containment vessel, such as the containment vessel 54 in FIG. 3, and the reactor vessel 52. The containment region 54 surrounds the reactor vessel 52 and may be substantially dry prior to a high pressure event. This coolant, such as coolant 100, may be released to the containment vessel 54 in the form of steam 41 or steam. In one embodiment, steam released from the auxiliary cooling system 30 of FIG. 1 due to a fault or overall pressure loss may also be vented to the containment vessel 54.

In operation 830, the water vapor 41 condenses on an inner wall, such as the inner wall 55 of the containment vessel 54. The water vapor 41 may be condensed into a liquid, such as water.

At operation 840, decay heat is transferred to the liquid medium surrounding the containment vessel 54. The decay heat may be transferred via condensation of water vapor 41 and convection and conduction of the condensed liquid.

In operation 850, the pressure within the containment region 44 is limited or maintained within design limits by condensation of the coolant on the inner walls. A steam flow limiter, such as the flow limiter 58 in FIG. 3, may be sized to limit the rate of pressure rise within the containment vessel 54. The rate of pressure rise may be largely offset by condensation of the water vapor 41 into a liquid. The steam flow limiter 58 may be selectively or intermittently opened such that the pressure within the containment vessel 54 is limited to a maximum value and allowed to decompress when the flow limiter 58 is closed.

The condensation of the water vapor 41 may reduce the pressure within the containment region 44 by approximately the same amount that the released coolant increases the pressure within the containment region 44. The coolant 100 may be released to the containment region 44 in the form of steam 41 or steam, and the decay heat of the reactor core 6 may be removed from the power module assembly 50 by condensation of the steam 41 on the inner wall 55 of the containment vessel 54.

Although the embodiments provided herein primarily describe a pressurized water reactor, it will be apparent to those skilled in the art that these embodiments may be applied to other types of nuclear power systems as described or with some obvious variation. For example, these embodiments or variants thereof may also be made available to a boiling water reactor. A boiling water reactor may require a larger vessel to produce the same energy output.

The rate at which coolant is released into the containment vessel, the rate at which coolant condenses to a liquid, and the rate of pressure rise within the containment vessel, as well as other rates and values described herein, are provided by way of example only. Other rates and values may be determined by experimentation, such as constructing a full scale model or a proportional model of a nuclear reactor.

Having described and illustrated the principles of the invention in the form of preferred embodiments thereof, it should be apparent that the invention may be modified in arrangement and detail without departing from such principles. We claim all modifications and variations coming within the spirit and scope of the following claims.

Claims (18)

1. A power module assembly comprising:

a containment vessel; and

a reactor vessel housed in the containment vessel, wherein a containment region is located between the reactor vessel and the containment vessel, wherein the reactor vessel is configured to controllably release coolant into the containment region during a limp-home operation of the power module assembly, wherein at least a portion of the containment region is maintained at a pressure below atmospheric pressure during normal operation of the power module assembly, and wherein the pressure below atmospheric pressure substantially prevents convective heat transfer between the reactor vessel and the containment vessel, wherein an entire interior surface of the containment vessel is dry prior to the limp-home operation.

2. The power module assembly according to claim 1 wherein said containment vessel is surrounded by a heat sink.

3. The power module assembly according to claim 2, further comprising a heat sink attached to an outer wall of the containment vessel and in contact with the heat sink.

4. The power module assembly according to claim 2, wherein the heat sink comprises water or gas.

5. The power module assembly according to claim 1, wherein the containment region comprises a first containment region and a second containment region, wherein the first containment region is maintained at atmospheric pressure, wherein the second containment region is maintained at a pressure below atmospheric pressure, and wherein both the first containment region and the second containment region are internally dry prior to the emergency operation.

6. The power module assembly according to claim 5 wherein the coolant is released into the first containment region in the form of steam during the emergency operation, and wherein the steam condenses in the containment as liquid coolant.

7. The power module assembly according to claim 6, further comprising one or more valves connecting the first containment region to the second containment region, wherein the one or more valves are operably configured to transfer the liquid coolant from the first containment region to the second containment region.

8. The power module assembly according to claim 6, wherein the reactor vessel is insulated by two types of thermal insulation that differ between the first containment region and the second containment region, and wherein the reactor vessel is insulated by reflective insulation located in the second containment region.

9. The power module assembly according to claim 1, further comprising a reflective insulator located between the reactor vessel and the containment vessel, wherein substantially all of the thermal insulation of the reactor vessel is provided by the combination of the sub-atmospheric pressure and the reflective insulator.

10. A power module assembly comprising:

means for circulating primary coolant through a reactor core, wherein the reactor core and the primary coolant are contained in a reactor vessel;

means for controllably releasing the primary coolant in vapor form into a containment vessel in response to a high pressure event within the reactor vessel, wherein at least a portion of the containment vessel is maintained at a partial vacuum prior to releasing the primary coolant in vapor form into the containment vessel, wherein the vapor condenses on an interior surface of the containment vessel to form a pool of primary coolant, and wherein the partial vacuum operates substantially to prevent convective heat transfer between the reactor vessel and the containment vessel; and

means for circulating the pool of primary coolant back into the reactor vessel and through the reactor core, wherein an entire interior surface of the containment vessel is dry prior to the high pressure event, and wherein the pool of primary coolant is not present prior to the high pressure event.

11. The power module assembly according to claim 10, wherein the condensed steam forms a pool of primary coolant extending between an outer wall of the reactor vessel and an inner surface of the containment vessel, and wherein this pool of primary coolant consists entirely of the condensed steam.

12. The power module assembly according to claim 10 wherein the condensation of the steam reduces the pressure within the containment by the same amount that the released steam increases the pressure within the containment.

13. The power module assembly according to claim 12 wherein the containment vessel is maintained at a pressure above atmospheric pressure following the high pressure event.

14. The power module assembly according to claim 10, wherein the containment vessel substantially surrounds the reactor vessel.

15. The power module assembly according to claim 10, wherein substantially all of the thermal insulation of the reactor vessel is provided by the partial vacuum prior to releasing the primary coolant into the containment vessel in vapor form.

16. The power module assembly according to claim 10, wherein the outer surface of the reactor vessel comprises a steel shell, and wherein no insulating material is placed between the steel shell and the partial vacuum.

17. The power module assembly according to claim 10 wherein the containment includes a first containment region and a second containment region, wherein the first containment region is maintained at atmospheric pressure, wherein the second containment region is maintained at a pressure below atmospheric pressure, and wherein both the first containment region and the second containment region are internally dry prior to the high pressure condition.

18. The power module assembly according to claim 17, wherein during the high pressure condition, the coolant is released into the first containment region in the form of steam, wherein the steam condenses into a liquid coolant in the first containment region, and wherein the power module assembly further comprises means for transferring the liquid coolant from the first containment region to the second containment region.