JPS63169593A - Nuclear reaction reactor system and manufacture thereof - Google Patents

Abstract

(57) [Summary] This bulletin contains application data before electronic filing, so abstract data is not recorded.

Description

DETAILED DESCRIPTION OF THE INVENTION Field of the Invention The present invention relates to the generation of thermal energy by nuclear reactors (nuclear reactors), and more particularly to nuclear reactors exhibiting improved safety characteristics, as well as nuclear reactors. Concerning how to improve the safety characteristics of.

BACKGROUND OF THE INVENTION Two very important uses of the thermal energy obtained using nuclear power are the generation of electrical power and the generation of industrial process heat.

As far as electricity generation is concerned, the focus of today's technology is on large, unitary, water-cooled systems, with the exception of the large carbon dioxide-cooled versions used in the UK. It is located in a nuclear reactor. Such water-cooled reactors have a maximum operating temperature of about 600"F, which is lower than those for industrial process heat applications, which are normally accomplished at higher temperatures. Furthermore, even in their use for electricity generation, large, unitary, water-cooled reactors exhibit considerable disadvantageous characteristics. For example: The high cost of maintaining acceptable operating conditions and the potential for exposing humans to radiation in the vicinity of the reactor are characteristics that make such Significantly reduces the usefulness of nuclear reactors.The high cost of maintaining safe operating conditions greatly reduces the economic viability of using such reactors. Also, the relatively high cost of protecting humans from radiation exposure is
This will have an adverse impact on operating and maintenance costs.

Additionally, as can be understood by those skilled in the art, when a water-cooled nuclear reactor is subjected to loss of forced cooling, the thermal energy input (decaying heat from the reactor fuel, The heating rate of the reactor's metal-coated fuel, based on stored heat and heat from the reaction of the metal and water, becomes too rapid. Temperatures which result in damage to the cladding are reached in seconds, resulting in the release of fission products from the fuel elements of the reactor.

Instead of problematic large, single-body water-cooled reactors, the technology uses large, hot, and helium-cooled reactors.
It can be directed to gas cooled and graphite regulated nuclear reactors. These reactors have advanced beyond carbon dioxide gas cooled and graphite regulated designs whose temperatures are constrained by the properties of carbon dioxide. They have the advantage that they can be operated at high temperatures up to about 1750'F, which allows them to be used under conditions normally required for efficient power generation. and otherwise meet a wide range of industrial processing heat generation requirements.

These large, helium gas cooled and graphite regulated reactors have a much slower heating rate, as opposed to the water cooled variety discussed above. The temperature at which there is a risk of release of the products of nuclear fission is reached only after a few hours from the time of loss of forced circulation. Furthermore, since no metal cladding is employed within these high temperature reactors to preserve the products of nuclear fission,
The risk of damage to the metal coating at high temperatures is obviated. The mode of failure encountered in such high-temperature reactors is gradual rather than rapid, cataclysmic failure;
Initially, the products of fission are slowly diffused from the fuel particles found in the reactor core through a ceramic barrier (geometric stability of the fuel element is maintained).

In no small part, the hot helium gas cooled reactors discussed in the previous section also have some drawbacks. First, although these reactors are more resistant to the effects of loss of forced circulation than water-cooled reactors, they also suffer from an uncontrolled warming of the atoms due to the products of fission. A state is reached where it can be released from the furnace core. Also, because of their large dimensions, these reactors have limited
cannot be manufactured in a protected factory environment and must be assembled on an open field site. This requirement would significantly increase the cost of manufacturing the reactor. moreover,
Each large nuclear reactor is identified by the use it is placed in and the JJ! Must be custom made for the occasion. This factor also significantly increases the cost of manufacturing the reactor, since standardization of installations 1 becomes impossible.

Therefore, the state of the art leaves much to be desired.

OBJECT OF THE INVENTION The object of the invention is to provide a heat source generated by nuclear energy, which can be significantly cooled by providing passive cooling means independent of operator action. by it compensating for the loss of the condition;
It is inherently safe compared to currently available heat sources.

Another object of the invention is to provide a heat source generated by nuclear energy, which heat source is equipped with passive safety means in the form of a continuously operating heat sink; The heat pool has the maximum combustion F4 temperature of the heat source,
This is to limit the value to a value at which nuclear fission products are not released.

Yet another object of the present invention is to provide a heat source generated by nuclear energy, which is destructive to the environment due to leakage of products of nuclear fission, even when the pressure of the cooling system is increased. It allows continuous operation of the heat source at reduced power levels without posing a significant risk of damage.

Another object of the invention is to provide radioactive material leaking from said nuclear reactor system, such as radioactive material in contaminated cooling material, or radioactive material from shredded fuel elements released from said fuel handling system after fuel element rupture. The objective is to provide a barrier to polluting interchange between the materials (materials) and the atmosphere.

Yet another object of the invention is to provide a nuclear energy generated heat source that is more economically constructed, operated, maintained, and retired than currently available large nuclear reactors. .

These and other objects are met by the present invention.

DESCRIPTION AND ADVANTAGES OF THE INVENTION In one of its features, the present invention is a structure in a high temperature and gas cooled nuclear reactor system;
The structure includes passive heat sink means for absorbing decaying thermal energy generated within a reactor core contained within the nuclear reactor system and when the reactor is in a critical power generating condition. for absorbing the contributing thermal energy so that the loss of forced circulation of the gaseous coolant, or the release of products of fission resulting from such loss in connection with the depressurization of the coolant, is prevented. and means for removing thermal energy from the heat sink means at a rate sufficient to maintain the capacity of the heat sink means.

Among other features, the present invention provides high-temperature, gas-cooled and
A method of manufacturing a nuclear reactor system, the method comprising: placing passive heat sink means in communication with the nuclear reactor, such that the means is connected to a nuclear reactor housed within the nuclear reactor; positioned to absorb decaying thermal energy from the core; and loss of forced circulation of the gas coolant or depressurization of the coolant when the nuclear reactor is in critical power generation conditions. so that the release of fission products resulting from such loss in coupling is prevented.
means for removing thermal energy from the thermal sink means at a sufficient rate to maintain the ability of the thermal sink means to absorb decreasing thermal energy; and operably associated with the thermal sink means. The invention also includes the following:

In yet another feature of the present invention, the present invention is compact and compact.
A method for preventing the escape of fission products from a high-temperature, gas-cooled, nuclear reactor due to overheating, the method employing passive heat sink means placed in communication with said nuclear reactor. absorbing a fraction of the decaying thermal energy generated within the reactor core contained within the nuclear reactor; maintaining the ability of said thermal reservoir means to absorb decaying thermal energy so that loss of forced circulation of the coolant or release of products of fission resulting from such loss in leading to depressurization of the coolant is prevented; and removing thermal energy from the heat reservoir means at a rate sufficient to allow the thermal energy to be removed from the heat reservoir means.

A number of advantages result from implementing the invention. The operation of said nuclear reactor system of the present invention provides an inherently safe operation of the nuclear reactor, both from the point of view of general public health and safety, and a sustained safety of the investment in the reactor plant itself. bring. This is because the release of the products of nuclear fission is prevented only by passive means. In other words,
According to the present invention, the need to dissipate the decaying thermal energy from the reactor core in order to prevent the build-up of temperatures at which the products of nuclear fission are released is explained by the need to dissipate the decaying thermal energy from the core into the This is accomplished by transferring it to a hot pool. The passive heat dissipation feature operates without human or automatic manipulation and is therefore immune to the vagaries of human or mechanical error. Moreover, the nuclear reactor of the present invention is equipped with a forced circulation of the cooling system, since the decaying thermal energy generated by the reactor core is dissipated without the contribution of the forced cooling system of the present invention. It can also be operated on a reduced scale in the event of a loss of cooling and/or an increase in pressurization of the cooling system. Furthermore, the reactor system provided in accordance with the present invention, because of its shape, is more easily and economically constructed, is more easily and reliably tested over its lifetime, and can be used in larger, more specialized They can be maintained, repaired, and retired more conveniently than standardized nuclear reactors.

In the following sections, the invention will be described in more detail to illustrate some of its embodiments.

DESCRIPTION OF THE PREFERRED EMBODIMENTS The nuclear reactor system of the present invention typically includes a nuclear reactor pressure vessel and a plurality of reactors contained within the pressure vessel and arranged to form a cobblestone platform. It includes a reactor core with a fuel element and a pressurized and forced gas circulation system. The reactor system is advantageously adapted for use in combination with one or more other similar nuclear reactor systems - when a need exists for energy that exceeds the output of the same system. Also, the dimensions of the system are sufficiently small in the advantageous embodiments comprising, so that its manufacture can suitably be carried out under controlled factory conditions rather than on-site;
Therefore saving costs.

A basic feature of the invention is that a passive system for dissipating the decaying thermal energy generated in the reactor core complements the operation of the gas cooling system used to remove heat from the reactor core. The purpose is to provide means. This feature of the invention is important because of the following characteristics of high temperature, gas-cooled nuclear reactors: More specifically, when the reactor is in critical power generation conditions, if the coolant circulation is stopped or, in addition, the coolant is depressurized, then the (reducing the molecular weight of the coolant and correspondingly reducing the ability of the coolant to absorb heat), the temperature of the fuel increases. The temperature of the reactivation rate reaches the temperature at which the reactor is shut down and the products of nuclear fission begin to escape from the spherical fuel particles used in the fuel elements of the reactor core, for example 1600°C. It is possible.

Escape of the products of fission from the fuel particles must obviously be avoided. This is because this increases the likelihood of them escaping into the atmosphere and increases their potential safety hazard to the public.

However, with the present invention, the provision of said passive cooling means is effective even if forced coolant circulation is stopped when said reactor is in a critical, power-producing state, as well as in said cooling system. Even if the fuel is further depressurized, the increase in temperature of the fuel (assuming no control ronds are inserted) will push the temperature of the reactor core beyond the temperature at which the products of fission are released. has the effect of making it insufficient. Therefore, by providing the above-mentioned passive cooling means, the present invention provides the following advantages:
The thermal energy of the reactor core, which if not dissipated would cause the release of fission products from the fuel element, is removed without the need for manually or automatically actuated safety measures and loss of fission products. Great technological progress has been made in that treatment can be performed without the presence of any cancer.

The charts in Figures 1 and 2 clearly demonstrate the above. FIG. 1 is a plot of fuel temperature in a hypothetical nuclear reactor system according to the present invention versus elapsed time since loss of forced cooling. Line 1 shows the temperature limit, at or above which the diffusion of the products of nuclear fission from the nuclear fuel particles begins, which is here 1600<0>C. Curve 2 shows the temperature in the nuclear fuel particles when the forced gas coolant circulation has ended but the coolant is properly pressurized. Curve 3 depicts the temperature reached within the nuclear fuel particles when forced circulation is terminated and when the primary cooling system is depressurized. It is clear from FIG. 1 that in no case does the temperature of the nuclear fuel element reach the 1600° C. limit. Figure 2 shows that following complete loss of forced cooling,
The radial contours of the maximum fuel temperature reached at various locations in a hypothetical nuclear reactor system according to the invention are shown (the system comprises a cylindrical core cooled by an axial flow of the coolant). ). Each of the selected locations is a location of a component that is typically integrated within a nuclear reactor system according to the present invention (see 3 to 3 below).
(See Figure 6).

Accordingly, FIG. 2 is a plot of such temperature versus radial distance of the location indicated on the chart from the center of the core. From the chart, it can be seen that within the core, within the adjacent graphite reflector material, within the ceramic in the core housing, within the gas space between the core housing and the pressure vessel wall, within the pressurized solid wall itself, The temperature in the air space between the wall and the inner wall of the containment vessel, in the inner wall of the containment vessel, and in the water held in the containment tank is ascertained. Curve 6 shows those maximum temperatures when forced coolant circulation has ended but the system remains pressurized; curve 7 shows forced coolant circulation indicates their maximum temperature reached when the system is to be depressurized. Again, their maximum fuel temperature does not reach the 1600° C. limit at which the products of each fission begin to diffuse out of the fuel particles.

Passive heat sink means are provided in accordance with the present invention by merging a heat sink within a high temperature, gas-cooled, nuclear reactor system, the heat sink being provided when said reactor is in a critical, power producing state. or when the reactor is shut down and emitting waning heat, regardless of whether the forced gas cooling system of the reactor is fully operational or not.
It constantly absorbs the thermal energy released by the reactor core. The heat reservoir generally comprises a liquid, typically water, which is liquid at the temperatures encountered around each reactor system and which is accessible to the heat reservoir by natural convection, conduction, and radiation. Other materials that vaporize upon absorption of the transferred thermal energy are also suitable. The liquid is positioned to absorb decaying thermal energy emitted by the reactor core. This is achieved by placing a containment vessel around the reactor pressure vessel housing the reactor core, part of which is actually a tank means integral with the containment vessel itself. It is. The tank means suitably comprises an inner wall and an outer wall, between which there is a volume to be occupied by the liquid. The reactor pressure vessel housing the reactor core is preferably circular in cross-section, and the tank means forming part of the containment vessel are accordingly:
It is arranged around the reactor pressure vessel which has an annular configuration.

The above-discussed structure provides a reservoir for thermal energy, particularly decaying thermal energy, which is generated within the reactor core and directed to the reactor pressure vessel. and into the chamber between the containment vessel and the reactor pressure vessel, and pass through the inner wall of the tank portion of the containment vessel. To continue the effective absorption of such thermal energy by the heat reservoir, the absorbed thermal energy is removed from the liquid, thereby preserving its further absorption capacity. This is done by providing means for removing that thermal energy in communication with said heat sink means, such as said tank and the liquid present therein. The removal means is suitably a condensation means;
There will typically be one or more condensing devices to which the vapor released by the heat pool is directed by suitable conduits, vents, or the like. A suitable condensing device is, for example, a natural convection cooled air condenser, although other types of conventional condensing devices can also be utilized in the present invention.

Said condensing means is generally dimensioned such that thermal energy is removed from said vapor in sufficient quantities such that the resulting liquid has a continuous stream of attenuating thermal energy absorbed by said heat pool. can be returned to said tank means in order to establish boiling equilibrium conditions to enable efficient removal. Return of the condensed liquid is achieved by providing suitable conduits or vents or the like so that the liquid can be directed back to the tank means of the containment vessel.

Examples of suitable condensed Hn/recirculation systems are conventional 1-vibe and 2-vibe systems.

In the one-vibe system, vapor is conducted to the condensing means and liquid is recirculated via the same vibrator to the tank means holding the liquid, thereby controlling the flow of vapor and the return of liquid. The flow becomes countercurrent. 2-
In a vibrator system, a separate vibrator is provided for each flow of vapor toward the condenser and return flow of liquid toward the heat reservoir.

The boiling equilibrium conditions described above can be varied over a wide range and are suitably performed at atmospheric or sub-atmospheric pressures. In a preferred embodiment of the invention,
Boiling equilibrium is established at about 150°F due to the water acting as the heat sink.

Thus, a continuously operating passive thermal energy dissipation device is provided.

According to the above, a method for manufacturing a high-temperature, gas-cooled, nuclear reactor system so as to prevent the accidental release of products of nuclear fission due to overheating is based on the reactor pressure vessel (the reactor core). It will be readily appreciated that the invention advantageously comprises arranging the container around the holder (containing the holder). The containment vessel, comprising tank means including inner and outer walls, is positioned to absorb decaying thermal energy from the reactor core by means of a liquid that the tank means is adapted to retain. The method further comprises: providing in operative association with the tank means means for removing heat from vapor released by the liquid as a result of absorption of the decaying thermal energy; and providing means in operative association with the heat removal means for returning liquid resulting from the removal of heat from the vapor to the tank means.

Also in accordance with the above, the present invention provides an advantageous method of preventing the release of products of nuclear fission from a high temperature, gas-cooled, nuclear reactor due to overheating. The method involves transferring the decaying thermal energy generated within the reactor core into a liquid disposed around the reactor pressure vessel (containing the reactor core) as previously described. removing thermal energy from the vapor released by the liquid resulting from the absorption of said attenuated thermal energy; and removing the liquid resulting from the removal of thermal energy from said vapor from said liquid. and a step of returning it to the main body.

These and more detailed embodiments of the invention are further described below with reference to some of the accompanying figures of the drawings.

Figure 3 shows a nuclear reactor system 1 that is heated to high temperatures, cooled by helium, and controlled by graphite.
0 is disclosed. The reactor system 10 includes a cylindrical reactor pressure vessel 12 (the pressure inside the one shown is maintained at 50 bar).
2 is generally cylindrical in shape and made of steel;
It has a diameter of approximately 550 ctn and a height of approximately 2,400 G. Inside the reactor pressure vessel 12, there is a cylindrical core vessel 14.
is placed and is also made of steel. Same core container 14
The diameter is about 500cxr and the height is about 1,700c!
It is R. A core 15 is disposed centrally within the core container 14, and the core 15 is of cylindrical shape, includes a spherical fuel element 16, and has a diameter of approximately 2.5 inches.

The core 15 is advantageously of the cobblestone head type. The cobblestone foundation combines a stable structure in a simple single stage for regulating equipment, suitable gas channels for cooling, and fuel supply and fuel removal. The diameter of the core is such that the nuclear fuel temperature is low enough to eliminate significant release of fission products even in the presence of forced coolant circulation loss and/or depressurization of the cooling system. Conveniently, it is adapted to maintain.

The design of the spherical nuclear fuel element 16 is also an important point.

The fission fuel material is in the form of coated particles. It is evenly distributed over the central matrix of the fuel element balls. The fuel particles are made of pyrocarbon to form a pressure-tight barrier against the release of fission products.
ocarbon) and silicon carbide and embedded in a bed of graphite poles in the core, thereby geometrically positioning the radioactive fuel particles; Achieve regulation of nuclear reactions. One advantage of using such an element is that it eliminates the use of metal cladding materials on the nuclear caloric element, which are rapidly melted and release the products of fission. The result is In contrast, the spherical fuel elements containing multi-coated fuel particles that are merged according to the present invention lack metallic cladding material to retain the products of nuclear fission. It has thermal resistance tolerances that significantly exceed those of fuel elements with The failure of the multi-coated fuel particle is not a gradual, sudden cataclysm, but rather a slow initiation of the diffusion of fission products through the particle coating, followed by a geometric Starting with the fuel element maintained stable.

Other advantages of using such a high temperature, helium cooled and graphite conditioned reactor using multi-stage coated fuel particles and graphite fuel elements can be exploited. The fuel has a wide range of uranium enrichment. Through the use of such fuels, the breeder resistance of the present invention is increased compared to light water reactors. This requires 20
As a result, enrichment levels of % or less are generally considered to be "insensitive" to at least uranium-235. Furthermore, a high percentage of up to 90% of the fission-resulting plutonium produced in the reactor is burned before the spent fuel is released. Additionally, the remaining fission plutonium in the spent fuel is mixed with high-grade plutonium isotopes, which are pure atoms, and the critical mass of such remaining fission plutonium is Reactor-type plutonium requires up to five times the size to be separated, which is difficult. Furthermore, the separation of isotopes from uranium in the supplied fuel, and the separation of plutonium in the consumed fuel, will necessarily yield fissionable material suitable for use in a nuclear explosion, and such Separation is difficult because it requires very complex and intricate equipment and is radiologically hot in both cases. Also,
The spherical fuel element ball typically has a diameter of about 2.5 mm.
inch, their large size and relatively small number make them easier to kill than the much smaller nuclear fuel spherules used, for example, in water-cooled nuclear reactors.

A ceramic shell 20 (having a thickness of about 250 mm and an outer diameter of about 500 mm) is housed within the core container 14, and the mesenterium 2
A graphite reflector 18 (having a thickness of approximately 75 cIR and an outer diameter of approximately 450 z) is housed inside the 0. control rondo 22
is located in the reactor pressure vessel for insertion into the graphite reflector 18, the control rod being made of any suitable material that acts as a neutron absorber, thereby , such as hafnium
The nuclear fission reaction in the core is controlled. The core vessel 14 is positioned within the reactor pressure vessel 12, thereby defining a gas space 24 that extends radially from about 250 to about 275 CIm from the center of the core. The storage container 26 has an inner wall 30.
together with the top and bottom part also constitute a pressure vessel (in which the pressure is maintained at approximately 1 bar), which houses the reactor pressure vessel. Said section 30 has a height of approximately 4,300 cII. The portion is located around the reactor pressure vessel 12 and is connected to the reactor pressure vessel 1.
2 and therefore about 28 mm from the center of said core.
said 2 extending radially from 0 to approximately 380 cm.
A small chamber 28 is formed between the two containers. The containment volume 3 itself has an internal tank wall 30 approximately 380 cI4 from the center of the core and approximately 5003 cI4 from the center of the core.
and an external tank wall 32 at a point , which cooperate with each other to form a tank 34 , which includes water 36 (approximately 380 m from the center of the core) as shown in FIG. extending radially from approximately 5001:lIl). The interior of tank 34 communicates with a condenser 40 via a single pipe conduit system 38;
The condenser 40 is cooled with air by natural convection.

In both this and other embodiments of the invention,
Preferably, the reactor pressure vessel core accommodating body and the accommodating vessel are not insulated at least at a position along a traveling route to a heat pool of decaying thermal energy of the reactor core,
It is easily understood that this minimizes the rate of transfer of thermal energy. The reactor pressure vessel, core enclosure and containment vessel are typically manufactured from metal, for example steel.

An advantageous feature of the invention is that the containment vessel 26 and a portion of the chamber 28 of the reactor pressure vessel 1211J are located below the reactor pressure vessel. In some embodiments that incorporate this feature, the volume of the portion is large enough to accommodate any volume of the water (or other liquid), such that the amount of water (or other liquid) is
Although it is possible for leakage from the tank means into the chamber, the leaked liquid is retained below the reactor pressure vessel and at a stable level below the reactor pressure vessel level. It will have a liquid level.

The reactor core, surrounding components, reactor pressure vessel, and containment vessel are operatively associated with a reactor gas cooling system, whereby thermal energy generated within the core is It is removed and transported to other locations for purposes such as generating electricity, providing heat for industrial processes, etc. Cooling of the reactor core 15 is largely achieved by providing a relatively low temperature, part-gas coolant (inert to the core components and conditions). Helium at a temperature of approximately 300°C (although other gases that absorb thermal energy may also be used). The helium is typically passed through the core in an axial direction (ie, in the same direction as the longitudinal axis of the core). Therefore, the relatively cold reactor cooling gas in FIG. 3 has a diameter of about 185 cr+
f into the reactor core in a conduit 44. The relatively cool gas is circulated upwardly through the passage or passages 42 and then passed downwardly in the direction of axial flow through the core 15 held within the core housing 14. Ru. The helium gas is a spherical fuel element 1.
6, the sensitive heat generated by the fissile material in their fuel elements becomes "hot" and its temperature increases to about 900°C. absorb heat.

The helium gas appears at the bottom of core 15 and is conducted into conduit 4G, conduit 46 having a diameter of about 110 mm and being disposed concentrically within conduit 44.

The helium, which is currently hot, is in conduit 4.
6 to the outside of the reactor pressure vessel 12 and into a mixing device 48, which forms part of a heat exchange system 49, which is connected to the pressure vessel 51 and It is housed within an enclosed portion of a containment vessel 26, designated by the reference numeral 26[3, which has a height of approximately 4 cm, 300 crR and a diameter of approximately 500 cm.
It has a cylindrical shape of 1).

As shown, primary "hot" helium (or other coolant) gas is preferably kept out of contact with pressure containment components of the reactor system, such as the reactor pressure vessel. .

Therefore, in the system of FIG.
As the "hot" helium gas emerges from core 15, it flows into conduit 46 and through conduit 4G, so that
It can no longer come into contact with the reactor pressure vessel 12, for example in the gas space 24.

In the heat exchange system 49, the primary cooling helium is used in this case for the purpose of removing thermal energy from the same cooling gas and transferring that energy somewhere.
It is brought into heat-transferring relationship with a secondary cooling gas body that is helium at 0° C. (but alternatively, other gas or gases as the primary coolant).

As can be seen from FIG. 3, the heat exchange system is installed in the illustrated embodiment outside the reactor pressure vessel 12. However, in some embodiments of the invention, the same heat exchange system may be
It is also advantageous to locate it within the reactor pressure vessel as shown in the figure.

The heat exchange system 49 includes a support cover piece 50 depending from the right side of the heat exchanger shell or flow skirt 66. A secondary "cold" helium gas header 56 is integral with the support cover 50. Heat exchange involves transferring the primary "hot" helium gas from mixing device 48 to volume 52 (
approximately 260 cm in diameter and approximately 1700 cm in height) in an upward direction, while simultaneously passing secondary "cold" helium gas from conduit 55 to header 56 in a downward direction into volume 52. As a result, the two gas bodies are in a counterflow relationship with each other. This includes a bundle of one or more fully removable and replaceable spiral (although other shapes are also useful) tubes (not shown for simplicity purposes) in volume 52. The plurality of tubes are suspended from a support cover 50, and the primary "hot" helium gas is passed upwardly around the spiral tubes. while the secondary "cold" gas is passed downwardly through the spiral tube, and the thermal energy of the primary "hot" helium gas is transferred to the secondary "cold" helium gas. transmitted to the gas. The temperature of the secondary helium gas is about 200
°C to about 900 °C, and the temperature of the primary helium gas correspondingly increases from about 900 °C to about 30 °C.
The temperature is lowered to 0°C.

The primary helium gas so cooled is directed below the support cover 50 and guided through an annular region [64] having a diameter of approximately 300 mm and a height of 1,700 cpt) and a flow skirt 66 (approximately 400 CIR in diameter and 1,700 atr in height). The primary "cold" helium gas travels downwardly through annular space 64 to conduit 68;
And through conduit 68 is it a gas circulation device? Conducted to ff (air blower) 70. Circulator 70 returns the primary "cold" helium gas to outer conduit 44;
Through the conduit 44 it is conducted back to the reactor core 15 in the core housing 14, thereby absorbing thermal energy again from said spherical fuel element 16. gas circulation system ra
The motor driving 70 is preferably an electric motor fitted with gas or magnetic bearings capable of operating without lubricant or the like. The device is mounted on the intermediate heat exchange device within the pressurized containment vessel section 26B. This allows the gas circulation device 70 to be placed within the confines of a 1tJ pressurized containment vessel. This is because it eliminates the need for moving parts to remain constant. In an advantageous embodiment of the invention, the need for closed moving parts that enter the pressure boundaries of the nuclear reactor system is eliminated.

At the same time, said secondary helium gas, which at the moment is already heated to a temperature of about 900C, flows into a secondary "hot" helium gas header 58 through which it has a diameter of about 100C! R and height approximately 1,9
00 cN) into a secondary "hot" gas line 60. The same secondary "hot" helium gas is conducted upwardly through tube 60 and to another location outside the heat exchange system where it is used for power generation or industrial processing heat, etc. adopted for the preparation of

As is clear from the above discussion, the primary and secondary gas coolants, although in a heat transfer relationship, are not physically mixed. Due to this separation of the primary and secondary cooling systems, which is a preferred embodiment of the present invention, this is due to the products of fission leaking out of the fuel element 16 and into the circulating helium cooling gas. The possibility of contamination of secondary helium gas is minimized. Similarly, the reactor primary helium gas coolant is free from contamination by such means even if the secondary helium gas coolant becomes contaminated on the process side of the gas cooling loop. protected.

In another advantageous embodiment, the primary and secondary gas coolants are maintained at a higher pressure than normal.

These pressures are selected to optimize overall heat transfer efficiency between the fuel element 16 and the primary coolant, and between the primary and secondary coolants, and to reduce core pressure drop and gas It is selected in consideration of the pump operating force of the circulation device 70. In one illustration of a preferred embodiment of the invention, a header 56 is fed into the header 56 and a header 58 is fed into the header 58.
The pressure of the secondary coolant removed from the reactor core is maintained at a higher level than that of the primary coolant circulated through the reactor core. Therefore, if there is a leak between the two systems, flow will be directed from the secondary side to the primary side to best prevent radioactive contamination of the secondary system to the extent possible.

Furthermore, the storage container 26 is itself a pressure container,
It therefore serves to contain all contaminated reactor cooling gases escaping from the reactor cooling system. Accordingly, the inner wall 30 of the tank means 34 is
An initial contamination barrier for such leaked cooling gas (
Its outer wall 32 also acts as a cross-contamination barrier in case the same inner wall leaks, and as a crash barrier.

During normal operation of the nuclear reactor system 10 according to the present invention, the forced circulation of gas coolant in heat transfer relationship with the reactor core 15 contributes to the large amount of thermal energy generated in the core. It is done to absorb parts. Nevertheless, a smaller amount of thermal energy is transferred from the core to the walls of the reactor pressure vessel 12. The reactor pressure vessel walls are heated to approximately 200°C (392'F), which is slightly lower than the cooling gas inlet temperature. Thermal energy penetrating into the uninsulated reactor pressure vessel wall is transferred primarily by convection and radiation across the space between that wall and the uninsulated contaminated vessel wall, and is transferred to the containment vessel and the reactor vessel by convection and radiation. The space within the chamber between the pressure vessels is typically occupied by air, but occupation by some other similar gas or mixture of gases is also permissible. The thermal energy then travels through the uninsulated inner wall of the tank portion of the containment vessel. The flow of heat due to conduction through this wall is absorbed by the water in the annularly shaped tank;
Quantify the water and then bring it to a boil. The vapor generated by the liquid rises to the condensing unit 40.

After removal of air from the condenser, the boiling/condensation saturation temperature and pressure stabilize at a value determined by the condenser's capacity for the heat load taken from the heat reservoir. The dynamics of this system are similar to those of conventional heating systems. The stabilizing temperature in this example is about 65°C (150'F) to maximize the sensitive heat absorption capacity of the water above this temperature.

Therefore, under normal operating conditions, the nuclear reactor system of the invention incorporates means for transferring heat from said reactor pressure vessel wall to the atmosphere, said means operating continuously and
It is fully powered by natural thermal forces, thereby eliminating the need for automatic start-up or operator initiation in case of an emergency.

Of course, the above continuous operation results in continuous loss of heat from the system. However, the losses are small and minimized. This is because the reactor pressure vessel is cooled by cooling gas at a relatively low temperature. That small heat loss cools the reactor even when the gas coolant pressure decreases, however, under emergency conditions where forced gas circulation is lost, as explained below. This is more than offset by the benefits of providing a passive means of

If the forced circulation of gas coolant is interrupted when the reactor is critical and generating power, the temperature of the fuel elements of the reactor (assuming no control rods are inserted) will be Temperatures that negate the re-operation rate inherent in the reactor design rise to the point where the reactor is shut down. A sufficiently large amount of decaying thermal energy from the reactor core is conducted away from the reactor fuel element to the heat pool, and then to the exterior of the reactor system and into the atmosphere. , the temperature of the reactor system is not sufficiently increased even when forced circulation of the gas coolant is lost and the pressure of the gas coolant is reduced, thereby reducing the humidity of the fuel The temperature at which products begin to be released is exceeded.

Particularly in the case of a pressure reduction of the gas coolant, i.e. a significant loss of reactor cooling from the gas cooling system, an advantageous feature of embodiments of the invention is that between the containment vessel 2G and the reactor pressure vessel 12 The volume of the chamber 28 is small enough that leakage from the cooling system into the chamber - even uncontrolled leakage - will not completely depressurize the system. Rather, since the containment vessel is a pressure vessel and the chamber is of advantageously small volume, an equilibrium pressure is established in the chamber and in the reactor cooling system, which is advantageous. is sufficiently higher than atmospheric pressure. Preferably, the volume of the chamber is approximately the same as, or smaller than, the volume of the primary cooling system, so that at least half of the initial coolant pressure is maintained. Thus, a sufficiently dense gaseous coolant is maintained such that heat transfer by the cooling system still contributes to the removal of thermal energy from the reactor fuel elements, so that at a reduced level the reactor Allows continuous operation. Furthermore, the pressure in the chamber between the containment vessel and the reactor pressure vessel is increased, and therefore the density of the gas occupying the same chamber is increased, so that the pressure between the reactor pressure vessel wall and the containment vessel wall is increased. The thermal conductivity of the material is increased, and this increased conductivity increases the efficiency of heat transfer from the reactor core to the thermal warmer. Therefore, the operation of the passive heat dissipation means is also enhanced.

Other advantageous features of the embodiment of the invention shown in FIG.
A fuel handling space 76 is provided below the storage container 26. The fuel handling space communicates with the reactor core housed within the core housing 14 via a conduit 74 . The fuel handling space 76 contains various conduits and vents and mechanical devices adapted to receive the spherical fuel element from the bottom of the cobblestone bed core within the core housing 14. further adapted to inspect them and circulate them back into the rough bed core in the upper part of the core receptacle 14 or to divert them for disposal of spent fuel. (the vent passages and mechanical devices in the fuel handling space and the conduit means for recirculating the spherical fuel elements are not shown for simplicity). With such a configuration, rupture of conduit 74 or other fuel handling equipment is a certainly possible accident condition that must be taken into account. To explain in more detail, rupture of the above-mentioned hard object,
i.e. large enough that said spherical fuel element blows away said fuel handling device with sufficient velocity;
A rupture that causes them to shatter upon impact results in the release of fission products. However, the provision of the containment vessel 26 according to the invention provides protection against contamination replacing the atmosphere in that case. Because the water-filled tank 34 acts as a biological shield, the balance of the containment vessel is also effective in containing radioactive contamination.

Another noteworthy feature is that the chamber 28 is also suitably shaped, so that both the inner surface of the tank 34 and the outer surface of the reactor pressure vessel 12 are completely sealed for periodic non-destructive test results. It becomes something that can be approached.

Of course, in some instances, the shape of the chamber allows for such non-destructive testing, and the chamber is configured such that the chamber contains leaked water in a volume smaller than the level of the reactor pressure vessel. It may also be true that the desirability of shaping the chamber is a consideration comparable to the minimization of the volume of said chamber in order to maximize the equilibrium pressure in the case of uncontrolled leakage of the primary cooling gas. However, in such circumstances it is permissible to depart from the option of maintaining the volume of the chamber at an amount comparable to that of the volume of the primary cooling gas. This is because a low coolant leak equilibrium pressure is allowed.

In any case, it is desirable to maintain such equilibrium pressure at the same value as practicable by minimizing the volume of the chamber to the extent possible.

An alternative embodiment of the invention is shown in FIGS. 4 and 5. A nuclear reactor (nuclear reactor) system 100 includes a reactor pressure vessel 102 made of non-insulated steel, and the reactor pressure vessel 102 has an overall cylindrical shape,
Therein, in this example the pressure is maintained at 50 bar. A generally cylindrical core housing 104 is disposed within the reactor pressure vessel, which is also made of steel and is not insulated. A plurality of spherical fuel elements 106 are disposed within the core housing, and the fuel elements 106 have fuel particles embedded therein and coated in multiple stages, and these have an annular shape. They are arranged in a cobblestone bed (foundation)-like core 107.

Near the core is a graphite reflector 108, which is itself surrounded by a ceramic shell 110. A control rod 112 made of any suitable material, for example haunium or boron carbide, can be inserted into the graphite reflector 108 between the reactor pressure vessel 102 and the core housing 104. , a gas space 114. Non-insulated steel container 11G
is itself a pressure vessel (in which the pressure is maintained at 1 bar in this example) and is arranged around the reactor pressure vessel 102.
2, thereby creating a chamber 1 between said two containers.
18 is formed. The storage container has an internal tank wall 1
20 and an external tank wall 122, which cooperate to form a tank 124 containing water 126, which is disposed around the reactor pressure vessel and therefore the core 107. The inside of the tank 124 is
Two pipe systems namely pipes 128 and 1
30 communicates with a condensing device 132. Alternatively, a 1-pipe system can be used. The condensing device is cooled with air by natural convection.

The reactor pressure vessel 102 and containment vessel 116 are different from the reactor pressure vessel 12 and the containment vessel 26A shown in FIG. 3 and described in connection with FIG. is also somewhat larger, but of similar size dimensions.

When the reactor is in a critical power producing state, a small portion of the thermal energy generated in the core by the spherical fuel elements is absorbed by the 1N and ceramic materials (108
# and 110 respectively), through the core containment body 104, across the gas space 114, and through the reactor pressure vessel 1.
02 wall, across the chamber 118, and inside wall 1
20 and thereby absorbed by water 126 in tank 124. The isothermal absorption causes steam to be generated, which is communicated via via 128 to condenser 132. After the appreciable latent heat is removed from the vapor in condenser 132, the condensed liquid water is returned to tank 124 via pipe 130. The condensing device 132 is approximately 150°
It is suitably sized to achieve boiling balance at F.

The primary cooling gas, helium, is supplied to the gas circulation system @ 13
4, gas space 114 in the direction indicated by the arrow.
is forced downward through the Identical "cold J helium gas, i.e. relatively low temperature (e.g. about 300 degrees Celsius)
) passes through passages within the ceramic and graphite materials (110i and 108, respectively) and then radially into the core 107 containing the spherical fuel elements 106.
and into heat exchange region 144 along the path indicated by arrow 136 . After passing through the core, the primary helium gas is a "hot" or relatively high temperature (eg, 900° C.) gas, removing thermal energy from the core. In the heat exchange zone [144], the "hot" primary helium gas is placed in heat transfer relationship with a cold secondary helium gas, ie, one that is at a relatively low temperature (eg, about 200° C.). The thermal energy absorbed by the primary helium gas is therefore transferred to the secondary helium gas, thereby cooling the primary helium gas, which then passes through the annular core/heat exchange structure. is conducted into passageway 138 at the center of . The primary helium, which is currently “cold” (relatively low temperature),
Gas is conducted up passageway 138 in the direction indicated by the arrow within passageway 138 and into circulation device 134 for recirculation.

Secondary "cold" helium gas is conducted to the heat exchange area through conduit 140 in the direction indicated by the arrow. It is maintained at a slightly higher pressure than the primary gas circuit. The secondary "cold" helium
Gas is directed into a heat exchange zone [144] so that it can absorb heat from the primary "hot" helium gas emerging from the reactor core. The secondary helium gas moves through the heat exchange region in a U-shaped path as shown by arrow 14G. This path of travel is achieved in practice by conducting said secondary helium gas through passages 154 in a graphite heat exchanger, as shown in cross-section in FIG. . When the secondary helium gas appears at the bottom of the heat exchange region, it is heated by the same energy while absorbing the thermal energy of the primary [hot J gas helium] supplied to the heat exchange region.

The secondary "hot" helium gas is then transferred to conduit 14
2 and through conduit 142 and to the exterior of the nuclear reactor system, such as for use in power generation or industrial processing heat. If there is a leakage from the primary to the secondary gas heat exchanger, the leakage will become part of the secondary cooling gas inside and therefore
Minimizes the potential for contamination of the primary gas by leakage of fission products.

As in the embodiment discussed in connection with FIG.
The volume of the chamber 118 is sufficiently small (preferably approximately the same volume as that of the primary coolant) so that leakage (even uncontrolled leakage) of the primary helium gas coolant cannot occur. Even so, a subatmospheric pressure (preferably about half of the initial primary coolant pressure) will be maintained within the primary cooling system. Also, the chamber 118 can be shaped such that a sufficiently large portion thereof is located below the level of the reactor pressure vessel, so that the water 126 leaks from the tank 124 into the chamber 118. The stabilized level of volume water will then be lower than that of the reactor pressure vessel 102.

Additionally, nuclear reactor system 100 is provided with piping or conduits and other fuel handling equipment for removing and/or recirculating spherical fuel elements 106 from the reactor core. Upon removal, these fuel elements travel along path 148 to conduit 150, which communicates the fuel elements to fuel handling space 152, where they are connected to fuel handling equipment (not shown for simplicity). Inspections, etc. will be conducted by

Similar to the embodiment shown in FIG. 3, the containment vessel 116 is not only part of a passive heat dissipation structure, but also constitutes a water-filled tank, which contains biological Acting as a shield and barrier, thus resulting in release of products of nuclear fission from the fuel element, leakage of contaminated primary or secondary cooling gas, fragmentation of spherical fuel elements due to rupture of the fuel handling equipment, etc. , to accommodate the radioactive material that is now present in the small chamber.

In operation, the system shown in FIGS. 4 and 5 functions in the same general manner as described for the system shown in FIG. 3, thereby providing passive thermal energy dissipation. and prevent the release of products of fission even when coupled to a vacuum in the primary cooling system despite the loss of forced gas cooling.

FIG. 6 shows yet another embodiment of the invention. Nuclear reactor system 200 includes a core 202 comprising a rough bed reactor containing spherical graphite fuel elements, the graphite fuel elements being It has fuel particles embedded within a spherical fuel element 204 and coated with multi-stage ceramic.

The reactor core 202 is surrounded by a graphite reflector 206 and a ceramic membrane 208, which are in turn housed within a core housing 210. Reactor pressure vessel 2
12 (in which the pressure is maintained at approximately 50 bar in this case) are arranged around said core housing 210, thereby forming a gas space 214 between them. The control rod 216 is connected to the graphite reflector 20
6.

A containment vessel 218 (also a pressure vessel, in which a pressure of approximately 1 par is maintained) is arranged around said reactor pressure vessel 212, which is in an annular configuration, thereby A small chamber 222 is formed between them. The container 218 has an inner wall 250 and an outer wall 25.
2, which cooperate to form a tank 254. A liquid, in this case water, is disposed within the tank 254. The chamber 222 has a sufficiently large portion located below the reactor pressure vessel;
The tank 254 can be configured to contain water that escapes therein as a result of a leak (even an uncontrolled leak), and the level of water that leaks into the chamber is controlled by the reactor pressure vessel. becomes stable below .

The reactor pressure vessel and containment vessel include the reactor pressure vessel 512 shown in and described in connection with FIG. 3;
It is larger than the portion of the storage container 26A that surrounds it, but has dimensions of approximately the same size.

A gas circulation device ff (blowing device 234) is installed above the cobble bed reactor core 202 and within the confines of the containment vessel 212. It is driven by an electric motor, which is connected to the 11ii ring system. and is hermetically sealed within the pressure boundaries of the reactor and uses magnetic bearings. Within the reactor, the primary "cold" helium gas (introduced at a relatively low temperature, e.g., about 300° C.) is forced to move along an axial path that is The same "hot" helium gas, which becomes "hot" due to its absorption of energy, is then circulated into heat exchange region 238 along the path indicated by arrow 240.

Dogenko FJ! The device is made of non-metallic materials, which are suitable for such devices in other embodiments of the invention. Secondary “cold” helium gas (relatively low temperature, eg, about 200° C.) is conducted through conduit 242 into heat exchange region 238 .

The primary "hot" helium gas and the secondary "cold" helium gas are placed in a heat-transferring relationship within an equivalent exchange region, so that the thermal energy absorbed by the primary helium gas is transferred to the secondary helium gas. The primary "cold" helium gas then emerges from the heat exchange region 238 as shown by arrow 248 and is directed into passageway 246 through which it passes through arrow g]248 The circulation device 234 is electrically connected as shown by
be returned to. At the same time, secondary "hot" helium gas is removed from heat exchange region 238 via conduit 244 and extracted to a remote location where it is used for power generation or industrial process heat generation. Adopted for preparation etc.

The interior of tank 254 communicates via a one-pipe system with condenser 226, which is cooled with air by natural convection. Of course, a two-pipe system could be used instead. When thermal energy is transferred from the core 202 to the water in tank 254, steam is generated which condenses Ha 22
6 to pipe 228, in a condenser 226 the same vapor is cooled and condensed to remove heat (which is released to the atmosphere, which is hot air), and then as a liquid (again through via 228) to a tank. 254. Thus, a passive, continuously operating heat dissipation system is provided.

Also, in this example, the conduit 232 is connected to the core 202.
and the fuel handling area [230]. The spherical fuel element 204 can be removed from the cylindrical core 202 via a conduit 232, thus providing a fuel handling area within the fuel handling area and in communication with the upper end of the core. It is received, inspected, and circulated by equipment (not shown for simplicity).

Small again! ! ! 218 is small enough so that even if there is an uncontrolled leakage of some cooling gas into the chamber 218, its equilibrium pressure will be greater than atmospheric pressure. Preferably, the chamber 218 is approximately the same volume as that of the primary cooling system, so that the balancing pressure is approximately 1.5 times that of the original pressure of the primary coolant.

Similar to the other embodiments shown in the previous figures, the water-filled tank surrounding the core constitutes a biological shield, in addition to passive heat dissipation, and provides atmospheric protection. prevent the release of radioactive contamination to Similarly, the containment vessel constitutes a barrier against environmental contamination by radioactive materials escaping into the small v222. This is due to the risk of nuclear fission products escaping into the atmosphere and the primary or secondary cooling of helium.
Minimizing the risk of gas contamination or rupture of the fuel handling equipment resulting in evacuation and shredding of the fuel elements so as to release radioactive material.

The containment structure described herein is also suitably configured to provide effective missile protection for the nuclear reactor system. The shell of the tank means as described above can be made sufficiently strong and heavy to provide considerable protection against collisions. Alternatively, the system can be housed in a building constructed from interfitting pre-stressed concrete elements, or the system can be
They can also be placed in underground silos covered with heavy missile-resistant concrete slabs or the like.

The present invention therefore provides an inherently safe nuclear reactor system, which can be activated by automatic start-up or by the use of purely passive heat dissipation means that are independent of operator action. , permitting a loss of forced coolant circulation and/or a reduction in the pressure of said coolant. The present invention also provides a nuclear heat source that, because of its small physical size, is more economical to construct, operate, maintain, and decommission under controlled conditions. suitable for in-factory manufacture, and inherently safe. As a result, the cost of complying with government p14III is reduced and the radiological cleanliness and breeding resistance of the nuclear reactor system is increased over conventional nuclear reactor systems. Furthermore, the present invention has a nuclear reactor of 45! and this reactor is subject to uncontrolled leakage of the primary coolant, as well as to the fuel circulation, which leads to the destruction of the fuel element during an impact, leading to its evacuation and the consequent release of radioactive material. In the event of a rupture of a utility line or other fuel handling equipment, the possibility of radioactivity being interchanged with the surrounding environment is greatly minimized. The present invention therefore represents a significant advance in the art.

The foregoing terms and expressions employed are used as terms of description and not of limitation, and the use of such terms and expressions includes the illustrations and expressions. It is recognized that there is no intention to exclude all equivalents of the described features, or parts thereof, and that various modifications are possible within the scope of the invention.

[Brief explanation of the drawing]

FIG. 1 is a plot of fuel temperature versus time elapsed since loss of forced cooling for the high temperature, gas-cooled, nuclear reactor system of the present invention. FIG. 2 is a plot of temperature within a nuclear reactor system according to the present invention versus radial distance from the longitudinal axis of the core. FIG. 3 is a cross-sectional schematic diagram of the nuclear reactor system of the present invention. FIG. 4 is a cross-sectional schematic diagram of another nuclear reactor system according to the invention. FIG. 5 is a schematic diagram of a cross-section of the system of FIG. 4 taken along line A-A. FIG. 6 is a schematic cross-sectional view of yet another embodiment of the invention. 10... Nuclear reactor (nuclear reactor) system 12... Reactor pressure vessel 14... Core vessel 15... Core
16...Fuel element 18...Reflector
20...Ceramic acid 22...Control Rondo
24... Gas space 26... Storage container
28... Small room 30... Inner wall 32.
・・External tank wall 34・Tank 36・
... Water 38 ... Conduit system 40 ... Condensing device 42 ... Passage 44.46 ... Conduit 48 ... Mixing device 49 ... Heat exchange system 50 ... Support cover piece 51. ...Pressure vessel 5
2...Volume 54...Heat exchange equipment equipment 5
5... conduit 56... secondary "cold" helium gas header 6
0... Secondary "hot" gas pipe 64... Annular region 66... Flow skirt 68... Conduit 70... Circulation device 7
4... Conduit 16... Fuel handling space 100... Nuclear reactor system 102...
Reactor pressure vessel 104... Core container 106...
Fuel element 107... Rough stone base core 108.
...Reflector 110...Ceramic acid 11
2... Control rod 114... Gas space 116
...Storage container 118...Small room 120...
Internal tank wall 122...External tank wall 124...
・Tank 126...Water 128, 130
...Pipe 132...Condensation'lA position 134...
・Gas circulation lrt13g...Passage 140...Conduit
144... Heat exchange area 150... Conduit
152...Fuel handling space 154...
Passage 200...Nuclear reactor (nuclear reactor) system 202...
Core 204...Fuel element 206...
Reflector 208...Ceramic film 210
... Core container 212 ... Reactor pressure vessel 2
14... Gas space 216 River control Rondo 218
...Storage container 222...Small room 226...
Condensing device 228...Vibe 232...Conduit 234...Gas 1lSrjA device 238
... Heat exchange area 242 ... Conduit 24400.
Conduit 246...Passway 250...Inner wall 252...External wall 254...Tank (voluntary) procedure Nem Author: Commissioner of the Patent Office February 2, 1988
4th day 1? Displaying S items
Patent Application No. 61-315975 2, Name of the invention Nuclear reactor system and its manufacturing method 3, Relationship with the case of the person making the amendment Address of patent applicant: State of Connecticut, United States of America 06340
Groton Poconnock Road 591 Name Proto-Power Corporation 4, Agent Address: 106 7th floor, Uraiya Building, 5-2-1 Roppongi, Minato-ku, Tokyo! (479) 2367 (7318)
Patent Attorney Yanagi 1) Seishi (and 2 others) 8. Contents of amendment The handwritten details will be corrected into a typed engraving. 1 Poor) Procedural amendment Old 1, Indication of the case 1986 Patent Application No. 315.975 2, J1 Ming's name Nuclear reactor system and its manufacturing method 3, Relationship with the person making the amendment Patent applicant name Proto - Power Corporation Representative John El Helm 4, Agent address: 6, 7th floor, Hauraiya Pill, 5-2-1 Roppongi, Minato-ku, Tokyo Number of inventions increased by amendment
None 7. Target of correction: Drawing 8, details of correction: Correct hand-drawn drawings into inked drawings. (No change in content)

Claims (28)

[Claims] (1) In a high temperature, gas-cooled, nuclear reactor system suitable for use as a reference unit in combination with one or more similar nuclear reactor systems, a nuclear reactor core contained within said reactor system. passive heat sink means for absorbing the decaying thermal energy generated in the reactor; and when said reactor is in a critical power generating condition, forced circulation loss of gaseous coolant, or loss of coolant. said heat reservoir means at a sufficient rate to maintain the ability of said heat reservoir means to absorb decaying thermal energy so that release of products of fission resulting from such loss in connection with the reduced pressure of said heat sink means is prevented. A nuclear reactor system comprising: means for removing thermal energy from a heat reservoir means; (2) the passive heat storage means is a tank means,
Claim 1, characterized in that the tank means is adapted to hold a liquid and is positioned for absorption by the liquid of decaying thermal energy emitted by the reactor core. The nuclear reaction system according to item 1. (3) A high-temperature, gas-cooled, nuclear reactor system suitable for use as a reference unit in combination with one or more other similar nuclear reactor systems, which is identical to the reactor pressure vessel. A nuclear reaction that includes a nuclear reactor core contained within a reactor pressure vessel and having a plurality of fuel elements arranged as a cobblestone reactor and a cooling system with pressurized and forced circulating gas. In the system, a containment vessel is arranged around the reactor pressure vessel and is spaced apart from the reactor pressure vessel, thereby forming a chamber between the plurality of vessels, and a portion of the chamber is formed on an inner wall. and tank means including an outer wall, said tank means being adapted to hold a liquid and to absorb decaying thermal energy generated in said reactor core and exiting said reactor pressure vessel. The containment vessel is adapted to contain leaked reactor gas coolant positioned for absorption by such liquid and is itself placed in heat transfer relationship with the reactor core. means for removing thermal energy from the vapor released by the liquid as a result of absorption of the attenuated thermal energy; and means for removing thermal energy from the vapor. as a result, comprising means for returning liquid to said tank means, wherein said tank means of said containment vessel is under boiling equilibrium conditions established in cooperation with said thermal energy removal means and liquid return means. , to prevent the loss of forced circulation of gas coolant, or the release of products of fission resulting from such loss in conjunction with depressurization of the coolant, when the reactor is in critical and power producing conditions. A nuclear reactor system adapted to retain an effective amount of said liquid so as to absorb a sufficient amount of decaying thermal energy. (4) gas cooling of the reactor core, wherein the gas cooling system of the nuclear reactor is located at least partially within the reactor pressure vessel and is adapted to remove thermal energy generated in the core; means for establishing a heat transfer relationship between a primary cooling gas and the reactor core to remove thermal energy generated in the reactor core from the reactor core; and means for establishing a heat transfer relationship between a secondary cooling gas and the primary cooling gas to remove heat from the primary cooling gas. Nuclear reactor system as described in Section. (5) The nuclear reactor system according to claim 4, wherein the gas cooling system of the nuclear reactor includes a nonmetallic gas-to-gas heat exchange device. (6) further comprising means for maintaining the secondary cooling gas at a higher pressure than the primary cooling gas so that any leakage flow between the two gases enters the primary cooling gas; A nuclear reactor system according to claim 4, characterized in that: (7) A nuclear reaction according to claim 4, wherein the means for creating a heat transfer relationship between the primary and secondary cooling gases is within the confines of the reactor pressure vessel. Furnace system. (8) A nuclear reactor according to claim 4, wherein the means for creating a heat transfer relationship between the primary and secondary cooling gases is located outside the reactor pressure vessel. system. (9) The nuclear reactor system according to claim 4, further comprising means for directing the flow of the primary cooling gas in an axial direction or a radial direction with respect to the core. (10) The core has a cylindrical outer shape, and the means for directing the flow of the primary cooling gas directs the flow in an axial direction with respect to the core. The nuclear reactor system according to item 9. (11) The core has a cylindrical outer shape, and the means for directing the flow of the primary cooling gas directs the flow radially with respect to the core. The nuclear reactor system according to item 9. (12) The nuclear reactor system according to claim 4, wherein the primary and secondary cooling gases are helium. (13) The nuclear reactor system according to claim 3, wherein the reactor core has a cylindrical or annular outer surface. (14) The means for condensing the coolant is suitably dimensioned so that boiling equilibrium is established at a temperature of about 150°F.
Nuclear reactor system as described in Section. (15) the gas cooling system maintains a primary cooling gas in heat transfer relationship with the reactor core to remove thermal energy generated in the core with a gas-to-gas heat exchange device; first setting means for setting an adapted volume; and a volume adapted for holding a secondary cooling gas in a heat transfer relationship with said primary cooling gas to remove thermal energy from said primary cooling gas. second setting means for setting, the volume of the chamber between the reactor pressure vessel and the containment vessel being sufficiently small such that when there is an uncontrolled leakage of primary cooling gas into the chamber; 4. The nuclear reactor system according to claim 3, wherein the value is higher than that set by the first setting means. (16) A portion of the small chamber between the containment vessel and the reactor pressure vessel is arranged below the level of the reactor pressure vessel, and the portion of the small chamber is sufficiently large, so that the small chamber Claim 3, characterized in that uncontrolled leakage of the liquid held in the tank means into the chamber results in a stable liquid level in the chamber below the level of the reactor pressure vessel. Nuclear reactor system as described in Section. (17) further comprising a fuel handling area containing conduit means and other fuel handling equipment for removing or recirculating fuel elements from the reactor core; characterized in that it is adapted to constitute a barrier against the release of radioactive material escaping from said fuel element in the event of rupture and consequent expulsion of said fuel element therefrom and shredding. A nuclear reactor system according to claim 3. (18) A small reference unit nuclear reactor system suitable for producing high temperatures of electrical power or industrial process heat and adapted to prevent the accidental release of the products of nuclear fission, the nuclear reactor a pressure vessel; and a core containing body housed within the reactor pressure vessel and separated from the reactor pressure vessel, the core containing body containing fuel particles coated in multiple stages. a core container containing a plurality of spherical fuel elements arranged in a cobblestone reactor to form a core; and means for controlling the reaction within the core. and means for gaseous cooling of the reactor core, the means being at least partially disposed within the reactor pressure vessel and adapted to remove thermal energy generated in the core; the means comprising a heat exchange device for the gas; and a first setting means housed in the gas cooling means, the means for directing the primary cooling gas in the volume set by the first setting means to the nuclear reactor. a first setting means for maintaining the core in a heat transfer relationship; and a second setting means housed in the gas cooling means, the secondary setting means in a volume set by the second setting means. the second setting means for maintaining the cooling gas in a heat transfer relationship with the primary cooling gas;
a containment vessel disposed around the reactor pressure vessel and spaced from the reactor pressure vessel, thereby
forming a chamber between the two containers, the receiving container including tank means having an inner wall and an outer wall, the tank means being adapted to hold a liquid, and the tank means having a liquid in the core; and a containment vessel positioned for absorption by the liquid of decaying thermal energy exiting the reactor pressure vessel, the containment vessel itself being in heat transfer relationship with the reactor core. a pressure vessel adapted to contain leaked reactor gas coolant, the volume of the chamber between the reactor pressure vessel and the containment vessel being sufficiently small;
Therefore, even when there is an uncontrolled leakage of primary cooling gas into said chamber, the equilibrium pressure of said primary cooling gas within the volume set by said first setting means is still preferably higher than atmospheric pressure. , the volume of that part of the chamber between the containment vessel and the reactor pressure vessel that is lower than the reactor pressure vessel is sufficiently large so that uncontrolled leakage of the liquid into the chamber is resulting in a stable liquid level in said chamber below the level of the pressure vessel, and means for condensing the vapor released by said liquid as a result of absorption of said decaying thermal energy; and from said vapor. Means are provided for recirculating the condensed liquid to said liquid, said storage vessel tank means being capable of recirculating said liquid under conditions of boiling equilibrium established in cooperation with said condensing means and said recirculating means. , to prevent the loss of forced circulation of gas coolant or the release of products of fission resulting from such loss in conjunction with depressurization of the coolant when the reactor is in critical and power generating conditions; A small reference unit nuclear reactor system, characterized in that it is adapted to hold an effective amount of said liquid in order to absorb a sufficient amount of decaying thermal energy to (19) Manufacture a high temperature, gas-cooled, nuclear reactor suitable for use as a reference unit in conjunction with one or more similar nuclear reactors to prevent the release of products of nuclear fission due to overheating. a method for disposing passive heat sink means in communication with said nuclear reactor, said means being adapted to remove decaying thermal energy from a reactor core disposed within said nuclear reactor; and means for removing thermal energy from the thermal sink means at a sufficient rate so as to maintain the ability of the thermal sink means to absorb decreasing thermal energy. means in operative communication with the reactor, thereby causing loss of forced circulation of gaseous coolant, or in connection with depressurization of the coolant, when said reactor is in a critical power generating condition. the release of products of fission resulting from such losses is prevented. (20) the passive heat sink means is a tank means adapted to hold a liquid and absorb by the liquid decaying thermal energy emitted from the reactor core; 20. A method according to claim 19, characterized in that the method is located for: (21) A high-temperature, gas-cooled, nuclear reactor suitable for use as a reference unit in combination with one or more other similar nuclear reactors to prevent the release of products of nuclear fission due to overheating. A method of manufacturing a nuclear reactor, the reactor core having a reactor pressure vessel and a plurality of fuel elements contained within the reactor pressure vessel and arranged as a cobblestone reactor; a pressurized and forcedly circulated gas cooling system, in which containment vessels are spaced apart around the reactor pressure vessel;
Thereby, a chamber is formed between the same plurality of containers, and therefore,
A portion of said containment vessel comprising tank means comprising an inner wall and an outer wall is positioned for containing thermal energy dissipated from said reactor core by means of a liquid that said tank means is adapted to retain. and providing means in operative association with the tank means for removing thermal energy from the vapor released by the liquid as a result of absorption of the attenuated thermal energy; providing means for returning liquid resulting from the removal of thermal energy from the vapor to the tank means in operative communication with the tank means and the thermal energy removal means; The tank means is configured to provide for forced release of gas coolant when the reactor is in critical power generation conditions under conditions of boiling equilibrium established in cooperation with the heat removal and liquid return means. absorbing a sufficient amount of decaying thermal energy to prevent loss of circulation, or release of products of fission resulting from such loss in connection with depressurization of the coolant;
the containment vessel is adapted to hold an effective amount of the liquid, and the containment vessel is itself capable of containing leaked reactor gas coolant in heat transfer relationship with the reactor core. A method characterized in that the pressure vessel has: (22) establishing a volume within the gas cooling system adapted to maintain a primary cooling gas in heat transfer relationship with the reactor core for removing decaying thermal energy generated in the core; a first setting means for setting a volume adapted to hold a secondary cooling gas in a heat transfer relationship with said primary cooling gas for removing thermal energy from said primary cooling gas; further comprising merging the chambers between the reactor pressure vessel and the containment vessel, the volume of the chamber being small enough so that even when there is an uncontrolled leakage of primary cooling gas into the chamber, 22. A method as claimed in claim 21, characterized in that the primary coolant equilibrium pressure in the volume set by the first setting means is still preferably above atmospheric pressure. (23) The accommodation vessel is arranged around the reactor pressure vessel, so that a part of the small chamber between the plurality of vessels is located lower than the level of the reactor pressure vessel, and a stable liquid level in the chamber which is such that the chamber portion is sufficiently large so that uncontrolled leakage of the liquid into the chamber that the tank means is adapted to retain is below the level of the reactor pressure; 22. The method of claim 21, further comprising causing the result to be. (24) The release of products of nuclear fission due to overheating from a high temperature, gas-cooled, nuclear reactor suitable for use as a reference unit in combination with one or more similar nuclear reactors. A method of preventing the use of passive heat sink means arranged in communication with said nuclear reactor to absorb the amount of decaying thermal energy generated in a reactor core contained within said nuclear reactor. to do and
and the loss of forced circulation of gas coolant, or the release of products of nuclear fission resulting from such loss in conjunction with depressurization of the coolant, when said nuclear reactor is in critical power generating conditions. removing thermal energy from the heat sink means at a sufficient rate to maintain the capacity of the heat sink means for absorbing declining thermal energy so as to prevent Method. (25) the passive heat sink means is a tank means adapted to hold a liquid and absorb by the liquid decaying thermal energy emitted by the reactor core; 25. A method according to claim 24, characterized in that the method is located for: (26) a reactor pressure vessel, a reactor core having a plurality of fuel elements disposed within the reactor pressure vessel and arranged as a cobblestone reactor, and a pressurized and forced circulating gas cooling system; The products of fission are released due to overheating from a high-temperature, gas-cooled, nuclear reactor suitable for use as a reference unit in combination with one or more similar nuclear reactors containing A method for preventing the deterioration of thermal energy generated within the reactor core into a liquid disposed around the reactor pressure vessel while being spaced from the reactor pressure vessel. absorbing; and removing thermal energy from the vapor released by the liquid as a result of the absorption of the attenuated thermal energy;
and returning liquid to the liquid as a result of the removal of thermal energy from the vapor, the amount of the liquid being established in cooperation with the removal of the thermal energy and the returning of the liquid. Due to loss of forced circulation of gaseous coolant, or such loss in connection with depressurization of the coolant, when the reactor is in critical power generating conditions, under conditions of boiling equilibrium, the amount of attenuating thermal energy is sufficient to prevent the release of products of nuclear fission. (27) The nuclear reactor further performs heat transfer with a containment vessel disposed around and spaced apart from the reactor pressure vessel and the reactor core to remove thermal energy generated in the core. a gas cooling system containing a first setting means for setting a volume adapted to hold a primary cooling gas in relation thereto; 27. The method of claim 26, further comprising the step of providing electrical communication into the space between the vessel and the reactor pressure vessel. (28) When the liquid leaks, retaining the leaked liquid in the space between the plurality of vessels so that the level of the leaked liquid is lower than that of the reactor pressure vessel. 27. The method of claim 26, further comprising: