GB1582107A - Nuclear reactor - Google Patents

Abstract

In the water-cooled nuclear reactor, the vertically arranged control rods (50) pass through the outlet plenum (26) into the fission chamber (12), these control rods being surrounded by thimbles. A major portion of the cooling water flow is fed directly to the floor of the fission chamber (12), so that flow through the latter takes place in an upwards direction. A small portion pressurises a space (72) in the upper part of the tank and flows downwards through the rod thimbles and thus reaches the major portion of the cooling water flow in the lower part of the fission chamber. The result is to provide a cooling arrangement in which the control rod cooling is less critical with respect to the flow control and the performance in the case of an emergency shutdown of the reactor. <IMAGE>

Description

(54) NUCLEAR REACTOR (71) We, COMBUSTION ENGINEERING, INC., a corporation organized and existing under the laws of the State of Delaware, United States of America, of 1000 Prospect Hill Road, Windsor, Connecticut, United States of America, do hereby declare the invention, for which we pray that a patent may be granted to us, and the method by which it is to be performed, to be particularly described in and by the following statement: This invention relates to nuclear reactors and particularly to water cooled pressurized water nuclear reactors.

A water cooled pressurized water nuclear reactor conventionally includes a core formed of vertically supported fuel elements and vertically movable control rods passing therethrough. These control rods are surrounded by guide tubes at least through the core to assure proper guidance of their movement. Flow of the coolant water is upward through the core to insure stability of flow in the event of any localized steam or overheating.

While the control rods contain no fuel they do absorb neutrons and thereby generate some heat. Cooling of the control rods is, therefore, required. The conventional method of cooling these rods involves passing a portion of the flow upwardly through the control rod guide tubes which may then exit either in the outlet chamber or in an upper portion of the reactor vessel from which location the flow passes to the outlet.

The flow passing through these guide tubes is in parallel with the flow actually passing over and cooling the fuel assemblies. It, therefore, must be severely restricted to avoid an undue reduction in the thermal performance of the core. This flow must pass through the guide tubes when the control rods are withdrawn as well as when they are inserted. The flow path has a relatively low pressure drop when they are withdrawn, and a concomitant increase in flow. In order to restrict these variations in flow, restricted orifices must be placed at the inlets of the guide tubes. This cannot avoid the increase in core by-pass of coolant when the rods are withdrawn but it does minimize the extent to which the flow increases. The use of restricted orifices involves not only the expense of installing these but also the possibility of plugging, which is inherent in any flow restriction which is put within a nuclear reactor.

The selection of the particular by-pass flow quantity through the control rod guide tubes requires a critical allocation of flow, since there must be sufficient flow properly to cool the rods in the fully inserted position, but any excess flow used needlessly degrades the thermal performance of the reactor core.

Since flow is upwardly along the control rods there is an upward force due to the drag of the fluid flow as well as the pressure difference between the bottom and upper portions of the control rods. The force resists the downward movement required in scramming a reactor, thereby lengthening scram time and increasing the forces required to drive the control rods down, beyond what they would be in the absence of such flow arrangements.

In a conventional arrangement, the pressure below the core is higher than the pressure at the outlet of the core due to the friction drop of the flow passing therethrough. This results in a significant upward force on the core in the order of 3,000,000 newtons for a 280 kilopascal pressure drop. Since the entire upper portion of a conventional reactor vessel is at the outlet pressure this force can be resisted only by structures which transmit the force to the reactor vessel or reactor head.

In the conventional arrangement, the upper portion of the reactor vessel is not only at outlet pressure but also at outlet temperature. The core support barrel is the structure which separates the two pressure and temperature volumes. The support barrel is generally supported at the top of the reactor vessel body immediately adjacent the bolted joint between the head and the body. The complex structure in this area must not only tolerate the physical forces due to the internal pressure as transmitted through the bolts but must also simultaneously tolerate the thermal stresses due to the temperature difference on the two sides of the core barrel at the joint area.

According to the present invention there is provided a water cooled pressurised water nuclear reactor comprising a core, vertically extending guide tubes within said core, vertically movable control rods within said guide tubes, means for supplying a first major quantity of water coolant adjacent the bottom of said core, for upward flow therethrough, and means for supplying a second minor quantity of water coolant to the guide tubes adjacent their upper ends, for downward flow therethrough, said guide tubes having openings adjacent their lower ends whereby in use, coolant flowing downwardly through the guide tubes may join with the coolant supplied adjacent the bottom of said core so as to flow therewith upwardly through the core.

Preferably a common supply of water coolant is divided into said first major quantity of water coolant and said second minor quantity of coolant.

Preferably the reactor includes means defining an outlet chamber above the core, for coolant flowing upwardly through the core in use.

In a preferred arrangement the reactor comprises a reactor vessel body, a core support barrel surrounding said core and forming an outer periphery of said outlet chamber, said barrel supporting said core and defining an inlet chamber therebelow, said barrel being supported within said reactor vessel whereby an annular space is defined therebetween, said annular space and said inlet chamber being in fluid communication, the means defining said outlet chamber including a seal plate spaced above said core, a head of the reactor vessel defining a second inlet chamber above the seal plate, and there being an aperture in said barrel such that said annular space and said second inlet chamber are in fluid communication, said second inlet chamber being in flow communication with said guide tubes.

In use of such a preferred embodiment the major portion of the water flow follows the conventional flow path, upwards through the core. Water passes into the vessel and downwardly between the core support barrel and the vessel entering the core at the bottom, and then passes upwardly therethrough. A smaller portion of the flow, however, passes through the core support barrel to the upper portion of the reactor vessel thereby producing a pressure level in the top of the vessel which is significantly above the core outlet pressure. The flow from this location passes downwardly through the guide tubes to cool the control rods and joins the major portion of the flow near the bottom of the core. This minor portion of flow joins the major portion at this location so that the total flow then passes upwardly through the core in contact with the fuel assemblies. There is constant circulation of both portions of coolant during operation of the reactor.

The force required to scram the control rods is reduced as a result of this flow path.

Since the flow is downwardly through the control rod guide tubes all drag forces aid in scramming control rods. Furthermore, since the pressure at the top of the control rods is nearer the inlet pressure than outlet pressure there is an additional pressure differential to aid in the scram of the control rods.

This arrangement also avoids or minimizes by-pass of coolant flow around the core. Since all the flow which passes over and cools the control rods joins the main flow before passage through the core there is no by-pass of the core. The only by-pass that could occur is that due to leakage at any sealed joints in the structure. Such leakage would only be a function of one's ability to effect tight seals and not a function of any flow required for cooling. The seal provided by normal fits between the fuel assemblies and their guide structure is sufficiently good to reduce leakage flow to a fraction of that which is currently accepted in conventional control rod cooling arrangements.

Since by-pass of the flow which passes over the control rods is avoided, this decreases the criticality of the design to allocate flow to cooling the control rod. Substantial excess flow can be used to cool the control rods since it has no deleterious effect on reactor performance. Therefore, restricted orifices are not required in the guide tubes for the purpose of limiting flow.

This preferred structure also provides a pressurized chamber in the upper portion of the reactor vessel. The pressure here is approximately the inlet pressure to the vessel, as compared to the outlet pressure in prior art designs. This pressure exerts a substantial downward force on the seal plate which separates this pressurized chamber from the outlet chamber. Since the seal plate can be connected to other structures such as the core support barrel it reduces or eliminates the additional force required to hold down the core support barrel. Furthermore, this core structure hold down force is a function of the actual reactor coolant flow.

Therefore uncertainties in the coolant flow, in design or operation, are automatically compensated by appropriate hold down force variations.

Since the coolant inlet temperature is maintained not only in the annular space between the core support barrel and the reactor vessel but also in the second inlet chamber above the seal plate, the temperature difference at the closure for the vessel head is reduced. This reduces thermal stresses in the bolts during steady state operation and minimizes them during transient operation.

An embodiment of the invention will now be described by way of example and with reference to the accompanying drawings, in which : Figure 1 is a sectional elevation of the general arrangement of a nuclear reactor in accordance with the invention which illustrates the general structure and the coolant flow paths therethrough; and Figure 2 is an isometric view of detail in the region of the outlet chamber.

Referring now to the drawings, a reactor vessel body 2 and a reactor vessel head 4 are joined by a bolted connection at flange 6. The reactor vessel body has an inlet opening 8 and an outlet opening 10 for flow of coolant water therethrough.

A core 12 is comprised of a plurality of fuel assemblies 14, each of which is comprised of a plurality of elongate fuel rods.

The core is supported on a core support assembly 16 which is in turn supported by a core support barrel 18. This core support barrel is supported by a flange 20 from the reactor vessel body 2 at a location adjacent the flange 6.

Immediately above the core 12 is a fuel assembly alignment plate 22 which serves to engage the upper ends of the fuel assemblies and to maintain alignment thereof. A seal plate structure 24 is located above the alignment plate, thereby defining an outlet chamber 26.

After the coolant enters through inlet opening 8 a first quantity comprising the major part of the flow passes downwardly through annular space 28 between the reactor vessel and the core support barrel. This flow passes downwardly through a flow skirt 30 into an inlet chamber 32 located below the core 12. The flow then passes upwardly through the core and through openings in the alignment plate 22 into the outlet chamber 26. From here the flow passes out through outlet opening 10 to a steam generator (not shown).

Each of the fuel assemblies 14 contains within its structure four control rod guide tubes 40 which pass through the entire length of the fuel assembly. These guide tubes extend upwardly above an upper fuel assembly end plate 42. The extensions are surrounded by hold down springs 44 which bear against a fuel assembly upper end fitting 46.

These end fittings in turn bear against the fuel assembly alignment plate 22 whereby the fuel assemblies 14 are held down through the compressive action of the springs.

Elongate control rods 48 are vertically movable within the guide tubes 40 of the fuel assemblies. Each of these rods individually extends to a position above the seal plate 24 at which location they may be joined in subgroupings to a control rod extension 50.

In addition to a plurality of flow holes 52, the alignment plate 22 also has openings 54 through which the control rods pass.

The upper ends of the guide tubes 40 pass i to these openings with a machined close fit. These joints should be such as to take horizontal forces so that the fuel assemblies can be aligned, and must permit vertical movement to allow for expansion of the different fuel assemblies. Since coolant leakage at these joints would by-pass the core, minimizing leakage is desirable. Conventional fits used for alignment should, however, be sufficient to maintain by-pass leakage well below that of prior art designs.

Control rod shroud tubes 56 pass through the outlet chamber 26 and may be welded to the alignment plate 22 and the seal plate structure 24. These shroud tubes surround and protect the control rods from the effects of cross flow through the chamber 26.

Extending above the seal plate 24 is a control assembly shroud 58. This surrounds a group of control rods which are joined to a single control rod extension. This shroud protects the control rods from localized transverse flow effects.

Since the seal plate 24 is used not only to seal but is also part of the structural arrangement for the upper guide assembly it is supported from barrel 60 to form a more rigid structure. Furthermore, it permits the entire structure including the fuel assembly alignment plate 22 to be removed when refueling to expose the fuel assemblies. This barrel 60 is supported by flanges 62 resting on flanges 20 of the core support barrel.

An upper guide structure support plate 64 is open to permit flow therethrough.

A flow opening 70 is provided through the core support barrel and also through the upper guide assembly barrel so that a second mnior portion of the flow entering the reactor vessel passes into a second inlet chamber 72. The control assembly shrouds 58 are open at their upper ends and may have openings at various locations throughout their length whereby the minor portion of flow passes downwardly inside these shrouds. The flow then passes downwardly through the control element shroud tubes 56 into the fuel assembly control rod guide tubes 40. This second minor portion of flow continues through the length of the fuel assemblies inside the guide tube to a location near the bottom of core 12 where it passes outwardly joining the first main portion of flow. These two flows are then combined and the total quantity passes up wardly through the core 12 and outlet chamber 26. There is a constant circulation of both portions of coolant.

It can be seen that the two parallel flow paths exist between the inlet and the bottom of core 12. The pressure drop is essentially established by the larger first portion of flow passing down through the annular space 28.

The remaining portion of the flow passing through the other path experiences the same pressure drop with the flow being established by the geometry of the flow path. It is preferred that the portion of this flow path from the inlet 8 to the pressurized second inlet chamber 72 be of low resistance and, therefore, have a relatively low pressure drop.

The portion of the flow path through the assembly shroud and ultimately through the guide tubes 40 should have a major portion of the available pressure drop. This tends to maintain the pressure in the pressurized chamber 72 at a relatively high level.

It also results in improved distribution between the various control rod guide tubes.

The flow passing through the guide tubes should be sufficient to remove all the heat generated within the control rods. Since none of the flow by-passes the core, this flow may be conveniently selected on the high side thereby resulting in increased design tolerances.

Since flow is downwardly along the control rods the drag forces tend to aid in reactor scram. Furthermore, while the lower ends of the control rods are exposed to core inlet pressure the upper ends are exposed to the higher pressures in the second inlet chamber 72 thereby further establish irg a pressure differential tending to force the control rods down. Both of these characteristics aid in reducing scram time and in reducing the drive forces required.

The relatively high pressure in the second inlet chamber 72, which approximates to the inlet pressure to the reactor, results in a force on the upper side of the seal plate structure 24. The opposite side of that plate is exposed to the outlet pressure in chamber 26. If plates 24 and 22 along with the control rod shroud tubes 56 are considered to be a unitary structure the opposing force would be the pressure immediately below the fuel assembly alignment plate 22. This pressure is only slightly above the pressure in the outlet chamber 26. The pressure differential across either of these structures then is approximately equal to the pressure drop through the reactor vessel, which would be expected to be in the order of 280 kilopascals. If the plates have a diameter in the order of 3.7 metres, this amounts to 3,000,000 newtons of downward force. The core support barrel and the upper guide structure barrel of conventional designs require substantial structure to withstand the upward force produced in the core and on the other reactor elements due to the upward flow therethrough. This downward force due to the pressure difference counteracts the upward force thereby significantly reducing the amount of structure which is required to hold the reactor internals down against the reactor vessel itself. The forces tending to raise the components are a function of the flow through the reactor. It should be noted that the downward force generated by the pressure differential is of course a function of this pressure differential which in turn is a function of the flow through the reactor vessel. Therefore, the force resisting the upward thrust varies in accordance with the same parameter which increases the upward thrust and, therefore, tends to be selfcompensating with variations of flow through the reactor and with variations in deposits which may occur generally throughout the flow path.

Not only is the pressure at inlet 8 and in chamber 72 approximately equal but the temperature of the fluid is substantially equal in both locations. It follows, therefore, that during steady state operation there is no significant temperature difference across the flanges 20, 62 and 6 due to fluid temperature differences. This reduces thermal stresses in this area where pressure induced stresses are already high due to the complex nature of a bolted connection.

WHAT WE CLAIM IS:- 1. A water cooled pressurised water nuclear reactor comprising a core, vertically extending guide tubes within said core, vertically movable control rods within said guide tubes, means for supplying a first major quantity of water coolant adjacent the bottom of said core, for upward flow therethrough, and means for supplying a second minor quantity of water coolant to the guide tubes adjacent their upper ends, for downward flow therethrough, said guide tubes having openings adjacent their lower ends whereby in use, coolant flowing downwardly through the guide tubes may join with the coolant supplied adjacent the bottom of said core so as to flow therewith upwardly through the core.

2. A nuclear reactor as claimed in claim 1, in which a common supply of water coolant is divided into said first major quantity of water coolant and said second minor quantity of coolant.

3. A nuclear reactor as claimed in claim 1 or 2, including means defining an outlet chamber above the core, for coolant flowing upwardly through the core in use.

4. A nuclear reactor as claimed in claim 3, comprising a reactor vessel body, a core support barrel surrounding said core and

**WARNING** end of DESC field may overlap start of CLMS **.

Claims (6)

**WARNING** start of CLMS field may overlap end of DESC **. wardly through the core 12 and outlet chamber 26. There is a constant circulation of both portions of coolant. It can be seen that the two parallel flow paths exist between the inlet and the bottom of core 12. The pressure drop is essentially established by the larger first portion of flow passing down through the annular space 28. The remaining portion of the flow passing through the other path experiences the same pressure drop with the flow being established by the geometry of the flow path. It is preferred that the portion of this flow path from the inlet 8 to the pressurized second inlet chamber 72 be of low resistance and, therefore, have a relatively low pressure drop. The portion of the flow path through the assembly shroud and ultimately through the guide tubes 40 should have a major portion of the available pressure drop. This tends to maintain the pressure in the pressurized chamber 72 at a relatively high level. It also results in improved distribution between the various control rod guide tubes. The flow passing through the guide tubes should be sufficient to remove all the heat generated within the control rods. Since none of the flow by-passes the core, this flow may be conveniently selected on the high side thereby resulting in increased design tolerances. Since flow is downwardly along the control rods the drag forces tend to aid in reactor scram. Furthermore, while the lower ends of the control rods are exposed to core inlet pressure the upper ends are exposed to the higher pressures in the second inlet chamber 72 thereby further establish irg a pressure differential tending to force the control rods down. Both of these characteristics aid in reducing scram time and in reducing the drive forces required. The relatively high pressure in the second inlet chamber 72, which approximates to the inlet pressure to the reactor, results in a force on the upper side of the seal plate structure 24. The opposite side of that plate is exposed to the outlet pressure in chamber 26. If plates 24 and 22 along with the control rod shroud tubes 56 are considered to be a unitary structure the opposing force would be the pressure immediately below the fuel assembly alignment plate 22. This pressure is only slightly above the pressure in the outlet chamber 26. The pressure differential across either of these structures then is approximately equal to the pressure drop through the reactor vessel, which would be expected to be in the order of 280 kilopascals. If the plates have a diameter in the order of 3.7 metres, this amounts to 3,000,000 newtons of downward force. The core support barrel and the upper guide structure barrel of conventional designs require substantial structure to withstand the upward force produced in the core and on the other reactor elements due to the upward flow therethrough. This downward force due to the pressure difference counteracts the upward force thereby significantly reducing the amount of structure which is required to hold the reactor internals down against the reactor vessel itself. The forces tending to raise the components are a function of the flow through the reactor. It should be noted that the downward force generated by the pressure differential is of course a function of this pressure differential which in turn is a function of the flow through the reactor vessel. Therefore, the force resisting the upward thrust varies in accordance with the same parameter which increases the upward thrust and, therefore, tends to be selfcompensating with variations of flow through the reactor and with variations in deposits which may occur generally throughout the flow path. Not only is the pressure at inlet 8 and in chamber 72 approximately equal but the temperature of the fluid is substantially equal in both locations. It follows, therefore, that during steady state operation there is no significant temperature difference across the flanges 20, 62 and 6 due to fluid temperature differences. This reduces thermal stresses in this area where pressure induced stresses are already high due to the complex nature of a bolted connection. WHAT WE CLAIM IS:-

1. A water cooled pressurised water nuclear reactor comprising a core, vertically extending guide tubes within said core, vertically movable control rods within said guide tubes, means for supplying a first major quantity of water coolant adjacent the bottom of said core, for upward flow therethrough, and means for supplying a second minor quantity of water coolant to the guide tubes adjacent their upper ends, for downward flow therethrough, said guide tubes having openings adjacent their lower ends whereby in use, coolant flowing downwardly through the guide tubes may join with the coolant supplied adjacent the bottom of said core so as to flow therewith upwardly through the core.

2. A nuclear reactor as claimed in claim 1, in which a common supply of water coolant is divided into said first major quantity of water coolant and said second minor quantity of coolant.

3. A nuclear reactor as claimed in claim 1 or 2, including means defining an outlet chamber above the core, for coolant flowing upwardly through the core in use.

4. A nuclear reactor as claimed in claim 3, comprising a reactor vessel body, a core support barrel surrounding said core and

forming an outer periphery of said outlet chamber, said barrel supporting said core and defining an inlet chamber therebelow, said barrel being supported within said reactor vessel whereby an annular space is defined therebetween, said annular space and said inlet chamber being in fluid communication, the means defining said outlet chamber including a seal plate spaced above said core, a head of the reactor vessel defining a second inlet chamber above the seal plate, and there being an aperture in said barrel such that said annular space and said second inlet chamber are in fluid communication, said second inlet chamber being in flow communication with said guide tubes.

5. A nuclear reactor as claimed in claim 4 wherein said seal plate is supported by said barrel.

6. A nuclear reactor substantially as hereinbefore described with reference to the accompanying drawings. - -