CA1091827A - Pressurized water reactor flow arrangement - Google Patents
Pressurized water reactor flow arrangementInfo
- Publication number
- CA1091827A CA1091827A CA295,607A CA295607A CA1091827A CA 1091827 A CA1091827 A CA 1091827A CA 295607 A CA295607 A CA 295607A CA 1091827 A CA1091827 A CA 1091827A
- Authority
- CA
- Canada
- Prior art keywords
- core
- reactor vessel
- guide tubes
- reactor
- pressurized
- Prior art date
- Legal status (The legal status is an assumption and is not a legal conclusion. Google has not performed a legal analysis and makes no representation as to the accuracy of the status listed.)
- Expired
Links
Classifications
-
- G—PHYSICS
- G21—NUCLEAR PHYSICS; NUCLEAR ENGINEERING
- G21C—NUCLEAR REACTORS
- G21C1/00—Reactor types
- G21C1/04—Thermal reactors ; Epithermal reactors
- G21C1/06—Heterogeneous reactors, i.e. in which fuel and moderator are separated
- G21C1/08—Heterogeneous reactors, i.e. in which fuel and moderator are separated moderator being highly pressurised, e.g. boiling water reactor, integral super-heat reactor, pressurised water reactor
- G21C1/086—Pressurised water reactors
-
- G—PHYSICS
- G21—NUCLEAR PHYSICS; NUCLEAR ENGINEERING
- G21C—NUCLEAR REACTORS
- G21C15/00—Cooling arrangements within the pressure vessel containing the core; Selection of specific coolants
- G21C15/02—Arrangements or disposition of passages in which heat is transferred to the coolant; Coolant flow control devices
-
- Y—GENERAL TAGGING OF NEW TECHNOLOGICAL DEVELOPMENTS; GENERAL TAGGING OF CROSS-SECTIONAL TECHNOLOGIES SPANNING OVER SEVERAL SECTIONS OF THE IPC; TECHNICAL SUBJECTS COVERED BY FORMER USPC CROSS-REFERENCE ART COLLECTIONS [XRACs] AND DIGESTS
- Y02—TECHNOLOGIES OR APPLICATIONS FOR MITIGATION OR ADAPTATION AGAINST CLIMATE CHANGE
- Y02E—REDUCTION OF GREENHOUSE GAS [GHG] EMISSIONS, RELATED TO ENERGY GENERATION, TRANSMISSION OR DISTRIBUTION
- Y02E30/00—Energy generation of nuclear origin
- Y02E30/30—Nuclear fission reactors
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- Physics & Mathematics (AREA)
- Engineering & Computer Science (AREA)
- Plasma & Fusion (AREA)
- General Engineering & Computer Science (AREA)
- High Energy & Nuclear Physics (AREA)
- Monitoring And Testing Of Nuclear Reactors (AREA)
- Structure Of Emergency Protection For Nuclear Reactors (AREA)
Abstract
Abstract of the Disclosure A pressurized water nuclear reactor of the type with vertical control rods passing through the core surrounded by control rod guide tubes.
A major portion of the water flow is passed directly to the bottom of the core for upward flow therethrough. A minor portion pressurizes a volume at the top of the vessel and passes downwardly through the control rod guide tubes to Join the major portion of flow at the lower end of the core.
A major portion of the water flow is passed directly to the bottom of the core for upward flow therethrough. A minor portion pressurizes a volume at the top of the vessel and passes downwardly through the control rod guide tubes to Join the major portion of flow at the lower end of the core.
Description
Background of the Invention This invention relates to a pressurized water cooled nuclear reacltor and in particular to a flow path therethrough.
A pre6surized water cooled nuclear reactor conventionslly includes a core formed of vertically supported ~uel elements snd vertically movable control rod~ passing therethrough. These control rods are surrounded by guide `
tubes at least through the core to as3ure proper guidance of their movement.
Flow of the coolant water i8 upward through the core to insure stability of n ow in the event of any localized steam or overheating.
While the control rods contain no ~uel 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 portlon of the flow upwardly through the control rod guide tubes which may then exlt elther ln the outlet plenum or in an upper portion of the reactor ves~el from which location the ~low pa3~e3 to the outlet.
The ~low pas ing through the3e guide tubes is ln parallel with the flow actually passlng over and cooling the fuel assemblies. It, therefore, must be severely restrictea to avoid an undue reduction in the ther~al performance of the core. Thi~ 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, orifices must be placed at the inlet of the guide tubes. This cannot avoid the increase in core by-pass when the rods are withdrawn but it does minimize the extent to which the n ow increases. The use Or orifices involves not only the expense of installing these but also the potentisl of plugging which is inherent in any flow restriction which i8 put within a nuclear reactor.
The selectioh of the particular by-pa3s flow quantity through the control rod 6~ide tubes requires a critical allocation of flow, since there must be sufficient flow to properly cool the rods in the fully inserted position, but any excess flow used needlessly degrades the thermal performance of the reactor core.
C700430 -2- ~d~
, Since flow is upwardly alo~g the control rods there is an upward force due to the drag of the fluid flow as well as the pressure difference betwe!en the bottom and upper portion of the control rod. The force resists the downward movement reguired in scramming a reactor, thereby lengthening scram time and increasing the forces required to drire the control rods down beyond what they would be in the absence of such flow arrangements.
In the convention~l arrangement, the pressure belcw the core is higher in the pressureat the outlet of the core due to the friction drop of the n ow 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 i9 at the outlet presfiure this ~orce can be resisted oDly by structures which transmit the force to the reactor vessel of reactor head.
In the conventional arr~ngement, the up~er portion of the reactor vecsel 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 ad~acent the bolted ~oint between the head and the body. The complex ~tructure in this area must not only tolerate the physical force~ 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 ~oint area.
Summary of the Invention In the nuclear reactor according to the present inYention the ma~or portion of the water n ow follows the conventional flow path. It 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 minor portion of the n ow, however, passes through the core support barrel to the upper portion of the reactor vessel thereby effecting a pressure level in the top of the vessel which is significantly above the core outlet pressure. The flow from this location passes downwardly ~09i827 through the guide tubes to cool the control rods and ~oins the ma~or portion of the flow near the bottom of the core. This minor portion of flow ~oins the m~or portion at this location 80 that the total flow passes upwardly through the core in contact with the fuel assemblies.
The rorce 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 st the top of the control rod approaches inlet pressure rather than outlet pressure there is an additional pressure differential to aid in the scram Or the control rods.
This arrangement also avoids or mimimizes by-pass of flow around the core, Since all the flow which passes over and cools the control rods ;~
~oins the main flow before passage through the core there is no by-pass of the core. The o~ly by-pass that could occur is that due to loakage at any ~ealed ~olnts ~n the structure, Such leakage would only be a nunction of one's ability to effect tight seals and not a function o~ any flow required for cooling. Tho seal provided by normal fits between the fuel assemblies and their guide structure is sufficiently good to reduce leakage n ow to a fraction of that which is currently accepted in con~entional control rod cooling arrangements Since by-pass of the flow which pssses 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 ~t has no deleterious effect on reactor performance.
Therefore, orifices are not reguired in the control roa channels for the purposeof limiting flow.
The structure also provides a pressurized vol~me in the upper portion of the reactor vessel. This is approximately the inlet pressure to the vessel as compared to the outlet pressure in prior art designs.
The presence of this pressure exerts a substantial downward force on the seal plate which separates this pressurized dome from the outlet plenum.
Since the seal plate can be connected to other structure~ such as the core ~760430 ~4~
1~9i8Z7 support barrel it reduces or elimlnates the supplemental force required to hold down the core support barrel. Furthermore, this core structure hold down force i8 a runction of the actual reactor coolant flow. Therefore, uncertalnties in the coolant n ow, in design or operation, are automatically compl~nsated by appropriate hold down force variation.
Since the inlet temperature exists not only in the annular space batween the core support barrel and the reactor vessel but also in the dome, the temperature difference at the reactor vessel closure is reduced. This reduces thermal stresses in the bolts during steady state operation and minimizes them during transient operation. ~ ;
Brief DescriPtion of the Drawin~s Figure l is a sectional elevation of the general arrangement of a nuclear reactor which illustrates the general structure and the flow paths therethrough.
Figure 2 is an isometrlc view of a detail in the area of the outlet plenum.
De3cri~tion of the Preferred Embodiment A reactor vessel body ~ and a reactor vessel head 4 are ~oined by a bolted connection at n ange 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 elongated fuel rods. The core is supported on the core support assembly 16 which is in turn supported by the core support barrel 18. This core support barrel is supported ~y flange 20 from the reactor vessel body 2 at a location ad~acent the flange 6.
Immediately above the core 12 is a fuel assemDly ali~nment 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 the outlet plenum 26.
After the coolant enters through inlet opening ô a first quantity comprising the bulk of the flow passes downwardly through the annular spsce 28 between the reactor vessel and the core support barrel. This flow C760~30 -5-passes downwardly through the flow skirt 30 into an inlet plenum 32 located below the core 12. The flow passes upwardly through the core and through openings in thea~ignment plate 22 into the outlet plenum 26. From here the flow passes out through outlet opening 10 to a steam generator (not shown).
Each of the fuel assemblies 14 contain within their structure four control rod guide tubes 40 which pass through the entirelength of the fuel assembly. These guide tubes extend upwardly above the upper fuel assembly end plate 42. The extensions are surrounded by hold down springs 44 which bear a6ainst the fuel assembly upper end fitting 46. These end fittings in turn bear against the fuel assembly alignment plate 22 whereby the fuel assemblie~ 14 are held down through the compressive action of the springs.
Finger shaped control rods 48 are vertically movable within the guide tubes 40 of the ~uel assemblies. Each o~ these rods individually extends to a~ elevatlon above the seal plate 24 at which location they may be Joined ln subgroupings to the control rod extension 50.
In addition to the ~low holes 52, the alignment plate 22 also has openings 54 through which the control rods pas6. The extensions of the guide tube 40 pass into these openings with a machined close ~it. This Joint should be such as to take horizontal forces 80 that the ~uel assmeblies can be aligned, and must permit vertical movement to allow for expansion of the different ~uel assemblies. Since leakage at this Joint bypasses the core, minimizing leakage is efficacious in carrying out this invention.
Conventior,al fits used for alignment are, however, 3ufficient to maintain by-pass leakage well below the core by-pass o~ prior art designs.
Control rod shroud tubes 56 pass through the outlet plenum 26 and may be welded to the alignment plate 22 and the seal plate structure 24.
These shroud tubes surround and protect the control roas from the ef~ects of cross flow through the plenum 26.
Extending above the seal plate 24 is the control assembly shroud 58. Ihis surrounds a group of control rods which are Joined to a sinsle control rod extension. This shroud protects the control rods from localized transverse flow effects.
109i8Z7 Since the seal plate 24 is used not only as a seal plate but ~lso as part of the structural arrangement for the upper guide assembly it is sup~orted from barrel 60 to form a more rigid structure. Furthermore, it permits the entire structure includlng the fuel assembly alignment plate 22 to b~ romoved when refueling to expose the fuel assemblies. This barrel 60 is supported by flanges 62 resting on n anges 20 of the core support barrel.
The upper guide gtructure support plate 6~ 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 minor portion of the flow entering the reactor vessel passes into the pressurized chamber 72. The control assembly shrouds 58 are open at the upper end and may have openings at varlous locations throughout the length whereby the minor portion of flow passes downwardly inside these shrouds. The n ow then passes downwardly through the control elemsnt shroud tube 56 into the fuel a~sembly control rod guide tubes 40. This second minor portion of ~low continues through the length of the fuel a~semblies inside the guide tube to a location near the bottom of core 12 where it passes outwardly ~oining the first main portion of n ow. These two flows are then combined and the total gua~tity passes upwardly through the core 12 and outlet plenum 26.
It can be seen the two parallel flow paths exist between the inlet 8 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 chamber 72 be of low resistance and, therefore, ha~e 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 ma~or portion of the available pressure drop. This tends to maintain the pressure in the pressurized plenum 72 at a relatively high C760430 -7~
109182~7 pressure level. It also results in improved distribution between the various control rod guide tubes.
The design ~low passing through the guide tubes should be sufficient to remove all the heat Benerated withln the control rods. Since none of the flow by-passes the core, this n ow may be conveniently selected on the high side thereby resulting in increased design tolerance Slnce n ow is downwardly along the control rods the drag forces tend to aid in reactor scram. Furthermore, while the lower end or the control rod is exposed to core inlet pressure the upper end is exposed to the higher pressure in the pressurized chamber 72 thereby further establishing a pressure differential tending to force the control rods down. ~oth of these characteristics aid in reducing scram time and in reducing the drive ;~
forces required.
The relativoly high pressurs in the pres~urizèd chamber 72, which approxlmate~ the inlet pressure to the reactor, exerts its force on the upper side or the seal plate structure 24. The opposite side of that plate is expoaed to the outlet pressure in plenum 26. Ir plates 24 and 22 along with the control rod shroud tubes 56 are consiaered 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 plenu~ 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 meters, 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 snd on the other reactor elements due to the upward n ow 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.
~O91~Z 7 It should be noted that the downward force generated by the pressure differential i~ of course a function of this pressure differential which in turn Ls a function of the flow through the reactor vessel. ~herefore, the ~rce resisting the upward thrust raries in accordance with the same par&meter which increases the upward thrust and, therefore, tends to be sel~-compen3ating with variations of flow through the re~ctor and with variations in deposits which may occur generally throughout the n ow path.
Not only is the pressure at inlet 8 and in plenum 72 approximately equal but the temperature of the fluid is equal in both locations. It follows, therefore, that d~ring steady state operation there is no significant temperature difference across the n anges 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.
While a preferred embodiment of the invention has been illustrated and described, it is understood that this is merely illustrative and not restrictive and that variations and modifications may be made therein without departing from the spirit and scope of the invention.
What is claimed is:
A pre6surized water cooled nuclear reactor conventionslly includes a core formed of vertically supported ~uel elements snd vertically movable control rod~ passing therethrough. These control rods are surrounded by guide `
tubes at least through the core to as3ure proper guidance of their movement.
Flow of the coolant water i8 upward through the core to insure stability of n ow in the event of any localized steam or overheating.
While the control rods contain no ~uel 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 portlon of the flow upwardly through the control rod guide tubes which may then exlt elther ln the outlet plenum or in an upper portion of the reactor ves~el from which location the ~low pa3~e3 to the outlet.
The ~low pas ing through the3e guide tubes is ln parallel with the flow actually passlng over and cooling the fuel assemblies. It, therefore, must be severely restrictea to avoid an undue reduction in the ther~al performance of the core. Thi~ 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, orifices must be placed at the inlet of the guide tubes. This cannot avoid the increase in core by-pass when the rods are withdrawn but it does minimize the extent to which the n ow increases. The use Or orifices involves not only the expense of installing these but also the potentisl of plugging which is inherent in any flow restriction which i8 put within a nuclear reactor.
The selectioh of the particular by-pa3s flow quantity through the control rod 6~ide tubes requires a critical allocation of flow, since there must be sufficient flow to properly cool the rods in the fully inserted position, but any excess flow used needlessly degrades the thermal performance of the reactor core.
C700430 -2- ~d~
, Since flow is upwardly alo~g the control rods there is an upward force due to the drag of the fluid flow as well as the pressure difference betwe!en the bottom and upper portion of the control rod. The force resists the downward movement reguired in scramming a reactor, thereby lengthening scram time and increasing the forces required to drire the control rods down beyond what they would be in the absence of such flow arrangements.
In the convention~l arrangement, the pressure belcw the core is higher in the pressureat the outlet of the core due to the friction drop of the n ow 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 i9 at the outlet presfiure this ~orce can be resisted oDly by structures which transmit the force to the reactor vessel of reactor head.
In the conventional arr~ngement, the up~er portion of the reactor vecsel 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 ad~acent the bolted ~oint between the head and the body. The complex ~tructure in this area must not only tolerate the physical force~ 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 ~oint area.
Summary of the Invention In the nuclear reactor according to the present inYention the ma~or portion of the water n ow follows the conventional flow path. It 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 minor portion of the n ow, however, passes through the core support barrel to the upper portion of the reactor vessel thereby effecting a pressure level in the top of the vessel which is significantly above the core outlet pressure. The flow from this location passes downwardly ~09i827 through the guide tubes to cool the control rods and ~oins the ma~or portion of the flow near the bottom of the core. This minor portion of flow ~oins the m~or portion at this location 80 that the total flow passes upwardly through the core in contact with the fuel assemblies.
The rorce 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 st the top of the control rod approaches inlet pressure rather than outlet pressure there is an additional pressure differential to aid in the scram Or the control rods.
This arrangement also avoids or mimimizes by-pass of flow around the core, Since all the flow which passes over and cools the control rods ;~
~oins the main flow before passage through the core there is no by-pass of the core. The o~ly by-pass that could occur is that due to loakage at any ~ealed ~olnts ~n the structure, Such leakage would only be a nunction of one's ability to effect tight seals and not a function o~ any flow required for cooling. Tho seal provided by normal fits between the fuel assemblies and their guide structure is sufficiently good to reduce leakage n ow to a fraction of that which is currently accepted in con~entional control rod cooling arrangements Since by-pass of the flow which pssses 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 ~t has no deleterious effect on reactor performance.
Therefore, orifices are not reguired in the control roa channels for the purposeof limiting flow.
The structure also provides a pressurized vol~me in the upper portion of the reactor vessel. This is approximately the inlet pressure to the vessel as compared to the outlet pressure in prior art designs.
The presence of this pressure exerts a substantial downward force on the seal plate which separates this pressurized dome from the outlet plenum.
Since the seal plate can be connected to other structure~ such as the core ~760430 ~4~
1~9i8Z7 support barrel it reduces or elimlnates the supplemental force required to hold down the core support barrel. Furthermore, this core structure hold down force i8 a runction of the actual reactor coolant flow. Therefore, uncertalnties in the coolant n ow, in design or operation, are automatically compl~nsated by appropriate hold down force variation.
Since the inlet temperature exists not only in the annular space batween the core support barrel and the reactor vessel but also in the dome, the temperature difference at the reactor vessel closure is reduced. This reduces thermal stresses in the bolts during steady state operation and minimizes them during transient operation. ~ ;
Brief DescriPtion of the Drawin~s Figure l is a sectional elevation of the general arrangement of a nuclear reactor which illustrates the general structure and the flow paths therethrough.
Figure 2 is an isometrlc view of a detail in the area of the outlet plenum.
De3cri~tion of the Preferred Embodiment A reactor vessel body ~ and a reactor vessel head 4 are ~oined by a bolted connection at n ange 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 elongated fuel rods. The core is supported on the core support assembly 16 which is in turn supported by the core support barrel 18. This core support barrel is supported ~y flange 20 from the reactor vessel body 2 at a location ad~acent the flange 6.
Immediately above the core 12 is a fuel assemDly ali~nment 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 the outlet plenum 26.
After the coolant enters through inlet opening ô a first quantity comprising the bulk of the flow passes downwardly through the annular spsce 28 between the reactor vessel and the core support barrel. This flow C760~30 -5-passes downwardly through the flow skirt 30 into an inlet plenum 32 located below the core 12. The flow passes upwardly through the core and through openings in thea~ignment plate 22 into the outlet plenum 26. From here the flow passes out through outlet opening 10 to a steam generator (not shown).
Each of the fuel assemblies 14 contain within their structure four control rod guide tubes 40 which pass through the entirelength of the fuel assembly. These guide tubes extend upwardly above the upper fuel assembly end plate 42. The extensions are surrounded by hold down springs 44 which bear a6ainst the fuel assembly upper end fitting 46. These end fittings in turn bear against the fuel assembly alignment plate 22 whereby the fuel assemblie~ 14 are held down through the compressive action of the springs.
Finger shaped control rods 48 are vertically movable within the guide tubes 40 of the ~uel assemblies. Each o~ these rods individually extends to a~ elevatlon above the seal plate 24 at which location they may be Joined ln subgroupings to the control rod extension 50.
In addition to the ~low holes 52, the alignment plate 22 also has openings 54 through which the control rods pas6. The extensions of the guide tube 40 pass into these openings with a machined close ~it. This Joint should be such as to take horizontal forces 80 that the ~uel assmeblies can be aligned, and must permit vertical movement to allow for expansion of the different ~uel assemblies. Since leakage at this Joint bypasses the core, minimizing leakage is efficacious in carrying out this invention.
Conventior,al fits used for alignment are, however, 3ufficient to maintain by-pass leakage well below the core by-pass o~ prior art designs.
Control rod shroud tubes 56 pass through the outlet plenum 26 and may be welded to the alignment plate 22 and the seal plate structure 24.
These shroud tubes surround and protect the control roas from the ef~ects of cross flow through the plenum 26.
Extending above the seal plate 24 is the control assembly shroud 58. Ihis surrounds a group of control rods which are Joined to a sinsle control rod extension. This shroud protects the control rods from localized transverse flow effects.
109i8Z7 Since the seal plate 24 is used not only as a seal plate but ~lso as part of the structural arrangement for the upper guide assembly it is sup~orted from barrel 60 to form a more rigid structure. Furthermore, it permits the entire structure includlng the fuel assembly alignment plate 22 to b~ romoved when refueling to expose the fuel assemblies. This barrel 60 is supported by flanges 62 resting on n anges 20 of the core support barrel.
The upper guide gtructure support plate 6~ 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 minor portion of the flow entering the reactor vessel passes into the pressurized chamber 72. The control assembly shrouds 58 are open at the upper end and may have openings at varlous locations throughout the length whereby the minor portion of flow passes downwardly inside these shrouds. The n ow then passes downwardly through the control elemsnt shroud tube 56 into the fuel a~sembly control rod guide tubes 40. This second minor portion of ~low continues through the length of the fuel a~semblies inside the guide tube to a location near the bottom of core 12 where it passes outwardly ~oining the first main portion of n ow. These two flows are then combined and the total gua~tity passes upwardly through the core 12 and outlet plenum 26.
It can be seen the two parallel flow paths exist between the inlet 8 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 chamber 72 be of low resistance and, therefore, ha~e 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 ma~or portion of the available pressure drop. This tends to maintain the pressure in the pressurized plenum 72 at a relatively high C760430 -7~
109182~7 pressure level. It also results in improved distribution between the various control rod guide tubes.
The design ~low passing through the guide tubes should be sufficient to remove all the heat Benerated withln the control rods. Since none of the flow by-passes the core, this n ow may be conveniently selected on the high side thereby resulting in increased design tolerance Slnce n ow is downwardly along the control rods the drag forces tend to aid in reactor scram. Furthermore, while the lower end or the control rod is exposed to core inlet pressure the upper end is exposed to the higher pressure in the pressurized chamber 72 thereby further establishing a pressure differential tending to force the control rods down. ~oth of these characteristics aid in reducing scram time and in reducing the drive ;~
forces required.
The relativoly high pressurs in the pres~urizèd chamber 72, which approxlmate~ the inlet pressure to the reactor, exerts its force on the upper side or the seal plate structure 24. The opposite side of that plate is expoaed to the outlet pressure in plenum 26. Ir plates 24 and 22 along with the control rod shroud tubes 56 are consiaered 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 plenu~ 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 meters, 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 snd on the other reactor elements due to the upward n ow 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.
~O91~Z 7 It should be noted that the downward force generated by the pressure differential i~ of course a function of this pressure differential which in turn Ls a function of the flow through the reactor vessel. ~herefore, the ~rce resisting the upward thrust raries in accordance with the same par&meter which increases the upward thrust and, therefore, tends to be sel~-compen3ating with variations of flow through the re~ctor and with variations in deposits which may occur generally throughout the n ow path.
Not only is the pressure at inlet 8 and in plenum 72 approximately equal but the temperature of the fluid is equal in both locations. It follows, therefore, that d~ring steady state operation there is no significant temperature difference across the n anges 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.
While a preferred embodiment of the invention has been illustrated and described, it is understood that this is merely illustrative and not restrictive and that variations and modifications may be made therein without departing from the spirit and scope of the invention.
What is claimed is:
Claims (8)
PROPERTY OR PRIVILEGE IS CLAIMED ARE DEFINED AS FOLLOWS:
1. In a water cooled pressurized water nuclear reactor having a flow of water coolant supplied to a location near said reactor, said reactor having a core, water coolant flowing upwardly through said core, vertical guide tubes within said core, and vertically movable control rods within said guide tubes the improvement comprising:
means for conducting a first major quantity of the water coolant from said location directly to the lower end of said core for upward flow therethrough; and means for conducting a second minor quantity of the water coolant from said location through a fixed resistance flow path. to the upper end of said guide tubes for downward flow therethrough, said guide tubes having openings in the lower portion thereof, in fluid communication with said first quantity of water, whereby siad first and second quantities of water coolant both pass upwardly through said core.
means for conducting a first major quantity of the water coolant from said location directly to the lower end of said core for upward flow therethrough; and means for conducting a second minor quantity of the water coolant from said location through a fixed resistance flow path. to the upper end of said guide tubes for downward flow therethrough, said guide tubes having openings in the lower portion thereof, in fluid communication with said first quantity of water, whereby siad first and second quantities of water coolant both pass upwardly through said core.
2. An apparatus as in claim 1 having also:
a seal plate structure spaced above said core, defining an outlet plenum above said core;
said guide tubes also passing vertically through the plenum.
a seal plate structure spaced above said core, defining an outlet plenum above said core;
said guide tubes also passing vertically through the plenum.
3. An apparatus as in claim 2 having also:
a reactor vessel body;
a core support barrel surrounding said core and forming the outer periphery of the outlet plenum, said barrel supporting said core and thereby forming an inlet plenum there below, said core support barrel also supported within said reactor vessel body thereby forming an annular space therebetween, the annular space and the inlet plenum being in fluid communication whereby said first quantity of water coolant passes therethrough;
a reactor vessel head defining a pressurized volume between said head and said seal plate;
an opening through said core support barrel near the upper end thereof, whereby the annular space and the pressurized volume are in fluid communication for passage of said second quantity of water coolant therethrough.
a reactor vessel body;
a core support barrel surrounding said core and forming the outer periphery of the outlet plenum, said barrel supporting said core and thereby forming an inlet plenum there below, said core support barrel also supported within said reactor vessel body thereby forming an annular space therebetween, the annular space and the inlet plenum being in fluid communication whereby said first quantity of water coolant passes therethrough;
a reactor vessel head defining a pressurized volume between said head and said seal plate;
an opening through said core support barrel near the upper end thereof, whereby the annular space and the pressurized volume are in fluid communication for passage of said second quantity of water coolant therethrough.
4. An apparatus as in claim 3:
wherein said seal plate structure is supported from said core support barrel.
wherein said seal plate structure is supported from said core support barrel.
5. A pressurized water nuclear reactor comprising:
a reactor vessel body having an inlet opening and an outlet opening;
a reactor vessel head secured to the top of said reactor vessel body;
a core supported within said reactor vessel body;
vertically movable control rods passing through said core;
guide tubes surrounding said control rods and vertically extending through said core;
an outlet chamber above said core;
a pressurized chamber located above said outlet chamber, the control rod guide tubes passing vertically through said outlet chamber into said pressurized chamber;
a first main water flow path from the inlet opening downwardly around the periphery of said core to the lower portion of said core;
a parallel water flow path of unvarying restriction from the inlet opening through said pressurized chamber and thence continuing through said guide tubes to the lower portion of said core; and a combined water flow path upwardly from the lower portion of the core through said core.
a reactor vessel body having an inlet opening and an outlet opening;
a reactor vessel head secured to the top of said reactor vessel body;
a core supported within said reactor vessel body;
vertically movable control rods passing through said core;
guide tubes surrounding said control rods and vertically extending through said core;
an outlet chamber above said core;
a pressurized chamber located above said outlet chamber, the control rod guide tubes passing vertically through said outlet chamber into said pressurized chamber;
a first main water flow path from the inlet opening downwardly around the periphery of said core to the lower portion of said core;
a parallel water flow path of unvarying restriction from the inlet opening through said pressurized chamber and thence continuing through said guide tubes to the lower portion of said core; and a combined water flow path upwardly from the lower portion of the core through said core.
6. An apparatus as in claim 5:
wherein said parallel water flow path has a low resistance from the inlet to said pressurized chamber and a high resistance from said pressurized chamber to the lower portion of said core.
wherein said parallel water flow path has a low resistance from the inlet to said pressurized chamber and a high resistance from said pressurized chamber to the lower portion of said core.
7. An apparatus as in claim 6:
wherein said outlet chamber includes a seal plate separating the outlet chamber from said pressurized chamber;
said seal plate being supported from said reactor vessel body.
wherein said outlet chamber includes a seal plate separating the outlet chamber from said pressurized chamber;
said seal plate being supported from said reactor vessel body.
8. An apparatus as in claim 7:
having also a core support barrel for supporting said core within said reactor vessel body; and said seal plate being supported on said core support barrel, thereby being in turn supported by said reactor vessel body.
having also a core support barrel for supporting said core within said reactor vessel body; and said seal plate being supported on said core support barrel, thereby being in turn supported by said reactor vessel body.
Applications Claiming Priority (2)
| Application Number | Priority Date | Filing Date | Title |
|---|---|---|---|
| US77346577A | 1977-03-02 | 1977-03-02 | |
| US773,465 | 1977-03-02 |
Publications (1)
| Publication Number | Publication Date |
|---|---|
| CA1091827A true CA1091827A (en) | 1980-12-16 |
Family
ID=25098369
Family Applications (1)
| Application Number | Title | Priority Date | Filing Date |
|---|---|---|---|
| CA295,607A Expired CA1091827A (en) | 1977-03-02 | 1978-01-25 | Pressurized water reactor flow arrangement |
Country Status (14)
| Country | Link |
|---|---|
| JP (1) | JPS591995B2 (en) |
| AU (1) | AU514977B2 (en) |
| BE (1) | BE864438A (en) |
| BR (1) | BR7801231A (en) |
| CA (1) | CA1091827A (en) |
| CH (1) | CH629328A5 (en) |
| DE (1) | DE2804937C3 (en) |
| ES (1) | ES467218A1 (en) |
| FR (1) | FR2382747A1 (en) |
| GB (1) | GB1582107A (en) |
| IT (1) | IT1092993B (en) |
| NL (1) | NL7802250A (en) |
| PT (1) | PT67726A (en) |
| SE (1) | SE430108B (en) |
Families Citing this family (6)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| FR2529705A1 (en) * | 1982-07-01 | 1984-01-06 | Framatome Sa | DEVICE FOR VERIFYING THE DISCONNECTION OF THE CONTROL CLUSTERS OF A NUCLEAR REACTOR |
| DE3373586D1 (en) * | 1983-05-13 | 1987-10-15 | Westinghouse Electric Corp | Nuclear reactor |
| FR2595501B1 (en) * | 1986-03-07 | 1988-06-10 | Framatome Sa | INTERNAL EQUIPMENT OF NUCLEAR REACTORS WITH EXTENDED TANK |
| JPH067180B2 (en) * | 1987-10-19 | 1994-01-26 | 動力炉・核燃料開発事業団 | Reactor with integrated pressure vessel structure |
| FR2627321B1 (en) * | 1988-02-11 | 1992-08-14 | Framatome Sa | SUPERIOR INTERNAL EQUIPMENT OF NUCLEAR REACTOR PROVIDED WITH A FLOW SEPARATION DEVICE |
| CN111370148B (en) * | 2018-12-25 | 2024-05-14 | 国家电投集团科学技术研究院有限公司 | Two sets of shutdown mechanisms of reactor and reactor |
Family Cites Families (8)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| BE586969A (en) * | 1959-01-29 | |||
| GB883717A (en) * | 1959-04-03 | 1961-12-06 | Babcock & Wilcox Ltd | Improvements in nuclear reactors |
| US3481832A (en) * | 1967-04-14 | 1969-12-02 | Combustion Eng | Nuclear reactor core and control element arrangement |
| US3770583A (en) * | 1971-05-20 | 1973-11-06 | Combustion Eng | Fuel assembly hold-down device |
| UST911015I4 (en) * | 1971-12-21 | 1973-06-26 | Nuclear core positioning system | |
| BE793195A (en) * | 1971-12-23 | 1973-04-16 | Combustion Eng | NUCLEAR REACTOR EQUIPPED WITH AN INDIVIDUAL HYDRAULIC DEVICE TO ACTUATE EACH CONTROL BAR |
| US3853703A (en) * | 1972-07-03 | 1974-12-10 | Combustion Eng | Fuel assembly hold-up device |
| DE2237208A1 (en) * | 1972-07-28 | 1974-02-07 | Siemens Ag | NUCLEAR REACTOR |
-
1978
- 1978-01-25 CA CA295,607A patent/CA1091827A/en not_active Expired
- 1978-02-06 DE DE2804937A patent/DE2804937C3/en not_active Expired
- 1978-02-21 CH CH187078A patent/CH629328A5/en not_active IP Right Cessation
- 1978-02-22 ES ES467218A patent/ES467218A1/en not_active Expired
- 1978-02-28 SE SE7802267A patent/SE430108B/en not_active IP Right Cessation
- 1978-02-28 AU AU33692/78A patent/AU514977B2/en not_active Expired
- 1978-03-01 JP JP53022156A patent/JPS591995B2/en not_active Expired
- 1978-03-01 PT PT67726A patent/PT67726A/en unknown
- 1978-03-01 BE BE185572A patent/BE864438A/en not_active IP Right Cessation
- 1978-03-01 BR BR7801231A patent/BR7801231A/en unknown
- 1978-03-01 FR FR7805872A patent/FR2382747A1/en active Granted
- 1978-03-01 NL NL7802250A patent/NL7802250A/en unknown
- 1978-03-02 GB GB8352/78A patent/GB1582107A/en not_active Expired
- 1978-03-03 IT IT20830/78A patent/IT1092993B/en active
Also Published As
| Publication number | Publication date |
|---|---|
| BR7801231A (en) | 1978-09-26 |
| AU514977B2 (en) | 1981-03-12 |
| SE430108B (en) | 1983-10-17 |
| AU3369278A (en) | 1979-09-06 |
| NL7802250A (en) | 1978-09-05 |
| IT7820830A0 (en) | 1978-03-03 |
| CH629328A5 (en) | 1982-04-15 |
| FR2382747A1 (en) | 1978-09-29 |
| DE2804937B2 (en) | 1980-04-03 |
| FR2382747B1 (en) | 1982-01-29 |
| JPS591995B2 (en) | 1984-01-14 |
| SE7802267L (en) | 1978-09-03 |
| IT1092993B (en) | 1985-07-12 |
| JPS53107595A (en) | 1978-09-19 |
| PT67726A (en) | 1978-04-01 |
| ES467218A1 (en) | 1979-02-01 |
| BE864438A (en) | 1978-07-03 |
| DE2804937A1 (en) | 1978-09-07 |
| GB1582107A (en) | 1980-12-31 |
| DE2804937C3 (en) | 1980-12-04 |
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Legal Events
| Date | Code | Title | Description |
|---|---|---|---|
| MKEX | Expiry |