Note : Les descriptions sont présentées dans la langue officielle dans laquelle elles ont été soumises.
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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when the control rods are withdrawn as well as when they are inserted.
The flow path has a relatively low pressure drop when they are withdrawn,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avoid the increase in core by-pass when the rods are withdrawn but it does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.
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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
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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
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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
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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.
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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
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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.
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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:
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