Note: Descriptions are shown in the official language in which they were submitted.
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GUIDE ASSEMBLY
OF THE CORIUM LOCALIZING AND COOLING SYSTEM
OF A NUCLEAR REACTOR
Technical field of the invention
The invention is applicable to the corium localizing and cooling systems of a
nuclear reactor designed for localization of severe beyond design-basis
accidents, in
particular, to the devices for directing corium of a nuclear reactor to the
corium trap.
The accidents with core meltdown, which may take place during multiple
failure of the core cooling system, constitute the greatest radiation hazard.
During such accidents the core melt ¨ corium ¨ by melting the core structures
and reactor pressure vessel, flows out beyond its limits, and as a consequence
of the
decay heat retained in it may disturb the NPP containment integrity ¨ the last
barrier
on the escape routes of radioactive products to the environment.
It is required to localize corium escaping from the reactor pressure vessel
for
excluding this, and provide its continuous cooling up to complete
crystallization of all
the corium components. The corium trap performs this function, which after
entry of
corium into it prevents the NPP containment damage, thereby protecting the
public
and environment against radiation impact during severe accidents of the
nuclear
reactors by cooling and subsequent crystallization of corium.
After the reactor vessel rupture corium enters the guide assembly, which is
usually executed in funnel shape installed on the cantilever truss, and
designed for
changing the corium movement direction from the place of its escape from the
reactor
pressure vessel towards the reactor cavity axis for guaranteed input of corium
to the
maintenance platform. By burning the service platform, corium enters into the
corium
trap, where it enters into interaction with the filler by gradually heating
the corium
trap casing. In addition, during the melt-through of the reactor pressure
vessel, the
head of the reactor pressure vessel may completely separate following which
the head
of the reactor pressure vessel falls on the guide assembly, considerably
reducing or
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completely blocking the entry of corium into the corium trap. This may lead to
accumulation of corium in the guide assembly area, increase of corium
temperature,
burning through of the dry protection base and surrounding structural
concrete,
collapse of dry protection into corium, chemical interaction of serpentine
concrete of
.. dry protection with corium with generation of a large quantity of hydrogen,
other
non-condensing gases and aerosols. The generation of large quantity of
hydrogen,
other non-condensing gases and aerosols shall lead to considerable increase of
the
risks of hydrogen explosions and above design pressure buildup in the
containment
which consequently may lead to containment damage and escape of beyond design
.. quantity of radioactive fission products outside the containment.
Prior art
The guide assembly [1] (RF Patent No. 2253914, priority dated 18/08/2003) of
the nuclear reactor corium localizing and cooling system installed below the
reactor
pressure vessel head and resting on the cantilever truss executed in funnel
form
comprising of the cylindrical and conical parts, surfaces thereof are covered
with
heat-resistant concrete, apertures made in the center of the conical part is
well-
known.
One disadvantage of the guide assembly is the insufficient heat insulation of
the walls of conical and cylindrical parts. In the event of quick entry of
corium from
the reactor pressure vessel on separation of the head with full cross-section
considering the acceleration created by residual pressure inside the reactor
pressure
vessel, and considering the separated head in the process of movement,
blocking of
aperture executed in the center of the conical part is possible. This may lead
to
corium accumulation in the conical part area of the guide assembly, and thus
to
increase of temperature in this area. An increase of temperature may lead to
melt-
through of the walls of both the conical and cylindrical parts of the guide
assembly,
following which corium enters outside the corium trap, namely into the
structural and
serpentine concrete, which on failure generates large quantity of hydrogen and
non-
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condensing gases following which risks of hydrogen explosions and beyond
design
pressure rise occur in the containment. This may lead to containment failure
and
escape of beyond-design quantity of radioactive fission products outside the
containment limits.
Another problem of the guide assembly is the lack of reallocation (levelling)
mechanism of corium jet streams. This shall lead to the fact that the impact
thermal
and mechanical loads are concentrated in the upper and middle area of the
cylindrical
part. The concentration of impact thermal and mechanical stresses may lead to
failure
of the guide assembly and corium entry into the structural and serpentine
concrete
with their subsequent failure and generation of hydrogen and non-condensing
gases
following which risks of hydrogen explosions and beyond design-basis pressure
rise
in the containment occur. This may lead to containment failure and escape of
beyond-
design quantity of radioactive fission products outside the containment
limits.
The guide assembly [2] (Corium localizing device, 7th International Research
and practical conference "Safety assurance of NPP with VVER", OKB Gidropress,
Podolsk, Russia, May 17-20, 2011) of the corium localizing and cooling system
of
the nuclear reactor comprising of cylindrical and conical parts with aperture
in the
center thereof, bearing ribs passing from the central orifice to the boundary
of the
cylindrical part.
One disadvantage of the guide assembly is the insufficient heat insulation of
the walls of conical and cylindrical parts. In the event of quick entry of
corium from
the reactor pressure vessel on separation of the head with full cross-section
considering the acceleration created by residual pressure inside the reactor
pressure
vessel, and considering the separated head in the process of movement,
blocking of
aperture executed in the center of the conical part is possible. This may lead
to
corium accumulation in the conical part area of the guide assembly, and thus
to
increase of temperature in this area. An increase of temperature may lead to
melt-
through of the walls of both the conical and cylindrical parts of the guide
assembly,
following which risks of hydrogen explosions and beyond-design pressure rise
in the
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containment occur. This may lead to containment failure and escape of beyond-
design quantity of radioactive fission products outside the containment
limits.
Another problem of the guide assembly is the lack of reallocation (levelling)
mechanism of corium jet streams. This shall lead to the fact that the impact
thermal
and mechanical loads are concentrated in the upper and middle area of the
cylindrical
part. The concentration of impact thermal and mechanical stresses may lead to
damage of the guide assembly and entry of corium into the structural and
serpentine
concrete with their subsequent damage and generation of hydrogen and non-
condensing gases, following which risks of hydrogen explosions and beyond-
design
basis pressure rise occur in the containment. This may lead to containment
failure and
escape of beyond-design quantity of radioactive fission products outside the
containment limits.
The guide assembly [3, 4, 5] [RF Patent No. 2576516, priority dated
16.12.2014; RF Patent No. 2576517, priority dated 16.12.2014; RF Patent No.
2575878, priority dated 16.12.2014] of the corium localizing and cooling
system of
the nuclear reactor comprising of the cylindrical and conical parts with
aperture in the
center thereof, bearing ribs passing from the central aperture to the upper
edge of the
cylindrical part and dividing the cylindrical and conical parts into sectors,
covered
with layers of sacrificial and heat-resistant concrete is the closest analog
to the
claimed invention.
Such a guide assembly is designed for directing corium (melt) after damage or
melt-through of reactor to the corium trap, retention of large-sized debris of
the
reactor internals, fuel assemblies and reactor pressure vessel head against
fall of
corium into the trap, protection of the cantilever-truss and its
communications against
damage on corium input from the reactor pressure vessel to the core catcher,
guarding
the reactor pit against direct contact with corium.
The bearing ribs hold down the reactor pressure vessel casing with corium that
does not allow the head in the process of its damage or severe plastic
deformation to
overlap the cross-sections of the sectors and violate the corium trickling
process.
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Sacrificial concrete by dissolving in corium provides increase of the cross
section in the guide plate sectors when blockades are formed (on corium
setting in
one or several sectors) that allows not allow overheating and damage of the
bearing
ribs, i.e. complete blocking of the cross section, and as a consequence damage
of the
guide plate. Heat-resistant concrete provides structural strength on reduction
of the
sacrificial concrete thickness. This concrete protects the underlying
equipment
against corium impact not allowing the corium to melt-through or damage the
guide
plate.
One disadvantage of the guide assembly is the inability of the two-layer
sacrificial concrete to provide increase of the cross-section in the guide
plate sectors
on simultaneous input of a large volume of metal and oxide melt, for example
on
separation of reactor pressure vessel head by full cross section or on its
sectoral
damage. In this case the simultaneous interaction of two types of overheated
corium
(metallic and oxide) with sacrificial concrete (based on aluminium and ferrous
oxides) shall lead to quick release of oxygen, rapid oxidation, aerosol and
slag
formation with complete overlap of the flow cross section. Taking into
consideration
that the hot gas-vapor and aerosol interaction products of sacrificial
concrete with
metal or oxide corium components rise up, and their movement is directed
against the
corium flow, in a squeezed space between the reactor pressure vessel head and
heat-
resistant concrete (based on aluminium oxide) hydrodynamic blockage of
cellular
sacrificial concrete is formed preventing corium movement. On formation of
stagnant
zone the heat resistant concrete quickly gets overheated and enters into
chemical
reaction with the corium components increasing the gas-aerosol counter-flow.
One more disadvantage of the guide assembly is the insufficient thermal
insulation of the walls of conical and cylindrical part. In the event of quick
entry of
corium from the reactor pressure vessel on separation of the head with full
cross-
section considering the acceleration created by residual pressure inside the
reactor
pressure vessel, and considering the separated head in the process of
movement,
blocking of aperture executed in the center of the conical part is possible.
This may
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lead to corium accumulation in the conical part area of the guide assembly,
and thus
to increase of temperature in this area. An increase of temperature may lead
to melt-
through of the walls of both the conical and cylindrical parts of the guide
assembly,
following which corium enters outside the corium trap, namely into the
structural and
serpentine concrete, which on failure generates large quantity of hydrogen and
non-
condensing gases following which risks of hydrogen explosions and beyond
design
pressure rise occur in the containment. This may lead to containment failure
and
escape of beyond-design quantity of radioactive fission products outside the
containment limits.
Another problem of the guide assembly is the lack of reallocation (levelling)
mechanism of corium jet streams. This shall lead to the fact that the impact
thermal
and mechanical loads are concentrated in the upper and middle area of the
cylindrical
part. The concentration of impact thermal and mechanical stresses may lead to
damage of the guide assembly and entry of corium into the structural and
serpentine
concrete with their subsequent damage and generation of hydrogen and non-
condensing gases, following which risks of hydrogen explosions and beyond-
design
basis pressure rise occur in the containment. This may lead to containment
failure and
escape of beyond-design quantity of radioactive fission products outside the
containment limits.
Disclosure of the invention
The technical result of the claimed invention consists in enhancing the safety
of
nuclear power plant by enhancing the reliability of the corium localizing and
cooling
system of the nuclear reactor.
The tasks for resolving thereof the invention is directed consist in providing
the
following operation conditions of the corium localizing and cooling system of
the
nuclear reactor:
- excluding the blocking of the aperture made in the center of the conical
part;
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- excluding the entry of nuclear reactor corium into the structural and
serpentine concretes of the reactor cavity with the subsequent generation of
hydrogen
and non-condensing gases.
The set tasks are resolved by the fact that the guide assembly (1) of the
corium
localizing and cooling system of a nuclear reactor, installed under the
reactor pressure
vessel and resting on the cantilever truss, containing the cylindrical part
(2) and
conical part (3) with aperture (4) executed in it, bearing ribs (5), located
radially
relative to the aperture (4) and separating walls of the cylindrical (2) and
conical (3)
parts for the sectors (7), characterized in that it additionally contains the
load-bearing
frame, consisting of the external upper thrust ring (8), external lower thrust
ring (9),
internal central thrust ring (10), external upper shell (11), middle thrust
ring (12),
separated into sectors by the bearing ribs (5) and having aperture (14) in the
upper
part, external lower thrust shell (15), foundation (16), bearing stiffeners
(17), upper
tilted plate (18), connecting the conical head (19), bearing ribs (5) and
middle thrust
ring (12), lower tilted plate (20), connecting conical head (19), bearing ribs
(5),
middle thrust ring (12) and external upper thrust ring (11), thermal plate
metal shields
(23), installed on bearing stiffeners (17) and installed with gap (22) along
the internal
surface of middle thrust shell (12), and along the upper tilted plate (18),
dismountable
thermal plate metal shield (13), installed on bearing stiffeners (17) and
covering the
aperture (4), cooling canal (21), outgoing from the header (6) and passing
between
the upper and lower tilted plates (18 and 20), and between the middle and
external
upper thrust rings (12 and 11), connected through the aperture (14) with gap
(22)
forming a space between the thermal plate metal shield (23) and upper tilted
plate
(18), in addition, the space (24) limited by the pedestal (16), conical head
(19), lower
tilted plate (20), part of the upper thrust ring (11), external lower thrust
ring
(9),external lower thrust shell (11) and middle thrust shell (12), and the
space (26)
between the upper and lower tilted plates (18 and 20) is filled with concrete
or
ceramic material (27), leak-tight head (28), connected with the external lower
thrust
shell (15) and bearing ribs (17).
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One feature of the claimed invention is the availability in the Guide assembly
(1) of the corium localizing and cooling system of the load bearing frame,
consisting
of the external upper thrust ring (8), external lower thrust ring (9),
internal central
thrust ring (10), external upper shell (11), middle thrust ring (12),
separated into
sectors by bearing ribs (5) and having aperture (14) in the upper part,
external lower
thrust shell (15), foundation (16), bearing stiffeners (17), upper tilted
plate (18),
connecting the conical head (19), bearing ribs (5) and middle thrust ring
(12), lower
tilted plate (20), connecting conical head (19), bearing ribs (5), middle
thrust ring
(12) and external upper thrust ring (11). In accordance with the claimed
invention, the
availability of load-bearing frame allows provide retention of large-sized
debris of
internals and reactor pressure vessel head against fall into the corium trap
thus the
shell of the corium trap is protected against damages.
One more feature of the claimed invention is the availability in the guide
assembly (1) of the thermal plate metal shield (23), installed on bearing
stiffeners
(17) and installed with gap (22) along the internal surface of the middle
thrust shell
(12), and along the upper tilted plates (18), of the dismountable thermal
plate metal
shield (13), installed on the bearing stiffeners (17) and covering the
aperture (4). The
presence of thermal plate metal shields (23) allows provide gravity trickling
to
corium filler after damage or melt-through of reactor pressure vessel,
protection of
the cantilever-truss and its communications against damage during melt
movement,
excluding direct contact of corium with the reactor cavity equipment and
structural
concrete, exclusion of direct radiant action on the part of corium on the
reactor cavity
equipment and fittings of the reactor pressure vessel due to exclusion of the
formation
of blockades related to blocking of the cross-section by corium, by quick
increase of
the effective cross-section provided by levelling and melting of the thin
elements of
thermal plate metal shield (23).
One more distinctive feature of the claimed invention is the availability of
the
cooling channel (21) coming out of the header (6) and passing between the
upper and
lower tilted plates (18 and 20), and between the middle and external upper
bearing
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shells (12 and 11), connected through the aperture (14) with gap (22), forming
a
space between the thermal plate metal shield (23) and middle load bearing
shell (12),
and between the thermal plate metal shield 923) and upper tilted plate (18) in
the
guide assembly (1) of the corium localizing and cooling system. The
availability of
cooling channel (21) provides thermal stabilization of the entire guide
assembly (1)
during reactor power operation in normal operation conditions.
One more distinctive feature of the claimed invention is that in the guide
assembly (1) of the corium localizing and cooling system of the nuclear
reactor, the
space (24) limited by the base (16), conical head (19), lower tilted plate
(20), part of
the upper external bearing shell (11), external lower bearing ring (9),
external lower
bearing shell (15), and the space (25) between the external upper bearing
shell (11)
and middle bearing shell (12), as well as the space (26) between the upper and
lower
tilted plates (18 and 20) is filled with concrete or ceramic material (27).
The use of
concrete and ceramic material (27) in the specified spaces allows provide
thermo-
mechanical protection of the load-bearing elements of the guide assembly (1)
from
damage than retention is provided of the reactor pressure vessel and its large
fragments on reactor pressure vessel damage, large-sized fragments of the
internals
against fall into the corium is provided, protection of corium casing against
damages
on fall of large fragments is provided, protection of corium trap casing
against
damages on fall of large fragment is provided, protection of the cantilever-
truss and
its communications against damage during corium movement is provided,
exclusion
of direct contact of corium with reactor cavity equipment and structural
concrete is
provided.
One more distinctive feature of the claimed invention is the availability of
leak-
tight head (28), connected with the external lower bearing shell (15) and
knife edges
(17) in the guide assembly (1) of the corium localizing and cooling system of
the
nuclear reactor. The availability of a leak-tight head (28) allows provide
water
drainage from the head (28) surface, consequently absence of steam explosions
at the
time of corium input to the filler, and retention of the integrity of filler
and structural
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materials in the process of the entire period of normal operation and during
operational occurrences and during design-basis accident.
In aggregate, such a design of the guide assembly allows:
- provide gradual intake of corium (melt) after damage or melt-through of
the
reactor into the corium trap;
- provide protection of concrete cavity and dry protection with serpentine
concrete against direct contact with corium.
Brief description of drawings
The guide assembly of the corium localizing and cooling system of the nuclear
reactor executed in accordance with the claimed invention is shown in Fig. 1.
The sectional view of the guide assembly of the corium localizing and cooling
system of the nuclear reactor, executed in accordance with the claimed
invention is
shown in Fig. 2.
The guide assembly fragment of the corium localizing and cooling system of
the nuclear reactor, executed in accordance with the claimed invention is
shown in
Fig. 3.
Embodiments of the invention
As shown in Fig. 1, the guide assembly (1) of the corium localizing and
cooling system of the nuclear reactor installed below the reactor pressure
vessel and
resting upon the cantilever truss contains the cylindrical part (2) and
conical part (3).
An aperture (4) is made in the base of the conical part (3). Bearing ribs (5)
pass along
the conical and cylindrical parts (2 and 3) located radially with respect to
the aperture
(4). The bearing ribs (5) divide the walls of the cylindrical (2) and conical
(3) parts
into sectors (7). The guide assembly (1) contains the load-bearing frame,
which
consists of the following basic (load bearing) elements: external upper load-
bearing
ring (8), external lower load-bearing ring (9), internal central load-bearing
ring (10),
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external upper load-bearing shell (11), middle load-bearing shell (12). Middle
load-
bearing shell (11) is divided into sectors by the bearing ribs (5) similar to
the wall of
the cylindrical part (2). The bearing frame is also composed of the external
lower
load bearing shell (15), base (16), knife edges (17), upper tilted plate (18).
The upper
tilted plate (18) connects the conical head (19), bearing ribs (5) and middle
load-
bearing shell (12). The lower tilted plate (20) connects the conical head
(19), bearing
ribs (5), middle load-bearing shell (12) and external upper load-bearing shell
(11).
Apart from the load-bearing elements, the following thermal elements are used
as part of the guide assembly (1): thennal plate metal shields (23),
dismountable
thermal plate metal shield (13). The thermal plate metal shields (23) are
installed on
the knife edges (17), and with gap (22) along the internal surface of the
middle load
bearing shell (12) and along the upper tilted plate (18). The dismountable
thermal
plate metal shield (13) is installed on the knife edges (17) and closes the
aperture (4).
The cooling channel (21) passes between the upper and lower tilted plates (18
and 20) and between the middle and external upper load-bearing shells (12 and
11).
The cooling channel (21) exits from the header (6) and connects through the
aperture
(14) with gap (22) forming the space between the thermal plate metal shield
(23) and
middle load bearing shell (12), as well as between the thermal plate metal
shield (23)
and upper tilted plate (18).
The space (24) limited by the base (16), conical head (19), lower inclined
plate
(20), part of the upper external load-bearing shell (11), external lower
bearing ring
(9), external lower load-bearing shell (15) and space (25) between the
external upper
load-bearing shell (11) and middle load-bearing shell (12), as well as the
space (26)
between the upper and lower tilted plates (18 and 20) is filled with concrete
or
ceramic material (27).
A leak-tight head (28) is welded below to the external lower load-bearing
shell
(15) and knife edges (17).
The claimed guide assembly functions as follows.
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As shown in Fig. 1 - 3, the guide assembly (1) installed on the cantilever
truss
below the reactor pressure vessel head, in accordance with the gist of the
claimed
invention, performs the thermal barrier functions between the reactor pressure
vessel
and the reactor cavity equipment in its lower part, and between the reactor
pressure
vessel head and corium trap located below the guide assembly (1). The
availability of
thermal barrier during normal operation allows provide thermal insulation of
the
reactor pressure vessel head, and during severe accident at the time of
reactor
pressure vessel damage by corium provide conditions for diagnosis of the start
of
melt entry into the trap.
Heat insulation consisting of plate metal thermal shields (23), executed in
the
form of packets assembled from dimple and non-dimple thin stainless steel
sheets is
installed on the guide plate for providing thermal insulation of the reactor
pressure
vessel head during normal operation. Such packets are installed on the walls
(6) of
the cylindrical and conical parts (2 and 3), and on the inner surface of the
middle
load-bearing shell (12) and upper tilted plate (18) using the fasteners
providing
thermal displacements of heat insulating packets and guide plate frame with
respect
to each other during normal operation, operational occurrence and design-basis
accident.
The dismountable thermal plate metal shield (13) is installed directly below
the
reactor pressure vessel head pole that provides if required access to the
external
surface of the reactor pressure vessel. An hatch with displacing inset is
executed for
access to the dismountable thermal plate metal shied (13) in the lower part of
the
guide assembly (1) on the service platform side. Such a design allows exclude
water
accumulation in the hatch during operational occurrence, during design-basis
and
beyond design-basis accidents.
The space between the load-bearing elements (5, 8, 11,9, 15, 16, 19, 18, 12)
of
the guide assembly is filled with heat-resistant concrete for providing
thermal
insulation of structural concrete and cantilever truss during beyond design-
basis
accident. The load-bearing elements (5, 8, 11, 9, 15, 10) and concrete and
ceramic
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material (27) form according to their function the guide assembly in the form
of a
funnel, providing coverage of the lower part of the reactor pressure vessel
above the
connection plane of the head with the cylindrical part (2). In the process of
corium
exit, the guide assembly (1) can be subjected both to a relatively slow
loading under
plastic deformations of the reactor pressure vessel, and to impact loading
when the
head of the reactor pressure vessel is torn off due to the residual pressure.
These loads
are taken up by the guide assembly, formed by the load-bearing elements (5, 8,
11, 9,
15, 10) and concrete and ceramic material (27). Such a design shall provide:
- free-flow drainage to the corium filler after damage or melt-through of
reactor pressure vessel;
- retention of large-sized debris of internals and head of the reactor
pressure
vessel against fall into the corium trap;
- protection of corium trap casing against damages on fall of large
fragments;
- protection of the cantilever truss and its communications against damage
on corium movement.
- excluding direct contact of corium with the reactor cavity equipment and
construction concrete;
- exclusion of direct radiant impact on the part of the corium on reactor
cavity equipment and reactor pressure vessel fittings.
The layers of sacrificial material (concrete or ceramic) are located under the
tilted surfaces of the guide assembly ¨ under the upper and lower tilted
plates (18 and
20). directly below the upper tilted plate (18) is the sacrificial layer
prepared for
example based on aluminium and ferrous oxides, and under the lower tilted
plate (20)
is the thermally durable heat-proof layer, made for example based on aluminium
oxide.
The sacrificial material located under the upper tilted plate (18), by melting
in
the corium ensures increase of the cross-section in the guide assembly (1)
sectors, if
the increase of the effective cross-section provided by the flattening and
melting of
the thin elements of the plate metal shied (23) was not sufficient for example
on
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outflow of the melt from the reactor pressure vessel with large flow exceeding
the
cross-section throughput capacity of the guide assembly (1) or on outflow of
corium
with core debris covering the cross-section and preventing free outflow of
corium.
The dissolving of the sacrificial material allows to not allow overheat and
damage of
bearing ribs (5). The complete blocking of the cross-section is possible on
damage of
the bearing ribs (5), and sectoral damage of the guide assembly (1) as a
consequence
of this.
The thermal-resistant heat-proof layer located below the lower tilted plate
(20)
provides strength and stability of the structure on reduction of the thickness
of
sacrificial material located between the upper and lower tilted plates (18 and
20).
Thermal resistant concrete protects the underlying equipment against the
corium
impact, by not allowing the corium to sectoral through melt-through or damage
the
guide assembly.
The guide assembly (1) on reactor pressure vessel damage takes on itself the
dynamic loads occurring:
- on lateral outflow of corium under the action of residual pressure in the
reactor pressure vessel;
- on increase of the lateral cavity cross-section in the reactor pressure
vessel
and change of its profile in the process of corium outflow;
- on detachment of reactor pressure vessel head parts following plastic
deformation under the action of thermal and mechanical loads and residual
pressure;
- on detachment of the reactor pressure vessel head parts following impulse
pressure rise inside the pressure vessel (on flooding the corium with water)
and their
shock diffusion about the guide vanes;
- during external effects and auto shocks in the process of beyond design-
basis accident propagation.
Before the start of corium inflow the filler present in the trap casing is
tightly
closed by the head (28) of the guide assembly (1) that provides:
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- water drainage from the head surface (28) and as a consequence of this no
steam explosions at the time of corium inflow to the accumulator.
- retention of integrity of the accumulator and structural materials in the
process of the total period of normal operation, and on abnormal operation and
during
design-basis accident.
The following is performed for providing unconstrained inflow of corium:
- leak-tight head (28) is executed in the form of an easily damageable
membrane;
- thermal plate metal shields (13 and 23) are executed with easily
damageable high-temperature corium so as to not prevent its displacement. On
melting of the thermal insulation the cross-section for the tricking of corium
along the
surface of the guide assembly increases several times. Various degree of
increase of
the cross-section is provided for the vertical and tilted thermal plate metal
shields
(23) that is related to different geometry of the channels fonned by the
vertical
bearing fins;
- An aperture (4) is made in the central part of the guide assembly for
corium
passage, sizes thereof is limited by the scatter of solid and liquid fragments
of the
core in the process of its outflow from the reactor pressure vessel.
Thus, the thermal plate metal shields (23) and sacrificial material installed
below the upper and lower tilted plates (18 and 20), used as guide assembly
(1) of the
core localizing and cooling system of the nuclear reactor perform anti-impact,
channel forming and protection functions.
The plate metallic heat shields (23) provide initial damping of the impact
load
on the part of the separated sectors of the damaged head considering the
acceleration
created by residual pressure inside the reactor pressure vessel. Moreover, the
crushed
plate metal heat shields (23) provide the initial protection of the guide
assembly (1)
and against impact action of the corium jet at small residual pressure in the
reactor
pressure vessel.
Date Recue/Date Received 2021-12-30
CA 03145726 2021-12-30
On strong dynamic response on the part of the separated sectors of the
damaged head of the reactor pressure vessel, the concrete or ceramic materials
(27)
forming the protective layer around the critically important load-bearing
elements (5,
11, 15, 9) of the guide assembly (10) takes the impact load, moreover the
bearing fins
(5) may be partially melted, especially this concerns the tilted part
protected by layers
of sacrificial material under the upper and lower tilted plates (18 and 20).
Together with the load-bearing elements (5, 8, 11, 9, 15, 18, 20, 12) of the
guide assembly (1) the concrete or ceramic material (27) creates impenetrable
barriers for the flying objects and corium jet.
Thus, the thermal plate metal shield (23) and concrete or ceramic material
(27)
forming the protective layers of load-bearing elements (5, 9, 11, 12, 15) of
the guide
assembly (1) ensure breaking and blocking of large fragments of reactor
pressure
vessel and its internals, at the same time providing sequential input of
corium,
fragments of internals and head of the nuclear reactor pressure vessel in the
corium
trap.
The removable thermal plate metal shields (23) provide increase of the cross-
section for displacement of corium in each radial vertical and tilted sectors
and in
azimuth direction on horizontal flow of corium.
During severe thermo-mechanical impact on the part of corium outflowing
from the reactor pressure vessel the cross-section increases in the guide
assembly (1)
for displacement of corium by thermo-chemical interaction of the concrete or
ceramic
material (27) with corium, besides the chemical activity and thermo-mechanical
impact on the load-bearing frame of the guide assembly (1) are reduced that
retention
of its integrity.
Thus, the thermal plate metal shields (23) and concrete or ceramic material
(27) forming the protective layers of the load-bearing elements (5, 9, 11, 12,
15) of
the guide assembly (1) provides protection of the structural and serpentine
concretes
of the reactor cavity against interaction with corium.
16
Date Recue/Date Received 2021-12-30
CA 03145726 2021-12-30
The concrete or ceramic material (27) foiming the protective layers around the
critically important load-bearing elements (5,11, 15, 9) of the guide assembly
(1)
create thermal and chemical barriers preventing the surface and structural
damage of
the load-bearing elements (5, 8, 11, 9, 15, 18, 20, 12) of the guide assembly
(1) on
thermal and thermo-mechanical impacts on the part of corium jet, for which
purpose
the thermal resistance of concrete or ceramic material (27) is selected
different in
different directions of corium flow that provides earlier damage of the
sacrificial
material under the upper tilted plate (18) located close to the reactor
pressure vessel
than is achieved by quicker escape of corium and reduction of thermo-chemical
and
thermo-mechanical impacts on the critically important load-bearing elements
(5, 6, 9,
7, 11, 14, 10) of the guide assembly (1).
Thus, the concrete or ceramic material (27) forming the protective layers of
the
load-bearing elements (5, 6, 9, 7, 11, 14, 10) of the guide assembly (1)
provides their
strength on lateral melt-through of the reactor pressure vessel and as a
consequence
protection of the structural and serpentine concretes of the reactor cavity
against
interaction with corium.
The use of guide assembly (1) having load-bearing frame equipped additionally
with thermal elements allowed provide gradual input of corium (melt) after
damage
or melt-through of the reactor pressure vessel, retention of large-sized
fragments of
the internals, fuel assemblies and head of the reactor pressure vessel against
fall into
the corium trap, protection of the cantilever truss and its communications
against
damage on corium input from the reactor pressure vessel to the corium trap,
without
blocking of the central aperture made in the conical part, preservation of the
concrete
cavity and dry protection with serpentine concrete against direct contact with
corium.
Sources of information:
1. RF patent No. 2253914, IPC G21C 9/016, priority dated 18.08.2003
2. Corium localizing device, 7th International Research and Training
Conference "Safety assurance of NPP with VVER", OKB Gidropress, Podolsk,
Russia, May 17-20, 2011.
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3. RF patent No. 2576516, IPC G21C 9/016, priority dated 16.12.2014;
4. RF patent No. 2576517, IPC G21C 9/016, priority dated16.12.2014;
5. RF patent No. 2575878, IPC G21C 9/016, priority dated 16.12.2014.
18
Date Recue/Date Received 2021-12-30
