Reactor core
The main design data is shown in table 08 and table 09;
Table 8: Main design data for the core
Table 9: AGR Core Design Parameters
The reactor moderator is a sixteen-sided stack of graphite
bricks. The bricks are interconnected with graphite keys to give the moderator
stability and to maintain the vertical channels on their correct pitch, despite
dimensional changes due to irradiation, pressure loads and thermal stresses
(Figure 08).
Figure 8: Interconnection
of graphite bricks with keys
The complete reactor core consists of an inner cylinder of
graphite moderator containing 332 fuel channels. It is surrounded by a graphite
reflector and a steel shield. The graphite structure is maintained in position
by a steel restraint tank which surrounds the graphite and which is supported
on a system of steel plates (Figure 5.3).
The primary system for Control and shutdown of the reactor
consists of 89 absorber rods and drives housed in standpipes in the top part of
the reactor vessel. The Control rods are located in interstitial positions,
i.e. in off-lattice positions (Figure 09).
As a back up
against the extremely remote possibility of a fault in the primary system which
will prevent a substantial number of Control rods from entering the core when required,
a secondary shutdown and hold-down system is provided. The secondary shutdown
system comprises 163 interstitial core channels into which nitrogen can be
injected from beneath the core (Nitrogen absorbs neutrons to a much greater
extent than carbon dioxide). Gradually the nitrogen flows from the interstitial
channels into the reentrant passages and through the fuel channels. Thus, a
nitrogen concentration is building up in the coolant gas circuit until it is
sufficient to hold the reactor in a shutdown condition.
The store of nitrogen, which consists of banks of high
pressure cylinders, is common to both reactor units. It holds sufficient
nitrogen gas for the shutdown and subsequent hold down one reactor, provided
the reactor remains pressurized. Thus a fast shutdown is achieved.
Furthermore, a boron bead injection system is also provided
in 32 of the 163 interstitial channels designed to give long-term-hold-down
capabilities in the extremely unlikely situation where an insufficient number
of Control rods have been inserted into the core and depressurization of the
core is required; in this case the pressure of the nitrogen injection system
cannot be maintained.
Figure 9: One Quarter Cross Section of reactor core
Fuel Assemblies
The main design data for fuel assemblies are shown in Table
10. The fuel in AGR’s consists of slightly enriched U02 in the form
of cylindrical pellets with a central hole. These are contained within
stainless-steel cladding tubes, each of which is about 900 mm long. A fuel element consists of 36 fuel
pins surrounded by two concentric graphite sleeves. The fuel pins are supported
by top and bottom grids which are fixed to the outer graphite sleeve (Figure 10
and Figure 11). The possible bowing of the pins is limited by support braces.
The support grids, braces, and inner sleeves are secured in position by a
screwed graphite retaining ring at the top. The complete unit forms a fuel element
(Figure 10). Eight of these fuel elements are linked together with a tie bar to
form a fuel stringer assembly. Each of the 332 fuel channels is provided with a
fuel stringer assembly.
Table 10: Main design data for fuel elements
The function of the double sleeve in the fuel element is to
provide a static gas gap between inner and outer sleeve to reduce leakage of
heat from the hot coolant to the graphite moderator.
The fuel stringer assembly and its associated plug unit form
a composite fuel assembly that is handled by the fuelling machine and loaded
into the reactor as one unit. Since it is important that the cladding exhibit
good heat transfer properties, the cladding is provided with small transverse
ribs on the outer surface (Figure 11), and is compressed onto
the pellets during manufacture to minimize the clearance gab between pellet and
cladding.
Figure 10: Dimensions
of an AGR fuel element
Figure 11: AGR fuel Element
The remaining space is filled with helium, an inert gas with
good heat-conducting properties.
The reactor has to single channel access, i.e. each channel
is extended upwards with its own separate opening in the concrete vessel. This
permits the gas temperature in each channel to be measured and to adjust the
flow of gas coolant remotely. Finally it permits refueling both when the
reactor is on and off load and for any pressure from atmospheric to normal
operating pressure, by a fuelling machine that handles complete
fuel assemblies from the top of the vessel.
Figure 12: Detailed
view of AGR fuel element
The fuel assembly plug consists of a closure unit, a
biological shield plug, a valve (gag) unit and actuator and a neutron scatter
plug. During normal operation, the fuel assembly plugs unit acts as the seal of
the reactor pressure vessel at each fuel standpipe. They are provided with
closure/locking mechanisms, which are operated remotely.
The biological shield plug is designed primarily to limit
neutron and gamma-radiation through the standpipe. It consists of two
mild-steel blocks joined together by a mild-steel tube. A steel ring
loosely mounted on the lower block reduces radiation streaming through the
annular gab between the plug and the standpipe liner. The valve unit is
situated in the lower part of the fuel assembly plug. It contains a duct for the
hot coolant gas between the fuel stringer assembly and the outlet ports above
the gas baffle dome. It includes a flow Control valve (gag) for adjustment of
coolant flow through the individual channels. The valve is coupled by a shaft
passing through the biological shield plug to a motor-driven valve actuator,
the basic function of which is to set the valve position. The actuator is
operated by remote Control from the central Control room.
Below the gag unit a neutron scatter plug is located to
prevent neutrons streaming up the channel. The tie bar, attached to the top of
the gag unit carry both that unit and the fuel stringer assembly during refueling
operations.
Refueling and Fuel handling system
Access to the reactor for refueling is provided by
standpipes located in the top cap of the reactor pressure vessel, one standpipe
for each reactor channel. One refueling machine, designed to handle both fuel
and Control assemblies, serves both reactors. The machine runs on a travelling
gantry that spans the width of the charge hall and is supported on rails which run
along the full length of the hall. This gives the machine access to both
reactors and to the central service block. In this block there are facilities
for storage of new fuel elements and fuel stringer components and for their
assembly into complete fuel assemblies. It also includes facilities for
temporary storage and dismantling of irradiated fuel assemblies. The fuel storage
pound, used for the longer term storage of irradiated fuel elements, is also
located in the central block.
The refueling machine
essentially a hoist contained within a shielded pressure vessel is provided with a telescopic snout which can
be extended to connect, seal and lock on to a short extension tube fitted to the
standpipe being serviced. Within the refueling machine pressure vessel a three compartment
turret can be rotated to align any of the compartments with the machine snout.
One turret tube is for withdrawing of used fuel, one carry the new fuel and the
last carry a spare plug unit. The top section of the pressure vessel above the
turret contains the hoist drive shaft, which passes through seals in the
vessel. At the end of the shaft is located the machine grab, suspended in
roller chains. The grab is operated electrically by solenoids, and it can be lowered
through the turret tube aligned with the machine snout to pick up fuel assemblies.
The movements of the machine and the connections to the
standpipes are controlled from a platform at the bottom of the machine. All
other operations are controlled from a platform on the machine located just
above the gantry. The refueling program requires about 5 fuel assemblies to be
replaced per month.
Gas Controlling System
Carbon dioxide gas is used to transfer the heat produced in
the reactor to the boilers. The gas is pumped through the channels of the
reactor at high pressure by gas circulators, its main flow paths are shown in
Figure 13.
Figure 13: Gas flow distribution in core and vessel
The gas circulator pumps the cooled gas from the bottom of
the boilers and into the space below the core. About half of this gas flows
directly to the fuel channel inlets, while the remainder, known as the
re-entrant flow, passes up through the annulus surrounding the core along the
inner surface of the gas baffle to the top baffle. It returns downwards through
passages between the graphite moderator and the graphite sleeves of the fuel
elements to rejoin the main coolant flow at the bottom of the fuel channels. (Probable
some kind of orifice is used at the bottom of the fuel channels to split the
gas flow into the re-entrant flow and the fuel channel flow).
The re-entrant flow thus cools the graphite bricks, the core
restraint system and the gas baffle. The combined flow passes up the fuel
channels and through the guide tubes. Then the hot gas flows into the space
above the gas baffle and down through the boilers, where it is cooled, before
re-entering the gas circulators below the boilers.
The main reason for the re-entrant flow from the top of the
core to the bottom is to keep the moderator temperature below 450oC
to avoid excessive thermal oxidation of the graphite bricks, and to limit
temperature gradients within a brick to about 50oC. This
is most economically achieved by a re-entrant coolant flow, in which part of
the coolant will pass downwards between the bricks before entering the bottom
of the fuel channels. However, it does complicate the internal layout of the
plant within the pressure vessel vault by necessitating a gas baffle around the
core. Fuel element guide tubes, located at the top of each channel, are used to
duct the hot gas through the space below the gas baffle before it is discharged
into the space above.
The core and the surrounding graphite reflector and shield
are completely enclosed in the gas baffle which has a diameter of 13.7 and
which is provided with a tri-spherical head. The baffle has to withstand the full
core pressure differential of 1.9 kg/cm2 and its temperature is kept down to 325"
C by insulation on the topside so that mild steel can be used. Table 11 shows
the heat balance scheme of a typical AGR plant.
Table 11: Heat
balance for an AGR plant
Controlling Systems of an AGR
Primary Reactivity Control System
The primary system for Control and shutdown of the reactor
consists of 89 absorber rods and drives housed in standpipes in the top
cap of the reactor vessel. 44 of these are black rods (Figure 14) of which 7
act as sensor rods for detecting any guide tube misalignment that may occur
between the graphite moderator and the steel structures above it. The remaining
45 absorber rods are grey regulating rods, of which 16 are used as a
safety group. This safety group can be moved out of the core when the reactor
is shut down so that it can be moved into the core in case of an inadvertent
criticality. Each black rod consists of eight cylindrical sections linked
together by joints.
Figure 14: AGR Control Rod
Each of the lower six sections consists of a 9 % Cr, 1 % Mo
steel sheath containing four tubular inserts of stainless steel with a 4.4 %
boron content to ensure blackness to thermal neutrons. Between the tubular
inserts are two solid, cylindrical, graphite inserts to reduce neutron
streaming.
The upper two sections, which form part of the top reflector
when fully inserted, contain full-length, solid graphite inserts only. The 45
grey regulating rods are of a design similar to the black rods. However, the
lower six sections contain tubes of stainless steel without boron, but with
graphite inserts arranged as in the black rods.
The Control assembly consists of a Control rod, a Control
plug unit, a Control rod actuator and standpipe closure unit and its housing.
The complete Control assembly is designed for removal by the refueling machine
both when the reactor is operating and when it is shut down.
The Control plug unit is designed to reduce to acceptable
levels radiation streaming from the core through the standpipe penetrations in
the vessel roof. It consists of steel plug with a central hole through which
passes the Control rod suspension chain.
The Control rod actuators raise or lower the Control rods.
Each actuator is provided with motor operated winding gear and suspension chain
storage, electromagnetic clutch, hand winding drive to the clutch, rod position
indicator and limit switches. The actuator and rod drive is designed for
frequent smal1 movements. 'The Control rod speed is controlled by regulating
the electricity supply to the induction motor. In the event of a reactor trip, the
clutch is de-energized to allow rod insertion by gravity. The insertion rate is
controlled by a carbon disc brake.
Secondary Shutdown system
As a
backup against the extremely remote possibility of a fault in the primary
system preventing a substantial number of Control rods from entering the core
when required, two secondary shutdown and hold-down system are provided.
Fast shutdown is achieved by a system that automatically
injects nitrogen from beneath the core into 163 interstitial core channels.
Nitrogen absorbs neutrons to a much larger extent than carbon dioxide. The
nitrogen storage arrangements are designed to provide nitrogen injection in two
stages. When the trip valves are opened, a high initial nitrogen flow rate is
provided by the first stage storage. This initial flow purges the 163
interstitial channels of carbon dioxide and fills them with nitrogen. Flow from
the second stage provide make-up to each channel as the nitrogen flows from the
channels into the reentrant passages and through the fuel channels, thus
gradually building up the nitrogen concentration in the coolant gas circuit
until it is sufficient to hold the shutdown core in a sub-critical condition
for several hours.
A boron
bead injection system is also provided designed to give long-term hold-down in the
extremely unlikely situation where an insufficient number of rods have been
inserted into the core and when the reactor is depressurized, whereby the
nitrogen pressure is reduced.
Boron glass beads with a diameter of 3 mm are injected into 32 of the 163 secondary shutdown
channels. Each of these channels has an associated bead delivery pipe, with one
end terminated at the top of the channel and other end connected to one of the
bead storage hoppers. CO2 gas is used to inject the beads
pneumatically from the bottom of the hopper to the top of the channel. The
beads run downwards into the channel from the open end of the delivery pipe
until the channel is filled.
Bead delivery is initiated by the manual operation of valves
situated adjacent to the hoppers. Key interlocks Control the valve operation
sequence, while additional locks prevent unauthorized release of the beads.
This system holds the reactor in a shutdown condition indefinitely.
Main Coolant
System for AGR
Carbon dioxide gas is used to transfer heat from the reactor
to the boilers. The gas is pumped through the channels of the reactor by gas
circulators at a pressure of about 40 bars.
Each reactor has 8 gas circulators driven by
induction motors, Figure 15. Each circulator, complete with motor and Control
gear, is a totally closed unit located in a horizontal penetration at the
bottom of the reactor pressure vessel.
In addition to its normal duties, the circulator unit, its
mounting system and shaft labyrinth act as a secondary containment system,
should the penetration closure fail. The mounting system is pre-tensioned to
provide nominally constant loading of the motor stator frame under all
operating and fault conditions leading to depressurization. The motor is
provided with a variable frequency power supply to enable operation at lower
speeds especially at reactor trips. If a reactor trip occurs, the blower speed
drops to 450 rev/min, but increases automatically to 3000 rev/min in the case
of accidental depressurization of the reactor.
The normal regulation of the flow is via variable inlet
guide vanes, which also Control the reverse flow when the pump motor has
stopped.
Table 12: Design Data for gas circulator
The reason for the use of a totally enclosed gas circulator
design is partly to make swift removal and replacement of circulator units with
a minimum loss of reactor output possible and partly to avoid the high
pressure, rotating, oil-fed gas seal which has been used so far on circulators
for Magnox reactors.
Figure 15: AGR Gas Circulator
CO2 Supply
A carbon
dioxide supply system is located on site. Its purpose is to provide storage capacity
for liquid carbon dioxide and supplies of gaseous carbon dioxide for each
reactor, the refueling machine and auxiliary plant facilities during normal operation
and fault conditions. The composition of the gas coolant is maintained within
defined operational limits by the reactor coolant processing system. A fraction of the reactor coolant
flow is passed continuously through the processing plant and after treatment is
returned to the main coolant circuit at a circulator inlet, the circulator
providing the driving force.
The plant is also provided with a reactor coolant discharge
system for the controlled discharge of contaminated gas from the reactor and
associated equipment. It comprises
·
The reactor vessel blow down and purge system -
one per reactor
·
The auxiliary blow down system - one per station
·
The reactor vessel safety relief-valve system -
two per reactor
Radioactive Waste Associate with AGR
During power operation the major part of radioactive wastes
is produced in the irradiated fuel and contained in the fuel matrix. However,
certain other wastes produced at nuclear power stations in gaseous, liquid or
solid form may be radioactive to some degree. The radioactivity is due to
neutron irradiation of materials in the reactor.
The management of medium and low-level wastes is an
important part of nuclear power station design and operation. The general
approach is in the case of low-level liquid and gaseous wastes to filter dilute
and disperse to the environment. Solid wastes, such as redundant plant items,
filter dusts and sludge, ion-exchange resins, and discharged protective
clothing are stored in special buildings.
The irradiated fuel is transferred to a cooling pond at the
site before it is transported to Windscale for further storage and eventual
reprocessing. The resulting high-level waste is stored at Windscale pending
ultimate disposal.
Liquid Waste
The principal sources of radioactive liquid effluents are:
·
Water from the reactor coolant driers
·
Soluble and insoluble activity from the
irradiated-hel cooling pond
·
Soluble activity from the sludge and resin tanks
·
Washing water from plant and hel flask
decontamination and drainage from reactor areas
Tritium is produced in the graphite core and reflector from
the reaction,
Where the Li-atoms are present in the graphite as
impurities. The tritium atoms exchange with hydrogen in the methane present in
the coolant and is finally removed in the coolant driers as tritiated water.
Gaseous Waste
The main source of radioactive releases to the atmosphere is
the radioactivity associated with the reactor carbon dioxide coolant. The
coolant is released to the atmosphere when the reactor or refueling machine is
depressurized (blow down). Additionally, as in any large pressurized vessel,
leakage occurs through glands and seals. Other sources of radioactive releases
include ventilation air from contaminated areas and the air used to purge the
reactor pressure vessel during periods when man-access is required within it.
The major contributors to the radioactive discharge to the
atmosphere are:
41Ar, I4C, 16N, 3H and 35S
High Temperature Reactor
High Temperature Reactors HTR, first developed during the 1970s
and 1980s in Germany and the USA, may be doing a comeback based on their high
thermal efficiency and their very high degree of “intrinsic” safety. These
characteristics derive from the use of helium gas as coolant (Melese and Katz
1984), graphite as moderator, and, above all, a very unusual type of fuel.
Fuel Elements of HTRs (Particles,
Pebbles and Prisms)
What constitutes the specificity of the HTRs and gives them
their qualities is their fuel. It was invented in Harwell, UK, during the
mnid-1950s. Wholly refractory and helium cooled, the core is made of tiny
fissile particles, less than 1 mm diameter, dispersed within a graphite
moderator.
The kernel of each individual particle is coated (Fig. 16) by catalytic cracking in
fluidized bed, with a number of concentric layers, like the sugar coatings of
the almond in a dragée: inner
layers of pyrocarbon which protect a layer of silicon carbide SIC from the hot
kernel and outer layers of dense pyrocarbon which can withstand the pressure of
fission gases up to very high burnups. The SiC layer is a leak tight barrier to
contain the fission products: it plays the role of the cladding in a
conventional fuel pin. The outermost carbon layer facilitates the agglomeration
of the particles inside “compacts” or pebbles (Fig. 17).
Extremely divided and fully refractory, this fuel enables
the reactor to operate with very high coolant temperatures (we shall see how
high later) and therefore with an excellent thermal efficiency while the center
of the particle remains relatively cold. The coated particle is, indeed, a very
special breed of fuel element:
·
There are several tens of billions particles in
a reactor core. It is therefore a mass produced object, whose quality can only
be assessed by statistical tools (no fewer than 1011 individual
claddings constitute the first barrier against radioactivity dispersal, versus
the 2x105 pin claddings of a PWR).
·
There is an almost unlimited flexibility in the
core composition. You can freely select the nature (fissile, fertile, burnable
poison, mixture) and dimension (i.e., self-protection) of the kernels. You can
adjust the particle concentration within the graphite matrix of the compact or
pebble, as well as their distribution by size (double heterogeneity).HTRs can
therefore be adapted to any fuel cycle whatsoever.
Figure 16: Scanning electron micrography of a coated
particle
The actual flexibility offered to the designer can be
illustrated by the two types of fuel elements used in the HTR prototypes,
prisms in Fort Saint Vrain and pebbles in THTR, not to mention many other types
tested in Dragon (annular, teledial, etc.).
Figure 17: The two families of fuel elements
(compacts-in-prism and pebbles)
First HTR Demos: Dragon, AVR, Peach Bottoms
Dragon
It is around 1956, while the UK was launching its big Magnox
program, that the Harwell discovery was developed inside the Dragon Project, an
ad hoc OECD enterprise located on the UKAEA Winfrith site. A demonstration
facility was built and operated at Winfrith as soon as 1964, and established
successfully the HTR feasibility. In addition to building and operating the
reactor, the 12-country Dragon team paved the way for future HTRs by exploring
reactor designs and testing a number of fuel cycles (low enriched uranium LEU,
thorium/235U, and plutonium, both with oxide or carbide kernels). Being
multinational, Dragon Project introduced the HTR to the whole Europe and
triggered interest in the USA, then Japan (Table 13).
Table 13: Characteristics of HTR demos
AVR
Dragon partner through Euratorm, Germany developed HTRs as
its first purely national design. As soon as 1967, the AVR, a very innovative
demonstration reactor, began operation in Julich, where it operated very successfully
for more than 20 years.
Figure 18: AVR HTR Pebble
Both core and steam generators were contained in a single
steel double walled pressure vessel. The helium coolant was circulating upward,
as its outlet temperature was increased from 750 to 8500C, and then
to 9500C during its last two years of operation. The temperature was
even pushed to 1,0500C in the last days before shutdown.
The main innovation of AVR was its spherical fuel element,
the 6cm diameter graphite “pebble” inside which coated particles were
agglomerated (Fig. 18). Hundred
thousand pebbles are heaped inside a funnel shaped graphite cavity. The control
rods move in channels within the graphite reflector (Fig. 19).
Six hundred pebbles a day were continuously extracted from
the bottom of the funnel, and tested for physical integrity and burnup. Ninety
percent were recycled on top of the heap, with the required complement of fresh
pebbles: an intact pebble traveled therefore ten times through the core before
disposal. Each pebble contained on average 1 g of HEU and 6 g of thorium, in particles
with a BISO all pyrocarbon coating. Burnups as high as 150 GWd/tons were
routinely reached in AVR.
Figure 19: The Julie
AVR
Peach Bottoms
Fifty-three electricity producers, with support from the US government, very soon entered the HTR
race and built a demonstration reactor at Peach Bottom (Pennsylvania), which
reached its nominal 40 MWe power in
1967. The peach Bottom fuel element is close to the Dragon design, a long
hexagonal graphite prism with a pile of annular compacts inside. The core,
surrounded by a graphite reflector, is located at the bottom of a steel
pressure vessel. The first core was fabricated using particles with a coating
still imperfect. It was soon replaced by a core with much improved fission
products retention. Peach Bottom was decommissioned in 1974, just after the
start-up of the Fort Saint Vrain prototype.
The very successful operation of these three demos gave
great hopes concerning the future of the HTR families. Unfortunately, the
performances of their immediate successors were less bright.
Fort St
Vrain and THTR Prototypes
In 1968, 1 year only after the start-up of Peach Bottom,
general atomic started the construction on the Fort St Vrain site of a 330 MWe
HTR prototype. The operator was to be Public Service of Colorado, a small utility without previous
nuclear experience. Being a prototype, Fort Saint Vrain was built with federal
support (Table 14).
Table 14:
Characteristics of HTRs prototypes
The general layout
of the reactor (Fig. 20) is strongly inspired by the 500 MWe St Laurent UNGG design but with a reactor cavity
six times smaller.
The core is
composed of 1,483 prismatic fuel elements (Fig. 21) superposed on six layers.
Each fuel element is a hexagonal graphite prism in which cylindrical blind
channels are bored. Cylindrical compacts fill these channels, which are
surrounded by coolant channels in which helium circulates downwards, under 48
bars pressure.
Figure 20: Fort Saint
Vrain reactor layout (Melese & Katz 1982)
The compacts are
fabricated by mixing two types of particles: fissile particles with a kernel of
HEU dicarbide 235UC2 with TRISO coating including one SiC layer and fertile particles
ThC2 with a BISO
coating without silicon carbide. The core is axially and radially zoned, and it
is reloaded by 1/6th at each annual outage.
Twelve once-through
helical steam generator modules are located below the core.
Critical in 1974,
Fort St 'rain was connected to the grid in 1976, to be decommissioned in 1989
with a cumulative load factor of 30%. The fuel behaved successfully, but the
reactor’s overall design was rather a failure.
Figure 21:
HTR Prism
At the very
beginning of the 1970s, while Fort St Vrain was under construction and Peach
Bottom still operating, a few US utilities ordered from General Atomic, then a
subsidiary of Gulf Oil (and soon of Shell
as well), 8 large HTR rating 1,160
or 770 MWe, two very similar models with either three or two loops.
The general layout (fig.
22) is derived from the pods-type AGR, but with downward coolant
circulation to protect the upper structures and control rod mechanisms from the
hot helium. Prominent in this design stands the massive prestressed concrete
reactor vessel (PCRV), with vertical tendons and circumferential wire
wrappings. In the center of the PCRV, a main cavity contains the core while
peripheral cavities (the “pods”) contain the helical SGs and the helium
circulators. The core, extrapolated from Fort Saint Vraii, was supported by a
forest of graphite pillars above a lower plenum connected to the pods by hot ducts
with thermal insulation.
Figure 22: 1160MWe HTR Project (1973) (Melese &
Katz 1982)
The 1974 oil shock triggered overnight a rash of nuclear
project cancellations in the USA:
latest ordered, most HTR projects were victims of this epidemic. The vendor itself canceled in 1975 the last two survivors. One must remember that there has been no surviving nuclear order in the USA since 1973.
latest ordered, most HTR projects were victims of this epidemic. The vendor itself canceled in 1975 the last two survivors. One must remember that there has been no surviving nuclear order in the USA since 1973.
The
Schmehausen THTR
In 1970, The German industry received an order for a 300 MWe
prototype developed from AVR, to be built on the Schmehausen site.
The construction of THTR lasted 14 years, but the plant was
operated only for four years before definitive shutdown. A number of technical
difficulties, mostly due to the increase in size were met, but none appeared
insolvable: the control rods had to be inserted in the stack of pebbles instead
of in the reflector, the mass flow of helium was too big to allow counter
current circulation of helium and pebbles, fixation of the graphite to the wall
of the core cavity above the level or the top or the pebbles heap proved to be
uneasy (Fig. 23). But the real
reasons of THTR premature shutdown were the context of a German public opinion
becoming antinuclear, and power struggles between the Land and the Federal government
about licensing issues.
The mediocre performances of Fort St Vrain did not convince
utilities from other count tries do order HTR plants, but General Atomic, well
introduced in the Us Congress, managed to get year after year enough money on
the Us DOE budget to keep alive a small but highly competent team or engineers
and scientists. But the thorium cycle was abandoned because it needed highly
enriched uranium HEU to start the cycle and as a complement to because HTR are
not breeders. After 1974 and the Indian explosion, the civilian use or HEU
became taboo for nonproliferation reasons. Today, one would likely use
plutonium to start a thorium cycle.
Figure 23: THTR Schematics
Lesson learned from First HTRs
Despite the abortion of the US and German programs, the
results of this first part of the HTR saga were far from negligible.
On the plus side:
·
This type of reactor can reach a high thermal efficiency,
as good as that of the best gas turbines
·
Cold fuel, refractory core, high thermal
inertia, one-phase coolant chemically inert: all these elements result in a
high level of safety and forgiveness of operator mistakes
·
The particle-based fuel can accommodate any
possible fuel cycle
·
The first small demos have proven the concept
feasibility (and the rocket program described below demonstrated the existence
of huge margins)
·
HTR is one of the very few concepts to offer
real prospects of non electrical uses of fission (together with the Gas-cooled
breeder, which exists only on paper)
On the minus side:
·
The low core power density means a large vessel
and therefore a high capital cost
·
The GA 1160 Project did not have a secondary
containment
·
If core meltdown is beyond credibility (though
the SiC layer begins to deteriorate when the particle temperature exceeds 1,6000C),
a massive water ingress in the hot core might provoke a dangerous weakening by
corrosion of the core support pillars
·
The core itself is quite refractory, but the
long-term behavior of the materials outside the core exposed to hot helium is
of concern. This includes the concrete PCRV
·
Both prototypes did not meet with great success
Generation IV GFR (Gas Cooled Fast Reactor)
The Generation IV GFR project addresses a twofold challenge:
combining high thermodynamic efficiency through high temperatures, and high
neutronic efficiency (with significant economization of resources in the case
of the uranium-plutonium cycle) through fast spectrum conditions. It has
therefore been referred to as a “high-efficiency reactor,” constituting the second
wave of modern GCRs (beyond HTRs).
The specific advantages of the GFR are the following:
knowledge and operating experience acquired with GCRs, twofold concept allowing
the nuclearization of high-performance modern technologies developed outside
the nuclear field, and progressive transition via the HTR-type thermal GCR
fleet that will precede it.
Figure 24: ANTARES flow chart (from AREVA)
To address the twofold challenge of fast spectrum and high
temperature conditions, the GFR possesses advantages inherited from modern HTR
concepts, i.e., combination of a chemically inert coolant (helium) transparent
to neutrons (no capture, little diffusion, no activation, even at pressures of
several tens of bar) with a refractory and mechanically robust core using “cold
“fuel and locally confining FPs at high temperatures.
This combination makes it possible to benefit from the
decoupling of neutronics and thermal hydraulics, and thermo-mechanics and
chemistry. The design of nuclear reactors is determined by the analysis of
failure modes associated with couplings of neutronics, thermal hydraulics,
material mechanics, and chemistry. The benefits of said decouplings, associated
with a more efficient fuel, manifest themselves under both normal and accident
operating conditions.
The helium flow path in the core can be modified beyond a
minimum core volume without significant disturbance of spectrum, capture, and
leak conditions. Together with the possibility of significant increases in core
temperature, this property allows for reducing the pumping power under normal
operating conditions and favors gas convection in decay heat evacuation situations.
The practical exclusion of recriticality accidents through the
insertion of reactivity exceeding the delayed neutron fraction constitutes a
significant advantage for the design of a fast neutron reactor concept subject
to increased core sensitivity (namely due to the loading dominated by
plutonium, which reduces the delayed neutron fraction βeff, and also due
to the short lifetimes of prompt neutrons under fast spectrum conditions, i.e.,
approximately one microsecond).The increase in reactivity due to
depressurization can be limited by design to a value less than βeff
.
The use of a chemically inert coolant makes it possible to
benefit from the refractory and mechanical robustness qualities of the core. In
severe accident situations, an additional margin of a few 100o is
guaranteed beyond the fission gas containment limit (i.e., before extended core
degradation leading to a loss of geometry inhibiting core cooling in the long
term, or to a core collapse possibly resulting in a significant release of
energy due to recriticality).
Helium is not activated under neutron flux. It is chemically
inert and, if pure, does not contribute to structural corrosion or activation. This
advantage has been confirmed in HTRs. Combined with the HTR fuel containment
quality, it has led to very satisfactory operating experience in terms of
doses. It is particularly advantageous in the hypothesis of reactors operating on
a direct cycle with gas turbines.
It is therefore possible to benefit from the remarkable
increases in efficiency and competitiveness achieved by fossil fuel plants over
the past decades with conventional industrial coolants (gas and steam or
supercritical water). This is particularly clear in the case of gas turbines. The
GFR system combines high thermodynamic and neutronic efficiency. It is a modern
and competitive technology capable of following up on progress with fossil
thermal systems (particularly as regards coal, a potential competitor in the
long term). It guarantees a sustainable development of nuclear energy by
maximizing the use of uranium resources through industrially optimized
plutonium recycling.
Specific Problems Associated with GFR
These problems are due to the above-mentioned twofold
objective (high temperatures and fast spectra) and mainly concern the
following: fuel and structural materials under flux, economic fuel reprocessing
and fabrication, and evacuation of residual power under loss-of-pressure accident
conditions. They can be overcome through a combination of technological
innovation and optimized reactor design.
A steel-clad pellet-type fuel with large volumes of fission
gas expansion outside the core, such as that developed for SFRs, can be adapted
for a GFR core. However, it does not provide the second set of properties
sought, characteristic of micro-confining, refractory (cold), and mechanically
robust fuels such as the graphite matrix particle fuels tested up to very high
burnup fractions under thermal spectrum conditions in HTRs. Due to the damage
associated with fast spectrum irradiation, and given the power density sought,
these fuels are not usable as such in a GFR system.
In addition, imposing fission gas retention within the core
volume leads to a diluted core and makes it more difficult to obtain a hard
spectrum. Adapting such concepts, modifying the materials and ensuring
competitive fuel reprocessing and fabrication is therefore one of the greatest
challenges for the GFR. The same applies to the core structures and, more
generally, the flux-exposed structures.
The need to evacuate residual power under loss-of-pressure
accident conditions with loss of nominal forced gas convection contributes to
the design of the backup systems. The combination of high specific power (aimed
at minimizing the plutonium inventory required for a given power output) and
high concentration of fissile nuclei (aimed at hardening the spectrum) imposes
a power density of between 50 and 100 MWth/m3. Correlatively, the thermal
inertia of the core and structures (thermally coupled) is reduced as compared
to HTR systems. As a result, the GFR cannot copy the solution implemented in HTR
systems, which is primarily based on thermal inertia. It is necessary to use
gas convection, maintaining a backup pressure capable of ensuring minimum
thermal efficiency for the coolant.
In a high-power core, with moderate power density compared
to that of conventional water-cooled reactors, increasing the core fraction
reserved for the coolant has little impact on spectrum hardness and reactivity.
We can therefore consider a “porous” core with low hydraulic resistance but
still mechanically robust. Satisfactory gas convection for residual power
evacuation as per admissible core outlet temperatures can be ensured for a core
power of approximately one electric giga watt through the use of backup systems
pumping requiring approximately 100 kW, assisted by natural convection capable
of taking over after a few hours.
The Advantages of the GFR System Have Two Main
Origins
Firstly, the genealogy and operating experience of the
series are very significant. In addition to the AGR, this particularly includes
the AVR (pebble-bed HTR), which operated for approximately 20 years and
sustainably achieved core outlet temperatures of 950oC. It also
includes the reactors of the NERVA nuclear space propulsion program, which
achieved exceptional performance in terms of hydrogen outlet temperature (2,500oC)
and power density (4,000MW/m3) due to the absence of industrial
constraints regarding cost, lifetime, and safety. The most powerful reactor of
the series had a total power output of 4.3 GWth, close to that of the EPR (and the
largest ever built in the USA).
Secondly, significant scientific and technological progress
has been achieved regarding high temperature and fluence materials, and also
high-temperature mechanics. In addition, at the system level, the benefits of
the twofold concept enable the exploitation of high-temperature technologies,
particularly for gas turbines.
The GFR is still mostly a “paper-design,” which cannot be
fairly compared to the Sodium cooled SFR for instance. The fuel remains to be
designed, even though some preliminary tests were carried out in the Phenix
reactor during its very last years of operation. It is notably impossible to
venture any comment about its future economics.
Its prospects will depend upon the magnitude of the
so-called “Renaissance” expected to take place soon in nuclear power
development, because this magnitude shall, through the fear of uranium
scarcity, determine the timing of deployment of generation IV fast breeders. If
this deployment starts around 2040, it will be too early for the GFR to have
passed through the steps of demo plant and prototype and be ready for
commercialization. If the renaissance is slower, then, maybe, the intrinsic
qualities of the GFR will open opportunities for its deployment.
Conclusion
As we have seen, gas cooling was extensively used in the
early days of nuclear power. For reasons completely independent of their
technical characteristics, HTRs missed their commercial introduction in the
late 1970s and are still today considered as “promising” designs. It is the personal
opinion of the author that as pure electricity producers they will not compete
economically with LWRs, the dominant species in the nuclear “biotope.” Their
future may be as co-generators of electricity and process heat, notably to
produce hydrogen as feedstock for synthetic liquid fuels, which would provide
the opportunity for nuclear power to enter significantly the transportation
sector. They might, later on, share this “niche” with GFRs.
