US NUCLEAR POWER PROGRAM (SANITIZED)

• Document Type: CREST
• Collection: General CIA Records
• Document Number (FOIA) /ESDN (CREST): CIA-RDP86M00886R000400100006-4
• Release Decision: RIPPUB
• Original Classification: C
• Document Page Count: 11
• Document Creation Date: December 21, 2016
• Document Release Date: August 15, 2008
• Sequence Number: 6
• Publication Date: November 13, 1984
• Content Type: MEMO

 Approved For Release 2008/08/15 : CIA-RDP86M00886R000400100006-4
SCIENTIFIC
AMERICAN
YU/osw(C
Gas-cooled Nuclear Power Reactors
Although the U.S. has only one such reactor, they have served well
overseas. They have an attractive safety feature: a loss-of-coolant
accident such as the one at Three Mile Island is all but impossible
I n March, 1979, the nuclear power in- simple explanation. The pressurized- plant's rated capacity is 330 megawatts
dustry suffered a shock from which light-water reactor is a straightforward of electric power, or MWe; which is
it has not yet recovered: the acci- adaptation of the highly compact reac- about a third the output of a standard
dent that disabled one of the nuclear re- tor designed to propel the first nuclear- commercial power plant. The reactor
actors at Three Mile Island. it is ironic powered submarine, the U.S.S. Nauti- has been operating at up to 70 percent of
that an event that caused no discernible lus, launched in 1954. An electric-power rated power and has recently been re-
physical harm to anyone crippled the version of the submarine reactor, built leased by the Nuclear Regulatory Com-
prospect for expanding nuclear power at by the Westinghouse Electric Company, mission for testing at up to full power.
the very time the nation was becoming went into service at Shippingport, Pa., The Fort St. Vrain demonstration
generally aware of the need for new do- three years later. The General Electric plant was designed and built for the
mestic sources of energy. Company soon introduced a reactor de- Public Service Company of Colora-
Although the experience at Three sign of its own, the boiling-water reac- do by the General Atomic Company
Mile Island demonstrated to the satis- tor, in which the heat generated by nu- as a part of the Atomic Energy Com-
faction of technically qualified people clear fission was carried away from the mission's Power Reactor Demonstra.
that present-day water-cooled nuclear core by steam rather than by pressurized tion Program. This followed the successreactors offer no significant threat to the hot water. ful Peach Bottom of Ata omic 40-MWe We Power prototype, Station
health and safety of the general public,
it also showed that such accidents and In both types of reactor it is essential No. 1, on the system of the Philadelphia
equipment failures can jeopardize the I that the reactor core not be uncov- Electric Company. In its seven and a
operability of the plant and place at risk ered, even briefly, lest the temperature half years of operation, from 1967
the heavy capital investment it, repre- in the core quickly rise and melt the through 1974, the Peach Bottom reac-
sents. In the extreme case an accident metal jackets around the fuel pellets, tor was available for service 86 percent
such as the one at Three Mile Island can as indeed probably happened at Three of the time (except for scheduled shut-
threaten the financial survival of the op- Mile Island. Light-water reactors are downs related to the research and devel-
erifting utility. equipped with redundant safety features opment objectives of the reactor itself).
Perhaps the principal lesson of Three to cope with a "loss of coolant" acci- The comparable figure for all U.S. nu-
Mile Island is that the current genera- dent. In such accidents the emergency clear reactors is about 66 percent.
tion of nuclear power plants is vulnera- equipment is designed to flood the core The key safety features that differenti-
ble to certain rare events that can lead to with water from a plentiful and assured ate the helium-cooled reactor from wa-
a condition where the time available for source. When the normal coolant flow ter-cooled reactors are two. First, since
responding correctly can be less than a was interrupted at Three Mile Island, a the reactor core is cooled by a circu-
minute. In such low-probability events sequence of improbable events, includ- lating gas completely confined within a
if the appropriate actions are not under- ing apparent operator error, interrupted massive reactor vessel, the reactor can-
taken immediately, the consequences the delivery of the emergency cascade not lose its primary coolant because of a
can be extremely costly even when pub- of water for too long a time. rupture of pipes outside the vessel. Sec-
lic safety is not at issue. It is reasonable All but one of the 71 commercially ond, if the circulation of the gas is in-
to ask: Do we need to be content with licensed and operating nuclear power terrupted by some mishap to all of
nuclear reactors of a design such that plants in the U.S., which currently sup- the main helium-circulation system, the
operators must react correctly within a ply about 11 percent of the nation's elec- temperature within the reactor core ris-
minute in order to prevent damage to tric power, are light-water reactors. The es only slowly because the fuel elements
the reactor? The answer is no. exception is a helium-cooled reactor, are embedded in a massive matrix of
That being the case, how did the U.S. the Fort St. Vrain Nuclear Generating graphite, which serves as the moderator
nuclear power industry come to follow Station, which was accepted for service for slowing down neutrons and which
the path it did? The dominance in the in the summer of 1979 by the Pub- can absorb the heat released by fission
U.S. of the light-water reactor has a lic Service Company of Colorado. The products after the nuclear chain reac-
 Approved For Release 2008/08/15 CIA-RDP86M00886R000400100006-4
lion itself has been halted. In a helium-
cooled graphite-moderated reactor the
nuclear reaction is halted by the inser-
tion of control rods, similar to those in
all reactors. or alternatively by the in-
jection of small boron-containing balls
that. "poison" the reaction. In water-
cooled reactors the loss of the coolant,
which also acts as the moderator, stops
the reaction.
If the emergency cooling systems in a
light-water reactor should fail to func-
tion, the temperature in the reactor core
would rise even after the reaction had
been shut off because the fission prod-
ucts accumulated in the fuel elements
would continue to release energy at a
high rate. At the instant of shutdown,
decay heat amounts to about 7 percent
of the rated thermal output of the reac-
tor, or about 210 megawatts in a water-
cooled plant with a thermal rating of
3,000 megawatts (equivalent to an elec-
tric output of 1,000 MW). It is estimated
that in such a loss-of-coolant accident
the temperature of the cladding around
the fuel elements would reach 3,000 de-
grees Fahrenheit and fuel failure would
begin in as little as 50 seconds in a pres-
surized-water reactor and in less than
two minutes in a boiling-water reac-
tor. With a helium-cooled reactor, in a
comparable event involving a system
depressurization and the total failure
of the helium-circulation system, more
than an hour would be required for the
temperature inside the core to reach
3,000 degrees F. At that temperature
both the coated fuel particles and graph-
ite fuel elements in a helium-cooled re-
actor would not be affected. The fuel
particles and graphite can readily with-
stand temperatures of up to 4,000 de-
grees F., which would not be reached
until at least 10 hours had elapsed. In
short, there is ample time to institute a
variety of reasoned emergency me2-
sures for restoring the flow of helium
.coolant.
T he vfftues of gas-cooled graphite re-
actofS have been widely recognized
elsewhere in the world. In the 1950's and
1960's, when the U.S. had committed
itself to light-water reactors, Britain and
France developed gas-cooled graphite-
moderated reactors, in which the cool-
ant was carbon dioxide rather than heli-
um. Britain now has more than 40 gas-
cooled reactors in operation or under
construction, France has seven and It-
aly, Spain and Japan have one each.
More than 600 reactor-years of operat-
ing experience has been acquired with
the European gas-cooled reactors. Such
reactors have accounted for nearly a
fifth of the total nuclear power gener-
ated in western Europe, Japan and the
U.S. so far.
The British and French efforts were at
an early stage in 1956 when a group of
physicists, many of them with experi-
ONLY HTGR IN THE U.S is near Denver, Colo. It is the Fort St. Vrain Nuclear Generating
Station, designed by General Atomic and owned and operated by the Public Service Company'
of Colorado. The plant, which has a capacity of 330 MR e, was placed in operation three years
ago and has since supplied more than two billion kilowatt-hours of electricity. On several oc-
casions the forced circulation of coolant in the reactor core has been interrupted for periods
of as much as 15 minutes without doing any detectable harm to the core or to the fuel elements.
ence at the Los Alamos Scientific Labo-
ratory, gathered at La Jolla, Calif., to
consider the problem of designing a re-
actor that would be both more efficient
and inherently more "forgiving" than
the reactors then available. Among
those present were H. A. Bethe of Cor-
nell University, Freeman J. Dyson of
the Institute for Advanced Study, Peter
Fortescue of the Atomic Energy Re-
search Establishment at Harwell in En-
gland and Frederic de Hoffmann, who
was then president of General Atomic.
Out of these early deliberations, aided
by concepts from Britain and France,
evolved the concept of the high-temper-
ature gas-cooled reactor, or HTGR, test-
ed at Peach Bottom and on a larger
scale at Fort St. Vrain. Because the U.S.
has plentiful supplies of helium, that gas
could be selected as a coolant instead of
carbon dioxide. Helium has the impor-
tant advantage that it is stable to the
high radiation flux in the reactor, does
not become radioactive, is chemically
inert and has excellent heat-transfer
characteristics.
The attractive features of HTGR's
were summarized by Joseph M. Hen-
drie, chairman of the Nuclear Regulato-
ry Commission, in testimony before a
congressional subcommittee in March,
1980. Such reactors, he said, "have effi-
ciencies as good as the best fossil-fuel
plants and are substantially more effi-
cient than the water-cooled reactors.
They not only get better thermal effi-
ciency but also get better energy utiliza.
tion out of each pound of'uranium that
is mined, better, in fact, by probably 15
or 20 percent than the best estimates for
advanced light-water fuels." He added
that HTGR's "have some safety advan-
tages. They are machines in which you
don't have to do a lot of things in a hurry
if something goes wrong because the
core structure is a great massive pile of
graphite, a very high-temperature and
stable material, so that if you get a pow-
er dropoff or the plant circulators go
out, [you have time] to sit down and
think about what to do."
The first series of gas-cooled reactors
built in Britain were called Magnox
reactors because the fuel rods, which
contained natural unenriched uranium,
were clad in a magnesium alloy. The
reactor core, incorporating many tons
of graphite, was housed in a large and
expensive steel pressure vessel many
times bigger than the pressure vessels
needed for light-water reactors. Then in
1958 French engineers showed that the
steel vessel could be replaced with a ves-
sel of prestressed concrete that could
be constructed in sizes large enough
to house the entire reactor system, in-
cluding the steam generators. The
prestressed-concrete reactor vessel, or
PCRV, is :kept in compression at all
times by a network of redundant, ten-
sioned steel tendons that can be moni-
tored and retensioned or even replaced
if necessary. Tightness against leaks is
ensured by a steel liner affixed to the
inside of the PCRV, which acts only as a
membrane seal to contain the coolant.
The liner and the walls of the PCRV
are cooled by water circulating through
 Approved For Release 2008/08/15 : CIA-RDP86M00886R000400100006-4
tubes that arc welded to the outer sur- NO was uranium oxide. a ceramic. clad
face of the vessel. in' stainless steel, a change made possible
for all French and British gas-reactor
systems. The high degree of safety af-
forded by the concrete vessel contribut-
ed to the British decision to construct a
second generation of reactors known as
advanced gas reactors (AGR's) near ur-
ban sites. In this second generation the
by the adoption of slightly enriched ura-
nium. With the new fuel AGR's could
operate at higher temperatures than the
Magnox-fueled reactors and were able
to "burn" more of the uranium 235 in
the fuel before refueling became neces-
sary. With higher temperatures the effi-
ciency of electric-power generation was
raised from about 30 percent to a little
more than 40 percent.
In the U.S. the Atomic Energy Com-
mission to predecessor agency of the
Department of Energy) nurtured inter.
est in gas-cooled reactors in the 1950's
and 1960's by supportingihe study of
several advanced reactor concepts. One
of the AEC's main objectives was tore- '-
duce the amount of uranium required
REACTOR CORE AND STEAM GENERATORS of the HTGR
are enclosed in a massive prestressed-concrete reactor vessel (PCRV).
Fora reactor designed to generate 860 MWe the PCRV would be 102
feet in diameter and 95 feel high. (A pres?.urized-light-water reactor
of slightly larger capacity is shown at the same scale in the illustra-
lion on the opposite page.) The graphite core of the 860-M We HTGR
fills a cylindrical volume 26 feet in diameter and 21 feet high. Heli-
um at a pressure of 1,050 pounds per square inch is circulated through
some 27,000 vertical channels in the core by four primary circulators,
of which only two appear in this cross section. The helium emerges
from the reactor core at a temperature of 1,266 degrees Fahrenheit
and enters the base of the steam generators, where it makes two pas-
ses over an array of helical and straight steam coils. Water boils up-
ward through the coils and is further heated as it passes downward to
emerge as superheated steam with a temperature of 1.000 degrees F.
and a pressure of 2,500 pounds per square inch. Not shown are three
coolant loops in which water-cooled beat exchangers can remove
heat from circulating helium when the steam-generating loops are
out of service. After a reactor shutdown fusion products in the core
release heat at a rate that is high initially but declines exponentially.
 Approved For Release 2008/08/15 CIA-RDP86M00886R000400100006-4
per unit of electric power; at that time
uranium resources appeared scarce in
relation to the projected needs. As a re-
sult the stud> emphasized reactor
con-cepts that were either breeders or ad-
vanced converters. A breeder creates at
least one atom of new fuel for each atom
of fuel consumed. Advanced converters
generally create from .7 to one atom of
fuel for each atom consumed. Light-wa-
ter reactors yield between.5 and.6 atom
of fuel for each atom consumed. The
high-temperature gas-cooled graphite-
moderated reactor qualifies as an ad-
vanced converter. It was one of the de-
signs that survived the inevitable weed-
ing out. The HTGR had strong support
from the utility industry because it is
competitive in capital costs with light-
water reactors and because it exploits a
uranium-thorium fuel cycle with a low
uranium consumption and therefore low
fuel costs.
The continuing evolution of gas-reac-
tor technology in Europe and the
U.S. has led to a convergence in at least
two important particulars for the next
stage in the development of gas-cooled
reactors. Helium replaces carbon diox-
ide as a coolant and the reactor core is
charged with nuclear fuel in a unique
system that dispenses with the need for a
metal cladcing. The two features have
been demonstrated not only at Peach
Bottom and Fort St Vrain but also in
two European reactors. The British op-
erated a helium-cooled 20-MW thermal
test reactor in southern-England from
1965 to 1976. In Germany an HTGR of
15 MWe (called the AVR) has been gen-
erating electric power since 1967, with
the outlet gas temperature being as high
as 950 degrees Celsius. (The tempera-
ture of water leaving the core of a pres-
surized-light-water reactor is about 610
degrees F., or 321 degrees C.).A.300-
MWe plant based on the AVR experi-
ence is now under construction in Ger-
many and is scheduled for start-up in
1984'br 1985. In the U.S. the Fort St.
Vrairnreactor of 330 MWe has provided
more than two billion kilowatt-hours of
power since 1978 and has demonstrated
the fuel performance and safety charac-
teristics of a contemporary HTGR de-
sign. The reactor has been subjected to
test transients up to and including the
complete loss of forced-coolant circula-
tion with no adverse effects on the reac-
tor core or on other primary compo-
nents of the system.
On the basis of the Fort St. Wain ex-
perience General Atomic, in cooper-
ation with Gas-Cooled Reactor Asso-
ciates (an organization of U.S. utility
companies) and the Department of En-
ergy, has developed a reference design
for an HTGR of 860 MWe. The goal
has been a design that is simple and con-
servative and that places high empha-
sis on the safety and protection of capi-
REACTOR
PRESSURE
VESSEL
S
LIGHT-WATER
COOLANT INLET
Q
STEAM CONDENSATE
FROM TURBINES
PRESSURIZED-UGHT-WATER REACTOR has been the commonest type of nuclear
power reactor In the U.S., with 44 reactors now in operation. Another 24 reactors are of the
boiling-water type, to which beat is carried off from the core by steam rather than by pressssur-
ized heated water. The core of a pressurized-light-water reactor rated at 1,100 MWe is shown
here. It is bowed in a steel pressure vessel about 15 feet in diameter, 40 feet high and from six
to 11 inches thick; the vessel is designed to operate with an internal pressure of 2,250 pounds
per square inch. The coolant water leaves the reactor at 610 degrees F. and passes to four
steam-generating loops, only one of which is shown here. Steam emerges from the generator at
540 degrees F. and a pressure of 1,000 pounds. At this temperature and pressure the system's
thermal efficiency is only 32 to 33 percent, compared with 38.5 percent for an HTGR system.
tal investment. The reactor core is con-
tained within a multicavity prestressed-
concrete reactor vessel. Helium leaves
the core at 1,266 degrees F. (reduced
from 1,494 degrees F. at Fort St Vrain)
and passes through four primary cool-
ant loops, where steam is generated
at a temperature of 1,000 degrees F.
and a pressure of 2,500 pounds per
square inch.
Helium is forced through each cool-
ant loop by a circulator driven by an
electric motor. (The Fort St. Wain cir-
culators are driven b_Lsteam.) The core
also is provided wit a diary cool-
ing system consisting of three loops,
each sufficient by itself to deliver 100
percent of the required cooling when the
helium in the reactor vessel is at the nor-
mal working pressure of 1,050 pounds
per square inch, or 50 percent of the
cooling when the vessel. is depressur-
ized. The helium that passes through the
CT
OR
RE
auxiliary cooling system is cooled with
water circulated by electrically driven
pumps that can be powered, if need be,
by diesel generators.
The combination of a stable, inert gas
for a reactor coolant and a highly tem-
perature-resistant graphite core struc-
ture allows steam to be generated at the
high temperatures and pressures found
in the modern electric-power plants that
burn fossil fuel. The net electric-gener-
ating efficiency of the HTGR reference
design is 38.5 percent, slightly below the
39.2 percent achieved at Fort St. Wain.
The small reduction was made in the
interest of simplifying the steam-gener-
ating system and to furnish still further
operating and safety margins.
A fundamental property of the heli-
um coolant, a confined gas that cannot
possibly condense to liquid form in the
system, is that it follows a linear temper-
ature-pressure relation; therefore instru-
Approved For Release 2008/08/15 : CIA-RDP86M00886R000400100006-4
 Approved For Release 2008/08/15 CIA-RDP86M00886R000400100006-4
FUEL PARTICLE developed for HTGR systems is .03 inch in diameter, about the size of a
grain of sand. A cross section of the particle is enlarged 150 diameters at the top. The nuclear
fuel itself is the crystalline-like material in the center. It consists of uranium oxycarbide in
which for best performance the content of the fissionable isotope uranium 235 is enriched to
93 percent. Layers of carbon and silicon carbide are built up by a high-temperature process.
sure can provide independent cM_,}.,
on each other. Because there is no 1i;,.
uid-gas interface, as there is in boiiinr.
water reactors (and in pressurized.u at..
reactors under certain emergence co,
ditions), a single unambiguous signal_
pressure-always indicaris the presencr
and physical condition of the coo:aat
Rapid depressurization of the prima-
cooling system can be tolerated wiN,
out concern that voids have formed and
left part of the core uncovered, as car.
happen when pressure is released from
water? that is. above its atmospheric
boiling point.
The Fort St. Wain experience has ver.
ified several important safety and oper-
ating advantages of HTGR's. Operat.
ing and maintenance personnel have re-
ceived exposures to radiation far below
the limits established for nuclear plants.
Fewer than 10 workers out of a total of
several hundred have received amounts
of radiation that were even measurable.
e ment readings.of temperature and
The Fort St. Wain system has re-
sponded smoothly and gracefully to
load changes caused either by transient
excursions in the power-generating cy-
cle or by the temporary shutoff of equip-
ment within the plant. Because the core
of the HTGR is large and releases less
heat per unit volume than light-water
reactors do and because the massive
core, incorporating some 1,500 tons of
graphite, has a large capacity to absorb
heat if coolant flow is reduced or inter-
rupted, the reactor responds slowly to
an unexpected operational upset, allow.
ing the operators enough time to take
appropriate action: hours rather than
seconds.
At Fort St. Wain five such upsets have
interrupted the forced circulation of he-
lium for extended periods without giv-
ing rise to a measurable increase in the
temperature of the core or harming the
plant or the fuel in any way. The risk of
damaging the reactor or the reactor core
through operator error is virtually elim-
inated. Thanks to the HTGR's thermal
stability the system for bringing the ac-
tivity of the reactor to a halt by the inser-
tion of neutron absorbers and the sys-
tems for emergency cooling can be of
simple design. There is also ample time
for such systems to be actuated manual-
ly if it is allowed by regulation. One con-
sequence of the Three Mile Island ac-
cident is that the Nuclear Regulatory
Commission now requires the full-time
presence of an on-iite expert, called a
shift technical adviser, at nuclear power
plants. Fort St. Wain is the only reactor
exempted from this rule; an expert is not
required to remain on the site but is on
call to report within an hour.
The prestressed-concrete reactor ves-
sel is incorporated in the design as a ma-
jor safety feature. First, a catastrophic
rupture of the PCRV is such a remote
possibility that risk analysts character-
 - --
tze it as 0nn5 uw,...,..... ._..-.: ---
doe that give the PCRV its strength is
independent and redundant: the vessel is
in a constant state of compression. Sec-
ond. the PCRV is designed to withstand
an ultimate pressure of more than twice
the normal operating pressure, or some
2.400 pounds per square inch. Any
crack in the steel liner that might result
from excessive pressure can do no more
than give rise to a sloe gas leak: such
leaks tend to seal themselves when the
pressure is reduced slightly. Third. total
depressurization can result only if there
is a failure of one of. the pipe penetra-
tions or small service lines that pass
through the wall of the PCRV. Such a
hypothetical failure is an extremely low-
probability event. Moreover, at each
penetration site the vessel is equipped
with flow limiters that prevent the rapid
release of gas that could cause structur-
al damage to the core or to the cool-
ing system.
T he improved performance charac-
teristics of the HTGR also offer sev-
eral environmental advantages over the
current generation of reactors. Because
an HTGR operates at an efficiency of
about 39 percent compared with an effi-
ciency of about 33 percent for light-wa-
ter reactors. an HTGR releases about 25
percent less waste heat to be dissipated
into the sur:^unding environment. If the
heat, in the form of hot water, is reject-
ed into a nearby lake or river, concern
FUEL ROD AND FUEL BLOCK for an HTGR are shown respectively at the left and the
right The fuel rod, about 2.5 inches in length, consists of tens of thousands of fuel particles
bound in a graphite matrix. Each fuel block, which is approximately 14 inches across and 31
inches high, holds 1,656 fuel rods packed in hexagonal arrays. The numerous empty channels
in the block are paths for the flow of helium. The large central bole accommodates a mecha-
nism for inserting the fuel blocks in the core of the reactor. The core of an 860-MWe reactor
will require 3,512 blocks. Each 270-pound fuel block contains on the average 1.54 pounds of
U-235 and 35 pounds of thorium 232. In its four-year residence in the reactor such a block
would yield energy equivalent to 2,500 tons of coal or 12,000 barrels of fuel oil. If the unburned
U-235 and the U-233 created from thorium were recovered and recycled, the energy equiv.
alent of the original nuclear fuel would rise to some 11,000 tons of coal or 54,000 barrels of oiL
about raising temperatures to a point
harmful to the aquatic ecosystem is re- th
duced proportionately. If cooling tow- shipped off-site with little difficulty or tenance and refueling. For example, e
ers are used to dissipate the heat, they retained on-site. The solid wastes pro- entire primary coolant, helium, is con-
consume less water and can be smaller duced by an HTGR should total less fined within the prestressed-concrete re-
and less expensive. If cooling ponds are than 2,000 cubic feet per year. Some 80 actor vessel. The PCRV itself provides
used, an HTGR plant with about a third percent will consist of low-level waste all the necessary shielding for personnel.
more megawatts of capacity than a (such as paper, filter elements and spent so that maintenance work can be done
light-water plant can be sited on a pond resins) that is only slightly contaminated throughout the reactor building while
of a given size without exceeding a spec- and can be shipped off-site in drums for the plant is in operation. Because the
ified pond temperature. Where dry cool- burial or burning with virtually no effect entire secondary steam system is essen-
ing towers must be adoptedto meet en- on the environment. The remaining 20 tially free of radioactivity all equipment
vironmental regulations or fit available percent will be intermediate-level waste, in the steam cycle outside the PCRV,
water supplies, the loss of plant capacity consisting chiefly of reflector blocks, including the turbine-plant equipment,
in hot'weather will be only about half which must be periodically replaced. can be operated and maintained as it
as greflt with an HTGR as it is with exist- Such waste can be shipped off-site in would be in a plant fired with fossil fuel.
ing nuclear power plants. As a result an shielded 55-gallon transport casks for B e [useithe in oanunHTGt of flow
isng to
HTGR plant can be situated at a remote long-term safe disposal. elium-
gas-recovery owing
ication an
in cost. d site with a smaller pen- systems incorporated ind the standard toothe6ttu0 rrbines tinsaal'ight-waterflpowerr
and
arty
level of radioactivity in nor- HTGR plant should reduce the radioac- plant of the same output, all the equip-
The lest.
mal discharges from all nuclear pow- tive levels in released gases to several ment associated with the steam and
er plants is carefully monitored. An orders of magnitude below the current feed-water cycles of an HTGR plant is
HTGR plant inherently releases into the . Government regulation of.five milli- small and therefore easier to maintain.
plant process streams less radioactiv- rems per year. Tritium (the radioactive In general, maintenance, repair and han-
ity and at lower concentrations than a isotope of hydrogen) generated within dling costs are lower in an HTGR plant
light-water reactor does. In addition an the primary system of an HTGR is re- than they are in light-water plants be-
HTGR incorporates features that will moved in the helium-purification sys- cause helium, unlike water, is inert. non-
ensure that releases of radioactivity tem by an oxidizer that converts the trit- radioactive and noncorrosive.
from the plant to the environment are ium into tritiated water, which is subse- One big advantage of gas as a coolant
essentially zero. Routine decontamina- quently solidified and handled as solid is its transparency, which makes it pos-
tion procedures can be expected to pro- waste that can be readily isolated as the sible to inspect many areas within the
duce small volumes of low-level liquid tritium decays. (The half-life of tritium PCRV hesra iation Pshield CRV
wastes (less than 2,000 gallons per year is 12.26 years.) ing with a total activity of less than 150 The HTGR has evolved a number of makes it possible to carry out many in-
curies). Such small volumes can be features that simplify'operation, main- spection and maintenance tasks while
61
 STEAM TEI,4PERATURE
(DEGREES FAHRENHEIT(
1ODo
NET PLANT THERMAL
EFFICIENCY (PERCENT)
REACTOR-CORE POWER DENSITY
(WATTS PER CUBIC CENTIMETER)
07.1
MAKEUP COOLING WATER REQUIRED
(WET COOLING TOWERS WITH FORCED
DRAFT GALLONS PER MINUTE)
FUEL REQUIRED PER 10a WATTS
OF ELECTRICITY DURING 30
YEARS OF PLANT OPERATION
(SHORT TONS OF UaQ.)
PWR: URANIUM OF LOW
ENRICHMENT
HTGR: URANIUM OF MEDIUM
ENRICHMENT F7777774,510
PWR: LOW-ENRICHMENT URANIUM
WITH RECYCLE OF PLUTONIUM
HTGR: HIGH-ENRICHMENT URANIUM
WITH RECYCLE OF U-233
LIQUID
(CURIES PER YEAR)
GASEOUS
(CURIES PER YEAR)
ONCE-THROUGH
FUEL CYCLE
OPERATING CHARACTERISTICS of the 860-MWe HTGR (color) are compared with
those of a pressurized-water reactor of the same generating capacity (gray). The lower fuel
consumption of the HTGR can be attributed in part to higher thermal efficiency and in part to
the fact that for each atom of 1U-235 consumed in the HTGR about.7 atom of new fuel is cre-
ated. The pressurized-water reactor creates less than .5 atom of new fuel for each atom con-
sumed. With a once-through fuel cycle both systems convert a certain fraction of U-238 or
Th-232 atoms into isotopes of plutonium or uranium, some of which are beneficially consumed
before the fuel needs replacing. If the spent fuel could be recycled (which was contrary to the
policy of the last Administration), it would be preferable to fuel an HTGR with a mixture of
highly enriched uranium (about 93 percent U-235) and thorium. Some of the thorium would
be converted into fissionable U-233, which could be recovered and recycled to replace U-235
in subsequent fuel charges. Smaller volume of radioactive wastes from an HTGR results part-
ly from its higher efficiency and partly from advantages of helium over water as a coolant.
SOLID
(CUBIC METERS PER YEAR)
the reactor is runnin_. v hich reduces the
time the reactor must be taken out of
service for such purposes.
Essentially all structural members of
the PCRV. such as the vertical tendons
and the circumferential cable scrapping.
can be inspected visually while the reac-
tor is operating. Selected members are
continuously monitored for chances jr.
tension or strain that would indicate a
deterioration in performance. If neces-
sary. any structural member can be re-
placed. All external concrete surfaces,
except those immediately surrounding
the ports for the control rods, can be
inspected visually while the plant is run-
ning. The control-rod ports and the sur-
faces surrounding the site where the
control-rod drives penetrate the PCRV
can be readily inspected in the course
of refueling.
Recent refueling experience at Fort St.
Vrain has demonstrated the ease of
handling the HTGR's block-type fuel el-
ements. About 240, or a sixth, of the fuel
elements were removed from the core
and replaced with fresh fuel; the other
1,240 elements were left in place. The
refueling crew was exposed to such low
levels of radiation that measuring them
called for a microrem meter. By extrap-
olating from existing data one can cal-
culate that the sum of the integrated
man-rem exposure for the entire refuel-
ing operation following on the opera.
tion of the reactor at full power will be
less than five man-rem. Federal regula-
tions currently limit individual workers
to five rem over a period of a year.
Each HTGR fuel element is a graph-
ite block, hexagonal in cross section, 14
inches wide and 31 inches long. The
block is perforated lengthwise with 72
coolant channels and 138 blind holes
for fuel. Graphite is an ideal choice as
a moderator and a structural material
because its strength actually increases
with temperature. In the reference de-
sign the graphite fuel blocks are stacked
in columns of eight. This axially seg-
mented arrangement facilitates fabrica.
tion, handling and refueling.
The convenient block configuration
has been made possible by the develop-
ment of a specially coated fuel particle.
The kernel of each particle is a micro-
sphere of uranium oxycarbide (suitably
enriched in uranium 235) about .01 inch
in diameter. Around each kernel thin
layers of carbon, pyrolytic carbon and
silicon carbide are applied at high tem-
perature, yielding a tightly encased par-
ticle with a total diameter of about .03
inch. A similar form of encapsulation
is used for the thorium particles. The
technique ensures the containment of
the fission products. The tiny spheres
are tested in batches of 2.000 for struc-
tural integrity when they are exposed
to a radiation flux that simulates the in-
ternal environment of the reactor. The
particle-production process, which is
 ai nuautom::nom. :,cad the rigorous test-
ing procedure wsorl: together to achieve
a close control of quality
:kithough severe and unforeseen serv-
ice conditions in one region of the reac-
tor core might cause the particle coating
to fail and release fission products. the
failure would be limited to the area di-
rectly involved. In most reactors. where
the cladding of the fuel elements runs
the entire length of the reactor core. an
operating upset that ruptures a small
section of cladding could release fission
products from the entire length of the
fuel rod. The performance of the fuel
elements at Fort St. Vrain has fully met
design expectations. Indeed, the release
of fission products has been well below
the predicted levels. In sum, the fission-
product barriers in the HTGR fuel ele-
ment have been demonstrated to have a
high degree of reliability.
The properties of the HTGR make
it possible to exploit a wide variety
of nuclear fuel cycles with it. The cycle
that has been most intensively studied
and tested is the uranium-thorium one.
in which fully enriched uranium (93
percent U-235) serves as the primary
fissile material and thorium (Th-232)
serves as a "fertile" material. In the reac-
tor the thorium absorbs neutrons and is
ultimately :onverted into the fissile iso-
tope uranium 233, which can be recy-
cled in subsequent fuel reloadings. The
Fort St. Vrain reactor is fueled with ura-
nium enriched to 93.5 percent U-235,
in combination with thorium. The de-
sign of the plant allows the use of either
fully enriched or medium-enrichment
uranium (about 20 percent U-235). The
HTGR fuel-cycle costs', under the cur-
rent restraints on fuel reprocessing and
recycling, are essentially equivalent to
those of other commercial plants. Un-
less the policy is changed by the Adjnin-
istration spent fuel is to be stored indefi-
nitely, without the recovery either of
the unspent U-235 or of the U-233 or
plutonium created during the operation
of the reactor. This fuel cycle is com-
monly called the stowaway cycle.
If an HTGR were operated on a stow-
away uranium-thorium cycle with fully
enriched uranium, it would consume
about 20 percent less uranium over its
40-year life than a light-water reactor
would. If both types of reactor could be
operated with a full recovery of their
uranium and plutonium, the HTGR
would consume about 50 percent less
uranium. The HTGR therefore offers
the opportunity of saving substantial
amounts of uranium with either a stow-
away policy or a full-recycle one, pro-
vided the reactor is designed to accept
fully enriched fuel. The significance of
the potential uranium saving can be ap-
preciated when one considers that the
total fuel cost over the life of a nuclear
power plant is roughly equal to the total
initial cost of the plant.
1.2 2.4 - 6.0 60
MINUTES AFTER INSTANTANEOUS LOSS OF FORCED COOLING
INHERENT SAFETY OF AN HTGR is abown in graphs that compare the temperature in the
core of an HTGR, of a pressurized-water reactor and of a boiling-water reactor following a hy-
pothetical loss-of-coolant or loss-of-forced-ciirculation accident In the water-cooled reactors
the nuclear reaction Is baited automatically by the loss of water, which serves as a moderator.
In the HTGR the reaction must be stopped by the insertion of control rods that absorb neu-
trons. At the moment of shutdown decoying fission products in the fuel release heat at a rate
equivalent to 7 percent of the thermal output of the reactor. The beat release falls to 1 percent
in about two hours and to .5 percent in 24 hours. In the water-cooled reactors, in the absence
of emergency cooling, the temperature of the cladding of the fuel would rise in less than two
minutes to 3,000 degrees F., causing the cladding to fail. In the HTGR the mass of the graphite
moderator would absorb the heat released by fission products, so that 3,000 degrees would
not be reached for at least an hour. A temperature high enough to damage the graphite core
(about 4,000 degrees) would be attained only after at least 10 hours without forced cooling.
Over the past six years orders for
about 55 nuclear power plants have
been canceled. Only six years ago U.S.
utilities had demonstrated interest in
constructing 10 HTGR plants. Once the
Fort St. Vrain reactor has been brought
up to full power, which is scheduled for
this summer, and has demonstrated the
exceptional safety and reliability that its
designers confidently predict, it is rea-
sonable to assume that U.S. utilities will
look favorably on the HTGR when they
are again ready to place orders for nu-
clear power plants.
YEAR
AVAILABILITY OF
NUCLEAR STEAM
SUPPLY SYSTEM
(PERCENT)
PLANT
AVAILABILITY
(PERCENT)
OVERALL PLANT
CAPACITY FACTOR
(PERCENT)
1967
81
78
69
1968
Be
88
82
1969
86
84
67
1970
95
95
88
1971
90
87
78
1972
71
71
58
1973
95
94
76
1974
96
95
70
-CORE 1 AVERAGE
85
83
73
CORE 2 AVERAGE
89
88
74
TOTAL LIFETIME
88
86
74
STATION AVERAGE
RELIABILITY OF FIRST HTGR PLANT designed by General Atomic. the Peach Bottom
Atomic Power Station No. 1, is attested to by the statistics shown here. Apart from scheduled
down time or time lost for reasons unrelated to the reactor, the HTGR was available for supply-
ing steam for power generation 88 percent of the time. In achieving 74 percent of its rated elec-
tric-generating capacity over its seven-and-a-half-)'ear lifetime the Peach Bottom reactor ex-
ceeded the typical figure of 66 percent achieved by light-water reactors operated by utilities.
 Approved For Release 2008/08/15 : CIA-RDP86M00886R000400100006-4
CONCRETE
REACTOR
VESSEL(PCR/)
CIRCUMFERENTIA}
PRESTRESSING CABLES
HIGH-TEMPERATURE GAS-COOLED REACTOR (HTGR) is
a second-generation system more efficient than the 71 light-water
power reactors that now supph about 11 percent of U.S. electricity.
In this HTGR designed by the General Atomic Company the moder-
ator (the material that slows neutrons in the reactor core) is graphite
and the coolant is helium. In light-water reactors ordinary (but de-
mineralized and conditioned) water serves both as the moderator and
as the coolant. The HTGR shown would have an output of 860 mega-
mat efficiency of 38.5 percent, which is comparable to the efficiencl
of the best fossil-fuel plants and is higher than the 32 to 33 percent at-
tained by current light-water reactors. Because the core of the HTGR
contains nearly 1,500 tons of graphite, which has a high capacity for
absorbing heat, an HTGR is much less like] ' to be damaged than a
light-water reactor if there is an interruption in the flow of coolant
or a loss of coolant It was such an interruption that caused the acci-
warts of electricity (M We), slightly less than that of the largest power P. The reactor core and steam-generating system of the HTGR are
Approved For Release 2008/08/15: CIA-RDP86M00886R000400100006-4 to vessel with wails some 15 feet thick.
 Approved For Release 2008/08/15 : CIA-RDP86M00886R000400100006-4
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THE AUTHORS
HAROLD M. AGNEW ("Gas-
cooled Nuclear Power Reactors") is
president of the General Atomic Com-
pany, which he joined in 1979 after
more than eight years as director of the
Los Alamos Scientific Laboratory. His
work on the development of nuclear en-
ergy dates from the early 1940's. when
as a recent graduate of the University of
Denver he joined the small group that
worked with Enrico Fermi on the first
nuclear-fission chain reaction. In 1943
Agnew joined the Los Alamos laborato-
ry to participate in the development of
the atomic bomb. From 1946 to 1949 he
was again with Fermi at the University;
of Chicago, obtaining his Ph.D. there in
1949. Thereafter Agnew was at Los Ala-
mos except for three years (1961-64) as
scientific adviser to the Supreme Allied
Commander in Europe.
STEVEN WEINBERG ("The Decay
of the Proton") is Higgins Professor of
Physics at Harvard University and sen-
ior scientist at the Smithsonian Astro-
physical Observatory. He is currently
visiting professor at the University of
Texas at Austin. He did his undergradu-
ate work at Cornell University, being
graduated in 1954, studied for the next
year at the Niels Bohr Institute in Co-
penhagen and received his Ph.D. from
Princeton University in 1957. Thereaf-
ter he worked at Columbia University,
the Lawrence Radiation Laboratory of
the University of California, the Uni-
versityof California at Berkeley and the
Massachusetts Institute of Technology
before going to Harvard in 1973. He has
written two books: The First Three Min-
utes: A Modern View of the Origin of the
Universe and Gravitation and Cosmology:
Principles and Applications of the General
Theory of Relativity. He has received nu-
merous honors for his work on the theo-
ry of elementary particles, including five
honorary degrees and the 1979 Nobel
prize in physics, which he shared with
Sheldon Lee Glashow and Abdus Sa-
lam. Weinberg was recently elected a
foreign member of the Royal Society.
E. P. ABRAHAM ("The Beta-Lac-
tam Antibiotics") is professor emeritus
of chemical pathology at the Sir Wil-
liam Dunn School of Pathology of the
University of Oxford. He did his under-
graduate work atOxford and obtained
his D.Phil. there in 1938. After two
years as a Rockefeller Foundation trav-
eling fellow in Stockholm he returned to
Oxford to work on the isolation and
chemistry of penicillin with Howard
W. Florey, Ernst B. Chain and others.
In 1940 he and Chain discovered the
enzyme penicillinase;. in 1953 he and
G. G. OF. Newton isolated cephalosporin
C from an impure preparation of peni-
cillin A. Subsequent work,at Oxford
and in pharmacemical companies led to
the introduction of cephalosporins into
medicine. Abraham. a Fellow of the
Royal Society. was knighted last year.
His leisure interests include walking and
skiing.
CARL R. WOESE ("Archaebacte-
ria") is professor of microbiology and of
geneticsland development at the Univer-
sity of Illinois at Urbana-Champaign:
from 1972 to 1979 he held a third ap-
pointment as professor of biophysics.
His bachelor's degree (in mathematics
and physics) was awarded by Amherst
College in 1950 and his Ph.D. (in bio-
physics) by Yale University in 1953. He
then spent two years at the University of
Rochester School of Medicine and Den-
tistry before returning to Yale to do re-
search in biophysics. From 1960 to 1963
he worked as a biophysicist at the Gen-
eral Electric Research Laboratory, join.
ing the faculty of the University of Il-
linois in 1964. "My entire career," he
writes, "has been a deepening venture
into the recesses of evolution." Much
of his work has been on the evolution
of the mechanism whereby the genetic
code is translated in the cell by the ribo-
somes. Now, Woese says, "it is time to
press deeper, and my interest is turning
to the evolution of the ribosome itself."
ROBERT G. BLAND ("The Alloca.
tion of Resources by Linear Program-
ming") is assistant professor in the
School of Operations Research and In-
dustrial Engineering and the Center for
Applied Mathematics at Cornell Uni-
versity. "I studied at Cornell," he writes,
"and got my Ph.D. in operations re-
search there in 1974. I was assistant pro-
fessor of mathematical sciences at the
State University of New York at Bing-
hamton, research fellow at the Center
for Operations Research and Economet.
rics in Louvain and professor of man-
agement at the European Institute for
Advanced Studies in Management in
Brussels before returning to Cornell in
1978. My research interests are in the
theory and applications of graphs and
networks, mathematical programming
and discrete optimization."
BERND HEINRICH ("The Regula-
tion of, Temperature in the Honeybee
Swarm") is professor of zoology at the
University of Vermont. He writes: "I
grew up in rural Maine after coming to
this country (when I was 10 years old)
from Germany. I received bachelor's
and master's degrees at the University
of Maine and a Ph.D. from the Univer-
sity of California at Los Angeles, all in
zoology. For the next 10 years I was in
the entomology department at the Uni-