U.S. patent application number 11/106427 was filed with the patent office on 2006-03-16 for method of manufacturing a radioactive-substance storage member, billet for use in extrusion of the same, and square pipe.
This patent application is currently assigned to MITSUBISHI HEAVY INDUSTRIES, LTD.. Invention is credited to Toshiro Kobayashi, Kazuo Murakami, Katsunari Ohsono, Yasuhiro Sakaguchi, Toyoaki Yasui.
Application Number | 20060057013 11/106427 |
Document ID | / |
Family ID | 18971601 |
Filed Date | 2006-03-16 |
United States Patent
Application |
20060057013 |
Kind Code |
A1 |
Ohsono; Katsunari ; et
al. |
March 16, 2006 |
Method of manufacturing a radioactive-substance storage member,
billet for use in extrusion of the same, and square pipe
Abstract
An aluminum powder is mixed with a neutron absorber powder
through cold isostatic press to form a preliminary molding. The
preliminary molding is then subjected to sintering under no
pressure in vacuum. After sintering, a billet is subjected to
induction heating and hot extrusion to form a square pipe.
Inventors: |
Ohsono; Katsunari; (Hyogo,
JP) ; Murakami; Kazuo; (Hyogo, JP) ;
Sakaguchi; Yasuhiro; (Hyogo, JP) ; Kobayashi;
Toshiro; (Hiroshima, JP) ; Yasui; Toyoaki;
(Hiroshima, JP) |
Correspondence
Address: |
OBLON, SPIVAK, MCCLELLAND, MAIER & NEUSTADT, P.C.
1940 DUKE STREET
ALEXANDRIA
VA
22314
US
|
Assignee: |
MITSUBISHI HEAVY INDUSTRIES,
LTD.
Chiyoda-ku
JP
|
Family ID: |
18971601 |
Appl. No.: |
11/106427 |
Filed: |
April 15, 2005 |
Related U.S. Patent Documents
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Application
Number |
Filing Date |
Patent Number |
|
|
10125372 |
Apr 19, 2002 |
6902697 |
|
|
11106427 |
Apr 15, 2005 |
|
|
|
Current U.S.
Class: |
419/28 ;
419/32 |
Current CPC
Class: |
Y02E 30/30 20130101;
B22F 1/0003 20130101; G21F 1/08 20130101; G21F 5/012 20130101; B22F
2009/041 20130101; G21F 3/00 20130101; G21C 19/40 20130101; C22C
1/05 20130101; B22F 2998/10 20130101; B22F 2999/00 20130101; B22F
2998/10 20130101; B22F 1/0003 20130101; B22F 3/02 20130101; B22F
3/10 20130101; B22F 3/20 20130101; B22F 2999/00 20130101; B22F
1/0003 20130101; B22F 9/04 20130101 |
Class at
Publication: |
419/028 ;
419/032 |
International
Class: |
B22F 3/24 20060101
B22F003/24 |
Foreign Application Data
Date |
Code |
Application Number |
Apr 19, 2001 |
JP |
2001-121799 |
Claims
1-16. (canceled)
17. A method of manufacturing a radioactive-substance storage
member for use in storage of a radioactive substance, the method
comprising: mixing an aluminum powder with a neutron absorber
powder; pressing the mixed powder to form a preliminary molding;
and sintering the preliminary molding under no pressure in an inert
gas ambience.
18. The method according to claim 17, wherein the preliminary
molding has a weight density of more than 75% and less than
95%.
19. The method according to claim 17, wherein the mixing includes
mechanical alloying.
20. The method according to claim 19, wherein the mechanical
alloying includes ball milling, the ball milling using a ball
having a major constituent identical to an element previously added
to a sample in order to add the element from the ball into the
aluminum matrix by the wearing of the ball during the ball
milling.
21. The method according to claim 18, further comprising, after the
extruding, leaving the square pipe or plate in that extruded state
or naturally aging the square pipe or plate, the square pipe of
plate which configures a basket for accommodating an aggregate of
spent fuel or the rod to be inserted in a guide pipe for spent
fuel.
22. The method according to claim 19, further comprising, after the
extruding, leaving the square pipe or plate in that extruded state
or naturally aging the square pipe or plate, the square pipe of
plate which configures a basket for accommodating an aggregate of
spent fuel or the rod to be inserted in a guide pipe for spent
fuel.
23. A method of manufacturing a radioactive-substance storage
member for use in storage of a radioactive substance, the method
comprising: mixing an aluminum powder with a neutron absorber
powder; pressing the mixed powder to form a preliminary molding;
sintering the preliminary molding under no pressure in an inert gas
ambience to form a billet; heating the billet using an induction
heating unit; and extruding the induction-heated billet using dies
to form a square pipe or plate which configures a basket for
accommodating an aggregate of spent fuel or a rod to be inserted in
a guide pipe for spent fuel.
24. The method according to claim 23, wherein the preliminary
molding has a weight density of more than 75% and less than
95%.
25. The method according to claim 23, wherein the mixing includes
mechanical alloying.
26. The method according to claim 25, wherein the mechanical
alloying includes ball milling, the ball milling using a ball
having a major constituent identical to an element previously added
to a sample in order to add the element from the ball into the
aluminum matrix by the wearing of the ball during the ball
milling.
27. The method according to claim 23, further comprising, after the
extruding, leaving the square pipe or plate in that extruded state
or naturally aging the square pipe or plate, the square pipe of
plate which configures a basket for accommodating an aggregate of
spent fuel or the rod to be inserted in a guide pipe for spent
fuel.
28. A method of manufacturing a radioactive-substance storage
member for use in storage of a radioactive substance, the method
comprising: mixing an aluminum powder with a neutron absorber
powder; pressing the mixed powder to form a preliminary molding;
sintering the preliminary molding under no pressure in an inert gas
ambience to form a billet; extruding the billet using the heat
during the sintering to form a square pipe or plate which
configures a basket for accommodating an aggregate of spent fuel or
a rod to be inserted in a guide pipe for spent fuel.
29. The method according to claim 28, wherein the preliminary
molding has a weight density of more than 75% and less than
95%.
30. The method according to claim 28, wherein the mixing includes
mechanical alloying.
31. The method according to claim 30, wherein the mechanical
alloying includes ball milling, the ball milling using a ball
having a major constituent identical to an element previously added
to a sample in order to add the element from the ball into the
aluminum matrix by the wearing of the ball during the ball
milling.
32. The method according to claim 28, further comprising, after the
extruding, leaving the square pipe or plate in that extruded state
or naturally aging the square pipe or plate, the square pipe of
plate which configures a basket for accommodating an aggregate of
spent fuel or the rod to be inserted in a guide pipe for spent
fuel.
33-37. (canceled)
Description
FIELD OF THE INVENTION
[0001] The present invention relates to a method of manufacturing a
radioactive-substance storage member that is employed to configure
a cask or rack which accommodates and storing an aggregate of spent
nuclear fuel coming to an end of burn-up, and a billet for use in
extrusion of the member. It also relates to a square pipe which
accommodates an aggregate of spent fuel.
BACKGROUND OF THE INVENTION
[0002] An aggregate of useless nuclear fuel coming to an end of
burn-up at the termination of a nuclear fuel cycle is called the
spent nuclear fuel. Currently, the spent nuclear fuel is stored and
managed in a storage facility until it is reprocessed. For example,
in a storage system of fuel pool type, a SUS rack including bundles
of square pipes is sunken in a pool, and the aggregate of spent
nuclear fuel is accommodated in the square pipes to satisfy the
needs for cooling effect, shielding effect and non-criticality.
[0003] In recent years, a start has made at employing a boron-doped
stainless material for square pipes to configure a rack. The use of
such the square pipes can omit a neutron absorber material disposed
between the square pipes to eliminate a gap between the square
pipes. Therefore, it is possible to increase the number of square
pipes that can be inserted in a pit in the pool and accordingly
increase the number of the aggregates of the spent nuclear fuel to
be accommodated.
[0004] The above square pipes can be applied to various storage
systems including types of cask, lateral silo, pool and bold. It is
required to produce a large number of square pipes to configure
even a rack, and accordingly a technology capable of producing
square pipes efficiently is required. It is also required to absorb
neutrons radiated from the aggregate of spent nuclear fuel.
Therefore, the square pipes are required to have soundness in their
structures.
[0005] The square pipes are employed for storing the aggregate of
spent nuclear fuel. Other than the rack of square pipe type, a flat
plate type is also known, which requires efficiency in the
production and soundness in the structure. The present invention
relates to a method of manufacturing such square pipes, for
example.
SUMMARY OF THE INVENTION
[0006] The method of manufacturing a radioactive-substance storage
member according to one aspect of the present invention comprises
mixing an aluminum powder with a neutron absorber powder, pressing
the mixed powder to form a preliminary molding, and sintering the
preliminary molding under no pressure in vacuum.
[0007] The method of manufacturing a radioactive-substance storage
member according to another aspect of the present invention
comprises mixing an aluminum powder with a neutron absorber powder,
pressing the mixed powder to form a preliminary molding, sintering
the preliminary molding under no pressure in vacuum to form a
billet, heating the billet using an induction heating unit, and
extruding the induction-heated billet using dies to form a square
pipe or plate which configures a basket for accommodating an
aggregate of spent fuel or a rod to be inserted in a guide pipe for
spent fuel.
[0008] The method of manufacturing a radioactive-substance storage
member according to still another aspect of the present invention
comprises mixing an aluminum powder with a neutron absorber powder,
pressing the mixed powder to form a preliminary molding, sintering
the preliminary molding under no pressure in vacuum to form a
billet, and extruding the billet using the heat during the
sintering to form a square pipe or plate which configures a basket
for accommodating an aggregate of spent fuel or a rod to be
inserted in a guide pipe for spent fuel.
[0009] The method of manufacturing a radioactive-substance storage
member according to still another aspect of the present invention
comprises mixing an aluminum powder with a neutron absorber powder,
pressing the mixed powder to form a preliminary molding, and
sintering the preliminary molding under no pressure in an inert gas
ambience.
[0010] The method of manufacturing a radioactive-substance storage
member according to still another aspect of the present invention
comprises mixing an aluminum powder with a neutron absorber powder,
pressing the mixed powder to form a preliminary molding, sintering
the preliminary molding under no pressure in an inert gas ambience
to form a billet, heating the billet using an induction heating
unit, and extruding the induction-heated billet using dies to form
a square pipe or plate which configures a basket for accommodating
an aggregate of spent fuel or a rod to be inserted in a guide pipe
for spent fuel.
[0011] The method of manufacturing a radioactive-substance storage
member according to still another aspect of the present invention
comprises mixing an aluminum powder with a neutron absorber powder,
pressing the mixed powder to form a preliminary molding, sintering
the preliminary molding under no pressure in an inert gas ambience
to form a billet, extruding the billet using the heat during the
sintering to form a square pipe or plate which configures a basket
for accommodating an aggregate of spent fuel or a rod to be
inserted in a guide pipe for spent fuel.
[0012] The billet according to still another aspect of the present
invention comprises a mixed powder of an aluminum powder and a
neutron absorber powder, the mixed powder molded to have a weight
density ranging from 75% to 95% and sintered to fuse each powdery
particle to another.
[0013] The billet according to still another aspect of the present
invention comprises a mixed powder of an aluminum powder flattened
by mechanical alloying and a pulverized boron or boron-compound
folded and dispersed into the aluminum powder, the mixed powder
molded to have a weight density ranging from 75% to 95% and
sintered to fuse each powdery particle to another.
[0014] In the billet according to still another aspect of the
present invention, the billet is employed as a structural material
of storage or transportation containers for spent nuclear fuel, the
aluminum powder containing an additional element such as Zr for
imparting high strength.
[0015] In the square pipe according to still another aspect of the
present invention comprises, the square pipe is formed by mixing an
aluminum powder with a neutron absorber powder through cold
isostatic press or cold unidirectional press to form a preliminary
molding, vacuum sintering the preliminary molding under no pressure
to form a billet, and heating the billet using an induction heating
unit to extrude the square pipe.
[0016] The method of manufacturing a radioactive-substance storage
member according to still another aspect of the present invention
comprises mixing an aluminum powder with a neutron absorber powder,
pressing the mixed powder to form a preliminary molding, sintering
the preliminary molding with a vacuum hot press to form a billet,
heating the billet using an induction heating unit, and extruding
the induction-heated billet using dies to form a square pipe or
plate which configures a basket for accommodating an aggregate of
spent fuel or a rod to be inserted in a guide pipe for spent
fuel.
[0017] Other objects and features of this invention will become
apparent from the following description with reference to the
accompanying drawings.
BRIEF DESCRIPTION OF THE DRAWINGS
[0018] FIG. 1 is a cross-sectional view which shows a square
pipe,
[0019] FIG. 2 is a flowchart which shows a method of manufacturing
a square pipe according to Embodiment 1 of the present
invention,
[0020] FIG. 3 shows a configuration of dies in a porthole extruder
in an axial cross-sectional view (a) and a radial cross-sectional
view (b),
[0021] FIG. 4 shows an outlined configuration of a powder
manufacturing apparatus which performs a manufacturing method
according to Embodiment 2 of the present invention,
[0022] FIG. 5 is a perspective view which shows a rack of flat
plate type,
[0023] FIG. 6 shows an aggregate of spent fuel,
[0024] FIG. 7 is a perspective view which shows a cask,
[0025] FIG. 8 is a radial cross-sectional view of the cask shown in
FIG. 7,
[0026] FIG. 9 is an axial cross-sectional view of the cask shown in
FIG. 7,
[0027] FIG. 10 is a perspective view which shows a method of
inserting square pipes, and
[0028] FIG. 11 is a perspective view which shows a spent fuel pool
for PWR.
DETAILED DESCRIPTIONS
[0029] Embodiments of methods of manufacturing a
radioactive-substance storage member and billets for use in
extrusion of the member according to the present invention will be
described below with respect to embodiments in detail with
reference to the drawings. These embodiments are, however, not
intended to limit the scope of the invention.
[0030] A first embodiment will be explained below. FIG. 1 shows a
square pipe in a cross-sectional view. This square pipe 1 has a
square section and is composed of an aluminum complex or an
aluminum alloy including an Al or Al-alloy powder, to which a B or
B-compound powder having a neutron absorbing power is added.
Available neutron absorbers other than boron include cadmium,
hafnium and rare earth elements that have large cross-sectional
areas for neutron absorption. The B or B-compound is mainly
employed in the case of a boiling water reactor (BWR), while an
alloy of Ag--In--Cd is employed in the case of a press water
reactor (PWR). If B is employed as a dispersive medium, 7 wt. % or
less is preferable for easy processing. A composition of the
Ag--In--Cd alloy contains 15 wt. % of In and 5 wt. % of Cd in
general. Available rare earth elements include europium,
dysprosium, samarium and gadolinium in their oxidized form.
[0031] An example of a method of manufacturing the square pipe 1 is
described next specifically. FIG. 2 is a flowchart which shows a
method of manufacturing a square pipe according to Embodiment 1 of
the present invention. A rapid solidification processing such as an
atomization method is employed to produce an Al or Al-alloy powder
(Step S201) and prepare a B or B-compound powder (Step S202). These
powders are mixed for 10 to 20 minutes in a cross rotary mixer, a
V-mixer, a ribbon mixer or a pug mixer (Step S203) This mixing may
be performed in an argon ambient. The aluminum powder used has an
average particle diameter of about 35 .mu.m and B4C an average
particle diameter of about 10 .mu.m.
[0032] The Al or Al-alloy powder available as a base can be
selected among pure aluminum metals (JIS 1xxx series), Al--Cu
series aluminum alloys (JIS 2xxx series), Al--Mg series aluminum
alloys (JIS 5xxx series), Al--Mg--Si series aluminum alloys (JIS
6xxx series), Al--Zn--Mg series aluminum alloys (JIS 7xxx series),
and Al--Fe series aluminum alloys (1 to 10% by weight of Fe
content) as well as Al--Mn series aluminum alloys (JIS 3xxx
series). These can be selectively employed in accordance with
necessary properties such as strength, stretchability,
processability and heat resistance without particular
limitations.
[0033] B4C and B2O3 can be employed as the above B or B-compound.
Preferably, an amount of boron added to aluminum is more than 1.5
wt. % but less than 9 wt. %. An additional amount below 1.5 wt. %
cannot achieve a sufficient neutron absorbing power and an
additional amount above 9 wt. % lowers a stretch against
tensile.
[0034] Preferably, the mixed powder is contained and sealed with in
a rubber case, then high-pressure is applied uniformly from all
directions at normal temperature using CIP (Cold Isostatic Press)
to mold the powder (Step S204). A molding condition for CIP
includes a molding pressure of 1000 to 2000 kg/cm.sup.2. The CIP
process allows a powdery body to lose its volume by about 20% and
the preliminary molding to have a diameter of 600 mm and a length
of 1500 mm. The uniform pressurization from all directions by CIP
allows high dense moldings to be obtained with less variation in
molding densities. In the CIP process the preliminary molding is
managed to have a weight density of 75 to 95%.
[0035] As a preferred pressurization technology, instead of the
above CIP, a unidirectional high-pressure press can be employed to
form a preliminary molding. Specifically, the mixed powder is led
into a mold that is set in a press machine to form a preliminary
molding under a high molding pressure of from 5000 tons to 10000
tons. The use of such the extremely high-pressure for press is
effective to achieve uniform molding densities in the preliminary
molding. Preferably, the extent of the uniformity of molding
densities is substantially equal to that obtained in the above CIP
process. In this case, a target molding density may be employed as
the reference to determine the molding pressure. The mixed powder
is not required to be contained in a rubber case, which is
vaccumized, but in a mold to be pressed. Therefore, a relatively
easy preliminary molding operation can be performed compared to
CIP. The methods of producing preliminary moldings are not limited
in the CIP and the unidirectional press.
[0036] The preliminary molding is led into a sintering furnace and
sintered under no pressure after the furnace is vaccumized (Step
S205). During the vacuum sintering, a vacuum is kept about
10.sup.-1 Torr and a temperature set at 550 to 600.degree. C. A
time period for retaining the sintering temperature can be
appropriately set between 5 hours and 10 hours. The sintering
temperature is elevated step by step at a pitch of 100.degree. C. A
graphite heater provided in the sintering furnace is employed for
heating. The vacuum sintering can fuse temporally fixed powdery
particles with each other to form a neck, which turns into a billet
for use in extrusion. Instead of vacuumizing the sintering furnace,
an inert gas such as an argon gas and a helium gas may be filled
for sintering under on pressure. The conditions including the
sintering temperature and time are similar to those described
above.
[0037] During the sintering, as any pressurization technology such
as HIP and hot press is not employed in the above method, the
sintered body has a weight density rarely varied from that at the
time of preliminary molding, maintaining 75 to 95%. The vacuum
sintering can prevent billets from oxidizing and omit canning.
Therefore, it is possible to save spending on cans, eliminate the
need for steps of cutting the outer and end surfaces to remove the
cans, and additionally omit the accompanying process steps of
canning and so forth.
[0038] A porthole extruder is employed to hot extrude the billet
(Step S206). The extrusion condition in this case includes a
heating temperature of 500 to 520.degree. C. and an extrusion speed
of 5 m/min. This condition can be changed appropriately in
accordance with the content of B. The porthole extruder has an
extrusion force of 5000 to 6000 tons. FIG. 3 shows a configuration
of dies in the porthole extruder in an axial cross-sectional view
(a) and a radial cross-sectional view (b). This porthole extruder
300 comprises dies 301 and a container 302. A RF coil 303 is
located around the container 302 for induction heating. When a RF
current flows into the RF coil 303, a billet B in the dies 301 can
be inductively heated. The dies 301 include a female type 304 and a
male type 305. A mandrel 306 of the male die 305 is inserted into
the female die 304 to shape a bearing 307 in the extruded form of a
square. The mandrel 306 is supported by four bridges 308 extending
from the periphery of the male die 305 to form four ports 309 among
the bridges 308.
[0039] The induction heating generates an induced current in the
billet B to heat it. A heating target or the billet B is in a state
of the mixed powder fused in the step of vacuum sintering.
Accordingly, the induced current can be generated over the whole
billet B for efficient heating. Two preliminary moldings were
produced by CIP as practical samples with a weight of 2510 g,
.phi.-size of 89 mm.times.175 mm, volume of 1100 mm.sup.3 and
relative density of 85%. Both were compared with each other after
only one of them was sintered in vacuum. As a result of the
comparison, the sample compacted only by CIP has an electric
conductivity of 7% while the vacuum-sintered sample exhibits that
of 37%, which is more than five times the former.
[0040] When the samples were inductively heated, the
vacuum-sintered sample exhibited a temperature elevation as the
temperature elevation program for induction heating defines (to
elevate a temperature up to 520.degree. C. at 200.degree. C./min
and then retain it for a certain time period). It was found that
less variation was present in the sample at the edges and the
surface and internal center of the mid-portion, and that
temperatures were elevated almost uniformly at any locations. On
the other hand, the sample compacted only by CIP could not allow a
temperature to elevate as the temperature elevation program,
resulting in a temperature elevation rate of 50.degree. C./min at
most. This demonstrates that improvement of the electric
conductivity is associated with the time period for induction
heating during extrusion, and that the application of vacuum
sintering as is in the present invention allows the temperature
elevation to follow the temperature elevation program. It can be
finally concluded that the vacuum sintering extremely increases the
efficiency of the induction heating and improves the extrusion
speed of the billet advantageously.
[0041] The billet B inductively heated in the container is then
pushed from behind by a punch and split into four at the bridges
308 to pass through the ports 309. Consequently, they are
integrated together during passage from the ports 309 to the
bearings 307, which extrude the square pipe 1 in the form of a
certain extrusion shape. The weight density of the billet B is 75
to 95% while that of the square pipe 1 becomes almost 100% because
air gaps between the powdery particles are crushed at the time of
extrusion.
[0042] After the extrusion, tensile reformation is performed (Step
S207) and unsteady parts and evaluation parts are cut away to
complete products (Step S208). The completed square pipe 1 has a
square shape in section with an outer side of 162 mm and an inner
side of 151 mm as shown in FIG. 1. The above process steps are
effective when the steps of billet molding and extrusion are
performed at different places or different timings.
[0043] If the steps of vacuum sintering and extrusion are performed
closely in time, as is in a process line including a vacuum
sintering line and an extrusion line successively, a temperature is
elevated up to 550 to 600.degree. C. during vacuum sintering.
Therefore, after completion of the sintering, the billet may be
inserted into the container and extruded directly within a thermal
region at least 500.degree. C. that is the extrusion temperature.
Specifically, the billet is removed out of the vacuum furnace and,
before the temperature of the billet lowers, conveyed to the
extruder, which extrudes the square pipe 1. If the heated billet is
exposed to the atmosphere but only for a short time, the
oxidization effect can be almost neglected and the performance of
the square pipe 1 is hardly affected. Preferably, the billet is
extruded within 15 minutes after removed from the vacuum furnace.
In this case, the oxidization effect is rarely problematic. As
obvious from the above procedures, the induction heating eliminates
the need for re-heating of the billet and further simplifies the
process steps.
[0044] Also in this case, the vacuum sintering can prevent billets
from oxidizing and omit canning. Therefore, it is possible to save
spending on cans, eliminate the need for steps of cutting to remove
the cans, and additionally omit the accompanying process steps of
canning and so forth. Temporally stored for a short time in a
heat-insulated chamber that can keep the temperature during vacuum
sintering, the billet may be transferred into the container in the
extruder within a thermal region more than 500.degree. C. In this
case, the vacuum sintering line is not required to follow the
extrusion line and both may be spaced apart from each other without
any problems. If a distance between the vacuum sintering line and
the extrusion line is small and a time required for conveying the
billet is short, the heat from vacuum heating can be employed for
the extrusion similar to the above, needless to say. Further, hot
working such as forging may be carried out before the extrusion, in
order to increase a sintered density of the billet.
[0045] The extruder employed in the above example is of the
porthole type because it has a high compressibility and is suitable
for extruding a complicated shape composed of a soft material such
as aluminum. Though, the extruder is not limited in this type. For
example, a stationary or movable mandrel type may also be employed.
Besides the direct extrusion, isostatic press extrusion may be
applied. These can be selected appropriately within a possible
range that an ordinary skilled person in the art can consider. Even
though a batch process has a relatively low yield, it may be
applied to batch billets in the heating furnace instead of the
induction heating.
[0046] A second embodiment will be explained below. In this
embodiment, boron-doped aluminum alloy is employed as the material
to construct the square pipe 1 in the above. If the doped element,
B4C, has a large average particle diameter, it reduces the strength
of the square pipe 1. In contrast, If B4C has a small average
particle diameter, B4C particles aggregate together and
precipitate, resulting in a lowered neutron absorbing power and a
worsened processability. As described above, the average particle
diameter is determined 80 .mu.m for the Al powder and 9 .mu.m for
the B4C powder. The reason for the determination of 9 .mu.m for the
particle diameter of the B4C is that a smaller particle diameter
than that promotes the aggregation of the B4C powder and causes the
precipitation easily. In this Embodiment 2, instead of the mixer in
the Embodiment 1, high-energy ball milling (mechanical alloying) is
employed to achieve a fine and uniformly dispersed B4C powder.
[0047] In the high-energy ball milling, a tumbling mill, a rocking
mill and an attritor mill can be employed in general. The attritor
mill is exemplified below. FIG. 4 shows a configuration of the
attritor mill for use in a method of manufacturing a square pipe
according to Embodiment 2. The attritor mill 30 comprises a
container 31 with a volume of 150 liters. A water jacket 32 is
formed inside the wall plate of the container 31. A water supplier
33 such as a pump is employed to supply an appropriate amount of
cooling water into the water jacket 32. An attritor 34 is coupled
via a decelerator 36 to a drive motor 35 located above. In the
upper surface of the container 31, an inlet 37 and an outlet 38 are
provided to create an ambient of an inert gas or argon (Ar) within
the container 31. The inlet 37 is connected to a gas cylinder 39 of
argon gas and the outlet 38 is connected to a hose 40, which is led
into water to prevent reverse flow of the atmosphere.
Carbon-steel-based bearing steel (SUJ-2) or a ceramic ball is
employed for a ball 41 for use in this ball milling.
[0048] A condition was determined for producing a high-energy
powder in practice such that an amount of the balls 41 to be
contained in the container 31 is equal to 450 kg and a diameter of
the ball 41 equal to 3/8 inch. The number of revolutions of the
attritor 34 is determined 300 rpm and argon is continuously flowed
at 0.5 liter/min into the container 31 to create an inert gas
ambience inside. Prior to the ball milling, 10 to 50 cc of ethanol
or methanol relative to 1 kg of powder is deposited as an auxiliary
agent. An amount of powder to be deposited in the container 31 is
determined 15 kg, which includes 0.75 kg of B4C (5% by weight). In
practical use, Al powder has an average particle diameter of 35
.mu.m and B4C powder an average particle diameter of 9 .mu.m. A
time for ball milling can be selected appropriately from a range
between 1 hr and 10 hrs.
[0049] In the process of ball milling, the deposited aluminum
suffers impacts from the balls 41 and is crushed, folded and
flattened. As a result, an outer diameter of the aluminum is
widened in plane to about 80 .mu.m. In contrast, the B4C powder is
pulverized during ball milling, down-sized to have a particle
diameter of 0.5 to 1.0 .mu.m and uniformly mixed into the aluminum
matrix.
[0050] In the process of ball milling, the balls 41 collide with
each other and the components wore of the balls 41 may often be
mixed as impurities. Accordingly, if an element to be added as an
impurity is previously contained in the ball 41 as a component, the
element can be added in the process of ball milling. An example of
such the element is zirconia, alumina, or the like. After the
termination of ball milling, the high-energy powder is removed from
the container 31 and subjected to the hot press step and then the
extrusion step to form the square pipe 1 as shown in FIG. 1.
[0051] According to the method of manufacturing the square pipe 1,
the fine and uniform B4C powder can be dispersed in the matrix of
Al powder, resulting in an improved strength imparted to the square
pipe 1. Specifically, in comparison with the square pipe 1 obtained
by the method of Embodiment 1, the strength can be improved up to
about 1.2 to 1.5 times. In particular, it is useful as a square
pipe for a cask in PWR that has a large weight of the aggregate of
spent fuel. The fine and uniform dispersion of the highly hard B4C
powder in the matrix can prevent aggregation of the B4C powder and
improve the extrusion ability. It is also effective to reduce wear
of the dies for extrusion.
[0052] During ball milling, an organic solvent such as an alcohol
may be deposited to produce a compound of the organic solvent and
aluminum, which is added effectively to improve the strength and
stretchability of the square pipe 1.
[0053] A third embodiment will be explained below. In this
embodiment, a rack for accommodating the aggregate of spent fuel
may be of a flat plate type instead of the square pipe type. FIG. 5
shows a flat plate rack in a perspective view. This flat plate rack
60 employs the billets produced by the manufacturing method of
Embodiment 1 or 2, which are extruded to form flat members 61 each
having a width of 300 to 350 mm. Plural slits 62 are formed
successively in each of the flat members 61. The flat members 61
are engaged with each other in the longitudinal and lateral
directions at the slits 62 to form a grid-like cross section. As
the flat plate rack 60 has a smaller plate thickness than that of
the square pipe type, a larger amount of boron is dispersed in
aluminum. The flat plate rack 60 can be employed as a cask and a
rack for spent fuel pool.
[0054] The manufacturing method of the first or third embodiments
can also produce a plate, not depicted, which can be employed not
only in the flat plate rack 60 but also in a structural material
for a transportation container for low level wastes.
[0055] A fourth embodiment will be explained below. In the first
and second embodiments, the dispersion of boron in the square pipe
1 imparts the neutron absorbing power and prevents the spent fuel
aggregate from reaching the criticality. In the fourth embodiment,
instead of the square pipe 1, a round rod is molded to absorb
neutrons from the spent fuel aggregate. As show in FIG. 6, a round
rod 70 is inserted into a guide tube 72 for a cluster of control
rods (or a measurement tube) in a spent fuel aggregate 71. The
insertion of the round rod 70 ensures a certain neutron absorbing
power and eliminates the need for dispersion of a large amount of
boron in the square pipe 1. The manufacturing method of Embodiment
1 or 2 can be employed in production of the round rod 70 because
only shapes of dies in final extrusion steps differ from each
other.
[0056] A fifth embodiment will be explained below. As the fifth
embodiment, preferably, an additional element such as Zr and Ti is
added in the aluminum powder to impart a high strength. The content
of Zr in this case is determined more than 0.2 wt. % and less than
2.0 wt. %, more preferably more than 0.5 wt. % and less than 0.8
wt. %. The content of Ti is determined more than 0.2 wt. % and less
than 4.0 wt. %. Both Zr and Ti can be added together. The addition
of Zr or Ti improves mechanical properties such as a tensile
property and achieves a high processability. Accordingly, this
aluminum composite material can be employed as a structural
material for use in a storage or transportation container for spent
nuclear fuel and a structural member for nuclear-related
facilities.
[0057] The following reason is given to the additional amount of Zr
or Ti that should be kept within the range described above. In the
case of Zr, the content below 0.2 wt. % reduces an effect of
strength improvement and, in contrast, the content above 2.0 wt. %
reduces stretchability and tenacity and saturates the effect of
strength improvement. In the case of Ti, the content below 0.2 wt.
% causes an insufficient effect of strength improvement and, the
content above 4.0 wt. % makes it difficult to form a fine
intermetallic compound and has a trend to easily reduce tenacity
and saturate the effect of strength improvement. As for the Zr or
Zr-compound, a sponge form one is employed, for example, as well as
for the Ti or Ti-compound.
[0058] The moldings after the extrusion may be subjected to
heating, if required. For example, if an Al-alloy powder in
Al--Mg--Si series is employed as a base, T6 processing of JIS is
performed. If an Al-alloy powder in Al--Cu series is employed as a
base, T6 processing of JIS is also performed. If a powder of pure.
Al or of an Al-alloy in Al--Fe series is employed as a base,
heating is not required. This case corresponds to T1 or H112
processing of JIS.
[0059] In particular, T1 or H112 processing is performed on the
molding that contains the additional element for imparting the high
strength to obtain a material suitable for a structural material
for use in a storage or transportation container for the spent
nuclear fuel. As a result, the material obtained is stabile at high
temperature over a long period of time. In contrast, if artificial
aging such as T6 processing is performed on the molding, decrease
in strength due to over-aging may occur, and it becomes difficult
to obtain the material that is resistant to high temperature over a
long period of time.
[0060] The aluminum composite material according to this embodiment
is suitable for a structural material of a container for use in
storage or transportation of the spent nuclear fuel. The reason is
given below. When the spent nuclear fuel is stored or transported,
it is required to accommodate the spent nuclear fuel in a container
that has a function of shielding radioactivity from the spent
nuclear fuel. The container is assumed to fall to the ground in a
storage facility during storage due to some causes or drop down
from a carrier vehicle during transportation of the container. In
these situations, if the structural material of the storage or
transportation container is damaged, neutrons from the accommodated
spent nuclear fuel can not be absorbed. This leads to a possibility
to cause an accident, which can not be denied. Specifically, if a
basket for a cask is constructed as disclosed below, the basket
includes grid-like cells for accommodating the aggregates of the
spent nuclear fuel inserted, which has a weight of 150 kg or more
per aggregate in the case of BWR fuel. If these aggregates of the
spent nuclear fuel are accommodated 50 or more, the dropping or
falling of the cask imparts a considerably large impact on the
basket. Therefore, the basket is required to have a mechanical
property sufficiently durable against such the impact.
[0061] The basket is also required to have a power to absorb
neutrons radiated from the spent nuclear fuel during accommodation
of that spent nuclear fuel. A generally known neutron absorbing
material includes B or B-compound, which is mixed in an aluminum
matrix and employed as a neutron shielding material in common. If
the aluminum material that contains such B or B-compound is
employed to construct the basket for the cask, the basket itself is
required to have both a neutron absorbing power and a mechanical
strength as a structural material.
[0062] The storage period for the spent nuclear fuel accommodated
in the container is as extremely long as about 60 years. A
temperature elevates high (specifically, 150.degree. C. or more)
within the container for a long time due to the decay heat of the
spent nuclear fuel and the aluminum composite material inside is
exposed to a high-temperature environment for a long period. In
this case, over-aging of the structural material lowers its
mechanical strength with time. Therefore, the structural material
may fall to the ground during storage or drop down during
transportation 30 or 40 years later, for example. If such a
situation occurs, the structural material can not suffer the impact
and get damaged, resulting in a possible critical accident and the
like. Accordingly, the container for use in storage or
transportation of the spent nuclear fuel is required to use such a
material that rarely varies its mechanical property and can retain
original properties even under a high-temperature environment for a
long time.
[0063] For the major purpose to provide a mechanical property that
is not changed through a long-term use under a high-temperature
environment, the aluminum composite material according to this
embodiment includes a material composed of B or B-compound having a
neutron absorbing power contained in a matrix of Al or Al-alloy.
The material is sintered under pressure without artificially aging
or by naturally aging (including T1 processing) the sintered
material, or leave the sintered material in its extruded state. For
instance, a billet sintered under pressure is extruded and then
left in that extruded state or naturally aged to produce a square
pipe for use in constructing the basket in the cask as disclosed
below. In particular, the natural aging makes a hardly observable
reduction in the mechanical strength even after the long-term
exposure to the high-temperature environment while the original
strength is low, because unlike artificial aging, the natural aging
does not cause over-aging.
[0064] The mechanical strength may possibly be insufficient if the
structural material is only naturally aged. To raise the base in
the strength of the aluminum composite material, an additional
element is contained to impart a property of high strength. This
enables such a material to be realized for the first time, that has
a sufficient mechanical strength and a hardly observable reduction
in the mechanical strength even after the long-term exposure to the
high-temperature environment. Such the aluminum composite material
is extremely suitable for use in a structural material of a
container for storage or transportation of the spent nuclear
fuel.
[0065] The aluminum composite material was tested on its
properties. In this test, certain test pieces were retained for 100
hours at 180.degree. C., for 100 hours at 200.degree. C. and for
100 hours at 350.degree. C. to perform tensile tests on 0.2%
durability (MPa), tensile strength (MPa) and shear stretch (%). The
two formers were tested assuming the case immediately after
accommodation of the spent nuclear fuel and the latter was tested
assuming the case after 60-year accommodation of the spent nuclear
fuel. As for the latter, a 60-year test can not be performed in
practice. Instead, a test was performed at temperatures accelerated
up to 350.degree. C., which corresponds to a long-term retention of
60 years under a high-temperature environment of 200.degree. C. As
a result, in comparison of the case immediately after accommodation
of the spent nuclear fuel to the case after 60-year accommodation,
any variation was hardly observable in either of 0.2% durability
(MPa), tensile strength (MPa) and shear stretch (%). Thus, it was
found that the mechanical properties were unchanged.
[0066] In contrast, a T6-processed test piece and a test piece that
has been T1-processed (naturally aged) or left in its extruded
state after extruding were prepared and tested. As a result, it was
determined that the T6-processed test piece was easily affected
from a temperature elevation. This means that if the T6-processed
composite material is employed in the long-term use under the
high-temperature environment, the temperature elevation inside the
container affects on the structural material and degrades the
mechanical properties thereof.
[0067] If the aluminum composite material is not heated, it is
difficult to obtain a fine crystal grain and accordingly improve
the mechanical strength. The use of the aluminum composite material
as a structural material, however, requires a mechanical strength
to some extent. Therefore, the aluminum composite material of this
embodiment contains an additional element for imparting high
strength to compensate a lack in the mechanical strength. The
additional element herein described is added to compensate a lack
in the mechanical strength that is caused when no heating is
performed. This has a different purpose from that of an element
adding operation that is normally performed to pure aluminum.
[0068] Tensile tests were performed on the aluminum composite
material that contains the additional element in the case of the
use in a room-temperature environment and in the case corresponding
to the 60-year use under the high-temperature environment (see the
above description). As a result, improvements were observed in the
0.2% durability and the tensile strength while the shear stretch
lowers slightly. The numeral values each hardly vary from the case
corresponding to the 60-year storage, from which it was found that
the mechanical properties could be maintained.
[0069] In consideration of the above, the aluminum composite
material that is subjected to T1-processing (or natural aging that
includes a certain use immediately after the extrusion that
substantially effects as natural aging) is suitable for a
structural material of a storage or transportation container.
Because such the aluminum composite material has a mechanical
property that is not degraded even if it is exposed under a
long-term, high-temperature environment and also has a mechanical
strength suitable for a structural material.
[0070] Examples of the aluminum composite material include aluminum
in 6000 series. An aluminum-alloy in 6000 series is commonly known
as a heated (T6-processed) alloy. Through natural aging by leaving
the alloy in its extruded state after extruding, instead of
positive heating, it can be employed as a structural material
suitable for a storage or transportation container for the spent
nuclear fuel. In addition, non-heated alloy (for example of 3000 or
5000 series) can also be similarly employed.
[0071] A sixth embodiment will be explained below. In the sixth
embodiment, the square pipe 1 can be employed in a specific use,
which is exemplified next. FIG. 7 is a perspective view which shows
a cask. FIG. 8 is a radial cross-sectional view of the cask shown
in FIG. 7. FIG. 9 is an axial cross-sectional view of the cask
shown in FIG. 7. This cask 100 comprises a cylinder body 101 having
a cavity 102, of which inner surface is machined corresponding to
the outer shape of a basket 130.
[0072] The cylinder body 101 and a bottom plate 104 are castings of
carbon steel having a .gamma.-ray shield function. Stainless steel
may also be employed instead of carbon steel. The cylinder body 101
and the bottom plate 104 are coupled by welding. A metallic gasket
is provided between a primary lid 110 and the cylinder body 101 to
ensure the hermeticity as a pressure-proof container.
[0073] A resin 106 is filled between the cylinder body 101 and an
outer cylinder 105. The resin is a polymeric material that contains
a large amount of hydrogen and has a neutron shield function. A
plurality of inner fins 107 composed of copper is welded for
thermal conduction between the cylinder body 101 and the outer
cylinder 105. The resin 106 is injected in a fluidized state into
spaces formed by the inner fins 107, then cooled and solidified.
Preferably, the inner fins 107 are located at a high density on
parts with a large amount of heat to uniformly dissipate heat. A
thermal expansion margin 108 of several mm is provided between the
resin 106 and the outer cylinder 105.
[0074] A lid section 109 comprises the primary lid 110 and a
secondary lid 111. The primary lid 110 is a disk composed of
stainless steel or carbon steel that shields y-ray. The secondary
lid 111 is also a disk composed of stainless steel or carbon steel.
A resin 112 is filled as a neutron shield over the upper surface of
the secondary lid 111. The primary lid 110 and the secondary lid
111 are attached to the cylinder body 101 by a bolt 113 composed of
stainless steel or carbon steel. Between the primary 110 and
secondary 111 lids and the cylinder body 101, metallic gaskets are
provided, respectively, to keep the inner hermeticity. An auxiliary
shield 115 filled with a resin 114 is arranged around the lid
section 109.
[0075] At both sides of a cask body 116, trunnions 117 are provided
to suspend the cask 100. During transportation of the cask 100, the
auxiliary shield 115 is removed and a buffer 118 is attached
instead. The buffer 118 has such a structure that includes an outer
cylinder 120 made of stainless steel and a buffering material 119
composed of redwood and the like incorporated in the outer cylinder
120.
[0076] A basket 130 consists of 69 square pipes 1 that construct a
cell 131 for accommodating an aggregate of spent nuclear fuel. The
square pipes 1 employed are those manufactured by the manufacturing
method according to the above embodiments. FIG. 10 shows a method
of inserting the square pipes in a perspective view. The square
pipes 1 manufactured by the above steps are inserted in turn along
the machined shape inside the cavity 102.
[0077] As shown in FIGS. 10 and 8, at both sides of an array of
square pipes in the cavity 102 where the number of cells is equal
to 5 or 7, a dummy pipe 133 is inserted, respectively. The dummy
pipe 133 is employed for the purpose of reducing a weight of the
cylinder body 101 and uniforming a thickness of the cylinder body
101 as well as ensuring steadiness of the square pipes 1. The above
steps can be similarly applied to produce the dummy pipe 133 using
a boron-doped aluminum alloy. The dummy pipe 133, however, can be
omitted.
[0078] The spent nuclear fuel aggregate to be accommodated in the
cask 100 contains fissionable substances and fission products and
radiates radioactive rays along with decay heat. Therefore, the
cask 100 is required to keep a heat removal function, a shield
function and a criticality preventive function reliably during
storage (about 60 years). In the cask 100 according to this
embodiment, the cavity 102 in the cylinder body 101 is machined so
that the basket 130 configured from the square pipes 1 can be
inserted into the cavity 102, holding the outer side in a
substantially intimate contact state (without any large gaps). In
addition, the inner fins 107 are provided between the cylinder body
101 and the outer cylinder 105. Accordingly, heat from the fuel
rods is conducted through the square pipes 1 or the filled helium
gas to the cylinder body 101 and dissipated from the outer cylinder
105 mainly through the inner fins 107.
[0079] .gamma.-ray radiated from the spent nuclear fuel aggregate
can be shielded, for example, at the cylinder body 101, outer
cylinder 105 and lid section 109 each composed of carbon steel or
stainless steel. Neutrons can be shielded at the resin 106 to
eliminate the exposure influencing on radioactive-related workers.
Specifically, a shield function is designed to achieve a surface
dose equivalent ratio of 2 mSv/h or less and a dose equivalent
ratio of 100 .mu.Sv/h or less at a depth of 1 m below the surface.
As the boron-doped aluminum alloy is employed for the square pipes
1 to construct the cell 131, it is possible to absorb neutrons so
as not to reach criticality.
[0080] The inside of the cavity 102 in the cylinder body 101 is
machined such that the square pipes 1 for use in construction of
the outer circumference of the basket 130 can be inserted into the
cavity 102 in a substantially intimate contact state. Therefore,
according to the cask 100, an area of the square pipe facing to the
cavity can be widened and thermal conduction from the square pipe 1
can be improved. It is also possible to eliminate the space region
in the cavity 102 and accordingly realize a compact and light
cylinder body 101 if the number of the accommodated square pipes 1
is identical. In contrast, if the outer diameter of the cylinder
body 101 is held unchanged, the number of cells can be ensured to
the extent. Accordingly, it is possible to increase the number of
the accommodated aggregates of the spent nuclear fuel.
Specifically, the cask 100 is possible to accommodate the spent
nuclear fuel as large as 69 aggregates while the cask body 116 is
down-sized to have an outer diameter of 2560 mm and a weight of 120
tons.
[0081] In the sixth embodiment, the square pipes 1 are bundled to
construct the basket 130. Alternatively, the square pipes 1 may be
arranged in a stagger or checker pattern to construct the basket.
Preferably, edges of the square pipes 1 are provided with securing
structures for mating with each other.
[0082] In the cylinder body 101 the cask of the sixth embodiment
includes a configuration for housing a canister in the form of a
thin can (not depicted). The canister is employed to seal various
radioactive substances such as the above spent nuclear fuel
aggregates and nuclear wastes solidified in glass. The canister is
accommodated in the cylinder body 101 and transported or managed
per cask.
[0083] A seventh embodiment will be explained below. The seventh
embodiment relates to another usage example of the square pipe.
FIG. 11 shows a spent fuel pool for PWR in a perspective view. This
spent fuel pool 200 comprises a rack 202, which includes a
plurality of standing square pipes 1 manufactured by the
manufacturing method according to Embodiment 1 or 2, and support
plates 201 that support the upper and lower portions of the square
pipes 1. The rack 202 is located inside a pit 203 composed of
reinforced concrete. A lining 204 composed of stainless steel is
stuck on the inner surface of the pit 203 to prevent the pit water
from leaking.
[0084] The pit 203 is generally filled with an aqueous boric acid.
The spent fuel pool 200, as it is configured with the use of the
square pipes 1, has a high neutron absorbing power and can ensure
the soundness of the structure. Therefore, it is possible to
effectively prevent the spent nuclear fuel aggregate from reaching
criticality. Replacement of square pipes of B-SUS with those of
B--Al leads to a greatly reduced, abnormal load in the fuel pool
and to an improved security in the storage facility.
[0085] In the case of the use as the rack 202 in the fuel pool, the
B-SUS square pipe currently employed has a pH of about 3.0 while
the B--Al square pipe 1 has a pH of about -4.5. The aqueous boric
acid is acidic and thus causes crevice corrosion possibly.
Therefore, it is preferable to make an aqueous environment in the
fuel pool milder.
[0086] As described above, the method of manufacturing a
radioactive-substance storage member according to the present
invention includes the steps of, mixing an aluminum powder with a
neutron absorber powder, pressing the mixed powder to form a
preliminary molding, and sintering the preliminary molding under no
pressure in vacuum or in an inert gas ambience. This method is
effective to simplify the manufacture of the radioactive-substance
storage member.
[0087] The method of manufacturing a radioactive-substance storage
member according to the present invention also includes the steps
of, mixing an aluminum powder with a neutron absorber powder,
pressing the mixed powder to form a preliminary molding, sintering
the preliminary molding under no pressure in vacuum or in an inert
gas ambience to form a billet, heating the billet using an
induction heating unit, and extruding the induction-heated billet
using dies to form a square pipe or plate for configuring a basket
for accommodating an aggregate of spent fuel or a rod to be
inserted in a guide pipe for spent fuel. In the square pipe for
accommodating an aggregate of spent fuel, the square pipe is formed
by mixing an aluminum powder with a neutron absorber powder through
cold isostatic press or cold unidirectional press to form a
preliminary molding, vacuum sintering the preliminary molding under
no pressure to form a billet, and heating the billet using an
induction heating unit to extrude the square pipe. This is
effective to achieve efficient induction heating, a simplified
manufacturing process and an efficient extrusion of billets.
[0088] The method of manufacturing a radioactive-substance storage
member according to the present invention further includes the
steps of, mixing an aluminum powder with a neutron absorber powder,
pressing the mixed powder through cold isostatic press or cold
unidirectional press to form a preliminary molding, sintering the
preliminary molding under no pressure in vacuum to form a billet,
and extruding the billet using the heat during the sintering to
form a square pipe or plate for configuring a basket for
accommodating an aggregate of spent fuel or a rod to be inserted in
a guide pipe for spent fuel. This is effective to eliminate the
need for re-heating of the billet and simplifies the manufacturing
process. In the method of manufacturing a radioactive-substance
storage member according to the present invention, the preliminary
molding may be determined to have a weight density of more than 75%
and less than 95% to perform sound extrusion.
[0089] In the method of manufacturing a radioactive-substance
storage member according to the present invention, mechanical
alloying may be employed to mix the aluminum powder with the
neutron absorber powder. This allows the neutron absorber powder in
a fine state to be uniformly dispersed in the aluminum matrix,
resulting in an improved mechanical strength of the storage member
for the spent nuclear fuel.
[0090] The method of manufacturing a radioactive-substance storage
member according to the present invention, a ball used in ball
milling has a major constituent identical to an element that is
expected to add. Therefore, the additional element can be added to
a sample during the ball milling. This is effective to simplify the
manufacturing process.
[0091] The method of manufacturing a radioactive-substance storage
member according to the present invention, further includes, after
the extruding, the step of naturally aging the square pipe or plate
for configuring a basket for accommodating an aggregate of spent
fuel or the rod to be inserted in a guide pipe for spent fuel. This
is effective to obtain a suitable radioactive-substance storage
member.
[0092] The billet for use in extrusion of a radioactive-substance
storage member according to the present invention comprises a mixed
powder of an aluminum powder and a neutron absorber powder. The
mixed powder is molded to have a weight density ranging from 75% to
95% and sintered to fuse each powdery particle to another. This
allows the billet to be inductively heated easily and simplifies
the manufacturing process of the radioactive-substance storage
member.
[0093] The billet for use in extrusion of a radioactive-substance
storage member according to the present invention comprises a mixed
powder of an aluminum powder flattened by mechanical alloying and a
pulverized boron or boron-compound folded and dispersed into the
aluminum powder. The mixed powder is molded to have a weight
density ranging from 75% to 95% and sintered to fuse each powdery
particle to another. This is effective to prevent aggregation
during sintering. The radioactive-substance storage member extruded
using the billet has uniform components and improved the mechanical
strength.
[0094] The billet for use in extrusion of a radioactive-substance
storage member according to the present invention is employed as a
structural material of storage or transportation containers for
spent nuclear fuel. The aluminum powder contains an additive
element such as Zr for imparting high strength to improve the
mechanical strength.
[0095] The method of manufacturing a radioactive-substance storage
member for use in storage of a radioactive substance according to
the present invention comprises the steps of mixing an aluminum
powder with a neutron absorber powder, pressing the mixed powder to
form a preliminary molding, sintering the preliminary molding with
a vacuum hot press to form a billet, heating the billet using an
induction heating unit, and extruding the induction-heated billet
using dies to form a square pipe or plate which configures a basket
for accommodating an aggregate of spent fuel or a rod to be
inserted in a guide pipe for spent fuel. As a result, the
manufacturing process of the radioactive-substance storage member
can be simplified, and the billet can be extruded efficiently.
[0096] Although the invention has been described with respect to a
specific embodiment for a complete and clear disclosure, the
appended claims are not to be thus limited but are to be construed
as embodying all modifications and alternative constructions that
may occur to one skilled in the art which fairly fall within the
basic teaching herein set forth.
* * * * *