U.S. patent application number 12/435708 was filed with the patent office on 2010-05-06 for zirconium alloy composition for nuclear fuel cladding tube forming protective oxide film, zirconium alloy nuclear fuel cladding tube manufactured using the composition, and method of manufacturing the zirconium alloy nuclear fuel cladding tube.
This patent application is currently assigned to Korea Atomic Energy Research Institute. Invention is credited to Byoung Kwon Choi, Yong Hwan Jeong, Yang Il Jung, Hyun Gil Kim, Myung Ho Lee, Jeong-Yong Park, Sang Yoon Park.
Application Number | 20100108204 12/435708 |
Document ID | / |
Family ID | 41601897 |
Filed Date | 2010-05-06 |
United States Patent
Application |
20100108204 |
Kind Code |
A1 |
Park; Jeong-Yong ; et
al. |
May 6, 2010 |
ZIRCONIUM ALLOY COMPOSITION FOR NUCLEAR FUEL CLADDING TUBE FORMING
PROTECTIVE OXIDE FILM, ZIRCONIUM ALLOY NUCLEAR FUEL CLADDING TUBE
MANUFACTURED USING THE COMPOSITION, AND METHOD OF MANUFACTURING THE
ZIRCONIUM ALLOY NUCLEAR FUEL CLADDING TUBE
Abstract
Disclosed herein is a zirconium alloy composition for nuclear
fuel cladding tubes, comprising: 1.6.about.2.0 wt % of Nb;
0.05.about.0.14 wt % of Sn; 0.02.about.0.2 wt % of one or more
elements selected from the group consisting of Fe, Cr and Cu;
0.09.about.0.15 wt % of O; 0.008.about.0.012 wt % of Si; and a
balance of Zr, a nuclear fuel cladding tube comprising the
zirconium alloy composition, and a method of manufacturing the
nuclear fuel cladding tube. Since the nuclear fuel cladding tube
made of the zirconium alloy composition can maintain excellent
corrosion resistance by forming a protective oxide film thereon
under the conditions of high-temperature and high-pressure cooling
water and water vapor, it can be usefully used as a nuclear fuel
cladding tube for light water reactors or heavy water reactors,
thus improving the economical efficiency and safety of the use of
nuclear fuel.
Inventors: |
Park; Jeong-Yong; (Daejeon,
KR) ; Jeong; Yong Hwan; (Daejeon, KR) ; Park;
Sang Yoon; (Daejeon, KR) ; Lee; Myung Ho;
(Daejeon, KR) ; Choi; Byoung Kwon; (Daejeon,
KR) ; Kim; Hyun Gil; (Daejeon, KR) ; Jung;
Yang Il; (Daejeon, KR) |
Correspondence
Address: |
INTELLECTUAL PROPERTY GROUP;FREDRIKSON & BYRON, P.A.
200 SOUTH SIXTH STREET, SUITE 4000
MINNEAPOLIS
MN
55402
US
|
Assignee: |
Korea Atomic Energy Research
Institute
Daejeon
KR
Korea Hydro and Nuclear Power Co., Ltd
Seoul
KR
|
Family ID: |
41601897 |
Appl. No.: |
12/435708 |
Filed: |
May 5, 2009 |
Current U.S.
Class: |
148/557 ;
148/421; 420/422; 420/423 |
Current CPC
Class: |
Y02E 30/30 20130101;
C22F 1/186 20130101; C22C 16/00 20130101; Y02E 30/40 20130101; G21C
3/07 20130101; G21C 21/02 20130101 |
Class at
Publication: |
148/557 ;
420/422; 420/423; 148/421 |
International
Class: |
C22F 1/18 20060101
C22F001/18; C22C 16/00 20060101 C22C016/00 |
Foreign Application Data
Date |
Code |
Application Number |
May 9, 2008 |
KR |
10-2008-0043446 |
Claims
1. A zirconium alloy composition for nuclear fuel cladding tubes,
comprising: 1.6.about.2.0 wt % of Nb; 0.05.about.0.14 wt % of Sn;
0.02.about.0.2 wt % of one or two elements selected from the group
consisting of Fe, Cr and Cu; 0.09.about.0.15 wt % of O;
0.008.about.0.012 wt % of Si; and a balance of Zr.
2. The zirconium alloy composition according to claim 1, wherein
the composition comprises: 1.8 wt % of Nb; 0.1 wt % of Sn; 0.05 wt
% of one element selected from the group consisting of Fe, Cr and
Cu; 0.12 wt % of O; 0.01 wt % of Si; and a balance of Zr.
3. The zirconium alloy composition according to claim 1, wherein
the composition comprises: 1.6.about.2.0 wt % of Nb;
0.05.about.0.14 wt % of Sn; 0.02.about.0.2 wt % of Fe;
0.02.about.0.2 wt % of Cr or Cu; 0.09.about.0.15 wt % of O;
0.008.about.0.012 wt % of Si; and a balance of Zr.
4. The zirconium alloy composition according to claim 3, wherein
the composition comprises: 1.8 wt % of Nb; 0.1 wt % of Sn; 0.05 wt
% of Fe; 0.05 wt % of Cr or Cu; 0.12 wt % of O; 0.01 wt % of Si;
and a balance of Zr.
5. The zirconium alloy composition according to claim 3, wherein
the composition comprises: 1.8 wt % of Nb; 0.1 wt % of Sn; 0.1 wt %
of Fe; 0.1 wt % of Cr or Cu; 0.12 wt % of O; 0.01 wt % of Si; and a
balance of Zr.
6. The zirconium alloy composition according to claim 3, wherein
the composition comprises: 1.8 wt % of Nb; 0.1 wt % of Sn; 0.2 wt %
of Fe; 0.1 wt % of Cr or Cu; 0.12 wt % of O; 0.01 wt % of Si; and a
balance of Zr.
7. A nuclear fuel cladding tube comprising the zirconium alloy
composition of claim 1.
8. A method of manufacturing a zirconium alloy nuclear fuel
cladding tube, comprising: 1) vacuum-arc-remelting and cooling a
mixture of the elements constituting the zirconium alloy
composition of claim 1 to form an ingot; 2) forging the ingot at a
temperature of 1000.about.1200.degree. C.; 3)
solution-heat-treating the forged ingot at a temperature of
1000.about.1200.degree. C. for 10.about.40 minutes and then cooling
the solution-heat-treated ingot to a temperature of
300.about.400.degree. C. at a cooling rate of 300.about.400.degree.
C./s; 4) extruding the cooled ingot at a temperature of
600.about.640.degree. C. to form an extruded shell; 5) primarily
heat-treating the extruded shell at a temperature of 570.about.610
for 2.about.4 hours; 6) cold-working the primarily heat-treated
extruded shell 2 times and intermediately heat-treating it
1.about.4 times between the cold working processes at a temperature
of 570.about.610.degree. C. for 3.about.10 hours to prepare a
zirconium alloy nuclear fuel cladding tube; and 7) finally
heat-treating the prepared zirconium alloy nuclear fuel cladding
tube at a temperature of 470.about.580.degree. C. for 1.about.100
hours to manufacture a zirconium alloy nuclear fuel cladding
tube.
9. The method of manufacturing a zirconium alloy nuclear fuel
cladding tube according to claim 8, wherein, in step 1, the mixture
is repetitively vacuum-arc-remelted 3.about.6 times to form an
ingot.
10. The method of manufacturing a zirconium alloy nuclear fuel
cladding tube according to claim 9, wherein the mixture is
vacuum-arc-remelted and then cooled while injecting an inert gas to
form an ingot.
11. The method of manufacturing a zirconium alloy nuclear fuel
cladding tube according to claim 8, wherein, in step 7, the final
heat treatment is performed in a vacuum.
12. The method of manufacturing a zirconium alloy nuclear fuel
cladding tube according to claim 8, wherein, in step 7, the final
heat treatment is performed such that an average particle size of
.beta.-niobium precipitates is controlled to be 70 nm or less.
Description
CROSS-REFERENCES TO RELATED APPLICATION
[0001] This patent application claims the benefit of priority under
35 U.S.C. .sctn.119 of Korean Patent Application No.
10-2008-0043446 filed on May 9, 2008, the contents of which are
incorporated herein by reference.
BACKGROUND OF THE INVENTION
[0002] 1. Field of the Invention
[0003] The present invention relates to a zirconium alloy
composition for nuclear fuel cladding tubes, which can maintain
excellent corrosion resistance for a long period of time by forming
a protective oxide film under the operating environment of a light
water reactor or a heavy water reactor, a zirconium alloy nuclear
fuel cladding tube manufactured using the zirconium alloy
composition, and a method of manufacturing the zirconium alloy
nuclear fuel cladding tube.
[0004] 2. Description of the Related Art
[0005] A nuclear fuel cladding tube used for a nuclear fuel
assembly of a light water reactor or a heavy water reactor is
predominantly manufactured using a zirconium alloy. To date,
zircaloy-2 comprising 1.2.about.1.7 wt % of tin (Sn),
0.07.about.0.2 wt % of iron (Fe), 0.05.about.1.15 wt % of chromium
(Cr), 0.03.about.0.08 wt % of nickel (Ni), 900.about.1500 ppm of
oxygen (O) and a balance of zirconium (Zr), and zircaloy-4
comprising 1.20.about.1.70 wt % of tin (Sn), 0.18.about.0.24 wt %
of iron (Fe), 0.07.about.1.13 wt % of chromium (Cr), 0.07 wt % or
less of nickel (Ni), 900.about.1500 ppm of oxygen (O) and a balance
of zirconium (Zr) have been most widely used in the manufacture of
nuclear fuel cladding tubes.
[0006] However, recently in order to improve the economical
efficiency of a nuclear reactor, a high burn-up operation, in which
the burning period of nuclear fuel increases, has been employed.
Due to the high burn-up operation, the time for reacting nuclear
fuel with high-temperature and high-pressure cooling water and
water vapor has increased. Therefore, when the conventional
Zircaloy-2 or Zircaloy-4 is used as a raw material for a nuclear
fuel cladding tube, there is a problem in that the corrosion
phenomenon of the nuclear fuel becomes serious.
[0007] Therefore, research into developing materials which can be
used to manufacture high burnup nuclear fuel cladding tubes having
excellent corrosion resistance to high-temperature and
high-pressure cooling water and water vapor is being conducted.
Thus, nuclear fuel cladding tubes having better performance than
the conventional nuclear fuel cladding tubes made of zircalloy-2 or
zircalloy-4 are being developed.
[0008] As such, newly-developed zirconium alloy nuclear fuel
cladding tubes are most characterized by the fact that they contain
niobium (Nb) to improve corrosion resistance. However, the change
in the corrosion resistance of the Nb-containing zirconium alloy
nuclear fuel cladding tubes is very sensitive to the kind and
amount of added elements, the size and distribution of precipitates
present in a microstructure, and the like. Accordingly, in order to
manufacture a Nb-containing zirconium alloy nuclear fuel cladding
tube having excellent corrosion resistance, it is most important to
optimize the kind and amount of the elements added to a zirconium
alloy.
[0009] Conventional technologies related to Nb-containing zirconium
alloys used to manufacture nuclear fuel cladding tubes for nuclear
reactors are described as follows.
[0010] U.S. Pat. No. 5,838,753 and European Patent No. 1,111,623
disclose a method of preparing a zirconium alloy for nuclear fuel
cladding tubes and structural parts for high burnup, in which the
zirconium alloy comprises 0.5.about.3.25 wt % of niobium (Nb) and
0.3.about.1.8 wt % of tin (Sn). More specifically, these patents
disclose a method of manufacturing a nuclear fuel cladding tube,
comprising the steps of: heating a zirconium alloy billet to a
temperature above 950.degree. C. and then rapidly quenching the
billet to a temperature below the .alpha.-transformation
temperature in the (.alpha.+.beta.)-phase range; extruding the
quenched billet at a temperature below 600.degree. C. to form a
hollow billet; annealing the hollow billet by heating at a
temperature up to 590.degree. C.; pilgering the annealed hollow
billet; and finally annealing the pilgered hollow billet at a
temperature up to 590.degree. C. to form the nuclear fuel cladding
tube. Here, the zirconium alloy has a microstructure of
.beta.-niobium second phase precipitates distributed uniformly and
intergranularly forming radiation resistant second phase
precipitates in the alloy matrix so as to result in an increased
resistance to aqueous corrosion compared to that of Zircaloy when
irradiated to high fluence.
[0011] International Patent Publication No. 2001-061062 discloses a
method of manufacturing a nuclear fuel cladding tube including
0.6.about.2 wt % of niobium (Nb) and a small amount of tin (Sn), in
which the ratio of Sn/Fe is 0.25/0.5, 0.4/(0.35.about.0.5) or
0.5/(0.25.about.0.5), and the amount of Sn+Fe is 0.75 wt % or more.
This method comprises the steps of vacuum melting, forging,
hot/cold rolling and heat treatment. The final target of the method
is to uniformly distribute small-sized .beta.-niobium (.beta.-Nb)
precipitates and zirconium-niobium-iron (Zr--Nb--Fe) precipitates
in the zirconium alloy matrix.
[0012] Japanese Patent Publication No. 2001-208879 discloses a
nuclear fuel assembly member having a welding area. Here, a
zirconium alloy or a zircalloy including 0.2.about.1.5 wt % of
niobium (Nb) is annealed at a temperature of 400.about.620.degree.
C. in order to improve the corrosion resistance of the welding area
of the nuclear fuel assembly member.
[0013] International Patent Publication Nos. 2001-024193 and
2001-024194 disclose a zirconium alloy for nuclear reactor
components, comprising 0.02.about.1 wt % of iron (Fe),
0.8.about.2.3 wt % of niobium (Nb), 2000 ppm or less of tin (Sn),
2000 ppm or less of oxygen (O), 100 ppm or less of carbon (C),
5.about.35 ppm of sulfur (S), and 0.25 wt % or less of
chromium+vanadium (Cr+V).
[0014] As described above, in the field of nuclear fuel cladding
tubes, research into making a zirconium alloy nuclear fuel cladding
tube having excellent corrosion resistance using a
niobium-containing zirconium alloy by changing the kind and amount
of the elements added to the zirconium alloy has been continuously
conducted. However, in nuclear power plants, in order to meet
high-burnup long-period operation conditions, it is still required
to develop a zirconium alloy nuclear fuel cladding tube having more
excellent corrosion resistance than that of conventional nuclear
fuel cladding tubes. When a zirconium alloy nuclear fuel cladding
tube is used in the reactor environment, the corrosion phenomenon
thereof cannot be avoided, and thus an oxide film is formed on the
surface of the zirconium alloy nuclear fuel cladding tube. Although
it is well known that the corrosion resistance of a nuclear fuel
cladding tube is determined by the protection ability of an oxide
film formed in the early stage of corrosion, to date, zirconium
alloys for nuclear fuel cladding tubes have been developed based on
only experimental results, and technologies for improving the
protection ability of the oxide film have not yet been
developed.
[0015] Therefore, while the present inventors were conducting
research into niobium-containing zirconium alloy nuclear fuel
cladding tubes having excellent corrosion resistance, they
developed a zirconium alloy composition which can greatly improve
the corrosion resistance of nuclear fuel cladding tubes by forming
a protective oxide film thereon in the reactor environment, thereby
completing the present invention.
SUMMARY OF THE INVENTION
[0016] Accordingly, the present invention has been made to solve
the above-mentioned problems, and an object of the present
invention is to provide a zirconium alloy composition for nuclear
fuel cladding tubes, which can maintain excellent corrosion
resistance for a long period time by forming a protective oxide
film under the operating environment of a light water reactor or a
heavy water reactor.
[0017] Another object of the present invention is to provide a
zirconium alloy nuclear fuel cladding tube having excellent
corrosion resistance, which can meet high-burnup long-period
operation conditions, and a method of manufacturing the same.
[0018] In order to accomplish the above objects, the present
invention provides a zirconium alloy composition for nuclear fuel
cladding tubes, comprising: 1.6.about.2.0 wt % of Nb;
0.05.about.0.14 wt % of Sn; 0.02.about.0.2 wt % of one or two
elements selected from the group consisting of Fe, Cr and Cu;
0.09.about.0.15 wt % of 0; 0.008.about.0.012 wt % of Si; and a
balance of Zr, a nuclear fuel cladding tube comprising the
zirconium alloy composition, and a method of manufacturing the
nuclear fuel cladding tube.
BRIEF DESCRIPTION OF THE DRAWINGS
[0019] The above and other objects, features and advantages of the
present invention will be more clearly understood from the
following detailed description taken in conjunction with the
accompanying drawing, in which:
[0020] FIG. 1 is a graph showing the weight gain of the nuclear
fuel cladding tube, which was corroded for 480 days, according to
an embodiment of the present invention.
DESCRIPTION OF THE PREFERRED EMBODIMENTS
[0021] Hereinafter, preferred embodiments of the present invention
will be described in detail with reference to the attached
drawings.
[0022] The present invention provides a zirconium alloy composition
for nuclear fuel cladding tubes, including: 1.6.about.2.0 wt % of
Nb; 0.05.about.0.14 wt % of Sn; 0.02.about.0.2 wt % of one or two
elements selected from the group consisting of Fe, Cr and Cu;
0.09.about.0.15 wt % of O; 0.008.about.0.012 wt % of Si; and a
balance of Zr.
[0023] Here, the zirconium alloy composition may include: 1.8 wt %
of Nb; 0.1 wt % of Sn; 0.05 wt % of one element selected from the
group consisting of Fe, Cr and Cu; 0.12 wt % of O; 0.01 wt % of Si;
and a balance of Zr.
[0024] Further, the present invention provides a zirconium alloy
composition for nuclear fuel cladding tubes, including:
1.6.about.2.0 wt % of Nb; 0.05.about.0.14 wt % of Sn;
0.02.about.0.2 wt % of Fe; 0.02.about.0.2 wt % of Cr or Cu;
0.09.about.0.15 wt % of O; 0.008.about.0.012 wt % of Si; and a
balance of Zr.
[0025] Here, the zirconium alloy composition may include: 1.8 wt %
of Nb; 0.1 wt % of Sn; 0.05 wt % of Fe; 0.05 wt % of Cr or Cu; 0.12
wt % of O; 0.01 wt % of Si; and a balance of Zr.
[0026] The zirconium alloy composition may include: 1.8 wt % of Nb;
0.1 wt % of Sn; 0.1 wt % of Fe; 0.1 wt % of Cr or Cu; 0.12 wt % of
O; 0.01 wt % of Si; and a balance of Zr.
[0027] The zirconium alloy composition may include: 1.8 wt % of Nb;
0.1 wt % of Sn; 0.2 wt % of Fe; 0.1 wt % of Cr or Cu; 0.12 wt % of
O; 0.01 wt % of Si; and a balance of Zr.
[0028] Generally, it is known that an oxide film, which is formed
when a zirconium alloy nuclear fuel cladding tube is corroded under
a reactor environment, has a lamellar structure. The reason why the
oxide film has a lamellar structure is that crystal grains are
periodically and repetitively changed from columnar grains to
equiaxial grains when the oxide film is grown. This periodical and
repetitive procedure is referred to as `transition`. A protective
oxide film is generally characterized in that the thickness of each
layer of its lamellar structure is large, and the length and width
of columnar grains constituting each layer are large. When this
oxide film having such characteristics is formed in the early stage
of the corrosion of a nuclear fuel cladding tube, the diffusion
rate of oxygen ions through the oxide film is decreased, so that
the growth rate of the oxide film is also decreased, thereby
improving the corrosion resistance of the nuclear fuel cladding
tube.
[0029] The conditions that are required to form a protective oxide
film on a nuclear fuel cladding tube are determined by the kind and
composition ratio of the elements constituting a zirconium alloy.
Therefore, hereinafter, the characteristics and composition ratios
of the elements constituting a zirconium alloy composition for
nuclear fuel cladding tubes according to the present invention will
be described in detail in terms of the formation of a protective
oxide film.
[0030] (1) Niobium (Nb)
[0031] Niobium (Nb) is known as an element for stabilizing the beta
(.beta.) phase of zirconium. The effects of niobium (Nb)
influencing corrosion are different from each other in results.
Generally, it is known that when the amount of niobium (Nb) added
to a zirconium alloy composition is less than 0.5 wt % (low niobium
content), the corrosion resistance of the zirconium alloy is
greatly increased and the workability thereof is improved, whereas,
when the amount of niobium (Nb) added to the zirconium alloy
composition is more than 1.0 wt % (high niobium content), the
corrosion resistance of the zirconium alloy are increased. When
niobium (Nb) is added to a zirconium matrix in more than solid
solubility, solid solutions and precipitates are formed in the
zirconium matrix, thereby improving the mechanical properties of
zirconium.
[0032] It is generally accepted that the addition of niobium (Nb)
contributes greatly to the increase in corrosion resistance of a
zirconium alloy. Actually, several kinds of niobium-containing
zirconium alloy nuclear fuel cladding tubes were developed, and
have been practically employed as nuclear fuel cladding tubes in
nuclear power plants. However, the causes of the improvement in
corrosion resistance of a niobium-containing zirconium alloy have
not yet been clearly verified. Several causes of the increase in
corrosion resistance of the niobium-containing zirconium alloy have
been proposed, and among them, it is known that the strongest
contributor to the increase in corrosion resistance thereof is
.beta.-niobium precipitates, and that the corrosion resistance
thereof improves as the size of the .beta.-niobium precipitates
decreases. Sine the corrosion resistance of the zirconium alloy is
influenced by the characteristics of the oxide film formed in the
early stage of the corrosion of a nuclear fuel cladding tube, it
improves if the oxide film maintains the protective properties
against the diffusion of oxygen for a long period of time. From
this viewpoint, fine and uniform .beta.-niobium precipitates are
advantageous to the improvement in the corrosion resistance of the
zirconium alloy because they can make the internal stress of the
oxide film uniform and can maintain the protective properties of
the oxide film for a long period of time even when they penetrate
into the oxide film through the corrosion process of the zirconium
alloy. For this reason, the amount of niobium (Nb) may be
1.6.about.2.0 wt %, preferably 1.7.about.1.9 wt %.
[0033] (2) Tin (Sn)
[0034] Tin (Sn) is known as an element for stabilizing the alpha
(.alpha.) phase of zirconium, and serves to improve the mechanical
strength of a zirconium alloy by solid-solution hardening. It is
known that the crystal grain size of an oxide film is decreased by
the addition of tin (Sn). Further, decreasing the amount of tin
(Sn) added to a zirconium alloy composition advantageously aids the
formation of a protective oxide film, but when tin (Sn) is not
added to the zirconium alloy composition at all, the corrosion rate
of the zirconium alloy can be rapidly accelerated in LiOH corrosion
conditions. Accordingly, in the present invention, it is preferred
that the amount of tin (Sn) be 0.05.about.0.14 wt % as long as it
does not greatly influence the decrease in corrosion resistance of
the zirconium alloy.
[0035] (3) Iron (Fe), Chromium (Cr) and Copper (Cu) (Transition
Metal Elements)
[0036] Transition metals, such as iron (Fe), chromium (Cr), copper
(Cu) and the like, make the growth of an oxide film irregular.
However, due to this phenomenon, it is possible to prevent the
oxide film from growing in only one direction, so that it is
possible to prevent the oxide film from being abruptly destroyed.
Further, it is known that the deformability of the oxide film is
improved by the addition of these transition metals. However, when
the amount of the transition metal added to a zirconium alloy
composition is increased, the workability of the zirconium alloy at
the time of manufacturing a nuclear fuel cladding tube is
decreased. Therefore, in the present invention, the amount of the
transition metal may be 0.02.about.0.2 wt %, preferably
0.05.about.0.1 wt %, and the transition metal may be one or two
elements selected from among iron (Fe), chromium (Cr) and copper
(Cu).
[0037] (4) Silicon (Si) and Oxygen (O)
[0038] Silicon (Si) serves to decrease the absorptivity of hydrogen
in a zirconium matrix and to delay the transition phenomenon in
which the corrosion rate of a zirconium alloy rapidly increases as
time advances, and oxygen (O) serves to improve the mechanical
strength of the zirconium alloy because it dissolves in the
zirconium matrix and thus causes solid-solution hardening.
[0039] In the zirconium alloy composition according to the present
invention, the amount of silicon (Si), which is added to the
zirconium alloy composition in very small quantities, may be
0.008.about.0.012 wt %, preferably 0.009.about.0.011 wt %. Further,
the amount of oxygen (O) added thereto may be 0.09.about.0.15 wt %,
preferably 0.11.about.0.13 wt %. When the amount of silicon (Si)
deviates from this range, the corrosion resistance of the zirconium
alloy can be decreased, and when the amount of oxygen (O) deviates
from this range, the corrosion resistance of the zirconium alloy
can be decreased and the workability thereof can be
deteriorated.
[0040] In addition, the present invention provides a zirconium
alloy nuclear fuel cladding tube manufactured by the
above-mentioned zirconium alloy composition for nuclear fuel
cladding tubes.
[0041] Further, the present invention provides a method
manufacturing a zirconium alloy nuclear fuel cladding tube.
[0042] The method of manufacturing a zirconium alloy nuclear fuel
cladding tube according to the present invention includes the steps
of: 1) vacuum-arc-remelting and cooling a mixture of the elements
constituting the zirconium alloy composition to form an ingot; 2)
forging the ingot at a temperature of 1000.about.1200.degree. C.;
3) solution-heat-treating the forged ingot at a temperature of
1000.about.1200.degree. C. for 10.about.40 minutes and then cooling
the solution-heat-treated ingot to a temperature of
300.about.400.degree. C. at a cooling rate of 300.about.400.degree.
C./s; 4) extruding the cooled ingot at a temperature of
600.about.640.degree. C. to form an extruded shell; 5) primarily
heat-treating the extruded shell at a temperature of
570.about.610.degree. C. for 2.about.4 hours; 6) cold-working the
primarily heat-treated extruded shell 2.about.5 times and
intermediately heat-treating it 1.about.4 times between the cold
working processes at a temperature of 570.about.610.degree. C. for
3.about.10 hours to prepare a zirconium alloy nuclear fuel cladding
tube; and 7) finally heat-treating the prepared zirconium alloy
nuclear fuel cladding tube at a temperature of
470.about.580.degree. C. for 1.about.100 hours to manufacture a
zirconium alloy nuclear fuel cladding tube.
[0043] The steps of the method of manufacturing a zirconium alloy
nuclear fuel cladding tube according to the present invention will
be described in detail.
[0044] First, in step 1, an ingot is formed by mixing the elements
constituting the zirconium alloy composition according to the
present invention in a predetermined range and then
vacuum-arc-remelting and cooling the mixture.
[0045] The ingot may be formed through a vacuum arc remelting (VAR)
process. Specifically, a chamber is maintained in a vacuum of
1.times.10.sup.-5 torr, argon (Ar) gas is injected into the chamber
at a pressure of 0.1.about.0.3 torr, and then an electric current
of 500.about.1000 A is applied to the chamber to melt the mixture,
and then the melted mixture is cooled to form the ingot in the form
such as button. In this case, in order to prevent the segregation
of impurities or the nonuniform distribution of the zirconium alloy
composition in the button, the mixture may be repetitively remelted
3.about.6 times. In the cooling process, in order to prevent the
oxidization of the remelted mixture, it may be cooled while
injecting an inert gas.
[0046] Next, in step 2, the ingot formed in step 1 is forged in a
.beta.-phase region. In this step, in order to destroy the cast
structure in the ingot formed in step 1, the ingot may be forged in
the .beta.-phase region at a temperature of 1000.about.1200.degree.
C. When the forging temperature of the ingot is below 1000.degree.
C., there is a problem in that the cast structure in the ingot
cannot be easily destroyed, and when the forging temperature
thereof is above 1200 there is a problem in that the heat treatment
cost is increased.
[0047] Next, step 3 is a .beta.-quenching step of
solution-heat-treating the ingot forged in step 2 in the
.beta.-phase region and then rapidly cooling the
solution-heat-treated ingot. In this step, in order to uniformize
the zirconium alloy composition in the ingot and obtain fine
precipitates, the forged ingot is solution-heat-treated in the
.beta.-phase region and then rapidly cooled. In this case, in order
to prevent the oxidization of the forged ingot, the forged ingot
may be sealed with a stainless steel sheet and then
solution-heat-treated at a temperature of 1000.about.1200.degree.
C., preferably 1050.about.1100.degree. C., for 10.about.40 minutes,
preferably 20.about.30 minutes. After the solution-heat-treatment
of the forged ingot, the solution-heat-treated ingot may be cooled
using water in the .beta.-phase region to a temperature of
300.about.400.degree. C. at a cooling rate of 300.about.400.degree.
C./s.
[0048] Next, step 4 is a hot-working step of extruding the ingot
cooled in step 3. In this step, the ingot cooled in step 3 is
processed into a hollow billet, and then the hollow billet is
hot-extruded to form an extruded shell suitable for cold working.
In this case, the extrusion temperature may be
600.about.640.degree. C., preferably 625.about.635.degree. C. When
the extrusion temperature deviates from this range, it is difficult
to obtain an extruded shell suitable for the following step 5.
[0049] Next, in step 5, the extruded shell formed in step 4 is
primarily heat-treated. Specifically, the extruded shell may be
heat-treated at a temperature of 570.about.610.degree. C. for
2.about.4 hours, preferably at a temperature of
575.about.585.degree. C. for 2.5.about.3.5 hours.
[0050] Next, in step 6, the extruded shell primarily heat-treated
in step 5 is repetitively cold-worked and intermediately
heat-treated several times to prepare a zirconium alloy nuclear
fuel cladding tube. In this step, the cold-working and intermediate
heat treatment of the primarily heat-treated extruded shell may be
formed by cold-working the primarily heat-treated extruded shell
2.about.5 times and intermediately heat-treating it 1.about.4 times
between the cold working process. In this case, the intermediate
heat treatment of the extruded shell may be performed at a
temperature of 570.about.610.degree. C. for 3.about.10 hours. The
reason for repetitively performing the cold working and
intermediate heat treatment is to make a recrystallization texture
in a nuclear fuel cladding tube, to finely and uniformly distribute
.beta.-niobium precipitates and to bring the concentration of
niobium in the zirconium matrix to an equilibrium concentration of
0.3.about.0.6 wt %. Through the above processes, a nuclear fuel
cladding tube having an outer diameter of 9.5 mm and a thickness of
0.57 mm can be prepared.
[0051] Finally, in step 7, the zirconium alloy nuclear fuel
cladding tube prepared in step 6 is finally heat-treated. In this
case, the final heat treatment of the prepared zirconium alloy
nuclear fuel cladding tube may be performed at a temperature of
470.about.580.degree. C. for 1.about.100 hours. Due to the final
heat treatment, the concentration of niobium in an
.alpha.-zirconium matrix of the zirconium alloy nuclear fuel
cladding tube becomes 0.3.about.0.6 wt %, and precipitates
including .beta.-niobium precipitates are formed. A zirconium alloy
including the .beta.-niobium precipitates is required to be
heat-treated for a long period of time in order to form an
equilibrium texture therein through the final heat treatment.
However, in this case, since the corrosion resistance of the
zirconium alloy can be deteriorated due to the increase in the size
of the precipitates, the average particle size of the
.beta.-niobium precipitates may be controlled to be 70 nm or
less.
[0052] The zirconium alloy nuclear fuel cladding tube according to
the present invention can maintain excellent corrosion resistance
under the condition of high-burnup operation, so that it can be
usefully used as a nuclear fuel cladding tube for light water
reactors or heavy water reactors, thus improving the economical
efficiency and safety of nuclear fuel.
[0053] Hereinafter, the present invention will be described in more
detail with reference to the following Examples. Here, the
following Examples are set forth to illustrate the present
invention, and the scope of the present invention is not limited
thereto.
Example 1
Manufacture of a Zirconium Alloy Nuclear Fuel Cladding Tube
[0054] (1) Formation of an Ingot
[0055] An ingot was formed using a zirconium alloy composition
including 1.8 wt % of Nb, 0.1 wt % of Sn, 0.05 wt % of Fe, 0.12 wt
% of O, 0.01 wt % of Si and a balance of zirconium through a vacuum
arc remelting (VAR) process. Here, reactor-grade sponge zirconium
defined clearly in the ASTM B349 was used as the balance of
zirconium, and the elements constituting the zirconium alloy
composition had a high purity of 99.99%. Further, silicon (Si),
oxygen (O) and the sponge zirconium were primarily melted to
prepare a mother alloy, and then a desired amount of the mother
alloy was added at the time of remelting the ingot. In this case,
in order to prevent the segregation of impurities or the nonuniform
distribution of the zirconium alloy composition, the zirconium
alloy composition was repetitively remelted 4 times. Further, in
order to prevent the oxidization of the remelted zirconium alloy
composition, a chamber was maintained in a vacuum of
1.times.10.sup.-5 torr, argon (Ar) gas having high purity (99.99%)
was injected into the chamber and then an electric current of 500 A
was applied to the chamber to melt the mixture, and then the melted
mixture was cooled to form the ingot in a water-cooled copper
crucible having a diameter of 60 mm and including cooling water
having a pressure of 1 kgf/cm.sup.2.
[0056] (2) Forging of an Ingot in a .beta.-Phase Region
[0057] In order to destroy the cast structure in the ingot, the
ingot was forged in the .beta.-phase region at a temperature of
1100.degree. C.
[0058] (3) .beta.-Quenching
[0059] In order to uniformize the zirconium alloy composition in
the ingot, the ingot forged was solution-heat-treated in the
.beta.-phase region at a temperature of 1050.degree. C. for 20
minutes. Further, in order to prevent the oxidization of the ingot,
the ingot was cladded with a stainless steel sheet having a
thickness of 1 mm. After the solution heat treatment, the ingot was
rapidly cooled to a temperature of 400.degree. C. or less at a
cooling rate of 300.degree. C./s or more to form a martensite
structure or a Widmanstatten structure. Thereafter, in order to
remove moisture remaining in the cladded ingot, the cooled ingot
was dried at a temperature of 150 for 24 hours.
[0060] (4) Hot Working
[0061] The ingot .beta.-quenched was processed into a hollow
billet, and then the hollow billet was hot-extruded at a
temperature of 630.degree. C. to form an extruded shell suitable
for cold working.
[0062] (5) Primary Heat Treatment
[0063] The extruded shell formed was primarily heat-treated at a
temperature of 580.degree. C. for 3 hours.
[0064] (6) Cold Working and Intermediate Heat Treatment
[0065] The extruded shell primarily heat-treated was primarily
cold-worked and then intermediately heat-treated in vacuum at a
temperature of 580.degree. C. for 2 hours. Subsequently, the
extruded shell was secondarily cold-worked and then intermediately
heat-treated in a vacuum at a temperature of 580.degree. C. for 2
hours. Then, the extruded shell was tertiarily cold-worked and then
intermediately heat-treated in vacuum at a temperature of
580.degree. C. for 2 hours. Thereafter, the extruded shell was
finally cold-worked to prepare a nuclear fuel cladding tube having
an outer diameter of 9.5 mm and a thickness of 0.57 mm.
[0066] (7) Final Heat Treatment
[0067] The nuclear fuel cladding tube prepared was finally
heat-treated in a vacuum at a temperature of 470.about.580.degree.
C. for 10 hours to manufacture a zirconium alloy nuclear fuel
cladding tube.
Example 2
Manufacture of a Zirconium Alloy Nuclear Fuel Cladding Tube
[0068] A zirconium alloy nuclear fuel cladding tube was
manufactured in the same manner as in Example 1, except that the
zirconium alloy composition includes 1.8 wt % of Nb, 0.1 wt % of
Sn, 0.05 wt % of Cr, 0.12 wt % of O, 0.01 wt % of Si and a balance
of zirconium.
Example 3
Manufacture of a Zirconium Alloy Nuclear Fuel Cladding Tube
[0069] A zirconium alloy nuclear fuel cladding tube was
manufactured in the same manner as in Example 1, except that the
zirconium alloy composition includes 1.8 wt % of Nb, 0.1 wt % of
Sn, 0.05 wt % of Cu, 0.12 wt % of O, 0.01 wt % of Si and a balance
of zirconium.
Example 4
Manufacture of a Zirconium Alloy Nuclear Fuel Cladding Tube
[0070] A zirconium alloy nuclear fuel cladding tube was
manufactured in the same manner as in Example 1, except that the
zirconium alloy composition includes 1.8 wt % of Nb, 0.1 wt % of
Sn, 0.05 wt % of Fe, 0.05 wt % of Cr, 0.12 wt % of O, 0.01 wt % of
Si and a balance of zirconium.
Example 5
Manufacture of a Zirconium Alloy Nuclear Fuel Cladding Tube
[0071] A zirconium alloy nuclear fuel cladding tube was
manufactured in the same manner as in Example 1, except that the
zirconium alloy composition includes 1.8 wt % of Nb, 0.1 wt % of
Sn, 0.2 wt % of Fe, 0.1 wt % of Cr, 0.12 wt % of O, 0.01 wt % of Si
and a balance of zirconium.
Example 6
Manufacture of a Zirconium Alloy Nuclear Fuel Cladding Tube
[0072] A zirconium alloy nuclear fuel cladding tube was
manufactured in the same manner as in Example 1, except that the
zirconium alloy composition includes 1.8 wt % of Nb, 0.1 wt % of
Sn, 0.05 wt % of Fe, 0.05 wt % of Cu, 0.12 wt % of O, 0.01 wt % of
Si and a balance of zirconium.
Example 7
Manufacture of a Zirconium Alloy Nuclear Fuel Cladding Tube
[0073] A zirconium alloy nuclear fuel cladding tube was
manufactured in the same manner as in Example 1, except that the
zirconium alloy composition includes 1.8 wt % of Nb, 0.1 wt % of
Sn, 0.1 wt % of Fe, 0.1 wt % of Cu, 0.12 wt % of O, 0.01 wt % of Si
and a balance of zirconium.
Comparative Example 1
[0074] A zirconium alloy nuclear fuel cladding tube was
manufactured in the same manner as in Example 1, except that the
zirconium alloy composition includes 1.3 wt % of Sn, 0.2 wt % of
Fe, 0.1 wt % of Cr, 0.12 wt % of O, 0.01 wt % of Si and a balance
of zirconium.
Comparative Example 2
[0075] A zirconium alloy nuclear fuel cladding tube was
manufactured in the same manner as in Example 1, except that the
zirconium alloy composition includes 1.0 wt % of Nb, 1.0 wt % of
Sn, 0.1 wt % of Fe, 0.12 wt % of O, 0.01 wt % of Si and a balance
of zirconium.
Comparative Example 3
[0076] A zirconium alloy nuclear fuel cladding tube was
manufactured in the same manner as in Example 1, except that the
zirconium alloy composition includes 1.0 wt % of Nb, 0.12 wt % of
0, 0.01 wt % of Si and a balance of zirconium.
[0077] The above-mentioned zirconium alloy compositions are shown
in Table 1.
TABLE-US-00001 TABLE 1 Nb Sn Fe Cr Cu O Si Class. (wt %) (wt %) (wt
%) (wt %) (wt %) (wt %) (wt %) Zr Exp. 1 1.8 0.1 0.05 0.12 0.01
balance Exp. 2 1.8 0.1 0.05 0.12 0.01 balance Exp. 3 1.8 0.1 0.05
0.12 0.01 balance Exp. 4 1.8 0.1 0.05 0.05 0.12 0.01 balance Exp. 5
1.8 0.1 0.2 0.1 0.12 0.01 balance Exp. 6 1.8 0.1 0.05 0.05 0.12
0.01 balance Exp. 7 1.8 0.1 0.1 0.1 0.12 0.01 balance Comp. Exp. 1
1.3 0.2 0.1 0.12 0.01 balance Comp. Exp. 2 1.0 1.0 0.1 0.12 0.01
balance Comp. Exp. 3 1.0 0.12 0.01 balance
Experimental Example 1
Corrosion Experiments for Evaluating Corrosion Resistance
[0078] Corrosion experiments for evaluating the corrosion
resistance of the nuclear fuel cladding tubes manufactured
according to the present invention were conducted as follows.
[0079] The nuclear fuel cladding tubes manufactured in Examples 1
to 7 and Comparative Examples 1 to 3 were fabricated into corrosion
specimens having a length of 50 mm, and then the corrosion
specimens were immersed into a mixed solution of water (H.sub.2O),
nitric acid (HNO.sub.3) and hydrofluoric acid (HF) having
H.sub.2O:HNO.sub.3:HF=50:40:10 (v/v) to remove impurities and fine
defects from the surfaces of the corrosion specimens. Before the
surface-treated corrosion specimens were put into a corrosion test
apparatus, the surface areas and initial weights thereof were
measured. Subsequently, the corrosion specimens were put into the
corrosion test apparatus simulating a reactor environment using
water conditions of 360.degree. C. (18.9 MPa) and then corroded for
90 and 480 days, and then the increases in weights of the corrosion
specimens were measured, so that the weight gains per surface area
of the corrosion specimens were calculated, thereby quantitatively
evaluating the corrosion degree of the corrosion specimens. The
corrosion test results are shown in Table 2 and FIG. 1.
TABLE-US-00002 TABLE 2 Weight gain (mg/dm.sup.2) Class. 90 days 480
days Exp. 1 26.64 47.51 Exp. 2 28.66 47.60 Exp. 3 28.18 47.08 Exp.
4 29.12 49.96 Exp. 5 27.16 61.06 Exp. 6 27.25 46.98 Exp. 7 27.49
48.80 Comp. Exp. 1 28.38 144.17 Comp. Exp. 2 28.84 126.93 Comp.
Exp. 3 28.37 62.25
[0080] As shown in Table 2 and FIG. 1, after 90 days, the weight
gains of the nuclear fuel cladding tubes manufactured in Examples 1
to 7 are in a range of 26.64.about.29.12 mg/dm.sup.2, which differ
little from the weight gains (28.37.about.28.84 mg/dm.sup.2) of the
nuclear fuel cladding tubes manufactured in Comparative Examples 1
to 3. However, after 480 days, the weight gains of the nuclear fuel
cladding tubes manufactured in Examples 1 to 7 are in a range of
47.08.about.61.06 mg/dm.sup.2, which are 50% or more lower than the
weight gains (62.25.about.144.17 mg/dm.sup.2) of the nuclear fuel
cladding tubes manufactured in Comparative Examples 1 to 3.
Experimental Example 2
Observation of an Oxide Film Formed After the Corrosion of a
Nuclear Fuel Cladding Tubes
[0081] In order to evaluate the ability to protect against
corrosion of an oxide film formed on the nuclear fuel cladding tube
manufactured according to the present invention, the oxide film,
which was formed after the corrosion test, was observed using a
transmission electron microscope.
[0082] Specimens for observing the oxide film were made thin to
such a degree that they could be observed by the transmission
electron microscope using ion beams after the corrosion specimens
of Examples 1 to 7 and Comparative Examples 1 to 3, which had been
corroded for 90 days under a reactor atmosphere, were cut to a
thickness of 100 .mu.m. The oxide film specimens were observed by
the transmission electron microscope, and thus the crystal grain
size distribution of the oxide film was measured, thereby
evaluating the ability of the oxide film to protect against
corrosion. The results thereof are shown in Table 3.
TABLE-US-00003 TABLE 3 Fraction of Length of Width of columnar
grain columnar grain columnar grain Class. (%) (nm) (nm) Exp. 1 ~70
~500 ~60 Exp. 2 ~70 ~500 ~60 Exp. 3 ~70 ~500 ~60 Exp. 4 ~70 ~500
~60 Exp. 5 ~70 ~500 ~60 Exp. 6 ~70 ~500 ~60 Exp. 7 ~70 ~500 ~60
Comp. Exp. 1 ~20 ~150 ~20 Comp. Exp. 2 ~40 ~250 ~30 Comp. Exp. 3
~50 ~350 ~40
[0083] The corrosion resistance of the nuclear fuel cladding tube
is determined by the protective ability of the oxide film formed in
the early stage of the corrosion of the nuclear fuel cladding tube.
As shown in Table 3, since the oxide films formed on the nuclear
fuel cladding tubes manufactured in Examples 1 to 7 are protective
oxide films having a larger columnar grain fraction, length and
width than those of the oxide films formed on the nuclear fuel
cladding tubes manufactured in Comparative Examples 1 to 3, the
diffusion of oxygen through an oxide film can be prevented, thereby
maintaining excellent corrosion resistance.
[0084] As described above, since the nuclear fuel cladding tube
made of the zirconium alloy composition according to the present
invention can maintain excellent corrosion resistance by forming a
protective oxide film thereon under the conditions of
high-temperature and high-pressure cooling water and water vapor,
it can be usefully used as a nuclear fuel cladding tube for light
water reactors or heavy water reactors, thus improving the
economical efficiency and safety of the use of nuclear fuel.
[0085] Although the preferred embodiments of the present invention
have been disclosed for illustrative purposes, those skilled in the
art will appreciate that various modifications, additions and
substitutions are possible, without departing from the scope and
spirit of the invention as disclosed in the accompanying
claims.
* * * * *