Note: Descriptions are shown in the official language in which they were submitted.
~lAl)77 ;,
RD- 9 2 5 7 , i~,
NUCLEAR FUEL ELEMENT HAVING A COMPOSITE COATING i~
Background of the Invention 'i
S1; .
This invention relates broadly to nuclear fuel ele~
ments for use in the core of nuclear fis ion reactors. More '
particularly, the present invention relates to a zirconium
containing composite cladding for nuclear fuel having a copper );
coating on its inner surface in proximity to the fuel and an ,"
intermediate zirconium oxide boundary laye~
Nuclear reactors are presently being designed, con-
structed and operated in which the nuclear fuel is contained
in fuel elements which can have various geometric shapes, such
as plates, tube,, or rods. The fuel material is usually
enclo~ed in a low neutron absorbing corrosion-resistant, non- ~;
reactive, heat conductive container or cladding. The elements `
are assembled together in a lattice at fixed distances from "
each other in a coolant flow channel or region forming a fuel
assembly, and sufficient fuel assemblies are combined to form
the nuclear fission chain reacting assembly or reactor core
capable of a self-sustained fission reaction. The core in turn
i8 enclosed within a reactor vessel through which a coolant 1;~
is passed.
The cladding serves several purposes and two primary
purposes are: First, to prevent contact and chemical reactions
between the nuclear fuel and the coolant or the moderator if
A moderator is present, or both if both the coolant and the
moderator are present, and second, to prevent the radioactive ~.
fission produots, some of which are gases, from being released i;
from the fuel into the coolant or the moderator or both if
both the coolant and the moderator are present. Co~mon clad-
ding materialFs are steel, and its alloys, zirconium and its - }:
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RD-9257
alloys, niobium, (columbium) and its alloys, and the like. The
. . ~
failure of the cladding, i.e., a 1088 of the leak tightness, ,~,.,
can contaminate the coolant or moderator and the associated
system~ with radioactive fission products to a degree which l:
interferes with plant operation.
Problems have been encountered in the manufacture
and in ~e operation of nuclear fuel elements which employ
certain metal~ and alloys as the clad material due to mechanical i~
or chemical reactions of these cladding materials under certain
~10 circumstances. Zirconium and its alloys, under normal circum-
stance8, are excellent nuclear fuel claddings since they have
low neutron absorption cross sections are strong, ductile,
extremely stable and at temperatures below about 750F (about ~i.
398C) non-reactive in the presence of demineralized water
~lS and/or steam which are commonly used as reactor coolants and
moderators. ~,
However, fuel element performance has revealed a pro-
blem with defecting of the cladding due to the mechanical
lnteractions between the nuclear fuel and the cladding in the
presenCe of certain fission products produced by nuclear fisq7ion ~ `
reactions. It has been discovered that this undesirable per-
formance is promoted by localization of mechanical stresses ;-;
~due to fuel-cladding differential expansion) at cracks and
at pellet-pellet interfaces in the nuclear fuel. Corrosive ~i
fis~ion products are released from the nuclear fuel and are $~j~
pre8ent at pellet-pellet-interfaces and a~ the intersection
of the fuel cracks with the cladding surface. Fission products . i
are;created in the nuclear fuel during the fission chain t~;*
reaction duri~g operation of nuclear reactor. The localized ~ ~;t
~30; ~ stress is exaggerated by high friction between the fuel and
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9257 .i l '
the cladding. ,~
The zirconium alloy cladding of a nuclear fuel element ~',J.'`
isi exposed to fission products during irradiation in a nuclear
reactor. Sintered refractory and ceramic compositions, such as `,~,
uranium dioxide and other compositions used as nuclear fuel, ,
release quantities of the fission products during irradiation.
Certain of these fission product~ are capable of reacting with A'~
the zirconium or zirconium alloy cladding containing the nuclear '`,
fuel.
~10 Another approach to reactor design has been to coat '~'
the nuclear fuel material with a ceramic to prevent moisture
coming in contact with the nuclear fuel material as disclosed
in U.S. Pate`nt No. 3,108,936. U.S. Patent No. 3,085,059 preRents
a fuel element including a metal casing containing one or more
pellet~ of fissionable ceramic material and a layer of vitreous ~
materlal bonded to the ceramic pellets 90 that the layer is `"
between the casing and the nuclear fuel to assure uniformly good l~
heat conduction from the pellets to the casing. U.S. Patent No. ~'
2,873,238 presents jacketed fissionable slugs of uranium
cànned in a metal case in which the protective jackets or ~'
coverings for the slugs are a zinc-aluminum bonding layer. 1~
U.S. Patent ~o. 2,849,387 discloses a jacketed body sections of ii
nuclear fuel~which have been dipped into a molten bath of a "
~ bonding material giving an effective ther~sally conductive bond '.
1, 25 between the uranium body sections and the container (or clad~
ding). The coating is disclosed as any metal alloy having
¦ good thermal conduction properties with examples including j,;~
alumsinum-silicon and zinc-aluminum alloys. Japanese Patent ~"~
,,.. ,- .
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t,
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Publication No. SHO 47-14200 in which the coating of one of
two groups of pellets is coated with a layer of silicon carbide
and the other group is coated with a layer of pyrocarbon or
metal carbide.
The coating of a nuclear fuel material introduces
reliability problems in that achieving uniform coatings free
of faults is difficult. Further, the deterioration of the
coating can introduce problems with the performance life of the
nuclear fuel material.
In prior work, there was developed a method for
; preventing defects in nuclear fuel cladding, consisting of the
addition of a metal such as niobium to the fuel. The additive
can be in the form of a powder, provided the subsequent fuel
processing operation does not oxidize the metal. Or the additive
can be incorporated into the fuel element as wires, sheets or
other forms in, around or between fuel pellets.
Document GEAP-4555, dated February 1964, discloses
a composite cladding of a zirconium alloy with an inner lining
of stainless steel metallurgically bonded to the zirconium alloy,
~0 and the composite cladding is fabricated by use of extrusion
of a hollow billet of the zirconium alloy having an inner
lining of stainless steel. This cladding has the disadvantage
that the stainless steel layer involves a neutron absorption
penalty of about ten to fifteen times the penalty for a zir-
conium alloy layer of the same thickness.
U.S. Patent No. 3,502,549, Charveriat, issuedMarch 24, 1970, discloses a method of protecting
zirconium and its alloys by the electrolytic deposition
of chrome to provide a composite material useful ~r
nuclear reactors. A method for electrolytic deposition of
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RD-9257 ,~
copper on Zircaloy-2 surfaces and subsequent heat treatment
for the purpose of obtaining surface aiffusion of the electro- '.'7~.
lytically deposiited metal is presented in Energia Nucleare,
Volume 11, number 9 (September 1964) at pages 505-508. In `"
5Stability and Compatibility of Hydrogen Barriers Applied to '~'
: ?'
Zirconium Alloy~, by F. srossa et al (European Atomic Energy ,J'
Community, Joint Nuclear Research Center, EUR 4098e 1969),
methods of deposition of different coatings and their effic-
iency as hydrogen diffusion barriers are described along with
10an Al-Si coating as the most promi6ing barrier against hydrogen ~'
difusion. Methods for electroplating nickel on zirconium
and zirconium tin alloys and heat treating these alloys to
produce alloy-diffusion bonds are disclosed in Electroplating ,~'
on Zirconium and Zirconium-Tin, by W.C. Schickner et al (BMl-757,
15Technical Information Service, 1952). U.S. Patent No. 3,625,821 , ?
pre~ents a fuel element for nuclear reactor having a fuel clad-
ding tube with the inner surface of the tube being coated with
a retaining metal of low neutron capture cross section such as
nlckel and having finely dispersed particles of a burnable
;:1:
20poison disposed therein. Reactor Development Program Progress i -
Report of August, 1973 (ANL-RDP-l9) discloses a chemical getter ,~
arrangement of a sacrificial layer of chromium on the inner ~,
~urface of a stainless steel cladding.
Another approach has been to introdùce a barrier ~'
between the nuclear fuel material and the cladding, a~ dis-
J 1 .
clo~ed in U.S. Patent No. 3,230,150 (copper foil), German Patent ;
Publication DAS 1,238,115 (titanium layer), U.S. Patent No. ,'-
3,212,988 (sheath of zirconium, aluminum or beryllium), U.S. .
Patent ~o. 3,018,238 (barrier of crystalline carbon between the ;).
UO2 and the zirconium cladding, and U.S. Patent No. 3,088,893 ~.'!;
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1~14~ RD-9257
(Stainless steel foil. While the barrier concept proves pro-
mising, some of the foregoing references involve incompatible
materials with either the nuclear fuel (e.g., carbon can combine
with oxygen from the nuclear fuel), or the cladding (e.g., copper
and other metals can react with the cladding, altering the pro-
perties of the cladding), or the nuclear fission reaction (e.g.,
by acting as neutron absorbers). None of the listed references
disclose solutions to the recently discovered problem of
localized chemical-mechanical interactions between the nuclear
fuel and the cladding.
Further approaches to the barrier concept are
disclosed in United States patent 3,969,186 issued July 13,
1976 to Thompson et al (refractory metal such as molybdenum,
tungsten, rhenium, niobium and alloys thereof in the
form of a tube or foil of single or multiple layers or a
coating on the internal surface of the cladding), and
United States patent 3,925,151 issued December 9, 1975 to
Klepfer (liner of zirconium, niobium or alloys thereof
I between the nuclear fuel and the cladding with a
1 20 coating of a high lubricity material between the liner and
the cladding).
An additional effort to the solution of protecting
the zirconium or zirconium alloy cladding container is
shown in United States patent 4,029,545 issued June 14, 1977
to ~ordon et al and assigned to the same assignee
as the present invention. In this application, a layer,
; such as chromium, is electroplated onto a zirconium or
zirconium alloy substrate, followed by the electroplating of
copper onto the chromium layer. However, it has been found to
be economically unattractive to electroplate the zirconium or
zirconium alloy cladding, which hereinafter may be referred to
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6 -
~:114(~7~
RD-9257
:~?3
as the "zirconium cladding", witH chromium rendering the overall ',~
procedure less promising than originally anticipated. An alter-
native procedure is shown by Gordon et al u.s. patent No.4,022,662
~ which shows a nuclear fuel element having a metal liner, such ,''
as a copper liner, between the cladding and the nuclear fuel ,:
and diffusion barrier, such as a chromium coating between the ;~
liner and the cladding. Again, the Gordon et al nualear fuel
el-ment i8 uneconomic because electrodeposition is required ~l'
and a copper liner has to be fabricated.Research effort has '~
therefore been continually directed toward an economic solution j~
of the ~oblem of preventing perforations or failures in the
.....
cladding substrate resulting from metal embrittlement or stress
corrosion cracking involving fuel pellet-cladding interaction. ,j;
;, . .
' Summary of the Invention ;,~'''" '
Th- present invention is based on the discovery that "
a sub~tantial reduction in metal embrittlement or stress cor- '
rosion cracking from fuel pellet-cladding interaction can be ,~ '
achieved by the employment of a copper layer or liner in prox- '.
lmity ~ the nuclear fuel and an intermediate zirconium oxide
barrier layer between the copper layer and the zirconium clad- l''''
ding substrate. Advantageously, the intermediate zirconium
l oxide barrier''layer has been foun'd to be an excellent copper ''
¦~ diffusion barrier. In addition, although copper cannot be
directly electroplated onto non-conducting zirconium oxide, it -t"
1~ 25 has been found that modification of the zirconium cladding
¦ ~ 8urface prior to oxidation, allows for copper deposition by "'
electroless p'lating. ,i'
One aspect of the invention therefore i8 directed to
¦~ a nuclear fuel element comprising "',
: 30 ~ ~A) a central core of nuclear fuel material, ~.
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~114077 ~;
RD--9257 Z/'1
(B) an elongated composite cladding containing fj;
the nuclear fuel material comprising a -
- zirconiumZ or zirconium alloy substrate, ,~
having on its inside surface in proximity
to the nuclear fuel material, a layer of metal ~1,
selected from the group consisting of copper,
nickel, iron and alloys thereof and an ,.
intermediate zirconium oxide barrier ,
between the zirconium or zirconium alloy
substrate and the metal layer.
A further aspect of the invention is directed to a
method for making a composite zirconium or zirconium alloy con-
tainer for nuclear fuel material to produce a nuclear fuel ele-
ment which comprises
l~ 15 (1) etching or roughening the surface of the zircon-
ium or zirconiumZ alloy container, ,
(2) oxidizing the surface of the resulting nuclear ,S'
fuel container of (1) to produce a zirconiumZ or zirconium
alloy container having a zirconium oxide coating,
'~
(3? activating the zirconium oxide coated surface of
the nuclear fuel container of (2) to allow for the metallic
coating of such surface by electroless deposition, and
4 ? further coating the zirconium o~ide layer on the ~ -
~ 25 inside surface of the nuclear fuel container with a metal.
i Description of the Drawings
Figure 1 pre~ents a partial cutaway sectional view of ,~
a nuclear fuel assembly containing nuclear fuel element~ con- i
~tructed according to the teaching of this invention. "
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Figure 2 presents an enlarged cross sectional view
of the nuclear fuel element illustrating the teaching of this
invention. '~.5~!
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Description of the Invention
Referring now more particularly to Figure 1, there i
::,
iB 8hown a partially cutaway sectional view of a nuclear fuel '`",
assembly 10. This fuel assembly 10 consists of a tubular flow
channel 11 of generally square cross section provided at its ',l,
upper end with lifting bale 12 and at its lower end with a 'i
no8e piece (not shown due to the lower portion of assembly 10
being omitted). The upper end of channel 11 is open at 13 and ~'~
the lower end of the nose piece is provided with coolant flow ,~,
openings. An array of fuel elements or rods 14 is enclosed in ',~.'
channel 11 and sUpported therein by means of upper end plate
lS lS and a lower end plate (not shown due to the lower portion
being omitted). The liquid coolant ordinarily enters ~hrough ~'
the openings in the lower end of the nose piece, passes up~
wardly around fuel elements 14, and discharges at upper outlet ' ,
13 ln a partially vaporized condition for boiling reactors or ,
l20 in~an unvaporized condition for pressurized reactors at an 1,
j levated temperature. ,1
l~ i The nuclear fuel elements or rods 14 are sealed at ~';.,'
, . . .,
their end8 by mean8 of end plugs 18 welded to the cladding 17,
¦ ~ whlch may include studs 19 to facilitate the mounting of the !j
~25 ~uel rod in the assembly. A void space or plenum 20 is pro~
. ~ vided at one end of the element to permit longitudinal expan~
. sion of the fuel material and accumulstion of gases released l.,
rom the~fuel material. A nuclear fuel material retainer - ~ "
means 24 in the form of a helical member is positioned within ~;~
~;30~ pace 20 to provlde restraint against the axial movement of i~
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RD-9257
the pellet column, especially during handling and tran8porta- 5r
tion of the fuel element. ~,"
The fuel element is designed to provide an excellent ~ t~
thermal conductance between the fuel and the cladding material, ~ '
S and to avoid bowing and vibration which is occasionally caused
by flow of the coolant at high velocity. ',,
A nuclear fuel element or rod 14 is shown in a par-
tial section in Figure 1 constructed according to the teachings j,,
of thi8 invention. The fuel element 14 includes a core or cen- J~'
tral cylindrical portion of nuclear fuel material 16, here ,'
.. .
shown as a plurality of fuel pellets of fissionable and/or ,'t'~
fertile material positioned within a structural cladding or
container 17. In some cases the fuel pellets may be of various
8hapes such as cylindrical pellets or spheres, and in other
lS cases different fuel forms such as a particulate fuel may be ,1?
used. The physical form of the fuel is immaterial to this
invention. Various nuclear fuel materials may be used includ-
~ ing uranium compounds, plutonium compounds, thorium compounds,
¦ and mixtures thereof. A preferred fuel i8 uranium dioxide or ~.
~20 a mixture comprising uranium dioxide and plutonium dioxide. ,[~
Referring now to Figure 2, the nuclear fuel material
J~ 16 forming the central core of the fuel element 14 is sur- ,;
¦~ rounded by a cladding 17 hereinafter in this description also '~;~
referred to a~ a composite and a8 a composite~cladding. The
~25 composite cladding 17 ha8 a zirconium or zirconium alloy such ''
a8 Zir¢aloy-2 8ub8trate at 21. The substrate has attached
on th- in8ide ~urface thereof, a diffusion barrier 22 80 that
th diffu~ion barrier 22~ forms a shield preventing any dif- -~
fw ion~of;~oth-r sp-cies through the diffusion barrier 22 to .,
~30~ t~he~subs~rate 21. The diffu~ion barrier 22 is preferably about ~;i?
} X 10 5 inch to about S x 10 5 inch in thickness and is '.
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RD-9257
comprised of zirconium dioxide. The diffusion barrier protects '`,-~'
the substrate at 21 from contact and reaction with the metallic 1'.
layer at 23. ~.. `,f
The diffusion barrier 22 has attached thereon a metal
layer 23 so that the metal layer 23 covers the diffusion barr- ~t
ier 22 and also forms a shield for the substrate against fis~
sion products and gaseous impurities emanating from the nuclear
fuel material held in the container. The metal layer is about
2 x lO 4 to about 4 x lO inch in thickness and is composed of
a low neutron penalty metal which is preferably copper, but
;.;:
can include a metal selected from the group consisting of copper, '~ -
nickel, iron and alloys thereof. The copper layer serves aR a
primary or preferential reaction site for fiRsion products 1;
and also acts as a barrier to protect the substrate from contact ~ ,
l15 and reaction with deleterious fission products.
¦ The purity of the copper layer i~ important from a ~:'
neutron penalty aspect. The total impurities in the two layers .,
are ~imited to a boron equivalent of 40 parts per million or '~;
less. In addition, impurities should be kept at a level of
less than one weight percent and preferably below lO00 parts
per million to maintain high ductility and good thermal con- "`-
ductance. '"
The composite cladding of the nuclear fuel element - "
of thi~ invention has the diffusion barrier bonded to the sub- i;;
"
~ 25 strate in a strong bond and the metal layer bonded to the dif- :~
j fu6ion barrier in a strong bond. Te~ts to show the bond ~'
etrength between the diffusion barrier and the sub~trate show
?
that the diffusion barrier remains firmly affixed when bent in
the elaotic region or when permanently strained to about 2%. ~rl
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RD-9257 I!J-
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The copper layer is more resistant to the deleterious ,~,~
;'1.-"
effects of radiation hardening and damage than zirconium and
zirconium alloys under the conditions found in commercial
nuclear fission reactors, e.f3. at temperatures of 500F to
750F. Thus, copper has more ability to withstand plastic
deformation without mechanical failure than zirconium and zir- l;
conium alloys under operating nuclear reactor conditions. In
addition, copper can deform plastically from pellet-induced ~/;
stresses during power transients, relieving pellet-induced
stresses. In addition, these metals will not rupture mechanic-
ally and thus will also shield the zirconium alloy substrate , .
from the deleterious action of fission products. .
It has been di~covered that a metal layer of the order
of about .0001 inch to about .001 inch bonded to the diffusion
barrler which in turn is bonded to the substrate of zirconium
or a zirconium alloy provides stress reduction and chemical ;;~
resistance sufficient to prevent nucleation of failures in the
substrate of the cladding. The metal layer provides signifi-
cant chemical reSistance to fission products and gases that may
be present in the nuclear fuel element and prevents these fis- ,,
sion products and gases from contacting the substrate of the
composite cladding protected by the metal barrier.
It has been discovered, for example, the copper layer
.,, 1
does not oxidize to any appreciable extent, and the stoichio-
me~ry of the UO2 fuel can be stabilized. Without the copper
layer, the ziraonium or zirconium alloy would react with the ,-~
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oxide nuclear fuel forming ZrO2,-thus changing the stoichio- ,.,
metry of the oxide nuclear fuel. The chemical state of various
fiss,ion products i8 a very strong function of the oxide nuclear ,"''
fuel stoichiometry. For example, at higher oxygen to uranium ;;~ .t
ratios,, cesium forms a compound with the UO2 fuel. At lower ,!~ ~,
ratio8, thi8 compound is not stable and cesium can migrate to :!
the lower temperature regions of the fuel rod (e.g., inner ,
surface of the ~ adding). Cesium, either alone or i~ combin- '
stion with other fission products, may then promote stress cor-
~;10 ro8ion of the cladding. In a fuel rod with an uncoated cladding, '`"
J
even if the oxide nuclear fuel has a high initial oxygen to `""
I uranium ratio, the oxygen consumed by the oxidation of the zir- ~i"i,
I conium alloy will lower this ratio, and cesium can then be re-
lea8ed to migrate to the cladding 8urface. With the present ,.,
~15 inv-ntion using a diffusion barrier and a metal layer, the ratio ,~
Will remain nearly con&tant or change at a reduced rate. Thus, ',~
' ~ . ,;, .
an oxlde nuclear fuel with any de8ired s,tiochiometry can be used ~''
~ in the compo-ite cladding with the expectation that this ~toichio-
¦ motry will remain con8tant or change with time at a much slower ,
~20 rate.
In the practice of the invention, the zirconium or i,.;
. i~
; ;~ zirconium alloy container, referred to hereinafter as the zir~
c,onLum &ubstrate, zirconium container or zirconium tube can be i,i"
;~; ~ converted to the composite cladding con8i6ting~0f the zirconium .~,
25; ~container with a copper coating on it8 inside surface in con- .,
taCt with an~intermediate zirconium oxide boundary layer by ',~i.
initl~lly modifying the inside surface of the zirconium, con- ~,~"
talh-r.~ ~odification of the inside surface of the zirconium ,~
contàlner~oan bs achieved by either a grit blasting or roller - ~.~,
30 ~ milling technlque or by using a particular etchant. After ~he j~
-13-
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1~ 1 4 ~ RD-9257
zirconium surface has been modified, it is oxidized. The
oxidized surface of the zirconium substrate is then activated
to allow for the electroless plating of a metal such as copper
onto the zirconium oxide.
If the inside surface of the zirconium tube is modi-
fied by the surface roughening technique, the zirconium surface
can be roughened by grit blasting with an aluminum oxide grit
or by internal roller milling using weighted aluminum o~ide
tubing having an outside diameter of from about 8 to 10 milli-
meters and an inside diameter of from about 5 to 7 millimeters.
Roller milling of the zirconium tube can be achieved with wet
powdered aluminum oxide by plugging the ends of the tube and
rolling the tube for 24 to 72 hours at from 12 to 20 RPM.
When employing the etching method to modify the inside
surface of the zirconium tube, the inside of the tube is prefer-
ably initially cleaned with a detergent, exposed to a bright
dip solution, and thereafter washed. A preferred etchant is
shown by United States patent 4,017,368 issued April 12, 1977
to Daniel E. Wax and Robert L. Cowan, II and assigned to
the same assignee as the present invention. A typical etching
; procedure would be to contact the zirconium alloy with an
aged aqueous activating solution comprising from ~bout 10 to 20
grams per liter of ammonium bifluoride and from about Q.75 to
about 2.0 grams per liter of sulfuric acid. The solution can
be aged by immersion of a piece of zirconium having an area of
100 sq. centimeters, per liter of solution, for 10 minutes.
The etched surface of the zirconium tube can then optionally be
scaled to effect the removal of lcosely adhering film.
Oxidation of the above described surface roughened
3Q zirconium tube or etched and scaled zirconium tube can be
' ' .
~1114(1~ ,~:
RD-9257
accomplished by exposure to an oxygen atmosphere at 300C to 1
500C over a period of from 1 to 100 hours. Alternatively, sur-
face oxidation can be effected by treating the inside surface
of the zirconium tube after modification with steam at a tem- ~,
perature of from 350-450C over a period of from 5 to 50 hours.
Experience has shown that activation of the oxidized
surface of the zirconium tube can be achieved by employing
salts of tin and salts of various noble metals. A preferred
.,;. .
combination is alkaline solutions of stannous tin, such as ':
sodium ~tannite and palladium chloride. However, other noble
metàl salts can be used, such as silver nitrate, platinum "
chloride, gold chloride, alkaline solution~ of precious metal~
such as sodium aurate, sodium palladate, sodium platinate.
A typical activating mixture is shown by C.R. Shipley U.S.
patent 3,011,920 or E. Saubestre Technical Proceedings, American -
Electroplating Society 1959. The oxidized zirconium ~urface is
treated with Cuposit Catalyst 9F, a product of the Shipley
Company of Newton, Mass. The treated zirconium oxide can then
be rin-ed further with water and treated with Cuposit Acceler- -
I~ 20 ator 19, also a product of the Shipley Company.
! ~ The electroless plating of the activated zirconium
oxide ooated zirconium substrate can be achieved by standard
proo-dures, such as by allowing the plating solution to flow l;
unlformly through-or over the zirconium substrate to achieve
2~ a uniform buildup of metal on the article. Although copper
i8 the ~oferred metal, other metals such as nickel or iron ~'
also~can be ~ated onto the surface of the zirconium oxide to
lachieve effective results. ~ ~-
For -lectroless plating, an aqueous bath of the fol- ~;~
.~
lowing composition can be used: 600 ml of H20, 141.5 gram~ of
: -15~
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:: ' 't't~ i;;t'ir~ t'!';tiS li ! 7i~ ! t; ;!"~ 's 5 ;i? ~ 5);i ~ t ~ji,.i
1114~7 RD-9257 "~
sodium potassium tartrate ~KNaC4H406.4H20), 41.5 grams of sodium
hydroxide (NaOH), 29 grams of copper sulfate (CuSo4.5H20) plus ~j1
H20 to make 1 liter. Immediately prior to use, 16.7 ml of a
73~ formaldehyde solution (H2CO) can be added to the bath. This "
is a version of well known Fehling' 8 copper plating bath. Other
proprietary electroless copper plating formulations can be
I employed, such as those identified as MacDermid 9038, Shipley ,!
CP74 and Sel-Rex CU510. The plating bath is agitated and passed
uniformly over the article to be plated while being maintained !'
at a temperature of about 25 to about 75C. This procedure
produces a very good as-plated adherence with substantially
no porosity. In order to insure that the plated article can
be used at elevated temperatures without any substantial 1098
of ~dhesion, the plated aritcle is out-gassed in either argon
or vacuum at a temperature of about 300 to about 400F (149 to ,
204C). In this out-ga59ing, the temperature is raised from j~
ambient to the final temperature at a rate of about 50F to
125F per hour.
During the electroless plating of copper on the .
aritlce, a considerable quantity of hydrogen gas can be evolved.
Inasmuch as hydrogen gas can interfere with the electroless
plating process, since it has a tendency to adhere to the wall
of the tube, hydrogen removal can be facilitated by pumping '
the plating solution through the tube. In ad~ition, the tube ~;
, can be electrole~s plated while it is in a vertical position.
i For plating nickel on zirconium, an aqueous bath of
the following composition is employed. 30 grams/liter of
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RD-9257
nickel chloride (NiC12.6H2O), 10 grams/liter of sodium hypopho~-
phite (NaH2P02.H2O), 12.6 grams/liter of sodium citrate
(Na3C6H5O7.2H2O), 5 grams/liter of sodium acetate (NaC2H3O) and
sufficient sodium hydroxide (NaOH) to give a pH in the range of
4 to 6. Other proprietary electroless nickel plating formula-
tions can be employed such as those identified as Enplate 410
and Enplate 416. The plating bath is agitated and passed uni-
formaly over the article to be plated while being maintained at
a temperature of about 194 to about 212F t90 to 100C) with a
preferred target temperature being 95-2C. This procedure pro-
duces a very good as-plated adherence with no porosity. In
, order to insure that the plated article can be used at elev-
I ated temperatures without any substantial loss of adhesion,1 the same out-gassing procedure employed above for copper is1 15 used.
The articles treated by the process of this invention
can be zirconiummaterial taken directly from milling opera-
tions or can be articles subjected to prior mechanical clean-
ing (e.g., grit blasting) or chemically cleaned articles (e.g.,
cleaned by acid and/or alkaline etching).
In order that those skilled in the art will be better
able to practice the invention, the following examples are
glven by way of illustration and not by way of limitation.
; 25 Example 1.
A zircaloy-2 tube, 5 inches long having a 0.490 inch
OD and 0.425 ID i9 cleaned in a detergent solution for 10 min-
. .
utes in a 50 watt ultrasonic cleaner. It is then rinsed 10
minutes ~n distilled water. There is then pumped through the
~ tube at a rate of about 1000 ml/min for 2 minutes a bright
:
~ -17
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4~7~
RD-9257
polish ~olution consisting of 500 ml of H20, 500 ml of concen-
trated nitric acid and 10 grams of ammonium bifluoride.
The tube is then rinsed with water and neutralized in an aqueous
sodium hydroxide solution. After a 5 minute rinse in distilled
water, the tube is etched for 1 minute in the ultrasonic
cleaner, using a solution of 1000 ml of water, 15 grams of
ammonium ~flurodie and 0.5 ml o sulfuric acid. The etching
solution has been previously mixed and aged for 10 minutes by
contacting it with a piece of Zircaloy-2 tube having an area
of about 100 square centimeterR. The ultrasonic cleaner effects
the removal of any loose scaly material which is formed during
etching. After etching, the sample is then rinsed for about
1 minute in distilled water and thereafter dried using dry
nitrogen. The tube is then put in a furnace for oxidizing.
Tho tube i~ oxidized 24 hour8 at 400C using an oxy~en flow of
about 0.2 cubic feet per hour. When the tube has cooled, it
is removed from the furnace and cleaned again in an aqueous
~odium hydroxide solution for 5 minutes in the ultrasonic
cleaner, followed by a 10 minute rinse in distilled water.
I 20 The tube i-q then activated by initiaily pumping
¦~ through it a solution of Cuposit Catalyst 9F manufactured by
the Shipley Company of Newton, Mass., at a rate of 1000 ml/min
for a period of 3 minutes and then rinsed for 3 minutes. There
~ then pumped through the Zircaloy-2 tube, a ffolution of
¦~ 25 Cuposit Accelerator 19 for 6 minutes at a rate of about 1000
i~ ml/min, followéd by a 10 minute rinse in distilled water.
; The tube is then plated for 2 hours at 60C in Metex #9038
plating bath, a commercial product manufactured by MacDermid
Inc.,~of Waterford, Conn. The plating bath.is pumped through
the sample tube at a rate of 1000 ml/min from a vessel having
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RD-9257
a thermostatic control. There ic obtained a composite Zircaloy-2
tube cladding coated on its inside surface with about 3.8 x 10 4
inch of copper and intermediate boundary layer of about 4 x 10 5
I inch of zirconium oxide. The aforementioned Zircaloy-2 tube
S composite is then loaded in accordance with standard techniques
using 0.4 inch x 1.5 inch uranium oxide pellets to produce a
nuclear fuel element useable in the core of a nuclear reactor.
In order to demonstrate the outstanding ability of
zirconium oxide as a barrier layer between a copper coating
and a zirconium substrate as a means for reducing metal embrit-
tlement or failure under reactor condition~, i.e.,
temperatures, such as in excess of 290C while in contact with
cadmium dissolved in ce9ium, etc., a series of 1/2 inch long
Zircaloy-2 tensile samples were prepared having a 1/8 inch
gauge section. The tensile samples were evaluated on an
Instron tensile tester at 300C while in a bath of liquid
cesium saturated with cadmium. Some of the tensile samples
were hoat treated at about 580C for 2-1/4 hours in argon or in
vacuum prior to the aforementioned tensile test in liquid cesium.
The tensile samples evaluated were (A) uncoated
Zircaloy-2 , (B) copper coated Zircaloy-2 and (C)
Zircaloy-2 coated with copper and having an intermediate
boundary layer of zirconium oxide between the copper and the
l zircaloy-2 substrate. The following table shows the results
¦25 obtained, where "yes" under "Heat Treatment" indicates that
th- tenslle sample was exposed 2 1/4 hours to a temperature
of 580C in argon or in vacuo prior to the Instron tensile
test.
! :~
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RD-9257
Heat Treatment Plastic Strain at Fracture
A No 0% ..
Yes % ;
B No 1.5%
Yes % ,
C No 1.5%
Yes 3.8%
The above results establish that Zircaloy-2 tensile
sample ~C) coated with copper and with an intermediate boundary
layer of zirconium oxide exhibited the largest plaqtic strain
at fracture. Surprisingly, the 3.8~ plastic strain at fracture
was even larger under the hostile environment of liquid cesium
~aturated with cadmium after heat treatment as compared
to the plastic strain at fracture of the tensile sample which
had not been heat treated. These technical facts would suggest
that a nuclear fuel element made in accordance with the present
invention under actual service conditions over an extended
period of time would exhibit a superior resistance to failure.
The zirconium cladding would resist embrittlement to a greater
extent since it would be protected by the copper barrier which
in turn would be prevented by the zirconium oxide barrier from
diffusing into the zirconium substrate. Those ~killed in the
, art also know that even a 1% plastic strain at fracture would
,l lndicate resi6tance to cracting of a significant degree. A1BO
' significant i8 the failure exhibited by the (B) tensile sample
protected only by a copper baxrier after heat treatment. The
dif~usion of copper into the zirconium substrate when heated
i to 580C resulted in embrittlement and failure as indicated
by the 0~ pla~tic 3train at fracture since there wa~ no zircon-
ium oxidet barrier.
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RD--9257
Example 2.
The procedure of Example 1 was repeated except that
ins~ead of etching the zirconium ~ube prior to oxidation, a
1 x 1.5 cm flat coupon was grit blasted by mechanical attrition
with aluminum oxide grit of 90 mesh size for 10 seconds. The
grit blasted coupon was then oxidized at 400C for 24 hours in
accordance with the procedure of Example 1.
Example 3.
The procedure of Example 2 was repeated except a
Zircaloy-2 tube was used in place of the flat coupon. Surface
roughening was achieved by roller milling, using as the roller,
an aluminum oxide tube having a 0.31 inch OD and 0.28 inch ID.
The aluminum oxide tube was filled with mercury to give it
added weight and placed inside the Zircaloy-2 tube along with
wet aluminum oxide grit, 90 mesh size as previously describ~d.
The tube was rolled with the ends stoppered to prevent loss of
the grit and water for 64 hours at 128 RPM. The tube was then
washed with distilled water and surface oxidized at 400C for
24 hour~ as previously described.
~ he above oxidized samples were then activated in
accordance with the procedure of Example 1 followed by electro-
le~s plating with copper. The resulting Zircaloy-2 samples
re~embled each other in appearance as well as resembling the
Zircaloy-2 tube coated with copper and zirconium oxide as
described in Example 1.
Although the above examples are directed to only a
few of the very many variables which can be used in the method
of the presant invention to provide a variety of useful nuclear
fuel elements and cladding for co~taining nuclear fuel, it
should be understood that a much broader variety of materials
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RD-9257
; and procedures can be utilized as set forth in the description
preceding these examples.
-22-
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