CA1114077A

CA1114077A - Nuclear fuel element having a composite coating

Info

Publication number: CA1114077A
Application number: CA312,786A
Authority: CA
Inventors: Lawrence H. King; Willard T. Grubb
Original assignee: General Electric Co
Current assignee: General Electric Co
Priority date: 1978-10-05
Filing date: 1978-10-05
Publication date: 1981-12-08

Abstract

Abstract of the Disclosure A nuclear fuel element consisting of a zirconium or ziroonium alloy container and nuclear fuel pellets is provided for use in the core of a nuclear reactor. The zirconium or zirconium alloy container has an inner coating of copper in proximity to the nuclear fuel, and is separated from the zir-conium or zirconium alloy by an intermediate zirconium oxide diffusion barrier layer. The copper layer and the intermediate zirconium oxide diffusion barrier of the composite cladding reduce perforations or failure in the zirconium or zirconium alloy cladding substrate caused by stress corrosion cracking or metal embrittlement. Good bonding of the copper to the oxided zirconium and zirconium alloy prevents scaling of copper from the composite cladding during the loading of the fuel element with fuel pellets.

Description

~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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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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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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~ A~ RD-9257 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 J~

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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 B

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:~?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;
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' 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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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~!
-, ., 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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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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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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The copper layer is more resistant to the deleterious ,~,~
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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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RD-9257 ',', 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, ',~
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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,.;
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; ;~ 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~

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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 ' ' .

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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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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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1~4~'7q 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 1' ' .

.:

4~7~

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 .' :

: , - . ~ . . . , . -:. . - --, -- . ~.. . . . . .. . . . .

~1~4~m 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.
! :~

, ..

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.

, .

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 ~1 ', ' . , j.;, ;~) il t~t .~ t'll~'i! li'~i ,;" tt'tJ~ Ti~iit~ 'J~ 't''i~:r )?it ': J ~;!i;iil i't-' ; and procedures can be utilized as set forth in the description preceding these examples.

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Claims

The embodiments of the invention in which an exclu-sive property or privilege is claimed are defined as follows:

1. An elongated container for nuclear fuel material, said container comprising zirconium and having an inner surface coated with a metal and a zirconium oxide diffusion barrier between the zirconium container inner surface and the metal coating which is selected from the class consisting of copper, nickel, iron and alloys thereof.

2. The container of claim 1, in which the zirconium container comprises a zirconium alloy.

3. The container of claim 1 or 2, in which the metal coating comprises copper.

4. A nuclear fuel element comprising:
(a) a central core of nuclear fuel material, and (b) an elongated container for the nuclear fuel material, said container comprising zirconium and having an inner surface coated with a metal and a zirconium oxide diffusion barrier between the zirconium container inner surface and the metal coating which is selected from the class consisting of copper, nickel, iron and alloys thereof.

5. The nuclear fuel element of claim 4, further comprising a cavity and a nuclear fuel material retaining means in the form of a helical member positioned in the cavity.

6. The nuclear fuel element of claim 4, in which the metal coating comprises copper.

7. The nuclear fuel element of claim 4, in which the zirconium container comprises a zirconium alloy.

8. The nuclear fuel element of claim 4, in which the nuclear fuel material is selected from the group consisting of uranium compounds, plutonium compounds and mixtures thereof.

9. The nuclear fuel element of claim 4, in which the nuclear fuel material comprises uranium dioxide.

10. The nuclear fuel element of claim 4, in which the nuclear fuel material is a mixture comprising uranium dioxide and plutonium dioxide.