CA1198231A - Zirconium alloy barrier having improved corrosion resistance - Google Patents
Zirconium alloy barrier having improved corrosion resistanceInfo
- Publication number
- CA1198231A CA1198231A CA000423387A CA423387A CA1198231A CA 1198231 A CA1198231 A CA 1198231A CA 000423387 A CA000423387 A CA 000423387A CA 423387 A CA423387 A CA 423387A CA 1198231 A CA1198231 A CA 1198231A
- Authority
- CA
- Canada
- Prior art keywords
- zirconium
- zirconium alloy
- composite cladding
- substrate
- liner
- Prior art date
- Legal status (The legal status is an assumption and is not a legal conclusion. Google has not performed a legal analysis and makes no representation as to the accuracy of the status listed.)
- Expired
Links
Classifications
-
- B—PERFORMING OPERATIONS; TRANSPORTING
- B32—LAYERED PRODUCTS
- B32B—LAYERED PRODUCTS, i.e. PRODUCTS BUILT-UP OF STRATA OF FLAT OR NON-FLAT, e.g. CELLULAR OR HONEYCOMB, FORM
- B32B15/00—Layered products comprising a layer of metal
- B32B15/01—Layered products comprising a layer of metal all layers being exclusively metallic
-
- C—CHEMISTRY; METALLURGY
- C22—METALLURGY; FERROUS OR NON-FERROUS ALLOYS; TREATMENT OF ALLOYS OR NON-FERROUS METALS
- C22C—ALLOYS
- C22C16/00—Alloys based on zirconium
-
- G—PHYSICS
- G21—NUCLEAR PHYSICS; NUCLEAR ENGINEERING
- G21C—NUCLEAR REACTORS
- G21C3/00—Reactor fuel elements and their assemblies; Selection of substances for use as reactor fuel elements
- G21C3/02—Fuel elements
- G21C3/04—Constructional details
- G21C3/16—Details of the construction within the casing
- G21C3/20—Details of the construction within the casing with coating on fuel or on inside of casing; with non-active interlayer between casing and active material with multiple casings or multiple active layers
-
- Y—GENERAL TAGGING OF NEW TECHNOLOGICAL DEVELOPMENTS; GENERAL TAGGING OF CROSS-SECTIONAL TECHNOLOGIES SPANNING OVER SEVERAL SECTIONS OF THE IPC; TECHNICAL SUBJECTS COVERED BY FORMER USPC CROSS-REFERENCE ART COLLECTIONS [XRACs] AND DIGESTS
- Y02—TECHNOLOGIES OR APPLICATIONS FOR MITIGATION OR ADAPTATION AGAINST CLIMATE CHANGE
- Y02E—REDUCTION OF GREENHOUSE GAS [GHG] EMISSIONS, RELATED TO ENERGY GENERATION, TRANSMISSION OR DISTRIBUTION
- Y02E30/00—Energy generation of nuclear origin
- Y02E30/30—Nuclear fission reactors
Abstract
ZIRCONIUM ALLOY BARRIER HAVING
IMPROVED CORROSION RESISTANCE
ABSTRACT OF THE DISCLOSURE
A nuclear fuel element for use in the core of a nuclear reactor which has a composite cladding container having a substrate and a dilute zirconium alloy liner bonded to the inside surface of the substrate. The dilute zirconium alloy liner forms about 1 to about 20 percent of the thickness of the cladding and comprises from about 0.1% to about 0.5%
by weight niobium and preferably from about 0.2% to about 0.4% by weight niobium, the balance being zirconium. The dilute zirconium alloy liner shields the substrate from impurities or fissin products from the nuclear fuel material and protects the substrate from stress corrosion and stress cracking.
The dilute zirconium alloy liner displays greater corrosion resistance, especially to oxidation by hot water and steam than unalloyed zirconium. The substrate material is selected from conventional cladding materials, and preferably is a zirconium alloy having a higher alloy content than the dilute zirconium alloy liner.
IMPROVED CORROSION RESISTANCE
ABSTRACT OF THE DISCLOSURE
A nuclear fuel element for use in the core of a nuclear reactor which has a composite cladding container having a substrate and a dilute zirconium alloy liner bonded to the inside surface of the substrate. The dilute zirconium alloy liner forms about 1 to about 20 percent of the thickness of the cladding and comprises from about 0.1% to about 0.5%
by weight niobium and preferably from about 0.2% to about 0.4% by weight niobium, the balance being zirconium. The dilute zirconium alloy liner shields the substrate from impurities or fissin products from the nuclear fuel material and protects the substrate from stress corrosion and stress cracking.
The dilute zirconium alloy liner displays greater corrosion resistance, especially to oxidation by hot water and steam than unalloyed zirconium. The substrate material is selected from conventional cladding materials, and preferably is a zirconium alloy having a higher alloy content than the dilute zirconium alloy liner.
Description
~3~1~3~
ZIRCONIUM ALLOY BARRIER HAVING
IMPROVED_CORROSI'ON RESIST~NCE
Field o'f the'Invention This invention relates broadly to an improvement in nuclear fuel elements for use in the core of nuclear fission reactors and, more particularly, to an improved nuclear fuel element havlng a composite cladding container having a metal liner of dilute ~irconium alloy comprising zirconium and niobium bonded to the inside surface of a zirconium alloy cladding substrate.
Background'of''the- Inve'nti'on Nuclear reactors are presently being designed, constructed, and operated in which -the nuclear fuel is contained in fuel elements which can have various geometric shapes, such as plates, tubes, or rods. The fuel material is usually enclosed in a corrosion-resistant, non-reactive, heat conductive container or cladding. The fuel elements are assembled together in a lattice at fixed distances from each other in a coolan-t flow channel or region forming a fuel assem~ly, and sufficient fuel assemblies are combined to form -the nuclear fission chain reacting assembly or reac-tor core capable of a self-sustained fission reaction.
The core, in -turn, is enclosed within a reac-tor vessel -through which a coolant is passed.
The cladding serves several purposes and -two primary purpose, are: first, to prevent contact and chemical reactions between the nuclear fuel and the 3g~
24-NT-O~a~C13 coolant or the moderator i~ a moderator is present, or both i~ both the coolant and the moderator are present;
and second; to prevent the radioactive fission products, some of which are gases, from being released from the fuel into the coolant or the moderator or both iE both the coolant and the moderator are present. Common cladding materials are stainless steel~ aluminum and its alloys, zirconium and its alloys, niobium (columbium), certain magnesium alloys, and others.
The failure of the cladding, i.e., a loss ~ the leak tightness, can con~aminate the coolant or moderator and the associated systems with long-lived radioactive products to a degree which interferes with plant operation.
Problems have been encountered in the manufacture and in the operation of nuclear fuel elements which employ certain metals and alloys as the cladding material due to mechanical or chemical reactions of these cladding materials under certain circumstances.
Zirconium and its alloys, under normal circumstances, make excellent nuclear fuel c]addings since they have low neutron absorption cross-sections and at temperatures below about 750F (about 398 C) are strong, ductile, extremely stable and relatively non-reactive in the presence of demineralized water or steam which are commonly used as reactor coolants and moderators.
~ owever, ~uel element performance has revealed a problem with the brittle splitting of the cladding due to the combined interactions between the nuclear fuel, the cladding and the fission products produced during nuclear fission reactions. It has been discovered that this ~Indesirable perEormance is promoted by localized mechanical stresses due to fuel cladding differential expansion o:E fuel and cladding (stresses in the cladding are concentrated at cracks in the nuclear fuel).
Corrosive fission products are released ~rom the nuclear fuel and are present at the intersection of the uel cracks with the cladding surface. Such fission products are created in the nuclear fuel during the fission chain reaction during operation of a nuclear reactor.
The localized stress is exaggerated by high friction between the fuel and the cladding.
Within the confines of a sealed fuel element, hydrogen gas can be generated by the slow reaction between the cladding and residual water inside the cladding. This hydrogen gas may build up to levels which, under certain conditions, can result in localized hydriding of the cladding with concurrent local deterioration in the mechanieal properties of the cladding. The cladding is also adversely affected by such gases as oxygen, nitrogen, carbon monoxide, and carbon dioxide over a wide range of temperatures.
The zireonium cladding of a nuclear fuel element is exposed to one or more of the gases listed above and fission products during irradiation in a nuclear reactor and this occurs in spite of the fact that these gases may not be present in the reactor coolant or moderator, and further may have been excluded as far as possible from the ambient atmosphere during manufacture of the cladding and the fuel element.
Sintered refractory and eeramic compositions, such as uranium dioxide and other compositions used as nuelear fuel, release measurable quantities of the aforementioned gases upon heating, such as during fuel element manufacture and further release fission produets during irradiation. Partieulate refractory and eeramic compositions, such as uranium dioxide powder and other powders used as nuclear fuel~ have been known to release even larger quantities of the aforementioned gases during irradiation. These released gases are capable of reacting with the zireonium eladding con-taining the nuele~r fuel.
3~
Thus, in liyht of the foregoing, i-t has been found desirable to minimize attack of the cladding from water, water vapor and other yases, especially hydrogen, which are reactive with the cladding from inside the fuel element throughou-t the time the fuel element is used in the operation of nuclear power plants. One such approach has been to find materials which will chemically react rapidly with -the water, water vapor and other gases to eliminate these from the interior of the cladding. Such materials are cal]ed getters.
Another approach has been to coat the nuclear fuel material with any of a variety of materials to prevent moisture coming in contact with the nuclear fuel material.
The coating of nuclear fuel material introduces reliability problems in that achieving uniform coatings free of faults is difficulto Further, the deterioration of the coating can introduce problems with the long-lived performance of the nuclear fuel material.
General Electric Atomic Power Document 4555 of February 1964, at GE NEB0 Library, 175 Curtner Ave., San Jose, Calif. 95125, discloses a composite cladding of a zirconium alloy with an inner lining of stainless steel metallurgically bonded to the zirconium alloy, and the composite cladding is fabricated by extrusion of a hollow billet of the zirconium alloy having an inner lining of stainless steel. This cladding has ihe disadvantages that the stainless steel develops brittle phases, and -the s-tainless steel layer involves a neutron absorption penalty of about ten to fifteen ti.mes the penal-ty for a zirconium alloy of the same thickness.
U.S~ Patent No. 3,502,5~9, issued March 24, 1970 to Charveriat, discloses a method for protecting zirconium and its alloys by the electrolytic deposition oE chronium to provide a composite material useful Eor nucJear reactors. A method for electrolytic deposition of copper on Zircaloy-2 surfaces and
ZIRCONIUM ALLOY BARRIER HAVING
IMPROVED_CORROSI'ON RESIST~NCE
Field o'f the'Invention This invention relates broadly to an improvement in nuclear fuel elements for use in the core of nuclear fission reactors and, more particularly, to an improved nuclear fuel element havlng a composite cladding container having a metal liner of dilute ~irconium alloy comprising zirconium and niobium bonded to the inside surface of a zirconium alloy cladding substrate.
Background'of''the- Inve'nti'on Nuclear reactors are presently being designed, constructed, and operated in which -the nuclear fuel is contained in fuel elements which can have various geometric shapes, such as plates, tubes, or rods. The fuel material is usually enclosed in a corrosion-resistant, non-reactive, heat conductive container or cladding. The fuel elements are assembled together in a lattice at fixed distances from each other in a coolan-t flow channel or region forming a fuel assem~ly, and sufficient fuel assemblies are combined to form -the nuclear fission chain reacting assembly or reac-tor core capable of a self-sustained fission reaction.
The core, in -turn, is enclosed within a reac-tor vessel -through which a coolant is passed.
The cladding serves several purposes and -two primary purpose, are: first, to prevent contact and chemical reactions between the nuclear fuel and the 3g~
24-NT-O~a~C13 coolant or the moderator i~ a moderator is present, or both i~ both the coolant and the moderator are present;
and second; to prevent the radioactive fission products, some of which are gases, from being released from the fuel into the coolant or the moderator or both iE both the coolant and the moderator are present. Common cladding materials are stainless steel~ aluminum and its alloys, zirconium and its alloys, niobium (columbium), certain magnesium alloys, and others.
The failure of the cladding, i.e., a loss ~ the leak tightness, can con~aminate the coolant or moderator and the associated systems with long-lived radioactive products to a degree which interferes with plant operation.
Problems have been encountered in the manufacture and in the operation of nuclear fuel elements which employ certain metals and alloys as the cladding material due to mechanical or chemical reactions of these cladding materials under certain circumstances.
Zirconium and its alloys, under normal circumstances, make excellent nuclear fuel c]addings since they have low neutron absorption cross-sections and at temperatures below about 750F (about 398 C) are strong, ductile, extremely stable and relatively non-reactive in the presence of demineralized water or steam which are commonly used as reactor coolants and moderators.
~ owever, ~uel element performance has revealed a problem with the brittle splitting of the cladding due to the combined interactions between the nuclear fuel, the cladding and the fission products produced during nuclear fission reactions. It has been discovered that this ~Indesirable perEormance is promoted by localized mechanical stresses due to fuel cladding differential expansion o:E fuel and cladding (stresses in the cladding are concentrated at cracks in the nuclear fuel).
Corrosive fission products are released ~rom the nuclear fuel and are present at the intersection of the uel cracks with the cladding surface. Such fission products are created in the nuclear fuel during the fission chain reaction during operation of a nuclear reactor.
The localized stress is exaggerated by high friction between the fuel and the cladding.
Within the confines of a sealed fuel element, hydrogen gas can be generated by the slow reaction between the cladding and residual water inside the cladding. This hydrogen gas may build up to levels which, under certain conditions, can result in localized hydriding of the cladding with concurrent local deterioration in the mechanieal properties of the cladding. The cladding is also adversely affected by such gases as oxygen, nitrogen, carbon monoxide, and carbon dioxide over a wide range of temperatures.
The zireonium cladding of a nuclear fuel element is exposed to one or more of the gases listed above and fission products during irradiation in a nuclear reactor and this occurs in spite of the fact that these gases may not be present in the reactor coolant or moderator, and further may have been excluded as far as possible from the ambient atmosphere during manufacture of the cladding and the fuel element.
Sintered refractory and eeramic compositions, such as uranium dioxide and other compositions used as nuelear fuel, release measurable quantities of the aforementioned gases upon heating, such as during fuel element manufacture and further release fission produets during irradiation. Partieulate refractory and eeramic compositions, such as uranium dioxide powder and other powders used as nuclear fuel~ have been known to release even larger quantities of the aforementioned gases during irradiation. These released gases are capable of reacting with the zireonium eladding con-taining the nuele~r fuel.
3~
Thus, in liyht of the foregoing, i-t has been found desirable to minimize attack of the cladding from water, water vapor and other yases, especially hydrogen, which are reactive with the cladding from inside the fuel element throughou-t the time the fuel element is used in the operation of nuclear power plants. One such approach has been to find materials which will chemically react rapidly with -the water, water vapor and other gases to eliminate these from the interior of the cladding. Such materials are cal]ed getters.
Another approach has been to coat the nuclear fuel material with any of a variety of materials to prevent moisture coming in contact with the nuclear fuel material.
The coating of nuclear fuel material introduces reliability problems in that achieving uniform coatings free of faults is difficulto Further, the deterioration of the coating can introduce problems with the long-lived performance of the nuclear fuel material.
General Electric Atomic Power Document 4555 of February 1964, at GE NEB0 Library, 175 Curtner Ave., San Jose, Calif. 95125, discloses a composite cladding of a zirconium alloy with an inner lining of stainless steel metallurgically bonded to the zirconium alloy, and the composite cladding is fabricated by extrusion of a hollow billet of the zirconium alloy having an inner lining of stainless steel. This cladding has ihe disadvantages that the stainless steel develops brittle phases, and -the s-tainless steel layer involves a neutron absorption penalty of about ten to fifteen ti.mes the penal-ty for a zirconium alloy of the same thickness.
U.S~ Patent No. 3,502,5~9, issued March 24, 1970 to Charveriat, discloses a method for protecting zirconium and its alloys by the electrolytic deposition oE chronium to provide a composite material useful Eor nucJear reactors. A method for electrolytic deposition of copper on Zircaloy-2 surfaces and
2~l subsequent heat treatment for the purpose of obtaining surface diffusion of -the electrolytically deposited metal is presented in Ene-rgia Nucleare, Volume 11, No~ 9 (September, 196~) at pages 505-508. In Stability and Compatibility of Hydrogen Barrie~ plied to _ . _ Zirconium Alloys, by ~. ~rossa et al (European Atomic Energy Communi-ty~ ~oint Nuclear Research Center, EUR 4098e, 1969), methods of deposition of different coatings and their efficiency as hydrogen diffusion barriers are described along wi-th an Al-Si coating as the most promising barrier against hydrogen diffusion.
Methods for electroplating nickel on zirconium and zirconium-tin alloys and heat treating these alloys to produce alloy-diffusion bonds are disclosed in Electropl~a'ting o'n Zirconium 'and Zirconlum=Tin, by W.C. Schickner et al (BMI-757, Technical Information Service, 19521.
U.S. Patent No. 3,625,821, issued December 7, 1971, to Ricks, presents a fuel element for a nuclear reactor having a fuel cladding tube with the inner surface of the tube being coated with a metal of low neutron capture cross-section such as nickel and having finely dispersed particles of a burnable poison disposed therein. Reac'tor Devel'opment Program Progress Report of August, 1973 (ANL-RDP-l9) discloses a chemical getter arrangement of a sacrificial layer of chromium on the inner surface of a stainless steel cladding.
Another approach has been to introduce a barr:ier between the nuclear fuel material and the cladding holding the nuclear fuel material as disclosed in U.S. Patent No. 3,230,150, :issued January 18, 1966 to Martin et al, tcopper foil); German Patent Publication DAS 1,238,115 (titanium layer); U.S. Patent No.
Methods for electroplating nickel on zirconium and zirconium-tin alloys and heat treating these alloys to produce alloy-diffusion bonds are disclosed in Electropl~a'ting o'n Zirconium 'and Zirconlum=Tin, by W.C. Schickner et al (BMI-757, Technical Information Service, 19521.
U.S. Patent No. 3,625,821, issued December 7, 1971, to Ricks, presents a fuel element for a nuclear reactor having a fuel cladding tube with the inner surface of the tube being coated with a metal of low neutron capture cross-section such as nickel and having finely dispersed particles of a burnable poison disposed therein. Reac'tor Devel'opment Program Progress Report of August, 1973 (ANL-RDP-l9) discloses a chemical getter arrangement of a sacrificial layer of chromium on the inner surface of a stainless steel cladding.
Another approach has been to introduce a barr:ier between the nuclear fuel material and the cladding holding the nuclear fuel material as disclosed in U.S. Patent No. 3,230,150, :issued January 18, 1966 to Martin et al, tcopper foil); German Patent Publication DAS 1,238,115 (titanium layer); U.S. Patent No.
3,212,988,issuecl October 19, 1965 to Ringot et al (sheath of zirconium, aluminum or beryllium); U.S.
Patent No. 3,018,238, issued ~anuary 23, 1962 to 3~
layer et al (barrier of crystalline carbon between the U2 and the zirconium alloy cladding); and U.S. Patent No. 3,088,893~ issued May 7, 1963 to Spalaris (stainless steel foil). While the barrier concept proves pro~ising~ 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 properties of the cladding), or the nuclear fission reaction (e.g., by acting as neutron absorbers).
None of the listed references disclose solutions to the problem of localized chemical-mechanical interactions between the nuclear fuel and the cladding.
Further approaches to the barrier concept are disclosed in U.S. Patent No. 3,969,186, issued July 13, 1976 to Thompson, (refractory metal such as molybdenuml tungsten, rhenium, niobium and alloys thereof in the ~orm of a tube or foil of ~ingle or multiple layers or a coating on the internal surface of the cladding), and U.S. Patent No. 3,925,151, issued December 9, 1975 to Klepfer (liner of zirconium, niobium, or alloys thereof between the nuclear fuel and the cladding with a coating of a high lubricity material between liner and the cladding).
U.S. Patent No. 4,045,288, issued August 30, 1977 to Armijo discloses a composite cladding of a zirconium alloy substrate with a metal barrier metallurgically bonded to the substrate and an inner layer of zirconium alloy metallurgically bonded to the metal barrier. ~he barrier is selected from a group of niobium, aluminum, copper, nickel, stainless steel, and iron. The buri.ed metal barrier reduces corrosion due to fission product:s and corrosive gases, but is subject to stress corrosion crackin~ and liquid metal embrit-tlement.
3~
U.S. Patent No, 4,200,~9~, issued ~pril 29, 1980 -to Armijo et al discloses a composite cladding of a zirconium alloy substrate with a sponge zirconium liner.
The soft zirconium liner minimizes localized stress, and reduces stress corrosion cracking and liquid metal embrittlement, but is subject to loss due to honing and the like during fabrication and due to oxidation.
Furthermore, if a breach in the cladding should occur, allowing water and/or steam to enter the fuel rod, the zirconium liner tends to oxidize rapidly.
Accordingly~ it has remained desirable to develop nuclear fuel elements minimizing the problems discussed above.
Summary of the Invention . . . _ A particularly effective nuclear fuel element for use in the core of a nuclear reactor has a composite cladding having an inner liner of dilute zirconium alloy metallurgically bonded to the inside surface of the substrate. The dilute zirconium alloy comprises from about 0.1% to about 0.5~ by weight niobium and preferably from about 0.2~ to about 0.4~ by weight niobium, the balance being zirconium.
The substrate of the cladding is completely unchanged in design and function from previous practice for a nuclear reactor and is selected from conventional cladding materials such as zirconium alloys. A zirconium alloy cladding substrate has a higher alloy content than the dilute zirconium alloy liner.
The clilute zirconium alloy liner forms a continuous shield be-tween the suhstrate and the nuclear fuel material held in the cladding, as well as shielding the zirconium alloy or other substrate cladding from fission products and gases. The dilute 3S zirconium a]loy liner forms Erom aboutl to about 20 percent of the thickness of the cladding.
;23~
24~NT-04493 The liner remains soft, relative to the substrate, during irradiation and minimizes localized strain inside the nuclear fuel element~ thus serving to protect the cladding from stress corrosion cracking or liquid metal embrittlement. The dllute zirconium alloy liner gives a preferential reaction site for reaction with volatile impurities or fission products present inside the nuclear fuel element and~
in this manner, serves to protect the cladding substrate from attack by the volatile impurities or fission products.
This invention has a striking advantage that the substxate of the cladding is protected from stress corrosion cracking and liquid metal embrittlement, in addition to contact with fission products, corrosive gases, etc., by the dilute zirconium alloy liner and the lin~r does not introduce any appreciable neutron capture penalties, heat transfer penalties, or fuel/
liner incompatibility problems. In addition, the liner provides superior resistance to hot water or steam o~idation as compared to unalloyed zirconium in the event of a breach in the cladding.
Description of the; Drawings _ ... . . .... _ _ _ The foregoing and other objects of this invention will become apparent to persons skilled in the art from reading the following specification and the appended claims with reference to the accompanying drawings, wherein:
FIG. 1 is a partial cukaway sectional view of a nuclear fuel assembly containing nuclear fuel elements constructed acc:ording to the teaching of this invention;
and FIG. 2 is an enlarged transverse cross-sectional view of the nuclear fuel element in FIG. 2 illustrating the teaching oE this invention.
3~
24-NT~04493 _.9_ Description of the Invention Referring now more particularly to FIG. 1, there is shown a partially cutaway sectional view of a nuclear fuel assembly 10. This ~uel assembly 10 consists of a tubular flow channel 11 of generally square cross section provided at its upper end with a lifting bail 12 and at :its lower end with a. nose piece (not shown due to the lower portion of assembly 10 being omitted). The upper end of channel 11 is open at outlet 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 the channel 11 and supported therein by means of an upper end plate 15 and a lower end plate (not. shown due to the lower portion beinc3 omitted). The liquid coolant ordinarily enters through the openings in -the lower end of the nose piece, passes upwardly around fuel elements 14, and discharges through the upper outlet 13 at an elevated temperature in a partially vaporized condition for boiling reactors or in an unvaporized condition for pressurized reactors.
The nuclear fuel elements or rods 14 are sealed at their ends by means of end plugs 18 welded to the cladding 17, which may include studs 19 to facilitate the mounting of the fuel rod in the assembly.
A void space or plenum 20 is provided at one end of the element to permit longit.udinal expansion of the fuel material and accumulat.ion of gases released from the ~uel material. A nuclear fuel material retainer means 24 in the form of a helical member i.s positioned within space 20 to provide restrain-t against the axial movement of the pellet colllmn, espec.ially duri.ng handling and transportation of the fuel element.
The fuel element is designed to provide an excellent thermal contact between the cladding and the fuel material, a minimum of parasitic neutron 23~L
absorption and resistance to bowing and vibration which is occasionally caused by flow of the coolant at high velocity.
A nuclear fuel element or rod 14 constructed according to the teachings of this invention is shown in a partial sectior) in FIG. 1. The fuel element includes a core or central cylindrical portion of nuclear fuel material 16, here shown as a fertile material positioned within a structural cladding or container 17. In some cases, the fuel pellets may be of various shapes such as cylindrical pellets or spheres and, in other cases, different fuel forms such as a particulate fuel may be used.
The physical form of the fuel is immaterial to this invention. Various nuclear fuel materials may be used including uranium compounds, plutonium compounds, thorium compounds, and mixtures thereof. A preferred fuel is uranium dioxide or a mi~ture comprising uranium dioxide and plutonium dioxide.
Referring now to FIG. 2~ the nuclear fuel material 16 forming the central core of the fuel element 14 is surrounded by a cladding 17 which, in this invention, is also referred to as a composite cladding container.
The composite cladding container encloses the fissile core so as to leave a gap 2~ between the core and the cladding during use in a nuclear reactor. The composite cladding container has an external substrate 21 selected from conventional cladding materials such as a stainless steel and zirconium alloys and, in a preferred embodiment of this invention, the substrate is a zirconium alloy such as Zircaloy-2.
The substrate 21 has metallurgically bonded on the inside circumference thereof a dilute zirconium alloy liner 22 so that the zirconium alloy liner forms a continuous shield of the substrate from the nuclear fuel material 16 inside the composite cladding.
The dilute zirconium alloy liner preferably forms from about 1 to about 20 percent of the thickness of the cladding. As used herein, dilute zirconium alloy means a zirconium alloy with an alloy content sufficiently low to display greater ductility than the substrate material.
A dilute zirconium alloy liner forminy less than about 1 percent of the thickness of the cladding would be difficult to achieve in commercial production, and a dilute zirconium alloy liner forming more than 20 percent of the thickness of the cladding provides no additional benefit for the added thickness. Further a liner more than about 20% of the thickness of the cladding means a concomitant reduction in thickness of the substrate and possible weakening of the cladding.
The dilute zirconium alloy liner serves as a prefarential reaction site for gaseous impurities and fission products and protects the substrate portion of the claddin~ from contact and reaction with such impurities and fission products and ameliorates the occurence of localized stresses.
The dilute zirconium alloy liner comprises from about 0.1% to about 0.5~ by weight niobium, and preferably from about 0.2% to 0.4~ by weight niobium, the balance being zirconium.
A dilute zirconium alloy comprising from about 0.1~ to about 0.5% by weight niobium exhibits increased resistance to corrosion or oxidation by contact with hot water and steam as compared with unalloyed zirconium. Dilute zirconium alloys comprising less than about 0.1~ by weight niobium exhibit no significant increase in corrosion resistance and are difficult to achieve in commercial production.
Niobium is soluble in zirconium in the range from about 0.1% to above 0.5~ by we:ight. While some 3~
24-NT-0~493 solid solution strengthening does occu:r due to the solubility of niobium, the amount ls sufficiently low to enable the dilute zirconium alloy to resist fuel rod failure from pellet-cladding interaction.
Above about 0.5% by weight, niobium forms precipitates which increase the strength of the zirconium alloy and significantly decreases i-ts ductility or plasticity. An upper limit of about 0.5% by weight is therefore preferred to insure that the dilute zirconium alloy remains highly ductile and resistant to radiation hardening which enables the dilute zirconium alloy liner, after prolonged irradiation, to maintain desirable structural properties such as yield strength and hardness at levels considerably lower than those of conventional zirconium alloys. In effect the dil.ute zirconium alloy liner does not harden as much as conventional zirconium alloys when subjected to irradiation and thisl together with its initial low yield strength, 20. enables the dilute zirconium alloy liner to deform plastically and relieve pellet-induced stresses in the cladding. Such stresses can be brought about, for example, by swelling of the pellets of nuclear fuel at reactor operating temperature (300 C to 350C) so that the pellet comes into contact with the cladding.
A dilute zirconium alloy liner comprising about 0.2% to about 0.~% by weight niobium is particularly preferred because a dilute zirconium alloy in this range exhibits the preferred combination of corrosion resistance and ductility. ~elow about 0.2% nic~bium .in the zirconium, the corrosion resis-tance beyins to approach that of sponge zirconium.
It is particularly preferred to h~e a ma.ximum niobium content of about 0.4% to assure t.hat the liner does not lie outside the solid soluhility range ~38~3~
regardless of thermal exposure of the fuel rod over long periods of time~ thereby assuring continued ductility.
The dilute zirconium alloy liner comprising about 0.1% to about 0.5~ by weight niobium, and preferably from about 0~2~ to about 0.4% by weight niohium and forming from about 1 to 20 percent of the thickness of the claddîng and preferably Erom about 5 to 15 percent of the cladding bonded to a conventional zirconium alloy substrate provides stress reduction while improving corrosion resistance, especially resistance to oxidation by hot water and steam in the event of a cladding breach.
The purity of the zirconium metal that is alloyed with niobium is important and serves to impart particular properties to the dilute zirconium alloy liner. Generally, there is at least 1000 parts per million (ppm~ by weight and less than 5000 ppm impurities in the zirconium metal and preferably less than 4200 ppm. Of these oxygen may vary up l:o about 1000 ppm.
The composite cladding of the nuclear fuel element of this invention has a dilute zirconium alloy liner metallurgically bonded to the substrate.
Metallographic examination shows that there is sufficient cross-diffusion between the substrate and the zirconium liner to form metallurgical bonds, but insuEficient cross-difEusion to significantly alloy with the dilute ~irconium alloy liner itselE.
Among the conventional zirconium alloys serving as suitable substrates are Zircaloy-2 and Zircaloy-4. Zircaloy-2 has on a weight basis about 1.5 percent tin; 0.12 percent iron; 0.09 percent chromium and 0~005 percent nickel and is extemsively employed in water--cooled reactors. Zircaloy-4 has less nickel than Zircaloy-2l but contains slightly more iron than Zircaloy-2. The composite cladding used 323~
in the nuclear fuel elements oE this invention can be fabricated by any of t~e following methods.
In one method/ a tube of -the dilute zirconium alloy liner material is inserted into a hollow billet of the material selected to be the substrate, and then the assembly is subjected to explosive bonding of the tube to the billet~ The composite is ex-truded using conventional tube shell extrusion at elevated temperatures of about 1000F to lA00F (about 538 C
to 760C). Then the extruded composite is subjected to a process involving conventional tube reduction until the desired size of cladding is achieved. The relative wall thickness of the hollow billet and the dilute zirconium alloy liner tube are selected to give the desired thickness ratios in the finished cladding tube.
In another method, a tube of the dilute zirconium alloy liner material is inserted into a hollow billet of the material selected to be the substrate, and then the assernbly is subjected to a heating step (such as at 750C for 8 hours) under compressive stress to assure good metal-to~metal contact and diffusion bonding between the tube and the billet. The diffusion bonded composite is extruded using conventional tube shell extrusion such as described above in the immedia-tely preceding paragraph. Then the extruded cornposite is subje!cted to a process involving conventional tube reduction until the desired size of cladding is achieved.
In still another method, a tube of the!
dilute zirconium alloy liner material is inserted into a hollow bille-t of the material selected to be the substrate~ and the assembly is extruded using conventional tube shell extrusion as described above.
Then the extrud,ed composite is subjected to a process involvingconvention~l tube reduction until the desired size of cladding is achieved.
24-NT-04~93 The foregoing processes of fabricating t:he composite cladding of this invention give economies over other processes used in fabricating cladding such as electroplating or vapor c1eposition. A nuclear fuel element can be forged by making a composite cladding container which is open at one end, the cladding container having a substrate and an inner dilute zirconium alloy liner metallurgically bonded to the inside surface of the substrate. The element is completed by fil.ling the composite cladding container with nuclear fuel material~ leaving a cavity at the open end~ inserting a nuclear fuel material retaining means into the cavity, applying an enclosure to the open end of the container leaving the cavity in communication with the nuclear fuel, and then bonding the end of the cladding container to said enclosure to form a tight seal therebetween.
The present invention offers several advantages promoting a long operating life ~or a rluclear fuel element, including the reduction of chemical interaction o~ the cladding, the minimization of localized stress on the æirconium alloy substrate portion of the cladding, the minimization of stress corrosion on the zirconi.um alloy substrate portion of the cladding, and the reduction of the probability of a splitting failure occurring in the zirconium alloy substrate.
In addi-tion to minimizing stress and stress corrosion on the substrate, the dilute zirconium alloy liner is resistant to oxidation by steam and hot water in theevent that the cladding is breached, whereas unalloyed zirconium oxidizes very rapidly under these conditions. The dilute zirconium alloy exh.ibits plasticity similar to unalloyed zirconium and provides the benefits thereof while also provid:ing increased resistance to corrosion, especially to oxidation by hot water and steam.
3~
24-NT-0~493 ~16-An important property ~ the composite cladding of this invention is that the foregoing improvements are achieved with no substan-tial additional neutron penalty. Such a cladding is readily accepted in nuclear reactors since the cladding would have no eutectic foxmation during a loss-of-coolant accident or an accident involving the dropping of a nuclear control rodA Furtherl the composite cladding has a very small heat transfer penalty in that there is no thermal barrier to transfer of heat such as results in the situation where a separate foil or liner is inserted in a fuel element. Also~ the composite cladding of this invention is inspectahle by conventional non-destructive testing methods during various stages of fabrication.
As will be apparent to those skilled in the art, various modifications and changes may be made in the invention described herein. It is accordingly the intention that the invention be construed in the broadestmanner within the spirit and scope as set forth in the accompanying claims.
Patent No. 3,018,238, issued ~anuary 23, 1962 to 3~
layer et al (barrier of crystalline carbon between the U2 and the zirconium alloy cladding); and U.S. Patent No. 3,088,893~ issued May 7, 1963 to Spalaris (stainless steel foil). While the barrier concept proves pro~ising~ 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 properties of the cladding), or the nuclear fission reaction (e.g., by acting as neutron absorbers).
None of the listed references disclose solutions to the problem of localized chemical-mechanical interactions between the nuclear fuel and the cladding.
Further approaches to the barrier concept are disclosed in U.S. Patent No. 3,969,186, issued July 13, 1976 to Thompson, (refractory metal such as molybdenuml tungsten, rhenium, niobium and alloys thereof in the ~orm of a tube or foil of ~ingle or multiple layers or a coating on the internal surface of the cladding), and U.S. Patent No. 3,925,151, issued December 9, 1975 to Klepfer (liner of zirconium, niobium, or alloys thereof between the nuclear fuel and the cladding with a coating of a high lubricity material between liner and the cladding).
U.S. Patent No. 4,045,288, issued August 30, 1977 to Armijo discloses a composite cladding of a zirconium alloy substrate with a metal barrier metallurgically bonded to the substrate and an inner layer of zirconium alloy metallurgically bonded to the metal barrier. ~he barrier is selected from a group of niobium, aluminum, copper, nickel, stainless steel, and iron. The buri.ed metal barrier reduces corrosion due to fission product:s and corrosive gases, but is subject to stress corrosion crackin~ and liquid metal embrit-tlement.
3~
U.S. Patent No, 4,200,~9~, issued ~pril 29, 1980 -to Armijo et al discloses a composite cladding of a zirconium alloy substrate with a sponge zirconium liner.
The soft zirconium liner minimizes localized stress, and reduces stress corrosion cracking and liquid metal embrittlement, but is subject to loss due to honing and the like during fabrication and due to oxidation.
Furthermore, if a breach in the cladding should occur, allowing water and/or steam to enter the fuel rod, the zirconium liner tends to oxidize rapidly.
Accordingly~ it has remained desirable to develop nuclear fuel elements minimizing the problems discussed above.
Summary of the Invention . . . _ A particularly effective nuclear fuel element for use in the core of a nuclear reactor has a composite cladding having an inner liner of dilute zirconium alloy metallurgically bonded to the inside surface of the substrate. The dilute zirconium alloy comprises from about 0.1% to about 0.5~ by weight niobium and preferably from about 0.2~ to about 0.4~ by weight niobium, the balance being zirconium.
The substrate of the cladding is completely unchanged in design and function from previous practice for a nuclear reactor and is selected from conventional cladding materials such as zirconium alloys. A zirconium alloy cladding substrate has a higher alloy content than the dilute zirconium alloy liner.
The clilute zirconium alloy liner forms a continuous shield be-tween the suhstrate and the nuclear fuel material held in the cladding, as well as shielding the zirconium alloy or other substrate cladding from fission products and gases. The dilute 3S zirconium a]loy liner forms Erom aboutl to about 20 percent of the thickness of the cladding.
;23~
24~NT-04493 The liner remains soft, relative to the substrate, during irradiation and minimizes localized strain inside the nuclear fuel element~ thus serving to protect the cladding from stress corrosion cracking or liquid metal embrittlement. The dllute zirconium alloy liner gives a preferential reaction site for reaction with volatile impurities or fission products present inside the nuclear fuel element and~
in this manner, serves to protect the cladding substrate from attack by the volatile impurities or fission products.
This invention has a striking advantage that the substxate of the cladding is protected from stress corrosion cracking and liquid metal embrittlement, in addition to contact with fission products, corrosive gases, etc., by the dilute zirconium alloy liner and the lin~r does not introduce any appreciable neutron capture penalties, heat transfer penalties, or fuel/
liner incompatibility problems. In addition, the liner provides superior resistance to hot water or steam o~idation as compared to unalloyed zirconium in the event of a breach in the cladding.
Description of the; Drawings _ ... . . .... _ _ _ The foregoing and other objects of this invention will become apparent to persons skilled in the art from reading the following specification and the appended claims with reference to the accompanying drawings, wherein:
FIG. 1 is a partial cukaway sectional view of a nuclear fuel assembly containing nuclear fuel elements constructed acc:ording to the teaching of this invention;
and FIG. 2 is an enlarged transverse cross-sectional view of the nuclear fuel element in FIG. 2 illustrating the teaching oE this invention.
3~
24-NT~04493 _.9_ Description of the Invention Referring now more particularly to FIG. 1, there is shown a partially cutaway sectional view of a nuclear fuel assembly 10. This ~uel assembly 10 consists of a tubular flow channel 11 of generally square cross section provided at its upper end with a lifting bail 12 and at :its lower end with a. nose piece (not shown due to the lower portion of assembly 10 being omitted). The upper end of channel 11 is open at outlet 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 the channel 11 and supported therein by means of an upper end plate 15 and a lower end plate (not. shown due to the lower portion beinc3 omitted). The liquid coolant ordinarily enters through the openings in -the lower end of the nose piece, passes upwardly around fuel elements 14, and discharges through the upper outlet 13 at an elevated temperature in a partially vaporized condition for boiling reactors or in an unvaporized condition for pressurized reactors.
The nuclear fuel elements or rods 14 are sealed at their ends by means of end plugs 18 welded to the cladding 17, which may include studs 19 to facilitate the mounting of the fuel rod in the assembly.
A void space or plenum 20 is provided at one end of the element to permit longit.udinal expansion of the fuel material and accumulat.ion of gases released from the ~uel material. A nuclear fuel material retainer means 24 in the form of a helical member i.s positioned within space 20 to provide restrain-t against the axial movement of the pellet colllmn, espec.ially duri.ng handling and transportation of the fuel element.
The fuel element is designed to provide an excellent thermal contact between the cladding and the fuel material, a minimum of parasitic neutron 23~L
absorption and resistance to bowing and vibration which is occasionally caused by flow of the coolant at high velocity.
A nuclear fuel element or rod 14 constructed according to the teachings of this invention is shown in a partial sectior) in FIG. 1. The fuel element includes a core or central cylindrical portion of nuclear fuel material 16, here shown as a fertile material positioned within a structural cladding or container 17. In some cases, the fuel pellets may be of various shapes such as cylindrical pellets or spheres and, in other cases, different fuel forms such as a particulate fuel may be used.
The physical form of the fuel is immaterial to this invention. Various nuclear fuel materials may be used including uranium compounds, plutonium compounds, thorium compounds, and mixtures thereof. A preferred fuel is uranium dioxide or a mi~ture comprising uranium dioxide and plutonium dioxide.
Referring now to FIG. 2~ the nuclear fuel material 16 forming the central core of the fuel element 14 is surrounded by a cladding 17 which, in this invention, is also referred to as a composite cladding container.
The composite cladding container encloses the fissile core so as to leave a gap 2~ between the core and the cladding during use in a nuclear reactor. The composite cladding container has an external substrate 21 selected from conventional cladding materials such as a stainless steel and zirconium alloys and, in a preferred embodiment of this invention, the substrate is a zirconium alloy such as Zircaloy-2.
The substrate 21 has metallurgically bonded on the inside circumference thereof a dilute zirconium alloy liner 22 so that the zirconium alloy liner forms a continuous shield of the substrate from the nuclear fuel material 16 inside the composite cladding.
The dilute zirconium alloy liner preferably forms from about 1 to about 20 percent of the thickness of the cladding. As used herein, dilute zirconium alloy means a zirconium alloy with an alloy content sufficiently low to display greater ductility than the substrate material.
A dilute zirconium alloy liner forminy less than about 1 percent of the thickness of the cladding would be difficult to achieve in commercial production, and a dilute zirconium alloy liner forming more than 20 percent of the thickness of the cladding provides no additional benefit for the added thickness. Further a liner more than about 20% of the thickness of the cladding means a concomitant reduction in thickness of the substrate and possible weakening of the cladding.
The dilute zirconium alloy liner serves as a prefarential reaction site for gaseous impurities and fission products and protects the substrate portion of the claddin~ from contact and reaction with such impurities and fission products and ameliorates the occurence of localized stresses.
The dilute zirconium alloy liner comprises from about 0.1% to about 0.5~ by weight niobium, and preferably from about 0.2% to 0.4~ by weight niobium, the balance being zirconium.
A dilute zirconium alloy comprising from about 0.1~ to about 0.5% by weight niobium exhibits increased resistance to corrosion or oxidation by contact with hot water and steam as compared with unalloyed zirconium. Dilute zirconium alloys comprising less than about 0.1~ by weight niobium exhibit no significant increase in corrosion resistance and are difficult to achieve in commercial production.
Niobium is soluble in zirconium in the range from about 0.1% to above 0.5~ by we:ight. While some 3~
24-NT-0~493 solid solution strengthening does occu:r due to the solubility of niobium, the amount ls sufficiently low to enable the dilute zirconium alloy to resist fuel rod failure from pellet-cladding interaction.
Above about 0.5% by weight, niobium forms precipitates which increase the strength of the zirconium alloy and significantly decreases i-ts ductility or plasticity. An upper limit of about 0.5% by weight is therefore preferred to insure that the dilute zirconium alloy remains highly ductile and resistant to radiation hardening which enables the dilute zirconium alloy liner, after prolonged irradiation, to maintain desirable structural properties such as yield strength and hardness at levels considerably lower than those of conventional zirconium alloys. In effect the dil.ute zirconium alloy liner does not harden as much as conventional zirconium alloys when subjected to irradiation and thisl together with its initial low yield strength, 20. enables the dilute zirconium alloy liner to deform plastically and relieve pellet-induced stresses in the cladding. Such stresses can be brought about, for example, by swelling of the pellets of nuclear fuel at reactor operating temperature (300 C to 350C) so that the pellet comes into contact with the cladding.
A dilute zirconium alloy liner comprising about 0.2% to about 0.~% by weight niobium is particularly preferred because a dilute zirconium alloy in this range exhibits the preferred combination of corrosion resistance and ductility. ~elow about 0.2% nic~bium .in the zirconium, the corrosion resis-tance beyins to approach that of sponge zirconium.
It is particularly preferred to h~e a ma.ximum niobium content of about 0.4% to assure t.hat the liner does not lie outside the solid soluhility range ~38~3~
regardless of thermal exposure of the fuel rod over long periods of time~ thereby assuring continued ductility.
The dilute zirconium alloy liner comprising about 0.1% to about 0.5~ by weight niobium, and preferably from about 0~2~ to about 0.4% by weight niohium and forming from about 1 to 20 percent of the thickness of the claddîng and preferably Erom about 5 to 15 percent of the cladding bonded to a conventional zirconium alloy substrate provides stress reduction while improving corrosion resistance, especially resistance to oxidation by hot water and steam in the event of a cladding breach.
The purity of the zirconium metal that is alloyed with niobium is important and serves to impart particular properties to the dilute zirconium alloy liner. Generally, there is at least 1000 parts per million (ppm~ by weight and less than 5000 ppm impurities in the zirconium metal and preferably less than 4200 ppm. Of these oxygen may vary up l:o about 1000 ppm.
The composite cladding of the nuclear fuel element of this invention has a dilute zirconium alloy liner metallurgically bonded to the substrate.
Metallographic examination shows that there is sufficient cross-diffusion between the substrate and the zirconium liner to form metallurgical bonds, but insuEficient cross-difEusion to significantly alloy with the dilute ~irconium alloy liner itselE.
Among the conventional zirconium alloys serving as suitable substrates are Zircaloy-2 and Zircaloy-4. Zircaloy-2 has on a weight basis about 1.5 percent tin; 0.12 percent iron; 0.09 percent chromium and 0~005 percent nickel and is extemsively employed in water--cooled reactors. Zircaloy-4 has less nickel than Zircaloy-2l but contains slightly more iron than Zircaloy-2. The composite cladding used 323~
in the nuclear fuel elements oE this invention can be fabricated by any of t~e following methods.
In one method/ a tube of -the dilute zirconium alloy liner material is inserted into a hollow billet of the material selected to be the substrate, and then the assembly is subjected to explosive bonding of the tube to the billet~ The composite is ex-truded using conventional tube shell extrusion at elevated temperatures of about 1000F to lA00F (about 538 C
to 760C). Then the extruded composite is subjected to a process involving conventional tube reduction until the desired size of cladding is achieved. The relative wall thickness of the hollow billet and the dilute zirconium alloy liner tube are selected to give the desired thickness ratios in the finished cladding tube.
In another method, a tube of the dilute zirconium alloy liner material is inserted into a hollow billet of the material selected to be the substrate, and then the assernbly is subjected to a heating step (such as at 750C for 8 hours) under compressive stress to assure good metal-to~metal contact and diffusion bonding between the tube and the billet. The diffusion bonded composite is extruded using conventional tube shell extrusion such as described above in the immedia-tely preceding paragraph. Then the extruded cornposite is subje!cted to a process involving conventional tube reduction until the desired size of cladding is achieved.
In still another method, a tube of the!
dilute zirconium alloy liner material is inserted into a hollow bille-t of the material selected to be the substrate~ and the assembly is extruded using conventional tube shell extrusion as described above.
Then the extrud,ed composite is subjected to a process involvingconvention~l tube reduction until the desired size of cladding is achieved.
24-NT-04~93 The foregoing processes of fabricating t:he composite cladding of this invention give economies over other processes used in fabricating cladding such as electroplating or vapor c1eposition. A nuclear fuel element can be forged by making a composite cladding container which is open at one end, the cladding container having a substrate and an inner dilute zirconium alloy liner metallurgically bonded to the inside surface of the substrate. The element is completed by fil.ling the composite cladding container with nuclear fuel material~ leaving a cavity at the open end~ inserting a nuclear fuel material retaining means into the cavity, applying an enclosure to the open end of the container leaving the cavity in communication with the nuclear fuel, and then bonding the end of the cladding container to said enclosure to form a tight seal therebetween.
The present invention offers several advantages promoting a long operating life ~or a rluclear fuel element, including the reduction of chemical interaction o~ the cladding, the minimization of localized stress on the æirconium alloy substrate portion of the cladding, the minimization of stress corrosion on the zirconi.um alloy substrate portion of the cladding, and the reduction of the probability of a splitting failure occurring in the zirconium alloy substrate.
In addi-tion to minimizing stress and stress corrosion on the substrate, the dilute zirconium alloy liner is resistant to oxidation by steam and hot water in theevent that the cladding is breached, whereas unalloyed zirconium oxidizes very rapidly under these conditions. The dilute zirconium alloy exh.ibits plasticity similar to unalloyed zirconium and provides the benefits thereof while also provid:ing increased resistance to corrosion, especially to oxidation by hot water and steam.
3~
24-NT-0~493 ~16-An important property ~ the composite cladding of this invention is that the foregoing improvements are achieved with no substan-tial additional neutron penalty. Such a cladding is readily accepted in nuclear reactors since the cladding would have no eutectic foxmation during a loss-of-coolant accident or an accident involving the dropping of a nuclear control rodA Furtherl the composite cladding has a very small heat transfer penalty in that there is no thermal barrier to transfer of heat such as results in the situation where a separate foil or liner is inserted in a fuel element. Also~ the composite cladding of this invention is inspectahle by conventional non-destructive testing methods during various stages of fabrication.
As will be apparent to those skilled in the art, various modifications and changes may be made in the invention described herein. It is accordingly the intention that the invention be construed in the broadestmanner within the spirit and scope as set forth in the accompanying claims.
Claims (38)
1. A nuclear fuel element comprising:
a central core of a body of nuclear fuel material selected from the group consisting of compounds of uranium, plutonium, thorium, and mixtures thereof; and an elongated composite cladding container enclosing said core consisting essentially of an outer portion forming a substrate and a dilute zirconium alloy liner consisting essentially of zirconium and at least about 0.1% by weight of niobium metallurgically bonded on the inside surface of the substrate, said dilute zirconium alloy liner comprising from at least about 1 to about 20 percent of the thickness of the composite cladding container.
a central core of a body of nuclear fuel material selected from the group consisting of compounds of uranium, plutonium, thorium, and mixtures thereof; and an elongated composite cladding container enclosing said core consisting essentially of an outer portion forming a substrate and a dilute zirconium alloy liner consisting essentially of zirconium and at least about 0.1% by weight of niobium metallurgically bonded on the inside surface of the substrate, said dilute zirconium alloy liner comprising from at least about 1 to about 20 percent of the thickness of the composite cladding container.
2. The nuclear fuel element of claim 1 in which the dilute zirconium alloy liner consists essentially of from about 0.1% to about 0.5% by weight niobium, the balance being zirconium.
3. The nuclear fuel element of claim 1 in which the dilute zirconium alloy liner consists essentially of from about 0.2% to about 0.4% by weight niobium, the balance being zirconium.
4. The nuclear fuel element of claim 1 in which the dilute zirconium alloy liner comprises from about 5 to about 15 percent of the thickness of the composite container.
5. A composite cladding container for nuclear reactors consisting essentially of a zirconium alloy outer portion forming a substrate and a dilute zirconium alloy liner consisting essentially of at least about 0.1% by weight of niobium in solid solution in zirconium, the dilute zirconium alloy liner being metallurgically bonded on the inside surface of the substrate, the dilute zirconium alloy liner comprising from about 5 to about 15 percent of the thickness of the composite cladding container.
6. A composite cladding container according to claim 5 in which the dilute zirconium alloy liner consists essentially of from about 0.1% to about 0.5% by weight niobium, the balance being zirconium.
7. A composite cladding container according to claim 5 in which the dilute zirconium alloy liner consists essentially of from about 0.2% to about 0.4% by weight niobium, the balance being zirconium.
8. A composite cladding container for nuclear reactors comprising a zirconium alloy outer portion forming a substrate and a dilute zirconium alloy liner consisting essentially of from about 0.1% to about 0.5%
by weight niobium, the balance being zirconium, bonded on the inside surface of the substrate.
by weight niobium, the balance being zirconium, bonded on the inside surface of the substrate.
9. A composite cladding container according to claim 8 in which the dilute zirconium alloy liner consists essentially of from about 0.2% to about 0.4% by weight niobium, the balance being zirconium.
10. A composite cladding container according to claim 8 in which the dilute zirconium alloy liner comprises at least about ] to about 20 percent of the thickness of the composite cladding container.
11. A composite cladding container according to claim 8 in which the dilute zirconium alloy liner comprises from about 5% to about 15% of the thickness of the composite cladding container.
12. A nuclear fuel element which comprises an elongated composite cladding container consisting essentially of an outer portion formed of a zirconium forming a substrate and a continuous dilute zirconium alloy liner consisting essentially of from about 0.1% to about 0.5% by weight niobium, the balance being zirconium;
the dilute zirconium alloy liner comprising from about 5 to 15 percent of the thickness of the composite cladding container metallurgically bonded on the inner surface of the substrate, a central core of nuclear fuel material selected from the group consisting of compounds of uranium, plutonium, thorium, and mixtures thereof, disposed in and partially filling said container and leaving an internal cavity in the container, an enclosure integrally secured and sealed at each end of said container, and a nuclear fuel material retaining means positioned in the cavity, said cladding container enclosing said core so as to leave a gap between said core and said cladding during use in a nuclear reactor.
the dilute zirconium alloy liner comprising from about 5 to 15 percent of the thickness of the composite cladding container metallurgically bonded on the inner surface of the substrate, a central core of nuclear fuel material selected from the group consisting of compounds of uranium, plutonium, thorium, and mixtures thereof, disposed in and partially filling said container and leaving an internal cavity in the container, an enclosure integrally secured and sealed at each end of said container, and a nuclear fuel material retaining means positioned in the cavity, said cladding container enclosing said core so as to leave a gap between said core and said cladding during use in a nuclear reactor.
13. A nuclear fuel element as recited in claim 12 wherein the dilute zirconium alloy liner consists essentially of from about 0.2% to about 0.4% by weight niobium, the balance being zirconium.
14. In a hollow composite cladding container for nuclear fuel for use in a nuclear reactor consisting essentially of an outer substrate of zirconium alloy and an inner protective liner having a thickness in the range of at least about 1 to about 20 percent of the thickness of the composite cladding container metallurgically bonded to the inner surface of the substrate, the improve-ment wherein the liner comprises a dilute zirconium alloy consisting essentially of zirconium and from about 0.1%
to about 0.5% by weight niobium.
to about 0.5% by weight niobium.
15. A composite cladding container as recited in claim 14 wherein the dilute zirconium alloy contains from about 0.2% to about 0.4% niobium.
16. A composite cladding container as claimed in claim 14 wherein the dilute zirconium alloy liner has a thickness in the range of from about 5 to about 15 percent of the thickness of the composite cladding container.
17. A composite cladding container for nuclear reactors consisting essentially of a zirconium alloy outer portion forming a substrate and a dilute zirconium alloy liner metallurgically bonded on the inside surface of the substrate, the alloy liner consisting essentially of niobium in amount of at least about 0.1% by weight in solid solution in zirconium and having a lower alloy content than the substrate.
18. A composite cladding container according to claim 17 wherein the niobium content of the liner is in the range of from about 0.1% to 0.5% by weight.
19. A composite cladding container according to claim 17 wherein the niobium content of the liner is in the range of from about 0.2% to 0.4% by weight.
20. A composite cladding container according to claim 17 wherein the alloy liner has a thickness in the range of from at least about 1 percent to about 20 percent of the thickness of the composite cladding container.
21. A composite cladding container according to claim 17 wherein the alloy liner has a thickness in the range of from about 5 percent to about 15 percent of the thickness of the composite cladding container.
22. A composite cladding container for nuclear reactors consisting essentially of a zirconium alloy outer portion forming a substrate and a zirconium alloy liner having a lower alloy content than the substrate formed of at least about 0.1% by weight niobium in solid solution in zirconium, the zirconium alloy liner being metallurgically bonded on the inside surface of the substrate, the zirconium alloy liner comprising from about 5 to about 15 percent of the thickness of the composite cladding container.
23. A composite cladding container according to claim 22 in which the zirconium alloy liner consists essentially of from about 0.1% to about 0.5% by weight niobium, the balance being zirconium.
24. A composite cladding container according to claim 22 in which the zirconium alloy liner consists essentially of from about 0.2% to about 0.4% by weight niobium, the balance being zirconium.
25. A composite cladding container for nuclear reactors consisting essentially of a zirconium alloy outer portion forming a substrate and a zirconium alloy liner consisting essentially of from about 0.1% to about 0.5%
by weight niobium, the balance being zirconium, metallurgically bonded on the inside surface of the substrate.
by weight niobium, the balance being zirconium, metallurgically bonded on the inside surface of the substrate.
26. A composite cladding container according to claim 25 in which the zirconium alloy liner comprises from about 0.2% to about 0.4% by weight niobium, the balance being zirconium.
27. A composite cladding container according to claim 25 in which the zirconium alloy liner comprises from about 1 to about 20 percent of the thickness of the composite cladding container.
28. A composite cladding container according to claim 25 in which the zirconium alloy liner comprises from about 5 to about 15 percent of the thickness of the composite cladding container.
29. A nuclear fuel element which comprises an elongated composite cladding container consisting essentially of an outer portion formed of a zirconium alloy forming a substrate and a continuous zirconium alloy liner having a lower alloy content than the substrate consisting essentially of from about 0.1% to about 0.5% by weight niobium, the balance being zirconium; the zirconium alloy liner comprising from about 5 to about 15 percent of the thickness of the composite cladding container metallurgically bonded on the inner surface of the substrate, a central core of nuclear fuel material. selected from the group consisting of compounds of uranium, plutonium, thorium, and mixtures thereof, disposed in and partially filling said container and leaving an internal cavity in the container, an enclosure integrally secured and sealed at each end of said container, and a nuclear fuel material retaining means positioned in the cavity, said cladding container enclosing said core so as to leave a gap between said core and said cladding during use in a nuclear reactor.
30. A nuclear fuel element as recited in claim 29 wherein the zirconium alloy liner consists essentially of from about 0.2% to about 0.4% by weight niobium, the balance being zirconium.
31. In a hollow composite cladding container for nuclear fuel for use in a nuclear reactor consisting essentially of an outer substrate of zirconium alloy and an inner protective liner having a thickness in the range of at least about l to about 20 percent of the thickness of the composite cladding container metallurgically bonded to the inner surface of the substrate, the improvement wherein the liner is a zirconium alloy having a lower alloy content than the substrate consisting essentially of zirconium and from about 0.1% to about 0.5% by weight niobium.
32. A composite cladding container as recited in claim 31 wherein the zirconium alloy comprises from about 0.2% to about 0.4% niobium.
33. A composite cladding container as claimed in claim 31 wherein the zirconium alloy liner has a thickness in the range of from about 5 to about 15 percent of the thickness of the composite cladding container.
34. A composite cladding container for nuclear reactors consisting essentially of a zirconium alloy outer portion forming a substrate and a zirconium alloy liner having a lower alloy content than the substrate metallurgically bonded on the inside surface of the substrate, the alloy liner consisting essentially of at least about 0.1% by weight niobium in solid solution in zirconium.
35. A composite cladding container according to claim 34 wherein the niobium content of the liner is in the range of from about 0.1% to 0.5% by weight.
36. A composite cladding container according to claim 34 wherein the niobium content of the liner is in the range of from about 0.2% to 0.4% by weight.
37. A composite cladding container according to claim 34 wherein the alloy liner has a thickness in the range of at least about 9 percent to about 20 percent of the thickness of the composite cladding container.
38. A composite cladding container according to claim 34 wherein the alloy liner has a thickness in the range of from about 5 percent. to about 15 percent of the thickness of the composite cladding container.
Applications Claiming Priority (2)
| Application Number | Priority Date | Filing Date | Title |
|---|---|---|---|
| US36395682A | 1982-03-31 | 1982-03-31 | |
| US363,956 | 1982-03-31 |
Publications (1)
| Publication Number | Publication Date |
|---|---|
| CA1198231A true CA1198231A (en) | 1985-12-17 |
Family
ID=23432437
Family Applications (1)
| Application Number | Title | Priority Date | Filing Date |
|---|---|---|---|
| CA000423387A Expired CA1198231A (en) | 1982-03-31 | 1983-03-11 | Zirconium alloy barrier having improved corrosion resistance |
Country Status (7)
| Country | Link |
|---|---|
| JP (1) | JPS58195185A (en) |
| BE (1) | BE896318A (en) |
| CA (1) | CA1198231A (en) |
| DE (1) | DE3310054A1 (en) |
| ES (1) | ES8605119A1 (en) |
| IT (1) | IT1160767B (en) |
| SE (1) | SE462307B (en) |
Cited By (3)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| US4775508A (en) * | 1985-03-08 | 1988-10-04 | Westinghouse Electric Corp. | Zirconium alloy fuel cladding resistant to PCI crack propagation |
| US4933136A (en) * | 1985-03-08 | 1990-06-12 | Westinghouse Electric Corp. | Water reactor fuel cladding |
| US4971753A (en) * | 1989-06-23 | 1990-11-20 | General Electric Company | Nuclear fuel element, and method of forming same |
Families Citing this family (10)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| JPS5958389A (en) * | 1982-09-29 | 1984-04-04 | 日本核燃料開発株式会社 | Nuclear fuel element |
| US4675153A (en) * | 1984-03-14 | 1987-06-23 | Westinghouse Electric Corp. | Zirconium alloy fuel cladding resistant to PCI crack propagation |
| US4613479A (en) * | 1984-03-14 | 1986-09-23 | Westinghouse Electric Corp. | Water reactor fuel cladding |
| US4664881A (en) * | 1984-03-14 | 1987-05-12 | Westinghouse Electric Corp. | Zirconium base fuel cladding resistant to PCI crack propagation |
| JPS6224182A (en) * | 1985-03-08 | 1987-02-02 | ウエスチングハウス・エレクトリック・コ−ポレ−ション | Nuclear fuel coated tube |
| JPS61217793A (en) * | 1985-03-08 | 1986-09-27 | ウエスチングハウス・エレクトリック・コ−ポレ−ション | Nuclear fuel coated tube |
| CN86101123A (en) * | 1985-03-08 | 1987-01-21 | 西屋电气公司 | Vessel of water reactor fuel |
| EP0301295B1 (en) * | 1987-07-21 | 1991-07-24 | Siemens Aktiengesellschaft | Fuel rod for a nuclear reactor fuel assembly |
| US4894203A (en) * | 1988-02-05 | 1990-01-16 | General Electric Company | Nuclear fuel element having oxidation resistant cladding |
| SE513185C2 (en) | 1998-12-11 | 2000-07-24 | Asea Atom Ab | Zirconium-based alloy and component of a nuclear power plant |
-
1983
- 1983-03-02 ES ES520223A patent/ES8605119A1/en not_active Expired
- 1983-03-11 CA CA000423387A patent/CA1198231A/en not_active Expired
- 1983-03-19 DE DE19833310054 patent/DE3310054A1/en not_active Ceased
- 1983-03-25 IT IT20296/83A patent/IT1160767B/en active
- 1983-03-30 BE BE0/210441A patent/BE896318A/en not_active IP Right Cessation
- 1983-03-30 SE SE8301815A patent/SE462307B/en not_active IP Right Cessation
- 1983-03-31 JP JP58054059A patent/JPS58195185A/en active Granted
Cited By (3)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| US4775508A (en) * | 1985-03-08 | 1988-10-04 | Westinghouse Electric Corp. | Zirconium alloy fuel cladding resistant to PCI crack propagation |
| US4933136A (en) * | 1985-03-08 | 1990-06-12 | Westinghouse Electric Corp. | Water reactor fuel cladding |
| US4971753A (en) * | 1989-06-23 | 1990-11-20 | General Electric Company | Nuclear fuel element, and method of forming same |
Also Published As
| Publication number | Publication date |
|---|---|
| BE896318A (en) | 1983-09-30 |
| JPH033917B2 (en) | 1991-01-21 |
| SE8301815L (en) | 1983-10-01 |
| IT1160767B (en) | 1987-03-11 |
| IT8320296A0 (en) | 1983-03-25 |
| SE8301815D0 (en) | 1983-03-30 |
| ES8605119A1 (en) | 1986-02-16 |
| ES520223A0 (en) | 1986-02-16 |
| DE3310054A1 (en) | 1983-10-13 |
| JPS58195185A (en) | 1983-11-14 |
| SE462307B (en) | 1990-05-28 |
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Legal Events
| Date | Code | Title | Description |
|---|---|---|---|
| MKEC | Expiry (correction) | ||
| MKEX | Expiry |