CN101805842B - Zirconium-tin-niobium corrosion-resistant zirconium-base alloy for nuclear fuel cans - Google Patents
Zirconium-tin-niobium corrosion-resistant zirconium-base alloy for nuclear fuel cans Download PDFInfo
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Abstract
The invention discloses a zirconium-tin-niobium corrosion-resistant zirconium-base alloy for nuclear fuel cans, which comprises the following components in percentage by mass: 0.6-1.4% of Sn, 0.10-1.5% of Nb, 0.1-0.5% of Fe, 0.02-0.3% of Cr, 0.005-0.5% of MgO, 0-0.5% of CeO2, 0-0.5% of Y2O3, 0-0.015% of SiO2, 0-0.03% of V2O3, 0.07-0.15% of O, and balance of Zr and inevitable impurities. Compared with the prior Zr-4 alloy, the zirconium-base alloy of the invention has better high temperature creep resistance and excellent corrosion resistance, and is suitable for can materials, grids and otherstructural members for reactor fuel rods in a nuclear power station.
Description
Technical field
The invention belongs to the zirconium-based alloy material technical field; Relate to a kind of corrosion-resistant Zr-based alloy that can be used as fuel rod clad material, grid and structural part in light-water and the heavy water nuclear power plant nuclear reactor, being specifically related to a kind of used by nuclear fuel jacketing zirconium-Xi-niobium that utilizes oxide compound to strengthen is corrosion-resistant Zr-based alloy.
Background technology
The thermal neutron absorption cross section of zirconium is very little, and has good high-temperature resistant water corrosive nature and mechanical property, and therefore zirconium alloy is widely used as the can material of fuel stick and the structural element of nuclear reactor core in water cooled nuclear reactor.Along with the power producer technology towards improving fuel burnup and reducing fuel cycle cost; Improve reactor thermo-efficiency; The direction that improves safe reliability develops; Demands for higher performance to key core parts fuel element can material zirconium alloy comprises corrosive nature, hydrogen sucking function, mechanical property and irradiation dimensional stability etc.Fuel element under the condition (irradiation, high temperature, high pressure and complicated stress), place creep and fatigue under arms.Creep property is one of the major issue that will consider when in the water-cooled power reactor, working of zirconium alloy, a large amount of research has been carried out in the creep of zirconium alloy both at home and abroad.Early development goes out in the sixties in last century zirconium alloy such as Zr-4 alloy, it has excellent physical strength, creep resistance, heat conductivity and low neutron-absorption cross-section under the reactor working temperature, and uses so far widely.Because the burnup value of the fuel for nuclear power plant that the Zr-4 alloy that conventional Zr-Sn is can satisfy is generally 33GWd/tU; Therefore, in order to satisfy the requirement that high burnup and long lifetime push away core, on the one hand; Many countries have all carried out the corrosion research that improves the Zr-4 alloy since the seventies in 20th century; Study the better novel zirconium alloy of performance on the other hand, the exploitation of novel zirconium alloy tends to reduce or eliminate the content and adding niobium (Nb) of tin (Sn), and wherein the most outstanding achievement is to have developed low tin Zr-4 alloy; Or being referred to as to optimize the Zr-4 alloy, design burn-up can reach 45GWd/tU.
US Westinghouse company has developed Zirlo the seventies
TMAlloy (Zr1.0%Nb1.0%Sn1.0%Fe), nineteen ninety-five reaches industrial scale applications.This alloy adopts the involucrum pipe of low temperature process β quench treatment production subsequently, and microstructure contains the tiny second phase particle that is evenly distributed.Under reactor operation; The all more conventional Zr-4 of water-fast side corrosive nature, fuel stick irradiation growth and creep-resistant property and the low tin Zr-4 of Zirlo alloy are superior; When burnup reaches 37.8GWd/tU; The erosion rate of Zirlo alloy is lower by 67% than conventional Zr-4, and lower by 58% than low tin Zr-4, irradiation growth is than conventional Zr-4 low 60%.Use Zirlo
TMThe assembly of alloy manufacturing reached 55GWd/tU in 1992, compared with standard package, and the Fuel cycle expense descends 13%～14%.
The seventies, the FSU developed E635 alloy (Zr1.3%Sn1.0%Nb0.35%Fe).The microstructure of this alloy is mainly by the α crystal grain and the second (distribution density (2～4) * 10 mutually
13) form.Constituent particle has three kinds of patterns: mainly be close-packed hexagonal structure Zr (Nb, Fe)
2Phase, also have tetragonal lattice (Zr, Nb)
2Fe phase and rhombic system (Zr, Nb)
3The Fe phase.At 360 ℃, 18.6MPa contains in the water of 70ppm Li, and the solidity to corrosion of autoclave test E635 alloy obviously is superior to the Zr-4 alloy, also is superior to the Zr1.0%Nb alloy.At 400 ℃, corrosion resisting property in the 10.3MPa water vapor and Zirlo alloy phase are worked as.The E635 alloy is done reactor fuel element involucrum and VVER and RBMK reactor core assembly, test data in the existing heap fully.
M5
TM(Zr1.0%Nb0.125%O) be the ZrNb alloy of French Fa Jiema company exploitation, be used as the involucrum pipe of design burn-up for the AFA-3G fuel assembly of (55～60) GWd/tU.The anti-uniform corrosion performance of this alloy has been improved 2 times than the MV of optimizing Zr-4, and oxidation rate is little under high burnup, and the data dispersiveness is little, inhales hydrogen and also lacks than optimizing Zr-4, and the fuel stick irradiation growth is than optimizing low 1 times of Zr-4.
Related to a kind of have good solidity to corrosion and high-intensity zirconium alloy among the patent CN1087037C of Korea Atomic Energy Research Institute's application, each component concentration of zirconium alloy is by percentage to the quality: Nb:0.3～0.6%, Sn:0.7～1.0%; Be selected from Mo; A kind of element among Cu, the Mn, content are 0.05～0.4%, oxygen 600～1400ppm; Wherein also can add Fe 0.2～0.5% or Cr 0.05～0.25%, make product have suitable corrosion resisting property.
Patent CN97110736.X discloses zirconium base alloy and the method for manufacture and the application of a kind of creep resistance and water and steam corrosion; The sulphur that comprises 8～100ppm (is preferably 8～30ppm) and surpass the zirconium alloy of 96% zirconium and following for the dependent claims of 8 kinds of alloys.
Alloy 1: the zirconium alloy that comprises 1.2～1.7%Sn, 0.18～0.25%Fe, 0.05～0.15%Ni and 0.05～0.15%Cr.
Alloy 2: the zirconium alloy that comprises 1.2～1.7%Sn, 0.07～0.2%Fe, 0.05～0.15%Ni and 0.05～0.15%Cr.
Alloy 3: the zirconium alloy that comprises 0.7～1.3Nb and 900～1600ppmO.
Alloy 4: the zirconium alloy that comprises 0.3～1.4Sn, 0.4～1%Fe, 0.2～0.7%V and 500～1800ppm O.
Alloy 5: the zirconium alloy that comprises 0.7～1.3Nb, 0.8～1.5%Sn, 0.1～0.6%Fe, 0.01～0.2Cr and 500～1800ppm O.
Alloy 6: the zirconium alloy that comprises 0.1～0.3%Nb, 0.7～1.25%Sn, 0.1～0.3%Fe, 0.05～0.2%Cr, 0.01～0.02%Ni and 500～1800ppm O.
Alloy 7: the zirconium alloy that comprises 2.2～2.8%Nb.
Alloy 8: the zirconium alloy that comprises 0.3～0.7%Sn, 0.3～0.7%Fe, 0.1～0.4%Cr, 0.01～0.04%Ni, 70～120ppm Si and 500～1800ppm O.
Mentioned a kind of zirconium base alloy among the patent CN1150562C, except unavoidable impurities, also comprise by quality: Fe 0.02～1%; Nb 0.8～2.3%; Be lower than the Sn of 2000ppm, be lower than the O of 2000ppm, be lower than the C of 100ppm; The S of 5～35ppm and Cr, V summation are 0.01～0.25%, and content of niobium deducts 0.5% and adds the chromium of inessential interpolation and/or the ratio of vanadium composition is higher than 2.5 with iron level.
U.S. Pat 4963323 has been adjusted the alloy compositions of conventional Zr-4 alloy, and improving the corrosion resistance nature of alloy, this patent reduces the content of Sn, adds the loss of strength that Nb causes owing to the minimizing of Sn with compensation, and guarantees that nitrogen content is lower than 60ppm.The composition range of alloy is: Sn 0.2～1.15%, and Nb 0.05～0.5%, and Fe 0.19～0.6%, and Cr 0.07～0.4% and N are less than 60ppm.
U.S. Pat 5017336 adds Nb, Ta, V and Mo on Zr-4 alloying constituent basis, propose a kind of Sn 0.2～0.9% that comprises, and Fe 0.18～0.6%; Cr 0.07～0.4%; Nb 0.05～0.5%, and Ta 0.01～0.2%, the zirconium alloy of V 0.05～1% and Mo 0.05～1%.
In order to improve the creep-resistant property of zirconium alloy; Existing patent tends in zirconium alloy, add small amount of sulfur (S) and silicon (Si) more; But these two kinds of elements are prone in melting and hot-work subsequently and heat treatment process that easily crystal boundary takes place gathers phenomenon partially, causes the decline of corrosive nature.
In sum, people are the corrosion resistance nature and the growth of anti-neutron irradiation that improve constantly zirconium alloy, irradiation creep performance, anti-hydrogen sucking function etc. to the ultimate aim of being pursued of used by nuclear reactor Zirconium alloy material.For this reason, the present invention studies the alloy compositions proportioning, proposes new alloying constituent, and exploitation has the more zirconium alloy of good corrosion resistance.
Summary of the invention
The objective of the invention is in order to overcome the deficiency of prior art; It is corrosion-resistant Zr-based alloy that a kind of used by nuclear fuel jacketing zirconium-Xi-niobium is provided; This zirconium base alloy has better high temperature creep-resisting performance and solidity to corrosion than existing Zr-4 alloy, is applicable to can material, grid and other structural part of nuclear power plant reactor fuel stick.
For realizing above-mentioned purpose; The technical scheme that the present invention adopts is: a kind of used by nuclear fuel jacketing zirconium-Xi-niobium is a corrosion-resistant Zr-based alloy, and this alloying constituent content is by mass percentage: Sn0.6～1.4%, and Nb 0.10～1.5%; Fe 0.1～0.5%; Cr 0.02～0.3%, and MgO 0.005～0.5%, CeO
20～0.5%, Y
2O
30～0.5%, SiO
20～0.015%, V
2O
30～0.03%, O 0.07～0.15%, and surplus is Zr and unavoidable impurities.
The mass content of said Nb is 0.15～0.45%.
The mass content of said Nb is 0.80～1.2%.
The content of said MgO is by mass percentage: 0.015～0.025%.
Said CeO
2Content be by mass percentage: 0.05～0.07%.
Said Y
2O
3Content be by mass percentage: 0.04～0.06%.
Said SiO
2Content be by mass percentage: 0.008～0.01%.
Said V
2O
30.008～0.01%.
Said MgO, CeO
2, Y
2O
3, SiO
2And V
2O
3The associating addition can not surpass 0.65% by mass percentage.
To confirm the reason of main alloy element and consumption details are as follows at present:
(1)Nb
Known Nb is a kind of β phase stable element in the zirconium.Research shows; When adding the Nb of a small amount of (less than 0.15%), the corrosion resistance nature of Zirconium alloy material just can be improved, when the Nb content height to 1.2% that adds; The corrosion resistance nature of alloy also can be greatly improved, and the mechanics and the anti-hydrogen sucking function of alloy also are improved simultaneously simultaneously.
(2)Sn
Sn is a kind of α phase stable element, and can improve the intensity and the corrosive nature of zirconium alloy, but adding a spot of Sn can not reach needed intensity and creep-resistant property.The Sn optimum amount scope that the present invention confirms is 0.6～1.4% (mass percent).
(3) Fe and Cr
Fe and Cr can improve the corrosion resistance nature and the tensile property of alloy, but in alloy involved in the present invention, as the Fe that adds during less than 0.1% (mass percent) with greater than 0.5% (mass percent), all can not obviously improve its corrosion resistance.Though the Cr element can improve the corrosion resistance nature of alloy, and the Zr that Cr and Fe form in alloy material (Fe, Cr)
2But can significantly reduce the anti-hydrogen sucking function of alloy mutually, so the corrosion-resistant and anti-hydrogen sucking function of the content range of Cr and Fe and the remarkably influenced of proportioning meeting between the two alloy.In alloy involved in the present invention, the content range of Fe is 0.1～0.5% (mass percent), and the content range of Cr is 0.02～0.3% (mass percent).
Research thinks that the creep mechanism of zirconium alloy has two kinds: dislocation mechanism and flooding mechanism.The creep that zirconium alloy produces under low temperature (reactor working temperature) and condition of high ground stress is considered to dislocation mechanism control usually, promptly is out of shape through stack (CPG) place creep of climb of dislocation and dislocation glide.Flooding mechanism is the mode of texturing that occurs under the comparatively high temps.Therefore the method that improves the zirconium alloy creep-resistant property mainly contains following two kinds:
A, solution strengthening effect.Alloying element has considerable influence to the creep property of zirconium alloy, and tin, niobium, the oxygen of solid solution in matrix can improve the strength of zirconium alloy.Wherein element nb is a very important element, and it is the β phase stable element of zirconium, has research to think and adds corrosion resistance nature and the processibility that minor N b (less than 0.5wt%) just can improve alloy.Other research thinks that the zirconium alloy that contains Nb1.0wt% has best corrosion resistance nature.In zirconium, add the Nb of 0.10～1.5wt% among the present invention, the creep speed of zirconium alloy is reduced with the increase that is solid-solubilized in the content of niobium in the matrix, its reason is the solution strengthening effect of niobium atom.
B, precipitation strength effect.The crystal boundary place that often is prone in zirconium alloy through microalloying, multi-element alloyed mode generates some disperses distributions, thermally-stabilised high precipitation strength phase; It still can be lived the intracrystalline dislocation and hinder intercrystalline slip by pinning when high temperature; Play the effect of effective inhibition, machinery obstruction, thereby improve the hot strength and the creep-resistant property of zirconium alloy.Fully utilize each grain-size, grain shape is also influential to creep property; Existing research shows: under the constant situation of volume share; The precipitated phase finer particles is little to be that the quantity of precipitated phase particle is many more; The strength of alloy is high more, and with the increase of dislocation desity in the matrix, its creep resistance will improve.Add polynary a spot of MgO, CeO among the present invention
2, Y
2O
3, SiO
2And barium oxide, and, the oxide compound even dispersion is distributed through heat processing technique and heat treated adjustment, combination and rational Match, improve the zirconium base alloy creep-resistant property and optimize its corrosion resistance nature.
In addition, because the solution strengthening effect, the O that adds 700ppm～1500ppm can make alloy have enough mechanical propertys and creep-resistant property.But O surpasses the workability that 1600ppm can reduce alloy.
The present invention compared with prior art has the following advantages: zirconium alloy of the present invention has the corrosion resistance nature that good creep-resistant property is become reconciled; This zirconium base alloy has more excellent comprehensive performances than existing Zr-4 alloy, is applicable to can material, grid and other construction package of nuclear power plant reactor fuel stick.
Embodiment below in conjunction with concrete is described further the present invention, and embodiment is just to explanation of the present invention and non-limiting.
Embodiment
Embodiment
With nuclear level zirc sponge, Nb, MgO, CeO
2, Y
2O
3, SiO
2, V
2O
3, elements such as O are prepared burden by prescription with the form of master alloy and are pressed into electrode, adopt vacuum consumable electrode arc furnace to carry out three meltings and process each 3kg alloy pig of 6 kinds of alloying constituents; Chemical composition analysis is carried out in sampling to ingot casting, and alloying constituent is seen table 1.Alloy pig forged at 980 ℃～1050 ℃ be processed into the base material; Water medium quenches after 1015 ℃～1075 ℃ β phase region heating again; The base material is being lower than 620 ℃ of hot rollings, and deflection is greater than 60%, after 600 ℃ are carried out process annealing; Then through repeatedly cold rolling; Fire time deflection is greater than 50%, adopt with hot rolling after identical annealing temperature carry out process annealing, process sheet material; End article is handled through 580 ℃ of recrystallization annealings, promptly makes this zirconium alloy sheet material.
Characteristics in the embodiment of the invention are: 1) characteristic alloy formula; 2) in the following process process of base material after the β heat phase is quenched; Adopt the big strain complete processing of low temperature, Heating temperature is no more than 620 ℃, and fire time variable is greater than 50%; Help obtaining second phase that small and dispersed distributes, can further improve the corrosion resistance nature of alloy like this.
8 kinds of zirconium base alloy sheet materials of acquisition the present invention are carried out the corrosive nature test.Corrosion test is carried out at autoclave, and etching condition is 360 ℃, 18.6MPa deionized water; 400 ℃, 10.3MPa deionized water steam, etching time is 100 days, and table 1 has provided the composition proportion of every kind of zirconium base alloy.When table 2 has been listed the embodiment of the invention in the surrosion under the above-mentioned etching condition with at 375 ℃, creep strain rate under 117MPa and the 137MPa stress.As a comparison, the testing data of the same test conditions of Zr-4 alloy is listed in table 2 too.
Table 1, zirconium base alloy composition of the present invention are for example
Table 2, zirconium base alloy sheet material of the present invention and Zr-4 alloy corrosion test, creep property are relatively
Can find out that from the foregoing description zirconium base alloy of the present invention and existing Zr-4 alloy phase be than there being more good high temperature creep-resisting performance, and in high temperature pure water and high-temperature steam, all have excellent corrosion resistance.Thereby zirconium base alloy of the present invention can be as coating layer, grid and other structural part material of nuclear reactor core fuel stick.
Claims (8)
1. used by nuclear fuel jacketing zirconium-Xi-niobium is a corrosion-resistant Zr-based alloy, and this alloying constituent content is by mass percentage: Sn 0.6～1.4%, and Nb 0.10～1.5%, and Fe 0.1～0.5%, and Cr 0.02～0.3%, and MgO 0.005～0.5%, CeO
20～0.5%, Y
2O
30～0.5%, SiO
20～0.015%, V
2O
30～0.03%, O 0.07～0.15%, and surplus is Zr and unavoidable impurities; Said MgO, CeO
2, Y
2O
3, SiO
2And V
2O
3The associating addition can not surpass 0.65% by mass percentage;
The preparation method is: (1) is by the prescription batching and be pressed into electrode, adopts vacuum consumable electrode arc furnace to carry out three meltings and processes alloy pig;
(2) alloy pig described in the step (1) is forged at 980 ℃～1050 ℃ be processed into the base material, through 1015 ℃～1075 ℃ β phase regions heating back water medium quenchings, the base material is being lower than 620 ℃ of hot rollings again; Deflection is greater than 60%, after 600 ℃ carried out process annealing, then through repeatedly cold rolling; Fire time deflection is greater than 50%, adopt with hot rolling after identical annealing temperature carry out process annealing, process sheet material; End article is handled through 580 ℃ of recrystallization annealings, promptly makes zirconium alloy sheet material.
2. a kind of used by nuclear fuel jacketing zirconium-Xi according to claim 1-niobium is a corrosion-resistant Zr-based alloy, and the mass content that it is characterized in that said Nb is 0.15～0.45%.
3. a kind of used by nuclear fuel jacketing zirconium-Xi according to claim 1-niobium is a corrosion-resistant Zr-based alloy, and the mass content that it is characterized in that said Nb is 0.80～1.2%.
4. a kind of used by nuclear fuel jacketing zirconium-Xi according to claim 1-niobium is a corrosion-resistant Zr-based alloy, it is characterized in that, the content of said MgO is by mass percentage: 0.015～0.025%.
5. a kind of used by nuclear fuel jacketing zirconium-Xi according to claim 1-niobium is a corrosion-resistant Zr-based alloy, it is characterized in that said CeO
2Content be by mass percentage: 0.05～0.07%.
6. a kind of used by nuclear fuel jacketing zirconium-Xi according to claim 1-niobium is a corrosion-resistant Zr-based alloy, it is characterized in that said Y
2O
3Content be by mass percentage: 0.04～0.06%.
7. a kind of used by nuclear fuel jacketing zirconium-Xi according to claim 1-niobium is a corrosion-resistant Zr-based alloy, it is characterized in that said SiO
2Content be by mass percentage: 0.008～0.01%.
8. a kind of used by nuclear fuel jacketing zirconium-Xi according to claim 1-niobium is a corrosion-resistant Zr-based alloy, it is characterized in that said V
2O
30.008～0.01%.
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| US8971476B2 (en) | 2012-11-07 | 2015-03-03 | Westinghouse Electric Company Llc | Deposition of integrated protective material into zirconium cladding for nuclear reactors by high-velocity thermal application |
| CN104451260B (en) * | 2014-11-29 | 2016-06-29 | 西部新锆核材料科技有限公司 | A kind of nuclear reactor fuel can zirconium-niobium alloy containing ferrimanganic |
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| SE530673C2 (en) * | 2006-08-24 | 2008-08-05 | Westinghouse Electric Sweden | Water reactor fuel cladding tube used in pressurized water reactor and boiled water reactor, comprises outer layer of zirconium based alloy which is metallurgically bonded to inner layer of another zirconium based alloy |
| CN101413073B (en) * | 2008-12-03 | 2011-03-16 | 西北有色金属研究院 | Magnesium-containing zirconium-niobium alloy for nuclear reactor fuel can |
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Effective date of registration: 20130508 Address after: Economic and Technological Development Zone Jingwei Metro New Material Industrial Park in Shaanxi Province Jing high road 710200 in the middle of Xi'an City Patentee after: WESTERN ENERGY MATERIAL TECHNOLOGIES CO.,LTD. Address before: 710016 Shaanxi province Xi'an Weiyang Weiyang Road No. 96 Patentee before: NORTHWEST INSTITUTE FOR NONFERROUS METAL RESEARCH |
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| CP03 | Change of name, title or address |
Address after: 710299 No. 19, east section of Jinggao North Road, Jingwei new town, Xi'an Economic and Technological Development Zone, Xi'an City, Shaanxi Province Patentee after: Xi'an Western New Zirconium Technology Co.,Ltd. Address before: 710,200 Middle section of Jinggao West Road, Jingwei New City New Material Industrial Park, Xi'an Economic and Technological Development Zone, Shaanxi Province Patentee before: WESTERN ENERGY MATERIAL TECHNOLOGIES CO.,LTD. |