CN101654752A - Zirconium-tin-niobium system zirconium alloy used by nuclear reactor - Google Patents
Zirconium-tin-niobium system zirconium alloy used by nuclear reactor Download PDFInfo
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- CN101654752A CN101654752A CN200910023986A CN200910023986A CN101654752A CN 101654752 A CN101654752 A CN 101654752A CN 200910023986 A CN200910023986 A CN 200910023986A CN 200910023986 A CN200910023986 A CN 200910023986A CN 101654752 A CN101654752 A CN 101654752A
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Abstract
The invention discloses a zirconium-tin-niobium system zirconium alloy used by a nuclear reactor, which is characterized in that the component content of the alloy according to mass percent is as follows: 0.6-1.4% of Sn, 0.15-1.2% of Nb, 0.1-0.5% of Fe, 0.02-0.3% of Cr, 0.004-1.0% of Te, 5-150ppm of S, 700-1600ppm of O, 0-0.3% of Bi, 0-0.3% of Cu, 0-0.3% of Mg, and remainder of Zr and inevitable impurities. The zirconium alloy of the invention has outstanding tensile property, and has excellent anti-corrosion property and high temperature creep distortion resistance both in high temperature high pressure water and vapor.
Description
Technical field
The invention belongs to the material technology field, particularly relate to a kind of used by nuclear reactor zirconium-Xi-Zirconium-tin-niobium system zirconium alloy that is used for making nuclear reactor fuel can pipe, location grid, end plug and other structured material.
Background technology
Zirconium base alloy has the little and excellent creep resistance of thermal neutron absorption cross section, anti-irradiation growth, corrosion resistance nature and proper mechanical capacity, is that present optimum is used for one of material under the harsh working condition of nuclear reactor, is often used as reactor structural material.
At present ripe, the most most widely used is the zirconium base alloy that is called as Zr-2, Zr-4 alloy, but, require zirconium base alloy as reactor structural material must have over-all propertieies such as better anti-corrosion, creep resistance, anti-irradiation growth along with the development of nuclear fuel assembly to long-lived phase, high burnup direction.Because the burnup design load 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 of high burnup and long lifetime reactor core, since the seventies in 20th century, both at home and abroad the research and development of novel zirconium alloy is attached great importance to, its general thought is to carry out the adjustment of alloying constituent content and add other alloying elements on Zr-Nb system and Zr-Sn-Nb are the basis of alloy, and perhaps the two carries out simultaneously to reach the purpose of raising alloy monolithic performance.As everyone knows, niobium improves solidity to corrosion and the absorption that reduces hydrogen except being used to, and can also be used to improve physical strength and creep property.Therefore, exploitation recently and novel zirconium alloy cladding nuclear fuels material that successfully be used for commercial Nuclear power plants is characterized in that containing niobium.
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/MTU, the erosion rate of Zirlo alloy is than conventional Zr-4 low 67%, 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/MTU in 1992, compared with standard package, and the Fuel cycle expense descends 13%～14%.
The seventies, USSR (Union of Soviet Socialist Republics) was 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 better than the Zr-4 alloy, also is better than 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/MTU.The anti-uniform corrosion performance of this alloy has been improved 2 times than the mean value 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.
Japan has developed NDA and MDA alloy, and all belonging to ZrSnFeCrNb is alloy.(erosion rate is lower by 30%～40% than Zr-4 for interpolation B, out-pile corrosion test Li), and hydrogen is also low, and irradiation growth is also little 360 ℃ of one circuit cools agent simulated conditions.
A kind of have good solidity to corrosion and high-intensity zirconium alloy have been related 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.
Mentioned a kind of zirconium base alloy among the patent CN1150562C, except unavoidable impurities, also comprise by weight: 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 sum, people are to improve constantly the corrosion resistance nature of zirconium alloy and the growth of anti-neutron irradiation, 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 deficiencies in the prior art, a kind of all have in high-temperature high pressure water and the steam excellent corrosion resistance and the used by nuclear reactor zirconium-Xi-Zirconium-tin-niobium system zirconium alloy of high temperature creep-resisting performance are provided.
For solving the problems of the technologies described above, the technical solution used in the present invention is: a kind of used by nuclear reactor zirconium-Xi-Zirconium-tin-niobium system zirconium alloy is characterized in that this alloying constituent content is by mass percentage: Sn0.6～1.4%, Nb 0.15～1.2%, Fe 0.1～0.5%, and Cr 0.02～0.3%, and Te 0.004～1.0%, S5～150ppm, O 700～1600ppm, Bi 0～0.3%, and Cu 0～0.3%, Mg 0～0.3%, and surplus is Zr and unavoidable impurities.
The mass content of described element S is 15～40ppm.
The mass content of described element nb is 0.30～0.50%.
The mass content of described element nb is 0.90～1.2%.
The mass content of described element Cu is 0.004～0.3%.
The mass content of described element T e is 0.05～0.2%.
The mass content of described element Bi is 0.01～0.3%.
The interpolation quality summation of described element T e, Bi, Cu and Mg is not more than 1.0%.
Though the zirconium metal all has gratifying corrosion resistance nature in high-temperature high pressure water and steam, but the corrosion resistance nature of pure metal very easily is subjected to the wherein influence of impurity element, cause the unstable of performance, in addition, the stretching of pure metal, fatigue and creep-resistant property can not satisfy the requirement that engineering is used, therefore, the present invention is by having added elements such as a certain amount of Nb, Sn, Fe, Cr, Te, Bi, Cu, Mg, O, S in the zirconium metal, with the particularly corrosion-resistant and tensile property of over-all properties of raising alloy, thereby reach the requirement that engineering is used.
Details are as follows now to determine the reason of each alloying element and consumption:
(1)Nb
Known Nb is a kind of β phase stable element in the zirconium.Studies show that, 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 determines is 0.6～1.4% (weight).
(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% (weight percent) with greater than 0.5% (weight 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% (weight percent), and the content range of Cr is 0.02～0.3% (weight percent).
(4)Cu
Add a spot of Cu in the alloy and can improve its corrosion resistance nature, when the interpolation scope of Cu is 0.004～0.3% (weight percent), can obviously improve the corrosion resistance nature of alloy.
(5)S
Content does not influence corrosive property and the impurity element that helps to improve creep-resistant property being lower than 30ppm.When the sulphur that adds surpasses 20ppm, no longer reduce creep strain speed.Therefore, in order to improve the creep-resistant property of alloy, the content that preferred the present invention controls sulphur is about 25ppm.
(6)O
The O element can form interstitial solid solution in zirconium alloy, this sosoloid can improve the alloy physical strength, and still, the strengthening effect of crossing low O content is not obvious, do not reach required performance requriements, and too high O content can reduce the workability of alloy.The determined optimum content scope of the present invention is 700～1400ppm (weight percent).
(7) Bi also is a kind of α phase stable element, and has low neutron-absorption cross-section (82 millibarn), adds the appearance that a small amount of Bi can suppress nodular corrosion in zirconium alloy.The addition of setting Sn among the present invention is 0.6～1.4wt%, and the addition of Bi is 0.01～0.3wt%.
(8) Te: add corrosion resistance nature that can improve zirconium alloy and the anti-hydrogen sucking function of micro-Te, the addition of Te is 0.05～0.2wt.% usually.
Taking all factors into consideration above-mentioned each factor just can prepare and have excellent tensile properties and zirconium alloy anticorrosive, creep-resistant property.
The present invention compared with prior art has the following advantages: zirconium alloy of the present invention has excellent tensile properties, all has excellent corrosion resistance and high temperature creep-resisting performance in high-temperature high pressure water and steam.
The invention will be further described below in conjunction with specific embodiment, and embodiment is just to explanation of the present invention and non-limiting.
Embodiment
Embodiment
In zirconium sponge, add Nb, Sn, Fe, Cr, Te, Cu, Bi, the Mg element of requirement with the form of master alloy, in zirconium sponge, add the S and the O of requirement with oxide form, zirconium sponge is pressed into definite shape and electrodes sized, adopts vacuum consumable electrode arc furnace to carry out three meltings and make each 5kg alloy pig of 7 kinds of alloying constituents; Chemical composition analysis is carried out in sampling to ingot casting, and alloying constituent sees Table 1.Alloy cast ingot 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 650 ℃ of hot rollings, 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, make 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 650 ℃, 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.
Prepare 7 kinds of exemplary alloy samples that composition meets scope described in claims altogether by above-mentioned technology, the concrete composition of alloy sees Table 1.
7 kinds of exemplary alloy sheet material samples are carried out the room temperature tensile performance test and carry out 400 ℃, the test of 10.3Mpa steam corrosion and 360 ℃, 18.6Mpa neutral water corrosion test in autoclave, the time of corrosion test is 100 days.
The stretching of each alloy correspondence and 400 ℃, 360 ℃ corrosive natures see Table 2 and table 3, and with the performance of Zr-alloy under the same terms as a comparison, so that effect of the present invention to be described.
7 kinds of zirconium base alloy compositions of table 1 the present invention
The room temperature tensile performance of 7 kinds of zirconium base alloy sheet materials of table 2 the present invention (recrystallization annealing attitude)
| Embodiment | ??σ b(MPa) | ??σ 0.2(MPa) | ??δ 5(％) |
| ??1 | ??580 | ??420 | ??36 |
| ??2 | ??595 | ??435 | ??35 |
| ??3 | ??590 | ??430 | ??35 |
| ??4 | ??615 | ??438 | ??32 |
| ??5 | ??620 | ??440 | ??31 |
| ??6 | ??595 | ??423 | ??37 |
| ??7 | ??565 | ??410 | ??39 |
| ??Zr-4 | ??510 | ??339 | ??32 |
7 kinds of zirconium base alloy sheet material of table 3 the present invention corrosive nature
From above-mentioned example as can be seen, with Zr-4 alloy phase ratio, zirconium base alloy of the present invention has higher mechanical property, all has excellent corrosion resistance in high temperature pure water and high-temperature steam.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 reactor zirconium-Xi-Zirconium-tin-niobium system zirconium alloy, it is characterized in that this alloying constituent content is by mass percentage: Sn 0.6～1.4%, and Nb 0.15～1.2%, and Fe 0.1～0.5%, Cr 0.02～0.3%, Te 0.004～1.0%, S 5～150ppm, O 700～1600ppm, Bi 0～0.3%, Cu 0～0.3%, and Mg 0～0.3%, and surplus is Zr and unavoidable impurities.
2. a kind of used by nuclear reactor zirconium-Xi according to claim 1-Zirconium-tin-niobium system zirconium alloy, the mass content that it is characterized in that described element S is 15～40ppm.
3. a kind of used by nuclear reactor zirconium-Xi according to claim 1-Zirconium-tin-niobium system zirconium alloy, the mass content that it is characterized in that described element nb is 0.30～0.50%.
4. a kind of used by nuclear reactor zirconium-Xi according to claim 1-Zirconium-tin-niobium system zirconium alloy, the mass content that it is characterized in that described element nb is 0.90～1.2%.
5. a kind of used by nuclear reactor zirconium-Xi according to claim 1-Zirconium-tin-niobium system zirconium alloy, the mass content that it is characterized in that described element Cu is 0.004～0.3%.
6. a kind of used by nuclear reactor zirconium-Xi according to claim 1-Zirconium-tin-niobium system zirconium alloy, the mass content that it is characterized in that described element T e is 0.05～0.2%.
7. a kind of used by nuclear reactor zirconium-Xi according to claim 1-Zirconium-tin-niobium system zirconium alloy, the mass content that it is characterized in that described element Bi is 0.01～0.3%.
8. a kind of used by nuclear reactor zirconium-Xi according to claim 1-Zirconium-tin-niobium system zirconium alloy is characterized in that the interpolation quality summation of described element T e, Bi, Cu and Mg is not more than 1.0%.
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Cited By (6)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| CN102433465A (en) * | 2011-12-14 | 2012-05-02 | 国核宝钛锆业股份公司 | Bismuth-zirconium alloy |
| CN103451476A (en) * | 2013-09-05 | 2013-12-18 | 上海大学 | Sulfur-containing zircaloy for nuclear power plant fuel cladding |
| CN103469010A (en) * | 2013-09-05 | 2013-12-25 | 上海大学 | Sulfur-containing low-Nb zirconium-tin-niobium alloy for fuel cladding of nuclear power plant |
| CN103898368A (en) * | 2012-12-27 | 2014-07-02 | 中国核动力研究设计院 | Zirconium-based alloy for nuclear fuel assembly |
| CN103898363A (en) * | 2012-12-27 | 2014-07-02 | 中国核动力研究设计院 | Zirconium alloy for nuclear power |
| CN103898364A (en) * | 2012-12-27 | 2014-07-02 | 中国核动力研究设计院 | Zirconium alloy for nuclear reactor |
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2009
- 2009-09-22 CN CN200910023986A patent/CN101654752A/en active Pending
Cited By (8)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| CN102433465A (en) * | 2011-12-14 | 2012-05-02 | 国核宝钛锆业股份公司 | Bismuth-zirconium alloy |
| CN103898368A (en) * | 2012-12-27 | 2014-07-02 | 中国核动力研究设计院 | Zirconium-based alloy for nuclear fuel assembly |
| CN103898363A (en) * | 2012-12-27 | 2014-07-02 | 中国核动力研究设计院 | Zirconium alloy for nuclear power |
| CN103898364A (en) * | 2012-12-27 | 2014-07-02 | 中国核动力研究设计院 | Zirconium alloy for nuclear reactor |
| CN103898368B (en) * | 2012-12-27 | 2017-05-17 | 中国核动力研究设计院 | Zirconium-based alloy for nuclear fuel assembly |
| CN103451476A (en) * | 2013-09-05 | 2013-12-18 | 上海大学 | Sulfur-containing zircaloy for nuclear power plant fuel cladding |
| CN103469010A (en) * | 2013-09-05 | 2013-12-25 | 上海大学 | Sulfur-containing low-Nb zirconium-tin-niobium alloy for fuel cladding of nuclear power plant |
| CN103469010B (en) * | 2013-09-05 | 2016-04-27 | 上海大学 | The zirconium tin niobium alloy of the low Nb of fuel for nuclear power plant involucrum sulfur-bearing |
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Open date: 20100224 |