CN101265538A - Zirconium-base alloy used for light-water reactor - Google Patents
Zirconium-base alloy used for light-water reactor Download PDFInfo
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Abstract
The invention relates to a zirconium alloy material, particularly to a zirconium-base alloy used for a light water reactor. The zirconium-base alloy is composed of (wt %) Sn 0.80-1.20, Nb 0.90-1.25, Fe 0.12-0.45, O 0.06-0.15, C less than 0.15, N less than 0.008, and zirconium and impurities in balance. By adding Fe and O two elements based on the current Zr-Sn-Nb alloy, the zirconium-base alloy with good corrosion-resisting property in all corrosion tests using 360 DEG C deionized water, 360 DEG C lithium-containing water solution, 400 DEG C steam and 500 DEG C can be prepared. The inventive zirconium-base alloy can be used as structural material of a nuclear reactor core.
Description
Technical field
The present invention relates to a kind of Zirconium alloy material, relate in particular to a kind of zirconium base alloy that is used for light-water reactor.
Background technology
At light-water reactor, comprise in the evolution of boiling-water reactor and pressurized-water reactor that fuel design as fuel element can, screen work, guide pipe etc., has proposed very high requirement to the reactor core structure parts.Current, these parts are made by Zr-2 and Zr-4 alloy usually.The design of high fuel burnup, requirement prolong the residence time and the raising coolant temperature of these parts in heap, thereby make the zirconium alloy parts be faced with more harsh corrosion and suction hydrogen environment.These high requests have promoted to improve the Study on Corrosion Resistance of Zr-2 and Zr-4 alloy, promoted to have superior corrosion resistance more can and the exploitation of the novel zirconium alloy of anti-hydrogen sucking function.
In the light-water reactor environment, the reaction of zirconium water can take place in zirconium alloy, forms ZrO at the zirconium alloy parts surface
2Film.Commitment in oxidation forms fine and close black oxide film, has protectiveness, and oxide film has monocline, four directions, cube heterogeneous structure.Along with the carrying out of oxidation, rate of oxidation can be transferred, and turnover rear oxidation film skin cavity or crackle constantly occur and loses protectiveness, and the new compact oxidation layer of can constantly growing on matrix and the interfacial oxide film.Therefore the corrosion characteristics of zirconium alloy is exactly the repetitive process of the turnover of the growth of zone of oxidation on matrix and the interfacial oxide film and surface oxide layer, the porous oxide skin of the unprotect that the final generation of this process is thicker.And, in the boiling water environment, also nodular corrosion can occur, thereby limit zirconium alloy cladding work-ing life.
Owing to contain the lithium hydroxide of adjusting the pH value in the pressurized water reactor refrigerant, and contain the boric acid of controlling initial reactivity, B
10Through (n, α) corrosion of zirconium alloy has been quickened in the existence of the lithium that produces of reaction decomposes, the acceleration problem of zirconium alloy component corrosion occurs causing under the extreme condition of high lithium concentration so need consider regional area.
Although corrosion resistance nature is made moderate progress by studying improved Zr-4 alloy, but higher lithium concentration in the higher burnup of the requirement of nuclear-power reactor development, longer refulling cycle, higher coolant temperature, the refrigerant, longer residence time in the reactor core, these high requests have increased the corrosion load of zirconium alloy parts.
To the high request that fuel sheath proposes, launched the research of novel zirconium alloy at the Nuclear Power Technology development in the world.GEORGEP.SABOL as US Westinghouse company in the 8th zirconium alloy international symposium has reported " development of high burnup involucrum alloy " (" Development of a Cladding Alloy for High Burnup ", Zirconium in the Nuclear Industry:EighthInternational Symposium, ASTM STP 1023, L.F.P.Van Swan and C, M, Eucken, Eds., American Society for Testing and Materials, Philadelphia, 1989,227-244), announced the result of study of the Zr-Nb-Sn-Fe alloy that is referred to as ZIRLO, its nominal composition (nominal chemical composition) is Zr-1.0wt%Nb-1.0wt%Sn-0.1wt%Fe.This alloy has improved corrosion resistance nature.GEORGE P.SABOL has reported " the in-pile corrosion behavior of ZIRLO and Zr-4 alloy " (" In-Reactor Corrosion Performance of ZIRLO and Zircaloy-4 " once more in the tenth zirconium alloy international symposium, Zirconiumin the Nuclear Industry:Tenth International Symposium, ASTM STP 1245, A.M.Garde and E.R.Bradley, Eds., American Society for Testing and Materials, Philadelphia, 1994,724-744), showed that ZIRLO has corrosion resistance nature and anti-hydrogen and the creep-resistant property inhaled in the better heap than Zircaloy-4.Muscovite Nikulina in the 11 zirconium alloy international symposium, A.V. reported " as the E635 zirconium alloy of VVER and RBMK reactor fuel rod involucrum and component materials " (" Zirconium Alloy E635 as a Material for Fuel Rod Cladding and Other Components of VVER and RBMK Cores ", Zirconium in the Nuclear Industry:Eleventh InternationalSymposium, ASTM STP 1295, E.R.Bradley and G.P.Sabol, Eds., American Society for Testing and Materials, Philadelphia, 1996,785-804), the composition of having announced E635 is Zr-1.0～1.4wt%Nb-0.9～1.1wt%Sn-0.3～0.5wt%Fe.The out-pile performance of this alloy is better than Zircaloy-4 and E110 alloy.
The patent of invention (CN 1125885C) of the full Russia of Russia inorganic materials research institute provides a kind of zirconium base alloy (percentage composition is together following by weight) to contain: the Nb of 0.50-3.0; 0.50-2.0 Sn; 0.30-1.0 Fe; 0.002-0.2 Cr; 0.03-0.04 C; 0.04-0.15 O; 0.002-0.15 Si; 0.001-0.4 W, Mo or V; Surplus is Zr.The patent of invention of US Westinghouse company (CN1404532) provides a kind of erosion resistance zirconium base alloy that is used in the nuclear fuel coating, is to be made by the zirconium alloy of low tin content, and the zirconium alloy of low tin content is basically by following elementary composition: the Nb of 0.60-2.0; When Sn was 0.25, Fe was 0.50; When Sn was 0.40, Fe was 0.35-0.50; When Sn was 0.50, Fe was 0.25-0.50; When Sn was 0.70, Fe was 0.05-0.50; When Sn was 1.0, Fe was 0.05-0.50; Wherein, the weight percentage sum of Fe and Sn is greater than 0.75, and other other component is no more than 0.50, and surplus is Zr.
The Korea Atomic Energy Research Institute of Korea S has applied for the multinomial zirconium alloy patent that is used for the reactor core structure material in China, and these patents are added element to improve its corrosion resistance nature on the basis of existing zirconium alloy.As denomination of invention is the patent of invention of " as the novel zirconium alloy of nuclear fuel coating layer " (Granted publication CN 1087037C), claiming provides a kind of have excellent corrosion resistance and high-intensity advanced zirconium alloy, in its explanation, understand the content range of every kind of alloying element specifically, and the reason of determining corresponding content range, make product have suitable corrosion resistance nature, be unlikely to lose processibility again.Simultaneously, the patent of Korea Atomic Energy Research Institute's application also in its background technology, has been mentioned multiple zirconium alloy.
The anti-all even nodular corrosion that it has been generally acknowledged that the zirconium alloy that is used for the reactor core structure material is most important corrosive nature requirement.Corrosive nature at out-pile check zirconium alloy comprises: 360 ℃ of deionized waters; 360 ℃ contain the lithium aqueous solution; The corrosion test of 400 ℃, 500 ℃ steam.Although people can both accept following viewpoint: the material that test is up to the standards in 360 ℃ of aqueous solution and 400 ℃ of steam can be used for pressurized-water reactor, 360 ℃ of qualified then more being applicable in the high lithium concentration operating mode of pressurized-water reactor of experimental examination that contain in the lithium aqueous solution, the experimental examination in steam more than 500 ℃ qualified then applicable in boiling-water reactor; And, in the above-mentioned disclosed document, all bright by the test illustration, relevant zirconium alloy zirconium-2 and zirconium-4 alloy more in the past has more excellent performance, but whether these alloys can really be applied in the middle of the practice, and it is still unknown to show satisfactory technique effect.And above-mentioned document does not all provide the anti-nodular corrosion behavior of relevant alloy in 500 ℃ of steam yet.
Summary of the invention
The purpose of this invention is to provide the zirconium base alloy that is used for light-water reactor a kind of novelty, that have the good corrosion resistance energy.
The zirconium base alloy that is used for light-water reactor of the present invention is called the N36 alloy, and percentage composition meter by weight is made up of following ingredients:
Sn:0.80-1.20, Nb:0.90-1.25, Fe:0.12-0.45, O:0.06-0.15, C: less than 0.015, N: less than 0.008, surplus is zirconium and impurity.
According to the test detected result, the technical scheme that the present invention is further limited is: percentage composition meter by weight, and composed as follows: Sn:0.90-1.10, Nb:0.90-1.10, Fe:0.25-0.35, O:0.06-0.15, C: less than 0.015, N: less than 0.008, surplus is zirconium and impurity.
The above-mentioned preparation method who is used for the zirconium base alloy of light-water reactor provided by the present invention in turn includes the following steps:
(1) various components is mixed according to said ratio;
(2) component that mixes is made electrode, adopt vacuum consumable electrode arc furnace to carry out melting, make alloy cast ingot;
(3) alloy cast ingot is forged into the base material of desired shape at 900 ℃ of-1100 ℃ of β phase regions;
(4) with the base material 1000 ℃ of-1050 ℃ of β heat phase homogenizing and quench treatment;
(5) the base material after will quenching carries out hot-work at 600 ℃ of-700 ℃ of alpha phase zones;
(6) with the base material after the hot-work repeatedly cold working with 560 ℃ of-650 ℃ of α mutually temperature range carry out process annealing, until being processed into required section bar;
(7) in 440 ℃-620 ℃, carry out stress relieving or recrystallization annealing and handle, obtain the zirconium alloy finished-product material.
The present invention is on Zr-Sn-Nb alloy basis, add other and be used to improve the composition of alloy property, and selected suitable component concentration, especially for the addition of Nb and O, changed in the prior art its understanding aspect the content restriction in zirconium alloy, alloy property provided by the invention satisfies the requirement of light-water reactor high burnup to core structural material.Improved in the out-pile pure water anti-uniform corrosion performance in lithium hydroxide aqueous solution particularly by the pipe plate product of this prototype alloy preparation, improved the anti-nodular corrosion performance in high-temperature steam.By the test detected result in the embodiment, can think these alloys in reactor, use have better anti-all even nodular corrosion performance, higher creep resistance and fatigue characteristic, anti-irradiation growth performance.
The material of complete processing provided by the invention preparation can guarantee to have good use properties in the reactor core rigorous environment by the equiaxial α-Zr crystal grain and equally distributed tiny second microtexture formed of particle mutually.
Embodiment
Below by embodiment the present invention is described in more detail.
Table 1 is the composition of alloy provided by the present invention, in the table 13
*Be Zr-4 alloy composition and corresponding experimental examination result, each content is the weight percent of respective components in alloy in the table 1.
Table 1 alloy composition provided by the present invention
| The alloy sequence number | Sn | Nb | Fe | Cr | O | C | N | Zr and impurity |
| 1 | 0.92 | 1.02 | 0.12 | --- | 0.15 | 0.012 | 0.008 | Surplus |
| 2 | 0.81 | 1.20 | 0.44 | --- | 0.10 | 0.014 | 0.007 | Surplus |
| 3 | 0.96 | 1.02 | 0.39 | --- | 0.08 | 0.013 | 0.005 | Surplus |
| 4 | 0.89 | 1.25 | 0.27 | --- | 0.06 | 0.012 | 0.010 | Surplus |
| 5 | 1.18 | 1.10 | 0.24 | --- | 0.09 | 0.010 | 0.009 | Surplus |
| 6 | 1.04 | 1.16 | 0.31 | --- | 0.10 | 0.013 | 0.004 | Surplus |
| 7 | 0.88 | 1.07 | 0.27 | --- | 0.14 | 0.011 | 0.005 | Surplus |
| 8 | 1.01 | 0.90 | 0.30 | --- | 0.12 | 0.014 | 0.007 | Surplus |
| 9 | 0.98 | 1.04 | 0.35 | --- | 0.09 | 0.012 | 0.010 | Surplus |
| 10 | 0.91 | 0.97 | 0.31 | --- | 0.13 | 0.015 | 0.009 | Surplus |
| 11 | 0.86 | 1.20 | 0.12 | --- | 0.10 | 0.020 | 0.012 | Surplus |
| 12 | 1.00 | 1.00 | 0.30 | --- | 0.10 | 0.010 | 0.006 | Surplus |
| 13 * | 1.50 | 0.00 | 0.22 | 0.12 | 0.09 | 0.014 | 0.008 | Surplus |
To being used for the Zirconium alloy material of nuclear reactor, the corrosion resistance nature of alloy is the factor of overriding concern, to consider when production cost and workability are selected alloying element on this basis, therefore, need study each alloying element in great detail to the influence of erosion resistance, mechanical property and creep behaviour and the amount ranges of alloy system and every kind of alloying element.Zirconium base alloy of the present invention has better anti-all even nodular corrosion performance, has higher creep resistance and fatigue characteristic, has anti-irradiation growth performance, and particular case is as follows:
(1) zirconium (Zr)
By the consideration to the neutron absorption factor, the present invention selects zirconium as fundamental element, also considers the neutron absorbing state of adding other alloying elements in the basic zirconium to simultaneously.
(2) tin (Sn)
Tin can stabilised zirconia α-phase, can increase its intensity, and can offset nitrogen the corrosive deleterious effect.When the tin consumption after a little while, can not reach required effect.Sn adds content at 0.80-1.20 weight % among the present invention, and it can guarantee that alloy has superior corrosion resistance energy and good mechanical performance.
(3) niobium (Nb)
Niobium can stabilised zirconia β-phase, niobium has higher strengthening effect to zirconium.The niobium consumption is too much to the thermal treatment sensitivity.Nb adds content at 0.90-1.25 weight % among the present invention, it can guarantee that alloy has in superior corrosion resistance energy and the good mechanical performance in pure water and lithium hydroxide aqueous solution, processing characteristics is also uninfluenced, need be limited in below 0.6% unlike content in the alloy of mentioning in the prior art.
(4) iron (Fe)
Iron can improve alloy corrosion resistance and mechanical property, but the consumption of iron is too much or very fewly all adverse influence can be arranged.The content that Fe adds among the present invention is at 0.12-0.45 weight %, and it can guarantee that alloy has the superior corrosion resistance energy in pure water and lithium hydroxide aqueous solution.
(5) oxygen (O)
Oxygen can stabilised zirconia α-phase, add oxygen in the alloy and can improve yield strength.The content that oxygen adds among the present invention is at 0.06-0.15 weight %, and it can guarantee that alloy has enough mechanical propertys and creep-resistant property.The increase of oxygen level greatly reduces the control difficulty in the material processing.
(6) carbon (C)
Carbon in the alloy exists as the unavoidable impurities element and content when higher, can reduce the corrosion resistance of alloy.The weight percent of C is less than 0.015% among the present invention, and it can guarantee that alloy has the superior corrosion resistance energy in high-temperature water and steam.
(7) nitrogen (N)
Nitrogen in the alloy exists as the unavoidable impurities element and content when higher, can reduce the corrosion resistance of alloy.The weight percent of N is less than 0.008% among the present invention, and it can guarantee that alloy has the superior corrosion resistance energy in high-temperature water and steam.
The zirconium base alloy that the present invention is used for the reactor core structure material prepares by the following method:
(1) various components is mixed according to said ratio;
(2) component that mixes is made electrode, adopt vacuum consumable electrode arc furnace to carry out melting, make alloy cast ingot;
(3) alloy cast ingot is forged into the base material of desired shape at 900 ℃ of-1100 ℃ of β phase regions;
(4) with the base material 1000 ℃ of-1050 ℃ of β heat phase homogenizing and quench treatment;
(5) the base material after will quenching carries out hot-work at 600 ℃ of-700 ℃ of alpha phase zones;
(6) with repeatedly cold working and carry out process annealing of the base material after the hot-work, until being processed into required section bar 560 ℃ of-650 ℃ of temperature ranges;
(7) in 440 ℃-620 ℃, carry out stress relieving or recrystallization annealing and handle, obtain the zirconium alloy finished-product material.
By the equiaxial α-Zr crystal grain and equally distributed tiny second microtexture formed of particle mutually, can guarantee in the reactor core rigorous environment, to have good use properties by the material of above-mentioned complete processing preparation.By the alloy material of method for preparing, its performance test results is shown in table 2 and table 3.
The erosion rate of table 2 alloy material provided by the present invention in high-temperature water
The erosion rate of table 3 alloy material provided by the present invention in water vapour
From table 2 and table 3 kind as can be seen, except that the higher alloy of N and C content, remaining alloy and has shown excellent corrosion resisting performance in 400 ℃ and 500 ℃ of steam all in 360 ℃ of pure water, lithium hydroxide aqueous solution.
Claims (3)
1. zirconium base alloy that is used for light-water reactor, percentage composition meter by weight, form by following ingredients:
Sn:0.80-1.20, Nb:0.90-1.25, Fe:0.12-0.45, O:0.06-0.15, C: less than 0.015, N: less than 0.008, surplus is zirconium and impurity.
2. the zirconium base alloy that is used for light-water reactor as claimed in claim 1 is characterized in that percentage composition meter by weight, and is composed as follows: Sn:0.90-1.10, Nb:0.90-1.10, Fe:0.25-0.35, O:0.08-0.12, C: less than 0.015, N: less than 0.008, surplus is zirconium and impurity.
3. the zirconium base alloy that is used for light-water reactor as claimed in claim 2 is characterized in that percentage composition meter by weight, and is composed as follows: Sn:1.00, and Nb:1.00, Fe:0.30, O:0.10, C: less than 0.015, N: less than 0.008, surplus is zirconium and impurity.
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| CN103194650A (en) * | 2013-04-10 | 2013-07-10 | 苏州热工研究院有限公司 | Preparation method of Zr-1Nb alloy |
| CN103898369A (en) * | 2012-12-27 | 2014-07-02 | 中国核动力研究设计院 | Zirconium alloy for nuclear reactor |
| CN103898363A (en) * | 2012-12-27 | 2014-07-02 | 中国核动力研究设计院 | Zirconium alloy for nuclear power |
| CN105400997A (en) * | 2015-12-09 | 2016-03-16 | 上海大学 | Germanium-bismuth-containing zirconium-niobium alloy for nuclear power station fuel cladding |
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| CN112775202A (en) * | 2020-12-23 | 2021-05-11 | 西部新锆核材料科技有限公司 | Preparation method of zirconium or zirconium alloy rolled section |
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| Publication number | Priority date | Publication date | Assignee | Title |
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| EP0643144B1 (en) * | 1993-03-04 | 1997-12-29 | Vsesojuzny Nauchno-Issledovatelsky Institut Neorga Nicheskikh Materialov Imeni Akademika A.A. Bochvara, | Zirconium-based material, article made of the said material for use in the active zones of atomic reactors, and a process for obtaining such articles |
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| GB2523976A (en) * | 2012-12-27 | 2015-09-09 | Nuclear Power Inst China | Zirconium alloy for nuclear power |
| CN103194650B (en) * | 2013-04-10 | 2015-08-19 | 苏州热工研究院有限公司 | A kind of preparation method of Zr-1Nb alloy |
| CN103194650A (en) * | 2013-04-10 | 2013-07-10 | 苏州热工研究院有限公司 | Preparation method of Zr-1Nb alloy |
| CN105400997A (en) * | 2015-12-09 | 2016-03-16 | 上海大学 | Germanium-bismuth-containing zirconium-niobium alloy for nuclear power station fuel cladding |
| CN107686902A (en) * | 2017-07-10 | 2018-02-13 | 中国核动力研究设计院 | A kind of nuclear grade zirconium alloy cast ingot preparation method |
| CN107686902B (en) * | 2017-07-10 | 2019-08-13 | 中国核动力研究设计院 | A kind of nuclear grade zirconium alloy cast ingot preparation method |
| CN112775202A (en) * | 2020-12-23 | 2021-05-11 | 西部新锆核材料科技有限公司 | Preparation method of zirconium or zirconium alloy rolled section |
| CN112775203A (en) * | 2020-12-23 | 2021-05-11 | 西部新锆核材料科技有限公司 | Preparation method of zirconium or zirconium alloy extruded section |
| CN112775203B (en) * | 2020-12-23 | 2024-01-19 | 西部新锆核材料科技有限公司 | Preparation method of zirconium or zirconium alloy extrusion profile |
| WO2023087898A1 (en) * | 2021-11-18 | 2023-05-25 | 中国核动力研究设计院 | Fuel assembly grid, fuel assembly, and pressurized water reactor core |
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