CN101586201B - Nuclear-used zirconium alloy with excellent corrosion resistance - Google Patents
Nuclear-used zirconium alloy with excellent corrosion resistance Download PDFInfo
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- CN101586201B CN101586201B CN2009100538846A CN200910053884A CN101586201B CN 101586201 B CN101586201 B CN 101586201B CN 2009100538846 A CN2009100538846 A CN 2009100538846A CN 200910053884 A CN200910053884 A CN 200910053884A CN 101586201 B CN101586201 B CN 101586201B
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Abstract
The invention relates to a zirconium alloy material for cladding a nuclear fuel element for a pressurized-water reactor (PWR for short), belonging to the technical field of zirconium alloy material. Main compositions of the alloy and weight percentages thereof are as follows: 0.8-1.4% of Sn, 01- 0.3% of Nb, 0.3-0.5% of Fe, 0.07- 0.25% of Cr, 0.05- 0.3% of Cu, 0.013- 0.05% of Si, and the balance isZr. The alloy of the invention is obtained by adding trace alloy element Si based on the compositions of prior zirconium alloy, and then optimizing the content proportion of Sn, Nb, Fe, Cr and Cu. The alloy provided by the invention performs excellent corrosion resistance both in the 0.01 M LiOH aqueous water at 360 DEG C with pressure of 18.6 MPa and in the overheated steam of 400 DEG C, performing better than the unoptimized zirconium alloy, ZIRLO and Zr-4 alloy. The invention can be used as the material for core structure such as cladding of fuel element and framework in PWR power station.
Description
Technical field
The present invention relates to the zirconium-based alloy material that a kind of pressurized-water reactor nuclear power plant fuel assembly involucrum and screen work are used, belong to the Zirconium alloy material technical field.
Background technology
Zirconium alloy is a kind of important core structural material in the Nuclear power plants, as the involucrum of nuclear fuel element.The water-fast side corrosive nature of zirconium alloy cladding is to influence fuel element topmost factor in work-ing life.In the development and optimizing components of zirconium alloy, usually earlier filter out the alloy of fine corrosion resistance, and then make fuel stick and be placed on and carry out the irradiation test in the test reactor by out-pile autoclave corrosion test, understand its corrosion behavior in heap.Owing in a loop water, added H during the pressurized-water reactor nuclear power plant operation
3BO
3, use
10B controls and regulates superfluous nuclear reactivity as burnable poison, so need to adopt alkaline waters (pH:7.1～7.2) with the release that reduces various steel corrosion of component products in the loop and the migration of radioactive substance, reduce staff's raying dosage level.The method that LiOH is added in present most of pressurized-water reactor employing is regulated the pH value in the loop water.In addition, the superheated vapour corrosion test of 400 ℃/10.3MPa commonly used is checked the corrosion resistance nature of Zr-4 compo pipe in the production.Therefore, being used for out-pile now checks the test of zirconium alloy corrosion resistance nature mainly to adopt the superheated vapour of the 360 ℃/18.6MPa/0.01M LiOH aqueous solution and 400 ℃/10.3MPa.
In order to reduce the cost of nuclear power, require further to improve the burnup of nuclear fuel, so just need to prolong the time that nuclear fuel assembly stops in reactor core, this corrosion resistance nature to can material is had higher requirement, thereby having promoted the development of zirconium alloy, various countries have carried out the R﹠D work of high-performance zirconium alloy one after another.The zirconium alloy of exploitation mainly contains Zr-Sn, Zr-Nb and Zr-Sn-Nb three big series in the world at present.After having added alloying elements such as Fe, Cr, Ni, Cu on this basis, formed zirconium alloys such as the Zr-2, the Zr-4 that have used, Zr-2.5Nb, E110, M5, ZIRLO, E635, and the zirconium alloy (composition sees Table 1) such as N18, N36, HANA with application prospect.Zr-2 and Zr-4 are the alloys of developing the earliest, belong to Zr-Sn system, and wherein Zr-2 is the can material that is used for boiling-water reactor, and Zr-4 is the can material that is used for pressurized-water reactor.When burnup when 33GWd/tU is following, conventional Zr-4 alloy cladding can meet the demands; When burnup was brought up to 40～50GWd/tU, modified version Zr-4 alloy (comprise and optimize the hot-work system and adopt the low tin content composition) involucrum just can meet the demands.Yet when burnup reached 60GWd/tU, modified version Zr-4 alloy can not meet the demands, must adopt new zirconium alloy to make involucrum.The corrosion resistance nature of ZIRLO alloy in 360 ℃/18.6MPa/0.01M of the out-pile LiOH aqueous solution of US Westinghouse company's exploitation obviously is better than Zr-4 alloy (Sabol, G.P., Kilp, G.R., Balfour, M.G., et al., Development of a cladding alloy for higher burnup.Zirconium in the Nuclear Industry:Eighth International Symposium, ASTM STP 1023,1989, pp.227-244.); Then the ZIRLO alloy being made fuel element tests in the BR3 test reactor, after average burn-up reached 71GWd/tU, the oxide thickness of ZIRLO alloy uniform corrosion was littler by 50% than Zr-4 alloy, and anti-irradiation growth and irradiation creep are also good than Zr-4 alloy, shown that very superior corrosion resistance can (Sabol in the heap, G.P., Comstock, R.J., Weiner, R.A., et al, In-reactor corrosion performance of ZIRLO
TMAnd Zircaloy-4.Zirconium in the NuclearIndustry:Tenth International Symposium, ASTM STP 1245,1994, pp.724-744.).This illustrates that also the corrosion behavior in 360 ℃/18.6MPa/0.01M of the out-pile LiOH aqueous solution can reflect preferably that really Zr-Sn-Nb is the corrosion situation of alloy in heap.
Table 1 is current mainly at the nuclear-used zirconium alloy of using and grinding
| Title | Alloy nominal composition (weight percentage %, down together) | Exploitation country | Remarks |
| Zr-2 | Zr-1.5Sn-0.2Fe-0.1Cr-0.05Ni | The U.S. | Use |
| Zr-4 | Zr-1Sn-1Nb-0.1Fe | The U.S. | Use |
| Zr-2.5Nb | Zr-2.5Nb | Canada | Use |
| Zr-1Nb | Zr-1Nb | USSR (Union of Soviet Socialist Republics) | Use |
| ZIRLO | Zr-1.0Sn-1.0Nb-0.1Fe | The U.S. | Use |
| M5 | Zr-1.0Nb-0.16O | France | Use |
| E635 | Zr-1.2Sn-1Nb-0.4Fe | Russia | Use |
| N18 | Zr-1Sn-0.4Nb-0.3Fe-0.1Cr | China | Grinding |
| N36 | Zr-1Sn-1Nb-0.3Fe | China | Grinding |
| HANA6 | Zr-1.1Nb-0.05Cu | Korea S | Grinding |
| HANA3 | Zr-1.5Nb-0.4Sn-0.1Fe-0.1Cu | Korea S | Grinding |
| HANA4 | Zr-1.5Nb-0.4Sn-0.2Fe-0.1Cr | Korea S | Grinding |
Existing result of study shows: now in the zirconium alloy of development and application, the proportioning of its composition might not be in optimum range, as on ZIRLO alloying constituent basis, Sn content being reduced to 0.75%, can also further improve corrosion resistance nature (Yueh, the H.K. of zirconium alloy, Kesterson, R.L., Comstock, R.J., et al., Improved ZIRLO
TMCladdingperformance through chemistry and process modifications[C] .Zirconium in the Nuclear Industry:Fourteenth International Symposium, ASTM STP 1467,2004, pp.330-346.).Korea Atomic Energy Research Institute provides a kind of zirconium alloy as fuel rod clad in the patent CN1128235C that China is authorized, the weight percentage of its component is Sn:0.8～1.6, Nb:0.05～0.3, Fe:0.25～0.5, Cr:0.05～0.25, any 0.05～0.20 among Mo, Mn or the Cu, O:0.06～0.14, Zr: surplus.Enumerated the weightening finish that the Zr-1.4Sn-0.22Nb-0.45Fe-0.22Cr-0.11Cu-0.1O alloy corroded 100 days in its preferred embodiment and (be respectively 33.2mg.dm in 360 ℃/18.6MPa deionized water and 400 ℃/10.3MPa superheated vapour
-2And 70.2mg.dm
-2) (be respectively 50.1mg.dm than Zr-4 alloy
-2And 85.8mg.dm
-2) low, think that thus this is a kind of zirconium alloy of fine corrosion resistance.In general, the surrosion data that need to obtain more than 300 days could be made evaluation to the quality of zirconium alloy corrosion resistance nature, because along with etching time prolongs, the good and bad order of different-alloy corrosion resistance nature changes through regular meeting.Obviously, this corrosion test time of 100 days is obviously too short, and judging whether of so making accurately also need further checking.In addition, do not carry out the corrosion test in the 360 ℃/18.6MPa/0.01M LiOH aqueous solution among this embodiment, be unable to find out the corrosion resistance nature of this alloy under this water chemistry condition.Therefore, on the basis of existing zirconium alloy, optimize the different proportionings of alloying constituent and also can develop the better zirconium alloy of corrosion resistance nature, to satisfy the needs that burnup improves constantly or/and add other kind alloying element.
Summary of the invention
The zirconium alloy that the purpose of this invention is to provide a kind of fine corrosion resistance, described zirconium alloy can be used as the material of core structure bodies such as fuel element can, screen work in the Nuclear power plants pressurized-water reactor.
The objective of the invention is the proportioning by changing existing zirconium alloy interalloy element and add that trace alloying element Si realizes, its technical scheme is as follows:
A kind of nuclear-used zirconium alloy of fine corrosion resistance, its chemical constitution and weight percentage are as follows:
0.8～1.4%Sn, 0.1～0.3%Nb, 0.3～0.5%Fe, 0.07～0.25%Cr, 0.05～0.3%Cu, 0.013～0.050%Si, surplus is Zr, and foreign matter content meets the standard of nuclear with the Zr-4 alloy, is: Al≤0.0075%, B≤0.00005%, Cd≤0.00005%, C≤0.015%, Co≤0.002%, Hf≤0.01%, H≤0.0025%, Mg≤0.002%, Mn≤0.005%, Mo≤0.005%, Ni≤0.007%, N≤0.008%, Ti≤0.005%, U≤0.00035%, W≤0.01%.
Above-mentioned zirconium alloy, the preferred weight percentage composition of Si is 0.013～0.025%.
Effect of the present invention is: alloy provided by the invention all shows very superior corrosion resistance energy in the 360 ℃/18.6MPa/0.01M LiOH aqueous solution and 400 ℃/10.3MPa superheated vapour, obviously be better than Zr-4 alloy and ZIRLO alloy, the alloy approaching with alloying constituent of the present invention (Zr-1.4Sn-0.22Nb-0.45Fe-0.22Cr-0.11Cu-0.1O) that also is better than providing among the patent CN1128235C reached the purpose of optimizing components.
Description of drawings
Fig. 1 ZRSHU1 alloy of the present invention (Zr-1.3Sn-0.2Nb-0.4Fe-0.1Cr-0.1Cu-0.02Si) and Zr-4 alloy and ZIRLO alloy are at (a) 360 ℃/18.6MPa/0.01M LiOH aqueous solution and (b) the surrosion curve in the 400 ℃/10.3MPa superheated vapour.
Embodiment
Below in conjunction with embodiment a kind of zirconium alloy that is used for the fine corrosion resistance of nuclear reactor structured material provided by the present invention is described in further detail.
Embodiment 1
Alloying constituent (weight percentage) is Sn:1.3%, Nb:0.2%, and Fe:0.4%, Cr:0.1%, Cu:0.1%, Si:0.02%, Zr are surplus (alloy is defined as the ZRSHU1 alloy); Foreign matter content meets the standard of present nuclear with the Zr-4 alloy.
Concrete preparation process is as follows:
(1) presses SHU1 alloying constituent prescription batching with nuclear level zirconium sponge and pure raw material (Sn, Nb, Fe, Cr, Cu and Si), with the alloy pig of vacuum non-consumable arc furnace melting into about the 100g weight, fill the high-purity argon gas protection during melting, and alloy is stood up melt back make alloy pig 5 times;
(2) above-mentioned alloy pig is carried out repeatedly hot pressing under 700 ℃, be processed into the base material, purpose is broken thick as-cast grain structure;
(3) with after 700 ℃ of hot rollings, grease is removed in first scale removal, pickling after the hot rolling, again air cooling after 1030～1050 ℃ β phase homogenizing is handled 0.5h in a vacuum;
(4) carry out behind the base material air cooling 2 times cold rolling, each cold roling reduction is 40%, carry out 600 ℃ of process annealing 1h in a vacuum between cold rolling per twice, make sheet material, carry out 600 ℃ of recrystallization annealing 1h at last in a vacuum, all carry out scale removal, the processing of pickling removal grease before each process annealing or the recrystallization annealing, promptly make the ZRSHU1 alloy material.
To together put into autoclave by the ZRSHU1 alloy and the Zr-4 alloy sample of above-mentioned prepared, in the 360 ℃/18.6MPa/0.01M LiOH aqueous solution and 400 ℃/10.3MPa superheated vapour, carry out corrosion test, investigate their corrosion behavior, the surrosion curve as shown in Figure 1, (data are selected from Sabol to have provided the surrosion curve that is used for correlated ZIRLO alloy among the figure simultaneously, G.P., Comstock, R.J., Weiner, R.A., et al, In-reactor corrosion performance ofZIRLO
TMAnd Zircaloy-4.Zirconium in the Nuclear Industry:Tenth International Symposium, ASTM STP 1245,1994, pp.724-744.).From accompanying drawing 1 as can be seen: when corroding 450 days the 360 ℃/LiOH aqueous solution, the weightening finish (240mg.dm of ZRSHU1 alloy of the present invention
-2) than ZIRLO alloy (280mg.dm
-2) little by 14%; And 220 days weightening finish of Zr-4 alloy corrosion is just up to 450mg.dm
-2(accompanying drawing 1a).When corroding 476 days in 400 ℃ of superheated vapours, the weightening finish (309mg.dm of ZRSHU1 alloy of the present invention
-2) than ZIRLO alloy (500mg.dm
-2) little by 38%, than Zr-4 alloy (330mg.dm
-2) little by 6% (accompanying drawing 1b).In addition, with patent CN1128235C in the alloy approaching (Zr-1.4Sn-0.22Nb-0.45Fe-0.22Cr-0.11Cu-0.1O) that provide with alloying constituent of the present invention in 400 ℃ of superheated vapours, corrode 100 days corrosion resistance nature (weightening finish be 70.2mg.dm
-2) compare, ZRSHU1 alloy of the present invention also demonstrates more excellent corrosion resistance nature, and (weightening finish is 60mg.dm
-2).As seen, on the alloying constituent basis that in patent CN1128235C, provides, the present invention is by adding micro-Si, Cr content takes off limit and inhales hydrogen to reduce, adjust Sn, Nb, Fe, Cr, Cu ratio optimization alloying constituent, because the interaction of micro-Si and Sn, Nb, Fe, Cr, Cu, significantly improved the corrosion resistance nature of alloy, reached the purpose of alloying constituent optimization.
Characteristics of the present invention are: pass through to add micro-Si on the basis of existing zircaloy composition, the content proportioning of optimizing Sn, Nb, Fe, Cr and Cu has obtained the very good zircaloy of a kind of decay resistance, has realized the purpose that alloying component is optimized.
Claims (2)
1. the nuclear-used zirconium alloy of a fine corrosion resistance is characterized in that its chemical constitution and weight percentage are as follows: 0.8～1.4%Sn, 0.1～0.3%Nb, 0.3～0.5%Fe, 0.07～0.25%Cr, 0.05～0.3%Cu, 0.013～0.050%Si, surplus is Zr, and foreign matter content meets the standard of nuclear with the Zr-4 alloy, be: Al≤0.0075%, B≤0.00005%, Cd≤0.00005%, C≤0.015%, Co≤0.002%, Hf≤0.01%, H≤0.0025%, Mg≤0.002%, Mn≤0.005%, Mo≤0.005%, Ni≤0.007%, N≤0.008%, Ti≤0.005%, U≤0.00035%, W≤0.01%.
2. zirconium alloy according to claim 1, the weight percentage that it is characterized in that Si is 0.013～0.025%.
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| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| CN101984114A (en) * | 2010-07-05 | 2011-03-09 | 大连理工大学 | Low-elastic modulus high-strength BCC Zr-Ti-Nb alloy |
| CN102140596B (en) * | 2011-01-12 | 2012-11-21 | 苏州热工研究院有限公司 | Zirconium-based alloy used for nuclear reactor |
| CN102766778B (en) * | 2011-05-04 | 2015-05-06 | 上海大学 | Zircaloy for fuel cladding at nuclear power station |
| CN102660699B (en) * | 2012-05-16 | 2014-02-12 | 上海大学 | Zr-Sn-Nb-Fe-Si alloy for fuel cladding of nuclear power station |
| CN103898360B (en) * | 2012-12-27 | 2016-08-31 | 中国核动力研究设计院 | A kind of nuclear reactor core zircaloy |
| CN105925846B (en) * | 2016-06-24 | 2018-02-23 | 西部新锆核材料科技有限公司 | A kind of Zr Sn Nb Hf alloy bar materials and preparation method thereof |
| JP7263745B2 (en) * | 2018-11-30 | 2023-04-25 | 株式会社プロテリアル | Zr alloys, Zr alloy products and Zr alloy parts |
| CN111254315A (en) * | 2020-03-30 | 2020-06-09 | 上海核工程研究设计院有限公司 | Furuncle-corrosion-resistant Zr-Sn-Fe-Cr-O alloy and preparation method thereof |
| CN111676389A (en) * | 2020-06-30 | 2020-09-18 | 上海大学 | Zirconium alloy cladding material for small water-cooled nuclear reactor and preparation method thereof |
Citations (2)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| US5211774A (en) * | 1991-09-18 | 1993-05-18 | Combustion Engineering, Inc. | Zirconium alloy with superior ductility |
| CN1128235C (en) * | 1998-02-04 | 2003-11-19 | 韩国原子力研究所 | Novel zirconium alloy used as fuel-rod coating layer |
-
2009
- 2009-06-26 CN CN2009100538846A patent/CN101586201B/en active Active
Patent Citations (2)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| US5211774A (en) * | 1991-09-18 | 1993-05-18 | Combustion Engineering, Inc. | Zirconium alloy with superior ductility |
| CN1128235C (en) * | 1998-02-04 | 2003-11-19 | 韩国原子力研究所 | Novel zirconium alloy used as fuel-rod coating layer |
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