CN107034385A - For Reactor fuel element cladding material zirconium alloy in non-deoxidization by adding hydrogen presurized water reactor - Google Patents
For Reactor fuel element cladding material zirconium alloy in non-deoxidization by adding hydrogen presurized water reactor Download PDFInfo
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- CN107034385A CN107034385A CN201710077447.2A CN201710077447A CN107034385A CN 107034385 A CN107034385 A CN 107034385A CN 201710077447 A CN201710077447 A CN 201710077447A CN 107034385 A CN107034385 A CN 107034385A
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- deoxidization
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- 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
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- 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
- G21C3/06—Casings; Jackets
- G21C3/07—Casings; Jackets characterised by their material, e.g. alloys
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- 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
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Abstract
The present invention relates to Reactor fuel element cladding zircaloy in the non-deoxidization by adding hydrogen presurized water reactor of one kind, belong to Zirconium alloy material technical field.The chemical composition of the zircaloy is by weight percentage：0.73%~1.1%Sn, 0.25%~0.6%Fe, 0.1%~0.25%Cr, surplus are Zr and inevitable impurity.The zircaloy of the present invention is free of niobium element, therefore it is insensitive to the dissolved oxygen in corrosive medium, this alloy when corroding respectively during 4 kinds of etching conditions include 500 DEG C/10.3 MPa superheated steams, 400 DEG C/10.3 MPa superheated steams, 360 DEG C/18.6 MPa/0.01 M LiOH aqueous solution and 360 DEG C/18.6 MPa deionized waters simultaneously, all show very excellent decay resistance, hence it is evident that better than the alloys of Zr 4.As Reactor fuel element cladding material in non-deoxidization by adding hydrogen presurized water reactor.
Description
Technical field
It is used for Reactor fuel element cladding material zirconium alloy in non-deoxidization by adding hydrogen presurized water reactor the present invention relates to one kind, belongs to zirconium conjunction
Golden field of material technology.
Background technology
Zircaloy has special nuclearity can (thermal neutron absorption cross section is 0.18 barn), excellent decay resistance and suitable
In mechanical property, and be widely used as the cladding materials of water cooled nuclear reactor fuel element, be in pressurized-water reactor nuclear power plant very
Important structural material.Zirconium alloy cladding is operationally aoxidized because of the corrosion by high-temperature high pressure water, makes zirconium alloy cladding
Effective thickness is thinned, and influences its service life.Zr-4 (Zr-1.5Sn-0.2Fe-0.1Cr) alloys are from middle 1960s
Have started to be widely used in water cooling reactor of nuclear power plant, and show excellent decay resistance.But in order to further carry
The burnup of high nuclear fuel, reduces nuclear power cost, it is necessary to take the extension refulling cycle, improves the measures such as coolant temperature, this is just
Propose higher requirement to the water-fast side corrosive nature of zirconium alloy cladding, the most Zr-4 alloys of commercial it is corrosion-resistant
Performance can not meet the requirement that burnup is further increased to 55 GWd/tU.
In the presurized water reactor of non-deoxidization by adding hydrogen, oxygen content in cooling water will be apparently higher than in the presurized water reactor of deoxidization by adding hydrogen
Oxygen content, material impact will be produced to the decay resistance of zircaloy.In the presurized water reactor of non-deoxidization by adding hydrogen, Zr-4 alloys can be sent out
Raw obvious nodular corrosion；Corrosion test in the MPa superheated steams of heap external application 500 DEG C/10.3 characterizes nodular corrosion row
To find that nodular corrosion also occurs in Zr-4 alloys.In addition, Zr-4 alloys are in the MP a/0.01 M of out-pile 360 DEG C/18.6
Also obvious corrosion can occur in the LiOH aqueous solution to accelerate.Addition alloying elements nb can be obviously improved the shape of resistance to furuncle of Zr-4 alloys
Corrosive nature, but the zircaloy containing niobium is more sensitive to the dissolved oxygen concentration in corrosive medium.Dissolved oxygen in water mainly comes
The RADIATION DECOMPOSITION reaction of reactor core reclaimed water is come from, the presence of dissolved oxygen can have adverse effect on to the corrosion behavior of zircaloy.One
As in presurized water reactor primary Ioops by deoxidization by adding hydrogen, can be by dissolved oxygen content control less than 5.0 × 10-9(mass fraction), even
It is lower, but in some specific use presurized water reactors, do not carrying out deoxidization by adding hydrogen processing, dissolved oxygen content up to 0.2 ×
10-6(mass fraction), hence it is evident that higher than the presurized water reactor of deoxidization by adding hydrogen.It will be shown to dissolved oxygen containing 0.3% niobium in zircaloy
Sensitiveness.
The content of the invention
It is an object of the invention to provide a kind of zircaloy without niobium element, nodular corrosion does not occur for this zircaloy, its
Decay resistance is because insensitive to the dissolved oxygen in corrosive medium without niobium and corrosion-resistant under 4 kinds of etching conditions of out-pile
Performance is all substantially better than commercial Zr-4 alloys, the Reactor fuel element cladding of the specific use presurized water reactor as non-deoxidization by adding hydrogen
Material.
The purpose of the present invention is by further on the basis of commercial fuel for nuclear power plant involucrum Zr-4 alloying components
Theil indices are reduced while iron and chromium are come what is realized in raising alloy, its technical scheme is as follows：
One kind is used for Reactor fuel element cladding material zirconium alloy in non-deoxidization by adding hydrogen presurized water reactor, and the chemical composition of the zircaloy is with weight
Measuring percentages is：0.73%~1.1%Sn, 0.25%~0.6%Fe, 0.1%~0.25%Cr, surplus are Zr and inevitably miscellaneous
Matter.
Above-mentioned to be used for Reactor fuel element cladding material zirconium alloy in non-deoxidization by adding hydrogen presurized water reactor, its alloying element is with weight hundred
Divide and be than meter preferred scope：0.73%~1.0%Sn, 0.25%~0.5%Fe, 0.14%~0.23%Cr, surplus are Zr and can not kept away
The impurity exempted from.
Above-mentioned to be used for Reactor fuel element cladding material zirconium alloy in non-deoxidization by adding hydrogen presurized water reactor, its alloying element is with weight hundred
Divide and be than meter preferred scope：0.75%~0.9%Sn, 0.3%~0.45%Fe, 0.15%~0.2%Cr, surplus are Zr and inevitable
Impurity.
Above-mentioned to be used for Reactor fuel element cladding material zirconium alloy in non-deoxidization by adding hydrogen presurized water reactor, its alloying element is with weight hundred
Divide and be than meter preferred scope：0.75%~0.8%Sn, 0.36%~0.4%Fe, 0.15~0.18%Cr, surplus are Zr and inevitable
Impurity.
Above-mentioned to be used for Reactor fuel element cladding material zirconium alloy in non-deoxidization by adding hydrogen presurized water reactor, its alloying element is with weight hundred
Ratio is divided to be calculated as：0.78%Sn, 0.38%Fe, 0.16%Cr, surplus are Zr and inevitable impurity.
The present invention by the basis of Zr-4 alloying components further reduction Theil indices while improve alloy in iron and
The content of chromium, due to the reciprocation of these alloying elements, brings the advantageous effects of the present invention.The effect of the present invention：
The application example that the present invention is provided shows, alloy 4 kinds of etching conditions include 500 DEG C/10.3 MPa superheated steams, 400 DEG C/
In 10.3 MPa superheated steams, 360 DEG C/18.6 MPa/0.01 M LiOH aqueous solution and 360 DEG C/18.6 MPa deionized waters
When corroding respectively, very excellent decay resistance is all shown, hence it is evident that better than Zr-4 alloys.For example in 500 DEG C of superheated steams
During 500 h of middle corrosion, there is not nodular corrosion spot in Zr alloy surface of the present invention, and surrosion reaches 155 mg.dm-2, and Zr-
113 mg.dm are reached during 47 h of alloy corrosion-2, corrosion sample surfaces oxide-film also peeled off.The alloy of the present invention
Composition has excellent decay resistance, particularly resistance to Nodular Corrosion, with good application prospect.
Brief description of the drawings
Fig. 1 is alloy of the present invention and surrosion curve of the Zr-4 alloys in 500 DEG C/10.3 MPa superheated steams；
Fig. 2 inventions alloy and surrosion curve of the Zr-4 alloys in 400 DEG C/10.3 MPa superheated steams；
Fig. 3 inventions alloy and surrosion curve of the Zr-4 alloys in 360 DEG C/18.6 MPa/0.01 M LiOH aqueous solution.
Embodiment
The zirconium tin system alloy of the fine corrosion resistance of the present invention is described in further detail with reference to embodiment, but
The invention is not restricted to following examples.
Embodiment one：Referring to table 1, there is shown the composition of zirconium tin system's alloy of the present invention and Zr-4 alloys composition：
。
Prepared in accordance with the following steps with the alloy material constituted in table 1：
(1) above-mentioned formula dispensing is pressed, commission factory is prepared into the alloy pig of about 20 kg weights with common process；
(2) above-mentioned alloy pig is forged into blank material at 950 ~ 1050 DEG C, while broken thick as-cast grain structure；
(3) blank material is after scale removal and pickling, in a vacuum through 1030~1050 DEG C of β phases Homogenization Treatments 0.5
Air cooling after~1 h；Through 700 ~ 800 DEG C of hot rollings after, first scale removal after hot rolling；
(4) repeatedly cold rolling and 580 DEG C of intermediate annealings are carried out after blank material air cooling, 580 DEG C of recrystallizations are finally carried out in a vacuum and are moved back
2 h of fire, are prepared into thick 2.8 mm sheet alloy;
(5) above-mentioned sheet material is cut into the mm of the mm of 22 mm × 15 × 2.8 sheet sample with wire electric discharge, corrosion is prepared into and uses
Sample.
The alloy 1 prepared by above-mentioned technique is together put into autoclave with the Zr-4 alloy samples Jing Guo same preparation technology
In, corrosion test is carried out in 500 DEG C/10.3 MPa superheated steams, their corrosion behavior is investigated：Fig. 1 gives two kinds of conjunctions
Surrosion curve of the gold in 500 DEG C of superheated steams, when corroding in 500 DEG C of superheated steams during 500 h, zirconium of the present invention is closed
There is not nodular corrosion spot in gold surface, and surrosion reaches 155 mg.dm-2, and reach 113 during Zr-4 7 h of alloy corrosion
mg.dm-2, sample shows that oxide-film is also peeled off.The alloy of the present invention shows excellent in 500 DEG C of superheated steams
Resistance to Nodular Corrosion.
Prepared by alloy 1 and technique into identical invention alloy together to put with the Zr-4 alloy samples Jing Guo same preparation technology
Enter in autoclave, carry out corrosion test in 400 DEG C/10.3 MPa superheated steams, investigate their corrosion behavior：Fig. 2 is provided
After surrosion curves of two kinds of alloys in 400 DEG C of superheated steams, corrosion 400 days, zircaloy of the invention is closed with Zr-4
The surrosion of gold is respectively 127 mg.dm-2And 200mg.dm-2.Corrosion resistant of the alloy of the present invention in 400 DEG C of superheated steams
Corrosion can be substantially better than Zr-4 alloys.
Alloy 1 is together put into autoclave with the Zr-4 alloy samples Jing Guo same preparation technology, 360 DEG C/18.6
Corrosion test is carried out in the MPa/0.01 M LiOH aqueous solution, their corrosion behavior is investigated：Fig. 3 gives two kinds of alloys 360
DEG C/the 18.6 surrosion curve in the MPa/0.01 M LiOH aqueous solution, and after corrosion 190 days, zircaloy and Zr- of the invention
The surrosion of 4 alloys is respectively 67 mg.dm-2With 752 mg.dm-2, Zr-4 alloy samples oxidation film outer surface turn white, aoxidize
Film is peeled off.The alloy of the present invention was continued to corrode to 340 days, its surrosion is only 358 mg.dm-2.The conjunction of the present invention
Decay resistance of the gold in 360 DEG C/18.6 MPa/0.01 M LiOH aqueous solution is substantially better than Zr-4 alloys.
Alloy 1 is together put into autoclave with the Zr-4 alloy samples Jing Guo same preparation technology, 360 DEG C/18.6
Corrosion test is carried out in MPa deionized waters, their corrosion behavior is investigated：Corrode in 360 DEG C/18.6 MPa deionized waters
After 190 days, the surrosion of zircaloy Yu Zr-4 alloys of the invention is respectively 43 mg.dm-2With 57 mg.dm-2, the present invention
Decay resistance of the alloy in 360 DEG C/18.6 MPa deionized waters also superior to Zr-4 alloys.
As can be seen here, decay resistance of the alloy of the present invention under 4 kinds of etching conditions is all substantially better than Zr-4 alloys, especially
It is that, without occurring nodular corrosion, the use of specific use presurized water reactor cladding nuclear fuels material can be met in 500 DEG C of superheated steams
It is required that.
Above-described embodiment, simply the present invention section Example, not for limit the present invention practical range, therefore it is all with
The equivalence changes that content described in the claims in the present invention is done, all should be included within scope of the invention as claimed.
Claims (5)
1. one kind is used for Reactor fuel element cladding material zirconium alloy in non-deoxidization by adding hydrogen presurized water reactor, it is characterised in that the zircaloy
Chemical composition is by weight percentage：0.73%~1.1%Sn, 0.25%~0.6%Fe, 0.1%~0.25%Cr, surplus be Zr and
Inevitable impurity.
2. it is used for Reactor fuel element cladding material zirconium alloy, its feature in non-deoxidization by adding hydrogen presurized water reactor as described in claim 1
It is：By weight percentage, 0.73%~1.0%Sn, 0.25%~0.5%Fe, 0.14%~0.23%Cr, surplus are Zr and can not
The impurity avoided.
3. it is used for Reactor fuel element cladding material zirconium alloy, its feature in non-deoxidization by adding hydrogen presurized water reactor as described in claim 1
It is：By weight percentage, 0.75%~0.9%Sn, 0.3%~0.45%Fe, 0.15%~0.2%Cr, surplus are Zr and can not
The impurity avoided.
4. it is used for Reactor fuel element cladding material zirconium alloy, its feature in non-deoxidization by adding hydrogen presurized water reactor as described in claim 1
It is：By weight percentage, 0.75%~0.8%Sn, 0.36%~0.4%Fe, 0.15~0.18%Cr, surplus are Zr and can not
The impurity avoided.
5. it is used for Reactor fuel element cladding material zirconium alloy, its feature in non-deoxidization by adding hydrogen presurized water reactor as described in claim 1
It is：By weight percentage, 0.78%Sn, 0.38%Fe, 0.16%Cr, surplus are Zr and inevitable impurity.
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| CN201710077447.2A CN107034385A (en) | 2017-02-14 | 2017-02-14 | For Reactor fuel element cladding material zirconium alloy in non-deoxidization by adding hydrogen presurized water reactor |
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| CN201710077447.2A CN107034385A (en) | 2017-02-14 | 2017-02-14 | For Reactor fuel element cladding material zirconium alloy in non-deoxidization by adding hydrogen presurized water reactor |
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Cited By (1)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| 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 |
-
2017
- 2017-02-14 CN CN201710077447.2A patent/CN107034385A/en active Pending
Cited By (1)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| 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 |
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Application publication date: 20170811 |