GB2523975A - Zirconium alloy for nuclear power reactor core - Google Patents

Zirconium alloy for nuclear power reactor core Download PDF

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GB2523975A
GB2523975A GB1512801.0A GB201512801A GB2523975A GB 2523975 A GB2523975 A GB 2523975A GB 201512801 A GB201512801 A GB 201512801A GB 2523975 A GB2523975 A GB 2523975A
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alloy
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nuclear power
reactor core
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GB2523975B (en
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Wenjin Zhao
Zhongbo Yang
Zhi Miao
Xun Dai
Wei Yi
Zhaohua Huang
Jun Qiu
Chunrong Xu
Zhihai Liao
Pengfei Wang
Qionggen Dong
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Nuclear Power Institute of China
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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
    • C—CHEMISTRY; METALLURGY
    • C22—METALLURGY; FERROUS OR NON-FERROUS ALLOYS; TREATMENT OF ALLOYS OR NON-FERROUS METALS
    • C22F—CHANGING THE PHYSICAL STRUCTURE OF NON-FERROUS METALS AND NON-FERROUS ALLOYS
    • C22F1/00—Changing the physical structure of non-ferrous metals or alloys by heat treatment or by hot or cold working
    • C22F1/16—Changing the physical structure of non-ferrous metals or alloys by heat treatment or by hot or cold working of other metals or alloys based thereon
    • C22F1/18—High-melting or refractory metals or alloys based thereon
    • C—CHEMISTRY; METALLURGY
    • C22—METALLURGY; FERROUS OR NON-FERROUS ALLOYS; TREATMENT OF ALLOYS OR NON-FERROUS METALS
    • C22F—CHANGING THE PHYSICAL STRUCTURE OF NON-FERROUS METALS AND NON-FERROUS ALLOYS
    • C22F1/00—Changing the physical structure of non-ferrous metals or alloys by heat treatment or by hot or cold working
    • C22F1/16—Changing the physical structure of non-ferrous metals or alloys by heat treatment or by hot or cold working of other metals or alloys based thereon
    • C22F1/18—High-melting or refractory metals or alloys based thereon
    • C22F1/186—High-melting or refractory metals or alloys based thereon of zirconium or alloys based thereon
    • G—PHYSICS
    • G21—NUCLEAR PHYSICS; NUCLEAR ENGINEERING
    • G21C—NUCLEAR REACTORS
    • G21C13/00—Pressure vessels; Containment vessels; Containment in general
    • G21C13/02—Details
    • G—PHYSICS
    • G21—NUCLEAR PHYSICS; NUCLEAR ENGINEERING
    • G21C—NUCLEAR REACTORS
    • G21C21/00—Apparatus or processes specially adapted to the manufacture of reactors or parts thereof
    • 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
    • G—PHYSICS
    • G21—NUCLEAR PHYSICS; NUCLEAR ENGINEERING
    • G21C—NUCLEAR REACTORS
    • G21C5/00—Moderator or core structure; Selection of materials for use as moderator
    • G21C5/02—Details
    • 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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  • Engineering & Computer Science (AREA)
  • Physics & Mathematics (AREA)
  • Chemical & Material Sciences (AREA)
  • Plasma & Fusion (AREA)
  • General Engineering & Computer Science (AREA)
  • High Energy & Nuclear Physics (AREA)
  • Metallurgy (AREA)
  • Materials Engineering (AREA)
  • Mechanical Engineering (AREA)
  • Organic Chemistry (AREA)
  • Manufacturing & Machinery (AREA)
  • Thermal Sciences (AREA)
  • Crystallography & Structural Chemistry (AREA)
  • Powder Metallurgy (AREA)
  • Manufacture And Refinement Of Metals (AREA)

Abstract

Disclosed is a zirconium alloy material for a nuclear power reactor core, falling within the technical field of special alloy materials, and comprising the following components in percentage by weight: 0.4-0.8 percent of Sn, 0.75-1.1 percent of Nb, 0.20-0.50 percent of Fe + Cr, 0.20-0.35 percent of Fe/(Fe + Nb), 0.01-0.1 percent of Cu or Bi or Ge, 0.002-0.02 percent of Si or S, 0.06-0.15 percent of O, less than 0.008 percent of C, less than 0.006 percent of N, and the balance being Zr. On the basis of a Zr-Sn-Nb system alloy, other components for improving the performance of the alloy are added, not only improving the corrosion resistance of the alloy, but also improving the mechanical properties and radiation resistance of the alloy, so that the requirements of high burnup of the nuclear power reactor on the structural materials of the core are met.

Description

ficat ion Zirconium Alloy for Nuclear Power Reactor Core
Thchnical Field
The present invention belongs to the technical field of special alloys, specifically to a zirconium alloy fur nuclear power reactor core.
uround Art Due to its low neutron absorption cross section, excellent corrosion resistance and mechanical properties. etc, zirconium alloy has been widely used as nuclear reactor fuel element cladding an.d other reactor components. in the development process of pressurized water reactors, the fuel design needs high requirements for the reactor core structural components, such as fuel element cladding, grids, guide tubes, etc. In early period, these parts were usually made of Zr4 alloy. Design of high fuel hurnup requires extension of duration time of these corn...i)one*nts in the reactor and increasing the temperature of coolants, is resulting in that zirconium alloy components face more rigorous corrosive environment, These high requirements promote the research of improvement corrosion resistance property of Zr-4 alloy and the development of new zirconium alloys with more excellent corrosion resistance.
In light of the high requirements for *iuel cladding proposed by nuclear power technology. international study of new zirconium alloys is developed. For example, at the Tenth International Symposium on Zirconium Alloys, GEORGE R SABOL reported "[n4teactor Corrosion Performance of Z1RLO and ZircaIoy4" ("inReactor Corrosion Perulbnnance of ZJRLO and Zircaioy4", Zirconium in the Nuclear industry: Tenth International Symposium, ASTM STP 1245, AM Garde and ER. Bradley, Eds, American Society for Testing and Materials, Phiiadefphia, 1994. pp.724-144), demonstrating that ZIR1.X) has better in'i'eactor corrosion resistance than Zirca.ioy4. At the l International Symposium on Zirconium Alloys, Nikuhna, .A\T from Russia reported "Zirconium Alloy E635 as a Material for Fuel Rod Cladding and Other Components of VVER an.d i.BMK Cor& ("Zirconium Alloy E635 as a Material for Fuel Rod Cladding and Other Components of VVER and RBMK Cores", Zirconium in the Nuclear industry: Eleventh International Symposium, ASTM STP 1295, ER Bradley and GP Sahol, Eds, American Society for Testing and Materials, Philadelphia, 1996, pp.785404)., announcing that the components of E635 include Zri £ 1,4wt%Nb..0,9'H Jwt%Sn -M3---0.Swt%Fe. [he out.-ofpiie properties of this alloy are superior to Zircaloy4 and EllO alloy. At the Twelfth International Is Symposium on Zircon. iirn Alloy, French Jean.Pau] Mard.on reported "Influence of Composition and Fabrication Process on Outof-Piie and In-PiIe Properties otM5 Alloy" ("influence of Composition and Fabrication Process on OutofPile and In Pile Properties of MS Alloy, Zirconium in the Nuclear Industry: Twelfth international Symposium, ASTM S'FP 1354, Sabol, 0, F, Moan. GD, 13th, American Society for Testing and Materials, West Conshohocken, 2000, pp.505 524. ), announcing MS alloy (Zr-lNb-O) with corrosion resistance superior to Zircaloy4 in high bumup (> 65GWd) condition, At the Sixteenth International A, Symposium on Zirconium Alloy, American A.M.Garde reported "Advanced Zirconium Alloy for PWR Application" ("Advanced Zirconium Alloy for PWR Application, Zirconium in the Nuclear Industry: sixteenth international Symposium, ASTM SW 1529, 2010, pp.784 -826), announcing X5A alloy $ (Zr-O.5Sn-0.3Nb-0.35Fe-O.25Cr) with in-pile and out-of-pile properties superior to ZIRLO alloy.
Studies have shown that the existing ratio of zirconium atloy components may not be within the optimal range. Lowering content of Sn in ZIRLO alloy can further improve its corrosion resistance (Yueh, HK, Kesterson, RL, Comstock,.., 0 Ri, et at, Improved ZLRLOTM cladding performance through chemistiy and process modifications Zirconium in the Nuclear Industry:. Fourteenth International Symposium, ASTM STP1467, 2004, pp 330-346); adding a little Cu (0.O5wt%) to Zr-Nb alloy to form HANA-6 alloy, which also has extremely excellent corrosion resistance (Park JY, Choi, BK, Yoo, Si Jeong YH, Corrosion i 5 behavior and oxide properties of Zr-1.1 wt% Nb-0,05 wt%. Cu alloy, I. NucI Mater, 359(2006) 59-68). MS alloy shows bending of fuel rods or fuel assemblies and poor growth performance and resistance to radiation during in-pile operation process. Thus, France adds a little Sn and Fe based on M5 alloy components so as to maintain the excellent corrosion resistance of the alloy and significantly 2o improve the mechanical properties, especially the radiation growth and creep properties. Therefore, optimizing alloy composition ratio or adding other alloy elements based on the existing zirconium alloys can develop zirconium alloys *4.
with more excellent corrosion resistance to meet the increasing needs of fuel burn4lp.
Further more, after the alloy composition is determined, the use of appropriate thermal processing can further improve the COrrosion resistance of the alloy. For zirconium alloys containing high content of Nb, including ZIRLO, MS and N36, etc., after the thermal processing temperature is increased, corrosion resistance will become weaker due to due to roughness or the second phase and uneven distribution as well as supersaturated solid solution Nb. Thus, "cold processing" is emphasized. (Mardon, iF, Charquet, D., and Senevat, 3., Influence in of composition and fabrication process on outoftpi1e and inpile properties of M5 alloy Zirconium in the Nuclear Industry:. T\velfih International Symposium, ASTM STP 1354, 2000, pp 505524), Low temperature process at low thermal extrusion temperature and annealing temperature can obtai.n dispersed fine second phase structures, greatly improving the corrosion and mechanical properties of is alloys, especially resistance to corrosion.
The main consideration in PWR is uniform corrosion of zirconium alloys. it is generally considered that those qualified in corrosion testing in outofreactor 360E water solution and 400 LI vapor can be used for pressurized water reactor and those qualified in outofireactor 360L1 lithiumcontaining solution are even more appli cable in the PWR conditions of high lithium concentration.
The present invention aims to provide a novel zirconium alloy with, good corrosion resistance for the nuclear power reactor core.
To achieve thi.s object, the present invention adopts the technical solution as follows: A zirconium alloy for nuclear power i.eactor core, comprisin.g the following components according to percentage by weight: O4OO,8Owt% of Sn, Q.7541Owt% of Nb, O,20-O,5Owt% of Fe+Cr, O,20-OlSwt% of Fe/(Nb+Fe), OM6-Ol5wt% of 0. less than O.OOSwt% of C, less than QOO6wt% of N and the rest of Zr, A zirconium alloy for nuclear power reactor core, comprising the following io components according to percentage by weight: O,4-O.Swt% of Sn, O75-4lOwt% of Nb, O.20M.5Owt% of Fe+Cr, ft2OO.35wt% of Fe/(Nb±Fe), OO6O. 15wt% of 0, ftOl-Olwt% of Cu, Bi or Ge, less than O.OO8wt% of C, less than O,006wt% of N and the rest of Zr..
A zirconium alloy for nuclear power reactor core, comprising the following components according to percentage by weight: O,4O.8wt% of Sn, O754.1Owt% of Nb, O?2OO5O4% of Fe+Cr, O2O-O.35wt% of Fe/( b±Fe), O.002M,O2wt% of Si or S. OA)6-OJSwt?/o of C), less than OMOSwt% of C, less than OOO5wt% of N and the rest of Zr.
A zirconium alloy fo:r nuclear power reactor core, comprising the following components according to percentage by weight: O4OM8Owt% of Sn, O.75-'l.lOwt% of Nb, O.2OO,5Owt% of Fe+Cr, O2OO.35wt% of Fe/(NbH-fe), OOl-OJwt% of Cu, Bi or Ge, O.002-O,O2wt% of Si or 5, OO&Ol5wt% of 0, less than O.OO8wt% of C, less than O.OO6wt% of N and the rest of Zr.
A zirconium alloy for nuclear power reactor core, comprising the following components according to percentage by weight: O.40-O.6Owt% of Sn, O.90-1..lOwt% of Nb, O.20-O.5Owt% of Fe+Cr, O.20-O.35wt% of FeI(Nb+Fe), s O.O1-O.lOwt% of Cu, Hi or Ge, O.002-O.O2Owt% of Si or 5, O.06-O.lSwt% of 0, less than O.OO8wt% of C, less than O.OO6wt% of N and the rest of Zr.
A zirconium alloy for nuclear power reactor core, comprising the following components according to percentage by weight: O.40-O.6Owt% of Sn, O.90-1.lOwt% of Nb, O.20-O.5Owt% of Fe+Cr, O.20-.O.35wt% of FeI(Nb+Fe), to O.O1-O.lwt% of Cu, Hi or Ge, O.Ol-O.O2wt°A of Si or S, O.06-O.l5wt% of 0, less than O.OOSwt% of C, less than O.OO6wt% of N and the rest of Zr.
A zirconium alloy for nuclear power reactor core, comprising the following components according to percentage by weight O.60-O.8Owt% of Sn, O.75.-1.OOwt% of Nb, O.20-O.5Owt% of Fe+Cr, O.20-O.35wt% of Fe/(Nb+Fe), O.O1-O.lOwt% of Cu, Hi or Ge, O.002-O.O2Owt% of Si or S. O.06-O.lSwt% of 0, less than O.OO8wt% of C, less than O.OO6wt% of N and the rest of Zr.
A zirconium alloy for nuclear power reactor core, comprising the following components according to percentage by weight: O.60-O.SOwt% of Sn, O.75-1.OOwt% of Nb, O.20-O.SOwt% of Fe-f Cr, O.20-O.35wt% of Fe/(Nb+Fe), O,O1-O.lowt% of Cu, Hi or Ge, O.005-O.OlSwt% of Si or 5, O.06-O.lSwt% of 0, less than O.OO8wt% of C, less than O.OO6wt% of N and the rest of Th A zirconium alloy for nuclear power, comprising the following components according to percentage by weight: ft7Owt?/o of Sn, L 00w1% of Nb, 0,3Owt% of Fe, OM5wt% of C, OMiwt% of Cu, Si, Hi or Ge, OAOwt% of 0, less than 0.OO8wt% of C, less than 0MO6wt% ofN and the rest of Zr.
A method for preparing said zirconium alloy for nuclear power reactor core, s comprising the following steps: (1) Mixing the various components of zirconium alloy in accordance with the formulation amount; (2) Melting in vacuum consumable electric arc furnace to produce alloy ingots; a (3) Forging the alloy ingot in 950'C 108000 f3 phase region into billet of the desired. shape; (4) Heating the billet in 1000°C hOOt phase region evenly and quenching; (5) Per:ibrming thermal processing on the quenched billet in 600t 650t a phase region; (6) Performing cold processing on the billet after thermal processing and performing intermediate annealing at 550 t 620°C; (7) Performing stress relief annealing or recrystallization annealing at 460°C 600°C to obtain said zirconium alloy material.
In the present invention., based on ZrSn.Nh alloy, other ingredients are added to improve the properties of the alloy and the appropriate component content is selected, Especially, the control of addition of Sn, Nb, Fe, Cr and Cu or Ri not only improves the corrosion resistance of the alloy but also improves the mechanical properties and anti-radiation properties of the alloy. The properties of the alloy provided in the present invention meet the requirements of high burnup of nuclear reactor for core structural materials. The prototype alloy increases the uniformity corrosion resistance in the pure water outside the reactor, particularly in lithium hydroxide solution. According to the test results of the embodiments, these alloys can be considered to show more excellent resistance to uniformity corrosion, high resistance to creep and fatigue as well as anti-irradiation growth property in case of application in the reactor.
Ic Embodiments The present invention is described in more details in the following embodiments.
For zirconium alloys for nuclear reactors, corrosion resistance of the alloy is the primary consideration. On this basis, production cost and workability should be considered at the time of selecting alloy elements. Therefore, it requires detailed study of the impact of each alloy element on corrosion resistance, mechanical properties and creep, alloy system and the amount of each alloy element. The zirconium alloys of the present invention have better uniformity and nodular corrosion resistance, higher resistance top and fatigue properties and anti-irradiation growth property as follows: (1)Zr Considering the neutron absorption factor, the present invention selects Zr as the basic elements, with consideration into neutron absorption of other alloy elements added to the basic zirconium.
(2) Sn Sn can stabilize the a-phase of zirconium, increase its strength and $ counteract the harmftul effects of nitrogen on corrosion. Small amount of Sn can not achieve the desired effect. Sn content of the present invention is 0.40-0..80.OOwt%, which can ensure the alloy to have excellent corrosion resistance and good mechanical properties. (3)Nb
Nb can stabilize n-phase of zirconium and has high strengthening effect on zirconium. Excessive amount of niobium is sensitive to heat treatment. The Nb content of the present invention is 0.75-1.10 wt%, which can ensure the alloy to show excellent corrosion resistance and good mechanical properties in pure water and lithium hydroxide solution.
is (4) Fe and Cr Fe and Cr can improve corrosion resistance and mechanical properties of the alloy but too much or too little amount of Fe and Cr will have adverse impact. The sum amount of Fe and Cr in the present invention is 0.20-0.50 wt%, which can ensure the alloy to show excellent corrosion resistance in pure water and lithium hydroxide solution. (5)Cu
Cu can improve corrosion resistance of the alloy but too much amount of Cu will have adverse impact. The amount of Cu in the present invention is less than 0.1 wt%, which can ensure the alloy to show excellent corrosion resistance in pure water and lithium hydroxide solution.
(6) Di Bi can improve corrosion resistance of the alloy but too much amount of El will have adverse impact. The amount of Di in the present invention is less than 0.lwt%, which can ensure the alloy to show excellent corrosion resistance in pure water and lithium hydroxide solution.
(7) Ge Ge can improve corrosion resistance of the alloy but too much amount of Ge will have adverse impact. The amount of Ge in the present invention is less than 0.1 wt%, which can ensure the alloy to show excellent corrosion resistance in pure water and lithium hydroxide solution. (8)Si
is Si can affect the uniform distribution of precipitates. Thus, too much amount of Si will have adverse impact. The amount of Si in the present invention is less than O.O2wt%, which can ensure the alloy to show excellent corrosion resistance in lithium hydroxide solution. (9) S
Addition of appropriate amount of S to the alloy can improve creep strength and corrosion resistance of the alloy. However, too much S has adverse effect.
The amount of S in the present invention is less than 0.O2wt%, which can ensure the alloy to show excellent corrosion resistance in high4emperature steam.
(10) 0 0 can. stabilize the apiiase of zirconium arid addition of 0 to the alloy can.
increase the yield strength. The amount of 0 in the present invention is 0.060, 15 wt%, which can ensure the alloy to show adequate mechanical properties and creep resistance. The increase of 0 greatly reduces control difficulty in the processing process.
(Ii) C C is an inevitable impurity element in the alloy, and high content of C will ic reduce the corrosion resistance of the alloy. The amount of C in the present invention is less than 0.008 wt%, which can ensure the alloy 1.0 show excellent corrosion resistance in high temperature water and steam, (12)N N is an inevitable impurity element in the alloy, and high content of N will reduce the corrosion resistance of the alloy. The amount of N: in the present invention is less than 0.006 wt%, which can ensure the alloy to show excellent corrosion resistance in high temperature water and steam.
A. zirconium alloy for nuclear power reactor core, comprising the following components according to percentage by weight: 0.400.80wt% of Sn, 0,75i,lOwt% of Nb, 0,204L50wt% of Fe+Cr, 0.200.35wt% of Fe/(Nh+Fe).
0.0&0lSwt% of 0, less than 0.OOSwt% of C, less than 0,006wt% of N and the rest of Zr.
A zirconium alloy for nuclear power reactor core, comprising the following components according to percentage by weight: O.4-O.8wt% of Sn, O.75-l.lOwt% of Nb, O.20-O.5Owt% of Fe+Cr, O.20-O.35wt% of Fe/(Nb+Fe), O.06-O.l5wt% of 0, O.Ol-O,lwt% of Cu, Bi or Ge, less than O.OO8wt% of C, less than O.OO6wt% of N s and the restofzr.
A zirconium alloy for nuclear power reactor core, comprising the following components according to percentage by weight: O.4-O.Swt% of Sn, 0.75-1 * lOwt% of Nb, O.20-O.SOwt% of Fe-f-Cr, O.20-O.35wt% of FeI(Nb+Fe), 0.002-O.O2wt% of Si or S. O,06-O.lSwt% of 0, less than O.OO8wt% of C, less than O.OO6wt% of N and the rest of Zr.
A zirconium alloy for nuclear power reactor core, comprising the following components according to percentage by weight: 0.40-O.SOwt% of Sn, O.75-1.lOwt% of Nb, O.20-O.5Owt% of Fe+Cr 0.20-O.35wt% of Fe/(Nb+Fe), O.Ol-O.lwt% of Cu, Bi or Ge, 0.002-0.O2wt% of Si or 5, O.06-O.l5wt% of 0, less IS than O.OOSwt% of C, less than O.OO6wt% of N and the rest of Zt A zirconium alloy for nuclear power reactor core, comprising the follpwing components according to percentage by weight: O.40-O.6Owt% of Sn, O.90-1.lOwt% of Nb, O.20-O3Owt% of Fe+Cr, O.20-O.35wt% of Fe/(Nb+Fe), O.O1-O.IOwt% of Cu, Si or Ge, 0.002-0.O2Owt% of Si or S, O.06-O.lSwt% of 0, less than O.OOSwt% of C, less than O.OO6wt% of N and the rest of Zr.
A zirconium alloy for nuclear power reactor core, comprising the following components according to percentage by weight: O.40-O.6Owt% of Sn, O.90-1.lOwt% of Nb, 0.20-0.SOwt% of Fe+Cr, 0.20-O.35wt% of Fe/(Nb+Fe), O.Ol-O.lwt% of Cu, Di or Ge, O.O1-0.O2wt% of Si or S. O.06-O.lSwt% of 0, less than O.OOSwt% of C, less than 0,006wt% of N and the rest of Zr.
A zirconium alloy for nuclear power reactor core, comprising the following components according to percentage by weight: O.60-O.8Owt% of Sn, 0.75-1.OOwt% of Nb, O.20-O.SOwt% of Fe+Cr, O.20-O.35wt% of Fe/(Nb+Fe), 0.Ol-0.lOwt% of Cti, Si or Ge, O.002-0.O2Owt% of Si or 5, O.06-O.lSwt% of 0, less than O.OOSwt% of C, less than O.OO6wt% of N and the rest of Zr.
A zirconium alloy for nuclear power reactor core, comprising the following components according to percentage by weight: O.60-0.SOwt% of Sn, 0.75-1.OOwt% of Nb, O.20-O.5Owt% of Fe+Cr, O.20-O.35wt°% of Fe/(Nb+Fe), O.01-O.lOwt% of Cu, Di or Ge, O.005-O.OlSwt% of Si or 5, O.06-0.l5wt% of 0, less than O.OOSwt% of C, less than 0.OO6wt% of N and the rest of Zr.
A zirconium alloy for nuclear power, comprising the following components is according to percentage by weight: O.7Owt% of Sn, i.OOwt% of Nb, O.3Owt% of Fe, 0.OSwt% of Cr, 0.Olwt% of Cu, Si, Di or Ge, 0.lOwt% of 0, less than 0.OOSwt% of C, less than O.OO6wt% of N and the rest of Zt The zirconium-based alloy for pressurized water nuclear reactor core structural materials provided in the present invention optimizes component ratio of Zr-Sn-Nb alloy and adds a little Cr, Di, Cu and other elements in order to improve the corrosion resistance of the alloy.
Table 1 shows the components of the alloy, wherein l4 and 15* are respectively components and corresponding experimental results for Zr-4 alloy and N36 alloy, and the contents refers to weight percentage of the components in the alloy.
Thble 1 Components of the Alloy of the Present Invention Alloy components (wt%) Alloy. -___ --________ Cu,Bt. Zrand No. Sn Nb Fe Cr Siors 0 C N or Ge bnpvxities 1 0.40 0.81 0.30 0.15 0.01 -0.15 0.007 0.005 Rest 2 0.80 1.10 0.31 0.05 0.07 0.10 0.006 0.005 Rest 3 0.52 1.02 0.35 0.10 --0.08 0.007 0.005 Rest - 4 0.48 0.80 026 0.22 0.10 0.005 0.06 0.006 0.005 Rest 0.60 1.00 0.29 0.21 ---0.09 0,006 0.004 Rest 6 0.68 -1.06 0.23 -0.08 o.oi 0.10 0.007 0.004 Rest 7 0.70 1.00 0.30 0.05 0.005 0.10 0.007 0.005 Rest 8 0.75 0.75 0.25 --0.07 o.oo2 -0.12 0.001 0.004 Rest 9 0.42 1.10 0.31 0.10 -0.015 0.09 0.007 0.005 Rest 0.60 1.00 0.25 0.20 -0.10 0.15 0.005 0.005 Rest 11 0.75 -0.76 0.28 0.08 0.09 0.015 0.10 0.006 0.004 Rest 12 0.5 0.70 0.33 0.05 ---0.09 0.007 0.005 Rest 13 0.70 1.00 0.30 0.05 -0.005 0.10 0.007 0.005 Rest 14* 1.27 -0.22 0.12 -----0.09 0.014 0.008 Rest 15* 1,05 1.10 0.32 ---0.09 0.014 0.008 Rest s A method for preparing said zirconium alloy for nuclear power reactor core, comprising the following steps: (I) Mixing the various components of zirconium alloy in accordance with the formulation amount; (2) Melting in vacuum consumable electric arc furnace to produce alloy ingots; (3) Forging the alloy ingot in 950t -1080°C phase region into billet of the desired shape; (4) Heating the billet in 1000°C 1100°C [ phase region evenly and quenching; (5) Performing themmi processing on the quenched billet in 600°C 650°C a phase region; (6) Performing cold processin.g on the billet after therma.l processing and.
perthrming intermediate annealing at 550°C 620°C; (7) Performing stress relief annealing or recrystallization annealing at 460°C 600°C to obtain said zirconium alloy material.
Microstructure composed of the equlaxed a-Zr grain and fine second phase particles evenly distributed prepared by the process described above can ensure good performance in harsh reactor core environment. The performance test results s of alloy prepared in the above method are shown in Table 2 3 and 4. The specific test conditions described in Table 2 are as follows: 360°C and 18.6MPa deionized water; the test conditions in Tahl.e 3 are as follows: 360t and 18,6MPa 70.xg/g lithium-containing aqueou.s solution (adding to the deionized water in the form of lithium hydroxide); the test conditions in Table 4 are as follows: 400°C and 103MPa deionized water vapor. The corrosion period is 300 days in 360°C and 400°C steam. respectively. The table shows the corrosion rate of each alloy (mg/dm2/d) in order l:o facilitate comparison of the relative properties of the alloy, and the relative corrosion rate is also given in the table. According to Table (2, 3 and 4), all the alloys show good corrosion resistance in 360'C pure water and lithium hydroxide solution as well as 400t steam.
Table 2 Corrosion rate of alloy of the present invention after corrosion for 300 daylin 360tdeionized water 36tit118.6MPa pure Alloy components (wt%) wa Alloy ----_____ -____ ________ __________ ____-Cu, Cormsion nO, Sior Zrand Sn Nb Fe Cr Blot 0 C N imt ____ ___ (mgld&Id) rate 1 0.40 0.81 030 0.15 0.01 0.15 0.007 0.005 Rest -0.18 1.00 2 0.80 1.10 031 0.05 0.07,... 0.10 0.006 0.005 Rest 0.33 LW -3 0.52 1.02 0.35 0.10 0.08 0.007 0.005 Rest 0.24 1.33 4 0.48 0.80 0.26 0.22 0.10 0,005 0.06 0.006 0.005 Rest -0.23 1.28 0.60 1.00 0.29 0.21 0.09 0.006 0.004 ReSt 0.26 1.44 6 0.68 1.06 0.23 0.08 0.01 0.10 0.007 0.004 Rest 0.29 1.61 7 0.70 1.00 0.30 0.05 0.005 0.10 0.007 0.005 Rest 0.32 1.78 8 0.75 0.75 0.25 0.07 0002 0.12 0.007 0.004 Rest -0.32 1.7$ 9 0.42 1.10 0.31 0.10 0,015 0.09 0.007 0.005 Itest oz 1.28 0.60 1.00 0.25 020 0.10 --0.15 0.005 0.005 Rest 0.31 1.72 II 0.75 0.76 0.28 0.08 0.09 0.015 0.10 0.006 0.004 Rest 032 1.7$ 12 0.5 0.70 0.33 0.05 0.09 0.007 0.005 Rest o.n i.n 13 0.70 ri 0.30 0.05 -. 0.007 0.005 Rest 0.23 1.28 14' 1.27 --0.22 0.11 -0.09 0.014 0.008 Rest on in 15' 1.05 1.10 0.32 --0.09 0,014 0.008 Rest 0.28 1.22 Table 3 Corrosion rate of alloy of the present invention after corrosion in 3600 70pgg lithium-con taking water for 300 days 360t11&6MPeJ70j.tgIg Alloy components (wt%) lithium-containing Alloy water No. Cu, Coninion Sior Zrand Sn Nb Fe Cr Bior 0 C N mit relative S impurities ____ ___ (ng/d&/d) rate 1 0.40 0.81 0.30 0.15 0.01 0.15 0,007 0.005 Rest 034 1.00 2 0.80 1.10 0.31 0.05 0.07 0.10 0.006 0.005 Rest 0.39 1.1$ 3 0.52 1.02 0.35 0.10 --0.08 0.007 0.005 Rest 032 0.94 4 0.41 0.80 0.26 0.22 0.10 0.005 0.06 0.006 0.005 0.35 1.03 0.60 1.00 0.29 0.21 -0.09 0.006 0.004 Rest -038 1.12 6 0.68 1.06 0.23 -0.08 0.01 0.10 0.007 0.004 Rest 0.37 1,09 7 0.70 1.00 0.30 0.05 0.005 0.10 0.007 0.005 0.40 l.t8 8 0.75 0.75 0.25 -0.07 0.002 0.12 0.007 0.004 Rest 0.37 1.09 9 0.42 1.10 0.31 ---0.015 0.09 0.007 0.005 Rest 0.32 0.94 0.60 1,00 0.2$ 0.20 0.10 0.1$ 0.00$ 0.00$ 0.36 1.06 II 0.75 0.76 0.28 0.08 0.09 0.015 0.10 0.006 0.004 Rest 0.38 1.12 12 0.5 0.70 0.33 0.05 ---0.09 0.007 0.005 Rest 0.32 0.94 13 0.70 1.00 0,30 0,05 ---0.005 0.10 0.007 0.005 Rest 0.32 0.94 14 1.27 -0.22 0.12 --0.09 0.014 0.008 Rest 4.52 13.29 15* 1.05 1.10 0.32 ----0.09 0.014 0.008 Rest 0.41 1.21 Table 4 Corrosion rate of alloy of the present invention after corrosion in 400 steam for 300 days Alley cemponents (wt%) 400t steam AHoy Lu, , Conosion No. Si CT Zr aicL Sn Nb Fe Cr Si or 0 C N, ,. rate relalive Ce (ngIdrn2/d) rate I 0,40 0,81 0,30 0.15 0.01 015 0.007 0.005 Rest 0,70 1.00 2 0.80 1.10 0.31 0.05 0.07 010 0,006 0,005 Rest 0.54 1.20 -.-,.,.,.,.... .,,,.......,......-...., 3 0.521 02 035 0.10 0.08 0.007 0.005 Rest 0.76 1,09 4 0.48 0.80 0.26 0,22 0.10 0.06 0.006 0.005 Rest 0,72 1.03 0.60 1.00 0.29 0.21..---0.09 0.006 0.004 Rest 0.74 1.06 6038 o.2 ioI flQ04 7 030 1.00 0,30 0.05 0,005 0.10 0,007 0005 Rest 0.78 LII 8 075 07 02S I H 0 11 0(2 012 Y'07 0004 Rest 076 9 0,42 1,10 0,31 0.015 0.09 0.007 0.005 Rest 0.71 LW 0.60 1.00 0,25 0.20 0.10 0,15 0,005 0.005 Rest 0.75 1.07 ii 0,75 0.76 0.28 0.08 0.09 0.015 010 0,006 0.004 Rest 0.3 0,43 12 0.5 0.70 0,33 0,05 0.09 0.007 0.00) 1est 0.74 1.06 13 0.70 1,00 0.30 0.05 0.005 0.10 0.007 0.005 Rest 0.72 1,03 .1 H: 103 L47j In summary, according to embodiments of the present invention, the alloy of the present invention shows very good corrosion resistance in above three s hydrochernical conditions, significantly belier than that of ZN4 alloy and N36(Zr-I.OSm. I.ONb-0.3Fe) alloy developed by China. The corrosion rate of the present alloy in 360 [1/18.6 MPa LiOH solution for 300 days is lower than that of N36 alloy by 21%; corrosion rate in 3600/18.6 MPa deionized water for 300 days is lower than that of N36 alloy by 35%; and corrosion rate in 4000/103 MPa superheated steam for 300 days is lower titan that of N36 alloy by 23% The present invention uses preferred component range of Sn, Nb, Fe, Cr, Cu s or Bi. The interaction of the alloy elements in this range and low temperature processing produces unexpected results, mainly including: 1) the alloy of the present invention shows very good corrosion resistance in above three hydrochemical conditions and is siguificantly better than that of N36 alloy and Zr-4 alloy; 2) The alloy of the present invention is treated by low-temperature processing to obtain fine and dispersed second phase so as to improve the mechanical properties of the alloy (such as creep and fatigue properties) and anti-irradiation growth property.

Claims (10)

  1. Claims 1. A zirconium alloy for nuclear power reactor core, comprising the following components according to percentage by weight: O.40-O.8Owt% of Sn, O.75-1..lOwt% of Nb, O.lO-O.5Owt% of Fe+Cr, O.20-O.35wt% of Fe/(Nb+Fe), s O.06-O.l5wt% of 0, less than O.OOSwt% of C, less than O.OOówt% of N and the rest of Zr.
  2. 2. A zirconium alloy for nuclear power reactor core, comprising the following components according to percentage by weight: O.40-O.SOwt% of Sn, O.75-l.lOwt% of Nb, O.1O-O.SOwt% of Fe+Cr, O.20-O.35wt% of FeI(Nb+Fe), O.06-O.]5wt% of 0, O.Ol-O.lOwt% of Cu, Ri or Ge, less than O.OOSwt% of C, less than O.OO6wt% of N and the rest of Zr.
  3. 3. A zirconium alloy for nuclear power reactor core, comprising the following components according to percentage by weight: O.40-O.SOwt% of Sn, O.75-1.lOwt% of Nb, O.lO-O.5Owt% of Fe+Cr, O.20-O.3Swt% of FeI(Nb+Fe), Is O.002-O.O2Owt% of Si or S, O.064.lSwt% of 0, less than O.OO8wt% of C, less than O.OO6wt% of N and the rest of Zr.
  4. 4. A zirconium alloy for nuclear power reactor core, comprising the following components according to percentage by weight: O.40-O.EOwt% of Sn, O.75-l.lOwt% of Nb, O.1O-O.5Owt% of Fe+Cr, O.20-O.35wt% of Fe/(Nb+Fe), O.O1-O.lOwt% of Cu, Bi or Ge, O.002-O.O2Owt% of Si or 5, O.06-O.l5wt% of 0, less than O.OOSwt% of C, less than O.OO6wt% of N and the rest of Zr.
  5. 5. The zirconium alloy for nuclear power reactor core according to claim 4, comprising the following components according to percentage by weight: O.40-0.60w1% of Sn, O.90-1,lOwt% of Nb, 0.10-0..SOwt% of Fe+Cr, 0.20-O.35wt% of Fe/(Nb+Fe), 0.Ol-O.lOwt% of Cu, Bi or Ge, 0.002-O.O2Owt% of Si or S. 0.06-0.l5wt% of 0, less than O.OOSwt% of C, less than O.OO6wt% of N s and the rest of Zr and impurities.
  6. 6. The zirconium alloy for nuclear power reactor core according to claim 5, comprising the following components according to percentage by weight: 0.40-0.6Owt% of Sn, 0,90-i.IOwt% of Nb, 0.10-0.5Owt% of Fe+Cr, 0.20-0.35wt% of Fe/(Nb+Fe), 0.01-O.lwt% of Cu, 81 or Ge, 0.01-0.O2wt% of Si or S. J0 0.06-0. lSwt% of 0, less than 0.OOSwt% of C, less than 0.OO6wt% of N and the rest of Zr and impurities.
  7. 7. The zirconium alloy for nuclear power reactor core according to claim 4, comprising the following components according to percentage by weight: O.60-0.8Owt% of Sn, 0.75-1.OOwt% of Nb, 0. lO-O.SOwt% of Fe+Cr, 0.20-0.35wt% of Fe/(Nb+Fe), 0.01-0.lOwt% of Cu, 131 or Ge, O.002-O.O2Owt% of Si or 0.06-0.1 5wt% of 0, less than 0.OO8wt% of C, less than 0.OO6wt% of N and the rest of Zr and impurities.
  8. 8. The zirconium alloy for nuclear power reactor core according to claim 7, comprising the following components according to percentage by weight: 0.60-0.SOwt% of Sn, 0.75-1.OOwt% of Nb, 0.10-O.5Owt% of Fe+Cr, 0.20-0.35wt% of FeI(Nb+Fe), 0.01-O.l0wt% of Cu, Bi or Ge, 0.O05-0.Ol5wt% of Si or S, 0.06-0.1 Swt% of 0, less than 0.OO8wt% of C, less than 0,006wt% of N and the rest of Zr and impurities.
  9. 9. A zirconium alloy for nuclear power reactor core, comprising the following components according to percentage by weight: O70wt% of Sn, LOOwt% of Nb, 0.3Owt% of Fe, 0.O5wt% of Cr, 0.Olwt% of Cu, Si, Bi or Ge, 0.iOwt% of 0. less than (LOOSwt% of C, less than ftOO6wt% ofN and the rest of Zr.
  10. 10. A method for preparing the zirconium alloy fo:r nuclear power reactor core according to any one of claims I to 9, comprising the following steps: (I) Mixing the various components of zirconium alloy in accordance with ic the formulation amount; (2) Melting in. vacuum consumable electric arc thmace to produce alloy ingots; (3) Forging the alloy ingot in 950t 1080t l phase region into billet of the desired shape; (4) Heating the billet in 1000°C 1100°C [3 phase region evenly and quenching; (5) Performing thermal processing on the quenched billet in 600°C 650°C a phase region; (6) Performing cold processing on the billet after them. .ai processing and performing intermediate annealing at 550°C 620°C; (7) Performing stress relief annealing or recrystallization annealing at 460°C 600°C to obtain, said zirconium alloy material.
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