US20060225815A1 - Zirconium alloy and components for the core of light water-cooled nuclear reactors - Google Patents
Zirconium alloy and components for the core of light water-cooled nuclear reactors Download PDFInfo
- Publication number
- US20060225815A1 US20060225815A1 US11/333,643 US33364306A US2006225815A1 US 20060225815 A1 US20060225815 A1 US 20060225815A1 US 33364306 A US33364306 A US 33364306A US 2006225815 A1 US2006225815 A1 US 2006225815A1
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- Prior art keywords
- core
- zirconium
- alloy
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- reactor
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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
-
- 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
-
- 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
Definitions
- the invention lies in the nuclear reactor technology field. More specifically, the invention relates to a zirconium alloy, or zirconium-based alloy, and to structural parts made from an alloy of this type for the core of light-water-cooled nuclear reactors, in particular of pressurized-water reactors. Structural parts of this type are in particular fuel cladding tubes, spacers and control rod guide tubes.
- zirconium which has a low neutron absorption, is used as base metal for the structural parts in reactor cores.
- reactor-pure zirconium sponge the composition of which is governed by standards.
- Zircaloy-2 for boiling-water reactors
- Zircaloy-4 for pressurized-water reactors
- other zirconium-based alloys for example those known from U.S. Pat. Nos. 5,940,464 and 4,938,920 (corresp. German published patent application DE 38 05 124 A1); U.S. Pat. Nos. 5,230,758 and 5,112,573 (cf. German DE 690 10 115 T2), and from international PCT publication WO 01/24193 A1, are nowadays generally used for the above-mentioned purpose.
- Binary Zr—Nb alloys are also used to a lesser extent.
- Reactor-pure zirconium (maximum contents in ppm): Al B Cd C Cl H Hf Fe O Si 75 0.5 0.5 250 1300 25 100 1500 1600 20
- a zirconium alloy comprising, in percent by mass:
- an amount of vanadium with a content of at least 0.07% is provided.
- a sum of the contents of Sn, Nb, Fe, and V is at most 1.3%.
- the novel alloy is composed of a matrix of reactor-pure zirconium together with 0.2 to 0.5% of Sn, 0.2 to 0.5% of Nb, 0.05 to 0.40% of Fe and 0 to 0.20% of V, with the carbon content being restricted to at most 120 ppm, and a range of from 80 to 120 ppm for Si and from 0.12 to 0.20% for O being maintained. It has been found that alloys of this type can be used to produce components, such as cladding tubes, spacers, guide tubes and further structural elements of a fuel assembly, for the core of light water reactors, in particular of pressurized water reactors.
- the components have an improved resistance to corrosion compared to components made from Zircaloy-4, while maintaining substantially the same production and the same heat treatment. This property is particularly pronounced if the sum of the alloying constituents Sn, Nb, Fe and V must drop if the total amount of Sn and Nb increases.
- Vanadium is an addition which is not absolutely imperative with a view to improving the corrosion properties. For example, it is possible to increase the corrosion resistance with a high burn-up using Sn contents of 0.4% to 0.5%. However, if some of the iron is replaced by V or if V is added to the alloy in small quantities (0.02 to 0.20%), the hydrogen pickup factor (HPUF) and therefore the formation of hydrides, which in addition to embrittling the material also cause material growth, is reduced.
- HPUF hydrogen pickup factor
- Nb to the alloy in an amount of up to 0.8%, preferably up to its solubility limit, i.e. up to 0.5%. If this limit is not significantly exceeded, there is no risk of uncontrolled phase transitions, which result at relatively high temperatures, e.g. when welding spacers or cladding tubes to their end stoppers, on account of the complicated phase diagrams of ZrNb. Consequently, it should not be necessary for the alloy according to the invention to be subjected to a further heat treatment following welding.
- the alloys are relatively insensitive to the effects of high heating surface stresses and local boiling processes at the interface with water.
- they have a low radiation-induced growth.
- a component for the core of a light water reactor in particular, a pressurized-water reactor, the is formed of the above-outlined zirconium-based alloy.
- the component is produced while maintaining a cumulative annealing parameter of (10-40) E-18 h.
- FIGURE of the drawing is a chart in which the thickness of a resultant oxide layer is plotted over burn-up.
- Remainder reactor-pure unalloyed zirconium in each, with permitted foreign substances or impurities.
- ingots of the alloys A to D are melted in vacuo in a number of melting steps and then forged in the ⁇ -range of the alloys below the melting point.
- the forgings are heated again to a temperature in the ⁇ -range and then quenched in a water bath with a cooling rate of at least 30 K/s.
- the forgings are then forged to form rods.
- the forged rods are machined and cut into pieces which are used to extrude tubes.
- an anneal is carried out after the extrusion.
- the tubes treated in this way are pilgered in a number of steps by cold-forming to form cladding tubes.
- an intermediate anneal is carried out in vacuo at temperatures of approximately 700° C., which brings about recovery and recrystalization.
- the final deformation, which leads to the definitive cross section of the cladding tube is followed by a final anneal at approximately 600° C.
- Control rod guide tubes are also produced by the same process.
- the forging is hot-rolled (once or in a number of steps with anneals between them) to form plates.
- the temperatures are selected in such a way that they are in the ⁇ -range of the alloys.
- the plate is cold-rolled in a number of steps to form a metal sheet of the desired thickness.
- a vacuum anneal is carried out, which can also take place as a continuous process and brings about complete recrystalization.
- spacers, guide tubes and fuel rods produced in this way are used in a pressurized-water reactor, these components have better corrosion properties, in particular after a prolonged operating period, compared to components made from conventional Zircaloy-4 with a low tin content (low tin Zirc-4), as can be established from empirical calculations.
- the results of these calculations are disclosed in the diagram of the FIGURE.
- the burn-up is plotted on the abscissa and the oxide layer thickness on the ordinate. It can be seen that the alloys according to the invention can remain in the reactor for approximately twice as long (6 cycles) as the conventional alloys (3 cycles) before they have to be replaced for corrosion reasons. As noted above, all percentages cited herein are in percent by mass.
Abstract
Description
- This is a continuing application, under 35 U.S.C. § 120, of copending international application PCT/EP2004/007822, filed Jul. 15, 2004, which designated the United States; this application also claims the priority, under 35 U.S.C. § 119, of German patent application No. 103 32 239.6, filed Jul. 16, 2003; the prior applications are herewith incorporated by reference in their entirety.
- The invention lies in the nuclear reactor technology field. More specifically, the invention relates to a zirconium alloy, or zirconium-based alloy, and to structural parts made from an alloy of this type for the core of light-water-cooled nuclear reactors, in particular of pressurized-water reactors. Structural parts of this type are in particular fuel cladding tubes, spacers and control rod guide tubes.
- For physical reasons, zirconium, which has a low neutron absorption, is used as base metal for the structural parts in reactor cores. On account of the separation of the neutron absorber hafnium, it is customary to use reactor-pure zirconium sponge, the composition of which is governed by standards.
- Zircaloy-2 (for boiling-water reactors) and Zircaloy-4 (for pressurized-water reactors) or other zirconium-based alloys, for example those known from U.S. Pat. Nos. 5,940,464 and 4,938,920 (corresp. German published patent application DE 38 05 124 A1); U.S. Pat. Nos. 5,230,758 and 5,112,573 (cf. German DE 690 10 115 T2), and from international PCT publication WO 01/24193 A1, are nowadays generally used for the above-mentioned purpose. Binary Zr—Nb alloys are also used to a lesser extent.
- The following table gives the compositions of zirconium sponge and the standardized alloys which have hitherto been customary in Western engineering. In this context, it should be mentioned that nowadays some of the permitted impurities can be set in a particularly controlled way or even, by using suitable additions, set to specific values. By way of example, on account of its hardening action on zirconium, oxygen was originally controlled only to levels corresponding to manufacturing requirements, but nowadays it is actually used deliberately as a hardening addition.
- Table
- Reactor-pure zirconium (maximum contents in ppm):
Al B Cd C Cl H Hf Fe O Si 75 0.5 0.5 250 1300 25 100 1500 1600 20 - Composition of Zircaloy and ZrNb alloys (in % by mass):
Other Sn Fe Cr Ni stipulations Zircaloy-2: 1.2-1.7 0.07-0.20 0.05-0.15 0.03-0.08 0.18-0.36 FeCrNi Zircaloy-4: 1.2-1.7 0.18-0.24 0.07-0.13 ≦0.007 0.28-0.37 FeCr Zr-2, ≦0.05 ≦0.150 ≦0.02 ≦0.007 2.40-2.80% 5% Nb: Nb - It is accordingly an object of the invention to provide a zirconium alloy, which is further improved relative to the above-mentioned materials and structural parts of the general type and which, in particular, provides for further improved zircaloy structural parts for light water reactor cores.
- With the foregoing and other objects in view there is provided, in accordance with the invention, a zirconium alloy, comprising, in percent by mass:
-
- Sn: 0.2-0.5%
- Nb: 0.2-0.5%
- Fe: 0.05-0.40%
- V: 0-0.20%
- O: 0.12-0.20%
- Si: 80-120 ppm
- C: ≦120 ppm
and a remainder of reactor-pure zirconium together with standard impurities.
- In accordance with an added feature of the invention, there is provided an amount of vanadium with a content of at least 0.07%.
- In accordance with an additional feature of the invention, a sum of the contents of Sn, Nb, Fe, and V is at most 1.3%.
- In other words, the novel alloy is composed of a matrix of reactor-pure zirconium together with 0.2 to 0.5% of Sn, 0.2 to 0.5% of Nb, 0.05 to 0.40% of Fe and 0 to 0.20% of V, with the carbon content being restricted to at most 120 ppm, and a range of from 80 to 120 ppm for Si and from 0.12 to 0.20% for O being maintained. It has been found that alloys of this type can be used to produce components, such as cladding tubes, spacers, guide tubes and further structural elements of a fuel assembly, for the core of light water reactors, in particular of pressurized water reactors. The components have an improved resistance to corrosion compared to components made from Zircaloy-4, while maintaining substantially the same production and the same heat treatment. This property is particularly pronounced if the sum of the alloying constituents Sn, Nb, Fe and V must drop if the total amount of Sn and Nb increases.
- Values higher than 0.5% of Sn, for example up to 0.75%, have an adverse effect on the corrosion resistance, increase the radiation-induced growth, while the mechanical properties are significantly improved, which means that the proposed value of at most 0.5% represents a good compromise. The minimum Sn content at which components with good mechanical properties can still be produced is 0.2%.
- Vanadium is an addition which is not absolutely imperative with a view to improving the corrosion properties. For example, it is possible to increase the corrosion resistance with a high burn-up using Sn contents of 0.4% to 0.5%. However, if some of the iron is replaced by V or if V is added to the alloy in small quantities (0.02 to 0.20%), the hydrogen pickup factor (HPUF) and therefore the formation of hydrides, which in addition to embrittling the material also cause material growth, is reduced.
- To achieve an optimum creep rupture strength and at the same time a yield strength with a high value, it is possible to add Nb to the alloy in an amount of up to 0.8%, preferably up to its solubility limit, i.e. up to 0.5%. If this limit is not significantly exceeded, there is no risk of uncontrolled phase transitions, which result at relatively high temperatures, e.g. when welding spacers or cladding tubes to their end stoppers, on account of the complicated phase diagrams of ZrNb. Consequently, it should not be necessary for the alloy according to the invention to be subjected to a further heat treatment following welding.
- Furthermore, the alloys are relatively insensitive to the effects of high heating surface stresses and local boiling processes at the interface with water. In this context, in particular a low uptake of lithium and a low level of nodular corrosion—as is found with cladding tubes made from Zircaloy-4 under standard pressurized-water conditions—are observed. Moreover, they have a low radiation-induced growth.
- With the above and other objects in view there is also provided, in accordance with the invention, a component for the core of a light water reactor, in particular, a pressurized-water reactor, the is formed of the above-outlined zirconium-based alloy.
- In a preferred embodiment of the invention, the component is produced while maintaining a cumulative annealing parameter of (10-40) E-18 h.
- Other features which are considered as characteristic for the invention are set forth in the appended claims.
- Although the invention is illustrated and described herein as embodied in a zirconium alloy and components for the core of light-water-cooled nuclear reactors, it is nevertheless not intended to be limited to the details shown, since various modifications and structural changes may be made therein without departing from the spirit of the invention and within the scope and range of equivalents of the claims.
- The construction and method of operation of the invention, however, together with additional objects and advantages thereof will be best understood from the following description of specific embodiments of the invention.
- The sole FIGURE of the drawing is a chart in which the thickness of a resultant oxide layer is plotted over burn-up.
- The following table illustrates four exemplary embodiments of the invention:
Sn(%) Nb(%) Fe(%) V(%) O(%) Si(ppm) C(ppm) A 0.30 0.25 0.35 0.16 0.14 110 100 B 0.30 0.45 0.15 0.10 0.14 110 100 C 0.40 0.45 0.10 0.07 0.14 110 100 D 0.30 0.75 0.13 0.07 0.14 110 100 - Remainder: reactor-pure unalloyed zirconium in each, with permitted foreign substances or impurities.
- To produce cladding tubes, ingots of the alloys A to D are melted in vacuo in a number of melting steps and then forged in the β-range of the alloys below the melting point. The forgings are heated again to a temperature in the β-range and then quenched in a water bath with a cooling rate of at least 30 K/s. The forgings are then forged to form rods.
- The forged rods are machined and cut into pieces which are used to extrude tubes. To obtain a fully recrystalized microstructure, an anneal is carried out after the extrusion. The tubes treated in this way are pilgered in a number of steps by cold-forming to form cladding tubes. Prior to each deformation operation, an intermediate anneal is carried out in vacuo at temperatures of approximately 700° C., which brings about recovery and recrystalization. The final deformation, which leads to the definitive cross section of the cladding tube, is followed by a final anneal at approximately 600° C. In this way, a low creep deformation with a high yield strength is set for the intended reactor use. A cumulative annealing parameter in the range A=10-40 E-18 h is maintained during production. It is in addition optionally possible to carry out an anneal in the alpha range following production of the forged rods.
- The cladding tubes which have been produced in the manner outlined are finally filled with fuel pellets and welded to end stoppers in a gastight manner at both ends. This concludes the production of the fuel rods. Control rod guide tubes are also produced by the same process.
- In another exemplary embodiment, following corresponding heating and quenching of an ingot of the same composition, the forging is hot-rolled (once or in a number of steps with anneals between them) to form plates. For the hot-forming and intermediate annealing steps, the temperatures are selected in such a way that they are in the α-range of the alloys. Then, the plate is cold-rolled in a number of steps to form a metal sheet of the desired thickness. Between the deformation steps and following the final deformation, a vacuum anneal is carried out, which can also take place as a continuous process and brings about complete recrystalization. These metal sheets are processed further to form spacers.
- If spacers, guide tubes and fuel rods produced in this way are used in a pressurized-water reactor, these components have better corrosion properties, in particular after a prolonged operating period, compared to components made from conventional Zircaloy-4 with a low tin content (low tin Zirc-4), as can be established from empirical calculations. The results of these calculations are disclosed in the diagram of the FIGURE. The burn-up is plotted on the abscissa and the oxide layer thickness on the ordinate. It can be seen that the alloys according to the invention can remain in the reactor for approximately twice as long (6 cycles) as the conventional alloys (3 cycles) before they have to be replaced for corrosion reasons. As noted above, all percentages cited herein are in percent by mass.
Claims (6)
Applications Claiming Priority (3)
Application Number | Priority Date | Filing Date | Title |
---|---|---|---|
DE10332239A DE10332239B3 (en) | 2003-07-16 | 2003-07-16 | Zirconium alloy and components for the core of light water cooled nuclear reactors |
DEDE10332239.6 | 2003-07-16 | ||
PCT/EP2004/007822 WO2005007908A2 (en) | 2003-07-16 | 2004-07-15 | Zirconium alloy and components for the core of light water cooled nuclear reactors |
Related Parent Applications (1)
Application Number | Title | Priority Date | Filing Date |
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PCT/EP2004/007822 Continuation WO2005007908A2 (en) | 2003-07-16 | 2004-07-15 | Zirconium alloy and components for the core of light water cooled nuclear reactors |
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US20060225815A1 true US20060225815A1 (en) | 2006-10-12 |
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US11/333,643 Abandoned US20060225815A1 (en) | 2003-07-16 | 2006-01-17 | Zirconium alloy and components for the core of light water-cooled nuclear reactors |
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US (1) | US20060225815A1 (en) |
EP (1) | EP1627090B1 (en) |
JP (1) | JP4417378B2 (en) |
KR (1) | KR100766202B1 (en) |
CN (1) | CN100372954C (en) |
AT (1) | ATE343655T1 (en) |
DE (2) | DE10332239B3 (en) |
ES (1) | ES2274482T3 (en) |
TW (1) | TWI315343B (en) |
WO (1) | WO2005007908A2 (en) |
ZA (1) | ZA200509729B (en) |
Cited By (3)
Publication number | Priority date | Publication date | Assignee | Title |
---|---|---|---|---|
US9284629B2 (en) | 2004-03-23 | 2016-03-15 | Westinghouse Electric Company Llc | Zirconium alloys with improved corrosion/creep resistance due to final heat treatments |
US10119181B2 (en) | 2013-01-11 | 2018-11-06 | Areva Np | Treatment process for a zirconium alloy, zirconium alloy resulting from this process and parts of nuclear reactors made of this alloy |
US10221475B2 (en) | 2004-03-23 | 2019-03-05 | Westinghouse Electric Company Llc | Zirconium alloys with improved corrosion/creep resistance |
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GT200500281A (en) | 2004-10-22 | 2006-04-24 | Novartis Ag | ORGANIC COMPOUNDS. |
GB0424284D0 (en) | 2004-11-02 | 2004-12-01 | Novartis Ag | Organic compounds |
JP5373599B2 (en) | 2006-04-21 | 2013-12-18 | ノバルティス アーゲー | Purine derivatives for use as adenosine A2A receptor agonists |
CN101270426B (en) * | 2008-03-24 | 2010-06-23 | 中国核动力研究设计院 | Zirconium based alloy for nuclear reactor |
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 |
CN102864338B (en) * | 2012-09-04 | 2014-06-18 | 上海核工程研究设计院 | Corrosion resistant zirconium-based alloy used for high burnup and preparation method thereof |
CN103898362B (en) * | 2012-12-27 | 2016-08-10 | 中国核动力研究设计院 | A kind of water cooled nuclear reactor zirconium-base alloy |
CN103898369A (en) * | 2012-12-27 | 2014-07-02 | 中国核动力研究设计院 | Zirconium alloy for nuclear reactor |
CN104745875A (en) * | 2013-12-30 | 2015-07-01 | 上海核工程研究设计院 | Zirconium alloy material for light water reactor under higher burnup |
CN113249616A (en) * | 2021-04-08 | 2021-08-13 | 岭澳核电有限公司 | Zirconium alloy for fuel assembly, preparation method thereof and cladding tube of fuel assembly |
CN113201665A (en) * | 2021-04-08 | 2021-08-03 | 中广核研究院有限公司 | Zirconium alloy for fuel assembly cladding, manufacturing method thereof and fuel assembly cladding tube |
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2003
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-
2004
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- 2004-07-15 AT AT04763227T patent/ATE343655T1/en not_active IP Right Cessation
- 2004-07-15 KR KR1020067000890A patent/KR100766202B1/en active IP Right Grant
- 2004-07-15 CN CNB2004800152598A patent/CN100372954C/en not_active Expired - Fee Related
- 2004-07-15 EP EP04763227A patent/EP1627090B1/en active Active
- 2004-07-15 TW TW093121130A patent/TWI315343B/en not_active IP Right Cessation
- 2004-07-15 WO PCT/EP2004/007822 patent/WO2005007908A2/en active IP Right Grant
- 2004-07-15 DE DE502004001861T patent/DE502004001861D1/en active Active
- 2004-07-15 JP JP2006519872A patent/JP4417378B2/en not_active Expired - Fee Related
-
2005
- 2005-12-01 ZA ZA200509729A patent/ZA200509729B/en unknown
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2006
- 2006-01-17 US US11/333,643 patent/US20060225815A1/en not_active Abandoned
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US9284629B2 (en) | 2004-03-23 | 2016-03-15 | Westinghouse Electric Company Llc | Zirconium alloys with improved corrosion/creep resistance due to final heat treatments |
US9725791B2 (en) | 2004-03-23 | 2017-08-08 | Westinghouse Electric Company Llc | Zirconium alloys with improved corrosion/creep resistance due to final heat treatments |
US10221475B2 (en) | 2004-03-23 | 2019-03-05 | Westinghouse Electric Company Llc | Zirconium alloys with improved corrosion/creep resistance |
US10119181B2 (en) | 2013-01-11 | 2018-11-06 | Areva Np | Treatment process for a zirconium alloy, zirconium alloy resulting from this process and parts of nuclear reactors made of this alloy |
Also Published As
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JP2009513821A (en) | 2009-04-02 |
KR100766202B1 (en) | 2007-10-10 |
ZA200509729B (en) | 2006-10-25 |
EP1627090B1 (en) | 2006-10-25 |
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DE502004001861D1 (en) | 2006-12-07 |
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WO2005007908A2 (en) | 2005-01-27 |
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TW200510550A (en) | 2005-03-16 |
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