KR100709389B1 - Process for the manufacture of zirconium?alloy cladding tube having excellent corrosion resistance for a nuclear fuel rod - Google Patents
Process for the manufacture of zirconium?alloy cladding tube having excellent corrosion resistance for a nuclear fuel rod Download PDFInfo
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- KR100709389B1 KR100709389B1 KR1020030012616A KR20030012616A KR100709389B1 KR 100709389 B1 KR100709389 B1 KR 100709389B1 KR 1020030012616 A KR1020030012616 A KR 1020030012616A KR 20030012616 A KR20030012616 A KR 20030012616A KR 100709389 B1 KR100709389 B1 KR 100709389B1
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- 238000005253 cladding Methods 0.000 title claims abstract description 51
- 238000005260 corrosion Methods 0.000 title claims abstract description 34
- 230000007797 corrosion Effects 0.000 title claims abstract description 34
- 239000000956 alloy Substances 0.000 title claims abstract description 33
- 238000004519 manufacturing process Methods 0.000 title claims abstract description 8
- 238000000034 method Methods 0.000 title claims description 6
- QCWXUUIWCKQGHC-UHFFFAOYSA-N Zirconium Chemical compound [Zr] QCWXUUIWCKQGHC-UHFFFAOYSA-N 0.000 title claims 2
- 229910052726 zirconium Inorganic materials 0.000 title claims 2
- 239000003758 nuclear fuel Substances 0.000 title abstract description 20
- 229910045601 alloy Inorganic materials 0.000 title description 14
- 230000008569 process Effects 0.000 title description 2
- 229910001093 Zr alloy Inorganic materials 0.000 claims abstract description 39
- 239000000446 fuel Substances 0.000 claims abstract description 17
- 238000010438 heat treatment Methods 0.000 claims description 13
- 239000000203 mixture Substances 0.000 claims description 8
- 230000001590 oxidative effect Effects 0.000 claims description 4
- RVTZCBVAJQQJTK-UHFFFAOYSA-N oxygen(2-);zirconium(4+) Chemical compound [O-2].[O-2].[Zr+4] RVTZCBVAJQQJTK-UHFFFAOYSA-N 0.000 abstract description 8
- 229910001928 zirconium oxide Inorganic materials 0.000 abstract description 8
- 230000001681 protective effect Effects 0.000 abstract description 4
- 239000011248 coating agent Substances 0.000 description 10
- 238000000576 coating method Methods 0.000 description 10
- 230000000694 effects Effects 0.000 description 8
- 230000003647 oxidation Effects 0.000 description 8
- 238000007254 oxidation reaction Methods 0.000 description 8
- 239000010955 niobium Substances 0.000 description 7
- XLYOFNOQVPJJNP-UHFFFAOYSA-N water Substances O XLYOFNOQVPJJNP-UHFFFAOYSA-N 0.000 description 6
- 229910052758 niobium Inorganic materials 0.000 description 5
- 239000000463 material Substances 0.000 description 4
- 238000012360 testing method Methods 0.000 description 4
- 229910052718 tin Inorganic materials 0.000 description 4
- 239000007789 gas Substances 0.000 description 3
- 230000006872 improvement Effects 0.000 description 3
- 230000001133 acceleration Effects 0.000 description 2
- 230000015572 biosynthetic process Effects 0.000 description 2
- 230000008859 change Effects 0.000 description 2
- 229910052804 chromium Inorganic materials 0.000 description 2
- 238000002485 combustion reaction Methods 0.000 description 2
- 239000002826 coolant Substances 0.000 description 2
- 239000011162 core material Substances 0.000 description 2
- 238000011161 development Methods 0.000 description 2
- 238000005516 engineering process Methods 0.000 description 2
- 230000004992 fission Effects 0.000 description 2
- 229910052742 iron Inorganic materials 0.000 description 2
- 239000008188 pellet Substances 0.000 description 2
- 229910052720 vanadium Inorganic materials 0.000 description 2
- 229910052684 Cerium Inorganic materials 0.000 description 1
- 238000010521 absorption reaction Methods 0.000 description 1
- 238000005275 alloying Methods 0.000 description 1
- QVGXLLKOCUKJST-UHFFFAOYSA-N atomic oxygen Chemical compound [O] QVGXLLKOCUKJST-UHFFFAOYSA-N 0.000 description 1
- 238000006243 chemical reaction Methods 0.000 description 1
- 238000005097 cold rolling Methods 0.000 description 1
- 238000013461 design Methods 0.000 description 1
- 230000001627 detrimental effect Effects 0.000 description 1
- 230000005674 electromagnetic induction Effects 0.000 description 1
- 230000008676 import Effects 0.000 description 1
- 239000011261 inert gas Substances 0.000 description 1
- 230000004807 localization Effects 0.000 description 1
- 229910052751 metal Inorganic materials 0.000 description 1
- 239000002184 metal Substances 0.000 description 1
- 239000011259 mixed solution Substances 0.000 description 1
- 229910052759 nickel Inorganic materials 0.000 description 1
- GUCVJGMIXFAOAE-UHFFFAOYSA-N niobium atom Chemical compound [Nb] GUCVJGMIXFAOAE-UHFFFAOYSA-N 0.000 description 1
- 239000001301 oxygen Substances 0.000 description 1
- 229910052760 oxygen Inorganic materials 0.000 description 1
- 230000010287 polarization Effects 0.000 description 1
- 238000011160 research Methods 0.000 description 1
- 238000005096 rolling process Methods 0.000 description 1
- 239000000243 solution Substances 0.000 description 1
- 229910052727 yttrium Inorganic materials 0.000 description 1
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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
-
- C—CHEMISTRY; METALLURGY
- C23—COATING METALLIC MATERIAL; COATING MATERIAL WITH METALLIC MATERIAL; CHEMICAL SURFACE TREATMENT; DIFFUSION TREATMENT OF METALLIC MATERIAL; COATING BY VACUUM EVAPORATION, BY SPUTTERING, BY ION IMPLANTATION OR BY CHEMICAL VAPOUR DEPOSITION, IN GENERAL; INHIBITING CORROSION OF METALLIC MATERIAL OR INCRUSTATION IN GENERAL
- C23C—COATING METALLIC MATERIAL; COATING MATERIAL WITH METALLIC MATERIAL; SURFACE TREATMENT OF METALLIC MATERIAL BY DIFFUSION INTO THE SURFACE, BY CHEMICAL CONVERSION OR SUBSTITUTION; COATING BY VACUUM EVAPORATION, BY SPUTTERING, BY ION IMPLANTATION OR BY CHEMICAL VAPOUR DEPOSITION, IN GENERAL
- C23C8/00—Solid state diffusion of only non-metal elements into metallic material surfaces; Chemical surface treatment of metallic material by reaction of the surface with a reactive gas, leaving reaction products of surface material in the coating, e.g. conversion coatings, passivation of metals
- C23C8/06—Solid state diffusion of only non-metal elements into metallic material surfaces; Chemical surface treatment of metallic material by reaction of the surface with a reactive gas, leaving reaction products of surface material in the coating, e.g. conversion coatings, passivation of metals using gases
- C23C8/08—Solid state diffusion of only non-metal elements into metallic material surfaces; Chemical surface treatment of metallic material by reaction of the surface with a reactive gas, leaving reaction products of surface material in the coating, e.g. conversion coatings, passivation of metals using gases only one element being applied
- C23C8/10—Oxidising
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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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- Chemical & Material Sciences (AREA)
- Metallurgy (AREA)
- Physics & Mathematics (AREA)
- Plasma & Fusion (AREA)
- High Energy & Nuclear Physics (AREA)
- General Engineering & Computer Science (AREA)
- Materials Engineering (AREA)
- Mechanical Engineering (AREA)
- Organic Chemistry (AREA)
- Chemical Kinetics & Catalysis (AREA)
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Abstract
본 발명은 핵반응기의 핵연료봉에 사용되는 지르코늄 합금재의 피복관(Nuclear Fuel Cladding)에 관한 것이다. 이 피복관은 핵연료를 수용하기 위한 신장된 중공의 금속성 관으로, 상기 금속성관은 지르코늄 합금재이며, 상기 관의 외측면에는 두께 0.5~3.7㎛의 지르코늄 산화피막이 형성된 것이다. 이 피복관의 제조방법 역시 제공된다. 피복관의 외측면에 고온에서 형성된 보호성 산화피막에 의해 내부식성이 개선된다. The present invention relates to a nuclear fuel cladding of a zirconium alloy material used for nuclear fuel rods of a nuclear reactor. The cladding tube is an elongated hollow metallic tube for accommodating nuclear fuel. The metallic tube is a zirconium alloy material, and a zirconium oxide film having a thickness of 0.5 to 3.7 μm is formed on the outer surface of the tube. A method for producing this cladding tube is also provided. Corrosion resistance is improved by a protective oxide film formed at a high temperature on the outer surface of the cladding tube.
핵연료봉, 피복관, 지르코늄 합금, 지르코늄 산화피막, 내부식성Fuel rods, cladding, zirconium alloy, zirconium oxide, corrosion resistance
Description
도 1은 핵연료봉의 일례도, 1 is an example of a nuclear fuel rod,
도 2는 본 발명의 핵연료 피복관의 단면도,2 is a cross-sectional view of a nuclear fuel cladding tube of the present invention;
도 3은 지르코늄 합금재의 핵연료 피복관을 360℃의 물 분위기에서 부식 실험한 그래프이다.3 is a graph of corrosion test of a nuclear fuel cladding tube made of a zirconium alloy material in a water atmosphere at 360 ° C.
* 도면의 주요부분에 대한 부호의 설명 *Explanation of symbols on the main parts of the drawings
10....핵연료 봉 20.... 핵연료 펠릿 10 ....
30.... 피복관 30 .... cladding
31....중공의 지르코늄 합금재 관 32..지르코늄 산화피막 31..Zirconium alloy tube of hollow 32..Zirconium oxide film
40.... 스프링 50....플러그 40 ...
본 발명은 핵반응기의 핵연료봉에 사용되는 지르코늄 합금재의 피복관(Nuclear Fuel Cladding)에 관한 것으로, 보다 상세하게는 지르코늄 합금재질의 관 외측면에 산화피막(ZrO2)이 형성된 내부식성이 우수한 피복관과 그 제조방법에 관한 것이다. The present invention relates to a cladding tube (Nuclear Fuel Cladding) of the zirconium alloy material used for the nuclear fuel rods of the nuclear reactor, and more particularly, a cladding tube having excellent corrosion resistance formed of an oxide film (ZrO 2 ) on the outer surface of the zirconium alloy material; It relates to a manufacturing method.
원자로의 핵연료 집합체는 다수개의 핵연료봉으로 구성된다. 도 1에는 이러한 핵연료봉이 나타나 있다. 핵연료봉(10)은 핵연료 펠릿(20)을 수용하기 위한 신장된 중공의 피복관(30)과 플러그(50)을 구비하고 있다. 핵연료봉의 피복관은 핵분열 생성물이 냉각제로 방출되는 것을 방지하고, 연료와 냉각제간의 접촉 및 화학반응을 방지하는 기능을 한다. 피복관은 345℃에 달하는 가압수중에서의 내식성이 필수적이다. 또한, 핵분열 연료물질을 경제적으로 사용하기 위하여 중성자 흡수단면적이 작은 것이 좋다. The nuclear fuel assembly in a reactor consists of a number of fuel rods. 1 shows such a fuel rod. The
지르코늄 합금이 1960년 개발된 후 현재까지 피복관 재료로 사용되어 오고 있다. 하지만 일반 경수로 (Light Water Reactor)에서 운전조건이 고pH 운전, 고연소도 운전 등의 가혹한 조건으로 바뀌어가면서 기존의 핵연료 피복관인 Zircaloy-4 합금의 사용은 한계점에 도달했다고 할 수 있다. Zirconium alloys were developed in 1960 and have been used as cladding materials to date. However, as the operating conditions in light water reactors are changed to harsh conditions such as high pH operation and high combustion operation, the use of the conventional fuel cladding Zircaloy-4 alloy has reached its limit.
고pH 운전, 고연소도 운전은 피복관의 부식을 가속시켜 장주기 운전시 피복관에 치명적 영향을 끼치므로 부식 저항성을 향상시키기 위한 기술개발이 원전 선 진국을 중심으로 대대적으로 이루어지고 있는 상황이다. High-pH operation and high-combustion operation accelerate the corrosion of the cladding pipe and have a fatal effect on the cladding pipe during long cycle operation. Therefore, the development of technology to improve the corrosion resistance is being carried out mainly in advanced countries of nuclear power plants.
미국의 Westinghouse사는 기존의 Zircaloy-4 합금에 Nb을 첨가하여 ZIRLO(Zr-1Nb-1Sn-0.1Fe) 합금을 개발하였으며, 이 합금은 내식성이 매우 우수한 것으로 보고된 바 있다. 독일의 KWU에서 개발한 Duplex피복관은 피복관 내면에는 기존의 Zircaloy-4 합금을 사용하고 외부측(두께의 10%)에는 내식성을 향상시킬 수 있는 다른 합금을 사용한 것으로 현재 상용 원자로에서 시험 연소 중이다. 일본에서는 Mitsubishi사를 중심으로 Sn을 줄이고 Nb을 약간 첨가한 VAZ(Zr-0.5Sn-0.1Nb-0.2Fe-0.1Cr)합금이 개발가능성이 있는 것으로 평가되어 이 합금에 대하여 연구가 진행 중이다. 프랑스에서는 Sn과 Nb을 제거하고 V를 첨가한 Zr-0.25Fe-0.25V 합금에 대해 연구를 하고 있다. 러시아에서는 Zr-1Nb합금을 핵연료 피복관으로 사용하여 왔으나 최근에는 Nb와 Sn이 혼합된 Zr-1Sn-1Nb-0.5Fe합금을 개발하여 현재 상용로에 사용 여부를 검토 중에 있다. 하지만 이들 합금은 기존의 Zircaloy-4에 비해 특정한 성능면에서는 개선되었으나, 피복관에서 요구되는 모든 조건을 충족하지 못하고 있는 실정이다. 특히, 현재 개발중인 지르코늄 합금들은 첨가원소의 중성자효과, 가격, 내식성, 유해함 등을 고려하여 Sn, Fe, Cr, V, Nb등의 합금원소를 조절하는 기술이 대부분이다. Westinghouse of the United States has developed a ZIRLO (Zr-1Nb-1Sn-0.1Fe) alloy by adding Nb to an existing Zircaloy-4 alloy, which has been reported to be very corrosion resistant. Duplex sheath, developed by Germany's KWU, uses a conventional Zircaloy-4 alloy on the inside of the cladding and another alloy on the outside (10% of the thickness) to improve corrosion resistance. In Japan, VAZ (Zr-0.5Sn-0.1Nb-0.2Fe-0.1Cr) alloys with reduced Sn and slightly added Nb have been evaluated as a possibility of development. In France, Zr-0.25Fe-0.25V alloys with Sn and Nb removed and V added are being studied. In Russia, Zr-1Nb alloys have been used as fuel cladding tubes, but recently, Zr-1Sn-1Nb-0.5Fe alloys containing Nb and Sn have been developed and are currently being considered for commercial use. However, these alloys have been improved in specific performance compared to the existing Zircaloy-4, but do not meet all the requirements of the cladding. In particular, most of the zirconium alloys currently being developed are technologies for controlling alloying elements such as Sn, Fe, Cr, V, and Nb in consideration of neutron effect, price, corrosion resistance, and harmfulness of added elements.
이외에도 대한민국 공개특허공보 2001-0047592호(니오븀이 첨가된 핵연료피복관용 지르코늄 합금의 조성물), 2000-0056306호(핵연료 피복관용 지르코늄 합금조성물 및 제조방법), 2000-0026542호(내부식성과 기계적 특성이 우수한 지르코늄 합금 조성물), 1999-0069103호(우수한 부식 저항성과 고강도를 갖는 지르코늄 합금 조성물), 1999-0069104호(저 부식성과 고강도를 갖는 지르코늄 합금 조성물)에는 핵연료봉 피복관의 지르코늄 합금 조성물이 개시되어 있다. In addition, Korean Unexamined Patent Publication No. 2001-0047592 (composition of a zirconium alloy containing niobium for nuclear fuel cladding), 2000-0056306 (composition of zirconium alloy composition for nuclear fuel cladding and manufacturing method), 2000-0026542 (corrosion resistance and mechanical properties Excellent Zirconium Alloy Compositions), 1999-0069103 (zirconium alloy compositions having excellent corrosion resistance and high strength), and 1999-0069104 (zirconium alloy compositions having low corrosion and high strength) disclose zirconium alloy compositions of nuclear fuel rod cladding. .
이와 같이, 지금까지 핵연료봉 피복관의 성능향상은 대부분 지르코늄 합금 설계에 국한되고 있으며, 부식가속에 대한 근본적인 해결책을 제시하지 못하고 있다. 따라서, 원자력 발전소의 운전 조건 변화 추세를 감안한 부식가속문제를 해결할 수 있는 새로운 피복관의 개발이 시급한 실정이다. As such, the performance improvement of the nuclear fuel rod cladding tube is limited to the design of zirconium alloy, and thus, it cannot provide a fundamental solution to corrosion acceleration. Therefore, it is urgent to develop a new cladding tube that can solve the corrosion acceleration problem in consideration of the change in the operating conditions of the nuclear power plant.
본 발명에서는 상기 목적을 달성하기 위한 연구과정에서 안출된 것으로, 지르코늄 합금재의 피복관 외측면에 형성된 보호성 지르코늄 산화피막에 의해 내부식성이 개선된 피복관과 그 제조방법을 제공하는데 그 목적이 있다.
The present invention has been made in the course of research to achieve the above object, it is an object of the present invention to provide a coating tube and a method of manufacturing the corrosion resistance improved by a protective zirconium oxide film formed on the outer surface of the coating tube of zirconium alloy material.
상기 목적을 달성하기 위한 본 발명의 피복관은, The cladding tube of the present invention for achieving the above object,
핵연료를 수용하기 위한 신장된 중공의 금속성 관으로, 상기 금속성 관은 지르코늄 합금재이며, 상기 지르코늄 합금재관의 외측면에는 두께 0.5~3.7㎛의 지르코늄 산화피막이 형성되는 것을 포함하여 구성된다. An elongated hollow metallic tube for accommodating nuclear fuel, wherein the metallic tube is a zirconium alloy material, and a zirconium oxide film having a thickness of 0.5 to 3.7 μm is formed on an outer surface of the zirconium alloy material tube.
또한, 본 발명의 피복관 제조방법은, 지르코늄 합금재의 중공의 튜브형인 핵연료 피복관을 산화성 분위기의 650~1100℃에서 열처리하여 피복관의 외측면에 두께 0.5~3.7㎛의 산화피막을 형성하는 것이다. The method for producing a cladding tube according to the present invention is to heat-treat a hollow tubular nuclear fuel cladding tube made of a zirconium alloy material at 650 to 1100 ° C. in an oxidizing atmosphere to form an oxide film having a thickness of 0.5 to 3.7 μm on the outer surface of the cladding tube.
이하, 본 발명을 상세히 설명한다. Hereinafter, the present invention will be described in detail.
지르코늄 합금은 활성이 강한 금속이므로 공기 또는 물의 존재하에서 표면에 산화피막이 형성된다. 따라서, 원자로내 약 350℃ 가압수의 환경에서 지르코늄 합금의 피복관에는 자연적으로 산화피막이 형성된다. 그러나, 원자로내에서 피복관의 사용중에 피복관의 외측면에 형성된 산화피막은 내식성이 떨어지고 부식진행이 현저하게 되면서 산화피막의 박리가 일어난다. 따라서, 피복관의 내부식성 측면에서 그 외측면에서의 산화피막의 형성은 긍정적으로 보고 있지 않다. Since zirconium alloy is a highly active metal, an oxide film is formed on the surface in the presence of air or water. Therefore, an oxide film is naturally formed in the coating tube of the zirconium alloy in an environment of about 350 ° C. pressurized water in the reactor. However, during use of the cladding tube in the reactor, the oxide film formed on the outer surface of the cladding tube is inferior in corrosion resistance and corrosion progresses, and peeling of the oxide film occurs. Therefore, in view of the corrosion resistance of the cladding tube, the formation of an oxide film on its outer surface is not viewed positively.
그런데, 본 발명자들은 피복관의 사용환경 보다 높은 고온의 열처리과정에서 지르코늄 합금의 표면에 형성되는 산화피막은 내부식성에 유리하게 작용한다는 사실에 주목하고, 본 발명을 완성하게 이른 것이다. 본 발명의 피복관은 그 외측면에 고온에서 의도적으로 형성된 산화피막을 갖는데, 특징이 있다. However, the present inventors have noted that the oxide film formed on the surface of the zirconium alloy in the heat treatment process at a higher temperature than the use environment of the cladding tube advantageously works for corrosion resistance, and has completed the present invention. The cladding tube of the present invention is characterized by having an oxide film intentionally formed at a high temperature on its outer surface.
도 2에는 외측면에 지르코늄 산화피막(32)이 형성된 피복관(30)이 제시되어 있다. 2 shows a
본 발명의 대상이 되는 피복관은 핵연료를 수용하기 위한 신장된 중공의 금속성 관으로, 상기 금속성 관은 지르코늄 합금재이다. 지르코늄 합금재는 Zr을 주성분으로 하고, 여기에 필요에 따라 Nb, Sn, Fe, Cr, V, Y, Ce, Ni 등의 첨가원소를 포함한다. 이러한 지르코늄 합금은 널리 알려져 있으며, 본 발명에서 지르코늄 합금을 특별히 제한하지 않는다. 본 발명에서는 원자력 발전소의 노심재료로 주로 이용되는 Zircally-4(Sn:1.20~1.70중량%, Fe:0.18~0.24중량%, Cr:0.07~1.13중량%, O:900~1500ppm, 나머지 Zr)의 피복관에 산화피막을 적용하는 것이 가장 바람직하다. The cladding tube of the present invention is an elongated hollow metallic tube for accommodating nuclear fuel, and the metallic tube is a zirconium alloy material. The zirconium alloy material contains Zr as a main component, and includes additional elements such as Nb, Sn, Fe, Cr, V, Y, Ce, and Ni as necessary. Such zirconium alloys are well known and do not particularly limit the zirconium alloy in the present invention. In the present invention, Zircally-4 (Sn: 1.20 ~ 1.70% by weight, Fe: 0.18 ~ 0.24% by weight, Cr: 0.07 ~ 1.13% by weight, O: 900 ~ 1500ppm, the remaining Zr) mainly used as the core material of nuclear power plants It is most preferable to apply an oxide film to the coating tube.
본 발명의 피복관은 지르코늄 합금재관의 외측면의 전체에 걸쳐 두께 0.5~3.7㎛의 지르코늄 산화피막이 형성되는 것이 바람직하다. 산화피막의 두께가 0.5㎛미만이거나 3.7㎛ 초과의 경우에는 내부식성이 떨어진다. 산화피막의 두께가 두꺼워지면 오히려 내부식성에 불리하다. 산화피막의 두께가 0.7~1.9㎛일때 내부식성이 가장 우수하다. 본 발명의 피복관은 핵연료봉으로서 원자로의 노심에 들어가기전 그 외측면 전체에 걸쳐 산화피막이 형성된 것이다. 이러한 적정 두께의 산화피막은 핵연료봉의 가혹한 부식환경에서 피복관을 보호한다. In the coated pipe of the present invention, it is preferable that a zirconium oxide film having a thickness of 0.5 to 3.7 µm is formed over the entire outer surface of the zirconium alloy material pipe. If the thickness of the oxide film is less than 0.5 µm or more than 3.7 µm, the corrosion resistance is poor. If the thickness of the oxide film is thick, it is rather detrimental to corrosion resistance. The corrosion resistance is the best when the thickness of the oxide film is 0.7 ~ 1.9㎛. In the cladding tube of the present invention, an oxide film is formed over the entire outer surface of the reactor as a fuel rod before entering the core of the reactor. This proper thickness of oxide film protects the cladding tube from the harsh corrosive environment of the fuel rods.
본 발명에 따라 지르코늄 합금재의 피복관에 산화피막을 형성하는 방법은 다음과 같다. According to the present invention, a method of forming an oxide film on a coating tube of a zirconium alloy material is as follows.
지르코늄 합금재의 핵연료 피복관을 산화성 분위기의 고온에서 열처리한다. 가열은 피복관을 적정한 고온으로 유지할 수 있는 것이면 가능하며, 예를 들어 로에 의한 가열, 전자유도가열, 전기저항가열 등이 있다. 산화성 분위기는 산소함유 혼합가스로서, Ar 등의 불활성가스를 혼합가스로 사용하는 것이 바람직하다. 열처리시간, 분위기, 온도 등의 조건은 피복관의 외측면에서 보호성 지르코늄 산화피막을 형성하도록 적절히 선택된다. The fuel cladding tube of the zirconium alloy material is heat-treated at high temperature in an oxidizing atmosphere. The heating can be carried out as long as the coating tube can be maintained at an appropriate high temperature. Examples thereof include heating by a furnace, electromagnetic induction heating, and electrical resistance heating. The oxidative atmosphere is an oxygen-containing mixed gas, and it is preferable to use an inert gas such as Ar as the mixed gas. Conditions such as heat treatment time, atmosphere, temperature, and the like are appropriately selected to form a protective zirconium oxide film on the outer surface of the coating tube.
본 발명에서 열처리온도는 500~1100℃가 바람직하다. 열처리 온도가 1100℃초과의 온도에서는 3.7 ㎛ 두께의 산화막을 얻는데 걸리는 시간이 약 30초 정도로 짧다. 따라서, 3.7㎛ 두께 이하의 얇은 산화막을 얻기 위해 30초이내의 시간동안 열처리를 해야하는데, 이는 장비와 열처리 기술상 많은 제약이 있어 제조기술적으로 바람직하지 않다. 또한, 500℃ 이상에서 산화피막을 형성할 때 내부식성 효과가 보여지는데, 내부식성 개선효과가 확연해지는 650℃ 이상의 온도에서 산화피막을 형성하는 것이 바람직하다. 가장 바람직하게는 880~1100℃의 온도에서 열처리하는 것이다. In the present invention, the heat treatment temperature is preferably 500 ~ 1100 ℃. When the heat treatment temperature is higher than 1100 占 폚, the time taken to obtain an oxide film having a thickness of 3.7 µm is as short as about 30 seconds. Therefore, in order to obtain a thin oxide film having a thickness of 3.7 μm or less, the heat treatment must be performed for a time of less than 30 seconds. In addition, the corrosion resistance effect is seen when forming the oxide film at 500 ℃ or more, it is preferable to form the oxide film at a temperature of 650 ℃ or more that the effect of improving the corrosion resistance is evident. Most preferably, the heat treatment at a temperature of 880 ~ 1100 ℃.
이하, 본 발명을 실시예를 통하여 보다 구체적으로 설명한다. Hereinafter, the present invention will be described in more detail with reference to Examples.
[실시예]EXAMPLE
관상로에서 Zircaloy-4의 피복관 외측면에 700-1100℃의 온도범위에서 산화 피막을 형성시켰다. 산화피막의 두께는 각 열처리 온도에서 Ar/O2의 혼합기체를 이용하여 산화시간을 변화하는 방법으로 조절하였다. 표 1은 산화시간에 따른 산화막 두께의 변화를 보여주고 있다.An oxide film was formed on the outer surface of the coated tube of Zircaloy-4 in the tubular furnace at a temperature range of 700-1100 ° C. The thickness of the oxide film was controlled by varying the oxidation time using a mixed gas of Ar / O 2 at each heat treatment temperature. Table 1 shows the change in oxide film thickness with time of oxidation.
표 1의 조건으로 산화피막을 형성한 피복관에 대해 조직관찰과 산화시험을 행하였다. The coated tube on which the oxide film was formed under the conditions shown in Table 1 was subjected to a tissue observation and an oxidation test.
(1) 조직관찰(1) organizational observation
최종 열처리 조직은 재료의 특성에 직접적인 영향을 줄 수 있기 때문에 미세조직 관찰은 냉간압연 후 700 ℃에서 1시간 열처리한 시편에 대하여 HF(10%)+(45%)+(45%)의 혼합 용액으로 부식시킨 후 편광 현미경으로 관찰하였다. 합금의 최종조직은 재결정된 조직을 나타내었고 압연으로 인한 연신조직은 관찰되지 않았다. 이러한 재결정 조직은 연신 조직보다 합금의 내부식성 향상에 기여하는 것으로 알려져 있다.The final heat treatment can directly affect the properties of the material, so microscopic observations are performed with a mixed solution of HF (10%) + (45%) + (45%) for specimens heat treated at 700 ° C for 1 hour after cold rolling. Corrosion was observed and observed with a polarization microscope. The final structure of the alloy showed a recrystallized structure and no stretched structure due to rolling was observed. Such recrystallized structures are known to contribute to the improvement of corrosion resistance of alloys over stretched structures.
(2) 산화 시험(2) oxidation test
산화시험은 ASTM G-2방법에 따라 360 ℃, 2660 psi의 고온, 고압의 물 분위기에서 실험을 행하였다. 산화에 따른 무게 증가량을 20일까지 측정하여 도 2에 나타내었다. Oxidation test was carried out in a high temperature, high pressure water atmosphere of 360 ℃, 2660 psi in accordance with ASTM G-2 method. The weight increase due to oxidation was measured up to 20 days and is shown in FIG. 2.
도 3에 나타난 바와 같이, 본 발명에 따라 고온에서 산화피막을 형성한 지르코늄 합금재의 피복관은 종래의 Zircaloy-4에 비하여 낮은 무게 증가량을 보여 향상된 산화 저항성을 나타내었다. 산화피막은 0.5~3.7㎛의 범위에서 내부식성의 개선효과가 나타났다. 1.0~2.3㎛의 범위에서 대체적으로 내부식성 증진효과가 좋았다. 이 결과를 볼 때, 최적의 산화피막 두께는 0.7~1.9㎛로 판단되며, 가장 최적의 산화피막 두께는 1.0~1.3㎛이다. As shown in FIG. 3, the coated tube of the zirconium alloy material in which the oxide film was formed at a high temperature according to the present invention showed a lower weight increase compared to the conventional Zircaloy-4, thereby showing improved oxidation resistance. The oxide film showed the improvement of corrosion resistance in the range of 0.5 ~ 3.7㎛. Corrosion resistance enhancement effect was good in the range of 1.0 ~ 2.3㎛. In view of these results, the optimum oxide thickness is judged to be 0.7-1.9 탆, and the most optimal oxide thickness is 1.0-1.3 탆.
본 발명은, 상기 실시형태에 한정되는 것은 아니며, 상기 실시형태는 하나의 예시이다. 본 발명의 특허청구 범위에 기재된 기술적 사상과 실질적으로 동일한 구성을 갖고, 유사한 작용효과를 제공하는 것은, 어느 것이라도 본 발명의 기술적 범위에 포함된다. 예를 들면, 상기 실시형태에서는 Zircaloy-4의 피복관을 이용하지만, 본 발명이 반드시 이와 같은 지르코늄 합금재일 필요는 없고, 피복관에 적용되는 지르코늄 합금이면 어떠한 것이든 사용할 수 있다. This invention is not limited to the said embodiment, The said embodiment is an illustration. Anything having a configuration substantially the same as the technical idea described in the claims of the present invention and providing a similar effect is included in the technical scope of the present invention. For example, in the said embodiment, although the coating tube of Zircaloy-4 is used, this invention does not necessarily need to be such a zirconium alloy material, Any thing can be used as long as it is a zirconium alloy applied to a coating tube.
상술한 바와 같이, 본 발명에 따라 보호성 산화피막을 갖는 지르코늄 합금재의 피복관은 가혹한 부식 환경에서도 핵연료봉의 내부식성을 증진시킬 수 있는 유용한 효과가 있는 것이다. 이러한 핵연료봉은 국산화에 기여하여 피복관재인 Zircaloy-4합금을 대체할 경우 연 100억 이상의 수입대체 효과가 예상된다. As described above, the coating tube of the zirconium alloy material having a protective oxide film according to the present invention has a useful effect of improving the corrosion resistance of the nuclear fuel rod even in the harsh corrosive environment. Such fuel rods will contribute to the localization, and the replacement of the Zircaloy-4 alloy, which is a cladding material, is expected to have an annual import replacement effect of more than 10 billion won.
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| KR1020030012616A KR100709389B1 (en) | 2003-02-28 | 2003-02-28 | Process for the manufacture of zirconium?alloy cladding tube having excellent corrosion resistance for a nuclear fuel rod |
| PCT/KR2004/000223 WO2004077452A1 (en) | 2003-02-28 | 2004-02-06 | Zirconim alloy nuclear fuel cladding tube having excellent corrosion resistance and process for the manufacture of the cladding tube |
| JP2005518751A JP4389252B2 (en) | 2003-02-28 | 2004-02-06 | A method for producing a zirconium alloy nuclear fuel cladding having excellent corrosion resistance. |
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| KR102049430B1 (en) * | 2017-03-29 | 2019-11-27 | 울산과학기술원 | Nuclear fuel cladding tube and manufacturing method of the same |
| CN112899682A (en) * | 2021-01-20 | 2021-06-04 | 哈尔滨工业大学 | Preparation method of zirconium cladding protective coating |
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|---|---|---|---|---|
| KR890002898A (en) * | 1987-07-21 | 1989-04-11 | 쉬타트뮐러, 쉬미트 | Fuel rods for nuclear reactor fuel assemblies |
| JPH06201869A (en) * | 1992-12-28 | 1994-07-22 | Nippon Nuclear Fuel Dev Co Ltd | Nuclear fuel rod |
| JPH09113682A (en) * | 1995-10-23 | 1997-05-02 | Hitachi Ltd | Zirconium alloy for reactor |
| US5768332A (en) | 1997-03-27 | 1998-06-16 | Siemens Power Corporation | Nuclear fuel rod for pressurized water reactor |
| JPH11118968A (en) | 1997-10-09 | 1999-04-30 | Mitsubishi Materials Corp | Composite cladding tube for nuclear fuel |
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| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| US5699396A (en) * | 1994-11-21 | 1997-12-16 | General Electric Company | Corrosion resistant zirconium alloy for extended-life fuel cladding |
| JPH08194081A (en) * | 1995-01-18 | 1996-07-30 | Hitachi Ltd | Fuel assembly and fuel cladding pipe |
| JP2001004768A (en) * | 1999-06-24 | 2001-01-12 | Hitachi Ltd | Nuclear-fuel cladding tube and its manufacture |
-
2003
- 2003-02-28 KR KR1020030012616A patent/KR100709389B1/en not_active IP Right Cessation
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2004
- 2004-02-06 WO PCT/KR2004/000223 patent/WO2004077452A1/en active Application Filing
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Patent Citations (5)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| KR890002898A (en) * | 1987-07-21 | 1989-04-11 | 쉬타트뮐러, 쉬미트 | Fuel rods for nuclear reactor fuel assemblies |
| JPH06201869A (en) * | 1992-12-28 | 1994-07-22 | Nippon Nuclear Fuel Dev Co Ltd | Nuclear fuel rod |
| JPH09113682A (en) * | 1995-10-23 | 1997-05-02 | Hitachi Ltd | Zirconium alloy for reactor |
| US5768332A (en) | 1997-03-27 | 1998-06-16 | Siemens Power Corporation | Nuclear fuel rod for pressurized water reactor |
| JPH11118968A (en) | 1997-10-09 | 1999-04-30 | Mitsubishi Materials Corp | Composite cladding tube for nuclear fuel |
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| Title |
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| KR20040077162A (en) | 2004-09-04 |
| JP4389252B2 (en) | 2009-12-24 |
| WO2004077452A1 (en) | 2004-09-10 |
| JP2006514312A (en) | 2006-04-27 |
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