WO2008090269A1 - Procede de conception d'un assemblage de combustible optimise en fonction des contraintes d'utilisation en reacteur nucleaire a eau legere, et assemblage de combustible en resultant - Google Patents
Procede de conception d'un assemblage de combustible optimise en fonction des contraintes d'utilisation en reacteur nucleaire a eau legere, et assemblage de combustible en resultant Download PDFInfo
- Publication number
- WO2008090269A1 WO2008090269A1 PCT/FR2007/002018 FR2007002018W WO2008090269A1 WO 2008090269 A1 WO2008090269 A1 WO 2008090269A1 FR 2007002018 W FR2007002018 W FR 2007002018W WO 2008090269 A1 WO2008090269 A1 WO 2008090269A1
- Authority
- WO
- WIPO (PCT)
- Prior art keywords
- content
- whose
- exceed
- elements
- components
- Prior art date
Links
- 239000000446 fuel Substances 0.000 title claims abstract description 22
- 238000000034 method Methods 0.000 title claims abstract description 16
- XLYOFNOQVPJJNP-UHFFFAOYSA-N water Substances O XLYOFNOQVPJJNP-UHFFFAOYSA-N 0.000 title claims description 12
- 229910045601 alloy Inorganic materials 0.000 claims abstract description 81
- 239000000956 alloy Substances 0.000 claims abstract description 81
- 229910001093 Zr alloy Inorganic materials 0.000 claims abstract description 17
- 230000006835 compression Effects 0.000 claims description 32
- 238000007906 compression Methods 0.000 claims description 32
- 238000005260 corrosion Methods 0.000 description 22
- 230000007797 corrosion Effects 0.000 description 22
- 238000004845 hydriding Methods 0.000 description 6
- 239000000463 material Substances 0.000 description 5
- 239000000126 substance Substances 0.000 description 3
- 238000005275 alloying Methods 0.000 description 2
- 230000000712 assembly Effects 0.000 description 2
- 238000000429 assembly Methods 0.000 description 2
- 238000009835 boiling Methods 0.000 description 2
- 238000004364 calculation method Methods 0.000 description 2
- 230000004907 flux Effects 0.000 description 2
- 239000000203 mixture Substances 0.000 description 2
- 230000015572 biosynthetic process Effects 0.000 description 1
- 230000008094 contradictory effect Effects 0.000 description 1
- 230000002089 crippling effect Effects 0.000 description 1
- 239000012530 fluid Substances 0.000 description 1
- 239000012535 impurity Substances 0.000 description 1
- 238000005457 optimization Methods 0.000 description 1
- 239000008188 pellet Substances 0.000 description 1
- 238000004088 simulation Methods 0.000 description 1
- 150000003754 zirconium Chemical class 0.000 description 1
- 229910052726 zirconium Inorganic materials 0.000 description 1
Classifications
-
- 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
-
- 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
- G21C3/00—Reactor fuel elements and their assemblies; Selection of substances for use as reactor fuel elements
- G21C3/42—Selection of substances for use as reactor fuel
- G21C3/58—Solid reactor fuel Pellets made of fissile material
- G21C3/60—Metallic fuel; Intermetallic dispersions
-
- 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 relates to the design of fuel assemblies for light water nuclear reactors.
- the zirconium alloy components of the fuel assemblies are subjected to severe stresses that cause their deformation.
- the main components involved are the mixing grids, the guide tubes, the sheaths containing the fuel pellets and the housings.
- Deformation of the structural components of the assembly can cause many operating problems.
- the overall deformation of the assembly caused essentially by that of the guide tubes or the housing, may interfere with the operation of the control clusters that allow the control of the reactor.
- the deformations of the components such as the grids or the casing increase the friction and the risk of snagging. This often leads the operator to reduce the handling speeds, thus increasing the duration of unavailability of the reactor.
- the deformation of the components no longer enables them to perform their function safely, and may lead the operator to prematurely discharge the fuel assembly concerned.
- the object of the invention is to propose a method making it possible to optimize the choice of materials used to produce the various structural components of the fuel assembly according to the specific operating conditions of the reactor or the family of reactors considered in order to to minimize the total deformation of the components. It is therefore necessary to define the chemical composition of the Zr alloys used to produce the different structural components of the fuel assembly that will make it possible to achieve the performance expected by the operator, in terms of maneuverability, ease of operation and operating, life time ...
- the subject of the invention is a method for designing a fuel assembly for a light-water nuclear reactor comprising zirconium alloy structure components, characterized in that: the average uniaxial tensile stresses are calculated or compression to which will be subject said components during the life of the assembly;
- At least some of said components subjected to an axial tensile stress or across between 10 and 20 MPa will be made of an alloy whose content of elements other than Zr does not exceed 5% and whose Sn content is between 0.50% and 1, 25%.
- said zirconium alloys in which said components will be made are selected according to the following criteria:
- At least some of said components subjected to an axial tensile stress or across between 10 and 20 MPa will be made of an alloy whose content of elements other than Zr does not exceed 5% and whose Sn content is between 0.50 and 1, 25%;
- At least some of said components subjected to an axial tensile stress or across between 10 and 20 MPa will be made of an alloy whose content of elements other than Zr does not exceed 5% and whose Sn content is between 0.75% and 1, 25%.
- an alloy whose content of elements other than Zr does not exceed 5% and whose Sn content is between 0.75% and 1, 25%.
- At least some of those subjected to an axial tensile stress or across ⁇ between 0 and + 10 MPa will be made of an alloy of which the content of elements other than Zr does not exceed 5% and whose Sn content is equal to 0,1. ⁇ % or deviates from the thus calculated value of + 20% thereof;
- At least some of those subjected to an axial compression stress or across ⁇ ranging between 0 and -10 MPa will be made of an alloy whose content of elements other than Zr does not exceed 5% and whose Sn content is equal to 0.025. ⁇ % or deviates from the thus calculated value of + 20% thereof;
- At least some of said components have a total content of elements other than Zr not exceeding 3.5%. At least some of said components may have a Nb content of 0.5 to 3%.
- the invention also relates to a fuel assembly for light water nuclear reactor comprising Zr alloy components subjected to axial stresses or through compression or traction ⁇ , characterized in that said components are made of selected alloys by application of the above method.
- the invention is based on the initial reasoning according to which the Sn content of the alloy in which a component is made has a marked influence on its properties, and the choice of this content must be carried out according to thermal, mechanical and physico-chemical stresses to which the component is expected to be subjected during the use of the reactor, optimally also taking into account the more or less rescristallized or expanded state of the component.
- FIG. 1 shows the minimum, maximum and preferred contents of Sn that, according to the invention, the components of the fuel assembly as a function of the axial tensile or compressive stress to which they are subjected, in the case where the alloy is in the recrystallized state;
- FIG. 2 which shows the minimum, maximum and preferred levels of Sn which, according to the invention, are imposed on the components of the fuel assembly as a function of the axial tensile or compressive stress to which they are subjected, in the case where the alloy is in the relaxed state;
- FIG. 3 which shows the minimum and maximum Sn levels that, according to the invention, are imposed on the components of the fuel assembly as a function of the axial tensile or compressive stress to which they are subjected, in the most general case. What will be said is valid for Zr alloys whose content of alloying elements other than Zr does not exceed 5%, preferably does not exceed 3.5%.
- the Sn content of a Zr alloy has a marked influence on both its corrosion behavior and its creep resistance, which is one of the most important mechanical characteristics to consider when assessing the behavior of a component. .
- the invention is based on the concept that the Sn content of the various components of a reactor must be optimized so that the component corrodes little and is slightly subject to deformation, under the precise conditions where it will be used. This optimization must then be refined thanks to the choice of the contents of the other elements, in particular of O and S which have a significant influence on the creep and Fe which has a significant influence on the corrosion.
- the inventors carried out a modeling of the creep behavior of Zr alloys containing at most 5% of elements other than Zr, as a function of: the uniaxial stress ⁇ of compression or traction applied to the component considered , in the range -20MPa to + 20MPa, averaged over the life of the assembly;
- the inventors also took into account the influence of the temperature and the physico-chemical conditions of use of these alloys in the usual environments of the reactors. In particular, it was necessary to take into account the magnification under irradiation caused by the neutron flux, and the constraints caused by the formation of the layer of oxide due to corrosion of the material. Also, the hyd r ration, which causes a magnification of the material, and the friction of the fluid on the assembly, have been taken into account.
- temperatures to which the components are subjected they have been considered to be typically 280 to 360 ° C. for a pressurized water reactor and 280 to 300 ° C. for a boiling water reactor.
- Figure 1 relates to alloys in the fully recrystallized state
- Figure 2 relates to alloys in the relaxed state.
- the curves referenced 1 correspond to the minimum contents of Sn to be imposed according to the invention.
- the curves 2 correspond to the maximum levels of Sn to be imposed according to the invention when a high resistance of the component to corrosion and hydriding is sought; the curves 2 'correspond to a variant of the curves 2 corresponding to the case where a high resistance of the component to corrosion and hydriding is not particularly sought after.
- the curves 3 correspond to the Sn contents considered optimal.
- the uniaxial stress in question may be a longitudinal (axial) compression or tensile stress, as is the case on guide tubes, ducts and housings, or a compressive or tensile stress, such as this is the case on the grids.
- the preferential choice criteria are as follows, applied to alloys whose total content of elements other than Zr does not exceed 5%, preferably 3.5%.
- it is equal to 0.1. ⁇ % or deviates from the thus calculated value of + 20% of it. If this calculated value is less than 0.1%, 0.15% is taken as the optimum upper limit of the Sn content.
- the Sn content is between 0.50% and 1.70% when a high corrosion resistance is not present. not particularly desired, or between 0.50 and 1, 25% when a high resistance to corrosion is sought. Optimally, it is equal to 1% + 0.2%.
- the preferred choice criteria are as follows, applied to alloys whose total content of elements other than Zr does not exceed 5%, preferably not more than 3,5 %.
- Sn 0.1. ⁇ % or deviates from this value by + 20% of it. If this calculated value is less than 0.1%, 0.15% is taken as the optimum upper limit of the Sn content.
- the Sn content is between 0.75 and 1.70% if a high resistance to corrosion is not particularly sought, or between 0.75% and 1.25% if a high resistance to corrosion is sought. Optimally, the Sn content is equal to 1% + 0.2%.
- the alloy When the alloy is in a partially recrystallized state, it can be given, for a given compressive or axial tensile stress, a intermediate value between that defined as above for a fully recrystallized alloy and that defined for an alloy in the relaxed state. As a first approximation, it will be possible to assimilate an alloy recrystallized to more than 50% to a fully recrystallized alloy, and an alloy recrystallized to less than 50% to a relaxed alloy.
- the Sn content is between 0.50% (curve 1 of FIG. 1) and 1.70% (curve 2 'of FIG. 2) when a high resistance to corrosion to the component is not particularly desired and between 0.50% (curve 1 of FIG. 1) and 1.25% (curve 2 of FIG. High corrosion resistance of the component is sought.
- the optimal Sn content is at most 0.15%, and can go down to simple traces as impurities resulting from the elaboration of the alloy. In general, the optimum Sn content is higher in the relaxed states than in the recrystallized states due to the higher creep rate.
- the invention applies, as mentioned, to Zr alloys containing up to 5% (better, up to 3.5%) of alloying elements other than Zr.
- the fuel assembly produced according to the invention may jointly use structural components meeting one or the other of the compositional criteria according to the invention.
Landscapes
- Engineering & Computer Science (AREA)
- Physics & Mathematics (AREA)
- Chemical & Material Sciences (AREA)
- Metallurgy (AREA)
- Plasma & Fusion (AREA)
- High Energy & Nuclear Physics (AREA)
- General Engineering & Computer Science (AREA)
- Materials Engineering (AREA)
- Mechanical Engineering (AREA)
- Organic Chemistry (AREA)
- Dispersion Chemistry (AREA)
- Fuel-Injection Apparatus (AREA)
- Heat Treatment Of Steel (AREA)
- Monitoring And Testing Of Nuclear Reactors (AREA)
Abstract
Description
Claims
Priority Applications (6)
Application Number | Priority Date | Filing Date | Title |
---|---|---|---|
CN2007800509255A CN101601101B (zh) | 2006-12-11 | 2007-12-07 | 根据轻水核反应堆的使用应力优化的燃料组件的设计方法以及由此得到的燃料组件 |
US12/448,157 US8576977B2 (en) | 2006-12-11 | 2007-12-07 | Method for designing a fuel assembly optimized as a function of the stresses in use in light-water nuclear reactors, and resulting fuel assembly |
JP2009540807A JP5305299B2 (ja) | 2006-12-11 | 2007-12-07 | 軽水炉中で使用される応力の関数で最適化される核燃料集合体の設計方法およびその結果の核燃料集合体 |
ES07871813.7T ES2524616T3 (es) | 2006-12-11 | 2007-12-07 | Procedimiento de concepción de un ensamblaje de combustible optimizado en función de los esfuerzos de utilización en un reactor nuclear de agua ligera |
EP07871813.7A EP2126926B1 (fr) | 2006-12-11 | 2007-12-07 | Procede de conception d'un assemblage de combustible optimise en fonction des contraintes d'utilisation en reacteur nucleaire a eau legere |
KR1020097014408A KR101441944B1 (ko) | 2006-12-11 | 2007-12-07 | 경수원자로에서 사용 조건의 함수로서 최적화된 연료 조립체의 제조 방법 및 그에 따른 연료 조립체 |
Applications Claiming Priority (2)
Application Number | Priority Date | Filing Date | Title |
---|---|---|---|
FR0610785A FR2909798A1 (fr) | 2006-12-11 | 2006-12-11 | Procede de conception d'un assemblage de combustible optimise en fonction des contraintes d'utilisation en reacteur nucleaire a eau legere,et assemblage de combustible en resultant. |
FR0610785 | 2006-12-11 |
Publications (1)
Publication Number | Publication Date |
---|---|
WO2008090269A1 true WO2008090269A1 (fr) | 2008-07-31 |
Family
ID=38480651
Family Applications (1)
Application Number | Title | Priority Date | Filing Date |
---|---|---|---|
PCT/FR2007/002018 WO2008090269A1 (fr) | 2006-12-11 | 2007-12-07 | Procede de conception d'un assemblage de combustible optimise en fonction des contraintes d'utilisation en reacteur nucleaire a eau legere, et assemblage de combustible en resultant |
Country Status (9)
Country | Link |
---|---|
US (1) | US8576977B2 (fr) |
EP (1) | EP2126926B1 (fr) |
JP (1) | JP5305299B2 (fr) |
KR (1) | KR101441944B1 (fr) |
CN (1) | CN101601101B (fr) |
ES (1) | ES2524616T3 (fr) |
FR (1) | FR2909798A1 (fr) |
WO (1) | WO2008090269A1 (fr) |
ZA (1) | ZA200904062B (fr) |
Families Citing this family (1)
Publication number | Priority date | Publication date | Assignee | Title |
---|---|---|---|---|
US10745789B2 (en) * | 2015-06-02 | 2020-08-18 | Ltag Systems Llc | Activated aluminum fuel |
Citations (7)
Publication number | Priority date | Publication date | Assignee | Title |
---|---|---|---|---|
EP0399223A1 (fr) * | 1989-05-25 | 1990-11-28 | General Electric Company | Revêtement résistant à la corrosion, pour barres de combustible nucléaire |
EP0802264A1 (fr) * | 1996-04-16 | 1997-10-22 | Compagnie Européenne du Zirconium CEZUS | Alliage à base de zirconium résistant au fluage et à la corrosion par l'eau et la vapeur, procédé de fabrication, et utilisation dans un réacteur nucléaire |
JPH10273746A (ja) * | 1997-01-28 | 1998-10-13 | Sumitomo Metal Ind Ltd | 冷間加工性と耐食性に優れたジルコニウム合金、この合金を用いた核燃料被覆用二重管およびその製造方法 |
FR2769637A1 (fr) * | 1997-10-13 | 1999-04-16 | Mitsubishi Materials Corp | Procede pour fabriquer un alliage de zirconium pour gainage combustible de reacteur nucleaire ayant une excellente resistance a la corrosion et des proprietes de fluage |
JPH11286736A (ja) * | 1998-02-04 | 1999-10-19 | Korea Atom Energ Res Inst | 核燃料被覆管用ジルコニウム合金組成物 |
US5972288A (en) * | 1998-02-04 | 1999-10-26 | Korea Atomic Energy Research Institute | Composition of zirconium alloy having high corrosion resistance and high strength |
EP1688508A1 (fr) * | 2005-02-07 | 2006-08-09 | Korea Atomic Energy Research Institute | Alliage à base de zirconium, présentant une bonne résistance au fluage |
Family Cites Families (38)
Publication number | Priority date | Publication date | Assignee | Title |
---|---|---|---|---|
US3769159A (en) * | 1968-06-24 | 1973-10-30 | Combustion Eng | Fuel element grid support for nuclear reactor |
US3791466A (en) * | 1969-05-19 | 1974-02-12 | Westinghouse Electric Corp | Low parasitic capture fuel assembly structure |
US3770583A (en) * | 1971-05-20 | 1973-11-06 | Combustion Eng | Fuel assembly hold-down device |
FR2219978B1 (fr) * | 1973-03-02 | 1976-04-30 | Commissariat Energie Atomique | |
US4197145A (en) * | 1974-12-23 | 1980-04-08 | General Electric Company | Zirconium-base alloy structural component for nuclear reactor and method |
US4058436A (en) * | 1975-12-05 | 1977-11-15 | Combustion Engineering Inc. | Nuclear reactor seismic fuel assembly grid |
US4212686A (en) * | 1978-03-03 | 1980-07-15 | Ab Atomenergi | Zirconium alloys |
US4295935A (en) * | 1979-03-29 | 1981-10-20 | Combustion Engineering, Inc. | Bimetallic spacer means for a nuclear fuel assembly |
FR2479536A1 (fr) * | 1980-03-26 | 1981-10-02 | Commissariat Energie Atomique | Perfectionnements aux tubes guides des assemblages combustibles pour reacteur nucleaire et procede de demontage de ces tubes guides |
US4418036A (en) * | 1980-12-16 | 1983-11-29 | Westinghouse Electric Corp. | Fuel assembly for a nuclear reactor |
US4584030A (en) * | 1982-01-29 | 1986-04-22 | Westinghouse Electric Corp. | Zirconium alloy products and fabrication processes |
US4717434A (en) * | 1982-01-29 | 1988-01-05 | Westinghouse Electric Corp. | Zirconium alloy products |
US4707330A (en) * | 1985-01-08 | 1987-11-17 | Westinghouse Electric Corp. | Zirconium metal matrix-silicon carbide composite nuclear reactor components |
US4678632A (en) * | 1985-06-05 | 1987-07-07 | Westinghouse Electric Corp. | Nuclear fuel assembly grid with predetermined grain orientation |
US4879093A (en) * | 1988-06-10 | 1989-11-07 | Combustion Engineering, Inc. | Ductile irradiated zirconium alloy |
FR2642215B1 (fr) * | 1989-01-23 | 1992-10-02 | Framatome Sa | Crayon pour assemblage combustible d'un reacteur nucleaire resistant a la corrosion et a l'usure |
US4986957A (en) * | 1989-05-25 | 1991-01-22 | General Electric Company | Corrosion resistant zirconium alloys containing copper, nickel and iron |
US5073336A (en) * | 1989-05-25 | 1991-12-17 | General Electric Company | Corrosion resistant zirconium alloys containing copper, nickel and iron |
US5245645A (en) * | 1991-02-04 | 1993-09-14 | Siemens Aktiengesellschaft | Structural part for a nuclear reactor fuel assembly and method for producing this structural part |
US5130083A (en) * | 1991-08-15 | 1992-07-14 | General Electric Company | Hydride resistant spacer formed from interlocking strips |
FI923892A (fi) * | 1991-09-16 | 1993-03-17 | Siemens Power Corp | Strukturella element foer en kaernreaktors braenslestavsmontering |
US5190721A (en) * | 1991-12-23 | 1993-03-02 | General Electric Company | Zirconium-bismuth-niobium alloy for nuclear fuel cladding barrier |
FR2686445B1 (fr) * | 1992-01-17 | 1994-04-08 | Framatome Sa | Crayon de combustible nucleaire et procede de fabrication de la gaine d'un tel crayon. |
DE9206038U1 (de) * | 1992-02-28 | 1992-07-16 | Siemens AG, 80333 München | Werkstoff und Strukturteil aus modifiziertem Zirkaloy |
US5267290A (en) * | 1992-06-30 | 1993-11-30 | Combustion Engineering, Inc. | Zirconium alloy absorber layer |
US5267284A (en) * | 1992-06-30 | 1993-11-30 | Combustion Engineering Inc. | Zirconium alloy containing isotopic erbium |
US5278882A (en) * | 1992-12-30 | 1994-01-11 | Combustion Engineering, Inc. | Zirconium alloy with superior corrosion resistance |
KR100231081B1 (ko) * | 1994-08-31 | 1999-11-15 | 멀홀랜드 존 에이치. | 텅스텐과 니켈을 함유하는 지르코늄 합금 |
US5699396A (en) * | 1994-11-21 | 1997-12-16 | General Electric Company | Corrosion resistant zirconium alloy for extended-life fuel cladding |
FR2737335B1 (fr) * | 1995-07-27 | 1997-10-10 | Framatome Sa | Tube pour assemblage de combustible nucleaire et procede de fabrication d'un tel tube |
US5838753A (en) * | 1997-08-01 | 1998-11-17 | Siemens Power Corporation | Method of manufacturing zirconium niobium tin alloys for nuclear fuel rods and structural parts for high burnup |
US5844959A (en) * | 1997-08-01 | 1998-12-01 | Siemens Power Corporation | Zirconium niobium tin alloys for nuclear fuel rods and structural parts for high burnup |
US5835550A (en) * | 1997-08-28 | 1998-11-10 | Siemens Power Corporation | Method of manufacturing zirconium tin iron alloys for nuclear fuel rods and structural parts for high burnup |
US5854818A (en) * | 1997-08-28 | 1998-12-29 | Siemens Power Corporation | Zirconium tin iron alloys for nuclear fuel rods and structural parts for high burnup |
FR2799210B1 (fr) * | 1999-09-30 | 2001-11-30 | Framatome Sa | Alliage a base de zirconium et procede de fabrication de composant pour assemblage de combustible nucleaire en un tel alliage |
KR100441562B1 (ko) * | 2001-05-07 | 2004-07-23 | 한국수력원자력 주식회사 | 우수한 내식성과 기계적 특성을 갖는 지르코늄 합금핵연료 피복관 및 그 제조 방법 |
DE10146128B4 (de) * | 2001-09-19 | 2005-03-03 | Framatome Anp Gmbh | Brennelement für einen Druckwasserreaktor |
US8043448B2 (en) * | 2004-09-08 | 2011-10-25 | Global Nuclear Fuel-Americas, Llc | Non-heat treated zirconium alloy fuel cladding and a method of manufacturing the same |
-
2006
- 2006-12-11 FR FR0610785A patent/FR2909798A1/fr not_active Withdrawn
-
2007
- 2007-12-07 JP JP2009540807A patent/JP5305299B2/ja not_active Expired - Fee Related
- 2007-12-07 EP EP07871813.7A patent/EP2126926B1/fr not_active Not-in-force
- 2007-12-07 ES ES07871813.7T patent/ES2524616T3/es active Active
- 2007-12-07 CN CN2007800509255A patent/CN101601101B/zh not_active Expired - Fee Related
- 2007-12-07 WO PCT/FR2007/002018 patent/WO2008090269A1/fr active Application Filing
- 2007-12-07 US US12/448,157 patent/US8576977B2/en not_active Expired - Fee Related
- 2007-12-07 KR KR1020097014408A patent/KR101441944B1/ko not_active IP Right Cessation
-
2009
- 2009-06-10 ZA ZA200904062A patent/ZA200904062B/xx unknown
Patent Citations (7)
Publication number | Priority date | Publication date | Assignee | Title |
---|---|---|---|---|
EP0399223A1 (fr) * | 1989-05-25 | 1990-11-28 | General Electric Company | Revêtement résistant à la corrosion, pour barres de combustible nucléaire |
EP0802264A1 (fr) * | 1996-04-16 | 1997-10-22 | Compagnie Européenne du Zirconium CEZUS | Alliage à base de zirconium résistant au fluage et à la corrosion par l'eau et la vapeur, procédé de fabrication, et utilisation dans un réacteur nucléaire |
JPH10273746A (ja) * | 1997-01-28 | 1998-10-13 | Sumitomo Metal Ind Ltd | 冷間加工性と耐食性に優れたジルコニウム合金、この合金を用いた核燃料被覆用二重管およびその製造方法 |
FR2769637A1 (fr) * | 1997-10-13 | 1999-04-16 | Mitsubishi Materials Corp | Procede pour fabriquer un alliage de zirconium pour gainage combustible de reacteur nucleaire ayant une excellente resistance a la corrosion et des proprietes de fluage |
JPH11286736A (ja) * | 1998-02-04 | 1999-10-19 | Korea Atom Energ Res Inst | 核燃料被覆管用ジルコニウム合金組成物 |
US5972288A (en) * | 1998-02-04 | 1999-10-26 | Korea Atomic Energy Research Institute | Composition of zirconium alloy having high corrosion resistance and high strength |
EP1688508A1 (fr) * | 2005-02-07 | 2006-08-09 | Korea Atomic Energy Research Institute | Alliage à base de zirconium, présentant une bonne résistance au fluage |
Non-Patent Citations (1)
Title |
---|
BRADLEY, E. ROSS ET AL: "Zirconium in the Nuclear Industry: Eleventh International Symposium. (ASTM Symposium on Zirconium in the Nuclear Industry, held in September 1995, in Garmisch-Partenkirchen, Germany.) [In: ASTM Spec. Tech. Publ., 1996; STP 1295]", 1996, PUBLISHER: (ASTM, PHILADELPHIA, PA.), PP. 12-32, XP002457537 * |
Also Published As
Publication number | Publication date |
---|---|
JP2010512530A (ja) | 2010-04-22 |
ZA200904062B (en) | 2010-04-28 |
EP2126926B1 (fr) | 2014-11-05 |
US8576977B2 (en) | 2013-11-05 |
JP5305299B2 (ja) | 2013-10-02 |
KR20090088945A (ko) | 2009-08-20 |
CN101601101A (zh) | 2009-12-09 |
CN101601101B (zh) | 2012-12-12 |
EP2126926A1 (fr) | 2009-12-02 |
KR101441944B1 (ko) | 2014-09-18 |
US20100091932A1 (en) | 2010-04-15 |
ES2524616T3 (es) | 2014-12-10 |
FR2909798A1 (fr) | 2008-06-13 |
Similar Documents
Publication | Publication Date | Title |
---|---|---|
EP2099943B2 (fr) | Alliage de zirconium resistant a la corrosion en ombres portees pour composant d'assemblage de combustible pour reacteur a eau bouillante, composant realise en cet alliage, assemblage de combustible et son utilisation | |
EP0380381B1 (fr) | Crayon pour assemblage combustible d'un réacteur nucléaire résistant à la corrosion et à l'usure | |
FR2978697A1 (fr) | Tube multicouche ameliore en materiau composite a matrice ceramique, gaine de combustible nucleaire en resultant et procedes de fabrication associes | |
EP1913600A1 (fr) | Element combustible de type plaque macrostructuree | |
FR2953637A1 (fr) | Crayon de combustible nucleaire et procede de fabrication de pastilles d'un tel crayon | |
FR2686445A1 (fr) | Crayon de combustible nucleaire et procede de fabrication de la gaine d'un tel crayon. | |
BE1004393A5 (fr) | Element tubulaire en acier inoxydable presentant une resistance a l'usure amelioree. | |
FR2817385A1 (fr) | Pastille de combustible nucleaire oxyde et crayon comportant un empilement de telles pastilles | |
EP2126926B1 (fr) | Procede de conception d'un assemblage de combustible optimise en fonction des contraintes d'utilisation en reacteur nucleaire a eau legere | |
JP2006226905A (ja) | 金属燃料高速炉炉心 | |
EP0242251A1 (fr) | Organe de structure en alliage austénitique nickel-chrome-fer | |
FR2898727A1 (fr) | Barre de combustible nucleaire annulaire pouvant etre regulee en flux de chaleur de tubes interne et externe | |
Booker | Analytical description of the effects of melting practice and heat treatment on the creep properties of a 2 1/4 Cr-1 Mo steel | |
Andresen | Why historical material degradation experience might not represent future response | |
WO2021004943A1 (fr) | Composant tubulaire de réacteur nucléaire à eau pressurisée et procédé de fabrication de ce composant | |
JPS6333535A (ja) | 原子炉用ジルコニウム合金 | |
Morize et al. | Behavior under irradiation of zirconium alloy strips | |
JPH03214092A (ja) | 核燃料要素 | |
Gittus | Theoretical magnitude of the stresses produced in fuel cladding by the expansion of cracked pellets: Effect of interfacial dislocations (interfaceons) and pellet relaxation | |
FR2849865A1 (fr) | Procede de fabrication d'un demi-produit en alliage de zirconium pour l'elaboration d'un produit plat et utilisation | |
FR2503917A1 (fr) | Assemblage combustible nucleaire pour surregenerateurs rapides | |
JPS6238388A (ja) | 原子燃料用複合被覆管 | |
Yang et al. | Development Status of Accident Tolerant Fuels for Light Water Reactors in Korea | |
Mercer et al. | An investigation of fatigue crack growth in Inconel 718 | |
Sejnoha et al. | Hydrogen in CANDU fuel elements |
Legal Events
Date | Code | Title | Description |
---|---|---|---|
WWE | Wipo information: entry into national phase |
Ref document number: 200780050925.5 Country of ref document: CN |
|
121 | Ep: the epo has been informed by wipo that ep was designated in this application |
Ref document number: 07871813 Country of ref document: EP Kind code of ref document: A1 |
|
ENP | Entry into the national phase |
Ref document number: 2009540807 Country of ref document: JP Kind code of ref document: A |
|
WWE | Wipo information: entry into national phase |
Ref document number: 2007871813 Country of ref document: EP |
|
NENP | Non-entry into the national phase |
Ref country code: DE |
|
WWE | Wipo information: entry into national phase |
Ref document number: 1020097014408 Country of ref document: KR |
|
WWE | Wipo information: entry into national phase |
Ref document number: 12448157 Country of ref document: US |