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Thursday, July 30, 2020 | History

1 edition of Mechanical properties of unirradiated and irradiated Zr-1% Nb cladding found in the catalog.

Mechanical properties of unirradiated and irradiated Zr-1% Nb cladding

Mechanical properties of unirradiated and irradiated Zr-1% Nb cladding

procedures and results of low temperature biaxial burst tests and axial tensile tests

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Published by Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission in Washington, D.C .
Written in English

    Subjects:
  • Nuclear fuel claddings -- Testing.,
  • Zirconium alloys -- Effect of radiation on.,
  • Niobium alloys -- Effect of radiation on.

  • Edition Notes

    Statementprepared by E. Kaplar ... [et al.].
    SeriesInternational agreement report ;, NUREG/IA-0199
    ContributionsKaplar, E. P.
    Classifications
    LC ClassificationsTK9207.5 .M43 2001
    The Physical Object
    Paginationix, 4, 12, 9, 1 p. :
    Number of Pages12
    ID Numbers
    Open LibraryOL3992516M
    LC Control Number2001330393

    Specimen geometries have been developed to determine the mechanical properties of irradiated Zircaloy cladding subjected to the mechanical conditions and temperatures associated with reactivity-initiated accidents (RIA) and loss-of-coolant accidents (LOCA). Miniature ring-stretch specimens were designed to induce both uniaxial and plane-strain states of stress in the transverse (hoop Cited by: @article{osti_, title = {Development of data base with mechanical properties of un- and pre-irradiated VVER cladding}, author = {Asmolov, V and Yegorova, L and Kaplar, E and Lioutov, K and Smirnov, V and Prokhorov, V and Goryachev, A}, abstractNote = {Analysis of recent RIA test with PWR and VVER high burnup fuel, performed at CABRI, NSRR, IGR reactors has shown that the data base .

    Abstract. In a round robin effort between the US Nuclear Regulatory Commission, Institut de Protection et de Surete Nucleaire in France, and the Russian Research Centre-Kurchatov Institute, Argonne National Laboratory conducted 16 modified ring stretch tensile tests on unirradiated samples of Zr-1Nb cladding, which is used in Russian VVER reactors. @article{osti_, title = {Modified ring stretch tensile testing of Zr-1Nb cladding}, author = {Cohen, A. B. and Majumdar, S. and Ruther, W. E. and Billone, M. C. and Chung, H. M. and Neimark, L. A.}, abstractNote = {In a round robin effort between the US Nuclear Regulatory Commission, Institut de Protection et de Surete Nucleaire in France, and the Russian Research Centre-Kurchatov.

    Irradiation embrittlement ofaustenitic stainless steels K. Q. Bagley, J. W. Barnaby, and A. S. Fraser UKAEA, DERE, Thurso Introduction Experimental Procedures Mechanical Test Results Metallography and Electron Microscopy Damage Mechanisms The effects of fast neutron irradiation on the high temperature mechanical properties of a group.   results of out-of-pile mechanical tests of non-irradiated and irradiated Zr-1%Nb claddings. This volume of the report is of independent significance and contains an overview of the research program as well as results of investigations carried out to study the behavior of VVER fuel rods under reactivity accident conditions.


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Mechanical properties of unirradiated and irradiated Zr-1% Nb cladding Download PDF EPUB FB2

Number of tests Temperature (K) Unirradiated Irradiated 1 - 1 1 2 1 2 2 2 1 1 1 E___ 1 7 8. It should be noted that cladding specimens for mechanical tests of both types were made of a material simi lar to that of the cladding specimens used in the first two stages of.

Mechanical Properties of Unirradiated and Irradiated Zr-1% Nb Cladding: Procedures and Results of Low Temperature Biaxial Burst Tests and Axial Tensile Tests (NUREG/IA, IPSNNSI RRC ) On this page: Publication Information; Abstract; Download complete document. NUREG/IA (PDF MB) Publication Information.

This report contains the description of the test conditions, procedures, and main results of the following types of mechanical tests of unirradiated and irradiated Zr-1 %Nb cladding: "* biaxial burst tests with liquid pressurized tube specimens in the temperature range K; "* uniaxial tensile tests in axial direction with tube specimens in the temperature range K.

Results of. Mechanical properties of unirradiated and irradiated Zr-1% Nb cladding. (OCoLC) Online version: Kaplar, E. (Evgeniĭ Petrovich). Mechanical properties of unirradiated and irradiated Zr-1% Nb cladding. (OCoLC) Material Type: Government publication, National government publication: Document Type: Book: All Authors.

Mechanical properties of unirradiated and irradiated Zr-1% Nb cladding. (OCoLC) Print version: Mechanical properties of unirradiated and irradiated Zr-1% Nb cladding.

(OCoLC) Material Type: Document, Government publication, National government publication, Internet resource: Document Type: Internet Resource, Computer File.

The present study deals with the proton irradiation of Zr–1 wt.% Nb alloy, which is a established fuel cladding material in the Russian Pressurized Water Reactor (PWR) of VVER type, and is also being used as a fuel clad in the Indian VVER-PWR [1], [40].

This alloy has shown superior properties with respect to resistance to corrosion, Cited by: Comparison of the ductility of hydrided Zr-1wt% Nb and Zrwt% Nb at 25 ºC and ºC.

At two test temperatures both Zr-1wt% Nb and Zrwt% Nb heat treated during 24h to ºC do not show significant variations in mechanical properties, but it could be seen a hardening produced by the hydrogen by: 4.

Comparison of Mechanical Properties of Zr-1%Nb and Zr%Nb Alloys The Zr%Nb alloy was studied for the cases of non-irradiated and irradiated. The Zr-1%Nb alloy was studied only the case of unirradiated 2. Experimental procedure (Materials and Methods) Materials were provided by the company Teledyne Wah Chang by: 2.

Structural and mechanical properties of γ-irradiated Zr/Nb multilayer nanocomposites Article (PDF Available) in Materials Letters October with Reads How we measure 'reads'.

The present study rationalizes the effect of proton irradiation on the microstructure and mechanical properties of Zr-1 wt.% Nb alloy which is used as fuel cladding in VVER light water : S. Saini, N. Gayathri, N. Gayathri, S.K. Sharma, Aruna Devi, A.P.

Srivastava, S. Neogy, P. Mukherjee. The creep rates of ZrNbCu cladding tubes stress-relieved at °C were found to be appreciably higher than those of ZrNbCu annealed at °C.

The dislocation densities ρ have been measured from the relation ρ=(ρ D ρ S) 1/2, where ρ D =3/D av 2 (dislocation density due to domain size) and ρ S =k〈ϵ l 〉 2 /b 2 (dislocation density due to strain), k is a material constant and b ̄ is the modulus of Burger's vector, 1/3[1 1 2 ̄ 0], and are listed in Table has been observed that the order of dislocation density of the Cited by: Thus, the study of.

their microstructure and mechanical properties, which are affected by radiation throughout useful life, is essential. either for security reasons or eventually for the life extension of the utility. One of the effects of irradiation which mostly affects mechanical behavior is irradiation hardening.

The high resistance to nodular corrosion and irradiation-induced creep and growth shown by Zr-1% Sn-1% Nb% Fe compared with Zircaloy or binary Zr Nb alloys requires a scientific explanation of. APT also revealed that Fe- and Nb -enriched nano-clusters (less than 20 nm diameter) are present in the Zr matrix for ZrNb and ZrNb.

Irradiation was found to reduce the matrix Nb. hydrogenated water, was the preferred cladding in Western PWRs. At the 2nd U.N. Conf. on the Peaceful Uses of Atomic Energy in Geneva in the Union of Soviet Socialist Republics (USSR) revealed extens ive corrosion data from their development of Zr-Nb alloys for fuel cladding and structural alloys in their water-cooled reactors [Ambartsumyan.

Zr–1Nb samples were irradiated with MeV O 5 + ions at different doses ranging from 5 × 10 17 to 8 × 10 18 O 5 + /m 2. X-ray diffraction line profile analysis was performed to characterize the microstructural parameters of these samples.

Average domain size, microstrain and dislocation density were estimated as a function of dose. An anomaly was observed in the values of these. Specimen geometries have been developed to determine the mechanical properties of irradiated Zircaloy cladding subjected to the mechanical conditions and temperatures associated with reactivity-initiated accidents (RIA) and loss-of-coolant accidents (LOCA).Cited by: Ring compression tests on the oxidized ZrNbCu cladding tubes were performed to examine the mechanical reliability of Zr-Nb-Cu cladding tubes.

In compression of the non-oxidized cladding tubes, large load drop associated with the cladding failure appeared at the total displacement of mm. After oxidation at °C for 24 hrs and °C for 3 hrs, small drop of load resulting from Cited by: 1.

Fig. 5Effect of irradiation A (°C) on the tensile properties of low boron cold worked and aged L, including the influence of helium variation Fig. 6Isochronal (24 h) annealing of displacement damage in L and L as measured by the hardness effect Fig.

7Precipitation of 30 Å helium bubbles on the dislocation network of 20% cold. @article{osti_, title = {Mechanical properties of irradiated fast breeder reactor cladding and ducts}, author = {Johnson, G D and Hunter, C W}, abstractNote = {Austenitic stainless steels are being used for various core components in Liquid Metal Fast Breeder Reactors.

Twenty percent cold worked Type stainless steel is being used for both fuel pin cladding and ducts in the Fast Flux.irradiation to support the enhanced corrosion resistance. Therefore, the Nb redistribution in the solid solution upon irradiation has become a critical knowledge gap to understand its in-reactor corrosion kinetics.

In this study, the microstructure and microchemistry of irradiated ZrNb have been characterized using (Scanning).Mechanical properties of unirradiated and irradiated Zr-1% Nb cladding: procedures and results of low temperature biaxial burst tests and axial tensile tests / by: Kaplar, E.

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