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ASTM Selected Technical Papers
Zirconium in the Nuclear Industry: 20th International Symposium
Editor
Suresh K. Yagnik,
Suresh K. Yagnik
Symposium Chairperson and STP Editor
1
Electric Power Research Institute (EPRI)
, Palo Alto, CA,
US
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Michael Preuss
Michael Preuss
Symposium Chair and STP Editor
2;
The University of Manchester Manchester
, GB
Monash University
, Clayton/Melbourne,
AU
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ISBN:
978-0-8031-7737-6
No. of Pages:
928
Publisher:
ASTM International
Publication date:
2023
eBook Chapter
Delayed Hydride Cracking of Fuel Rod Claddings Made of Three Zirconium Alloys
By
Nikolay S. Saburov
,
Nikolay S. Saburov
1
JSC A. A. Bochvar High-Technology Scientific Research Institute of Inorganic Materials
, Rogova St., 5A
, P.O. Box 369, Moscow,
RU
123098
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Vladimir A. Markelov
,
Vladimir A. Markelov
1
JSC A. A. Bochvar High-Technology Scientific Research Institute of Inorganic Materials
, Rogova St., 5A
, P.O. Box 369, Moscow,
RU
123098
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Vladimir V. Novikov
Vladimir V. Novikov
1
JSC A. A. Bochvar High-Technology Scientific Research Institute of Inorganic Materials
, Rogova St., 5A
, P.O. Box 369, Moscow,
RU
123098
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Page Count:
23
-
Published:2023
Citation
Saburov, NS, Markelov, VA, & Novikov, VV. "Delayed Hydride Cracking of Fuel Rod Claddings Made of Three Zirconium Alloys." Zirconium in the Nuclear Industry: 20th International Symposium. Ed. Yagnik, SK, & Preuss, M. 100 Barr Harbor Drive, PO Box C700, West Conshohocken, PA 19428-2959 : ASTM International, 2023.
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The threshold stress intensity factor for delayed hydride cracking (DHC; KIH), the rate of stable crack growth (VDHC), and their temperature dependences were determined for unirradiated fuel cladding tubes in comparable metallurgical states made of three alloys representing different alloying systems of zirconium: Zircaloy-4, E635M, and E110opt. The temperature dependences of KIH and VDHC have the same behavior for all alloys but are quantitatively different. VDHC and its temperature dependence decrease for the alloys in the above sequence, whereas KIH increases at minimum values of 5, 6.7, and 11.5 MPa√m, respectively, for Zircaloy-4, E635M, and E110opt. The maximum temperature, above which there is a deviation of VDHC from the Arrhenius behavior with a sharp decrease in it and a concomitant increase in KIH, is 292, 280, and 231°C for Zircaloy-4, E635M, and E110opt, respectively. The obtained results, including an analysis of striations in the fractography of hydride cracks, indicate an increase in the resistance to DHC of the alloys in the order Zircaloy-4, E635M, and E110opt, with the obvious superiority of the latter.
References
1.
Coleman
C. E.
, “Cracking of Hydride-Forming Metals and Alloys
,” in Comprehensive Structural Integrity
, ed. Milne
I.
, Ritchie
R. O.
, and Karihaloo
B.
(Amsterdam, The Netherlands
: Elsevier
: 2003
), 103–161.2.
Perryman
E. C.
W.
, “Pickering Pressure Tube Cracking Experience
,” Nuclear Energy
17
(1978
): 95–105.3.
Field
G. J.
, Dunn
J. T.
, and Cheadle
B. A.
, “Analysis of the Pressure Tube Failure at Pickering NGS ‘A’ Unit 2
,” Canadian Metallurgical Quarterly
24
(1985
): 181–188.4.
Platonov
P. A.
, Ryazantseva
A. V.
, Saenko
G. P.
, Knizhnikov
Y. N.
, and Viktorov
V. F.
, “The Study of Cause of Cracking in Zirconium Alloy Channel Tubes
” (paper presentation, ASTM Zirconium in the Nuclear Industry—Eighth International Symposium, San Diego, CA
, June 19–23, 1988
).5.
Jonsson
A.
, Hallstadius
L.
, Grapengiesser
B.
, and Lysell
G.
, “Failure of a Barrier Rod in Oskarshamn-3
” (paper presentation, International Topical Meeting on LWR Fuel Performance, Avignon, France
, April 21–24, 1991
).6.
Schrire
D.
, Grapengiesser
B.
, Lundholm
I.
, Lysell
G.
, Frenning
G.
, Ronnberg
G.
, and Jonsson
A.
, “Secondary Defect Behaviour in ABB BWR Fuel
” (paper presentation, International Topical Meeting on LWR Fuel Performance, West Palm Beach, FL
, April 17–21, 1994
).7.
Armijo
J. S.
, “Performance of Failed BWR Fuel
” (paper presentation, International Topical Meeting on LWR Fuel Performance, West Palm Beach, FL
, April 17–21, 1994
).8.
Lysell
G.
and Grigoriev
V.
, “Characteristics of Axial Splits in Failed BWR Fuel Rods
,” in Ninth International Symposium on Environment Degradation of Materials in Nuclear Power Systems—Water Reactors
, ed. Bruemmer
S.
, Ford
P.
, and Was
G.
(Hoboken, NJ
: Wiley
, 1999
), 1.169–1.175.9.
Edsinger
K.
, Davies
J. H.
, and Adamson
R. B.
, “Degraded Fuel Cladding Fractography and Fracture Behavior
,” in Zirconium in the Nuclear Industry: 12th International Symposium
, ed. Sabol
G. P.
and Moan
G. D.
(West Conshohocken, PA
: ASTM International
, 2000
), 316–339, 10.
Edsinger
K.
, “A Review of Fuel Degradation in BWRs
” (paper presentation, International Topical Meeting on LWR Fuel Performance, Park City, UT
, April 10–13, 2000
).11.
Shimada
S.
, Eton
E.
, Hayashi
H.
, and Tukuta
Y.
, “A Metallographic and Fractographic Study of Outside-In Cracking Caused by Power Ramp Tests
,” Journal of Nuclear Materials
327
(2004
): 97–113.12.
Chung
H. M.
, “Understanding Hydride and Hydride-Related Processes in High-Burnup Cladding in Spent-Fuel-Storage and Accident Situations
” (paper presentation, International Topical Meeting on LWR Fuel Performance, Orlando, FL
, September 19–22, 2004
).13.
Kim
Y. S.
, “Delayed Hydride Cracking of Spent Fuel Rods in Dry Storage
,” Journal of Nuclear Materials
378
(2008
): 30–34.14.
Coleman
C. E.
, Markelov
V. A.
, Roth
M.
, Makarevicius
V.
, He
Z.
, Chakravartty
J. K.
, Alvarez-Holston
A.-M.
, Ali
L.
, Ramanathan
L.
, and Inozemtsev
V.
, “Is Spent Nuclear Fuel Immune from Delayed Hydride Cracking during Dry Storage? An IAEA Coordinated Research Project
,” in Zirconium in the Nuclear Industry: 18th International Symposium
, ed. Comstock
R. J.
and Motta
A. T.
(West Conshohocken, PA
: ASTM International
, 2018
), 1224–1251, 15.
Efsing
P.
and Petersson
K.
, “Delayed Hydride Cracking in Irradiated Zircaloy Cladding
,” in Zirconium in the Nuclear Industry: 12th International Symposium
, ed. Sabol
G. P.
and Moan
G. D.
(West Conshohocken, PA
: ASTM International
, 2000
), 340–355, 16.
Shek
G. K.
and Graham
D. B.
, “Effects of Loading and Thermal Maneuvers on Delayed Hydride Cracking in Zr-2.5Nb Alloys
,” in Zirconium in the Nuclear Industry: Eighth International Symposium
, ed. Van Swam
L. F.
P.
and Eucken
C. M.
(West Conshohocken, PA
: ASTM International
, 1989
), 89–110, 17.
Coleman
C. E.
and Inozemtsev
V. V.
, “Measurement of Rates of Delayed Hydride Cracking (DHC) in Zr-2.5Nb Alloys—An IAEA Coordinated Research Project
,” Journal of ASTM International
5
(2008
): JAI101091, 18.
Markelov
V. A.
, “Delayed Hydride Cracking of Zirconium Alloys: Appearance Conditions and Basic Laws” (in Russian)
, Deformatsiya i Razrushenie Materialov
1
(2010
): 31–37.19.
Delayed Hydride Cracking of Zirconium Alloy Fuel Cladding, IAEA-TECDOC 1649 (Vienna, Austria:
International Atomic Energy Agency
, 2010
).20.
Coleman
C. E.
, Grigoriev
V.
, Inozemtsev
V. V.
, Markelov
V.
, Roth
M.
, Makarevicius
V.
, Kim
Y. S.
et al, “The Effect of Microstructure on Delayed Hydride Cracking Behavior of Zircaloy-4 Fuel Cladding—An International Atomic Energy Agency Coordinated Research Program
,” Journal of ASTM International
7
(2010
): JAI103008, 21.
Markelov
V.
, Gusev
A.
, Kotov
P.
, and Novikov
V.
, “Temperature Dependence of Delayed Hydride Cracking Velocity in Fuel Claddings Made from Zirconium Alloys of Different Compositions
” (paper presentation, TopFuel 2012, Manchester, UK
, September 2–6, 2012
).22.
Evaluation on Conditions for Hydrogen Induced Degradation of Zirconium Alloys during Fuel Operation and Storage, IAEA-TECDOC 1781 (Vienna, Austria:
International Atomic Energy Agency
, 2015
).23.
Markelov
V.
, Saburov
N.
, Bekrenev
S.
, and Novikov
V.
, “Determination of Threshold Stress Intensity Factor K1H in DHC Tests of Fuel Claddings by Method of Constant Displacement
” (paper presentation, TopFuel 2015, Zurich, Switzerland
, September 13–17, 2015
).24.
Peregud
M.
, Markelov
V.
, Novikov
V.
, Gusev
A.
, Kon’kov
V.
, Pimenov
Y.
, Agapitov
V.
, and Shtutsa
M.
, “Characteristics and Properties of Cladding Tubes for WWER-1000 Higher Uranium Content Fuel Rods
” (paper presentation, WWER Fuel Performance, Modelling and Experimental Support, Helena
, Bulgaria, September 26–October 4, 2009
).25.
Lavchiev
K.
, Konig
M.
, Jadernas
D.
, Karlsson
J.
, Novikov
V. V.
, Markelov
V. A.
, Shevyakov
A. Y.
et al, “Post Irradiation Examination of Cladding from TVS-K Fuel Rods Operated in Ringhals 3 for Three Cycles
” (paper presentation, TopFuel 2021, Santander, Spain
, October 24–28, 2021
).26.
Vasilchenko
I.
, Kushmanov
S.
, and Vyalitsin
V.
, “Objectives of Further Development of WWER Reactor Cores
” (paper presentation, 12th International Conference WWER Fuel Performance, Modelling and Experimental Support, Nesebar, Bulgaria
, September 16–23, 2017
).27.
Markelov
V.
, Novikov
V.
, Saburov
N.
, Gusev
A.
, Kon’kov
V.
, Peregud
M.
, Dolgov
A.
, Volkov
B.
, and Andersson
V.
, “Irradiation Test under Advanced PWR Conditions in the Halden Reactor and Post-Irradiation Examination of Fuel Rod Claddings from Different Zirconium Alloys
” (paper presentation, TopFuel Reactor Fuel Performance 2018, Prague, Czech Republic
, September 30–October 4, 2018
).28.
Grigoriev
V.
, Josefsson
B.
, Lind
A.
, and Rosborg
B.
, “A Pin-Loading Tension Test for Evaluation of Thin-Walled Tubular Materials
,” Scripta Metallurgica et Materialia
33
, no. 1
(1995
): 109–114.29.
Grigoriev
V.
and Jakobsson
R.
, DHC Axial Crack Velocity Measurements in Zirconium Alloy Fuel Cladding, STUDSVIK/N-05/281 (Nukoping, Sweden: Studsvik Nuclear AB
, 2005
).30.
Coleman
C. E.
and Cox
B.
, “Cracking Zirconium Alloys in Hydrogen
,” in Zirconium in the Nuclear Industry: 6th International Symposium
, ed. Franklin
D. G.
and Adamson
R. B.
(West Conshohocken, PA
: ASTM International
, 1984
), 675–690.31.
Standard Test Method for Determining Threshold Stress Intensity Factor for Environment-Assisted Cracking of Metallic Materials
, ASTM E1681-03 (West Conshohocken, PA
: ASTM International
, approved September 10, 2003
).32.
Standard Test Method Measurement of Fracture Toughness
, ASTM E1820-15a (West Conshohocken, PA
: ASTM International
, approved October 15, 2015
), 33.
Smith
R. R.
, and Eadie
R. L.
, “High Temperature Limit for Delayed Hydride Cracking
,” Scripta Metallurgica et Materialia
22
, no. 6
(1988
): 833–836.34.
Sagat
S.
and Puls
M. P.
, “Temperature Limit for Delayed Hydride Cracking in Zr-2.5Nb Alloys
” (paper presentation, International Conference on Structural Mechanics in Reactor Technology, Prague, Czech Republic
, August 17–22, 2003
).35.
Levi
M. R.
and Puls
M. P.
, “DHC Behaviour of Irradiated Zr-2.5Nb Pressure Tubes up to 365 ºC
” (paper presentation, International Conference on Structural Mechanics in Reactor Technology, Beijing, China
, August 7–12, 2005
).36.
Coleman
C.
, Grigoriev
V.
, Inozemtsev
V.
, Markelov
V.
, Roth
M.
, Makarevicius
V.
, Kim
Y. S.
et al, “Delayed Hydride Cracking in Zircaloy Fuel Cladding—An IAEA Coordinated Research Programme
,” Nuclear Engineering Technology
41
(2009
): 171–178.37.
Puls
M. P.
, “Delayed Hydride Cracking: Theory and Experiment
,” in The Effect of Hydrogen and Hydrides on the Integrity of Zirconium Alloy Components
(London, UK
: Springer
, 2012
), 333–408.38.
Nuttall
K.
and Rogowski
A. J.
, “Some Fractographic Aspects of Hydrogen-Induced Delayed Cracking in Zr-2.5wt%Nb Alloys
,” Journal of Nuclear Materials
80
(1979
): 279–290.39.
Broek
D.
, Elementary Engineering Fracture Mechanics
(Leiden, The Netherlands
: Martinus Nijhoff Publishers
, 1982
).
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