Is spent nuclear fuel immune from delayed hydride cracking during dry storage? An IAEA coordinated research project

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dc.contributor.authorCOLEMAN, CHRISTOPHER E.
dc.contributor.authorMARKELOV, VLADIMIR A.
dc.contributor.authorROTH, MARIA
dc.contributor.authorMAKAREVICIUS, VIDAS
dc.contributor.authorHE, ZHANG
dc.contributor.authorCHAKRAVARTTY, JAYANTA K.
dc.contributor.authorALVAREZ-HOLSTON, ANNA-MARIA
dc.contributor.authorALI, LIAQAT
dc.contributor.authorRAMANATHAN, LALGUDI
dc.contributor.authorINOZEMTSEV, VICTOR
dc.contributor.editorCOMSTOCK, ROBERT J.
dc.contributor.editorMOTTA, ARTHUR T.
dc.coverageInternacionalpt_BR
dc.date.accessioned2019-03-07T11:41:10Z
dc.date.available2019-03-07T11:41:10Z
dc.date.issued2018pt_BR
dc.description.abstractDelayed hydride cracking (DHC) has been responsible for cracking in zirconium alloy pressure tubes and fuel cladding and is a concern for spent fuel storage. For cracking to start, sufficient hydrogen must be present for hydride to form at a flaw tip and the local tensile stress must be sufficiently large to crack the hydride (a crack will not extend if the threshold in the stress intensity factor, KIH, is not exceeded. A high-temperature limit exists when the yield stress of the cladding alloy becomes too low to crack the hydride. In this paper we describe measurements of KIH and the crack growth rate, V, in unirradiated Zircaloy-4 fuel cladding containing approximately 130 ppm hydrogen in the cold-worked stress–relieved condition representing pressurized water reactors (PWRs) and pressurized heavy-water (PHWR) reactors. Four methods are used to evaluate KIH. The test specimen and fixture used in these methods was the pin-loading tension configuration. The test temperature ranged from 227 to 315 C. The mean value of KIH below 280 C had little temperature dependence; it was about 5.5 MPaHm in the PWR cladding and slightly higher at 7 MPaHm in the PHWR material. At higher test temperatures, KIH increased dramatically to more than 12 MPaHm, whereas the crack growth rate declined toward zero. This behavior suggests that unirradiated Zircaloy-4 fuel cladding is immune from DHC above about 320 C; this temperature may be increased to 360 C by irradiation. The implications for spent fuel storage are that during early storage when the temperatures are high, any flaw will not extend by DHC, whereas at low temperatures, after many years of storage, flaws would have to be very large, approaching through wall, before being extended by DHC. To date, spent nuclear fuel is not known to have failed by DHC during storage, confirming the inference.pt_BR
dc.format.extent1224-1251pt_BR
dc.identifier.citationCOLEMAN, CHRISTOPHER E.; MARKELOV, VLADIMIR A.; ROTH, MARIA; MAKAREVICIUS, VIDAS; HE, ZHANG; CHAKRAVARTTY, JAYANTA K.; ALVAREZ-HOLSTON, ANNA-MARIA; ALI, LIAQAT; RAMANATHAN, LALGUDI; INOZEMTSEV, VICTOR. Is spent nuclear fuel immune from delayed hydride cracking during dry storage? An IAEA coordinated research project. In: COMSTOCK, ROBERT J. (ed.); MOTTA, ARTHUR T. (ed.). <b>Zirconium in the Nuclear Industry: 18th International Symposium</b>. West Conshohocken, PA: ASTM International, 2018. p. 1224-1251. (Selected Technical Papers, STP1597). DOI: <a href="https://dx.doi.org/10.1520/STP159720160048">10.1520/STP159720160048</a>. Disponível em: http://repositorio.ipen.br/handle/123456789/29752.
dc.identifier.doi10.1520/STP159720160048pt_BR
dc.identifier.orcid0000-0003-4822-8840pt_BR
dc.identifier.orcidhttps://orcid.org/0000-0003-4822-8840
dc.identifier.urihttp://repositorio.ipen.br/handle/123456789/29752
dc.localWest Conshohocken, PApt_BR
dc.publisherASTM Internationalpt_BR
dc.relation.ispartofseriesSelected Technical Papers, STP1597pt_BR
dc.rightsclosedAccesspt_BR
dc.subjectzircaloy 4
dc.subjectcladding
dc.subjectdry storage
dc.subjecttemperature dependence
dc.subjectimmunity
dc.subjecthydrides
dc.subjectcracking
dc.titleIs spent nuclear fuel immune from delayed hydride cracking during dry storage? An IAEA coordinated research projectpt_BR
dc.title.livroZirconium in the Nuclear Industry: 18th International Symposiumpt_BR
dc.typeCapítulo de livropt_BR
dspace.entity.typePublication
ipen.autorLALGUDI VENKATARAMAN RAMANATHAN
ipen.codigoautor95
ipen.contributor.ipenauthorLALGUDI VENKATARAMAN RAMANATHAN
ipen.date.recebimento19-03pt_BR
ipen.identifier.ipendoc25520pt_BR
ipen.type.genreCapítulo
relation.isAuthorOfPublication8e95d525-c559-413e-ac38-1e26a8fd67a3
relation.isAuthorOfPublication.latestForDiscovery8e95d525-c559-413e-ac38-1e26a8fd67a3
sigepi.autor.atividadeRAMANATHAN, LALGUDI:95:730:Npt_BR
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