The terminal solid solubility of hydrogen in irradiated Zircaloy-2 and microscopic modeling of hydride behavior K. Une a, * , S. Ishimoto a , Y. Etoh a , K. Ito b , K. Ogata c , T. Baba c , K. Kamimura c , Y. Kobayashi d a Nippon Nuclear Fuel Development, Co., Ltd., 2163 Narita-cho, Oarai-machi, Ibaraki-ken 311-1313, Japan b Global Nuclear Fuel Japan Co., Ltd., 3-1 Uchikawa 2-chome, Yokosuka-shi, Kanagawa-ken 239-0836, Japan c Japan Nuclear Energy Safety Organization, 3-17-1 Toranomon, Minato-ku, Tokyo 105-0001, Japan d M.O.X. Corporation, 1828-520 Hirasu-cho, Mito-shi, Ibaraki-ken 310-0853, Japan article info abstract Differential scanning calorimetry (DSC) has been applied to elucidate the terminal solid solubility (TSS) of hydrogen in Zircaloy-2 cladding tubes and spacer bands irradiated in commercial BWRs. While recovery of irradiation defects during the first heating stage of as-irradiated specimens made the DSC peak of hydride dissolution dull or broader, no significant difference was detected in the TSS between unirradi- ated and irradiated Zircaloy-2, irrespective of fast neutron fluence. The effect of post-irradiation anneal- ing on TSS was also examined. The results suggest almost no interaction between irradiation defects and dissolved hydrogen or hydrides at temperatures around 300 °C. Using the present TSS data and reported hydrogen- and hydride-related properties, a microscopic analysis code HYMAC for analyzing hydride behavior in cladding tube with textured grains was constructed. Stress-induced preferential precipitation and dissolution of hydrides were reproduced by adopting a TSS sub-model in which the solubilities decrease in proportion to stress normal to the habit plane in grains and to grain faces. Analyzed results by the code were consistent with typical experimental results of hydride behavior. Ó 2009 Elsevier B.V. All rights reserved. 1. Introduction Hydride-related embrittlement of fuel cladding and structural material made of Zr alloys is one of the most important issues lim- iting light water reactor fuel performance at high burnup; for example, the outside-in type cracking in high burnup BWR clad- ding arising at power ramp tests [1]. When considering hydride behavior in Zr alloys, the terminal solid solubility (TSS) of hydrogen for hydride dissolution (TSSD) and hydride precipitation (TSSP) is a prime factor, and for further detailed quantitative analyses, a mechanistic and microscopic model containing TSS data is needed. There is only limited TSS data for irradiated Zr alloys, compared to the data for unirradiated materials. Within the authors’ knowledge, there are two papers [2,3] regarding TSS measurements for irradi- ated materials, both carried out by means of differential scanning calorimetry (DSC). Though these studies recognized the increase of hydrogen solubility in irradiated Zircaloys, the increments dif- fered considerably in each. As a mechanism for the increase of sol- ubility, trapping of dissolved hydrogen by irradiation damage was suggested. In contrast, assuming the existence of some interactions between microscopic hydrides and irradiation defects [4], TSSD of irradiated material would become lower than that of unirradiated material. In this work, DSC measurements were made for the purpose of confirming whether any meaningful changes in TSS data, or signif- icant trapping effects, are apparent in Zircaloy-2 irradiated in com- mercial BWRs. Though part of the data has been already reported as a short communication [5], a full set of the data and detailed dis- cussion are presented in this paper. Special attention is paid to the DSC responses, i.e. the competing phenomena of inherent endo- thermic reaction due to hydride dissolution and exothermic reac- tion due to recovery of irradiation defects. Moreover, using the present TSS data and reported hydrogen- and hydride-related properties, a microscopic model of hydride behavior was con- structed for the purpose of analyzing hydride-related phenomena of stress reorientation of hydride, thermal diffusion of hydrogen, and delayed hydride cracking (DHC). In the model, the concept of stress-induced decrease of TSS plays an important role. 2. Experimental 2.1. Materials Irradiated Zircaloy-2 specimens approximately 4 mm square (30–80 mg) were prepared from cladding tubes and spacer bands, which had been irradiated during 3 and 5 cycles in commercial BWRs. The alloy compositions of Zircaloy-2 in wt% were 1.30– 1.37 Sn, 0.16–0.18 Fe, 0.10–0.11 Cr, 0.06–0.07 Ni and 0.11–0.13 0022-3115/$ - see front matter Ó 2009 Elsevier B.V. All rights reserved. doi:10.1016/j.jnucmat.2009.01.017 * Corresponding author. Tel.: +81 29 267 9011; fax: +81 29 267 9014. E-mail address: une@nfd.co.jp (K. Une). Journal of Nuclear Materials 389 (2009) 127–136 Contents lists available at ScienceDirect Journal of Nuclear Materials journal homepage: www.elsevier.com/locate/jnucmat