• Nuclear Engineering and Technology
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• v.52 no.7
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• pp.1486-1494
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• 2020

To evaluate transport-induced cladding embrittlement after interim-dry storage, ring compression tests were carried out at room temperature(RT) and 135 ℃. The ring compression test specimens were prepared by simulating the interim-dry storage conditions that include four peak cladding temperatures of 250, 300, 350 and 400 ℃, two tensile hoop stresses of 80 and 100 MPa, two hydrogen contents of 250 and 500 wt.ppm-H and a cooling rate of 0.3 ℃/min. Radial hydride fractions of the ring specimens vary depending on those interim-dry storage conditions. The RT compression tests generated lower offset strains than the 135 ℃ ones. In addition, the RT and 135 ℃ compression tests indicate that a higher peak cladding temperature, a higher tensile hoop stress and the lower hydrogen content generated a lower offset strain. Based on the embrittlement criterion of 2.0% offset strain, an allowable peak temperature during the interim-dry storage may be proposed to be less than 350 ℃ under the tensile hoop stress of 80 MPa at the terminal cool-down temperature of 135 ℃.

• Journal of the Korean Society for Heat Treatment
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• v.17 no.4
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• pp.216-222
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• 2004

Study was focused on the quenching embrittlement property of Zircaloy-4 cladding simulated Loss Of Coolant Accident (LOCA) environment in terms of high temperature oxidation and phase transformation. Property in LOCA condition of advanced cladding that contained Nb element was also investigated. Claddings were oxidized at given temperature and given time followed by water quenching. The results showed that ${\beta}$ phase which formed at quenching stage has an influence on cladding property. In case of advanced cladding, Nb retards cladding oxidation, thus enhances quenching resistance.

• Journal of the Korean Society for Heat Treatment
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• v.18 no.2
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• pp.112-118
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• 2005

Study was focused on the effect of the cooling rate on the embrittlement behavior of Zircaloy-4 cladding simulated Loss Of Coolant Accident (LOCA) environment. Claddings were oxidized at given temperature and given time followed by various water quenching in the range of $0.6^{\circ}C$ and $100^{\circ}C$ per second. Cladding failed after water quenching above the threshold oxidation. Threshold oxidation was decreased as the cooling rate increased, which is due to the matensite structure formed during fast cooling rate.

• Korean Journal of Materials Research
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• v.15 no.2
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• pp.126-133
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• 2005

Mechanical properties of zirconium cladding were evaluated by ring tension test to simulate Reactivity-Initiated Accident (RIA) at high burnup situation as an out-reactor test. Zircaloy-4 cladding was hydrided up to 1000 ppm as well as oxidized up to $100\;{\mu}m$ to simulate high-burnup situation. After simulated high-burnup treatment, ring tension test was carried out from 0.01 to 1/sec to correlate with actual RIA event. The results showed that ductility and circumferential toughness decreased with the hydrogen content and oxide thickness. Hydride generated inside cladding acted as brittle failure. Oxygen influenced cladding tube by the reduction of load bearing area, oxygen embrittlement, and thermal aging. Correlation between in-reactor RIA parameter like fuel enthalpy and out-reactor toughness was performed and showed a reasonable result.

• Nuclear Engineering and Technology
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• v.53 no.2
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• pp.474-483
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• 2021

To establish the exclusive role of hydrogen on burst behaviour of Zircaloy-4 during loss-of-coolant accident transients, an extensive single-rod burst tests were conducted on both unirradiated as-received and hydrogenated Zircaloy-4 cladding tubes at different heating rates and internal overpressures. The visual observations of cladding tubes during bursting as well as post-burst are presented in detail to understand the effect of hydrogen concentration, heating rate, and internal pressure. Impact of hydrogen on burst parameters-burst stress, burst strain, burst temperature-during loss-of-coolant accident transients are compared and discussed. Rupture at multiple locations for hydrogenated cladding at lower internal pressure and higher heating rate is reported for the very first time. A novel burst criterion accounting hydrogen concentration in nuclear fuel cladding is proposed.

• Nuclear Engineering and Technology
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• v.53 no.11
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• pp.3741-3758
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• 2021

Outcomes of the FALCON code analysis-related part of the STARS-ENSI Service Project on Evaluation of the new U.S.NRC RIA Fuel Safety Criteria and Application to the Swiss Reactors are presented. Substantial conservatism of the updated safety limits for high-temperature and PCMI cladding failure, as proposed in the NRC Regulatory Guide RG 1.236, is confirmed. Applicability of the updated failure limits to fuel safety analysis in the Swiss PWRs, as applied to standard fuel designs using UO2 fuel pellets and SRA Zry-4 as cladding materials is discussed. Conducting of new integral RIA tests with irradiated samples using doped- and gadolinia fuel pellets to support appropriate fuel safety criteria for RIA events is recommended.

• Korean Journal of Metals and Materials
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• v.48 no.5
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• pp.410-416
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• 2010

Studies were carried out to establish the technology for sodium-clad compatibility and to analyze the compatibility behavior of the Sodium-cooled Fast Reactor (SFR) cladding material under a flowing sodium environment. The natural circulation facility caused by the thermal convection of the liquid sodium was constructed and the 316-series stainless steels were exposed at $650{^{\circ}C}$ liquid sodium for 1458 hours. The weight change and related microstructural change were analyzed. The results showed that the quasi-dynamic facility represented by the natural convection exhibited similar results compared to the conventional dynamic facility. Selective leaching and local depletion of the chromium, re-distribution of the carbide, and the decarburization process took place in the 316-series stainless steel under a flowing sodium environment. This process decreased as the sodium flowed along the channel, which was caused by the change in the dissolved oxygen and carbon activity in the liquid sodium.

• Proceedings of the Korean Nuclear Society Conference
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• 2002.10a
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• pp.263-263
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• 2002

• Nuclear Engineering and Technology
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• v.38 no.5
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• pp.405-410
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• 2006

Microstructural features and their related mechanical property changes in the 309L cladding and the heat affected zone (HAZ) of SA508 cl.3 steel were investigated through the use of TEM, tensile and small punch (SP) tests. The specimens were irradiated at 563 K up to the neutron fluences of $5.79{\times}10^{19}n/cm^2$ (>1MeV). The microstructure of the clad was mainly composed of a fcc ${\gamma}-phase$, a low percentage of bcc ${\delta}-ferrite$, and a brittle ${\sigma}-phase$. Along the weld fusion line there formed a heavy carbide precipitation with a width of $20{\sim}40{\mu}m$, showing preferential cracking during plastic deformation. The yield stress and ductile-to-brittle transition temperature (DBTT) of the irradiated clads increased. The origin of the hardening and the shift of the DBTT are discussed in terms of the irradiation-produced defect clusters of a fine size and brittle ${\sigma}-phase$.

• Nuclear Engineering and Technology
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• v.39 no.5
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• pp.591-602
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• 2007

The results of Reactivity-Initiated Accidents (RIA) experiments have been analysed and the main variables affecting the fuel failure propensity identified. Fuel burn-up aggravates the mechanical loading of the cladding, while corrosion, or better the hydrogen absorbed in the cladding as a consequence of corrosion, may under some conditions make the cladding brittle and more susceptible to failure. Experiments point out that corrosion impairs the fuel resistance for RIA transient occurring at cold conditions, whereas there is no evidence of important embrittlement effects at hot conditions, unless the cladding was degraded by oxide spalling. A fuel failure threshold correlation has been derived and compared with experimental data relevant for BWR and PWR fuel. The correlation can be applied to both cold and hot RIA transients, account taken for the lower ductility at cold conditions and for the different initial enthalpy. It can also be used for non-zero power transients, provided that a term accounting for the start-up power is incorporated. The proposed threshold is easy to use and reproduces the results obtained in the CABRI and NSRR tests in a rather satisfactory manner. The behaviour of advanced PWR alloys and of MOX fuel is discussed in light of the correlation predictions. Finally, a probabilistic approach has been developed in order to account for the small scatter of the failure predictions. This approach completes the RIA failure assessment in that after determining a best estimate failure threshold, a failure probability is inferred based on the spreading of data around the calculated best estimate value.