• Nuclear Engineering and Technology
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• v.49 no.4
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• pp.663-667
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• 2017

Delayed hydride cracking (DHC) was first observed in pressure tubes in Canadian CANDU reactors. In light water reactors, DHC was not observed until the late 1990s in high-burnup boiling water reactor (BWR) fuel cladding. In recent years, the focus on DHC has resurfaced in light of the increased interest in the cladding integrity during interim conditions. In principle, all spent fuel in the wet pools has sufficient hydrogen content for DHC to operate below $300^{\circ}C$. It is therefore of importance to establish the critical parameters for DHC to operate. This work studies the threshold stress intensity factor ($K_{IH}$) to initiate DHC as a function of temperature in Zry-4 for temperatures between $227^{\circ}C$ and $315^{\circ}C$. The experimental technique used in this study was the pin-loading testing technique. To determine the $K_{IH}$, an unloading method was used where the load was successively reduced in a stepwise manner until no cracking was observed during 24 hours. The results showed that there was moderate temperature behavior at lower temperatures. Around $300^{\circ}C$, there was a sharp increase in $K_{IH}$ indicating the upper temperature limit for DHC. The value for $K_{IH}$ at $227^{\circ}C$ was determined to be $2.6{\pm}0.3MPa$ ${\surd}$m.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.17 no.4
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• pp.375-387
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• 2019

In this study, the radiation dose rates for the design basis fuel of 360 assemblies CANDU spent nuclear fuel transportation cask were evaluated, by measuring radiation source terms for the design basis fuel of a pressurized heavy water reactor. Additionally, radiological safety evaluation was carried out and the validity of the results was determined by radiological technical standards. To select the design basis fuel, which was the radiation source term for the spent fuel transportation cask, the design basis fuels from two spent fuel storage facilities were stored in a spent fuel transportation cask operating in Wolsung NPP. The design basis fuel for each transportation and storage system was based on the burnup of spent fuel, minimum cooling period, and time of transportation to the intermediate storage facility. A burnup of 7,800 MWD/MTU and a minimum cooling period of 6 years were set as the design basis fuel. The radiation source terms of the design basis fuel were evaluated using the ORIGEN-ARP computer module of SCALE computer code. The radiation shielding of the cask was evaluated using the MCNP6 computer code. In addition, the evaluation of the radiation dose rate outside the transport cask required by the technical standard was classified into normal and accident conditions. Thus, the maximum radiation dose rates calculated at the surface of the cask and at a point 2 m from the surface of the cask under normal transportation conditions were respectively 0.330 mSv·h^-1 and 0.065 mSv·h^-1. The maximum radiation dose rate 1 m from the surface of the cask under accident conditions was calculated as 0.321 mSv·h^-1. Thus, it was confirmed that the spent fuel cask of the large capacity heavy water reactor had secured the radiation safety.

• Nuclear Engineering and Technology
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• v.28 no.4
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• pp.349-358
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• 1996

The leaching behavior of uranium from unirradiated CANDU UO$_2$ fuel pellet in the spent fuel wet storage and disposal conditions has been investigated. A modified IAEA leach test method was used, and then the extent of leaching was monitored by analysis for uranium in the leachant. The leach test has been performed in various leachants(demineralized water and boric acid solution at pH=6, synthetic granite groundwater) for a long-term period of 5.4 years, and the effect of temperature on the leach rate of uranium has been analyzed. The leach rates of uranium at $25^{\circ}C$ were dependent on the leachants. Over initial 100 days of leach periods, the leach rate in groundwater was the highest in three leachants and no significant differences of leach rates ore observed in the demineralized oater and boric acid solution. But these leach rates in three leachants around 2,000 days at $25^{\circ}C$ appeared to be reached the steady rates in the range of 1~5$\times$10$^{-8}$ g/$\textrm{cm}^2$ day. The leach rate of uranium in groundwater shooed to be independent of the temperature, but those in both demineralized water and boric acid solution increased with temperature. These results show that the leaching behavior of uranium from UO$_2$ fuel in both the demineralized water ann boric acid may be controlled tv the surface oxidative.dissolution reaction of UO$_2$ and the leach rate of uranium in groundwater at room temperature could mainly be controlled by the complex reaction of dissolved uranyl ions with carbonate ions and no variation of leach rate of UO$_2$ in groundwater with temperature may be due to the local deposition of passivating uranyl phases on the surface.