Abstract
Pressure tubes of Advanced Thermal Reactor (boiling-light-water-cooled, heavy-water-moderated, pressure tube-type reactor) in Japan are made of heat-treated Zr-2·5 wt% Nb alloy and both ends are mechanically joined with stainless steel extension tubes. Sharp artificial cracks were introduced in the rolled joint region of pressure tube specimens. The cracks were propagated, and penetrated the tube wall by fatigue and DHC in a high-temperature, high-pressure water loop. From the results, it was shown that the LBB phenomena were valid for the rolled joint region of the pressure tube under the reactor operating conditions and that the critical crack length was more than 50 mm. Calculations were performed for the subsequent leak rate, using critical flow data.
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