Kuan-Che Lan

The nickel-base Alloy 617 has been considered as the lead candidate structural material for the intermediate heat exchanger (IHX) of the Very-High-Temperature Reactor (VHTR). In order to assess the long-term performance of Alloy 617, thermal aging experiments up to 10,000 h in duration were performed at 1000 °C. Subsequently, in-situ synchrotron wide-angle X-ray scattering (WAXS) tensile tests were carried out at ambient temperature. M23C6 carbides were identified as the primary precipitates, while a smaller amount of M6C was also observed. The aging effects were quantified in several aspects: (1) macroscopic tensile properties, (2) volume fraction of the M23C6 phase, (3) the lattice strain evolution of both the matrix and the M23C6 precipitates, and (4) the dislocation density evolution during plastic deformation. The property?microstructure relationship is described with a focus on the evolution of the M23C6 phase. For aging up to 3000 h, the yield strength (YS) and ultimate tensile strength (UTS) showed little variation, with average values being 454 MPa and 787 MPa, respectively. At 10,000 h, the YS and UTS reduced to 380 MPa and 720 MPa, respectively. The reduction in YS and UTS is mainly due to the coarsening of the M23C6 precipitates. After long term aging, the volume fraction of the M23C6 phase reached a plateau and its maximum internal stress was reduced, implying that under large internal stresses the carbides were more susceptible to fracture or decohesion from the matrix. Finally, the calculated dislocation densities were in good agreement with transmission electron microscopy (TEM) measurements. The square roots of the dislocation densities and the true stresses displayed typical linear behavior and no significant change was observed in the alloys in different aging conditions.

This paper studies the regulatory condition specified in the U.S. Nuclear Regulatory Commission's (NRC) NUREG-2215 review guideline, which does not allow the reflooding of previously dried high-burnup spent nuclear fuel (HBF), to assess the applicability of this condition for the retrieval process of HBF during the operation of indoor dry storage facilities in Taiwan, as well as its applicability to other light water reactor HBF in similar scenarios. By referencing NUREG CR-7198, revision 1, NUREG-2224, the NRCs responses to public comments, and research findings from the test plan for U.S. high burnup dry storage research project (HDRP), it explores the regulatory basis for not allowable reflooding of the previously dried HBF. Findings indicate that intact HBF can be subjected to at least one thermal cycle exceeding 65 °C without compromising the safety margin of the cladding during storage or transportation design-basis drop accidents. Recent license amendment reviews indicate that the NRC has lifted the reflooding restriction; however, HBF which has been stored for over 20 years requires additional analyses per NUREG-2224. Based on the assessment of the U.S.-Taiwan regulatory background, HBF characteristics related to radial hydride formation (including burnup levels, cladding alloy materials, maximum cladding temperature, hydride reorientation threshold stress, and number of thermal cycling), retrieval facilities, and retrieval operations, this study confirms that the regulatory condition in NUREG-2215 and methods recommended in NUREG-2224 are applicable to Taiwan's indoor dry storage facilities and other light water reactor HBF.

This paper studies the regulatory condition for reflooding of the previously dried high burnup fuel is not allowable in the NUREG-2215 review guidance issued by the NRC, to evaluate the regulatory applicability of this condition to the retrieval operations in Taiwan’s indoor dry storage facilities.

The discussion content firstly explains the research background and purpose of this paper, and then introduces NUREG-2215. And refer to NUREG CR-7198, Revision 1, NUREG-2224 and other research reports, as well as NRC’s response to public comments, to discuss the regulatory basis for reflooding of the previously dried high burnup fuel is not allowable. And from the NUREG-2224 evaluation results, it is known that intact high burnup fuel can withstand at least one thermal cycle exceeding 65°C without affecting the safety analysis of transportation or storage based on design basis accidents. The latest NRC review operation has also eliminated the restriction of reflooding, but high burnup fuel stored for more than 20 years needs to undergo additional supplementary analysis according to the recommended method.

Subsequently, the feasibility of applying this regulatory condition to Taiwan is evaluated based on items such as the regulatory situation, characteristics of spent nuclear fuel cladding, retrieval facilities, and retrieval operations.