[en] In order to provide basic data for selecting candidate sites for high-level radioactive waste geological disposal, the drilling project is being carried out to investigate the deep environment for each type of bedrock existing on the Korean Peninsula. Securing 10 boreholes with a depth of 750 m is the final goal of the drilling project, which is planned for 5 years from 2020 to 2024. Two boreholes are being secured for sedimentary, plutonic, volcanic, metamorphic rock, respectively, and two additional boreholes will be secured for the rock types that require further investigation. As of December 2022, three boreholes were secured for sedimentary rocks, two for plutonic rocks, and one for metamorphic rocks. The boreholes were targeted for shale, granite and gneiss, respectively. To provide precise and reliable basic data, multidisciplinary (geological, geophysical, geochemical, hydrogeological, rock mechanics, etc.) studies have been conducted using drilling holes and cores. These research results are expected to be used as important basic data for selecting candidate sites for high-level radioactive waste geological disposal in Korea.

[en] Eddy current test(ECT) is used to inspect not only the failed fuel rods but also peripheral rods during repairing of the failed fuel rods, to detect internal defects in irradiated fuel rods which could not be detected by ultrasonic test and visual test, and to obtain the data for determining the root cause of fuel rod failure. This study evaluates the effect of properties of test article, irradiated fuel rods, on the impedance diagram in order to reduce the difficulty of ECT signal analysis. The optimum eddy current probe design conditions for inspecting the irradiated fuel rods, is estimate by using experimental equations and the probe is manufactured based on the estimated conditions. The performance of developed eddy current probe and the optimum conditions is proved through characteristic comparison experiment with the probe purchased from the foreign vendor

[en] Korea Nuclear Fuel Co., Ltd. (KNFC) is supplying Westinghouse 14 x 14, 16 x 16 and 17 x 17 type fuels and W-CE 16 x 16 type fuels to 16 PWR power plants and also 37-bundle type PHWR fuel to 4 CANDU plant in Korea. KNFC is developing high burnup, high power and high integrity fuel in order to improve plants efficiency, safety and economics. KNFC developed fuel repair technology to repair failed fuels such as Westinghouse 14 x 14, 16 x 16 and 17 x 17 type fuels and W-CE 16 x 16 type fuels. And also developed pool side fuel inspection technology to examine irradiation performance of existing as well as on-developing fuels. For the first step, single rod inspection technology to measure oxide layer thickness, fretting wear depth, diameter and defect measurements of single rod was developed and used for irradiation performance examination of Zirlo clad rods of one cycle burned Plus7 LTA (Lead Test Assembly, similar to WCE-type fuel assemblies) in October 2002 and second cycle burned Plus7 LTA in May 2004 at YGN4. For second step, fuel assembly inspection and function test technologies such as assembly bowing, twisting, and growing measurements, grid width and position measurement, peripheral rod growing and diameter and inside rod oxide layer thickness measurement have been developed and used for irradiation performance examination of two Plus-7 LTA, at UGN3 in April 2004. Now, KNFC is developing fuel ultrasonic cleaning technology to resolve axial offset anomaly that is occurred during operation of cycle in longer cycled core of PWR. (authors)

[en] In this study, the state-of-the art for estimating LERF is considered for the regulatory risk-informed decisions. The consideration is mainly focused on (1) the relationship between Level 2 PSA and LERF evaluation methodology, (2) the standard requirements in terms of modeling preparation and the acceptance criteria based on the application capability II of ASME PRA standard, and (3) some pending issues for developing and proposing a simplified LERF model. This study is preliminarily presented and will be updated for establishing detailed evaluation scheme of extended MPAS (multi-purpose probabilistic analysis of safety) model and preparing the technical basis

[en] An 'equivalent rectangular crack' approach was employed to predict rupture pressures and leak rates through laboratory generated stress corrosion cracks and steam generator tubes removed from the McGuire Nuclear Station.Specimen flaws were sized by post-test fractography in addition to a pre-test advanced eddy current technique.The predicted and observed test data on rupture and leak rate are compared.In general,the test failure pressures and leak rates are closer to those predicted on the basis of fractography than on nondestructive evaluation (NDE). However,the predictions based on NDE results are encouraging,particularly because they have the potential to determine a more detailed geometry of ligamented cracks,from which failure pressure and leak rate can be more accurately predicted.One test specimen displayed a time-dependent increase of leak rate under constant pressure

[en] This report summarizes models for the prediction of failure pressures and leak rates under normal operation and design-basis accident conditions in steam generator tubes with axial and circumferential cracks. These models were first validated through failure and leak rate tests at room temperature and at 282 C on tubes with rectangular, triangular, and trapezoidal notches fabricated by electrodischarge machining. They were then compared with failure and leak rate tests conducted on tubes with laboratory-generated outer-diameter stress corrosion cracks and steam generator tubes with field-induced stress corrosion cracks, which have highly complex morphology. Complex crack profiles are addressed using a model based on the concept of equivalent rectangular cracks. The predictions of the models are in reasonable agreement with test results, the time-dependent initiation and increase of leak rates observed in some tests cannot be predicted by the model

No abstract available

[en] An international Heated Crevice Seminar, sponsored by the Division of Engineering Technology, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, Argonne National Laboratory, and the Electric Power Research Institute, was held at Argonne National Laboratory on October 7-11, 2002. The objective of the seminar was to provide a working forum for the exchange of information by contributing experts on current issues related to corrosion in heated crevices, particularly as it relates to the integrity of PWR steam generator tubes. Forty-five persons from six countries attended the seminar, including representatives from government agencies, private industry and consultants, government research laboratories, nuclear vendors, and electrical utilities. The seminar opened with keynote talks on secondary-side crevice environments associated with IGA and IGSCC of mill-annealed Alloy 600 steam generator tubes and the submodes of corrosion in heat transfer crevices. This was followed by technical sessions on (1) Corrosion in Crevice Geometries, (2) Experimental Methods, (3) Results from Experimental Studies, and (4) Modeling. The seminar concluded with a panel discussion on the present understanding of corrosive processes in heated crevices and future research needs

[en] In this study, we aimed at the development of fabrication technology and property evaluation of triplex SiC composite tube, the evaluation and assessment of SiC corrosion behavior under PWR water conditions, and the development of SiC joining technology for end-plug joining. Multi-layered SiC composite tubes were fabricated by CVD and CVI methods for the application of PWR nuclear fuel cladding. High-quality and uniform SiC triplex composite tubes could be fabricated and the SiC phase was highly stoichiometric without any presence of free silicon or free carbon. The triplex composite tubes had hoop strengths of 250-300 MPa and Weibull modulus of 11.05, a high degree of uniformity in the chemical and mechanical properties. The dissolved hydrogen significantly affected the corrosion behavior of SiC under PWR water condition, which was first observed in this study. SiC plates were diffusion bonded using a hot press method, pressureless joining method, and laser beam scanning method. It was confirmed that the joint property was promising with the torsional strength of about 100 MPa, in which the fracture occurred at the base material instead of the joint region.

No abstract available