Collaborative R&D Work Between Korea and USA for Cooling Rate Effect of Hydride Reorientation

Donghak Kook,Hongryoul Oh,Daeho Kim,Yanghyun Koo 한국방사성폐기물학회 2022 한국방사성폐기물학회 학술논문요약집 Vol.20 No.1

Laboratory testing to simulate the drying of spent fuel is most often done using a cooling rate of approximately 5°C per hour because there are so many restricted test conditions like R&D project duration limit, budget and temporary electronic supply blackout at laboratory building. However, in a real dry cask storage system, the fuel cools much slower. Early data from KAERI on unirradiated, pre-hydrided cladding has shown that slower cooling may result in more brittle behavior than is currently observed based on these short-term tests. Given the potential safety and future handling implications of failed fuel, it is important to determine if the material properties of spent fuel cladding measured in these laboratory tests are the same as would be observed on fuel that has undergone a much longer, slower cooling, which may provide more time for hydrides to precipitate in the radial direction. KAERI and PNNL have started a collaborative I-NERI R&D project on this topic and each organization will perform tests on unirradiated & irradiated cladding under various hoop stress and cooling rate combinations. Scope of collaborative work is to evaluate long-term cooling (slow cooling rate) on hydride reorientation and subsequent material properties of cladding to determine if past and current research activities on spent nuclear fuel are bounding. The results will be used to direct future testing and help predict cladding performance over a wide range of burnups during extended storage and transportation.

Current Status of Spent Nuclear Fuel Integrity Evaluation R&D in the Major Countries

Donghak Kook,Jaeho Yang,Yanghyun Koo 한국방사성폐기물학회 2018 한국방사성폐기물학회 학술논문요약집 Vol.2018 No.10

Stepwise Approach on Spent Nuclear Fuel Integrity Evaluation R&D Work in Korea

Donghak Kook,Jaeyong Kim,Yongsik Yang,Changwhan Shin 한국방사성폐기물학회 2017 한국방사성폐기물학회 학술논문요약집 Vol.2017 No.05

REVIEW OF SPENT FUEL INTEGRITY EVALUATION FOR DRY STORAGE

Kook, Donghak,Choi, Jongwon,Kim, Juseong,Kim, Yongsoo Korean Nuclear Society 2013 Nuclear Engineering and Technology Vol.45 No.1

Among the several options to solve PWR spent fuel accumulation problem in Korea, the dry storage method could be the most realistic and applicable solution in the near future. As the basic objectives of dry storage are to prevent a gross rupture of spent fuel during operation and to keep its retrievability until transportation, at the same time the importance of a spent fuel integrity evaluation that can estimate its condition at the final stage of dry storage is very high. According to the national need and technology progress, two representative nations of spent fuel dry storage, the USA and Japan, have established different system temperature criteria, which is the only controllable factor in a dry storage system. However, there are no technical criteria for this evaluation in Korea yet, it is necessary to review the previously well-organized methodologies of advanced countries and to set up our own domestic evaluation direction due to the nation's need for dry storage. To satisfy this necessity, building a domestic spent fuel test database should be the first step. Based on those data, it is highly recommended to compare domestic data range with foreign results, to build our own criteria, and to expand on evaluation work into recently issued integrity problems by using a comprehensive integrity evaluation code.

Preliminary Evaluation of Cr-coated ATF Spent Fuel Integrity for Dry Storage and Transportation

Donghak Kook,Yong-sik Yang 한국방사성폐기물학회 2021 한국방사성폐기물학회 학술논문요약집 Vol.2021 No.11

PYROPROCESS WASTE DISPOSAL SYSTEM DESIGN AND DOSE CALCULATION

DONGHAK KOOK,DONG-KEUN CHO,Min-Soo Lee,JONGYOUL LEE,HEUI-JOO CHOI,김용수 한국원자력학회 2012 Nuclear Engineering and Technology Vol.44 No.5

PWR spent fuels produced in the Republic of Korea are expected to be recycled by pyroprocess in the long term future. Even though pyroprocess waste amounts can be smaller than that of PWR spent fuel assembly in case of direct disposal, this process essentially will produce various and unique radioactive wastes. The goals of this article are to characterize these wastes, calculate the amount of wastes, design disposal systems for each waste and evaluate the radiation safety of each system by dose assessment. The absorbed dose results of the metal and ceramic waste for the engineering barrier system (EBS) showed 2.21 x 10-2 Gy/h and 1.15 x 10-2 Gy/h, which are lower than the recommended value of 1 Gy/h. These results confirmed that the newly proposed disposal systems have a safety margin for the radiation produced from each waste.

WASTE CLASSIFICATION OF 17X17 KOFA SPENT FUEL ASSEMBLY HARDWARE

조동건,DONGHAK KOOK,JONGWON CHOI,HEUI-JOO CHOI 한국원자력학회 2011 Nuclear Engineering and Technology Vol.43 No.2

Metal waste generated from the pyroprocessing of 10 MtU of spent fuel was classified by comparing the specific activityof a relevant radionuclide with the limit value of the specific activity specified in the Korean acceptance criteria for a low-and intermediate-level waste repository. A Korean Optimized Fuel Assembly design with a 1717 array, an initial enrichmentof4.5 weight-percent, discharge burn-up of 55 GWD/MtU, and a 10-year cooling time was considered. Initially, the mass andvolume of each structural component of the assembly were calculated in detail, and a source term analysis was subsequentlyperformed using ORIGEN-S for these components. An activation cross-section library generated by the KENO-VI/ORIGEN-S module was utilized for top-end and bottom-end pieces. As a result, an Inconel grid plate, a SUS plenum spring, a SUSguide tube subpart, SUS top-end and bottom-end pieces, and an Inconel top-end leaf spring were determined to be unacceptablefor the Gyeongju low- and intermediate-level waste repository, as these waste products exceeded the acceptance criteria. Incontrast, a Zircaloy grid plate and guide tube can be placed in the Gyeongju repository. Non-contaminated Zircaloy claddingoccupying 76% of the metal waste was found to have a lower level of specific activity than the limit value. However, Zircaloycladding contaminated by fission products and actinides during the decladding process of pyroprocessing was revealed tohave 52 and 2 times higher specific activity levels than the limit values for alpha and 90Sr, respectively. Finally, it was foundthat 88.7% of the metal waste from the 1717 Korean Optimized Fuel Assembly design should be disposed of in a deepgeological repository. Therefore, it can be summarized that separation technology with a higher decontamination factor fortransuranics and strontium should be developed for the efficient management of metal waste resulting from pyroprocessing.

Suggestion of Simple Rod Internal Pressure and Free Void Volume Correlations in Spent Fuel Rod

Yongsik Yang,Donghak Kook 한국방사성폐기물학회 2022 한국방사성폐기물학회 학술논문요약집 Vol.20 No.2

A rod internal pressure increased by fission gas release is major factor that causes degradation during dry storage of spent fuel. Because rod internal pressure is greatly affected by fuel design, operation power history, it is essential to perform complex calculation using performance code to accurately predict rod internal pressure as function of burnup. However, because it is difficult to apply a complex method into dry storage design and to determine rod internal pressure based on conservative way this study presents a simple correlation that can predict an approximate rod internal pressure as function of burnup For the development of simple correlation, rod internal pressure and fuel rod void volume data measured through about 400 PIE (Post Irradiation Examination) data were used. The developed simple correlation can cover various fuel rod arrays, discharged fuel average burnup, operation history, cladding type, burnup range, and information on Westinghouse type fuel rods such as Spain ENUSA, USA EPRI/ANL/ORNL/PNNL, WEC, etc. In this paper, the data of simple correlation determination is briefly introduced, and the data analysis process and results are summarized. Two correlations that can conservatively determine rod internal pressure and free void volume in fuel rod according to fuel rod average burnup were presented, and the effect of initial He fill pressure was evaluated. In particular, the results of Post Irradiation Examination for 46 fuel rods conducted in Korea are also included, so it is expected that newly presented correlations can be used easily in various ways in the domestic research, industry, and academia.

A STUDY ON THE INITIAL CHARACTERISTICS OF DOMESTIC SPENT NUCLEAR FUELS FOR LONG TERM DRY STORAGE

Kim, Juseong,Yoon, Hakkyu,Kook, Donghak,Kim, Yongsoo Korean Nuclear Society 2013 Nuclear Engineering and Technology Vol.45 No.3

During the last three decades, South Korean nuclear power plants have discharged about 5,950 tons of spent fuel and the maximum burn-up reached 55 GWd/MTU in 2002. This study was performed to support the development of Korean dry spent fuel storage alternatives. First, we chose V5H-$17{\times}17$ and KSFA-$16{\times}16$ as representative domestic spent fuels, considering current accumulation and the future generation of the spent fuels. Examination reveals that their average burn-ups have already increased from 33 to 51 GWd/MTU and from 34.8 to 48.5 GWd/MTU, respectively. Evaluation of the fuel characteristics shows that at the average burn-up of 42 GWd/MTU, the oxide thickness, hydrogen content, and hoop stress ranged from $30{\sim}60{\mu}m$, 250 ~ 500 ppm, and 50 ~ 75 MPa, respectively. But when burn-up exceeds 55 GWd/MTU, those characteristics can increase up to 100 ${\mu}m$, 800 ppm, and 120 MPa, respectively, depending on the power history. These results demonstrate that most Korean spent nuclear fuels are expected to remain within safe bounds during long-term dry storage, however, the excessive hoop stress and hydrogen concentration may trigger the degradation of the spent fuel integrity early during the long-term dry storage in the case of high burn-up spent fuels exceeding 45 GWd/MTU.

Experimental Technology of Long-term Cooling Effect for Hydride Reorientation Test of Non-irradiated Zircaloy-4 Cladding

Hongryoul Oh,Daeho Kim,Donghak Kook 한국방사성폐기물학회 2022 한국방사성폐기물학회 학술논문요약집 Vol.20 No.1

A long-term cooling effect on hydride reorientation of a cladding tube can affect the integrity of spent nuclear fuel transportation and long-term storage. In this study, experimental setup for investigating the degree of radial reorientation of hydrides in the circumferential direction during the long-term cooling was established. The experimental setup was designed to be simplified since the long-term evaluation requires a long term period such as 12, 18 and 24 months when the cladding tube specimen is gradually cooled down from 400°C to 100°C. For the test, hydrogen-charged specimens of 100 ppm, 200 ppm, and 500 ppm were prepared. The specimen was sealed with fixtures and check valve, and was pressurized up to 90 Mpa. To heat the specimen, a box-type furnace was used while the temperature of the specimen was measured from thermocouples attached to the specimen. After the heat treatment, the long-term cooling was performed by developing temperature control program to investigate several cooling rate conditions of the specimen. As a reference case, microstructure and brittle property of the hydrogen-charged specimens of 100 ppm, 200 ppm, and 500 ppm without the long-term cooling was observed. In the case of the hydrogen content, it was uniformly distributed in circumferential direction although it was non-uniform in the axial direction. In the case of the brittle property, a compression test was performed. For the future work, the microstructure and brittle property of the hydrogencharged specimens after the several long-cooling conditions were investigated. Then, the degree of radial reorientation of hydrides in the circumferential direction during the long-term cooling was studied.