• Nuclear Engineering and Technology
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• 제45권1호
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• pp.29-40
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• 2013

Two different kinds of nuclear power plants produce a substantial amount of spent fuel annually in Korea. According to the current projection, it is expected that around 60,000 MtU of spent fuel will be produced from 36 PWR and APR reactors and 4 CANDU reactors by the end of 2089. In 2006, KAERI proposed a conceptual design of a geological disposal system (called KRS, Korean Reference disposal System for spent fuel) for PWR and CANDU spent fuel, as a product of a 4-year research project from 2003 to 2006. The major result of the research was that it was feasible to construct a direct disposal system for 20,000 MtU of PWR spent fuels and 16,000 MtU of CANDU spent fuel in the Korean peninsula. Recently, KAERI and MEST launched a project to develop an advanced fuel cycle based on the pyroprocessing of PWR spent fuel to reduce the amount of HLW and reuse the valuable fissile material in PWR spent fuel. Thus, KAERI has developed a geological disposal system for high-level waste from the pyroprocessing of PWR spent fuel since 2007. However, since no decision was made for the CANDU spent fuel, KAERI improved the disposal density of KRS by introducing several improved concepts for the disposal canister. In this paper, the geological disposal systems developed so far are briefly outlined. The amount and characteristics of spent fuel and HLW, 4 kinds of disposal canisters, the characteristics of a buffer with domestic Ca-bentonite, and the results of a thermal design of deposition holes and disposal tunnels are described. The different disposal systems are compared in terms of their disposal density.

• Nuclear Engineering and Technology
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• 제17권3호
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• pp.185-192
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• 1985

가압경수로에서 나오는 사용후 핵연료의 fissile 양은 CANDU형 원자로에 쓰는 천연우라늄의 농축도 보다 높다. 따라서 핵연료 활용을 다양화하고 점차 누적되고 있는 가압경수로의 사용후 핵 연료의 저장문제를 부분적으로나마 해결하기 위하여, 가압경수로의 사용후 책 연료를 CANDU 형 원자로에 사용하는 방안을 검토 하였다. 가압경수로에서 나온 사용후 핵 연료에서 가공되는 혼합핵연료(Mixed Oxide Fuel)를 CANDU형 원자로에 장전하였을 경우, WIMS/D 코드를 이용하여 핵적특성을 분석하였다. 그리고 본 분석에서는 현 CANDU형 원자로의 반응도 조절장치를 변경시키지 않고 혼합핵 연료를 CANDU형 원자로에 사용할 수있는 방안만 조사하였다.

• 방사선산업학회지
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• 제18권3호
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• pp.255-266
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• 2024

The uranium nuclear fuel used in nuclear power generation needs to be replaced with new fuel after a certain period. In South Korea, the spent nuclear fuel generated during this process is temporarily stored within the nuclear power plant site, and there are ongoing issues with the saturation of storage capacity. To address these problems, the South Korea government has established a plan to manage high-level radioactive waste, including provisions for securing interim storage facilities. An interim storage facility is designed to safely store spent nuclear fuel for certain period before its permanent disposal. This study analyzed leading international cases of interim storage facilities that are technically feasible and can reduce the operating period of temporary storage facilities for spent nuclear fuel within nuclear power plant sites. It also presented the technical concepts required for the operation of interim storage facilities for spent fuel from PWR(Pressurized Water Reactor), reflecting the situation in South Korea.

• Nuclear Engineering and Technology
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• 제40권6호
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• pp.517-526
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• 2008

The widespread concern regarding the management of spent fuel that mainly contributes to nuclear waste has led to the development of the sodium-cooled fast reactor (SFR) as one of the most promising future types of reactors at both national and international levels. Various reactor deployment scenarios with SFR introductions with different conversion ratios in the existing PWR-dominant nuclear fleet have been assessed to optimize the SFR deployment strategy to replace PWRs with the view toward a reduction in the level of spent fuel as well as efficient uranium utilization through its reuse in a closed fuel cycle. An efficient reactor deployment strategy with the SFR introduction starting in 2040 has been drawn based on an SFR deployment strategy in which burners are deployed prior to breakeven reactors to reduce the amount of PWR spent fuel substantially at the early deployment stage. The PWR spent fuel disposal is reduced in this way by 98% and the cumulative uranium demand for PWRs to 2100 is projected to be 445 ktU, implying a uranium savings of 115 ktU. The SFR mix ratio in the nuclear fleet near the year 2100 is estimated to be approximately 35-40%. PWRs will remain as a main power reactor type until 2100 and SFRs will support waste minimization and fuel utilization.

• Nuclear Engineering and Technology
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• 제54권12호
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• pp.4499-4513
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• 2022

Spent fuel behavior of dry storage was simulated in a continuous state from steady-state operation by modifying FRAPCON-4.0 to incorporate spent fuel-specific fuel behavior models. Spent fuel behavior of a typical PWR was compared with that of NuScale Power Module (NPM^TM). Current PWR discharge burnup (60 MWd/kgU) gives a sufficient margin to the hoop stress limit of 90 MPa. Most hydrogen precipitation occurs in the first 50 years of dry storage, thereby no extra phenomenological safety factor is identified for extended dry storage up to 100 years. Regulation for spent fuel management can be significantly alleviated for LWR-based SMRs. Hydride embrittlement safety criterion is irrelevant to NuScale spent fuels; they have sufficiently lower plenum pressure and hydrogen contents compared to those of PWRs. Cladding creep out during dry storage reduces the subchannel area with burnup. The most deformed cladding outer diameter after 100 years of dry storage is found to be 9.64 mm for discharge burnup of 70 MWd/kgU. It may deteriorate heat transfer of dry storage by increasing flow resistance and decreasing the view factor of radiative heat transfer. Self-regulated by decreasing rod internal pressure with opening gap, cladding creep out closely reaches the saturated point after ~50 years of dry storage.

• Nuclear Engineering and Technology
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• 제52권3호
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• pp.614-620
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• 2020

In a previous study, delayed hydride cracking (DHC) assessment of pressurized water reactor (PWR) spent fuel during dry storage using the threshold stress intensity factor (K_IH) was performed. However, there were a few limitations in the analysis of the cladding properties, such as oxide thickness and mechanical properties. In this study, those models were modified to include test data for irradiated materials, and the cladding creep model was introduced to improve the reliability of the DHC assessment. In this study, DHC susceptibility of PWR spent fuel during dry storage depending on the axial elevation was evaluated with the improved assessment methodology. In addition, the sensitivity of affecting parameters such as fuel burnup, hydride thickness, and crack aspect ratio are presented.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2001년도 춘계학술발표회요약집
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• pp.371.1-371
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• 2001

• 방사성폐기물학회지
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• 제20권2호
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• pp.231-241
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• 2022

Several countries, including Korea, are considering the direct disposal of spent nuclear fuels. The radiological safety assessment results published after a geological repository closure indicate that the instant release is the main radiation source rather than the congruent release. Three Safety Case reports recently published were reviewed and the IRF values of seven long-lived radionuclides, including relevant experimental results, were compared. According to the literature review, the IRF values of both the CANDU and low burnup PWR spent fuel have been experimentally measured and used reasonably. In particular, the IRF values of volatile long-lived nuclides, such as ¹²⁹I and ¹³⁵Cs, were estimated from the FGR value. Because experimental leaching data regarding high burnup spent nuclear fuels are extremely scarce, a mathematical modelling approach proposed by Johnson and McGinnes was successfully applied to the domestic high burnup PWR spent nuclear fuel to derive the IRF values of iodine and cesium. The best estimate of the IRF was 5.5% at a discharge burnup of 55 GWd tHM^-1.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1995년도 추계학술발표회논문집(1)
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• pp.109-114
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• 1995

A preliminary study of the heterogeneity effect of spent P% fuel in CANDU was made using a reduced spent PWR fuel data base. The instantaneous core simulation has shown that the refueling ripple in the CANDU reactor is large if the spent PWR fuel is directly used. But the fuel heterogeneity effect can be reduced appreciably by blending spent PWR fuel with a small amount of fresh UO$_2$. The refueling simulation has shown that the operating margins of 6.0% and 8.7% are achievable for the peak channel and bundle powers, respectively, with the blended fuel.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1998년도 춘계학술발표회논문집(2)
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• pp.470-475
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• 1998

The marginal nuclear criticality analysis for the high density spent fuel storage at a PWR plant was carried out by using the HELIOS and CASMO-3 codes. More than 20 % of the calculated reactivity saving effect is observed in this analysis. This mainly comes from the adoption of some important fission products and B-10 in the criticality analysis. By taking burnup and boron credits, the high capacity of the spent fuel storage rack can be more fully utilized, reducing the space of storage. Larger storage for a given inventory of spent fuel should result in remarkable cost savings and mort importantly reduce the risks to the public and occupational workers.