• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.7 no.1
• /
• pp.49-62
• /
• 2009

Social Acceptance is important to decide policy of spent nuclear fuel management. The idea of a public sphere as a receptacle of dynamic process is the core in this discussion. The purpose of this study is to examine the concept, participants, the conflict-structure and agreeable conditions of a public sphere. A public sphere means in this paper, mechanism and systems that various stakeholders' and public's participation with spontaneous will can affect decision-making process. For good designing and implementing a public sphere, it is necessary to analysis and cope with political, foreign and security, economic, sociocultural environments, the law and systems around spent nuclear fuel management.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.4 no.3
• /
• pp.285-299
• /
• 2006

At the end of 2005, 443 nuclear reactors were operating in 32 countries worldwide. They had provided about 3,000 TWh, which was just over 16 percent of global electricity supply. With the generating capacity of 368 GWe in 2004, the spent fuel generation rate worldwide, now becomes at about 11,000 tHM/y. Projections indicate that cumulative amounts to be generated by the year 2020, the time when most of the existing NPP will be closed to the end of their licensed lifetime, may be close to 445,000 tHM. In this regard, spent fuel management is a common issue in all countries with nuclear reactors. Whatever their national policy and/or strategy is selected for the backend of the nuclear fuel cycle, the management of spent fuel will contribute an impending and imminent issues to be resolved in the foreseeable future. The 2nd Review Meeting of the Contracting Parties to the Joint Convention was held in Vienna from 15 to 24 May 2006. The meeting gave an opportunity to exchange information on the national policy and strategy of spent fuel management of the Contracting Parties, to discuss their situations, prospects and the major factors influencing the national policies in this field and to identify the most important directions that national efforts and international co-operation in this area should be taken. In this paper, an overview of national and global trends of spent fuel management is discussed. In addition, some directions are identified and recent activities of each Member States in the subject area are summarized.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.7 no.4
• /
• pp.229-236
• /
• 2009

There are two types of spent fuels generated from nuclear power plants, CANDU type and PWR type. PWR spent fuels which include a lot of reusable material can be considered to be recycled. CANDU spent fuels are considered to directly disposed in deep geological formation, since they have little reusable material. In this study, based on the Korean Reference spent fuel disposal System(KRS) which is to dispose both PWR and CANDU spent fuels, the more effective CANDU spent fuel disposal systems have been developed. To do this, the disposal canister has been modified to hold the storage basket which can load 60 spent fuel bundles. From these modified disposal canisters, the disposal systems to meet the thermal requirement for which the temperature of the buffer materials should not be over $100^{\circ}C$ have been proposed. These new disposals have made it possible to introduce the concept of long tenn storage and retrievabililty and that of the two-layered disposal canister emplacement in one disposal hole. These disposal concepts have been compared and analyzed with the KRS CANDU spent fuel disposal system in terms of disposal effectiveness. New CANDU spent fuel disposal concepts obtained in this study seem to improve thermal effectiveness, U-density, disposal area, excavation volume, and closure material volume up to 30 - 40 %.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.5 no.1
• /
• pp.53-64
• /
• 2007

The dynamic release behavior of Cs from high burn-up spent PWR fuel was experimentally performed under the conditions of a thermal treatment process such as voloxidation and sintering conditions. In voloxidation process, influence of the oxidation and reduction atmosphere on the Cs release characteristic using fragment type of spent fuel heated up to $1,500^{\circ}C$ was compared. In sintering process, temperature history effect on Cs release behavior was evaluated using green pellet under 4% $H_2/Ar$ environment. Temperature range for complete Cs release from spent fuel fragment under voloxidation condition was about $800^{\circ}C{\sim}1,200^{\circ}C$, but that of green pellet under the reduction atmosphere was $1,100^{\circ}C{\sim}1,400^{\circ}C$. Key parameters on Cs release behavior from spent fuel was powder formation as well as the diffusion rate of Cs compound to grain boundary and fuel surface.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.4 no.1
• /
• pp.87-93
• /
• 2006

Inventories, and characteristics such as dimension, fuel rod array, weight, $^{235}U$ enrichment, and discharge burnup of spent nuclear fuel (SNF) generated from existing and planed nuclear power plants based on National 2nd Basic Plan for Electric Power Demand and Supply were investigated and projected to support geological disposal system design. The historical and projected inventory by the end 2057 is expected to be 20,500 and 14,800MTU for PWR and CANDU spent nuclear fuel, respectively. The quantity of SNF with initial $^{235}U$ enrichment of 4.5 wt.% and below was shown to be 96.5% in total. Average burnup of SNF revealed $\sim36$ GWD/MTU and $\sim40$ GWD/MTU for the period of 1994-1999 and 2000-2003, respectively. It is expected that the average burnup of SNF will be $\sim45$ GWD/MTU at the end of 2000's. From the comprehensive study, it was concluded that the imaginary SNF with $16\times16$ Korean Standard Fuel Assembly, cross section of $21.4cm\times21.4cm$, length of 453cm, mass of 672 kg, initial $^{235}U$ enrichment of 4.5 wt.%, discharge burnup of 55 GWD/MTU could cover almost all SNFs to be produced by 2057.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.6 no.2
• /
• pp.119-128
• /
• 2008

Thermal assessment of a new CANDU spent fuel disposal system, which improves the retrievability of the spent fuel and enhances the densification factor compared with the Korean Reference disposal System, is carried out in this study. The canisters for CANDU spent fuels are stored for long term and cooled by natural convection in the proposed disposal system for the retrievability. The steady state thermal analyses for proposed CANDU disposal system are carried out with the ANSYS 10.0 CFX code. The thermal analyses are performed through two steps. At the first step, the sensitivity of the disposal tunnel spacing is analysed. The differences of maximum temperatures by several tunnel spacings are calculated at three points in the disposal tunnel. The result shows that the differences of the temperature at the three points are almost negligible because 99% of the decay heat is removed by natural convection. At the second procedure, 60m tunnel spacing with a ventilation system instead of natural convection is considered. The result is applied to the calculation of the canister surface temperature in disposal tunnel as boundary conditions. Consequently, the average and the maximum surface temperature of disposal canisters are $79.9^{\circ}C$ and $119^{\circ}C$, respectively. The inner maximum temperature of a basket in the disposal canister is calculated as $140.9^{\circ}C$. The maximum temperature of the basket meets the thermal requirement for the CANDU spent fuel cladding.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.2 no.3
• /
• pp.155-164
• /
• 2004

Advanced spent fuel Conditioning Process(ACP) project which is under development for efficient spent fuel management has finished process feasibility study and is preparing $\alpha$-${\gamma}$ type hot cell construction for process experimentation. Radiation dose evaluation for the radioactive nuclides were preliminarily performed for normal operation and accident case with the basic concept design report, the meteorological data and the recent site specific data. According to the production and release rate of nuclides, dose evaluations for residents around facility were performed. The evaluation result shows a safe margin for regulation limits and SAR(Safety Analysis Report) limit of IMEF(Irradiated Material Examination Facility) where this facility will be constructed.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.4 no.3
• /
• pp.245-253
• /
• 2006

A test of the 1/3 scale model was conducted to verify the tip-over analysis of a dry. concrete storage cask under a hypothetical accident condition. The tip-over analysis was executed using the velocity at each point as the initial conditions of the model just before the impact. The initial velocity was determined from the initial angular velocity, which would make the equivalent kinetic energy to the potential energy. To confirm the structural integrity of the canister, the visual testing and the non-detective testings such as Liquid Penetrant testing and Ultrasonic Testing were conducted. The lid of a storage cask was plastically deformed near the impact point. The structural integrity of storage cask was maintained. To verify the tip-over analysis the strains and the accelerations acquired by the tip-over test were compared with those by the analyses. The results of the analysis were larger than the test results about two times.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.5 no.4
• /
• pp.351-357
• /
• 2007

The Japanese Government selected Mutsu, Aomori Prefecture as a provisional spent-fuel repository site. This comes as a result of the prefecture's five-year campaign to host the site since 2000. Korea stores spent nuclear fuel within sites of nuclear power plants, and expects the storage capacity to reach its limit by the year 2016. This compels Korea to learn the cases of Japan. Having successfully hosted Gyeongju as a site for low-to-intermediate-level nuclear waste repository, Korea has already learned the potential process of hosting spent fuel storage site. The striking difference between the two countries in the process of hosting the site is that the Korean government had to offer the local city a large amount of subsidy for hosting through competitive citizens' referendum among candidate cities while it was the leadership of the local municipality that enabled the controversial decision in Japan. It is also a distinguishable characteristics of Japan that not a huge subsidy is provided to the local host city. I hope this study offers an idea to Korea's future effort to select a spent-fuel host site.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.13 no.2
• /
• pp.141-154
• /
• 2015

In general, conventional criticality analysis for spent fuel transport/storage systems have been performed based on the assumption of fresh fuel concerning the potential uncertainties from number density calculations of actinide nuclides and fission products in spent fuel. However, these evaluation methods cause financial losses due to an excessive criticality margin. In order to overcome this disadvantage, many studies have recently been conducted to design and commercialize a transportation and storage cask applied to the Burnup Credit (BUC). This study conducted an assessment to ensure criticality safety for reactor operating parameters, axial burn-up profiles and misload accident conditions, which are the factors that are likely to affect criticality safety when the BUC is applied to the dual-purpose cask under development at the KOrea RADioactive waste agency (KORAD). As a result, it was found that criticality resulting from specific power, changed substantially and relied on conditions of low enrichment and high burn-up. Considering the end effect in the case of high burn-up produced a positive-definite result. In particular, the increment of maximum effective multiplication factors due to misloading was 0.18467, confirming that misload is a factor that must be taken into account when applying the BUC. The results of this study may therefore be utilized as references in developing technologies to apply the BUC to domestic models and operational procedures or preventing any misload accidents during the process of spent fuel loading.