Ki-g Seo

HEAT REMOVAL TEST USING A HALF SCALE STORAGE CASK

Spent nuclear fuel generated at nuclear power plants must be safely stored during interim storage periods. A dry storage cask to safely store the spent nuclear fuel should be able to adequately emit the decay heat from the spent nuclear fuel. Therefore, heat removal tests using a half scale dry storage cask have been performed to estimate the heat transfer characteristics of a dry storage cask under normal, off-normal, and accident conditions. In the normal condition, the heat transfer rate to an ambient atmosphere by convective air through a passive heat removal system reached 83%. Accordingly, the passive heat removal system is designed well and works adequately. In the off-normal condition, the influence of a half blockage in the inlet on the temperature appears minimal. In the accident condition, the temperature rose for 12 hours after the accident, but the temperature rise steadied after 36 hours.

Beneficial influence of the weldment rupture of cask impact limiter case on the impact absorbing behavior

In the cask impact limiter design, the functions of steel case should be evaluated for the protection of the filler materials and the impact energy absorption by the buckling deformation in both the fire and impact accidents. The objective of this paper is to evaluate the beneficial influence of the intermittent weldment of the cask impact limiter case on the cask impact behavior. This paper describes the test results for the joint strength evaluation of intermittent weldment and the cask impact analysis considering the weldment rupture of the impact limiter case. The weldment rupture of the impact limiter case causes to lose their constraining effect for the wood blocks, which are filled into the metal incasement between the case and the gussets. The weldment rupture of the impact limiter case reduces the impact force which acts on the impact target significantly in vertical and horizontal drop impacts. Therefore the beneficial effect of weldment rupture should be considered in the impact limiter design and the cask impact analysis.

IMPACT ANALYSES AND TESTS OF CONCRETE OVERPACKS OF SPENT NUCLEAR FUEL STORAGE CASKS

A concrete cask is an option for spent nuclear fuel interim storage. A concrete cask usually consists of a metallic canister which confines the spent nuclear fuel assemblies and a concrete overpack. When the overpack undergoes a missile impact, which might be caused by a tornado or an aircraft crash, it should sustain an acceptable level of structural integrity so that its radiation shielding capability and the retrievability of the canister are maintained. A missile impact against a concrete overpack produces two damage modes, local damage and global damage. In conventional approaches [1], those two damage modes are decoupled and evaluated separately. The local damage of concrete is usually evaluated by empirical formulas, while the global damage is evaluated by finite element analysis. However, this decoupled approach may lead to a very conservative estimation of both damages. In this research, finite element analysis with material failure models and element erosion is applied to the evaluation of local and global damage of concrete overpacks under high speed missile impacts. Two types of concrete overpacks with different configurations are considered. The numerical simulation results are compared with test results, and it is shown that the finite element analysis predicts both local and global damage qualitatively well, but the quantitative accuracy of the results are highly dependent on the fine-tuning of material and failure parameters.

ARISING TECHNICAL ISSUES IN THE DEVELOPMENT OF A TRANSPORTATION AND STORAGE SYSTEM OF SPENT NUCLEAR FUEL IN KOREA

In Korea, although the concept of dry storage system for PWR spent fuels first emerged in the early 1990s, wet storage inside nuclear reactor buildings remains the dominant storage paradigm. Furthermore, as the amount of discharged fuel from nuclear power plants increases, nuclear power plants are confronted with the problem of meeting storage capacity demand. Various measures have been taken to resolve this problem. Dry storage systems along with transportation of spent fuel either on-site or off-site are regarded as the most feasible measure. In order to develop dry storage and transportation system safety analyses, development of design techniques, full scale performance tests, and research on key material degradation should be conducted. This paper deals with two topics, structural analysis methodology to assess cumulative damage to transportation packages and the effects of an aircraft engine crash on a dual purpose cask. These newly emerging issues are selected from among the many technical issues related to the development of transportation and storage systems of spent fuels. In the design process, appropriate analytical methods, procedures, and tools are used in conjunction with a suitably selected test procedure and assumptions such as jet engine simulation for postulated design events and a beyond design basis accident.

Tritium Research Activities in KAERI

Abstract Korea has twenty nuclear power plants and a nuclear research reactor in operation. Out of the twenty plants, four are CANDU reactors at the Wolsong Nuclear Power Site. In the CANDU reactors, deuterium (heavy water) is used as a moderator and as the primary heat transport from the nuclear fuel. The nuclear research reactor, HANARO, in KAERI (Korea Atomic Energy Research Institute) uses heavy water as a neutron reflector. Tritium is formed by a neutron capture from the deuterium. If left to accumulate, tritium oxide would become a hazard to the operating staff and public. The primary purpose of a Tritium Removal Facility (TRF) is to reduce tritium concentration in a heavy water moderator. Operation of a TRF commenced at the Wolsong Nuclear Power Site on July 26th, 2007. Korea shared in the construction of the ITER fuel cycle plant with the EU, Japan, and US, and is responsible for the supply of an SDS (Tritium Storage and Delivery System). KAERI has been developing tritium technologies related to the Wolsong TRF, HANARO, and nuclear fusion fuel systems. We thus present details on the recent development status of the tritium systems.

Development of Tritium Transport Package for ITER Supply

For tritium supply to the fusion reactor of ITER (The Way to New Energy) [1], tritium need to be transported from tritium production sites, mainly the CANDU type reactor sites to the tritium plant building of ITER. Korea Atomic Energy Research Institute (KAERI) was commissioned the work of developing the transport package for tritium by ITER Organization and the first stage of the development has been just finished. The developed package was designed to carry 70 g of tritium and classified as a type B(U) package, which should comply with the requirements stipulated in IAEA Safety Standard Series [2]. The package is composed of a storage vessel, a containment vessel, an overpack and an aluminum liner which is a unique feature of the package. The aluminum liner between the storage vessel and the containment vessel is for containment control under the repetitive use of the package. The package has enough pressure resistance for 5 year in-site storage and the structural and thermal integrity under the hypothetical accident conditions has been demonstrated through a series of analyses.Copyright

Porous Media Modelling and Verification of Thermal Analysis for Inlet and Outlet Ducts of Spent Fuel Storage Cask

사용후핵연료 저장용기의 공기 흡입구 및 배기구에는 외부환경으로부터 이물질의 유입을 방지하기 위하여 bird screen이 설치되며, bird screen에서는 공기의 유동 저항이 발생하게 된다. 본 연구에서는 Bird screen mesh의 단순화 모델을 이용한 열해석을 수행하기 위하여 다공성매질 해석모델을 개발하였다. CFD 해석을 ...

Influence of Temperature on Chloride Ion Diffusion of Concrete

The long term integrity of concrete cask is very important for spent nuclear fuel dry storage system. However, there are serious concerns about early deterioration of concrete cask from creaking and corrosion of reinforcing steel by chloride ion because the cask is usually located in seaside, expecially by combined deterioration such as chloride ion and heat, carbonation. This study is to investigate the relation between temperature and chloride ion diffusion of concrete. Immersion tests using 3.5% NaCl solution that were controlled in four level of temperature, i.e. 20, 40, 65, and 90℃, were conducted for four months. The chloride ion diffusion coefficient of concrete was predicted based on the results of profiles of Cl- ion concentration with the depth direction of concrete specimens using the method of potentiometric titration by AgNO3. Test results indicate that the diffusion coefficient of chloride ion increases remarkably with increasing temperature, and there was a linear relation between the natural logarithm values of the diffusion coefficients and the reciprocal of the temperature from the Arrhenius plots. Activation energy of concrete in this study was about 46.6 (W/C = 40%), 41.7 (W/C = 50%), 30.7 (W/C = 60%) kJ/mol under a temperature of up to 90℃, and concrete with lower water-cement ratio has a tendency towards having higher temperature dependency.

Safety Evaluation of Radioactive Material Transport Package under Stacking Test Condition

Radioactive waste transport package was developed to transport eight drums of low and intermediate level waste(LILW) in accordance with the IAEA and domestic related regulations. The package is classified with industrial package IP-2. IP-2 package is required to undergo a free drop test and a stacking test. After free drop and stacking tests, it should prevent the loss or dispersal of radioactive contents, and loss of shielding integrity which would result in more than 20 % increase in the radiation level at any external surface of the package. The objective of this study is to establish the safety test method and procedure for stacking test and to prove the structural integrities of the IP-2 package. Stacking test and analysis were performed with a compressive load equal to five times the weight of the package for a period of 24 hours using a full scale model. Strains and displacements were measured at the corner fitting of the package during the stacking test. The measured strains and displacements were compared with the analysis results, and there were good agreements. It is very difficult to measure the deflection at the container base, so the maximum deflection of the container base was calculated by the analysis method. The maximum displacement at the corner fitting and deflection at the container base were less than their allowable values. Dimensions of the test model, thickness of shielding material and bolt torque were measured before and after the stacking test. Throughout the stacking test, it was found that there were no loss or dispersal of radioactive contents and no loss of shielding integrity. Thus, the package was shown to comply with the requirements to maintain structural integrity under the stacking condition.

Evaluation of Cumulative Damage From Drop to Puncture Conditions

During safety assessments of transport packages, cumulative damages are naturally accumulated for assessments performed using physical tests specimens. However, the cumulative damages are not as easily accounted when assessments are by numerical simulations. While analysts are comfortable with simulating packages for single events, it is not yet common practice to incorporate the effect handed over from a former event to the next, in a series of sequential load events. Thus, many numerical simulations in SAR (Safety Analysis Report) represent just a single event in the series of sequential event comprising the required accident condition. These single event numerical simulations are then based on initial conditions different from the analogous physical test specimen, which could contribute to a growing disparity in results between assessments by physically testing compared to numerical simulation. The reason why analyses do not consider the cumulative damage is difficulties in delivering the final result of the previous analysis to the current analysis.The hypothetical accident conditions described in the IAEA regulations include drop, puncture, fire, and water immersion conditions, which should be sequentially simulated. There can be cumulative damage between two accident conditions, such as drop and puncture, puncture and fire, and so forth. In this study, as the first step to consider cumulative damage, an analysis technology to perform a puncture analysis incorporating the final response field from a prior drop analysis is proposed. The necessity and validity of the proposed analysis technology are evaluated by a comparison with the results obtained by performing each analysis independently.Copyright