It has been investigated on the management of the nuclides in KAERI. Strontium-90 is a high heatgenerating nuclide in spent nuclear fuel. It is needed to separate the salt from the salt solution for the recovery of strontium after the chlorination of the strontium oxide in molten salt. A vacuum distillation technology was used for the separation of strontium from the molten salt. It was investigated on operating conditions of reactive distillation process for the recovery of the strontium from the salt solution. At a reduced pressure, considerable amount of the carbonation agents such as K2CO3 and Li2CO3 were reduced during heating in the distiller due to the thermal decomposition. Therefore, the two step process was proposed, which is composed of a reaction step at an atmospheric pressure and a salt distillation step at a reduced pressure. In the reaction step, the condition of low temperature and high pressure is suitable to suppress the decomposition of the carbonation agent. In the salt distillation step, reduced pressure is preferable at a suitable temperature depending on the evaporation rate of the salt.

Instead of using expensive platinum, carbon anodes could potentially be utilized in the process of reducing oxides in LiCl-Li2O molten salt at a high cell potential. However, this high potential leads to the generation of a mixture of anodic gases containing toxic and corrosive gases such as chlorine (Cl2), oxygen (O2), carbon monoxide (CO), and carbon dioxide (CO2). To better understand this gas mixture, we conducted real-time analyses of the gases generated on the carbon anode during the TiO reduction reaction in the molten salt at 650°C, using a MAX-300-LG gas analyzer. Our results indicate that the ratio of CO/O2/CO2/Cl2 in the gas mixture is significantly influenced by the composition of the salt, and that removing the sources of oxygen ions in the salt increases the likelihood of generating toxic and corrosive Cl2 gas.

Separating nuclides from spent nuclear fuel is crucial to reduce the final disposal area. The use of molten salt offers a potential method for nuclide separation without requiring electricity, similar to the oxide reduction process in pyroprocessing. In this study, a molten salt leaching technique was evaluated for its ability to separate nuclides from simulated oxide fuel in MgCl2 molten salts at 800°C. The simulated oxide fuel contained 2wt% Sr, 3wt% Ba, 2wt% Ce, 3wt% Nd, 3wt% Zr, 2wt% Mo, and 89wt% U. The separation of Sr from the simulated oxide fuel was achieved by loading it into a porous alumina basket and immersing it in the molten salt. The concentration of Sr in the salt was measured using ICP analysis after sampling the salt outside the basket with a dip-stick technique. The separated nuclides were analyzed with ICP-OES up to a duration of 156 hours. The results indicate that Ba and Sr can be successfully separated from the simulated fuel in MgCl2, while Ce, Nd, and U were not effectively separated.

When damaged nuclear fuel is stripped and re-fabricated into stabilized pellets, it is necessary to analyze the characteristics of the stabilized pellets, such as density, leaching behavior, and compressive strength, for final disposal. In this study, simulated nuclear fuel with UO2 and burn-up of 35 GWd/tU and 55 GWd/tU was used to measure the compressive strength of the stabilization pellet. In order to change the density of the sintered pellet, a sintered pellet was prepared by heat treatment at 1,550°C and 1,700°C for 6 hours in a reducing atmosphere of 4% H2/Ar. In the case of UO2, the density was 10.4 g/cm3 (94.5% of T.D.) and 10.6 g/cm3 (96.6% of T.D.) depending on the sintering temperature (1,550°C, 1,700°C). In the case of simulated fuel with a burn-up of 35 GWd/tU, the density was 8.8 g/cm3 (80.9% of T.D.) and 10.2 g/cm3 (93.6% of T.D.) depending on the sintering temperature (1,550°C, 1,700°C). In the case of simulated fuel with a burn-up of 55 GWd/tU, the density was 8.3 g/cm3 (77.0% of T.D.) and 10.0 g/cm3 (92.3% of T.D.) depending on the sintering temperature (1,550°C, 1,700°C). It was found that the compressive strength of simulated nuclear fuel decreased with increasing burn-up and increased with increasing density. In the case of UO2, the compressive strengths were 717.8 MPa and 897.4 MPa when the densities were 10.4 g/cm3 and 10.6 g.cm3, respectively. In the case of simulated nuclear fuel with a burn-up of 35 GWd/tU, the compressive strengths were 472.1 MPa and 732.3 MPa when the densities were 8.8 g/cm3 and 10.2 g/cm3. In the case of simulated nuclear fuel with a burn-up of 55 GWd/tU, the compressive strengths were 301.4 MPa and 515.5 MPa when the densities were 8.3 g/cm3 and 10.0 g/cm3, respectively.

Spent fuel from the Wolsong CANDU reactor has been stored in above-ground dry storage canisters. Wolsong concrete dry storage canisters (silos) are around 6 m high, 3 m in outside diameter, and have shielding comprised of around 1 m of concrete and 10 mm of steel liner. The storage configuration is such that a number of fuel bundles are placed inside a cylindrical steel container known as a Fuel Basket. The canisters hold up to 9 baskets each that are 304 L stainless steel, around 42” in diameter, 22” in height, and hold 60 fuel bundles each. The operating license for the dry storage canisters needs to be extended. It is desired to perform in-situ inspections of the fuel baskets to very their condition is suitable for retrieval (if necessary) and that the temperature within the fuel baskets is as predicted in the canister’s design basis. KHNP-CNL (Canadian Nuclear Lab.) has set-up the design requirements to perform the in-situ inspections in the dry storage canisters. This Design Requirements applies to the design of the dry storage canister inspection system.

Spent nuclear fuels released from the reactor are stored in cooling pools and then stored in dry storage casks. During the transition from the wet storage to dry storage cask, a vacuum drying process is used to remove residual water in the cask. During the vacuum drying process, gas pressure is reduced to below 400 Pa to promote evaporation and water removal. KAERI is developing a PWR single assembly (PLUS7) test equipment to simulate the thermal flow in spent fuel assembly. In this study, the thermal conductivity of air at low pressure was derived to perform the thermal analysis of the canister in vacuum. In addition, thermal analyses were performed for the canister with backfill gases of helium, air, and a vacuum in the vertical orientation using the COBRA-SFS code. At low pressure, the thermal conductivity of air depends on pressure and temperature. The reduced thermal conductivity, kr (W/m-K) was calculated using the curve fit for air at reduced pressure in thin gaps presented in the General Electric Fluid Flow Handbook. 􀝇􀯥/􀝇􀬴 = 􀬵 􀬵􀬾 􀮼􀯍/􀯉􀰋 Where, k0 is the thermal conductivity at atmospheric pressure (W/m-K), P is the reduced (vacuum) pressure (Pa), δ is the gap size (m), T is the temperature (K), and C is the Lasance constant (7.657E-5 N/m-K). The thermal conductivity of air decreases as the pressure decreases. The reduced thermal conductivity of air at pressures of 400 Pa and 40 Pa was calculated to be 0.97 and 0.77, respectively. For the analysis in vacuum, no enhancement of the convective heat transfer was assumed (Nu=1.0). For the helium backfill, the peak cladding temperature was the lowest and the axial temperature profile was the flattest due to the higher thermal conductivity and lower density of the helium. For the vacuum backfill, the peak cladding temperature was the highest and temperature gradient was the sharpest due to the only radiative heat transfer effect in the fuel assembly.

A radiation shielding resin with thermal stability and high radiation shielding effect has been developed for the neutron shielding resin filled in the shielding shell of dry storage/transport cask for spent nuclear fuel. Among the most commercially available neutron shielding resins, epoxy and aluminum hydroxide boron carbide are used. But in case of the resin, hydrogen content enhances the neutron shielding effect through optimization of aluminum hydroxide, zinc borate, boron carbide, and flame retardant. We developed a radiation shielding material that can increase the boron content and have thermal stability. Flame retardancy was evaluated for thermal stability, and neutron shielding evaluation was conducted in a research reactor to prove the shielding effect. As a result of the UL94 vertical burning test, a grade of V-0 was received. Therefore, it was confirmed that it had flame retardancy. According to an experiment to measure the shielding rate of the resin against neutron rays using NRF (Neutron Radiography Facility), a shielding rate of 91.54% was confirmed for the existing resin composition and a shielding rate of 96.30% for the developed resin composition. A 40 M SANS (40 M Small Angle Neutron Scattering Instrument) neutron shielding rate test was performed. Assuming aging conditions (6 hours, 180 degrees), the shielding rate was analyzed after heating. As a result of the experiment, the developed products with 99.8740% and 99.9644% showed the same or higher performance.

The damaged spent fuel rods must be stabilized by encapsulation or dry re-fabrication technologies before geological disposal. For applying the dry re-fabrication technology, we manufactured a vertical type furnace to perform the fuel material recovery from damaged fuel rods by oxidative decladding technology. As driving forces to accelerate oxidative decladding rate, magnetic vibration and pulse hammering generated by a pneumatic cylinder were used in this study. The oxidative decladding efficiency and recovery rate of fuel oxide powder with rod-cut length, oxidation temperature and time, oxygen concentration, and gas mixtures were investigated using simfuel rod-cuts in a vertical furnace for fuel material recovery and powder quality improvement. The oxidative decladding was performed for 2.5-10 h as following operation parameters: simfuel rod-cut length of 50-200 mm, oxidative temperature from 450 to 580°C, oxygen concentration of 49.5 or 75.6%, and gas mixtures in O2/Ar or O2/N2. In magnetic vibration, oxidative decladding was progressed only at bottom portion of fuel rodcut. Whereas, oxidative decladding in pulse hammering was occurred at both top and bottom portions of fuel-rod. In pulse hammering method, the oxidative decladding conditions to declad rod-cuts of 50- 200 mm in length were established to achieve both decladding efficiency of ~100% and fuel material recovery rate of > 99%. These conditions were as follows: oxidation temperature and time at 500°C and 2.5-10 h, oxygen concentration at 75.6% under O2/N2 gas mixtures. As operation conditions for a pneumatic cylinder, stroking, actuating, and waiting times were 0.5, 3, and 12 s.

The Ag0-containing sorbents synthesized by Na, Al, and Si alkoxides have higher maximum iodine capture capacity and textural properties than zeolite-based Ag0-containing sorbents. However, these sorbents were prepared in the form of granules via a step for cutting cylindrical alcogels. Since asmade sorbents decreased packing density, they must be additionally crushed and then classified into an appropriate size for increasing packing density. The bead formation in the step of sol-gelation could bring about the simplification of sorbent preparation process and an improvement of packing density. In the Na, Al, and Si alkoxides as starting materials, sol solution was hydrophilic and lower density than vegetable oil, which transformed sol droplets to sol-gel beads. Thus, in these precursors, sol droplets, which must be sprayed by single nozzle placed at bottom side of oil column, can rise up through oil column. Acetic acid (HOAc) was used as the catalyst for the hydrolysis of Na alkoxide (TEOS) and gelation of the Na+AlSi-OH alcosol. For obtaining sol-gel beads, experiments were performed by the flowrate change of sol solution and HOAc at different nozzle sizes using soybean oil column of 1 m in length. At a sol/HOAc flowrate ratio of 3.85, some Na+AlSi-OH alcogel beads were obtained. After the Ag/Na ion-exchange, Ag content in Ag+AlSi-OH hydrogel was low due to reaction between Na+ and HOAc during sol-gelation and aging step. The Ag+AlSi-OH hydrogel with high Ag content could be prepared by Na addition. After the solvent exchange and drying at ambient pressure, the bead sorbents had higher Ag0 content and larger pore size than granular sorbents. However, further experiments are needed to increase yield rate in bead sorbent.

The stabilization techniques are highly required for damaged nuclear fuel to strengthen safety in terms of transportation, storage, and disposal. This technique includes recovering fuel materials from spent fuel, fabrication of stabilized pellets, and fabrication of fuel rods. Thus, it is important to identify the leaching behavior of the stabilized pellets to verify their stability in humid environments which are similar to storage conditions. In this study, we introduce various leaching experiment methods to evaluate the leaching behavior of the stabilized pellets, and determine the most suitable leaching test methods for the pellets. Also, we establish the leaching test conditions with various factors that can affect the dissolution and leaching behavior of the stabilized pellets. Accordingly, we prepare the simulated high- (55 GWd/tU) and low- (35 GWd/tU) burnup nuclear fuel (SIMFUEL) and pure UO2 pellets sintered at 1,550°C and 1,700°C, respectively. Each pellet is placed in a vessel and filled with DI water and perform the leaching test at three different temperature to verify the leaching mechanism at different temperature range. Based on the standard leaching test method (ASTM C1308-21), the test solution is removed from the pellet after specific time intervals and replaced in the fresh water, and the vessel is placed back into the controlled-temperature ovens. The test solutions are analyzed by using ICP-MS.

On-site storage facility using concrete silo dry storage systems for spent nuclear fuel at Wolsong NPP site came into operation in 1992 and was expanded four times, and a total of 300 silo dry storage systems are currently in operation. The design lifetime of silo dry storage systems has been licensed for 50 years. As the dry storage systems are subject to time constraints for a limited lifetime, countries operating the dry storage systems are working to ensure the long-term integrity of dry storage systems and IAEA also recommends that the dry storage systems be assessed for long-term storage. To demonstrate the long-term integrity due to material degradation during the licensed design lifetime, the structural integrity of silo dry storage systems was evaluated by considering the material degradation characteristics of concrete. The concrete compressive strength results measured so far by the rebound hammer method, which is an internationally standardized nondestructive test method for converting hardness into compressive strength using the correlation between rebound number and strength at the time of a Schmidt hammer strike, were analyzed in accordance with Wolsong NPP’s procedure to quantify the degradation characteristics, and the prediction of concrete strengths for 20 years and 50 years after construction of the silo dry storage systems was determined, respectively. Based on these residual compressive strengths, structural analyses of the silo dry storage systems were carried out under normal, off-normal and accident conditions of the related regulations, and the structural integrity of silo dry storage systems was reevaluated. It was confirmed the silo dry storage systems are able to maintain structural integrity up to the design lifetime of 50 years even if the concrete is deteriorated.

The spent fuel storage canister is generally made of austenitic stainless-steel and has the role of an important barrier to encapsulate spent fuels and radioactive materials. Canister near coastal area has welding lines, which have high residual tensile stresses after welding process. Interaction between austenitic stainless steel and chloride environment forms detrimental condition causing chloride induced stress corrosion cracking (CISCC) in canister. Reducing or eliminating tensile stress on canister can significantly decrease probability of crack initiation. Surface stress improvement works by inducing plastic strain which results in elastic relaxation that generates compressive stresses. Surface stress improvement methods such as burnishing process can effectively prevent for CISCC of canister surfaces. In this study, burnishing treatment has been evaluated to control residual tensile stress practically applicable to atmospheric CISCC for aging management of steel canisters. Burnishing process was selected as a prevention technology to CISCC of stainless steel canisters to improve resistance of CISCC through enhancement of surface roughness and generation of compressive residual stress. SUS 316 SAW (Submerged Arc Welding) specimens were burnished with flat roller and round roller after manufactured and assembled on CNC machine using base plate. The burnishing test results showed that the surface roughness of SUS 316 SAW welded specimens after roller burnishing of pass No. 5 was improved with 85% with flat roller and 93% with round roller, individually. Surface roughness showed the best state when burnished at pressure of 115 kgf, feeding rate of 40 m/stroke and pass No. of 5 turns with round roller. The surface of SUS 316 SAW welded specimens had much high residual compressive stress than yield stress of SUS 316 materials with roller burnishing treatment, independently of kinds of roller. The surface of the welded specimen by round roller burnishing showed smaller compressive stress and deeper stress region than in the surface of flat round roller burnishing. The roller burnished canister with good surface roughness could reduce the number of crack initiation sites and the high residual compressive stress formed on the welded surface might prevent the crack initiation by reducing or eliminating tensile residual stress in the weld zone, finally leads to excellent CISCC resistance. The crack growth behavior of SUS 316 welded specimens will need to investigate to evaluate the corrosion integrity of the canister materials under chloride atmosphere according to burnishing treatment.

The purpose of this study is to provide technical issues in upgrade and modification of fuel handling equipment at operating nuclear power plants. The improvement for safety function and performance enhancement of fuel handling equipment has been going on for 20 years since the early 2000’s. This improvement is recently focused on the replacement of components through the performance analysis and the operation and maintenance plan based on replacement cycle of its component. Additionally, it is required to secure spare parts so that it can be operated at all times with compatibility and standardization to other domestic nuclear power plants. The fuel handling equipment is consisted of refueling machine, upender and carriage of fuel transfer system, spent fuel handling machine, new fuel elevator and various tools, and the equipment are linked in systematic interlocks. Fuel handling is a critical task during a nuclear power plant refueling outage. Even minor component defects may stop operation of the whole system and have a significant impact on the overall system process. To achieve this goal, major components that are expected to be replaced for reliable operation are summarized as follows; 1) motor assembly with AC servomotors and driver for bridge, trolley and hoist of refueling machine and spent fuel handling machine, 2) winch motor and drive for upender and carriage of fuel transfer system, 3) operator control console with a HMI PC base PLC (Programmable Logic Controller) control system, 4) positioning and load weighing sensors such as an encoder and a load cell with its support for periodic calibration and maintenance, 5) main power drapped style festoon cable assembly for bridge of refueling machine, 6) pneumatic control assembly for gripper operation of refueling machine, 7) active components (e. g., air motor, hydraulic cylinder and limit switch) to be removable and reinstallable without requiring the water level to be lowered. It is advisable to utilize such various information as it can help to improve reliability of fuel handling as a critical path in upgrade and modification of fuel handling equipment at operating nuclear power plants.

After Fukushima nuclear power plant accident in 2011, Concerns about accident of spent fuel pool increase. In Korea, the time of saturation of spent fuel pool is coming, but regulatory measures and safety evaluation are insufficient when occurring spent fuel pool accident. Thus, it is necessary to review of spent fuel pool accident in foreign countries to establish regulatory measures and safety evaluation of spent fuel pool accident suitable for domestic spent fuel pool. Therefore, we reviewed spent fuel pool accident that occurred at Fukushima Unit 4, SONGS Unit 2 and PAKS. In Japan, spent fuel pool accident occurred at Fukushima NPP in 2011. Tsunami was cause of the accident. Station Black Out occurred at Fukushima NPP and Emergency Diesel Generator lost their functions due to Tsunami. As a result, Loss of cooling happened in spent fuel pool at Fukushima NPP. For Unit 4, wall of spent fuel pool in Unit 4 was damaged due to hydrogen explosive, so loss of coolant in spent fuel pool of Unit 4 occurred. After the accident, the temperature of spent fuel pool increases to 75°C, but there was no damage to the spent fuel. In USA, spent fuel pool accident occurred at SONGS Unit 2 in 2013. The debris of nearby ocean is cause of the accident. The debris entered the system through a damaged Salt Water Cooling pump suction strainer. The debris obstructed flow through the Component Cooling Water heat exchanger and operation of Salt Water Cooling. The maximum spent fuel pool temperature during this event was 25.6°C. It was a value that satisfied the technical specifications of the SONGS NPP. In Ukraine, spent fuel pool accident occurred at PAKS in 2003. Unintentionally opened valve of cleaning tank is cause of the accident. Loss of coolant occurred in spent fuel pool of PAKS. Due to loss of coolant, spent fuels were exposed to the vapor state atmosphere, and oxidation occurred in the cladding tube of the spent fuel that rose to 1,400°C. In this study, Review of spent fuel pool accident in major foreign countries was conducted as basic studies for establishing regulatory measures and safety evaluation of spent fuel pool in Korea. Causes of each accident were different by structure of spent fuel pools. Result of this study will be contributed to establish safety measures of spent fuel pool accident suitable for domestic spent fuel pool facility.

As temporary storage facilities for spent nuclear fuel (SNF) are becoming saturated, there is a growing interest in finding solutions for treating SNF, which is recognized as an urgent task. Although direct disposal is a common method for handling SNF, it results in the entire fuel assembly being classified as high-level waste, which increases the burden of disposal. Therefore, it is necessary to develop SNF treatment technologies that can minimize the disposal burden while improving long-term storage safety, and this requires continuous efforts from a national policy perspective. In this context, this study focused on reducing the volume of high-level waste from light water reactor fuel by separating uranium, which represents the majority of SNF. We confirmed the chlorination characteristics of uranium (U), rare earth (RE), and strontium (Sr) oxides with ammonium chloride (NH4Cl) in previous study. Therefore, we prepared U-RE-SrOx simulated fuel by pelletizing each elements which was sintered at high temperature. The sintered fuel was again powdered by heating under air environment. The powdered fuel was reacted with NH4Cl to selectively chlorinate the RE and Sr elements for the separation. We will share and discuss the detailed results of our study.

As a method for chlorinating spent nuclear fuel, a method of using ZrCl4 in high-temperature molten salt is known. However, ZrCl4 has a sublimation property that vaporizes at a temperature similar to the melting temperature of molten salt. Since solubility of ZrCl4 in molten salt is very low, it is difficult to dissolve a large amount of ZrCl4 in molten salt. However, once ZrCl4 can be dissolved together with the molten salt, it remains in the molten salt without vaporizing. That is, it is known that when vaporized ZrCl4 reacts with molten salt in a sealed reactor, it dissolves into the molten salt, and ZrCl4 above the solubility remains in the molten salt in the form of M2ZrCl6. Here, M represents an alkali element. Therefore, in this study, a flange-type sealed reactor was fabricated to dissolve a large amount of ZrCl4 in LiCl-KCl salt, and LiCl-KCl salt in which ZrCl4 was dissolved as K2ZrCl6 was prepared. LiCl-KCl, KCl, and ZrCl4 salts were charged into alumina crucibles and placed in a sealed reactor. The reactor was heated to 500°C and the reaction time was about 20 hours. The temperature of the reactor surface was about 480°C. After completion of the reactions, each crucible was recovered from the inside of the reactor. All of the ZrCl4 vaporized and there was no residue in the crucible. Both KCl and LiCl-KCl salts appear to have dissolved and then cooled, with respective weight gains. XRD analysis was performed to observe the structure of the recovered salts, and ICP analysis was performed to measure the Zr element content in each salt. As a result of XRD analysis, the structure of K2ZrCl6 was found in the KCl salt, but not in the LiCl-KCl salt. As a results of ICP analysis, it was found that the LiCl-KCl salt contained about 33wt% of ZrCl4, and about 25wt% was dissolved in the KCl salt. In other words, it was shown that ZrCl4 above the solubility can be dissolved in the LiCl-KCl molten salt.

The hydride reorientation (HR) of the post-irradiated nuclear fuel cladding after use affects the integrity of the spent nuclear fuel. During the dry storage process, which is an intermediate storage method, it was found that the hydride in the circumferential direction is rearranged into radial hydride, and this is believed to be due to factors such as hoop stress, peak temperature, accumulated hydrogen concentration, and cooling rate during the storage period. f(HR) = f(Tmax) + f(σH) + f(CH) + f(△T) + f(10Cy) + f(cooling rate) + ...... To simulate long-term dry storage of spent nuclear fuel, the hydride reorientation behavior was evaluated using unirradiated Zircaloy-4 (CWSRA) cladding with hydrogen charged under various hoop stresses (70, 80, 90, and 110 MPa) at long-term cooling periods (3, 6, and 12 months). Test results showed that as the cooling time increased, the sample with 90 MPa hoop stress at a maximum temperature of 400°C approached the ductility recommendation limit of 2%. In a 90 MPa hoop stress specimen with 3 months cooling period at peak temperature of 400°C, the offset strain was 4.24% at room temperature RCT, while it showed the result of 2.86% for the cooling period of 12 months. On the other hand, the specimen with hoop stress of 110 MPa and cooling period of 12 months showed result of 1.4%. The test results need to take into account errors in hydrogen charging and hydrogen analysis, and it is necessary to consider reproducibility through repeated tests. These results indicate the need for continued attention to the evaluation of the effects of hydride reorientation due to long-term cooling in the context of the integrity of spent fuel.

In the process of spent fuel dry storage, which is an intermediate management method, it was found that hydrides in the circumferential direction rearranged into radial hydrides. Various factors, such as hoop stress, peak temperature, cooling rate during the storage period, and hydrogen concentration accumulated during the burnup process, significantly affect the susceptibility of spent fuel cladding. In recent studies based on the hydrogen solubility value of about 210 ppm corresponding to the peak temperature of 400°C, if the threshold stress decreases as the hydrogen concentration increases in the low hydrogen range under 210 ppm, the threshold stress increases as the hydrogen concentration increases in the low hydrogen range under 210 ppm. The fundamental cause of this trend is the diffusion of hydrogen into the high-stress region due to the stress gradient formed in the specimen, and hydrogen compounds which remain undissolved in the circumferential direction, even at the peak temperature, play a crucial role to determine the magnitude of the threshold stress. This study evaluated the behavior of hydride reorientation under various hoop stress conditions (70, 80, 90, and 110 MPa) using unirradiated Zircaloy-4(CWSRA) cladding tubes under long-term cooling conditions (3, 6, and 12 months). The results of analyzing the offset strain by hydrogen concentration for long-term cooling showed that specimens with low hydrogen concentration exhibited higher integrity than specimens with high hydrogen concentration at hoop stresses of 90 and 110 MPa. The HR test using irradiated fuel cladding showed that specimens with low hydrogen concentrations exhibited relatively higher susceptibility. To quantify these results, it is necessary to research further in detail by repeated tests.

Spent nuclear fuels should be safely stored until being disposed and dry storage system is predominantly used to retain the fuels. During long-term storage, there are several mechanisms that could result in the degradation of spent nuclear fuels, and the temperature is the most important parameter to predict and estimate the degradation behaviors. Therefore, thermal analysis to estimate temperatures of spent nuclear fuel and the storage system should be performed to evaluate whether the temperatures exceed safety limit. Recently, thermal hydraulic analysis with CFD codes is widely used to investigate the temperature of spent nuclear fuel in dry storage. Herein, Explicit CFD analysis model is introduced and validated by estimating the thermal hydraulic response of the dry storage system that is Dry Cask Simulator (DCS). Extended Storage Collaborating Program (ESCP) led by the Electric Power Research Institute (EPRI) is organized to assess degradation effects of spent nuclear fuel during long-term dry storage, and DCS is the first phase of the program. The dry storage system, containing a single BWR assembly in a canister, was designed to produce validation-quality data for thermal analysis model. ANSYS FLUENT is used to simulate DCS, and the test condition of 0.5 kW decay heat and 100 kPa helium pressure was investigated in this study. In case of peak cladding temperature (PCT), PCT from the experiment was 376 K while that of CFD was 374 K. It implies CFD simulation gives good agreement with experimental measurement. Peak temperatures of channel can, basket, canister and shell predicted by CFD simulation also show good prediction and the discrepancies were less than 7 K while measurements uncertainty was 7 K.

Once systems, structures and components (SSCs) of dry storage systems are classified with respect to safety function or safety significance (i.e., safety classification), appropriate engineering rules can be applied to ensure that they are designed, manufactured, maintained, managed (e.g. aging management) etc. In Unites States, the systems, structures and components (SSCs) consisting DSSs are classified into two or several grades (i.e., class A, B and C or not important to safety, and important to safety (ITS) or not important to safety (NITS)) with respect to intended safety function and safety significance. This classification methods were based on Regulatory Guide 7.10 (i.e., guidance for use in developing quality assurance programs for packaging). Also, in Korea, SSCs of DSSs should be classified into ITS and NITS in much the same as method based on Regulatory Guide 7.10. In that guidance, for providing graded approach to manage the SSCs of packaging, they were trying to classifying SSCs in accordance with radiological consequences. But there was limitations that the provided classification criteria was still qualitative, so that it was not enough for managing the SSCs according to graded approach. On the other hand, in some other nuclear facilities (i.e., nuclear power plant, radioactive waste management facility and disposal facility etc.), quantitative criteria relevant to radiological consequence (i.e., radiation doses to workers or to the public) or inventory of radioactivity are existed so that it can be applied for classifying safety classes. In summary, the study on the application safety classification that applied quantitative criteria to perform safety classification of SSCs in DSS is inadequate or insufficient. The purpose of this study is proposing the preliminary framework for estimating safety significance of SSCs in DSS which can be utilized in our further advanced studies. In this study, a framework was established to estimate the safety significance of SSCs related to radiation shielding and confinement using MCNP® 6.2 and Microsoft Excel. Referring to the methodology of IAEA Specific Safety Guide 30, we assumed severity for failures of components that could lead to degradation of the SSC’s performance. The safety class of SSC was decided based on the impact of SSC’s failure on consequences.