The 2-round Delphi survey and Focus Group Interview (FGI) survey method, in this study, are sequentially applied for the level analysis of the high-level radioactive waste (HLW) management technologies, that are classified into transport/storage, site evaluation, and disposal categories. The 2- round Delphi survey was conducted on domestic 56 experts in the HLW field in Korea, and survey answers were managed with questionnaires distributed by e-mail. In the FGI survey, domestic 24 experts from management field were formed into three groups to conduct in-depth interviews. Past research achievements including journal papers, intellectual properties and the expert opinions presented at expert hearing on HLW technology were used as reference materials. As a result of the survey, in this study, the average domestic technology level compared to the leading countries was 83.1% in transport area, 79.6% in storage area, 62.2% in site evaluation area, and 57.4% in disposal area, respectively. When compared to the former level analysis results in 2017, technology level of transport-storage area increased by 8.6%, and the site evaluation-disposal technology area decreased by 7.27%. The highest factor that increase the level of technology in the transport-storage field was due to the increased R&D program resulting on journal papers, intellectual properties. In addition, the decrease factor in the level of technology in the site evaluation-disposal field was mainly due to relatively low R&D program when compared to the leading countries. Suggested method for the level survey can be used to find out the basic data of the lower tech technologies, to estimate the efficient research budgets and to prepare the R&D human resources. With this regards, R&D roadmap can be matured with suggested prediction method for the domestic technology level on HLW.

Despite the increasing interest in Deep Borehole Disposal (DBD) for its capability of minimizing disposal area, detailed research about DBD operation system design should be conducted before the DBD can be implemented. Recently, DBD operation system applying wireline emplacement (WE) technique is under study due to its high flexibility and capability of minimizing surface equipment. In this study, a conceptual WE system, and operation procdure is introduced. The conceptual WE system consists of 3 main stations, which from the top are hoisting station (HS), canister connection station (CCS) and basement (BS). In HS, WE is controlled and monitored. The WE is controlled using wireline drum winch and sheaves, and load on wireline is measured using a load cell. HS also has a pressure control system (PCS), which monitors internal pressure of the system, and a lubricator, which act as housing for joint device, allowing the joint device to be easily inserted into the borehole. The joint device is used to connect the disposal canister to wireline for emplacement/retrieval. In CCS, a rail transporter brings a transport cask containing disposal canisters, then the transport cask is connected to the hoisting system and a PCS in the BS. The main component located at canister station are a sliding shielding door (SSD), and a slip. The SSD is used to prevent canister from falling into borehole during the connecting operation and prevent radiation from BS to affect the workers. The slip is located beneath the SSD and is used to hold the disposal canister before it is lowered into the borehole. In BS, PCS is installed to prevent overflow and blowout of borehole fluid. The PCS consists of wireline pressure valve, christmas tree and BOP, which all are a type of pressure valve to seal the borehole and release pressure inside the borehole. The WE procedure starts with transporting transport cask to CCS. The transport cask is connected to lubricator, and PCS. Joint device is lowered down to be connected with disposal canisters, then pulled up to check the load on the wireline. After the check-up, SSD is opened, and disposal canister is lowered into the borehole. When desired depth is reached, joint device is disconnected and retrieved for next emplacement. In this study, the conceptual deep borehole disposal system design implementing WE technique is introduced. Based on this study, further detailed design could be derived in future, and feasibility could be tested.

Compacted bentonite buffer materials are a key component of the engineered barrier system for high-level radioactive waste disposal. The bentonite buffer is saturated via groundwater flow through the excavation damaged zone in the adjacent rock mass. Bentonite saturation results in bentonite swelling, gelation and intrusion into the nearby rock discontinuities. Groundwater flow can cause bentonite erosion and transportation of bentonite colloids. This bentonite mass loss can negatively impact the long-term integrity of the engineered barrier system. Hence, it is necessary to understand the effects of erosion on the properties of the bentonite buffer. In this study, a series of artificial fracture erosion experiments are conducted to investigate the erosion characteristics of compacted Ca-bentonite buffer materials for different initial dry density conditions. Compacted bentonite blocks and bentonite pellets were manufactured using the cold isostatic pressing technique and granulation compactor respectively. The specimens were placed in a custommade transparent artificial fracture cell and the bentonite intrusion characteristics were monitored for two months under free swelling conditions with no groundwater flow. The radial expansion of the bentonite specimens within the artificial fracture was measured using a digital camera. In addition, the swelling pressure, displacement, and saturation were determined using a load cell-piston system, LVDT, and electrical resistivity electrodes respectively. A hydro-mechanical-chemical coupled dynamic bentonite diffusion model was applied to model the bentonite erosion characteristics using COMSOL Multiphysics.

The underground environment has an advantage to minimize the external influences because it is isolated space with surrounded rock medium. Therefore, underground rock has been used recently as the target for a disposal system of spent fuel with high-level radioactive. The disposal system mainly consists of natural barrier (i.e., surrounded rock medium) and engineered barrier (i.e., concrete lining, plug, backfill, canister, and buffer). In particular, the engineered barrier is important for long-term storage because it has to preferentially block the leakage of radioactive nuclide. Non-destructive technologies (NDT) have been utilized to monitor the state of disposal system for considering the limitation in deep depth conditions such as limited environment for direct damage inspection. Acoustic emission (AE) monitoring technique is an effective method to monitor the damage (crack) magnitude, history (i.e., crack evolution), and location using high-frequency elastic waves. To apply the AE monitoring method in the disposal system, the characteristics of damaged materials should be considered. The concrete lining has multi-failure behavior (i.e., brittle and ductile) resulted from composition as cement and reinforcing steel bar. Therefore, it important to investigate the AE characteristics according to the failure level of reinforced concrete for damage monitoring of the disposal systems. In this study, the four-point bending tests were carried out to measure the AE signals from the cracking of reinforce concrete specimens in laboratory. The test specimens were prepared with different strength. After the experiment, the AE characteristics were analyzed using the AE parameters with loading and failure state in the curve of time-stress. This study will be helpful for damage monitoring using AE technique in the field of high-level radioactive disposal system.

When a rapid groundwater inflow is introduced from the adjacent rock mass in the early stage of disposal, hydraulic pressure build-up occurs, which may cause piping erosion at the buffer material itself and the interface of the gap-filling material. Such piping erosion in compacted bentonite buffer via interaction between the buffer and the adjacent rock mass may deteriorate the performance of the buffer material. Therefore, it is necessary to understand the conditions and scenarios in which the piping phenomenon around the buffer material occurs for the long-term health of the repository. In this study, laboratory-scale experimental tests of piping erosion in buffer and interfacial rock was introduced. ø 100 mm × 200 mm height compacted bentonite specimens were placed in a cylindrical acetal cell, and the distilled water was continuously injected at a flow rate of 0.068 L/min using a dual syringe pump. The inflow of water was generated from the bottom and side cell of buffer material. During water injection, injected water pressure and amount were measured with visual observation. The results showed that the external saturation of buffer firstly occurs followed by piping crack generation along the wetting front. The additional piping channels were generated and merged with others. As the injection stopped, the swelling and self-sealing behavior of buffer material were observed. Moreover, X-ray CT scanning of the cell was conducted after the piping simulation to analyze the piping channels and saturation depth. The results highlight the piping erosion phenomenon mainly occurs due to the presence of a gap outside the buffer material. Further experimental cases is need to comprehensively understand piping phenomena in buffer material for assessing the long-term stability of underground radioactive waste disposal systems.

The backfill refills the deep geological disposal system after the installation of buffer in the disposal hole. SKB and Posiva have established the safety function for the backfill such as hydraulic conductivity of 10-10 m/s and swelling pressure of 0.2 MPa. The study on the thermal properties is required for the evaluation of performance design and long-term stability of backfill, since the thermal condition affects the hydraulic and mechanical behavior of backfill. Thermal conductivity is a key characteristic of thermal properties due to heat dissipation from spent fuel. In this study, thermal conductivities of bentonite-sand mixed blocks were measured. The silica sands were used instead of the crushed rock with bentonil-WRK, one of the candidate bentonite of the Korean repository system. The effects of size distribution and mass ratio of sand were evaluated. Four different size of silica sand (i.e., 0.18-0.25, 0.7-1.12, 1.6-2.5, 2.5-5.0 mm) and five mixing ratio (i.e., 1:9, 2:8, 3:7, 4:6, 5:5 of bentonite and sand) were used for characterization of thermal conductivity. As a result, the thermal conductivities were measured ranging from 1.6 to 3.1 W/m∙K depending on the size and mass ratio of the sand. The smaller the size or higher the mixing ratio of sand or the higher the water contents, the higher the thermal conductivity on the surface of backfill block. The higher compressing pressure induce higher thermal conductivity. Meanwhile, the feasibility study of backfill block productivity was reviewed according to the variables of this study. The excessive sand ratio and water contents lead to poor quality that results in the failure of the block. In Korea, the research of backfill is only now in fundamental steps, thus the results of this study are expected to use for setup the experimental conditions of hydraulic and mechanical performance, and can be used for the design of safety function and evaluation of long-term stability for deep geological disposal system.

The high-level waste disposal system is an underground structure exposed to complex environmental conditions such as high temperature, radiation, and groundwater. The high-level waste disposal causes structural cracks and deterioration over time. However, since the high-level waste disposal system is a structure that should be operated for a very long time, developing a high-durability monitoring sensor to detect cracks and deterioration is essential. The durability of the sensor can be evaluated by predicting the expected life through the accelerated life test, one of the reliability qualification tests. The most important factor in the accelerated life test design is setting the harsh stress level. This study figured out the harsh stress level of the piezoelectric sensor, which is commonly used for underground structure monitoring. It is possible to determine the appropriate stress level for the accelerated life test by investigating the harsh stress level for the temperature factor. It will contribute to more accurate life expectancy prediction.

The structural integrity of concrete silos is important from the perspective of long-term operation of radioactive waste repository. Recently, the application of acoustic emission (AE) is considered as a promising technology for the systematic real-time health monitoring of concrete-like brittle material. In this study, the characteristics of AE wave propagation through concrete silo of Gyeongju radioactive waste repository were evaluated under the effects of groundwater and temperature for the quantitative damage assessment. The attenuation coefficients and absolute energies of AE waves were measured for the temperature cases of 15, 45, 75°C under dry and saturated concrete specimens, which were manufactured based on the concrete mix same as that of Gyeongju concrete silo. The geometric spreading and material loss were taken into account with regard to the wave attenuation coefficient. The attenuation coefficient shows a decreasing pattern with temperature rise for both dry and saturated specimens. The AE waves in saturated condition attenuate faster than those in dry condition. It is found that the effect of water content has a greater impact on the wave attenuation than the temperature. The results from this study will be used as valuable information for estimating the quantitative damage at the location micro-cracks are generated rather than the AE sensor location.

In ROK, when designing a spent nuclear fuel (SNF) storage facility and cask, criticality safety analysis is performed assuming that the SNF is a fresh fuel in order to ensure conservatism. Storage and transportation capacity can be increased by more than 30% by applying the burnup credit, but it has not been applied to the management of SNF. On the other hand, currently in criticality safety analysis, average burnup value is applied to axial burnup profiles, and it is not conservative because burnup of the middle of SNF is greater than average value. Thus, measuring burnup of SNF with high accuracy contributes to the economics and safety of the management of SNF. In this paper, nondestructive burnup evaluation methods for SNF are reviewed in order to study how to measure burnup more accurately. Gamma ray spectrometry and neutron counting have been used as non-destructive burnup evaluation methods of SNF. Gamma spectrum analysis uses the ratio of Cs-134/Cs-137 or Eu-154/Cs-137. The ratio of Cs-134/Cs-137 is used to SNF with cooling time less than 20 years, and the ratio of Eu- 154/Cs-137 is used to SNF with cooling time more than 20 years due to their half-life. In spectrum analysis, detector sensors with high efficiency and energy resolution are needed to clarify each spectrum. High-purity germanium (HPGe) detector has high energy resolution. However, it is not suitable for the analysis of the SNF in the spent fuel pool because it requires separate cooling system and large volume. Thus, CdZnTe (CZT) detector, which has medium energy resolution, is used as a detector of gamma ray spectrometry for the analysis of the SNF in the spent fuel pool. Recently, LaBr3 detector has been commercialized. Although it is difficult to compare clearly due to different conditions such as detector volume and crystal size, LaBr3 detector showed better resolution than CZT in the entire energy region. Neutron counting method has a large error compared to gamma spectrometry because the neutron flux is lower than gamma ray, and neutron absorption reaction, induced fission, and pool environment have to be considered. Large quantity of gamma energy is deposited in the detector by the fission fragments near the SNF. Therefore, fission chambers, which have the highest insensitivity to gamma rays, must be used as neutron detector in order to avoid noise from gamma rays.

A tensile test is performed to obtain the mechanical property data of the spent fuel cladding. In general, the elastic modulus, elongation, yield stress, tensile stress, etc. are obtained by axial tensile test of cladding attaching an extensometer. However, due to the limitation in the number of specimens for spent nuclear fuel that can be made, the ring tensile test (RTT) whose required length of the specimen is short is mainly performed. In the case of RTT, an extensometer or strain gauge cannot be attached because the gauge part of the specimen is formed around the cladding and is short. In addition, since a load is applied in the radial direction of the cladding, a curved portion of the circular cladding is spread out and becomes straight, and then the cladding is tensioned. For this reason, it is difficult to obtain the stress-strain curve directly from the RTT results. Isight, which is used to identify the optimization design parameters, was used to build an optimization process that minimizes the difference between the RTT and the analysis to estimate the material property. For this, the elastic modulus, plastic strain, and the radius of the RTT jig were taken as fixed variables. As variables, isotropic hardening data and plastic stress were taken. The objective function was taken as the minimization of the area difference of the load-displacement curve obtained from the tests and analysis, of the difference in the magnitude of the maximum reaction force, and of the difference in the location where the maximum reaction force occurred. Optimization workflow was configured in the following order. First, using the calculator component, plastic stress design variables were created. Next, ABAQUS was placed to perform analysis using design variables, and the reaction force or displacement was calculated. After that, the reaction force was calculated considering the 1/4 symmetry condition using the script component. After that, the data matching component performed quantitative comparison of test and analysis data. Finally, by utilizing the exploration component, the plastic stress design variable that minimizes the difference in the objective function was obtained by automatically changing six optimization algorithms. In this paper, the constructed optimization process and the obtained plastic stress by applying it to the SUS316 RTT results are briefly described. The established optimization process can be utilized to obtain mechanical property from the results of the cladding RTT of spent nuclear fuel or new material.

For the transport of spent nuclear fuel, it is necessary to evaluate the amount of radioactivity for each assembly and the total amount of radioactivity for each cask. Currently, KHNP is evaluating the radioactivity using the Express mode of the OrigenArp program in the SCALE6.1 code. Express mode is a method to evaluate the radioactivity assuming that it has been burned with the same power per cycle, and Detail mode is a method to evaluate the actual combustion history such as power and cooling time for each cycle. For a total of 3,795 assemblies, including 1,391 assembliess for Kori Unit 1, 1,427 assemblies for Hanbit Unit 2, and 977 assemblies for Hanul Unit 3, the radioactivity was evaluated in Express mode and Detail mode, respectively, and the results were compared. As a result of the evaluation, it was confirmed that the results of the Express mode were evaluated more conservatively by 2.5~12.9% than that of the Detail mode. Accordingly, KHNP established a plan to change the evaluation method from Express mode to Detail mode in order to improve the accuracy of the radioactivity assessment results and eliminate conservatism.

The skeleton of fuel assembly is composed of top nozzle, bottom nozzle, grids, and guide tubes. In the reactor core, all the parts of the fuel assembly suffer degradations due to the condition of high temperature, pressure and water environment. Therefore, many material properties of high temperature mechanical strength, corrosion and irradiation resistance have been considered to choose the material for fuel assembly parts in the fuel development stage. The guide tubes have important roles to connect each parts and support the load of fuel assembly while the fuel is lifted. In Westinghouse 14×14 standard fuel assembly, Zircaloy-4 was used for the material of the guide tubes. Zircaloy-4 has a resistance to water corrosion and maintain good mechanical properties after the discharge from the core, so this alloy is also utilized for a fuel rod cladding material although the microstructure is slightly different due to the heat treatment difference. Thus, it is expected that there is no issue regarding the guide tube integrity after the discharge and during the storage in the pool, especially in case of low burn-up. However, the surface oxidation and resultant hydrogen pick-up can affect to the embrittlement to the Zr alloy. So, it is needed to know the actual status of spent fuel assembly by performing post-irradiation examination. In this study, the degradation level of the guide Tubes in low burn-up spent fuel assembly was investigated using the KAERI PIE facility in order to make some data which can be utilized to the baseline for evaluating the integrity of the spent fuel skeleton.

The management before disposal of spent nuclear fuel is an essential process for safe management. It is important to determine the amount of nuclide inventory in order to ensure the integrity of spent nuclear fuel, as radiation generated from the nuclides is generated along with residual heat in the spent nuclear fuel. Based on the data on the characteristics of spent nuclear fuel generated in Korea, the correlation equation between burnup and enrichment was derived by referring to overseas cases (Sweden). Source term analysis was performed using the SCALE ORIGEN ARP code by securing the burnup history of nuclear fuel. Calculation was performed by inputting the combustion history of the fuel WH14×14 and WH17×17 as a reference for CE16×16 spent fuel. Through this study, the relationship was identified using the burnup, enrichment, and cooling time factors that influence the characteristics of spent nuclear fuel. In addition, the total source and spectrum data from neutrons and gamma sources were used to find out the characteristics of fuel.

As drone technology and industry develop around the world, the use of drones are increasing in number and expanding to different fields. On the other hand, illegal flight and terrorist incidents using drones are also increasing day by day. In Korea, it is reflected in the “Design Basis Threat (DBT)”, which is the standard for designing and evaluating the physical protection system of nuclear power plants in accordance with the “Act on Physical Protection and Radiological Emergency”, that nuclear power plants continue to establish physical protection against drone threats. A total of 141 drone attacks or incidents have occurred around the world since 2015. Cases related to the Russian-Ukraine war, in which so many cases occurred, were excluded. There were 112 cases (79%) of terrorism or suspected terrorism using a single drone. There were 4 cases of terrorism using more than 5 drones, and a total of 20 drones were used to attack an oil facility in Yemen (2019). By region, a total of 111 incidents occurred in Middle East & North Africa. By country, there were 49 cases in Iraq, 35 cases in Saudi Arabia, and 8 cases in Syria. Among major countries, three cases occured in Korea, five in the United States, two in England, Canada, and Italy, and one in Japan and France. Since 2021, there have been 15 drone attacks. Multiple drones were used in attacks targeting military or large-scale Important National Facilities such as the Saudi oil refinery, Indian Kashmir air base, and reconnaissance of Iranian Natanz nuclear and surrounding military facilities. Also in 89% of the cases, the drones were loaded with explosives in order to cause large-scale damage. Accordingly, nuclear power plants, which are important national facilities, need to establish a system that can detect and respond to multiple drones. Furthermore, additional protective measures are needed for areas that are expected to be severely damaged which can be established by evaluating the impact of explosives on major points in the plant. In additionthere is a high possibility of terrorism by organizations aiming for national turmoil rather than individual terrorists. So it is important to identify signs of terrorism in advance and prepare through cooperation with related agencies.

UAVs (Unmanned Aerial Vehicle) are a rising threat to national facilities due to their cheap price and accessibility. Incidents such as the terrorism attack in Saudi Arabia’s oil facilities and the paralysis of the airport system in England’s Gatwick airport shows the need for integrating CUAS (Counter- Unmanned Aerial Systems) in important national facilities. Recently efforts have been made to evaluate the technical performance of the CUAS. Especially SNL (Sandia National Laboratory) modified the methodology used for PPS (Physical Protection Systems) to develop a performance metrics for CUAS. The performance metrics can be used to effectively analyze the facilities capability of countering drone attacks in a probabilistic way. In this study, we managed to derive the safety boundary of a reference nuclear power plant model based on its current CUAS and protection capabilities with a simplified methodology. Based on the outermost boundary of the model, the time table of the UAS consist of 4 variables which are the assessment time, transmission time, neutralization time and the maximum vehicle velocity. Dividing the maximum velocity to the net time derived, we estimated the minimum sensing point of the CUAS which is the minimum safety boundary of the facility to safely manage the UAV attack. Two practice cases were evaluated with the methodology which is based on the UAV groups classified by the United States DOD (Department Of Defense) that matches the classification of the UAV in Korea. Each variable was assumed to fit the process of a realistic nuclear power plant. Using the variables, we calculated the minimum safety boundary of the facility. With the methodology introduced in this study, regulators and stakeholders can easily evaluate the capability of the facilities CUAS for a design basis UAV attack. Also it can be used as a simple tool to analyze the facilities vulnerability for specific UAV specifications and a guideline to check the protective procedures of the facility.

The purpose of the present research is to verify the design characteristics of the SMART facility for the application of the IAEA’s safeguards-by-design (SBD) concept to small modular reactor (SMR) and to establish a foundation for SBD to be faithfully implemented as early as possible from the design stage. International Atomic Energy Agency (IAEA) is planning to facilitate the verification activities of inspectors by developing a safeguards approach to the reactor as early as possible and preparing a safeguards technical report (STR) before commercial operation of SMR begins. To this end, the IAEA is developing various approaches to the application of SBD to SMR with countries such as Republic of Korea, Russian Federation, China, the United States, and Canada through the Member State Support Program (MSSP). In order to review the unique design information of SMART facilities, the only deployable SMR in Korea, and to establish safeguards from the early design stages of SMART, it is necessary to carry out the task through cooperation with the Korea Atomic Energy Research Institute (KAERI) and Korea Institute of Nuclear Nonproliferation and Control (KINAC). IAEA agreed with the KINAC and KAERI to the direction of the project and to prepare both the Design Information Questionnaire (DIQ) and the Safeguards Technical Report (STR) for SMART facilities sequentially. The DIQ is a collection of questions to understand the characteristics of the reactor facilities that must be considered in applying safeguards. The STR is a document referenced by IAEA inspectors when verifying safeguards. Those draft versions were prepared and submitted to the IAEA. After review opinions were received, additional revision was conducted. In 2022, the IAEA holds the consultancy meeting on SBD for SMART. The purpose of the meeting is to review the draft DIQ and STR prepared by designers and discuss the future work plan of the task with designer and the task point of contact in order to safeguards can be considered at the early stage of the design. The results will be beneficial to the efficient safeguards verification activities of IAEA inspectors in the future.

The acoustic emission (AE) is proposed as a feasible method for the real-time monitoring of the structural damage evolution in concrete materials that are typically used in the storage of nuclear wastes. However, the characteristics of AE signals emitted from concrete structures subjected to various environmental conditions are poorly identified. Therefore, this study examines the AE characteristics of the concrete structures during uniaxial compression, where the storage temperature and immersion conditions of the concrete specimens varied from 15℃ to 75℃ and from completely dry to water-immersion, respectively. Compared with the dry specimens, the water-immersed specimens exhibited significantly reduced uniaxial compressive strengths by approximately 26%, total AE energy by approximately 90%, and max RA value by approximately 70%. As the treatment temperature increased, the strength and AE parameters, such as AE count, AE energy, and RA value, of the dry specimens increased; however, the temperature effect was only minimal for the immersed specimens. This study suggests that the AE technique can capture the mechanical damage evolution of concrete materials, but their AE characteristics can vary with respect to the storage conditions.

With the rapid development of China's economy and the gradual maturity of high-speed rail technology, China's high-speed rail tourism is also developing. With the slow recovery of the COVID-19 epidemic and the slow recovery of the tourism industry, the high-speed rail tourism industry has also continued to grow, and the combination of high-speed rail and tourism has been studied in depth to meet market demand. In this diverse society, the development of high-speed rail tourism will be promoted, employment will be increased, and consumption will be stimulated, thereby promoting a better development of the society. This paper explains the current situation of high-speed rail tourism and the specific impact of high-speed rail on tourism, as well as the current status of high-speed rail tourism, in order to better promote the development of high-speed rail tourism, thereby promoting national economic and social development.

Human intestinal flora is very diverse; with lactic acid bacteria (LAB) existing as part of the most vital gut microbes that improve host health. The application of LAB as a whole organism or its metabolite in the case of probiotic and bacteriocin respectively is extensive. Thus, the need to always bio-prospect for newer strains of LAB is essential. This study focused on isolating LAB from kunun-zaki and kindirmo, a fermented non-alcoholic beverage of non-dairy and dairy sources respectively, explored their physiological and biochemical properties, antibiotics sensitivity pattern and identified based on their 16S rRNA sequencing. A total of eighty isolates were selected sixty-six from kunun-zaki and fourteen from kindirmo in which 93.7% were bacilli and 6.3% were spherical in shape having 68.75% and 30% homofermentative and heterofermentative pathway respectively. All isolates have the ability to utilize glucose to produce lactic acid while their tolerance to pH 3 and salt concentration at 2%, 4% and 6.5% varied widely. Thirty-four isolates based on their physiological and biochemical properties were selected for molecular identification to ascertain their genera and species. Limosilactobacillus fermentum (68%); Lactiplantibacillus plantarum (6%) and Weissella confusa (3%) were confirmed species isolated. Thus, it was concluded that traditional fermented foods such as kunun-zaki and kindirmo are a good source to bio-prospect for LAB for product development, starter culture and probiotic study.

Electroanalytical study for the rotating cylinder electrode in molten LiCl-KCl eutectic salt (58– 42mol%) containing MgCl2 (0.1wt%) at 600°C is conducted. The researches of rotating cylinder electrode have been widely conducted for the century. The advantage of the electrode is that it can mitigate the unintended natural convection by providing a controlled diffusion boundary layer thickness. However, the experimental data for the high temperature molten salts is barely existed. The study adopts the electrochemical techniques such as cyclic voltammetry for the static cell and linear sweep voltammetry for the dynamic cell to calculate the diffusion coefficient. The peak current density and limiting current density are measured according to the scan rate. In order to evaluate the mass transfer under hydrodynamic flow condition, the revolution speeds of cylindrical electrode are varied from 10 rpm to 500 rpm which are corresponded to the Reynolds number of 4 and 185 respectively. The flow regime covers from the laminar to semi-turbulent regime (transient) as the critical Reynolds number Recrit is 200. The limiting current density shows a linear trend with the revolution speed and agrees well with the existing mass transfer correlations. For the extended flow regime, a new mass transfer correlation is suggested as the relation of non-dimensional numbers (Sh = aRebScc) based on the dimensionless analysis.