There are a number of methods to evaluate the quality of squid. However, when purchasing the fish, consumers and retails rely only on the sensory test and flavor in the field. Therefore, this study was aimed to prove relationship between scientific indicator and sensory test. Total viable cell count (TVC), viable cell count of Pseudomonas spp., pH and volatile basic nitrogen (VBN) were selected as scientific indicators and mesured during the storage of squid at different temperature. The squid was storaged at 3 different temperature (5oC, 15oC, 20oC). Off flavor determination time was measured by R-index, and kinetic modeling was conducted. Activation energies of offflavor determination time, TVC, Pseudomonas spp, VBN, and pH were 51.210 kJ/mol, 42.88 kJ/mol, 50.283 kJ/mol, 72.594 kJ/mol and 41.99 kJ/mol respectively. Activation energy of off-flavor determination time was approximated to viable cell count of Pseudomonas spp., TVC, pH and VBN as an order. Especially, viable cell count of Pseudomonas spp. had best match of the activation energy. Therefore, it was judged that indicator of off-flavor determine time was viable cell count of Pseudomonas spp..

본 연구는 기능성 물질인 알로에를 종류별로 첨가함에 따라 유화소시지의 이화학적 특성과 저장성, 그리고 관능성에 미치는 영향을 조사하기 위해 수행되었다. 알로에 베라(AV구)와 알로에 사포나리아(AS구)는 각 3%씩 첨가하였으며, 모든 처리구들은 냉장온도 4±1℃에서 4주간 저장하면서 주 1회 간격으로 실험이 진행되었다. pH는 AV구와 AS구가 대조구보다 유의적(p<0.05)으로 낮았고, 저장기간의 경과함에 따라 대조구 및 모든 처리구는 증가하였다. 보수성에서 AS구는 대조구와 AV구보다 유의적(p<0.05)으로 높은 값을 나타냈으며, 저장기간의 경과함에 따라 대조구 및 모든 처리구는 유의적(p<0.05)으로 감소하였다. 적색도는 AS구가 대조구 및 AV구 보다 낮은 값을 나타내었고, 황색도는 AS구가 대조구, AV구보다 높은 값을 나타내었다. 전단가는 AS구가 대조구보다 낮은 값을 나타냈으며, AV구와는 유의적인 차이가 없었다. 지방 산패도에서 AS구는 유의적 (p<0.05)으로 가장 낮은 값을 나타내었고, 관능 평가의 색은 AV구가 가장 높은 값을 나타냈으며, 선호도는 AV구가 대조구보다 높은 값을 나타냈고 AS구와는 유의적인 차이가 없었다. 이는 알로에 사포나리아의 첨가가 높은 보수성, 알로에 베라 첨가보다 낮은 지방산패도, 유사한 관능성을 나타내 기능성 소시지의 개발에 충분한 경쟁력을 가지고 있으며 향후 알로에 사포나리아를 첨가한 육제품의 일반화 및 대량생산 체계의 확립에 긍정적인 영향을 준다고 판단된다.

This study was conducted to establish an appropriate period of use of sawdust spawn at low temperatures and a nutrient supplement medium for cultivation of Lentinula edodes ‘Hwadam’. Of the nutrient supplements, the total yield of rice bran (5%) + corn flour (5%) treatments were 673.3 g, which was higher than rice bran (551.6 g) and wheat bran (546.7 g) treatments, respectively. As shown by the growth of Lentinula edodes ‘hwadam' during to the sawdust spawn storage period (at 4oC), the period of spawn running, browning, fruiting body formation, and development was 27 d, 81 d, 5 d, and 11-13 d, respectively, regardless of the length of the storage period at 4 oC. After 3 months of storage of sawdust spawn, the number of fruiting bodies and yield decreased as the storage period increased. Therefore, the period of use of sawdust spawn (at 4 oC) for the stable production of fruiting bodies of Lentinula edodes ‘Hwadam’ was a maximum of 3 months.

This study evaluated the Protaetia brevitarsis larvae powder’s characteristic changes using hot air drying (60±2.5oC, 12 h) with different pre-treatment methods, including two sacrifice methods, two storage temperatures, and two defatting processes. Appearance, yield, moisture contents, pH, color, proximate analysis, volatile basic nitrogen level, DPPH radical scavenging activity, and total phenol content were assessed. Results revealed that a combination of blanching, defatting, and -20oC storage temperature resulted in higher total phenol contents, lower water contents, and lower volatile basic nitrogen levels than other methods. Defatted treatment resulted in a higher L-value than the non-defatted treatment. Taken together, these results indicate that a combination of -20oC storage, blanching, and defatting is the optimal pre-treatment method for obtaining P. brevitarsis larvae powder with high total phenol content, low water content, and low volatile basic nitrogen, taking into account cost efficiency considerations.

In Korea, research on the introduction of dry storage facility is being conducted as an alternative to saturation of temporary storage facilities for spent nuclear fuel. The introduction of dry storage facilities requires a radiological impact assessment on the workers of the facility, and for this, an appropriate exposure scenario must be derived through work procedure analysis. In this study, the procedure for storing spent nuclear fuel in dry storage facilities was analyzed based on the case of evaluating the radiological impact of workers in dry storage facilities abroad. We investigated cases of radiological impact assessment on workers at on-site dry storage facilities by PNNL, Dominion, and P. F. Weck. PNNL and Dominion analyzed the storage work procedure of the VSC (Vertical Storage Cask) method using CASTOR V/21, TN-32, respectively, and conducted a radiological impact assessment. P. F. Weck analyzed the storage work procedure of various spent nuclear fuel casks for VSC and HSM (Horizontal Storage Module), conducted a radiological impact assessment. As a result of comparing the procedure for storing spent nuclear fuel by case, it was found that the storage procedure was determined by the storage method and the cask type. In the case of VSC method, canister-type casks and basket-type casks are used, and the storage procedure are partially different according to each. Canister-type cask requires repackaging from transfer overpack to storage overpack, but basket-type cask doesn’t require that procedure. In the case of the HSM method, only the canister type cask was found to be used. However, the storage procedure was different depending on the type of HSM system. Depending on the type of HSM system, the necessity of cask for on-site transport was different. In this study, we investigated and analyzed the work procedure according to the storage method of dry storage facilities abroad. It was found that the dry storage procedure of spent nuclear fuel different according to the storage method and type of cask. The results of this study can be used as basic when deriving the exposure scenario for spent nuclear fuel dry storage workers suitable for the domestic situation.

To rationalize the protection of spent nuclear fuel transport storage cask, we intend to investigate the status of domestic and foreign safety regulations and related technologies to develop sabotage scenarios and analyze the protection performance and radiation impact of transport storage cask. It is essential to conduct an aircraft collision safety evaluation on spent nuclear fuel transportation and storage casks in Korea due to changes in laws and regulations related to nuclear power plant design and demand for enhanced safety. Domestic and foreign research on the protection performance of spent nuclear fuel transport storage cask was based on 9.11 events, and the results of all studies show that the speed of the aircraft and leakage of nuclear materials are insignificant. The Sandia National Laboratory (SNL) calculates Aerosol emissions from spent fuel damage in the event of sabotage and calculates Source Term based on the Durbin-Luna model. In this paper, radiation sensitivity analysis was performed due to damage to the carrier according to the size of the accident, assuming that there was a hole enough to basket from the external shell among the collision scenarios identified for domestic cask models.

The mechanical safety of the container designed according to the IP-2 type technology standard was analyzed for the temporary storage and transportation of Very-Low-Level-Waste (VLLW) for liquid occurring at the nuclear facilities decommissioning site. The container was designed and manufactured as a composite shielding container with the effect of storing and shielding liquid radioactive waste using High Density Polyethylene (HDPE) and eco-friendly shielding material (BaSO4) with corrosion and chemical resistance. The main material of the composite shielding container is HDPE and BaSO4, the material of the cover, cage and pallet is SUS304, and the angle guard is elastic rubber. The test and analysis requirements were analyzed for structural analysis of container drop and lamination test. As test requirements for IP-2 type transport containers should be verified by performing drop and lamination tests. There should be no loss or dispersion of contents through the 1.2 m high free-fall drop and lamination test for a load five times the amount of transported material. ABAQUS/Explicit, a commercial finite element analysis program, was used for structural analysis of the drop and lamination test of the transport and storage container. (Drop test) It was confirmed that the container was most affected when it falls from a 45-degree slope. Although plastic deformation was observed at the edge axis of the cover, it was evaluated that the range of plastic deformation was limited to the cover and cage, and stress within the elastic limit occurred in the inner container. In the analysis results for other falling direction conditions, it was evaluated that stress within the elastic limit was generated in the inner container except for minor plastic deformation. In the case of on-site simulation evaluation, deformation of the inner container and frame due to the drop impact occurred, but leakage and loss of contents, which are major evaluation indicators, did not occur. (Lamination test) The maximum stress was calculated to be 19.9 MPa under the lamination condition for a load 5 times the container weight, and the maximum stress point appeared at the corner axis of the pallet. The calculated value for the maximum stress is about 10%, assuming the conservative yield strength of SUS304 is 200 MPa. It was evaluated that stress within the limit occurred. In the case of on-site simulation evaluation, it was confirmed that there was no container deformation or loss of contents due to the load.

To reduce the environmental burden caused by the disposal of spent nuclear fuel and maximize the utilization of the repository facility, waste burden minimization technology is currently being developed at the Korea Atomic Energy Research Institute (KEARI). The technology includes a nuclide management process that can maximize disposal efficiency by selectively separating and collecting major nuclides in spent nuclear fuel. In addition, for efficient storage facility utilization, the short-term decay heat generated by spent nuclear fuel must be removed from the waste stream. To minimize the short-term thermal load on the repository facility, it is necessary to separate heat generating nuclides such as Cs-137 and Sr-90 from the spent fuel. In particular, Sr-90 must be separated because it generates high heat during the decay process. KAERI has developed a technology for separating Sr nuclides from Group II nuclides separated through the nuclide management process. In this study, we prepared Sr ceramic waste form, SrTiO3, by using the solid-state reaction method for long-term storage for the decay of separated Sr nuclides and evaluated the physicochemical properties of the waste form. Also, the radiological and thermal characteristics of the Sr waste form were evaluated by predicting the composition of Sr nuclides separated through the nuclide management process, and the estimation of centerline temperature was carried out using the experimental thermal data and steady state conduction equation in a long and solid cylinder type waste form. These results provided fundamental data for long-term storage and management of Sr waste.

The conventional research trend on spent fuel was safety analysis based on mechanical perspective. Analysis of spent fuel cladding is based on the temperature of cladding and pressure inside cladding. To improve fuel cladding analysis, precise and accurate thermal safety evaluation is required. In this study a database which is about thermal conductivity and emissivity for the thermal modeling was established for a long-term safety analysis of spent fuel. As a result, we confirmed that the thermal conductivity of zirconium hydride was not accounted in conventional model such as FRAPCON and MATPRO. The conductivity of zirconium and its oxide was evaluated only as a function of temperature. However, the behavior of heat conductivity and emissivity is determined by the change of the material properties. The material properties depend on the microstructural characteristic. It can be seen that this conventional approach does not consider the microstructure change behavior according to vacuum drying process or burn-up induced degradation phenomena. To improve the thermal properties of spent nuclear fuel cladding, the measurement experiments of heat conduction and emissivity are required according to spent fuel experience and status such as the number of vacuum drying, cooling rate, burn up, hydrogen concentration and oxidation degree. In previous domestic reports and papers, we found that relative data between thermal properties and spent fuel experience and status does not exist. Recently, in order to understand the failure mechanism of hydrogen embrittlement, many studies have been conducted by accounting and spent fuel experience and status in a mechanical perspective. If microstructure information could be obtained from these studies, the modeling of thermal conductivity and emissivity will be possible indirectly. According to a recent abroad paper, it was confirmed that the thermal conductivity decreased by about 30% due to irradiation damage. The radiation damage effects on thermal conductivity also has not been studied in zirconium oxide and hydride. These un-revealed phenomena will be considered for the thermal safety model of spent fuel.

To dry storage of spent nuclear fuel withdrawn the wet storage, all moisture inside the dry storage container must be removed to ensure the long-term integrity and retrievability. Substantial amounts of residual water in dry storage container may have potential impacts on the fuel, cladding, and other components in the dry storage system, such as fuel degradation and cladding corrosion, embrittlement, and breaching. The drying could perform as a vacuum drying process or a forced helium dehydration process. In NUREG-1536, the evacuation of most water contained within the canister is recommended a pressure of 0.4 kPa (3 torr) to be held in the canister for at least 30 minutes while isolated from active vacuum pumping as a measure of sufficient dryness in the canister. Monitoring the moisture content in gas removed from the canister is considered as a means of evaluating adequate dryness. Dew point monitoring and special techniques could be used to evaluate this adequacy. Various studies are continuing for quantitative evaluation of residual moisture inside the dry storage system. Andrawes proposed a methodology for determining trace water contents in gaseous mixtures, utilizing gas chromatography together with a helium ionization source. A microwave plasma source and emission spectrometry were utilized to determine trace amounts of bound water in solid samples using peak areas of atomic oxygen (O) and hydrogen (H) emissions. Bryans measured the gas samples taken from the High Burn-Up Demonstration Cask at three intervals: 5 hours, 5 days, and 12 days after the completion of drying and backfilling in the North Anna power Station. To measure water content, a Vaisala humidity probe was used. Final results indicated that the cask gas water content built up over 12 days to a value of 17,400 ppmv ±10%, equivalent to approximately 100 g of water within the entire cask gas phase. Tahiyats also proposed a methodology that involves a direct current (dc) driven plasma discharge and optical emission spectroscopy for detecting and quantifying water vapor in a flowing gas stream under both trace and high water vapor loading conditions. For detecting water vapor concentration, the emission from H at 656.2 nm was employed. The H emission is the red visible spectral line generated by a hydrogen atom when an electron falls from the third lowest to the second lowest energy level, this suggests that the normalized H intensity can be used as a marker for water vapor detection and quantification. Several of the attempts are continuing to quantify water contents in dry storage system. Lessons learned by Case studies would be provided insights into how to improve future measurements.

In order to construct and operate the dry storage systems, it is essential to confirm the safety of the systems through safety analysis. If the dry storage cask is damaged due to an accident, a large amount of radioactive material may be leaked to the outside and cause radiation exposure to surrounding workers and nearby public, so the effect thereof should be evaluated. Many input parameter are required in the confinement evaluation for accident condition, and in this study, the change in the confinement evaluation result according to the change of major input parameter is to be studied. In this study, we selected fractions of radioactive materials available for release from spent fuel, cooling time, and distance to exclusive area boundary as the major input parameter. In general, the release fraction suggested by NUREG-1536 has been used, but NUREG-2224 provides the fraction for high burn-up spent fuel in fire and impact accident conditions, unlike NUREG-1536 which provide a single value. In the case of the distance to exclusive area boundary, 100 to 800 m was considered, and in the case of the cooling time, 10 to 50 years was considered in this study. In order to compare the dose change by the parameter, we set up the hypothetical storage system. A storage cask of the system contain 21 PWR spent fuel assemblies with an initial enrichment of 4.5wt%, burnup of 45,000 MWD/MTU. During the accident condition, it is assumed that the cask is leaked at 1.0×10−7cm3·sec−1. Since the main dose criterion for accident conditions is 50 mSv of effective dose, effective doses are calculated in this study. In an accident condition, transuranic particulate contribute most of the doses, so the doses are determined according to the fraction for the particulate. Therefore, it was confirmed that the dose was almost the same as the fraction for the accident conditions in NUREG-1536 and the fraction for the impact accident conditions in NUREG-2224 is 3×10−5, but the dose was also 100 times higher as the fraction for the fire accident conditions in NUREG-2224 is 3×10−3. In the case of the cooling time, it was confirmed that the dose change according to the cooling time was not significant because the dose contribution of transuranic elements having very long half-life was very large. In the case of the distance, it was confirmed that the dose decreased exponentially as the atmospheric dispersion factor decreased exponentially with the distance.

The design of nuclear fuel storage and handling area includes the activities related to the storage and inspection before fuel loading, transfer into the reactor, removal of irradiated fuel to the spent fuel storage rack, underwater handling and storage, and handling into a shipping cask. The purpose of this study is to provide the design requirements for the spent fuel pool to be prevented from the loss of cooling water and for heavy load control to prevent any load drop resulting in damage to safetyrelated systems during heavy load handling in accordance with the regulatory guidelines. And another purpose is to review the sizing of minimum wet storage capacity in the spent fuel pool based on the maximum refueling batch from the core during refueling plus a full core off-load of fuel assemblies and the minimum discharge burnup spent fuel storage during the design life of plant requested by the utility. As the results of this study, the current general arrangement for the spent fuel storage and handling area and the minimum storage capacity are evaluated. These can be good recommendations to enhance more safe and efficient if implemented to the new nuclear power plants.

The spent fuel dry storage canister is generally made of austenitic stainless-steel and has the role of an important barrier to encapsulate spent fuels and radioactive materials. The canister on the dry storage system has several welding lines in the wall and lid, which have high residual tensile stresses after welding procedure. Interaction between stainless steel and chloride environment from a sea results in an aged-related degradation phenomenon causing chloride induced stress corrosion cracking (CISCC) in the dry storage system. A pending issue to the interim storage of spent fuel awaiting repository disposal is their susceptibility to CISCC of stainless steel canisters. The available mitigation technology should be studied sufficiently to prevent the degradation phenomenon. This paper assesses stress-based mitigation to control residual tensile stress practically applicable to the atmospheric CISCC for the aging management of the stainless steel canisters. There are major components, that is, elevated tensile stress, susceptible material and corrosive environment that must be simultaneously present for CISCC degradation to occur. Surface stress improvement can effectively mitigate the potential for CISCC of the canister external surfaces. The potential deleterious effect of the additional work is negated by the presence of compressive residual stress, which removes the tensile stress needed for CISCC to occur. Surface stress improvement methods such as shock-based peening, shot peening and low plasticity burnishing can be applied for surface stress improvement of the canisters. Stress relaxation processes and advanced welding methods such as laser beam welding and friction stir welding can be also available to mitigate the susceptibility to CISCC. As the result assessing the stress-based mitigation technologies, promising candidate methods could be selected to reduce the residual tensile stresses and to control an aged-related degradation condition causing CISCC in the spent fuel dry storage canister.

Concrete structures of spent nuclear fuel interim storage facility should maintain their shielding ability and structural integrity during normal, off-normal and accident conditions. The concrete structures may deteriorate if the interim storage facility operates for more than several decades. Even if deterioration occurs, the concrete structures must maintain its unique functions (shielding and structural integrity). Therefore, it is necessary to establish an analysis methodology that can evaluate whether the deteriorated concrete structure maintains its integrity under not only normal or off-normal condition but also accident condition. In accident conditions such as tip over and aircraft collision, both static material properties and dynamic properties of the concrete are required to evaluate the structural integrity of the concrete structures. Unlike the calculated damage results for the static deformation of the concrete structure, it is very difficult to accurately estimate the damage values of the degraded concrete structures where an aircraft collides at a high strain rate. Therefore, the present authors have a plan to establish a database of the dynamic material properties of deteriorated concrete and implement to a Finite Element Analysis model. Prior to that, dynamic increase factors described in a few technical specifications were investigated. The dynamic increase factor represents the ratio of the dynamic to static strength and is normally reported as function of strain rate. In ACI-349, only the strain rate is used as a variable in the empirical formula obtained from the test results of specified concrete strengths of 28 to 42 MPa. The maximum value of dynamic increase factor is limited to 1.25 in the axial direction and 1.10 in the shear direction. On the other hand, in the case of the CEB model, static strength is included as variables in addition to the strain rate, and a constitutive equation in which the slope changes from the strain rate of 30 /s is proposed. As plotting the two dynamic increase factor models, in the case of ACI, it is drawn as a single line, but in the case of CEB, it is plotted as multiple lines depending on the static strength. The test methods and specimen sizes of the previously performed tests, which measured the concrete dynamic properties, were also investigated. When the strain rate is less than 10 /s, hydraulic or drop hammer machines were generally used and the length of the specimens was more than twice the diameter in most cases. However, in the case of Split Hopkinson Pressure Bar tests, the small size specimens are preferred to minimize the inertia effect, so the specimens were small and the length was less than twice the diameter. We will construct the dynamic properties DB with our planned deteriorate concrete specimen test, and also include the dynamic property data already built in the previous studies.

The amount of temporarily stored spent nuclear fuel in South Korea will be reaching saturation in a near future. Therefore, it is an urgent issue to construct a spent nuclear fuel storage system. In order to construct the storage system, some coastal environmental characteristics such as temperature, pH, and chemical composition of sea water in South Korea have to be evaluated and predicted because they can affect in deterioration of the storage system. However, in South Korea, the coastal environmental characteristics of area where the storage system is likely to be built are not well established until now. In this study, a time-series deep-learning algorithm is developed using the Long-Short Term Memory (LSTM) algorithm to predict and evaluate the coastal environmental characteristics based on the wellestablished data from Korea Meteorological Administration (KMA) and Ministry of Oceans and Fisheries (MOF). As a result, by developing the predictive model to evaluate the coastal environmental characteristics, we intend to apply it for site evaluation to construct the spent nuclear fuel storage system or many other applications related to the nuclear as well.

In South Korea, the master plan for high-level radioactive waste management, announced in 2016, suggested the construction and operation of intermediate storage facilities on a permanent disposal site and specified the adoption of dry storage in consideration of the ease of operation and expansion. As of 2021, the government is again reviewing its overarching policy on the back-end fuel cycles, including intermediate storage and permanent disposal. In the case of dry storage facilities, safety evaluation is being conducted using a combination of deterministic and probabilistic approaches, similar to that of nuclear power plants. The two methods are complementary, of which Probabilistic Safety Assessment (PSA) has the advantage of being able to identify key scenarios affecting safety, but its use in storage facilities has not been highlighted so far. However, depending on the spent fuel management phases such as loading, transportation, and storage, it may be not enough to capture effective and efficient safety evaluation only deterministically, and probabilistic methods may contribute to the evaluation of long-term operation or external events such as an earthquake. There have already been cases where PSA has been performed on a part of the nuclear fuel cycle through previous studies. This paper created the safety assessment model based on open sources such as the released EPRI reports, by targeting arbitrary intermediate storage facilities. The model considered the scenarios for loading, transportation, and storage, with human error respectively. It will be able to be modified and improved to fit domestic and specific intermediate storage facilities in the future.

This presentation summarizes recent research on estimating the mechanical loading environment of spent nuclear fuel (SNF) during normal storage and transportation scenarios sponsored by the US Department of Energy Spent Fuel and Waste Science and Technology (SFWST) program. Normal conditions of truck, ship, and railroad transportation of SNF were studied with testing and numerical modeling to determine that the shock and vibration loads applied to SNF during transportation are not expected to challenge SNF cladding integrity or the fatigue life of cladding. The 30 cm package drop scenario was studied with experiments and modeling to determine that mechanical loads during a 30 cm SNF package drop scenario are only expected to challenge SNF cladding integrity under worstcase conditions at elevated temperatures. The SFWST program is currently preparing seismic shake table testing to record SNF mechanical loads in a dry storage earthquake scenario. This presentation summarizes the findings of the transportation and package drop research and details the progress made on the current seismic test.

Sandia National Laboratories is the lead laboratory for the United States Department of Energy for the research and development (R&D) efforts to support the technical basis for the long-term storage, subsequent transportation, and permanent disposal of commercial spent nuclear fuel and high-level waste. Sandia does not design nuclear facilities; Sandia performs R&D to help ensure facilities and the fuel cycle are safe, sustainable, and secure. This talk will focus on the spent fuel storage and transportation programs that contribute to this work. The goal in spent fuel storage and transportation R&D is to understand the mechanical integrity of the fuel, cladding, and storage system beyond interim storage and into disposal time frames. Our research is focused on understanding the high burn-up cladding integrity over time, understanding the thermal behavior during drying and storage, understanding potential cladding oxidation pathways, and quantifying in the external loads experienced during transportation, handling, and seismic events. Additionally, this work includes extensive work to understand the basic science of canister stress corrosion cracking and the potential consequences of a through wall canister crack.

Meat affects color and quality by metabolite concentrations. Meat produces metabolites, and metabolites are caused by a variety of causes. Meat also produces metabolites by oxidation, which is an inevitable chemical process that meat undergoes which is resulting information of various chemical compounds. Thus, the aim of this study was to profiling the change of metabolites of M. longissimus lumborum during the storage at 4°C. Instrumental color measurements were showed decreasing chroma value, redness and yellowness (P<0.05) during storage, while non-significance (P>0.05) changes found in lightness value. Above all, hue angle was highest at 21 d of storage (P<0.05). The lipid and protein oxidation of muscles was measured by TBARS value significantly increased (P<0.05), thiol and carbonyl groups were also increased significantly (P<0.05) during the display. Total 19 of 60 identified compounds appeared to have a significant difference by storage time (P<0.05). Hue angle had a significant correlation with specific metabolites such as carbon disulfide, 3-methyl-1-butanol, 2-ethyl-1-hexanol, lactic acid and palmitic acid (P<0.05). Results of the current study provide the conversion of volatile and non-volatile metabolites and their correlation with oxidative indicators for changes in meat quality during aerobic storage.