Within the air purification system of a nuclear power plant, specific radioactive isotopes are extracted from gases through adsorption onto activated carbon. To properly dispose of used activated carbon, it is essential to determine the concentration of radioactive nuclides within it. This study discusses the application of the pyrolysis method for analyzing the concentrations of 3H and 14C in spent activated carbon. The pyrolysis was conducted using Raddec’s Pyrolyser, with adjustments made to parameters such as temperature profiles, airflow rates, sample quantities, and trapping solution volumes. The evaluation method for the pyrolysis of activated carbon to analyze 3H and 14C involved adding 3H and 14C sources to the activated carbon before use and subsequently assessing the recovery rates of the added sources in comparison to the analysis results.

In order to establish disposal plans for sludge, which is one of the untreated waste materials from domestic nuclear power plants, it is necessary to determine the radioactivity concentration of radioactive isotopes. In this study, we aim to evaluate the gross alpha radioactivity of sludge containing radioactive contaminants after pre-treatment, in order to assess the level of sludge waste and obtain analytical data for discussing disposal methods. Samples of sludge generated from nuclear power plants were pre-treated, solutionized, and prepared as analysis samples for evaluating the gross alpha radioactivity.

Typically, the bottom of the effluent treatment facility at a nuclear power plant contains sediment, which is low-contamination waste consisting of sludge, gravel, sand, and other materials from which radioactive contaminants have been removed. Among these sediments, sludge is an irregular solid form consisting of small particles that are coagulated together, with radioactive isotopes containing cobalt attached. Currently, there is a record of disposing of dry active waste from domestic nuclear power plants, and efforts are underway to gather basic data for the disposal of untreated waste such as sludge, spent filter, and spent resin. In particular, the classification and disposal methods of waste will be determined based on the radioactivity concentration. Therefore, plans are being made to determine the radioactivity concentration of radioactive isotopes and establish disposal plans for sludge samples. In this study, pre-treatment and solutionization were carried out for the analysis of radioactive isotopes in sludge sampels from nuclear power plants. The deviation of the gamma radioisotope analysis results was derived to obtain an optimal sample quantity that represents the sludge.

For the disposition of radioactive wastes generated from nuclear power plant, radioisotope inventory must be analyzed to determine an activity concentration of radionuclides. Radionuclides in low- and intermediate-low-level of radioactive wastes, however, can be easily classified to easyto- measure (ETM) and difficult-to-measure (DTM) nuclides. ETM nuclides are gamma emitting nuclides that is relatively easy to measure because they do not need to be destroyed for the preprocessing. On the other hands, DTM nuclides are alpha and beta emitting nuclides that need to be destroyed for the preprocessing and also need chemical separation. Currently, measurement methods for DTM nuclides are developed and in this paper measurement methods of Fe-55, Ni-59, Ni-63, Sr-90 and Tc-99 will be introduced.

To analyze the radioactivity of 3H and 14C in miscellaneous radioactive wastes generated from nuclear power plants, a wet digestion method using sulfuric acid is currently used. However, sulfuric acid is classified as a special management material, and there is no disposal method for contaminated radioactive waste. Therefore, research on a thermal decomposition method that can analyze the DAW radioactive waste samples without using sulfuric acid is necessary. In this study, we will cover the final sample amount, sample injection method, and prevention of organic ignition to meet the minimum detection limit requirements of the analysis equipment. Through this research, optimal conditions for the thermal decomposition method for analyzing the radioactivity of 3H and 14C in DAW radioactive wastes generated from nuclear power plants can be derived.

For the final disposal of radioactive waste, concentration of gamma nuclides such as Co-58, Co-60, Cs-137, Nb-94 have to be determined to meet nuclear regulatory requirements. In general, gamma nuclide analysis can be performed with simple sample pretreatment without complicated chemical separation processes due to the characteristics of the nuclide and high resolution of the measuring equipment. However, when the concentration of Co-60 is high in a specific radioactive waste generated at the NPP, the background is increased by the compton continuum of Co-60. That makes it difficult to evaluate accurately Nb-94, which is in the lower energy band than the gamma ray energy region of Co-60 and especially Cs-137, which is used as a key nuclide of scaling factor. In this study, We consider the problem of MDA dissatisfaction or overestimation due to the increased background by Co-60.

When self-disposing of radioactive waste, it is important to follow the acceptable concentration standards for each nuclide set by the Nuclear Safety and Security Commission (NSSC). Gamma-emitting nuclides can be easily analyzed with a simple pretreatment process, but beta-emitting nuclides require a chemical separation procedure to be analyzed for radiochemistry analysis. When analyzing betaemitting nuclides for the purpose of self-disposal, there may be difficulties in radiation detection after the chemical separation process. This is because the concentration of beta nuclides in the sample may be low and some of them may be lost during the chemical separation. Therefore, measurement method of gross-beta activity can be used instead of that of each nuclide to access the compliance of selfdisposal criteria. While a proportional counter is commonly used to measure gross-beta activity, liquid scintillation counting can also be used to measure gross-beta, and we plan to compare the results of both methods.

Many radionuclides emit two or more gamma rays in a cascade once they decay. At this time, gamma rays are detected at the same time, and the signals are overlapped and measured as one added signal. This is called the summing coincidence effect, and it causes an error of more than 10% depending on the detection efficiency, measurement conditions, and target nuclide. It is known to be greater as the efficiency of the detector increases and as the distance between the source and the detector decreases. It is necessary to consider the summing coincidence effect since the efficiency of the HPGe detector owned by the KHNP CRI is as high as 65%. In this study, We would like to propose an appropriate gamma nuclide analysis method for radioactive waste generated from NPP by evaluating the influence on the summing coincidence effect.

To analyze the activity concentration of radionuclides in radioactive sludge samples generated from low- and intermediate-low-level radioactive waste from domestic nuclear power plant, a pretreatment process that dissolves and homogenizes the sample is essential. However, this pretreatment process requires the use of hydrofluoric acid, which makes analysis difficult and challenges users to handle harmful chemicals. Therefore, we aim to minimize the use of hydrofluoric acid by measuring gamma nuclides in the sludge sample without pretreatment process and compare the differences of measurement results according to the sample matrix with and without pretreatment process. We will collect about 0.1 g of the sludge sample, and dissolve it using an acid treatment process after using microwave decomposition. We will then use gamma spectroscopy to check the concentration of nuclides present in the sludge before and after dissolution and consider the effect of the sample matrix.

To optimize initial cooling conditions, forced-air cooling was applied to freshly harvested oak mushrooms at 2 levels (0oC for 30 minutes, at 0oC for 1 hour) followed by room cooling at 3 levels (-3oC for 1 day, 0oC for 1 day, 3oC for 1 day). After initial cooling, the oak mushrooms were packaged with PVC film, then held in a storage room at 1oC for 6 weeks. Quality characteristics and percentage marketability were then investigated. As a control, Mushrooms were placed in storage with no initial cooling. The quality factors impacting marketability of fresh oak mushrooms were color change and appearance of decay. Off-odor did not occur or developed only slightly, so it did not affect oak mushroom quality within 6 weeks of low temperature storage. In all treatment groups, the shelf life in which 100% marketability was maintained was up to 3 weeks. At week 5, percent marketability of the 3 treatment groups 1 hour room cooling treatment at 0°C, 1 hour forced air cooling, and control was 100%. 80% In the group that underwent 30 min forced air cooling retained 80% marketability, and the group exposed to 1 day in room cooling at -3oC retained 86.7% marketability. At week 6 of 1oC storage, the marketability ratio was 80% in the 1 day room cooling at 0oC group, 66.7% in the 1 day room cooling at 3oC group, 46.7% in the 1 hour forced air cooling group, and 33% or less in all other treatment groups. Therefore, the most suitable initial cooling parameter to extend shelf-life of oak mushrooms is 1 day of in room cooling at 0oC immediately after harvest.

Radioactive wastes that are generated as a result of operating NPPs, contain 63Ni and 59Ni that should be analyzed in accordance with the notice of Nuclear Safety and Security Commission (NSSC) for the acceptance of Korea Radioactive Waste Agency (KORAD). Analyzing 63Ni and 59Ni has few challenges to determine activities of each nuclide in radioactive waste sample that contains both nuclides. As is well known, 63Ni can be analyzed by liquid scintillation counter (LSC) detecting its emitted beta rays, however, beta rays emitted from 59Ni are overlapped on the spectrum. Therefore, to discriminate those two nuclides, spectrum channel should be divided according to its dedicating part of the spectrum. For instance, 59Ni contribute to spectrum channel 30–250, on the other hand, 63Ni contributes to spectrum channel 30–450. In other word, 63Ni solely can be analyzed on the channel from 260 to 450. To analyze both 63Ni and 59Ni using this channel division method, detection efficiency must be measured in advance; efficiency of 63Ni and 59Ni at ch. 30–250, and efficiency of 63Ni at ch. 260–450, then the activity can be calculated using the corresponding efficiency. In this study, for verifying the feasibility of channel division method, 5 simulated samples were prepared with different ratio of 63Ni/59Ni. The ratio varies as 1, 2, 10, 20 and 100 spiking standard source of 63Ni and 59Ni. Each sample was mixed with scintillation cocktail and detected for 90 minutes by LSC (300SL, Hidex) after the stabilization of solutions. As a result, calculated 63Ni activities for all sample were averaged as 97% of spiked activity. However, calculated 59Ni activity were 101%, 103%, 128%, 140%, 260%, respectively. The result indicates that 59Ni cannot be discriminated by channel division method when it exists in the sample with high 63Ni over 10 times then 59Ni such as radioactive waste sample. However, the results also show that the channel division method for analyzing 63Ni activity was successful verifying it can determine the activity of 63Ni regardless of the affect of 59Ni on the spectrum.

Radioactivity of radiostrontiums, Sr-89 and Sr-90, which are both pure beta-emitters, are generally measured via Cherenkov counting. However, the determination of Cherenkov counting efficiencies of radiostrontiums requires a complicated procedure due to the presence of Y-90 (also a pure betaemitter) which is the daughter nuclide of Sr-90. In this study, we have developed a machine learning approach using a linear regression model which allows an easier and simultaneous determination of the Cherenkov counting efficiencies of the radiostrontiums. The linear regression model was employed because total net Cherenkov count (Ct) from the three beta-emitters at time t after the separation of Y- 90, can be expressed as a linear combination of their respective time-varying radioactivities with their respective coefficients (parameters) being their counting efficiencies: Ct = εSr-90[ASr-90·exp(–λSr-90·t)] + εSr-89[ASr-89·exp(–λSr-89·t)] + εY-90[ASr-90·exp(1–λSr-90·t)], where ε is a counting efficiency, A is an initial activity, λ is a decay constant and t is time after the separation of Y-90, Thus, if we train the model with multiple Cherenkov counts measured from the three beta emitters, then we can obtain their estimates for counting efficiencies (so-called parameters) straightforward. For this, the model has been trained by two methods: Ordinary Least Squares (OLS) and Bayesian linear regression (BLR), for which two software packages, PyMC3 and Stan were employed to compare their performances. The results showed that the accuracy of the OLS was worse than that of the BLR. Particularly, the counting efficiency of Sr-90 was estimated to be smaller than 0, which is an unrealistic value. On the other hand, the estimates of the BLR gave realistic values which are close to the true values. Additionally, the BLR was able to provide a distribution for each counting efficiency (so-called “posterior”) from which various types of inference can be made including median and credible interval in the Bayesian statistics which is analogous to, but different from confidence interval in the Frequentist statistics. In the results of the BLR, the Stan package gave more accurate estimates than the PyMC3 package. Therefore, it is expected that counting efficiencies of the radiostrontiums including radioyttrium can be determined at the same time, more easily and accurately, by using the BLR with the Stan package and that the activities of radiostrontium also can be determined more easily by using the BLR if we know their counting efficiencies in advance. It is worth noting that the usage of the linear regression model in this study was different from the usual one where the trained model is used to predict a response value (count) from a set of unseen regressor values (activities).

큰느타리의 수출시 유통 기한 연장을 목적으로 수확시 품질 등급, 대와 자실체 사이를 잘라낸 손질 처리 유무, 그리고 관행 OPP 봉지에 끈묶음한 포장 방법을 개선시켜 트레이용기에 넣은 후 밀봉처리시 효과를 구명하고자 하였다. 수확시 품질 등급은 수확전 재배사의 온도를 9~11˚C 낮추어 적응시킨 버섯을 특품으로, 관행 13~15˚C로 적응시킨 버섯을 상품으로 설정하였다. 선별한 특품과 상품 버섯을 이용하여 손질 및 포장방법으로 3처리구를 두었다. 첫째는 절단 손질 후 OPP 봉지에 넣어 끈묶음한 포장(Cut & OPP), 둘째는 손질하지 않고 OPP 봉지에 넣어 끈묶음한 포장(Uncut & OPP), 마지막으로 개선포장방 법으로 절단 손질한 후 트레이용기에 넣고 밀봉한 포장 (Cut & Tray)이었다. 포장 완료한 버섯 처리구를 0℃ 저장고에 42일간 보관하면서 포장 내부의 기체 조성, 품질 요인의 변화, 신선 품질에 대한 관능평가를 실시하였다. 특품과 상품의 버섯 모두 Cut & Tray, Cut & OPP, 그리고 Uncut & OPP 처리 순으로 전반적으로 신선도가 높게 유지되었다. 특품 버섯의 유통 수명은 Cut & Tray 처리의 경우 30일, Cut & OPP 처리의 경우 28일, Uncut & OPP 처리의 경우 21일이었고, 상품 버섯의 유통 수명은 Cut & Tray 처리시 22일, Cut & OPP 처리시 17일, 그리고 Uncut & OPP 처리시 14 일이었다. 신선 버섯의 품질에 영향을 미치는 요인은 갓과 대의 갈변과 부패 지수였다. 특히, 버섯 대의 아랫부분의 갈변과 그에 연관된 표피색 a*값과 b*값의 변화가 품질 저하의 주요인이었다.

신선 버섯의 수확 후 관리 기술은 2000년대까지 출원의 증가와 감소를 반복하는 경향을 보이다가 2010년대에 들어서 꾸준한 증가세를 나타내고 있어, 신선 농산물의 수확 후 관리 기술이 시대의 흐름에 부합하는 기술임을 알 수 있다. 주요 출원인에 대한 특허출원은 일본이 전체 출원 중에서 가장 많은 특허를 출원하였으며 대부분의 국가에서 내국인의 특허출원 비중이 높게 나타나 자국 중심의 출원 경향을 보이고 있다. 신선 버섯의 수확 후 포장 기술은 초반에 출원인 및 출원 건수가 감소하였다가, 최근 출원인 및 출원 건수가 증가하여 회복기 단계에 있으며, 저장기술은 최근 출원인 수와 출원 건수가 대폭 증가하여 성장기의 단계로 평가된다. 신선 버섯의 수확 후 관리 기술은 주로 다국적 기업과 국가 연구기관에서 다수의 특허를 보유하고 있는 것으로 나타났으며, 이는 해당 기술의 연구개발 및 상업화가 활발하게 이루어지고 있음을 시사하였다.

본 연구는 몇가지 전처리제 및 보존용액 처리가 국내 육성 절화 프리지아 ‘Gold Rich’의 수확후 절화수명과 개화 품질에 미치는 영향을 구명하고자 수행하였다. 10% sucrose, 200mg･L-1 Al2(SO4)3, 200mg･L-1 GA3, 150mg･L-1 citric acid가 혼용된 전처리 용액에서 절화 관상 기간 동안 수분 흡수량과 생체중은 유의적으로 증가한 것으로 조사되었지만 절화수명에는 큰 영향을 미치지 않았다. 보존용액으로 5% sucrose와 살균제인 300mg･L-1 8-HQ를 혼용 처리하였을 때 절화수명 연장(1일 이상), 높은 개화율(82.3~89.1%) 및 만개율(71.5~84.2%) 등의 개화 품질이 향상되었다. 이러한 결과들을 통해 10% sucrose, 200mg･L-1 Al2(SO4)3, 200mg･L-1 GA3, 150mg･L-1 citric acid를 혼합한 전처리제와 5% sucrose, 300mg･L-1 8-HQ를 혼합한 보존용액 처리는 절화수명 연장, 개화율 및 만개율 증가에 효과적일 것이라고 판단된다. 특히, sucrose가 포함된 보존용액은 절화 프리지아 ‘Gold Rich’ 의 상품성 향상에 있어 매우 유용할 것이다.