The primary objective of the present study is the characterization of the hybrids of dikaryon-monokaryon(di-mono) and monokaryon-monokaryon(mono-mono) crosses in mushroom breeding. We employed this technique for developing develop superior species from Pleurotus spp. varieties with 56 Di-mono intraspecific hybrids of 14 combinations and 85 mono-mono intraspecific hybrids of 7 combinations between six Pleurotus ostreatus varieties and one Pleurotus florida variety. In this study, the results of analysis on hybridization rate, nuclear and mitochondrial DNA patterns, and colors and yields of fruit-bodies, are presented as follows. In di-mono crosses, hybrids between Pleurotus ostreatus and Pleurotus florida showed 100% of crossability as seen in those between Pleurotus ostreatus and Pleurotus ostreatus indicating that nuclei and mitochondria of a dikaryon migrated to a recipient of monokaryon. The mitochondrial DNA patterns of the hybrid strain were composed of 75% dikaryon donors and 25% monokaryon recipient. The crossability between mono-mono crossing ranges between 50 and 93.75%. 82.4% of the hybrid strain showed mitochondrial DNA patterns predominated by either parent, while the remaining 17.6% had recombinant or half-and-half combined patterns of both parents.

The primary objective of the present study is the characterization of the somatic hybrids of dikaryon-monokaryon (di-mono) crosses in mushroom breeding. We employed this technique for developing superior strain from Pleurotus ostreatus strains with 56 intraspecific hybrids of 14 combinations between six Pleurotus ostreatus strains and one Pleurotus florida strain. In this study, the results of analysis on hybridization rate, nuclear DNA patterns, and colors and morphology of fruit-bodies, are presented as follows. In di-mono crosses, somatic hybrids among Pleurotus strains showed 100% of crossability as seen in those among Pleurotus strains indicating that nuclei of a dikaryon migrated to a recipient. 89.3% of the somatic hybrids among Pleurotus strains were similar to the donor dikaryons, and 10.7% had combined DNA patterns of both parents. In the 14.3% di-mono cross between P. ostreatus and P. florida, the nuclear DNA patterns of the all hybrid strain showed the same or similar patterns compared to the donor dikaryons. 75.0% of the hybrid between P. ostreatus and P. ostreatus were similar to the donor dikaryons; 10.7% had combined DNA patterns of both parents. 82.2% of fruiting body morphology of the hybrids among Pleurotus strains were similar to the dikaryons, and 17.8% had combined DNA patterns of both parents. All hybrid strains between dikaryon P. florida and monokaryon P. ostreatus showed the fruiting body whose colors were similar to those of the dikaryon, while the hybrids between dikaryon P. ostreatus and monokaryon P. florida were all showed combined colors of both parents but are more similar to the dikaryon. Therefore, the fruiting body color of P. florida tends to be generally dominant. The present study was able to find out and suggest superior hybrid strains by identifying the nuclear DNA patterns of hybrids between Pleurotus strains as well as the characteristics of their fruiting bodies. This study expects that the advantages of the di-mono crossing are needs to be fully utilized in mushroom breeding and it is better to develop superior strains of Pleurotus strains.

This study was conducted to develop superior hybrids of Pleurotus ostreatus with di-mono and mono-monoka-ryon crosses. Random Amplified Polymorphic DNA-PCR (RAPD-PCR) was used to compare its mitochondrial DNA profiles of hybrids using specific primers designed from microsatellite markers of Pleurotus salmoneo-stramineus. A total of fifty-six dikaryon-monokaryon hybrids were sampled for RAPD-PCR experiments and the results show that twenty-four hybrids were dikaryon and thirty-one hybrids were monokaryon. Interestingly, one hybrid was an intermediate form with the DNA profiles that are different from those of its parents. The DNA profiles from eighty-eight monokaryon-monokaryon hybrids were also analyzed by RAPD-PCR. The results of mitochondria DNA profiles show that seventy-one hybrids are the same to one of their parents, but seventeen hybrids show DNA profiles of both parents.

For safe disposal of radioactive wastes, accurate analysis of nuclear isotopes is important. It is known that there are 14 nuclides that have to identify nuclide-specific concentration levels. 63Ni, one of non-volatile nuclear isotopes which is included in those 14 nuclides, has to follow chemical separation for exact analysis. As various analysis methods were developed, various methods for analyzing 63Ni also emerged. Past method has used measurement specimens of 59Ni, after 59Ni measurement has been done. It used HClO4, known as strong oxidizing agent, to dissolve DMG, an organic substance used to form 59Ni precipitates. Nowadays, we analyze 59Ni and 63Ni simultaneously, which enables short analysis time, without use of HClO4. But high accuracy is just as important as short measurement time and efficiency. So, this paper compare 63Ni specific activity value used new method with the value, past method used, using real sample’s data. As a result, all sample data from new method’s relative 63Ni specific activity is within the uncertainty range of past ones based on past specific activity value. Consistency of new method’s result and past method’s data increased the reliability of the data and accuracy of those methods.

Concrete is the primary building material for nuclear facilities, making it one of the most common forms of radioactive waste generated when decommissioning a nuclear facility. Of the total waste generated at the Connecticut Yankee and Maine Yankee nuclear power plants in the United States, concrete waste accounts for 83.5% of the total for Connecticut Yankee and 52% for Maine Yankee. In order to dispose of the low- to medium-level radioactive concrete waste generated during the decommissioning of nuclear power plants, it is necessary to analyze the radioactivity concentration of gamma nuclides such as Co-58, Co-60, Cs-137, and Ce-144. Gamma-ray spectroscopy is commonly used method to measure the radioactivity concentration of gamma nuclides in the radioactive waste; however, due to the nature of gamma detectors, gamma rays from sequentially decaying nuclides such as Co-60 or Y-88 are subject to True Coincidence Summing (TCS). TCS reduces the Full Energy Peak Efficiency (FEPE) of specific gamma ray and it can cause underestimation of radioactivity concentration. Therefor the TCS effect must be compensated for in order to accurately assess the radioactivity of the sample. In addition, samples with high density and large volume will experience a certain level of self-shielding effect of gamma rays, so this must also be compensated for. The Radioactive Waste Chemical Analysis Center at the Korea Atomic Energy Research Institute performs nuclide analysis for the final disposal of low- and intermediate-level concrete waste. Since a large number of samples must be analyzed within the facility, the analytical method must simultaneously satisfy accuracy and speed. In this study, we report on the results of evaluating the accuracy of the radioactivity concentration correction by applying an efficiency transfer method that appears to satisfy these requirements to concrete standard reference material.

Alpha activities can be used for categorization, transportation, and disposal of radioactive waste generated from the operation of nuclear facilities including nuclear power plants. In order to transport and dispose of such low- and intermediate-level radioactive waste (LILW) to the Wolsong LILW Disposal Center (WLDC) at Gyeongju, the gross alpha concentration of an individual drum should be determined according to the acceptance criteria. In addition, when the gross alpha concentration exceeds 10 Bq/g, the inventory of the comprising alpha emitters in the waste is to be identified. Gross alpha measurements using a proportional counter are usually straightforward, inexpensive, and high-throughput, so they are broadly used to assay the total alpha activity for environmental, health physics, and emergency-response assessments. However, several factors are thoughtfully considered to obtain a reliable approximate for the entire alpha emitters in a sample, which include the alpha particle energy of a particular radionuclide, the radionuclide that is used as a calibration standard, the uniformity of film in a planchet, time between sample collection and sample preparation, and time between sample preparation and counting. Korea Atomic Energy Research Institute (KAERI) have evaluated the inventory of radionuclides in low-level radioactive waste drums to send every year hundreds of them to the WLDC. In this presentation, we revisit the gross alpha measurement results of the drums transported to WLDC in the past few years and compare them with the concentrations of alpha emitters measured from alpha spectrometry and gamma spectrometry. This study offers an insight into the gross alpha measurement for radioactive waste regarding calibration source, self-absorption effect, composition of alpha emitters, etc.

To achieve permanent disposal of radioactive waste drums, the radionuclides analysis process is essential. A variety of waste types are generated through the operation of nuclear facilities, with dry active waste (DAW) being the most abundant. To perform radionuclides analysis, sample pretreatment technology is required to transform solid samples into solutions. In this study, we developed a dry ashing-microwave digestion method and secured the reliability of the analysis results through a validity evaluation. Additionally, we conducted a comparative analysis of the radioactivity of 94Nb nuclides with and without the chemical separation process, which reduced the minimum detectable activity (MDA) level by more than 65-fold for a certain sample.

The Korea Atomic Energy Research Institute (KAERI) employs a methodology for evaluating the concentration of radionuclides, dividing them into volatile and non-volatile nuclides based on their characteristics, to ensure the permanent disposal of internally generated radioactive waste. Gamma spectroscopy enables the detection and radiation concentration determination of individual nuclides in samples containing multiple gamma-emitting nuclides. Due to the stochastic nature of radioactive decay, the generated radiation signal can interact with the detector faster than the detected signal processing time, causing dead time in the gamma spectroscopy process. Radioactive waste samples typically exhibit higher radiation levels than environmental samples, leading to long dead times during the measurement process, consequently reducing the accuracy of the analysis. Therefore, dead time must be considered when analyzing radioactive waste samples. During the measurement process, dead time may vary between a few seconds to several tens of thousands of seconds. More long dead time may also result in a temporal loss in the analysis stage, requiring more time than the actual measurement time. Long dead time samples undergo re-measurement after dilution to facilitate the analysis. As the prepared solution is also utilized in the nuclide separation processes, minimizing sample loss during dilution is crucial. Hence, predicting the possibility of dead time exceeding the target sample in advance and determining the corresponding dilution factor can prevent delays in the analysis process and the loss of samples due to dilution. In this study, to improve the issues related to gamma analysis, by using data generated during the analysis process, investigated methods to predict long dead time samples in advance and determining criteria for dilution factors. As a result of comparing the dead time data of 5% or long with the dose of the solution sample, it was concluded that analysis should be performed after dilution when it is about 0.4 μSv/h or high. However, some samples required dilution even at doses below 0.4 μSv/h. Also, re-measurement after dilution, the sample with a dead time of less than 32% was measured with less than 5% when diluted 10 times, and more than 32% required more than 10 times dilution. We suppose that with additional data collection for analyzing these samples in the future, if we can establish clearer criteria, we can predict long dead time samples in advance and solve the problem of analysis delay and sample loss.

Korea Atomic Energy Research Institute (KAERI) is planning to disposal of the radioactive contaminated cement waste form to the final disposal facility. The final disposal facility require evaluation of immersion, compressive strength, and radionuclide inventory of radioactive wastes to meet the acceptance criteria for safe disposal. According to the LILW acceptance criteria of the Nuclear Safety and Security Commission ok Korea (NSSC), the disposal limit radioactivity of 129I (3.70×101 Bq/g) is lower than other radionuclides. 129I emits low energy as its disposal limit is low, so it is difficult to analyze in the presence of 137Cs and 60Co which emit high energy. Therefore, it is essential to an accurately separate and analyze iodine in radioactive waste. In this study, we focused on the determination of 129I in cement waste form containing 137Cs, 60Co. We added 1 g of 129I(11.084 Bg), 137Cs(1,300 Bq) and 60Co(402 Bq) to cement waste form, respectively. The separation of 129I in cement waste form was carried out using an acid leaching method. And, we confirmed the specific activity of 137Cs and 60Co at each separation step. It was observed that an acid leaching method showed the remove efficiency 137Cs(99.97%) and 60Co(99.94%), respectively. In addition, 129I was also analyzed at approximately 96.44% in simulated contaminated cement waste form. In conclusion, through this experiment, it was confirmed that 129I could be successfully separated and analyzed by using the acid leaching method in cement waste form containing excessive 137Cs and 60Co.

The decommissioning of nuclear facilities produces various types of radiologically contaminated waste. In addition, dismantlement activities, including cutting, packing, and clean-up at the facility site, result in secondary radioactive waste such as filters, resin, plastic, and clothing. Determining of the radionuclide content of this waste is an important step for the determination of a suitable management strategy including classification and disposal. In this work, we radiochemically characterized the radionuclide activities of filters used during the decommissioning of Korea Research Reactors (KRRs) 1 and 2. The results indicate that the filter samples contained mainly 3H (500–3,600 Bq·g−1), 14C (7.5–29 Bq·g−1), 55Fe (1.1– 7.1 Bq·g−1), 59Ni (0.60–1.0 Bq·g−1), 60Co (0.74–70 Bq·g−1), 63Ni (0.60–94 Bq·g−1), 90Sr (0.25–5.0 Bq·g−1), 137Cs (0.64–8.7 Bq·g−1), and 152Eu (0.19–2.9) Bq·g−1. In addition, the gross alpha radioactivity of the samples was measured to be between 0.32–1.1 Bq·g−1. The radionuclide concentrations were below the concentration limit stated in the low- and intermediatelevel waste acceptance criteria of the Nuclear Safety and Security Commission, and used for the disposal of the KRRs waste drums to a repository site.

We established pretreatment method of solidified cement ion-exchange resin samples generated before 2003 in nuclear power plants for measurement of non-volatile radionuclide activity. A microwave digestion system (MDS) with mixed acid (HCl-HNO3-HF-H2O2) was used to dissolve cement and to desorb non-volatile elements such as Ce, Co, Cs, Fe, Nb, Ni, Re, Sr and U from mixed ion-exchange resin. The content of Ce, Co, Fe, Nb, Ni, Re, Sr, U and Cs after pretreatment of cement plus mixed ion-exchange resin was measured by ICP-AES and ICP-MS, respectively. As iron and strontium are also present in cement, their content after dissolving a certain amount of cement was measured by ICP-AES. All elements except Nb were quantitatively recovered. Especially since the Nb recovery was low at 72.0±2.5%, the MDS following addition of the mixed acid to the resin was operated once more for desorbing Nb from it. Finally the recovery of Nb was over 95%. This sample pretreatment method will be applied to solidified cement ion-exchange resin samples generated in nuclear power plants for assessment of radionuclide inventory.

We conducted multi-elements determination of reference material certified by the Inorganic Ventures, IV-26, using iCAP 7400 ICP-OES of Thermo Fisher Scientific. And we statistically evaluated analysis results by introducing the in-house proficiency evaluation method implemented at the Ministry of Food and Drug Safety. Ca, Co, Fe, Mg, Ni, and V were selected as target elements, and extended uncertainty was estimated at a confidence level of about 95% and coverage factor k = 2. Five parameters incurred at manufacturing process (standard solution, calibration curve, repeated measurement and dilution factor of the test sample) were considered when determining the uncertainty. En-score can be calculated using the formula En=(x-X)/(Ulab 2+Uref 2)1/2 described in KS Q ISO 13528, where x, Ulab, X, and Uref are the test results, the uncertainty of the result, and the certified value and the uncertainty of the value. And if the absolute value |En| is less than 1, it can be evaluated as a satisfied value. As a result of ICP-OES analysis, each concentration of the elements to be measured was almost similar to the certified concentration of the reference material, and the uncertainty was slightly different. Also since evaluation on multi-elements determination had an En-score within 1, it was confirmed that the analysis results satisfied En evaluation.

The massive amount of radioactive waste will generated during decommissioning of nuclear. Among the radioactive waste from these disposal process, 50-55 million tons of concrete waste are included. For safe disposal, it is very important to accurately analyze the concentration of radionuclides, especially 129I and 131I, contaminated concrete. 129I, a long-lived radioisotope of iodine (t1/2=1.57 × 107 y), and 131I (t1/2=8.04 d) are generated from the fission of uranium in nuclear reactors. In Korea, according to the Nuclear Safety and Security Commission (NSSC) radioactive clearance level guide, the limit for radioactive clearance level of 129I is less than 0.01 (Bq/g). Iodine can be absorbed, accumulate in organisms, and exhibit low energy emission compared with cesium, and cobalt. Therefore, it is essential to an accurately separate and analyze iodine radioactive waste. In this study, we focused on the determination of iodine in simulated cement waste form containing KI for the recovery of iodine. We performed cement waste form sieved through a different particle size (0.5 mm < ɸ < 6.35 mm). For the separation of iodine from solid samples with low iodine content, such as soil, sediment, and cement, for sample decomposition associated with solvent extraction using CHCl3 for separation of iodine from the matrix. The separation of iodine in cement waste particles was therefore carried out using an acid leaching method using KI containing cement particles. We observed that cement particle size decreased at 6.35 mm to 0.5 mm with iodine yield decrease at 0.840±0.011 to 0.582±0.010. Thus, in this study, the acid leaching method enables to determination Iodine in cement.

Combustion method has been widely used in the analysis of 3H and 14C in various types of radioactive wastes since X. Hou reported the analysis of 3H and 14C in graphite and concrete for decommissioning of nuclear reactor. In this work, it was demonstrated that the validation result of combustion method for the simultaneous analysis of 3H and 14C in non-combustible radioactive wastes. To validate the combustion method, 3H and 14C spiked to sea sand and soil were prepared and each simulated sample was combusted with the accordance to a combustion temperature program. The validation of combustion method was assessed by the radioactivity recovery, radioactivity deviation, and relative standard deviation of measured radioactivity. The results of radioactivity recovery, radioactivity deviation, and relative standard deviation of 14C were 100~105%, less than 7%, less than 5%, respectively. In addition, 3H showed about 90% of radioactivity recovery, less than 20% of radioactivity deviation, and less than 5% of relative standard deviation. It will be provided that the validation results of combustion method in detail.