Pseudo-pheochromocytoma is defined as an adrenal tumor that has the typical characteristics of pheochromocytoma likely hypertension but is histopathologically diagnosed to other types of adrenal tumors. 11-year-old, castrated male Maltese presented for a left adrenal mass with hypertension. According to clinical signs and radiological features, a temporary diagnosis was made of pheochromocytoma. The adrenalectomy was performed and the histopathological examination of the resected adrenal gland was diagnosed as cortical adenocarcinoma. This report describes the case of pseudo-pheochromocytoma in a dog and implies the salience that even if pheochromocytoma is diagnosed based on all test results, pseudo-pheochromocytoma should not be excluded.

Organisms constituting a large proportion of marine ecosystems, ranging from bacteria to fish, exhibit fluorescence and bioluminescence. A variety of marine organisms utilize these biochemically generated light sources for feeding, reproduction, communication, and defense. Since the discovery of green fluorescent protein and the luciferin-luciferase system more than a century ago, numerous studies have been conducted to characterize their function and regulatory mechanism. The unique properties of fluorescent and bioluminescent proteins offer great potential for their use in a broad range of applications. This short review briefly describes the functions and characteristics of fluorescent and bioluminescent proteins, in addition to summarizing the recent status of their applications.

The biological efficiencies of promising Pleurotus spp. were evaluated. Pleurotus ostreatus, Pleurotus tuberregium,and Pleurotus sajor-cajuwere investigated at temperatures of 22oC, 20oC, and 18oC to determine their ability to adapt to temperatures that are likely to be found in subtropical regions. The experiment was conducted using a substrate comprising Popular sawdust 60% + cotton seed 20% + beet pulp 20% under a relative humidity of 65%. The results indicated that there were significant differences in terms of biological efficiency between the species: P. tuber regium 128.84% at 22oC, P. ostreatus 108.41% at 22oC, and P. sajor-caju is 80% at 20oC. The common temperatures at which all species showed the highest biological efficiencies were 22oC and 20oC. Therefore, the production of the evaluated species was equal to or surpassed biological efficiencies reported in tropical environments, thereby demonstrating a potential opportunity to adapt to tropical environments without compromising quality standards.

Gamma spectrometry is one of the main analysis methods used to obtain information about unknown radioactive materials. In gamma-ray energy spectrometry, even for the same gamma-ray spectrum, the analysis results may be slightly different depending on the skill of the analyst. Therefore, it is important to increase the proficiency of the analyst in order to derive accurate analysis results. This paper describes the development of the virtual spectrum simulator program for gamma spectrometry training. This simulator program consists of an instructor module and trainee module program based on an integrated server, in which the instructor transmits a virtual spectrum of arbitrarily specified measurement conditions to the students, allowing each student to submit analysis results. It can reproduce a virtual gamma-ray energy spectrum based on virtual reality and augmented reality technique and includes analysis function for the spectrum, allowing users to experience realistic measurement and analysis online. The virtual gamma-ray energy spectrum DB program manages a database including theoretical data obtained by Monte Carlo simulation and actual measured data, which are the basis for creating a virtual spectrum. The currently developed database contains data on HPGe laboratory measurement as well as in-situ measurements (ground surface, decommissioned facility wall, radiowaste drum) of portable HPGe detectors, LaBr3(Ce) detector and NaI detector. The analysis function can be applied not only to the virtual spectrum, but also to the input measured spectrum. The parameters of the peak analysis algorithm are customizable so that even low-resolution spectra can be properly analyzed. The validity of the database and analysis algorithm was verified by comparing with the results derived by the existing analysis programs. In the future, the application of various in-situ gamma spectrometers will be implemented to improve the profiling of the depth distribution of deposited nuclides through dose rate assessment, and the applicability of the completed simulator in actual in-situ gamma spectrometry will be verified.

Nuclear power plants decommissioning is planned to be started in middle of the 2020. It is necessary to develop safety evaluation and verification technology during decommissioning to ensure the safety of security monitoring measures and maintenance measures, appropriate emergency plans and preparations for decommissioning, and the use of proven engineering when establishing decommissioning plan. For this purpose, a nuclear power plant decommissioning plan is prepared in several stages before decommissioning. When a lifetime of a nuclear power plant has reached, it needs to be decommissioned and therefore operator company should submit decommissioning plans to the National Safety and Security Commission. And safety analysis should be included in this document and it is explained in chapter 6. According to the NSSC Notice No. 2021-10, it is largely divided into principles and standards, exposure scenarios, dose assessment, residual radioactivity, abnormal events, and risk analysis. When unexpected radiological accident is happened, both public and occupational dose analysis should be conducted. However, research on the former can be found easily on the other hands, research on the latter is not active. In this paper, method of choosing scenarios of accidents during the decommissioning the nuclear power plants is briefly introduced. Accidents during nuclear power plants decommissioning cases in USA is chosen and its risk is evaluated by using risk matrix and ranked by AHP method. During the decommissioning phases, varieties of radioactive waste is expected to be generated such as contaminated concrete and metal. On the other hand, Dry Active Waste (DAW) is generated and its amount is and its amount is 7,353 drums. Characteristic of DAW is highly flammable compared to concrete or metal. Moreover, depending on method of radioactive waste conditioning and type of radioactive nuclides, release rate of the nuclides varies. Thus this type of radioactive waste is critical to fire accidents and such accident can occur extra dose exposure which exceeds the guideline of the regulatory body to workers. Therefore, in this paper, occupational dose exposure during the fire accident is conducted.

Radioactive source terms are important factor in design, licensing and operation of SMR (Small Modular Reactor). In this study, regulatory requirements and evaluation methodology for normal operation on NuScale SMR, which received standard design certification approval on September 11, 2020 from US NRC, are reviewed. The radioactive waste management system of nuclear power reactor should be designed to limit radionuclide concentration in effluents and keep radioactive effluents at restricted area boundary ALARA according to 10 CFR 20 and 10 CFR 50 Appendix I. Also, in general, the coolant source term to calculate the off-site radiological consequences for normal operation of SMR should be determined by using models and parameters that are consistent with regulatory guide 1.112, NUREG- 0017 and the guidance provided in ANSI/ANS-18.1-1999, and the result should be corrected by reflecting the design characteristics of SMR. The coolant source term of NuScale, unlike the case of large NPPs, cannot rely solely on empirical source term data, because the NuScale source term is based on first principle physics, operational experience from recent industry, and lessons learned from large PWR operation. Fission products in reactor coolant are conservatively calculated using first principle physics in SCALE Code assuming 60 GWD/MTU. The release of fission products from fuel to primary coolant based on industry operational experience is determined as fuel failure fraction of 0.0066% for normal operation source term and 0.066% for design basis source term while coolant source term of large NPP is calculated by using ANSI/ANS-18.1 for normal operation and fuel failure fraction of 1% for design basis source term. Water activation products in reactor coolant are calculated from first principles physics and corrosion activation products are calculated by utilizing current large PWR operating data (ANSI/ANS 18.1- 1999) and adjusted to NuScale plant parameters. Also, because ANSI/ANS 18.1-1999 is not based on first principle physics models for CRUD generation, buildup, transport, plate-out, or solubility, NuScale has incorporated lessons learned by using ERPI’s primary water chemistry and steam generator guidelines to ensure source term is conservative and design of materials used cobalt reduction philosophy to help ensure the coolant source term are conservative. Based on the coolant source term calculated according to the above-described method, the annual releases of radioactive materials in gaseous and liquid effluents from NuScale reactor are evaluated. Currently, Small Modular Reactors such as ARA, SMART 100 are under review for licensing in Korea. This study will be helpful to understand how the reactor coolant system source terms are defined and evaluated for SMR.

Organic complexing agents may affect the mobility of radionuclides at low- and intermediate-level radioactive waste repositories. Especially, isosaccharinic acid (ISA) is the main cellulose degradation product under high pH conditions in cement pore water. ISA can combine with radionuclides and form stable complexes that adversely influence adsorption in the concrete phase, resulting in radionuclides to leach to the near- and far-fields of repositories. This study focuses on investigating the sorption of ISA onto engineered barriers such as concrete, thereby studying adsorption isotherms of ISA on concrete and comparing various isotherm models with the experimental data. The adsorption experiment was conducted in three background solutions, groundwater (adjusted to pH 13 using NaOH), State 1 (artificial cement pore water, pH 13.3), and State 2 (artificial cement pore water, pH 12.5), in a batch system at a temperature of 20°C. Concrete was characterized using BET, Zeta-potential analyzer, XRD, XRF, and SEM-EDS. ISA concentrations were detected using HPLC. The experimental data were best fitted to one-site Langmuir isotherm; On the other hand, either two-site isotherm or Freundlich isotherm couldn’t give reasonable fitting to the experimental data. The observed ISA sorption behavior on concrete is crucial for the disposal of radioactive waste because it can significantly lower the concentration of ISA in the pore water. Although one-site Langmuir isotherm might effectively represent the sorption behavior of ISA on concrete, the underlying mechanism is still unknown, and further investigation should be done in the near future.

Cellulose-based wastes can be degraded into short-chain organic acids at the cementitious radioactive waste repository. Isosaccharinic acid (ISA), one of the main degradation products, can form the chelate complex with metals and radionuclides, and these complexes have a potential that can accelerate to move the radionuclides to far-field from the repository. This study characterized the amount of generated ISA from typical cellulosic materials in the repository. Two different degradation experiments were conducted under alkaline conditions (saturated with Ca(OH)2 at pH 12.4): i) cellulosic material mixture under an opened condition (partially aerobic), and ii) cellulosic material under an anaerobic condition in a nitrogen-purged glove box. In the first case, three different types of cellulosic materials–paper, cotton, and wood– were mixed at the same ratio, and the experiments were carried out at three different temperatures (20°C, 40°C, and 60°C). It revealed that both the cellulose degradation rate and generated ISA concentration were high at high reaction temperatures, and various soluble degradation products such as formic acid and lactic acid were generated. The cellulose degradation in this work seems to still stay at a peeling-off process. In the second study, each type of cellulosic material was applied in its own batch experiments, and the amount of generated ISA was in the order of paper > wood > cotton. The above two experiments are supposed to be a long-term study until the generated ISA reaches an equilibrium state.

There are generally two kinds of spent filter; one is spent filter media for mainly gaseous purification such as HEPA filter, the other is spent filter cartridge for liquid purification such as CVCS BRS cartridge type filter. The spent filter cartridge from liquid purification system has been storing in special shielding space in auxiliary building in NPPs since the beginning of 2006 according to the long term storage strategy for decaying short lived radionuclide and gaining the time for selecting practical treatment technology before final packaging. The spent filter cartridges generated Kori-1 reactor vary in their sizes as in length from 913 mm to 290 mm and range in radiation level from several hundred mSv per hour to below mSv per hour . It is high time that the spent filter cartridge is treated and packaged because LILW repository in Wolsung area is operating and Kori-1 reactor is scheduled to decommission. The spent filter cartridge is one of the wet solid wastes required of solidification. It is difficult for the spent filter cartridge to solidify because of their shape, structure, physical and chemical characteristics in addition to having high radiation level. NSSC notice defines that solidification of wet solid wastes include that solid material such as spent filter is encapsulated with cement, etc. as a form of macro-encapsulation. The radioactive waste acceptance criteria describes that non-homogeneous waste having above 74,000 Bq/g such as spent filter, dry active waste should be encapsulated with qualified material. Homogeneous waste such as spent resin, sludge, concentrated waste (liquid waste evaporator bottoms), etc. should be solidified complied with requirements except that spent filter which is allowed to encapsulate. It is needed to guide to the practice of these two requirements for spent filter. The sampling and test method is different between homogeneous solidification waste form and spent filter cartridge encapsulation waste form. For example, how core sample can be taken and how void space can be measured among spent filter cartridge in encapsulation waste form. The technical evaluation report for spent filter cartridge polymer encapsulation by US NRC has been reviewed and the technical position of US NRC was identified. As a result of review, improvement fields of waste acceptance criteria for spent filters are pointed out, and the technical position of US NRC for spent filter cartridge solidification is summarized. The recommendation on improvement directions for spent filter cartridge encapsulation is suggested.

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.