석유 정제 시설 등에서 발생하는 함유폐수의 처리는 폐수의 유류 허용한계를 넘기지 않기 위해 중요한 공정이다. 세라믹 멤브레인은 유류 처리에서의 높은 효율, 내화학성, 내열성, 기계적 안정성, 그리고 단순한 작동 원리 등의 장점을 가지 고 있어 함유폐수의 처리에 효과적이다. 그러나 세라믹 멤브레인은 원재료의 높은 가격 때문에 널리 사용되는 데에 한계가 있다. 최근에는 이를 해소하기 위해 플라이 애시나 점토를 사용하는 노력도 있었다. 이 리뷰는 세라믹 멤브레인의 효율과 제 작을 실리콘, 알루미나, 그리고 폐석탄회의 재료로 나누었다.

췌석은 만성췌장염에서 흔히 동반되는 소견으로 반복되는 복통과 췌장염의 원인이 될 수 있다. 췌석은 췌관 협착과 자주 동반되고 췌관내 박혀 있는 경우가 많아 바스켓을 이용하여 제거하기 어렵다. 체외충격파쇄석술(ESWL)로 치료하기도 하나 반복 시술이 필요하고 성공률도 높지 않았다. 최근에 개발된 SpyGlass™ DS II (Boston Scientific, Marlborough, MA, USA)는 직경이 3.5 mm로 가늘어 췌관이 확장되어 있을 때 안으로 삽입이 가능하게 되었다. 그리고 직접 췌석을 보면서 전기수압쇄석술(EHL)이나 레이저 유도 쇄석술(laser lithotripsy)를 시행하며 췌석을 제거해 볼 수 있게 되었다. 본고에서는 SpyGlass™ DS II와 EHL을 이용하여 10 mm 이상의 다발성 췌석을 제거하는 방법을 소개하고자 한다.

EBV-positive mucocutaneous ulcer (EBVMCU) is an indolent, superficial lymphoproliferative disorder that occurs in either iatrogenic or age-associated immunocompromised patients with latent Epstein-Barr virus (EBV) infection. Although EBVMCU is histologically similar with other lymphoproliferative disorders, such as EBV-positive diffuse large B cell lymphoma, the diseases are classified as distinct entities by the World Health Organization with different clinical manifestations, prognosis, and genetic profiles. EBVMCU commonly shows spontaneous regression by conservative management, reduction or cessation of immunosuppressive treatment, but local progression is possible. Complete remission of disease can be achieved with combination of surgical resection, chemotherapy and local radiation therapy. Herein, we report 2 cases of oral superficial lymphoproliferative disorders arising adjacent to the gnathic bone with striking differences in prognosis and bone involvement. One of the cases induced extensive osteomyelitis in the underlying bone. Furthermore, we discussed the differential diagnosis of EBVMCU and reviewed the former literature.

Malignant melanoma is a highly malignant tumor derived from melanocyte. Malignant melanoma of the oral cavity occurs mainly in the palatine mucosa and the maxillary gingiva in men in their 50s. Malignant melanoma can be divided into pigmented and non-pigmented(amelanotic). Among them, non-pigmented malignant melanoma accounts for 2-8% of all malignant melanomas. Pigmented malignant melanoma is detected through changes in pigmentation, whereas non-pigmented malignant melanoma is characterized by no pattern of color change. In this study, at the initial visit, a malignant lesion was suspected and a biopsy was performed. According to the biopsy results, it was diagnosed as polymorphic sarcoma, but the histological examination performed during the operation revealed that it was amelanotic melanoma. As such, the differential diagnosis is important because there is no clinical change in non-pigmented malignant melanoma. Diseases to be differentially diagnosed when non-pigmented malignant melanoma occurs in the oral cavity include squamous cell carcinoma, lymphoma, sarcoma, inflammation, and osteomyelitis. In this study, we report a case that showed the histopathological characteristics of malignant melanoma without superficial pigmentation.

For the safety assessment of the high-level radioactive waste (HLRW) disposal, the thermodynamic data such as solubility products, formation constants of complexes, redox equilibrium constants of radionuclides, and their reaction enthalpy and entropy are required. In order to recommend and summarize the reliable data, thermodynamic databases (TDB) have been persistently developed through the OECD-NEA TDB projects and an updated TDB of actinides has been recently published in 2020. To date, reliable data for Pu reactions are scarce due to the possibility of coexistence of four different oxidation states, Pu(III-VI) by redox equilibria in solutions. To determine the thermodynamic data for the reaction of each Pu oxidation state, it is necessary to precisely control the oxidation state and quantitatively analyze all reactants, products and bi-products by using highly sensitive speciation techniques. Since 2004, the nuclear chemistry research team in KAERI has been focused on developing techniques for the sensitive chemical speciation by using laser-based spectroscopy and determining thermodynamic data of actinides such as U, Pu, Am. In this paper, chemical speciation and thermodynamic studies on Pu in KAERI are reviewed. A combination of a commercial spectrophotometer and a capillary cell was adopted for a sensitive chemical speciation of Pu(III-VI) in solutions. A sensitive detection of trace amount of Pu colloids was carried out with the laser-induced breakdown detection (LIBD) system. Pu(VI) complexation with hydroxide or carbonate ions were investigated under strong oxidation conditions controlled with hypochlorite (NaOCl). The solubility product of Pu(OH)3(am) and formation constant of Pu(III)-OH speices were determined by a combination of wet-chemistry experiments and several analysis methods of spectrophotometry, LIBD, radiometry under a strong reducing condition controlled by electrochemistry. More recently, we reported the reaction enthalpy and entropy data for the formation of Pu(OH)2+ and the dissolution of Pu(OH)3(am). A preliminary data for reaction between Pu(III) and organic matter will be presented.

The correlation between accident management plan and radiation emergency plan of Shin-Kori Units 3 and 4 was compared and analyzed from the point of view of the adequacy of facilities, equipments, organization and manpower which are necessary for the related emergency response. It was found the equipment of accident management plan and emergency response facility of radiation emergency plan had different technical contents and scope of application, so there was no risk of mutual conflict and overlapping functions. However, since the accident impact assessment code in accident management plan and computer program of radiation emergency plan were different, it was necessary to ensure the agreement or linkage of the evaluation between them. When a radiation emergency is issued in accident management plan, the composition and mission of the accident response organization were mostly consistent with the contents of the radiation emergency plan, but some corrections and improvement items were identified. Accident management plan specified that the disaster response safety center belonged to the emergency operations facility (EOF), but the radiation emergency plan did not mention it at all. The main tasks of disaster response safety center were the movement, arrangement and connection of mobile emergency response facilities, on-site construction of other emergency response facilities, and on-site road restoration. According to the accident management plan, the movement, arrangement, and connection of mobile facilities (i.e., mobile generators, mobile pumps, multi-purpose communication relay facilities), which were considered very important for the prevention and mitigation of serious accidents, were under the supervision of the disaster response safety center. It was stipulated that the operation was carried out with the cooperation of a regular emergency organization, and that the start, operation and stop of mobile equipments were to be performed under the supervision of the emergency operation team supported by the regular emergency organization. Since this organization structure and assignment of duties could not be confirmed in radiation emergency plan, it was necessary to revise and improve the radiation emergency plan for the successful operation of mobile equipments and to link them with the accident management plan.

In accordance with the notification of the Nuclear Safety and Security Commission (NSSC), environmental impact assessments around nuclear power plants are conducted annually and the results are disclosed to the public. The effects of direct radiation exposure from nuclear power plants as well as liquid effluents and gaseous effluents are taken into consideration in the evaluation of dose calculation for residents. In the United States, regulatory guidelines on direct radiation exposure are described in Reg. Guide 4.1, and the effects of direct radiation are evaluated through regulatory guidelines in Korea. We are going to review optimal evaluation method by reviewing the direct radiation exposure evaluation method currently being conducted in domestic nuclear power plants and the direct radiation exposure evaluation method in overseas nuclear power plants such as in the United States.

When the leakage of radioactive material or radiation to the environment or a concern, it is important to accurately understand the impact on the environment. Therefore, environmental effects evaluation using modeling based on meteorological data and source-term data is carried out, or environmental radiation monitoring which is an emergency response activity that directly measures dose is performed. As lessons learned from the Fukushima accident, environmental effects evaluation and modeling cannot utilize during the emergency and decision-making process for protective action for the public. Thus, rapid environmental radiation monitoring is required. In Korea, when an emergency is issued at a nuclear facility, urgent environmental radiation monitoring is conducted based on the national nuclear emergency preparedness and response plan, which can provide important information for decisionmaking on public protective actions. A review of strategies for urgent environmental radiation monitoring is important in performing efficient emergency responses. The main purpose of urgent environmental radiation monitoring is to gather data for decisionmaking on public protective actions to minimize the damage from the accident. For effective data collection and distribution, support from the national and local government and local public organizations and radiation expertise groups, and nuclear facility licensee are required. In addition, an emergency environmental radiation monitoring manual is required to immediately perform environmental monitoring in an emergency situation. The manual for emergency monitoring should include the activities to be conducted according to the phases of the emergency. The phases of the emergency are divided into pre-leakage, post-leakage, intermediate, and recovery. The reasons for establishing strategies are government and public information, the implementation of urgent population protection countermeasures, predicting and tracking plume trajectory, and detection of any release, the protection of emergency and recovery workers, the implementation of agricultural countermeasures and food restrictions, the implementation of intermediate- and recovery-phase countermeasures, contamination control. Besides meteorological data, ambient dose rate and dose, airborne radionuclide concentration, environmental deposition, food, water, and environmental contamination, individual dose, and object surface contamination data are also required for making information for the public.

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.

In this research, KPS manufactured Full System Decontamination (FSD) equipment, which is consisted of Oxidizing Agent Manufacturing System (OAMS), Chemical Injection System (CIS), RadWaste Treatment System (RWTS), Chemical Waste Decomposition & Treatment System (CWDS) and conducted demonstration test to prepare Decontamination and Decommissioning (D&D) project of Kori nuclear power plant in Korea. Each equipment of FSD was modularized due to the limited size of equipment hatch of Kori nuclear power plant. To simulate the expected circumstances in nuclear power plant such as usage of heater or position of each equipment, additional equipment was used. The chemical concentration and flow rate of process water for FSD were used as mentioned in the previous study by KHNP CRI. FSD was conducted for three cycles and each cycle was consisted of oxidation, reduction, chemical decomposition and purification. Oxidation and reduction process were conducted at 90°C. Chemical decomposition and purification process were conducted at 40°C due to the damage of UV lamp and IX by the heat. Total volume of process water for FSD demonstration test was 2.5 m2. KPS conducted decontamination performance review by calculating thickness reduction and weight loss of installed specimen. Operational review was conducted as if FSD test was conducted in the field based on the result of demonstration test. One of the most prioritized features is the workers’ safety. Also, the appropriate position of equipment needs to be considered to meet the required specification of component.

In preparation for the decommissioning of Kori unit 1 of the nuclear power plant (NPP), new containers of package, transportation, and disposal are being developed that reflect the type, generation amount, and radiological characteristics of decommissioning waste. The containers under development have internal volumes of 1 m3 ~ 14 m3 and loading weights of 1 ton ~ 35 tons, which are larger in size and have a higher loadable weight compared to the 200 L drum and IP-2 type transport container currently being used for packaging and transporting waste. So, there is a limit to handling new containers using existing transport systems (cranes, spreaders, forklifts, transport vehicles, etc.). Therefore, in this study, the status of handling equipment in NPP and disposal facilities was reviewed, the flow from the generation to disposal of decommissioning waste was analyzed, and the possibility of handling new container or the necessity of introducing new systems were derived. Except for some high-dose/high-radioactive wastes among decommissioning wastes, all wastes are finally disposed of through decommissioning area, temporary storage facility, waste treatment facility, waste storage facility, and receipt and storage building. The decommissioning area, temporary storage facility, and waste treatment facility are newly established areas for the decommissioning and should be equipped with a spreader/crane with a lifting weight of 15 tons, 35 tons, and 40 tons in consideration of the weight of the package to be handled in the zone. The waste storage facility has a 7.5 tons crane, so it can handle only some of the lower weight of the 5 to 35 tons package that is expected to be handled. Therefore, additional installation of spreaders/cranes, each with a lifting capacity of 15 tons and 40 tons, is required. The maximum loading weight of forklifts and transport vehicles operating at NPP, and disposal facilities is 10 tons and 12.6 tons, respectively. To transport the package, the facility must additionally install 15 tons and 40 tons forklifts, and 40 tons transport vehicles. Since the lifting weight of the crane installed on the transport vessel is also low at 12.5 tons, it is necessary to change the design of the existing or replace it with 40 tons to handle high-weight package. The results of this study will be used as basic data for the establishment of transport systems in the relevant area and facility, and design requirements for each equipment will be derived through additional research.

The goal of the decommissioning of nuclear facilities is to remove the regulations from the Nuclear Safety Act. The media that can be considered at the time of remediation stage may usually include soils, buildings, and underground materials. In addition, underground materials may largely be the groundwater, buried pipes, and concrete structures. In fact, it can be seen that calculations of the Derived Concentration Guideline Level (DCGL) and ALARA action levels was conducted in the case of overseas decommissioning experiences of Nuclear Power Plants (NPPs). Therefore, the aim of this study is to review the remediation activities and scenarios applied for the calculation of ALARA action level from the overseas decommissioned nuclear power plants. Media that can be considered for DCGL calculation at the time of license termination may differ from site to site. If the DCGL for the target media was derived, whether additional remediation actions are required under the DCGL value from the ALARA perspective was identified by calculating the ALARA action levels in the case of the U.S. The activities to determine whether additional clean-up is justified under the regulatory criteria are remediation actions which is dependent on the material contaminated. Therefore, the typical materials that can be subjected to remediation are soils and structure basements in the overseas cases. Remediation actions involved in the decommissioning process on the structure surfaces can be typically considered to be scabbling, shaving, needle guns, chipping, sponge and abrasive blasting, pressure washing, washing and wiping, grit blasting, and removal of contaminated concrete. For the cost-benefit analysis of the media subject to DCGL calculation, it is necessary to assume a scenario for the remediation actions of the target media. The scenarios can be largely divided into two types. Those are basement fill and building occupancy scenario. In basement fill mode, buildings and structures on the site are removed, and the effect of receptors from the contamination of the remaining structures is considered. In the building occupancy mode, it is assumed that the standing building remains on the site after the remediation stage. It is a situation to evaluate how the effect of additional remediation actions changes as the receptors occupy inside of the contaminated building. Therefore, parameters such as population density, area being evaluated, monetary discount rate, numbers of years, etc. can be set and assessed according to the scenarios.

This study introduces the licensing process carried out by the regulatory body for construction and operation of the 2nd phase low level radioactive waste disposal facility in Gyeongju. Also, this study presents the experience and lessons learned from this regulatory review for preparing the license review for the next 3rd phase landfill disposal facility. Korea Radioactive Waste Agency (KORAD) submitted a license application to Nuclear Safety and Security commission (NSSC) on December 24, 2015 to obtain permit for construction and operation of the national engineered shallow land disposal facility at Wolsong, Gyeongju. NSSC and Korea Institute of Nuclear Safety (KINS) started the regulatory review process with an initial docket review of the KORAD application including Safety Analysis Report, Radiological Environmental Report and Safety Administration Rules. After reflecting the results of the docket review, the safety review of revised 10 application documents began on November 29, 2016. Total 856 queries and requests for additional information were elicited by thorough technical review until November 16, 2021. As the Gyeongju and Pohang earthquakes occurred in September 2016 and November 2017, respectively, the seismic design of the disposal facility for vault and underground gallery was enhanced from 0.2 g to 0.3 g and the site safety evaluation including groundwater characteristics was re-investigated due to earthquake-induced fault. Also, post-closure safety assessments related to normal/abnormal/human intrusion scenarios were re-performed for reflecting the results of site and design characteristics. Finally, NSSC decided to grant a license of the 2nd phase low level radioactive waste disposal facility under the Nuclear Safety Laws in July 2022. This study introduces important issues and major improvements in terms of safety during the review process and presents the lessons learned from the experience of regulatory review process.

In 2022, new regulatory guidelines were announced in relation to the off-site dose calculation (ODC), and accordingly, measures to improve the off-site does calculation program (ODCP), kdose60, were reviewed. The main consideration is, first, that if multiple nuclear facilities are operated on the same site, the boundaries of the restricted areas shall be set as the overlapping outer boundaries of the restricted areas determined by calculation for each nuclear facility. Second, the external exposure caused by direct radiation from a number of nuclear facilities in the same site must be partially or fully applied depending on the facility and site characteristics. Third, the dose conversion coefficient should be evaluated by checking whether the effect of the daughter nuclides is properly reflected. Fourth, the soil contamination period is a factor to consider that radioactive substances deposited on the surface, such as particulate nuclides, affect residents over a long period of time. Fifth, due to the recent construction of Shin-Kori Units 5 and 6, there is a change in the site boundary of the Kori/Saeul site, so as the site boundary is expanded, it is required to add an exposure dose assessment point due to gas effluents and change the exposure dose assessment point according to crop intake. Therefore, through this study, the direction for improving the ODCP will be prepared by reviewing the recent revision of the regulatory guidelines.

There are various types of level gauging method such as using float, differential pressure, hypersonic, displacement and so on. In this study, among them, the method utilizing the differential pressure was reviewed. The strengths include: the differential pressure type level gauge can measure the level without direct contact of the sensor with media. That is to say, the level can be measured even if the sensor is far away from the tank. And regardless of the size of the tank, the level can be measured if the pneumatic pipes are installed. The weaknesses include: the sensor needs intermedium to recognize the level. The intermedium utilizes a fluid, which is compressed air. It is difficult to handle that compressed air has the properties of a gas. And to make compressed air needs compressor, tank and pneumatic pipes. So if you have many tanks, you need to install exponentially the pneumatic pipes. As well, level measurement range is limited to the points where the pneumatic pipes of the tank is installed. And if a compressed air that supplies to the sensor leaks, uncertainty will increase. A compressed air is colorless and odorless, so it’s difficult to pinpoint the leak. Finally, events like cracks and clogging can cause inaccurate measurement. Rather than using only differential pressure, it is better to use another measurement method according to the situation of the facility.

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.

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.