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.

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.

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.

In the field of 3H decontamination technology, the number of patent applications worldwide has been steadily increasing since 2012 after the Fukushima nuclear accident. In particular, Japan has a relatively large number of intellectual property rights in the field of 3H processing technology, and it seems to have entered a mature stage in which the growth rate of patent applications is slightly reduced. In Japan, tritium is being decontaminated through the Semi-Pilot-class complex process (ROSATOM, Russia) using vacuum distillation and hydrogen isotope exchange reaction, and the Combined Electrolysis Catalytic Exchange (CECE, Kurion, U.S.) process. However, it is not enough to handle the increasing number of HTOs every year, so the decision to release them to the sea has been made. Another commercial technology in foreign countries is the vapor phase catalyst exchange process (VPCE) in operation at the Darlington Nuclear Power Plant in Canada. This process is a case of applying tritium exchange technology using a catalyst in a high-temperature vapor state. The only commercially available tritium removal technology in Korea is the Wolseong Nuclear Power Plant’s Removal Facility (TRF). However, TRF is a process for removing HTO from D2O of pure water, so it is suitable only for heavy water with high tritium concentration, and is not suitable for seawater caused by Fukushima nuclear power plant’s serious accident, and surface water and groundwater contaminated by environmental outflow of tritium. Until now, such as low-temperature decompression distillation method, water-hydrogen isotope exchange method, gas hydrate method, acid and alkali treatment method, adsorption method using inorganic adsorbent (zeolite, activated carbon), separator method using electrolysis, ion exchange adsorption method using ion exchange resin, etc. have been studied as leading technologies for tritium decontamination. However, any single technology alone has problems such as energy efficiency and processing capacity in processing tritium, and needs to be supplemented. Therefore, in this study, four core technologies with potential for development were selected to select the elemental technology field of pilot facilities for treating tritium, and specialized research teams from four universities are conducting technology development. It was verified that, although each process has different operating conditions, tritium removal performance is up to 60% in the multi-stage zeolite membrane process, 30% in the metal oxide & electrochemical treatment process, 43% in the process using hydrophilic inorganic adsorbent, and 8% in the process using functional ion exchange resin. After that, in order to fuse with the pretreatment process technology for treating various water quality tritium contaminated water conducted in previous studies, the hybrid composite process was designed by reflecting the characteristics of each technology. The first goal is to create a Pilot hybrid tritium removal facility with 70% tritium removal efficiency and a flow rate of 10 L/hr, and eventually develop a 100 L/hr flow tritium removal system with 80% tritium removal efficiency through performance improvement and scale-up. It is also considering technology for the postprocessing process in the future.

Developments in cancer therapies and diagnostic techniques have improved the long-term survival of cancer patients. Certain cancer treatments, such as radiotherapy, often harm normal tissue as well as the specifically targeted cancer cells. High doses of radiation induce bone loss. This study investigated the effects of pentoxifylline (PTX) on radiation-induced bone loss in C3H/HeN mice. C3H/HeN mice were divided into sham and irradiation (3 Gy, gamma-ray, IR) groups. The irradiated mice were treated for 12 weeks with vehicle, PTX (p.o.) or PTX (s.c.). Grip strength, uterus weight, serum alkaline phosphatase (ALP) and tartrate-resistant acid phosphatase (TRAP) level were measured. Tibiae were analyzed using micro-computed tomography. There were no significant differences in the degree of grip strength, body weight and uterine weight between IR group and PTX-treated group. Treatment of PTX significantly preserved trabecular bone volume, trabecular number, trabecular separation and bone mineral density of proximal tibia metaphysic. The administration of PTX lowered serum TRAP in IR mice, suggesting that PTX can reduce the bone resorptive rate in mice. Our experimental data support the protective role of PTX against bone loss in irradiated mice. Based on the findings of this study, development of PTXbased treatments is anticipated to address bone loss after radiotherapy. Prospective dose escalation studies are required to determine the appropriate dosage of PTX.

Magnesium hydroxide sulfate hydrate (MHSH) whiskers were synthesized via a hydrothermal reaction by using MgO as the reactant as well as the acid solution. The effects of the H2SO4 amount and reaction time at the same temperature were studied. In general, MHSH whiskers were prepared using MgSO4 in aqueous ammonia. In this work, to reduce the formation of impurities and increase the purity of MHSH, we employed a synthesis technique that did not require the addition of a basic solution. Furthermore, the pH value, which was controlled by the H2SO4 amount, acted as an important factor for the formation of high-purity MHSH. MgO was used as the raw material because it easily reacts in water and forms Mg+ and MgOH+ ions that bind with SO4 2- ions to produce MHSH. Their morphologies and structures were determined using X-ray diffraction (XRD) and scanning electron microscopy (SEM).

본 연구에서는 국내외 저탄소 녹색성장을 위한 대안으로서 수소에너지와 그 이용 기술에 대한 관심이 높아지는 추세에 발맞춰 무탄소 연료인 수소를 LNG 의 주성분인 메탄, 메탄-프로판, 메탄-프로판-에탄 동축류 확산화염 내에 첨가하여 화염형상 및 연소생성물에 미치는 영향을 확인하였다. 상온상압 조건의 확산화염에 수소를 단계적으로 첨가하여 실제 생성되는 연소생성물의 변화 추이를 가스 분석기를 이용하여 실험적으로 관찰하였고 확산화염의 형상은 디지털카메라를 이용하여 단계적으로 관찰 하였다. 실험결과에서 확산화염에 수소를 첨가함에 따라 질소산화물의 생성량이 선형에 가깝게 증가하는 경향을 보였다. 이것은 수소의 상대적으로 높은 단열화염온도와 빠른 연소속도가 Thermal NOx의 생성을 촉진했기 때문이다. 반면 이산화탄소의 생성량은 감소하는 경향이 나타났는데 수소를 첨가함에 따라 메탄, 메탄-프로판, 메탄-에탄-프로판의 혼합 확산화염에 포함되어있는 전체 탄소비율이 줄어들어 이산화탄소의 생성량이 감소한 것이다. 이는 선박에서 LNG-수소의 혼합 연료사용으로 인해 온실가스인 이산화탄소를 저감할 수 있는 하나의 방안으로 고려될 수 있다는 것을 의미한다.