초록 ▼

1. 방사선 사고시 주요 환경내 확산영향 결정기술 개발
○ 기상청의 대용량 수치예보자료 자동 송ㆍ수신 프로토콜을 개발하고 수치예보자료 수신서버를 구축하여 대용량의 수치예보자료를 안정적으로 수신하고, 자료저장/검색 모듈 및 자료 색인화/동기화 DB 모듈을 개발함.
○ 기상자료를 이용한 대기 혼합고 산정 및 해양자료를 이용한 해양 성층고 산정 모듈을 개발함.
○ 지구규모 대기중 확산범위 예측을 위한 3차원 대기확산 수치모델 LADAS-Global을 개발하고, 후쿠시마 사고시 대기로 누출된 방사성핵종(Xe-133, I-1

1. 방사선 사고시 주요 환경내 확산영향 결정기술 개발
○ 기상청의 대용량 수치예보자료 자동 송ㆍ수신 프로토콜을 개발하고 수치예보자료 수신서버를 구축하여 대용량의 수치예보자료를 안정적으로 수신하고, 자료저장/검색 모듈 및 자료 색인화/동기화 DB 모듈을 개발함.
○ 기상자료를 이용한 대기 혼합고 산정 및 해양자료를 이용한 해양 성층고 산정 모듈을 개발함.
○ 지구규모 대기중 확산범위 예측을 위한 3차원 대기확산 수치모델 LADAS-Global을 개발하고, 후쿠시마 사고시 대기로 누출된 방사성핵종(Xe-133, I-131, Cs-137)의 확산분포 산정 및 관측자료와 검증을 수행함.
○ 대기중 침적 핵종농도 연계 지표면 오염도 평가를 위한 외부피폭모델 개발하고, 지구규모 Cs-137 지표면 침적에 의한 유효선량 평가 및 분석을 수행함.
○ 지구규모 해양환경내 확산범위 예측을 위한 3차원 해양확산 수치모델 LORAS-Global을 개발하고, 해양환경내 핵종농도 연계 초기 및 장기 선량평가 모델을 개발함.
○ 지역규모 해양환경내 핵종 확산모델 검증을 위한 수리모형장치를 설계하고 제작함.

2. 방사선 사고 생태환경 방호기술 개발
본 연구는 원자력사고시 생태환경 장단기선량평가기술개발, 국내고유 생태환경 핵종전이 특성인자 측정 및 자료 구축, 식물을 이용한 방사능오염환경완화기술개발, 야생생물종 방사선 생물학적 특성 조사 4개 세부목표로 연구가 수행되었다. 각 세부목표별 주요 연구개발결과는 다음과 같다.
○ 원자력사고시 생태환경 장단기선량평가기술개발
⦁ 국제 표준방법 (취급핵종 확대, 단계적 평가방법, 불확실도 평가 모듈 포함)의 생태계 선량평가코드를 아시아 최초로 개발
⦁ 코드의 성능검증을 위해 IAEA 국제공동연구에 참가하여 세계 유수의 코드와 시나리오 테스트 비교 연구 수행하였으며, 참가자들과 공동으로 SCI 급 논문 작성을 통해 코드의 세계화 추진 및 국제전문가 네트워크 구축
⦁ 후쿠시마사고 평가의 적용연구를 수행하여 코드의 실용성 입증
⦁ 유엔 산하 UNSCEAR 후쿠시마 사고 방사선 영향평가 전문가로 참여하여 국내코드를 활용한 국내 결과 입력 반영
○ 국내고유 생태환경 핵종전이 특성인자 측정 및 자료 구축
⦁ 국내 원자력시설 주변 생태계 (해양, 육수(민물), 강 하구 및 육상 생태계)로부터 야생동식물, 토양, 물 시료 채취
⦁ 월성 및 영광 원자력발전소 주변 반경 10km 이내 약 130 생물종에 대한 총 26 가지 주요 핵종의 전이인자 (약 5,500여개) 데이터베이스를 구축
○ 식물을 이용한 방사능오염환경완화기술개발
⦁ 방사능오염 환경복원을 위하여 기존 기술과 식물이용기술의 비교분석을 통하여 기술현안 및 후보작물 도출
⦁ 국립수목원, 충남대학교와 협력하여 국내 야생 수생식물의 발아 및 성장에 대한 국내전문기관 간 협력체계 구축
⦁ 식물 생장상 및 포트실험을 통해 후보작물의 오염 수 중 ¹³⁷Cs 흡수 특성 실험적 조사
○ 야생생물종 방사선 생물학적 특성 조사
⦁ 국내 생태계 방사선 생물지표(biomarker)를 개발하기 야생 등줄쥐를 모델동물로 선정
⦁ 실험에 소요되는 야생 등줄쥐 포획 및 교배를 통한 실험용 F1개체수 확보 (5마리/실험군)와 함께 분류학적 특성 조사
⦁ 방사선 조사 (0~3 Gy)에 따른 모델동물의 염색체 이상 및 손상 반응 자료 데이터 확보

3. 방사선환경방호체제 지원 인체피폭해석 최적화기술개발
○ 사고시 대기확산평가 최적화기술
⦁ 사고시 대기확산평가 모델간 특성 및 결과비교
⦁ 사고시 대기확산평가 최적모델개발
⦁ 입력변수 민감도분석 및 부지특성 최적 변수값 선정
⦁ 악기상조건 대기확산 실험설계 및 장비제작
○ 대기확산검증실험 및 사고시 인체피폭해석 최적화기술
⦁ 악기상조건에 대한 대기확산 검증실험(Ⅰ)
⦁ 국내 방사선방호평가체계에 기반한 사고시 인체선량평가 최적모델개발
⦁ 사고시 인체선량평가 최적 변수값 선정
⦁ 국내고유 지형‧환경적 특성에 기반한 평가결과해석기술
○ 국내 기술기준 기반 사고시 국내고유 인체피폭해석 기술개발
⦁ 악기상조건에 대한 대기확산 검증실험(Ⅱ)
⦁ 사고시 인체선량평가결과 종합비교검증
⦁ 국외 신기준의 변화에 따른 국내 원전영향분석
⦁ 사고시 인체피폭해석 전산프로그램 개발
⦁ 사고시 인체피폭평가 권고지침 작성

4. 비상대응 환경방사성유출물 탐지기술 개발
○ 연구개발 목표
⦁ 사고대응 고방사능현장 연속 분석시스템 개발
⦁ 주요 누출핵종 긴급분석기술 개발
⦁ 해양 방사성유출물 탐지기술 개발
○ 연구수행 내용 및 결과
⦁ 사고대응 현장 연속 소형 검출시스템 설계
⦁ 식품긴급 분석을 위한 전처리 기술개발
⦁ 해양시료 전처리 장치 개발
⦁ 현장 연속 소형 검출시스템 구축
⦁ 긴급분석을 위한 자동축차분리 시스템 구축
⦁ 해양시료 축차분리 기술개발
⦁ 현장 연속 소형 검출시스템 성능평가
⦁ 주요핵종 신속측정기술 개발
⦁ 해양시료 미량핵종 측정법 개발

5. 방사선방호용 선량측정 및 평가기술개발
○ 방사선 사고 대응 방사선량 회구 평가(retrospective dosimetry) 기술 확보
⦁ 개인 휴대형 전자기기 내부 소자를 이용한 방사선량 회구평가 기술 확보
- 최소검출선량: 10 mGy 이하(저항소자: 6.6 mGy, 인덕터소자: 2.4 mGy)
- 방사선량 회구 평가 절차서 개발
- 휴대전화 이용 선량회구평가 실증 실험 실시: 기준선량대비 5%-40% 오차 이내로 회 구 평가가능
⦁ 건축물 구성 물질을 이용한 방사선량 회구 평가 기술 확보
- 사고선량 평가를 위한 건축물 구성 물질의 분류 및 선정: 적벽돌, 타일, 변기
- 사고선량 신속 평가 프로토콜 개발: 최소검출선량 약 10 mGy
○ 방사성핵종 섭취 신모델 해석 및 적용체계 개발
⦁ OIR 내부피폭 평가모델을 분석하고 자체 개발한 잔류함수산출 모듈을 이용하여 OIR을 적용한 핵종별 섭취 잔류 및 배설함수 산출 및 데이터베이스 구축
○ 방사선방호/방재/보안용 측정기 성능시험 체계구축
⦁ 개인 및 주변 선량(률) 측정기 시험 및 교정을 위한 기준 엑스선장 확대 구축
⦁ 방사능방재/보안용 방사선/능 측정시스템 현장 성능시험 시스템 구축

(출처 : 보고서 요약서 3p)

Abstract ▼

IV. Results
1. Development of the assessment technology of dispersion effects in the environment for the radiological accident
In this study, a server was setup to receive the numerical forecast data without any data loss from KMA and preprocess/DB modules have been developed to treat and mana

IV. Results
1. Development of the assessment technology of dispersion effects in the environment for the radiological accident
In this study, a server was setup to receive the numerical forecast data without any data loss from KMA and preprocess/DB modules have been developed to treat and manage big data in connection with atmospheric and marine dispersion models. Especially, a data server has been operated to receive UM data to send in real time on the basis of MOU between KAERI and KMA in 2012. The operation of the atmospheric dispersion model has been simplified by development of the shell program to control the calculation date and time for automatic connection with the model.
Three dimensional atmospheric dispersion model named LADAS-Global has been developed to evaluate the effects to have an influence in Korea by the radionuclides released into the air after the Fukushima accident. Developed LADAS-Global was applied to simulate the dispersion patterns of Xe-133, I-131 and Cs-137 released into the air after the accident, and it was validated through comparing with the measurements at some points in the world and Korea. In the Fukushima accident 2011, the regional atmospheric dispersion model named LADAS-Regional in Korea was only operated for the Northeast Asian region and the radiological effects was evaluated to the human and environment due to the radioactive materials released into the air. Therefore, it is necessary to develop LADAS-Global of the global scale to evaluate the radiological assessment to have an influence on the human and environment in Korea by the radionuclides released into the air from the radiological accidents in the world. Radiological assessments including internal and external dose were performed in connection with the air concentrations, and the deposited concentrations on the surface were calculated.
Three dimensional marine dispersion model named LORAS-Global has been also developed to estimate the dispersion characteristics of the radionuclides released into the sea. LORAS-Global could simulate the concentrations of the radionuclide in the seawater, suspended matter and sediment. Also, it was constructed in parallel mode for fast simulation. The dispersion pattern of Cs-137 released into the sea after the Fukushima accident was simulated and predicted that the patch of radioactive materials could be reached near the south sea in Korea. The oceanic compartment model was also developed to understand the dose and cancer risks to have an influence on human from ingestion of the contaminated seafood by the radionuclides released into the sea. The comparative results showed a good agreement between the calculations of the model and measurements.
Hydraulic scale instrument based on the distortion and similarity theory was designed and manufactured to validate the regional marine dispersion model. It has included the detection instruments to measure velocity and height of solitary wave like tsunami. Wolsung nuclear site in hydraulic scale model was selected to investigate the phenomena of sedimentation, erosion, sinking of sea floor and inundation from tsunami. Hydraulic scale model was mainly composed of wave generator, rotating coastal structure, water tank, detective instruments of velocity and soil pressure.

2. Development of technologies to protect environment from the radiation risk of emergency exposures
2.1 Development of wildlife dose assessment code.
The Korean wildlife dose assessment code K-BIOTA has been upgraded, and its performance was tested through the analysis of the radiation risk of marine biota in the port of the FDNPS as a result of the Fukushima nuclear accident.
The upgraded version of the code can handle the 77 radionuclides to cover a wide variety of conceivable exposure situations including those arising from routine discharges, potential releases from repository for the radioactive waste, and a nuclear accident. Also, the function of screening level dose assessment is included in the new version of the code. The screening-level assessment is a preliminary step to determine whether the detailed assessment (Level 3 assessment) is needed. The upgraded version of K-BIOTA considers 11 groups of organisms that have similar habitats in three ecosystems (terrestrial, marine, and freshwater). For each group of organism, the screening assessment of two steps are performed step-by-step. Level 1 assessment calculates the risk quotients using the measured media concentration and pre-calculated environmental media concentration limit (EMCL) that have been derived from a bench-mark screening dose. If the calculated risk quotient is greater than 1, the assessment proceed to the level 2. Level 2 assessment calculates dose rates but allows the user to examine and edit some major parameters used in the calculation including concentrations ratios and distribution coefficients. In general, the level 2 assessment is less conservative than the level 1 assessment.
In level 3 assessment which is a site specific analysis, a module for the uncertainty analysis of dose rate on major model parameters (concentration ratio, Kd, and environmental medium concentration) has been added. Four probability density functions of normal, log-normal, uniform and exponential distribution and Latine Hypercube Sampling method are applied to the sensitivity analysis of major input parameters. The upgraded version of the code was registered on the Korea Copyright Commission (Reg No. C-2013-015027).
The code has been applied to analyze the radiation risk of marine biota in the port of the FDNPS as a result of the Fukushima nuclear accident. The analysis results showed that the accumulated dose for 3 months after the accident was about 4 to 4.5 Gy, indicating the possibility of occurrence of an acute radiation effect in the early phase after the Fukushima accident; however, the total dose rate for most organisms studied was usually below the UNSCEAR (United Nations Scientific Committee on the Effects of Atomic Radiation)’s bench mark level for chronic exposure except for the initial phase of the accident, suggesting a very limited radiological effect on the marine biota at the population level.

2.2 Development of comprehensive database for radionuclide transfer characteristic of domestic ecosystems.
During the first three years of the five-year research, a total of 5500 TF values were measured for about 130 wildlife species and 26 different elements through the collections of wildlife (plants and animals) and media (water and soil) samples in different types of ecosystems (marine, freshwater, estuary and terrestrial). The field works were carried out twice a year around the Wolseong and Younggwang NPPs. Wildlife samples were collected by means of trapping, netting, fishing, diving, shooting and hand-catching by employed divers and hunters, and self-catching using with or without instruments according to the characteristics of the organisms. Water samples were collected using a van Dorn sampler or collected directly into sample bottles. Soil was sampled using a cylindrical sampler of a height of 10 cm.
In the study of marine and freshwater ecosystems, about 1500 TF values were obtained for more than 40 wildlife species (fish, mollusca, crustacean, seaweed, vascular plant) and 22 nuclides (K, Ca, Na, Mg, Al, Fe, Ti, Sr, Mn, Cr, Cu, Zn, Li, Ni, Co, V, Rb, Cs, Ba, Pb, Th, U). In the study of estuaries and adjacent seas, more than 60 wildlife species were sampled near four estuaries located around the NPP sites, and a total of 2300 TF values were produced for 26 nuclides including K, Na, Sr, Mn, Zn, Co, Cs, U. The TF values were mostly higher than 1, reflecting the phenomenon of biological concentration.
Of the presently obtained TF values for the Korean aquatic ecosystems, the Al and U TF values for freshwater fish differs from the corresponding IAEA generic values (geometric menas) by factors of up to about several hundreds. Also for seawater and brackish-water organisms, there are many cases where differences by factors of several tens to several hundreds can be found. Therefore, it is important to use domestic or site-specific data for enhancing the reliability of the assessment.
In the study of terrestrial ecosystems, a total of 1700 TF values were measured for about 40 wildlife species including mammals, reptiles, amphibians, birds, insects, trees, shrubs, composites and grasses, and 24 nuclides (Ca, K, Mg, Na, P, S, Al, Sr, Fe, Mn, Li, Ba, Cr, Co, Zn, Th, U et al.). For terrestrial ecosystems, where soil acts as the medium for radionuclide transfer to wildlife, almost all of the measured TF values were lower than 1 except for Ca, Mg, P and S. Many TF values for terrestrial ecosystems also differed from the corresponding IAEA generic values by factors of several tens to several hundreds.

2.3 Development of phyto-technology to remedy the contaminated environment.
The state of art of the phytoremediation was investigated, and a number of criteria was developed to select an appropriate plant to be used to remove radionuclide from the environment contaminated with radionuclides. Based on the criteria, a few aquatic and terrestrial plants that are able to absorb radioactive cesium highly have been selected as the candidate plants for the experimental investigation during 2013 ~ 2016.
- Aquatic plants : Bulrush, Water parsley, Raupo, Wild rice, Lotus, Water hyacinth, Duckweed, Hydrilla verticillate
- Terrestrial plants: Poplar, Willow, Corn, Sorghum, Sunflower, Rice
The seeds of water parsley, raupo, and wild rice were germinated and maintained for a month in hydroponic conditions. One-month-old plants with some leaves and well-developed roots were selected for experiments on uptake of ¹³⁷Cs.
The ¹³⁷Cs uptake by water parsley, raupo, and wild rice was studied while cultivating the plants in a hydroponic medium. The result of the present study showed that more than 70% of the initial Cs in solution was removed by the candidate plants in 60 hours, indicating the potential possibility to remove ¹³⁷Cs successfully from a contaminated solution. The absorption of Cs by the candidate plants from the solution was more influenced by the pH than initial activity of Cs. Of all the studied candidate plants, the highest uptake of Cs was obtained by water parsley. The water parsley absorbed more than 98% of ¹³⁷Cs in solution for 120 hours,which can be considered as one of strong candidates of aquatic plant to remove of ¹³⁷Cs in solution.

2.4 Investigation of the biological characteristics on radiation of domestic wildlife.
Based on specific selection criteria such that it; has a small body structure; inhabits over widely-distributed areas; travels within a limited range; lives on local grains and has a high number of population, and taxonomy on the basis of external and skull forms, Apodemus agrarius coreae (A. a. coreae), which is a species of wild mouse was selected as a reference model animal for the study, and they have been bred in a laboratory to secure enough population for the experiment. A chromosomal aberration method was tested to evaluate the biological effects caused by radiation using commercial mouse strain of C57BL/6, and the same method was subsequently applied to F1 of A. a. coreae.
The obtained dose-effect for the chromosomal aberrations are
Y=- 0.0005+0.0004X, R²=0.75 for ring
Y=- 0.0261+0.0196X, R²=0.75 for gap
Y= -0.0017+0.0017X (R²=0.99) for micronucleus in polychromatic erythrocytes
It has been proved that the dose for the radiation exposure less than 1 Gy can be evaluated by combining each dose-effect relationship for the gap chromosomal aberration, the frequency of micronucleus in polychromatic erythrocytes, and apoptotic splenocytes.

3. Optimization of Human Exposure Analysis to Support Radiological Environmental Protection
Based on the Korean technical standards, the integrated accident consequence assessment model (ACCESS) considering the characteristics of both PWR and CANDU has been developed. This code is based on the Visual Basic language and the monitor system is designed by the graphic user interface (GUI). To validate the atmospheric dispersion factor model, the field experiments under unfavorable meteorological conditions in atmospheric dispersion (i.e., low wind speed and stable atmosphere) were conducted by using the tracer SF₆ and the 150 air samplers. They were performed on the flat terrain and the complex terrain, 2 times and 4 times, respectively. The predicted results of the model results were shown to be more conservative than the observation results of the experiments. An user’s guide of the ACCESS code has been made for its effective application. In addition, the further improvement for results analysis has been proposed considering institutional, geological and socio-environmental characteristics of Korea.

4. Techniques for the detection and identification of released radioactive materials for emergency responses
4.1 Development of continuous detection system in high radiation areas
A continuous gamma radiation detection system based on the CZT sensor was designed and developed for the emergency environmental monitoring in high radiation areas from high level radioactive materials as well as nuclear disaster. As assembling and improving the electric circuits for signal processing and system control by domestic technology except for the detector sensor, the energy spectrum with high resolution was achieved. An IRIS collimator was developed to make the quick and efficient measurement in high radiation area, This made the areal classification according to the radiation level practical. It was possible to remotely interface and control the system of hand-held CZT detector with Bluetooth communication. The sustainable measurement in the loss of the outer electric power was enhanced with internal battery of a detector. The dose rate spectroscopy, which is a new concept for the simultaneous measurement of the individual dose rate and radioactivity for detected gamma nuclides, was first developed and experimentally verified. In addition, a program for the calculation of the calibration factor in the in-situ measurement was developed to convert the measured cps from the detector into the radioactivity.

4.2 Development of rapid analysis methods for certain released radionuclides
The sample preparation and radiochemical separation methods on food stuffs was developed for the rapid analysis of certain released radionuclides. In terms of the pre-treatment time, the current developed methods was tremendously improved than the typical dry-digestion (10 days → < 12 hrs). An automated sequential separation system for the radionuclides (radio strontium and plutonium) was designed and fabricated. The developed system can process 8 samples at a time for certain radionuclide. In addition, it can handle four samples at a time for more than two radionuclides. The performance of the system was evaluated using simulated samples. All process is operated in a fully computer controlled mode with a user-friendly software. We developed the rapid determination method of Sr-89/90 activity for mixed samples with Sr-89/90. The determination of Plutonium in environment was developed using LSC (Liquid Scintillation Counter) and alpha spectrometry for emergency responses.

4.3 Identification technique for radioactive materials released into the sea
The sequential separation method of Sr-89/90, Pu, Tc-99 and I-129 in seawater was developed with anion-exchange resin and co-precipitation. We tested the performance of the recommended procedure with 1 ∼ 10 L seawater. The automated pre-treatment system for large volume of samples was developed. It can process 8 samples at a time. It can handle 20 L of seawater for the concentration of Sr-89/90 from sample. All process is operated in computer controlled mode. We developed the determination of I-129 in environmental samples. I-129 is very useful indicator to reconstrut the contamination map caused by I-131 at nuclear accident. We suggested the determination of I-129 using RNAA, ICP-MS and LSC.

5. New technology development for radiation dose measurement and evaluation based on the operational quantity
5.1 Establishment of retrospective dosimetry technology for radiation accident dosimetry
1) Development of retrospective dosimetry technology using personal electronic devices (smart phones)
⋅ Dose dependence of OSL from resistors, inductors, and IC chips
- Linear in the tested dose range (8.7 mGy - 8909 mGy) on the whole
⋅ Fading test of OSL from the samples
- Inducttors: about 50 % in 36 hours after the irradiation
- Resistors: about 80 % in 36 hours after the irradiation
⋅ Estimation of minimum detectable dose for resistors and inductors
- Inducttors: 2.4 mGy; Resistors: 6.6 mGy
⋅ Development of a protocol for retrospective dosimetry using electronic devices
- Procedures from sample preparation and dose estimation
⋅ Field test of the developed procedures for retrospective dosimetry
- It was possible to estimated the irradiated dose respectively in deviation from 5 % to 40 % (depends on dose range)
2) Development of retrospective accident dosimetry technique using building materials
⋅ Selection of building materials for retrospective accident dosimetry
- Fired bricks, tiles, roof tiles, toilet porcelain
⋅ Sample extraction of quartz inclusion and core-disc method for TL/OSL/POSL measurement
⋅ Reliability test of equivalent dose obtained from building materials
- Verification of SAR-OSL method using quartz grains(90-250 ㎛)
- Verification of SAAD-POSL method using core-disc sample(Ø8 mm)
⋅ Development of accidental dosimetry as fast assessment using SAAD-POSL method
- Regeneration test of equivalent dose up to 7 Gy
- Minimum detectable dose of 10 mGy in SAAD-POSL method
- Measurement time as short as about 2 hours in SAAD-POSL method
3) Designing and assembling of OSL spectrum measurement equipment and development of required software for OSL/TL analysis
⋅ Photon measurement devices: CCD and PMT
⋅ Spectrum wavelength band: 325 nm - 800 nm
⋅ Stimulation light sources: Green and Red LED, Green Laser
⋅ Measurement of OSL spectrum from Al₂O₃:C and LiAlO₂ materials
⋅ Development of an algorithm for analysis of OSL, TL based on trap interaction model

5.2 Interpretation of new models for intake of radionuclides and it’s application to internal dosimetry system
1) Analysis and application of HATM
• Analysis of HATM
- Analysis of difference between HATM and GI tract model
- Establishment of database for transfer rates by age group and food type
- Analysis of transfer rate calculation methodology for absorption from alimentary regions other than small intestine
• Application of HATM
- Calculation of intake retention and excretion functions for 100 days after ingestion of ⁹⁰Sr and comparison with existing functions
- Calculation of total disintegration number for 50 years (U₅₀) in each alimentary tract compartment after ingestion of ⁹⁰Sr and confirmation of agreement with the result in ICRP publication 100
- Calculation of intake retention and excretion functions with Cd retention on teeth and Fe retention on small intestine wall
2) Analysis of OIR and establishment of application system
• Analysis of revised HRTM
- Analysis of changed fractional deposition in extrathoracic region (ET)
- Analysis of changed particle transport compartment model and establishment of relevant database
• Development of intake retention functions calculation module using OIR
- Establishment of intake retention and excretion functions calculation methodology
- Development of transfer rate matrix and initial value vector creation module (matrix_creator)
- Development of retention and excretion calculation module (IRF_calculator)
- Application of Intel Kernel Library to improve calculation speed for matrix
- Development of expm module for C++ to calculate matrix exponential
3) Calculation of retention functions for each radionuclide using OIR
• Analysis of radionuclides in OIR part 2 and calculation of intake retention functions
- Analysis of systemic biokinetic model for 14 radionuclides (H, C, P, S, Ca, Fe, Co, Zn, Sr, Y, Zr, Nb, Mo, Tc)
- Establishment of transfer rates for systemic biokinetic model and relevant parameters
- Calculation of intake retention/excretion functions for each radionuclide and intake route
• Analysis of radionuclides in OIR part 3 and calculation of intake retention functions
- Analysis of systemic biokinetic model for 14 radionuclides (Ru, Sb, Te, I, Cs, Ba, Ir, Pb, Bi, Po, Rn, Ra, Th, U)
- Establishment of transfer rates for systemic biokinetic model and relevant parameters
- Calculation of intake retention/excretion functions for each radionuclide and intake route

5.3 Establishment of radiological performance testing scheme of radiation detection systems used for radiation protection, radiological emergency and security
1) Production of the reference X-ray fields (ISO LK qualities) and gamma ray field for test and calibration of personal electronic dosimeters and workplace monitors.
• Production of the reference X-ray fields (ISO LK qualities)
- ISO LK beam series : LK35, LK55, LK70, LK100, LK125, LK170, LK210, LK240.
- Quantification of the produced LK X-ray beams using the measurement method of half value layer (HVL).
- Homogeneity of LK X-ray beams were satisfied within 5% according to ISO4037.
- Air-kerma rates of eight LK X-ray beams were measured using ionization chamber calibrated at KRISS (Korea Research Institute Standard and Science) and JAEA (Japan Atomic Energy Agency).
- Air-kerma rates of KAERI’s LK X-ray (0.1－0.24 mGy.h^-1.mA^-1) were satisfied with ISO criterion (< 0.3 mGy.h^-1.mA^-1)
• Production of the low air-kerma rate gamma ray fields using a radiation source of ²⁴¹Am
- Air-kerma rates of radiation fields of ²⁴¹Am were 0.43 mGy.h^-1 @5 cm, 0.11 mGy.h^-1 @10 cm.
2) Establishment of field performance testing systems used for radiation protection, radiological emergency and security at workplace.
• Evaluation and preparation of performance test manuals and protocols of radiation detection portal monitors and hand-held radionuclide identifiers.
• Establishment of the radionuclide check source set for evaluation and performance test of portal and hand-held monitors.
- Gamma sources: ⁵⁷Co, ¹³³Ba, ¹³⁷Cs, and ⁶⁰Co, and radioactivities are 3.5, 0.7, 0.56, 0.12 MBq, respectively.
- Neutron source: ²⁵²Cf, neutron emission rate is 1.4×10⁴ n/s.
• Production of a check source transfer equipment: moving speed and a height of source position can be adjusted and the function of automatic round-trip was provided.

(출처 : SUMMARY 34p)

목차 Contents

• 표지 ... 1
• 제 출 문 ... 2
• 보고서 요약서 ... 3
• 요 약 문 ... 7
• SUMMARY ... 27
• CONTENTS ... 49
• 목차 ... 50
• 제 1 장 연구개발과제의 개요 ... 51
• 제 2 장 국내외 기술개발 현황 ... 59
• 제 1 절 국외 기술개발 현황 ... 59
• 제 2 절 국내 기술개발 현황 ... 66
• 제 3 장 연구개발 수행 내용 및 결과 ... 71
• 제 4 장 목표달성도 및 관련분야에의 기여도 ... 87
• 제 1 절 연구개발 목표 달성도 ... 87
• 제 2 절 관련 분야 기여도 ... 100
• 제 5 장 연구개발결과의 활용계획 ... 107
• 제 6 장 연구개발과정에서 수집한 해외과학기술정보 ... 115
• 제 7 장 참고문헌 ... 119
• 끝페이지 ... 121