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REFERENCE LINKING PLATFORM OF KOREA S&T JOURNALS
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Journal of Radiation Protection and Research
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Journal DOI :
Korean Association for Radiation Protection
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Volume & Issues
Volume 26, Issue 4 - Dec 2001
Volume 26, Issue 3 - Sep 2001
Volume 26, Issue 2 - Jun 2001
Volume 26, Issue 1 - Mar 2001
Selecting the target year
STUDY ON X-RAYS AND NEUTRONS LEAKED FROM A 45 MeV ELECTRON LINAC FACILITY
Sawamura, Sadashi ; Kitaichi, Masatoshi ; Nojiri, Ichiro ; Yamada, Takuma ; Kaneko, Junichi ; Sawamura, Teruko ;
Journal of Radiation Protection and Research, volume 26, issue 3, 2001, Pages 133~137
Spatial and time distributions of x-rays and neutrons from Hokkaido University 45 MeV electron linac facility were measured and compared with the calculation. In the calculation, x-rays in a Pb-target were evaluated using the EGS-code. The x-rays and the neutrons from the target to the facility building boundary and skyshine process outside the facility building were simulated with the EGS and the MCNP respectively.
STUDIES ON THE BIOLOGICAL HALF-LIVES OF TRITIUM RELEASED AT WOLSONG NUCLEAR POWER PLANTS
Kim, H.G. ; Eum, H.M. ; Cha, S.C. ; Kim, M.C. ;
Journal of Radiation Protection and Research, volume 26, issue 3, 2001, Pages 139~142
The one of important parameter involved in the calculation of internal radiation dose to the human body is the biological half-life of the radionuclide. The biological half-life is population specific and may differ from one population group to another. So the effective half-life of tritium exposure based on urinal bioassay measurement of Wolsong Nuclear Power Plants was investigated and studied.
THERMOLUMINESCENCE DOSIMETRIC PROPERTIES OF Ge- AND Er-DOPED OPTICAL FIBRES AND THEIR APPLICATION IN THE MEASUREMENT OF DEPTH -DOSE IN SOLID WATER PHANTHOM
Amin, Y.M. ; Abdulla, Y.A. ; Khoo, B.H. ;
Journal of Radiation Protection and Research, volume 26, issue 3, 2001, Pages 143~147
The dosimetric properties of Ge- and Er-doped optical fibres are studied. The Ge-doped fibre is found to be more sensitive to radiation and there is little fading of TL signal compared with Er-doped fibre. The Ge- and Er-doped fibres showed a linear response over a range of
to about 120 Gy and
to about 250Gy respectively. The Ge-doped fibre is found to be dose-rate independent both for photons and electron beams of energy ranging from 6 to 10 MeV and 6 to 12 MeV respectively. The fibre is energy independent for energy greater than
for photon or 0.1 MeV for electron beam. From the depth-dose measurement, it was found that the position of maximum dose, dmax, increased with increasing energy ranging from
for 6 MeV and 10 MeV photons respectively. The central axis percentage depth dose at 10 cm depth was found to be in good agreement with the value obtained using ionization chamber.
AN EVALUATION OF RADIATION DOSES RESULTING FROM THE MEDICAL USE OF HIGH-ENERGY BETA-RAY SOURCES
Park, Jae-Woo ; Kim, Hyun-Jo ;
Journal of Radiation Protection and Research, volume 26, issue 3, 2001, Pages 149~154
Calculational models to evaluate radiation doses resulting from the medical use of high energy beta-ray sources are presented. The radioactive sources considered are Sr-90/Y-90 used as ophthalmic applicator, Re-188 used for treating restenosis of coronary artery, and Ho-166 used for treating hepatic tumors. Typical therapeutic situations which might induce relatively high radiation doses the medical person involved were considered to compute by using MCNP-4C Monte Carlo code the radiation doses. The calculation results suggest that for all of the cases considered, the evaluated radiation doses are negligible compared to the dose limits. It is also found that the effect of Bremsstrahlung radiations on the total dose is insignificant, and hence the conventional lead gown is also effective in shielding beta-rays.
A STUDY ON THE SIMULTANEOUS MEASUREMENTS OF BETA EMITTING ISOTOPES
Lee, Goung-Jin ; Kim, Seoung-Pyung ;
Journal of Radiation Protection and Research, volume 26, issue 3, 2001, Pages 155~159
Beta radiation is measured for an environmental monitoring purpose or for an internal radiation exposure monitoring of nuclear power plant's worker. In korea, strontium 89 and strontium 90 is measured for an environmental monitoring purpose. Also tritium and carbon 14 contained in urine is measured for an internal radiation exposure monitoring of nuclear power plant's worker. Because above isotopes emits low energy beta radiations having a wide range of energy, very complicated isotope separation preprocess is needed. In this study, two mixed beta emitting isotopes are measured simultaneously using a liquid scintillation counter(LSC) and analyzed by using a developed statistical method. Banded least square method is used to analyze the mixed spectrum, and the goodness-of-fitness test is proposed. Test results show that the developed procedure can be very useful for analyzing a mixed beta emitting isotopes.
DEVELOPMENT OF POSITION-SENSITIVE PROTON RECOIL TELESCOPE (PSPRT)
Miura, Takako ; Baba, Mamoru ; Kawata, Naoki ; Sanami, Toshiya ;
Journal of Radiation Protection and Research, volume 26, issue 3, 2001, Pages 161~165
We have developed a position-sensitive proton recoil telescope (PSPRT) which employs a position-sensitive photomultiplier (PS-PMT) and a scintillator for both a radiator and a proton-detector. This system is expected to achieve high energy resolution under a large solid angle, because it enables to obtain the information not only on the proton energy but also the recoil angle from the position data for both detectors. The response of the PSPRT for 14.1 MeV mono-energetic neutrons was measured, and the PSPRT proved to be operating as expected.
THIN-FILM-COATED DETECTORS FOR NEUTRON DETECTION
McGregor Douglas S. ; Gersch Holly K. ; Sanders Jeffrey D. ; Klann Raymond T. ; Lindsay John T. ;
Journal of Radiation Protection and Research, volume 26, issue 3, 2001, Pages 167~175
Semiconductor diode detectors coated with neutron reactive material are presently under investigation for various uses, such as remote sensing of thermal neutrons, fast neutron counting, and thermal neutron radiography. Theory indicates that single-coated devices can yield thermal neutron efficiencies from 4% to 11 %, which is supported by experimental evidence. Radiation endurance measurements indicate that the devices function well up to a limiting thermal neutron fluence of
, beyond which noticeable degradation occurs. Thermal neutron contrast images of step wedges and simple phantoms, taken with dual in-line pixel devices, show promise for thermal neutron imaging detectors.
STUDY ON THE ELECTRON GENERATION BY A MICRO-CHANNEL PLATE BASED ON EGS4 CALCULATIONS AND THE UNIVERSAL YIELD CURVE
Moon, B.S. ; Han, S.H. ; Kim, Y.K. ; Chung, C.E. ;
Journal of Radiation Protection and Research, volume 26, issue 3, 2001, Pages 177~181
The conversion efficiency of a cesium iodine coated micro-channel plate is studied. We use the EGS4 code to transport photons and generated electrons until their energies become less than 1keV and 10keV respectively. Among the generated electrons, the emission from the secondary electrons located within the escape depth of 56nm from the photo-converter boundary is estimated by integrating the product of the secondary electrons with a probability depending only on their geometric locations. The secondary electron emission from the generated electrons of energy higher than 100eV is estimated by the 'universal yield curve'. The sum of these provides an estimate for the secondary electron yield and we show that results of applying this algorithm agree with known experimental results. Using this algorithm, we computed secondary electron emissions from a micro-channel plate used in a gas electron multiplier detector that is currently being developed at Korea Atomic Energy Research Institute.
PRIMORDIAL RADIONUCLIDES DISTRIBUTION AND DOSE EVALUATION IN UDAGAMANDALAM REGION OF NILGIRIS IN INDIA
Manikandan, N.Muguntha ; Selvasekarapandian, S. ; Sivakumar, R. ; Meenakshisundaram, V. ; Raghunath, V.M. ;
Journal of Radiation Protection and Research, volume 26, issue 3, 2001, Pages 183~190
The activity concentration of primordial radionuclides i.e.,
, in soil samples collected from Udagamandalam environment, have been measured by employing NaI (Tl) Gamma ray Spectrometer. The absorbed gamma dose rate has also been simultaneously measured by using both Environmental Radiation Dosimeter at each soil sampling location (ambient gamma dose) as well as from the gamma dose derived from the activity concentration of the primordial radionuclides. The results of activity concentration of each radio nuclides in soil, absorbed dose rate in air due to soil activity and possible cosmic radiation at each location along with human effective dose equivalent for Udagamandalam environment are presented and discussed.
A SIMPLE AND QUANTITATIVE DETERMINATION OF PU ISOTOPES IN SOIL SAMPLES
Lee, Myung-Ho ; Choi, Geun-Sik ; Chung, Kun-Ho ; Cho, Young-Hyun ; Lee, Chang-Woo ;
Journal of Radiation Protection and Research, volume 26, issue 3, 2001, Pages 191~195
An accurate and simple analytical technique for low levels of fallout Pu in the environment was developed using an anion exchange resin. To develop the reliable determination of Pu isotopes in soil samples, decomposition of soil matrix, plutonium oxidation state adjustment on the anion exchange column and source preparation of Pu isotopes have been carried out. The optimum method of Pu isotopes with an anion exchange has been validated by application to IAEA-Reference soils.
DEVELOPMENT OF A FRAMEWORK FOR ASSESSING RADIATION SOURCE TERMS IN NUCLEAR POWER PLANTS
Jae, Moo-Sung ; Park, Shane ; Kang, Kyung-Min ; Jeun, Gyoo-Dong ;
Journal of Radiation Protection and Research, volume 26, issue 3, 2001, Pages 197~201
A risk analysis consists of a triplet,
, where Si is the scenario identification; Pi is the probability of each scenario; and Xi is the consequences of each scenario. A new computing framework, OMAM (ORIGEN-MAAP4-MMCS), has been developed and applied for assessing the risk of a reference plant as well as radiation source terms using the concept of risk triplet. The result of this study using the OMAM framework presented in this paper, can contribute to producing domestic nuclear power plant's risk data base as well as to establishing severe accident management plans.
GAMMA-SPECTROMETRY IN ENVIRONMENTAL MONITORING OF NUCLEAR POWER
Cechak, Tomas ; Gerndt, Josef ; Kluson, Jaroslav ; Musilek, Ladislav ; Thinova, Lenka ;
Journal of Radiation Protection and Research, volume 26, issue 3, 2001, Pages 203~206
The mathematical processing (unfolding) of pulse height spectra from a scintillation detector helps to calculate the photon fluence rate energy distribution in a measured photon field. The data processing is based on the knowledge of detection system response function and directional dependence respectively. The experimental results of the photon fields measurements in the vicinity of the spent fuel temporary storage and inside the storage hall are presented. The containers Castor 440 are used for temporary storing of the burnt up fuel assemblies in the Czech nuclear power plant Dukovany. A set of periodical measurements was performed in order to get basic information on the time dependence of the photon fields spatial distributions and spectral characteristics in the temporary storage hall and its vicinity. The photon fields were measured by the scintillation system. The obtained photon fields spatial distributions and spectral characteristics present the information on the radiation hazard in the storage.
RADIATION SAFETY ASSESSMENT FOR KN-12 SPENT NUCLEAR FUEL TRANSPORT CASK USING MONTE CARLO SIMULATION
Kim, J.K. ; Kim, G.H. ; Shin, C.H. ; Choi, H.S. ;
Journal of Radiation Protection and Research, volume 26, issue 3, 2001, Pages 207~214
The KN-12 spent nuclear fuel (SNF) transport cask is designed for transportation of up to 12 assemblies and is in standby status for being licensed in accordance with Korea Atomic Energy Act. To evaluate radiation shielding and criticality safety of the KN-12 cask, each case of study was carried out using MCNP4B Code. MCNP code is verified by performing benchmark calculation for the KSC-4 SNF cask designed in 1989. As a result of radiation safety evaluation for the KN-12 cask, calculated dose rates always satisfied the standards at the cask surface, at 2m from the surface in normal transport condition, and at 1 m from the surface in hypothetical accident condition. Maximum dose rate was always arisen on the side of the cask. For normal transport condition, photons primarily contribute to dose rate between two kinds of released sources, neutrons and photons, from spent nuclear fuel but for hypothetical accident condition, contrary case was resulted. The level of calculated dose rate was 27.8% of the limit at the cask surface, 89.3% at 2 m from the cask surface, and 25.1% at 1 m from the cask surface. For criticality analysis, keff resulting from the criticality analysis considering the condition of optimum partial flooding with fresh water is 0.89708(0.00065. The results confirm the standards recommended by all regulations on radiation safety.
DEVELOPMENT OF POINT KERNEL SHIELDING ANALYSIS COMPUTER PROGRAM IMPLEMENTING RECENT NUCLEAR DATA AND GRAPHIC USER INTERFACES
Kang, Sang-Ho ; Lee, Seung-Gi ; Chung, Chan-Young ; Lee, Choon-Sik ; Lee, Jai-Ki ;
Journal of Radiation Protection and Research, volume 26, issue 3, 2001, Pages 215~224
In order to comply with revised national regulationson radiological protection and to implement recent nuclear data and dose conversion factors, KOPEC developed a new point kernel gamma and beta ray shielding analysis computer program. This new code, named VisualShield, adopted mass attenuation coefficient and buildup factors from recent ANSI/ANS standards and flux-to-dose conversion factors from the International Commission on Radiological Protection (ICRP) Publication 74 for estimation of effective/equivalent dose recommended in ICRP 60. VisualShieid utilizes graphical user interfaces and 3-D visualization of the geometric configuration for preparing input data sets and analyzing results, which leads users to error free processing with visual effects. Code validation and data analysis were performed by comparing the results of various calculations to the data outputs of previous programs such as MCNP 4B, ISOSHLD-II, QAD-CGGP, etc.
SHIELDING DESIGN ANALYSES FOR SMART CORE WITH 49-CEDM
Kim, Kyo-Youn ; Kim, Ha-Yong ; Cho, Byung-Oh ; Zee, Sung-Quun ; Chang, Moon-Hee ;
Journal of Radiation Protection and Research, volume 26, issue 3, 2001, Pages 225~229
In Korea, an advanced reactor system of 330MWt power called SMART (System integrated Modular Advanced ReacTor) is being developed by KAERI to supply energy for seawater desalination as well as electricity generation. A shielding design of the SMART core with 49 CEDM is established by a two-dimensional discrete ordinates radiation transport analyses. The DORT two-dimensional discrete ordinates transport code is used to evaluate the SMART shielding designs. Three axial regions represent the SMART reactor assembly, each of which is modeled in the R-Z geometry. The BUGLE-96 library is used in the analyses, which consists of 47 neutron and 20 gamma energy groups. The results indicate that the maximum neutron fluence at the bottom of reactor vessel is
and that on the radial surface of reactor vessel is
. These results meet the requirement,
, in 10 CFR 50.61 and the integrity of SMART reactor vessel during the lifetime of the reactor is confirmed.
RADIATION SAFETY STUDIES AT TOHOKU UNIVERSITY CYRIC
Yamadera M. Baba A. ; Miura T. ; Aoki T. ; Hagiwara M. ; Kawata N. ;
Journal of Radiation Protection and Research, volume 26, issue 3, 2001, Pages 231~236
A brief introduction is presented on the radiation safety studies at Tohoku University Cyclotron & Radioisotope Center. Studies on two subject are described; (1) measurement of the thick target neutron yield and radioisotope production / activation cross section for ten's of MeV neutrons and ions using K=110 Tohoku University cyclotron to provide basicdata for accelerator shielding, and (2) development of techniques for high sensitive radiation detection and profile measurement using an Imaging Plate which is a high sensitive two-dimensional radiation sensor. Application of the Imaging Plate techniques to localization of very weak radioactivity and to neutron profile measurement is described.
EVALUATION OF THE UNCERTAINTIES IN THE MODELING AND SOURCE DISTRIBUTION FOR PRESSURE VESSEL NEUTRON FLUENCE CALCULATIONS
Kim, Yong-Il ; Hwang, Hae-Ryong ;
Journal of Radiation Protection and Research, volume 26, issue 3, 2001, Pages 237~241
The uncertainties associated with fluence calculation at the pressure vessel have been evaluated for the Korean Next Generation Reactor, APR1400. To obtain uncertainties, sensitivity analyses were performed for each of the parameters important to calculated fast neutron fluence. Among the important parameters to the overall uncertainties, reactor modeling and core neutron source were examined. Mechanical tolerances, composition and density variations in the reactor materials as well as application of
geometry in rectilinear region contribute to uncertainty in the reactor modeling. Depletion and buildup of fissile nuclides, instrument error related to core power level, uncertainty of fuel pin burnup, and variation of long-term axial peaking factors are main contributors to the core neutron source uncertainty. The sensitivity analyses have shown that the uncertainty in the fluence calculation at the reactor pressure vessel is +12%.
CHARACTERISTICS OF THE KAERI NEUTRON REFERENCE FIELDS FOR THE CALIBRATION OF NEUTRON MONITORING INSTRUMENTS
Kim, Bong-Hwan ; Kim, Jang-Lyul ; Chang, Si-Young ; Cho, Gyu-Seong ;
Journal of Radiation Protection and Research, volume 26, issue 3, 2001, Pages 243~248
Neutron reference fields of Korea Atomic Energy Research Institute (KAERI) for calibrating neutron measuring devices to be used in radiation workplace monitoring consist of two kinds of neutron spectra, the direct and the scattered neutron fields, which are produced by using radionuclide neutron sources, 252Cf and 241AmBe sources. Necessary parameters for calibration such as the anisotropy factor of each neutron source and the room-scattered fraction of some neutron surveymeters in the KAERI calibration facility were determined by calculation or measurement. Spectral measurement of scattered neutron fields were performed at each reference calibration point using a Bonner Multi-sphere Spectrometer (BMS) and the dosimetric quantities for calibration also estimated from the neutron energy spectra which were unfolded using the BUNKI code.
APPLICATION OF WHOLE BODY COUNTER TO NEUTRON DOSE ASSESSMENT IN CRITICALITY ACCIDENTS
Kurihara, O. ; Tsujimura, N. ; Takasaki, K. ; Momose, T. ; Maruo, Y. ;
Journal of Radiation Protection and Research, volume 26, issue 3, 2001, Pages 249~253
Neutron dose assessment in criticality accidents using Whole Body Counter (WBC) was proved to be an effective method as rapid neutron dose estimation at the JCO criticality accident in Tokai-mura. The 1.36MeV gamma-ray of
in a body can be detected easily by a germanium detector. The Minimum Detectable Activity (MDA) of
is approximately 50Bq for 10miniute measurement by the germanium-type whole body counter at JNC Tokai Works. Neutron energy spectra at the typical shielding conditions in criticality accidents were calculated and the conversion factor, whole body activity-to-organ mass weighted neutron absorbed dose, corresponding to each condition were determined. The conversion factor for uncollied fission spectrum is 7.7
INFORMATION SYSTEM ON INTEGRATED RADIATION SAFETY (ISIRS) AND ORPHAN SOURCES CONTROL IN KOREA
Lee, Dewhey ;
Journal of Radiation Protection and Research, volume 26, issue 3, 2001, Pages 255~263
An Orphan Sources Control program controlled by a web based information system in Korea has been developed to satisfy the national demand on a total management of and integrated radiation safety. There are, currently, three approaches going on to control and manage the orphan sources in Korea. First, Korean regulatory authority has been conducting scrutinizing investigation on and thoroughly monitoring the possession of unlicensed radioactive sources from the late 1990s. Second, the regulatory authority will fully operate an information system on radiation safety to effectively trace and monitor all radioactive sources in the country by the mid 2001. Finally, the regulatory authority strongly advises steel mill companies to closely scrutinize and inspect the scrap metals when they attempt to reutilize metals to prevent from being contaminated by uncontrolled sources through the scrap monitoring systems.
NEUTRON ELASTIC AND NON-ELASTIC SCATTERING STUDIES IN TENS OF MeV REGION
Baba Mamoru ; Ibaraki Masanobu ; Miura Takako ; Aoki Takao ; Nakashima Hiroshi ; Tanaka Shin-ichiro Meigo Susumu ;
Journal of Radiation Protection and Research, volume 26, issue 3, 2001, Pages 265~270
Experimental data have been obtained on the neutron elastic scattering cross sections for 55, 65 and 75 MeV neutrons, and non-elastic scattering cross sections for 40 to 80 MeV neutrons using the
neutron source at TIARA of Japan Atomic Energy Research Institute and the TOF method. Data were obtained for C, Si, Fe, Zr, and Pb of natural elements. Elastic scattering data were obtained for 25 laboratory angles between 2.6 and 53.0 that clarified the angular distributions and angle integrated values. The data obtained were compared favorably with recent LA150 data library.
MONTE CARLO SIMULATION FOR CORRECTION OF IONIZATION CHAMBER WALL
Kurosawa, Tadahiro ; Takata, Nobuhisa ; Koyama, Yasuji ;
Journal of Radiation Protection and Research, volume 26, issue 3, 2001, Pages 271~273
In precise measurement of air kerma with cavity ionization chambers, the effect of wall attenuation and scatter are corrected by Kwall and that of nonuniformity by Knu. Using the EGS4 code, we calculated these two correction factors. Correction factors calculated for two different-sized cylindrical ionization chamber differ by up to 0.7% from those obtained by measurements.
ENRICHMENT USING THE SEMI-PEAK-RATIO TECHNIQUE WITH CdZnTe GAMMA-RAY DETECTOR
Ha, J.H. ; Ko, W.I. ; Lee, S.Y. ; Song, D.Y. ; Kim, H.D. ; Yang, M.S. ;
Journal of Radiation Protection and Research, volume 26, issue 3, 2001, Pages 275~279
In uranium enrichment plants and nuclear fuel fabrication facilities, exact measurement of fissile isotope enrichment of uranium is required for material accounting in international safeguards inspection as well as process quality control. The purpose of this study was to develop a simple measurement system which can portably be used at nuclear fuel fabrication plants especially dealing with low enriched uranium. For this purpose, a small size CZT (CdZnTe) detector was used, and the detector performance in low uranium gamma/X -rays energy range was investigated by use of various enriched uranium oxide samples. New enrichment measurement technique and analysis method for low enriched uranium oxide, so-called, 'semi-peak ratio technique' was developed. The newly developed method was considered as an alternative technique for the low enrichment and would be useful to account nuclear material in safeguarding activity at nuclear fuel fabrication facility.
REAL-TIME PERSONAL DOSE MEASUREMENT AND MANAGEMENT SYSTEM RESEARCH IN CHINA
Zhang, Z.Y. ; Cheng, C. ; Liu, Z.S. ; Yang, H.T. ; Deng, C.M. ; Zhang, X. ; Guo, Z.J. ;
Journal of Radiation Protection and Research, volume 26, issue 3, 2001, Pages 281~286
The composition and design of a real-time personal dose measurement and management system are described in this paper. Accordingly, some pertinent hardware circuits and software codes including their operation modes have also been presented.
SEPARATION OF GAMMA-RAYS PRODUCTION FROM
REACTIONS USING DOPPLER SHIFT EFFECT
Kim, Y.K. ; Ha, J.H. ; Youn, M. ; Han, S.H. ; Chung, C.E. ; Moon, B.S. ;
Journal of Radiation Protection and Research, volume 26, issue 3, 2001, Pages 287~290
The 9.17MeV gamma-rays from the
reactions were measured. The incident 9.17MeV gamma-ray was produced from the
reaction at Ep=1.75MeV resonance. The 1.75MeV proton beam was accelerated using the 3MV SNU-AMS Tandetron and 1.7MV KIGAM Tandem accelerators. The enriched 13C target was
self-supporting foil, and we used liquid nitrogen as a resonant absorption target. We used a HP-Ge detector with 30% efficiency and less 2keV energy resolution. We developed new method to detect the scattered 9.17MeV gamma-ray from the nitrogen target by using the energy difference between the Doppler shifted gamma-ray from the
reaction and the resonant absorbed and rescattered gamma-ray from the
ESTIMATION OF OFF-SITE DOSE AND RELEASE CONCENTRATION OF RADIOACTIVE LIQUID EFFLUENTS FROM RADWASTE TREATMENT SYSTEM IN KORI 3&4
Kim, H.S. ; Son, J.K. ; Kim, K.D. ; Ha, J.H. ; Song, M.J. ;
Journal of Radiation Protection and Research, volume 26, issue 3, 2001, Pages 291~298
The designed release rate of liquid effluents from radwaste treatment system should be calculated and evaluated during normal operation, including anticipated operational occurrence and be assured that the release concentration and off-site dose at unrestricted area do not exceed the limits of regulation. The expected annual release rate and off-site dose for the currently operating nuclear power plants in Korea had been calculated and evaluated using PWR-GALE and LADTAP-II which was based on USNRC Regulatory Guide 1.109. Recently, the MOST Notice 2001-2 related to release concentration and off-site dose at unrestricted area was revised to reflect the concept of ICRP-60. It is necessary for KORI 3&4 to re-calculate the release concentration and off-site dose and to compare these results with the limits of regulation. As the results of assessment, we confirmed that the release concentrations were less than its limits of MOST Notice 2001-2 and the off-site dose at unrestricted area using K-DOSE60 was 3.61E-03 mSv/yr to the age of five for the effective dose, and 4.10E-2 mSv/yr to thyroid of the age of five for the organ equivalent dose. We also confirmed the off-site dose was within the limits of MOST Notice 2001-2. Therefore, the release concentration and off-site dose re-evaluated at unrestricted area in KORI 3&4 were well below the regulation limits of MOST Notice 2001-2.
CHEST WALL THICKNESS MEASUREMENTS AND THE DOSIMETRIC IMPLICATIONS FOR MALE RADIATION WORKERS AT THE KAERI
Lee, Tae-Young ; Lee, Jong-Il ; Chang, Si-Young ; Kim, Jong-Kyung ;
Journal of Radiation Protection and Research, volume 26, issue 3, 2001, Pages 299~303
Using ultrasound techniques, the Korea Atomic Energy Research Institute has measured chest wall thicknesses of a group of male workers at the Korea Atomic Energy Research Institute. A site-specific biometric equation has been developed for these workers. Chest wall thickness is an important modifier on lung counting efficiency. These data have been put into the perspective of the ICRP recommended dose limits for occupationally exposed workers: 100 mSv in a 5-year period with a maximum of 50 mSv in anyone year. For measured chest wall thicknesses of 1.9 cm to 4.1 cm and a 30 min counting time, the achievable MDAs for natural uranium in the KAERI lung counter vary from 5.75 mg to 11.28 mg. These values are close to, or even exceed, the predicted amounts of natural uranium that will remain in the lung (absorption type M and S) after an intake equal to the Annual Limit on Intake corresponding to a committed dose of 20 mSv. This paper shows that the KAERI lung counter probably cannot detect an intake of Type S natural uranium in a worker with a chest wall thickness equal to the average value (2.7 cm) under routine counting conditions.
SITE-SPECIFIC ATMOSPHERIC DISPERSION CHARACTERISTICS OF KOREAN NUCLEAR POWER PLANT SITES
Han, M.H. ; Kim, E.H. ; Suh, K.S. ; Hwang, W.T. ; Choi, Y.G. ;
Journal of Radiation Protection and Research, volume 26, issue 3, 2001, Pages 305~309
Site-specific atmospheric dispersion characteristics have been analyzed. The northwest and the southwest wind prevail on nuclear sites of Korea. The annual isobaric surface averaged for twenty years around Korean peninsula shows that west wind prevails. The prevailing west wind is profitable in the viewpoint of radiation protection because three of four nuclear sites are located in the east side. Large scale field tracer experiments over nuclear sites have been conducted for the purpose of analyzing the atmospheric dispersion characteristics and validating a real-time atmospheric dispersion and dose assessment system FADAS. To analyze the site-specific atmospheric dispersion characteristics is essential for making effective countermeasures against a nuclear emergency.
COMPARATIVE RADON MEASUREMENTS
Mahat Rosli H. ; Amin Yusoff M. ; Abdullah Aswadi ;
Journal of Radiation Protection and Research, volume 26, issue 3, 2001, Pages 311~313
NEUTRON FIELD OF THE EARTH, ORIGIN AND DYNAMICS
Kuzhevskij, B.M. ; Nechaev, O.Yu. ; Panasyuk, M.I. ; Sigaeva, E.A. ; Volodichev, N.N. ; Zakharov, V.A. ;
Journal of Radiation Protection and Research, volume 26, issue 3, 2001, Pages 315~319
It is shown, that both cosmic radiation (external source) and natural radioactive gases (inner source) are sources of neutrons near the Earth crust. Correlation between the Earth crust dynamics and variations of thermal and slow neutron flux near the Earth surface is studied. It is shown, that variations of neutron flux near the Earth crust can be used for short-term predicting of natural hazards.
X-RAY FLUORESCENCE IN RESEARCH ON THE CULTURAL HERITAGE
Cechak, Tomas ; Kopecka, Ivana ; Musilek, Ladislav ;
Journal of Radiation Protection and Research, volume 26, issue 3, 2001, Pages 321~326
Radionuclide X-ray fluorescence analysis is a method, which has many advantages for analysing various historic artefacts, as it is relatively cheap, sensitive and non-destructive, and it allows measurements in-situ. However, this analysis has also certain limitations especially concerning sensitivity to chemical elements only, irrespective of the compounds or chemical forms in which these elements have been bonded. In addition, light elements emitting very soft X-rays cannot be measured, and in order to detect a wide range of elements, it is necessary to carry out repeated measurements with different radiation sources. Despite these limitations, valuable information can be obtained about the composition of historic materials and data about the origin and age of these artefacts can be derived. Analyses of wall paintings, ancient metal sculptures or other objects of art provide the basis for historic considerations documented in our results for some objects belonging to the Czech cultural heritage. The results are promising. Thus it is expected that our laboratory will expand its work into more fields of the fine and applied arts.
PULSED NEUTRON FACILITY BASED ON AN ELECTRON LINAC
Kim, Guin-Yun ; Son, Dong-Chul ; Lee, Young-Seok ; Ko, In-Soo ; Cho, Moo-Hyun ; Namkung, Won ; Chang, Jong-Hwa ;
Journal of Radiation Protection and Research, volume 26, issue 3, 2001, Pages 327~331
The Pohang Neutron Facility based on an electron linac was constructed in order to construct the infrastructure for nuclear data production in Korea. It consists of a 100-MeV electron linac, a water-cooled Ta target, and an 11-m time-of-flight path. We measured the time-of-flight path length, the neutron energy spectra for different water levels inside the moderator, and the neutron total cross sections of polyethylene and copper by the transmission method.
PHOTO-NEUTRON SOURCE USING 2 GEV ELECTRON LINAC FOR RADIATION SHIELDING RESEARCH
Lee, Hee-Seock ; Bak, Joo-Shik ; Chung, Chin-Wha ; Ban, Syuichi ; Shin, Kazuo ; Sato, Tatsuhiko ;
Journal of Radiation Protection and Research, volume 26, issue 3, 2001, Pages 333~335
The 2 GeV electron linac, the injector of the Pohang Light Source, was used as a photo-neutron source for radiation shielding research. The operational beam parameters are the nominal electron intensity of
, the repetition rate of 10 Hz, and the beam pulse length of 1.0 nsec. One electron beam line was modified in order to install the target systems for producing pulsed photo-neutrons. The neutron spectrum and intensity were investigated by the time-of-flight technique. The reliable maximum energy of the measured neutrons was about 500 MeV. The number of neutrons above 20 MeV produced by one 1 GeV electron in a thick Pb target was about
at 90 degrees to the beam axis. The status of the photo-neutron source and the application research are presented.
A STUDY ON ICRP RECOMMENDATIONS
Wang, Hengde ;
Journal of Radiation Protection and Research, volume 26, issue 3, 2001, Pages 337~340
This paper reviews briefly the ICRP recommendations published before 1977, in 1977 (ICRP 26), in 1990 (ICRP 60) and in the near future (around 2005) mainly in the philosophy and principles. The great progress is appreciated. A discussion is presented at the end.
CURRENT STATUS OF ACCELERATOR RADIATION SAFETY IN JAPAN
Nakamura, Takashi ;
Journal of Radiation Protection and Research, volume 26, issue 3, 2001, Pages 341~345