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Sanada, Yukihisa; Nemoto, Makoto*; Suzuki, Kei*; Kawai, Keiichi*; Momose, Takumaro
Proceedings of 12th International Congress of the International Radiation Protection Association (IRPA-12) (CD-ROM), 7 Pages, 2008/10
The purpose of this study is to show the design principle and the determination procedures for the adequate relocation of the CADs in TRP. When the alarm trip point was 2.0 mGy/h, it is possible to detect a criticality accident even under conservative conditions (i.e. 
: 10 m distances from the equipment to the CAD and 140 cm concrete shield between them; neutron: 15 m distances and both 60 cm concrete shield and 60 cm steel shield). Consequently, it was found the seven CADs at these locations could monitor PCDF adequately. This study was reflected in the determination of the new CAD location.
Tsujimura, Norio; Yoshida, Tadayoshi; Nunomiya, Tomoya*; Aoyama, Kei*
Proceedings of 12th International Congress of the International Radiation Protection Association (IRPA-12) (CD-ROM), 9 Pages, 2008/10
The authors developed a new neutron-measuring instrument that was designed to measure a neutron personal dose equivalent, H
(10). This instrument is composed of a conventional moderator-based neutron dose equivalent meter and a neutron shield made of borated polyethylene which covers a backward hemisphere to adjust the angular dependence. The whole design was determined on the basis of MCNP calculations so as to have response characteristics which would generally match both energy and angular dependencies of H(10). The verification experiments for energy and angular responses were done using monoenergetic neutrons produced by the accelerator and polyenergetic neutrons from the moderated 
Cf source. As results of experiments, it is found that this new instrument has reasonably acceptable responses as a reference neutron dosemeter.
Tsujimura, Norio; Yoshida, Tadayoshi; Takada, Chie; Nunomiya, Tomoya*; Aoyama, Kei*
Proceedings of 12th International Congress of the International Radiation Protection Association (IRPA-12) (CD-ROM), 8 Pages, 2008/10
The authors assessed performance of electronic personal neutron dosemeters (EPD). The EPD selected in this study is NRY21, manufactured by Fuji Electric Systems Co., Ltd. This dosemeter uses two silicon semiconductor detectors to provide wide measurable energy ranges from thermal to several tens of MeV. In the test, we focused on the energy and angular dependences of dose equivalent responses because it was of great importance in neutron dosimetry in work environments. Measurements of dose equivalent responses were performed in the moderated neutron field reproducing neutron spectra likely to be encountered in workplaces of MOX fuel facilities. Neutron dose equivalents given by EPD were compared with reference dose values. As a result, dose equivalent responses of this EPD were evaluated to be 0.7 to 1.1 for normal incidence of neutron irradiation. This is satisfactory performance required for personal neutron monitoring.
Takahara, Shogo; Homma, Toshimitsu
Proceedings of 12th International Congress of the International Radiation Protection Association (IRPA-12) (CD-ROM), 8 Pages, 2008/10
Relocation is one of the long-term protective actions in case of nuclear emergency to mitigate the consequences of an accidental release of radionuclides. The strategy of relocation is characterized by its protective benefit, cost and the corresponding residual dose in planning. This paper describes the application of a probabilistic accident consequence assessment model to the calculation of these quantities and the planning of relocation. Calculations of the consequence have been made of a postulated accident with source terms derived from a generic level 2 PSA. The results provided the insights for the development optimum dose criteria for introducing and terminating relocation.
Yasuda, Hiroshi*; Sato, Tatsuhiko; Terakado, Masato*
Proceedings of 12th International Congress of the International Radiation Protection Association (IRPA-12) (CD-ROM), 6 Pages, 2008/10
An Excel VBA program to calculate aviation route doses is developed for personal use by general public. This program named as "JISCARD EX" runs on a personal computer installed with the Microsoft Excel 2003/2007. Cosmic radiation dose rates at aviation altitude are calculated using the PARMA code developed in Japan. Calculated results are automatically summarized in another Excel worksheet as a report containing airport names, flying time, route dose and time plots of altitude, Rc and effective dose rate. This program will be open to the public from a web site of National Institute of Radiological Sciences.
Ishimori, Yuu
Proceedings of 12th International Congress of the International Radiation Protection Association (IRPA-12) (CD-ROM), 10 Pages, 2008/10
The purpose of this paper is to review the current status of the techniques which are used at the Japan Atomic Energy Agency (JAEA) to evaluate the environmental impacts of radon released from closed mine sites. The impacts of radon from the sites are evaluated at the JAEA by two way approaches: environmental monitoring and dispersion model use. Environmental monitoring program of the JAEA basically consists of the passive monitoring of radon and the active monitoring of F value. From a viewpoint of conservative evaluation, the plume model assuming Gaussian distribution was utilized to calculate the impact of radon from the sites. Our assessments, both environmental monitoring and dispersion model use, showed that the effective doses caused by closed uranium mine sites are less than 1 mSv/y, the effective dose limit for the public, outside of the sites. These results were supported by the supplemental investigations.
Tsuda, Keisuke; Fukushi, Masahiro*; Myojoyama, Atsushi*; Kitamura, Hideaki*; Inoue, Kazumasa*; Nakaya, Giichiro*; Kimura, Junichi*; Sawaguchi, Masato*; Kinase, Sakae; Saito, Kimiaki
no journal, ,
Positron emission tomography (PET) is very effective in the diagnosis and management of patients with various types of cancers. PET scanning with the tracer FDG is widely used in clinical PET. However, the effective dose constant of the positron emitter is about eight times of nuclide 
Tc used in a nuclear medicine diagnosis widely. Severe radiation protection is necessary for development of the examination with a positron emitter. Radiation protection in the PET institution therefore and safe security are problems. Hence, lead glass has attracted considerable attention as the radiation shielding materials of the PET institution. In the present study, we received a request of the radiation shielding ability evaluation of two kinds of lead glass made in Pilkington plc. The aim of the present work is the radiation shielding ability evaluation for positron emitter such as 
F(511 keV) of the lead glass. The shielding ability evaluation has been studied in the experimental and the Monte Carlo simulation evaluation. Consequently, effective dose transmission factor with experimental and the Monte Carlo simulation value was calculated. There were enough protection effects to evaluated lead glass. Furthermore, radiation shielding ability evaluation for 
Cs(662 keV), 
Co(1.17, 1.33 MeV) was performed, and it was shown that there was a similar protection effect.
Petoussi-Henss, N.*; Schlattl, H.*; Zankl, M.*; Becker, J.*; Saito, Kimiaki; Endo, Akira
no journal, ,
no abstracts in English
