Takatoshi Hattori

Calculation of isotope-specific exemption levels for surface contamination

Isotope-specific exemption levels for surface contamination are calculated for representative radionuclides in general nuclear power plants by developing a deterministic dose assessment model for surface contamination that can be applied to radiation, transport and waste safety, and a practical idea of judging exemption for gross surface contamination by measuring gross gamma-ray emission has been proposed. In the dose assessment model, the objects with surface contamination are classified into three types: manually handled, closely handled and remotely handled objects, and the exemption criteria are chosen to be 0.01mSv/yr in the case of using realistic exposure parameters and 1mSv/yr in the case of using low-probability exposure parameters in accordance with the IAEA Safety Standards Series No. RS-G-1.7. Taking into account the distribution area of surface contamination assumed in the dose assessment model, instead of using the evaluation area of 100cm(2) without variation, the exemption levels for gross surface contamination are found to be higher than those obtained by the conventional method for some radionuclides such as Mn-54, Co-60, Zn-65, Nb-94, Cs-134, Cs-137, Eu-152 and Eu-154.

Verification of screening level for decontamination implemented after Fukushima nuclear accident.

The screening level for decontamination that has been applied for the surface of the human body and contaminated handled objects after the Fukushima nuclear accident was verified by assessing the doses that arise from external irradiation, ingestion, inhalation and skin contamination. The result shows that the annual effective dose that arises from handled objects contaminated with the screening level for decontamination (i.e. 100 000 counts per minute) is <1 mSv y−1, which can be considered as the intervention exemption level in accordance with the International Commission on Radiological Protection recommendations. Furthermore, the screening level is also found to protect the skin from the incidence of a deterministic effect because the absorbed dose of the skin that arises from direct deposition on the surface of the human body is calculated to be lower than the threshold of the deterministic effect assuming a practical exposure duration.

Radiation protection lessons learned from the TEPCO Fukushima No.1 NPS accident

Lessons learned from the TEPCO Fukushima No.1 NPS accident are discussed from the viewpoint of radiation protection in the situation of nuclear emergency. It became clear from the discussion that the protective measures should be practiced by taking into account the time profiles of the radiological disaster after the nuclear accident and that the land and coastal sea areas monitoring had to be practiced immediately after the nuclear accident and the communication methods to tell the public about the radiation information and the meaning of protective measures should be developed for mitigation of the sociological aspects of disaster impacts. And it was pointed out from the view point of practicing countermeasures that application of the reference levels, above which it was judged to be inappropriate to plan to allow exposure to occur, played an important role for practicing protective measures in an optimized way and that the quantities and units used for quantifying radiation exposure of individuals in terms of radiation doses have caused considerable communication problems. Finally, the occupational exposures and the public exposures that have been reported so far are shown, and it is concluded that there is no conclusive evidence on low dose exposures that would justify a modification of the radiation risk recommended by the International Commission on Radiological Protection.

RECONSIDERATION OF THE MINIMUM DOSE CONSTRAINT FOR PUBLIC EXPOSURES IN RADIOLOGICAL PROTECTION

By using a probabilistic approach, the effects of the dose distribution of radiation due to man-made radioactive nuclides when added to those of natural background radiation have been studied. These results show that additional exposure to man-made radiation of up to 0.5 mSv y(-1) (as a dose constraint) would not significantly change the distribution of total public doses. Taking into consideration such probabilistic analysis and rationales of derivations of exemption and clearance levels, it can be concluded that the minimum dose constraint that requires optimisation in radiation protection, should be set to 0.1 mSv y(-1), which is one-order magnitude higher than 0.01 mSv y(-1), the current dose criterion for exemption and clearance.

The Long-Term Performance of Concrete in Nuclear Applications

The Idaho National Laboratory investigated long-term concrete performance for nuclear applications. Scientists searched the open literature for information on the effects of heat and radiation. They also examined the concrete shield of the Pacific Sierra Nuclear Ventilated Storage Cask (VSC-17) for concrete deterioration and loss of shielding. The literature search revealed that low doses, <109 n/cm2 or <1011 Rad gamma, of radiation over periods less than 50 years do not appear to significantly affect concrete. Exposure over 100 years was not studied. The effects of higher doses of radiation are not as clear. Generally, the threshold of degradation is 95°C, and degradation increases with increasing temperature and time. In 15 years of VSC-17 use, there is no apparent effect from environment, radiation, or temperature that has adversely affected the shielding or structural functions, although there has been some minor cracking of the concrete.© 2005 ASME

Development of Waste Monitor of Clearance Level to Ensure Social Reliance on Recycled Metal From Nuclear Facilities

Metal and concrete wastes in the decommissioning of nuclear facilities are classified according to their radioactivity level after decontamination. Radioactive waste below the clearance level (e.g., 0.4Bq.g−1 for Co-60 in Japan) can be disposed of as general industrial waste or recycled. Metal wastes mainly originate from equipment in buildings, except for the metal bars in reinforced concrete. Since contaminated equipment must be decontaminated after dismantling, the main target of measurement would be fragments of equipment, of various shapes, numbers and sizes. In order to transport such metallic fragments out of controlled areas, a surface contamination survey must be performed to confirm that the contamination level is below the legal standard level (e.g., 4Bq.cm−2 for beta or gamma emitters in Japan) in addition to satisfying the clearance level. Taking account of social reliance on recycled metal after inspection of the clearance level and the surface contamination level, it is important to remove the possibility of overlooking contamination above these levels in the recycled metal. The measurement of beta rays is suitable for determining surface contamination on metal because almost none of the beta particles from inside the metal can be detected and the detected radiation can be mostly limited to that from the surface. This is the reason why a survey meter for measuring surface contamination has a detector with a higher sensitivity for beta particles than for gamma rays. Considering the characteristics of the survey meter, it may be difficult to measure the contamination level of the surface of a metal fragment, particularly when the surface is not flat. Moreover, in the case of internal contamination of a small metal pipe, measurement is impossible. The permeability of gamma rays is much greater than that of beta particles. Therefore, gamma rays can be detected even from internal contamination in metal. For gamma ray measurement, accurate and easy calibration of the actual radioactivity level and count rate obtained using a measurement instrument is important. If gamma ray measurement can confirm that the radioactivity level is less than about 400Bq, both the clearance level and the surface contamination level could be inspected simultaneously. In addition, the great amount of labor needed for manual inspection using a survey meter could be saved, and there will be no possibility of missing hot spots of radioactivity due to human error. In this study, a new technique for precise and automatic measurement of gamma emitters in metal waste has been developed using 3D noncontact shape measurement and Monte-Carlo calculation techniques to objectively confirm that the specific radioactivity level of metal waste satisfies the clearance level and furthermore, that the surface contamination level of the metal waste is below the legal standard level. The technique can yield a calibration factor for every measurement target automatically and realizes automatic correction for the reduction of the background count rate in gamma measurements due to the self-shielding effect of the measurement target. A practical monitor (Clearance Automatic Laser Inspection System, CLALIS) has been developed. The accuracy of the automatic calibration and correction of background reduction of the practical monitor has been clarified using mock metal wastes of various shapes, numbers and sizes. It was found that the values measured using the present monitor and the actual radioactivity level agreed within +/−20%, and the corrected and actual background reductions agreed within +/−2%. The detection limit of the present monitor was estimated as being 100Bq for Co-60, taking into consideration the calibration error and correction error of the reduction of the background count rate. The monitor accomplished precise measurements with a 100sec (30sec for gamma ray measurement, 30sec for background measurement) process time per inspection. This indicates that approximately 5 tons of metal waste can be measured per day (1,000 tons per year) in 20kg batches at that process speed.

RADIOLOGICAL PROTECTION FROM RADIOACTIVE WASTE MANAGEMENT IN EXISTING EXPOSURE SITUATIONS RESULTING FROM A NUCLEAR ACCIDENT

In environmental remediation after nuclear accidents, radioactive wastes have to be appropriately managed in existing exposure situations with contamination resulting from the emission of radionuclides by such accidents. In this paper, a framework of radiation protection from radioactive waste management in existing exposure situations for application to the practical and reasonable waste management in contaminated areas, referring to related ICRP recommendations was proposed. In the proposed concept, intermediate reference levels for waste management are adopted gradually according to the progress of the reduction in the existing ambient dose in the environment on the basis of the principles of justification and optimisation by taking into account the practicability of the management of radioactive waste and environmental remediation. It is essential to include the participation of relevant stakeholders living in existing exposure situations in the selection of reference levels for the existing ambient dose and waste management.

Validity of generic scenarios used in derivation of exemption levels for surface contamination considering transport-specific aspects

The exemption levels for surface contamination in units of Bq cm(-2) were derived by developing a new universal dose assessment model that consists of three generic scenarios assessed by considering manually, closely and remotely handled objects. In this paper, as part of the process of verifying the validity of these generic scenarios, annual doses that arise from transport-specific aspects are calculated. The maximum annual doses are found to be lower than 10 µSv, which is the bottom line of the exemption dose criterion. The result verifies the validity of the generic scenarios used in the previous derivation of exemption levels for surface contamination.

Operational level for unconditional release of contaminated property from affected areas around Fukushima Daiichi nuclear power plant.

This paper focuses on the surface contamination control of slightly contaminated property after the Fukushima nuclear accident. The operational level for the unconditional release of contaminated properties is calculated in counts per minute (cpm) to enable the use of a typical Geiger-Muller (GM) survey meter with a 50-mm bore, on the basis of the surficial clearance level of 10 Bq cm−2 for 134Cs and 137Cs derived in the previous studies of the authors. By applying a factor for the conversion of the unit surface contamination to the count rate of a survey meter widely used after the Fukushima accident, the operational level for the unconditional release of contaminated properties was calculated to be 2300 cpm on average and 23 000 cpm at the highest-contamination part. The calculated numerical values of the operational levels are effective as long as the typical GM survey meter is used in the radiation measurement.

Lower Bound of Dose Constraints in Radiological Protection of the Public Taking into Consideration Dose Criteria for Exemption

The International Atomic Energy Agency (IAEA) has published a safety guide, RS-G-1.7, that gives specific values of activity concentration for radionuclides of artificial origin that may be used for bulk amounts of material for the purpose of applying exemption. The primary radiological basis for establishing these values is that the effective doses received by individuals should be of the order of 0.01 mSv or less per year. On the other hand, the International Commission on Radiological Protection (ICRP) has published the new concept of a representative person in Publication 101. This representative person is a hypothetical person exposed to a dose that is representative of the most highly exposed persons in the population. On the basis of the new concept of the ICRP, it is reasonable that, theoretically, the 95th percentile of the distribution of the dose received by the population is lower than the dose constraint, which indicates that the main part of the dose distribution is considerably lower than the dose constraint. Taking into consideration the rationale of IAEA/RS-G-1.7 and the new recommendations for probabilistic dose assessment in ICRP/Pub.101, it is possible to propose that the minimum dose constraint for the optimization of radiological protection of the public should be set to 0.1 mSv/y.