Taro Shimada

Decommissioning of Nuclear Facilities

Decommissioning is a series of measures taken after the main activities associated with a licensed activity or reactor have been terminated and before the regulations set forth in the Act on the Regulation of Nuclear Source Material, Nuclear Fuel Material and Reactors [1] (hereinafter referred to as “Reactor Regulation Act”) are fulfilled, including the transfer of nuclear fuel material, elimination of contamination caused by nuclear fuel material, and disposal of nuclear fuel material or other materials contaminated with nuclear fuel material. Therefore, the dismantling of facilities, which is undertaken after the main activities associated with a licensed activity or reactor have been terminated, is also included in decommissioning. Decommissioning is thus a process to reduce the residual radioactivity of such facilities to the levels necessary for fulfilling the regulations set forth in the Reactor Regulation Act. Because these measures produce various types of radioactive wastes in large amounts in a short period of time, the concept of radioactive waste management needs to be actively incorporated into the planning and implementation of decommissioning. If a decommissioning plan is not adequately formulated, there is a possibility that material that does not need to be handled as radioactive wastes may be improperly classified and handled as such. Furthermore, depending on the dismantling method selected, the amount of secondary wastes generated may increase or decrease and the disposal method for the wastes may also vary. It is therefore important to develop a decommissioning plan based on analytical evaluation, operating history surveys, measurement evaluation and other advance surveys as well as the latest dismantling technology studies. As explained above, there is a close relationship between decommissioning and radioactive waste management.

Improvement and testing of radiation source models in DecDose for public dose assessments during decommissioning of nuclear facilities

Radiation source models in a code called DecDose were improved in this study. DecDose had been developed for assessing public and worker exposure doses during the decommissioning of nuclear facilities. A segmentation model evaluating the length, volume, and surface area of kerfs in the object to be dismantled was improved to deal with seven shapes of objects simulating most of the components and the structures in nuclear facilities, while the previous model could treat only two of them. Models for the evaluation of the external dose by direct and skyshine radiation were also improved to deal with the distribution of waste containers temporarily placed in the building and the quantity of radionuclides stored in the individual container. Good agreement was observed between actual and calculated kerf volumes in cutting the reactor pressure vessel, the waste collector tank, and the channel box of the Japan Power Demonstration Reactor (JPDR). It is an indication of the validity of the model improved in this study. On the other hand, some discrepancies were observed between actual and calculated quantities of radionuclides discharged into the ocean during the JPDR dismantling project, indicating the necessity of further validation of the model.

Study on application of kriging to evaluation of radioactivity concentration for ensuring compliance with the criterion of site release

In the field of safety regulation systems for the decommissioning of nuclear facilities, a reliable method for ensuring compliance with the criterion of site release is an important technical issue to be resolved in Japan. Considering that kriging can consider the spatial correlation of radioactivity concentrations between different points, we propose a method of applying kriging to ensure compliance with the site release criterion. Estimated radioactivity concentrations exhibit uncertainty, which results in a certain probability of the occurrence of decision errors regarding site release. We describe a method for calculating the uncertainty as a function of the number of measurement points and establish a minimum number of measurement points required for a given error probability. We applied the proposed method and a conventional statistical method to two sample cases. It was observed that the proposed method appropriately estimated the mean radioactivity concentration even if the measurement points were distributed inhomogeneously. This method was observed to lead to an efficient measurement requiring fewer measurement points relative to the conventional statistical method when spatial correlation existed.

Plasma Arc Cutting Experiments Using Radioactive Materials for Evaluation of Airborne Dispersion Ratio

Experiments for airborne dispersion ratio of radionuclides during plasma arc cutting were carried out in a contamination control enclosure, using stored radioactive metal wastes arising from the decommissioning activities of Japan Power Demonstration Reactor, which was a boiling water type reactor. Neutron induced-activated piping and surface contaminated piping were segmented into pieces using air plasma arc cutting, using a current power was 100A. In addition, similar experiments for contaminated piping of the Advanced Thermal Reactor, Fugen were carried out. As a result, dispersion ratios for activated piping were 0.2 to 0.7% of Co-60 and 0.4% of Ni-63 under the condition with a covered cap on the head. And those for surface contaminated piping were from 18 to 23%. In addition, those for vertically segmented piping which simulated flat plate were from 34 to 43%. There was no difference of dispersion ratios between stainless steel and carbon steel base materials. All values obtained were smaller than the Handbook recommended value of 70% for contaminated materials. Filtering collection efficiencies of the coarse dust filter were approximately 40% for activated piping and approximately 55 to 80% for surface contaminated piping. However there was no effect for collection of aerosols smaller than 1 μm. Size distribution analysis indicated a greater concentration of radionuclides in particles smaller than 0.1μm when compared with larger particles. In addition, there was a tendency that the Ni-63 was concentrated to the particles smaller than 0.3 μm compared with the Co-60. The results support data obtained in the previous studies using non-radioactive materials.© 2009 ASME

Determination of Environmental Uranium Concentration by Utilizing Gamma-Ray Emission From the Progeny Radionuclides

Nuclear facility sites such as an enrichment plant and fabrication plant are allowed to be released from the regulatory control of nuclear safety after the plants are decommissioned. It is necessary to confirm that the site has been decontaminated successfully, prior to be released. A determination method for U-238 concentration of background level in environment and for probate of vast site areas was proposed, in which the gamma-ray emission from the progeny radionuclides of U-238, such as Th-234, Pa-234m, Ra-226, are measured by in-situ gamma-ray spectrometry with a portable germanium semiconductor detector (portable Ge detector). Validity of the determination method of U-238 concentration from the progeny radionuclides was examined by the comparison between the U-238 concentration determined by the in-situ measurement with the portable Ge detector and that directly measured by ICP-MS. The U-238 concentration by the in-situ measurement was determined from peak counting rate at 186 keV of the gamma-ray emission corresponding to Ra-226. The determined U-238 concentration was in the order of 0.01 Bq/g in radioactive concentration, and was in comparable level with the concentrations decided by the ICP-MS. The proposed method utilizing gamma-ray estimation from the progeny radionuclides may be available for the U-238 concentration determination in vast land areas.Copyright

Development of Safety Assessment Code for Decommissioning of Nuclear Facilities (DecDose)

A safety assessment code, DecDose, for decommissioning of nuclear facilities has been developed, based on the experiences of the decommissioning project of Japan Power Demonstration Reactor (JPDR) at Japan Atomic Energy Research Institute (now Japan Atomic Energy Agency). DecDose evaluates the annual exposure dose of the public and workers according to the progress of decommissioning of the plant, and also evaluates the public dose at accidental situations including fire and explosion. The public dose at normal situations during decommissioning is evaluated from the amount of radionuclides discharged from the plant to the atmosphere and the ocean. The amounts of radionuclides discharged depend on which and how activated and/or contaminated components and structures are dismantled. The amount is predicted by using the radioactive inventory given by the plant. The filtration efficiency of the ventilation system and decontamination factors of the liquid waste treatment system of the plant are also considered. Both of the internal dose caused by inhalation and ingestion of agricultural crops and seafood, and the external dose by radioactive aerosols airborne and radioactive deposition at soil surfaces are calculated for all of possible pathways. Also included in the external dose are direct radiation and skyshine radiation from waste containers which are packed and temporarily stored in the in-site building. For external dose of workers, the radiation dose rate from dismantling contaminated components and structures is calculated using the dose rate library which was previously evaluated by a point kernel shielding code. In this condition, radiation sources are regarded to be consisted of two parts; one is a dismantling object of interest, and the other is the sum of surrounding objects. Difference in job type or position is taken into account; workers for cutting are situated closer to a dismantling component, other workers help them at some distance, and the supervisor watches their activities from away. For worker’s internal dose, the radionuclide concentrations in air for individual radionuclides are calculated from a dismantling condition, e.g. cutting speed, cutting length of the dismantling component and exhaust velocity. A calculation model for working time on dismantling was developed using more segmented WBS (work breakdown structure). DecDose was partially verified by comparison with measured the external dose of workers during JPDR Decommissioning Project. The DecDose is expected to contribute to utilities in formulating rational dismantling plans and to the safety authority in estimating conservativeness in safety assessment of licensing application or risk-based regulatory criteria.Copyright