Abstract
The decontamination and facing process at premises (outdoors) of TEPCO's Fukushima Daiichi Nuclear Power Station (referred to as FDNPS) has made it possible to work in standard work clothing without a full face mask in more than 96% of the site, but high levels of contamination from melted nuclear fuel still remain in the reactor buildings. In the FDNPS, since Cs-137 is the dominant radionuclide in most of the workplace, all workers who enter the controlled area are monitored individually for internal exposure by whole-body counter (for screening purposes). From 2024 or later, the gradual removal of fuel debris is planned, which will increase work inside the reactor building. To ensure the radiation safety of workers in such special operations, it is essential to establish an internal dose monitoring system that properly uses a combination of in-vivo and in-vitro bioassay. This article discusses monitoring methods by calculating the radionuclides that affect dose during an internal exposure event based on the FDNPS workplace measurement data. In addition, the approach for quality assurance of individual monitoring in the FDNPS will be presented.
1. PRESENT STATUS OF FDNPS
1.1. Outline of work environment
On March 11, 2011, a magnitude 9.0 earthquake centred on the Pacific side of Japan triggered a massive tsunami of unprecedented scale that struck the Tohoku coast. The tsunami caused the loss of safety-critical functions such as water injection into the reactors and plant monitoring, which led to the evaporation of water from inside the pressure vessels of Units 1, 2, 3, resulting in an accident (severe accident) that damaged the reactor cores. The hydrogen generated by the water–zircaloy reaction between the exposed fuel rod cladding and water vapour accumulated in the reactor building, leading to hydrogen explosions in Units 1 and 3 (IAEA, 2015).
The reactor building (R/B) is contaminated with highly concentrated radionuclides from the explosion, and molten nuclear fuel debris remains in the RPV and PCV. To remove decay heat from this nuclear fuel debris, continuous circulation of water cooling has been conducted. The water used for cooling contains high levels of actinides, Cs, and Sr due to contact with nuclear fuel debris. After the Cs and Sr concentrations are reduced using adsorption equipment, the residual salt is removed by desalination equipment (reverse osmosis membrane filters) (Fig. 1). Some of this treated water is recirculated back into the reactor building for cooling nuclear fuel debris, and some is treated in the multi-nuclide removal facility (ALPS) for further radionuclide removal process. In the ALPS treatment process, actinide separation by iron co-precipitation is followed by Sr separation using carbonate formation, and at the final step, radioactive materials other than tritium are removed using adsorbents. The slurry and spent adsorbent generated during this process are stored in a high-integrity container (HIC). With these processes as a background, the FDNPS premises have a work environment where highly radioactive materials such as concentrated liquid waste generated from desalination equipment and slurry generated from the ALPS treatment process, in addition to actinides in the reactor building.
Treatment process of stagnant water.
1.2. Radiation protection programme for internal exposure
In FDNPS, an average of approximately 3700 people are working on a daily basis in FY2022. Currently, the rubbles from the explosion have been removed, and the work environment has been improved. Approximately 96% of the controlled area has been designated as the G zone, where work can be performed in standard work clothing (Fig. 2). The other areas according to radiation risk are called the Y zone and R zone. Some instances of the work in the Y and R zones include ‘removal of fuel from the spent fuel pools of Units 1 and 2’ and ‘transfer of highly concentrated Sr carbonate slurry’. In these operations, remote control and robots are used to minimise the risk of exposure to workers. However, there are situations where workers themselves enter the highly contaminated areas, and even with full consideration of safety measures, they still have the possibility of internal exposure.
Improvement and change of controlled zone (facings, decontamination, removal of contaminated materials).
When work is planned at FDNPS, a radiation work authorization (RWA) is prepared. To establish RWA, the optimisation of radiation protection is discussed in the following order: engineering measures, administrative measures, and the use of personal protective equipment. Optimisation of protective equipment for radiation dose reduction measures is based on dose equivalent rate (mSv h−1), surface contamination density (Bq cm−2), and airborne radionuclide concentration (Bq cm−3) (Fig. 3). Additionally, the concentration of radionuclides in the air during the work is constantly monitored by continuous dust monitors to ensure that the working environment is safe with the selected respiratory protective equipment. If the value of the continuous dust monitor exceeds the predetermined control standard and the alarm sounds, the work should be suspended, and measures such as changing the zone or protective equipment should be taken.
Radiation protection equipment by category.
The individual monitoring programme for internal exposure at FDNPS consists of routine monitoring and special monitoring. Routine monitoring has focused mainly on Cs-137, and WBC measurements have been conducted for all workers before and after assigning them as radiation workers and every 3 months (every month for women) during their work. Cases requiring special monitoring have been rare, and no significant internal exposure which exceeds recording level has been reported since October 2011.
2. DISCUSSION OF EFFECTIVE DOSE IN THE TREATMENT PROCESS OF STAGNANT WATER
Internal exposure occurs when radioactive materials that enter the body via inhalation, orally, through wounds, or dermally accumulate in the body and affect target organs. The major target of Pu, as represented by actinides abundantly contained in fuel debris, is the liver and skeleton, which have high effective dose coefficients (ICRP, 1997, 2019), resulting in a large committed effective dose (CED) even with small amounts of intake.
Radionuclides taken into the body at the incident can be measured by in-vivo bioassays, such as WBC and lung monitors, and in-vitro bioassays by analysing faeces and urine. Gamma-ray emitters such as Cs-137 are generally measured by WBC. Pu-239 measurements of 17.06 keV Lα x-rays are usually shielded by the chest wall, making direct measurement by lung monitors practically impossible. Therefore, 59.5 keV gamma-rays from coexisting Am-241 are used to measure lungs, but the gamma-ray emission rate is small, and the detection limit becomes high. For alpha emitters such as actinides and beta emitters such as Sr/Y-90, in-vitro bioassays are commonly used.
It is important to remove radionuclides from the body before they accumulate in target organs in medical treatment for internal exposure. Therefore, in-vivo bioassay plays an important role in obtaining rapid dose estimates for medical decisions. Since most of the contaminated areas in the FDNPS workplace contain Cs-137 as a dominant radionuclide, the initial dose assessment is based on estimating the radioactivity level in the body using WBC and by the ratio of Cs-137 and other radionuclides. In the case of a suspected incident exceeding recording level (2 mSv), an additional measurement is performed in order to make medical decisions and finalise the dose assessment (Fig. 4).
Dose assessment for the workers.
Therefore, both the preparation of analytical facilities and the establishment of methods for prompt response are necessary in additional monitoring (special monitoring), and it is important to create a monitoring procedure for each workplace before starting high-risk work. In addition, since diverse radionuclides are present at an unspecified ratio in the workplace, it is essential to identify actinides in advance based on work environment data for prompt response.
Among the work environments with high radioactive contamination, four samples were selected from the ‘Fukushima Daiichi Radwaste Analytical Data Library’ (https://frandli-db.jaea.go.jp/FRAnDLi/index.php?country=e) where workers may enter (Asami et al., 2017). Figure 5 shows the ratios of radioactivity for each nuclide which were calculated in Eq. 1 below. Sample C is from an area that is particularly contaminated with high alpha emitters and is normally restricted from entering. Since there were no measurement results for Y-90, it was assumed to be in radiative equilibrium with Sr-90. The measured values of Pu-241 in samples A, B, and D were predicted based on the ratio of Pu-238 in the inventory calculation results:
Ratio of radioactivity at each sampling point.
Cs-137 was the major radionuclide in the R/B loose contamination, and even in sample C, an area with high contamination by alpha emitters, the ratio of alpha emitters was small. In this kind of environment, the screening method for internal exposure will be Cs-137 measurement using WBC. In the work environment where Sr/Y-90 are major nuclides, such as sample D, screening using WBC is not available. Therefore, in addition to the measurement of airborne radioactive concentrations, screening by nasal smears of workers is conducted.
The ratio of CED in the case of inhalation of the samples in Fig. 5 was then calculated in Eq. 2 below and shown in Fig. 6. For the effective dose coefficients, all unspecified forms were selected for beta and gamma-ray emitters and uranium, since the chemical forms are unknown. For Eu-154, type F was chosen because there was no selection for all unspecified forms. For alpha emitters such as Pu, ‘Pu in mixed oxide’ or type S was selected in consideration of the core meltdown situation at high temperatures (selected from OIR data viewer):
Ratio of committed effective dose at each sampling points.
Due to the high CED of alpha emitters, sample C has a high CED ratio of Pu and other alpha emitters, indicating that using DTPA is necessary for internal decontamination in case of internal exposure. In addition, the target radionuclides for special monitoring are Pu, Am, Cm, Cs, and Sr, and systematic chemical analysis will be necessary.
2.1. Routine and special monitoring
The internal exposure monitoring method for FDNPS with the work environment described above is shown in Table 1. Routine monitoring using WBC as the target radionuclide of Cs-137 is performed every 3 months (every month for women) for all workers who enter the controlled area. If there is any suspicion of internal exposure, additional in-vivo and in-vitro bioassays are performed as special monitoring, depending on the nuclide of interest. For areas where Sr-90 is a major target radionuclide, monitoring is conducted by controlling airborne radiation concentrations, and urine analysis is performed if internal intake is suspected.
Monitoring methods of the worker at FDNPS.
Figure 7 shows a block diagram of a rapid urine analytical method for Sr which is performed in a suspected internal exposure situation and conventional method which is commonly required long time for analysis.
Rapid and conventional bioassay method for Sr.
2.2. Quality assurance and improvement
Quality management system is systematised and constructed based on regulatory requirements. Since there has been no accreditation programme for individual monitoring of internal exposure in Japan, we are going to improve the quality of our operation with participation of international intercomparison programme and the help of experts outside Japan as follows:
PROCORAD (in-vitro bioassay) Calibration and measurement using BOMAB phantoms (in-vivo bioassay): special thanks to Guy Backstrom (DOE RESL, backstlg@id.doe.gov) (Fig. 8) Internal
WBC (plastic scintillator) and BOMAB phantom.
3. CONCLUSIONS
By zoning and monitoring the FDNPS workplace, where various radionuclides are present, appropriate management methods have been established and improved. To strengthen the internal exposure control system for future debris retrieval, the operation of the in-house in-vitro bioassay facility and continuous training of experts will be conducted.