Dose and Risk Calculations for Decontamination of a Hot Cell

Dose and Risk Calculations for Decontamination of a Hot Cell Amr Abdelhady and T. Mongy Reactors dept., Nuclear Research Center, Atomic Energy Authority, Cairo, Egypt Received: 1 /2/ 2014 Accepted: 6 /3/ 2014 ABSTRACT Transporting and processing of radioisotopes and irradiated targets inside hot cells generate a significant contamination. The majority of contamination comes from dispersion of radioactive materials during processing the samples after irradiation. Processing includes opening, extracting the irradiated samples, and preparing the samples in a shield prior to transportation. A model of dispersion of radioactive products inside the cell is postulated. Before decontaminating the cell, the expected dose received by the worker must be evaluated. A RESRAD-BUILD code is used in this study to calculate the dose and the corresponding risk. The calculated dose received during the decontamination process is more than the permissible dose and many proposals are presented in the study to decrease the level of received doses. Key Words: Decontamination / calculation / Hot cell INTRODUCTION A hot cell is a cell that receives the targets and capsules after irradiation in the material testing reactor (MTR) core. The irradiated capsules will be opened and their contents will be fractioned and conveniently processed for internal distributions (1). During opening and processing the irradiated capsules, the internal components of the cell (surfaces and equipment) may be contaminated causing a high radiation level inside the cell and this area is inaccessible. So, cleaning up has to be done in order to decrease the dose level inside the cell to the permissible dose. The hot cell is contaminated with a lot of radioactive sources and activated materials. The total activity of materials abandoned in this cell is not known certainly, but it is expected to be around 6 Ci per operation run. Major radioactive contaminants generated by this activity are Ir-192, Cr-51, and I-125 which considered the most probable isotopes produced in radioisotopes production program in the MTR (1). The goal of this study is calculating the expected dose level and corresponding risk inside the cell during decontamination process. The dose rates and the associated risks related to the clean-up and decontamination are calculated using RESRAD Build 3.5 code. Corresponding author e-mail: amr.abdelhady@gmail.com

HOT CELL DESCRIPTION: As shown in Fig-1, the cell shielding consists of lead walls 100 mm thick in the work area and by reactor block concrete at the rear wall. The ceiling is made out of small carbon steel slabs. Walls are built in bricks with "chevron" type double embeddings. The radiological shielding also continues around the corners. The junctions of lead walls and concrete are made by embedding a portion of the walls into the concrete. This insures that no leaks result from shielding discontinuities. Standard type viewers are provided, built in lead glass with a thickness equivalent to the thickness of the lead walls. The cell work area includes an airtight box to confine aerosols resulting from fractionating of irradiated material or obtained during the development of chemical processes carried out in this cell. Air extracted from the movable container is recycled into the interior of the airtight box.the cell is provided with a double filtering system; a hot absolute filter inside the cell and a second absolute filter outside. The cell filters are connected to the main ventilation system of the reactor. The cell is also provided with telemanuipulators in order to manipulating the irradiated materials inside the cell. INVENTORY DETERMINATION Spreading and dispersion of radioactive material inside the cell depend on its form. Radioactive material in forms of powder, gas and liquid can easily be spread and is called dispersible radioactive material. So, the form of each isotope in this study must be determined. The main radioisotopes produced in MTR are Cr-51 with activity of 500 mci/run, I-125 with activity of 5 Ci/run, and Ir-192 with activity of 280 mci/run. As shown in table 1, Cr-51, I-125, and Ir-192 are irradiated in forms of powder, gas, and solid (wire) respectively. So, Cr-51 and I-125 are considered to be dispersible materials. Dispersion of radioactive Cr-51 (with half life of 27.7 days) in form of powder causes a contamination for the tools, equipment, and internal surfaces of the cell. Dispersion of I-125, which is a radioactive gas with half life of 60.14 days, inside the cell may inhaled by the worker during contamination process (1). Dispersion of Ir-192 in form of solid wire present mainly an external exposure for the worker (1).The total activity inside the cell must be evaluated before decontamination by accumulating all the radioactive materials activities during the age of cell operation. In this case and due to the lack of information, the total activity inside the cell is supposed and considered to be the accumulation of the maximum activity for each isotope for each run. So, the calculated dose by the code will be the maximum expected dose received by the worker during decontamination process. Table (1): inventory of isotopes in hot cell Isotope form activity Half life Cr-51 powder 500 mci/run 27.7 d I-125 gas 5 Ci/run 60.14 d Ir-192 solid wire 280 mci/run 74 d 010

01 Structure 02 Shield thick 100 mm 03 Airtight box 04 Leveling system 05 Auxiliary door 06 Shielding door 07 Passant for service 08 Articulated tong 09 Internal terminal board 10 Shields hoist device 11 Ventilation system Fig (1): diagram of the hot cell

CODE OF SOLUTION The needed code to calculate the received dose for the worker during the decontamination process must be able to compute the dose from external and internal exposure. RESRAD-BUILD is a sophisticated code developed by Argonne National Laboratory. The code is a model for analyzing the radiological doses resulting from the remediation and occupancy of buildings contaminated with radioactive material. The pathways of the code include direct external exposure from surface source, inhalation of airborne radioactive particulates, inadvertent ingestion of source material directly, inadvertent ingestion of deposited materials, exposure to deposited materials, exposure due to air submersion, inhalation of aerosol indoor radon progeny (2,3). The radiation risk can be computed by using the U.S. Environmental Protection Agency (EPA) risk coefficients with the exposure rate (for the external radiation pathways) or the total intake amount (for internal exposure pathways (2). The EPA risk coefficients are estimates of risk per unit of exposure to radiation or intake of radionuclides that use age- and sex-specific coefficients for individual organs, along with organ-specific dose conversion factors (DCF). The EPA risk coefficients are characterized as best-estimate values of the age-averaged lifetime excess cancer incidence risk or cancer fatality risk per unit of intake or exposure for the radionuclide of concern. Detailed information on the derivation of EPA risk coefficients and their application can be found in several EPA documents (4,5). It is assumed, in RESRAD-BUILD code, that the depositing velocity is 0.01 m/s, resuspension rate is 5 10-7 s -1, breathing rate is 18 m 3 /d, ingestion rate is 1 10-4 m 3 /h, and different times of exposure range between 5 minutes and 6 hours (3). The receptor will receive total dose including; external dose, deposition dose, immersion dose, inhalation dose, and ingestion dose. RESULTS Assuming that the cell decontamination is urgent and cleaning cannot be postponed, the received dose certainly increases with increasing the time of exposure and the worker will receive 6.28 msv during one hour of decontamination process inside the hot cell as shown in Fig-2. Fig-2 represents a relation between the worker received dose inside the cell and the time of exposure. The relation is derived from running RESRAD-BUILD code at different time of exposure. From radioprotection point of view, the calculated dose per hour is very important to comparing with the permissible dose limits. The received dose per one hour is 6.28 msv with a corresponding risk of 4. 10-4 (5,6) as shown in table 2. Fig (2): worker received dose and corresponding risk inside the hot cell at various time intervals of decontamination.

Table (2): the received dose inside the cell during decontamination for different exposure times with corresponding risk time of exposure (day) dose (msv) Risk 0.003472 (5 min) 0.524 3.42E-05 0.0052 (7.5 min) 0.784 5.13E-05 0.0104 (15 min) 1.57 1.03E-04 0.02 (0.5 hr) 3.02 1.97E-04 0.0416 (1 hr) 6.28 4.10E-04 0.1 15.1 9.86E-04 0.25 37.7 2.46E-03 0.5 75.2 4.92E-03 1 150 9.81E-03 5 732 4.79E-02 10 1420 9.30E-02 20 2680 1.76E-01 30 3790 2.49E-01 The expected dose rate is very higher than the permissible dose limits for the worker during operation (10 μsv/h) (6,7). The main dose is coming from I-125 with percentage of 77.44 % of the total received dose and the residual contributions are coming from Ir-192 and Cr-51 with percentages of 21.61% and 1.03 % respectively of the total dose as shown on Fig- 3. From the point of pathways, the external exposure pathway is the dominant and contributes with 70.8 % from the total pathways. The residual pathways are ingestion, inhalation, deposition, and immersion which contribute with percentages of 20.04%, 9.08%, 0.12%, and 0.01% respectively. As mentioned above, the received dose is higher than the permissible dose and so if the cell cleanup is urgent, using the manipulator of the cell and remote handling equipment is necessary to decrease the received dose (8). Fig (3): nuclides contributions percentages of the total dose

Another proposal is suggested in order to decrease the received dose by postponing the decontamination process to the future. Fig- (4) shows the received dose for one hour of decontamination at different decay times. If the decontamination is postponed for 1, 2, 3, 6, 12 months the worker will receive 4.37, 3.04, 2, 0.5, 4.5 10-2 msv as shown in table 3. Postponing the decontamination depends on the conditions and the demand of using hot cell. The radioprotection responsible determines the decay time at which the worker can do the task of decontamination depending on the permissible dose limits. Fig (4): worker received dose for one hour of exposure versus decay time in the hot cell. Table (3): dose received during one hour exposure with decay time Received dose for 1 hr exposure (msv) 0.00E+00 6.28 2.74E-03 6.21 5.48E-03 6.13 8.22E-02 4.37 1.64E-01 3.04 1.00E+00 4.47E-02 2.00E+00 1.06E-03 Decay time (yr)

The last proposal for decreasing the received dose by using many workers in executing the task of decontamination and so decreasing the time of exposure for each worker and so decreasing the received dose per worker. CONCLUSION The doses and risks for worker received during decontamination of hot cell after hypothetical dispersion of radioisotopes inside hot cell are calculated using RESRAD-BUILD code. Due to high dose/risks involved, complying with the ALARA principle, some proposals and unconventional solutions are suggested in order to decrease the received dose during decontamination process; 1. If the cell cleanup and decontamination process is urgent, using the manipulator of the cell and remote handling equipment is necessary to decrease the received dose by the worker. 2. Decreasing the time of exposure per worker by using many workers during the decontamination process and so decreasing the personal received dose and corresponding risk. 3. If the decontamination process is not urgent, it can be postponed to the future. REFERENCES (1) Amr Abdelhady, Determination of the maximum individual dose exposure resulting from a hypothetical LEU plate-melt accident, Annals of Nuclear Energy 56, 189 193, (2013). (2) C. Yu, D.J. LePoire, J.J. Cheng, E. Gnanapragasam, S. Kamboj, J. Arnish, B.M. Biwer, A.J.Zielen, W.A. Williams, A. Wallo III, H.T. Peterson, Jr., User s Manual for RESRADBUILD Version 3, Environmental Assessment Division, Argonne National Laboratory (ANL), ANL/EAD/03-1, June 2003. (3) Biwer, B. M., S. Kamboj, J. Arnish, C. Yu, and S. Y. Chen, Technical Basis for Calculating Radiation Doses for the Building Occupancy Scenario Using the Probabilistic RESRAD-BUILD 3.0 Code, NUREG/CR-6755, U.S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Research, Washington, D.C., February. [SRDB Ref ID: 91669], 2002. (4) U.S. Environmental Protection Agency, Exposure Factor Handbook, EPA/600/P95/002Fa, Office of Research and Development, National Center for Environmental Assessment, 1997. (5) K.F. Eckerman, et al., Cancer Risk Coefficients for Environmental Exposure to Radionuclides, EPA 402-R-99-001, Federal Guidance Report No. 13, prepared by Oak Ridge National Laboratory, Oak Ridge, Tenn., for U.S. Environmental Protection Agency, Office of Radiation and Indoor Air, 1999. (6) International Commission on Radiation Protection (ICRP), Recommendations of the International Commission on Radiation Protection ICRP Publication 60, Pergamon Press, Oxford, England (1991). (7) IAEA, FAO, ILO, OECD, WHO, International Basic Standards for Protection against Ionizing Radiation and for the Safety of Radiation Sources, Safety Series no. 115, IAEA, Vienna (1996). (8) M. Dragusin, et al., IFIN-HH VVR-S Research Reactor Decommissioning Plan, Revision 10, Institute of Physics and Nuclear Engineering Horia Hulubei (IFIN-HH), Centre of Decommissioning and Radioactive Waste Management (CDMR), February 2010.