Risk Assessment of Low Level Radioactive Waste
in Near Surface Disposal Facilities
S. SUJITHA a, Sampurna DATTA b and G. L. Sivakumar BABU c
aSenior Research fellow, Department of Civil Engineering, Indian Institute of Science, India bSenior Research fellow, Department of Civil Engineering, Indian Institute of Science,
Bangalore, India
cProfessor, Department of Civil Engineering, Indian Institute of Science,Bangalore, India
Abstract. In this paper, a radiological model for the risk assessment is developed for the near surface disposal facilities of Cesium, Strontium and Cobalt. In addition to the time-dependent annual release rate and dose rate of the radionuclides, the risk to the critical individual is also estimated. The results show that the highest value of annual dose rate is less than the dose limit to the critical individual. The highest value of dose rate appears at about 200years, and at this time, dominant radionuclides are Cs-137 and Sr-90. The risk obtained is much lower than the specified tolerable risk .
Keywords. Risk assessment, near surface disposal, uncertainty analysis.
1. Introduction
In India, the Radioactive Storage and Waste Management site (RSMS) has been mainly receiving the waste in the form of spent resin, sludge and cakes and packaged spent radiation sources from industries and various laboratories. The spent resin and sludge cakes contain roughly 90% 137Cs and 10% 90Sr. For present safety assessment 100% 60Co has been assumed in spent sources (Rakesh et al. 2005). Near surface disposal is an option used by many countries for the disposal of radioactive waste containing mainly short lived radionuclides and low concentrations of long lived radionuclides, IAEA (1994, 1999). Two disposal methods are practiced for Near Surface Disposal Facilities (NSDF), they are single dump mode and multiple dump mode. The single dump mode can represent an unplanned disposal scenario, whose operation period will be over within 10 or 20 years. In countries like India, where the disposal facilities are close to nuclear power plants, the disposal operation of low-level radioactive waste starts with the commissioning of nuclear power plants and continues for a long period ( > 50 to 70 years) until the permanent shutdown of the plants. Such a disposal mode is referred as multiple dump mode. The multiple dump mode
requires evaluation of radionuclide source term during the dumping period and after termination of disposal (post dumping period) Nair et al., (1999). In this study, multiple dump mode is considered.
The disposal site considered in this study is a typical engineered near surface disposal facility for radioactive waste. The engineered barrier system includes top cover, waste form, waste container, backfill, bottom cover and near field geosphere. The waste is immobilised in a solidified matrix. The spent resin is stored in 200 litre steel drum and is not immobilised. The top cover and bottom cover is made of reinforced cement.
The methodology of simplified mathematical modeling for the risk assessment can be an effective tool for the decision making of the radioactive waste repository selection and the management of the repository system, Kim et al (1993). Safety of radioactive waste disposal facilities can be assessed by conventional transport and dispersion models IAEA (1995) in which source term is computed based on diffusion controlled release and dissolution controlled release.
The low and intermediate level waste from wide range of activities, such as the operation and decommissioning of nuclear fuel cycle
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facilities and the use of various sealed and unsealed radiation sources in broad range of medical, industrial, research and other activities has different characteristics, which can influence its acceptability for near surface disposal facilities, IAEA (2003). In the present study, emphasis is placed on the migration of radionuclides from the disposal facility through the geosphere and the resultant dose rate and risk to the critical individual based on the failure rates of barriers.
2. Repository Model
The multi barrier system includes top cover, waste container, waste form, backfill material, bottom cover and the near field geosphere in a sequence. After proper conditioning, the radioactive waste in the solidified form (waste form) is packed in steel drums (waste container) and buried in the facility. The top concrete cover ensures a long term protection from infiltration due to rainfall. The waste containers are disposed over a backfill made of soil mixed with clay and the bottom concrete cover which enhances the isolation capacity. The final release of the radioactive waste to the groundwater is retarded by near field geosphere (unsaturated zone). The schematic diagram of engineered barriers of the repository is shown in Figure 1 (Cadini et al., 2012).
Figure 1. Schematic diagram of engineered barriers of the repository.
Though the barrier system is designed for safe environment, a failure scenario may be encountered due to infiltration. The radionuclide release to the groundwater is estimated by considering the sequential failure of the barrier system by means of infiltration from rainfall waters. The failure of the top cover (Barrier a) begins with infiltration due to rainfall. This would result in contact of the waste container (Barrier b) with water and corroding the mild steel. As the corrosion proceeds, water will interact with the solidified waste (Barrier c) resulting in leaching of radionuclides from the waste form. The leached radioactivity will begin to migrate through the backfill (Barrier d) and after the failure of the bottom cover (Barrier e) reaches the geosphere (Barrier f).
The failure of the ith barrier is modelled as a stochastic event whose time of occurrence is distributed as exponential distribution of constant failure rate λ, Nair et al (1999).
) exp( ) (t t f =λi −λi , t >0, i=a,b,…,f, λi >0 (1) The distribution of the time of release of the radionuclides in to the groundwater can be determined analytically as,
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⎟
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⎞
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⎛
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⎞
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⎛
∑ − ∏ − ∏ = = ≠ = f a i i i j j i f a i i s t e t f ) ( ) ( λ λ λ λ (2) where fs(t) is the exponential failure probability density of the barrier system (y-1),which can be determined analytically using Eq 2.For multiple dump mode, the release rate during dumping period is,
) ( ) ( ) (T S T f T Rd = d s (3) where Sd(T)=(Q/λp)(1−exp(−λpT)) is the inventory (Bq) during dumping period T (50 years) and Q is the annual disposal rate (Bq/y) of the radionuclide. The release rate for post dumping period is evaluated as,
) ( ) ( ) (t S t f t T Rp = p s + (4) where Sp(t)=Sd(T)exp(−λpt) is the inventory (Bq) of the radionuclide after time t in years from closure of the disposal facility i.e., post dumping period. The time dependent concentration of the radionuclide in the groundwater for multiple dump mode during dumping period can be evaluate as convolution integral,
(5)
Similarly the concentration after the post closure period can be evaluated as,
(6) The two dimensional solution of
concentration of radionuclide in groundwater for the instantaneous release of unit activity is considered in the model as Freeze and Cherry (1979),
(7) where Dx is the retarded longitudinal dispersion coefficient (cm2/y), Dy is the retarded lateral dispersion coefficient (cm2/y), Ux is the retarded groundwater velocity (cm/y), A is the cross sectional area of aquifer (cm2), Rg is the retardation factor which is 1+(Kdρb/θg) where Kd is the distribution coefficient (ml/g), ρb is the bulk density (g/cc), Hg is the aquifer thickness (cm) and θg is the effective porosity.
The radiation dose due to the radionuclide through the drinking water pathway is calculated as the product of concentration of radionuclide in the ground water, drinking water intake and the ingestion dose coefficient. The ingestion dose coefficients IAEA (1996) applicable to general population along with the water intake of 2.2 l/day is used in the evaluation. The total risk factor to the public as recommended by ICRP
(1990) is 7.3x10-5 mSv-1. This risk factor includes risk due to fatal cancer, non-fatal cancer and severe hereditary effects. Hence the product of the risk factor and the dose received gives the risk to the critical individual.
Figure 2. Process to evaluate risk to the critical individual
A MATLAB program is generated for the probabilistic safety assessment of NSDF by solving the convolution integral using Simpson’s rule. The output of the program evaluates the radioactive release rate, concentration of radionuclide in the groundwater, radioactive dose to the critical individual through groundwater for the radionuclide migration immediately below the unsaturated zone. The radionuclide dependent parameters used in the model are given in Table 1. The mean time to failure (MTTF) for all the barriers is mentioned in Table 2. The independent parameters of the radionuclides are given in Table 3.
Table 1. Radionuclide dependent parameters, Nair et al., (1999) Nuclide Half life (y) Inventory (Bq) Ingestion dose coefficient (Sv/Bq) Distribution coefficient Kd(ml/g), BARC report Cs-137 30.2 5.46x1013 1.30x10-8 3580 Sr-90 28.8 2.64x1013 2.80x10-8 4255 Co-60 5.20 5.55x1012 3.40x10-9 4421 τ τ τ C x y d t R t y x C g t d gd( , , ) ( ) ( , , ) 0∫ − = τ τ τ τ τ τC xyt d R t C xy d T R t y x C g t p g T d gp(, ,) ( ) (, , ) ( ) (, , ) 0 0
∫
∫
− + + − =Sequential failure of barriers
Radioactive release rate = source term x repository failure
Radionuclide concentration in groundwater
Dose rate= radionuclide concentration in groundwater x drinking water consumption x
ingestion dose
Risk to the critical individual = dose rate x risk factor
Table 2. MTTF of all barriers, BARC report Notation Barrier MTTF (years) A Top cover 236 B Waste container 10 C Waste form 300 D Backfill 400 E Bottom cover 155
F Near field geosphere RdTr
Rd represents the retardation factor and Tr the travel time
(=z/Uz) in years.
Table 3. Radionuclide independent parameters, BARC report
Parameter Unit Value
Bulk density (ρb) g/cc 1.26
Porosity (θg) - 0.3
Longitudinal distance parallel to the flow (x)
cm 0 Groundwater velocity cm/s 1.157x10-4 Dispersivity (α) cm 100 Thickness of unsaturated zone (z) cm 200
Water intake l/day 2.2
Risk factor mSv-1 7.3x10-5
Aquifer thickness (H) cm 600
Aquifer cross sectional area (A) cm2 1.0 x 106 Seepage velocity in unsaturated zone (Uz) cm/s 1.157 x 10-8 Lateral distance perpendicular to the flow (y) cm 0
3. Results and Discussion
The release rate of the radionuclide in to groundwater (Bq/y) computed using Eq 3 and Eq 4 is presented in the Figure 3. The highest release rate is delivered 137Cs because its inventory value is high and retardation factor is low in the unsaturated zone between the facility and water table. The lowest release rate is observed for
60Co, which has low inventory and high
retardation in unsaturated zone.
The concentration of the radionuclides in the groundwater solved using the Eq 5 and Eq 6 is shown in Figure 4. The concentration depends upon the radionuclide inventory, half life, and sorption capacity to reach the unsaturated zone. In this study, the longitudinal distance and lateral distance is taken as zero in order to predict the concentration of the radionuclide transport
Figure 3. Release rate in to groundwater below the unsaturated zone
Figure 4. Concentration plots below the unsaturated zone
immediately below the unsaturated zone. The maximum concentration is delivered by 137Cs (9.82x10-2 Bq/ml), followed by 90Sr (3x10-2 Bq/ml) and 60Co (9.01x10-5 Bq/ml) respectively. The maximum concentrations of these radionuclides occur between 50 and 200years.
The time history of effective dose (mSv/y) through drinking water immediately below the unsaturated zone is shown in Figure 5.
The maximum effective dose (1.03mSv/y) is contributed by 137Cs at 200 years after disposal followed by 90Sr (6.74x10-1mSv/y at 200years) and 60Co (2.46x10-4mSv/y at 50years). The dose delivered by 60Co is low because of low inventory and high Kd value.
The maximum concentration and maximum dose rate is given in the Table 4 for multiple dump mode. 1.0E-01 1.0E+00 1.0E+01 1.0E+02 1.0E+03 1.0E+04
1.0E+00 1.0E+01 1.0E+02 1.0E+03 1.0E+04
Annual Release Rate (Bq/GW e.y ) Time (years) Cs-137 Sr-90 Co-60 1.0E-06 1.0E-04 1.0E-02 1.0E+00 1.0E+02
1.0E+00 1.0E+01 1.0E+02 1.0E+03 1.0E+04
Concentration (Bq/m l) Time (years) Cs-137 Sr-90 Co-60
Figure 5. Dose rate plots through the drinking water pathway
Table 4. Maximum concentration and maximum dose rate Radionuclide Maximum concentration (Bq/ml) Maximum dose rate (mSv/y) 137Cs 9.82x10-2 1.03 90Sr 3x10-2 6.74x10-1 60Co 9.01x10-5 2.46x10-4
The risk from the near surface disposal of low-level radioactive waste is computed for multiple dump mode as the product of dose received (mSv/y) and the risk factor of 7.3x10-5 mSv-1. The maximum risk obtained over time is given in Table 5.
Table 5. Risk assessment.
Radionuclide Maximum dose rate (mSv/y) Obtained risk (y-1) Risk observed, Nair et al(1999) (y-1) 137Cs 1.03 7.48x10-5 1x10-3 – 1x10-4 90Sr 6.74x10-1 4.92x10-5 60Co 2.46x10-4 1.80x10-8
The maximum risk obtained over time is lower than the risk observed from industrial accidents and natural catastrophes. The length of travel of the radionuclides was observed to be 3m below the unsaturated zone on further analysis. Beyond 3m, the radionuclides become insignificant since these are short lived and highly sorbing radionuclides.
4. Conclusions
A risk analysis methodology has been developed for safety assessment of near surface disposal facilities for Cesium, Strontium and Cobalt. The developed model can generate the release rate, concentration of radionuclide in groundwater, the annual effective dose to a critical individual and risk due to waste disposal practice. The risk obtained is less than the risk observed from industrial accidents and natural catastrophes.
Acknowledgements
The work presented in the paper is part of the research work in the project entitled "Development of probabilistic design and analysis procedures in radioactive waste disposal in NSDF and design of NSDF closure" supported by Board of Research in Nuclear Sciences, Department of Atomic Energy, Government of India, Mumbai. Their financial support is gratefully acknowledged.
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