(9C/(9t)t = a(x,t) (92C/3x2)t + b(x,t) (9C/9 x)t + c(x,t)ct + d(x,t)
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1 ABSTRACT Safe management Including disposal of radioactive wastes from the various parts of the nuclear fuel cycle is an important aspect of nuclear technology development. The problem of managing radioactive wastes, having various chemical and physical forms, is expected to grow with the development of nuclear technology. Underground disposal of low-level radioactive wastes, appropriately immobilized and/or packaged, is generally agreed to be an adequate way of providing the necessary protection for humans and the environment. One of the most important problems associated with this method of waste management has been the possibility of radioactive material becoming mobile. In order to ensure that radioactive waste management facilities are located and designed in such a way as to minimize any possible hazard,. it is necessaiy to understand the mechanism of contaminant transport and the influence of site characteristics on the transport mechanism. With growing public concern and awareness, safety assessment of disposal facilities would assume increasing importance and significance in the years to come. Safety assessment is a multi-disciplinary effort which, in the context of radioactive waste disposal, is a synthesis of expertise from various branches of science viz., radiochemistry, geology, hydrology, geochemistiy, geophysics, modeling and health-physics. The first and foremost task is centered around definition of the initial state of the system which includes detailed site characterisation. Site characterisation is needed for developing an appropriate conceptual model and also for providing necessary input parameters for sensitivity analysis and numerical code calibration. The main purpose of safety assessment is to check the position as regards meeting the regulatory requirements by the disposal operations at a waste disposal facility. In case the projected impacts or results are not meeting the regulatory requirement, the assessment is repeated using parameters reflecting improved physical characteristics of the site such as those obtained by soil conditioning
2 methods or by improving the waste solidification methods etc., as the situation demands. This warrants detailed site characterisation to obtain a reliable safety assessment. A few other soil parameters other than those required in safety assessment have also to be estimated to take effective measures to improve the overall integrity of the site, and eventually help meet the regulatory requirement. The processes and factors that govern the movement of radionuclides in a hydrogeological environment need to be identified and evaluated in order to (a) estimate the length of time for the contaminant to reach the accessible environment and (b) to forecast the concentration levels resulting from radionuclide releases from a nuclear facility into a water-bearing formation. The estimate/forecast is usually carried out by modelling the radionuclide transport from the site specific release point to potential outlets to the human environment. Among the many natural processes that influence the movement of radionuclides and their concentration distribution in the subsurface, certain physical and chemical processes play a major role. A reasonably accurate conceptualization of such physical processes of pollutant migration can be made by considering two basic mechanisms, advection and dispersion. Generally, the most important chemical and nuclear processes to be considered in radioactive-contaminant migration are the influence of ion exchange and radioactive decay. Both of these processes are natural attenuation mechanisms. The principal transport routes to be considered are: (a) the movement of radionuclides through the unsaturated zone and their interaction with the soil or rock. (b) the transport, dilution, dispersion, sorption and decay of radionuclides in the saturated zone. To predict the rates of transport and of intake of radionuclides by humans, mathematical models are necessary. These phenomena can
3 vi be described by a set of differential equations which can readily be derived and which are amenable for solution using numerical techniques. The results obtained depend largely on the quality of input data. The present study has the objective of quantitative forecasting of the performance of the ground disposal system for nuclear wastes (i.e. the evaluation of the radiological impact of the disposal practice at the disposal facility) located at Kalpakkam, Chengai-MGR District, Tamil Nadu, India. The study attempts to compare the forecast results with regulatory limits stipulated by the Code of Federal Regulations of USA 10 CFR 61 and national regulatory requirements stipulated by the Atomic Energy Regulatory Board (AERB) in India. The facility at CWMF is designed to segregate, collect, treat and dispose of low and intermediate level radioactive waste arising from the operation of 2x230 MW(e) Pressurised Heavy Water Reactors (PHWRs) and other installations. Over 95% of the waste generated are stored in a retrievable manner in engineered containment structures. The major release scenario considered at the site includes failure of engineered barriers and subsequent leaching of radionuclides present in the waste by the waters infiltrating into the disposal facility due to precipitation and the leachate reaching the groundwater after passage through the unsaturated zone. A one-dimensional (1-D) numerical code of radionuclide migration using finite-difference method (FDM) has been developed as a part of this study to understand the transport and migration of radionuclide in the saturated region, based on the governing advection-dispersion equation : (9C/(9t)t = a(x,t) (92C/3x2)t + b(x,t) (9C/9 x)t + c(x,t)ct + d(x,t) C represents the concentration of radionuclide or pollutant, a.b.c and d are the coefficients represented by a = Dc/Rr, b = -GWV/Rf, c=-rdc; d = 0 (or source term as the case may be)
4 vu where Dc Rf Kd = TD GWV = RDC = Dispersion coefficient (Li^Tf1) Retardation factor (Dimensionless) 1 + (1-porosity)* Kd * TD/(porosity) Distribution coefficient (L 1) True density (ML'3) Groundwater velocity (LTl ) and Radioactive decay constant ft1). The safety assessment studies for the disposal facility at CWMF were carried out in terms of five isotopes of concern viz., 3H, 14C, 90Sr, 137Cs and 129I. These isotopes were selected based on quantity present in wastes disposed, radiotoxicity and half-lives. The conceptual model broadly envisaged transport modeling in both the unsaturated and saturated zones: however, a detailed modeling of transport in the unsaturated region was not attempted. Appropriate source-term and arrival times of radionuclides in the saturated region of aquifer were computed based on groundwater velocity in the unsaturated region. The 1-D Numerical Code has been employed to forecast the transport of the five radionuclides viz., 3H, 14C, 90Sr, 137Cs and 129I relevant in the context of waste disposal at Kalpakkam towards a hypothetical well 500 m from the disposal facility in the direction of the flow field. The dose received by an individual by consumption of 2 L/d of this well water containing the predicted concentrations of radionuclides at the well has been obtained by using the dose conversion factors (Sv/Bq) of each radionuclide provided by the internationally accepted recommendations of the International Communication on Radiation Protection (ICRP) which has also been incorporated in the code. With this methodology, the performance of the disposal facility has been evaluated by comparing the results obtained from using the above code with the Code of Federal Regulations of USA (10 CFR 61) and national regulations stipulated by the AERB for acceptance criteria. To determine whether the level of uncertainty in the results obtained would be acceptable, sensitivity/uncertainty analysis has also been attempted.
5 viii This study has provided a methodology for safety assessment which has been carried out using simple mathematical models. The methodology has been applied to the safety assessment of the radioactive waste disposal operations at CWMF, Kalpakkam. The significant results obtained are recounted in the following : (1) The time of experiencing maximum dose by ingestion of water from a well 500 m downstream from the disposal site, in respect of the nuclides 3H, 14C, 90Sr, 137Cs and 129I, is 29 years, 35 years, 164 years, 981 years and 35 years respectively. The time of maximum dose follows the order : 137Cs ^"So129! >14C>3H. (2) The maximum dose received itself follows the order 14C > 129I >90Sr >3H >137Cs. (3) Under the failure scenario with a very low probability, the maximum dose received from exposure to 14C appears to be high and the time for maximum dose is also of the order of only a few tens of years; hence this isotope requires special attention in the disposal operations from long-term hazard considerations. (4) The maximum doses that might be received due to the other four isotopes meet the regulatory requirement as specified in 10 CFR 61 (5) The site recipient capacity or site burden estimates for disposal at Kalpakkam were 4.9 TBq, 2.42 x 10'2TBq, 8.84 x 10+13TBq, 7.71 x 10"1 TBq and 5.59 x 10'5 TBq per year respectively for 3H, 14C, 137Cs, 90Sr and 129I. All the estimated quantities except that of 137Cs show that H, C, Sr and I disposal needs continuous and careful consideration because AERB stipulates an annual limit of 10 TBq for waste disposal from all sources. The very large permissible limit determined for Cesium only indicates that the attenuation provided by the soil environment for Cesium is quite significant even at highly conservative retardation factors. The site recipient capacities computed based on AERB limits become five times more stringent and the site recipient
6 ix estimates are: 9.8 x 10'1 TBq, 4.84 xlo'3 TBq, 1.76xlO+13TBq, 1.54 x 101 TBq and 1.11 xlo"5 TBq per year respectively for 3H, 14C, 137Cs, 90Sr and 129I. An overall conclusion of the study is that the Kalpakkam site has favourable characteristics to attenuate the radionuclide migration and also provide assurance that the disposal scenario would meet the internationally accepted Code of Federal Regulations of USA (10 CFR Part 61) and national regulations of AERB provided the inventories of 14C and 129I are carefully monitored and controlled. The three other O 1 0*7 Q/\ isotopes considered, viz., H, Cs and Sr would meet the regulatory requirement quite well even under worst case scenario. This study has shown that stringent surveillance has to be carried out for 14C and 129I at the Kalpakkam site and perhaps advanced immobilisation methods have to be adopted for better containment. The present study has contributed to the evaluation of the overall performance of the radioactive waste disposal facility at Kalpakkam. Use of the 1-D code developed in the study has also helped in identifying the long-term impact of 14C and 129I disposal. Site recipient capacity or site burden for the five nuclides viz., 3H, 14C, 137Cs, 90Sr and 129I could also be evaluated using the 1-D numerical code. Such an exercise can provide quick guidelines regarding quantities to be disposed and related safety requirements. The experience gained at this disposal site could be advantageously applied to other similar sites to carry out safety assesment.
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