EXPERIENCES FROM THE SOURCE-TERM ANALYSIS OF A LOW AND INTERMEDIATE LEVEL RADWASTE DISPOSAL FACILITY

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1 EXPERIENCES FROM THE SOURCE-TERM ANALYSIS OF A LOW AND INTERMEDIATE LEVEL RADWASTE DISPOSAL FACILITY Jin Beak Park, Joo-Wan Park, Eun-Young Lee and Chang-Lak Kim Korea Hydro & Nuclear Power Co., Ltd. (KHNP) Nuclear Environment Technology Institute (NETEC), P.O. BOX 149, Yusong-gu, Daejeon, , KOREA ABSTRACT Enhancement of a computer code SAGE for evaluation of the Korean concept for a LILW waste disposal facility is discussed. Several features of source term analysis are embedded into SAGE to analyze: 1) effects of degradation mode of an engineered barrier, 2) effects of dispersion phenomena in the unsaturated zone and 3) effects of time dependent sorption coefficient in the unsaturated zone. IAEA s Vault Safety Case (VSC) approach is used to demonstrate the ability of this assessment code. Results of MASCOT are used for comparison purposes. These enhancements of the safety assessment code, SAGE, can contribute to realistic evaluation of the Korean concept of the LILW disposal project in the near future. INTRODUCTION It is generally accepted that eventually radionuclides in the radioactive waste will be released and that these radionuclides will be transported to the accessible environment, initially through an engineeringbased barrier system, the ground water system in the geosphere, and subsequent release into the biosphere at the disposal site. Therefore, the long-term safety assessment of near-surface radioactive waste disposal should be described by modeling potential releases of radionuclides from 1) an engineered barrier system, 2) their movement through the geosphere, 3) the resulting distribution of radiological contamination in the biosphere, and 4) the consequent radiological dose to humans. As for the near-surface radioactive waste disposal facility in which low and intermediate-level radioactive wastes (LILW) are disposed of, the design of an engineering-based barrier system is intensively analyzed to demonstrate that the disposal practice is acceptable. This is often achieved by showing compliance with regulatory criteria. In this practice, safety assessment is a key component in evaluating the acceptability of a disposal practice. International and national efforts (1-4) have been devoted to the development and application of safety assessment methodology for the near-surface disposal of low and intermediate-level radioactive waste. An 1

2 initial attempt at developing improved confidence in safety assessment approaches for near-surface disposal facilities was the International Atomic Energy Agencies s (IAEA) Co-ordinated Research Programme (CRP) on the Near-Surface Radioactive Waste Disposal Safety Assessment Reliability Study (NSARS) (1,3), which developed confidence in the analytical process by conducting inter-comparisons between approaches for specific test cases that represented typical safety assessment problems. To enhance the experience of NSARS and to review the assessment approaches and tools for proposed and existing near-surface radioactive waste disposal facilities, the new CRP on the Improvement of Safety Assessment Methodologies for Near-Surface Waste Disposal Facilities (ISAM) was launched. The Vault Safety Case (VSC) approach was used for the reference near-surface radioactive waste disposal facility in Vaalputs (South Africa) and showed that particular objectives were fulfilled and a number of limitations were recognized. System-level assessment tools, named as SAGE (SAGENF and SAGEFF), were developed by NETEC, KAERI and Monitor Scientific LLC. as the safety assessment code to be employed on the Korean concept of a LILW disposal facility. SAGENF calculates radionuclide releases from waste packages placed in a vault and transport through the engineered barrier system (EBS) and unsaturated layers before reaching the saturated zone. SAGEFF calculates transport of radionuclides in the far field and subsequent release into the biosphere. Following the development and testing of SAGE in FY 2001, NETEC/KAERI enhanced the functionality and features of the SAGE assessment codes in FY In this study, enhanced features of SAGE, especially SAGENF, are discussed to highlight the performance of the Korean safety assessment code. The problem that calculated during IAEA s Vault Safety Case (VSC) is used to show three enhanced features: 1) effects of degradation mode of the engineered vault system, 2) effects of dispersion phenomenon in the unsaturated zone and 3) effects of time-dependent sorption coefficients in the unsaturated zone. VAULT SAFETY CASE (VSC) FOR LIQUID RELEASE CALCULATIONS FOR THE DESIGN SCENARIO Conceptual Model for the Near-field Consistent with the design scenario description of IAEA (4-6), the disposal facility is a set of 20 concrete vaults located above ground level. The waste disposal area contains two lines of 10 vaults. Each vault including four 200l drums contains 4,692 concrete cubes in array. Approximate dimensions of the disposal area are 170 by 210 m giving a surface area of 35,700 m ,000 m 3 of grouted waste is disposed in standard 200l drums and placed into concrete cubes, and grout filled in between the drums. 2

3 The facility has a total of about 750,000 drums (37,500 per vault). Each vault has internal dimensions of 9m high by 20.5m wide by 83m long allowing concrete cubes to be stacked in an array 4 high x 17 wide x 69 long. It is assumed that the near-field barriers degrade with time. The drums are assumed to remain intact for 100 years and the concrete cubes are assumed to physically fail after 300 years of closure and chemically degrade over a 1,000 year period from site closure. The cap is assumed to be maintained during the 100- year active institutional control period, but then starts to degrade so that it allows 50 % of the total precipitation to pass up to 500 years, and no longer limits the rate of water infiltration after 500 years. The near field barrier is also assumed to be degraded chemically after 500 years, so that the distribution coefficient for degraded vault is used for the safety assessment thereafter. The release of radionuclides from the engineered vault and their transport through the unsaturated zone are analyzed in IAEA s Vault Safety Case (VSC) approach. Table 1 lists the radionuclide inventory at facility closure and decay chains considered in this safety assessment. Physical and hydraulic properties of the unsaturated zones are listed in Table 2. Table 1. Radionuclide inventory at facility closure and decay chains Nuclides Inventory (Bq) Decay Chain H-3 1E+15 C-14 1E+13 Ni-63 1E+10 Sr-90 1E+15 Tc-99 1E+14 I-129 3E+10 Cs-137 6E+09 U-234 8E+15 Th-230 Ra-266 Pb-210 U-238 5E+10 U-234 Th-230 Ra-226 Ph-210 Po-210 Pu-238 2E+10 U-234 Th-230 Ra-226 Ph-210 Po-210 Pu-239 3E+10 U-235 Pa-231 Ac-227 Pu-241 6E+11 Am-241 Np-237 Pa-233 U-233 Th-229 Am-241 2E+10 Np-237 Pa-233 U-233 Th-229 TOTAL 1E+16 ENHANCEMENT OF THE SAFETY ASSESSMENT FOR THE NEAR-SURFACE DISPOSAL FACILITY Effects of degradation mode of the engineered vault system The Engineering Barrier System (EBS) is designed to protect the radioactive waste matrix from surface 3

4 water infiltration. After site closure, it is assumed that the near-field barriers degrade with time. The drums are assumed to remain intact for 100 years after site closure, and then start to degrade until the 500 year. After 500 years, the engineering vault no longer acts as a barrier system. This degradation mode follows infiltration of surface water into the engineered vault system. Table 2. Physical and Hydraulic Properties of Geosphere a (4) Lithology Bulk Total Water Longitudinal Thickness Density Porosity Content Dispersivity (m) (kg/m 3 ) (-) (-) (m) c Red sand/calcrete b Brown sand/gritty clay White Kaolinite clay Wheathered Granite a Site specific data b Includes the sand base to the disposal facility c Calculated assuming that longitudinal Dispersivity is 10% of the length of the flow path(thickness) This degradation mode and resulting infiltration are considered in the enhanced SAGENF program. Fig. 1. shows the inter-comparison of vault degradation mode effects and subsequent infiltration into the engineered barrier system between SAGENF and MASCOT (7-9), which consider that there is no degradation of the vault system from site closure until 500 years. Radionuclide fluxes at the outlet of Weathered granite are calculated in both SAGENF and MASCOT throughout this paper. In Fig. 1., the vault degradation mode of SAGENF shows earlier arrival of radionuclides and shows a lower peak flux to the saturated zone for all radionuclides of interest when compared with the result of MASCOT. Effects of dispersion phenomenon in the unsaturated zone By allowing input of dispersivity (dispersion length in meters) for the unsaturated zone, the dispersion phenomenon is implemented in SAGENF. The dispersion coefficient for a given zone, i, is calculated by: D = D + α v (Eq.1) i eff, i i o where Di is the dispersion coefficient [m 2 /yr], Deff, i is the effective molecular diffusion coefficient [m 2 /yr], αiis the Dispersivity [m], and vo is the ambient infiltration rate [m/yr]. This dispersion coefficient is then used to calculate the mass transfer coefficient between compartments. 4

5 1x10 0 1x10-1 1x10-2 Vault Degradation Model: SAGENF MASCOT I-129 U-238 Radionuclide Flux (mol/yr) 1x10-3 1x10-4 1x10-5 1x10-6 1x10-7 1x10-8 1x10-9 Tc-99 U-234 C-14 1x Time after disposal (year) Th-230 Fig. 1. Comparison of radionuclide fluxes: Effect of degradation mode of engineered vault system between SAGENF and MASCOT 1x10 0 1x10-1 1x10-2 Dispersion in Unsaturated Zone: With Dispersion by SAGENF Without Dispersion by SAGENF 1x10-3 Radionuclide Flux (mol/yr) 1x10-4 1x10-5 1x10-6 1x10-7 1x10-8 1x10-9 1x10-10 Tc-99 I-129 U-238 U-234 C Time after disposal (year) Fig. 2. Comparison of radionuclide fluxes: Effect of dispersion phenomena in unsaturated zone 5

6 Fig. 2. shows the release rate comparison between two examples: one without dispersion and one with dispersion. In this example, there are four unsaturated layers, and their dispersivities are assumed to be: 0.27, 0.85, as indicated in Table 2 respectively. The ambient flow infiltration rate is m/yr. With the values assigned to each layer as shown above, the case with dispersion result in an earlier arrival of radionuclide at the saturated zone. The dispersivity for the waste, the EBS, and the saturated zone is set to 0.0 in this calculation. Dispersion phenomenon considers the spreading of individual radionuclides in porous media. Fig. 2. shows this spreading of each radionuclide of interest as expected. Earlier arrival of radionuclides to the saturated zone is illustrated. As for peak fluxes of radionuclides, dispersion does not contribute to as significant a reduction as the vault degradation mode. Effects of time-dependent sorption coefficients in the unsaturated zone SAGENF is enhanced to address the capability of allowing time dependent sorption coefficients in the near field. This feature can be used to model groundwater geochemistry changes due to alkaline plumes released from the dissolving cement waste matrix as well as the concrete vault structure. To model this effect, the approach is the same as that used in modeling time dependent infiltration rate inside the vault. Hence, the model assumes that K d for each element and in each material in the near field is a piecewise step function of time. Sorption coefficients whose time histories follow the piecewise step function are given in general by: n 1 K ( t) = K [ h( t t ) h( t t )] + K [ h( t t ) h( t t )] + K [ h( t t ) h( t t )], (Eq. 2) d d,1 s 1 d, i+ 1 i i+ 1 d, n+ 1 n e i= 1 where ts and te are the starting and ending times of computation [yr], respectively, n is the number of switching times between ts and t e, and h(x) is the unit step function. Fig. 3. shows the effects of time dependent sorption coefficients in the unsaturated zone. In this calculation only one switching time at 500 yr is considered. Before and after this degradation switching time (500yr), sorption coefficients listed in Table 3 are applied separately as the non-degraded and degraded value. The left column in each unsaturated layer is the non-degraded value and the right column is the degraded value, respectively, in Table 3. Geochemical changes due to alkaline plumes from the concrete barrier are specifically considered. The equilibrium sorption coefficients of the concrete barrier are reduced at the assumed switching time as shown in Table 3. 6

7 Element Table 3. Unsaturated Zone Distribution coefficients for Near-field (m 3 /kg) Concrete Nondegraded degraded Red sand/ Calcrete Nondegraded degraded Brown sand/ Gritty clay Nondegraded degraded White Kaolinite clay Nondegraded degraded Weathered Granite Nondegraded degraded H 0.0e+0 0.0e+0 0.0e+0 0.0e+0 0.0e+0 0.0e+0 0.0e+0 0.0e+0 0.0e+0 0.0e+0 Cl 2.0e+0 2.0e-1 5.0e-3 5.0e-3 5.0e-3 5.0e-3 1.0e-3 1.0e-3 5.0e-3 5.0e-3 Ni 1.0e-1 1.0e-2 4.0e-1 4.0e-1 4.0e-1 4.0e-1 6.0e-1 4.0e-1 4.0e-1 4.0e-1 Sr 1.0e-3 1.0e-3 8.8e-3 8.8e-3 7.1e-3 7.1e-3 8.3e-3 8.3e-3 5.5e-3 5.5e-3 Tc 1.0e-3 0.0e+0 1.0e-4 1.0e-4 1.0e-4 1.0e-4 1.0e-3 1.0e-3 1.0e-4 1.0e-4 I 1.0e-2 1.0e-3 1.0e-3 1.0e-3 1.0e-3 1.0e-3 1.0e-3 1.0e-3 1.0e-3 1.0e-3 Cs 2.0e-2 2.0e-2 5.4e-1 5.4e-1 3.4e-1 3.4e-1 2.2e-1 2.2e-1 2.6e-1 2.6e-1 Pb 5.0e-1 5.0e-2 3.0e-1 3.0e-1 3.0e-1 3.0e-1 5.0e-1 5.0e-1 3.0e-1 3.0e-1 Po 0.0e+0 0.0e+0 1.5e-1 1.5e-1 1.5e-1 1.5e-1 3.0e+0 3.0e+0 1.5e-1 1.5e-1 Ra 5.0e-2 5.0e-2 5.0e-1 5.0e-1 5.0e-1 5.0e-1 9.0e+0 9.0e+0 5.0e-1 5.0e-1 Ac 1.0e+0 2.0e-1 3.4e-1 3.4e-1 3.4e-1 3.4e-1 7.6e+0 7.6e+0 3.4e-1 3.4e-1 Th 5.0e+0 1.0e+0 3.0e+0 3.0e+0 3.0e+0 3.0e+0 6.0e+0 6.0e+0 3.0e+0 3.0e+0 Pa 5.0e+0 1.0e-1 3.4e-1 3.4e-1 3.4e-1 3.4e-1 7.6e+0 7.6e+0 3.4e-1 3.4e-1 U 2.0e+0 1.0e-1 2.5e-3 2.5e-3 6.8e-3 6.8e-3 1.4e-3 1.4e-3 3.0e-3 3.0e-3 Np 5.0e+0 1.0e-1 3.4e-1 3.4e-1 3.4e-1 3.4e-1 7.6e+0 7.6e+0 3.4e-1 3.4e-1 Pu 5.0e+0 1.0e-1 3.4e-1 3.4e-1 3.4e-1 3.4e-1 7.6e+0 7.6e+0 3.4e-1 3.4e-1 Am 1.0e+0 2.0e-1 3.4e-1 3.4e-1 3.4e-1 3.4e-1 7.6e+0 7.6e+0 3.4e-1 3.4e-1 1x10 0 1x10-1 1x10-2 Time dependent sorption coeff., K d : K d change at 500 yr Constant K d U-238 1x10-3 Radionuclide Flux (mol/yr) 1x10-4 1x10-5 1x10-6 1x10-7 1x10-8 1x10-9 Tc-99 I-129 U-234 C-14 1x Time after disposal (year) Fig. 3. Comparison of radionuclide fluxes: Effect of time dependent K d in unsaturated zone 7

8 When the higher equilibrium sorption in the concrete barrier changes at the switching time to a reduced sorption value, some of radionuclides sorbed onto the concrete solid can migrate into the porous media. These enhanced concentrations cause higher peak fluxes compared to the case of no switching time. (10) This means that the radionuclide flux after the peak value tails off more rapidly than that of the no switching case. This physical phenomenon is depicted in Fig. 3. In this section, independent evaluation of several features embedded in the newly developed safety assessment code, SAGENF such as: 1) effects of degradation mode of the engineered vault in Fig. 1.; 2) effects of dispersion phenomenon in the unsaturated zone in Fig. 2.; and finally, 3) effects of time dependent sorption coefficients in the unsaturated zone in Fig. 3. Fig. 4. shows the simultaneous effects of the individually evaluated features discussed. CONCLUSIONS Enhancement of the SAGE computer code for evaluation of the Korean concept for a LILW waste disposal facility is discussed. The program SAGE is being developed by NETEC, KAERI and Monitor Scientific LLC. Several features are embedded into SAGE to analyze: 1) effects of degradation mode of the engineered barrier; 2) effects of dispersion phenomena in the unsaturated zone; and 3) effects of time dependent sorption coefficients in the unsaturated zone. IAEA s VSC approach is used to demonstrate the ability of the assessment code. Results of MASCOT are used for comparison purposes with the results of SAGE, especially SAGENF. The vault degradation mode of SAGENF shows an earlier arrival of radionuclides and lower peak flux to the saturated zone for all radionuclides of interest when compared with the result of MASCOT. Earlier arrival due to the dispersion phenomenon is illustrated. As for peak fluxes, dispersion does not contribute to a significant peak reduction compared to the results of the vault degradation mode. When higher equilibrium sorption in the concrete barrier changes at the switching time to a reduced sorption value, some of the radionuclides sorbed onto the concrete solid can migrate into the porous media. These concentrations cause higher peak fluxes with narrower flux bands compared to those of the no switching case. These enhancement of the safety assessment code, SAGE, can contribute to more realistic evaluations of the Korean concept of the LILW disposal project in the near future. ACKNOWLEDGEMENT The authors would like to acknowledge the technical support of program SAGE by Dr. Wei Zhou and Dr. Matt Kozak, Monitor Scientific LLC., USA. This work is carried out as a part of the National Research Laboratory (NRL) program financially supported by the Ministry of Science and Technology, Korea. 8

9 Radionuclide Flux (mol/yr) Radionuclide Flux (mol/yr) 1x10 0 1x10-2 1x10-4 1x10-6 1x10-8 1x x10 0 1x10-2 1x10-4 1x10-6 1x10-8 1x10-10 SAGENF: C-14 Tc-99 I-129 U-238 U-234 Th-230 Ra-226 Pb-210 MASCOT C-14 TC-99 I-129 U-238 U-234 TH-230 RA-226 PB Time after disposal (year) Fig. 4. Comparison of radionuclide fluxes: Full effects of 1) vault degradation mode, 2) dispersion phenomena and 3) time dependent sorption coefficient in unsaturated zones 9

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