RELATIONSHIP BETWEEN NUCLEAR FUEL CYCLE PARAMETERS AND HLW REPOSITORY PERFORMANCE

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1 RELATIONSHIP BETWEEN NUCLEAR FUEL CYCLE PARAMETERS AND HLW REPOSITORY PERFORMANCE Joonhong Ahn Department of Nuclear Engineering University of California, Bereley Introduction and Bacground Environmental impacts arising from deep-geologic disposal of high-level radioactive wastes (HLW) containing long-lived radionuclides are considered to be contentious, due to the long time-scale involved. The feasibility of geologic disposal of HLW (including spent fuel) has been confirmed by previous performance-assessment studies [1, 2, 3] carried out by several countries for a variety of host geologic formations and different forms of HLW. The difficulties in geologic disposal have also been recognized, such as understanding the nature and rates of geologic processes, predicting long-term environmental changes, predicting repository and waste-pacage performance, and predicting long-term human behavior for the purposes of ris estimation, for more than ten thousand years [4]. A future nuclear power system is not liely to be accepted by the public if it generates as much radioactive waste as the current system does. For an enduring nuclear fuel cycle to be accepted by the public, a clear vision should be presented of the quantity and the toxicity of wastes arising from the enduring system. These should be significantly smaller than those needed by the current system to support a long-term commitment to nuclear fission. Concern over proliferation and weapons-usable nuclear material will increase as more countries have nuclear power systems, as radiation barriers decline in spent fuel, and as the number of waste-storage locations increases [5]. More countries, especially those countries recognized as Newly Industrialized Economies in Asia, are liely to have nuclear power capacity in the future [6], because carbon dioxide emission must be limited while maing economic growth. Proliferation-resistant recycling methods that permit termination of safeguards monitoring on resulting HLW provide the best method to control proliferation ris from long-term use of fission energy, but require systematic life-cycle study. Partitioning-and-transmutation (P/T) systems designed to maximize the rate of transmutation of problematic long-lived nuclides [7, 8, 9] could be a remedy for reducing the difficulties of geologic disposal. However, previous analyses for the impacts of P/T systems on the performance of geologic disposal [1, 11, 12] show that effects of P/T systems are not ustifiable due to their high costs as they provide only 61

2 modest improvements in radiological safety. This conclusion is based on the observation that, due to low solubilities and high sorption distribution coefficients of actinides in geologic formations, decrease in the masses of actinides in the waste does not linearly decrease the actinides' radiological impact on future public. This insensitivity of predicted repository performance to fuel cycle schemes is because repository performance is currently measured by exposure dose rates determined primarily by concentrations of radionuclides in groundwater, whereas effects of fuel-cycle schemes manifest as reductions in masses of radionuclides contained in the waste streams. Recently, a mass flow analysis and a repository-performance-assessment study [13, 14] have been performed for the Accelerator-driven Transmutation of Waste (ATW) system, conceived of by a team at Los Alamos National Laboratory [8]. By the mass flow analysis, estimated is the mass of individual radionuclides that would be destined for the geologic repository, as well as the mass of each radionuclide transmuted per passage through the reactor. With the calculated waste mass and waste composition, the radiological impact of geologic disposal of the waste was estimated by mass transport analysis. Neptunium-237 was found to be a dominant contributor to the long-term radiological hazard in the case that the radioactive wastes generated from the P/T system are disposed of in the proposed Yucca Mountain Repository (YMR) [13, 14]. For significant reduction of the mass of 237 Np that is to be disposed of in the repository, not only 237 Np but also 241 Am and 241 Pu must be fissioned effectively by the transmuter, because 241 Pu decays with the half-life of 13.2 yr to 241 Am, which subsequently decays to 237 Np with the half-life of 458 yr. These previous studies, however, were done by crudely coupling existing models for the fuel cycle and the repository performance. To better understand the impacts of P/T systems on geologic disposal, we need to develop an integrated model, into which relevant parameters are incorporated. With such models, new measures for repository performance could be proposed, in addition to the exposure dose rate by radionuclides taen by human beings. New measures, especially those based on masses of radionuclides to be disposed of in a geologic repository, are expected to express effects of fuel cycle system parameters on repository performance more sensitively. In this study, a mass flow analysis has been performed for the evolution of actinide masses in a hypothetical P/T system with the nuclide chain, β-decay 238 U(n,γ) 239 Pu(n,γ) 24 Pu(n,γ) 241 Pu 241 Am 237 Np. Then, with the result of the mass flow analysis, a repository performance model has been established. The masses of a radionuclide retained in the repository region and in the far-field region are proposed as mass-based repository performance measures. It is investigated if repository performance can be differentiated by these measures for wastes with different characteristics generated by different fuel cycle schemes. For illustration, the masses of 237 Np have been calculated. α-decay Mass Flow Model for P/T Schemes 62

3 Model The model concept is depicted in Fig. 1. In the recycling scheme, in the first cycle, it is assumed that after irradiation for a period, T, a fraction ƒ of the core is discharged. The discharged material undergoes removal of fission products and fabrication of new fuel assemblies, where actinides are lost as wastes with the assumed mass fraction α. The mass of removed fission products and the mass of actinides lost as waste during partitioning and fuel fabrication are made up at the beginning of the second cycle by fresh material consisting of actinides with the initial compositions M o [mol/g of heavy metal (HM)]. Time for partitioning and fuel-assembly fabrication is not included in the figure or the calculations. Fuels added to the core at the beginning of each cycle are assumed to be homogeneously mixed with the existing fuels. The governing equations for the concentration, M () t [mol/g-hm], of radionuclide in the -th cycle which is related to time variable t by ( 1 ) T < t < T, = 1, 2,..., are written as dm = P 1M 1 DM, = 1, 2,..., 6, P =, dt (1) subect to M (( 1 ) T)= M, = 1, 2,... (2) The concentration M of radionuclide at the beginning of the -th cycle is given as 1 M = 1 f M 1 T ( ) ( ) ( ) 1 + f ( 1 α) M (( 1) T)+ M o 1 1 α ( ) 6 l= 1 1 Ml (( 1) T) 6, o M l l= 1 =1, 2,..., 6, = 2,..., M 1 o = M. (2) For 239 Pu ( = 2), 24 Pu ( = 3), and 241 Pu ( = 4), the production coefficients P 1 [yr 1, 1 1 ] are equal to σ n γ Φ, where σ n, γ [cm 2 ] is an effective one-group capture crosssection of the precursor, and Φ [cm -2 yr -1 ] the total neutron flux in the transmuter. For 241 Am ( = 5) and 237 Np ( = 6), the production coefficients are equal to the decay constant of their precursors, λ 1. The destruction coefficients, D [yr 1 ], for the first three members are σ a Φ, where σ a is the effective one-group absorption cross-section of radionuclide. For the last three members, the destruction coefficients are written as λ + σ aφ. The MCNP code and its associated cross-section library [15] are used to generate the σ s and Φ. The flux amplitude is set to provide an average power density corresponding to a beginning-of-life specific power of 75 W/g HM. The transmuter is assumed to operate at the constant power density with a hard neutron spectrum. The value of Φ is assumed to be 9.8E15 cm -2 yr -1. For the governing equations (1), analytical solutions are readily available [16]. 63

4 The production and destruction coefficients of the six members [17] are shown in Table 1. M o are determined by assuming that 99.8% of uranium is removed from LWR spent fuel that has undergone 3 years of cooling. The transmuter core is assumed to contain 3 tons of heavy metals. With ƒ = 1/3 assumed, one ton of irradiated fuel is discharged from the core at the end of each irradiation period. We assume that T = 1/3 year or 3 year, and α =.32 [13, 14] or 1. With α = 1, a once-through scheme, such as the one proposed by [18], can be investigated. Recycling is assumed to continue until 63, MT of spent fuel has been utilized. At the end of 1 th cycle, Homogeneous Core Discharged Fuel 1 M (( 1) T) 1 f Remaining Fuel 1 M (( 1) T) 1 f f f Returning Fuel R + 6 l = 1 M M l W α Loss aswaste 1 L = α M (( 1) T) Recycling 1 R = ( 1 α ) M (( 1) T) Maeup 6 6 = 1 = 1 W = M R Reprocessed LWR spent fuel At the beginning of th cycle, the concentration of -th member is 1 M = ( 1 f) M (( 1) T) + f R + M 6 W M l l = 1 Fig. 1 Mass-flow model Table 1 Parameters for Recycle and Transmutation Actinides Inventory Concentration (mol/g HM) M o Absor ption Cross Section ( barn) σ a Destruction Coefficient (1/yr) Capture Cross Section (barn) 1 σ n, γ Decay Constant (1/yr) λ 1 Production Coefficient (1/yr) P 1 1 U ÐÐÐ ÐÐÐ 2 Pu (1.55E-1) Pu (2.88E-5) Pu (1.6E-4) Am ÐÐÐ 4.83E Np ÐÐÐ 1.6E-3.16 Figure 2 shows the evolution of the masses of the six isotopes in the transmuter. The zigzag profile is obtained due to the addition of mae-up fuel between two consecutive cycles. The mass at the bottom of one zigzag is the mass at the end of one irradiation period, or M 1( ( 1) T). The mass at the top of a zigzag is the mass at the beginning of the irradiation period, or M, given by eq. (2). D 64

5 In the top-left figure (T = 1/3 year and α =.32), the 238 U concentration shows an increasing tendency because more mass of 238 U is added than that is transmuted between two consecutive cycles. The concentrations of the other nuclides evolve towards an unchanging pattern as the number of cycles increases. This implies that the radionuclide added at the beginning of each cycle is transmuted in the same cycle, or approximately at a constant rate. The concentrations of 239 Pu, 241 Am, and 237 Np decrease for the first 1 cycles, whereas those of 24 Pu and 241 Pu increase. In the top-right figure (T = 1/3 year and α = 1), the discharged target is directly transferred to a repository. After 1 yr (or 3 cycles), the unchanging pattern of concentrations is established for all 6 radionuclides. Two figures at the bottom are for the cases with T = 3 years. Due to the longer irradiation time, the masses at the bottom of each zigzag are substantially smaller than those with T = 1/3 yr. Profiles for two different α values with the same T = 3 years loo similar to each other. Because greater masses of actinides are transmuted in each irradiation period with the longer T, the initial compositions M of the target for the next irradiation period are dominated by the mae-up material f=1/3, T=1/3 year, α =.32 Pu-239 Pu-24 U f=1/3, T=1/3 year, α = 1 Pu-239 Pu Am-241 Pu Am-241 Np-237 U-238 Np Cycle nuber (3 cycles = 1 year) Pu Cycle number (3 cycles = 1 year) Mass in transmuter core, mol f=1/3, T=3 years, α =.32 Pu-239 Pu-24 U-238 Pu-241 Am-241 Np Cycle number (1 cycle = 3 years) Mass in transmuter core, mol f=1/3, T=3 years, α = 1 Pu Pu-24 U Pu-241 Am Np Cycle number (1 cycle = 3 years) Fig. 2 Masses of six isotopes in transmuter core. Waste Accumulation For the repository performance, 241 Pu, 241 Am, and 237 Np are of maor concern. The mass of each radionuclide destined to a repository is obtained by the mass M ( 1) T at the bottom of each zigzag multiplied by the fraction, α. 1( ) 65

6 Table 2 shows the cumulative masses of these radionuclides lost as waste. The cumulative mass is obtained for each radionuclide by summing the mass αm 1( ( 1) T) over. The values for the case labeled "No P/T" are the masses of the six radionuclides included in 63, tons of LWR spent fuel. The sum of the cumulative masses of 241 Pu, 241 Am, and 237 Np in the waste stream for α = 1 are about two orders of magnitude greater than those for α =.32 (the right-most column). For the same α, the sum is smaller for the longer irradiation period, T. Table 2 Cumulative Masses of Radionuclides in Waste Stream α T cycle Cumulative radionuclide mass in waste (mol) Sum of 241 Pu, 241 Am, (yr) # U-238 Pu-239 Pu-24 Pu-241 Am-241 Np-237 and 237 Np (mol) No P/T 5.E+5 1.4E+6 5.9E+5 1.E+5 2.5E+5 1.4E+5 4.9E E+4 7.4E+3 1.1E+4 2.1E+3 1.3E+3 1.5E+3 5.E E+5 8.E+5 5.6E+5 9.9E+4 1.5E+5 1.1E+5 3.6E E+3 4.1E+2 9.1E+2 1.9E+2 6.8E+1 1.1E+2 3.8E E+5 9.3E+4 2.1E+5 4.4E+4 1.8E+4 2.9E+4 9.1E+4 Repository Performance Model Model The obective of this section is to establish a mathematical model to define a mass-based repository performance measure. We consider the waste region and the geologic medium, as shown in Fig. 3. The waste region has the cross-sectional area of A [m 2 ] and the length, L [m] initially. The region consists of the solid phase and the pores completely filled with water. The porosity is ε. Initially, the radionuclide concentration in the solid phase is homogeneously S o [mol/m 3 ]. The radionuclide in the solid phase of the waste region dissolves into the water phase after t =. The dissolution rate of the radionuclide is f d (x, t) [mol/m 3 yr]. Water in the waste region flows at the Darcy velocity of v D [m/yr]. Because the concentration C is assumed to be always zero at the left side of the waste, where uncontaminated water flows in, the dissolution of the radionuclide is fastest at the left-side boundary, as is assumed by eq. (6) later. The waste region shrins with time. When the left side, x = l( t), coincides with the right-side boundary, x = L, then the entire waste has dissolved out. If we neglect the decay precursor, dispersion and sorption retardation of the radionuclide in the waste region, the transport equations for the radionuclide in the waste region can be written as ε C C + vd + λεc = fd( x, t), t >, l ( t) < x < L, (3) t x ds ( 1 ε) + λ( 1 ε) S = fd ( x, t), t >, l ( t) < x < L, (4) dt 66

7 Cx (, ) =, Sx (, ) = So, l( t) < x< L, l( ) =, subect to (5) C(, t) =, t > where C(x, t) and S(x, t) are the concentrations [mol/m 3 ] of the radionuclide in the water phase and in the solid phase of the waste region, respectively. λ is the decay constant of the radionuclide [yr 1 ]. Equation (6) is assumed for the dissolution rate, f d (x, t), considering that the radionuclide is hardly soluble into water. { } > < < f ( x, t ) = F C * d C ( x, t ), t, l ( t ) x L. (6) This means that the dissolution of the radionuclide into water is limited by its solubility C*, and is driven by the concentration difference between the water-phase concentration C(x, t) at the local point and the solubility. The quantity, F, is the fractional release rate [yr 1 ] of the radionuclide. The radionuclide in the water phase of the waste region is released into the geologic medium by advection. Assuming that the porosity of the geologic medium is ε', and that the sorption retardation is characterized by the capacity factor, α p, defined in Figure 2, the transport equation for the concentration of the radionuclide in the water in the pore of the geologic medium can be written as α N N + v + p D λα N =, t >, x > L, (7) p t x Nx (, ) =, x> L subect to (8) NLt (, ) = CLt (, ), t> The second boundary condition assures the concentration continuity at the interface between the waste region and the geologic medium. We select an arbitrary point at x = X in the geologic medium. The region, L < x < X, is considered to be inside the repository, whereas the region, x > X, is exterior to the repository, or the far-field region (Fig. 3). Then, the mass, M r (t) [mol], of the radionuclide in the repository region, i.e., < x < X, is obtained as M( t) = 1 ε Sxtdx (, ) εcxtdx (, ) α Nxtdx (, ), (9) r L ( ) + + L whereas the mass, M f (t) [mol], of the radionuclide existing in the far-field region is written as L Darcy velocity, v D Darcy velocity, v D Fig. 3 Porosity, ε Initial conc. In solid, S o A Inside repository Far field X waste Geologic medium Configuration for repository performance model X L p Porosity, ε Capacity factor, α p = ε + (1 ε )ρkd 67

8 Mf() t = α pn( x,) t dx. (1) X Analytical solutions for C, S, and N have been obtained. It has been confirmed analytically that the sum of two masses M f + M f is equal to S o AL(1 ε) exp(-λt), where S o AL(1 ε) is the initial mass of the radionuclide in the solid phase of the waste. Waste Form Characterization Table 3 shows the numerical values assumed for the numerical results of the repository performance assessment, shown in Figures 4 to 7. In this illustration, 237 Np is considered. The mass of 237 Np included in the waste is obtained by lumping the masses of its two precursors, which is given in the last column of Table 2. Table 3 Input Data for Repository Performance Assessment Parameters α =.32 α = 1 T = 3 yr T = 1/3 yr T = 3 yr T = 1/3 yr Total Mass of Np-237 in repository (mol) (from Table 2) 3.8E+2 5.E+3 9.1E+4 3.6E+5 Number of canisters 1 79 Mass of Np-237 in one canister (mol) 3.8E+2 5.E E E+3 Canister cross-sectional area (m 2 ) A Canister thicness (m) L Canister volume (m 3 ) V Initial Np-237 concentration in waste solid (mol/m 3 ) S o Porosity of waste ε.3 Darcy velocity (m/yr) v D.3 Pore velocity in waste (m/yr) v 1. Porosity of geologic medium ε'.1 Density of geologic medium (g/m 3 ) ρ 22 Half-life of Np-237 (yr) 2.14E+6 Decay constant of Np-237 (yr -1 ) λ 3.24E-7 Fractional release rate for Np-237 (yr -1 ) f 1.E-5 1.E-5 d 1.E-4 1.E-4 1.E-5 1.E-5 1.E-5 Solubility of Np (mol/m 3 ) C* 1.E-2 Sorption distribution coefficient for Np (m 3 /g) Kd 1.E-2 After we have quantified the mass of radionuclides in the waste stream by the mass flow analysis, we need to perform a waste characterization analysis, prior to the repository performance assessment. In such characterization, the waste matrix, the volume of the waste, the radionuclide concentration in the waste matrix, and the mechanism and rate of the waste matrix dissolution should be identified. These are primarily determined by the waste treatment process, or the partitioning process. It is assumed for α = 1 that the fuel discharged from the transmuter is directly disposed of in the repository. The waste canister is assumed to contain 9 ton of the discharged fuel in the volume of 8.9 m 3 (1.5 m in diameter and 4.9 m in height). These values are the same as those of the Multi-Purpose Canister (MPC) considered for the disposal of LWR spent fuel at the YMR [3]. Because one ton is assumed to be discharged per T years, the total mass of the discharged fuel after having utilized 68

9 63, tons of LWR spent fuel by the assumed P/T schemes is obtained by one ton times the number (713, see Table 2) of cycles. The number (79) of canisters is obtained by dividing the total mass of the discharged fuel by 9 ton. Each canister is assumed to be placed in the repository separately, as LWR spent fuel canisters are planned to be placed in the YMR. It is assumed for α =.32 that the waste could be categorized in TRU waste. TRU waste is characterized by relatively large volume with a low concentration of radionuclide. Based on the experience with LWR spent fuel reprocessing, it is assumed that.38 m 3 [19] of TRU waste is generated by processing of one ton of discharged fuel. Because one ton of discharged fuel is processed per cycle, the total volume of the waste is obtained by.38 m 3 times the number of cycles. Furthermore, it is assumed that all the waste is stored in a large silo-type cavern. Values of other parameters, such as the Darcy velocity, porosities, the solubility and the sorption distribution coefficient of neptunium, have been chosen as shown in Table 3 for an illustrative purpose. Numerical Illustrations for Repository Performance Figure 4 shows the mass release rate, εavd N( X, t) [mol/yr] of neptunium at the repository boundary, x = X. This is considered to be a concentration-based performance measure. If we assume that the water at this location is taen by human beings, this release rate can be converted to a dose rate. Figure 4 shows that the release rate does not differ for differing waste characterizations. Thus, it is observed that the concentration-based performance measure is insensitive to fuel cycle schemes. This is observation is commonly made by the previous studies for P/T schemes [1, 11, 12]. Figure 5 shows the masses of neptunium in the repository region, M r (t), and in the far-field region, M f (t). The solid curves represent the sum, M r (t)+m f (t). Notice that the sum decreases with time exponentially, as has been confirmed by the analytical solutions. The sum is determined solely by the decay constant and the initial mass of the radionuclide in the waste. Thus, the uncertainty associated with the solid curves could be very small. The mass, M f (t), in the far-field region appears at 3, years, increasing with time. The curve for M f (t) eventually meets with the solid curve, meaning that all the radionuclide existing at that time exists in the far field. Because the volume of the waste is identical, and because the concentration in the water phase is limited by the solubility, two curves of M f (t) for α = 1 are identical. However, the repository performance is different for α =.32, because the waste dimensions and the initial concentration of the radionuclide in the waste solid are different. The more frequently the recycling is made, the worse the repository performance would be. Figure 6 shows the effect of the fractional release rate on the release of neptunium for the cases with α =.32. Here, two values are assumed for the fractional release rate: yr 1 and yr 1. The mass of neptunium in the far-field region could be reduced by a factor of 5 or so by improving the fractional release rate by one order of magnitude from to yr 1. The improved 69

10 performance is comparable to the performance for the case with α = 1. This figure indicates that improving the performance of a repository for diluted waste might not be so effective. Figure 7 shows the effect of the waste dimensions. The fractional release rate is fixed at yr 1. Decreasing the cross-sectional area normal to the water flow improves the repository performance. Release rate, mol/yr 1.E-2 1.E-3 1.E-4 1.E-5 1.E-6 α = 1 α =.32, f d = 1e-5 /yr T = 3 yr T= 1/3 yr Mass of Np-237, mol 1.E+6 1.E+5 1.E+4 1.E+3 1.E+2 1.E+1 α=1, T=1/3yr α=1, T= 3yr α=.32, T= 1/3yr α=.32, T= 3yr Mf M f+m r Mr 1.E-7 1.E+3 1.E+4 1.E+5 1.E+6 1.E+7 1.E+8 Time, year 1.E+ 1.E+3 1.E+4 1.E+5 1.E+6 1.E+7 1.E+8 Time, Year Fig. 4 Np mass release rate at Fig. 5 Mass of Np inside and the repository boundary outside of the repository Mass of Np-237, mol 1.E+6 1.E+5 1.E+4 1.E+3 1.E+2 1.E+1 α=1, T=1/3yr α=.32, T= 1/3yr α=.32, T= 3yr Fractional release rate 1E-4 /yr 1E-5 /yr 1.E+ 1.E+3 1.E+4 1.E+5 1.E+6 1.E+7 1.E+8 Time, year Mass of Np-237, mol 1.E+6 1.E+5 1.E+4 1.E+3 1.E+2 1.E+1 α=.32, T= 1/3yr 13 m 2 x 1 m 35.5 m 2 x 1 m 1.E+ 1.E+3 1.E+4 1.E+5 1.E+6 1.E+7 1.E+8 Time, year Fig. 6 Effect of the fractional Fig. 7 Effect of the waste release rate on the release volume on the release Concluding Remars A mass flow model has been established for the six-member transmutation chain including 237 Np for a hypothetical P/T system with a homogenized core. Four different schemes have been considered by combination of two different target irradiation times, T, and two different waste loss fractions, α. Waste from the schemes with α = 1 is assumed to have characteristics similar to the LWR spent fuel, whereas waste from the schemes with α =.32 is assumed to have characteristics similar to TRU waste. The masses of radionuclides in the repository region and in the far-field region have been defined by establishing a mathematical model for radionuclide transport, which incorporates shrining source conditions and solubility-limited release. These masses 7

11 have been proposed as the performance measures. It has been confirmed that the conventional performance measure based on the radionuclide concentration in groundwater is insensitive to the difference in PT schemes, because the radionuclide concentration in groundwater is assumed to be limited by its solubility. The repository performance measure proposed in this study could differentiate the PT schemes. The mass in the far field for the schemes with α = 1 is smaller than that for the schemes with α =.32. From the viewpoint of proliferation resistance, the schemes with α =.32 are preferable to the schemes with α = 1 because of the smaller mass in the repository. If the fractional release rate could be decreased (more leach-resistant waste form would be developed), or if the volume of the waste could be reduced, repository performance could be optimized from the viewpoint of radiological impact as well as proliferation resistance. References 1. SVENSK KARNBRANSLEHANTERING AB, SKB91: Final disposal of spent nuclear fuel. Importance of the bedroc for safety, SKB Tech. report 92-2, POWER REACTOR AND NUCLEAR FUEL DEVELOPMENT CORPORATION, Technical Report on Research and Development for Geological Disposal of High-Level Radioactive Wastes, PNC TN , Office of Civilian Radioactive Waste Management, Department of Energy, Viability Assessment of a Repository at Yucca Mountain, DOE/RW-58/V3, Board on Radioactive Waste Management, Commission on Geosciences, Environment, and Resources, National Research Council, Discussion Paper Prepared for the Worshop, "Disposition of High-Level radioactive Waste Through Geological Isolation, Development, Current Status, and technical and Policy Challenges," National Academies, Irvine, California, November 4-5, 1999, National Academy Press, Peterson, P. F., Relative Attractiveness of Reprocessing Waste and Spent Fuel: Implica-tions for Long-Term Safeguards Technical Requirements, Proc. the Int. Symp. on Energy Future in the Asia/Pacific Region, Toai University, Japan, March 27 28, 1998, Honolulu, Ahn, J., and W. E. Kastenberg, ed., Proceedings of the Nuclear Engineering Session, Industrial Liaison Program, 19th Annual Conference, March 12, 1997, edited by J. Ahn and W. E. Kastenberg, Bereley, California, Griffith, J. D., US Department of Energy, Office of Nuclear Energy, Actinide Recycle: Presentation to National Research Council Committee on Future Nuclear Power Development, January 29, For example, Arthur, E. D., Overview of a New Concept for Acclerator-Based Transmutation of Nuclear Waste, Trans. Am. Nuc. Soc., , June 1991, or 71

12 Los Alamos National Lab., Presentations Prepared for the MIT Technical Review, available at 9. Muaiyama, T., Overview of R&D Program on Nuclide Partitioning and Transmutation OMEGA in Japan, IAEA Specialist Meeting on Use of FBRs for Actinide Transmutation, Obnins, Ramspott, L. D., et al., Impacts of New Developments in Partitioning and Transmutation on the Disposal of High-level nuclear Waste in a Mined Geologic Repository, UCRL ID-1923, Lawrence Livermore National Lab., March Pigford, T. H., Actinide Burning and Waste Disposal, UCB-NE-4176 An invited Review for the MIT International Conference on the Next Generation of Nuclear Power Technology, October 5, Office of Civilian Radioactive Waste Management, Department of Energy, A Roadmap for Developing Accelerator Transmutation of Waste (ATW) Technology, A Report to Congress, DOE/RW-519, October Ahn, J., Preliminary Assessment of the Effects of ATW System Application on YM Repository Performance, Global 99, International Conference on Future Nuclear Systems, August 29-September 3, 1999, Jacson Hole, Wyoming (1999). 14. Ahn, J., P. Chambré, E. Greenspan, W. E. Kastenberg, M. D. Lowenthal, B. Par, and N. Stone, Impacts of Waste Transmutation on Repository Hazards, Global 99, International Conference on Future Nuclear Systems, August 29-September 3, 1999, Jacson Hole, Wyoming (1999). 15. Briesmeister, J. F., MCNP - A General Monte Carlo N-Particle Transport Code, Version 4A, Los Alamos National Laboratory, Report LA (1993). 16. Benedict., M., T. H. Pigford, and H. Levi, Nuclear Chemical Engineering, Second Edition, McGraw-Hill (1981) 17. Greenspan, E., and N. Stone, Personal Communication. We benefited from discussions and advice. 18. Bowman, C. L., Once-through Thermal-Spectrum Accelerator-Driven System for LWR Waste Destruction Without Reprocessing: Tier-1 description, ADNA Corporation, ADNA/ Pradel. P., Waste management optimization: The COGEMA answer, Nuclear Engineering and Design, 176,