HARMONIZED CONNECTION OF WASTE DISPOSAL AND PARTITIONING & TRANSMUTATION

HARMONIZED CONNECTION OF WASTE DISPOSAL AND PARTITIONING & TRANSMUTATION Toshiaki Ohe Department of Nuclear Engineering, Tokai University Japan Ohe@keyaki.cc.u-tokai.ac.jp Abstract High-level radioactive waste repository is designed from aspects such as mechanical stability, thermal load, and chemical environment. Some partitioning processes were developed and used to classified the nuclides into several groups from the point of process viability but not fully investigated from the impact due to leaching and migration of nuclides from the HLW repository so far. The report indicated the appropriate grouping of the reprocessed elements from both thermal and migration aspects: A(heat dominant: Sr, Eu, Cm, Am),B(heat & migration dominant: Cs), C(migration dominant: Np, Tc, Zr, Se, Pd, Sn), D(not attractive). Then comparison was then made between the present grouping and some partitioning scheme. Introduction The concept of partitioning/transmutation (P/T) has been discussed internationally [1] and domestically [] to overcome the difficulties arising from the long-term credibility of the deep geologic disposal. A number of research activities have been carried out so far and the discussions mainly focus on the technical feasibility based on the reactor physics and/or the engineering chemistry. However, it is not clear whether these activities proceed toward the final goal to sweep away the fear of the long-term issue of geologic disposal. Ultimate zero release of radio-nuclides out of the fuel cycle is practically impossible if the partial loss is taken into account [3] thus final disposal of the process wastes is, more or less, not avoidable. This means the consideration is necessary whether the process is available from both points of technical viability and disposal acceptability. However, the latter point has not been fully investigated. The reason of such gap may be caused by the lack of the tight connection among the specialists in the fields of both P/T and the geologic disposal. This paper is trying to present the practical goal of P/T process from the viewpoint of the deep geologic disposal in order to bridge the gap between P/T and disposal. Grouping of Nuclides Two aspects are considered to minimize the waste load; thermal impact due to heat generation nuclides and radiological impacts due to migration of potentially haz- 73

ardous nuclides. Migration Impact Assuming the dispersion coefficient D is approximated by D = LU p / where U p is the groundwater pore velocity and L is the migration distance, the equations of the peak nuclide concentration in groundwater are simply derived for two ultimate boundary conditions; instantaneous discharge of soluble nuclide eq.(1), and solubility-limited dissolution of insoluble nuclides eq.(), sol Cmax( L) L = exp( A L) (1) M S π [ ( )] ( ) = + insol Cmax L C0 exp 5 1 04. A L 1 () where, C max : maximum nuclide concentration L: distance from the source M: nuclide inventory S: discharge cross-section C o : nuclide solubility λ: decay factor R : retardation factor U p : groundwater pore velocity. A = λr/u p The above equations are characterized by the factor A and similar tendency for the two boundary conditions are found. This similarity implies that nuclides can be classified into groups by estimating the factor A of each nuclide even when the boundary condition is different. In particular, the factor A is divided into the nuclide dependent term λr and geologic term U p and the latter is independent of nuclide properties. Thus λr (or R/T 1/ ) term is used for the later nuclide grouping. Thermal impacts Equation (3) approximately calculates the temperature rise due to heat generation nuclides. This equation is based on the non-stationary analytic equation [3] and indicates the highest temperature rise of the surrounding geologic medium when the spherical heat source having the radius R is placed in the medium with the thermal conductivity K, max. K T 3 1 0 78 0. 1( R λ ), Wo 4πR 3 (3) where, K : thermal conductivity of surrounding medium T max : maximum temperature rise with single heat source W o : initial heat generation in a single waste R : radius of spherical heat source. Grouping of Nuclides with Impact Factors Nuclides in a spent fuel are classified into several groups by using the Eqs. (1) 74

and (3) from the viewpoints of migration and thermal impacts, respectively. The maximum temperature rise of each nuclide calculated by Eq.(3) with assumptions of 150L/MTU waste volume. In practical manner, the partitioning process can be developed by employing the elemental separation technique. The grouping was thus revised by the element basis. The maximum impact factor among the isotopes of the same element is selected and shown again with the similar figure. In this classification, peak concentration of nuclide is calculated by Eq. (1) assuming the migration length L=0[m] and unit discharge cross-section S=1[m ]. Decay after discharge from reactor 4 years 5 50years 4 3 1 0-1 - -3-4 -5-6 -7-8 -11 Ru(+Rh) Eu Ce - A. heat dominant Am Sr(+Y) Cm Sb Pu Sm Nb -9 Pm(Kd?) D. not attractive -8-7 -6 B. heat & migration dominat -5 C. migration dominat Sn -4-3 Se Pd Radiological Impact C max (L) / S I Cs(+Ba) Fig. 1 Elemental Impact factors of both thermal and radiological aspects According to the results shown in Fig.1 the four elemental groups are extracted: heat dominant (A), both heat & migration dominant (B), migration dominant (C), and not attractive (D). When the group A is separated from another groups, it is not necessary to dispose it in deep geologic formation after suitable decay period. The group B and C can be disposed together with moderate heat generation of Cs. The remaining group D, there are not important nuclides other than Am, Cm and Pu. Curium-45, Am-41 and Pu-41 can produce Np-37 and further separation of these elements from another FP elements is required. As a result, the following four groups are expected; I : a group including Sr, Eu, Cm, Am II : a group including Cs Np - Zr -1 Tc 0 1 75

III: a group including Np (Cm, Pu, Am), I, Pd, Se, Sn, Tc, Zr IV: a group including other nuclides. The comparison is made among another grouping concepts by Japan Atomic Energy Research Institute [6] (JAERI) and by Tokyo Institute of Technology [7] (TITEC). The two proposed groupings are different from the present grouping. The former grouping consists of two heat dominant elements Sr and Cs in a group but the latter grouping separates these elements. These differences may rise due to the individual object; JAERI grouping focuses on the partitioning, TITEC grouping on the reprocessing, and the present study on the disposal. The disagreement implies that further discussions from different angles including acceptability and feasibility are needed. Table 1 Comparison of Nuclide Groupings Present grouping JAERI grouping TITECH grouping I. Sr, Eu, Cm, Am Sr, Cs Te, Se, Ba, Y, Sr, Pd, Rh II. Cs Tc, Pd, Rh, Rh Np, Sn, Cs III. Np(Cm, Pu, Am), I, Pd, Se, Sn, Tc, Zr MA Cm, Am, RE IV. others Other FPs Nb, Sb, Pu, Tc Total Volume of Repository The volume of the HLW repository is simply governed by the interval of the emplacement between wastes. Dense disposal may cause high temperature field around the disposal pit thus the interval must be expanded due to the thermal stability of buffer material. Partitioning is considered to reduce the thermal load in the repository by separating the heat generation nuclides such as Sr-90 and Cs-137 from HLW. The effect of partitioning is examined to check the total volume of the repository for the partitioned groups. The JAERI grouping is considered as an example because the preliminary specifications of partitioned groups had been reported [6]. The temperature field was calculated by assuming the single heat source [5] then the method of superposition was used to taken into account the contribution from the neighboring wastes. The minimum interval value is set here as 3m because of the mechanical stability of the disposal pits. In the case of the Sr+Cs group, the minimum cooling time to keep the 3m interval is 70 years. The necessary disposal pitch of each group is calculated as shown in Table. The total disposal area required for three groups is approximately 60% of the current HLW disposal design [4]. It should be noted that the minor actinides are to be recycled in appropriate reactors and not considered here. When the recovered Sr + Cs group is temporary used for heat source for about 00 years, dense disposal may be possible and the necessary disposal area will be reduced to1/40 for the Sr+Cs group. In this calculation, the issues for the long-term temporary use are not considered; (1) the radiation effect to the waste form, () the thermal alteration of the waste form itself due to high temperature field. These issues have to be solved before the temporary use is expected. 76

Table Redaction of Disposal Area by Introducing the Partitioning Proposed by JAERI Sr + Cs group Tc+ PGM group Other FPs group HLW Waste Volume [L/MTU] 14 0.39 31 10 Inventory* [MTU/waste] 1.69 5.87 3.9 1.5 Overpack none yes yes yes Bentonite [cm] 30 70 70 70 Depth [m] 00 00 Pitch [m] 3 3 3 3.5 Cooling time[y] 70 0 5 50 Area [m/mtu] 16.7 4.8 7.1 48.11 Total [m/mtu] 8.74 48.11 Reduction [%] 60 0 Waste Form Ion Exchange Mineral Metal? (.3kW) Glass (5wt%) Glass (.3kW) * The inventory was determined to choose the smaller value either the initial heat generation of.3kw/waste or 5wt% as same as the glass solidified HLW. Radiological Impacts Radiological impacts are calculated by using either Eq.(1) or () depending on the element solubility. The peak concentration divided by the annual limit of intake (ALI) is termed as the peak hazard impact factor and the sum of the impact factors of each separated group is summarized in Fig.. In this figure the overview of the previous JAERI and TITEC groupings in Table 1 are shown and the results implies both JAERI and TITEC groupings are not fully attractive for geologic disposal because of the insufficient separation of Cs-135, Zr-93 in groups and 3. 1 Cs-135 Cs-135 0 Cs-135 Se-79 Zr-93 TITEC JAERI -1 Zr-93 Tc-99 Se-79 Pd-7 Zr-93 Tc-99 Zr-93 Tc-99 Se-79 Pd-7 - Fig. 1 3 HLW Group The total peak hazard impact factors of JAERI and TITEC groupings 77

Elimination of Isotopes Transmutation is designed to reduce the implicit hazard defined as M[Bq]/ ALI[Bq/y]. This term however dose not represents the impact of the disposal and contradiction is found because of the different object of the impacts. For example, Am-41 would have high implicit hazard values but have not significant migration impact because of its low solubility. Thus, the practical goal to reduce the migration impact must be assigned from the viewpoint of not only implicit hazard but also geologic disposal. Elimination rate The elimination of initial nuclide inventory of the insoluble nuclide dose not drastically change the nuclide concentration in groundwater as seen in Eq. (). If the inventory is sufficiently less than that to attain the steady-state condition, Eq. () is not adequate and Eq.(1) must be used instead. Thus the required rate of isotopes elimination can be simply assigned from the critical concentration as defined below. The two time periods are introduced for convenience: time of leach of nuclide T L and the time for steady-state T S. The time of leach is simply calculated by Eq. (4) by inserting the approximate expression of the integral in the numerator. T ( ) + L λt λt M e dt L M 1 e 0 1 TL = = (4) CSU CSUλ CSUλ o d o d o d M λ The time for steady-state is approximated by inserting the limit of the complementary error function (approx. 3) in the analytical solution for the solubility-limited boundary condition (for example, Eq. (7-133) in Ref. [9] ) ; D D L T 3 + 9 + µ Steady (5) µ U p D where U d : Darcy Velocity (= U p ε), and µ = +4λ. R R If T L >>T S, the elimination is not attractive because the maximum concentration does not depend on the inventory. On the contrary, the elimination may be effective to reduce the migration impact if T L <<T S. The critical inventory M C to use Eq. (1) for insoluble nuclide is thus determined by assuming T L =T S. λtsteady Mc = CoSUdTSteady T e λ 1 Steady (6) The required elimination rates of several nuclides are shown in Table 3 as an example. These values are the minimum requirement to reduce the peak concentration in underground water. Effect of Decay Chain 78

The large initial inventory of both Cm-45 and Am-41 produce Np-37 and the resulting radiological impact of Np-37 may not be negligible. Approximate solutions for decay chain are derived to take into account the nuclide production along with the migration. The maximum concentration of daughter nuclide can be derived by inserting Eq.() to the dispersion-convection equation and the maximum concentration is derived by the Laplace transformation C x C ( ) λ R 1 1 exp( Bx )+ { exp( Bx 1 ) exp( Bx )}. (7) C0 C0 λr λ1r1 The maximum concentration of the grand-daughter nuclide is also derived by inserting Eq.(7) to the convection equation for simplicity [8] then the solution is obtained by the Laplace transform; C3( x) C0 Cγ A exp( Ax 3 )+ exp( Ax 3 ) exp 5 1+ 0. 4Ax 1 C30 C30 B A3 Cγ A + exp( Ax 3 ) exp 5( 1+ 04. Ax 1 1) C B A { [ ]} 30 1 3 ( ) { [ ( )]} λ R 5 1+ 0. 4Ax 1 n n λ1r1 where An =, Bn =, γ =. U / ε x λ R λ R d 1 1 By using Eqs.(7) and (8), the effect of decay chain can simply be estimated. For example, Am-41 may leach for a few hundred thousand years according to the result in Table 3. The continuous release of Am-41 for infinite time may increases the maximum Np-37 concentration by a factor of approximately 0 in the case shown in Table 3. This may also gives an alternative requirement of elimination rate for transmutation. Table 3 Required Elimination Rate Determined by the Critical Inventory Cm-45 Am-41 Np-37 Tc-99 Initial Inventory* [Bq] 1.0E+ 4.19E+13 1.76E+ 6.36E+11 Sorption Coefficient** K d [m 3 /kg] 5.0 5.0 1.0.4E-03 Time of Leach T L [y].45e+04 1.4E+03 6.17E+06 6.15E+05 Solubility [mol/m 3 ] 1.0E-04 1.0E-04 6.5E-06 3.3E-05 Time for Steady-state** T S [y].53e+06 5.04E+05 4.8E+07 3.11E+05 Critical Inventory M C [Bq/MTU] --- --- --- 3.7E+04 Elimination rate --- --- --- 5.14E-08 * Total inventory of each nuclide corresponds to the 300 MTU PWR Spent fuel; burn-up of 45000MWd/MTU with initial enrichment 4.5%. **Groundwater flow condition; U p =1[m/y], ε=0.01, L=0[m] (8) 79

Conclusion This paper indicates the simple and practical goal of the partitioning and transmutation from the viewpoint of geologic disposal. A simple procedure to define the element groups for partitioning from both thermal and migration impacts is formulated. Elimination rate of the transmutation to reduce the maximum radiological impact was determined by comparing two time periods; the time for the steady state and the time of leach of insoluble nuclide. The critical concentration on which the elimination becomes attractive for geologic disposal was simply formulated and this factor suggests the minimum requirement of the transmutation rate. The comparison of the present study and the existent investigations of partitioning indicated there still remains contradiction among them and further discussions is necessary from both technical feasibility and acceptability points of view. References 1. OECD/NEA, Actinide and Fission Product Partitioning and Transmutation -Status and Assessment Report, OECD/NEA, (1999). IAEA, Safety and Environmental Aspects of Partitioning and Transmutation of Actinides and Fission Products:, IAEA, Jan,(1995).. H. Yamana, M.Shiotsuki, Recovery Yield of Nuclides Required for Satisfactory Actinide Recycling, Radioactive Waste Research, (1/), 35(1996). 3. H. Yamana, M.Shiotsuki, Recovery Yield of Nuclides Required for Satisfactory Actinide Recycling, Radioactive Waste Research, (1/), 35(1996). 4. Japan Nuclear Fuel Cycle Development Institute, Project Overview Report, H1 Project to Establish Technical Basis for HLW Disposal in Japan, JNC TN1400 99-0, May 1999. 5. I.R. Mufti, "Geothermal Aspects of Radioactive Waste Disposal into the Subsurface", J.Geophys.Res., 76(35), 8568(1971). 6. Y.Morita and M.Kubota, Wet Partitioning and Waste Treatment, Radioactive Waste Research, (1/), 75(1996). (in Japanese) 7. Y.Asano, et al., Study on a Nuclear Fuel Reprocessing System Based on the Precipitation Method in Mild Aqueous Solutions, Nucl. Technol., 10,198(1997). 8. V.C.Rogers, Migration of Radionuclide Chains in Groundwater, Nucl.Technol., 40,315(1978). 9. J.Bear, Hydraulics of Groundwater, McGraw-Hill, p. 68, (1979). 80