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Conference Paper TRANSMUTATION OF AMERICIUM AND IN A LANTHANIDE MATRIX COMPANY WIDE CW-123700-CONF-018 Revision 0

Prepared by Rédigé par

Hyland Bronwyn

Reviewed by Vérifié par

Edwards Geoffrey W R

Approved by Approuvé par

Hyland Bronwyn

2012/07/04 2012/07/04 UNRESTRICTED ILLIMITÉ

Atomic Energy of Énergie Atomique du Canada Limited Canada Limitée

Chalk River, Chalk River (Ontario) Canada K0J 1J0 Canada K0J 1J0 Proceedings of Global 2011 Nagoya, Japan, December 11-15, 2011 UNRESTRICTED CW-123700-CONF-018

Transmutation of Americium and Curium in a Lanthanide Matrix

B. HYLAND1, E.D COLLINS2, R. J. ELLIS2, G. DEL CUL2 and M. MAGILL1 1Chalk River Laboratories, Atomic Energy of Canada Limited, Canada, K0J 1J0 Tel: +1(613)584-9243 ex. 44707 , Fax: +1(613)584-8198 , Email: [email protected] 2Oak Ridge National Laboratory (ORNL), Oak Ridge, Tennessee, U.S.A.

Abstract – With world stockpiles of used nuclear increasing, the long-term remediation of used fuel is a growing concern. Many of the transuranic (TRU) in nuclear spent fuel produce decay heat for long durations, resulting in significant nuclear waste management challenges. Partitioning and transmutation of the spent fuel is one possibility to deal with the long-lived transuranic actinides. It is desirable to develop a partitioning and transmutation scheme that is as straightforward as possible for both the reprocessing step and the transmutation in reactor. This study looks at a reprocessing scheme that is simplified to keep the americium, curium and lanthanides together in one recycle stream. These actinides can be transmuted to fissionable and then to shorter-lived fission products in a CANDU reactor. Many of the design features of the CANDU reactor make it uniquely adaptable to transmutation. The small, simple fuel bundle eases the fabrication and handling of active . allows precise management of core reactivity and the high economy of the CANDU reactor results in high TRU destruction to fissile-loading ratio. The transmutation scheme described in this paper places the americium/curium/lanthanide mixture, diluted with an inert matrix, into the centre pin of a CANFLEX fuel bundle. This centre pin can be removed and reinserted into a new fuel bundle for further irradiation to achieve higher total transmutation of the actinides. This scheme provides the dual benefit of transmuting long-lived actinides into shorter-lived fission products, while at the same time reducing the void reactivity coefficient of the reactor.

I. INTRODUCTION americium transmutation schemes to study fuel that has decayed for a significant length of time after exit. This Management of spent will continue to be study uses 10-year-cooled fuel, so that there is a a challenge in the future, as stockpiles increase worldwide. significant build-up of Am-241 to be transmuted. Many of the transuranic (TRU) nuclides are long-lived, Studies conducted previously [1-7] have and continue to produce heat for thousands of years. This examined scenarios in which americium has been long-term decay heat presents a challenge to the disposal separated from spent fuel, either by itself, with curium, or of (SNF). Reducing the decay heat of with all of the TRU nuclides. This study looks at a spent nuclear fuel from LWRs will increase the capacity of simplified scheme in which the americium, curium and long-term geological disposal sites. Am-241, with a half lanthanides are all kept together in one stream. life of 432 years, is a major contributor to the long-term The CANDU reactor offers attractive solutions for decay heat. Its primary contribution is between 100 and effectively dealing with used nuclear fuel from a light 1000 years after the fuel exits the reactor. Am-241 is reactor (LWR) fleet. Many of the design features of produced in the reactor during irradiation, but most of the the CANDU reactor make it uniquely adaptable to actinide Am-241 in SNF results from the decay of Pu-241 (half life transmutation as well as utilization of LWR used fuel with of 14 years) after the fuel has been discharged from the minimal reprocessing. The most significant feature is the reactor. Consequently, it is beneficial when investigating high resulting from the

CANDU is a registered trademark of Atomic Energy of Canada Limited (AECL).

Proceedings of Global 2011 Nagoya, Japan, December 11-15, 2011 UNRESTRICTED CW-123700-CONF-018

moderator, which allows a high TRU transmutation rate available for transmutation rather than being parasitically relative to the fissile loading because more are absorbed in the moderator. Online refuelling also allows

Cm-245 Heavier Cm-242 Cm-244 (n, γ) (n, γ) 162.8 d 18.10 y 8500y mass curium    (n, f ) isotopes

- 83% Am-242m Am-244m Am-241 141 y 26 m (n, γ) IT Am-243 - 432.2 y (n, γ) (n, γ) 7370 y  (n, f ) Am-242 (n, f )  Am-244 16.02 h 10.1 h - -, EC 

EC (17%) Pu-238 Pu-239 Pu-240 Pu-241 Pu-242 Pu-243 (n, γ) (n, γ) (n, γ) (n, γ) (n, γ) 87.7 y 2.4E5 y 6563 y 14.4 y 3.7E5 y 5 h   (n, f )  - (n, f )  -

Figure 1. The transmutation pathways of Am-241. precise management of core reactivity, and further decay to Cm-244. Cm-244 has a relatively short half-life, increases the neutron economy relative to batch and alpha decays to Pu-240. Am-242m can also decay by refuelling. Lastly, the small and simple fuel bundle electron capture to Pu-242. The isotopes Cm-242, Cm- simplifies the fabrication and handling of active fuels. 244 and Pu-238 all have an impact on the decay heat of CANDU fuel bundles are short in length (49.52 cm) and the spent fuel. light in weight (~21 kg), consisting of either 37 pins or Am-242m, Cm-245, Pu-239, and Pu-241 are the 43 pins (CANFLEX® fuel) which simplifies the fissile isotopes. The other isotopes act as a poison, fabrication and handling of the bundles. These capturing neutrons and reducing the void characteristics also enable a CANDU fuel bundle to reactivity (CVR) of the bundle (when located in the function as a target carrier with minimal or no design central element). change to the bundle. With regard to curium production, for schemes with relatively short irradiation time (a few years) the curium 1.1 Transmutation Pathways isotopes that are created are the lower mass, short-lived curium isotopes, Cm-242 and Cm-244. These isotopes The transmutation of Am-241 follows several pathways have half-lives on the same time scale as fission products, that affect the decay heat production of the spent fuel, and once produced in the used fuel the curium could be and result in the production of and stored and decayed in a manner similar to fission . In the first step, a neutron captures onto products, and not put into long-term storage or further Am-241, creating Am-242 or Am-242m. transmuted. Several different pathways are available after the initial . Am-242m has a high fission 1. REPROCESSING/RECYCLING SCHEME cross-section, so by this path the Am can be transmuted by fission. In the second pathway the Am-242 beta The proposed reprocessing/recycling scheme for TRU decays into Cm-242. The Cm-242 then alpha decays actinides is illustrated in Figure 2. The separations steps with a relatively short half-life (163 days), and some of would be simplified significantly by not requiring the the original americium will end up as Pu-238. The complex steps needed to separate the chemically similar, Am-242m can also neutron capture to Am-243, and a trivalent TRU actinides, americium and curium, from the second neutron capture creates Am-244 or Am-244m. trivalent lanthanide fission products. The overall scheme The Am-244 nuclides both have short half-lives and beta fits the definition of “modified open cycle” since ~10% to 20% of the americium-curium would eventually be CANFLEX® is a registered trademark of AECL and the Korea Atomic Energy Research Institute (KAERI).

Proceedings of Global 2011 Nagoya, Japan, December 11-15, 2011 UNRESTRICTED CW-123700-CONF-018

disposed, along with all of the fission product elements, to the high level waste emplaced in a geologic repository.

Separations U-Pu-Np MOX Fuel SNF LWRs Process “A” Product Blending Fabrication

U-Pu-Np Am-Cm FPs

“Burnable Poison” Separations Separations Am-Cm-Ln FPs Target Rod HWRs Process “B” Process “C” Fabrication

All FPs + Other FPs HLW Solidification and Disposal Residual Am-Cm

Figure 2. Reprocessing/Recycling Scheme Separations Process A (Figure 2) would likely be a contamination in this model of zirconia. The solvent extraction process using tri-normal butyl purpose of the zirconia is to dilute the Am/Cm/Ln phosphate as the extractant to recover the recycle fuel mixture, in order to investigate the effect of varying the components, , plutonium, and neptunium. amount in the centre pin. The amount of Am/Cm/Ln in Several options for Separations Process B will be the centre pin was varied between 5% and 60% by considered. One option is to use the TRUEX process, volume. Other methods to vary the quantity of followed by a cation resin loading/ calcination process, Am/Cm/Ln could be employed, such as reducing the although it is possible that the TRUEX process could be density of the centre pin, or dilution using recycled omitted. In the latter version, the monovalent and uranium. divalent fission products, including the highly radioactive Recycled uranium (RU) with an enrichment of 1.0% U- cesium and strontium, would remain with the liquid 235 comprised the three other rings of fuel. The whole waste. The trivalent americium-curium-lanthanide reactor was loaded with this fuel, and operated at fission products, plus tetravalent fission standard operating powers. The isotopic composition of product, would be converted to mixed oxide the PWR SNF corresponds to a 10 year cooling time, and microspheres, suitable for recycle target fabrication and was calculated using the TRITON, ORIGEN-S and re-irradiation as “burnable poison rods”. ORIGEN-ARP modules of SCALE 5.1 [9, 10]. The LWR had an initial enrichment of 4 wt% U-235 and an After irradiation, the spent “burnable poison rods” would exit of 50 MWd/kg initial heavy elements (IHE) be stored for an appropriate time and then processed by [11]. Only nuclides with significant neutron cross- means of Separations Process C (Figure 2). Separations sections, are contained in the WIMS-AECL library that Process C would likely be a solvent extraction process was used for the study; other lanthanides present in the similar to Separations Process A, except operated on a spent fuel have been ignored in this work. The smaller scale for the purpose of recovering the Am/Cm/Ln composition used in the first cycle is shown plutonium, generated by transmutation of americium- in Table 1. 241. The wastes from Separations Process C would contain the residual (~10-20%) americium and curium, plus all fission products.

2. FUEL DESIGN

A standard CANFLEX bundle design was used for the model, see Figure 3 [8]. For the initial core, the centre pin was Am, Cm, La, Ce, Pr, Nd, Pm, Sm, Eu, Gd, Tb, Dy, Ho, and Er, mixed with zirconia (ZrO2). There is no

Proceedings of Global 2011 Nagoya, Japan, December 11-15, 2011 UNRESTRICTED CW-123700-CONF-018

containing fresh LEU in the remaining pins. For the initial study no decay time is used between irradiations For the 60% Am/Cm/Ln case the decay time between successive irradiations was varied. Decay times of 2, 5, and 30 years were modeled to examine the effect of decay time on the transmutation of the various transuranic nuclides. 3. RESULTS

All models for this report were based on lattice cell calculations performed using WIMS-AECL v. Figure 3. Fuel Design using a CANFLEX bundle 3.1.2.1[12] with the ENDF/B-VII-based library [13]. geometry with a centre pin comprised of Am, Cm and Ln The burnup calculations were performed using a constant power of 32 W/g IHE (initial heavy element) applied at each burnup step. Leakage and absorption by unmodelled reactor components were assumed to be worth 45 mk in total, which is typical of a CANDU 6 Table 1. Isotopic Composition of the Americium, Curium reactor with the adjustor rods inserted. and Lanthanides in the Centre Pin For this study the following conventions were used: the % by % by % by percent difference is, Nuclide weight Nuclide weight Nuclide weight Final  Initial Am-241 4.7294 Nd-146 6.7930 Gd-154 0.1710 %difference 100% Initial . Am-242m 0.0077 Nd-148 3.4660 Gd-155 0.0435 Thus, a negative number is destruction and a positive Am-243 1.6710 Nd-150 1.7199 Gd-156 1.0355 number is creation. Percent difference per cycle refers to Cm-243 0.0039 Pm-147 0.1012 Gd-157 0.0013 the difference from the start of a cycle to the end, Cm-244 0.5109 Sm-147 1.9771 Gd-158 0.2218 whereas cumulative percent difference refers to the Cm-245 0.0511 Sm-148 1.3336 Gd-160 0.0112 difference in composition from the fresh centre pin to the Cm-246 0.0065 Sm-149 0.0272 Tb-159 0.0257 end of a given cycle. For the mass destroyed per year, a positive number is destruction and a negative number is Cm-247 0.0001 Sm-150 3.0960 Dy-160 0.0037 creation. La-139 11.4135 Sm-151 0.1054 Dy-161 0.0035 Ce-140 12.0208 Sm-152 0.8458 Dy-162 0.0027 3.1 Burnup Ce-142 10.4744 Sm-154 0.3637 Dy-163 0.0020 Ce-144 0.0004 Eu-151 0.0086 Dy-164 0.0005 Table 2 gives the effect of Am, Cm, and Ln concentration Pr-141 10.4431 Eu-152 0.0002 Ho-165 0.0009 in the initial centre pin, and of the irradiation cycle of that pin, on the exit burnup of the fuel bundle. The Nd-142 0.2501 Eu-153 1.2015 Er-166 0.0003 associated graph is shown in Figure 4. This shows a Nd-143 6.8181 Eu-154 0.1176 decrease in burnup as the concentration of Am/Cm/Ln is Nd-144 12.8596 Eu-155 0.0127 increased. This is expected, as an increase of Am/Cm/Ln Nd-145 6.0455 Gd-152 0.0005 is an increase in parasitic neutron absorption. There is an increase in burnup as the pin is recycled into a new To obtain greater transmutation of Am/Cm, the centre bundle; this is expected as with each recycle the centre pin cases were designed so that the centre pin is recycled pin will contain less of the Am/Cm/Ln neutron poison, into a fresh bundle. Demountable bundles have been in allowing the fuel to obtain a higher burnup. The changes use at the National Research Universal (NRU) research in exit burnup are relatively small for successive recycles reactor, located at the , for many of the centre pin. The 60% loading case, corresponding years. This demountable element fuel concept is to the largest change in burnup with recycle, shows an employed in this study. While this technology is well increase in burnup of 2.4 MWd/kg(IHE) between the first proven for the research reactor application, further to fourth cycles. development would be needed to implement this concept in power reactors. After the first irradiation, the centre Table 2. Exit burnup (MWd/kg) for varying centre pin pin would be removed and placed in a new bundle concentrations for each cycle, and the decrease in exit

Proceedings of Global 2011 Nagoya, Japan, December 11-15, 2011 UNRESTRICTED CW-123700-CONF-018

burnup relative to a reference bundle with RU in the concentration of Am/Cm/Ln in the centre pin adds centre pin neutron poison to the centre of the bundle, which Exit Burnup Decrease Relative decreases the CVR. Each successive irradiation will (MWd/kg) to all RU Bundle have a lower amount of neutron absorber in the centre (MWd/kg) pin, thus there will be less of a reduction of the coolant % Initial Irradiation Cycle Irradiation Cycle void reactivity. In this concept the reactor would contain Am/Cm/Ln 1 2 3 4 1 2 3 4 bundles with the centre pin at different irradiations (i.e., 5 16.1 16.6 16.6 16.7 1.1 0.6 0.6 0.5 some bundles would be undergoing the first cycle, and 10 15.1 15.6 16.1 16.1 2.1 1.6 1.1 1.1 others the second or third irradiation cycles), such that 15 14.7 15.1 15.6 16.1 2.5 2.1 1.6 1.1 this effect averages out over the whole reactor. The 20 13.7 14.7 15.1 15.6 3.5 2.5 2.1 1.6 changes in CVR are relatively small for successive 25 13.2 14.2 14.7 15.1 4.0 3.0 2.5 2.1 recycles of the centre pin. The 60% loading case, which 30 12.8 13.7 14.2 14.7 4.4 3.5 3.0 2.5 also corresponds to the largest change in burnup with 35 11.8 13.2 13.7 14.2 5.4 4.0 3.5 3.0 recycle, shows a decrease relative to a reference bundle 40 11.3 12.8 13.2 13.7 5.9 4.4 4.0 3.5 45 11.3 12.3 12.8 13.7 5.9 4.9 4.4 3.5 composed entirely of RU of 4.5 to 6.3 mk. This case also 50 10.8 11.8 12.3 13.2 6.4 5.4 4.9 4.0 has the largest change in burnup with recycle, with an 55 10.4 11.3 12.3 12.8 6.8 5.9 4.9 4.4 increase in CVR of 1.8 mk from the first to fourth cycles. 60 9.9 10.8 11.8 12.3 7.3 6.4 5.4 4.9 Table 3. The decrease in coolant void reactivity for varying centre pin concentrations for each cycle, relative to a reference bundle with RU in the centre pin Decrease in CVR Relative to all RU Bundle (mk) % Initial Irradiation Cycle Am/Cm/Ln 1 2 3 4 5 1.2 1.0 0.9 0.8 10 1.8 1.4 1.2 1.1 15 2.4 1.9 1.6 1.4 20 2.9 2.3 1.9 1.7 25 3.5 2.8 2.3 2.0 30 4.0 3.3 2.7 2.3 35 4.4 3.7 3.1 2.7 40 4.9 4.2 3.6 3.1 45 5.3 4.5 3.9 3.4 50 5.7 4.9 4.3 3.8 55 6.0 5.3 4.7 4.1 60 6.3 5.7 5.1 4.5

Figure 4 Exit burnup (MWd/kg) for varying centre pin concentrations for each cycle

3.2 Coolant Void Reactivity

Table 3 gives the effect of Am, Cm, and Ln concentration in the initial centre pin and the irradiation cycle of that pin on the coolant void reactivity of the fuel bundle, relative to a bundle with RU in the centre pin. The associated graph is shown in Figure 5. Increasing the

Proceedings of Global 2011 Nagoya, Japan, December 11-15, 2011 UNRESTRICTED CW-123700-CONF-018

reactors, as possible (choose a greater initial concentration). The percentage of the initial americium that is transmuted decreases as the initial amount increases. After four cycles 94% and 84% of the Am is transmuted if the centre pin has 5% and 60% Am/Cm/Ln initially. However, with a higher initial amount of Am/Cm/Ln, there is more mass in the reactor, and a higher mass of Am is transmuted per year. The four-cycle average mass of Am transmuted is 1.2 kg/year and 20.6 kg/year for 5% and 60% initial loadings, respectively.

Table 4. Percent difference by cycle and cumulative percent change of americium for varying centre pin concentrations for each cycle Initial % Percent change over Cumulative AmCmL the cycle Percentage Change n 1 2 3 4 1 2 3 4 5 -74.6 -51.8 -34.2 -27.3 -74.6 -87.4 -91.6 -93.9 10 -71.1 -52.6 -35.9 -27.7 -71.1 -85.9 -90.8 -93.3 15 -68.3 -52.7 -36.5 -28.4 -68.3 -84.5 -90.0 -92.7 20 -64.2 -53.1 -37.5 -28.7 -64.2 -82.7 -88.9 -92.0 Figure 5. Change in CVR relative to a reference bundle 25 -61.2 -52.4 -38.1 -28.9 -61.2 -80.9 -87.9 -91.3 with RU in the centre pin for varying centre pin 30 -58.1 -51.5 -38.6 -29.2 -58.1 -79.0 -86.9 -90.6 concentrations after each cycle 35 -53.5 -50.8 -39.5 -29.9 -53.5 -76.4 -85.4 -89.6 40 -50.4 -49.3 -39.6 -30.4 -50.4 -74.1 -84.0 -88.7

45 -48.8 -47.2 -39.1 -31.4 -48.8 -72.2 -82.7 -87.9 3.3 Transmutation of Americium 50 -45.7 -45.4 -38.8 -31.9 -45.7 -69.5 -80.9 -86.8

The percentage change over each cycle and the 55 -42.8 -43.4 -39.3 -31.9 -42.8 -66.8 -79.4 -85.7 cumulative percentage change in americium 60 -39.9 -41.3 -38.5 -32.2 -39.9 -63.8 -77.3 -84.3 concentration for varying initial Am, Cm, and Ln concentration in the centre pin for each cycle are given in 3.4 Transmutation of Curium Table 4. Table 5 gives the effect of Am, Cm, and Ln concentration in the centre pin and the irradiation cycle Table 6 gives the effect of Am, Cm, and Ln concentration of that pin on the rate of americium transmutation. The in the centre pin and the irradiation cycle of that pin on relationship between transmutation rate, percentage the average mass of Cm transmuted per year. The transmuted, and the cycle number is shown in Figure 6 associated graph is given in Figure 7. for a few sample initial concentrations. The total mass produced is relatively small. Cycles 2 As is typical with transmutation studies, there is a to 4 show net destruction of curium, however the small competition between the fraction of americium that can amounts transmuted in the subsequent irradiations do not be transmuted and the mass of americium (kg/year) that add up to the amount produced in the first irradiation, so can be transmuted. That is, if there is a small amount of that the net result at the end of all cycles is a net gain of Am loaded, then a larger fraction of the Am can be Cm. The four-cycle average mass production of Cm is transmuted than for a higher initial concentration. 0.20 kg/year and 3.95 kg/year for 5% and 60% initial However, a higher initial concentration will allow a loadings, respectively. greater rate of transmutation. The initial concentration selected for the fuel cycle would depend on the goal, Table 5. Rate of americium transmuted (kg/year) for whether it is desired to have as little Am as possible varying centre pin concentrations after each cycle remaining (choose a smaller initial concentration), or to Initial % Initial Mass Rate of Am Transmutation 4- Cycle transmute as much mass as possible as fast, or in as few AmCmL (kg/reactor) (kg/year) per cycle Averag n 1 2 3 4 e

Proceedings of Global 2011 Nagoya, Japan, December 11-15, 2011 UNRESTRICTED CW-123700-CONF-018

5 7.19 3.88 0.69 0.22 0.11 1.22 35 4.50 -10.45 1.19 0.92 0.06 -2.07 10 14.39 7.86 1.68 0.54 0.27 2.59 40 5.14 -12.02 1.03 1.14 0.15 -2.43 15 21.58 11.70 2.86 0.93 0.45 3.98 45 5.79 -13.13 0.91 1.30 0.23 -2.67 20 28.77 15.69 4.48 1.47 0.69 5.58 50 6.43 -14.62 0.53 1.49 0.38 -3.05 25 35.97 19.36 6.20 2.12 0.98 7.16 55 7.07 -16.12 0.02 1.60 0.53 -3.49 30 43.16 22.87 8.17 2.93 1.33 8.83 60 7.73 -17.61 -0.59 1.67 0.73 -3.95 35 50.35 26.59 10.77 4.07 1.83 10.82 40 57.55 29.82 13.22 5.32 2.42 12.69 45 64.74 32.45 15.28 6.57 3.04 14.34 50 71.93 35.31 17.96 8.25 3.91 16.36 55 79.13 38.02 20.73 10.02 4.84 18.40 60 86.30 40.53 23.58 12.14 6.09 20.59

Figure 7. Mass of curium (kg/year) transmuted per year for varying centre pin concentrations after each cycle

3.5 Plutonium Production Figure 6. Rate of americium transmuted and the cumulative percent transmuted for each cycle for 5%, The mass of plutonium that is produced in the centre pin 30%, and 60% initial concentration of Am/Cm/Ln. only is given in Table 7 and Figure 8. There is an increase in plutonium mass, as would be expected (see Table 6. Mass of curium (kg/year) transmuted per year discussion in Section 1.1). The increase occurs primarily for varying centre pin concentrations after each cycle in the first cycle. For the cases with low initial amounts Rate of Cm of Am/Cm/Ln (5 and 10%), there is a destruction of Pu 4- Initial % Initial Mass Transmutation Cycle in cycles 2-4. For the cases with 15 to 45% initial AmCmLn (kg/reactor) (kg/year) per cycle Avg. Am/Cm/Ln there is a destruction of Pu in cycles 3 and 4, 1 2 3 4 and for the 50 to 60% initial Am/Cm/Ln there is 5 0.64 -1.22 0.28 0.04 0.11 -0.20 destruction of Pu in cycle 4. Note that to be consistent 10 1.29 -2.59 0.60 0.09 0.02 -0.47 with the rest of the document a negative number 15 1.93 -3.97 0.88 0.18 0.01 -0.72 corresponds to an increase and a negative number is a 20 2.57 -5.59 1.12 0.31 -0.01 -1.04 decrease in mass. In the early cycles the plutonium is mainly Pu-238, produced from the alpha decay of Cm- 25 3.21 -7.09 1.28 0.47 -0.01 -1.34 242 and Pu-240, from the electron capture decay of Am- 30 3.86 -8.59 1.36 0.66 0.00 -1.64 242. With subsequent irradiation the amount of Pu-239,

Proceedings of Global 2011 Nagoya, Japan, December 11-15, 2011 UNRESTRICTED CW-123700-CONF-018

Pu-240, and Pu-241 increase, and some of the initial Am- there is little difference in Pu production in reactor for 241 is fissioned as Pu-239 and Pu-241. these three cases.

Table 7. Mass of plutonium transmuted (kg/year) for each cycle for the various initial loadings of Am/Cm/Ln. Rate of Pu Transmutation Initial % (kg/year) per cycle 4-Cycle AmCmLn Average 1 2 3 4 5 -1.8 0.2 0.6 0.4 -0.1 10 -3.6 0.1 1.3 0.9 -0.4 15 -5.5 -0.4 1.9 1.4 -0.6 20 -7.4 -1.4 2.3 2.1 -1.1 25 -9.1 -2.6 2.5 2.8 -1.6 30 -10.7 -4.1 2.5 3.5 -2.2 35 -12.3 -6.2 2.0 4.1 -3.1 40 -13.7 -8.2 1.2 4.4 -4.0 45 -14.9 -9.8 0.4 4.7 -4.9 50 -16.0 -11.9 -1.0 4.5 -6.1 55 -17.0 -14.1 -2.6 4.2 -7.4 60 -16.9 -16.2 -4.6 3.4 -8.6

4.6 Impact of Decay Time

If this fuel cycle option of transmuting americium with the curium and lanthanides in the centre pin of the Figure 8. Mass of Pu produced per year for each cycle for bundle, and recycling the poison pin into fresh bundles, various initial concentrations of Am/Cm/Ln. were to be implemented in a power reactor, there would be some time in which the Am/Cm/Ln is out of the There is fractionally more curium produced during the reactor between cycles. This decay period has not been irradiation for the longer decay times, as there is less taken into account in the calculations above. The effect curium present in the reactor at the start of the of a two year, a five year, and a thirty year decay period irradiation, since it has decayed prior to the irradiation. are examined in this section. These calculations have This phenomenon is illustrated in Figure 9 and Table 10, been performed for the 60 vol% loading case only. All of which show the amount of Cm present at each stage of the data that was shown in the previous sections is the fuel cycle. Long decay times between cycles allows tabulated, Table 8, Table 9, and Table 10, for the varying the curium to decay, and there is less increase in curium decay times, but graphs are shown only for the amounts overall through the whole fuel cycle. The decay time of of plutonium and curium present in each stage of the fuel 30 years allows for enough decay of Cm-244 between cycle, for which there is significant change in the results cycles so that there is a net reduction of heavy curium for the different decay times, Figure 9 and Figure 10. isotopes over the fuel cycle. A long decay time serves to The half lives of Cm-242 and Cm-244 are 163 days break the chain to producing higher mass curium and 18 years, respectively. For the case with no decay isotopes, by decaying Cm-244 instead of allowing it to time, all decay of Cm-242 happens during the capture a neutron to Cm-245. Instead it decays to Pu- irradiation. The decay of Cm-242 is the source of the 240, where it can capture a neutron to become Pu-241, differences in plutonium production between the no where it will either fission, or decay to Am-241, and decay time and any of the models with a decay time. begin the transmutation cycle over again. There is more plutonium being produced in reactor (Figure 10) for the no decay time case because some of the initial Cm-242 is still decaying in the reactor during Table 8. Exit burnup (MWd/kg) and the change in the second cycle. For the other 3 cases, the initial curium coolant void reactivity relative to a reference bundle with decays outside of the reactor before the second cycle, so the increase in Pu happens during the decay time, and so

Proceedings of Global 2011 Nagoya, Japan, December 11-15, 2011 UNRESTRICTED CW-123700-CONF-018

RU in the centre pin for varying decay times between cycles for the 60 vol% initial loading Decay Change in CVR Time Exit Burnup (MWd/kg) Relative to all RU between Bundle (mk) Cycles (years) 1 2 3 4 1 2 3 4 0 9.9 10.8 11.8 12.3 -6.3 -5.7 -5.1 -4.5 2 9.9 10.8 11.8 12.3 -6.3 -5.7 -5.1 -4.5 5 9.9 10.8 11.8 12.3 -6.3 -5.6 -5.1 -4.6 30 9.9 10.8 11.8 12.3 -6.3 -5.7 -5.1 -4.5

Table 9. Percent difference by cycle and cumulative percent change of americium for varying decay times between cycles for the 60 vol% initial loading Decay Cumulative Percent change of Am Time Percentage Change of over the cycle between Am Cycles (years) 1 2 3 4 1 2 3 4 0 -38.3 -41.3 -38.5 -32.2 -38.3 -63.8 -77.3 -84.3 2 -38.3 -39.6 -37.1 -31.1 -38.3 -62.9 -76.7 -83.9 5 -38.3 -39.6 -37.0 -31.2 -38.3 -63.0 -76.7 -83.8 30 -38.3 -39.5 -36.9 -32.4 -38.3 -63.9 -77.0 -83.5

Table 10. Rate of americium, curium, and plutonium Figure 9 Mass of curium (kg/reactor) present at various transmuted (kg/year) for varying decay times between stages in the fuel cycle for varying decay times. cycles for the 60 vol% initial loading Decay Rate of Am Transmutation 4- Cycle Time (kg/year) per cycle Average between Cycles (years) 1 2 3 4 0 38.95 23.58 12.14 6.09 20.19 2 38.95 22.55 11.71 5.96 19.79 5 38.95 22.50 11.65 6.00 19.77 30 38.95 21.87 11.49 6.47 19.70 Rate of Cm Transmutation (kg/year) per cycle 0 -17.29 -0.59 1.67 0.73 -3.87 2 -17.29 -9.69 -5.40 -3.12 -8.87 5 -17.29 -10.16 -5.77 -3.51 -9.18 30 -17.29 -10.32 -6.39 -4.48 -9.62 Rate of Pu Transmutation (kg/year) per cycle 0 -16.9 -16.2 -4.6 3.4 -8.6 2 -16.9 -5.7 3.3 7.2 -3.0 5 -16.9 -5.1 3.8 7.5 -2.7 30 -16.9 -4.5 3.4 6.0 -3.0

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cycle the bundle can achieve a higher output burnup, while the total transmuted Am, Cm, and Ln continues to increase, albeit at a less efficient rate. A long decay time serves to break the chain to producing higher mass curium isotopes. Through this design it is possible to have a fuel cycle that results in a net decrease in heavy mass curium (mass > 244) isotopes, and has a significant overall transmutation of minor actinides.

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