a physics study on thorium fuel recycling in a candu reactor using dry ...

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recycle the thorium fuel through the dry process option in the CANDU reactor, which in turn significantly improves natural uranium savings and diminishes spent ...
A PHYSICS STUDY ON THORIUM FUEL RECYCLING IN A CANDU REACTOR USING DRY PROCESS TECHNOLOGY

FUEL CYCLE AND MANAGEMENT KEYWORDS: thorium recycle, dry process, CANDU reactor

HANGBOK CHOI* and CHANG JE PARK Korea Atomic Energy Research Institute, P.O. Box 105, Yusong, Daejon, 305-600, Korea

Received February 21, 2005 Accepted for Publication March 31, 2005

Dry process fuel technology has high proliferation resistance, which is one of the important goals of the Generation-IV nuclear energy system developments. It is expected that dry process fuel technology can be applied not only to existing but also to future nuclear systems. In this study, the homogeneous ThO2-UO2 fuel cycle and the heterogeneous ThO2-DUPIC fuel cycle options of a Canada deuterium uranium (CANDU) reactor were assessed, which included a neutronic feasibility analysis of recycling spent fuels. Parametric calculations were also performed for reactivity coefficients and isotopic content changes for various initial fuel conditions. The results of the physics calculations have shown that it is feasible to recycle the thorium fuel through the dry process option in the CANDU reactor, which in turn significantly improves natural uranium savings and diminishes spent fuel. However, further investigation of the dry process option, which is technically and economically feasible for thoriumabundant dioxide fuel, is required.

I. INTRODUCTION Thorium fuel has been studied as an alternative to conventional nuclear fuels in the pressurized water reactor ~PWR! as well as the Canada deuterium uranium ~CANDU! reactor to expand energy resources and to provide a greater degree of energy self-reliance. The thorium fuel cycle is also considered for Generation-IV nuclear energy systems owing to its proliferation resistance, which is one of the goals for the Generation-IV nuclear energy systems.1 Thorium fuel produces fewer minor ac*E-mail: [email protected] 132

tinides than does uranium because of its lower atomic number. It also produces much less plutonium in comparison to uranium fuel and consequently is more proliferation resistant than slightly enriched uranium fuel. In addition, the presence of 232 U in the spent thorium fuel enhances the proliferation resistance because 232 U makes 233 U less attractive for diversion because of its strong alpha-particle emissions and the gamma ray associated with the 232 U decay chain. Since the 1970s, Atomic Energy of Canada Limited ~AECL! has studied many aspects of the thorium fuel cycle for the CANDU reactor, including fuel cycle analysis, reactor physics, fuel fabrication, irradiation, and waste management.2– 4 Both the once-through and recycling fuel cycles were investigated through various fuel management simulations. From these studies, AECL concluded that the use of thorium fuel in CANDU reactors ensures the long-term supply of nuclear fuel when a proven and reliable reactor technology is used. In this study, we extend the previous research on a thorium-based fuel cycle to a multiple recycling fuel cycle through dry process technology.5,6 The dry processes considered in this study are the “dry reprocess” developed for transmutation of the actinides in oxide fuel 7–9 or the “thermomechanical process” developed for the direct use of spent PWR fuel in the CANDU reactor ~DUPIC! fuel cycle.10,11 This study considers two fuel types: a homogeneous fuel bundle with a thorium and uranium mixture and a heterogeneous fuel bundle with thorium and DUPIC fuels. For the homogeneous fuel model, parametric calculations were performed for the thorium-to-uranium volume fraction and the fission product removal rate of the dry process to assess the neutronic feasibility of the recycling thorium fuel cycle. For the heterogeneous fuel model, parametric calculations were also performed for the fission product removal rate. For both fuel cycle models, the fuel cycle costs ~FCCs! were estimated to compare the economic advantage of the multiple recycling thorium fuel cycle. NUCLEAR TECHNOLOGY

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Fig. 1. CANDU fuel bundle models.

II. FUEL CYCLE ANALYSIS MODEL The reactor system considered in this study is a 713MW~electric! CANDU ~CANDU-6! reactor, which has a very high thermal flux level ~1.9 ⫻ 10 14 n0cm 2 {s! due to heavy water moderation. The CANDU-6 reactor was originally designed to use natural uranium as the fuel and pressurized heavy water as the coolant. This saves the high initial capital expense of uranium enrichment and fuel-reprocessing plants, although heavy water production plants are required. However, an important feature of the CANDU reactor concept is that it can evolve to use different coolants and fuels, resulting in an improvement of the fuel cycle. For example, the use of thorium fuel can substantially reduce the uranium requirements. There are 380 fuel channels in the CANDU-6 reactor, and each channel contains 12 fuel bundles in a horizontal channel. The standard CANDU fuel bundle has 37 fuel elements as shown in Fig. 1. A 43-element CANDU flexible ~CANFLEX! fuel bundle was also developed, and its use was begun in CANDU plants owing to its high thermal-hydraulic performance.12 From the physics viewpoint, however, these two fuel bundles are not very different from each other. II.A. Fuel Cycle Model The homogeneous thorium-uranium ~ThO2-UO2 ! fuel was designed for a closed fuel cycle as shown in Fig. 2. In this fuel cycle model, the thorium and uranium are

Th FUEL RECYCLING IN CANDU REACTORS

homogeneously mixed and burned in the reactor. The fission products are assumed to be removed from the spent fuel through the dry process. Then, the spent fuel is mixed with 20 wt% slightly enriched uranium ~SEU! for the next fuel cycle. In this way, it is possible to keep most of the actinides in the reactor system throughout the plant’s lifetime. It is therefore expected that the total amount of high-level waste is appreciably reduced when compared to the conventional once-through fuel cycle, and the amount of higher actinides is considerably reduced too. The heterogeneous thorium-DUPIC fuel cycle was designed to transmute the PWR spent fuel in the CANDU reactor. The fuel bundle has both thorium and PWR spentfuel elements in a fuel bundle cluster. The thorium fuel is located in the inner region of the fuel bundle and continuously recycled. The PWR spent fuel ~or DUPIC fuel! is located in the outer region of the fuel bundle and replaced after each fuel cycle. Therefore, a partially closed fuel cycle is constructed for the thorium-DUPIC fuel as shown in Fig. 3. In the homogeneous and heterogeneous thorium fuel cycles, the spent fuel is recycled through the dry reprocess or the thermo-mechanical process. The dry reprocess developed by Russian scientists is an electrorefining technology that recycles uranium and plutonium oxide fuel by utilizing the molten chloride media. The thermomechanical process simply relies on the oxidation and reduction of the oxide fuel 13,14 ~OREOX!, and therefore, all the actinides and most of the fission products reappear in the recycled fuel. In order to apply the OREOX process to the thorium-based fuel, this study proposes to mix the thorium fuel with a certain amount of UO 2 so that the mixture is voloxidized owing to the oxidation of the UO 2 . Because the process does not include any aqueous material and a separation step, this process is inherently the most proliferation resistant. II.B. Fuel Discharge Burnup Model In the CANDU reactor analysis, it is possible to estimate the fuel discharge burnup using only the lattice code once the reactor characteristics ~reactor size, effective fuel length, reactivity device, etc.! are appropriately described in the lattice calculation. Because the CANDU reactor is refueled daily, the fuel burnup is distributed from the fresh to the discharged state over the entire core.

Fig. 2. The closed thorium fuel cycle ~homogeneous recycle!. NUCLEAR TECHNOLOGY

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Fig. 3. The partially closed thorium fuel cycle ~heterogeneous recycle!.

The way to estimate the discharge burnup is to solve an integral equation of the excess reactivity for wdis ~Ref. 15!:



Wdis

@k eff ~w! ⫺ 1# dw ⫽ 0 ,

~1!

0

where wdis is the discharge burnup. Even if the fuel is different from natural uranium, the discharge burnup can also be estimated by the integral equation ~1!. However, the effect of the neutron spectrum change on the neutron leakage and absorption should be considered. Therefore, a simple formulation was derived to calculate the effective multiplication factor that can be estimated from the lattice calculation as follows: k eff ⫽

Production rate

.

Loss rate

~2!

Here, the loss term can be divided into three components: neutron absorption by the reactivity devices and structural materials, neutron absorption by the fuel cell, and neutron leakage. For convenience, the k eff can be written as follows: 1 k eff



Ad k` Ac



L k` Ac



1 k`

,

~3!

where A d , A c , and L are the parasitic absorptions in the reactivity devices and structural material, absorptions in the lattice cell, and the leakage term, respectively. The k ` and A c can be calculated from the lattice calculation. The leakage term can also be estimated using the buckling approximation. Because the parasitic absorptions occur mostly in the moderator region where the thermal flux is dominant, the fast absorptions can be neglected in A d . Therefore, Eq. ~3! can be rewritten as follows in a twogroup form: 1 k eff

冉 冊

⫽ S da2



Vd Vc

w2d k ` ~S ca1 w1c



III. HOMOGENEOUS RECYCLING OF THE THORIUM/URANIUM FUEL One of the most important features of the closed fuel cycle analysis is to maintain the material balance during the recycling. The fuel discharge burnup and isotopic number density can be obtained using the simplified model described in Sec. II. In this study, the WIMS-AECL transport code 16 was used for the lattice calculation and material balance analysis. WIMS-AECL uses an 89-group cross-section library that is based on ENDF0B-V nuclear data. The thorium series isotopes such as 232 Th, 233 Pa, and 233 U have four temperature data and the resonance tables for the fission and capture reactions. For the physics analysis of the closed fuel cycle, the analysis model and assumptions were made as follows: 1. The CANDU-6 reactor was used as the reference core. The reactivity loss due to the reactivity device was estimated based on the natural uranium CANDU-6 reactor analysis results. 2. The 43-element fuel bundle design was chosen.

c S a2 w2c !

~D1 B 2 w1c ⫹ D2 B 2 w2c ! k ` ~S ca1 w1c ⫹ S ca2 w2c !



1 k`

3. The fuel material is a mixture of ThO 2 and UO 2 . 4. The enrichment of the uranium feed is 20 wt%. ,

~4!

where d and c stand for the reactivity device and lattice cell, respectively. 134

The thermal flux of the device ~w2d ! can be replaced by the edge flux of the lattice cell. The effective thermal absorption cross section of the device ~S de ⫽ S da2 Vd 0Vc ! can be estimated from the reference core calculation. When the discharge burnup of the natural uranium fuel is 7300 MWd0t, the estimated effective thermal absorption cross section of the device is 1.096 ⫻ 10⫺3 cm⫺1. If the fuel is different from the natural uranium, the lattice properties can be fully incorporated into Eq. ~4! as well as the effective device cross section obtained from the reference core, and the discharge burnup of that specific fuel can be estimated without a need for a full-core fuel management simulation.

5. By the dry reprocess, all the actinides are recycled, while the fission products are removed. The fuel mass is kept constant by feeding thorium and uranium fuel. NUCLEAR TECHNOLOGY

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TABLE I

III.A. Parametric Calculation of the Material Flow Based on the analysis model and assumptions described earlier, a series of parametric calculations was performed on the uranium fraction, 235 U enrichment of the fresh fuel, and the fission product removal rate of the recycled fuel, and the results were produced for the material balance of the recycled fuel. The cases selected for the parametric calculations are as follows:

Summary of Isotopic Mass Change for the ThO 2-UO 2 Fuel* 233 U

235 U

239 Pu

ThO 2-9%UO 2

0 1 2 3 4 5

0.0 0.128 0.181 0.200 0.207 0.209

0.335 0.147 0.085 0.065 0.059 0.059

0.0 0.007 0.008 0.008 0.009 0.009

0.139 0.145 0.148 0.147 0.150 0.152

0.124 0.096 0.094 0.098 0.092 0.093

ThO 2-10%UO 2

0 1 2 3 4 5

0.0 0.169 0.201 0.204 0.203 0.200

0.372 0.135 0.101 0.100 0.104 0.108

0.0 0.008 0.009 0.010 0.011 0.012

0.346 0.358 0.364 0.372 0.381 0.390

0.121 0.095 0.091 0.082 0.073 0.063

ThO 2-11%UO 2

0 1 2 3 4 5

0.0 0.183 0.201 0.198 0.193 0.187

0.409 0.156 0.140 0.146 0.153 0.157

0.0 0.008 0.010 0.012 0.014 0.015

0.560 0.579 0.595 0.613 0.623 0.616

0.078 0.050 0.031 0.013 0.0 0.0

case A1: sensitivity to the UO 2 volume fraction case A2: sensitivity to the initial

235 U

enrichment

case A3: sensitivity to the fission product removal rate. III.A.1. Effect of the UO 2 Volume Fraction The volume fractions of UO 2 considered in this study are 9, 10, and 11% with an initial 235 U enrichment of 20 wt%. The estimated discharge burnups are 14 000, 26 000, and 36 000 MWd0t for the ThO2-9%UO2 , ThO2-10%UO2 , and ThO 2-11%UO 2 fuels, respectively. The variations of the infinite multiplication factor ~k ` ! are shown in Fig. 4, and the isotopic mass changes are given in Table I. It can be seen that the neutronic property represented by k ` converges immediately after the first recycle. For the ThO 2-9%UO 2 case, the amount of uranium feed is 0.152 kg0bundle0recycle, which corresponds to 5.89 kg of the natural uranium when the tail enrichment is 0.2 wt%. For the ThO 2-10%UO 2 and ThO 2-11%UO 2 fuels, the equivalent natural uranium feeds are 15.4 and 23.9 kg, respectively. Considering that the fuel mass of the 43-element fuel bundle is 18.6 kg, the natural uranium utilization defined as the energy produced per natural uranium consumed can be estimated as given in Table II for the three cases.

Fig. 4. Infinite multiplication factors of the recycled ThO 2UO 2 fuel. NUCLEAR TECHNOLOGY

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Uranium Thorium Feed Feed

Recycle

*In units of kg0bundle.

The natural uranium utilization of the ThO 2-9%UO 2 , ThO 2-10%UO 2 , and ThO 2-11%UO 2 fuels are 44.2, 31.4, and 28.0 MWd0kg, respectively. Table II also shows the natural uranium utilization normalized to the value of the standard natural uranium for direct comparisons. The results indicate that the recycling of the thorium-uranium fuel in the CANDU-6 reactor is feasible as far as the mass balance is concerned, and the uranium consumption of the recycled ThO 2-UO 2 fuel is much less than that of the conventional natural uranium fuel. Specifically, the natural uranium utilization of the ThO 2-UO 2 fuel is 3.8 times higher than that of the standard natural uranium fuel. It was also found that the high-burnup fuel ~e.g., ThO 211%UO 2 ! consumes more uranium resources when compared to the low-burnup fuel ~e.g., ThO2-9%UO2 ! because it burns more 235 U rather than utilizing the 233 U bred from 232 Th ~see Table I for comparison of fissile content!. Figure 5 shows the fissile content change of the recycled ThO 2-UO 2 fuel as a function of irradiation. In the equilibrium thorium-uranium fuel cycle, the fissile isotope 235 U is continuously supplied in order to maintain the excess reactivity. In general, the fissile content of the fuel is proportional to the burnup reactivity swing, and therefore, the ThO 2-9%UO 2 fuel has the smallest change of the fissile content among the three fuel types. The plutonium contents increase as the fuel is recycled; however, the plutonium buildup at the end of an equilibrium cycle ~after 14 recycles! is ;0.013 kg0bundle for the ThO 2-9%UO 2 fuel, which is even smaller than that of the natural uranium fuel at the discharge burnup ~;0.045 kg0bundle!. 135

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TABLE II Comparison of Natural Uranium Utilization for the ThO 2-UO 2 Fuel Discharge Burnup ~MWd0t! Natural uranium ~43 elements!

7300

Uranium Feed ~kg0bundle! 18.6

Equivalent Natural Uranium Feed ~kg0bundle!

Natural Uranium Utilization ~MWd0kg!

Normalized Natural Uranium Utilization

18.6

7.3

1.0

Case A1 ~UO 2 fraction! 9% 10% 11%

14 000 26 000 36 000

0.152 0.398 0.616

5.89 15.4 23.9

44.2 31.4 28.0

6.1 4.3 3.8

Case A2 ~ 235 U enrichment! 15 wt% 10 wt% 5 wt%

14 000 14 000 14 000

0.165 0.183 0.219

6.39 7.09 8.49

40.8 36.7 30.7

5.6 5.0 4.2

Case A3 ~fission product removal! 100% 90% 80%

14 000 14 000 14 000

0.152 0.181 0.195

5.89 7.01 7.56

44.2 37.1 34.4

6.1 5.1 4.7

Fig. 5. Fissile content of the recycled ThO 2-UO 2 fuel.

The neutron capture of 233 Pa is important in the thorium fuel reactor because it intercepts the production of 233 U from 232 Th neutron capture. In principle, the fluxdependent production of 233 U competes with the burning of the 235 U and eventually reduces the burnup reactivity swing to a certain extent. Figure 6 shows that the 233 Pa concentration reaches an asymptotic value as the fuel is irradiated, which indicates that the 233 U concentration of the recycled ThO 2-UO 2 fuel remains constant. III.A.2. Effect of the Initial 235 U Enrichment of the Fresh Fuel The parametric calculation was also performed for the initial 235 U enrichment of the thorium-uranium fuel 136

Fig. 6. The

233

Pa content of the ThO 2-9%UO 2 fuel.

~case A2!. The purpose of this calculation is to see how much of the uranium volume can be tolerated in the fuel mixture without deteriorating the recycling capability. It is known that thorium dioxide is chemically stable, which makes the dry process unfavorable. Therefore, the sensitivity calculations were performed to estimate the allowable amount of UO 2 in the thorium-based fuel from the reactor physics viewpoint. The initial 235 U enrichments considered are 5, 10, and 15 wt%; the estimated discharge burnup is fixed to 14 000 MWd0t, which is the discharge burnup of the ThO 2-9%UO 2 fuel. In order to obtain the target discharge burnup, the initial UO 2 volume fraction was searched by a trial-and-error method. The final values of the volume NUCLEAR TECHNOLOGY

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fraction are 32.5, 18, and 12% for the initial enrichments of 5, 10, and 15 wt%, respectively. The mass balance and natural uranium utilization of these fuels are also summarized in Table II and compared to the results of the natural uranium and ThO2-9%UO2 fuels. The results show that the recycling of the thorium-uranium fuel with a higher uranium fraction is feasible when considering the natural uranium utilization. For the cases considered here, it was possible to obtain a normalized natural uranium utilization greater than 4.2. It is worth noting that the natural uranium utilization is higher when the uranium fraction is smaller. In other words, the natural uranium utilization increases if the thorium fraction increases even though the initial uranium enrichment is high. III.A.3. Effect of Fission Product Removal The fission product content is an important factor when recycling the oxide fuels through the dry process because in principle the thermomechanical process leaves many of the neutron-absorbing fission products in the fuel, which in turn has negative effects on both the neutronic and material performance of the recycled fuel. In this study, the effect of the fission product removal rate was assessed for the recycled thorium-uranium fuel. For convenience, ThO 2-9%UO 2 fuel was chosen for the parametric calculations. The fission product removal rates used in this study were 100, 90, and 80%. The parametric calculations were performed under the condition that the discharge burnup of the recycled fuel is 14 000 MWd0t, which is the discharge burnup of the ThO 2-9%UO 2 fuel. If the fission product removal is ,80%, the amount of fissile uranium feed that makes up for the removed fission products is not sufficient to reach the burnup target. The mass balance and natural uranium utilization of these fuels are summarized in Table II and compared to the results of the natural uranium fuel. The results show that the recycling of the thorium-uranium fuel with a certain amount of fission products is feasible as far as the mass balance is concerned. The natural uranium utilization is 4.7 times greater than that of the natural uranium fuel if the fission product content is kept below 20%. III.B. Neutronic Characteristics of the Recycled Fuel There are many physics parameters that represent the safety characteristics of the CANDU reactor such as the void reactivity and temperature coefficient. These physics parameters can be readily generated by the perturbation option of the WIMS-AECL code, and the neutronic characteristics of the thorium-uranium fuel can be assessed. III.B.1. Fuels with Different UO 2 Volume Fractions The safety-related parameters were calculated for the three cases ~ThO 2-9%UO 2 , ThO 2-10%UO 2 , and ThO 2-11%UO 2 !, and the results are summarized in NUCLEAR TECHNOLOGY

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Table III. The results were obtained at the midburnup of one fuel cycle after a sufficient number of recycles ~defined as the equilibrium burnup of the recycled fuel in this study!. The equilibrium burnup of the recycled fuel is ;200 000 MWd0t. The safety-related parameters of the ThO 2-UO 2 fuel are also compared with those of the natural uranium fuel at the equilibrium burnup. For the natural uranium fuel, the equilibrium burnup is equivalent to the midburnup of the once-through cycle. The fuel temperature coefficient ~FTC! of the natural uranium fuel is near zero at the equilibrium burnup, while the FTC of the thorium-uranium fuel is slightly negative. It is understood that the 239 Pu fission reaction is mitigated to a certain extent in the thorium-uranium fuel owing to the preferential fission reaction by 233 U, which can reduce the FTC even though the 238 U content is low. Most of the fission reactions are from 233 U for the thorium-uranium fuel ~72%!, while the fission reaction by 239 Pu is 44% for the natural uranium fuel. As the fuel temperature increases, the ~renormalized! fission reactions by the fissile uranium isotopes tend to decrease while those by the fissile plutonium isotopes tend to increase. As a result, the FTC of the thorium-uranium fuel is smaller than that of the natural uranium fuel. However, the FTCs of both the natural uranium and recycled thorium-uranium fuels are very small. Figure 7 shows the FTC as a function of the accumulated fuel burnup, which is maintained negative for the recycled thorium-uranium fuel. The coolant temperature coefficients ~CTCs! are similar for both the natural uranium and thorium-uranium TABLE III Comparison of Safety-Related Parameters for the ThO 2-UO 2 Fuel at the Equilibrium Burnup Void FTC CTC MTC Reactivity ~mk 0K! ~mk 0K! ~mk 0K! ~mk! Natural uranium

;0.0

0.057

0.035

15.0

Case A1 ~UO 2 fraction! 9% 10% 11%

⫺0.010 ⫺0.010 ⫺0.010

0.046 0.048 0.049

0.006 0.004 0.004

15.3 15.8 16.2

Case A2 ~ 235 U enrichment! 15 wt% ⫺0.010 10 wt% ⫺0.010 5 wt% ⫺0.009

0.046 0.049 0.052

0.007 0.009 0.011

15.1 15.6 16.3

Case A3 ~fission products removal! 100% ⫺0.010 90% ⫺0.010 80% ⫺0.010

0.046 0.046 0.046

0.006 0.005 0.003

15.3 15.3 15.4

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Fig. 7. Variation of the FTC.

fuels. The coolant temperature increase has a positive effect on the reactivity because of the reduced heavy water density that decreases the upscattering of the thermal neutrons and enables more penetration of the thermal neutrons into the fuel bundle cluster. This phenomenon slightly increases both the fission and capture reaction rates of the uranium and plutonium in the inner fuel elements of the fuel bundle, which results in a positive CTC due to the excess reactivity of the inner fuel elements. Figure 8 shows the CTC, which is positive at the beginning and increases slightly with the recycling. The moderator temperature coefficient ~MTC! is appreciably reduced for the thorium-uranium fuel. The thermal spectrum of the thorium-uranium fuel is much less than that of the natural uranium fuel. Compared to the natural uranium fuel, the peak cell-average thermal flux is reduced by 26% for the thorium-uranium fuel at ;0.1 eV. If the moderator temperature increases, the neutron spectrum of the thermal energy region increases more for the

Fig. 8. Variation of the CTC. 138

natural uranium fuel compared to the thorium-uranium fuel at the equilibrium state, which results in a larger MTC value for the natural uranium fuel. Figure 9 shows the MTC, which is negative at the beginning but becomes slightly positive with the recycling. The void reactivity is one of the most important physics parameters of the CANDU fuel because it is positive and determines the magnitude of the power pulse upon a loss-of-coolant accident. When the coolant channel is voided, more thermal neutrons travel deep into the fuel bundle center as was seen for the CTC. Then, the magnitude of the void reactivity is determined by the excess reactivity of the inner fuel elements of the fuel bundle. For the cases considered here, it turned out that the void reactivity is similar for both the natural uranium and thorium-uranium fuels. Figure 10 shows the void reactivity as a function of the burnup. The void reactivity at the midburnup of each recycle is almost constant at ;15 mk. The isotopic contents of the recycled fuel are summarized in Table IV for the equilibrium cycle. As was discussed earlier, the natural uranium fuel builds up 239 Pu as the fuel is irradiated, while the thorium-uranium fuel builds up 233 U. As the thorium-uranium fuel is continuously recycled, the 233 U concentration reaches an asymptotic value, which results in steady behavior of the physics parameters of the thorium-uranium fuel. The amount of 232 U is negligible even though the radioactive decay of 232 U is very important in the recycled fuel. The radiation effect of other isotopes ~e.g., 234 U and 236 U! as well as fission products is estimated in Sec. III.C using a full decay chain of the ORIGEN-2 code.17 III.B.2. Fuels with Different Initial

235 U

Enrichment

The safety-related parameters of the fuels with different initial 235 U enrichments ~15, 10, and 5 wt%! are given in Table III too. Note that the 235 U enrichment of

Fig. 9. Variation of the MTC. NUCLEAR TECHNOLOGY

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20 wt%. The only difference in these three fuels is the amount of initial uranium loading. The calculations showed that the multiple recycling of the thoriumuranium fuel can be sustained with sufficient fuel burnup until the initial 235 U enrichment is reduced to ;5 wt%. If the fuel has low-enrichment uranium ~e.g., ThO 2 32.5%UO 2 !, the fuel will build up more 239 Pu when compared to the higher-enrichment fuel ~e.g., ThO 2-9%UO 2 !, and the contribution of 233 U will decrease. For the cases considered in this study, however, it turned out that the effects of the initial uranium enrichment on the physics parameters are not that great. As the initial uranium enrichment is reduced, the FTC, CTC, MTC, and the void reactivity tend to slightly increase. Fig. 10. Variation of the coolant void reactivity.

the uranium feed for the recycled fuel is still 20 wt% though the initial 235 U enrichment changes. The results show similar behaviors when compared with the ThO 29%UO 2 fuel, which has an initial 235 U enrichment of

TABLE IV Comparison of Isotopic Content of the ThO 2-UO 2 Fuel with Different Uranium Fractions at the Equilibrium Cycle* Nuclide

If the fission products are recycled together with the actinides through the dry process, this will affect not only the mass balance but also the physics parameters of the recycled fuel. Table III shows the safety-related parameters of the ThO 2-9%UO 2 fuels with different fission product removal rates. The results, however, showed that the safety-related parameters are not very dependent on the amount of fission products in the recycled fuel. In fact, these parameters are mostly dependent on the major fissile and fertile isotopes, not on the fission products. For the case considered in this study ~the maximum value of 20% fission product content!, the poisoning effect ~thermal neutron absorption! was well managed by adjusting the amount of the 20 wt% enriched uranium feed. The results also showed that the physics parameters of the recycled fuel were not linearly dependent on the amount of fission products in the fuel as far as the comparisons were made at the equilibrium burnup.

Natural Uranium

ThO 29%UO 2

ThO 210%UO 2

ThO 211%UO 2

0.0 0.0

13.717 13.528

13.078 12.936

12.624 12.205

0.0 0.0

0.201 0.199

0.193 0.189

0.187 0.182

0.0 0.0

0.097 0.098

0.093 0.095

0.091 0.093

0.133 0.044

0.073 0.040

0.116 0.037

0.157 0.034

0.0 0.013

0.097 0.102

0.112 0.123

0.121 0.137

TABLE V

18.468 18.337

2.622 2.591

3.226 3.162

3.653 3.559

Radioactivity of the ThO 2-UO 2 Fuel for the Equilibrium Cycle*

0.0 0.045

0.013 0.013

0.014 0.015

0.015 0.017

0.0 0.004

0.004 0.004

0.004 0.004

0.004 0.005

232 Th

Charge Discharge 233 U Charge Discharge 234 U Charge Discharge 235 U Charge Discharge 236 U Charge Discharge 238 U Charge Discharge 239 Pu Charge Discharge 241 Pu Charge Discharge

III.B.3. Fuels with Different Fission Product Removal Rates

The radiation level of the thorium-uranium fuel was estimated by the ORIGEN2 code and compared to that of the natural uranium fuel. The calculation was performed for the equilibrium cycle of the ThO 2-9%UO 2 fuel, and the results are given in Table V. The result shows that the

Burnup State

Natural Uranium

ThO 2-9%UO 2

Charge Discharge After cooling for 5 yr

0.0 2.79 ⫻ 10 6 2.63 ⫻ 10 3

3.94 ⫻ 10 1 3.17 ⫻ 10 6 4.46 ⫻ 10 3

*In units of Ci0bundle.

*In units of kg0bundle. NUCLEAR TECHNOLOGY

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Fi ~t ! ⫽ fuel cycle component ~i ! cost at time t

TABLE VI Input Values for the Fuel Cycle Cost Analysis Loss Lead0 Rate Lag Unit ~%! ~months! Cost a

Component

Uranium ~$0lb U3O 8 ! Thorium ~$0lb ThO 2 ! Conversion ~$0kg HM! 0.5 Enrichment ~SWU! Fabrication ~$0kg HM! 0.5 Refabrication ~$0kg HM! 0.5 Transportation0storage ~$0kg HM! Disposal ~$0kg HM! a b

⫺17 ⫺17 ⫺13 ⫺12 ⫺10 ⫺10 60 360

25.0 20.0 b 8.0 100.0 65.0 616.0 13.0 73.0

Based on the price as of 1994 OECD0NEA report. 80% of uranium purchase cost considering natural resource.

radiation level of the thorium-uranium fuel is higher than that of the natural uranium fuel by 14% when discharged from the core. The radioactivity of the discharged fuel is mostly determined by short-lived actinides such as 233 Th ~T102 ⫽ 22.4 min! and 233 Pa ~T102 ⫽ 27 days! for the thorium-uranium and 239 U ~T102 ⫽ 23.5 min! and 239 Np ~T102 ⫽ 2.35 days! for the natural uranium fuel, respectively. If the spent fuel is cooled for 5 yr, most of the radioactivity is from fission products such as 137 Cs, 90 Sr, and 90 Y for both the thorium-uranium and natural uranium fuels. III.D. Fuel Cycle Cost The FCC of the thorium-uranium fuel was estimated by utilizing the unit cost data developed for the DUPIC fuel cycle analysis.18 Table VI summarizes the input data for the FCC calculation such as the loss rate, the lead0lag time, and the unit cost of each fuel cycle component. The FCC was estimated by the levelized lifetime cost model provided by the Organization for Economic Cooperation and Development0Nuclear Energy Agency 19 ~OECD0 NEA!, in which the levelized FCC is calculated as follows: t⫽t0⫹L⫹T2

FCC ⫽

(i t⫽t(⫺T 0

1

t⫽t0⫹L

(

t⫽t0

Fi ~t ! ~1 ⫹ r! t⫺t0

,

~5!

E ~t ! ~1 ⫹ r! t⫺t0

where t ⫽ particular time in the fuel cycle span t0 ⫽ base year of the money count L ⫽ reactor lifetime T1 ⫽ maximum value of the lead time in the frontend fuel cycle T2 ⫽ maximum value of the lag time in the backend fuel cycle 140

E ~t ! ⫽ electric power generated from the nuclear fuel cycle at time t r ⫽ discount rate. For the standard CANDU-6 reactor, the once-through FCC is 2.79 mills0kW{h, and the fuel purchase cost is the most expensive component ~46% of the FCC!. The FCC of the homogeneous thorium-uranium fuel was calculated for different conditions as given in Table VII. As the uranium fraction increases, the fuel burnup increases, and the FCC decreases accordingly. The FCC can be reduced to 2.79 mills0kW{h if the uranium fraction is 11% and the corresponding fuel burnup is 36 000 MWd0t. For the effect of the initial 235 U enrichment on the FCC, the amount of SEU required for continuous recycling increases if the initial 235 U enrichment decreases. However, the enrichment cost is greatly saved if the initial 235 U enrichment decreases. As a result, the FCC is saved by 27% for the case of 5 wt% initial 235 U enrichment, when compared to the case of 20 wt% initial 235 U enrichment. For the fission product removal rates considered in this study, the effect on the FCC is negligible. It should be noted that the fuel burnup of cases A2 and A3 was fixed at 14 000 MWd0t. It can be therefore seen that the FCC is mostly determined by the fuel burnup, which can be found from the fuel fabrication cost. Regarding the fuel burnup ~case A1!, the FCC is appreciably reduced ~;2.77 mills0kW{h! when the UO 2 fraction increased from 9 to 10%, which corresponds to the fuel burnup change from 14 000 to 26 000 MWd0t. From the viewpoint of the FCC, therefore, it is required to appreciably increase the fuel burnup or introduce an inexpensive refabrication process in order to compete with the existing natural uranium CANDU fuel cycle. However, emphasis should also be given to the fuel cycle concept that produces no high-level waste.

IV. HETEROGENEOUS RECYCLING OF THE THORIUM/DUPIC FUEL For the heterogeneous fuel bundle option, two kinds of fuel elements are considered: thorium and DUPIC fuels. The outer two rings of the fuel bundle are loaded with DUPIC fuel, while the inner two rings are loaded with thorium fuel for multiple recycling. Because the thorium fuel has no fissile isotope, a driver fuel is required to maintain a chain reaction in the fuel bundle, which is achieved by the DUPIC fuel with a fixed fuel composition.20 In this fuel cycle model, the standard 37element fuel bundle was used in order to balance the amount of the thorium and the DUPIC fuels. It should be noted, however, that the optimum fuel bundle geometry shall be studied in the future for the thorium-based fuel. NUCLEAR TECHNOLOGY

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TABLE VII Levelized Costs of the ThO 2-UO 2 Fuel Cycle*

Uranium

Thorium

Conversion

Enrichment

0.139

Fabrication

Transport0 Storage0 Disposal

Total

1.114

0.387

2.792

Natural uranium

1.152

Case A1 ~UO 2 fraction! 9% 10% 11%

0.824 0.569 0.326

0.133 0.087 0.049

0.121 0.082 0.047

1.317 0.910 0.521

4.236 2.226 1.843

0.016 0.009 0.005

6.648 3.883 2.791

Case A2 ~ 235 U enrichment! 15 wt% 10 wt% 5 wt%

0.618 0.410 0.190

0.133 0.133 0.134

0.096 0.071 0.044

1.806 1.217 0.608

4.236 4.236 4.236

0.016 0.016 0.016

6.058 5.470 4.861

Case A3 ~fission product removal! 100% 90% 80%

0.824 0.809 0.786

0.133 0.133 0.134

0.121 0.119 0.116

1.317 1.294 1.257

4.236 4.236 4.236

0.016 0.016 0.016

6.648 6.609 6.545

*In units of mills0kW{h.

IV.A. Parametric Calculation of the Material Flow For the thorium-DUPIC fuel cycle, a fixed fuel composition is used for the DUPIC fuel, while the thorium fuel is continuously recycled through the dry process. Because the DUPIC fuel provides a uniform neutronic property, the recycling characteristics of the thoriumDUPIC fuel are determined by a thermomechanical process that can remove some of the fission products from the spent thorium fuel. Therefore, the process-related variable to be studied is the removal rate of the fission products. Because the excess reactivity of the fuel bundle is mostly provided by the DUPIC fuel, it is not necessary to remove the bulk of the fission products. Therefore, the parametric calculation was performed for the removal of the rare earth elements ~Nd, Ce, La, Pr, Pm, Sm, Eu, Gd, and Dy!. As was done for the homogeneous fuel model, the thorium fuel is mixed with a small amount of uranium to facilitate the thermomechanical process. So, the parametric calculation was also performed for the uranium fraction in the thorium fuel. In this case, natural uranium was used to avoid the additional enrichment cost. The parametric calculations can be categorized as follows:

earth elements from the recycled thorium fuel. The multiplication factors converge immediately after the first cycle, which is similar to the behavior of the homogeneous thorium-uranium fuel. If the rare earths are not removed, the fuel cycle burnup at the equilibrium state is ;19 100 MWd0t. If the rare earth removal rate increases to 10, 20, and 30%, the fuel cycle burnup also increases to 19 500, 19 700, and 19 900 MWd0t, respectively. The transmutation of the higher actinides in the thorium-DUPIC fuel is summarized in Table VIII. For the minor actinides, the isotopic mass of 237 Np and 243Am slightly increases for each recycle. But, 241Am ~halflife ⫽ approximately 433 yr!, which decays to 237 Np, is significantly reduced when compared to other actinides.

case B1: sensitivity to the rare earth removal from the recycled thorium fuel case B2: sensitivity to the initial uranium fraction in the thorium fuel with a rare earth removal rate of 30%. IV.A.1. Effect of Rare Earth Removal Variations of the infinite multiplication factors are shown in Fig. 11 for various removal rates of the rare NUCLEAR TECHNOLOGY

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Fig. 11. Infinite multiplication factors of the thorium-DUPIC fuel with different rare earth removal rates. 141

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TABLE VIII Comparison of the Minor Actinide Content of the Thorium-DUPIC Fuel at the Equilibrium Cycle* 237

Case B1 ~rare earth removal from the thorium fuel! 10% Charge Discharge 20% Charge Discharge 30% Charge Discharge Case B2 ~initial U fraction in the thorium fuel! 10% Charge Discharge 20% Charge Discharge 30% Charge Discharge

Np

241

Am

243

Am

Minor Actinides

Fissile Plutonium

Fissile Total

5.34 5.41

9.69 1.54

1.13 3.15

16.22 11.44

76.0 59.1

307.9 151.1

5.39 5.45

9.69 1.50

1.13 3.21

16.22 11.45

76.0 59.0

317.5 155.0

5.39 5.46

9.69 1.47

1.13 3.24

16.22 11.47

76.0 59.0

311.5 150.7

5.45 5.52

9.71 1.54

1.41 3.51

16.59 11.87

77.6 60.6

311.4 153.1

5.52 5.59

9.73 1.61

1.67 3.77

16.94 12.27

79.2 62.1

305.3 147.0

5.58 5.65

9.75 1.68

1.92 4.01

17.28 12.67

80.8 63.6

298.9 143.4

*In units of g0bundle.

As a result, the total minor actinide inventory is reduced by ;30%. For the residual fissile isotopes in the thoriumDUPIC fuel, the fissile plutonium inventory is reduced by ;20%. However, it should be noted that the transmutation of minor actinides and fissile plutonium occurs mostly in the DUPIC fuel, which is made from PWR spent fuels. Therefore, the thorium-DUPIC fuel concept can provide the potential for the fissile self-sustaining recycling of the thorium fuel as well as the transmutation of higher actinides in the existing spent fuels.

and the rare earth removal rate is maximized from the viewpoint of transmuting transuranics. IV.B. Neutronic Characteristics of the Recycled Fuel The safety-related physics parameters of the thoriumDUPIC fuel were also evaluated at the equilibrium burnup. The parametric calculations were performed for the rare

IV.A.2. Effect of the Initial Uranium Fraction In order to improve the voloxidation capability of the recycled thorium fuel, the natural uranium was initially mixed with the thorium fuel with volume fractions of 10, 20, and 30% for the case of a rare earth removal rate of 30%. Figure 12 shows the infinite multiplication factors for the three uranium volume fractions. The isotopic mass change for the 30% removal rate of rare earth is also given in Table VIII. For the cases considered in this study, the fuel cycle burnup at the equilibrium state was ;19 100 MWd0t, which was not dependent on the initial uranium fraction. However, the amount of minor actinides and residual fissile in the DUPIC fuel elements was appreciably reduced. Therefore, it is recommended that the initial uranium fraction be kept as low as possible without losing the voloxidation capability of the thorium fuel, 142

Fig. 12. Infinite multiplication factors of the thorium-DUPIC fuel with different initial uranium fractions. NUCLEAR TECHNOLOGY

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TABLE IX Safety-Related Parameters of the Thorium-DUPIC Fuel at the Equilibrium Burnup FTC ~mk 0K!

CTC ~mk 0K!

MTC ~mk 0K!

Void Reactivity ~mk!

Natural uranium ~37 elements!

;0.0

0.068

0.035

14.2

Case B1 ~rare earth removal from the thorium fuel! 10% 20% 30%

⫺0.004 ⫺0.004 ⫺0.004

0.063 0.064 0.064

0.022 0.024 0.024

15.6 15.8 15.9

Case B2 ~initial U fraction in the thorium fuel! 10% 20% 30%

⫺0.003 ⫺0.003 ⫺0.003

0.064 0.064 0.064

0.024 0.023 0.023

15.8 15.7 15.5

earth removal rate ~case B1! and the initial uranium fraction of the thorium fuel ~case B2!. IV.B.1. Safety Parameters for Various Rare Earth Removal Rates Table IX shows the key safety-related physics parameters of the thorium-DUPIC fuel for various rare earth removal rates at the equilibrium burnup. It can be seen that the safety-related parameters represented by the temperature coefficients are close to those of the natural uranium fuel at the equilibrium burnup. Compared to the homogeneous thorium-uranium fuel, however, the temperature coefficients of the thorium-DUPIC fuel are a little higher, which is due to the high excess reactivity provided by the DUPIC fuel located at the outer fuel rings. The inbred 233 U in the thorium fuel also contributes positively to the void reactivity, but the void reactivity of the thorium-DUPIC fuel is close to that of the thorium-uranium fuel. IV.B.2. Safety Parameters for Various Initial Uranium Fractions Tables IX also shows the safety-related parameters at the equilibrium burnup for various initial uranium fractions. It is observed that the safety-related physics parameters are not sensitive to the initial uranium loading. Similar to the case of the sensitivity to the rare earth removal rate, the temperature coefficients of the thoriumDUPIC fuel with a different initial uranium fraction are slightly lower than those of the natural uranium fuel but are higher than those of the homogeneous thoriumuranium fuel. IV.C. Radioactivity of the Thorium/DUPIC Fuel The radiation level of the thorium-DUPIC fuel for the equilibrium cycle is given in Table X with the result of the DUPIC fuel for comparison. For the thoriumNUCLEAR TECHNOLOGY

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DUPIC fuel bundle, 19% of the fuel material is thorium, and the rest of it is PWR spent fuel. The thorium fuel is initially mixed with natural uranium by 30%. It is also assumed that 30% of rare earths is removed from the irradiated thorium fuel after cooling for 5 yr. Note that the thorium-uranium fuel used in Sec. III.C contains 20 wt% SEU by 9%. The results show that the radiation level of the thorium-DUPIC fuel is similar to that of the DUPIC fuel when discharged from the core. For the charged ~fresh! fuel of the equilibrium cycle, however, the radiation level of the thorium-DUPIC fuel is higher than that of the DUPIC fuel by 54% because it contains recycled thorium fuel of which only 30% of rare earths is removed. At the discharge state, the radiation level of the thorium-DUPIC fuel is determined by 239 U and 239 Np followed by 233 Th and 233 Pa because most of fuel material is uranium based. If the spent fuel is cooled for 5 yr, the radiation level is dominated by fission products such as 137 Cs, 90 Sr, and 90 Y. IV.D. Fuel Cycle Cost To calculate the thorium-DUPIC FCC, we assumed that the PWR spent fuel was taken as “free.” Table XI TABLE X Radioactivity of the Thorium-DUPIC Fuel for the Equilibrium Cycle*

Burnup State

DUPIC Fuel

Thorium-DUPIC Fuel ~Initial U fraction ⫽ 30%!

Charge Discharge After cooling for 5 yr

2.82 ⫻ 10 3 2.67 ⫻ 10 6 7.82 ⫻ 10 3

4.35 ⫻ 10 3 2.86 ⫻ 10 6 9.26 ⫻ 10 3

*In units of Ci0bundle. 143

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TABLE XI Levelized Costs of the Thorium-DUPIC Fuel Cycle*

Uranium Case B1 ~rare earth removal from the thorium fuel! 10% 20% 30% Case B2 ~initial U fraction in the thorium fuel! 10% 20% 30%

0.005 0.009 0.014

Thorium

Conversion

Fabrication

Transport0 Storage0 Disposal

0.038 0.037 0.037

0.006 0.006 0.006

5.843 5.820 5.814

0.182 0.181 0.181

6.069 6.044 6.039

0.034 0.030 0.026

0.006 0.006 0.006

5.683 5.510 5.362

0.178 0.172 0.168

5.906 5.727 5.576

Total

*In units of mills0kW{h.

shows the FCC with various rare earth removal rates and initial uranium fractions. The FCC decreases as the rare earth removal rate increases because the burnup increases slightly, but the difference is negligible. As the initial uranium fraction increases in the thorium fuel, the FCC increases slightly because of the additional processes of the uranium fuel. It can also be seen that the cost of the heterogeneous thorium-DUPIC fuel cycle is mostly determined by the fuel burnup, which is ;19 100 MWd0t.

V. SUMMARY, CONCLUSION, AND FUTURE WORK In this study, a series of parametric calculations were performed to obtain the physics characteristics of the thorium-based fuel cycle. Specifically for the homogeneous thorium-uranium fuel, this study investigated the effects of the uranium fraction, initial enrichment of the uranium, and the fission product removal. The results of the mass balance calculation showed that it is feasible to construct a closed fuel cycle by utilizing the thoriumuranium fuel in the CANDU-6 reactor. The physics analyses also showed that the safety-related parameters of the thorium-uranium fuel do not deteriorate when compared to those of the natural uranium fuel. The heterogeneous thorium-DUPIC fuels were also assessed for the partial recycling of the thorium fuel through the thermomechanical dry process. The simulation showed that the recycling fuel cycle can be established by the DUPIC fuel without sacrificing safety-related reactor physics parameters. The preliminary results showed the technical feasibility of the homogeneous thorium-uranium fuel as well as the heterogeneous thorium-DUPIC fuel bundle for the CANDU reactor when considering the mass balance of the recycling fuel cycle and the safety-related reactor physics parameters. The recycling of the spent fuel al144

lows a longer fuel cycle period and a high fuel burnup when compared to the conventional once-through fuel cycle, which is an incentive to the fuel cycle economics. The thorium-based fuel bundle concept in conjunction with the dry process technology can also provide a safeguardable way of transmuting both the residual fissile of the PWR spent fuel and the inbred fissile of the thorium fuel. In conclusion, it is feasible to recycle the thoriumbased fuel continuously in the CANDU reactor as far as the mass balance and neutronic characteristics are concerned if the initial 235 U enrichment and the fission products removal rate are well controlled in the feasible ranges. Though this study has investigated the key parameters of the recycling thorium fuel cycle, there are more issues that should be considered in the future as follows: 1. The fuel element power distribution needs to be examined, which is related to the fuel performance and thermal-hydraulic characteristics of the fuel bundle. 2. The application of the thermomechanical process to the treatment of the thorium-based fuel should be examined for both the powder preparation and fission product removal. 3. It will be necessary to estimate the channel and bundle power distribution of the CANDU core loaded with the thorium-based fuel. 4. The thorium fuel can also be recycled in a mixed channel form: The thorium fuel is loaded in one channel, and other fuel ~DUPIC, SEU, or mixed oxide fuel! can be loaded in another channel. 5. It is expected that the cost of the CANDU recycling thorium fuel cycle is higher than that of the conventional natural uranium CANDU fuel cycle unless the fuel burnup is more than doubled because of the high fuel fabrication cost. NUCLEAR TECHNOLOGY

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6. However, closed and partial recycling thoriumbased fuel cycles are feasible without sacrificing the major physics parameters of the CANDU reactor. It is also believed that the recycling thorium-based fuel cycle can appreciably reduce the high-level waste that should be otherwise geologically disposed of.

9. A. V. BYCHKOV, O. V. SKIBA, A. A. MAYORSHIN, V. A. KISLY, S. K. VAVILOV, M. V. KORMILITZYN, L. S. DEMIDOVA, L. G. BABIKOV, and R. A. KUZNETSOV, “Fuel Cycle of Actinide Burner-Reactor. Review of Investigations by ^^DOVITA&& Program,” Proc. Int. Conf. Future Nuclear Systems, Challenge Towards Second Nuclear Era with Advanced Fuel Cycles, Yokohama, Japan, October 5–10, 1997, p. 657, American Nuclear Society ~1997!.

ACKNOWLEDGMENT

10. J. S. LEE, K. C. SONG, M. S. YANG, K. S. CHUN, B. W. RHEE, J. S. HONG, H. S. PARK, and H. KEIL, “Research and Development Program of KAERI for DUPIC ~Direct Use of Spent PWR Fuel in CANDU Reactors!,” Proc. Int. Conf. Future Nuclear Systems: Emerging Fuel Cycles and Waste Disposal Options (GLOBAL’93), Seattle, Washington, September 12–17, 1993, p. 733, American Nuclear Society ~1993!.

This work has been carried out under the Nuclear Research and Development program of the Korea Ministry of Science and Technology.

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