Proceedings of ICONE23 rd 23 International Conference on Nuclear Engineering May 17-21, 2015, Chiba, Japan ICONE23-1561 USE OF PA-231 FOR AXIAL POWER DISTRIBUTION FLATTENING OF THORIUM FUEL CANDLE HIGH TEMPERATURE GAS-COOLED REACTORS Peng Hong Liem Nippon Advanced Information Service (NAIS Co., Inc.) 416 Muramatsu, Tokaimura, Ibaraki, Japan Hoai Nam Tran Institute of Research and Development, Duy Tan University, K7/25 Quang Trung, Da Nang, Vietnam Hiroshi Sekimoto Tokyo City University 1-28-1 Tamazutsumi, Setagaya-ku, Tokyo, Japan Keywords: Pa-231, thorium fuel cycle, CANDLE, HTGR, axial power distribution flattening, small sized long life reactor with CANDLE active core height of 800 cm and reactor thermal power of 30 MWth. ABSTRACT The innovative CANDLE (Constant Axial shape of Neutron flux, nuclide densities and power shape During Life of Energy producing reactor) burnup strategy has been successfully applied to both fast and thermal reactors. In particular for thermal reactor applications, CANDLE blocktype high temperature gas-cooled reactors (HTGRs) with either uranium or thorium fuel cycle had been proposed and investigated for their simple and safe reactor operation, and the ease of designing a long life reactor. Small sized long life CANDLE HTGRs with thorium fuel shows superior burnup performance than the ones with uranium fuel but their axial power peaks are relatively higher which may not be advantageous during a depressurization accident. In this work, we proposed and investigated the use of Pa-231 mixed homogeneously in the (Th-232/U-233)O2 fuel kernel of the TRISO particles to obtain lower axial power peaks. Addition of Pa-231 decreases the required amount of natural gadolinium burnable poison in the fresh fuel for establishing a valid CANDLE HTGR design since Pa-231 has a large thermal absorption cross section. Besides the role as a burnable poison nuclide, Pa-231 also serves as a fertile nuclide during the CANDLE burning since Pa-231 is finally transmuted to a fissile U-233 nuclide. A promising analysis result shows that for U233 enrichment of 15 w/o and Pa-231 addition of 7.50 w/o, the axial power peak is decreased from 5.9 to 3.6 W/cm 3 while the averaged burnup is increased from 138 to 149 GWd/t. This extends the core life time about 16 %, i.e. from 35 to 41 years NOMENCLATURE kinf : infinite medium neutron multiplication factor keff : effective neutron multiplication factor r (cm) : radial direction in 2-D R-Z reactor geometry z (cm) : axial direction in 2-D R-Z reactor geometry 1. INTRODUCTION The innovative CANDLE (Constant Axial shape of Neutron flux, nuclide densities and power shape During Life of Energy producing reactor) burning scheme was proposed by Sekimoto et al. [1] originally aimed for fast reactors. Under this burning scheme, the burning region moves autonomously with a constant velocity along the core axis from bottom to top (or from top to bottom) as shown in Fig. 1. As shown in the figure, the core can be roughly divided into three regions: (1) fresh fuel region (kinf <1), (2) burning region (kinf >1) and (3) spent fuel region. When the burning scheme is applied to a block-type HTGR, burnable poison (for e.g. natural gadolinium) is used to adjust the kinf of the fresh fuel to be sub-critical. The CANDLE HTGR burn-up process is as follows (Fig. 2). Neutrons leaked from the burning region into the fresh fuel region will be absorbed by the burnable poison and the burning region will move slowly into the fresh fuel region with depleted burnable poison. In the burning region, depletion of fissile material for energy production is accompanied by conversion of fertile material into fissile material. The spent fuel region is the region left by the burning region which contains mainly fission products and depleted fuel. Copyright © 2014-2015 by JSME Start On load Finish Fresh Fuel Region Burning Region Spent Fuel Region Fig. 1 CANDLE burnup concept. BP Fissile Fuel BP Fissile Fuel Neutron Flux Neutron Flux Burning Direction FP Equilibrium Condition FP End of Life BP：Burnable Poison FP：Fission Products Fig. 2 CANDLE burnup application to HTGR. After the CANDLE HTGR is operated for a certain core life time, the reactor can be shut down for refueling. If the core active height is design properly the CANDLE concept may feature a long life HTGR design as will be shown later. For a unique combination of core geometry and fresh fuel composition, one can find an equilibrium critical condition where CANDLE burning scheme is realized. Under the equilibrium condition, the moving (axial) velocity of the burning region is constant. Analytical codes for obtaining either the equilibrium condition or for simulating the reactor start-up have been developed. The details of the computational procedures are not given here and readers should consult other references [1, 2, 3, 4]. In our previous works as well as in the present work we consider small sized long life prismatic/block-type HTGRs adopting CANDLE burning scheme which can take full advantages of CANDLE properties: 1. Constant reactor parameters (e.g. power peaking, reactivity coefficients etc.) during reactor operation. This will simplify not only the reactor design itself and its licensing process but also simplify its reactor operation and maintenance. 2. No requirement for burn-up reactivity control mechanism. Besides simplifying the reactor design, a severe control rod ejection accident during full power operation (under nominal pressure) can be avoided. 3. Proportionality of core height to reactor core life. A long life core can be easily designed by adjusting the core height. 4. Sub-criticality of fresh fuel. No criticality accident will occur during transportation and storage of fresh fuels. In addition, application of CANDLE burning scheme to small sized long life HTGRs can be realized by the present coated fuel particle and HTGR reactor technologies. As for the fuel cycle, CANDLE HTGRs can be applied for uranium fuel cycle [2], thorium fuel cycle [4] as well as for effective incineration of weapon grade plutonium [5]. In our previous work [4] we showed that CANDLE HTGRs with thorium fuel gave better burnup performance than the ones with uranium fuel. This fact is also true for other thorium fueled HTGRs of the pebble bed type with once-through-then-out (OTTO), multipass as well as peu-a-peu (PAP) schema as described, analyzed and reviewed comprehensively by Liem et al. [6]. Disregarding the type of reactor and fueling scheme, the thorium CANDLE HTGRs show relatively higher axial power peaking factors which may not be advantageous during a depressurization accident, especially if one try to increase the reactor thermal power. In this work we investigated the use of Pa-231 mixed homogeneously in the fuel kernels for flattening the axial power peaking factor of the thorium CANDLE HTGRs. The motivation of the present investigation is as follows. Pa-231 has a large thermal neutron absorption cross section (201.7 barns), which is about one or two orders greater than that of Th-232 (7.4 barns) and U-238 (2.7 barns). After absorbing two thermal neutrons, Pa-231 is transmuted to U-233 fissile isotope. Therefore, Pa-231 has a potential to be used as burnable poison (BP) in the early stage and fertile fuel in the later stage of burnup. The U-233 bred from Pa-231 will contribute to increase the reactivity at the later stage of burnup, i.e. broadening the burning region which in turn will lower the axial power Copyright © 2014-2015 by JSME peaking factors. A feasibility study of U-Th-Pa fuel in pebble bed reactors was investigated, in which additional Pa-231 content of 3-5% increases fuel burnup performance of about 30% [7]. A difficulty is that natural abundance of Pa-231 is very little due to its short half-life (3.28×104 years) compared to other rare earth elements. However, Pa-231 can be produced from Th-231 via a (n, 2n) reaction followed by a beta decay with a half-life of about 1 d [8]. This reaction has threshold energy of about 7 MeV, and would be suitable with fast neutrons from an accelerator driven system or a fusion reactor [9]. 2.2 CANDLE Burnup Equilibrium and Critical Search Procedure In order to obtain a CANDLE burnup equilibrium and critical condition, in the first stage, one has to determine the fresh fuel kernel composition, i.e. the fissile U-233 (enrichment), natural Gd burnable poison and the small addition of Pa-231 weight fractions (the rest is Th-232), and search the CANDLE burnup equilibrium condition. 2. CALCULATION MODELS AND CONDITIONS An in-house analytical tool for obtaining the CANDLE burnup equilibrium condition has long been developed in our previous works for both fast and thermal reactors. The code takes into account the burning region movement, nuclides burnup and criticality equations simultaneously. The details of the computational procedures are given in references [1, 2, 3, 4]. In this section the group-constants preparation, CANDLE burnup equilibrium and critical search procedure as well as the calculation conditions are discussed. 2.1 Group Constants Preparation The analytical tool used for obtaining the CANDLE burnup equilibrium condition, in principal, needs (1) effective microscopic cross sections and (2) burnup related data such as fission product yields, decay constants, branching ratios and depletion chain. The effective microscopic cross sections were prepared using the collision probability (PIJ) module of SRAC code system [10] with a SRAC library based on the JENDL-3.3 evaluated nuclear data [11]. The burnup related data were taken directly from the SRAC library. Figs. 3 and 4 show the SRAC calculation model for the group-constants generation. Since we adopted the JAEA High Temperature Engineering Test Reactor (HTTR, 30 MWth) type HTGR fuel, in the 2-D hexagonal fuel lattice, the cell is divided into annulus fuel compact, graphite sleeve, annulus helium coolant channel and graphite block. Use of TRISO coated fuel particles in the HTGR fuel compact demands double heterogeneity calculation feature which is provided by SRAC code system in its PIJ module. Furthermore, in the resonance energy region, the ultra-fine energy group capability of SRAC (PEACO module with its MCROSS ultrafine group library) was utilized to obtain accurate effective accurate cross sections on the energy region. After obtaining the effective microscopic cross sections in 107 energy group (the largest number for SRAC code and library) then the cross sections were collapsed into 4 energy group (Table 1) to be used for the CANDLE burnup calculations. The depletion chain which consists of 29 heavy metal, 66 important fission product (including one pseudo fission product) nuclides and 16 burnable poison nuclides is based on SRAC’s THCM66FP chain [10]. Fig. 3 TRISO coated fuel particle and HTTR type fuel compact Fig. 4 Cell calculation model by SRAC2006 code system Copyright © 2014-2015 by JSME In this stage, the obtained equilibrium condition may give an effective neutron multiplication factor (keff) which is not critical. In the second stage, only the natural gadolinium burnable poison concentration is adjusted to obtain a critical (keff=1.0) and equilibrium condition of CANDLE burnup. This two-stage iterative procedure is terminated if the keff is near 1.0 within 1 % convergence criterion. There is a possibility that a certain composition of U-233 enrichment and Pa-231 weight fraction will not give a critical and equilibrium condition of CANDLE burnup. Table 1. Neutron energy group structure for CANDLE burnup calculations (4-group) Group 1 2 3 4 Fission Slowdown Resonance Thermal Energy (eV) Upper Lower 1.0000E+07 1.1109E+05 1.1109E+05 2.9023E+01 2.9023E+01 2.3824E+00 2.3824E+00 1.0000E-05 Lethargy Upper Lower 0 4.5000 4.5000 12.7500 12.7500 15.2500 15.2500 27.6310 Table 2. Design parameters of small sized long life thorium CANDLE high temperature gas-cooled reactor Thermal power (MWth) 30 Core Diameter (cm) 230 Active height (cm) 800 Radial reflector (graphite) Thickness (cm) 100 Coated fuel particle Fuel (U-233/Th-232)O2 U-233 enrichment (w/o) 10 and 15 Burnable poison material Natural gadolinium Type TRISO Kernel diameter (mm) 0.608 Particle diameter (mm) 0.940 Coating material PyC / PyC / SiC / PyC Thickness (mm) 0.060 / 0.030 / 0.030 / 0.046 Density (g/cm3) 1.143 / 1.878 / 3.201 / 1.869 Packing fraction (v/o) 30.0 Fuel compact JAEA HTTR type Inner diameter (cm) 1.00 Outer diameter (cm) 2.60 Graphite sleeve Inner diameter (cm) 2.60 Outer diameter (cm) 3.40 Coolant annulus channel Inner diameter (cm) 3.40 Outer diameter (cm) 4.10 Fuel pitch Flat to flat distance (cm) 6.60 2.3 Calculation Conditions In order to investigate the effect of Pa-231 addition on the axial power peaking of a thorium CANDLE HTGR, we first set the Pa-231 concentration equal to zero to obtain a comparison design case for each U-233 enrichment (10 and 15 w/o), followed by several cases of Pa-231 concentration values. It should be noted that the Pa-231 is mixed homogeneously in the TRISO fuel kernels. The reactor design parameters shown in Table 2 are identical with our previous work for thorium CANDLE HTGR by Ismail et al. [4]. These design parameters represent a small sized long life thorium CANDLE HTGR with thermal power of 30 MWth. The active core height is set to be 800 cm to obtain continuous operation for more than 25 years before refueling operation is needed. However as mentioned before, the core life is easily extended since it is proportional to its active core height. 3. RESULTS AND DISCUSSIONS 3.1 Thorium CANDLE-HTGR without Pa-231 First we consider the critical and equilibrium condition of thorium CANDLE HTGR without Pa-231 addition for U-233 enrichment of 15 and 10 w/o. The natural Gd burnable poison concentrations needed for the two enrichments were found to be 8.75 and 5.40 w/o, respectively. For U-233 enrichment of 15 w/o, the burning region moving velocity is around 22.74 cm/year so that with an active core height of 800 cm the core life time can reach approximately 35 years. A considerably high discharge burnup of 138 GWd/t can be attained. The axial power peak is around 5.93 W/cm3. For U-233 enrichment of 10 w/o, the core life time is found to be about 25 years while the attained discharge burnup is around 96 GWd/t. However, the axial power peak is higher than the one with U-233 enrichment of 15 w/o, i.e. 6.10 W/cm3. The axial power peaks for the two cases are also the maximum power peaks of the whole core and their locations are in the radial core center (r=0 cm). At the core and graphite reflector radial boundaries (r=115 cm) we also observed increased values of power density but still lower than the peak in the radial core center. This can be understood since we do not adopt different fresh fuel compositions in the radial direction. Several important nuclide distributions, thermal neutron flux distribution, power density and kinf in the axial direction at the radial core center (r=0 cm) are shown in Figs. 5 and 6. In these figures, the left hand side of the axial position corresponds to the fresh fuel region. It can be observed that the kinf values are less than unity in the fresh fuel region while larger than unity at the major part of the burning region. In the vicinity of burning region, one can observe the thermal neutron flux peaks which are responsible for the axial power density peaks. In the burning region where the neutron flux is high so that the burnable poison concentration (represented in the figures by Gd-157) depletes sharply as well as the fissile U-233 concentration. On the other hand, one can observe the U-234 and U-235 concentrations build up at the burning region and finally saturate at the spent fuel region. The buildup of Pa-233 concentrations is attributed to Th232 neutron absorptions but the concentrations rapidly decrease because of its short half life (around 27 days). Copyright © 2014-2015 by JSME 15 w/o U-233, 8.75 w/o Nat. Gd (No Pa-231) 10 w/o U-233, 5.40 Nat. Gad (No Pa-231) 8.0E+13 Pa-231 7.0E+13 Nuclide Density (atoms/cm3) 6.0E+13 U-232 1.0E+19 5.0E+13 U-233 4.0E+13 U-234 3.0E+13 U-235 1.0E+17 Gd-157 2.0E+13 1.0E+16 Thermal Flux 1.0E+13 1.0E+15 100 200 300 400 500 600 700 1.0E+18 3.0E+13 U-235 2.0E+13 Gd-157 Thermal Flux 100 200 300 400 500 600 700 800 Axial Position (cm) (a)Nuclide densities (left axis) and thermal neutron flux (right axis) 10 w/o U-233, 5.40 Nat. Gd (No Pa-231) 8.00 Power Density Thermal Neutron Flux 7.00 1.30 Infinite Multiplication Factor 6.00 1.20 5.00 1.10 4.00 1.00 3.00 0.90 2.00 0.80 1.00 0.70 0.00 0.60 500 600 700 800 Axial Position (cm) Power Density (W/cm3), Neutron Flux (x10 13 n/cm2.s) Power Density (W/cm3), Neutron Flux (x10 13 n/cm2.s) U-234 0.0E+00 0 1.40 400 4.0E+13 1.0E+13 15 w/o U-233, 8.75 w/o Nat. Gd (No Pa-231) 300 U-233 1.0E+16 800 8.00 200 U-232 5.0E+13 1.0E+17 (a)Nuclide densities (left axis) and thermal neutron flux (right axis) 100 Pa-233 6.0E+13 1.0E+19 Axial Position (cm) 0 Pa-231 1.0E+20 1.0E+15 0.0E+00 0 8.0E+13 7.0E+13 Pa-233 1.0E+20 1.0E+18 1.0E+21 Nuclide Density (atoms/cm3) 1.0E+21 1.40 Power Density Thermal Neutron Flux 7.00 1.30 Infinite Multiplication Factor 6.00 1.20 5.00 1.10 4.00 1.00 3.00 0.90 2.00 0.80 1.00 0.70 0.00 0.60 0 100 200 300 400 500 600 700 800 Axial Position (cm) (b)Power density, thermal neutron flux (left axis) and kinf (right axis) (b)Power density, thermal neutron flux (left axis) and kinf (right axis) Fig. 5. Critical and equilibrium condition of thorium CANDLE HTGR with U-233 enrichment of 15 w/o and burnable poison concentration of 8.75 w/o (without Pa-231 addition). Fig. 6. Critical and equilibrium condition of thorium CANDLE HTGR with U-233 enrichment of 10 w/o and burnable poison concentration of 5.40 w/o (without Pa-231 addition). The decay of Pa-233 contributes to the production of the fissile U-233. A very small amount of Pa-231 in the burning region mostly from the (n,2n) reaction of Th-232 can be observed. Its product after absorbing one neutron followed by a beta decay, i.e. the U-232 can also be observed saturated in the spent fuel region, however the amount is also small so that U-232 contribution for the fissile U-233 production via neutron absorption is negligible. The effects of Pa-231 addition on the axial power peak and other burnup characteristics of the thorium CANDLE HTGR are summarized in Table 3. For example, in the case of 15 w/o enriched U-233 with 7.50 w/o Pa-231 concentration, the annual requirement of Pa-231 is around 64 kg/year at nominal power. Several important nuclide distributions, thermal neutron flux distribution, power density and kinf in the axial direction at the radial core center are shown in Fig. 7 for enrichment of 15 w/o and Pa-231 weight fraction of 7.50 w/o, and in Fig. 8 for enrichment of 10 w/o and Pa-231 weight fraction of 5.00 w/o. From Table 3 one can observe that a significant improvement on the maximum power density can be achieved by Pa-231 addition for both U-233 enrichments, i.e. a reduction of more than 40%. 3.2 Effects of Pa-231 Addition for Flattening the Axial Power Distribution For U-233 enrichment of 15 w/o, three Pa-231 concentrations were evaluated, i.e. 2.50, 5.00 and 7.50 w/o, while for enrichment of 10 w/o, one Pa-231 concentration was evaluated (5.00 w/o). Copyright © 2014-2015 by JSME 15 w/o U-233, 2.00 w/o Nat. Gd (7.5 w/o Pa-231 addition) 10 w/o U-233, 0.40 w/o Nat. Gd (5.00 w/o Pa-231 addition) 8.0E+13 Pa-231 7.0E+13 Nuclide Density (atoms/cm3) 6.0E+13 U-232 1.0E+19 5.0E+13 U-233 4.0E+13 U-234 3.0E+13 U-235 1.0E+17 Gd-157 2.0E+13 1.0E+16 Thermal Flux 1.0E+13 1.0E+15 100 200 300 400 500 600 700 1.0E+18 U-235 2.0E+13 Gd-157 Thermal Flux 100 200 300 400 500 600 700 800 (a)Nuclide densities (left axis) and thermal neutron flux (right axis) 10 w/o U-233, 0.40 w/o Nat. Gd (5.00 w/o Pa-231 addition) 8.00 Thermal Neutron Flux 1.30 Infinite Multiplication Factor 6.00 1.20 5.00 1.10 4.00 1.00 3.00 0.90 2.00 0.80 1.00 0.70 0.00 0.60 500 600 700 800 Axial Position (cm) Power Density (W/cm3), Neutron Flux (x10 13 n/cm2.s) Power Density (W/cm3), Neutron Flux (x10 13 n/cm2.s) 3.0E+13 Axial Position (cm) Power Density 400 U-234 0.0E+00 0 1.40 300 4.0E+13 1.0E+13 15 w/o U-233, 2.00 w/o Nat. Gd (7.5 w/o Pa-231 addition) 200 U-233 1.0E+16 800 8.00 100 U-232 5.0E+13 1.0E+17 (a)Nuclide densities (left axis) and thermal neutron flux (right axis) 0 Pa-233 6.0E+13 1.0E+19 Axial Position (cm) 7.00 Pa-231 1.0E+20 1.0E+15 0.0E+00 0 8.0E+13 7.0E+13 Pa-233 1.0E+20 1.0E+18 1.0E+21 Nuclide Density (atoms/cm3) 1.0E+21 1.40 Power Density Thermal Neutron Flux 7.00 1.30 Infinite Multiplication Factor 6.00 1.20 5.00 1.10 4.00 1.00 3.00 0.90 2.00 0.80 1.00 0.70 0.00 0.60 0 100 200 300 400 500 600 700 800 Axial Position (cm) (b)Power density, thermal neutron flux (left axis) and kinf (right axis) (b)Power density, thermal neutron flux (left axis) and kinf (right axis) Fig. 7. Axial distribution of important nuclides and thermal neutron flux for U-233 enrichment of 15 w/o and burnable poison concentration of 2.00 w/o (with Pa-231 addition of 7.5 w/o). Fig. 8. Axial distribution of important nuclides and thermal neutron flux for U-233 enrichment of 10 w/o and burnable poison concentration of 0.40 w/o (with Pa-231 addition of 5.0 w/o). The addition of Pa-231 also contributes in enhancing the thorium CANDLE burnup performance as can be observed from the increased discharged burnup and slower velocity of burning region movement. These promising effects of Pa-231 addition are discussed below. Pa-231, with a higher absorption cross section relative to Th-232, is transmuted to Pa-232 by a neutron absorption which is shortly followed by a decay reaction (half-life of about 1.3 day) into U-232 (see Figs. 7 and 8). Since the fission cross section of U-232 is in the same order with its absorption cross section in thermal energy region, its fission rate and transmutation rate to the fissile U-233 are comparable. Table 3. Effects of Pa-231 addition on the maximum power density and other burnup characteristics. Burning Max. Nat. Ave. U-233 Pa-231 Region Power Gd Burnup (w/o) (w/o) Velocity Density (w/o) (GWd/t) (cm/year) (W/cm3) 15.0 0.00 8.75 138 22.7 5.93 2.50 5.50 149 20.4 5.86 5.00 4.00 142 21.0 4.79 7.50 2.00 149 19.6 3.57 10.0 0.00 5.40 95 31.8 6.10 5.00 0.40 97 29.7 3.26 Copyright © 2014-2015 by JSME As discussed in the previous subsection, the fissile U-233 is also bred from Th-232 via Pa-233 but presumably slower due to the longer half-life of Pa-233 (26 days). The Pa-231 addition relaxes the rapid depletion of U-233 especially in the burning region which in turn will widen the full height maximum width of the moving burning wave. This phenomenon is responsible for the decrease of the maximum power density in the axial moving direction of the burning wave as clearly shown in Figs. 7 and 8. From the figures, one can also observe that the Pa-231 concentration depletion profile resembles the burnable poison’s profile which indicates that Pa231 plays also the essential role as a burnable poison in the CANDLE burnup phenomenon. However, although not shown here, our parametric calculation results showed that it is impossible to obtain an equilibrium yet critical thorium CANDLE HTGR by eliminating the burnable poison (natural Gd) completely (i.e. using only Pa-231). Hence, Pa-231 can only reduce the amount of burnable poison needed for a physically realizable thorium CANDLE HTGR. Within the U233 enrichment considered here (15 and 10 w/o), the Pa-231 addition is also contributing in slowing down the velocity of the burning wave so that double fold benefits are achieved simultaneously i.e. larger discharged burnup and longer core life time for the same active core height. From Figs. 7 and 8, one can also observe that the kinf of the spent fuel are slightly greater than one. This indicates that a criticality safety measure must be taken into account for handling (storage and transportation) of the spent fuel. 4. CONCLUSIONS The effects of Pa-231 addition on decreasing the maximum power peak in the thorium CANDLE HTGRs have been investigated for U-233 enrichment of 15 and 10 w/o. For U-233 enrichment of 15 w/o, three Pa-231 concentrations were evaluated, i.e. 2.50, 5.00 and 7.50 w/o, while for enrichment of 10 w/o, one Pa-231 concentration was evaluated (5.00 w/o). Comparing with the original thorium CANDLE HTGR (without Pa-231 addition), for all cases Pa-231 addition decreases the maximum power density and at the same time increases the discharge burnup and core life time. From the numerical simulations in the present work, it is shown that Pa231 plays partly the role of a burnable poison in the early stage and of a fertile fuel in the later stage of burnup. The U-233 bred from Pa-231 contributes to the reactivity increase at the later stage of burnup, i.e. broadening the burning region (wave) which in turn lowers the axial power peaks. [2] Ohoka Y. and Sekimoto H., “Application of CANDLE Burnup to Block-Type High Temperature Gas Cooled Reactor”, Nuclear Engineering Design, Vol. 229 No. 1, pp. 15-23 (2004). [3] Ohoka Y., Watanabe T and Sekimoto H., “Simulation Study on CANDLE Burnup Applied to Block-Type High Temperature Gas Cooled Reactor”, Progress in Nuclear Energy, Vol. 47, No. 1-4, pp. 292-299 (2005). [4] Ismail, Ohoka Y., Liem P.H. and Sekimoto H., “Long Life Small CANDLE-HTGRs with Thorium,” Annals of Nuclear Energy 34, pp. 120-129 (2007). [5] Ohoka Y. and Sekimoto H., “Application of CANDLE Burnup to Block-Type High Temperature Gas Cooled Reactor for Incinerating Weapon Grade Plutonium,” Proceedings of GENES4/ANP2003, September 15-19, 2003, Kyoto, Japan. [6] Liem P. H., Ismail and Sekimoto H., “Small High Temperature Gas-Cooled Reactors with Innovative Nuclear Burning,” Progress in Nuclear Energy Vol. 50, pp. 251-256 (2008). [7] Tran H.N. and Liem P.H., “Neutronic feasibility study of U-Th-Pa based high burnup fuel for pebble bed reactors” Progress in Nuclear Energy Vol. 80, pp.17-23 (2015). [8] Shmelev, A., Saito, M., Artisyuk, V., “Multi-component self-consistent nuclear energy system: on proliferation resistance aspect,” Proceeding of 2nd Annu. JNC Int. Forum on the Peaceful Use of Nuclear Energy, Tokyo, Japan, Feb. 21-22, (2000). [9] Imamura, T., Saito, M., Yoshida, T., Artisyuk, V., “Potential of Pa for gas cooled long-life core,” J. Nucl. Sci. Technol. Vol. 39, pp. 226-233, (2002) . [10] Okumura K., Kugo T., Kaneko K. and Tsuchihashi K., “SRAC2006: A Comprehensive Neutronics Calculation Code System,” JAEA-Data/Code 2007-004 (2007). [11] Shibata K. Kawano T., Nakagawa T., Iwamoto O., Katakura J., Fukahori T. et al., “Japanese Evaluated Nuclear Data Library Version 3 Revision-3: JENDL-3.3,” Journal of Science and Technology 39, pp. 1125-1136 (2002). ACKNOWLEDGMENTS Some figures related to CANDLE burnup and cell calculation model used in the manuscript were provided by Drs. Ohoka Y. and Ismail. REFERENCES [1] Sekimoto H., Ryu K and Yoshimura Y., “CANDLE: The New Burnup Strategy”, Nuclear Science and Engineering, Vol. 139, No. 306 (2001). Copyright © 2014-2015 by JSME

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