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Neutronic feasibility study of UeThePa based high burnup fuel for pebble bed reactors Hoai-Nam Tran a, *, Peng Hong Liem b a b

Institute of Research and Development, Duy Tan University, K7/25 Quang Trung, Da Nang, Viet Nam Nippon Advanced Information Service (NAIS Co., Inc.) Scientiﬁc Computational Division, 416 Muramatsu, Tokaimura, Ibaraki, Japan

a r t i c l e i n f o

a b s t r a c t

Article history: Received 22 September 2014 Received in revised form 28 November 2014 Accepted 30 November 2014 Available online

Neutronic feasibility study of UeThePa based high burnup fuel for the pebble bed reactors has been carried out. An additional content of 231 Pa in a mixed 233 Ue232 Th fuel pebble was optimized for simultaneously controlling long-term excess reactivity and attaining high fuel burnup. Because of a large thermal absorption cross section, 231 Pa has a potential to absorb thermal neutrons and act as a burnable absorber in the early stage of burnup. In the latter stage, 231 Pa is transmuted to 233 U and act as a fertile fuel. Parametric study was performed for heavy metal (HM) amount, 233 U enrichment and 231 Pa content. The objective is to attain a target burnup greater than 100 GWd/T with a low uranium enrichment and a small content of 231 Pa. Six fuel compositions with the HM amount of 9e18 g per pebble, the 233 U enrichment of 8 and 10%, and the 231 Pa content of 3.2e5.5% were selected for discussion and comparison. The burnup levels of about 130e170 GWd/t are achieved for the six fuel compositions. The results of temperature coefﬁcient show that negative reactivity feedback during burnup with the increase of temperature is assured, which is one of the important safety parameters. © 2014 Elsevier Ltd. All rights reserved.

Keywords: UeThePa Thorium fuel High burnup Pebble bed reactor

1. Introduction Thorium fuel cycle has been recognized with a number of advantages such as natural abundance, low radiotoxicity waste, chemical and radiational inert of oxide compound (ThO2), high thermal conductivity, low thermal expansion coefﬁcient and proliferation resistance, etc (IAEA-TECDOC-1450, 2005). Natural thorium, mostly in form of fertile 232 Th isotope, is three to four times more than uranium (IAEA-TECDOC-1450, 2005). After absorbing one neutron and followed by decay reactions, 232 Th is transmuted to ﬁssile 233 U isotope. Since the absorption crosssection for thermal neutrons of 232 Th (7.4 b) is about 3 times greater than that of 238 U (2.7 b), it has a greater conversion ratio (converted to 233 U) compared to 238 U (converted to 239 Pu), and therefore, having a potential to be used as a fertile fuel. It also means that more number of thermal neutrons is used to convert 232 Th. Hence, 233 Ue232 Th fuel could be a good option for this purpose. Compared to 235 U and 239 Pu, 233 U has the greatest neutron yield per ﬁssion (h), which is greater than 2.0 over a wide

* Corresponding author. Tel.: þ84 511 3827111 (809); fax: þ84 511 3650443. E-mail address: [email protected] (H.-N. Tran). http://dx.doi.org/10.1016/j.pnucene.2014.11.024 0149-1970/© 2014 Elsevier Ltd. All rights reserved.

range of a thermal neutron spectrum. The average h factor is 2.29 for 233 U, while it is about 2.05 for 235 U, and 1.80 for Pu in a thermal reactor (IAEA-TECDOC-1645, 2010). The High-Temperature Gas-cooled Reactor (HTGR), a graphitemoderated He-cooled reactor, shows distinguishing features for the use of Th-based fuel (IAEA-TECDOC-1155, 2000; IAEA-TECDOC1349, 2003; Lung and Gremm, 1998; Mazzini et al., 2009; Liem et al., 2008). The fuel in form of TRISO-coated particles embedded in graphite matrix has potential to conﬁne ﬁssion products for a long burnup (up to 800 GWd/T) (IAEA-TECDOC-1349, 2003; Mazzini et al., 2009). The use of thorium fuel in the pebble bed reactor (PBR), one of the two HTGR technologies, has been started for over 50 years, such as in highly enriched uranium pebbles in the 15 MWe AVR from 1967 to 1988 and in the 300 MWe THTR in the 1980s (Lung and Gremm, 1998). In a previous work conducted in 1990s, design procedures for very small and small-sized modular pebble bed HTGRs with uranium and thorium fuels were proposed and discussed (Liem, 1996) and it was shown that even for these reactor scales thorium fuel burnup performance is signiﬁcantly superior to the uranium fuel. In PBRs, multi-pass refueling scheme is the most common, in which the fuel pebbles are reloaded into the core for several cycles until they reach the designed target burnup. Once-through-then-

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out (OTTO) scheme is an alternative option, in which only fresh fuel pebbles are loaded during core operation and the target burnup of the fuel pebble is achieved during one irradiation period (Hansen et al., 1972). The OTTO scheme is simplest because of no need of refueling mechanism. However, the biggest problem of the OTTO scheme is the high power peak at the core top making it unpractical for medium and large sizes of the reactor. This is because the fresh fuel pebbles are concentrated at the core top, while the irradiated pebbles are concentrated at the core bottom. In a previous 500 MWth core design with OTTO refueling, the power peaking factor is 3.62 at the core top. The power peaking factor is reduced to 2.08 by applying a multi-pass refueling scheme, in which each pebble passes through the core four times (Hansen et al., 1972). For very small and small-sized PBRs similar trends are observed (Liem, 1996; Liem et al., 2008). A comprehensive review on the conventional PBR refueling schema (which includes OTTO as well as multipass) can be found in Liem et al. (2008). A uniform axial power distribution and a low power peaking factor can be obtained through controlling a constant axial k∞ distribution in the core. In the OTTO scheme, this is equivalent to maintain constant k∞ of the fuel pebble during burnup in the lattice model (Tran and Kato, 2009). New fuel pebbles designed with socalled spherical burnable poison (BP) particles could help to control the axial power proﬁle of the OTTO scheme effectively (Tran et al., 2008; Tran and Kato, 2009). However, the adoption of BP leads to the decrease of fuel burnup performance. Comparison between the OTTO scheme using new fuel pebbles designed with B4C and Gd2 O3 particles and the reference multi-pass scheme of PBMR-400 shows that the core performance is comparable but the fuel burnup of the OTTO scheme is about 20e21% less than that of the multi-pass scheme (Tran, 2012; Tran and Hoang, 2012). Hence, it is of interest to investigate a possible fertile material that acts as BP in the early burnup stage for the purpose of long-term reactivity control. Then, this material is transmuted to ﬁssile fuel in the later stage to increase neutron economy, and as a result, the fuel burnup performance increases. The concept of using fertile material for simultaneously controlling excess reactivity and attaining a high burnup fuel is similar to the use of minor actinides, especially 237 Np, for lengthening the core lifetime of fast breeder reactors (Tran and Kato, 2008). 237 Np absorbs fast neutrons in the early stage and then transmuted to 239 Pu, which act as ﬁssile fuel afterward. The use of 237 Np for long life core of water-cooled reactors was also investigated (Nikitin et al., 1999). Similar to 237 Np, 231 Pa has a large thermal neutron absorption cross section (201.7 b), which is about one or two orders greater than that of 232 Th (7.4 b) and 238 U (2.7 b). After absorbing two thermal neutrons, 231 Pa is transmuted to 233 U ﬁssile isotope. Therefore, 231 Pa has a potential to be used as burnable poison (BP) in the early stage and fertile fuel in the later stage of burnup. The 233 U bred from 231 Pa will contribute to increase the reactivity at the end of cycle and lengthen the cycle life-time. The conversion ratio of 231 Pa is also increased when using with 233 Ue232 Th fuel because of its large h factor. A difﬁculty is that the natural abundance of 231 Pa is very little due to its short half-life (3:28 104 yr) compared to other rareearth elements. However, 231 Pa can be produced from 232 Th via a (n, 2n) reaction to transmute to 231 Th and followed by a beta decay with a haft-life of about 1 d (Shmelev et al., 2000). 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 (Imamura et al., 2002). A feasibility study of using 231 Pa for designing a long-life fuel of PBRs was performed to achieve a high burnup up to 700 GWd/t of 233 Ue231 Pa fuel with equal mass fractions (Imamura et al., 2002). However, these mass fractions seem too large for both 231 Pa and 233 U when considering the total amount of 231 Pa and the enrichment of 233 U to be supplied.

Uranium fuel with 9.6 wt% 235 U enrichment of PBMR-400 achieves a target burnup of about 90e95 GWd/T (Boer et al., 2009; Mulder and Teuchert, 2008). Thorium fuel mixed with Pu or minor actinides could achieve a target burnup of about 100e200 GWd/T (Acir and Coskun, 2012, 2015). In the present work, we aim at investigating a conceptual design of UeThePa based fuel with a target burnup greater than 100 GWd/T for the use in PBRs with the OTTO scheme. An additional content of 231 Pa in the mixed fuel of 233 Ue232 Th is determined and optimized for controlling excess reactivity during burnup. The objectives are to ﬂatten the k∞ curves during burnup and increase the fuel burnup performance. Because of a little natural abundance, the additional content of 231 Pa in the fuel is limited within a few percent. Calculations for the temperature coefﬁcients of the new fuel have also been conducted and discussed. 2. Nuclear characteristics of

231 Pa

Fig. 1 shows a comparison of the neutron absorption crosssections of 232 Th, 238 U and 231 Pa in a wide range of neutron energy according to JENDL-3.3 library (Shibata et al., 2002). One can see that the absorption cross-section of 231 Pa is one or two orders greater than that of 232 Th and 238 U in the thermal energy range. Fig. 2 displays the main reaction chain of 231 Pa and 232 Th with neutron in nuclear reactors. 232 Th produces ﬁrst 233 Th by absorbing a neutron which almost immediately decays to 233 Pa with a halflife of 21.83 min 233 Pa is then transmuted to 233 U via a decay reaction with a half-life of 27 d. This rather long half-life of 233 Pa can result in a reactivity surge a long time after reactor shutdown due to ﬁssile 233 U production and this must be taken into account in safety features (Lung and Gremm, 1998). When 231 Pa is added into the fuel, 231 Pa is transmuted to 232 Pa by a neutron absorption which is shortly followed by a decay reaction to transfer to 232 U with a half-life of 1.31 d. Since the ﬁssion cross section of 232 U is in the same order with its absorption cross section in thermal energy, its ﬁssion rate and transmutation rate to 233 U are comparable. The bred 232 U and 233 U afterward will contribute to increase the fuel burnup. One of the difﬁculties of UeTh fuel cycle is that some decay daughters of 232 U and 232 Th are hard gamma emitting isotopes (2e2.6 MeV). The presence of the gamma emitting isotopes

Fig. 1. Neutron absorption cross sections of JENDL-3.3 library (Shibata et al., 2002).

231 Pa, 232 Th

and

238 U

according to the

H.-N. Tran, P.H. Liem / Progress in Nuclear Energy 80 (2015) 17e23

Fig. 2. Reaction chain of

requires a fuel fabrication technique remotely in a gamma-shielded environment (Lung and Gremm, 1998). This makes the process complicated and expensive. The problem could become more complicated with the appurtenance of 231 Pa content since it produces a large amount of 232 U during irradiation compared to original UeTh fuel. It is expected that the advantages of the TRISO fuel of HTR could help to solve this problem (Lung and Gremm, 1998). 3. Calculation model Numerical calculations for conceptual design of the UeThePa based fuel pebble for PBR have been carried out using a lattice calculation model. The lattice model consists of one central fuel pebble surrounded by a coolant layer as illustrated in Fig. 3. The fuel pebble with an outer diameter of 60 mm consists of many TRISOcoated fuel particles distributed stochastically in a graphite matrix with an outer diameter of 50 mm. The number of TRISO-coated fuel particles in each fuel pebble is determined via the total amount of heavy metal (HM) loaded into the fuel pebble and the diameter of the fuel kernel. The thickness of the coolant layer of 5.35 mm is determined so that the volume ratio of the fuel pebble to the total volume is 0.61, which is equal to the fuel packing fraction in the core. The TRISO-coated fuel particle consists of a fuel kernel coated by four layers of pyrolytic carbon and SiC for reﬁning ﬁssion

231 Pa

and

19

232 Th.

products and supporting fuel integrity. The kernel contains a mixture of UeThePa fuel. For designing the fuel pebble with UeThePa fuel, we started with a reference conﬁguration design parameters of the fuel pebble used in the PBMR-400 as displayed in Table 1. The lattice calculations have been performed using the MVP code (Nagaya et al., 2005) and the JENDL-3.3 library (Shibata et al., 2002). MVP is a general purpose Monte Carlo transport code for neutron and photon developed by Japan Atomic Energy Agency (JAEA). In the lattice calculations, a reﬂective boundary condition was applied. The number of histories of 1:5 106 was chosen to achieve a statistic error of k∞ within 0.025%. The STG model of MVP was used to simulate the stochastic distribution of the fuel particles. 4. Results and discussions 4.1. Fuel burnup performance In the present work, neutronics design of UeThePa based fuel pebble has been conducted for attaining a target burnup greater than 100 GWd/T. In the ﬁrst step of the design process, the UeTh consumption in a fuel pebble is estimated via a parametric study, in which the total HM amount per pebble and the enrichment of 233 U are considered. The fuel is assumed to consist of only 233 U and 232 Th isotopes. In the second step, the additional content of 231 Pa in the mixed 233 Ue232 Th fuel is determined and optimized for controlling the excess reactivity and increasing the fuel burnup performance. 231 Pa is mixed uniformly in the 233 Ue232 Th fuel and its content is limited within a few percent due to a small abundance. For the desire target burnup of greater than 100 GWd/T, six fuel compositions corresponding to the HM amount of 9, 12, 15 and 18 g and the enrichment of 233 U of 8% and 10% have been selected for further design and comparison. Low enrichment of 233 U (less than 12%) is selected for the purpose of proliferation resistance. Fig. 4 Table 1 Parameters of a fuel pebble.

Fig. 3. Lattice calculation model in the MVP code.

Parameters

Value

Fuel Coolant Fuel density (g=cm3 ) Pebble diameter (mm) Free fuel zone diameter (mm) Graphite matrix density (g=cm3 ) Fuel kernel diameter (mm) Fuel temperature (K) Coating material Thickness of coating layers (mm) Density of coating material (g=cm3 )

UeThePa He 10.2 60 50 1.75 500 1200 PyC/PyC/SiC/PyC 95/40/65/40 1.05/1.90/3.18/1.90

20

H.-N. Tran, P.H. Liem / Progress in Nuclear Energy 80 (2015) 17e23

displays the evolution of the k∞ of the 233 Ue232 Th fuel pebble with the six fuel compositions. It is the fact that the discharged burnup of the fuel in PBR depends on the refueling scheme and the core performance including the leakage effect of the core. In a multipass scheme, the k∞ of the fuel at the discharged burnup could be lower than 1.0 since the core is mixed with fuel pebbles at many different burnup levels. The fuel pebble in the multi-pass scheme of PBMR-400 with six passes per each pebble can achieve the discharged burnup of 90e95 GWd/T which corresponds to the k∞ of about 0.95 (Boer et al., 2009; Mulder and Teuchert, 2008). The same fuel but with additional BP particles used in the OTTO scheme can achieve the discharged burnup of 75 GWd/T corresponding to the k∞ of about 1.0 (Tran and Hoang, 2012). This UeTh based fuel could achieve a higher burnup if it is used in the multi-pass scheme, and of course the fuel pebble would have to be reloaded for a greater number of cycles compared to that in the PBMR-400 since it has a greater initial excess reactivity. In the present work, for simplicity in comparison it is assumed that the maximum attainable burnup of the fuel is when the k∞ decreases to 1.0. For the case of 9 g HM and 10% 233 U, the burnup of 115 GWd/T can be achieved and the excess reactivity is the greatest compared to other cases as shown in Fig. 4. Increasing the amount of HM while keeping the same enrichment of 233 U leads to the decrease of the excess reactivity and the increase of burnup level. However, the increase of fuel burnup is only about 5 GWd/T when the HM increases to 12 g and 10 GWd/T when the HM increases to 15 g. This means that the burnup increases only about 5% with the increase of 3 g HM or the increase of 30% of the total HM amount. The trend is similar to the cases of 8% 233 U, but the increase of burnup is smaller, about 3 GWd/T or 2e3% corresponding to the increase of 3 g HM. Comparing between the two enrichments of 8% 233 U and 10% 233 U, when the HM is 12 g the higher burnup of about 15% can be achieved by increasing 2% enrichment of 233 U. In particular, the burnup of 100 GWd/T is achieved for 12 g HM and 8% 233 U, but it is about 120 GWd/T when increase 233 U to 10%. Thus, increasing 233 U enrichment would lead to the increase of fuel burnup more effectively. In order to optimize the content of 231 Pa in the fuel, 231 Pa is mixed uniformly in 233 Ue232 Th fuel, and its content is varied within a few percent. The objective is to maintain a ﬂat k∞ curve around the value of 1.1 in the early burnup stage for minimizing the initial excess reactivity. This value is assumed to assure the criticality when the leakage effect is taken into account in a full core

Fig. 5. k∞ curves with burnup of UeThePa fuel pebble.

model. In the later stage of burnup, the k∞ value should be as high as possible for attaining a high discharged burnup. Fig. 5 illustrates the k∞ of the fuel pebble as a function of burnup with the optimal content of 231 Pa. Table 2 summarizes the design parameters of the fuel with the six fuel compositions of the total HM amount and 233 U enrichment. For all cases, the target burnup of about 130e170 GWd/T can be achieved as displayed in Fig. 5 and Table 2. The target burnup is increased by about 32e37% compared to the fuel without 231 Pa. The content of 231 Pa is determined within 3.2% and 5.5% of the total HM amount. The highest burnup of 132 GWd/T is obtained for 12 g HM and 8% U while it is 170 GWd/T for 15 g HM and 10% U. This is also the highest achievable burnup among the six fuel compositions. For the case of 10% 233 U, the 231 Pa contents of 5.5, 5 and 4.5% are determined corresponding to the HM of 9, 12 and 15 g, respectively. This means that a fuel composition with a lower HM amount requires more 231 Pa content to compensate the greater initial excess reactivity. Comparing the maximum attainable burnup of the six fuel compositions, similar to the fuel without 231 Pa one can see that increasing the HM amount does not lead to the increase of fuel burnup effectively. For the fuel with 8% 233 U, increasing the HM from 12 g to 18 g (50%) leads to the increase of fuel burnup of only 10 GWd/T (about 8%). Similarly, in the fuel with 10% 233 U, when the HM increases from 9 g to 15 g (67%), the fuel burnup increases only 12 GWd/T corresponding to about 8%. In contradiction, increasing 233 U enrichment leads to the increase of burnup more effectively. Increasing 233 U enrichment by 2% leads to the increase of burnup of about 18e25%. If the efﬁciency of fuel utilization is considered as the ratio of the fuel target burnup to the consumed 233 U amount or the total HM amount, among the six fuel compositions, the ﬁrst case with 9 g HM and 10% of 233 U is the most effective in fuel utilization. This case consumes least HM and ﬁssile amounts but Table 2 Summary of the design parameters of the UeThePa fuel pebble.

Fig. 4. k∞ curves with burnup of UeTh fuel pebble.

Heavy metal (g)

233 U

9 12 12 15 15 18

10 8 10 8 10 8

(wt%)

231 Pa

5.5 3.5 5 3.4 4.5 3.2

(wt%)

Burnup (GWd/T) 158 132 166 139 170 142

H.-N. Tran, P.H. Liem / Progress in Nuclear Energy 80 (2015) 17e23

Fig. 6. Atomic number densities of

232 Th, 231 Pa, 232 U

and

233 U

21

isotopes as a function of burnup.

Fig. 7. Temperature coefﬁcient of the UeThePa fuel pebble.

achieve the burnup of 158 GWd/T. The greatest amount of (5.5%) is required for compensating the excess reactivity.

231 Pa

4.2. Atomic number densities Fig. 6 illustrates the change of the atomic number densities of and 233 U isotopes during burnup in the six fuel

232 Th, 231 Pa, 232 U

compositions. It can be seen that the density of 232 Th decreases linearly with burnup, while that of 231 Pa and 233 U decrease exponentially. 231 Pa is almost burnt at the target burnup of the fuel as shown in Fig. 6. For the fuel with higher content of 231 Pa, 231 Pa density decreases slower. The density of 232 U increases during burnup in the early stage, and then decreases after reaching a maximum value. The maximum value appears at about 75 GWd/T

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H.-N. Tran, P.H. Liem / Progress in Nuclear Energy 80 (2015) 17e23

Fig. 8. Neutron spectra in the fuel region of the UeThePa fuel pebble at 0 and 150 GWd/T.

for the fuel compositions with 8% of 233 U of which the content of 231 Pa are 3.2e3.5%. While this value appears at about 100 GWd/T for the three other cases with 10% 233 U when the 231 Pa contents are 4.5, 5 and 5.5%, respectively. The density of 232 U is the greatest in the case of 5.5% 231 Pa. The trend is similar to the other cases that the higher content of 231 Pa leads to the higher density 232 U in the fuel. This means that the evolution of 232 U depends mainly on the content of 231 Pa in the fuel compositions. The formation of 232 U is one of the disadvantage of the Th-based fuel related to safety issue due to its strong gamma decay. However, this issue would be solved by the use of TRISO-coated fuel particles in PBR. 4.3. Temperature coefﬁcient Calculations of temperature coefﬁcients of the UeThePa fuel have been performed to evaluated effect on reactivity feedback when the temperature of fuel, moderator and coolant change. In the calculation model, the temperature was assumed to increase by 100 K, and then the temperature coefﬁcient was calculated by the change of reactivity compared to the normal state. The isothermal coefﬁcient was also calculated by the change of the temperatures of all fuel, moderator and coolant together. Fig. 7 displays the dependence of temperature coefﬁcient on burnup of the UeThePa based fuel with the six fuel compositions. Comparing among the six fuel compositions, the isothermal temperature coefﬁcient is slightly less negative for the cases with greater HM amount or higher 233 U enrichment. The effect of fuel and coolant temperature is stable during burnup, but the coefﬁcient of moderator temperature increases with burnup. The isothermal temperature coefﬁcient becomes less negative and reaches zero or slightly positive at 200 GWd/T for all selected fuel compositions. One of the reason is the shift of neutron spectra during burnup. Fig. 8 shows a comparison of the neutron spectra in the fuel region at the beginning of cycle and at the burnup of 150 GWd/T for the six fuel compositions. It can be seen that the spectra are shifted to thermal range during burnup. Decreasing the 233 U enrichment leads to slightly thermalized spectra. The most important result is that the negative temperature coefﬁcients during burnup are ensured for all fuel compositions. 5. Conclusions Neutronic feasibility study of UeThePa based fuel for the OTTO refueling scheme of PBRs has been performed. An additional content of 231 Pa in the low enriched 233 Ue232 Th fuel as BP and fertile fuel has been optimized for simultaneously controlling long-term excess reactivity and attaining high burnup. Parametric study was performed for the total HM amount, 233 U enrichment and 231 Pa

content in order to achieve a target burnup greater than 100 GWd/ T. Six fuel compositions corresponding to the HM of 9e18 g per pebble, 233 U enrichment of 8e10% and the optimal 231 Pa content of 3.2e5.5%, respectively have been selected for further comparison. The target burnup of 130e170 GWd/T is achieved for all the six fuel compositions. These burnups are increased by about 32e37% compared to that of the fuel without 231 Pa. The highest burnup of 170 GWd/T is obtained for the fuel with 15 g HM, 10% 233 U and 4.5% 231 Pa. However, the fuel with 9 g HM, 10% 233 U and 5.5% 231 Pa, which achieves the target burnup of 158 GWd/T, is considered to be the most effective in fuel utilization. Temperature coefﬁcient calculations show that the negative reactivity feedback during burnup is assured for all the fuel compositions. Acknowledgment The authors are grateful to Mr. V.K. Hoang of Institute for Nuclear Science and Technology for his assistance on MVP calculations. This work has been supported by National Foundation for Science and Technology Development via project No. 103.042014.79. References Acir, A., Coskun, H., 2012. Neutronic analysis of the PBMR-400 full core using thorium fuel mixed with plutonium or minor actinides. Ann. Nucl. Energy 48, 45e50. Acir, A., Coskun, H., 2015. Monte carlo calculations on transmutation of plutonium and minor actinides of pebble bed high temperature reactor. Prog. Nucl. Energy 78, 380e387. http://dx.doi.org/10.1016/j.pnucene.2013.11.009. Boer, B., Kloosterman, J., Lathouwers, D., van der Hagen, T., 2009. In-core fuel management optimization of pebble-bed reactors. Ann. Nucl. Energy 36, 1049e1058. Hansen, U., Schulten, R., Teuchert, E., 1972. Physical properties of the once through then out pebble-bed reactor. Nucl. Sci. Eng. 47, 132e139. IAEA-TECDOC-1155, 2000. Thorium Based Fuel Options for the Generation of Electricity. Technical Report. International Atomic Energy Agency. IAEA-TECDOC-1349, 2003. Potential of Thorium Based Fuel Cycles to Constrain Plutonium and Reduce Long Lived Waste Toxicity. Technical Report. International Atomic Energy Agency. IAEA-TECDOC-1450, 2005. Thorium Fuel Cycle-potential Beneﬁts and Challenges. Technical Report. International Atomic Energy Agency. IAEA-TECDOC-1645, 2010. High Temperature Gas Cooled Reactor Fuels and Materials. Technical Report. International Atomic Energy Agency. Imamura, T., Saito, M., Yoshida, T., Artisyuk, V., 2002. Potential of 231Pa for gas cooled long-life core. J. Nucl. Sci. Technol. 39, 226e233. Liem, P.H., 1996. Design procedures for small pebble-bed high temperature reactors. Ann. Nucl. Energy 23, 207e215. Liem, P.H., Ismail, Sekimoto, H., 2008. Small high temperature gas-cooled reactors with innovative nuclear burning. Prog. Nucl. Energy 50, 251e256. Lung, M., Gremm, O., 1998. Perspectives of the thorium fuel cycle. Nucl. Eng. Des. 180, 133e146. Mazzini, G., Bomboni, E., Cerullo, N., Fridman, E., Lomonaco, G., Shwageraus, E., 2009. The use of th in HTR: state of the art and implementation in Th/Pu fuel cycles. Sci. Technol. Nucl. Install. 2009 http://dx.doi.org/10.1155/2009/749736. Article ID 749736, 13 pages.

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