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Analysis of thorium and uranium based nuclear fuel options in Fluoride salt-cooled High-temperature Reactor
Progress in Nuclear Energy 78 (2015) 285e290
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Analysis of thorium and uranium based nuclear fuel options in Fluoride salt-cooled High-temperature Reactor X.X. Li a, b, c, X.Z. Cai a, b, c, D.Z. Jiang a, b, c, Y.W. Ma a, b, c, J.F. Huang a, b, c, C.Y. Zou a, b, c, C.G. Yu a, b, c, J.L. Han a, b, c, J.G. Chen a, b, c, * a b c
Shanghai Institute of Applied Physics, Chinese Academy of Sciences, Shanghai 201800, China CAS Center for Excellence in TMSR Energy System, Chinese Academy of Sciences, Shanghai 201800, China Key Laboratory of Nuclear Radiation and Nuclear Energy Technology, Chinese Academy of Sciences, Shanghai 201800, China
a r t i c l e i n f o
a b s t r a c t
Article history: Received 4 July 2014 Received in revised form 22 September 2014 Accepted 9 October 2014 Available online 1 November 2014
In order to utilize thorium in Fluoride salt-cooled High-temperature Reactor (FHR), neutronics analysis for thorium based fuels (233 U þ Th, 235 U þ Th and 239 Pu þ Th) is carried out in a whole-core model of pebble bed FHR. Uranium (238 U) based fuels with the above three fissile nuclides are also analyzed for comparison. The atomic density of fissile material is kept constant at start-up for the six fuel types. Neutron characteristics including neutron spectrum, effective multiplicity factor (keff), temperature coefficient of reactivity (TCR), conversion ratio (CR) and burnup for the six fuel options are discussed. With the same fissile nuclide, the thorium based fuels have a higher initial keff than the uranium based fuels due to the smaller resonance absorption of 232 Th. As for fissile material, 233 U is the best candidate as a driver fuel in thermal and epithermal spectra due to its effective number of fission neutrons (h). Besides, the 232 Th/233 U fuel can be extended to radioactive waste management benefited from deeper burnup and lower level of radio-toxicity. The FHR core with thermal and epithermal spectra is more suitable for TheU fuel cycle than UePu fuel cycle. These analyses can provide an approach for the further optimizations of thorium utilization in FHR. © 2014 Elsevier Ltd. All rights reserved.
Keywords: Thorium Neutronics FHR
1. Introduction Most of present commercial reactors using enriched uranium (about 5%) with once-through fuel cycle will consume up the whole world's estimated uranium in a few decades along with the growing demand of energy (David, 2005). It is therefore important to develop advanced reactors with high efficiency of fuel utilization and corresponding fuel cycles (includes fuel types and fuel cycle modes) to solve the problem of nuclear fuel shortage. As an alternative nuclear fuel resource, thorium attracts more and more attentions to ensure sustainable energy generation. Thorium is a fertile fuel which is reported to be 3 ~ 4 times as abundant in the earth's crust as uranium (Wickleder et al., 2006). It will greatly expand the nuclear fuel resources through the conversion behavior from 232 Th to 233 U by a neutron capture. TheU fuel cycle offers attractive features (Cycle, 1450), including better breeding capability in thermal reactors and lower radio-toxicity levels in nuclear waste.
* Corresponding author. Shanghai Institute of Applied Physics, Chinese Academy of Sciences, Shanghai 201800, China. Tel.: þ86 21 39194027. E-mail address: [email protected] (J.G. Chen). http://dx.doi.org/10.1016/j.pnucene.2014.10.004 0149-1970/© 2014 Elsevier Ltd. All rights reserved.
Many countries and organizations have already started the research of strategy and engineering for thorium utilization since mid 1950s (Mathieu et al., 2006; García et al., 2013; Nuttin et al., 2005). The majority of options for the use of thorium previously proposed have relied on highly innovative reactor concepts (Lung, 1997), such as Molten Salt Reactor (MSR) (Merle-Lucotte et al., 2008), Liquid Metal Fast Breeder Reactor (LMFBR) (Ramanna and Lee, 1986), Super Critical Water Reactor (SCWR) (Chaudri et al., 2013) and Accelerator Driven Systems (ADS) (Salvatores et al., 1997). And the studies of thorium utilization with existing reactors, in most cases PWR specific, were also summarized (Puill, 1999). To take the advantages of thorium, especially its great stability at high temperature and breeding capacity in thermal and epithermal spectra, thorium based fuels were tested in High-temperature gascooled Reactors (HTRs). Most of the researches and operating experiences of thorium utilization in HTR are focused on high enriched uranium (HEU) (Baumer and Kalinowski, 1991; Habush and Harris, 1968) or low enriched uranium (LEU) (Pohl, 2006; Ding and Kloosterman, 2014). Fluoride salt-cooled High-temperature Reactor (FHR) (Ingersoll et al., 2004) synthesizes the advantages of MSR and HTR, which
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has similar neutronics with HTR, fueled with TRISO particle, differentiated only by its coolant with fluoride salt instead of helium gas. FHR can support conventional LEU fuel cycle and also advanced fuel options, such as thorium. The study on the neutronics performance of FHR has been concentrated on uranium (François-Paul and Fabien, 2006; Fratoni, 2008; U. M. Facilitators and M. Facilitators, 2013), while that of thorium is not enough. To achieve a high thorium utilization in FHR, University of California, Berkeley (UCB) is developing a pebble fueled FHR (PB-FHR) with a thorium blanket (Cisneros et al., 2012). In January 2011, the Chinese Academy of Science (CAS) launched the TMSR (Thorium-based Molten-Salt Reactor nuclear energy system) project with efforts for thorium utilization both in FHR (solid fuel) and MSR (liquid fuel) (Serp et al., 2014). A whole-core model of PB-FHR was researched in UCB for uranium with varying graphite-to-heavy metal ratio (C/HM) to obtain deep burnup and sufficient negative temperature coefficient of reactivity (TCR) (François-Paul and Fabien, 2006). Since the TMSR project intends to achieve effective utilization of thorium (Jiang et al., 2012), it is necessary to study the potential utilization of thorium in FHR with a similar core geometry to UCB. Thorium and fissile materials are assumed to be homogenous mixed in each fuel kernel in a form of (X, Th) O2 where X represents fissile materials (233 U, 235 U or 239 Pu). Three different uranium based fuels in a form of (X, 238 U) O2 are also analyzed for comparison. It might not be somewhat suitable to analyze all the six type of fuels based on the UCB core and the related pebble geometries, which were just designed for the 235 U þ 238 U fuel. However, this kind of comparison under the same conditions should be able to provide an approach for the further optimization of fuel utilization in FHR. The optimization approaches of thorium fueled pebble are recommended from this analysis and the related work is being carried out. Section 2 introduces the core description and simulation software. Section 3 presents the analysis at the beginning of life (BOL). Section 4 discusses the time-dependent characteristics. Section 5 gives the conclusions. 2. Analyses methodology 2.1. Fuel types For convenience, the six fuel forms considered in this work are abbreviated as Th2/U3, Th2/U5, Th2/Pu9, U8/U3, U8/U5 and U8/ Pu9. The atomic density of fissile material and the molar ratio of fissile to fertile fuels are approximately same in each fuel type. The C/HM of thorium fuels are around 370 while those of uranium fuels are about 375. 2.2. FHR core geometry Fig. 1 shows the geometry model of FHR. The fuel pebbles are assumed to be regularly arranged in the columnar hexagonal lattices (Ilas et al., 2006) with a packing factor of 60% while the rest 40% is filled with FLiBe salt. Each pebble has a fuel mass of 11 g and contains 16,000 TRISO particles. In order to save computing time, it is assumed that the TRISO particles are uniformly distributed inside the pebble. Such fuels, embedded in graphite matrix that is stable at high temperature, allow irradiation for a long period and a deep burn to exploit fission energy. The fluoride salt density is 1.9 g/cm3 (François-Paul and Fabien, 2006) at 1050 K (assumed operation temperature), in which the enrichment of 7Li is 99.995%.
Fig. 1. Sketch of FHR.
2.3. Neutronics simulation software Existing general-purpose codes are considered to be inadequate for FHR-specific phenomenon. The development programs for HTRs have already invested in advancing methods to account for the double heterogeneity (Schultz et al., 2010), such as the SCALE code (Goluoglu and Williams, 2005). SCALE (Scale, 2011) is a comprehensive modeling and simulating tool suited for nuclear criticality and safety analysis, in which a special DOUBLEHET is selected to deal with the double heterogeneous nature and the resonance selfshielding effect (Williams et al., 2005). The verification and validation of SCALE for modeling and analysis of HTRs (HTR-10, HTTR etc.) shows a relatively good agreement between the SCALE and MCNP calculations (Ilas et al.; Ilas, 2013; Kim, 2013). Both the Advanced High Temperature Reactor at the Oak Ridge National Laboratory (Ryan Kelly, 2013) and the Fluoride-salt-cooled, Hightemperature Reactor (Allen et al., 2013) at UCB are analyzed by SCALE. Therefore, SCALE code is regarded as a suitable simulation tool at present for FHR. In this work, SCALE version 6.1 is used for describing the thorium and uranium based FHR. A 238-group ENDF/B-VII library is selected for time-dependent cross-section processing which reveals fuel composition variation during irradiation. Additionally, the depletion is performed at a constant power (1000 MWth) and 388 nuclides are tracked in trace quantities. To improve the results accuracy, the calculation steps with a smaller time are needed especially at the beginning of the simulation to build in the equilibrium concentration of major fission products. Meanwhile, increasing the magnitude of time steps as the depletion progresses is also desired to save computing time (Allen and Knight, 2010). Under the above considerations, the FHR core is depleted for almost 3 years in 16 steps. Each depletion step is scheduled to skip 50 cycles and run a total of 200 cycles with nominally 2000 neutrons per cycle. The typical computing time of one depletion step is about 15 min. 3. Neutronics characteristics of FHR core with the six fuel types 3.1. Neutron spectrum and criticality The neutron spectra of the six fuel types at BOL are shown in Fig. 2. With the similar C/HM, the spectra of different fuels reveal different inherent characteristics. There are resonance dips around 0.1 MeV energy range for each curve, corresponding to the inelastic
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Table 2 TCR of FHR cores with the six fuels at BOL. Fuel type
FTCR (pcm/K)
MTCR (pcm/K)
CTCR (pcm/K)
TCR (pcm/K)
Th2/U3 Th2/U5 Th2/Pu9 U8/U3 U8/U5 U8/Pu9
0.855 1.014 1.012 2.391 2.550 2.884
0.073 0.509 0.511 0.121 0.673 0.640
0.388 0.152 0.428 0.071 0.235 0.901
0.281 1.321 1.871 2.383 3.514 4.340
3.2. Temperature coefficient of reactivity
Fig. 2. Neutron spectra of FHR for studied fuel options at BOL.
scattering resonance of FLiBe, particularly 19F and 6Li (http:// www.nndc.bnl.gov/exfor/endf00.jsp). The resonance dips between 1 eV and 1 keV appear in each fuel option due to the resonance absorption of 232 Th or 238 U. Besides, the spectra of the 232 Th loading fuels are softer than 238 U because of the stronger resonance absorption of 238 U. The stronger capture ability of 233 U makes the spectrum of 233 U slightly harder than that of 235 U. Note that the neutron spectra of the 239 Pu loading fuels are significantly harder than the other cases, which mainly result from the larger absorption cross section of 239 Pu around 0.3 eV (http:// www.nndc.bnl.gov/exfor/endf00.jsp). These spectrum characteristics for different heavy metal fuels also exist in other types of reactors (Chaudri et al., 2013; Salvatores et al., 1997). Table 1 presents the neutronics parameters of the FHR core with the six fuel options. n denotes the average number of produced neutron per fission. Wf is for the weight gain by fission and (n, xn) of heavy metals per source neutron. Wcfertile is for the weight of capture of fertile nuclides, and Wother for the sum of the weight of capture of other nuclides (includes fissile material) and the weight of escape (includes the neutron leakage from reactor and the capture during moderation process). keff is a result of competition between neutron production and disappearance, which approximately equals the product of n and the fission weight Wf. With the same fissile material X, the Wcfertile is lower for 232 Th than 238 U while Wother is almost similar. Considering the neutron balance, the net neutron production is higher in the 232 Th based fuels which leads to a larger keff. With the same fertile nuclide, the keff of the 233 U loading fuel is largest because the h (the number of fission neutrons per neutron absorbed) of 233 U is 10e20% higher than those of 235 U and 239 Pu in thermal and epithermal spectra. In another words, at the same level of fissile enrichment, the keff at BOL is highest for Th2/U3, which leads to a deeper burnup operation and will be further discussed in Section 4.1.
Table 1 Neutronics parameters of FHR at BOL with the six fuels. Fuel type
keff
n
Wf
Wcfertile
Wother
Th2/U3 Th2/U5 Th2/Pu9 U8/U3 U8/U5 U8/Pu9
1.71 1.49 1.49 1.61 1.39 1.38
2.50 2.44 2.86 2.50 2.44 2.86
68.4% 61.0% 52.1% 64.3% 57.0% 48.2%
12.7% 13.3% 10.4% 17.7% 18.6% 16.6%
18.9% 25.7% 37.5% 18.0% 24.4% 35.2%
A negative temperature coefficient of reactivity (TCR) is necessary for reactor nuclear safety. Table 2 gives TCR, including fuel, coolant and moderator, separately corresponding to FTCR, CTCR and MTCR. FTCR mainly results from the resonance absorption of fertile nuclide, and also represents the competition between the capture reaction of fertile nuclide and the fission reaction of fissile material. The lower resolved resonance region of a nuclide, the greater value of FTCR can be obtained on account of more apparent Doppler effect (Chaudri et al., 2013). The Evaluated Nuclear Data File on the website (http://www.nndc.bnl.gov/exfor/endf00.jsp) shows that the sequence of the lowest energy boundaries of the resolved resonance energy region of the nuclides studied here is: 239 Pu 235 U < 233 U < 238 U < 232 Th. Thus, the absolute values of FTCR for the 238 U based fuels are higher than the 232 Th based ones. With the same fertile nuclide, the ordering of the FTCR values is 239 Pu > 235 U > 233 U. The irregularity of FTCR (1.014 for Th2/U5 and 1.012 for Th2/Pu9) is from the unresolved resonance disturbing. MTCR and CTCR are caused by the influence of moderator and coolant respectively on the fuel resonance absorption by spectrum shift. The MTCR value is very small because the changes of the graphite density can be ignored and the spectrum shift is very little with varying temperature (Nuttin et al., 2005). The competition between the neutron absorption and the moderation effects of FLiBe leads to small negative values and even positive values for CTCR. The farther the unresolved resonance energy region of fissile is from that of fertile, the larger difference between the fission and capture cross sections, and therefore the smaller the negative feedback of moderator and coolant will be (see the Th2/U3 case for example). The change trends of MTCR and CTCR are similar with FTCR since the spectrum shift directly affects the macroscopic cross sections of nuclides. For the Th2/U3 fuel, the TCR is somewhat too small from the viewpoint of safety, it is therefore needed to reduce the resonance fission reaction of 233 U and simultaneously increase the capture reactions of 232 Th and 233 U to improve FTCR and MTCR. Meanwhile, a harder spectrum is suggested to weaken the positive effect of FLiBe to improve CTCR. 4. Burnup characteristics of FHR Burnup behavior is one major characteristic of nuclear fuel to determine the economic feasibility and proliferation resistance. Benefited from the stability of the fuel pebbles and thorium at high temperature, a deeper burnup can be obtained for thorium fuels. Using 232 Th and 238 U as fertile fuels, the conversion ratio (CR) can be expressed as
CR ¼
Rc ð232 Th þ 234 U þ 238 U þ 240 Pu 233 PaÞ ; Ra ð233 U þ 235 U þ 239 Pu þ 241 PuÞ
(1)
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where the numerator (Rc) represents the capture reaction rate while the denominator (Ra) denotes both the capture and fission reaction rates. Compared with 239 Np with a 2.35-day half-life, the relatively long half-life (27.0 days) of 233 Pa leads to a neutron absorption to form 234 U and means that the capture reaction of 233 Pa should be subtracted in the numerator.
Table 3 Burnup comparison for studied fuel options. Fuel type
Mfertile/Mfissile at BOL (kg)
Mfertile/Mfissile at EOL (kg)
EFPD (days)
Burnup (GWd/tHM)
232 Th/233 U
4404/874 4397/882 4384/897 4417/874 4411/882 4397/897
4200/241 4216/292 4215/299 4124/299 4175/414 4223/552
725 600 630 750 550 400
137 114 119 142 104 76
232 Th/235 U 232 Th/239 Pu
4.1. Burnup analysis from neutronics aspect
238 U/233 U 238 U/235 U 238 U/239 Pu
The variations of keff and CR with burnup are shown in Fig. 3. From the top panel in Fig. 3, it can be seen that the 233 U loading fuel types reach deeper burnups while that with U8/Pu9 gets the smallest burnup. With the burnup evolution, the keff of the 239 Pu-loading fuels changes the relative order compared with that of the 235 U-loading fuels; the keff of the former decreases faster and then slower than the other fuel options. This is mainly because 239 Pu absorbs two neutrons and then generates 241 Pu which has a delayed contribution to the neutron fission fraction. Since the stronger resonance capture is beneficial to the conversion of fertile nuclide, the CR of the thorium based fuels are lower than that of the 238 U based ones which have a larger resonance energy portion (bottom in Fig. 3). In this sense, the moderator needs to be reduced to improve the CR of Th2/U3. Table 3 shows the mass loading of fertile and fissile (includes 233 U, 235 U, 239 Pu and 241 Pu), the effective full power days (EFPD) and burnup. The variation of fertile per day can be considered as the consumption of the bred fissile material. Under the same operation power, more fissile material (contains the initial loading and bred) is consumed per day for the fuel types with a larger capture-tofission ratio. As shown in Table 3 (the forth column), the lowest fissile is burnt for Th2/U3 (1.15 kg/day) and the highest fissile is consumed for U8/Pu9 (1.30 kg/day). The 233 U loading fuels reach deeper burnups both for Th2/U3 (137 GWd/tHM) and U8/U3 (142 GWd/tHM) than the 235 U and 239 Pu loading fuels. To understand the above differences, the absorption estimates of the six fuel options are shown in Fig. 4. The notation PRO
represents the regenerated neutrons except the one used to maintain fission reaction while DIS means the neutron disappearance including the capture and escape probabilities. The difference between PRO and DIS is considered as the exceeded neutrons. Therefore, the higher the difference, the larger the initial keff (second column in Table 1), the more neutrons left for the next fission reaction (i.e. Wf in Table 1), and then the deeper burnup will be obtained (last column in Table 3), and vice versa. Furthermore, the higher difference means that less neutrons are captured by fertile manifested as smaller CR (bottom in Fig. 3). With the smaller CR, the less bred fissile materials used for the supplement for fission reaction has an influence on burnup. This also explains why U8/U3 with a smaller difference reaches a slight higher burnup than Th2/ U3. Considering the deeper burnup for Th2/U3, increasing molar ratio of 232 Th and 233 U is suggested to reduce a bit burnup but improve TCR and CR. Fig. 5 displays the neutron capture proportions of different materials for Th2/U3 and Th2/Pu9. Among them, the capture of actinides occupies a large proportion, especially for Th2/Pu9 because of the larger capture cross section of 239 Pu. The capture ratio of FLiBe in Th2/U3 becomes larger and increases faster with the burnup evolution while that of Th2/Pu9 keeps stable. Along with the deeper burnup, the neutron spectrum becomes softer due to the consumption of fissile fuel, which leads to an enhancement of the mean capture cross section of FLiBe. Th2/Pu9 displays a lower
Fig. 3. Variation of keff (top) and CR (bottom) with burnup for the six fuel options.
Fig. 4. Neutron absorption estimates for thorium fuels (top) and uranium fuels (bottom).
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Fig. 6. The radio-toxicities of the six fuels after shutdown.
Fig. 5. Capture ratio of multi-material variation with time for Th2/U3 (top) and Th2/ Pu9 (bottom).
burnup due to its smaller initial difference between neutron production and disappearance, although it has a slow capture growth due to the less softening of neutron spectrum. Since the situations of the other four fuel options are similar to Th2/U3 (such as Th2/U5) or to Th2/Pu9 (such as U8/U3, U8/U5 and U8/Pu9), the results are not presented here.
thorium. The radio-toxicity of thorium in Th2/U3 is almost similar with Th2/U5 and Th2/Pu9, however the corresponding radiotoxicity proportion of thorium in Th2/U3 is much higher than those in Th2/U5 and Th2/Pu9 due to the much higher radio-toxicity of plutonium. Additionally, the radio-toxicities of U8/Pu9 and Th2/ Pu9 are very similar due to the main contribution of 239 Pu even though their burnup difference is very large. The same situation happens in the cases of U8/U3 and U8/U5. In contrast, the radiotoxicity of Th2/U3 is more than one order of magnitude lower than that of Th2/U5 despite of their almost similar burnups. The radio-toxicity of spent nuclear fuel is therefore determined by both fuel type and discharged burnup.
4.2. Radio-toxicity
5. Conclusions
One of the motivations for thorium utilization is its potential reductions of transuranic (TRU) waste generation and of radiotoxicity level. The radio-toxicity R of nuclides i with inventories Ni and decay constants li evolves as detailed in Eq. (2) (Nuttin et al., 2005). Dose coefficients ri are given by the International Commission on Radiological Protection (Compilation of ingestion, 1995). Calculations of waste radio-toxicity R(t) are done using the code ORIGEN-s.
Thorium utilization in FHR is investigated by using a full core model. Neutronics behavior is analyzed in terms of keff, neutron spectrum, temperature coefficient of reactivity and burnup. Conclusions drawn from this study are as follows:
RðtÞ ¼
X
ri li Ni ðtÞ:
(2)
i
The radio-toxicities of the six fuel options after shutdown are compared in Fig. 6. The Th2/U3 fuel with the deepest burnup shows the lowest radio-toxicity up to cooling time of 1000 years whereas the U8/Pu9 fuel with the least burnup shows the highest radiotoxicity. At cooling times above 10,000 years, the differences among the six fuel options become smaller. It is important to note that the radio-toxicity of fuels involving 232 Th or 233 U presents an upward trend at cooling time of around 1000 years due to the increasing uranium radio-toxicity (mainly 234 U). The radio-toxicity proportions of different nuclear elements at shutdown as a function of cooling time are shown in Table 4. The value of zero corresponds to the radio-toxicity less than 0.1%. The radio-toxicities for the fuels involving 239 Pu, 238 U or 235 U are firstly from plutonium, and secondly from uranium (such as Th2/U5 and U8/U5) or americium (such as Th2/Pu9, U8/U3 and U8/Pu9). The radio-toxicity of the Th2/U3 fuel chiefly stems from uranium and
The keff of Th2/U3 is about 1.1e1.2 times the other fuel types with the same inventory of fissile and fertile at the beginning of life. Compared with the other fuel types especially U8/U5 and U8/Pu9, Th2/U3 obtains a relatively deeper burnup. Introducing fertile 232 Th and fissile 233 U alters the TCR of relevance to safe operation of FHR to a minor degree with small negative values (0.281 for the Th2/U3 fuel). The coolant TCR and the moderator TCR tend to be positive when 233 U is the dominant fissile component. Further optimization of the core geometry and the fuel pebble composition should be carried out to increase the absolute value of negative TCR.
Table 4 The radio-toxicity proportions of different parent nuclides for the six fuel options at reactor shutdown. Fuel type
Th (%)
U (%)
Np (%)
Pu (%)
Am (%)
Th2/U3 Th2/U5 Th2/Pu9 U8/U3 U8/U5 U8/Pu9
28.6 1.4 0.1 0.0 0.0 0.0
61.8 7.2 0.4 1.0 2.2 0.0
0.3 2.0 0.0 0.2 0.8 0.0
9.3 89.4 95.5 95.8 95.4 97.8
0.0 0.0 4.0 3.0 1.6 2.2
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The radio-toxicity of spent nuclear fuel depends, to a great extent, on the fuel type rather than the discharged burnup. The radio-toxicity of the Th2/U3 fuel is almost two orders of magnitude lower than those fuels involving 239 Pu due to the lower minor actinides produced from 232 Th and 233 U. Moreover, the radio-toxicity of Th2/U3 is more than one times that of U8/ U3 even though these two fuel types have the almost similar burnups. It can be concluded that the thorium based fuel, especially Th2/ U3 may be a best option for FHR. Further optimization, however, is needed for the molar ratio of 232 Th and 233 U and C/HM (i.e. neutron spectrum) to obtain the balance between TCR, CR and burnup. The thorium utilization is expected to be further improved by considering elaborate fuel pebble arrangement as well as fuel pebble circulating.
Acknowledgments This work is supported by the Chinese TMSR Strategic Pioneer Science and Technology Project under Grant No. XDA02010000 and the National Natural Science Foundation of China under Grant No. 91326201.
References Allen, K., Knight, T., 2010. Destruction rate analysis of transuranic targets in sodiumcooled fast reactor (SFR) assemblies using MCNPX and SCALE 6.0. Prog. Nucl. Energy 52 (4), 387. Allen, T., et al., 2013. Fluoride-salt-cooled, High-temperature Reactor (FHR) Methods and Experiments Program White Paper. Baumer, R., Kalinowski, I., 1991. THTR commissioning and operating experience. Energy 16 (1), 59. Chaudri, K.S., et al., 2013. Coupled neutronics/thermal hydraulics evaluation for thorium based fuels in thermal spectrum SCWR. Prog. Nucl. Energy 68, 55. Cisneros, A.T., et al., 2012. Pebble fuel design for the PB-FHR. In: Proceedings of the International Congress on Advances in Nuclear Power Plants (ICAPP12). Compilation of Ingestion and Inhalation Dose Coefficients, vol. 76, 1995. International Commission on Radiological Protection, Publication. Cycle, T.F., Potential Benefits and Challenges, International Atomic Energy Agency (IAEA) Technical Document 1450. David, S., 2005. Future scenarios for fission based reactors. Nucl. Phys. A 751, 429. Ding, M., Kloosterman, J.L., 2014. Thorium utilization in a small long-life HTR. part I: Th/U MOX fuel blocks. Nucl. Eng. Des. 267, 238. François-Paul, F.E., Fabien, Koenig, E., 2006. Preliminary Study of the Pebble-bed Advanced High Temperature Reactor. Master's thesis.
Fratoni, M., 2008. Development and Applications of Methodologies for the Neutronic Design of the Pebble Bed Advanced High Temperature Reactor (PBAHTR). ProQuest. García, C., et al., 2013. Evaluation of uranium thorium and plutonium thorium fuel cycles in a very high temperature hybrid system. Prog. Nucl. Energy 66, 61. Goluoglu, S., Williams, M.L., 2005. Modeling doubly heterogeneous systems in scale. Trans. Am. Nucl. Soc. 93, 963. Habush, A., Harris, A., 1968. 330-MW(e) fort st. vrain high-temperature gas-cooled reactor. Nucl. Eng. Des. 7 (4), 312. http://www.nndc.bnl.gov/exfor/endf00.jsp. Ilas, G., et al., 2006. On the few-group cross-section generation methodology for PBR analysis. Ann. Nucl. Energy 33, 1058. Ilas, G., et al., Validation of Scale for High Temperature Gas-cooled Reactors Analysis, NUREG/CR-7107, U.S. Nuclear Regulatory Commission http://dx.doi.org/10. 2172/1048709. Ilas, D., 2013. Scale code validation for prismatic high-temperature gas-cooled reactors. Nucl. Tech. 183 (3), 379. Ingersoll, D., et al., 2004. Status of Preconceptual Design of the Advanced Hightemperature Reactor (AHTR). Department of Energy, United States. Jiang, M., Xu, H., Dai, Z., 2012. Advanced fission energy program-TMSR nuclear energy system. Bull. Chin. Acad. Sci. 27, 366. Kim, J., 2013. Assessment of double heterogeneity treatment capability in scale. In: 22nd International Conference Nuclear Energy for New Europe. Lung, M., 1997. A Present Review of the Thorium Nuclear Fuel Cycles. Tech. rep.. Euratom. Mathieu, L., et al., 2006. The thorium molten salt reactor: moving on from the MSBR. Prog. Nucl. Energy 48 (7), 664. Merle-Lucotte, et al., 2008. Optimization and simplification of the concept of nonmoderated thorium molten salt reactor. In: Proceedings of International Conference on the Physics of Reactors, Nuclear Power: a Sustainable Resource. Nuttin, A., et al., 2005. Potential of thorium molten salt reactors detailed calculations and concept evolution with a view to large scale energy production. Prog. Nucl. Energy 46 (1), 77. Pohl, P., 2006. The Importance of the AVR Pebble-bed Reactor for the Future of Nuclear Power. Tech. rep.. American Nuclear Society, United States. Puill, A., 1999. Thorium Utilization in PWRs, IAEA-TECDOC-1319, p. 185. Ramanna, R., Lee, S., 1986. The thorium cycle for fast breeder reactors. Pramana 27, 129. Ryan Kelly, D.I., 2013. Verification of a depletion method in SCALE for the advanced high temperature reactor. Nucl. Technol. 183 (3), 391. Salvatores, M., et al., 1997. The potential of accelerator-driven systems for transmutation or power production using thorium or uranium fuel cycles. Nucl. Sci. Eng. 126, 333. Scale: a Comprehensive Modeling and Simulation Suite for Nuclear Safety Analysis and Design, ORNL/TM-2005/39, June 2011. Oak Ridge National Laboratory. Schultz, R.R., et al., 2010. Next Generation Nuclear Plant Methods Technical Program Plan, INL/EXT-06e11804. Idaho National Laboratory. Serp, J., et al., 2014. The molten salt reactor (MSR) in generation IV: overview and perspectives. Prog. Nucl. Energy 77, 308. U. M. Facilitators, M. Facilitators, 2013. Fluoride-Salt-Cooled High Temperature Reactor (FHR) Materials. Fuels and Components White Paper. Wickleder, M.S., et al., 2006. Thorium. In: The Chemistry of the Actinide and Transactinide Elements. Springer, p. 52. Williams, M., et al., 2005. Recent enhancements to the SCALE 5 resonance selfshielding methodology. ANS Trans. 92, 751.
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Analysis of thorium and uranium based nuclear fuel options in Fluoride salt-cooled High-temperature Reactor X.X. Li a, b, c, X.Z. Cai a, b, c, D.Z. Jiang a, b, c, Y.W. Ma a, b, c, J.F. Huang a, b, c, C.Y. Zou a, b, c, C.G. Yu a, b, c, J.L. Han a, b, c, J.G. Chen a, b, c, * a b c
Shanghai Institute of Applied Physics, Chinese Academy of Sciences, Shanghai 201800, China CAS Center for Excellence in TMSR Energy System, Chinese Academy of Sciences, Shanghai 201800, China Key Laboratory of Nuclear Radiation and Nuclear Energy Technology, Chinese Academy of Sciences, Shanghai 201800, China
a r t i c l e i n f o
a b s t r a c t
Article history: Received 4 July 2014 Received in revised form 22 September 2014 Accepted 9 October 2014 Available online 1 November 2014
In order to utilize thorium in Fluoride salt-cooled High-temperature Reactor (FHR), neutronics analysis for thorium based fuels (233 U þ Th, 235 U þ Th and 239 Pu þ Th) is carried out in a whole-core model of pebble bed FHR. Uranium (238 U) based fuels with the above three fissile nuclides are also analyzed for comparison. The atomic density of fissile material is kept constant at start-up for the six fuel types. Neutron characteristics including neutron spectrum, effective multiplicity factor (keff), temperature coefficient of reactivity (TCR), conversion ratio (CR) and burnup for the six fuel options are discussed. With the same fissile nuclide, the thorium based fuels have a higher initial keff than the uranium based fuels due to the smaller resonance absorption of 232 Th. As for fissile material, 233 U is the best candidate as a driver fuel in thermal and epithermal spectra due to its effective number of fission neutrons (h). Besides, the 232 Th/233 U fuel can be extended to radioactive waste management benefited from deeper burnup and lower level of radio-toxicity. The FHR core with thermal and epithermal spectra is more suitable for TheU fuel cycle than UePu fuel cycle. These analyses can provide an approach for the further optimizations of thorium utilization in FHR. © 2014 Elsevier Ltd. All rights reserved.
Keywords: Thorium Neutronics FHR
1. Introduction Most of present commercial reactors using enriched uranium (about 5%) with once-through fuel cycle will consume up the whole world's estimated uranium in a few decades along with the growing demand of energy (David, 2005). It is therefore important to develop advanced reactors with high efficiency of fuel utilization and corresponding fuel cycles (includes fuel types and fuel cycle modes) to solve the problem of nuclear fuel shortage. As an alternative nuclear fuel resource, thorium attracts more and more attentions to ensure sustainable energy generation. Thorium is a fertile fuel which is reported to be 3 ~ 4 times as abundant in the earth's crust as uranium (Wickleder et al., 2006). It will greatly expand the nuclear fuel resources through the conversion behavior from 232 Th to 233 U by a neutron capture. TheU fuel cycle offers attractive features (Cycle, 1450), including better breeding capability in thermal reactors and lower radio-toxicity levels in nuclear waste.
* Corresponding author. Shanghai Institute of Applied Physics, Chinese Academy of Sciences, Shanghai 201800, China. Tel.: þ86 21 39194027. E-mail address: [email protected] (J.G. Chen). http://dx.doi.org/10.1016/j.pnucene.2014.10.004 0149-1970/© 2014 Elsevier Ltd. All rights reserved.
Many countries and organizations have already started the research of strategy and engineering for thorium utilization since mid 1950s (Mathieu et al., 2006; García et al., 2013; Nuttin et al., 2005). The majority of options for the use of thorium previously proposed have relied on highly innovative reactor concepts (Lung, 1997), such as Molten Salt Reactor (MSR) (Merle-Lucotte et al., 2008), Liquid Metal Fast Breeder Reactor (LMFBR) (Ramanna and Lee, 1986), Super Critical Water Reactor (SCWR) (Chaudri et al., 2013) and Accelerator Driven Systems (ADS) (Salvatores et al., 1997). And the studies of thorium utilization with existing reactors, in most cases PWR specific, were also summarized (Puill, 1999). To take the advantages of thorium, especially its great stability at high temperature and breeding capacity in thermal and epithermal spectra, thorium based fuels were tested in High-temperature gascooled Reactors (HTRs). Most of the researches and operating experiences of thorium utilization in HTR are focused on high enriched uranium (HEU) (Baumer and Kalinowski, 1991; Habush and Harris, 1968) or low enriched uranium (LEU) (Pohl, 2006; Ding and Kloosterman, 2014). Fluoride salt-cooled High-temperature Reactor (FHR) (Ingersoll et al., 2004) synthesizes the advantages of MSR and HTR, which
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has similar neutronics with HTR, fueled with TRISO particle, differentiated only by its coolant with fluoride salt instead of helium gas. FHR can support conventional LEU fuel cycle and also advanced fuel options, such as thorium. The study on the neutronics performance of FHR has been concentrated on uranium (François-Paul and Fabien, 2006; Fratoni, 2008; U. M. Facilitators and M. Facilitators, 2013), while that of thorium is not enough. To achieve a high thorium utilization in FHR, University of California, Berkeley (UCB) is developing a pebble fueled FHR (PB-FHR) with a thorium blanket (Cisneros et al., 2012). In January 2011, the Chinese Academy of Science (CAS) launched the TMSR (Thorium-based Molten-Salt Reactor nuclear energy system) project with efforts for thorium utilization both in FHR (solid fuel) and MSR (liquid fuel) (Serp et al., 2014). A whole-core model of PB-FHR was researched in UCB for uranium with varying graphite-to-heavy metal ratio (C/HM) to obtain deep burnup and sufficient negative temperature coefficient of reactivity (TCR) (François-Paul and Fabien, 2006). Since the TMSR project intends to achieve effective utilization of thorium (Jiang et al., 2012), it is necessary to study the potential utilization of thorium in FHR with a similar core geometry to UCB. Thorium and fissile materials are assumed to be homogenous mixed in each fuel kernel in a form of (X, Th) O2 where X represents fissile materials (233 U, 235 U or 239 Pu). Three different uranium based fuels in a form of (X, 238 U) O2 are also analyzed for comparison. It might not be somewhat suitable to analyze all the six type of fuels based on the UCB core and the related pebble geometries, which were just designed for the 235 U þ 238 U fuel. However, this kind of comparison under the same conditions should be able to provide an approach for the further optimization of fuel utilization in FHR. The optimization approaches of thorium fueled pebble are recommended from this analysis and the related work is being carried out. Section 2 introduces the core description and simulation software. Section 3 presents the analysis at the beginning of life (BOL). Section 4 discusses the time-dependent characteristics. Section 5 gives the conclusions. 2. Analyses methodology 2.1. Fuel types For convenience, the six fuel forms considered in this work are abbreviated as Th2/U3, Th2/U5, Th2/Pu9, U8/U3, U8/U5 and U8/ Pu9. The atomic density of fissile material and the molar ratio of fissile to fertile fuels are approximately same in each fuel type. The C/HM of thorium fuels are around 370 while those of uranium fuels are about 375. 2.2. FHR core geometry Fig. 1 shows the geometry model of FHR. The fuel pebbles are assumed to be regularly arranged in the columnar hexagonal lattices (Ilas et al., 2006) with a packing factor of 60% while the rest 40% is filled with FLiBe salt. Each pebble has a fuel mass of 11 g and contains 16,000 TRISO particles. In order to save computing time, it is assumed that the TRISO particles are uniformly distributed inside the pebble. Such fuels, embedded in graphite matrix that is stable at high temperature, allow irradiation for a long period and a deep burn to exploit fission energy. The fluoride salt density is 1.9 g/cm3 (François-Paul and Fabien, 2006) at 1050 K (assumed operation temperature), in which the enrichment of 7Li is 99.995%.
Fig. 1. Sketch of FHR.
2.3. Neutronics simulation software Existing general-purpose codes are considered to be inadequate for FHR-specific phenomenon. The development programs for HTRs have already invested in advancing methods to account for the double heterogeneity (Schultz et al., 2010), such as the SCALE code (Goluoglu and Williams, 2005). SCALE (Scale, 2011) is a comprehensive modeling and simulating tool suited for nuclear criticality and safety analysis, in which a special DOUBLEHET is selected to deal with the double heterogeneous nature and the resonance selfshielding effect (Williams et al., 2005). The verification and validation of SCALE for modeling and analysis of HTRs (HTR-10, HTTR etc.) shows a relatively good agreement between the SCALE and MCNP calculations (Ilas et al.; Ilas, 2013; Kim, 2013). Both the Advanced High Temperature Reactor at the Oak Ridge National Laboratory (Ryan Kelly, 2013) and the Fluoride-salt-cooled, Hightemperature Reactor (Allen et al., 2013) at UCB are analyzed by SCALE. Therefore, SCALE code is regarded as a suitable simulation tool at present for FHR. In this work, SCALE version 6.1 is used for describing the thorium and uranium based FHR. A 238-group ENDF/B-VII library is selected for time-dependent cross-section processing which reveals fuel composition variation during irradiation. Additionally, the depletion is performed at a constant power (1000 MWth) and 388 nuclides are tracked in trace quantities. To improve the results accuracy, the calculation steps with a smaller time are needed especially at the beginning of the simulation to build in the equilibrium concentration of major fission products. Meanwhile, increasing the magnitude of time steps as the depletion progresses is also desired to save computing time (Allen and Knight, 2010). Under the above considerations, the FHR core is depleted for almost 3 years in 16 steps. Each depletion step is scheduled to skip 50 cycles and run a total of 200 cycles with nominally 2000 neutrons per cycle. The typical computing time of one depletion step is about 15 min. 3. Neutronics characteristics of FHR core with the six fuel types 3.1. Neutron spectrum and criticality The neutron spectra of the six fuel types at BOL are shown in Fig. 2. With the similar C/HM, the spectra of different fuels reveal different inherent characteristics. There are resonance dips around 0.1 MeV energy range for each curve, corresponding to the inelastic
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Table 2 TCR of FHR cores with the six fuels at BOL. Fuel type
FTCR (pcm/K)
MTCR (pcm/K)
CTCR (pcm/K)
TCR (pcm/K)
Th2/U3 Th2/U5 Th2/Pu9 U8/U3 U8/U5 U8/Pu9
0.855 1.014 1.012 2.391 2.550 2.884
0.073 0.509 0.511 0.121 0.673 0.640
0.388 0.152 0.428 0.071 0.235 0.901
0.281 1.321 1.871 2.383 3.514 4.340
3.2. Temperature coefficient of reactivity
Fig. 2. Neutron spectra of FHR for studied fuel options at BOL.
scattering resonance of FLiBe, particularly 19F and 6Li (http:// www.nndc.bnl.gov/exfor/endf00.jsp). The resonance dips between 1 eV and 1 keV appear in each fuel option due to the resonance absorption of 232 Th or 238 U. Besides, the spectra of the 232 Th loading fuels are softer than 238 U because of the stronger resonance absorption of 238 U. The stronger capture ability of 233 U makes the spectrum of 233 U slightly harder than that of 235 U. Note that the neutron spectra of the 239 Pu loading fuels are significantly harder than the other cases, which mainly result from the larger absorption cross section of 239 Pu around 0.3 eV (http:// www.nndc.bnl.gov/exfor/endf00.jsp). These spectrum characteristics for different heavy metal fuels also exist in other types of reactors (Chaudri et al., 2013; Salvatores et al., 1997). Table 1 presents the neutronics parameters of the FHR core with the six fuel options. n denotes the average number of produced neutron per fission. Wf is for the weight gain by fission and (n, xn) of heavy metals per source neutron. Wcfertile is for the weight of capture of fertile nuclides, and Wother for the sum of the weight of capture of other nuclides (includes fissile material) and the weight of escape (includes the neutron leakage from reactor and the capture during moderation process). keff is a result of competition between neutron production and disappearance, which approximately equals the product of n and the fission weight Wf. With the same fissile material X, the Wcfertile is lower for 232 Th than 238 U while Wother is almost similar. Considering the neutron balance, the net neutron production is higher in the 232 Th based fuels which leads to a larger keff. With the same fertile nuclide, the keff of the 233 U loading fuel is largest because the h (the number of fission neutrons per neutron absorbed) of 233 U is 10e20% higher than those of 235 U and 239 Pu in thermal and epithermal spectra. In another words, at the same level of fissile enrichment, the keff at BOL is highest for Th2/U3, which leads to a deeper burnup operation and will be further discussed in Section 4.1.
Table 1 Neutronics parameters of FHR at BOL with the six fuels. Fuel type
keff
n
Wf
Wcfertile
Wother
Th2/U3 Th2/U5 Th2/Pu9 U8/U3 U8/U5 U8/Pu9
1.71 1.49 1.49 1.61 1.39 1.38
2.50 2.44 2.86 2.50 2.44 2.86
68.4% 61.0% 52.1% 64.3% 57.0% 48.2%
12.7% 13.3% 10.4% 17.7% 18.6% 16.6%
18.9% 25.7% 37.5% 18.0% 24.4% 35.2%
A negative temperature coefficient of reactivity (TCR) is necessary for reactor nuclear safety. Table 2 gives TCR, including fuel, coolant and moderator, separately corresponding to FTCR, CTCR and MTCR. FTCR mainly results from the resonance absorption of fertile nuclide, and also represents the competition between the capture reaction of fertile nuclide and the fission reaction of fissile material. The lower resolved resonance region of a nuclide, the greater value of FTCR can be obtained on account of more apparent Doppler effect (Chaudri et al., 2013). The Evaluated Nuclear Data File on the website (http://www.nndc.bnl.gov/exfor/endf00.jsp) shows that the sequence of the lowest energy boundaries of the resolved resonance energy region of the nuclides studied here is: 239 Pu 235 U < 233 U < 238 U < 232 Th. Thus, the absolute values of FTCR for the 238 U based fuels are higher than the 232 Th based ones. With the same fertile nuclide, the ordering of the FTCR values is 239 Pu > 235 U > 233 U. The irregularity of FTCR (1.014 for Th2/U5 and 1.012 for Th2/Pu9) is from the unresolved resonance disturbing. MTCR and CTCR are caused by the influence of moderator and coolant respectively on the fuel resonance absorption by spectrum shift. The MTCR value is very small because the changes of the graphite density can be ignored and the spectrum shift is very little with varying temperature (Nuttin et al., 2005). The competition between the neutron absorption and the moderation effects of FLiBe leads to small negative values and even positive values for CTCR. The farther the unresolved resonance energy region of fissile is from that of fertile, the larger difference between the fission and capture cross sections, and therefore the smaller the negative feedback of moderator and coolant will be (see the Th2/U3 case for example). The change trends of MTCR and CTCR are similar with FTCR since the spectrum shift directly affects the macroscopic cross sections of nuclides. For the Th2/U3 fuel, the TCR is somewhat too small from the viewpoint of safety, it is therefore needed to reduce the resonance fission reaction of 233 U and simultaneously increase the capture reactions of 232 Th and 233 U to improve FTCR and MTCR. Meanwhile, a harder spectrum is suggested to weaken the positive effect of FLiBe to improve CTCR. 4. Burnup characteristics of FHR Burnup behavior is one major characteristic of nuclear fuel to determine the economic feasibility and proliferation resistance. Benefited from the stability of the fuel pebbles and thorium at high temperature, a deeper burnup can be obtained for thorium fuels. Using 232 Th and 238 U as fertile fuels, the conversion ratio (CR) can be expressed as
CR ¼
Rc ð232 Th þ 234 U þ 238 U þ 240 Pu 233 PaÞ ; Ra ð233 U þ 235 U þ 239 Pu þ 241 PuÞ
(1)
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where the numerator (Rc) represents the capture reaction rate while the denominator (Ra) denotes both the capture and fission reaction rates. Compared with 239 Np with a 2.35-day half-life, the relatively long half-life (27.0 days) of 233 Pa leads to a neutron absorption to form 234 U and means that the capture reaction of 233 Pa should be subtracted in the numerator.
Table 3 Burnup comparison for studied fuel options. Fuel type
Mfertile/Mfissile at BOL (kg)
Mfertile/Mfissile at EOL (kg)
EFPD (days)
Burnup (GWd/tHM)
232 Th/233 U
4404/874 4397/882 4384/897 4417/874 4411/882 4397/897
4200/241 4216/292 4215/299 4124/299 4175/414 4223/552
725 600 630 750 550 400
137 114 119 142 104 76
232 Th/235 U 232 Th/239 Pu
4.1. Burnup analysis from neutronics aspect
238 U/233 U 238 U/235 U 238 U/239 Pu
The variations of keff and CR with burnup are shown in Fig. 3. From the top panel in Fig. 3, it can be seen that the 233 U loading fuel types reach deeper burnups while that with U8/Pu9 gets the smallest burnup. With the burnup evolution, the keff of the 239 Pu-loading fuels changes the relative order compared with that of the 235 U-loading fuels; the keff of the former decreases faster and then slower than the other fuel options. This is mainly because 239 Pu absorbs two neutrons and then generates 241 Pu which has a delayed contribution to the neutron fission fraction. Since the stronger resonance capture is beneficial to the conversion of fertile nuclide, the CR of the thorium based fuels are lower than that of the 238 U based ones which have a larger resonance energy portion (bottom in Fig. 3). In this sense, the moderator needs to be reduced to improve the CR of Th2/U3. Table 3 shows the mass loading of fertile and fissile (includes 233 U, 235 U, 239 Pu and 241 Pu), the effective full power days (EFPD) and burnup. The variation of fertile per day can be considered as the consumption of the bred fissile material. Under the same operation power, more fissile material (contains the initial loading and bred) is consumed per day for the fuel types with a larger capture-tofission ratio. As shown in Table 3 (the forth column), the lowest fissile is burnt for Th2/U3 (1.15 kg/day) and the highest fissile is consumed for U8/Pu9 (1.30 kg/day). The 233 U loading fuels reach deeper burnups both for Th2/U3 (137 GWd/tHM) and U8/U3 (142 GWd/tHM) than the 235 U and 239 Pu loading fuels. To understand the above differences, the absorption estimates of the six fuel options are shown in Fig. 4. The notation PRO
represents the regenerated neutrons except the one used to maintain fission reaction while DIS means the neutron disappearance including the capture and escape probabilities. The difference between PRO and DIS is considered as the exceeded neutrons. Therefore, the higher the difference, the larger the initial keff (second column in Table 1), the more neutrons left for the next fission reaction (i.e. Wf in Table 1), and then the deeper burnup will be obtained (last column in Table 3), and vice versa. Furthermore, the higher difference means that less neutrons are captured by fertile manifested as smaller CR (bottom in Fig. 3). With the smaller CR, the less bred fissile materials used for the supplement for fission reaction has an influence on burnup. This also explains why U8/U3 with a smaller difference reaches a slight higher burnup than Th2/ U3. Considering the deeper burnup for Th2/U3, increasing molar ratio of 232 Th and 233 U is suggested to reduce a bit burnup but improve TCR and CR. Fig. 5 displays the neutron capture proportions of different materials for Th2/U3 and Th2/Pu9. Among them, the capture of actinides occupies a large proportion, especially for Th2/Pu9 because of the larger capture cross section of 239 Pu. The capture ratio of FLiBe in Th2/U3 becomes larger and increases faster with the burnup evolution while that of Th2/Pu9 keeps stable. Along with the deeper burnup, the neutron spectrum becomes softer due to the consumption of fissile fuel, which leads to an enhancement of the mean capture cross section of FLiBe. Th2/Pu9 displays a lower
Fig. 3. Variation of keff (top) and CR (bottom) with burnup for the six fuel options.
Fig. 4. Neutron absorption estimates for thorium fuels (top) and uranium fuels (bottom).
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Fig. 6. The radio-toxicities of the six fuels after shutdown.
Fig. 5. Capture ratio of multi-material variation with time for Th2/U3 (top) and Th2/ Pu9 (bottom).
burnup due to its smaller initial difference between neutron production and disappearance, although it has a slow capture growth due to the less softening of neutron spectrum. Since the situations of the other four fuel options are similar to Th2/U3 (such as Th2/U5) or to Th2/Pu9 (such as U8/U3, U8/U5 and U8/Pu9), the results are not presented here.
thorium. The radio-toxicity of thorium in Th2/U3 is almost similar with Th2/U5 and Th2/Pu9, however the corresponding radiotoxicity proportion of thorium in Th2/U3 is much higher than those in Th2/U5 and Th2/Pu9 due to the much higher radio-toxicity of plutonium. Additionally, the radio-toxicities of U8/Pu9 and Th2/ Pu9 are very similar due to the main contribution of 239 Pu even though their burnup difference is very large. The same situation happens in the cases of U8/U3 and U8/U5. In contrast, the radiotoxicity of Th2/U3 is more than one order of magnitude lower than that of Th2/U5 despite of their almost similar burnups. The radio-toxicity of spent nuclear fuel is therefore determined by both fuel type and discharged burnup.
4.2. Radio-toxicity
5. Conclusions
One of the motivations for thorium utilization is its potential reductions of transuranic (TRU) waste generation and of radiotoxicity level. The radio-toxicity R of nuclides i with inventories Ni and decay constants li evolves as detailed in Eq. (2) (Nuttin et al., 2005). Dose coefficients ri are given by the International Commission on Radiological Protection (Compilation of ingestion, 1995). Calculations of waste radio-toxicity R(t) are done using the code ORIGEN-s.
Thorium utilization in FHR is investigated by using a full core model. Neutronics behavior is analyzed in terms of keff, neutron spectrum, temperature coefficient of reactivity and burnup. Conclusions drawn from this study are as follows:
RðtÞ ¼
X
ri li Ni ðtÞ:
(2)
i
The radio-toxicities of the six fuel options after shutdown are compared in Fig. 6. The Th2/U3 fuel with the deepest burnup shows the lowest radio-toxicity up to cooling time of 1000 years whereas the U8/Pu9 fuel with the least burnup shows the highest radiotoxicity. At cooling times above 10,000 years, the differences among the six fuel options become smaller. It is important to note that the radio-toxicity of fuels involving 232 Th or 233 U presents an upward trend at cooling time of around 1000 years due to the increasing uranium radio-toxicity (mainly 234 U). The radio-toxicity proportions of different nuclear elements at shutdown as a function of cooling time are shown in Table 4. The value of zero corresponds to the radio-toxicity less than 0.1%. The radio-toxicities for the fuels involving 239 Pu, 238 U or 235 U are firstly from plutonium, and secondly from uranium (such as Th2/U5 and U8/U5) or americium (such as Th2/Pu9, U8/U3 and U8/Pu9). The radio-toxicity of the Th2/U3 fuel chiefly stems from uranium and
The keff of Th2/U3 is about 1.1e1.2 times the other fuel types with the same inventory of fissile and fertile at the beginning of life. Compared with the other fuel types especially U8/U5 and U8/Pu9, Th2/U3 obtains a relatively deeper burnup. Introducing fertile 232 Th and fissile 233 U alters the TCR of relevance to safe operation of FHR to a minor degree with small negative values (0.281 for the Th2/U3 fuel). The coolant TCR and the moderator TCR tend to be positive when 233 U is the dominant fissile component. Further optimization of the core geometry and the fuel pebble composition should be carried out to increase the absolute value of negative TCR.
Table 4 The radio-toxicity proportions of different parent nuclides for the six fuel options at reactor shutdown. Fuel type
Th (%)
U (%)
Np (%)
Pu (%)
Am (%)
Th2/U3 Th2/U5 Th2/Pu9 U8/U3 U8/U5 U8/Pu9
28.6 1.4 0.1 0.0 0.0 0.0
61.8 7.2 0.4 1.0 2.2 0.0
0.3 2.0 0.0 0.2 0.8 0.0
9.3 89.4 95.5 95.8 95.4 97.8
0.0 0.0 4.0 3.0 1.6 2.2
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The radio-toxicity of spent nuclear fuel depends, to a great extent, on the fuel type rather than the discharged burnup. The radio-toxicity of the Th2/U3 fuel is almost two orders of magnitude lower than those fuels involving 239 Pu due to the lower minor actinides produced from 232 Th and 233 U. Moreover, the radio-toxicity of Th2/U3 is more than one times that of U8/ U3 even though these two fuel types have the almost similar burnups. It can be concluded that the thorium based fuel, especially Th2/ U3 may be a best option for FHR. Further optimization, however, is needed for the molar ratio of 232 Th and 233 U and C/HM (i.e. neutron spectrum) to obtain the balance between TCR, CR and burnup. The thorium utilization is expected to be further improved by considering elaborate fuel pebble arrangement as well as fuel pebble circulating.
Acknowledgments This work is supported by the Chinese TMSR Strategic Pioneer Science and Technology Project under Grant No. XDA02010000 and the National Natural Science Foundation of China under Grant No. 91326201.
References Allen, K., Knight, T., 2010. Destruction rate analysis of transuranic targets in sodiumcooled fast reactor (SFR) assemblies using MCNPX and SCALE 6.0. Prog. Nucl. Energy 52 (4), 387. Allen, T., et al., 2013. Fluoride-salt-cooled, High-temperature Reactor (FHR) Methods and Experiments Program White Paper. Baumer, R., Kalinowski, I., 1991. THTR commissioning and operating experience. Energy 16 (1), 59. Chaudri, K.S., et al., 2013. Coupled neutronics/thermal hydraulics evaluation for thorium based fuels in thermal spectrum SCWR. Prog. Nucl. Energy 68, 55. Cisneros, A.T., et al., 2012. Pebble fuel design for the PB-FHR. In: Proceedings of the International Congress on Advances in Nuclear Power Plants (ICAPP12). Compilation of Ingestion and Inhalation Dose Coefficients, vol. 76, 1995. International Commission on Radiological Protection, Publication. Cycle, T.F., Potential Benefits and Challenges, International Atomic Energy Agency (IAEA) Technical Document 1450. David, S., 2005. Future scenarios for fission based reactors. Nucl. Phys. A 751, 429. Ding, M., Kloosterman, J.L., 2014. Thorium utilization in a small long-life HTR. part I: Th/U MOX fuel blocks. Nucl. Eng. Des. 267, 238. François-Paul, F.E., Fabien, Koenig, E., 2006. Preliminary Study of the Pebble-bed Advanced High Temperature Reactor. Master's thesis.
Fratoni, M., 2008. Development and Applications of Methodologies for the Neutronic Design of the Pebble Bed Advanced High Temperature Reactor (PBAHTR). ProQuest. García, C., et al., 2013. Evaluation of uranium thorium and plutonium thorium fuel cycles in a very high temperature hybrid system. Prog. Nucl. Energy 66, 61. Goluoglu, S., Williams, M.L., 2005. Modeling doubly heterogeneous systems in scale. Trans. Am. Nucl. Soc. 93, 963. Habush, A., Harris, A., 1968. 330-MW(e) fort st. vrain high-temperature gas-cooled reactor. Nucl. Eng. Des. 7 (4), 312. http://www.nndc.bnl.gov/exfor/endf00.jsp. Ilas, G., et al., 2006. On the few-group cross-section generation methodology for PBR analysis. Ann. Nucl. Energy 33, 1058. Ilas, G., et al., Validation of Scale for High Temperature Gas-cooled Reactors Analysis, NUREG/CR-7107, U.S. Nuclear Regulatory Commission http://dx.doi.org/10. 2172/1048709. Ilas, D., 2013. Scale code validation for prismatic high-temperature gas-cooled reactors. Nucl. Tech. 183 (3), 379. Ingersoll, D., et al., 2004. Status of Preconceptual Design of the Advanced Hightemperature Reactor (AHTR). Department of Energy, United States. Jiang, M., Xu, H., Dai, Z., 2012. Advanced fission energy program-TMSR nuclear energy system. Bull. Chin. Acad. Sci. 27, 366. Kim, J., 2013. Assessment of double heterogeneity treatment capability in scale. In: 22nd International Conference Nuclear Energy for New Europe. Lung, M., 1997. A Present Review of the Thorium Nuclear Fuel Cycles. Tech. rep.. Euratom. Mathieu, L., et al., 2006. The thorium molten salt reactor: moving on from the MSBR. Prog. Nucl. Energy 48 (7), 664. Merle-Lucotte, et al., 2008. Optimization and simplification of the concept of nonmoderated thorium molten salt reactor. In: Proceedings of International Conference on the Physics of Reactors, Nuclear Power: a Sustainable Resource. Nuttin, A., et al., 2005. Potential of thorium molten salt reactors detailed calculations and concept evolution with a view to large scale energy production. Prog. Nucl. Energy 46 (1), 77. Pohl, P., 2006. The Importance of the AVR Pebble-bed Reactor for the Future of Nuclear Power. Tech. rep.. American Nuclear Society, United States. Puill, A., 1999. Thorium Utilization in PWRs, IAEA-TECDOC-1319, p. 185. Ramanna, R., Lee, S., 1986. The thorium cycle for fast breeder reactors. Pramana 27, 129. Ryan Kelly, D.I., 2013. Verification of a depletion method in SCALE for the advanced high temperature reactor. Nucl. Technol. 183 (3), 391. Salvatores, M., et al., 1997. The potential of accelerator-driven systems for transmutation or power production using thorium or uranium fuel cycles. Nucl. Sci. Eng. 126, 333. Scale: a Comprehensive Modeling and Simulation Suite for Nuclear Safety Analysis and Design, ORNL/TM-2005/39, June 2011. Oak Ridge National Laboratory. Schultz, R.R., et al., 2010. Next Generation Nuclear Plant Methods Technical Program Plan, INL/EXT-06e11804. Idaho National Laboratory. Serp, J., et al., 2014. The molten salt reactor (MSR) in generation IV: overview and perspectives. Prog. Nucl. Energy 77, 308. U. M. Facilitators, M. Facilitators, 2013. Fluoride-Salt-Cooled High Temperature Reactor (FHR) Materials. Fuels and Components White Paper. Wickleder, M.S., et al., 2006. Thorium. In: The Chemistry of the Actinide and Transactinide Elements. Springer, p. 52. Williams, M., et al., 2005. Recent enhancements to the SCALE 5 resonance selfshielding methodology. ANS Trans. 92, 751.