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Evaluation of the Th–U mixed oxide fuel neutronic characteristics in a CANDU 6® reactor
Annals of Nuclear Energy 129 (2019) 240–248
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Annals of Nuclear Energy journal homepage: www.elsevier.com/locate/anucene
Evaluation of the Th–U mixed oxide fuel neutronic characteristics in a CANDU 6Ò reactor Iuliana Visan ⇑, Andreea Moise Institute for Nuclear Research Pitesti, Romania
a r t i c l e
i n f o
Article history: Received 1 November 2018 Received in revised form 15 January 2019 Accepted 26 January 2019
Keywords: Thorium fuel CANDU Neutronic evaluation Higher burnup Plutonium reduction SERPENT
a b s t r a c t The nuclear waste reduction is a long-term key issue for the nuclear energy, ensuring a low impact on the environment and a sustainable development of this technology. Addressing the fuel resources better utilization, the use of thorium in nuclear reactors has gained the attention of the scientific community. In order to identify the most appropriate configuration to be used in a CANDU 6 reactor, with respect to extended burnup and power shape flattening, a neutronic evaluation for thorium-based fuel has been conducted. The uniform power distribution in the fuel bundle has been obtained by mixing 50% ThO2 with low enriched uranium in form of UO2 in the central pin, with a radially increasing of the ThO2 fraction up to 78% in the outer ring. For this fuel configuration, three cases have been analysed, varying the enrichment in the fissile 235U (6.4%, 6.6% and 6.8%, respectively). The fuel burnup was performed using the Continuous-energy Monte Carlo Reactor Physics Burnup Calculation Code – SERPENT 2. The simulations have shown that the average exit burnup can be extended from the typical 7-7.5 MWd/kg-U up to 20 MWd/kg-HM. For the investigated cases, a significant decreasing in the overall production of plutonium isotopes and highly radiotoxic minor actinides in the spent fuel has been revealed. Ó 2019 Elsevier Ltd. All rights reserved.
1. Introduction In over 50 years of nuclear fission power, humanity has been known many benefits from nuclear energy: high energy density, stable base load of energy, environmentally friendly source of energy etc. While taking the advantages of the nuclear power, there are also some negative effects. The high radio-toxicity and heat generation caused by minor actinides (MA) represent major concerns in the management of the spent nuclear fuel (SNF). Beside the increased requirements on nuclear safety, the sustainable development of nuclear energy is based on the reduction of SNF amounts, enhancing the proliferation resistance and a better utilization (efficiency and economic competitiveness) of the fuel resources. Along the time, significant amounts of civil plutonium have been accumulated especially from LWRs and CANDU spent fuel. The isotope 239 Pu is the main fissile material in nuclear weapons manufacturing, while 238,240,241,242Pu generates technical difficulties in the use of nuclear weapons.
⇑ Corresponding author. E-mail address: [email protected] (I. Visan). https://doi.org/10.1016/j.anucene.2019.01.046 0306-4549/Ó 2019 Elsevier Ltd. All rights reserved.
The possibility of reaching higher burnup and the power flattening with a longer refuelling time is of particular interest in the nuclear reactors operation. Recent studies were carried out on the use of alternative fuels in the present or evolutionary Gen III+ nuclear reactors (NEA, 2015). The CANDU nuclear energy system has the ability to accommodate a wide variety of fuel types (Constantin et al., 2003) as DUPIC (Direct Use of spent PWR fuel In CANDU reactor) fuel, NUE (Natural Uranium Equivalent) fuel, and thorium–based fuel (Bradley et al., 1997). The nuclear fuel cycle flexibility of the CANDU type reactors is mainly based on a high neutron economy (heavy water moderator is a low neutron absorber) and the on-power refuelling. The investigations on the possibility to use the once-through thorium (OTT) fuel cycle in CANDU reactors are based on this flexibility of the CANDU system and considering the attractive features of thorium (Boczar et al., 2002), mentioned as follows: - The world thorium reserves are estimated to be approximately three times more abundant than those of uranium (Puill, 2002), (Sahin et al., 2008); the half-life of 232Th (T1/2 = 1.4 1010 years) is three times higher than that of 238U (T1/2 = 4.5 109 years). Unlike uranium, the exploitation of thorium recently started
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in Asia. As the technologies for the Th-fuel cycle have not yet achieved their maturity, both in Front-end and Back-end activities, we cannot talk about a large operating global market. - Although thorium is not a fissile material, it can be used in a mixture with different fissile materials, such as 233,235U, or 239 Pu, respectively. In the nuclear reactor, the fertile 232Th captures a neutron, then via two b decays, is transformed to 233U, which is a fissile isotope. In the Th fuel cycle, 232Th isotope plays an analogous role as 238U isotope in the U fuel cycle. Moreover, from the four practical fissile isotopes, 233U has the highest
value of the average reproduction factor g in a thermal spectrum (see Table 1 reproduced from reference (Milgram, 1984)). As known, the neutron energy spectrum in CANDU is well thermalized. - The thermo-physical properties show that ThO2 has higher thermal conductivity comparatively with the UO2 one, providing a melting point 500 °C higher than that of UO2 (Puill, 2002). The lack of oxidation is conferred by the chemical stability of ThO2 (Boczar et al., 2002). These properties provide benefits for normal operation and/or postulated accidents. - Thorium based fuel may play an important role in the reduction of plutonium inventories. By capturing a neutron, 238U in uranium fuel cycle produces 239Pu, while 232Th generates 233U. Generally, it is accepted the fact that 233U has better nonproliferation characteristics than 239Pu. The Significant Quantity of 233U is similar to that of 239Pu (8 kg), but 233U benefits by the self-protection due to the contamination with 232U which decays to a high energy gamma ray’s emitter, 208Tl (2.6 MeV per decay), but also to 212Bi and 220Rn which create some problems (Nuttin et al., 2012). The decay heat of these gammas emitters makes Th-based fuel unattractive for nuclear weapon fabrication. Although the presence of 232U makes U233 hard to handle and easy to detect, and confers proliferation resistance to the Th fuel cycle, it results in increased costs. - Higher conversion ratios (CR) is a key aspect of the sustainability of advanced nuclear systems. CR is based on the reaction rate of fissile and fertile nuclides and some contribution of intermediate nuclides (Permana et al., 2011). Thorium based fuel has the advantage of achieving higher conversion ratios due to the neutronic properties of 233U in thermal neutron energy spectrum.
2. Roadmap study The objective of this study is to find an alternative to standard CANDU 6Ò fuel able to offer improved neutronic characteristics and to mitigate the total amount of produced SNF by increasing the fuel burnup. A good fuel performance, both in terms of safety and fuel cost efficiency, can be reached by the power shape flattening. Several trials have been conducted to find a fuel composition able to reach the above objective based on a fuel bundle design containing a composition of low enriched uranium (LEU) in form of UO2 and ThO2 (named here T37), without major changes on the core design (geometry, structural and control materials, coolant etc.).
At the fuel bundle level, the flattening strategy consists in using different compositions of thorium and uranium mixed-oxide fuel (based on different ratios between ThO2 and UO2) in each fuel bundle ring, as follows: -
50% 56% 67% 78%
ThO2 ThO2 ThO2 ThO2
and and and and
50% 44% 33% 22%
UO2 UO2 UO2 UO2
in in in in
the the the the
central element; inner ring; intermediate ring; outer ring.
For this configuration, the correlation between power flattening, extended burnup and appropriate criticality results can be obtained for an enrichment in 235U over 6.4%. In this study, three cases have been analysed varying the enrichment in the fissile 235U: - Case 1: 6.4% enrichment in - Case 2: 6.6% enrichment in - Case 3: 6.8% enrichment in
235
U; U; 235 U. 235
The Continuous-energy Monte Carlo Reactor Physics Burnup Calculation Code, SERPENT 2 version 2.1.27 (Leppanen, 2015) is used for 3D lattice modelling and performing of the neutronic analysis. SERPENT 2 calculations were performed using ACE format cross section library based on JEFF-3.1.1 (Santamarina et al., 2009). Thermal and bound-atom scattering data for light and heavy water are included in SERPENT 2 code. Additional input, such as decay data library and fission yield data library, is also needed to run burnup calculation. In this paper, the decay and fission yield data are also based on JEFF-3.1.1. The simulation consists in an infinite 3D geometry defined by a cuboidal CANDU-6 lattice cell containing the fuel bundle surrounded by the moderator (see the left part in Fig. 1). A radial section of the lattice cell simulated by the SERPENT 2 code is shown in the right part of the Fig. 1 (a different material card is used to represent each ring in the fuel bundle). The reflective boundary condition (in this simulation) takes effect in all directions. This simple approximation is considered representative for the average core. Consequently, relevant indications can be obtained for the use of thorium–uranium mixed fuel in CANDU. Throughout the burnup calculation, the average power of a reference CANDU fuel bundle is used for the source rate normalization. In this paper, a high-power bundle of 800 kW (about 80% from the total core fuel bundles are characterized by lower powers) has been studied. The hottest bundle power, under the threshold defect curve (keeping the integrity of the clad), is 900 kW (around 20% from the total core fuel bundles powers are under this value). Rarely, a small percentage (less than 1%) of bundles may exceed 900 kW. CANDU 6 fuel bundle data, other than thorium-based fuel, are taken from reference (Roy et al., 1994), being given in Table 2. 3. Results and discussions 3.1. Criticality and burnup
Table 1 g in a thermal spectrum for
233,235
U and
239,241
Pu (Milgram, 1984).
Isotopes
Thermal neutron energy
Fast neutron energy
233
2.28 2.07 2.11 2.15
2.31 1.93 2.49 2.72
U U Pu 241 Pu 235 239
The infinite multiplication factor (kinf) offers indications on the multiplicative properties for the ‘‘ideal” situation (without neutrons leakage) of the lattice cell (Rouben, 2002). In CANDU 6 fuelled with natural uranium, kinf corresponding roughly to a critical reactor reaches the value of 1.05. On account of on-power refuelling, the fuel is maintained in the reactor beyond this value (even if the fuel becomes a neutron absorber) until the averaged core kinf is about 1.05 (Rouben, 2003). The average
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Fig. 1. Simulated model for CANDU 6 lattice cell: mirrored 3D view (left), radial section (right).
Table 2 Data for CANDU 6 lattice cell.
Fuel pin Fuel Cladding Coolant Pressure tube Annular gas Calandria tube Moderator *
Material
Temperature [K]
Density [g/cm3]
ThO2/UO2 Zircaloy-4 D2 O Zr-Nb CO2 Zircaloy-2 D2 O
1000 560 560 560 345 345 345
9.71/10.65* 6.44 0.81 6.57 0.0012 6.44 1.083
Inner radius [mm]
Outer radius [mm]
6.12
6.11 6.55
51.69
56.03
64.48
65.88
Considered here 97% from the theoretical density, based on the natural uranium density used in CANDU.
discharge burnup obtained from SERPENT 2 calculation for a CANDU standard bundle is 7.4 MWd/kg-U, corresponding to a fuel residence time of around 181 days. The depletion curves obtained for the considered T37 fuel bundle cases, given in comparison with that corresponding to the reference bundle (CANDU standard fuel bundle with 37 elements
fuelled with natural uranium) are shown in Fig. 2. The fuel burnup range, starting from fresh fuel at Beginning-Of-Cycle (BOC) until the burnup corresponds to the discharged fuel at End-Of-Cycle (EOC), was used. With respect to the operational criticality, kinf = 1.05, for the fresh fuel, T37 bundles are 80–100 mk supercritical comparatively with the reference bundle.
Fig. 2. Infinite multiplication constant variation with fuel burnup for the considered T37 bundle cases in comparison with Unat reference bundle case.
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The SERPENT results for T37 bundles show a smooth exponential decreasing of reactivity with the increasing irradiation. The absence of the phenomenon known as the ‘‘plutonium peak”, present in the case of Unat, is explained by the lower weight of 238U in thorium-based fuel. Small increase of 235U enrichment extends the exit fuel burnup (from 17 MWd/kg-HM to 18.2 MWd/kg-HM and 20 MWd/kg-HM, respectively) and consequently, increases the fuel residence time from 392 days to 418 days and 457 days, respectively.
becomes lower in peripheral elements (due to the higher thermal neutron flux) than in the central pin. Proportional with the 235U fission events, 239,241Pu isotopes are continuously produced from 238 U. The higher energy released per Pu fission event (from the higher mass of 239,241Pu), correlated with higher fission cross-sections in thermal energy range (rf,th(239Pu) = 742.5b, rf,th(241Pu) = 1009b versus rf,th(235U) = 583.5b) (Sahin et al., 2004), replaces the 235U consumption. The stronger non-uniformity with burnup for T37 fuel bundle cases can be explained by the accumulation of 233U which has a
higher value of g in a thermal spectrum than
3.2. Bundle power distribution Quantifying the power profiles is crucial for fuel performance analyses. Unbalanced power generation in the fuel bundle can lead to thermal deformation of pellets, inducing cladding strains that cause cracking (Rolstad and Borreses, 1977). In terms of fuel cost efficiency, a uniform burnup ensures the well-balanced fuel utilization. The variation of the T37 fuel bundle composition has been introduced to flatten the radial power shape factors for an averaged bundle. The pin shape factors obtained by normalisation the individual pin power to the mean pin power for T37 in comparison with Unat, are shown at BOC in Fig. 3, and at EOC in Fig. 4. The ‘‘ideal” situation, expressing the uniform power distribution (pin factor, PF = 1), is represented with red line. The radial power distributions at BOC and EOC, given as the average situation corresponding to each ring, are shown in Figs. 5 and 6, in which the ‘‘ideal” situation, expressing the uniform radial power distribution (radial factor, RF = 1), is represented with red line. With respect to the Unat reference bundle case, the proposed T37 configuration induces a balanced radial power distribution by increasing the power corresponding to the central pin, inner and intermediate ring elements and decreasing the power on the outer ring. The maximum values of the pin power factors at BOC are 1.07 for the proposed T37 bundle cases and 1.15 for Unat bundle case. At EOC, the situation is also improved by using thorium based fuel; thereby the maximum values of the pin power factors are 1.13 (case 1 and case 3) and 1.12 (case 2), comparatively with 1.16 for Unat bundle case. The non-uniformity of the Unat bundle power distribution remains at about the same level over the entire residence time in the reactor core. This behaviour is mainly explained by the accumulation of 239Pu; during irradiation, the atomic density of 235U
239
Pu.
3.3. Reactivity effects In this section, the Doppler reactivity (qD) and the infinite lattice coolant void reactivity (CVR) have been evaluated. The Doppler broadening of resonances is an important feedback effect depending on densities of fissile and fertile materials, on spectrum and on fuel temperature. The Doppler effect involves the reactions of absorption, fission and elastic scattering. The most important characteristic of nuclear Doppler effect is represented by the immediate response in reactivity after a change in the fuel temperature. For the safe operation of the reactor, the global effect of the Doppler broadening of resonances, as well as any other reactivity effect, must provide null reactivity variation, with tendency to zero from negative values. To simulate the Doppler feedback, fuel temperature was increased from the nominal temperature (1000 K) to 1800 K, from fresh fuel to the discharged fuel for each specific case. The replacement of the fertile 238U by the fertile 232Th contributes to the reactivity reduction as the temperature increases, while the added fissile 235U at BOC, and the produced 233U + Pu during the fuel irradiation, lead to a higher reactivity at any temperature increase. The net effect between fissions, elastic scattering events and captures defines the Doppler reactivity effect. SERPENT 2 results for Doppler constant at BOC and EOC are given in Table 3. As it can be observed, at BOC, the Doppler reactivity of T37 bundle cases is higher in absolute value than that characterizing the Unat reference fuel case. The coolant void reactivity expresses the variation of reactivity resulting from the partial or total loss of the coolant. The loss of the coolant is expressed by the void fraction, fv. The standard CANDU reactor fuelled with natural uranium and moderated with heavy water has a positive CVR, amplifying the reactor dynamics
Fig. 3. Pin power (normalised to the mean power) at BOC.
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Fig. 4. Pin power (normalised to the mean power) at EOC.
Fig. 5. Radial power distribution (average pin per ring) at BOC.
Fig. 6. Radial power distribution (average pin per ring) at EOC.
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I. Visan, A. Moise / Annals of Nuclear Energy 129 (2019) 240–248 Table 3 Doppler Reactivity values. Doppler constant, pcm
Unat Case 1 Case 2 Case 3
k1 (Tnominal)
k1 (TDoppler)
BOC
EOC
BOC
EOC
BOC
EOC
1308 1439 1581 1586
590 536 475 742
1.123 1.149 1.160 1.172
1.006 1.006 1.007 1.007
1.113 1.138 1.148 1.160
1.003 1.002 1.004 1.002
problems. It should be mentioned that, practically, the coolant void evaluation in the infinite lattice approach (characterised by the missing of the neutron leakage) provides information on the coolant density effect. In this study, the CVR has been evaluated considering the total loss of the coolant (fv 99.99%). Since the neutrons leakages are neglected, this is a conservative assumption. The voided condition has been simulated by decreasing the density of the coolant. Table 4 gives the CVR values obtained at BOC and EOC. Similar to Unat bundle case, the CVR values of T37 bundle cases decrease with burnup and keep positive values. The obtained results show that the T37 configurations are characterized at BOC by positive values,
Table 4 Coolant Void Reactivity values (pcm).
Unat Case 1 Case 2 Case 3
BOC
EOC
1389 1736 1786 1700
959 930 1063 923
higher than the reference configuration Unat. The greater positive void reactivity of T37 bundle cases can be explained by the hardened neutron spectrum which favours the fission events in 232Th and mainly in 238U. On the contrary, at EOC, the results for T37 and Unat bundle cases are somehow equivalent and of reduced magnitude due to the negative impact of the capture events of the fission products, accumulated during the fuel irradiation. 3.4. Mass inventory of the irradiated fuel The operation of a nuclear power plant produces large amounts of SNF. The composition of the irradiated fuel removed from the reactor core depends on the general core design and performances, as these relate to the initial quantity of the fissile material and the energy extracted from the fuel. Fissile isotopes, such as 233,235U and 239,241 Pu, are transmuted by nuclear fission reactions in fission products. Through neutron capture followed by b decays, fertile elements, such as 232Th, 238U, and 240Pu, will produce new fissile isotopes of 233U, 239Pu and 241Pu, respectively. The neutron reactions and radioactive decays for the major isotopes involved in the reactor operation using Th-U fuel are shown in Fig. 7 (Sahin et al., 2008).
Fig. 7. Main actinides isotopes involved in the reactor operation with Th-U fuel.
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The mass of the SNF inventory has been evaluated here, mainly to estimate the content of Pu, MA and U. The content of Pu and MA in the spent fuel is presented at discharge for the reactor fuelled with Th-based fuel in comparison with the reference case of the reactor fuelled with standard UO2 fuel. Fig. 8 illustrates the accumulation of the total Pu and main Pu isotopes. Each T37 bundle case produces a considerably lower quantity of Pu in comparison with the reference Unat fuel (about 60% less than in natural uranium). Among the most radiotoxic MA in the SNF, we can mention 241,243 Am and 237Np. The accumulation of the total Np and main Np isotopes is shown in Fig. 9. The T37 bundle cases produce significantly lower amounts of 239Np in comparison with the case of Unat bundle case. It has to be noted that although 239Np has a short half-life (T1/2 = 2.35 days), it disintegrates to 239Pu. The amount of 237 Np is about 150% higher for the studied T37 cases while the total Np production per bundle is approximately 50% less than the reference Unat. case. The inventories of Am and Cm major isotopes and the total mass are given in Figs. 10, respectively. The largest contribution (99.5%) in the total mass of Am comes from 241Am and 243Am, both for T37 cases and for the reference Unat. As can be observed,
the production of Am per bundle is higher for T37 cases than for Unat. A similar situation appears in the case of Cm; the main contributors (99%) to the total mass are 242Cm and 244Cm. In the present study, the SERPENT 2 cell calculations results on the irradiated fuel mass inventory are extrapolated for 30 years CANDU 6 reactor operation, at 90% capacity factor (according to the PRIS database, (IAEA, 2018), for Unit 1 Cernavoda NPP, Romania). Table 5 presents comparatively the calculated mass of the SNF inventory generated in 30 years of CANDU 6 reactor operation considering, in a conservative assumption, that the reactor is fuelled with Th-U fuel (T37 bundle cases) and standard CANDU UO2 fuel (Unat bundle case), respectively. The bundle mass was assumed to be around 20 kg. The total amount of the discharged fuel during the CANDU 6 reactor operating lifetime is over two times higher for Unat fuelled reactor than for Th-based fuel cases. In 30 years of CANDU 6 reactor operation with the standard Unat fuel, about 770 kg of the generated SNF mass is comprised by the Np, Am and Cm. The overall mass of MA in the SNF considerably decreases to around 180 kg for the CANDU reactor fuelled with Th-based fuel (T37 cases). Table 6 gives the major isotopes production of minor actinides in 30 years operating reactor.
Fig. 8. Inventory of Pu: the major isotopes and total mass per discharged bundle.
Fig. 9. Inventory of Np: the major isotopes and total mass per discharged bundle.
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Fig. 10. Inventory of Am: the major isotopes and total mass per discharged bundle.
Fig. 11. Inventory of Cm: the major isotopes and total mass per discharged bundle.
Table 5 Total SNF mass in 30 years operating reactor.
In Table 7, the T37 bundle cases and the reference Unat bundle case are compared in terms of main fissile material (233,235U, and 239,241 Pu) production. Unlike Unat bundle case, the Th-U spent fuel discharged from the CANDU 6 reactor after the considered operating lifetime (30 years) contains less Pu and MA, but significant amounts of 233U, thus increasing the total amount of weapon usable material (233U + 239Pu). Obviously, the higher fuel burnup obtained in Case 3 leads to a decreasing in the amount of fissile materials in comparison with Case 1 and Case 2.
mass of SNF (tonnes) Unat Case 1 Case 2 Case 3
5531 2397 2247 2054
Table 6 The major isotopes production (kg) of minor actinides in 30 years operating reactor. 237
Unat Case 1 Case 2 Case 3
Np
106.1 109.9 112.9 116.7
239
Np
241
655.5 61.6 5738 52.6
4.4 2.7 2.9 3.2
Am
243
242
6.1 4.4 4.9 5.9
1.0 0.8 0.9 1.0
Am
Cm
244
Cm
0.4 0.4 0.5 0.6
Table 7 Fissile material production (tonnes) in 30 years operating reactor.
Total MA 774.3 180.3 180.3 180.5
233
235
0 13.2 12.7 12.1
12.0 14.7 13.7 12.0
U
Unat Case 1 Case 2 Case 3
U
239
Pu
12.0 2.2 2.1 1.9
241
Pu
0.8 0.2 0.2 0.2
233
Total
12.0 15.3 14.8 14.1
U+239Pu
24.8 30.3 28.7 26.3
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4. Conclusions
Acknowledgements
The current study represents an investigation on the neutronic performances of a Th-U fuel bundle to be used in standard CANDU 6 reactor. Three (Th,U)O2 fuel options with different 235U enrichment, lower than 7%, have been analysed in comparison with the Unat bundle, considered as reference. The variation of Th/U ratios for each bundle ring has been used to obtain an appropriate configuration which allows the extended burnup and the uniform fuel utilization by the power flattening. The extended burnup leads to a significant reduction of the nuclear waste amount for the same reactor operating lifetime. The configurations of the analysed T37 bundle led to an improvement of the pin power factors from 1.15 to 1.07 (at BOC), and from 1.15 to 1.12 (at EOC), also increasing the average exit fuel burnup. Nonetheless, a total ‘‘loss” of the improved pin power factors for intermediate ring and a strong reduction (greater than 50%) for the outer ring was observed at EOC, with respect to the Unat bundle case. According to the fuel burnup, the fuel residence time in the reactor core has been extended from 181 days (for Unat bundle case) to 392 days (T37 bundle case 1), 418 days (T37 bundle case 2) and 457 days (T37 bundle case 3), respectively. Similar to the reference Unat bundle case, the T37 bundle cases keep the negative Doppler reactivity, ensuring prompt reactivity feedback in case of a temperature transient, as well as a positive coolant void reactivity. For the T37 bundle cases, CVR was higher than the CVR obtained for the reference Unat bundle case, as follows: by 22.4% up to 28.6% at BOC, and variable between 3.9% up to +10.8% at EOC. The partial substitution of uranium with thorium leads to the expected reduction of Pu and MA in the SNF, while the LEU use results in the increasing of the fuel residence time, thus meeting the paper objectives of SNF amount reduction, as well as a better utilization of the nuclear fuel resources. As the SNF composition affects the decay heat and the interim storage, reduction of the minor actinides amounts is important in order to mitigate their impact on the spent fuel overall radiotoxicity. The investigated T37 fuel bundles decrease the overall production of plutonium isotopes and the highly radiotoxic minor actinides, but contain 232U which decays to high energy gamma ray’s emitters. The contamination of U233 from Th-based spent fuel with 232 U should improve the proliferation resistance, but can imply additional radioprotection measures and increased costs. The interest for 233U as a nuclear fuel is given by the superior conversion ratios in thermal systems. The discharged T37 fuel bundle cases contain significant amount of 233U which, eventually, can be recycled along with thorium, fission products and other actinides. The production of fissile material with attractive nuclear weapon characteristics (233U + 239Pu) is higher in T37 bundle cases than in the already proliferating configuration Unat.
Authors would like to address their thanks to several colleagues, Senior Researchers from the Institute for Nuclear Research (RATEN ICN) Pitesti, Romania, as follows: to Mr. G. Olteanu, the Nuclear Fuel Research Program coordinator, for his support, and to Dr. C. A. Margeanu, Dr. M. Constantin, and Mr. A. Rizoiu, for the valuable discussions. The authors would like to extend their gratitude to Dr. G. Glinatsis, former Senior Researcher at ENEA, Italy, for the constructive criticism on the present work. Appendix A. Supplementary data Supplementary data to this article can be found online at https://doi.org/10.1016/j.anucene.2019.01.046. References Boczar, P. et al., 2002. Thorium Fuel-cycle Studies for CANDU Reactors, Thorium Fuel Utilization: Options and Trends. International Atomic Energy Agency, Vienna. Bradley, K., Boczar, P., Fehrenbach, P., 1997. CANDU Advanced Fuel Cycles: Key to Energy Sustainability. PEBN-BATAN, Jakarta. Constantin, M., Gugiu, D., Balaceanu, V., 2003. Void reactivity and pin power calculation for a CANDU cell using the SEU-43 fuel bundle. Ann. Nucl. Energy 30, 301–316. IAEA PRIS – The Power Reactor 2018 Information System. [Online] Available at: https://pris.iaea.org. Leppanen, J., 2015. Serpent – A Continous-energy Monte Carlo Reactor Physics Burnup Calculation Code (June 18, 2015), s.l. VTT Technical Research Centre of Finland. Milgram, M., 1984. Thorium Fuel Cycles in CANDU Reactors: A Review, s.l.: Atomic Energy of Canada Report, AECL-8326. NEA, 2015. Perspective on the Use of Thorium in the Nuclear Fuel Cycle, Paris: OECD NEA No. 7224. Nuttin, A. et al., 2012. Comparative analysis of high conversion achievable in thorium-fueled slightly modified CANDU and PWR reactors. Ann. Nucl. Energy 40, 171–189. Permana, S., Takaki, N., Sekimoto, H., 2011. Breeding and void reactivity analysis on heavy metal closed-cycle water cooled thorium reactor. Ann. Nucl. Energy 38, 337–347. Puill, A., 2002. Thorium Utilization in PWRs Neutronic Studies, Thorium Fuel Utilization: Options and Trends. International Atomic Energy Agency, Vienna. Rolstad, E., Borreses, S., 1977. Fuel Reliability in Light-Water Reactors. International Atomic Energy, Vienna, pp. 581–594. Rouben, B., 2002. Introduction to Reactor Physics. Course Notes, Atomic Energy of Canada Limited. [Online] Available at:. Rouben, B., 2003. CANDU Fuel Management. Atomic Energy of Canada Limited, s.l.. Roy, R., Marleau, G., Tajmouati, J., Rozon, D., 1994. Modelling of CANDU reactivity control devices with the lattice code DRAGON. Ann. Nucl. Energy 21 (2), 115– 132. Sahin, S. et al., 2008. CANDU reactor as minor actinide/thorium burner with uniform power density in the fuel bundle. Ann. Nucl. Energy 35, 690–703. Sahin, S., Yildiz, K., Acir, A., 2004. Power flattening in the fuel bundle of a CANDU reactor. Nucl. Eng. Des. 232, 7–18. Santamarina, A. et al., 2009. The JEFF-3.1.1 Nuclear Data Library, s.l. OECD NEA.
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Annals of Nuclear Energy journal homepage: www.elsevier.com/locate/anucene
Evaluation of the Th–U mixed oxide fuel neutronic characteristics in a CANDU 6Ò reactor Iuliana Visan ⇑, Andreea Moise Institute for Nuclear Research Pitesti, Romania
a r t i c l e
i n f o
Article history: Received 1 November 2018 Received in revised form 15 January 2019 Accepted 26 January 2019
Keywords: Thorium fuel CANDU Neutronic evaluation Higher burnup Plutonium reduction SERPENT
a b s t r a c t The nuclear waste reduction is a long-term key issue for the nuclear energy, ensuring a low impact on the environment and a sustainable development of this technology. Addressing the fuel resources better utilization, the use of thorium in nuclear reactors has gained the attention of the scientific community. In order to identify the most appropriate configuration to be used in a CANDU 6 reactor, with respect to extended burnup and power shape flattening, a neutronic evaluation for thorium-based fuel has been conducted. The uniform power distribution in the fuel bundle has been obtained by mixing 50% ThO2 with low enriched uranium in form of UO2 in the central pin, with a radially increasing of the ThO2 fraction up to 78% in the outer ring. For this fuel configuration, three cases have been analysed, varying the enrichment in the fissile 235U (6.4%, 6.6% and 6.8%, respectively). The fuel burnup was performed using the Continuous-energy Monte Carlo Reactor Physics Burnup Calculation Code – SERPENT 2. The simulations have shown that the average exit burnup can be extended from the typical 7-7.5 MWd/kg-U up to 20 MWd/kg-HM. For the investigated cases, a significant decreasing in the overall production of plutonium isotopes and highly radiotoxic minor actinides in the spent fuel has been revealed. Ó 2019 Elsevier Ltd. All rights reserved.
1. Introduction In over 50 years of nuclear fission power, humanity has been known many benefits from nuclear energy: high energy density, stable base load of energy, environmentally friendly source of energy etc. While taking the advantages of the nuclear power, there are also some negative effects. The high radio-toxicity and heat generation caused by minor actinides (MA) represent major concerns in the management of the spent nuclear fuel (SNF). Beside the increased requirements on nuclear safety, the sustainable development of nuclear energy is based on the reduction of SNF amounts, enhancing the proliferation resistance and a better utilization (efficiency and economic competitiveness) of the fuel resources. Along the time, significant amounts of civil plutonium have been accumulated especially from LWRs and CANDU spent fuel. The isotope 239 Pu is the main fissile material in nuclear weapons manufacturing, while 238,240,241,242Pu generates technical difficulties in the use of nuclear weapons.
⇑ Corresponding author. E-mail address: [email protected] (I. Visan). https://doi.org/10.1016/j.anucene.2019.01.046 0306-4549/Ó 2019 Elsevier Ltd. All rights reserved.
The possibility of reaching higher burnup and the power flattening with a longer refuelling time is of particular interest in the nuclear reactors operation. Recent studies were carried out on the use of alternative fuels in the present or evolutionary Gen III+ nuclear reactors (NEA, 2015). The CANDU nuclear energy system has the ability to accommodate a wide variety of fuel types (Constantin et al., 2003) as DUPIC (Direct Use of spent PWR fuel In CANDU reactor) fuel, NUE (Natural Uranium Equivalent) fuel, and thorium–based fuel (Bradley et al., 1997). The nuclear fuel cycle flexibility of the CANDU type reactors is mainly based on a high neutron economy (heavy water moderator is a low neutron absorber) and the on-power refuelling. The investigations on the possibility to use the once-through thorium (OTT) fuel cycle in CANDU reactors are based on this flexibility of the CANDU system and considering the attractive features of thorium (Boczar et al., 2002), mentioned as follows: - The world thorium reserves are estimated to be approximately three times more abundant than those of uranium (Puill, 2002), (Sahin et al., 2008); the half-life of 232Th (T1/2 = 1.4 1010 years) is three times higher than that of 238U (T1/2 = 4.5 109 years). Unlike uranium, the exploitation of thorium recently started
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in Asia. As the technologies for the Th-fuel cycle have not yet achieved their maturity, both in Front-end and Back-end activities, we cannot talk about a large operating global market. - Although thorium is not a fissile material, it can be used in a mixture with different fissile materials, such as 233,235U, or 239 Pu, respectively. In the nuclear reactor, the fertile 232Th captures a neutron, then via two b decays, is transformed to 233U, which is a fissile isotope. In the Th fuel cycle, 232Th isotope plays an analogous role as 238U isotope in the U fuel cycle. Moreover, from the four practical fissile isotopes, 233U has the highest
value of the average reproduction factor g in a thermal spectrum (see Table 1 reproduced from reference (Milgram, 1984)). As known, the neutron energy spectrum in CANDU is well thermalized. - The thermo-physical properties show that ThO2 has higher thermal conductivity comparatively with the UO2 one, providing a melting point 500 °C higher than that of UO2 (Puill, 2002). The lack of oxidation is conferred by the chemical stability of ThO2 (Boczar et al., 2002). These properties provide benefits for normal operation and/or postulated accidents. - Thorium based fuel may play an important role in the reduction of plutonium inventories. By capturing a neutron, 238U in uranium fuel cycle produces 239Pu, while 232Th generates 233U. Generally, it is accepted the fact that 233U has better nonproliferation characteristics than 239Pu. The Significant Quantity of 233U is similar to that of 239Pu (8 kg), but 233U benefits by the self-protection due to the contamination with 232U which decays to a high energy gamma ray’s emitter, 208Tl (2.6 MeV per decay), but also to 212Bi and 220Rn which create some problems (Nuttin et al., 2012). The decay heat of these gammas emitters makes Th-based fuel unattractive for nuclear weapon fabrication. Although the presence of 232U makes U233 hard to handle and easy to detect, and confers proliferation resistance to the Th fuel cycle, it results in increased costs. - Higher conversion ratios (CR) is a key aspect of the sustainability of advanced nuclear systems. CR is based on the reaction rate of fissile and fertile nuclides and some contribution of intermediate nuclides (Permana et al., 2011). Thorium based fuel has the advantage of achieving higher conversion ratios due to the neutronic properties of 233U in thermal neutron energy spectrum.
2. Roadmap study The objective of this study is to find an alternative to standard CANDU 6Ò fuel able to offer improved neutronic characteristics and to mitigate the total amount of produced SNF by increasing the fuel burnup. A good fuel performance, both in terms of safety and fuel cost efficiency, can be reached by the power shape flattening. Several trials have been conducted to find a fuel composition able to reach the above objective based on a fuel bundle design containing a composition of low enriched uranium (LEU) in form of UO2 and ThO2 (named here T37), without major changes on the core design (geometry, structural and control materials, coolant etc.).
At the fuel bundle level, the flattening strategy consists in using different compositions of thorium and uranium mixed-oxide fuel (based on different ratios between ThO2 and UO2) in each fuel bundle ring, as follows: -
50% 56% 67% 78%
ThO2 ThO2 ThO2 ThO2
and and and and
50% 44% 33% 22%
UO2 UO2 UO2 UO2
in in in in
the the the the
central element; inner ring; intermediate ring; outer ring.
For this configuration, the correlation between power flattening, extended burnup and appropriate criticality results can be obtained for an enrichment in 235U over 6.4%. In this study, three cases have been analysed varying the enrichment in the fissile 235U: - Case 1: 6.4% enrichment in - Case 2: 6.6% enrichment in - Case 3: 6.8% enrichment in
235
U; U; 235 U. 235
The Continuous-energy Monte Carlo Reactor Physics Burnup Calculation Code, SERPENT 2 version 2.1.27 (Leppanen, 2015) is used for 3D lattice modelling and performing of the neutronic analysis. SERPENT 2 calculations were performed using ACE format cross section library based on JEFF-3.1.1 (Santamarina et al., 2009). Thermal and bound-atom scattering data for light and heavy water are included in SERPENT 2 code. Additional input, such as decay data library and fission yield data library, is also needed to run burnup calculation. In this paper, the decay and fission yield data are also based on JEFF-3.1.1. The simulation consists in an infinite 3D geometry defined by a cuboidal CANDU-6 lattice cell containing the fuel bundle surrounded by the moderator (see the left part in Fig. 1). A radial section of the lattice cell simulated by the SERPENT 2 code is shown in the right part of the Fig. 1 (a different material card is used to represent each ring in the fuel bundle). The reflective boundary condition (in this simulation) takes effect in all directions. This simple approximation is considered representative for the average core. Consequently, relevant indications can be obtained for the use of thorium–uranium mixed fuel in CANDU. Throughout the burnup calculation, the average power of a reference CANDU fuel bundle is used for the source rate normalization. In this paper, a high-power bundle of 800 kW (about 80% from the total core fuel bundles are characterized by lower powers) has been studied. The hottest bundle power, under the threshold defect curve (keeping the integrity of the clad), is 900 kW (around 20% from the total core fuel bundles powers are under this value). Rarely, a small percentage (less than 1%) of bundles may exceed 900 kW. CANDU 6 fuel bundle data, other than thorium-based fuel, are taken from reference (Roy et al., 1994), being given in Table 2. 3. Results and discussions 3.1. Criticality and burnup
Table 1 g in a thermal spectrum for
233,235
U and
239,241
Pu (Milgram, 1984).
Isotopes
Thermal neutron energy
Fast neutron energy
233
2.28 2.07 2.11 2.15
2.31 1.93 2.49 2.72
U U Pu 241 Pu 235 239
The infinite multiplication factor (kinf) offers indications on the multiplicative properties for the ‘‘ideal” situation (without neutrons leakage) of the lattice cell (Rouben, 2002). In CANDU 6 fuelled with natural uranium, kinf corresponding roughly to a critical reactor reaches the value of 1.05. On account of on-power refuelling, the fuel is maintained in the reactor beyond this value (even if the fuel becomes a neutron absorber) until the averaged core kinf is about 1.05 (Rouben, 2003). The average
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Fig. 1. Simulated model for CANDU 6 lattice cell: mirrored 3D view (left), radial section (right).
Table 2 Data for CANDU 6 lattice cell.
Fuel pin Fuel Cladding Coolant Pressure tube Annular gas Calandria tube Moderator *
Material
Temperature [K]
Density [g/cm3]
ThO2/UO2 Zircaloy-4 D2 O Zr-Nb CO2 Zircaloy-2 D2 O
1000 560 560 560 345 345 345
9.71/10.65* 6.44 0.81 6.57 0.0012 6.44 1.083
Inner radius [mm]
Outer radius [mm]
6.12
6.11 6.55
51.69
56.03
64.48
65.88
Considered here 97% from the theoretical density, based on the natural uranium density used in CANDU.
discharge burnup obtained from SERPENT 2 calculation for a CANDU standard bundle is 7.4 MWd/kg-U, corresponding to a fuel residence time of around 181 days. The depletion curves obtained for the considered T37 fuel bundle cases, given in comparison with that corresponding to the reference bundle (CANDU standard fuel bundle with 37 elements
fuelled with natural uranium) are shown in Fig. 2. The fuel burnup range, starting from fresh fuel at Beginning-Of-Cycle (BOC) until the burnup corresponds to the discharged fuel at End-Of-Cycle (EOC), was used. With respect to the operational criticality, kinf = 1.05, for the fresh fuel, T37 bundles are 80–100 mk supercritical comparatively with the reference bundle.
Fig. 2. Infinite multiplication constant variation with fuel burnup for the considered T37 bundle cases in comparison with Unat reference bundle case.
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The SERPENT results for T37 bundles show a smooth exponential decreasing of reactivity with the increasing irradiation. The absence of the phenomenon known as the ‘‘plutonium peak”, present in the case of Unat, is explained by the lower weight of 238U in thorium-based fuel. Small increase of 235U enrichment extends the exit fuel burnup (from 17 MWd/kg-HM to 18.2 MWd/kg-HM and 20 MWd/kg-HM, respectively) and consequently, increases the fuel residence time from 392 days to 418 days and 457 days, respectively.
becomes lower in peripheral elements (due to the higher thermal neutron flux) than in the central pin. Proportional with the 235U fission events, 239,241Pu isotopes are continuously produced from 238 U. The higher energy released per Pu fission event (from the higher mass of 239,241Pu), correlated with higher fission cross-sections in thermal energy range (rf,th(239Pu) = 742.5b, rf,th(241Pu) = 1009b versus rf,th(235U) = 583.5b) (Sahin et al., 2004), replaces the 235U consumption. The stronger non-uniformity with burnup for T37 fuel bundle cases can be explained by the accumulation of 233U which has a
higher value of g in a thermal spectrum than
3.2. Bundle power distribution Quantifying the power profiles is crucial for fuel performance analyses. Unbalanced power generation in the fuel bundle can lead to thermal deformation of pellets, inducing cladding strains that cause cracking (Rolstad and Borreses, 1977). In terms of fuel cost efficiency, a uniform burnup ensures the well-balanced fuel utilization. The variation of the T37 fuel bundle composition has been introduced to flatten the radial power shape factors for an averaged bundle. The pin shape factors obtained by normalisation the individual pin power to the mean pin power for T37 in comparison with Unat, are shown at BOC in Fig. 3, and at EOC in Fig. 4. The ‘‘ideal” situation, expressing the uniform power distribution (pin factor, PF = 1), is represented with red line. The radial power distributions at BOC and EOC, given as the average situation corresponding to each ring, are shown in Figs. 5 and 6, in which the ‘‘ideal” situation, expressing the uniform radial power distribution (radial factor, RF = 1), is represented with red line. With respect to the Unat reference bundle case, the proposed T37 configuration induces a balanced radial power distribution by increasing the power corresponding to the central pin, inner and intermediate ring elements and decreasing the power on the outer ring. The maximum values of the pin power factors at BOC are 1.07 for the proposed T37 bundle cases and 1.15 for Unat bundle case. At EOC, the situation is also improved by using thorium based fuel; thereby the maximum values of the pin power factors are 1.13 (case 1 and case 3) and 1.12 (case 2), comparatively with 1.16 for Unat bundle case. The non-uniformity of the Unat bundle power distribution remains at about the same level over the entire residence time in the reactor core. This behaviour is mainly explained by the accumulation of 239Pu; during irradiation, the atomic density of 235U
239
Pu.
3.3. Reactivity effects In this section, the Doppler reactivity (qD) and the infinite lattice coolant void reactivity (CVR) have been evaluated. The Doppler broadening of resonances is an important feedback effect depending on densities of fissile and fertile materials, on spectrum and on fuel temperature. The Doppler effect involves the reactions of absorption, fission and elastic scattering. The most important characteristic of nuclear Doppler effect is represented by the immediate response in reactivity after a change in the fuel temperature. For the safe operation of the reactor, the global effect of the Doppler broadening of resonances, as well as any other reactivity effect, must provide null reactivity variation, with tendency to zero from negative values. To simulate the Doppler feedback, fuel temperature was increased from the nominal temperature (1000 K) to 1800 K, from fresh fuel to the discharged fuel for each specific case. The replacement of the fertile 238U by the fertile 232Th contributes to the reactivity reduction as the temperature increases, while the added fissile 235U at BOC, and the produced 233U + Pu during the fuel irradiation, lead to a higher reactivity at any temperature increase. The net effect between fissions, elastic scattering events and captures defines the Doppler reactivity effect. SERPENT 2 results for Doppler constant at BOC and EOC are given in Table 3. As it can be observed, at BOC, the Doppler reactivity of T37 bundle cases is higher in absolute value than that characterizing the Unat reference fuel case. The coolant void reactivity expresses the variation of reactivity resulting from the partial or total loss of the coolant. The loss of the coolant is expressed by the void fraction, fv. The standard CANDU reactor fuelled with natural uranium and moderated with heavy water has a positive CVR, amplifying the reactor dynamics
Fig. 3. Pin power (normalised to the mean power) at BOC.
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Fig. 4. Pin power (normalised to the mean power) at EOC.
Fig. 5. Radial power distribution (average pin per ring) at BOC.
Fig. 6. Radial power distribution (average pin per ring) at EOC.
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Unat Case 1 Case 2 Case 3
k1 (Tnominal)
k1 (TDoppler)
BOC
EOC
BOC
EOC
BOC
EOC
1308 1439 1581 1586
590 536 475 742
1.123 1.149 1.160 1.172
1.006 1.006 1.007 1.007
1.113 1.138 1.148 1.160
1.003 1.002 1.004 1.002
problems. It should be mentioned that, practically, the coolant void evaluation in the infinite lattice approach (characterised by the missing of the neutron leakage) provides information on the coolant density effect. In this study, the CVR has been evaluated considering the total loss of the coolant (fv 99.99%). Since the neutrons leakages are neglected, this is a conservative assumption. The voided condition has been simulated by decreasing the density of the coolant. Table 4 gives the CVR values obtained at BOC and EOC. Similar to Unat bundle case, the CVR values of T37 bundle cases decrease with burnup and keep positive values. The obtained results show that the T37 configurations are characterized at BOC by positive values,
Table 4 Coolant Void Reactivity values (pcm).
Unat Case 1 Case 2 Case 3
BOC
EOC
1389 1736 1786 1700
959 930 1063 923
higher than the reference configuration Unat. The greater positive void reactivity of T37 bundle cases can be explained by the hardened neutron spectrum which favours the fission events in 232Th and mainly in 238U. On the contrary, at EOC, the results for T37 and Unat bundle cases are somehow equivalent and of reduced magnitude due to the negative impact of the capture events of the fission products, accumulated during the fuel irradiation. 3.4. Mass inventory of the irradiated fuel The operation of a nuclear power plant produces large amounts of SNF. The composition of the irradiated fuel removed from the reactor core depends on the general core design and performances, as these relate to the initial quantity of the fissile material and the energy extracted from the fuel. Fissile isotopes, such as 233,235U and 239,241 Pu, are transmuted by nuclear fission reactions in fission products. Through neutron capture followed by b decays, fertile elements, such as 232Th, 238U, and 240Pu, will produce new fissile isotopes of 233U, 239Pu and 241Pu, respectively. The neutron reactions and radioactive decays for the major isotopes involved in the reactor operation using Th-U fuel are shown in Fig. 7 (Sahin et al., 2008).
Fig. 7. Main actinides isotopes involved in the reactor operation with Th-U fuel.
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The mass of the SNF inventory has been evaluated here, mainly to estimate the content of Pu, MA and U. The content of Pu and MA in the spent fuel is presented at discharge for the reactor fuelled with Th-based fuel in comparison with the reference case of the reactor fuelled with standard UO2 fuel. Fig. 8 illustrates the accumulation of the total Pu and main Pu isotopes. Each T37 bundle case produces a considerably lower quantity of Pu in comparison with the reference Unat fuel (about 60% less than in natural uranium). Among the most radiotoxic MA in the SNF, we can mention 241,243 Am and 237Np. The accumulation of the total Np and main Np isotopes is shown in Fig. 9. The T37 bundle cases produce significantly lower amounts of 239Np in comparison with the case of Unat bundle case. It has to be noted that although 239Np has a short half-life (T1/2 = 2.35 days), it disintegrates to 239Pu. The amount of 237 Np is about 150% higher for the studied T37 cases while the total Np production per bundle is approximately 50% less than the reference Unat. case. The inventories of Am and Cm major isotopes and the total mass are given in Figs. 10, respectively. The largest contribution (99.5%) in the total mass of Am comes from 241Am and 243Am, both for T37 cases and for the reference Unat. As can be observed,
the production of Am per bundle is higher for T37 cases than for Unat. A similar situation appears in the case of Cm; the main contributors (99%) to the total mass are 242Cm and 244Cm. In the present study, the SERPENT 2 cell calculations results on the irradiated fuel mass inventory are extrapolated for 30 years CANDU 6 reactor operation, at 90% capacity factor (according to the PRIS database, (IAEA, 2018), for Unit 1 Cernavoda NPP, Romania). Table 5 presents comparatively the calculated mass of the SNF inventory generated in 30 years of CANDU 6 reactor operation considering, in a conservative assumption, that the reactor is fuelled with Th-U fuel (T37 bundle cases) and standard CANDU UO2 fuel (Unat bundle case), respectively. The bundle mass was assumed to be around 20 kg. The total amount of the discharged fuel during the CANDU 6 reactor operating lifetime is over two times higher for Unat fuelled reactor than for Th-based fuel cases. In 30 years of CANDU 6 reactor operation with the standard Unat fuel, about 770 kg of the generated SNF mass is comprised by the Np, Am and Cm. The overall mass of MA in the SNF considerably decreases to around 180 kg for the CANDU reactor fuelled with Th-based fuel (T37 cases). Table 6 gives the major isotopes production of minor actinides in 30 years operating reactor.
Fig. 8. Inventory of Pu: the major isotopes and total mass per discharged bundle.
Fig. 9. Inventory of Np: the major isotopes and total mass per discharged bundle.
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Fig. 10. Inventory of Am: the major isotopes and total mass per discharged bundle.
Fig. 11. Inventory of Cm: the major isotopes and total mass per discharged bundle.
Table 5 Total SNF mass in 30 years operating reactor.
In Table 7, the T37 bundle cases and the reference Unat bundle case are compared in terms of main fissile material (233,235U, and 239,241 Pu) production. Unlike Unat bundle case, the Th-U spent fuel discharged from the CANDU 6 reactor after the considered operating lifetime (30 years) contains less Pu and MA, but significant amounts of 233U, thus increasing the total amount of weapon usable material (233U + 239Pu). Obviously, the higher fuel burnup obtained in Case 3 leads to a decreasing in the amount of fissile materials in comparison with Case 1 and Case 2.
mass of SNF (tonnes) Unat Case 1 Case 2 Case 3
5531 2397 2247 2054
Table 6 The major isotopes production (kg) of minor actinides in 30 years operating reactor. 237
Unat Case 1 Case 2 Case 3
Np
106.1 109.9 112.9 116.7
239
Np
241
655.5 61.6 5738 52.6
4.4 2.7 2.9 3.2
Am
243
242
6.1 4.4 4.9 5.9
1.0 0.8 0.9 1.0
Am
Cm
244
Cm
0.4 0.4 0.5 0.6
Table 7 Fissile material production (tonnes) in 30 years operating reactor.
Total MA 774.3 180.3 180.3 180.5
233
235
0 13.2 12.7 12.1
12.0 14.7 13.7 12.0
U
Unat Case 1 Case 2 Case 3
U
239
Pu
12.0 2.2 2.1 1.9
241
Pu
0.8 0.2 0.2 0.2
233
Total
12.0 15.3 14.8 14.1
U+239Pu
24.8 30.3 28.7 26.3
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4. Conclusions
Acknowledgements
The current study represents an investigation on the neutronic performances of a Th-U fuel bundle to be used in standard CANDU 6 reactor. Three (Th,U)O2 fuel options with different 235U enrichment, lower than 7%, have been analysed in comparison with the Unat bundle, considered as reference. The variation of Th/U ratios for each bundle ring has been used to obtain an appropriate configuration which allows the extended burnup and the uniform fuel utilization by the power flattening. The extended burnup leads to a significant reduction of the nuclear waste amount for the same reactor operating lifetime. The configurations of the analysed T37 bundle led to an improvement of the pin power factors from 1.15 to 1.07 (at BOC), and from 1.15 to 1.12 (at EOC), also increasing the average exit fuel burnup. Nonetheless, a total ‘‘loss” of the improved pin power factors for intermediate ring and a strong reduction (greater than 50%) for the outer ring was observed at EOC, with respect to the Unat bundle case. According to the fuel burnup, the fuel residence time in the reactor core has been extended from 181 days (for Unat bundle case) to 392 days (T37 bundle case 1), 418 days (T37 bundle case 2) and 457 days (T37 bundle case 3), respectively. Similar to the reference Unat bundle case, the T37 bundle cases keep the negative Doppler reactivity, ensuring prompt reactivity feedback in case of a temperature transient, as well as a positive coolant void reactivity. For the T37 bundle cases, CVR was higher than the CVR obtained for the reference Unat bundle case, as follows: by 22.4% up to 28.6% at BOC, and variable between 3.9% up to +10.8% at EOC. The partial substitution of uranium with thorium leads to the expected reduction of Pu and MA in the SNF, while the LEU use results in the increasing of the fuel residence time, thus meeting the paper objectives of SNF amount reduction, as well as a better utilization of the nuclear fuel resources. As the SNF composition affects the decay heat and the interim storage, reduction of the minor actinides amounts is important in order to mitigate their impact on the spent fuel overall radiotoxicity. The investigated T37 fuel bundles decrease the overall production of plutonium isotopes and the highly radiotoxic minor actinides, but contain 232U which decays to high energy gamma ray’s emitters. The contamination of U233 from Th-based spent fuel with 232 U should improve the proliferation resistance, but can imply additional radioprotection measures and increased costs. The interest for 233U as a nuclear fuel is given by the superior conversion ratios in thermal systems. The discharged T37 fuel bundle cases contain significant amount of 233U which, eventually, can be recycled along with thorium, fission products and other actinides. The production of fissile material with attractive nuclear weapon characteristics (233U + 239Pu) is higher in T37 bundle cases than in the already proliferating configuration Unat.
Authors would like to address their thanks to several colleagues, Senior Researchers from the Institute for Nuclear Research (RATEN ICN) Pitesti, Romania, as follows: to Mr. G. Olteanu, the Nuclear Fuel Research Program coordinator, for his support, and to Dr. C. A. Margeanu, Dr. M. Constantin, and Mr. A. Rizoiu, for the valuable discussions. The authors would like to extend their gratitude to Dr. G. Glinatsis, former Senior Researcher at ENEA, Italy, for the constructive criticism on the present work. Appendix A. Supplementary data Supplementary data to this article can be found online at https://doi.org/10.1016/j.anucene.2019.01.046. References Boczar, P. et al., 2002. Thorium Fuel-cycle Studies for CANDU Reactors, Thorium Fuel Utilization: Options and Trends. International Atomic Energy Agency, Vienna. Bradley, K., Boczar, P., Fehrenbach, P., 1997. CANDU Advanced Fuel Cycles: Key to Energy Sustainability. PEBN-BATAN, Jakarta. Constantin, M., Gugiu, D., Balaceanu, V., 2003. Void reactivity and pin power calculation for a CANDU cell using the SEU-43 fuel bundle. Ann. Nucl. Energy 30, 301–316. IAEA PRIS – The Power Reactor 2018 Information System. [Online] Available at: https://pris.iaea.org. Leppanen, J., 2015. Serpent – A Continous-energy Monte Carlo Reactor Physics Burnup Calculation Code (June 18, 2015), s.l. VTT Technical Research Centre of Finland. Milgram, M., 1984. Thorium Fuel Cycles in CANDU Reactors: A Review, s.l.: Atomic Energy of Canada Report, AECL-8326. NEA, 2015. Perspective on the Use of Thorium in the Nuclear Fuel Cycle, Paris: OECD NEA No. 7224. Nuttin, A. et al., 2012. Comparative analysis of high conversion achievable in thorium-fueled slightly modified CANDU and PWR reactors. Ann. Nucl. Energy 40, 171–189. Permana, S., Takaki, N., Sekimoto, H., 2011. Breeding and void reactivity analysis on heavy metal closed-cycle water cooled thorium reactor. Ann. Nucl. Energy 38, 337–347. Puill, A., 2002. Thorium Utilization in PWRs Neutronic Studies, Thorium Fuel Utilization: Options and Trends. International Atomic Energy Agency, Vienna. Rolstad, E., Borreses, S., 1977. Fuel Reliability in Light-Water Reactors. International Atomic Energy, Vienna, pp. 581–594. Rouben, B., 2002. Introduction to Reactor Physics. Course Notes, Atomic Energy of Canada Limited. [Online] Available at: