Thorium Fuel Performance in a Tight Pitch LWR Lattice

Transcription

1 Thorium Fuel Performance in a Tight Pitch LWR Lattice Taek Kyum Kim and Thomas J. Downar School of Nuclear Engineering Purdue University Abstract Research on the utilization of thorium-based fuels in the intermediate neutron spectrum of a tight pitch LWR lattice is reported in this paper. The analysis was performed using the Studsvik/Scandpower lattice physics code HELIOS. The results show that thorium-based fuels in the intermediate spectrum of tight pitch LWR have considerable advantages in terms of conversion ratio, reactivity control, non-proliferation characteristics and a reduced production of long-lived radiotoxic wastes. Due to the high conversion ratio of thorium-based fuels in intermediate spectrum reactors, the total fissile inventory required to achieve a given fuel burnup is only 11% to 17% higher than that of U 238 fertile fuels. However, unlike U 238 fertile fuels, the void reactivity coefficient with thorium-based fuels is negative in an intermediate spectrum reactor. This provides motivation for replacing U 238 with Th 232 in advanced high conversion intermediate spectrum LWRs (AHCLWR), such as the Reduced-moderator reactor (RMWR) or the Supercritical reactor (SCR). I. Introduction Historically, the principal focus of research on thorium-based fuels has been on applications to conventional light water reactor core designs [1,2]. This was motivated primarily by the attractive neutronics properties of the fissile U 233 isotope which is bred by neutron capture in Th 232. At thermal and epithermal neutron energies, U 233 has a higher neutron yield (η) per fission than other common fissile isotopes, U 235 and Pu 239 and therefore was attractive for such innovative fuel cycles as a thermal breeder. More generally, however, thorium-based fuels have several superior physical properties relative to the corresponding uranium-based fuels. From the fuel materials standpoint, ThO 2 -rich 1

2 fuels are better suited to withstand the rigors of high burnup than all UO 2 fuels, [3] and the higher conductivity of thorium-based fuels than that of uranium-based fuel and the lower fission gas release rate from the fuel matrix contribute to the reliability and safety margin of the core. Because thorium does not have a fissile isotope, it is necessary to utilize thorium, at least initially, with some fissile material other than U 233. Many of the advantages of thorium-based fuels are most evident only after thorium has been irradiated and reprocessed, and therefore most of the reactor concepts utilizing throrium fuels were performed under the assumption of reprocessing. W.J. Oosterkamp [1] showed that the thorium cycle in PWR has a potential for reducing the uranium resources with no substantial economic penalty. D.B. Trauger [2] studied the use of thorium-based fuels in various reactor concepts, such as the Light Water Reactors (LWRs), Heavy Water Reactors (HWRs), High Temperature Gas Cooled Reactors (HTGRs), and Molten Salt Reactors (MSRs). N. L. Shapiro [4] evaluated alternate thorium fuel cycles in Combustion Engineering Standard System 80 plants to determine the economic potential and technical feasibility of thorium-based fuel in present PWRs. P. R. Kastern [5] evaluated the role of thorium fuel in four-types of power reactors: LWRs, HTGRs, HWRs and FBRs. International Nuclear Fuel Cycle Evaluation (INFCE) of IAEA [6,7] also studied the thorium fuel cycle in order to minimize the danger of proliferation of nuclear weapons without jeopardizing energy supplies or the development of nuclear power. All of this previous work demonstrated that thorium-based fuel cycles were generally comparable to the UO 2 fuel cycle from the standpoint of overall technical feasibility in current PWRs. However, these studies were performed under the assumptions of reprocessing and the use of highly enriched U 235. These assumptions have become less valid because of concerns about nuclear proliferation issues. More recently the direction of thorium-based fuel studies have turned to the usage of thorium for incinerating the surplus or weapons-grade plutonium, and for reducing the inventory of long-term toxicity isotopes without substantial modifications to current LWR systems and without requiring fuel reprocessing. A. Galperin, A. Radkowsky, M. Todosow, and H. K. Joo, et al, [8~13] showed that the performance of PWRs with thorium-based fuels are similar to that of current reactors fuel with uranium, and also showed that thorium-based fuels have advantages compared to uranium-based fuel in terms of reactivity coefficients, reduced production of long-term toxicity isotopes, proliferation resistance, and the incineration of 2

3 plutonium. However, the discharge burnup and the cycle length of cores loaded with thorium-based fuels were shorter than corresponding uranium-based cores due to the relatively low fissile cross-section of U 233 and the high thermal absorption of thorium. Therefore, to achieve a high burnup with thorium-based fuels in LWRs required higher fuel enrichments which led to higher fuel cycle costs. Some recent studies which have been designed to achieve improved economics by increasing the burnup of thorium-based fuels, [14-18] still show that in conventional LWR lattices thorium-based fuel will always have lower reactivity and therefore will not be able to achieve significantly higher burnups than uranium-based fuels with a similar fissile enrichment. Some researchers then began to focus attention on the performance of thorium in tighter pitch fuel lattices and investigate whether thorium-based fuels possess advantages in intermediate neutron spectra. [14-20] The results of some preliminary studies showed that thorium-based fuels do have several attractive characteristics in the tight pitch lattice designs such as a more negative void coefficient and a higher fuel conversion ratio than corresponding uranium based fuels. [19,20] These preliminary studies motivated the research performed in this work which was to investigate the use of thorium in advanced high conversion light water reactor concepts such as the Reduced Moderation Water Reactor (RMWR) [21,22] and the Supercritical Reactor (SCR). [23-25] Most high conversion light water reactors concepts fueled with plutonium in a tight pitch lattice have struggled with insuring a negative void coefficient of reactivity and have had to introduce some mechanical measures to augment neutron leakage effects such as void tubes within the fuel assemblies. One of the motivations for the work here was to investigate whether thorium fuels in an intermediate spectrum possessed inherent neutronics properties that would insure negative void reactivity and thereby obviate the need for any mechanical measures to insure safe reactor operation. The advanced high conversion light water reactor (AHCLWR) design will be introduced in the following section and section 3 will then describe the neutronics properties of thorium-based fuel. The results of calculations with the HELIOS code on tight pitch thorium fueled lattices will be presented in section 4 and conclusions will be provided in section 5. II. Advanced High Conversion Light Water Reactors One of the principal motivations for high conversion light water reactors (HCLWR) [26,27] 3

4 is to enhance the utilization of uranium resources by increasing the conversion of fertile to fissile fuels. High conversion in a LWR is usually obtained by reducing neutron moderation so that the conversion of uranium into plutonium can be enhanced through increased neutron absorption by U 238. During the 1980 s, High Conversion Boiling Water Reactor (HCBWR) studies focused on the extension of standard light water technology toward improved fuel utilization with minimum modification of current LWR system designs. However, research on advanced high conversion light water reactors has been focused more on the enhancement of fuel utilization, on the improvement of reactor safety and economics, and on minimizing the production of long-lived radioactive wastes. Therefore AHCLWRs have adopted more innovative fuel and core design concepts than traditional HCLWR research. One example of a advanced high conversion reactor is the Supercritical Water Reactor (SCR) which operates above the critical pressure and near the pseudocritical temperature. [23] Since supercritical water does not exhibit a change of phase, the concept of boiling does not exist in SCR and it is possible to implement a direct cycle which reduces capital costs and the overall size of the plant. The low-density of supercritical water leads to higher burnup and improved fuel cycle economics. Another example of a AHCLWR is the reduced moderation water reactor (RMWR) that has been developed by JAERI. [21-22] The RMWR is a water-cooled reactor with an intermediate neutron spectrum resulting from reduced neutron moderation because of a tight pitch lattice and a reduced water volume fraction. Because of the intermediate spectrum, the RMWR achieves a considerably higher conversion of U 238 into Pu 239 and thus achieves one of the primary objectives of the AHCLWR. These two examples of AHCLWR have different neutron spectra than a typical LWR or Fast breeder reactor (FBR) as shown in Figure 1. As shown in Figure 1, the thermal flux of the AHCLWR is considerably smaller than the corresponding value for the LWR, but the AHCLWR flux in the intermediate energy region (e.g., from 10 ev to about 30keV) is larger than the values of either the LWR or FBR and thus is referred to as an intermediate spectrum. The very different spectrum of the AHCLWR gives rise to very different neutronic performance than either the LWR or FBR. This will be discussed in detail in section 3. During the past several years, Professor Oka s research group in Japan has led the development of supercritical reactor technology [23-25] and they have proposed several candidate supercritical reactor designs. They have shown that SCR concepts can be implemented with improved power plant economics primarily because of the considerable 4

5 increase in the plant thermal efficiency. However, they have noted some difficulties with reactivity control in the SCR, particularly with the void coefficients of the SCR when it is loaded with plutonium fuel. This has required the use of some innovative mechanical methods such as the use of ZrH layers to insure negative void coefficients. The RMWR designs have also experienced similar problems such as a positive void coefficient in plutonium fueled cores. [21,22] The RMWR also adopted innovations such as a heterogeneous core configuration in order to achieve a negative void coefficient. Specifically, the RMWR uses a double layer pancake type core with 3-region blankets axially to enhance neutron leakage during voiding. The BWR type RMWR also uses void tube assemblies to further enhance leakage as shown in Figure 2. Both the SCR and the RMWR designs satisfy safety concerns by introducing complex configurations both in the fuel assemblies and in the core configuration. However this additional complexity may introduce additional costs and may introduce additional uncertainties in predicting the core behavior. The strong heterogeneities in the fuel assembly and in the core may result in a loss of accuracy when using the current standard nuclear design methodologies. Specifically the streaming effect in the void tube or ZrH layers present challenges to existing neutronics and thermal-hydraulics methods that will require additional methods development and experimental benchmarking before actual implementation. The development of the AHCLWR would be considerably simplified if negative reactivity feedback and core safety could be insured based solely on the neutronics properties of the fuel itself. III. Th 232 and U 238 Fertile Fuel Properties in Intermediate Spectrum LWR The microscopic absorption and fission cross-sections of the two principal fertile isotopes (i.e. Th 232 and U 238 ) are shown in Figures 3 and 4 and their neutronics characteristics are summarized in Table 1. One of the most important differences in the two isotopes is that the resonance integral of Th 232 is about 3 times smaller than that of U 238. This has an important implication for the void reactivity in a tight pitch lattice as will be discussed later. It is important to note that despite the large differences in the resonance integral, the effective resonance capture of both fertile fuels is comparable because resonance self-shielding reduces the effective resonance capture of U 238. [4] The thermal capture cross section of Th 232 is about 3 times larger than the 5

6 corresponding value of U 238 as shown in Table 1 and Figure 3. This leads to much higher fissile enrichment requirements for Th 232 fueled cores in thermal spectrum reactors. To a lesser extent the enrichment requirements are also higher in a tight pitch lattice since the capture cross sections of Th 232 is also slightly larger than those in U 238 in intermediate and fast energy regions as shown in Table 1. Finally, the fast fission cross section of Th 232 is smaller in intermediate spectrum lattices which further reduces initial k inf values in thorium-based fuels. All of these effects are substantial disincentives for using thorium as fertile material instead of U 238 since the fissile enrichment requirements for thorium based fuel are significantly higher in intermediate spectrum reactors. The magnitudes of these effects were quantified by analyzing several fuel pin designs with different pitches and fuel compositions using the HELIOS lattice physics code. The validation of HELIOS for intermediate spectrum LWRs loaded with thorium has been reported in previous work. [28,29] The fuel compositions and geometric properties analyzed here are shown in Table 2, where the total fissile isotope concentration was fixed at 6.0% in all cases. Several fuel types were examined: uranium only (U), uranium/thorium (TU), uranium/plutonium (MOX), and uranium/thorium/ plutonium (TMOX). Two pin geometries are shown for each fuel type, a square pitch corresponding to a current BWR-6 fuel design and a tight pitch hexagonal design corresponding to the proposed AHCLWR. [21] The depletion performance of the different fuel designs and pin types are also summarized in Table 2. In all cases the initial k inf values for the thorium-based pins are smaller than the corresponding U 238 fertile fuel pins because of the larger capture cross section of thorium. In a thermal spectrum fuel pin with U 235 fissile, the penalty is about 0.12% k, whereas in an intermediate spectrum fuel pin with U 235 fissile the penalty is even larger at about 0.14% k. The detailed isotopic reaction rates at the beginning and end of the pin cell burnup cycle are shown in Table 3. The relative reaction rate used in this Table is defined as the isotopic reaction rate per total fuel absorption rate. As shown in Table 3, the fission reaction ratio of U 235 in the square thermal lattice decreases from 51.1 % of U fuel to 48.8% of TU fuel at BOC. This is primarily a result of the incremental increase in the capture rate from replacing the fertile isotopes of U 238 with Th 232. Similar trends are also found in the MOX and TMOX fuels. However, a more important neutronics measure of fuel cycle economics is the conversion ratio, which is typically defined as the ratio of the fissile material bred through 6

7 capture in the fertile to the fissile material consumed. As shown in Table 2, the thoriumbased fuels have a higher conversion ratio than the corresponding cases of the uraniumonly fuel designs (e.g. at BOC the conversion ratio of the thorium square thermal lattice (0.55) is about 20% higher than the uranium only square thermal lattice (0.46)). This is particularly true in the intermediate spectrum fuel pins in which the absorption in fissile materials is reduced but the fertile capture is increased (e.g. at BOC the conversion ratio of the thorium tight pitch lattice (0.83) is about 80% higher than the uranium only square lattice (0.46)). This provides the obvious incentive for the development of an intermediate spectrum fuel pin and for high conversion reactors which can have significantly longer fuel depletion cycles and provide corresponding reductions in power generation costs. The neutronics advantages of thorium-based fuels over uranium-based fuels in an intermediate spectrum fuel pin can be understood using the results shown in Table 3. For example, the improved conversion ratio can be explained by comparing the fissile and fertile neutron captures. As shown in Table 3, the fertile capture in the uranium square pin design at BOC is 29.8% compared to 44.2% at BOC for the thorium tight pitch pin (16.0% U 238 and 28.2% Th 232 ). Conversely, the fission cross section of Th 232 is very small and almost all neutrons absorbed in Th 232 are converted directly to fissile U 233. However, in U 238 about 13% of the neutrons absorbed cause fission. This explains the much larger conversion ratio of the thorium-based tight lattice design. Thorium-based fuel designs also have the potential of enhanced safety characteristics in intermediate spectrum designs. Because of the reduced resonance integral and the higher fast fission threshold in Th 232 compared to U 238, the void coefficient of the thorium pins is larger and negative in the tight lattice cases (e.g., as shown in Table 2, in the poison-free case at EOC, the void coefficient of the thorium tight pitch lattice (-284 pcm) is more negative than the uranium only tight pitch lattice (-125 pcm)). Also shown in Table 2 is that the void coefficient for the tight pitch MOX lattice is positive at EOC (+83 pcm), whereas the void coefficient for the same pin designs loaded with TMOX fuel is negative at EOC (-94 pcm). These behaviors are very important for the utilization of plutonium in tight lattice intermediate spectrum reactors. The reasons for a more negative void coefficient with thorium in a hexagonal tight pitch TMOX fuel pin can be most easily understood by comparing the relative isotopic neutron production rates as shown in Table 4 and by comparing η values of the major fissile isotopes, U 233, U 235 and Pu 239 in each of the neutron spectra as shown in Figure 5. 7

8 The isotopic relative reaction rates used in Table 4 have the same meaning as those used in Table 3. Because the k inf is defined as the neutron production rate per neutron absorption rate, the void coefficient can be estimated by using the change of the relative isotopic neutron production rate. In general, the fertile material contributes to the positive void coefficient and fissile material contributes to the negative void coefficient (e.g. as shown in Table 4, for the square thermal U fuel pin the relative production rate change of fertile and fissile are 3.48% and 22.25%, respectively, resulting in a net change of 18.77%). The primary reasons for positive and negative contributions are the increase of fast fission in the fertile materials and the decrease of thermal and resonance fission in the fissile materials, respectively. Typically, the negative contribution of the fissile material exceeds the positive contribution of fertile material in thermal spectrum reactors and therefore the net void coefficient becomes negative. This is also true in intermediate spectrum lattices fueled with U 235 fissile fuels. However, the void coefficient of MOX fuel may have positive values in an intermediate spectrum during the burnup cycle period because the negative contribution of the fissile material does not overcome the positive effect of fertile materials as shown in Table 4. Since η of Pu 239 increases very quickly from intermediate neutron energy ranges as shown in Figure 5, the spectrum hardening in the intermediate energy region gives rise to an increase in the neutron production rate and thereby the negative contribution of the fissile material becomes smaller in MOX fuel. For example, the negative neutron production rate change of Pu 239 in hexagonal tight lattice is 0.39% but 10.26% in square lattice. So, the net relative neutron production rate change in the hexagonal tight lattice MOX fuel pin becomes positive (e.g. 4.19%) as shown in Table 4. However, by just replacing U 238 with Th 232, a negative void coefficient is maintained throughout the burnup cycle. This occurs because of the decreased positive contribution of fertile materials (e.g. from 5.31% of MOX fuel to 2.11% of TMOX fuel) and because of the increased negative contribution of fissile due to addition negative contribution of U 233 (e.g. from 1.13% of MOX fuel to 7.30% of TMOX fuel). IV. Advanced High Conversion LWRs with Thorium-based Fuels The Reduced Moderation Water Reactor (RMWR) and the Supercritical Reactor (SCR) have been proposed with several different fuel compositions. [21-25] One of the important issues addressed in this paper has been on the use of MOX fuel compositions in BWRs 8

9 because of possible positive void coefficient problems and the potential of using thorium fuel compositions to insure a negative void coefficient. A subsequent paper will provide a more definitive analysis of the safety issues by performing full core analysis of hexagonal tight lattice BWR cores. The purpose of the work here has been to examine in detail the fuel isotopes for which fuel assembly-wise level analysis using the HELIOS code is sufficient. The specific design data for the RMWR and SCR fuel assemblies analyzed in this work are shown in Table 5. A comparison of the assembly reactivity and proliferation resistance properties of the two AHCLWR concepts is summarized in Table 6. In Figure 6, the k inf versus burnup of the RMWR assembly is shown with and without thorium. In this Figure, RMWR and RMWR-Void denote the original fuels proposed in Ref. 21 without and with void tubes within the fuel assembly, respectively. Also in Figure 6, TPu denotes the thorium-based fuel, which just replaces uranium with thorium in the RMWR. The 9.4% and 11.0% fissile denotes the total amount of fissile isotopes initially loaded in the RMWR. The same information is shown in Figure 7 for the SCR. As discussed in the previous sections, the initial k inf values of thorium-based fuels are smaller than the corresponding values of MOX fuels and their critical burnups are also shorter for the thorium-based fuels with the same fissile loading. Therefore, in order to achieve the same cycle length as MOX fuel, thorium-based fuels require additional fissile inventory or the use of innovative fuel configurations such as micro-heterogeneity. [30] In either case the use of thorium-based fuels can not be justified on the basis of improved fuel cycle economics alone. However, because of the higher conversion ratio of thoriumbased fuels as shown in Table 6, the incremental increase in fissile inventory required to achieve parity in cycle length is not excessive. As shown in Figures 6 and 7, the enrichment increases from 9.4% fissile to 11% fissile in the RMWR and from 10% to 11.1% in the SCR to achieve the same burnup with thorium as with standard MOX cores. This is only a 17% and 11% change in fissile inventory for RMWR and SCR, respectively. On the other hand, the attraction of thorium fuels is more because of improved reactivity control performance (e.g. reduced burnup reactivity swing), the guarantee of a negative void coefficient, and the enhanced proliferation resistance properties. Because of the greater conversion ratios of thorium-based fuels, the reactivity swings for the AHCLWR designs (18.9% and 15.4% for the RMWR and SCR, respectively) are smaller than the values for MOX fuels (e.g. 22.4% and 18.5% for RMWR and SCR, respectively) as shown in Table 6. The smaller burnup reactivity swing in thorium-based fuels is attractive from a 9

10 safety standpoint because of the reduced reactivity control burden and the possibility for an increased shutdown margin. However, an even more attractive reactor safety advantage is the improved void coefficient in the RMWR and SCR when they are fueled with thorium-based fuels. The void coefficients of the RMWR with MOX and TPu fuels as a function of fuel burnup are shown in Figure 8. Even though the void coefficient of a whole core calculation may be negative due to the axial leakage effect, [21] both the void coefficient of the RMWR and RMWR-Void become positive in the assembly level calculations as shown in Figure 8. However, the thorium based fuel maintains a negative void coefficient throughout the burnup cycle. Similar results for the SCR are shown in Figure 9, a negative void coefficient is observed in the thorium-based fuel and a positive void coefficient in the MOX fuel. In general, the use of thorium-based fuels will provide a more negative void coefficient in tight pitch lattice cores and obviate the need for innovative methods to insure a negative void coefficient in uranium based cores. An additional attraction of thorium-based fuels is the enhancement of proliferation resistance properties, as well as a reduction in the amount of the long-term toxicity isotopes such as Np 237. The proliferation resistance properties of plutonium are usually measured by the spontaneous fission rate (SFR) and specific decay heat (SDH). [31] Table 6 indicates that in the RMWR, thorium based fuels increase the SFR by as much as 56% (e.g. from 450.5/g.sec to 703.6/g.sec in SCR) and the SDH by as much as 49% (e.g. from 18.5 W/kg to 27.5 W/kg in RMWR). An additional important metric for proliferation resistance is the amount of U 232 in the fuel [32] because it is a precursor of the high energy gamma emitter, Tl 208. Only thorium based fuels have the possibility of producing U 232 and meeting the IAEA standards for fuel self protection. [32] Finally, thorium based fuels provide a more attractive waste form because there is a substantial reduction in the amount of long lived isotopes produced in the thorium burnup cycle because long-term toxicity isotopes such as Np 237 and also because thorium oxide appears to provide a chemically more stable waste form than uranium oxide. [34] V. Conclusion and Continuing Work Research on the utilization of thorium-based fuels in the intermediate spectrum of a tight pitch lattice LWR has been discussed in this paper. The analysis was performed using the Studsvik/Scandpower lattice physics code HELIOS. The results show that thoriumbased fuels in an intermediate spectrum LWR have considerable advantages in terms of 10

11 conversion ratio, reactivity control, non-proliferation characteristics and a reduced production of long-lived radiotoxic wastes. Due to the high conversion ratio of thoriumbased fuels in intermediate spectrum reactors, the total fissile inventory required to achieve a given fuel burnup is only 11% to 17% higher than that of U 238 fertile fuels. However, unlike U 238 fertile fuels, the void reactivity coefficient with thorium-based fuels is negative in an intermediate spectrum reactor. This provides motivation for replacing U 238 with Th 232 in advanced high conversion LWRs (AHCLWR), such as the Reduced-moderator reactor (RMWR) or the Supercritical reactor (SCR). While the work here has shown there is motivation for considering thorium based fuels in tight pitch lattice intermediate spectrum LWRs, the calculations performed here used only fuel assembly models and therefore provide only preliminary confirmation of the potential benefits. Full three-dimensional, multi-group core calculations must be performed with well-benchmarked core simulators to verify the behavior of AHCLWRs fueled with thorium. These calculations should include such transients as LOCA which is a particular concern in tight lattice reactors. [26,27,33] Such analysis is the logical next step in thorium research and is the focus of ongoing DOE sponsored research at Purdue. [35] 11

12 References 1. W. J. Oosterkamp, The Potential of the Thorium Cycle in PWRs, Annals of Nuclear Energy, Vol (1978). 2. D. B. Trauger, Thorium Utilization, Annals of Nuclear Energy, Vol. 5, (1978). 3. S. M. McDeavitt, et al, Fuel for a Once-Through Cycle-(Th,U)O2, in a Metal Matrix, Proposal for NERI project. 4. N.L. Shapiro, et al, Assessment of Thorium Fuel Cycles in Pressurized Water Reactors, EPRI NP-359, Final Report, (February 1977). 5. P. R. Kastern, Assessment of the thorium fuel cycle in power reactors, ORNL/TM-5565, Oak Ridge National Laboratory, (1977). 6. Advanced Fuel Cycle and Reactor Concepts, Report of INFCE working group 8, IAEA, Vienna (1980). 7. Thorium-Based Nuclear Fuel: Current Status and Perspectives, Proceedings of a Technical Committee meeting on utilization of Thorium-based nuclear fuel, IEAE- TECDOC-412, IAEA, Vienna, 2-4, December, T. Nishigori, et al, Some Nuclear Characteristics of Thorium Fueled Light Water Reactors, Journal of Nuclear Science and Technology, Vol. 25, ,(1988). 9. S. Gungor, Thorium Utilization in Unmodified Pressurized Water Reactor, Annals of Nuclear Energy, Vol. 17, (1990) 10. A. Galperin, Utilization of Thorium in Light Water Reactors, Nuclear Science and Engineering, Vol. 86, 112(1994) 11. A. Radkowsky, et al, The Nonproliferative Light Water Thorium Reactor: A New Approach to Light Water Reactor Core Technology, Nuclear Technology. Vol. 124, 215 (1998). 12. H. K. Joo, T. K. Kim, et al, Potential of Thorium-based Fuel Cycle for PWR Core to Reduce Plutonium and Long-Term Toxicity, KAERI/TR-1208/99, Korea Atomic Energy Research Institute (1999). 13. A. Galperin, et al, A Pressurized Water Reactor Design for Plutonium Incineration: Fuel Cycle Options, Nuclear Technology, Vol. 117, 125(1997) 14. Xianfeng Zhao, et al, Rationale for Reconsidering the Thorium Cycle in Light Water Reactors, Transaction of American Nuclear Society, 80, 43 (June, 1999). 12

13 15. J. Stephen Herring, et al, Advanced. Lower-Cost, Proliferation-Resistant, Uranium- Thorium Dioxide Fuels for LWRs, Transaction of American Nuclear Society, 80, 45 (June, 1999). 16. M. Todosow, et al, A Novel Nonproliferation Thorium-Based Seed-Blanket Fuel Concept for PWRs, Transaction of American Nuclear Society, 80, 46 (June, 1999). 17. Marc Caner, et al, ThO 2 -UO 2 Annular Pins for High Burnup Fuels, Annals of Nuclear Energy, vol. 27, (2000). 18. Advanced Proliferation Resistant, Lower Cost, Uranium-Thorium Dioxide Fuels for Light Water Reactors, INEEL/EXT , Progress Report of Nuclear Energy Research Initiative, (2000). 19. T. Downar, et al, Feasibility Study of a Plutonium-Thorium Fuel Cycle for a High Conversion Boiling Water Reactor, Transaction of American Nuclear Society, 83, 192 (Nov., 2000). 20. T. K. Kim, et al, Thorium fuel in Tight Pitch LWR Lattice, Transaction of American Nuclear Society, (June, 2001, to be published). 21. T. Iwamura, et al, Research on Reduced-Moderator Water Reactor (RMWR), JAERIresearch T. Okubo, Conceptual Designing of Reduced-Moderation Water Reactors (1) Design for BWR-Type Reactors, ICONE-8, Baltimore, USA (2000). 23. Y. Oka, Concept and Design of a Supercritical Pressure, Direct-cycle Light Water Reactor, Nuclear Technology, Vol 103, 295 (1993). 24. Y. Oka, System Design of Direct-cycle Supercritical-Water-Cooled Fast Reactors, Nuclear Technology, Vol 109, 295 (1995). 25. Y. Oka, et al, UO2 Core Design of a Direct-Cycle Fast Converter Reactor Cooled by Supercritical Water, Nuclear Technology, Vol. 114, 273(1996). 26. Eturo Saji, et al, Feasibility Studies on High Conversion Pressurized Water Reactors with Semitight Core Configurations, Nuclear Technology, Vol. 80, 18 (1988). 27. Junich Yamashita, et al, Development of a High-Conversion Boiling Water Reactor, Nuclear Technology, Vol. 96, 20 (1991). 28. Xianfen Zhao, et al, Comparison of Code Results for PWR Thorium/Uranium Pin Cell Burnup, MIT-NFC-TR-027, Center for Advanced Nuclear Systems, MIT(2000). 29. T. Donwar, et al, A Proliferation Resistant Hexagonal Tight Lattice BWR fueled Core for Increased Burnup and Reduced Fuel Storage Requirement, Proposal for NERI 13

14 project. 30. D. Wang, et al, Design and Performance Assessment of a PWR Whole-Assembly Seed and Blanket Thorium Based Fuel Cycle, MIT-NFC-TR-026, Center for Advanced Nuclear Systems, MIT(2000). 31. J. Carson Mark, Explosive Properties of Reactor-Grade Plutonium, Science and Global Security, Vol. 4, (1993). 32. J. M. Kang, et al, U 232 and the Proliferation-Resistance of U 233 in Spent Fuel, Science and Global Security (accepted). 33. V. O. Uotinen, et al, Technical Feasibility of Pressurized Water Reactor Design with a Low-Water-Volume-Fraction Lattice, NP-1833, Babcock & Wilcox Company (1981). 34. J. Tulenko, et al., DOE Nuclear Engineering Research Initiative (NERI), DE-FG03-99SF21913, T. Downar, et al., DOE Nuclear Engineering Research Initiative (NERI), DE-FG03-99SF21890,

15 Table 1 Summary of neutronics properties of Th 232 and U 238 Capture Fission U 238 Th 232 Maxwell avg. at ev b b Resonance integral b b Fission spectrum average mb mb Maxwell avg. at ev µb ~0 Resonance integral b ~0 Fission spectrum average mb mb *) Reference : Table of Nuclides, Korea Atomic Energy Research Institute, Table 2 Several Fuel Pin Compositions and Nuclear Characteristics (BOC/EOC a) ) Fuel Type U TU MOX TMOX Fuel Composition UO 2 (Th+U)O 2 MOX b) 6% U % Th, 6% U 235 6% fissile ThO 2 + MOX b) 65% Th, 6% fissile V m /V f c) Pin geometry Square Hexagonal Square Hexagonal Square Hexagonal Square Hexagonal Initial k inf Conversion ratio 0.46/ / / / / / / /0.99 Void Coefficient (pcm) -396/ / / / / / / / -94 a) EOC burnup = 50 GWD/t b) Pu vector (pu238/pu239/pu240/pu241/pu242) = 1/59/21/14/5 c) V m /V f = is similar to BWR-6(J.J. Duderstadt, Nuclear Reactor Analysis, John Wiley & Sons, Inc. 1976) fuel pins and is intermediate spectrum BWR. Here, V f includes the volume of fuel and clad. 15

16 Table 3 Competition of the isotopic relative reaction rate in %, (BOC/EOC) Fuel Type U TU MOX TMOX Pin geometry Square Hexagonal Square Hexagonal Square Hexagonal Square Hexagonal Cap. 21.8/ / / /23.4 Th 232 Fis. 0.7/ / / / 0.9 Fertile Cap 29.8/ / / / / / / /12.4 U 238 Fis 4.5/ / / / / / / / 1.36 Cap - / / / / / / / / 6.4 Pu 240 Fis - / / / / / / / / 0.6 Cap - / / / / 1.9 U 233 Fis - / / / /12.7 Fissile Cap 14.3/ / / / / / / / 0.4 U 235 Fis 51.1/ / / / / / / / 0.5 Cap - / / / / / / / / 6.7 Pu 239 Fis - / / / / / / / /12.0 Cap - / / / / / / / / 1.3 Pu 241 Fis - / / / / / / / / 5.0 a) Isotopic relative reaction ratio is defined by the isotopic reaction rate per total fuel absorption reaction rate: isotope fuel Cap. = 100 Σ c Σ, Fis. = a 100, isotope fuel Σ f Σa Table 4 Comparison of the isotopic relative neutron production rate a) change from 40% to 80% voided at EOC (unit: %). U TU MOX TMOX Pin geometry Square Hexagonal Square Hexagonal Square Hexagonal Square Hexagonal Th U U U Pu Pu Pu

17 others Net a) Isotopic production rate = 100 isotope fuel υσ f Σa Table 5 Primary design characteristics of advanced high conversion LWRs RMWR a) SCR b) Fuel type MOX MOX Thermal power (MWt) Core equivalent diameter (m) ~ Core Active height (m) Number of Non-void assembly Number of Void assembly /102(inner/outer) - Assembly Shape Hexagonal Hexagonal Fuel Density 90% TD 90% TD Fuel pellet diameter (cm) Fuel cladding thickness (cm) Fuel rod pitch (cm) Assembly pitch (cm) Vm/Vf in fuel cell Average fissile content (wt %) /15.4(inner/outer) Specific power density (W/g) Average linear power (W/cm) ~ a) Ref. 21 & 22 b) Ref

18 Table 6 Comparison of reactivity and proliferation resistance properties RMWR SCR MOX (Th+Pu)O 2 MOX (Th+Pu)O 2 Initial fissile content (%) Conversion ratio (BOC/EOC a) ) 0.67/ / / /0.81 Reactivity Swing (%) b) Spontaneous fission rate (#/g.sec) Specific decay heat (W/kg) U 232 contamination (%) c) Np 237 weight percent (%) d) a) EOC= 50GWD/t b) Reactivity swing = 100x(1-1/k BOC ) c) U 232 contamination = 100 x mass of U 232 / mass of total heavy metal d) Np237 does not include its precursor. 18

19 FBR Normalized Eφ(E) BWR RMWR SCR Neutron energy(ev) Figure 1 Comparison of AHCLWR spectra for typical BWR and FBR 19

20 (a) Non-Void Fuel Assembly (b) Void Fuel Assembly Figure 2 Fuel Assemblies design by JAERI for RMWR 21) 20

21 Th 232 U 238 Th 232 U 238 Figure 3 Comparison of absorption cross section of Th 232 and U

22 U 238 Th 232 Figure 4 Microscopic fission cross section of Th 232 and U

23 intermediate Spectrum Normalized Eφ(E) Thermal Spectrum η of Pu 239 η of U 233 η of U η Energy(eV) Figure 5 Comparison of spectrum and h values 23

24 Infinite multiplication factor RMWR, 9.4% fissile RMWR-Void, 9.4% fissile TPu, 11.0% fissile TPu, 9.4% fissile Burnup(GWD/t) Figure 6 Infinite multiplication factor behavior of RMWR with/without Thorium 24

25 1.30 Ininite Multiplication Factor SCR, 10.0% fissile TPu, 11.1% fissile TPu, 10.0% fissile Burnup(GWD/t) Figure 7. Infinite multiplication behavior of SCR with/without Thorium 25

26 Void coefficient (pcm) RMWR, 9.4% fissile RMWR-Void, 9.4% fissile TPu, 11.0% fissile Burnup(GWD/t) Figure 8. Void Coefficient behavior of RMWR with/without Thorium 26

27 60 Void Coefficient(pcm) SCR, 10.0% fissile TPu, 11.1% fissile TPu, 10.0% fissile Burnup(GWD/t) Figure 9. Void Coefficient behavior of SCR with/without Thorium 27