Eugene Shwageraus - Academia.edu (original) (raw)

Papers by Eugene Shwageraus

Progress in Nuclear Energy, 2011

This scoping study proposes using mixed nitride fuel in Pu-based high conversion LWR designs in o... more This scoping study proposes using mixed nitride fuel in Pu-based high conversion LWR designs in order to increase the breeding ratio. The higher density fuel reduces the hydrogen-to-heavy metal ratio in the reactor which results in a harder spectrum in which breeding is more effective. A Resource-renewable Boiling Water Reactor (RBWR) assembly was modeled in MCNP to demonstrate this effect in a typical high conversion LWR design. It was determined that changing the fuel from (U,TRU)O 2 to (U,TRU)N in the assembly can increase its fissile inventory ratio (fissile Pu mass divided by initial fissile Pu mass) from 1.04 to up to 1.17.

The objective of this work was to assess the potential of thorium based fuel to minimise Pu and M... more The objective of this work was to assess the potential of thorium based fuel to minimise Pu and MA production in Pressurised Water Reactors (PWRs). The assessment was carried out by examining destruction rates and residual amounts of Pu and MA in the fuel used for transmutation. In particular, sensitivity of these two parameters to the fuel lattice Hydrogen to Heavy Metal (H/HM) ratio and to the fuel composition was systematically investigated. All burn-up calculations were performed using CASMO4 -the fuel assembly burn-up code. The results indicate that up to 1 000 kg of reactor grade Pu can potentially be burned in thorium based fuel assemblies per GW e Year. Up to 75% of initial Pu can be destroyed per path. Addition of MA to the fuel mixture degrades the burning efficiency. The theoretically achievable limit for total TRU destruction per path is 50%. Efficient MA and Pu destruction in thorium based fuel generally requires a higher degree of neutron moderation and, therefore, higher fuel lattice H/HM ratio than typically used in the current generation of PWRs. Reactivity coefficients evaluation demonstrated the feasibility of designing a Th-Pu-MA fuel with negative Doppler and moderator temperature coefficients.

One of the main goals of Generation IV nuclear energy systems is to minimize and manage nuclear w... more One of the main goals of Generation IV nuclear energy systems is to minimize and manage nuclear wastes through partitioning and transmutation of transuranic elements (TRU).

ABSTRACT The once through nuclear fuel cycle adopted by the majority of countries with operating ... more ABSTRACT The once through nuclear fuel cycle adopted by the majority of countries with operating commercial power reactors imposes a number of concerns. The radioactive waste created in the once through nuclear fuel cycle has to be isolated from the environment for thousands of years. In addition, plutonium and other actinides, after the decay of fission products, could become targets for weapon proliferators. Furthermore, only a small fraction of the energy potential in the fuel is being used. All these concerns can be addressed if a closed fuel cycle strategy is considered offering the possibility for partitioning and transmutation of long lived radioactive waste, enhanced proliferation resistance, and improved utilization of natural resources. It is generally believed that dedicated advanced reactor systems have to be designed in order to perform the task of nuclear waste transmutation effectively. The development and deployment of such innovative systems is technically and economically challenging. In this thesis, a possibility of constraining the generation of long lived radioactive waste through multi-recycling of Trans-uranic actinides (TRU) in existing Light Water Reactors (LWR has been studied. Thorium based and fertile free fuels (FFF) were analyzed as the most attractive candidates for TRU burning in LWRs. Although both fuel types can destroy TRU at comparable rates (about 1150 kg/GWe-Year in FFF and up to 900 kg/GWe-Year in Th) and achieve comparable fractional TRU burnup (close to 50a/o), the Th fuel requires significantly higher neutron moderation than practically feasible in a typical LWR lattice to achieve such performance. (Cont.) On the other hand, the FFF exhibits nearly optimal TRU destruction performance in a typical LWR fuel lattice geometry. Increased TRU presence in LWR core leads to neutron spectrum hardening, which results in reduced control materials reactivity worth. The magnitude of this reduction is directly related to the amount of TRU in the core. A potential for positive void reactivity feedback limits the maximum TRU loading. Th and conventional mixed oxide (MOX) fuels require higher than FFF TRU loading to sustain a standard 18 fuel cycle length due to neutron captures in Th232 and U238 respectively. Therefore, TRU containing Th and U cores have lower control materials worth and greater potential for a positive void coefficient than FFF core. However, the significantly reduced fuel Doppler coefficient of the fully FFF loaded core and the lower delayed neutron fraction lead to questions about the FFF performance in reactivity initiated accidents. The Combined Non-Fertile and UO2 (CONFU) assembly concept is proposed for multi- recycling of TRU in existing PWRs. The assembly assumes a heterogeneous structure where about 20% of the UO2 fuel pins on the assembly periphery are replaced with FFF pins hosting TRU generated in the previous cycle. The possibility of achieving zero TRU net is demonstrated. The concept takes advantage of superior TRU destruction performance in FFF allowing minimization of TRU inventory. At the same time, the core physics is still dominated by UO2 fuel allowing maintenance of core safety and control characteristics comparable to all-UO2. Thesis (Ph. D.)--Massachusetts Institute of Technology, Dept. of Nuclear Engineering, February 2004. Includes bibliographical references (p. 249-262).

The COmbined Non-Fertile and UO 2 (CONFU) fuel system for sustainable actinide management has the... more The COmbined Non-Fertile and UO 2 (CONFU) fuel system for sustainable actinide management has the potential to reduce accumulation of actinides in the next 50 years under the assumption of a growing worldwide nuclear energy power demand. The system, which is compatible with currently operating Light Water Reactors (LWRs), employs CONFU fuel assemblies where about 20% of the UO 2 fuel pins are replaced with fertile free fuel (FFF) hosting the TRU.

ABSTRACT Response of a PWR core loaded with Combined Non-Fertile and UO{sub 2} (CONFU) fuel assem... more ABSTRACT Response of a PWR core loaded with Combined Non-Fertile and UO{sub 2} (CONFU) fuel assemblies to control rod ejection accident was evaluated. A number of core arrangements and TRU fuel compositions were considered and the results were compared with the performance of a reference all-UO{sub 2} core. The comparison was based on the results of a simple point kinetics model with thermal reactivity feedbacks and temperature dependant materials properties. The reactivity coefficients and core average kinetics parameters were obtained from the full core 3D neutronic simulations. The results show that application of the CONFU assembly concept causes only minor deviation of fuel performance characteristics in reactivity initiated accidents. This is a consequence of relatively small loadings of TRU in the CONFU assembly and therefore dominating role of conventional UO{sub 2} fuel in the neutronic performance of the core. (authors)

Nuclear Instruments and Methods in Physics Research Section A Accelerators Spectrometers Detectors and Associated Equipment

There is a growing interest in using Am as a nuclear fuel. The advantages of Am as a nuclear fuel... more There is a growing interest in using Am as a nuclear fuel. The advantages of Am as a nuclear fuel derive from the fact that Am has the highest thermal "ssion cross section. The thermal capture cross section is relatively low and the number of neutrons per thermal "ssion is high. These nuclear properties make it possible to obtain nuclear criticality with ultra-thin fuel elements. The possibility of having ultra-thin fuel elements enables the use of these "ssion products directly, without the necessity of converting their energy to heat, as is done in conventional reactors. There are three options of using such highly energetic and highly ionized "ssion products.

ABSTRACT A thorium-based fuel cycle for light water reactors will reduce the plutonium generation... more ABSTRACT A thorium-based fuel cycle for light water reactors will reduce the plutonium generation rate and enhance the proliferation resistance of the spent fuel. However, priming the thorium cycle with 235U is necessary, and the 235U fraction in the uranium must be limited to below 20% to minimize proliferation concerns. Thus, a once-through thorium-uranium dioxide (ThO2-UO2) fuel cycle of no less than 25% uranium becomes necessary for normal pressurized water reactor (PWR) operating cycle lengths. Spatial separation of the uranium and thorium parts of the fuel can improve the achievable burnup of the thorium-uranium fuel designs through more effective breeding of 233U from the 232Th. Focus is on microheterogeneous fuel designs for PWRs, where the spatial separation of the uranium and thorium is on the order of a few millimetres to a few centimetres, including duplex pellet, axially microheterogeneous fuel, and a checkerboard of uranium and thorium pins. A special effort was made to understand the underlying reactor physics mechanisms responsible for enhancing the achievable burnup at spatial separation of the two fuels. The neutron spectral shift was identified as the primary reason for the enhancement of burnup capabilities. Mutual resonance shielding of uranium and thorium is also a factor; however, it is small in magnitude. It is shown that the microheterogeneous fuel can achieve higher burnups, by up to 15%, than the reference all-uranium fuel. However, denaturing of the 233U in the thorium portion of the fuel with small amounts of uranium significantly impairs this enhancement. The denaturing is also necessary to meet conventional PWR thermal limits by improving the power share of the thorium region at the beginning of fuel irradiation. Meeting thermal-hydraulic design requirements by some of the microheterogeneous fuels while still meeting or exceeding the burnup of the all-uranium case is shown to be potentially feasible. However, the large power imbalance between the uranium and thorium regions creates several design challenges, such as higher fission gas release and cladding temperature gradients. A reduction of plutonium generation by a factor of 3 in comparison with all-uranium PWR fuel using the same initial 235U content was estimated. In contrast to homogeneously mixed U-Th fuel, microheterogeneous fuel has a potential for economic performance comparable to the all-UO2 fuel provided that the microheterogeneous fuel incremental manufacturing costs are negligibly small.

This study explores the basic possibility of achieving a self-sustainable Th-233 U fuel cycle tha... more This study explores the basic possibility of achieving a self-sustainable Th-233 U fuel cycle that can be adopted in the current generation of Pressurized Water Reactors. The study outlines possible core design strategies of a typical PWR in order to achieve Breeding Ratio (BR) close to unity. Major design tradeoffs in the core design are discussed. Preliminary neutronic analysis performed on the fuel assembly level with BOXER computer code suggests that net breeding of 233 U is feasible in principle within a typical PWR operating envelope. However, some reduction in the core power density and/or shorter than typical fuel cycle length would most likely be required in order to achieve such performance.

ABSTRACT A new combined Non Fertile and Uranium (CONFU) fuel assembly is proposed to limit the ac... more ABSTRACT A new combined Non Fertile and Uranium (CONFU) fuel assembly is proposed to limit the actinides that need long-term high-level waste storage from the pressurized water reactor (PWR) fuel cycle. In the CONFU assembly concept, ∼20% of the UO2 fuel pins are replaced with fertile free fuel hosting the transuranic elements (TRUs) generated in the previous cycle. This leads to a fuel cycle sustainable with respect to net TRU generation, and the amount and radiotoxicity of the nuclear waste can be significantly reduced in comparison with the conventional once-through UO2 fuel cycle. It is shown that under the constraints of acceptable power peaking limits, the CONFU assembly exhibits negative reactivity feedback coefficients comparable in values to those of the reference UO2 fuel. Feasibility of the PWR core operation and control with complete TRU recycle has been shown based on full-core three-dimensional neutronic simulation. However, gradual buildup of small amounts of Cm and Cf challenges fuel reprocessing and fabrication due to the high spontaneous fission rates of these nuclides and heat generation by some Pu, Am, and Cm isotopes. Feasibility of the processing steps becomes more attainable if the time between discharge and reprocessing is 20 yr or longer.

ABSTRACT The homogeneous ThO2-UO2 fuel cycle option for a pressurized water reactor (PWR) of curr... more ABSTRACT The homogeneous ThO2-UO2 fuel cycle option for a pressurized water reactor (PWR) of current technology is investigated. The fuel cycle assessment was carried out by calculating the main performance parameters: natural uranium and separative work requirements, fuel cycle cost, and proliferation potential of the spent fuel. These performance parameters were compared with a corresponding slightly enriched (all-U) fuel cycle applied to a PWR of current technology. The main conclusion derived from this comparison is that fuel cycle requirements and fuel cycle cost for the mixed Th/U fuel are higher in comparison with those of the all-U fuel. A comparison and analysis of the quantity and isotopic composition of discharged Pu indicate that the Th/U fuel cycle provides only a moderate improvement of the proliferation resistance. Thus, the overall conclusion of the investigation is that there is no economic justification to introduce Th into a light water reactor fuel cycle as a homogeneous ThO2-UO2 mixture.

ABSTRACT Coupled Monte Carlo depletion systems provide a versatile and an accurate tool for analy... more ABSTRACT Coupled Monte Carlo depletion systems provide a versatile and an accurate tool for analyzing advanced thermal and fast reactor designs for a variety of fuel compositions and geometries. The main drawback of Monte Carlo-based systems is a long calculation time imposing significant restrictions on the complexity and amount of design-oriented calculations. This paper presents an alternative approach to interfacing the Monte Carlo and depletion modules aimed at addressing this problem. The main idea is to calculate the one-group cross sections for all relevant isotopes required by the depletion module in a separate module external to Monte Carlo calculations. Thus, the Monte Carlo module will produce the criticality and neutron spectrum only, without tallying of the individual isotope reaction rates. The onegroup cross section for all isotopes will be generated in a separate module by collapsing a universal multigroup (MG) cross-section library using the Monte Carlo calculated flux. Here, the term "universal" means that a single MG cross-section set will be applicable for all reactor systems and is independent of reactor characteristics such as a neutron spectrum; fuel composition; and fuel cell, assembly, and core geometries. This approach was originally proposed by Haeck et al. and implemented in the ALEPH code. Implementation of the proposed approach to Monte Carlo burnup interfacing was carried out through the BGCORE system. One-group cross sections generated by the BGCORE system were compared with those tallied directly by the MCNP code. Analysis of this comparison was carried out and led to the conclusion that in order to achieve the accuracy required for a reliable core and fuel cycle analysis, accounting for the background cross section (σ0) in the unresolved resonance energy region is essential. An extension of the one-group cross-section generation model was implemented and tested by tabulating and interpolating by a simplified σ0 model. A significant improvement of the one-group cross-section accuracy was demonstrated.

ABSTRACT This paper presents results of a feasibility study aimed at developing a zero-transurani... more ABSTRACT This paper presents results of a feasibility study aimed at developing a zero-transuranic-discharge fuel cycle based on the U-Th-TRU ternary cycle. The design objective is to find a fuel composition (mixture of thorium, enriched uranium, and recycled transuranic components) and fuel management strategy resulting in an equilibrium charge-discharge mass flow. In such a fuel cycle scheme, the quantity and isotopic vector of the transuranium (TRU) component is identical at the charge and discharge time points, thus allowing the whole amount of the TRU at the end of the fuel irradiation period to be separated and reloaded into the following cycle. The TRU reprocessing activity losses are the only waste stream that will require permanent geological storage, virtually eliminating the long-term radiological waste of the commercial nuclear fuel cycle. A detailed three-dimensional full pressurized water reactor (PWR) core model was used to analyze the proposed fuel composition and management strategy. The results demonstrate the neutronic feasibility of the fuel cycle with zero-TRU discharge. The amount of TRU and enriched uranium loaded reach equilibrium after about four TRU recycles. The reactivity coefficients were found to be within a range typical for a reference PWR core. The soluble boron worth is reduced by a factor of ∼2 from a typical PWR value. Nevertheless, the results indicate the feasibility of an 18-month fuel cycle design with an acceptable beginning-of-cycle soluble boron concentration even without application of burnable poisons.

ABSTRACT The growing interest in innovative reactors and advanced fuel cycle designs requires mor... more ABSTRACT The growing interest in innovative reactors and advanced fuel cycle designs requires more accurate prediction of various transuranic actinide concentrations during irradiation or following discharge because of their effect on reactivity or spent-fuel emissions, such as gamma and neutron activity and decay heat. In this respect, many of the important actinides originate from the 241Am(n,γ) reaction, which leads to either the ground or the metastable state of 242Am. The branching ratio for this reaction depends on the incident neutron energy and has very large uncertainty in the current evaluated nuclear data files. This study examines the effect of accounting for the energy dependence of the 241Am(n,γ) reaction branching ratio calculated from different evaluated data files for different reactor and fuel types on the reactivity and concentrations of some important actinides. The results of the study confirm that the uncertainty in knowing the 241Am(n,γ) reaction branching ratio has a negligible effect on the characteristics of conventional light water reactor fuel. However, in advanced reactors with large loadings of actinides in general, and 241Am in particular, the branching ratio data calculated from the different data files may lead to significant differences in the prediction of the fuel criticality and isotopic composition. Moreover, it was found that neutron energy spectrum weighting of the branching ratio in each analyzed case is particularly important and may result in up to a factor of 2 difference in the branching ratio value. Currently, most of the neutronic codes have a single branching ratio value in their data libraries, which is sometimes difficult or impossible to update in accordance with the neutron spectrum shape for the analyzed system.

ABSTRACT There is a growing need for very small nuclear reactors for space applications and as po... more ABSTRACT There is a growing need for very small nuclear reactors for space applications and as portable high-intensity neutron sources. This technical note investigates the question of what is the smallest possible thermal reactor. It was found that the smallest reactor is a spherically shaped solution of 242mAm(NO3)3 in water. The weight of such a reactor is 4.95 kg with 0.7 kg of 242mAm nuclear fuel. The radius of the reactor in this case is 9.6 cm.

ABSTRACT BGCore is a software package for comprehensive computer simulation of nuclear reactor sy... more ABSTRACT BGCore is a software package for comprehensive computer simulation of nuclear reactor systems and their fuel cycles. The BGCore interfaces Monte Carlo particles transport code MCNP4C with a SARAF module - an independently developed code for calculating in-core fuel composition and spent fuel emissions following discharge. In BGCore system, depletion coupling methodology is based on the multi-group approach that significantly reduces computation time and allows tracking of large number of nuclides during calculations. In this study, burnup calculation capabilities of BGCore system were validated against well established and verified, computer codes for thermal and fast spectrum lattices. Very good agreement in k eigenvalue and nuclide densities prediction was observed for all cases under consideration. In addition, decay heat prediction capabilities of the BGCore system were benchmarked against the most recent edition of ANS Standard methodology for UO2 fuel decay power prediction in LWRs. It was found that the difference between ANS standard data and that predicted by the BGCore does not exceed 5%.

ABSTRACT This paper investigates the basic feasibility of using reactor-grade Pu in fertile-free ... more ABSTRACT This paper investigates the basic feasibility of using reactor-grade Pu in fertile-free fuel (FFF) matrix in pressurized water reactors (PWRs). Several important issues were investigated in this work: the Pu loading required to achieve a specific interrefueling interval, the impact of inert matrix composition on reactivity constrained length of cycle, and the potential of utilizing burnable poisons (BPs) to alleviate degradation of the reactivity control mechanism and temperature coefficients. Although the subject was addressed in the past, no systematic approach for assessment of BP utilization in FFF cores was published. In this work, we examine all commercially available BP materials in all geometrical arrangements currently used by the nuclear industry with regards to their potential to alleviate the problems associated with the use of FFF in PWRs. The recently proposed MgO-ZrO2 solid-state solution fuel matrix, which appears to be very promising in terms of thermal properties and radiation damage resistance, was used as a reference matrix material in this work. The neutronic impact of the relative amounts of MgO and ZrO2 in the matrix were also studied. The analysis was performed with a neutron transport and fuel assembly burnup code BOXER. A modified linear reactivity model was applied to the two-dimensional single fuel assembly results to approximate the full core characteristics. Based on the results of the performed analyses, the Pu-loaded FFF core demonstrated potential feasibility to be used in existing PWRs. Major FFF core design problems may be significantly mitigated through the correct choice of BP design. It was found that a combination of BP materials and geometries may be required to meet all FFF design goals. The use of enriched (in most effective isotope) BPs, such as 167Er and 157Gd, may further improve the BP effectiveness and reduce the fuel cycle length penalty associated with their use.

ABSTRACT Up to 50% increase in the power density of the existing pressurized water reactor (PWR)-... more ABSTRACT Up to 50% increase in the power density of the existing pressurized water reactor (PWR)-type reactors can be achieved by the use of internally and externally cooled annular fuel geometry. As a result, the accumulated stock-piles of Pu, especially if incorporated infertile-free inert matrix, can be burnt at a substantially higher rate as compared with the conventional mixed oxide-fueled reactors operating at standard power density. In this work, we explore the basic feasibility of a PWR core fully loaded with Pu incorporated infertile-free fuel of annular internally and externally cooled geometry and operating at 150% of nominal power density. We evaluate basic burnable poison designs, fuel management strategies, and reactivity feedback coefficients. The three-dimensional full core neutronic analysis performed with Studsvik Core Management System showed that the design of such a Pu-loaded annular fuel core is feasible but significantly more challenging than the Pu fertile-free core with solid fuel pins operating at nominal power density. The main difficulty arises from the fact that the annular fuel core requires at least 50% higher initial Pu loading in order to maintain the standard fuel cycle length of 18 months. Such a high Pu loading results in hardening of the neutron spectrum and consequent reduction in reactivity worth of all reactivity control mechanisms and, in some cases, positive moderator temperature coefficient (MTC). The use of isotopically enriched Gd and Er burnable poisons was found to be beneficial with respect to maximizing Pu burnup and reducing power peaking factors. Overall, the annular fertile-free Pu-loaded high-power-density core appears to be feasible, although it still has relatively high power peaking and potential for slightly positive MTC at beginning of cycle. However, we estimate that limiting the power density to 140% of the nominal case would assure acceptable core power peaking and negative MTC at all times during the cycle.

ABSTRACT There is growing interest in the use of 242mAm as a nuclear fuel. Because of its very hi... more ABSTRACT There is growing interest in the use of 242mAm as a nuclear fuel. Because of its very high thermal fission cross section and its large number of neutrons released per fission, it can be used for various unique applications, such as space propulsion, medical applications, and compact energy sources. Since the thermal absorption cross section of 242mAm is very high, the best way to obtain 242mAm is by the capture of fast or epithermal neutrons in 241Am. However, fast spectrum reactors are not readily available. In this paper, we explore the possibility of producing 242mAm in existing pressurized water reactors (PWRs) with minimal interference in reactor performance. As suggested in previous studies on the subject, the 242mAm breeding targets are shielded with strong thermal absorbers in order to suppress the thermal neutron flux that causes 242mAm destruction. Since 242mAm enrichment within the Am target mainly depends on the neutron energy distribution, which in turn depends on the Am target thickness and on the neutron filter cutoff energy (thermal absorber type), this unique Am target design was developed. In our study, Cd, Sm, and Gd were considered as thermal neutron filters, as suggested by Cesana et al. The most favorable results were obtained by irradiating Am targets covered either with Gd or Cd. In these cases, up to 8.65% enrichment of 242mAm is obtained after 4.5 yr (three successive PWR fuel cycles) of irradiation. It was also found that significant quantities [up to 1.3 kg/GW (electric)-yr] of 242mAm can be obtained in PWR reactors without notable interference with reactor performance. However, in order to maintain the original fuel cycle length, the enrichment of the driver (UO 2) fuel must be increased by ∼1%, raised from the conventional 4.5 to 5.5%, depending on the thermal neutron filter used. The most important reactivity feedback coefficients for fuel assemblies containing the 242mAm breeding targets were evaluated and found to be close to those of a standard PWR. Another product of neutron capture in the 241Am reaction is 238Pu. It was found that in a typical 1000 MW (electric) PWR core with one-third of the fuel assemblies containing 241Am targets, up to 15.1 kg of 238Pu enriched to 80% can be produced per year.

ABSTRACT This paper demonstrates that the traditional method of few-group homogenized cross secti... more ABSTRACT This paper demonstrates that the traditional method of few-group homogenized cross section generation for full core analyses may not be sufficient for high conversion LWR designs. In these type of reactors, such as the Hitachi RBWR, separate fissile and blanket zones are required for breeding and for managing void reactivity feedback and therefore leads to a highly axially heterogeneous assembly. Two methods for homogenization were investigated, where a conventional two-dimensional geometry was used and compared to a full three-dimensional assembly calculation. In the two-dimensional calculation, each zone was decoupled from other zones by assuming zero net current boundary conditions, whereas in the three-dimensional calculation the presence of other zones influences the generation of homogenized cross sections. Errors in flux energy spectra were seen, which leads to differences in 2-group homogenized cross sections of up to 50%. The differences in the homogenized parameters were highest in interface zones and near the top of the assembly due to the presence of a reflector and high coolant void fraction. It was determined that these errors may be significant and propagate to the full core analysis of these types of advanced LWRs.

Progress in Nuclear Energy, 2011

This scoping study proposes using mixed nitride fuel in Pu-based high conversion LWR designs in o... more This scoping study proposes using mixed nitride fuel in Pu-based high conversion LWR designs in order to increase the breeding ratio. The higher density fuel reduces the hydrogen-to-heavy metal ratio in the reactor which results in a harder spectrum in which breeding is more effective. A Resource-renewable Boiling Water Reactor (RBWR) assembly was modeled in MCNP to demonstrate this effect in a typical high conversion LWR design. It was determined that changing the fuel from (U,TRU)O 2 to (U,TRU)N in the assembly can increase its fissile inventory ratio (fissile Pu mass divided by initial fissile Pu mass) from 1.04 to up to 1.17.

The objective of this work was to assess the potential of thorium based fuel to minimise Pu and M... more The objective of this work was to assess the potential of thorium based fuel to minimise Pu and MA production in Pressurised Water Reactors (PWRs). The assessment was carried out by examining destruction rates and residual amounts of Pu and MA in the fuel used for transmutation. In particular, sensitivity of these two parameters to the fuel lattice Hydrogen to Heavy Metal (H/HM) ratio and to the fuel composition was systematically investigated. All burn-up calculations were performed using CASMO4 -the fuel assembly burn-up code. The results indicate that up to 1 000 kg of reactor grade Pu can potentially be burned in thorium based fuel assemblies per GW e Year. Up to 75% of initial Pu can be destroyed per path. Addition of MA to the fuel mixture degrades the burning efficiency. The theoretically achievable limit for total TRU destruction per path is 50%. Efficient MA and Pu destruction in thorium based fuel generally requires a higher degree of neutron moderation and, therefore, higher fuel lattice H/HM ratio than typically used in the current generation of PWRs. Reactivity coefficients evaluation demonstrated the feasibility of designing a Th-Pu-MA fuel with negative Doppler and moderator temperature coefficients.

One of the main goals of Generation IV nuclear energy systems is to minimize and manage nuclear w... more One of the main goals of Generation IV nuclear energy systems is to minimize and manage nuclear wastes through partitioning and transmutation of transuranic elements (TRU).

ABSTRACT The once through nuclear fuel cycle adopted by the majority of countries with operating ... more ABSTRACT The once through nuclear fuel cycle adopted by the majority of countries with operating commercial power reactors imposes a number of concerns. The radioactive waste created in the once through nuclear fuel cycle has to be isolated from the environment for thousands of years. In addition, plutonium and other actinides, after the decay of fission products, could become targets for weapon proliferators. Furthermore, only a small fraction of the energy potential in the fuel is being used. All these concerns can be addressed if a closed fuel cycle strategy is considered offering the possibility for partitioning and transmutation of long lived radioactive waste, enhanced proliferation resistance, and improved utilization of natural resources. It is generally believed that dedicated advanced reactor systems have to be designed in order to perform the task of nuclear waste transmutation effectively. The development and deployment of such innovative systems is technically and economically challenging. In this thesis, a possibility of constraining the generation of long lived radioactive waste through multi-recycling of Trans-uranic actinides (TRU) in existing Light Water Reactors (LWR has been studied. Thorium based and fertile free fuels (FFF) were analyzed as the most attractive candidates for TRU burning in LWRs. Although both fuel types can destroy TRU at comparable rates (about 1150 kg/GWe-Year in FFF and up to 900 kg/GWe-Year in Th) and achieve comparable fractional TRU burnup (close to 50a/o), the Th fuel requires significantly higher neutron moderation than practically feasible in a typical LWR lattice to achieve such performance. (Cont.) On the other hand, the FFF exhibits nearly optimal TRU destruction performance in a typical LWR fuel lattice geometry. Increased TRU presence in LWR core leads to neutron spectrum hardening, which results in reduced control materials reactivity worth. The magnitude of this reduction is directly related to the amount of TRU in the core. A potential for positive void reactivity feedback limits the maximum TRU loading. Th and conventional mixed oxide (MOX) fuels require higher than FFF TRU loading to sustain a standard 18 fuel cycle length due to neutron captures in Th232 and U238 respectively. Therefore, TRU containing Th and U cores have lower control materials worth and greater potential for a positive void coefficient than FFF core. However, the significantly reduced fuel Doppler coefficient of the fully FFF loaded core and the lower delayed neutron fraction lead to questions about the FFF performance in reactivity initiated accidents. The Combined Non-Fertile and UO2 (CONFU) assembly concept is proposed for multi- recycling of TRU in existing PWRs. The assembly assumes a heterogeneous structure where about 20% of the UO2 fuel pins on the assembly periphery are replaced with FFF pins hosting TRU generated in the previous cycle. The possibility of achieving zero TRU net is demonstrated. The concept takes advantage of superior TRU destruction performance in FFF allowing minimization of TRU inventory. At the same time, the core physics is still dominated by UO2 fuel allowing maintenance of core safety and control characteristics comparable to all-UO2. Thesis (Ph. D.)--Massachusetts Institute of Technology, Dept. of Nuclear Engineering, February 2004. Includes bibliographical references (p. 249-262).

The COmbined Non-Fertile and UO 2 (CONFU) fuel system for sustainable actinide management has the... more The COmbined Non-Fertile and UO 2 (CONFU) fuel system for sustainable actinide management has the potential to reduce accumulation of actinides in the next 50 years under the assumption of a growing worldwide nuclear energy power demand. The system, which is compatible with currently operating Light Water Reactors (LWRs), employs CONFU fuel assemblies where about 20% of the UO 2 fuel pins are replaced with fertile free fuel (FFF) hosting the TRU.

ABSTRACT Response of a PWR core loaded with Combined Non-Fertile and UO{sub 2} (CONFU) fuel assem... more ABSTRACT Response of a PWR core loaded with Combined Non-Fertile and UO{sub 2} (CONFU) fuel assemblies to control rod ejection accident was evaluated. A number of core arrangements and TRU fuel compositions were considered and the results were compared with the performance of a reference all-UO{sub 2} core. The comparison was based on the results of a simple point kinetics model with thermal reactivity feedbacks and temperature dependant materials properties. The reactivity coefficients and core average kinetics parameters were obtained from the full core 3D neutronic simulations. The results show that application of the CONFU assembly concept causes only minor deviation of fuel performance characteristics in reactivity initiated accidents. This is a consequence of relatively small loadings of TRU in the CONFU assembly and therefore dominating role of conventional UO{sub 2} fuel in the neutronic performance of the core. (authors)

Nuclear Instruments and Methods in Physics Research Section A Accelerators Spectrometers Detectors and Associated Equipment

There is a growing interest in using Am as a nuclear fuel. The advantages of Am as a nuclear fuel... more There is a growing interest in using Am as a nuclear fuel. The advantages of Am as a nuclear fuel derive from the fact that Am has the highest thermal "ssion cross section. The thermal capture cross section is relatively low and the number of neutrons per thermal "ssion is high. These nuclear properties make it possible to obtain nuclear criticality with ultra-thin fuel elements. The possibility of having ultra-thin fuel elements enables the use of these "ssion products directly, without the necessity of converting their energy to heat, as is done in conventional reactors. There are three options of using such highly energetic and highly ionized "ssion products.

ABSTRACT A thorium-based fuel cycle for light water reactors will reduce the plutonium generation... more ABSTRACT A thorium-based fuel cycle for light water reactors will reduce the plutonium generation rate and enhance the proliferation resistance of the spent fuel. However, priming the thorium cycle with 235U is necessary, and the 235U fraction in the uranium must be limited to below 20% to minimize proliferation concerns. Thus, a once-through thorium-uranium dioxide (ThO2-UO2) fuel cycle of no less than 25% uranium becomes necessary for normal pressurized water reactor (PWR) operating cycle lengths. Spatial separation of the uranium and thorium parts of the fuel can improve the achievable burnup of the thorium-uranium fuel designs through more effective breeding of 233U from the 232Th. Focus is on microheterogeneous fuel designs for PWRs, where the spatial separation of the uranium and thorium is on the order of a few millimetres to a few centimetres, including duplex pellet, axially microheterogeneous fuel, and a checkerboard of uranium and thorium pins. A special effort was made to understand the underlying reactor physics mechanisms responsible for enhancing the achievable burnup at spatial separation of the two fuels. The neutron spectral shift was identified as the primary reason for the enhancement of burnup capabilities. Mutual resonance shielding of uranium and thorium is also a factor; however, it is small in magnitude. It is shown that the microheterogeneous fuel can achieve higher burnups, by up to 15%, than the reference all-uranium fuel. However, denaturing of the 233U in the thorium portion of the fuel with small amounts of uranium significantly impairs this enhancement. The denaturing is also necessary to meet conventional PWR thermal limits by improving the power share of the thorium region at the beginning of fuel irradiation. Meeting thermal-hydraulic design requirements by some of the microheterogeneous fuels while still meeting or exceeding the burnup of the all-uranium case is shown to be potentially feasible. However, the large power imbalance between the uranium and thorium regions creates several design challenges, such as higher fission gas release and cladding temperature gradients. A reduction of plutonium generation by a factor of 3 in comparison with all-uranium PWR fuel using the same initial 235U content was estimated. In contrast to homogeneously mixed U-Th fuel, microheterogeneous fuel has a potential for economic performance comparable to the all-UO2 fuel provided that the microheterogeneous fuel incremental manufacturing costs are negligibly small.

This study explores the basic possibility of achieving a self-sustainable Th-233 U fuel cycle tha... more This study explores the basic possibility of achieving a self-sustainable Th-233 U fuel cycle that can be adopted in the current generation of Pressurized Water Reactors. The study outlines possible core design strategies of a typical PWR in order to achieve Breeding Ratio (BR) close to unity. Major design tradeoffs in the core design are discussed. Preliminary neutronic analysis performed on the fuel assembly level with BOXER computer code suggests that net breeding of 233 U is feasible in principle within a typical PWR operating envelope. However, some reduction in the core power density and/or shorter than typical fuel cycle length would most likely be required in order to achieve such performance.

ABSTRACT A new combined Non Fertile and Uranium (CONFU) fuel assembly is proposed to limit the ac... more ABSTRACT A new combined Non Fertile and Uranium (CONFU) fuel assembly is proposed to limit the actinides that need long-term high-level waste storage from the pressurized water reactor (PWR) fuel cycle. In the CONFU assembly concept, ∼20% of the UO2 fuel pins are replaced with fertile free fuel hosting the transuranic elements (TRUs) generated in the previous cycle. This leads to a fuel cycle sustainable with respect to net TRU generation, and the amount and radiotoxicity of the nuclear waste can be significantly reduced in comparison with the conventional once-through UO2 fuel cycle. It is shown that under the constraints of acceptable power peaking limits, the CONFU assembly exhibits negative reactivity feedback coefficients comparable in values to those of the reference UO2 fuel. Feasibility of the PWR core operation and control with complete TRU recycle has been shown based on full-core three-dimensional neutronic simulation. However, gradual buildup of small amounts of Cm and Cf challenges fuel reprocessing and fabrication due to the high spontaneous fission rates of these nuclides and heat generation by some Pu, Am, and Cm isotopes. Feasibility of the processing steps becomes more attainable if the time between discharge and reprocessing is 20 yr or longer.

ABSTRACT The homogeneous ThO2-UO2 fuel cycle option for a pressurized water reactor (PWR) of curr... more ABSTRACT The homogeneous ThO2-UO2 fuel cycle option for a pressurized water reactor (PWR) of current technology is investigated. The fuel cycle assessment was carried out by calculating the main performance parameters: natural uranium and separative work requirements, fuel cycle cost, and proliferation potential of the spent fuel. These performance parameters were compared with a corresponding slightly enriched (all-U) fuel cycle applied to a PWR of current technology. The main conclusion derived from this comparison is that fuel cycle requirements and fuel cycle cost for the mixed Th/U fuel are higher in comparison with those of the all-U fuel. A comparison and analysis of the quantity and isotopic composition of discharged Pu indicate that the Th/U fuel cycle provides only a moderate improvement of the proliferation resistance. Thus, the overall conclusion of the investigation is that there is no economic justification to introduce Th into a light water reactor fuel cycle as a homogeneous ThO2-UO2 mixture.

ABSTRACT Coupled Monte Carlo depletion systems provide a versatile and an accurate tool for analy... more ABSTRACT Coupled Monte Carlo depletion systems provide a versatile and an accurate tool for analyzing advanced thermal and fast reactor designs for a variety of fuel compositions and geometries. The main drawback of Monte Carlo-based systems is a long calculation time imposing significant restrictions on the complexity and amount of design-oriented calculations. This paper presents an alternative approach to interfacing the Monte Carlo and depletion modules aimed at addressing this problem. The main idea is to calculate the one-group cross sections for all relevant isotopes required by the depletion module in a separate module external to Monte Carlo calculations. Thus, the Monte Carlo module will produce the criticality and neutron spectrum only, without tallying of the individual isotope reaction rates. The onegroup cross section for all isotopes will be generated in a separate module by collapsing a universal multigroup (MG) cross-section library using the Monte Carlo calculated flux. Here, the term "universal" means that a single MG cross-section set will be applicable for all reactor systems and is independent of reactor characteristics such as a neutron spectrum; fuel composition; and fuel cell, assembly, and core geometries. This approach was originally proposed by Haeck et al. and implemented in the ALEPH code. Implementation of the proposed approach to Monte Carlo burnup interfacing was carried out through the BGCORE system. One-group cross sections generated by the BGCORE system were compared with those tallied directly by the MCNP code. Analysis of this comparison was carried out and led to the conclusion that in order to achieve the accuracy required for a reliable core and fuel cycle analysis, accounting for the background cross section (σ0) in the unresolved resonance energy region is essential. An extension of the one-group cross-section generation model was implemented and tested by tabulating and interpolating by a simplified σ0 model. A significant improvement of the one-group cross-section accuracy was demonstrated.

ABSTRACT This paper presents results of a feasibility study aimed at developing a zero-transurani... more ABSTRACT This paper presents results of a feasibility study aimed at developing a zero-transuranic-discharge fuel cycle based on the U-Th-TRU ternary cycle. The design objective is to find a fuel composition (mixture of thorium, enriched uranium, and recycled transuranic components) and fuel management strategy resulting in an equilibrium charge-discharge mass flow. In such a fuel cycle scheme, the quantity and isotopic vector of the transuranium (TRU) component is identical at the charge and discharge time points, thus allowing the whole amount of the TRU at the end of the fuel irradiation period to be separated and reloaded into the following cycle. The TRU reprocessing activity losses are the only waste stream that will require permanent geological storage, virtually eliminating the long-term radiological waste of the commercial nuclear fuel cycle. A detailed three-dimensional full pressurized water reactor (PWR) core model was used to analyze the proposed fuel composition and management strategy. The results demonstrate the neutronic feasibility of the fuel cycle with zero-TRU discharge. The amount of TRU and enriched uranium loaded reach equilibrium after about four TRU recycles. The reactivity coefficients were found to be within a range typical for a reference PWR core. The soluble boron worth is reduced by a factor of ∼2 from a typical PWR value. Nevertheless, the results indicate the feasibility of an 18-month fuel cycle design with an acceptable beginning-of-cycle soluble boron concentration even without application of burnable poisons.

ABSTRACT The growing interest in innovative reactors and advanced fuel cycle designs requires mor... more ABSTRACT The growing interest in innovative reactors and advanced fuel cycle designs requires more accurate prediction of various transuranic actinide concentrations during irradiation or following discharge because of their effect on reactivity or spent-fuel emissions, such as gamma and neutron activity and decay heat. In this respect, many of the important actinides originate from the 241Am(n,γ) reaction, which leads to either the ground or the metastable state of 242Am. The branching ratio for this reaction depends on the incident neutron energy and has very large uncertainty in the current evaluated nuclear data files. This study examines the effect of accounting for the energy dependence of the 241Am(n,γ) reaction branching ratio calculated from different evaluated data files for different reactor and fuel types on the reactivity and concentrations of some important actinides. The results of the study confirm that the uncertainty in knowing the 241Am(n,γ) reaction branching ratio has a negligible effect on the characteristics of conventional light water reactor fuel. However, in advanced reactors with large loadings of actinides in general, and 241Am in particular, the branching ratio data calculated from the different data files may lead to significant differences in the prediction of the fuel criticality and isotopic composition. Moreover, it was found that neutron energy spectrum weighting of the branching ratio in each analyzed case is particularly important and may result in up to a factor of 2 difference in the branching ratio value. Currently, most of the neutronic codes have a single branching ratio value in their data libraries, which is sometimes difficult or impossible to update in accordance with the neutron spectrum shape for the analyzed system.

ABSTRACT There is a growing need for very small nuclear reactors for space applications and as po... more ABSTRACT There is a growing need for very small nuclear reactors for space applications and as portable high-intensity neutron sources. This technical note investigates the question of what is the smallest possible thermal reactor. It was found that the smallest reactor is a spherically shaped solution of 242mAm(NO3)3 in water. The weight of such a reactor is 4.95 kg with 0.7 kg of 242mAm nuclear fuel. The radius of the reactor in this case is 9.6 cm.

ABSTRACT BGCore is a software package for comprehensive computer simulation of nuclear reactor sy... more ABSTRACT BGCore is a software package for comprehensive computer simulation of nuclear reactor systems and their fuel cycles. The BGCore interfaces Monte Carlo particles transport code MCNP4C with a SARAF module - an independently developed code for calculating in-core fuel composition and spent fuel emissions following discharge. In BGCore system, depletion coupling methodology is based on the multi-group approach that significantly reduces computation time and allows tracking of large number of nuclides during calculations. In this study, burnup calculation capabilities of BGCore system were validated against well established and verified, computer codes for thermal and fast spectrum lattices. Very good agreement in k eigenvalue and nuclide densities prediction was observed for all cases under consideration. In addition, decay heat prediction capabilities of the BGCore system were benchmarked against the most recent edition of ANS Standard methodology for UO2 fuel decay power prediction in LWRs. It was found that the difference between ANS standard data and that predicted by the BGCore does not exceed 5%.

ABSTRACT This paper investigates the basic feasibility of using reactor-grade Pu in fertile-free ... more ABSTRACT This paper investigates the basic feasibility of using reactor-grade Pu in fertile-free fuel (FFF) matrix in pressurized water reactors (PWRs). Several important issues were investigated in this work: the Pu loading required to achieve a specific interrefueling interval, the impact of inert matrix composition on reactivity constrained length of cycle, and the potential of utilizing burnable poisons (BPs) to alleviate degradation of the reactivity control mechanism and temperature coefficients. Although the subject was addressed in the past, no systematic approach for assessment of BP utilization in FFF cores was published. In this work, we examine all commercially available BP materials in all geometrical arrangements currently used by the nuclear industry with regards to their potential to alleviate the problems associated with the use of FFF in PWRs. The recently proposed MgO-ZrO2 solid-state solution fuel matrix, which appears to be very promising in terms of thermal properties and radiation damage resistance, was used as a reference matrix material in this work. The neutronic impact of the relative amounts of MgO and ZrO2 in the matrix were also studied. The analysis was performed with a neutron transport and fuel assembly burnup code BOXER. A modified linear reactivity model was applied to the two-dimensional single fuel assembly results to approximate the full core characteristics. Based on the results of the performed analyses, the Pu-loaded FFF core demonstrated potential feasibility to be used in existing PWRs. Major FFF core design problems may be significantly mitigated through the correct choice of BP design. It was found that a combination of BP materials and geometries may be required to meet all FFF design goals. The use of enriched (in most effective isotope) BPs, such as 167Er and 157Gd, may further improve the BP effectiveness and reduce the fuel cycle length penalty associated with their use.

ABSTRACT Up to 50% increase in the power density of the existing pressurized water reactor (PWR)-... more ABSTRACT Up to 50% increase in the power density of the existing pressurized water reactor (PWR)-type reactors can be achieved by the use of internally and externally cooled annular fuel geometry. As a result, the accumulated stock-piles of Pu, especially if incorporated infertile-free inert matrix, can be burnt at a substantially higher rate as compared with the conventional mixed oxide-fueled reactors operating at standard power density. In this work, we explore the basic feasibility of a PWR core fully loaded with Pu incorporated infertile-free fuel of annular internally and externally cooled geometry and operating at 150% of nominal power density. We evaluate basic burnable poison designs, fuel management strategies, and reactivity feedback coefficients. The three-dimensional full core neutronic analysis performed with Studsvik Core Management System showed that the design of such a Pu-loaded annular fuel core is feasible but significantly more challenging than the Pu fertile-free core with solid fuel pins operating at nominal power density. The main difficulty arises from the fact that the annular fuel core requires at least 50% higher initial Pu loading in order to maintain the standard fuel cycle length of 18 months. Such a high Pu loading results in hardening of the neutron spectrum and consequent reduction in reactivity worth of all reactivity control mechanisms and, in some cases, positive moderator temperature coefficient (MTC). The use of isotopically enriched Gd and Er burnable poisons was found to be beneficial with respect to maximizing Pu burnup and reducing power peaking factors. Overall, the annular fertile-free Pu-loaded high-power-density core appears to be feasible, although it still has relatively high power peaking and potential for slightly positive MTC at beginning of cycle. However, we estimate that limiting the power density to 140% of the nominal case would assure acceptable core power peaking and negative MTC at all times during the cycle.

ABSTRACT There is growing interest in the use of 242mAm as a nuclear fuel. Because of its very hi... more ABSTRACT There is growing interest in the use of 242mAm as a nuclear fuel. Because of its very high thermal fission cross section and its large number of neutrons released per fission, it can be used for various unique applications, such as space propulsion, medical applications, and compact energy sources. Since the thermal absorption cross section of 242mAm is very high, the best way to obtain 242mAm is by the capture of fast or epithermal neutrons in 241Am. However, fast spectrum reactors are not readily available. In this paper, we explore the possibility of producing 242mAm in existing pressurized water reactors (PWRs) with minimal interference in reactor performance. As suggested in previous studies on the subject, the 242mAm breeding targets are shielded with strong thermal absorbers in order to suppress the thermal neutron flux that causes 242mAm destruction. Since 242mAm enrichment within the Am target mainly depends on the neutron energy distribution, which in turn depends on the Am target thickness and on the neutron filter cutoff energy (thermal absorber type), this unique Am target design was developed. In our study, Cd, Sm, and Gd were considered as thermal neutron filters, as suggested by Cesana et al. The most favorable results were obtained by irradiating Am targets covered either with Gd or Cd. In these cases, up to 8.65% enrichment of 242mAm is obtained after 4.5 yr (three successive PWR fuel cycles) of irradiation. It was also found that significant quantities [up to 1.3 kg/GW (electric)-yr] of 242mAm can be obtained in PWR reactors without notable interference with reactor performance. However, in order to maintain the original fuel cycle length, the enrichment of the driver (UO 2) fuel must be increased by ∼1%, raised from the conventional 4.5 to 5.5%, depending on the thermal neutron filter used. The most important reactivity feedback coefficients for fuel assemblies containing the 242mAm breeding targets were evaluated and found to be close to those of a standard PWR. Another product of neutron capture in the 241Am reaction is 238Pu. It was found that in a typical 1000 MW (electric) PWR core with one-third of the fuel assemblies containing 241Am targets, up to 15.1 kg of 238Pu enriched to 80% can be produced per year.

ABSTRACT This paper demonstrates that the traditional method of few-group homogenized cross secti... more ABSTRACT This paper demonstrates that the traditional method of few-group homogenized cross section generation for full core analyses may not be sufficient for high conversion LWR designs. In these type of reactors, such as the Hitachi RBWR, separate fissile and blanket zones are required for breeding and for managing void reactivity feedback and therefore leads to a highly axially heterogeneous assembly. Two methods for homogenization were investigated, where a conventional two-dimensional geometry was used and compared to a full three-dimensional assembly calculation. In the two-dimensional calculation, each zone was decoupled from other zones by assuming zero net current boundary conditions, whereas in the three-dimensional calculation the presence of other zones influences the generation of homogenized cross sections. Errors in flux energy spectra were seen, which leads to differences in 2-group homogenized cross sections of up to 50%. The differences in the homogenized parameters were highest in interface zones and near the top of the assembly due to the presence of a reflector and high coolant void fraction. It was determined that these errors may be significant and propagate to the full core analysis of these types of advanced LWRs.