Composition Optimization of Thorium-uranium Pressurized Water Reactor Cores PDF Download

Author: Kevin Taylor Clarno
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Category :
Languages : en
Pages : 125

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Author: Dean Wang
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Category :
Languages : en
Pages : 252

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(Cont.) Fuel performance was analyzed using FRAPCON. The radioactivity and decay heat from the spent seed and blanket fuel were studied using MIT's MCODE (which couples MCNP and ORIGEN) to do depletion calculations, and ORIGEN to analyze the spent fuel characteristics after discharge. The analyses show that the WASB core can satisfy the requirements of fuel cycle length and safety margins of conventional PWRs. The coefficients of reactivity are comparable to currently operating PWRs. However, the reduction in effective delayed neutron fraction (eff) requires careful review of the control systems because of its importance to short term power transients. Whole core analyses show that the total control rod worth of the WASB core is about 1/3 less than those of a typical PWR for a standard arrangement of Ag-In-Cd control rods in the core. The use of enriched boron in the control rods can effectively improve the control rod worth. The control rods have higher worth in the seed than in the blanket. Therefore, a new loading pattern has been designed so that almost all the control rods will be located in seed assemblies. However, the new pattern requires a redesign of the vessel head of the reactor, which is an added cost in case of retrofitting in existing PWRs. Though the WASB core has high power peaking factors, acceptable MDNBR in the core can be achieved under conservative assumptions by using grids with large local pressure loss coefficient in the blanket. However, the core pressure drop will increase by 70% ...

Author: L. E. Strawbridge
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Category : Nuclear reactors
Languages : en
Pages : 246

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Languages : en
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Two of the major barriers to the expansion of worldwide adoption of nuclear power are related to proliferation potential of the nuclear fuel cycle and issues associated with the final disposal of spent fuel. The Radkowsky Thorium Fuel (RTF) concept proposed by Professor A. Radkowsky offers a partial solution to these problems. The main idea of the concept is the utilization of the seed-blanket unit (SBU) fuel assembly geometry which is a direct replacement for a 'conventional' assembly in either a Russian pressurized water reactor (VVER-1000) or a Western pressurized water reactor (PWR). The seed-blanket fuel assembly consists of a fissile (U) zone, known as seed, and a fertile (Th) zone known as blanket. The separation of fissile and fertile allows separate fuel management schemes for the thorium part of the fuel (a subcritical 'blanket') and the 'driving' part of the core (a supercritical 'seed'). The design objective for the blanket is an efficient generation and in-situ fissioning of the U233 isotope, while the design objective for the seed is to supply neutrons to the blanket in a most economic way, i.e. with minimal investment of natural uranium. The introduction of thorium as a fertile component in the nuclear fuel cycle significantly reduces the quantity of plutonium production and modifies its isotopic composition, reducing the overall proliferation potential of the fuel cycle. Thorium based spent fuel also contains fewer higher actinides, hence reducing the long-term radioactivity of the spent fuel. The analyses show that the RTF core can satisfy the requirements of fuel cycle length, and the safety margins of conventional pressurized water reactors. The coefficients of reactivity are comparable to currently operating VVER's/PWR's. The major feature of the RTF cycle is related to the total amount of spent fuel discharged for each cycle from the reactor core. The fuel management scheme adopted for RTF core designs allows a significant decrease in the amount of discharged spent fuel, for a given energy production, compared with standard VVER/PWR. The total Pu production rate of RTF cycles is only 30 % of standard reactor. In addition, the isotopic compositions of the RTF's and standard reactor grade Pu are markedly different due to the very high burnup accumulated by the RTF spent fuel.

Author: Ildo Luis Sauer
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Category : Pressurized water reactors
Languages : en
Pages : 283

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The primary objective of this research was the development of a comprehensive, rapid and conceptually simple methodology for PWR core reload pattern and fuel composition optimization, capable of systematic incorporation of constraints, in which cycle burnup is defined as the optimality criterion. A coarse mesh nodal method for PWR core analysis was formulated by coupling the one-and-one-half-group diffusion theory model for spatial power calculations with the linear reactivity versus burnup model (LRM) for depletion calculations. The accuracy and suitability of this model was determined through comparisons of its results with those of state-of-the-art core analysis methods. The simplicity of the LRM-based core model allowed the direct analytical computation of the derivatives necessary in the steepest gradient type optimization methods applied in the present work, and its versatility permitted use of the analytical and computational methods for a variety of applications, ranging from core reload pattern searches to burnable poison (BP) and composition optimization. Algorithms for identification of unconstrained maximum-burnup core reload patterns and for optimal BP allocation were successfully implemented and tested, and the basis for systematic incorporation of constraints on power peaking was developed. The potential application of the methodology to fuel composition optimization was also examined. Most of the methodological developments have been embodied in the LRM-NODAL code which was programmed in the course of this research. From the numerical and analytical results it was found that the optimal core configurations are arranged such as to produce power histories and profiles in which the most reactive assemblies are at their highest allowable power at EOC (thus maximizing their importance) and where the converse applies to the least reactive; these preferred profiles also produce relatively higher leakage at EOC, evolving to the lowest possible leakage at EOC, but always consistent with the maximization of the core reactivity importance.

Author: Phillip Michael Gorman
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Category :
Languages : en
Pages : 177

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The Resource-renewable Boiling Water Reactors (RBWRs) are a set of light water reactors (LWRs) proposed by Hitachi which use a triangular lattice and high void fraction to incinerate fuel with an epithermal spectrum, which is highly atypical of LWRs. The RBWRs operate on a closed fuel cycle, which is impossible with a typical thermal spectrum reactor, in order to accomplish missions normally reserved for sodium fast reactors (SFRs) - either fuel self-sufficiency or waste incineration. The RBWRs also axially segregate the fuel into alternating fissile "seed" regions and fertile "blanket" regions in order to enhance breeding and leakage probability upon coolant voiding. This dissertation focuses on thorium design variants of the RBWR: the self-sufficient RBWR-SS and the RBWR-TR, which consumes reprocessed transuranic (TRU) waste from PWR used nuclear fuel. These designs were based off of the Hitachi-designed RBWR-AC and the RBWR-TB2, respectively, which use depleted uranium (DU) as the primary fertile fuel. The DU-fueled RBWRs use a pair of axially segregated seed sections in order to achieve a negative void coefficient; however, several concerns were raised with this multi-seed approach, including difficulty with controlling the reactor and unacceptably high axial power peaking. Since thorium-uranium fuel tends to have much more negative void feedback than uranium-plutonium fuels, the thorium RBWRs were designed to use a single elongated seed to avoid these issues. A series of parametric studies were performed in order to find the design space for the thorium RBWRs, and optimize the designs while meeting the required safety constraints. The RBWR-SS was optimized to maximize the discharge burnup, while the RBWR-TR was optimized to maximize the TRU transmutation rate. These parametric studies were performed on an assembly level model using the MocDown simulator, which calculates an equilibrium fuel composition with a specified reprocessing scheme. A full core model was then created for each design, using the Serpent/PARCS 3-D core simulator, and the full core performance was assessed. The RBWR-SS benefited from a harder spectrum than the RBWR-TR; a hard spectrum promotes breeding and increases the discharge burnup, but reduces the TRU transmutation rate. This led the RBWR-SS to have a very tight lattice, which has a lot of experimental uncertainty in the thermal hydraulic correlations. Two different RBWR-SS designs were created assuming different thermal hydraulic assumptions: the RBWR-SSH used the same assumptions as Hitachi used for the RBWR-AC, while the RBWR-SSM used more conservative correlations recommended by collaborators at MIT. However, the void feedback of the pure Th-fed system was too strongly negative, even with a single elongated seed. Therefore, instead of using just thorium, the self-sustaining designs were fed with a mix of between 30% and 50% DU and the rest thorium in order to keep the void feedback as close to zero as possible. This was not necessary for the RBWR-TR, as the external TRU feed fulfilled a similar role. Unfortunately, it was found that the RBWR-SSM could not sustain a critical cycle without either significantly downgrading the power or supplying an external feed of fissile material. While the RBWR-SSH and the RBWR-TR could reach similar burnups and transmutation rates to their DU-fueled counterparts as designed by Hitachi, the thorium designs were unable to simultaneously have negative void feedback and sufficient shutdown margin to shut down the core. The multi-seed approach of the Hitachi designs allowed their reactors to have much lower magnitudes of Doppler feedback than the single-seed designs, which helps them to have sufficient shutdown margin. It is expected that thorium-fueled RBWRs designed to have multiple seeds would permit adequate shutdown margin, although care would need to be taken in order to avoid running into the same issues as the DU fueled RBWRs. Alternatively, it may be possible to increase the amount of boron in the control blades by changing the assembly and core design. Nonetheless, the uncertainties in the multiplication factor due to nuclear data and void fraction uncertainty were assessed for the RBWR-SSH and the RBWR-TR, as well as for the RBWR-TB2. In addition, the uncertainty associated with the change in reactor states (such as the reactivity insertion in flooding the core) due to nuclear data uncertainties was quantified. The thorium RBWRs have much larger uncertainty of their DU-fueled counterparts as designed by Hitachi, as the fission cross section of 233U has very large uncertainty in the epithermal energy range. The uncertainty in the multiplication factor at reference conditions was about 1350 pcm for the RBWR-SSH, while it was about 900 pcm for the RBWR-TR. The uncertainty in the void coefficient of reactivity for both reactors is between 8 and 10 pcm/% void, which is on the same order of magnitude as the full core value. Finally, since sharp linear heat rate spikes were observed in the RBWR-TB2 simulation, the RBWR-TB2 unit cell was simulated using a much finer mesh than is possible using deterministic codes. It was found that the thermal neutrons reflecting back from the reflectors and the blankets were causing extreme spikes in the power density near the axial boundaries of the seeds, which were artificially smoothed out when using coarser meshes. It is anticipated that these spikes will cause melting in both seeds in the RBWR-TB2, unless design changes - such as reducing the enrichment level near the axial boundaries of the seeds - are made.

Author:
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Category : Electronic dissertations
Languages : en
Pages : 270

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Reactivity coefficients -- Doppler coefficient (DC) -- Moderator temperature coefficient (MTC) -- Critical Boron Concentration (CBC) -- UO2 (Uranium dioxide) -- MOX (Mixed Oxide fuel) -- Thorium -- Fuel pin -- PWR (Pressurized Water Reactor) -- MCNP (Monte Carlo N Particle Code)

Author: Altamash Kamal
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Category : Light water reactors
Languages : en
Pages : 318

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Systematic procedures have been developed and applied to assess the uranium utilization potential of a broad range of options involving the selective use of thorium in Pressurized Water Reactors (PWRs) operating on the once-through cycle. The methods used rely on state-of-the-art physics methods coupled with batch-wise core depletion models based on the "group-and-one-half" theory. The possible roles for thorium that were investigated are: as internal and radial blanket material, as thorium pins dispersed within uranium fuel assemblies, its use in PWRs operating on spectral shift control, and its reconstitution and reinsertion as radial blanket assemblies. The use of smaller assemblies in PWRs (for cores with and without thorium) was also investigated, as well as options which can be regarded as reasonable substitutes for employing thorium. The analyses were performed for both current (3-batch, discharge burnup n 30 GWD/MT) and high-burnup (5-batch, discharge burnup% 50 GWD/MT) PWR cores in their steady-state. It was found that except for special circumstances (dry lattices and/or high burnup), the use of thorium does not save uranium compared to the conventional all-uranium PWRs. When savings are achieved (typically 1-3%, but as high as 9% in special circumstances), they can be, for the most part, equaled or exceeded by easier means: in particular, by the re-use of spent fuel. On the other hand, up to 15 or 20% thorium could be added into PWRs without significant losses in uranium utilization, if policies called for the build up of a U-233 inventory for later use in the recycle mode. It was also found that, regardless of the deployment of thorium, the use of smaller fuel assemblies with the concurrent deployment of radial blankets is an effective uranium conservation strategy, with accompanying power-shaping advantages.

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Languages : en
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Typical pressurized water reactors, although loaded with uranium fuel, produce 225 to 275 kg of plutonium per gigawatt year of operation. Although the spent fuel is highly radioactive, it nevertheless offers a potential proliferation pathway because the plutonium is relatively easy to separate, amounts to many critical masses, and aside from the alpha (n reaction on the 240 Pu isotope) does not present any significant intrinsic barrier to weapon assembly.

Author:
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Category :
Languages : en
Pages : 146

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Issues affecting the implementation, public perception and acceptance of nuclear power include: proliferation, radioactive waste, safety, and economics. The thorium cycle directly addresses the proliferation and waste issues, but optimization studies of core design and fuel management are needed to ensure that it fits within acceptable safety and economic margins. Typical pressurized water reactors, although loaded with uranium fuel, produce 225 to 275 kg of plutonium per gigawatt-year of operation. Although the spent fuel is highly radioactive, it nevertheless offers a potential proliferation pathway because the plutonium is relatively easy to separate, amounts to many critical masses, and does not present any significant intrinsic barrier to weapon assembly. Uranium 233, on the other hand, produced by the irradiation of thorium, although it too can be used in weapons, may be ''denatured'' by the addition of natural, depleted or low enriched uranium. Furthermore, it appears that the chemical behavior of thoria or thoria-urania fuel makes it a more stable medium for the geological disposal of the spent fuel. It is therefore particularly well suited for a once-through fuel cycle. The use of thorium as a fertile material in nuclear fuel has been of interest since the dawn of nuclear power technology due to its abundance and to potential neutronic advantages. Early projects include homogeneous mixtures of thorium and uranium oxides in the BORAX-IV, Indian Point I, and Elk River reactors, as well as heterogeneous mixtures in the Shippingport seed-blanket reactor. However these projects were developed under considerably different circumstances than those which prevail at present. The earlier applications preceded the current proscription, for non-proliferation purposes, of the use of uranium enriched to more than 20 w/o in 235U, and has in practice generally prohibited the use of uranium highly enriched in 235U. They were designed when the expected burnup of light water fuel was on the order of 25 MWD/kgU--about half the present day value--and when it was expected that the spent fuel would be recycled to recover its fissile content.