Detailed design of242mAm breeding in pressurized water reactors

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 (UO2) 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.

Fertile-free fuels in pressurized water reactors: Design challenges and solutions

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.

Assessment of homogeneous thorium/uranium fuel for pressurized water reactors

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.

Microheterogeneous Thoria-Urania Fuels for Pressurized Water Reactors

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.

FUELGEN: effective evolutionary design of refuellings for pressurized water reactors

The paper describes the design of an efficient and robust genetic algorithm for the nuclear fuel loading problem (i.e., refuellings: the in-core fuel management problem) - a complex combinatorial, multimodal optimisation., Evolutionary computation as performed by FUELGEN replaces heuristic search of the kind performed by the FUELCON expert system (CAI 12/4), to solve the same problem. In contrast to the traditional genetic algorithm which makes strong requirements on the representation used and its parameter setting in order to be efficient, the results of recent research results on new, robust genetic algorithms show that representations unsuitable for the traditional genetic algorithm can still be used to good effect with little parameter adjustment. The representation presented here is a simple symbolic one with no linkage attributes, making the genetic algorithm particularly easy to apply to fuel loading problems with differing core structures and assembly inventories. A nonlinear fitness function has been constructed to direct the search efficiently in the presence of the many local optima that result from the constraint on solutions.

Pressurized water reactors : a literature search /

"June 1959."

Small size pressurized water reactor.

"Including errata of March 7, 1960 and Supplement No. 1, Revised March 11, 1960."

Analysis of a power failure incident for the preliminary conceptual design of the small pressurized water reactor /

"Contract No. AT(40-1)-2589."

The Shippingport pressurized water reactor : a bibliography of report literature /

A total of 193 annotated references to unclassified reports on the design, development and construction of the Shippingport Pressurized Water Reactor is presented. Author, subject, and report number indexes are included.

Recycle of transuranium elements in light water reactors for reduction of geological storage requirements

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.

Muitiobjective pressurized water reactor reload core design by nondominated genetic algorithm search

The design of pressurized water reactor reload cores is not only a formidable optimization problem but also, in many instances, a multiobjective problem. A genetic algorithm (GA) designed to perform true multiobjective optimization on such problems is described. Genetic algorithms simulate natural evolution. They differ from most optimization techniques by searching from one group of solutions to another, rather than from one solution to another. New solutions are generated by breeding from existing solutions. By selecting better (in a multiobjective sense) solutions as parents more often, the population can be evolved to reveal the trade-off surface between the competing objectives. An example illustrating the effectiveness of this novel method is presented and analyzed. It is found that in solving a reload design problem the algorithm evaluates a similar number of loading patterns to other state-of-the-art methods, but in the process reveals much more information about the nature of the problem being solved. The actual computational cost incurred depends: on the core simulator used; the GA itself is code independent.

The performance of closed reactor grade plutonium-thorium fuel cycles in reduced-moderation pressurised water reactors

The production of long-lived transuranic (TRU) waste is a major disadvantage of fission-based nuclear power. Previous work has indicated that TRU waste can be virtually eliminated in a pressurised water reactor (PWR) fuelled with a mixture of thorium and TRU waste, when all actinides are returned to the reactor after reprocessing. However, the optimal configuration for a fuel assembly operating this fuel cycle is likely to differ from the current configuration. In this paper, the differences in performance obtained in a reduced-moderation PWR operating this fuel cycle were investigated using WIMS. The chosen configuration allowed an increase of at least 20% in attainable burn-up for a given TRU enrichment. This will be especially important if the practical limit on TRU enrichment is low. The moderator reactivity coefficients limit the enrichment possible in the reactor, and this limit is particularly severe if a negative void coefficient is required for a fully voided core. Several strategies have been identified to mitigate this. Specifically, the control system should be designed to avoid a detrimental effect on moderator reactivity coefficients. The economic viability of this concept is likely to be dependent on the achievable thermal-hydraulic operating conditions. © 2012 Elsevier Ltd. All rights reserved.