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Approximately one year before a single UO2 fuel pellet is pressed and fresh fuel assemblies manufactured, design calculations are performed to effectively guarantee the performance of a light water reactor (LWR) core for an upcoming fuel cycle. Energy requirements must be fulfilled in conjunction with all other reactivity, thermal, and operational limits. Furthermore, several months prior to a reactor startup, the proposed core must be licensed by the regulating authority. Therefore, all built-in design conservatisms (i.e., target margins) established a priori must satisfy all safety, operational, and regulatory constraints, a posteriori. The magnitude of target margins directly impacts cycle energy efficiency, which is why this design cushion is often referred to as the "cost of margin" because it ultimately affects the cost per generated kilowatt-hour by a LWR. This article illustrates the modern role of nuclear fuel management optimization in the LWR core reload design process, highlighting some of the history and recent advancements in the field, particularly, in the area of pin-by-pin optimization. Also, important limitations are highlighted to help define the new level of sophistication which the field must conquer for designers to ultimately be able to optimize LWRs "one fuel pin at a time." [ABSTRACT FROM AUTHOR]

Abstract: Various arrangements of beryllium loadings to create an internal neutron reflector in the flux trap region of the Oak Ridge National Laboratory’s High Flux Isotope Reactor (HFIR) have been investigated. In particular, the impact upon fuel cycle length has been studied by performing calculations using the HFIR MCNP-based model HFV4.0. This study included examining perturbations in reactivity, flux, and power distribution caused by the various beryllium loadings. The HFIR Cycle 400 core configuration was used as a reference to calculate the impact of beryllium loadings upon cycle length. Three different configurations of beryllium loadings were investigated and compared against the Cycle 400 benchmark calculations; Cases 1 through 3 modeled combinations of 12 and 18 beryllium rods loaded into unused experimental sites. Calculated eigenvalues have shown that potential increases in reactivity between 0.56 and 0.79 dollars are attainable, depending on the various beryllium configurations. These results correspond to possible increases in fuel cycle length ranging between 2.3% and 3.3%. On the basis of their practicality, cost versus benefit, and greater potential for implementation, Cases 2 and 3 (both with 18 beryllium rods) were studied further and are herein reported in greater detail. Neutron flux distributions for Cases 2 and 3 were calculated at the horizontal mid-plane of the flux trap region, which showed no significant changes in the thermal flux magnitude and radial profile in comparison to Cycle 400. Likewise, safety analysis related parameters were contrasted, revealing power increments of up to 2% near the inside edge of the inner fuel element, well below the maximum acceptable value of 9%, a standing guideline employed for experiments at the HFIR. Additionally, the average neutron heat generation rate in beryllium rods and the maximum heat generation rate were evaluated to confirm that the design provides adequate coolant flow inside the rod and around the beryllium targets to carry away any excess heat. Case 2 with 18 beryllium rods was recommended for implementation at the HFIR. In fact, funding has been allocated for this project and beryllium rod fabrication is scheduled. The proposed beryllium loading does not involve permanent design changes and is predicted to yield a meaningful increase in fuel cycle length that could translate into potential annual savings in direct fuel costs of approximately $235,000, or of more than five million dollars over the projected life of this reactor. [Copyright &y& Elsevier]

• FeCrAl, cladding/channel box halved, required ∼0.6% more 235U average enrichment. • Geometric/enrichment-altered full core models reached 527 effective full power days. • Hybrid design—FeCrAl cladding and SiC/Ni/Cr channel box—matched cycle length. • Thermomechanical performance of a Zircaloy-cladded fuel rod was compared to FeCrAl. The impact of replacing Zircaloy with FeCrAl, a candidate enhanced accident-tolerant fuel cladding material, was evaluated for 10 × 10 boiling water reactor fuel bundles. Results from a series of full-core parametric studies estimated that replacing UO 2 /Zircaloy with UO 2 /FeCrAl would require an average enrichment increase of 0.6% 235U throughout the fuel lattice with the cladding and channel box thicknesses halved and fuel pellet diameter increased. Full-core results indicated that UO 2 /FeCrAl models with these geometric/enrichment specifications matched the base UO 2 /Zircaloy cycle length of 527 effective full power days. Optimization studies of the full-core design established loading and control blade patterns for both Zircaloy and FeCrAl models. A side study was conducted modeling a hybrid fuel bundle consisting of FeCrAl cladding and a SiC/Ni/Cr channel box. By halving the cladding thickness, the enrichment level required was less than that of the Zircaloy base case design after performing loading pattern optimization of the hybrid bundle core. Lastly, the thermomechanical performance of a Zircaloy-cladded fuel rod was compared to that of a FeCrAl system. Results from this analysis show that, if starting from the same fuel-cladding gap thickness, a FeCrAl-clad fuel rod operates with a greater average fuel centerline temperature, comparable axial elongation and radial displacement, and longer time to gap closure compared to a Zircaloy-clad fuel rod. This fuel performance analysis was primarily based on the commercial Kanthal APMT FeCrAl alloy but also used available data for the C35M FeCrAl alloy developed at Oak Ridge National Laboratory. [ABSTRACT FROM AUTHOR]