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CDMA/OFDM signals far various values of M with QPSK transmission and N = 128. In this Figure, the analytic results of the onginal OFDM signals with the SLM approach arc also given for using one, two and four statistically independent sequences corresponding to S= I, 2 and 4, respectively. The 0.1% PAR of the transformed CDMAIOFDM signals is reduced by 2-3 dB compared to that o f the onginal CDMA/OFDM signals for the multiuser case with half user capacity. As descnbed in the previous Section, further PAR reduction can be obtained by modifying the SLM approach. The 0.1% PAR ofthe OFDM-CDMA signals camhincd with the modified SLM approach using four statistically independent sequences o f S = 4 is reduced by up to 1 dB compared to that of the OFDM-CDMA signals with the anginal SLM approach employing two sequences of S=2. maintaining approximately same transmitter complexity. Unfmhmately, one additional side information hit is required. However, under the condition of the same hits of the side information of log,S= 1, the complexity of the modified SLM approach with S= 2 is reduced by 25% compared with that of the original SLM approach with S=2, giving the nearly sanie PAR performance. The transmitter complexity is funher reduced for larger value of S and in the case of S = 4 the complexity reduction o f 37.5% is obtained.

Idaho National Laboratory (INL) is tasked with the development of reactor physics analysis capability for the Next Generation Nuclear Power (NGNP) project. In order to examine INL's current prismatic reactor deterministic analysis tools, the project is conducting a benchmark exercise based on modeling the High Temperature Test Reactor (HTTR). This exercise entails the development of a model for the initial criticality, a 19-fuel column thin annular core, and the fully loaded core critical condition with 30 fuel columns. Special emphasis is devoted to the annular core modeling, which shares more characteristics with the NGNP base design. The DRAGON code is used in this study because it offers significant ease and versatility in modeling prismatic designs. Despite some geometric limitations, the code performs quite well compared to other lattice physics codes. DRAGON can generate transport solutions via collision probability (CP), method of characteristics (MOC), and discrete ordinates (Sn). A fine group cross-section library based on the SHEM 281 energy structure is used in the DRAGON calculations. HEXPEDITE is the hexagonal-z full-core solver used in this study and is based on the Green's Function solution of the transverse-integrated equations. In addition, two Monte Carlo (MC) based codes, MCNP5 and PSG2/SERPENT, as well as the deterministic transport code INSTANT, provide benchmarking capability for the DRAGON and HEXPEDITE. The results from this study show reasonable agreement in the calculation of the core multiplication factor with the MC methods, but a consistent bias of 2-3% with the experimental values is obtained. This systematic error has also been observed in other HTTR benchmark efforts and is well documented in the literature. The uncertainty in the graphite impurity appears to be the main source of the error, whereas inaccuracies in the ENDF/B-VII graphite and U 235 cross-sections have a secondary effect. The isothermal temperature coefficients calculated with the fully loaded core configuration agree well with other benchmark participants but are 40% higher than the experimental values. This discrepancy with the measurement partially stems from the fact that during the experiments the control rods were adjusted to maintain criticality, whereas in the model, the rod positions were fixed. In addition, this work includes a brief study of a cross-section generation approach that seeks to decouple the domain in order to account for neighbor effects. This spectral interpenetration is a dominant effect in annular HTR physics. This analysis methodology should be further explored in order to reduce the error that is systematically propagated in the traditional generation of cross-sections. vii viii CONTENTS Abstract

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A 1000 MWth commercial-scale Sodium Fast Reactor (SFR) design with a conversion ratio (CR) of 0.50 was selected in this study to perform perturbations on the external feed coming from Light Water Reactor Spent Nuclear Fuel (LWR SNF) and separation groupings in the reprocessing scheme. A secondary SFR design with a higher conversion ratio (CR=0.75) was also analyzed as a possible alternative, although no perturbations were applied to this model.Metal and oxide fuel SFR designs were both included in the analysis. Initial results showed good agreement between the UREX+1a base cases and data previously published in the literature for the SFR conceptual design. The initial set of perturbations involved varying the external feed to study the so-called 'vintage problem', which addresses the large variation in burnup and cooling needed to be accommodated in the SFR. Three sets of external feed isotopic vectors were generated for the cases of a low burnup and long cooling time (33 MWd/kg, 30 year cooled LWR SNF), high burnup and medium cooling time (51 MWd/kg, 10 year cooled LWR SNF), and the reference high burnup and short cooling time (51 MWd/kg, 5 year cooled LWR SNF.) Results show that the choice of external feed has little impact on the TRU enrichment, burnup, or cycle length of either the metal or oxide fuel SFR because all the TRU vectors having similar fissile plutonium content. Also, the slightly larger presence of americium in the low burnup, long cooling time vector increases its consumption rate in the SFR, and thus increases the production of curium 242 and 244 compared to a high burnup, short cooling time TRU vector.The second set of perturbations involved varying the external feed and reprocessing of the TRU groupings for the metal and oxide SFR designs. Four separation technologies were applied to the LWR SNF; PUREX, UREX+2/+3, UREX+4, and UREX+1a. In the case of metal fuel, this perturbation only affects the feed of isotopes coming from the separation facility, while the electrochemical reprocessing recycles all TRU isotopes from the SFR back into the reactor core as fuel. This is different from the oxide case, in which the four separation technologies (PUREX and UREX+) may also be applied to the reprocessing of the SFR fuel. In the case of PUREX, for example, the neptunium, americium, curium, berkelium, and californium are separated from the discharged fuel reprocessing and assumed to be disposed of, thus creating fresh plutonium-only oxide fuel.The effects of the choice of separation and reprocessing strategy on the neutron emission, gamma energy, and decay heat at beginning-of-equilibrium cycle (BOEC) and the decay heat at end-ofequilibrium cycle (EOEC) were also incorporated into this study. The effects of different 'groupings' were found to have a minimal effect on the parameters mentioned above for the metal fuel SFR, since all TRU isotopes are homogeneously recycled back into the core. In the case of an oxide fuel SFR, the BOEC charge neutron emission, gamma energy, and decay heat all decr...