The subject of this thesis is the development of a nodal discretization of the low-order quasi-diffusion (QDLO) equations for global reactor core calculations. The advantage of quasi-diffusion (QD) is that it is able to capture transport effects at the surface between unlike fuel assemblies better than the diffusion approximation. We discretize QDLO equations with the advanced nodal methodology described by Palmtag (Pal 1997) for diffusion. The fast and thermal neutron fluxes are presented as 2-D, non-separable expansions of polynomial and hyperbolic functions. The fast flux expansion consists of polynomial functions, while the thermal flux is expanded in a combination of polynomial and hyperbolic functions. The advantage of using hyperbolic functions in the thermal flux expansion lies in the accuracy with which hyperbolic functions can represent the large gradients at the interface between unlike fuel assemblies. The hyperbolic expansion functions proposed in (Pal 1997) are the analytic solutions of the zero-source diffusion equation for the thermal flux. The specific form of the QDLO equations requires the derivation of new hyperbolic basis functions which are different from those proposed for the diffusion equation. We have developed a discretization of the QDLO equations with node-averaged cross-sections and Eddington tensor components, solving the 2-D equations using the weighted residual method (Ame 1992). These node-averaged data are assumed known from single assembly transport calculations. We wrote a code in "Mathematica" that solves k-eigenvalue problems and calculates neutron fluxes in 2-D Cartesian coordinates. Numerical test problems show that the model proposed here can reproduce the results of both the simple diffusion problems presented in (Pal 1997) and those with analytic solutions. While the QDLO calculations performed on one-node, zero-current, boundary condition diffusion problems and two-node, zero-current boundary condition problems with UO₂-UO₂ assemblies are in excellent agreement with the benchmark and analytic solutions, UO₂-MOX configurations show more important discrepancies that are due to the single-assembly homogenized cross-sections used in the calculations. The results of the multiple-node problems show similar discrepancies in power distribution with the results reported in (Pal 1997). Multiple-node k-eigenvalue problems exhibit larger discrepancies, but these can be diminished by using adjusted diffusion coefficients (Pal 1997). The results of several "transport" problems demonstrate the influence of Eddington functionals on homogenized flux, power distribution, and multiplication factor k. / Graduation date: 2003

CANDU (CANada Deuterium Uranium) reactor utilizes Pressure Tube (PT) fuel channel design and heavy water as a coolant. Fuel channel annulus gas acts as an insulator to minimize heat losses from the coolant to the moderator. Since fuel bundles are continuously under high neutron fluxes, annulus gas nuclides undergo different nuclear transformations generating new composition of the gas that might have different physical properties which are undesirable for the annulus system. In addition, gas nuclides become radioactive and lead to an increase of the radioactive material inventory in the reactor and consequently to an increase of radiation levels. Pressure Tube Reactor (PTR) and Pressure Tube Supercritical Water Reactor (PT SCWR) fuel channel models have been developed in Monte Carlo N-Particle (MCNP) code. Neutron fluxes in the fuel channel annulus gas have been obtained by simulating different types of neutron sources in both PTR and PT SCWR fuel channels. Transmutation rates of annulus gases have been calculated for different gases (CO2, N2, Ar and Kr) at different pressures and temperatures in both fuel channels. The variation of the transmutation rates, neutron fluxes and gas densities in the annulus gas have been investigated in PTR and PT SCWR fuel channels at constant pressures and different temperatures. MCNP code along with NIST REFPROP [14] and other software tools have been used to conduct the calculations. / UOIT

Measurements of the neutron emission, resulting from nuclear fusion reactions between the hydrogen isotopes deuterium and tritium, can provide a wealth of information on the confinement properties of fusion plasmas and how these are affected by Magneto-Hydro-Dynamic (MHD) instabilities. This thesis describes work aimed to develop neutron measurement techniques for nuclear fusion plasma experiments, specifically regarding the performance and design of collimated neutron flux monitors (neutron cameras) for the Mega Ampere Spherical Tokamak, MAST, and for MAST Upgrade. The first part of the thesis focuses on the characterization of a prototype neutron camera installed at MAST and provides an account of the very first measurements of the neutron emissivity along its collimated fields of view. It is shown that the camera has sufficient temporal and spatial resolution to measure the effect of MHD instabilities on the neutron emissivity. The neutron camera fulfils the requirement on the measurements of the neutron count rate profile with less than 10 % statistical uncertainty in a time resolution of 1 ms. The instrument's more rudimentary capabilities to provide information on the neutron energy distribution are also presented and discussed. The encouraging results obtained with the prototype neutron camera show the potential of a collimated neutron flux monitor at MAST and suggest that an upgraded instrument for MAST Upgrade will provide crucial information on fast ions behavior and other relevant physics issues. The design of such an upgraded instrument for MAST Upgrade is discussed in the second part of the thesis. Two design options are explored, one consisting of two collimator arrays in the horizontal direction, another more traditional design with lines-of-sight in the poloidal cross section plane. On the basis of the experience gained with the prototype neutron camera and on the exploratory design and estimated performance for the upgraded camera presented here, a conceptual design of a neutron camera upgrade is proposed.

No description available.

The reported lifetime in an in-beam neutron lifetime experiment performed at NIST was tn = (886.3 ± 3.4) s. The largest source of uncertainty was the efficiency of the neutron flux monitor (0.3% relative uncertainty). The flux monitor operates by counting charged particles produced when neutrons impinge on a 6Li foil. Its efficiency was calculated from the 6Li thermal neutron cross section, the solid angle subtended by the charged particle detectors, and the amount of neutron-absorbing material present on the foil. An absolute black neutron detector for cold neutron beams has been developed to measure the efficiency without the need to know these quantities. The flux monitor efficiency is measured to a precision of 0.052% using this direct calibration technique. This calibration removes the largest barrier to a 1 s neutron lifetime measurement with the beam technique. It is hoped that this data can also be used to re-evaluate the current NIST neutron lifetime value, reduce its uncertainty, and remove the dependence on evaluated nuclear data files. There is also the possibility for a direct measurement of the 6Li thermal neutron cross section.

Thesis (M.S. in Physics)--Naval Postgraduate School, January 1964. / Thesis Advisor(s): Hawns, William W. "January 1964." Description based on title screen as viewed on June 2, 2010 DTIC Descriptor(s): (Neutron Flux), (Research Reactors, Measurement, Neutron Activation, Reactor Lattice Parameters, Reactor Control, Power, Reactor Operation, Reactor Power Density, Scintillation Counters, Foils (Materials), Gold, Gamma Rays, Nuclear Radiation Spectrometers, Indium, Monitors, Errors, Probability, Statistical Analysis. DTIC Identifier(s): AGN-201 Reactors. Includes bibliographical references (p. 20). Also available in print.

Thesis (M.S. in Nuclear Engineering)--Naval Postgraduate School, January 1963. / Thesis Advisor(s): Gerba, Alex. "January 1963." Description based on title screen as viewed on June 2, 2010. DTIC Descriptor(s): Research Reactors, Control Systems, Neutron Detectors, Neutron Flux, Reaction Kinetics, Equations, Simulation, Production Reactors, Test Reactors, Computer Programming, Analog Computers, Digital Computers, Nuclear Reactions, Performance (Engineering), Specifications, Neutron Activation. DTIC Identifier(s): AGN-201 Reactors. Includes bibliographical references (p. 54-55). Also available in print.

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