▼ This research is based on the concept that the diversion of nuclear fuel pins from Light Water Reactor (LWR) spent fuel assemblies is feasible by a careful comparison of spontaneous fission neutron and gamma levels in the guide tube locations of the fuel assemblies. The goal is to be able to determine whether some of the assembly fuel pins are either missing or have been replaced with dummy or fresh fuel pins. It is known that for typical commercial power spent fuel assemblies, the dominant spontaneous neutron emissions come from Cm-242 and Cm-244. Because of the shorter half-life of Cm-242 (0.45 yr) relative to that of Cm-244 (18.1 yr), Cm-244 is practically the only neutron source contributing to the neutron source term after the spent fuel assemblies are more than two years old. Initially, this research focused upon developing MCNP5 models of PWR fuel assemblies, modeling their depletion using the MONTEBURNS code, and by carrying out a preliminary depletion of a ¼ model 17x17 assembly from the TAKAHAMA-3 PWR. Later, the depletion and more accurate isotopic distribution in the pins at discharge was modeled using the TRITON depletion module of the SCALE computer code. Benchmarking comparisons were performed with the MONTEBURNS and TRITON results. Subsequently, the neutron flux in each of the guide tubes of the TAKAHAMA-3 PWR assembly at two years after discharge as calculated by the MCNP5 computer code was determined for various scenarios. Cases were considered for all spent fuel pins present and for replacement of a single pin at a position near the center of the assembly (10,9) and at the corner (17,1). Some scenarios were duplicated with a gamma flux calculation for high energies associated with Cm-244. For each case, the difference between the flux (neutron or gamma) for all spent fuel pins and with a pin removed or replaced is calculated for each guide tube. Different detection criteria were established. The first was whether the relative error of the difference was less than 1.00, allowing for the existence of the difference within the margin of error. The second was whether the difference between the two values was big enough to prevent their error bars from overlapping. Error analysis was performed both using a one second count and pseudo-Maxwell statistics for a projected 60 second count, giving four criteria for detection. The number of guide tubes meeting these criteria was compared and graphed for each case. Further analysis at extremes of high and low enrichment and long and short burnup times was done using data from assemblies at the Beaver Valley 1 and 2 PWR. In all neutron flux cases, at least two guide tube locations meet all the criteria for detection of pin diversion. At least one location in almost all of the gamma flux cases does. These results show that placing detectors in the empty guide tubes of spent fuel bundles to identify possible pin diversion is feasible. Advisors/Committee Members: Christenson, John.

▼ The Pebble Bed Reactor (PBR) is a helium-cooled, graphite-moderated high temperature nuclear power reactor. In addition to its inherently safe design, a unique feature of this reactor is its multipass fuel cycle in which graphite fuel pebbles (of varying enrichment) are randomly loaded and continuously circulated through the core until they reach their prescribed end-of-life burnup limit (~ 80,000-100,000 MWD/MTU). Unlike the situation with conventional light water reactors (LWRs), depending solely on computational methods to perform in-core fuel management will be highly inaccurate. As a result, an on-line measurement approach becomes the only accurate method to assess whether a particular pebble has reached its end-of-life burnup limit. In this work, an investigation was performed to assess the feasibility of passive gamma-ray spectrometry assay as an approach for on-line interrogation of PBR fuel for the simultaneous determination of burnup and enrichment on a pebble-by-pebble basis. Due to the unavailability of irradiated or fresh pebbles, Monte Carlo simulations were used to study the gamma-ray spectra of the PBR fuel at various levels of burnup. A pebble depletion calculation was performed using the ORIGEN code, which yielded the gamma-ray source term that was introduced into the input of an MCNP simulation. The MCNP simulation assumed the use of a high-purity coaxial germanium detector. Due to the lack of one-group high temperature reactor cross sections for ORIGEN, a heterogeneous MCNP model was developed to describe a typical PBR core. Subsequently, the code MONTEBURNS was used to couple the MCNP model and ORIGEN. This approach allowed the development of the burnup-dependent, one-group spectral-averaged PBR cross sections to be used in the ORIGEN pebble depletion calculation. Based on the above studies, a relative approach for performing the measurements was established. The approach is based on using the relative activities of Np-239/I-132 in combination with the relative activities of Cs-134/Co-60 (Co-60 is introduced as a dopant) to yield the burnup and enrichment for each pebble. Furthermore, a direct consequence of the relative approach is the ability to apply a self-calibration scheme using the multiple gamma lines of Ba-La-140 to establish the relative efficiency curve of the HPGe detector. An assessment of the expected uncertainty components in this approach showed that a maximum uncertainty of less than 5% should be feasible. To confirm the above findings, gamma-ray scans were performed on irradiated PULSTAR reactor fuel assemblies at North Carolina Sate University. The measurements used a 40% efficient n-type coaxial HPGe detector connected to an ORTEC DSPECplus digital Gamma-Ray Spectrometer, and a data acquisition computer. The obtained results showed consistency with the predictions of the simulations including the observation of the I-132, Cs-134, Np-239 uncontaminated gamma lines. In addition, the Ba-La-140 lines were clearly observed confirming the ability to perform relative calibration of the spectrometer. Advisors/Committee Members: Hawari, Ayman I.

▼ Different types of radiation detectors are routinely used for the dosimetry of photon beams. Finite detector sizes have certain effects to the broadening of the measured beam penumbra. The problem is more important in small field measurement, such as stereotactic radiosurgery, small beamlet IMRT, etc. The dosimetry associated with small fields is very difficult because of the steep dose gradients and the lack of lateral electronic equilibrium conditions that complicate the interpretation of the dose measurement. Many Researchers have investigated this problem from different points of view utilizing, for example, extrapolation method, analytical method. But their studies were all measurements based. In this study, we investigated the problem using Monte Carlo simulation method. Compared with practical measurements, the advantages of using Monte Carlo simulation are: 1. Simulation can be performed in a scenario where radiation dosimetry is technically difficult or even impossible to accomplish; 2. Possible systematic errors, e.g., setup errors, reading errors, can be eliminated; 3. Simulation of radiation detectors which are not readily available allowed the study of a wider range of detector sizes. In this study we used Monte Carlo methods to develop and apply detector response functions (DRFs) for three types of clinically available radiation detectors and two theoretical detectors. Detector response functions were determined by deconvolving known values of input (simulated true data from Monte Carlo simulation) and output (simulated empirical data from Monte Carlo simulation or empirical data from radiation dosimetry). Deconvolved detector response functions were applied to typical stereotactic radiosurgery fields to obtain the true beam profile. This application was then benchmarked by both Monte Carlo simulation method and dosimetry methods, which include diode dosimetry, radiographic film dosimetry, and Gafchromic film dosimetry. The results of this research demonstrate: 1. Detector response function of cylindrical detectors can be approximately represented as a Gaussian distribution dependent upon the radius of the detector; 2. Deconvolution method can create a more realistic beam profile by reducing the detector size effect, however it can not completely remove this effect limited by the inaccuracy derived from the Fourier transform-based nature of this procedure; 3. Diode dosimetry and Gafchromic film dosimetry both yield satisfactory beam profiles in small field relative measurements and are the preferred measurement techniques. Advisors/Committee Members: Spitz, Dr. Henry B.

▼ This work presents a method to empirically determine the photon spectrum of a megavoltage bremsstrahlung beam. The method makes use of the fact that scatter cross sections vary in a known fashion with incident photon energy. The distribution of scatter produced by a scattering object placed in a good geometry photon beam was measured. The scatter distribution was simulated for a series of monoenergetic good geometry photon beams. A system of linear equations was generated to combine the polyenergetic measurements with the monoenergetic simulations. Regularization techniques were applied to solve the system for the incident photon spectrum. Monte Carlo simulations were used to predict the ratio of primary to scattered photons for narrow mono-energetic photon beams at 9 different locations, with 10 degree increments and 15 cm from a scattering material. Measurements were performed in the same geometry using the photon beams produced by linear accelerators. A linear matrix system, A×F=T, was developed to describe the scattering interactions and their relationship to the primary spectrum. A is the monoenergetic scatter kernel determined from the Monte Carlo simulations, F is the incident photon spectrum, and T represents the scatter distribution characterized by empirical measurement. Direct matrix inversion methods produce results that are not physically consistent due to errors inherent in the system. Tikhonov regularization methods were applied to address the effects of these errors and solve the system for physically consistent bremsstrahlung spectra. The results of this research provide a method to empirically characterize the primary radiation energy spectrum produced by a linear accelerator. Key words: photon energy, spectrum, scatter analysis, regularization, Tikhonov Advisors/Committee Members: Spitz, Henry.

▼ The Monte Carlo code MCNP4B was used to investigate the radiation dose distribution in several areas of intravascular brachytherapy that would be very difficult or impossible to investigate experimentally or analytically. A model for a liquid-filled balloon catheter was developed and validated by comparing the results of Monte Carlo simulations with experimental measurements made in a tissue equivalent phantom. The dose distribution in the coronary vessel wall from a liquid-filled balloon catheter containing the following radioisotopes was examined: Y-90, Re-188, P-32, Re-186, Sm-153, In-111, and Tc-99m. At 0.5 mm from the vessel surface, the beta emitters deliver a higher dose per unit cumulated activity than the gamma emitters. The gamma emitters, however, deliver a dose that is more uniform throughout the vessel wall. Eight sizes of balloon catheters were modeled and the effect of balloon catheter size on the radiation dose distribution was examined. Effects that perturb the radiation dose distribution for a liquid-filled balloon catheter were investigated. These perturbing effects were: (1) an air bubble within the balloon catheter, (2) contrast media within the balloon catheter, (3) a Palmaz-Schatz stent surrounding the balloon catheter, and (4) the deflation of the balloon catheter. The radiation dose distribution produced from a P-32 coated Palmaz-Schatz stent was investigated by developing a model with P-32 deposited on the stent surface to a depth of one micron, and calculating the dose delivered to the coronary vessel wall. High dose fluctuations were observed near the surface of the stent, but these fluctuations leveled off at depth in the vessel wall. The radiation dose distribution for the case of direct injection of radioisotopes into the coronary vessel wall using the Infiltrator angioplasty balloon catheter (IABC) was also investigated. The source distribution produced by the IABC was modeled for two configurations within the vessel wall: (1) uniform to a depth of 0.5 mm into the vessel wall, and (2) confined to discrete pools surrounding the delivery injection needles. The dose distribution for the following radioisotopes was examined: Re-188, Re-186, Sm-153, In-111, I-123, and Tc-99m. Advisors/Committee Members: Spitz, Dr. Henry B.

▼ This work describes the magnetic resonance (MR) imaging techniques and image processing algorithms developed for radiation dosimetry with the superheated emulsion chamber. The chamber contains an emulsion of chloropentafluoroethane droplets in a tissue-equivalent glycerin-based gel. The droplets are highly superheated and expand into vapor bubbles upon exposure to irradiation. Brachytherapy sources can be inserted into the superheated emulsion chamber to create distributions of bubbles. The distribution of bubbles is then representative of the dose distribution to which the emulsion is exposed. Cumulating data from multiple independent exposures is required to calculate statistically significant bubble densities. MR imaging is well suited to determining the bubble distribution. Susceptibility gradients at the interfaces between bubbles and gel are exploited to enhance contrast so microscopic bubbles can be imaged using relatively large voxel sizes. A conventional three-dimensional gradient echo imaging method is developed and applied to multiple independent irradiations of the superheated emulsion chamber from an 125I source. An image post-processing technique is developed to semi-automatically segment the bubbles from the images and to assess dose distributions based on the measured bubble densities. Relative bubble densities compare favorably to relative radial dose distributions calculated as recommended by Task Group 43 (TG43) of the American Association of Physicists in Medicine as well as Monte Carlo radiation transport simulations. A three-dimensional, segmented, double sampled, echo-planar imaging (EPI) technique is subsequently developed and applied to an 125I source. Combining two-dimensional EPI with a conventional phase encode in the third dimension provides for rapid acquisition of susceptibility weighted images. Segmentation reduces artifacts produced by magnetic field inhomogeneities, while double sampling removes Nyquist ghosting. Post-processing is performed to segment the bubbles and to generate two-dimensional relative bubble density curves. Monte Carlo generated corrections are applied to convert bubble density to dose to water. The results are compared to TG43 using a distance to agreement map. It is shown that the superheated emulsions can be extended from performing point radiation dosimetry to generating one-dimensional and two-dimensional radiation dose maps. Advisors/Committee Members: Spitz, Henry.

▼ This research utilized Monte Carlo N-Particle version 4C (MCNP4C) to simulate K X-ray fluorescent (K XRF) measurements of stable lead in bone. Simulations were performed to investigate the effects that overlying tissue thickness, bone-calcium content, and shape of the calibration standard have on detector response in XRF measurements at the human tibia. Additional simulations of a knee phantom considered uncertainty associated with rotation about the patella during XRF measurements. Simulations tallied the distribution of energy deposited in a high-purity germanium detector originating from collimated 88 keV 109 Cd photons in backscatter geometry. Benchmark measurements were performed on simple and anthropometric XRF calibration phantoms of the human leg and knee developed at the University of Cincinnati with materials proven to exhibit radiological characteristics equivalent to human tissue and bone. Initial benchmark comparisons revealed that MCNP4C limits coherent scatter of photons to six inverse angstroms of momentum transfer and a Modified MCNP4C was developed to circumvent the limitation. Subsequent benchmark measurements demonstrated that Modified MCNP4C adequately models photon interactions associated with in vivo K XRF of lead in bone. Further simulations of a simple leg geometry possessing tissue thicknesses from 0 to 10 mm revealed increasing overlying tissue thickness from 5 to 10 mm reduced predicted lead concentrations an average 1.15% per 1 mm increase in tissue thickness (p<0.0001). An anthropometric leg phantom was mathematically defined in MCNP to more accurately reflect the human form. A simulated one percent increase in calcium content (by mass) of the anthropometric leg phantom’s cortical bone demonstrated to significantly reduce the K XRF normalized ratio by 4.5% (p<0.0001). Comparison of the simple and anthropometric calibration phantoms also suggested that cylindrical calibration standards can underestimate lead content of a human leg up to 4%. The patellar bone structure in which the fluorescent photons originate was found to vary dramatically with measurement angle. The relative contribution of lead signal from the patella declined from 65% to 27% when rotated 30°. However, rotation of the source-detector about the patella from 0 to 45° demonstrated no significant effect on the net K XRF response at the knee. Advisors/Committee Members: Spitz, Dr. Henry B.

▼ In this research a Stochastic Parameter Simulation System (SPSS) computer program employing wavelet techniques was developed. The SPSS was designed to fulfill two key functional requirements:1. To be able to analyze any steady state plant signal, decompose it into its deterministic and stochastic components, and then reconstruct a new, simulated signal that possesses exactly the same statistical noise characteristics as the actual signal; and 2. To be able to filter out the principal serially-correlated, deterministic components from the analyzed signal so that the remaining stochastic signal can be analyzed with signal validation tools that are designed for signals drawn from independent random distributions. The results obtained using SPSS were compared to those obtained using the Argonne National Laboratory Reactor Parameter Simulation System (RPSS) which uses a Fourier transform methodology to achieve the same objectives. RPSS and SPSS results were compared for three sets of stationary signals, representing sensor readings independently recorded at three nuclear power plants. For all of the recorded signals, the wavelet technique provided a better approximation of the original signal than the Fourier procedure. For each signal, many wavelet-based decompositions were found by the SPSS methodology, all of which produced white and normally distributed signal residuals. In most cases, the Fourier-based analysis failed to completely eliminate the original signal serial-correlation from the residuals. The reconstructed signals produced by SPSS are also statistically closer to the original signal than the RPSS reconstructed signal. Another phase of the research demonstrated that SPSS could be used to enhance the reliability of the Multivariate Sensor Estimation Technique (MSET). MSET uses the Sequential Probability Ratio Test (SPRT) for its fault detection algorithm. By eliminating the MSET residual serial-correlation in the MSET training phase, the SPRT user-defined false alarm rates can be met, even for signals which contain serially-correlated components. Advisors/Committee Members: Christenson, Dr. John.

▼ A new method was developed to measure the effectiveness of water in reducing the release of radon emanating from 226 Ra-bearing sand into air. Fick’s law on diffusion was used to model the transport of radon in water including the impact associated with radioactive decay. A multi-region, one-dimensional, steady-state transport model was used to analyze the movement of radon through a sequential column of air, water and air. An effective diffusion coefficient was determined by varying the thickness of the water column to predict the transport of 222 Rn through particular thickness of water. A one-region, one-dimensional transient diffusion equation was developed to investigate the build up of radon at the end of the water column until a steady-state, equilibrium condition was achieved. This build up with time is characteristic of the transport rate of radon in water and established the basis for estimating the effective diffusion coefficient for 222 Rn in water. The results suggest that convective forces other than molecular diffusion impact the transport of 222 Rn through the water barrier. An effective diffusion coefficient is defined that includes effects of molecular diffusion and convection to describe the transport of radon in water. Several experimental arrangements were evaluated to examine the influence of physical parameters on the radon transport. The effective diffusion coefficients measured in these experiments are 6.8×10 -4 ± 28% and 3.5×10 -4 ± 34% cm 2  sec -1 for the steady-state and transient diffusion approaches, respectively. Water barriers ranging in thickness from 30 – 50 cm reduce the amount of radon released from the radium-bearing source material by a factor of 0.3 – 0.1, respectively. Advisors/Committee Members: Spitz, Dr. Henry B.

▼ The goal of this research was to determine whether neutron dose reconstruction could be improved through re-analysis of historic NTA films worn by workers in the 1950 through the 1970s. To improve neutron dose reconstruction, the underlying neutron energy spectra is critical in determining the organ dose due to energy dependence of the dose conversion factor as well as the application of radiation weighting factors used in epidemiology and probability of causation calculations. Monte Carlo models of proton-recoil track length distributions were developed and benchmarked against measurement data for both NTA and Ilford films. These models, when applied to several NTA film dosimeter configurations, demonstrated that proton-recoil track length distributions change based upon incident neutron energy. The neutron energy spectra changes that result from the general work environment such as source term and shielding can subsequently be modeled to predict the response of the NTA film dosimeter. An Automatic NTA Film Analyzer has been designed and developed to determine if the difference in proton-recoil track length distributions predicted by the Monte Carlo models could be measured and whether these differences could be correlated to the incident neutron energy spectra. The design required the development of a 2D-3D hybrid track recognition algorithm for a three dimensional analysis of the NTA film in order to accurately determine the proton-recoil track length for subsequent neutron energy determination. NTA films exposed to a plutonium fluoride (PuF4) and polonium boron (PoB) calibration sources were measured and compared. The proton-recoil track lengths were used to reconstruct the incident neutron energy spectra demonstrating the functionality of the analyzer and that reconstruction of the neutron energy spectra from NTA films is feasible. These measurements were compared to the Monte Carlo models and confirmed the applicability of using models to determine the NTA film response. Based on the Monte Carlo Modeling and the Automatic NTA Film Analyzer, the neutron energy spectra to which an individual worker was exposed can be retrospectively determined through re-analysis of historic NTA dosimeter films thus improving retrospective neutron dose assessments. Advisors/Committee Members: Spitz, Henry.

▼ The ability to accurately predict the multiplication factor (k eff ) of a nuclear reactor core as a function of exposure continues to be an elusive task for core designers despite decades of advances in computational methods. The difference between a predicted eigenvalue (target) and the actual eigenvalue at critical reactor conditions is herein referred to as the “eigenvalue drift.” This dissertation studies exposure-dependent eigenvalue drift using MCNP-based fuel management analysis of the ORNL High Flux Isotope Reactor core. Spatial-dependent burnup is evaluated using the MONTEBURNS and ALEPH codes to link MCNP to ORIGEN to help analyze the behavior of k eff as a function of fuel exposure. Understanding the exposure-dependent eigenvalue drift of a nuclear reactor is of particular relevance when trying to predict the impact of major design changes upon fuel cycle behavior and length. In this research, the design of an advanced HFIR core with a fuel loading of 12 kg of 235 U is contrasted against the current loading of 9.4 kg. The goal of applying exposure dependent eigenvalue characterization is to produce a more accurate prediction of the fuel cycle length than prior analysis techniques, and to improve our understanding of the reactivity behavior of the core throughout the cycle. This investigation predicted a fuel cycle length of 40 days, representing a 50% increase in the cycle length in response to a 25% increase in fuel loading. The average burnup increased by about 48 MWd/kg U and it was confirmed that the excess reactivity can be controlled with the present design and arrangement of control elements throughout the core’s life. Another major design change studied was the effect of installing an internal beryllium reflector upon cycle length. Exposure dependent eigenvalue predictions indicate that the actual benefit could be twice as large as that originally assessed via beginning-of-life (BOL) analyses. Advisors/Committee Members: Maldonado, G. Ivan.

▼ The Boltzmann equation, which is often used to describe neutral particle transport process, is impossible to solve analytically in most cases. Solving the equation by a numerically method could also be very time-consuming for some problems. Therefore, various approximation methods have been developed for particle transport calculations. In this dissertation, a new Flux-Limited diffusion method (DM) is investigated as a low-order approximation to the transport equation. Specifically, the new Flux-Limited diffusion method (DM) is derived based on the Minerbo maximum entropy Eddington factor and corresponding boundary conditions are developed. Performance of the DM method is tested for a variety of particle transport problems: including steady-state problems, transient problems, one-dimensional problems, two-dimensional problems, flux distribution problems, eigen-value problems, one-group problems, and two-group problems. The tests demonstrate that compared with other approximations, the DM method has the following good features. (1) The DM method is a low order approximation as the classic diffusion method but can improve the accuracy of classic diffusion method by a factor up to 10. (2) The DM method has relatively fast computing speed compared with other high order approximations and its computation cost is not too expensive compared with the classic diffusion method. (3) The DM method can predict proper traveling speed of particles in time-dependent problems. (4) The DM method is valid for neutral particle transport in systems containing strong absorbers, regions near particle sources or boundaries, and systems where strongly anisotropic scattering of particles occur. (5) The DM method is superior to other Flux-Limited diffusion methods available in the literature. (6) The solutions of the DM method are numerically stable in all the circumstances. Overall, the DM method is promising as a low-order approximation for particle transport calculations in some applications. Advisors/Committee Members: Dr. Su, Bingjing.

▼ On-line signal validation is essential for safe and economic operations of a complicated industrial system such as a nuclear power plant. Various signal validation methods based on empirical signal estimation have been developed and successfully used. The first part of the thesis addresses a common and unavoidable problem for these methods -fault propagation, which causes false identification of healthy signals as faulty ones because of the faults existing in other signals. This effect is especially serious when faults occur in multiple signals and/or during system transient. A sub-grouping technique is presented in the thesis to prevent the effect of fault propagation in general signal validation methods. Specifically, two methods, Subgroups Consistency Check (SCC) and Subgroups Voting (SV), are developed. Their effectiveness is demonstrated by using a well-known Multivariate State Estimation technique (MSET) as a general method of signal estimation. To further improve the performance of MSET estimation, a procedure called Feedback Once (FBO) is also developed. All these new methods are tested and compared with MSET by using real transient data from a reactor startup process in a nuclear power plant. The results show that false identification of signals caused by fault propagation is significantly reduced by the two sub-grouping methods and the FBO method is able to improve the performance of MSET estimation to some extent. The results demonstrate that implementation of these new methods can lead to an improved signal validation technique that remains effective even when faults occur in multiple signals during system transients. The other major contribution is on the improvement of statistical test used for signal validation. Sequential Probability Ratio Test (SPRT) is a popular method that has been widely used in many signal validation methods. However, the assumption of SPRT is too stringent to satisfy in practice, which may cause the false identification rate exceeding the preset tolerance. In this thesis, a Sequential Rank-sum Probability Ratio Test (SRPRT) method is developed. This method is similar to SPRT in procedure but is based on a much weaker assumption that can be easily satisfied. The demonstrations show that SRPRT yields a smaller false identification rate than SPRT and is always below the preset tolerance. Advisors/Committee Members: Su, Dr. Bingjing.

▼ Modular Pebble Bed Reactor (MPBR) is a high temperature gas-cooled nuclear power reactor. In addition to its inherently safe design, a unique feature of this reactor is its multi-pass fuel circulation in which the fuel pebbles are randomly loaded and continuously cycled through the core until they reach their prescribed End-of-Life burnup limit. Unlike the situation with a conventional light water reactor, depending solely on computational methods to perform in-core fuel management for MPBR will be highly inaccurate. An on-line measurement system is needed to accurately assess whether a given pebble has reached its End-of-Life burnup limit and thereby provide an on-line, automated go/no-go decision on fuel disposition on a pebble-by-pebble basis. This project investigated the feasibility of using the passive neutron counting and active neutron/gamma counting for the on line fuel burnup measurement for MPBR. To investigate whether there is a correlation between neutron emission and fuel burnup, the MPBR fuel depletion was simulated under different irradiation conditions by ORIGEN2. It was found that the neutron emission from an irradiated pebble increases with burnup super-linearly and reaches to 10 4 neutron/sec/pebble at the discharge burnup. The photon emission from an irradiated pebble was found to be in the order of 10 13 photon/sec/pebble at all burnup levels. Analysis shows that the neutron emission rate of an irradiated pebble is sensitive to its burnup history and the spectral-averaged one-group cross sections used in the depletion calculations, which consequently leads to large uncertainty in the correlation between neutron emission and burnup. At low burnup levels, the uncertainty in the neutron emission/burnup correlation is too high and the neutron emission rate is too low so that it is impossible to determine a pebble’s burnup by on-line neutron counting at low burnup levels. At high burnup levels, the uncertainty in the neutron emission rate becomes less but is still large in quantity. However, considering the super-linear feature of the correlation, the uncertainty in burnup determination was found to be ~ 7% at the discharge burnup, which is acceptable. Therefore, total neutron emission rate of a pebble can be used as a burnup indicator to determine whether a pebble should be discharged or not. The feasibility of using passive neutron counting methods for the on-line burnup measurement was investigated by using a general Monte Carlo code, MCNP, to assess the detectability of the neutron emission and the capability to discriminate gamma noise by commonly used neutron detectors. It was found that both He-3 and BF 3 detector systems are able to satisfy the requirement on detection efficiency; but their gamma discrimination capability is only marginal. Even using thick gamma shielding, these two types of detectors shall deteriorate in performance after a limited period of operation time because of excess accumulative gamma exposures. Thus, two or more detector systems must be used alternatively for continuous measurement. On the other hand, fission counters were found that they can effectively discriminate gamma interference for this on-line application even without using gamma shield. However, detection efficiency of fission counters is low; thus a multi-fission-counter system (using at least 12 commercially available fission chambers) must be used to achieve the required detection efficiency. Overall, passive neutron counting could be used to provide an on-line, go/no-go decision on fuel disposition on a pebble-by-pebble basis for MPBR, if the detection system is well designed. Using active interrogation methods for this on-line burnup determination of fuel pebbles was also briefly studied. Results show that both active neutron counting and active gamma counting do not work for this application because the induced fission neutrons are overwhelmed by scattered source neutrons and the prompt fission gamma emission rate is much less than that of the decay gammas. In terms of gamma spectrometry measurement, the prompt fission gammas are stronger than the decay gammas in the high energy ranges; however, due to the extremely strong gamma emission at low energy and low gamma detection efficiency at high energy, it seems (though not confirmed) that active gamma spectrometry measurement is not promising either for this on-line burnup determination application. Advisors/Committee Members: Su, Dr. Bingjing.