Eric C. Miller

Computational evaluation of neutron multiplicity measurements of polyethylene-reflected plutonium metal

Abstract Simulations of neutron multiplicity measurements of a highly multiplicative plutonium sphere measured with a moderated array of 3He proportional counters have consistently overpredicted the mean and variance of the measured multiplicity distribution. In contrast, identical experiments using a 252Cf source have been accurately simulated. This paper outlines a sensitivity analysis of several key parameters that could account for the overprediction in the simulation of the plutonium sphere. Parameters that were analyzed include source-detector distance, detector dead time, variations in density and volume of the plutonium, and the value of for v̄ 239Pu-induced fission. Of these parameters, the only factor that accounted for the overprediction within reasonable bounds was a change in the value of the 239Pu v̄. The sensitivity analysis showed that a small change (1.14% reduction) in the value of v̄ dramatically improved the simulated results.

Neutron and gamma-ray cross-correlation measurements of MOX fuel using liquid scintillators

The detection of special nuclear material (SNM) is a very difficult problem: typically, the signals are weak thus difficult to distinguish from the background and in many practical applications additional shielding can be expected. In order to detect SNM more accurately, new methods to isolate the source signal are needed. Using cross-correlation functions it may be possible to identify the time-correlated particles from fission events above the background. To examine this feasibility, measurements were performed on 1-kg MOX fuel samples, a 252Cf source, and an AmBe source. Using pulse shape discrimination the particle types were isolated and the individual correlation contributions were determined. By comparing the (n, n) peak of the correlation curves a distinct difference is observed between a spontaneous fission source and an (α, n) source.

Validation of MCNPX-PoliMi fission models

We present new results on the measurement of correlated, outgoing neutrons from spontaneous fission events in a Cf-252 source. 16 EJ-309 liquid scintillation detectors are used to measure neutron-neutron correlations for various detector angles. Anisotropy in neutron emission is observed. The results are compared to MCNPX-PoliMi simulations and good agreement is observed.

The estimation of neutron energy spectra of nuclear materials by passive measurements for nuclear nonproliferation applications

The development of technologies to accurately characterize nuclear materials is becoming an area of significant importance in the United States and internationally. The development of new measurement systems with improved characterization capabilities is a top priority. Specifically, the study of neutron detection via passive measurement yields many opportunities. In this research, passive measurements were performed on mixed-oxide samples (MOX) and several isotopic neutron sources at the Joint Research Centre in Ispra, Italy in June of 2010. These measurements are used to test the capabilities of a neutron detection system and analysis algorithms developed at the University of Michigan. The measurements are also utilized to evaluate the use of liquid scintillation detectors for the determination of neutron energy distributions emitted from MOX samples and from various isotopic neutron sources. This measurement campaign provided the opportunity to identify and evaluate new candidates for the replacement of current detector technologies.

Measurement of detector resolution for neutral particle detection with liquid scintillators

The characterization of the photon energy resolution of a liquid scintillation detector is desirable for simulating photon pulse height distributions in Monte Carlo codes such as MCNPX-PoliMi. Previous studies in literature determined a function characterizing the photon energy resolution for four photon detectors using the wide-angle Compton coincidence technique. The goal of this work is to utilize this technique to measure the photon resolution for the EJ-309 organic scintillation detector. The photon energy resolution was measured for an EJ-309 organic scintillation detector using a high-purity germanium (HPGe) detector and a 100-μCi 137Cs source. The photon energy resolution of the EJ-309 was measured to be between 16 and 28% for photon energies between 390 and 280 keV.

Simulations of Neutron Multiplicity Measurements with MCNP-PoliMi

The heightened focus on nuclear safeguards and accountability has increased the need to develop and verify simulation tools for modeling these applications. The ability to accurately simulate safeguards techniques, such as neutron multiplicity counting, aids in the design and development of future systems. This work focuses on validating the ability of the Monte Carlo code MCNPX-PoliMi to reproduce measured neutron multiplicity results for a highly multiplicative sample. The benchmark experiment for this validation consists of a 4.5-kg sphere of plutonium metal that was moderated by various thicknesses of polyethylene. The detector system was the nPod, which contains a bank of 15 3He detectors. Simulations of the experiments were compared to the actual measurements and several sources of potential bias in the simulation were evaluated. The analysis included the effects of detector dead time, source-detector distance, density, and adjustments made to the value of {nu}-bar in the data libraries. Based on this analysis it was observed that a 1.14% decrease in the evaluated value of {nu}-bar for 239Pu in the ENDF-VII library substantially improved the accuracy of the simulation.

Plutonium sphere multiplicity simulations with MCNP-PoliMi

In order to improve the characterization methods for fissile materials, effort must be made to validate the computer codes that are used to simulate the behavior of these systems. For this work, measurements of a 4.4-kg sphere of weapons grade plutonium metal were taken. A detector array of 3He tubes embedded in polyethylene was used to measure the multiplicity of the system. The experiment was then modeled using the MCNP-PoliMi code and the simulated results were compared to the measured results using the Feynman-Y metric. MCNP-PoliMi is able to correctly predict the measured value of the Feynman-Y vale within 5% for all of the moderated cases and within 22% for the bare sphere.

Time dependent neutron detector response simulation for 252 Cf

Liquid scintillator detectors are valuable for detection of special nuclear material since they are capable of detecting both neutrons and gamma rays. Furthermore, scintillators can also provide energy information which can help in identification and characterization of the source. In order to deign scintillation based measurement systems appropriate simulation tools are needed. MCNPX-PoliMi is capable of simulating scintillator detector response, however, simulations have traditionally been run in analog mode which leads to longer computation times. In this paper nonanalog MCNPX-PoliMi mode which uses variance reduction techniques is applied and tested. The nonanalog MCNPX-PoliMi simulation test case uses source biasing variance reduction technique to efficiently simulate pulse height distribution and then time-of-flight for a heavily shielded case with a 252Cf source. Neutron time-of-flight measurements are essential to nonproliferation and safeguards efforts to identify and characterize special nuclear material. The resulting simulation is marked by an improvement in the figure of merit (FOM) by a factor of 3.4.