A Novel Compact Pyroelectric X-Ray and Neutron Source

This research was focused on the utilization of pyroelectric crystals for generation of radiation. When in constant temperature pyroelectric crystals are spontaneously polarized. The polarization causes internal charges to accumulate near the crystal faces and masking charges from the environment are attracted to the crystal faces and neutralize the charge. When a pyroelectric crystal is heated or cooled it becomes depolarized and the surface charges become available. If the heating or cooling is done on a crystal in vacuum where no masking charges are available, the crystal becomes a charged capacitor and because of its small capacitance large potential develops across the faces of the crystal

Improvements to Nuclear Data and Its Uncertainties by Theoretical Modeling

This project addresses three important gaps in existing evaluated nuclear data libraries that represent a significant hindrance against highly advanced modeling and simulation capabilities for the Advanced Fuel Cycle Initiative (AFCI). This project will: Develop advanced theoretical tools to compute prompt fission neutrons and gamma-ray characteristics well beyond average spectra and multiplicity, and produce new evaluated files of U and Pu isotopes, along with some minor actinides; Perform state-of-the-art fission cross-section modeling and calculations using global and microscopic model input parameters, leading to truly predictive fission cross-sections capabilities. Consistent calculations for a suite of Pu isotopes will be performed; Implement innovative data assimilation tools, which will reflect the nuclear data evaluation process much more accurately, and lead to a new generation of uncertainty quantification files. New covariance matrices will be obtained for Pu isotopes and compared to existing ones. The deployment of a fleet of safe and efficient advanced reactors that minimize radiotoxic waste and are proliferation-resistant is a clear and ambitious goal of AFCI. While in the past the design, construction and operation of a reactor were supported through empirical trials, this new phase in nuclear energy production is expected to rely heavily on advanced modeling and simulation capabilities. To be truly successful, a program for advanced simulations of innovative reactors will have to develop advanced multi-physics capabilities, to be run on massively parallel super- computers, and to incorporate adequate and precise underlying physics. And all these areas have to be developed simultaneously to achieve those ambitious goals. Of particular interest are reliable fission cross-section uncertainty estimates (including important correlations) and evaluations of prompt fission neutrons and gamma-ray spectra and uncertainties

Continuous p-n junction with extremely low leakage current for micro- structured solid-state neutron detector applications

ABSTRACT Considerable progress has been achieved recently to enhance the thermal neutron detection efficiency of solid-state neutron detectors that incorporate neutron sensitive materials such as 10 B and 6 LiF in Si micro-structured p-n junction diode. Here, we describe the design, fabrication process optimization and characterization of an enriched boron filled honeycomb structured neutron detector with a continuous p + -n junction. Boron deposition and diffusion processes were carried out using a low pressure chemical vapor deposition to study the effect of diffusion temperature on current density-voltage characteristics of p + -n diodes. TSUPREM-4 was used to simulate the thickness and surface doping concentration of p + -Si layers. MEDICI was used to simulate the depletion width and the capacitance of the microstructured devices with continuous p + -n junction. Finally, current density-voltage and pulse height distribution of fabricated devices with 2.5×2.5 mm 2 size were studied. A very low leakage current density of ~2×10 -8 A/cm 2 at -1 V (for both planar and honeycomb structured devices) and a bias-independent thermal neutron detection efficiency of ~26% under zero bias voltage were achieved for an enriched boron filled honeycomb structured neutron detector with a continuous p + -n junction

Pyroelectric Crystal Accelerator In The Department Of Physics And Nuclear Engineering At West Point

Abstract. The Nuclear Science and Engineering Research Center (NSERC), a Defense Threat Reduction Agency (DTRA) office located at the United States Military Academy (USMA), sponsors and manages cadet and faculty research in support of DTRA objectives. The NSERC has created an experimental pyroelectric crystal accelerator program to enhance undergraduate education at USMA in the Department of Physics and Nuclear Engineering. This program provides cadets with hands-on experience in designing their own experiments using an inexpensive tabletop accelerator. This device uses pyroelectric crystals to ionize and accelerate gas ions to energies of ~100 keV. Within the next year, cadets and faculty at USMA will use this device to create neutrons through the deuterium-deuterium (D-D) fusion process, effectively creating a compact, portable neutron generator. The double crystal pyroelectric accelerator will also be used by students to investigate neutron, x-ray, and ion spectroscopy

Preliminary Results from Pyroelectric Crystal Accelerator

Abstract. The Nuclear Science and Engineering Research Center (NSERC), a Defense Threat Reduction Agency (DTRA) office located at the United States Military Academy (USMA), sponsors and manages cadet and faculty research in support of DTRA objectives. Cadets in the Department of Physics and Nuclear Engineering at USMA are using pyroelectric crystals to ionize and accelerate residual gas trapped inside a vacuum system. A system using two lithium tantalate crystals with associated diagnostics was designed and is now operational. X-ray energies of approximately 150 keV have been achieved. Future work will focus on developing a portable neutron generator using the D-D nuclear fusion process

Lead Slowing Down Spectrometer Status Report

This report documents the progress that has been completed in the first half of FY2012 in the MPACT-funded Lead Slowing Down Spectrometer project. Significant progress has been made on the algorithm development. We have an improve understanding of the experimental responses in LSDS for fuel-related material. The calibration of the ultra-depleted uranium foils was completed, but the results are inconsistent from measurement to measurement. Future work includes developing a conceptual model of an LSDS system to assay plutonium in used fuel, improving agreement between simulations and measurement, design of a thorium fission chamber, and evaluation of additional detector techniques

Lead Slowing-Down Spectrometry for Spent Fuel Assay: FY11 Status Report

Executive Summary Developing a method for the accurate, direct, and independent assay of the fissile isotopes in bulk materials (such as used fuel) from next-generation domestic nuclear fuel cycles is a goal of the Office of Nuclear Energy, Fuel Cycle R&D, Material Protection and Control Technology (MPACT) Campaign. To meet this goal, MPACT supports a multi-institutional collaboration to study the feasibility of Lead Slowing Down Spectroscopy (LSDS). This technique is an active nondestructive assay method that has the potential to provide independent, direct measurement of Pu and U isotopic masses in used fuel with an uncertainty considerably lower than the approximately 10% typical of today’s confirmatory assay methods. This document is a progress report for FY2011 collaboration activities. Progress made by the collaboration in FY2011 continues to indicate the promise of LSDS techniques applied to used fuel. PNNL developed an empirical model based on calibration of the LSDS to responses generated from well-characterized used fuel. The empirical model demonstrated the potential for the direct and independent assay of the sum of the masses of 239Pu and 241Pu to within approximately 3% over a wide used fuel parameter space. Similar results were obtained using a perturbation approach developed by LANL. Benchmark measurements have been successfully conducted at LANL and at RPI using their respective LSDS instruments. The ISU and UNLV collaborative effort is focused on the fabrication and testing of prototype fission chambers lined with ultra-depleted 238U and 232Th, and uranium deposition on a stainless steel disc using spiked U3O8 from room temperature ionic liquid was successful, with improving thickness obtained. In FY2012, the collaboration plans a broad array of activities. PNNL will focus on optimizing its empirical model and minimizing its reliance on calibration data, as well continuing efforts on developing an analytical model. Additional measurements are planned at LANL and RPI. LANL measurements will include a Pu sample, which is expected to provide more counts at longer slowing-down times to help identify discrepancies between experimental data and MCNPX simulations. RPI measurements will include the assay of an entire fresh fuel assembly for the study of self-shielding effects as well as the ability to detect diversion by detecting a missing fuel pin in the fuel assembly. The development of threshold neutron sensors will continue, and UNLV will calibrate existing ultra-depleted uranium deposits at ISU

Fuel Assembly Self Shielding of Interrogation Neutrons in a Lead Slowing-Down Spectrometer

Recent attention to nuclear safeguards has stepped up the need for additional non proliferation safety measures. One of these safeguards is the tracking of 239Pu and other fissile materials in spent nuclear fuel. Noninvasive methods are being investigated, including neutron interrogation. RPI has modeled a spent fuel assembly assay method in its lead slowing-down spectrometer. The fuel assembly is interrogated with neutrons from a neutron source in the center of the lead. As the interrogation neutrons slow down in the lead, they create fissions in the fuel assembly. An array of 238U detectors can then detect the fission neutrons from the 235U and 239Pu in the fuel as a function of slowing down time. The focus of this MCNP modeling is to determine the sensitivity and self shielding effects in a 16 x 16 pin fuel assembly. The results show significant shielding of interrogation neutrons from the fuel pins located further from the source up to 80%. The shielding is more significant for slower neutrons than fast. Also, secondary fissions in the assembly greatly affect the detector response and create a nonlinear response to quantities of fissile materials. The system easily identified missing fuel pins, but the response is not proportional to the quantity of fuel missing and depends on the location of the missing pins. The error in determining the quantity of 239Pu was greater than 100% when using a linear fitting model. New fitting procedures and sources of data used to benchmark measurements must be further investigated

Improvements to Nuclear Data and Its Uncertainties by Theoretical Modeling

This project addresses three important gaps in existing evaluated nuclear data libraries that represent a significant hindrance against highly advanced modeling and simulation capabilities for the Advanced Fuel Cycle Initiative (AFCI). This project will: Develop advanced theoretical tools to compute prompt fission neutrons and gamma-ray characteristics well beyond average spectra and multiplicity, and produce new evaluated files of U and Pu isotopes, along with some minor actinides; Perform state-of-the-art fission cross-section modeling and calculations using global and microscopic model input parameters, leading to truly predictive fission cross-sections capabilities. Consistent calculations for a suite of Pu isotopes will be performed; Implement innovative data assimilation tools, which will reflect the nuclear data evaluation process much more accurately, and lead to a new generation of uncertainty quantification files. New covariance matrices will be obtained for Pu isotopes and compared to existing ones. The deployment of a fleet of safe and efficient advanced reactors that minimize radiotoxic waste and are proliferation-resistant is a clear and ambitious goal of AFCI. While in the past the design, construction and operation of a reactor were supported through empirical trials, this new phase in nuclear energy production is expected to rely heavily on advanced modeling and simulation capabilities. To be truly successful, a program for advanced simulations of innovative reactors will have to develop advanced multi-physics capabilities, to be run on massively parallel super- computers, and to incorporate adequate and precise underlying physics. And all these areas have to be developed simultaneously to achieve those ambitious goals. Of particular interest are reliable fission cross-section uncertainty estimates (including important correlations) and evaluations of prompt fission neutrons and gamma-ray spectra and uncertainties