Calvin E. Moss

Differential neutron production cross sections and neutron yields from stopping-length targets for 113-MeV protons

We have measured differential (P,xin) cross sections, d/sup 2/sigma/d..cap omega..dE/sub n/, from thin targets and absolute neutron yields from stopping-length targets at angles of 7.5/degree/, 30/degree/, 60/degree/, and 150/degree/, for the 113--MeV proton bombardment of elemental beryllium, carbon, aluminum, iron, and depleted uranium. Additional cross-section measurements are reported for oxygen, tungsten, and lead. We used time-of-flight techniques to identify and discriminate against backgrounds and to determine the neutron energy spectrum. Comparison of the experimental data with intranuclear-cascade evaporation-model calculations with the code HETC showed discrepancies as high as a factor of 7 in the differential cross sections. These discrepancies in the differential cross sections make it possible to identify some of the good agreement seen in the stopping-length yield comparisons as fortuitous cancellation of incorrect production estimates in different energy regimes. 13 refs., 20 figs., 4 tabs.

Differential neutron production cross sections for 256-MeV protons

In this paper differential (p, xn) cross sections d{sup 2}{sigma}/d{Omega}E{sub n}, from the thin targets of beryllium, carbon, oxygen, aluminum, iron, lead, and {sup 238}U for 256-MeV protons are measured. Time-of-flight techniques are used to identify and discriminate against backgrounds and to determine the neutron energy spectrum. Comparison of the experimental data with intranuclear-cascade evaporation-model calculations using the HETC code showed discrepancies of as much as a factor of 7, notably at 7.5 and 150 deg.

CdZnTe gamma ray spectrometer for orbital planetary missions

We present the design and analysis of a new gamma ray spectrometer for planetary science that uses an array of CdZnTe detectors to achieve the detection efficiency needed for orbital measurements. The use of CdZnTe will provide significantly improved pulse height resolution relative to scintillation-based detectors, with commensurate improvement in the accuracy of elemental abundances determined by gamma ray and neutron spectroscopy. The spectrometer can be flown either on the instrument deck of the spacecraft or on a boom. For deck-mounted systems, a BGO anticoincidence shield is included in the design to suppress the response of the CdZnTe detector to gamma rays that originate in the spacecraft. The BGO shield also serves as a backup spectrometer, providing heritage from earlier planetary science missions and reducing the risk associated with the implementation of new technology.

Development of high efficiency, multi-element CdZnTe detectors for portable measurement applications

We describe the development of detector arrays and electronics for large-volume, hand-held CdZnTe detectors with the same counting efficiency as portable NaI(Tl) detectors presently used for nuclear material measurement applications. The pulse-height resolution of the multi-element detectors is at least three times better than NaI(Tl) over a wide energy range (from 100 keV to several MeV), enabling more accurate measurements of gamma-rays emitted by special nuclear material. Arrays of up to eight coplanar grid detectors can be combined to make detectors ranging in size from 4 to 14 cm3. Because the number of spectroscopy channels is small, low-power, hand-held detectors can be manufactured with conventional printed circuit board technology, thus keeping the cost of multi-element detectors to a minimum. The design and performance of an 8-element detector is presented.

Gamma-ray line intensities for depleted uranium

Measurements of the gamma-ray line intensities from depleted uranium allowed us to determine which of two conflicting previous experiments was correct. For the 1001-keV line we obtain a branching ratio of 0.834 +- 0.007, in good agreement with one of the previous experiments. A table compares our intensities for several lines with those obtained in previous experiments. 5 refs., 2 figs., 1 tab.

Neutron Yields from Stopping-Length Targets for 256-MeV Protons

In this paper absolute neutron yields from stopping-length targets at angles of 7.5, 30, 60, and 150 deg for the 256-MeV proton bombardment of elemental beryllium, carbon, aluminum, and iron are measured. Time-of-flight techniques are used to identify and discriminate against backgrounds and to determine the neutron energy spectrum. Comparison of the experimental data with intranuclear-cascade evaporation-model calculations using the HETC code showed good agreement, indicating that transport probably dominates production effects in the calculations.

Efficiency of a Bismuth-Germanate Scintillator: Comparison of Monte Carlo Calculations with Measurements

Monte Carlo calculations of a bismuth-germanate scintillators efficiency agree closely with experimental measurements. For this comparison, we studied the absolute gamma-ray photopeak efficiency of a scintillator (7.62 cm long by 7.62 cm in diameter) at several gamma-ray energies from 166 to 2615 keV at distances from 30.5 to 152.4 cm. Computer calculations were done in a three-dimensional cylindrical geometry with the Monte Carlo coupled photon-electron code CYLTRAN. For the experiment we measured 11 sources with simple spectra and precisely known strengths. The average deviation between the calculations and the measurements is 3%. Our calculated results also closely agree with recently published calculated results.

Photon and Neutron Active Interrogation of Highly Enriched Uranium

The physics of photon and neutron active interrogation of highly enriched uranium (HEU) using the delayed neutron reinterrogation method is described in this paper. Two sets of active interrogation experiments were performed using a set of subcritical configurations of concentric HEU metal hemishells. One set of measurements utilized a pulsed 14‐MeV neutron generator as the active source. The second set of measurements utilized a linear accelerator‐based bremsstrahlung photon source as an active interrogation source. The neutron responses were measured for both sets of experiments. The operational details and results for both measurement sets are described.

Warhead counting using neutron scintillators: detector development, testing, and demonstration

The authors describe the development of a fast-neutron detector based on a boron-loaded plastic scintillator used previously for space applications. The detector rejects gammas and scattered low-energy neutrons, and its segmentation allows narrow fan-shaped collimation within +or-20 degrees horizontally and +or-50 degrees vertically. Testing includes distinguishing between mockups with either two or three warheads and locating the ten warheads on a silo-based Peacekeeper missile. For warhead-counting applications, the results presented indicate that collimated proton-recoil scintillations can, at least in some circumstances, provide the position resolution needed to resolve individual warheads without revealing intrusive information about their design.<<ETX>>

Unfolding bismuth-germanate pulse-height distributions to determine gamma-ray flux spectra and dose rates

Abstract We report a new procedure for unfolding gamma-ray pulse-height distributions acquired with bismuth-germanate detectors. The equipment used for acquiring the distributions consists of a LeCroy 3500 data acquisition and analysis system and eight bismuth-germanate scintillation detectors 7.62 cm in diameter and 7.62 cm long. The system was calibrated and characterized from 0.12 to 8.28 MeV by using gamma-ray spectra from a variety of radioactive sources and from the 14 N(p,γ) 15 O reaction. By fitting these pulse-height distributions with a function containing 17 parameters, we determined theoretical response functions and used them to obtain the gamma-ray flux spectra at multiple space points from a variety of radioactive objects of interest to nuclear safeguards. We used two flux-spectrum-to-dose-rate conversion curves to obtain dose rates. For a composite source, consisting of several sources with accurately known strengths, the result of our procedure agreed with the expected value to within less than 10%. Direct use of measured spectra and the flux-spectrum-to-dose-rate curves to obtain dose rates avoids the errors that can arise because of spectrum dependence in simple gamma-ray dosimeter instruments.