MCNPX simulation studies have been performed to model and optimize the photoneutron target design and support the mini-MIPS experiments being conducted at ANL. Three target geometries and two materials (DU and Ta) were simulated with MCNPX at 35 MeV electron energy to optimize photoneutron production. The results indicated that at 20 kW beam power, the total energy integrated neutron fluxes of 8×10 11 n/cm 2 s is achievable, which makes a number of photoneutron-based applications feasible including mini-MIPS.

Component radioactivation is an important problem in accelerator facilities, impacting operations, maintenance, decommissioning, and disposal. Radionuclide inventories are calculated for an 8-cm-diam, 30.9-cm-long lead target irradiated by 660-MeV protons using the particle transport code MCNPX and the transmutation codes CINDER’90, ORIHET-3, and SP-FISPACT. The results using the various codes and data libraries are compared with experimental measurements. Comparisons are also made between the outputs of the three codes for nuclides not represented in the measurements. For more than half the nuclides studied, the codes agree with the measurements within a factor of 2, and nearly all agree within a factor of 10. The present set of codes and nuclear data files are largely adequate for calculating radioactivation in accelerator facilities, but there is room for substantial improvement for selected radionuclides.

This paper describes the results of SANS measurements of small samples using the very cold neutron (VCN) beam of the PF2 instrument at the Institut Laue Langevin (ILL), France. In addition to a classical SANS pinhole collimation, the experiment used a polarizing supermirror as a monochromator and a magnetic sextupole lens to focus the neutron beam in order to gain intensity and avoid any material in the neutron beam besides the sample.

Paul Scherrer Institute, Villigen, Switzerland The radiation effects module (REM) at Argonne’s Intense Pulsed Neutron Source has been used for a wide variety of irradiation experiments over the facility’s lifetime. Recently planned and executed experiments include studies of radiation damage in optical fibers and scintillators for use in high-energy physics detectors, irradiation of optical components for SNS neutron scattering instruments, irradiation of dilute fissile mixtures to study separations processes, and irradiation of foils and wires for neutronic characterization of the facility. Characterization of the neutron flux profile in the REM is essential for planning experiments and analyzing the resulting data. In this paper we compare the neutron fluxes obtained from recent MCNPX calculations with the results of previous experiments which used a set of activation foils to measure the neutron flux from high energies down to thermal at two locations in the REM. We also compare recent activation measurements in nickel and bismuth samples with predictions using the CINDER90 code and neutron fluxes generated using MCNPX.

The motivation for our study is to establish the prospects for a neutron source providing intense pulsed beams with spectra as cold as is realistic. The scientific motivation is to serve applications in nanoscience, biology and technology.

The relevant facts concerning the Argonne National Laboratory – Intense Pulsed Neutron Source (ANL/IPNS) and the Idaho National Laboratory (INL) apparatus for use at the ANL/IPNS facility to measure differential neutron interaction cross sections of interest for advanced reactor physics applications are presented. The INL apparatus, which consists of an array of multiple types of multiple detectors operated in coincidence, signal electronics, and a data acquisition system, is presented as an application of new means and methods to measure the relevant parameters described. The immediate measurement goals involve measurement of neutron induced interaction cross sections for 240Pu and 242Pu with 241Pu, 241Am, with measurements for other nuclides of interest for advanced reactor physics applications to follow later. Specific uncertainties and error limits are presented and methods for controlling these uncertainties are described. The post experiment analysis using data sorts and data selection from a large, self-consistent data set to produce spectra that will be analyzed for direct results and used to determine cross sections is also discussed.

The FIGARO technique uses 6–7 MeV γ-rays produced in the 19F(p, αγ)16O reaction to detect materials used in nuclear weapons or associated with their production. These γ-rays induce neutron emission via the photoneutron and photofission processes in nuclear materials. Previous experiments have shown that FIGARO gives responses specific to nuclear materials with little or no response to common benign materials. The technique is also resistant to both photon and neutron shielding countermeasures. We present preliminary results from modeling studies of neutron detection rates with simulated air cargo and intermodal shipping containers. A general methodology to compare operating performance based on receiver–operator–characteristic curves is also discussed.

Various schemes have been proposed for neutron interrogation of packages, luggage, or containers with the intent of locating concealed contraband items such as conventional explosives, drugs, or restricted special nuclear materials. Relatively intense and energetic neutron sources are usually required in these applications in order to penetrate the scanned objects and to provide unambiguous characteristic signals that are well above background, thereby minimizing both false positives and negatives. Consequently, neutron irradiation of the materials in tested objects during the interrogation process could lead to the production of significant residual radioactivity. This, in turn, might either limit or prevent the application of these methods in those situations where there is a potential for unacceptable public exposure to the induced secondary radiations. The present study aims to identify those particular neutron-induced reactions that might generate significant activation. This is accomplished by conducting a thorough survey of the current status of the pertinent cross section data available from the major general purpose and special purpose data files for neutron energies up to 15 MeV. This conference paper provides an overview of this ongoing project and discusses the status of some of the more prominent candidate reactions that have been identified to date.

We measured the photon yields for proton energies between 1.5 and 4.25MeV using both CaF2 and MgF2 solid targets and SF6 gas targets. Photon yields were measured using a 7.62cm×7.62cm NaI scintillator detector. Detector response functions for these three individual γ-rays were calculated using the Monte Carlo code MCNP-4C. The relative intensities of the three γ-rays were determined by a least-squares fit of these response functions to the data in a selected region of the pulse-height spectrum. A maximum photon yield of 6.0×107γ/μC/sr (at 0°) was determined for the sum of these three γ-rays at an effective proton energy of 4.0MeV. The contribution of the individual lines to the total photon yield depends strongly on the incident proton energy.

The potential diversion of nuclear materials is a major international concern. Fissile (e.g., U, Pu) and other nuclear materials (e.g., D, Be) can be detected using 6–7MeV gamma-rays produced in the 19F(p,αγ) 16O reaction. These gamma-rays can induce neutron emission via photoneutron and photofission processes in nuclear materials. However, they are not energetic enough to generate significant numbers of neutrons from common benign materials. Neutrons are counted using an array of BF3 tubes in a polyethylene moderator. A strong increase in neutron count rates is seen when irradiating depleted uranium, Be, D2O, and 6Li, with little or no increase for other materials (e.g., H2O, SS, Cu, Al, C, 7Li). Experiments using both photon and neutron shielding show that the technique is resistant to countermeasures. We have reduced the neutron background from proton beam reactions (thus increasing the system's sensitivity) and have tested a high-current gas cell which should be capable of operating at proton beam currents of up to 100μA.

In the few-MeV energy range below the (p,n) threshold, gamma rays are produced from proton bombardment of aluminum via the 27Al(p,p′γ)27Al, 27Al(p,α,γ)24Mg, and 27Al(p,γ)28Si reactions. Thick-target γ-ray yield data are important for various applications, including radiation shielding near accelerators and γ-ray background interference in experiments that focus on other materials. We measured p+Al γ-ray spectra over the photon energy range 0.5–15MeV and at 0° and 90° laboratory angle relative to the incident proton beam. In the experiment, mono-energetic proton beams of 1.75, 2, 2.5, 3, 3.5, and 4MeV irradiated thick stopping targets of pure aluminum. The resulting gamma rays were detected by 7.62×7.62cm2 NaI scintillation detectors. The spectra were unfolded using detector response functions calculated with the Monte Carlo code MCNP-4C. We report γ-ray spectra and tabulated γ-ray yields that can be used in applications that involve protons incident on thick aluminum targets at energies ⩽4MeV.