Woo Y. Yoon

Pulsed photon interrogation with neutron-induced gamma-ray spectrometry for cargo inspections

An accelerator-based, active, pulsed, interrogation system capable of non-destructive, elemental analysis from secondary gamma-ray emissions is being developed for various inspection applications. The system consists of a very narrow pulsed, electron accelerator (> 8 MeV) for photoneutron production and a multiple channel, gamma-ray detection system capable of ultra-fast detection. This system has applicability to cargo container inspections. Advantages of the system include use of highly penetrative, energetic, interrogating X-rays, compatibility with existing X-ray inspection systems, capability of large neutron production yield from the pulsed accelerator, ability to induce a volumetric, tailored neutron source spectrum within or very near the containers, and most importantly, the spectrometry capability of a ultra-fast, gamma-ray detection system. The portable detection system is described which has been designed to acquire multiple, single gamma-ray events (up to 40 MHz between each accelerator pulse) from neutron inelastic scattering (starting from within 100 ns after an accelerator pulse) and neutron capture interactions. The system and its overall operational requirements are described. Experimental results, using a 50 ps electron accelerator pulse and selected inspected objects, are presented along with numerical predictions and system characterizations.

COMBINE7.0 - A Portable ENDF/B-VII.0 Based Neutron Spectrum and Cross-Section Generation Program

COMBINE7.0 is a FORTRAN 90 computer code that generates multigroup neutron constants for use in the deterministic diffusion and transport theory neutronics analysis. The cross-section database used by COMBINE7.0 is derived from the Evaluated Nuclear Data Files (ENDF/B-VII.0). The neutron energy range covered is from 20 MeV to 1.0E-5 eV. The Los Alamos National Laboratory NJOY code is used as the processing code to generate a 167 finegroup cross-section library in MATXS format for Bondarenko self-shielding treatment. Resolved resonance parameters are extracted from ENDF/B-VII.0 File 2 for a separate library to be used in an alternate Nordheim self-shielding treatment in the resolved resonance energy range. The equations solved for energy dependent neutron spectrum in the 167 fine-group structure are the B-3 or B-1 approximations to the transport equation. The fine group cross sections needed for the spectrum calculation are first prepared by Bondarenko selfshielding interpolation in terms of background cross section and temperature. The geometric lump effect, when present, is accounted for by augmenting the background cross section. Nordheim self-shielded fine group cross sections for a material having resolved resonance parameters overwrite correspondingly the existing self-shielded fine group cross sections when this option is used. The fine group cross sections in the thermal energy range are replaced by those selfshielded with the Amouyal/Benoist/Horowitz method in the three region geometry when this option is requested. COMBINE7.0 coalesces fine group cross sections into broad group macroscopic and microscopic constants. The coalescing is performed by utilizing fine-group fluxes and/or currents obtained by spectrum calculation as the weighting functions. The multigroup constant may be output in any of several standard formats including ANISN 14** free format, CCCC ISOTXS format, and AMPX working library format. ANISN-PC, a onedimensional, discrete-ordinate transport code, is incoprated into COMBINE7.0. As an option, the 167 fine-group constants generated by COMBINE portion in the program can be used to cacluate regionwise spectra in the ANISN portion, all internally to reflect the one-dimensional transport correction. Results for the criticality validation calculations are included as a part of verification and validation.

An Accelerator-Based Epithermal Photoneutron Source for BNCT

Therapeutically-useful epithermal-neutron beams for Boron Neutron Capture Therapy (BNCT) are currently generated by nuclear reactors. Various accelerator-based neutron sources1–3 for BNCT have been proposed and some low-intensity prototypes of such sources, generally featuring the use of proton beams and beryllium or lithium targets have been constructed. Scaling of most of these proton devices for therapeutic applications will require the resolution of some rather difficult issues associated with target cooling. This paper describes an alternate approach to the realization of a clinically-useful accelerator-based source of epithermal neutrons for BNCT that reconciles the often-conflicting objectives of target cooling, neutron beam intensity, and neutron beam spectral purity via a two-stage photoneutron production process.

Bremsstrahlung versus Monoenergetic Photon Dose and Photonuclear Stimulation Comparisons at Long Standoff Distances

Energetic photon sources with energies greater than 6 MeV continue to be recognized as viable source for various types of inspection applications, especially those related to nuclear and/or explosive material detection. These energetic photons can be produced as a continuum of energies (i.e., bremsstrahlung) or as a set of one or more discrete photon energies (i.e., monoenergetic). This paper will provide a follow‐on extension of the photon dose comparison presented at the 9th International Conference on Applications of Nuclear Techniques (June 2008). Our previous paper showed the comparative advantages and disadvantages of the photon doses provided by these two energetic interrogation sources and highlighted the higher energy advantage of the bremsstrahlung source, especially at long standoff distances (i.e., distance from source to the inspected object). This paper will pursue higher energy photon inspection advantage (up to 100 MeV) by providing dose and stimulated photonuclear interaction predictions ...

Performance evaluation of the passive-active neutron (PAN) radioassay system at Idaho National Engineering Laboratory

The Idaho National Engineering Laboratory is using the passive active neutron (PAN) radioassay system to assay 208 liter drums for plutonium in transuranic waste. PAN systems similar to INELs are in use to perform like tasks at other US DOE sites and later versions of the PAN are in use throughout the world. The PAN system, developed by Los Alamos National Laboratory, uses both passive neutron coincidence counting and thermal neutron interrogation/differential die-away to determine the amount of plutonium in a waste drum. This paper discusses a modified statistical sampling and verification approach to determine the total uncertainty of a PAN systems measurement when applied in the assay of real waste. In this approach a statistical review is performed on real waste drum radiographs, waste generator data and PAN assay data to identify the variations of the physical, chemical and measurement characteristics pertaining to a waste category. Using data from the statistical review, the performance of the PAN assay systems is simulated using computer models of the assay system with regard to the waste form under review. The simulated responses, in terms of plutonium mass in these cases, are compared with the known input masses to assess total uncertainty. To demonstrate the method, the result of this approach will be presented for PAN measurements on two waste categories at INEL.