Thermal Flux Maximization from a DD Neutron Generator

Journal Article · Wed Jun 15 04:00:00 EDT 2016 · Transactions of the American Nuclear Society

OSTI ID:22991901

Metwally, Walid A.; Ballaith, Mohammad ^[1]; Chen, Allan X. ^[2]

---

1. Nuclear Engineering Department, University of Sharjah, Sharjah (United Arab Emirates)
2. Adelphi Technology, 2003 E Bayshore Rd., Redwood City, CA 94063 (United States)

The University of Sharjah recently acquired a model DD-109.4M neutron generator from Adelphi Technology. The generator is moderated and has three neutron ports which are capable of providing fast and thermal neutrons. The neutron ports are plugged with polyethylene layers of different thickness in order to control the neutron flux at different regions in the port. Neutron Activation Analysis (NAA) is an important analysis technique to determine the amounts of isotopes in a sample at different concentrations. For many isotopes the ability to precisely determine their concentrations is enhanced by utilizing thermal neutrons in the activation process; where the capture cross section is greatest. It is thus important to subject the sample to the maximum thermal flux in order to minimize the time required to activate the sample and consequently extend the lifetime of the neutron generator. A detailed MCNP model of the neutron generator has been developed for the purpose of determining the moderator thickness required to maximize the thermal neutron flux in the neutron port. The maximization of the neutron flux must therefore take into account both the forward and backscattered neutrons reaching the sample. Also provided in this work is a mapping of the thermal flux at different moderator thicknesses in order to study the variation of the flux over the area of the sample. The developed model will be used as the basis for optimizing experiments and predicting their results. A study was performed to determine the moderator thickness needed to maximize the thermal neutron flux that can be obtained from a DD neutron generator. This has a significant role in the efficient utilization of the neutron generator. The maximum thermal neutron flux was found at approximately a 3.2 cm moderator thickness. The thermal neutron flux is resulting from both the directly moderated neutrons and the backscattered neutrons from the surrounding polyethylene. The results obtained in this work will be verified experimentally by irradiating samples at different moderator thickness and performing a neutron activation analysis for these samples. (authors)

• OSTI ID: 22991901
• Journal Information: Transactions of the American Nuclear Society, Journal Name: Transactions of the American Nuclear Society Journal Issue: 1 Vol. 114; ISSN 0003-018X
• Country of Publication: United States
• Language: English

Related Subjects

07 ISOTOPE AND RADIATION SOURCES
46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND TECHNOLOGY
CROSS SECTIONS
ISOTOPES
MODERATORS
NEUTRON ACTIVATION ANALYSIS
NEUTRON FLUX
NEUTRON GENERATORS
POLYETHYLENES
THERMAL NEUTRONS
THICKNESS