Ned Xoubi

Assumptions and Criteria for Performing a Feasability Study of the Conversion of the High Flux Isotope Reactor Core to Use Low-Enriched Uranium Fuel

A computational study will be initiated during fiscal year 2006 to examine the feasibility of converting the High Flux Isotope Reactor from highly enriched uranium fuel to low-enriched uranium. The study will be limited to steady-state, nominal operation, reactor physics and thermal-hydraulic analyses of a uranium-molybdenum alloy that would be substituted for the current fuel powder--U{sub 3}O{sub 8} mixed with aluminum. The purposes of this document are to (1) define the scope of studies to be conducted, (2) define the methodologies to be used to conduct the studies, (3) define the assumptions that serve as input to the methodologies, (4) provide an efficient means for communication with the Department of Energy and American research reactor operators, and (5) expedite review and commentary by those parties.

Design, Development and Installation of Jordan Subcritical Assembly

Following its announcement in 2007 to pursue a nuclear power program and in the absence of any nuclear facility essential for the education, training, and research, Jordan decided to build a subcritical reactor as its first nuclear facility. Jordan Subcritical Assembly (JSA) is uranium fueled light water moderated and reflected subcritical reactor driven by a plutonium-beryllium source, and the core consists of 313 LEU fuel rods, loaded into a water-filled vessel in a square lattice of 19.11u2009mm pitch. The fuel rods are based on PWR fuel structural pattern type, made of uranium oxide (UO2) with 3.4u2009wt% 235U enrichment in zirconium alloy (Zr-4) cladding. Design, optimization, and verification were performed using MCNP5 nuclear code; the computed effective multiplication factor is 0.95923. The JSA is designed to fulfill the training needs of students and is equipped to perform all of the fundamental experiments required for a typical nuclear engineering university program. This paper presents the design, development, modeling, core analysis, and utilization of Jordan’s first nuclear facility and why this simplified low cost reactor presents an attractive choice to fulfill the preliminary experimental needs of nuclear engineering education in developing countries.

MCNP prediction of measured subcritical reactor parameters

At the University of Cincinnati Nuclear Engineering Program, the resurrection of a once moth-balled subcritical reactor facility now provides an ideal laboratory experience to its students. Its simplicity and accessibility are among its key features. This article focuses upon supplementing the experimental aspects of this facility via the development of a three-dimensional, full detail, MCNP model of this reactor, while emphasizing the validation of this computational tool against laboratory measurements. The subcritical reactor parameters herein compared include: neutron and photon flux distributions, fission rate, and the subcritical multiplication factor (k-effective), among other relevant parameters of interest

Neutronic design study of accelerator driven system (ADS) for Jordan subcritical reactor as a neutron source for nuclear research

In this paper, a preliminary neutronic design study of an accelerator driven subcritical system for Jordan Subcritical Assembly (JSA) is presented. The conceptual design of coupling the JSA core with proton accelerator and spallation target is investigated, and its feasibility as a neutron source for nuclear research, and possibly for target irradiation and isotope production evaluated. 3D MCNPX model of the JSA reactor, the accelerator beam, and the Pb target was developed, based on actual reactor parameters. MCNPX calculations were carried out to estimate the absolute radial and axial neutron flux in the reactor, and to calculate the multiplication factor Keff and heat generated in the reactor. Numerical results showed an enormous increase in the neutron flux, by seven orders of magnitude, compared to the current JSA core design using Pu-Be source. In this research the results obtained are discussed and compared with those of the JSA, and do confirm the feasibility of utilizing the JSA as a viable nuclear research facility with adequate neutron flux.

Neutronic Analysis of an Advanced Fuel Design Concept for the High Flux Isotope Reactor

Abstract This study presents the neutronic analysis of an advanced fuel design concept for the Oak Ridge National Laboratory (ORNL) High Flux Isotope Reactor (HFIR) that could significantly extend the current fuel cycle length under the existing design and safety criteria. A key advantage of the fuel design herein proposed is that it would not require structural changes to the present HFIR core, in other words, maintaining the same rated power and fuel geometry (i.e., fuel plate thickness and coolant channel dimensions). Of particular practical importance, as well, is the fact that the proposed change could be justified within the bounds of the existing nuclear safety basis. The simulations herein reported employed transport theory–based and exposure-dependent eigenvalue characterization to help improve the prediction of key fuel cycle parameters. These parameters were estimated by coupling a benchmarked three-dimensional MCNP5 model of the HFIR core to the depletion code ORIGEN via the MONTEBURNS interface. The design of an advanced HFIR core with an improved fuel loading is an idea that evolved from early studies by R. D. Cheverton, formerly of ORNL. This study contrasts a modified and increased core loading of 12 kg of 235U against the current core loading of 9.4 kg. The simulations performed predict a cycle length of 39 days for the proposed fuel design, which represents a 50% increase in the cycle length in response to a 25% increase in fissile loading, with an average fuel burnup increase of ~23%. The results suggest that the excess reactivity can be controlled with the present design and arrangement of control elements throughout the core’s life. Also, the new power distribution is comparable or even improved relative to the current power distribution, displaying lower peak to average fission rate densities across the inner fuel element’s centerline and bottom cells. In fact, the fission rate density in the outer fuel element also decreased at these key locations for the proposed design. Overall, it is estimated that the advanced core design could increase the availability of the HFIR facility by ~50% and generate ~33% more neutrons annually, which is expected to yield sizeable savings during the remaining life of HFIR, currently expected to operate through 2014. This study emphasizes the neutronics evaluation of a new fuel design. Although a number of other performance parameters of the proposed design check favorably against the current design, and most of the core design features remain identical to the reference, it is acknowledged that additional evaluations would be required to fully justify the thermal-hydraulic and thermal-mechanical performance of a new fuel design, including checks for cladding corrosion performance as well as for industrial and economic feasibility.