Berkeley Nuclear Research Center

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Recommended Events

The Spent Nuclear Fuel and Nonproliferation

Monday, April 4, 2011, 4-5pm, 3105 Etcheverry Hall

Dr. Bekhzod YuldashevYuldashev_photo

CISAC, Stanford University, Stanford

 

The review of the activities on repatriation of highly enriched spent fuel from research reactors as well as data on the conversion of research reactors to low enriched fuel will be presented.

 

 

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Delayed gamma assay for nuclear safeguards

MONDAY, MARCH 7, 4-5pm 3105 Etcheverry Hall

Vladimir MozinV_Mozin

PhD candidate, graduate research assistant

The UC Berkeley Nuclear Engineering Department in collaboration with Los Alamos National Laboratory and Lawrence Berkeley National Laboratory participates in the US DOE Next Generation Safeguards Initiative with a focus on developing advanced instruments and methods in nuclear material control and accountancy. Within this context, a delayed gamma non-destructive assay (NDA) technique is being investigated as a means to directly quantify both the fissile and fertile content of spent nuclear fuel, and as a general safeguards tool that can be easily integrated with other active interrogation instruments. In support of this research, a newly developed modeling technique was introduced, offering a versatile capability for time- and spatially-dependent, prompt and delayed discrete gamma source term and detector response calculations. The new modeling approach was validated in a series of experiments involving accelerator-driven neutron sources and samples of fissile and fertile materials and their combinations with varying parameters for interrogation setups.

Reduced-Order Physics Models For Computationally Efficient Charged Particle Transport

Monday February 14, 2011, 4-5pm, 3105 Etcheverry Hall

PrinjaDr. Anil Prinja, University of New Mexico

High-energy charged particles (electrons, light and heavy ions, from 10's keV to GeV and above) are ubiquitous in nuclear science and engineering, medical physics, space science, and advanced materials applications, but the computational complexity associated with the transport of charged particles can greatly exceed that of neutrons, photons and other neutral particles. The primary reason is that long range, Coulomb-field mediated elastic and inelastic collisions of energetic charged particles with target nuclei and electrons are characterized by extremely small collision mean free paths and near-singular differential cross sections. This extreme physics renders the computational modeling of the analog or true problem prohibitively expensive in both stochastic (Monte Carlo) and deterministic numerical settings. The condensed history (CH) Monte Carlo method, widely employed in electromagnetic and hadronic shower codes, attempts to circumvent this practical difficulty by advancing the particle in fixed large steps but inherent flaws limit the ultimate accuracy possible with this method.

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An Innovative and Advanced Neutron Transport Method for Whole Reactor Core Criticality Analysis

February 28, 2011, from 4-5pm  3105 Etcheverry Hall

Dr. Farzad Rahnemarahnema

Georgia Institute of Technology

 

A new coarse-mesh radiation transport (COMET) method for modeling and simulation of realistic reactor cores (e.g., operating water reactors) is presented at this colloquium. This innovative method has Monte Carlo accuracy while having computational efficiency that is several orders of magnitude better than achievable by stochastic and fine-mesh deterministic transport methods. Benchmark results in several whole-core problems typical of operating reactors are presented to demonstrate the accuracy and efficient of the method.

The new method overcomes many of the limitations inherent in current whole-core (loosely coupled transport/diffusion theory) methods used in the industry. Notable limitations/approximations are single lattice transport theory calculations with approximate boundary conditions (e.g., full specular reflection), cross section homogenization, ad hoc de-homogenization (fuel pin reconstruction) and whole-core homogenized diffusion theory calculations. These approximations breakdown with increasing assembly and core heterogeneities, features encountered in advanced and next generation reactor designs.

We first present an overview of current industry methods, research directions and critical gaps in the context of advanced and Generation IV nuclear reactors. The limitations of current methods and reactor design trends are highlighted as motivation for the developments of the advanced radiation transport methods by the Computational Reactor and Medical Physics Group (CRMPG) at Georgia Tech.

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