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- by and +1
- •
- Radiation, Virtual Environments, Virtual Learning Environments, Virtual Laboratory

This paper summarizes the results of modeling methodologies developed for the zero-power (100 W) teaching and research reactor CROCUS located in the Laboratory for Reactor Physics and Systems Behavior (LRS) at the Swiss Federal Institute of Technology in Lausanne (EPFL). The study gives evidence that the Monte Carlo code Serpent can be used effectively as a lattice physics tool for small reactors. CROCUS’ core has an irregular geometry with two fuel zones of different lattice pitches. This and the reactor’s small size necessitate the use of nonstandard cross-section homogenization techniques when modeling the full core with a 3D nodal diffusion code (e.g. PARCS). The primary goal of this work is the development of these techniques for steady-state neutronics and future transient neutronics analyses of not only CROCUS, but research reactors in general. In addition, the modeling methods can provide useful insight for analyzing small modular reactor concepts based on light water technology. Static computational models of CROCUS with the codes Serpent and MCNP5 are presented and methodologies are analyzed for using Serpent and SerpentXS to prepare macroscopic homogenized group cross-sections for a pin-by-pin model of CROCUS with PARCS. The most accurate homogenization scheme lead to a difference in terms of keff of 385 pcm between the Serpent and PARCS model, while the MCNP5 and Serpent models differed in terms of keff by 13 pcm (within the statistical error of each simulation). Comparisons of the axial power profiles between the Serpent model as a reference and a set of PARCS models using different homogenization techniques showed a consistent root-mean-square deviation of ∼8%, indicating that the differences are not due to the homogenization technique but rather arise from the definition of the diffusion coefficients produced by Serpent. A comparison of the radial power profiles between the best PARCS model and full-core Serpent model showed largest relative differences in terms of power prediction at the core periphery, which is believed to be the product of the geometry simplifications made, the diffusion coefficients produced by Serpent, and the two-group energy structure used. The worth of a single control rod reproduced in PARCS showed a difference of ~33 pcm from its 169 pcm worth simulated in Serpent.

- by Daniel Siefman and +1
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- Reactor physics (Physics), Nuclear Reactors, Reactors, Research Reactor

In order to analyze the steady state and transient behavior of the CROCUS reactor, several methods and models need to be developed in the areas of reactor physics, thermal-hydraulics, and multiphysics coupling. The long-term objectives of this project are to work towards the development of a modern method for the safety analysis of research reactors and to update the Final Safety Analysis Report of the CROCUS reactor. A first part of the paper deals with generation of a core simulator nuclear data library for the CROCUS reactor using the Serpent 2 Monte Carlo code and also with reactor core modeling using the PARCS code. PARCS eigenvalue, radial power distribution, and control rod reactivity worth results were benchmarked against Serpent 2 full-core model results. Using the Serpent 2 model as reference, PARCS eigenvalue predictions were within 240 pcm, radial power was within 3% in the central region of the core, and control rod reactivity worth was within 2%. A second part reviews the current methodology used for the safety analysis of the CROCUS reactor and presents the envisioned approach for the multiphysics modeling of the reactor.

- by Daniel Siefman and +1
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- Reactor physics (Physics), Nuclear Reactors, Reactor Design, Reactor Physics and Engineering
- by Sari Alkhatib and +1
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- Nuclear Engineering, Nuclear, Nuclear Fuel for Research Reactors, Research Reactor

A one dimensional steady state thermal analytical model has
been developed to study the heat transfer and temperature distributions
in a quartz ampoule filled with MoO3 powder. The
source heat generation inside the ampoule is released from the
high neutron flux (1.4 · 1014 neutrons/cm2s) interaction with
MoO3. Natural and forced convections heat transfer boundary
conditions are adopted during the irradiation process. The
peak temperatures in MoO3 powder and quartz are calculated
and compared with their melting temperatures to ensure the irradiation
safety criteria.

- by Amr Abdelhady
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- Neutron Flux, Thermal Modeling, Neutron Irradiation Effect, Research Reactor
- by ZEESHAN AKRAM
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- Engineering, Risk Analysis, Reliability, PSA
- by Asif Anwar
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- Temperature, Correlation coefficient, Research Reactor

... Young G. Jo, William J. Spiesman, and Naeem M. Abdurrahman ... The simplified model geometry is shown in Figure 2. Since it took very long time to run the detailed TRIGA core model to get results with acceptable accuracy, such a model... more

- by Naeem Abdurrahman
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- Nuclear Engineering, Materials Science, Computed Tomography, Monte Carlo

This paper presents the current status and feasibility of extended use of WWR-M research reactor; the extention of the reactor life depends strongly on the vessel condition. Generic information regarding works performed during reactor upgrade is also presented. In the present paper, the concept of the reactor renovation is proposed; and the key element is the replacement of the reactor vessel. Estimation of the technical implementation for such replacement is given. The concept schedule sequence is determined with forthcoming result of the reactor life-time extension for another 40-50 years.

- by Co. SEP
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- Research Reactor