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Title: High-Fidelity Lattice Physics Capabilities of the SCALE Code System Using TRITON

Abstract

Increasing complexity in reactor designs suggests a need to reexamine of methods applied in spent-fuel characterization. The ability to accurately predict the nuclide composition of depleted reactor fuel is important in a wide variety of applications. These applications include, but are not limited to, the design, licensing, and operation of commercial/research reactors and spent-fuel transport/storage systems. New complex design projects such as space reactors and Generation IV power reactors also require calculational methods that provide accurate prediction of the isotopic inventory. New high-fidelity physics methods will be required to better understand the physics associated with both evolutionary and revolutionary reactor concepts as they depart from traditional and well-understood light-water reactor designs. The TRITON sequence of the SCALE code system provides a powerful, robust, and rigorous approach for reactor physics analysis. This paper provides a detailed description of TRITON in terms of its key components used in reactor calculations.

Authors:
 [1]
  1. ORNL
Publication Date:
Research Org.:
Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)
Sponsoring Org.:
Work for Others (WFO)
OSTI Identifier:
968024
DOE Contract Number:
DE-AC05-00OR22725
Resource Type:
Conference
Resource Relation:
Conference: ANS Winter 2007 Meeting, Washington, DC, USA, 20071111, 20071115
Country of Publication:
United States
Language:
English
Subject:
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS; 21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS; DESIGN; FORECASTING; ISOTOPES; LICENSING; NUCLEAR FUELS; PHYSICS; POWER REACTORS; REACTOR PHYSICS

Citation Formats

DeHart, Mark D. High-Fidelity Lattice Physics Capabilities of the SCALE Code System Using TRITON. United States: N. p., 2007. Web.
DeHart, Mark D. High-Fidelity Lattice Physics Capabilities of the SCALE Code System Using TRITON. United States.
DeHart, Mark D. Mon . "High-Fidelity Lattice Physics Capabilities of the SCALE Code System Using TRITON". United States. doi:.
@article{osti_968024,
title = {High-Fidelity Lattice Physics Capabilities of the SCALE Code System Using TRITON},
author = {DeHart, Mark D},
abstractNote = {Increasing complexity in reactor designs suggests a need to reexamine of methods applied in spent-fuel characterization. The ability to accurately predict the nuclide composition of depleted reactor fuel is important in a wide variety of applications. These applications include, but are not limited to, the design, licensing, and operation of commercial/research reactors and spent-fuel transport/storage systems. New complex design projects such as space reactors and Generation IV power reactors also require calculational methods that provide accurate prediction of the isotopic inventory. New high-fidelity physics methods will be required to better understand the physics associated with both evolutionary and revolutionary reactor concepts as they depart from traditional and well-understood light-water reactor designs. The TRITON sequence of the SCALE code system provides a powerful, robust, and rigorous approach for reactor physics analysis. This paper provides a detailed description of TRITON in terms of its key components used in reactor calculations.},
doi = {},
journal = {},
number = ,
volume = ,
place = {United States},
year = {Mon Jan 01 00:00:00 EST 2007},
month = {Mon Jan 01 00:00:00 EST 2007}
}

Conference:
Other availability
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  • This paper describes ongoing calculations used to validate the TRITON depletion module in SCALE for light water reactor (LWR) fuel lattices. TRITON has been developed to provide improved resolution for lattice physics mixed-oxide fuel assemblies as programs to burn such fuel in the United States begin to come online. Results are provided for coupled TRITON/PARCS analyses of an LWR core in which TRITON was employed for generation of appropriately weighted few-group nodal cross-sectional sets for use in core-level calculations using PARCS. Additional results are provided for code-to-code comparisons for TRITON and a suite of other depletion packages in the modelingmore » of a conceptual next-generation boiling water reactor fuel assembly design. Results indicate that the set of SCALE functional modules used within TRITON provide an accurate means for lattice physics calculations. Because the transport solution within TRITON provides a generalized-geometry capability, this capability is extensible to a wide variety of non-traditional and advanced fuel assembly designs. (authors)« less
  • Reactor physics numerical benchmarks have been performed at the Brookhaven National Laboratory (BNL) with the software package SCALE 5.0 and its TRITON module to assess their capability to predict neutronics parameters for mixed oxide (MOX) fuels. The results of such calculations are herein presented. Specifically, BNL results for neutron multiplication factors (kINF), neutron fluxes and fuel burnup have been added to published OECD/NEA benchmarks for MOX fuels and particular emphasis has been given to the impact of cross-section libraries and their energy structure on the results. Among the OECD/NEA published benchmarks two have been considered here: the first one modelsmore » a fuel pin surrounded by moderator, in which two different MOX fuels can be introduced, and for each one of them kINF and neutron fluxes as a function of burnup are calculated. The second one includes both a fuel pin case and a macro-cell case (a heterogeneous 30 by 30 configuration of fuel pins), for which the void coefficient is determined by calculating kINF at zero burnup as a function of moderation. The calculations are repeated for several combinations of MOX and uranium oxide fuels using several different cross-section libraries. The final results have been compared with each other. This study shows that SCALE 5.0 (with TRITON) overall performs in line with the other codes in the benchmark, but the results are dependent on the energy group structure of the cross section libraries used. For instance, when fissile plutonium is increased in the fuel, TRITON results become slightly divergent with burnup (with respect to the other codes in the benchmark) and if the standard 44-group library provided with SCALE 5.0 is used void coefficient calculations become inadequate for very low void (below 10% of the operating value of moderator density). Moreover, the prediction capabilities of the code are shown to be dependent on the MOX fuel enrichment and the MOX isotopic composition. (authors)« less
  • Traditional nuclear reactor system analysis codes such as RELAP and TRAC employ an operator split methodology. In this approach, each of the physics (fluid flow, heat conduction and neutron diffusion) is solved separately and the coupling terms are done explicitly. This approach limits accuracy (first order in time at best) and makes the codes slow in running since the explicit coupling imposes stability restrictions on the time step size. These codes have been extensively tested and validated for the existing LWRs. However, for GEN IV nuclear reactor designs which tend to have long lasting transients resulting from passive safety systems,more » the performance is questionable and modern high fidelity simulation tools will be required. The requirement for accurate predictability is the motivation for a large scale overhaul of all of the models and assumptions in transient nuclear reactor safety simulation software. At INL we have launched an effort with the long term goal of developing a high fidelity system analysis code that employs modern physical models, numerical methods, and computer science for transient safety analysis of GEN IV nuclear reactors. Modern parallel solution algorithms will be employed through utilizing the nonlinear solution software package PETSc developed by Argonne National Laboratory. The physical models to be developed will have physically realistic length scales and time scales. The solution algorithm will be based on the physics-based preconditioned Jacobian-free Newton-Krylov solution methods. In this approach all of the physical models are solved implicitly and simultaneously in a single nonlinear system. This includes the coolant flow, nonlinear heat conduction, neutron kinetics, and thermal radiation, etc. Including modern physical models and accurate space and time discretizations will allow the simulation capability to be second order accurate in space and in time. This paper presents the current status of the development efforts as well as some results from analyzing a simplified primary system model of GNEP’s advanced burner test reactor (ABTR) designed by Argonne. Various transient analyses are performed with this simplified ABTR model to study two fundamental issues related to system analysis codes – accuracy of numeric algorithm and efficiency. The accuracy study is carried by comparing the second order method with the first order method. The results show that numerical errors in the first order method are large and it is very difficult to distinguish numerical errors from physical modeling errors. On the other hand, second order method yields small numerical errors and it is very easy to spot physical modeling errors. The efficiency study is carried out by comparing the time steps for the fully implicit solution algorithm versus CFL stability limit methods. The dynamic time steps used in a fully implicit method will adjust the time step to resolve the time scale during the various stages of a long lasting transient. This will make a computer code based on fully implicit methods run more efficiently versus a CFL stability limit method code like RELAP, in which a particle of fluid cannot cross a control volume in a single time step.« less