Theory of Westinghouse advanced nodal code for MOX applications

Journal Article · Tue Dec 31 04:00:00 EST 1996 · Transactions of the American Nuclear Society

OSTI ID:426404

Shatilla, Y A; Chao, Y A ^[1]; Tahara, Y ^[2]

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1. Westinghouse Electric Corp., Pittsburgh, PA (United States)
2. Mitsubishi Heavy Industries, Ltd., Yokohama (Japan)

The Westinghouse Advanced Nodal Code (ANC) has been modified to incorporate new theoretical improvements in its nodal solution method. These theoretical improvements include the employment of a hybrid of the Polynomial Nodal Expansion Method and the Analytic Nodal Expansion Method in solving the nodal diffusion equation and the use of the burnup gradient treatment to accurately capture uneven assembly burnup. In today`s environment, nuclear electric utilities, especially in Europe, are used to loading their cores with mixed oxides (MOX) of enrichments that give the same lifetime average reactivity of regular UO{sub 2} fuel assembly. This did not pose any difficulties for core design methodologies that are used to analyze those cores. However, the usage of weapons-grade plutonium, or even recycled plutonium with higher plutonium, in core design will challenge existing design methods. Also, there has been a strong inclination among utilities to run their reactors longer and burn their fuel higher to improve their fuel cycle economy. These two challenges and many others have put a lot of pressure on core analysis methods in their attempt to maintain the same superior performance that they enjoyed in the past. This paper outlines Westinghouse efforts in the area of nodal method development. In particular, ANC has several theoretical enhancements that enable accurate predictions of design parameters for cores that have mixed loadings of regular UO{sub 2} and MOX and/or high-burnup fuel. These enhancements include a consistent corner-point flux calculation for pin-power reconstruction and spectral interaction correction of assembly cross sections for accurate environment effects evaluation. The basic theoretical foundation of these enhancements is outlined in this paper

• OSTI ID: 426404
• Report Number(s): CONF-961103--
• Journal Information: Transactions of the American Nuclear Society, Journal Name: Transactions of the American Nuclear Society Vol. 75; ISSN TANSAO; ISSN 0003-018X
• Country of Publication: United States
• Language: English

Related Subjects

22 GENERAL STUDIES OF NUCLEAR REACTORS
A CODES
COMPUTERIZED SIMULATION
MIXED OXIDE FUELS
NODAL EXPANSION METHOD
REACTOR CORES