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Home / Publications / Journals / Nuclear Technology / Volume 187 / Number 1 / Pages 103-116

Impact of Local Burnup on Prediction of Power Density in the NIST Research Reactor

Nicholas R. Brown, Albert L. Hanson, and David J. Diamond

Nuclear Technology / Volume 187 / Number 1 / July 2014 / Pages 103-116

Technical Note / Fuel Cycle and Management / dx.doi.org/10.13182/NT13-20

This study addresses the overprediction of local power when the burnup distribution in each half-element of the National Institute of Standards and Technology (NIST) research reactor, known as the NBSR, is assumed to be uniform—a constraint in the full-core model used for neutronic analysis. A single-element model was utilized to quantify the impact of axial and platewise burnup on the power distribution within the NBSR fuel elements for both high-enriched uranium (HEU) and (proposed) low-enriched uranium (LEU) fuel. To validate this approach, key parameters in the single-element model were compared to parameters from an equilibrium core model, specifically, neutron energy spectrum, power distribution, and integral 235U vector. The power distribution changes significantly when incorporating local burnup effects and has lower power peaking relative to the uniform burnup case. In the uniform burnup case, the axial relative power peaking is overpredicted by as much as 59% in the HEU single element and 46% in the LEU single element. In the uniform burnup case, the platewise power peaking is overpredicted by as much as 23% in the HEU single element and 18% in the LEU single element. The degree of overprediction increases as a function of burnup cycle, with the greatest overprediction at the end of fuel element life. However, the overprediction in local power is always conservative in terms of the minimum critical heat flux ratio, a key safety parameter that depends on the local heat flux condition. The thermal flux peak is always in the midplane gap; this causes the local cumulative burnup near the midplane gap to be significantly higher than the fuel element average. Uniform burnup distribution throughout a half-element also causes a bias in fuel element reactivity worth particularly near end of life, primarily due to the importance of the fissile inventory in the midplane gap region. Despite this bias, comparisons of cycle length exhibit very good agreement between the core model with uniform burnup and the NBSR, which has many decades of operational experience with HEU fuel.

 
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