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Title: Proposed re-evaluation of the 154Eu thermal ( n, γ) capture cross-section based on spent fuel benchmarking studies

154Eu is a nuclide of considerable importance to both non-destructive measurements of used nuclear fuel assembly burnup as well as for calculating the radiation source term for used fuel storage and transportation. But, recent evidence from code validation studies of spent fuel benchmarks have revealed evidence of a systemic bias in predicted 154Eu inventories when using ENDF/B-VII.0 and ENDF/B-VII.1 nuclear data libraries, wherein Eu-154 is consistently over-predicted on the order of 10% or more. Further, this bias is found to correlate with sample burnup, resulting in a larger departure from experimental measurements for higher sample burnups. Here, the bias in Eu-154 is characterized across eleven spent fuel destructive assay benchmarks from five different assemblies. Based on these studies, possible amendments to the ENDF/B-VII.0 and VII.1 evaluations of the 154Eu (n,γ) 155Eu are explored. By amending the location of the first resolved resonance for the 154Eu radiative capture cross-section (centered at 0.195 eV in ENDF/B-VII.0 and VII.1) to 0.188 eV and adjusting the neutron capture width proportional to $$\sqrt1/E$$, the amended cross-section evaluation was found to reduce the bias in predicted 154Eu inventories by approximately 5–7%. And while the amended capture cross-section still results in a residual over-prediction of 154Eu (ranging from 2% to 9%), the effect is substantially attenuated compared with the nominal ENDF/B-VII.0 and VII.1 evaluations.
  1. Univ. of Tennessee, Knoxville, TN (United States). Dept. of Nuclear Engineering
Publication Date:
Grant/Contract Number:
Published Article
Journal Name:
Annals of Nuclear Energy (Oxford)
Additional Journal Information:
Journal Name: Annals of Nuclear Energy (Oxford); Journal Volume: 99; Journal Issue: C; Journal ID: ISSN 0306-4549
Research Org:
Univ. of Tennessee, Knoxville, TN (United States)
Sponsoring Org:
USDOE Office of Nuclear Energy (NE), Fuel Cycle Technologies (NE-5)
Country of Publication:
United States
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS; Eu-154; Nuclear data; ENDF; Burnup; Nuclear fuel depletion
OSTI Identifier:
Alternate Identifier(s):
OSTI ID: 1364131