Skip to main content
U.S. Department of Energy
Office of Scientific and Technical Information

Monte Carlo Perturbation Analysis of Fuel Temperature Variations in the MCNP Model of the Annular Core Research Reactor

Journal Article · · Nuclear Technology
The Annular Core Research Reactor (ACRR) Monte Carlo N-Particle (MCNP) model is used by ACRR reactor operators and experiment designers at Sandia National Laboratories for a variety of computational calculations ranging from reactor kinetics parameter estimates and safety analyses to experimental planning. To understand the dominant source of uncertainty within the MCNP model, perturbations in temperature were applied to individual ACRR MCNP fuel rods. Fuel rod temperatures were randomly sampled from a uniform distribution from operational temperatures to quantify temperature-related uncertainty effects. Stochastic mixing was used to blend the cross sections of the desired temperatures using the MCNP continuous and Thermal Neutron Scattering Treatment [S(α,β)] libraries in ENDF/B-VII.1. Furthermore, this uncertainty analysis produced a 640 row × 640 column correlation and covariance matrix of the neutron energy spectra. Positive covariance was produced around the 1-MeV region and the 0.2-eV region. Correlation was found in the thermal and fast energy regions, but no correlation was observed in the slowing-down energy region because interactions in this region are not dominated by fuel.
Research Organization:
Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)
Sponsoring Organization:
USDOE National Nuclear Security Administration (NNSA)
Grant/Contract Number:
NA0003525
OSTI ID:
3019216
Report Number(s):
SAND--2024-11009J; 1699187
Journal Information:
Nuclear Technology, Journal Name: Nuclear Technology Journal Issue: 6 Vol. 210; ISSN 0029-5450; ISSN 1943-7471
Publisher:
Taylor & FrancisCopyright Statement
Country of Publication:
United States
Language:
English

References (1)

Stochastic mixing of bound thermal scattering data in MONK journal February 2020

Similar Records

MCNP/MCNPX model of the annular core research reactor.
Technical Report · Sun Oct 01 00:00:00 EDT 2006 · OSTI ID:895071

Random Variation of ACRR Core Report
Technical Report · Wed Apr 01 00:00:00 EDT 2020 · OSTI ID:1617559

Method for Calculating Delayed Gamma-Ray Response in the ACRR Central Cavity and FREC-II Cavity Using MCNP
Technical Report · Mon Dec 31 23:00:00 EST 2018 · OSTI ID:1762940