Program - Brookhaven National Laboratory

Benchmarking of ENDF/B-VII and JENDL-4.0 in the Fast Neutron Range
Mark Cornock
AWE.Plc
The recent releases of the ENDF/B-VII.1 [1] and JENDL-4 [2] libraries represent a significant upgrade in
data for a number of nuclides. As part of its data release process the AWE Nuclear Data team assesses
the internationally available data, through rigorous validation and verification. A part of this validation
process involves Benchmarking of new data using a suite of ICSBEP [3] Benchmarks. K-effectives are
calculated and compared to the published benchmark values. A number of other metrics are compared
to determine which data shows the ”best fit” to the benchmark results. These metrics include a χ2 like
measure, χ2 (kcalc−kexp)/(δkexp) 2
sys = Σ n and a measure of the average difference between calculated and benchmark
k- effectives 〈|∆|〉 = Σ |kcalc−kexp|
n This benchmarking is then used to help make an informed decision
as to which data is recommended for use by AWE’s user base. This initial work focuses on fast benchmark
systems as defined in the ICSBEP Handbook. K-effectives are calculated for a number of fast systems with
a wide range of materials. Validation metrics are calculated and compared for both libraries and compared
to their predecessors. Calculations are in the main performed using a proprietary 1D Sn neutron transport
code and comparisons are made with MCNP [4] calculations. Benchmark calculations will be shown and
discussed; discussion will focus on the general fit of each of the calculations to the benchmark results with
particluar emphasis placed on bare spheres such as JEZEBEL and GODIVA.
British Crown Owned Copyright 2012/AWE Published with the permission of the Controller of Her
Britannic Majesty’s Stationary Office. [1] “ENDF/B-VII.1 Nuclear Data for Science and Technology:
Cross Sections, Covariances, Fission Product Yields and Decay Data”, M.B. Chadwick, M.W. Herman,
P. Oblozinsky et al., NDS112, 2887 (2011). [2] “JENDL-4.0: A New Library for Nuclear Science and
Engineering”, K Shibata, O Iwamoto, T Nagakawa et al., J Nucl Sci Technol, 48(1), 1-30, (2011). [3]
“International handbook of evaluated criticality saftey benchmark experiments”, J.B Briggs et al., Tech
Report NEA/NSC/DOC(95)04/I, (2004). [4] “MCNP- A General Monte Carlo N-Particle Transport Code,
Version 5”, X-5 Monte Carlo Team, LANL Tech Report LA-UR-03-1987, (2003).
NE 4 11:40 AM
239 Pu Prompt Fission Neutron Spectra Impact on a Set of Criticality and Experimental
Reactors Benchmarks
Yannick Peneliau, Julien Henaff
Commissariat à l’Energie Atomique et aux Energies Alternatives - French Atomic Energy Commission -
FRANCE
Many nuclear data are investigated today to improve the calculation predictions of the new neutron transport
simulation codes. With the new generation of nuclear power plants (GEN IV project), one expects
to reduce the calculation uncertainties, mainly the uncertainties coming from nuclear data which are still
very important before adjustment. In France, some future nuclear power plants will use MOX fuel, either
in Sodium Fast Reactors or in Gas Cooled Fast Reactors. The knowledge of 239 Pu cross sections and other
nuclear data is crucial issue to reach this in order to decrease these uncertainties. The Prompt Fission
Neutron Spectra (PFNS) for this isotope are part of this challenge (an IAEA working group is dedicated
to PFNS) and the work presented here deals with this particular topic. The main international data (i.e.
JEFF-3.1.1, ENDFB-7.0, JENDL-4.0, BRC-2009) have been considered and compared with two different
spectra, the ones coming from the works of Maslov and Kornilov. Other spectra are also examined, coming
from the fission process Monte Carlo simulation code, FIFRELIN. The spectra are first compared by
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