Overview of recent U235 neutron cross section evaluation work Page: 4 of 10
This article is part of the collection entitled: Office of Scientific & Technical Information Technical Reports and was provided to Digital Library by the UNT Libraries Government Documents Department.
The following text was automatically extracted from the image on this page using optical character recognition software:
OVERVIEW OF RECENT U235 NEUTRON CROSS SECTION EVALUATION WORK
Lockheed Martin Corporation
P. O. Box 1072
Schenectady, New York 12301-1072
This report is an overview (through 1997) of the U235
neutron cross section evaluation work at Oak Ridge
National Laboratory (ORNL), AEA Technology (Harwell)
and Lockheed Martin Corp.-Schenectady (LMS), which has
influenced, or appeared in, ENDF/B-VI through Release 5.
The discussion is restricted to the thermal and resolved
resonance regions, apart from some questions about the
unresolved region which still need investigation. The
important role which benchmark testing has played will be
The original Releasea of ENDF/B-VI.0 U235 in 1990
was a collaboration among the Oak Ridge, Los Alamos,
and Argonne National Laboratories, the last two providing
the high-energy cross sections, and ORNL the resonance
analysis. The latter culminated in an extensive Reich-
Moore multilevel fit to the resolved-resonance region 0-
2250 eV.' The work was carried out with a powerful new
resonance-fitting code, SAMMY, which incorporated
Bayesian methodology into the least-squares fitting, making
it possible to deal sequentially, yet consistently, with the
many large data sets requiring analysis.' The evaluation
techniques were a major advance over the earlier single-
level treatments in ENDF/B-I through V, the latter of which
ended the resolved region at 82 eV and treated the thermal
region (0-1 eV) as tabulated data in File 3. The ENDF/B-
a Successive versions of ENDF/B-VI materials are
designated as Mods. U235 has undergone modification in
every Release of ENDF/B-VI, so the Mod number, which
started as 1 in Release 0, is one greater than the Release
number. We use the latter, as more familiar to most users.
(The number of Releases, and the fact that different ones
may contain quite different cross sections, makes it
important for users and authors of technical articles to
insure adequate identification of which ENDF/B-VI
materials they used.)
VI evaluators recognized that the experimentally-observed
structure in the cross sections above 100-150 eV was
largely due to clumps of resonances, but it was their opinion
that fitting it as if it were resolved "pseudo-resonances"
would preserve the structure and permit users to derive
more accurate estimates for self-shielded multigroup cross
sections than could be inferred from a traditional treatment
in tens of average unresolved resonance parameters. This
important question has never been settled, and remains one
of the tasks for future investigation.
In the Cross Section Evaluation Working Group
(CSEWG) review procedure, which was quite extensive,
involving special meetings at Oak Ridge, the Thermal
Benchmark Testing Subcommittee, under the chairmanship
of J. R. Hardy and M. L. Williams, observed that there were
two "differential-integral" discrepancies which affected the
1. The 2200 m/s values for capture and fission, and
their associated Westcott g-factors, produced a Maxwellian-
averaged q (v-fission/absorption) which was lower than the
value inferred from eigenvalue calculations of thermal
reactors. Quantitatively, the effect was measured by the
parameter Ki = v-fission minus absorption (all quantities
Maxwellian-averaged). The integral value was 722.7 3.9
whereas the evaluation gave about 719. The 2200 m/s
values were the then-recommended thermal constants that
resulted from an extensive round of least-squares fitting to
pointwise and reactor-averaged thermal data, following a
tradition started many years earlier by G. H. Westcott. The
CSEWG discussions of the least-squares results were
spirited. Within the CSEWG Task Force on Thermal
Constants, led by B. R. Leonard and J. R. Stehn, some
believed that least-squares was the best one could do, while
others believed that the results were too sensitive to the
choice of input uncertainties, and that the output
uncertainties were unrealistically low. It was noted that the
main input data at the low-K1 end were the Chalk River
reactor-average-alpha measurements, which were high and
carried low uncertainties, and some low nubar data. A
complete discussion of the relevant data can be found in
Here’s what’s next.
This article can be searched. Note: Results may vary based on the legibility of text within the document.
Tools / Downloads
Get a copy of this page or view the extracted text.
Citing and Sharing
Basic information for referencing this web page. We also provide extended guidance on usage rights, references, copying or embedding.
Reference the current page of this Article.
Lubitz, C. Overview of recent U235 neutron cross section evaluation work, article, October 1, 1998; Schenectady, New York. (digital.library.unt.edu/ark:/67531/metadc677717/m1/4/: accessed January 17, 2019), University of North Texas Libraries, Digital Library, digital.library.unt.edu; crediting UNT Libraries Government Documents Department.