The CIELO Collaboration:
Neutron Reactions on 1 H, 16 O, 56 Fe, 235,238 U, and 239 Pu
M.B. Chadwick,1, ∗ E. Dupont,2 E. Bauge,3 A. Blokhin,4 O. Bouland,5 D.A. Brown,6 R. Capote,7 A. Carlson,8
Y. Danon,9 C. De Saint Jean,5 M. Dunn,10 U. Fischer,11 R.A. Forrest,7 S.C. Frankle,1 T. Fukahori,12Z. Ge,13 S.M. Grimes,14 G.M. Hale,1 M. Herman,6 A. Ignatyuk,4 M. Ishikawa,12 N. Iwamoto,12
O. Iwamoto,12 M. Jandel,1 R. Jacqmin,1 T. Kawano,1 S. Kunieda,12 A. Kahler,1 B. Kiedrowski,1 I. Kodeli,15
A.J. Koning,16 L. Leal,10 Y.O. Lee,17 J.P. Lestone,1 C. Lubitz,18 M. MacInnes,1 D. McNabb,19
R. McKnight,20 M. Moxon,21 S. Mughabghab,6 G. Noguere,5 G. Palmiotti,22 A. Plompen,23 B. Pritychenko,6
V. Pronyaev,4 D. Rochman,16 P. Romain,3 D. Roubtsov,24 P. Schillebeeckx,23 M. Salvatores,5 S. Simakov,7
E.Sh. Soukhovitskiı̃,25 J.C. Sublet,26 P. Talou,1 I. Thompson,19 A. Trkov,15 R. Vogt,19 and S. van der Marck16
1

Los Alamos National Laboratory, Los Alamos, NM 87545, USA
2
Nuclear Energy Agency, OECD, Issy-les-Moulineaux, France
3
CEA, DAM Ile de France, F-91297 Arpajon, France
4
Institute of Physics and Power Engineering, Obninsk, Russia
5
CEA, Nuclear Energy Division, Cadarache, Saint-Paul-lez-Durance, France
6
Brookhaven National Laboratory, Upton, NY 11973-5000, USA
7
International Atomic Energy Agency, Vienna-A-1400, PO Box 100, Austria
8
National Institute of Standards and Technology, Gaithersburg, MD, USA
9
Rensselaer Polytechnic Institute, Troy, NY, USA
10
Oak Ridge National Laboratory, Oak Ridge, TN 37831, USA
11
Karlsruhe Institute of Technology, Karlsruhe, Germany
12
Japan Atomic Energy Agency, Tokai-mura, Japan
13
China Institute of Atomic Energy, Beijing, China
14
Ohio University, Athens, OH 45701, USA
15
Jozef Stefan Institute, Ljubljana, Slovenia
16
Nuclear Research and Consultancy Group, Petten, The Netherlands
17
Korea Atomic Energy Research Institute, Daejeon, South Korea
18
Knolls Atomic Power Laboratory, Schenectady, NY, USA
19
Lawrence Livermore National Laboratory, Livermore, CA, USA
20
Argonne National Laboratory, Argonne, USA
21
Abingdon, Oxfordshire, United Kingdom
22
Idaho National Laboratory, Idaho Falls, ID, USA
23
EC-JRC-IRMM, Geel, Belgium
24
Chalk River Laboratories, AECL, Chalk River, Ontario, Canada
25
Joint Institute for Energy and Nuclear Research, 220109, Minsk-Sosny, Belarus
26
Culham Centre for Fusion Energy, Abingdon, Oxfordshire, United Kingdom(Dated: October 4, 2013)
CIELO (Collaborative International Evaluated Library Organization) provides a new working
paradigm to facilitate evaluated nuclear reaction data advances. It brings together experts fromacross the international nuclear reaction data community to identify and document discrepancies
among existing evaluated data libraries, measured data, and model calculation interpretations, andaims to make progress in reconciling these discrepancies to create more accurate ENDF-formatted
files. The focus will initially be on a small number of the highest-priority isotopes, namely 1 H, 16 O,
56
Fe, 235,238 U, and 239 Pu. This paper identifies discrepancies between various evaluations of thehighest priority isotopes, and was commissioned by the OECD’s Nuclear Energy Agency WPEC(Working Party on International Nuclear Data Evaluation Co-operation) during a meeting held in
May 2012. The evaluated data for these materials in the existing nuclear data libraries — ENDF/BVII.1, JEFF-3.1, JENDL-4.0, CENDL-3.1, ROSFOND, IRDFF 1.0 — are reviewed, discrepancies
are identified, and some integral properties are given. The paper summarizes a program of nuclear
science and computational work needed to create the new CIELO nuclear data evaluations.

∗ Corresponding author: mbchadwick@lanl.gov

The CIELO Collaboration . . .

NUCLEAR DATA SHEETS

M.B. Chadwick

References

CONTENTS

I. INTRODUCTION

2

II. LIGHT ELEMENTS
A. 1 H
1. Summary of the evaluations
2. Elastic scattering
B. 16 O
1. Summary of the evaluations
2. (n,α) reaction
3. Radiative capture
4. Elastic scattering

5
5
5
5
5
5
5
6
6

III. STRUCTURAL MATERIALS
A. 56 Fe
1. Summary of the evaluations
2. Inelastic scattering
3. (n,xn) and (n,xp) reactions
4. (n,α) reaction

7
7
7
7
8
8

IV. ACTINIDES
A. General comments
B. Compensating errors
C. Fission cross sections
D. Prompt fission neutron spectra
E. Prompt fission gamma-ray spectra
F. Inelastic scattering
G. Radiative capture
H. 235 U
1. Resolved resonance parameters
2. Radiative capture
3. Inelastic scattering
4. (n,2n) reaction
5. Average number of neutrons per fission ν
6. PFNS integral validation
I. 238 U
1. Radiative capture
2. Elastic and inelastic scattering
3. (n, 2n) reaction
4. Average number of neutrons per fission ν
J. 239 Pu
1. Resolved resonance parameters
2. Radiative capture
3. Inelastic scattering
4. (n, 2n) reaction
5. Average number of neutrons per fission ν
6. PFNS integral validation

8
8
9
9
9
10
10
11
11
11
11
12
12
12
12
14
14
14
14
15
15
15
15
16
16
16
16

V. VALIDATION BENCHMARKS

17

VI. ANALYSIS OF INTEGRAL QUANTITIES
A. Thermal cross sections
B. Westcott Factors
C. Resonance Integrals
D. Maxwellian-averaged cross sections
E. 252 Cf(sf) cross sections

17
18
18
18
19
20

VII. CONCLUSIONS

21

21

A. Updated Results of LANL Integral
Experiments

I.

23

INTRODUCTION

Outstanding progress has been made around the worldin nuclear reaction and decay data evaluation. The
quality of the main evaluated data libraries, such as
ENDF/B-VII.1 [1], JEFF-3.1 [2, 3], JENDL-4.0 [4],BROND/ROSFOND [5], and CENDL-3.1 [6], is high, andfor the most part the libraries perform well in neutronics simulations for fission and fusion energy applications(though covariance data that represent uncertainties areless advanced). However, our current understanding is insuﬃcient in many essential areas, some user needs remain
inadequately addressed, and a new working paradigm isneeded to expedite future evaluated nuclear reaction dataadvances. We see this as being facilitated by (a) poolingexpertise from across the world through creation of collaborative teams, and (b) using new computational techniques for optimization, sensitivity analyses, and uncertainty quantification (UQ). Stronger international collaborations will provide a new framework for nuclear data
evaluation, and will help establish the highest fidelity general purpose nuclear database for all nuclear science communities around the world.
It is recognized that for many important applications,for example nuclear criticality calculations, the existingevaluated data perform well in transport simulations owing, in part, to compensating errors in the databases.
Diﬀerent cross section libraries may predict almost thesame keﬀ for benchmark experiments, but for very different reasons at a microscopic level [7, 8]. Such errors must be minimized since simulation predictions awayfrom calibration points (corresponding to the benchmark
experiments) can rapidly become erroneous if the underlying physical data used in a simulation are incorrect.
Also, cross sections for transmutation reactions, including fission, capture, and (n, 2n), are inadequately known
for certain applications. And in many cases scatteringcross sections — elastic and inelastic, and secondary neutron energy and angular distributions — are inadequately
known for transport calculations.
In this paper we suggest that a new paradigm is needed
to more rapidly advance our understanding for the evaluation of nuclear reaction cross sections. Closer international cooperation is needed, where the world’s experts
for various capabilities are brought together to solve theproblems and to provide peer review on proposed solutions. We suggest the name for this collaborative effort, and for the evaluated data libraries that results, be
CIELO (Collaborative International Evaluated Library
Organization) — the name has the merit of not being
closely related to the name of any of the existing evaluation projects, and it emphasizes the centrality of collab2

The CIELO Collaboration . . .

NUCLEAR DATA SHEETS

M.B. Chadwick

oration. We anticipate that the OECD’s Nuclear Energy
Agency and the IAEA’s Nuclear Data Section will provide important coordination resources for this project,
and the resulting product will be analogous to the “standards cross section files”, but for general purpose transport applications (criticality, activation, shielding etc.).
The principal motivations for CIELO are:
• The product — CIELO — will be higher quality,
benefiting from the combined eﬀort of the world’s
experts in the fields of nuclear data measurement,
modeling, evaluation, data processing, and integral
validation.
• The scope and complexity of evaluated cross section libraries have become so large that we should
consider sharing responsibilities, and resources, fordeveloping the files.

FIG. 1. Calculated/measured root mean square deviation (in%) from unity for nuclear criticality, keﬀ , for a suite of 119 critical assemblies, as a function of ENDF library version number.

• Such collaborations of experts around the world areless likely to unintentionally omit key measured differential or integral data, or physical insights, intheir considerations.

understanding for the participating organizations, though
we do not attempt to address such issues here.
One way to measure the overall progress that has been
made in advancing the accuracy of evaluated data libraries is to consider comparisons of calculated and measured nuclear criticality (keﬀ ). This has been done at
Los Alamos for a suite of 119 critical assemblies [9] using
the MCNP6 simulation code [10], for various data setsthat go back over three decades in time to ENDF/B-V.
This suite of critical assemblies includes a selection of all
the diﬀerent neutron spectrum types (fast, intermediate,
thermal), actinide materials (HEU, DU, Pu, ...), reflector materials, and so on. The root-mean-square (rms)deviation in C/E from unity is shown as a function of
the library version, see Fig. 1. Whilst the error is seen
to get smaller with each release of ENDF, it is evident
that it is diﬃcult to make substantial improvements withthe methods we are presently using – we are facing diminishing returns. CIELO, using new evaluation and optimization methods may enable more significant future
improvements. Of course we recognize that considerations of keﬀ are inadequate to fully quantify the qualityof evaluated cross section sets, and one should also look at
the evaluated cross sections compared to diﬀerential laboratory measurements, where substantial advances havebeen made over the last decades.
It is useful to list just a few of the successful international collaborative projects that demonstrate the feasibility, and the value, of such eﬀorts.

• Errors, and the possibility of poor evaluation decisions, will be minimized through increased peer
review. This is particularly important for technicalareas where there are few specialists working.
• The diversity of evaluated data libraries has been
useful as a tool to communicate open questions regarding our knowledge of cross sections, but newcovariances in ENDF, JEFF, JENDL, etc. have recently become available, and updated covariances
should characterize our uncertainties.
• Pooling our resources to develop high fidelity evaluated data files may help attract new users and key
experts who are not typically involved in nuclear
data activities.
• The current ENDF, JENDL, JEFF, BROND, andCENDL evaluated data libraries have reached alevel of maturity that allows us to prepare the next
step that will respond to users’ long-term needs ofever more complete and accurate nuclear data forintensive numerical simulations.
Of course, to date cross section evaluation eﬀorts havealready benefited in various ways from international collaborations, especially collaborations under the auspicesof the IAEA and the WPEC (Working Party on International Nuclear Data Evaluation Co-operation) of theOECD Nuclear Energy Agency. Furthermore, there has
been much “borrowing” where one evaluation project haslargely, or entirely, adopted an evaluation from anotherproject. So what we are proposing is not new, although
it does represent a significant increase in intensity. It is
likely that we will need to develop completely new working procedures for eﬀectively and eﬃciently working together in these international collaborations. This may require establishing a governance model and memoranda of• The IAEA/WPEC/CSEWG standard eﬀort wascomprised of leading researchers from around theworld [11]. Some of the resulting standards havebeen adopted by many projects (e.g. ENDF/BVII.1, JENDL-4.0, JEFF-3.1), and the IAEA has
established a long-term commitment to supporting
this eﬀort.
3

The CIELO Collaboration . . .

NUCLEAR DATA SHEETS• Several special-purpose nuclear data files havebeen produced under IAEA coordination, such as
the International Reactor Dosimetry and Fusion
File (IRDFF) [12], the photonuclear data library,
and the Fusion Evaluated Nuclear Data Library(FENDL) [13].

M.B. Chadwick

be determined later from user community input, from the
WPEC High Priority Request List (HPRL), and from priorities identified by WPEC Subgroup 26, 33 and follow-on
eﬀorts.
The second goal is for this international collaboration to create demonstrably more accurate evaluations
such that the regional eﬀorts (ENDF, JEFF, JENDL,
BROND/ROSFOND, CENDL, TENDL etc.) will naturally want to adopt CIELO evaluations (or partial evaluations). Over time this may lead to internationally consistent, and internationally maintained, evaluated databasesthat best represent these constants of nature. But before
such a scenario can be planned , we — as an international
community — first need to demonstrate that the CIELO
concept is indeed feasible.
In this paper we describe the status of existing files ineach evaluation project nucleus by nucleus, together witha summary of some outstanding issues that need to besolved. The issues we intend to summarize are: (a) discrepancies amongst the evaluated nuclear data libraries(ENDF, JENDL, JEFF, etc.); (b) notable experimental
data gaps/discrepancies limiting our understanding; and(c) notable diﬀerences in theory and interpretation. In
these descriptions, the extent to which the current evaluations already involve much collaboration and borrowing
from other evaluations is evident. The thermal cross sections, Westcott factors, resonance integrals, Maxwellianaveraged cross sections (MACS) at 30 keV, and 252 Cf(sf)neutron spectrum averaged cross sections, are given inthis paper to help identify and illustrate diﬀerences anddiscrepancies.
In addition to this document, we provide a compilationof figures created by the National Nuclear Data Center(NNDC) of BNL in an automated way from the various
evaluated databases. As this project evolves, additional
figures will be created to help identify and illustrate discrepancies that need to be resolved. This supplement isavailable to collaborators on the NEA website [17].
Evaluations of cross section uncertainties and correlations (covariances) have become an increasingly important component of nuclear data evaluation work. Such covariance data are important in subsequent studies of theoverall uncertainty budget of a neutron transport integrated simulation, for example for criticality or for transmutation rate predictions. There has been much recent
work in the nuclear data and transport community on covariances, and the recent releases of evaluated databases
include fairly comprehensive covariance data. However,such capabilities are in their early stages and significant
improvements are needed to ensure the uncertainties are
credible, and complete. CIELO collaborations will include this important topic, although because of limited
space we do not address covariances in any detail in thispaper.

• A new 232 Th evaluation was undertaken by an
IAEA Coordinated Research Project (CRP) andsubsequently adopted by ENDF.
• Evaluations recommended by WPEC’s Subgroup
23 (Evaluated Data Library for the Bulk of Fission Products) [14] have been adopted by ENDF,
and provide guidance for future evaluations in otherprojects.
• WPEC Subgroup 4 (238 U Capture and Inelastic
Cross-sections) [15] helped solve a long-standing
problem on the capture cross section in the fast
energy range.
• LANL, LLNL, and the CEA have had longstanding collaborations on nuclear reactions in thefast range. Collaborations between ORNL and
CEA/Cadarache have also led to jointly developed
resonance analyses for many actinides. Two ofthese, 235,238 U, were done under the auspices of
WPEC subgroups 18 and 22 in collaboration with
Harwell, KAPL, LANL, and other laboratories.
ENDF/B-VI Release 5 of 235 U was adjusted to calculate HEU thermal solution benchmark eigenvalues correctly, and 238 U extended those “targeted”
methods to LEU systems. These have been adopted
by many other projects. Because the average eigenvalue of uranium-fueled assemblies is now so close
to unity, CIELO will have the additional constraintof maintaining that desirable condition.
• The Reference Input Parameter Library (RIPL)[16] illustrates how teams of scientists across theworld can come together (under the auspice of the
IAEA in this case) to create a valuable reference
work summarizing our understanding of nuclear reaction physics.
The first goal of CIELO is shorter term: bringing together experts from across the international nuclear reaction data community to identify discrepancies (and document reasons for the discrepancies) among existing evaluated data libraries, measured data, and model calculationinterpretations (described herein), and to make progressin reconciling these discrepancies to create more accurate ENDF-formatted files. The focus will initially be
on a small number of the highest-priority isotopes (1 H,
16O, 56 Fe, 235,238 U, 239 Pu). If successful, this will subsequently be expanded to include an ever-broadening suiteof nuclei. We plan to place the second priority on nuclei
such as 2 H, 6,7 Li, 9 Be, 10,11 B, 12 C, 23 Na, 52 Cr, 58,60 Ni,
Mo isotopes, 240,241 Pu, and 241 Am, but the specifics will
4

The CIELO Collaboration . . .
II.

1

H

Summary of the evaluations

The n-p cross section is a primary standard in nuclear
physics, and the cross section represents a basic tool forstudying the nuclear force [11, 18]. The n-p cross section plays an essential role in many neutronics simulationsof transport and criticality, as well as absolute measurements of neutron-induced reactions.
The R-matrix analysis for the n-p cross section was
performed by G. Hale of LANL as part of the IAEAWPEC-CSEWG standards eﬀort [11], and it was compiled into the most recent ENDF/B-VII.0 evaluation,
which is identical to ENDF/B-VII.1. This was adopted
by JENDL-4.0 and ROSFOND. JEFF-3.1 still uses an
earlier version of this work from ENDF/B-VI.8. The
most recent ENDF/B-VII.0 work by Hale benefits from
inclusion of more data (in addition to n-p and p-p data)
which includes n-p capture and phototisintegration channels, and Ohio University n-p elastic scattering angular
distribution data at 10 MeV [19]. The latter measurement led to ENDF/B-VII.0 upgrades which correct theangular distribution in ENDF/B-VI.8 that had too large
a backward-forward center-of-mass (cm) ratio in the angular distribution at 10 MeV.

2.

M.B. Chadwick

is 1% lower than those of Arndt et al. and the Nijmegen group. A more precise determination (1-2%) of
both the total and double-diﬀerential elastic scatteringcross-sections is requested at high incident energy (1020 MeV) with emphasis on the data at small cm angles (see High Priority Request List at http://www.oecdnea.org/dbdata/hprl/hprlview.pl?ID=463).

LIGHT ELEMENTS
A.

1.

NUCLEAR DATA SHEETS

B.
1.

16

O

Summary of the evaluations

The following description of the oxygen evaluations in
use around the world illustrates the interconnected natureof such eﬀorts. The ENDF/B-VII.0 oxygen evaluation,
developed by Hale et al. at Los Alamos and by Lubitz
and Caro (KAPL), has been adopted by many projects.
In particular, JEFF-3.1 has adopted this work, which included an upgrade to the (n,α) reaction compared withthe earlier ENDF/B-VI.8 file, as has the CENDL-3.1 library. Also, the ROSFOND evaluation adopts ENDF/BVII.0, though an improvement to radiative capture wasmade by use of the JENDL-3.3 cross section for capture.
And indeed, the new ENDF/B-VII.1 file, although identical to VII.0 in other respects, adopts the JENDL-4.0 capture cross section since ENDF/B-VII.0 needed to be modified to account for deviations away from 1/v at higher
energies important in nucleosynthesis applications. Although the JENDL-4.0 evaluation is largely independentof the US ENDF work, it did adopt the ENDF/B-VII.0(n,α) cross section below 6.5 MeV.

Elastic scattering

As noted above, the existing evaluations have largely
converged. Future work will most usefully include further refinements to the R-matrix analysis through the
IAEA-WPEC-CSEWG collaboration. Some possibilities
are described below.
Some recent high-accuracy angular distribution measurements have been made by Boukharouba et al. [19] at
Ohio University at 14.9 MeV. The motivation was to better understand this important cross section, and to assess
the accuracy of the ENDF/B-VII and phase-shift analysis
cross sections. The authors concluded that ENDF/B-VII
agrees well with their data, but open questions have hindered a more rigorous test. In particular, very few data
have been measured at small cm angles — the lowestangle data dates from 1967 — and this leads to ambiguities as to whether the shape of the angular distribution
monotonically decreases with decreasing angle (P1 component in the Legendre expanded angular distribution) or
whether there is a backward- and forward-angle increase(P2 ). Future small-angle precision data are required tohelp determine this angular distribution cross section todiscriminate between the ENDF/B-VII and the Arndt etal. and the Nijmegen group predictions [20, 21], and toadvance future evaluations. Furthermore, at 14.9 MeV
the total elastic scattering cross section in ENDF/B-VII2.

(n,α) reaction

Discrepancies of up to 30% in both measured and evaluated 16 O(n, α) data have been a long-standing issue in thefield of fission applications. These discrepancies aﬀect theprediction of keﬀ for current and innovative reactors, theprediction of helium production in reactors, and the calibration of reference neutron-source strength in metrology
measurements. New measurements and evaluations havebeen requested in the range 2.5-20 MeV to reduce theuncertainty down to 5-10% (see the High Priority Request List and documents therein at http://www.oecdnea.org/dbdata/hprl/hprlview.pl?ID=417).
The existing 16 O(n,α) cross section in ENDF/B-VII.0
and VII.1 is consistent with some data below 6 MeV (thenew IRMM and IPPE measurements by Giorginis [22]and the Bair and Haas ORNL data [23] with their recommended 20% decrease) but is discrepant with these newdata above 6 MeV. A developmental/test set of multichannel R-matrix calculations by Hale suggested a fit todata that was about 30% higher for (n,α) in the 3–6 MeV
range (in contradiction to the new IRMM/IPPE data) —this was not adopted for VII.1, as it was decided morework is needed to understand this channel.
5The CIELO Collaboration . . .

NUCLEAR DATA SHEETS
0.3

The (n,α) channel does aﬀect criticality — forexample, the lower (n,α) cross section adopted for
ENDF/B-VII.0 impacted LCT (Low-enriched-uranium,
Compound, Thermal spectrum) benchmarks by increasing keﬀ slightly [24]. An earlier version of ENDF/BVI had a higher (n, α) cross section (that is, it better
matched the original Bair and Haas data without following their guidance to lower the cross section), and thisresulted in H. Huria of Westinghouse finding a loweringof his calculated criticality by 100 – 200 pcm for a commercial reactor design. This led J. Weinman of KAPL
to calculate criticality for various homogeneous spheres(in the HST series 1, 9-13, 32, 42, 43) with the resultthat a decreasing trend as a function of oxygen absorption was observed for ENDF/B-VI.8, but less of a trend
was observed for ENDF/B-V. This was part of the reason for Page et al. decreasing the (n, α) cross section
for ENDF/B-VII.0 (and VII.1). These integral calculations need to be redone, and an assessment made as towhether they can be used to guide evaluation decisions
on the (n, α) cross section.
Recently finalized IRMM measured data have been released by Giorginis. The author concludes that ENDF/BVII.1=VII.0 is accurate below 6 MeV but most likely
needs to be modified above 6 MeV (higher from 6.0 to
8.8 MeV; lower above 8.8 MeV, see Fig. 2); the author
even provides a scaling factor to apply to ENDF/B-VII.0
above 6 MeV (above 6 MeV the existing ENDF/B-VII.0
evaluation follows earlier ENDF analyses in matching theolder US data by Davis et al., but these data are discrepant with the newer Geel-IRMM and IPPE data). The
IRMM data are in good agreement with the Harissopulos et al. [25] data below 5.2 MeV, but at 6.4 MeV the
IRMM data are larger and agree with the corrected Bairand Haas 1973 Oak Ridge data (Bair and Haas recommended a 15–20% reduction to their original data). At
5.4 MeV the IRMM data agree with the IPPE data, but
at 6.4 MeV the IPPE data are significantly higher; Giorginis et al. suggest that a problem exists with the IPPE
data.
Future (n, α) evaluation work on oxygen is needed toaddress at least two issues. Firstly, as described above
there are some discrepancies in the measurements for(n, α) direct and inverse channels. If the recent IRMM
data are correct, ENDF needs to be modified above
6 MeV. Secondly, there are R-matrix theory insights that
could be in conflict with these IRMM data. The R-matrix
analysis by Hale, which is also informed by the total crosssection data, presently predicts a higher (n, α) cross section than ENDF for the few MeVs above threshold, as
does preliminary R-matrix work by Kunieda and by Leal.

3.

M.B. Chadwick
ENDF/B-VI
ENDF/B-VII.1
IRMM (2008)

Cross Section [b]

0.25

0.2

0.15

0.1

0.05

0
3.5

0.3

4
4.5
5
Neutron Incident Energy [MeV]ENDF/B-VI
ENDF/B-VII.1
IRMM (2006,2007)

0.25

Cross Section [b]

5.5

0.2

0.15

0.1

0.05

0
5

6
7
8
9
Neutron Incident Energy [MeV]10

FIG. 2. 16 O(n, α) showing recent IRMM measurements by
Giorginis et al. They suggest that the ENDF/B-VII.1 needs
to be increased above 6 MeV. The (n,α0 ) cross sections in
ENDF/B-VI and ENDF/B-VII.1 are broadened by a Gaussian
to reflect the detector response.

pecially at 30 keV for nuclear astrophysics applications.
The capture data in JENDL include the direct capture
and a description of the resonance process performed by
Mengoni making use of the Igashira [26] data. JEFF3.1 and CENDL have not yet adopted this upgrade. As
the radiative capture cross section is expected to have aresonance structure for energies above 0.1 MeV, a future
upgrade above this energy would be beneficial.

Radiative capture

4.

ENDF/B-VII.1 adopted JENDL-4.0 data from below
thermal to higher energies. This was motivated by adesire to have more accurate values above 0.1 keV, es-

Elastic scattering

The thermal scattering cross sections in the libraries,
which are summarized later, do not agree with the experimental data of Dilg et al. [27] of 3.761 ± 0.007 b, and6

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NUCLEAR DATA SHEETS

Mughabghab [28] adopted the same value with a slightly
smaller uncertainty of 3.761 ± 0.006 b. In the modern
evaluations, ENDF/B-VII.1, JEFF-3.1, CENDL-3.1, andROSFOND have 3.852 b, and JENDL-4.0 has 3.841 b.
As Plompen [29], Hale, and Lubitz have noted, there
appears to be a discrepancy between various data sets.
Ohkubo (1984) and Johnson’s (1974) measurements(above 1 keV) give a higher value compared with othermeasurements by Dilg, Koester, Block, and Melkonian.
Lubitz has noted that discrepancies in various data sets
could be related to the role of Doppler broadening of aconstant cross section, which has not always been appreciated, and which gives a 3% eﬀect for oyxgen at room
temperature [30, 31]. Also, Plompen’s weighted average of measurements of the coherent scattering length
gives 5.825(1) fm, slightly higher than Atlas’s 5.805(4)fm, but considerably smaller than the ENDF/B-VII.1 derived value of 5.875 - and Plompen and Lubitz note thatthese lower values for the coherent scattering length areconsistent with a smaller total elastic cross section at low
energies (5.825 fm gives a cross section of 3.787 b; 5.805
fm gives a cross section of 3.761 b) [29]. A high-accuracy
measurement here would be most useful.
New R-matrix analyses that aim to resolve the differences are being performed by Hale and by Kunieda.
These analyses, which include cross section and scattering length measurements, will correct measured data to
0 K values as needed, and bridge information that is contained in the higher energy and lower energy data. Kozier
and Roubtsov [32] have noted this 3% discrepancy andthat some heavy water criticality benchmarks seem tosupport a lower thermal scattering cross section (the impact on heavy water reactors is larger than on LWRs).

III.

VII.1 and CENDL have the same resonance parameter
set. Other evaluations (JENDL, JEFF, and ROSFOND)
adopt a modified version of the resolved resonances by
Fröhner, performed for the JEF-2.2 evaluation.
In the MeV energy region, the fluctuation behavior
seen in the experimental total cross sections, which an
optical model cannot reproduce, should exist in the evaluated files, as this is crucial for neutron transport andshielding calculations. Usually the total cross sectionsin this energy region are obtained by tracing the experimental data available. For the other reaction channels,
the Hauser-Feshbach model calculations are used for theevaluation, though the model codes employed are diﬀerent.
As iron is one of the most important materials in experimental accelerator facilities, many evaluations go beyond 20 MeV. ENDF/B-VII.1 includes the LA-150 evaluation that goes up to 150 MeV, and JEFF-3.1 contains
TALYS calculations all the way up to 200 MeV. Other
files (JENDL-4.0, CENDL-3.1, and ROSFOND) just go
up to 20 MeV, but JENDL-4.0 has a separate file for highenergy applications (JENDL/HE).

2.

1.

56

Inelastic scattering

The nuclear data uncertainty assessment performed by
WPEC Subgroup 26 for innovative reactor systems showsthat the knowledge of the inelastic scattering cross section of 56 Fe should be improved to meet the target accuracy requirements for these systems. New measurements and evaluations have been requested in the range
0.5-20 MeV to reduce the uncertainty down to 2-10% depending on the energy region and the reactor considered(see the High Priority Request List at http://www.oecdnea.org/dbdata/hprl/hprlview.pl?ID=454).
There are significant diﬀerences in evaluated inelastic
scattering for iron. The diﬀerence in the inelastic scattering among the libraries mainly come from experimental
data used. ENDF/B-VII.1, JENDL-4.0, and CENDL3.1 include a resonance-like shape taken from experimental data up to 2.1 MeV. ROSFOND goes up to 5 MeV.
JEFF-3.1 considers high-resolution experimental data, sothat the fluctuation continues up to 10 MeV. Importantly,
these diﬀerences are not only related to fluctuations thatcome from diﬀerent experiments but there are significant diﬀerences in the multi-group averaged cross sections. This is evident by looking at the ratio of JENDL3.3 and JEFF-3.1 to ENDF/B-VII.0 (= ENDF/B-VII.1)
with the associated cross section uncertainties in a 40energy group representation. Below 2 MeV the diﬀerences reach 28%. JENDL-3.3 and ENDF/B-VII.1 diﬀer
by only a few percent and agree within their uncertainties. However, there are significant discrepancies between
these two evaluations and JEFF-3.1 below 5 MeV. JEFF3.1 adopts small uncertainties. The original ENDF/BVII.0 uncertainties between 1 and 2 MeV, which wereabout 3.3%, were increased to 16% in COMMARA-2 to

STRUCTURAL MATERIALS
A.

M.B. Chadwick

Fe

Summary of the evaluations

Although we saw a lot of “borrowing” from other evaluations in the 1 H and 16 O evaluations, the 56 Fe evaluationsin the libraries are largely independent, with some exceptions such as the resolved resonance parameters. The
evaluations rely heavily on the optical model and statistical model calculations, where the secondary particle
energy and angular distributions play an important rolein radiation shielding calculations.
The evaluations can be separated into four energy
ranges: (a) the resolved resonance region up to 850 keV,(b) from 850 keV to about 7 MeV where fluctuation still
persists in the measured total cross section, (c) fromabout 7 MeV to 20 MeV, and (d) above 20 MeV. The
separation of regions (b) and (c) is only for the total andelastic scattering cross sections.
F. Perey and C. Perey of ORNL evaluated the resolved
resonance parameters for ENDF/B-VI, and ENDF/B7

The CIELO Collaboration . . .

NUCLEAR DATA SHEETS

2

Inelastic Scattering [b]tering. Nelson (LANL) has gamma-production data foriron. Plompen (Geel) has inelastic data from gammaproduction measurements, measured recently from 800
keV to 5 MeV, that will also be valuable, and Schillebeeckx has made some new measurements. The analysis of these data, as well as SINBAD shielding benchmarks (EURACOS, ASPIS, IPPE, OKTAVIAN, etc.) isplanned. Also, an OM is being developed by the IAEA
for the analysis of iron scattering. To date, most evaluations use a spherical model but the soft rotor model
proposed by Soukhovitskii [35] may be more appropriate, and the IAEA collaboration is seeking to develop a
soft-rotor coupled-channels optical model that describes
all scattering data. Such a model may also improve evaluated data representations of the unresolved resonanceregion. Finally, we note that the CEA is performing sensitivity studies to assess the importance of scattering angular distributions in the resonance region.

ENDF/B-VII.1
JENDL-4.0
JEFF-3.1
CENDL-3.1
RUSFOND

1.5

1

0.5

0
0

5
10
15
Neutron Incident Energy [MeV]

M.B. Chadwick

20

FIG. 3. Comparison of the evaluated inelastic scattering crosssections for 56 Fe.

3.

(n,xn) and (n,xp) reactions

The evaluated data for double-diﬀerential neutron
or proton emission cross-sections in the 20–200 MeV
range rely on pre-equilibrium models essentially constrained by proton-induced (p,xn) and (p,xp) experimental spectra.
The lack of experimental datafor double-diﬀerential (n,xn) spectra may introduce
a bias in the model calculation predictions (seethe High Priority Request List at http://www.oecdnea.org/dbdata/hprl/hprlview.pl?ID=423.)

reflect the large discrepancies between the various evaluations.
Above 10 MeV all libraries give a similar shape for the
total inelastic scattering cross sections and the diﬀerences
there are not so significant. For example, at 14 MeV they
are 767 mb (CENDL), 774 mb (ENDF), 804 mb (JEFF),
773 mb (JENDL), and 800 mb (ROSFOND). The total
inelastic scattering cross sections in these libraries arecompared in Fig. 3.
Recent measurements by Wenner et al. [33] of neutron transmission through iron spheres at Ohio University, for quasi-monoenergetic neutrons ranging from ∼1
to ∼10 MeV, suggest a deficiency in the iron nonelastic cross section that progressively gets worse with increasing neutron energy. This was ascribed to deficiencies in the evaluated ENDF total inelastic cross section.
The suggested reduction in the inelastic cross sections islarge: the authors suggest decreases of 21%, 29%, and
35% at 6.2, 8.2, and 10.8 MeV respectively. At present
ENDF/B-VII.1, JEFF-3.1, and JENDL-4.0 have similar
total nonelastic cross sections in the 6–10-MeV range (approximately 1.4 b); Wenner et al. suggest that the correct value over this energy range decreases from about
1.1 b to 1.0 b, and they note that their model with Dietrich [34] (using insights from Wick’s limit) supports sucha reduced nonelastic cross section. Work is needed to
study this question, and to apply optical model insights.
These Ohio measurements confirm the ENDF/B-VII.1
total cross section to better than 1% in the 6 – 8 MeV
range.
RPI has high-resolution transmission data up to 2
MeV, and scattering data (“quasi-diﬀerential data”),
which need an MCNP calculation to compare with thedata, for 0.5 to 20 MeV, and these point to improvements needed in ENDF for elastic and inelastic scat-4.

(n,α) reaction

There was a long-standing issue concerning the cluster emission in the pre-equilibrium process, where phenomenological alpha-particle knock-out models do not reproduce the recent measurements of alpha-particle productions at LANSCE/LANL. Since gas production in the
structural material causes serious embrittlement problems in fusion reactors or other high energy applications, a better modeling of the cluster emission is crucial.
This problem was addressed by applying an improved
Iwamoto-Harada model [36], and the data files for theseisotopes were upgraded by including Kunieda’s new calculations. The new model also impacts the calculatedalpha-particle energy spectra. ENDF/B-VII.1 includes
these updates, and these high energy (n,α) cross sections
were also incorporated into FENDL-3.0 at the IAEA.

IV.
A.

ACTINIDES

General comments

Before discussing each major actinide in turn, it is useful to address some general issues that apply to all the8The CIELO Collaboration . . .

NUCLEAR DATA SHEETS

existing actinide evaluations.
The major evaluated libraries (ENDF/B-VII.0,
ENDF/B-VII.1, JENDL-4.0, JEFF-3.1 etc.) predictmeasured criticality extremely well (for many assemblies,
but not all) when used in MCNP / TRIPOLI / MVP
radiation transport simulations. This has been documented by many authors [1, 2, 4, 24, 37, 38], and mostrecently in a paper by van der Marck that compares
these libraries for criticality applications [39]. However,such good performance in integral testing creates a false
sense of optimism. In particular, a more careful look at
the integral testing comparisons, and the fundamental
and evaluated cross section data, points to various
problems.

M.B. Chadwick
B.

Compensating errors

An understanding has been established by the evaluation communities, and especially by Bauge et al. at
the CEA [8], that the present fast neutron evaluationsfor 235,238 U, 239 Pu perform well in fast criticality simulations (Jezebel, Godiva etc.) — but for the wrong reasons!
That is, they embody compensating errors between theroles of fission (cross sections, average number of prompt
neutrons, and neutron spectra), capture, inelastic scattering and elastic scattering. A major challenge to ourcommunity is to strive to remove these compensating errors.
In fact, this situation has been recognized for a much
longer period of time. The WPEC (Working Party on International Nuclear Data Evaluation Co-operation) was
created in the late 1980s under OECD/NEA, and thesummary notes from one of the first meetings [40] documented priority areas to tackle. In those notes, Kikuchi
described the compensating eﬀects of inelastic scatteringand prompt fission neutron spectra on calculated criticality. This same point has been made more recently byMaslov, Ignatyuk et al. [41].
This topic was also discussed in detail in the IAEA report edited by Plompen et al. [7]. Quantitative advances
have recently been developed by Romain of the CEA
through the use of “Dalitz plots,” a technique used inhigh energy physics. This approach quantifies the relative
probabilities of three reaction mechanisms that comprisethe total nonelastic cross section below the (n, 2n) threshold — fission, inelastic, and capture — and facilitates avisual intercomparison of diﬀerent evaluated cross section
libraries. These plots indeed show that, especially for
239
Pu, ENDF/B-VII.1 and JEFF-3.1 diﬀer significantly,
yet both predict the criticality of keﬀ =1.000 accurately
for Jezebel. For 238 U, though, the plot shows that diﬀerences between ENDF/B-VII.1, JENDL-4.0, and JEFF3.1 are much smaller.

• Compensating errors
Significant compensating errors must be present in
most, and likely all, of the evaluations — see thenext subsection for details.
• Calibration
Calibration has been used in some cases in the evaluated databases to better match measured criticality of integral systems. Thus, agreement between
simulated and measured criticality keﬀ is not as impressive as it might seem, though of course a common set of evaluated data was used for all the neutronics simulations of diﬀerent critical assemblies(that is, calibration was not done on an assemblyby-assembly basis!). Where some calibration was
done it usually involved some reasonable physics
assumptions and was not entirely ad hoc, and crosssections were usually adjusted within their uncertainty levels. Such a tiny calibration may also compensate some deficiencies in the data processing andtransport calculations. The work by Iwamoto, described in the recent IAEA report [7], nicely shows(Fig. 4 in that report, p. 19) that fast neutron
assembly uncertainties in calculated keﬀ , obtainedby propagating the evaluated data covariances, aretypically of the order of 1%, whilst the measured
keﬀ have uncertainties of about 0.1 – 0.2%, yet thecalculations agree with the measured data (within
the measured uncertainties for a high fraction ofcases). This could only happen if some calibration
has been done to better match the integral experiments.

C.

Fission cross sections

ENDF/B-VII.1, which has the same major actinide
fission cross sections as ENDF/B-VII.0, uses the IAEAWPEC-CSEWG standards as described by Carlson et al.
[11], with some modifications to the unresolved resonanceregion. JEFF-3.1 and JENDL-4.0 use their own evaluation; in some cases there are diﬀerences with the standards values that lie outside the 1–σ uncertainties for the
standards, and these discrepancies should be resolved.

• Discrepancies
Plots of cross sections prepared by NNDC [17]clearly illustrate numerous discrepancies between
the various evaluations. These discrepancies reflect
both discrepancies amongst various measurements,
and diﬀerences in theory, model parameters, andcode calculations. The latter point has been documented in detail in the aforementioned IAEA report on actinide inelastic scattering and capture [7],where diﬀerences due to optical model formalisms
and parameters were explored.

D.

Prompt fission neutron spectra

Although the PFNS (prompt fission neutron spectrum)for major actinides in ENDF, JEFF, and JENDL are
not identical, they tend to be rather similar as they
all trace their origins to the Madland-Nix Los Alamos
9The CIELO Collaboration . . .

NUCLEAR DATA SHEETS

model. ENDF/B-VII.1 contains the updated MadlandNix model evaluated for ENDF/B-VII.0 [37]. The PFNS
data in JENDL-4.0 below 5 MeV are identical to thosein JENDL-3.3, evaluated by Ohsawa et al. [42], and
CCONE calculations were adopted above 5 MeV. In
JEFF-3.1, the PFNS of 235 U is the same as in ENDF/BVI, and new Madland-Nix-type model calculations performed by Vladuca and Tudora [43] were adopted for 238 U
and 239 Pu.
Major eﬀorts are underway to advance our understanding of PFNS. These involve the “Chi-nu” measurement capabilities being developed by LANL, LLNL, andthe CEA at LANSCE, theoretical development at many
places, and a wide body of work coordinated by an IAEA
CRP [44].
Most recent evaluations of 239 Pu PFNS are based on
the Madland-Nix model in its original or some revised
version that takes into account more parameters to betterfit detailed prompt neutron data, e.g., diﬀerent neutron
multiplicities from light and heavy fragments, diﬀerent
temperatures in the two fragments, anisotropy of emission of the neutrons, etc. At higher energies, JENDL4.0 includes pre-equilibrium components of the spectrum.
Significant experimental and theoretical eﬀorts are underway to shed some light on the low- and high-energy tailsof the n+239 Pu PFNS, which should help resolve some of
the remaining uncertainties.
Issues to be solved include the following.

M.B. Chadwick

suggest that the Cf results might be reliable, butthat larger uncertainties may be present in the thermal uranium dosimetry analysis owing to the extent
to which the uranium plate in the experiment truly
creates a thermal PFNS spectrum. Work beginningon a new IAEA CRP for validation of dosimetry
data will help resolve this.
• LANL released data for NUEX measured PFNS appear to corroborate ENDF/B-VII PFNS data for
235
U and 239 Pu from about 1 to 7 MeV outgoingenergies, for fast neutron incident energies [1, 47](see later in this paper, Figs. 5, 7). Are the verysmall uncertainties quoted reliable?
• How do neutrons from multi-chance fission change
the PFNS shape? Kawano et al. [48] incorporated the FKK theory into the prompt fission neutron spectrum calculations, and the CCONE calculations in JENDL-4.0 combine the pre-equilibrium
process with the Los Alamos model [49] where themulti-chance fission occurs. Owing to the preequilibrium emission, the PFNS at higher energies
is anisotropic. In general, refined theoretical treatments are needed to describe experimental PFNS
above the second-chance emission threshold for actinides.
E.

• Does the shape of the PFNS need to be higher below an MeV (say 10% higher near 0.1 MeV outgoingneutron energy) for essentially all incident energies
and all actinides, as suggested by some experimental data sets and by Maslov? At the present time
this remains an open question, needing more experimental work (to ensure that erroneous multiplescattering backgrounds are not biasing the data)
and more theoretical analyses.

Prompt fission gamma-ray spectra

Evaluated prompt fission gamma-ray spectra (PFGS)
tend to be based on very old measurements, and updates are needed to reflect new measurements, including those made recently at LANSCE/DANCE.
New reactor design work has identified uncertainty
requirements of 7.5% for gamma-ray heating in peripheral non-fueled zones, yet current discrepancies in
C/E for integral benchmarks can be as high as 1028% (see High Priority Request List http://www.oecdnea.org/dbdata/hprl/hprlview.pl?ID=422 (and 421)).
New evaluations should be based on all available experimental information, as well as on model calculations suchas CGMF (Los Alamos) [50–52] , FREYA (LLNL-LBNL)[53, 54], and GEF (Schmidt) [55].

• If the above proposal is adopted, this will impact
thermal system 235 U and 239 Pu criticality (e.g., theaverage energy of 235 U PFNS at thermal might decrease from 2.03 to 1.99 MeV). Can good thermal
criticality performance be re-established via othernuclear data changes?
• Dosimetry threshold data in fast LANL criticalassemblies (Jezebel and Pu-Flattop; Godiva andFlattop-25) may provide guidance on the shape ofthe high emission energy PFNS for fast incident energies [45] (see later in this paper, Figs. 5, 7). Additional work is needed to quantify the uncertainties
in this analysis.

F.

Inelastic scattering

There are very significant diﬀerences amongst the various evaluations. For example, for the total inelastic
scattering cross section of 239 Pu, ENDF/B-VII.1 andJENDL-4.0 are close in the 0.5–3 MeV region. However
JEFF-3.1 is significantly lower. For 238 U, the various
evaluations are reasonably close between 0.7 and 3 MeV.
However ENDF/B-VII.1 is notably higher in the 0.2–0.6
MeV range (see the useful review by Plompen et al. [7]).
Since inelastic scattering to the low-lying states of actinides is very diﬃcult to separate experimentally from• Are we confident that Zolotarev’s [46] evaluationsof dosimetry cross section activations in 235 U PFNS
at thermal energy, and 252 Cf spontaneous fission
neutron spectra, point to the correctness from 1
to 10 MeV outgoing neutron energies of the existing ENDF/B-VII.1 evaluations? Initial discussions
10

The CIELO Collaboration . . .

NUCLEAR DATA SHEETS

elastic scattering, the evaluation of inelastic scatteringcross sections relies on model calculations, where many
uncertain factors are involved, such as the optical potential parameters, nuclear deformation, width fluctuation
corrections, fission competition, etc. It is essential to understand diﬀerences in the statistical model calculations
and the model parameters adopted by various groups.
Diﬀerences in partitioning the total incoming flux into
the elastic and inelastic channels may also change thescattering angular distributions.
Discrepancies exist between evaluated and measured 14
MeV incident neutron double-diﬀerential neutron emission data [56, 57] for 238 U, especially for neutron energies above 10 MeV where a significant underestimationis seen near Q = −2 MeV. One might attribute thisto unknown collective strength that increases the inelastic scattering to continuum states, and it was handled
in ENDF/B-VII.0 [58] and JEFF-3.1 using an empiricalapproach with pseudo-levels to represent data. Insights
were obtained not only from Baba’s 238 U data [56, 57]
but also from similar 235,238 U data from Kammerdiener[59], and from analyzing LLNL 14 MeV pulsed spheremeasurements [45]. The pseudo-level was also adopted
in JENDL-3.2 by Kawano et al. [60] in order to better
fit such data. However, these pseudo-levels were removed
in JENDL-3.3 [61] as they are less justifiable from a fundamental theoretical view. A recent MSD (multi-step
direct) calculation based on the quasi-particle random
phase approximation by Dupuis et al. [62] reproduced thespectra at 90◦ and 120◦ without any adjustable parameters, although disagreement still remains at 30◦ . Also,a MSD treatment by Wienke et al. [63] emphasizing theimportance of deformations and the collective structureof actinide targets provides an alternative approach. Ongoing work at Livermore is also focused on the theoreticalunderstanding of such processes through the calculationof transition densities to QRPA excited states.
Since neutron transport properties (such as room returns, multiple scattering, etc.), as well as the underlying
cross sections, are required to model neutron transmission measurements, a better modeling of the experiments
is also essential to resolve this problem [64]. In addition,
PFNS representations that contain a pre-equilibrium prefission angular component that is anisotropic should be
implemented in all evaluations.

G.

M.B. Chadwick

tion can sometimes be inferred reliably from simultaneous
fits to the total and partial cross sections.

H.
1.

235

U

Resolved resonance parameters

All the existing libraries adopted the resolved resonance parameters for 235 U from ORNL [65]. The upper energy boundary of the resolved region, which is
2.25 keV, was lowered to 500 eV in JENDL-4.0 andpoint-wise cross sections are provided in the 500 eV
to 2.25 keV energy range based on the previous resolved resonance parameter set from ORNL [66]. Hencethe cross sections below 500 eV in all the libraries
are identical. New work on 235 U is needed to solve
a number of open problems, as discussed below (seealso the High Priority Request List at http://www.oecdnea.org/dbdata/hprl/hprlview.pl?ID=430).

2.

Radiative capture

The capture cross section in the resolved resonance region is given by the resonance parameters from ORNL
[65]. In the fast energy range, although unresolved resonance parameters were also provided by ORNL, the evaluations are based on available experimental data. The
evaluation in ENDF/B-VII.1 is based on an analysis ofmeasured capture and α-value (capture to fission ratio) data above 2.25 keV. ROSFOND adopted ENDF/BVII.1. JEFF-3.1 adopted ENDF/B-VI.8, which resulted
in very similar cross sections above 25 keV.
Questions associated with the 235 U capture cross section should be resolved, especially in the 0.5–2 keV energy
range where the recent JENDL-4.0 evaluation lowered thecross section by more than 25% based on integral reactor (sodium void) testing [67]. Also, their analysis led toincreased capture compared to ENDF and JEFF in the
3–5 keV region, and for 100–1000 keV. New data fromLANSCE (LANL) by Jandel et al. [68] with the DANCE
detector partly support the Japanese conjecture (from 1
to 5 keV), but also suggest that a ∼10% capture increaseis needed from 10 to 70 keV for all evaluations, as shownin Fig. 4. Recent measurements by Danon et al. at RPI
also support a lower capture cross section in the 100 eV–2 keV region.
Other more integral measurements of capture also
point to work needed to resolve discrepancies. Measured PROFIL [69] fast reactor measurements of 235 U
capture are 3–5% above MCNP simulations that use the
ENDF/B-VII.1 capture cross sections [24]. Yet the recent accelerator mass spectrometry data by Wallner et
al. would instead suggest the capture data in ENDF/BVII.1 should be 5–8% lower at 25 keV and 423 keV [1].

Radiative capture

In general it is diﬃcult to measure capture cross sections for fissile actinides, owing to the diﬃculty in separating the capture gamma-rays from the prompt fission gamma-rays, though recent progress has been madeat LANSCE (LANL), Rensselaer (RPI), and n-TOF(CERN) to solve this diﬃculty. The activation method
avoids this diﬃculty, but typically this is only possible
for limited incident energies (e.g. thermal) and quasimonoenergetic neutron sources. The capture cross sec11

σ(n,γ) [barns]

The CIELO Collaboration . . .

Jandel, 2012
Corvi, 1982
Kononov, 1975
Hopkins, 1962
Muradyan, 1980
Wallner, 2011
Jandel, 2012 <σ(n,γ)> @ 25 keV
ENDF/B-VII.1
JENDL-4

Ref-DANCE [%]

50
25
0

4.

-25
-3
10

-2

ENDF/B-VII.1, JEFF-3.1 (same as ENDF/B-VI), andJENDL-4.0 are in fair agreement, and are consistent withdata of Frehaut [72], as well as the measurements from
Younes et al. using the GEANIE detector at LANSCE together with model calculation extrapolations. Note that
CEA reported a normalization factor that should be applied to the data of Frehaut for all Gd-tank measurementsof ν, (n, 2n) and (n, 3n), and Vonach suggested a valueof 1.078 for this normalization [73].

Gwin, 1976
Bolotskij, 1973
De Saussure, 1962
Vorotnikov, 1971
Dvukhsherstnov, 1975
Corvi, 1982
Jandel, 2012
Diven, 1958
Hopkins, 1962
ENDF/B-VII.1 (640 gps)
JENDL-4 (640 gps)

0.8

0.6

0.4

5.

0.2

0.01
0.1
Incident Neutron Energy (MeV)

1

FIG. 4. 235 U capture (upper-panel) and alpha (capture to
fission ratio, lower panel), showing recent LANL (DANCE)
measurements by Jandel et al.

3.

Average number of neutrons per fission ν

A common method of evaluating ν is a direct fittingof a simple functional form to experimental data available, taking into account uncertainties associated withthe data. The ENDF/B-VII.1 ν evaluation comes from a
covariance analysis of the measured data, though in thefast region the evaluation was modified slightly (within
the uncertainties) to optimize a match to the fast criticalassembly benchmarks [58]. The evaluations of prompt
and total fission multiplicity, ν p , ν t = ν p + ν d , appear to be very similar, although JENDL-4.0 lies under
ENDF/B-VII.1 in the 10–15 MeV region, probably due tothe energy grid given in the file (JENDL-4.0 gives only
five energy points above 5 MeV, whilst the ENDF/BVII.1 energy grid is denser).
The total thermal ν value in ENDF/B-VII.1 (2.4367)
diﬀers slightly from the evaluated constant by the standards group (2.4355) owing to the calibration to thermal
reactor benchmarks. JENDL-4.0 has a very similar value(2.4363) which is based on the experimental data of Gwin
et al. [75, 76].

0
0.001

(n,2n) reaction

1

-1

10
10
Incident Neutron Energy [MeV]1
U Capture-To-Fission Ratio α

M.B. Chadwick

not only in the total inelastic cross section but also inthe partial cross sections to individual levels and theirangular distributions. Experts in coupled channel andHauser-Feshbach calculations need to resolve these differences.
We note that MacFarlane has suggested that the inelastic scattering for incident energies ∼0–4 MeV in 235 U
might need to be smaller, based on the systematic undercalculation (by a few percent) of spectral indices 238 U
fission/235 U fission (fission ratio) in Godiva and Flattop25 — less inelastic scattering would lead to a hotter neutron spectrum in the assembly, and consequently a higher
238
U fission/235 U fission ratio. This needs to be studied,along with the role of the PFNS shape on such spectral
indices.

1

0.1

235

NUCLEAR DATA SHEETS

Inelastic scattering

From threshold to a few MeV there are significant differences between JENDL-4 and ENDF/B-VII.1 (JEFF3.1 is the same as ENDF/B-VI) up to 50 keV, exhibiting
diﬀerences in the treatment of the low-lying rotationalband built upon the isomeric state. All evaluations lie significantly higher than some of the measurements in the 1–
2 MeV range (but we note that the direct measurement of
total inelastic scattering has a large uncertainty). ROSFOND takes ENDF/B-VII.1, with some modifications to
individual levels. JENDL-4.0’s independent analysis appears similar to ENDF/B-VII.1 from 50 keV up to a few
MeV, but then diﬀerences are seen at higher energies.
These diﬀerences have a large impact on the fast criticality (Godiva), leading to a 540 pcm swing in calculated
criticality, as shown by Go Chiba [71]. Diﬀerences exist

6.

PFNS integral validation

Los Alamos has been exploring the use of dosimetry(n, 2n) detectors in fast critical assemblies to provide
12

The CIELO Collaboration . . .

NEUX-U
237Np(n,f)
238U(n,2n)
169Tm(n,2n)
204Pb(n,2n)
32S(n,p)1.25

239Pu(n,f)
238U(n,f)
238U(n,2n)
191Ir(n,2n)75As(n,2n)51V(n,ɲͿ

239Pu(n,f)
238U(n,f)
203Tl(n,2n)
197Au(n,2n)
90Zr(n,2n)
C-NR (8U(n,2n))

1.00

0.75

"Closed" data points are from Godiva, "open" data points are from Flattop-25
0.50
0.0

2.5

5.0

7.5

10.0

12.5

M.B. Chadwick

TABLE I. Calculated and measured reaction rates (in ratio to 235 U(n, f )) for various threshold reactions in the central region of fast Los Alamos critical assemblies. The data
for the uranium-235 assemblies (Godiva and Flattop-25) arepresented in Fig. 5, whereas data for the 239 Pu assemblies(Jezebel and Flattop-Pu) are presented in Fig. 7. See the Appendix for more details on these data. Calculations come fromMCNP5 simulations using ENDF/B-VII.1 data, and include
cross section uncertainties where available, but not contributions from the PFNS spectra uncertainties. The measured
data given are averages of multiple measurements if morethan one exist, and uncertainties include a random component obtained from replicate experiments (3%) and a systematic component that comes from the 14.1 MeV cross section
uncertainty in the calibration process (see text and Table IV).

Selected Spectral Index Data for the Central Region of Godiva
or Flattop-25 (with ENDF/B-VII.1 Cross Sections)1.50

Calculated / Measured

NUCLEAR DATA SHEETS

15.0

Average Reaction Energy (MeV)

Assembly/
Reaction
GODIVA
238
U(n, 2n)
Flattop-25
238
U(n, 2n)203
Tl(n, 2n)169
Tm(n, 2n)191
Ir(n, 2n)197
Au(n, 2n)204
Pb(n, 2n)75
As(n, 2n)90Zr(n, 2n)32
S(n, p)51
V(n, α)Jezebel238
U(n, 2n)169
Tm(n, 2n)191
Ir(n, 2n)
Flattop-Pu
238
U(n, 2n)203
Tl(n, 2n)169
Tm(n, 2n)191
Ir(n, 2n)

FIG. 5. Integral validation testing of dosimetry threshold reactions in uranium fast critical assemblies, providing a testof the 235 U prompt fission neutron spectrum for higher neutron emission energies. The x-axis shows the average incident
neutron energy at which the reaction occurs. Deviations from
C/E=1 could point to (n, 2n) cross section deficiencies as well
as PFNS deficiencies. The uncertainties shown include bothmeasurement uncertainties and calculation uncertainties fromthe evaluated cross sections but do not include uncertainties
in the evaluated PFNS spectrum. Numerical values for thedata and calculations are given in Table I. Uranium NUEXPFNS data that have been recently released by Los Alamos
are also shown [47].

insights into the shape of the high-energy prompt fission neutron spectrum. The MCNP code can simulate
the neutron transport in fast assemblies, such as Jezebel(a Pu sphere), Flattop-25, and Flattop-Pu (spheres of235
U and 239 Pu surrounded by a 238 U tamper), using
ENDF/B-VII.1 cross sections. We expect the MCNPpredicted high energy neutron spectrum above a few MeV
to be highly dependent on the shape of the underlying
PFNS in the ENDF/B-VII.1 database. Discrepancies
between predictions and measurements of the dosimetry(n, 2n) rates point to possible deficiencies in the evaluated
PFNS. By using activation cross sections from (n, 2n)dosimetry detectors to characterize the high-energy spectrum, we can seek to validate the evaluated PFNS. Thisapproach depends upon having accurate (n, 2n) cross sections for the dosimetry reactions, and therefore (a) we use
dosimetry materials for whose (n, 2n) cross sections havebeen measured by many groups and have relatively small
uncertainties and (b) we use as many detectors as we canso that individual detector (n, 2n) cross section biases arelikely to cancel out.
The appendix summarizes the historical Los Alamos
fast critical assembly measurements that we use for thisanalysis, and relevant averages of these data (where multiple measurements were made) are presented in Table I
along with our MCNP5 calculations. These measured

Calculation

Experiment

7.84E-3 ± 13%7.73E-3 ± 4%

7.08E-3 ± 13%
1.70E-3 ± 13%
1.78E-3 ± 3%
1.76E-3 ± 12%
1.43E-3 ± 5%
1.66E-5 ± 12%
1.43E-4 ± 7%
4.40E-5 ± 3%
3.11E-2
1.18E-5 ± 12%6.94E-3 ± 4%
1.52E-3 ± 6%
1.47E-3 ± 4%
1.89E-3 ± 4%
1.61E-3 ± 3%
1.64E-5 ± 6%
1.50E-4 ± 6%4.85E-5 ± 3%
3.06E-2 ± 6%
1.11E-5 ± 6%

1.32E-2 ± 12%3.72E-3 ± 3%
3.76E-3 ± 11%

1.06E-2 ± 4%
3.13E-3 ± 3%
3.21E-3 ± 4%

1.10E-2 ± 12%2.92E-3 ± 13%
3.05E-3 ± 3%
3.08E-3 ± 11%9.02E-3 ± 4%2.27E-3 ± 6%
2.43E-3 ± 3%
2.83E-3 ± 4%

data have been reassessed and updated to modern calibration standards. Fig. 5 shows our results for uranium
critical systems. The (n, 2n) and (n, f ) reactions are mostsensitive to the 235 U nuclear data, with the (n, 2n) reactions being particularly sensitive to the PFNS. Although
there are some inconsistencies between the feedback fromthe diﬀerent detectors, especially from Zr and Tm, overall the validation comparisons shown in Fig. 5 indicatethat the high-energy PFNS spectrum (for fast incident
neutrons) appears to be reasonable, and there is no clear
indication that the high-energy tail should be higher orlower. We note that this conclusion diﬀers from our earlier results which were based on a smaller number of detectors, and on dosimetry activation data before the calibration basis was updated (our earlier studies [45] suggested the need for a softer 235 U PFNS above 8 MeV
13The CIELO Collaboration . . .

NUCLEAR DATA SHEETS

for fast incident neutrons). The principal change is thatour 191 Ir(n, 2n) measured data has been reassessed to belarger than before (see Appendix). Our present conclusion also appears to diﬀer from preliminary CEA results[77] which had also suggested the need for a softer 235 U
PFNS, from dosimetry feedback from a fast critical assembly in France. Future work is needed to understandthe LANL and the CEA diﬀerent results, and indeed a
new IAEA CRP on dosimetry data validation will address
this subject.
Later, in Section IV J 6, we provide analogous information on testing the 239 Pu PFNS for fast incident neutrons.

I.
1.

238

M.B. Chadwick
2.

Elastic and inelastic scattering

As was the case for 235 U, there are significant diﬀerences between the diﬀerent cross section libraries, bothfor elastic and inelastic cross sections and for angular
distributions. Cross section diﬀerences are evident in the
total inelastic as well as in the partial cross sections to
individual levels. This can lead to significant impacts
on calculated criticality [71] for fast systems. Indeed,feedback from integral reactor sensitivity studies in theUSA, Japan, Europe, and WPEC Subgroup 26, 33 haveconsistently pointed to the need for more accurate 238 U
inelastic cross sections and angular distributions [79–82](see also High Priority Request List at http://www.oecdnea.org/dbdata/hprl/hprlview.pl?ID=435).
Recent “quasi-diﬀerential” measurements from RPI
that are sensitive to elastic and inelastic scattering in238
U provide useful guidance on how to improve the scattering cross sections. These data are from 0.5 up to
20 MeV incident energies. ENDF/B-VII does not include anisotropic angular distributions in compound elastic scattering (JENDL has this and does better at backward angles). At forward angles the libraries performbetter than at backward-angles, where clear deficiencies
can be seen. For ENDF/B-VII the problem seems to
be in its angular distribution representations, while for
JEFF-3.1 the problem seems to be in its cross sections(JEFF-3.1 has an anisotropic elastic compound component). Also, ENDF/B-VI appears to perform better than
ENDF/B-VII.1. The VII.1 (=VII.0) results were basedon a pragmatic choice a decade ago by Young, MacFarlane and Chadwick, using Maslov’s optical model that
led to good performance in critical assembly simulations,but improvements may now be needed. The IAEA has
ongoing work (Capote, Trkov) on calculations and benchmarks based on a new dispersive coupled-channel potential that couples almost all levels below 1 MeV. A developmental test evaluated file has already been madeand tested with encouraging results against a suite of
twelve ICSBEP benchmarks that are especially sensitive
to elastic scattering. Work at Livermore and Los Alamos
on coupled-channel scattering for the actinides will also
be relevant to this project, as Thompson, Dietrich, andKawano showed that great care is needed to ensure thatconvergence is achieved in the scattering calculation.

U

Radiative capture

The resolved resonance parameters evaluated by
ORNL and the CEA [78] are given up to 20 keV
in ENDF/B-VII.1, JEFF-3.1, JENDL-4.0, and ROSFOND. CENDL-3.1 adopted the resonance parameters in
JENDL-3.3. Above the resolved resonance region, there
is now better consensus amongst the various libraries, together with consistency with the IAEA-WPEC-CSEWG
standards result. The evaluated uncertainties are rather
small — 2% in some cases in the fast energy range. The
recent AMS data by Wallner et al. at 25 and 426 keV
support these evaluations (see Ref. [1] p. 2970).
The international communities’ success in determining this important cross section accurately, through both
WPEC Subgroup 4, 22 and through the IAEA-WPECCSEWG standards eﬀort, is one of the notable accomplishments in recent decades. However, there is one issue that needs to be resolved. The ENDF/B-VII evaluation made some small modifications to the standards
capture result: for example, in the 20 – 100 keV region, avalue that is a few percent above the standards value wasadopted, motivated by the shape predicted by GNASH
model calculations [70] as well as by improved MCNP
simulation performance for the Bigten critical assembly.
But as this small modification exceeded the uncertainties
of the standards evaluation, future work is needed to resolve these diﬀerences. Moreover, the nuclear data uncertainty assessment performed by WPEC Subgroup 26 forinnovative reactor systems shows that the uncertainty inthe radiative capture cross section of 238 U should be further reduced to 1-3% in the energy range from 20 eV to 25
keV (see High Priority Request List at http://www.oecdnea.org/dbdata/hprl/hprlview.pl?ID=456). The recent
ENDF/B-VII.1 evaluation of this cross section uncertainty (3-4% in this energy range) suggests that additional measurements in the resonance regions are stillneeded. Indeed, new measurements are being planned
by Cano Ott et al. in Europe, at the CERN n TOF facility and at IRMM GELINA, with the goal of further
improving our understanding of the 238 U capture crosssection from a few eV to several hundred keV [7].

3.

(n, 2n) reaction

There is fair agreement among the libraries, although
ENDF/B-VII.1 rises from a threshold about 10% higher
than JENDL-4.0 in the 7–9 MeV region. This was motivated in ENDF by a desire to match the LANL Knight
data [83]. Most data appear to lie below Knight though
the data of Kornilov et al. [84] are also higher in thethreshold region and close to ENDF/B-VII.1. It will be
important to resolve such (n, 2n) threshold discrepancies,
because the 238 U(n, 2n) reaction is used as a dosimetry
14

The CIELO Collaboration . . .

NUCLEAR DATA SHEETS

M.B. Chadwick
J.

3
Frehaut (1969)
Nurpeisov (1975)
Malynovskyj (1983)
other
JENDL-4.0
ENDF/B-VII.1
JEFF-3.1
CENDL-3.1

2.9
2.8

1.

νp

2.6
2.5
2.4
2.3
2.2
1

2
3
Neutron Energy [MeV]4

FIG. 6. Comparison of the evaluated ν p for 238 U with experimental data.

reaction in assessments of the higher-energy tail of theprompt fission spectrum, a topic of current intense study.
Recent (n, 2n) measurements by Zhu et al. [85] agree
well with ENDF/B-VII.1. For example, the experimental
value at 14.1 MeV is 858 ± 34 mb, which confirms the
ENDF value at 14.1 MeV of 849 mb, while CENDL-3.1
has 871 mb, JEFF-3.1 has 849 mb, and JENDL-4.0 has
943 mb.

Resolved resonance parameters

2.

4.

Pu

Derrien et al. of ORNL performed resolved resonanceanalysis in the energy range 0 – 2.5 keV [87]. JENDL4.0 adopted these same resonance parameters. JEFF3.1.1 includes some modifications to these resolved resonances for a better match to some integral keﬀ data.
These modification were motivated by the problem of an
over-prediction of calculated keﬀ in 239 Pu solution critical assemblies [1, 24] (see also the High Priority Request List and references therein at http://www.oecdnea.org/dbdata/hprl/hprlview.pl?ID=427). WPEC Subgroup 34 “Coordinated evaluation of 239 Pu in the resonance region” was formed to help address these issues.
As an example of the magnitude of this problem, inreflected critical assemblies with high concentrations of239
Pu within a thermal neutron spectrum (e.g. someof the PU-SOL-THERM benchmarks) there is an overall
trend in C/E of keﬀ of 1.005 – 1.01, using ENDF/B-VII.1(=VII.0) data [24].
An improved modeling of plutonium solution criticalassemblies is likely dependent not only upon resonanceparameter analyses but also on the values for ν and the
PFNS shape, as reported by Subgroup 34. The goal for
Subgroup 34 is to improve our modeling of such integrated criticality experiments whilst advancing the understanding, and evaluation, of the underlying nuclear
data.
It is also possible that angular distributions from resonance fission neutrons are needed for high-fidelity simulations of criticality.

2.7

0

239

Radiative capture

The nuclear data uncertainty assessment performed by
WPEC Subgroup 26 for innovative reactor systems showsthat the knowledge of the radiative capture cross sectionof 239 Pu should be improved to meet the target accuracy
requirements for these systems. New measurements andevaluations have been requested from 2 keV up to about
1.5 MeV to reduce the uncertainty down to 3-7% depending on the energy region and the reactor considered(see the High Priority Request List at http://www.oecdnea.org/dbdata/hprl/hprlview.pl?ID=451).
There are significant (on the order of 10%) diﬀerencesamongst the evaluations in the fast region, and above
0.6 MeV ENDF/B-VII.1 falls well below the Los Alamos
data of Hopkins and Diven [88]. PROFIL [69] fast reactor
data analyses suggest ENDF/B-VII.1 may be on average
10% too low in the fast region, while JEFF-3.1 performs
better here. These PROFIL data may be the only integral data on 239 Pu capture in the fast range as neither
Los Alamos nor Japan (we understand) have their own
integral measurements (such measurements can be diﬃcult owing to the ingoing 240 Pu often present in 239 Pu

Average number of neutrons per fission ν

There appear to be some significant diﬀerences in theevaluations below 4 MeV that should be resolved. As seenin Fig. 6, ENDF/B-VII.1 and CENDL-3.1 have a kink at
2.9 MeV, while JENDL-4.0 and JEFF-3.1 have a linear
extrapolation toward zero energy. These representations
coincide at 2 MeV but large diﬀerences are evident at
3 MeV. Since the average neutron energy causing fission
in 238 U in critical assemblies such as Bigten is about 3
MeV (owing to the ∼ 1 MeV fission threshold), as shownin the Appendix of Selby’s paper [86], this discrepancy
impacts criticality calculations and should be resolved.
The linear interpolation toward zero energy defined in
ENDF/B-VII.1 results in too high a value for ν at thethermal energy. This was also supported by the MadlandNix model calculation, which gives lower ν, similar to
JENDL-4.0 and JEFF-3.1 values. But because the 238 U
fission cross section at thermal energies is of the order ofµb, this does not have an impact on practical calculations.
15

The CIELO Collaboration . . .

NUCLEAR DATA SHEETS

samples). It is striking that in the important 200 keV - 1
MeV region there is only one measurement, that of Hopkins and Diven from Los Alamos from the early 1960s!
ENDF/B in particular needs to be updated. Future Los
Alamos DANCE experiments that build on the recent
detector advances made for 235 U capture hope to better
determine the cross section. These experiments should
also strive to reach higher accuracies in the hundreds ofkeV region as well as at lower energies.
Also, nuclear reaction theory is especially importantfor evaluating the capture cross section because we haveso little data above 100 keV. Such calculations must treat
the fission competition channel, and associated channel
width fluctuations, accurately. The role of (n,γf) processes need to be carefully considered.

3.

adopted the data in JENDL-3.3, and ROSFOND took
the JEFF-3.1 file. ENDF/B-VII.1 applied a small modification to match Jezebel keﬀ . In one region the tweak lay
outside the 1-σ error bar. This tweak should be removed
in the future. Also, recent work by WPEC Subgroup 34
may lead to an updated recommendation on the 239 Pu ν
at lower energies.
The total thermal ν value in ENDF/B-VII.1 (2.879)
diﬀers slightly from the evaluated constant by the standards group (2.884 ± 0.16%) [37]. JEFF-3.1, and ROSFOND have 2.8778 (ROSFOND is the same as JEFF3.1), JENDL-4.0 has 2.8786, and CENDL-3.1 adopted the
JENDL-3.3 value, which is 2.8843. Given the well-known
over-prediction of keﬀ in thermal plutonium critical assemblies, there is a need to study the limits of how low
an evaluated thermal ν could be. The evaluated values
in ENDF/B-VII.1, JEFF-3.1, and JENDL-4.0 are already1-σ below the standards evaluation. Does the standards
constant need re-evaluating? We note that the optimized
search approach by Rochman and Koning [93] was able
to match thermal solution Pu criticality better, albeit
at the cost of using a thermal ν value that is 0.3% below
ENDF/B-VII.1, which is 3 standard deviations below thestandard.

Inelastic scattering

As discussed earlier [7], there are significant diﬀerencesamongst the evaluations for 239 Pu in the fast energy
range, with ENDF/B-VII.1 and JENDL-4.0 lying significantly above the JEFF-3.1 evaluation. Data reported
by Batchelor and Wyld [89] and Andreev [90] suggest asmaller total inelastic scattering for 239 Pu, although thesedata were not measured directly. Maslov and Ignatyuk
have also suggested that ENDF/B-VII.1 and JENDL-4.0
inelastic scattering cross sections in the fast range are
too large, and they postulate that they compensate an
ENDF/B-VII.1 PFNS spectrum that is too low below 1
MeV. Ideally, new accurate measurements would be performed to better determine inelastic scattering, but weare not aware of any immediate prospects for this. Workis needed by experts in coupled channels and HauserFeshbach reaction theory to better understand plutonium
scattering reactions.

4.

6.

PFNS integral validation

See Section IV H 6 for a summary of our work to test
the PFNS using Los Alamos fast critical assemblies. Here
we provide our results for the 239 Pu PFNS based on
historical dosimetry activation measurements made inJezebel and Flattop-Pu.
We expect the MCNP-predicted high-energy neutron
spectrum above a few MeV to be highly dependent on
the shape of the underlying PFNS in the ENDF/B-VII.1
database; likewise, discrepancies between predictions andmeasurements of the high energy spectrum point to possible deficiencies in the evaluated PFNS.
The appendix summarizes the historical Los Alamos
fast critical assembly measurements that we use for thisanalysis, and relevant averages of these data (where multiple measurements were made) are presented in Table I
along with our MCNP5 calculations. These data havebeen reassessed and updated to be based on modern calibration standards. Fig. 7 shows our results for plutonium
fast critical systems. The (n, 2n) and (n, f ) reactions aremost sensitive to the 239 Pu nuclear data, with the (n, 2n)reactions being particularly sensitive to the PFNS. These
data suggest that for outgoing neutron energies above 10
MeV the ENDF/B-VII.1 239 Pu PFNS may need to besofter, as we suggested in Ref. [45]. Therefore, unlike
the case for 235 U PFNS discussed in Section IV H 6, theprocess of updating our dosimetry measured values andadding additional materials (thallium in this case) has notchanged our earlier conclusions. But this kind of dosimetry testing for the PFNS should still be treated cautiously
because we have fewer data for plutonium assemblies than(n, 2n) reaction

ENDF/B-VII.1 and JENDL-4.0 are in fair agreement,
and match the GEANIE data [91] and the activation dataof Lougheed et al. [92] near 14 MeV. JEFF-3.1 rises
more rapidly from threshold, and lies below the data ofLougheed et al. at 14 MeV [92]. PROFIL measurements[69] tend to suggest a smaller (n, 2n) cross section, at leastin its rise from threshold [24]. However, other integral
measurements support the ENDF/B-VII.1 239 Pu(n, 2n)evaluation as shown in Fig. 7 (the larger error bar on the
239
Pu(n, 2n) data point in the figure comes from our large
uncertainty in the evaluated (n, 2n) cross section near its
threshold, not measurement uncertainties).

5.

M.B. Chadwick

Average number of neutrons per fission ν

Independent statistical analyses were performed for
ENDF/B-VII.1, JEFF-3.1, and JENDL-4.0. CENDL-3.1
16

The CIELO Collaboration . . .

NUEX-Pu
237Np(n,f)
238U(n,f)
203Tl(n,2n)
191Ir(n,2n)
C-NR (8U(n,2n))

1.25

239Pu(n,f)
237Np(n,f)
238U(n,2n)
169Tm(n,2n)
191Ir(n,2n)

239Pu(n,f)
238U(n,f)
238U(n,2n)
169Tm(n,2n)
C-NR (9Pu(n,2n))

1.00

0.75

"Closed" data points are from Jezebel; "open" data points are from Flattop-Pu.
0.50
0.0

2.5

5.0

7.5

10.0

12.5

15.0

Average Reaction Energy (MeV)

FIG. 7. Integral validation testing of dosimetry threshold reactions in plutonium fast critical assemblies, providing a testof the 239 Pu prompt fission neutron spectrum at higher neutron emission energies. The x-axis shows the average incident
neutron energy at which the reaction occurs. Deviations from
C/E=1 could point to (n, 2n) cross section deficiencies as well
as PFNS deficiencies. The uncertainties shown include bothmeasurement uncertainties and calculation uncertainties fromthe evaluated cross sections but do not include uncertainties
in the evaluated PFNS spectrum. Numerical values for thedata and calculations are given in Table I. Plutonium NUEXPFNS data that have been recently released by Los Alamos
are also shown [47].

for uranium assemblies and it is possible that the overcalculation in Fig. 7 is instead due to (n, 2n) cross section
deficiencies rather than to PFNS deficiencies. Indeed, the
NUEX data shown in the figure appear to be in conflictwith the dosimetry data, and instead support the existing
ENDF/B-VII PFNS evaluation up to 10 MeV. Again, a
new IAEA CRP on dosimetry data validation will address
this subject.

V.

M.B. Chadwick

since we will aim to avoid making any ad hoc evaluation decisions motivated solely by a desire to match an
integral experiment. Rather, our goal is to obtain improved integral performance whilst at the same time alsousing improved microscopic nuclear physics in our ENDFformatted CIELO files.
In addition to studying a wide range of keﬀ integral
criticality experiments that use diﬀerent neutron spectra (fast, intermediate, thermal) and diﬀerent materials(HEU, LEU, Pu, solutions, etc.), we will also include
beta-eﬀ, Rossi-alpha, reaction rate measurements (fission
ratios, (n, 2n) dosimetry responses, etc.), and shielding
and neutron transmisison experiments. Data from PROFIL and MANTRA post-irradiation experiments can provide valuable guidance for capture and (n, 2n) reactions.
Other “semi-integral” experiments that isolate certain
nuclear reaction quantities will also be studied, such as
RPI’s experiments that are sensitive to neutron scattering angular distributions, and subcritical neutron multiplication experiments that may be especially sensitive toν.
Some of the recent data adjustment investigations will
also play an important role in guiding CIELO work
[81, 82]. These eﬀorts have used recently created evaluated libraries of cross sections and covariances, and haveinvestigated feedback from a wide range of critical assemblies that explore diﬀerent materials and neutron energy
spectra. The goal of such projects is to identify possible
physically-important guidance that comes from these integral data (criticality, reaction rates, and so on), so as
to improved the underlying nuclear cross sections and remove compensating errors. The work of the NEA/WPEC
Subgroup 39 in this area will therefore be closely coordinated with our CIELO collaboration goals.

Selected Spectral Index Data for the Central Region of Jezebelor Flattop-Pu (with ENDF/B-VII.1 Cross Sections)1.50

Calculated / Measured

NUCLEAR DATA SHEETSVI.

ANALYSIS OF INTEGRAL QUANTITIESPrecise knowledge of neutron physics quantities plays
an important role in the assessment of neutron evaluations for CIELO. Their numerical values can be extracted
from the international collection of evaluated reaction libraries (ENDF/B-VII.1 [1], JEFF-3.1 [2], JENDL-4.0 [4],ROSFOND [5], CENDL-3.1 [6], and EAF-2010 activation
file [94]) and compared.
In the present work, we consider neutron fission (n, f )and capture (n, γ) reactions and several parameters thatare important for nuclear science and technology applications. Numerical values of thermal neutron cross sections (σ 2200 ), Westcott factors (gw ), resonance integrals(RI), Maxwellian-averaged cross sections, and 252 Cf spontaneous fission (sf) neutron spectrum averaged cross sections were produced in a systematic approach for nuclei of interest using the evaluated nuclear reaction data,
Doppler broadened at 293.6 K. The complete descriptionof this work for Z = 1–100 is given in Refs. [95, 96]. In
this section, we will concentrate on the analysis of 1 H,
16
O, 56 Fe, 235 U, 238 U, and 239 Pu.

VALIDATION BENCHMARKSThe CIELO project will pay close attention to integral validation testing of the new evaluations, with thegoal of creating a suite of new evaluated files that perform very well in simulations of an agreed set of validation experiments. Our goal is for validation comparisonsthat are as good, and hopefully better, than the current
suite of evaluated nuclear data sets. Van der Marck [39]recently provided an assessment of how ENDF/B-VII.1,
JENDL-4.0, and JEFF-3.1 perform in MCNP6 simulations of over 2000 criticality (keﬀ ) ICSBEP experiments,and of hundreds of neutron transmission and shielding experiments. Overall these libraries performed rather well,and we recognize that making improvements over the current libraries will be quite an accomplishment, especially
17

The CIELO Collaboration . . .

NUCLEAR DATA SHEETS

M.B. Chadwick

thermal energy region, such as the 0.29562 eV resonancein 239 Pu [28], are often responsible for Westcott factor
temperature variations.
A.

1.08

Thermal cross sections

1.06

Fission Westcott Factors

Neutron thermal cross section values for fission andcapture are shown in Fig. 8, and Tables II - III. These
values are compared with Atlas of Neutron Resonancesand Neutron Cross Section Standards evaluations [11, 28].
Further analysis of fission cross sections shows a disagreement between Atlas of Neutron Resonances [28] andENDF/B-VII.1 thermal fission cross sections for 238 U. In
this case the fission threshold is well above the thermalenergy, but sub-threshold thermal neutron fission takes
place. The experimental measurement of D’Hondt et al.
[98] (11±2 micro-barns) is a lower value than that givenin the ENDF/B-VII.1 library, indicating that more work
is needed here.

1.04

1.02

ENDF/B-VII.1
JEFF-3.1.2
JENDL-4.0
ROSFOND-2010
CENDL-3.1
EAF-2010
Atlas

1

0.98

0.96

92-U -235

92-U -238

94-Pu-239

10

Thermal Fission CS/ENDF/B-VII.1

FIG. 9. Thermal neutron fission Westcott factors.

Deviations from unity for thermal fission Westcott factors are shown in Fig. 9. For 235 U, the fission Westcott factor is in agreement with the Atlas value of
0.9771±0.0008 [28].
1

ENDF/B-VII.1
JEFF-3.1.2
JENDL-4.0
ROSFOND-2010
CENDL-3.1
EAF-2010
Atlas
Standards

C.

The epicadmium dilute resonance integral [28, 97] for aparticular reaction σR (E) in a 1/E spectrum is expressed
by
∫ ∞
dE
RI =
σR (E)
,
(1)
E
Ec

0.1
92-U -235

92-U -238

94-Pu-239

FIG. 8. Ratio of thermal neutron fission cross sections to
ENDF/B-VII.1.

where Ec is determined by cadmium cut-oﬀ energy(Ec =0.5 eV). Fig. 10, and Tables II - III show resonanceintegrals for fission, and capture. To satisfy the requirements of evaluation and research communities resonanceintegrals are calculated with the aid of Eq. (1) with an
upper energy limits of 100 keV and 20 MeV, respectively.
For capture reactions, the dominant contribution to thisintegral comes from the energy region below a few keV. In
contrast, for threshold reactions and sub-threshold fission
this is not the case, since the major contribution comesfrom the energy region above the threshold energy.
Most of the integrals reproduce the available data witha few exceptions. Table III shows evaluated library capture integral values for 16 O, and they are in fair agreement. At the same time the ENDF/B-VII.1 16 O RI valueof 7.380×10−4 b strongly disagrees with the calculated
Atlas value of 2.700×10−4 ±3.000x10−5 b. The lattervalue does not include direct capture eﬀects which wereincorporated into the ENDF/B-VII.1 evaluation following JENDL-4.0.

Thermal capture cross sections are consistent with
ENDF/B-VII.1 values with the exception of 16 O, 238 U
evaluations in EAF-2010.
B.

Resonance Integrals

Westcott Factors

The Westcott g-factor, gw , is the ratio of theMaxwellian-averaged cross section to the 2200 m/s (thermal) cross section. The gw -factor is temperature dependent and its value is close to 1 for most nuclei whereσ ∼ 1/v. Calculated Westcott factors are shown in
Fig. 9, and Tables II - III.
Complete calculation of capture and fission gw factors
reveals that most of them are close to 1 with the exceptionof non-1/υ σ(n, γ) nuclei: 113 Cd, 135 Xe, 149 Sm, 151 Eu,
176
Lu, 182 Ta, 239 Pu, 249 Bk [28]. Strong resonances in the18

The CIELO Collaboration . . .

NUCLEAR DATA SHEETS

M.B. Chadwick

TABLE II. Comparison of Evaluated Libraries Neutron Fission Integral Quantities and Experimental Benchmarks (C-calculated
value). The uncertainties on the evaluated quantities were systematically calculated for ENDF/B-VII.1 as an illustration of thecurrent knowledge of evaluated data.
Quantity

Library

ENDF Materials
U(1.680±1.048)E-5
2.651E-5
1.680E-5
1.680E-5
1.178E-5
1.178E-5
3.000E-6C

235

238

U
ENDF/B-VII.1 (5.850±0.020)E+2
JEFF-3.1.1
5.850E+2
σ 2200JENDL-4.0 5.851E+2(b)
ROSFOND 5.851E+2
CENDL-3.1 5.850E+2
EAF-2010
5.832E+2
Atlas [28](5.826±0.011)E+2
Standards [11] (5.843±0.010)E+2
ENDF/B-VII.1 9.767E-1
JEFF-3.1.1
9.767E-1
JENDL-4.0 9.765E-1
gw
ROSFOND 9.765E-1
CENDL-3.1 9.767E-1
EAF-2010
9.748E-1
Atlas [28](9.771±0.008)E-1
ENDF/B-VII.1 (2.687±0.008)E+2
100keV∫JEFF-3.1.1
2.688E+2
RI
JENDL-4.0 2.686E+2
0.5eV
ROSFOND 2.689E+2(b)
CENDL-3.1 2.688E+2
EAF-2010
2.716E+2
ENDF/B-VII.1 (2.760±0.008)E+2
20M∫ eV
JEFF-3.1.1
2.760E+2
RI
JENDL-4.0 2.757E+2
0.5eV
ROSFOND 2.762E+2(b)
CENDL-3.1 2.760E+2
EAF-2010
2.788E+2
Atlas [28](2.750±0.050)E+2
ENDF/B-VII.1 (2.204±0.013)E+0
MACS,
JEFF-3.1.1
2.213E+0
JENDL-4.0 2.169E+0
kT = 30keV
ROSFOND
2.219E+0(b)
CENDL-3.1 2.216E+0
EAF-2010
2.213E+0
ENDF/B-VII.1 (1.209±0.019)E+0σ
JEFF-3.1.1
1.203E+0252
JENDL-4.0 1.202E+0
Cf(sf)ROSFOND
1.211E+0(b)
CENDL-3.1 1.202E+0
EAF-2010
1.200E+0
EXFOR [99] (1.215±0.022)E+0

1.002E+01.001E+01.002E+01.002E+01.002E+0
1.000E+0(2.340±0.428)E-3
1.217E-3
2.349E-3
2.340E-3
1.895E-3
2.159E-3(2.054±0.012)E+0
2.059E+02.029E+02.054E+02.022E+02.024E+0(1.63±0.16)E-3C(7.988±0.336)E-5
7.173E-5
7.566E-5
7.989E-5
7.532E-5
1.206E-4(3.117±0.040)E-1
3.102E-1
3.094E-1
3.115E-1
3.097E-1
3.078E-1(3.109±0.140)E-1

It is important to note that for 238 U target nucleus,
where sub-threshold fission was observed, the calculated
fission integrals Icf reported in the Atlas of Neutron Resonances [28] correspond to subthreshold fission values.

D.

Capture RI/ENDF/B-VII.1

1.0

0.8

0.6
ENDF/B-VII.1
JEFF-3.1.2
JENDL-4.0
ROSFOND-2010
CENDL-3.1
EAF-2010
Atlas

0.2

1-H - 1Pu
(7.479±0.081)E+2
7.471E+2
7.474E+2
7.479E+2
7.471E+2
7.477E+2(7.481±0.020)E+2(7.500±0.018)E+2
1.055E+01.050E+0
1.058E+0
1.055E+0
1.054E+0
1.054E+0(1.055±0.001)E+0(2.931±0.021)E+2
2.932E+2
2.930E+2
2.928E+2
2.893E+2
2.926E+2(3.027±0.022)E+2
3.028E+2
3.026E+2
3.025E+2
2.990E+2
3.024E+2(3.030±0.100)E+2(1.822±0.006)E+0
1.809E+0
1.836E+01.808E+01.843E+01.806E+0(1.774±0.027)E+0
1.774E+0
1.777E+01.771E+01.786E+01.786E+0(1.790±0.041)E+0

Maxwellian-averaged cross sections

Maxwellian-averaged cross sections have been calculated for the temperature (kT ) of 30 keV. This temperature is the commonly accepted value in stellar nucleosynthesis s-process compilations. Analysis of calculated
MACS and experimental data provides an important toolfor ENDF quality assurance.
The neutron capture MACS are shown in Fig. 11.
There is no reliable source for fission MACS besides theevaluated nuclear library values. The stellar nucleosynthesis KADoNiS library [100] contains only neutron capture MACS for the 1 H and 210 Po range of nuclei. The
analysis of fission cross sections in Table II indicates potential issues with the JEFF-3.1 library for 238 U, where
sub-threshold fission cross section is underestimated.
A detailed analysis of Fig. 11 demonstrates the nuclear
astrophysics potential of ENDF libraries as a complementary source of evaluated cross sections and reaction rates[96]. There are noticeable diﬀerences between KADoNiS[100] and ENDF/B-VII.1 libraries for light and medium
nuclei. The 1 H deviation is due to diﬀerences between
center of mass (ENDF calculated) and lab reference system (KADoNiS) cross section values, and therefore, thisis not a discrepancy. Owing to a lack of resonances in the

1.2

0.4

239

8-O - 16 26-Fe- 56 92-U -235 92-U -238 94-Pu-239

FIG. 10. Ratio of neutron capture resonance integrals to
ENDF/B-VII.1.

19

The CIELO Collaboration . . .

NUCLEAR DATA SHEETS

M.B. Chadwick

TABLE III. Comparison of Evaluated Libraries Neutron Capture Integral Quantities and Experimental Benchmarks (Rrelative measurement, C-calculated value). The uncertainties on the evaluated quantities were systematically calculated for
ENDF/B-VII.1 as an illustration of the current knowledge of evaluated data.
Quantity

Library

1H(3.320±0.085)E-1
3.320E-1
3.320E-1
3.320E-1
3.320E-1
3.320E-1(3.326±0.007)E-1

ENDF/B-VII.1
JEFF-3.1.1
JENDL-4.0
ROSFONDCENDL-3.1
EAF-2010
Atlas [28]Standards [11]ENDF/B-VII.1
1.001E+0
JEFF-3.1.1
1.001E+0
JENDL-4.0
1.001E+0
gw
ROSFOND
1.001E+0
CENDL-3.1
1.001E+0
EAF-2010
9.995E-1
Atlas [28]
ENDF/B-VII.1 (1.490±0.038)E-1
100keV∫JEFF-3.1.1
1.490E-1
RI
JENDL-4.0
1.490E-1
0.5eV
ROSFOND
1.490E-1(b)
CENDL-3.1
1.512E-1
EAF-2010
1.489E-1
ENDF/B-VII.1(1.492±0.038)E-1
20M∫ eV
JEFF-3.1.1
1.492E-1
RI
JENDL-4.0
1.492E-1
0.5eV
ROSFOND
1.492E-1(b)
CENDL-3.1
1.515E-1
EAF-2010
1.491E-1
Atlas [28]
ENDF/B-VII.1 (1.525±0.059)E-4
MACS,
JEFF-3.1.1
1.524E-4
JENDL-4.0
1.525E-4
kT = 30keV
ROSFOND
1.525E-4(b)
CENDL-3.1
1.525E-4
EAF-2010
1.523E-4
KADONIS [100] (2.540±0.200)E-4
ENDF/B-VII.1 (3.872±1.452)E-5
σ
JEFF-3.1.1
3.865E-5
252
JENDL-4.0
3.871E-5
Cf(sf)ROSFOND
3.871E-5(b)
CENDL-3.1
3.764E-5
EAF-2010
3.864E-5
EXFOR [99] (7.418±0.526)E-5R
σ 2200(b)

ENDF Materials
56
235
238
239
O
Fe
U
U
Pu
(1.900±0.190)E-4 (2.589±0.140)E+0 (9.869±0.162)E+1 (2.863±0.053)E+0 (2.707±0.041)E+2
1.900E-4
2.586E+09.869E+1
2.684E+02.728E+2
1.899E-4
2.591E+09.871E+1
2.683E+02.715E+2
1.899E-4
2.591E+09.871E+1
2.683E+02.707E+2
1.900E-4
2.589E+09.869E+1
2.718E+02.704E+2
2.020E-4
2.591E+09.895E+1
2.720E+02.705E+2(1.900±0.200)E-4 (2.590±0.140)E+0 (9.880±0.080)E+1 (2.680±0.019)E+0 (2.693±0.029)E+2(9.940±0.072)E+1 (2.677±0.013)E+0 (2.715±0.021)E+2
1.001E+0
1.000E+09.913E-1
1.002E+01.146E+01.001E+0
1.000E+09.913E-1
1.002E+01.146E+01.001E+01.001E+09.911E-1
1.002E+0
1.142E+01.001E+01.001E+09.911E-1
1.002E+01.146E+01.001E+0
1.000E+09.913E-1
1.003E+01.146E+09.996E-1
9.987E-1
9.833E-1
1.001E+01.144E+0(9.956±0.016)E-1(1.0009E+0C )1.130E+0(1.525±0.153)E-4 (1.329±0.133)E+0 (1.398±0.015)E+2 (2.752±0.040)E+2 (1.811±0.016)E+2
8.538E-5
1.327E+01.398E+2
2.750E+2
1.811E+2
1.588E-4
1.330E+01.383E+2
2.752E+2
1.793E+2
1.588E-4
1.330E+01.398E+2
2.752E+2
1.811E+2
8.538E-5
1.329E+01.398E+2
2.777E+2
1.846E+2
1.724E-4
1.328E+01.320E+2
2.777E+2
1.841E+2(7.380±1.658)E-4 (1.346±0.134)E+0 (1.404±0.016)E+2 (2.756±0.040)E+2 (1.814±0.016)E+2
8.556E-5
1.343E+01.405E+2
2.753E+2
1.815E+2
7.427E-4
1.349E+01.390E+2
2.756E+2
1.797E+2
7.427E-4
1.350E+01.404E+2
2.756E+2
1.815E+2
8.556E-5
1.347E+01.404E+2
2.780E+2
1.849E+2
8.522E-4
1.346E+01.326E+2
2.781E+2
1.846E+2C(2.700±0.300)E-4(1.460±0.060)E+2 (2.770±0.030)E+2 (1.800±0.200)E+2(3.154±0.325)E-5 (1.151±0.118)E-2 (6.926±2.244)E-1 (4.004±0.062)E-1 (5.278±0.549)E-1
1.695E-7
1.148E-2
6.947E-1
3.970E-1
5.519E-1
3.155E-5
1.184E-2
7.048E-1
3.989E-1
5.400E-1
3.15E-5
1.217E-2
6.926E-1
4.004E-1
5.517E-1
1.695E-7
1.151E-2
6.953E-1
3.992E-1
5.165E-1
3.585E-5
1.151E-2
6.930E-1
4.011E-1
5.651E-1(3.800±0.400)E-5 (1.170±0.050)E-2(1.119±1.073)E-4 (2.893±0.255)E-3 (9.025±1.870)E-2 (6.720±0.142)E-2 (3.870±0.679)E-2
2.827E-8
2.634E-3
9.031E-2
6.632E-2
5.194E-2
1.117E-4
3.316E-3
8.470E-2
6.802E-2
5.105E-2
1.117E-4
3.174E-3
9.012E-2
6.716E-2
5.181E-2
2.827E-8
3.125E-3
7.611E-2
6.765E-2
5.118E-2
7.929E-5
2.893E-3
9.017E-2
6.675E-2
6.880E-2
16

radiative capture cross sections of JEFF-3.1 and CENDL3.1 16 O evaluations, the corresponding neutron capture
MACS strongly deviate from the rest.

E.

Capture MACS (30 keV)/ENDF/B-VII.1

1

0.1

ENDF/B-VII.1
JEFF-3.1.2
JENDL-4.0
ROSFOND-2010
CENDL-3.1
EAF-2010
KADONIS1-H - 1Cf(sf ) cross sections

The californium neutron source provides a pure fission
neutron spectrum. This is very useful for validation testsof evaluated data in the fast region, even though it is not
representative of a fast reactor spectrum (being hotter).
Selected fission neutron spectrum-averaged experimental
data have been compiled in the EXFOR database [99].
Many experimental spectrum-averaged cross sections
have been evaluated by the dosimetry community [12,46, 101]. In this work, we used Mannhart’s evaluation[101] of the 252 Cf(sf) neutron spectrum. As is shownin Tables II - III, the calculated californium spectrumaveraged cross sections exhibit reasonable agreement withmeasurements. However, spectrum-averaged cross section measurements of this type below 2 MeV are very
diﬃcult due to increased neutron scattering and background; measurements above 8 MeV become diﬃcult due
to reduced neutron flux. Therefore both capture andfission measurements in a Cf spectrum should be takenwith care. New spectrum-averaged cross section measurements in a Cf-spectrum are encouraged, to improve ourdatabase and to provide a better validation of our crosssection evaluations. We also note that small diﬀerencesin various computational methods for calculating the cal-

10

0.01

252

8-O - 16 26-Fe- 56 92-U -235 92-U -238 94-Pu-239

FIG. 11. Ratio of calculated MACS at kT =30 keV and
KADoNiS data to ENDF/B-VII.1 for neutron-induced capture reactions.

20

The CIELO Collaboration . . .

NUCLEAR DATA SHEETS

M.B. Chadwick

56

Fe, 235,238 U, and 239 Pu among the evaluated nuclear
data libraries (ENDF/B-VII.1, JEFF-3.1, JENDL-4.0,
CENDL-3.1, and ROSFOND). We anticipate that thiscollaboration will help identify future experiments thatare needed — from cross section measurements (e.g., ascollated in the NEA’s High Priority Request List) to more
integral experiments. The next step is to put together
teams of specialists to resolve these discrepancies. This
is now being done in the framework of the NEA WPEC
in close collaboration with the IAEA-NDS for standards
evaluations.

ifornium spectrum-averaged cross sections, together withvariations in measured values, leads to some small diﬀerences in Table II (from Pritychenko) relative to Capote
et al.’s values [102] that also uses ENDF/B-VII.0 (whichis the same as VII.1 for fission). For example, for fission, Capote et al. have: 235 U(n, f ) calc. 1.225±0.4% b,exp. 1.210±1.2% b; 238 U(n, f ) calc. 0.3185±0.6% b, exp.
0.3257±1.6% b; 239 Pu(n, f ) calc. 1.796±0.5% b, exp.
1.812±1.4% b.
We also note that for capture, the 239 Pu(n,γ) reactionneeds work and ENDF/B-VII.1 appears to be discrepantcompared to the other evaluations; however, there are few
measured data to compare against.

We are proposing a closer international cooperationfor evaluating nuclear data. This will ultimately benefit nuclear data users through the creation of a higher
fidelity database. Our long-term objective is to create
the international nuclear data library CIELO (Collaborative International Evaluated Library Organization),and as the first step we focused in this paper on indentifying discrepancies in the evaluations for 1 H, 16 O,

We thank P. Oblozinsky and A. Nichols for usefulcomments. This work was carried out under the auspices of the National Nuclear Security Administration of
the US Department of Energy at Los Alamos National
Laboratory under Contract No. DE-AC52-06NA25396,
at Brookhaven National Laboratory under Contract No.
DE-AC02-98CH10886, at Oak Ridge National Laboratory under Contract No. DE-AC05-00OR22725, and at
Lawrence Livermore National Laboratory under Contract
DE-AC52-07NA27344. R. Vogt also acknowledges the
Physics Department, University of California at Davisfor support.

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May 1991, p.349 (1991).
[58] P. G. Young et al., Nucl. Data Sheets 108, 2589
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[59] J.L. Kammerdiener, Lawrence Livermore National Laboratory report UCRL-51232 (1972).
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on Nuclear Data, 18–19 Nov., 1993, JAERI, Tokai,
Japan, Ed. M. Kawai and T. Fukahori, JAERI-M 94019, pp.290–299 (1994).
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(2002).
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Data Sheets 115, xxx (2014).

22

The CIELO Collaboration . . .

NUCLEAR DATA SHEETSAppendix A: Updated Results of LANL Integral
Experiments

TABLE VI. Reaction rate measurements in plutonium
Flattop-Pu critical assemblies.
Assembly/date
Flattop-Pu
8/10/1967

In this appendix we provide measured experimental
data obtained at Los Alamos National Laboratory on
activation and fission cross sections within fast criticalassemblies. These data were used in Figs. 5 and 7.

8/5/1970
TABLE IV. 14.1 MeV cross sections (barns) used in thecalibration process for determining activation cross sections
in critical assemblies. The pre-1963 and post-1963 values
were those in use at the time by Los Alamos radiochemists.
The uncertainties adopted were taken from various sources:
ENDF/B-VII.1, IRDFF, or internal LANL estimates.

8/5/1963
3/25/1962

Reaction
Measurement
Pu(n, f )/235 U(n, f )
1.415
169
Tm(n, 2n)/235 U(n, f )2.433E-03
203
Tl(n, 2n)/235 U(n, f )2.270E-03
107
Ag(n, 2n)106m Ag/235 U(n, f ) 1.973E-04
238
U(n, f )/235 U(n, f )(1.484E-01)a
191
235
Ir(n, 2n)/ U(n, f )2.831E-03
238
U(n, 2n)/235 U(n, f )9.041E-03
238
U(n, f )/235 U(n, f )1.712E-01
238
U(n, 2n)/235 U(n, f )8.989E-03
239

a We do not recommend using this value as it appears to be

erroneous and is discrepant with the 0.1712 value shown. This
spectral index value should be larger than the values in
Flattop-25 but it is not.

Reaction

Los Alamos
ENDF/B
pre-1963
post-1963 VII.1
Unc.(%)
Sc(n, 2n)
Sc
0.115
0.105
0.104
5.0
51
V(n, α)48 Sc
0.01570.01525.0
75As(n, 2n)74 As
1.050
0.994
5.0
89
Y(n, 2n)88 Y
0.670
0.845
0.850
2.1
90
Zr(n, 2n)89 Zr
0.643
0.590
0.617
0.91
103
Rh(n, 2n)102g Rh 0.750
0.783
0.741
5.0
107
Ag(n, 2n)106m Ag
0.573
0.520
5.0
169
Tm(n, 2n)168 Tm
1.96
1.980
1.7
175
Lu(n, 2n)174 Lu
1.789
2.122
5.0
191
1.995
2.066
3.0
Ir(n, 2n)190 Ira
197
Au(n, 2n)196 Au
2.214
2.132
1.1
203
Tl(n, 2n)202 Tl 1.428
2.090
2.005
5.0
204
Pb(n, 2n)203 Pb 1.746
2.193
5.0
238
U(n, 2n)237 U
0.895
0.850
2.4
45

M.B. Chadwick

44m

TABLE VII. Reaction rate measurements in the HEU Godiva
critical assembly.
Assembly/date
Reaction
Measurement
238
GodivaU(n, 2n)/235 U(n, f ) 7.729E-03
238
6/30/1959
U(n, f )/235 U(n, f ) 1.629E-01
239
Pu(n, f )/235 U(n, f ) 1.365

TABLE VIII. Reaction rate measurements in the HEU
Flattop-25 critical assemblies.
Assembly/date
Reaction
Measurement75
Flattop-25
As(n, 2n)/235 U(n, f )1.543E-04
103
3/5/1962
Rh(n, 2n)102g Rh/235 U(n, f ) 2.618E-04
204
Pb(n, 2n)/235 U(n, f )1.635E-05
51
V(n, α)/235 U(n, f )1.142E-05
45
Sc(n, 2n)44m Sc/235 U(n, f ) 9.017E-06
75
3/30/1962
As(n, 2n)/235 U(n, f )1.452E-04
51
V(n, α)/235 U(n, f )1.087E-05
45
Sc(n, 2n)44m Sc/235 U(n, f ) 8.889E-06
239
4/29/1968
Pu(n, f )/235 U(n, f )
1.382
169
Tm(n, 2n)/235 U(n, f )1.468E-03
203
Tl(n, 2n)/235 U(n, f )1.535E-03
107
Ag(n, 2n)106m Ag/235 U(n, f ) 1.336E-04
197
4/7/1969
Au(n, 2n)/235 U(n, f )1.614E-03
191
Ir(n, 2n)/235 U(n, f )1.828E-03

a 190g+m1+m2+8.6%m3 is measured, see Ref. [45], Fig. 13.

TABLE V. Reaction rate measurements in plutonium Jezebel
critical assemblies.
Assembly/date
Reaction
Measurement239
JezebelPu(n, f )/235 U(n, f ) 1.430
169
10/23/1973
Tm(n, 2n)/235 U(n, f ) 3.125E-03
191
Ir(n, 2n)/235 U(n, f ) 3.206E-03
238
U(n, 2n)/235 U(n, f ) 1.058E-02a
238
U(n, f )/235 U(n, f ) 2.121E-01
a Values we gave in NDS 108, 2587 (2007) Table IX (∼1.42E-02)

for 238 U(n, 2n) were erroneous. We apologize – those values
were for Jezebel-33.

TABLE IX. Reaction rate measurements in the HEU Flattop25 critical assembly traverse, 2/20/1961 (distance measured
from center in cm).

By comparing MCNP model calculations of thesereactions (using ENDF/B-VII.1) with our measurements, we can assess the accuracy of the ENDF/B-VII.1(n, 2n) cross sections, the prompt fission neutron spectra (PFNS), and the neutronic transport modeling. The
value of these data in particular relates to insights thatare provided on the accuracy of our PFNS. This is because we focus on the use of activation dosimetry reactions where the (n, 2n) cross sections are well known.
Also, we aim to use a number of diﬀerent dosimetry reactions so that even if one evaluated cross section is biased
in some way, we benefit from using feedback from theensemble of reactions.
The measurements reported here are old and were carried out in the 1960s and 1970s. Some of them have

Distance
Reaction
Measurement
238
0.4
U(n, 2n)/235 U(n, f ) 6.975E-03
238
1.03
U(n, 2n)/235 U(n, f ) 7.199E-03
238
3.69
U(n, 2n)/235 U(n, f ) 6.364E-03
238
5.6
U(n, 2n)/235 U(n, f ) 5.679E-03
238
6.32
U(n, 2n)/235 U(n, f ) 4.562E-03
238
7.7
U(n, 2n)/235 U(n, f ) 3.059E-03
238
12.78
U(n, 2n)/235 U(n, f ) 1.243E-03
238
0.4
U(n, f )/235 U(n, f ) 0.1543
238
1.03
U(n, f )/235 U(n, f ) 0.1560
238
3.69
U(n, f )/235 U(n, f ) 0.1435
238
5.6
U(n, f )/235 U(n, f ) 0.1266
238
6.32
U(n, f )/235 U(n, f ) 0.1005
238
7.7
U(n, f )/235 U(n, f ) 0.0710
238
12.78
U(n, f )/235 U(n, f ) 0.0279

23

The CIELO Collaboration . . .

NUCLEAR DATA SHEETS

M.B. Chadwick

TABLE X. Reaction rate measurements in the HEU Flattop25 critical assembly traverse, 3/13/1961 (distance measured
from center in cm).

TABLE XII. Reaction rate measurements of 203 Tl(n, 2n) in
the HEU Flattop-25 critical assembly traverse, 3/13/1961(distance measured from center in cm).

Distance
Reaction
Measurement
238
0.01
U(n, 2n)/235 U(n, f ) 6.826E-03
238
1.02
U(n, 2n)/235 U(n, f ) 6.765E-03
238
3.7
U(n, 2n)/235 U(n, f ) 6.459E-03
238
5.16
U(n, 2n)/235 U(n, f ) 5.963E-03
238
6.31
U(n, 2n)/235 U(n, f ) 4.531E-03
238
9.62
U(n, 2n)/235 U(n, f ) 1.641E-03
238
14.75
U(n, 2n)/235 U(n, f ) 8.676E-04
238
0.01
U(n, f )/235 U(n, f ) 0.1502
238
1.02
U(n, f )/235 U(n, f ) 0.1512
238
3.7
U(n, f )/235 U(n, f ) 0.1421
238
5.16
U(n, f )/235 U(n, f ) 0.1289
238
6.31
U(n, f )/235 U(n, f ) 0.1002
238
9.62
U(n, f )/235 U(n, f ) 0.0428
238
14.75
U(n, f )/235 U(n, f ) 0.0206

Distance
Reaction
Measurement
203
0.01
Tl(n, 2n)/235 U(n, f )1.495E-03
203
1.02
Tl(n, 2n)/235 U(n, f )1.465E-03
203
3.7
Tl(n, 2n)/235 U(n, f )1.401E-03
203
5.16
Tl(n, 2n)/235 U(n, f )1.403E-03
203
6.31
Tl(n, 2n)/235 U(n, f )1.011E-03
203
9.62
Tl(n, 2n)/235 U(n, f )4.432E-04
203
14.75
Tl(n, 2n)/235 U(n, f )2.119E-04

TABLE XIII. Reaction rate measurements of 90 Zr(n, 2n) in
the HEU Flattop-25 critical assembly traverse, 2/20/1961(distance measured from center in cm).
Distance
Reaction
Measurement
90
0.4
Zr(n, 2n)/235 U(n, f )4.778E-05
90
1.03
Zr(n, 2n)/235 U(n, f )5.153E-05
90
3.69
Zr(n, 2n)/235 U(n, f )4.678E-05
90
5.6
Zr(n, 2n)/235 U(n, f )4.004E-05
90
6.32
Zr(n, 2n)/235 U(n, f )3.432E-05
90
7.7
Zr(n, 2n)/235 U(n, f )2.277E-05
90
12.78
Zr(n, 2n)/235 U(n, f )1.017E-05

TABLE XI. Reaction rate measurements of 203 Tl(n, 2n) in the
HEU Flattop-25 critical assembly traverse, 2/20/1961 (distance measured from center in cm).
Distance
Reaction
Measurement
203
0.4
Tl(n, 2n)/235 U(n, f )1.558E-03
203
1.03
Tl(n, 2n)/235 U(n, f )1.571E-03
203
3.69
Tl(n, 2n)/235 U(n, f )1.466E-03
203
5.6
Tl(n, 2n)/235 U(n, f )1.251E-03
203
6.32
Tl(n, 2n)/235 U(n, f )1.059E-03
203
7.7
Tl(n, 2n)/235 U(n, f )6.884E-04
203
12.78
Tl(n, 2n)/235 U(n, f )3.471E-04

scale for the historical reported values. But since ourknowledge of the 14.1 MeV cross sections has changed
since the 1960s and 1970s, we update the original reported data to our new 14.1 MeV cross sections. Table IV shows the historical cross sections assumed as wellas the modern ENDF/B-VII.1 values we have calibrated
against. A second aspect is that we believe our historical
radiochemical estimates of 235 U fission (in a fission spectrum) were too high by 2%, as discussed by Selby [86],and therefore our updated values account for this change.
Table V gives measured data in the center of theJezebel assembly, a plutonium sphere. Table VI gives
measured data in the center of the Flattop-Pu assembly,
a plutonium core surrounded by depleted uranium. Table VII give measured data in the center of the Godiva
assembly, a highly-enriched uranium sphere. Table VIII
give measured data in the center of the Flattop-25 assembly, a highly-enriched uranium core surrounded bydepleted uranium.
The subsequent tables provide data through a traverse

been reported before in our Nuclear Data Sheets papers[45, 58] and by Frankle [103], but in the course of thiswork we have found some additional measurements that
we report here. Most importantly, we update the “historical” LANL reported experimental values to new values
based on use of modern calibration standards, and we
also provide uncertainty assessments on the data, see Table I. This work is reported in more detail in a laboratory
report by MacInnes et al. [104]. The measurements arereported in ratio to 235 U fission at the same location, a
traditional Los Alamos approach that has the advantage
of not requiring an absolute measurement of the neutron
fluence to use the data. It does require that we also model
fission in MCNP, but because 235 U(n, f ) is a standard
and we have much experience modeling it in our criticality calculations, the associated uncertainties in this partof the calculation are rather small, less than about 2%.
The procedure we have gone through to update theoriginal reported measured values is as follows. One aspect is that the original radiochemical measurements determined count rates from the activation product decays
following an irradition, and to convert these to an absolute physical scale the original scientists performed concurrent irradiations in a 14.1 MeV neutron beam, where
the 14.1 MeV cross section was “known”, measuring theinduced count rates there as well. This calibration process, using the known 14.1 MeV cross sections, set theTABLE XIV. Reaction rate measurements of 90 Zr(n, 2n) in
the HEU Flattop-25 critical assembly traverse, 3/13/1961(distance measured from center in cm).
Distance
Reaction
Measurement90
0.01
Zr(n, 2n)/235 U(n, f )4.765E-05
90
1.02
Zr(n, 2n)/235 U(n, f )4.705E-05
90
3.7
Zr(n, 2n)/235 U(n, f )4.400E-05
90
5.16
Zr(n, 2n)/235 U(n, f )5.081E-05
90
6.31
Zr(n, 2n)/235 U(n, f )3.480E-05
90
9.62
Zr(n, 2n)/235 U(n, f )1.414E-05
90
14.75
Zr(n, 2n)/235 U(n, f )1.862E-05

24

The CIELO Collaboration . . .

NUCLEAR DATA SHEETS

M.B. Chadwick

TABLE XV. Reaction rate measurements of 191 Ir(n, 2n) in the
HEU Flattop-25 critical assembly traverse, 8/18/1970 (distance measured from center in cm for samples co-located withthe spectral index data shown in the following table).

TABLE XVI. Reaction rate measurements of 238 U(n, f ) in the
HEU Flattop-25 critical assembly traverse, 8/18/1970 (distance measured from center in cm for samples co-located with
the iridium data in the previous table).

Distance
Reaction
Measurement191
1.1
Ir(n, 2n)/235 U(n, f )1.904E-03
191
0.96
Ir(n, 2n)/235 U(n, f )1.936E-03
191
3.66
Ir(n, 2n)/235 U(n, f )1.799E-03
191
5.02
Ir(n, 2n)/235 U(n, f )1.715E-03
191
5.76
Ir(n, 2n)/235 U(n, f )1.515E-03
191
8.36
Ir(n, 2n)/235 U(n, f )9.289E-04
191
11.0
Ir(n, 2n)/235 U(n, f )5.544E-04
191
14.8
Ir(n, 2n)/235 U(n, f )3.766E-04

Distance
Reaction
Measurement
238
1.1
U(n, f )/235 U(n, f )0.1485238
0.96
U(n, f )/235 U(n, f )0.1495
238
3.66
U(n, f )/235 U(n, f )0.1455238
5.02
U(n, f )/235 U(n, f )0.1315238
5.76
U(n, f )/235 U(n, f )0.1144
238
8.36
U(n, f )/235 U(n, f )0.0583
238
11.0
U(n, f )/235 U(n, f )0.0361
238
14.8
U(n, f )/235 U(n, f )0.0223of Flattop-25 from the central region out into the tamper.
Tables IX and X give data for 238 U(n, 2n) and (n, f ).
Tables XI and XII give data for 203 Tl(n, 2n). Tables XIII
and XIV give data for 90 Zr(n, 2n). Tables XV and XVIgive data for 191 Ir(n, 2n) co-located with 238 U samples
used to measure the spectral index.

25