Issue 
EPJ Nuclear Sci. Technol.
Volume 4, 2018
Special Issue on 4th International Workshop on Nuclear Data Covariances, October 2–6, 2017, Aix en Provence, France – CW2017



Article Number  41  
Number of page(s)  9  
Section  Applied Covariances  
DOI  https://doi.org/10.1051/epjn/2018035  
Published online  14 November 2018 
https://doi.org/10.1051/epjn/2018035
Regular Article
Use of integral data assimilation and differential measurements as a contribution to improve ^{235}U and ^{238}U cross sections evaluations in the fast and epithermal energy range
^{1}
CEA, DEN, DER, SPRC Cadarache,
13108
St PaulLezDurance, France
^{2}
ED352 Doctoral School, AMU, Luminy Campus,
13288
Marseille, France
^{*} email: virginie.huy@cea.fr
Received:
7
December
2017
Received in final form:
1
March
2018
Accepted:
22
May
2018
Published online: 14 November 2018
Critical mass calculations of various HEUfueled fast reactors result in large discrepancies in C/E values, depending on the nuclear data library used and the configuration modeled. Thus, it seems relevant to use integral experiments to try to reassess cross sections that might be responsible for such a dispersion in critical mass results. This work makes use of the Generalized Least Square method to solve Bayes equation, as implemented in the CONRAD code. Experimental database used includes ICSBEP Uranium based critical experiments and benefits from recent reanalyses of MASURCA and FCAIX criticality experiments (with MonteCarlo calculations) and of PROFIL irradiation experiments. These last ones provide very specific information on ^{235}U and ^{238}U capture cross sections. Due to high experimental uncertainties associated to fission spectra, we chose to consider either fitting these data or set them to JEFF3.1.1 evaluations. The work focused on JEFF3.1.1 ^{235}U and ^{238}U evaluations and results presented in this paper for ^{235}U capture and ^{238}U capture, and inelastic cross sections are compared to recent differential experiment or recent evaluations. Our integral experiment assimilation work notably suggests a 30% decrease for ^{235}U capture around 1–2.25 keV, a 10% increase in the unresolved resonance range when using JEFF3.1.1 as “a priori” data. These results are in agreement with recent microscopic measurements from Danon et al. [Nucl. Sci. Eng. 187, 291 (2017)] and Jandel et al. [Phys. Rev. Lett. 109, 202506 (2012)]. For ^{238}U cross sections, results are highly dependent on fission spectra.
© V. Huy et al., published by EDP Sciences, 2018
This is an Open Access article distributed under the terms of the Creative Commons Attribution License (http://creativecommons.org/licenses/by/4.0), which permits unrestricted use, distribution, and reproduction in any medium, provided the original work is properly cited.
1 Introduction
Critical mass calculations of various HEUfueled fast reactors result in large discrepancies in C/E values, depending on flux spectra, fuel enrichment, structural materials present and so on. These C/E values, calculated with the MonteCarlo code TRIPOLI4 [1], are shown in Figure 1. Table 1 gives some specifications about fuel and structural materials present in each configuration.
Figure 1 underlines that critical mass C/E values for Uraniumfueled configurations of Fast Reactors calculated with JEFF3.2 library are systematically overestimated (except for BIGTEN and GODIVA) and are larger than those calculated with JEFF3.1.1. Discrepancy between the two sets of calculations goes from ∼250 to ∼630 pcm for BIGTEN. Moreover, large C/E values are observed for FCAIX configurations (overestimation up to ∼800 pcm), BIGTEN and GODIVA when using the JEFF3.1.1 library. Except for FLATTOP^{235}U, all critical masses for configurations with HEU fuel exceed experimental uncertainties when using JEFF3.1.1. Although MASURCA 1B and FCAIX configurations [2] have similar spectra (as they both contain graphite) but significantly different Uranium enrichments and geometries, the large discrepancy observed in their C/E values (using either JEFF3.1.1 or JEFF3.2) rise concerns of possible compensating errors between ^{235}U and ^{238}U evaluations in the JEFF libraries in the fast and epithermal energy range.
Fig. 1
Critical mass C/Es compared with experimental uncertainties for Uranium configurations (using JEFF libraries). 
Specifications on fuel enrichment and structural materials for the different configurations.
2 Integral experiments assimilation
Considering the very large C/E values presented in Section 1, it seems relevant to use integral data assimilation to identify which nuclear data are responsible for these discrepancies. This was performed using the CONRAD code from CEA [3], which can solve analytically Bayes' theorem.
2.1 Bayesian inference
As a reminder, Bayes' theorem [4] generalized to continuous probability densities is given: (1) where the vector x contains the parameters to be reassessed (in our case, the 33group cross sections) in the view of new observations enclosed in the vector y. U gathers all the “background” information, that is, hypotheses or approximations made to obtain the values for x and y.
In practice, probability densities associated to each multigroup crosssection are assumed to be Gaussian distributions, as this choice maximizes the entropy [5]. Using Laplace approximation [6], we then assume that the posterior probability density function solution of equation (1) can be wellapproximated by a Gaussian distribution: (2) where E is a vector containing integral measurements values, C is a vector containing associated calculated values, M_{σ} is the covariance matrix associated to nuclear data σ, M_{E} is the covariance matrix associated to C/E values.
For a Gaussian distribution, the central value is associated to its maximum. Thus, optimal solutions for σ and associated covariances, M_{σ} are determined by minimizing a cost function (using Generalized Least Square method): (3)
2.2 Integral data assimilation strategy and results for posterior C/E
The JEFF3.1.1 library was chosen as the a priori as it gives more satisfying results than JEFF3.2 for Uranium configurations sensitive to the fast energy range, as seen in Figure 1. For our assimilation work, we used critical mass C/E of MASURCA 1B, FCAIX cores 1–7, FLATTOP^{235}U and GODIVA, as well as variations of concentration ratios C/E from PROFIL2A irradiation experiments [7,8]. Experimental correlation matrix for FCAIX configurations is provided in reference [2]. Experimental correlation matrix for PROFIL experiments is given in Table 2. In this table, “^{8}U” refers to the ratio variation and “^{5}U” refers to the ratio variation .
Experimental correlation matrix for PROFIL2A C/E.
C/E used in the assimilation work were calculated using the MonteCarlo code TRIPOLI4 (except for PROFIL's variation of concentrations ratios, calculated with ECCO/ERANOS) and 33group sensitivity coefficients to nuclear data were calculated using the ECCO/ERANOS code [9]. For nuclear data covariance matrices, we used COMACV1.0 [10], except for ^{235}U ν for which we used the COMMARA2.0 matrix [11].
Critical mass C/E values for these configurations provide a great variety of sensitivity profiles to ^{235}U capture and ^{238}U capture and inelastic cross sections (this is shown for ^{235}U capture in Fig. 2). Using all these C/E values with their associated sensitivity coefficients in a single assimilation calculation allows us to make the most of both the redundant or complementary information they provide for the whole fast energy range.
Fig. 2
33group sensitivity profiles of several critical masses to ^{235}U capture. 
Notably, the simultaneous use of GODIVA and FLATTOP^{235}U critical masses can help avoiding compensations between ^{235}U and ^{238}U cross sections, as these fast spectrum critical configurations are similar, except for the presence of natural Uranium reflector in FLATTOP^{235}U. Critical mass sensitivities to ^{238}U inelastic and capture and ^{235}U capture cross sections for these two configurations are shown in Figure 3. One can see that critical mass sensitivities to ^{235}U cross sections are similar whereas sensitivity coefficients to ^{238}U cross sections are important for FLATTOP^{235}U and low for GODIVA.
Fig. 3
Sensitivity coefficients of FLATTOP^{235}U and GODIVA critical masses to ^{235}U capture and ^{238}U inelastic and capture cross sections. 
Impact on MASURCA 1B and FCAIX 1 to 3 critical masses when using carbon evaluation of JENDL4.0 instead of JEFF3.1.1.
The nuclear data fitted through assimilation are ^{235}U and ^{238}U capture, elastic, inelastic 33groups cross sections as well as their fission spectrum χ (unless specified otherwise) and multiplicity ν. ^{235}U and ^{238}U fission were not fitted, as JEFF3.1.1 evaluations are in good agreement with Neutron Standard from IAEA [12] for these cross sections. Also, it should be noted that assimilation work does not take into account sensitivities to angular distributions as no covariance matrices are currently available for these data. Taking into account these approximations through marginalization is the topic of future works.
In this integral data assimilation work, an effort was made to try to reduce risks of compensating errors by relying on the Neutron Crosssection Standards [12] for ^{235}U and ^{238}U fission cross sections and by using PROFIL2A C/E (which add a specific constraint on ^{235}U or ^{238}U capture cross sections).
Nevertheless, as this will be shown in the following sections, high uncertainties associated to fission spectra can have a significant impact on assimilation result. Also, as differences in JEFF3.1.1 and JENDL4.0 carbon evaluations were found to have a nonnegligible impact for some critical masses used in this work (Tab. 3), we ran CONRAD calculations for both of these options. For these reasons, the results presented in Section 3 are sets of trends that include the four alternatives considered: fission spectra fitted or not and carbon evaluation either from JEFF3.1.1 or JENDL4.0. Assimilation trends are presented in this manner to stress that the variability in the results due to these choices can be seen as additional uncertainties.
Experimental correlations between FCAIX critical mass C/E were taken into account using the matrix provided by JAEA [2]. Also, correlations between PROFIL irradiation experiments were calculated. Figure 4 displays postassimilation C/E for critical masses compared with prior JEFF3.1.1C/E values for the case where fission spectra are set to JEFF3.1.1 and JEFF3.1.1 graphite evaluation is used. A priori and a posteriori C/E values for the PROFIL irradiation experiment are given in Table 4, along with experimental uncertainties.
Fig. 4
Comparison between prior (JEFF3.1.1) and posterior C/E values. 
Prior and posterior C/E values for PROFIL2A variation of concentrations ratios and associated uncertainties.
Postassimilation C/E values are wellincluded in 1σ experimental uncertainties, except for MASURCA 1B and FCAIX 6, which however remain in 2σ total uncertainties. This means there exists an optimal set of cross sections for the experimental database taken into account, and no inconsistency between C/E had been found.
3 Comparison of assimilation trends with differential measurements
To discuss the reliability of the trends on cross sections suggested through the integral data assimilation, we compared them to recent differential measurements from the EXFOR database [13] when they are available or recent evaluations otherwise. In this section, trends are given relative to JEFF3.1.1.
3.1 ^{235}U capture cross section
Assimilation results suggest a significant modification for ^{235}U capture: a ∼30% decrease around 1–2 keV and a ∼10% increase in the unresolved resonance range (URR) when using JEFF3.1.1 as “a priori” data. This is shown in Figure 5, along with prior and posterior uncertainties. One can notice that from 1 to 500 keV, posterior uncertainties are sufficiently low to consider assimilation trends as possible recommendations for a change in ^{235}U capture cross section. As mentioned earlier, the two curves displayed in Figure 5 represent an envelope, in which the assimilation results for the following four cases are included: uncertainties on graphite evaluation choice (JEFF3.1.1 or JENDL4.0) and fission spectra (fitted or set to JEFF3.1.1). For ^{235}U capture cross sections, differences in posterior uncertainties for these four cases do not exceed 0.5% in the energy range of interest. Thus, only one curve is displayed in Figure 5.
Focusing on the end of the resolved resonances range (RRR) from 1 to 2.25 keV, we compared our assimilation trends in this energy range with recent differential measurements made at RPI. Figure 6 displays results of these measurements as published in reference [14] (as they are not currently available in the EXFOR database) with a comparison to ENDF/BVII and JENDL4.0. One has to note that for ^{235}U capture cross section, JEFF3.1.1 and ENDF/BVII.1 evaluations are identical. This graph of Figure 6 shows that our assimilation results are in good agreement with Danon measurements at RPI as they suggest a ∼33% decrease of ^{235}U capture cross section from JEFF3.1.1 at around 2 keV. This issue on ^{235}U capture was already addressed in WPEC Subgroup 29 [15], which underlined an overestimation of this cross section in the end of the RRR in the JEFF3.1 evaluation.
In the URR, from 10 to 100 keV, most recent measurements performed by Jandel et al. [16] at LANSCE with the DANCE detector are consistent with assimilation trends from 3 keV to 1 MeV (Fig. 7).
Comparing now assimilation results to JEFF3.3t3 [17] (in Fig. 8), one can see that they agree well in the end of the RRR (considering that assimilation results uncertainties in this range is around 9%). In the URR, from 10 to 100 keV, JEFF3.3t3 evaluation suggests a higher increase from JEFF3.1.1 (around 20%) than our assimilation results.
Figure 9 shows a comparison between Jandel et al. [16] measurements, JEFF3.3t3 [17] and JEFF3.1.1 evaluations. Compared to Jandel measurements, it seems that JEFF3.3t3 ^{235}U capture cross section evaluation is slightly higher whereas JEFF3.1.1 appears to underestimate this cross section in the 10–100 keV energy range.
Fig. 5
Trends from assimilation work for ^{235}U capture (relative to JEFF3.1.1) compared with a priori and a posteriori nuclear data uncertainties. The two red dotted curves represent an envelope gathering all the trends suggested by assimilation results (that includes cases with fission spectra fitted or not, and with graphite evaluation from JEFF3.1.1 or JENDL4.0). 
Fig. 6
Results of differential measurements from Danon et al. [14] for ^{235}U capture from 0.5 to 3 keV, compared with ENDF/BVII.1 and JENDL4.0 evaluations. 
Fig. 7
Results of differential measurements from Jandel et al. [16] for ^{235}U capture from 3 keV to 1 MeV. Comparison with assimilation results applied to JEFF3.1.1 pointwise evaluations (red continuous line). 
Fig. 8
33group assimilation results for ^{235}U capture compared with “a priori” JEFF3.1.1 and JEFF3.3t3 evaluation. Posterior uncertainties for assimilation results are in dotted line. 
Fig. 9
Comparison of Jandel et al. measurements [16] to JEFF3.3t3 and JEFF3.1.1 evaluations for ^{235}U capture cross sections. 
3.2 ^{238}U capture cross section
Unlike ^{235}U capture, trends for ^{238}U capture are highly dependent on fission spectra values. As it can be seen in Figure 10, in the case where fission spectra are fitted through assimilation, resulting trends on ^{238}U capture are included in posterior uncertainties. When fission spectra are not fitted and set to JEFF3.1.1, trends suggested (−4% up to −7% from JEFF3.1.1) by the assimilation work are higher than posterior uncertainties from 10 keV to 3 MeV. Dependency of the results on fission spectra values is also reflected by the differences in posterior uncertainties for the two cases (Fig. 10). A posteriori uncertainties for ^{238}U capture are noticeably higher in the case where fission spectra are fitted. However, one can notice that the choice for graphite evaluation has little impact on the results in both cases.
The dependency of assimilation results for ^{238}U capture cross section can be explained by the fact that sensitivity coefficients of PROFIL ratio variations are at the same level as critical masses sensitivity coefficients for this cross section. Moreover, from 100 keV to 1 MeV, these sensitivity coefficients are noticeably lower than those of some critical masses. This is not the case for whose sensitivity profile dominates all the critical mass sensitivity profiles to ^{235}U capture. The constraint brought by PROFIL2A C/E on ^{238}U capture is thus less important than for ^{235}U capture. This is shown in Figure 11. Also, a priori correlations between ^{238}U cross sections might amplify the impact of fission spectra on assimilation results.
In the end, the great impact of fission spectra on ^{238}U capture results suggests possible compensations between ^{238}U capture and ^{238}U and ^{235}U fission spectra in our assimilation work. This assimilation results for ^{238}U capture cross section are all the more questioning as these can have a significant impact on fast reactor calculations. For instance, the trend suggested by the assimilation (for the case where fission spectra are set to JEFF3.1.1) has an impact of around +500 pcm on the reactivity of the SFR ASTRID. Details of this impact per energy group (for a 33group sensitivity calculation) are given in Table 5. Thus, considering the high sensitivity of some fast reactors critical masses to this cross section, assimilation results should be clarified, for instance by using a wider experimental database for the assimilation.
Fig. 10
Trends from assimilation work for ^{238}U capture (relative to JEFF3.1.1) compared with a priori and a posteriori nuclear data uncertainties. For both cases (fission spectra fitted or not), the two dotted lines have to be seen as an additional uncertainty associated to the choice of graphite evaluation. 
Fig. 11
Comparison of sensitivity profiles of PROFIL2A C/E, and FCAIX 7 and MASURCA 1B critical masses to ^{235}U and ^{238}U capture. 
Relative impact on ASTRID critical mass of the trends suggested by assimilation when fissions spectra are set to JEFF3.1.1 evaluations. Only trends superior to posterior uncertainties were considered.
3.3 ^{238}U inelastic cross section
As for ^{238}U capture cross section, trends for ^{238}U inelastic depend on whether fission spectra are fitted through assimilation or set to JEFF3.1.1. Indeed, some of the critical configurations that are the most sensitive to ^{238}U inelastic cross are also the most sensitive to ^{238}U fission spectrum (FCAIX 6, FCAIX 7 and FLATTOP^{235}U). Besides, all critical configurations are highly sensitive to ^{235}U capture.
All sets of trends for ^{238}U inelastic are shown in Figure 12, along with associated uncertainties. A posteriori uncertainties are sufficiently low in the plateau region (∼1 to 6 MeV) to consider assimilation trends as possible recommendations. For this energy range, assimilation results propose a 4%–8% decrease (from JEFF3.1.1 ^{238}U inelastic cross section) depending on whether fission spectra are fitted or not. For ^{238}U inelastic cross sections, differences in posterior uncertainties for these four cases do not exceed 0.5% in the energy range of interest. Thus, only one curve is displayed in Figure 12.
Assimilation results are compared to CIELO [18] (evaluation version of September the 29th, 2017), JEFF3.1.1 and JEFF3.3t3 [17] evaluations in Figure 13. Focusing on the plateau region, we observe that CIELO and JEFF3.3t3 evaluations are both lower than JEFF3.1.1 in this region, but the level of decrease is different.
Once again, the dependency of assimilation results for ^{238}U inelastic cross sections on fission spectra is a hint of possible compensation errors in the results. Assimilation work can be improved with the use of a wider database including more C/Es sensitive to ^{238}U inelastic cross sections.
Fig. 12
Trends from assimilation work for ^{238}U inelastic (relative to JEFF3.1.1) compared with a priori and a posteriori nuclear data uncertainties. For both cases (fission spectra fitted or not), the two dotted lines have to be seen as an additional uncertainty associated to the choice of graphite evaluation. 
Fig. 13
33group assimilation results (case where fission spectra are not fitted and graphite evaluation used is from JEFF3.1.1) for ^{238}U inelastic compared with “a priori” JEFF3.1.1, CIELO and JEFF3.3t3 evaluation. Posterior uncertainties for assimilation results are in dotted line. 
4 Conclusion
C/E values from several critical masses calculations and from PROFIL irradiation experiments were used in a Bayesian inference approach as implemented in the CONRAD code to investigate cross sections that might need reassessment. These C/E values provide a great variety of sensitivity profiles to ^{235}U and ^{238}U cross sections, including capture and inelastic.
Trends suggested for ^{235}U capture, which are in agreement with recent differential measurements made at RPI and LANSCE, confirm that significant modifications are needed for this cross section in JEFF3.1.1 (∼30% decrease around 1–2.25 keV and ∼10% increase in the 10–100 keV energy range). This issue was already addressed in WPEC Subgroup 29, which underlined an overestimation of this cross section in the end of the RRR [15]. JEFF3.3t3 seems to go in the right direction with a decrease of around 25% from JEFF3.1.1 in the end of RRR and an increase up to 20% in the URR. Comparisons of integral data assimilation results with recent differential measurements constitute a key step in our study as sources of uncertainties are different.
For ^{238}U cross sections, results are highly dependent on whether fission spectra are fitted or not. For ^{238}U capture cross section, the integral data assimilation suggests a 4%–7% decrease of the cross section from 10 keV to 3 MeV in the case where fission spectra are set to JEFF3.1.1 evaluations. Such modifications on ^{238}U capture can have a significant impact on critical mass calculations of Fast Reactors. Thus, these results should be further confirmed by assimilation results using a wider experimental database.
For ^{238}U inelastic cross sections, integral data assimilation suggests a 4% to 8% decrease in the plateau region (from around 1 to 6 MeV), depending on whether fission spectra are fitted or not. JEFF3.3t3 and CIELO evaluations also point toward a decrease from JEFF3.1.1 in this energy region but at different levels. Previous work from Santamarina [19], using the RDN code and targeted on integral measurements with a strong sensitivity to ^{238}U inelastic cross section (including Pufueled systems), suggested a reduction trend of −11% ± 3% (in a case where ^{238}U fission spectra were not reestimated).
In the end, this assimilation work focusing on ^{235}U and ^{238}U nuclear data with a reduced database enables us to deduce possible trends on these data independently from Pu isotopes nuclear data. Results presented in this work have to be confirmed by the addition of other integral experiments. Notably, trends on ^{238}U capture and inelastic cross sections might possibly exhibit compensating errors. Besides, posterior uncertainties from this work are probably underestimated: indeed, we did not take into account uncertainty from nuclear data which are not fitted (structural material, fission cross sections, etc.). An attempt to take into account these approximations through marginalization is under study.
Author contribution statement
The results presented in this paper were produced in the framework of V. Huy PhD work. G. Rimpault and G. Noguere have contributed to this work by providing supervisory support and expert viewpoints.
Acknowledgments
The authors express their gratitude to S. Okajima, K. Tsujimoto and M. Fukushima, from JAEA for providing detailed information on the FCAIX experiments. The authors wish to thank J. Tommasi and E. Privas for their detailed work on the PROFIL experiments. Virginie Huy thanks EDF and CEA for their common financial support of her Ph.D.
References
 E. Brun, TRIPOLI4, CEA, EDF and AREVA reference Monte Carlo code, in Joint International Conference on Supercomputing in Nuclear Applications and Monte Carlo (2015), Vol. 82, pp. 151–160 [Google Scholar]
 M. Fukushima, Y. Kitamura, T. Kugo, S. Okajima, Benchmark models for criticalities of FCAIX assemblies with systematically changed neutron spectra, J. Nucl. Sci. Technol. 53, 406 (2016) [CrossRef] [Google Scholar]
 C. De Saint Jean et al., Uncertainty evaluation of nuclear reaction model parameters using integral and microscopic measurements with the CONRAD code, in ND2010 Conference (2010) [Google Scholar]
 T. Bayes, An essay toward solving a problem in the Doctrine of chances, Philos. Trans. R. Soc. Lond. 53, 370 (1763) [Google Scholar]
 C.E. Shannon, A mathematical theory of communication, Bell Syst. Tech. J. 27, 379 (1948) [CrossRef] [MathSciNet] [Google Scholar]
 A. AzevedoFilho, R.D. Shachter, Laplace's method approximations for probabilistic inference in belief networks with continuous variables, in Uncertainty in Artificial Intelligence Proceedings of the Tenth Conference, Seattle, Washington, USA (1994) pp. 28–36 [Google Scholar]
 E. Privas, P. Archier, C. De Saint Jean, G. Noguère, J. Tommasi, The use of nuclear data as nuisance parameters in the integral data assimilation of the PROFIL experiments, Nucl. Sci. Eng. 182, 377 (2016) [CrossRef] [Google Scholar]
 J. Tommasi, G. Noguere, Analysis of the PROFIL and PROFIL2 sample irradiation experiments in Phénix for JEFF3.1 nuclear data validation, Nucl. Sci. Eng. 160, 232 (2008) [CrossRef] [Google Scholar]
 G. Rimpault, D. Plisson, J. Tommasi, R. Jacqmin, The ERANOS code and data system for fast reactor neutronic analyses, in PHYSOR'02, Seoul, KOREA (2002) [Google Scholar]
 P. Archier, C. De Saint Jean, G. Noguere, O. Litaize, P. Leconte, C. Bouret, COMAC: nuclear data covariance matrices library for reactor applications, in PHYSOR 2014–The Role of Reactor Physics Toward a Sustainable Future, Kyoto, Japan (2014) [Google Scholar]
 M. Herman et al., COMMARA2.0 Neutron Cross Section Covariance Library, BNL948302011, 2011 [Google Scholar]
 A.D. Carlson et al., International evaluation of neutron cross section standards, Nucl. Data Sheets 110, 3215 (2009) [CrossRef] [Google Scholar]
 N. Otuka, E. Dupont, Towards a more complete and accurate Experimental Nuclear Reaction Data Library (EXFOR): International Collaboration Between Nuclear Reaction Data Centres (NRDC), Nucl. Data Sheets 120, 272 (2014) [CrossRef] [Google Scholar]
 Y. Danon et al., Simultaneous measurement of ^{235}U fission and capture cross sections from 0.01 eV to 3 keV using a gamma multiplicity detector, Nucl. Sci. Eng. 187, 291 (2017) [CrossRef] [Google Scholar]
 O. Iwamoto, R. McKnight, International Evaluation Cooperation Volume 29: Uranium235 Capture Crosssection in the keV to MeV Energy Region, NEA, NEA/WPEC29, 2011 [Google Scholar]
 M. Jandel et al., New precision measurements of the ^{235}U(n,g) cross section, Phys. Rev. Lett. 109, 202506 (2012) [CrossRef] [PubMed] [Google Scholar]
 JEFF3.3 Nuclear Data library can be downloaded on the NEA website: https://www.oecdnea.org/dbdata/JEFF33/ [Google Scholar]
 M.B. Chadwick et al., The CIELO collaboration: neutron reactions on ^{1}H, ^{16}O, ^{56}Fe, ^{235,238}U, and ^{239}Pu, Nucl. Data Sheets 118, 1 (2014) [CrossRef] [Google Scholar]
 A. Santamarina, Improvement of ^{238}U inelastic scattering cross section for an accurate calculation of large commercial reactors, Nucl. Data Sheets 118, 118 (2014) [CrossRef] [Google Scholar]
Cite this article as: Virginie Huy, Gilles Noguère, Gérald Rimpault, Use of integral data assimilation and differential measurements as a contribution to improve ^{235}U and ^{238}U cross sections evaluations in the fast and epithermal energy range, EPJ Nuclear Sci. Technol. 4, 41 (2018)
All Tables
Specifications on fuel enrichment and structural materials for the different configurations.
Impact on MASURCA 1B and FCAIX 1 to 3 critical masses when using carbon evaluation of JENDL4.0 instead of JEFF3.1.1.
Prior and posterior C/E values for PROFIL2A variation of concentrations ratios and associated uncertainties.
Relative impact on ASTRID critical mass of the trends suggested by assimilation when fissions spectra are set to JEFF3.1.1 evaluations. Only trends superior to posterior uncertainties were considered.
All Figures
Fig. 1
Critical mass C/Es compared with experimental uncertainties for Uranium configurations (using JEFF libraries). 

In the text 
Fig. 2
33group sensitivity profiles of several critical masses to ^{235}U capture. 

In the text 
Fig. 3
Sensitivity coefficients of FLATTOP^{235}U and GODIVA critical masses to ^{235}U capture and ^{238}U inelastic and capture cross sections. 

In the text 
Fig. 4
Comparison between prior (JEFF3.1.1) and posterior C/E values. 

In the text 
Fig. 5
Trends from assimilation work for ^{235}U capture (relative to JEFF3.1.1) compared with a priori and a posteriori nuclear data uncertainties. The two red dotted curves represent an envelope gathering all the trends suggested by assimilation results (that includes cases with fission spectra fitted or not, and with graphite evaluation from JEFF3.1.1 or JENDL4.0). 

In the text 
Fig. 6
Results of differential measurements from Danon et al. [14] for ^{235}U capture from 0.5 to 3 keV, compared with ENDF/BVII.1 and JENDL4.0 evaluations. 

In the text 
Fig. 7
Results of differential measurements from Jandel et al. [16] for ^{235}U capture from 3 keV to 1 MeV. Comparison with assimilation results applied to JEFF3.1.1 pointwise evaluations (red continuous line). 

In the text 
Fig. 8
33group assimilation results for ^{235}U capture compared with “a priori” JEFF3.1.1 and JEFF3.3t3 evaluation. Posterior uncertainties for assimilation results are in dotted line. 

In the text 
Fig. 9
Comparison of Jandel et al. measurements [16] to JEFF3.3t3 and JEFF3.1.1 evaluations for ^{235}U capture cross sections. 

In the text 
Fig. 10
Trends from assimilation work for ^{238}U capture (relative to JEFF3.1.1) compared with a priori and a posteriori nuclear data uncertainties. For both cases (fission spectra fitted or not), the two dotted lines have to be seen as an additional uncertainty associated to the choice of graphite evaluation. 

In the text 
Fig. 11
Comparison of sensitivity profiles of PROFIL2A C/E, and FCAIX 7 and MASURCA 1B critical masses to ^{235}U and ^{238}U capture. 

In the text 
Fig. 12
Trends from assimilation work for ^{238}U inelastic (relative to JEFF3.1.1) compared with a priori and a posteriori nuclear data uncertainties. For both cases (fission spectra fitted or not), the two dotted lines have to be seen as an additional uncertainty associated to the choice of graphite evaluation. 

In the text 
Fig. 13
33group assimilation results (case where fission spectra are not fitted and graphite evaluation used is from JEFF3.1.1) for ^{238}U inelastic compared with “a priori” JEFF3.1.1, CIELO and JEFF3.3t3 evaluation. Posterior uncertainties for assimilation results are in dotted line. 

In the text 
Current usage metrics show cumulative count of Article Views (fulltext article views including HTML views, PDF and ePub downloads, according to the available data) and Abstracts Views on Vision4Press platform.
Data correspond to usage on the plateform after 2015. The current usage metrics is available 4896 hours after online publication and is updated daily on week days.
Initial download of the metrics may take a while.