Improving nuclear data accuracy of 241 Am and 237 Np capture cross sections

In the framework of the OECD/NEA WPEC subgroup 41, ways to improve neutron induced capture cross sections for 241Am and 237Np are being sought. Decay data, energy dependent cross section data and neutron spectrum averaged data are important for that purpose and were investigated. New timeof-flight measurements were performed and analyzed, and considerable effort was put into development of methods for analysis of spectrum averaged data and re-analysis of existing experimental data.


Introduction
There is a serious gap between required accuracy and current accuracy on nuclear data for developing innovative nuclear reactor systems.To bridge this gap, an international joint activity entitled "Improving Nuclear Data Accuracy of 241 Am and 237 Np capture cross sections (INDA)" is performed under WPEC.In this joint study, the forefront knowledge of energy dependent data, spectrum averaged data, relevant decay data, and evaluations are intended to be integrated on the capture cross sections of 237 Np and 241 Am.Each kind of data by state-of-art technique has been reviewed at first, and key issues on systematic errors have been identified for each kind of data to be solved to bridge the gap.
It was recognized that the forefront knowledge of each kind of data are valuable each other for improving data accuracy.Furthermore, key issues have been identified for each kind of data toward further improvement of nuclear data.In chapter 2-4, the forefront knowledge of each kind of data and key issues on systematic errors are summarized.In chapter 5, an example of successful achievement is described obtained by integrating the forefront knowledge.In chapter 6, the benefit of these collaborative works is discussed from the view point of evaluating nuclear data.a e-mail: gasper.zerovnik@ec.europa.eu

Current status of decay data
The relevant nuclear structure data for 237 Np and 241 Am measurements, are decay data for 233 Pa, 237,238 Np, 238 Pu, and 241,242g,242m Am, 242 Cm, respectively.The main databases for radioactive decay data are ENSDF [1] and DDEP [2].In Tables 1 and 2, most prominent X-, γ -ray, and α lines are summarized.Candidates for the emissions for activation experiments were summarized, and uncertainties of the current emission probabilities were explained in the framework of the WPEC subgroup 41.The difficulty due to the contaminated emissions will be a candidate for unrecognized systematic error.

Current status of energy dependent cross section measurements
Recent energy dependent cross section measurements were reviewed from GELINA, DANCE, n TOF, and ANNRI (Fig. 1 for 241 Am(n,γ )).Origins of systematic errors and important correction factors have been systematically identified, for example, sample impurity, sample amount, flux, detection efficiency, neutron selfshielding & multiple-scattering factors, and normalization.The importance of deducing the Westcott g-factor from energy dependent data was also pointed out, which is reflected in the spectrum averaged data.
The derived values for thermal (0.0253 eV) capture cross section for 241

Current status of spectrum averaged experiments
Reported results from integral measurements of 241 Am neutron capture cross sections over the last 60 years scatter by >40% as can be seen from Fig. 2. A source of systematic bias could not be allocated, hence a concerted action for improving of nuclear data has been initiated with the scope to augment accuracy and lower uncertainty substantially.The major underlying reasons for bias in 241 Am capture cross sections are low energy resonances and bound levels with variable influence to the measurement technique applied.

Feedback to spectrum averaged experiments
A systematic study was performed to estimate the possible biases in thermal capture cross section values derived from neutron activation measurements using conventional methods such as the k 0 standardization method [7] and the Westcott convention [8] as implemented e.g., in Refs.[9,10].Both methods are crude since they approximate the neutron spectrum ϕ(E) with only two free parameters (three in case variations in spectrum temperature are allowed).A general method based on Monte Carlo (MC) calculated reactor spectra (Fig. 3) and JEFF-3.2cross section library was used to calculate reference reaction rates for the measured reactions and standards ( 197 Au(n,γ ), 59 Co(n,γ ), etc.), irradiated bare and under Cd/Gd filters.Due to limited space, results for 237 Np and 241 Am only are presented here (Table 3 and Table 4, respectively).Adopting the reaction rates from Table 3 and Table 4 as reference "experimental" data, k 0 and Westcott methods have been used to derive the capture cross section σ 0 at thermal energy (25.3 meV).The biases produced due to the limitations in methodologies are presented in Table 5 and Table 6.
For 241 Am, a systematic trend of significant overestimation of σ 0 for activation method using Cd filters with an effective cut-off energy of around 0.55 eV can be  observed.The magnitude overestimation mainly depends on the ratio of the thermal spectrum peak to the epithermal spectrum component.This is due to inability of the analytical methods to take into account the contribution of the epithermal neutron spectrum component to the reaction rate between about 0.1 eV and 0.55 eV, which is for most nuclides insignificant.However the latter for 237 Np and especially 241 Am it is very important due to low-energy resonances below and around 0.5 eV (Fig. 4).Expectedly, the bias is larger if the epithermal component is stronger (corresponding to a lower thermal Maxwellian peak).With the use of Gd filter (effective cut-off energy around 0.1 eV), the problem of the bias is practically nonexistent.This method was used e.g., by Nakamura [9].An independent study [11], based on analytical corrections to the Westcott formalism due to deviation of the neutron spectrum from idealized shape, yields similar correction factors for 241 Am(n,γ ) cross section at 25.3 meV as the above described numerical method.The results of both studies applied to activation measurements on 241 Am are summarized in Table 7.The results in the column "MC corrected σ 0 " were also corrected for the deviation of the transmission filter from ideal and the generalized Westcott factor.The MC correction relies on JEFF-3.2library, while the corrections in Ref. [11] rely on JENDL-4.0 library.

Feedback to evaluation
From the joint study, it was shown that integration of knowledge from the independent specialities enables us not only to crosscheck between data obtained by different techniques but also to improve the measurement accuracy of each other.
The main outcome of the study in terms of feedback to evaluations will be to give recommendations to JEFF, JENDL and ENDF projects in order to take into account recent findings of the WPEC SG-41 and implement them into the new releases of the nuclear data libraries.

Conclusions
Neutron induced capture cross sections of 237 Np and 241 Am were studied in order to improve the quality of nuclear data.A detailed look into available experimental data was performed, including time-of-flight, reactor activation data, integral experiments and measurement in cold neutron beams.For a consistent analysis of experimental data, reliable and accurate decay data are also required.
Re-analysis of the existing activation data in reactor spectra is currently under way in order to revise the value of the neutron induced cross section of 241 Am at thermal energy.Combined with TOF measurements and measurements with cold neutrons, a new recommended value will be presented as one of the main results of the WPEC SG-41 activities.
This contribution presents the outline on the international joint activity coordinated via the OECD/NEA Working Party on Evaluation Cooperation (WPEC) Subgroup 41.

Figure 1 .
Figure 1.Comparison of recent measurements of neutron capture cross sections for 241 Am, normalized to 749 b at 0.0253 eV.

Figure 2 .
Figure 2. 241 Am thermal neutron capture cross sections as calculated from measurements performed at the cold PGAA beam in Garching, FRM II, along with other values from literature.The ENDF/B-VII.1 value is given for comparison (line at 684.3 b).

Figure 4 .
Figure 4. Comparison of Monte Carlo calculated (reference) and approximate (k 0 and Westcott convention) neutron spectra for a typical thermal research reactor irradiation channel (JSI TRIGA IC40).The differences in spectra clearly overlap with the low energy resonances in 241 Am and 237 Np neutron induced cross sections inducing biases in the derivation of the thermal cross section from the reaction rates.

Table 1 .
Main decay data, important for 237 Np measurements.

Table 2 .
Main decay data, important for 241 Am measurements.

Table 3 .
Reaction rates R for the 237 Np(n,γ ) reaction from normalized MC spectra and assuming JEFF-3.2cross sections.

Table 4 .
Reaction rates for the 241 Am(n,γ ) reaction from normalized MC spectra and assuming JEFF-3.2cross sections.
Figure 3. Monte Carlo calculated spectra in irradiation channels of JSI TRIGA (below) and KUR (above) reactors.

Table 5 .
Biases in the derived σ 0 for 237 Np(n,γ ), relative to the thermal reference value 181 b due to approximations in methodologies.

Table 6 .
Biases in the derived σ 0 for 241 Am(n,γ ), relative to the thermal reference value 748 b due to approximations in methodologies.

Table 7 .
Original and corrected values for 241 Am(n,γ ) cross section at 25.3 meV from neutron activation measurement.