Neutron cross section measurements for BUC approaches

. Criticality safety analysis is required at various stages of the back-end of the fuel cycle, i.e. reprocessing, transport, storage and disposal of spent nuclear fuel (SNF). To account for the reduction in reactivity due to fuel burnup, the Burn-Up Credit (BUC) concept was introduced. Evidently, this concept depends on the quality of nuclear data, in particular the absorption cross sections of some key nuclides. A dedicated programme has been established at the GELINA facility of the JRC-Geel to produce accurate cross section data and validate the evaluated nuclear data libraries for neutron interactions with fission fragments that are relevant for a BUC approach. In this work, cross section data for 103 Rh and 155 Gd are presented and the results are compared with the main evaluation libraries, showing good agreement in the thermal energy region with ENDF/B-VIII.0 and JEFF-3.3, but not with JENDL-4.0 for 103 Rh.


Introduction
The need of criticality safety analysis at various stages of the nuclear fuel cycle, such as reprocessing, transport, storage and final disposal of spent nuclear fuel motivates the current work. Traditionally was assumed that spent fuel is as reactive as fresh fuel [1]. This is known as the fresh fuel assumption, and avoids a number of calculation and verification problems. However, it has a large impact on the efficiency.
Since the 80's, an effort is being made to use more realistic and less conservative estimates of the nuclear reactivity of SNF by accounting for the reduction in reactivity due to fuel burnup. This approach is known as Burn-Up Credit (BUC). BUC has been successfully applied to spent fuel storage pools, resulting in increased capacity and permitting the storage of spent fuel with higher initial enrichments.
The complexity of reactivity calculations that include BUC together with related nuclear data and validation requirements, significantly increases with the number of nuclides. The majority of the produced nuclides will have a negligible effect on the SNF reactivity due to their low quantity, small neutron reaction cross section and/or short half-live. Hence, a careful selection of nuclides that are import for criticality studies is required.
Different approaches are considered while selecting the nuclides of importance for reactivity studies including BUC: an approach that accounts for the depletion of net fissile nuclides, that is, considering changes in the concentrations of 235 U and 239,241 Pu; * Corresponding author: carlos.paradela-dobarro@ec.europa.eu an actinide-only approach, in which the concentrations of key fissile and neutron absorbing actinides are considered; an actinide plus fission products approach that considers changes in the concentration of key fissile nuclides and neutron absorbing actinides and fission products; and an approach that considers the full nuclide inventory.
Carmouze et al. [2] studied the interest of a BUC approach for the storage of LWR-MOX spent fuel in a pool. The nuclide inventory was determined for a total burnup of 92 GWd/t and a 5-year cooling time. For some nuclides the negative reactivity worth resulting from this study is reported in Table 1. These values represent an indication of the importance of these nuclides for criticality safety studies including BUC. The absorption cross sections of these nuclides not only determine the reactivity worth in spent fuel, but they also have an influence on the final nuclide inventory. Therefore, the accuracy of these neutron absorption cross sections is of primary importance for any criticality analysis based on a BUC approach.

Measurements at GELINA
To improve the status of cross sections for fission products that are important for criticality safety studies, a dedicated programme was defined as part of a collaboration between the CEA Cadarache and JRC-Geel. It includes experiments at the time-of-flight facility GELINA to characterise the samples that were used for the MINERVE experiments of Ref. Error! Reference source not found. and to produce total and capture cross section data to improve the evaluated data in the resonance region. This collaborative effort triggered the interest of other institutes and organisation such as the IFIN-HH (Romania), INFN Bologna (Italy), INRNE (Bulgary) and SungkyunKwan University (Republic of Korea) by participating in experiments at GELINA and assisting in the data reduction and analysis. GELINA is a neutron Time-Of-Flight (TOF) facility designed for high-resolution cross section measurements in the resonance region. It is a multi-user facility providing a white neutron beam from 10 meV up to 20 MeV for 12 flight paths with measurement stations that have special equipment for total and reaction cross section measurements in the resonance region. More information on the experimental capabilities of GELINA facility are discussed in Ref. [4]. For the BUC programme three stations have been used, a capture measurement station at FP5-10m and two transmission stations at FP13-10m and one at FP4-50m.

Capture measurements
Neutron capture measurements at FP5-10m use a setup based on two C6D6 liquid scintillators to detect photons emitted by neutron induced capture events. Applying the total energy detection principle in combination with the pulse height weighting technique [5], the efficiency for detecting a capture event becomes directly proportional to the sum of energies of the gamma-ray cascade, so independent of the decay mode. The energy dependent neutron fluence rate at the station is determined by a 10 Bloaded ionisation chamber.

Transmission measurements
In the transmission measurements, neutrons are detected using a 6 Li glass scintillator placed further away from the sample. The experimental transmission is obtained from the ratio of TOF spectra measured with and without the sample in beam, after subtracting the background contribution. This contribution is estimated by applying the black resonance filter technique [5]. Since the experimental transmission is a ratio of TOFspectra the dependence on the incoming neutron fluence rate and the detection efficiency cancels out. In, addition, the experimental transmission is directly related to the total cross section. Hence, transmission measurements are the most accurate cross section measurements.

Experimental data on fission product
From the list shown in the Table 1, experimental data have been produced at GELINA for 109 Ag [6], 99 Tc [7], 103 Rh and 155 Gd [8]. As data for the first two nuclides have already been published, we will focus on the results obtained for rhodium and gadolinium.

103 Rh measurements
For BUC the interest on 103 Rh is focused on the thermal region and the strong low energy resonances. Dilg and Mannhart [10] determined the capture cross section at thermal energy by transmission measurements. Their value (144.8 (7) b) is at present the most accurate value published in the literature. However, it can be biased due to a deviation from the 1/v dependence of the cross section, primarily due to the contribution of the 1.2 eV resonance. The correction applied by Dilg and Mannhart was +3.95 %; whereas from the parameters in JEFF.3.1 a correction of +4.00% is calculated. Applying this correction, the capture cross section at thermal energy becomes 144.9 (7). Most of the data reported are within the quoted uncertainties in agreement with Dilg and Mannhart. The large difference with the value obtained by Lee et al. [11] (133.0 (9) b) would suggest that their data suffer from a substantial bias effect.  We performed an experimental campaign at GELINA to provide transmission data in the resolved resonance energy region for different rhodium sample thicknesses. The results, together with a detailed description of the experiments [12,13], have been submitted to the EXFOR. The experimental transmissions derived from these measurements are in good agreement with those calculated with the JEFF-3.3 resonance parameters while they are discrepant with the transmissions using the parameters in JENDL-4. The latter are strongly based on the work of Lee et al. [10]. To crosscheck these results, capture cross section measurements were performed and data analysis is currently ongoing.  155 Gd has a very large absorption cross section for thermal neutrons, although smaller than that of 157 Gd. At present only two sets of energy dependent cross section measurements have been carried out to assess the resonance parameters for 155 Gd and 157 Gd in the resolved energy range. These are the transmission measurements of Møller et al [14], using enriched Gd2O3 powder samples in 1960 and capture and transmission measurements of Leinweber et al. [15], using natural metallic and liquid samples. In the EXFOR data library, only the data from Leinweber are available for a re-evaluation based on a resonance shape analysis.

155 Gd measurements
The study of the gadolinium cross sections at GELINA was carried out in collaboration with INFN Bologna. Two 155 Gd metallic samples of thicknesses 0.04 and 0.004 mm were measured in transmission. The thin sample was used to study the thermal neutron energy region, while the thick one the resolved resonance region above the strong low energy resonances. The data, together with a detailed description of the experiment in Ref. [16], were submitted to the EXFOR data library. The 155 Gd capture cross section was determined at the CERN n_TOF facility using the same sample [17]. The project is still ongoing, including the analysis of additional transmission and capture experiments using nat Gd and enriched 154 Gd samples performed at GELINA and n_TOF [18], respectively.

Summary and outlook
This work presents the measurements carried out at GELINA to study neutron absorption interaction cross sections with fission fragments that are of interest for BUC calculations, in particular, for 103 Rh and 155 Gd. The main objective is to verify and validate the corresponding data in the evaluated data libraries. Our results agree with the thermal neutron energy cross sections provided by ENDF/B-VIII.0 and JEFF-3.3, while are discrepant with JENDL-4.0 in the case of rhodium. These data together with the work of Noguere et al. [19] provide nuclear data that are essential for the implementation of BUC approaches for criticality safety analyses.