Nuclear Data Sheets最新文献

The absolute γ-ray branching ratios (emission probabilities) for 19 prominent peaks in the α decay of ²³³U are reported with energies ranging from 97 to 366 keV. The branching ratios were obtained from α-γ coincidence measurements, normalized to the α singles. The high-purity ²³³U material ($>99.98\text{\%}$) was separated from its decay products before preparation of a 4.74-$\text{mg}/{\text{cm}}^2$ uranyl nitrate foil with a 6-mm diameter. This is the first report on the absolute normalization of the branching ratios since 1984 and it is the only measure by α-γ coincidences, where the $\alpha \text{-}\gamma /\alpha$ ratio removes several potential systematic biases. Overall, the results are consistent with the adopted literature and they validate the quantities most relevant to the non-destructive assay of fissile ²³³U material, which is relevant to the thorium fuel cycle and associated safeguards.

The Fusion Evaluated Nuclear Data Library (FENDL) is a comprehensive and validated collection of nuclear cross section data coordinated by the International Atomic Energy Agency (IAEA) Nuclear Data Section (NDS). FENDL assembles the best nuclear data for fusion applications selected from available nuclear data libraries and has been under development for decades. FENDL contains sub-libraries for incident neutron, proton, and deuteron cross sections including general purpose and activation files used for particle transport and nuclide inventory calculations.

In this work, we describe the history, selection of evaluations for the various sub-libraries (neutron, proton, deuteron) with the focus on transport and reactor dosimetry applications, the processing of the nuclear data for application codes (e.g. MCNP), and the development of the TENDL-2017 library which is the currently recommended activation library for FENDL. We briefly describe the IAEA IRDFF library as the recommended library for dosimetry fusion applications. We also present work on validation of the neutron sub-library using a variety of fusion relevant computational and experimental benchmarks using the MCNP transport code and ACE-formatted cross section libraries. A variety of cross section libraries are used for the validation work including FENDL-2.1, FENDL-3.1d, FENDL-3.2, ENDF/B-VIII.0, and JEFF-3.2 with the emphasis on the FENDL libraries.

The results of the validation using computational benchmarks showed generally good agreement among the tested neutron cross section libraries for neutron flux, nuclear heating, and primary displacement damage (dpa). Gas production (H/He) in structural materials showed substantial differences to the reference FENDL-2.1 library. The results of the experimental validation showed that the performance of FENDL-3.2b is at least as good and in most cases better than FENDL-2.1.

Future work will consider improved evaluations developed by the International Nuclear Data Evaluation Network (INDEN) for materials such as O, Cu, W, Li, B, and F. Additionally, work will need to be done to investigate differences in gas production in structural materials. Covariance matrices will need to be developed or updated as availability of consistent and comprehensive uncertainty information will be needed as fusion technology and facility construction matures. Finally, additional validation work for high energy neutrons, protons and deuterons, as well as validation work for the activation library will be needed.

Precise normalization of ToF yield measurements of neutron induced fission in fissile targets is challenging, but the appropriate normalization is also of critical importance for nuclear energy and criticality safety, among other applications. A typical normalization relies on the Thermal Neutron Constants (TNC) recommended at the thermal point (fission ${\sigma }_f^0$ and capture ${\sigma }_{\gamma }^0$ thermal cross sections) by the Neutron Standards. However, many ToF experiments do not collect data down to the thermal energy and use normalization cross-section integrals defined at different arbitrary energy intervals. Normalization fission cross-section integrals $I_3$ are recommended in between-valleys energy regions 8.1–14.7 eV, 7.8–11 eV, 9–20 eV, and 11.7–19.5 eV for fissile targets ²³³U, ²³⁵U, ²³⁹Pu, and ²⁴¹Pu with values equal to 689.0(10.8), 245.7(4.1), 1059(6), and 1378(33) $b\cdot eV$, respectively. The ₂₃₅U normalization integral $I_3$ derived in this work of 245.7(4.1) $b\cdot eV$ is in good agreement within quoted uncertainties with the Neutron Standards value of 247.5(3.0) $b\cdot eV$.

Additional cross-section integrals $I_1$ in the thermal region from 20–60 meV are derived to fix both the normalization and the slope of the fission cross section at the thermal point providing additional constraints for R-matrix evaluations of experimental fission yields. Ratios ${\sigma }_f^0/I_1$ and $I_3/I_1$ feature very low uncertainty due to the strong positive correlations between the numerator and the denominator and are comprehensively derived for the first time. Integral ratios $I_3/I_1$ of 39.31(54), 13.08(20), 41.65(22), and 40.46(85) are recommended as reference for fissile targets ²³³U, ²³⁵U, ²³⁹Pu, and ²⁴¹Pu, respectively. Similarly, ratios