neutron diffraction revealing the achievement of excellent combination of strength and ductility in metastable austenitic steel by grain refinement

Mao, W.; Gong, W.; Harjo, S.; Morooka, Satoshi; Gao, S.*; Kawasaki, Takuro; Tsuji, Nobuhiro*

Journal of Materials Science & Technology, 176, p.69 - 82, 2024/03

https://doi.org/10.1016/j.jmst.2023.07.036

The yield stress of Fe-24Ni-0.3C (wt.%) metastable austenitic steel increased 3.5 times (158 551 MPa) when the average grain size decreased from 35 m (coarse-grained [CG]) to 0.5 m (ultrafine-grained [UFG]), whereas the tensile elongation was kept large (0.87 0.82). neutron diffraction measurements of the CG and UFG Fe-24Ni-0.3C steels were performed during tensile deformation at room temperature to quantitatively elucidate the influence of grain size on the mechanical properties and deformation mechanisms. The initial stages of plastic deformation in the CG and UFG samples were dominated by dislocation slip, with deformation-induced martensitic transformation (DIMT) also occurring in the later stage of deformation. Results show that grain refinement increases the initiation stress of DIMT largely and suppresses the rate of DIMT concerning the strain, which is attributed to the following effects. (i) Grain refinement increased the stabilization of austenite and considerably delayed the initiation of DIMT in the 111//LD (LD: loading direction) austenite grains, which were the most stable grains for DIMT. As a result, most of the 111//LD austenite grains in the UFG specimens failed to transform into martensite. (ii) Grain refinement also suppressed the autocatalytic effect of the martensitic transformation. Nevertheless, the DIMT with the low transformation rate in the UFG specimen was more efficient in increasing the flow stress and more appropriate to maintain uniform deformation than that in the CG specimen during deformation. The above phenomena mutually contributed to the excellent combination of strength and ductility of the UFG metastable austenitic steel.

Double-differential cross sections for charged particle emissions from particle impinging on Al at 230 MeV/u

Furuta, Toshimasa*; Uozumi, Yusuke*; Yamaguchi, Yuji; Iwamoto, Yosuke; Koba, Yusuke*; Velicheva, E.*; Kalinnikov, V.*; Tsamalaidze, Z.*; Evtoukhovitch, P.*

Journal of Nuclear Science and Technology, 61(2), p.230 - 236, 2024/02

https://doi.org/10.1080/00223131.2023.2294192

Charged particle production from particle fragmentation reactions was investigated experimentally by measurement of 230-MeV/u particles bombarding an aluminum target. Double differential cross sections were measured for each ejectile of p, d, t, He, and He at laboratory angles between 15 and 60 deg. The results of analyzed data found the following common characteristics: (1) spectra of proton- and neutron-emission are similar in high energy region at forward angle, (2) triton-to-He ratio of -breakup yield is 1:2, which is similar to lower incident energy experiment, and (3) the shape of broad peak formed by He and particles could be explained by the process with collision between induced particle and target nucleus.

Uncertainty reduction of sodium void reactivity using data from a sodium shielding experiment

Maruyama, Shuhei; Endo, Tomohiro*; Yamamoto, Akio*

Journal of Nuclear Science and Technology, 61(1), p.31 - 43, 2024/01

https://doi.org/10.1080/00223131.2023.2244512

This study investigated the feasibility of reducing the uncertainty associated with fast-reactor-core design by sharing an experimental database between different fields (e.g., reactor physics and radiation shielding) using data assimilation techniques. As the first step in this study, we focused on the ORNL sodium shielding experiment and investigated the possibility of using the experimental data to reduce the uncertainty in sodium void reactivity (SVR), which is the most important safety parameter for sodium-cooled fast reactors. A sensitivity analysis based on the Generalized Perturbation Theory was performed for the sodium shielding experiment. Using the sensitivity coefficients evaluated here and those of the sodium void reactivity previously evaluated by the JAEA, we showed that sodium shielding experimental data can contribute to the uncertainty reduction of SVR by adopting the cross-section adjustment method. Based on this study, the uncertainty reduction effect is expected to be significant, especially for SVR dominated by neutron-leakage phenomena. Although new reactor physics experimental data on SVR may be difficult to obtain, the results of this study suggest that data from sodium shielding experiments can partially substitute for this role. This study demonstrated the value of the mutual use of integral experimental data in fast reactor designs.

Measurement of the neutron capture cross section of Re in the keV energy region

Katabuchi, Tatsuya*; Sato, Yaoki*; Takebe, Karin*; Igashira, Masayuki*; Umezawa, Seigo*; Fujioka, Ryo*; Saito, Tatsuhiro*; Iwamoto, Nobuyuki

Journal of Nuclear Science and Technology, 6 Pages, 2024/00

https://doi.org/10.1080/00223131.2023.2298749

A Study on the fracture pattern change of high-burnup fuel cladding failed by pellet-cladding mechanical interaction failure under reactivity-initiated accident conditions

Li, F.; Mihara, Takeshi; Udagawa, Yutaka

Journal of Nuclear Science and Technology, 11 Pages, 2024/00

https://doi.org/10.1080/00223131.2024.2313553

Simulated performance evaluation of d-Be compact fast neutron source

Nakayama, Shinsuke

Journal of Nuclear Science and Technology, 60(12), p.1447 - 1453, 2023/12

https://doi.org/10.1080/00223131.2023.2225498

The d+Be neutron source is a candidate for transportable neutron source for on-site nondestructive inspection of infrastructure facilities such as bridges, tunnels and so on. The applicability of the d+Be neutron source to a transportable fast neutron source is explored by Monte Carlo particle transport simulations with PHITS and JENDL-5. The simulation results show that by increasing the shielding thickness by about 1.5 times, it is possible to realize the d+Be neutron source with the comparable performance to another candidate, the 2.5-MeV p+Li neutron source, at lower beam energy.

Development of new treatment of fuel isotope vector in the core disruptive accident analysis of fast reactors

Tagami, Hirotaka; Ishida, Shinya; Tobita, Yoshiharu

Journal of Nuclear Science and Technology, 60(12), p.1548 - 1562, 2023/12

https://doi.org/10.1080/00223131.2023.2224333

In a design of future Sodium-cooled Fast Reactor, there is a demand for evaluation of sequences and consequences of core disruptive accidents. Future SFRs include a unique core design with axially or horizontally heterogeneous core arrangement having complex fuel isotope distribution. A new model to flexibly represent fuel isotope distribution, called the Pu-vector model, has been developed in this study for inclusion in the SIMMER-III and SIMMER-IV codes (simply called as SIMMER). The model calculates movement of individual fuel isotopes, assuming they always accompany the convecting fuel in the fluid-dynamics model. The accuracy of the Pu-vector model was confirmed by comparing with the standard Monte Carlo static neutronics calculation. The new model can improve some of the limitations in the current SIMMER code, in which the fuel isotopes are represented only by two groups, fertile and fissile fuels. Assignment of a number of fuel isotopes to the two groups requires a detailed examination of different combinations of fuel isotopes to determine an optimized combination. The Pu-vector model can eliminate this complicated procedure to be performed prior to a SIMMER analysis, and more importantly provides accurate spatial distribution of fuel isotopes and thus will improve the applicability of SIMMER to the analyses of future large heterogeneous reactors.

Development of an integrated non-destructive analysis system, Active-N

Tsuchiya, Harufumi; Toh, Yosuke; Ozu, Akira; Furutaka, Kazuyoshi; Kitatani, Fumito; Maeda, Makoto; Komeda, Masao

Journal of Nuclear Science and Technology, 60(11), p.1301 - 1312, 2023/11

https://doi.org/10.1080/00223131.2023.2192529

Sintering and microstructural behaviors of mechanically blended Nd/Sm-doped MOX

Hirooka, Shun; Horii, Yuta; Sunaoshi, Takeo*; Uno, Hiroki*; Yamada, Tadahisa*; Vauchy, R.; Hayashizaki, Kohei; Nakamichi, Shinya; Murakami, Tatsutoshi; Kato, Masato

Journal of Nuclear Science and Technology, 60(11), p.1313 - 1323, 2023/11

https://doi.org/10.1080/00223131.2023.2192725

Additive MOX pellets are fabricated by a conventional dry powder metallurgy method. NdO and SmO are chosen as the additive materials to simulate the corresponding soluble fission products dispersed in MOX. Shrinkage curves of the MOX pellets are obtained by dilatometry, which reveal that the sintering temperature is shifted toward a value higher than that of the respective regular MOX. The additives, however, promote grain growth and densification, which can be explained by the effect of oxidized uranium cations covering to a pentavalent state. Ceramography reveals large agglomerates after sintering, and Electron Probe Micro-Analysis confirms that inhomogeneous elemental distribution, whereas XRD reveals a single face-centered cubic phase. Finally, by grinding and re-sintering the specimens, the cation distribution homogeneity is significantly improved, which can simulate spent nuclear fuels with soluble fission products.

Convergence behavior of statistical uncertainty in probability table for cross section in unresolved resonance region

Tada, Kenichi; Endo, Tomohiro*

Journal of Nuclear Science and Technology, 60(11), p.1397 - 1405, 2023/11

https://doi.org/10.1080/00223131.2023.2204102

The probability table method is a well-known method for addressing self-shielding effects in the unresolved resonance region. A long computational time is required to generate the probability table. The effective way to reduce the generation time of the probability table is the reduction of the number of ladders. The purpose of this study is the estimation of the optimal number of ladders using the statistical uncertainty in the probability table. To this end, the statistical uncertainty quantification method of the probability table was developed and the convergence behavior of the statistical uncertainty was investigated. The product of the probability table and the average cross section was considered the target of the statistical uncertainty. The convergence rate was affected by the average level spacing and reduced neutron width. The generation time of the probability table was less than half when the input parameter was changed from the number of ladders to the tolerance value.