Gigacycle fatigue strength evaluation of welded 316L stainless steels for mercury target vessel

Naoe, Takashi; McClintock, D.*

JPS Conference Proceedings (Internet), 45, p.011169_1 - 011169_7, 2026/06

https://doi.org/10.7566/JPSCP.45.011169

Mercury target vessel for the pulsed spallation neutron source suffers two kinds of cyclic stress during operation. One is the thermal stress due to the internal heating and swings by proton beam trip. The other is the impulsive stress by the pressure waves generated by the proton beam injection. The total number of loading cycles for the former is , and the later is for 2 years operation in the J-PARC mercury target vessel. The target vessel is made of 316L stainless steels and assembled by electron beam welding (EBW) and gas tungsten arc welding (GTAW). However, fatigue data of welded 316L stainless steels up to gigacycle is insufficient. Ultrasonic fatigue test, applying load cycle by utilizing ultrasonic resonance, for the welded metals were performed to investigate the effect of welding on gigacycle fatigue behavior. Fatigue strength of EBW and EB+GTAW specimens including weld bead effect will be discussed.

An Analytical study on Russia's civilian nuclear energy utilization

Kawasaki, Nobuchika

JAEA-Review 2025-043, 74 Pages, 2025/10

JAEA-Review-2025-043.pdf:2.45MB

Russia is one of the most advanced countries in the civilian use of nuclear energy. However, understanding the internal mechanisms of its nuclear program remains difficult due to various reasons. Therefore, this study presents a historical overview of Russia's nuclear energy utilization, fuel supply, fuel manufacturing capabilities, and concepts regarding reprocessing and the nuclear fuel cycle. From this overview, insights have been extracted and analyzed. These insights are then organized under two strategic perspectives: "Strategic diversity and continuity in developments and demonstrations" and "Diversity in utilizations and deployments," with considerations of implications for Japan, as below. Russia's nuclear energy policy strategically utilizes a variety of reactor types and fuel cycle technologies to expand nuclear power generation both domestically and internationally. Currently, nuclear power, centered on light-water reactors (VVER series), accounts for about 20% of Russia's electricity supply, and there are plans to increase this share to 25% by 2045. A wide range of reactors, from large-scale to medium and small modular reactors, are being constructed in Russia. Russia is also actively developing fast reactor technologies, and focusing on the reprocessing and recycling of spent fuel. Internationally, VVER-1200 reactors are under construction in several countries, and cooperation with China is deepening in the field of fast reactors. Notably, Russia offers an integrated, or selectively customizable, package of nuclear technology services on the international stage. These include not only reactor deployment, but also fuel supply, reprocessing, waste management, and even the provision of radioisotopes. Rather than simply exporting products or technology, Russia fosters long-term relationships and trust by flexibly responding to the conditions and needs of partner countries. For this reason, Russia promotes the technology developments in advance within the country in areas anticipated for future overseas deployment. It carefully selects target technologies and services and systematically rolls them out. This flexible strategy, combining "technological diversity" and "strategic consistency", enables cooperation with countries across various geopolitical contexts. For Japan, this strategic approach offers valuable lessons on how to engage in comprehensive international nuclear cooperation, not merely through technology exports, but through integrated approaches that encompass the entire fuel cycle, and by combining elements such as fast reactors and RI supply.

Initial benchmark comparison of the open-source Cyclus and NMB fuel cycle simulators

Bachmann, A. M.*; Nishihara, Kenji; Richards, S.*; Abe, Takumi; Feng, B.*

Progress in Nuclear Science and Technology (Internet), 8, p.11 - 16, 2025/09

https://doi.org/10.15669/pnst.8.11

Scenario analysis of future nuclear energy use in Japan, 3; Promotion of Plutonium utilization by RBWR-Backfit

Nishihara, Kenji; Oizumi, Akito; Hino, Tetsushi*; Soneda, Hideo*

Progress in Nuclear Science and Technology (Internet), 7, p.305 - 310, 2025/03

https://doi.org/10.15669/pnst.7.305

Hitachi-GE is conducting research and development of new fuel design named resource-renewable boiling water reactor-backfit (RBWR-BF) to enhance plutonium utilization in BWRs. In this study, the effects of installing RBWR-BF were estimated using the NMB code, a nuclear fuel cycle simulator, under the assumption of a future nuclear energy utilization scenario based on light water reactor (LWR) in Japan. As a result, it was shown that burden of MOX loading to LWRs can be minimized and fissionability of remaining plutonium in 2100 was improved by RBWR-BF introduction from 2040.

Initial verification of Cyclus and NMB fuel cycle simulators

Bachmann, A. M.*; Richards, S.*; Feng, B.*; Nishihara, Kenji; Abe, Takumi

Proceedings of International Conference on Nuclear Fuel Cycle (GLOBAL2024) (Internet), 4 Pages, 2024/10

This work demonstrates the value of code verification as an initial step in utilizing fuel cycle simulation. Cyclus and NMB are open-source fuel cycle simulators that provide computational modeling of nuclear fuel cycle alternatives and were chosen by Argonne National Laboratory and Japan Atomic Energy Agency (JAEA), respectively, for a multi-year collaboration on fuel cycle benchmarks. Both are relatively new and can be improved after conducting a rigorous code-to-code comparison. Initial verification of these simulators was performed using a set of hypothetical scenarios for once-through and multi-recycle fuel cycles. The results of this work identify how differences in scenario definitions and the modeling methodologies of the two simulators lead to differences in results in material inventories, mass flows, and other important metrics for fuel cycle assessments.

Neutron flux estimation and neutronics characteristics calculation in post-JMTR conceptual study

Oizumi, Akito; Akie, Hiroshi

JAEA-Technology 2023-017, 93 Pages, 2023/12

JAEA-Technology-2023-017.pdf:8.45MB

After the decision of decommissioning JMTR (Japan Materials Testing Reactor), Japan Atomic Energy Agency investigated the possibility to construct a new irradiation test reactor to succeed JMTR (post-JMTR), and the final report of the investigated result was submitted to the Ministry of Education, Culture, Sports, Science and Technology on March 30th 2021. This investigation was carried out in 4 steps of (1) selection of reactor type, (2) reactor core plans studies, (3) neutronic studies, (4) thermal studies, and was finally (5) considered and evaluated. This JAEA-Technology report summarizes the process and the results of (3) neutronic studies. Neutron fluxes were calculated at irradiation sample positions in the investigated cores, the standard core and the compact core, and the calculated fluxes satisfied the required irradiation capability. It was also evaluated the two investigated cores' continuous reactor operation time in days in one refueling cycle, and the results guaranteed an operation days equality with that of existing JMTR. In addition, neutronic characteristics of the cores were estimated, such as power distribution in the core, control rod reactivity worth, reactivity coefficients, distribution of fuel burnup rate of each fuel element, and kinetics parameters. The evaluated neutronic characteristics were used in the post-JMTR final investigation report to confirm the neutronic feasibility by comparing with the neutronic limiting values of existing JMTR, and to estimate the cooling capability to make the core thermally feasible.

Reactivity estimation based on the linear equation of characteristic time profile of power in subcritical quasi-steady state

Yamane, Yuichi

Journal of Nuclear Science and Technology, 59(11), p.1331 - 1344, 2022/11

https://doi.org/10.1080/00223131.2022.2053221

The reactivity was estimated from a time profile of neutron count rate or a simulated data in a quasi-steady state after sudden change of reactivity or external neutron source strength. The estimation was based on the equation of power in subcritical quasi-steady state. The purpose of the study is to develop the method of timely reactivity estimation from complicated time profile of neutron count rate. The developed method was applied to the data simulating neutron count rate created by using one-point kinetics code, AGNES, and Poisson-distributed random noise and to the transient subcritical experiment data measured by using TRACY. The result shows that the difference of the estimated and reference value was within about 5% or less for ( -1) for simulated data and within about 7% or less for -1.4 and -3.1 for the experimental data. It was also shown that the possibility of the reactivity estimation several ten seconds after the status change.

Water experiments on thermal striping phenomena at the core outlet of an advanced sodium-cooled fast reactor, 2; Proposal of countermeasures to mitigate temperature fluctuations around radial blanket fuel assemblies

Kobayashi, Jun; Aizawa, Kosuke; Ezure, Toshiki; Kurihara, Akikazu; Tanaka, Masaaki

Hozengaku, 20(3), p.97 - 101, 2021/10

Focusing on the thermal striping phenomena that occurs at a bottom of the internal structure of an advanced sodium-cooled fast reactor (Advanced-SFR) that has been designed by the Japan Atomic Energy Agency, a water experiment using a 1/3 scale 60 degree sector model simulating the upper plenum of the Advanced-SFR has been conducted to examine countermeasures for the significant temperature fluctuation generated around the bottom of Upper Internal Structure (UIS). In the previous paper, we reported the effect of measures to mitigate temperature fluctuations around the control rod channels. In this paper, the same test section was used, and a water experiment was conducted to obtain the characteristics of temperature fluctuations around the radial blanket fuel assembly. And the shape of the Core Instrumentation Support Plate (CIP) was modified, and it was confirmed that it was highly effective in alleviating temperature fluctuations around the radial blanket fuel assembly.

In situ diffraction characterization on microstructure evolution in austenitic stainless steel during cyclic plastic deformation and its relation to the mechanical response

Kumagai, Masayoshi*; Akita, Koichi*; Kuroda, Masatoshi*; Harjo, S.

Materials Science & Engineering A, 820, p.141582_1 - 141582_9, 2021/07

https://doi.org/10.1016/j.msea.2021.141582

Effect of nitrogen oxides on decomposition behavior of gaseous ruthenium tetroxide

Yoshida, Naoki; Amano, Yuki; Ono, Takuya; Yoshida, Ryoichiro; Abe, Hitoshi

JAEA-Research 2020-014, 33 Pages, 2020/12

JAEA-Research-2020-014.pdf:3.66MB

Considering the boiling and drying accident of high-level liquid waste in fuel reprocessing plant, Ruthenium (Ru) is an important element. It is because Ru would form volatile compounds such as ruthenium tetroxide (RuO) and could be released into the environment with other coexisting gasses such as nitric oxides (NOx) such as nitric oxide (NO) and nitrogen dioxide (NO). To contribute to the safety evaluation of this accident, we experimentally evaluated the effect of NOx on the decomposition and chemical change behavior of the gaseous RuO (RuO(g)). As a result, the RuO(g) decomposed over time under the atmospheric gasses with NO or NO, however, the decomposition rate was slower than the results of experiments without NOx. These results showed that the NOx stabilized RuO(g).