Boundary layer measurements for validating CFD condensation model and analysis based on heat and mass transfer analogy in laminar flow condition

Soma, Shu; Ishigaki, Masahiro*; Abe, Satoshi; Shibamoto, Yasuteru

Nuclear Engineering and Technology, 10 Pages, 2024/00

https://doi.org/10.1016/j.net.2024.02.011

Large-eddy simulation on gas mixing induced by the high-buoyancy flow in the CIGMA facility

Abe, Satoshi; Shibamoto, Yasuteru

Nuclear Engineering and Technology, 55(5), p.1742 - 1756, 2023/05

https://doi.org/10.1016/j.net.2023.01.019

Raman spectroscopy of eutectic melting between boride granule and stainless steel for sodium-cooled fast reactors

Fukai, Hirofumi*; Furuya, Masahiro*; Yamano, Hidemasa

Nuclear Engineering and Technology, 55(3), p.902 - 907, 2023/03

https://doi.org/10.1016/j.net.2022.11.010

This paper addresses reaction products and their distribution of the eutectic melting/solidifying reaction of boron carbide (BC) and stainless-steel (SS). The influence of the existence of carbon on the BC-SS eutectic reaction was investigated by comparing the iron boride (FeB)-SS reaction by Raman spectroscopy with Multivariate Curve Resolution (MCR) analysis. The scanning electron microscopy with dispersive X-ray spectrometer was also used to investigate the elemental information of the pure metals such as Cr, Ni, and Fe. In the BC-SS samples, a new layer was formed between BC/SS interface, and the layer was confirmed that the formed layer corresponded to amorphous carbon (graphite) or FeB or FeB. In contrast, a new layer was not clearly formed between FeB and SS interface in the FeB-SS samples.

Development and validation of fuel stub motion model for the disrupted core of a sodium-cooled fast reactor

Kawada, Kenichi; Suzuki, Toru*

Nuclear Engineering and Technology, 53(12), p.3930 - 3943, 2021/12

https://doi.org/10.1016/j.net.2021.06.018

To improve the capability of the SAS4A code, which simulates the initiating phase of core disruptive accidents for MOX-fueled Sodium-cooled Fast Reactors (SFRs), the authors have investigated in detail the physical phenomena under unprotected loss-of-flow (ULOF) conditions in a previous paper. As the conclusion of the last article, fuel stub motion, in which the residual fuel pellets would move toward the core central region after fuel pin disruption, was identified as one of the key phenomena to be appropriately simulated for the initiating phase of ULOF. In the present paper, a simple model describing fuel stub motion, which was not modeled in the previous SAS4A code, was newly proposed. The applicability of the proposed model was validated through a series of analyses for the CABRI experiments, by which the stub motion would be represented with reasonable conservativeness for the reactivity evaluation of disrupted core.

ROSA/LSTF test and RELAP5 analyses on PWR cold leg small-break LOCA with accident management measure and PKL counterpart test

Takeda, Takeshi; Otsu, Iwao

Nuclear Engineering and Technology, 49(5), p.928 - 940, 2017/08

https://doi.org/10.1016/j.net.2017.03.004

A Preliminary evaluation of unprotected loss-of-flow accident for a prototype fast-breeder reactor

Suzuki, Toru; Tobita, Yoshiharu; Kawada, Kenichi; Tagami, Hirotaka; Sogabe, Joji; Matsuba, Kenichi; Ito, Kei; Ohshima, Hiroyuki

Nuclear Engineering and Technology, 47(3), p.240 - 252, 2015/04

https://doi.org/10.1016/j.net.2015.03.001

International collaboration in assessment of radiological impacts arising from releases to the biosphere after disposal of radioactive waste into geological repositories

Smith, G.*; Kato, Tomoko

Nuclear Engineering and Technology, 42(1), p.1 - 8, 2010/02

Geological disposal is designed to provide safe containment of radioactive waste for very long times, with the containment provided by a combination of engineered and geological barriers. In the extreme long term, after many thousands of years or longer, residual amounts of long-lived radionulides such as Cl-36, but also radionuclides in the natural decay chains, may be released into the environment normally accessed and used by humans, termed here, the biosphere. It is necessary to ensure that any such releases meet radiation protection objectives through the development of a safety case, which will include assessment of radiation doses to humans. The design of such dose calculations over such long timeframes is not straightforward, because of the range of potentially relevant assumptions which could be made, concerning environmental change and changes in human behavior. The issue has therefore been subject to international cooperation for many years. This paper summarizes the evolution and results of that collaboration leading up to the present day, taking account of developments in international recommendations on radiation protection objectives and the more recent greater focus on preparation of site specific safety cases.

High temperature oxidation of Nb-containing Zr alloy cladding in LOCA conditions

Chuto, Toshinori; Nagase, Fumihisa; Fuketa, Toyoshi

Nuclear Engineering and Technology, 41(2), p.163 - 170, 2009/03

In order to evaluate high-temperature oxidation behavior of the advanced alloy cladding under LOCA conditions, isothermal oxidation tests in steam were performed with cladding specimens prepared from high burnup PWR fuel rods (79 MWd/kg). Cladding materials were M5 and ZIRLO which are Nb-containing alloys. Ring-shaped specimens were isothermally oxidized in flowing steam at temperatures from 1173 to 1473 K for the duration between 120 and 4000 s. Protective effect of the pre-formed corrosion layer is seen for the shorter time range at the lower temperatures. Influence of pre-hydriding is not significant for the examined range. Alloy composition change generally has small influence on the oxidation in the examined temperature range, though M5 shows obviously smaller oxidation constant at 1273 K. Consequently, the oxidation rates of the high burnup M5 and ZIRLO cladding are comparable or lower than that of unirradiated Zircaloy-4 cladding.

A Next generation sodium-cooled fast reactor concept and its R&D program

Ichimiya, Masakazu; Mizuno, Tomoyasu; Kotake, Shoji

Nuclear Engineering and Technology, 39(3), p.171 - 186, 2007/06

Critical issues in the development targets for the future fast reactor (FR) cycle system, including odium-cooled FR were to ensure safety assurance, efficient utilization of resources, reduction of environmental burden, assurance of nuclear non-proliferation, and economic competitiveness. A promising design concept of sodium-cooled fast reactor JSFR is proposed aiming at fully satisfaction of the development targets for the next generation nuclear energy system. A roadmap toward JSFR commercializationis described, to be followed up in a new framework of the Fast reactor Cycle Technology development (FaCT) Project launched in 2006.

Key R&D activities supporting disposal of radioactive waste; Responding to the challenges of the 21st century

Miyamoto, Yoichi; Umeki, Hiroyuki; Osawa, Hideaki; Naito, Morimasa; Nakano, Katsushi; Makino, Hitoshi; Shimizu, Kazuhiko; Seo, Toshihiro

Nuclear Engineering and Technology, 38(6), p.505 - 534, 2006/08

Ensuring sufficient supplies of clean, economic and acceptable energy is a critical global challenge for the 21st century. There seems little alternative to a greatly expanded role for nuclear power, but implementation of this option will depend on ensuring that all resulting wastes can be disposed of safely. Although there is a consensus on the fundamental feasibility of such disposal by experts in the field, concepts have to be developed to make them more practical to implement and, in particular, more acceptable to key stakeholders. By considering global trends and using illustrative examples from Japan, key areas for future R&D are identified and potential areas where the synergies of international collaboration would be beneficial are highlighted.