Characteristics in trace elements compositions of tephras (B-Tm and To-a) for identification tools

Nara, Fumiko*; Yokoyama, Tatsunori; Yamasaki, Shinichi*; Minami, Masayo*; Asahara, Yoshihiro*; Watanabe, Takahiro; Yamada, Kazuyoshi*; Tsuchiya, Noriyoshi*; Yasuda, Yoshinori*

Geochemical Journal, 55(3), p.117 - 133, 2021/00

https://doi.org/10.2343/geochemj.2.0619

The absolute date of the Millennium Eruption (ME) of Changbaishan Volcano is widely recognized as AD 946. The Baegdosan-Tomakomai (B-Tm) tephra dispersed during the ME is a robust-age key bed. In order to identify the tephra, refractive index and major-element compositions of volcanic glass shards are conventionally used. However, trace-element analysis has been rarely carried out, especially for rare-earth elements (REEs) and for tephra layer bulk sediments. Here we present the datasets of major- and trace-element compositions datasets for the glass shards and bulk sediments of the B-Tm and Towada caldera eruptions (To-a) tephra deposits from the Lake Ogawara sediment core, Tohoku region, northern Japan. The depth profiles of the major and trace elements show the significant peaks for the KO and some trace elements (Zn, Rb, Zr, Nb, Sn, Y, La, Ce, Nd, Th, and U) at the B-Tm tephra layer in the Lake Ogawara sediment core, but no peaks of these elements at the To-a tephra layer. High concentrations of the trace elements in the B-Tm tephra layer were observed in individual glass shards as well as in the bulk sediment. These concentrations are highlighted by the elemental abundance pattern normalized by the crustal abundance. The elemental pattern in individual glass shards from other Japanese tephras showed significant differences from those of the B-Tm tephra, especially in REEs compositions. The trace-element compositions of the glass shards and bulk sediment show strong advantages for distinguishing the B-Tm tephra from other Japanese tephras.

Trial visualization of fast reactor design knowledge

Yoshikawa, Shinji; Minami, Masaki*; Takahashi, Tadao*

Journal of Nuclear Science and Technology, 48(4), p.709 - 714, 2011/04

https://doi.org/10.3327/jnst.48.709

In design problems of large-scale systems like fast breeder reactors, inter-relations among design specifications are very important where a selected specification option is transferred to other specification selections as a premise to be taken account in engineering judgments. These inter-relations are also important in design case studies with hypothetical adoption of rejected design options for evaluation of deviation propagations among design specifications. Some of these rejected options have potential worth for future reconsideration by some circumstance changes (e.g., advanced simulations to exclude needs for mock-up tests, etc), to contribute for flexibility in system designs. In this study, a computer software is built to visualize design problem structure by representing engineering knowledge nodes on individual specification selections along with inter-relations of design specifications, to validate the knowledge representation method and to derive open problems.

Data description for coordinated research project on benchmark analyses of sodium natural convection in the upper plenum of the Monju reactor vessel under supervisory of Technical Working Group on Fast Reactors, International Atomic Energy Agency

Yoshikawa, Shinji; Minami, Masaki*

JAEA-Data/Code 2008-024, 28 Pages, 2009/01

JAEA-Data-Code-2008-024.pdf:5.83MB

A series of information required for numerical simulation of sodium thermal stratification observed at the plant trip test of "Monju" conducted in 1995 is provided, which consists of the test outline, geometry data of the reactor vessel upper plenum between the reactor core top and reactor outlet nozzles, and flow inlet boundary conditions at the reactor core top surface.

Development of "FBR plant engineering system"

Minami, Masaki*; Sakata, Hideaki; Yoshikawa, Shinji; Yamada, Fumiaki

JNC TN4410 2005-001, 123 Pages, 2005/03

JNC-TN4410-2005-001.pdf:6.81MB

A software system for straightforward and quick conceptual studies and technical evaluations of fast breeder reactor plants has been developed, mainly targeting the Japanese Demonstration Fast Breeder Reactor Monju. The studies and evaluations by this system used to be limited within steady and nominal conditions, excluding influences by changing specification values in accidental conditions. In this fiscal year, a new software component has been included in the system for simplified evaluation of transient characteristics, which is an essential for complete design of an FBR plant. This new evaluation function enabled to detect specifications to vary over acceptable range in transient conditions, and to notify necessity for re-adjustment of steady state design specification values. This system was also utilized to generate a sensitivity survey program in order to evaluate appropriateness of Monju design. The appropriateness of Monju design was evaluated in two ways. The first one is to follow specification selecting sequence as the as-built Monju has actually been designed. The second one is hypothetical deviations of major specifications of Monju and observation of the influences on other specifications. As a result, Monju design was confirmed to be adequate from the view point of need to meet design limitations at the design stage of Monju. This system is expected to be systematically re-arranged, because the system has now considerably complicated configuration after additions of various programs for many objects. This arrangement will facilitate future contribution of this system to technical studies in order for development of FBRs.

Data description for the second Research Coordination Meeting (RCM) of the IAEA Coordinated Research Project (CRP) on "Benchmark Analyses of Sodium Natural Convection in the Upper Plenum of the MONJU Reactor Vessel"

Yoshikawa, Shinji; Minami, Masaki*

no journal, , 

This document provides a set of technical data for the thermal hydraulic analysis of liquid sodium in the upper plenum of Monju reactor vessel, consisting of the additional information which JAEA stated to be offered at first RCM held in September 2008 in Vienna, and the table of temperature versus time measured by the vertical array of thermocouples inserted in the reactor upper plenum during the turbine trip test of the system start-up tests conducted in December 1995.

Validation of natural circulation heat removal evaluation method by using EBR-II shutdown heat removal test data

Doda, Norihiro; Igawa, Kenichi*; Minami, Masaki*; Iwasaki, Takashi*; Ohira, Hiroaki

no journal, , 

Sodium-cooled fast reactors have been developed aiming at introducing natural circulation decay heat removal systems by utilizing the characteristic of having a large coolant temperature difference between at the inlet and at the outlet of reactor vessel. In this study, as part of validation for core hot spot evaluation method, which is required for adoption of natural circulation decay heat removal systems, EBR-II (Experimental Breeder Reactor II) shutdown heat removal test was simulated. The simulation results demonstrated that the evaluation method sufficiently predicts the whole plant thermal hydraulic behaviors and the maximum coolant temperature in a fuel subassembly in natural circulation decay heat removal.