目　　次

■ 紹　　介
高速実験炉「常陽」MARICO-2試料部の回収 ( 3 )
古賀和浩 / 大原紀和 /猪　博一 / 近藤勝美 / 伊東秀明 / 芦田貴志 / 中村俊之

新形式触媒式水素再結合装置の開発状況の紹介 ( 9 )
平田慎吾 / 毛利智聡 / 五十嵐　実 / 佐藤　学 / 上地　優 / 西畑保雄 / 日野竜太郎

ナトリウム工学研究施設内装設備の設計・製作 ( 13 )
内山尚基 / 朽木憲一 / 佐藤康士

■ 論　　文
HTGR燃料ブロック冷却流路の流動特性の研究 ( 20 )
辻　延昌 / 大橋一孝 / 田澤勇次郎 / 橘　幸男 / 大橋弘史 / 高松邦吉

社会インフラ評価方式案の原子力発電所への適用 ( 25 )
市村富保

FAPIGの機構 ( 33 )

表紙デザイン：渡辺　望

CONTENTS

■ Introduction
Experimental Fast Reactor "Joyo"Retrieval
for the Bent MARICO-2 Test Subassembly Using Remote Control Devices ( 3 )
K. Koga / N. Oohara / H. Ino / K. Kondo / H. Ito / T. Ashida / T. Nakamura

Development Status of New Type Passive Autocatalytic Recombiner ( 9 )
S. Hirata / T. Mohri / M. Igarashi / M. Satoh / Y. Kamiji / Y. Nishihata / R. Hino

Design and Fabrication of Sodium Test Facility for Fast Breeder Reactor ( 13 )
N. Uchiyama / N. Kuchiki / K. Satou

■ Report
Study of the Flow Characteristics of Coolant Channel of Fuel Blocks for HTGR ( 20 )
N. Tsuji / K. Ohashi / Y. Tazawa / Y. Tachibana / H. Ohashi / K. Takamatsu

Applicability of the Proposed Evaluation Method
for Social Infrastructures to Nuclear Power Plants ( 25 )
T. Ichimura

Cover Design：Nozomu Watanabe

SYNOPSES

Kazuhiro Koga, Norikazu Oohara, Hiroichi Ino, Katsumi Kondo, Hideaki Ito, Takashi Ashida, Toshiyuki Nakamura
Experimental Fast Reactor "Joyo"
Retrieval for the Bent MARICO-2 Test Subassembly Using Remote Control Devices
FAPIG No. 190 pp.3 to 8 (2015)

In the experimental fast reactor "Joyo", the incident that the MARICO-2 test subassembly was bent in the reactor vessel occurred in 2007.
Fuji Electric has designed and manufactured the devices to retrieve the bent MARICO-2 test subassembly. Then, Fuji Electric and Japan Atomic Energy Agency have jointly accomplished retrieval works for the bent MARICO-2 test subassembly using remote control devices in "Joyo" in September, 2014.
This manuscript introduces the outline of design and manufacture of the MARICO-2 retrieval devices, and retrieval works in "Joyo".

KEYWORDS：fast reactor, Joyo, MARICO-2, test subassembly, retrieval works

Shingo Hirata,　Tomoaki Mohri, Minoru Igarashi, Manabu Satoh, Yu Kamiji, Yusuo Nishihata, Ryutaro Hino
Development Status of New Type Passive Autocatalytic Recombiner
FAPIG No. 190 pp.9 to 12 (2015)

A new type passive autocatalytic recombiner (PAR) to mitigate released hydrogen gas at the accident of nuclear power plants has been developed. In order to realize easy and anywhere installation, this new PAR has advanced automotive catalysts with features such as light and small, high durability, and high performance. Downsizing and light-weighting of recombiner has been conducted with thermal hydraulic and structural analysis, and the PAR concept proposed is revealed feasible for hydrogen mitigation in nuclear power plants.

KEYWORDS：passive autocatalytic recombiner, nuclear power plant, hydrogen mitigation

Naoki Uchiyama, Norikazu Kuchiki, Kouji Satou
Design and Fabrication of Sodium Test Facility for Fast Breeder Reactor
FAPIG No. 190 pp.13 to 19 (2015)

JAEA has established SERF (Sodium Engineering Research Facility) facility for research and development of advanced sodium handling technology in Tsuruga City, Fukui Prefecture. We introduce design, fabrication and installation of SERF carried out by KHI through fiscal year 2011 to 2014.

KEYWORDS：sodium handling technology, FBR, MONJU

Nobumasa Tsuji, Kazutaka Ohashi, Yujirou Tazawa, Yukio Tachibana, Hirofumi Ohashi, Kuniyoshi Takamatsu,
Study of the Flow Characteristics of Coolant Channel of Fuel Blocks for HTGR
FAPIG No. 190 pp.20 to 24 (2015)

Passive heat removal performance is of primary concern for enhanced inherent safety of HTGR. In a loss of forced cooling accident, decay heat of fuels must be removed to graphite blocks by radiation, thermal conduction and natural convection in block-type HTGR. Because the temperature of fuels is strongly affected by natural convection of coolant in core region, it becomes important to estimate the behavior of natural convection in core region precisely. The numerical studies are performed using thermal hydraulic CFD code for one column-model of fuel blocks which is represented explicitly as individual coolant channels in fuel block. The thermal hydraulic analyses are conducted for normal operation and loss of forced cooling accident conditions, as results, the flow and heat transfer characteristics of fuel blocks are quantitatively evaluated both in forced convection mode and natural convection mode.

KEYWORDS：HTGR, HTTR, fuel block, natural convection, CFD code

Tomiyasu Ichimura
Applicability of the Proposed Evaluation Method for Social Infrastructures to Nuclear Power Plants
FAPIG No. 190 pp.25 to 32 (2015)

This paper propose an evaluation method for social infrastructures and the applicability of the proposed evaluation method is verified by the trial application of the method to a nuclear power plant which is one of social infrastructures.
The method proposes common criteria to the four evaluation aspects and their evaluation criteria. By the application of this method to the nuclear power plant as one of social infrastructures, sample criteria from the four evaluation aspects were able to be clarified and also it was verified that current activities in a nuclear power were reached to the highest standard level by the evaluation using the proposed evaluation method for social infrastructure.
Through this verification, it is recognized that this evaluation method can provide certain effectiveness for the evaluation of social infrastructures.

KEYWORDS：social infrastructure, environment, maintenance, hazard, maturity model, evaluation figures, ISO