Development of transient behavior analysis code for metal fuel fast reactor during initiating phase of core disruptive accident

太田 宏一*; 尾形 孝成*; 山野 秀将; 二神 敏; 島田 貞衣*; 山田 由美*

Proceedings of 30th International Conference on Nuclear Engineering (ICONE30) (Internet), 8 Pages, 2023/05

The experimental analyses of the U-Pu-Zr fuel pin behavior during transient overpower (TOP) tests were performed by CANIS, and the residual cladding wall thickness and molten region of the fuel alloy after the tests, and the reactivity inserted by molten fuel extrusion before the fuel pin failure were compared with the experimental results. On the basis of these analysis results, detailed calculation models were developed for and implemented into CANIS to be made it possible to consider changes in the local properties of the fuel alloys due to redistribution of fuel constituents during steady-state irradiation and in the cladding thinning rate depending on the fuel-cladding interface temperature. The modified CANIS properly predicted fuel behavior and resulting reactivity changes before fuel pin failure in TOP events.

Engineering formulation of the irradiation growth behavior of zirconium-based alloys for light water reactors

垣内 一雄; 天谷 政樹; 宇田川 豊

Journal of Nuclear Materials, 573, p.154110_1 - 154110_7, 2023/01

https://doi.org/10.1016/j.jnucmat.2022.154110

The irradiation growth behavior of coupon specimens prepared from improved Zr-based alloys for light-water reactor fuel cladding, which have various additive elements and fabrication conditions, was investigated by conducting an irradiation test at 573 and 593 K under typical PWR coolant conditions up to a fast-neutron fluence of 7.810 (n/cm , E 1 MeV) in the Halden reactor in Norway. Based on the dimensional change data measured at interim and final inspections, the amounts of irradiation growth of the improved Zr-based alloys were formulated from the viewpoint of engineering. The trends of the parameters which express the effects of additive elements on irradiation growth behavior were in good agreement with those previously reported, and it was found that the amount of irradiation growth can be expressed by using a summation rule of the effect of each additive element on irradiation growth.

Irradiation growth behavior and effect of hydrogen absorption of Zr-based cladding alloys for PWR

垣内 一雄; 天谷 政樹; 宇田川 豊

Annals of Nuclear Energy, 171, p.109004_1 - 109004_9, 2022/06

https://doi.org/10.1016/j.anucene.2022.109004

In order to understand the dimensional stability of the fuel rod during long-term use in commercial LWRs, an irradiation growth testing in the Halden reactor of Norway was conducted on various fuel cladding materials including the improved Zr alloy. In this paper, the effect of hydrogen, which was absorbed in the cladding tube due to corrosion, on the irradiation growth behavior was evaluated. Comparison between the specimens with or without pre-charged hydrogen revealed that the effect of hydrogen absorption, accelerating irradiation growth, became significant when the hydrogen content exceeded the hydrogen solubility limit in the corresponding irradiation temperature. Analysis based on this understanding derived growth acceleration effect (0.060.01)%/100 ppm, whose denominator is defined as the amount of absorbed hydrogen involved in hydride precipitation under irradiation as a relevant parameter.

Nuclide production cross sections of Ni and Zr irradiated with 0.4-, 1.3-, 2.2-, and 3.0-GeV protons

竹下 隼人; 明午 伸一郎; 松田 洋樹; 岩元 大樹; 前川 藤夫; 渡辺 幸信*

JPS Conference Proceedings (Internet), 33, p.011045_1 - 011045_6, 2021/03

https://doi.org/10.7566/JPSCP.33.011045

加速器駆動核変換システム(ADS)における核設計の高度化のため、ADSで使われる材料であるNiとZrについて、数GeVエネルギー領域における陽子入射の核種生成断面積測定を行い、核設計に用いる計算コードPHITSによる計算値やJENDL/HE-2007との比較検討を行った。

炉心溶融物の粘性及び表面張力同時測定技術の開発(委託研究); 令和元年度英知を結集した原子力科学技術・人材育成推進事業

廃炉環境国際共同研究センター; 大阪大学*

JAEA-Review 2020-038, 41 Pages, 2020/12

JAEA-Review-2020-038.pdf:3.28MB

日本原子力研究開発機構(JAEA)廃炉環境国際共同研究センター(CLADS)では、令和元年度英知を結集した原子力科学技術・人材育成推進事業(以下、「本事業」という)を実施している。本事業は、東京電力ホールディングス福島第一原子力発電所の廃炉等をはじめとした原子力分野の課題解決に貢献するため、国内外の英知を結集し、様々な分野の知見や経験を、従前の機関や分野の壁を越えて緊密に融合・連携させた基礎的・基盤的研究及び人材育成を推進することを目的としている。平成30年度の新規採択課題から実施主体を文部科学省からJAEAに移行することで、JAEAとアカデミアとの連携を強化し、廃炉に資する中長期的な研究開発・人材育成をより安定的かつ継続的に実施する体制を構築した。本研究は、研究課題のうち、平成30年度に採択された「炉心溶融物の粘性及び表面張力同時測定技術の開発」の令和元年度の研究成果について取りまとめたものである。炉心溶融物である(U,Zr)Oやボライドは非常に高温であるために、通常の測定方法では容器との反応が避けられず、熱物性の測定は困難である。本研究では、ガス浮遊法を用いて浮遊させた試料を加熱溶融させることで液滴とし、その液滴を基板に衝突させる。その衝突の一瞬の挙動から、粘性と表面張力を同時に導出する新しい技術を開発する。

Irradiation growth behavior of improved Zr-based alloys for fuel cladding

天谷 政樹; 垣内 一雄; 三原 武

Proceedings of International Nuclear Fuel Cycle Conference / Light Water Reactor Fuel Performance Conference (Global/Top Fuel 2019) (USB Flash Drive), p.1048 - 1056, 2019/09

New fuel cladding alloys of which composition was changed from conventional ones have been developed by nuclear fuel vendors and utilities. Since the irradiation growth of fuel cladding is one of the most important parameters which determine the dimensional stability of fuel rod and/or fuel assembly during normal operation, the irradiation growth behavior of the improved Zr-based alloys for light-water reactor fuel cladding was investigated. The coupon specimens were prepared from fuel cladding tubes with various kinds of improved Zr-based alloys. The specimens were loaded into test rigs and had been irradiated in the Halden reactor in Norway under several coolant temperature conditions up to a fast-neutron fluence of 7.810 (n/cm, E 1 MeV). Irradiation conditions such as specimen temperatures had been continuously monitored during the irradiation. During and after the irradiation, the amount of irradiation growth of each specimen was evaluated as a part of the interim and final inspections. The effect of the difference in alloy composition on the amount of irradiation growth seemed insignificant if the other conditions e.g. the final heat treatment condition at fabrication and the irradiation temperature were the same.

High-temperature interaction between zirconium and UO

白数 訓子; 鈴木 晶大*; 永江 勇二; 倉田 正輝

Proceedings of International Topical Workshop on Fukushima Decommissioning Research (FDR 2019) (Internet), 4 Pages, 2019/05

ジルカロイ被覆管とUO燃料の高温における溶融過程解析モデルの高度化に資するため、ZrとUOの高温反応試験を実施した。UOるつぼに、Zr試料を装荷し2173K近傍で加熱を行い、生成した反応相の成長状況や溶融状態、組織変化の観察を行った。試料の中間領域には、上方へ直線状に伸びる相が観測された。この相は、U-Zrの金属溶体相と考えられ、Zr試料中、酸素濃度が少ない方へ選択的に成長したと考えられる。

Microscopic analyses of complexes formed in adsorbent for Mo and Zr separation chromatography

安倍 諒治*; 名越 航平*; 新井 剛*; 渡部 創; 佐野 雄一; 松浦 治明*; 高木 秀彰*; 清水 伸隆*; 江夏 昌志*; 佐藤 隆博*

Nuclear Instruments and Methods in Physics Research B, 404, p.173 - 178, 2017/08

https://doi.org/10.1016/j.nimb.2017.02.082

Molybdenum and zirconium obstruct the efficient recovery of minor actinides (MA(III): Am(III) and Cm(III)) by extraction chromatography; hence, the removal of these elements prior to MA(III) recovery is desirable. The use of an adsorbent impregnated with bis(2-ethylhexyl) hydrogen phosphate (HDEHP) for Mo and Zr decontamination was evaluated in this report. The adsorption/elution and column separation experiments showed that Mo and Zr in the simulated HLLW were selectively adsorbed on the particles, and that Mo was eluted by HO. EXAFS analysis and SAXS patterns of the adsorbent containing Zr revealed that the Zr-HDEHP complex had a crystal-like periodic structure similar to the structure of the precipitate produced in the solvent extraction system. Micro-PIXE analysis revealed that distribution of the residual Zr on the adsorbent was uniform.

Development of an extraction chromatography method for the analysis of Zr, Nb, and Mo in radioactive contaminated water generated at the Fukushima Daiichi Nuclear Power Station

島田 亜佐子; 亀尾 裕

Journal of Radioanalytical and Nuclear Chemistry, 310(3), p.1317 - 1323, 2016/12

https://doi.org/10.1007/s10967-016-5008-x

A new mutual separation method was developed to enable the analysis of Zr, Nb, and Mo in radioactive contaminated water from the Fukushima Daiichi Nuclear Power Station (FDNPS). First, distribution coefficient ( ) values between Tetra Valent Actinide (TEVA) resin, and HF, HF/HNO, and HF/HCl solutions were obtained to select the conditions for chromatographic separation. The separation performance of the developed method was validated with a control sample, which contained elements that could be interfere with the measurement of Zr, Nb, and Mo, (Cr, Co, Ni, Ge, and Se) and non-radioactive elements of high-dose nuclides in the contaminated water (Cs, Sr, and Y). The scheme was also evaluated with a simulated seawater sample because seawater was injected into the damaged reactors to assist cooling in the early stages of reactor failure. Finally, the chromatographic separation method was applied to the analysis ofZr, Nb, and Mo in contaminated water sampled at the FDNPS.

Extraction capacity of diglycolamide derivatives for Ca(II), Nd(III) and Zr(IV) from nitric acid to -dodecane containing a solvent modifier

佐々木 祐二; Zhu, Z.-X.; 須郷 由美; 鈴木 英哉*; 木村 貴海

Analytical Sciences, 21(10), p.1171 - 1175, 2005/10

https://doi.org/10.2116/analsci.21.1171

新しい抽出剤であるジグリコールアミド化合物の抽出特性研究の一環として、硝酸-ドデカン系で+2, +3及び+4価の陽イオンの抽出容量を調べた。金属イオンの最大抽出量(LOC)は硝酸濃度、抽出剤濃度に大きく依存した。LOCを増加する目的で、添加剤や疎水性の高いDGA化合物(側鎖アルキル基の長いDGA)を利用した。その結果、LOCは添加剤のDHOAモノアミド濃度やDGAのアルキル基長さに依存して増加することを明らかにした。