Scaling issues for the experimental characterization of reactor coolant system in integral test facilities and role of system code as extrapolation tool

Mascari, F.*; 中村 秀夫; Umminger, K.*; De Rosa, F.*; D'Auria, F.*

Proceedings of 16th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-16) (USB Flash Drive), p.4921 - 4934, 2015/08

The phenomenological analyses and thermal hydraulic characterization of a nuclear reactor are the basis for its design and safety evaluation. Scaled down tests of Integral Effect Test (IET) and Separate Effect Test (SET) are feasible to develop database. Though several scaling methods such as Power/Volume, Three level scaling and H2TS have been developed and applied to the IET and SET design, direct extrapolation of the data to prototype is in general difficult due to unavoidable scaling distortions. Constraints in construction and funding for test facility demand that a scaling compromise is inevitable further. Scaling approaches such as preservation of time, pressure and power etc. have to be adopted in the facility design. This paper analyzes some IET scaling approaches, starting from a brief analysis of the main characteristics of IETs and SETFs. Scaling approaches and their constraints in ROSA-III, FIST and PIPER-ONE facility are used to analyze their impact to the experimental prediction in Small Break LOCA counterpart tests. The liquid level behavior in the core are discussed for facility scaling-up limits.

Experiment Data of ROSA-III Integral Test RUN 710; Full ECCS Actuation,without Heat Generation in Channel A

小泉 安郎; 田坂 完二; 安達 公道; 安濃田 良成; 早田 邦久; 鈴木 光弘; 岡崎 元昭; 傍島 真; 村田 秀男; 斯波 正誼

JAERI-M 9249, 122 Pages, 1981/01

JAERI-M-9249.pdf:3.05MB

本報は、ROSA-III実験装置における実験RUN 710の実験データレポートである。RUN 710はBWRの再循環ポンプ吸込側配管の200%両端破断を模擬した実験である。実験はすべてのECCSを作動させ、炉心4チャンネルのうち、1チャンネルを発熱させないで行なわれた。主な実験初期条件は系圧力7.35MPa、炉心入口サブクーリング10.8K、炉心入口流量31.3kg/s、炉心発熱量2.42MWである。最高被覆管温度は609K、Position3で得られた。全ての模擬燃料棒はECCS作動後クエンチし、ECCSの有効性が確かめられた。