Bubble behavior in natural circulation of two-phase flow employed for integrated modular water reactor (IMR), analyzed by ALPHA-FLOW code
Creators
- 1. Mitsubishi Heavy Industries, Ltd. 3-1, Minatomirai 3-Chome, Nishi-ku, Yokohama 220-8401 (Japan)
- 2. Kyoto University Yoshida, Sakyo-ku, Kyoto 606-8501 (Japan)
- 3. Central Research Institute of Electric Power Industry 2-11-1, Iwado Kita, Komae-shi, Tokyo 201-8511 (Japan)
- 4. The Japan Atomic Power Co. Mitoshiro Bldg., 1-1, Kanda-Mitoshiro-Cho, Chiyoda-ku, Tokyo 101-0053 (Japan)
Description
Full text of publication follows: Integrated Modular Water Reactor (IMR) is a small-scale reactor to satisfy variable demand of electric power in the near future, which is now being developed by our project consisting of Kyoto University, Central Research Institute of Electric Power Industry, the Japan Atomic Power Company and Mitsubishi Heavy Industries. IMR is designed based on the accomplished design of the Japanese PWR, but is characterized by employing natural circulation of two-phase flow under 15.5 MPa of pressure as the primary cooling system, to realize effective heat transfer through the built-in-vessel steam generators utilizing both sensible and latent heat and small configuration of the reactor vessel without pipes and pumps. The system is named Hybrid Heat Transfer System (HHTS). Natural circulation of two-phase flow in HHTS should be stable with enough flow rate to avoid departure from nucleate boiling (DNB) on fuel. Bubbles in two-phase flow play important roles for the stable natural circulation, and bubble behavior in the natural circulation has been analyzed by a well known computational fluid dynamics code 'Alpha-Flow', which can treat hydraulic dynamics with multi-dimensional two-fluid model. Three dimensional behavior of bubbles in the up-flow and downflow is investigated through the analysis, taking account of internal structures in the IMR reactor vessel. Consequently, water-vapor separation at the water surface in the natural circulation depending on distribution of flow velocity and void fraction at core outlet is confirmed. And also correlation of bubble behavior and the internal structures is investigated, which would lead to optimum design of the internal structures. (authors)
Availability note (English)
Available in abstract form only, full text entered in this recordAdditional details
Publishing Information
- Imprint Pagination
- 1 p.
- Report number
- INIS-FR--3726
Conference
- Title
- 11. international topical meeting on nuclear reactor thermal hydraulics (Nureth 11)
- Dates
- 2-6 Oct 2005
- Place
- Avignon (France)
INIS
- Country of Publication
- France
- Country of Input or Organization
- France
- INIS RN
- 36087665
- Subject category
- S42: ENGINEERING; S21: SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS;
- Resource subtype / Literary indicator
- Conference, Non-conventional Literature
- Descriptors DEI
- A CODES; BUBBLES; COMPUTERIZED SIMULATION; FLOW MODELS; NATURAL CONVECTION; PWR TYPE REACTORS; THERMAL HYDRAULICS; TWO-PHASE FLOW; VELOCITY; VOID FRACTION
- Descriptors DEC
- COMPUTER CODES; CONVECTION; ENERGY TRANSFER; ENRICHED URANIUM REACTORS; FLUID FLOW; FLUID MECHANICS; HEAT TRANSFER; HYDRAULICS; MASS TRANSFER; MATHEMATICAL MODELS; MECHANICS; POWER REACTORS; REACTORS; SIMULATION; THERMAL REACTORS; WATER COOLED REACTORS; WATER MODERATED REACTORS