EXPERIMENTAL STUDY ON THE EFFECT OF NaOH AND Na3PO4 ON SiO2 NANOFLUID SATURATED POOL BOILING CHF
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- Yu Jian
- School of Nuclear Science and Technology, Xi’an Jiaotong University
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- Zhang Yapei
- School of Nuclear Science and Technology, Xi’an Jiaotong University
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- Su G.H.
- School of Nuclear Science and Technology, Xi’an Jiaotong University
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- Tian Wenxi
- School of Nuclear Science and Technology, Xi’an Jiaotong University
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- Qiu Suizheng
- School of Nuclear Science and Technology, Xi’an Jiaotong University
抄録
<p>In-vessel retention through external reactor vessel cooling (IVR-ERVC) is one of the key severe accident mitigation strategies, which takes away the heat of melt through the reactor pressure vessel external cooling and prevents the leakage of the radioactive materials. Studies have shown that SiO2 nanofluids can significantly improve the critical heat flux (CHF) of the heating surface, and applying nanofluids to external cooling of pressure vessels can reduce pressure vessel failure risk. Sodium hydroxide (NaOH) as an additive in the spray water would change the chemical environment of cooling fluid, so this work investigated the effects of NaOH and Na3PO4 solution at PH 9.7 on SiO2 nanofluid saturated pool boiling CHF under different concentrations and different downward heating surface orientation angle. The experimental results show that under the same concentration, the CHF of SiO2 nanofluid based on NaOH and Na3PO4 solutions are nearly the same. Compared with SiO2/H2O nanofluid, the reduction in CHF is 15% at 0.01vol%. Same as SiO2/H2O nanofluid, CHF increased with the orientation angle of heating surface in NaOH and Na3PO4 environment. What’s more, scanning electron microscope (SEM) images were taken on the nanoparticle deposition layer formed on heating surface. The roughness, thickness, and the pattern of the nanoparticle deposition layer was different under different chemical environment, which lead to a decrease in CHF.</p>
収録刊行物
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- Proceedings of the ... International Conference on Nuclear Engineering. Book of abstracts : ICONE
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Proceedings of the ... International Conference on Nuclear Engineering. Book of abstracts : ICONE 2023.30 (0), 1247-, 2023
一般社団法人 日本機械学会
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詳細情報 詳細情報について
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- CRID
- 1390298278366374784
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- ISSN
- 24242934
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- 本文言語コード
- en
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- データソース種別
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- JaLC
- Crossref
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- 抄録ライセンスフラグ
- 使用不可