CFD investigation on thermal-hydraulic characteristics of a helical cruciform fuel bundle
The thermal-hydraulic performance of nuclear fuel has a very important effect on the safety and economy of a reactor. The helical cruciform fuel (HCF) is an innovative fuel design with a lot of potential advantages. HCF assembly is a self-supporting assembly. It needn't spacer grid which is usu...
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| Veröffentlicht in: | Progress in nuclear energy (New series) 2022-06, Vol.148, p.104228, Article 104228 |
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| creator | Zhao, Hangbin Zhang, Qi Gu, Hanyang Xiao, Yao Liu, Maolong |
| description | The thermal-hydraulic performance of nuclear fuel has a very important effect on the safety and economy of a reactor. The helical cruciform fuel (HCF) is an innovative fuel design with a lot of potential advantages. HCF assembly is a self-supporting assembly. It needn't spacer grid which is usually used in the cylindrical fuel assembly. It has larger surface-to-volume ratio in comparison to the cylindrical fuel, which can increase the power output without necessarily altering the operating surface heat flux. In this research, the thermal-hydraulic characteristics of a 4 × 4 HCF bundle were investigated by numerical method. A numerical model of the bundle was developed and validated with experimental data. Based on this model, five flow conditions were respectively calculated. The transverse flow in the HCF bundle was analyzed, and the temperature distributions of the bundle and the water were discussed. The results show that the transverse mixing of water in the bundle is enhanced due to the helical geometry of the HCF rod. The mixing intensity periodically varies with twist angle increasing, and every twist angle of 90° is a variation cycle. The temperature distribution on the rod surface helically varies along the flow direction, and the temperature in the valley region is obviously higher than that in the lobe region. Additionally, the water temperature in the back of the lobe is higher than that in the front of the lobe. |
| doi_str_mv | 10.1016/j.pnucene.2022.104228 |
| format | Article |
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The helical cruciform fuel (HCF) is an innovative fuel design with a lot of potential advantages. HCF assembly is a self-supporting assembly. It needn't spacer grid which is usually used in the cylindrical fuel assembly. It has larger surface-to-volume ratio in comparison to the cylindrical fuel, which can increase the power output without necessarily altering the operating surface heat flux. In this research, the thermal-hydraulic characteristics of a 4 × 4 HCF bundle were investigated by numerical method. A numerical model of the bundle was developed and validated with experimental data. Based on this model, five flow conditions were respectively calculated. The transverse flow in the HCF bundle was analyzed, and the temperature distributions of the bundle and the water were discussed. The results show that the transverse mixing of water in the bundle is enhanced due to the helical geometry of the HCF rod. The mixing intensity periodically varies with twist angle increasing, and every twist angle of 90° is a variation cycle. The temperature distribution on the rod surface helically varies along the flow direction, and the temperature in the valley region is obviously higher than that in the lobe region. Additionally, the water temperature in the back of the lobe is higher than that in the front of the lobe.</description><identifier>ISSN: 0149-1970</identifier><identifier>EISSN: 1878-4224</identifier><identifier>DOI: 10.1016/j.pnucene.2022.104228</identifier><language>eng</language><publisher>Oxford: Elsevier Ltd</publisher><subject>Assembly ; Heat conductivity ; Heat flux ; Heat transfer ; Helical cruciform fuel ; Mathematical models ; Nuclear fuel ; Nuclear fuels ; Nuclear safety ; Numerical methods ; Numerical models ; Self-supporting assembly ; Temperature distribution ; Thermal-hydraulic characteristics ; Transverse mixing ; Water temperature</subject><ispartof>Progress in nuclear energy (New series), 2022-06, Vol.148, p.104228, Article 104228</ispartof><rights>2022</rights><rights>Copyright Elsevier BV Jun 2022</rights><lds50>peer_reviewed</lds50><woscitedreferencessubscribed>false</woscitedreferencessubscribed><citedby>FETCH-LOGICAL-c337t-71aff16fdf0cd58271f3816c901bdd39be00161d82411b0b8c807196324fe1db3</citedby><cites>FETCH-LOGICAL-c337t-71aff16fdf0cd58271f3816c901bdd39be00161d82411b0b8c807196324fe1db3</cites></display><links><openurl>$$Topenurl_article</openurl><openurlfulltext>$$Topenurlfull_article</openurlfulltext><thumbnail>$$Tsyndetics_thumb_exl</thumbnail><linktohtml>$$Uhttps://dx.doi.org/10.1016/j.pnucene.2022.104228$$EHTML$$P50$$Gelsevier$$H</linktohtml><link.rule.ids>314,777,781,3537,27905,27906,45976</link.rule.ids></links><search><creatorcontrib>Zhao, Hangbin</creatorcontrib><creatorcontrib>Zhang, Qi</creatorcontrib><creatorcontrib>Gu, Hanyang</creatorcontrib><creatorcontrib>Xiao, Yao</creatorcontrib><creatorcontrib>Liu, Maolong</creatorcontrib><title>CFD investigation on thermal-hydraulic characteristics of a helical cruciform fuel bundle</title><title>Progress in nuclear energy (New series)</title><description>The thermal-hydraulic performance of nuclear fuel has a very important effect on the safety and economy of a reactor. The helical cruciform fuel (HCF) is an innovative fuel design with a lot of potential advantages. HCF assembly is a self-supporting assembly. It needn't spacer grid which is usually used in the cylindrical fuel assembly. It has larger surface-to-volume ratio in comparison to the cylindrical fuel, which can increase the power output without necessarily altering the operating surface heat flux. In this research, the thermal-hydraulic characteristics of a 4 × 4 HCF bundle were investigated by numerical method. A numerical model of the bundle was developed and validated with experimental data. Based on this model, five flow conditions were respectively calculated. The transverse flow in the HCF bundle was analyzed, and the temperature distributions of the bundle and the water were discussed. The results show that the transverse mixing of water in the bundle is enhanced due to the helical geometry of the HCF rod. The mixing intensity periodically varies with twist angle increasing, and every twist angle of 90° is a variation cycle. The temperature distribution on the rod surface helically varies along the flow direction, and the temperature in the valley region is obviously higher than that in the lobe region. Additionally, the water temperature in the back of the lobe is higher than that in the front of the lobe.</description><subject>Assembly</subject><subject>Heat conductivity</subject><subject>Heat flux</subject><subject>Heat transfer</subject><subject>Helical cruciform fuel</subject><subject>Mathematical models</subject><subject>Nuclear fuel</subject><subject>Nuclear fuels</subject><subject>Nuclear safety</subject><subject>Numerical methods</subject><subject>Numerical models</subject><subject>Self-supporting assembly</subject><subject>Temperature distribution</subject><subject>Thermal-hydraulic characteristics</subject><subject>Transverse mixing</subject><subject>Water temperature</subject><issn>0149-1970</issn><issn>1878-4224</issn><fulltext>true</fulltext><rsrctype>article</rsrctype><creationdate>2022</creationdate><recordtype>article</recordtype><recordid>eNqFkEtLxDAQx4MouD4-ghDw3DWTvtKTyOqqsOBFD55CmkxsSrddk3Zhv71ZundhYIaZ_7x-hNwBWwKD4qFd7vpJY49LzjiPuYxzcUYWIEqRxDg7JwsGWZVAVbJLchVCyxiUkOcL8r1aP1PX7zGM7keNbuhptLFBv1Vd0hyMV1PnNNWN8kqP6F0U6kAHSxVtMJZUR7WftLOD31I7YUfrqTcd3pALq7qAtyd_Tb7WL5-rt2Tz8fq-etokOk3LMSlBWQuFNZZpkwtegk0FFLpiUBuTVjXGWwswgmcANauFFqyEqkh5ZhFMnV6T-3nuzg-_U_xDtsPk-7hS8iISSHnkEFX5rNJ-CMGjlTvvtsofJDB5pChbeaIojxTlTDH2Pc59GF_YO_QyaIe9RuM86lGawf0z4Q_4w338</recordid><startdate>202206</startdate><enddate>202206</enddate><creator>Zhao, Hangbin</creator><creator>Zhang, Qi</creator><creator>Gu, Hanyang</creator><creator>Xiao, Yao</creator><creator>Liu, Maolong</creator><general>Elsevier Ltd</general><general>Elsevier BV</general><scope>AAYXX</scope><scope>CITATION</scope><scope>7TB</scope><scope>8FD</scope><scope>FR3</scope><scope>KR7</scope></search><sort><creationdate>202206</creationdate><title>CFD investigation on thermal-hydraulic characteristics of a helical cruciform fuel bundle</title><author>Zhao, Hangbin ; Zhang, Qi ; Gu, Hanyang ; Xiao, Yao ; Liu, Maolong</author></sort><facets><frbrtype>5</frbrtype><frbrgroupid>cdi_FETCH-LOGICAL-c337t-71aff16fdf0cd58271f3816c901bdd39be00161d82411b0b8c807196324fe1db3</frbrgroupid><rsrctype>articles</rsrctype><prefilter>articles</prefilter><language>eng</language><creationdate>2022</creationdate><topic>Assembly</topic><topic>Heat conductivity</topic><topic>Heat flux</topic><topic>Heat transfer</topic><topic>Helical cruciform fuel</topic><topic>Mathematical models</topic><topic>Nuclear fuel</topic><topic>Nuclear fuels</topic><topic>Nuclear safety</topic><topic>Numerical methods</topic><topic>Numerical models</topic><topic>Self-supporting assembly</topic><topic>Temperature distribution</topic><topic>Thermal-hydraulic characteristics</topic><topic>Transverse mixing</topic><topic>Water temperature</topic><toplevel>peer_reviewed</toplevel><toplevel>online_resources</toplevel><creatorcontrib>Zhao, Hangbin</creatorcontrib><creatorcontrib>Zhang, Qi</creatorcontrib><creatorcontrib>Gu, Hanyang</creatorcontrib><creatorcontrib>Xiao, Yao</creatorcontrib><creatorcontrib>Liu, Maolong</creatorcontrib><collection>CrossRef</collection><collection>Mechanical & Transportation Engineering Abstracts</collection><collection>Technology Research Database</collection><collection>Engineering Research Database</collection><collection>Civil Engineering Abstracts</collection><jtitle>Progress in nuclear energy (New series)</jtitle></facets><delivery><delcategory>Remote Search Resource</delcategory><fulltext>fulltext</fulltext></delivery><addata><au>Zhao, Hangbin</au><au>Zhang, Qi</au><au>Gu, Hanyang</au><au>Xiao, Yao</au><au>Liu, Maolong</au><format>journal</format><genre>article</genre><ristype>JOUR</ristype><atitle>CFD investigation on thermal-hydraulic characteristics of a helical cruciform fuel bundle</atitle><jtitle>Progress in nuclear energy (New series)</jtitle><date>2022-06</date><risdate>2022</risdate><volume>148</volume><spage>104228</spage><pages>104228-</pages><artnum>104228</artnum><issn>0149-1970</issn><eissn>1878-4224</eissn><abstract>The thermal-hydraulic performance of nuclear fuel has a very important effect on the safety and economy of a reactor. The helical cruciform fuel (HCF) is an innovative fuel design with a lot of potential advantages. HCF assembly is a self-supporting assembly. It needn't spacer grid which is usually used in the cylindrical fuel assembly. It has larger surface-to-volume ratio in comparison to the cylindrical fuel, which can increase the power output without necessarily altering the operating surface heat flux. In this research, the thermal-hydraulic characteristics of a 4 × 4 HCF bundle were investigated by numerical method. A numerical model of the bundle was developed and validated with experimental data. Based on this model, five flow conditions were respectively calculated. The transverse flow in the HCF bundle was analyzed, and the temperature distributions of the bundle and the water were discussed. The results show that the transverse mixing of water in the bundle is enhanced due to the helical geometry of the HCF rod. The mixing intensity periodically varies with twist angle increasing, and every twist angle of 90° is a variation cycle. The temperature distribution on the rod surface helically varies along the flow direction, and the temperature in the valley region is obviously higher than that in the lobe region. Additionally, the water temperature in the back of the lobe is higher than that in the front of the lobe.</abstract><cop>Oxford</cop><pub>Elsevier Ltd</pub><doi>10.1016/j.pnucene.2022.104228</doi></addata></record> |
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| ispartof | Progress in nuclear energy (New series), 2022-06, Vol.148, p.104228, Article 104228 |
| issn | 0149-1970 1878-4224 |
| language | eng |
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| source | Elsevier ScienceDirect Journals |
| subjects | Assembly Heat conductivity Heat flux Heat transfer Helical cruciform fuel Mathematical models Nuclear fuel Nuclear fuels Nuclear safety Numerical methods Numerical models Self-supporting assembly Temperature distribution Thermal-hydraulic characteristics Transverse mixing Water temperature |
| title | CFD investigation on thermal-hydraulic characteristics of a helical cruciform fuel bundle |
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