超临界压力水在竖直上升管内的传热研究
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摘要
超临界水冷堆(SCWR)是六种第四代核电站概念堆系统中唯一的水冷堆,它在经济性、经验延续性、技术成熟性和可持续性等方面具有独特的优势,是大功率压水堆技术发展的必然趋势。而掌握超临界压力流体传热特性是超临界水冷堆研发的基础和必要前提之一。
本文根据搜集到的实验数据讨论了超临界流体在管内强制对流换热模型和各控制参数对换热系数的影响,根据物性在拟临界附近急剧变化的特点分析了浮力作用和热加速效应等换热重要影响因素的机理和影响。
针对超临界压力水在竖直上升管内的换热工况,从公开文献中广泛搜集实验数据,建立并整理了宽范围换热实验数据库;在对传热机理与关联形式分析讨论的基础上,引入两组无量纲数表征超临界流体强烈物性变化及其次级效应对传热的影响,引入物性比进行关联式的物性修正,并最终确定关联式形式;采用SRC和误差比较相结合的方式进行敏感性分析,确定关联式最终引入的变量;采用主成分分析的方法去除引入变量之间的多重共线性,进而通过对实验数据的多元回归建立了两种形式的宽范围换热关联式,用以预测多种工况下超临界压力水的换热趋势和规律。误差评价表明,这组关联式在预测准确度高、应用范围广的同时,较好地预测了数据库中传热恶化的趋势和Nu值。
然而,经验关联式虽然便于工程设计和直接应用,但并不能完全满足用于系统设计和安全分析的系统分析程序对计算效率和迭代收敛性的需求。为此,本文还针对超临界压力水在竖直上升管内的换热工况构建了便于使用、易于更新的查询表以进行换热预测。基于对应的数据库和回归模型讨论,结合数据分区的讨论,运用响应面法建立超临界水换热特性和控制参数之间简单的映射关系,分别建立了两个换热查询表,并初步讨论了尺寸效应。误差评价表明,这一预测方法的误差在可接受范围内。若数据库在参数范围和数据分布均匀性等方面能有所改善,查询表的预测范围和预测准确度将得到极大的提高。
本文的工作建立了超临界压力水在竖直上升管内强制对流换热实验数据库,完善了换热规律的研究框架,为今后进一步更全面、更深入探索超临界流体对流换热规律奠定了基础。
As the only type of water-cooled reactor among the six advanced nuclear energy systems proposed by the Generation-IV International Forum, Supercritical Water-Cooled Reactor (SCWR) has the prominent advantages of economy, continuity of experience, technology maturity and sustainability. It comes into an essential choice of development of high-power pressure water reactor technology. Among the fundamental topics and necessary basis of SCWR R&D, study on heat transfer characteristics of water at supercritical pressures has got great attention.
In this thesis, according to test data, discussions on heat transfer model of water of supercritical pressures in vertical upward tubes, and the effects of various factors on heat transfer are performed. According to sharply change of thermal-physical properties near the pseudo-critical region, analysis on action of buoyancy and acceleration effects is conducted.
In this thesis, available experimental data of forced convective heat transfer of supercritical water in vertical upward tubes are extensively collected to establish a wide-ranged data bank; based on analysis of mechanism of heat transfer and the form of correlation, two kinds of dimensionless number groups are introduced to reflect special influents on heat transfer which are induced by the drastic variation of supercritical water properties and related secondary effects. With Standard regression coefficient (SRC) and comparisons of error, sensitivity analysis is performed and the form of correlation is identified. Applying principal factor analysis, two types of wide-ranged heat transfer correlations are regressed through principal component analysis. The evaluation of prediction error shows that the two correlations arrive both wide-ranged and high accuracy and work well within heat transfer deterioration region.
However, correlation is not enough for system analysis code for its relatively slow calculation efficient and iteration convergence. Therefore, look-up table is introduced to predict heat transfer behavior, which is easy to use and easy to update. Based on a correspond databank and study of regression model, two look-up table are established with discussion on test data groups, applying respond surface method (RSM). Meanwhile, based on the established look-up table, tube size effect on supercritical water heat transfer in vertical upward flow is preliminarily correlated and investigated. Further, assessments of the look-up table are carried out and delivered, and the errors are acceptable. With the increase of test data in various other practical engineered regimes, the look-up table is expected to be further improved.
With the work of this thesis, experimental databanks of forced convective heat transfer of water in vertical upward tubes at supercritical pressures, as well as improvement of heat transfer research method, is constructed, which might lay a foundation for further comprehensive and intensive analysis.
本文根据搜集到的实验数据讨论了超临界流体在管内强制对流换热模型和各控制参数对换热系数的影响,根据物性在拟临界附近急剧变化的特点分析了浮力作用和热加速效应等换热重要影响因素的机理和影响。
针对超临界压力水在竖直上升管内的换热工况,从公开文献中广泛搜集实验数据,建立并整理了宽范围换热实验数据库;在对传热机理与关联形式分析讨论的基础上,引入两组无量纲数表征超临界流体强烈物性变化及其次级效应对传热的影响,引入物性比进行关联式的物性修正,并最终确定关联式形式;采用SRC和误差比较相结合的方式进行敏感性分析,确定关联式最终引入的变量;采用主成分分析的方法去除引入变量之间的多重共线性,进而通过对实验数据的多元回归建立了两种形式的宽范围换热关联式,用以预测多种工况下超临界压力水的换热趋势和规律。误差评价表明,这组关联式在预测准确度高、应用范围广的同时,较好地预测了数据库中传热恶化的趋势和Nu值。
然而,经验关联式虽然便于工程设计和直接应用,但并不能完全满足用于系统设计和安全分析的系统分析程序对计算效率和迭代收敛性的需求。为此,本文还针对超临界压力水在竖直上升管内的换热工况构建了便于使用、易于更新的查询表以进行换热预测。基于对应的数据库和回归模型讨论,结合数据分区的讨论,运用响应面法建立超临界水换热特性和控制参数之间简单的映射关系,分别建立了两个换热查询表,并初步讨论了尺寸效应。误差评价表明,这一预测方法的误差在可接受范围内。若数据库在参数范围和数据分布均匀性等方面能有所改善,查询表的预测范围和预测准确度将得到极大的提高。
本文的工作建立了超临界压力水在竖直上升管内强制对流换热实验数据库,完善了换热规律的研究框架,为今后进一步更全面、更深入探索超临界流体对流换热规律奠定了基础。
As the only type of water-cooled reactor among the six advanced nuclear energy systems proposed by the Generation-IV International Forum, Supercritical Water-Cooled Reactor (SCWR) has the prominent advantages of economy, continuity of experience, technology maturity and sustainability. It comes into an essential choice of development of high-power pressure water reactor technology. Among the fundamental topics and necessary basis of SCWR R&D, study on heat transfer characteristics of water at supercritical pressures has got great attention.
In this thesis, according to test data, discussions on heat transfer model of water of supercritical pressures in vertical upward tubes, and the effects of various factors on heat transfer are performed. According to sharply change of thermal-physical properties near the pseudo-critical region, analysis on action of buoyancy and acceleration effects is conducted.
In this thesis, available experimental data of forced convective heat transfer of supercritical water in vertical upward tubes are extensively collected to establish a wide-ranged data bank; based on analysis of mechanism of heat transfer and the form of correlation, two kinds of dimensionless number groups are introduced to reflect special influents on heat transfer which are induced by the drastic variation of supercritical water properties and related secondary effects. With Standard regression coefficient (SRC) and comparisons of error, sensitivity analysis is performed and the form of correlation is identified. Applying principal factor analysis, two types of wide-ranged heat transfer correlations are regressed through principal component analysis. The evaluation of prediction error shows that the two correlations arrive both wide-ranged and high accuracy and work well within heat transfer deterioration region.
However, correlation is not enough for system analysis code for its relatively slow calculation efficient and iteration convergence. Therefore, look-up table is introduced to predict heat transfer behavior, which is easy to use and easy to update. Based on a correspond databank and study of regression model, two look-up table are established with discussion on test data groups, applying respond surface method (RSM). Meanwhile, based on the established look-up table, tube size effect on supercritical water heat transfer in vertical upward flow is preliminarily correlated and investigated. Further, assessments of the look-up table are carried out and delivered, and the errors are acceptable. With the increase of test data in various other practical engineered regimes, the look-up table is expected to be further improved.
With the work of this thesis, experimental databanks of forced convective heat transfer of water in vertical upward tubes at supercritical pressures, as well as improvement of heat transfer research method, is constructed, which might lay a foundation for further comprehensive and intensive analysis.
引文
[1]. Pioro, I. L.; Duffey, R. B.; Engineers, A. S. o. M., Heat transfer and hydraulic resistance at supercritical pressures in power engineering applications[M]. ASME Press: 2007.
[2]. Naidin, M.; Pioro, I.; Duffey, R., et al. In Super-critical water-cooled nuclear reactors (SCWRs): thermodynamic cycle options and thermal aspects of pressure-channel design, 2009; 2009; p 134e5.
[3]. Li, C.; Shan, J.; Leung, L. K. H., Subchannel analysis of CANDU-SCWR fuel[J]. Progress in Nuclear Energy 2009, 51 (8), 799-804.
[4]. Shitsman, M. E., Impairment of the heat transmission at supercritical pressures[J]. High Temperatures 1963, 1(2), 237-244.
[5]. Swenson, Heat transfer to supercritical water in smooth-bore tubes[J]. Journal of Heat Transfer 1965.11, 477~483.
[6]. Hall, W. B.; Jackson, J.; Engineers, A. S. o. M. In Laminarization of a turbulent pipe flow by buoyancy forces, 1969; ASME: 1969.
[7]. J.W.Ackerman, Pseudoboiling heat transfer to supercritical pressure water in smooth and ribbed tubes[J]. Transactions of the ASME 1970.8, 490-498.
[8]. Vikhrev, Y. V., Barulin, Yu.D., and Kon'kov, A.S., A study of heat transfer in vertical tubes at supercritical pressures[J]. Thermal Engineering 1967, 14(9), 116-119.
[9]. K. Yamagata, K. N., S. Hasegawa, T. Fujii and S. Yoshida,, Forced Convection Heat Transfer to Supercritical Water Flowing in Tubes[J]. Heat Mass Transfer 1972, 15, 2575~2593.
[10]. Jackson, J. D., Forced convection heat transfer to fluids at supercritical pressure[J]. Turbulent forced convection in channels and bundels 1978, 2, 563~612.
[11].胡志宏.超临界和近临界压力区垂直上升及倾斜管传热特性研究[D].西安:西安交通大学, 2001.
[12]. Yamashita, T.; Mori, H.; Yoshida, S., et al., Heat transfer and pressure drop of a supercritical pressure fluid flowing in a tube of small diameter[J]. Memoirs of the Faculty of Engineering, Kyushu University 2003, 63 (4), 227-244.
[13]. D.C.Groeneveld, L. K. H. L., P.L.Kirillov, The 1995 look-up table for critical heat flux in tubes[J]. Nuclear Engineering and Design 1996.
[14]. H.Griem, A new procedure for the prediction of forced convection heat transfer at near and supercritical pressure[J]. Heat and Mass Transfer 1996, 31, 301-305.
[15]. Shitsman, M. E., Temperature conditions in tubes at supercritical pressures[J]. Thermal Engineering 1968, 15(5), 72-77.
[16].徐峰,超临界压力下水在管内流动与传热特性研究[J].西安交通大学硕士学位论文2004.
[17]. Herkenrath, H.; Euratom, W?rmeübergang an Wasser bei erzwungener Str?mung imDruckbereich von 140 bis 250 bar[M]. EURATOM: 1967.
[18]. Alferov, N. S., Rybin, R.A., and Balunov, B.F., Heat transfer with turbulent water flow in a vertical tube under conditions of appreciable influence of free convection[J]. Thermal Engineering 1969, 16(12), 90-95.
[19]. Mokry, S.; Pioro, I.; Kirillov, P., et al., Supercritical-water heat transfer in a vertical bare tube[J]. Nuclear Engineering and Design 2010, 240 (3), 568-576.
[20].潘杰;杨冬;董自春, et al.,垂直上升光管内超临界水的传热特性试验研究[J].核动力工程2011, 32 (001), 75-80.
[21]. McEligot, D.; Coon, C.; Perkins, H., Relaminarization in tubes[J]. Int. J. Heat Mass Transfer 1970, 13, 431-433.
[22].何晓群,多元统计分析[M].中国人民大学出版社: 2004; Vol. 1.
[23].张文彤, SPSS统计分析高级教程[M].高等教育出版社: 2004.
[24]. H.Zahlan, D. C. G., S.Tavoularis, S.Mokry and I.Pioro, Assessment of supercritical heat transfer prediction methods. In The 5th Int. Sym. SCWR (ISSCWR-5)[C], Vancouver, British Columbia, Canada, 2011.
[25]. X. Cheng, T. S., A simplified method for heat transfer prediction of supercritical fluids in circular tubes[J]. Annals of Nuclear Energy 2009, 1120~1128.
[26]. Loewenberg, M. F., Supercritical water heat transfer in vertical tubes: A look-up table[J]. Progress in Nuclear Energy 2008, 532~538.
[27]. Styrikovich, M. A., Margulova, T.Kh., and Miropol'skii, Z.L., Problems in the development of designs of supercritical boilers[J]. Thermal Engineering 1967, 14(6), 5-9.
[28]. X. C. Huang and S. C. Cheng, Simple method for smoothing multidimensional experimental data with application to the CHF and postdryout look-up tables[J]. Numerical Heat Transfer 1994, Part B: 26,425-438
[29]. A. Tanase, S. C. Cheng, D. C. Groeneveld, J. Q. Shan, Diameter effect on critical heat transfer [J]. Nuclear Engineering and Design 2009,289-294
[30]. Zhu X. J, Bi Q. C, and Chen T. K.“An investigation on heat transfer characteristics of steam-water at different pressure in vertical upward tube”. 3rd Int. Symposium on SCWR– Design and Technology, March 12-15, 2007
[31]. A.A.Bishop, R.O.Sandberg, L.S.Tong, 1964.“High temperature supercritical pressure water loop”
[32]. Petukhov, B.; Kurganov, V.; Ankudinov, V., Heat transfer and flow resistance in the turbulent pipe flow of a fluid with near-critical state parameters[J]. Teplofizika Vysokikh Temperatur 1983, 21, 92-100.
[33].周强泰,超临界压力水的管内强迫对流传热[J].华中科技大学学报(自然科学版) 1983, 1.
[34]. Kuang, B.; Zhang, Y.; Cheng, X. In A new, wide-ranged heat transfer correlation of water at supercritical pressures in vertical upward ducts, 2008; 2008.
[2]. Naidin, M.; Pioro, I.; Duffey, R., et al. In Super-critical water-cooled nuclear reactors (SCWRs): thermodynamic cycle options and thermal aspects of pressure-channel design, 2009; 2009; p 134e5.
[3]. Li, C.; Shan, J.; Leung, L. K. H., Subchannel analysis of CANDU-SCWR fuel[J]. Progress in Nuclear Energy 2009, 51 (8), 799-804.
[4]. Shitsman, M. E., Impairment of the heat transmission at supercritical pressures[J]. High Temperatures 1963, 1(2), 237-244.
[5]. Swenson, Heat transfer to supercritical water in smooth-bore tubes[J]. Journal of Heat Transfer 1965.11, 477~483.
[6]. Hall, W. B.; Jackson, J.; Engineers, A. S. o. M. In Laminarization of a turbulent pipe flow by buoyancy forces, 1969; ASME: 1969.
[7]. J.W.Ackerman, Pseudoboiling heat transfer to supercritical pressure water in smooth and ribbed tubes[J]. Transactions of the ASME 1970.8, 490-498.
[8]. Vikhrev, Y. V., Barulin, Yu.D., and Kon'kov, A.S., A study of heat transfer in vertical tubes at supercritical pressures[J]. Thermal Engineering 1967, 14(9), 116-119.
[9]. K. Yamagata, K. N., S. Hasegawa, T. Fujii and S. Yoshida,, Forced Convection Heat Transfer to Supercritical Water Flowing in Tubes[J]. Heat Mass Transfer 1972, 15, 2575~2593.
[10]. Jackson, J. D., Forced convection heat transfer to fluids at supercritical pressure[J]. Turbulent forced convection in channels and bundels 1978, 2, 563~612.
[11].胡志宏.超临界和近临界压力区垂直上升及倾斜管传热特性研究[D].西安:西安交通大学, 2001.
[12]. Yamashita, T.; Mori, H.; Yoshida, S., et al., Heat transfer and pressure drop of a supercritical pressure fluid flowing in a tube of small diameter[J]. Memoirs of the Faculty of Engineering, Kyushu University 2003, 63 (4), 227-244.
[13]. D.C.Groeneveld, L. K. H. L., P.L.Kirillov, The 1995 look-up table for critical heat flux in tubes[J]. Nuclear Engineering and Design 1996.
[14]. H.Griem, A new procedure for the prediction of forced convection heat transfer at near and supercritical pressure[J]. Heat and Mass Transfer 1996, 31, 301-305.
[15]. Shitsman, M. E., Temperature conditions in tubes at supercritical pressures[J]. Thermal Engineering 1968, 15(5), 72-77.
[16].徐峰,超临界压力下水在管内流动与传热特性研究[J].西安交通大学硕士学位论文2004.
[17]. Herkenrath, H.; Euratom, W?rmeübergang an Wasser bei erzwungener Str?mung imDruckbereich von 140 bis 250 bar[M]. EURATOM: 1967.
[18]. Alferov, N. S., Rybin, R.A., and Balunov, B.F., Heat transfer with turbulent water flow in a vertical tube under conditions of appreciable influence of free convection[J]. Thermal Engineering 1969, 16(12), 90-95.
[19]. Mokry, S.; Pioro, I.; Kirillov, P., et al., Supercritical-water heat transfer in a vertical bare tube[J]. Nuclear Engineering and Design 2010, 240 (3), 568-576.
[20].潘杰;杨冬;董自春, et al.,垂直上升光管内超临界水的传热特性试验研究[J].核动力工程2011, 32 (001), 75-80.
[21]. McEligot, D.; Coon, C.; Perkins, H., Relaminarization in tubes[J]. Int. J. Heat Mass Transfer 1970, 13, 431-433.
[22].何晓群,多元统计分析[M].中国人民大学出版社: 2004; Vol. 1.
[23].张文彤, SPSS统计分析高级教程[M].高等教育出版社: 2004.
[24]. H.Zahlan, D. C. G., S.Tavoularis, S.Mokry and I.Pioro, Assessment of supercritical heat transfer prediction methods. In The 5th Int. Sym. SCWR (ISSCWR-5)[C], Vancouver, British Columbia, Canada, 2011.
[25]. X. Cheng, T. S., A simplified method for heat transfer prediction of supercritical fluids in circular tubes[J]. Annals of Nuclear Energy 2009, 1120~1128.
[26]. Loewenberg, M. F., Supercritical water heat transfer in vertical tubes: A look-up table[J]. Progress in Nuclear Energy 2008, 532~538.
[27]. Styrikovich, M. A., Margulova, T.Kh., and Miropol'skii, Z.L., Problems in the development of designs of supercritical boilers[J]. Thermal Engineering 1967, 14(6), 5-9.
[28]. X. C. Huang and S. C. Cheng, Simple method for smoothing multidimensional experimental data with application to the CHF and postdryout look-up tables[J]. Numerical Heat Transfer 1994, Part B: 26,425-438
[29]. A. Tanase, S. C. Cheng, D. C. Groeneveld, J. Q. Shan, Diameter effect on critical heat transfer [J]. Nuclear Engineering and Design 2009,289-294
[30]. Zhu X. J, Bi Q. C, and Chen T. K.“An investigation on heat transfer characteristics of steam-water at different pressure in vertical upward tube”. 3rd Int. Symposium on SCWR– Design and Technology, March 12-15, 2007
[31]. A.A.Bishop, R.O.Sandberg, L.S.Tong, 1964.“High temperature supercritical pressure water loop”
[32]. Petukhov, B.; Kurganov, V.; Ankudinov, V., Heat transfer and flow resistance in the turbulent pipe flow of a fluid with near-critical state parameters[J]. Teplofizika Vysokikh Temperatur 1983, 21, 92-100.
[33].周强泰,超临界压力水的管内强迫对流传热[J].华中科技大学学报(自然科学版) 1983, 1.
[34]. Kuang, B.; Zhang, Y.; Cheng, X. In A new, wide-ranged heat transfer correlation of water at supercritical pressures in vertical upward ducts, 2008; 2008.