Cargando…
Heat transfer and fluid flow in nuclear systems /
Heat Transfer and Fluid in Flow Nuclear Systems discusses topics that bridge the gap between the fundamental principles and the designed practices. The book is comprised of six chapters that cover analysis of the predicting thermal-hydraulics performance of large nuclear reactors and associated heat...
| Clasificación: | Libro Electrónico |
|---|---|
| Otros Autores: | |
| Formato: | Electrónico eBook |
| Idioma: | Inglés |
| Publicado: |
New York :
Pergamon Press,
�1981.
|
| Temas: | |
| Acceso en línea: | Texto completo |
MARC
| LEADER | 00000cam a2200000 a 4500 | ||
|---|---|---|---|
| 001 | SCIDIR_ocn560401712 | ||
| 003 | OCoLC | ||
| 005 | 20231117032922.0 | ||
| 006 | m o d | ||
| 007 | cr bn||||||abp | ||
| 007 | cr bn||||||ada | ||
| 008 | 100318s1981 nyua ob 001 0 eng d | ||
| 040 | |a OCLCE |b eng |e pn |c OCLCE |d OCLCQ |d OCLCF |d UIU |d N$T |d E7B |d EBLCP |d IDEBK |d DEBSZ |d YDXCP |d OCLCQ |d MERUC |d OCLCQ |d UKAHL |d OCLCQ |d OCLCO |d OCLCQ |d OCLCO | ||
| 019 | |a 297897190 |a 894791001 |a 899000865 |a 974618098 |a 974661226 |a 1100937293 | ||
| 020 | |a 9781483150789 |q (electronic bk.) | ||
| 020 | |a 148315078X |q (electronic bk.) | ||
| 020 | |z 0080271812 | ||
| 020 | |z 9780080271811 | ||
| 035 | |a (OCoLC)560401712 |z (OCoLC)297897190 |z (OCoLC)894791001 |z (OCoLC)899000865 |z (OCoLC)974618098 |z (OCoLC)974661226 |z (OCoLC)1100937293 | ||
| 042 | |a dlr | ||
| 050 | 4 | |a TK9153 |b .H4 1981 | |
| 072 | 7 | |a TEC |x 009070 |2 bisacsh | |
| 082 | 0 | 4 | |a 621.48/31 |2 19 |
| 245 | 0 | 0 | |a Heat transfer and fluid flow in nuclear systems / |c edited by Henri Fenech. |
| 260 | |a New York : |b Pergamon Press, |c �1981. | ||
| 300 | |a 1 online resource (vii, 582 pages) : |b illustrations | ||
| 336 | |a text |b txt |2 rdacontent | ||
| 337 | |a computer |b c |2 rdamedia | ||
| 338 | |a online resource |b cr |2 rdacarrier | ||
| 504 | |a Includes bibliographical references and index. | ||
| 506 | |3 Use copy |f Restrictions unspecified |2 star |5 MiAaHDL | ||
| 533 | |a Electronic reproduction. |b [Place of publication not identified] : |c HathiTrust Digital Library, |d 2010. |5 MiAaHDL | ||
| 538 | |a Master and use copy. Digital master created according to Benchmark for Faithful Digital Reproductions of Monographs and Serials, Version 1. Digital Library Federation, December 2002. |u http://purl.oclc.org/DLF/benchrepro0212 |5 MiAaHDL | ||
| 583 | 1 | |a digitized |c 2010 |h HathiTrust Digital Library |l committed to preserve |2 pda |5 MiAaHDL | |
| 588 | 0 | |a Print version record. | |
| 505 | 0 | |a Front Cover; Heat Transfer and Fluid in Flow Nuclear Systems; Copyright Page ; Table of Contents; LIST OF CONTRIBUTORS; PREFACE; CHAPTER 1. GENERAL CONSIDERATIONS ON THERMAL DESIGN AND PERFORMANCE REQUIREMENTS OF NUCLEAR REACTOR CORES; 1.1 INTRODUCTION; 1.2 DESIGN REQUIREMENTS FOR NUCLEAR REACTORS; 1.3 BASIC CONSIDERATIONS AFFECTING THE DESIGN AND SAFETY OF NUCLEAR REACTORS; REFERENCES; PROBLEMS; CHAPTER 2. PRESSURIZED SUBCOOLED LIGHT WATER SYSTEMS; 1.0 INTRODUCTION; 2.0 PWR DESCRIPTION; 3.0 THERMAL HYDRAULIC DESIGN REQUIREMENTS. | |
| 505 | 8 | |a 4.0 core thermal performance evaluation -- thermal margin methodology5.0 conclusion; appendix a: conservatipn equations for subchannel analysis methods; a.1 definition of terms and volume averaged conservation equations of the porous body approach (following ref. a.1); a.2 definition of terms in the subchannel approach; a.3 derivation of the subchannel conservation equations; references; appendix references; table 8 references; problems; chapter 3. boiling water reactor systems; 1.0 description of bwr system; 2.0 reactor configuration and flow paths; 3.0 two-phase flow and heat transfer. | |
| 505 | 8 | |a 4.0 BWR THERMAL HYDRAULIC PERFORMANCE5.0 ANALYSIS OF OPERATIONAL AND ABNORMAL TRANSIENTS; 6.0 LOSS-OF-COOLANT ACCIDENT (LOCH) ANALYSIS; 7.0 STABILITY; 8.0 CONCLUSION; REFERENCES; PROBLEMS; CHAPTER 4, PART 1: LIQUID METAL COOLED SYSTEMS. ONE-PHASE HEAT TRANSFER FLUID FLOW; 1.0 FEATURES OF LMFBR DESIGNS; 2.0 THERMOHYDRAULICS OF THE SUBASSEMBLY; 3.0 COMPLEX BUOYANT FLOWS OF LIQUID METALS; 4.0 THERMOHYDRAULIC STUDIES OF THE PLENUMS; APPENDIX; REFERENCES; PROBLEMS; CHAPTER 4, Part 2: LIQUID METAL COOLED SYSTEMS. SODIUM BOILING DYNAMICS; 1.0 INTRODUCTION; 2.0 FUNDAMENTAL CONSIDERATIONS. | |
| 505 | 8 | |a 3.0 transient sodium voiding models4.0 applications; 5.0 concluding remarks; appendix: two-phase pressure drop; references; problems; chapter 5, part 1: helium cooled systems. high temperature gas-cooled reactor (htgr); 1.0 introduction; 2.0 core thermal design basis; 3.0 design data; 4.0 core thermal and flow analysis; 5.0 systems transient performance; references; chapter 5, part 2: helium cooled systems. htr -- pebble design; 1.0 introduction; 2.0 pressure drop coefficient; 3.0 forced convection heat transfer; 4.0 heat conductivity in packed beds; references; problems. | |
| 505 | 8 | |a CHAPTER 5, PART 3: HELIUM COOLED SYSTEMS. THE GAS-COOLED FAST BREEDER REACTOR1.0 INTRODUCTION; 2.0 SYSTEM DESCRIPTION; 3.0 REACTOR CORE; 4.0 THERMAL-HYDRAULIC DESIGN CONSIDERATIONS FOR GCFR ASSEMBLIES; 5.0 THERMAL-HYDRAULIC CORRELATIONS FOR FUEL ASSEMBLY; 6.0 THERMAL-HYDRAULIC CORRELATIONS FOR RADIAL BLANKET ASSEMBLIES; 7.0 COMPUTER CODES FOR ASSEMBLY ANALYSIS; 8.0 TRANSIENT ANALYSIS; REFERENCES; PROBLEMS; CHAPTER 6, PART 1: THE THERMAL HYDRAULICS OF STEAM GENERATORS FOR PWRs; 1.0 INTRODUCTION; 2.0 THERMAL DESIGN; 3.0 STEAM GENERATOR ANALYSIS AND THERMAL PERFORMANCE. | |
| 520 | |a Heat Transfer and Fluid in Flow Nuclear Systems discusses topics that bridge the gap between the fundamental principles and the designed practices. The book is comprised of six chapters that cover analysis of the predicting thermal-hydraulics performance of large nuclear reactors and associated heat-exchangers or steam generators of various nuclear systems. Chapter 1 tackles the general considerations on thermal design and performance requirements of nuclear reactor cores. The second chapter deals with pressurized subcooled light water systems, and the third chapter covers boiling water reacto. | ||
| 650 | 0 | |a Nuclear power plants |x Thermodynamics. | |
| 650 | 6 | |a Centrales nucl�eaires |x Thermodynamique. |0 (CaQQLa)000298178 | |
| 650 | 7 | |a TECHNOLOGY & ENGINEERING |x Mechanical. |2 bisacsh | |
| 650 | 7 | |a Nuclear power plants |x Thermodynamics |2 fast |0 (OCoLC)fst01040671 | |
| 700 | 1 | |a Fenech, Henri, |d 1925- | |
| 776 | 0 | 8 | |i Print version: |t Heat transfer and fluid flow in nuclear systems. |d New York : Pergamon Press, �1981 |w (DLC) 81008670 |w (OCoLC)7572715 |
| 856 | 4 | 0 | |u https://sciencedirect.uam.elogim.com/science/book/9780080271811 |z Texto completo |