rdf-syntax-grammar
mods:
  schema location: http://www.loc.gov/mods/v3 http://www.loc.gov/standards/mods/v3/mods-3-2.xsd
  title info:
    title: Heat transfer and fluid flow in nuclear systems
  name:
    name part: Fenech, Henri
    name part: 1925-
  name:
    name part: ScienceDirect (Online service)
  type of resource: text
  genre: bibliography
  genre: Electronic books.
  origin info:
    place:
      place term: nyu
    place:
      place term: New York
    publisher: Pergamon Press
    date issued: ©1981
    date issued: 1981
    issuance: monographic
  language:
    language term: eng
  physical description:
    reformatting quality: access
    form: remote
    extent: 1 online resource (vii, 582 pages) : illustrations
  abstract: 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.
  table of contents: 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.
  table of contents: 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.
  table of contents: 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.
  table of contents: 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.
  table of contents: 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.
  note: edited by Henri Fenech.
  note: 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. http://purl.oclc.org/DLF/benchrepro0212 MiAaHDL
  note: digitized 2010 HathiTrust Digital Library committed to preserve pda MiAaHDL
  note: Includes bibliographical references and index.
  subject:
    topic: Nuclear power plants
    topic: Thermodynamics
  subject:
    topic: Centrales nucléaires
    topic: Thermodynamique
  subject:
    topic: TECHNOLOGY & ENGINEERING
    topic: Mechanical
  subject:
    topic: Nuclear power plants
    topic: Thermodynamics
  subject:
    topic: Technologie nucléaire
  subject:
    topic: Transfert de chaleur
  subject:
    topic: Nuclear power plants
    topic: Thermodynamics
  subject:
    topic: Nuclear reactors
  subject:
    topic: Nuclear power plants
    topic: Thermodynamics
  classification: TK9153 .H4 1981
  classification: 621.48/31
  classification: 36
  classification: 52.55
  related item:
    title info:
      title: Heat transfer and fluid flow in nuclear systems
    origin info:
      publisher: New York : Pergamon Press, ©1981
    identifier: (DLC) 81008670
    identifier: (OCoLC)7572715
  related item:
    identifier: (DLC) 81008670
    identifier: (OCoLC)7572715
  identifier: 9781483150789
  identifier: 148315078X
  identifier: 0080271812
  identifier: 9780080271811
  identifier: http://proxy.library.tamu.edu/login
  record info:
    record content source: OCLCE
    record creation date: 100318
    record change date: 20260326172126.1
    record identifier: in00005740143
    language of cataloging:
      language term: eng