Forced-flow interchannel mixing model for fuel rod assemblies utilizing a helical wire-wrap spacer system-Reference-Cited by-同舟云学术

Forced-flow interchannel mixing model for fuel rod assemblies utilizing a helical wire-wrap spacer system

Published:1972-08 Issue:1 Volume:22 Page:43-50
ISSN:0029-5493
Container-title:Nuclear Engineering and Design
language:en
Short-container-title:Nuclear Engineering and Design

Author:

Ginsberg Theodore

Publisher

Elsevier BV

Subject

Mechanical Engineering,Waste Management and Disposal,Safety, Risk, Reliability and Quality,General Materials Science,Nuclear Energy and Engineering,Nuclear and High Energy Physics

Reference14 articles.

1. COBRA-II: A Digital Computer Program for Thermal-Hydraulic Subchannel Analysis of Rod Bundle Nuclear Fuel Elements;Rowe;BNWL-1229,1970

2. Hambo — A Computer Programme for Subchannel Analysis of the Hydraulic and Burnout Charateristics of Rod Clusters, (Part I) General Description;Bowing;AEEW-R524,1967

3. Heat-Transfer Analysis of Gas- or Liquid-Cooled Reactor Passages;Katchee;IDO-28595,1965

4. Experimental Study of Coolant Cross-Mixing in Multirod Bundles, Consisting of Unfinned, One-, Three-, and Six- Fin Fuel Rods;Baumann;Atomkernenergie,1969

5. Heat Transfer and Hydraulic Characteristics of SRE Fuel Element;Shimizaki,1957

Cited by 12 articles. 订阅此论文施引文献订阅此论文施引文献，注册后可以免费订阅5篇论文的施引文献，订阅后可以查看论文全部施引文献

1. Analysis of pressure losses and flow distribution in wire-wrapped hexagonal rod bundles for licensing. Part I: The Pacio-Chen-Todreas Detailed model (PCTD);Nuclear Engineering and Design;2022-03

2. Investigation on subchannel flow distribution in wire-wrapped 37 and 61-pin bundle using computational fluid dynamics;Nuclear Engineering and Design;2020-12

3. Natural convection heat transfer from a vertical single cylinder with helical wire spacer in liquid sodium;Nuclear Engineering and Design;2019-01

4. Analysis of inter-channel sweeping flow in wire wrapped 19-rod bundle;Nuclear Engineering and Design;2018-07

5. A review of sub-channel thermal hydraulic codes for nuclear reactor core and future directions;Nuclear Engineering and Design;2018-06