Summary Report On Design And Development Of High Temperature Gas-Cooled Power Pile

doi:10.2172/4359623

Title: Summary Report On Design And Development Of High Temperature Gas-Cooled Power Pile

Full Record
Other Related Research

Abstract

This report presents a description of a design for an experimental nuclear power plant utilizing a high temperature gas-cooled power pile as the energy source. The plant consists of the pile, a heat exchanger or boiler, a conventional steam turbine generator and their associated auxiliaries. Helium gas under pressure transfers heat from the pile to the boiler which generates steam for driving the generator. The plant is rated at a normal output of 12,000 kilowatts of heat and an electrical output of 2400 kilowatts. Provision is made for operation up to 20,000 kilowatts of heat (4000 kilowatts of electrical output) in the event operation of the plants proves this possible.

Authors:

McCullough, C. R. ^[1]

Clinton Labs., Oak Ridge, TN (United States)

Publication Date:: 1947-09-15

Research Org.:: Clinton Labs., Oak Ridge, TN (United States)

Sponsoring Org.:: USDOE

OSTI Identifier:: 4359623

Report Number(s):: MonN-383

NSA Number:: NSA-11-013961

DOE Contract Number:: W-35-058-ENG-71

Resource Type:: Technical Report

Resource Relation:: Other Information: Decl. Feb. 27, 1957. Orig. Receipt Date: 31-DEC-57

Country of Publication:: United States

Language:: English

Subject:: 71 CLASSICAL AND QUANTUM MECHANICS, GENERAL PHYSICS; BERYLLIUM MODERATOR; BERYLLIUM OXIDES; CONFIGURATION; COOLANT LOOPS; CYLINDERS; ELECTRICITY; FABRICATION; FUEL ELEMENTS; GAS COOLANT; GAS FLOW; GENERATORS; GRAPHITE MODERATOR; HEAT EXCHANGERS; HEAT TRANSFER; HELIUM; HIGH TEMPERATURE; MECHANICAL STRUCTURES; MODERATORS; PLANNING; POWER PLANTS; PRESSURE; REACTOR CORE; REACTOR SAFETY; STEAM; TEMPERATURE; THERMAL NEUTRONS; TURBINES; URANIUM; URANIUM 235; Beryllium Moderator; Beryllium Oxides; Configuration; Coolant Loops; Cylinders; Electricity; Fabrication; Fuel Elements; Gas Coolant; Gas Flow; Generators; Graphite Moderator; Heat Exchangers; Heat Transfer; Helium; High Temperature; Mechanical Structures; Moderators; Planning; Power Plants; Pressure; Reactor Core; Reactor Safety; Steam; Temperature; Thermal Neutrons; Turbines; Uranium; Uranium 235

Citation Formats


                    McCullough, C. R.. Summary Report On Design And Development Of High Temperature Gas-Cooled Power Pile.  United States: N. p., 1947. 
        Web.  doi:10.2172/4359623.

Copy to clipboard


                    McCullough, C. R.. Summary Report On Design And Development Of High Temperature Gas-Cooled Power Pile.  United States.  doi:10.2172/4359623.

Copy to clipboard


                    McCullough, C. R.. Mon .  
        "Summary Report On Design And Development Of High Temperature Gas-Cooled Power Pile".  United States.  
        doi:10.2172/4359623.  https://www.osti.gov/servlets/purl/4359623.

Copy to clipboard


                    
@article{osti_4359623,

  title        = {Summary Report On Design And Development Of High Temperature Gas-Cooled Power Pile},

  author       = {McCullough, C. R.},

  abstractNote = {This report presents a description of a design for an experimental nuclear power plant utilizing a high temperature gas-cooled power pile as the energy source. The plant consists of the pile, a heat exchanger or boiler, a conventional steam turbine generator and their associated auxiliaries. Helium gas under pressure transfers heat from the pile to the boiler which generates steam for driving the generator. The plant is rated at a normal output of 12,000 kilowatts of heat and an electrical output of 2400 kilowatts. Provision is made for operation up to 20,000 kilowatts of heat (4000 kilowatts of electrical output) in the event operation of the plants proves this possible.},

  doi          = {10.2172/4359623},

  journal      = {},

  number       = ,

  volume       = ,

  place        = {United States},

  year         = {1947},

  month        = {9}

}

Copy to clipboard

Technical Report:

View Technical Report (92.10 MB)

DOI: 10.2172/4359623

Save / Share:

Export Metadata

Save to My Library

Similar records in OSTI.GOV collections:

40-MW(E) PROTOTYPE HIGH-TEMPERATURE GAS-COOLED REACTOR POSTCONSTRUCTION RESEARCH AND DEVELOPMENT PROGRAM. Summary Progress Report for the Period Ending July 31, 1964

Technical Report None
40-MW(E) PROTOTYPE HIGH-TEMPERATURE GAS-COOLED REACTOR RESEARCH AND DEVELOPMENT PROGRAM. Summary Report for the Period January 1, 1959-December 31, 1959 and Quarterly Progress Report for the Period October 1, 1959-December 31, 1959

Technical Report None

The HTGR prototype plant (Peach Bottom Power Reactor) is being designed to produce steam at l450 psi and 1000 deg F and to have a net capacity of 40 Mw(e). The fuel temperatures and gas pressures will be approximately the same as those required for larger plants. The reactor data and operating conditions for the graphite-clad core are given. The reactor and primary coolant systems are described. The prospects for development of the graphite-clad fuel element in time for use in the first loading of the reactor were improved by important advances in methods of fabrication and testing of bothmore »« less
DOI: 10.2172/4105200

Full Text Available
40-MW(E) PROTOTYPE HIGH-TEMPERATURE GAS-COOLED REACTOR POSTCONSTRUCTION RESEARCH AND DEVELOPMENT PROGRAM. Summary Progress Report for the Period Ending, October 1964

Technical Report None
DOI: 10.2172/4580708

Full Text Available
GAS-COOLED HIGH-TEMPERATURE NUCLEAR REACTOR DESIGN TECHNOLOGY. I. INTRODUCTION. 2. INFLUENCE OF PERFORMANCE REQUIREMENTS ON POWER PLANT COMPONENT SPECIFICATIONS. 3. REACTOR DESIGN

Technical Report Moon, C.W. ; Motsinger, R.E. ; Crook, J.M. ; ...

Design technology for gas-cooled high-temperature nuclear reactors is described by recounting the solutions to reactor design problems used for the direct air cycle portion of the national Aircraft Nuclear Propulsion (ANP) Program. Major shield and control system design problem solutions as related to the reactor are also recounted. General characteristics of nuclear-powered gas- turbine cycles of gas-cooled reactors and of shields are discussed and means are indicated for relating those characteristics to basic requirements of the nuclear reactor. A general discussion of reactor design, including initial and final design sequences is also presented. (auth)
High-temperature gas-cooled reactor steam cycle/cogeneration. Design and technology development plan for Nuclear Steam Supply System. Volume I: Summary

Technical Report Not Available

The HTGR-SC/C Lead Project is proposed as a joint Industry/Government venture for the development and construction of an HTGR-SC/C Lead Plant. The Design and Technology Development Plan is presented in three volumes of which this document is Volume I, Summary. Section 2 of this document presents the basis for the remaining development work in terms of the evolution in requirements imposed by the User, Regulator and Vendor institutions since the Fort St. Vrain demonstration plant was committed. Development costs amounting to an estimated $483M are discussed in context with the effort to: design and document the NSSS of a commercialmore »« less

Similar Records