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Title: URANIUM ALLOY FUEL ELEMENT FABRICATION DEVELOPMENT FOR HNPF CORE I

Abstract

A fuel element, consisting of a cluster of stainlesssteel-clad sodium- bonded, U - 10 wt.% Mo fuel rods, has been proposed for the first core loading in the Hallam Nuclear Power Facility. Analytical data on density, soundness, chemical composition, dimensions, and surface finish were obtained from more than 100 heats of fuel slugs which were vacuum cast in graphite molds. The quality levels obtained were satisfactory, and casting was proven to be a feasible fabrication method for U - 10 wt.% Mo fuel slugs. Extrusion and liquid pouring were both investigated as methods of handling sodium to bond fuel rods. The extrusion technique was shown to be impractical because of difficulty in supplying a sufficient volume of sodium to fill the rod. Satisfactory bonding was obtained using the liquid technique. Two full-scale fuelelement mock-ups were fabricated. Each element contained 19 fuel rods located by spacers made from spot welded strips and enclosed in a process tube. Assembly problems encountered in the first mock-up were corrected by modifying the fuel-rod spacers. Methods were developed for safe handling of the fuel rods and other long (15 ft) components. The second mock-up demonstrated that the fuel element was suitable for fabrication and assembly.more » Physical testing was performed on fuel-rod end closures, cladding support springs, the variable-orifice thermocouple, and the fuel-rod spacers. The end closures and support springs were shown to have adequate strength. The thermocouple functioned well at temperatures and flexure conditions stimulating reactor operation. A minimum material thickness was determined for the fuel-rod spacers. (auth)« less

Authors:
Publication Date:
Research Org.:
Atomics International. Div. of North American Aviation, Inc., Canoga Park, Calif.
Sponsoring Org.:
USDOE
OSTI Identifier:
4146802
Report Number(s):
NAA-SR-5291
NSA Number:
NSA-15-003046
DOE Contract Number:  
AT-11-1-GEN-8
Resource Type:
Technical Report
Resource Relation:
Other Information: Orig. Receipt Date: 31-DEC-61
Country of Publication:
United States
Language:
English
Subject:
METALS, CERAMICS, AND OTHER MATERIALS; BONDING; CASTING; CLOSURES; CONFIGURATION; DEFECTS; DENSITY; EXTRUSION; FUEL CANS; FUEL ELEMENTS; GRAPHITE; HALLAM NUCLEAR POWER FACILITY; MECHANICAL STRUCTURES; MELTING; MOCKUP; MOLYBDENUM ALLOYS; SODIUM; SPACERS; SPRINGS; STAINLESS STEELS; SURFACES; TESTING; THERMOCOUPLES; THICKNESS; URANIUM ALLOYS; VACUUM

Citation Formats

Cobb, S M. URANIUM ALLOY FUEL ELEMENT FABRICATION DEVELOPMENT FOR HNPF CORE I. United States: N. p., 1960. Web. doi:10.2172/4146802.
Cobb, S M. URANIUM ALLOY FUEL ELEMENT FABRICATION DEVELOPMENT FOR HNPF CORE I. United States. https://doi.org/10.2172/4146802
Cobb, S M. Tue . "URANIUM ALLOY FUEL ELEMENT FABRICATION DEVELOPMENT FOR HNPF CORE I". United States. https://doi.org/10.2172/4146802. https://www.osti.gov/servlets/purl/4146802.
@article{osti_4146802,
title = {URANIUM ALLOY FUEL ELEMENT FABRICATION DEVELOPMENT FOR HNPF CORE I},
author = {Cobb, S M},
abstractNote = {A fuel element, consisting of a cluster of stainlesssteel-clad sodium- bonded, U - 10 wt.% Mo fuel rods, has been proposed for the first core loading in the Hallam Nuclear Power Facility. Analytical data on density, soundness, chemical composition, dimensions, and surface finish were obtained from more than 100 heats of fuel slugs which were vacuum cast in graphite molds. The quality levels obtained were satisfactory, and casting was proven to be a feasible fabrication method for U - 10 wt.% Mo fuel slugs. Extrusion and liquid pouring were both investigated as methods of handling sodium to bond fuel rods. The extrusion technique was shown to be impractical because of difficulty in supplying a sufficient volume of sodium to fill the rod. Satisfactory bonding was obtained using the liquid technique. Two full-scale fuelelement mock-ups were fabricated. Each element contained 19 fuel rods located by spacers made from spot welded strips and enclosed in a process tube. Assembly problems encountered in the first mock-up were corrected by modifying the fuel-rod spacers. Methods were developed for safe handling of the fuel rods and other long (15 ft) components. The second mock-up demonstrated that the fuel element was suitable for fabrication and assembly. Physical testing was performed on fuel-rod end closures, cladding support springs, the variable-orifice thermocouple, and the fuel-rod spacers. The end closures and support springs were shown to have adequate strength. The thermocouple functioned well at temperatures and flexure conditions stimulating reactor operation. A minimum material thickness was determined for the fuel-rod spacers. (auth)},
doi = {10.2172/4146802},
url = {https://www.osti.gov/biblio/4146802}, journal = {},
number = ,
volume = ,
place = {United States},
year = {1960},
month = {11}
}