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Title: AIRCRAFT NUCLEAR PROPULSION PROJECT QUARTERLY PROGRESS REPORT FOR PERIOD ENDING MARCH 10, 1956

Abstract

9 8 9 9 < 9 < < > @ : : 9 ; 9 4 ayouts were completed on all major subassemblies of the ART. Recent calculations have indicated that the heat flux through the core shells of the ART will be higher than previously estimated. The thickness of the lead portion of the reactor shield was reduced from 7 to 4.3 in. to obtain space for increased neutron shielding around the pumps. Studies of flow in a full-scale plastic model of the core were continued. Flow diagrams and instrumentation lists were prepared for the Engineering Test Unit (ETU). which is a nonnuclear mockup of the ART. Modifications were made in the proposed ART control system on the basis of information obtained with the reactor simulator. Three separate calculations of the activity that will be produced in the NaK in the fuel-to-NaK heat exchangers of the ART were norraalized to the current ART design. An estimate was made of the activity to be expected in the ART radiators because of the deposition on the radiator tubes or header walls of constituents of Inconel that will become activated in the heat exchangers and will be carried to the radiators inmore » the NaK streann. A calculation of the selfabsorption of decay gamrua rays in the ART fuel dump tank indicated a value of 90%. and therefore ccoling facilities must be provided to remove almost all the decay heat. Disassembly and sectioning of in- pile loop No. 3 which operated in the MTR were completed. Synthetic lubricants in the UCON LB series were investigated to determine their stability as pump lubricating fluids in the ART. Additional tests of the performance of the ART sodium pump with water were made. Difficulties were encountered in high- temperature tests of the fuel pump with NaK. Operations were continued on heat exchanger stands. A stand for testing the thermal stability of the outer core shell of the ART is in operation. Materials Research. Additional phase- equilibrium studies were carried out in several fluoride systems in onder to gain a better understanding of the structure and to aid in devising improved fuels. A material containing NaF--ZrF/sub 4/--UF/sub 4/ was found to show promise as a lower melting. nonsegregating substitute for Na/sub 2/UF/sub 6/ as a fuel concentrate for reactor systeras. Studies were made on the UF/sub 4/--UO/sub 2/ system which led to a value of 1034 deg C for the melting pcint of pure UF/sub 4/. Experimental work was carried out in the following Zr-bearing systems: KF--ZrF/ sub 4/, RbF -- ZrF/sub 4/. and NaF --KF -- ZrF/sub 4/. Thermal analysis data were obtained on the RbF--BeF/sub 4/ system. The program for studying chemical reactions of molten fluorides was continued. Studies of the reactions of Ta and W with UF/sub 4/ in NaK--LiF--KF eutectic showed that neither metal was stable and that they offered little or no advantage over Cr. The effects on corrosion of varying the ratio of hot-leg surface area to volume of fluoride fuel mixture in a forcedcirculation Inconel loop were investigated. Exaramations of two forced-circulation loops that had operated with the fuel mixture NaF--KF--LiF (UF/ sub 4/ + UF/sub 3/) clearly demonstrated that the presence of U/sup 3+/ in such a mixture is very effective in reducing hot-leg attack. Two Inconel forced- circulation loops in which NaK was circulated completed 1000 hr of operation with a temperature gradient of 300"F and a maximum fluid teraperature of 1500 deg F. Exaruinations of two Hastelloy B thermal connection loops operated for 500 hr with an alkali metal fluoride mixture containing UF/sub 4/ at a maximum temperature of 1500 deg F showed the maximum attack to be less than 2 mils. Inconel tube-to-header joints brazed with Ni--Cr--Ge--Si low-cross-section aIloys were tested in Na. NaK, and NaF-- ZrF/sub 4/--UF/sub 4/ in seewaw apparatus. One specimen of Sm/sub 2/O/sub 3/ and two specimens of a commercial rare-earth oxide mixture (Sm/sub 2/O/sub 3/ + Gd/sub 2/O/sub 3/) were tested in static sodium in Inconel containers at 1500 deg F. Two type 316 stainless steel thermal- convection loops were operated with Na to study the effect of a diffusion cold trap on the amount of corrosion and mass transfer observed in« less

Publication Date:
Research Org.:
Oak Ridge National Lab., Tenn.
OSTI Identifier:
4147988
Report Number(s):
ORNL-2061(Pts. I, II, and III)
NSA Number:
NSA-14-023757
DOE Contract Number:  
W-7405-ENG-26
Resource Type:
Technical Report
Resource Relation:
Other Information: Decl. with deletions Nov. 10, 1959. Orig. Receipt Date: 31-DEC-60
Country of Publication:
United States
Language:
English
Subject:
REACTOR TECHNOLOGY; BERYLLIUM FLUORIDES- RUBIDIUM FLUORIDES; FLUORIDES- FUELS- FUSED SALT FUEL- HIGH TEMPERATURE- - MELTING POINTS- MIXING- PHASE DIAGRAMS- SODIUM COMPOUNDS- SODIUM FLUORIDES- STABILITY- URANIUM COMPOUNDS- URANIUM DIOXIDE- URANIUM FLUORIDES- ZIRCONIUM FLUORIDES; POTASSIUM FLUORIDES- SODIUM FLUORIDES- ZIRCONIUM FLUORIDES; POTASSIUM FLUORIDES- ZIRCONIUM FLUORIDES

Citation Formats

. AIRCRAFT NUCLEAR PROPULSION PROJECT QUARTERLY PROGRESS REPORT FOR PERIOD ENDING MARCH 10, 1956. United States: N. p., 1956. Web.
. AIRCRAFT NUCLEAR PROPULSION PROJECT QUARTERLY PROGRESS REPORT FOR PERIOD ENDING MARCH 10, 1956. United States.
. Wed . "AIRCRAFT NUCLEAR PROPULSION PROJECT QUARTERLY PROGRESS REPORT FOR PERIOD ENDING MARCH 10, 1956". United States.
@article{osti_4147988,
title = {AIRCRAFT NUCLEAR PROPULSION PROJECT QUARTERLY PROGRESS REPORT FOR PERIOD ENDING MARCH 10, 1956},
author = {},
abstractNote = {9 8 9 9 < 9 < < > @ : : 9 ; 9 4 ayouts were completed on all major subassemblies of the ART. Recent calculations have indicated that the heat flux through the core shells of the ART will be higher than previously estimated. The thickness of the lead portion of the reactor shield was reduced from 7 to 4.3 in. to obtain space for increased neutron shielding around the pumps. Studies of flow in a full-scale plastic model of the core were continued. Flow diagrams and instrumentation lists were prepared for the Engineering Test Unit (ETU). which is a nonnuclear mockup of the ART. Modifications were made in the proposed ART control system on the basis of information obtained with the reactor simulator. Three separate calculations of the activity that will be produced in the NaK in the fuel-to-NaK heat exchangers of the ART were norraalized to the current ART design. An estimate was made of the activity to be expected in the ART radiators because of the deposition on the radiator tubes or header walls of constituents of Inconel that will become activated in the heat exchangers and will be carried to the radiators in the NaK streann. A calculation of the selfabsorption of decay gamrua rays in the ART fuel dump tank indicated a value of 90%. and therefore ccoling facilities must be provided to remove almost all the decay heat. Disassembly and sectioning of in- pile loop No. 3 which operated in the MTR were completed. Synthetic lubricants in the UCON LB series were investigated to determine their stability as pump lubricating fluids in the ART. Additional tests of the performance of the ART sodium pump with water were made. Difficulties were encountered in high- temperature tests of the fuel pump with NaK. Operations were continued on heat exchanger stands. A stand for testing the thermal stability of the outer core shell of the ART is in operation. Materials Research. Additional phase- equilibrium studies were carried out in several fluoride systems in onder to gain a better understanding of the structure and to aid in devising improved fuels. A material containing NaF--ZrF/sub 4/--UF/sub 4/ was found to show promise as a lower melting. nonsegregating substitute for Na/sub 2/UF/sub 6/ as a fuel concentrate for reactor systeras. Studies were made on the UF/sub 4/--UO/sub 2/ system which led to a value of 1034 deg C for the melting pcint of pure UF/sub 4/. Experimental work was carried out in the following Zr-bearing systems: KF--ZrF/ sub 4/, RbF -- ZrF/sub 4/. and NaF --KF -- ZrF/sub 4/. Thermal analysis data were obtained on the RbF--BeF/sub 4/ system. The program for studying chemical reactions of molten fluorides was continued. Studies of the reactions of Ta and W with UF/sub 4/ in NaK--LiF--KF eutectic showed that neither metal was stable and that they offered little or no advantage over Cr. The effects on corrosion of varying the ratio of hot-leg surface area to volume of fluoride fuel mixture in a forcedcirculation Inconel loop were investigated. Exaramations of two forced-circulation loops that had operated with the fuel mixture NaF--KF--LiF (UF/ sub 4/ + UF/sub 3/) clearly demonstrated that the presence of U/sup 3+/ in such a mixture is very effective in reducing hot-leg attack. Two Inconel forced- circulation loops in which NaK was circulated completed 1000 hr of operation with a temperature gradient of 300"F and a maximum fluid teraperature of 1500 deg F. Exaruinations of two Hastelloy B thermal connection loops operated for 500 hr with an alkali metal fluoride mixture containing UF/sub 4/ at a maximum temperature of 1500 deg F showed the maximum attack to be less than 2 mils. Inconel tube-to-header joints brazed with Ni--Cr--Ge--Si low-cross-section aIloys were tested in Na. NaK, and NaF-- ZrF/sub 4/--UF/sub 4/ in seewaw apparatus. One specimen of Sm/sub 2/O/sub 3/ and two specimens of a commercial rare-earth oxide mixture (Sm/sub 2/O/sub 3/ + Gd/sub 2/O/sub 3/) were tested in static sodium in Inconel containers at 1500 deg F. Two type 316 stainless steel thermal- convection loops were operated with Na to study the effect of a diffusion cold trap on the amount of corrosion and mass transfer observed in},
doi = {},
url = {https://www.osti.gov/biblio/4147988}, journal = {},
number = ,
volume = ,
place = {United States},
year = {1956},
month = {6}
}

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