NUCLEAR SUPERHEAT DEVELOPMENT PROGRAM. DESIGN, FABRICATION AND IRRADIATION TEST OF A DOUBLE-ANNULAR BOILING SUPERHEATING FUEL ELEMENT

Technical Report · Fri Jun 28 00:00:00 EDT 1963

OSTI ID:4644896

Weems, S J; Turner, S E; Wiley, J S; Lichtenberger, H; Nulk, D E

The experimental fuel element for the in-pile test is a shortened version of the outer annulus section (about 1.5in. OD by 1-in. 1D) of the double-annular combination boiling-superheating fuel element of the type used in the reference design of the NUSU 200-Mw(e) power plant. Th external clad for this outer annular section is 0.028-in.thick type 347 stainless steel in direct contact with the boiling -water moderator; the internal clad is 0.025-in.thick Inconel-X tubing which is in contact with superheater steam. The fabrication test activity was directed at the production of an adequate end closure for the outer and inner cladding tubes. The preparation of a welded end closure produced satisfactory results. The thermal cycle test work consisted of electrically heating and thermally cycling the inner cladding in air between about 700 and cladding at about 600 deg F in a liquid metal environment. This test activity was terminated after the last test fuel element had successfully withstood 825 thermal cycles. The in-pile test, while of relatively short duration (total reactor exposure was 42 days, maximum fuel burnup was about 750 Mwd/t U), was conducted under severe conditions. Although the experimental fuel element was subjected to extensive thermal cycling during the in-pile test, no excessive damage or distortion was found. There were 20 scrams from power levels with high clad temperatures, and there were numerous power level changes. During the in-pile test, the Inconel-X was hardened, and therefore strengthened, in regions where the high operating temperatures would have otherwise caused a reduction of strength. This strength enhancement, caused by age hardening at temperatures around 1300 deg F resulted in a more uniform distribution of strain during thermal cyciing. Corrosion oi the steamcooled Inconel-X was limited to a uniform scale about 0.8 thick with no evidence of localized corrosion or cracking However, during the latter part of the irradiation test, the activity of the steam coolant was measurably higher down stream of the fuel element. This higher downstream activity can be satisfactorily explained by either or both of two mechanisms; the test fuel element may have lost clad ding integrity, or there may have been water in-leakage from the reactor moderator. The amount of water inleakage increased (as measured by Na/sup 2/ throughput) at the same time the downstream activity increased above the upstream activity. Post-irradiation leak tests yielded no in dication of a cladding leak. Visual and metallographic examinations showed no indication that the cladding had undergone any significant damage which might have resulted in a failure. It was concluded that the feasibility of the NUSU fuel element concept is demonstrated, at least for short term irradiation. Evidence to date indicates that use of such fuel elements could significantly decrease both the capital and fuel cycle costs of nuclear power reactors. (auth)

• Research Organization: General Nuclear Engineering Corp., Dunedin, Fla.; Combustion Engineering, Inc. Nuclear Div., Windsor, Conn.
• DOE Contract Number: AT(11-1)-795
• NSA Number: NSA-17-035099
• OSTI ID: 4644896
• Report Number(s): GNEC-288
• Resource Relation: Other Information: Orig. Receipt Date: 31-DEC-63
• Country of Publication: United States
• Language: English

Related Subjects

REACTOR TECHNOLOGY
ALUMINUM ALLOYS
BOILING
BURNUP
CHROMIUM ALLOYS
CORROSION
CRACKS
ECONOMICS
FABRICATION
FAILURES
FUEL CANS
FUEL ELEMENTS
HARDENING
HEATING
INCONEL ALLOYS
IRRADIATION
LEAKS
LIQUID METAL COOLANT
METALLOGRAPHY
NICKEL ALLOYS
NIOBIUM ALLOYS
NUSU
PLANNING
POWER PLANTS
RADIATION EFFECTS
RADIOACTIVITY
SHUTDOWN
SODIUM 24
STAINLESS STEELS
STEAM
SUPERHEATING
TESTING
TITANIUM ALLOYS
TUBES
WATER MODERATOR
WELDING