DESCRIPTION AND ANALYSIS OF NUSU CRITICAL EXPERIMENTS WITH ANNULAR MULTI- TUBE SUPERHEATER FUEL ELEMENTS

Technical Report · Wed Aug 01 04:00:00 EDT 1962

OSTI ID:4728306

Klann, P G; Dechand, C O; Gormley, M F; Crudele, J S

In previous phases of the Nuclear Superheat Program (NUSU), measurements were made with critical arrays of rod type and of double annular type fuel elements in both clean arrays and in geometries simulating superheating core designs. The assemblies were light water moderated and the steam channels contained either water (flooded) or air (voided). Experiments and analysis were continued on critical cores built of annular multitube fuel elements spaced on a triangular lattice, with two lattice spacings. These fuel elements consist of two concentric stainless steel tubes in which the annular region between the tubes is filled with 2.55 wt% enriched UO/sub 2/ vibratory compacted to 84% theoretical density. The fused oxide was re-claimed from the double annular elements and was re-used without any re-processing. The fuel annulus is penetrated by the steam superheater channels which are 15 full length stainless steel tubes arranged in a circular pattern. In the operating reactor, water is evaporated on the inner and outer clad surfaces of the fuel element. The annular multitube fuel element design is considered superior to the double annular fuel element from the standpoint of performance, reliability, and cost of fabrication and was therefore suggested as an alternate fuel element for the NUSU Reference Design. Data were obtained for checking the validity of the physics analysis methods which were developed for the superheat fuel elements previously investigated, and also to guide any further development of analysis techniques. Comparisons were made between calculation and measurement of criticality and power distribution, and also of the lattice parameters connected with the resonance escape probability and thermal utilization. Measurements were made at water-to-oxide volume ratios of 2.210, 1.485, and 1.300, which correspond to the hydrogen-to-U/sup 235/ atomic ratios encountered in operating cores. Agreement on reactivity to within one percent was obtained between calculation and measurement for clean critical configurations at these water-to-oxide volume ratios, with the steam tubes alternately flooded and voided, and for critical configurations containing heterogeneities such as water gaps formed by removing fuel elements or by the addition of poison elements. Reasonable agreement between measured and calculated power distributions around centrally located core heterogeneities was also obtained with an over-estimation in power of about 18% noted for fuel elements located at the vertices of rather thick, 2.86 in., Y- shaped water gaps at the center of the core. The tendency to over-estimate the power near water gaps was previously noted in the analysis of double annular cores. It was shown that the predicted power in a fuel element is insensitive to the thermal cross section schenie. Also, it was shown that the over-estimate is not caused by the procedure of homogenizing the fuel rods adjacent to water channels as though they were in an infinite lattice. Measurements were made of the dimensional and fuel loading variances along the length as well as around the annular circumference near the axial midsection of the fuel elements used in the critical experiments, and the effect of these non-uniformities on the physics characteristics of the core was determined. The variation of the azimuthal power generation around fuel elements situated adjacent to large water gaps was also measured to determine azimuthal power peaking factors which must be taken into account in the thermal analysis. In addition, criticality measurements were made using flux trap or rectifier control elements. The flux trap control element is essentially a tubular, water-filled control rod consisting of an annulus of neutron-absorbing control material surrounding a central water region. The potential advantages of this element lie in the additional control worth obtained, due to capture of neutrons slowed down in the water hole. Elements of this type were specified in the NUSU reference design. (auth)

• Research Organization: Combustion Engineering, Ind., Windson, Conn.
• NSA Number: NSA-17-015388
• OSTI ID: 4728306
• Report Number(s): CEND-167
• Country of Publication: United States
• Language: English

Related Subjects

BOILING
CAPTURE
COMPACTING
CONFIGURATION
CONTROL ELEMENTS
CRITICAL ASSEMBLIES
CRITICALITY
CROSS SECTIONS
DENSITY
DISTRIBUTION
ENRICHMENT
EQUATIONS
ERRORS
FABRICATION
FLUID FLOW
FUEL ELEMENTS
MEASURED VALUES
NEUTRON FLUX
NEUTRONS
NUSU
POWER
REACTIVITY
REACTOR CORE
REACTOR TECHNOLOGY
REACTORS
RESONANCE ESCAPE PROBABILITY
RODS
SLOWDOWN
STAINLESS STEELS
STEAM
SUPERHEATING
THERMAL NEUTRONS
THERMAL UTILIZATION
TUBES
URANIUM DIOXIDE
VARIATIONS
WATER COOLANT
ZONES