CAROLINAS VIRGINIA NUCLEAR POWER ASSOCIATES, INC., RESEARCH AND DEVELOPMENT PROGRAM QUARTERLY PROGRESS REPORT, APRIL-MAY-JUNE 1960

Technical Report · Tue Oct 31 00:00:00 EST 1961

OSTI ID:4106878

; ; 9 7 C ; ? 7 9 : 8 7 5 < NDC code. This code is used (for the IBM-704) to calculate lattice parameters for cells of pressure tube reactors. The cell model is composed of four concentric homogeneous regions, namely, the fuel-coolant region, thermal baffle region, tube region, and moderator region. Various fuel cycling schemes wvere studied using the CYCLONE code. Tworegion and three-region fuel cycling were considered and burn-up, radial hot channel factors, and power distribution were determined. A test unit was designed and fabricated to ascertain the effectiveness of the CVTR thermal baffles. The test facility is flexible. Some tests were run and the results obtained were within 20% of Jakob correlation. More refined calculations of the heat losses in the pressure tube were made. The heat losses were determined to be 27% higher than originally calculated. Testing of the in-pile size Zircaloy-stainless steel Conoseal joint was concluded. The joint performed well under temperature and pressure conditions expected during normal CVTR operation. A change was made to the reference design, specifying Conoseal joints rather than seal-welded joints. Progress on the ultrasonic inspection of pressure tubes was confined to the development of a suitable wedge for producing the desired shear wave angles in zirconium and its alloys. It appears that a lead alloy will be most suitable for this purpose. The CVTR Reference Design II clad thickness of 0.033 in. was established on the basis of minimum properties of Zircalov2 in the annealed condition. All material and equipment required for the tests to determine the thermal resistance between the Zircaloy cladding and the UO/sub 2/ fuel have been procured. The test apparatus is in the final stages of fabrication. The Zircaloy specimens are completed and thermocouple holes are being drilled into the UO/sub 2/ specimens. Fabrication of the in-pile loop for the CVTR Irradiation Program was completed. Work on the interconnecting piping was completed and the in-pile loop was installed in the Westinghouse Test Reactor. The design of the pressure tube header assembly was completed. All the calculated stresses were less than the allowable stresses. New layout drawings of the reactor compartment were completed showing the top-suspended moderator tank, the selfstanding side and bottom thermal neutron shields, and other recent modifications. The refueling concept was changed from one wherein fuel assemblies would be moved between U-tubes in the core to one wherein entire U- tubes with fuel would be moved. The refueling machine design was revised to incorporate this change in concept. The system proposed for detection of leakage of heavy water in the stcam generator utilizes a continuous on-stream infrared spectrometer plus periodic laboratory counting of the H/sup 3/ activity in the secondary water. The pectrometer will measure leakage rates as low as 0.09 lb/ hr and the H/sup 3/ monitor as low as 0.001 lb/hr. A study was completed to determine the effect of using direct river water cooling in the component cooling system on the ability to detect heavy water leakage. The heavy water lost before detection was determined as a function of leak rate. A simple mathematical model for the oil-fired superheater was obtained. A detailed over-all control study was initiated based on a constant cold leg program and taking into account the pressurizer and superheater. A set of runs is being made to check the control characteristics of the plant and its inherent stability, A study of the reactivity requirements for an effective scram was completed. It was concluded that an insertion of approximately --3.5% DELTA k is adequate for an effective scram during a period of time from 0 to 5 sec following the insertion. Inertia and scram requirements in case of a complete loss of flow accident were determined. Assuming inertia is added to only one of the two primary coolant pumps, 2.77 x 10/sup 6/ ft-lb of stored

• Research Organization: Westinghouse Electric Corp. Atomic Power Dept., Pittsburgh
• NSA Number: NSA-15-008339
• OSTI ID: 4106878
• Report Number(s): CVNA-59
• Resource Relation: Other Information: Orig. Receipt Date: 31-DEC-61
• Country of Publication: Country unknown/Code not available
• Language: English

Related Subjects

REACTOR TECHNOLOGY
ACCIDENTS
ALUMINUM
BARRIERS
BURNUP
CADMIUM
COMPUTERS
CONTROL
CONTROL ELEMENTS
COOLANT LOOPS
COOLANTS
COOLING
CVTR
DETECTION
EFFICIENCY
FUEL ELEMENTS
FUELS
GEOMETRY
HEAVY WATER
IBM 704
INSPECTION
JACOB CORRELATION
LEAKS
MATHEMATICS
MEASURED VALUES
MODERATORS
NEUTRONS
PETROLEUM
PLANNING
POWER
PRESSURE
PUMPS
RADIOACTIVITY
REACTIVITY
REACTOR FUELING
REACTOR SAFETY
REACTORS
RECOVERY
REFLECTORS
RODS
SEALS
SHIELDING
SHUTDOWN
SPECTROMETERS
SPHERES
STABILITY
STEAM
SUPERHEATING
TESTING
THERMAL NEUTRONS
TRITIUM
TUBES
WASTE PROCESSING
WTR
ZONES