PRESSURIZED WATER REACTOR (PWR) PROJECT FOR THE PERIOD JUNE 24, 1960 TO AUGUST 23, 1960
= C 8 9 5 A 9 : ; : 4 9 8 : C nuclear power range linear level amplifier to resolve the problem of (1) OPERATE-TEST zero shift and (2) circuit transient. A design modification was prepared. The allowable pressure drop on the air locks and butterfly valves during the performance of DLCS-357 was increased from 0.5 psig/hr to 0.8 psig/hr and 4.0 psig/hr. respectively. This increase will eliminate the excessive maintenance required each time test DLCS-357 is performed. A design modifying the charging- pump relief valve discharge lines was completed. This design eliminated the necessity of freeze plugging the charging-pump suction lines in order to isolate relief valves 08-HlS-6 and 08-HlS-7. A design adding permanent local-flow indicators for each major component requiring cooling water was completed. This design enables the station operator to regulate the flow of cooling water to equipment. Steady-state and transient power testing of the Babcock and Wilcox and Foster Wheeler Steam Generator was conducted at Shippingport which will provide data for use in PWR Core 2 modifications. Preliminary results from operation of the PWR plant with the purification system bypassed showed some increase in radiation levels in the reactor-vessel head area but no other significant changes in plant radiation levels. Long-lived fission-product levels increased in a predictable manner. The CR-V in-pile test loop was successfully decontaminated from activated corrosion products using the alkaline permaganate ammonium citrate (APACI process to yield a decontamination factor of 10. The functional requirements for the plant container air cooling system were developed. Reactor Engineering - PWR Core 1: The design of the fuel assembly holders for the M-130 container were completed. The first three calibrations of the core thermocouples were evaluated. Fewer thermocouples showed calibration shifts with time and with temperature cycling in each successive calibration. Reactor pressure drop and core-flow instrumentation were monitored periodically since Seed 2 startup. No significant buildup of pressure drop or decrease in flow were observed. The refueling system description for PWR Core 2 was completed in accordance with established reference method using wet refueling. Preliminary testing of the full-scale carbon-steel model of the PWR Core 2 bottom support was concluded. The initial test was a determination of the load-carrying ability of the weldment alone. Phase I testing of RCH-1 was concluded. After shutdown, components were disassembled and inspected. It was concluded that, with minor exceptions, all components were tested without detrimental effects. Metallurgy of Core Materials: An interim examination of irradiation test plates containing ZrO/sub 2/-34 wt.% UO/sub 2/ and ZrO/sub 2/-46 wt.% UO/sub 2/ was completed after exposures of 25.3 x 10/sup 20/ and 18.1 x 10/sup 20/ fissions/cc. respectively, in the VH-3 Loop of the MTR. These fuel plates showed average increases in thickness of 0.004 in. and 0.0026 in., respectively, which would represent 7.0% and 4.7% increase in fuel swelling, respectively, if the changes are due entirely to the fuel. Density measurements on bulk B/sub 4/C irradiated to 1.02 x 10/sup 22/ B/sup 10/ fissions/cc showed density decreases of about 30%. Means of avoiding exaggerated grain growth during pressure-bonding were developed. A postpressure-bonding heat treatment showed proimise of converting Type C bonds to predominantly Type B bonds. Thermal-expansion measurements showed that the expansivities of ZrO/sub 2/-25 wt.% UO/sub 2/ and ZrO/sub 2/-34 wt.% fuel compositions were about 10% greater than that of pure UO/sub 2/. A 3-day corrosion test in 756-F steam on platelets of ZrO/sub 2/-25 wt.% UO/sub 2/ and ZrO/sub 2/-34 wt.% UO/sub 2/ showed the same negligible weight changes as were observed for a three-daytest in 650 deg F water. Diffusion anneal data for Xe/ sup 133/ ZrO/sub 2/- 25 U.% UO/sub 2/ indicated that diffusion coefficients are
- Research Organization:
- Westinghouse Electric Corp. Bettis Atomic Power Lab., Pittsburgh
- DOE Contract Number:
- AT-11-1-GEN-14
- NSA Number:
- NSA-14-026477
- OSTI ID:
- 4128910
- Report Number(s):
- WAPD-MRP-87
- Resource Relation:
- Other Information: Orig. Receipt Date: 31-DEC-60
- Country of Publication:
- United States
- Language:
- English
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Related Subjects
ACTIVATION- BLANKETS- COMPUTERS- CONFIGURATION- CONTROL ELEMENTS- DENSITY- DISTRIBUTION- EQUATIONS- FAST NEUTRONS- GROUP THEORY- LIFETIME- MOCKUP- MODERATORS- NEUTRON FLUX- PLUTONIUM- PROGRAMMING- QUALITATIVE ANALYSIS- QUANTITY RATIO- REACTIVITY- REACTOR CORE- SHIPPINGPORT- TEMPERATURE- URANIUM- URANIUM DIOXIDE- VARIATIONS
AIR- AMMONIUM COMPOUNDS- AMPLIFIERS- CLEANING- COOLING- CORROSION- DECONTAMINATION- ELECTRONIC EQUIPMENT- FISSION PRODUCTS- GENERATORS- IN PILE LOOPS- LIQUID FLOW- MAINTENANCE- MANGANESE COMPOUNDS- ORGANIC COMPOUNDS- POWER PLANTS- PRESSURE- PRESSURE VESSELS- PUMPS- PWR- RADIATION