ML19317D746
| ML19317D746 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 02/15/1973 |
| From: | Thies A DUKE POWER CO. |
| To: | Anthony Giambusso US ATOMIC ENERGY COMMISSION (AEC) |
| Shared Package | |
| ML19317D747 | List: |
| References | |
| NUDOCS 7912100513 | |
| Download: ML19317D746 (2) | |
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t-Duxz POWER COMPANY Powza Burr orwo 4aa SouTu Cnemen StazzT, CHAR 14TTz, N. C. asaos
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- ioeN, PeoOdC?lON AND TRANSes49tsON February 15, 1973
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Mr. Angelo Giambusso Deputy Director for Reactor Projects ro \\ g [
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,l United States Atomic Energy Commission Washington, D. C.
20545 Re: Oconee Nuclear Station, Unit 1 Docket No. 50-269
Dear Mr. Giambusso:
Pursuant to 10CFR50.59, Duke Power Company is hereby requesting that the Oconee Nuclear Station Unit 1 Technical Specifications be revised.
Oconee Unit 1 is under operation pursuant to License No. DPR-38.
Attached are 22 copies of a report entitled, "Results of Oconee 1 Hot Functional Testing Internals Vibration Monitoring Program." We believe that this report provides sufficient information to confirm the redesign of the reactor internals as presented in Babcock & Wilcox Topical Report BAW-10051 and shows that the prototype vibration measurement program identified in Babcock & Wilcox Topical Report BAW-10038 has been i
successfully completed.
Therefore, Duke Power Company requests that Technical Specification 3.11. " Maximum Power Restriction," which is attached to License DPR-38 as part of Appendix A, be rescinded.
Sincerely, A. C. Thies
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Page 2 February 15, 1973 l
A. C. THIES, being duly sworn, states that he is Senior Vice President of Duke Power Company; that he is authorized on the part of said Company to sign and file with the Atomic Energy Commission this application for change in Technical Specifications; and that all statements and matters set forth therein are true and correct to the best of his knowledge.
A. C. Thies Senior Vice President ATTEST:
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sistant Secretary Subscribed and sworn to before me on this 15th day of February, 1973.
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RESULTS OF OCONEE I HOT FUNCTIONAL TESTING g
INTERNALS VIBRATION HONITORING PROGRAM
1.0 INTRODUCTION
BAW-10038, Rev. 1, " Prototype Vibration Measurement Program for Reactor Internals (177-Fuel Assembly Plant)," outlined the program to be imple-mented at Oconee Nuclear Station during hot functional testing (HFT) of Unit 1.
During the period November 1972 to January 1973, this program was successfully carried out.
The purpose of this report is to summarize the results of the vibration t
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monitoring program to confirm the design of the reactor internals as presented in BAW-10051, Rev.1, " Design of Reactor Internals and Incore Instrument Nozzles for Flow-Induced Vibration."
2.0 COLLECTION OF DATA The response of the reactor internals and incore instrument nozzles was monitored by 40 strain gages and 12 accelerometers.
In addition, there were pressure cells attached to the core support shield and thermal shield to determine the static and dynsmic pressure variations. Pressure sensing lines were also connected to an incore instrument nozzle and mating guide tube to provide data for calculating the cross-flow velocities in the vessel lower head. Figure 1 shows the instrument locations.
Strain gages were placed on two incore instrument nozzles, on two incore instrument guide tubes (one guasetted and one non-pasastrad), two thermal shield lower support bolts, six thermal shield support pads, at two locations on the plenum cylinder between the outlet holes, on one surveillance specimen holder tube, and on one shroud tube. Accelerometers were placed on the thermal shield, inside a guide tube to record the motions of the flow distributor, and on the core screen. External accelerometers were placed on the vessel support skirt, and on a vessel head stud.
The accelerometer on the screen was utilized to monitor for loose parts, and those external on the reactor vessel were used for reference.
Seven pressure transducers were located around the outside dismeter of the thermal shield, and four around the outside diameter of the core support shield. Pressure sensing lines were placed, four each in an incore instrument nozzle and four each in an incore guide tube.
All of the sensor output signals and the conditioning electronics were in good working order when the reactor coolant pumps were initially operated. For the duration of the preoperational testing period, the sensors and the data acquisition system continued to work with very good reliability and accuracy.
Some data channels, however, did become inoperable during the test. At the end of HFT, there were 14 channels (mostly strain gages) that were inoperative. The failure of these signal, channels did not, however, prevent the obtaining of sufficient data to assess the structural adequacy of the internals.
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During HFT a total of 280 hours0.00324 days <br />0.0778 hours <br />4.62963e-4 weeks <br />1.0654e-4 months <br /> were accumulated (logged) with four reactor coolant pumps operating (full flow conditions) at conditions 0
of approximately 530 F and 2155 psig.
Consequently, the thermal shield which had the lowest measured structural frequency of 12 Hz, was sub-7 jected to more than 10 cycles.
The reactor internals were also subjected to an additional 280 hours0.00324 days <br />0.0778 hours <br />4.62963e-4 weeks <br />1.0654e-4 months <br /> of conditions other than full flow operation.
This was sufficient to accumulate an additional 107 cycles of vibration on the thermal shield.
Outputs of the sensors recorded on magnetic tape were continuously monitored by an on-line computer.
3.0 RESULTS Table 1 compares the stresses and deflections measured during the hot functional testing vibration monitoring program with the predicted and allowable values as presented in BAW-10038. As indicated by the large ratio of allowable values to measured values, a substantial acceptance margin exists for all components.
It is noted that the predicted responses of the various components were calculated using conservative criteria.
Consequently, the generally lower values of measured response are not unexpected.
4.0 POST HOT FUNCTIONAL TESTING INSPECTION Following the conclusion of HFT and the concurrent vibration monitoring program, the reactor internals were removed from the vessel and the post-test inspection program was implemented.
The purpose of this program was to visually inspect all major internals, surfaces and/or parts for any indications of distress, loose parts, cracking, fretting, or distortion as a recult of HFT.
The results of this inspection program indicated that the reactor internals sustained no structural damage as a result of HFT.
It was determined that no deterioration that might affect the structural integrity of the internals had occurred.
Only expected, minor indi-cations of fretting and wear were observed on metal-to-metal contact surfaces; e.g.,
internals vent valve seats.
It was discovered, however, pr'or to HFT, that the surveillance specimen holder tubes required more torque than anticipated to rotate them for internals insertion into the reactor vessel.
The same situation existed after HFT.
This situation was investigated as part of the inspection routine and it was determined that bearing and tube misalignment were the cause of the increased torques.
Corrective actions were taken:
(a) The surveillance tubes were straightened to improve alignment.
Because of schedule, the identical surveillance tubes for Oconee 2 were installed on Oconee 1.
The Oconee 1 tubes will be used on Oconee 2.
l (b) The installation procedure was modified to improve the alignment.
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This situation did not affect the structural adequacy of the surveillance tubes.
5.0 CONCLUSION
S Evaluation of the measured vibrational responses and the direct visual examination of the reactor internals has revealed that the structures experienced extremely low stresses during operation and sustained no structural damage. All data indicate that the internals will perform in the intended manner with no adverse deterioration occurring over the extended life of the unit.
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VIBRATION TEST RESULTS 1
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Measured Values Predicted Acceptance Ratio of Allowable Component RMS Peak-to-Peak +2 Values Criteria To Measured (Peak-to-Peak 12) i Nozzle
<100 psi
<100 psi 3,000 psi 5,800 psi
>58.0
,1 Guide Tube:
Gussetted
<100 psi
<100 psi 2,050 psi 7,000 psi
>70.0 Non-Gussetted
<100 psi
<100 psi 2,700 psi 8,600 psi
>86.0 l
l Flow Distributor
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.003 in.
.011 in.
.025 in.
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e Holder Tube 250 psi 300 psi 10,000 psi 13,500 psi 45.0 e
Upper Thermal Shield Support 200 psi 300 psi 4,150 psi 13,500 psi 45.0
.i Lower Thermal Shield Bolt 760 psi 3,900 psi 3,500 psi 7,500 psi 2.0 j
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e GENERAL ARRANGEMENT OF REACTOR VESSEL AND INTEE"ALS S110 WING INSTRUMENTATION LOCATIONS i
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