IR 05000424/1987033
| ML20215A633 | |
| Person / Time | |
|---|---|
| Site: | Vogtle  |
| Issue date: | 06/03/1987 |
| From: | Burnett P, Jape F NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20214W931 | List: |
| References | |
| 50-424-87-33, NUDOCS 8706170010 | |
| Download: ML20215A633 (4) | |
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UNITE 3 STATES j meeg,#g,
NUCLE A'l CEOULATORY COMMIS$10N y
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101 MAMitTTA STMEET,N.W.
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Report No.:
50-424/87-33 l
l Licensee: Georgia Power Company l
P. O. Box 4545 l
Atlanta, GA 30302 Docket No.:
50-424 License No.: NPF-68
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Facility Name: Vogtle 1 Inspection conducted: Ma 5-8, 1987
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1nspector
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Oate Signed A/N l
Approved by:
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l F. Jape, 5e tion te r /
l Engineering Branch t
Division of Reactor Safety i
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$UMMARY
Scope:
This routine, unannounced inspection addressed the area of completed startup tests.
Results: No violations or deviations were identified.
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REPORT DETAILS 1.
Persons Contacted Licensee Employees
- T. Greene, Plant Manager
- C, E. Belflower, Quality Assurance Site Manager W. L. Burmeister, Operations Supervisor
- J. F. D'Amico, Nuclear ?afety and Compliance Manager R. J. Florian, Reactor I:ngineering Supervisor
- W. C. Gabbard, Senior Regulatory Specialist
- C. A. Griffin, Plant Engineer, Independent Safety Engineering Group T. S. Hargis, Operations Shift Supervisor
- J. Hartka, Senior Nuclear Engineer, Licensing W. F. Kitchens, Manager of Operations D. W. Schretbor, Operations Supervisor Other Itcensee employees contacted included engineers, operators, and office personnel.
Other Organizations W. Pheonix, Consul Tec R. Done, Westinghouse R. Porter, Core NRC Resident Inspectors J. F. Rogge, Senior Resident Inspector, Operations
- R. J. Schepens, Resident Inspector
- Attended exit interview 2.
Exit Interview The inspection scope and findings woro summarized on May 8,1987, with those persons indicated in paragraph I above. The inspector described the areas inspected and discussed in detail the inspection findings.
No dissenting comments were recolved from the Itcensee. The licenseo did not identify as proprietary any of the materials provided to or reviewed by the inspector during this inspection.
3.
Licenseo Action on Previous Enforcement Matters (0pon) Violation 424/87-24 01: Failure to perform an adequate surveillance of reactor coolant system leakage. The licensee had revised the procedure in response to the oral nottfication of the violation, but the now version still did not properly account for changes in pressurizer level.
The
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inspector discussed the necessary calculations at some length with the licensee staff, and at the exit interview voiced concern with time required to date to obtain a procedure that properly accounted for the difference in water properties in the sub-systems included in the surveillance.
4.
Unresolved Items No unresolved item was identified during this inspection.
5.
Review of Low Power Physics Test (72572)
Completed startup test 1-6SF-04, Pseudo Rod Ejection Test, was reviewed.
The procedure was performed on March 20, 1987 and the results accepted by
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the general manager on May 6, 1987. The measured worth of the ejected rod at zero power was 492 pcm, and was conservatively adjusted to 541 pcm, which was less than the acceptance criterton (upper limit vilue) of 860
pcm. The measured incore quadrant flux tilt was 1.03, which was in excess of the acceptance criterion of 1.02, but less than the safety review criterion limit of 1.04.
No violations or deviations were identified.
6.
Review of Power Escalation Tests (72600, 72608, 72616)
The following completed power escalation tests were reviewed:
a.
1-6SC-01, Power Coefficient Determination, was performed at 36% rated thermal power (RTP) On April 3,1987, at 48% RTP on April 18,1987, and at 75% RTP on May 3, 1987. (It is also scheduled to be performed at 90%, once that power plateau is reached). The measured parameter is the ratio of isothermal temperature coefficient to the doppler power coefficient.
In sequence the measured ratios, each an average of six or more observations, were:
-2.2, -1.69, and -1.14 degree F/%
power.
Each result satisfied the acceptance criterion of agreeing with the predicted ratio by 0.5 b.
1-65C-02, Load Swing Test, was performed at 30% power on April 10, 1987.
All acceptance criteria were satisfied.
The test is also scheduled to be performed at 75 and 100% power; so review has not been completed. The swings, both positive and negative were at least 10% rated thermal power, c.
1-600-08, Remote Shutdown Test, was performed initially on April 8, 1987, and was witnessed by region-based inspectors (see inspection report no. 50-424/87-28). They questioned the licensee's conclusion that the reactor had been held stable for 30 minutes in hot standby, when during that time average temperature dropped 46 degrees F.
The licensee agreed to re perform the portion of the test that demonstrated stability, and the appropriate sections of the test were
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completed successfully on April 14, 1987.
During the retest the average temperature was stable within a band of I degree F.
Plant management had not completed review of the test at the time of this inspection, d.
1-600-09, Loss of Offsite Power at Greater Than 10% Power Test, was performed on March 28, 1987, and witnessed by region-based inspectors as reported in inspection report no. 50-424/87-28. The plant review board (PRB) completed its review on April 21, 1987, and the results were accepted by the General Manager on April 24, 1987. Test change notice (TCN) 1-600-09-01 allowed prestarting of the diesel generator for the security computer. The Acceptance Criteria (section 9) do not address safety injection as a test consideration, but step 6.22 states, "If safety injection (S.I.) occurs, refer to 19000-1 for S.I.
termination. Otherwise mark this step N/A." The step is initialed and dated. According to Data Sheet 7.1, pressurizer pressure dropped from 2230 to 2080 and then stabilized at 2240 psig, and pressurizer level dropped from 23 to 22 and stabilized at 24 %.
Mainsteam header pressure rose from 1030 to 1040 and then stabilized at 400 psig.
Steam generator pressures averaged 1020 prior to the trip, quickly dropped an average of 12 psig and stabilized at an average of 888 psig. The typical steam generator level dropped from 48 to 31 %
and stabilized at 41%. None of the preceding parameter changes would initiate S.I.
Discussions with plant personnel and inspectors witnessing the test confirmed that no S.I. signal was received. The licensee was made aware of this minor test discrepancy, and started the paper work necessary to correct it.
e.
The following continuing tests were scanned to confirm they were up-to-date for the current status of the test program:
1)
1-600-06, Dynamic Response Test, 2)
1-600-13, Power Ascension Test Sequence, 3)
1-6AE-02, Calibration of Steam and Feedwater Flow Instrumenta-tion at Power, (4)
1-6AE-03, Feedwater Test, and (5)
1-6RJ-01, At-Power Intercomparison of Reactor Protection System Input and Plant Computer Output No violations or deviations were identified.
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