ML20246M290
| ML20246M290 | |
| Person / Time | |
|---|---|
| Site: | Beaver Valley |
| Issue date: | 05/04/1989 |
| From: | DUQUESNE LIGHT CO. |
| To: | |
| Shared Package | |
| ML20246M242 | List: |
| References | |
| NUDOCS 8905190081 | |
| Download: ML20246M290 (6) | |
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BEAVER VALLEY I POWER STATION UNIT 2 l
STARTUP REPORT 4 SUPPLEMENT 5 ,
DUQUESNE LIGHT COMPANY OHIO EDISON COMPANY THE CLEVELAND ELECTRIC ILLUMINATING COMPANY THE TOLEDO EDISON COMPANY lEP2%888A8s88,8j32
TABLE OF CONTENTS P_ age i TABLE OF CONTENTS i
1.0 INTRODUCTION
1 2.0 DEFERRED POST-COMMERCIAL OPERATIONS TEST PROCEDURES 1 2.1 Fuel Handling Equipment (P0-2.66.01) 1 2.2 Cranes and Lifting Equipment (P0-2.66.02) 2 2.3 Automatic Steam Generator Water Level Control Test 3 (SOV-2,240.01) 3.0 POST-COMMERCIAL OPERATIONS CHRON0 LOG 4 1
1.0 INTRODUCTION
The fifth supplement to the Beaver Valley Power Station Unit 2 (BVPS-2)
Startup Report covers the period of testing deferred after Commercial Operations between November 11, 1988 and February 9, 1969. This supplement report is prepared in accordance with Regulatory Guide 1.16, " Reporting of Operating Information - Appendix A, Technical Specifications", and addresses the results of deferred testing identified in the BVPS-2 Final Safety Analysis- Report during this period. Those tests which were deferred, but not completed by February 9, 1989, will also be identified in this report, but a comprehensive discussion on these tests will be provided when they are completed. Another supplement report will cover those tests completed subsequent to February 9, 1989.
The Unit ran at approximately 100% power for 88 days since returning to full power on October 12, 1988 following a load reduction to 75% power for '"B" Condenser waterbox maintenance. On January 7, 1989 at 2343 hours0.0271 days <br />0.651 hours <br />0.00387 weeks <br />8.915115e-4 months <br />, a reduction from 97% to 30% reactor power was commenced to allow maintenance on the main feedwater regulating valves. Reactor power was returned to approximately 100% power on January 10, 1989. On January 27, 1989, Unit 2 began a planned power reduction to 92% power due to the end of cycle schedule. The Unit continued to operate at approximately 92% power operation through the remainder of this report's period. Section 3.0 of this supplement report contains a Post-Commercial Operations Chronolog for additional information.
2.0 DEFERRED POST-COMMERCIAL OPERATIONS TEST PROCEDURES The test procedures found in this section could not be completed prior to Commercial Operations, and were deferred as concurred with by the Onsite !
Safety Committee and formally identified to the NRC. This section will
. discuss test results for those test procedures completed after Commercial Operations between November 11, 1988 and February 9, 1989. Another supplement report will be issued covering the period of February 9, 1989. to May 10, 1989, 2.1 Fuel Handling Equipment Test (P0-2.66.01)
This test verified control logic and demonstrated operability of the Fuel Handling Equipment and provided a functional demonstration of a simulated fuel transfer between the Fuel Building and Reactor l satisfactorily l Containment. Fuel Handling Equipment testing was l completed to support core load although portions of the test were not done due t3 unavailchility of equipment. Thore portions of the test not available for testing prior to core load were cortpleted prior to Commercial Operations except for testEng of the Rod Cluster Control (RCC) Change Fixture (Retest 06-005). A discussion on that portion of the test completed prior to Commerciai Operations can be found. In Section 8.4 ef the original Startup Report. Testing of the RCC Change l- Fixture (Retest 06-005) was deferred with a required completion date of prior to the first refueling. Completion of this retest cannot be done until during the first refueling outage when Westinghouse performs their tool checkouts prior to moving fuel. The RCC Change Fixture was not used during initial fuel load because there was no irradiated fuel present. Completion of this test has been deferred until prior to the end of the first refueling outage.
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'The period of time covered by this supplement report did not include the first refueling outage, and testing of the RCC Change Fixture had
- not- yet. been started. The results of Retest 06-005 will be discussed in another supplement report subsequent to completion of the retest.
2.2 Cranes and Lifting Equipment Test (PO-2.66.02) i Control. logic testing for the Reactor Containment Polar Crane and operability of .its auxiliary hoist under load was successfully completed prior to core load. Polar Crane main hoist operability under load was checked during its load test and satisfactorily completed during reactor vessel head installation. Retest 66-007 was generated to test an indicating light on the telescoping work platform which was inoperative during control logic testing. Outstanding rework on the Spent Fuel Cask Crane was not completed prior to core loat because it was thought to be safer to do it without fuel in the Fuel Storage area.
Retest 66-007 and the remainder of the test which tests the Spent Fuel Cask Crane' were deferred with a required completion date of prior to the first refueling.
Testing of the Spent Fuel Cask Crane was completed on December 20, 1988. The control logics of the crane performed satisfactorily meeting all applicable acceptance criteria. Operability of the Main Hoist with a 100 ton load (its heaviest normal load) was satisfactorily demonstrated. The Main Hoist maximum normal lowering speed with a 100 ton load suspended was slightly high out of the acceptance criteria range. The Main Trolley normal mode maximum forward and reverse speeds unloaded were slightly high out of the acceptance criteria range. Both out of range conditions were evaluated by Engineering and determined to be acceptable. Operability of the Auxiliary Hoist at 100% rated load (30 tons) was satisfactorily demonstrated. The Auxiliary Hoist maximum normal lowering speed with a 30 ton load suspended was high out of the acceptance criteria range. The Auxiliary Trolley fast speed reverse and slow speed forward and reverse with the rated load (30'. tons) suspended were all slightly high out of the acceptance criteria range.
Both out of range conditions were evaluated by Engineering and determined to be acceptable.~. In addition, the Auxiliary Trolley slow speed motor thermal overloads had tripping problems that were troubleshoot by Maintenance. Investigation revealed that both fast and slow speed amperages were normal, but the thermal overloads were significantly undersized for the slow speed winding and slightly undersized for the- fast speed winding. It is anticipated that Engineering will approve of thermal overload size increases for both speeds. Proper installation and performance testing of the anticipated thermal overload replacements will be handled by the Post Maintenance Testing Program.
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I Retest 66-007 (Verification of the Reactor Containment Crano Telescoping Work Platform (TWP) Elevation 866'-6" indicating light) will be handled by the Post Maintenance Testing Program. This light failed to illuminate as required during the final stages of preoperational testing prior to initial core loading. The light bulb was checked to ensure that it was not burned out. The purpose of this indicating light is to indicate when the TWP is positioned to permit full extension for gaining access to the inside central Reactor Containment dome area and the upper spray header. Vendor drawings were 2-L . . __ . . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __________________________l
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reviewed to determine the potential causes and consequences of a failed
- .Elev. 866'-6" indicating light.. One of two contacts' in position ~ switch
. LS6 controls the indicating light. The other contact automatically stops the TWP- if in full extension just before'it reaches the upper spray header area to prevent impact. The TVP must be lowered at -this point to Elev. 847'-9" or less before proceeding further forward. The automatic stop mentioned above was verified' to function properly indicating- that position switch LS6'is operable. The problem with the-indicating light is probably due to a bad light socket, contact or wire connection.. ln any case, it is not affecting the safety interlocks.of the TWP and is considered a minor inconvenience. Manpower and Reactor Containment atmospheric . conditions have failed to provide an opportunity to correct this problem prior to the first refueling.
Troubleshooting and repairs have been scheduled.for the first refueling outage and should adequately correct and verify proper operation'of the indicating light under post maintenance testing. All required testing has been considered satisfactorily completed.
2.3 Automatic Steam Cenerator Water Level Control Test (S0V-2.24C.01)
This test was performed at various modes and power levels during the Startup Testing Program. A detailed discussion on that portion of the test completed prior to Commercial Operations can be found in Section.
7.3.3 of the original Startup' Report.
During StartupL at the 30, 50, 75, and 100% power plateaus, each steam and feedwater flow transmitter output signal was compared to the
. associated calorimetric flow computed from test gauges in accordance with IST-2.02.06, " Thermal Power Calorimetric Test". At each power level, the feedwater and steam flow rates which were calculated did not agree with.the calorimetric flow rate criteria. A copy of the test data was forwarded to . Instrumentation & Control (I&C) to make the necessary adjustments to both the feedwater and steam. flow 1 transmitters. These adjustments and the remainder of the' test were not completed prior to Commercial Operations, and were deferred with a required completion date of the first refueling. The last section in this test obtained additional flow data at 100% power and determined the need for further adjustments to demonstrate. repeatability of the feedwater and steam flow transmitters at 30, 50, 75, and 100% power during a subsequent plant startup. I&C could not make the necessary adjustments to both the feedwater and steam flow transmitters prior to the first refueling. The tast was deferred so that I&C could make the ner.essary adjustments during the first refueling outage. The remainder of this test which will demonstrate repeatability of the feedwater and steam flow transmitters will be performed during startup following the first refueling outage. This test is expected to be completed by August 31, 1989 following the final 100% power verification of repeatability.
Note that I&C had determined that there was no problem with a steam flow / feed flow mismatch and recommended that plant operation continue until final adjustments and checks can be performed. Test results will be discussed in another supplement report upon completion of the remainder of testing following the first refueling outage.
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