ML20245H782
| ML20245H782 | |
| Person / Time | |
|---|---|
| Site: | Beaver Valley |
| Issue date: | 08/04/1989 |
| From: | Sieber J DUQUESNE LIGHT CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NUDOCS 8908170277 | |
| Download: ML20245H782 (12) | |
Text
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9
'g- Braver Valley Power Station Shippingport. PA 15077 4 004 JOHN D SIEBER 44121 643-5502 Vice Presedent
- Nuclear Operations August 4, 1989 U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555
Reference:
Beaver Valley Power Station, Unit No. 2 Docket-No. 50-412, License No. .NPF-73 BVPS-2 Startup Report, Supplement 6 Gentlemen:
In accordance.with Regulatory Guide 1.16, " Reporting of Operating Information - Appendix A,. Technical Specifications", the attached Summary Report of plant startup testing is being submitted. This report is due 90 days following ~ submittal of the BVPS-2 Startup Report, Supplement 5, for Beaver Valley Power Station,. Unit No. 2.
This report addresses the results of all startup testing identified in the Beaver Valley Power Station Unit No. 2 Final Safety Analysis Report during the period from February 9, 1989 to June 9, 1989. This l
is the final Supplement Report due to completion of the startup test program.
l Very truly yours, !
~{
. D. Sieber ;
' cVice President Nuclear Group Attachment 1 cc: Mr. J. Beall, Sr. Resident Inspector Mr. W. T. Russell, NRC Region I Administrator Mr. P. Tam, Sr. Project Manager l
l 8908370277 890804 PDR P
ADDCM 050004i2 pyy. f676 I L
... . COMMONWEALTH OF PENNSYLVANIA)
) SS:
COUNTY OF BEAVER *)
On- this day of hM , 1989, before' me, D /// M, a Notary. blic in and for!said
/ i Commonwealth and County, personally appeared J. D. Sieber, who.being, duly sworn, deposed, and said that- (1) he is Vice President of Duquesne ' Light, (2) he is duly authorized to execute and file the foregoing Submittal.on behalf of said company, and (3) the statements set .forth. in the Submittal are true and correct to the best of his j knowledge, information and belief.
i l
U
/ Netana! Seal Sheila M. Fattore, Notary Public .
i Shippingport Boro, Beaver County
._ My Commscon Expres Oct.23,1989 Meneer, Pennspvenis hescaten of Notanes i
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._._____._______.___..._______._..._____.___-________-.j
- r a
BEAVER VALLEY POWER STATION UNIT 2 STARTUP REPORT SUPPLEMENT 6 l
DUQUESNE LIGHT COMPANY OHIO EDISON COMPANY THE CLEVELAND ELECTRIC ILLUMINATING COMPANY THE TOLEDO EDISON COMPANY
i
-TABLE OF CONTENTS P_agg i TABLE OF CONTENTS .i
1.0 INTRODUCTION
1 2.0 DEFERRED POST-COMMERCIAL OPERATIONS T' dST PROCEDURES 1 2.1 Fuel Handling Equipment Test (P0-2.66.01) 1 2.2 hutomatic Steam Generator Water Level Control Test 3 (SOV-2,24C.01) 3.0 POST-COMMERCIAL OPERATIONS CHRON0 LOG 6 I
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1.0 INTRODUCTION
The sixth and final supplement to the Beaver Valley Power Station Unit 2 (BVPS-2) Startup Report covers the period of testing deferred after Commercial Operations between February 9, 1989 and June 9, 1989. 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 Anal ~ysis Report during this period. This is the final Supplement l Report due to completion of all startup testing. Results of all remaining testing completed since the last Supplement Report will be discussed in this report.
l The Unit continued to operate at approximately 92% power due to the end of cycle power schedule. On February 12, 1989, power was reduced to approximately 65% due to trouble controlling the main feed regulating valves. A subsequent reactor trip / turbine trip occurred at 0120 hours0.00139 days <br />0.0333 hours <br />1.984127e-4 weeks <br />4.566e-5 months <br /> due to "C" steam generator hi-hi level caused by the main feed regulating valve problems. The Unit returned to approximately 58% power on February 19, 1989 following main feed regulating valve repairs. On March 17, 1989, the Unit commenced a load reduction in preparation for the Unit 2 first refueling outage. The Unit entered Mode 6 on March 26, 1989 and re-entered Mods 5 on April 27, 1989 following completion of refueling activities. Problems with the Reactor Vessel Level Indication System (RVLIS) and a resistance temperature detector (RTD) manifold delayed startup after the first refueling outage. The Unit achieved 100% power on June 9, 1989 and completed the last remaining test deferred during the Startup Test Program.
Section 3.0 of this supplement report contains a Post-Commercial Operations Chronolog for additional information.
2.0 DEFERRED POST-CHEMICAL 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 the remaining test procedures completed after Commercial Operations between February 9, 1989 and June 9, 1989.
2.1 Fuel Handling Equipment Test (PO-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 Containment. Fuel Handling Equipment testing was satisfactorily completed to support core load although portions of the test were not done due to unavailability of equipment. Those portions of the test not available for testing prior to core load were completed prior to Commercial Operations except for testing of the Rod Cluster Control Assembly (RCCA) Change Fixture (Retest 66-005). A discussion on that portion of the test completed prior to Commercial Operations can be found in Section 8.4 of the original Startup Report. Testing of the RCCA Change Fixture (Retest 66-005) was deferred with a required completion date of prior to the first refueling. Completion of this retest could not be done until during the first refueling outage when Westinghouse performed their tool checkouts prior to moving fuel. The RCCA Change Fixture was not used during initial fuel load because there
. was no irradiated fuel present. Completion of this test was deferred ;
until prior to the e:nd of the first refueling outage.
Operability and control logic testing of the Rod Cluster Control Assembly (RCCA) Change Fixture was verified on March 26, 1989 during the first refueling outage. The RCCA Change Fixture was operated by the Westinghouse Refueling Crew in accordance with the Beaver Valley Unit 2 Refueling Cycle I-II Checkout Procedure in lieu of performing retest'66-005 per P0-2.66.01. The Westinghouse checkout procedure performed basically the same equipment movements and verifications to satisfy pre-operational testing requirements.
The following equipment operability problems occurred during testing.
Tha RCCA Chenge Fixturg carriage assembly and guide tube assembly were misaligned such that exchanging a RCCA (commonly referred to as a control rod) between the guide tube assembly and carriage assembly could not be properly performed. A RCCA could be removed and reinserted into the dummy fuel assembly located in the carriage ,
assembly west compartment (i.e., first compartment) or into the RCCA 1 storage compartment (i.e., second compartment), but the re-insertion i did not occur smoothly due to a misalignment of approximately 3/16 inch in these areas. However, in the carriage assembly east compartment (i.e., third compartment), the misalignment was slightly worse 1 (approximately 1/4 inch) and enough to prevent reinsertion of the RCCA )
in the dummy fuel assembly. A maintenance work request was issued to j correct the misalignment problem and will be completed in accordance
'l with Station maintenance procedures. The gripper hoist bottom limit failed to automatically stop the RCCA Change Fixture winch when the gripper reached the bottom of its travel. A maintenance work request .
was issued to adjust the limit switch to ensure proper performance . j However, this adjustment could not be properly performed before i completion of the first refueling. Proper adjustment requires hoisting-the dummy fuel assembly into the RCCA Change Fixture guide tube in order to apply the appropriate load on its rigging. Usage of the dummy fuel assembly in the RCCA Change Fixture guide was not advisable due to the misalignment problem which could have caused the dummy fuel assembly to become hung up between the guide tube and carriage assembly. Adjustment of the gripper hoist bottom limit will be completed in accordance with Station maintenance procedures following resolution of the misalignment problem. i The RCCA Change Fixture was not required for any fuel movements during the first refueling outage nor is it anticipated for use during the second refueling outage. The station has a RCCA Change Tool located in the Fuel Building which is used in place of the RCCA Change Fixture for RCCA changes during complete core offloads (and during shuffles, if required). Due to the Westinghouse 17 x 17 fuel design which will be in use until Cycle 4, complete core offloads must be performed. The complete core offload avoids fuel assembly grid strap interaction which frequently occurs for this fuel design. Lack of need for the RCCA Change Fixture because of similar equipment located in the Fuel Building; high radiation levels (3 to 40 REM / hour) which eventually i
developed in the area where repair work would occur; and the need to avoid impacting the first refueling outage schedule combined to make resolution of the misalignment problem impractical until the third l
refueling outage. Responsibility for ensuring resolution of the 1
c
, . misalignment problem as well as adjustment of the gripper hoist bottom limit prior. to completion of the third refueling outage has been l
transferred to. the.' Refueling Group for completion 'per: Station maintenance ' procedures. Testing'.per PO-2.66.01 to ~ satisfy UFSAR Section 14.2.12.77.1 requirements is considered complete and further progress reporting in accordance with this report 'is no longer necessary.
2.2 Automatic Steam Generator Water Level Control Test.(SOV-2.240.01)
. I This test was performed at various modes and power levels during the i Startup-Testing Program. A detailed discussion on that portion of the .
test completed prior to Commercial Operation can be found in Section 7.3.3.of the original Startup Report.
During Startup 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 -- H necessary adjustments to both; the feedwater- and steam' flow 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. I&C had, determined )
that there was no problem with a' steam flow / feed flow mismatch- and recommended that plant operation continue until fina1' adjustments and checks could be performed. The last section in this test obtained additional flow data at approximately 100% power.'and determined the I
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 test was deferred with a required completion date of August 31, 1989, so that I&C could make the necessary adjustments during the first refueling outage. The remainder of this test which demonstrated repeatability of the feedwater and steam flow transmitters was performed during startup following the first refueling outage and is discussed below.
The last section of this procedure was performed initially at 99.5%
power on December 28, 1988. . It was repeated during the .startup j following the completion of the-first refueling outage at power levels of 30% (on 6/3/89), 50% (on 6/5/89), 75% ,(on 6/6/89), and 98.5% (on- :
6/9/89)~ . A calorimetric flow for each loop was' computed from data '
gathered using test gauges. Steam and feedwater flow transmitter output signal data was taken. simultaneously with the calorimetric flow j data'and the values compared to determine if any instrument adjustments were necessary to match the transmitter output to the calorimetric flow within the statistical tolerance of the permanently installed ' plant instruments.
From the initial 99.5% power run on December 28, 1988, the data .
revealed that in each protection channel, at least one transmitter output signal was out of tolerance and did not meet the test acceptance
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- i. criteria of +/-27,950 pph from the calorimetric flow- for feedwater flow, and +/-34,731 pph from the calorimetric flow for steam flow. The.
results are summarized below:
DIFFERENCE FLOW FROM PROTECTION CHANNEL TRANSMITTER CALORIMETRIC REMARKS Loop 1 Protection III- 2FWS*FT477- + 40.106 pph Out of tolerance 2 MSS *FT474 + 30,710 pph Within tolerance Loop-1 Protection IV 2FWS*FT476' -
8,387 pph Within tolerance 2 MSS *FT475 + 143,210 pph Out'of tolerance Loop 2 Protection III 2FWS*FT487 - 3,767 pph Within tolerance 2 MSS *FT484 + 100,601 pph Out of tolerance Loop 2 Protection IV 2FWS*FT486 -+ 30,332 pph Out of tolerance 2 MSS *FT485 + 105,601 pph Out of tolerance Loop 3 Protection III 2FWS*FT497. '
9,763 pph Within tolerance 2 MSS *FT494 + 156,743 pph Out of tolerance Loop 3 Protection IV 2FWS*FT496 + :16,971 pph Within tolerance 2 MSS *FT495 -
48,757 pph Out of tolerance All' data for this run was submitted to the Instrument & Control (I&C)
Department for use in making the required adjustments. 'During the first refueling outage, the feedwater . flow ' transmitters were -
recalibrates and the steam flow transmitters were rescaled and recalibrates. In addition, adjustments were made to the multiplier / divider card gain settings for both steam flow loops.-
Upon plant startup following the first refueling outage, the test was run again at 30, 50, 75, and 98.5% power. The results at 30, 50, and 75% power have been disregarded for purposes of instrument fine tuning for the following reasons:
(1) The instrumentation associated with the steam / feed flow protection loops is designed, scaled, and calibrated to achieve maximum resolution and accuracy at full power. Any adjustments made based on data from another power level could affect the loop performance at 100% power where the plant is intended to run.
(2) At 30 and 50% power the feedwater flow differential pressures (d/p) were below the range of the Tobar d/p . transmitters which were scaled for 100 to 500 inches of water equal to 1 to 5 volts dc. At 30 and 50% power most of the data did not meet the l acceptance criteria. At each power level of 75 and 98.5% power, three of - the transmitter . flows fell outside the acceptance criteria range.
The following is a summary of the results at 98'5% power which will be used for making final channel adjustments:
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1 i
'. DIFFERENCE FLOW FROM PROTECTION CHANNEL TRANSMITTER CALORIMETRIC. REMARKS l
Loop 1 Protection III 2FWS*FT477 + 7,501 pph Within tolerance.
2 MSS *FT474- + 21,483 pph Within tolerance Loop 1 Protection IV 2FWS*FT476 - 8,540 pph Within' tolerance 2 MSS *FT475 + 23,483 pph Within tolerance Loop 2 Protection III 2FWS*FT487 -
9,794 pph Within tolerance.
2 MSS *FT484 +. 17,686 pph W.tthin tolerance Loop 2 Protection IV 2FWS*FT486 -
~27,725 pph 'Within tolerance 2 MSS *FT485 + 75,186 pph Out of tolerance Loop 3 Protection III 2FWS*FT497 + 8,134 pph Within tolerance 2 MSS *FT494 + 74,594 pph Out'of tolerance.
Loop 3 Protection IV. 2FWS*FT496 + 17,849 pph 'Within tolerance 2 MSS *FT495 + 60,594 pph Out of tolerance-Maintenance work requests were issued to make the final . loop.
adjustments for [2 MSS *FT485, 494,' and 495), and will be completed per Station maintenance procedures.
In addition, .the feedwater control portions of PO-2.01A.03,
" Verification of Reactor Plant Setpoints" pertaining to steam / feed -flow mismatch were also finalized by this test.- All acceptance criteria for these setpoints (steam / feed flow mismatch reactor trip, steam / feed flow mismatch alarm, and steam generator level deviation) have been met. A discussion on the results of PO-2.01A.03 can be'found in Section 2.2 of the Startup Report, Supplement 4.
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