ML20235N089

Startup Rept Suppl 4 for Beaver Valley Power Station, Unit 2

ML20235N089
Person / Time
Site:	Beaver Valley
Issue date:	02/03/1989
From:	Sieber J DUQUESNE LIGHT CO.
To:	NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM)
References
NUDOCS 8903010028
Download: ML20235N089 (18)

Text

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Shippingport, PA 150776 JOHN D SfEBE A Vete Prescent . Nuclear Group M12) 643 5255 February 3, 1989 i

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 4 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 3, 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 August 13, 1988 to November 11, 1988.

Completion of the startup test program has not been fully accomplished due to some deferred testing. A supplemental report will be issued within 90 days to update the progress of the startup test program.

Very truly yours, k -

J. D. Sieber Vice President Muclear Group Attachment cc: Mr. J. Beall, Sr, Resident Inspector Mr. W. T. Russell, NRC Region I Administrator Mr. P. Tam, Project Manager g]}& b

' I 8903010028 890203 PDR ADOCK 03000412 P PDC '

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'i COMMONWEALTH OF PENNSYLVANIA)

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COUNTY OF BEAVER )

On this //7 , day of I , 1 ,

before me, m /fA/f),NdthlL&, a Notary Pub h in and for said-Commonwealth a nd 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 knowledge, information and belief.

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Neta ia! Seal

$beda M.Fattore. Notary Pubic

$h'ppingport Boro, Beaver County My Commrssion Ex;nres Oct 23,1989

, Member, Pennsylvania AncoatonolNotanes

a e BEAVER VALLEY POWER STATION UNIT 2 STARTUP REPORT SUPPLEMENT 4 DUQUESNE LIGHT COMPANY OHIO EDISON COMPANY THE CLEVELAND ELECTRIC ILLUMINATING COMPANY THE TOLEDO EDISON COMPANY

1 TABLE OF CONTENTS Page i TABLE OF-CONTENTS i 11 LIST OF ABBREVIATIONS 11

1.0 INTRODUCTION

1 2.0 DEFERRED POST-COMMERCIAL OPERATI'.!S TEST PROCEDURES 1 2.'1 Verification of Plant Performance Following Plant 2 Load Rejection / Trip From Power (IST-2.04.06) 2.2 Verification of Reactor Plant Setpoints 4 (PO-2.01A.03) 1 2.3 Fuel Pool Cooling System Test (PO-2.20.01) 4 2.4 Spent Fuel Pool and Refueling Cavity Leak Test 5 (PO-2.20.02) 2.5 Fuel Handling Equipment (PO-2.66.01) 5 2.6 Cranes and Lifting Equipment (PO-2.66.02) 6 2.7 Verification of Performance Calculations 6 (S0V-2.05A.03) 2.8 Automatic Stear Generator Water Level Control Test 7 (SOV-2.24C.01) 2.9 Turbine Overspeed Trip Test 8 (S0V-2.26B.02) 2.10 Final Discussion on RCS Flow and AT Data 8 Obtained Af ter Commercial Operations 3.0 POST-COMMERCIAL OPERATIONS CHRON0 LOG 12

' LIST OF ABBREVIATIONS s

BOP Balance of Plant BVPS Beaver Valley Power Station CREBAPS Control Room Emergency Bottle Air Pressurization System EFPH Effective Full Power Hours ESF- Engineered Safety Features F Degrees Fahrenheit GPM Gallons Per Minute HP High Pressure I6C Instrumentation and Control JC0 Justification for Continued Operation KWH Kilowatt-hour OST Operations Surveillance Test RCS Reactor Coolant System RTD Resistance Temperature Detector Tavg Average Reactor Coolant Temperature

.T/G Turbine Generator Th Reactor Coolant Hot Leg Temperature AT Change in Temperature UFSAR Updated Final Safety Analysis Report l

1.0 INTRODUCTION

The fourth supplement to the Beaver Valley Power Station Unit 2 (BVPS-2)

S'artup t Report covers the period of testing deferred after Commercial l Operations between August 13, 1988 and November 11, 1988. This supplement  !

report is prepared in acccrdance 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 November 11, 1988, 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 November 11, 1988.

The Unit ran at approximately 100% power for 21 days since the last reactor trip on July 27, 1988. On August 18, 1988, reactor power was reduced to 30% power while awaiting confirmation that an oil sample taken from the main trans former indicated imminent failure. Backup oil samples were taken for analysis to Westinghouse Lab and determined to be good.

Reactor power was returned to approximately 100% at 2115 hours0.0245 days <br />0.588 hours <br />0.0035 weeks <br />8.047575e-4 months <br />. At 1902 hours0.022 days <br />0.528 hours <br />0.00314 weeks <br />7.23711e-4 months <br /> on August 22, 1988, the Unit 2 control room received a chlorine alarm on one detector. At 1913 hours0.0221 days <br />0.531 hours <br />0.00316 weeks <br />7.278965e-4 months <br />, the Unit 1 control room received 2/3 chlorine alarms and control room isolation commenced. Both control rooms were checked for chlorine and no indication was noted. The control room air bottles were manually isolated and the plant prepared for shutdown in accordance with Technical Specifications 3.0.3 and 3.7.7. An Unusual Event was declared at 1955 hours0.0226 days <br />0.543 hours <br />0.00323 weeks <br />7.438775e-4 months <br /> and later terminated at 2040 hours0.0236 days <br />0.567 hours <br />0.00337 weeks <br />7.7622e-4 months <br />. Unit 2 entered Mode 3 while repressurization of the CREBAPS air bottles continued. The Unit continued to cooldown to Mode 4 while eleven (11) containment isolation valves were being verified operable. Following verification, the Unit returned to approximately 100% power on August 26, 1988. The reactor tripped at 0306 hours0.00354 days <br />0.085 hours <br />5.059524e-4 weeks <br />1.16433e-4 months <br /> on September 20, 1988 due to a spike received on N41 (A Loop) resulting in a 2/3 overpower / delta temperature reactor trip. This occurred during the performance of OST 2.02.01, " Nuclear Power Range Channel Functional Test", while testing N43. The Unit returned to approximately 100% power on September 22, 1988.

From September 30 to October 2, 1988 and from October 7 to 12, 1988, the Unit reduced power to 75*, to perform maintenance on the "B" condenser waterbox. The Unit returned to approximately 100% power on October 12, 1988 and continued at full 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. All remaining test procedures which could not be completed in the time frame of this supplement report have a scheduled completion date of the first refueling.

This section will discuss test results for those test procedures completed after Commercial Operations between August 13, 1988 and November 11, 1988.

Another supplement report will be issued covering the period of November 11, 1988 tci February 9, 1989.

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2.1 Verification of Plant Performance Following Plant Load Rejection / Trip From Power (IST-2.04.06)

Beaver Valley Power Station, Unit 2, was designed with the capability of sustaining a 100% full load rejection wicnout a turbine trip or a reactor trip. This test performed a full load rejection from approximately 100% power prior to Commercial Operations, but did not satisfy all acceptance criteria. A detailed discussion about the l full load rejection can be found in Section 7.2.5 of the original Startup Report.

The other portion of this test was to perform a turbine trip from 100% power. Data obtained from an inadvertent turbine trip from approximately 100% power prior to Commercial Operations was evaluated and found to be acceptable to satisfy the performance of the turbine trip portion of this test. A discussion about the turbine trip can be found in Section 7.2.5 of the original Startup Report.

As discussed in Section 7.2.5 of the original Startup Report, data collected during the inadvertent turbine trip and during the unsuccessful full load rejection turbine trips, could not be used for the Reactor Coolant Hot Leg Resistance Temperature Detector (RTD) time response evaluation. Retest 04-002 was generated to'obtain the required RTD response time during the next inadvertent reactor trip from. 100% power. Since the plant was declared Commercial on the day this retest was initiated, the retest and IST-2.04.06 were both deferred until prior to the first refueling. I&C was directed to connect a memory type strip chart recorder to the Th RTD outputs at the primary process racks to gather the required data for Westinghouse during the next reactor trip from 100% power. The recorder would automatically start upon a reactor trip, and Westinghouse would use this information to determine plant RTD time response. An inadvertent reactor trip occurred on November 17, 1987, j subsequent to Commercial Operutions, but prior to installing che strip chart recorder. Consequently, the RTD response time data was not obtained during this trip.

The Startup Report, Supplement 1, Section 2.3 discusses the results of data obtained during an inadvertent reactor trip from approximately 96% power on January 27, 1988. This data was evaluated by Westinghouse and determined not usabic for the purpose of RTD response time evaluation. On April 4, 1988, an inadvertent reactor trip from approximately 100% power occurred due to a low RCS flow in the "A" loop. Since the analysis assumptions for the January 27, 1988 trip were the same for this trip, the data was determined not usable for the purpose of RTD response time evaluation. The analysis assumptions for development of RTD response time criteria is that the full reactor coolant pump flows are maintained in all loops for the duration of the required RTD response time (7-10 seconds). The flow increase or decrease in any one loop can thus affect the RTD bypass flow and can invalidate the acceptance criteria.

On July 27, 1988 a reactor trip occurred that resulted in the collection of plant data that could be used in the evaluation of overall RTD response times. The data showed that overall RTD response times exceeded the startup test acceptance criteria by

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approximately 0.25 seconds. Westinghouse performed an initial review l

of the plant trip data and concluded that, based on experience, the re,sponse times may be acceptable since the methodology used to develop the test acceptance criteria typically includes several tenths of a second margin. On several plants, Westinghouse has been abic to recalculate the acceptance criteria based on the actual plant configuration at the time of the test, thereby removing margin from the startup test acceptance criteria. One of the most significant sources of margin is the assumed fuel heat transfer which is a function of fuel burnup. In general, the more burnup, the greater the fuel heat transfer and the lower the response time' acceptance criteria. The analysis per formed to determine the startup test acceptance criteria assumes that the fuel burnup at the time of the test is 3000 EFPH. Since fuel heat transfer increases from initial criticality to 3000 EFPH, this is a conservative assumption for the startup test and provides a window for which the acceptance criteria is applicable.

Westinghouse subsequently requested additional data from the subject reactor trip, including fuel burnup, and performed a more in depth evaluation of the Beaver Valley Unit 2 data and the plant  ;

configuration at the time of trip. This evaluation showed that the j source of margin (i.e., fuel heat transfer) that is usually available may not be available for Beaver Valley since the reactor trip occurred .at 6,083 EFPH. Consequently, additional evaluation and calculation of the overall RTD response time acceptance criteria was being performed by Westinghouse. In the interim (during August of 1988), the I&C Group determined the final full power AT values and upon rescaling the AT summing amplifiers, turbine runback signals were experienced due to the reduced setpoints and the signal spikes that were being amplified from the lead / lag cards. In order to prevent the turbine runback condition, it was necessary to reduce plant power to approximately 98% until the RTD response time data could be evaluated and justification obtained from Westinghouse to I

increase the overpower AT turbine runback and reactor trip setpoints by 1.5%. On September 2, 1988, Westinghouse issued a Justification for Continued Operation (JCO) to allow the plant to operate with the overpower AT setpoints 1.5% below the Technical Specification limit of 107.81%. Following Engineering concurrence with the Westinghouse JCO, the overpower AT setpoints were changed. This corrected the turbine runback problem and the plant was returned to 100% power.

Westinghouse also said that overtemperature AT reactor trip setpoints should remain conservatively set at 3% below Technical Specification values in accordance with the BVPS Unit 2 Westinghouse Precautions, Limitations, and Setpoints Manual since this protection system is taken credit for in the mitigation of a boron dilution transient.

On September 20, 1988, another inadvertent plant trip occurred and additional overall RTD response time data was obtained. This data was also forwarded to Westinghouse for inclusion in their ongoing evaluation.

Westinghouse evaluation of the RTD response time data resulted in recommendations regarding UFSAR and Technical Specification changes.

Westinghouse concluded that the in-situ overall RTD response time is 5.5 seconds in lieu of the design basis allocation of 4 seconds. The 4

additional 1.5 seconds delay has been attributed to the RTD time constant, thus increasing the RTD time constant allocation from 2

, senonds to 3.5 seconds. The resultant in-situ overpower /overtemperature AT reactor trip response time is 7.5 seconds with an associated 5.5 seconds overpower /overtemperature AT channel response time. Based on this, Westinghouse concluded that the results from the July 27 and September 20, 1988 reactor trips were acceptable.

2.2 Verification of Reactor Plant Setpoints (P0-2.01A.03)

I The initial reactor plant setpoints were verified by using the latest plant documentation and/or actual instrumentation module settings ensuring the- equipment was aligned for the initial startup setpoint values. All setpoints fell within the required Technical Specification limitations. Final setpoint verification was started with the plant at 100% power, but could not be completed prior to Commercial Operations because several systems did not have the final setpoint values in place. The remainder of this test was deferred with a required completion date of the first refueling.

The second portion of this test was completed on October 12, 1988 and successfully verified that the final reactor plant setpoints had been  :

incorporated into the reactor plant equipment by confirming that the required source documents had been implemented and by recording the values contained in those source documents into this test. The values contained in Section VII.B of this test are considered final for the purposes of this test. However, they are subject to change due to the necessity for further adjustments at some future time.

This test has been considered completed, although final steam /feedwater flow mismatch data collection and adjustments to instrumentation were .not completed by this test. The final verification of these setpoints will be completed in SOV-2.240.01,

" Automatic Steam Generator Water Level Controls Test".

2.3 Fuel Pool Cooling System Test (PO-2.20.01)

This test was satisfactorily completed prior to core load with the exception of Retest 20-003 which will measure Fuel Pool Cooling Pump 21A vibration the next time the Spent Fuel Pool is filled.

Engineering resolution for high pump vibration during original testing modified the motor support stand by welding on stiffening plates. The pump motor was retested yielding acceptable vibration levels. The pump and motor were then recoupled, but the Spent Fuel Pool had already been drained. Refilling of the Spent Fuel Pool was not planned until prior to the first refueling so the test was deferred with a required completion date of prior to the first refueling.

Reteet 20-003 was successfully completed on October 18, 1988 following filling of the Spent Fuel Pool. The "A" Fuel Pool Cooling l Pump was operated in the normal configuration through the Fuel Pool 11 eat Exchanger and recirculated back to the Spent Fuel Pool.

Following bearing temperature stabilization a complete set of vibration data was obtained. All vibration data met the acceptance criteria limits signifying satisfactory pump operation.

2.4 Spent Fuel Pool and Refueling Cavity Leak Test (PO-2.20.02)

Th,is test was satisfactorily completed prior to core load with the exception of Retests20-001 and 20-002. Retest 20-001 was generated wnen the gates between the Fuel Pool and Cask Area and between the l Fuel Pool and Fuel Transfer Canal exhibited considerable leakage. at I the bottom seals. Repairs were made to both gates but the retest l could not be performed until the next time the Spent Fuel Pool was filled. Retest 20-002 was generated when the Spent Fuel Pool .

instrumentation test could not be performed due to unavailability of  :

the Spent Fuel _ Pool 1cyc1 transmitters. This retest required the Spent Fuel Pool to be filled to its high level. Refilling of the  ;

Spent Fuel Pool was not planned until prior to the first refueling so both retests were deferred with a required completion date of prior to the first refueling.

Retests20-001 and 20-002 were successfully performed following filling of'the Spent Full Pool on October 17, 1988. With the Spent Fuel Pool filled to the high level alarm condition, the wier gate seals between the Spent Fuel Pool and the Fuel Transfer Canal, and between the Spent Fuel Pool and the Cask Area were inspected for leakage every ten (10) minutes for one hour. The test acceptance criteria was met, as neither gate exhibited any leakage. On October 17, 1988, the Spent Fuel Pool temperature alarms were successfully verified using a decade box to simulate pool temperature. The high temperature alarm and reset values for both Trains A and B were measured. On November 2, 1988, the Spent Fuel Pool high and low level alarm and reset values were measured for both Trains A and B by raising and lowering the level in the Spent Fuel Pool. All temperature and Icvel data measured met the test acceptance criteria.

2.5 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 RCC Change Fixture (Retest 06-005). 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 RCC Change Fixture (Retest 06-005) was deferred with a required completion date of prior to the first refueling.

At the time of this supplement report, 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.

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2.6 ' Cranes and Lifting Equipment Test (PO-2.66.02)

, 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 l

generated to test an indicating light on the telescoping work l platform which was inoperative when.in the maximum down position.

Outstanding rework on the Spent Fuel Cask Crane was not completed l prior to core load 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 Fuci Cask Crane, were deferred with a required completion date of prior to the first refueling.

At the time of this supplement report, electrical maintenance repair of the indicating light on the Polar Crane telescoping work platform had not yet been performed. Testing of the Spent Fuel Cask Crane was completed with some minor deficiencies which must still be resolved.

The results of Retest 06-007 and Spent Fuel Cask Crane testing will be discussed in another supplement report subsequent to completion of testing and clearing of test deficiencies.

2.7 Verification of Performance Calculations (SOV-2.05A.03)

This- test was performed at 30, 50, 75 and 100% power plateaus during Startup, and verified that the computer generated performance calculations, Nuclear Steam Supply System (NSSS), and Balance of -l Plant (BOP) programs were accurate at these power levels.

Deficiencies at lower power levels were cleared at subsequent power 1cvels or Maintenance Work Requests were issued against deficiencies  !

to perform troubleshooting with any required retests to be performed by existing plant procedures.

This procedure only satisfied a portion of the BVPS-2 Final Safety I Analysis Report (FSAR) test objective that was required to be performed. This limited scope was justified by the fact that a comprehensive set of initial operating procedures for verifying the accuracy of the plant performance calculations was being developed e and performed by the Computer Integration Group. These tests perform thorough verifications of all the programs and calculations related to the plant performance applications of the process computer system.  ;

The tests are designed in accordance with accepted software validation and verification standards, using test data sets to prove the accuracy of the program output. The accuracy of the field sensors was verified by the Phase 1 testing program. The accuracy of the raw value conversion into engineering units was verified by SOV-2.05A.03 as mentioned above. The total of final test results after completion, will fully satisfy all FSAR test objectives. The Computer Integration Group continued testing per the Initial Operating Procedures. Deferral to beyond Commercial Operations was approved with the estimated completion date for this testing prior to initial criticality after the first refueling.

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4 The Computer Integration Group has satisfactorily completed all testing required to satisfy the UFSAR. The following functions were satisfactorily tested:

Not KWH Logging Boron Study Mode Heatup-Cooldown Rate Reactor. Protection System Monitoring Containment Leak Rate Test Boron / Xenon / Reactivity Calculations ESF Valve Actuation Monitoring Containment / Plant Systems Monitoring Rod Position Monitoring incore Thermocouple Monitoring Accumulator Monitoring Primary / Secondary Side Thermal Calculations  !

T/G Stator Winding Temperatures Axial Flux Difference BOP Performance Calculations 2.8 Automatic Steam Generator Water Level Control Test (SOV-2.24C.01)

This test was performed at various modes and power levels during the Startup Testing Program. A detailed discussion on that portion or the test completed prior to Commercial Operations 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 I&C to make the 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. The last section in this test will obtain additional flow data at 100% power to determine the need for further adjustments, and to demonstrate repeatability of the feedwater and steam flow transmitters at 30, 50, 75, and 100% power during a subsequent plant startup.

At the time of this report, I&C had not made the necessary adjustments to both the feedwater and steam flow transmitters. This still precludes the completion of the remainder of this test. 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. j l

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2.9 Turbine Overspeed Trip Test (S0V-2.26B.02)

The Turbine Overspeed Trip Test was performed after completion of low power physics testing and following initial turbine generator synchronization during startup. A complete discussion on the testing performed can be found in Section 7.5.3 of the original Startup Report. During this testing, the e'lectrical overspeed trip device failed to function. Maintenance discovered a bad electrical overspeed pickup and replaced it, but did not perform a calibration of the overspeed trip chassis hardware. During a subsequent plant start-up, the electrical overspeed device tripped the turbine when no ,

overspeed condition existed. The electrical overspeed trip was defeated pending a final solution to the problem associated with this device by Westinghouse and Maintenance.

It has been determined that an unknown source of " noise" has been interfering with the signal going to the electrical overspeed trip chassis so thic trip circuit has been permanently disabled. A second electrical overspeed trip device is still operational. Two electronic turbine overspeed trip devices were originally provided at .

Unit 2. One trip circuit included a speed pickup at the turning gear j and provided a signal to the overspeed trip chassis within the emergency trip cabinet. This trip circuit was to initiate a turbine trip at 111% turbine overspeed. The second trip circuit which is still operational includes a speed pickup at the turbine stub shaft and provides a signal to the emergency trip cabinet at 111.5% turbine  !

overspeed. Reference to the disabled 111% overspeed trip device will be deleted from the UFSAR pending approval of a Technical Specifica-tion Change. A mechanical overspeed trip device is also operational. l 2.10 Final Discussion ca RCS Flow and AT Data Obtained After Commercial Operations Final discussion / resolutions of RCS flow and AT problems referred to in test results for IST-2.06.04, " Reactor Coolant System Flow Measurement Test" and IST-2.02.05, " Alignment of Process Temperature Instruments" (Sections 4.6 and 7.4.3 of the original Startup Report, respectively) are presented as follows:

Both tests were performed at various power plateaus during startup and were dependent on calculated secondary thermal power and primary temperature instrumentation accuracy for obtaining proper results.

RCS flow was determined by dividing the secondary thermal power (less non-reactor heat additions) by the primary side change in enthalpy for each reactor coolant loop. Extrapolated 100 percent power primary AT obviously required thermal power and primary temperature instrumentation for proper determination. When final calculated RCS flow was determined at the (nominal) 100 percent power plateau in IST-2.06.04, the calculated value was 279,074 GPM. The value exceeded the test acceptance criteria minimum flow requirement of 274,800 GPM which included a 3.5 percent flow measurement uncertainty (based on BVPS-2 Technical Specification 3.2.5). However, when this calculated flow was compared to the Westinghouse best estimate prediction (listed in the BVPS-2 UFSAR) of 290,400 GPM, it was determined to be approximately 4 percent low. Similarly, when the extrapolated full power AT was determined in IST-2.02.05, the value

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. obtained was approximately 65F, This value was less than the Westinghouse design maximum AT of 67F based on design minimum RCS f1pw rate but exceeded the best estimate AT of 61F. Since both disparities could have been related to possible disparities in thermal power calorimetric, additional testing was performed after Commercial Operations to further evaluate any impact on RCS flow measurement and full power AT determination. i Additional data was obtained at approximately 100% power on December 9, 1987 through December 15, 1987, using more accurate measurement techniques. The data obtained determined feedwater flow venturi tapset differences, feedwater flow comparisons between condensate and feedwater flow venturi devices, AT and RCS flow rate. Westinghouse reviewed the data stating that it accurately represented the plant condition, but there was nothing in the data that could explain the apparent low RCS flow condition. The process spare, protection and control channel temperature readings agreed satisfactorily which indicated at that time that no problems existed with the calibration and alignment of the process temperature instrumentation. The data collected on the feedwater flow venturi tapsets and comparJson to the feedwater pump suction venturis indicated a possible measurement error of 1-1.5%, not enough to explain the RCS flow being approximately 4% low. Engineering suggested performing a visual inspection of the venturi internal surfaces. Further evaluations continued by Engineering and Westinghouse.

During the January / February 1988 outage, the discharge feedwater flow venturi taps were blown down using condensate pump pressure in an attempt to clean the sensing lines. On February 2, 1988, a boroscope inspection of the feedwater flow venturi taps was performed. All of the instrument tap connections, except the high pressure (HP) tap on tap set number 1 on loop A and one of the tap sets on loop B feedwater lines, were in good repair and free of obstruction. The #1 tap set HP connection on loop A contained a deposit of foreign material (either corrosion products or a solidified crud / sludge deposit). The HP tap set on loop B contained the same deposits but to a lesser degree (approximately 20% of the quantity of loop A and located in three separate areas).

The vendor (BIF Industries) indicated that deposits or corrosion products at the entrance to the sensing tube would cause erroneous readings. The simultaneous tap set flow measurements previously taken on December 9, 1987, indicated that the A loop had a mismatch of 1.5%, B loop had a mismatch of 0.67%, and C loop had a mismatch of 0%. The deposits on the HP sensing lines could be the cause of the lower flow readings experienced on the A and B feedwater lines.

During startup after the outage, the instrument taps were flushed with a one minute high pressure blowdown using feedwater pump pressure. On February 26, 1988, with the plant near 100% power, comparison of flows measured from each tap set on a given venturi were obtained with results similar to those obtained on December 9, 1987 (i.e., the A loop had a mismatch of 1.5%, B loop had a mismatch l of 0.86%, and C loop had a mismatch of 0.2%). Engineering was requested to evaluate / develop an acceptable method for removing the foreign material (crud) identified during the boroscope inspection of the A and B loop venturis. The response by Engineering stated that i

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I l the " crud" in the venturis would be rodded out during 1R to determine j if the crud is the reason for the feedwater flow discrepancy. In the interim, the flowmeter discrepancy would be offset by use of new "K" values.

New "K" values were implemented, and on April 19, 1988, a new set or i

data was obtained during a calorimetric to verify 100% power operation. The calorimetric data was verified to be accurate at 100%

power. RCS hot ar.d cold Icg temperature data was recorded by reading the RTD's in each loop. The RCS flow rate was calculated to be 282,641 GPM, still exceeding the BVPS-2 Technical Specification i requirement value of 2 274,800 GPM but short by 2.7% of the best

! estimate prediction. The AT now calculated out to be approximately 63F instead of 65F at 100% power; an improvement, but still exceeding the best estimate AT of 61F. I&C review of primary temperature instrumentation data indicated some disparities which could have affected both calculated RCS flow and AT.

Further troubleshooting of the primary temperature continued. Data to provide reliable correction factors for each primary loop RTD was finally obtained.

On July 27, 1988 with the plant at approximately 100% power, a final calorimetric was run while additional primary temperature instrumentation data was obtained for determining full power AT values. The new extrapolated full power AT values varied -between 63.5F to 64.6F. When the full power AT data was utilized to obtain RCS calculated flow, a flow rate of 280,527 GPM was obtained. These values for full power AT and RCS calculated flow are considered to be based on the most accurate and reliable calorimetric and primary temperature instrumentation data obtained.

On August 29, 1988, Westinghouse issued a report on " Evaluation of Low RCS Flow". It provided a new best estimate flow value for BVPS-2 of 95,450 GPM/ Loop or 286,350 GPM total RCS flow. Based on the RCS measured flow data provided to Westinghouse for their evaluation (92,350 GPM/ Loop or 277,050 GPM total derived from the December 9, 1987 - December 15, 1987 data), the flow was considered sufficient since it met Technical Specification requirements. If the more recent RCS calculated flow (280,527 GPM) from the July 27, 1988 data is compared to the new best estimate flow value, the comparison improves such that RCS calculated flow is only 2.0% low and well above the Technical Specification lower limit of 274,800 GPM.

Following rescaliag of protection and control AT summing amplifiers i with the new full power AT values obtained from the July 27, 1988 data, the plant began experiencing periodic overpower AT runback signals. In accordance with BVPS-2 Westinghouse Precautions, Limitations and Setpoints, the overpower and overtemperature AT trip and runback setpoints had been set conservatively 3% below Technical Specification values until RTD time response testing analysis was complete. This resulted in the Overpower AT runback points set at 102.81%. With the new full power AT summing amplifiers rescaled, the corresponding protection channels were operating close enough to the runback points such that signal spikes from the lead-lag cards would cause runbacks, especially when increases in Tavg would cause

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setpoints to automatically decrease, During a BVPS-2 reactor trip on July 27, 1988"(nct related to the full power AT data gathering

,af,orement i oned), RTD time response data was obtained. Analysis of this data indicated that RTD sensor time constants could be as large as 3 seconds when 2 seconds maximum was required. Westinghouse evaluated the safety ramifications of this disparity cnd provided technical justification for raising the Overpower AT trip points from 3.0% to 1.5% below the Technical Specification values.

Overtemperature AT setpoints remain 3% below Technical Specification values since it is taken credit for in the mitigation of a boron dilution accident. Overpower AT runback and trip setpoints were recalibrates for 1.5*. below Technical Specification values. BVPS-2 is no longer experiencing Overpower AT runbacks.

On September 20, 1988, another reactor trip occurred and additional RTD response time data was obtained. Based on evaluation by l

Westinghouse, revised RTD sensor time constants of 3.5 seconds were more appropriate, a 0.5 second increase over the previously determined maximum RTD sensor time constants of 3 seconds discussed above. For final discussions / resolutions of this matter, refer to Section 2.1 of this report for IST-2.04.06, " Verification of Plant Performance Following Plant Load Rejection / Trip from Power".

Investigation of RCS flow and AT problems is considered complete.

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