ML20065F842

Suppl 3 to Initial Startup Rept for June-Aug 1990. W/

ML20065F842
Person / Time
Site:	Seabrook
Issue date:	09/27/1990
From:	Feigenbaum T PUBLIC SERVICE CO. OF NEW HAMPSHIRE
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NYN-90178, NUDOCS 9010120024
Download: ML20065F842 (72)

Text

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'4 A New Hampshire .

Ya 1eal C. N oeninowm Senior Vice President and Chief Operating Officer NYN. 90178 September 2'. 1990 United States Nuclear Regulatory Commission Washington, D.C. 20555 Attention: Document Control Desk i s

References:

(a) Facility Operating License No. NPF 86, Docket No. 50 443 (b) NHY Letter NYN 90070, ' Initial Startup Report' dated March 13,1990, T. C. Felgenbaum to USNRC (c) NHY Letter NYN.90126, ' Supplement I to the Initial Startup Report' dated June 13, 1990, T. C. Felgenbaum to USNRC (d) NHY Letter NYN 90167, ' Supplement 2 to the Initial Startup Report' dated September 13, 1990, T. C. Felgenbaum to USNRC l

Subject:

Supplement 3 to the Initial Startup Report Genticmen:

In accordance with the requirements of Technical Specification 6.8.1.1 enclosed is Supplement 3 to the Initial Startup Report submitted via Reference (b) and supplemented via References (c) and (d). Supplement 3 to the Initial Startup Report covers the period from June 1990 through August 1990. The Power Ascension Test Program was completed on August 18, 1990.

Should you have any questions regarding 0,is report please contact _Mr.' James M.

l'eschel, Regulatory Compliance Manager at (603) 474 9521 extension ~3772.

Very ruly yours, f >

T Ad/ ($s#1 R g Ted C. Feigenba

$""O os a Enclosure Sh TCP:ALL/ssl do n.% , .

ny U ,., f *a l b

Re I5 (p

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&e New Hampshire Yankee Division of Public Service Company of New Hampshhe I\

P.O. Box 300 e Seabrook, NH 03874

• Telephone (603) 474 9521

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United States Nuclear Regulatory Commission September 27, 1990 Attention:- Document Control Desk Page two cc: Mr. Thor.as T. Martin Regional Administrator United States Nuclear Regulatory Commission

- Region L 475 Alleudale Road -

King of Prussim, PA 19406 Mr. Noel Dudley NRC Senior Resident inspector P.O. Box 1149 Seabrook, NH 03874 l

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NEW RAMPSHIRE YANKEE SEABROOK STATION SUPPLEMENT 3_

in IN171A U TARTUP REPORT 12 1h.t UNIIED STAffd NUCLEAR REGULATORY COMMISSION OPERATING LICENSE: NPF 86 NRC DOCKET NO. 50-443 For the Period June. 1990 throunh Aucust. 1990

et l TABLE OF CONTENTS f Seetion f. ggt List of' Figures lii l List of Tables v-t List of Acronyms vi l 1.0 Introduction 1-1 i

2.0 Startup Test Program Overview, Supplement 3 2-1  !

3.0 Seabrook Startup Chronology, Supplement 3 3-1 j 4.0 Summary of Initial Startup Report, Supplement 3 4-1 ,

5.0 Power Ascension Testing 5-1 5.1 ST-22, Natural Circulation Test 5-2 f 5.2 ST-24, Automatic Reactor control 5-7 5.3 ST-25,- Automatic Steam Generator Level Control. 5-8 ,

5.4 $7-29 Core Performance Evaluation 5-14 5.5 ST-30 Power Coefficient Heasurement 5-16  :

5.6 ST-33, Shutdown from Outside the Control Room 5-17 i 5.7 ST-34 Load Swing Test 5-19 5.8 ST-35, Large Load Reduction 5-29 .

5.9 ST-37 Hoisture Carryover Measurement 5-32 1 5.10 ST-30, Unit Trip'from 1001 Power 5-34 t 5.11 ST-39, Loss of Offsite Power Test 5-41  !

5.12 ST-40 NSSS Acceptance Test 5-47 -

5.13 ST-48 Turbine Generator Startup Test 5-49  !

6.0 Instrument Calibration and Alignment 6-1 1

6.1 ST-13, Operational Alignment of Nuclear Instrumentation 6-2 6.2 87-14.1, Operational Alignment of Process Temperature 6-9 Instrumentation 6.3 ST-15 Reactor Plant System Setpoint Verification 6-11 6.4 ST-26 Thermal Power Measurement and Statepoint Data 6-12 Collection

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6.5 ST-27, Startup Adjustments of Reactor Control System 6-14 6.6 ST-28, Calibration of Steam and Feedwater Flow 6-19 Instrumentation 6.7 ST-36, Axial Flux Difference Instrumentation Calibration 6 23 i

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. . TABLE OF CONTENTS (Continued) l i

Section M  !

7.0 General Plant Testing 7-1 j 7.1 .ST-41, Radiation Survey 7-2 ]

7.2 $7-42 Water Chemistry Control 7-3  ;

7.3 ST-43 Process Computer 7-6 7.4 ST-44 Loose Parts Monitoring 7-8

• 7.5 ST-45 Process Effluent Radiation Monitoring System 7-11 7.6 ST-46 Ventilation System Operability Test 7-13 7.7 ST-49 Circulating Water System Thermal-Hydraulic Test 7-15 7.8 ST-$1 Power Ascension Dynamic Vibration Test 7-16  !

7.9 ST-52 Thermal Expansion 7-17 7.10 ST-56 Piping Vibration Testing 7-20 a

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. - LIST OF FIGURES ]

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ST-13 Fig. 1 Total Detector Current (N41) vs Reactor Power 6-5 j Fig. 2 Total Detector Current (N42) vs Reactor Power 66 l i

Fig. 3 Total Detector Current (N43) vs Reactor Power 6-7 1 Fig. 4 Total Detector Current'(N44) vs Reactor Power 6-8

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ST-22 Fig. 1 SG2 (NR) Level vs Time 5 5~

i Fig. 2 L2 (WR) THOT and TCOLD vs Time 56 ST-25 Fig. 1 SG2 (NR) Level and Feed Pump Speed vs Time 5-11  !

Fig. 2 Steam Header Pressure and Feed Pump Discharge 5-12  ;

Pressure vs Time Fig. 3 Feed Pump A and Feed Pump B Speed vs Time 5-13 ,

ST-27 Fig. 1 RCS Temperature vs Power 6-16  ;

Fig. 2 SG Pressure vs Power 6-17 Fig. 3 Turbine Impulse. Pressure vs Power 6-18  :

ST-28 Fig. 1 Steam Flow / Feed Flow Mismatch (PT-510. FT-512) at 6-21 752 RTP vs Power Fig. 2 Steam Flow / Feed Flow Mismatch (PT-511. FT-513) at 6-22 '

75! RTP vs Power ST-34 Fig. 1 Pressuriner Pressure (Load Decrease at 1002) vs 5-21 Time Fig. 2 '

Pressurizer Pressure (Load Increase at 100Z) vs 5-22 Time Fig. 3 SG 1,2,3,4 Pressure (Load Decrease at 1002) vs 5-23 Time l Fig. 4 SG 1,2,3,4 Pressure (Load Increase at 100Z) vs 5-24 Time Fig. 5 TAyg L1,L2,L3,L4 (Load Decrease at 100I) vs Time 5-25 1

Fig. 6 TAyo L1.L2,LS.L4 (Load Increase at 1001) vs Time 5-26 [

Fig. 7 SG2 (NR) Level (Load Decrease at 1002) vs Time 5-27 'l Fig. 8 SG2 (NR) Level (Load Increase at 1002) vs Time 5-28 1

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LIST OF FIGURES (Continued)

E.AES.

ST-36 Fig. 2 Axial Flux Deviation at 752 vs Time 6-25 Fig. 2 N41 Incore AFD vs Excore AFD at 75% 6-26 Fig. 3 N41 Detector Currents vs Incore AFD at 752 6-27 Fig. 4- N42 Incore AFD vs Excore AFD at 752 6-28 Fig. 5 N42 Detector Currents vs Incore AFD at 752 6-29 Fig. 6 N43 Incore AFD vs Excore'AFD at 75% 6-30 Fig. 7 N43 Detector Currents vs Incore AFD at 752 6-31 Fig. 8 N44 Incore AFD vs Excore AFD at 752 6-32 Fig. 9 N44 Detector Currents vs Incore AFD at 751 6-33 ST-38 Fig. 1 TAyo L1.L2,L3,L4 vs Time i

5-36 Fig. 2 SG 1,2,3,4 Pressure vs Time  !

5-37 Fig. 3 Pressurizer Pressure vs Time 5-38 Fig. 4 Pressurizer Level vs Time 5-39 Fig. 5 SG 1,2,3,4 (WR) Level vs Time 5-40 ST-39 Fig. 1 Pressurizer Preseure vs Time 5 44 Fig. 2 Pressurizer Level vs Time 5-45 I i.

j Fig. 3 l

L4 (WR) THOT and TCOLD vs Time 5-46 l

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LIST OF TABLES I.1&f.

ST-14.1 Table 1, Out-of-Tolerance Conditions 6-10 ST-22 Table 1, Natural Circulation Subcooling Margin 5-4 ST-25 Table 1. Test Data, Steam Generator Automatic Level Control 5-10 ST-29 Table 1. Core Performance Parameters 5-15 ST-30 Table 1 Power Coefficient Measurement 5-16 ST-35 Table 1. Large Load Reduction 5-31 ST-37 Table 1. Moisture Cartwerar Results 5-33 ST-42 Table 1._ Process Instrument / Grab Sample Comparisons 7-4 Table 2, Out-Of-Specification Results, 1002 RTP 7-5 ST-44 Table 1. LPMS Alert Setpoints 7 ST-45 Table 1. Process and Effluent Radiation Monitoring Test 7-12 Exceptions ST-52 Table 1. Summary of Thermal Expansion Problems 7-19 i

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• .,* LIST OF ACRONYMS ACOT - Analog Channel Operational. Test l AFD - Axial Flux Difference  ;

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AO - Auxiliary Operator ARI - All Rods Inserted I ARO - All Rods Out l i

ASDV - Atmospheric Steam Dump Valve CB.,- Control Bank (A.B.C. or D) l CE_- Combustion Engineering CIV- Combined Intermediate Valve  !

CRD - Control Rod Drive t

CRDM - Control Rod Drive Mechanism CV - Control Valve l

CVCS - Chemical and Volume Control System DRPI - Digital Rod Position Indication EBOP - Emergency Bearing 011 Pump ECCS - Emergency Component. Cooling System EFPH - Effective Full Power Minute EFW - Emergency Feedwater EHC - Electrohydraulic Control ELOP - Emergency Lube 011 Pump ESOP - Emergency Seal Oil Pump FCFM - Full Core Flux Hap EPS - Emergency Power Sequencer FSAR - Final Safety Analysis Report FTC - Fuel Temperature Coefficient GETARS - General Electric Transient Analysis Recording System HFT - Hot Functional Test i

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o- 4 LIST OF ACRONYMS (Continued)

HSB - Hot Standby '

HX - Heat Exchanger  !

HZP - Hot Zero Power ,

IGC - Instrumentation and Control ICRR - Inverse Count Rate Ratio  ;

IV - Intercept Valve IR - Intermediate Range ITC - Isothermal Temperature coefficient  ;

LPMS - Loose Parts Monitoring System LPDS - Loose Part Detection System MCB - Main Control Board 'l 1

' MIDS - Movable Incore Detector System MSIV - Main Steam Isolation Valve MTC - Moderator Temperature Coefficient ,

MWE - Megawatts Electric MWT - Megawatts Thermal NDR - Nuclear Design Report NIS - Nuclear Instrumentation System i

NPDES - National Pollutant Discharge Elimination System NR - Narrow Range NRC - Nuclear Regulatory Commission NSSS - Nuclear Steam Supply System OTDeltaT - Over Temperature Delta-Temperature PATP - Power Ascension Test Program PCV - Pressure Control Valve PCCW - Primary Component Cooling Water l

PLS - Precautions, Limitations and Setpoints i i

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e 4 ., L IST OF ACRONYMS (Continued) 7 PLU - Power Load Unbalance PORV - Power Operated Relief Valve PR - Power Range RAT - Reserve Auxiliary Transformer RCCA - Rod Cluster Control Assembly RCS - Reactor Coolant System RDMS - Radiation Data Management System RHR - Reactor Heat Removal RM0 - Remote Manual Operation RSS - Remote Safe Shutdown RTD - Resistance Temperature Detector RTP - Rated Thermal Power RVLIS - Reactor Vessel. Level Indication System SB_ - Shutdown Bank (A.B.C.D and E)

SG_ - Steam Generator (A,B.C and D)

SGFP - Steam Generator Feed Pump SORC - Station Operating Review Committee SR - Source Range SSPS - Solid State Protective System SSCP - Seabrook Station Chemistry Program Manual TBWD - Thrust Bearing Wear Detector TEC - Technology for Energy Corp..

DAT - Unit Auxiliary Transformer UE&C - United Engineers and Constructors UPS - Uninterruptible Power Source URAL - Underexcited Reactive Ampere Limit WR - Vide Range vili

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.,1. 0 INTRODUCTION f The Initial Startup Report was submitted to the Nuclear Regulatory l' Commission in March,1990 and covered startup activities through completion of low power physics testing (June 1989). Supplement 1 reported testing which took place in the interval July 1989 through May 1990. Supplement 2 ,

was a summary document covering the tests reported in detail in Supplement

3. Supplement 3 covers the remainder of the Power Ascension Test Program  ;

from June 1990 through August 1990, and completes the initial startup i documentation as required by NRC Reg. Guide'1.16, Section C, Part Ic.  ;

Approximately eight months elapsed between completion of the low power ,

physics tests and receipt of a full power licenses the full power license 2 was received on March 15, 1990 and the power ascension test program was undertaken promptly thereafter. .

Af ter completion of preparations to begin the ascension to the . SOE power level test plateau, the first section of procedure ST-48 Turbine  :

Generator Startup Test was run and ST-48.1 Turbine Generator Torsional Response Test started. An unsatisfactory resonance was detected by the latter test, and the PATP was suspended for approximately one month to allow '

GE turbine personnel to modify low pressure turbine rotor 'C'. Upon ,

completion of the modifications, a retest of the turbine, using a revised ST-48.1, was conducted and satisfactory performance obtained.

Supplement 1 to the Initial Startup Report covers testing _through ST-48.1. The turbine was synchrunited onto the grid, but actual ascension to 302 power was not included.

A summary of testing was reported in Supplement 2 which covered entrance into the 302 power level test plateau through the NSSS Acceptance '

Test (ST-40), the final test in the PAT sequence.

Supplement 3 covers the Supplement 2 test period in detail, and is the final supplement to the Initial Startup Report.

The following tests are included in Supplement 3:

ST-13 Operational Alignment of Nuclear Instrumentation ST-14.1, Operational Alignment of the Process Temperature Instrumentation ST-15 Reactor Plant System Setpoint Verification ST-22 Natural Circulation Test l ST-24 Automatic Reactor Control ST-25, Automatic Steam Generator Level Control ST-26 Thermal Power Measurement and Statepoint Data Collection ST-27, Startup Adjustments of Reactor Control System ST-28, Calibration of Steam and Feedwater Flow Instrumentation j ST-29, Core Performance Evaluation 1 ST-30, ' Power Coefficient Heasurement ST-33, Shutdown from Outside the Control Room )

ST-34, Load Swing Test l

ST-35 Large Load Reduction  :

~ST-36, Axial Flux Difference Instrumentation Calibration l i

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1.0 INTRODUCTION

(Continued)

Tests included in Supplement 3 (Continued):

ST-37 Steam Generator Moisture Carryover Measurement ST-38 Unit Trip from 1002 Power ST-39 Loss of Offsite Power Test ST-40 NSSS Acceptance Test ST-41 Radiation Survey ST-42 Water Chemistry Control ST-43 Process Computer ST-44 Loose Parts Monitoring ST-45 Process Effluent Radiation Monitoring System ST-46 Ventilation System Operability Test ST-48 Turbine Generator Startup Test

• ST-49 Circulating Water System Thermal-Hydraulic Test ST-51, Power Ascension Dynamic Vibration Test ST-52 Thermal Expansion ST-56 Piping Vibration Testing

• Test deferred. Testing will be performed prior to operation of the circulating water heat treatment.

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. 2.0 STARTUP TEST PROGRAM OVERVIEW. SUPPLEMENT 3 The Initial Startup Report, submitted to the . Nuclear Regulatory  !

Commission on March 13, 1990, covered that portion of the startup sequence  ;

through low power physics testing. Supplement 1, was prepared to report  !

testing which took place in the three months after the initial report.

Supplement 2 was a summary document covering the tests reported in detail in l Supplement 3. Supplement 3 reports those startup activities after Supplement 1, through completion of the Power Ascension Test Program.

Startup teart 'h.ich had sections previously reported, but carry over i

.into the test sequences covered in Supplement 3, are included asLcomplete 1 tests.

A full power license was received on March 15, 1990. f Portions of several startup tests involving alignments and setpoints were required in the startup sequence prior to increasing power to the 302 power level plateau. These tests were completed and the Station Operating '

Review Comittee (SORC) authorized entry into the 301 test sequence on March  ;

25, 1990. During power ascension, the initial turbine generator tests were scheduled in the 8Z-201 power range. A turbine generator torsional response test had been reconsnended by General Electric (GE), the turbine vendor, and after turbine rolls to validate turbine protective systems and make adjustments to control systems, the torsional response test was undertaken.

The torsional res9onse test disclosed an undesirable resonance, requiring modification tv the 'C' low pressure turbine. Power ascension ;

testing was interrupted on April 27, 1990 for turbine modification and resumed on May 25, 1990.

When the turbine was reassembled after modification, a revised torsional response test was performed to verify correction of the resonance problem. Testing through completion of ST-48.1. Turbine Generator Torsional Response Test, was covered in Supplement 1. Supplement 2 was a summary document covering the remainder of the PAT sequences reported in detail in Supplement 3, the final supplement in the Initial Startup Report.

Supplement 3 reports ascension to the 302 power level test plateau and the remaining tests in the power ascension sequence. The 101-30% turbine tests, preliminary to testing at 302 power, were completed on June 4, 1990.

Section 5.0, Power Ascension Testing, includes those tests related to primary and secondary plant performance. Many tests in this sequence, usually those identified with one aspect of plant systems, such as S/G 1evel: )

control, were conducted at every. plateau, and related control systems readjusted as required. Others, such as the loss of offsite power, and unit ,

trip from 1002, were performed only once, and evaluated many plant systems j and system interactions. 4 At the 30% and subsequent power level test plateaus, several startup I tests were conducted to verify or readjust instrument ranges or setpoints.

In some cases data from these tests were entered into other tests: in

. general, tests in this category are found in Section 6.0, Instrument Calibration and Alignment.

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, . 2.0 S1ARTUP TEST PROGRAM OVERVIEW. SUPPLEMENT 3 (Continued)

Other tests to validate plant support systems, such as the radiation shield design and ventilation systems: are included in Section 7.0, General  ;

Plant Testing, and were sequenced at appropriate test plateaus. For example, ST-41, Radiation Survey, was performed at the 50% and 1002 power '

level test plateaus, but not at 302, where reduced radiation levels would mean fewer useful data points.

I In general, testing of the primary and secondary plant systems went smoothly, with only minor problems and interruptions. Turbine adjustment and fine tuning, primarily at the 30! plateau, was the major challenge early  !

in the test sequence. Later, condensate pump heater drain pump, and feedwater heater interactions contributed to control problems in the secondary plant, limited progress. i Testing of the NSSS portion of the plant yielded expected results, almost without exception. The availability of GETARS for data acquisition provided test personnel with fast and very accurate information. At no time did data analysis contribute significantly to delays in the program  !

sequence.

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The NSSS Acceptance Test, ST-40, was started on August 5, 1990 at 1700 hours0.0197 days <br />0.472 hours <br />0.00281 weeks <br />6.4685e-4 months <br />, and completed on August 17, 1990 at 1800 hours0.0208 days <br />0.5 hours <br />0.00298 weeks <br />6.849e-4 months <br />. The test was interrupted for two hours on ~ August 6 for stop and control valve surveillance tests at a power level of < 952, and for approximately 1 days on August 13 for repair of a leak in an EHC line. Later, on August 16, a steam leak in a S/G blowdown line-required isolation of the flash tank, but '

the line was repaired without reducing power.

Results of the acceptance test were reviewed following the warranty run, and the Power Ascension Test Program was officially completed at 2400 .

hours, August 18, 1990.

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,l .'3.0 SEABROOK STARTUP CHRONOLOGY. SUPPLEMENT 3 i

This chronology documents the Power Ascension Test Program (PATP) from ,

test resumption following main turbine modifications to completion of PATP.

Alignment and setpoint adjustments per ST-13 ST-14.1~and.ST-15 occur frequently in the test sequence. These entries have not been included in the chronology.

ELLE IZtal ,

5/30/90 Testing underway: transfer from bypass to main feedwater valves ,

(ST-25), overspeed testing (ST-48). ,

Test interruption for turbine overspeed trip rework.

3 6/4/90 Completion of 102-30! load data collection (ST-48). {

Ascension to 302 power.

6/5/90 Verification of S/G feedwater pump auto speed control (ST-25): '

completion of 302 thermal expansion data (ST-52). '

6/6/90 Turbine shutdowns arcing in generator isophase bus duct. ,

6/7/90 Verification of main feedwater reg valve stability (ST-25) =

preparation of MIDS for flux mapping (ST-29).

6/9/90 Water chemistry sampling (ST-42): flux mapping; at' 30" completed (ST-29) and first power coefficient determinatien (ST-30) completed valve and feedwater pump testing at this riateau (ST-25): initial MPCS data taken (ST-43). J 6/10/90 Load swings of 10% (ST-34): automatic reactor control verification (ST-24).

SORC approval received for ascension to the 502 power level test plateau.

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-l 6/11/90 Completion of 30Z-502 load data collection (ST-48): shield survey -)

underway (ST-41). l l

Fifteen hours S/G chemistry holdup.

6/12/90 Statepoint data collection (ST-26): steam feedwater flow i calibration (ST-28).

6/13/90 Adjustments to reactor control system (ST-27): completed thermal expansion observations (ST-52): core performance evaluation completed (ST-29): radiation surveys (ST-41) and chemistry sampling (ST-42) underway.

Feedwater heater level fluctuations prevent operation of MFP-B per 1 ST-25. Operations starts second heater drain pump to improve l heater drain tank level control. .)

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. . h.0 SEABROOK STARTUP CHRONOLOGY. SUPPLEMENT 3 (Continued) j 6/14/90 calibration of AFD instrumentation (ST-36): loose parts monitoring ,

(ST-44) and ventilation system operability test (ST-46) underway. j Power decrease of 100 MWe transient due to turbine control problem.

6/15/90 Power coefficient determination (ST-30): 102 load swings (ST-34).

6/16/90 Shutdovn from outside the control room (ST-33): MPCS data acquisition complete (ST-43).

6/18/90 Completed 50!' power level AFD calibrations (ST-36).

6/20/90 SORC approval received for ascension to 752 power'1evel plateau.

Unplanned turbine trip / reactor trip due to a fault in generator  :

protective relaying.

6/25/90 Restored criticality.

• 6/27/90 Repeated ST-48 302-502 load data collection. ,

Oscillations of feedwater heater level when second MFP (A) placed in service (second heater drain pump placed in service to dampen ,

oscillations).

6/28/90 Additional flux map performed.

6/29/90 Flux mapping in progress.

6/30/90 Heater drain piping leak, manual turbine shutdown.

7/5/90 Commenced testing at 75Z power: statepoint data (ST-26) and S/G 1evel control (ST-25) completed.

Unplanned reactor trip due to high vibration on- EHC system pressure switches. "

7/7/90 Reactor entered Mode 1.

7/8/90 Returned to 752 power.

7/9/90 Completed 502-701 load data collection (ST-48): thermal expansion evaluation (ST-52): steam and feedwater flow calibration (ST-28):

water chemistry sampling (ST-42).

7/10/90 Flux mapping (ST-29 and ST-36). j 7/11/90 During power coefficient measurement (ST-30), throttle pressure limiter interference caused megawatt reduction, and a test-interruption.

Completed 102 load swing at 752 (ST-34).

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3,0 SEABROOK STARTUP CHRONOLOGY. SUPPLEH17T 3 (continued) 7/12/90- Power coefficient determination (ST-30): large load reduction j (ST-35),

s 7/13/90 AFD calibration (ST-36) plant computer validation'(ST-43).

, 7/14/90 TREF program change completed reactor control system adjustments completed for 75% (ST-27).

I SORC approval received for ascension to the 100% power level test plateau. PATP Management hold for 901 testing.

70% 90% load data collection.(ST-48): flux mapping (ST-29).

I 7/16/90 Statepoint data collection at 90% power (ST-26). j Reduced power to 75% temporarily, for Operations to test Main l Steam Control Valves, process temperaturex alignment (ST-14.1) and -

steam /feedwater flow calibration (ST-28).

7/19/90 Returned power level to 90% and conducted _ additional process temperature alignment (ST-14.1).

1 PATP Management approval received.for ascension to the 100% power level test plateau. ]

I 7/20/90 Feedwater heater oscillations forced power reduction to 90%.

After approximately six -hours, return to 100% started: high feedwater flow oscillations required manual control of S/Gs . A.

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B and D. New gain settings requested for controllers. ,

7/21/90 Returned to 100% power level test plateau.

7/22/90 100% load data collection (ST-48): verification of feedwater pump speed control (ST-25) statepoint data collection (ST-26):

chemistry sampling (ST-42): shield survey (ST-41); moisture carryover test (ST-37) baseline loose parts monitoring (ST-44):

ventilation system (ST-46).

7/23/90 Thermal expansion (ST-52) and piping vibration (ST-56) completed:

process effluent data collection (ST-45) core performance flux maps (ST-29).

7/26/90 Load swings (ST-34) and Large Load Rejection (ST-35) accumulated >

60 AFD penalty minutes _ required power to remain below 50% for 24-hours.

7/28/90 Returned to 100% power level preparations for unit trip from 1002 (ST-38).

7/29/90 Unit trip from 100% power (ST-38) coordinated with process ,

computer (ST-43), LPMS (ST-44) and vibration measurements (ST-51):

natural circulation test (ST-22).

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. $,0 SEABROOK STARTUP CHRONOLOGY. SUPPLEMENT 3 (Continued) 1 7/31/90 During restart, high vibration on turbine bearings required power  ;

reduction and then a manual turbine shutdown. After four hours on I turning gear, turbine was resynchronized. I l

8/1/90 Loss of offsite power test from 202 RTP (ST-39).  !

8/3/90 Returned plant to services preparation for ST-25.1, single  !

feedwater pump capacity test.  ;

8/4/90 ST-25.1 (Main Feed Pump Flow Capacity Test) interrupted (and later cancelled) when MFP A suction pressure dropped and was accompanied  !

by feedwater heater level instabilities.

8/5/90 Attempt to reach 1002 RTP on two condensate pumps: third pump (in automatic) started on low suction pressure at 882 RTP. f Commenced 250 hour0.00289 days <br />0.0694 hours <br />4.133598e-4 weeks <br />9.5125e-5 months <br /> warranty run (ST-40).

1 8/7/90 Control system adjustments (ST-27).

Precision calorimetric measurement (ST-40).  ;

8/13/90 EHC piping leaks turbine shutdown to effect repair, f 8/14/90 Generator back on line, j 8/15/90 Recommenced warranty run. ,

8/16/90 Two inch crack found in S/G C blowdown piping reducer flash tank ,

isolated for repair. .i Completed process computer data (ST-43). -

8/17/90 Blowdown returned to service.

Warranty run completed (ST-40).

8/18/90 Review of warranty rung startup program officially completed.

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, ile.0 SUMARY OF INITIAk STARTUP REPORT. SUPPLEMENT 3 The Power Ascension Test phase of initial startup was initiated with  !

the receipt of a full power license on March 15, 1990. Supplement 3 reports testing beginning with the approach to the 302 power level test plateau and ,

. continuing to the completion of the PAT sequence.

l Supplement 2 was a summary document covering the tests reported in i detail in Supplement 3. I l

Those tests which were conducted at several power level test' plateaus, and were reported in the' original startup report and/or in Supplement 1, are  !

reported in Supplement 3 completely, for continuity. In some cases, less ,

detail is included for sections previously reported. Following is a list of i startup' tests which were completed . prior to the time frame covered by i supplement 3: these terts are discussed in the earlier documents and are not  !

discussed herein:

ST-2 Primary Source Installation ,

ST-3, Core Loading Prerequisites  :

ST-4, ' Initial Core Loading i ST-5, Control Rod Drive Mechanism Operational Test J ST-6, Rod Control System 1 ST-7, Rod Drop Time Measurements

• ST-8 Rod Position Indication ,,

ST-9, Pressurizer. Spray and Heater Capability  :

ST-10 RTD Bypass Loop Flow Verification l ST-11 Reactor Coolant System Flow Heasurement '

ST-12 Reactor Coolant System Flow Coastdown ST-14.2, Resistance Temperature Detector and Incore ,

Thermocouple Cross Calibration '!

ST-16, Initial Criticality ST-17, Boron Endpoint Measurement ST-18 Isotherral Temperature Coefficient  !

ST-19, Flux Distribution Measurements at Low Power q ST-20 Control Rod Worth Heasurements '

ST-20.1 Additional Control Rod Worth Heasurements ST-21, Pseudo Rod Ejection Test ST-23, Dynamic Automatic Steam Dump Control d ST-48.1 Turbine Generator Torsional Response Test '

ST-50, Movable Incore Detector System l ST-53 Turbine Driven Emergency Feedwater Start l Verification J ST-55, Steam Dump System Test ,

In the 302 sequence, reactor power was increased enough to bring the:

turbine on line (8-101), and after a detailed checkout, to allow synchronization of the generator to the grid. Only after completion of the turbine generator qualification testing did actual ascension to and testing at 302 power begin. As was reported in ' Supplement 1, power ascension-testing was interrupted on April 27, 1990 for turbine modification, and  ;

resumed on May 25, 1990. '

i 4-1

Power pla teaus ascension 302 t testing, he longest testing moved later, as turbine interruption rapidly 8 hold of two power was overspeed problems about through t previously shiftsescalation to therequiredfour 502 days. he several test At the foreign not in was necessary as turbiplateau service three days, start and aoff matter. was underway ew days systems by the usejust coming Water o received steam ne drains a \

and were n li chemistry testing (ST 42flushed p,iping systemchemistry and other s which were frequentlyoftrailer-mountedCleanup auxiliary,ne. )

of of secondary residual results were Two changed as eystems typical of the needed. was s carbon filter days,due econd atunplanned reactor s trip and resinaccelerated beds, to result of a

thefault752 power leve,l the first at approxi Turbine high were set backs vibration in EHin C

test plateau, interrumately 302 power generator protective and 1

Fine tuning of the responsible for test e ays d switches, lsystem and 3pressurerelaying,pted days as a tes

\ feedwaterwas the primary acti i secondary plant, of 1-2 shifts.respectively.

{

v lve a

heater drain instabilityv pumps ank ty outside the testiparticularly ilevel-heater level required drain t the feed pumps atte required condensate although the pumps nto care, additionalservice.

and then feedwa fluctuations, terInitially ntion,n were system needed, to Smooth and for full operation two withproblems wo in bringing theat power main feedwater extrapola tionsInstrumentation was designed pump operation.

operation for thretwomaintain feedw suction pr, essure,e after plant the 902 topower full poweradjustments le were such as the T efficien vel test gyg specifications.cy and lowered plateau, withongoing with few problprogram, ba sed on The turbine impulsethe MSRs in ems.

Opera tion, loss of major plant pressure service improved problems. all offsite transient tests, down to power, design Report, When were such as problems Initial attempted completed as large lood conjunction with a document),in naturthe low power ph planned reduction, with onl and with thesteam dump valve. al circulationysics test unit (ST-22), sequence (y minor availa bilityTesting of the from 1002 powThe test wasfailed Startup tripNSSS fast and of GETARS portion of er (ST-38).successfullydueperformed to in contribute very accuratefor data the plant yielded significantly to delinformation. At acquisitiontest provided expected results no designTests to time di personnel with and and ventilationvalidate rt plant suppoays sequence .d data in the program analysis systems systems, were such as the sequenced at appropriatradiation shield e test plateaus.

4-2

o

• l l

4.0 S1994ARY OF INITTAf STARTUP RDORT. SUPPLEMENT 3 iContinued) l Power ascension testing moved rapidly through the several test plateaus: the longest interruption was about four days. At the start of ,

302 testing, turbine overspeed problems required three days, and a few days later, as power escalation to the 50t plateau was underway, a chemistry hold of two shifts was necessary as turbine drains and other piping systems previously not in service received steam and were flushed of residual foreign matter. Water chemistry testing (ST-42) results were typical of systems just coming on line. Cleanup ~of secondary systems was accelerated by the use of auxiliary, trailer-mounted carbon filter and resin beds,  ;

which were frequently changed as needed. -

i Two unplanned reactor trips, the first at approximately 302 power and the second at the 75Z power level test plateau, interrupted testing for 4 l days, due to a fault in generator protective relaying, and 3 days as a result of high vibration in EHC system pressure switches, respectively.  ;

Turbine setbacks were responsible for test delays of 1-2 shifts.

Fine tuning of the secondary plant, particularly the feedwater train, ,

was the primary _ activity outside the testing sequence itself. Initially ,

feedwater valve instability required attention, then feedwater heater ,

level-heater drain tank level fluctuations, and problems in bringing two heater drain pumps into service. Smooth operation with two main feedwater ,

pumps required additional care, and for full power operation, three condensate pumps were needed, to maintain feedwater pump suction pressure.

although the system was designed for two pump operation. '

Instrumentation adjustments such as the TAyo program, based on '

extrapolations to full power, were ongoing with few problems. Operation, af ter the 901 power level test plateau, with the MSRs in service improved  :

plant efficiency and lowered turbine impulse pressure down to design specifications. ..,

The major plant transient tests, such as large load reduction, and loss of all offsite power, were completed as planned with only minor problems. When attempted in the low power physics test sequence (Startup Report, Initial document), natural circulation (ST-22), failed due to ,

problems with a steam dump valve. The test was successfully performed in conjunction with the unit trip from 100% power (ST-38).

Testing of the NSSS portion of the plant yielded expected results and l availability of GETARS for data acquisition provided test. personnel with fast and very accurate information. At no time did data analysis 3 contribute significantly to delays in the program sequence. j Tests to validate plant support systems, such as the radiation shield design and ventilation systems were sequenced at appropriate test plateaus.

r 4-2 l i i

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. '4.0 SLM4ARY OF INITIAL STARTUP REPORT. SUPPLEMENT 3 (Continued)

The NSSS Acceptance Test, ST-40, was started on August 5, 1990 at 1700 hours0.0197 days <br />0.472 hours <br />0.00281 weeks <br />6.4685e-4 months <br /> .and completed on August 17, 1990 at 1800 hours0.0208 days <br />0.5 hours <br />0.00298 weeks <br />6.849e-4 months <br />. The test was interrupted for two hours on August 6 for stop and control valve surveillance tests at a power level of < 952, and for approximately 1\ days on August 13 for repair of a leak in an EHC line. Later, on August 16, a steam leak in a S/G blowdown line required isolation of'the flash tank, but the line was repaired without reducing power.

Results of the acceptance test were reviewed following the warranty run, and the Power Ascension Test Program was officially completed at 2400 hours0.0278 days <br />0.667 hours <br />0.00397 weeks <br />9.132e-4 months <br /> August 18, 1990.

i 4-3

^I

. .'5.0 POWER ASCENSION TESTING Contents:

5.1 ST-22 Natural Circulation, Test i 5.2 ST-24 Automatic Reactor Control  !

i 5.3 ST Automatic Steam Generator' Level Control 5.4 ST-29 Core Performance Evaluation 3.5 ST-30 Power coefficient Measurement j 5.6 ST-33 Shutdown from outside.the Control Room 5.7 ST-34 Load Swing Test 5.8 ST-35 Large Load Reduction  ;

5.9 ST-37 Moisture Carryover Measurement t

5.10 ST-38 Unit Trip from 100% Power 1 5.11 ST-39 Loss of Offsite Power Test

=5.12 ST-40 NSSS Acceptance Test l 5.13 ST-48 Turbine Generator Startup Test k

I'

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'5 1 ST-22 . NATURAL CIRCULATION TESI obiective r

The objective of this test was a demonstration of heat removal from the reactor coolant system using natural circulation and determination of l several natural circulation characteristics. '

The natural circulation test is described in FSAR, Section 14 Table l 14.2-5; Sheet 25. ,

Discussion  !

Natural circulation requires residual heat in the reactor core to establish. convection flow. ST-22 was initially scheduled in the . low power physics test sequence, with the necessary core heat provided by maintaining >

reactor power at approximately 37. Failure of a steam dump valve to function properly prior' to reaching natural circulation conditions required a rescheduling the test in the PAT sequence.

ST-22 was conducted in conjunction with ST-38, Unit Trip from 1002 Power.

I The natural circulation test demonstrated the followings i

. The reactor coolant system can transition from forced to )

natural circulation.

. Natural circulation is established and paintained as indicated by stable RCS temperature indications.

• A uniform reactor coolant flow distribution under natural circulation -

conditions as indicated by incore thermocouple temperature data.

• A determination of the length of time necessary to stabilize natural circulation.

In addition, the test provided data to verify simulator modeling and to support results of transient analysis.

The test was initiated following the trip from 1002 power.

A loss of forced flow was simulated by simultaneously tripping all reactor coolant pumps. Manual manipulation of the pressurizer pressure control systems utilizing auxiliary sprays adjustments of charging and letdown flows and ASDV use ensured stable plant conditions during natural circulation operation. '

Results i The acceptance criteria were mets demonstration of natural circulation by a stable RCS temperature, and a coolant flow distribution . se shown by incore thermocouple maps.

Following RCP trip, the transition to natural circulation , vent j smoothly, requiring approximately eleven minutes.

i i l

5-2 i j

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• -a 5.1 ST-22. NATURAL CIRCULATION TEST fContinued)

Auxiliary spray for pressuriser control was set up three minutes after RCP trip, and as shown by Figures 1.2, 3 and 4, prepared from GETARS data a smooth transition to a natural. circulation state was obtained.

Pressurizer pressure peaked 'at THOT increased, and then stabilised at a value about 20 psig higher than the initial value. ' ASDVs cycled - while natural circulation conditions existed, and as noted - provided ~ a stable means of automatic heat rejection. RCPs were restarted' at approximately 50.3 minutes after pump trip.

The final conditions specified for the test were met as follows:

(1) Level of Decay Heat (Q) - Q was calculated to be~56 MWth prior to trip of the RCPs, using the known flow rate and T y3(NR).

A 1

(2) RCS Flow Rate under Natural Circulation - The RCS flow rate under natural circulation was calculated using Q (56 MWth) and TAyg(WR). A flow rate of 4.92 of full flow resulted. It is noted that the use of TA yg(WR) introduces a relatively large ,

uncertainty in'the result. TAVG(NR) was not used because of its l unreliability under natural circulation conditions.

(3) Uniformity of Coolant Flow within the Reactor Core - From the l incore T/C temperature distributions taken with pumps running and after establishing natural circulation,' a uniform coolant flow 4 I

with natural circulation was verified.

(4) Time for Natural Circulation to Stabilize - Approximately eleven minutes after RCP natural circulation had been established.

1 (5) Ability of Subcooling Monitors to Accurately Display Saturation- I As shown in Table 1, the subcooling margin displayed was in -I agreement with the margin determined using RCS (WR) pressure and Core Quad Max Temp.

I 4

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, 5.kST-22.NATURALCIRCULATIONTEST(Continued)

TABLE 1 ST-22 NATURAL CIRCULATION SUBC00 LING MARGIN I

RVLIS RVLIS Calculated GETARS RCS (WR) Core Quad T (Sat. Sub. Sub. Margin Time Pressure Max Temp Curve) Margin Margin 'from (A2970) (A2927) ('F) ('F) ('F) Curve (psig) ('F) (*F) 9:30 2285.0 579.9 655.8 76.6 78.4 75.9 9:32 2281.9 579.5 655.5 77.1 79.2 76.0 9:34 2283.5 581.19 655.6 76.9 78.7 74.4 9:36:15 2286.6 579.9 655.9 76.8 79.1 76.0 9:38:15 2286.6 579.5 655.9 76.9- 80.0 76.4 9:39:15 2288.2 579.9 656.0 76.7 79.1 76.1  !

9:42:10 2286.6 579.5 655.9 77.0 79.8 76.4 9:44:15 2286.6 579.5 655.9 77.3 79.7 76.4 1

9:46:20 2286.6 579.9 655.9 76.9 80.0 76.0  !

I 9:48 2286.6 580.2 655.9 76.6 79.7 75.7 '

9:50:10 2286.6 579.9 655.9 77.3 79.7 76.0 9:52:05 2285.0 579.9 655.7 77.5 80.3 75.8 i

9:54:20 2285.0 579.5 655.7 77.2 79.6 76.2 09:56:05 2285.0 578.2 655.7 77.5 80.0 77.5 9:58 2285.0 578.6 655.7 77.3 77.9 77.2 10:00:10 2285.0 579.9 655.7 77.2 79.4 75.8 5-4

i ST-22 NATURAL CIRCULATION R# 1 TIME 9:12:35:889 DATE 7:29:90 6ETBNS-ISEBFNIC TRIP CH # 126 AT 9:12:35:989 DATE. 7:29:90

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5.2 ST-24. AUTOMATIC REACTOR CONTROJ-Obiective The objective of this test was- a demonstration ~ that the Automatic Reactor Control System is able to maintain the ' average reactor coolant:

temperature within acceptable steady state-limits.

The procedure _ is described in FSAR, Section 14. Table 1L4.2-5, Sheet 27 $-

Discussion Procedure ST-24 was performed with the reactor at the 30Z - pover level-test plateau..in three sections: _

1) Stable operation, following a switch-to automatic control.

2) Automatic restoration of stable operation from a. temperature; imbalance of TAVG > TREF by approximately 6'F.

3) Automatic restoration of stable operation from ' a temperature (

1mbalance of TAVG < TREF by approximately 6*F.

The change-in TAVG was produced by moving Control Bank D and_ stability determined by monitoring !!i: TAVG/ TREF. A maximum deviation of i 1.5'F _was -

permitted for satisfactory performance.

Results All acceptance criteria were mett Automatic plant control systems respond properly.

Test Results Satisfactory performance.

No manual intervention required to bring . TAVG to and maintain within i 1.5'F of TREF' Test Result; No manual intervention.

On increase. TAVG returned to 1.0*F of' TREF On decrease, T A yg returned to O'.4*F'of TREF

• Combined TERROR signal returned to

• 1 F of TREF*

Test Result; On increase, TAVG returned to 0.8'F of TREF On decrease, TAVG returned to 0.8'F of-TREF There were no test exceptions.

5-7

, . 5.3 ST-25. STEAM GENERATOR AUTOMATIC LEVEL CONTROL Oblective This procedure demonstrated the'= stability of the Automatic Steam.

Generator Level Control System under ' simulated _ transient conditions, andt proper operation of the main feedwater pump speed control.. System stability during transfer from bypass feedwater regulating valves to main feedwater regulating valves was included.

The steam generator automatic level control tests are described in FSAR Section 14. Table 14.2-5 Sheet 28.

Discussion Level control of the steam generators and speed control of feedwater pumps is validated in ST-25 by testing at reactor power levels of l!-42,82-102, 82-202, 302, 502, 75Z,,and 1002.

At a power level of approximately 31, each feedwater bypass valve-automatic controller was tested, using station operating procedures, y to i demonstrate stability during a steady state manual to automatic transfer, as well as stability , during steady state operations. The changeover from startup feedwater pump to one main feedwater - pump was next demonstrated:

using station procedures. Again, stability during a steady state manual to automatic transfer, and steady state operation in automatic control was  !

verified for main feedwater pump operation. The changeover from _ startup _j feedwater. pump. to the other main feedwater . pump, and the stability 1 verifications was demonstrated at the end of the test, by repeating-- a - <j section of the procedure.  ;

1 At 8Z-102 power, after the feedwater regulating block < valves were t opened, and stable level control demonstrated with - feedwater . regulating {

bypass valves in automatic, the ability of-each of these valves to restore l and maintain steam generator narrow range level was demonstrated when 1 the _'

narrow range level was successively raised and lowered.

A transfer from bypass to main feedwater regulating valves was made at 82-202 power, and after stable level control was demonstrated in automatic again, and the power level raised to 302, the narrow K range -level was successively raised and lowered and the ' ability to ' restore and ' maintain .

level demonstrated under these conditions.

Main feedwater pump (MFP) automatic speed control was demonstrated at.

the 302, 502, 75Z and 1002 power level test plateaus'by monitoring feedwater' 1 pump parameters and' steam generator levels.

5-8

..a

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e

.. . 5.3 ST-25. STEAM GENERATOR AUTOMATIC LEVEL CONTROL (Continued) k Results-

-l All acceptance-cr.iteria were met I

. No manual: intervention was required af ter initiating automatic; Control.

., steam generator level returned to and remaindd within

• 22 of i the reference level, within 3 times the level' controller time constant, following transfers and-simulated level transients.

. Steam : generator level overshoot' (undershoot) was less than 42 following a level increas'e (decrease).  ;

e Feedwater pump- discharge pressure oscillations were_ less thanL* . l 3Z of the final value.  !

. At the 100Z. power level'. test plateau, ' the- main feedwater regulating. valve stem position stabilized at'less,than 85% open.

A test data sumnary is given in Table 1. System behavior at full power i is-shcwn in Figures 1 and 2.= i Two test exceptions were taken. One excepted. a missing data sheet:

which was located 'later; the.second was taken to account for an erroneous" The reading >failedit o meet acceptance criteria,

~

valve position indication. 't but the actual position, as determined.by' valve positioner, was,752,'meetingi the criterion.

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. . 5.3 ST-25. STEAM GENERATOR AUTOMATIC LEVEL CONTROL ~(Continued)

' TABLE 11 i TEST DATA, ST-25 STEAM GENERATOR AUTOMATIC LEVEL CONTROL Automatic Level Control on Bypass Valves: SG Level at 50%

Controllers (2.Open): .

Acceptance Criterion:

No manual intervention.

Maximum 39%

Minimum 362 -None required.

Main Feedwater Pump ~A (or B) operating in automatics S/G Level at 502 Acceptance Criterion:-

No manual intervention.

None required.

Feedwater regulating block _ valve opent feedwater regulating bypass valves in automatics S/G Level at 502 Acceptance Criterion: ,

No manual' intervention.

None: required.

Main Feedwater Regulating Valve Controller Stability, in Automatics 302 Power Level (2) Recovery Level FW Pump Test Plateau' Change Initial Timef overshoot- ' Discharge

/ Final (Seconds) Undershoot- Pressure.

Valve, Osc. t FW-LK-510 Raised 50.8 51.6 2340 -1.5 None Lowered 49.5 49.7 2400' O.4 ~None-520 Raised 49.1- 50.0 2190 1.8 None Lowered 49.7 48.5- 2240 -2.4 None 530 Raised 49.6 50.2- 2310 1.2 None Lowered 50.0 48.6 2170- -2.8 .None 540 Raised 49.3 50.3 2460 2.0 None Lowered 50.0 48.6 1860 -2.8 None Acceptance Max.' Min.

Criteria 52 -48 ~3000 <42 <'*3%.

Computed Difference Between Feedwater - Pump Discharge Pressure and . - S team Header Pressures-Power Level. Computed Delta-P: Master' Speed.

Controller Setpoint: 1 SOI 100 psi. 90 psi l 50 125 110 I 75 163.4 162.4 , i 100 196.5 195.6 l Main Feedwater valve position (reg. valve controller M/A station):

Valve ID Controller Output (2 Open)

FW-FCV-510 77 l FW-FCV-520' :76 o l FW-FCV-530 78 '

FW-FCV-540 75*

• Control Room reading 852, see Results for explanation.

5-10 l

_ -. . - - . ~ . _ . = - . . . - . _

ST-25 NFP A AND B, AUTO 100% R# 24 TIME 23:59:31:563 DATE 7:21:90 gel #RS-I6##FNIC TRIP CH #-127 RT 0: 0: 5:663 DATE 7:22:90 ,

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5.4 ST-29. CORE PERFORMANCE EVALUATION Obiective-The> objective .of- this procedure was verification - of1 proper ; core

. performance through acquisition and analysis of incore flux and thermocouple maps.

.The' procedure'is described in FSAR, Section 14. Table 14.2-5 Sheet.32.

Discussion

< ST-29 was performed in its entirety at the 30Z, 50Z, 75Z, 902, and'1002 power level test plateaus and . utilized' Reactor Engineering . surveillance procedures throughout.

Full core flux maps - (PCFM) .were taken -- and analyzed using Reactor Engineering ~ procedures. = In1 the analysis, the resulting measured versus predicted assembly power distribution; was compared within the INCORE 7 computer code.

A quadrant = power tilt ratio (QPTR) .- surveillance was taken'. with each flux map at greater than SOI power - to ensure that technical specification requirements were met.

Analysis using additional Reactor Engineering procedures yielded - the-heat - flux hot channel Z), and _ the nuclear enthalpy, . rise hot channel factor, FDELTA-H.factor,. The Fq(kingpea factors were then used.to' support power ascension to the next power level test plateau.

Departure from Nucleate Boiling (DNB) data was transcribed from the-  !

results obtained at the same . test plateau by ST-26 Thermal Power Measurement and Statepoint Data Collection.

Results All acceptance criteria were met:

The core performance parameters of Fq(Z), FDELTA.H, QPTR, and DNB parameters meet technical specification requirements.

Discrepancies in the measured to predicted l assembly- power distribution shall'be less than 10%.

Results of the core performance parameter measurements are given in Table #1.

Core performance results at the 50% power level test plateau were-acceptable for escalation to the 75% plateau. However, a technical specification requirement in,the station surveillance procedure used in ST-29, required a remeasurement of FXY at 652 RTP. The necessary flux map was taken at 65Z with acceptable results.

5-14

.t

. 1

,,' ,,5,4'ST-29.

CORE PERFORMANCE EVALUATION (Continued) j l

The value' of- . FXY at the 100Z power level: testi plateau-l exceeded the l full-power limit .of 1.55 -(Measured: Value. -1.572). The value was. I consistently high at- lower power levels also: however, at lower' levels it I did not limit escalation to the next test plateau. ]

A test exception was- taken to accept' the result based on the value of F 2.08, Upper Limit = ?.32), and.a Beginning ~of Cycle ~ 1 ~ j F /FXY Evaluation w = hich determined that ' there was sufficient margin between (Measured Fq _

te design- Fo for all possible operating conditions and; the technical specification limit.

TABLE 1 ST-29 CORE PERFORMANCE PARAMETERS Power Level 30Z. 50Z 752. 902 1002 Bank D Position 210 210 191 201 212 FXY 1.596 ~1.567 1.558 1.565 1.572 Fq 2.189 2.120' ;2.089 2.080 -2.077 FDELTA-H 1.454 1.430 1.390 1.396 1.404-Maximum 1.014 1.017 1.016 1.014 1.016 Incore Tilt-l l.

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.'S.5 ST-30. POWER COEFFICIENT MEASUREMENT ObiectIve-  !

The Power Coefficient Measurement verified the nuclear des'ign predictions of . the ' Doppler-only' power coefficient,- through correlative-measurements of RCS temperatures and core thermal power. output.  ?

The. measurement is described in FSAR, Section 14 Table 14.2-5, Sheet.33.

Discussion a The Doppler-coeffic'ient of reactivity is-that portion of the reactivity.

feedback. due to temperature changes in the fuel. The Doppler temperature i coefficient, pcm/'F, relates the reactivity change to the change in average:

temperature of the fuel. The ' Doppler-only" power coefficient. pcm/Z' pwr,

relates to the change in power which produced the-temperature change.

'

ST-30 was performed at the 30Z, 502, 752'and 100Z power level ~ test' plateaus.  ;

i

-c In the test. TAVG, delta-T, and reactor power !were measured , for a ]

series of three small--(= 32) load increases (and decreases) by changing the-turbine generator' output with control rod position held constant. The power >

coefficient was then inferred from a quantity called the Doppler coefficient- a P

verification factor, C , defined as the ratio' of the, change ~1n . core average .

temperature to the change in core' power due to the Doppler effect.

The predicted'value of'CP was determined from'the Nuclear Design Report (NDR), WCAP-10982.

i Results At all test plateaus .the acceptance criterion was met, the : average measured Doppler coefficient verification factor shall. be within 't :0.5'F/I of the predicted Doppler coefficient verification factor.'

TABLE.1' ST-30 POWER COEFFICIENT MEASUREMENT Difference Between 1 Doppler Coefficient

-Measured and" Predicted Power Level Verification: Factors.

30Z -0.2478"F/Z:

50% :0.1732 752 0.0116 100% 0.1121 j 5-16 l.

i k

~

(

. '5 6 ST 33

. - . SHUTDOWN FROM-0UTSIDE THE CONTROL ROOM l 1

Obiective l The procedure demonstrated that the reactor could e be < tripped from.a_  ;

. location' external to the control room, thati operations could be transferred l to the remote . safe shutdown' (RSS) facility, l and the plant - brought' to' hot 1 standby with the normal shift compliment of personnel'. l l

The procedure is described c.n FSAR Section 14, Table 14.2-5, Sheet 36. '

Discussion-1 ST-33 was performed - ct - the end of the 50Z L power ' level test- plateau.

The power level was-reduced.to approximately-20Z rated-thermal _ power as an '

initial condition' prior to shutdown.

7 Two operating crews, funder the direction of a:1 single Shift Superintendent, performed-the test. Ultimate command ~and control authority.

remained .in the control . room with the normal watch crew. The second crew-initiated the trip and performed; the shutdown. from the RSS facility,t while =

the control room crew. observed and monitored plant status-using all.of the instrumentation available at the main control board (MCB). The reactor coolant pumps remained in operation throughout the procedure.

The remote ehutdown utilized Train B and-selected portions of Tra h A.

RSS~ controls. Some Train A controls were selected for use to minimize the~ '

amount of protection and control-functions bypassed when local control was established. Communications were established using the RSS sound-powered-phone channel and two-way radios. .

Results The acceptance criteria for the test:

1. The reactor has been-tripped from outside the-control' room.

2. The unit has -been . maintained . at . stable (TA yo > 480*F), . hot standby conditions for at least'30-minutes.

3. - Operations to: control RCS temperature and to maintain the reactor in a safe shutdown condition were performed without . assistance l from the Control Room Crew.  !

All acceptance criteria were met.

. O 1

1 1

5-17

_ _ ~ _ .-

1 -

,,_ . 5.6'ST-33. SHUTDOWN FROM OUTSIDE THE CONTROL ROOM-(Continued)

The -reactor . trip was initiated > from the Train . A 'Switchgear Room by -

' depressing; the - trip levers for ' reactor- trip and . bypass breakers.- = The i

MSIV's were closed .from the B RSS panels the. resulting transient, was easily-  ;

controlled'and plant = parameters remained relatively stables' ,

SG' Pressure = Small' change. -20 psi ~to +30-psi.  !

THOT (WR).- Decrease,-5-15'F Pressurizer Levelt- Decrease,'10%  :

Steam Generator Levels _Small change -2%.to +4%/(NR)  ;

Automatic EFW Actuation -~No The Train A steam driven. EFW pump was manually started ; from the RSS -

panels.-but the, motor driven EFW pump was. - not - required and was .never started.' Steam generator levels were controlled at the B RSS. panel'(B'and D generators) and the A~ RSS panel (A and- C generators); .the atmospheric steam .

dumps-(ASDVs) were-not used until about' 25 minutes into the recovery..'and -

then only minimal jogging was required.'-

Plant conditions-at the RSS panels - on completion - of' the thirty minute stability period were as follows:

THOT - L1 = 575'F L4 = 560*F .

TCOLD - L1 - 570'F L4 - 560*F-SG Pressure - SGA 1095 psig, SGB 1080 psig.

SGC 1100 psig, SGD 1070 psig.

SG. Levels --SGA 862, SGB 80%.

SGC 862, SGD 85%.

Pressurizer Level 25%. i RCS Pressure . 2195-2200 psig. '

RCS Cooldown Rate - Approximately 10'/ hour.. I Two minor problems were observed- during the test. .

Three computer.

points were not recorded-by the MPCS, and operation of-the RSS Panel PCCW HX temperature control valve hand controller was reverse acting! from'what hads been expected, but posed no control problem. ,

]

1 There were-no test exceptions. l 1

1 I,

5-18 1 1

,-. .. 5.7 ST-34. LOAD SWING TEST

.ObiectiveL q The objective . ofl this procedure. was a : demonstration of ' proper plant- '

transient and automatic control system . performanceL for a 102 step . load .

, change introduced at the: turbine generator. J The procedure;is described-in-FSAR.' Section 14,. Table 14.2-5 Sheet 37. l 1

l

-\

Discussion The load _ swing : test was performed at the 302,l 502, 75% and 100Z power level test plateaus.

With the plant stable and controla systems ~ in , automatic, a step sload' ,

change was introduced by manual manipulation of turbine = generator controls -

associated with the load set' reference signals. between : primary and standby s controls. By : use - of the standby - . load set potentiometer, a deviation; ,

between the standby and -primary control signals: was Lintroduced, which  :

developed .an equivalent- 10% turbine power signal. mismatch.- ..Once : the mismatch had been- developed, transfer to' standby control caused a-stepwise- >

load change. ,

The transient (initially a load decrease) was recorded on GETARS and an MPCS trend block, and after system stability' had -been observed ~ for approximately ten minutes, transfer back to primary control producedi a reverse load change (increase).

Results Successful completion of the-load swing test is. based on the:fol' lowing- ,

acceptance criteria:

• No reactor or turbine trip. q

. No safety injection (SI). :l l

• No lifting of steam generator ASDVs or safety. valves. -l No' lifting of pressurizer PORVs or safety valves, 'j

• No manual intervention to reach steady state. l

• No sustained or diverging oscillations.in plant parameters, j Nuclear power' overshoot (undershoot) < 32: rated thermal. power. l No manual intervention for TAVG 1 1.5*F TREF after load swing.- ,s During data evaluation, the combined TERROR. signal,,which is TAyg

- Power Mismatch-(*F), was returned-to within i l'F of. TREF after-the load swing. .

. Feedwater pump discharge pressure oscillations are less 'than i- 32 of the final'value, two minutes after aLeteam flow change. i w

5-19 I

a

i

.. .. 1

.>5.7 ST-34'. LOAD SWING TEST (Continued)

I

.g All acceptance criteria were met except for the following  ;

At the 50% power level test plateaus-I TAVG returnsito

• 1.5'F of TREF- .

Actual return (load increase) approximately 1.6*F .

TERROR returns to

• l'F of TREF r Actual return (load increase) approximately'1.5'F= l At the 1002 power level test plateau No manual intervention: .

Loop 2, S/G' Level Control placed in manual (load decrease).- 1 TAVG returns to *~1.5'F of TREFI .

Actual return (load increase) approximately 3.2*F.

Test exceptions were written to address the- deviations. - At the 1002 conditions ; manual control was takeni but GETARS . data - showed . - that the.

oscillation was converging at the times the failure ofLTAVG to. return, was due to' control rods reaching a. fully withdrawn-position. "

j In . addition . to meeting ' acceptance criteria.: the procedure requires.

certain conditions to be reported to Westinghouse for review.- The following -

discrepancies, at the specified power levels, were reported to Westinghouse.. y Westinghouse, after review, found the values acceptable:

At the SOI power level test plateau:

-1 SG (NR) Level Variations: 1 Range 6.82'to 7.8% (Load Increase)  !

6.8% to 7.4Z (Load-Decrease)

Reporting Requirement: 2.5%

L SG Press. Over/Undershoots. >

. Range 36.2 to 37.5 psig (Increase)

Reporting Requirement > 25 psi. "i d

At the 50%. power level test plateau The above parameters again exceeded the Westinghouse limits.

At the 75% power level test plateaus.

SG (NR) Level Variations: ' 'l Range 12.5Z (Load Increase)*

Reporting Requirement: 2 5I

• The power level swing was specified to be 10%.

Actually, the value at this plateau was 16%.

Typical transient behavior at the 1002 - power . level test plateau is.

shown in Figures 1-8. -l l

5-20' l' l l

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-. c . . -

,5.8 STI35. LARGE LOAD REDUCTION Obiective The large. load reduction test demonstrated proper automatic response.of'-

the plant control- systems and - proper plant transient response to a_large

load reduction;of approximately 50%.

ST-35 is described-in FSAR, Section 14,= Table 14.2-5! Sheet 38.

Discussion The~1arge' load reduction te'st was conducted with the plant operating at' the 75Z and'1002 power level' test' plateaus.

A largeJload reduction, subjecting the! plant to the design rampa rate;

(1332/ minute), utilizes the ' Decrease Loadpushbutton on.the: turbine' load -

selector. .With' plant systems. stable, controlisystemsiin' automatic,-and the':

turbine generator in the manual mode, the1' Decrease Load _" pushbutton was depressed and held until the load redu.ction setpoint (indicated on~the Load' j Set Meter)Lwas decreased by 45%. The 45% decrease..was selected to minimize '1 the= possibility of an overshoot beyond 50%. _'-

In . addition to the acceptance criteria for.the test (Results below)',

deviation - of parameters beyond stated limits . required notification of ; the 4

' Westinghouse representative.

t i

Results All acceptance criteria were met  ;

• The reactor and turbine did not trip.

. Safety injection was not initiated, j

e Pressure responso and visual observation' verify that the steam generator safety valves did not;11ft.

No manual intervention was required to bring plant . conditions ~ to '

steady state.

• Plant parameters did 'not incur sustained or- diverging-oscillations. #

The feedwater pump discharge pressure oscillations are leso than- '

• 32 of the final value, two minutes after a steam flow change. .  ;.1 Plant performance during the two . load reductions was as anticipated.a The load rediction at the 75Z power level' test plateau was from 772'toi31%~

(Reduction - 540 mwe), and at 1002,' from 100% to 48% (Reduction = 620 mwe)..v ,

e  :

5-29 ,

1

,[ . 5.8 ST-35. LARGE LOAD REDUCTION (Continued){

During the load reduction from 100Z, feedpump suction pressure during--

the ' transient ' fell below- the feedpump trip setpoint, - but no trip occurred.

The trip circuit,- which utilizes 2 out; of 3 logic, was. examined af ter the

. test, - and the - failure determined to have . resulted from two. problems: the-trip setpoint was adjusted to below'the desired point, and an incorrect head pressure _' correction was used. Necessary changes in the adjustment' process i have been made to prevent a recurrence and the setpoint is being lowered.

~

A ' test exception was written to address the fallure to ' trip. Test- l results areigiven in Table 1. l I

-l 1

1 i

N

' Y, l

l l

J l

l l

1 1

~5-30  :

a-*

2

• ~

4 -'q. .,/

1 5.8 ST-35. LARGE LOAD REDUCTION (Continued)

TABLE 1 ST-35 LARGE LOAD REDUCTION Measured Values' Acceptable 1 '

Parameter 75% -100%- Deviation Primary Pressure Swings (from initial value). +30psig/-93psig. +69psig/-99psig +100/-150 psig SG (NR) Level $

+09%/-11% +12%/-10%.

• 25% '

TAVG undershoot below 4 final, steady state value None .None 3*F ^I TAVG Peaked above initial value- 3*F 3'F- 8'F  !

TAVG oscillations during Approx. Approx. No '

steam dump operations 0. 5 ' F, 0.5'F Oscillations .;

I TAVG oscillations after 5*F.

steam dump operation ended None None'  ;

Peak / Valley- ,

TAVG within l'F TREF after No manual ,

transient < 0.8'F- J< 0.5'F intervention '

With TAVG restored to l'F of TREF, auto control maintains Approx.- Approx.

TAyo with respect to TREF.

• 0.5'F i 0.5'F' i 1.5'F. -i Automatic rod control inserts Approx. Approx.. *Approximately rods at maximum speed 70 secs. 80 secs. 30 seconds =

Steam dump system Yes Yes Did not cycle on and off i Duration of steam dump system operation *Approximately-

.7.3 min. 12 min.. 8 minutes '

• • <

• 3% final Feedwater pump discharge ,

pressure oscillations value, 2 min,- '

1.22. 2.0% after change i Notes Parameters the last item.in Table 1 require vendor notification only, except for

• Westinghouse notified

• • Acceptance criterion-5-31 i

', . 1. ST-37. MOISTURE CARRYOVER MEASUREMENT i i

'Obiective l

Steam ' generator moisture carryover was determined- in this; test by -i accurately measuring the moisture . leaving each steam generator after:

injecting r sonradioactive chemical tracer, lithium-6, directly into the , 'i steam gene cors via'the main feedwater system.

The moisture carryover measurement is described in FSAR, Chapter 14,.

Table 14'.2-5, Sheet 40.

1 Discussion The traditional method for measuring moisture carryover utilizes 'a short-lived radioactive tracer. Na-24. The problems associated :with.

trans port , i handling , and exposure to a high level radioactive? source ' are eliminated using a nonradioactive isotopic' tracer (11thium-6 hydroxide mono-hydrate). . Isotopic ' dilution mass- spectroscopy :was used -by ! Combustion-Engineering (CE), under contract, to - measure the concentration of Li-6 carried over.

A valid or representative . sample of moisture carryover - cannot abe ' '

obtained from the main steam line sampling nozzles, as moisture droplets tend to flow along the walls of the steam lines. Isokinetic sampling.

nozzles =are not normally available, and were not in thisl instance. In ST- l 37, samples were drawn from a common header in the feedwater. line, and by proportioning the feedwater tracer concentration between the -steam.

i generators,-using the individual steam line samples, the performance of each.

steam generator was estimated.

A temporary modification was made to permit j individual sampling of the main steam' lines.

f i The tracer was injected af ter a preconditioning purge 4 of: sample lines -

for twelve hours, and a one-hour stable power. condition. 'After a half-hour-valt to ensure an equilibrium condition, samples were taken from each S/G blowdown sample point. The blowdown sample from each generator was required to yield at least 5 ppb. lithium-6, prior to general sampling.- 1 Six sample sets were drawn from each S/G-blowdown line, each main steam  ;

line and the conunon main feedwater line. The samples were; transferred to CE analysis. -

t i.

.I 2

5-32 F

a-

'x.,

'6 g,

• 5.9 ST-37. MOISTURE CARRYOVER MEASUREMENT'(Continued)

Results The acceptance criterion was met; moisture . carryover of;- each steam generator has been calculated by CE and determined to befs.0.252.

r In Table 1,. the sample ' concentrations ~and' moisture carryovers reported by CE are given.

TABLE 1 ST-37 MOISTURE CARRYOVER RESULTS

~

Sample 1- Sample 2

'S/G A Concentration -

. Blowdown 17.425 ppb 16.715 ppb Mainsteam 0.004; ppb 0.003 ppb Moisture Carryover - 0.03t! 0.01 2.

S/G B Concentration -

• Blowdown  !

16.769f. ppb '16'049 ppb' Mainsteam 0.008 ppb =0.008 ppb Moisture Carryover - 0.06-2 0.03 %

S/G C Concentration -

Blowdown 17.882 ppb -!

16.912xppb '

Hainsteam 0.004 ppb 0.001 ppb Moisture Carryover - 0.03 I 'O.00 %

S/G D Concentration -

Blowdown 14.686 ppb. 14.396 ppb Mainsteam ~0.002 ppb Moisture Carryover - 0.003 ppb 0.02 24 -0.01.2 3

Main Feedwater Concentration - 0.007 ppb. 0.003 ppb  ;

There were no test exceptions. i

.i

o

., 5-33

5.10 ST-38. UNIT TRIP FROM 1001 POWER Obieetive The procedure demonstrated proper plant response to a trip from 100!

power, and verified that the actual overall hot leg resistance temperature detector (RTD) response time is conservative with respect to the value used in the accident analysis.

The test is described in FSAR. Chapter 14. Table 14.2-5. Sheet 41.

Discussi.2E With the plant in steady state operation at the 2002 power level test plateau, a unit trip was initiated by manually opening the generator breaker from the main control board (MCB). This action caused the main generator breaker to trip open and the turbine to trip with a resultant reactor trip. f Prior to test initiation, the following systems were placed in automatic:

• Steam Generator bevel Control e

Feedwater Pump Speed Control l

Steam Dump Control (T Ayg Mode)

+ ASDVs (Set Point 1125 psig)

Pressuriser Pressure Control (Set Point 2235 psig)

The UATs were aligned as the source of power for the onsite distribution i system.

Results All acceptance criteria were mets i

• Pressuriser safety valves do not lift I Steam generator safety valves do not lift l

• Safety injection was not initiated OverallhotlegRTDresponsetime*k6.7 seconds Heasured Response Time = 5.5 seconds.

• The interval of time measured between the taint where the i neutron flux has decreased by 502 from its initial value to the point where the hot leg temperature signal has decreased by one-third of the initial loop delta-T value (*F).

5-34

A S.10 $T-38. UNIT TRIP FROM 1001 POWER (Continued)

Plant performance during the test was generally as expected.

Reptesentative system behavior during the transient is shown in Figures 1 2, 3, 4 and 5.

Two operational responses requiring evaluation were identified:

A P.14 feedwater isolstion signal occurred due to S/G narrow range level spiking high. The action caused the feedwater regulating valves to rapidly close, causing a number of feedwater heater -

relief valves to lift. ~A design . change to address the. problem 'is in preparation.

When transferring the steam dump system to the steam pressure mode from the TA yg mode, a 352 demand was observed which lasted 5 seconde. A procedural change has been initie t.e6 to ensure a bumpless transfer.

In addition, the following operational response was reported to-Vestinghouse Pressuriner pressure dropped:31 psi below the expected minimum of 2000 psig.

5-35 i

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, . 5.11 ST-39. LOSS OF OFFSITE POWER  !

Obieetive t

The Loss of of fsite Power Test demonstrated response of the plant and  ;

emergency electrical power system to meet design performance requirements  ;

under the condition of a loss of all offsite power, coincident with the loss of the main generator. Also, stable, shutdown maintenance of the reactor plant under natural circulation conditions with a loss of offsite power was t l

shown.

I Discussion ,

The loss of offsite power was initiated, from a power level just above the P-9 setpoint (approximately 20! RTP), by simultaneously tripping the turbine generator and opening a 345 KV control breaker. The incoming supply breakers for the reserve auxiliary transformers - (RATS) had been ';

placed in ' pull-to-lock' to prevent an automatic transfer. A. reactor trip resulted and the reactor coolant pumps tripped on underfrequency.

r The test method permitted bus voltages to decay and be sensed by all of I the various undervoltage protection circuits. First level undervoltage '

protection (less than 72% rated voltage for > 1.2 seconds), caused emergency power sequencer (EPS) actuation, which then initiated an emergency start of the diesel generators.

The emergency power sequencers (EPS) sequenced the required loads onto the diesel generators. The uninterruptible power sources (UPS) shifted to  ;

the DC backup supply with the trip and loss of offsite power (LOP), and then back to AC from the diesel generators.

During recovery, pressurizer backup heaters, auxiliary spray and-PORVs ,

were available to control primary pressure. Atmospheric steam dump valves i (ASDVs) were available, but not the steam dumps to the condenser.

Results All acceptance criteria were met. The criteria and related  ;

performance are given as follows: ,

• First level undervoltage protection causes the start of both '

emergency diesel generators, and both emergency power sequencers i (EPS) perform their automatic sequencing correctly, as recorded on Form ST-39A.

Main plant computer system (MPCS) logger data is used to complete the above form, which requires.the start time for tie diesels and sequencing steps. In the two-second period following test initiation, the MPCS degraded which caused the prime Lost to fail over to -the backup host. During the fallover (2 6 minutes depending on the MPCS feature) diesel and sequencer innrmation was lost or unreliable.

5-41 r

)

H

, . '5.11 ST- 3 9. LOSS'0F OFFSITE POWER (Continued)

The possibility of . f ailover was anticipated and, as a failover f contingency, a trained observer with a calibrated stopwatch had i been stationed at the MCB electrical section tc visually verify startup of the diesels and ' the sequencer steps. The diesels [

started and sequenced essentially together: the bounding start time for the diesels was 8.44 seconds and panel lights verified that all sequence steps occurred in order.

A test exception was written to cover the loss of MPCS data during the failover. +

+ The diesel generator starting times are less than 10 seconds. t Maximum starting time was 8.44 seconds.

i

• Satisfactory operation without offsite AC electrical power was achieved.for a minimum of 30 minutes.-

The plant was tripped at 0941 hours0.0109 days <br />0.261 hours <br />0.00156 weeks <br />3.580505e-4 months <br /> and recovery began at 1014 hours0.0117 days <br />0.282 hours <br />0.00168 weeks <br />3.85827e-4 months <br />.

• Evaluation of recorded data and plant responses confirm proper

• dynamic system responses resulting f rom a loss =of offsite AC electrical power.

Proper dynamic system response was confirmed. Representative plant response is shown on Figures 1, 2 and 3. Plant responses which were not anticipated are noted below.

Vital Busses 1A, 1B and 1C properly shif ted to the DC backup supply, but did not automatically shift back to AC when the diesels had powered the  :

emergency busses (E-5 and E-6) . This is a characteristic of . these units. -

The inverters were manually transferred.

• DC Voltage and current indications on 14CB were lost. An' investigation f of the problem disclosed that the transducers feeding the indicators are AC i supplied. An engineering evaluation has_been requested. l The current limit circuit of non-vital Battery Charger 2A' failed early '

into the test, causing the Bus 12A supply breaker from the charger to overload and open. As a result, Bus 12A voltage decreased from approximately 150 VDC to 120 VDC due to the load of the DC powered bearing,  !

seal and MFP lube oil pumps. The DC pumps were no longer required when the t emergency busses were powered, and the voltage rate of decrease slowed' significantly. With the restoration of offsite power, and no current limit, the bus was recharged without problem. ',

The Control room non-vital lighting inverter did not operate and was found to be. misaligned. Loss of the inverter did not affect operator-performance during the test. A procedure upgrade will correct the problem. '

s 5-42 i

i

• $,11 ST-39. LOSS OF OFFSITE POWER (Continued) ,

The pressuriser Group A backup heaters could not be manually re-energized from the MCB following Remote Manual Operation (RMO) reset._ A '

minor wiring discrepancy was found to be the cause and was corrected.

r During restoration of offsite power, a Train A RAT closing circuit inop  :

alarm was received. A blown secondary PT fuse was the cause. l i

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, .'5.12 S't-40. NSSS ACCEPTANCE TEST Obiective ]

)

The Nuclear Steam Supply System (NSSS) test demonstrated the I reliability of the NSSS by maintaining the plant, at 1002 (+01/-52) reactor i power, for 250 hours0.00289 days <br />0.0694 hours <br />4.133598e-4 weeks <br />9.5125e-5 months <br /> without a load reduction or a plant trip resulting from l an NSSS nelfunction. Also, the'NSSS output was measured at (or near) its I warranted rating. I l

The acceptance test is described in FSAR, Chapter 14 Table 14.2-5 I Sheet 43. J l

Discussion l i

To demonstrate acceptable .NSSS performance, the plant was required to I maintain 952 to 100% of full power for 250 hours0.00289 days <br />0.0694 hours <br />4.133598e-4 weeks <br />9.5125e-5 months <br />. Full power rated l conditions were stated as:

• Thermal output of NSSS: 3425 Hwt Thermal output of reactor core: 3411 Hwt i Steam flow from NSSS (No blowdown): 15,140,000 lb/hr

• Steam generator outlet pressure: 1000 psia e

Assumed feedwater temperature 440'F

• Maximum moisture content 0.252 The warranted NSSS thermal output was 3425 Hwt, and the acceptable performance was-982, or 3357 Hwt.

On an hourly basis, during the 250 hours0.00289 days <br />0.0694 hours <br />4.133598e-4 weeks <br />9.5125e-5 months <br />, power level data was taken using the NIS power range channels, station recorder charts and data log printouts. After approximately 60 hours6.944444e-4 days <br />0.0167 hours <br />9.920635e-5 weeks <br />2.283e-5 months <br />, the performance measurements were taken, to verify the warranted thermal output. Four. sets of hourly data were taken, with blowdown secured, for the precision calorimetric station '

procedure.

Results The acceptance criteria were mets plant operated at 1002 (+02/-52) for 250 hours0.00289 days <br />0.0694 hours <br />4.133598e-4 weeks <br />9.5125e-5 months <br /> without a load reduction or plant trip resulting f rom a NSSS malfunction, and NSSS thermal output as determined by a performance measurement, is 2 3357 Hwt.

Performance measurements yielded the following:

Average reactor power level from calorimetrics = 99.782 Power deviation during performance measurements = 0.211

+

Measured NSSS thermal output * = 3422 Hwt

• Includes 18.4 Hwt net input from the RCS 5-47

1

(

5.12 ST-40. NSS$ ACCEPTANCE TEST fContinued) i The test began on August 5, 1990 at 1700 hours0.0197 days <br />0.472 hours <br />0.00281 weeks <br />6.4685e-4 months <br />, and ended on August 17, 1990 at 1800 hours0.0208 days <br />0.5 hours <br />0.00298 weeks <br />6.849e-4 months <br />. The test was interrupted for two hours on August 6 for  ;

stop and control valve surveillance tests at a power level of < 951, and for '

approximately 1 days . on August 13 for repair of a leak in an EHC line.

Power was reduced and' the turbine tripped for the latter interruption. On August 16 a steam leak in a S/G blowdown line required isolation of the flash tank, but the line was repaired without reducing' power.

There were no test exceptions. ]

e i

k i

i 5 .

,' ,.5.13 ST-48. TURBINE GENERATOR STARTUP TEST Obinetive The objective of the turbine generator startup test was acquisition of baseline operating parameters for the turbine generator and . associated components, and operational data at each of the power level test plateaus for evaluating unit performance.

This procedure is described in FSAR, Chapter 14. Table 14.2-5 Sheet $1.

Discussion The turbine generator startup test demonstrated the following:

The loss of primary or backup speed signals will not trip the turbine, but loss of both speed signale causes a turbine trip.

The Backup Overspeed Trip and Emergency Trip circuits function as designated.

• The turbine-generator is capable of operating at various loads without exceeding any manufacturers' design limitations.

GE Startup Engineers assisted PAT personnel.throughout the test, and on occasion requested that additional measurements be made, or extended the time for gathering data. As was - reported in Supplement 1 Startup Test Report, ST-48 was interrupted, after rolling the turbine and validating protective systems, to conduct ST-48.1. Turbine Generator Torsional Response Test. An undesirable resonance was found, necessitating a month-long PATP interruption for modifications to the 'C' low pressure turbine. i l

No-load data was recorded for turbine' steam conditions,. lubrication and control systems, and generator parameters.

Following initial synchronization and overspeed testing, the load was increased, and at selected power levels, subsystems of the turbine generator such as EHC and Alterex were adjusted and operation of protective systems verified.

Turbine generator tests were conducted at the 30%, 50%, 751, 90% and 1001 power level test plateaus. At each level, steady-state data was collected: where necessary, power load unbalance (PLU) checks, underexcited reactive ampere limit (URAL) checks and trip tests of the thrust bearing wear detector (TBWD) were included.

Results The acceptance criteria were met's the turbine generator has been synchronized to the grid, and all required operational data has been.

collected and evaluated as satisfactory.

5-49

, $.$3 ST-48. TURBINE GENERATOR STARTUP TEST (Continued) ,

In the initial tests, af ter ' no-load data was taken and initial synchronization and overspeed testing completed, some corrective measures were taken to address problems identified. These included ' backseating intercept valves (IV), readjusting the EHC speed error signal, resolution of thermal expansion concerns, repair of the mechanical overspeed trip, and installation of flow orifices in the fast acting solenoids of the EHC trip circuit.

At the 502 and higher power level test' plateaus, the required turbine generator performance data was collected, and where indicated PLU, TBWD and URAL checks made.

Eight test exceptions were taken: nine RTDs embedded in generator windings are defective, but do not prevent adequate monitoring of winding temperatures: six exceptions were written against inoperative instruments or readouts and one excepted trend data which was misplaced.

1 l

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'l 5-50

6,0 INSTRUMENT CALIBRATION AND ALIGNMENT i

Contents:

6.1 ST-13 Operational Alignment of Nuclear Instrumentation 6.2 ST-14.1 Operational Alignment of Process Temperature .

Instrumentation 6.3 ST-15 Reactor Plant System Setpoint Verification

• l 6.4 ST-26 Thermal Power Measurement and- Statepoint Data ,

Collection i

6.5 STa27 Startup Adjustments of Reactor Control ^ System ,

6.6 ST-28 Calibration' of Steam and Feedwater Flow Instrumentation 6.7 ST-36 Axial Flux Difference Instrumentation Calibration

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4 6-1  :

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A  ;

• 6.1 ST-13. OPERATIONAL ALIGNMENT OF NUCf m INSTRUMENTATION

}

Obiective The objective of this procedure was determination of various voltage, trip, alarm, operational and overlap settings for the source, intermediate i and power range nuclear instrumentation. Portions of ST-13 are performed at I a number of power levels and at each of the power level test plateaus in Power Ascension Testing.

The procedure is described in FSAR, Section 14. Table 14.2 5, Sheet 16. l l

Discussion Calibration of nuclear instrumentation, including alarm settings, trip points, and operational ranges, cannot be properly completed until the system is functioning on line, in its intended operational ranges. At each test condition, the nuclear instruments are adjusted,.using.the'best available conservative information. Initially, setpoint data furnished by Westinghouse was used, and, as higher neutron' fluxes became available, these values were superseded by actual measured data.

Test conditions and the adjustments made during ST-13 were Prior to Criticality -

• IR Channels high flux' rod stop bistable and high flux bistable.

• PR Channels scaled for. total full power using 400 .microamps. ,

verified / scaled f(delta-I) summing amplifiers, scaled ' AFD (from Reactor Engineering).

Approach to Criticality -

e MCB shutdown monitor teot ,

102 to 152 Power -

• PR Channels - Indications checked against heat balance.-

302 Power Level Test Plateau -

• IR-PR overlap data (from ST-26)-

PR Channels, total full power detector currents by extrapolation using calorimetric power data (from ST-26).

502 Power Level Test Plateau -

• IR-PR overlap data (from ST-26)

PR Channels, total full power detector currents by extrapolation using calorimetric power data (from ST-26): rescaling using AFD data (from ST-36).

• Gammametric Channels, calorime d e adjustment Note: Gammametrics is the trade name for the post accident excore detectors.

75% Power Level Test Plateau - I Same as 502 without Gammametric adjustment l

6-2

.

• 6.1 ST-13. OPERATIONAL ALIGNMENT OF NUC1 h8 INSTRUMENTATION fContinuedi 1002 Power Level Test Plateau -

• IR and PR saturation curves,'IR-PR overlap .

9

+ PR Channels, total full power currents by measurement. J

• IR' Channels, bistable adjustment if required i

+ Gansnametric Channels, adjustment Shutdown from 1001 Power In conjunction With (ST-38) - l

• SR Channels, saturation and integral bias curves j

• IR Channels, compensating voltage and bistable adjustment >

+ .Gansnametric Channels, discriminator adjustment, if required.

Results All acceptance criteria were met with certain exceptions (indicated by asterisks):-

, . i

• Shutdown Monitor Alarm setpoints were 1.5 times the previously recorded countrate i 10!.* .

. Shutdown monitor countrates at alarm were equal to the previous alarm

• 102.* .

. Overlap data was obtained between IR and PR channels at . the ,

individual test plateaus.

• Plots of PR channel total detector currents vs calorimetric power- '

exhibit linear response from 0% to 1002 power.*

• Final operational settings have been documented for SR IR and PR ,

channels and meet the range limitations of T.S. 3.3.1 and the Westinghouse NI Manual.

Gammametric detectors have been adjusted at the 100% power level ,'

test plateau.

The failure to meet acceptance criteria for the shutdown monitor resulted from testing requirements which did not simulate . actual plant  ;

conditions for normal service. A detailed discussion of the test, conducted .

during the initial PAT criticality, is given in Supplement 1 to the Initial l Startup Report. A test exception was taken. ,

Per procedure, following the shutdown f rom 1002 power (ST-38), the- q compensating voltages and bistable alarm settings for the IR channels were to be adjusted. The adjustments were not successful, and were attempted a second time following the loss of offsite power test (trip from 202 power.

ST-39). Westinghouse was consulted after the first attempt . . . and again, after the second failure to adjust. Since the . P-6 permissive function (Energitation of SRs) by the irs, and proper SR to IR overlap was observed during the two trips and subsequent startups, the instrument settings were left as is, and a test exception taken.

1 6-3

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7 1

,' , 6,1 ST-13. OPERATIONAL ALIGNMENT (Continued)

Figures 1, 2, 3 and 4 ' are graphs of the PR channels versus reactor i power. A linear behavior was observed until the 2001 power level test plateaus at 1001, a lower value of current for each detector was noted than would have been expected. The non-linearity was attributed to adjustments made in the turbine impulse pressure and TAVG program at 751 power -($7 27)  ;

which caused T A yo to be lower than expected at 1001 reactor power. A test- i exception was taken.

i l

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f l-6-4

c .-

TOTAL DETECTOR CURRENT VS RX POWER ST-13 N41 (100% RTP) FIGURE 1 800 T

0 700 -

A ,

000 -

C II A

N 500 -

N E

L i

<100 -

C /

U /

R 300 -

E N /

T 200 -

/

! I U '

u In 100 -

P s

0 O 20 40 60 80 100 REACTOR POWER (%) ,

6-5 -

1 e

i TOTAL DETECTOR CURRENT VS RX POWER j ST-13 N42 (100% RTP) FIGURE 2 1

800 --

l T _

9p 700 -

A L

000 -

C 11 A

N 500 -

N E

L 400 -

C -

U h300 - ~

E N

T 200 -

+/

u a

ni 100 -

P s

0 ' ' '

0 20 40 00 80 100 REACTOR POWER (%)

l 6-6 ni..ama i dumi l

TOTAL DETECTOR- CURRENT VS RX. POWER ST-13 N43 (100%- RTP) FIGURE 3 000 T

0 700 -

4 4

A L

600 -

H =

A N 500 -

N E

L 400 -

C U '{

g 300 -

E N ~.

T 200 -

u a

nT 100 -

P s /

0'l I O 20 40- 00 80 100 REACTOR POWER (%)

6-7

u TOTAL DETECTOR CURRENT VS RX POWER ST-13 'N44 (1007. RTP) FIGURE 4 800 -

T Pp 700 -

g a / -

L 000 -

C 11 ,

A N 500 -

N E

L 400 -

C '

U 5

jf300 -

E N e T 200 -

11 a

m

~

100 -

P s

OL '

O 20 40 60 80 -100 REACTOR POWER (%)

6-8

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f . 5.2 ST-14.1. OPERATIONAL ALIGNMENT OF PROCESS TEHPERATURE INSTRUMENTATION )

Obiective j The objective of the procedure was proper alignment.of the delta-T and TAVG instrumentation channels at all test plateaus.

The procedure is described in FSAR Section 14, Table'14.2-5, Sheet 17. I Discussi2a

' Initial alignments were carried out during Low Power Physics Testing.

With the plant in the hot standby condition and temperature stabilized, the THOT and TCOLD R/E Converters, and the T Ayg and delta-T Summing' Amplifiers were aligned using station procedures. Temperatures. were then measuredt :

values converted to corresponding engineering units; and temperature and .

delta-T values calculated. Results; were checked against- allowable tolerances.

• Further alignment of the process temperature instrumentation was-dependent on data obtained during the performance of ST-26 Thermal Power Measurement and Statepoint Data Collection. At power level test plateaus of 301, 502, 752, 902 and 1002, the following data was obtained from ST-26: i e Total reactor power (2) from the calorimetric analysis.

l

• THOT and TCOLD (*F)-from the operating and installed spare RTD's.

• TA yg ('F) from the summing amplifier output. -

• Delta-T (!) from the sunning amplifier.

The data was then used to calculate . the THOT and TCOLD difference,.

TAVG, and delta-T A yg for all four loops. If values obtained failed to meet specified tolerances, then corrective actions were taken using appropriate IEC Department procedures.

At the 752 power level test plateau, the~ fluid specific enthalples for each of the TH0T/TCOLD temperatures were determined for each of the four loops at a nominal RCS pressure value of 2235 psig. The enthalples were correlated with the calorimetric values of power and. extrapolated to 1002 -

power. From these results, extrapolated full power delta-T and TAVG values were calculated and the delta-T summing amplifiers rescaled.  ;

The procedure. provided for additional rescaling, as necessary. at the 90% and 1002 power level test plateaus. l P

6-9

, 5.2 ST-14.1 OPERATIONAL ALIGNMENT OF PROCESS TEMPERATURE INSTR. (Continued) l Results l All acceptance criteria (applies to full power results only), were mets.  ;

e The TAVG (summing amp) from each channel is within

• 0.5'F.of the value calculated from the TH0T and TCOLD R/E. converter outputs. l

• The delta-T (2) from each: channel is within i 12 of the l calorimetric power.

4

• The Tyor and TCOLD from the R/E converters are within 't 1.2'F of the installed spare RTD values. .

The measured full-power average core delta-T was' determined to be  !

56.2*F. r Process temperature parameters which were found to be out-of-tolerance,- l i

and the actions taken, are listed in Table 1.

TABLE 1 e ST-14.1 OUT-OF-TOLERANCE CONDITIONS  !

Power Lvl Parameter / Problem Value Tolerance Action  ;

302 Loop 2 (NR) TCOLD None - 0 Difference -1.636*F i 1.2'F. Test'Excep-Loop 2 (NR) Calculated tion Written Delta-TAyo 0.961'F i 0.5'F 302 Results*. ']

502 None  ;

752 Loop 3 Extrapolated .

Rescale TAVG I Full Power TAVG 589.62'F 5 588.5'F and Turbine

= Impulse Presy sure (per _. l ST-27) Test i Exception Written: .

I 902 Loop 2 Calorimetric Power _

- Delta-T (2 Power) -1.1252

• 12 Rescaled_ J Loop 4 Calorimetric Power  !

- Delta-T (2 Power) -2.5422 i 12 Rescaled 1 1002 Loop 2 Calorimetric Power _

- Delta-T (2 Power) 1.1542 1 12 Rescaled l Loop 3 Calorimetric Power

- Delta-T (2 Power) 1.1022 1 12 Rescaled

• Difference not observed at higher test plateaus.

6-10

. .6.3 ST-15. REACTOR, PLANT SYSTEMS SETPQINT VERIFICATION Obiective i J

Procedure ST-15 provides verification that the initial setpoint adjustments have been nede prior to startup, and serves to document setpoint modifications made during startup testing.

This procedure is described in FSAR, Section 14, Table 14.2-5, Sheet 18.

Discussion i

Initial setpoints were verified for the following plant components and systems:

1. Safeguards

2. Reactor Coolant Pumps '

3. Nuclear Instrumentation (Excore) -

4. Delta T - TAVG

5. Precaurizer Pressure

6. Cold Overpressurization Mitigation System (COMS)

7. Pressurizer Level

8. Charging Flow

9. Rod Control

10. RCS Flow

11. Feedwater Flow

12. Steam Generator Level

13. Steam Line Pressure

14. Steam Dump System

15. Steam Generator Relief Valve Control (ASDVs)

16. Turbine Impulse Chamber Pressure  !

Prior to criticality, ST-15 served as a detailed listing of setpoints for plant instrumentation to be compared to the Westinghouse Precautions, Limitations and Setpoints (PLS) values, and Technical Specifications.

During Low Power Physics Tests, had any setpoint changes been required, '

these would have been documented by this procedure. No - changes were required.

All setpoint changes made during Power Ascension Testing were documented by.ST-15, and the final PAT requirement prior to the 1002 power

~

plateau results review was a verification that any changes made were incorporated into both the plant hardware and station procedural text.

Results The acceptance criterion was met all setpoint changes during PAT were documented in the procedure.

A total of 17 setpoint changes were necessary and were entered into ST-15: of those, 7 changes were later superseded by test results at higher power level test plateaus.

6-11

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' 6.4' $7-26. THEkMAL POWER MEASUREMENT AND STATEPOINT DATA C01120TIg, .

Obiective The. objective of this ' procedure was a calorimetric - determination of reactor power, and verification of main steam and feedwater performance ~from various primary and secondary process data.

The procedure is described in FSAR, Section 14, Table 14.2-5 Sheet 29.

Discussion i

Primary and secondary. process parameters were measured, using station l procedures, and from these data a calorimetric determination of power was .j made. Calorimetric determinations of ' thermal power were made at the-302, l 502, 75I, 90% and 1002 power level test plateaus.

ST-26 specified the following stability requirements for a calorimetric. i power level determinations i

i e

RCS temperature (T A yo) changing less than l'F/ hour.

'l Core power (power range) changing < 0.5Z/ hour.

Steam generator water level'at 502 (48-52%).

Pressurizer pressure at 2235 psig (2210-2260 psig).

]

Pressurizer level +0/-22 of programmed level.

• Blowdown secured.

+ Charging and letdown flow constant. .)

This procedure rarved as - the data gathering procedure for a number of the '

instrumentation calibration and alignment' procedures.

Operation of main steam and feedwater systems were observed-throughout-the entire test program and demonstrated satisfactory performance.

4 Results

• All acceptance criteria were met.

CETARS.

At hot zero power (HZP) an initial set of process data was taken'using "

( At each of the power level test plateaus, data was collected and analysis performed using station procedures RN 1730, ' Precision' '

Calorimetric /RCS Flow Rate Measurement, and RN-1731, Secondary Heat Balance.

A problem was encountered at 302 power with the method for locally determining feedwater temperatures a wiring change in the test instrumentation circuitry was implemented to correct the problem.

i 6-12 i

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• 6.4 ST-26. THERMAL POWER MEASUREMENT AND STATEPOINT ~ DATA C0112CTION 'l (Continued) i Tho ' calculatedL RCS flow rate. (required : only at the-502 power-level

-test plateau) was RCS Flow Rate Measurement Surveillance = 416,771 gpm Technical Specification Limit.= > . 391.000 gpm Calculated results based on the collected data were consistent with  ;

plant conditions at the time. There were'two test exceptions: both were of -

a minor procedural nature.

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i f' , 6.5 ST-27. STARTUP ADJUSTMENTS OF THE REACTOR CONTROL- SYSTEM Descriotign The procedure. obtained and evaluated the data necessary to determine the TAVG program that will result in the highest possible steam pressure to assure optimum plant efficiency, while maintaining pressure for_ the turbine I and TAVG within requirod lindtations. The ' system pressure and temperature.

data provided a basis for adjustments of the reactor control system (TREF on the T Ayo programming module). ]

The procedure is described in FSAR, Section 14, Table 14.2-5,-

Sheet 30. 1 Discussion The following data, obtained from ST-26 Thermal' Power Measurement and Statepoint Data Collection - at hot zero power, 302 power and all subsequent power level test. plateaus, were used in this procedure:

Loops 1,2,3,4 - THOT TCOLD and TAvo.

SG Pressure Calorimetric Reactor Power Turbine Impulse Pressure The data were averaged and used to generate graphs of:

RCS Temperature vs Power S/G Pressure vs Power Turbine Impulse Pressure vs Power At the 502 power level test plateau, extrapolated full power values, obtained- f rom the graphs, were compared- to: the design full power . steam generator pressure and' the design (or' current span) turbine impulse pressure. If'the allowable tolerances (* 10 psi and i 50 psi, respectively),

were exceeded, the turbine impulse pressure and full . power TAVG ivere adjusted to the extrapolated values. The process was repeated' at each ,

subsequent power level test plateau (75Z, 902 and 1002). 1The allowable -

i tolerance for turbine impulse pressure, at 1002,.was i 10 psi. .

Results The acceptance criteria were. met l

• The final full power steam generator pressures are 985

• 10 psig.

l

+

Thegfinal full power TREF _does not exceed 588.5'F.

i -

No-adjustments at the 50Z power level test plateau were required. Full' power extrapolations of average S/G pressure 'and turbine impulse' pressure were within required tolerances.

6-14 l

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6.5 ST-27. STARTUP ADJUSTMENTS OF THE REACTOR CONTROL SYSTEM (Continued)

At the 752 power level - test . plateau, full power extrapolated - values j were:

• RCS Temperature (Average TAVG) = 590'F' s S/G Pressure = 1035 psia

. Turbine Impulse Pressure - 730 psia and rescaling of turbine impulse pressure and full opower. T; AVG Program- was ]

necessary.

At 902 no adjustments were required.  ;

l The moisture. separator-reheaters (MSR) were:placed in service prior to I testing at the 1002 power-level test plateau.:which resulted in the turbine ,

impulse pressure being reduced f rom ' the full power scaling . of 730, ps  ;

-(based on- operation without MSRs). Rescaling of turbine impulse pressu,la re ;

and the full power TAyo-program resulted in final values of:

Full Power TAVG = $87.5"F

. S/G Pressure, A = 988.5 psig; B = 984.5 psig '

C = 990.4 psig D = 986.0 psig' J

. Turbine Impulse Pressure =:691.4 psia

. S/G Saturation Temperature =~544.8'F t

~

Graphs of RCS Temperature S/G Pressure' and ' Turbine ImpulseLPressure -

versus Reactor Power are given in Figures 1, 2 and 3 respectively..

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6-15 1

.y RCS TEMPERATURE VS POWER  ;

ST-27 FIGURE 1 600 -

595 -

R C 590 -

S  :

T 585 -

E-M p 580 -

=

E R

A 575 -

T U

/

=

R 570 -

E I 565 -

G /

5GO -

F u.

555 -

l- DESIGN TAVG

-550 ' '

0: 20 40 60 ' 80 100' REACTOR POWER (%)

-t 6-16 j

SG PRESSURE; VS POWER-ST-27 FIGURE 2-1200

.1190 -

1180 -

1170 -

1160 -

1150 -

S 1140 - '

G 1130 -

p- 1120 -

R 1110' -

E 1100 -

S S 1000 -

U 1080 -

It 1070 -

l E

1060 -

P 1050 -

S 1040 -

I A

1030 - -

El I,

1020 -

990 -

l DESIGN CURVE 980 -

1 970 ' ' ' ' ' ' ' ' l 0 10 20 30 40 50 60' 70- 80 90 100 j REACTOR POWER (%). j i

y l

6-17 j

• f TURBINE IMPULSE: PRESSURE VS PWR  !

ST-27 FIGURE 3- '

000 750 -

T U 700 -

R -

=-

D G50 -

I

-N 600 -

550 - -

I 500 -

+ r y

450 -

h L 400 -

S E

4 350 -

P 300 -

R E

S 250 -

g[

S 200 -

P 150 -

S .

I 100 - '

A  ! DESIGN CURVE 50 -

0

' ' ' ' ' '~ i

'O 10 20 30 -40 50 60 70 80 90 100 REACTOR POWER '(%)

6-18

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' e6.6 ST-28. CALIBRATION OF STEAM AND FEEDWATER FLOW INSTRUMENTATION 4

l Obiective J l

,The objective - of this test ; wasi calibration of. the main steam' flow transmitters based on feedwater flow measurements, q

~

l This procedure.is described in FSAR Section 14. Figure.14.2-5, . '

Shee t 31. - 'i Discussion l

The -procedure utilized the installed station feedwater?. flow- -

instrumentation as the measured parameter of steam flow to empirically, determine values used to- calibrate the main steam. flow transmitters. Data -

i f rom ST-26 Thermal Power Measurement. and Statepoint Data Collection, -was used to determine ~ steam flow and new calibration values. ST-28 was an-ongoing startup test data was collected at each test plateau: l At hot ~ zero power - - Main - steam' flow transmitter calibrations were -

performed to monitor lany zero' shift changes from cold conditions.

At each power level test plateau - Steam and feed' flow calibration. data- j were collected at the 302, 502, 752, 902 and 100Z power level test '

plateaus.

l At 'the 1002 power level test plateau - Steam , flow and ; feed e flow l

transmitter outputs were compared; steam flow output was corrected to ,

l agree with feed flow within the required accuracy.

Results The acceptance criterion was met: 'the steam ' flow transmitter outputs -

l for each channel have been matched to the associated feed flow transmitter-l outputs to within i 100,000 pph (* 22).

The original-acceptance criterion, i.25,000 pph'(*'0.5%), could not be '

met, and the higher tolerance was introduced by procedure' change af ter review of the basis for the initial value with Westinghouse... Theireview of j

, the basis for the i 0.51, determined that the . value was -derived . from a o li safety analysis performed for a Hi Steam Flow . Rate Circuit - protective i function, utilized at some other Westinghouse plants. Seabrook does -. not - si l- utilize this protective function.. -The Seabrook system uses - steam flow!  !

values along with feed flow to produce a mismatch error' signal for S/G 1evel control circuits. The i 22 acceptance limit was recommended to, replace the

• 0.52 value.

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- ' -. 6.6 ST-28. CALIBRATION OF STFM AND FEEDWATER FLOW INSTRUMENTATION i (Continued)

'At each . power level test plateau... for ' each loop., the~ steam flow ,

1 transmitter data, for both transmitters .in' thei same loop, wasf required to agree within 1 12.(* 50.000 pph). The steam flows'did not.always meet these limitse.

-Power Level- Loop' Flow Difference-30Z 1: 88,400 pph

-2 74,300 pph. i 502 .1 82,000 pph- i 2 71,000 pph l 3 53,000 pph-75% 2 59.900:pph

-4 60,600.pph a 902 2- 62,000 pph 1

-100Z' 2 -- 68,600 pph '

The procedure- required adjustments to steam flow = transmitters if the difference limit could not be met. However, when steam flow and feed' flow for a given loop were plotted versus reactor power, in'all cases,_the steam-  ;)

flow / feed flow mismatch at the extrapolated maximum-flow rate was less'than l

the allowable mismatch,' 700,000 pph (for power: level test. plateaus < 1002) .

Pour test exceptions .were written, . with Westinghouse concurrence.- to; continue to the next test plateau without adjustment, based on the allowable; mismatch.

I At the 1002 power level test plateau, the eight steam' flow transmitters 1 were respanned to new full power delta-P values',' and additional data taken~. J The maximum flow difference between steam flow transmitters.in the same-loop t

1 was then 36,700 pph.

1 Examples (Loop 1) of the. final Steam / Feed Flow . . Mismatch graphical .

results are given in Figures 1 and 2.

l l

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l 6-20 a

STEAM ' FLOW / FEED FLOW MISMATCH ST-28 (100% RTP) FIGURE 1 i 5

4.5 -

'4 -

F ,

h.3.5 W

3 -

X 2.5 -

1 0 3 1

2 -

P 1.5 -

P H

t .,

0.5 -

0-O 10 20 30 40 50 6 0 ~- 70 80 190. ,100:-

i REACTOR POWER (%)  ;.

=

FT-510 FW FLOW I FT-512 STM FLOW Y

6-21

e _ '*

lp -.

t STEAM ; FLOW / FEED FLOW : MISMATCH-ST-28.(100% RTP) FIGUREi2 5

4.5 -

4 -

F F

0 3.5 -

w q 3 -

X l 1

1 2.5 -

O .]

E -l 2 -

6 s  !

.i P 1.5 -

P i H  !

1

.j q

i 0.5 -

1 0-

0. 10 20 30 40 50 60 70 80 90- '100- -l 4.

REACTOR POWER (%) }

=

FT-511 FW FLOW  ! FT-513 STM FLOW j 6-22.

n

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.'6.7 ST-36. AXIAL FLUX DIFFERENCE INSTRUMENTATION CALIBRATION c5 Obiective; '

Startup,' calibration of the- excore ' power- range- detectors,; and.

calibration d'ata for. the overtemperature delta-T f(Delta-I). summing amplifier gains are provided by :this test. . In faddition data. for

• calibration -of Main' Control Board. (MCB)- axial flux . difference - ( AFD): -

'g. Indicators and the AFD monitoring program on the_ Main Plant' Computer System (MPCS)Lis obtained.

Discussion The AFD test was performed at_ the '502, 752iand 100Z power level test' plateaus. The three sections of the. complete test ares

1. A preliminary incore-excore calibration, using a ~three point?

method for determining the relationship between .incore and excore -

AFD, prior to escalation to and above 50% power.- ,

2. Incore-excore calibration. using a multi-point' method ,for determining - the ' relationship. between incore ' and excore AFD

~

at greater than-752 RTP.

4

3. A'one-point calibration verification at 100Z RTP.

The relationship between incore and excore axial flux difference was=

determined by plotting the incore flux , difference. versus- excore . detector =

currents. From this plot, slopes and intercepts permitted calculation of  ;

summing amplifier gains, MPCS AFD monitoring program constants and alignment of the MCB AFD instrumentation. -

Station operating procedures for incore-excore - calibration, incore- j excore surveillance, flux mapping, and axial flux difference control were )

utilized'in the test procedure. l l

i L Results The acceptance criteria. were mets determination' of. the relationship between incore axial offset and AFD, and calibration of the' AFD instrumentation.

At the 50% power level test plateau, after an initial full core flux

, map (FCFM) =vas taken, a negative AFD was- produced- by dilution and

'l L

compensating insertion of the controlling rod bank. An initial' quarter core l flux map (QCFM) was taken, followed by. additional' dilution. ~A second QCFM-was . then taken. and the controlling ' bank was borated' back to its initial t position. .From the data , extrapolated full power currents for NIS as ~ a-function of AFD, and the gain of the ' f(delta-I) ' portion . of the over _ ;i; temperature delta-temperaturo (OTdeltaT) protection system were determined.

The results were ~t hen used- by I&C to calibrate power range (PR) a'nd (OTdeltaT) inputs.

6-23 l

l 1

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l a a #

6,? ST-36~. AXIAL' FLUX DIFFERENCE INSTRUMENTATION CALIBRAiION fContinued)

.- *=

, 3

^

i At the 752 power level ~ test plateau. - an FCFM was followed; by: dilution and' control rod insertion until the indicated AFD- was' near- the-predetermined : limit. - The initial > QCFM was taken. and dilution continued.

Af ter 'a 2-hour hold, a : second QCFM . was- taken, = boration was atarted and:

continued until:',the; controlling: bank -(Bank -D).was near. its starting position. Five additional QCFMs-were taken as.the axial xenon oscillation-progressed._ The . oscillation varied between --161 and +3% AFD.-- 'When data collection =.was completed, the xenon oscillation was terminated. Some i scatter in the incore to 'excore data was noted.: and some- data; points were -]

discarded however, the requirement of four useable ' points was always met.- '

From the data .IEC carried out-the necessary calibrations as noted above.

Results' of the calibrations at the; 752 power level test plateau are-given in Figures 1-9.

At the 100% power level test plateau,-- a' FCFM was takenito verify the previous calibrations. .The calibrations are acceptable if the incore-excore- --,

comparison differs by < 323 the maximum actual difference was 0.552.

s 1

k I

i t

h 6-24 >

I y 4 m- -w-w- ~ e, -

w 9- w

.g ST-36 AXIAL FLUX DIFFERENCE VS TIME N43 (75% RTP) FIGURE 1 10 8 -

6 -

4 -

4

~

0 D -2

& ^^^At I

E R

3:

~0 ~

[

C -10 -

E3 -12 -

'h -1_4 -

)- --16 -

-18 -

-20 -

-22 -

-24 ' ' ' ' '~

0 2 4 6 '8 10 12 14.16 18'20 22 24:26 28 30 32 34 36.38 TIME, HOURSL(7/10-11/90) bWT 'eti'k_=' 9--i.O'ha.Tt'.-WP.1 & .% i.=4%saa.1'hih"'haaiis.*-% .1 d141HL s W_*'M'""'"h'"""""""-"'U""""

"'_ "'W e'm_9 _. d._ y-

O-W

- ge- a, e, ST-36 DETECTOR CURRENTS VS INCORE AFD

~

N41 (75% RTP) . FIGURE 2 D

E ,175 -

T E 150 -

C T 125 -

0 R 100 -

T C 375 -

R 350 -'

R E 325 -

300 -

275 -

250 -

n1 225 -

P s ' ' ' ' ' ' ' ' '

-200

-30 -20 10 -5 0- 5 10. 15 20 25 30 INCORE JAFD (%)l

_._______.._._____..____LI ' "%  % W"'* * " ~ F '*p '4 F D- _-

• _.c d __.u- 4- -_W

ST-36 DETECTOR CURRENTS VS INCORE AFD -

N41 (75% RTP) FIGURE '3 450 D

E 430 -

f 410 -

0 390 -

370 -

T C S U 350 -

R R - 330 -

E -

N 310 -

T S 290 -

u 270 -

-a-m 250 -

P s- 230 ' ' ' ' ' ' ' ' '

-30 -20 -15 -10 -5 .0 5 10 15' '20 25 30 INCORE -AFD (%)

s

. . ~ _ -_ --

+ .~,...a, ^ :n-

, ,,.,,r .

. -an.

y ST-36 INCORE AFD VS EXCORE AFD

'N42. (75% RTP) FIGURE -4 6

4 -

N 2 -

C 9 O

R -2 -

E

_4 _

p A -G -

g F _g _

-10 -

y -12 -

E _14 -

R C -16 -

E _13 _

N T -20 -

) -22 -

' ' ' '- ' '4

' ' '  : r

-24

-20 -18 -16 -14 -12 -10 -8 -G -- r -2 0 ~2

EXCOREl AFD (PERCENT)-

. _ . . - = .. . . - - - - -. - . .

ST-36 DETECTOR CURRENTS VS INCOREI AFD

~

N42 (75% RTP) FIGURE. 5 D

E 500 -

f 475 -

o 450 -

425 -

C i U 400 -

M R

.R 375 -

E N 350 -

T S 325 -

u .300 .-

a in 275 -

p .

s ' ' ' ' ' ' ' '

250 ~'

-30 --25' -20 -15 -10 -5 0 5 10 :15 20 25 30-

'INCORE . AFD' (%)-

1;u . . , . _ . . n._ ,_,_ ___ ,,. _

~ _ . .~ _ c._. _ ._.__.2_

p .

ST-36 INCORE AFD VS EXCORE AFD

~

N43 (75% RTP) FIGURE .6 g

4 -

N 2 -

.C 0

^

O R -2 -

E

._4 _

A -6 -

[ h -10

-h -12 --

.E: -14 -

.R-C -16 -

E _13 - -

N

.r . -20 -- --

~

) -

' l ' ' ' ' ' ' ' '

-22. - 16 -10 14: -12 -10' -8 -6 2 0 i

-EXCORE' AFD (PERCENT)

-~ =. .

^

ST-36 DETECTOR CURRENTS VS INCORE AFD

.N43 (75% RTP) FIGURE 7 D

E- 450 -

T E 430 -

C T 410 -

0 It 390 -

4 o C' 370 -

' U M It 350 -

330 --

T 310 -

290 -

270 -

I" 250 -

s ' ' ' ' ' ' ' ' '

.230

-30 -25 -20 -10 -5 0 5 10 15 20 25 30 INCORE AFD (-%)

-- -- - - - = -- . c - -

= . . - - - - - -.

~

ST-36 INCORE AFD VS EXCORE AFD N44 (75% RTP) FIGURE 8 6

4 -

N 2 -

C 9

+

0 R -2 -

E

_4 _

m A -6 -

0 F D -0 ~

-10 -

i> -12' -

.E -

R C -16 -

E. _13 -

N T -20 -

) -22 -

-24

-22 18 -16 12 -10 -8 -6 -4 -2 0 EXCORE AFD (PERCENT)

. , - - , _ _. - . .. .: e . . . . >

c~;- . . . - .._ -.

~; .

~

ST-36 DETECTOR CURRENTS VS INCORE AFD N44 (75% RTP) FIGURE 9 D

E 450 -

T E 430 -

C T 410 -

0 R 390 -

C 370 -

i U O a 350 -

330 -

310 -

290 -

-270 -

250 -

p s 230 ' ~'~ ' ' ' ' ' ' ' '

-30 :-25 -20: -15 -10 -5 0 5 10 15 20 -25 30 INCORE - AFD (%)

c

..' .. 7.0 GENERAL PLANT TESTING Contentet J

7.1 ST-41 Radiation Survey.

7.2 ST-42 ' Water Chemistry control' 1

7.3 ST-43 Process Computer 7.4 ST-44 Loose Parts Monitoring i

7.5 ST-45 Process-Effluent Radiation Monitoring System--- }

7.6 ST-46 ' Ventilation System Operability Test t

7.7 ST-49 Circulating Water System, Thermal-Hydraulic Test l

-r 7.8' ST-51 ' Power Ascension. Dynamic Vibration Test ,

7.9 ST-52 Thermal Expansion j 7.10 ST-56 Piping Vibration. Testing '

.j Y

a 4

E L

t

.1 7-1

(

~,[.

+

m j 7.1 ST-41.' RADIATION SURVEY

'j Obiective-l The objective of this test was determination of. neutron and gamma dose rate levels and verification of operation'of' selected, radiation monitors by.- ,,

comparison of monitor response to survey readings. )

1 Discussion Radiation . surveys- were: conducted':to? verify. that- the- radiation-protection design features of the . facility, as described. in the: FSAR;3 havel been met. Data was taken'at the 502 and 100Z. power level test plateaus..

Health- Physics. personnel, . usingi survey ' instruments and. the ' normal Health Physics shield survey procedure,' conducted:the tests.- Survey maps'of ,

the areas to be evaluated were utilized, : and any discrepancies noted - and :

evaluated. During the ' surveys, - they incore detection' system was caution.

tagged to prevent operation of.the system with survey. personnel in the area.

Results i All. acceptance criteria were met:

1. Neutron and gansna radiation dose rates have been. measured at the-required locations, and high < radiation' areasL have; been - properly identified.

2. Except for documented: discrepancies :allimeasured.dopo rates =are within zone criteria.

3. The response' of the radiation monitors agrees with the t survey results - within

• 20% or are dispositioned;Lper the 'surveyL procedure._

i Seven -shield survey discrepancies were' found in l the ~ final (100%)

l survey.- All were dispositioned in'the survey procedure. .Three-were duetto streaming at doorways, not due to defects in shielding. . . One,: a. higher than -

(

expected reading, is located where: the shield thickness was decreased 1 to permit maintenance on pressurizer -heaters. 'The problem is'.very localized-and would not affect the . general ' area dose rate. The , remaining three discrepancies, survey readings which exceeded the'. greater.than 20Z criteria, resulted when the survey instrument was reading below atminimum sen'sitivity of the monitor.

i 1

7-2 l

a1

. 7.2 ST-42. WATER CHEMISTRY CONTROL

~Obiective The procedure -demonstrated that chemical and radiochemical control and analysis systems function to maintain primary and secondary water chemistry.

within the requirements of the Station Chemistry Control Program.-

Water chemistry control is described in FSAR, Section 14, Table 14.2-5,.

Sheet 45.

Discussion With the plant operating under steady-state conditions at the-30Z,J 50%,1 ,

752 and 100Z power level test plateaus, samples of reactor coolant steam i generator water; and feedwaterfwere obtained and analyzed using the Primary- ,

Chemistry Control and the Secondary . Chemistry Control - portions - of p the ~. '

Seabrook Station Chemistry Program.- A minimum of four hours stable power ,

'I

~

conditions at the specified power level testLplateau was required' .

Readings of selected secondary system on-line analyzers were compared to respective sample analyses to determine in-plant analyzer reading' accuracy.

Results All acceptance criteria were . met; as discussed - below. . chemistry in several systems was out of specification, but was acceptable ~by Westinghouse for a. plant at this operating stage.

Agreement between selected: secondary system on-line analyzers and the-respective sample analyses was very good. Table =1' lists values.obtained~and.  !

acceptable tolerance. -

At the 302 power level test plateau. cation. conductivity and sulfates in the steam generators, and the specific cation conductivity in the umin l steam, feedwater and condensate. systems were above the limits in the L

Westinghouse Secondary: Water Chemistry Manual. The;. source of theses contaminates -was identified. as original ' system preservatives.. With I Westinghouse and Chendstry Department approval, a test exception- was l prepared. The mechanism for removal is continued' operation.

Again, data taken at the 502, 75Z and- 1002 powerL levels indicated 1a q continued' problem- with. contamination Edue to system . preservatives.

Westinghouse, after a-review of the results,-noted that these were typical? I of similar plants at this operating stage. The Chemistry Department,'with Westinghouse concurrence has approved the results. 1 Plant chemistry - out-of-specification results at . the ' 100% power level -

test plateau are given in Table 2. l l

l

'a l

7-3 l l

l l

-q

, ,l

ss - .

, 7.2 ST-42; WATER CHEMISTRY CONTROL-(Continued) i TABLE l' ST-42 PROCESS INSTRUMENT / GRAB SAMPLE COMPARISONS J l

Power Leveli 100Z Sample Analyzer Sample.

• Point Parameter Reading: Analysis' - Tolerance-Blowdown- Na + 14.0 111.2 s-20 ppb-S/G A pH' 8. 9' 8.96-

• 0.5 pH .

Cation'Cond 3.4 3.43'

• 0.5 umho s 20 ppb

~

S/G B Na + 9.5 10.8 pH 8.8- 9;04

• 0.5 pH '

Cation Cond 3.5 -3.46 1.0.5 umho S/G C Na + 12.0 11.0 5 20 ppb' pH 8.8 9.04 .

• 0.5.pH 4 Cation Cond 3.4 3;4 4

• 0.5 umho S/G D Na + 10.0 9.3 s 20 ppb . ,

pH 8.9 9.07. i 0.5 pH- .i i

Cation Cond 3.1 -2.96

• O'.5 umho '

Cond Pump 0xygen 2.0 <5.0-

• 5' ppb Discharge '

pH 9.0 L 9' 2

.

• 0.5 pH Na+ 0.75 <1.0

• 3 ppb Cation Cond 0.24 0.243- i 0.1 umho .

Condensate Hydrazine 26.0- 26.0

• 20%

Htr 22 Outlet ' '

Feedwater Silica 9.0 <10.0

• 10%

Outlet 26'Htr Cation Cond 0.42 0.44 i 10%

pH 9.0 9.14 t 0.5 pH 1 0xygen <1.0 <5.0 1 5 ppb c

Hydrazine 20.0 20.0

• 202

• All tolerances given in 2 are of the full scale reading for that '

instrument.

7-4 4

r

a-

• 7.2 ST-42. WATER CHEMISTRY CONTROL'(Continued)?

-TABLE 2' .

. ST-42 OUT-OF-SPECIFICATION RESULTS. 1002 RTP-Cation Measured SSCP*-

Conductivity Values Limit 1 (uS/cm)

Main Steam 0.45 -0.2.

Feedwater 0.37-0.42 0.3 Steam Gen.- 3.0-3.7 0.8- '

Sulfate

(ppb)  ;

Steam Gen. 23.6 19.3 20' )

Silica (ppb)

'I Steam Gen. 570-520 300 s

• Seabrook Station Chemistry Program Manual'

t t

a l

i i

i e

7-5 La

,' '. 7.3 ST-43. PROCESS COMPUTER l

Obiective h i

Procedure , ST-43. verified that- the . Main Plant Computer System (MPCS).

receives correct _ inputs from process = variables _ and_ performs related calculations correctly. In conjunction with selected transient tests, the; procedure evaluated the response of the Safety _ Parameter Display . System 'l (SPDS)-during translent conditions.

The procedure is described in FSAR, Section 14. Table 14.2-5, Sheet 46.

i

-i Discussion  !

At the 30%,; 502, 75% and 1002 power level test - plateaus, computer values . of _ plant ~ process parameters.: indicated on l'.PCS, were compared to '~

other indications. to validate; the computer, _ In general, MPCS values = were' compared to Main Control Board!(MCB) hard-wired indications.

1 Where - the ' process instruments are RTDs, _ the 'only; indicators are on ]

MPCS in the case of containment ' air temperature' RTDs, .testingJand j calibration was completed (prior to ST-43),' using Integrated Leak Rate. Test j (ILRT) procedures: feedwater heater RTDs were verified _during the performance of ST-26, Thermal Power- 1 Measurement and Statepoint Data Collection, also prior to performance of ST-43. .

The MPCS sof tware verification required comparison of : the program {

output to the appropriate station procedures which are utilized when the plant computer is out of service,  ;

j The verification schedule is givent below:

Test Plateau - 30Z 50% 752 1002-

=

RCS Leakage Monitor Q X X X< X-

• SPDS .;

X X X' X

• Secondary ifeat Balance X. X X -

X

. Xenon / Samarium Monitor X X X- -.X' "

• Core Burnup Monitor X

. Containment Average Air Temperature .X j X X-Monitor l i

i

• RCS Delta-T and.TAVG Monitor X X'

. Rod Deviation Honitor and' Report X X j e

Condenser Performance Report X X i Feedwater ifeater Performance Report X X Turbine Performance Report 7 X e AFD'lonitor X X e

QPT Ratio Monitor and Report X X .

Response time of SPDS during transient conditions . was evaluated by.

accumulating data indicative of HPCS system load during periods of expected heavy clarm activity. Four transient startup tests were utilized: 10% load l

swing from 100% power (ST-34), large load reduction from 100Z power (ST-35),

unit trip from 1002 power (ST-38) and loss of offsite power (ST-39).

7-6

?

l

• . 7.3 ST-43. PROCESS COMPUTER (Continued)

Results All acceptance criteria were met, except for those parameters noted below, where test exceptions were taken.

The following criteria apply to the 302, 502 752 and 1002-power level test plateaus

+

All MPCS indications and Main Control Board (MCB) indications agree within the tolerance specified.

• All MPCS calculations are being performed correctly as demonstrated by the program verification attachments.

Additional acceptance criteria apply to the transient tests (above) y used to evaluate the response time of SPDS during transient conditions: 'A

+

The average SPDS response time on the STA work station is less than 10 seconds.

No MPCS fallovers occurred due to lack of CPU availability.

A number of test exceptions were taken at each test plateau, several \

addressed the same problems at successive power levels. Most exceptions-were related to data acquisition and instrument problems: all of this nature were easily resolved.

Two exceptions remain open. One, requires software changes to the turbine and condensar performance reports: the second was a computer failover which occurred shortly af ter the start of ST-39. Loss of Offsite Power. Both exceptions are now under evaluation.

7-7

I 7.4 ST-44. LOOSE PARTS MONITORING Obiective The procedure obtained RCS base 3tne noise level and signature data from the TEC Loose Parts Monitoring System (LPMS) during steady state conditions at the 502 and 1002 power level test plateaus. Additional data was obtained in conjunction with the performance of three transient tests.

ST-44 is described in FSAR, Section 14. Table 14.2-5. Sheet 47.

Discussion The LPMS provides a means for detection of loose metallic parts in the RCS. Twelve sensors (accelerometers) and associated circuitry continuously monitor noise levels at the reactor vessel and steam generators. If the noise level in a channel increases beyond a predetermined alert setpoint, an MPCS alarm alerts control room personnel.

With the plant operating in steady state, at the 50! and 1002 power level test plateaus, the procedure utilized two station procedures, an operational test and a quarterly surveillance test to obtain the necessary noise level and signature data. The measurements taken yielded alert setpoints for the LPMS channels, which were subsequently entered into ST-15 Reactor Plant System Setpoint Verification.

The LPMS sensor response was further evaluated during transient testet the load swing test (ST-34), at 50! and 1001 RTP for decreasing and  !

increasing transients, and the large load reduction (ST-35) and unit trip at 1001 RTP (ST-38).

Etellt.

All acceptance criteria were met except for one channelt Steady state and transient baseline data were obtained Hat the specified test plateaus.

Background values measured during normal plant operation were less-than the maximum allowable background values for each individual LPDS channel as defined in the quarterly surveillance.

Final alert setpoints have been determined.*

• LPMS alert level information submitted to meet the requirements of Reg. Guide 1.133. Part C.3.a. is given in Table 1.

At the 50Z power level test plateau, with the plant in steady state, the required operational test and quarterly surveillance was performed.

LPMS was found to be operating within its dynamic range, and no setpoint adjustments were required.

7-8  ;

1

. ,7.kET-44.LOOSEPARTSMONITORING(Continued) l

. i Also at this power level, steady state and transient baseline noi,se j level and signature data were recorded during the performance of ST-34, Load Swing Test. All LPMS auto functions operated as required during the transients.

Similar steady state results were obtained at the 2002 power level test '

platesu. At 100!, steady state and transient baseline noise level and signature data were recorded during the performance of three transient r

tests:

• ST-34, Load Swing Test

• ST-35, Large Load Reduction

• ST-38 Unit Trip from 1002 Power All LPMS auto functions again operated as required during the transients. .

The channel which failed acceptance, did not meet the background value  !

requirement at both the 50! and 1001 power level test plateaus. The ,

c quarterly surveillance acceptance criterion for LPMS is identical to the ,

above background value requirement. The surveillance specified that in this '

circumstance, a new alert setpoint should be calculated per the 'LPMS -

Technical Manual'. The calculation was made, and per vendor. recomunendation, the value was lef t unchanged (1.0), since no alarms were occurring during normal operation. A test exception was written to address the problem. '

r i

6

?

I 1

7-9 i

. . l

, 7.4 ST-44. LOOSE PARTS MONITORING (Continued) i

.

• j TABLE l' )

ST-44 LPMS ALERT SETPOINTS ,

i 1

ALERT SETPOINTS i FOR POWER OPERATION j I

CHANNEL ID SENS0E LOCATION INITIAL FINAL i

VB-YH-6824-1 Reactor Vessel Htad 1 1  !

VB-YH-6824 2 Reactor Vessel Head 1 -

1 j VB-YH-6825-1 Reactor Vessel Bottom 1 1 VB-YH-6825-2 Reactor Vessel Bottom i 1 l

VB-YH-6826 1 SG A Below Tube Sheet 1 1

• VB-YH-6826-2 SG A Above Tube Sheet 1 1 ,

VB YH-6827-1 SG B Below Tube Sheet 1 1 {

VB-(M-6827-2 SG B Above Tube Sheet 1 1 VB-YH-6828-1 SG C Below Tube Sheet 1 1 VB-YH-6828-2 SG C Above Tube Sheet i 1 VB-YH-6829-1 SG D Below Tube Sheet 1 1 -

VB-YH-6829-2 SG D Above Tube Sheet i 1 t

NOTF.: Alert setpoint values were initially set to a value of (1)-for all ,

LPMS channels as recommended by the vendor. Changes to these values '

are based on the results of RN 1714, Loose-Part Det.ection System Quarterly Surveillance, performed during this procedure and an evaluation of tabulated alarm data.

4

}

7-10

'l

')

. 7.5 $7-45. PROCESS AND EFFLUENT RADIATION MONITORING SYSTEM TEST Obiective The procedure verified that the process and effluent radiation monitors respond correctly to actual sample activity determined by radiochemical analysis, j A description of the procedure is found in FSAR, Chapter 14 Table 14.2-5. Sheet 48.

Discussi2E l The Radiation Data Management System (RDMS), through the RM-11 console. l provides readings from the process and effluent radiation monitors. RDMS is ]

designed to continuously monitor selected process and effluent steams ]

wherever the potential for a significant release of radioactivity exists 1 during normal operation ' including anticipated operational occurrences, and j during postulated accidents.

At the 502 and 100% power level test plateaus, Chemistry Department ;i personnel collected samples, coordinated with measurements from RDMS., After radiochemical sample analysis, the results were compared to the RDMS values to verify performance of the RDMS system. j I

Results The acceptance criterion, monitor _results agree.with laboratory sample analysis to within a factor of 2, was not expected to be met under all l

conditions. Sample analysis could not always be directly compared to RDMS ',

readings, particularly at beginning of core life. Results were acceptable even though RDMS reported values and sample analysis differed, provided there was a reasonable explanation for the difference.

The acceptance criterion was met for all monitors except those listed in Table 1. The monitor location (system) and an explanation for the i difference for each exception is includ6d.  !

Three test exceptions were written to address these failures to meet-acceptable limits.

7-11 i

. . .. 1

. 7.5 $7-45. PROCESS AND EFFLUENT RADIATION MONITORING SYSTEM TEST (Continued) )

1 TABLE'1' f ST-45 PROCESS AND EFFLUENT RADIATION MONITORING .

TEST EXCEPTIONS  !

Monitor M Exclanation RM-6500 Boron Waste Poor sampling point locations dilution has Storage occurred after monitor, but before sampling  ?

Tank Inlet point. Evaluation determined that dilution  ;

occurred due to a mispositioned valve, the i result of a drawing error. A DCR is in j preparation to correct the drawing.  ;

l RM-6502 Inlet to Defective instruments work request issued to >

Carbon Delay repair.

Bed Room RM-6504 Waste Gas Discrepancy because short-lived isotopes Compressor decay before sample analysis can be Discharge completed.

RM-6509 Liquid Waste Level of activity measured was below monitor i Test Tank sensitivity limit. i Discharge to -'

Circulating ,

Water System i (CWS)

RM-6520e2 RC Letdown. Same as RM-6504.

Gross Activity  !

Monitor RM-6514 Liquid Waste- A Temporary Modification which is not from Evapora- scheduled for removal until the first 4 tors refueling outage prevents sample flow to the monitor.

NOTE: Additional monitors, RM-6490, RM-6501, RM-6502, RM-6515. RM-6516 RM-6519, and RM-6528 were excepted at the 50% power level ,

plateau, because levels of activity measured were below monitor ,

sensitivity limits. i l

+

4 7-12 h

-,%+ -- . - , , ~ . . - . ~ ,, . - - -. , -- r-. .

7.6 ST-46. VENTILATION SYSTEM OPrnARILITY TEST Obiective Heating, ventilation and air conditioning ' systems , were monitored to demonstrate that the systems maintain their service environment areas within design limits under_ normal plant operating conditions. A comparison of permanent room' temperature indicator readings with survey instruments was made.

The procedure is described in FSAR, Section 15. Table 14.2-5, Sheet 49.

Discussion With the reactor in steady state, at the 502 and 1002 power level test plateaus, a test team monitored, using temperature and humidity measuring devices, ventilation systems operating in their normal operating mode.-

The areas monitored were included in the following structures:

• Containment Building

• Containment Enclosure Diesel Generator Building

+

Waste Process Building (Tank Farm)

Service Water Pumphouse and Cooling Tower Control Building Emergency Feedwater Pumphouse

• Fuel Storage Building Main Steam and Feedwater Pipe Chases Primary Auxiliary Building

. Equipment Vaults For the identified environmental zones in the above areas, six representative ambient area temperature and/or humidity readings were taken at 502 powers based on the results, only one set of readings was taken at 1002 power.

t Results The acceptance criteria:

Test data has determined that the ventilation systems are capable of maintaining equipment space environmental conditions, based on temperature and humidity, within FSAR specified design values.

A comparison of permanent plant temperature indicator readings to <

measured Technical readings verifies that plant equipment used to satisfy  !

Specifications is sensing a representative area temperature, were met with three exceptions as noted below.

7-13

, _7.6 ST-46. VENTILATION SYSTEM OPER ARILITY TEST (Continued)

Hessurements were made ~ in the indicated . areas except for the east and west main steam and feedwater pipe chases. Temporary fans were installed in these areas to maintain acceptable temperatures. A modification to change design air. flows for the pipe chase cooling fans is in process, and the pipe chase area will be monitored after it is approved and installed. The modification will be made when the plant is in a refueling or maintenance outage. . Test exceptions were taken at the 502 and 1002 plateaus for the areas where temporary fans were in use.

Certain environmental zones failed to meet design requirements:

Containment 2 Areas Control Building 10 Areas East Pipe Chase 10 Areas West Pipe Chase 5 Areas -

A test exception was written for the above areas, and a Request for Engineering Services (RES) submitted for engineering evaluation.

7-14 l

1 l

9 7.7 ST-49. CIRCULATING WATER SYSTEM THrRMAL. HYDRAULIC TEST-(DEFrantn)

Obiective The procedure was prepared to demonstrate that the thermal-hydraulic characteristics of the Circulating Water System are such that the intake tunnel can be treated with 110'F to 120*F water for 1 to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> with the plant at 751 RTP. The procedure was also to demonstrate that heat treatment can be performed within the constraints of the National Pollutant Discharge Elimination System (NPDES) permit, and determine the effects on the power plant.

The circulating water system thermal-hydraulic test is described in FSAR, Chapter 14. Table 14.2-5, Sheet 52.

Discussion The major source of fouling in the circulating water system tunnels is expected to be the common mussel. Experiments on mussels typical of the area demonstrated that a temperature of 110'F for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> would produce 1001 mortality. The NPDES permit for the test limited the warm water discharge to 120*F for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, and is the basis for the acceptance criteria.

Heat treatment of the intake tunnel requires establishment of flow reversal and warm water recirculation to reach the required temperature condition, followed by the necessary treatment period.

The chlorination system installed af ter the procedure had been planned appears to be effectively eliminating the buildup of marine organisms in the intake tunnels a decision was made to defer the procedure.

ST-49 will be performed prior to any heat treatment operation. It should be noted that Technical Specifications do not contemplate operation in the heat treatment mode. NHY plans to request a technical specification change which will explicitly address heat treatment.

Heat treatment of the intake tunnel is a design feature of the Seabrook plant, not assumed in safety analysis. Therefore, the decision to defer ST-49 vill not affect a successful completion of PATP or the overall Startup Test Program.

7-15 I

. . . . . . . m

I 7.8 ST-51. POWER ASCENSION DYNAMIC TEST '

_0biective The objective of this procedure was measuroment of the dynamic response of certain main steam, feedwater and pressurizer relief systems under transient conditions. ,

t Dynamic testing is described in FSAR, Section 3.9(B).2, and ' Table 3.9(B)-1.

i Discussion

?

Displacement transducers, installed on the piping and components to be monitored, measured the-behavior of the system during specified transients. ,

The data was recorded on GETARS for later analysis. l During the precritical test program, dynamic testing of the pressurizer ,

relief system, individual operation of the - condenser steam dump and atmospheric steam dump valves, and trip of the emergency feedwater pump' -

Terry Turbine was conducted under ST-51. ,

i The feedwater pump dynamic testing was performed in. conjunction with '

ST-53, Turbine Driven Emergency Feedwater Pump Start Verification. The steam dump portion was performed in conjunction with ST-55, Steam Dump System Test.  !

In the Power Ascension Test Program, additional dynamic response testing, coordinated with ST-38. Unit Trip from 1002 Power, was conducted.

The transients monitored were:

1. Main Steam System - Turbine trip, and simultaneous operation of the condenser steam dump valves.

2. Feedwater System - Closure of the feedwater containment isolation valve and trip of the steam generator main feedwater pump.

Results The acceptance criteria was a review and_ verification by New Hampshire l Yankee Engineering that the measured stresses do not exceed code limits.

All measurements met this criteria.

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. . 7.9 ST-52. POWER ASCENSION THERMAL EXPANSION TEST ,

Obiective The objective of this test was a demonstration that piping systems were l free to thermally expand consistent with design. These measurements confirmed that associated restraints and supports allow the required thermal movement.

The test is described in FSAR, Section 3.5.3.4.d. and Section 14 I Table 14.2-3 Sheet 6.  ?

Discussion r Thermal expansion data was obtained from displacement measuring [

transducers and by visual observation of spring hangers snubbers, and pipe whip restraints. Walkdowns were performed to identify areas of potential restraint to free movement. ,

The following systems were monitored for baseline, no-load (557'F),

30Z, 502, 752 and 100Z conditions:

1. Snubbers:

Auxiliary Steam Condensate- * -

Primary Component Cooling Chemical and Volume Control  ;

i Diesel Generator Feedwater '

Main Steam l Main Steam Drains Nitrogen Gas

• Reactor Coolant Residual Heat Removal Steam Generator Blowdown Spent Fuel Pool Cooling Safety Injection Service Water Waste Processing Liquid Drains t

2. Spring Hangers *:

Condensate Extraction Steam Feedwater Heater Drains Main Steam Main Steam Drains .

Moisture Separator & Reheater Drains / Sampling System

• Adjustments were made, during the test sequence. to spring hangers which were not within their hot and cold settings.

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. ,,'$.9ST-52.POWERASCENSIONTHERMALEXPANSIONTEST(Continued) l

3. Pipe Whip Restraints **: I Feedwater i Hain Steam l

• • Baseline and 1002 test conditions only.

The following system was monitored during a turbine driven EFW pump l runt

1. Snubbers:

Main Steam (associated with EFW pump)

Results f Acceptance criteria for thermal movements were specified for-Westinghouse (NSSS) scope, for UEEC scope, and for completion of NNY ,

Engineering review:

t

• Piping and components are free to expand without restriction other than by design during heatup'and operation of the reactor coolant and associated systems.

• The measured thermal movements shall be within

• 50% of the analytical value or

• _0.25 inches, whichever is greater for movements up to 1 inch. For analytical movements greater than 1 inch, the measured thermal movements shall.be within i 25% of the analytical value, t

• Problem log discrepancies have been resolved.

• NHY Engineering evaluation is complete for data obtained outside the acceptance criteria.- .

All acceptance criteria were met.

Eighty-six problem log sheets were developed during performance of ST-

52. The procedure required preparation of a problem log sheet, and NHY i

Engineering resolution for each problem identified at each power level test plateau. Thus, in some cases, four problem log sheets were prepared and resolved for a single monitored system. A summary of actual problems is given in Table 1.

No thermal expansion problems were identified during the EFW pump measurements.

I A test exception was written for test equipment which is not accessible for removal under present plant operating conditions. 'I 7-18

• /.9 ST-52.~ POWER' ASCENSION THERMAL EXPANSION TEST (Continued)

TABLE 1

SUMMARY

OF THERMAL EXPANSION PROBLEMS Problem Reoorted Number of Cases Predicted growth does not match actual growth within acceptable tolerance. 13*

Bolting on instrument loose 4 Instrument broken 2 Instrument displaced slightly 2 Counterweight striking structure 2 Spring can still has shipping lugs 1 Spring can no longer installed 1 Clamp holding instrument moved 1 Valve body in contact with grating 1 Unable to move as designed .j,,

Total 28

• Thirteen instrumented locations responsible for 71 problem sheets.

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• ,7910 ST-56. PIPING VIBRATION TESTING g .'

Obiective The objective of this test was verification that the vibration level of selected portions of the condensate, feedwater and main steam systems within the containment MS/FW chases, and turbine building are within design limite

• under steady state conditions at 1001 RTP. ,

. i The procedure is required by FSAR Section 3.9(B).2.la and Reg. Guide 1.68, Revision 2, Appendix A. Sec. 5.0.0, t

Discussion ,.

Piping vibration data was obtained with hand-held vibration meters at potential high vibration areas on condensate, feedwater and main steam systems. Points where high vibration was expected, such as pumps, control valves, heat exchangers, etc. received particular attention.

The data was recorded and determined for acceptability by comparing measured values with previously determined acceptability limits.

Unacceptable results were evaluated for resolution by NHY Engineering. -

Results t The acceptance criterion was mets acceptability of the data by NHY Engineering when compared to analytically predicted limits.

All vibration amplitudes measured were small compared to the limiting i values specified in the procedure.

l The procedure specified an initial condition of two condensate pumps operating; normal plant operating conditions at the 1002 power level test plateau was with three condensate pumps operating. 1 Twenty-five data points on condensate systems were remeasured when it  ;

was found that one vibration meter f ailed a post-measurement calibration i test. All remeasured data points were acceptable. A test exception was l written to permit remeasurements. j 1

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