ML20153C747

Cycle 1 Startup Rept Final Suppl

ML20153C747
Person / Time
Site:	Braidwood
Issue date:	08/31/1988
From:	COMMONWEALTH EDISON CO.
To:
Shared Package
ML20153C732	List: ML20153C729 ML20153C747
References
NUDOCS 8809010277
Download: ML20153C747 (49)

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COMMONWEALTH EDISON COMPANY BRAIDWOOD UNIT 1 CYCLE 1 l

STARTUP REPORT-FINAL SUPPLEMENT ,

AUGUST 1988 i

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8809010277 DR 000025 ~ '

ADOCK 05000456 l PNU i 1

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t-g TABLE OF CONTENTE Section Title EA&c.

List of Tables 2 1.0 Introduction 3 2.0 Discussion of Braidwood Startup Program 4 3.0 Braidwood Startup Tests at Full Power 9 P

4.0 References 48 h

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LIST OF TABLES TABLE TITLE PAGI '

2.0-1 Braidwocd Unit 1 Major Milestone 6 2.0-2 Braidwood Operational Modes 7 2.1 Full Power Flux Map Results

3.0 List of Test Summaries 10 3.3 Steam Generator Level Control Summary 14 3.4-1 Primary Makeup Water Chemistry Summary 16 3.4-2 Reactor Coolant Water Chemistry Summary 17 i

3.4-3 Steam Generator Water Chemistry Summary 18 i 3.4-4 Main Feedwater Chemis-cy Summary 19 3.12-1 10% Load Decrease at 75% Power Summary (Repeated Test) 28 3.12-2 10% Load Increase at 75% Power Summary '

(Repeated Test) 29 3

3.12-3 10% Load Decrease at Full Power Summary 30 3.12-4 10% Load Increase at Full Power Summary 31 3.13 Full Power Plant Trip Summary 33 3.14 50% Load Reduction at Full Power Summary 35 1 3.16 Thermal Power Measurement and Statepoint Data Collectf.on Summary 38 i

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s 1.0 - INTRQDUCTION This final supplemental startup report describes the required testing at Braidwood Station Unit 1 at the 100% power testing plateau. It satisfies the requirement of the Braidwood Unit 1 Technical Specifications that a Startup Report be submitted to the NRC every 90 days until completion of the Startup 'i, sting Program.

Braidwood Station, located in northeastern Illinois, uti*ites a four loop Westingbouse pressurized water reactor system.

Westinghouse Electric Corporation, Sargent & Lundy, and the Coamonwealth Edison Company jointly participated in the design and construction of Braidwood Station. The plant is operated by Commonwealth Edison Company with Sargent & Lundy as the Architect - Engineer.

The Nuclear Steam Supply System is designed for a reactor power output of 3411 MWt. The equivalent warranted gross and approximate net electrical output are 1175 MWe and 1120 MWe, respectively. Cooling for the plant is provided by a large man-made cooling lake of approximately 2500 acres constructed over a previously strip-mined area. Essential service cooling is provided by a 99-acre auxiliary cooling pond which is integral with the main pond.

This is the Final Startup Report Supplement which covers testing ,

through the completion of the Braidwood Unit 1 Startup Testing Program.

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s 2.0 - DISCUSSION OF BRAIDWOOD STARTUP PROGRAM The Braidwood Unit 1 startup testing program cousieted of single and multi-system tests that occurred commencing with initial fuel loading and continuing through full power. These tests demonstrated overall plant performance and included such activities as fuel loading, precritica,' testing, low power tests, and power ascension tests. Testing sequence documents  ;

were utilized for each plateau to coordinate the sequence of testing activities at that plateau, in the section that follows, a description or the completed 100% ,

power sequence of testing is provided. Also included as a part :

of Section 2.0 is a table showing major milestones'for Braidwood Unit 1 and a list of operational modes as dhfined by the plant -

Technical Specifications.

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s TkBLE2.0-1 BRAIDWOOD UNIT 1 MAJOR MILESTONES MAJ.0A_MILESIONEI DATE Fuel Load License Issued 10/17/86 Fuel Load Commenced 10/25/86 5% Power License Issued 03/21/87 Initial Criticality 05/29/87 5% License (Low Power) Tests Completed 06/23/87 Full Power License Received 06/30/87 :

Entered Mode 1 07/07/87 Initial Synchronization to Grid 07/12/87 30% Tests Completed 09/19/87 ,

50% Tests Completed 10/23/87 l Plant In Service

• 11/20/87 75% Tests Completed 12/02/87 i 90% Tests Completed 12/10/87 Surveil 189ce Outage 01/01/88 thru 03/21/88 Post Outage Tests Completed 06/30/88 100% Tests Completed 07/29/C8 Commercial Service 07/29/88

• Illinois Commerce Commission definition  ;

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TABLE 2.0-2 QEERAIIDNAL MODES REACTIVITY  % RATED AVERAGE COOLANT liDDE CONDITI.QL_Fg T TPMAL POWER

• TEMPERATURE-

1. POWER OPERATION 1 0.,99 > 5% 1 350*F

2. STARTUP 1 0.99 1 5% 1 350'F

3. HOT STANDBY < 0.99 0 2 350'*F

4. HOT SHUTDOWN < 0.99 0 350'F>Tavg>200*F

5. COLD SHUTDOWN < 0.99 0 1 200*F

6. REFUELING ** 1 0.95 0 1 140'F

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• Excluding decay heat.

• • Fuel in the reactor vessel with the vessel head closure bolts less than fully tensioned or with the head removed.

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2.1 - TEST _SEQUIRCL . T 1007,_P_0MIR The Test Sequence at 100% Power was utilized to define the activities which constituted the startup testing program during power escalation following RTD replacement up to 100% power and testing at approximately 100% of rated thermal power. This document ensured that the Test Sequence at 90% Power had been completed and the results approved prior to increasing power above the 90% testing plateau. Prior to increasing power for this test sequance, reactor core flux map results from a 90%

power baseline map were verified acceptable. The flux map results were also extrapolated to 108% power to ensure parameters indicative of DNBR and linear heat rate were acceptable for power ascension to the 100% testing' plateau.

The Unit was shut down following the completion of a portion of the 100% plateau testing for modification work. Retesting following Reactor Coolant System Narrow Range Resistance Temperature Detector (RTD) replacements and ether modifications was performed in Modes 3 and 4 n.ior to reactor restart and at the 50% and 75% power levels duliag escalation back to 100%

power.

Plant operating procedures were utilized where appropriate to establish plant conditions and to change reactor power. During the initial ascension to the 100% plateau, power was stabilized near the 98% level to accommodate testing at this plateau.

Results of individual tests completed during this power ascension and while at the 100% plateau are discussed in Section 3.0 of this report. At the 100% testing plateau, a comparison was made of selected control room indications with the process computer to demonstrate com Data was also taken at the 100%plateau puter monitoring capability.

and reviewed to demonstrata the operability of the inadequate core cooling monitoring system. Reactor cavity sump leakage data was obtained and a leakage rate of 0 gpm was determined. Primary containment ventilation system data, auxiliary building ventilation system data, and steam tunnel ventilation system data was obtained to i verify the adequacy of these systems to provide the necessary cooling and ventilation in their respective areas. A discussion of ventilation system performance is also presented in Section 3.0 of this report. Containment penetration cooling temperature data was also obtained while at the 100% plateau and all readings were found to be well below the 120*F upper limit.

A flux map was taken during the 100% power plateau and the results were examined to determine acceptability. Table 2.1 is a tabulation of these flux map results.

Upon completion of this testing phase, the plant was restored as directed by the Shift Engineer.

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IARLE 2.1 FULL POWER FLUX MAP RESULTS ACTUAL LIMIT Peak Linear Heat Rate (KW/ft) 10.93 12.6 Extrapolated to 100% Power 11.27 FDHN 1.4411 1.5640 ,

Fxy - rodded N/A N/A Fxy - unrodded 1.5209 1.5593 Quadrant Power Tilt Ratios 0.987.8 ._L.00Ai 1.02 ,

0.9988 1.0089 I

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. o 3.0 - BRAIDWOOD STARTUP TESTS AT FULL POWER TABLE 3.0 LIST OF TEST SUMMARIES 3.1 Degassing the Reactor Coolant System, CV-30 3.2 Pipe Vibration, EM-30C 3.3 Steam Generator Level Controller Response, TW-30D 3.4 Chemistry and Radio Chemistry Criteria for Monitoring Water Quality During Startup and Power Ascension, PS-32 3.5 Radiation Surveys During Power Ascension, PS-33 3.6 Process and Effluent Monitors and Failed Fuel Monitor Checks, PS-34 3.7 Reactor Coolant System Flow Measurement, RC-31B 3.8 Heat Capacity Verification for Auxiliary Building HVAC System, VA-30 3.9 Heat Capacity Verification for Switchgear Heat Removal Ventilation System VX-30 3.10 Heat Capacity Verification for Miscellaneous Electrical Equipment Room Ventilation System VE-30 3.11 Heat Capacity Verification for Remote Shutdown Control Room Ventilttion, VI-80 3.12 Load Swing Test, NR-36 3.13 Full Power Plant Trip, NR-39 3.14 Large Load R? duction NR-37 3.15 Calibration of Steam and Feedwater Flow, FW-31 3.16 Thermal Power Measurement and Statepoint Data Collection, IT-32A/B/C/D/E 3.17 Incore Thermocouple /RTD Cross Calibration, IT-33 3.18 Operational Alignment of Excore Nuclear Instrumentation NR-34D/E/F 3.19 Axial Flux Difference Instrumentation Adjustment at Full Power, NR-35B 3.20 Reactor Loose Parts Monitoring, LM-30 3.21 Startup Adjustments of Reactor Control Systems RD-80 3.22 Nuclear Steam Supply System Acceptance Test, TG-80 3.23 Process Computer Verification - Thermal Power, CX-80 3.24 Process Computer Verification - Incore Thermocouple, CX-81 3.25 Process Computer Verification - Boron Follow, CX-82 3.26 Process Computer Verification - Flux Mapping, CX-83 l

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3.1 - DEGASSING THE_ REACTOR __ COOLANT SYSTEM. CV-30 OBJECTIVE The purpose of this procedure was to verify the removal of excess hydrogen and other non-condensable gases from the reactor coolant system (RCS).

IEST MEIHODOLQGI

With the plant in Hot Standby Condition after the Full Power  ;

Plant Trip test the RCS was desassed via the Waste Gas system from the Volume Control Tank. The process of degassing actually involved a transfer of RCS water to the Volume Control Tank and ,

then venting off the gases coming from solution to the Waste Gas system, j SIRRfARY OF PJSJ1LIS

The acceptance criteria for this test was to reduce the dissolved gas concentration of the RCS by at least 10 cc/kg H 2 O from its initial value of greater than 20cc/kg H 20.

4 Through several degassing iterations, the concentration was  !

reduced from 27.1 cc/kg H2 O to 7.5 cc/kg H 2 O representing a a change of 19.6 cc/kg H 20.

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3. 2 - PIPE VIBRATION. Eti-3AC QU ICIIYE The Pipe Vibration test procedure demonstrated that the peak stresses resulting from steady state flow induced vibration were within allowable design limits. The test was performed in an effort to qualify or accept the piping associated with the Main Steam and Main Feedwater systems.

IEST_METHODOLOGI The Pipe Vibration test utilized normal operating conditions on the Main Steam and Main Feedwater systems at the 90-100% power level in order to inspect and monitor the effects of flow induced vibration. The inspection consisted of walk downs of the affected piping and through the use of portab~1e vibration analyzers. Readings were taken on each selected piping system portion and anywhere else levels were deemed excessive. A simple beam analogy or finite element computer analysis was then applied in the areas of concern arriving at an allowable deflection limit and a comparison was made between the theoretical limit and the actual reading. Based on the outcome, vibration levels less than the allowable limit met the Acceptance Criteria. Vibration monitoring was also performed using temporarily mounted accelerometers on piping located in inaccessible areas and the readings stored in a data collection system. The data was then forwarded to offsite Engineering for their analysis and approval.

SIIMMARY OF RESULIS The tests performed at the 90-100% power level generated one calculation where the level of vibration was questionable with respect to the piping subsystems uncer inspection. The piping associated with the 1A Motor Driven Main Feedwater Pump was found to have excessive vibration at high pump flowrates. The piping was provisionally accepted for flowrates of u? to 13000 gpm and further evaluations are in progress. All other vlbration readings obtained through system walkdowns were analyzed to have met their respective allowable limits and therefore the Acceptance Criteria was met. Results from the remote vibration monitoring process met all Acceptance Criteria limits as confirmed by offsite Engineering.

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3.3 - STEAM GENERATOR LEVEL CONTR0f.TJR RESPONSE. FW-30D OBJECTIVE Steam generator level control stability was demonstrated periodically throughout power ascension. Level conttol stability of the four steam generators was previously demonstrated while operating on the main f eedwater regulatir.g valves and the feedwater bypass regulating valves. Level control stability was also previously demonstrated while transferring feedwater flow between the feedwater bypass regulating valves and the main feedwater regulating valves and >

while swapping main feedwater pumps. This execution of the FW-30D test at 100% power was the final performance of steam generator level controller response testing.

TEST METHODOLOGY In order to verify level control stability while operating on the main feedwater regulating valves, a 5% level dev$ation was ,

established in each steam generator manually. The control system was then transferred to the automatic control position.

The actual steam generator level was monitored to determine if it returned to the nominal programmed level of 66'A within a t specified time frame.

SUMMARY

OF RESULTS '

When given a 5% level deviation (high or low), the main feedwater regulating valves returned the steam generator level to the programmed level within 3 reset time constants as expected. This test was performed at approximately 98% power.

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TABLE 3.3 LIE g GENERATOR LEVEL CONIRQL

SUMMARY

MAIN TEEDWATER REGULATING VALVE '

LEVEL C0KIROL RESPONSE PERFORMED AT 98% POWER FW-30D TEST j ACCEPTANCE ACTUAL TIME STEAM LEVEL CRITERIA IN RESPONSE

_ GENERATOR DEVIATION MINUTES (3 times constants) IN MINUTES t 1A 5% 0 1 84.8 39

' 5% o 1 84.8 45 ,

1B 5% 0 1 98.0 42 5% o 1 98.0 46 1C 5% 0 1 78.1 40 5% o 1 78.1 47.5 1D 5% 0 1 84.3 53 5% o 1 84.3 51 J

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3.4 - CHEtiLSTRY_AND RADIOCHEMISTRI_CMITERIA FOR MQUITORING WATER QUALITY DURING STARTUP AND POWER ASCENSION. PS-32 DBJECTIYE j This test was performed to verify that the water quality within  !

the primary water make-up system, the reactor coolant system,  !

the steam generator blowdown system, and steam generator i feedwater system met the specified chemistry requirements.

IEST_ METHODOLOGY The test was performed by obtaining samples of the primary make-up water, reactor coolant, steam generator blowdown, and l

' main feedwater systems from the appropriate sample panels l throughout the plant. Chemical analyses were then performed on  ;

each sample and the results tabulated.  !

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SUMMARY

OF RESULIS {

s During this final execution of this test, all required

Acceptance Criteria were adequately met for each system that was sampled. No corrective actions in plant operation were needed  !

to meet the Acceptance Criteria.

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guidelines stated within the test. The values given in these tables are the results of the samples obtained from each system.

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i sequence.

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IARLE 3.4-1 ERIMARY MAKEUP WATER. CHEMISTRY

SUMMARY

CHEMISTRY CRITERIA /

PARAMETER GUIDELINE TEST RESULT 1

Chloride (ppb) --- 1 Fluoride (ppb) --- <1 Total ,

Chloride and Fluoride 1100 ppb <2 i

l Silica 1100 ppb 30 1 ,

pH 6 25 C 6.0-8.0 6.6 Specific l Conductivity Micromho/cm 6 25'c 11.0 0.6 r

i Total Solids <1.0 ppm 0.366 Suspended l Solids 10.05 ppm <0.05 j

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Sodium 110 ppb 2 l Aluminum (20 ppb 1 Calcium 15 ppb 4 l

Magnesium 15 ppb 2 Potassium 110 ppb 5 15 2149m(082388)

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TABLE 3.4-2 REACTOR COOLANT WATER CliEMISTRY

SUMMARY

CHEMISTRY CRITERIA / ,

PARAMETER GUIDELINE TEST RESULT Chloride <150 ppb 28 Fluoride <l50 ppb <1 Dissolved ,

Ox9sen (100 ppb 0 Silica <1000 ppb 108 I

Li,thium 0.7-2.2 ppm 1.3 Hydrogen 25-50 cc/kg H 2O 30.3 Suspended j Solids <0.200 ppm <0.2 q

Aluminum <50 ppb 1.7  :

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l IABLE 3.4-3 STEAM GENERATOR WATER CHEMISTRY f"MMARY CHEMISTRY CRITERIA / TEST RESULTS

_EAIMMEIER GUIDELINE S/G 1A S!G 1B S/G 1C S/G 1D Cation Conductivity  !

Micromho/cm 4 25'C 10.8 0.5 0.53 0.53 0.63  ;

pH 6 25 C 19.0 9.07 9.03 9.02 9.19 .

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--w------,a--. -- .-,ee f -- ...- c-.p -,,..,ow-e_- , p9,,__,,m99-. ,--n,,,---,-,*gge %9.%. . ., -, -

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TABLE 3.4-4 MAIN FEEDWATER CHEMISTRY

SUMMARY

CHEMISTRY CRITERIA /  !

PAR &iETER GUIDELINE TEST RESULT Hydrazine >20 ppb 24 r

Dissolved Oxygen 1 3 ppb 0.11  ;

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3.5 - RADIATION SURVEYS DURING POWER ASCENSION. PS-33 ORJECTIVE The Radiation Surveys During Power Ascension test was performed to determine dose levels at specified points throughout the plant and to verify the effectiveness of radiation shielding to gamma and neutron radiation.

IESIJiE'DIODOLOSI Gamma radiation dose rate values were established by surveying with portable survey instrumentation in the Service, Radwaste, Fuel Handling Turbine, Auxiliary, and the Unit 1 Containment Buildings. Neutron radiation dose rate values were established in the Unit 1 Containment and certain penetration areas.

SUMMARLQLRES11LIS The effectiveness of neutron radiation shielding in Unit 1 Containment was found to be inadequate during this performance of this test. This was also the case for Byron Units 1 and 2.

Radiation surveys for selected base points would be repeated after any neutron shielding modifications are completed.

The effectiveness of gamma shielding and the general determination of dose levels were found adequate during performances of the test. At nominally full power, gamma dose rates ranged from < 0.1 mR/hr in the Service and Turbine Buildings to 16 mR/hr at one Auxiliary Building location. Only two points exceeded 5 mR/hr; the one at 16 mR/hr and one at 7 mR/ hour. Auxiliary Building Neutron dose rates ranged from

<0.5 mrem /hr to 10 mrem /hr at one location.

In the Containment Building, gamma done rates ranged from < 0.1 mR/hr to 80 mR/hr. Only nine points exceeded 5 mR/hr; one at 6 mR/hr, two at 7 mn/hr, two at 14 mR/hr and one each at 8, 10, 60 and 80 mR/hr. Neutron dose rates ira the Unit 1 Containment Building ranged from <0,5 mrem /hr to 400 mrem /hr. Only six points exceeded 20 mrem /hr; one each at 25, 40, 60, 80, 300, and 400 mrem /hr. All measured dose rates have been dispositioned as acceptable for continued plant operation.

This test was executed with Unit 1 at approximately 96% power, 19 2149m(Od2388)

1 3.6 - PROCESS AND EFFLUENT MONITORS AND FAILED FUEL MONITOR CHECKS <

PS-34 QBJECTIVE The Process and Effluent Monitors and Failed Fuel Monitor Checks startup test was performed to verify proper responses of all process and effluent monitors and the failed fuel monitor to Known sources of radiation.

I4ST METHODOLOGY Radioactive sources with known activities were exposed to the J Process and effluent monitor and failed fuel monitor detectors.

The observed detector responses were compared to the calculated expected detector responses.

SUMMARY

OF RESULTS l

i All process and effluent monitors and failed fuel monitor detector responses were within i 20% of the expected detector

responses during the performance of this test, thus meeting the  ;

test Acce Five detectors that initially failed the 120% ptance Criteria. criteria were repaired and retested subsequent to test execution.

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3.7 - REACTOR COOLANT SYSTEM (RCS) FLOW MEASUREMENT. RC-31B QBJECTIVE The RCS Flow Measurement test procedure was performed at various reactor power levels (50, 75% and 100% power) to determine the RCS flowrate for each of the 4 loops and also the total RCS flowrate. The during the 100% performances of the test at 50% and 75% powersequence we RC-31B tests. They were repeated during the power ascension following RCS Narrow R&nge RTD replacement.

IEST METHODOLOGY While the plant was at 50, 75, and 100% reactor power level data was obtained to determine the RCS flowtate. This data consisted of RCS hot and cold leg spare RTD resistance readings. Both normal and reverse polarity readings were taken. These RTD resistance readings were converted to temperatures ('F) and then, using Steam Tables, to hot & cold leg enthalpies and cold leg specific volumes. This data combined with calorimetric power values from the appropriate Startup Test IT-32C, IT-72D, or IT-32E, were used to calculate the flowrates. This method of calculation was used each time the test was performed at the various power levels.

The actual calculated flowrates for the various power levels were as follows:

RCS LOOP FLOWMIES (GPM)

TOTAL RCS

% POWER LQQP_l LQQP_2 LOOP 3 LOOP 4 FLOWRATE 50% 102,002 99,509 103,582 101,704 406,797 75% 101.878 98,906 103,099 99,769 403,652 100% 104,060 100,280 101,707 99,255 405,302 All the results met the minimum flow requirements for total RCS flowrate.

SUMMARLDLEESULTS The required total RCS flowrate as determined by calorimetric measurement varied for the different power levels and were as follows:

% POWER TOTAL FLOWRAIE MSIS OF REO.UIREliEllI 50% 1 377,600 gpm 100% of Thermal Design Value 75% 1 390,400 gpm Technical Specification minimum value 100% 1 390,400 gpm Technical Specification minimum value 21 2149m(082388)

l l . l 3.8 NEAT CAPACITY VERIFICATION FOR AUXILIARY BUILDING KVAC SYSTEM, l VA-30 l l

OBJECTIVE I The Heat Capacity Verification for .4nxiliary Building HVAC j System procedure was performed to acquire heat removal capacity l l data for the Auxiliary Building ventilation system. The data  ;

obtained in the procedure will be analyzed to verify that the  :

j design heat loads can be removed. Local temperature data was l also taken to verify VA system cooling capability.  !

TEST METHODOLOGY  !

In order to obtain data for the heat removal capacity of the

Auxiliary Building ventilation system, the HVAC system was .

l divided into subsystems and tested independently. The l temperatures, heat loads, cooling air flow rates and water flow j rates for each subsystem were measured. The data for each j subsystem will be analyzed to determine that the design heat  !

loads can be removed, j EUMMARY OF RESULTS i

The test data yielded temperatures that were analyzed to verify i the heat removal requirements for plant opcrations. All room ,

temperature readings were below their allowed maximum values  !

except the IB CV Charging Pump room. .The overall temperature of this room was less than the allowed maximum of 122'F, but two temperature acquisition pointe adjacent to the pump motor air  !

discharges were recorded as 124'F. The effect of these l localized high temperatures is under investigation by offsite  !

Engineering. All test data will be analyzed by offsite Engineering to show that this HVAC system was able to remove the i design heat loads.  :

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3.9 HEAT _ CAPACITY VERIFICATION FOR SWITCHGEAR HEAT REMOVAL VItiTILATION VX-30 012ECTIVE Subsection 9.4 of the Heat Capacity Verification for Switchgear Heat Removal Ventilation procedure was retested during the 100%

. sequence to provide additional heat removal capacity data for the switchgear heat removal ventilation system. The data obtained in the procedure will be analyzed to verify that the design heat loads can be removed.

TEST METHODOLQGI In order to obtain data for the heat removal capacity of the

switchgear heat removal ventilation system, the system was divided into subsystems and tested independently. This retest was only of the Division 12 Power Cable Spreading Room. The 1

data for the Division 12 Power Cable Spreading Room subsystem will be analyzed to determine that the design heat loads can be removed.  ;

SITMMARY OF RESl&TS Test data will be analyzed by offsite Engineering to show that this subsystem was able to remove the design heat loads. The test Acceptance Criteria of room temperatures less than or equal to 108'T was met as the maximum room temperature found was 89'F.

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3.10 - HEALCAEACITY VERIUCATION FOR _MI1CILLAllEOUS ELECTRICAL EQUIPMENT ROOM VENTILATION SYSTEM. VE-30 OBJECTIVE Subsection 9.1 of the Heat Capacity Verification for Miscellaneous Electrical Equipment Room Ventilation System procedure was retested during the 100% sequence to provide heat removal capacity data for the miscellaneous electrical equipment room ventilation system. The data obtained in the procedure will be analyzed to verify that the design heat loads can be removed.

IEST METHODOLOGI In order to obtain data for the heat removal capacity of the miscellaneous electrical equipment room ventilation system, the system was divided into subsystems and tested independently.

This retest was only of the Division 11 Miscellaneous Electrical Equipment Room and Battery Area 111. The data for the Division 11 subsystem will be analyzed to determine that the design heat loads can be removed.

SUMMARLDLEESULIS The data will be analyzed by offsite Engineering to show that the Division 11 subsystem of miscellaneous electrical equipment room (MEER) ventilation was able to remove the design heat loads. The test Acceptance Criteria of room temperatures less than or equal to 108'F was met as the maximum room temperature found wcs less than 94'F.

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3.11 - HEAT CAPACITY VERIFICATION FOR REMOTE SHUTDOWN CONTROL ROOM l VENTILATION. VI-80 1 QBJECTIVE The Heat Capacity Verification for Remote Shutdown Control Room Ventilation procedure was performed to provide heat removal capacity data for the radwaste panel area and the Unit 1 and Unit 2 remote shutdown control room's ventilation system. The i data obtained in the procedure will be analyzed to verify that the design heat loads can be removed. ,

IEST METHODOLOGY Temperatures, heat loads, cooling air flowrates, and refrigeration unit parameters were measured for the radwaste panel area and the Unit 1 and Unit 2 remote shutdown control rooms. Extrapolations to design conditions were then made to verify that the design heat load could be removed.

SUMMARY

OF RESULTS The data will be analyzed by offsite Engineering to show that this ventilation system was able to remove the design heat loads. The test Acceptance Criteria of remote shutdown control room temperatures less than or equal to 80'F was met as the maximum room temperature found was 78'F.

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l L12M0AD_SWIHG TEST NR-36 0UICI1YE This test was performed to demonstrate the dynamic recponse of the Reactor Coolant System (RCS) and the Rod Control System to automatically bring the plant to staady state conditions following a 10% reduction in turbine load, and to a 10% increase in turbine load. The test execution at 75% power was a repeat of an earlier execution per a request from offsite Engineering.

IESIJiETE0D0LOGI With plant conditions stable at the appropriate power level (75%

or 100%) a 10% step load decrease was programmed into the turbine's Digital Electro-Hydraulic Control (D"HC) system at a rate of 2350 MW per minute. The 10% decrease was initiated and plant parameters were allowed to stabilize. After stabilization, a 10% step load increase was programmed into the DEHC System and initiated. Plant parameters were again allowed to stabilize. During the course of the t?st, strip chart recordings of essential plant parameters were taken so that their responses could be analyzed. The parameters monitored included RCS Tave, Tref, pressurizer pressure and level, steam generator pressure and levels, steam and f9edwater flows, control rod positions and speed, OTWT and OPWT setpoints, reactor power, feedwater pump speeds and discharge pressure and feedwater regulating valve demand positions. ,

SUMMARY

_OF_RESULTS The load decreases .2nd increases did not cause the reactor to trip nor the turbine to trip. The steam generator safety valves and pressurizer safety valves did not lift during any of the load swings and nuclear power over/undershoot was less than 3%

in all of the executions. No manual intervention was required.

From the acceleroneter data, it was deterreined that no damaging water hammer occurred during the load swings at 100% power. No accelerometer data was taken for this 75% power retest because it had been obtained previously during the first 75% power executione See Tables 3.12-1 through 3.12-4 for additional details.

26 2149m(082388)

.. _ . .. - . . =

IeBLE 3.12-1 101_ LOAD DECREASE AT 75% POWER

SUMMARY

(REPEATED TEST) l l

INITIAL FINAL l CONDITION CQNDITION Generator Load (MWe) 890- 762 l t

Nuclear Power (%) 75 63 Tave Loop 1A ('F) 581 575 l Tave Loop 1B ('F) 581 577 f Tave Loop IC (*F) 581 577 l Tave Loop ID ('F) 580 577 Tref ('F) 580 577 j

! Delta T Loop 1A (%) 76 68  !

OFWT Loop 1A (%) 107 107 i

OTWT Loop 1A (%) 136 147

Pressurizer Pressure (psig) 2230 2230 l 1

4 Pressurizr Level (1) 52 46 Steam GenerJtor Level Loop 1A (%) 64 64  ;

s Steam Generator Level Loop 1B (%) 65 65 +

,) Steam Generator Level Loop 1C (%) 65 65 i Steam Generator Level Loop 1D (%) 65 65 '

Steam Header Pressure (psig) 1008 1008 ,

i 4 St..a Flow Loop 1A (pph) 2.9E6 2.5E6 t

, Staam Flow Loop 1B (pph) 2.8E6 2.4E6 .

Steam Flow Loop IC (pph) 2.8E6 2.4E6 i Steam Flow Loop 1D (pph) 2.8E6 2.4E6 ,

l Feedwater Flow Loop 1A (pph) 2.8E6 2.4E6 Feedwater Flow Loop 1B ( h) 3.0E6 2.6E6 >

Feedwater Flow Loop 1C ( h) 2.8E6 2.5E6 .

Feedwater Flow Loop 1D (pph) 2.7E6 2.3E6 Feedwater Temperature Loop 1A ('F) 414 403  ;

Feedwater Temperature Loop 1B ('F) 414 402 f l Feedwater Temperature Loop 1C (*F) 414 403 ,

Feedwater Temperature Loop 1D ('F) 414 403 i Teed Pump Discharge Pressure (psis) 1175 1156 ,

l i Control Bank D Position (steps) 180 151  !

Control Bank C Position (6teps) 228 228 {

Boron Concentration (ppm) 757 757 I E

27 2149m(082388) )

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TABLE 3.12-2 10% LOAD INCREASE AT 75% POWER

SUMMARY

(REPEATED TES.I.).

INITIAL FINAL CONDITIQH CONDITION Generator Load (MWe) 775 920 Nuclear Power (%) 63 77 Tave Loop 1A-('F) 577 582 d

Tave Loop 1B (*F) 577 582 Tave Loop 1C ('F) 577 582 Tave Loop 1D ('F) 576 581 Tref (*F) 577 581 Delta T Loop 1A (%) 68 79 OPWT Loop 1A (%) 107 107 f

OTWT Loop 1A (%) 147 133 Pressurizer Pressure (psig) 2230 2230

( ,

Pressurizer Level (%) 46 53 i Steam Generator Level Loop 1A (%) 64 64 l Steam Generator Level Loop 1B (%) 65 65 Steam Generator Level Loop 1C (%) 65 65  :

Steam Generator Level Loop 1D (1) 65 65 ,

Steam Header Pressure (psis) 1008 1006

. Steam Flow Loop 1A (pph) 2.5E6 3.0E6 ,

1 Steam Flow Loop 1B (pph) 2.5E6 2.9E6 ,

Steam Flow Loop 1C (pph) 2.4E6 2.9E6 i Steam Flow Loop 1D (pph) 2.4E6 2.9E6 i l l Feedwater Flow Loop 1A (pph) 2.4E6 2.9E6 ,

i Teedwater Flow Loop 1B (pph) 2.6E6 2.9E6 e 1 Feedwater Flow Loop 1C (pph) 2.5E6 2.9E6 '

j Feedwater Flow Loop 1D (pph) 2.3E6 2.7E6 ,

r 1

Feedwater Temperature Loop 1A ('F) 403 417 l I Feedwater Temperature Loop 1B ('F) 402 416 .

Feedwater Temperature Loop 1C ('F) 403 417 l Feedwater Temperature Loop 1D (*F) 403 417 Feed Pump Discharge Pressure (psis) 1156 1175 Control Bank D Position (steps) 151 220

' Control Bank C Position (steps) 228 228 Boron Concentration (ppm) 757 757 l 28

~

2149m(082388) l

, _ . - . . -_ . . _ - - _ _ _ _ _ - . - . _ _ _ _ _ __. _ _ _ _ _ ___.. _ ._. -. _ a

TABLE 3.12-3 10% LOAD DECREASE AT FULL POWER SUMMARI INITIAL FINAL CONDITION CONDITION Generator Load (MWe) 1150 1000 98 87 \

Nuclear Power (%)

Tave Loop 1A (*F) 583 580 Tave Loop 1B ('F) 583 580 Tave Loop 1C (*F) 583 580 Tave Loop 1D ('F) 583 580 Tref ('F) 584 580 j

Delta T Loop 1A (%) 98 88 OPWT Loop 1A (%) 107 107

{

OTWT Loop 1A (%) 125 125 Pressurizer Pressure (psig) 2235 2235 Pressurizer Level (%) 55 55 i Steam Generator Level Loop 1A (%) 64 64

Steam Generator Level Loop 1B (%) 65 65 Steam Generator Level Loop 1C (%) 65 65 Steam Generator Level Loop 1D (%) 65 65 i l Steam Header Pressure (psig) 956 970 Steam Flow Loop 1A (pph) 3,5E6 3.2E6 Steam Flow Loop 1B (pph) 3.6E6 3.2E6 Steam Flow Loop 1C (pph) 3.6E6 3.2E6 Steam Flow Loop 1D (pph) 3.6E6 3.2E6  ;

Feedwater Flow Loop 1A (pph) 3.6E6 3.3E6 Feedwater Flow Loop 1B (pph) 3.6E6 3.2E6 Feedwater Flow Loop 1C (pph) 3.7E6 3.3E6 j Feedwater Flow Loop 1D (pph) 3.7E6 3.2E6 ,

Feedwater Temperatare Loop 1A ('F) 435 426 1 Feedwater Temperature Loop 1B ('F) 434 425 Feedwater Temperature Loop 1C ('F) 434 426 Feedwater Temperature Loop 1D ('F) 434 426 ,

a Feed Pump Discharge Pressure (psig) 1155 1150 Control Bank D Position (staps) 203 170 Control Bank C Position (steps) 228 228 Boron Concentration (ppm) 650 650 29 2149m(082388)

T BLE 3.12-4 10% LOAD _ INCREASE _AT FULL POWER

SUMMARY

INITIAL FINAL QNDITION GHDll10N ,

Generator Load (MWe) 1000 1150 Nuclear Power (%) 87 98 Tave Loop 1A (*F) 580 583 Tave Loop 1B (*F) 580 583 Tave Loop 1C ('F) 580 583 Tave Loop 1D ('F) 580 583 Tref ('F) 581 584 i

Delta T Loop 1A (%) 90 98 OPWT Loop 1A (%) 106 106 OTWT Loop 1A (%) 126 126 4 Pressurizer Pressure (psig) 2240 2230 i Pressurizer Level (%) 52 57 Steam Generator Level Loop 1A (%) 64 64 Steam Generator Level Loop 1B (%) 65 65 Steam Generator Level Loop 1C (%) 65 65 Steam Generator Level Loop 1D (%) 65 65 Steam Header Pressure (psig) 968 956 Steam Flow Loop 1A (pph) 3.2E6 3.6E6 Steam Flow Loop 1B (pph) 3.2E6 3.6E6 Steam Flow Loop 1C (pph) 3.3E6 3.7E6  ;

Steam Flow Loop lo (pph) 3.3E6 3.7E6 Feedwatur Flow Loop 1A (pph) 3.3E6 3.7E6 F2edwater Flow Loop 1B (pph) 3.2E6 3.6E6 ,

Feedwater Flow Loop 1C (pph) 3.3E6 3.7E6 -

Feedwater Flow Loop 1D (pph) 3.2E6 3.6E6 Feedwater Temperature Loop 1A ('F) 426 434 I Feedwater Temperature Loop 1B ('F) 425 433 .

Feedwater Temperature Loop IC ('F) 426 434 Feedwater Temperature Loop 1D ('F) 426 434 l

Feed Pump Discharge Pressure (psig) 1150 1155 Control Bank D Position (steps) 170 223 l Control Bank C Position (steps) 228 228 i Boron Concentration (ppm) 650 650 30 2149m(082388)  !

i l

3.13 - HILLE0WER PLANT TRIP. NR-39 OBJECTIVu Thin test was performed to verify the ability of the primary and secondary plant and the plant automatic control systems to suttain a generator trip from full power and to bring the plant to stable conditione following the transient. The response time of the reactor coolant hot I?g Narrow Range RTDs and the amount of turbine overspeed that occurred upon a generator trip were also determined.

IEST METHODOLOGY From a stable plant condition at approximately full power, a generator trip was initiated by opening the main generator output breakers. This caused a turbine trip and a reactor trip. The operators were instructed to follow the regular Station Emergency Procedures to bring the plant to stable conditions. The data trending was terminated whr4 Tr.ve was stabilized at approximately 557'F (no load Tave).

SUMMARY

_0F RESE IS All of the test Acceptance Criteria were met. The response time of the Hot Leg 'iarrow Range RfDs for Loop 1A (without pressurizer) was 7.0 seconde and for Loop 1D (with pressurizer),

8.2 seconds. This satisfied the tiiae response requirements for Loop 1A of 1 7.8 seconds and for Loop 1D of 1 9.2 seconds.

These response times also resolved concerns regarding inadequate RTD manifold flowrates in the BwSU RC-30 teet. The RC-30 test acceptance was deferred to performance of this NR-39 test.

Because the RTD response times determined by the NR-39 test were adequate, the RTD uanifold flowrates were judged to be acceptable. The maximum turbine speed was found to be 1925 rpm. There was no unacceptable water hammer experienced and safety injection was not initiated. No pre;surizer safety valves lif ted, and no steam generator safety vaJves lif ted.

Several plant parameters exceeded the expected rang.ms for this test but were all found acceptable in test review. See Table 3,13 for additional details.

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2149m(082388) l

TABLE 3.13 FULL POWELPLANT TRIP

SUMMARY

INITIAt. FINAL CONDYTION CONDITION Generator Load (MWe) 1208 25 Nuclear Power (%) 99 0 Auctioneered High Tave (*F) 586 558 Tref (*F) 586 557 Deltc.T Loop 1A (%)* 104 4 OPWT Loop 1A (%) 108 108 OTWT Loop 1A (%) 124 >150 Pressurizer Pressure (psig) 2231 2235 Pressurizer Level (%) 58 27 Steam Generator Level Loop 1A (%) 64 50 Steam Generator Level Loop 1B (%) 66 50 Steam Generator Level Loop 1C (%) 66 48 Steam Generator Level Loop 1D (%) 65 50 Steam Header Pressure (psig) 965 1095 Steam Flow Loop 1A (pph) 3.9E6 0.4E6 Steam Flow Loop 1B (pph) 3.0E6 0.0E6 Steam Flow Loop 1C (pph) 3.8E6 0.0E6 Steam Flow Loop 1D (pph) 3,9EG 0.0E6 Feedwater Flow Loop 1A (ppb.) 3.7E6 0.0E6 Feedwater Flow Loop 13 (pph) 3.8E6 0.0E6 Feedwater Flow Loop IC (pph) 3.736 0.0E6 3,7E6 0.0E6 Feedwater Flow Loop 1D (pph)

Feedwater Temperature Loop 1A ('F) 436 75 Feedwater Temperaturs Loop 1B (*F) 434 75 Feedwater Temperature Loop 1C ('F) 434 75 Feedwater Temperature Loop ID ('F) 435 75 Feed Pump Aiccharge Pressure (psi s) 1160 1060 Control Bank D Positt>n (stapt') 216 0 Ttro'

. Bank C Pocition (steps) 228 0 Boron Concentration (ppm) 697 714

• Delte T was conservatively scaled at the time of test such that indicated delta T was greater than true reactor power 32 2.149m(082388)

-.- -, , . , . - , c -. . . , - g .a,.~

~ , , . - - ~ . - - - - ,m., .-,. , - - ,

l 3.14 - LARGE LOAD REDUCTION. NR-31 l

OBJECTIVE l l

the ability of the plant The purpose to sustain andofstabilize this test was to verify following a 50% step load reduction from approximately 100% power without operator intervention.

Acceptance criteria focused on the plant response during the test: no reactor or turbine trip, no damaging water hammer transients, no ESF actuations and no lifting of pressurizer or steam generator safety valves.

IEST METHODOLOGY With the plant stabilized at approximately full power, the DEHC system was programmed for a 50% (588 MWe) turbine load reduction at a 2350 MWe/ minute load reduction rate. The load reduction was initiated and key plant parametcrs were monitored during the transient. The parameters monitored included feedwater and reactor coolant system temperatures and pressures, reactor power, steam generator and pressurizer water levels and pressures, control rod positions, boron concentration, electrical output, steam and feedwater flows and feedwater regulating valve demand positions.

SUMMARY

OF RESULTS Following the load ter:ction Tave stabilized in 10 minutes.

Manual emergency boration was require / ring the test to respond to the Lo-2 Rod Insertion Lima ilarm. Approximately 300 gallons of boric acid were add (1- ha other manual actions were required. This action was deemed acceptable by test review. The transient did not cause either the reactor or the turbine to trip. Safety valves on the pressurizer and steam generators did not lift. Accelerometer data indicated no damaging water hammer occurred. There were no ESF actuations.

See Table 3.14 for additional details.

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l 33 2149m(082388)  ;

Q.

TABLE 3.14 50% LOAD REDUCTION AT FULL POWER SU121ARY INITIAL FINAL CONDITION CONDITION Generator Load (MWe) 1165 580 Nuclear Power (%) 100 49 Auctioneered High Tave ('F) 586 569 Tref (*F) 386 570 Delta T Loop 1A (%) 102 58 OPWT Loop 1A (%) 107 107 OTWT Loop 1A (%) 122 >150

~

Pressurizer Pressure (psig) 2230 2230 Pressuriser Level (%) 55 38 Steam Generator Level Loop 1A (%) 65 65 Steam Generator Level Loop 1B (%) 65 65 Steam Generator Level Loop 1C (%) 65 65 Steam Generator Level Loop 1D (%) 65 65 Steam Header Pre. sore (psig) 972 1015 Steam Flow Loop 1A (pph) 3.7E6 1.9E6 Steam Flow Loop 1B (pph) 3.8E6 1.9E6 Steam Flow Loop 1C (pph) 3.7E6 1.8E6 Steam Flow Loop 1D (pph) 3.7E6 1.8E6 Feedwater Flow Loop 1A (pph) 3.7E6 1.9E6 Feedwater Flow Loop 1B (pph) 3.8E6 1.9E6 Feedwater Flow Loop 1C (pph) 3.7E6 1.9E6 Feedwater Flow Loop 1D (pph) 3.7E6 1.8E6 Feedwater Temperature Loop 1A ('F) 436 384 Feedwater Temperature Loop 13 (*F) 436 384 Feedwater Temperature Loop 1C (*F) 436 384 Feedwater Temperature Loop 1D ('F) 436 384 Feed Pump Discharge Pressure (psig) 1180 1140 Control Bank D Position (steps) 221 154 Control Bank C Position (steps) 228 228 Boron Concentration (ppm) 649 649 34 2149m/^82388)

3.15 - CALIBRATION OF STEAM AND FEEDWATER FLOW. FW-31 OBJECTIVE The Calibration of Steam and Feedwater Flow test was performed during power ascensica to verify that the feedwater flow and steam flow instrvmentation was properly calibrated.

TEST METHODOLOGY With the plant stable at approximately 0%, 30%, 50%, 75%, and 100% power, plant parameters concerning steam and feedwater flows were recorded. These parameters included steam flows, feedwater flows, feedwater tempering flows, feedwater temperatures, and steam generator pressures. Differential pressure gauges were installed on the precision main feedwater flow venturies to accurately ceasure feedwater flows. Steam generator blowdown was isolated for the duration of the test.

The precision feedwater flow measurement was then compared to the electrical output of the steam and feedwater flow transmitters and square root extractora. The acceptance of this startup test was based on these comparisons.

SUMMARY

OF RESULTS During the executions of this test a number of the compari ons failed to meet the stringent test acceptance criteria. These differences were either corrected through transmitter and instrument loop rescaling and recalibrations or remeasured with refined data aq'tisition techniques. This test executed during the 100% power sequence provided the final calibration verifications based on actual full powe' flow data and additional data taken at lower power levels. Data acquired at  !

lower cower levels served to scale steam and feed'aater flow '

instrume'6ation progressively during power ascension using extrapolaced values.

35 2149m(082388)

3.16 - Tli3EMAL POWER MEASUEEMENT AND STATEPOINT DATA COLLECTION.

IT-32A/B/_C/D/E QBJECTIVE The three basic objectives of this test series were to periodically determine thermal power using calorimetric data, collect control and protection instrumentation data at steady state power levels (statepoints), and to verify and align WT and Tave instrumentation. There were no acceptance criteria for the tests in this series. Table 3.17 lists the various tests and intent of each. The IT-32A, C and D tests were repeats of previously performed tests following RCS Narrow Range RTD replacement. The IT-32B test was performed at 98% power prior to the RTD replacement and was not repeated because it did not acquire statepoint data. The IT-32E test was performed following RTD replacement.

TEST METHOD 0LQGI Statepoint data was collected at the approximate 50%, 75%, and 100% power levels with a WT instrumentation alignment done at 75% power. Thermal power calorimetrics were performed at the approximate 50, 75, 98 and 100% power levels. Calorimetric data included feedwater temperature and main feedwater flow venturi WP, steam pressure, atmospheric pressure, feedwater tempering flow and steam generator blowdown flow. The IT-32A test does not take calorimetric or statepoint data and was performed to verify RTD calibration accuracy.

For calorimetric measurements three readings were taken of each of the parameters within a 20 minute time period to assure good quality of calorimetric calculation input values. Alignment of the WT process instrumentation was accomplished by plotting these values gathered at various power levels as a function of the respective calculated calorimetric power. Trend lines fitted through the points were extrapolated to 100% power to predict the 100% WT and Tave values.

SUMMARY

OF RESULTS Power ascension testing during power ascension to 100%

progressed well with only minor instrument anomalies which were documented and resolved. These out of expected range anomalies were in general due to iterative calibrations during power ascension and a conservatism of 1.5'F in the initial scaling of full power WT prior to this post RTD outage power ascension.

The four RCS loop full power WTs were rescaled at 75% power as I follows:  !

WT ('F)

LOOP 1A 56.8 j LOOP 1B 56.4 LOOP 1C 57.4 LOOP 1D 57.7 36 l 2149m(082388) j

~

IABLE 3.16 IHERMAL POWER MEAS.UREMENT AND STATEPOINT DATA COLLECTION SUMM6El APPROXIMATE INST. ALIGNMENT IESI IHERMAL PWR STATEP_0 INT _ DATA OR ALIGN. CHECK CALORIMEIRIC IT-32A 27%* X IT-32C 50% X X IT-32D 75% X X X IT- 32B 987. X IT-32E 100% X X

• Performance of IT-32A is independent of plant power level. The test was actually performed at 27% power..

7 37 '

2149m(082388)

3.17 - INCORE THERMOCOUPLE /RTD CROSS CALIBRATION. IT-31 OBJECTIVE The objective of the test was to verify reactor coolant system RTD performance over a range of temperatures. Additionally a cross calibration data base was developed for the incore thermocouple system. This test execution was a repeat of a previously performed test following replacement of RCS Narrow Range RTDs.

TEST METHODOLOGI Data was gathered at four temperature plateaus; 250', 340', 450' and 557'F. The plant conditions required at each plateau included RCS temperature stabilization allowing for no more than 10.5'F drift, steam generator levels within 104 of one another, all four reactor coolant pumps running and the plant in either mode 3 or mode 4. For the RTDs, four sets of data were taken at each plateau by monitoring RTD resistance with a four wire resistance meter. For the thermocouples, two sets of data were taken at each plateau from indicators located in the main control room and process computer trending logs. At each plateau a correction factor was calculated based on the average of the narrow range RTDs.

SUMMARY

OF RESJEIji Calculations of the difference between RCS stabilized temperature and each RTD's temperature calibration curve were performed. The RTD calibration curves were developed using preliminary calibration data derived by Westinghouse from manufacturer's data along with test data. All narrow range RTDs were found to be within acceptable limits.

The wide range RTDs and thermocouples were found to provide acceptable temperature indication. Two of the sixty-five thermocouples were found to be greatly in error and were removed from scan. Sufficient thermocouples remain to support unit operation. Two wide range RTD elements were found to be in error, but do not provide temperature interlock functions or main control board displays. They will be repaired when practical to do so.  !

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3.18 - OPEEAIl011AL ALIGNMENT OF EXCORE NUCfFAR INSTRUMENTATION.

NR-34D/E/F QBJECTIVE This series of tests was. performed to verify that the excore nuclear instrumentatic.: tystem was functioning per design and capable of detecting alar.71ng and mitigating reactivity excursions.

TEST METHODOLOGY Selected parameters and alarms were evaluated, monitored, and determined during variors testing phases.

Test NR-34D. DuriD&_fower Escalation. The overlap between the IR and PR channels pas determintd, the high level trip setpoints for the PR channels were verified, and the PR detector response versus core power was checked for linearity by aligning the channels with the calorimetric thermal power as calculated in the associated Thermal Power Measurement test.

Imat_NR-34E. At Full Power. The characteristic plateau curves for the IR and PR channels were determined and the operating voltages were set and verified to be in the plateau region of these curves.

lesLNR _34F. Af tRLShutdown froIn Powe_r_QgelatioRS_QLat_J,fSSi 800_ MWD /MTU. The operating voltages for the SR channele and the compensating voltages for the IR channels were determined and set following the Full Power Plant Trip test. ,

SUMMARY

OF RESULTS The setpoints of the PR channel trips were verified as meeting all associated acceptance criteria. A minimum overlap of 1.5 decades was observed on all IR/PR channel combinations.

Specifically, the overlaps for all four IR N35/PR channel coubinations were observed to be 1.8 decades and for all four IR N36/PR channel combinations .7 decades. All detector operating voltages were verified to be withi. the expected ranges and fell well within the plateau region of the characteristic curves.

All Power Range and Intermediate Range detector high voltages werc 9et at 800 1 5 VDC. Source Range detector high voltages were set at 1810 VDC (N31) and 1805 VDC (N32). The compensating  ;

voltages for the IR detectors were set at -20.41 VDC (N35) and

-21.33 VDC (N36).

39 1 2149m(082388)

l 3.19 - AXIAL FLUX DIFFERENCE INSTRUMENTATION ADJUSTMENT AT FULL )

POWER.NR-35H 1 QAJECTIVE This test was performed to verify that the 100% calibration values extrapolated from the most recent lower power level data were within 11% accuracy when compared to actual full power data. If not, then fine tuning adjustments would be made to correct each nuclear instrumentation channel to be within 11% of the actual full power values.

TEST METHODOLOGI A flux map taken at approximately fiill pnwer was used to gather actual full power values for core sverage axial offset, average power level, and top and bottom p*.wer r:ange detector currents.

The equations for expected detector cu:: rent as a function of axial offset developed at the most recent axial flux difference calibration were employed using the actual 100% axici ;*fset value from the full power flux map results to predict 100%

detector currents. The actual extrapolated full power detector currents were then compared to the predicted full power detector currents. If different by more than 11%, calculations of new expected calibration currents would be performed using the actual full, power data combined with the most recent calibration results.

SUMMARY

OF RESULTS The flux map us0d was taken at 99.5% power. The actual 99.5%

power detector currents deviated from the predicted 99.5% power currents as extrapolated from lower power calculations by greater than 11%. The calculations of new calibration currents were performed and the fine tuning adjustments made.

)

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l 3.20 - REACTOR LOOSE PARTS MONITORING. LM-30 OBJECTIVE The test objective was to gather background noise frequency response data at full power. This data will be used as a reference baseline for alarm setpoints and when analyzing suspected loose parts in the NSSS.

TEST METHODOLOGY At each power level throughout power ascension a background noise recording of each of the 24 loose parts accelerometer channels was made. This full power test execution'provided the final set of background noise data.

SUMMARY

OF RES11TI Data from all 24 channels was obtained at full power. The data collected in the test provided a baseline for each of the 24 accelerometer channels.

Data was previously not obtained for channel 20 during the LM-30 tests conducted in the 30%, 50%, 75% and 90% test sequences.

This channel was repaired and the required data was retaken during the power ascension back to 100% power.

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3.21 - STARTUP ADJUSTMENTS OF REACTOR CONTROL SYSTEMS. RD-JS OBJECTIVE The .bjective of this test was to determine the Tave program nhich would result in the design full load steam pressure. This would ensure optimum plant eff:.ciency without exceeding pressure limitations for the turbine or temperature limits with respect to maximum allowab) Tave. This was accomplished by rescaling the reference Tave (Tref) program and turbine impulse pressure instrumentation as necessary.

TEST METHODOLQGI Data for this test was obtained from executions of the Thermal Power Measurement and Statepoint Data Collection tests, IT-32D at 75% and IT-32E at the 100% power level. Data utilized for this tesc included RCS Tave, Thot. Tcold, steam generator pressure, and turbine impulse chamber pressure. At 75% power, the available data was extrapolated to 100% to determine if any Tref or impulse pressure rescaling would be necessary. The same data was reviewed at 100% power for any additional required rescaling. If any scaling changes were made, additional plant data would be taken to verify the adequacy of the scaling.

SUMMARY

OF RESULTS Following replacement of RCS narrow range RTDs and adjustments of Tref made outside of the testing program, the RD-80 test was repeated at 75% power. At 75% power, Tave was extrapolated to a full power value of 585'F which was verified to be below the design maximum of 588.4'F. Steam generator ytsssure was extrapolated to a full power value of 987 psta which matched the design value of 990 i 10 psia. Turbine impulse chamber pressure was extrapolated to 660 psig at 100% power (792 psig at 120%

power). Because of the acceptable extrapolations, ne control system adjustments were made.

At full power, Tave was found to be 586.0'F which was below the design maximum of 588.4'F. Steam generator pressure was 998 psia which matched the design value of 990 i 10 psia. Turbine impulse chamber pressure was found to be 675.8 psig at 100%

power (811 psig at 120% power). This 100% power turbine impulse pressure of 675.8 psig was found to deviate by more than i 10 psig from the 660 psig instrument scaling. Impulse pressure was rescaled to this new value and test data was retaken. The final 100% power Tave was found to be 585.0'F, Steam Generator pressure was found to be 990 psia, and impulse pressure 674 psig. These values satisfied all applicable criteria.

42 2149m(082388)

o

, e 3.22 - NUCLFAR STEAM SUPPLY STEAM ACCEPTANCE TEST. TG-80 i QBJECTIVE This test was performed to demonstrate the reliability of the i Nuclear Steam Supply System (NSSS) to maintain its warranted output of 3425 MWth (+0, -5%) for 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> without a load reduction or plant trip resulting from an NSSS malfunction.

TEST METHODOLOGY ,

The test was initiated with the plant operating within 5% of its rated NSSS output as determined by a calorimetric calculation. .

Plant conditions were stabilized at their design values for t 35-65 hours with the stabilization verified by calorimetric data acquisition over two consecutive half-hour periods. A four hour thermal performance measurement was then initiated to determine actual reactor power by collecting calorimetric data every 5 minutes and computing power hourly. The remaining hours of the test were then completed to make a total of 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br />.

SUMMARY

OF RESULTS l Steam generator moisture carryover measurements have not yet been made for Braidwood Unit 1. The four hourly calorimetrics demonstrated NSSS output to be an average of 3396 MWth ,

correcponding to 99.2% power assuming no moisture carryover  !

(100% steam quality). The four hourly calorimetries demonstrated NSSS output to be an average of 3388 MWth corresponding to 98.94 power at a maximum warranted moisture carryover of 0.25% (99.75% steam quality). The minimum duration of 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> was met with no load reductions below 95% power and no plant trip, t i

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0 3.23 - PROCESS COMPUTER VERIFICATION - THERMAL POWER. CX-80 OBJEGIIYE The Process Computer Verification of Thermal Power was performed to verify the ability of the process computer to receive and reduce calorimetric data.

IEST METHODOLOGY The process computer verification of thermal power was performed by performing a precision secondary plant calorimetric measurement while executing the process computer calorimetric program and comparing the two outputs.

SUMMARY

OF RESULTS The acceptance criteria for this test consisted of a i 2% power agreement between the precision calorimetric performed by hand and the computer calorimetric power results. The computer results met the acceptance criteria in comparison with the precision calorimetric results at 98% power. The comparison showed a difference of 0.51%. Based on the results of this test at full power, this computer program was accepted as satisfactorily demonstrated.

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3.24 - PROCESS COMPUTER VERIFICATION - INCORE THERMOCOUPLE. CX-81 OBJECTIVE The Process Computer Verification of Incore Thermocouples was performed to verify the ability of the process computer to receive and reduce data from the incore thermocouples.

TEST METHODOLOGY The process computer verification of incore thermocouples was performed by taking a core exit thermocouple map. Thermocouple temperature distributions were than used to hand calculate fuel assembly relative powers based on enthalpy changes and to calculate core power tilts. These hand calculated thermocouple based results were compared against similar thermocouple calculations performed by the process computer.

SID2MRY OF RESULTE The acceptance criteria for this test consisted of a 1 0.01 agreement between the hand calculated values and the computer calculated values. The Acceptance Criteria were met at full power. The comparison showed a maximum difference of 0.002.

Based on the results of this test at full power, this computer program was accepted as satisfactorily demonstrated.

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3.25 - PROCESS COMPUTER VERIFICATION - BORON FOLLOW. CX-82 l OBJECTIV.E The Process Computer Verification of Boron Follow was performed to ve- / the ability of the process computer to monitor Reactor Coolt..c System (RCS) Boron and Lithium Concentrations changes.

IEST METHODOLOGY The process computer verification of boron follow was performed by taking RCS chemistry samples, analyzing them for boron and lithium and inputting these initial values into the computer program. Following an RCS chemistry change, samples were again taken and analyzed in the laboratory. These final chemistry results were compared to the final values from the computer program.

SUMMARY

OF RESULTS The acceptance criteria for this test consisted of a t 5% boron and lithium concentration agreement between the laboratory data and the computer data. The test acceptance criteria of 15% was met for this test execution at full power. The comparisons showed a maximum boron deviation of 3.6% and a maximum lithium deviation of 4.2%. Based on the results of this test at full power, this computer program was accepted as satisfactorily demonstrated.

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o o e i 4 3.26 - PROCESS COMPUTER VERIFICATION - FLUX MAPPING. CX-83 DBlEC.T1.YE The Process Computer Verification of Flux Mapping was performed to verify the ability of the process computer to receive and reduce data from the incore movable detectors.

IEST METHODOLOGY The process computer verification of flux mapping was performed by taking one pass of a flux map and comparing the voltages received at the process computer against the traces on the strip chart recorders at the movable detector panel. For the selected pass the strip chart recordings for all six detectors were hand digitized using mylar overlays. These hand digitized values were compared against the raw voltages recorded by the computer movable detector software.

SUMMARY

OF RESULTS The acceptance criteria for this test consisted of a 37. of full scale voltage agreement between the hand digitized data and the computer data. The test acceptance criteria of 13% was met at full power. The comparisons showed a maximum deviation of 2.6%. Based on the results of this test at full power, this computer program was accepted as satisfactorily demonstrated.

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- e 4.0 - REFERENCES

1) Braidwood Station Final Safety Analysis Report

2) Regulatory Guide 1.68, Revision 2

3) Braidwood Station Technical Specifications l l

4) Braidwood Station Operating License NPF-70

5) Braidwood Station Operating License NPF-72

6) WCAP 10935, Core Physics Parameters and Plant Operations Data for the Braidwood Generating Station Unit 1, Cycle 1

7) Westinghouse NSSS Startup Manual

8) Byron Unit 1 Cycle 1 Startup Report

9) Byron Unit 2, Cycle 1 Startup Report

10) Braidwood Unit 1, Cycle 1 Startup Report (Partial Report February 1988)

11) Braidwood Unit 1, Cycle 1 Startup Report Supplement (May 1988) l 4

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