ML20205M241

Commonwealth Edison Co,Byron Unit 1,Cycle 1,Startup Rept

ML20205M241
Person / Time
Site:	Byron
Issue date:	10/25/1985
From:	Querio R COMMONWEALTH EDISON CO.
To:	James Keppler NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
References
#24-032, #24-32, BYRON-85-1421, NUDOCS 8604150135
Download: ML20205M241 (152)

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COMMONWEALTH EDISON COMPANY BYRON UNIT 1 CYCLE 1 3 STARTUP REPORT 3

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8604150135 851025 DR ADOCK 0500 g"4 3 (0851M/0080M)

O ACKNOWLEDGEMENTS The following people aided in the preparation of this report:

S. Altmayer R. Campbell O

R. Choinard C. Kilbride D. Parr B. Koehler P. Reister T. Melloch O

S. Shaneyfelt K. Passmore M. Simmons M. Robinson G. Stauffer J. Ventimiglia O

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O-TABLE OF CONTENTS Section Title Page List of Tables 4 O

List of Figures 5 1.0 Introduction 6 2.0 Discussion of Byron Startup Program 7 3.0 Discussion of Byron Startup Tests 24 3.1 Core Loading 26 3.2 Post Core Loading System Testing 42 3.3 Physics Testing 93 3.4 Transient Testing 122

.5 Instrumentati n and Calibration Testing 139 O

4.0 References 151 O

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O LIST OF TABLES TABLE TITLE PAGE 2.0-1 Byron Unit 1 Major Milestone 8 O

2.0-2 Byron Operational Modes 9 2.3-1 HZP Flux Map Results 13 2.4-1 30% Power Flux Map Results 15 2.5-1 50% Power Flux Map Results 17 2.6-1 75% Power Flux Map Results 19 2.7-1 90% Power Flux Map Results 21 2.8-1 100% Power Flux Map Results 23 3.0-1 List of Test Summaries 24 3.3.4 Measured vs. Predicted Isothermal Temperature Coefficient 97 3.3.5-1 Measured vs. Predicted Doppler Only Power Coefficient 99 3.3.9-1 Measured vs. Predicted Bank Worths 111 3.3.10-1 Measured vs. Predicted Boron Endpoint Concentrations and Boron Reactivity Worths 119 3.4.2-1 10% Load Decrease at 35% Power Summary 125

.4.2-2 10% Load Increase at 35% Power Summary 0 126 3.4.2-3 10% Load Decrease at 75% Power Summary 127 3.4.2-4 10% Load Increase at 75% Power Summary 128 g 3.4.2-5 10% Load Decrease at 100% Power Summary 129 3.4.2-6 10% Load Increase at 100% Power Summary 130 3.4.3-1 Trip from 100% Power Summary 132 0 3 4 4-1 Trip from 25% Power Summary 134 3.4.6-1 Large Load Reduction at 75% Power Summary 137 3.4.6-2 Large Load Reduction at 100% Power Summary 138 O

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O LIST OF FIGURES FIGURE TITLE PAGE 3.1.1-1 Randomized Core Loading Pattern 28 0

3.1.1-2 Randomized Core Loading Pattern 29 3.1.1-3 Randomized Core Loading Pattern 30 3.1.1-4 Randomized Core Loading Pattern 31 O

3.1.1-5 Randomized Core Loading Pattern 32 3.1.1-6 Randomized i3 ore Loading Pattern-Final Configuration 33 3.1.1-7 Shutdown and Control Rod Locations 34 0

3.1.1-8 Burnable Poison Rod Locations 35 3.2.20-1 Pressurizer Heaters Effectiveness 73 3.2.20-2 Pressurizer Spray Effectiveness 74 0

3.3.6-1 ICRR vs. Shutdown Bank Position During Approach to 102 Criticality 3.3.6-2 ICRR vs. Control Bank Position During Approach to 103 Criticality 3.3.6-3 ICRR vs. Primary Water Added During Approach to Criticality 104 3.3.6-4 Source Range Detector Response During Approach to 105 Criticality 3.3.6-5 Intermediate Range Detector Response During Approach to 106 Criticality

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3.3.9-1 Control Bank D Integral Worth vs. Steps Withdrawn 112 3.3.9-2 Control Bank C Integral Worth vs. Steps Withdrawn 113 3.3.9-3 Control Bank B Integral Worth vs. Steps Withdrawn 114

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3.3.9-4 Control Bank A Integral Worth vs. Steps Withdrawn 115 3.3.9-5 Differential and Integral worth of Control Banks in Overlap 116

) 3.3.9-6 Differential Boron Worth over the Control Banks 117 g' (0851M/0080M)

l 0 - INTRODUCTION This report describes the required testing at Byron Station Unit 1 from the time the first fuel assembly was loaded into the reactor until the plant was placed in operation. It satisfies the requirement of the Byron

) Unit 1 Technical Specifications that a Startup Report be submitted to the NRC after completion of the Startup Testing Program.

Byron Station, located in northern Illinois, utilizes a four loop Westinghouse pressurized water reactor system. Westinghouse Electric Corporation, Sargent & Lundy, and the Commonwealth Edison Company jointly

() participated in the design and construction of Byron Unit 1. The plant is operated by Commonwealth Edison Company with Sargent & Lundy as the Architect - Engineer.

The nuclear steam supply system is designed for a power output of 3411 MWt. The equivalent warran,ted gross and approximate net electrical

() output are 1175 MWe and 1120 MWe, respectively. Cooling for the plant is provided by a natural draft cooling tower for nonessential service cooling, and by mechanical draft cooling towers for essential cooling,

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M 2.0 - DISCUSSION OF BYRON STARTUP PROGRAM 1 1

The Byron Unit 1 startup testing program consisted of single and multisystem tests that occurred commencing with initial fuel loading and continuing through full power. These tests demonstrated overall plant j performance and included such activities as precritical testing, low  ;

() power tests, and power ascension tests. Testing sequence documents were 1 utilized for each plateau to coordinate the sequence of testing i activities at that plateau.

1 In the subsections that follow, a description of the sequence of testing j

() at each plateau is provided. The descriptions include additional details concerning special license conditions and commitments made to the Nuclear j

Regulatory Commission prior to completion of the startup testing program, )

where applicable. Also included as a part of Section 2.0 is a table l showing major milestones for Byron Unit 1 which occurred during the I startup program and a list of operational modes as defined by the  !

,() Technical Specifications. l i

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BYRON UNIT 1 MAJOR MILESTONES O

MAJOR MILESTONES DATE 5% Power License Issued 10/31/84 Fuel Load 11/02/84 O

Initial Criticality 02/02/85 Full Power License Issued 02/14/85 5% Tests Completed 02/20/85 O

Entered Mode 1 02/24/85 Initial Synchronization to Grid 03/01/85 30% Test Completed 04/15/85 O i Plant In Service

• 04/22/85 50% Test Completed 05/23/85 75% Test Completed 06/18/85 90% Test Completed 06/22/85 100% Test Completed 09/10/85 Commercial Operation 09/16/85

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O TABLE 2.0 - 2 OPERATIONAL MODES O

REACTIVITY  % RATED AVERAGE COOLANT MODE CONDITION, K,gg THERMAL POWER

• TEMPERATURE

1. POWER OPERATION > 0.99 > 5% > 350*F o

2. STARTUP > 0.99 i 5% > 350*F

3. HOT STANDBY < 0.99 0 > 350*F

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

5. COLD SHUTDOWN < 0.99 0 1 200*F

6. REFUELING ** i 0.95 0 1 140*F O

• Excluding decay heat.

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

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O 2.1 - lurTIAL cons toAn snousucs The Core Load Sequence Document was utilized to coordinate the sequence

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of operations associated with the initial core loading program. This sequence included scheduling of the individual'startup tests associated g with core loading. This document specified as prerequisites which testing had to be completed prior to commencement of core loading, the required status of the plant systems necessary to support core loading, and the reactor vessel status. A log was also included in the sequence document to verify containment evacuation alarm operability throughout core loading. This document also provided the criteria for stopping core g loading, the criteria for emergency boration, and the actions to be followed prior to the resumption of core loading, in the event core loading was stopped prior to completion. Results of individual tests completed during the core loading sequence are discussed in Section 3.1 of this report. Upon completion of core loading, plant systems were aligned as directed by the Shift Engineer.

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O 2.2 - POST CORE LOADING PRECRITICAL TEST SEQUENCE (PCLPC)

The PCLPC Sequence Document was utilized to define the sequence of tests and operations to be performed between completion of initial core loading and prior to initial criticality. This document ensured that core load C) testing had been successfully completed and results approved prior to continuation of the testing program. This document scheduled the performance of precritical tests to ensure the necessary testing was completed prior to initial criticality. Plant operating procedures were utilized where appropriate to establish necessary plant conditions.

Results of individual tests completed during the post core load

) precritical testing phase are discussed primarily in Section 3.2 of this report. Upon completion of this testing phase, plant systems were restored as directed by the Shift Engineer.

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O 2.3 - INITIAL CRITICALITY & LOW POWER TEST SEvumau s (IC & LPT)

The IC & LPT Sequence Document was utilized to define the sequence of tests and operations, beginning with initial criticality, which constituted the low power testing program. This document ensured that g post core loading procritical testing had been completed and results approved prior to continuation of the testing program. Prior to commencement of dilution to initial criticality, source range nuclear instrumentation channels were verified to have a signal to noise ratio greater than 2 and power range high level trip setpoints were conservatively set to 20 1% of full power. Plant operating procedures lg were utilized where appropriate to establish plant conditions.

This sequence document performed a comparison of selected control room indication with the process computer to demonstrate computer monitoring capability. This document obtained a full core flux map with all rods out, aflux map with control barJc D fully inserted with remaining banks O withdrawn and a flux map with the control banks at their hot zero power insertion limits. (Refer to Table 2.3-1 for a tabulation of the flux map results obtained.) This procedure also measured the worth of RCCA F-10, the predicted most reactive stuck rod, and shutdown banks A and B. This is discussed in Section 3.3.12. Results of individual tests completed during the initial criticality and low power test sequence are discussed O primarily in Section 3.3 of this report.

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

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TABLE 2.3-1 HZP FLUX MAP RESULTS ACTUAL MAXIMUM O

Peak Linear Heat Rate (KW/ft) 0.375 12.6 Extrapolated to 100% 12.514 12.6 FDHN 1.555 2.001 O Fxy - rodded N/A N/A Fxy - unrodded 1.6733 1.8507 Quadrant Po.ser Tilt Ratios .9812 1.0095 1.02

.9967 1.0126 O

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O 2.4 - Test sEoUENCE AT 30% POWER The Test Sequence at 30% Power was utilized to define the activities which constituted the startup testing program from Hot Zero Power, after i completion of low power testing, up to and including the 30% power testin9 program. This document ensured that the Initial Criticality and O

Low Power Test Sequence had been completed and results approved prior to entry into Mode 1. prior to increasing power for this test sequence, power range high level trip setpoints were conservatively set to 49 1 1%

of full power and reactor core flux map results from a low power map were verified acceptable. The results were also extrapolated to 50% power to O ensure parameters indicative f DNBR and linear heat rate were acceptable for power ascension to the 30% testing plateau, Refer to Table 2.4-1 for a tabulation of flux map results. Also prior to increasing power above 5% it was verified that the high radiation sampling system panel had been demonstrated to meet the requirements of NUREG-0636.

g Until a precision calorimetric could be performed in the 20-30% power range, reactor power level was monitored by using reactor coolant system delta temperature indication where 100% power was conservatively equated to a 51*F core temperature difference. plant operating procedures were utilized where appropriate to establish plant conditions and to change reactor power. During ascension to the 30% plateau, power was stabilized O near the 5%, 10%, and 20% levels to accommodate testing at these plateaus. Results of individual tests completed during power ascension to and while at the 30% plateau are discussed in Sections 3.2 through 3.5 of this report.

At the 30% power testing plateau, this sequence document performed a O comparison of selected control room indications with the process computer to demonstrate computer monitoring capability. Work requests were initiated as appropriate to resolve discrepancies between computer points and control board indications that failed to agree to within 3% of each other. Data was also taken at the 30% plateau and reviewed to demonstrate the operability of the inadequate core cooling monitoring O system. Reactor cavity sump leakage data was obtained and a leakage rate of 0 gpm was determined. Power was reduced to the 25% level and a plant trip test was performed to complete testing for this sequence. The plant was then restored as directed by the Shift Engineer.

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TABLE 2.4-1 30% FLUX MAP RESULTS g ACTUAL MAXIMUM '

Peak Linear Heat Rate (KW/ft) 3.430 12.6 Extrapolated to 100% 11.432 12.6 FDHN 1.4535 1.8755 Fxy - rodded 1.6129 1.9494 Fxy - unrodded 1.6007 1.7670 Quadrant Power Tilt Ratics .9876 1.0062 1.02

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9 2.5 - TEST SEvura,m. AT 50% POWER 1

The Test Sequence at 50% Power was utilized to define the activities which constituted the startup testing program during escalation from 30%

to 50% power and at approximately 50% of rated thermal power. This document ensured that the Test Sequence at 30% Power had been completed

'O and the results approved prior to increasing power above the 30% testing plateau. Prior to increasing power for this test sequence, power range high level trip setpoints were conservatively set to 69 t 1% power and reactor core flux map results from a 30% power baseline map were verified acceptable. The flux map results were also extrapolated to 70% power to

{) ensure parameters indicative of DNBR and linear heat rate were acceptable for power ascension to the 50% testing plateau. Refer to Table 2.5-1 for j a tabulation of flux map results.

! Plant operating procedures were utilized where appropriate to establish plant conditions and to change reactor power,. During ascension to the

() 50% plateau, power was stabilized near the 35% and 40% levels to

, accommodate testing at ',hese plateaus. Results of individual tests completed during power ascension to and while at the 50% plateau are j discussed in Sections 3.2 through 3.5 of this report. At the 50% testing l plateau, a comparison was made of selected control room indications with j the process computer to demonstrate computer monitoring capability.

'() Several discrepancies of greater than 3% between computer points and control board indications were noted for RCS temperature parameters. )

Resolution was deferred until the 75% testing plateau, at which time the RCS temperature loops were to be recalibrated. Data was also taken at the 50% plateau and reviewed to demonstrate the operability of the inadequate core cooling monitoring system. Reactor cavity sump leakage

() data was obtained and a leakage rate of 0 gpm was deteomined. Primary i containment ventilation system data, auxiliary building ventilation i system data, and steam tunnel ventilation system data was obtained and analyzed to verify the adequacy of these systems to provide the necessary cooling and ventilation in their respective areas. Primary containment ventilation data obtained indicated that 2 reactor containment fan

() coolers in operation may not be able to maintain the containment building

average temperature below the 120*F limit. It was determined that 3 l reactor containment fan coolers would be operated as necessary to remove

{ the heat load. A further discussion of ventilation systems performance i j is presented in Section 3.2 of this report. l

$3 Upon completion of this testing phase, the plant was restored as dire:ted l by the Shift Engineer.

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l O TAsLE 2.5-1 i i 1 50% FLUX MAP RESULTS l ACTUAL MAXIMUM l l

Peak Linear Heat Rate (KW/ft) 5.404 12.6 Extrapolated to 100% 10.961 12.6 FDHN- 1.4537 1.7858 O Fxy - rodded N/A N/A Fxy - unrodded 1.5397 1.7072 l

Quadrant Power Tilt Ratios .9859 1.0010 1.02 1.0027 1.0104

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O 2.6 - TEST sew ucs AT 75% POWER The Test Sequence at 75% Power was utilized to define the activities which constituted the startup testing program during escalation from 50%

i to 75% power and at approximately 75%.of rated thermal power. This d cument ensured that the Test Sequence at 50% Power had been completed O and the results approved prior to increasing power above the 50% testing plateau. Prior to increasing power for this test sequence, power range high level trip setpoints were conservatively set to 94 ! 1% power and reactor core flux map results from a 50% power baseline map were verified acceptable. The flux map results were also extrapolated to 94% power to

() ensure parameters indicative of DNBR and linear heat rate were acceptable for power ascension to the 75% testing plateau. Refer to Table 2.6-1 for a tabulation of flux map results.

Plant operating procedures were utilized where appropriate to establish

! plant conditions and to change reactor pow'er. Results of individual

'() tests completed while at the 75% plateau are discussed in Sections 3.2, 3.4 and 3.5 of this report. At the 75% testing plateau, a comparison was made of selected control room indications with the process computer to demonstrate computer monitoring capability. No significant discrepancies were noted. Data was also taken at the 75% testing plateau and reviewed to demonstrate 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. A log of primary containment ventilation system data was also obtained.

Also while at the 75% testing plateau, an axial xenon oscillation was induced so that flux map data could be obtained to be used for

() calibration of the execre axial flux difference instrumentation.

Indicated excore delta flux (AI) was driven to approximately -16.5% by insertion of control bank D concurrent with a boron dilution. A full i core flux map was then obtained and after approximately 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> control bank D was borated out to its original position of greater than 180 steps. This induced an axial xenon oscillation and AI was allowed to j()

drift to approximately +3% at which time control bank D was inserted by boron dilution to drive AI back to its target value of approximately i

-4%. AI was than allowed to drift another 7% in the negative direction to approximately -11% at which time control bank D was withdrawn to l reposition AI back to the target value of approximately -4%. This i maneuver induced an axial xenon oscillation and subsequently dampened the

() oscillation thereby successfully demonstrating axial xenon oscillation 3

suppression control.

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

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TABLE 2.6-1  !

75% FLUX MAP RESULTS l l

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Peak Linear Heat Rate (KW/ft) 8.123 12.6 Extrapolated to 100% Power 10.947 12.6 FDHN 1.4534 1.6700 O N/A N/A Fxy - rodded l

Fxy - unrodded 1.5258 1.6300 Quadrant Power Tilt Ratios .9855 1.0051 1.02 )

O .9973 1.0122 l 1

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O 2.7 - TEST SEQUENCE AT 90% POWER The Test Sequence at 90% Power was utilized to define the activities which constituted the startup testing program during escalation from 75%

to 90% power and at approximately 90% of rated thermal power. This document ensured that the Test Sequence at 75% Power had been completed

() and the results approved prior to increasing power above the 75% testing plateau. Prior to increasir.g power for this test sequence, power range high level trip setpoints were set to 108 + 1% power and reactor core flux map results from a 75% power baseline map were verified acceptable.

The flux map results were also extrapolated to 109% power to ensure

() parameters indicative of DNBR and linear heat rate were acceptable for power ascension to the 90% testing plateau. Refer to Table 2.7-1 for a tabulation of flux map results.

Plant operating procedures were utilized where appropriate to establish plant conditions and t'o change reactor power. Results of individual

() tests completed while at the 90% plateau are discussed in Sections 3.3 and 3.5 of this report. While at the 90% testing plateau a log of primary containment ventilation system data was obtained. Also while at the 90% testing plateau, a baseline flux map was obtained for use in satisfying the initial conditions of the upcoming 100% sequence document.

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

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O TABLE 2.7-1 90% FLUX MAP RESULTS ACTUAL MAXIMUM O

Peak Linear Heat Rate (KW/ft) 9.581 12.6 Extrapolated to 100% 10.646 12.6 FDHN 1.4381 1.5965 O Fxy - rodded N/A N/A Fxy - unrodded 1.5029 1.5810 Quadrant Power Tilt Ratios .9870 1.0060 1.02 O .9944 1.0126 O i

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O '2.8 - TEST SEQUENCE AT 100% POWER The Test Sequence at 100% Power was utilized to define the activities which constituted the startup testing program during escalation from 90%-

to 100% power and 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 j plateau. Prior to increasing power for this test sequence, reactor core flux map results from a 90% power baseline map were verified acceptable.

The flux map results were also extrapolated to 109% power to ensure parameters indicative of DNBR and linear heat rate were acceptable for

.() power ascension to the 100% testing plateau.

Plant operating procedures were utilized where appropriate to establish plant conditions and to change reactor power. During ascension to the 100% plateau, power was stabilized near the 98% level to accomodate 1 testing at this plateau. Results of individual tests completed during

() this power ascension and while at the 100% plateau are discussed in i Sections 3.2, 3.4 and 3.5 of this report. At the 100% testing plateau, a comparison was made of selected control room indications with the process computer to demonstrate computer monitoring capability. Data was also taken at the 100% plateau and reviewed to demonstrate 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 and analyzed to verify the adequacy of these systems to provide i the necessary cooling and ventilation in their respective areas. Primary 4

containment ventilation data obtained during the 100% testing sequence

() was consistent with data obtained during the 50% testing saquence in that j 2 reactor containment fan coolers in operation may not be capable of

maintaining containment building average temperatures below the 120*F j limit. A third reactor containment fan cooler will be operated as i necessary to remove the heat load. A further discussion of ventilation

, systems performance is presented in Section 3.2 of this report.

() Containment penetration cooling temperature data was also obtained while at the 100% plateau and all readings were below the 120*F upper limit.

I A flux map was taken near the 100% power plateau and its results were examined to determine acceptability. The peak value for Fxy exceeded its expected value and several measured versus predicted reaction rate errors

() exceeded 10%. The quality of this flux map was determined to be suspect and another map was then taken at approximately the same power level.

The data obtained from the second map was found acceptable. Refer to

! Table 2.8-1 for a tabulation of flux map results.

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

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O TABLE 2.8-1 100% FLUX MAP RESULTS ACTUAL MAXIMUM g

Peak Linear Heat Rate (KW/ft) 10.671 12.6 FDHN 1.4500 1.5082 Fxy - rodded N/A N/A O

Fxy - unrodded 1.5152 1.5221 Quadrant Power Tilt Ratios .9848 1.0020 1.02

.9984 1.0149 O

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O TABLE 3.0-1 3.1 CORE LOADING i i

3.1.1 Initial Core Loading and Fuel Transfer Record, 2.32.32

() 3.1.2 Core Loading Instrumentation (High Voltage / Discriminator / Neutron Check) 2.52.31A-B 3.1.3 Operational Alignment of Excore Nuclear Instrumentation,

2.52.35A-B l 3.1.4 Reactor System Sampling for Core Load, 2.61.30 3.1.5 Radiation Surveys Prior to Core Load, 2.61.31 0

3.2 POST CORE LOADING SYSTEM TESTING l

! 3.2.1 Degassing the Reactor Coolant System, 2.18.30 Pipe Vibration, 2.28.30A-C 3.2.2 3.2.3 Steam Generator Level Controller Response, 2.34.30A-D

() 3.2.4 Thermal Expansion - Feedwater, 2.34.32A-B 3.2.5 Main Feedwater (Performance Verification of Waterhamner Prevention System - Upper Nozzle), 2.34.33A-B 3.2.6 Incore Flux Mapping System Checkout, 2.45.30 3.2.7 Digital Rod Position Indication System Checkout, 2.59.30 3.2.8 Chemistry and Radio Chemistry Criteria for Monitoring Water 13 Quality During Startup and Power Ascension, 2.61.32 3.2.9 Radiation Surveys During Power Ascension, 2.61.33 3.2.10 Process and Effluent Monitors and Failed Fuel Monitor Checks, 2.61.34 3.2.11 RTD Bypass Loop Flow Verification, 2.63.30

, 3.2.12 Reactor Coolant System Flow Measurement, 2.63.31A-B

() 3.2.13 RC Flow Coastdown, 2.63.32 3.2.14 Reactor Coolant System Leak Testing, 2.63.33 3.2.15 Control Rod Drive Mechanism Operational Test, 2.64.31 1 3.2.16 Rod Control System Checkout, 2.64.32 3.2.17 Automatic Reactor Control, 2.64.34 I

3.2.18 Rod Drop Time Measurement Test, 2.64.35

() 3.2.19 Reactor Protection Logic, 2.68.30 3.2.20 Pressuriser Sprays, Heaters and Bypass Flow Adjustments, 2.69.30 1

3.2.21 Auxiliary Building Ventilation, 2.84.11' 3.2.22 Heat Capacity Verification for Auxiliary Building Ventilation System, 2.84.30 3.2.23 Heat Capacity Verification for Control Room HVAC System, 2.85.30

,() 3.2.24 Heat Capacity Verification for Diesel Generator Ventilation, l 2.86.30 3.2.25 Heat Capacity Verification for Primary Containment Ventilation System, 2.93.30 3.2.26 Heat Capacity Verification for Safety Valve Enclosure Room 4 Ventilation System, 2.97.44 C) 3.2.27 Heat Capacity Verification for Switchgear Heat Removal Ventilation System, 2.99.30 4

3.2.28 Heat Capacity Verification for Miscellaneous Electrical Equipment Room Ventilation System, 2.128.30 j 3.2.29 Heat Capacity Verification for Remote Shutdown Control Room Ventilation, 2.129.44

) 3.2.30- Steam Generator Moisture Carryover Measurement, 2.51.45 (0851M/0080M)

O TABLE 3.0-1 (Continutd)

3.2.31 Dynamic Automatic Steam Dump Control, 2.51.46 3.2.32 Safety Parameter Display System, SPP 85-10,11 (License Cond.)

3.2.33 VC Positive Pressure Test, R-284 (License Cond. Attach 1, Item C)

() 3.3.34 Aux Bldg Vent. Charcoal Filter Plenums - Thermistors (License Cond. C.6.D) 3.2.35 Control Room Ventilation System Accident Req. (License Cond. C.13) i 3.3 PHYSICS TESTING

() 3.3.1 Incore Flux Mapping at Low Power, 2.45.31 3.3.2 Incore Moveable Detector and Incore Thermocouple Mapping at Power, 2.45.32B-C 3.3.3 Pseudo Rod Ejection, 2.45.33A-B 3.3.4 Isothermal Temperature Coefficient Measurement, 2.47.30 3.3.5 Power Coefficient Determination, 2.47.31

() 3.3.6 Initial Criticality, 2.52.32 3.3.7 Determination of Core Power Range for Low Power Physics Testing, 2.52.33 3.3.8 Reactivity Computer Checkout, 2.52.34 3.3.9 Bank North Measurement at Zero Power, 2.64.30A-B

! 3.3.10 Boron Endpoint Determination, 2.64.33 C) 3.3.11 Pseudo Rod Drop, 2.64.36 3.3.12 Miscellaneous Physics Test Results, 2.32.33 3.4 TRANSIENT TESTING 3.4.1 Loss of Offsite Power, 2.05.30 f) 3.4.2 Load Swing Test, 2.52.37 3.4.3 Full Power Plant Trip, 2.52.38

3.4.4 Plant Trip from 25% Power, 2.52.39 l 3.4.5 Shutdown From Outside the Control Room, 2.63.35 l 3.4.6 Large Load Reduction, 2.64.39

O 3.5 INSTRUMENTATION AND CALIBRATION TESTING 1

4 l 3.5.1 Process Computer Verification - Flux Mapping, 2.20.35 1 3.5.2 Calibration of Steam and Feedwater Flow, 2.34.31 3.5.3 Thermal Power Measurement and Statepoint Data Collection, 2.47.32A-E

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I 3.5.4 Incore Thermocouple /RTD Cross Calibration, 2.47.33

, 3.5.5 Operational Alignment of Excore Nuclear Instrumentation, I 2.52.35C-F 3.5.6 Axial Flux Difference Instrumentation Calibration, 2.52.36A-B 3.5.7 Reactor Loose Parts Monitor, 2.133.30B

') 3.5.8 3.5.9 Startup Adjustments of Reactor Control Systems, 2.64.44 Nuclear Steam Supply System Acceptance Test, 2.80.49 O

(0851M/0080M)

O 3.1 CORE LOADING 3.1.1 - INITIAL CORE LOADING 2.32.32 1

OBJECTIVES O

The Initial Core Loading Test Procedure was performed to ensure that the l nuclear fuel assemblies were loaded in a safe and cautious manner such  !

that an inadvertent criticality was avoided. This procedure was also  !

utilized to verify placement of the fuel assemblies into their proper  :

core locations upon completion of core load. l TEST METHODOLOGY The test procedure began by loading the temporary core loading instrumentation into their initial positions and determining background count rates for all source range and temporary nuclear instrumentation C) channels. The two primary source bearing assemblies and six additional assemblies, comprising the ' source nucleus', were loaded next. Audible indication of neutron population changes from one of the two installed plant channels, was maintained in both the control room and containment for the duration of the core loading process. After the source nucleus assemblies were loaded, the average of ten sets of counts measured over a

() 100 second time period was taken for all five nuclear channels used in the core loading process (two source range and three temporary channels). The first reference value, for use in inverse count rate ratio monitoring, was determined from these sets of counts after appropriate background values had been subtracted. Subsequent reference values were calculated whenever core loading was suspended for eight E) hours or longer, a temporary detector was noved, or a primary source bearing assembly was moved to a different core location.

Predictions were made prior to fuel load to verify that the reactor would l remain at least ten percent shutdown throughout the loading process.  !

Inverse count rate ratio monitoring was used following each fuel assembly

() move to ensure that the reactor was not approaching criticality. To )

ensure reliability in the monitoring, a minimum of two of the five nuclear instrumentation channels were required to be responding to source neutron population changes throughout core loading. Data obtained during inverse count rate ratio monitoring was compared to actual data supplied by Westinghouse Electric Corporation from the Salem Unit Two core

() loading. Salem Unit Two data, having a similar core loading pattern to that used at Byron, allowed core loading personnel to verify that the core was being loaded in a safe and cautious manner.

Upon completion of loading, the core was mapped using a video camera to verify proper placement of assemblies into the reactor vessel.

()

1 O

C3 4 1

(0851M/0080M) 1

O 3.1.1 - INITIAL CORE LOADING, 2.32.32 (Continuid)

SUMMARY

OF RESULTS Core loading was completed in a safe and cautious manner as required by

')

e the acceptance criteria of the core loading procedure. Problems encountered during the test were primarly associated with non-responding or malfunctioning neutron detectors.

No ise spiking problems on the two installed source range channels caused several delays during fuel load. The observed spikes were of varying magnitudes and durations, and most resulted in actuation of the

() containment evaulation alarm. One occurrence of the spiking caused actuation of the boron dilution protection system, resulting in charging pump suction switchover to the refueling water storage tank. The root

, cause for the spiking problem was traced primarily to electrical noise.

Cleaning, reterminating and reconnecting the triaxial connectors,

() movement of the source range preamplifiers, and rerouting of source range cables and conduits away from identified noise sources were all corrective actions performed to mitigate the effects of the spiking problems.

i

.O O

'O l

I

-O l

O O

C) (0851M/0080M)

.O FIGURE 3.1.1 - 1 BYRON UNIT l CYCLE I RANDOMlZED CORE LOADING PATTERN O iso. (E)

S R- N31 C C30 C46 C48 I I l

'O B C35 A27 C37 A 2

3

.,9 4 -

5 6

.O 7 l 8

9

,O 10 C44 ll

.O 12 I

13 L 14 .

'O 15 SR-N32 R P N M L K J H G F E D C B A O

.O* (W) i i

O Notes The ecpaipment hatch is located at 90' inside Unit i Containment.

• Denotes the location of a primary source.

O

O FIGURE 3.1.1 -2 BYRON' UNIT I CYCLE I RANDOMlZED CORE LOADING PATTERN O 180' (E)

S R- N31 C30 C46 C48 C03 I

,Q B C35 A27 C37 A30 A 2 .

A49 B30 A03 B36 3

B12 A42 B39 A35 A29 B01 A01 B45 5

, B43 A52 B64 A44 6

A13 B55 A56 B11

.,0 7 C B52 A55 B54 A64 8

1 9

O 10 C44 ll -

1 O 12 = -

l l

l 13 l CST l4 . l O

15 SR-N32 R P N M L K J H G F E D C B A O o (w)

O * **

The e@ipment hatch is located at 90' inside Unit i Containment.

• Denotes the location of a primary source.

4 O

l O FIGUR E 3.1.1-3 BYRON UNIT l CYCLE I RANDOMlZED CORE LOADING PATTERN

.O 180* (E)

S R- N31 C30 C46 C48 C03 Q B C35 A27 C37 A30 A A49 B30 A03 D46 l B12 A42 B39 A35 A29 B01 A01 B45 5

B43 A52 B64 A44 6

., A13 B55 A56 B11 C B52 AS$ B54 A64 A10 BIS A20 B10 9

.O Bis A57 B47 A34 10 A08 B09 A17 B08

ll J B04 A04 B02 A02 l A59 B22 A61 B21 I

C44 C57 A37 O 15 SR-N32 R P N M L K J H G F E D C B A O oo(w) i

'O h The epipment hatch is located at 90' inside Unit I Containment.

• Denotes the location of a primary source.

O

,0 FIGUR E 3.1.1 - 4 BYRON UNIT l CYCLE I i

RANDOMlZED CORE LOADING PATTERN O 180* (E)

SR- N 31 C30 C46 C48 C03 C59 I

Q B C35 A27 C37 A30 C26 A49 B30 A03 B46 A60 3

B12 A42 B39 A35 B41 4

A29 B01 A01 B45 B24 5

B43 A52 B64 A44 B42 6

A13 B55 A56 B11 All B28 A28 C49 C33 0 7 C B52 A55 B54 A64 B63 A06 B06 A63 A 8

A10 B19 A20 B10 A32 B20 A31 C62 C07 9

Q B15 A57 B47 A34 B36 10 A08 B09 A17 B08 B17 11 B04 A04 B02 A02 B23 O I2 l

A59 B22 A61 B21 A22  !

13 L l C44 A41 C57 A37 C18 14 .

O C58 Cc0 c22 C21 C40 g

SR-N32 R P N M L K J H G F E D C B A O oo(w) l 9 Notes The equipment hatch is located at 90* Inside Unit i Containment.

• Denotes the location of a primary source.

! ;O

O FI G U R E 3.1.1 -5 -

BYRON UNIT l CYCLE I 1

l RANDOMlZED CORE LOADING PATTERN O 180* (E)

S R- N31 C30 C46 C48 C03 C59 I

C35 A27 C37 A30 C26 C29 C01 Q B A49 B30 A03 B46 A60 B48 C14 C53 3

B12 A42 B39 A35 B41 A38 D35 C43 4

A23 B01 A01 B45 B24 B14 A46 C38 C50 5

B43 A52 B64 A44 B42 A65 B38 A43 C55 6

A13 BS$ A56 B11 All B28 A28 C49 C33

,0 7 B52 AS$ B54 AS4 B63 A06 B06 A63 A 8

A10 B19 A20 B10 A32 B20 A31 C62 C07 9

O B15 A57 B47 A34 B36 A16 B56 A25 C10 10 A08 B02 A17 B08 B17 B44 A45 C31 C15 ll B04 A04 B02 A02 B23 A23 B61 C36 O 12 -

[- - -

A59 E22 A61 B21 A22 B40 C16 C25 13 C44 A41 C57 A37 C18 C04 C02 14

• O C C58 C60 C22 C21 C40 SR-N32 1

M L R P N K J H G F E D C B A O o.(w)

O Notes The epipment hatch is located at 90' inside Unit I containment.

~

• Denotes the location of a primary source.

FIGURE 3.1.I-6 O BYRON UNIT l CYCLE I l RANDOMlZED CORE LOADING PATTERN l 180' (E)

O SR- N31 C5 C27 C30 C46 C48 C3 C58 I

C12 C39 C41 A5 C35 A27 C37 A30 C26 C29 Cl O 2 C45 C24 B5 A47 827 A49 B30 A3 B46 A60 B48 C14 C53 C51 B3 A12 B59 A40 B12 A42 B39 A35 B41 A38 B35 C43 4

0 ^ " " ^ "' ^' " ^ " '

5 C13 A24 B53 A14 B51 AS4 B43 A52 B64 A44 B42 A65 B38 A43 CS$

6 C11 C28 A39 B50 A51 B13 A13 B55 A56 Bil All B28 A28 C49 C33 7

O C52 A50 B49 A33 B33 A53 B52 A55 B54 A64 B63 A6 B6 A63 C54 l

C63 C6 A7 B62 A19 B7 A10 B19 A20 BIO A32 B20 A31 C62 C7 9

C19 A15 B32 A48 B31 A35 B15 A57 B47 A34 B36 A16 B56 A25 C10 l C20 C32 A9 B34 B16 B25 A8 B9 A17 D8 B17 B44 A45 C31 C15 ll C47 B57 A58 B18 Alf B4 A4 B2 42 B23 A23 861 C3G 12 --

.O C56 , C34 B58 A62 B50 A59 B22 A61 B21 A22 B40 C16 C35 i 13 C9 CS C17 A21 C44 A41 C57 A37 CIS C4 C2 C23 C61 C58 C60 C22 C21 C40 SR-N32 R P N M L K J H G F- E D C B A O O'(W)

A - REGION I ( 2.1 w/o )

O- B - REGION 2 ( 2.6 w/o )

C - REGION 3 (3.lw/o )

i O FINAL CORE CONFIGURATION O

FIGURE 3.1.1-7 Shutdown and Control Rod Locations

.O R P N M L K J  !! G F E D C B A 180' 1

'O 2 3 B C B S A g 3 5 5 3 3 0 8 8 C O 4 Sg D S D E Sg

~5 S C I D

O B C A C B 7

i S

B 3 8

8 9(P C S E A D A S E C 270*

10

9 Sg Sg

, 10 B C A C B

U

3 i 11 D 3 C

3 D S 12 A E D S A

.;O 13 3 3 3 3 C 8 8 0 1

14 3 B C B A Sg 15

.O D*

CONTROL NUMBER SMUTDOWN NUMBER

_ SANK OF 2005 ,

SANr er n00s A 4 O s 8A 3 a s, e C 8 8 C I D $ y 3,

TOTAL 25 3 E

TOTAL 28 I

O FIGURE 3.1.1-8 O M L K J H F R P N G E D C B A 180' 1 6 12A 12A O

23 2 5 20 24 is 20 5 3 5 24 24 20 24 24 5 jO 4 24 20 45 24 24 20 24 i

F 20 20 12 24 20 24 12 20 20 4

6 24 24 24 24 24 24

O 7 12A 24 24 24 24 24 24 24 12A

8. 90' 6 20 20 24 24 20 20 6 270' 9 9 12A 24 24 24 24 24 24 24 12,A 10 24 24 24 24 24 24 11 ;20 20 '12 24 20 24 12 20 20

,0 24 20 24 24 45 20 24 12 5 24 24 20 24 24 5 13 ,

5 20 24 20 5 14 f3 15 12A 6 12A O'

O Number Indicates Number of 1934 BP, Rods Burnable Poison Rods 12.5 w/o B23 0 5 Indicates Source Rod

'O BURftABLE POISON ROD LOADIi:G PATTER!1 4

C) l 3.1.2.a - CORE LOADING INSTRUMENTATION (HIGH VOLT 7GE/ DISCRIMINATOR / NEUTRON i CHECK), 2.52.31A OBJECTIVE

() The core loading instrumentation test procedure was performed prior to core loading to determine the proper operating and discriminator voltage settings for the temporary core loading instrumentation and to verify that the instruments provided indication when a neutron source was present.

O TEST METHODOLOGY Following the initial installation of the equipment, the temporary detectors were positioned near a neutron source. Using the neutron source, an optimum operating voltage was selected for each of the three detectors to ensure that minor fluctuations in detector power supply

() voltages would not adversely affect detector output. With the individual detector operating voltages selected, discriminator voltages were determined based on detector characteristic curves.

SUMMARY

OF RESULTS

~,(3 Upon completion of the procedure, operating voltages were determined to be 2100 volts for all three temporary nuclear instrumentation channels and discriminator voltages were set at 2.25 volts for detectors A and B and 2.75 volts for channel C.

Several problems were noted with the temporary nuclear instrumentation

() during core load. Since two of the original detectors were replaced, only one of the original detectors remained in place upon completion of.

core load. The two replacement detectors did not have individual voltage plateau data supplied with them, therefore acceptance criteria of the procedure was not technically met. (Acceptacce Criteria required comparison of on-site generated voltage plateau data to vendor supplied C) shop test results.) Offsite engineering review determined the adequacy of the onsite generated voltage plateau data for the two detectors in conjunction with vendor technical guidance and stated that the intent of the precedure's acceptance criteria had been met. I

'O O

() l (0855M/0080M)

3.1.2.b - CORE LOADING INSTRUMENTATION (NEUTRON CHECK DURING CORE LOADING), 2.52.31B OBJECTIVE

.'O The core loading instrumentation test procedure was performed to verify that the temporary instruments provided indication when a neutron source was present prior to resuming core loading following an eight hour delay in loading.

TEST METHODOLOGY j O

Following the suspension of core loading for eight hours, a submersible neutron source was lowered into the vessel and moved toward each of the three temporary detectors. As the source approached each detector the indica,ted count rate was observed.

'O

SUMMARY

OF RESULTS The test procedure was executed in its entirety five times during the t core loading activity. During each performance it was observed that as the source approached each detector the channels count rate increased, )

indicating that the detector was responding.

,0  !

t i

t I

O  !

O 1

) I

O l

O i

O a

(0855M/0080M)

3.1.3 - OPERATIONAL ALIGNMENT OF EXCORE NUCLEAR INSTRUMENTATION (PRIOR TO CORE LOAD, DURING CORE LOAD), 2.52.35A-B l i

OBJECTIVE i

This test was performed to verify that the Source Range (SR) and

() Intermediate Range (IR) excore instrumentation channels were functioning as designed and capable of detecting and alarming an excessive insertion j of positive reactivity. Test 2.52.35B was specifically used to re-verify 4 the response of SR channels to a neutron source when core load procedures experienced a delay of eight hours or more.

l O

TEST METHODOLOGY l

I Operational verification of the high level trips, high flux indications l and containment evacuation horns was assured by observing annunciator i window's on the main control board and SR/IR channel drawer. Each SR 1

() channel responded by indicating a positive change in the count rate when  ;

i a neutron source was introduced near the detector associated with the i channel. SR channel circuit components were adjusted.

SUMMARY

OF RESULTS j() The level trip bistables for SR channels N31 and N32 tripped at 105 cps t and 4.9 X 104 cps, respectively. The high flux at shutdown bistables

for these channels tripped at 9.6 cps and 10.0 cps, respectively. Proper

actuation of main control board annunciators, level trip and high flux at i shutdown drawer windows, and containment evacuation horn were observed.

.() The level trip bistables for IR channels N35 and N36 both tripped at 10-4 amps and the power above P-6 bistable for both channels tripped at 10-10 amps. Proper actuation of power above permissive P-6 and high level trip drawer windows was observed. These numeric values met the acceptance criteria.

i C) Testing conducted prior to core load utilized a 7.5 curie Pu-Be neutron source. With ventilation covers in place, only a marginal count rate of

)

less than 2.0 cps was observed when the neutron source was placed inside the reactor vessel near the detector. Consequently, prior to the setting of SR high voltages it was necessary to remove the ventilation covers to

! allow insertion of the neutron source into the detector wells. With this C) configuration, count rates of 4000 cps for channel N31 and 10,000 cps for channel N32 were observed and were within the acceptance criteria. To eliminate continual removal of the ventilation covers during core I

loading, subsequent detector response verification utilized a different i neutron source methodology. Specifically, the vertical movement of a fuel assembly lying between the primary Californium source and the SR

() detector' acted as a neutron source by subcritical multiplication of the Californium source neutrons. With this methodology, statistically valid j positive count rate changes of over 2.3 cps to 3.1 cps were observed on the three occasions where the core loading procedure was delayed for more than eight hours.

O i

) (0855M/0080M)

0 3.1.4 - REACTOR SYSTEMS CHEMICAL SAMPLING FOR CORE LOAD, 2.61.30 OBJECTIVE This test was performed to verify correct and uniform boron

() concentrations in unisolated portions of the reactor coolant system (RCS) and the directly connected portions of the auxiliary systems as required for core;1oading. This test was also designed to ensure that the possibility of an inadvertent dilution of the RCS during core loading was minimized.

TEST METHOLODOGY O

Prior to the commencement of core loading, the RCS was sampled and verified to meet the specified water chemistry criteria stated in the test. ,This procedure was then repeated daily throughout the core loading process.

O Each of the RCS cold legs, the safety injection system accumulators, the safety injection system, the containment spray system, and the spent fuel pool were sampled, if they contained water, and that water was verified to contain at least 2000 ppm boron. An exception to this was the U-1 boric acid storage tank which had a limit of > 7000 ppm boron.

O Following the initial verification of the chemistry in the reactor coolant system, four samples were taken at equidistant depths from the reactor vessel along with a sample from the operating residual heat removal train. These samples were then analyzed for boron to verify a uniform boron concentration throughout the entire system. All samples

() were verified to be within 30 ppm of each other. Within four hours after the RCS boron was verified to be at a uniform concentration, the refueling canal inside and outside of containment, the operating residual heat removal (RHR) train, and the CVCS domineralizers influent and effluent were sampled and analyzed for boron to verify that all water within these systems were > 2000 ppm boron. Sampling continued hourly

() until the boron concentration in the CVCS effluent was within 30 ppm of the operating RHR train. At this time sampling continued on the refueling canal and the RHR train every four hours throughout the core loading process.

SUMMARY

OF RESULTS O

During the execution of this test which lasted throughout the core loading process, all acceptance criteria were adequately met for each system that was sampled. No corrective actions in the core loading process were needed to meet the acceptance of this test. The ID accumulator remained drained throughout the the core loading process

() along with the U-2 boric acid storage tank and the spent fuel pool. All other systems that were sampled were shown to contain a boron concentration of at least 2000 ppm. The U-1 boric acid storage tank had a concentration of 7276 ppm boron. These results assured that there was no inadvertant dilution of the RCS boron level during core load.

O C) (0855M/0080M) l

l 0

3.1.4 - REACTOR SYSTEMS CHEMICAL SAMPLING FOR CORE LOAD, 2.61.30 (Continu-d)

At no time during the execution of this test did the water in the RCS exceed the limits of 0.150 ppm chloride and fluoride. The lowest boron concentration found within that system during core load was 2061 ppm on

() 11/4/84. After 11/8/85 the boron concentration within the RCS remained above 2100 ppm with a maximum value reaching 2238 ppm on 11/15/84.

The refueling canal inside and outside of containment had a boron concentration of between 2070 ppm and 2209 ppm throughout the execution of this test. (The refueling canal boron concentration was intentionally

() increased so as to ensure that the concentration would not fall below the 2000 ppm limit.)

The rise in the reactor coolant system boron concentration as the test progressed was due to an operator initiation of emergency boration on 11/9/86 as a result of a source range noise spike. A total of

() approximately 1200 gallons of 7300 ppm boric acid from the boric acid storage tank was injected. Additional boration of the reactor coolant system periodically occurred throughout the remainder of core load.

O O

O O

O O

(0855M/0080M)

O 3.1.5 - RADIATION SURVEYS PRIOR TO CORE LOADING, 2.61.31 OBJECTIVE i The radiation surveys prior to core loading test was performed to O establish baseline data to be compared with radiation measurements during power ascension and commercial operation.

TEST METHODOLOGY Baseline radiation values were established by using precalibrated O portable survey instrumentation to measure dose rates at a series of specified points throughout Byron Station. Gamma readings were taken throughout the Service, Radwaste, Turbine, Auxiliary, and Fuel Handling Buildings and the Unit 1 Containment.

Upon co'mpletion, all survey instrumentation was post-calibrated to ensure O measurement validity.

SUMMARY

OF RESULTS 1

All of the specified points were surveyed for gamma radiation. All I readings were 0.02 mr/hr or less. This value is more than one order of O magnitude less than the acceptance criteria limit from FSAR Table 12.3-1. None of the radiation base point measurements required further investigation.

O O

1 O

O O

(0855M/0080M)

3.2 - POST CORE LOADING SYSTEM TESTING 3.2.1 - DEGASSING THE REACTOR COOLANT SYSTEM, 2.18.30 OBJECTIVE O

The purpose of this procedure was to verify the removal of excess hydrogen and other non-condensible gases from the reactor coolant system (RCS).

TEST METHODOLOGY O

With the plant in Hot Standby Condition after a reactor trip on OTAT due to two circulating water pumps tripping, the RCS was degassed via the waste gas system from the volume control tank. A nitrogen blanket was placed in the volume control tank to aid in the removal of the excess hydrogen and non-condensible gases from the RCS. The process of '

O degassing actually involved a transfer of pressurizer level to the volume control tank and then venting off the gases to the waste gas system. The Radiation-Chemistry Department used a gas chromatograph to verify that each RCS obtained had been degassed further. This was accomplished using normal operating procedure 1 BOP CV-18.

O

SUMMARY

OF RESULTS The acceptance criteria for this test was to reduce the dissolved gas concentration of the RCS by at least 10 cc/kg H 2O from its initial valve. Through several degassing efforts, the concentration i was reduced from 29.18 cc/kg H2 O to 16.9 cc/kg H 2O O - a difference of 12.28 cc/kg H20. The test was successful.

O O

O O

O (o855M/co8cM)

O 3.2.23 - PIPE VIBRATION (PRE CRITICAL), 2.28.30A OBJECTIVE The Pipe Vibration test procedure demonstrated that the peak stresses

() resulting from steady state flow induced vibration were within allowable design limits. The scope of the test was limited to portions of the chemical & volume control and main steam systems.

TEST METHODOLOGY

() The systems involved were operated under normal, steady state design conditions during which a visual inspection of the piping was conducted.

The walkdown divided the system into smaller piping subsystems between restraints in order to utilize simple beam analogy to determine deflection limits. Portable vibration analyzers were used to obtain vibratio'n levels and a comparison was made between the calculated C) deflection limit and the actual reading. Based on the outcome, vibration levels less than the allowable limit met acceptance criteria and those-that exceeded the acceptance criteria prompted further analysis by offsite engineering.

SUMMARY

OF RESULTS O

The test generated three sets of calculations where the levels of vibration were questionable with respect to the piping subsystems under inspection. The final results of the analysis indicated that the

. subsystems met acceptance criteria in that the levels of vibration did not exceed the allowable design limits.

O O

O O

(0855M/0080M)

3.2.2b - PIPE VIBRATION (20% REACTOR POWER), 2.28.30B OBJECTIVE The Pipe Vibration test procedure demonstrated that the peak stresses

() resulting from steady state flow induced vibration were within allowable design limits. The scope of this test was limited to the main feedwater system operating during low power lineup.

TEST METHODOLOGY

() While operating the feedwater system in accordance with general operating procedure BGP 100-3, the affected piping feeding the steam generators was

visually inspected for excessive vibration. The walkdown divided the system into smaller piping subsystems between restraints utilizing simple beam analogy to determine deflection limits. Portable vibration analyzers were used to obtain vibration levels and a comparison was made

() between the calculated deflection and the actual reading. Based on the outcome, vibration levels less than the allowable limit met the I

acceptance criteria and those that exceeded the acceptance criteria prompted further analysis by offsite engineering.

3

SUMMARY

OF RESULTS 0

4 The test generated two sets of calculations where the levels of vibration

experienced were questionable with respect to the piping subsystems under inspection. The final results of the analysis indicated that the subsystems met the acceptance criteria in that the levels of vibration did not exceed the allowable design limit.

O .

O

'o O

O i

3 t0855M/0080M)

l

.O - 3.2.2c - PIPE VIBRATION (30, 50, 75 and 90-100% REACTOR POWER), 2.28.30C OBJECTIVE l

The Pipe Vibration test procedure demonstrated that the peak stresses i resulting from steady state flow induced vibration were within allowable j f) design limits. The test was performed in an effort to qualify or accept- ,

the piping; associated with the main steam and feedwater systems.  ;

1 TEST METHODOLOGY l

The Pipe Vibration test utilized normal operating conditions on the main I

() steam and feedwater systems at various power levels in order to inspect and monitor the effects of flow induced vibration. The inspection consisted of walking down the affected piping and through the use of portable vibration analyzers readings were taken where levels rare deemed excessive. A simple beam analogy 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 and those that exceeded the acceptance criteria ,

prompted further analysis by offsite engineering. Vibration monitoring was also performed using temporarily mounted accelerometers on piping

'() located in areas inaccessible and the readings stored in a data collection system. The data was then forwarded to offsite engineering for their initial analysis and approval.

SUMMARY

OF RESULTS

() The tests performed at the various power levels generated a number of calculations where the levels of vibration were questionable with respect to the piping subsystems under the inspection process. The vibration readings obtained through system walkdowns were less than their respective allowable limits and therefore the acceptance criteria was met. Results from the remote vibration monitoring process met all 1

() acceptance criteria limits and was confirmed by offsite engineering.

i, O

I 1

i O

l i

O C) (0855M/0080M)

l 3.2.3 - STEAM GENERATOR LEVEL CONTROLLER RESPONSE, 2.34.30A-D l

OBJECTIVE Steam generator level control was demonstrated throughout power

'g ascension. Changing feedwater flow configurations and major power j changes necessitated the need for multiple performances of these tests. j

! Level control stability of the four steam generators was demonstrated j while operating on the feedwater bypass regulating valves and the main l feedwater regulating valves. Level control stability was also demonstrated while transferring feedwater flow between the feedwater O bypass regulating valves and the main feedwater regulating valves and  !

while swapping main feedwater pumps.  ;

I TEST METHODOLOGY In order'to verify level control stability while operating on the bypass

'O or main feedwater regulating valves, a 5% level deviation was manually

established in each steam generator. The control system was then l transferred to the automatic control position. The actual steam generator level was monitored to determine if it returned to the programmed level of 66% within a specific time frame.

O In order to verify level control stability while transferring between the feedwater bypass regulating valves and the main feedwater regulating

, valves, steam generator level was monitored while transferring from one to the other and then back again. This was done to ensure stability when transferring to the main feedwater regulating valves in the case of power I

ascension or to the bypass regulating valves in the case of power O descension.

A swapping maneuver among feedwater pumps was performed to verify the i pressure stability of the level control system. Every combination of our three feedwater pumps was tested to ensure stability with any pump supplying feedwater.

SUMMARY

OF RESULTS When given a 5% level deviation (high or low), the bypass regulating valves returned steam generator level to the programmed level within 37.5 minutes as expected. This was done at 7% power.

O When given a 5% level deviation (high or low), the main feedwater regulating valves returned the steam generator level to the programmed level within 93 minutes as expected. This was done at 30, 50, 75, and 100% power. While performing the 30 and 100% tests, the acceptance criterion reguarding expected main feedwater regulating valve position O was not met.

During the 30% power performance, all four valves were found to be open more than expected. Corrective action involved recognizing the correct valve opening characteristics. Byron's valves are equal percentage, not linear. The acceptance criteria was adjusted accordingly.

I O (J855M/0080M)

O 3.2.3 - STEAM GENERATOR LEVEL CONTROLLER RESPONSE, 2.34.30A-D (Continu*d)

Also in conjunction with the acceptance criteria change, a steam header vs feedwater header differential pressure setpoint program adjustment was made. This adjustment forced the valves to close slightly.

O' During the 100% power performance, the A regulating valve opened more than expected. Corrective action was taken by a complete grooming of all four main feedwater regulating valves. This activity involved stroke setting and positioner calibration. The corrective actions rectified the problems identified in the failed acceptance criteria.

O After transferring from the feedwater bypass regulating valves to the main regulating valves, the steam generator level returned to and remained within 2.0% of the programmed level within 93 minutes. After transferring from the main feedwater regulating valves to the feedwater bypass re@21ating valves, the steam generator level returned to and

() remained within 2.0% of the programmed level within 37.5 minutes, as expected. This was done at 20% power.

When transferring among the feedwater pumps, the feedwater header pressure oscillations were less than 3% of operating pressure. These maneuvers were performed at 48% power. Difficulties were found in

() controlling the pumps, but this was rectified through the dynamic tuning of the respective control systems.

() '

O

() ,

1 1

0 O

(0855M/0080M)

TABLE 3.2.3-1 BYPASS REGULATING VALVE LEVEL CONTROL RESPONSE PERFORMED AT 7% POWER O

ACCEPTANCE ACTUAL TIME STEAM LEVEL CRITERIA IN RESPONSE GENERATOR DEVIATION MINUTES IN MINUTES A 5% t i 37.5 16.7 l 5% 4 1 37.5 33.3 B 5*4 t i 37.5 16.7 5% 4 1 37.5 20.0 C 5% t i 37.5 20.0 0 5% 4 1 37.5 25.0 D 5% t i 37.5 26.7 5% 4 1 37.5 26.7 TRANSFER RESPONSE TRANSFER RESPONSE O FROM BYPASS TO MAIN FROM MAIN TO BYPASS l FEEDWATER REGULATING VALVE FEEDWATER REGULATING VALVE AT 20% POWER AT 20% POWER ACCEPTANCE ACTUAL TIME ACCEPTANCE ACTUAL TIME

O STEAM CRITERIA IN RESPONSE IN STEAM CRITERIA IN RESPONSE IN GENERATOR MINUTES MINUTES GENERATOR MINUTES MINUTES A i 93 37.3 A i 37.5 35.8

B i 93 29.7 B i 37.5 30.8 C < 93 34.3 C < 37.5 35.8 O D [93 45.0 D [37.5 31.7

,0 1

'O O (0855M/0080M) 1

O 3.2.4 - THERMAL EXPANSION-FEEDWATER, 2.34.32A-B OBJECTIVE Thermal expansion testing of the feedwater system was conducted to verify

() that components and piping could expand without restriction of movement upon system heatup. It was also conducted to confirm the correct functioning.of component supports, piping supports and restraints.

TEST METHODOLOGY

() At feedwater system ambient and hot conditions, system walkdowns were performed. Piping and components were visually examined and specific snubber positions recorded. Interferences were identified and dispositioned by the design engineers. When necessary, system walkdowns were again conducted following the resolution of interferences. All piping mov'ements were evaluated by the design engineers.

O

SUMMARY

OF RESULTS The piping and components were not to be constrained from expanding and actual thermal expansion movements could not vary from predicted thermal movements by more than ! 25% or 1/4 inch, whichever was greater.

() During the course of system walkdowns, several minor interferences were determined. These interferences were evaluated by the design engineers and determined to be acceptable as is, or specific corrective action was recommended. All recommended corrective actions were performed. Some portions of the feedwater system were again examined and measured following the removal of interferences. Movement of components not

() within the 25% or i 1/4 inch criterion were evaluated by the design engineers on a case-by-case basis. All thermal expansion movements were found to be acceptable.

O O

O O

(0$55M/0080M)

1 3.2.5a - WATER HAMMER PREVENTION, 2.34.33A OBJECTIVE l The upper nozzle of the steam generator was tested for damaging water O hamer following:

1) The initiation of auxiliary feedwater through an uncovered upper nozzle, and

2) Re-initiation of the purge flow through an uncovered upper nozzle.

.O TEST METHODOLOGY With the reactor coolant system at no load temperature and pressure, the steam generator level was decreased to 55% to uncover the upper nozzle.

After 30 minutes, a purge flow through the upper nozzle was established.

O An engineer qualified to perform vibration test measurements was i positioned to observe the upper nozzle water hamer. After that transient was observed, again steam generator level was lowered to 55%.

Auxiliary feedwater was initiated with the engineer surveying for any water hammer.  ;

O

SUMMARY

OF RESULTS l l

No damaging water hammer was observed during this test by the qualified engineer. This test demonstrated water hamer prevention as outlined in ASLB testimony.

10 -

1 O

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1 0 1 O

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O t

(OE55M/0080M)

O 3.2.5b - WATER HAMMER PREVENTION, 2.34.33B OBJECTIVE The purpose of this test was to observe and record the transient l

following the admission of feedwater into the preheater region of the '

E) steam generator. This test was performed at 28% reactor power. l TEST METHODOLOGY Twenty four accelerometers were attached at various points along the main

.q) feedwater lines to record pipe vibration. The admission of feedwater into the preheater region of the steam generator was accomplished by opening the feedwater isolation valves in accordance with Byron General Operating Procedure BGP 100-3, " Power Ascension 5% to 100%". Offsite engineering performed an evaluation of damaging water hammer using the accelerometer' data taken during the performance of the test.

.O

SUMMARY

OF RESULTS Two difficulties were encountered in the performance of this test. The first was the inability to obtain required feedwater isolation valve bypass purge flow. A design error restricting the flow was determined to

() be the cause. With the concurrence of Westinghouse, this test continued with less purge flow (90 gpm vs 130 gpm).

The second problem encountered was the underranging of the peak-to-peak calibration of the accelerometers. This caused the data from the initial performance of this test to be inconclusive. After expanding the range,

() the test was reperformed.

The data from the second performance was successfully recorded and j analyzed. It showed no damaging water hammer in any portion of the four l main feedwater lines. This test demonstrated water hammer prevention as outlined in NUREG 1606.

C) l l i O

O C) .

53 (0855M/0080M)

3.2.6 - INCORE FLUX MAPPING SYSTEM SYSTEM CHECKOUT, 2.45.30 OBJECTIVE' The purpose of this procedure was to demonstrate the proper operation of the flux mapping system which includes the leak detection system and the

{) CO2 gas purge system. In addition, top and bottom of core limits were set and the actual drive cables and detectors were installed.

TEST METHODOLOGY

'() Using dummy drive cables, the top and bottom of core limits were established for normal, emergency, calibrate and storage modes by slowly driving the dummy detectors to the top of the core (or storage position)

where clutch slippage was observed. The position was then recorded from 1 the encoder display. The top limit was obtained by subtracting two

inches from the recorded position and the bottom limit was obtained by I subtracting 170 inches from the top limit. The leak detection system was l

!() tested by filling the drain header with domineralized water and allowing

the leak detection level switch to actuate, thereby draining the water. l The withdraw limit switch was bypassed to verify the safety limit switch would prevent the detector from being taken up onto the reel. All push-to-test lights were verified. The CO2 gas purge system was l l() checked to verify a positive pressure existed in the system when the I detector cables were withdrawn. I i

l l

SUMMARY

OF RESULTS 4

i Proper operation of all indicating lights were verified along with the 1

'() proper operation of the leak detection system as described in the previous section. The CO2 gas purge system was verified to produce a positive pressure in the system when the detector cables were withdrawn. l The dummy detectors were inserted into all core locations with the exception of core location N-8. With this path blocked, the system still satisfies the minimum number of thimbles for core mapping as required by

() Technical Specifications. All top and bottom limits were established.

The safety limit switches were demonstrated operable as previously j discussed. As a final step, the actual detector cables were installed on j the drive units.

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1

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1

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! -52 O (0655M/0080M) .

l l

0 3.2.7 - DIGITAL ROD POSITION INDICATION SYSTEM CHECKOUT, 2.59.30 OBJECTIVES The digital rod position indication system checkout test procedure was O performed to verify the following indication and alarm functions for each individual RCCA: system' accuracy (full and half), position associated rod deviatiort alarms, rod at bottom indication and data transmission. In the i process of demonstrating the indication and alarms above, it was shown that all RCCAs operated over their full range of travel.

O TEST METHODOLOGY Two methods of testing were used to fully prove the rod position i indication system in this test procedure. The first type of. testing 4

involved injecting simulated rod position data from the test switches in the dat'a cabinets and observing the resulting LED indication locally and O in the control room. The second type of testing involved using actual rod position data by moving individual RCCAs and observing indications and alarms that resulted. Individual rods were disconnected from their group'using the manual lift disconnect switches in the main control room. All RCCAs were tested in this manner.

O

SUMMARY

OF RESULTS j

Accuracy requirements for the rod position indication system from the

. vendor technical manual were i 4 steps in comparison with associated step

! counters. All control banks met this criterion over the entire range of g rod travel from rod bottom (0 steps) to 228 steps. Shutdown banks met this criterion in the ranges from rod bottom to 18 steps and from 210 to 228 steps. The transition region of each shutdown bank was shown to indicate correctly between 21 1 4 steps and 207 i 4 steps as indicated on the step counters. Rod bottom indication was shown to occur for all shutdown and control rods at 3 i 1 steps on the step counters. The Rod

'g Versus Rod Deviation, Shutdown Rod Off Top, Rod Deviation and Rod At Bottom annunciator bypass were demonstrated to operate correctly as well. Additional testing of the Shutdown Rod Off Top alarm and the l Computer Alarm Rod Deviation Sequence NIS Power Range Tilts alarm will be done during the next outage and is being tracked by station action items.

O 1

1 1

1 0

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'O f

4 O (0855M/0080M)

_ ~ . _ __ __ __ . _ _ _

i 3.2.8 - CHEMISTRY AND RADIOCHEMISTRY CRITERIA FOR MONITORING WATER QUALITY DURING STARTUP AND POWER ASCENSION, 2.61.32 OBJECTIVE g 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 feedwater system met the i chemistry requirements specified in the Technical Specifications and/or

! the Westinghouse NSSS guidelines. The test was performed at Pre Heat-up, Heat-up Prior to Criticality, at Criticality, 30%, 50%, 75%, and 100%

0 Power.

TEST METHODOLOGY The testing was performed by obtaining five repetitive samples of the i priniary water make-up, reactor coolant, steam generator blowdown, and O steam generator feedwater systems from the appropriate sample panels throughout the plant. Chemical analyses were then performed on every sample from each system. The.results of these analyses were tabulated, averaged and the standard deviation calculated.

Although steam generator blowdown and steam generator feedwater samples

'O were obtained at each performance of the test, the samples and subsequent ,

l analyses were only required when the reactor was at-greater than 30%

power.

l

SUMMARY

OF RESULTS t

'O During the seven executions of this test, all acceptance criteria were

' adequately met for each system that was sampled. No corrective actions

in plant operation were needed to meet the acceptance criteria. However, I the criteria for silica was changed to (1000 ppb by Westinghouse and approved by offsite engineering. On occasions, one of the five repetitive samples had to be reanalyzed because a' result was not

.O consistent with those of the other four. Upon reanalysis the sample was i shown to be within specifications.

l I j During the performance of this test at the 75% power plateau, a problem l

! was identified in meeting the acceptance criteria for dissolved hydrogen j in the reactor coolant system. The problem was traced to a faulty vent O valve on the volume control tank. Following the repair of this vent l valve, the reactor coolant system was reanalyzed for hydrogen and found )

to meet the stated acceptance criteria within the test. 1 Tables 3.2.8-1 through 3.2.8-4 contain a summary of the results for each system sampled along with the acceptance criteria or guidelines stated

O within the test. The values given in these tables are the averaged j results of the five samples obtained from each system at each execution

of the test.

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! l (0855M/0080M) j 1

1

E) TABLE 3.2.8-1

SUMMARY

OF REACTOR COOLANT CHEMISTRY DURING EACH EXECUTION OF STARTUP TEST 2.61.32 g CHEMISTRY CRITERIA / MODE MODE MODE PARAMETER GUIDELINE 5 3 2 30% 50% 75% 100%

Chloride- <150 ppb <10 10.4 <10 17.2 <10 <10 <10 Fluoride <l50 ppb <50 <50 <50 <50 <50 <50 (50 J

()

Dissolved

• Oxygen <100 ppb 5653 12 20 5 5 5 6 O

Silica <1000 ppb 34.4 238.4 64 100.4 73.4 55.8 209 Lithium 0.7-2.2 ppm .538 1.46 1.76 1.8 1.52 1.37 1.4 O

l Hydrogen ** 25-50 cc/kg <1.0 17.7 21.5 37.7 27.6 36.2 33.51 Suspended Solids <l.0 ppm (0.5 0.538 <0.5 <0.5 <0.5 <0.5 <0.5 O

Aluminum (50 ppb <10 <10 <10 (10 <10 <10 <10 Calcium (50 ppb 23 11.4 8.8 1.3 19.2 <20 (20  ;

,C)

Magnesium (50 ppb 11 2 <2 2.2 <2 2.2 <2 l I

• When Tave > 180*F 0

• • When RCS > 1% Reactor Power

'O

'Q C) (0855M/0080M)

TABLE 3.2.8-2

SUMMARY

OF PRIMARY WATER MAKEUP CHEMISTRY DURING EACH EXECUTION OF STARTUP TEST 2.61.32 O CHEMISTRY CRITERIA / MODE MODE MODE PARAMETER GUIDELINE 5 3 2 30% 50% 75% 100%

Totai Chloride and Fluoride <100 ppb 8.37 6.69 <5.5 9.01 60.5 10.1 9.30 O

Silica <100 ppb (20 (20 <20 <20 (20 (20 34.4 pH'8 25'C 6.0-8.0 7.04 6.99 6.94 7.12 6.86 7.25 6.54

.O Specific Con-luctivity Micromho/cm 3

6 25"C <l.0 0.443 0.449 0.369 0.727 0.55 0.908 0.603 O

Total Solids (1.0 ppm <0.5 <0.5 <0.5 <0.5 0.5 <0.5 <0.5 Suspended O Solids <0.1 ppm 0.22* <0.05 <0.05 <0.05 <0.05 <0.05 <0.05 Aluminum (20 ppb <10 <10 <10 <10 (10 (10 <10 0 Calcium (20 ppb 6.0 4.4 2.2 <0.1 0.82 7.8 <0.1 Magnesium (20 ppb 4.4 3.0 1.4 3.0 <0.05 2.0 0.84 O potassium (10 ppb 2.6 2.4 0.83 1.0 2.2 1.6 0.84

• Five additional samples were obtained and analyzed and were all found to be ,

within specification. Based on the additional samples, the results were i accepted.

l

~o 0 (0855M/0080M) l

TABLE 3.2.8-3

SUMMARY

OF STEAM GENERATOR WATER CHEMISTRY DURING EACH EXECUTION OF

! STARTUp TEST 2.61.32 g CHEMISTRY CRITERIA / MODE MODE MODE PARAMETER GUIDELINE

• 5 3 2 30% 50% 75% 100%

Cation Conductivity Micromho/cm O 6 25'C <2.0 3.51 2.53 0.812 1.79 1.65 1.56 1.21 pH @ 25'C >9.0 9.85 9.08 9.22 8.63 9.12 9.16 9.20 O

• Guidelines not applicable until > 30% power P

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-~ ,,, ~~-- e- - - . - - - - - - , - , , , - - . ,, -

w - - - - - - -. --,n , - - , - , - - -

TABLE 3.2.8-4

SUMMARY

OF FEEDWATER CHEMISTRY DURING EACH EXECUTION OF STARTUP TEST

. 2.61.32 O CHEMISTRY CRITERIA / MODE MODE MODE i

PARAMETER GUIDELINE

• 5 3 2 30% 50% 75% 100%

Condensate

! Dissolved Oxygen ppb N/A N/A N/A 51.2 2.02 15.3 2.52 O

Hydrazine [1] + 5 ppb <5.0 <100 55.6 395 484 740 460 Dissolved O Oxygen ** <100 ppb 12710 1218 770 2.1 2.3 0.93 1.13 (1) Condensate Dissolved Oxygen

• Guidelines not applicable until > 30% Power

• • At >30% power, guideline is <5.0 ppb i

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(0855M/0080M)

o 3.2.9 - RADIATION SURVEYS DURING POWER ASCENSION (3-10%, 48-52%, 90-100%

i i Power), 2.61.33 f OBJECTIVE

.O 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 ganea and neutron radiation.

TEST METHODOLOGY O

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

radiation dose rate values were established in the Unit 1 Containment and l certain penetration areas. ~

O

SUMMARY

OF RESULTS The effectiveness of neutron radiation shielding in Containment was found

to be inadequate during all performances of this test. A modification to lower the reactor vessel nozzle shield cover plates with be implemented

..O at the next appropriate outage. The nozzle cover plates had been raised l approximately 14 inches and this gap had caused the ineffectiveness of neutron shielding. Radiation surveys for selected base points will be l

i repeated after neutron shielding modifications are completed.

I i The effectiveness of gamma shielding and the general determination of O dose levels were found adequate during all performances of the test.

Acceptable gamma dose rates ranged from 0.2 mR/hr in the Service Building

! to 1.0 mR/he in the Turbine Building and certain Auxiliary Building

locations to 100 mR/hr for a select number of points on equipment inside j cubicles in the Auxiliary Building. All measured ganea dose rates were at least 5% less than their respective acceptable level at 100% power.

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,o (0855M/0080M) l

O 3.2.10 - PROCESS AND EFFLUENT MONITORS AND FAILED FUEL MONITOR CHECKS (5%,

30%, and 100% Power), 2.61.34 OBJECTIVE f[) 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 a known source of radiation.

TEST METHODOLOGY C) Radioactive sources of known activities (as verified on counting room instrumentation) were exposed to the process and effluent monitor and failed fuel monitor detectors. The observed detector responses (as obtained on portable digital display instrumentation) were compared to the calculated expected detector responses to the radioactive sources to determine percent differences.

SUMMARY

OF RESULTS .,

All tested process and effluent monitor and failed fuel monitoc detector

! responses were within i 20% of the expected detector responses for all performances of this test, thus meeting the test acceptance criteria of

[3 1 20% agreement. Two detectors were unable to be tested at higher power levels due to detector failures and parts unavailability to make the needed repairs. Station action items were written to follow the repairs and subsequent testing of these detectors.

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() (0855M/0080M)

O 3.2.11 - RTD BYPASS LOOP FLOW VERIFICATION, 2.63.30 OBJECTIVE The RTD Bypass Loop Flow Verification test procedure was performed to (3 verify the actual hot leg RID bypass loop flowrates were greater than the flowrate required to meet a 1.0 second transit time, which is used in the design assumptions.

TEST METHODOLOGY

,C) Required minimum flow for a one second transit time was first calculated from actual pipe dimensions. Next, actual flowrates were determined for all four hot & cold leg RTD's by isolating the hot leg and recording the cold leg value and then repeating the steps for determining the hot leg

' flow. These measured values were then compared to the minimum valves to verify acceptability.

SUMMARY

OF RESULTS a

The calculated flow rates derived from the measured values exceeded the minimum required flow for the hot leg RTD's in all cases. Loop one was the least conservative with a margin of 53.75 gpm. Loops two, three and C) four exceeded the minimum required value with flow rates in excess of the minimum by 76.02 gpm, 64.77 gpm and 69.55 gpm, respectively.

Westinghouse subsequently issued a supplemental calculation which derived a new required minimum hot leg RTD flow of 110 gpm. Offsite engineering verified the calculation and the acceptance of the test. All four loops had hot leg flows greater than 110 gpm.

)

HOT LEGS I Difference Actual Calculated Required Cal.-Reg. Transit Loop No Flow 1 (qpm) Flow 2 (qpm) (Greater than 0) Time (sec)

)

1 118 64.25 53.75 0.54 l 2 140 63.98 76.02 0.46 3 142.5 77.73 64.77 0.55 4 135 65.45 69.55 0.48 l l

.O

1. Based on total, cold & hot leg measured flows.

l

2. Based on actual pipe lengths and diameters to give a one second transit time.

O O

i

() (0855M/0080M)

O 3.2.12s - REACTOR COOLANT SYSTEM (RCS) FLOW MEASUREMENT (HOT STANDBY), 2.63.31A i

OBJECTIVE The RCS Flow Measurement at Hot Standby test procedure was performed to C,) determine the RCS flowrates for each of the 4 loops and then the total flowrate. Also the reactor coolant pumps (RCPs) vibration measurements were taken from the installed vibration pickups during performance of the test.

TEST METHODOLOGY Prior to criticality data was obtained from the installed elbow tap differential pressure (d/p) instrumentation and used to find the RCS flowrates. Data was recorded every minute for ten minutes which included the following: RCS cold leg RTD resistance readings, and d/p transmitter output voltage. Then the RTD resistance readings were converted to f) temperature (*F) and the d/p transmitter output voltage readings were l converted to in. H 2O at 68'F. Each loop had three flow transmitters from which a d/p measurement was taken. The d/p readings (in. H2O at 68'F) were used to determine three flowrates for each loop by using the 4

Elbow Tap vs Reactor Coolant Cold Leg Volumetric Flow Rate figure. These three flowrates were averaged to obtain the loop average flow, and then c) t all four loops average flowrates were added to obtain the total RCS flowrate.

SUMMARY

OF RESULTS g) The total RCS flowrate must be equal to or greater than 339,840 gpm, which is 90% of the thermal design flow, as determined by elbow tap d/p prior to criticality. The expected flowrate for each loop was greater than or equal to 84,960 gpm with the actual measured average loop flowrates being: Loop 1 = 103,833 gpm; Loop 2 = 103,133 gpm; Loop 3 =

108,133 gpm; and Loop 4 = 102,933 gym. The sum of the average loop g) flowrates gives the total RCS flowrate of 418,032 gpm.

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C) '

3.2.12b - REACTOR COOLANT SYSTEM (RCS) FLOW MEASUREMENT (30, 50, 75 & 100%

POWER), 2.63.31B OBJECTIVE

() The RCS Flow Measurement test procedure was performed at various reactor i power levels (30, 50, 75, & 100% power) to determine the RCS flowrate for i each of the 4 loops and also the total RCS flowrate.

TEST METHODOLOGY '

C) While the plant was at 30, 50, 75 & 100% reactor power level data was obtained to determine the RCS flowrate. This data consisted of RCS hot and cold leg RIDS both nominal and reverse polarity resistance readings.

These RID resistance readings were converted to temperature (*F) and then by the use of Steam Tables, hot & cold leg enthalpy and cold leg specific volume were determined. Thi's data along with thermal power values from C) startup test 2.47.32C, D, or E, as appropriate, were used to calculate the flowrates. This method of calculation was used each time the test was performed at the various power levels. During the 50, 75, & 100%

power tests the ID loop'RCS cold leg temperature was taken from the spare RTD because the normal RTD had failed. Then during the 75 & 100% power

,3 tests the 1C loop RCS hot leg temperature was taken from its spare RTD

's because the normal RTD had failed. After completion of all the tests the IC hot leg and the 1D cold leg RTD's were replaced with new ones and calibrations were performed. The system was returned to normal and no further testing was performed.

The actual calculation flowrates for the various power levels were as I) follows:

RCS LOOP FLOWRATES (GPM)

TOTAL RCS

% P WER L P1 L P2 LOOP 3 LOOP 4 FLOWRATE O

30% 106,465.0 107,689.0 112,646.0 99,816.0 426,616.0 50% 105,303.4 103,065.4 105,289.4 99,116.6 412,774.0 75% 103,690.8 100,801.5 103,910.2 101,528.7 409,931.0 100% 101,652.7 100,798.2 102,163.0 100,348.1 404,962.0

'O All the results were consistent with the expected values and met the minimum flow requirements for total RCS flowrate.

SUMMARY

OF RESULTS

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

% POWER TOTAL FLOWRATE  % THERMAL DESIGN VALUE .

1 30% 1 339,840 gpm 90% l

() 50% 1 377,600 gpm 100%

75% 1 390,400 gpm 100% 2 390,400 gpm

() (0855M/0080M)

m U 3.2.13 - RC FLOW COASTDOWN, 2.63.32 i

OBJECTIVE The RC Flow Coastdown test was performed with the unit at Hot Standby to z) verify the measured core flow exceeds or is equal to the flow assumed in the accident analysis. In addition, time delays for low flow,.

undervoltage and underfrequency trips were verified to be within acceptable limits.

TEST METHODOLOGY

') Strip chart recorders were connected to the solid state protection system to monitor reactor coolant flow characteristics and reactor trip breaker positions. Data from the strip charts were then plotted on various graphs to verify acceptability of the measured flow values and time delays.

O

SUMMARY

OF RESULTS The time delays for the low flow, underfrequency and undervoltage trips ,

all met the required maximum values of 1.0, 0.6 and 1.5 seconds, respectively. Actual values obtained were 0.94, 0.185 and 0.655 seconds, 3 respectively.

The measured core flow for the first 10 seconds of the event did not meet the required curve in the FSAR. However, Westinghouse provided a new, up to date analysis technique which was reviewed by offsite engineering.

The measured flow was verified adequate per this new technique and a

) revision to the FSAR was submitted.

In addition the reactor coolant ramps were verified to trip within 100 i msec of each other. Actual valuu obtained was 20 msec. This was acceptable.

3 O

O (0855M/0080M)

I r3 3.2.14 - REACTOR COOLANT SYSTEM LEAK TEST, 2.63.33 OBJECTIVE The purpose of this procedure was to verify the reactor coolant system 3 (RCS) leak tightness after the system had been closed.

TEST METHODOLOGY With the plant in in Hot Standby conditions, prior to initial criticality, the reactor coolant system was tested to verify leak O tightness. The reactor coolant pressure was increased to between 2345 and 2355 psig for testing purposes. This was accomplished through closure of the pressurizer PORVs and block valves and by energizing the pressurizer heaters. Due to a misinterpretation this closure of both pressurizer PORVs and the block valves resulted in a Technical Specification Limiting Condition for Operation to be exceeded when the O block valves were not de-energized within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. When the problem was observed. the test was exited and the aforementioned valves were returned to operable status. When the test was re-entered, the Technical Specification Action Statement was followed by de-energizing the block valves. After the pressure had stabilized, the leak test was conducted by thorough visual inspection. In this inspection, the reactor pressure 3 vessel, pressurizer and all four reactor coolant loops were verified for leak tightness. In addition, the reactor cavity sump, reactor coolant drain tank and containment floor sump levels were measured. Also, the unidentified, identified, and controlled leakage rates were determined.

SUMMARY

OF RESULTS 3

The acceptance criteria for this test was to verify that RCS leakage would be: less than or equal to 1 gpm unidentified leakage, less than or equal to 10 gym identified leakage, less than or equal to 40 gym  ;

controlled leakage, and no pressure boundary leakage. Through the visual l inspection, no pressure boundary leakage was observed. Included in this I 3 were observations of the containment floor sump and the reactor cavity sump. No leakage was indicated (ie, no increase in levels), and during the test duration, neither of the containment floor sump pumps nor the reactor cavity sump pump were run. The unidentified leakage was determined to be 0.395 gym. The identified leakage was found to be'O.34 gpe. Finally, the controlled leakage was measured for each reactor O coolant pump seal injectior flow and totalled to 38.7 gym. This test thus verified acceptable leak tightness of the reactor coolant system.

O (0855M/0080M)

3.2.15 - CONTROL ROD DRIVE MECHANISM OPERATION TEST, 2.64.31 OBJECTIVE The test objectives were to verify the proper slave cycler timing, to 3~

perform an operational check of each control rod drive mechanism (CRDM) with a rod cluster control assembly (RCCA) attached prior to initial use of the mechanism under the conditions stated below and to demonstrate that the auto and manual rod withdrawal blocks functioned as designed.

TEST METHODOLOGY q

This test was performed under two plant conditions: Mode 5 - cold, no flow, and Mode 3 hot, full flow. The proper operation of the slave cycler timing was verified under cold, no flow conditions. The CRDM operational check was performed under both conditions, and the auto and manual rod withdrawal blocks were tested only under hot, full flow

] conditions.

Slave cycler timing was performed with all rods positioned at the core bottom. Initially, a single rod was withdrawn 50 steps to ensure the rodlet tips were at least 6 steps above the constricted thimble tube dashpot region. The rod was withdrawn 6 steps and then inserted 6 steps,

2) while slave cycler timing data was recorded. The data monitored included lift coil currents, movable and stationary gripper currents, and a sound pickup from a carbon microphone attached to the top end of the rod travel housing. The tested rod was inserted to the bottom of the core and the above procedure was repeated for each power cabinet.

3 The CRDM operational check was performed with all rods initially at the core bottom. A single rod was then withdrawn 48 steps. Next, data was recorded while the rod was withdrawn 12 steps, and then inserted 12 steps. The same parameters were monitored here as in the slave cycler timing procedure. This was repeated for both control and shutdown rods.

[) Auto and manual rod withdrawal blocks were also tested. A signal simulating 50*4 nuclear power was fed to the power range channels, and the appropriate signals for turbine first stage pressure and Tave generated.

By generating a T error signal and using jumpers and tripping various bistables, the following rod blocks & alarms were checked:

I) 1) Low Power Block of Automatic Rod Withdrawal (C-5)

2) Auto Rod Withdrawal Block When Control Bank D is Above Withdrawal Limit (C-11)

3) Intermediate Range High Flux Rod Stop (C-1)

4) Power Range High Flux Rod Stop (C-2)

, 5) OTAT Rod Stop & Turbine Runback (C-3)

J 6) OPAT Rod Stop & Turbine Runback (C-4)

7) Low & Low Low Insertion Limits for Control Banks A, B, C and D O

3 (0855M/0080M)

O 3.2.15 - CONTROL ROD DRIVE MECHANISM OPERATION TEST, 2.64.31 (Continu-d)

SUMMARY

OF RESULTS The slave cycler timing traces were found to be acceptable for all five

[) power cabinets. The operability of the CRDM was verified for all 53 CRDMs by demonstration of control rod insertion and withdrawal. Several of the traces indicated some of the coil voltages fell outside ranges stated in the acceptance criteria. However, the voltages were determined to be acceptable by Westinghouse and the results were accepted by offsite engineering review. The section of the test dealing with rod insertion

') limits (RILs) had to be reperformed due to a miscalibration of the RIL computer. After recalibration all points were found to be correct and acceptable.

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(0855M/0080M)

O 3.2.16 - ROD CONTROL SYSTEM CHECKOUT, 2.64.32 OBJECTIVE The rod control system checkout test demonstrated that the rod control b system performed the required control and indication functions in the areas listed below and verified it was ready for use just prior to initial criticality.

1. Individual Bank Rod Motion Test

2. Bank Overlap Operation Checkout O 3. Non-Urgent and Urgent Alarms for Power Cabinets and Logic Cabinets

4. D.C. Hold Cabinet TEST METHODOLOGY l Each rod control cluster (RCC) bank was singly withdrawn from the core 48

'O steps, inserted to 8 steps, and then fully inserted to O steps. This was done to verify the proper operation of the demand step counter, digital rod position indication (DRPI) system and proper annunciator response.

Bank overlap was set to 12 steps and all shutdown banks were withdrawn to

[

20 steps. Control banks were withdrawn u t*il control bank D (CBD) was at 20 steps to demonstrate proper overlap. All RCC banks were then O reinserted. The non-urgent alarm was tested for each cabinet by pulling fuses on one of the two redundant power supplies. The urgent alarm was tested for each cabinet by pulling a circuit card. The D.C. hold cabinet was checked to verify it was capable of supplying the necessary current to the stationary gripper coils.

O

SUMMARY

OF RESULTS The rod motion indicator (ROD IN/ ROD OUT) operated satisfactorily for all nine RCC banks. The demand step counters functioned properly for all RCC banks. Rod speed indication was acceptable for each RCC bank and all speeds were acceptable. Bank overlap performed properly. Non-urgent and O

urgent annunciation in the control room was completed successfully. The proper current reduction was shown upon an urgent failure. The D.C. hold cabinet exceeded its acceptance criteria by supplying too high a voltage to the stationary gripper coil. A station action item was generated, and then reviewed by offsite engineering. The test results were accepted band n Wutinghoun evaluation.

O l

l l

O O 4 (0855M/0080M)

O 3.2.17 - AUTCMATIC REACTOR CONTROL, 2.64.34 OBJECTIVE This procedure demonstrated the capability of the automatic reactor

() control system to maintain reactor coolant system average temperature within acceptable tolerance of Tref under steady state and transient conditions.

TEST METHODOLOGY C) With reactor powe,r stabilized at approximately 30% and Tave matched to Tref, rod control was placed in automatic to monitor for oscillations.

After approximately fifteen minutes, Tave was increased to 6*F higher than Tref by manual withdrawal of control bank D. Rod control was then placed in autcmatic and Tave allowed to stabilize. After Tave stabilized, rod control was placed in manual to decrease Tave 6'F lower '

C) than Tref by insertion of control bank D. Rod control was placed ba'ck in automatic and Tave was allowed to stabilize.

SUMMARY

OF RESULTS During steady state operution, it wac found that Tave maintained Tref C) with no problems being encountered. When Tave was increased, it took approximately 5 minutes to return Tave to within 1.5'F of Tref. When Tave was decreased, it took approximately 2.5 minutes to return Tave to within 1.5*F of Tref. Both of these times were well within the acceptance criteria time of 10 minutes. Because no problems or C) abnormalities were identified, no adjustments to the system were necessary.

O O

O O

(0855M/0080M)

C) 3.2.18 - ROD DROP TIME MEASUREMENT, 2.64.35 I l

l OBJECTIVE The purpose of this procedure was to determine the amount of time

() required to drep each RCCA from its fully withdrawn position, 228 steps, to the point of entry into the dashpot region. This was done in accordance with Technical Specifications 3.1.3.4 which states the maximum drop time as 2.2 seconds for hot full flow conditions.

TEST METHODOLOGY C)

With all rods fully inserted and the RCS boron concentration greater than 2000 ppm, one bank of rods was withdrawn to 228 steps. At that point, the digital rod position indication (DRPI) was deenergized and fuses pulled to release the stationary grippet for one RCCA and drop that rod.

• A carbon microphone was attached to the top of the control rod drive '

C) mechanism to obtain a sound trace on a strip chart recorder. Other signals monitored were from the Data A cabinet only, the Data B cabinet only, Data A & B, and the stationary gripper coil voltage. Drop times were evaluated from the Data A & B voltage trace. Any drop time that fell outside of two standard deviations from the average was dropped three additional times. The proper operation of the deshpot was also

() verified from the traces. This test was performed at four plant conditions: cold no flow, cold full flow, hot no flow, and hot full flow conditions.

SUMMARY

OF RESULTS C)

The initial test procedure attempted to utilize the automatic rod drop test system to drop the rods and calculate the drop times. After several mechanical problems were resolved, the system was found to have a software error in the calculation of the drop times. As a result, the

" fuse pulling" method was utilized to gather the required data. All rods C) in all test conditions dropped in less than the Technical Specification requirement of 2.2 seconds. The following times were recorded for the conditions listed.

AVERAGE FASTEST SLOWEST Cold No' Flow 1.30 seconds 1.27 seconds 1.31 seconds C)

Cold Full Flow 1.58 seconds 1.53 seconds 1.67 seconds Hot No Flow 1.27 seconds 1.25 seconds 1.29 seconds Hot Full Flow 1.46 seconds 1.41 seconds 1.52 seconds At all four test conditions, the voltage traces were examined and proper operation of the dashpot region was verified.

O O

(0855M/0080M)

O 3.2.19 - REACTOR PROTECTION LOGIC, 2.68.30 OBJECTIVE I The objective of the test was to verify proper operation of the automatic

() and manual reactor trip circuitry and to verify proper operation of the reactor trip breakers. The test was also designed to verify proper operation of the permissive and block circuitry.

TEST METHODOLOGY

() The test verified on both trains, with the use of the semi-automatic logic tester, proper operation of the reactor trip system logic combinations employed in the solid state proteccion system (SSPS). The test also verified proper operation of the reactor trip breakers in response to automatic and manual reactor trip signals. Additionally the test verified proper operation of the P-4 permissive and the resistance '

(3 from logic ground to chassis ground in each train of the SSPS.

SUMMARY

OF RESULTS The test proved SSPS logic combinations for reactor trip inputs as well as permissive and block inputs. Also, the test showed that the reactor C) trip breakers were operational when a signal was received from either an automatic or manual trip signal. Finally the test verified proper operation of the P-4 permissive and the SSPS ground circuits.

O O

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(0855M/0080M)

0 3.2.20 - PRESSURIZER SPRAYS, HEATERS AND BYPASS FLOW ADJUSTMENTS, 2.69.30 l

OBJECTIVE ]

The Pressurizer Sprays, Heaters and Bypass Flow Adjustments test was

() performed to verify pressurizer spray and heater effectiveness. In addition, spray line bypass valves were adjusted to maintain spray line j temperature above 540*F. i TEST METHODOLOGY

) In order to set the spray line flow, the valves were closed and the temperatures allowed to stabilize. The valves were then opened in 1/16 ,

turn increments until a satisfactory temperature reading was achieved. -

To verify spray effectiveness, the heaters were manually isolated and the spray valves placed in the full open position. Pressurizer parameters '

were monitored via strip chart recorders. These parameters were then C) analyzed and plotted to verify the pressure transient fell within the .

allowable limits. l To verify heater effectiveness, the spray valves were manually isolated and the heaters were placed to the full on position. A few of the  ;

heaters were found inoperable due to wiring problems. Repairs were 3 performed and the test section was re-executed. Pressure parameters were monitored via strip chart recorders. These parameters were then analyzed and plotted to verify the pressure transient fell within the allowable limits.

SUMMARY

OF RESULTS 3

The spray valves were set to 1/8 turn and 1/16 turn to allow enough bypass flow to maintain spray line temperature above 540'F. The pressure transient resulting from the spray effectiveness test fell within the required band.  :

C) The pressure transient resulting from the heater effectiveness tests fell outside the required band. Offsite engineering re-evaluated the data and made adjustments for absolute pressure. The new plot then fell within ,

the required range. The test results were considered acceptable. ,

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O O O O U U O U U U U l FIGURE 3.2.20-1

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) 3.2.21 - AUXILIARY BUILDING HVAC SYSTEM PREOPERATIONAL TEST, 2.84.11 OBJECTIVE The Auxiliary Building HVAC System Preoperational test procedure was [

] performed to satisfy the commitment that preoperational tests be performed on all safety-related equipment prior to being declared I operational. The specific objectives were to meet the acceptance l criteria as follows:

s

1. Verify that the Auxiliary Building HVAC system maintains the O Auxiliary Building main areas at a negative pressure relative to atmosphere under normal operating conditions.

2. Verify that the following Auxiliary Building cubicles are maintained greater than or equal to 1/4 inch water gauge negative pressure relative to atmosphere:

O Centrifugal Charging Pump 1A Room Centrifugal Charging Pump 1B Room Safety Injection Pump 1A Room Safety Injection Pump 1B Room Residual Heat Removal Pump 1A Room O Residual Heat Removal Pump 1B Room Residual Heat Exchanger lA Room Residual Heat Exchanger IB Room Containment Spray Pump 1A Room Containment Spray Pump 1B Room '

O 3. Verify that each of the three non-acceesible area exhaust plenums can fully serve the exhaust from the non-accessible spaces in charcoal booster fan bypass mode. Each plenum should exhaust 66,900 cfm 10%. i

4. Verify that each of four accessil)le area exhaust plenums can fully 3 serve the exhaust from the accessible spaces. Each plenum should exhaust 54,000 cfm ! 10%.

5. Verify that each of the six Auxiliary Building non-accessible area exhaust plenum charcoal booster fans can fully serve the spaces g exhausted by the system. Each of the booster fans should exhaust 66,900 cfm i 10%.

6. Verify that the Auxiliary Building charcoal booster fans start on a safety injection actuation signal with fans OVA 03CA, OVA 03CC, and OVA 03CE delay for 13-17 seconds on start (back-up fans if the lead

,g fan does not start).

7. Verify that the Auxiliary Building charcoal booster fans maintain the dirty areas of of the Auxiliary Building under negative pressure relative to atmosphere when the supply and exhaust fans are_off.

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(0855M/0080M) l .- . - - . - ,. , - -

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.O 3.2.21 - AUXILIARY BUILDING HVAC SYSTEM PREOPERATIONAL TEST, 2.84.11 (Continued)

8. Verify that the Auxiliary Building exhaust air filter train bypass dampers close when their associated charcoal booster fan starts.

'O

9. Verify that either of the two Fuel Handling Building exhaust air filter trains can fully serve the spaces exhausted by the system.

Each filter train should exhaust 21,000 cfm i 10% for all appropriate flow paths.

O 10. Verify that the Fuel Handling Building exhaust air filter train bypass dampers close when their associated charcoal booster fan starts.

11. Verify that the Fuel Handling Building charcoal booster fans start on a high radiation signal. Fan OVA 04CA delayed for l'3-17 seconds O on start (back-up fan if ova 04CB does not start).

12. Verify that the Fuel Handling Building charcoal booster fans start on a SI actuation signal. Fan OVA 04CA delayed for 13-17 seconds on start (back-up fan if OVA 04CB does not start).

O 13. verify that the Fuel Handling Building exhaust air filter train i charcoal isolation dampers open when their associated charcoal j booster fan starts.

14. Breakers shall close or trip when actuated from either the local, or remote shutdown control switch operation as designated.

15. Breakers with protective functions shall trip when that protective function is actuated.

4

16. Verify the "Go-Go" circuits (600 relays) perform to actuate a O **f'9"*"d* f**t"#**

TEST METHODOLOGY l All equipment, dampers, instrumentation, control and alarm functions of the Auxiliary Building HVAC system were tested for proper operation.

T88 tin 9 inv lved 8imulatin9 8i9na18 t verify Proper alarms and O actuations. Integrated testing was performed with the equipment.

operating so that air flow rates and system pressures could be measured.

SUMMARY

OF RESULTS 10 The review of this procedure included a review of retest procedures R-287~

and R-288 which were requested by offsite engineering. All equipment, dampers, instrumentation, control, and alarm functions, associated with .

the Auxiliary Building HVAC system performed acceptably during testing.

All air flowrates, pressures, and ESF actuations were found acceptable for Unit 1 operation by offsite engineering.

O O 76-(0855M/0080M)

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) 3.2.22 - HEAT CAPACITY VERIFICATION FOR AUXILIARY BUILDING HVAC SYSTEM, 2.84.30 l 1

OBJECTIVE  !

The Heat Capacity Verification for Auxiliary Building HVAC System 9- procedure was performed to provide heat removal capacity data for the Auxiliary Building ventilation system. The data obtained in the procedure was analyzed to verify that the design heat loads could be l removed.

TEST METHODOLOGY )

O In order to obtain data for the heat rea. oval capacity of the Auxiliary Building ventilation system, the HVAC system was divided into subsystems and tested independently. The temperatures, heat loads, cooling air flow rates and water flow rates for each subsystem were measured. The data for each subsystem was then analyzed to determine that the design heat J loads could be removed.

SUMMARY

OF RESULTS Initially, the Heat Capacity for Auxiliary Building HVAC System procedure was partially executed and submitted to offsite engineering for 3 approval. The data obtained in the initial execution was determined to be inconclusive by offsite engineering. Subsequent test data was obtained after procedure changes were made that was determined to be acceptable by offsite engineering. The analyzed data yielded temperatures that were below the limits set by the acceptance criteria as follows:

3 SUBSYSTEM ANALYZED TEMP (*F) ACCEPTANCE CRITERIA (*F) 1A SX Pump Room 110 1 122 1B SX Pump Room 113 i 122  !

1A CS Pump Room 116 i_122 l 1B CS Pump Room 116 ~< 122 l 3 1A RHR Pump Room 117 i 122 1B RHR Pump Room 117 1 122 1A CV Pump Room 111 i 122 1B CV Pump Room 108 1 122 1A SI Pump Room 108 .i 122 1B SI Pump Room 115 ~< 122 O Pos. Disp. Charging Pump Room 118 1 122 Diesel Driven Aux Feedwater 113 1 133 Motor Driven Aux Feedwater 110 1 133 O Component Cooling Pump Area 98 $ 104 0

1 1

(0855M/0080M)

O 3.2.23 - HEAT CAPACITY VERIFICATION FOR CONTROL ROOM HVAC SYSTEM, 2.85.30 OBJECTIVE The Heat Capacity Verification for Control Room HVAC System procedure was

() performed to provide heat removal capacity data for the control room ventilation system. The data obtained in the procedure was analyzed to verify that the design heat loads could be removed.

TEST METHODOLOGY C) In order to obtain data for the heat removal capacity of the control room ventilation system, each train (A & B) was tested independently. The temperatures, heat loads, and cooling air and water flow rates for each train were measured. The data for each train was then analyzed to determine that the design heat loads could be removed.

'O

SUMMARY

OF RESULTS Analyzed data showed that both trains of the control room HVAC system were able to remove the design heat loads as follows:

"A" chiller calculated heat removal rate: 2.76 x 106 BTUH C) "A" coil calculated heat removal rate: 2.48 x 106 BTUH "A" train heat load (corrected): 2.1 x 106 BTUH "B" chiller calculated heat removal rate: 2.76 x 106 BTUH "B" coil calculated heat removal rate: 2.48 x 106 BTUH "B" train heat load (corrected): 2.1 x 106 BTUH

)

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.O (0855M/0080M)

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3.2.24 - HEAT CAPACITY VERIFICATION FOR DIESEL GENERATOR VENTILATION, 2.86.30 OBJECTIVE The Heat Capacity Verification for Diesel Generator Ventilation procedure 3 was performed to provide heat removal capacity data for the diesel generator ventilation system. The data obtained in the procedure was analyzed to verify that the design heat loads could be removed.

TEST METHODOLOGY

] In order to obtain data for the heat removal capacity of the diesel generator ventilation system, each diesel generator room was tested independently with the associated diesel generator operating. The temperatures, heat loads, and cooling air temperatures for each diesel generator room were measured with the ventilation system in a recirculation mode. The data for each diesel generator room ventilation 3 system was then analyzed to determine that the design heat loads could be -

removed.

SUMMARY

OF RESULTS Offsite engineering analyzed the data and determined that the systems J would be able to remove the design heat loads. Since the diesel generators were operating under load and the ventilation fans were operating in 100% recirculation mode, the systems were operated near design conditions. The lA diesel generator ventilation system maximum room temperature was 134*F with a supply air temperature of ll4*F. Since ll4*F is 9*F above the 95"F design outside air temperature, the room

-) temperature for the 1A diesel generator ventilation system was capable of maintaining less than the 130*F temperature limit. The IB diesel generator ventilation system maximum room temperature was 108.8*F with a supply air temperature of 88*F. Since 88'F is 7*F below the 95*F design outside air temperature, the room temperature for the IB diesel generator

_ ventilation system was capable of maintaining less than the 130*F l

'J temperature limit.

O O

O (0855M/0080M)

l 3.2.25 - HEAT CAPACITY VERIFICATION FOR PRIMARY CONTAINMENT VENTILATION, 2.93.30 j 1

OBJECTIVE O The Heat Capacity Verification for Primary Containment Ventilation procedure was performed to provide heat removal capacity data for the l primary containment ventilation system. The data obtained in the procedure was analyzed to verify that the design heat loads could be removed.

l O TEST METHODOLOGY In order to obtain data for the heat removal capacity of the primary containment ventilation system each electrically independent train of  ;

primary containment ventilation was tested with the other train off. The temperatures, heat loads, and cooling air and water temperatures for each O train of primary containment ventilation were measured at 5% reactor power. The data for each train of primary containment ventilation was then analyzed and extrapolated to 100% reactor power to determine that the design heat loads could be removed.

SUMMARY

OF RESULTS O

Offsite engineering analyzed the data obtained and requested that more data be taken in Retest R-281. Both tests were then analyzed as a whole and the conclusion was made that the system appeared capable of removing the containment heat load. Offsite engineering then requested that various parameters of the primary containment ventilation system and

.O containment heat loads be monitored during the startup/ power ascension phase as follows:

Reactor Containment Fan Cooler (RCFC) Inlet Temperatures

{ Reactor Power Level i Containment Chiller Amperage O Containment Internal Pressure I CRDM Exhaust Fan Discharge Temperatures Offsite engineering requested that the above data be taken daily and submitted to offsite engineering during the startup/ power ascension phase, ffsite engineering als requested that data for cooling water O flow rates, water temperatures, and computer input air temperatures would j

be taken at 50% and 100% power. The data taken at 50% and 100% was l

! submitted with the respective sequence document.

The data revealed that the containment bulk temperature is being kept below the 120*F Technical Specification limit. However, the method of O obtaining the containment temperature specified by the Technical J

Specifications is conservative in that the containment temperature calculated per the Technical Specifications is higher than the true containment bulk temperature. Therefore, three RCFCs will be required to maintain containment temperature within the limitations of the Technical Specifications.

'O "O (0855M/0080M) l

3.2.26 - HEAT CAPACITY VERIFICATION FOR SAFETY VALVE ENCLOSURE ROOM VENTILATION SYSTEM, 2.97.44 OBJECTIVE

() The Heat Capacity Verification for Safety Valve Enclosure Room Ventilation System procedure was performed to provide heat removal capacity data for the safety valve enclosure room ventilation system.

The data obtained in the procedure was analyzed to verify that the design heat loads could be removed.

TEST METHODOLOGY O

The heat removal capacity procedure was performed at 5% i 2% reactor power and extrapolations were made to 100% power conditions. The temperatures, heat loads, and cooling air flowrates for the safety valve enclosure room ventilation system were measured. The data obtained was

() then analyzed to determine that the design heat load could be removed.

SUMMARY

OF RESULTS The data obtained was submitted to offsite engineering for analysis.

Offsite engineering requested more data which was supplied by retest

() R-283. Extrapolations to design conditions determined that temperatures would be kept below the acceptance criteria value of 122*F.

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3.2.27 - HEAT CAPACITY VERIFICATION FOR SWITCHGEAR HEAT REMOVAL VENTILATION 2.99.30 i

OBJECTIVE

.() The Heat Capacity Verification for Switchgear Heat Removal Ventilation procedure was performed to provide heat removal capacity data for the switchgear heat removal ventilation system. The data obtained in the procedure was analyzed to verify that the design heat loads could be removed.

O TEST METHODOLOGY In order to obtain data for the heat removal capacity of the switchgear heat removal ventilation system, the system was divided into subsystems i and tested independently. The temperatures, heat loads, and cooling air l flowrates'for each subsystem (Div 11 and'Div 12) were measured. The data

() for each subsystem was then analyzed to determine that the design heat loads could be removed.

SUMMARY

OF RESULTS I i

. Test data was submitted to offsite engineering for analysis. The

.)

analysis concluded that extrapolated room temperatures for the subsystems l were between 98*F and 10l*F which met the acceptance criteria of less i than or equal to 104*F.

l c) I O

O O

i o (0855M/0080M)

O 3.2.28 - HEAT CAPACITY VERIFICATION FOR MISCELLANEOUS ELECTRICAL EQUIPMENT ROOM VENTILATION SYSTEM, 2.128.30 OBJECTIVE

• O The Heat capacity Verification for Miscellaneous Electrical Equipment Room Ventilation System procedure was performed to provide heat removal capacity data for the miscellaneous electrical equipment room ventilation system. The data obtained in the procedure was analyzed to verify that the design heat loads could be removed.

O TEST METHODOLOGY 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. The temperatures for each subsystem were measured. Th'e data for each subsystem (Div 11

.O ~

and Div 12) was then analyzed to determine that the design heat loads could be removed.

SUMMARY

OF RESULTS The data, as analyzed by offsite engineering showed that each subsystem O of miscellaneous electrical equipment room (MEER) ventilation was able to remove the design heat loads as follows:

SUBSYSTEM CALCULATED (*F) DESIGN (*F)

Division 11 MEER 98 < 104 O Division 12 MEER 100 [104 O

O O

O O (0855M/0080M)

O 3.2.29 - HEAT CAPACITY VERIFICATION FOR REMOTE SHUTDOWN CONTROL ROOM VENTILATION, 2.129.44 OBJECTIVE C) The Heat Capacity Verification for Remote Shutdown Control Room I Ventilation procedure was performed to provide heat removal capacity data j for the Unit 1 remote shutdown control room ventilation system. The data '

obtained in the procedure was analyzed to verify that the design heat loads could be removed.

O TEST METHODOLOGY Temperatures, heat loads, cooling air flowrates, and refrigeration unit parameters were measured for the Unit I remote shutdown control room.

Extrapolations to design conditions were then made to determine that the design heat load could be removed.'

SUMMARY

OF RESULTS The data obtained was submitted to offsite engineering for analysis.

More data was requested and supplied via a phone conversation to offsite engineering. Extrapolations to design conditions showed that the

() temperature within the Unit I remote shutdown control room would be kept below the acceptance criteria maximum of 75'F. I I

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(0855M/0080M)

4 C)  !

3.2.30 - STEAM GENERATOR MOISTURE CARRYOVER, 2.51.45 OBJECTIVE j i

a To measure the steam generator moisture carryover (ie. steam quality) by C) use of a non-volatile, water-soluble, radioactive tracer.

TEST METHODOLOGY Initially, the secondary side of the plant was isolated (leaks repaired or isolated, hotwell makeup and steam generator blowdown isolated). Then

() one curie of Na24 was injected into the condensate system and allowed to mix for approximately tw) hours. After sufficient mixing time three sets of samples were drawn from each of the four steam generator blowdown sample connections and from the common main feedwater header. Using

, equations which ratio the activity levels of the blowdown samples to the

,s feedwater samples, the average ' moisture carryover of four steam

generators was determined.

SUMMARY

OF RESULTS This test was run on three separate occasions. The data from the second run was not considered valid due to improper steam generator blowdown

) valve lineups.

The other two test runs produced the data tabulated below. Steam l

generator level and power level were varied to check the sensitivity of moisture carryover to these parameters. Until steam generator modifications are effected to reduce moisture carryover to below 0.25% at

() full power, Byron Unit 1 is being operated at or below 92% power in order to protect the turbine blading from erosion.

]

4 Moisture Carryover Test Results i

() Reactor Power Steam Generator Level Average Moisture (From NIS - %) (Narrow Range - %[_ Carryover - %

! 100 66 0.78 97.8 66 0.71 94.5 66 0.31

() 92 66 0.21 94.5 61 0.14 98.8 61 0.20 l

l O

i 1

i C) (0855M/0080M)

O 3.2.31 - DYNAMIC AUTOMATIC STEAM DUMPS CONTROL, 2.51.46 '

OBJECTIVE  !

The Dynamic Automatic Steam Dump Control test was performed to verify

() proper response of the steam dump valves when controlled by the steam pressure controller, the turbine trip controller, and the load rejection controller.

TEST METHODOLOGY I

C) The steam pressure controller was tested by raising and lowering reactor  !

power between 0 and 6% power and monitoring steam dump response. The '

response was also monitored during the turbine roll up to synchronous speed. In all cases the steam pressure controller was to modulate the steam dumps to maintain steam pressure at 1092 1 13 psig. The turbine trip and load rejection controller were tested by changing power level C? between 0% and 7% power and monitoring reactor coolant average temperature, steam dump demand, and steam header pressure.

SUMMARY

OF RESULTS The steam pressure controller maintained steam pressure within the 1092 i C) 13 psig acceptable band with the greatest deviation being 4 psig. The turbine trip and load rejection control maintained a stable plant as )

expected. However, the actual Tave maintained by the steam dumps was as much as 10% higher than expected. This test was closed and accepted by issuing a station action item to investigate the capacity and overall operation of the steam dumps. As a result of additional testing, the C) capacity of the steam dump valves was verified to be equivalent to 40%

steam flow, the positioners for each of the 12 steam dumps were adjusted -

to allow full valve stroke, and a modification of the air supply to the 1 valves was implemented to allow faster stroke times.

Final startup testing, such as the 50% Load Reduction from 100% power,

() verified proper operation of the steam dumps during plant transients and stabilization periods'.

O O

O

-Q (0855M/0080M)

O 3.2.32 - SAFE 1*Y PARAMETER DISPLAY SYSTEM (NARROW RANGE ICONIC), SPP-85-10 OBJECTIVE The Safety Parameter Display System (SPDS) special procedure was

() performed to verify proper operation of the narrow range iconic according ,

to specifications provided for the system. The procedure was performed to verify proper operation:of the iconic through manipulation of inputs which exercised the software to all bounds stated in the specification.

TEST METHODOLOGY IO The narrow range iconic contains eight " spokes" which provide indication of vital plant parameters. To exercise each spoke values were entered at the process computer and recorded as they appeared on the graphics i screen. Each spoke was exercised such that both low and high alarms were verified for all inputs.' Spokes with redundant inputs were tested to

,)

I verify proper averaging of inputs and proper quality checking of inputs.

Inputs were dummied in for each spoke such that output was verified for j analog output values displayed on the* graphic, all color conventions and proper iconic deflection.

]

SUMMARY

OF RESULTS

.O This procedure was run concurrently with the 30% testing sequence.

1 No acceptance criteria were stated in the procedure. During the running of the procedure 13 deficient conditions were noted. Deficient conditions requiring changes to the SPDS program consisted of the I) following problems:

All values being displayed on the screen were being displayed as "PPP". This was resolved by a patch to 6 core locations which

! alleviated the problem.

) -

Containment floor drain sump levels active value failed to turn to red when in alarm. This problem required the loading of 14 database

', points to resolve it.

Particulate radiation status failed to change color properly when in alarm. This problem was corrected by a revision to the graphics

}) subsystem. l High tic mark on the radiation spoke failed to turn red when in high j alarm. This problem was resolved by reloading the graphics software on the process computer with a revision incorporated to correct this

() problem.

Radiation spoke went into high alarm despite the fact there were no inputs in alarm. This problem was resolved by reloading the graphic software with revisions incorporated to correct the problem.

O O (0855M/0080M)

3 3.2.32e - SAFETY PARAMETER DISPLAY SYSTEM (NARROW RANGE ICONIC),. SPP-85-10 i (Continued) l l

With the inputs to main steam loop A radiation in alarm main steam l loop B indicated alarm. This problem was alleviated by a patch to I

{~j the graphics software on the process computer. l With all inputs to. steam generator level at 66% the active value displayed on the screen was -3. Revision to the graphics software was required to correct the problem.

13 -

With all inputs to the steam generator level at 66% the active value was displayed in magenta. The revision required to alleviate this problem was provided by the revision for the previously listed problem.

Upon completion of this procedure along with the successful completion of J SPP-85-11, Byron Station was able to satisfy licensing commitment #5 of Attachment 2 to the full power license.

3 1

3 J

J 3

3 3 (0855M/0080M)

3.2.32b - SAFETY PARAMETER DISPLAY SYSTEM (WIDE RANGE ICONIC), SPP-85-11 OBJECTIVE 1

The Safety Parameter Display System special procedure was performed to

() verify proper operation of the wide range iconic according to specifications provided for the system. The procedure was performed to verify proper operation of the iconic through manipulation of inputs which +xercised the software to all bounds stated in the specification.

TEST METHODOLOGY O

l The wide range iconic contains eight spokes which provide indication of vital plant parameters. To exercise each spoke values were entered at l the process computer and recorded as they appeared on the graphics screen. Each spoke was exercised such that both low and high alarms were verified for all i'nputs. Spokes with redundant inputs were tested to

() verify proper averaging of inputs and proper quality checking of inputs.

Inputs were dummied in for each spoke such that output was verified for analog output values displayed on the graphic, all color conventions and proper Iconic deflection.

i i

SUMMARY

OF RESULTS O

This procedure was run concurrently with the 30% testing sequence.

No acceptance criteria were stated in the procedure. During the running of the procedure 15 deficient conditions were noted. Deficient conditions requiring changes to the SPDS program consisted of the

() following problems:

't The active value for containment pressure failed to turn red when above alarm limit of 3.4. This problem was resolved by a modification to the database and modification of a constant to the program.

The high tic mark failed to turn red when the spoke was in high alarm. This problem was alleviated by the same patches provided to clear the problem previously mentioned.

With the reactor coolant wide range pressure inputs at their

.C) reference value, the active value was red. To correct this problem four data base changes were required.

Upon completion of this procedure along with the successful completion of

, SPP-85-10 Byron Station was able to satisfy licensing commitment #5 of Attachment 2 to the Byron full power license.

)

i O

1

-g (0855M/0080M)

O 3.2.33 - LICENSE CONDITION ATT. 1, ITEM C (2.C.1), R-284 OMT:

The Byron Integrated Control Room pressure Test - Retest R-284 was

() performed to satisfy license condition Attachment 1, Item C (2.C.1). The procedure was to verify that the control room and upper cable spreading room could be maintained at proper pressures with the Control Room (VC),

Auxiliary Building (VA), Switchgear Room (VX), Miscellaneous Electrical Equipment Room (VV) ventilation systems operating in various modes.

O TEST METHODOLOGY Each control room ventilation system train was run in nine different operating combinations with the VA, VE, VX, and VV systems. During each combination the control room pressure with respect to surrounding areas was measured.

• C3

SUMMARY

OF RESULTS The pressures for each combination of the VC, VA, VE, VX and VV systems were compared to the acceptance criteria and found acceptable by offsite engineering. The control room pressure with respect to the turbine and C) auxiliary building was above 0.125 inches water gauge in all combinations of operation. Also, the upper cable spreading room was maintained at a pressure above positive 0.02 inches water gauge with respect to the turbine and auxiliary building in all test combinations as required by Technical Specifications.

O O

O O

O

() (0855M/J080M)

1

') 3.2.34 - LICENSE CONDITION C.6.D TESTED IN 2.84.11 SECTIONS 9.1, 9.2, 9.3 and 9.8 OBJECTIVE J The continuous thermistors for the Auxiliary Building ventilation system charcoal filter plenums had to be installed and tested to satisfy License Condition C.6.D.

TEST METHODOLOGY J Sections 9.1, 9.2, 9.3, and 9.8 of the Auxiliary Building Ventilation System Preoperational Test served to test the charcoal filter temperature switches. With the thermistors installed in the filter plenums, the temperature sensing elements and alarms were tested. Testing consisted of heating the element with a heat gun until local panel and main control board alarms annunciated. Once the alarms were annunciated the elements

'] were allowed to cool until the alarms reset.

SUMMARY

OF RESULTS  !

l l

All continuous thermistors of the Auxiliary Building charcoal filter plenums were tested prior to July 1, 1985. There were no deficient J conditions involving the functioning of the elements and alarms of the thermistors.

'3

3 J

D

- s J

m, "

(0856M)

O 3.2.35 - LICENSE CONDITION C.13 OBJECTIVE Modifications to ensure that the control room ventilation system may be C) used during an accident to protect operators within the criteria specified in 10CFR50, Appendix A, General Design Criteria 19 had to be completed prior to July 1,1985 in accordance with License Condition C.13.

TEST METHODOLOGY C3 The License Condition was satisfied in the following ways:

1. Control room recirculation charcoal plenums were tested to the guidelines of ANSI N510 by the construction acceptance tests.

2. Bubble tight dampers on the control room minimum outside air and

() makeup air intakes were installed and tested for zero leakage.

3. The main purge and maximum outside air dampers were blanked-off to prevent leakage until 54" bubble tight dampers can be installed.

SUMMARY

OF RESULTS Alltesting, modifications,andinstallationsEerecompletedtosatisfy the License Condition prior to July 1,1985.

O l

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I l

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'O c) -92 (0856M) 1

O 3.3 - PHYSICS TESTING I i

3.3.1 - INCORE FLUX MAPPING AT LOW POWER, 2.45.31 OBJECTIVE O

The objectives of this test were to obtain or verify high voltage

, plateaus for all six detectors required to perform an incore flux map, to verify the performance of the incore movable detector flux mapping (IC) system and to obtain a flux map at low power (less than 5*4). This included an initial operational alignment and checkout of the IC :3ystem

[3 and associated automatic data logging devices.

TEST METHODOLOGY The detectors were inserted into the core to the approximate maximum flux location in the core, and the applied voltage was incremented to

() determine the plateau curve (defined as the range of applied detector voltage over which the detector output is nearly constant). The operating voltage selacted was the mid-range of the plateau region. With

! the detectors set at their operating voltages, the sequence for obtaining a full core flux map was performed. Twelve passes through the core, each with one detector in the calibrate position, J10, were obtained. The

) data collected during the flux map was processed using an offsite incore program.

SUMMARY

OF RESULTS The plateau regions were determined and the operating voltages were

) established as follows: 81V, 90V, 100V, 81V, 90V, 90V for detectors A-F respectively. Analysis of the flux map data was as follows. The strip chart recorders were normalized. 57 of the 58 core locations were mapped, meetir.g the Technical Specification minimum of 45 locations (Technical Specification 3.3.3.2). The incore flux tilt was calculated z) as 1.01 where the acceptance criteria limit was less than 1.04. FDHN =

1.55 which was less than the 1.87 upper limit listed in the acceptance

criteria. The maximum value of Eg(Z) = 2.49 was less than the value of j 4.64 (K(Z)) = 3.0 (Where K(Z) is given in Technical Specification 3.2.2).

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1 i

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.O (0856M)

O 3.3.2 -INCORE MOVABLE DETECTOR AND THERMOCOUPLE MAPPING AT POWER (C'iWT'R CORE AND PARTI _Ak CORE), 2.45.32B-C -

OBJECTIVE O The purpose of this procedn: 9 was to provide a method for obtaining

)

quarter core flux maps and partial core flux maps during transient conditions of power.

TEST METHODOLOGY O In the 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> previous to the test a voltage plateau procedure had been executed to determine the operating vc1tages for the incore detectors. A full core flux mapping procedure had been performed less than 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> previous to the quarter core flux mapping (QCFM) or partial core flux map to provide a baseline for the hot channel factor comparisons. Both tests '

were performed as part of the Pseudo Rod Ejection at 30% power (see O Section 3.3.3). The QCFM was performed with RCCA D-12 withdrawn to 186, 221 and 228 steps. The flux maps were analyzed using the offsite IBM computer.

SUMMARY

OF RESULTS

-O All the flux maps required to support evaluation of the pseudo rod ejection test were obtained. The results of all the quarter core and partial core flux maps taken were determined acceptable by offsite engineering review.

O

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.O

O O 4 (0856M)

Q 3.3.3 - PSEUDO ROD EJECTION AT HOT ZERO POWER AND 30% POWER, 2.4$.33A-B OBJECTIVE This procedure verified that F T, at both hot zero power (HZP) and O 30% power, with rod control cl3 ster assembly (RCCA) D-12 fully withdrawn, and the total worth of the rod were less than the limits prescribed in Section 15 of the Final Safety Analysis Report.

TEST METHODOLOGY O For the HZP test (performed at 1% power) the control banks were positioned at their EZP insertion limits. RCCA D-12 was then withdrawn to 228 steps, during which the worth of RCCA D-12 was measured by RCS boron adjustment utilizing the rod worth tent procedure 2.64.30A and the boron endpoint test, 2.54.33. With RCCA D-12 at 228 steps, a full core flux map was obtained. '

O ,

At 30% power, the control banks were positioned at their hot full power (HFP) insertion limits. The worth of RCCA D-12 was measured by moving RCCA D-12 to the fully withdrawn position while holding turbine load constant which caused a corresponding increase in Tave. After Tave O

stabilized, RCCA D-12 was incerted to its original position. The measured change in Tave before and after rod withdrawal was multiplied by the isothermal temperature coefficient to obtain the worth of the pseudo ejected rod. RCCA D-12 was then borated out sita partial core flux maps being taken at 186 steps and 211 steps withdrawn. When RCCA D-12 reached 228 steps, a full core flux map was obtained. D-12 was then returned to its bank and verified to move with its bank.

SUMMARY

OF RESULTS Hot Zero Power Results The rod worth of RCCA D-12, from the Eank Worth Measurement at Zero Power test was found to be 0.487% AK/K, which was less than the acceptance criteria limit of 0.79% AK/K. ThehotchannelfactorF{,was calculated to be 9.1429 which was less than the acceptance criteria maximum of 11.5. Several deficiencies were written on this procedure which were subsequently resolved by offsite engineering.

O 30% Power Results RCCA D-12 had a worth of 0.02187% AK/K, meeting acceptance criteria of not greater than 0.025% AK/K. The hot channel factor, F , was calculatedtobe2.179andwasnotgreaterthantheTech{ical n

'O Specifications limit of 3.016 at 12 feet from core bottom and 29% power.

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o  !

O (0856M)

I 1

0 3.3.4 - ISOTHERMAL TEMPERATURE COEFFICIENT MEASUREMENT, 2.47.30 OBJECTIVE This test was performed to determine the isothermal temperature

() coefficient of reactivity, and to derive the moderator temperature coefficient of reactivity at the beginning of core life for the first fuel cycle.

TEST METHODOLOGY C) The isothermal temperature coefficient (ITC) was determined by calculating the change in reactivity induced by changing the temperature of the moderator, clad and fuel and dividing by the change in temperature. This test was performed at hot zero power conditions starting with a reactor coolant system (RCS) cooldown of 3*F at a rate of 10*F/hr. After a stabilization period, an RCS heatup was initiated for a '

C) 3*F increase at a rate of 10*F/hr. An X-Y plot of reactivity vs temperature was maintained throughout the cooldown and heatup portions of the test. Each cooldown and heatup was performed two times for each of the following rod positions: Control Banks D & C in (with B 2 200 steps), Centrol Bank D in (with C 2 200 steps) and all rods out (Control Bank D > 200 steps).

O

SUMMARY

OF RESULTS All values of the measured ITC were within 1 pcm/*F of one another and all were within i 3.0 pcm/*F of the acceptance criteria values. During the all rods out execution of the test the moderator temperature

() coefficient (MTC) of reactivity was determined to be -0.30 pcm/*F meeting the acceptance criteria and the Technical Specification requirement of MTC < 0 pcm/*F. Table 3.3.4-1 summarizes the results of the isothermal temperature coefficient measurements.

.O 1

1 O

O O

(0856M)

O TABLE 3.3.4-1 Measured vs Predicted Isothermal Temperature Coefficient O

Rod ITC ITC Configuration Expected Measured MTC O ARO -2.87 3.0 pcm/*F -2.15 -0.30 l CBD In -4.29 3.0 pcm/*F -3.55 -1.70 CBD&C In -8.367 1 3.0 pcm/*F -7.60 -5.75 ,

O O

O O

O O

O ,

O (0856M)

l O

3.3.5 - POWER COEFFICIENT DETERMINATION (30%, 50%, 75%, 90%), 2.47.31 OBJECTIVE The objective of this procedure was to verify the design predictions of C) the Doppler only power coefficient during the power ascension startup testing sequences at 30%, 50%, 75% and 90% power.

, TEST METHODOLOGY l

l The Doppler power coefficient verification was performed by varying plant C) load plant load approximately 47 MWe (26 MWe at 30% power) utilizing the turbine controls on the main turbine DEHC panel. Three successive load decreases / increases were performed, allowing Tave and AT to stabilize prior to each load swing. No changes in rod position or RCS boron concentration were made during the load changes.

• i

.O

SUMMARY

OF RESULTS After the load swings were performed, values for Tave and AT were taken from the strip chart recorders. Using this data, and the power readings obtained prior to and after the load swings, the average measured Doppler coefficient verification factor (Cm ) was calculated, where C m is C) effectively the measured ratio of change in Tave vs change in reactor power. This value of C mwas then compared with the predicted Doppler coefficient verification factor (Cp ). Cp was calculated using the Doppler only power coefficient the isothermal temperature coefficient and the difference in the measured isothermal temperature coefficient versus b) the predicted isothermal temperature coefficient at ARO, HZP. The values of Cmand Cpwere compared to determine if the acceptance criteria of not greater than a 0.5'F/% power difference were met. See Table 3.3.5-1 for the actual values.

O O

O O

[] (0856M)

TABLE 3.3.5-1 Measured vs Predicted Doppler Only Power Coefficient Power Measured Predicted Difference O

30% 2.33 *F/% Power 2.58 'F/% Power 0.25 *F/% Power 50% 1.66 1.84 0.18 75% 1.24 1.30 0.06 O

90% 1.08 1.00 0.08 1

.O O

O O

O O

O (0856M)

O 3.3.6 - INITIAL CRITICALITY, 2.52.32 OBJECTIVE This test procedure provided a method by which initial criticality could

'() be obtained in a cautious and controlled manner. The sequence, frequency, and core conditions for collecting nuclear data was provided as well as the method of analysis of this data. In addition, a manual reactor trip was demonstrated prior to initial criticality in accordance with the test procedure.

l O TEST METHODOLOGY Initial conditions were established with the RCS at an average temperature of 556.6*F, RCS pressure at 2256 psig, RCS boron concentration greater than 2000 ppm, and all RCC banks fully inserted.

() Programs were initiated to monitor neutron flux, reactivity and various

! other plant parameters for the duration of the ' test.

Reference counts were determined for each source range channel by collecting 10 sets of reference count data over 100 second intervals.

These values were to be used in the ICRR (Inverse Count Rate Ratio)

C) calculations after reactivity additions were made.

Shutdown banks were then withdrawn in alphabetical order. The withdrawals were made in increments of 50 steps or less and the value of the ICRR was determined prior to continuing.

() Control banks were then manually withdrawn in an overlap configuration.

Rod withdrawal was completed when CBD was positioned at 160 steps. A manual reactor trip was required to be performed to verify the control j rod drive mechanisms would unlatch upon opening of the trip breakers. )

l Subsequent withdrawals of control and shutdown banks were done in 228

$) step increments with the exception of CBD which was only pulled to 160 steps. At no time did the ICRR value calculated after any bank withdrawal vary by more than i 5% from the corresponding ICRR value obtained during the previous withdrawal prior to the manual trip.

The remaining reactivity insertion required to achieve criticality was

) made by diluting the RCS boron concentration by additions of primary I water to the RCS. The dilution rate was initially 60 gpm until the ICRR

, value fell below 0.2 and was then reduced to 30 gpm. The ICRR was renormalized at this point as well. This dilution was maintained until the renormalized ICRR value fell below 0.3 at which time the RCS dilution was terminated to allow for mixing. Criticalit was achieved during this

() time period. Flux level was established at 10- 5 X 10-9 amps using CBD.

'O -100-(0856M)

O 3.3.6 - INITIAL CRITICALITY, 2.52.32 (Continu-d)

SUMMARY

OF RESULTS Acceptance Criteria was met in that the reactor was successfully tripped, ,

O- criticality was achieved within the approximate range of the predicted i boron concentration, and neutron flux level was established within specified bounds on the Intermediate Range NIS channels.

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l 0 -101-(0856M)

O O O O O O O O O O O FIGURE 3.3.6-1 ICRR vs SHUTDOWN BANK POSITION 1.5 -

1.4 1.3 1.2 1.1

" * ** < I v V VV 1 0

= " .+

• t m x & A p m R + x + x' 4 x  ; +t 0.9 0.8 m

m 9 0.7 0.6 0.5 O.4

\

0.3 O.2 O.1 0.0 l SBB SBC SBD SBE ,

SBA STEPS WITHDRAWN

+ SR N31 x SR N32

-102-

O O O O O O O O O O O FIGURE 3.3.6-2 l ICRR vs CONTROL BANK POSITION 1.5 1.4 1.3 1.2 1.1 1.0 x x ,

f 0.9 ji Y +

x g 0.8 m g 4 y y y 9 0.7 0.6 0.5 0.4 0.3 0.2 0.1

! CBA CBC 258 2ba

CBB CBD  !

2h8 ggg STEPS WITHDRAWN

+ SR N31 x SR N32

-103-

' ~

O d 'O O ~d O O O O O O FIGURE 3.3.6-3 ICRR vs PRIMARY WATER ADDED DURING APPROACH TO CRITICALITY 1.5 1.4 1.3 1.2 1.1 1.0 giiaaaaaaac.= asaaaaats t.aaaaaaaaaa aamaaaaaaa aaa O.9 OE +--

O.8 80( 3b m

E oogE An#

9 *

+0g 0.6 *b is '

m O.5 s;;;mg g sim-O.4 ""6 0.3 -6 g 0.2 8",

0.1

0.0 0 20 40 (Thousands)

PRIMARY WATER ADDED [qallons]

o SR N31 + SR N32 a IR b35 v IR N36 I

i -104-l _- ___ __ -_ ___ _ ______ _ ___________ ___ __ __ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _____

FIGURE 3.3.6-4 l

Source Range Detector Response During Approach to Criticality m00s2 =003 RELATIVE TIME 000.0 768.7 0.00 5.00 10.00 15.00 20.00 25.00 30.00 (MINS)0 36.0 40.00 45.00 50.00 55.00 60.00

[

720.0 601.0 540.0 606.3 560.0 829.7 480.0 454.0 l 400.0 370.3 320.0 302.7 a

240.0 227.0 160.0 158.3 80.00 75.67 I _ _ _ _

/

0.0000 0.0000 '  ;

POINT ID DESCRIPTION UNITS SYN 80L l 8YRON UNIT 1 l N0031 SOURCE RNG CH-31 LOG G DKCS O N0032 SOURCE RNG CH-32 LOG G DKCS A START TIME 02FE885 23:00:00

-105-

FIGURE 3.3.6-5 Intermediate Range Detector Response During Approach to Criticality m0030 wo035 flELATIVE TIME 0.n 5.n so.n is.n 20.00 25.n 30.00 (NINSj0 as. 40.00 45.n so.00 ss.n so.n 8.as00E-Os 0.s500E-Os 0.8357E-Of 0.1357E-05 0.1204E-01 0.1206E-01

9. 30sM-0 8 0,t055E-0

8.0040E-02 0. 90 46E-02

^ _ - < - - -. p 0.7530E-02 0.7530E-02 O.0030E-02 0.0030E-02 0.452X-02 0.4523E-02 I

0.30s u-02 0.30 u-02 0.1500E-02 0.1500E-02 0.0000 0.0000 POINT ID DESCRIPTION UNITS SYMBOL l8YRON UNIT 1 l N0035 INTERMEDIATE RNG CH-35 LOG G MCAMP O N0036 INTERMEDIATE RNG CH-36 LOG G HCAMP A 4

START TIME 02FEB85 23:00:00

-106-

3.3.7 - DETERMINATION OF CORE POWER RANGE FOR LOW POWER PHYSICS TESTING, 2.52.33 OBJECTIVE g This procedure was used to determine the power level (neutron flux level) at which detectable reactivity feedback effects from nuclear fuel heating occurred and to establish the range of neutron flux in which zero power reactivity measurements were performed.

TEST METHODOLOGY O

Initial conditions were established with the RCS at an average temperature of 556.5*F, RCS pressure at.2230 psig and the RCS boron concentration at 1088 ppm. The reactor was critical with the flux at 1.0 X 10-8 amps on both Intermediate Range channels.

O Initially, the reactivity computer was verified to be set up on power range channel N-44 which was taken out of service. Reactivity computer outputs of reactivity and flux were displayed on a two-pen strip chart recorder. A second strip chart recorder displayed RCS Tave and neutron flux. The Tave input on the second strip chart was from the 7300 process instrumentation racks.

O ,

A reactor period measurement was then made on the reactivity computer and compared to the value from the Nuclear Design Report (WCAP 10315).

Next, the determination of the power range for low power physics testing was made. Control bank D was withdrawn to achieve a positive reactivity O addition of 30 10 pcm. Reactivity and flux level were then observed to determine the point of adding nuclear heat as indicated by negative reactivity addition from the Doppler fuel temperature coefficient.

Another indication of nuclear heating that was monitored was RCS Tave which was expected to increase as a ra'uit of the increased temperature of the fuel.

O

SUMMARY

OF RESULTS The reactor period measurement was found to be within 3.3% of the design value, which was within the 14*4 tolerance.

.O The reactor power level at which detectable reactivity feedback effects from nuclear heating occurred was determined to be 1 X 10-6 amps on the Intermediate Range (IR) NIS channels. This was within the expected range of 5 X 10-7 to 5 X 10-6 amps.

The neutron flux level at which zero power reactivity measurements were O to be performed was determined to be 1.0 X 10-8 to 1.0 x 10-7 amps on the IR channels and 1.3 X 10-8 to 1.3 X 10-7 acips on the reactivity comguter. These were both within the expected range of 5 X 10-9 to 5 X 10- amps.

O O (0856M)

3.3.8 - REACTIVITY COMPUTER CHECKOUT, 2.52.34 4 OBJECTIVE I

The Reactivity Computer Checkout test procedure was performed to d*" "****** Proper operation of the Westinghouse reactivity computer

-O Model 7746A through a dynamic test using neutron flux signals.

. TEST METHODOLOGY A twenty five pcm reactivity increase as shown ort the reactivity computer )

strip chart recorder was. initiated by withdrawing control bank D. A O  !

! stopwatch was used to determine the reactor doubling time by measuring j the time required for the flux to increase from a starting value to a j value twice as large. This doubling time was used to determine the theoretical reactivity increase. The theoretical reactivity increase was )

compared to the reactivity indica'ted o'n the reactivity computer strip O chart recorder. This was repeated for reactivity increases of i approximately fifty pcm and seventy five pcm.

i i

SUMMARY

OF RESULTS )

l The acceptable average absolute deviation between indicated and i

~O theoretical reactivity was less than 4%. The absolute deviations between I indicated and theoretical reactivity were 1.6%, 1.2% and 0.1% for I approximate reactivity gains of 25 pcm, 50 pcm and 75 pcm respectively. l The average absolute deviation between indicated and theoretical reactivity was 1.0%.

O l l

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l O 3.3.9n - BANK WORTH MEASUREMENT AT ZERO POWER. 2.64.30A OBJECTIVE l The Bank Worth Measurement at Zero Power test procedure was performed to 5

determine the differential and integral worth of individual banks and to

0 determine the differential boron worth for the control banks for

! comparison to the design predictions;made in WCAP 10315, "The Nuclear Design and Core Physics Characteristics of the Byron Unit 1 Nuclear Power

! Plant Cycle 1".

TEST METHODOLOGY O

l In order to perform the bank worth measurement either a reactor coolant i system dilution or boration was initiated and the resulting reactivity

) change compensated for by inserting or withdrawing the bank to be I measured. During the course of the' boron concentration change, flux and

'O reactivity signals were recorded using the Westinghouse reactivity computer. From the reactivity trace a total reactivity change for the 3

movement of the bank was determined. This value was then combined with 4

data from the appropriate boron endpoint measurement to determine the total bank worth.

I Q In order to determine the differential boron worth over the control i banks, a least squares fit was applied to the plot of totalized integral i bank worth versus endpoint boron concentration. The slope from the

. equation for the line, determined by the least squares fit was the

! differential boron worth.

,0

SUMMARY

OF RESULTS 1

! For the bank worth measurements the acceptable variation from the design

! predictions was 110%. For each of the measurements the bank worth was i found to be within this limit. During the control bank D and C 4 measurements small xenon transients were in progress resulting from the 1

0 increased reactor power needed to obtain the ARO and D-in flux maps. The reactivity addition rate from the xenon removal was within the

Westinghouse reconunended limit for xenon stability based on reactivity i changes observed prior to initiation of the dilutions.

! The average of the absolute values of percent error was 2.5%. The best

,0 agreement was found in the measurement of shutdown bank D which was within 0.5% of its predicted value. The worst agreement was found in the j measurement of shutdown bank E which was within 7.9% of its predicted i value. A summary of the bank worth measurements and the predicted values 1 appears in Table 3.3.9-1 and plots of integral and differential bank i

worths are found in Figures 3.3.9-1 through 3.3.9-4.

O For the determination of differential boron worth over the control banks

! the acceptable variation from the design prediction was 110%. ~The determined value of -11.63 pcm/ ppm is within 10% of the predicted value of -11.51 pcm/ ppm (See Figure 3.3.9-5).

o l

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-109-(0856M)

.- . =.

3.3.9b - BANK WORTH WITH OVERLAP MEASUREMENT AT ZERO POWER, 2.64.30B OBJECTIVE j The Bank Worth With Overlap Measurement test procedure was performed to determine the integral and differential worths of the control banks in an

'}) overlap configuration and to compare these values with Westinghouse design predictions.

TEST METHODOLOGY

() The bank worth measurements were made in this procedure by initiating a reactor coolant system boration and compensating for this reduction in reactivity by manual withdrawal of the control banks in normal overlap.

Flux and reactivity were monitored on the Westinghouse reactivity

, computer and were used to determine when sufficient reactivity compensation had been made. The reactivity computer also facilitated f3 measurement of the reactivity change made by each increment of rod i motion. Boration was continued until control bank D was positioned at

116.5 steps withdrawn after which no further changes in reactivity were made. The remaining data for control bank D was obtained from the executed test 2.64.30A, " Bank Worth Measurement at Zero Power" for control bank D.

O Integral worth was then calculated by adding up the increments of i

^

reactivity over the entire range of b.?.nk travel. Differential worth was calculated by dividing the reactivity increments by the number of control

> bank steps the increment represented. Both types of bank worths were j plotted to illustrate their position dependence and the core properties

() they demonstrate (See Figure 3.3.9-6).

SUMMARY

OF RESULTS )

i The integral reactivity worth of the control banks with overlap was

! expected to be 3660 pcm t 10%. The calculated value was 3597 pcm, which

.C) was within 2% of the predicted value.

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() (0856M)

O' TABLE 3.3.9-1 Measured Versus Predicted Bank Worths MEASURED PREDICTED WORTH [pcm] WORTH [pcm]

O SBA *721.8 1700 t 170 SBB 562.0 570 1 57 SBC 1215.0 1180 118 SBD 801.3 800 t 80 ,

O SBE 571.6 530 1 53 CBA 615.2 600 ! 60 CBB 1275.6 1280 1 128 CBC 1109.3 1140 + 114 CBD 847.8 648 ! 65 O RCCA D-12 Ejected 487.0 780 maximum

• Due to the measurement technique, offsite engineering determined that the original predicted worth was incorrect and supplied a revised prediction of approximately 700 pcm.

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g ._ g_ .. _

g ._

g

.g 3 g g g g 1

FIGURE 3.3.9-1 4

INTEGRAL WORTH vs STEPS WITHDRAWN CONTROL BANK D 1.4 1.3 1.2 1.1 3o 1.0 em 0.9 l 0.8 x% f 00 0.7

• s -

~

8.0 0.6 '

$D Q o 0.5 s

z

~ O.4 (

0.3 l

0.2 i

O.1  %, -

O.O O 40 80 120 160 200 240

STEPS WITHDRAWN j -112-

.g g. -g. -g g- -g g

g g -o - g-FIGURE 3.3.9-2 INTEGRAL WORTH vs STEPS WITHDRAWN CONTROL BANK C 1.4 1.3 1.2 11 s

E 10 (

0 4 0.9  ;

I7 V 0.8 o

Om 0.7 o

i - Jr O.6 l D i O O.5 u Z

~ O.4 i

0.3 k

0.2

\ -

O.1 u

m 0.0 O 40 80 120 160 200 240 3

STEPS WITHDRAWN l -113-

.g----- g g

g g

.--g .g - .. -.

g g_ -g.. .g i

FIGURE 3. 3. 9 -3 INTEGRAL WORTH vs STEPS WITHDRAWN CONTROL BANK B 1.4 2

x

~

1.1 h

T 1.0 O.9 I9 m 0.8 k i Oocs 0.7

, d;$c O.6 q '

mm 0 0.5 (

d z O.4 ,

0.3 3

l 0.2 Ag 0.1

%m

~

0.0 ^O O 40 80 120 160 200 240 l STEPS WITHDRAWN i

-114-1

- _ _ .g.. .- g . -.

-g . _ - . -- -- - _-_- .. -_ - -

4 i

i FIGUR E 3.3.9-4 INTEGRAL WORTH vs STEPS WITHDRAWN CONTROL BANK A 1.4 1.3

1.2

1.1 I ^

E 1.0 O

& O.9 m

Im 0.8

>- C EO oa O.7

y3 2 2

,! a.S O.6 20,-

l

=t o 0.5 sa_

i w b-3 k z O.4

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O*3

'a i

! O.2 %g .

l O.0 l O 40 80 120 160 200 240

! STEPS WITHDRAWN I

-115-

O O O O O O O O, O O O FI G UR E 3.3.9-5 DIFFERENTIAL AND INTEGRAL WORTH OF CONTROL BANKS IN OVERLAP

~. a

~.. __ '

9

s. I un ~

F

_N.

7 t 'N E N 2""-

'N R A aA G "~

1 N

aT R 'N. I A \<' A L '~~ . L w ._.

h 'N s -

w

.~ 's  %

H / '\

h H f m ,

,~~.. A '

P

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! 'N \

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S

, - - . _ - ,2 T

.., .. 2. E

, , >=,,. , , P w.. . "w.-e ' '

G

• 23,. JM w= a w.m .

CONTROL BANK POSITIONS

-116-

FIGURE 3.3.9-6 DIFFERENTIAL BORON WORTH OVER THE CONTROL BANKS 4.5 4

4.0 E gx

.l E 3.5 N

% 3.0 \

, 84 4

mg

_.s a 2.5 t

5o 0C N 2.0 '

E 9 1.5 l 2

, a

v1 1.0 N

O.5 700 800 900' 1000 1100 1200 BORON ENDPOINT CONCENTRATION [ ppm )

o -11.63 pcm/ ppm

^

-117-1

F

'O

~

3.3.10 - BORON ENDPOINT DETERMINATION, 2.64.33 l j OBJECTIVE i~

This test determined the rod worths at the extreme ends of bank travel (at the near fully withdrawn or near fully inserted position) for use in

'O the Bank Worth Measurement Test, 2.64.30A. In addition, the just critical reactor coolant system (RCS) boron concentration was determined at various RCC bank configurations along with the all rods out critical I boron concentration and the boron reactivity worth for each control bank.

lO TEST METHODOLOGY l

With RCS loop and pressurizer boron concentration having been sampled at

least three times at approximately thirty minute intervals and within 10 t ppm from the average, RCC banks were positioned to within 50 pcm of the desire'd endpoint configuration with the reactor critical in the zero O Power physics testing range. The controlling bank was then inserted or withdrawn to reach the desired endpoint configuration while neutron flux, i reactivity, RCS Tave and pressurizer level were monitored with the strip j chart recorders. When the neutron flux level had increased to greater than 30 percent of full scale and reactivity was constant, the controlling bank was repositioned to re-establish the initial flux l

.O level. This process was then repeated twice for each RCC configuration, j

SUMMARY

OF RESULTS ,

)

The rod worths at the extreme ends of travel were determined and added to j the worths measured in the bank worth measurement test procedure. The )

=O all rods out just critical RCS boron concentration was found to be 1096 l ppm which met the acceptance criterion of 1046 50 ppm. Boron

! reactivity worths are shown in Table 3.3.10-1. The boron reactivity

worth of control bank D was found to be outside the Westinghouse design 1 predictions. This data was reviewed by offsite engineering and i Westinghouse and the determination was made that it did not impact safety

,0 analyses. Based on these reviews, the test was approved.

1

O l

i 4

O

,'O 4

'g -118-i (0856M) i i

TABLE 3.3.10-1 MEASURED VERSUS PREDICTED BORON ENDPOINT CONCENTRATIONS AND BORON REACTIVITY WORTHS 7^

ACTUAL BEP PREDICTED BEP BORON REACTIVITY CONCENTRATION CONCENTRATION WORTH (ppm]

[ ppm] [ ppm]

CBA OUT 817 782 1 78 54 J

CBB OUT 921 892 89 104 CBC OUT 1023 992 1 99 102 CBD OUT' 1096 1046 50 73 D

D 0

D D

0 0

O -119-(0856M)

f O 3.3.11 - PSEUDO ROD DROP, 2.64.36 l l

OBJECTIVE 1

Under an asymmetric power distribution, the DNBR was verified to be l O within the Technical Specification limits. RCCA H-04 was inserted into the core where incore flux map, incore thermocouple and excore neutron  ;

detector data was collected and the response of the excore neutron '

detector was determined for the asymmetric power distribution.

TEST METHODOLOGY O

With a stable power of 50%, a full core flux map was taken. Rods D-08, M-08 and H-12 were disconnected from shutdown bank E in the fully withdrawn position. By reactor coolant dilution, H-04 was inserted into the core. At the following step intervals, partial core flux maps, excore neutron detector response and power level were recorded: 220, O 216, 203, 178, 153, 128, 103, 78, 53, 28 and 0 steps. With rod H-04 fully inserted, a full core flux map was taken and by boron concentration adjustment rod H-04 was withdrawn and all rods were reconnected. The nuclear enthalpy rise hot channel factors needed to assure the limiting DNBR was not reached were obtained from the output of the Incore computer run.

O

SUMMARY

OF RESULTS ,

This test was in progress when detector A became stuck in the core while performing a partial core flux map. After several strategies had been i attempted to make necessary repairs, rod H-04 was removed and the test O exited. Upon reperformance of the test, the nuclear enthalpy rise hot channel factor FDHN, was found to be 1.616, where the Technical i Specification limit was 1.7843 (= 1.55 X (1.0 + 0.3 (1-% Power)] ). The heat flux hot channel factor Fg(Z), was found to be 2.2156 at axial location 31. The Technical Specification limit for Fg(Z) at axial point 31 was 4.640. All acceptance criteria were met during the second O execution of the test.

O 1

I O

O (0856M)

3.3.12 - MISCELLANEOUS PHYSICS TEST RESULTS

OBJECTIVES i,

f Additional testing covered in the Initial Criticality and Low Power Test g Sequence included measuring the worth of RCCA F-10, the predicted most reactive stuck rod, and measuring the worth of shutdown banks A and B.

i

, METHODOLOGY

$ To determine the worth of RCCA F-10, a rod swap was done with shutdown

'O bank B. RCCA F-10 was withdrawn while shutdown bank B was inserted to

! maintain just critics 1 conditions. When RCCA F-10 was fully withdrawn, a

- dilution was initiated to measure the remaining worth of shutdown bank B

! and the worth of shutdown bank A.

i j SOMMARY OF RESULTS

.O j A tabulation of the measurement results, along with associated acceptance j criteria values, appears below.

Bank or Rod Measured Acceptance Criteria Measured Value O RCCA F-10 1180 t 118 pcm 389.5 pcm i Shutdown Bank A 1700 1 170 pcm 721.75 pcm j Shutdown Bank B 570 1 57 pcm 562 pcm i

RCCA F-10 and shutdown bank A failed to meet their acceptance criteria.

. Based on an offsite engineering review, it was determined that the O approach taken by the station, with respect to data evaluation of RCCA F-10 worth did not result in values that provided a valid comparison to

the acceptance criteria. After looking at boron data obtained in the test procedure and accounting for the remaining worth of shutdown bank A, j~

it was concluded that the RCCA F-10 boron endpoint was 484 ppm, comparing favorably to a predicted endpoint of 412 ppm. While this falls outside O the i 10% tolerance, it demonstrates that the worth of the predicted most j

reactive stuck rod, RCCA F-10, is conservative, and N-1 shutdown margin requirements can be met throughout the cycle life.

I A similar evaluation of data supporting the determination of shutdown bank A worth, made by offsite engineering stated that the station's use

C of a 1700 pcm value for acceptance criteria was incorrect based on the i fact that bank measurement was made with RCCA F-10 fully withdrawn. With l RCCA F-10 fully withdrawn, spatial effects in the neutron' flux were i introduced.

I After taking these spatial affects into consideration, a predicted value O of 700 pcm for shutdown bank A was used for comparison to the measured value of 678 pcm. Based on these evaluations, the results were accepted by offsite engineering.

i L

g -121-

(0856M)

-w- -v- ,,--.ww..- -, ,- ,,..-g, , e, y- ,n. ._.-..m -m.. --,-.,.m,_ n--- - . ,-,,n, .. ,_,.m. .y, g 7. .y. , r- - . , n. m. _,

-- - ~. .. - - - -- -- .-. . .- - .

)

4 I

O

! 3.4 - TRANSIENT TESTING 3.4.1 - LOSS OF OFFSITE POWER, 2.05.30 j OBJECTIVE l

O i j The Loss of Offsite Power test was performed to verify the plant's

ability to safely sustain a loss of offsite power condition for a minimum

of thirty minutes.  ;

4

TEST METHODOLOGY O

! In order to establish the units ability to safely sustain a loss of l i

j offsite power the following RCS and steam generator parameters were i recorded: RCS loop wide range temperatures, RCS Tave, pressurizer (PZR) j pressure FZR level, steam generator (SG) level, and SG pressure. The emergency ciesel generators were monitored along with the engineered

,0 safeguards features (ESP) loads to verify proper diesel starting, ESF bus i load shedding, and proper sequence starting of the required ESF loads.

I The operating personnel were provided guidance for dealing with the 30 minute transient and the event was initiated by isolating the Unit.1 1 power transformers from the offsite power grid.

i O

SUMMARY

OF RESULTS ,

j The test procedure was successfully executed on March'14, 1985. The test i procedure specified six acceptance criteria parameters to evaluate plant

{ performance during the transient. Four of the parameters dealt with plant component operations and are as follows: All required loads on the

O ESF buses were shed prior to the diesel generators energizing the ESF I buses. The following ESF loads were properly sequenced onto the ESF buses

both trains of centrifugal charging pumps, both trains of f component cooling pumps, both trains of essential service water pumps, i

! both trains of control room chillers, and the motor-driven auxiliary feed

! water pump. The diesel-driven auxiliary feedwater pump automatically 0 started following the initiation of the loss of offsite power, and the j fourth criterion was met by having at least one safe shutdown train l 1 operable on emergency power during the 30 minute transient. The i a

remaining two acceptance criteria dealt with the voltage and frequency parameters of the diesel generators and the minimum / maximum values for the RCS parameters mentioned earlier. The diesel generators started and

O energized the ESF buses consistent with the Technical Specification j requirements of 4160 t 420 volts and 60 t 1.2 HZ withir.10 seconds after receiving the undervoltage start signal, j The acceptance criteria for the RCS parameters were as follows

RCS Tave g 550*-590*F, PZR pressure 1840-2315 psig, PZR level less than 90%, SG 1evel greater than 42% and SG pressure less-than 1160 psig. The acceptance criteria for PZR level, SG 1evel, and SG pressure were met i during the transient, PZR pressure reached 2340 psig due to residual j heating effects after pressurizer heaters were energized during the i transient as expected. The calculated minimum RCS Tave went below 550*F

! d"* t 1 "er than anticiP ated levels of decay heat. Sufficient data was O

i available to establish RCS temperature trends which satisfied concerns j regarding minimum temperature for criticality and negative moderator i temperature reactivity coefficient considerations. This satisfied the intent of the acceptance criteria.

'O -122-

- (0856M) i f -

O 3.4 - TRANSIENT TESTING 3.4.1 - LOSS OF OFFSITE POWER, 2.05.30 (Continued)

There were two actions taken by personnel which were not specified in the O test procedure. An emergency boration was manually initiated during the

, transient per the normal conduct of the Reactor Trip Recovery procedura because rod position could not be verified at bottom. Rod position could not be verified because digital rod position indication (DRPI) is powered from a non-ESF bus. Emergency boration was not required to control reactivity because the NIS source range instruments indicated that the O reactor was tripped and the rods were inserted: however the recovery procedure specified borating. DRPI was energized later in the transient and the emergency boration was suspended. The second action was to open two SG PORVs to reduce PZR pressure. This action came in response to the

, PZR pressure reaching 2348 psig as mentioned earlier.

O The operations personnel responded to the transient in accordance with approved Station procedures and training, however, this caused the plant to respond in a manner not predicted by the test procedure. The station procedures were written to respond to an actual plant transient and were not written for test conditions which require data acquisition.

O This caused minor discrepancies between actual plant response and the response anticipated by the test procedure. The plant response and test results were reviewed in light of the actions taken by operating personnel. It was determined that the plant responded per design. The test results were found acceptable and the test verified that the unit could safely sustain a 30 minute loss of offsite power transient.

Following the conclusion of the 30 minute test, there were problems encountered during plant recovery which stemmed from the main steam isolation valves (MSIVs). The unit entered an outage condition, repaired the problems with the MSIVs and reported the problems with the MSIVs in accordance with 10CFR Part 21 requirements.

O O

O O -123-(0856M)

O 3.4.2 - LOAD SWING TEST (35%, 75% AND 100%), 2.52.37 OBJECTIVE This test was performed to demonstrate the dynamic response of the g) Reactor Coolant System (RCS) and the Rod Control System to automatically bring the plant to steady state conditions following a 10% reduction in turbine load, followed by a 10% increase in turbine load.

TEST METHODOLOGY e)

With plant conditions stable at the appropriate power level (35%, 75% or 100%) a 10% step load decrease was programmed into the turbine's Digital Electro-Hydraulic Control (DEHC) 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 test, strip chart recordings of essential plant parameters were taken so that their response could be analyzed. The parameters monitored included RCS Tave, Tref, pressurizer pressure and level, steam generator pressure and levels, steam and feedwater flows, rod position and rod speed, OTAT and OpaT, reactor power, feedwater pump speed and discharge pressure and feedwater

() regulating valve demand position.

SUMMARY

OF RESULTS The load decreases and increases did not cause either the reactor to trip or 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.

During the load increase at the 35% execution, manual intervention was required after 12 minutes in order to terminate a sustained oscillation in the loop 1D fsedwater regulating valve 1FW-540. The steam generator level controller for loop 1D was tuned and verification of its proper C) dynamic response was accomplished at the 75% execution of the test. At the 100% execution boric acid was added during the load decrease and primary water was added during the load increase in order to maintain AI within the target band (this was not acceptance criteria and had no effect on the acceptance of the test). From the accelerometer data, it was determined that no damaging water hammer occurred during any of the C) load swings at any of the power levels. During the load reduction, Tave was observed to increase initially and during the load increase, Tave was observed to initially decrease (contrary to the expected). Subsequent review determined that the expected trends were incorrect and the observed data was acceptable. The test was accepted based on the data obtained.

O O

') (0856M)

-124-

TABLE 3.4.2-1 10% LOAD DECREASE AT 35% POWER SUWARY INITIAL FINAL O CONDITION CONDITION Generator Load (MWe) 383 260 Nuclear Power (%) 36 26.5 O Tave Loop 1A (*F) 566 563 Tave Loop 1B (*F) 562 560 Tave Loop 1C (*F) 565 563 Tave Loop 1D (*F) 565 562 Tref ('F) 568 563 O

Delta T Loop 1A (%) 32 24 OPAT Loop 1A (%) 106 107 OTAT Loop 1A (%) >150 >150 0

Pressurizer Pressure (psig) 2253 2252 Pressurl=er Level (%) 37 34 Steam Generator Level Loop 1A (%) 66.5 66.5 O Steam Generator Level Loop 1B (t) 63 65 Steam Generator Level Loop 1C (%) 63 66 Steam Generator Level Loop 1D (%) 62 65 Steam Header Pressure (psig) 1090 1150 O Steam Flow Loop 1A (pph) 1.20E6 0.65E6 Steam Flow Loop 1B (pph) 1.30E6 0.92E6 Steam Flow Loop 1C (pph) 1.10E6 0.80E6 Steam Flow Loop ID (pph) 1.15E6 0.78E6 Feedwater Flow Loop 1A (pph) 1.20E6 0.88E6 O Feedwater Flow Loop 1B (pph) 1.20E6 0.90E6 Feedwater Flow Loop 1C (pph) 1.10E6 0.79E6 Feedwater Flow Loop ID (pph) 1.15E6 0.96E6 Feedwater Temperature Loop 1A (*F) 354 344.5 Feedwater Temperature Loop 1B ('F) 354 344.3 O

Feedwater Temperature Loop 1C (*F) 354 344.3 Feedwater Temperature Loop 1D ('F) 354 344.3 Feed Pump Discharge Pressure (psig) 1118 1168 l O Control Bank D Position (steps) 174 158 Control Bank C Position (steps) 228 228 Boron Concentration (ppm) 785 785 (0856M)

O TABLE 3.4.2-2 10% LOAD INCREASE AT 35% POWER

SUMMARY

INITIAL FINAL O CowDITIon ConDITrow Generator Load (MWe) 230 340 Nuclear Power (%) 24 34 g Tave Loop 1A (*F) 564 566 Tave Loop 1B (*F) 561 566.5 Tave Loop 1C (*F) 564 564 Tave Loop ID (*F) 564.5 566.5 Tref (*F) 577 582 O

Delta T Loop 1A (%) 22 30 OPAT (%) 108 108 oTAT (%) >150 >150

'O Pressurizer Pressure (psig) 2250 2252 Pressurizer Level (%) 34 34 l

l Steam Generator Level Loop 1A (%) 67 68 O , Steam Generator Level Loop 1B (%) 66 66 Steam Generator Level Loop IC (%) 66 66 Steam Generator Level Loop 1D (%) 64 66 Steam Header Pressure (psig) 1143 1125 O Steam Flow Loop 1A (pph) 0.75E6 1.15E6 Steam Flow Loop 1B (pph) 0.80E6 1.20E6 Steam Flow Loop 1C (pph) 0.70E6 1.05E6 Steam Flow Loop 1D (pph) 0.70E6. 1.05E6 Feedwater Flow Loop 1A (pph) 0.85E6 1.12E6 O reedwater riow Loop 1B (pph) 0.80E6 1.10E6 Feedwater Flow Loop 1C (pph) 0.70E6 1.05E6 Feedwater Flow Loop 1D (pph) 0.90E6 1.20E6 Feedwater Temperature Loop 1A (*F) 341.6 351 i Feedwater Temperature Loop 1B (*F) 341.5 351.3 O reedwater Temperature Loop 1C (*r> 341.4 351 1

Feedwater Temperature Loop 1D (*F) 341.4 351.2 Feed Pump Discharge Pressure (psig) 1186 -

Control Bank D Position (steps) 130 142 O Control Bank C Position (steps) 228 228 Boron Concentration (ppm) 785 764 (0856M)

O TABtz 3.4.2-3 10% LOAD DECREASE AT 75% POWER

SUMMARY

INITIAL FINAL O CONDITION CONDITION Generator Load (MWe) 820 710 Nuclear Power (%) 73 61 O Tave L P 1A (*F) 583 579 Tave Loop 1B (*F) 584 580 Tave Loop 1C (*F) 582 579 Tave Loop 1D (*F) 580 577 Tref (*F) 581 577 O

Delta T (%) 74 74 OPAT Loop 1A (%) 107 107 OTAT Loop 1A (%) 135 143 O

Pressurizer Pressure (psig) 2230 2230 Pressurizer Level (%) 55 49 Steam Generator Level Loop 1A (%) 66 66 O Steam Generator Level Loop 1B (*) 65 66 Steam Generator Level Loop IC (%) 66 66 Steam Generator Level Loop ID (%) 66 66 Steam Header Pressure (psig) 995 1000 O Steam Flow Loop 1A (pph) 2.75E6 2.40E6 Steam Flow Loop 1B (pph) 2.70E6 2.35E6 Steam Flow Loop 1C (pph) 2.65E6 2.25E6 Steam Flow Loop 1D (pph) 2.65E6 2.30E6 Feedwater Flow Loop 1A (pph) 2.65E6 2.50E6 O Feedwater Flow Loop 1B (pph) 2.65E6 2.50E6 Feedwater Flow Loop 1C (pph) 2.70E6 2.55E6 Feedwater Flow Loop 1D (pph) 2.80E6 2.60E6 Feedwater Temperature Loop 1A (*F) 411 400 Feedwater Temperature Loop 1B (*F) 411 400 0 Feedwater Temperature Loop 1C (*F) 411 400 Feedwater Temperature Loop 1D (*F) 411 400 Feed Pump Discharge Pressure (psig) 1170 1160 O

Control Bank D Position (steps) 181 141 Control Bank C Position (steps) 228 228 Boron Concentration (ppm) 754 754 O (0856M)

O TABLE 3.4.2-4 10% LOAD INCREASE AT 75% POWER

SUMMARY

INITIAL FINAL CONDITION CONDITION O

Generator Load (MWe) 710 820 Nuclear Power (%) 61 73 O Tave Loop 1A (*F) 576 583 Tave Loop 1B (*F) 578 584 Tave Loop 1C (*F) 576 582 Tave Loop 1D (*F) 575 580 Tref (*F) 577 581 O

Delta T Loop 1A (%) 74 74 OPAT (%) 107 107 OTAT (%) 145 135

'O Pressurizer Pressure (psig) 2230 2230 Pressurizer Level (%) 45 52 Steam Generator Level Loop 1A (%) 65 66 O Steam Generator Level Loop 1B (%) 66 66 Steam Generator Level Loop 1C (%) 66 66 Steam Generator Level Loop 1D (%) 66 66 Steam Header Pressure (psig) 985 995 O Steam Flow Loop 1A (pph) 2.40E6 2.75E6 i Steam Flow Loop 1B (pph) 2.35E6 2.70E6 Steam Flow Loop 1C (pph) 2.25E6 2.65E6 Steam Flow Loop 1D (pph) 2.30E6 2.65E6 Feedwater Flow Loop 1A (pph) 2.50E6 2.60E6 O Feedwater Flow Loop 1B (pph) 2.50E6 2.65E6 Feedwater Flow Loop 1C (pph) 2.55E6 2.70E6 Feedwater Flow Loop 1D (pph) 2.60E6 2.80E6 Feedwater Temperature Loop 1A (*F) 400 411

1. • • d"***# ***P*#**"#* ' 8 l' ( F) 400 41 0 Feedwater Temperature Loop 1C (*F) 400 410 Feedwater Temperature Loop ID (*F) 400 410 Feed Pump Discharge Pressure (psig) 1140 1165 Control Bank D Position (steps) 144 180

'O Control Bank C Position (steps) 228 228 Boron Concentration (ppm) 754 754 (0856M)

TABLE 3.4.2-5 10% LOAD DECREASE AT 100% POWER

SUMMARY

4 INITIAL FINAL g CONDITION CONDITION Generator Load (MWe) 1140 1035 Nuclear Power (%) 99 89 O T*v' Loop 1A (*F) 587.5 586 Tave Loop 1B (*F) 586.5 585.5 Tave Loop IC (*F) 587.5 586.5 ,

Tave Loop ID (*F) 586.5 585 Tref (*F) 587 586 O

Delta T Loop 1A (%) 98 -98 ,

i i OPAT Loop 1A (%) 106 106 i

OTAT Loop 1A (%) 120 126 O

Pressurizer Pressure (psig) 2240 2235 Pressurizer Level (%) 62 58 Steam Generator Level Loop 1A (%) 68 68 O Steam Generator Level Loop 1B (%) 70 70 j Steam Generator Level Loop 1C (%) 67 67 i Steam Generator Level Loop 1D (%) 67 67 l

Steam Header Pressure (psig) 987 1045 O Steam Flow Loop 1A (pph) 3.90E6 3.55E6 Steam Flow Loop 1B (pph) 3.80E6 3.40E6 Steam Flow Loop IC (pph) 3.80E6 3.40E6 Steam Flow Loop 1D (pph) 3.80E6 3.40E6 Feedwater Flow Loop 1A (pph) 3.80E6 3.45E6

.O Feedwater Flow Loop 1B (pph) 3.80E6 3.35E6 Feedwater Flow Loop 1C (pph) 3.80E6 3.40E6 Feedwater Flow Loop ID (pph) 3.85E6 3.45E6 Feedwater Temperature Loop 1A (*F) 436 431 Feedwater Temperature Loop 1B (*F) 436 431 O Feedwater Temperature Loop IC (*F) 436 431 Feedwater Temperature Loop 1D (*F) 436 431-Feed Pump Discharge Pressure (psig) 1190 1205 Control Bank D Position (steps) 228 172

.O Control Bank C Position (steps) 228 228 Boron Concentration (ppm) 711 711 1

'g -129-

'(0856M)

i O TABLE 3.4.2-6 10% LOAD INCREASE AT 100% POWER

SUMMARY

INITIAL FINAL CONDITION CONDITION g

Generator Load (MWe) 1035 1140 Nuclear Power (%) 89.5 99 T* L p 1A (*F) 585 585 O Tave Loop 1B ('F) 584 584.5 Tave Loop 1C (*F) 585 585.5 Tave Loop 1D (*F) 584 584 Tref (*F) 586 587

.O Delta T Loop 1A (%) 98 98 l OPAT Loop 1A (%) 106 106 OTAT Loop 1A (%) 126 120

,0 Pressurizer Pressure (psig) 2235 2238 Pressurizer Level (%) 58 59 Steam Generator Level Loop 1A (%) 68 68 O Steam Generator Level Loop 1B (%) 70 70 Steam Generator Level Loop 1C (%) 67 67 Steam Generator Level Loop 1D (%) 67 67 Steam Header Pressure (psig) 995 920 O Steam Flow Loop 1A (pph) 3.55E6 3.95E6 i Steam Flow Loop 1B (pph) 3.00E6 3.75E6 Steam Flow Loop 1C (pph) 3.40E6 3.80E6 Steam Flow Loop ID (pph) 3.40E6 3.80E6 Feedwater Flow Loop 1A (pph) 3.45E6 3.80E6 ,

O Feedwater Flow Loop 1B (pph) 3.35E6 3.75E6 Feedwater Flow Loop 1C (pph) 3.40E6 3.80E6

- Feedwater Flow Loop 1D (pph) 3.45E6 3.85E6 Feedwater Temperature Loop 1A (*F) 430 436 )

Feedwater Temperature Loop 1B (*F) 429 436 )

.O Feedwater Temperature Loop 1C (*F) 430 436  ;

Feedwater Temperature Loop 1D (*F) 429 436 Feed Pump Discharge Pressure (psig) 1205 1150 Control Bank D Position (steps) 183 228 O Control Bank C Position (steps) 228 228 Boron Concentration (ppm) 711 711 O (0856M) l I

O 3.4.3 - FULL POWER PLANT TRIP, 2.52.38 OBJECTIVE 1

i This test was performed to verify the ability of the primary and  !

j) secondary plant and the plant automatic control systems to sustain a generator trip from full power and to bring the plant to stable j

4 conditions following the transient. The response time of the reactor coolant hot leg Narrow Range RTDs and the amount of turbine overspeed l that occurred upon a generator trip were also determined. '

TEST METHODOLOGY O

From a stable plant power of approximately 100% a generator trip was initiated by opening breakers in the switchyard.. This caused a turbine trip and a reactor trip. The operator was instructed to follow the Station Emergency Procedure to bring the plant to stable conditions. The

() data trending was terminated when Tave was stabilized at 557'F (no load Tave).

SUMMARY

OF RESULTS The response time of the hot leg narrow range RTDs for loop A (without

-() pressurizer) was 6.8 seconds and for loop D (with pressurizer), 7.85 seconds. The maximum turbine speed was found to be 1900 rpm. There was no unacceptable water hammer experienced and safety injection was not initiated. No pressurizer safety valves lifted, but two safety valves on steam generator C did lift, not meeting the acceptance criteria (no steam generator safety valves were to lift during the transient). The station

'() review determined that the lifting of the safety valves on steam generator C was acceptable since the power operated relief valve (PORV) on steam generator C had been blocked for modification (the surge pressure from the turbine trip had to be relieved and the PORVs had previously demonstrated their proper functioning). The lifting of the particular safety valves was deemed acceptable under the conditions of

'() the test by station review and agreed upon by offsite engineering review.

'O

.O

O 1

C) (0856M)

TABLE 3.4.3-1 TRIP FROM 100% POWER

SUMMARY

INITIAL FINAL CONDITION CONDITION O

Generator Load (MWe) 1200 0 Nuclear Power (%) 99 0 i

Tave Loop 1A (Auctioneered High) (*F) 589 557 O

Tref (*F) 588 557 Delta T Loop 1A (%) 100 1 OPAT Loop 1A (%) 107 107 O

OTAT Loop 1A (%) 118 150 l

Pressurizer Pressure (psig) 2235 2210 l I

O Pressurize:r Level (%) 61 g 21 '

)

Steam Generator Level Loop 1A (%) 67 54 Steam Generator Level Loop 1B (%) 70 62 Steam Generator Level Loop 1C (%) 68 56 Steam Generator Level Loop 1D (%) 66 60 0

Steam Header Pressure (psig) 970 1030 Steam Flow Loop 1A (pph) 3.80E6 0.00E6 Steam Flow Loop 1B (pph) 3.70E6 0.35E6 Steam Flow Loop IC (pph) 3.80E6 0.00E6 O Steam Flow Loop 1D (pph) 3.70E6 0.00E6 Feedwater Flow Loop 1A (pph) 3.70E6 0.40E6 Feedwater Flow Loop 1B (pph) 3.50E6 0.20E6 l Feedwater Flow Loop IC (pph) 3.80E6 0.00E6 Feedwater Flow Loop 1D (pph) 3.70E6 0.40E6 Feedwater Temperature Loop 1A ("F) 434.5 430.7 Feedwater Temperature Loop 1B (*F) 434.5 431.9 Feedwater Temperature Loop 1C (*F) 434.5 431.3 Feedwater Temperature Loop 1D (*F) 434.5 431.8 O Feed Pump Discharge Pressure (psig) 1200 1160 Control Bank D Position (steps) 210 0 Control Bank C Position (steps) 228 0 Boron Concentration (ppm) 711 711 (0856M)

'g 3.4.4 - PLANT TRIP FROM 25% POWER, 2.52.39 OBJECTIVE This test was executed to verify that primary and secondary plant and the

() plant automatic control systems could sustain a turbine trip from 25%

power and bring the plant to stable conditions without the use of the turbine bypass valves.

TEST METHODOLOGY i l

() With reactor power stable at approximately 25% power, a manual Turbine  ;

Trip / Reactor Trip was initiated. Unit Operators utilized plant emergency procedure 1BEP-ESO.1 to recover the plant following the reactor trip. l The test was concluded when a stable reactor coolant system average '

temperature was obtained. y i

O

SUMMARY

or REsutTs The test acceptance criteria was met by ensuring that the pressurizer safety valves did not lift, the pressurizer power operated relief valves did not lift, safety injection did not occur, no unacceptable water hammer in the steam generator or feedwater system occurred and that the

() i steam generator power operated relief valves operated as designed to bring the plant to hot no load conditions. Pressurizer pressure remained between 2180 psig and 2250 psig with pressurizer level remaining below 36%. The maximum steam pressure observed during the transient was 1160 psig. The minimum steam generator level observed during the transient was 54%. Table 3.4.4-1, on the following page, lists initial and final

() values for plant parameters trended during the test.

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() TABLE 3.4.4-1 TRIP FROM 25% POWER SUl24ARY INITIAL FINAL CONDITION CONDITION

})

Ger.erator Load (MWe) 150 0 Nuclear Power (%) 23.5 0.5 t

Tave (Auctioneered High) (*F) 562 558

()

Tref (*F) 563 557 Delta T Loop 1A (%) 29 1.5 .

OPAT Loop 1A (%) 107 107

()

OTAT Loop 1A (%) 150 150 Pressurizer Pressure (psig) 2240 2240

() Pressurizer Level (%) 32 30 Steam Generator Level Loop 1A (%) 66 64 Steam Generator Level Loop 1B (%) 66 55 Steam Generator Level Loop 1C (%) 66 55 Steam Generator Level Loop 1D (%) 65 55 O

Steam Header Pressure (psig) 980 1080 Steam Flow Loop 1A (pph) 0.60E6 0.00E6 Steam Flow Loop 1B (pph) 0.65E6 0.10E6 Steam Flow Loop 1C (pph) 0.60E6 0.00E6 C) Steam Flow Loop 1D (pph) 0.90E6 0.50E6 Feedwater Flow Loop 1A (pph) 0.70E6 0.00E6 Feedwater Flow Loop 1B (pph) 0.70E6 0.00E6 Feedwater Flow Loop 1C (pph) 0.60E6 0.00E6 Feedwater Flow Loop 1D (pph) 0.90E6 0.40E6 Feed Pump Discharge Pressure (psig) 1180 1360 Control Bank D Position (steps) 189 0 Control Bank C Position (steps) 228 0 0

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.o.

3.4.5 - SHUTDOW FROM OUTSIDE THE CONTROL ROOM,2.63.35 OBJECTIVE Shutdown From Outside The Control Room. verified that the unit could be O taken from 20% reactor power to Hot Standby conditions from outside the control room with a minimum shift crew. In addition, the potential to safety cool the plant to cold shutdown conditions from outside the control room was also demonstrated.

TEST METHODOLOGY

'o Utilizing abnormal operating procedure 1 BOA PRI-5, Rev. 5, the plant was manually tripped via the reactor trip breakers. The Unit Operator and Station Control Room Engineer proceeded to the remote shutdown panel to assume local control of plant equipment. Cooldown then proceeded without '

incident. Upon completion, control of the plant was transferred back to O the main control room.

SUMMARY

OF RESULTS Shutdown from outside the control room to hot standby conditions was accomplished with the minimum shift crew and maintained for a minimum of O 30 minutes. Actual time recorded was 34 minutes. The residual heat removal system was placed in operation and it demonstrated that the plant could be cooled at a rate of < 50*F in one hour. The test was performed successfully.

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3.4.6 - LARGE LOAD REDUCTION (75% & 100%), 2.64.39 OBJECTIVE i

i The purpose of this test was to verify the ability of the plant to

() sustain and stabilize following a 50% step load reduction from power levels of 75% & 100% without operator intervention. Acceptance criteria focused on the plant response during the test: no reactor or turbine

, trip, no damaging water haarner transients, no ESF actuations and no pressurizer or steam generator safety valves lifted.

O TEST METHODOLOGY 4

With the plant stabilized at 75% or 100% power, the DEHC system was programmed for a 2350 MWe/ Minute load reduction. Having notified the load dispatcher to assure the system would accept a 50% load reduction a '

558 MWe (50%) turbine load reduction was initiated. The following

() crucial data was trended throughout the transient: feedwater and reactor coolant system temperatures and pressures, reactor power, steam generator and pressurizer water levels and pressures, rod positions, boron

concentration, electrical output, steam and feedwater flows and feedwater regulating valve demand position.

'O

SUMMARY

OF RESULTS l The test was executed at 75% power three times due to manual intervention required to prevent a reactor trip on high steam generator level.

Multiple executions were also required at the 100% plateau due to feedwater regulating valve instability during the transient. Through  ;

() setpoint changes and field modifications acceptable plant performance was achieved. At the 75% level, Tave became stable after 11 minutes 10 seconds and at the 100% power level, 9.3 minutes were required to stabilize Tave. Manual emergency boration was also required during the 100% execution to maintain AI within Technical Specification limits.

This action was deemed acceptable by station 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 I

indicated no damaging water hammer occurred during either test sequerce.

There were no ESF actuations.

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l TABLE 3.4.6-1 ,

l LARGE LOAD REDUCTION AT 75% POWER

SUMMARY

INITIAL FINAL  ;

CONDITION CONDITION  !

O Generator Load (MWe) 825 280 Nuclear Power (%) 73 29

'O Tave (Auctioneered High) ( F) 581.5 566 l Tref (*F) 560.5 565.5 Delta T Loop 1A (%) 74 39 ,

Q OPAT Loop 1A (%) 107 107 OTAT Loop 1A (%) 136 150 Pressurizer Pressurc (psig) 2233 2235

~O Pressurizer Level (%) 53.5 36 Steam Generator Level Loop 1A (%) 66 66 Steam Generator Level Loop 1B (%) 66 66 Steam Generator Level Loop 1C (%) 66 66 Steam Generator Level Loop 1D (%) 66 66

O i Steam Header Pressure (psig) 985 1007 i

l Steam Flow Loop 1A (pph) 2.75E6 1.10E6 Steam Flow Loop 1B (pph) 2.70E6 1.25E6 Steam Flow Loop 1C (pph) 2.65E6 1.05E6

.O Steam Flow Loop 1D (pph) 2.70E6 1.00E6-Feedwater Flow Loop 1A (pph) 2.70E6 1.18E6 Feedwater Flow Loop 1B (pph) 2.70E6- 1.10E6 Feedwater Flow Loop 1C (pph) 2.75E6 1.12E6 Feedwater Flow Loop ID (pph) 2.75E6 1.27E6 Feedwater Temperature Loop 1A (*F) 411.8 344.9 Feedwater Temperature Loop 1B (*F) 411.8 344.9 Feedwater Temperature Loop IC (*F) 411.7 344.8 Feedwater Temperature Loop 1D ('F) 411.9 345.0 0 Feed Pump Discharge Pressure (psig) 1155 1108 j Control Bank D Position (steps) 172 85 Control Bank C Position (steps) 228 200 B r n Concentration < ppm) 748 748 O

(0856M)

TABLE 3.4.6-2 LARGE LOAD REDUCTION AT 100% POWER

SUMMARY

INITIAL FINAL CONDITION CONDITION O

Generator Load (MWe) 1178 527 Nuclear Power (%) 98 46.5 O Tave (Auctioneered High) (*F) 586 569 Tref (*F) 585 570 Delta T Loop 1A (%) 95 42 ,

'O OPAT Loop 1A (%) 107 107 OTAT Loop 1A (%) 128 150 Pressurizer Pressure (psig) 2245 2225 O Pressurizer Level (%) 55 46 Steam Generator Level Loop 1A (%) 66 66 Steam Generator Level Loop 1B (%) 70 70 Steam Generator Level Loop 1C (%) 66 66 Steam Generator Level Loop 1D (%) 66 65 Steam Flow Loop 1A (pph) 3.70E6 1.80E6 Steam Flow Loop IB (pph) 3.60E6 1.90E6 Steam Flow Loop IC (pph) 3.70E6 1.80E6 Steam Flow Loop IC (pph) 3.70E6 1.70E6 O Feedwater Flow Loop 1A (pph) 3.50E6 1.90E6 Feedwater Flow Loop IB (pph) 3.50E6 1.80E6 Feedwater Flow Loop 1C (pph) 3.60E6 1.80E6 Feedwater Flow Loop ID (pph) 3.60E6 1.90E6 Feedwater Temperature Loop 1A (*F) 432.6 386.8

.O Feedwater Temperature Loop 1B (*F) 432.9 387.0 Feedwater Temperature Loop 1C (*F) 432.7 386.8 Feedwater Temperature Loop 1D ('F) 423.5 386.8 FW Pump Discharge Header Pressuer (psig) 1181 1156 O CBD Position (steps) 197 95.5-CBC Position (steos) 228 213 Boron Concentration (ppm) 720 720

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3 5 - INSTRUMENTATION AND CALIBRATION TESTING _ .

3.5.1 - COMPUTER VERIFICATION - FLUX MAP 2.20.35 OBJECTIVE

-O-The Computer Verification of Flux Mapping was performed to verify the ability of_the process computer to receive and reduce data from the incore detectors. The test also included a section to compare selected output from the onsite Incore Code against the corporate mainframe Incore code.

O TEST METHODOLOGY The computer verification flux mapping was performed by taking 1 pass of a 12 pass flux map and comparing the voltage received at the process compu.er versus the voltages recorded by a strip chart recorder'at the

'() moveable detector panel. Upon completion of a flux map, 1 pass was selected for comparison. For this pass the strip chart recordings for all six detectors were hand digitized using Mylar overlays. These j digitized values were compared against the raw voltages reproduced by the moveable detector software output. The output from the moveable detector program was then run on both the process computer and mainframe Incore

'() codes. The output from these two programs was then compared for selected parameters.

SUMMARY

OF RESULTS The acceptance criteria for this test consisted of a 1% of full scale C) voltage agreement between the hand digitized data and the computer data.

The computer data failed to meet the 1% agreement criteria in comparison with the hand digitized data. A special procedure was run to further verify the computers ability to receive incore output. This procedure verified the computer's ability to accurately perceive the output from the moveable detector panel. This procedure was reviewed prior to and

() following execution by offsite engineering. During the course of power ascension the test was run at 0%, 30%, 50%, 75 and 100% power.

i Acceptance criteria was not met at all power levels. Based on the previously mentioned special procedure the station feels that we have satisfactorily demonstrated the accuracy of the computer. The only remaining item of concern centered around the deviation between onsite C) and offsite calculated core average axial offset . This deviation was 54 percent at 100 percent power. This item was a non acceptance criteria problem and is being tracked by the station.

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. - . . . . , _ - - , , ,- . , - . . - - . . , _ . .- - ,_ - - .,~

I O 3.5.2 - CALIBRATION OF STEAM AND FEEDWATER FLOW, 2.34.31 1

OBJECTIVE I The Calibration of Steam and Feedwater Flow test was performed to verify l g) that the feedwater flow and steam flow instrumentation was calibrated properly.

TEST METHODOLOGY With the plant stable at 0%, 30%, 50%, 75% 90%, and 100% power, plant

() parameters concerning steam and feedwater flow were recorded. These parameters included steam flow, feedwater flow, tempering flow, feedwater temperature, and steam generator pressure. Differential pressure gauges were also installed on the main feedwater venturi flow nozzle to accurately measure feedwater flow. Steam generator blowdown was isolated '

for the duration of the test. ,

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

O

SUMMARY

OF RESULTS The final test results indicated that the feedwater and steam flow transmitters were calibrated to agree within 1% of full scale when compared to the opposite channel on the same steam generator. The feedwater flow transmitters were also calibrated to within .75% of the C) test instrument monitoring the main feedwater flow venturi. The comparison between the steam and feedwater flow transmitters was found to be 1.2%, 3.8%, 2.9%, and 2.4% on the 1A, IB, 1C and ID steam generators respectively. The difference between steam and feedwater could not be reduced due to the level oscillations which presently exist in the steam generators at this time. A special test procedure was written and

() executed to obtain the steam and feedwater flow transmitter output on a strip chart recorder. Doing this allowed direct comparison between steam flow transmitters at the same point in time. The feedwater flow transmitters were also compared by this method. After the resolution to the level oscillation phenomenon has been successfully implemented, a better comparison between steam and feedwater flow will be made.

C) Rescaling of the steam flow transmitters to better match feed flow may be determined to be necessary at that time. A station action item was generated to investigate the mismatch after the level oscillation condition is resolved.

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O 3.5.3 - THERMAL POWER MEASUREMENT AND STATEPOINT DATA COLLECTION (PRE-CRT.TICAL, 0%, 30%, 50%, 75%, 90%, and 100%), 2.47.32A-E OBJECTIVE

() The three basic objectives of this test series were to determine thermal power using calorimetric data, collect control and protection instrumentation data at steady - state power levels (statepoints) and align AT instrumentation and check Tave alignment at 100% power by extrapolation. There were no acceptance criteria for any test in this series. Table 3.5.3-1 illustrates the various tests and intent of each.

O TEST METHODOLOGY Statepoint data was collected at the 0%, 30%, 50%, 75%, 90% and 100%

power levels while delta T and Tave instrumentation alignment was done at

, the pre-critical, 75% and 100% sequences. Thermal power'calorimetrics

() were performed at each execution mentioned above and 40% and 98% power.

Calorimetric data included feedwater temperature and delta pressure, steam pressure, atmospheric pressure, tempering flow and steam generator blowdown flow.

The average of three separate readings taken within 20 minutes of each

) other for each of these parameters assured statistically valid calorimetric input values. Alignment of the delta T and Tave process instrumentation was accomplished by plotting these values gathered at various power levels against the respective calculated calorimetric power. A trend line fitted through the points was extrapolated to 100%

power to predict the 100% delta T and Tave values.

SUMMARY

OF RESULTS The tests conducted at pre-critical and 0% power identified minor out-of-tolerance conditions on the various delta T summing amplifiers.

Administrative controls were implemented and actions were taken to

() properly align delta T loops and preclude the reactor from reaching the point of nuclear heat addition of 10-6 amps on the IR channels until

this was done. l Power ascension testing between 0% power and 100% power progressed well with only minor instrument anomalies which were documented and resolved.

f) These out-of-range anomalies were due to iterative calibrations,.

hysteresis effects and an overly conservative full power delta T of 51*F.

O 53 l (0856M)

O 3.5.3 - THERMAL POWER MEASUREMENT AND STATEPOINT DATA COLLECTION (PRE-CRITICAL, 0%, 30%, 50%, 75%, 90%, and 100%), 2.47.32A-E (Continued)

At the 50% power test sequence more accurate delta pressure gauges were installed for feedwater flowrate measurement. During the initial 75%

() power sequence the data scatter was too broad for proper analysis due to unstable plant conditions. Reperformance of the 75% test during a more stable period was utilized for calculating delta T and Tave values. A change in methodology from that noted above was also employed.

Westinghouse recommended using an extrapolation of the change in reactor

() . coolant enthalpy to predict delta T. Earlier values were corrected for new RTD cross-calibration data (see Section 3.5.4) and the 100%

extrapolated values for delta T and Tave were as follows:

Delta T (*F) Tave (*F) '

LOOP A 55.5 585.8

() LOOP B 55.8 585.7 LOOP C 54.7 586.1 LOOP D 55.2 584.8 The station's documented concurrence with the use of nominal full power Tave = 588.4*F rather than extrapolated Tave values for delta T C) calculations was based on the expectation that turbine impulse pressure channels would be rescaled subsequent to the test thus bringing the full power Tave closer to the nominal value than to the extrapolated values.

During the 100% sequence with calculated calorimetric power at 97.8%, the new 100% delta T values were extrapolated and set in the delta T loops as 56.19*F, 56.91*F, 55.88*F and 56.30*F for loops lA through 1D, C) respectively. Also, all Tave values were verified below the design maximum value of 588.4*F.

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TABLE 3.5.3-1 Thermal Power Measurement and Statepoint Data Collection C) Summary of Objectives TEST SEQUENCE THERMAL PWR STATEPOINT INST. ALIGNMENT 2.47.32A Pre-Crit X 2.47.32C 30% 0% X

.C) 2.47.32C 30% 30% X 2.47.32B 50% 40%

2.47.32C 50% 50% X 2.47.32D 75% 75% X X 2.47.32C 90% 90% X '

2.47.32B 100% 98%

C) 2.47.32E 100% 100% X X e

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O 3.5.4 - INCORE THERMOCOUPLE /RTD CROSS CALIBRATION, 2.47.33 OBJECTIVE The objective of the test was to verify reactor coolant system RTD

() performance over a range of temperatures. Additionally a cross I calibration data base was developed for the incore thermocouple system l and reactor vessel level indication system.  !

TEST METHODOLOGY

() 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 +2.5'F drift, steam generator levels within 10% of one another, all four reactor coolant pumps running and the plant in mode 3, 4 or 5. For the RTDs, four sets of data were taken at each plateau by monitoring RTD resistance at the C) master test cards 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.

O

SUMMARY

OF RESULTS The cross calibration was performed twice as initial data review by Westinghouse identified excessive temperature deviations at each plateau. Additional cross calibrations were reperformed at 380*, 450*,

and 557*F upon completion of the 50% testing. Data was collected within

,C) 10 minutes to minimize the effects of temperature variations.

Westinghouse reviewed the new data and developed new resistance calibration data sheets for the instruments. Based on Westinghouse calculations the narrow range RTD amplifiers were calibrated. All devices tested were determined to be acceptable.

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3.5.5 - OPERATIONAL ALIGNMENT OF EXCORE NUCLEAR INSTRUMENTATION (PRIOR TO.

INITIAL CRITICALITY, DURING POWER ESCALATION, AT FULL POWER, AFTER SHUTDOWN FROM POWER OPERATIONS OF AT LEAST 800 MWD /MTU), 2.52.35C-F OBJECTIVE O

This series of tests was performed to verify that the excore nuclear i instrumentation system was functioning per design and capable of detecting, alarming and mitigating unplanned reactivity excursions.

TEST METHODOLOGY O

{

Selected parameters and alarms were evaluated monitored, and determined during various testing phases.

Prior to Initial Criticality the proper setpoints, alarms and trip

functions were verified for the Source Range'(SR), Intermediate Range

() (IR) and Power Range (PR) channels.

During Power Escalation the overlap between the SR and IR channels and )

l the IR and PR channels was determined. The flux deviation alarm settings and high level trip setpoints for the IR and PR channels were verified. l Finally, the PR detector response versus core power was checked for

~

1

,3 ( linearity by aligning the channels with the calorimetric thermal power calculated in the thermal power measurement test.

At Full Power the characteristic curves for the IR and PR channels were determined and the operating voltages were chosen and verified to be in the plateau region of these curves.

O After Shutdown from Power Operations of at Least 800 MWD /MTU the operating voltages for the SR channels and the compensating voltages for the IR channels were determined and set.

I

SUMMARY

OF RESULTS 1 The setpoints of the SR channel trips, IR channel trips, and PR channel trips were verified as meeting all associated acceptance criteria. A minimum overlap of 1.5 decades was observed on all SR/IR and IR/PR channel combinations. Specifically, the overlaps for the four SR/IR )

channel combinations were observed to be above 1.8 decades and the 4

C) overlaps for all eight IR/PR channel combinations were observed to be

above 3.7 decades. All detector operating voltages were verified to be within the expected ranges and fell on the plateau region of the i characteristic curves. Power range detector voltages were set at 766 VDC j (N41), 797 VDC (N42), 797 VDC (N43), 800 VDC (N44). Intermediate range detector voltages were set at 799 VDC (N35) and 800 VDC (N36). Source

') range voltages were set at 1875 VDC (N31) and 1812 VDC (N32). The compensating voltages for the IR detectors were set at -17.71 VDC (N35) and -25.23 VDC (N36). These values were outside the expected range of

-50 VDC to -100 VDC. The values were verified with Westinghouse to reflect acceptable detector operation consistent with a relatively low level of gamma radiation early in plant life.

()

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3.5.69 - AXIAL FLUX DIFFERENCE INSTRUMENTATION CALIBRATION AT 50% OR 75%

POWER, 2.52.36A i

OBJECTIVE This test was performed to assure that a linear relationship exists

() between the excore neutron detector currents and the incore axial flux distribution. Once established, this linearity was used to accomplish various calibrations of the excore channels, the OTAT inputs, axial flux difference meters and plant process computer.

O TEST METHODOLOGY Baseline flux distribution data was obtained with three preliminary flux maps at 50% reactor power and with eleven flux maps,at 75% reactor power. The incore axial offsets determined from the flux map data were graphically compared to both the excore a'xial offsets and excore channel l() normalized top and bottom detector currents. Equations were developed for the least squares fit lines through the observed data points. These equations and their constants (ie. slope, y intercept) were subsequently I

used to determine expected currents, pseudo resistances and voltages measured at equipment test points. The measured outputs were then verified or adjusted with various potentiometers to meet the calculated

() values.

SUMMARY

OF RESULTS The various signals generated from the incore flux map data were successfully used in calibrating the OTdT setpoints, computer points l(3 and axial flux difference meters at both the 50% and 75% reactor power

levels.

i i Total core power normalized to 100% reactor power as measured by the excore channels was verified to be within 1% accuracy at 75% power and 5%

at 50% power. Voltages to the pen recorders and computer points which

() were calculated from measured detector currents were verified to be within 2% accuracy. The OTAT setpoints were verified to be within 1%

accuracy.

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3 5.6b - AXIAL FLUX DIEFERENCE ADJUSTMENT AT 100% POWER, 2.52.36B .

OBJECTIVE' This test was performed to verify that the 100% calibration values

() extrapolated from actual 75% power data were within 11% accuracy when  !

compared to actual 100% (+0%, -2%) power data. If not, then fine tune l adjustments were made to. correct each nuclear instrumentation channel to

]

,i within 11% of the actual 100% power values. 1 1

i TEST METHODOLOGY O

j A nominal 100% power flux map was taken to gather actual full power values for core average axial offset, average power level and top and bottom power range detector currents. The equation for expected detector i current as a function of axial offset developed at the 75% power level I

was employed using the actual 100% axi'al offset from the 100% flux map to

() predict 100% detector currents. The actual 100% detector currents were then compared to the predicted 100% detector currents. Selection of a calibration current was performed by normalizing the actual and predicted detector currents (extrapolated for 120% power) to an expected current.

i. This current was formulated from a linear equation with the same slope as

. at 100% power but with the yintercept and incore axial offset l()

extrapolated to 120%.

4

SUMMARY

OF RESULTS

! The quantitative comparison of the actual 98% power detector currents and predicted currents showed seven (7) of the eight (8) top and bottom l)( predicted currents to be out of the i 1% accuracy. Top detector current average error was 1.1% with a maximum value of 1.6% and a minimum of l 0.1%. Bottom detector current average error was 4.8% with a maximum value of 5.1% and minimum of 4.4%. These anomalies were corrected by fine adjustments of the power range detector instrumentation to normalized 120% power detector currents. The station accepted the

$) results.

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O E PARTS MONITORIg;, 2.133.30B OBJECTIVE The test objective was to gather noise frequency response data at 0%,

() 30%, 50%, 75%, 90% and 100% reactor power. This data will be used as a reference baseline when analyzing suspected loose parts in the NSSS.

TEST METHODOLOGY At each power level a recording of each of the twenty-four loose parts

() accelerometers was made. A frequency spectrum analysis of each recording was performed and graphical charts for the system's response were developed.

SUMMARY

OF RESULTS

() The data collected in the test will provide a baseline for each of the twenty-four accelerometers. Two sensor problems were identified. The first problem was with sensor #21. This sensor did not respond as expected at 30% power, however at higher power levels its response was l acceptable. At 90% and 100% reactor power sensor #8 did not respond as

()

expected. Due to the similar locations of sensor #7 and #8 and similar response characteristics noted at 50% reactor power additional testing of sensor #8 was not required.

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3.5.8 - STARTUP ADJUSTMENTS OF REACTOR CONTROL SYSTEMS, 2.64.44 OBJECTIVE The objective of this test was to determine the Tave program which would result in the design steam pressure at full load and therefore optimum

() plant efficiency without exceeding pressure limitations for the turbine or temperature limits for the maximum allowable Tave. This was i accomplished by. making adjustments to the reference Tave (Tref) program.

TEST METHODOLOGY

.O Data for this test was obtained from executions of the Thermal Power Measurement and Statepoint Data Collection test, 2.47.32A-E at 0%, 30%,

50%, 75%, 90%, and 100% power. Data utilized for this test 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 rescaling would be necessary. The same data was reviewed at 100% power for rescaling required. After any scaling changes were completed, additional data was taken to verify the adequacy of the scaling.

SUMMARY

OF RESULTS O

At 75% power, Tave extrapolated to 585.8'F which was verified to be below the design maximum of 588.4*F. Steam generator pressure was extrapolated to 974 psia. Turbine impulse chamber pressure was extrapolated to 770.3 1

psig at 120% power. Since, at the time of the test impulse pressure was scaled to 846 psig, impulse pressure was rescaled. At 100% power, Tave

,() extrapolated to 587.2 *F. This was verified to be below the design maximum of 588.4 'F. Steam generator pressure was extrapolated to 1004

, psia which was outside the range of 990 + 10 psia. Based on these 1 results, a new target Tref was determined to be 587.4*F and was input into the appropriate instrument loops. This reduced steam pressure to within the above mentioned range. In addition, turbine impulse chamber C) pressure was rescaled to approximately 820 psig at 120% power. Following all of the scaling changes, additional statepoint data was obtained to verify the adequacy of the changes.

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O 3.5.9 - NUCLEAR STEAM SUPPLY SYSTEM ACCEPTANCE TEST, 2.80.49 OBJECTIVE This test was performed to demonstrate the reliability of the Nuclear

"'(T Steam Supply System (NSSS) to maintain its warranted output of 3425 MWth 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 at within five (5) percent of its rated output as determined by a calorimetric calculation.

Plant conditions were stabilized at their design values for 40-50 hours when the stabilization was verified by calorimetric data aquisition for two consecutive half hour periods. Steam quality was unknown at this time since the Steam Generator Moisture Carryover test, 2.51.45 had not *

() yet been performed. It was therefore assumed to be at its design value.

A four hour performance measurement was initiated to determine actual reactor power by collecting calorimetric data each hour. The remaining hours of the test were then completed.

SUMMARY

OF RESULTS O

The four hourly calorimetrics demonstrated NSSS output to be an average of 3386 MWth corresponding to 98.9% power. 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 in two segments by completing 125 total hours. The two segments were composed of 70 hours8.101852e-4 days <br />0.0194 hours <br />1.157407e-4 weeks <br />2.6635e-5 months <br /> and 55 hours6.365741e-4 days <br />0.0153 hours <br />9.093915e-5 weeks <br />2.09275e-5 months <br /> respectively. During the initial run, the performance measurement was completed. The 100 hour0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> run

() had to be terminated due to Technical Specification limitations on a residual heat removal pump. The remaining run time was completed at a later date. It should be noted that while this test demonstrated the reliability of the NSSS in developing its warranted output, it did so at i a steam quality of 99.22% instead of the contract value of 99.75%.

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O REFERENCES

1) Byron Station Final Safety Analysis Report

2) Regulatory Guide 1.68, Revision 2 0

3) Byron Station Technical Specifications

4) Byron Station Operating License NPF-23

5) Byron Station Operating License NPF-37 O

6) WCAP 10315 the Nuclear Design and Core Physics Characteristics of the Byron Unit 1 Nuclear Power Plant Cycle 1, Revision 1 p

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,em C:mmormcith Edison Byron Nuclear Station ud 7// d

(' __. Byron,

} 4453 North German Church Road Illinois 61010 October 25,1985 EdS/k LTR: BYRON 85-1421

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FILE: 2.7.1000 g

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-n_;Anu - p s Mr. James G. Keppler Regional Administrator M-g  %

Region III U. S. Nuclear Regulatory Commission 799 Roosevelt Road Glen Ellyn, IL 60137

SUBJECT:

Byron Nuclear Power Station Unit 1 Cycle 1 Startup Report Docket No. 50-454

Dear Mr. Kepnler:

The enclosed Startup Report for Byron Unit 1, Cycle 1 is being submitted to you in accordance with Technical Specifications 6.9.1.1, 6.9.1.2 and 6.9.1.3.

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R. E. Querio Station Manager Byron Nuclear Power Station REQ /PDR/bf

Enclosure:

Byron Unit 1, Cycle 1 Startup Report ec: D. Farrar OCT 311985

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1. 24/032

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