TXX-9332, Cycle 1 Startup Rept

Cycle 1 Startup Rept

ML20082Q739
Person / Time
Site:	Comanche Peak
Issue date:	09/15/1993
From:	TEXAS UTILITIES ELECTRIC CO. (TU ELECTRIC)
To:
Shared Package
ML20082Q725	List: ML20082Q722 TXX-9332, Cycle 1 Startup Rept
References
TXX-93325, NUDOCS 9505010068
Download: ML20082Q739 (300)

Text

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ENCLOSURE TO TXX-93325 COMANCHE PEAK STEAM ELECTRIC STATIC UNIT 2 CYCLE 1 STARTUP REPORT  :

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p 9505010068 DR 930915 ADOCK 05000446 i PDR i

n TABLE OF CONTENTS Section Title Page Title Page 1

Table of Contents 2

List of Tables 3 List of Figures 5

1.0 Introduction 7

2.0 Discussion of the Initial Startup Program 10 3.0 Discussion of the Initial Startup Tests 25 3.1 Core Loading 28 3.2 System Testing After Core Load 44 and at Various Power Levels 3.3 Physics Testing 87 3.4 Transient Testing 107 3.5 Instrumentation and Calibration Testing 129 3.6 Deferred Preoperational Testing 167 4.0 References 189 Attachment A Comanche Peak Steam Electric Station 190 Unit 2 Loose Parts Monitoring System Special Report LIST OF TABLES TABLE TITLE PAGE 1.0-1 Cross Reference of FSAR Table 14.2-3 8 and Unit 2 Cycle 1 Startup Report 2.0-1 Comanche Peak Unit 2 Major Milestones 11 2.0-2 Operational Modes 12 2.3-1 HZP Physics Testing Results 18 l

-20 2.4-1 30% & 50% Power Flux Map Results 2.5-1 75% Power Flux Map Results 22 2.6-1 100% Power Flux Map Results 24 3.0-1 List of Test Summaries 25 3.2.2-1 Steam Generator Level Control Summary 49 3.2.5-1 RCS Chemistry Summary 57 3.2.5-2 Steam Generator Chemistry Summary 58 l 3.3.7-1 Measured and Inferred vs Predicted Rod Bank Worths 104 3.4.2-1 Design Load Swing Tests Summary 113 3.4,2-2 10% Load Decrease at 50% Power Summary 114 1.4.2-3 10% Load Increase at 50% Power Summary 115 3.4.2-4 10% Load Decrease at 75% Power Summary 116 3.4.2-5 10% Load Increase at 75% Power Summary 117 3.4.3-1 Trip From 100% Power Summary 120 3.4.6-1 Large Load Reduction Tests Eummary 127 3.4.6-2 Large Load Reduction Summary 128 4

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LIST OF TABLES (Continued)

TABLE TITLE E]LQX 3.5.1-1 Calibration of Steam Flow Transmitters 133 .

3.5.2-1 Process Temperature /N16 Tests vs. Plant Conditions Matrix 141  !

3.5.3-1 Nuclear Instrumentation Results Summary 145 3.5.4-1 Incore/Excore Detector Calibration Summary 152 .

3.5.6-1 Startup Adjustments Summary 156 f

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LIST OF FIGURES FIGURE TITLE PAGE 2.0-1 ISU Program Summary 13 3.1.1-1 Unit 2 Core Loading Pattern - 30 Initial Nucleus of Assemblies 3.1.1-2 Unit 2 Core Loading Pattern - 31 Partial Bridge Across Core 3.1.1-3 Unit 2 Core Load Pattern - 32 i Completed Bridge Across Core 3.1.1-4 Unit 2 Core Loading Pattern - 33 Partial Completion of Core  ;

3.1.1-5 Unit 2 Core Loading Pattern - 34 Final Configuration i 3.1.3-1 ICRR vs Fuel Assemblies Loaded - 39 Source Range Channels 3.1.3-2 ICRR vs Fuel Assemblies Loaded - 40 Unit 2 Temporary Detectors 3.2.4-1 Movable Incore Detector Path Locations 55 l 3.2.5-1 Lithium vs. Boron Curve 59 3.2.14-1 Pressure Response to Opening Both 81 Spray Valves l 3.3.2-1 ICRR During RCC Bank Withdrawal 92 3.3.2-2 ICRR During RCS Boron Dilution 93 3.3.6-1 Reactivity vs. Temperature 101 3.3.7-1 Differential and Integral Rod Worth 105 Rod Swap Reference Bank 3.5.3-1 N41 Power Range Current vs. Calorimetric 147 Power fjl j

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l LIST OF FIGURES .l (Continued) l FIGURE TITLE PAGE *!

3.5.3-2. N35 Intermediate Range Current vs. 148 .

Calorimetric Power i 3.5.6-1 Example Program Tref vs. Cslorimetric Power 157. 'I 3.5.6-2 Example Impulse Pressure vs. Calorimetric 158 Power 3.5.6-3 Example Pressurizer Level vs. Calorimetric 159 l Power  !

t 3.5.6-4 Example Steam Pressure vs. Power 160

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L 1.O - INTRODUCTION i

This report describes the required testing at Comanche Peak Steam Electric Station, Unit 2, from the preparations for loading the first fuel assembly into the reactor until the plant was place in commercial operation. This report satisfies the requirement of the  ;

Comanche Peak Technical Specifications, Section 6.9.1.1, that a Startup Report be submitted to the NRC within 90 days after completion of the Startup Testing Program. Attachment A is the '

Loose Parts Monitoring System Startup Report submitted per requirements of Regulatory Guide 1.133. l Comanche Peak Steam Electric Station, location in North Central Texas, is a two unit nuclear power plant. Unit one completed initial startup in 1990 and was declared to be in commercial-  ;

operation on August 13, 1990. Each unit utilizes a four loop .

I Westinghouse Pressurized Water Reactor as the Nuclear Steam Supply System. Westinghouse Electric Corporation, Stone & Webster Engineering Corp., Gibbs & Hill, Inc., Impell Corp., Ebasco,jointly Brown Bechtel and the TU Electric Company '

& Root, Inc.,

participated in the design and construction of Comanche Peak. The l plant is operated by the TU Electric Company. l The Nuclear Steam Supply System is designed for a thermal power output of 3425 MWth (3411 MWth reactor power). The not electrical output is 1150 MWe. Cooling for the plant is provided by the Squaw' ,

creek Reservoir, a 135,062 acre-foot man-made lake. Post design l basis accident cooling is provided by a separate 367 acre-foot Safe Shutdown Impoundment.

Table 1.0-1 provides a cross reference between the %est summaries  ;

in the Final Safety Analysis Report and the se% ions of this report.

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Table 1.0-1 Cross Reference of FSAR Table 14.2-3 and Unit 2 Cycle 1 Startup Report FSAR Table 14.2-3 Startup SHEET NUMBER TITLE Report Section 1 Cover Sheet N/A 2&2a Reactor Coolant System Flow Test 3.2.8, 3.2.14 3 Reactor Coolant System Flow 3.2.9 Coastdown Test 4 Control Rod Drive Tests 3.2.11, 3.2.12, 3.5.9 5 Rod Position Indication 3.2.11 6&7 Reactor Trip System 3.2.13 8 Auxiliary Startup Instrumentation 3.1.2 Test 9&10 Calibration of Process Temperature and Nuclear Instrumentation 3.5.2, 3.5.3 11 Chemical Tests 3.2.5 12 Radiation Surveys 3.2.6 13&l3a Process and Effluent Radiation No Unit 2 ISU Monitoring Test test required 14 Moderator Temperature Reactivity Coefficient 3.3.6 15 Control Rod Reactivity Worths 3.3.7 l 16 Boron Reactivity Worth 3.3.7, 3.3.8 17 Core Reactivity Balance 3.3.5 18 Loss of Offsite Power 3.4.1 i 19 Rod Drop Tests 3.2.12 20&21 Flux Distribution Measurements 2.4,2.5,2.6 l

i Table 1.0-1 v Cross Reference of FSAR Table 14.2-3 and Unit 2 Cycle 1 Startup Report (Continued)

FSAR Table 14.2-3 Startup SHEET NUMBER TITLE Report Section 22 Core Performance Evaluation 2.4,2.5,2.6, 3.3.5,3.5.3, 3.5.5 23&24 Unit Load Transients 3.4.2,3.4.3, 3.4.6 25&26 Remote Shutdown 3.4.4, 3.4.5 ;

27 Intentionally Left Blank N/A 28 Turbine Trip / Generator Load Rejection 3.4.3 29&30 Reactor Coolant Leak Test 3.2.10 31&32 Rod Control System Test 3.2.12 33 Automatic Control System Test 3.5.9 i

34&35 Incore Nuclear Instrumentation 3.2.4

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2.0 - DISCUSSION OF THE INITIAL STARTUP PROGRAM The Comanche Peak Unit 2 initial startup testing program consisted of single and multi-system tests that were performed commencing with initial fuel operation.

loading and continuing through full power The intent of these tests is to assure that tests deferred the plant from the preoperational test program are performed; that is safely brought to rated capacity; performance that plant is satisfactory in terms of established design criteria; and to demonstrate, where practical, that the plant is accidents.of These capable withstanding anticipated transients and postulated tests demonstrated overa'.1 plant performance and <

included such activities and power ascension tests.

as precritical testing, low power tests, Testing sequence documents were utilized for each plateau activities at that plateau.

to coordinate the sequence of testing In the subsections that follow, a description of the testing at each plateau is provided. The descriptions include additional details concerning special license conditions and commitments made to the Nuclear Regulatory Commission prior to completion of the startup testing program, where applicable. Also included as a part of Section 2.0 are tables and figures showing major milestones for Comanche Peak Unit 2 which occurred during the initial startup program and a list of operational modes as defined by the Technical Specifications.

The duration of the Initial Startup Program, from receipt of the lower power license on 2/02/93 until the declaration of commercial operations on 8/03/93 was 182 days. This duration also included a 51 day surveillance outage that was performed during the 75% power test sequence.

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TABLE 2 . 0-1, COMANCHE PEAK UNIT 2 MAJOR MILESTONES MAJOR MILESTONE _S DATE 5% Power License Received 2/02/93 Fuel Load Started 2/04/93 Fuel Load Completed 2/07/93 Initial Criticality 3/24/93 5% License (Low Power) Tests Completed 3/27/93 Full Power License Received 4/06/93 i Entered Mode 1 4/06/93 Initial Synchronization to Grid 4/09/93 30% Power Reached 4/13/93  ;

I 50% Power Reached 4/17/93 l l

75% Power Reached 5/15/93 l 1

Start of Surveillance Outage 5/20/93 Completion of Surveillance Outage 7/10/93 (return to 75% powor) 100% Power Reached 7/18/93 Test Review Group Approves Startup Test Program 8/02/93 V.P. Nuclear Operations Declares Completion of Startup Test Program and Commencement of Commercial Operation 8/03/93 l 4

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i TABLE 2.0-3 OPERATIONAL MODES i

AVERAGE REACTIVITY  % RATE COOLANT MODE CONDITION, Keff THERMAL POWER

• TEMPERATURE

1. POWER OPERATION 2 0.99 > 5% 2 350*F

2. STARTUP 2 0.99 s 5% 2 350*F l

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

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

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>200*F

5. COLD SHUTDOWN < 0.99 0 s 200*F  !

6. REFUELING ** s 0.95 0 s 140*F >

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• Excluding decay heat.  !

• • Fuel in the reactor vessel with the vessel head closure bolts i

less than fully tensioned or with the head removed.

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Figure 2.0-1 ISU PROGRAM

SUMMARY

100 %

100- ISU

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h*75- @ Planned Post Test Outage 7/27 - all ~

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2.1 - INITIAL FUEL LOAD SEOUENCE - ISU-001B OBJECTIVE The Initial Load Sequence document definen the sequence of testing and other operations to prepare for and perform initial core j i

loading. This test partially satisfies activities described in FSAR Section 14.2.10.1. f 1

TEST METHODOLOGY The Fuel Load Sequence Document is used to coordinate the sequence l of operations associated with the initial core loading program.  ;

This sequence includes scheduling of the individual startup tests i and selected key permanent plant procedures associated with core loading. This document specifies 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. This document also provides the criteria for stopping core loading, the criteria for emergency boration, and the actions to be followed prior to the resumption of core loading in the event loading was stopped prior to completion.

SUMMARY

OF RESULTS Initial core loading of 193 fuel assemblies took 93 hours0.00108 days <br />0.0258 hours <br />1.537698e-4 weeks <br />3.53865e-5 months <br />. Prior to the start of core loading, the condition of the reactor vessel and associat*ed components, the reactor coolant system, instrumentation, and administrative controls were verified to be acceptable. The sequence procedure verified that reactor coolant l system chemistry was properly established and maintained and verified timely nuclear instrumentation neutron response checks.

The procedure also ensured that a final general fuel assembly visual inspection was performed. Fuel loading operations were performed using permanent plant procedures. 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 Supervisor.

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' 2.2 - POST CORE LOAD PRECRITICAL TEST SEOUENCE (PCLPC) - ISU-010B 3 t

OBJECTIVE  :

The PCLPC Sequence Document defines the sequence of tests and ,

- operations to be performed between completion of initial core j loading and prior to initial criticality. This testing is

• performed in Technical Specification Modes 5, 4 and 3.

TEST METHODOLOGY t

This document ensures that core load testing had been successfully completed and results approved prior to continuation of the testing ' ,

program. This document schedules the perfbrmance of precritical 1 tests to ensure the necessary testing is completed prior to initial  !

criticality. This procedure . governs the sequence of testing through Modes 5, 4 and 3. Plant operating procedures are utilized where appropriate to establish necessary plant conditions.

SUMMARY

OF RESULTS l l

Results of individual tests completed during ' the post core load i precritical testing phase are discussed primarily in Sections 3.2 j

and 3.6 of this report. A daily log of RCS and pressurizer boron concentration was kept to ensure adequate shutdown margin during 'j testing. RCS baron concentration varied between 2050 ppm and 2113  :

ppm. This insured that the boron concentration was greater than the .

2000 ppm refueling concentration at all times.- Upon completion of this testing phase, plant systems were aligned as directed by the Shift Supervisor.

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1 2.3 - INITIAT. CRITICALITY & LOW POWER TEST SEQUENCE (IC & LPT) - 1 ISU-101B  !

OBJECTIVE l The IC & LPT Sequence Document-defines the sequence.of tests and.

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operations, beginning with initial criticality, which. constitute '

the low power physics testing program. This program of low power physics testing verifies the design of the reactor by performing a- i series of selected measurements including control bank worths and  :

Moderator Temperature Coefficient. This test sequence partially  !

satisfies activities described in FSAR Sections- 14.2.10.2 and  !

14.2.10.3. i TEST METHODOLOGY I

This document ensures.that post core loading procritical testing {

has 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 are  ;

verified to have a signal to noise ratio greater than 2.and power range high level trip setpoints are conservatively set to 5 20% of  ;

full power. This procedure sequences the low power physics testing into an officient order and ensures that'all required testing is performed. Surveillance Requirements for Technical Specification i

special Test Exception 3.10.3 usage are also controlled by this i test. This Technical Specification Special-Test Exception permits  !

physics testing in limited non-routine reactor controls  !

configurations.

j A reactivity monitoring system is set up using a power range NIS l channel detector output to monitor core flux. This device is a i digital computer that calculates reactivity values based on the rate of change of core flux levels. This device is used in core  :

physics testing to make measurements of control rod, soluble boron {

and moderator temperature worths.

Plant operating procedt.res are utilized where appropriate to establish and maintain plant conditions.  ;

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.I 2.3 - INITIAL CRITICALITY & LOW POWER TEST SEOUENCE (IC & LPT) -

3 ISU-101B (Continued) ,

SUMMARY

OF RESULTS Results .of individual tests completed during the initial criticality and low power test sequence are discussed primarily in

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Section- 3.3 of this report. A tabulation _of key physics measurement results is also included in Table 2.3-l'. All. required tests were performed. Initial criticality was achieved without '

incident on 3/24/93. A low power physics testing power range was a determined and the reactivity. monitoring system was. verified to.be operating properly. Boron endpoint concentration measurements and associated core reactivity balance calculations were performed.-

The Isothermal Temperature Coefficient was then measured in order to calculate the Moderator Temperature coefficient. -The Moderator '

Temperature coefficient, while slightly positive, was less positive .

than the Technical Specification limit and rod withdrawal limits l were not required to be imposed. Control rod worths were verified using the bank exchange method (rod swap method). All required testing was satisfactorily completed.

Upon completion of this testing phase, the plant was aligned as directed by the Shift Supervisor.

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TABLE 2.3-1 >

HZP PHYSICS TESTING RESULTS l MISC. PHYSICS TESTING RESULTS Actual Allowed Rance  ;

All Rods Out Critical Boron (ppm) 1022.5 942 to 1042 '

Reference-Bank-In Critical .

i Boron (ppm) 959.5 878 to 978  !

Icothermal Temperature -1.08 -2.82 to +1.18 Coefficient (pcm/*F)

Moderator Temperature +0.75 < +5 Coefficient (pcm/*F)

Differential Boron Worth (pcm/ ppm) -12.92 -10. 92 to -14. 78 l Reactivity Computer Error 0.66% -4% to +4%

Source Range / Intermediate Range NIS Overlap (decades) 2.57 > 1.5 Reference Rod Bank Worth Error -2.3% -10% to +10% l All Other Banks Worth Error (max.) -6.8% -15% to +15% j Total Rod Bank Worth Error + 2.9% -10% to +10%  ;

r NOTE: pcm means percent millirho, equivalent to a reactivity value  !

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2.4 - 50% REACTOR POWER TEST SEOUENCE - ISU-240B OBJECTIVE I I

i The 50% Reactor Power Test Sequence document defines the activities l which constitute the startup testing program between 0% and 50% ,

power and at approximately 50% of rated thermal power. l This test partially satisfies activities described in FSAR Table 14.2-3, Sheets 20-22 and Section 14.2.10.4.

TEST METHODOLOGY This document ensures that the low power physics testing has been i completed and the results approved prior to increasing power. i Prior to increasing power for this test sequence, power range high I level trip setpoints are conservatively set to < 70% power.

The l 30% power baseline flux map results are also extrapolated to 70% )

power to ensure parameters indicative of DNBR and linear heat rate are acceptable up to the level afforded protection by the power range trip setpoints.

Plant operating procedures are utilized where appropriate to establish plant conditions and to change reactor power. During this testing sequence, following completion of 50% power testing '

and instrumentation tuning, power is stabilized near the 20-25%

range to accommodate testing at those power levels.

SUMMARY

OF RESULTS Results of individual tests completed up to and while at the 50%  ;

pcwer plateau are discussed primarily in Sections 3.2, 3.4 and 3.5 of this report. Flux maps were taken at 26.90% and 46.35% power with satisfactory results as summarized in Table 2.4-1. Prior to increasing power to 50%, the 26.90% power flux map results were also verified acceptable when extrapolated to 70% power, the value of the Power Range NIS trip setpoints. All required testing was completed.

Upon completion of this testing phase, the plant was aligned as directed by the Shift Supervisor.

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TABLE 2.4-1 30% POWER FLUX MAP RESULTS.

Actual Maximum Limit Reactor Power 26.90% 30%

Reaction Rate Error 6.25% 10%

FDHN 1.5311 1.8899 FQ(Z) 2.1767 4.64 x K(Z)*

Quadrant Power Tilt Ratios 1.0063 ' 1.0103 1.02 0.9779 l 1.0055  ;

1.0194 0.9741[,' 01 0.9964 '

50% POWER FLUX MAP RESULTS Actual Marimum Limit Reactor Power 46.35% N/A Reaction Rate Error 5.63% 10%

FDHN 1.4925 1.7995  ;

FQ(Z) 2.1396 4.64 x K(Z)*

Quadrant Power Tilt Ratios 1.3031 ! 1.0122 1.02 0.9800 l 1.0048 1.0181

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0.9742 ,? 1.0113  !

0.9964

• K(Z) is from the Unit 2, Cycle 1 Core Operating Limits Report, Figure 3

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I 2.5 - 75% REACTOR POWER TEST SEOUENCE - ISU-260B j OBJECTIVE The 75% Reactor Power Test Sequence document defines the activities l which constitute the startup testing program during escalation from  !

50% to 75% power and at approximately 75% of rated thermal power.  ;

This test partially satisfies activities described by FSAR. Table  !

14.2-3, Sheet 22 and Section 14.2.10.4.

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TEST METHODOLOGY i This document ensures that the 50% Reactor Power Test Sequence has been completed 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 are conservatively set to 195% power and reactor core flux map results from a 50% power map are verified acceptable. The flux map results are also extrapolated to 95% power to ensure parameters indicative of DNBR and linear heat rate are acceptable up to the level afforded protection by the power range trip setpoints. l Plant operating procedures are utilized where appropriate to establish plant conditions and to change reactor power.

SUMMARY

OF RESULTS Results of individual tests completed while at the 75% plateau are discussed primarily in Sections 3.2, 3.4 and 3.5 of this report.

A preliminary incore/excore cross-calibration had been performed prior to exceeding 50% power, where Technical Specification AFD -

limits began to apply. At 78% power, a full incore/excore cross-calibration was performed using a large axial xenon transient to establish appropriate core conditions. There were no significant changes made to the AFD calibrations with respect to the preliminary values. -

A steady-state flux map was taken at 78.30% power with satisf actory results as summarized in Table 2.5-1. Prior to increasing power to 75%, the 46.35% power flux map results from the 50% power test sequence were also verified acceptable when extrapolated to 95%

power, the value of the Power Range NIS trip setpoints. All required testing was satisfactorily completed.

During performance of this sequence, Unit 2 underwent a surveillance outage. This outage started on 5/20/93, following a forced unit shutdown unrelated to startup testing activities, and 75% power plateau testing was resumed on 7/10/93, a duration of 51 days.

Upon completion of this testing phase, the plant was aligned as directed by the Shift Supervisor.

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TABLE 2.5-1 75% POWER FLUX MAP RESULTS .,

Actual Maximum Limit Reactor Power 78.38% N/A. t Reaction Rate Error 5.11% 10%

FDHN 1.4408- 1.6509-1 FQ(Z) 2.0878 2.9630 x K( Z) * ~

Quadrant Power Tilt Ratios 1.0024 ! 1.0086 1.02 [

O.9821 l 1.0069 1.0138 [

0.9770 1.0123 0.9968

• K(Z) is from the Unit 2 Cycle 1 Core Operating Limits Report,  !

Figure 3  !

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2.6 - 100% REACTOR POWER TEST SEOUENCE'- ISU-280B ,

O'BJECTIVE The 100% Reactor Power Test Sequence document defines the activities which constitute the startup testing program during  ;

escalation from 75% to 100% power and at close to, but not more than, 100% of rated thermal power. This test partially satisfies activities described in FSAR Table -14.2-3, Sheet 22 and Section 14.2.10.4.

6 TEST METHODOLOGY l This document ensures that the 75% Reactor Power Test sequence has been completed and the results approved prior to increasing power above the 75% testing plateau. Prior to increasing power above 75% l for this test sequence, reactor core flux map results from a 75%  ;

power baseline map are verified acceptable and the power range high t level trip setpoints are set to i 109%, their normal Technical Specification values. The flux map results are also extrapolated to 100% power to ensure parameters indicative of.DNBR and linear heat rate are acceptable for power ascension to the 100% testing plateau.

Plant operating procedures are utilized where appropriate to i establish plant conditions and to change reactor power. During L ascension to the 100% plateau, power is stabilized near the 90% ,

level to accommodate a limited scope of testing at that power i level.

SUMMARY

OF RESULTS l l

Results of individual tests completed during this power ascension '

and while at the 100% plateau are discussed primarily in Sections '

3.2, 3.4 and 3.5 of this report.

A flux map was taken at 99.67% power with satisfactory results, as  ;

summarized in Table 2.6-1. The 78.3% power flux map results, from  ;

the 75% power test sequence, were also verified acceptable when ,

extrapolated to 100% power. All required testing was completed. I Upon completion of this testing phase, the plant was aligned as  :

dirccted by the Shift Supervisor.

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TABLE 2.6-1  ;

100% POWER FLUX MAP RESULTS Actual Maximum Limit Reactor Power 99.67% N/A i Reaction Rate Error 4.85% 10%  ;

FDHN 1.4489 1.5510  !

FQ(Z) 2.0178 2.328 x K(Z)*

Quadrant Power Tilt Ratios 1.0031 ! 1.0101 1.02 0.9821 l 1.0048 1.0135 .

0.9791 '

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/ 0.9953 l

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• K(Z) is from the Unit 2, Cycle 1 Core Operating Limits Report, e Figure 3 ,

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3.0 DISCUSSION OF THE INITIAL STARTUP TESTS TABLE 3.0-1 List of Test Summaries 3.1 CORE LOADING 3.1.1 Development and Implementation of the Reload Fuel Shuffle Sequence Plan, RFO-106 3.1.2 Core Loading Instrumentation and Neutron Source Checks, ISU-003B 3.1.3 Inverse Count Rate Ratio Monitoring (Core Load Portion), NUC-111 3.1.4 RCS and Secondary Coolant Chemistry (Core Load Portion), ISU-006B 3.1.5 Verification of Core Loading Pattern, RFO-204 3.2 SYSTEM TESTING AFTER CORE LOAD AND AT VARIOUS POWER LEVELS 3.2.1 Piping Vibration Monitoring, ISU-212B 3.2.2 Steam Generator Level Control Test, ISU-207B 3.2.3 Thermal Expansion, Power Ascension Phase, ISU-308B 3.2.4 Incore Moveable Detector System Alignment, ISU-016B 3.2.5 RCS and Secondary Coolant Chemistry (Post Core Load), ISU-006B 3.2.6 Radiation Survey Tests, ISU-208B 3.2.7 Containment & Penetration Rooms Temperature Survey, ISU-282B 3.2.8 Reactor Coolant Flow Measurement, ISU-023B 3.2.9 Reactor Coolant System Flow Coa';tdown Test, ISU-024B 3.2.10 Reactor Coolant System Leakag'; Rate Test, ISU-022B 3.2.11 Cold Control Rod Operability testing, ISU-026B 3.2.12 Hot Control Rod Operability Jesting, ISU-027B 3.2.13 Reactor Trip System Tests, ISU-015B 3.2.14 Pressurizer Spray and Heater Capability, ISU-021B 3.2.15 Dynamic Automatic Steam Dump Control, ISU-205B 3.2.16 Turbine Generator Initial Synchronization and Overspeed Test, ISU-220B 3.2.17 Main Feedwater System Test, ISU-238B 3.3 PHYSICS TESTING 3.3.1 Inverse Count Rate Ratio Monitoring (Initial Criticality Portion), NUC-111 3.3.2 Initial Criticality, NUC-106B 3.3.3 Determination of Core Power Range for Physics Testing, NUC-109 3.3.4 ABB/CE Reactivity Monitoring System Checkout, NUC-210 3.3.5 Core Reactivity Balance, NUC-205

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i TABLE 3.0-1 (Continued) 3.3.6 Zero Power Isothermal and Moderator Temperature Coefficient Measurements, NUC-207 3.3.7 Rod' Swap Measurements, NUC-120 l 3.3.8 Boron Endpoint Determination and Differential Boron Worth, NUC-104 ,

3.4 TRANSIENT TESTING 3.4.1 Turbine Generator - Trip With Coincident Loss of 4 Offsite Power, ISU-222B 3.4.2 Design Load Swing Tests, ISU-231B 3.4.3 Dynamic Response to Full Load Rejection and Turbine ,

Trip, ISU-284B 3.4.4 Remote Shutdown Capability Test, ISU-223B '

3.4.5 Remote Shutdown Capability Test at Power, ISU-225B 3.4.5 Large Load Reduction Tests, ISU-263B i 3.( INSTRUMENTATION AND CALIBRATION TESTING 3.5.1 Calibration of Feedwater and Steam Flow  :

Instrumentation at Power, ISU-202B 3.5.2 Operational Alignaent of Process Temperature and N16 l Instrumentation, ISU-226B 3.5.3 Operational Alignment of Nuclear Instrumentation, .

ISU-204B 3.5.4 Incore/Excore Detector Calibration, NUC-203 3.5.5 Loose Parts Monitoring Baseline Data, ISU-211B .

3.5.6 Startup Adjustments of Reactor Control Systems, ISU-020B .

3.5.7 Full Power Performance Test, ISU-281B '

3.5.8 Plant Computer Software Verification, ISU-019B 3.5.9 Automatic Reactor Control System Test, ISU-203B 3.5.10 Precision Secondary Side Power -Calorimetric, [

PPT-P2-2050 3.6 DEFERRED PREOPERATIONAL TESTING r 3.6.1 Public Address and Emergency Evacuation Alarm

- System Test, PPT-TP-92B-1 3.6.2 Secondary Sampling Deferred Preop Test, PPT-TP-93B-13 3.6.3 Heating, Ventilation and Air Conditioning System Air Balance, X-ME-14

l

\

TABLE 3.0-1 (Continued)  !

I I

3.6.4 Main Steam Isolation Valve Test, 2HV-2334A, 1 P9T-52-9501A ,

3.6.5 Steam Dump Valve Testing, PPT-P2-2023 l 3.6.6 Thermal Expansion" Test of Extraction Steam to i Auxiliary Steam System, ISU-308B ,

3.6.7 Reactor Cavity Humidity Detectors, INC-4088B l 3.6.8 Pressurizer Power Operated Relief Valve Leak l l

Tightness, ISU-021B 3.6.9 Pressurizer Spray Valve Leakage, ISU-021B i 3.6.10 Chemical and Volume Control System (CVCS) Mixed Bed  !

Demineralizer 2-01 Pressure Drop, 4 1

WO #1-93-034962-00 3.6.11 Plant Computer Software Verification l

l I

b i

)

3 .1 - CORE LOADING 3.1.1 - Develooment and Imolementation of the Reload Fuel Shuffle Secuence Plan, RFO-106 OBJECTIVES This permanent plant procedure is performed to ensure that the nuclear fuel assemblies are loaded in a safe and cautious manner.

This procedure partially satisfies activities described in FSAR Section 14.2.10.1.

TEST METHODOLOGY The procedure is performed prior to the start of core loading to develop the detailed core loading sequer.ce sheets. Field use of the procedure begins with the loading of the temporary core loading instrumentation into its initial position and determination of background count rates for all source range and temporary nuclear instrumentation channels. The two primary source bearing assemblies and seven additional assemblies, comprising the " source nucleus", are. loaded. Audible indication of neutron population changes from one of the two installed source range plant channels is required to be maintained in both the control room and containment for the duration of the core loading process. After the source nucleus assemblies are loaded, count rate data is taken for the nuclear channels used in the core loading process (two

• source range and three temporary channels). The firnt reference value, for use in inverse count rate ratio monitoring, is determined from these counts after the appropriate background values have been subtracted. Subsequent reference values are calculated whenever core loading is suspended for eight hours or longer, a temporary detector is moved, or a primary source bearing fuel assembly is moved to a different core location.

Prior to fuel load, predictions were made for comparison to actual nuclear instrumentation response, to verify that the reactor would remain shutdown throughout the loading process. Inverse count rate ratio monitoring is used following each fuel assembly move to ensure that the reactor is not approaching criticality. To ensure reliability in the monitoring, a minimum of two of the five nuclear instrumentation channels are required to be responding to source neutron population changes throughout core loading. Data obtained during inverse count rate ratio monitoring is trended and extrapolated forward to permit evaluation of any indicated  !

criticality approach. Plant procedure NUC-111 is used to perform ,

the inverse count rate ratio measurements and extrapolations.

3.1.1 - Develcoment and Imolementation of the Reload Fuel Shuffle Secuence Plan, RFO-106 (Continued)

SUMMARY

OF RESULTS Core loading was completed in a safe and cautious manner as required by the acceptance criteria of the core loading procedure.

All 193 fuel assemblies were loaded in the core without incident. -

The fuel handling equipment performed well, with only one significant delay. . Core alterations were suspended for ,

approximately 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> due to problems encountered with the refueling machine controls. A vendor representative replaced the control circtit cards in the refueling machine to restore operability. ,

Refer to Figures 3.1.1-1 through 3.1.1-5 for a graphical description of how core loading progressed.

+-

l' l

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Figure 3.1.1-1 I l

UNIT 2 CORE LOADING PATTERN  ;

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Figure 3.1.1-5 UNIT 2 CORE LOADING PATTEPb FINAL CONFIGURATION i 37

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t i

t 1,1. 2 - CORE LOADING INSTRUMENTATION AND NEUTRON SOURCE l CHECKS - ISU-003B OBJECTIVE l The core loading instrumentation test is performed prior to core l loading to determine the proper operating and discriminator voltage i settings for the temporary core loading instrumentation and-to l

' verify that both the temporary and permanent nuclear monitoring. (

instrument channels respond properly to a neutron source. The test.  ;

is also performed to verify that both the temporary and permanent  ;

nuclear monitoring instrument channels respond properly to neutrons j prior to resuming core loading following any eight hour or longer I delay in loading. This test satisfies testing described by FSAR Table 14.2-3, Sheet 8 and Section 14.2.10.1. ,

TEST METHODOLOGY Following the initial installation of the equipmen*;, 'he c temporary \

detectors are positioned near a neutron source. Using the neutron >

source, an optimum. operating voltage is 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 i bias voltages are determined based on detector characteristic ,

curves. ,

Prior to core loading, all five channels (two installed source  :

range and three temporary core loading channels) are neutron '

response checked by moving a portable neutron source toward and away from each detector to verify detector response. l l

In the event of a delay in core loading of 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> or greater, this ,

test reverifies proper. detector neutron response by one of three  !

methods. One method uses a portable neutron source moved toward i and then away from a detector to verify detector response. The j second method is to use movement of an installed fuel assembly to  :

alter neutron flux at a detector by altering source to detector l neutronic coupling. The third method uses an evaluation of  !

counting statistics applied to detector output when in proximity to 4

a fixed neutron source. Nuclear decay is a random process and if i the detector output exhibits statistical behavior (standard i deviation, etc.) characteristic of a random process, then the-detector is judged to be responding to neutrons instead of 60 Hz or l other noise.  ;

r i

3.1.2 - CORE LOADING INSTRUMENTATION AND NEUTRON SOURCE CHECKS - ISU-003B (Continued)

SUMMARY

OF RESULTS Upon completion of the initial alignment procedure, operating voltages were determined to be 2050 volts for all three temporary nuclear instrumentation channels with discriminator bias voltages set at 3.0 volts for all three channels. Three additional detector tubes were also tested for use as spares, as necessary. All three spare tubes also had operating voltages of 2050 volts and 3.0 volt discriminator bias voltages. None of the spare tubes were used during core loading.

Prior to core loading, all five channels (2 installed source range and 3 temporary core loading channels) were neutron response checked using a portable neutron source. Adequate responses to the neutron source were observed when the source was placed near the various detectors.

The 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> delay portion of this test procedure was executed only once during the core loading activity. The response check was done satisf actorily using the statistical analysis method. No fuel assemblies were moved to verify detector neutron responses.

As allowed by the Initial Fuel Load Sequence - ISU-001B, the three temporary detectors were removed after 60 assemblies were loaded.

At this stage of the fuel loading activity, all loaded assemblies were neutronically coupled with respect to the two permanently installed source range detectors.

i t

3.1.3 - INVERSE COUNT RATE RATIO MONITORING,(Core Load Portion)

NUC-111 ,

OBJECTI'7E  ;

This pemanent plant procedure is performed to obtain and evaluate i nuclear monitoring data during core loading to ensure that core loading is done in a cautious and controlled manner. Thic procedure satisfies activities described in FSAR Section 14.2.10.1.

TEST METHODOLOGY  ;

Neutron count rate data from both installed source range NIS  !

channels and three temporary core load instrument channels is taken following each fuel assembly addition. The sources of the core neutron flux are the two installed Californium primary neutron .

sources with associated subcritical multiplication due to the loaded fuel lattice. As fuel is loaded, the core neutron flux  ;

changes due to changes in fuel lattice geometry and the addition of l uranium to the core. j To determine the effect of a single fuel assembly addition on core reactivity, count rate data after each fuel assembly is loaded is compared to a reference value to evaluate the effect of the additional fuel assembly. This comparison is performed as a ratio -

of the count rates to evaluate the fractional change. If this 2

ratio were to be very large, it would indicate that this fuel assembly addition brought the loaded fuel lattice significantly closer to criticality. For convenience, the procedure evaluates d the inverse of the count rate ratios (ICRR) such that an approach j to zero would indicate an approach to criticality. Additionally, this procedure trends the inverse count rate ratios and -

extrapolates the trends to evaluate whether or not additional fuel assembly loadings would be expected to result in an approach to i criticality.

Prior to the start of core loading, background counts are taken to j allow the elimination of general background radiation from the  !

calculations of ICRR values. Reference count rate data is taken initially after the first ten fuel assemblies are loaded. Eight of  ;

these fuel assemblies are loaded together to constitute a-" source 1 nucleus", providing a suberitical multiplied flux capable of being I used as a basis for meaningful comparisons. Reference values are l redetermined if a neutron source bearing fuel assembly is moved or l if a temporary detector is moved, both which cause a change in i source to fuel to detector geometry. New reference values are also obtained if neutron counting channel equipment or electronics settings are changed, to ensure that a valid reference value for count rate comparison is used. As a conservative measure, new reference count rate values are determined if core loading is delayed by 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> or more to ensure that any count rate changes l

3.1.3 - INVERSE COUNT RATE RATIO MONITORING,(Core Load Portion)

NUC-111 (Continued)

TEST METHODOLOGY (Continuedi over time are accounted for. Inverse count rate ratio data taking, calculations, plotting, trend evaluation and extrapolation are also repeated hourly during any core loading delay for general core monitoring and to aid in detection of any inadvertent RCS dilutions.

At all times a minimum of two selected channels of instrumentation are designated as " responding channels" . This designation is based on source to fuel to detector geometry considerations so as to avoid large local effects that may not be indicative of total core behavior.

Final reference count rate data is taken following the completion of core loading for use as baseline data to help verify source range NIS signal to noise ratio.

SUMMARY

OF RESULTS All count rate data was properly recorded and ICRRs were calculated, plotted, trended and extrapolated. Refer to Figures 3.1.3-1 and 3.1.3-2 for a graphic display of crocedure results during core loading. The inverse count rate ratie shows that core loading was performed in a cautious and controlled manner with no indicated unexpected approaches toward criticality. At no time did the extrapolated data from a responding channel indicate that criticality would be expected to occur with the loading of the next fuel assembly. Large changes in the inverse count rate ratios from one core loading step to the next step are due primarily to local geometric effects when a neutron source vas moved near a detector or when fuel was loaded between a source and a detector resulting in a large local countrate increase due to enhanced neutronic source to detector coupling via suberiticat cultiplication. These were local effects observed on only one or two channels at a time.

Monitoring data was properly taken and evaluated during core loading delays and reference count rates were properly recalculated. The background count rates were sufficiently low so as to be nearly negligible, also an indication of low neutron detector channel noise. Final reference count data was taken for both source range channels.

Figure 3.1.3-1 ICRR vs. FUEL ASSEMBLIES LOADED i.2 - -

• • f""*'"*""*'""""*"' -- - -- - -

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0.2 - - - -

0.1 50 70 90 13 0 150 70 19 0 10 30 11 0 Number of Fuel Assemblies Loaded N31 i N32

Figure 3.1.3-2 ICRR vs. FUEL ASSEMBLIES LOADED Unit 2 Temporary Detectors --

1.2 - - - - - - -

I l.1 -

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0.9 - - -

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0 10 30 50 70 90 11 0 130 150 17 0 190 Number of Fuel Assemblies Loaded Detector A i Detector C

~"~ .

i i

3.1.4 - RCS AND SECONDARY COOLANT CHEMISTRY (Core Load Portion) -

ISU-006B OBJECTIVE i

This. test is performed to verify correct and uniform boron  !

concentrations in portions of the reactor coolant system (RCS) and l the directly connect.ed portions of fluid systems as required for j core loading.- This test is also designed to help ensure that the  ;

possibility of an inadvertent dilution of the RCS during ccre j loading is minimized. This test satisfies activities Vascribad-in i FSAR Section 14.2.10.1. l TEST METHODOLOGY Prior to the commencement of core loading, the RCS is sampled and verified .to meet specified water chemistry criteria. As a .

prerequisite to RCS chemistry sampling, the borated water source,  !

the RCS loops, Chemical and Volume Control System piping, Safety i Injection System piping, and containment Spray System piping.were verified to have boron concentrations which would preclude inadvertent RCS dilutions. l Each of the RCS crostaver legs, the Residual Heat Removal (RHR) l system, the reactor wssel, the Volume Control Tank, the Safety Injection System accumulators, the boric acid tanks, and the i Refueling Water Storage Tank are sampled, and that water is- ._

verified to contain specified boron concentrations.

I Following the initial verification of the chemistry in the reactor coolant system, four samples are taken from the reactor vessel at equidistant depths along with a sample from the operating residual 'j heat removal-train. These samples are then analyzed for boron to verify a uniform boron concentration between the RCS and the RHR '

system (within a 30 ppm range). After the RCS and RHR is verified j to be at a uniform concentration, the operating residual heat  ;

renoval train is sampled and analyzed for boron to verify that the  ;

. water remains at > 2000 ppm boron.

Sampling continues every 12- >

1 hours1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> until the start of core loading. With the start of core  :

loading sampling continues on the operating RHR train every four hours throughout the core loading process. The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> ,

samples also include measurement of RHR inlet- temperature for use ,

in monitoring reactor coolant system temperature for compliance with Technical Specifications and to ensure temperature changes do l not adversely influence inverse count rate ratio monitoring.- The  !

spent fuel pool is dry during initial core load and the fuel i transfer system canal portions are not required to be borated. i i,

s 9 , - , .-,-,v. . ,-- --

+- r -

3.1.4 - RCS AND SECONDARY COOLANT CHEMISTRY (Core Load Portion) -

ISU-006B (Continued)

TEST METHODOLOGY - (Continued)

The criterion for the 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> samples is to ensure a minimum of 2000 ppm for shutdown margin and a maximum of 2150 ppm to not overly attenuate the neutron detector signals during core loading. Also, consecutive samples are not to differ by more than 20 ppm as a way of detecting any inadvertent dilution.

SUMMARY

OF RESULTS During the execution of this test which started before and lasted throughout the core loading process, all acceptance criteria were met for each system that was sampled. No corrective actions in the core loading process were needed to meet the acceptance criteria of this test. Detailed results obtained prior to core load are tabulated below:

Specified Actual Location Rance (com) Value (com)

Volume Control Tank 2000 - 2150 2021 RHR Train A 2000 - 2150 2023 RHR Train B 2000 - 2150 2016 Refueling Water Storage Tank 2000 - 2200 2081 Boric Acid Tank #1 > 7000 7539 Boric Acid Tank #2 > 7000 7327 Safety Injection Accumulator 1 1900 - 2200 2092 Safety Injection Accumulator 2 1900 - 2200 1981 Safety Injection Accumulator 3 1900 - 2200 2101 Safety Injection Accumulator 4 1900 - 2200 2106 RCS Loop 1 Crossover Leg > 2000 2063 RCS Loop 2 Crossover Leg > 2000 2061 RCS Loop 3 Crossover Leg > 2000 2031 RCS Loop 4 Crossover Leg > 2000 2060 Reactor Vessel Surface Within a 2038 Reactor Vessel 1/3 down 30 ppm 2048 Reactor Vessel 2/3 down range 2048 Reactor Vessel Bottom 2041 RHR Train B 2046 RCS/RHR uniformity values were within a 10 ppm range, well within the 30 ppm limit.  ;

The RHR samples prior to and during core loading varied within a range of 2031 ppm to 2060 ppm, well within the 2000 -

2150 ppm range limits. During core loading, no two consecutive samples deviated by more than 10 ppm. This satisfied the (20 ppm difference limit. All samples were from RHR Train B. During core  !

loading, RHR inlet temperature varied from an initial (and l maximum)value of 114.0*F to a minimum value of 104.6*F. l l

l

A - a a a-- A --< >An+ a a,-s----m .-x s46w 3.1.5 - VERIFICATION OF CORE LOADING PATTERN - RFO-204 OBJECTIVE This permanent plant procedure is performed to confirm that the loaded core matches the design loading pattern and to provide a videotape record of the as-loaded core.

TEST METHODOLOGY Using the manipulator crane (refueling machine) television camera ,

mast, an underwater TV camera is slowly traversed over.the entire loaded core allowing fuel assembly. and insert numbers, positions-and orientations to be observed on a TV monitor. This information is compared against the core loading pattern design information from the fuel vendor. . The TV signal is also . sent to a video recorder so a tape record of the as-loaded' core pattern is made.

The use of the camera within the reactor vessel constitutes a core alteration, so all required Mode 6 core alteration related Technical. Specifications are also verified by this procedure to have been satisfied.

SUMMARY

OF RESULTS The entire core was mapped using the underwater TV camera. All i fuel assemblies and inserts were found to be in their proper locations and orientations. A videotape record was made and reviewed to ensure that it was legible.

r

- 3. 2 SYSTEM TESTING AFTER CORE LOAD AND AT VARIOUS POWER LEVELS 3.2.1 - PIPING VIBRATION MONITORING - ISU-212B 1

OBJECTIVE This . test demonstrates that steady state flow induced' piping

. vibrations and transient response piping vibrations are within allowable design limits. The scope of the test is limited to portions of the Main Steam and Feedwater systems. These are systems which could not be fully tested.during-the Preoperational:

Test. Program due to plant conditions. This test partially satisfies the testing described by FSAR Table 14.2-2, Sheet 57 and Sections 3.9B.2.1.2 through 3.9B.2.1.4.

TEST METHODOLOGY The Main Steam and Feedwater systems are operated under normal, steady state conditions during which visual inspections of.the piping are conducted. Portable vibration analyzers are used.to obtain numerical values for selected vibration levels and comparisons. are made between the vibration velocities or displacements and the appropriate limits. The steady state testing of various subsystems is performed between 3% and 13%' reactor power and at 100% reactor power.

The transient response portion of the test combines data taken from selected remotely instrumented portions of the Main Steam and l

Feedwater systems during the imposed transient with concurrent

! visual observations of accessible piping system portions. Portions

of the Main Steam system are tested in response to a full power main turbine trip with portions of the Feedwater system tested in response to Main Feedwater Pump trip.

SUMMARY

OF RESULTS Steady-state vibration visual and velocity criterion were satisfied for all but thirteen measurement locations. One measurement was on i the Feedwater system recirculation piping to the condenser, eight l measurements were on the Main Steam turbine supply piping, and four  ;

measurements were on Main Steam to the Moisture Separator Reheater  ;

(MSR) piping. All measurements were found acceptable by l engineering calculation.

]

• FW Recirc to Condenser: The resultant test velocity of 0.45 inches /sec. zero peak was less than the calculated allowable velocity of 5.529 inches /sec. zero-peak.

i i

3.2.1~- PIPING VIBRATION MONITORING - ISU-212B (Continued)

SUMMARY

OF RESULTS (Continued)  ;

• MS Turbine Supply /MS to MSRs: The resultant test displacement  ;

of 0.088 inches for Drain Pots 2MS-03, 2MS-04, 2MS-05, 2MS-06 and 2MS-07 blowoff- lines was less than the calculated i allowable displacement of O'.115 inches. The resultant. test ,

displacements of 0.114 inches for Drain Pot 2MS-01 blowoff  !

line, 0.103 inches for Drain Pot 2MS-02 blowoff line and 0.111 i for Drain Pot 2MS-08 blowoff line were less than the ,

calculated allowable displacement of 0.115 inches. These calculated allowable displacements enveloped the measured displacements for Drain Pots 2MS-01, 2MS-02, 2MS-03 and 2MS-04 level sensing lines. J

• Transient vibrati,n visual test criterion were satisfied for all observed locat. ns. j

• Transient vibration instrumented test criterion were satisfied ,

at all but one measurement location. Feedwater system snubber  !

FW-2-095-415-C62K sustained a load of 834 lbs. The expected  ;

load was 569 lbs. Engineering evaluated and dispositioned <

this as acceptable based on the snubber upset design load of {

2056 lbs. One sensor on Feedwater (FW-2-019-405-C52K) and one sensor on Main Steam (MS-2-003-410-C72K)_ failed to function.  ;

Engineering evaluated and dispositioned the lack of two data .!

channels as acceptable based on review of data from sixteen  ;

functional Main Steam channels and twenty-three Feedwater  ;

channels. ,

i I

l 3.2.1 - PIPING VIBRATION MONITORING - ISU-212A (Continued)

SUMMARY

OF RESULTS (Continued)

Allowed Measured  ;

Pipe Line Velocity Velocity Location Number (inches /seci (inches /sec) 1A Miniflow 12FW-1-21-2002G 50.5 2.3 1A Miniflow 12FW-1-26-2002G 10.5 0.95 1B Miniflow 2FW-1-22-2002G 10.5 18 (2" Drain Line)

Allowed Actual Actual Displacement Displacement Displacement Location Ratio Ratio (inches)  ;

1A Miniflow 51.0 2.48 0.10 1A Miniflow 11.0 1.46 0.04 1B Miniflow 11.0 1.13 0.28 (2" Drain Line)

For the transient response portions of the test, there were no discrepancies noted with regard to the Main Feedwater Pump trip transient. There were two items noted in connection with the Full Power Turbine Trip transient. One instrumented snubber, MS-1-002-009-C72K (location TR-1-MS-25), exceeded its allowed loading criterion. Engineering evaluated and dispositioned this as acceptable because while the expected loading of 9253 lbs. was exceeded by 16% (10720 lbs), the support had available design margin. Even though the actual transient loading exceeded the ,

expected loading by 16%, when the transient loading is combined with predicted seismic loading the total load change is only an increase of 317 lbs, 0.6%. This is well within the 15% of total load snubber design margin available. A separate calculation indicated that the piping stresses in this area were 516750 psi which lies well within the 121000 psi allowable range.

Additionally, two remote sensors, at locations TR-1-MS-02 and TR MS-03, failed to function during the transient. Engineering evaluated and dispositioned this missing data as acceptable based on the data obtained from 17 other, functioning main steam system sensors.

Other than the above noted items, the remaining piping system portions all had vibration levels within the acceptable range.

3 i

3.'2.2 - STEAM GENERATOR LEVEL CONTROL TEST - ISU-207B ,

OBJECTIVE This test is performed to demonstrate steam generator level control-stability _ throughout power ascension. . Changing . feedwater flow .

configurations and major power changes necessitate the need for  ;

multiple performances of this test. Level control stability of the '

four steam generators is - demonstrated while operating on - the -

feedwater bypass control valves and the main feedwater control  ;

valves. <

TEST METHODOLOGY In order to verify level control stability while operating on the bypass or main feedwater control valves, a 5% level deviation is manually established in each steam generator. The control system is then transferred to the automatic control position. Steam ,

generators are tested sequentially, one at a time, _not simultaneously. The actual steam generator level is monitored to determine ' overshoot , undershoot and whether or not level returns to-and remains within the allowed band of 64.0 2% of narrow range ,

level within a specified time frame of 3 times the appropriate reset time. constant. The bypass valves are tested at approximately .

5% power. The main feedwater control valves are tested at  ;

approximately 50% power.

In order to verify correct main feedwater pressure and system j response during small transients on the main feedwater pump master controller, main feedwater header pressure is manually changed by l 10 to 25 psi. The pump controller is then transferred to the ,

automatic position and system. response is monitored. This testing  !

is done at approximate power levels of 5% and 50%. }

In order to verify level control stability while transferring between the feedwater bypass control valves and the main feedwater control valves, steam generator levels are monitored while performing this transfer at approximately 20% power.

At approximately 50%, 75% and 100% power, data is taken to verify .i expected main feedwater control valve positions, to verify proper l feedwater pump speed control operation on its sliding Ap program and to verify non-excessive feedwater header pressure oscillations.

At 75% and 100% power, data is also taken to verify proper steady state level control operation. f I

?

. . . ,. ... . = . . . ~ -. . .. . . . . . - -

i 3.2.2 - SThM GENERATOR LEVEL CONTROL TEST - ISU-207B (Continuedi  !

SUMMARY

OF RESULTS 8

. Refer to Table-3.2.2-1 for detailed-test results. l When given a 5% level deviation-(high or low), the bypass control  !

valves returned steam generator level to and remained within the-  ;

programmed level, 12%, within three time constants:with less-than- '

4% overshoot or undershoot, as expected. This was done at i approximately 5% power.

When given a 5% level deviation (high or low), the main feedwater.

control valves returned the steam generator leve1~to and remained- .;

within the programmed level 2% within three time constants with-  :

less than 4% overshoots or undershoots as expected. This was'done 1 at approximately'50% power. >

After transferring from the feedwater bypass control valves to the main control valves, steam generator level deviations were to return to and remain within 12.0% of the programmed level within l

three time constants. This was done at approximately 20% power..

t The feedwater header pressure oscillations were less than 3% of  !

operating pressure range at approximately 5%, . 50%, 75% and 100% l power.

j At approximately 75% and 100% power, all steam' generator steady i state levels were verified to remain within the expected 64% 12% l operating band..

j At approximately 5%, 50%, 75% and 100% power the sliding Ap program f value, used to control main feedwater pump speed, was verified to l be within 125psig of the actual Ap value At approximately 50%, 75% and 100% power, all feedwater control valves were verified to be within 110% of their. predicted positions. ,

A problem with feedwater pump speed control was noted and corrected ,

as a result of performance of this test. At approximately 5% ,

power, Main Feedwater Pump Speed Controller 2-SK-509A failed to i control in the_ automatic mode. Upon correction of the circuitry i problems, the controller functioned properly.

I t

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t TABLE 3.2.2-1  ;

STEAM GENERATOR LEVEL CONTROL

SUMMARY

BYPASS CONTROL VALVE-LEVEL CONTROL RESPONSE PERFORMED AT APPROXIMATELY 5% POWER '

MAX- MAX-ACTUAL OVERSHOOT / IMUM IMUM LEVEL ACCEPTANCE TIME UNDERSHOOT OVER- UNDER- :l STEAM DEVI- CRITERION RESPONSE LIMIT IN SHOOT SHOOT i GENERATOR ATION IN MINUTES IN MINUTES PERCENT PERCENT PERCENT 1 5% up 537.5 3.4 <4.0 1.5 0 5% down 137.5 3.0 <4.0 0 1.5 2 5% up 137.5 2.8 <4.0 1.75 0 [

5% down 137.5 2.9 <4.0 .0.2 1.0 3 5% up 537.5 3.4 <4.0 1.5 0 ,

5% down. <37.5 2.9 <4.0 0 1.5 4 5% up 537.5 3.2 <4.0 1.75 0 5% down 137.5 4.4 <4.0 0 1.0 LEVEL CONTROL RESPONSE AFTER BYPASS TO MAIN FEEDWATER CONTROL VALVE TRANSFER AT APPROXIMATELY 20% POWER ACCEPTANCE FINAL STEAM ACCEPTANCE ACTUAL TIME STEAM CRITERION GENERATOR CRITERION IN RESPONSE IN

._ GENERATOR IN LEVEL (%) LEVEL (%) MIN,UTES MINUTES 1 62-66 65 <37.5 0 2 62-66 65 <37.5 0 3 62-66 64 <37.5 0 4 62-66 63 <37.5 0 FW HEADER PRESSURE OSCILLATIONS Power Plateau Maximum Pressure Oscillation (esic/%)

Allowed Limit <45/3.0 5% 21/1.4 50%/ Pump A 13/0.9 50%/ Pump B 8/0.5 75% 6/0.4 100% 8/0.5

I I

TABLE 3.2.2-1 (CONTINUED)

FEEDWATER AP PROGRAM COMPARISON Power Plateau Maximum $P Deviation (esi)

Allowed Limit <25 5% 0 50%/ Pump A 3.91 50%/ Pump B 3.95 75% 6.35 100%

---

9.78 ------------..---------

MAIN FEEDWATER CONTROL VALVE LEVEL CONTROL RESPONSE AT APPROXIMATELY 50% POWER MAX- MAX-ACTUAL OVERSHOOT / IMUM IMUM LEVEL ACCEPTANCE TIME UNDERSHOOT OVER-STEAM DEVI- CRITERION UNDER-RESPONSE LIMIT IN SHOOT SHOOT GENERATOR ATION IN MINUTES IN MINUTBS PERCENT PERCENT PERCENT 1 5% up 537.5 7.2 <4.0 1.0 0 5% down 537.5 8.2 <4.0 0 1.5 2 5% up 137.5 6.5 <4.0 1.3 0 5% down 137.5 6.1 <4.0 0 1.3 3 5% up 537.5 7.7 <4.0 0.8 0 5% down 137.5 6.9 <4.0 0 1.3 4 5% up 137.5 8.4 <4.0 1.2 0 5% down 137.5

---

6.5 <4.0 0.5 0.7 FEEDWATER CONTROL VALVE POSITIONS AT VARIOUS POWER LEVELS ACTUAL VALVE POSITIONS IN %

Power Level Precicted Predicted 2-FCV-510 / Rance in % 2-FCV-520 / Rance in %

50%/ Pump A 46 41-61 50%/ Pump B 44 41-61 46 41-61 44 75% 41-61 58 50-70 59 100% 50-70 76 65-85 81 65-85 Power Level Predicted Predicted 2-FCV-530 / Rance in % 2-FCV-540 / Rance in %

50%/ Pump A 46 41-61 50%/ Pump B 42 41-61 46 41-61 42 75% 41-61 60 50-70 54 50-70 100% 81 65-85 75 65-85

'3.2.3 - THERMAT. WPANSION - POWER ASCENSION PHASE -'ISU-308B OBJECTIVE j Thermal expansion testing of plant systems is conducted to -

demonstrate that specified system piping experiences thermal expansion consistent with design, and that specified system support ,

components do not interfere with pipe thermal growth. This test covered portions of the plant that could not-be tested during the l Preoperational Test Program due to plant conditions. This test t satisfies activities described in FSAR Table 14.2-2, Sheets 52 and 52a and in FSAR Section 3.9B.2.1.1.

TEST METHODOLOGY l At ambient and hot conditions, system walkdowns are performed.

• Both the NSSS and selected secondary plant systems are evaluated.

Piping and components are visually observed, and selected snubber- '

and spring hanger positions recorded. Pipe whip restraints are verified not to interfere with the piping. Interferences are ';

identified and dispositioned by the design engineers. When necessary, system walkdowns are again conducted following the resolution of interferences. All piping movements are evaluated by  :

the design engineers. Selected locations are remotely instrumented ,

to measure piping movements for ALARA, safety and accessibility reasons. The walkdowns and remote data collection are performed at  ;

ambient temperatures (at or below 120*F) and at approximately 15%, 'j 30% and 100% reactor power.

SUMMARY

OF RESULTS i system walkdowns are conducted and instrumented measurements recorded to verify the following attributes

• The piping systems are free to expand thermally without l restriction other than by design. l

• Spring hanger movements remain within their working range and snubbers have not become fully extended or retracted. t

• Pipe whip restraints do not .c '#hre with the free j thermal movement of the piping.

• The measured thermal movement data shall be within t 1/4 l in. or ! 25% of the analytical predicted value, whichever is greater, or reconciled by engineering. ,

j l

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t 3.2.3 - THERMAL FYDANSION - POWER ASCENSION-PNARE - ISU-308B-(Continued)

SUMMARY

OF RESULTS (Continued)  ;

At ambient conditions, the following - types of problems were-identified and evaluated: i Piping movements restricted' by piping - or insulation contact

• Spring can outside of working range but not bottomed out' '

at 15% and 30% Reactor Power, the following types of problems were identified and evaluated: i e Piping movements restricted by piping or insulation contact

• Spring cans bottomed out .;

• Measured movements out-of-tolerance  :

at 100% Reactor Power, the following types of problems were '

identified and evaluated: i

• Piping movements restricted by piping or insulation  !

contact '

• Spring cans bottomed.out i e Measured movements out-of-tolerance

{

Heater Drain system transient damage included bent or I missing hanger rods and insulation damage. ,

Identified problems or out-of-tolerance conditions were evaluated l and dispositioned by engineering as acceptable, or corrective actions were implemented by work order or design modification. ,

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-r . , . . . , . - - . , .- -. _

3.2.4 - INCORE MOVEABLE DETECTOR SYSTEM ALIGNMENT - ISU-016B OBJECTIVE The purpose ~ of this procedure is to - demonstrate the- proper operation of the flux mapping system, including the-leak detection system. In addition, top and bottom of core limits are set and the actual drive cables and detectors are installed and verified to.  !

function properly. It also verifies the capability of the system to supply the appropriate digital ~and analog signals to the Plant {

Computer. This test satisfies activities described by FSAR Table 14.2-3, Sheets 34 and 35.

a TEST METHODOLOGY ,

Using a dummy drive cable, the top and bottom of core limits are  !

established for normal, emergency, calibrate and common modes and for the storage mode endpoint and insert limits by slowly driving the dummy detector to the top-of the core (or storage position) ..

where clutch slippage is observed. The position was then recorded ,

from the encoder display. The top limit is obtained by subtracting two inches from the recorded position and the bottom 11imit is obtained by subtracting 170 inches from the top limit. Storage mode insert limit is the endpoint minus 36 inches. Drive speed is measured by timing cable motion over a given distance to verify:the design speed of 14412 inches / minute. - The leak-detection system is . .i tested by filling the drain header with domineralized water and allowing the ' leak detection level switch to actuate, thereby  ;

draining the water and alarming. The withdraw and safety limit i switches are verified to prevent the detector from being taken up .

onto the reel. All push-to-test lights are verified. Simulated 1 signal transmissions to the Plant Computer and from the incore

~

system are made to verify proper computer data logging from the incore system. .

SUMMARY

OF RESULTS Figure 3.2.4-1 displays the Moveable Incore Detector Path -

Locations. Proper operation of all indicating lights were verified .i along with the proper operation of the leak detection system and ,

alarm as described in the previous section. The dummy detector was  ;

successfully inserted into all 58 core locations with proper drive speeds verified (see the table below). All top and bottom limits l were_ properly established. The limit switches were demonstrated  :

operable. The simulated data transmissions verified the ability of  !

the Plant Computer to receive signals from the incore flux mapping system and the ability of the incore system to~ supply proper ^

signals to the Plant Computer.

r r

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3.2.4 - INCORE MOVFARLE DETECTOR SYSTEM ALIGNMENT - ISU-016B (Continued)-

SUMMARY

-OF RESULTS (Continued)

Some problems were documented during the performance of this test.

These included defective switches, inoperable or sticking drives, bad versalect pins, drive motors tripped, bad connectors, and power supply failure.- Work orders were issued to the I&C, Electrical,

~ and Mechanical Maintenance Departments for repair and ratests were performed successfully. Problems also existed with the Plant )

Computer interface. Although the flux map traces correctly 1 displayed the simulated input signal, two problems were identified.

One problem was the Plant Computer's inability-to print a flux map ,

when using less than six detectors. The second problem was with how the Plant Computer. displays flux. map traces when using less i than the full complement of six detectors. These problems were corrected by changes to the Plant Computer software and ratests were successfully completed. l The detector cables could not access a number of core locations- I L

because of the cables sticking in the thimble-tubes and transfer devices. Westinghouse. was brought onsite to clean all the thimbles. A retest was successfully performed following the cleaning to prove the operability of the - drive system. All

-deficiencies have been corrected and the detector cables were able to access all designated core locations.

As a final step, the actual detector cables were installed on the drive units and a demonstration full core flux map was taken, even though no usable neutron flux had yet existed in the core.

Table of Detector Drive Someds Distance Time Actual Speed Allowed Drive (inches) (seconds) (inches / min) Rance A 200.1 83.5 143.78 142-146 B 200.1 83.47 143.83 142-146 C 200.1 83.58 143.4 142-146 D 200.1 83.5 143.78 142-146 E 200.1 83.58 143.6 142-146 F 200.1 83.59 143.6 142-146 4

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

Figure 3.2'.4-1 MOVABLE INCORE DETECTOR PATH LOCATIONS l

L R P N M L K J H G F E D C B .A .l l CET DET 1 l' C B _ ,

DET lCETI DET 2  !

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3.2.5 - RCS AND SECONDARY COOLANT CHEMISTRY (Post Core Load) -

ISU-006B OBJECTIVE This test is performed to verify that the water quality within the reactor coolant system and the steam generators meets the appropriate chemistry requirements. The test is performed at Cold Shutdown (Mode 5), Heat-up Prior to Criticality (Mode 3), at -

Criticality (Mode 2), and at approximately 50%, 75%, and 100%

Power. This test satisfies activities described by FSAR Table 14.2-3, sheet 11.

TEST METHODOLOGY The testing is performed by obtaining samples of the reactor coolant system and steam generators from the appropriate sample panels. Each sample is then chemically analyzed. The results of these analyses are tabulated and compared to the chemistry requirements.

SUMMARY

OF RESULTS During the executions of this test, all required Acceptance Criteria were adequately met for each system that was sampled. No corrective actions in plant operation were needed to meet the Acceptance Criteria. On occasions, one of the samples had to be reanalyzed or retaken because a result was not consistent with the others. Upon reanalysis the sample was shown to be within specifications.

Tables 3.2.5-1 and 3.2.5-2 contain a summary of the results for each system sampled along with the Acceptance Criteria or guidelines stated within the test.

i l

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l TABLE 3.2.5-1 i RCS CHEMISTRY

SUMMARY

I i

CHEMISTRY MODE MODE MODE  !

PARAMETER CRITERION 5 3 2 48% 76% 100% i i

Chloride <150 ppb 5.4 20.4 12.4 10.5 12.0 10.9 l l

s Fluoride <150 ppb <1 <1 <1 <2 <2 3.6 i Dissolved i Oxygen ** <100 ppb N/A 2 6 <2 <2 <2  ;

i Lithium *** 0.77 1.407 2.244 2.24 2.26 2.05 j Hydrogen 25-50

_ ______(( $__2f____[(f__[(f__ff;f___ff;f____ff;f___f$1I  !

Boron * >2000 ppm 2059 2086 1022 737 696 665 Gross <100/E 3.8 4.42 2.51 7.40 1.16 1.39 i Activity Ci/ml E-6 E-5 E-5 E-2 E-1 E-1 i Dose Equi- .,

valent <1.0 1.95 4.81 7.61  !

I-131 Ci/ml N/A <MDA <MDA E-5 E-5 E-5

• Mode 5 test portion sampled RHR instead of the RCS, due to system pressure, as allowed by the test procedure.  ;

Hydrogen and I-131 are not analyzed for RHR. ,

l

• • When Tave >250*F i

• • • In accordance with Lithium vs. Boron Curve above 1MW thermal (see Figure 3.2.5-1). RCS boron concentration was between f

. 400 ppm and 1200 ppm for all at-power test conditions (>l MW l l

thermal) l

• • • • When RCS >1MW thermal Reactor Power 1 1

N/A = Not applicable as no criterion is specified for this plant i condition

<MDA = Less than the Minimum Detectible Activity

t TABLE 3.2.5-2 -

STEAM GENERATOR CHEMISTRY

SUMMARY

CHEMISTRY MODE MODE MODE PARAMETER CRITERION 5 3 2 48% 76% 100%

Cation '

Conduc- 10.8 tivity* mho/cm 3.5 0.44 0.40 0.76 0.72 0.52 pH** 18.8 9.7 9.2 9.2 9.3 9.1 8.8 Sodium *** 120 ppb 4 1 1.4 6.1 2.2 8.7 Chloride ***120 ppb 30.6 7.1 3.9 4.2 2.8 9.3 Sulfate ***120 ppb 23.3 1.2 2.5 6.4 5.6 4.8 5

Silica 1300 ppb N/A 81 42 180 106 92 Hydrazine 175 ppm 50 0.3 0.10 N/A N/A N/A NOTE: The recorded value is the value from all 4 steam generators having the minimal margin to each criterion.

The silica criterion is not applicable in Modes 2, 3 & 5.

• No limit in Mode 5, Limit is 12.0 in Modes 2 & 3

• • Limit is 19.0 in Mode 5 during an outage of less than 7 days, 19.0 in Modes 2 & 3

• • • Limit is 11000 in Mode 5, 1100 in Modes 2 & 3 N/A = Not applicable as no criterion is specified for this plant condition

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3.2.6 - RADIATION SURVEY TESTS - ISU-208B OBJECTIVE The radiation survey test is performed to determine dose levels at specified points throughout the plant, to verify the effectiveness of radiation shielding, to identify any areas of streaming through shield walls and to verify proper posting of radiation areas. This test satisfies activities described by FSAR Table 14.2-3, sheet 12.

TEST METHODOLOGY Gamma and neutron radiation dose rate values are established by surveying with portable survey instrumentation in the Safeguards, Radwaste, Fuel, and Auxiliary Buildings, the Unit 2 Containment and penetration areas and the plant outside perimeter. Neutron radiation dose rate values are established in the Unit 2 Containment and certain penetration areas. Surveys are performed precritical, critical at 0-5%, 45-55% and 95-100% power. The most significant results are at the 95-100% power execution. The lower power data is primarily used to verify background radiation values and to identify potential problem areas prior to reaching full power.

SUMMARY

OF RESULTS The effectiveness of gamma shielding and the general determination of dose levels were found to be adequate during performances of the test. At nominally full power, 7 of the 96 Radiation Base Points (RBPs) could not be surveyed due to ALARA concerns. This was judged as acceptable based on acceptable trends in the lower power data and on the limited personnel access to these secured areas.

Also, 8 of the 96 RBPs exceeded their expected dose rate values, as listed below. These values were accepted because they exceeded their expected values by only a small margin and were well within their maximum allowed values.

Maximum Actual Expected Allowed RBP 4 (mR/hr) (mR/hr) (mR/hr) 10 1.2 0.4 25

. 11 0.5 0.4 25 12 0.7 0.4 25 13 5.0 0.4 25 24 0.6 0.3 10 26 0.5 0.3 5 52 2.0 1.0 10 58 2.5 1.5 10 All measured dose rates have been evaluated as acceptable for plant operation.

3.2.7 - CONTAINMENT & PENETRATION ROOMS TEMPERATURE SURVEY

- ISU-282B OBJECTIVE The Containment and Penetration Rooms Temperature Survey is performed to verify that the Reactor Coolant pipe penetrations, air supply to Reactor Vessel Supports, Neutron Detector Well discharge air, containment air, Steam Generator compartment air, Pressurizer room air, CRDM shroud air, CRDM platform area air, and FeedwAter and Main Steam penetration rooms are maintained at or below their design temperatures when the RCS is at normal no load operating temperature and also when the RCS is at 50% power and at nominal full power conditions. This test satisfies activities described in FSAR Section 9.4.A.

TEST METHODOLOGY The concrete temperature around each Reactor Coolant System (RCS) pipe penetration is measured with a thermocouple when the RCS is at normal no load operating temperature in Mode 3. Temperatures are recorded from permanent plant instrumentation for Neutron Detector Well exhaust air, CRDM shroud exhaust air and containment air.

Local readings using thermocouples or resistance temperature detectors are recorded for containment areas, Pressurizer room, Feedwater and Main Steam penetration areas, both inside and outside containment, and the Reactor Vessel Support supply air. The same measurements are later repeated with the reactor in operation at approximately 50% and 100% power.

SUMMARY

OF RESULTS After the required plant conditions were verified to have existed for a minimum of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, three sets of measurements were taken, each set at least two hours apart. The highest reading of each parameter was then compared to the acceptance criterion. All temperatures were within the acceptance criteria of the test and at approximately 50% and 100% power ext pt where noted.

TEST RESULTS Criterion Mode 3 50% Power 100% Power Concrete temperatures in each RCS Pipe Penetration are less 109.6*F 111.3*F 117.9'F than or equal to 200*F Containment average air temperature 90*F 90*F 85'F is less than or equal to 120*F

. . - -- - - - . .. _ . . . - _ - - - _ _ _ _ ~ , .

I 1

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3.2.7 - CONTAINMENT & PENETRATION ROOMS TEMPERATURE SURVEY

- ISU-282B (Continued) ,

SUMMARY

OF-RESULTS (Continued) j TEST RESULTS  :

' Criterion Mode 3 50% Power 100% Power '  !

i Steam Generator compartment air. 104.6*F 106.2'F 117.4*F temperatures are less than or .

equal to 120*F l Pressurizer-room temperatgre is 103.6*F 102.7'F 110.0*F less than or equal to 120 F f In containment, Main Steam and 98.9'T 101.9'F 106.8'F i Feedwater penetration area temp- ,

eratures are less than or equal j to 120*F Outside containment, Main Steam 93.6*F 101.8'F 108.8*F* l and Feedwater penetration room temp- i eratures are less than or equal l to 104*F  !

Neutron Detector Well and reactor 137'F 138'F 146*F  ;

vesd' ; 7 port area exhaust air j tempor 's is less than or equal to 150-1 j i

CRDM Shroud Exhaust air temper- 130*F 124*F 133*F  !

ature is less than or equal  ;

to 163*F l I

CRDM Platform area temperature 98.2*F 99.0*F 108.4*F l

?

is less than or equal to 140*F Reactor Vessel Support supply air 73*F N/A N/A-temperatures are less than or equal to 90*F. (Mode 3 only)

• The value of 108.8'F observed during testing at 95% to 100% [

power exceeded the allowed limit of 104*F. This limit is  !

imposed for Equipment Qualification concerns in this plant  :

area and the higher than expected temperature was evaluated by l Engineering to not pose an operability concern. The safety j related equipment in this area is operable up to 131*F.  ;

several problems occurred with the plant computer during l performance of this test. In all cases, missing or i irretrievable data was adequately compensated by information  !

from alternate sources, i t

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I 3.2.8 - PRACTOR COOT. ANT FLOW MFARUREMENT - ISU-023B l l

OBJECTIVE l The Reactor Coolant Flow Measurement test is performed to determina j the-Reactor Coolant System (RCS) flow rates for each of the 4 RCS <

loops, the total RCS flow rate, and to verify proper RCS flow i indications. This test is performed prior to initial criticality l (Mode 3) and during power ascension at approximately 75% and 100%  ;

power. This test partially satisfies activities described by FSAR  !

Table 14.2-3, sheets 2 and 2a and Technical Specification 3/4.2.5. 1 l

TEST METHODOLOGY Prior to criticality, data is obtained from the installed elbow tap differential pressure (d/p) instrumentation and used to calculate l the RCS loop flowrates. Average values for pressurizer pressure, RCS narrow range cold leg temperature and indicated RCS flow (in j

< percent) are determined concurrently. The temperature and pressure )

, readings a a unad to obtain cold leg specific volumes using Steam i Tables. The c' sated RCS flows (in percent) are converted to l inches of water based on the individual transmitter scaling. Each j loop has three flow transmitters from which a flow measurement is I

taken. The d/p values are used to determine three flow rate values for each loop using an equation for Reactor Coolant

• Cold Leg .

Volumetric Flow Rate as a function of Elbow Tap d/p and specific  ;

volume. These three flow rates are averaged to obtain the loop l average flow rate. The average flow rates from all four loops are  !

summed to obtain the total'RCS flow rate. j The flow transmitters are verified to be aligned and calibrated by review of the appropriate completed Instrumentation & Controls work i documents. RCS flow indications, processed'from the elbow tap d/p i transmitters, are read from the Plant Computer and verified to ]

indicate 100% 1 a specified error tolerance of 1.4%.  ;

With the plant at approximately 75% and 100% power, data is taken to determine the RCS loop flow rate. This data is a combination of ,

I a precision secondary plant calorimetric, cold leg RCS temperature values and N-16 Transit Time Flow Meter (TTFM) outputs. The TTFM is a direct flow measuring device using gamma detectors mounted on the outside of the RCS hot legs. RCS water flowing through the reactor has a portion of the Oxygen-16 nuclei present in the H.,0 molecules activated to Nitrogen-16 by the neutron absorption prot 5n I emission reaction. This N-16 leaving the reactor has a half-life of 7.10 seconds and emits gamma rays of 6.129 and 7.115 MeV. These gamma rays penetrate the RCS loop piping and are sensed by the N-16

gamma detectors. The detectors are located transversely to RCS loop flow and are collimated te observe fluctuations in the N-16 gamma activity as flow passes the detector. To measure RCS loop 3.2.8 - REACTOR COOLANT FLOW MEASUREMENT - ISU-023B (Continued)

TEST METHODOLOGY (Continued) flow the TTFM uses two pairs of gamma detectors located approximately 2 1/2 feet apart, 2 detectors upstream and 2 downstream of each other. _ Loop volumetric flow is calculated by multiplying the piping inside cross-sectional area by the fluid velocity. The fluid velocity is the known detector upstream-downstream spacing divided by the fluid transit time between them. ,

A statistical cross-correlation of the N-16 gamma signal between upstream-downstream detector pairs results in this transit time.

All possible upstream-downstream detector combinations are used to calculate transit times, then combined to form a mean transit time.

The cross-correlation data collection, analysis and calculation of volumetric flow rate is performed by the TTFM that is connected to the N-16 detector outputs for the given loop under test. The TTFM is moved from loop to loop sequentially and does not measure all 4 '.

RCS loop flows simultaneously. The RCS _;ot leg volumetric flow.

rates from the TTFM are converted to RCS cold leg flows using measured RCS cold leg temperature combined with a RCS hot leg temperature that is calculated from calorimetric power, cold leg -

temperature and hot leg vclumetric flow rate. These temperatures are used to calculate hot and cold leg specific volumes and the ratio of specific volumes is used to convert hot leg volumetric r flow to cold leg volumetric flow.

SUMMARY

OF RESULTS r At 0% power in Mode 3, all test criteria were met. Results were as follows:

• The RCS flow elbow tap d/p transmitters were verified to have been aligned for both zero and 100% flow prior to Mode 3  :

testing.

• The indicated percent RCS flows at normal RCS operating conditions in Mode 3 ranged from 99.5 to 100.3% which  ;

satisfied the specified 100% 1 1.4% flow range.

• The total RCS flow rate must be equal to or greater than 355,680 gpm (90% of the Thermal Design Flow) as determined by -

elbow tap d/p instruments prior to criticality. This was satisfied in Mode 3 at 0% power, where the total RCS flow rate was 431,155 gallons per minute.

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3.2.8 - RFACTOR COOLANT FLOW MEASUREMENT - ISU-023B (Continued)

SUMMARY

'OF RESULTS'(Continued)

At-power levels of'72%, 78% and 100%, RCS flow rate was determined i through the use of the N-16 Transit Time Flow Meter and a precision  !

secondary side calorimetric. When determining ~ the adequacy of .

determined flow rates, the RCS Lower. Plenum Flow Anomaly (LPFA) is i taken.into account. The LPFA is a core inlet flow maldistribution I attributed to the presence of a periodically occurring vortex type  ;

flow disturbance in the reactor vessel lower plenum. Because thel flow anomaly reduces the actual flow through the core,- the flow -

measured with the TTFM is penalized (reduced) by a conservative l amount (0.5%) before verification that the minimum flow requirement 1 is satisfied. However, for verification of the Review Criterion i for Mechanical Design Flow (maximum value), the actual measured flow, without LPFA penalty, is used. J i

l The following table represents the results from the TTFM flow  ;

measurements:

1

-Total RCS Total RCS Flow Without Flow With Minimum RCS Maximum RCS Reactor LPFA Penalty LPFA Penalty Flow Criteria Flow Criterion  !

Power (%) (ana) (anm) (ans) (ana) '

72 420,914 418,809 2395,200 s420,000 i 78 420,422 418,320 2395,200 s420,000  :

100 418,993 416,898 2395,200 5420,000  !

The minimum RCS flow criterion was easily satisfied for each of the flow measurements. The maximum flow criterion, based on mechanical design, was not achieved for the 72% and 75% power measurements. '

However, this criterion was satisfied during the 100% power test '

perJarmance. Because of this discrepancy, data from all three test '

perl'rmances was provided to the vendor for evaluation. The vendor <

subsequently determined that the measured RCS flow rate is l accep able. The vendor further stated that the 420,000 gpm limit applies to 100% power only. The 72% power results also satisfied-the zSluirements of Technical Specification 4.2.5.4 to have a flow rate of greater than or equal to 395,200 gpm, as determined by the TTFM, prior to exceeding 75% power.

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3.2.9'- REACTOR COOT. ANT SYSTEM FLOW COASTDOWN --ISU-024B OBJECTIVE The Reactor Coolant System Flow Coastdown test is performed with the unit in Hot Standby (Mode 3) to verify that the core flow during Reactor Coolant Pump (RCP) coastdown would exceed the flow assumed in the accident analysis during flow decay following a simultaneous trip of all four RCPs. In addition, the. low flow reactor trip time delay is verified to be within acceptable limits.

This test satisfies activities described by FSAR Table 14.2-3, '

sheet 3.

TEST METHODOLOGY Strip chart recorders are connected to the RCS elbow tap d/p transmitter outputs and the Solid State Protection System (SSPS) to ]

monitor Reactor Coolant System flow characteristics and Reactor l Trip Breaker positions as a function of time. . A P-8 permissive is  !

simulated (>48% power) to ensure that a single loop loss of flow results in generation of a reactor trip signal. All four Reactor Coolant Pumps are tripped by manual actuation of the RCP J

Underfrequency Trip relay. Flow and SSPS data are taken while the l RCS flow decays off. All four Reactor Coolant pumps are verified ,

to trip within 0.100 seconds of each other to ensure that the flow decay data corresponds to an essentially simultaneous loss of all-forced RCS flow. Data from the strip charts is'then statistically evaluated to verify acceptability of the measured flow values and b'

l related time delays.

SUMMARY

OF RESULTS The Flow Coastdown Time Constant was required to be ~ greater than or equal to 11.76 seconds. The measured value was 13.69 seconds. The total Low Flow trip time delay was required to be less than-or equal to 1.0 seconds. The measured value was 0.767 seconds. The

(

' Reactor Coolant Pumps were also verified to trip within 0.040 seconds of each other, which was well withit; the 0.100 second limit.

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3.2.10 - REACTOR COOLANT SYSTEM LEAKAGE RATE - ISU-022B l OBJECTIVE i

The purpose of this procedure is to verify the Reactor Coolant l System (RCS) leak tightness after the system has been closed. This test satisfies activities described by FSAR Table 14.2-3, Sheets 29 and 30.

TEST METHODOLOGY

( With the plant in Hot Standby (Mode 3) conditions, prior to initial criticality, the reactor coolant system is tested to verify leak l

tightness. After RCS pressure is stabilized, a visual leak test is

conducted with the reactor pressure vessel, pressurizer and all four reactor coolant loops verified to be leak tight. Also, the unidentified, identified, and controlled leakage rates are determined using normal operating Technical Specification surveillance techniques and results from OPT-303 and OPT-110B.

l Pressure isolation valve leakage is also verified, based on normal Technical Specification surveill.ance results from PPT-S2-7000A through PPT-S2-7007B. Primary tc secondary leakage is determined by measuring boron concentration of the steam generator liquid. j This calculation is based on RCS boron concentration, steam !

generator boron concentration, steam generator blowdown flowrate  !

and time. This primary to secondary leakrate is measured in gpd,  !

gallons per day. Under normal conditions, the minimum detectable boron concentration of 0.1 ppm would result in a calculated leakage  ;

rate of 2.88 gpd. I i

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3. 2.10 - REACTOR COOLANT SYSTEM LEAKAGE RATE - ISU-022B (Continued) ,

SUMMARY

OF RESULTS During the visual inspection, no pressure boundary leakage was observed nor was any leakage past'the Reactor Vessel flange seal observed from the flange seal leakoff. No boron was detected in the steam generators, so the conservative 0.1 ppm value was assumed. Leakage rate results are tabulated below:

Leakage Acceptance Rate TvDe CIiterion (com) Test Results (com)

Controlled i 40 33.4 Identified i 10 Unidentified i1 Pressure Isolation Valve 5 0.5/ nominal 3.04**

inch of valve size Primary to Secondary 1 500 gpd/ steam 12.88 gpd/ steam generator generator Total Primary to Secondary 51 10.008

• Total RCS leakage was 0.6822 gpm per OPT-303. This value satisfied both the Identified and Unidentified leakage criteria and

~

were not broken out as separate items.

• • The recorded value is the maximum value for any one valve. -

Individual valve leakage rates ranged from 0.0 to 3.04 gpm with a total of 19.038 gpm for 35 valves.

l This test verified acceptable leak tightness of the reactor coolant system.

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i 3.2.11 - COLD CONTROL ROD OPERARILITY TESTING - ISU-026B OBJECTIVE  ;>

The purpose ' of this. test is to verify. coil polarities, proper i Digital Rod Position IndicationL(DRPI) system operation, rod drop timing, alarm functions, DC Hold Cabinet operation, and proper. ,

slave cycler timing and to perform an . operational check of each i Control Rod ~ Drive Mechanism (CRDM) with a - rod cluster control  ;

assembly (RCCA) attached prior to initial use of the mechanism.  ;

This test partially satisfies activities described by FSAR Table ~

14.2-3, Sheets 4 and 5 and Technical Specifications 3/4.1.3.3 and i

3/4.10.5.

TEST METHODOLOGY Portions of this test may be is performed under two . plant  !

conditions: Mode 5 -

cold shutdown and Modes 3, hot standby. l Proper operation of coil polarities, DRPI operation, rod drop l timing, CRDM operation and slave cycler timing are verified under  !

Mode 5, conditions. The rod bottom, rod deviation, urgent and non- .!

urgent failure alarms and the DC Hold Cabinet are tested in either  !

Mode 3 or 5. l

. Coil polarities are verified to preclude individue.1 coil magnetic l fields from the stationary gripper, movable gripper and lift coils ,

from interfering with one another. This test uses two 6VDC batteries in series and a test switch to inject low voltage current I pulses into the moveable gripper coil and observes the direction of j current flow induced in the other two coils. Then the current is  ;

injected into the stationary gripper coil and the direction of  ;

induced current flows in the other two coils is again verified.  !

Each of the 53 CRDM coil. stacks is individually tested in this  !

manner.

Slave cycler timing and CRDM operational checks are performed ,

tarting with all RCCAs positioned at the core bottom. A selected 1 single bank is withdrawn 50 steps to ensure the RCCAs are above the dashpot region. Each RCCA in the withdrawn bank is then  :

individually withdrawn 5 steps and reinserted 5 steps. When all ,

RCCAs in a bank have been tested, the entire bank is reinserted to

. the bottom of the core. This is then repeated for each bank.

While the individual RCCAs are being withdrawn and inserted 5 steps, a visicorder is used to monitor lift coil, stationary gripper coil and movable gripper coil currents. An optional signal from a microphone attached to the CRDM housing may be used to help relate actual mechanical events to the coil currents. These visicorder traces are evaluated to verify that the coil current  !

signal traces were of the proper shape, the currents were of the proper magnitudes at the proper times, and to verify that the i

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3. 2.11 - COLD CONTROL ROD OPERABILITY TESTING - ISU-026B(Continued) .

TEST METHODOLOGY (Continued) events associated with -mechanism movements occur in the proper order. The traces for each RCCA are compared against vendor supplied criteria and model traces. Rod speeds are calculated from the period of successive rod steps as being the inverse of stepping frequency.

DRPI system operability is verified by monitoring DRPI Light Emitting Diode (LED) indications on the control board during bank withdrawal and comparing these indications against other indications of RCCA position; the plant computer, the demand step counters, and the rod control system pulse to analog converter. A selected bank of RCCAs is withdrawn to 231 steps, the mechanical RCCA limit of motion. Shutdown bank withdrawals are stopped at 18, 210, and 228 steps to record the various position indications listed above. Control bank withdrawals are stopped every 24 steps and at 228 steps to record this data. These periodic indication verifications are also used to demonstrate Technical Specification ,

operability of the DRPI system per Surveillance Requirement 4.1.3.3. .

With a selected rod bank fully withdrawn to 231 steps, the DRPI data cabinets in the containment building are de-energized. A strip chart recorder is hooked up to outputs from the DRPI data cabinets and the reactor trip breakers are then opened. As the RCCAs drop the, slightly magnetized, individual CRDM drive shafts .

which are connected to the RCCAs, also drop through the deenergized DRPI sensing coils and induce a current in these coils which is proportional to drop velocity. As the RCCA enters the dashpot region of the fuel assembly guide tubes, it is hydraulically braked, which also shows up as a significant velocity change in the  :

induced current signal. The chart recorder displays the induced current signals as a function of time from all RCCAs in the selected bank, a signal proportional to stationary gripper current, and an event mark for the opening of the reactor trip breakers.

From this information the rod drop time to dashpot entry can be evaluated. This rod drop timing testing is performed only one bank at a time, but is performed simultaneously for all RCCAs within a bank. The Surveillance Requirements for Technical Specification Special Test Exception 3.10.5 are .=1'.isfied within this test to allow the DRPI system to be de-onegized for rod drop timing measurements.

3. 2.11 - COLD CONTROL ROD OPERABILITY TESTING - ISU-026B(Continued)

TEST METHODOLOGY (Continued)

The rod deviation and dropped rod alarms are verified by withdrawing all shutdown RCCA banks to 228 steps, withdrawing Control Bank A to 18 steps, and then moving individual RCCAs as necessary to activate the particular alarm being tested. These initial positions are specified to clear alarms that normally exist with all RCCAs inserted. The rod deviation alarm is verified by deviating two RCCAs in Control Bank A by 12 steps or more and also ,

by partially inserting a Shutdown- Bank C RCCA from its fully withdrawn position. The dropped rod alarms are verified by inserting one RCCA from Control Bank A to near full insertion and then by inserting a second RCCA for the >2 rods at bottom alarm.

This alarm circuit logic is such that a successful test using any '

RCCA or pair of RCCAs verifies the alarm for all other associated RCCAs.

The non-urgent failure alarm is tested by removing the input power fuses for one power supply in each of the five rod drive system's power cabinets and the logic cabinet. The cabinets are tested sequentially, not simultaneously. The urgent failure alarm is tested by interrupting the lift coil firing circuit to all RCCAs powered by a single rod drive system power cabinet under test.

When the RCCAs associated with the cabinet under test are ordered to withdraw, the urgent failure alarm actuates in response to the missing lift coil current. The cabinets are tested sequentially, not simultaneously, using the permanently installed lift coil disconnect switches. The logic cabinet urgent failure alarm is tested by removing a preselected circuit board.

The DC Hold Cabinet serves as an alternate power source to hold a single group of up to four RCCAs in a withdrawn position to allow for maintenance on the stationary gripper power circuitry for that group. The DC Hold Cabinet is tested by switching it to hold a group of 4 withdrawn RCCAs, de-energizing the normal power circuitry for that group and verify.ng the RCCAs remain withdrawn.

SUMMARY

OF RESULTS

. Coil polarities were all verified to be correct.

The current and sound traces from all 53 RCCAs were all verified proper when tested. The traces were all of the proper shape with no significant anomalies. The timing of events and current magnitudes were all verified to be acceptable and concurred with by l 3.2.11 - COLD CONTROL ROD OPERABILITY TESTING - ISU-026B(Continued) k j

SUMMARY

OF RESULTS (Continued) the NSSS vendor. Actual rod speeds from evaluation of the inverse of the period of successive rod steps were as follows:

Expected Actual RCCA Bank Tvoe Soeed(steos/ min) Somed(steos/ min)

Control Bank 48 : 3 46.7 Shutdown Bank A or B 64 t 3 63.4 Shutdown Bank C,D or E 64 3 69.1 1 The tolerances on the expected speed values are for information I only because these rod speed values are to be rameasured at hot (Mode 3) RCS conditions as part of the ISU-027B test, section 3.2.12 of this Startup Report. These cold values are expected to differ from the expected speed values due to mechanical (thermal expansion) conditions associated with the low RCS temperature. For Shutdown Bank C, the time between rod steps was observed to be smaller than expected. Evaluation of the traces resulted in a measured rod speed of 69.1 steps per minute which exceeded the 64

3 expected range. An adjustment was made to the rod speed circuitry for that cabinet and the final value following this adjustment was obtained as part of the ISU-027B test.

The DRPI LED indications on the main control board and plant computer were verified to be within 4 steps of the rod drive ^,

system group step counter indications for all 53 RCCAs. The actual deviation was less than or equal to 3 steps. The pulse to analog converter indications were verified to be within 11 step of the group step counter indications for all 4 control banks. The actual agreement was either 0 or 1 step. The DRPI LED indication for rod bottom (RB) indicated at or prior to reaching zero steps, as indicated by the group step counter, during RCCA insertion. The RB LEDs all illuminated at an indicated 2 steps. This DRPI testing was performed in Mode 5.

Following correction of the plant computer software, the rod deviation alarms were verified to function properly for an actual li rod vs. rod deviation of 12 steps and a shutdown rod at 210 steps {

withdrawn. The rod bottom and >2 rods at bottom alarms were then also verified to function properly in response to actual RCCA i insertions. These alarms were tested in Mode 5.

The urgent and non-urgent alarms functioned properly in response to failed power supplies, missing lift coil currents and the missing circuit board. These alarms were tested in Mode 5.

The DC Hold Cabinet was verified to hold a group of 4 RCCAs in a withdrawn position for 10 minutes. This verification was performed in Mode 5.

3. 2.11 - COLD CONTROL ROD OPERABILITY TESTING - ISU-026B(Continued)

SUMMARY

OF RESULTS (Continued) i All rod drop times were verified to be less than the Technical Specification limit of 2.4 seconds. That limit does tot actually apply to this test performance because the limit is for a hot, full RCS flow test and these rod drops were done cold in Mcde 5. The average drop time was 1.326 seconds, ranging from a low of 1.310 seconds to a high of 1.353 seconds. The rod drop data was taken for baseline purposes and to verify rod drop test equipment operation only. Additionally, evaluation of the rod drop DRPI coil current trace shapes verified proper operation of the dashpot decelerating devices.

Two miscellaneous minor problems were also noted with respect to  !

initial rod drive and DRPI system operation during performance of this test:

The RCCA at core location D-12 did not withdraw with the rest of Shutdown Bank A due to lift coil firing problems.

The DRPI LED indications for the RCCAs at core locations F-8 and F-10 were erratic due to DRPI circuit board problems for the DRPI circuitry associated with these RCCAs.

Both of the above items were repaired under Work orders prior to the completion of this test. Satisfaction of the appropriate testing criteria verified the success of the repairs. ,

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3.2.12 - HOT CONTROL ROD OPERABILITY TESTING - ISU-027B OBJECTIVE The purpose of this test is to verify proper DRPI system operation, rod drop timing, rod speed and direction, overlap operation, manual operation and to perform an operational check of each CRDM with a RCCA attached in Mode 3 prior to initial criticality. The actual mechanical RCCA withdrawal limit is also verified. This test partially satisfies activities described by FSAR Table 14.2-3, Sheets 4, 19, 31 and 32 and Technical Specifications 3/4.1.3.4 and 3/4.10.5.

TEST METHODOLOGY Slave cycler timing, Control Rod Drive Mechanism (CRDM) operational checks, measurement of the mechanical withdrawal limit, DRPI system checks and rod drop timing is performed in an integrated fashion, on a sequential bank by bank basis. The selected bank is withdrawn to 228 steps, with the operation of every DRPI LED verified during i this withdrawal with respect to the group step counter indications.

There is an LED for every 6 steps of RCCA motion. Shutdown banks have no LEDs to represent position between 18 and 210 steps, only a transition region (TR) LED. Each individual RCCA in the withdrawn bank is inserted 5 steps and then withdrawn 10 steps, ending at an indicated position of 233 steps on the group step counters. During these 5 and 10 step movements, visicorder trace data is taken as was done in ISU-026B, refer to section 3.2.11 of this Startup Report. This trace data is also evaluated as was done in ISU-026B with respect to rod speeds and the timing and magnitudes of coil current changes. Sound traces are not taken due to microphone integrity concerns while at normal RCS operating temperature. The trace data is also evaluated to verify the mechanical RCCA withdrawal limit, typically either 230 or 231 steps. When the CRDM drive shaft reaches its mechanical limit of travel there are no more grooves on the Control Rod Drive Shaft available for the CRDM grippers to latch into. This shows up on the trace as a gripper current anomaly. The traces are evaluated near the top of travel, above 228 steps, with respect to where this anomaly occurs. This mechanical limit is typically 231 steps but can vary from reactor to reactor. Once the mechanical withdrawal limits have been determined for all RCCAs in a bank, that bank is dropped to measure rod drop times as was done in ISU-026B. This  :

set of rod drop times satisfies Surveillance Requirements for Technical Specification 3.1.3.4 and is performed at RCS hot, full flow conditions. The Surveillance Requirements for Technical  !

Specification Special Test Exception 3.10.5 are also satisfied l within this test. Any RCCA having a drop time deviating from the mean drop time by more than two standard deviations is redropped an additional three times to confirm its actual performance.

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3. 2.12 - HOT CONTROL ROD OPERABILITY TESTING - ISU-027B (Continued)'

' TEST METHODOLOGY (Continued) ,

-Rod speed and direction indications on the main control board are  !

verified while withdrawing and inserting various RCCA banks. j Control bank overlap is-verified by withdrawing the control banks i

in the. manual overlap mode instead of in the individual bank select mode of operation. As a prerequisite to this test portion, the i overlap switch settings are changed to lower, yet sequential, i values. This allows verification of overlap without the need for  !

complete withdrawal of the control banks. As the control. banks are [

withdrawn in manual overlap, _ data is recorded each time a bank i starts or stops motion. This data is compared to the ' switch settings. l The ability of an-urgent failure alarm to block RCCA motion is tested by creating an actual urgent failure alarm, by interrupting ,

lift coil signals using the permanently installed lift coil  ;

disconnect switches, and then attempting to move the RCCAs. The '

urgent failure alarm is then cleared and RCCA motion is verified to l have been restored. l

SUMMARY

OF RESULTS The current traces from.all 53 RCCAs were verified to be proper. {

The traces were of the proper shape with no notable anomalies. The j timing of events'and current magnitudes were all verified to be  ;

acceptable and were concurred with by the NSSS vendor. i Actual rod speeds were adjusted as necessary and finally verified i to all be within their expected ranges as follows. 3 l

Expected Actual {

RCCA Bank Tyne Somedisteos/ min) Soemd(steos/ min) .

Control Bank 48 12 46.9 Shutdown Bank A or B l 64 12 62.6 Shutdown Bank C,D or E 64 12 63.2 j The DRPI system indications were typically within 1 or 2 steps of the group step counter indication with only one group of 4 RCCAs j off by 3 steps at one rod position, Shutdown Bank D at 207 vs. 210  ;

steps. This satisfied the 14 step agrerment s criterion. i I

s The mechanical withdrawal limit (full out position) was established i to be 231 steps by inspection of visicorder trace data above 228  !

steps. This trace data was repeated for Shutdown Bank A due to i legibility problems with that portion of the original trace data j for that bank. This trace data was also repeated for Control Bank i A due to a visicorder paper jam.  !

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3. 2.12 - HOT CONTROL ROD OPERABILITY TESTING - ISU-027B (Continued)

SUMMARY

OF RESULTS (Continued)

Rod drop timing measurements were made for all 53 RCCAs from the 231 step full mechanical withdrawal position. All times were less than 2.4 seconds from decay of stationary gripper voltage to dashpot entry. The fastest RCCA took 1.380 seconds. The slowest RCCA took 1.490 seconds. The average RCCA drop time was 1.4331 seconds with a standard deviation of 0.03 seconds. Only four RCCAs were outside of the two standard deviation limits. They were redropped 3 times each with the following results:

Droo Times (seconds)

Droo Tvoe RCCA F-6 RCCA M-2 RCCA M-14 RCCA L-13 Original 1.390 1.490 1.475 1.380 Redrop #1 1.41 1.45 1.45 1.395 Redrop #2 1.39 1.45 1.44 1.39 Redrop #3 1.38 1.45 1.45 1.385 Except for RCCA F6, these redrops also satisfied the criterion that for each redropped RCCA, the three redrop times shall all be within a 0.02 second band. The rod drop traces for the F-6 redrops were examined by the NSSS vendor with no anomalies noted, the F-6 time range of 0.03 seconds, only slightly greater than 0.02 seconds, was evaluated to not be indicative of a RCCA problem. This rod drop timing test satisfied the Surveillance Requirements for Technical Specification 3.1.3.4 and Special Test Exception 3.10.5.

The rod speed and direction indications on the main control board were verified to be correct. The speed indications of either 48 or 64 steps / minute were correct. Control bank overlap was verified to occur exactly at the overlap switch settings with no deviation.

This satisfied the allowed il step deviation criterion.

The urgent f ailure was generated and was verified to inhibit manual RCCA motion. RCCA motion was restored following clearing of the alarm.

P 3.2.13 - REACTOR TRIP SYSTEM TESTS - ISU-015B OBJECTIVE The purpose of this test is to verify proper operation of the '

automatic and manual reactor trip breaker circuitry and to verify proper operation of the reactor trip breakers prior to initial  ;

criticality. This procedure also tests reactor trip bypass breaker functions and verifies proper unlatching of the control rods following opening of the reactor trip breakers. This test satisfies activities described by FSAR Table 14.2-3, Sheets 6 and ,

7.

TEST METHODOLOGY The Solid State Protection System (SSPS) general warning interlocks i associated with the trip breakers are uested by closing both '

reactor trip breakers (RTBs) and one of the trip bypass breakers (TBBs). The SSPS train opposite to the TBB that is closed is placed into test and it is verified that all three breakers then open automatically. This sequence is repeated for the other TBB i and SSPS train.

TBB interlocks are tested by closing one TBB and verifying that an attempt to close the second TBB results in the automatic opening of both TBBs. This sequence is repeated with the other TBB starting in the closed position.

Functional testing of RTB and TBB operation is performed by closing both RTBs and one TBB. A trip signal is then simulated on the SSPS train associated with the closed TBB. It is verified that the RTB corresponding to the tripped SSPS train opens and the other two breakers remain closed. This sequence is repeated for the other TBB and SSPS train.

Manual trip function is demonstrated by closing both RTBs and one TBB and generating a manual trip signal from a control board reactor trip switch. All three breakers are verified to open and the remaining TBB is closed and verified to open in response to a-  :

second actuation of the reactor trip switch. This sequence is '

repeated for the second main control board reactor trip switch.

verification of actual control rod release following RTB opening is tested by withdrawing all 53 control rods to 12 steps and manually initiating a trip signal using a main control board reactor trip switch. All control rods are verified to return to their fully inserted positions using the Digital Rod Position Indication System.

l 3.2.13 - REACTOR TRIP SYSTEM TESTS - ISU-015B (Continued)  ;

SUMMARY

OF RESULTS SSPS general warning ;.aterlocks were verified to properly result in ,

the opening of the RTBs and TBBs._ TBB interlocks were verified to properly prevent simultaneous closure of both TBBs. Proper function of RTB and TBB operation was verified, demonstrating that  :

the TBBs permit individual RTB trip testing without resulting in an actual reactor trip. The manual reactor trip switch trip function ,

was properly demonstrated for both main control board reactor trip switches. All 53 control rods were verified to unlatch and fall  !

from the 12 step position to the fully inserted position following l opening of the RTBs.

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3.2.14 - PRESSURIZER SPRAY AND HEATER CAPABILITY - ISU-021B OBJECTI'7E This test is performed to verify pressurizer spray effectiveness. In addition, the spray line bypass valves are adjusted to maintain spray line temperature above 540*F and valve leak tightness is verified. Pressurizer heater performance is verified during the preoperational test program. This test partially satisfies activities described by FSAR Table 14.2-3, Sheets 2 and 2a.

TEST METHODOLOGY In order to set the spray line bypass flows, the spray valves and spray bypass valves are closed and the line temperatures allowed to stabilize. The valves are then opened in up to approximately 1/4 turn increments until a satisfactory temperature reading is achieved. The spray line low temperature alarm is verified to actuate at 525 + 3, -0*F.

To verify spray effectiveness, the heaters are manually isolated and both spray valves are placed into the full open position.

Pressurizer parameters are monitored via the Plant Computer. These parameters are then analyzed and plotted to verify the pressure transient falls within the allowable limits.

To verify stable pressurizer pressure control ability, the spray valves and heaters are manually operated to adjust pressurizer pressure to approximately 2200 psig. The controls are placed in automatic and pressurizer pressure is verified to stabilize within the normal operating band of 2235 30 psig. A similar test is also performed starting at approximately 2300 psig.

SUMMARY

OF RESULTS The pressurizer spray bypass valves were properly set to ensure that adequate spray line temper.tures exist when the spray valves are closed. This prevents excessive spray line cooldown which can l cause potentially deleterious thermal effects on piping end l

components when sprays are activated. Valve 2RC-8051 was set to 1 3/8 turns open and valve 2RC-8052 to 1/16 turns open. These settings result in spray line temperatures of 533*F to 545'F, which also allows for sufficient margin above the 525'F low temperature alarm setpoint.

Pressurizer spray valve 2-PCV-0455C had seat leakage of a magnitude such that it masked the contribution from 2RC-8052 in keeping the spray line warm. This valve, 2RC-8052, will be reset following spray valve repair. The existing configuration adequately maintains spray line temperatures for operation. .

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- . v t'

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

3.2.14 - PRESSURIZER SPRAY AND' HEATER CAPABILITY -'ISU-021A  :

(Continued) [

SUMMARY

OF RESULTS (Continued) [

Testing also determined that these settings are the minimum-valve positions that can maintain line temperatures adequately above  ;

525 'F and that', at these valve settings, pressurizer control heater .

bank C would maintain pressurizer pressure _ by itself, without l excessive periodic backup heater bank actuation. Periodic backup-heater operation is not.a safety or operability. concern, only one ,

of efficiency.

The pressurizer pressure was verified to stabilize at 2235 30 psig ,

when controls were placed in automatic from starting points at  ;

approximately 2200 and 2300 psig. No sustained or diverging i oscillations were noted.

t The pressurizer PORVs were verified to have no detectable . seat i leakage. Spray valve 2-PCV-0455B had no detectable seat leakage. 3 Spray valve 2-PCV-0455C had some seat leakage, but of a magnitude  !

small enough to not adversely impact unit operation, as discussed i previously. .

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1 Figure 3.2.14-1 PRESSURE RESPONSE TO OPENING BOTH PRESSURIZER SPRAY VALVES (WITH ALLOWABLE DEVIATION) 0 -- a f

-50

' b E

w -100 "

O " = LOWER BAND o

ae 88 PRESSUP.E CHANGE o

E o

• UPPER BAND g -150 ._ __ _ _ _ . _,

a E u

-200 .

ll

.g g j g_ l - - -- g- _ . g __ _ - . - g -- l 0 10 20 30 40 50 60 TIME IN SECONDS

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3.2.15 - DYNAMIC AUTOMATIC STEAM DUMP CONTROL. ISU-205B l OBJECTIVE l

i The purpose of this procedure is to demonstrate the control capability of the Steam Dump System. The steam pressure, plant f trip, and load rejection controllers are tested to verify i Each controller is also tested to j consistency with design curves.

demonstrate stability following a small transient. I l

TEST METHODOLOGY j Prior to the transient testing portions of this procedure, the steam pressure, plant trip, and load rejection controllers are {

electronically checked using simulated plant signals and J parameters. This functional testing includes modulating all steam dump valves fully open and tripping open all steam dump valves using simulated plant trip and load rejection signals. The l controllers are recalibrated as necessary to correct unacceptable performance.

The steam pressure controller is tested by varying reactor power and allowing the steam dump valves to maintain steam header pressure. With the Steam Dump System in the Steam Pressure Mode, reactor power is increased approximately 3% by control rod withdrawal. Steam dump valves 2-PV-2369A, 2-PV-2369B and 2-PV-2369C will open and modulate in order to maintain steam header pressure stable and near 1092 psig. Reactor power is then decreased to its initial value by control rod insertion, and the j steam dump valves close. j l

The plant trip controller testing begins by simulating a reactor trip signal. With the Steam Dump System in the TAVG Mode and RCS temperature at 557'F, reactor power is increased by approximately 3% by control rod withdrawal. Steam dump valves will open smoothly in order to maintain RCS temperature stable and near 559'F.

Reactor power is then decreased by control rod insertion.

1 The load rejection controller testing begins by simulating a loss of load condition with a TREF signal of 552*F. With the Steam Dump System in the TAVG Mode and RCS temperature at 557'F, reactor power is increased by approximately 3% by control rod withdrawal. Steam dump valves will open smoothly in order to maintain RCS temperature I stable and near 558'F. Reactor power is then decreased by control rod insertion.

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3. 2.15 - DYilAMIC AUTOMATIC STEAM DUMP CONTROL, ISU-205B (Continued)

SUMMARY

OF RESULTS The functional e ack of the Steam Dump System had satisfactory results. While simulating main steam header pressure, each valve's position followed its design curve as a function of steam dump demand. For both the plant trip controller and the load rejection controller, steam dump demand followed 'its design curves for varying simulated RCS temperature.

For the steam pressure controller testing, results were satisfactory. Main steam header pressure: stabilized (i.e., no divergence) and was maintained between 1086 and 1089 psig. This is l

well within the 1072 to 1112 psig criterion.

For the plant trip controller testing, results were satisfactory.

Temperature control in the TAVG Mode was not divergent during the transient. RCS temperature was controlled between 555.8'F and 558.2*F, a 2.4*F band. Although this band is outcide the stated criteria of 556*F and 562*F, the controller operation was found to be acceptable because the starting temperature of the transient was low ' .. e., 555.8'F), as allowed by the procedure.

For che load rejection controller testing, results were satisfactory. RCS temperature stabilized (i.e. , no divergence) and was maintained between 556.8aF and 558.5*F. This is well within the 556*F to 560*F criterion.

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

t 3.2.16 - TURBINE GENERATOR INITIAL SYNCHRONIZATION AND i OVERSPEED TEST, ISU-220B  ;

OBJECTIVE ,

.i The purpose of this test.is to verify the main turbine trips on an actual overspeed. . .. In addition, ' this procedure demonstrates the  ;

ability to synchronize the main generator to the grid. l

?

TEST METHODOLOGY  ;

i The turbine is brought ' up to rated speed' (i.e. , 1800 RPM) and  !

vendor checks are performed. Turbine speed is then. increased to  !

the EHC hydraulic governor stop setting to verify the setpoint. In ,

order to further increase turbine speed, the trip test lever is-depressed. Overspeed trip testing is then performed.

The turbine is brought back to rated speed, final synchronization I checks are performed, and the main generator is synchronized to the  ;

grid. At this point, the EHC hydraulic fluid lines to the turbine control valves are monitored for excessive vibration. ,

SUMMARY

OF RESULTS  !

L. i The turbine was brought up to 1800 RPM and then increased to the  ?

EHC hydraulic. governor stop. setting, which was 1935 RPM. This was l,

~'

within the tolerance of 1916 to 1936 RPM.

The mechanical overspeed trip occurred at 1980 RPM. This satisfies l the acceptable range of 1980 to 1998 RPM.  ;

The main generator was successfully synchronized to the grid at  :

1943 hours0.0225 days <br />0.54 hours <br />0.00321 weeks <br />7.393115e-4 months <br /> on April 9, 1993. It was verified that the EHC  !

hydraulic fluid lines to the turbine control valves did not have  !

excessive vibration. l i

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3.2.17 - MAIN FEEDWATER SYSTEM TEST, ISU-238B OBJECTIVE This procedure monitors main feedwater system components during power ascension. The procedure verifies the interface criteria between the steam generators and the feedwater system, in particular the tempering flow at 75% power and the feedwater flow split at 100% power.

TEST METHODOLOGY In Mode 2, this procedure determines if there is any significant leakage past the Feedwater Isolation Valves (FIVs) and Feedwater Isolation Bypass Valves (FIBVs). With the FIVs and FIBVs closed, a main feedwater pump is started. Feedwater temperature is monitored upstream and downstream of the isolation valves.

Significant leakage is indicated by a temperature change in the section of feedwater piping between the isolation valves and the main steam generator nozzle.

At low power in Mode 1, this procedure measures the flushing flow around each FIV. Initially steam generator feedwater is being supplied from a main feedwater pump via the feedwater preheater bypass valves. The FIBVs are then opened one at a time.

Immediately after each valve is opened flow through the FIBV piping is measured with an ultrasonic flowmeter.

At approximately 75% power, tempering flow for each steam generator is determined by computer point readings and calculations.

Adjustments are made to the tempering flow control system if the flow rates are too high or too low..

At approximately 90% and 100% power, proper split flow rate to each steam generator is verified. The flow rates are determined by computer point readings and calculations. In addition, alarms for high steam generator main nozzle flow are verified to be clean.

SUMMARY

OF RESULTS With the FIVs and FIBVs closed, operation of a main feedwater pump did not result in significant leakage past these isolation valves.

Only slight temperature changes were noted downstream of the FIVs and FIBVs.

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l 3.2.17 - MAIN FEEDWATER SYSTEM TEST. ISU-238B (Continued) l

SUMMARY

OF RESULTS (Continued) l The flushing flow rate around each FIV is as follows:

FLUSHING ~ f STEAM FLOW RATE l GENERATOR (lbm/hr) 1 79,202 2 68,730 3 79,202 4 82,148 .

Each of these flow rates is within the acceptable range of 60,000 to 120,000 lbm/hr.

As determined at approximately 75% power, the tempering flow rate  :

for each steam generator is as follows:

FLUSHING '

STEAM FLOW RATE GENERATOR (lbm/hr)

-i 1 274,000 2 203,000 l 3 243,000 ,

4 264,000 .

I These values satisfy the requirement that the tempering flow be l within the range 37,900 to 378,500 lbm/hr when main nozzle flow i falls below 85% of the rated feedwater flow rate of 3.785 mpph. l At approximately 100% power, the feedwater flow rate to each steam l generator's main nozzle is as follows:

MAIN NOZZLE STEAM FLOW RATE GENERATOR (mech)

I 1 3.166

2 3.208~

3 3.230 j 4 3.219 l Each of these values is less than the maximum allowable flow rate of 3.390 mpph. This data shows that feedwater will flow to the auxiliary (upper) nozzle at high flow rates and that the split flow bypass valve will modulate to prevent feedwater flow to the main (lower) nozzle from exceeding the design maximum. In addition, the alarms for high steam generator main nozzle flow were verified to be clear at both 90% and 100% power.

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-3.3 PHYSICS TESTING l 3.3.1 - INVERSE COUNT RATE RATIO MONITORING, (Initial Criticality  ;

Portion) - NUC-111 OBJECTIVE '

This permanent plant procedure is performed to obtain and evaluate i

nuclear monitoring data during the approach to criticality to  ;

ensure that the approach is done in a cautious and controlled manner. This procedure satisfies activities described in FSAR Section 14.2.10.2. ,

TEST METHODOLOGY r

s Neutron count rate data, as an indicator of core nuclear flux, from  !

both installed source range NIS channels is taken periodically during core reactivity additions. The sources of the core neutron flux are the installed primary neutron sources with associated ,

suberitical multiplication due to the loaded fuel lattice. As control rods are withdrawn and, later, as boron is removed from the l; RCS water, the core neutron flux-and source range channel count rates increase due to the reduction of these neutron absorbers in the core. When the count rates become very large, this indicates i

i that the reactor is approaching criticality. To determine the  !

effect of a given change on core reactivity, count rate data taken ,

after a neutron absorber decrease is compared to a reference value.

This comparison is performed as a ratio of the count rates to 3 i

evaluate the fractional change. If this ratio were to be very j large, it would indicate that this neutron absorber decrease 4 brought the reactor significantly closer to criticality. For  ;

convenience, the procedure evaluates the inverse of the count rate -

ratios (ICRR) such that an approach to zero would indicate an i approach to criticality. Additionally, this procedure trends the inverse count rate ratios and extrapolates the trends to predict  !

what additional neutron absortar decrease would be expected to result in criticality. Prior to the start of the approach to criticality, background counts are taken to allow the verification l j

of adequate source range channel signal to noise ratios. This data taken at nominally 557'F in Mode 3 is compared against similar data '

taken cold in Mode 6 at the end of core loading. This preliminary t cold data was taken in the earlier performance of NUC-111. An '

equation relating the cold and hot count rates to the signal to noise ratio is supplied by the core designers. This equation is l used to verify that a signal to noise ratio of at least two exists. ,

Reference values are redetermined just prior to the start of the  !

dilution with control rod banks withdrawn and also to renormalize '

the ICRRs. When an ICRR value falls below 0.3, it is renormalized (

by using the latest average count rate as the new reference value, l if necessary. This effectively resets the ICRR plot to a value of 1.0. {

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3.3.1 - INVERSE COUNT RATE RATIO MONITORING, (Initial Criticality Portion) - NUC-111 (Continued)

TEST METHODOLOGY (Continued)

Renormalization improves the resolution of the plot and may also be done at the discretion of the test engineer.

Count rate data is taken periodically during the approach to criticality. Data is taken, the ICRR calculated, plotted, trended and extrapolated as a function of bank withdrawal following each incremental control rod bank withdrawal, as specified by NUC-106B.

This is every 116 steps for shutdown banks and approximately every 50 steps for the control banks as withdrawn in normal overlap. 4 Count rate data is also taken, and the ICRR calculated, plotted, j trended and extrapolated during the dilution of RCS boron l concentration to initial criticality. The ICRR values are plotted, I extrapolated and evaluated as both a function of elapsed time

! during dilution and as a function of the quantity of reactor makeup water added.

SUMMARY

OF RESULTS Refer to Figures 3.3.2-1 and 3.3.2-2 for ICRR curves during the approach to initial criticality.

, All count rate data was properly recorded and ICRRs were f calculated, plotted, trended and extrapolated. The ICRRs show that the approach to criticality was performed in a cautious and controlled manner with no indicated unexpected approaches toward criticality. Renormalization was not necessary.

Monitoring data was properly taken and evaluated during the approach to criticality. Reference count rates were properly recalculated. The signal to noise ratios for Source Range Channels N31 and N32 both easily satisfied the >2.0 criterion.

3.3.2 - INITIAL CRITICALITY - NUC-106B OBJECTIVE This permanent plant procedure provides a method by which initial criticality is attained in a deliberate and controlled manner.

This procedure is used to enter Mode 2 for the first time in the fuel cycle. The sequence, frequency, and conditions for collection of nuclear data are specified as well as the method of analysis of this data. Criteria for suspending the approach to criticality and for emergency boration are also specified. This procedure satisfies activities described in FSAR Section 14.2.10.2 and Technical Specifications 3/4.10.3 and 3/4.1.1.1.

TEST METHODOLOGY Initial conditions are established with the RCS at an average temperature of approximately 557'F, RCS pressure at approximately 2235 psig, all control rod banks fully inserted, and RCS boron concentration greater than 300 ppm above the predicted all-rods-out critical concentration.

Procedures are initiated to monitor neutron flux, boron concentration and various other plant parameters for the duration of the test.

Reference counts are determined for each source range channel per NUC-111 . These values are used in the ICRR (Inverse Count Rate Ratio) calculations performed following reactivity additions.

Physics Testing is declared to be in progress to permit usage of Technical Specification Special Test Exception 3.10.3 with respect to Mode 2 testing at off normal conditions.

Shutdown banks are then withdrawn in their normal, alphabetical order. The withdrawals are made in increments of 116 steps or less and the value of the ICRR is determined after each withdrawal, prior to subsequent withdrawals. These ICRR values are plotted against the cumulative shutdown bank position to trend and predict by extrapolation any potential unexpected approach to criticality.

Each bank is withdrawn to an indicated 232 steps, the rod drive step counters are reset to the actual mechanical withdrawal limit of 231 steps, and the bank is reinserted to 228 steps. This sets the control rods to their proper full out heights as part of the monthly control rod repositioning to reduce localized control rod cladding wear.

l

e L 3. 2 - INITI AI. CRITICALITY - NUC-106B (Continued)

TEST METHODOLCGY (Continued)

The Mode 2 entry checklist is verified to be completed and the  ;

control banks are then manually withdrawn in their normal overlap configuration, in nominally 50 step increments. Mode 2 is entered with the initial withdrawal of Control Bank A. Control bank withdrawal is completed when Control Bank D is positioned at 180 steps. During control bank withdrawal, proper bank overlap and rod  ;

insertion limit alarm functions are verified. ICRR monitoring, )

plotting and extrapolation is also performed, as was done l previously for the shutdown banks. 1 The remaining reactivity insertion required to achieve criticality l is made by diluting the RCS boron concentration by addition of unborated reacter makeup water to the RCS. Periodic ICRR monitoring during the dilution is performed per NUC-111 to plot, l trend, and extrapolate predictions of expected time and quantity of  :

water added for initial criticality. The dilution rate is i initially such that RCS boron concentration is reduced at a rate of up to approximately 50 ppm per hour. When the RCS boron -

i concentration is within 100 ppm of the estimated critical boron  ;

concentration, the dilution rate is reduced to result in up to an  ;

approximate 30 ppm per hour rate of change. The RCS dilution then ]

continues until critical conditions, or near critical conditions, 1 are achieved. If criticality is not achieved during the dilution or subsequent mixing, batch boron dilution or Control Bank D motion may be used to bring the reactor critical. The core flux level is then increased to and stabilized at approximately 10- amps, as indicated on the Intermediate Range NIS channels using Control Bank D motion.

SUMMARY

OF RESULTS The Acceptance Criteria was met in that criticality was achieved and the neutron flux level was established within specified bounds on the Intermediate Range NIS chan_yls. The neutron flux was increased and stabilized at 1.5x10 amps _gn both Intermediate Range channels. This was approximately 10 amps, as required.

Shutdown bank withdrawals were completed without incident. Mode 2 was entered at 0425 hours0.00492 days <br />0.118 hours <br />7.027116e-4 weeks <br />1.617125e-4 months <br /> on 3-24-93 with Technical Specification Special Test Exception 3.10.3 invoked at 0428 hrs. Boron dilution was initiated at 0900 hours0.0104 days <br />0.25 hours <br />0.00149 weeks <br />3.4245e-4 months <br /> on 3-24-93 at a' rate of 34 ppm per hour.

)

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3.3.2 - INITIAL CRITICALITY - NUC-106B (Continued)

SUMMARY

OF RESULTS(Continued)

Within 100 ppm of the estimated critical boron concentration, the dilution rate was slowed to 21 ppm per hour. The dilution was discontinued at 2042 hours0.0236 days <br />0.567 hours <br />0.00338 weeks <br />7.76981e-4 months <br /> on 3-24-93 when the ICRR was less than 0.01. After a short period of RCS mixing, criticality was achieved at_g046 hours on 3-24-93. . The flux was stabilized at approximately j 10 amps at 2128 hours0.0246 days <br />0.591 hours <br />0.00352 weeks <br />8.09704e-4 months <br /> on 3-24-93. The stable critical data was as follows:

~0 Intermediate Range NIS channel currents: 1.5x10 amps RCS temperature: 557'F Control Bank D position: 147 steps l RCS boron concentration: 1011 ppm.

Refer to Figures 3.3.2-1 and 3.3.2-2 for plots of ICRR versus bank withdrawal and reactor makeup water addition.

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Figure 3.3.2-1 ICRR During RCC Bank Withdrawal CPSES Unit 2, Cycle 1 initial Crit.

ICRR --

1. 2 3,3 _.

g C1 ==k mC i' + gg :4('

F " P -

1 O.9 - --

k 0.8

-~ -

O.7 - - -

0.e -- --

- -- -- - - I^ I- i 0.6 0.4 0.3 0.2 0.1 - - - - --

0 - - - -- - --

0 100 200 300 400 600 600 700 800 900 1000 1100 1200 1300 1400 1500 1600 1700 Total Steps Withdrawn

- - N-31 l - N-32 a

03/23/93 .

Figure 3.3.2-2 ICRR During RCS Boron Dilution ,

CPSES Unit 2, Cycle 1 initial Crit.

ICRR -- -- -- -- -

1. 2 1.1 - - - - -- -

0.9 ---

bi'.t. 4 ,gg o.s - -

T k.f kp kj-( - - - - - --

0.7 4d 4 'H+. s 0.6 --

,S {.-

0.6

--- 1:( --

0.4 -

i 0.3 A

0.2 O.1 --

i --

0 -

6 6 7 8 9 10 11 12 13 14 16 16 17 18 19 20 l 0 1 2 3 4 RMUW Added (gallons x 1000)

- - - N - 31 l- N-32 i

03/24/93

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

l 3.3.3 - DETERMINATION OF CORE POWER RANGE FOR PHYSICS TESTING, NUC-109 OBJECTIVE l This permanent plant procedure is used to determine the power level I I

(neutron flux level) at which detectable reactivity feedback l effects from nuclear fuel heating occur and to establish the range '

l of neutron flux in which zero power reactivity measurements are performed to avoid interference with these feedback effects.

TEST METHODOLOGY Initial conditions are established with the RCS at an average i temperature of approximately 557'F, RCS pressure at approximately8 2235 psig and the reactor critical with flux at approximately 10 amps on both Intermediate Range channels. Control Bank D is positioned such that approximately 40 pcm of reactivity worth is available to increase core reactivity. .

1 Initially, the reactivity monitoring system is set up using a power range channel detector, in this case channel N41, which is removed from service. Reactivity monitoring system outputs of reactivity and flux along with RCS cold leg temperature are displayed on strip

• chart recorders. The temperature input to the reactivity monitoring system comes from a process instrumentation rack f isolated output.

The determination of the power range for physics testing is made by withdrawing Control Bank D to achieve a positive reactivity addition of approximately 20 to 40 pcm. Reactivity and flux level are then observed to determine the point of adding nuclear heat as indicated by negative reactivity addition from the Doppler fuel temperature coefficient. RCS temperatures, pressurizer level, and startup rate are also monitored for an increase or decrease as an indication of nuclear heating. The flux is then teduced back to approximately 10-a amps and the measurement is repeated, at least once, to confirm the value. j

SUMMARY

OF RESULTS Two measurements were performed and the results were similar. The observed reactor power levels at which detectable reactivity I

feedback effects from nuclear heating occurred were as follows:

Indication Test 1 Test 2

~

N35 Intermediate Range Current (amps) 6 x 10 6. 5 x 10~

~ ~

N36 Intermediate Range Current (amps) 5 x 10 6 x 10

~

Picoammeter Flux Level (amps) 8.1 x 10 9 x 10~

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3.3.3 - DETERMINATION OF CORE POWER RANGE FOR PHYSICS TESTING, NUC-109 (Continued)

SUMMARY

OF RESULTS (Continued)

The neutron flux level range at which zero power reactivit measurements were to be performed was determined to be 0.20 x 10z amps to 2.0 x 10-7 amps as indicated on the reactivity monitorina system. This range corresponds to the 200 nanoamp (200 x 10 -*

amps) range setting on the reactivity monitoring system input ,

picoammeter. This range was suf ficiently below the nuclear heating level and above the gamma noise region.

Excessive noise on the reactivity signal, particularly at the low end of the flux range, was observed at the start of this test. The '

problem was traced to an input signal that is optionally used to compensate for gamma noise. Upon disconnection of this compensating input signal, the noise in the reactivity signal decreased substantially. The compensating input signal was unnecessary for further testing because the determined flux level range for zero power testing was above the gamma noise region.

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I 3.3.4 - ABB/CE REACTIVITY MONITORING SYSTEM CHECKOUT - NUC-210 OBJECTIVE This permanent plant procedure is performed to demonstrate proper operation of the reactivity computer through dynamic testing using actual neutron flux signals and core reactivity changes. This ensures that the reactivity computer is operating properly before it is used to measure reactor physics parameters.

TEST METHODOLOGY A reactivity increase of approximately 25 pcm, as shown on the reactivity computer, is initiated by withdrawal of Control Bank D.

A stopwatch is used to measure the time corresponding to a change in flux level in order to calculate reactor period. This period, P, is the time interval over which indicated core flux increases by a factor of e, with flux increasing on a stable period after the decay of f of initial transient ef fects. The period comes from the following equation it. terms of the initial flux, e t, final flux, a f, and measured time interval, t:

P = C/lnb0 1 This measured period is used to determine the theoretical reactivity increase using Startup and Operations Report predictions of reactivity as a function of reactor period. This prediction comes from solutions to the "inhour equation" using actual core physics constants from the Startup and Operations Report.

The predicted reactivity increase is compared to the reactivity indicated on the reactivity computer. This measurement may be repeated for reactivity increases of up to approximately +50 pcm.

An optional negative reactivity insertion of up to -20 pcm may also be performed.

3.3.4 -

ABB/CE REACTIVITY MONITORING SYSTEM CHECKOUT -

NUC-210 (Continued)

SUMMARY

OF RESULTS  !

Two runs were made, one each at approximately +25 and +40 pcm. No negative reactivity insertion runs were made. The acceptance criterion for this messurement is that the average of the absolute values of the reactivity differences be less than +4%. The results were as follows:

Measured Predicted Indicated Approximate Reactor Reactivity Reactivity Absolute '

i Reactivity Period Based on from Value of Insertion (ocm) (seconds) Period (Dem) Comouter (Demi %Dif ference 25 292.43 25.64 25.5 0.55 40 167.36 41.82 41.5 0.77 Average % Difference 0.66 i

i The average absolute difference of 0.66% satisfied the < +4% l criterion.

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3.3.5 - CORE REACTIVITY BA;ANCE - NUC-205 l OBJECTIVE The purpose of this permanent plant. procedure is to verify the design predictions of core reactivity during the power ascension l startup testing sequence. This ~ procedure satisfies activities described by FSAR Table 14.2-3, Sheet 17.

TEST METHODOLOGY This core reactivity verification is performed by comparing reactor  !

criticality parameters at zero power with those at full power.

Parameters measured include control bank positions, RCS temperature, RCS boron concentration, power level and core burnup.  ;

After compensating for differences in control bank position, boron j concentration, reactor power and Xenon and Samarium buildup, with '

respect to predicted core conditions, the actual critical boron  :

concentration present is compared to the design prediction. This verifies the accuracy of the design predictions of core reactivity. l

SUMMARY

OF RESULTS The Hot Zero Power, All Rods Out, Xenon'and Samarium free critical  ;

boron concentration was 1022.5 ppm. The Hot Full Power, All Rods-  !

Out, equilibrium Xenon and Samarium critical boron concentration j was 668.0 ppm at an average RCS temperature of 588.3*F and 1092.7 i MWD /MTU burnup. This represents a decrease of 354.5 ppm in boron i' concentration to get from Hot Zero Power, All Rods Out and no fission product poisons to just critical at Hot Full Power, All  :

Rods Out and equilibrium Xenon and Samarium. l t

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3. 3. 6 - ZERO POWER ISOTHERMAL AND MODERATOR TEMPERATURE COEFFICIENT MEASUREtiEFTS - NUC-207 OBJECTIVE This permanent plant procedure is performed to determine the Isothermal Temperature Coefficient (ITC) of reactivity and to infer, from this, the Moderator Temperature Coefficient (MTC) of reactivity at the beginning of core life. This procedure satisfies activities described by FSAR Table 14.2-3, Sheet 14 and Technical Specification 3/4.1.1.3.

TEST METHODOLOGY The ITC is determined by measuring the change in reactivity induced by simultaneously, uniformly changing the temperature of the moderator, cladding and fuel and dividing by the temperature change. The MTC is obtained by analytically removing a precalculated Doppler broadening fuel temperature coefficient factor from the ITC value to eliminate the cladding and fuel temperature change portions of the ITC. The remainder is the effect due to the moderator alone. ,

A voltage signal proportional to core reactivity is obtained from the reactivity computer output and a voltage signal proportional to RCS cold leg temperature is obtained from a process instrumentation rack isolated output. These signals are input to an X-Y plotter such that the slope of the X-Y plot corresponds to the ITC, change in core reactivity per unit change in RCS temperature. RCS temperature is slowly changed by manipulation of the rate of heat removal from the RCS by the secondary plant. The resulting reactivity as a function of the varying temperature is plotted and evaluated. By changing the RCS temperature slowly, the fuel, cladding and moderator temperatures all change at the same rate, nearly isothermal, with minimal temperature gradients. This measurement result must be compensated to eliminate the effect of Doppler resonance peak broadening in the fuel and the cladding to yield the effect of the moderator alone. The slow RCS temperature change permits the fuel temperature to change uniformly, isothermally, without the heat transfer that would result in a temperature profile to form within the fuel pellets. Because of this, the fuel temperature at a given time is essentially the same as the RCS temperature. This allows a fuel type and enrichment specific calculation of the Doppler broadening effect to be performed for the temperature regime at which the test is executed.

The correction for reactivity change due to cladding temperature change is very small and is included in the fuel value.

Unmeasurable variable fuel temperature distributions would render predictions of the isothermal temperature coefficient impossible.

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3. 3. 6 - ZERO POWER ISOTHERMAL' AND MODERATOR TEMPERATURE COEFFICIENT MEASUREMENTS - NUC-207 (Continued)

TEST METHODOLOGY (Continued) l 1

The calculated Doppler broadening coefficient of  ;

-1.83 pcm/*F is subtracted from the average ITC value from at least  !

two temperature changes to result in the MTC. The measured ITC values are evaluated to verify they are within 21 pcm/*F of each other, to demonstrate data consistency, and the average ITC value j is verified to be within 23 pcm/*F of the predicted ITC value of l

-0.82 pcm/*F. The MTC is verified to be less than +5 pcm/*F or, if  !

not, Rod Withdrawal Limits, using NUC-116, would be imposed to administrative 1y ensure that the MTC is maintained less than +5 pcm/*F.

This test is performed from Hot Zero Power conditions, nominally 557'F, starting with a Reactor Coolant System (RCS) cooldown of approximately 3*F at a rate of up to approximately 10*F/hr. After a stabilization period at this lower temperature, an RCS heatup is then initiated for an approximate 3*F increase, also at a rate of up to approximately 10*F/hr. A plot of reactivity vs. temperature is made for both the cooldown and heatup portions of the test. The cooldown and heatup are performed at the All Rods Out (Control Bank D2 200 steps) control rod configuration. Multiple cooldown and heatup cycles may be performed, if required to ensure data consistency.

i

SUMMARY

OF RESULTS The cooldown resulted in a measured ITC of -1.08 pcm/*F. The heatup also resulted in a measured ITC of -1.08 pcm/*F. Only one cooldown and heatup cycle was performed. The average ITC was -1.08 pcm/*F. The 11 pcm/*F criterion between the cooldown and heatup values was satisfied, as the values were identical in sign and magnitude. The 13 pcm/*F criterion between average measured ITC and the prediction of -0.82 pcm/*F was satisfied as they differed by only -0.26 pcm/*F. The calculated MTC value of -1.08 -(-1.83)

= +0.75 pcm/*F satisfied the less than +5 pcm/*F criterion so Rod withdrawal Limits were not imposed.

Control Bank D was at 2200 steps during this test performance with all other control rods fully withdrawn. Core neutron flux was maintained below the point of nuclear heat addition to preclude nuclear heating feedback effects from invalidating test results.

Refer to Figure 3.3.6-1 for the plot of reactivity vs. temperature.

-100-

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3.3.7 - ROD SWAP MEASUREMENT - NUC-120 OBJECTIVE This permanent plant procedure is performed to verify that the differential and integral worth of individual control rod ~ banks agree with the design predictions made in the Startup and Operations Report. This procedure satisfies activities described by FSAR Table 14.2-3, Sheet 15.

TEST METHODOLOGY The rod swap method, also called the bank exchange method, of determining bank worth only directly measures one bank. This bank is designated as the reference bank, and the worth of the other banks are inferred barad on the reference bank worth. Core design calculations determined Control Bank B to be the reference bank for this initial fuel cycle. Starting with the reactor critical and Control Bank D partially inserted to control core neutron flux, Control Bank D is first fully withdrawn. The reference bank, Control Bank B, is then immediately inserted to restore the core to the just critical condition. An endpoint measurement is made for the inserted worth of the reference bank by fully withdrawing it, measuring the core reactivity change with the reactivity monitoring system, and again reinserting the bank to the just critical position. Next, an RCS dilution at a rate of approximately 25 gpm is started. This is roughly equivalent to a 300 pcm/ hour positive reactivity addition rate. This positive reactivity addition is compensated by periodic insertions of the reference bank in -10 to

-20 pcm increments in order to maintain a near critical reactivity condition. The individual worth of these incremental insertions is measured with the reactivity monitoring system. The dilution is suspended with the reference bank nearly fully inserted. Another endpoint measurement is made for the last portion of reference bank worth by fully inserting the reference bank to core bottom and measuring this worth with the reactivity monitoring system. The sum of the incremental worths and the two endpoint worths is the integral worth of the reference bank. The differential worth is calculated as the incremental worth divided by the number of rod steps moved to result in each reactivity change. Both of these worth values are recorded and plotted.

With the reference bank nearly fully inserted and all other rod banks fully withdrawn, the integral worth of the other rod banks are individually verified by comparing their relative worths with respect to that of the just measured reference bank. A selected test bank is inserted to result in approximately 20 pcm of negative core reactivity. The reference bank is then immediately withdrawn to result in approximately 20 pcm of positive core reactivity.

This process is repeated until the test bank is fully inserted and the reference bank is adjusted to the just critical position.

-102-

3.3.7 - ROD SWAP MEASUREMENT - NUC-120 (Continued) l TEST METHODOLOGY (Continued)

The worth of the test bank is then inferred as being equal to the fractional portion of the reference bank worth that was withdrawn to compensate for the test bank's insertion. The test and

]

reference banks are returned to their initial positions, reference bank in and the test bank fully withdrawn, in the reverse order of l steps used for the measurement. The same process is then repeated for each remaining bank until all banks have been exchanged, or swapped, against the reference bank.

I Following the exchange of all banks against the reference bank,

! core conditions are restored. The reference bank is exchanged against Control Banks D and C until the reference bank is fully withdrawn. Control Banks D and C are then adjusted to their l

desired positions, and proper overlap, by boration or dilution.

1

SUMMARY

OF RESULTS The reference bank measured and predicted integral worths were to ~

differ by no more than 210%. They differed by only -2.3%. The remaining test banks measured and predicted worths were to differ j by no more than 15% or 100 pcm, whichever was greater. All test banks satisfied this criterion. The largest percent difference was

-6.8% on Shutdown Bank E. The largest reactivity difference was 28.6 pcm, also on Shutdown Bank E. The percent error between measured and predicted total worths was -2.9%. This satisfied the i criterion that the measured total worth be within 210% of the predicted total worth. A summary of the bank worth measurements and the predicted values appears in Table 3.3.7-1. Figure 3.3.7-1 is a plot of the integral and differential reference bank worths.

-103-

i

?

+

TABLE 3.3.7-1 Measured and Inferred Versus Predicted Rod Bank Worths ,

MEASURED / '

INFERRED PREDICTED ABSOLUTE BANK WORTH (ocm) WORTH (Demi- DIFFERENCE (ocm) % Dif ference ;

Shutdown A 647.9 661.8 -

-13.9 -2.1 Shutdown B 759.6 769.4 -9.8 -1.3 Shutdown C 478.9 494.0 -15.1 -3.0 Shutdown D 472.6 493.4 -20.8 -4.2 Shutdown E 390.4 419.0 -28.6 -6.8 Control A 327.6 347.6 -20.0 -5.8 Control B* 814.0 833.0 -19.0 -2.3 Control C 795.2 805.8 -10.6 -1.3 Control D 767.8 795.5 -27.7 -3.5 Total 5454.0 5619.5 -165.5 -2.9

• Control Bank B was the reference bank and has a measured worth.

All other banks have inferred worths.  ;

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3.3.8 - BORON ENDPOINT DETERMINATION AND DIFFERENTIAL BORON WORTH

- NUC-104 OBJECTIVE This permanent plant procedure is performed to determine the last portions of rod worth at the extreme ends of rod bank travel, at the near fully withdrawn or near fully inserted positions. This procedure also calculates the average differential boron worth over the reference bank based on endooint data at both ends of reference bank travel. In addition, the'just critical, All Rods Out (ARO),

Reactor Coolant System (RCS) boron concentration is determined.

This procedure satisfies activities described by FSAR Table 14.2-3, Sheet 16.

TEST METHODOLOGY The test starts with RCS temperatures and boron concentration verified stable and all control rods withdrawn except for Control Bank D, which is controlling flux at the just critical condition.

With no more than approximately 50 pcm of Control Bank D negative reactivity worth inserted, Control Bank D is then fully withdrawn to reach the desired ARO endpoint configuration while neutron flux and reactivity are monitored on a strip chart recorder. When the reactivity trace stabilizes, Control Bank D is repositioned to re-establish the initial flux level and core reactivity. This process is repeated at least one more time. The endpoint boron concentration is obtained by dividing the measured reactivity change due to Control Bank D withdrawal by the design prediction for differential boron worth at this particular rod bank configuration. This converts the measured pcm of reactivity worth from the reactivity monitoring system to ppm of equivalent boron worth. This boron worth value is combined with the actual measured boron concentration to vield the boron concentration that would exist if Control Bank D were fully withdrawn with the reactor just critical.

Differential boron worth over a particular bank is obtained by dividing the total integral worth of the selected rod bank by the difference in the endpoint boron concentrations, one for the bank fully withdrawn and one for the bank fully inserted. This results in a value of pcm/ ppm and is always negative.

SUMMARY

OF RESULTS The All Rods Out just critical RCS boron concentration was found to be 1022.5 ppm which satisfied the acceptance criterion of 992 +50 Ppm.

Differential boron worth was based on boron endpoint and reactivity data taken when the reference bank was diluted into the core during i NUC-120. The differential boron worth was determined to be -12.92 pcm/ ppm. The design value is -12.85 pcm/ ppm. This difference was only 0.5%, well within the :15% criterion.

-106-l 1

i 3.4 - TD N IENT' TESTING j 3.4.1 - TURBINE GENERATOR TRIP WITH COINCIDENT LOSS OF OFFSITE  !

POWER - ISU-222B l OBJECTIVE i

This test is performed to verify the plant's ability.to safely  !

sustain a turbine generator trip with' no offsite- power available - .

for at least thirty minutes. This test satisfies activities  !

described by FSAR Table 14.2-3, Sheet 18.

• TEST METHODOLOGY -!

The test is initiated with Unit 2 in Mode 1, at greater than 10% j reactor power, with the main generator output at approximately 130. '

MWe. All normal 6.9 kV electrical buses are initially energized by the Unit 2 Auxiliary Transformer (2UT). Their alternate source, Startup Transformer 2ST, is locked out, preventing a designed i automatic bus transfer to.this backup supply. Class 1E Safeguards l Busses 2EA1 and 2EA2 are initially aligned to Startup Transformer  ;

XST1, with their. alternate source, Startup Transformer XST2, j locked-out. The Unit 2 turbine is manually tripped and the offsite  ;

power feeder breakers to the Unit 2 Class 1E busses are opened.  !

This results in an immediate loss of Unit 2 Class 1E AC power and  ;

a loss of Unit 2 non-Class 1E AC power when the Unit 2 main j generator trips, approximately 10 seconds after the-turbine trip. ,

The 10 seconds is due to normal protective relaying time delays. I As part of the prerequisites, selected secondary plant equipment is l shut down or. transferred to uninterruptible or Unit 1 power i supplies to prevent possible damage as a result of the loss of  !

power. This equipment, typically lube oil pumps etc. , is non- '

safety related and would not be powered from the emergency diesel  ;

generators as part of this test. Therefore, these actions have no l adverse impact on test results.  ;

I Plant conditions are monitored to ensure that the standby emergency I diesel generators start and re-energize the safeguards buses _and that plant equipment functions properly to stabilize the Reactor Coolant System (RCS) in a Mode 3, hot standby condition and  ;

maintain it in that condition for at least 30 minutes. Also i monitored is the ability of the Steam Generator Atmospheric Relief l Valves.to control steam line pressures below 1185 psig for at least l 30 mHiu M . '

4 J

-107-i

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

t r

3.4.1 - TURBINE GENERATOR TRIP WITH COINCIDENT LOSS OF OFFSITE POWER - ISU-222B (Continued)

SUMMARY

OF RESULTS

'The alignment of Unit 2 power supply breakers was completed, the "

turbine was manually tripped, and. simultaneously, the proper offsite . feeder breakers were opened. . Both standby emergency diesel generators started and powered- the safeguards buses. The i safeguards sequencers both loaded the. required plant equipment onto l the safeguards buses at the proper times. Stabilization of.and i recovery from this event was performed in accordance with the '

permanent plant emergency operating and abnormal operating procedures. ,.

The safeguards buses were energized and plant equipment functioned properly to stabilize and maintain the RCS in a safe shutdown  !

condition. This conc _ tion was maintained for 35 minutes satisfying the >30 minute criterion. The non-Class 1E buses remained de--

energized for the duration of this test. RCS hot leg, cold leg and l core exit thermocouple readings were verified to stabilize-following the initiation of the transient indicating that naturel circulation cooling was established. RCS:subcooling was verified to be greater than 60*F.

Even though the Main Steam Isolation Valves'were closed early in i the event, in accordance with the permanent plant procedure used to .i stabilize the plant, main steam line pressures never exceeded the 1125 psig setpoint where the Steam Generator Atmospheric Relief .

Valves begin to automatically open. Therefore, the limit of 1185 1 psig for maximum main steam line pressure was never exceeded. 'The atmospheric relief valves were opened manually to initiate natural circulation cooling as prescribed by the permanent plant procedure l used for plant recovery. j The following is a summary of indicated plant conditions during the )

event: ,

J3.gE Maximum Value Minimum Value RCS Cold Leg Temp. 551.3*F 535.8'F ,

RCS Hot Leg Temp. 561.6*F 550.0*F ,

Pressurizar Pressure 2240 psig 2123 psig i Main Steamline Press. 1045 psig 892 psig l During the transient, only one unexpected event occurred. The .i Train A Blackout Sequencer display indicated that sequencer step 7 l occurred at 60 seconds instead of the expected 65 second delay. i Step 7 is associated with the start cf HVAC Centrifugal Water i Chiller 03. This had no overall impact on test results because  !

this chiller had been secured as part of the test prerequisites. J It was later determined by field troubleshooting -under a Work Order, and an evaluation of sequencer outputs, that the problem was associated with the sequencer display only, the relay functions associated with step 7 actually occurred at the proper 65 second time.

-108-

1 3.4.2 - DESIGN LOAD SWING TESTS - ISU-231B OBJECTIVE This test is performed _to demonstrate-the dynamic-response of the Reactor Coolant System '(RCS) . and- the Rod Control System to !

automatically bring the plant to steady state conditions following a rapid reduction in turbine load, and then to a rapid increase in turbine load.

magnitude.

The rapid load changes are approximately- 10% in 3 FSAR Table This test partially satisfies activities described by 14.2-3, Sheets 23 and 24. l TEST METHODOLOGY With plant conditions stable at approximately 50% and 75% power, a  ;

10% load decrease is manually initiated from the turbine-generator. i Electro-Hydraulic Controls (EHC) at a rate of approximately 200%

power / minute. Plant parameters are allowed to stabilize and, after ,

stabilization, a 10% load increase is manually initiated. Plant ,

parameters are again allowed to stabilize. The' load decrease is i performed by manually reducing the turbine generator load limit setpoint to a value approximately 10% in power below the initial l load reference operating power level. The . load increase is '

performed by manually raising the load limit setpoint back above ,

the original load reference operating power level. This allows the i load to increase back to its . original value at the start of the I test. The load limit setpoint adjustment occurs at a rate of  ;

approximately 200% power / minute and is performed by main control board manual push button operation of a motor _ driven potentiometer that is set to move at that rate. These push buttons are permanent plant control features and the related circuitry is closely associated with the built-in turbine generator runback circuits.

In fact, prior to initiation of the-10% load increase, the runback circuits are temporarily bypassed by actuating a switch inside the EHC cabinets. This is done to disable load increase inhibiting circuitry that is activated by the 10% load decrease via the shared runback circuit portions. The 10% power load changes are nominal values and are actually specified to be 10% 12% in magnitude. The 10% load change may result in reactor power changes of greater than 10% power due to relatively low plant efficiency at lower power levels.

During the course of the test, plant process computer and Data ,

Acquisition System (DAS) recordings of key plant parameters are taken so that plant response can be analyzed. The principal )

parameters monitored included RCS Tavg, Tcold, Tref, pressurizer i pressure and level, steam generator pressures and levels, steam and feedwater flows, control rod positions and speed, OTN16 and OPN16 setpoints, reactor power, feedwater pump speed and discharge pressure, N16 power, safety and relief valve positions, and steam dump valve positions.

-109-f$'"""-"y 1r? --ei- --y_. --- y w - y .-__ _,_,,___ _ m _ _,__ _ , _ _ , _ _ , , , , , _ _ , _ , _ , _ _ , , _ _ , , , _ _ _ _ _ _ , , _ _ , , , , , _ _

R 3.4.2 - DESIGN ~' LOAD SWING TESTS - ISU-231B (Continued) 4

SUMMARY

OF RESULTS The first test performed was at L the 50% power. plateau from 46% ,

reactor power. The second test execution was from.78% reactor '

power as part.of the 75% powerLsequence. Both test executions satisfied the following criteria:

• The load decreases and increases did not cause the reactor to l trip nor the turbine to trip. l

• Safety injection did not initiate.

• The steam generator. safety or atmospheric relief valves and pressurizer safety or power operated relief valves did not lift during any of the load swings. .

i e Nuclear power over/undershoot was less than 3%.

• No manual intervention was required to bring plant conditions ,

to steady state.

• Plant variables returned to steady state conditions without 1 sustained or diverging oscillations.

Numerical acceptance and review criteria are summarized on. Table 3.4.2-1.  ;

During the first test performance, from 46% power, the magnitude d of the load decrease was approximately 11% turbine generator power, ,

which adequately approximated a 10% load change. Equilibrium Tavg- ,

was reached approximately 9 minutes following initiation of the a load decrease. The subsequent load increase was also approximately 11% power in magnitude and equilibrium Tavg was also reached in approximately 9 minutes.

There were only two problems noted while performing the test at the ,

50% power plateau. The plant process computer failed to capture '

1 all of the data that it was believed to have been set up to monitor. This had no adverse impact on test results because the .  !

DAS was monitoring identical or related points. The DAS data was  !

sufficient to compensate for the missing plant computer data. The  ;

other problem was found during the analysis of the data trends j recorded in the test. The response of the main feedwater pump i

, speed controller was judged to have been sluggish. This controller  ;

was adjusted and later satisfactorily tested as part of the ISU-  !

207B test, Section 3.2.2 of this report.

I During the second test performance, from approximately 75% power, the load decrease was approximately 13% power in magnitude and equilibrium Tavg was reached in approximately 25 minutes. The load '

increase was also approximately 13% power in magnitude and equilibrium Tavg was also reached in approximately 25 minutes. i

-110-

. _ .~. . .- __- .

3.4.2 - DESIGN LOAD SWING TESTS - ISU-231B (Continued)

SUMMARY

OF RESULTS (Continued)

During the test performed at 78% power, there were four problems noted:

• The firat bank of three steam dump, or turbine bypass, valves momentarily opened as a result of the rapid load decrease. The steam dumps are armed by a 10% or greater decrease in turbine impulse chamber pressure. The turbine impulse chamber pressure is proportional to turbine generator power. The 75% power test had a load decrease of approximately 13% which allowed the steam dumps to arm. Once armed, the steam dump valves modulate to force average RCS temperature (Tavg) to within 5'F of the load varying reference RCS average temperature (Tref). Tavg was only marginally more than 5*F greater than Tref. Which is why only three of the twelve steam ,

dump valves modulated open for a short time, less then  !

one minute. This steam dump actuation did not invalidate the test results because the rod control system properly responded to and stabilized the load decrease transient.

The steam dumps aided the rod control system, as they are designed to do, during the initial portion of the transient. This test performance had a reactor power change in excess of 10%, the rod control system is designed to respond to absorb a nominal 10% power change.

The assistance of the steam dumps to absorb the excess over 10% was proper.

• Prior to starting the load decrease it was noted that Tavg was being controlled approximately 2.5'F above Tref as opposed to keeping the Tavg-Tref mismatch within a :

1.5'F range. This constant offset did not adversely impact test results based on the ability of the rod control system to respond to and stabilize Tavg during the load swings. The offset is to be corrected via a work order.

• Following the load decrease, and prior to the load increase, steam generator blowdown flow was reduced from 500 gpm to 130 gpm based on blowdown line waterhammer concerns and one extraction steam line automatically isolated due to high water level in one feedwater heater.

This did not adversely impact test results as they occurred during the stabilization period between the transients and, therefore, did not alter the transient plant performance parameters which are being tested here.

-111-

i i

3.4.2 - DESIGN LOAD SWING TESTS - ISU-231B (Continued)

SJH24ARY OF RESULTS (Continued)

• During the load increase portion of the test, Tavg decreased but - did not - f ully recover to match Tref : I 1.5'F. This was due to the controlling bank becoming  ;

nearly fully withdrawn during the load increase, thereby l possessing very little reactivity worth to help compensate for the power change. Control Bank D correctly stopped withdrawing at 222 steps, the automatic l rod control withdrawal stop. The plant stabilized at a i reduced Tavg to balance core reactivity. This had no -

adverse impact on overall test results based on proper responses of the rod control system with respect to demanded rod speed and direction as a result of the ,

imposed transient. ,

Refer to Tables 3.4.2-2 through 3.4.2-5 for additional detailed [

data.

1 4

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I s

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-112-

5 i

TABLE 3.4.2-1 i DESIGN LOAD SWING TESTS

SUMMARY

Nuclear  ;

Power. Pressurizer i Power Load Over/Under- Allowed Pressure Allowed' l Plateau (%) Swina shoot (%) Limit (%) Swina(osia) Limit (Dsia) j I

50 Decrease 2 <3 +15,-17 <1100 Increase 2 <3 +6,-22 <t100 ,

75 Decrease 2.5 <3 +13,-15 <t100 -

Increase 0 <3 +22,-29 <1100  !

6 i

i Steam Header i Steam Gen. Pressure Over/ 1 Power Load Undershoot Level Allowed Allowed Plateau (%) Swina Swinc(%) Limit (%) (osia) Limit (Dsia)  !

50 Decrease +4,-4 5 115 44 170 l Increase +7,-7 i 115 51 170

• 75 Decrease +3,-3 1 110 38 170 Increase +3,-3 1 110 15 170 Power Load Tavg Over/ Allowed  !

Plateau (%) Swina Undershoot(*F1 Limit (*F) ,

50 Decrease 0.5 12.0 Increase 1.0 12.0 ,

75 Decrease 1.0 12.0 Increase 0 12.0 t

i

-113-

TABLE 3.4.2-2 10% LOAD DECREASE AT 50% POWER

SUMMARY

INITIAL CONDITION FINAL CONDITION i Generator Load (MWe) 425 300 Nuclear Power (%) 44 32  ;

Tavg Auctioneered (*F) 570.5 565.9

{

570.6 566.9 Tref ( F)

N-16 Power (%) 47.3 35.9  :

OPN16 Setpoint (%) 112.1 L12.1 1 OTN16 Setpoint (%) 121.4 123.6 Pressurizer Pressure (psig) 2238 2240 l Pressurizer Level (%) 40.9 35.9 Steam Generator Level Loop 1 (%) 64 64 l Steam Generator Level Loop 2 (%) 64 64  ;

Steam Generator Level Loop 3 (%) 64 64 l Steam Generator Level Loop 4 (%) 64 64 {

Steam Header Pressure (psig) 1015 1025 Steam Flow Loop 1 (pounds / hour) 1.44E6 1.08E6 i Steam Flow Loop 2 (pounds / hour) 1.42E6 1.06E6 }

Steam Flow Loop 3 (pounds / hour) 1.47E6 1.11E6  :

Steam Flow Loop 4 (pounds / hour) 1.41E6 1.05E6 l Feedwater Flow Loop 1 (pounds / hour) 1.58E6 1.15E6 l Feedwater Flow Loop 2 (pounds / hour) 1.54E6 1.11E6 Feedwater Flow Loop 3 (pounds / hour) 1.62E6 1.20E6 Feedwater Flow Loop 4 (pounds / hour) 1.63E6 1.23E6 Feedwater Temperature Loop 1 (*F) 370 345 l Feedwater Temperature Loop 2 (*F) 365 345 Feedwater Temperature Loop 3 (*F) 370 345 Feedwater Temperature Loop 4 (*F) 375 350 Feed Pump Discharge Hdr Pressure (psig) 1116 1119  !

Control Bank D Position (steps) 176 145 j Control Bank C Position (steps) 227 227 Feedwater Pump 1-A Speed (rpm) 4612 4243 Feedwater Pump 2-B Speed ( rpm) 2062 2059 i i

-114-

7 i

TABLE 3.4.2-3 10% LOAD INCREASE AT 50% POWER

SUMMARY

l INITIAL CONDITION FINAL CONDITION Generator Load (MWe) 300 429 l Nuclear Power (%) 32 46 I

Tavg Auctioneered (*F) 565.9 569.8 566.9 570.6 Tref ( F)  :

N-16 Power (%) 35.9 47.6 l OPN16 Setpoint (%) 112.1 112.1  !

OTN16 Setpoint (%) 123.6 123.1 Pressurizer Pressure (psig) 2240 2237 l Pressurizer Level (%) 35.9 39.6 Steam Generator Level Loop 1 (%) 64 64  ;

Steam Generator Level Loop 2 (%) 64 64  !

Steam Generator Level Loop 3 (%) 64 64 l Steam Generator Level Loop 4 (%) 64 64 j Steam Header Pressure (psig) 1022 1020 4

Steam Flow Loop 1 (pounds / hour) 1.08E6 1.44E6 ,

Steam Flow Loop 2 (pounds / hour) 1.06E6 1.42E6  ;

Steam Flow Loop 3 (pounds / hour) 1.11E6 1.48E6 Steam Flow Loop 4 (pounds / hour) 1.05E6 1.42E6 Feedwater Flow Loop 1 (pounds / hour) 1.15E6 1.59E6 j Feedwater Flow Loop 2 (pounds / hour) 1.11E6 1.58E6  ;

Feedwater Flow Loop 3 (pounds / hour) 1.20E6 1.61E6 ,

Feedwater Flow Loop % (pounds / hour) 1.23E6 1.63E6 l Feedwater Temperature Loop 1 (*F) 345 365 f Feedwater Temperature Loop 2 (*F) 345 370 )

. Feedwater Temperature Loop 3 (*F) 345 370 Feedwater Temperature Loop 4 ('F) 350 375 l Feed Pump Discharge Hdr Pressure (psig) 1116 1120 Control Bank D Position (steps) 145 192 j Control Bank C Position (steps) 227 227 i i

Feedwater Pump 2-A Speed (rpm) 4243 4596 Feedwater Pump 2-B Speed (rpm) 2059 2060 l

i

-115- l I

L TABLE 3.4.2-4 10% LOAD DECREASE AT 75% POWER

SUMMARY

INITIAL CONDITION FINAL CONDITION Generator Load (MWe) 885 730 Nuclear Power (%) 78 66 Tavg Auctioneered (*F) 583.9 579.5 Tref (*F) 583.0 578.4 i

N-16 Power (%) 78.8 67.4 OPN16 Setpoint (%) 112.1 112.1 l OTN16 Setpoint (%) 112.1 116.0 Pressurizer Pressure (psig) 2224 2233 Pressurizer Level (%) 56.5 50.4 Steam Generator Level Loop 1 (%) 64 64 Steam Generator Level Loop 2 (%) 64 63 Steam Generator Level Loop 3 (%) 64 64 Steam Generator Level Loop 4 (%) 64 64 Steam Header Pressure (psig) 1018 1019 Steam Flow Loop 1 (pounds / hour) 2.91E6 2.44E6 Steam Flow Loop 2 (pounds / hour) 2.91E6 2.43E6 Steam Flow Loop 3 (pounds / hour) 2.89E6 2.37E6 Steam Flow Loop 4 (pounds / hour) 2.83E6 2.38E6 i

Feedwater Flow Loop 1 (pounds / hour) 2.96E6 2.43E6 Feedwater Flow Loop 2 (pounds / hour) 2.92E6 2.43E6 Feedwater Flow Loop 3 (pounds / hour) 2.90E6 2.43E6 ,

Feedwater Flow Loop 4 (pounds / hour) 2.88E6 2.36E6 Feedwater Temperature Loop 1 (*F) 422 400  ;

Feedwater Temperature Loop 2 (*F) 426 405 l Feedwater Temperature Loop 3 (*F) 427 405 l Feedwater Temperature Loop 4 (*F) 430 410 Feed Pump Discharge Hdr Pressure-(psig) 1149 1149 i Control Bank D Position (steps) 186 156 j Control Bank C Position (steps) 228 228 '

Feedwater Pump 2-A Speed (rpm) 4471 4197 Feedwater Pump 2-B Speed (rpm) 4440 4185

-116-

+-,.

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l 3.4.3 - DYNAMIC RESPONSE TO A FULL LOAD REJECTION AND TURBINE  !

i TRIP - ISU-284B OBJECTIVE l l

This test is performed to verify the ability of the primary and i secondary plant and the plant automatic control systems to sustain  ;

a generator trip from full power and to bring the plant to stable ,

conditions following the transient. The N-16 instrumentation 1 response time is also determined. This test satisfies activities described by FSAR Table 14.2-3, Sheets 23, 24 and 28.

TEST METHODOLOGY From a stable plant power of approximately 100%, a generator trip is initiated by opening both of the main generator output breakers.

This directly causes a turbine trip and a reactor trip. The operators follow the permanent plant Emergency Operating Procedures to bring the plant to stable conditions. The data trending is terminated when Tavg is stabilized at approximately 557'F (no-load T,yg).

SUMMARY

OF RESULTS Unit 2 Main Generator output breakers 8020 and 8030 were opened at 2055 hours0.0238 days <br />0.571 hours <br />0.0034 weeks <br />7.819275e-4 months <br /> on 7/28/93 and the RCS temperatures were determined to be stabilized at 2100 hours0.0243 days <br />0.583 hours <br />0.00347 weeks <br />7.9905e-4 months <br />. All of the following test acceptance criteria were met:

• The pressurizer and steam generator safety valves did not lift.

• Safety injection did not initiate.

• All control and shutdown rods released and dropped to the fully inserted position.

• The plant was stabilized in Mode 3.

• The steam dump valves modulated closed in the proper sequence.

• Feedwater isolation occurred 18 seconds into the plant transient, at 558.4*F. This was below the 564*F Lo-Lo Tavg interlock and prior to reaching the no-load average RCS temperature of 557'F, as expected.

• Average RCS temperature stabilized at 558"F. This aatisfied the >553*F review criteria.

The response time of the N-16 instrumentation was 1.97 seconds.

This satisfied the time response requirement of 52.17 seconds.

Nuclear flux dropped to less than 15% power in 1.9 seconds. This satisfied the 52 second response requirement.

-118-

3'4.3 - DYNAMIC RESPONSE TO A FULL LOAD REJECTION AND TURBINE TRIP - ISU-284B (Continued)

SUMMARY

OF RESULTS (Continued)

Narrow range steam generator levels remained within the narrow range span. This satisfied the review criterion that levels may <

drop out of span (<0%) but should return to span (>0%). {

Pressurizer level remained 124.75%. This satisfied - t.ne minimum level of 123% review criterion.

Only. one item was not satisfied. While pressurizer prissure remained 11929.4 psig, this was outside of the minimum p' cre . re of 21950 psig review criterion range. The reason for the low pressurizer pressure was a low transient Tavg that resul.ted trom a relatively large addition of cold Auxiliary Feedwater early ..nWhen the transient (within the first 15 seconds after- the trip),

combined with the relative lack of decay heat contaiced in the fresh fuel of this Cycle 1 core, Tavg dropped further tha- would be expected following a trip experienced on the unit when an. normal

- operation. This condition was evaluated by the NSSS ven.d3r and found acceptable.

Refer to Table 3.4.3-1 for detailed test results.

l l

9 e

-119-

TABLE 3.4.3-1 TRIP FROM 100% POWER

SUMMARY

INIT!AL CONDITION UJ{M, CONDITION Generator Load (MWe) 1140 0 Nuclear Power (%) 100 0 T

avg Auctioneered (*F) 589.3 557.8 589.2 557.7 Tref ( F)

N-16 Power (%) 100 0 OPN16 Setpoint (%) 112.1 112.1 OTN16 Setpoint (%) 115.0 118.7 Pressurizer Pressure (psig) 2244 2157 Pressurizer Level (%) 60.1 28.3 Steam Generator Level Loop 1 (%) 64 11 Steam Generator Level Loop 2 (%) 65 10 Steam Generator Level Loop 3 (%) 64 6 Steam Generator Level Loop 4 (%) 65 14 Steam Header Pressure (psig) 960 1070 Steam Flow Loop 1 (pounds / hour) 3.79E6 0.30E6 Steam Flow Loop 2 (pounds / hour) 3.72E6 -0.01E6 Steam Flow Loop 3 (pounds / hour) 3.79E6 0.20E6 Steam Flow Loop 4 (pounds / hour) 3.73E6 0.19E6 Feedwater Flow Loop 1 (pounds / hour) 3.80E6 0.00E6 Feedwater Flow Loop 2 (pounds / hour) 3.75E6 0.14E6 Feedwater Flow Loop 3 (pounds / hour) 3.73E6 0.01E6 Feedwater Flow Loop 4 (pounds / hour) 3.70E6 0.00E6 Feedwater Temperature Loop 1 (*F) 440 430 Feedwater Temperature Loop 2 (*F) 440 435 Feedwater Temperature Loop 3 (*F) 440 440  ;

Feedwater Temperature Loop 4 (*F) 450 445 l Feed Pump Discharge Hdr Pressure (psig) 1130 634 Control Bank D Position (steps) 210 0 Control Bank C Position (steps) 226 0 Feedwater Pump 2-A Speed (rpm) 5067 3 Feedwater Pump 2-B Speed (rpm) 5049 328

-120- )

3.4.4 - REMOTE SHUTDOWN CAPABILITY TEST - ISU-223B OBJECTIVE This portion of this testing that verifies that the unit can be taken from approximately 20% reactor power to Hot Standby conditions from outside the control room with a minimum shift crew is done in ISU-225B, Section 3.4.5 of this Startup Report. The potential to safely cool the unit to cold shutdown conditions from outside the control room is demonstrated here in this test. This test satisfies activities described by FSAR Table 14.2-3, Sheets 25 and 26 and Regulatory Guide 1.68.2. The remote cooldown to 350*F and remote switch over to Residual Heat Removal (RER) System cooling, and the cooldown using RHR is demonstrated by this test.

This test was performed as part of the Hot Functional Test during the Preoperational Test Program.

TEST METHODOLOGY Utilizing abnormal operating procedure SOI-HFT-ABN-905B, Loss of Control Room Habitability (a preliminary draft of the permanent ABN-905B procedure), the minimum shift crew establishes control of the reactor plant and stabilizes it in Mode 3 from the Remote Shutdown Panel (RSP). From the Mode 3 stabilized condition, a controlled cooldown of at least 50*F is initiated from the RSP to '

demonstrate cooldown capability. Upon completion of this remote cool down, the RCS is then cooled to approximately 350*F and depressurized to approximately 350 psig. Then the RHR system is remotely placed in service and is used to cool the RCS an additional 50*F. Plant control is then restored to the Main ,

Control Room from the RSP. A standby Operations crew remains in the Main Control Room throughout the test to assume plant control, if needed.

SUMMARY

OF RESULTS The minimum shift crew used procedure SOI-HFT-ABN-905A to transfer control to the RSP and to establish a stable, hot standby (Mode 3) condition by 1018 hrs on 8/17/92. The Reactcr Coolant System (RCS) was stable at approximately 550*F. Then, a controlled cooldown was started in accordance with SOI-HFT-ABN-905A. The cooldown was performed from 1125 hrs. to 1219 hrs. The cooldown was approximately 55'F. This satisfied the cooldown requirement of at least 50*F. The cooldown rate did not exceed Technical Specification limits at any time. .

At 1739 hours0.0201 days <br />0.483 hours <br />0.00288 weeks <br />6.616895e-4 months <br />, with the RCS cooled down to approximately 350*F and depressurized to approximately 350 psig, the RHR system was remotely placed in service from the RSP and locally at plant equipment.

l

-121-1 I

3.4.4 - REMOTE SHUTDOWN CAPABILITY TEST - ISU-223B (Continued)

SUMMARY

OF RESULTS (Continued)

At 1900 hours0.022 days <br />0.528 hours <br />0.00314 weeks <br />7.2295e-4 months <br />, a controlled cooldown was started via remote operation of the RHR system from an initial RCS temperature of approximately 310*F. This cooldown was terminated at 2010 hours0.0233 days <br />0.558 hours <br />0.00332 weeks <br />7.64805e-4 months <br /> with the RCS at approximately 255'F, which satisfied the cooldown requirement of at least 50*F. The cooldown rate did not exceed Technical Specification limits at any time.

Plant restoration began at 2010 hours0.0233 days <br />0.558 hours <br />0.00332 weeks <br />7.64805e-4 months <br /> and finished at 0225 hours0.0026 days <br />0.0625 hours <br />3.720238e-4 weeks <br />8.56125e-5 months <br /> on 8/18/92 with Unit 2 again being fully controlled from the Main Control Room.

All transfers of control to and from the RSP were properly done and the RSP equipment was verified to operate properly with the following minor exceptions:

• Steam Generator Level instrumentation in the Main Control Room and at the RSP disagreed by various small magnitudes, all less than 5%. This discrepancy was magnified at lower temperatures when indicated levels are administratively corrected for temperature. This has no adverse impact on overall test results because suf ficient latitude was available in the allowed level ranges to I operate within specification. A procedure change was

! also made to the test to clarify which indications were to be used as the correct S/G level values.

l

• The labeling of controller 2HC-0606 was found to be l inaccurate. Adjusting the controller to increase demand l resulted in valve closure. This was corrected after the

! test was completed and had no adverse impact on test l

results because this controller is adjusted as necessary l to control flows and temperatures, there is no requirement on which way to adjust it.

• Pressurizer Heater Backup Group A would not respond to Main Control Room control when control was transferred back from the RSP at the end of the test. Blown fuses in the control circuit were replaced and control was properly established. These fuses had never blown before as the result of control transfers. This had no adverse impact on test results because it occurred following testing, during restoration.

-122-t . _ _ _ _ _ _ _ _ _ _ _ - - - - - - - - - _ - - - - - - - .

3.4.4 - REMOTE SHUTDOWN CAPABILITY TEST-- ISU-223B (Continued)  !

SUMMARY

OF RESULTS (Continued)

• Three miscellaneous actions were performed by Main Control Room operators during the test:

a) The Pressurizer Relief Tank (PRT) was recirculated to cool its contents. Pressurizer PORV 2PCV-456 had seat leakage resulting in a heated water input to the PRT. This problem was corrected following test completion.

b) The Volume Control Tank (VCT) was periodically vented to reduce its internal pressure. Problems with the hydrogen regulator allowed gas to slowly l overpressurize the VCT. Nitrogen was being admitted to the VCT as a cover gas through the hydrogen supply lines. This had no adverse impact on test results, if no action had been taken, relief devices would have protected the tank from damage, c) Reactor Coolant Pump (RCP) #4 was stopped due to increasing vibration levels. This pump was repaired following test completion. This had no adverse impact on test results because this action was specifically allowed by the test procedure and because RCP #4 would have been shut down by RSP operators as part of the very next step of SOI-HFT-ABN-905B.  ;

The above four items did not compromise the overall performance of f the Remote Shutdown Panel instrumentation and controls or of the  ;

SOI-HFT-ABN-905B procedure.

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3.4.5 - PEMOTE SHUTDOWN CAPABILITY TEST AT POWER - ISU-225B OBJECTIVE This portion of this test verifies that the unit can be taken from approximately 20% reactor power to Hot Standby conditions from outside the control room with a minimum shift crew. The potential to safely cool the unit to cold shutdown conditions from outside the control room is done in ISU-223B, section 3.4.4 of this Startup Report. This test satisfies activities described by FSAR Table 14.2-3, Sheets 25 and 26 and Regulatory Guide 1.68.2.

TEST METHODOLOGY From the condition of greater than 10% generator load and less than 25% reactor power, the reactor is manually tripped locally from the reactor trip breakers. Utilizing abnormal operating procedure ABN-905A, Loss of Control Room Habitability, the minimum shift crew establishes control of the reactor plant and stabilizes it in Mode 3 from the Remote Shutdown Panel (RSP) for at least 30 minutes.

Upon completion, control of the plant is transferred back to the Main Control Room. A standby operations crew remains in the Main Control Room throughout the test to assume plant control, if needed.

SUMMARY

OF RESULTS The reactor was locally tripped at 1416 hrs on 5/6/93 from 18%

reactor power and 120 MWe generator load (10.3% generator load) .

The minimum shift crew used ABN-905A to establish a stable, hot standby (Mode 3) condition by 1440 hrs. The Reactor Coolant System (RCS) was stable at approximately 557'F. Following 33 minutes of data taking to ensure stable RCS conditions, the test was terminated and the plant was restored to Main Control Room control.

This stable 33 minutes satisfied the requirement to maintain a stable, hot standby condition for at least 30 minutes.

All transfers of control to and from the RSP were properly done and the RSP equipment was verified to operate properly with no exceptions.

-124-

s 3.4.6 - LARGE LOAD REDUCTION TESTS - ISU-263B OBJECTIVE This test is performed to demonstrate the dynamic response of plant systems to automatically bring the plant to steady state conditions following a rapid large reduction in turbine load, and then to  !

stabilize conditions at the reduced load. The magnitude of this load reduction is approximately 50% power. This test partially l satisfies activities described by FSAR Table 14.2-3, Sheets 23 and i 24.

TEST METHODOLOGY With plant conditions stable at approximately 100% power, an approximate 50% load decrease is manually initiated from the ,'

turbine generator Electro-Hydraulic Controls (EHC) at a rate of approximately 200% power / minute and plant parameters are allowed to stabilize. The load decrease is performed by manually reducing the i turbine-generator load limit setpoint to a value approximately 50%

in power below the initial load reference operating power level.

The load limit setpoint adjustment occurs at a rate of approximately 200% power / minute and is performed by main control board manual pushbutton operation of a motor driven potentiometer i that is set to move at that rate. These pushbuttons are permanent plant control features and the related circuitry is closely  :

associated with the built-in turbine generator runback circuits. }

The 50% power load change is a nominal value and is actually i specified to be > 48% in magnitude. The 50% load change may result in reactor power changes of less than 50% power due to relatively low plant efficiency at lower power levels. .

During the course of the test, plant process computer and Data Acquisition System (DAS) recordings of key plant parameters are taken so that plant response can be analyzed. The principal ,

parameters monitored included RCS Tavg, Tcold, Tref, pressurizer  ;

pressure and level, steam generator pressures and levels, steam and feedwater flows, control rod positions and speed, OTN16 and OPN16 setpoints, reactor power, feedwater pump speed and discharge pressure, N-16 power, safety and relief valve positions, and steam  !

dump valve positions.

SUMMARY

OF RESULTS The test was performed at the 100% power plateau from 99.8% reactor i power and satisfied the following criteria:

• The load decrease did not cause the reactor to trip nor the turbine to trip.

• Safety injection did not initiate.

i

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3.4.6 - LARGE LOAD REDUCTION TESTS - ISU-263B (Continued)

SUMMARY

OF RESULTS (Continued)

• The steam generator safety and pressurizer safety valves did not lift during the load reduction.

a No manual inter"ention was required to bring plant conditions to steady state.

• Plant variables returned to steady state conditions without sustained or diverging oscillations.

• Steam dump valves did not repeatedly cycle from open to closed position, although open position modulation did occur.

• After stabilization, Tavg controlled to within 1.4 *F of Tref, Steam Generator Levels to within a 63% to 64% range and the steam dump valves were fully closed. This satisfied the criteria that Tavg be within 1.5 *F of Tref, Steam Generator Levels be 64 : 3% and the steam dumps are closed. While one monitored channel indicated slightly greater than a 1.5aF Tavg-Tref mismatch, the Main Control Board indication used by the Reactor Operator was acceptable. A Work Order was issued to recalibrate the loop to eliminate any bias present in the Tave-Tref error signal. Work Order troubleshooting resulted in the discovery of a faulty lead / lag circuit board that caused the observed signal problems. This board was replaced and verified to be operating properly.

• During test performance the turbine-generator load was reduced by approximately 600 MWe, 52.5% of full power. This satisfied the criteria that the load be decreased by at least 560 MWe.

Numerical review criteria are summarized on Table 3.4.6-1. The test performance was without significant incident. No adjustments to control systems were found to be necessary following an evaluation of control system interactions during the load reduction transient.

Refer to Table 3.4.6-2 for additional detailed test data.

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TABLE 3.4.6-1 Larce Load Reduction Tests Summary Peak Auc- Auctioneered tioneered Expected Tavg Expected Resconse(*F) Undershoot(*F) Rescor;se (

• F )

Tavaf'F) 2.8 above <7 above 0 <3 below final initial value initial value value Pressurizer Pressure Expected Swinafosic) Resnonsefosic)

+61,-72 +100,-160 Steam Duration Steam Generator of Max. Expected Dump Expected Level Expected Rod Speed Time Duration Time Swinc(%) Resoonse(%) (seconds) (seconds) (minutes) (minutes)

+9,-11 +15,-20 120 approx. 30 4 <8 l

i 4

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TABLE 3.4.6-2 LARGE LOAD REDUCTION

SUMMARY

INITIAL CONDITION FINAL CONDITION Generator Load (MWe) 1140 540 Nuclear Power (%) 99.8 50.8 Tavg Auctioneered (*F) 589.3 570.2 Tref (*F) 589.2 571.9 N-16 Power (%) 100.5 56.1 OPN16 Setpoint (%) 112.3 112.3 OTN16 Setpoint (%) 113.5 129.0 Pressurizer Pressure (psig) 2231 2255 Pressurizer Level (%) 59.8 38.9 Steam Generator Level Loop 1 (%) 64 64 Steam Generator Level Loop 2 (%) 64 64 Steam Generator Level Loop 3 (%) 64 64 Steam Generator Level Loop 4 (%) 64 64 Steam Header Pressure (psig) 961 1018 Steam Flow Loop 1 (pounds / hour) 3.81E6 1.72E6 Steam Flow Loop 2 (pounds / hour) 3.83E6 1.73E6 Steam Flow Loop 3 (pounds / hour) 3.80E6 1.72E6 Steam Flow Loop 4 (pounds / hour) 3.80E6 1.68E6 i

Feedwater Flow Loop 1 (pounds / hour) 3.86E6 1.67E6 Feedwater Flow Loop 2 (pounds / hour) 3.83E6 1.69E6 Feedwater Flow Loop 3 (pounds / hour) 3.81E6 1.63E6 Feedwater Flow Loop 4 (pounds / hour) 3.79E6 1.64E6 Feedwater Temperature Loop 1 (*F) 445 260 Feedwater Temperature Loop 2 (*F) 445 255 Feedwater Temperature Loop 3 (*F) 450 255 Feedwater Temperature Loop 4 (*F) 450 260 Feed Pump Discharge Hdr Pressure (psig) 1133 1133 Control Bank D Position (steps) 205 121 Control Bank C Position (steps) 226 226 i

Feedwater Pump 2-A Speed (rpm) 5063 4105 Feedwater Pump 2-B Speed (rpm) 4983 4088  !

-128-

l 1

3.5 INSTRUMENTATION AND CALIBRATION TESTING ,

t 3.5.1 - CALIBRATION OF-FEEDWATER AND STEAM FLOW INSTRUMENTATION i AT POWER - ISU-202B l 1

OBJECTIVE f

The purpose of this test is to verify the calibration of Feedwater- I (FW) flow and Steam Flow (SF) instrumentation at each of 'the major .;

test plateaus. Calibration of the Feedwater flow instrumentation,  ;

because the characteristics of the Main Feedwater flow venturi are 1 well known, is relatively straightforward. The differential pressure d/p transmitters are zero and span checked, and the downstream electronics conversion cards are shown - to be in calibration. Steam flow is determined by measurement of steam generator pressure and the differential pressure developed by the ,

flow of steam across the steam generator steam exit nozzle and  !

associated piping to a downstream point on the main steamline. The i major task is that of determining both the zero and the span of the j steam flow differential pressure transmitters, because neither of  :

these quantities are known precisely prior to actual power l operations, j TEST METHODOLOGY The test is comprised of three related activities. First, at hot, zero power, zero flow conditions, both FW and SF flow transmitters.  ;

are verified to be at or adjusted to be at zero output. As power  !

is increased, the standard plant instrumentation is verified to be li within specified calibration tolerances with respect to the high accuracy d/p transmitters. Finally,.the as-built spans of the SF' measurement system are determined by requiring that the indicated SFs agree with the simultaneously measured FW flows.

i With the plant in Mode 3 at hot, zero power conditions, an average i reactor coolant temperature of approximately 557'F and steam j generator pressures of approximately 1100 psig, the SF flow transmitters should have a zero output. Because the condensing pot  ;

for the low pressure side of the SF dp cell is at an elevation approximately three feet above the high pressure condensing pot, a l zero offset has to be incorporated into the transmitter calibration i to account for this static head difference. Due to differences in  !

the thermal expansion of structures and piping to which these l condensing pots are attached, and also due to high pressure static l i

shift effects on the actual transmitters, the pre-test estimate of the zero offset usually has to be modified. The zero flow check of  ;

the FW flow transmitters is accomplished at low power operations t

(<10% reactor power) with the feedwater header pressurized by a ,

Main Feedwater pump. Because the FW d/p cell instrument taps are  !

-129-l

c l

3.5.1 - CALIBRATION OF FEEDWATER AND STEAM FLOW INSTRUMENTATION  !

AT POWER - ISU-202B (Continued) i i

TEST METHODOLOGY (Continued) installed in horizontal feedwater piping, the zero flow check is  !

performed primarily to account for high pressure static shift of  :

the transmitters.

At higher power levels up to and including 100%, data is taken ,

which ultimately is used to determine the full span of the SF d/p  !

transmitters as well as verifying the calibration of the FW flow instrumentation. With steam generator blowdown secured and the plant stable, a high accuracy measurement of FW flow is obtained.

The most important parameters are the FW venturi differential pressures and these are obtained using high accuracy plant a/p cells. The output of the standard plant d/p cells are compared to ,

these values and adjustments are made, if necessary. Feedwater pressures and temperatures are also required for the determination of FW flow. As stated previously, only two parameters are used to measure SF, steam generator pressure and SF differential pressure.

Unlike the FW permanent plant instrumentation, SF instrumentation  ;

has an allowance for density compensation based on steam generator  ;

pressure. Setting SF equal to FW flow, and extrapolating to full scale flows, allows the corresponding full span of the SF transmitter to be established. Recognizing that the most accurate

extrapolation is that using data from higher power levels, the SF transmitters are generally not respanned until the 75% power data ,

has been obtained. During steady operations at the 100% power  :

plateau, a final set of data is taken and used to verify the calibration of both the SF and FW flow instrumentation.

Adjustments and recalibration of the instrumente. tion is then >

undertaken, if found necessary.

SUMMARY

OF RESULTS Refer to Table 3.5.1-1 for detailed test results.

In Mode 3 at an RCS temperature of approximately 557'F and a corresponding steam generator pressure of approximately 1092 psig, .

adjustments were made so that the output of the SF transmitters properly represents the zero flow condition. This testing was successfully repeated during the Unit 2 outage which took place midway through the 75% power testing plateau.

With the Main Feedwater header pressurized by a Main Feedwater pump, and the equalizing valve on the FW flow transmitter manifold open, the NLP card output voltage is required to be 0.000 0.025 volts. The associated computer points required a reading of 0.0 :

2.5 inWC. Data from six of eight transmitters were within the tolerance. 2-FT-510 and 2-FT-530, the two out-of-tolerance transmitters, were recalibrated. The subsequent retest for these  ;

two transmitters was successful.

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1

3.5.1 - CAT.IBRATION OF FEEDWATER AND STFAM FLOW INSTRUMENTATION AT POWER - ISU-202B (Continued)

SUMMARY

OF RESULTS (Continued)

' At-power testing was performed, with- ratesting as - required, at >

approximate power levels of 30%, 50%, 75% and 100%. Prior to data acquisition at each power level, steam generator blowdown ~ was .

isolated and plant parameters were verified to be stable. Data acquisition required approximately 15 minutes.

-The 30% and 50% data. indicated that all eight of the SF transmitters would eventually have to be recalibrated. The SF/FW flow mismatches, however, were not so'large that they adversely impacted plant operations. SF/FW flow mismatches serve as inputs to the steam generator level control system. At 30% power, feedwater flow transmitter 2-FT-531 and steam pressure transmitter 2-PT-514 were outside tolerance. Following recalibration, both transmitters were successfully ratested. At 50% power, no instrumentation adjustments or recalibrations were necessary.

At 75% power, feedwater flow transmitters 2-FT-530 and 2-FT-531A l required recalibration. The retest for these transmitters was  !

successful. Review of the 75% power test results also revealed that all eight SF transmitters had sufficient data to calculate new ,

l full flow span values. As a result, new full flow spans . were l calculated using test data from 75%' power in combination with data'  !

from lower powers. The eight SF transmitters were then recalibrated using these new span values. A subsequent retest showed that the new spans were properly installed.

At 100% power, the test data led to a recalibration and new full flow span values for steam flow transmitters 2-FT-522 and 2-FT-523.

In addition, Feedwater Flow Loop F-531 needed an isolator card replaced. Retest results were satisfactory. The 100% power results (i.e., original test performance in combination with the retest) satisfied all test criteria. The most significant criteria were met as follows:

• The difference between the d/p from the standard feedwater flow tracumitters and the high accuracy flow transmitters was within 2'O.5% of the full span d/p of the standard feedwater flow transmitters. The maximum transmitter variance was -0.35%.

The difference between the calculated feedwater flow and the high accuracy computer generated flow was within :

1.0% of rated flow. The maximum steam generator loop variance was only 0.5%.

The difference between the calculated feedwater flow and the calculated steam flow was within 2.0% of the rated flow. The maximum transmitter variance was 1.24%.

-131-

i l

t 3.5.1 - CALIBRATION OF FEEDWATER AND STEAM FLOW INSTRUMENTATION  :

AT POWER - ISU-202B (Continued)

SUMMARY

OF RESULTS (Continued)

Table 3.5.1-1 provides the steam flow transmitter scaling changes.

A design change has been initiated in order to install the-final steam flow transmitter scaling into the computer data base.

In summary, the testing was viewed as having been successful with the overall final results being:

• The FW flow transmitters were left within their tightly specified calibration range

• The downstream FW flow instrumentation was shown to be within calibration

• The SF calibrations established in this test were '

acceptable, with only one transmitter requiring a new scaling.

P h

'e

-132-i

I i

i

^

Table 3.5.1-1 Calibration of Steam Flow Transmitters ,

All values are given in units of in WC. ,

Original 75% Power Final Scaling i TransmittgI Span Sean Soan Ranae  ;

2-FT-512 520 410 410 -28.2 TO 381.8 2-FT-513 520 410 410 -28.1 TO 381.9 [

2-FT-522 520 406 422 -27.3 TO'394.7 I 2-FT-523 520 406 422 -27.2 TO 394.8 .

2-FT-532 520 442 442 -28.3 TO 413.7  !

2-FT-533 520 442 442 -27.8 TO 414.2  !

f 2-FT-542 520 412 412 -28.4 TO 383.6 l 2-FT-543 520 372 372 -42.8 TO 329.2 l

i l

l 1

G

-133-

3.5.2 -

OPERATIONAL ALIGNMENT OF PROCESS TEMPERATURE AND N16 INSTRUMENTATION - ISU-226B OBJECTIVE This test is performed to align the N16 and Tavg process instrumentation, to verify the linearity of the N16 and Tavg instrumentation, to determine the optimum voltage setting for the N16 detector High Voltage power supplies, and to determine the N16 detector currents at various power levels. This test partially satisfies activities described by FSAR Table 14.2-3, Sheets 9, 10 and 22.

TEST METHODOLOGY At Comanche Peak, the Reactor Coolant System (RCS) Resistance Temperature Detector (RTD) manifold hot leg temperature (Thot) and cold leg temperature (Tcold) measurement instrumentation has been replaced by a Nitrogen-16 (N16) power monitor and an in-line Tcold RTD. The N16 power monitor measures the thermal power of the reactor by detecting the amount of N16 present in the coolant. The concentration of N16 in the coolant is directly proportional to the fission rate in the core and is detected by measuring the high energy gamma flux from the N16 decay which penetrates the walls of the hot leg piping. The fast response in-line Tcold RTD is in a thin wall thermowell installed in the cold leg piping. The process control system uses these inputs to generate a Tavg signal which is used for input to Rod Control, pressurizer level control, and steam dump control. An N16 Power signal is also generated which inputs ,

to the Reactor Protection System for Overtemperature and Overpower N16 reactor trips.

This test is a collection of several different tests of the N16, Tcold and Tavg process instrument loops which are performed throughout the startup program from Mode 3 through 100% power.

Refer to Table 3.5.2-1 for a matrix of which tests are performed at each plant condition.

The DETERMINATION / SETTING OF N16 DETECTOR HIGH VOLTAGE test is performed at approximately 50% reactor power. The N16 gamma detectors are tested one loop at a time. The current output from the N16 gamma detectors is measured by a picoammeter while the high voltage power supply output voltage is adjusted from 300 volts to  ;

1000 volts. This data is plotted to determine the plateau region of the curve, the region of minimum output current change for a l given voltage change. The power supply is then set to a value in this plateau region of the curve, nominally 800 volts. ,

The N16 CURRENT MEASUREMENT test is performed in Mode 3 and at the 30%, 50%, 75%, and 100% power plateaus. The input voltage and output voltage of each N16 power monitor module is recorded i l

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3.5.2 - OPERATIONAL ALIGNMENT OF PROCESS TEMPERATURE AND N16 INSTRUMENTATION - ISU-226B (Continued)

TEST METHODOLOGY (Continued) simultaneously with the reactor thermal power from a precision secondary calorimetric. At zero power, Mode 3, the power monitor module output is verified to be 0.000 +0.033 volts. There are no criteria at higher power levels.

The RCS COLD LEG TEMPERATURE CHECKS are performed in Mode 3. The active cold leg RTD temperature, as measured at the output of the NRA card in the process control rack, is compared to its associated spare RTD temperature, measured as resistance from the RTD. These temperatures are required to agree within 0.7'F for each RCS loop.

The VERIFICATION OF TAVG CIRCUITRY test is performed at every power plateau and in Mode 3. The process control system Tavg signal is compared to a calculated value generated by the following equation:

Tavg = Tcold + (K-9)(N16 Power) where Teold is the cold leg temperature from the active RTD circuitry, N16 is a power signal generated by the N16 process loop, and K9 is a constant equal to 1/2 the full power temperature dif ference of hot leg to cold leg. The calculated Tavg is verified to be within 0.5'F of the Tavg signal.

The NEUTRON STREAMING DETERMINATION test is performed at the 75%

power plateau. The N16 gamma detectors in the RCS hotAdditionally, legs monitor gamma rays from the decay of N16 in the RCS water.

gamma rays streaming directly from the upper portion of the reactor core and secondary gamma rays generated by streaming neutrons add to the N16 power signal. This contribution to the N16 power signal comes primarily from the top region of the core. The signals from the top two detectors in each nuclear instrumentation power range channel are used to compensate the associated N16 power signal.

During the Incore/Excore Detector Calibration test, NUC-203, an axial Xenon transient is initiated causing the neutron flux to shift axially in the core. The following data is taken during the transient: N16 power monitor output, the output from each of the top two power range detectors and precision calorimetric power measurements. Using the relationship:

Q = A (V"N16) +A2 IVA) *A 3 B) where Q is calorimetric power, V"N16 is the N16 Power monitor output, V is the top power range detector output and V B is A the next to tsp power range detector output, the constantsT A h d, n e tron u., a n d A are determined by linear regression analysis.

sdreaming compensation gains are then calculated andstreaming used to calibrate the process channels to negate the core effects.

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3. 5. 2 - OPERATIONAL ' ALIGNMENT OF PROCESS TEMPIRATURE AND N16  !

INSTRUMENTATION - ISU-226B (Continued) .l TEST METHODQLOGY (Continued)

The FULL POWER DELTA-T (K-9) VERIFICATION test is. performed at the  ;

75%.and 100% power plateau. Cold leg and hot leg temperatures are  !

determined from the PCS - flow test procedure, ISU-023B, with hot leg i temperature determined by iteration of TTFM measurement. of RCS  ;

flow, secondary precision . calorimetric power, and Tcold. Then,  ;

this temperature difference is compared to a temperature difference- l equivalent to twice the existing value of K-9. Any loop which i dif fers by more than 1% is recalibrated using -the new K-9 value  ;

determinod from the actual full power cold leg and deferred hot. leg '

temperatures. The initial K-9 estimated value is 28.100 *F. These. i measurements and adjustments are done as part of the ISU-023B test.

The N16 POWER CHECK - (K-8) ADJUSTMENT test is performed at every power plateau and in Mode 3. - The N16 power signal for each loop is.

compared to precision secondary calorimetric power. and adjustments  ;

are made to make them match. Below 75% power the gain of the N16 l power monitor module itself is adjusted. After the neutron streaming gains have been determined at 75% power, as' described l previously, any adjustments are made to the K-8 constant instead of  !

the power monitor module. At zero power, Mode 3, the N16 power q signal is verified ~to be 0.000 10.033 volts.

.The N16 POWER LINEARITY CHECKS are performed after the full power 3 N16 data has been taken and the N16 power instrumentation has been adjusted at full power. The power output from each loop is plotted q against calorimetric power and evaluated for linearity by linear  ;

regression. . Data is used from all power plateaus in the  :

regression. ,

I The N16 ELECTRICAL ZERO DETERMINATION test is performed at least i four hours after a reactor shutdown from 100% power. Input and i output voltages of each power monitor module are recorded. .The  ;

output voltages are verified or adjusted to.be 0.000 10.033 volts. 1 This ensures proper compensation.for the. background gamma flux.

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3.5.2 - OPERATIONAL ALIGNMENT OF PROCESS TEMPERATURE AND N16 INSTRUMENTATION - ISU-226B (Continued)

SUMMARY

OF RESULTS The-DETERMINATION / SETTING of N16 DETECTOR HIGH VOLTAGE test results indicated that the N16 detector current is independent of the power supply-high voltage setting between 300 and 1200 volts. The high voltage power supplies were set to 800 volts.

The N16 CURRENT MEASUREMENT test results were: I Power Plateau Mode 3 30% 50%

r Calorimetric Power (%) 0 27.48 47.74 Loop 1 input (volts) 0.000 -0.249 -0.421 Loop 1 output (volts) 0.000 1.967 3.335 Loop 2 input (volts) 0.000 -0.246 -0.417 Loop 2 output (volts) 0.000 1.828 3.313 '

Loop 3 input (volts) 0.000 -0.252 -0.413 Loop 3 output (volts) 0.000 1.938 3.286

• Loop 4 input (volts) 0.000 -0.252 -0.425 Loop 4 output (volts) 0.001 2.007 3.379 At zero power, Mode-3, the criterion is 0.000 10.033 volts Power Plateau 75% 100%

Calorimetric Power (%) 73.10 99.80 Loop 1 input (volts) -0.649 -0.8888  ;

Loop 1 output (volts) 5.130 6.731 Loop 2 input (volts) -0.640 -0.8806 Loop 2 output (volts) 5.090 6.684 Loop 3 input (volts) -0.637 -0.8745 Loop 3 output (volts) 5.055 6.616 ,

Loop 4 input (volts) -0.648 -0.8905 Loop 4 output (volts) 5.167 6.768 The loop inputs were all within 0.033 volts of 0.000 in Mode 3.

All of the data collected was acceptablo.

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3.5.2 - OPERATIONAL ALIGNMENT OF PROCESS TEMPERATURE AND N16- l INSTRUMENTATION - ISU-226B (Continued)

SUMMARY

OF RESULTS (Continued)

The RCS COLD LEG TEMPERATURE CHECKS test results were: -

RCS LOOP T, RTD AMPLIFIER I fSPARE Ie DELTA 1 557.02 556.68 -0.34 l 2 556.65 557.03 +0.38 )

3 556.98 556.92 -0.06 l 4 556.29 556.70 +0.41 The acceptance criterion of this test was that the spare Tcold RTD .

and the NRA card output from the active Tcold RTD for each loop shall indicate within 0.7'F of each other or less. All of the loop- i values were within this 0.7aF criterion.

i The VERIFICATION OF TAVG CIRCUITRY test results were:

Tava Error (*F)

Power Plateau Looo 1 Loco 2 Loon 3 Loco 4 Criterion '

Mode 3 0.00 +0.31 +0.06 +0.13 <10.5 30% -0.31 +0.38 -0.12 +0.39 < 0.5 50% -0.23 +0.26 +0.05 +0.26 <10.5 75% -0.16 +0.05 +0.14 +0.37 <10.5 100% -0.16 -0.11 -0.10 +0.05 <10.5 All loops calculated Tavg within the required 0.5'F accuracy.

The NEUTRON STREAMING DETERMINATION test results at the 75% power plateau were:

Item Loon 1 Looo 2 Looo 3 Loon 4 Ay 15.0261 15.3461 15.4185 15.3586 A -0.3987 -0.9405 -0.6981 -1.4224 2 0 A 0 0 0 i 3 -0.0926 Gg -0.0265 -0.0632 -0.0453 G 0 0 0 0 B

where G = A /A and G B = A /A , the gains for the top and next db tdp nuclear inEkrdhentation detector signal compensation. In practice, G, is set to 0 and the other ,

constants recalculated after @nfirmation that G B is very, very small in magnitude.

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I 3.5.2 - OPERATIONAL ALIGNMENT OF PROCESS TFMPERATURE AND N16 1 INSTRUMENTATION - ISU-226B (Continued) l

SUMMARY

OF RESULTS (Continued)

These values of gains were acceptable and used to calibrate the  !

neutron streaming compensation circuits.  ;

The K-9 constant used for preliminary calibration of N16 Tavg circuitry was 28.1*F. Since these extrapolated values were within 1.0*F of 28.1*F, no adjustments to K-9 were made at 75% power.

The VOLUMETRIC ENTHALPY (K-9) VERIFICATION test results were as follows where the % Error = ((K-9/K-9 calibrated) x 100) - 100 and the K-9 value should equal the Full Power AT/2:

75% (!NITIALi LOOP K-9 MEASURED (*F) K-9 CALIBRATED (*F) % ERROR ,

1 27.094 28.100 -3.580 l

2 27.249 28.100 -3.028 3 27.285 28.100 -2.900 l 4 27.556 28,100 -1.936 75% (FINAL)

LOOP K-9 MEASURED (*F) K-9 CALIBRATED (*F) 1_XRROR  :

i 1 27.130 27.094 0.148 ,

2 27.137 27.249 -0.415 3 27.161 27.285 -0.473 i 4 27.452 27.556 -0.392 100%

LOOP K-9 MFARURED (*F) K-9 CALIBRATED (SF) % ERROR '

1 27.019 27.094 -0.277 2 27.303 27.249 -0.198 1

3 27.406 27.285 -0.443 4 27.472 27.556 -0.305 The initial % errors for all four loops exceeded the allowed 1% and were recalibrated with new K-9 values. The VERIFICATION OF TAVG '

I CIRCUITRY was then reperformed at 75% power following the ,

adjustments to K-9 and this data is shown above. The 100% power l values did not require an adjustment. j l

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3.5.2 - OPERATIONAL ALIGNMENT OF PROCESS TEMPERATURE AND N16 INSTRUMENTATION - ISU-226B (Continued)

SUMMARY

OF RESULTS (Continued)

The N16 POWER CHECK - (K-8) ADJUSTMENT test results were:

For zero power, the N16 output voltage must be -0.033 volts to

+0.033 volts Test Loon 1 Looo 2 Looo 3 Looo 4 Mode 3 (volts) +0.002 +0.031 -0.016 -0.010 1

For at-power measurev. ants, if N16 power differs from calorimetric by 1% or more, the N16 gain is readjustad. (Listed Values are N16 ,

Poucr (%) Calorimetric Power (%))

Power Plateau Looo 1 Loco 2 Looo 3 Looo 4 30% -0.383 -0.773 -0.158 -0.293 l

50% -0.17 +0.06 -0.02 +0.01 75% -0.18 -0.33 -0.74 -0.33 100% +0.24 +0.49 +0.58 +0.45 No valves required readjustment as all difference were less than l 1%.

l The N16 ELECTRICAL ZERO DETERMINATION test results were:

Loop Inoutivolts) Outout(volts) l 1 -0.001 -0.005 l 2 -0.001 0.006 i 3 0.000 0.006 4 -0.001 0.008 )

The N16 Electrical Zero Determination was performed after the ISU-284B Full Load Rejection and Turbine Trip from 100% power. The output voltages were within the required 0.000 20.033 volts tolerance and no adjustments were made.

The N16 POWER LINEARITY CHECK results were as follows:

Loop Correlation Coefficient 1 0.999962 2 0.999955 3 0.999853 4 0.999947 All correlation coefficients satisfied the 20.99 review criterion by a wide margin, indicating very linear behavior of the N16 detectors.

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i TABLE 3.5.2-1 i

PROCESS TEMPERATURE /N16 TESTS VS. PLANT l CONDITIONS MATRIX MODE 3 30% 50% 75% 100%

DETERMINATION / SETTING X ,

OF N16 DETECTOR HIGH l VOLTAGE N16 CURRENT MEASUREMENTS X X X X X .

RCS COLD LEG TEMPERATURE X CHECKS VERIFICATION OF TAVG X X X X X  !

CIRCUITRY .

NEUTRON STREAMING X l DETERMINATION i FULL POWER DELTA-T(K-9) X X  !

VERIFICATION N16 POWER CHECK - K-8 X X X X _ _ __}l ADJUSTMENT 5

N16 POWER LINEARITY CHECKS X l

N16 ELECTRICAL ZERO Performed at zero power following  !

DETERMINATION 100% power testing i

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3. 5. 3 - OPERATIONAL ALIGNMENT OF NUCLEAR INSTRUMENTATION - ISU-204B OBJECTIVE This test is performed to verify that the excore Nuclear Instrumentation System (NIS) functions per design. This test partially satisfies activities described by FSAR Table 14.2-3,  !

Sheets 9 and 10.

TEST METHODOLOGY >

Selected parameters are evaluated, monitored, and determined during various testing phases.

Prior to and at the time of Initial Criticality, Source Range (SR) '

to Intermediate Range (IR) channel overlap data is taken to verify how much overlap exists between them. Data is recorded simultaneously from both SR and both IR channels as the reactor neutron flux increases during the approach to criticality. This data permits the calculation of whether or not the IR channels begin to indicate at a sufficiently low flu:t level such that adequate margin exists to be able to deenergize the SR channels prior to reaching the SR reactor trip setpoint. Data is also recorded simultaneously from both IR and all four Power Range (PR) channels. This data, when combined with similar data at full power, permits the calculation of IR and PR channel overlap.

During power escalation, at approximately 30%, 50%, 75%, and 90%

power, the % power outputs from all four PR channels are either verified to be or are aligned to be within 11% of reactor thermal power (calorimetric power). The 90% power execution provides additional assurance that the PR channels are properly calibrated so as not to exceed 100% power when power is increased from 90% to 100%. Additional data is also recorded for use in the IR and PR overlap calculations. The measured PR channel detector currents are also trended as a function of calorimetric power for use in verification of detector linearity.

At approximately full power, final data of the type taken during power ascension is recorded. The PR channels are either verified to be or are aligned to be within 11% of calorimetric power. The full power currents are combined with those during power ascension to verify PR detector linearity. The full power IR and PR current data is evaluated to demonstrate adequate IR and PR overlap, such that adequate margin exists to be able to block the IR reactor trip. The IR data is extrapolated to 100% power and this calculated 100% power value is used to compute the IR high level rod stop and IR high level trip setpoints and reset values.

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3. 5. 3 - OPERATIONAL ALIGNMENT OF NUCLEAR INSTRUMENTATION - ISU-204B (Continued)

TEST METHODOLOGY (Continued) {

After Shutdown from Power Operations of at least 800 MWD /MTU, the operating high voltages and discriminator bias voltages for the SR channels and the compensating voltages for the IR channels are redetermined and set. Prior to core loading, the SR channel high voltages and discriminator bias voltages were set using neutron sources to produce detector currents. This test portion has these voltages readjusted to properly correspond to actual shutdown reactor neutron and gamma spectra producing the detector currents.

The IR detector consists of two concentric detector volumes, one sensitive to neutrons and gamma rays and one sensitive only to gammas. The current outputs from the two detector volumes are placed in opposition to one another, i.e. bucked against each current signal components that are other, such that the proportional to gammas cancel and the net current corresponds only to neutrons. To compensate for size, geometry and efficiency differences between the two detector volumes, aIRbias current is compensating also applied between the two volumes. The voltages, which provide these proper bias currents, are initially set to -40 volts to ensure complete elimination of the gamma signal, even at the cost of losing a portion of the neutron signal.

This ensures that the channel output is forced low enough following a reactor trip to help ensure that the SR channels automatically reenergize. If improperly set, the large gamma signal present following a trip could cause the IR channel output to remain and prevent abnormally high for an extended period of The compensating time automatic reenergization of the SR channels.

voltages are reset to realistic values based on the use of actual reactor neutron and gamma spectra as detector inputs to ensure proper screening of the gamma signal.

SUMMARY

OF RESULTS A minimum overlap of 2.57 decades was observed on all SR/IR channel combinations. The minimum required overlap is 1 1/2 decades for SR l

to IR and IR to PR. The overlaps for all eight IR/PR channel Refer to combinations were observed to be moreThe thanhigh 2.1 decades. voltages for the SR Table 3.5.3-1 for detailed results.

channels were reset as were the discriminator bias voltages for the SR channels. The compensating voltages for the IR channels were also reset following a scheduled full power reactor trip, test procedure ISU-284B, with a core burnup of approximately 1217 MWD /MTU.

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3. 5. 3 - OPERATIONAL ALIGNMENT OF NUCLEAR INSTRUMENTATION - ISU-204B j (Continued)

SUMMARY

OF RESULTS (Continued)

The PR channel outputs were either verified to be or were aligned to be within 1% of calorimetric power at all power plateaus and at full power. The PR channels were verified to demonstrate .

acceptable linearity of outputs. Refer to Figure 3.5.3-1 for a  !

plot of channel N41 summed top and bottom detector currents as a l futetion of calorimetric power. Similar plots were made for the other three channels but are not included in this report. Refer to Table 3.5.3-1 for detailed results.

The 100% power IR channel outputs were calculated and the IR rod stop and trip setpoints and reset values were calculated. Refer to ,

Table 3.5.3-1 for detailed results. Refer to Figure 3.5.3-2 for a ,

plot of channel N35 IR detector current as a function of i calorimetric power.

No significant problems occurred during test performance.

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Table 3.5.3-1  !

Nuclear Instrumentation Results Summary  !

'i Testina Plateau '

121 1Q1 2B. 1Q1 100%'

Calorimetric Power (%)_ 27.00 50.43 73.24 IR N35 output (amps) 90.10 9 9. 7 d, IR N36 output (amps) 1.3E-4 2.7E-4 3.8E-4 1.3E-4 2.4E-4 4.1E-4 4.9E' 3.5E-4 4.0E-4 4.8E-1 PR N41 summed current (pamps) 196 PR N42 summed current 341 471 576 (yamps) 167 295 635{

PR N43 summed current (pamps) 411 503 187 326 451 551{

PR N44 summed current (yamps) 175 304 552 606]

419 517 57Ci PR PR N41 Power (%)* 29.0 50.0 N42 Power (%)* 27.5 73.0 90.0 PR N43 Power (%)* 50.3 73.0 90.0 99{

PR 27.0 50.2 99i N44 Power (%)* 28.0 73.0 90.0 50.3 73.0 90.0 99l 99-

• Values recorded are the "as found" values. The as left values '

either the same calorimetric power.as the as found values or adjusted to be with ,

k b

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t Table 3.5.3-1 5 Nuclear Instrumentation Results Summarv (Continued)

Source Range vs. Intermediate Range Overlaps  ;

Channels Overlao (decades) {

i

~

N31 vs. N35 2.57 N31 vs. N36 2.58 N32 vs. N35 2.59 l N32 vs. N36 2.60 Intermediate Range vs. Power Range Overlaps Channels Overlao (decades)  !

N35 vs. N41 2.6 2.1  !

N35 vs. N42 N35 vs. N43 2.1 [

N35'vs. N44 2.6 N36 vs. N41 2.7  ;

N36 vs. N42 2.2 N36 vs. N43 2.1 i N36 vs. N44 2.7  !

Intermediate Range Currents ,

l Channel  !

M M j Full Power Current (amps) 4.9E-4 4.8E-4 l High Level Trip Setpoint (amps) 1.2E-4 1.2E High Level Trip Technical Specification Limit (amps) 1.54E-4 1.51E-4 High Level Rod Stop (amps) 9.8E-5 9.6E-5 NOTE: The IR High Level Trip Setpoint is the current equivalent -

to <25% of full power, the Rod Stop is at 120% of full power and the Technical Specification Limit is at 131.5%

of full power. The reset values are nominally calculated to be 1/2 of the actuation values.

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Figure 3.5.3-1 N41 Power Range Current vs. Calorimetric Power 700 - -- - - - - - - - -- - -- - - - - - -

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

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~ sa o .24 - Ea3yO g5OW!"Iz -

b 3.5.4 - INCORE/EXCORE DETECTOR CALIBRATION - NUC-203 OBJECTIVE This permanent plant procedure is performed to assure that a linear relationship exists between the excore neutron currents and the incore Axial Flux Difference (AFD). Once established, this excore current / AFD relationship is used to perform various calibrations of the excore channels, the Overtemperature N16 (OTN16) AFD inputs, Axial Flux Difference indications and plant computer inputs. This procedure partially satisfies activities described by FSAR Table 14.2-3, Sheet 22 and Technical Specification 3/4.3.1.1.

TEST METHODOLOGY For the 50% power execution of this procedure, a base case full core flux map is taken at stable core conditions. A small reactor coolant dilution is made and Control Bank D is inserted 15 to 25 steps to compensate for this reactivity change, with reactor power held constant. The effect is to push neutron flux toward the bottom of the core which makes AFD more negative. With AFD more negative than the base case, a quarter core flux map is taken. The reactor coolant is then borated to restore Control Bank D to its original position. The small magnitude (approximately 5% AFD) and short duration (approximately 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />) of this AFD change does not result in any significant residual Xenon transient effects on the core. The following data is taken during both flux maps: l calorimetric power, excore nuclear detector currents and main .

control board and plant computer AFD and Axial Offset indications.

The flux map axial power distribution (top half of core vs. lower [

half of core) results are combined with the other data to compute l the preliminary calibration constants.

The incore flux map results are assumed to represent the true axial and are used as the basis for all AFD power distribution indications. However, the AFD indications come from excore power range detectors output. Due to changes in core radial power distributions, the flux that the excore detectors see may not l accurately represent the actual core averaged conditions. To l compensate for this, the calibrations of the excore detector based This AFD indications are based on incore flux map results.

procedure plots actual excore detector outputs as a function of incore flux map Aq. Aq is the relative top vs. bottom incore power i distribution while AFD is the excore detector measured top vs. i bottom core power distribution. When the channels are calibrated, l A q = AFD . The resulting plots allow calculation of the excore j currents that would be expected to be present if the core were to ,

be at selected incore Aq values. These selected incore aq values would be the calibration points. These values are supplied to Instrumentation and Controls for use with their normal plant

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3.5.4 - INCORE/EXCORE DETECTOR CALIBRATION - NUC-203 (Continued)

TEST METHODOLOGY (Continued) calibration procedures. They input the specified test current signals into the power range circuitry inputs and adjust the outputs and indications to correspond to the selected incore Aq.

The direct relationship between plant computer Axial Of fset and incore flux map Axial Offset is calculated as the slope of the curve for the plot of the computer values as a function of the incore flux map values. These slopes are input to the plant computer as conversion constants to convert the excore computer inputs to correspond to incore Aq values used for reactor monitoring.

The results of the excore current vs. incore AFD plots are also used to calculate the full span (120% power) currents that would exist at the 0% AFD condition for use in Quadrant Power Tilt Ratio (QPTR) calculations.

For the 75% power execution of this procedure, an axial Xenon oscillation is created by a significant insertion of Control Bank  ;

D (up to 40 steps) in response to a reactor coolant dilution, holding this inserted position for approximately two hours and then borating the reactor coolant to restore Control Bank D to its starting position. While Control Bank D is deeply inserted, Xenon is preferentially depleted and Iodine preferentially produced in the lower half of the core. When Control Bank D is withdrawn, the neutron flux shifts toward the top of the core over time as the Iodine decays to Xenon in the lower half of the core. While the flux moves upward, with the corresponding positive change in AFD, numerous quarter core flux maps are taken. Full core flur maps are taken prior to the Control Bank D insertion and with control Bank D at its maximum inserted position. The same plant data is taken during the flux maps as was done at 50% power. The same ,

calculations are also performed as at 50% power. The 75% power results are generally expected to be more accurate than those at 50% power due to a reduction of the adverse temperature redistribution effects with increasing core delta temperature and l the larger span of Aq values. Following completion of data acquisition, the axial Xenon transient is suppressed.

1 1

For the 100% power execution of this procedure, a full core flux map is performed at stable core conditions and the same plant data l is taken as done at lower power levels. A comparison is made between the indicated AFD values and the incore flux map AFD. If the comparison is satisfactory, no adjustments are made to the AFD circuitry. If the comparison is not satisfactory, the AFD circuitry would be recalibrated based on 100% power data.

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3.5.4 - INCORE/EXCORE DETECTOR CALIBRATION - NUC-203 (Continued) {

SUMMARY

OF RESULTS The excore detector data and l incore flux map results were '

successfully used to calculate the calibration parameters for the OTN16 inputs, plant computer - inputs and. Axial Flux. Difference i indications at both the 50% and 75% reactor. power levels. The 100% ;

power results were satisfactory with no recalibrations required. .

The 50% power test was performed at approximately 46% power. The  !

Aq corevalues flux map,for +3.3%.

the full core flux map was -5.8% and for, the quarter This 50% power test performance ensured that  ;

the AFD circuitry was properly calibrated prior to exceeding 50%  ;

power, where Technical Specification 3/4.2.1 first applies. ,

i During the 75% power test, performed at approximately 78.5% power, j ten to +

flux maps were obtained ranging from an incore q of -12.354% '

19.175%. This test performance was also used to satisfy Technical Specification Surveillance Requirement 4.3.1.7.

During the 100% power test, the differences between the excore AFD ,

values and the incore flux map Aq were all less than the allowed maximum of 3%  !

and, when statistically combined using the square root of the sum the allowed maximum of the squares method, the difference was less than  !

of 8%. Therefore, no instrumentation adjustments were necessary. This test performance was also used to satisfy Technical Specification Surveillance Requirement 4.3.1.1.2a.  ;

Refer to Table 3.5.4-1 for detailed test results. '

i I

l l

-151-f i

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b k

Table 3.5.4-1 Incore/Excore Detector Calibration Summarv l 50% Powgr (Incore Aq value in %) '

Slope of Incore NIS. Upper Detector Lower Detector vs. Excore  ;

Channel Currentf uamos) Current ( uamos) Axial offsg1

• N-41 (4.0183 x Incore Aq) (-2.7277 x Incore Aq) 1.0261

+423.400 +404.698 N-42 (3.2559 x Incore Aq) (-2.5275 x Incore Aq) 1.0278  ;

+340.570 +377.696 N-43 (3.6045 x Incore Aq) (-2.5253 x Incore Aq) 1.0763 ,

+399.721 +390.643 N-44 (3.4659 x Incore Aq) (-2.5292 x Incore Aq) 1.0249'

+370.535 +367.339 l 75% Power (Incore Aq value in %) Slope -

of Incore NIS Upper Detector Lower Detector vs. Excore j Channel Currentf uamos) Current ( uamosi' Axial offset N-41 (3.1724 x Incore Aq) (-2.2823 x Incore Aq) 1.1965 ,

+394.311 +379.427 -

N-42 (2.4931 x Incore Aq) (-2.3140 x Incore Aq) 1.1648 ,

+317.622 +354.355 N-43 (2.8868 x Incore Aq) (-2.2991 x Incore Aq) 1.1941 l

+372.360 +365.517 N-44 (2.7035 x Incore Aq) (-2.2015'x Incore.Aq) 1.1778 i

+346.081 +343.434 100% Power i

NIS Channel Excore AFD (%) Incore Aa(%) Difference (%) i N41 -4.62 -5.32 0.700 N42 -4.69 -5.32 0.691 f N43 -4.68 -5.32 0.692 N44 -4.76 -5.32 0.686 i Square root of the sum i of the 1.385% l Squares j i

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3.5.5 - LOOSE PARTS MONITORING BASELINE DATA - ISU-211B OBJECTIVE ,

i This test is performed to gather noise frequency response data at approximately 0%, 30%, 50%, 75% and 100% reactor power. This data is used as a reference baseline for analyzing suspected loose parts l in the NSSS and to verify proper alarm levels and noise filter l i

settings.

TEST METHODOLOGY At each specified plant condition, a background noise recording of )

each of the 20 installed Loose Parts Monitoring System (LPMS) l accelerometer channels is made using the permanently installed recording equipment. The recorded data is then played back through .

an oscilloscope to verify adequate signal quality. The data is '

later evaluated using a spectrum analyzer and a final summary report of LPMS performance is compiled, refer to Attachment A to ,

this Startup Report. l

SUMMARY

OF RESULTS A comprehensive summary of the results of LPMS testing including j preoperational impact testing results, filter settings and l background noise spectra is provided in Attachment A, pursuant to l Regulatory Guide 1.133 requirements for loose parts monitoring I system testing.

Baseline data was satisfactorily collected in this test to provide a baseline for each of the 20 accelerometer channels. Testing was performed without incident except for several minor problems as follows:

• During the 0% power portion of the test , sensor channels 3, 16 and 20 indicated erratically, indicative of malfunctioning sensors or sensor cables. These channels were later repaired and the data taken at 30% power and at higher power levels was significant to establish an dequate baseline response and to verify proper channel operation.

• During the 30% power portion of the test, sensor channel 1 was found to not have been identified as an active input to Loose Parts Monitor channel 1, the normal system alignment. The l system software was corrected and the proper data was taken.  !

• During the 100% power portion of the test, sensor channels 17 and 20 indicated erratically, indicative of malfunctioning sensors or sensor cables. These channels are to be repaired l during the next outage of sufficient duration due to ALARA l considerations. Spare sensors are already installed at these i locations and the previously collected data was adequate to l establish baseline response of these channels.  ;

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3.5.6 - STARTUP ADJUSTMENTS OF PEACTOR CONTROL SYSTEMS - ISU-020B OBJECTIVE This test is performed to determine the average RCS temperature (Tavg) value which results in establishment of the design steam pressure at full load, within the temperature limits for the maximum allowable Tavg. This is accomplished by making adjustments to the reference Tavg (Tref) program and rescaling the turbine impulse pressure instrumentation, as necessary. Pressurizer level is also verified to correspond to the proper programmed value as a function of power.

TEST METHODOLOGY At approximately 30%, 50%, 75%, 90% and 100% power, plant data is taken for use in evaluation and extrapolation of the Tref program and turbine impulse pressure (Pimp) program value. Data is also j taken to verify proper response of the pressurizer level control program as a function of power. This data consists of calorimetric power, Tavg values, pressurizer level and level setpoints, Tref, steam pressures, turbine impulse pressures and main generator electrical output.

A change in Tavg results in a change to average steam generator saturation temperature which directly affects steam generator saturation pressure. The rod control system automatically functions to maintain Tavg at, or very close to, Tref. The value of Tref increases above the 557'F no-load value as a programmed function of power. In order to optimize steam pressure, the test evaluates Tavg, Tref and steam generator pressure and calculates what change in Tref would be necessary to alter Tavg by the proper amount to result in the optimal steam generator pressure. The optimal steam generator pressure is assumed to be the full power design pressure of 1000 psia. There is an upper limit on Tref of 589.2*F, the highest Tref value assumed in the accident analyses.

Full power Tref is initially set to 589.2'F.

The power input to the Tref program comes from turbine impulse pressure. Pimp increases linearly with turbine-generator output and the predicted values may require rescaling to correspond to actual impulse pressures. Pimp data is taken and extrapolated to full power for comparison with the vendor supplied Pimp predictions. Any significant deviations in the Pimp program would have to be corrected by recalibration of the Pimp channel or taken into account in the calculation of the new Tref program. These Tref and Pimp extrapolations are made at 75%, 90% and 100% power only.

Actual pressurizer levels are compared to calculated program levels based on the actual power levels. The level control setpoint values are thus also verified to be proper.

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m __ __- _ - - . ... _ _- . ___ - - _ . ___ _ __ _

l 3.5.6 - STARTUP ADJUSTMENTS OF REACTOR CONTROL SYSTEMS - ISU-020B

-(Continued)

TEST METHODOLOGY.(Continued) {

Refer to Figures 3.5.6-1 through 3.5.6-3 for example plots of Tref, 1 Pimp 'and Pressurizer Level as functions of power and Figure 3.5.6-4 for an example plot of steam generator steam pressure vs. power. ,

SUMMARY

OF RESULTS f Data was taken at the approximate 30% and 50% power levels without J

1 any difficulties encountered.

At the 75% power plateau, Tref was extrapolated to.a full power -

value of 589'F which was verified to be below the design maximum of 589.2*F. Steam generator pressure was extrapolated to 995 psia  ;

which fell within-the allowed design range of 990 to 1010 psia. l Turbine impulse chamber pressure was extrapolated to 885 and 890 psia at 100% power. This compared well with the 870 to 890 psia l allowed range based on vendor supplied predictions. '

At 90% power, Tref was extrapolated to a full power value . of 588.9'F. This was again verified to be below the design maximum of 589.2*F. Steam generator pressure was extrapolated to 999.8 psia  !

which was also within the design range of 990 to 1010 psia.

Turbine impulse chamber pressure was extrapolated to 885 and 890 .

psia at 100% power. This also satisfied the 870 to 890 psia range.

At the 100% power plateau, Tref extrapolated co 589.214*F. While this was not below the design maximum of 589.2*F, the 0.014 of it which exceeded the limit corresponded to a span error of only 0.02%, well within allowed calibration tolerances. This had no 1 adverse impact on test results based on this evaluation. Steam  !

generator pressure was extrapolated to 990.4 psia which was within  !

the range of 990 to 1010 psia. Therefore,.an adjustment to the l Tref program value was not required. The full power extrapolated turbine impulse pressures of 923.0 and 933.7 psia was outside of the 870 to 890 psia allowed range. This 6% difference did not  ;

adversely affect the Tref program use of the Pimp signal, because ,

the output clips at 100%. Pimp power. Therefore, even though Pimp t power would indicate as 106% when actual power was at 100%, the Tref o q ut would still be the 100% value. This 6% error was i evaluated by the NSSS vendor for possible impact on safety analyses and was found to be of low significance such that the full power impulse pressure adjustments were not required to be performed.  ;

Pressurizer level was found to deviate from the calculated program i value by -0.8% at 30% power, -1.3% at 50% power, -1.8% at 75%

power, -2.1% at 90% power and by -0.2% at-1004 power. These all satisfied the -34 to +3% allowed deviation range.  ;

Refer to Table 3.5.6-1 for detailed results.  :

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r I

l TABLE 3.5.6-1 1

Startuo Adiustments Summarv 3p.1 191 2.5.1 90.1 100%

Calorimetric Power (%) 27.06 47.83 78.63 90.09 99.84 l

Tref (*F) 565.48 570.95 582.88 587.28 589.21 Tavg (*F) 565.47 571.33 582.57 586.97 589.28 ,

Pressurizer Level (%) 35.00 41.89 54.64- 59.71 60.32 Calculated Pressurizer Level (%) 34.20 40.58 52.79 57.57 60.10 <

Actual Pressurizer  :

Level Setpoint (%) 34.87 41.64 54.61 59.68 59.97 j Average impulse ,

pressure (psia) 232.74 381.77 707.84 827.67 926.90 l Average steam generator pressure (psia) 1065.0 1040.3 1016.7 1008.0 990.6  :

100% Extrapolated steam generator pressure (psia) N/A N/A 995.0 999.8 990.4 Gross Electric Output (MWe) 232.6 466.5 888.7 1033.4 1149.1 1

1 1

1 4

-156-

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-157-

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3.5.7 - FULL POWER PERFORMANCE TEST - ISU-281B OBJECTIVE This test is performed to demonstrate the reliability of the Nuclear Steam Supply System (NSSS) to maintain its warranted output of 3425 MWth (+0, -5%) for 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> without a load reduction or plant trip resulting from an NSSS malfunction, to demonstrate the ability of the plant to generate 1150 MWe (+0, -5%) for 100 consecutive hours and to demonstrate the ability of the NSSS to develop 3425 MWth (+0, -1%) at a steam generator pressure of >990 psia. _

TEST METHODOLOGY The test is initiated with the plant operating within 5% of its rated NSSS output as determined by power range nuclear instrumentation which is calibrated to correspond to calorimetric power and at a steam pressure of 990 to 1010 psia. Plant conditions are then adjusted to and stabilized at their design values for 100 hours with stability verified by periodic calorimetric data acquisition. Power is verified to be or is increased to be within 1% of the 3425 MWth design NSSS output.

During the 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br />, a four hour performance measurement is performed to verify actual NSSS power by collecting calorimetric data and computing power hourly.  !

SUMMARY

OF RESULTS The 100 hour0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> run was started at 2300 hours0.0266 days <br />0.639 hours <br />0.0038 weeks <br />8.7515e-4 months <br /> on 7-21-93 and completed at 0300 hours0.00347 days <br />0.0833 hours <br />4.960317e-4 weeks <br />1.1415e-4 months <br /> on 7-26-93. The minimum and maximum hourly NSSS power measurements were 98.86% and 99.57% of the rated power of 3425 MWth over the entire 100 hour0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> duration. There was no load reduction or plant trip during the 100 hour0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> duration. A net electrical output of between 1094 and 1107 MWe (95.13% to 96.26% of 1150 MWe) was demonstrated over the 100 hour0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> duration. This satisfied the 1150 MMe (+0, -5%) cr4terion. The average output over the 100 hour0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> duration was 1097.6 We, 95.44% of 1150 MWe. The  ;

four hourly calorimetrics demonstrated NSSS output to be between i 3415.16 and 3417.67 MWth, corresponding to a range of 99.71% to 99.79% power, at a steam generator outlet pressure of 1999.045 psia.

criterion.

This satisfied the 3425 MWth (+0, -1%) at 1990 psia

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3.5.7 - FULL POWER PERFORMANCE TEST - ISU-281B

SUMMARY

OF RESULTS (Continued)

During the 100 hour0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> run, the plant computer had three individual points fail to print out over very limited periods of time, typically less than 2.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />. Additionally, the computer itself failed for a twenty minute period of time. There was no adverse impact on test results based on data trend before and after the failures and on data recorded from alternate sources. Also, post-test calibration checks indicated that two main feedwater venturi flow transmitters were found slightly out-of-tolerance.

This had no adverse impact on test results based on the magnitude and direction of the out-of-tolerance condition. The transmitters resulted in calorimetric power values that were low by up to 0.01%

power, which is the conservative direction with respect to test acceptance criteria.

-162-j

3.5.8 - PLANT COMPUTER SOFTWARE VERIFICATION - ISU-019B OBJECTIVE The Plant Computer Software Verification test is performed to verify that the plant computer receives correct inputs from selected process variables in the field and to validate the Turbine Power Performance Calculations performed by the plant computer.

This verification and validation is performed by comparing the output from the plant computer to permanent plant instrumentation.

TEST METHODOLOGY At plant power testing plateaus at approximately 0%, 50%, 75% and 100% power, selected analog and digital plant parameters monitored by the plant computer are compared to Main Control Board instruments to ensure they agree within a specified tolerance of 12% of full scale. The 613 parameters selected were those judged to be most important to the operators in monitoring plant conditions and evaluating equipment performance. The tolerance is ,

based on the accuracy of the instrumentation and associated instrument loops.

During this test, all of the digital inputs to the plant computer were verified correct. During various power plateau performances of this test, a number of the analog signals could not be verified for various reasons including instruments out of service, instruments off-scale due to plant conditions, instruments out of calibration, and incorrect scaling or processing by the plant computer. Instruments were calibrated, or software corrected, as necessary and retesting was performed at higher power levels.

During this test, only three of the 613 computer addresses completely failed the instrument correlation. These items are to be resolved by Work Orders and retested as part of the Post Work ,

Testing Program. None of these three computer points adversely affects unit operation, all have redundant indication and do not impact computer based calculations.

Data taken associated with the Turbine Power Performance Calculations was transmitted to Engineering for use in rescaling ,

the Turbine Power meter. The existing scaling is adequate for use, the rescaling represents an optimization.

t

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3.5.9 - AUTOMATIC REACTOR CONTROL SYSTEM TEST - ISU-203B OBJECTIVE This procedure is performed to demonstrate the capability of the automatic reactor control system to maintain Reactor Coolant System average temperature (Tavg) within an acceptable tolerance about the reference Tavg (Tref) under steady state and transient conditions.

Tref is the programmed Tavg setpoint as a function of power. This procedure satisfies activities described by FSAR Table 14.2-3, Sheets 4 and 33.

TEST METHODOLOGY With reactor power stabilized at approximately 50% and Tavg matched j within 1.5'F of Tref, the rod control system is placed in automatic to monitor Tavg for oscillations. After approximately ten minutes, Tavg is manually increased to be approximately 5aF higher than Tref by manual withdrawal of Control Bank D. The rod control system is then placed in automatic and Tavg is allowed to return to and f stabilize within approximately 11. 5 'F of Tref by automatically controlled Control Bank D motion. After Tavg has again stabilized, rod control is then placed in manual to decrease Tavg to  !

approximately S*F lower than Tref by manual insertion of Control Bank D. The rod control system is then again placed back in automatic and Tavg again allowed to return to and stabilize within approximately 1.5'F of Tref. Various plant parameters and instrumentation signals within the automatic reactor control loops are monitored during these temperature transients. Values recorded are Tavg, Tref, nuclear flux, turbine power, turbine impulse presaure, steam header pressure, pressurizer pressure and rod control mismatch and error signals.

SUMMARY

OF RESULTS During steady state operation, it was found that Tavg was j maintained within 11.5'F of Tref with no problems. When Tavg was )

increased by 5'F, it took approximately 120 seconds to stabilize l

! Tavg to within 11.5'F of Tref. When Tavg was decreased by 5'F, it I took approximately 220 seconds to stabilize Tavg to within 11.5'F of Tref. ,

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3.5.10 - PRECISION SECONDARY SIDE POWER CALORIMETRIC - PPT-P2-2050-OBJECTIVE This test procedure is performed .to validate- the Plant Computer  ;

calorimetric program and to determine total reactor thermal power i through the use of a secondary-side precision calorimetric. '

TEST METHODOLOGY i Calorimetric power measurements are made at approximate-30%, 50%,-

75%, 90% and 100% power levels. Prior to data acquisition at each-power level, plant parameters are verified to be stable and blowdown is optionally isolated. The subsequent data acquisition period is generally 15 minutes. Calculations involve verifying ,

correctness of input parameters and reactor power level.- The most  !

significant parameters associated with the calculation are loop '

feedwater temperatures, feedwater header' pressure, loop feedwater flows based on the flow element venturies, loop steam . generator pressures, blowdown flow, and Leading-Edge-Flow-Meter (LEFM) feedwater flow.

The methods of determining reactor power include:

• Manual: manual calculation, using computer points for input

• Computer: Plant Computer calorimetric calculation

• LEFM: Manual calculation, using' computer points for input and the LEFM as the basis ~for feedwater flow

• Mini-DAS: off-line computer calculation, using transmitter output voltages as input

• Manual calculation, using Main control Board indications. .  !

Except where the LEFM is referencen, feedwater' flows are based on the transmitters associated with the flow element venturies.

The Plant Computer calorimetric calculation is the primary basis against which the Nuclear Instrumentation System power levels are compared. l I

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3.5.10 - PRECISION SECONDARY SIDE POWER CALORIMETRIC - PPT-P2-2050 (Continued)

SUMMARY

OF RESULTS Reactor thermal power was determined based on calorimetric measurements at approximately 30%, 50%, 75%, 90% and 100% power.

At each power level, calculated input parameters were consistent with computer indications. Reactor power levels, as calculated by the various methods described above, compared well throughout the power ascension. Results from the final performance of this test are as follows:

METHOD POWER (%) DIFFERENCE (%)

Manual 99.34 N/A Computer 99.31 -0.03 LEFM 99.65 0.31 Mini-DAS 99.34 0.00 The allowable tolerance is : 2%. The manual method which uses Main Control Board indications was verified during 30% and 50% power testing.

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_ , _ ._ ,_ . . _ _ __ __ _ . _ ~ .

i

-i L 6.- DEFERRED PREOPERATIONAL TESTING

3. 6.1 - PUBLIC ADDRESS AND ' EMERGENCY EVACUATION ATARM SYSTEM TEST

- PPT-TP-92B-1  !

i OBJECTI'7E  !

The public address and Emergency Evacuation Alarm System Test is  !

performed to demonstrate the capability of the- intraplar.c  :

page party /public addresa system and the emergency evacuation alarm l system to provide adequate communication and audibility. This test >;

satisfies activities described by FSAR Table 14.2-2, sheets 40 and  :

40a.

TEST METHODOLOGY This test is performed with sufficient equipment operating, in the area to be tested, to create an ambient noise level that would be j

expected during nocmal plant operation. Areas that contain emergency related equipment (i.e. DG's. SI, RHR, ev.c.) are tested i with equipment in operation. The Background Noise Level is j measured and recorded. A page is-initiated over the Gaitronics i apeaker system. The Voice Paging sound level in measured and i recorded. The Site-Evacuation Alarm signal is activated. The  !

Site-Evacuation Alarm sound level is measured'and recorded. The  :

Page/ Alarm speaker system shall not produce sound levels in excess l of 115dB. The Page/ Alarm speaker system shall provide output l ievels which are audible over the highest expected ambient noise i levels. l l

SUMMARY

OF RESULTS I The intraplant page-party /public address system provided voice paging output levels which were audible over the highest expected ambient noise levels at all locations tested except for the Train i A and B RHR and SI pump rooms, Room 285 (Lube Oil reservoir room),

and the pipe chase at the north end of the safeguards hallway elevation 790'. These deficient areas were identified under ONE Forms93-742 and 93-1086.

The Emergency Evacuation Alarm signals were audible during the highest expected anbient noise levels at all locations tested '

except for the Train A and B RHR and SI pump rooms, Room 285, and the pipe chase at the north end of the safeguards hallway elevation 790'. These deficient areas were identified und6r ONE Forms93-742 i and 93-1086.

ONE Form 93-742 addressed inadequate output levels of the Public i Address and Emergency Evacuation Alarm signals in the Train A and B RHR and SI pump rooms and the pipe chase at the north end of the Safeguards Building hallway elevation 790'.

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i f

3.6.1 - PUBLIC ADDRESS AND EMERGENCY EVACUATION ALARM SYSTEM TEST ,

- PPT-TP-92B-1 (Continued)

SUMMARY

OF RESULTS (Continued)  :

ONE Form 93-1086 addressed inadequate output levels of the Public Address and Emergency Evacuation Alarm signals in the Unit 2 i Turbine Lube Oil Reservoir Room (Room 285). It is expected that the disposition of ONE Forms93-742 and 93-1086 will result in a  !

design modification to install additional speakers.

Until such changes can be implemented, compensatory measures have  :

been established to dispatch a security officer to the affected  !

areas for personnel notification of an alarm condition. Reference .

Office Memorandum CPSES-9307394, CPSES-9307173 and letter, TSEC l 93050.

I TRG Action Item 93-052-01 was assigned to the completion of ONE Forms93-742 and 93-1086. A 10CFR50.59 Safety Evaluation for continued operation was approved.  :

i t

b s

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i

i 3.6.2 - SECONDARY SAMPLING DEFERRED PREOP TEST - PPT-TP-93B-13 OBJECTIVE i

.This test is performed to verify that the flows, temperatures and -

pressures attained at the secondary sampling station, for Main  :

Steam after the Moisture Separator Reheater (MSR) 2 B, meet design  !

criteria. This test satisfies activities described in FSAR Table ,

14.2-2, sheets 6 and 6a.

TEST METHODOLOGY A sample is obtained at thm secondary sampling station for Main Steam after the MSR 2-B. During the sample process the following are verified:

a) grab sample flow is 2 500 ml/ min b) sample temperature at 2-TI-3807J is s 77*F (85'F for grab sample) c) sample pressure at 2-PI-3805J is s 25 PSIG

SUMMARY

OF RESULTS l The test was successfully performed with one exception. Grab sample flow was below the Review Criterion of 500ml/ min. ONE Form -

93-1450 was issued to document the failed Review Criterion. The !

disposition for ONE Form 93-1450 stated that, the 500ml/ min grab sample flow specified in the Design Basis Document is not a j required sample flow. This flow is the design maximum flowrate for i sample cooler sizing.

i

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3.6.3 - Featina, Ventilation and Air Conditioninc System Air Balance - X-ME-1,1 OBJECTIVE .

This test is to demonstrate adequate flow to areas served by the '

Primary Plant Ventilation System. This test satisfies activities described by FSAR Table 14.2-2, sheets 29 and 29a.  ;

TEST METHODOLOGY l Verify the Primary Plant Ventilation System delivers design air flows by performing a system air balance in accordan:e with EGT-164, HVAC Air Balance.

SUMMARY

OF RESULTS This testing was successfully performed under W.O.# 1-92-031500-00. ,

Technical Evaluation #93-654 has been. evaluated for out of tolerance air flows. All items were determined to be acceptable.

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3.6.4 - MAIN STEAM ISOLATION VALVE TEST. 2HV-2334A - PPT-S2-9501A OBJECTIVE This test is to demonstrate the MSIV 2HV-2334A actuates to full closure A SSPS. within 5 seconds after receiving a close signal from Train This test. satisfies activities described by FSAR Table 14.2-2, sheets 50 and 50a.

TEST METHODOLOGY This test is performed in Mode 2, Mode 3 or in Mode 4 above 300*F.

A recorder is connected to the MSIV position indication circuits and to a test switch which has an input to the Train A Solid State Protection System that can actuate slave relay 2-K627-A. MS1V j 2HV-2334A is then closed by operation of the test switch and the l response time from the test switch actuation to the MSIV fully closed indication is determined from the recorder traces.

SUMMARY

OF RESULTS MSIV 2HV-2334A was response time tested from the Train A Solid State Protection System. The recorded closure time was 4.04 seconds.

criterion of 5.0 seconds. The valve satisfied the maximum allowed closure time

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3.6.5 - STEAM DUMP VALVE TESTING - PPT-P2-2023 OBJECTIVE This test is to demonstrate that the Steam Dump Valves open and close within the design stroke time. This test satisfies activities described by FSAR Section 10.4.4. ,

TEST METHODOLOGY l

The Steam Dump Valves are stroke tested hot with the downstream ,

block valves closed. The valves are stroke tested as follows:

a) each valve modulates from full closed to full open within 20 seconds b) each valve modulates from full open to full closed within 20 seconds c) each valve trips closed within 5 seconds, on a F;eactor Coolant Lo-Lo Tavg signal on Train A d) each valve trips closed within 5 seconds, on a Reactor Coolant '

Lo-Lo Tavg signal on Train B e) each valve trips open within 3 seconds, on a Plant Load i Rejection Signal.

SUMMARY

OF RESULTS The Steam Dump Valves were successfully stroke tested as follows:

a) Each valve modulated from full closed to full open within 20 seconds. The opening times ranged from 7.9 to 14.6 seconds.

b) Each valve modulated from full open to full closed within 20 seconds. The closing times ranged from 8.9 to 17.2 seconds. i c) Each valve tripped closed within 5 seconds, on a Reactor  !

Coolant Lo-Lo Tavg signal on Train A. The closing times (on a simulated Lo-Lo Tavg, Train A) ranged from 2.1 to 3.6 seconds.

d) Each valve tripped closed within 5 seconds, on a Reactor Coolant Lo-Lo Tavg signal on Train B. The closing times (on I a simulated Lo-Lo Tavg, Train B) ranged from 2.1 to 3.1 seconds. i i

e) Each valve tripped open within 3 seconds, on a Plant Load Rejection Signal. The opening times (on a simulated Load l Rejection) ranged from 1.4 to 1.9 seconds. ]

During the test performance at Step 8.18 valve 2-TV-2370H appeared  !

slow when tripped closed. The solenoids, tubing, diaphragm and j volume booster on valve 2-TV-2370H were inspected and cleaned under l WO# 1-93-042128-00. Post Work testing of WO# 1-93-042128-00 l tripped the steam dump valves closed several times from both A and B solenoids. All valves stroked within tolerance.

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l 3.6.5 - STEAM DUMP VALVE TESTING - PPT-P2-2023 (Continued)

SUMMARY

OF RESULTS (Continued)

During the performance of Step 8.24 the first bank of Steam Dump Valves partially opened when the mode selector switch was placed in Tavg. The drifting of the 2-TY-500 NMA card was investigated by WO# 3-93-326365-01. A ONE Form was written to address the card drift problems caused by unnecessary jumpers to unused circuits of card 2-TY-500. The unnecessary jumpers were removed per DCN 6025 RO. The steam dump mode switch was changed from pressure control to Tavg control to verify the steam dump valves remained closed, and the results were satisfactory.

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3.6.6 -

THERMAL EXPANSION TEST OF EXTRACTION STEAM TO AUXILI7E STEAM SYSTEM - ISU-308B OBJECTIVE To demonstrate that the Extraction Steam line to the Auxiliary Steam System:

a) experiences thermal expansion consistent with design b) support components do not interfere with pipe thermal growth.

This test satisfies activities described by FSAR Table 14.2.2, sheets 52 and 52a.

TEST METHODOLOGY This test is conducted at ambient temperature (at or below 120*F) and at power level of approximately 100% reactor power. Thermal expansion of piping and functioning of restraints are monitored at predetermined test points. Test data is then compared to predictions. If observed thermal expansion is not within the specified tolerance, support systems will be examined to determine the cause or to verify correct function. If binding is found, the restraints will be adjusted (to eliminate the unacceptable binding) or verified acceptable (through engineering analysis).

SUMMARY

OF PESULTS All required system walkdowns were performed. Problem Reports 062 and 063 were written to address problems noted in the field. Based on Engineering's evaluation, Problem Report 062, a spring hanger pin in contact with the bottom of the spring can, was determined to be " Acceptable as is". For Problem Report 063 Action Request

1. 146824 was initiated to correct a hard hit from a vertical hanger rod by replacing and moving the bent rod.

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3.6.7 - Reactor Cavity Humidity Detectors i OBJECTIVE i

The reactor cavity humidity detector in the CRDM area, 2-ME-5460-4, l is tested to demonstrate proper operation of the system including ^

the PACER (Programmable Automatic Containment Error Reduction) circuitr/. Proper operation of the humidity detector is demonstrated by comparing the. control room indication of containment dew point temperature with local. measurements using a suitable portable psychrometer. This test satisfies, in part, activities described in FSAR Table 14.2-2, sheet 23 and the deferred preoperational testing in system 2-6500, Containment Atmospheric Monitoring.  ;

l TEST METHODOLOGY j

The Containment Dew Point Channel 5460-4 is verified calibrated and i accurate by performance of INC-4088B, " Channel Calibration Containment Dew Point, Channel 5 460-4" . The main control board Dew Point Temperature indicator on CB-03A is recorded and compared to measurements by a portable humidity detector taken in the area of the CRDMs. The values are expected to agree within 2 Sk'F of each other.

SUMMARY

OF RESULTS l 1

The dew point channel was satisfactorily calibrated per INC-4088B.

The PACER module operation was satisf actorily tested per Work Order 4-93-038569. The CRDM area humidity was measured with a motorized )

sling psychrometer and converted to a dew point temperature of J approximately 55'F. The control room indication by channel 5460-4 (

was 53'F, which was within the criterion of Sk'F. l l

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3.6.8 - Pressurizer' Power Operated Relief Valve Leak Tichtness

- ISU-021B OBJECTIVE The Pressurizer PORVs 2-PCV-0455A and 2-PCV-0456 had detectable seat leakage during the hot functional testing. The valves have been repaired and a Westinghouse Field Change was implemented to install spacer rings in the valves. The objective of this test is to verify the valves are leak tight. This leakage test satisfies, in part, the deferred preoperational testing in System 2-5500, Reactor Coolant System.

TEST METHODOLOGY With Unit 2 in Mode 3 at Normal Operating Temperature (557'F) and Normal Operating Pressure (2235 psig), the temperature of the PORV discharge piping is measured and the annunciator for high outlet temperature is verified to be not actuated. Leakage past the PORVs seats would cause the discharge piping temperature to rise and cause the annunciator to alarm.

SUMMARY

OF RESULTS The discharge piping of the PORVs was measured to be at 104.2'F, well below the pressurizer vapor temperature of 650.8'F. The annunciator 2-ALB-5B(3-1), PRZR PORV OUT TEMP HI remained clear.

This satisfies the leak tightness criteria for the PORVs.

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3.6.9 - Pressurizer Sorav Valve Leakace - ISU-021B OBJECTIVE The pressurizer more sprayduring than expected valvesthe 2-PCV-0455B het and 2-PCV-0455C leaked have been reworked and are to be retested with Unitfunctional The valves testing.

2 at Normal Operating psig). Temperature (557'F) and Normal Operating Pressure (2235 This test satisfies, in part, the deferred preoperational testing in system 2-5500, Reactor Coolant System.

TEST METHODOLOGY With Unit 2Pressure Operating at Normal Operating Temperature (557'F) and Normal (2235 psig), the pressurizer spray valves 2-PCV-0455B and 2-PCV-0455C and the loop 1 spray bypass valve 2RC-8051 are closed.

SPRAY LN TEMP LO" will The annunciator alarm 2-ALB-SC as the loop 1 spray(3.2) "PRZR ANY line cools. This confirms 2RC-8051 2-PCV-0455B leakage is acceptably low. Bypass valve is then throttled open until the annunciator clears and the loop 1 spray line has sufficient flow to keep warm.

4 spray bypass valve 2RC-8052 is then closed and annunciator 2-ALB-The loop i SC (3.2) monitored. This annunciator will alarm as the loop 4 l spray line cools.

low. Bypass valveThis confirms 2-PCV-0455C leakage is acceptably 2RC-8052 is then throttled open until the annunciator keep warm. clears and the loop 4 spray line has sufficient flow to l

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SUMMARY

OF RESULTS l

Pressurizer spray valve l acceptably low leakage by 2-PCV-0455B the loop 1was demonstrated to have spray line temperature decreasing were closed.from 532*F to 498.5'F when the spray and bypass valves temperature of The annunciator 526.2*F and cleared2-ALB-5C (3.2) actuated at Loop 1 at 538.3*F.  !

2RC-8051 was throttled to 1% turns open as its final position. Bypass spray valve l Pressurizer spray valve )

2-PCV-0455C was demonstrated to have above 537'F with it and bypass4spray excessive leakage by the loop spray line temperature remaining valve 2RC-8052 closed. The leakage through 2-PCV-0455C is approximately 3 gpm. Normal continuous spray through the bypass valve is approximately 1 gpm.

This flow rate is too small to create transient or control problems.

New valve internals (ball and seat) will be installed when the pressurizer can be cooled and drained down to allow the work.

This is expected to be during the first refueling outage,  ;

In the interim, the leakage does not affect continued operation. i l

The spray bypass valve 2-RC-8052 was throttled to 1/16 turn open as its final position until the spray valve can be reworked.

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CHEMICAL AND VOLUME - CONTROL SYSTEM (CVCS) MIXED BED DEMINERALIZER 2-01 PRESSURE DROP - WO#1-93-034962-00 QBJECTIVE

To verify'that the CVCS Mixed Be'd Demineralizer 2-01 pressure drop i

.with letdown flow is consistent with design. This test satisfies activities described by FSAR Table 14.2-2, sheets 11 and lla. ,

TEST METHODOLOGY ,

This test is' conducted with a letdown flow in.the range of 112-129  !

gpm through CVCS Mixed Bed Demineralizer 2-01.- The pressure drop j across CVCS Mixed Bed Demineralizer 2-01 is verified to be less  !

than 12.8 psid.  !

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SUMMARY

OF RESULTS l This test was successfully performed under WO#1-93-034962-00. The  !

CVCS Mixed Bed Demineralizer 2-01 pressure drop was 12 psid with a +

letdown flowrate of 115 gpm.

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3. 6.11 - PTMT COMPUTER SOFTWARE VERIFICATION l 3.6.11.1 Flur Mannina Software Module - PPT-TP-93B-1 OBJECTIVE This test is performed to demonstrate that the Incore Movable  !

i, Detection System software module on the Unit 2 Plant Computer. l functions as required and calculates and displays accurate plant parameters.

14.2-2, sheet 38.

This test satisfies activities described by FSAR Table 1 l

TEST METHODOLOGY t

This test ' verifies that the Incore Movable Detection' System 5 software _ module on the Unit 2 Plant Computer records and displays accurate information from plant parameters using both direct and  ;

processed information. This is verified by successfully performing i the Incore Movable Detection software test and comparing the output j to the expected values. Also, this test verifies the Flux Mapping '

Module output is successfully downloaded on to a diskette in a  !

format that is acceptable to Site Reactor Engineering. {

SUMMARY

OF RESULTS k; i

The Incore Movable Detection software test demonstrated that the  :

software module is capable of accurately collecting data from a i pass map. The performance of the calculations in Section 8.3  ;

verified the accuracy of the software module. Technical i Evaluation, TE 93-720, was written for Engineering to evaluate the l correct technique for the software to determine and label bad data- 1 points. This TE does not impact functionality of the software 1 module or impact Nuclear Engineering. Site Reactor Engineering acknowledged that the output was in a format that was acceptable. c i

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3.6.11.2 Plant Comouter CPU Ucarade and Software Verification Deferred Test - PPT-TP-93B-3 System Integration Test, Section 8.2 Multiplexor Hardware-Test, Section 8.3 '

Plant Mode Test, Section 8.4 Source Range Group Alarm Inhibit Logic, Section 8.5  ;

OBJECTIVE This test demonstrates that the upgraded. Plant Computer hardware and software processes and displays proper plant parameters for the following items: t Upgraded Plant Computer Hardware and Software System Integration (SDR #3654)

Multiplexor Stability (SDR 73607)

Vided Hard Copier Machine (SDR #3653)

Additional Plant Mode Logic ,

Source Range Group Alarm Inhibit Logic (SDR #3676)  !

This test satisfies activities described by FSAR Table 14.2-2, ,

sheet 38.

i TEST METHODOLOGY This test verifies that the upgraded Plant Computer outputs and displays information from plant parameters using both direct and processed information. This is accomplished by exercising the Plant Computer by concurrently requesting large numbers of different Displays, Log Printouts, Trend Printouts, Alarm >

Printouts, Video Hard copies, Post Trip Reports, SOE Logs, and repetitive Software Calculations at several locations. Performance of the System Integration Test is verified by performing the procedural steps and comparing the printouts and responses.

SUMMARY

OF RESULTS ,

The System Integration Test was performed and various hardware problems were encountered. These problems were identified on TE 93-355 and have since been corrected. TE 93-356 was written for Engineering evaluation of the CPU Reserve Capacity not staying above 50% during the performance of the test. Engineering accepted the Test Results based on the reserve capacity remaining above 50%  ;

for the majority of the test. Engineering also noted that the hardware problems encountered during the test did not impact the test results, i

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3.6.11.2 - Plant Comouter CPU Ucarade and Software Verification Deferred Test - PPT-TP-93B-3 (Continued) [

SUMMARY

OF RESULTS (Continued)

To resolve the problem of the Unit 2 Plant Computer crashing when {

there is a loss of power or communication to a multiplexor, the '

Timeplex FDDI bridges were replaced with Fibronics FDDI bridges per

• Minor Modification 93-382 and DCN 6208. The Multiplexor Hardware Test was successfully run and verified that the problem was resolved. PCN PPT-TP-93B-3-RO-2 was implemented and successfully run which verified that the Fibronics bridges accurately transfers data from the multiplexors to the CPU. However, it was observed a -

few days later that the multiplexors were randomly dropping off-line at a rate of thirty times a day. ONE Form 93-1053 was written to identify the problem. The original problem had been resolved but a new problem had been created. ONE Form 93-1059 was written and DCN 6339 was implemented to replace communication cables which badly fit the Bridges. ONE Form' 93-1053 was written and DCN 6344 i was implemented to modify the f ront-end sof tware. On June 7, 1993,

• Retest 1 was successfully performed to rerun Section 8.3 which i verifies that the CPU can accurately monitor data from the multiplexors. However, it was observed a few days later that the multiplexors were still randomly dropping off-line but at a slower rate of five times a day. ONE Form 93-1053 is still open and Design Engineering and the vendor are still working on the resolution to the problem.

The Source Range Group Alarm Inhibit Logic test was successrully performed.

t The Test Review Criterion was met and no problems were observed.

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l 3.6.11.3 Plant Comouter Software Deferred Test - PPT-TP-93B-9 l Reactor Protection System Monitor, Section 8.2 Delta Flux Test Mode Verification , Section 8.3 Primary Plant Performance Deferred Testing, Section 8.4 Data Archive, Section 8.5 OBJECTIVE I This test demonstrates that the Plant Computer RPSM, DELTA I, and Primary Plant Performance Software processes, calculates and <

displays accurate plant parameters. Also, this test verifies that.

the Plant Computer can store seven (7) days of Archive data. This test satisfies activities described by FSAR Table 14.2-2, sheet 38. ,

TEST METHODOLOGY ,

This test verifies that the Plant Computer outputs and displays information from plant parameters using both direct and processed information. This is accomplished by successfully performing:

• The Reactor Protection System Monitoring test and verifying proper output to the alarm and trend printers and

• the Primary Plant Performance Software test and comparing the output to the expected values in the body of the procedure.

This test verifies that the Plant Computer displays the proper-Delta flux tent display and the test mode is ended automatically ,

when the 16 hour1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> limit is reached and the accumulation of penalty minutes automatically resumes. ,

This test verifies seven days of Archive data by successfully performing an Archive Retrieval Special Report Process for the previous seven days. ,

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SUMMARY

OF RESULTS The Reactor Protection System Monitoring test was performed and the following reactor trip groups displayed faulty logic:

Group 1 - Source Range High Flux Reactor Trip Group Group 2 - Intermediate Range High Flux Reactor Trip Group Group 3 - Power Range Low Flux Reactor Trip Group Group 7 - Pressurizer Low Pressure Safety Injection Group ,

Group 16 - RCS Low Loop Flow Reactor Trip Group Group 19 - Turbine Trip Group

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M 3.6.11.3 Plant ConDuter Software Deferre.

(Continued)

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SUMMARY

OF RESULTS (Continued)

A Technical Evaluation, TE 93-632, was :<

evaluation. Engineering determined t

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enhancement that is not functional on Un.

met TU's expectations, so the software moci the Plant Computer design. This test i applicable.

-"" j The Delta I test was successfully perfl resumed accumulating after the 16 hour1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> tt Flux tent was properly displayed.  !

The Primary Plant Performance Software, performed.

The results matched the exp!

allowable range. '

Before the performance of this test TE 93-435, was initiated for EngineE i

calorimetric ' calculations not operable loops. It was determined being pi, that '

performed only when four reactor coolant .

this is not a problem now.

Plant Computer documentation DCN to 5809 wa; remove ti loop calorimetric.

t The Plant Computer successfully stored dat )

seven (7) days. 1 the compression limits of individual addr;Af Archive File was 67% full at the end of ti a ,

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3.6.11.4 Diaital Outout for the Control Room Alarm Panel

- PPT-GO-1001

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OBJECTIVE This- test satisfies the Startup component functional test requirements of the Alarm Acknowledge digital output for the Control Room Alarm Panel. This test satisfies activities described by FSAR Table 14.2-2, sheet 38.

TEST METHODOLOGY  ;

This test verifies that the electrical control circuit functions in I accordance with latest design drawings including all inputs to the  !

computer. The control circuit functional test verifies each new .

alarm annunciates in the Control Room, both alarm acknowledged l pushbuttons illuminate, and alarms on the Plant Computer are .

acknowledged by depressing either pushbutton resulting in the  !

pushbuttons not being illuminated.

SUMMARY

OF RESULTS The initial performance of the test failed. DCN 6464, Rev. 2, was incorporated to correct software ' problems and implement wiring -

changes. The test was reperformed successfully. During the test j it was verified that both pushbuttons illuminated when a computer ,

alarm came in, the buzzer annunciated and then timed out, pressing either pushbutton acknowledged the alarms on the DAD display, The

~ pushbutton lights extinguished, computer input Y9999D toggled and indicated correct state, and the digital output D0013 operated  ;

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3.6.11.5 Modem Sharina Device for the Radiation Monitorina l Data Link - W.O. 1-93-038782-00 '

OBJECTIVE j l

This test is performed to verify the modem sharing device for.the radiation monitoring data link is operable and the data stream is ,

being received by the Plant Computer. This test satisfies !

activities described by FSAR Table 14.2-2, sheet 38. i TEST METHODOLOGY j l

This test verifies the data from the Radiation Monitoring Computer, PC-11, is received by the Plant Computer. The readings of the Plant Computer inputs are compared to the readings from the l Radiation Monitoring Computer.

SUMMARY

OF RESULTS '

The data was accurately received by the Plant Computer. The !

readings of the Plant Computer inputs were compared to the readings from the Radiation Monitoring Computer. The agreement of the i readings verified that the modem sharing device was working properly and the Plant Computer was receiving the data stream.

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3.6.11.6 Comoonent Testina - Various Documents OBJECTIVE L

To verify proper computer indication of field devices for Plant Computer inputs. These tests satisfy activities described by FSAR Table 14.2-2, sheet 38.

i TEST METHODOLOGY These tests verify Plant Computer indication of field processes by i calibration or comparison to plant indication. One hundred ninety- -

three (193) computer inputs are to be tested. L

SUMMARY

OF RESULTS One hundred eighty-nine (189) inputs were verified. Twenty-seven (27) inputs were tested through design modification testing, eighty-nine (89) were tested through Performance and Test functional tests, thirty-nine (39) were tested through I&C >

calibration procedures and functional tests, thirty-two (32) were -

tested through Startup preoperational and component tests, and two (2) were spared during the Startup program and do not need to be tested. .

Four inputs QO320D, QO321D, QO340D and Y8001A have not been tested  !

due to discrepancies in the field.  ;

Point QO320D, AUX XFMR 2UT PULSE MWH, is used by the computer to .

accumulate the output of transformer 2UT over the last hour or day.

Design Engineering is resolving pulse coefficient discrepancy per ONE Form 93-942.

Point QO321D, STARTUP XFMR XST1 PULSE MWH, is used by the computer to accumulate the output of transformer XST1 over the last hour or  !

day. Design Engineering is resolving pulse coefficient discrepancy per ONE Form 93-942.

Point QO340D, GEN GROSS PULSE MWH, is used by the computer to accumulate the output of the main generator over the last hour or ,

day. Design Engineering is resolving pulse coefficient discrepancy per ONE Form 93-942.

I&C is troubleshooting point Y8001A, GEN CORE MON-PARTICULATE CONC, under Work Order 1-93-046315. I i

TRG Open Item 28 was opened to track these Plte; Computer Points until the issue is resolved. l 1 l

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3.6.11.7 Control Room Printers Failover - Letter CPSES 9307342  !

-i OBJECTIVE i To verify the Control Room printer failover function is operable. l TEST METHODOLOGY f

None ,

SUMMARY

OF RESULTS Due to the nature of the printers involved, the vendor, SAIC, is unable to provide the printer failover function. Design Engineering Organization has decided to delete this rec:uirement for the Unit 2 Plant Computer and therefore no test .s required.  !

Engineering's position is outlined in letter CPSES-9307342. ,

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3.6.11.8 Boron Follow Software Module - Letter CPSES-9307514 j OBJECTIVE To verify the proper performance of the Boron Follow Software Module.

TEST METHODOLOGY i None

SUMMARY

OF RESULTS The Boron Follow Software Module did not meet the expectations of TU Electric. The module was deleted from the Plant Computer '

Design. Engineering position is outlined in letter CPSES-9307514.

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t 4.0 - REFERENCES J I

1) Comanche Peak Steam Electric Station Final. Safety Analysis  :

Report l

^

2) Regulatory Guide 1.68, Revision 2 i

3) Regulatory Guide'1.68.2, Revision 1

4) Regulatory Guide 1.133, Revision 1  !

5) Comanche Peak Technical Specifications l

6) Comanche Peak Operating License-NPF-88 I

7) Comanche Peak Operating' License NPF-89 i

8) CPSES Unit 2, Cycle 1 Startup and Operations Report, Rev. 1 I

9) Westingnouse NSSS Startup Manual  ;

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ATTACHMENT A TO TXX-93325 COMANCHE PEAK STEAM ELECTRIC STATIC!

UNIT 2 CYCLE 1 STARTUP REPORT b

190

2 ARJTRACT '

Regulating Guide 1.133, Revision 1, May 1981. Section c.3.a. (2)(a) requires the I power operation alert levels of the Loose Parts Monitoring System (LPMS) to be reported to the Nuclear Regulatory Commission following the initial startup program. This report was written to comply with this requirement.

The power operation alert levels for the individual LPMS channels were based on the comparison and analysis of; +

(1) The basic system sensitivity during plant shutdown, and (2) The background noise measured during normal plant power operation.

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LPMS Special Report Table of Contents Section ,

P15Lt 1.0 Introduction 1 1.1 System Description 1 Table 1 - CPSES LPM Sensor Location 2 i

j 2.0 Methodology 3 2.1 Data Acquisition 3 2.2 Data Analysis 3 ,

Table 2 Background Data 4 Table 3 Background G's to Impact Response 5 3.0 Data .

3.1 Impact Baseline Fensitivities 7-27 3.2 Impact Frequency Spectrum 28-35 l

l 3.3 Background Amplitude (Power Operation)36-136 i

3.4 Background Frequency Spectrum (100% Power) 137-157 j 3.5 Additional System settings 158 Table 4 Additional System settings 159 4.0 Results and Recommendations 160 )

i Table 5 Power Operation Alarm Settings 161 s

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I 1.0 INTPoDUCTroN The purpose of this report is to:

• Describe the loose-parts monitor system for CPSES.

Describe the methodology used to measure impact energy sensitivity.

Discuss reduct.on.

the results of LPMS associated test and data Document the LPMS power operation alert levels at completion of the ISU test program.  ;

1.1 System Description The Loose Parts Monitoring System employed at CPSES is the Babcock

& Wilcox (B&W) Loose Parts Monitor System IV. This system is designtd to n onitor, alarm and diagnose loose parts.

The Loose Parts Monitoring System uses an array of active and passive accelerometers to detect metal to metal impacts indicative of loose parts in the RCS. The accelerometers are permanently installed at selected locations where a loose part would tend to collect or impact. The locations of the accelerometers provide a high degree of reliability in the detection of metal to metal impact in the reactor coolant system. Two accelerometers are located at each of the following locations:

a Reactor Pressure Vessel - Upper Head Region ,

e Reactor Pressure Vessel - Lower Region --

e All Steam Generato:rs - Primary Ccolant Inlet Region

• one accelerometer is located at each of the following locations: 4

• All Reactor Coolant Pumps e All Steam Generators - Upper Region The accelerometers operate on the piezoelectric crystal ef fect with deformation of the crystal producing a measurable proportional charge. This signal is amplified, filtered, and conditioned to accentuate the frequency band known by measurement to correspond to metal to metal impacts. The system is designed to alarm the presence of unusual noises above normal background noise present in the plant. The system comprises eight active and twelve passive accelerometer channels. The active channels are monitored continuously for detection of loose parts. The passive channels are used for diagnostics (location determination) of a loose part that has been detected.

Sensor locations and LPM used for signal processing are shown in Table 1.

Page 1

TABLE 1 CPSES Loose Parts Monitoring Sensor Locations PROCESS-SENSOR MODULE TYPE LOCATION 1 LPM 1 ACTIVE SG "A" Inlet 1 2 PASSIVE SG "A" Inlet 2 3 PASSIVE RCP 201 4 PASSIVE SG "A" L Tap 5 LPM-2 ACTIVE SG "B" Inlet 1 6 PASSIVE SG "B" Inlet 2 7 PASSIVE RCP 202 8 PASSIVE SG "B" L/ Tao 9 LPM-3 ACTIVE SG "C" Inlet 1 10 PASSIVE SG "C" Inlet,2, 11 PASSIVE RCP 203 12 PASSIVE SG "C" L/ lap 13 LPM-4 ACTIVE SG "D" Initt 1 14 PASSIVE SG "D" Inlec 2 15 ._ PASSIVE RCP 204 16 PASSIVE SG "D" L/ Tap 17 LPM-5 ACTIVE RV IGT 017 18 LPM 6 ACTIVE RV ICT 018 19 LPM-7 ACTIVE RV TOP 019 l 20 LPM-8 ACTIVE RV TOP 020 Page 2

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2.0 METHODOLOGY 2.1 Data Acouisition f 2.1.1 Impact of theenergy data was collected during plant shutdtwn as part pre-operational i test program under procedure 2CP-PT-91-01 SFT. The data was collected using the integral ,

i computer system of the LPMS-IV. Sensitivity curves for each '

sensor / hammer combination were then generated. Impacts were imparted using calibrated weights.of 0.25 lba, 3 lba and 30 lbe within 3 feet of the sensor at a kinetic energy of approxi.ately 0.2 to 4.0 ft-lbs.

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2.1.2 Operational baseline (background noise) data was stored at reactor power levels of 0, 30, 50, 75 and 100% during the L initial plant startup program under procedure ISU-2118. l 2.2 Data Analysis 2.2.1 The system digitalprovided storage and display capacity of the B&W LPMS-IV- f sensitivity curve output for each sensor / hammer combination. This system was utilized in the ,

i same levels.

manner during startup to record the background-baseline the different These background baseline levels are tabulated for ,

startup power levels in Tsble 2. i Sensor response to each hammer's 0.5 ft-lb impact was then- !

extracted from the sensitivity curves of test 2CP-PT-91-01 and tabulated in Table 3. .

The 100% power background baselire levels of Table 2 were then "

transferred to backgroundtoratios.

Table 3 in order to calculate impact response J Where sensors had failed at 100% power, a representative level based upon trends at other power levels for that sensor was used instead. -

These ratios will then be used for setting alarm levels for active sensors.

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TABLE 2 POWER OPERATIONS BACKGROUND DATA

SUMMARY

BACKCROUND LEVELS (C's pk) l LOCATION

/ POWER LEVEL 0% =30t 2504 =75% =1004- i SCA IN1 0.040 0.041 0.042 0.040 0.041 l SCA IN2 0.048 0.049 0.050 0.050 0.049 l SGB IN1 0.043 0.045 0.045 0.044 0.045 SCB IN2 0.049 0.051 0.051 0.051 0.052 SCC IN1 0.048 0.051 0.051 0.050- 0.051  !

f SCC IN2 0.051 0.051 0.051 0.051 0.052 l SCD IN1 0.125 0.130 0.138 0.170 0.195  !

SCD IN2 0.113 0.110 0.112 0.111 0.114 I RCP 2-01 0.0081 0.065 0.065 0.062 0.063 l RCP 2-02 0.67 0.73 0.74 0.80- 1.03 l RCP 2-03 0.51 0.49 0.48 0.48 0.22 l RCP 2-04 0.44 0.43 0.43 0.43 0.35 SGA L/ TAP 0.035 0.041 0.040 0.040 0.04!

SCB L/ TAP 0.033 0.039 0.037 0.038 0.042 ,

SCC L/ TAP 0.028 0.03E 0.030 0.031 0.037 _

SCD L/ TAP 0.0041 0.039 0.37 0.033 0.039 RV ICT 017 0.055 0.058 ) 0.056 0.054 0.011 1 ]

RV ICT 018 0.055 0.056 0.055 0.058 0.057 I RV TOP 019 0.093 0.094 0.094 0.096 0.097  !

RV TOP 020 0.0041 0.246 0.255 0.265 0.044 1 I 1 - Sensor Malfunctioning l

Page 4 i

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TABLE 3 COMPARISONS AND RATIOS OF SENSITIVITY TO BACKGROUND LEVELS

• BACKGROUND G's TO IMPACT RESPONSE BACKGROUND 1/4 LB 3 LB 30 LB READING

/ LOCATION (G's) (G's) Ratio (G's) Ratio (G's) Ratio l

SGA IN1 0.041 0.70 17 0.15 3.6 0.04 .98 SGA IN2 0.049 1.2 24 0.48 9.8 0.6 12 SGB IN1 0.045 0.77 17 0.25 5.6 0.085 1.9 SGB IN2 0.052 0.3 5.8 0.24 4.6 0.085 1.6 SGC INI O.051 0.96 18 0.16 3.1 0.12 2.4 SGC IN2 0.052 0.43 9.2 0.25 4.8 0.06 1.2 SGD IN1 0.195 10 0.55 2.8

__1. '3 5 0.10 .51 SGD IN2 0.114 0.73 6.4 0.29 2.5 0.08 0.7 RCP 2-01 0.063 2.05 32 0.6 9.5 0.05 0.8 RCP 2-02 1.03 2.6 2.5 0.6 .58 0.08 .07 RCP 2~03 0.22 2.3 10 0.65 3.0 0.08 .36 RCP 2-04 0.35 3.05 8.7 1.14 3.2 0.07 0.20 ,

1 SGA L/ TAP O.045 0.8 17 0.8 17 0.5 11 SGB L/ TAP O.042 1.6 38 0.85 20 0.45 10 SGC L/ TAP O.037 1.5 40 0.7 18 l

0.35 9.4 '

SGD L/ TAP O.039 1.2 30 0.7 17 0.30 7.7 RV IGT 017 0.058' O.35 6.1 0.07 1.2 0.01 0.18 RV IGT 018 0.057 2.5 43 0.33 5.8 0.11 1.9 RV TOP 019 0.097 4.6 47 1.35 13 0.10 1.01 RV TOP O20 0.265' 1.41 5.3 .3 0.34 { 0.04 0.15 1 - Due to sensor malfunction, highest level of data collected at other power levels was used.

Page 5

2.2.2 In order to have a basic understanding of the LPMS signal processing at CPSES and the desired sensitivity requirements as stated in Reg. Guide 1.133 the following is provided:

NRC Regulatory Guide 1.133 makes the following statement {

concerning sensitivity.

"The on-line sensitivity of the automatic detection system should be such that, as a minimum, the system can detect a metallic loose part that weighs from .25 lb (0.11 kg) to 30 lb (13.6 Kg) and impacts with a kinetic energy of 0.5 f t-lb (0.68 joules) on the inside surf ace of the reactor coolant pressure boundary within 3 feet (0.91 meter) of a sensor. If the ,

recommended sensitivity cannot be achieved by automatic alert because of specific in-plant conditions, these conditions and the actual on-line sensitivity should be specified at the time .

the alert level is provided. As an example, one acceptable I method of verifying this on-line sensitivity is to demonstrate .

(1) the basic system sensitivity during the plant shutdown and (2) that the background noise measured during normal plant operation is sufficiently small that the signal associated with the specified detectable loose-part impact would be clearly discernicle in the presence of this background noise."

As the Commission position inf ers, the ability to detect a given impact energy is dependent on system background noise level and on the size of the impacting object. Typical background acceleration noise level (g) ranges from .04g to lg peak to peak. During testing and calibration of B&W LPM Systems, acceleration response to a 7.5 ft-lb impact nas been measured from 3.0g for a 0.25 lb bael down to 0.Olg for a 30 lb. ball. Thus, the lower level 0. 5 f t-lb impact responses could be masked by the background noise and therefore be undetectable. In this case the specific in-plant conditions and the actual on-line sensitivity would have to be specified.

High alarm detection in this system is performed as follows.

Continuous background levels are detected and processed through a circuit with a time constant while allows for relatively slow changes in background levels. This, in ef fect will slide the alarm threshold up or down depending upon the '

e change in background level, based upon the alarm level ratio of the particular sensor, as discussed earlier. When a signal which changes in a relatively short period of time (<10 ms) is i detected this signal is compared to the background level and  ;

alarm ratio. If the signal meets the alarm criteria, a  ;

digital signal processing function is performed to validate the signal as a true impact event.

I When an impact occurs on the Reactor Coolant System, a

' ringing' of the RCS structure occurs. This ' rj e;ing' (due to i the natural frequency of the structure at snat location)  !

generates an impulse with an exponential decay that is analyzed by the LPMS. Certain criteria are investigated in ,

order to determine whether the impulse does exhibit exponential decay criteria (and is therefore most likely a valid event) or not. If identified as a valid event 10 ms of pro trigger and 50 ms of post trigger data are recorded to disc and an alarm event flag is set. Should another valid event occur with 1 second, on any channel, the plant alarm is then set. Spurious electrical noise is then effectively filtered out as that will not meet the exponential decay criteria.

Page 6

l k

This system therefore uses a single channel alarm criteria and subsequent Digital Signal Processing (DSP) for alarm validation, as opposed to previous systems, such as Unit 1 at CPSES adjacent. channel alarm criteria.

3.0 DhIA 3.1 Imonet Baseline sensitivities, The following plots were made using the B&W LPMS IV during pre- I operational test 2CP-PT-91-Ol. These plots were utilized to generate the impact response to 0.5 ft/lb impacts. noted in Table 3, presented previously.

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3.5 Additienal system settines Table 4 delineates additicnal settings and alarms of the LPMS IV based upon data extracted from Table 2 and correlations between impact and background spectra. These settings are described as '

follows:

FS Range Full scale range, sets the channel range to the highest g value expected. A normal setting would be approximately 5-8 times the background level. '

This ,1stting establishes the coarse gain for the channel LPSC.

Fixed Fixed high alarm level (g's).

Low Alarm The ratio of the 0-1KHz RMS value over the Ratio 1-10KHz P.MS value required to cause a low alarm.

Active y/n Set by the user to inform the system which channels are operational.

Low Alarm The channel percentage of full scale range 1 KHz to 10 KHz setting that will trigger a low alarm if exceeded (decreasing value) by the background noise.

In addition to these settings, the minimum plant G'r pk-pk should be set at 0.1. This value is compared to the ur. filtered levels of each active sensor and inhibits all alarms during plant shutdown.  ;

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TABLE 4 ADDITIONAL SYSTEM SETTINGS Filter FS Fixed i Low Alarm Active Low J SENSOR Bandwidth Range .High Alarm I (kHz) (G's) (G's) Ratio 0.125' 4 Y 5% f SGA IN1 1-14 0.5 4 N 5% l j SGA IN2 1-14 0.5 0.125 4 Y 5%

SGB IN1 1-14 0.5 0.175 0.17F 4 N 5%  ;

SGB IN2 1-14 0.5 i

0.125 4 Y 5% l SGC IN1 1-14 0.5 i 0.5 0.125 4 N St SGC IN2 1-14 0.40 4 Y 5%  ! ,

SGD IN1 1-14 1.5 0.25 4 N 5%

SGD IN2 1-14 1.0 0.5 0.50 4 N 5% f RCP 2-01 1-14 6 8.0 0.50 4 N St RCP 2-02 1-14 O.50 4 N 5% j RCP 2-03 1-14 1.5 4 N 5%

RCP 2-04 1-14 2.5 0.50 0.5 0.1 4 N 54 f SGA L/ TAP 1-14 -

0.1 4 N 5%

SGB L/ TAP 1-14 0.5 4 N St SGC L/ TAP 1-14 0.5 0.1 0.1 4 N 5%  ;

SGD L/ TAP 1-14 0.5

.f 4 Y 5%  !

RV ICT 017 1-14 0.5 0.25 il 0.25 4 Y 5%

RV IGT 018 1-14 0.5 4 Y I 54 f RV TOP 019 1-14 0.5 0.3 RV TOP 020 1-14 2.0 0.3 4 Y 5t lf l

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4.0 RESULTS AND RECOHMENDATIONS

. Table 5 delineates The Hi Alarm t Background settings for all sensors.

The following considerations were accounted for when setting these levels. >

1) 3 lb hammer impact response to background ratios were considered as this enveloped the 1/4 lb hammer responses in all cases (see Table 3 ratios). As expected, the 30 lb hammer responses generally were very low and would in most cases not be discernible from background noise.

2) Impact response to background ration at similar sensor locations.  ;

Known operating conditions, such as' the incore guide tube to thimble

~

3) rattle on sensor RV IGT 017 and 018, or the generally noisy sensor  ;

environment at the RV TOP locations. i Specifically, as an example, for RV IGT 017, a setting of 120% (from Table  ;

3, 3 lb Hammer) would cause excessive alarms due to the incore thimble to I guide tube rattle. This channel, as an active channel as well, would have +

a setting of 5.80%, that of RV IGT 018. While both channels exhibit the incore rattle, 017 alarms frequently even at its present alarm ratio of _!

500%. Sensor 018 however, at 500% also, does not. As both are in approximately the same location (Bottom head region of the RV) a setting ,

of 5.8 for both is felt to be appropriate to alleviate excessive alarms  !

due to guide tube rattle, while still maintain sensitivity in this region  ;

to detect loose parts.

For the RV TOP Sensors 019 and 020, chose sensors have noisy environments  :

and for sensor 020 this causes a low ratio for the 3 and 30 lb hammers. t A setting of 500% is chosen for this sensor as an interim setting of 400% l still indicates several alarms per shift on this channel due only to its noisy nature. Sensor 019's ratio will be utilized from the 3 lb hammar .

response _ and will still provide sensitivity to loose parts in this region. {

For the steam generator inlet plenums, high flow velocities in these areas p will consequently generate high impact energies. Therefore a minimum of  !

a 300% alarm ratio for SGD IN1 (which has an indicated ratio of 280% for i 3 lb hammer hits, Table 3) and the actual ratios as per Table 3 for the

• three remaining inlet plenums is utilized.

These are the active sensors of the LPMS. Other sensors ratios are [

determined along the same lines although the remaining sensors are usually  ;

to be used for diagnostic purposes but can become part of the active  ;

sensor group should failures occur with the normal active sensor lineup.

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l TABLE 5 1

POWER OPERATIONS ALARM SETTINGS HI-ALARM % BACKGFOUND l

SGA IN1 360 )

i SGA IN2 980 SGB IN1 560 SGB IN2 460 SGC IN1 310 SGC IN2 480  :

SGD IN1 300 SGD IN2 300 RCP 201 950 RCP 202 300 RCP 203 300 RCP 204 320 SGA L/ TAP 1700 SGB L/ TAP . 2000 SGC L/ TAP 1800 SGD L/ TAP 1700 RV IGT 017 580 RV IGT 018 580 RV TOP 019 1300 RV TOP 020 500 s

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