ML20137T603

Startup Rept for Cycle 1

ML20137T603
Person / Time
Site:	Wolf Creek
Issue date:	11/27/1985
From:	Koester G KANSAS GAS & ELECTRIC CO.
To:	Martin R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
References
KMLNRC-85-259, NUDOCS 8512090065
Download: ML20137T603 (175)

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KAEAS GAS & EIECTRIC COMPANY NOLP CREEK GEERATING STATION STARPUP REPORP FOR CYCIE 1 NOVDIBER 27, 1985 ,

Docket No. STN 50-482 License No. NPF ,

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-8512090065 ADOCK O e5 h482 pg PDR P

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<h; TABLE OF CONTENTS PAGE NO.

. Table of Contents l i

' List of Tables iv List of Figures vi Introduction 1 1.0 Initial Core Loading 1.0-1 2.0 -Post Core Loading Precritical Testing 2.0-1 2.1 Control Rod Testing 2.0-3

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2.1.1 Cold and Hot Rod Control System Testing 2.0-5 2.1.2 - Rod Control System Test 2.0-10 2.2 Incore Movable De W Ler Systen 2.0-11 2.3 Pressurizer Continuous Spray Flow Setting 2.0-14 and Pressuriser Heater and Spray napability Tests 2.4 Reactor Coolant System Flow Measurement 2.0-18 2.5 Reactor Coolant Systen Flow Coastdown Test 2.0-20 2.6 RfD Bypass Flow Measurement 2.0-24 2.7 Preliminary Data Collection for Instrument 2.0-26 Calibration 2.8 Nuclear Instrunentation Syste 2.0-27 2.9 - RfD/IC Cross Calibration Tests 2.0-28

, 2.15 Thennocouple Core Subcooling Monitor 2.0-32

. Systen Test 2.11 Special Test Procedure For the Pressuriser 2.0-33 L Relief Valves i

2.12 Icose Parts Monitoring Systen 2.0-35 3.0 Initial criticality And Iow Power Test 3.0-1 Sequence l'

3.1 Initial Criticality 3.0-3 3.2 Control Rod Bank Worth Measurements 3.0-13

!_ 3.3 Isothermal Tesperature Coefficient 3.0-25 I:

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TABLE OF CONTENTS

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PAGE NO.

3.4 Boron Endpoint and Boron Worth Measurements 3.0-28 3.5 Pcuer Distribution Measurements 3.0-31 4.0 Power Ascension Testing 4.0-1 4.1 At Ibuer Physics Testing 4.0-2 4.1.1 Incore Movable Detector Mapping At Power 4.0-3 4.1.2 Axial Flux Difference Instrumentation 4.0-5 Calibration 4.1.3 Power Coefficient Determination 4.0-11 4.1.4 ?seudo Rod Ejection Test 4.0-13 4.2 Control Systen Dynamic Response 4.0-16

, 4.2.1 Cynamic Automatic Steam Dung Control 4.0-17 4.2.2 Automatic Reactor Control 4.0-19 4.2.3 Automatic Steam Generator Level Control 4.0-20 4.3 Transient and Trip Tests 4.0-22 4.3.1 Load Swing Tests '4.0-23 4.3.2 Large Load Reduction Tests 4.0-31 4.3.3 Shutdown Ard Maint.cnance Of Hot Standby External 4.0-34 To 'Ine Control Room 4.3.4 Rods Drop And Plant Trip 4.0-35 4.3.5- Plant Trip From 100 Percent Power 4.0-37 4.4 .Instrtamentation Calibration and Alignent 4.0-40 4.4.1 Thermal Power Measurement And Statepoint 4.0-41 Data Collection

, 4.4.2 Calibration Of Steam and Feedwater Flow 4.0-43 Instrumentation 4.4.3 Operational Alignnent of Nuclear Instrtimentation 4.0-44 4.4.4 Operational Aligrunent of Process Tenperature 4.0-53 Instrumentation x

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TABLE T CONTENTS PAT NO.

4.4.5 Startup Adjustments Of The Reactor Control 4.0-56 System 4.5 Steam Generator Moisture Carryover Measurement 4.0-58 4.6- NSSS Acceptance Test 4.0-60 4.7 Power Ascension ' thermal And Dynamic Test 4.0-62 4.8 Biological Shield' Testing 4.0-64 4.9 Plant Performance Test 4.0-65 4.19 Turbine Generator Tests- 4.0-66

~4.11 Special Tests 4.0-68 4.11.1 Moisture Separator Reheater Test 4.0-69 4.11.2 Reactor Vessel Level Instrumentation System 4.0-70 (RVLIS)

Appendix A: Chronology of The Post Fuel Load Startup Program A-1

. . Appendix B: Power Ascension Testing Synopsis B-1 Appendix C: -Unplanned Reactor- Trips. During Post Fuel Load C-1 Test Program 9

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4-LIST OF %BLES

~ TABLE TITLE PAT NO.

. 2.1.1-1 Rod Drop Time Sumary 2.0-7 2.4-1 . RCS Loop Flow Determination Prior To 2.0-19

. Initial Criticality 2.5-1 Flow Coastdown Rate Calculations 2.0-21 2.5-2 tow-Flow Reactor Trip Time Delay 2.0-23

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Calculations

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2. 6-l' RTD Bypass Flow Measurement Results 2.0-25

- 2.9-1 Fesults of Initial RTD/7C Cross 2.0-29 Calibration Test

- 2.9-2 Results of Second RTD/7C Cross 2.0-31 Calibration Test

. 2.11-1 Results of PORV Opening / Closing 2.0-34 Test

' 3.2-1 - Control Rod Bank Worth Stmmary 3.0-14 3.3-1 Isothermal Tenperature Coefficient 3.0-26 Results Stmmary 3.4-1  : Boron Endpoint Sumary 3.0-29

". 3. 4 -2

. Differential Boron Worth Sumary 3.0-30

-3.5-1 Fower Distribution Stamary 3.0-32

. 4.1.1-1 Incore Flux Map Stamary During Power 4.0-4 7.ccension 4.1.2-1 Incore/Excore Correction Factor 4.0-6 4.1.2-2 100% NIS Current Values 4.0-7 ,

,. 4.1.2-3' Gain Values for Delta I Function 4.0-9 Generator Delta q Valms L At Specified Power

.. ~ 4.1.2-4 4.0-10 Plateaus

. 4.1.3-1 Doppler Coefficient Verification Factors 4.0-12 iv m g y , , , , - + , g- m

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LIST OF TABLES TABLE TITLE PAGE NO.

4.1.4-l' D-12 Rod Worth From IEP RIL 4.0-14

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'4.1.4-2 Flux Map Results From Rod D-12 Ejection 4.0-15 4.3.1 Load Swirg Frm 30% to 20% Power 4.0-24 4.3.1-2 Load Swing From 20% to 30% Power. 4.0-25 4.3.1 Load Swirg Frm 75% to 65% Power 4.0-26 4.3.1-4 . Load Swing Frm 65% to 75% Power 4.0-27 4.3.1-5 Load Swing Fr m 100% to 90% Power 4.0-28 4.3.1-6' Load Swing Frm 90% to 100% Power 4.0-29

~4.3.2 Large Load Reduction Test Frm 75% 4.0-32

, Power 4.3.2-2 . Large Load Reduction 'Ibst Fra 100% 4.0-33 Power 4.3.4-1 Rods Drop.And Plant Trip Test Data 4.0-36 Sumary

.4.3.5-1 Plant Trip Frcm 100 Percent Power 4.0-39

. Test Sumary

. 4.4.1-1 ICS Flow Fra Calorimetric Measurment 4.0-42 4.4.3-1 Nuclear Instrmentation Overlap Data - 4.0-45 Source Range and Intermediate Range

,e 4.4.3-2 Nuclear Instrumentation Overlap Data - 4.0-46 Intermediate Range and Power Range 4.4.4-1 T eperature Alig nent Data at 100% Power 4.0-55

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4. 5-1. Steam Generator Moisture Carryover Test 4.0-59

1. Results V

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1 LIST OF FIGURES -

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FIGURE' ' TITLE PAGE NO.'

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L1.0-1. Core Loading Sequence - Legend 1.0-4 1.9 Core Meding Sequence Steps 9 to 7BL 1A5 _ ,

1.0-3: Core Loading Sequence StepsC7 to 7D 1.0-6 f1.5-4L  : Core Loading Sequence Steps 8'to 34 B 1*8-7 3 1.9 Core Loading _ Sequence Steps 35 to 55C .0-8 Core Loading Sequence Steps 55 to 56 B

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1.9-6 : 1*8-9 D

l'.'9-7 ! Core Loading Sequence Steps 57 to 86B 1.0-19 '

_1. 0 Core Loading Sequence Steps-87 to ll8 B 1. -11'.

l'. 9--9 ; Core Loading Sequence Steps 119 to 158 B 1.0-12 L... 1.9-19' Core Loading Sequence Steps 159 to 193 1.0-13 11.9-11 ' ICRR Plot For Core Loading Source Range'N31'~1.0-14

.1.0-12 ICRR Plot For Core Loading Source Range N32. 1.0 '

l.0-13 ICRR Plot For Core Loading Temporary.-- 1.0 -

Channel A c1.9-14 ICRR Plot For Core Mading Temporary 1.0-17 q Channel.B

.r 1.0-15 ICRR Plot For Core' Loading Tenporary 1.0-18 a Channel C-11.0-16 Wolf. Creek Generating Station 1.0-29 Cycle 1 Final Core Loading Map-

<2.1-l' Control Rod Locations 2.0-4 .!

2.0-12

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2.2.1. Movable Detector Locations-

2.3-1 Naninal

Pressure Response to Opening 2.0-16 of Both Pressurizer Spray Valves -l

-2.3-2 Pressure Response'to Actuation of 2.0-17 All Pressurizer lieaters

[ 3.1-1 ICRR During Rod Bank Withdrawal Channel N31 3.0-4 ,

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'" Q LIST-OF FICURES s

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FIGURE = TITLE s PAGE NO.

3.1-2 ICRR During. Ed Baiik Withdrawal Channel 3.0-5

.N32 5

.3.1-3 ICRR vs. JCS Boron Concentration 3.0-6 Channel,N31

^3.1-4 ICRR vs."ICS Boron Concentration 3.0-7 Channel N32 E

3.1-5 ICRR vs. '

Time of ICS Dilution Channel 3.0-8

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N31-

._ 3.1-6' ICRR vs. Time. of ICS Dilution Channel 3.0-9 N32 3.1-7 ICRR vs. Reactor Makeup Water Addition 3.0-10 Channel N31

," -3.1-8 ICRR vs. Reactor Makeup Water Addition 3.0-11

- Channel N32 3.2-1 Differential and Integral Bank Worth 3.0-15

- Plot (GD) 3.2-2 Differential and Integral Bank Worth 3.0-16 Plot (CBC) 3.2-3 ' Differential and Integral Bank Worth 3.0-17 Plot-(CBB) 3.2-4 Differential and_ Integral Bank Worth 3.0 ' Plot (CBA)

-3.2-51 Differential and Integral Bank Worth 3.0-19 Plot (SDE) 3.2-6 Differential and Integral Bank Worth 3.0-20 Ylot (SDD)

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3.2-7',3 3' -

,, , 'D iff6rential and. Integral Bank Worth 3.0-21

, ;. Plot- (SDC)

, 3.2-85 /; 'differentialandIntegralBankWorth 3.0-22

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Plot (E,iected Rod D-12) 3.2-9 -i Differential and Integral Bank Worth 3.0-23 Plot (SDB, F-10, Stuck Rod)

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LIST T FIGUPES

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t FIGURE TITLE PAGE-NO.

3.2-10 Differential and Integral Bank Worth 3.0-24

. Plot.(Overlap)-

3.3-l' Rod Withdrawal Limits 3.0-27 4.4.3-l' : Channel Current Vs. Reactor Power - 4.0-48 Channel N41-it -

4.4.3 Channel Current Vs. Reactor Power - 4.0-49 Channel N42.

4.4.3-3 Channel Current Vs. Reactor Power - 4.0-50 Channel N43 4.4.3-4 Channel' Current vs. Reactor Power - 4.0-51 Channel N44 e

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' INTRODUCTION This report presents.the results of' initial startup testing at the Wolf

' Creek Generating Station fr.om initial fuelcload until the completion of the

- Power' Ascension Test Program. The plant performed exceptionally well during the.startup phase allowing the entire post fuel load test program to be cxampleted .in 169 days; .

' Wolf Creek is .a Standardkzed Nuclear Unit Power Plant System (SNUPPS) unit . ,

located in Coffey County, Kansas. The Nuclear Steam Supply System (NSSS) 'is

- 's four loop, Westinghr>ose pressurized water reactor -(PWR) rated at 3411 megawatts- thermal (twr))(3425 IWP -including reactor coolant pmp '(ICP) heat). . General Electric provided the-turbine generator for Wolf Creek.

'Bechtel N wer Corporation was the architect for.the entire ~ power block.

Sargent & Lundy acted as architect-engineer- for. the' site-related portions of the project that were'not part of,the SNUPPS design. Daniel International Corporation. was the site constructor ~ for Wolf Creek.

1 License No. NPF-32 'was issued by 'the Nuclear Regulatory Conmission (NRC) on March 11, 1985, which autnorized Kansas Gas & Electric to proceed with

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initial fuel-loading and low power testing,: including initial criticality

.' Land . low. power physics tests, at power ' levels not :in excess of 5% rated -

thermal power -(RTP). The'first fuel assembly was inserted into the core on-March 12,1985, and fuel loading was' completed 'on' March L17,1985. : After the tinstallation of the. upper internals and the reactor vessel head, the.RCS was

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filled ~and' vented.' Cold plant testing was authorized on March 27, 1985, and

was gompleted on April 17, 1985. The RCS was at hot standby conditions (557 F, 2235 psig) on: April 30,,1985.. Post core loading precritical testing

- was c:xnpleted 'and approved 'on May 19,' 1985.

The reactor was taken critical on May 22,- 1985, andflow power physics l testing was conmenced.- Iow power physics testing was completed on June 3, 1935. On. June 4, 1985,Lthe NRC lifteduthe 5% power restriction and issued license number NPF-42 for full. power operation.

-L WolfiCreek entered Mode in (>5% power) for the first time on June 6,1985. ,

e 'The turbine generator was synchronized to the grid on June 13,.1985. The

-1000 plant. trip test, performed on August 29, 1985, was the final test in

. the Mwer Ascension Test Prograrc. Following a brief maintenance outage the unit was' declared 'consercial 'on September 3,1985, at 0114.

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F 1.9 INITIAL CORE IDADING We initial core loading test sequence consisted of activities required

. prior to and during the actual loading of fuel. Many of these tests were required .to be performed within certain time periods prior to the fuel load.

The Nuclear Instrumentation System (NIS) was tested well in advance of core load. Proper functioning of the system was verified including all alarm and trip mechanisms. Numerous deficiencies were encountered during the performance of this procedure, however, all were resolved prior to closing the procedure. Most of the deficiencies dealt with cmputer points not

- giving expected responses and were successfully resolved by having the computer group rebuild the points. Two cases of equipment failure were discovered during the test, one of which was a high voltage supply and the other a control board meter. Both pieces of equipment were replaced and successfully tested.

Following the functional check of the NIS, the source range preamplifier and

- pulse amplifier gains were adjusted for optimum settings. This was

. accomplished using a portable neutron source with a strength of approximately one curie. The neutron source was placed near the source range. detector housing and the high voltage bias adjusted to obtain data which was then plotted. By determining the point at which the curve deviated from a best fit straight line, optimum bias settings were chosen.

These settings were then verified by disabling the high voltage power supply bias supply and checking that the count rate settled out below 5 counts per second. Actual testing was performed using surveillance (STS) procedures which satisfied the same requirements as the startup procedures. No probles were experienced with those portions of the STS procedures that were related to the startup testing.

We remainder of the requirements of the core loading test sequence were restricted by time limits. Some of the limits were imposed by Technical Specifications, others by Westinghouse or plant administrative recomnendations.

Thc first such limit was that within 7 days prior to core load, temporary.

nuclear monitoring instrumentation should be setup and checked ard that valve lineups be performed in preparation'for chemistry sampling. We tenporary instrumentation package, which consisted of 3 neutron detectors and all associated equipment, was supplied by Westinghouse for use during core load. The equipment was setup on the refueling deck inside the containment building and settings adjusted in accordance with a Westinghouse pre-shipment calibration. A neutron source was placed near the detectors and the high voltage varied to obtain data for voltage plateaus. We results of the plateaus proved to be consistent with the pre-shipment calibration indicating no changes in settings were necessary. The high voltage operating setpoint to be used was determined to be 2100 volts on all three detectors. Due to the time requirements, this test was repeated three times because the operating license was not received as anticipated.

I 1.0-1

.As requ' ired by Plant Technical Specifications, containment ventilation, contai ment penetrations, and the refueling machine were proven operable within--199 hours of core load. Again, these procedures were repeated as the

-license was delayed.

Verification of valve lineups and boron concentration sampling comenced 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> prior to the anticipated core load time and was repeated several times due to delays in license receipt. Sample results indicated that the boron

. injection tank was out of specification high and the boric acid storage tank was out of specification low. . By recirculating the tanks, both were brought

=within specification. A saple from the excess letdown heat exchanger line could not be obtained since the reactor coolant systen fill elevation was below that of the~ sample point. .This line was then isolated and valves- ,

tagged shut to prevent any possible mixing of the two systens.

.-After being placed in the reactor vessel and within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> of the start of

, core load, the temporary detectors were response checked. This verified that'the count-rate would increase when exposed to a neutron source and was-accamplished by lowering a source into the vessel and placing it next to the

-detectors. The initial. position.of the detectors within the reactor vessel is;shown on Figure 1.0-2. Also during this 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> period, an analog channel operational. test was performed on.the NIS sourca range channels.

. On March 11,-1985,. operating license NPF-32 was issued to Wolf Creek.

. Prerequisites to the core loading procedure were cmpleted and a briefing was held with all involved personnel to confirm responsibilities.

Background count rates'were determined for all detectors, temporary and permanent. ICRR. monitoring was performed concurrently throughout the entire core load. As expected, very low background count rates were obtained which were on the order of 0.02 to 0.05 counts per'second.

As reference counts were taken, temporary detector A did not respond ,

reliably. Since tmporary detector B was'not considered to be a responding '

detector until late in the core loading sequence, it was switched with detector A thus making detector B the inoperable detector. . Detector B was never declared operable but there were at least two available responding detectors at all times.

After verifying that all requirements for core load were completed and' that the refueling equipment was operating properly, the Plant Manager's permission to begin core load was obtained. At 0747 on March 12, 1985,-the first fuel assembly, assembly C64, was removed from the spent fuel storage pool and placed into the upender. The first problem was encountered when I the transfer cart would not traverse all the way to the containment-

. . , -building. Assembly C04 was returned to the spent fuel pool in order to

' troubleshoot'the transfer cart. It was found that the energency pullout

. ~ cable was tangled.in'the cart device mechanism. A scuba diver was sent down

. .to remove the cable. Instead of replacing the cable,.it was decided to use the cart without the cable since this was a new core and with no high radiation present, a ' diver could, if necessary, go down to attach a new cable at a later-time. At 1216, after.a 4.5 hour5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> delay, assembly C04 was again removed from storage and this time.successfully loaded into core position L-15.

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As assably C04 was moved over the reactor vessel area, the high flux at C shutdown alarm sounded. The alarm was then. blocked as this was not unexpected. Assembly C04, as well as the second assably loaded, assembly.

C30, contained the primary sources. The sources within these assemblies were Californium - 252 .(Cf-252) which ware previously installed on January 12, 1985. - Since this was a new core, the water was only at the level of the nozzles within the reactor vessel and just above the transfer tube within the transfer canal. Therefore, there was no -moderator between the fuel assembly and the source range detector housing which allowed an increased number of neutrons to reach the detector. The high flux at shutdown alarm remained blocked until both assemblies were installed in the core.

-Core loading operations were suspended _after the first fuel assably was-unlatched in the core to investigate a problem.with the spent fuel bridge-crane.' The hoist of the crane was " chattering" as the fuel asse blies were being lifted. Since core loading was performed in a semi-dry condition, the lack of bouyancy presented an increased weight to be lifted by the hoist.

The original setpoints for the tx>ist were for wet conditions and were thus adjusted to compensate for the present dry condition. After doing'so, the crane was functionally checked and upon successful completion, core' loading resumed. Following this 3 hour3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> delay, the second assembly was removed from storage at 1602.

Since the first two assemblies loaded were source bearing assablies, 'the count rate significantly increased.as expected. The count rate-from source range channel .N31 increased to 6.24 counts.per second and N32 to 8.75 counts per second. Before proceeding with core loading, the high flux at shutdown setpoints were ' adjusted to five times these values. Core-loading then

- continued as illustrated in Figures 1.0-1 through 1.0-10.

As each fuel assembly was being inserted into the reactor vessel, outputs of responding temporary detectors were monitored _on a strip chart recorder at

. the core. loading station inside the containment building. Permanent plant 2 . instrumentation was monitored in the control- room. After verifying the core

~ was not approaching criticality, the fuel assemblies were unlatched from the

. manipulator crane. Count rate data from four operable detectors' (2

.tenporary, 2 permanent) were obtained by averaging the results of three -

i

- counting periods. This data was thel used to plot an inverse count rate

[ ratio (ICRR) curve.

The source bearing assemblies were loaded into the mre first, however, effective reactivity monitoring muld not be achieved until the sources were.

, moved to their~ final position and a cluster of assablies built around L then. Therefore, ICRR monitoring was not initiated until after step 19 of the loading sequence. Data was then obtained throughout the loading sequence and ICRR plots updated following each ste of the sequence. These

. plots are -included in Figures 1.0-11 through 1.0-15.

I' At no time was core loading interrupted due to high count rates or unexpected changes in the ICRR. However, there was some concern following

-step 87. After this assembly (assembly A04) was loaded, only on'e of the respondin) detectors was indicating greater than 2 counts per second. The Final' Safety Analysis Report requires at least two detectors indicate this count rate. This did not present an inmediate concern since another detector that was not identified as a responding detector did indicate this 1.0-3

v FIGURE 1.0-1 CORE IDADING ' SEQUENCE LEGEND e

Legend for Core Loading Figures Assembly loaded in permanent position in previous step.

] Assembly-loaded in tenporary position in previous step.

Assenbly' loaded into position during loading step Nurtber N.

' Location of Tenporary Detector A (B and C) .

• Assenbly with primary source insert.

. Not as yet loaded.

Note: Arrows indicate detector or fuel movenent.

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4 value thus satisfying the FSAR requirement. However,'the concern was that

. toward the end of the loading sequence, the tenporary detectors have to be removed in order to place the final fuel assemblies in the core. This would

-only leave the two NIS detectors which were not indicating the required 2 counts per second. The situation was evaluated and the requirenent of 2 counts per second was changed to 0.5 counts per second following a

. 10CFR50.59 evaluation.

Boron samples of the' reactor coolant system taken throughout the core load resultedinanaveragevalueof2104ppmwitharangeof20g8ppmto2120 ppg. Residyl Heat Renoval Systen temperatures averaged 95 F, ranging from 83 F to 101 F.

With the exception of the transfer cart and the spent fuel bridge crane hoist, no significant equipment problens occurred during the core loading

' operation. A few instances of minor difficulties were encountered with the tenporary monitoring equipment all of which were resolved in short periods of time.

Core loading was successfully completed at 0600 on March 17, 1985, as the last assembly was unlatched in core position L-15. The total elapsed time

' for loading the complete core of 193 fuel assemblies was 118.25 hours2.893519e-4 days <br />0.00694 hours <br />4.133598e-5 weeks <br />9.5125e-6 months <br />.

Including delay time the average assembly loading rate was 1.63 assemblies per hour. Core loading operations continued 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> a day using two 12

' hour shifts.

Innediately following completion of the core load, a video map of the core was taken to verify proper fuel assenbly positioning and that the proper

.. insert was contained in each fuel assembly. An underwater television camera and video cassette recorder were used for this verification. The video tape was than reviewed for double verification. The results'of the map are illustrated in Figure 1.0-16. Upon canpletion of the map review, the initial core loading operation was complete and accomplished in an exanplary

. manner.

1.0-19

m

FIGURE 1.0-16 WOLF CREEK GENERATION STATION - CYCLE 1 FINAL CORE LOADING MAP R :P> N. M L K J ~H G F E D C B A

/C27 C43 C60 C12 Col C50 GZ6 116DS B10P. 4305 B10P. 405 810P- 128DS 1 on 1t0 t?n C36 C32 C54 A38 C01 A22 - C04 A16 C28 CO2 C63 ASP 1D 2

89F60 R50 12P2D R17 20P90 R43 19PS1 R51 12P50 R34 C56 - CSI B46 A30 803 A40 B45 A42 819 A33 847 CIS C17 B9P50 92D5 20P25CR28 f 6P120 R14 16P17C R47 16P3C R13 20P24[ 7705 A9P40 3-

~ ~

C47 B28 B24 848 A43 B34 A52 B43 A01 B49 B59 839 C05 4 R26 20P2D R22 20P11[ 0551 16P28[ RS2 '6P130 3805 20P33C R35 20P8D R04 C06 C42 A49 B16 A51. B20 A32 861 A23 B09 A14 B22 A28 C25 C45 5 s 48DS 12P40 R12 20P22I 23D5 16P2D 25D5 16P18[ 46DS 16P310 7505 20P50 R02 12P10 79DS C38 A06 837 A05 B62 A41' B10 A26 831 ' A10 B63 A46 B14 A62 C34 6 A10 PAN R05 16P320 53DS 16P16I R29 20P190 R20 20P3 R31 :6P230 6605 16P80 R48 A10P2I C41 C03 A13 812 A53 BIS A15 B07 A03 B13 A63 606 A61 C55 C40 7

83DS 20P27[ R16 - 16P40 3305 20P32C 905 20P200 SDS 20P70 9705 16P11C R19 20P15E 2405 s

C22 A37' B18 A36 B38 A60 811' A29' B53 A35' B32 A21 857 A24 C13 8

A10P11 R53 16P300 R40 !6P260 R08 ?0P260 R39 - 20P170 R41 i6P100 ROI 16P200 R11 410P3D C20 C61 A58 B50 A64 B33 A47 817 A44 B02 A56 B64 A48 C11 C24 9

54DS 20P21t R30 16P22C 3405 20P60 4005 20P40 450S 20P23E 47D5 16P70 R36 20P18C 3905 C09 - A02 B56 A17 B35- A18 823 A11 801 A55 826 A04 805 - A39 C39 10 410P60 R25 16P250 8805 6P27C RIO ?OP34D R15 20P13[ R23 16P210 5605 16P29C R37 T10PSD

' -(i C53 COS A25 804 A50 842 A27 B58 A20 B52 A59 B40 A65 C64 C49 II

-C 12505 12P80 R09 20P310t1205 16P190 1705 16P50 4405 16P15[ 134D5?OP140 R49 12P60 1405 C19 - 854 836 851- A31 830 A12 860 A07 821 'B27 855 C18 12 R18 20P100 RS4 20P28[ 73DS 16P240 R07 16P60 OSS2 20P12C R33 20P16C R32 J

C35 C59 829 A08 B41 A57 B44 A45 B25 AS4 808 iC33 C46 R45 16PIC R27 I3 ~

A9P2C SSDS 20P30: RSS 16P90 '6P14D R06 20P290127DS 89PSD C16 C23 C21 A09 C31 A34 C30 A19 C62 C57 C48 ,

L~ I" A9P3[ R44 12P7D R38 20P10 R46' 19PS2 R03 12P3D R42 39P70 C44 Cid C52 gg p C37 C58 C29

,nn , 15 9805 310P70 3005 9005 110P80 6505 O

Top - Fuel Ass mbly ID Bottom - Insert ID

.,- 'XXDS - Thimble plug RXX.- Control. Rod Assembly WPXXD - Burnable poison with W poison rods (symetrical)

( XWPXD - Burnable poison with W poison rods (non-symetrical)-

XXPSX - Primary Source Ass mbly XSSX . Secondary Source Ass m bly

(

1.0-20

. . , = ... .

2.9 POST CORE IDADING PRECRITICAL TESTING After completion of initial core loading, preparations were begun to perform the post core loading precritical testing phase of the Startup Test Program. This test phase performed the final testing and alignment of various plant systes prior to initial criticality. This inv lved testing o

at cold shutdown conditions (RCS average temperature <200 F), testing during plant heatup and pressurizatign and testing at hot no! oad conditions (RCS 1

average t aperature 557 +0,-5 F, RCS pressure 2235 + 25 psig).

The upper internals were installed in the reactor vessel on March 17, 1985, and the reactor vessel head was set on March 18, 1985. With the completion of the tensioning of the reactor vessel head studs on March 21, 1985, the plant entered Mode 5 (Cold Shutdown). The fill and vent of the Reactor Coolant Systs was completed on March 29, 1985. During the Cold Shutdown period the following testing was performed:

1) - CRDM polarity checks,

2) Cold, no flow rod control system,

. 3) Cold, full flow rod control systen,

4) Initial incore novable detector tests,

5) Thermal and dynamic testing of the main steam and feedwater systens (initial data collection),

6) Initial testing of the reactor vessel level instrumentation systen (RVLIS).

Cold shutdown testing was completed on April 14, 1985, and the plant entered Mode 4 (Hot Shutdown) on April 17, 1985.

During the heatup phase, the following testing took place:

1) RCS RTD/IC cross calibration,

2) Continuation of the thermal and dynamic testing of the main steam and feedwater systens,

3) Special test of the pressurizer power operated relief valves (PORV) at operating RCS pressure (2235 psig), ,

4) RVLIS data collection.

The plant entered Mode 3 (Hot Standby) on April 26, 1985. Testing at 4500g RCS average tm perature was completed on April 28, 1985. Loose parts noise was noted on Channel 2 of the Loose Parts Monitor. Investigatignofthe noise continual during the heatup and subsequent testing at 557 F. The noise was determined to be caused by a vibrating thimble (tube 42) in the incore monitoring systen. This was evaluated as a minor problen and did not impact later testing.

2.0-1

?

0 The ICS was at 557 F, 2235 psig on April 30, 1985. Testing performed at

- this plateau included-

1) Setting of continuous spray flow valves,

2) Verification of pressurizer spray and heater effectiveness,

3) Verification of ICS flow rate, ,

4) Determination of transport time in the RCS RTD bypass loops,

5) Hot, full flow control rod system,

6) Hot, no flow control rod systen,

7) Checkout of the incore e vable detector system,

8) Measurenent of the reactor coolant loop flow coastcbwn time,

9) Precritical alignment of the Tavg and delta T instrumentation,

10) Precritical alignment of the Nuclear Instrumentation Systen (NIS),

- 11) Initial data collection at int no-load temperature and pressure for l

startup adjustments of the reactor control system.,

12) Checkout of the loose parts mnitoring system,

13) Checkout of the thermocouple core monitor system, 114) Background data collection for the biological shield testing, The final test of the rod control systen prior to initial criticality was completed on May 18, 1985. All precritical testing was completed and the Plant Safety Review Committee approved the test packages on May 19, 1985.

~

The following pages of this section contain detailed discussions of the post

- core loading precritical test program. Those ongoing procedures that were completed later in the Startup Test Program are discussed in Section 4 of this Startup Report:

1) Power Ascension Thermal and Dynamic Test - Section 4.7, ,

2) Biological- Shielding Testing - Section 4.8,

3) RVLIS - Section 4.11.2.

[. .

2.0-2

2.1 00!FrROL ROD TESTING i

A major cortion' of.-the tecting performed during the post core loading phase of the test program was involved with the various components of the rod control'syste. . The fall length control rod syst s consists of 53 rod cluster control assemblies -(ROCA) with each assembly consisting of individual rods constructed of hafnium encapsulated in cold worked stainless steel tubing. The assemblies are divided into 5 shutdown banks (SBA, SBB, SBC, SBD and SBE) and 4 control banks (CBA, CBB, CBC, CBD). Each bank,

. except shutdown' banks SBC, SBD and SBE, is divided into two groups.

Shutdown banks SBC,.SBD and SBE have only one group. Each group consists of 4 rods each except control banks CBA (2 rods in each group) and CBD (2 rods in group 1 and 3 rods in group 2). The rod banks are located in the core as

.shown in Figure 2.1-1.

The full length rods are moved by a Westinghouse magnetic jack type drive mechanism. Each control rod drive mechanism (CRDM) ~ contains three magnetic

. -induction coils which energize in a cyclic sequence to move the rods. Loss

-of power to these coils causes the rods to drop to the fully inserted position. The rod control system is designed to allow individual movenent of the various banks or movement of the control banks in the overlap mode.

l In the overlap mode, one control bank is withdrawn until it reaches a predeterminal setpoint where the next control bank in the sequence begins to move in synch'ronization with the first bank. During control bank withdrawal the sequence is CBA - CBD - CBC -.CBD. Control bank insertion.in overlap

- reverses the withdrawal sequence: CBD - CBC - CBB - CBA. During the stepping.of a bank, each rod group alternates motion to provide a more uniform reactivity change. In addition to rod movement by bank, an individual rod can be moved by use of the lift disconnect panelg This allows an out of position rod to be restored to the bank position.

e e

2.0-3

t

> FIGURE 2.1-2 ,

. CONTROL, ROD LOCATIONS

R P NM L K J H G F ED C B A - a s-1 2 4 @ @ @ $

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NO. CLUSTERS A 8 hSHUTDOWNBANK !D 4 E 4 A 4 B 8 Q CONTROL BANK C 8 D 5

. 2.0-4

2.1.1 COLD AIO HOP ROD COtfrROL SYSTD1 TESTING During the post core loading test sequence, major tests of the rod control systen were performed under four different sets of RCS conditions:

0

1) Cold, no flow (<200 F, >350 psig) 0

2) Cold, full flow (<200 F, >350 psig) 0

3) Hot, full flow (>551 F, 2235125 psig) 0

4) Hot, no flow (>500 F, 2235125 psig)

The cold, no flow and hot, full flow tests performed checks of CRDM operation and of rod position indication, verified rod movement over the stire range of travel, verified proper slave cycler timing (cold, no flow only) and determined the rod drop time for each individual rod. The effectiveness of the thinble dashpot region for decelerating the control rod was verified during each rod drop. The cold, full flow and hot, no flow tests measured rod drop times and verified thimble dashpot effectiveness only.

The initial testing performed was the slave cycler timing. One rod in a power cabinet was withdrawn to 54 steps and coil current traces (lift coil, m vable gripper coil and stationary gripper coll) were taken during a six step withdrawal and a six step insertion sequence. The rod was then inserted to rod bottom and the process was repeated until one rod in each power cabinet _ (five) had been tested. Each slave cycler operated satisfactorily.

The remaining testing was done by rod bank. For the cold, full flow and hot, no flow testing, only those steps necessary to measure the rod drop times were performed.

Each rod bank was withdrawn to 54 steps while taking rod position indication data at intermediate points. At 54 steps, each rod was individually withdrawn 6 steps and inserted 6 steps while recording coil currents. An additional 4 step withdrawal, 4 step insertion trace was taken for Westinghouse evaluation. These traces were used to determine that CRDM operation was satisfactory.

When coil current traces had been collected for each rod in the test bank, the bank was withdrawn to 228 steps while taking rod position data at intermediate points. Test equipment was then setup to monitor stationary gripper coil current and digital rod position indication (DRPI) coil output. Each rod was dropped individually by pulling the mvable gripper coil fuse and thm pulling a stationary gripper coil fuse while taking a visicorder trace. These traces were used to determine the drop time for each rod arxl the effectiveness of the thimble dashpot region to decelerate 2.0-5

4 the dropped rod.

After all rods in the test bank had been dropped, the bank was withdrawn to 228 steps ard then inserted until the rod bottm LED's energized to '

demonstrate satisfactory operation over the eitire range of travel (withdrawal and insertion).

This procedure was then repeated for each Rod Bank (control and shutdown) until all banks had been tests 3 under the applicable test conditions. After all the rods had been dropped, the standard deviation (sigma) was determined and any rod having a drop time outside a + two-sigma limit was droppal six additional times to ensure performance reliability during subsequent operation.

The following test results were obtained from this series of tests:

1) The slave cycler for each rod control syste power cabinet functioned satisfactorily during control rod withdrawal and insertion operations,

2) CRDM operability was dmonstrated at cold, no flow and hot, full flow 1 corations. All CRDM's operated satisfactorily,

3) The performance of the DRPI systs was dmonstrated at cold, no flow

- end hot, full flow conditions. The DRPI systs and related rod position indications satisfied all acceptance criteria,

~ - , -

M ) All rods operated satisfactorily when withdrawn and inserted over their entire range of travel at both cold, no flow and hot, full flow

, conditions, -

5)' The rod drop times for each individual full-length shutdown and control rod were less than 2.2~ seconds under all test conditions: cold, no flok; cold, full flow; hot, full flow; bot, no flow. The rod drop times are sumarized in Table 2.1.1-1,

6) Daring cold, full flow rod drop testing, rod D-2 (SBA) was outside the upper two-sigma limD:,'(1.47 seconds) with a drop time of 1.48 seconds.

Six additional rod drdps were performed with a minimum time of 1.48 ;

seconds and a maximum time of 1.51 seconds. This range of 0.030 scconds was greater then the 0.020 seconds allowable. After engiceering evalu ation, the drop tims for D-2 were determined to be accept 4Dle. The (i-2 drop time was within the two-sigma limit for the r eaining two tests at hot plant conditions.

~

. During hot, full flow rod drop testing rod B-10 (CBB) was outside the lower two sigma limit (1.33 seconds) with a drop time of 1.32 seconds.

, Rod B-10 was dropped six additional times with drop times of 1.33 to

. 1.34 seconds,

7) The rod drop traces were consistent with no signs of binding or other abnormaliti d under all test conditions. The rod deceleration through the thimble dashpot region was similiar for all rods at each set of test conditions. The thimble dachpot region was effective for decelerating the control rod during each rod drop.

./

2.0-6 s

P 6

TABLE 2.1.1-1 ROD DROP TIME SUP91ARY Rod Drop Time to Dashpot. Entry ROD' CORE - Cold, No Flow Cold,' Full Flow Hot, Full' Flow Hot, No Flow BANK Coord.

H-6 1.22 1.44 1.36 1.18 CBA H-10' .l.22 1.45 1.37 1.19--

F-8 1.20 1.43 1.35 1.18-K-8 1.21 1.46 1.33 1.19  !

F-2 1.21 1.44 1.36 1.29 B-10 1.22 1.40 1.32* 1. 29

-K-14 1.21 1.44 1.35 1.28 CBB P-6 1.21 1.43 1.36 1.21 B 1.19 1.42 1.36 1.21 F-14 1.20 1.41 1.35 '1.28 P-10 1.20 1.41 1.36 1.21 K-2 1.22 1.44 1.37 1.22 H-2 1.21 1.43 1.37 1.21 B-8 1.22 1.44 1.35 1.29 H-14 1.20 1.45 1.34 1.29 CBC P-8 1.21 1.41 1.34 1.21

+ F-6 1.21 1.42 1.37 1.21 F-10 1.21 1.44 1.37 1.29 K 1.22 1.43 1.36- 1.19 K-6 1.22 1.44 1.38 1.22' D-4 1.22 1.44. 1.35 1.21 M-12 1.20 1.43 1.35 1.21 CBD D-12 1.21 1.41 1.34 1.25 .

M-4 1.21 1.43 1.35 1.23 i H-8 1.20 1.45 1.37 1.19 D-2 1.19 1.48* 1.36 -1.21-SBA B-12 1.18 1.40 1.35 1.19 M-14 1.20 1.44 1.36 1.29 P-4 1.21 1.42 1.39 1.21 B-4 . 1.19 1.44 1.33 .l.281 D-14 1.19 1.45 1.38 1.21 P-12 1.18 1.41 1.36 1.22 M-2 1.20 1.46 1.39 1.22 2.0-7

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

.~; -

7 TABLE 2.1.1-1 (Cont) '

> ROD DROP TIME SIM ERY.

Rod D' rop Time to Dast E nt Entry t

RCD CORE Cold, No Flow Cold, Full Flow Hot, Full Flow- Hot, No Floti BANK Coord.

~G-3 1.21 - 1.44 1.36 1.22

.C-9 1.22 1.43 1.34 1.29 J-13 'l.29 1.44 1.36- 1.29  !

SBB- N-7 -1.2d 1.43 1.36 1.22 C 1.29 1.49 1.34 1.19

. G-13~ 'l.29- 1.43 1.36 1.19

L g .N-9 1.19 1.41 1.36 1.29 J-3 1.20 1.44 1.38 1.23

. E-3 1.19' -1.43 1.34 1.22 SBC- 'C 1.29- 1.41 1.33 1.25 l.19 1.21

~'

L-13 1.43 1.34

N-5 >1.29 1.43 1.34 1.23

.. C-5 1.29 1.42 1.36' 1.22

"' 1.29 SBD. .E .1.19 1.42 1.36 1.21

~

N-11 1.19- 1.49 -1.34 L-3 1.21 1.44 1.39 1.23 H-4 1.21 1.42 1.35 .1. 21 -

SBE- .D-8 l'.19 1.49 ' l.36 1.22 1.29

~

-H-12 1.21 1.46 1.36 M-8 1.22 1.42 1.35- 1.23

- Aver-age 1.29 1.43 -1.36 1.21

• Redropped six times F

3, i s

_. 2. 0-8 '

4

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7-_

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The following problems were encountered during the performance of these test

' procedures-

1) During withdrawal of CBD, DRPI indication for Rod K14 was lost.

Detector encoder card A406 in DRPI Data Cabinet A was found to be defective and was replaced (cold, no flow test),

2) During withdrawal of CBD, computer indication of rod position reained at zero. A power supply in . cabinet RJ048 was found powered down for

< investigation of 'an unrelated proble causing the zero _ indication on the computer. The_ computer data was collected after the power supply was energized (cold, no flow test),

3) During withdrawal of SBC, deand position as displayed by the computer indicated zero when all .other indications were -18 steps withdrawn.

During investigation,. the group step _ counter for SBC would not increent/decreent during rod mvment. The following itms were mrrected during the investigation:

a) The Kl9. relay in the Logic Cabinet was found to be sticking and was replaced, b) The slave cycler card for shutdown banks C, D and E -(Power Cabinet

' SCDE) had failed and was replaced, c) The A & B DRPI coil cables for rod Cll were interchanged at the Reactor Vessel head connections. These cables were reconnected in their correct locations, d) A cable was found terminated in RJ049 instead of' RJ046. Since this termination was correct per the schee drawings, a Teporary

?bdification was used to correct the termination. The Tenporary Modification was later made permanent.

1 The required rod position in31 cation was then collected when SBC was withdrawn for rod drop testing (cold, no flow test).

e

..D 2.0-9 a

4 2.1.2 ROD CONTROL SYSTEM TEST 1

s Prior to initial criticality, the full length rod control system was tested to verify satisfactory performance of the required control and indication functions and to verify that a manual rod block prevents manual withdrawal of the full length control rods. All the acceptance criteria of the test were met.

This final operational check was performed by rod bank. Each rod bank was withdrawn to either 18 steps (shutdown banks) or 48 steps (control banks) while monitoring proper operation of the group step counters, individual rod position indication (DRPI), rod speed indication and the console indicating lights. A set.of rod position indication dats (DRPI) was then taken. Each group in the withdrawn bank was individually put on the DC hold bus and the fusible disconnects for the appropriate power cabinet were opened. After verifying that rod position did not change, the fusible disconnects were closed and the DC hold switch for the rod group under test was returned to the'off position. The bank was then inserted to the zero step condition (rod bottom). This process was. repeated until all rod groups had been checked on the DC hold bus. The DC hold bus performed satisfactorily for all rod groups.- No problens were noted with Rod Position Indication or any

. other control or indication function.

' Bank overlap was checked using overlap setpoints that' allowed the control banks (CBA, CBB, CBC, CBD) to be withdrawn in MANUAL to approximately 30 steps. 'After the overlap and rod position data had been collected, the control rods were inserted in MANUAL to verify that the bank overlap function worked satisfactorily. The bank overlap system functioned to start and stop the control banks (CBA, CBB, CBC, CBD) at the correct setpoints.

A manual rod withdrawal block was simulated by lifting the lead of cable SSFS10AH at TP-PP-16 in RP040.- An attenpt was then made to withdraw Control Bank A in MANUAL. Control Bank A did not withdraw. The lead was then relanded and the ability to withdraw Cor.rrol Bank A was demonstrated.

s 2.0-10

2.2 INCORE MNABLE DETICIOR SYSTEM 1

The Incore Movable Detector Systen or Flux Mapping Systen is designed to provide a three-dimensional reactor core power profile through the use of movable fission chambers (neutron detectors) being moved axially in radial positions throughout the core. By placing the detectors into selected positions within the core, detector currents are provided and stored. This data is then processed yielding necessary information for reactor core surveill'ance.

The Flux Mapping Systen consists of two major itens: a flux mapping console and a detector drive systs. The detector drive systen consists of four (4) trains each providing a mechanical means of routing a detector into any one of 58 guide thimbles in the reactor core. The guide thimble positions are shown in Figure 2.2-1. Each detector is routed through a 6-path transfer device, a 15-path transfer device, and the seal table before reaching the guide thimble. Via the 6-path transfer device, each detector is capable of accessing another train's 15-path transfer device. The Flux Mapping Console provides a means of remotely controlling the detector drive systen including all the drive units and transfer devices. Two identical sections within the flux mapping console provides backup capability should one of the sections

. fail. The detector currents, which are a measure of neutron intensity, aru therefore core power, is reported to a CRT, printer, floppy disk, and the Westinghouse P2530 plant process computer.

After core loading, the guide thimbles were inserted in the core and final installation of the detector drive systen was completed.

The operation of the system was verified in two parts:

1) The operability of the detector drive systen was denonstrated.

2) The operability of the integral Flux Mapping System using the ronotely stationed flux mapping console was demonstrated.

OPERABILITY OF DETECTOR DRIVE SYSTEM The checkout of the drives and transfer mechanisms was done using a manual controller operated locally, and four dumny detectors (detector shaped but without co-axial output cable and fission chamber). Each one of the four detectors was run into all thimble combinations for that drive and the safety mechanisms were checked with regard to transfer rotation with a

. detector inserted 'and winding of retracted detectors onto the takeup reel.

The path length of each thimble combination was measured to give an initial path limit value for subsequent checkout of the integral systen. Problens

. were encountered with operation of the portable controller due to printed circuit card failures, but repairs were made to the necessary canponents to enable its use. Spring tension on the detector takeup reels required adjustment for proper detector withdrawal operation and a flexible braided storage path thinble required replacement. With these problems corrected, checkout of the detector drive systen was successfully completed.

2.0-11

FIGURE 2.2-1 MOVABLE DETECTOR LOCATIONS R P N M L K J H G F E D C B A 180*

1 o a 2- o o a 3 o o o o 4 o o o 5 o o o o

8 o o o o o 7 o o o o 8 so- o o o o o o o o 27o*

9 o o o o -

10 o o o

.11 o o o o o 12 o o o 13 .o o o a 14 o o o o 15 o o

. a.

o Movanle Detector -(58) 2.0-12

OPERABILITY OF ' die FLUX MAPPING SYSTEM The checkout of the integral system was done from the flux mapping consoles inside the control room. The menu driven consoles consist of the input

- computer terminal, printer, floppy disc unit, and data link to NSSS computer. The two redundant consoles were tested for all modes of operation. Initial checkout of the system was accomplished using the four dumy detectors to verify the thimble path lengths for all detector path combinations. Following path length verification, the dumy detectors were replaced with the operating fission chambers. During the che:kout, numerous computer card failures were encountered which were corrected by card replacenent or card / chassis reseating. Several problens with detector drive notors not being de-energized when the detector position encoder indicated no detector motion were corrected by adjustment of controlling relays.

Problems with data link failure to the Westinghouse P2500 process computer were corrected by giving flux mapping a higher priority level on the computer system. Due to system failures, it was not possible to verify all four detectors could individually obtain a full flux map from all 58 thinbles. 11owever, it was shown that the systen could be used to insert any one of the four detectors in any of the 58 thimbles. Subsequent work on the flux mapping systen verified the integral flux mapping system functioned as designed.

, During flux mapping in the power ascension program, the top of fuel and path length limits were revised based on grid strap data depressions to give proper detector position / core height correlation. Initial flux maps were run using only two detectors for a full flux map due to detector / computer card failures, but the necessary data for flux map analysis was obtained.

Card contacts and failures continued to be a source of problens during subsequent flux maps but reseating and/or replacement of the questionable cards enabled use of all four detectors during the test program.

e 9

2.0-13

d 2.3 PRESSURIZER 00lfrINUOUS SPRAY FIAN SE'I'fING AE PRESSURIZER HEATER A m SPRAY CAPABILITY TESTS The reactor coolant systen pressurizer establishes primary plant pressure by maintaining a saturated liquid and vapor environment in the pressurizer at the desired pressure. Activation of the two spray valves in the spray lines from two RCS cold legs to the pressurizer and imersion heaters within the pressurizer acts to control saturation pressure, thereby controlling PCS pressure. The continuous spray bypass valves are in parallel with the power operated spray valves. These valves provide a small continuous spray flow

- to warm the pressurizer spray lines and nozzle in order to limit thermal ,

stresses when the spray valves actuate and to assure that the boron concentration in the pressurizer is not dissimilar fran that in the reactor coolant loops.

PRESSURIZER CONTINUOUS SPRAY FIDW SEdTING This test was performed to establish a setting for the pressurizer continuous spray throttle valves to obtain ,an optimum continuous spray flow and to establish the setpoint for the pres'su'rizer spray line low tenperature

-alarms. Each continuous spray throttle valve was opened in discrete

.' increments while nonitoring pressurizer spray line temperature. Spray line tenperature was allowed to reach equilibrium prior to opening the valve further. Equilibrium spray line temperature was plotted against valve turns open. The continuous spray throttle valves were set at the break poi.nt on the curve where further valve opening had a minor effect on equilibrium spray line tenperature. ,

Valve Setting Equilibrium Spray Line Tenperature 0

BB-V082 51/4 turns opai 535 F BB-V083 5 turns open 532 F The spray line low tenperature alarm bistables were set at 10 + 5 0F below

~

the equilibrium spray line temperatures.

Bistable Setpoint 0

'IB-451 521 F TB-452 521 F

~ A review of the FCS. chemistry data denonstrated that the new settings for the continuous spray throttle valves assured that the boron concentration in the pressurizer was not dissimilar from that in the RCS. The completion of

- this test satisfied an outstanding test discrepancy from the Preoperational Hot Functional Test Program.

PRESSURIZER HEATER AND SPRAY CAPABILITY TEST The purpose of this test was to determine the rate of pressure reduction

- caused by the opening of both pressurizer spray valves and the rate of pressure increase caused by the operation of all the pressurizer heaters.

2.0-14

~While at hot, no-load conditions (5570 F, 2235 psig), the pressurizer spray

. valves were opened. fully. Pressurizer pressure, pressurizer level, pressurizer water temperature and spray line temperatures were recorded on a strip chart recorder. The transient was stopped when pressurizer pressure had fallen to.approximately 2000 psig g shutting the pressurizer spray valves and the RCS was returned to 557 F, 2235 psig. The results of this transient are shown as Figure 2.3-1.

While monitoring the same parameters as in the spraydown test, the pressurizer heaters were tested by energizing both banks of backup heaters and injecting a full demand signal to the control heaters. When pressurizer-pressure reached.23g0 psig, the pressurizer heaters were secured and the BCS was returned to 557 F, 2235 psig. The results of this transient are shown

. as Figure 2.3-2.

The pressurizer response to the opening of the pressurizer spray valves PCV455B and PCV455C was within the allowable range as shown on Figure 2.3-

1. However, an engineering evaluation of the data determined that both spray valves should'open within 5 seconds.

An additional test section was written and performed to-determine the stroke tig from full closed to full open for the pressurizer spray valves at 557 F, 2235 psig.' The opening times were:

Valve Opening Time PCV 455B 3.9 seconds-PCV 455C 6.97 seconds An additional engineering evaluation, determined that these opening times in conjuction with the previously determined pressure decrease rate were acceptable.

The pressurizer response to actuation of all pressurizer heaters was within the allowable bands as shown on Figure 2.3-2 indicating an acceptable response.

O e

2.0-15 n - - w-

i

~

i FIGURE 2.3-2

' NOMINAL

• PRESSURE RESPONSE TO OPENING OF BOTH PRESSURIZER SPRAYI VALVES' (Wi th ' Allowable Band) 2250, i 1.

, N.

! e N

_ a 2200 \ _. \ '

S \ ,

Ns

-f N o e \'

.L D .2150 - \- -

m - W .

\ T N s \ N

- N E \ N l

tt 2400 ' \ ~

N W

N \. % N

\ e  %' - Nominal Response

a \

D) ggyo \ \_ --Allowable Band

\

b st

'N \ N = Test Data 4 \ < N g . N 2000 s \ ,

0 20 40 80 80 100 120 140 TIME (seconds)

~

FIGURE 2.3-2 PRESSURIZER RESPONSE TO ACTUATION l

OF ALL PRESSURIZER HEATERS (With Allowable Band) 2350 <

/

/

, /

% 2330 ./

/

P'

?

=

S /

a N

~

23,0

/

/.

/ .

/

/

/ .

s Y

/ * /

- Nominal Responas

$ 2290 / s

'. /

N

/ --Allowable Band h

/ /,/

% / /

• Test Data

& 2270 / */ ,/

,' s' 2250 0 40 80 120 180 gog g4g TIMC (seconds)

2.4 RDCIOR 000LMrf SYSTEM FIDi MEASUREMENT I

Reactor Coolant Systen (RCS) flow indication is obtained from measurement of the differential pressure across the RCS coolant loop piping elbow which connects the steam generator piping and the reactor coolant pump suction (cross-over 'eg). Each RCS loop has three differential pressure flow transmitters which provide visual flow indication in the control. room, flow information to the plant computer, and voltage signals to the protection system. for the loss of flow rea tor trip. Wis test was performed at hot, no-load conditions (557 + 2oF, 2235 + 15 psig) to determine that RCS flow, as indicated by the loop elbow differential pressures, was equal to or greater than 90 percent of the thermal design value.

With four reactor coolant punps (RCP's) operating ~and the plant .at the required steady state conditions, data was collected from the. plant computer and the main control board indicators. This data was then averaged and

. converted to gpm. The total flow as determined from the plant canputer was

' 435,221 gpm as compared to the minimum required flow of 344,520 gpm. We results are shown in Table 2.4-1.

~As a result of this test, the twelve differential pressure flow transmitters were adjusted to indicate 100% flow at hot no-load conditions and -four ICP's operating. . Verification that FCS flow rate is greater than or equal to the thermal design flow rate using calorimetric data was done during the power ascension phase of the Startop Test Program and is discussed in Section 4.4.1 of this report.

s.

2.0-18

y TABLE 2.4-1

- RCS LOOP FLOW DETERMINATION PRIOR 'IO INITIAL' CRITICALITY Plant Computer. Main Control Board-Indicator Percent- GPM Percent GPM RCS 110.3 108,999 111.0 109,668' Loop 1-RCS - 110.1 108,749 109.8 108,515 Loop 2 RCS 109.1 108,594 110.0 108,680 Loop 3

ICS . 110.2 108,878 110.3 109,009 Loop 4

-Total - 435,221 - 435,873

, Flow -

Acceptance Criteria, Total Fl.ow > 344,520 gpm 4

s 2.0-19

y r

t 2.5 RERCNR COO [MF SYSTEM FIM COASTDutti TEST b

This test.was performed to measure the rate at which reacer coolant flow changes subsequent to a' simultaneous trip of all four reactor l coolant peps and to determine the reactor coolant system low-flow reactor trip time delay.

While' operating at hot, no-load conditions, the permissive P-8 was simulated by lifting leads'in the nuclear instrumentation cabinets. This ensured that low flow (<90%) in one reactor coolant loop would open ,the reactor -trip ibreakers. Reactor coolant'systen flow, reactor- coolant pmp breaker status and reactor trip breaker status was recorded on a high _ speed chart

~ recorder. - The four reactor coolant pumps were then simultaneously tripped by simulating an underfrequency mndition with a test circuit installed in -

the safeguards test cabinet (SB-030) .

~

The chart recorder trace was analyzed to determine normalized loop flow -

fractions at discrete points in time following the.RCP trips. At each time,'

the loop- flow fractions for all four loops were averaged to determine the g

core flow fraction. The relationship:

, F' (t) = - (- 1 -) -1 F(t) where F(t) .= core flow fraction  !

was.then used to. determine F'(t) (the inverse flow fraction) for each second during the first ten seconds following the reactor. coolant pump trip. Using ,

- the method of least. squares, F'(t) was fit as a function of time for 'the

. period of 2 seconds through 10 seconds:

N' (t) = At + B' '

where -A = slope B = Y-intercept

' ' The Westinghouse acceptance criterion' for the flow coastdown was:

TAU =1= w oas @ wn Parameter M

-A TAUg > 11'70

. seconds The test data TAUg was 12.56 seconds which was satisfactory. Table 2.5-1

• sumnarizes these calculations.

The chart recorder data was also used to determine the low flow reactor trip time delay. The loop inverse flow fractions were least squares fit for a

~ period from three to ten seconds following the reactor coolant' pump trip. -

The -least squares fitting parameters 'were then used to calculate the sensor delay time for each loop:

f' =-1/f where f = loop flow fraction f' (t) = At + B 2.0 _-. - _ . _ _._ - . . _ _ __ _ -__ _ . _. _

.. -- . - - - .. .. .. . -~

1.

4

- l 1

TABLE 2.5-1 I

FLOW COASTDOWN RATE CAICULATIONS I

7 Time' Core Average Flow F' (t) = - ( 1 ) *

-1

-l (sec) Fraction (F(t)) F(t)

G l.0 0 0.945 - 0.052 2 0.873 0.129 3 0.807 0.212 -

-4 0.749 .. 0.295-5 0.699.~ 0.378 6 0.660- - -

0.450 7- 0.619 0.536- '

8 -,~

0.588 . 0.608 9 0.566 - - -

0.691

-10 0.529- 0.768

- Method of Least Squares used to fit data:

[ F ' (t) = - At + B

~

'A = slope B = Y - intercept A = .0796 B = .0258

- TAUg = Flow Coastdown Parameter

=1 A

= 12.56 seconds

-Westinghouse acceptance cri ter on i TAUg > 11.70 seconds

. NOTE: Data fit for period from_2 through 10 seconds.

e a5 2.0-21

)

- Sensor delay' time =-T D. = - B)

A- .

' The longest sensor delay. time'was 0.533 seconds in loop 4. 'Cambining the sensor delay time (TD ) with.the time from when the first loop flow decreased to .the low-flow trip setpoi'nt- (90%) until the second reactor trip breaker i' - had opened :(T{ =.0.109 seconds) and the. gripper release time (Tg = 0.15 seconds) yielded a low-flow reactor trip time delay .of .0.792 seconds. The Westinghouse acceptance criterion was less than .or equal to 1.0 seconds, therefore the:results were satisfactory. Table. 2.5-2. sumnarizes these -

, . calculations.- -

The results of this test' confirmed, based on the Westinghouse acceptance i, -

criteria, that the flow coastdown following a trip of all.four reactor coolant pumps is. adequate to prevent the departure from nucleate boiling

_ ratio (DNBR) from decreasing below-the limiting value of 1.3.

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b

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+

• g e'

N

',' .?

. e -

4' 2.0-22

,- ., ,,u , ,. , .. . , , , _ _ , . . , _ _ . .

TABLE'2.5-2 II)W-FII)W REACTOR TRIP TIME DELAY CAIfULATIONS M

. Loop Inverse Flows - (1)

-Time Loop 1 Loop 2 Loop 3 Loop.4 (sec) i- 0- 1.0 1.0 1.0 1.0-l= 1.058 1.056 -1.065- 1.055 2- 1.139 1.136 1.158 1.148

~3 1.241 1.236 -1.243 1.234 4- 1.345 1.327 1.335- 1.332 5 -1.443 1.412 1.438 1.426 6 1.539- 1.502 1.508 1.518

, -7 1.659 1.574 1.623 1.616 8 1.709 1.679 .l.711 1.706 9 1.833 - 1.742 -1.816 1.809 10 1.914 1.880 .l.873 - 1.901 Ah B ';',,

0.096

-0.961 -

0.089 0.%7 0.092 0.969 0.095 0.949 Td '* O.406- 9.371 0.339 0.533 NOTES:

'l.. Inverse loop flow = f' = 1/f where f = loop flow fraction

2. A and'B are least squares fitting parameters for f' as a function of

, t: f' (t) = At + B. Data is fit from 3 through 10 seconds.

3. T D

e sor h Delay = G -'B)/A Time from first loop flow'at low flow trip setpoint until second reactor l

trip breaker trips: Ty = 0.109 sec Maximum sensor. delay time: T = 0.533 see D

Gripper ' release time (from rod drop data): T G = . sec Low-flow reactor trip time delay (T +T y D G): TLF = . sec Westinghouse acceptance criterion: Tg < l.0 seconds F

g 2.0-23 1

e L

2.6 R'ID BYPASS ETIM MEASUREMENP Each loop in the BCS contains a loop bypass manifold in which resistance tmperature detectors (RTD's) are placed to monitor hot and cold leg temperatures. Signals from the temperature detectors are used to compute minus the thereactorcoolantdeltaT(tmperatureofthehotleg,T"No,lant temperature of the cold leg, T and an average reactor tmperature (Tavg) . This infoM) ion is used by the reactor protection system as well as the process instrumentation system.

The manifolds for the cold leg ta peratures draw a sample from the RCS flow at the reactor coolant pump discharge. The manifolds for the hot leg tenperatures draw samples from RTD scoops in the reactor vessel outlet piping. The bypass lines join downstream of each set of t a perature detectors (hot and cold) and discharge into a comnon line. The combined bypass flow passes through a flow elenent before discharging into the suction side of the reactor coolant pump.

The purpose to this test was to measure the RTD bypass flow rate in each loop to ensure transport times of less than or equal to one second through the loops. The one second transport time was a design parameter for the overall RTD time response. The following combinations of flows were measured:

1) The combird hot and cold leg bypass flow rate was measured,

2) The cold leg bypass flow rate was measured with the hot leg isolated,

3) The hot leg bypass flow rate was measured with the cold leg isolated.

New low flow alarm setpoints were calculated at 90 percent of the total bypass flow for each loop. Since the sum of the individual hot and cold leg bypass flows with the other leg isolated was greater than the measure 3 total bypass flow, a correction was applied to obtain the actual flows based on the ratio of the measured hot and cold leg bypass flows. Table 2.6-1 sumnarizes the results and shows that the calculated bypass flows are greater than that required for a one secorx3 transport time through the bypass loops. In addition, with total bypass flow 'at the low flow alarm value of 90% and cold leg flow at its normal value, the hot leg flow still has a transport time less than 1 second.

The new low flow alarm setpoints were incorporated into the Wolf Creek

. Generating Station Total Plant Setpoint Document (TPSD) . The compldtion of this test satisfied an outstanding test discrepancy fran the Preoperational Hot Functional Test Program.

2.0-24

^

'l , . . .

a_' ,

t,,.

s TABLE 2.6  : R'ID ' BYPASS . F[OW - MEASUREMFRP . RESULTS (1) (2)

Total Calculated - Minimtsu' . (3)

. Loop Bypass' ' Tow Flow Leg- Required- Measured Calculated' Calculated Flow Hot.

No. Loop Flow- Alarm Setpoint ' Flow. Flow Flow- Leg Bypass at Low Flow (gpm) (gpm) (gpn) Alarm Setpoint- ,

1 265 238.5 Jiot 103.43 155 146.7 129.2 Cold- 67.62~ 125 118.3- -

m 2' 265 238.5 Hot 103.43 155- 149.4 122.9

'o

'k Cold 67.62 120 115.6 -

n 3 295 265.5 Hot 103.43 185 176.0- 146.5 -

Cold 67.62 125 119.0 -

4 282 253.8. ' Hot- -113.88 178 168.4 140.2 Cold 67.62 120 113.6 -

Notes:

(1) 90' percent of total bypass flow (2) Based on actual. pipe volumes and one second. transport time, .

(3) Total' Bypass Flow at: 90%,: cold leg flow constant ^

l t

+

h

-y y ,y--

7-, w- ., y . .- , - --, -, ,

2.7~ PRELIMINARY DATA COLLECTION FOR INSTRUMENT CALIBRATION

. While operating at hot, zero power conditions prior to initial criticality several preliminary data collection procedures were performed for procedures that were to be performed during the Power Ascension portion of the test program. The calibration procedures are discussed in more detail in Section 4'of this Startup Test Report.

AnalignmentwasperformedofthedegtaTandTavgprocessinstrumentation at hot isothermal conditions (557 12 F). Using. test cards, known.

resistances were input into ICS loop T and T instrument loops and the

- loop outputs were verified as well as E ta T. Y en at hot isothermal

. conditions actual instrument loop outputs were recorded. RCS spare T HOT and T outputs were compared to.the normal instruments with satisfactory rk$ts. Overtenperature delta T and overpressure delta T setpoints were also verified.

Data was collected for those components which are used to develop the rod speed control signals. This procedgre was performed with four RCP's operating and the BCS at 557 + 0,-5 F and 2235 125 psig. Parameters monitored for each loop were T Tav GeneratorPressure,TurbineImhs,e.Tk$s,ure,g,FeedwaterFlow, Steam and in addition, auctioneered NIS power, auctioneered Tavg and Tg were also monitored.

Test instrumentation was installed to monitor feedwater flow and steam generator pressure during later tests. In addition, the zero.of the steam

. flow and feedwater flow transmitters was verified at the process cabinets

' while the plant was in hot standby conditions. With the steam generator isolated, the transmitter voltages were measured at the process cabinets to ensuro no zero shift. One transmitter, AB-FT-542, had to be replaced because of static pressure shift.

o O

2.0-26

2.8 NUCLEAR INSTRUMDirATION SYSTD9 1

The nuclear instrumentation systen (NIS) was tested to verify that all voltage settings,. trip settings, and alarm settings in the NIS were within expected l tolerances and that all ranges were functioning properly. This was accmplished by conpleting a series of instrument and control channel calibration surveillance procedures written specifically to ensure the IHS is operable in accordance with Technical Specifications.

A total of twelve separate procedures were performed to complete this verification. One test was performed on each individual source range, intermediate range and power range drawer. The scaler timer and audio count rate drawer, canparator and rate drawer and the flux deviation and miscellaneous controls drawers were also checked out. No major problems or deficiencies were found during any of this testing.

A more detailed sunmary of all testing on the nuclear instrummtation system is given in section 4.4.3.

9 D

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_O e

2.0-27

2.9 RTD/IC CROSS CALIBRATION TESTS i

The purpose of this test procedure was to provide a functional check out of the reactor coolant systen resistance tenperature detectors (RTD's) and the incore thermocouples ('IC's) and to generate isothermal cross calibration data' for subsequent determination of individual RTD installation correction factors.

0 The0 test wgs performgd at four different temperature plateaus: 250 F, 345 F, 450 F and 557 F (within +5 F) . At each tenperature plateau, RTD resistances were read directly at the field wires using test boxes with multiposition switches and a digital _ voltmeter. Concurrently, 'IC readouts were obtained from the plant computer, and steam generator saturation pressures were obtained from tenporary test gauges installed on the main steam lines. The RTD resistances were then converted to temperatures using vendor curves; the saturation pressures were converted.to tenperatures using the ASME Steam Tables, and the 'IC temperatures were averaged. All of these tenperatures were compared and isothermal correction factors were calculated.

The initial performance of this test connenced on April 16, 1985 and was

.' essentially complete on April 30, 1985. The results are sumnarized in Table

~ 2.9-1. . The acceptance criteria for this test were:

0

1) All converted average RTD tenperatures were within +2 F of the calculated RCS average tenperature for the given temperature plateau.

t

2) All-average stgam saturation tenperatures for each tenperature plateau were within 12 F of the calculated BCS average tenperature.

3) .All thermgcouple average tenperatures'for each tenperature plateau are within 12 F of the calculated RCS tenperature.

All acceptgnce criteria were met except for the steam saturation tenperature at the 345 F plateau. Based on the accuracies of the Heise guages used to determine the steam pressure (11.5 psi) and the consistency of the RTD

-average, this was determined to be acceptable. .

Although acceptance criteria were met, as discussed above, during the initial performance of the test, further investigation of the data by

- Westinghouse resulted in a decision to repeat the cross-calibration test.

~

The main concern was whether the BCS was-in a truly steady state or isothermal condition when the initial data was collected. In the second performance of this test, additional precautions were taken to achieve the -

highest degree of steady state and isothermal conditions in'the RCS. In addition, data was collected with the R'ID-manifold-return valves in both the

- closed and the open position. The normal configuration is with the RTD-manifold-return valves open but Westinghouse determined that closing the valves would result in exposing both hot-leg and cold-leg narrow range RTD's to water at a single tenperature thereby g'iving better test data.

2.0-28

TABLE 2.9-1 RESULTS OF INITIAL RTD/'IC CROSS CALIBRATION TEST

'RTD No. Calculated Installation Correction Factors,' F 0 O 0 0 250 F 34S F 450 F 557 F TE-410A -0.5 0.0 -0.1 -0.1-TE-410B -0.7 -0.4 -0.1 -0.3 TE-411A -0.5 -0.1 -0.2 -0.1 TE-411B -0.7 -0.5 -0.1- -0.3 TE-413A -0.1 0.4 0.3 0.3 TE-423A 0.5 0.6 0.6 0.7 TE-433B 0.3 0.2 0.3 0.3-TE-4438 -0.3 0.0 0.2 -0.3 TE-430A 0.4 0.4 0.5 0.0 TE-430B 0.2 -0.5 -0.3 0.1 TE-431A 0.5 -0.5 -0.5 0.l' TE-431B 0.3 -1.4 -1.1 0.1

$ TE-420A 0.2 -0.2 -0.5 -0.1 i

TE-420B 0.0 -0.3 -0.2 0.1

$ TE-421 A 0.2 -0.1 -0.3 -0.1 TE-421B 0.1 -0.3 -0.1 0.1 TE-4138 -0.4 0.1 0.3 0.1 TE-433A 0.6 0.4 0.5 0.3 TE-440A 0.0 -0.4 -0.8 -0.2 TE-4408 -0.2 0.1 0.3 -0.4 TE-441A 0.0 0.7 0.7 0.0 TE-441B -0.2 0.9 0.9 -0.8 TE-4238 -0.1 -0.3 0.0 0.3 TE-443A 0.3 0.4 0.4 0.2 Avg p Tmp.

F 248.3 340.7 447.3 556.8 Steag Sat Temp F 248.6 339.4 444.8 556.0

~* * *

• 0 F

Avg 'IC Temp O 249.8 342.3 -447.6 556.7 F

T=T -T AW E -1.5 -1.6 -0.3 0.1 O

F

After reducing power from 30%, tgis secgnd test wgs performed on June 30, 1985. Data was collected at 375 F, 450 F and 557 F with the R7D-manifold-return valves both open and shut. The results of the second test are sumarized in Table 2.9-2. . The acceptance criteria remained the same for this test with the exception of:

1) All converted average R7D taperatures were within 11.7 F of the calcugated RCS average tsperature for the given tenperature plateau (12.0 F in the initial test) .

In gli cases, thgs more stringent acceptance criterion was met (TE-413B was 1.7 F at the 375 0

F plateau with the RTD bypass return valves open).- For 0 both the 450 F plateau cases, the average TC tempergture was more0 than 2.0 F less than the average R7D tenperature (close3 - 2.4 F, open - 2.8 F) . This wasdeterminedtogeacceptablesgncetheinstrumentaccuracyofthe thermocouple is 150F. At the 375 F plateau with RTD bypass return valves open, T was 2.2 F. A review of the data showed that this may have been causedbATy a slight non-isothermal condition in the RCS. Since T was in specificationfortheremainingsixcases,thiswasdeterminedn5kTto

~

be a problem.

The value of obtaining data with the RTD manifgld return valves closed was highly questionable. Note for example the 375"F data for the loop 1 RTD's.

Theo two hot lgg RTD's (-410A and -411A) were cooler than the average by 0.4 F and 0.6 F, respectively, while the two cold leg RTD's (-410B and -

411B) were higher in temperature than the FCS average tenperature. For the open-manifold case, this trend was reversed. In general, the quality.of the data for the open-manifold case was better than that of the closed-manifold Case.

The scatter in data (indicated by large installation correction factors) appears to decrease at the higher tenperature plateaus. This could have been due to several factors, two of the nost likely being:

1) A higher degree of isothermal conditions were obtained at the higher tenperatures.

2) The RID calculations are more accurate at higher temperatures.

In general, the test was satisfactorily completed. The most important accgptance criterion (individual RTD temperatures not different by more than 1.7 F from the average) was satisfied for all RTD's at all tenperature plateaus for both open and closed manifold cases. Those acceptance criteria nct satisfied at all points were determined to be acceptable. The test

. results were supplied to Westinghouse for possible determination of an improved calibration curve for the RTD's. Completion of this test satisfied an outstanding test discrepancy from the Preoperational Hot Functional Test

. Program.

2.0-30

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[

~ -- RESULTS T;S!!COto R1D/1C CROSS CALIBRATION TEST .

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.RTD No. . Calculated \ Installation Correction' Factors, Og .

j n

j' 375'T 450'T . 557'T t

-Valves CL Valves OP- Valves CL Valves OP Valves CL Valves OP- Valves OP TE-419A 9.4' -0.2' 9.7- -0 .1 ; - 9.8 9.1: ' 9. 2 l -410B -0.2 - 9.9 -0.3 -0.2 -0.6- -0.4 -0.1 - I

[~

.-411A 9.6 - -0.3- 9.7' -0.1- 0.8 9.9 -9.2

-4llB -0.1 0.1 -0.1 9.1 -0.3 -0.1 0.3 j -413A -9.2 1.1 9.9 0.3 >0.9- 0.3' -0.5 i -423A 9.2 0.4 9.2 0.5 0.l' 9.5 9.5 . - ,

-433B -0.5 -0.2 -0.3 0.0 -0.5 0.2 9.3-

• l -443B 0.1- 9.4 9.9 9.3 -0.3 0.0 9.9

-439A- -0.7 -0.6 -0.2 -0.3- 0.l' -0.1 0.0

-430B- -0.9 -0.7 -0 . 7 - -0.5- -0.8 -0.1 -0.1

-431A -9.5 -0.4 -0.1 -0.3 9.1 -0.1 -0.1 - -

P -431B- -1.1 -0.8 -0.8 -0.5- -0.8 -0.2 -0.2

! o -420A 0.1 -0.7 9.4 -0.4 0.9- -0.1 -0.2 l- M -420B -0.1 -0.1 -0. 2 - 9.9 '-0.3 0.0 -0.2 l: -421A 0.3 -0.5 9.4 -0.3 9.8 -0.1 --9.2 1

-421B 9.9 0.1 -0.1 0.1 - -0.2 9.9 -0.3 --

i -413B 1.4 1.7 0.6 9.9 -0.1 0.2 0.3

!' -433A -0.4 -0.1 -9.1 9.2 -0.2 0.3 0.3

-440A 9.4 -0.3 9.4 -0.2 0.7 9.9 -0.1-l -4498 -0.1 0.2 -0.2 0.1 -0.3 -0.2 -0.3

-441A 0.8 9.2 9.6: 9.0 9.8 9.1 0.0 j -441B 0.2 9.4 -0.3 0.0 -

-0.5 -0.6 i -423B 9.3 9.6 -0.4 -0.1 -0.6 -0.1 -0.3 i -443A 9.9 0.4 0.1 9.4 0.1 0.3 0.3

! Avg RTD ,

Temp 371.9 372.1 449.4' 449.7 556.9. 556.2 556.6 i

Steam Sat.

Tenp 369.9 369.9 448.1 '448.0- 556.6 555.7- 555.4 T=T - '

l Tsa f . 2.0 2.2  : 1.3 1.7 9.3 9.5 ' 1.2 Ave 1C Tcmp 371.4 379.7 447.0. 446.9 555.4. 554.4 554.2 i T=T g -

.' T 1c 9.5 ~1.4 , 2.4 2.8l 1.5. 1.5 2.4 T

e

2.19 THEHM0 COUPLE CORE SUBCOOLING MONITOR SYSTEM TEST The Thermocouple Core Subcooling Monitor System ('ICCN) consists of two trains which monitor fifty incore thermocouples, primary systen pressure and selected T and T RTD's. The 'ICCM therefore normally monitors primary system pre N re andC M perature. Mort: importantly, the microprocessor controlled system will calculate saturation temperature and pressure during an accident condition and will alarm when the margin to saturation is reduced to a preset level and again if the core ever reaches the saturation level.

The purpose of this test was to perform a preoperational type functional checkout of the 'ICCM. The checkout included verifying all thermocouples were functioning properly and the 'ICCM ffD displays, alarms, calculations, outputs and printers were working correctly.

With the plant at normal operating temperature and pressure, normal 'ICCM displays were verified to be functioning properly. Then, in order to be able to change input parameters, normal field inputs were disconnected and a signal injection test box was connected to the 'ICCM. Only one train at a time was taken out of service for testing so that the other train was always

operable. Using the test box, test engineers were able to inject a wide range of normal.and abnormal temperature and pressures. This method tested

. that IfD displays gave proper outputs, calculations performed by the 'IOCM were correct, alarms activated at the expected setpoints and verified that proper outputs were being sent to the plant canputer, analog indicators and the 'ICCM printers. A final check was performed after the signal test box was removed to ensure all the normal field inputs were reading properly.

During the test, Train B gave some unexpected displays when certain test signals were injected into the systen. After troubleshooting, it was discovered part of the A/D converter circuitry was out of calibration. A recalibration of Train B was performed and all the discrepancies were cleared.

- During the final verification of normal field inputs, it was discovered some of the thermocouples were not reading properly. Further investigation determined that three thermocouples were defective. Since the plant must be in Mode 6 in order to repair or replace the incore thermocouples, this problem will probably not be corrected until the first refueling. However, since Technical Specifications only require 2 thermocouples per core quadrant to be operable, the system still far exceeds the minimum requirements.

s 2.0-32 mee-.-r

+ .-

- .y =c 7 w

2.11 SPECIAL TEST PROCEDURE FOR

- '!HE PRESSURIZER RELIEF VALVES i

The pressurizer power operated relief valves (PORV's) are solenoid actuated valves which respord to a signal from a pressure sensing system or to manual control. Motor-operated block valves are provided to isolate each power operated relief valve if excessive leakage develops or if the PORV fails to close. The power-operated relief valves provide the safety-related means for reactor coolant systen depressurization to achieve cold shutdown.

During the preoperational hot functional test program, the following discrepancies were noted against the PORV's:

BB-PCV-455A - did not reliably and consistently close when operated from an initially ambient valve body tenperature BB-PCV-456A - did not reliably and consistently close when operated from an initially ambient valve body tenperature and leaked through the seat with the result that a water seal could not form in 4 the inlet piping.

, The purpose of thia test was to test the _PORV's after rework and demonstrate satisfactory operation. With RCS pressure at 2235 + 15 psig and a water

~

seal formed at the valve inlets, each PORV was tested individually by opening the valve in manual and allowing pressurizer pressure to decrease by 200 psi. The pressurizer heaters were deenergized during the test. The

. PORV was then closed. RCS pressure was then restored to 2235 psig and the other PORV was tested. A high speed chart reccrder was used to monitor valve opening and closing times.

Both BB-PCV-455A and BB-PCV-456A operated satisfactorily. The opening and closing times were within ' specification and there was no problen with valve closure. There were also no indications of valve seat leakage. The results are shown in Table 2.11-1. This test closed an outstanding test discrepancy from the Preoperational Hot Functional Test Program.

e 2.0-33 k

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• TABLE 2.11-1

- - RESULTS OF PORV OPENING /CLOSIM3 TEST

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-Valvej Opening . . ,

Acceptance Satisfactory'

' / Tii~c e

- " ' Critoria ~

Closure After-

's.- -

- 200 psig Pressure Drop *

y. ,

BB-PCV-455A- 0.2G seconds .

< 2 seconds Yes DB-ICV-456A 0.i seconds < 2 seccInds Yes s.

s, ,

s. ...- _o

• Both BB-PCV-455A and BB-PCV-456A closed in-less than 1 second.

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2.12 IDOSE PARTS MONrt0 RING SYSTEM The purpose _ of this procedure was to obtain baseline data from the loose parts. monitoring systen after the reactor core had been loadp. With plant conditions at normal _ operating tenperature and pressure (557 F, 2235 psig) ~

and four ICP's in operation, tape recordings of the noise from each of the

~

twelve accelerometers (channels) was obtained. A reference signal was introduced using the installed simulator while recording channels 1 through 74 . - Decibel levels were also obtained using the installed meter for each of the twelve accelerometers.

Channel. 2, which is one of two accelerome't ers mounted on incore thinble guide tubes at the bottom of 'the reactor vessel, indicated a high

- vibration. 'An alarm was present for vibration and loose parts at the loose parts. monitoring panel and the main control panel for channel 2.

After a review of the data by Westinghouse, it was determined that the noise was not characteristic of a loose part. For the following reasons, the

-noise was suspected to be normal thimble tube vibrations:

^

, 1) The noise only occurs on one (1) accelerometer as opposed to both accelerometers at the bottom of the vessel.

. 2) The_ noise stops when tha ICP's are turned off (lack of full flow) .

3)' Thinble tube vibration is consistent with qualitative experience at Lother Westinghouse plants with similar signals. .

Based on Westinghouse analysis of the test data and additional monitoring of.

. channel 2 with four ICP's in operation, the alert. level' alarm for channel- 2

.was increased from 1.813 volts to 3.0 volts.

4

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3.9' INITIAL CRITECRLITY AIO IOf IOfER TEST SEQUEEE l

l l

The initial' criticality and low power test segment of the startup program encompassed a number of activities ~ ranging from bringing the reactor l critica1'for the.first time to verifying design parameters of the core. An integrated test procedure was used to accanplish these activities. - It started by bringing the reactor critical .inmediately followed by determining the power range to be used for further testing. Nuclear Instrurnentation System checkouts were also incorporated into this portion of the test.

Control rod bank worths,. isothermal temperature coefficients, and boron i endpoints were measured for.various rod configurations. We worths of the

~

most reactive rod and a pseudo ejected rod were then determined followed by

. restoration of the reactor to a normal configuration. An additional rod-

. swap test was' performed to gather information to be used by-the IG&E Nuclear Fuels group. Wis dealt with swapping shutdown bank B, whose worth was

- known,.with the remaining control rod banks to determine their worth 1.e.,

insert shutdown bank A, withdraw shutdown bank B or vise versa. Data was gathered on site, however, the analysis was done by the fuels group as this test was conducted for information only. A combined effort of 10&E and Westinghouse personnel'was utilized to complete this extensive sequence of t'esting in a. timely, manner.

(

It'sh$uld.beIotedthatnaturalcirculationtestingwasnotperformedduring the low power test sequence. The performance of this test was comnitted to '

. by the first SNUPPS unit only as identified on page 640.6-l~ of Volume 11 of the Final Safety; Analysis Report. (FSAR) . -

- On May 22, 1985 at 0745, the plant was brought critical. Innediately .

- following, the'zero power ph W tP testing began.- All testing within this-segmmt was completed beloy ' pra ir level of 5 percent rated thermal power

' as' allowed by the operattr~ 3 cm w.

Acceptance criteria used for test results were' based on the core design

report which was provided by Westinghouse. A sumnary.of the results of the

' tests performed during this segment of the startup program follows:

' 1) - Isothermal- termerature coefficients at CBC and CBD. inserted, CBD

-inserted and t. all-rods-out configurations were measured to be within 1.5 ,

of't criteria of + 3.0 pcn/)'p

~

expected T. .A values thtis, positive moderator temperature meeting the acceptance coefficient was calculated for the all rods out configuration. .

Consequently, rod ' withdrawal limits were developed for use during the first cycle,.

2) Contro'l bank worths for control banks A through D were measured to be L-within 4.0% of the predicted values, well within the~ acceptance l criterion of 110%,

I 3) Shutdown Bank worths for shutdown banks C through E were measured to be within 2.6% of the predicted values, well within the' acceptance

- - criterion-of 110%,.

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' 4) . Total rod worth was measured to be within 5.4% of.the predicted value meeting the +10% acceptance criterion,.

e - 5) Critical boren concentrations were measured for six different control rod configurations. With the exception of the all-rods-out configuration (ARO),. all concentratilons met the acceptance criterion of'+10% of the predicted values. An evaluation of the all-rods-out.

case yielded no inpact on the safety analysis..from the critical boron concentration being 3 ppm out of tolerance high,

6) The differential boron worth was measured to be-within 0.19 pcav' ppm of -

the predicted value well within the +10% acceptance criterion,

- 7) Core sower distributions determined using flux maps.were acceptable- ~

.for t w configurations of all-rods-out, CBD-in, hot-zero power insertion limit, and pseudo ejected rod.

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3.1 INITIAL CRITICALITY I'

In preparation for bringing the reactor critical, the Nuclear Instrumentation Systen source range channels N31 and N32 were verifisi operational and, within twelve hours of criticality, analog channel operational test surveillance procedures were performed on each of the intermediate and power range channels. Reference counts were obtained for the source range channels using a 132 second interval and ten seperate counting periods for use in ICRR monitoring. The reference counts were 1181 counts for channel N31 and 1300 counts for channel N32.

The initial approach to criticality began at 1032 on May 21, 1985 at which time the reactor coolant systen (RCS) boron concentration was 2041 ppm and all control and shutdown banks were fully inserted. Beginning with Bank A, the shutdown banks were withdrawn in 50 step increnents stopping to obtain counts-for use in plotting inverse count rate ratios (ICRR's) . These ICRR plots for rod withdrawal verified the core would not be critical with the next 50 step withdrawal and are illustrated in Figures 3.1-1 and 3.1-2. Rod withdrawal continued with the control banks until control bank D was at 160 steps..

- Dilution to criticality began at 2040 on May 21, 1985 with a dilution rate of 60 gallons per minute. Boron concentration in the RCS was determined every 20 minutes. Plots of inverse count rate ratio versus time of dilution, RCS boron concentration, and makeup water addition were made during the approach to criticality in order to predict criticality. .These plots are shown in Figures 3.1-3 through 3.1-8. The dilution rate was changed to 30 gallons per minute at 0650 on May 22 and criticality was

. achieved at 0745 with boron concentration of 1343. Control bank D was then used to maintain the reactor just critical.

'Just after the reactor went critical, readings were taken to determine overlap between the NIS source and intermediate range channels. When ingpnediate range channels showed a positive indication, i.e, greater than 10 . amps, both source and intermediate range indications were recorded.

, Theindicagonswereagainrecordedwhentheintermediaterangeindication shows3 10- amps. It was not possible to get overlap regdings at any higherlevelsincethesourcerangereactortripisag10 counts per second even though the source range scale indicates up to 10 counts per second.

The overlap data is shown in Secticn 4.4 of the report and shows overlap between the source and intermediate range is greater than th'e reqaired 11/2 decades.

With the reactor stabilized and just critical, the range of core power for physics testing was determined. This was accanplished using a reactivity

- computer, supplied by Westinghouse, with an input signal coming from NIS power range channel N-42. Control bank D was withdrawn thereby increasing the flux level until the effects of nuclear heating were observed (i.e.

increpse in RCS average tenperature). This occurred at a power -7 1evel of 5.2 10- - amps on the reactivity computer picoamneter, and 8 x 10 amps on both NIS intermediate range channels N35 and N36. The tegting range wag declared to be 1/10 to 1/100 of these values or 5.2 x 10- to 5.2 x 10-amps on the reactiv.ity computer.

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4 After the testing range was determined, a reactivity computer checkout was performed. Positive and negative reactivity insertions were introduced by control rod movement with calculations of the change in reactivity made using the neutron flux doubling time. Comparing these calculated values to theoretical values-developed from the inhour equation, resulted in an average difference of 0.7%. Reactivity changes of approximately 25, 50 and 75 pan were used for the calculations.

Only. one significant problan was encountered during the approach to criticality. Water had been pumped from the spent fuel pool into the refueling water storage tank (WST) . Rod withdrawal was suspended until the boron concentration of the WST could be verified to be within Technical Specification requirements. After a 45 minute delay, verification was made and rod withdrawal resumed.

9

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3.0-12

3.2 00ttfRO[. ROD BANK WORTH MEASURDUNfS

}

Control rod bank reactivity worths were measured by monitoring reactivity changes associated with RCS boron and control rod bank exchanges.

Differential reactivity worths, being the ratio of the change in reactivity to the corresponding change in bank position, and integral reactivity worths,-being the total reactivity change due to the travel of the entire rod bank height, were obtainsi via these exchanges. After establishing a constant ES boron dilution or boration rate, the control rod banks were periodically-inserted or withdrawn to coupensate for the changing boron concentration. The changes in reactivity due to control rod bank movement

..were indicated on a strip chart recorder connected to the reactivity computer.

From an all-rods-out starting condition, boron endpoints and individual control rod bank reactivity worths for control banks A through D and shutdown banks C through E were obtained by RCS dilution. The worth of the most reactive rod, rod F-10, was then determined by withdrawing it and compensating with the insertion of shutdown banks A and B. When rod F-10 was full out, insertion of shutdown banks A and B continued by RCS dilution until.the banks were fully inserted.

- The reactor, which had previously been tripped to realign rod F-10, was then brought critical with shutdown banks withdrawn ard the control banks inserted. Initiating RCS boration, the control banks were withdrawn in order to measure their worth in the overlap mode. Upon completion, the rods were repositioned at the hot zero power insertion limit. By RCS boration, rod D-12 was then withdrawn to simulate an ejecta 3 rod and its reactivity worth measured. Rod D-12 was then realigned and the reactor manually tripped as this was the erd of rod worth measurenents.

A suumary of the results of the rod worth measurenents is presented in Table 3.2-1. The differential and integral reactivity worths of all cases have been plotted in Figures 3.2-1 through 3.2-10.

~

As indicatal in Table 3.2-1, all measured rod worths were within the acceptance criteria. One significant problem experienced during rod worth measurements was that at one point, individual rod groups within a rod bank became misaligned when switching back and forth from bank to bank on the selector switch. The largest misalignment encountered was four steps. Upon

_ completion of rod worth testing, the reactor was tripped and the step counters reset in order to realign the rods. The only other problen

encountered was during the worth measurements of the most reactive rod, F-

10. Technical Specification 4.10.1.2 requires that prior to this event, the rods must be tripped from the 50% withdrawn position to show insertion capability. When withdrawing control bank C to denonstrate this, a counts doubling occurred causing automatic boration of the ES. After terminating the boration, rod withdrawal continued slowly in anticipation of possible criticality. The reactor did go critical prior to reaching the 50% position but was quickly brought subcritical by rod insertion. Following an evaluation, which determined that control Bank C was worth more than predicted in this configuration (all other rods inserted), the RCS was

, borated conservatively and testing proceeded without any further difficulties.

3.0-13

~

TABLE 3.2-1 c':

1. WOLF CREEK GENERATING STATION -

. CYCLE 1 BOL MiYSICS TEST CONTROL ROD BANK WORTH EUPt%RY Bank / Rod Measured Worth Acceptance Criteria

.Configuratio.n* (pon) = (pan)

~

CBD' 650.4 650'1 65 CBC .(GD @ ' 0) -1194.3 1240 1: 124 CBB . (CBDiCBC 8 0): 1010- 970 1 97-

. CBA (GD,CBC,CBB00) 658 680 1 68

~ SDE (CB @ 0) 846.9 .870 i 87

' SDD (CB, SDE @ 0) 758 740.1 74, SDC ' (CB,SDE,SDD00) 954.7 960 1 %

ARI - 11 6322.5 .6680 1 668

. Ejected Rod D-12 548.5- < 860-

• CB = Control banks

, SD = Shutdown banks-

- ARI = All rods in' l

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3.3 ISOIRERMAL TEMPERA'IURE COEFFICIENT The isothermal temperature coefficient measurements we5e mp ished using a constant heatup or cooldown rate of approximately 10 F ger hour. With reactor power within the physics testigg range, 5.2 x 10- to 5.2 x 10" amperes, a cooldown of approximately 5 F was made by adjustment of the steam dump systs. Reactivity changes during the cooldown were recorded by the reactivity empgter. This process was then repeate3 for a heatup of approximately 5 F. Pressurizer level was maintained steady or slightly increasing during the process in order to eliminate boron reactivity effects due to outflow from the pressurizer.

Isothermal temperature coefficient measurements were performed at three different control rod configurations; all-rods-out; control bank D-in; and control banks D and C-in. A heatup and cooldown was performed for each coafiguration. The isothermal tmperature coefficient was taken to be the average of the values of the slopes of the heatup and cooldown plots from the reactivity cmputer.

The results of the isothermal tm persture coefficient measurements were all within the acceptance criteria as presented in Table 3.3-1. Using the value

. of the isothermal temperature coefficient, the mderator temperature coefficient (M'IC) was cala11ated for the all-rods-out configuration. This is simply the isothermal temperature coefficient minus the Doppler (fuel) t aperature coefficient. The MIC is requiral to be negative (i.e. as tmperature increases, negative reactivity is intr uced), however, the results of the calculation was a positive 1.03 F. This required rod withdrawal limits to be instatA which would preclude operation of the plant with a positive moderator tmperature coefficient. These limits are illustrated in Figure 3.3-1.

No major problems were mcountered during the isothermal temperature coefficient measurments and with rod withdrawal limits in pla e, all test results were satisfactory.

3.0-25

1 I s

v I

TABLE 3.3-1 iL *. -

.ISOIHERMAL TEMPERATURE COEFFICIENT RESULTS SUM 4ARY ^

Rod / Bank Measured Value ~Acceptanc riteria.

Configuration (pan F) ( )

.ARO -0.92 -2.03 + 3.0 GD 0 0 -2.05 -3.36 + 3.6 GD, CBC 0 0 -5.47 -6.66 + 3.0 a

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3.4 BORON EIOPOIlff Am BORON WORTH MEASUREMENFS

. Boron endpoints were measured in an integrated test procedure which also measured the control bank worths. With the reactor at hot zero power, RCS boron concentrations were measured at several rod bank configurations for emparison to design predictions. The RCS conditions were stabilized with the controlling rod bank at the desired endpoint position. Critical boron concentrations were then measured for the endpoints.

Results of the endpoint measurenents are presented in Table 3.4-1. With the exception of the all-rods-out configuration endpoint, .all measured endpoints were within 6.2 percent of the design predictions. In the all-rods-out case, the measured concentration was 3 ppm outside the allowable tolerance.

The situation was evaluated by Westinghouse with the result that no impact on the safety analysis was presented, therefore the measured endpoint was acceptable.

A differential boron worth was calculated using values of measured boron concentrations and rod worths. Dividing the total worth of the control banks (in overlap) by the difference of the critical boron concentrations from the all-rods-out to the control banks in configuration resulted in a differential boron worth of -10.08 pm/ ppm, well within the + 10% of design prediction tolerance as indicated in Table 3.4-2.

3.0-28

s.

TABLE 3.4 . .

BORON ENDPOItfr SUtNARY '

)

4 Control Rod Configuration Measured Value (ppm) Acceptance Criteria (ppm)

ARO_ 1352.2 1299 i 50 CBD @ 0 (ARO-CBD) x '65.3* 63 1 6 CBD, CBC @ 0 ' -

126.9'* ' 123 + 12

~

(CBD-OC)

CBD, CBC, CBB @ 0 87.9* ,

93 1 9

-(CBC-CBB) ,

GD,- CBC, CEL, CBA 0 0 6's . 3

• G5 1 7

, .(CBB-CBA)

ARI. .1 \ 605.9* '

625 1 63 (ARO -(ARI-1)) -

• Values-are the difference between successive endpoint measurements.

1

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• TABLE 3.4-2 DIFFERENTIAL BORON WORTH SUM 1ARY Differential Boron Worth = Measured Value Acceptance Criteria.

(pan / ppm) - (pan / ppm)

Control Bank Worth (overlap) -10.08 -10.27 + 1.03 CB(ARO) - CB (CBA @ 0) 6 6

e l

e'-

b

- - . ._. 3.0-30

3.5 POWER DISTRIWFION MEASURDENFS

'the core power distributions were measured using the incore flux mapping systen. Data from this system was then analyzed using the Westinghouse INCORE computer program.

Four flux maps were obtained during the low power physics testing segment of the startup program. The maps were taken at different control rod bank configurations which were a'l-rods-out, control bank D at 0 steps, hot zero power insertion limits, and a pseudo-ejected rod. Incore power tilts, reaction rates, and hot channel factors were reviewed for the flux maps all of which resulted in acceptable values. Results of the maps along with acceptable values are given in Table 3.5-1.

Analysis of the first flux maps indicated abnormalities between thimbles R-08 and N-08. After a detailed examination of plant documentation, it was determined that the two thimbles had been interchanged at the seal table.

'1he problem will be remedied during the first refueling by appropriate repositioning of the thimbles at the transfer device. The flux map data is currently being corrected for each map by manually interchanging the data for ' thimbles R-08 and N-08.

Several equipment poblems hampered the running of the flux maps. In most' cases, not all of the detector drives were available for operation.

However, constant maintenance and the redundant design of the system allowed the flux maps to be obtained, o ,

t.

I 3.0-31

~

I 4

TABLE 3.5-1 POWER DISTRIBUTION SUPMARY-Map l Control Rod Incore Tilt Reaction Rate- Enthalpy Rise Hot- Axial Offset:

Nunber Configuration Error Channel Factor- Fg(Z)

F A HN Measured Acceptable Measured k:ceptable Measured Acceptable Measured Acceptable Measured Acceptable ClM992 ARO 1.9157 <l.94' -6.9% < +10% 1.4395 1.39+0.14 2.3467- N/A -1.788 N/A:

CIM991 CBD @ 9 1.9125 <1'. 94 -8.8% < +10% 1.5792 1.57+0.16 2.7792' N/A -16.577 N/A 4

ClM993 Hot zero power 1.9174 <l.94: -9.5% < +19% 1.5775 N/A- 2.7654 N/A -36.928 .- N/A insertion limit ClM994 Pseudo rod 1.9513 N/A +

_50.8 N/A 3.8293 N/A 6.6999' ~

<7.03 -33.363 N/A Ejection

.M

.t r

te

r-4.5 PONER ASCENSION TESTING Following the completion of low power physics testing, the power ascension phase of the startup test program was begun. We testing in this phase of

, the test program included various at-power physics tests, control system dynamic response tests, overall transient and trip testing of the plant, and calibration and alignment of plant instrumentation and control systems.

Additional testing included a steam generator moisture carryover test, the NSSS acceptance test, thermal and dynamic testing of the main steam and main feedwater systems, biological shield surveys 'and turbine generator testing.

Several tests identified in the Final Safety Analysis Report (FSAR) were not performed. The loss of heater drain pmp test was comitted to be performed by the first SNUPPS unit only as identified on page 640.6-1 of the Volume 11 of the FSAR. The pseudo rod drop test was exmpted fra the Startup Program by the NRC due to potential Technical Specification violations during performance of the test.

We NRC lifted the 5% power restriction on June 4,1985. We low power physics test sequence was approved by the PSRC and the Plant Manager authorized the start of power ascension testing on June 5, 1985. We turbine generator was synchronized to the grid on June 13, 1985. The initial phase of power ascension testing was approved by the PSRC and, after Plant Manager authorization, power was increased to 30% on June 19, 1985.

We plant was shutdown from outside the control room on June 29, 1985.

Power was increased to 50% on July 6, 1985. The rods drop and plant trip

-test was performed on July 16, 1985, and testing at 50% power was completed on July 18, 1985. After PSBC review and approval of the. test procedures, the Plant Manager authorized the start of 75% power testing on July 19, 1985.

Testing at 75% power was started on July 20, 1985 and completed on July 29, 1985. The PSRC reviewed and approved the 75% power test procedures and the Plant Manager authorized the start of the 90% power test sequence on July 30, 1985. The plant was at,90% power on August 4, 1985, testing was completed on August 6,1985 and, after PSRC approval of the test procedures, the Plant Manager authorized the start of 100% power testing on August 8, 1985.

The plant was initially brought to 100% power at 1607 on August 8, 1985. A 100 hour0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> continuous run at 100% power was'ccmpleted at 2007 on August 12, 1985. Power was then decreased to 55% power while vibration problems with the "B" main feedwater pump were investigated. The plant was returned to 100% power on August 21, 1985.

%e 100% power plant trip test was successfully performed on August 28, 1985, and after a checkout of the NIS syst s to verify that the syst s had not degraded during 100% power operation, the 100% power test sequence was empleted. The remaining test packages were approved by tre PSFC on August 30, 1985. The unit was declared comercial at 0114, September 3,1985.

4.0-1

4.1 AT PONER PHYSICS TESTING

The At-Power Physics Tests are a series of tests to verify accuracy of the

physics models used in core design and accident analyses, .to verify that hot channel factors and control rod worths in a rod ejection accident are

conservative, and to obtain calibration data for the Nuclear Instrumentation Syste -(NIS). A sumary of the physics tests performed during power ascension'follows:

1)- Incore Movable Detector and Thermocouple Mapping at Power - Flux maps were taken at power levels of 30%, 50%, 75%, 90%, and 100% full power. The hot channel factors were measured to be within NCGS Technical Specifications. All measurenent parameters met their design and accident analysis acceptance criteria after clarification by

, Westinghouse,

2) Axial Flux Difference Instrumentation ~- Flux map results were used to calibrate the Nuclear Instrumentation System to axial power distribution in each section of the core. Resetting of the NIS .

circuitry was acceplished to duplicate certain key flux map parameter results, 3J Power Coefficient Determination - A comparison of the Doppler only Power Verification Factor showed the measured value was well within the design value tolerance of +5%,

4. ~ Pseudo Rod Ejection Test - With control rod D-12 pulled to 228 steps, core power distributions associated with a flux map taken at that time

. indicated hot channel factors were within the limits of NCGS Technical

' Specifications and the partial worth of the ejected rod was within the value for design tolerances.

4.0-2

4.1.1 IN00RE MOVABLE DETEIOR MAPPING AT POWER The core power distributions were measured during the power ascension portion of the startup testing program using the incore movable detector flux mapping system. . The results of the flux maps are give in Table 4.1.1-

1. The flux map taken with an ejected rod will; be discussed in Section -

4.1.4.

The flux maps taken and listed 'in Table -4.1.1-1 were taken over a range of 30% to 100% power at various control rod configurations. The flux maps were taken to provide results to verify the accuracy of the physics models used in the core design, to verify operation of the reactor within EGS Technical Specification power distribution requirements'during typical operation, to obtain calibration data for the Nuclear Instrumentation Systen (NIS), and to obtain baseline data for the target axial flux distribution surveillance.

As shown in Table 4.1.1-1, normally 58 thimbles are used for a full flux map. Several flux maps taken during the incore/excore detector calibration used only 16 thimbles because it was only desired to monitor the axial offset.

The measured power distribution parameters are compared with NCGS Technical Specification limits in Table 4.1.1-1.- The power distribution for all flux maps met their design, accident, and NCGS Technical Specification limits.

The predicted vs. measured axial powers were within the 10% value as required by design when clarification was obtained from Westinghouse indicating that this was only applicable for:the larger unrodded core region during normal. operation.

~

~

~ All power distribution measurement results were acceptable when compared with the design, accident analysis, and Technical Specification limits. The core exhibits a small natural tilt to quadrant II, NIS channel N43. The small natural tilt does not exceed the WCGS Technical Specification ~ limits.

r i

4.0-3

a ,

. .. - e . , , , -

-TABLE 4.1.1-1J

< s 10GS - INOORE FLUX MhP. SUmhRL DURItG ' POWER ASCENSION .

~

i Map No. ... Date %P . Core Rod- .F m .F A n, - . F.

No. . Thimbles ~B/D Pos'. .

Loc. Limit Max. AX. PT.

't Used N D/MIM Max. . Loc. Limit Max.  : Limit ClM005* 58 6/27/85 -39 134- HFPRIL 2.297G DS- - 4.649 1.4216 -D7 1.6986 1.5662 9- 1.959 ClM996* 58 6/27/85 39 _134 HFPRIL 2.2384 E14 4.649. 1.5304 E14 1.6986 1.8390 9 1.9494 D129228

~ClM997 58. 7/9/85 50 229 JD9292 2.0976 .D9 4.649 1.4228 D9 1.639- 1.4515 39 -- 1.795 i C1M008 58- 7/9/85 59 289 D9290 2.1117 D9- 4.640 1.4246 D9 1.639 1.6186 9 1.795- .

ClM009 58 7/21/85 68 370 D9202 2.0986 D12 3.412 1.3468 D9 1.5854 1.4752 39- 1.649

l. ClM010 58 7/26/85 -75 523 D@210 1.9891 D9 3.093 1.3538 D9 1.5645 1.3665 39 1.6275

!- C1M0128 16 7/26/85 75 533 D9152 2.2292 H11 3.993 1.395G Hll 1.5645 1.5242 9 1.7955 C1M013 58 7/27/85 75 533- D@l58 2.2523 D9 3.093 1.3798 D9, 1.5645 1.5307 9 1.7955 i C1M0198 16 7/27/85 75 -533- D0219- 1.9932 Hll 3.993 1.3491 Hll 1.5645 1.3585 36 1.6275 l- C1M020 58 . 7/27/85 75 537 D0210 2.9373 D4 3.093 1.3635 D4 1.5645 1.3721 17 1.6275

C1M921 58 8/6/85 92 -737 D@214 2.036G D9 2.522 1.3634 D9- 1.5138 1.367G 23 1.5748  :

CIM023 58 8/12/85 199 932- D0215 2.9313 HS 2.320 1.3621 D9 1.49 1.3663 16 1.55 t ClM024 58 8/22/85 les 1192 D0218 2.9250 HS 2.320 1.3632 H5 1.49 1.3703 16 1.55 o .

Radial Tilt. Incore AO (%)

Map AO .

j No. N41 N42 N43 N44 N41 . N42 N43: N44 5 ClM005* -12.214 0.9903 1.0137 1.0134 0.9826 -11.956 -12.591 -12.106 -12.240  ;

j- ClM006* -6.014 0.9461- 1.9935 0.9945 9. % 59 -8.484 -0.468 -7.571 -7.535 ClM007 -4.269 9.9896 1.0172 1.0128 0.9805 -4.382 -3.907 -4.618 -4.171 s j ClM008 -5.410 0.9894 1.0159 1.0130 0.9817 -4.497 -6.001 -5.299 -5.844 '

l ClM009 -7.762 0.9890 1.0105 1.0100 0.9905 -7.601 -8.084 -7.905 -7.458 ClM010 -6.468 9.9915 1.0084 1.0106 0.9895 -6.379 -6.629 -6.577 -6.287 C1M012# -21.813 l 1.000 1.099 1.999 1.990 -21.813 -21.813 -21.813 -21.813 CLM013 -25.133 0.9915 1.9973 1.0120 0.9892 -25.013 -25.406 -25.165 -24.949 '

ClM019# -1.171 1.000 1.000 1.000 1.000 -1.171 -1.171 -1.171 -1.171 ClM020 2.389 0.9907 1.0984 1.0111- 0.9898 2.493 2.381 2.349 2.334 ClM021 -10.035 0.9937 1.0982 1.0106 0.9875 -9.863 -10.049 -10.244 -9.985 i ClM023 -10. %2 0.9946 1.0065 1.0199 9.9881 -10.767 -10.996 -11.098 -10.986 C1M024 -9.926 l 9.9945 1.9060 1.0100 0.9895 -9.719 -9.924 -10.102 -9.958

• Flux Maps for eject'ed rod verification.

3

1. Quarter core flux maps. .

1 NCyfE: Map's #'s ClM0ll, ClM014-ClM018 were quarter core maps taken during the incore-excore calibration and not used. ClM922 is non-existent.

i-r - w ~ . - . . , . --. .e-. m. . . , , , .e , , - .. , ., ,

4.1.2 AXIAL FLUX DIFFERENCE INSTR [MEKfATION CALIBRATION This test consisted of three sections:

1) A preliminary Incore-Excore calibration after the trip at 50% Rated Thermal Power and prior to escalation above 50% power. It was done using the incore flux maps taken at 30% and 50% power,

2) The Incore-Excore calibration at 75% power using a series of full core and quarter core flux maps taken at the 75% power level,

3) A " fine tuning" of the Incore-Excore calibration at 100% power.

-The Axial Flux Difference Instrumentation Calibration is based on a number of flux maps taken at various axial offsets (AO) to determine a correlation between Incore and Excore detectors (AO is defined as difference of % power in top half of core and bottcm half of core divided by the sum of % power in top and bottom of core). Plots of Incore AO vs. Excore AO were generated for each excore channel. The slope of a least squares straight line drawn through the AO points yields the correcrion factor to correlate the NIS excore detectors-to the measured incore power distribution. The correction factor is equal to the least squares straight-line slope and is shown in Table 4.1.2-1 for each channel. The correction factor is entered into the plant computer at the computer addresses shown in Table 4.1.2-1.

Another plot of excore (NIS channel) top and bottom detector current vs. ,

incore AO is used to predict 100% power current values from each detector.

100% power currents are used because at that power level, delta q (defined as % power in top half of core minus % power in bottom half of core).is equal to AO and subsequent calculations can be minimized. The resulting least squares fit straight line is used to derive expected current values which are input into the NIS circuitry to check control board aM ccuputer values for consistency at specific AO values. Figure 4.1.2-2 shoQs 100%

power current values input into the NIS circuitry to calibrate the excore system to the incore measurement.

%e preliminary calibration at 50% power indicated that delta q values from the plant ccmputer did not favorably ccmpare with predicted values. A closer analysis showed the computer was using an average power from all NIS channels instead of only the channel being tested at this point in time, as assumed in calculating the predicted value. We predicted values were

. corrected to actual.NIS indications and close agreetient between predicted

.and actual computer values resulted. Since this was only a preliminary calibration, it was decided power escalation to 75% could be safely achieved, and a more accurate Axial Flux Calibration done at that power level.

A series of flux maps were taken at various AO values for the Axial Flux

- Calibration at 75% power by iMucing an axial xenon oscillation in the reactor core. %e flux maps used were maps ClM010, ClM012, ClM013, ClM019, aM ClM020 shown in Table 4.1.1-1. These maps allowed for use of five

. points to obtain the least squares straight line values to perform an

-accurate Axial Flux Calibration. Inputting the 100% power current values into the NIS circuitry resulted in a discrepancy between predicted and 4.0-5

TABLE 4.1.2-1~

.INCORE/EXCORE CORRECTION FACTOR POWER LEVEL Correction Factor K0554 K0552 K0551 K0553

-(N41) (N42) -(N43) (N44) 50% 1.350 2.192 -1.602 '1.525-75%> -1.8188 1.8418 - 1.7949 1.9342 1.8188 1.8418 1.7949 1.9342 100%-

F- [

0 4.0-6

TABLE 4.1.2-2 100% NIS CURRENT VALUES Power A.O. NIS Channel Currents (Ha) ~

Plateau (%) (%) 41 42 '

43 l 44 Top Bottom Top Bottm Top Bottom Top Bottom 50 -30 213.0 329.5 231.3 327.3 207.2 329.3 229.9 341.1

-75 135.8 430.5 164.2 375.4 117.2 395.4 168.7. 430.2

-35 204.4 340.9 223.8 332.6 197.2 336.6 223.1 351.0 0 264.5 262.2 276.0 295.2 267.2 285.2 270.7 281.7

+7 276.5 246.7 286.4 287.7 281.2 274.9 280.2 267.8

+75 393.2 93.9 387.7 214.9 417.2 174.9 372.7 133.2

+$0 316.0 194.9 320.7 263.1 327.2 241.1 311.5 222.3 75 -30 226.3 316.1 239.7 341.7 226.6 339.2 236. 1 328.4

-75 154.3 378.1 167.6 411.3 157.2 411.2 165.8 389.1

-35 218.3 323.0 231.7 349.5 218.8 347.2 228.3 335.1 0 274.4 274.7 287.9 295.4 272.8 291.2 283.1 287.9

+7 285.6 265.0 299.1 284.6 283.6 280.0 294.0 278.4

+75 394.4 171.2 408.1 179.5 388.5 171.1 400.3 186.6

+30 322.4 233.3- 336.0 249.0 319.1 243.1 330.0 247.4 100% 0 274.7 269.6 284.8 287.4 272.5 285.6 281.4 282.4 v

4, ,.

4 4.0-7

measured values from the delta I penalty generator and certain process emputer values for delta q. A new gain for the delta I penalty generator was calculated using voltage signals to prevent otherwise required gain changes each time an Axial Flux Calibration was done. The newly calculated gains produced the expected function generator output thereby satisfying the

. acceptance criteria. The delta I penalty function generator gain values for each NIS channel are shown in Table 4.1.2-3. The process computer flagged all values with AO +75% as unreliable. This was understandable since the

.cmputer was scaled for' voltages of +10 to -10 volts and 175% A0 values would correspond to voltage magnitudes greater than 10 volts. With the preceding discrepancies explained and/or corrected, the Axial Flux Calibration was satisfactorily cmpleted at 75% power and power escalation to 100% was allowable for Axial Flux Calibration verification purposes.

At 100% power, a flux map was taken (map ClM023 as shown on Table 4.1.1-1) to verify the Axial Flux Calibration values input at 75% power were 1) correct, or 2) needed " fine tuning". It was determined a " fine tuning" of the NIS was needed to give the required accuracy between the control board delta'I meters, process cmputer delta q, and flux map (Incore) A0 measurement. A one point correction for detector currents was done to give the required identical values, within acceptable tolerances, for process

. computer delta q, and control room delta I meter values (at 100% power percent delta I = percent delta q = AO).

With the " fine tuning" completed another flux map was taken at 100% power to verify that the " fine tuning" gave successful correlation between control room delta I indication, process computer delta q output, and flux map (Incore) A0 results. The results shown in Table 4.1.2-4 indicate the Axial Flux Calibration at 100% power was within the required accuracy.

The Axial Flux Difference Instrumentation Calibration showed improved accuracy each time it was checked from 50% power to 100% power. Each adjustment to the NIS circuitry brought the incore/ computer / control room meters into closer agrement until the final a3justment at 100% power verified incore delta q vs. control room meter delta q and computer delta q was within the design acceptance criteria of 1.5% and the control room meter delta q vs. cmputer delta q was within the design acceptance criteria of 10.5%.

4.0-8

TABLE 4.1.2-3~

GAIN VALUES EOR DELTA I FUNCTION GENERATOR 2

%P N41- N42- N43 N44- ,

4

. 50 1.661 2.480 - 1.978 2.073 75 110- +

-1. 0 l.0 1.0

' 100 1.832 1.845- 1.800 1.965:

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TABLE 4.1.2-4 DELTA q VALUES AT SPECIFIED POWER PIATEAUS.

N41 N42 N43 Power Meter / Cap Meter / Comp Meter / Comp Plateau Meas. Meter Comp. Error Meas. -Meter Cmp. Error- Meas. Meter Camp Error

+30 +30 +18.4 11.6 +30 +30 +15.3 14.7- +30 +30 +17.4 12.6 50% +7 - +7 +4.4 2.6 +7 +7 ,+3.8 3.2 +7 +7 +4.0 3.0 0 0 * -

0 0 0.1 -0.1 0 0 * -

-30 -30 -19.3 -10.7 -30 -30 -16.4 -13.6 -30 -30 -12.1 -17.9

+30 +30 +22.3 7.7 +30 +30 +22.2 7.8 +30 +30 +21.7 8.3 75% +7 +7 +5.3 1.7 +7 +7 +5.2 1.8 +7 +7 +5.1 1.9

. 0 0 0 0 0 0 0 0 0 0 0 0

-30 -30 -22.4 -7.6 -21.2 -8.8 -30 -30 -22.2 -7.8 100% -10.8 -9.0 -9.0 -0 -11.0 -9.0 -9.4 -0.4 -11.1 -9.0 -10.3 -1.3

-9.7 -9.8 -10.0 -0.2 -9.9 -10.1 -10.0 0.1 -10.1 -9.8 -10.1 -0.3 Acceptance Error 0.6 -

0.6 -

0.6

?

c) N44 -

Max.

Power Meter / Comp Meas./Cm p error Plateau Meas. Meter Cmp. Error Ch. No.

+30 +30 +14.9 15.1 15.1 N44

• Computer flagged as unreliable 50% +7 +7 +3.4 3.6 3.6 N44 0 0 0 0 - -

-30 -30 -14.0 -16.0 -17.9 N43

+30 +30 +22.2 7.8 8.3 N43 75% +7 +7 +5.3 1.7 1.9 N43 0 0 -11.9 11.9 -11.9 N44

-30 -30 -22.6 -7.4 -8.8 N42 100% -11.0 -11.0 -9.0 -2.0 -2.0 N44

-10.0 -11.1 -10.7 0.4 0.7 N44 Acceptable Error 0.6 1.5 -

b

i 4.1.3 PG4ER COEFFICIENT DErrEEMINATION Power coefficient measurement was done during the power ascension program to verify values'used:in the nuclear design and accident analysis prediction for the Doppler Only Power Coefficient.

.The measurement values were achieved at power plateaus of 30, 50, 75, and 90% power by causing a series of small load swings at the turbine-generator. These series.of load swings resulted in corresponding temperature swings in the Reactor Coolant System, but of opposite sign to the turbine-generator. load swing. The Reactor Coolant Systen tenperature change for each turbine-generator load change was recorded and used in a subsequent calculation to determine a Doppler Coefficient Verifigation Factor. 'Ihe measured Doppler Coefficient Verification Factor (C") is the ratio of the change in core average temperature ard the change in power due to.the doppler effect. 'Ihis ratio is equivalent to the predicted ratio of the Doppler Ong Power Coefficient and the isothermal temperature coefficient (C ) . Since for small changes of temperature, the void coefficient portion of the power coefficient is insignificant, the ratio is

-also proportional to the power coefficient.

'The predicted value of the verification factor was calculated using the design values for the Doppler Only and isothermal temperature coefficients.

The measured value of the Doppler Coefficient Verification Factor was

^

compared with the predicted value. The results of the comparison are shown in Table 4.1.3-1. The results show the values used for the nuclear design and. accident analysis prediction are very close to those measured and that .

the measured values are well within the required acceptance criteria.

4.0-11 I ~y

TABLE 4.1.3-1 DOPPLER COEFFICIEBTP VERIFICATION FACTORS l-Power C" C C" + C Acceptable Level (%)- ~O

( F/%) (OF/%)' [+CE 30 -2.35 2.34 0.01 < 0.5 50 ' -1.90 1.89 0.01 < 0.5

'75

-1.16 -1.21 0.05 < 0.5 90- -1.087 1.18 0.093 < 0.5 t

t

}

4.0-12 As

4.1.4 PSEDO ROD F.TBCTION TEST

. This test is used to verify that the values in the design and accident analysis for ejection of the most reactive rod from the Hot Full Power Rod Insertion Limit' (HFP RIL) are conservative.

Two main parameters are measural by this test for comparison with design values: 1) the positive reactivity worth added from the EP RIL to the full

-out position for control rod D-12 (designated as most reactive rod) and 2) the_ core power peaking factors resulting.from ejection of rod D-12 from the core with all other control rods at the W P RIL.

The worth of control rod D-12 was measured by pulling only rod D-12 from the HFP RIL to the full out. position and measuring the resulting change in moderator / coolant tenperature, Tavg. The mathenatical product of the isothermal tenperature coefficient and tenperature change results in the reactivity worth of rod D-12 from the HFP RIL to 228 steps withdrawn. The results and acceptance criteria for D-12 rod worth.are shown in Table 4.1.4-1.

Following D-12 rod worth determination, rod D-12 was reinserted to the HFP <

'RIL and the calculated worth was equated to boron worth. Rod D-12 was

- subsequently borated to the full out position with all other rods renaining at the HFP RIL and a full core flux map was obtained. Pertinent core peaking factor results from analysis of a flux map before and after rod D-12 ejection, are shown in Table 4.1.4-2.

4 e

e 4.0-13 Y

r.

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TABLE 4'.l.4-1

'D-12-ROD WORDI FROM HFP RIL-

& %P 30 .

, .a : d' elta rod position (steps) 67 reactivity change '(pan). 17.04

- 1.10 - times reactivity change (pan) -

18.74 acceptable reactivity change- < 230 r

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s 4.0-14 h: .-

9.

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. TABTE 4.1.4-2 s

. - FLUX HAP RESULTS FROM ROD D-12 FJK: TION

_ir

- Prior to D-12 ejection D-12 ejected

. -- Core Parar::eters : Measured. Limit Measured- Limit

~

F delta H Nuclear - -l.4216  :

1.6986- N/A -N/A Fxy 1.5662 - 1.95 N/A N/A

~

2.26 FQ-(Z)~

2.2070 4.64 2.2415 l.0137 . l.02

- OPTR N/A 'N/A-e N

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4.2 COM900 SYSTEM DDRMIC RESPONSE

--. Prior to the completion of .the 30% power plateau testing, the proper response of the reactor ' control systs, the steam dep control system and the steam generator level control was verified. ~ The purpose of these tests

. was to verify that the controller settings resulted in relative stable

. operation. The steam gmerator level control system section also discusses testing done at higher power levels. The overall plant response to.

. transient and trip testing is discussed in section 4.3.

4.0-16

s.

4.2.1 DYNAMIC AUFOMhTIC STEAM IXMP CONTROL me steam dmp -(turbine bypass) system consists of twelve valves _which are designed to handle 40% of full turbine steam flow at full load pressure.

Seven valves discharge into the low pressure condenser, four valves discharge into the intermediate condenser, and a single valve discharges into.tha high pressure condenser. The system is designed to:

1) ' Allow a 50 percent step electrical load reduction without reactor trip (Section 4.3.2) . _ %e system will allow a turbine and reactor trip

~ frm full power without lifting the main steam safety valves,

2) Control steam generator pressure at no-load conditions,

3) Bypass ' steam to the main condenser during plant startup and permit a normal manual cooldown of the ICS from a hot standby condition to a point consistent with the initiation of residual heat removal systen operation.

The steam dump system, during normal operating transients for which the plant. is designed, is automatically regulated by the ICS temperature control

'systen to maintain the programed coolant temperature (Tavg node) . We progranned coolant tenperature (Tref) is derived from the high pressure turbine first stage' impulse pressure, which is a load reference signal. We difference between Tref aM measured Tavg is used to activate the steam dmp system under autcmatic control. %e system operates the valves in two fundamental modes. In one mode, two groups of six valves each trip open sequentially in approximately 3 seconds. tis operational mode is activated during a large reactor-to-turbine power wismatch (load rejection controller). In the _second mode, four groups of three valves each modulate open sequentially in approximately 10 seconds -(plant trip controller) .

When the plant is at no load and there is no turbine load reference, the system is operated in the pressure control mode. The measured main steam header pressure is compared to the pressure set by the operator in the control room. The pressure control mode is also used for plant cooldown..

Both the Tavg mode and the steam pressure mode of control were individually tested. In the steam pressure mode, reactor power was raised from approximately 2 percent to approximately 10 percent to verify that the steam dump system would adjust valve position to maintain the steam header pressure at the set value.

In the Tavg mode, two individual tests were performed; one to verify proper plant trip controller response, and one to verify load rejection controller response. The plant trip controller response test was performed by simulating a P-4 reactor trip signal aM then adjusting Tavg by changing reactor power to verify proper automatic steam dump operation. The load rejection controller response test was performed by simulating a sudden loss of load at approximately 6 percent reactor power and verifying proper automatic. steam dep operation. During the testing, steam dump system parameters were monitored using strip chart recorders.

~

4.0-17 l d

-The following acceptance criteria were satisfied during the testing:

1)- The plant trip gontroller. (AB-E-500D) responded properly to maintain -

Tavg at 562 + 2 F at approximately 6% power. After steady state power was achieved there were no divergent oscillations in temperature,

' 2) The load rejection contrgiler (AB-K-500A) responded properly to maintain Tavg at 561-1 2 F at.approximately 6% power. After steady state power was achieved, there were no divergent oscillations in-tm perature, 3)- The steam header pressure controller . (AB-PK-507) responded properly to .

maintain pressure.at the normal no-load pressure of 1992 1 25 psig.-

No control systen adjustments were required _as a result of this test. The test was cmpleted on June. ll,1985, after several days-delay while adjustments were made to the automatic steam generator level control system.

4.0-18

4.2.2 AU10RTIC REACTOR 00tf1 POL To control ICS average temperature (Tavg) as power is increased or decreased a reference tenperature, Tref, corresponding to turbine load as indicated by first stage turbine impulse pressure, is provided to the reactor control systen. Tref is then compared to the actual Tavg in the RCS loops. The auctioneered high Tavg from .the four loop Tavg's- fs used as the basis for the comparison. If a difference of more than 1.5 E exists, the rod control system, when in the automatic mode, will insert or withdraw the control rods to align Tavg with Tref. The purpose of.this test was to verify initial satisfactory operation of the system in automatic. Additional testing to perform adjust.nents on the systen is discussed in Section 4.4.5.

To verify system operation, Tavg was increased by approximately 6 0F by withdrawing the control rods in MANUAL. Rod control was then placed in AUIO and various parameters were monitored on strip ch rt recorders as the plant responded. The process was then repeated for a 6 decrease in Tavg by inserting the control rods in MANUAL and then placing the rod control systen in AUIO.

This test was performed with the plant at steady state 30% power conditions on June 27, 1985. For both the increase and decrease in Tavg, control rods innediately started to move when the rod control systen was placed in AUTO.

Tavg was within 1.50 F of Tref in less than one minute in both cases. The following acceptance criteria were satisfied:

1) No manual intervention was required to bring plant conditions to equilibrium values following the initiation of the transients, 0

2) Tavg returned to within + l.5 F of Tref following the initiation of the transients, O

3) . Tenperature variations were less than S F peak-to-peak and temperature oscillations had a period of greater than one minute following the initiation of the transient, Whegthedatawasanalyzedsligh$temperatureoscillationsoflessthan 0.5 F with a period of approximately twenty seconds were noted both before and after the initiation of the transient. However, there was no tenperature oscillation caused by the reactor control systen during the transient, therefore, the period wm infinity and acceptance criterion 3) above was satisfied. Westinghouse concurred that the results were satisfactory.

4.0-19 m:We -

4.2.3 AIFICMATIC STEAM QNERA'IOR LEVEL 00tffROL TEST This series of tests was performed to verify the satisfactory operation of the various camponents of the automatic steam generator level control system at steady state conditions as well as under increasingly severe plant transients. The head capacity curves of both steam driven main feedwater pumps were also verified.

Initial testing was performed at approximately 10% reactor power to verify the ability of the bypass feedwater control valves to control steam generator level at zero electrical load in automatic. Level offsets were introduced in each steam generator and then the bynass feedwater control valves were placed in automatic and allowed to restore level. The nuclear feed forward signal in the bypass valve control circuitry was checked by performing 3% power decreases and increases and monitoring valve performance.

The next test was performed in the 10% - 19% power range to verify the initial satisfactory operation of the main feedwater control valves. The bypass valves were closed, the main feedwater control valves were in manual, the feedwater pmps were in auto and the pmp speed controls were in auto.

A level offset was introduced in each steam generator and the applicable i control valve was placed in automatic and allowed to restore level. 3% and 1% power changes were also introduced with the main feedaater control valves in automatic. Overall operation was satisfactory. Slight setting changes were made in the controls for the C steam generator to reduce coupling with the B steam generator. Two defective controller cards were replaced in the pamp speed control system.

Testing at 30% power included verification of the operating characteristics of the steam driven main fes3 Water pumps. Both pumps exceeded predicted performance. Main feedpump speed control performed satisfactorily under these conditions. A ten percent power decrease was performed and then a ten percent power increase. Overall operation of the steam generator level control systen was satisfactory although levels in the B and D steam generators went below 40% briefly at the beginning of. the transient.

Pump performance was again verified to be satisfactory at 50% power. Steam generator level control was monitored during a 25% power decrease and then a 15% power increase. The power increase was terminated at 40% due to a xenon transient. As a result of these transients adjustments were made to the main feedwater valve controllers ard the master pump speed controller.

Also, the feedwater pump delta P/ speed controller did not achieve the desired setpoint at 25% power for the differential between feediater header pressure and steam header pressure. The steam flow indication lu;ps were realigned.

At 75% power, steam generator level control was nonitored during the large load reduction test (Section 4.3.2) . The overall response of the control systems was-satisfactory during the transient, although levels went outside the 40-60% band and were outside the 45-55% band after 5 minutes. As noted previously at 25% power, the feedwater pump delta P/ speed controller was still not operating satisfactorily at 25% power thus requiring further adjustment.

4.0-20

During 100% power testing, data was' collected during the 10% load swing test

-(Section 4.3.1) . Control systen performance was satisfactory. The control systens were then monitored during the large load reduction test (Section 4.3.2) . Level did go outside the 40% -60% band with a low level of 34.5% in the "D" steam generator. . After review of the data, Westinghouse determined that response was satisfactory since no levels had been outside of initial level +15%.

The feedwater punp delta P/ speed controller performed satisfactorily during the transients.

In sumnary, the steam generator water level controls were satisfactorily adjusted to handle normal operation as well as major oscillations. The main feedwater punps head and capacity curves were better than expected.

. 4.0-21

4.3 TRANSIENT ABO TRIP ~ TESTS Plant response to transients of varying magnitude was determined during the power. ascension phase of the startup program. These tests were utilized to analyze the overall behavior of the major plant control systems during a power swing.or an actual plant trip. The need for further control system adjustments was determined by monitoring various plant parameters before, during and after each transient.

The specific tests performed consisted of a series of 10% power load swings at 30%, 75% and 100% power, a 50% load reduction at 75% and 100% power, a trip to determine ability to shutdown and maintain hot standby external to the control room, .a trip at 50% power to verify the ability of the nuclear instruments to detect dropped rods, and a unit trip at 100% power.

4.0-22

4.3.1 IDAD SWING TESTS The purpose of the load swing tests was to verify the proper nuclear plant transient' response, including autcxnatic control system performance, when 10%

. load changes, both decrease and increase, were introduced at the turbine

' generator. . Tests were performed at the 30%, 75% and 100% power test plateaus.

The tests were started with stable plant conditions at the desired test plateau and the following systens in automatic:

1) Steam Generator Main Feedwater Control,

2) Pressurizer Pressure Control, i [. .

~ 3) Pressurizer Heater Groups A and B,

4) Pressurizer Heater Group C in CLOSE,

5) " Pressurizer Level Control,

6) Steam Dump Control in Tavg Control Mode,

7) ! bin Feedwater Pump Turbine Speed Control.

After initial plant data was collected to verify plant stability, the electro-hydraulic controller (EHC) was used to achieve a 10% load decrease as rapidly as possible. When the plant was in a stable condition additional data.was collected.. Curing the transient certain parameters were monitored on multichannel stripichart recorders.

~

- 2 1he DC control 1er was then used' to increase the plant output as rapidly as possible to achiehe 'a 10% load increase and attain a final plant level at' approximately the original test plateau. After the pl.at was in a stable condition, a final set of data was collected.

The acceptance criteria for these tests were:

1) No reactor trip was generated,

2) No turbine trip was generated,

3) Safety injection was not initiated,

4) Neither the steam generator relief valves nor safety valves lifted,

5) Neither the pressurizer relief valves nor safety valves lifted,

6) No manual intervention was required to bring the plant conditions to steedy stato,

3) Nuclear power overshoot was less than 3% for load increase,

8) Nuclear power undershoot was less than 3% for load decrease.

The data frcm the three test plateaus is sumnarized in Tables 4.3.1-1 through 4.3.1-6.

4.0-23

/

TABLE 4.3.1 LOAD SWING FROM 30% 'IO 20% PONER Parameter Initial During Transient Final Minimum Maximum

- Plant Operating Level (Mie-Gross) 290 - - 198 Nuclear Power (%) - m 32- 20 32 22-Tavg - auctioneered (F) 566 561.5 -566 562 Tref ("F)- 566- -- - 563 Delta T - Loop 1 (%) 38, 25 38 27-Overpower Delta T Setpoint(%) 108- . - . -. 108 Overts perature Delta T Setpoint (%) 140- - - 147 Pressurizer Pressure-(psig) 2230 2219- -2283 2230 Pressurizer Level (%)- 35 29 36 30 Steam Header Pressure (psig)- 1030 1024 (1) 1068 (1) 1030 Steam Flow (lbpr x 10)

Loop 1 1.1 - -

0.7 Loop 2 1.15- - - 0.8 Loop 3 1.'2 - - 0.85 Loop 4- 1.0 - -

0.55 Narrow Range Steam Gerierator Level (%)

Loop 1 49 38 57 48 Loop 2 50 40 58 50 Loop 3 49 40 58 50 Loop 4, 49 39- 56 49 Fes3 water Teperature (F)

Loop 1 347.7 - - 316.8 Loop 2 347.8 - -

317.3 Loop 3 347.8 - -

317.0 Loop 4 347.2 - - 316.1 Feedwater (lbg/hr x 10Figw) Loop 1 1.0 - -

0.4 Loop 2 0.9 - -

0.4 Loop 3 1.0 - -

0.45 Loop'4 1.0 .-

0 0.5 Feedwater Punp Discharge (1) (1)

Pressure, psig 1120 1964 1155 1110 Feed Peo A Speed (RPM) 3900 3870 4200 3600 Feed Pump B Speed (RPM) . 1000 -(l) - (1) - 1000 Control Bank D Position (stms) 194/195 .- -

146/145 (1) Frm test recorders Time to reach equilibrium following load change: 4 minutes.

4.0-24

J '

jy j)/ .g . y 0O ,

' TABLE 4.3.1-2 f LOAD SWING FROM 20% 'ID 30% POWER

c Parameter Initial During Transient Final Mininum Maximum

g v Plant Operating Level Ji' (!*Te-Gross) -- 190 - - 290

, 3 Nuclear Power -(%) m 22- 22 32.5- 32 Tavg - auctioneered :(F) 562' 559 565.5 565.5 Tref ^  : (F) 563- . - 565.5 Delta T . Loop 1- (%) 27- 27> 38.5 38 Overpower < Delta T Setpoint(%) 108- . - 109 7

Overtemperature Delta T Setpointi -(%) 147- .. - 138 Presrurizer Pressure- (psig)' 2230- 2215 . 2275 2225 Pressurizer Level (%)' 30 26.5 36.5 34 Steam Header Pressure (psig) 1928 1030 -983 (1) 1930 (1) ,

if;t Steam Flow (lbg. x 10')

1.1 t;(._ Loop 1 0.7 -

~-

Loop 2 0.8 -- -. 1.15

'^3 ,' V.

. Loop 3 0.85 ~- -

1.2

'y - Loop 4 0.55 . - - )- -

1.0

+ Narrow Range Stean Generator Level (%)

.4B/

Loop 1 41 57.5 48

,' Loop 2- 50J 41.5 61 50 Loop 3- 50 40.5 58- 49

' Loop 4s 49, 43 .

59.5 49 Feslwater Tenperature (F) o# Loop 1- 316.8 " ,

- 347.9 W Loop 2 317.3 ,.

'348.0 Loop 3. '317.0 ,- -

-348.4 Loop 4 316.14 . - - -

347.6 FeedwaterFlgw '- * '

0.4 O.95

-(lb[nr x.10 ) . .Loop 1- -

.g , Loop 2 -0.4. - - 0.90

^- ;_

't. Loop 3. 0.45 - - :1.0 3 .: .

Loop 4 0.5- - - 1.0 Feedwater Ptrnp Discharge - . (1) (1) >

'1110 1921 :1196 1120-f .

> Pressure, psig- .

- U Feed Ptanp A Speed (RPM) 3600 .(1) 3720 -(1) 3785 3900 Feed Pump B Speed (RPM). 1000 -(l).1000 -(l) 1000 1000 Control Bank D Position

-(steps)' 146/145-

-- 194

.. J P ,1 '. I

, ,"1  :(1)'FroN1 test recorders 3

. Time!.to reach equilibrium following load change: 3 minutes.

(x>:  :.

~ ,

t jf.

n y 4.0 ' yQ_'-

^

Qs.n 5C f t 5 Ni-F

^ '

.1'

, . 4 ~.. t. . < . - . _ , , , - _ _ . ~ . . . . . . . ~,- . . - . .. . - -

TABLE 4.3.1-3 LOAD SWING FROM 75% 'IO 65% POWER Parameter Initial During Transient Final Minimum Maximum Plant Operating Level (We-Gross) 775 - - 650 Nuclear Power- (%) m 75.5 - 60 75.5 62.5 Tavg - auctioneered (%F) 579 575 581 576 Tref -("F) 581- - - 578 Delta T - Loop 1 (%) 75 64 76 67 Overpower Delta T Setpoint(%) 109- - - 109 Overtsperature Delta T Setpoint (%) 122 - - 128 Pressurizer Pressure (psig) 2250 2219 2288 2230 Pressurizer Level, (%) 54 45 56 46 Steam Header Pressure (psig) 1000 1014 (1) 1066 (1) 1000 Steam Flow (lbpr x 10")

Loop 1 2.7 - -

2.3 Loop 2 2.5 - -

2.2 Loop 3 2.7 - -

2.3 Loop 4 2.5 -- -

2.3 Marrow Range Steam Generator Level (%)

Loop 1 48 42 53 48 Loop 2 49 42 53 50 Loop 3 49 42 55 50~

Loop 4, 49- 42 52 48

.- Feedwater Teperature ("F)

Loop 1 416.6 - -

403.3 Loop 2 416.9 - -

403.6 Loop 3 416.8 - -

403.7 Loop 4 416.1 - -

402.8 FeedwaterFlgw)

(lby x 10 Loop'1 2.7 - -

2.3 Loop 2 2.7 - -

2. 2.

Loop 3 - 2. 8 - -

2.3 Loop 4 2.7- - -

2.3 Feedwater Pmp Discharge (1) (1)

Pressure, osig 1180- 1188 1204 1140 Feed Pep A Speed (RPM) 4400 (1) 4200 (1) 4608 4100 Feed Pump B Speed (RPM) 4600 -(l) 4392 (1) 4824 4300 Control Bank D Position (steps) 199/198 - -

146/145

- (1) Frcm test recorders Time to reach equilibrium following load change: 5 1/2 minutes.

?

4.0-26

TABr2 4.3.1 IDAD SWING FROM 65% 'IO 75% POWER Parameter Initial During Transient _

Final Minimum Maximum Plant Operating Level (Mie-Gross) 660 - - 760 Nuclear Power (%). . 62.5 62- 74.5 74.5 Tavg - auctioneered -(!F) 576 575 .

580 580 Tref- - (F ) 578 .. - 582 Delta T - Loop 1 (%)- 66 66 76 75 Overpower Delta T.Setpoint(%) 109 - - 109 Overts perature Delta T Setpoint (%) 117 - - 121 Pressurizer Pressure (psig) 2238 2220 2275 2235 Pressurizer Level (%) 47 44 -

53 53 Steam Header Pressure (psig) 1010 975 (1) 1014 (1) 1010 Steam Flow (lbpr x 10).

Loop 1 2.3 - -

2.85 Loop 2 2.25 - -

2.8 Loop 3 -

2.'45 - -

2.85 Loop 4 2.35- - - 2.8 Narrow Range Steam Geerator Level (%)

Loop 1 148 42 55 48 Loop 2 47 43 58 50 Loop 3 48 41 57 49 Loop 4 m '48 43 55 49 Feedwater T e perature ("F)

Loop 1 402.7 - -

415.1 Loop 2 403.0 - -

415.2 Loop 3' 403.0 - -

415.5 Loop 4 402.1 - -

414.5 FeedwaterFlgw)

(1by x 10 Loop 1 2.4 - -

2.8 Lcop 2 2.4 - -

2.8

. Loop 3 2.45 - -

2.85 Loop 4- 2.4 - -

2.3 Feedwater Pump Discharge (1) (1).

Pressure, osig 1150 1977 1116 1160 Feed Pump A Speed (RPM) 4100 (1)-4032 (1) 4392 4400 Feed Pump B Speed (RPM 4300 (1) 4104 (1) 4536 4500 Control Bank D Position (steos) 148 - -

199 ,

- (1) From test recorders Time to reach equilibrium following load change: 5 minutes.

~4.0-27

_.m ~

TABLE 4.3.1-5 LOAD SWING FROM 100% to 90% POWER

-Parameter' Initial During Transient Final Minimum Maximum Plant Operating Level (me-Gross) -1950 - - 930 Nuclear Power .(%) - .m - 99.8 87.5- 99.8 87.5 Tavg - auctioneered-(;F)- 587.5 585 587.5 585 Tref -("F) 588.5- -- -

585 Delta T.- Loop 1 (%). 101 90.5 101 90.5 Overpower Delta T Setpoint(%) 108- - - 108 Overtenperature Delta T Setpoint (%) 112 - - 115 Pressurizer Pressure (psig) 2230 2210 2300 2220 Pressurizer Level. (%)- 61 57- 65 57 Steam Header Pressure (psig)- -1000 1000 -(l) 1040 (1) 1010 Steam Flow (lbpr x 10')

Loop 1 2.99 - -

2.61 Loop 2 2.91 - -

2.53 Loop 3 2.95 - -

2.61 Loop 4 2.95 - - 2.57 Narrow Range Steam Generator Level (%)

Loop 1 .48 42 50 48 Loop 2 50 44 53 50

. Loop-3 48 45 53 49 Loop 4m- 49 46 52 49 Fes3 water Tenperature ("F)

Loop 1 439.1 - -

429.5 Loop 2 439.3 - -

429.7 Loop 3~ 439.5 - -

429.6 Loop 4 438.6 - -

428.7 FeedwaterFlgw)

(l p x 10 Lorp 1 2.99 - -

2.65 Loop 2 2.87 - -

2.49 Loop 3 2.95 - -

2.57 Loop 4 2.95 - -

2.53 Feedwater Punp Discharge (1)- (1)

Pressure, psig. -1200 < 1188 1240 1190 Feed Ptznp A Speed (RPM) -5000 (1) 4714 (1) 5143 4700-Feed Pump B Speed (RPM) 5200 .(1) 5000 (1) 5429 4900 Control Bank D Position (steps) 222 - -

162 (1) _ Fran test recorders Time to reach equilibrium following load change: 6 1/2 minutes.

4.0-28

TABLE 4.3.1-6 IDAD SWING FROM 90% to 100% POWER Parameter Initial During Transient Final Minimum Maximum Plant Operating Iavel (tWe-Gross) 935 - - 1910 Nuclear Power (%) m 89 89 93 93 Tavg - auctioneered (F) 585.5 581.5 582 582 Tref (F) 585 - - 587 Delta T - Loop 1 (%) 91 91 98 98 Overpower Delta T Setpoint(%) 109 - - 109 Overtenperature Delta T Setpoint (%) 116 - - 120 Pressurizer Pressure (psig) 2245 2220 2250 2250 Pressurizer Level (%) 59 42 59 54 Steam Header Pressure (psig) 1010 962 (1) 1014 (1) 962 Steam Flow (lbpr x 10)

Loop 1' 2.65 - -

2.91 Loop 2 2.57 - -

2.84 Loop 3 2.65 - - 2.84 Loop 4 2.61 - -

2.80 Narrow Range Steam Generator Level (%)

Loop 1 48 46 54 48 Loop 2 50 47 58 50 Loop 3 50 45 56 50 Loop 4, 49 47 55 49 Feedwater Tenperature (F)

Loop 1 429.7 - - 436.4 Loop 2 429.9 - - 436.8 Loop 3- 429.9 - - 436.5 Loop 4 429.2 - - 436.0 FeedwaterFlgw)

(1bg /hr x 10 Loop 1 2.65 - - 2.91 Loop 2 2.53 - - 2.80 Loop 3 2.57 - - 2.84 Loop 4 2.57 - - 2.84 Feedwater Pump Discharge (1) (1)

Pressure, psig 1200 1123 1188 1155 Feed Pump A Speed (RPM) -4750 (1) 4571 (1) 4714 4800 Feed Pump B Speed (RPM) 4950 (1) 4857 (1) 5143 5000 Control Bank D Position (steps) 194 - -

223 (1) From test recorders Time to reach equilibrium following load change: 6 1/2 minutes.

4.g-29

t

% e test at 30% power was' performed on June 29, 1985. All of the acceptance criteria as outlined above were satisfied. %ere were several parameters that went'outside their expected baM:

1) Maximurn pressurizer pressure was 53 psig above initial pressure (<50 psig expected),

2) Steam. generator levels were expected to be within + 10% of initial levels but steam generator A dropped 11% on the load decrease and steam generator B increased 11% on the load increase,

3) Tavg was not expected to overshoot (undershoot) itsffnalvalueon load increase- (decrease) but on the load decrease 0.5 F undershoot was noted.

Wese slight deviations from expected valves were reviewed by Westinghouse and determined to be acceptable.

W e test at 75% power was performed on July 28, 1985. All of the acceptance criteria as previously outlined.were satisfied. During the initial 10%

power decrease, the operators had to take manual control of feedwater pump speed control to maintain steam generator feed flow. After adjustments to the gain on the feed flow / steam flow mismatch cards (FY510D - FY540D), the

-10% power decrease aM the 10% power increase were performed without further difficulty. As for the-30% power test, several parameters, including steam pressure overshoot /undershoot, steam generator level swing and Tavg undershoot were outside the Westinghouse expected range by a slight amount but this was determined not.to impact the test results and no additional control system adjustments were required.

The test at 100% power was performed on August 22, 1985. All of the acceptance criteria as previously outlined were satisfied. We power increase was actually closer to 7% than 10%. Rods had to be withdrawn somewhat prior to the power increase to bring axial flux difference within Technical Specifications limits. This left insufficient rod worth to complete the 10% power increase. A review of the test _ data showed that all systens performed satisfactorily and that there was no need for further testing. As noted in the previous load swing tests, some parameters were slightly outside the Westinghouse expected values including RCS pressure swing and steam pressure overshoot. %ese were reviesed by Westinghouse and determined not to impact the test results. No control systen adjustments

~

were required.

4.0-30

4.3.2 - IARGE IDAD REDUCTION TESTS The large load reduction tests were performed during the 75 percent and 100 percent power testing plateaus to verify the ability of the primary plant, secondary plant and the automatic reactor control systems to sustain a 50 percent step load reduction. The data obtained was used to evaluate the

~

interaction between the control systens and to determine if any setp-int or gain adjustments were necessary.

During each load reduction test, selected plant parameters were trended on multi-channel strip chart records. Stable operation was verified with rod control, steam generator MEW control, pressurizer pressure and spray

. control, pressurizer level control and feedwater puup turbine speed control systens in automatic. Steam dump control was in the Tavg mode. Using the standby load set, generator load was reduced by 50 percent within approximately 1 minute with all test recorders in high speed. No manual intervention with the control systens was allowed during ard following the 50 percent load reduction. The plant t;as allowed to stabilize with control systens in automatic.

The large load reduction test at 75 percent power was performed on July 28, 1985. Without manual intervention, the automatic control systens did sustain the load reduction and allow the plant to be returned to stable conditions. All acceptance criteria were satisfied:

1) The reactor did not trip,

2) The turbine did not trip,

3) Safety injection did not initiate,

4) . Steam generator safety valves did not lift,

5) Pressurizer safety valves did not lift,

6) No manual intervention was required to reach equilibrium plant conditions.

'The actual load reduction was 627 W e (54.5%). As a result, the steam dump valves did not shut until after 9 minutes, 40 segonds after start of trgnsient (expected 8 minutes) and Tavg peaked 6 F above its initial value (5 F expected). A Westinghouse evaluation determined that both these values were consistent with a 54.5% power decrease. The test data is sumnarized in Table 4.3.2-1. No control systen setpoints or gains were changed as a result of this test.

The large load reduction test at 100 percent power was performed on August 29,1985. - Without manual intervention, the plant systems reached equilibrium conditions. All the major acceptance criteria were met as discussed above for the 75% power test. No control systen setpoints or gains were changed as a result of this test. The test data is sumnarized in Table 4.3.2-2.

4.0-31

TABLE 4.3.2-1 IARGE LOAD REDUCTION TEST FROM 75% POWER Parameter Initial During Transient Final Minimum Maximum Nuclear Power (%) m 75.5 20 75 26 Tavg- - auctioneered (F) 580 562 586- 564 Tref .(YF) >581 - - 565 Delta .T .(%) - 77 26- 77- 32 Overpower Delta Setpoint -(%) 109 . -

109 Over temperature Delta T

'Setpoint (%)- 121 - - 144 Pressurizer Pressure (psig) 2240 '

2140 2320 2240 Pressurizer Level . (%) 52 30 59 31 Steam Header Pressure (psig) - 1014 1014 -1118 1940 Steam Flow (lbpr. x 10)

Loop 1 2.75 - -

0.9 Loop 2 2.6 - -

1.0 Loop 3 2.8 - -

1.0 Loop 4- 2.7 - -

1.0 Narrow Range Steam Generator ' Level (%)

Loop 1 48 31 62 47 Loop 2 50 33 66 50 Loop 3 50 31 66.5 50 Loop A 50 34.5 64.5 49 Feedwater Tenperature(F)

Loop 1 416.4 - -

333.2 Loop 2 416.8 - -

333.4 Loop 3 416.6 - -

333.8 Loop 4 415.8 - - 332.9 Feedwater Flgw) Loop 1 (1bg /hr x 10 2.75 - -

0.8 Loop 2 2.75 - -

0.45 Loop 3 2.8 - -

0.65 Loop 4 2.7 - -

0.8 Feedwater Pump Discharge Pressure, psig 1162 1136 1240 1136 Control-Bank D Position (steps) 181 - -

95/94 Time to reach equilibrium following load change: 36 minutes.

4.0-32

. . . - - - - .~ - - . . .. . . - . . -. . . . . .

1 4 5 E k"

. . TABLE 4.3.2-2:

LARGE IDAD REDUCTION TEST FROM 100%' POWER

,. 1 -

4 4

Parameter Initial During Trmsient Final

- - - Minimtun Maximum-

~

Nuclear Power . (%) . m -99.5 53  ! 99.5 - 53 '

+

Tavg - auctioneered-(~F) -

588- - 573- 588- 573 Tref. -(F) - -588- - - - .

573 Delta.T- -(%)- - - 100. . -

.-- 62 ,

100 62 herpower Delta T Setpoint(%) 108 - -- ..

108 a- Over. temperature Delta T ,

Setpoint- -(%)- 110 -- - -

r-- 131 Pressurizer Pressure (psig). . 2235- --2155- ~ 2340- 2255 Pressurizer Level (%). -62 44 62 44 Steam Header Pressure (psig) - 1920 - 1920 -1118 1927 Stean Flow-( y r x 10")

Loop 1 3.8 - * ' - 2.1

. Loop 2 -3.6 - -

2.0 Loop 3.- 3.7 -  :- 2.2

( Loop 4 3.6 -

- 2.1 l Narrow Range Steam Generator. Level (%) ,

Loop 1 48 3115 56 .48 Loop 2 50- 33 57- 50 ,

• -Loop 3 49 33 . 58 50-Loop A 49 32.5 58 50 Feedwater Temperature ("F) .,;

Loop 1 438.2 - -

392.0 ,

Loop 2 ' 438.6 - -

392.3 '

Loop 3 438.5 - -

392.2

. Loop 4 '437.6- - -

391.3 +

l Fes3waterFlp)

(lbM/hr x 10 - Loop 1 ,

3.8 -' -

2.1  ;

{ .. Loop 2 3.6 - - -

1.9

. Loop 3 3.7 - -

2.0 ,

Loop 4 3.7 - -

2.0 g Feedwater Punp Discharge . .

Pressure, :sig J 1188' 1162 1331 1162 ,

Control Bant D Position (steps) 217 - - --

97 : -

Time to reach equilibrium following load change: Approximately 12 minutes L_, (based on steam dump denand trace).

l l

! 4.0-33 l s e , - - - , ,y-- ..-.,r. = r.w.e,,,,.,w,_,-,,, . - . ,..,-w.yy-, , ,- ~ ,m . , . . - - - - . .w,r.,_,_ , , , , , e-,-_,, -,w_,.-..-,x,--f

4.3.3 SHUIDOOM AIO MRIlffENANCE OF H0f STADOBY EXTERNAL '!O

'tHE CONTROL ROCM The purpose of thir test was to denonstrate that, using plant operating procedures (OFN Control Room Not Habitable), the plant can be taken from > 10 percent power to hot standby conditions and then maintained in hot standby for at least 30 minutes with a minimum shift crew using controls and instrumentation external to the Control Room. The minimum shift crew was that specified in OFN-13..

'1his test was performed on " June 29, 1985. During the evacuation of the Control Room by. the minimum shift crew, a standby crew remained in the Control Room to nonitor plant conditions. The standby crew was to take no actions during the transient. The reactor was tripped at 1111 by manually tripping the reactor trip breakers from the Control Room. The minimum shift crew then evacuated the Control Rocrn to take control of the plant at the Auxiliary Shutdown Panel (ASP) and other duty stations as outlined by OFN-13 and as directed by the Shift Supervisor. Hot standby conditions were established at 1147 and maintained until 1220.

l During this period, pressurizer pressure, pressurizer level, steam generator level and RCS tenperature were maintained from outside the Control Room.

The acceptance criteria of the procedure were met.

~

4.0-34

4.3.4 RODS DROP JWO PUNT TRIP This test was performed at the end of the 50 percent power test plateau to denonstrate operation of the negative rate trip circuitry by dropping two

. ROCA's from a comon rod group and to review plant response and control systems behavior to a plant trip fra an intermediate power level prior to the plant trip test from 100 percent power.

During the performance of the test, a high speed chart recorder was used to monitor the state of the reactor trip breakers, the rod-on-bottom alarms, nuclear instrumentation systen (NIS) power and negative rate trip bistable output. While at steady state plant conditions (50 + 5% power), the two "

control rods (D-4 and M-12) in group 1 of control bank D (CBD) were transferred to the DC hold bus. The drop of the two rods was initiated by removing the stationary gripper coil fuses for D-4 and M-12 in power cabinet 1AC and then deenergizing the DC hold bus.

The trip test was performed July 16, 1985 with satisfactory results. The two dropped rods did cause a reactor trip due to power range high negative rate and all rods dropped normally, The pressurizer safety valves did not lift, the steam generator safety valves did not lift, and safety injection was not initated, thus satisfying all the acceptance criteria for the test.

In addition, the reactor trip generat.ed a turbine trip. The steam dumps operated to reject heat to the condenser and feed flow, steam flow, and steam generator narrow range levels all went to zero as indicated on the process instrumentation.

The traces on the high speed chart recorder did not show a change of state for the NIS negative rate bistables because the recorder was set up with a 5 to' 120 VAC range. Investigation determined the NIS negative rate bistables tripped at 9 VAC therefore the event recorder bistable in the recorder did not show a change of state. However, the negative rate bistable trip was shown by the indicator lights on the front panel of the NIS negative rate trip drawer and by the first-out indicators on the main control board. Both of these had to be reset prior to subsequent startup.

Table 4.3.4-1 sumarizes the data from this test.

e 4.0-35

z.

TABLE 4.3.4-1

. RODS DROP AIO PIANT TRIP TEST DATA SU!HARY

?'

Plant .

.Before During Transient After Parameter -Trip Minimum- Maximum Trip Tavg Auctioneered, 0,# 568.5 555.0 568.5- 554.5 Tref ,T 571.5 -- -

558.9 Delta T - Loop 1 , t 55 2 55 .2

~

Overpower Delta T -

-Setpoint"- Loop 1, % 134 - - >150 Overtemperature Delta T Setpoint - Loop'1, t -

109 - -

108 Pressurizer.

Pressure, psig- 2225 2125 2245 2245 Pressurizer-Level,'%- 24 38- 25 -

38 ~

Narrow Range,

. - Steam Generator

/ ,

Invel - Loop-1, % 49 0 53 -14 3

0 S

G

[-,

T 4

4.0-36

~ f ,

y 1 a a f 4 .

4.3.5 PUWF TRIP FIDS laf Pusrumfr PONER

~ ..

y We' purpose of this. test was:

? 1). To verify the ability' of the primary and secondary plant and the plant automatic control systens to sustain a trip fra 100 percent

, power and;to bring the plant to stable conditions following.the

~

itransient,

. 2) To determine the overall response time of the reactor coolant hot leg resistance tenperature detectors and,

. 3) To evaluate the data resulting from this test to determine if changes

. in the control systen setpoints are warranted to improve transient response based on actu'al plant operation.

For the performance of this test, the plant was operated at 100 percent power with the following control systems in automatic:

' 1) Reactor Rod Control,

2) Steam Generator Main Feedwater Control,

[ 3) Pressurizer Pressure and Spray Control, 4)- Pressurizer Heater Groups A and B,.

. 5). Pressurizer Heater Group C in close position,.

6) , Pressurizer Level Control,

' 7) Steam Dep Control in Tavg control mode,

. '8)- Main.Feedwater Pump Turbine Controls.

' All shutdown and control rod banks were fully withdrawn'except control bank D which was positioned to maintain axial flux difference within the limits specified in E GS Technical Specifications.

.: 7 ,

- Various plant. parameters were input to strip chart recorders to monitor the-plant performance during the trip. 'After initial data was recorded and the.

test' recorders.were operational, the plant trip was' initiated by nomentarily

~

4 .

jumpering two terminals of'the Turbine Sequential Trip relay.(AR in Cabinet ~ ,

' MA 104B)'Which opened both Main Generator Output Breakers.

.After-the trip, the plant was restored to a stable condition using normal

.f, _ plant. operating procedures.

i, . Se plant was tripped from 100% power at 0512 on August 28, 1985. Using the opening of Generator' Output Breaker #2 as time zero, the turbine tripped at 0.375 seconds and the reactor trip breakers opened at 0.155 seconds. We

. steam dep valve demand signal went .to -100% (open) at 0.225 seconds ani then

~'

- . modulated' to 0%: (closed) at 46.885 seconds. Se steam dmp valves were.

s place!'in pressure' control mode at 0517. Feedwater isolation signal was received at 0514..

. We' following' major acceptance criteria ~of the test were met during the trip:

1) :All control rods dropped, 4.0-37

', , r p

,M ' 2)) Reactor' Coolant Systen Pressure remained less than 2450 psig.

.'  :(maximum' pressure of 2269 psig wasi r,eached at 16.5 minutes during w .,

plant recovery),--

3); Steam Header Pressure remained less than 1150 psig-(maximum pressure

- Lof?l099 psig was reached at 5.965 seconds),

14): ; Safety. Injection was not , initiated,

y .

7 5) Using Ltheistrip chart recordings, the overall Hot Leg R1D response (times,were determined for-each loop as:

_ Loop 1. '6.999 seconds

- _ Inop 5.999 seconds

,, Loop 3- -

5.639 seconds l Loop 4 .-5.889 seconds

-All. response; times were less than the required 8.4 seconds.

The overall Hot Img RTD response time ~ includes response of the RfD nitself,'and was defined for the purpose of this' test as the interval

~

. of timeLmeasured between the point Were the neutron flux has decraaaari by 50 percent of its initial value to the point where the s

T - hot leg temperature signal _ (as measured by the RTD: output) has i ' decreased by a value equivalent.to 33 1/3 percent of the initial.

~ loop delta T value.in degrees Fahrenheit.  ;

E - The remaining major acceptance criterion required that, neutron flux drop- '

E- .below 15 percent power within 2 seconds of the last generator output breaker-

, opening. -Since'the pen for Generator Output' Breaker.I1. stopped inking when

- :the recorder was shifted to high speed :just prior to the trip, it was :

4 inpossible to determine the exact time this breaker. opened. .However, a

roview of the' recorder traces and plant performance as discussed above indicate that the generator output breakers opened at-essentially the sane time. Raaari on the opening time of generator output breaker $2, Naclear,' '

i Flux was below.15 percent at 1.325 seconds. .

In addition,' a -review of the' strip chart recording showed' that the plant

' control systens responded satisfactorily to this transient and no changes to

. control system setpoints were required to improve plant response.

Table 4.3.5-l'sunnarizes the data from this test. .

1

-9 4.0-38 e

i s a~.. i._..._.. _-.-m. . , - , - - - . - _ . _._,,.,.,,,,__,m.._._,--_._ -..-.-.-I

. . - . ~ . . . _ .

,' - ,  : p. -

, 4-j < , .,

4

. ;+ t - -

3 TABLE 4.3.5-1 s PU MP TRIP.FROM 193 PERCENT POWER ',

TEST DATA SUP9fARY 3-

.,- - Plant. Before During Transient After-Parameter Trip Minimum Maxinum ' Trip TavgAnctioneered,k- 588.2 555- 588.2 555 ,

Tref, 7 - -588.5 -.

- 557 s Delta T - Loop 1, % 99.5 99.5 4- ,

Overpower Del:a T e Setpoint - Icop - 1, ~ % ~ 198 -108 Overtemperature Delta 1 '

Setpoint - Loop 1, t 119 - - -146 Pressurizer 2255 Pressure, psig '2230 ~ 1999 2255 ^ ,

Pressurizer-Level, t 62 27.5 62 27.5 Narrow Range Steam _

Generator Level -

Loop 1, %: 48 9 48 16

)L *,,

B 0

~'

., i b

r:6 - 4 ~ s L

5 Y

b s 1

~e:.

r

- }

4

,4.0-39

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L 4.4 INSTIDENFATION CALIBRATION AIO ALIGMENT I

During the power ascension-test program, a series of tests was performed to calibrate and align various plant control and instrumentation systens. The first test discussed determined thermal power as well as collecting statepoint data at various steady state power levels. The data collected was then used for the calibration of the steam and feedwater flow instrumentation _and startup adjustments of the reactor control systen. In addition, testing was performed to c,perationally align the nuclear instrumentation syston and the process temperature instrumentation.

4.0-40

4.4.1 1HEIDEL PONER MASUREMDff AIE) STATEPOINF DATA COLLETION This series of tests had three purposes:

1)' To provide.a method for determining reactor thermal power,

2) To collect control and protection instrumentation calibration data at -

steady state power levels (statepoints) during the power ascension test program,

3) To determine ES flowrate using calorimetric data (50%, 75%, 90%,

100%).

After the initial test setup was verified during the post core loading precritical testing (section 2.7), data was collected at 30%, 40%, 50%, 75%,

90%,' 98% and 100% power. The control and protection instrumentation data served as an input to the calibration of the steam and feedwater flow

' instrumentation (section 4.4.2) and to the startup adjustments of the

. reactor control systen (section 4.4.5) .

Test instrummtation was installed to nonitor feedwater flow to each steam generator and steam pressure in each steam generator. To improve the potential accuracy of the test data, the range of the feedwater flow test instruments was increased as power was increased. The initial set of instruments were 0-5 psid while the final set of instruments was 0-50 psid.

Using the information from the test instruments as well as feedwater temperature from process instrumstation and the ASME Steam Tables, the enthalpy rise across each stean generator was calculated. Steam quality was assumed to be 1.0 (see section 4.5 for actual moisture carryover) and for all but the 100% power test steam generator blowdown was isolated. After correcting the heat input to the steam generators for the actual EP heat input as determinsi during the preoperational hot functional test, the actual percent of design power was determined.

RCS flow was determined using calorimetric data at 50%, 75%, 90%, 98% and 100% power. The flow calculation used the thermal heat output for each ES loop as determined by the calorimetric, T and T for each loop, the ethalples of the water in the ES hot lekand coMegs, and the specific volume of the cold leg water to determine the RCS loop flow. The total DCS flow was the sum of the loop flows. The data is sumnarized in Table 4.4.1-

1. RCS flow was greater than the minimum allowable flow at all test plateaus.

4.0-41

TABLE 4.4.1-1

,- RCS FIIM FROM CAIDRIMETRIC MEASUREMENT Ncuninal Power Invel Loop 50% 75% 90% 98% 100%

1 106,795 gpm 105,648 gpn 108,107 gpm 105,480 gpm 103,890 gpn 2- 103,069 gpn 106,374 gpm 105,484 gpn 101,621 gpm 99,210 gpn 3 105,036 gpn 103,521 gpn 103,813 gpn 103,481 gpm 100,460 gpn

-4 103,796 gpn 101,778 gpn 103,665 gpn 102,939 gpn 102,150 gpm Total 418,696 gpn 417,321 gpn 421,069 gpn 413,521 gpm 405,710 gpn W tance Criteria >382,800 gpm >388,542 gpm >388,542 gpm * >388,542 gpm

• ~ The . test at 98% power was performed for information only, therefore there was no acceptance criteria in the procedure.

4.0-42

F 4.4.2 cat.IBRATION OF STEAM AIO FEIDETER FIDW INSTR [ MENTATION j

The purpose of this test was to collect data during the power ascension testing to allow the calibration of the stem flow transmitters against feedwater flow. Test instrummtation was used to measure the feedwater flow elenent differential pressure and the flow calculations were performed as part of the thermal power measurenent (section .4.4.1) . In addition, the

- flow' data collected was empared to design values for stem flow and feed flow.

During the post core loading precritical testing, the static zero shift of the installed instrumentation was verified. One stem flow transmitter had to be replaced (Section 2.7). The remainder of the~ testing used data collected at 30%, 50%, 75% and 100% power plateaus during the thermal power meanurement test (Section 4.4.1) .

At 30% and 50% power, data was accumulated and it was verified that the steam flow /feedwater flow mismatch alarm did not actuate. After the 75%

data was analyzed, the steam flow transmitters were respanned and additional data was collected. Also, the Tref program was adjusted prior to the second set of data resulting in higher steam pressure and lower flows for a given power level (section 4.4.5) . The steam flow /feedwater flow mismatch did not actuate.

The 100% power data was collected with steam generator blowdown in service.

After correcting for the blowdown flow, it was determined 'that the following

. acceptance criteria were satisfied:

1) The steam flow /feedwater flow mismatch alarm did not actuate,

2) Steam flow indication on the main control board was within + 4.0% of feedwater flow indication on the main control board,

3) Plots of feedwater flow frcn the test instrumentation versus feedwater flow from installed plant instrumentation were within 12.5% of the ideal (design) curves,

4) Plots of feedwater flow from.the test instrumentation versus steam flow from installed plant instrumentation were within 13.0% of the ideal . (design) curves.

Although all acceptance criteria were satisfied, the 100% power data was analyzed to determine optimum spans for the steam flow transmitters. The spans were adjusted after the completion of the test program and additional

- stem flow and feedwater transmitter output data was collected at 100%

power .

4.0-43

4.4.3 OPERATIONAL ALIGIMENT OF IRECLEAR INSTIOENTATION The Nuclear Instrumentation 'Systen (NIS), because of its importance in monitoring reactor operation, was checked and calibrated throughout the startup test program. Earlier sections of this report have detailed NIS testing which was performed before initial fuel load, during post-core load testing and during initial criticality and low power physics testing. This section sumnarizes all testing and calibration done on the NIS throughout the test program ' including the additional testing done during power ascension testing.

A PRIOR 'IO FUEL IDAD Testing before fuel load was done in three separate phases as described in Section 1.0. First, an extensive preoperational-type functional test was performed on all the circuits, alarms, bistables and meters in all three ranges of the NIS. Second, the high voltage, discriminator voltage and preamplifier settings were determined and the source ranges calibrated accordingly. Finally, just hours prior to the movenent of the first fuel assenbly, analog channel operational tests (ACOT) surveillance procedures were performed on each source range channel.

POST CORE IDAD TESTING As detailed in Section 2.8, additional functional testing was done on all of the NIS utilizing channel calibration surveillance procedures.

INITIAL CRITICALITY Just prior to initial criticality, analog channel operational test (ACOT) surveillance procedures were performed on each intermediate and power range channel. After initial criticality, readings were taken to determine the amount of overlap between the source and intermediate range. This data is shown in Table 4.4.3-1.

POWER ASCENSION TESTING The power ascension testing had three main objectives:

1) To determine overlap between the intermediate and power ranges,

2) To plot power range detector currents versus reactor power,

3) To adjust power range indication to agree with secondary calorimetric calculations.

Overlap testing began at 0% power and continued through 100%. Readings were taken on both intermediate and power range channels to verify that there is at.least l'l/2 decades of overlap between the two ranges. The overlap data is shown in Table 4.4.3-2.

Power range detector current and reactor power readings were taken from 10% ,

to 100% power to verify detector linearity. Figures 4.4.3-1 through 4.4.3-4 depict curves plotting detector current against power and shows that all 4.0-44

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

m b

' TABLE 34.4.3-1

. NUCLEAR INSTRUMElffATION OVERIAP DATA SOURCE RANGE AND. INTERMEDIATE RANGE s s f

.First Positive ' Intermediate-

) -

Indication On Rangegndicates

~y t

~

mediate Range . (>10~gter- Amps) - 10 Amps

,  : SOURCE RANGE Channel N31 -c -

~. Main Control-Board 8 x 10 2- - 1 x 10 4

~ NI' Drawer-

• 1.25 x 10 6.5 x 10 3 Channel'N32 - -

. Main Control Board < 8 x.10 2 , 1 x yg 4 a'

2

' NI Drawer 1.25 x le 7 x 10 3 '

f

.IlffERMEDIATE RANGE Channel N35 - .-

1 x.10~

1 x 10 Main Control Board .

.NI Drawer: 1 x 19~11- 1 x 10 ~1

~

Channel N36 - -

Maib Control Board 1 x 10~11 'l x 10-1E' 1 x 197 1

~

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_ . , _ _ . . - , . _ ~ , - - . .m me-. -y, _ _ ,_

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TABIE 4.'4.3-2i NUCLEARLINSTRUMENTATION OVER[AP DATA - * .

INTERMEDIATE RANGE AIO POWER RANGE INTERMEDIATE RANGE .

9% 10% - 30%

40%'

i Channel.N35 Main Control Board c7-x 19 I 1.7 x 19 3.9 x 19 3.5 x 10 -

NI Drawer 6.5 x-10 'S x 19' - - 1.5 x 19 2.8 x 19 i

Channel N36 -

Main Control Board 7 x 19 -4.9'x 19' l.9 x 10 3.9 x-19 i

-6 1,7 x ig 4 NI Drawer 6 x 10 5 x 10-5 2.9 x 10 1

A h

u POWER RANGE

Channel'N41 Main Control' Board 1% 19% 34% 43%

I

~

NI Drawer :1.5% 19.1% 34% 42.5%

Channel N42 Main Control Board 1% 10% 36% 42%

NI Drawer ** 1.5% 11% 37% 42.5%

i i Channe'l N43

! Main Control Board 1% 11% 36% 44%-

NI Drawer 1.5% 19.8% 35.25% 42.75%

Channel.N44

, Main Control Board 9% 19% 35.5% 42%

l NI Drawer 1.5%- 11% '36% ~42.5%

4 i

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

q.:{ si I <

_9-

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i. 1 TABLE 4.4.3-2 :(CDNT) i NUCLEAR ~ INSTRUMENTATION OVER[AP DATA -

INTERMEDIATE RANGE A20 IOWER RANGE 9

4 4 =

j ' INTERMEDIATE RANGE i , ,

50% 75%- ' 90 % ' 199%

i Channel N35 ~

Main Control Board 3.5 x 19-4 4 .5 x.19 6 x:19 6'x 10 '

-4 4.5.x'-19

-4 5.5-x 10 -4 .

2 NI Drawer 3 x 19

- '6 x-19 i Channel N36 4 4 4 4 Main Control-Board 3.5 x 19 5 x-19 6 x.19 7 x 19 NI Drawer 3.2 x-19 - 5.1 x le ~6 x 19~4 7 x.19  ;

4 POWER RANGE r >

l s

o ' Channel N41 '

l -E Main Control Board . 49%. -75%- 92%'. 100%~

s

1 NI Drawer 48%- 75% 92% 199%

t l

4 Channel N42 .

! Main Control Board 49%- 75% 92%- 100% ,

i

\ NI Drawer 48% 75% -

92% 199% ,

i ,

j- Channel N43  !

! Main Control. Board 49%. 76% 93% 191% t t

i NI Drawer 48%- 75% 92% 100%

~

Channel N44 i _ Main Control Board , 49%. 75% 92% 199%

4 NI Drawer- 48%. 76% .

921- 199%  ;

{.

4 . .

j *NorrE: Overlap readings at 9%, 19% and 30% were taken before Power Range Gain was adjusted to match .

secondary calor.icmetric.  :

i .;

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-n - - , c . , , - , , - - - -. -, ,. , . . - .

p i;; ,

FIGURE 4.4.3-1 CHANNEL CURRENT VS. REACTOR POWER V CHANNEL N41 600 n

500

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REACTOR POWER (percent) 4.0-48 L. -. -- - - _ . - - - - . - - - - - - - - - _ _ _ _ _ _ _ .;

m . . - . -- . ,

' e

^

s; < FIGURE-4.4.3-2.

CHANNEL. CURRENT ^ VS. REACTOR POWER 1

CHANNEL N42 800

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' TOTAL ;'  ; ,

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40 50 60 70 80 90 100 0 10 20 30

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REACTOR POWER (percent) 4.0-49

.. = _.

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FIGURE 4.4.3-3 CHANNEL CURRENT VS. REACTOR POWER CHANNEL M43 n

' 600

t X

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0 0 10 20 30 40 Sg 80 7 80 90 gog enM:-

REACTOR POWER (percent) 4.0-50

<<-

ni,

'_Q . ,

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,w FIGURE 4.4.3 e. ,, .

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. CHANNEL CURRENT VS. REACTOR POWER .

' CHANNEL N44 Y

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N 4.0-51 .

6 94

' ' ' A

, , -.n.. ,_ - . . _ . , m. -

m-detectors. responded linearly.

-Beginning at 30% power, the power range channels were adjusted to match reactor power as calculated using a secondary calorimetric. This adjustment' is required by Technical Specifications to be done on a daily basis.

Therefore, normally the operations surveillanc.e procedure was used to make this adjustment. A-few times, however, the adjustments were made based on a

- calorimetric value from the Startup Test Program. Either method was acceptable and the test program was made more flexible by allowing either type of calorimetric to be used to make the power range adjustments.

When the plant reached 100% power, additional testing was performed to make

' final adjustments to the NIS. The high voltage settings on both the intermediate and power range detectors were verified acceptable by plotting a plateau curve. The intermediate range 25% reactor trip bistables were also reset using the actual 100% power intermediate range current readings.

After the plant was tripped from 100% power, a test was performed to set the intermediate range compensation voltage and verified the source range detectors had survived 100% power operations by completing a voltage plateau check on each source range detector.

o 4.0-52 La

'C ,

.q i

.e

.h

. . 1

, p '4.4.4 OPERATIGIRL ALIGBBFF OF PROCESS I

-}/r y,

3  %, TBWERA M tB I : u # TION r

y ,

14 1

- Dt ,

I I %e process instrumentation receives tanperature data from the primary l y coolant systen R1D's. This data is.used to generate delta T and Tavg values forluse by the plant control and protection systens. - .

Prior ~to initial criticality while at hot, isothennal conditions (557 +2 7),

0

-an alignment was performed of the delta .T and Tavg process instrumentation. ,

!See section.2.7 for a discussion of this test.

\-

During the power' ascension test program, data was collected at a series of power plateaus to align the delta T and Tavg. instrumentation. The nominal power levels were 30%, 50%;s75%, 90% and '1000 With the plant at steady ,

state. conditions, voltage readings' were measured and recorded for each

.r protection channel T 'T and Tavg.at the applicable protection '

cabinet.: The recordEv,olh were then' co'nverted to. tenperatures and the

' measured Tavg was compared with the Tavg calculated from-the measured T and T and Tc spare E resis b . Also, at all power ledels except 100%, Twere measured-for each protection c

- RTD resistances were then converted to tanperature and canpared. to the appilcable loop T and Tg. The core exit 1C values were also recorded

.'as additional'in tion r ~

At 30%, 50%, 75% and 90% power levels the following acceptance criteria were satisfied:

0

1) The01 p.Tavg sumnator. output converted to F shall agree within +

0.5 F of the Tavg value calculated from the measurements of the loop operational TH T C

s,- .,

2) .The hoop operational T RTD output converted to F shall agree within

-+1.2'F of the value ted fran th

.Tnstalled spare TgRTD converted to.gF,measured output,0f the loop O

3) The foop operaticinal T RTD output converted to F shall agree within

+1.2 F of the value cchted from th measured output of the-loop Tnstalled spare TCR1D nyertal t F.

Ati75% power, using the temperature and calorimetric data' from the previous tests as well as the 75% power . test, the 100% ' power loop delta T's and Tavg's were' extrapolated:

~

- 4 J im m r[1 tat,OF '

Tavg, OF

.n /

Loop 1 56.28 583.3 7

, .r . t . ,

', M LLoop 2' 'Y 55.56 582.4 n G? .i

TLoop'3, '56.97 583.6 Loop 4 56.37- 583.6 b

_g

~

.s 4.0-53 ,'

f i . ,.'

g A, ,S :. M . a b ~_ h_. _ . , . , . _ - . - .. . _ _ . ~ . , . . _ . . _ . . _ . . _ , . _ . , _ , _ , , - , - -,

The. loop deltg T's compared reasonably well with the expected best estimate value of 56.8 F quoted in design documents. The extrapolated maximum loop Tavg for 100% power fell outside the original acceptance criteria but further test data at 75% power sMwed that it was satisfactory (See Section 4.4.5) . As a result of the extrapolated 100% power delta T's, new gains were calculated for the delta T sustator cards:

Loop New Delta T Sunmator Card Gain 1 1.45 2 1.44 3 1.40 4 1.41 0

At lg0% power, no loop Tavg exceeded 588.5 F and no loop delta T exceeded 59.4 F. However, to bring the loop delta T sumnator output converted to %

power within +1% of the calorimetric power (100.1%), it was necessary to increase the outputs'of loops 1, 3, and 4 delta T summators. This was accomplished by calculating new gains:

Loop Origir.a1 New Delta T Sumnator Power Card Gain 1 97.5% 1.445 3 98.63% 1.420 4 98.63% 1.451 In addition, although not required by the acceptance criteria, the output of the loop 2 delta T sumnator card was optimized by adjuating the gain to 1.459 (original power _99.15%) . Table 4.4.4-1 sumnarizes the tenperature data at the 100% power plateau.

4.0-54 y *

"W

1 TABLE 4.4.4-1 TH1PERATURE. ALIGNENP DATA AT 100% POWER Loop R/E Converter- Delta T Sunmator Delta T Tavg Stamator Tavg No. Ogtput Output m Calgulated ' Ogtgut - Calgulated

. F- Power (%) F -F- h 'F 1 T, m 615.3 97.5 54.87- 55.4 587.6 587.6 T' ~ ' 559.9 2 T,_ 614.4 99.15 55.09 54.9 587.1 587.0

- T Q 559.5 3 T, _ 616.5 >

~98.63- 56.19 56.5 588.2- 588.3 T Q' 560.0 a

. ~ ~o 4 ' T,_ 615.4 98.63 55.59 55.7 587.5- 587.6 k TC~06 559.7

~

Calorimetric Power = 100.1%

9

-r--

4.4.5 STARIUP ADJUS'INENTS OF 'niE REACIOR CQfrROL SYSTEM The reactor control syste in the automatic mode positions control rods to maintain Tavg in the reactor coolant syst s ag a reference ta perature, Tref. If Tavg is different from Tref by +1.5 F or more, the reactor control systs steps the control rods to restore Tavg to the desired taperature band. The reactor control syste uses the highest loop Tavg for comparison to Tref. Turbine power based on first stage turbine impulse pressure and nuclear power signals are used to form a power mismatch contribution to the rod control systs. The combina3 taperature error signal (Tavg-Tref) and the power mismatch signal determine rod speed as well as direction of motion.

The purpose of this series of tests was to determine the Tavg program resulting in the highest possible steam pressure and thus optimum plant efficiency without exceeding gressure limitations for the turbine, or the design full power Tavg (588.5 F) . As discussed in section 2.7, basgline data was collected in the postcore loading precritical tests at 557 F and 2235 psig. Additional data was collected at 30%, 50%, 75% and 100% power in conjunction with the thermal power measurements, section 4.4.1. Data collected for each loop included T Tavg, feedwater flow, and steam generator pressure. Firstshe,Tt N n,e impulse pressure was also recorded as well as Tref, auctioneered Tavg and auctioneered NIS power.

After the 75% data was collected, first stage turbine impulse pressure, Tavg and steam generator pressure were extrapolated to 100% power with the following results:

First stage turbine impulse pressure ~ 6850psia Tavg ~ 584 F Steam generator prescure 935 psia Since Tavg and steam generator pressure were extrapolated to be low, the gain and bias for the control cards TY-505A and TY-505E were checked. The bias on TY-505A was lower than the required value of 7.921. Additional data was then taken and extrapolated to 100% power:

First stage turbine impulse pressure ~ 691.8 gsia Tavg- ~ 587.5 F Steam generator pressure ~ 990.5 psia ,

At 100% power, the final seg of data was collected with Tref found to be slightly greater than 588.5 F. Tge output of control card TY-505A was adjusted to correct Tref to 588.5 F at nominal 100% power. After this correction, steam generator pressures were:

Loop 1 1007 psia Loop 2 1010 psia Loop 3 1004 psia Loop 4 1010 psia 4.0-56 a

.g . .

'J 'u - - At the same time, first stag'e turbine impulse pressure was an average of '

712.8 psia. Various control cards were adjusted to match the actual turbine -

impulse pressure vs.. plant power. The steam generator pressures at 100%

power were determined to' be satisfactory.

As a result of this test, the reactor control system was adjusted to provide a_ sufficient supply of steam at rated pressure to support 100% power operation.

1 t

l 4

?

4 a

~-

N r-4.0-57

+

W

' 4

4.5 STEAM GENERA'IOR MOISTURE CARRYOVER MEASUREMENT This test was performed to determine the average moisture carryover content in the steam fran the stean generators at 100% power. The radioactive tracer method was used to determine the moisture carryover.

With the plant operating at 100% steady state conditions, a one curie liquid radioactive tracer (sodium-24 in the form of a sodium nitrate solution) was mixed with approximately 20 gallons of danineralized water in a tenporary

- mixing tank and then injected into each steam generator feedwater line using tim four feedwater hydrazine ammonia addition pumps. The feedwater lines transported the radioactive tracer into each steam generator. A large volume of demineralized water was then injected to flush out the chenical addition lines. Steam generator blowdown flow had previously been secured to prevent dilution (and loss) of the radioactive tracer. Also, the condensate polishing systen and the condensate makeup reject line back to the condensate storage tank had previously been isolated to prevent dilution (and loss) of the radioactive tracer.

After a 30 minute stabilization period to allow mixing of the radioactive tracer in the steam generators and carryover of the radioactive tracer with moisture into the condensate /feedwater systems, three sets of samples were taken at 15 minute intervals from each of the four steam line probes, from each steam generator upper shell, and from the main feedwater systen. The samples were then analyzed for sodium -24 activity. Using the results of the sample analysis, a percent moisture carryover was calculated for each set of stean generator upper shell samples and steam line samples. Since the steam generator upper shell results were considered to be more accurate, the steamline sanples were used as a backup.

Tnis test was satisfactorily performed on August 10, 1985 with the plant at 99.74% power as determined by the plant calorimetric procedure. The moisture carryover results are sumnarized in Table 4.5-1. The average moisture carryover was 0.015% as compared to an acceptance criterion of 0.25%.

4.0-58 4

' TABLE 4.5-1 STEAM GENERATOR MOISTURE CARRYOVER TEST RESULTS I

Sample Point Sample Set Percent Carryover Steam Generator 1 .0166%

. Upper Shell 2- .0168% .

I, _ b 3 .

.0122%-

. . Average. . 015% -

Main Steam Probe *. 1 .0167%

2 .0168%

3 . 0122%

.r Average .015%

.

• Backup l

1 4.0-59 l-

4.6 NSSS ACCEPrANCE TEST The purpose of the NSSS Acceptance Test was 1) to demonstrate the availability ard reliability of the Nuclear Steam Supply Systen and 2) to measure the NSSS power output. We test did not verify any safety criteria, but rather verified the NSSS vendor had supplied an acceptable system for contractual and warranty purposes.

The reliability of the NSSS is demonstrated by maintaining the plant at rated output for 250 hours0.00289 days <br />0.0694 hours <br />4.133598e-4 weeks <br />9.5125e-5 months <br /> without incurring a load reduction or plant trip due to a NSSS malfunction. Meally, the 250 hours0.00289 days <br />0.0694 hours <br />4.133598e-4 weeks <br />9.5125e-5 months <br /> should be continuous uninterrupted operation. However, it is acceptable to have only 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> of continuous 100% power operation with the remaining 150 hours0.00174 days <br />0.0417 hours <br />2.480159e-4 weeks <br />5.7075e-5 months <br /> being accumulated time rather than continuous. Power may be reduced during the 150 hour0.00174 days <br />0.0417 hours <br />2.480159e-4 weeks <br />5.7075e-5 months <br /> portion, but the accumulation of time is stopped until power is returned to rated output.

We measurement of the NSSS output is achieved by calculating the enthalpy rise across the steam generators. This enthalpy rise is determined by measuring the inlet feedwater flow, temperature and pressure and its outlet steam pressure. The steam flow is considered to be equal to the feedwater flow since steam generator blowdown is isolated during the test.

The reliability portion of the test coranenced at 1607 on August 8,1985, and 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> of continuous operation was achieved at 2007 on August 12, 1985.

At no time during that period was power reduced below its acceptable value of 3425 + 0,-5% M9T. Imnediately following the completion of 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> of operatico, the plant was reduced to approximately 60% power to investigate vibration problems with one of the main feedwater pumps.

The punp vibration problen was corrected and at 0630 on August 21, 1985, power was returned to the acceptable range (3254-3425) for testing and the additional accunulation of 150 hours0.00174 days <br />0.0417 hours <br />2.480159e-4 weeks <br />5.7075e-5 months <br /> began. On August 22, 1985, power was reduced below the test band for 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> and 14 minutes during the performance;of another test procedure.

@e measurement of the NSSS output portion of this test was conducted on

. August 23, 1985. Four sets of feedwater and steam data were collected over a four hour period. The results proved tc be very consistent and well within the acceptance criteria of 3425 +0, -2% Mfr with the calculated values being 3417.5 Mff, 3410 M9T, 3422.4 M9T, and 3416.7 M9T. @ese values also easily met the additional requirements that each calculated value be within +1% of the average of the 4 calculated calorimetric values. The variation fran the average was less than 0.2%.

We total accumulation of 250 hours0.00289 days <br />0.0694 hours <br />4.133598e-4 weeks <br />9.5125e-5 months <br /> was officially signed complete at 1634 on_ August 27, 1985. Canputer trends and operations calorimetric procedures were used to verify that the power output remained inside the acceptable range during the test. The 250 hour0.00289 days <br />0.0694 hours <br />4.133598e-4 weeks <br />9.5125e-5 months <br /> run was completed without any NSSS malfunction and the only problem encountered during the test was due to secondary plant.

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J The t'esti was. considered successful in that it. proved the NSSS is capable of sustained operation at rated output ard is capable of producing an output at

.the warranted rating of 3425 IWP. Though'not a strict test acceptance criterion,. the test was also used to document that the secondary side of the plant was. capable of operating at 95% of its rated electrical output.

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4.7 PnWER ASCENSION 'n{El@ULL ABO DYNAMIC TEST The purpose of this test was to monitor those systems or portions of systems that could not be monitorad during the preoperational hot functional test thermal expansion program because systems did not reach normal operating tenperature ard also to renonitor those points that did not meet the acceptance criteria for the preoperational hot functional test. W e testing included:

1) Detonstrating that the following systems (or portions of systems) which were not monitored during hot functional testing are free to expand thermally as designed:

a) Main Steam System, from the main steam headers to the condenser via the condenser dtrup lines, b) Main Steam System, from the main steam header to the the steam generator fes3 water ptrup turbines, c) . Main Feedwater System, from the steam generator feedwater pumps to the steam generators,

2) Monitoring the dynamic response of the main steam system to a plant trip frcxn 100 percent power, .

3)- Remonitoring the snubbers and spring hangers whose measured movements during hot functional testing were outside of acceptance criteria or associated piping did not reach the normal operating temperature,

4) Visually monitoring (measurement, if required) steady-state vibration ofpressurizersurgepipingwithreacgorcoolgntsystemprimaryloop at normal operating mode (RCS at 557 F + 10 F, four ICP 's running).

For the thermal expansion portion of the test, 81. lanyard transducers and 37 resistance tenperature detectors (RID's) were installed at selected locations to measure piping movements and corresponding temperatures in the

- main steam and main feedwater systens. These instruments were connected to a data acquisition system (DAS) provided by West.inghouse where data was collected ard printed out at the selected tenperature plateaus. At these temperature plateaus, walkdown inspections were performed to verify that no piping was being restrained, other than by design, frcxn thermal growth, swing clearances were checked on all required snubbers and snubber and spring hanger settings were recorded.

For the dynamic response to the 100% power plant trip, 41 lanyard transducers and 4 pressure transducers were installed at selected locations in the main steam system. We instruments were connected to the Westinghouse DAS where the dynamic response for each channel was recorded on a EM tape recorder and a digital data acquisition system.

4.0-62

'Ihis testing was begun with the start of the post core loading precritical test program and continued through the power ascension test program until the 100% power plant trip was performed on August 28, 1985. During this time, all monitored systes were heated up .to normal operating temperature and cooled down to ambient, whicg was heated up to10340 + q one F, cooled downexception was the to ambient, heated O main up to 445 feedwater s

+ 10 F (normal operating temperature) and cooled down to 340 + 10 F. Pre-test ambient data, intermediate heat up plateau data (when plant conditions allowed), normal operating temperature (hot). data and post-test ambient data were collected. All data was reviewed and approved onsite by a Bechtel stress engineering team. All acceptance criteria pertaining to thermal expansion were satisfied. >

A visual inspection was performed of the pressurizer surge piping and displacenent and velocity data was taken at various locations using a vibration monitor. ' The ICS was at normal operating temperature and pressure with 4 ICP's operating. The highest peak to peak displacement was 3 mils with a velocity .of 0.1 in/sec giving a frequency = .1 in/sec = 33.3 Hz.

.003 in The data obtained were within allowable limits.

For the 100% power plant trip, data was collected on the FM tape recorder 2 minutes prior to the trip and 3 minutes following the trip. Data was collected on the digital data acquisition system approximately 30 seconds prior to the trip and 60 seconds after the trip. The time history plots for each channel were obtained for the period.of the monitored transient. These plots were' evaluated by Bechtel Stress Engineering. This evaluation concluded.that while all piping movements were within the allowance of applicable codes, some additional pipe supports should be added to the steam dump piping. This will be accomplished as permitted by plant conditions.

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4.8 BIorAGICM. SHIEED TESTING The purpose of this test was to measure and record the neutron and gama-ray radiation levels in accessible areas of the plant where radiation levels above background were anticipated and to determine locations, if any, where shielding was deficient thereby ensuring that plant personnel would not be subjected to overexposure from radiation as a result of inadequate shielding. The test procedure established the Health Physics requirments for the biological shield survey point selection and survey techniques as well as methods of documentation.

To meet t;he requirments of this test, a series of four biological shield surveys were performed during the period from May 17, 1985 to August 9, 1985.

The first survey was performed on May 17, 1985 prior to initial criticality. This Preoperational Survey was intended to provide baseline data and demonstrate that no sources of radiation were present that would effect subsequent surveys. The survey was successfully performed with no abnormal findings.

The Low Power Survey was performed on May 24, 1985, with reactor power at 3%. No unexpected radiation readings were noted. General readings taken inside steam generator labyrinths were in excess of 100 mrem per hour. The containment hatches were posted as High Radiation Areas in accordance with 10CFR20 and procedural requirements. The readings found were within the expected ranges due to N-16 shine fran primary piping.

The Intermediate Power Survey was originally started on July 6, 1985, in containment. This portion of the survey was terminated due to extensive flux mapping interference with gama readings. The survey outside containment was performed on July 8, 1985 with reactor power at 49.4%. One abnormal neutron reading near two electrical penetrations was noted. The area was resurveyed with another instrument and no neutron readings were detected. It is believed that D4 interference affected meter deflection.

The containment portion of the Intermediate Power Survey was completed on July 15, 1985, with reactor power at 48%. No readings beyond expected extrapolated values were noted.

The Containment portion of the High Power Survey was performed on August 8, 1985, with reactor power at 100%. No unusual radiation readings were noted. The balance of the survey was performed on August 9,1985 and no unusual readings outside containment were noted.

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• . '4.9 PLMfr PIRE0llMMCE TEST

'lhis test was used to monitor performance of various plant systems during the power ascension test program. -Specific test objectives were:

- 1) - To monitor'the balance of plant and electrical systems under loaded conditions,.

2) To obtain data to verify the ability of ventilation systes to maintain ambient taperature within design limits,

' 3) To verify evacuation alarm audibility'in high noise areas,

4) - To monitor concrete taperatures surrounding hot penetrations.

During the power ascension test program, baseline data was collected at' .

steady state conditions. Plant systes monitored during this test included:

Main Steam

-Feedwater Feedwater' Heater Extraction; Drains and Vents Station Service and Essential Service Water Transformer Electrical Load Centers AC Inverters 125 V AC and DC Contalment Cooling Auxiliary Building Ventilation Control Building Ventilation .

Steam Tunnel Ventilation Auxiliary Feedwater Pump Room Ventilation Auxiliary Boiler Room Ventilation Radwaste Building Ventilation Turbine Building Ventilation Camponent Cooling Water

. Steam Generator Blowdown:

Condensate Demineralizer Water h acceptance. criteria of the test were satisfied:

I

1) The audibility of the evacuation alarms was verified throughout the

-plant,

2) h contaimegt air coolers maintained containment air temperature less than 120 F throughout the power ascension test program.

In addition, baseline data was collected for many systems in the plant. One plant parameter was found to be significantly outside its predicted range.

Condensate pump C' suction pressure was 7.5 psia whereas 3.3 to 5.0 psia was expected. This condition is being investigated but does not impact syste operation.

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4.19 TURBINE GENERNIOR TESTS

%e Generalfdlectric turbine generator was checked out with a series of tests which monitored the turbim, generator aM associated cupport systes from the time the turbine was first brought on line until the plant was at 100% of rated power. Testing was performed with the generator off line and with the plant at 20%, 40%, 60%, 75%, 90% and 100% of rated power.

-The turbine was first brought to rated speed using nuclear power at 0308 on June 12, 1985. Turbine bearing vibration was monitored during acceleration to check for critical speeds. While at rated speed, several tests were performed; General Electric performed a checkout of the exciter, the turbine lube oil systs was checked, turbine steady ctate vibration was recorded, DiC control circuits were tested, the generator core nonitor was started up aM adjusted for proper flow, and a number of other parameters were monitored and recorded using permanent plant and test instrumentation and 4

the plant emputer.

After all exciter checks had been completed, the generator was first synchronized to the grid at 1857 on June 12, 1985. The unit trippad almost inmediately due to a minor problem with the- turbine control circuits. Se secoM time the generator was synched it again tripped due to a slightly different control circuit proble . At approximately 0233 on June 13, 1985

~the generator was successfolly synched and Wolf Creek began supplying electrical power for the first time using nuclear fuel. However, the turbine only ran for a few hours before it was shut down due to high vibration. It was determined that the vibration was due to the fact the turbine was operated nearly 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> without a load. When the generator was finally put on line and the turbine loaded down, it cooled cff rapidly causing the vibration. After the turbine had been off line and cooled off, it was brought back on line and the generator synched without any vibration problems.

With the generator on line and the load at approximately 20%, additional turbine generator monitoring was performed. All parameters which had been checked with the generator offline were again monitored and recorded. A final checkout of the exciter was performed and the generator hydrogen seal oil flow was measured. A check of the reverse power relay and turbine overspeed tests were also performed.

Turbine generator monitoring continued.at various testing plateaus through 100% power. Parameters which had previously been checked were again nonitored and recorded. In addition to those previously mentioned, these parameters included generator stator bar temperatures, DC control signal

-parameters, power load unbalance circuitry, thermal expansion movement,' seal and lube oil pressures, turbine performance data, MSR parameters aM a number of other miscellaneous parameters. As before, permanent plant and i test instrumentation as well as the plant computer were utilized for recording data.

See additional checks were done at 100% power. The main field and alterrex field carbon brush vibration was measured and a full load check and L recalibration of pressure transducers to reflect actual versus design 4.0-66

settings was performed.

Although a nunber of minor problems were found and subsequently corrected,

the only proble of any major significance was in the dynamic noise testing. Several different types of measurements were performed to check control valve instability and dynamic noise. The prob 1m was first encountered at 40% power and was seen in each succeeding plateau in varying degrees with each type of measur ment. The proble has been partially corrected by fixing a ground loop found in the EHC control cabinet. General

. Electric feels' the proble will be fully resolved after a filtering circuit is installed in the control valve signal amplifier circuitry.

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'4.11 SPIK:IAL TESTS Two special tests were performed during the post' fuel load test program..

The first monitored the performance of the moisture separator reheaters.

(MSR's) The second special test collected baseline data on the reactor vessel level' instrumentation systen (RVLIS) .

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m 4.11.1 MOIS'IURE SEPARMOR REHFATER TEST The General. Electric ~ turbine utilizes four moisture separator reheaters to reheat steam discharging fra the high pressure turbine before sending it to the three low pressure turbines. W e steam is reheated by tapping high taperature, high pressure stean off the main steam line.

We main steam is supplied to the reheaters in two stages. W e first stage has no control circuitry. It is manually turned on a M off by operators.

We second stage, however, has a more complex control circuit. Fran 5% to 65% turbine load the reheater outlet pressures are monitored to control main steam flow to the reheaters using a low load control valve. When the turbine load increases above 65% a high -load control valve then opens up.

(This is an on-off control valve, not proportional control).

Since the operation of the reheater was not being checked in the turbine generator testing, a special' test was written to monitor operation of the moisture separator reheaters (MSR's), especially the secoM stage reheat control circuitry while the turbine was being loaded.

The test was performed while the turbine was being loaded during startup and power ascension after the 50% plant trip. Both the first and second stage reheat supply lines were put into service using the normal plant operating procedures. Then, while the turbine was being loaded, data was recorded on moisture separator reheater valve positions, inlet tenperatures, inlet pressures, flow rates, outlet temperatures, outlet pressures, and controller outputs. The data was recorded every 5% to 75% of rated load. This data was then analyzed and compared to expecred values.

' The moisture separator arr3 first stage reheater drain tank condenser dump valves closed as expected at approximately 10% load. We second stage reheater drain tank condenser dump valves closed at approximately 20% load as expected. W e reheat supply line drains closed when the associated reheat supply lines were opened. The second stage reheat low load control circuitry controlled flow as expected as the turbine load was increased to 65%. And finally at approximately 65% load the high load valve opened up as it should have. W e only problems encountered were some computer points which did not indicate properly.

We test showed that the overall performance of the rgheaters maintained all the low pressure turbine inlet temperatures within 50 F as required by General Electric. Wough the reheaters met the minimum requirements, it is felt the overall system performance could be improved by tuning the-systen.

This tuning and additional nonitoring will be done as a part of normal plant performance testing and monitoring.:

4.0-69 P

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4.11.2 REACTOR VESSEL LEVEL INSTROMENTATION SYSTEM (RVLIS)

RVLIS is a redundant safety grade systen which provides reactor vessel water level indication. The systen utilizes two sets of two d/p cells. These cells measure the pressure differential between the bottom of the reactor vessel and the top of the vessel. _ Cells of differing ranges are utilized to cover different: flow behavior with and without EP operation:

1) Reactor Vessel Narrow Range (Delta P B This instrument provides an indication of reactor vessel level from the bottom of the reactor vessel to the top of the reactor during natural circulation conditions.

2) Reactor Vessel Wide Range (Delta PC}

This instrument provides an indication of reactor core and internals pressure drop for any combination of operating EP's. Comparison of the measured pressure drop with the normal, single-phase pressure drop will provide an approximate indication of 'the relative void content or density of the circulating fluid. The indication of coolant density is significant only when subcooling margin is near zero. This instrument monitors coolant conditions on a continuing basis during forced flow conditions.

To provide the required accuracy for level measurement, temperature measurements of the impulse lines are provided. These tenperatures, together with existing reactor coolant temperature measurements and wide-

- range RCS pressure, are enployed to compensate the d/p transmitter outputs for differences in systen density and reference leg density . This would be particularly important during the change in the environment inside the containment following an accident.

The RVLIS test collected baseline data on systen operation under a variety of plant conditions. Specific test objectives were to:

1) Check the RVLIS wide range compensating function by recording dynamic head and full range level indicator readings with all FCP's running over full BCS power / pressure / temperature,

2) Obtain plant specific RVLIS dynamic head and full range level indications for 0,1, 2, 3 and 4 FCP's operation at both cold shutcbwn and hot standby conditions,

3) Record RVLIS RTD and other inputs at selected points during heatup after fuel load,

4) Observe hydraulic isolater operation during heatup and initial plant operation.

4.0-70

Data collection started with the post core loading precritical test sequence

.and was completed with the plant at 100% power.

The hydraulic isolators operated per design. With the exception of one data point, all indications were within the expected ranges. The one exception-was for full range indication LI-1321-at a hot zero flow condition where the i expected range was'104 +6%. The actual recorded reading was 111%.

mre difficulty was experienced with agreement between the full range indicators and with agreenent between the dynamic range indicators. Thero are two indicators of each type and a two pen recorder with one pen assigned to each type. Required agreement for a given type of indication (full range or dynamic) was' within 2% for all .three indicators. In some cases the

- actual agreement was 3% or greater.

Although the system operates as designed, investigation .is continuing to determine the cause for excessive difference between the indicators of a given type. This work will be completed and the system fully operational by startup .following the first refueling.

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4.0-71

APPDOIX A CHRONO [DGY OF THE POST FUEL IDAD STAR 1UP PROGRAM March 11, 1985 -- NRC issued Low Power License authorizing initial fuel load and pre-critical testing, initial criticality and low power physics testing.

March 12, 1985 - Commenced fuel loading - Entered Mode 6.

March 17,'1985 - Completed fuel loading.

March 21, 1985 -

Reactor vessel studs tensioned - Entered Mode 5.

March 26, 1985 Reactor coolant system filled. ,

March 27, 1985 - Initial fuel load procedures approved by PSRC.

Received authorization from the Plant Manager to comence post core load precritical testing at 1300.

March 31, 1985 - Ccrmienced cold control rod testing.

April 10, 1985 - C mpleted cold control rod testing.

0 April 17,1985- -

Entered Mode 4 (>200 F) at 0750.

0 April'26, 1985 - Entered Mode 3 (>350 F) at 2300.

~ April 30,1985 - RCS At 5570F, 2235 psig at 0500.

May.4, 1985 - Comenced hot control rod testing.

I May 7,' 1985. - Cmpleted hot control rod testing.

A May 19, 1985 - Post core loading pre-critical testing approved by.

the PSRC. Received authorization from the Plant Manager to comence initial criticality and low power testing at 2330.

May 21, 1985 Comenced Mluting for initial criticality at 0840 -

Entered Mode 2.

May 22, 1985 - Reactor critical at 0745.

May 31, 1985 - Completed low power physics testing at 1130.

June 4, 1985 - NFC lifted 5% power restriction on license.

June 5, 1985 -- Initial criticality and low power physics testing approved by the PSRC. Received authorization from the Plant Manager to comence initial synchronization and 20% power test sequence at 1555.

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. June 6, 1985 -

Entered Mode 1 at 2222.

. June 13, 1985 - Turbine-generator synchronized to the grid at 0204.

June.18,- 1985 .- Initial-synchronization and 20% power test sequence complete at 2200.

June 19, 1985 - Testing approved by the PSRC. Received authorization from the Plant Manager to comence the power ascension and 50% power test sequence at 0830.

June 29, 1985 - Performed plant shutdown external to the control room.

July 6, 1985 - Plant at 50% power.

July 16, 1985 - Perfo'rmed rods drop and plant trip test.

July 18, 1985

- Ibwer ascension and 50% power testing complete at 0200.

July 19, 1985 - Testing approved by PSRC. Received authorization from the' Plant Manager. to comence the 75% power test sequence.at 1312.

July _20, 1985 - Plant at 75% power.

July 29, 1985 . . 75% power test sequence complete at 0645.

July 30, 1985- - Testing approved by the PSRC. Received authorization from the Plant Manager to comence the 90% power test sequence.

August 4, 1985 - Plant at 90% power.

August 6, 1985 - 90% power test sequence complete at 1630.

August 8, 1985 -- Testing approved by the PSRC. Received authorization from the Plant Manager to comence the 100% power test sequence.

- Plant at 100% power at 1607.

August 12, 1985 - 100 hour-' continuous run complete at 2007. Power reduced to ~55% because of main feed pump "B" vibration problens.

~ August 21, 1985 - Plant at 100% power.

e August.28, 1985- - Performed 100% plant trip at 0513. 100% test sequence complete.

August 30, 1985 - Testing approved by PSRC.

Septenber 3,1985 -

Unit declared comercial at.Oll4 A-2

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1 APPEtOIX C UNPLANNE REACTOR TRIPS

' DURING POST FLTL LOAD TEST PROGRAM Date/ Time Power Level Cause Low-low steam generator level due to feedwater control June 6, 1985/2247 6% - 7% oscillations Low-low steam generator level June 13, 1985/0401 12% due to main feedwater pump trip Inadvertently opened reactor June 23, 1985/1333 30% trip breaker during performance of surveillance procedure Low-low steam generator level July 9, 1985/1115 50% due to test recorder connections July 10, 1985/0820 ~45% Hi-hi steam generator level caused MEWI and low-low steam July 11, 1985/0230 12% - 15% generator level due to' feedwater control valve leakage Loose lug in PN07/PN08 caused July 23, 1985/0815 75% loss of S/G feedwater pump Positive rate trip on NIS July 31,1985/0030 75% channel N41 due to spike with another channel in test August 7, 1985/0626 91% Turbine trip due to hi-hi noisture separator reheater drain tank level TOTAL: 9 unplanned reactor trips C-1 L _ _ s

unsas cas ano necrmc cowaur 3 THE ELECTFMC COMAANY OLENN L. KOk&TER

. WCE PRES.OENT huCLEAR Novenber 27, 1985 22@E DW E V Mr. R.D. Martin, Regional Administrator U.S. Nuclear Regulatory Comission DEC - 3m Region IV , y 611 Ryan Plaza Drive, Suite 1000 " -

Arlington, Texas 76011 i

'10D EC 85-259 l Re: Docket No. STN 50-482 Subj: Startup Report for Wolf Creek Generating Station

Dear Mr. Martin:

The enclosed Startup Report is submitted pursuant to Technical Specification 6.9.1.

Yours very truly,

[ 24 Glenn L. Koester Vice President - Nuclear GE:see Attachment xc: PO'Connor (2), w/a Jraylor (36), w/a JCunmins, w/a 1 UN -

g 201 N. Market - Wichta, Kansss -Mail Address: RO. Box 208 i Wichita, Kamas 67201 - Telephyte: Area Code (316) 261-6451 t-