ML20012D474

New Hampshire Yankee Seabrook Station Initial Startup Rept for Oct 1986 - June 1989. W/900313 Ltr

ML20012D474
Person / Time
Site:	Seabrook
Issue date:	03/13/1990
From:	Feigenbaum T PUBLIC SERVICE CO. OF NEW HAMPSHIRE
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NYN-90070, NUDOCS 9003270386
Download: ML20012D474 (110)

Text

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Tod C. Presi Senior Vice FO < dent and ChielOperating Officer NYN-90070 March 13, 1990 United States Nuclear Regulatory Commission Washington, DC 20555 Attentions -Document Control Desk References Facility Operating License No. NPF-67. Docket No. 50-443 Subject Initial Startup Report Gentlemen:

In accordance with the requirements of Technical Specification 6.8.1.1, enclosed is the Initial Startup Report for Seabrook Station covering the period from October, 1986 through June, 1989. Supplemental reports will be submitted every three months until commencement of commercial = operation as required by this Technical Specification.

Should you have any question, please contact Mr. James M. Peschel, Regulatory Compliance Manager, at (603) 474-9521, extension 3772.

Very truly yours,

[- Ted C. Feigenbaum .

7, cci Mr. William T. Russell

g. Regional Administrator L,a United Stat's Nuclear Regulatory Conunission Region I ei 475 Allendale Road nQ King of Prussia, PA 19406 EO L oo Mr. Victor Nerses, Project Manager

$$ Project Directorate I-3 United States Nuclear Regulatory Commission v ' t,o o Division of Reactor Projects

$8 Washington, DC 20555 Y

M Mr. Noel oudley

$$ NRC.Senio? Repident Inspector /

• 1a P.O. Box 1149 ,

Seabrook, NH 03874 / l \

g , New Hompshire Yankee Division of Public Service Company of New Hampshire P.O. Box 300

• Seabrook, NH 03874

• Telephone (603) 474 9521 J 1

NEW HAMPSHIRE YANKEE SEABROOK STATION i

INITIAL STARTUP REPORT to the UNITED STATES NUCLEAR REGULATORY COMISSION OPERATING LICENSE: NPF 67 NRC DOCKET NO. 50-443 For the Period October, 1986 through June 1989 i

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TABLE OF CONTENTS Section IAgg, List of Tables III List of Figures IV '

List of Acronyms VI 1.0 Introduction 1-1 2.0 Startup Test Program Overview 2-1 6 3.0 Seabrook Startup Chronology 3-1 i

4.0 Initial Fuel Load 41 4.1 Summary of Initial Fuel Loading 4-1 4.2 ST-2. Primary Source Installation 42

• 4.3 ST-3, Core Loading Prerequisites 4-3 4.4 ST-4. Initial Core Loading 4-4 -

5.0 Precritical Testing 5-1 5.1 Precritical Test Summary 5-1 5.2 ST-50, Moveable Incore Detector System 5-2 t

5.3 ST-44, Loose Parts Monitoring System 5-3 5.4 ST-5, Control Rod Drive Mechanism Operational Test 5-4 5.5 ST-53. Turbine Driven Emergency Feedwater Start 5-6 Verification 5.6 ST-11, Reactor Coolant System Flow Measurement 5-7 5.7 ST-55, Steam Dump System Test 5-9 5.8 ST-10. RTD Bypass Loop Flow Verification 5-11  ;

5.9 ST-9 Pressurizer Spray and Heater Capability 5-12 i 5.10 ST-12. Reactor Coolant System Flow Coastdown 5-16 5.11 ST-8 Rod Position Indication 5-19 5.12 ST 7, Rod Drop Time Measurements 5-20 I

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TABLE OF CONTENTS (Continued)

• 5.13 ST-6, Rod Control System 5-23 )

5.14 ST-26. Thermal Power Measurement and Statepoint Data 5 24 Collection 6.0 Initial Criticality and Low Power Physics Testing 6-1 6.1 Summary of Initial Criticality and Low Power Physics 6-1 f Testing  !

6.2 ST-16 Initial Criticality 6-2 6.3 ST-17 Boron Endpoint Heasurement 6-7 6.4 ST-18. Isothermal Temperature Coefficient 69 l 6.5 ST-19, Flux Distribution Hessurements at Low Power 6-12 6.6 ST-20, Control Rod Worth Measurements 6 14 6.7 ST-20.1 Additional Control Rod Worth Measurements 6-32 6.8 ST-21, Pseudo Rod Ejection Test 6-33 t-6.9 ST-22. Natural Circulation Test 6-35 5 7.0 Instrument Calibration and Alignment 7-1 7.1 Instrument Calibration and Alignment Summary 7-1 7.2 ST-13. Operational Alignment of Nuclear Instrumentation 7-2 7.3 ST-14.1, Operational Alignment of Process Temperature 73 Instrumentation '

7.4 ST-14.2 Resistance Temperature Detector and Incore 7-4 Thermocouple Cross Calibration 7.5 ST-15. Reactor Plant Systems Setpoint Verification 7-6 8.0 General Plant Testing 8-1 8.1 General Plant Testing Summary 8-1 8.2 ST-41 Radiation Survey 8-2 8.3 ST-42, Water Chemistry Control 8-3 8.4 ST-45. Process and Effluent Radiation Monitoring System 8-4 8.5 ST-51. Power Ascension Dynamic Test 8-5 8.6 ST-52. Power Ascension Thermal Expansion Test 8-6 8.7 ST-56. Piping Vibration Testing 8-8 II

• 1 LIST OF TABLES ST-7 Table 1 Control Rod Drop Times 5 22 ST-9 Table 1 Pressuriter Spray and Heater Effectiveness 5-13 P

ST-10 Table 1 RTD Bypass Flow Rates 5-11 ST-11 Table 1 d/p Flow Calculations 5-8 i

ST-12 Table 1: Flow Coastdown Results 5-17 ST-17 Table 1 Boron Endpoint Test Results 6-8 ST 18 Table 1 Temperature Coefficient Predictions 6-10 -

and Results  :

$7-19 Table 1 Flux Map Parameters at Low Power 13 ,

ST-20 Table 1 Rod Worth Results 6-16 i

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i LIST OF FIGtRES i

ST-4 Fig.1 Core Load Verification Map 4-6 j Fig.2 Fuel Assemblies Loaded vs Time 4-7 i Fig.3 ICRR for N31 vs Fuel Assemblies Loaded 4-8 Fig.4 RCS Boron Concentration vs Time 4-9 l Fig.5 RHR Temperature vs Time 4-10 ST-9 Fig.1 Pressurizer Spray Effectiveness 5-14 )

Fig.2 Pressurizer Heater Effectiveness 5-15 I l

ST-12 Fig.1 Flow Coastdown at Hot Standby 5-18 1 ST-16 Fig.1 ICRR vs Sequential Rod Withdrawal 6-4 '

Fig.2 ICRR During Dilution vs Time 6-5 Fig.3 RCS Boron Concentration vs Time 6-6 ST-18 Fig.1 Control Bank D Operating Band 6-11 ST-20 Fig.1 CBD Integral Worth 6-17 Fig.2 CBD Differential Worth 6-18 Fig.3 CBC Integral Worth 6 19 Fig.4 CBC Differential Worth 6-20 Fig.5 CBB Integral Worth 6-21 Fig.6 CBB Differential Worth 6-22 Fig.7 CBA Integral Worth 6-23 Fig.8 CBA Differential Worth 6-24 '

Fig.9 SBE Integral Worth 6-25 Fig.10 SBE Differential Worth 6-26 Fig.11 SBD Integral Worth 6-27 Fig.12 SBD Differential Worth 6-28 IV

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LIST OF FIGURES. (Continued)  !

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Fig.13 SBC-Integral Worth 6-29 l

Fig.14 SBC Differential Worth 6-30 Fig.15 Control Bank Worth in Overlap 6 31 l Li ST-21 Fig.1 Pseudo Ejected Rod Flux Trace 6-34 ~

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[ ST-22 . Fig.1 Loop 1. Tgoy..T COLD, TA yg vs Time ' 6 36 i

. Fig.2 Loop-1. Delta-T vs Time 6-37 i

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0-LIST OF ACRONYMS ACOT - Analog Channel Operational Test ARI - All Rods Inserted l

ARO - All Rods Out ASDV - Atmospheric Steam Dump Valve CB_ - Control Bank (A.B.C. or D) ,

CRD - Control Rod Drive CRDM - Control Rod Drive Mechanism CVCS - Chemical and Volume Control System DRPI - Digital Rod Position Indication ECCS - Emergency Component Cooling System EFPM - Effective Full Power Minute EFW - Emergency Feedwater FSAR - Final Safety Analysis Report FTC - Fuel Temperature Coefficient i

GETARS - General Electric Transient Analysis Recording System :

HFT - Hot Functional Test HSB - Hot Standby i HEP - Hot Zero Power ICRR - Inverse Count Rate Ratio IR - Intermediate Range ITC - Isothermal Temperature Coefficient LPMS - Loose Parts Monitoring System MCB - Main Control Board MIDS - Movable Incore Detector System MSIV - Main Steam Isolation Valve l

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I LIST OF ACRONYMS (Continued)

MTC - Moderator Temperature Coefficient MWT - Megawatts Thermal NDR - Nuclear Design Report NHY - New Hampshire Yankee NI - Nuclear Instrumentation NRC - Nuclear Regulatory Commission  ;

NSSS - Nuclear Steam Supply System >

t PCV - Pressure Control Valve l PLS - Precautions, Limitations and Setpoints PORV - Power Operated Relief Valve RCCA - Rod Cluster Control Assembly RCS - Reactor Coolant System RDMS - Radiation Data Management System RHR - Reactor Heat Removal RTD - Resistance Temperature Detector SB_ - Shutdown Bank (A.B.C.D and E)

SFP - Spent Fuel Pool SORC - Station Operating Review Committee ,

SSPS - Solid State Protection System STP - Special Test Procedure UEEC - United Engineers and Constructors i

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VII 1 l

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• J 1.0. INTRODUCTION l

Seabrook Station received a license to load fuel and conduct precritical testing in October,1986. Due to the uniqueness of the Seabrook i licensing process, there have been extensive delays between fuel loading and completion of the low power physics tests. Precritical testing was completed  ;

in March, 1987, and two years elapsed before a low power (<52) license was aeceived. Initial criticality and low power physics testing were conducted in June, 1989.

The startup report covers the following testing sequences:  ;

1. Initial Fuel Loading and Precritical Testing. October, 1986 through March 1987. ,

2. Initial Criticality and Low Power Physics Testing, June 1989.

l All or portions of 33 startup tests, identified by the prefir ST, are l reported herein. Additional testing will take place upon receipt of a full I power license, and will be reported in supplements to this document.  ;

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2.0 STARTUP TEST PROGRAM OVERVILV The Startup Test Program was initiated on October 17, 1986 when the ,

first steps in procedure ST-3, Fuel Loading Prerequisites, were undertaken, j A sero power operating license was received that day, and the first fuel I assembly was loaded on October 22, 1986. Fuel loading proceeded, and with  !

only a few minor interruptions for mechanical problems, was completed on i October 29, 1986. Precritical testing commenced as soon as the reactor vessel upper internals and head installation was completed.

1 The sequence of precritical tests was developed to demonstrate that the j facility was in a final state of readiness for initial criticality and the conduct of low power physics testing. The precritical tests follow core I loading, and when completed, the facility is ready for initial criticality.

Data acquisition utilised a General Electric Transient Analysis Recording System (GETARS). The Seabrook installation utilises 132 Validyne amplifiers which provide signal conditioning and where necessary, analog to digAtal signal conversion. Test points may be scanned at a rate of up to 1000 times per second, and data is stored on disk for later transfer to tape for analysis and permanent retenticn.

The first precritical tests were the initial setup of the moveable incore detector system (ST-50), and baseline thormal expansion data (ST-52),

in mid-December, 1986. Shift crews were activated, and the precritical test program initiated in early January, 1987.

The actual order of testing is generally the order of listing in the Table of Contents, and reflects the sequence given in ST-1 Startup Test Program Administration. Certain tests are repeated under diff erent plant conditions, such as Water Chemistry Control (ST-42). Water chemistry samples are taken throughout the entire startup program, and the procedure is listed under General Plant Testing, Section 8.0. Another category of tests which are repeated throughout the startup program is Instrument Calibration and Alignment, Section 7.0.

Actual start of precritical testing was delayed for a month by pressuriser PORV and atmospheric dump valve (ASDV) repairs. The general precritical test sequence got underway in early February, 1987, as heatup using reactor coolant pumps began. Holds at 250, 350, 450 and $57 'F, permitted thermal expansion measurements (ST-52), and other sequence tests.

Testing on the turbine driven emergency feedwater pump (ST-53) and steam dump valves (ST-55) was conducted during heatup to qualify system performance carried over from HFT testing.

A significant portion of the time required for precritical testing was used for control rod system tests (ST-5 ST-6 ST-7, and ST-8). Precritical testing ended and the plant was cooled down in mid-March, 1987.

A low power license was issued on May 26, 1989.

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SEC,2.0 STARTUP PROGRAM OVERVIEW (Continued) .

1 Plant heatup was started when the license was received. Prior to SORC approval to take the reactor critical, prerequisites and some turbine driven {

EFW pump testing was required. These preliminaries were completed on June

12. Withdrawal of Shutdown Bank A began early on June 13. and the reactor was declared critical at 1723 hours0.0199 days <br />0.479 hours <br />0.00285 weeks <br />6.556015e-4 months <br />.

j 1

A license condition limited testing to an accumulated core exposure of j 45 Effective Full Power Minutes (EFPM). As is discussed in Sec. 6.1, one of 4 the first activities, following initial criticality, was calibration of l intermediate range (IR) detector currents with reactor coolant loop delta-T l measurements to provide a means of monitoring core exposure. Core exposure was then tracked on the plant computer.

Most of the low power testing is core physics: boron endpoints, control ,

rod worth measurements, and neutron flux distributions. Stuck and ejected 6 rod measurements were included. The test sequence proceeded relatively +

smoothly, with occasional interruptions. Mechanical problems were encountered with the moveable incore detector system (MIDS), which slowed, but did not prevent data acquisition.

The final test in the sequence. ST-22 Natural Circulation, was terminated prior to completion, due to failure of a steam dump valve controller. The test is rescheduled during the Power Ascension Test Program.

On June 23, 1989 testing was completed, and shortly thereafter the plant was cooled down to await receipt of a full power license.

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'3.0 SEABROOK STARTUP CHRONOLOGY The chronology documents the Startup Test Program from receipt of a license to load fuel, to completion of Low Power Physics Testing (52 license) j

,. EA.1.t. 31tD.1.

10/17/86 License to load fuel and conduct precritical tests received. ,

i 10/20/86 SORC approval to begit. Startup Test Program, i 10/22/86 Commenced fuel loading (ST-4) i 10/29/86 Loaded final fuel assembly: completed independent- core l verification.

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11/3/86 Vessel assembly completed.

i 11/7/86 Precritical startup tests commence. {

Pressuriser PORVs removed for repair.

l Insertion problem with movable incore detectors (ST-50). 1 12/29/86 Pressurizer PORV repairs completed. j CRD mechanical test (cold) underway (ST-5). j 1/6/87 Additional pressurizer PORV work.

2/10/87 Heating up, 250 'F hold for testing. I 2/11/87 Hold at 350 'F: enter Mode 3.

9 2/15/87 Complete tests at 450 'F hold point.

2/18/87 Pressuriser and RCS flow tests (ST-9 ST-11).

CRD testing underway (ST-5, ST-6 ST-7. ST-8).

2/19/87 Test of RTD bypass loop flow (ST-10): new cold leg orifices required.

2/24/87 Commence steam dump tests (ST-55).

Flow coastdown test (ST-12). .

3/12/87 Steam dump tests completed.

Preparing for EFW pump tests (ST-53).

3/17/87 EFW tests completed.

3/19/87 End of precritical tests.

Cooldown to await license for low power physics testing.

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t- l 3,0 SEABROOK STARTUP CHRONOLOGY (Continued) 5/26/89 License issued for low power physics testing.

i 6/2/89 Commenced plant heatup and additional IFW testing.

6/12/89 Received SORC approval to start initial approach to critical  ;

(ST-16).

6/13/89 Withdrawal of SDA at 0210 hours0.00243 days <br />0.0583 hours <br />3.472222e-4 weeks <br />7.9905e-5 months <br />.

Reactor critical at 1723 hours0.0199 days <br />0.479 hours <br />0.00285 weeks <br />6.556015e-4 months <br />. l l 6/14/89 Established Effective Full Power Minute (EFPM) conversion factor. 3 l Start low power physics tests: Boron endpoints (ST-17) and ARO isothennal temperature coefficient (ST-18) underway.

6/15/89 Connenced flux mapping (ST-19), and control rod worth measurements (ST-20). 1 6/17/89 Completed ST-17, and ST-18.

Continuing rod worth measurements.

6/19/89 Started pseudo ejected rod measurements (ST-21).

6/21/89 Completed ST-20.

Additional rod worth measurements (ST-20.1) underway.

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6'22/89 Attempted natural circulation test (ST-22).

6/.'3/89 Low power physics testing completed.

Awaiting receipt of full power license.

3-2  :

4 4.0 INITIAL FUEL LOAD 1 4.1 Summary of Initial Fuel Loadinn The initial fuel loading test sequence consisted of the following l startup test procedures:

l ST-2. Primary Source Installation '

ST-3, Core Loading Prerequisites ST-4. Initial Core Loading Installation of the two primary neutron sources took place on Oct. 20, 1986. There was some initial maintenance on required fuel handling systems, but the sources were then installed in fuel assemblies, C04 and C30, without further problems.

Procedure ST-3, Core Loading Prerequisites, as the name implies, was a check list of thirty items, including fuel inventory, reactor vessel preparation and a number of performance tests. Appendix A to the procedure was a Westinghouse Calibration and Checkout Procedure for the Core Loading Instrument System (temporary incore neutron detectors). The core loading prerequisites were completed on October 22, 1986.

Initial core loading began on October 22, 1986, at 1852, with the installation of fuel assembly C04, which contained a primary neutron source.

Fuel loading was dryg only the reactor vessel and transfer tube were flooded. Transfer of fuel assemblies proceeded as sequenced in ST-4. An exception was necessary when there was a problem inserting B05 in location C10: the sequence was changed to ' box' the location, and B05 was successfully inserted. Loading was again interrupted when C17 could not be unlatched at location B03: the problem was traced to the refueling machine position indicator.

Minor delays occurred on several occasions. The interruptions generally '

involved mechanical problems with the refueling machine and upender.

RHR temperature and RCS boron concentration remained relatively constant throughout core loading. These parameters are shown graphically in ST-4, Figures 4 and 5.

The-final fuel assembly, C44, was loaded on October 29, 1986 at 0712.

The core contains 193 fuel assemblies. The loading chronology is shown in ST-4, Figure 2.

On completion of initial core loading, a video-taped core verification was conducted to document proper fuel assembly placement. The verified core map is shown in Figure 1.

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4.2 ST-2. PRIMARY SOURCE INSTALLATION I l

Obiective The objective of the procedure was the proper handling and installation I of primary source assemblies into the required fuel assemblies. )

Primary source installation is described in FSAR, Section 14 Table 14.2-5. Sheet 5.

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Discussion Primary source installation was conducted just prior to initial fuel loading, and consisted of the installation of two Californium-252 sources from a shielded shipping cask, into fuel assemblies C04 and C30.

J The source, when installed, is attached to one position in a source assembly, which is identical in construction to a burnable poison assembly.

In preparation for installing the source, the source assembly was withdrawn from the fuel assembly to provide access to the fuel assembly guide thimble ,

which is to receive the source rod.

A source rod was removed from the shipping cask, and af ter a health physics leak check, was lowered into the specified fuel assembly guide thimble. In turn, the source assembly was reinserted into the fuel assembly guide thimbles, and the source rod attached to the source assembly with a nut, threaded onto the source rod, secured with a lock wire, and welded in place.

The process was repeated for the second source rod, end the fuel assemblies returned to storage.

Results After some initial delays associated with maintenance on the overhead auxiliary bridge crane, the SFP bridge crane, and the new fuel elevator, the source installation was successfully completed.

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o 4.3 ST-3. CORE LOADING PREREQUISITES Obiective '

Core Loading Prerequisites addresses plant systems, conditions and '

equipment necessary for a safe, controlled cure loading. Technical Specification requirements and system acceptance and performance tests are included.

The procedure is described in FSAR, Section 14, Table 14.2-5. Sheet 6.

Discussion Prerequisites for initial fuel loading consisted of thirty steps which required signoff. Included were fuel storage inventory, reactor vessel preparation, assembly and checkout of temporary core loading instrumentation, primary source installation (ST-2), performance tests addressing fuel handling, service water, RHR, CVCS, ECCS, and Instrument and Service Air. Sampling of reactor coolant and associated auxiliary systems verified uniform boron concentration. System lineups for those systems to be utilized in fuel loading completed the requirements.

Results The core loading prerequisites were completed, with no exceptions.

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4.4 ST-4. INITIAL CORE LOADING ,

Obiective i The objective of this procedure was detailed guidance for loading i reactor fuel assemblies from storage positions in the Fuel Storage Building  ;

to the reactor vessel in the Containment. Prerequisites, special precautions, and the sequencing assured loading in a safe, controlled manner.

l Initial core loading is described in FSAR, Section 14 Table 14.2-5, Sheet 7.

Discussion I Initial fuel loading required six days and one shift from insertion of  ;

the first assembly, C04, to the final, C41 The loading was performed with  !

the cavity dry only the resctor vessel and transfer tube were flooded.

Startup sources had been installed in the appropriate fuel assemblies in the ,

Spent Fuel Pool per ST-2, Primary Source Installation, and prerequisites  ;

completed using ST-3, Core Loading Prerequisites. In addition to the '

permanent source range nuclear instrumentation, three temporary incore neutron detectors were employed.

l The fuel was transferred from the spent fuel pool in the Fuel Storage Building, to the refueling cavity in the Containment via the normal fuel >

ttansfer system. Initial fuel loading utilized station fuel handling procedures. i The core load operation was based on the recommended Westinghouse loading sequences after a few assemblies were loaded around a startup source  ;

in the vicinity of one source range detector, a ' bridge' of fuel assemblies was loaded across the core to enable the second source range detector to i monitor reactivity. The remaining corners were then loaded. The temporary detectors were repositioned throughout the core loading sequence to ensure proper instrument response to verify that the core remained suberitical throughout the core loading process. t Inverse Count Rate Ratios (ICRR) vs number of fuel assemblies loaded, were graphed. The ICRRs were based on count rates from each of the Source Range channels. N31 and N32, and each of three temporary incore detectors.

The ICRR graphs were used to verify that the core remained suberitical.

Boron concentration and RCS temperature were monitored on an hourly basis whenever fuel loading was underway. If loading was interrupted. RCS temperature was monitored every hour, and boron concentration every four +

hours.

A core verification video tape, using an underwater camera and recorder, was made following completion of fuel loading. An independent review of the tape was conducted.

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~4,4 ST-4. INITIAL CORE LOADING fContinued) ,-

i Results )

Initial core loading proceeded in a timely fashion with no unusual l interruptions. The nuclear instrumentation, both source range and temporary, I provided the expected response throughout the process. A number of minor )

interruptions were encountered, generally involving mechanical problems with l the refueling machine and upender.  !

One fuel assembly, B05, could not be loaded in the sequence. The  ;

assembly was returned to the SFP, and inspection indicated a possible slight '

bowing. The loading sequence was modified to ' box' the position, and B05 was successfully loaded.

The following figures document the initial core loading process:

Figure 1, Core Load Verification Map Figure 2 Fuel Assemblies Loaded vs Time '

Figure 3. ICRR for N31 vs Fuel Assemblies Loaded l Figure 4 RCS Boron Concentration vs Time  !

Figure 5. RHR Temperature vs Time f

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ST-4 CORE LOADING VERIFICATION MAP I FIGURE 1 -

t 0 DEGREES 15 C29 C58 C37 C10 C52 C14 C44 '

14 C48 C57 C62 A19 C30 A34 C31 A09 C21 C23 C16  ;

13 C46 C33 B08 A54 B25 A45 B44 A57 B41 A08 B29 C59 C35 I

12 C18 B55 B27 B21 A07 B60 A12 B30 A31 B51 B36 Bf4 C19 11 C49 C64 A65 B40 A59 B52 A20 B58 A27 B42 A50 B04 A25 C08 C53 l

10 C39 A39 B05 A04 B26 A55 B01 All B23 A18 B35 A17 B56 A02 C09 9 C24 C11 A48 B64 A56 B02 A44 B17 A47 B33 A64 B50 A58 C61 C20 270 8 C13 A24 B57 A21 B32 A35 B53 A29 Bil A60 B38 A36 B18 A37 C22 90 7 C40 C55 A61 B06 A63 B13 A03 B07 A15 B15 A53 B12 A13 C03 C41 t 6 C34 A62 B14 A46 B63 A10 B31 A26 B10 A41 B62 A05 B37 A06 C38 5 C45 C25 A28 B22 A14 B09 A23 B61 A32 B20 A51 B16 A49 C42 C06 4 C05 B39 B59 B49 A01 B43 A52 B34 A43 B48 B24 B28 C47 3 C17 CIS B47 A33 B19 A42 B45 A40 B03 A30 B46 C51 C56 2 C63 CO2 C28 A16 C04 A22 C01 A38 C54 C32 C36 1 C26 C50 C07 C12 C60 C43 C27 A B C D E F G H J K L M N P R l 180 I

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ST-4 CHRONOLOGY OF LOADING  :

FIGURE 2 FUEL ASSEMBLIES LOADED i 200 190 -

• 180 -

(

f '

170 -

160 -

150 -

140 -

130 -

120 -

110 -

100 -

90 -

80 -

70 -

60 -

50 -

40 -

30 -

20 -

10 -

l 0 ' ' ' '

O 1 2 3 4 5 6 7 DAYS, FUEL LOAD 10/22-10/29 4-7

ST-4 ICRR VS FUEL ASSEMBLIES LOADED FIGURE 3 INVERSE COUNT RATE RATIO (ICRR)*  !

1.5 ,

t 1.25 - '

!\

t 3,Q/

1 0.75 -

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l 0.5 ' ' ' ' '

l 0 20 40 60 80 100 120 140 160 180 LOADING STEPS

=

DETECTOR N31

• DATA SHOWN, EVERY STH POINT 4-8

ST-4 RCS BORON CONCENTRATION -;

FIGURE 4  !

BORON CONCENTRATION (PPM) 2250 t

i 2150 -

rv. e . :=< N L W N "

2050 -

t 1950 i i i i i i i 0 1 2 3 4 5 6 7 DAYS, FUEL LOAD 10/22-10/29 4-9

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i ST-4 RHR TEMPERATURE l FIGURE 5 TEMPERATURE, DEGREES F 4

l 90 -

1 80 - == = - 1

• %*N _/e - - ,

s"  ;

70 -

60 -

l 50 ' ' ' i i L 0 1 2 3 4 5 6 7 DAYS, FUEL LOAD 10/22-10/29 .

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U 5.0 PRECRITICAL TESTING a

5.1 Preeritical Test Sn==ary The precritical test sequence is intended to ensure that reactor and related plant systems are in a final state of readiness for initial criticality and low power physics testing. This test phase commenced as soon l as vessel assembly had been completed following fuel load, and on completion, the plant was cooled down to avait receipt of a low power license.

Testing included in the precritical test sequence verified the following:

i. - Acceptable piping thermal expansion and vibration during heatup and operation of selected systems.

- Accessibility of the incore instrumentation thimbles by the Moveable incore Detector System (MIDS).

- Alert settings for the Loose Parts Monitoring System (LPMS).

- Operability of the control rod drive mechanisms, and rod control and rod position indication systems.

- Control rod drop timing under hot standby conditions.

- Turbine Driven EFW starting.

- Acceptable RCS flow.

- Acceptable RTD bypass loop flow.

- Steam dump system performance.

- Capability of the pressurizer sprays and heaters.

- Four pump flow ccastdown.

- Net RCS fixed heat input by determination of steam generator and chemical and volume control system (CVCS) heat removal rates.

After an initial delay for maintenance on power operated relief valves (PORVs) and atmospheric dump valves (ASDVs), heatup (with ' test holds" at 250, 335 and 450 'F for thermal expansion and vibration measurements), and testing at hot standby proceeded with few interruptions. The availability of steam to the secondary plant nede heatup of the secondary difficult, and limited conduct of tests which were carried over from Hot Functional Testing (HFT), (steam dump and EFW start verification). On several occasions, testing delays for steam dump and EFW adjustments were encountered.

l Precritical' testing ended in mid-March, 1987, and the plant was cooled down to await receipt of a license for low power testing.

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5.2 ST-50. MOVEABLE INCORE DETECTOR SYSTEM Obiective ,

Procedure ST-50 verified performance of the Moveable Incore Detector System (MIDS) in each mode of operation, determined initial limit switch settings for each path, and following initial criticality, determined detector operating voltages and intercalibration. t The procedure is described in FSAR, Section 14. Table 14 2-5. Sheet $3. ;

Discussion The MIDS provides core neutron flux distribution data to the plant computer system for determination of core power distributions. Six miniature fission chambers are each inserted in a pattern of thimbles in fuel assem-blies, via a transf er and switching tube system. A number of adjustments and checks are required for the mechanical system, as well as for the neutron detectors themselves. L Prior to initial criticality, a series of system performance functional checks were made using a dummy detector. Position indication on each drive was checked for modo, path length and locations thumbwheel settings were entered, and drive speed measured.

P The dummy detector was replaced with fission chambers, and, after initial criticality, voltage plateaus and operating voltages were determined I using station operating procedures.

Results All acceptance criteria were met. ,

During performance of ST-50, seven test exceptions were takens four path blockages, two failed indicator lights, and a faulty position indicator. All were resolved. Af ter the dununy detectors were replaced, some path limit settings required adjustment.

5-2

a $

, , 4 5,3 ST-44. LOOSE PARTS MONITORING SYSTEM )

Obiective This procedure establishes the following data and setpoint information for proper performance of the Loose Parts Monitoring System (LPMS):

1. Beseline data. '

2. Alert settings and adequate instrument sensitivity and response.

3. LPMS response during deliberate plant maneuvers to establish loose parts system disable levels to avoid spurious alarms.

4. LPMS sensitivity meeting Regulatory Guide 1.133 requirements.

LPMS is described in FSAR. Section 4.4.6.4, and Section 14. Table 14.2-5, Sheet 47.

Discussion Prior to ST-44, station operating procedures were used to setup and calibrate LPMS.

Initial measurements were made for this procedure at cold shutdown, under no-flow conditions. Baseline data (gain, threshold and background levels) were taken, and control rod test noise (ST-5 and ST-7) was observed on LPMS. Performance was also observed during startup of the reactor coolant pumps.

A second series of measurements were made under hot standby, full flow conditions. Baseline data was again taken, and alert setpoints calculated >

for each of the sixteen channels. Again, observation of LPMS response to control rod testing noise was made.

Additional LPMS testing under ST-44 occurs in the Power Ascension Test Program at the 50% and 1002 power test plateaus.

Results l

All acceptance criteria were met.

Two test exceptions were taken during the performance of ST-44. One l- resulted from insufficient background noise on channels in the no-flow i condition. A second involved K values (setpoints) for two channels. Both

! test exceptions were resolved during the Low Power Physics Tests.

1 5-3

{ 13 j

i 5.4 ST-5. CONTROL ROD DRIVE MECHANISM OPERATIONAL TEST i

Obieetive i

' The objective of the procedure was verification of proper operation and l timing of each slave cycler mechanism, and a demonstration of proper i operation of the control rod drive mechanisms (CRDMs). Proper operation of '

the CRDMs was shown by recording CRDM magnetic coil currents and audio signals. The slave cycler testing was conducted with the plant in cold  :

shutdown: the remainder of the procedure was carried out in both the cold J shutdown and hot stkndby condition.

The CRDM operational test is described in FSAR Section 14, Table 14.2-5, Sheet B. '

Discussion i i

Initial testing under ST-5, was conducted with the plant in cold

~

shutdown (Mode 5), following core loading.

Slave cycler timing was verified by stlecting one rod from a power cabinet, and observing the time sequence of events, while monitoring the CRDM lift, stationary and movable coil currents, and the output from a sound pickup at the top of the rod travel housing. All other rods in the group under test were prevented from moving by opening the appropriate lift coil disconnect switches. The timing process was repeated for one rod in each power cabinet.

• The CRDH operational check required withdrawing, in turn, each shutdown and control bank to 50 steps, disconnecting the lift coils of all mechanisms except the one being monitored, and withdrawing and inserting the selected rod 6 steps while recording the three coil currents.

Under hot standby conditions (Mode 3), the CRDM operational check was ,

reperformed as described in the previous paragraph.

Results All acceptance criteria were met.

Proper slave cycler timing was verified by comparison of event timing to Westinghouse reference timing data.

The CRDM operational test demonstrated that the lift, stationary and movable coils operated in a satisfactory manner when the shapes of current traces were compared to Westinghouse sample data.

In the Mode 5 sequence, replacement or repair of several defective microphones was necessary, otherwise, the tests were conducted without problems. ,

I l 5-4 l

4 I

5.4 ST-5. CONTROL ROD DRIVE MECHANISM OPERATIONAL TEST (Continued)

Under hot standby conditions (Mode 3).- satisfactory performance was again _ demonstrated, but some concerns were noted; potential evidence of l stationary gripper drag, an extra sound pulse on completion of a step, and in one case, a higher than normal stationary gripper current.

Westinghouse, after reviewing the problems, recommended additional current measurements for the most suspect group. The remeasurements were made using an alternate procedure supplied by Westinghouse. All currento were found to be within normal range, including the abnormally high stationary gripper case noted above.

After. the additional measurements, two questionable ' conditions remained: reduced current levels on some rods , . and the additional sound pulses. Westinghouse evaluated these conditions and indicated that these

-conditions will not affect rod control system performance.

5-5

5.5 ST-53.' TURBINE DRIVEN EMERGENCY FEEDWATER START VERIFICATION-Obiective

' The objective of this procedure was a demonotration that the turbine driven Emergency Feedwter Pump (E N) could meet certain performance requirements.

The ! EW test is tot deveribed in the Startup. Tests, FSAR Section 14.

The test was includcd .e the preeritical startup series to verify system performance f ollowing modiueations required after hot functional testing (HFT).

Discussion The performance requirements to be met in S4-53 were:

1. At normal system temperatures and pressure, the pump will start five times when the steam supply line metal temperature is at or below 150*F.

2. Both EW pumps can run on recirculation without overheating the suction line.

3. The turbine driven EW pump can obtain a turbine speed of 3570 rpm

-in 60 seconds or less from start initiation.

4. The air accumulator for turbine steam supply valve, MS-V-395, can maintain the valve closed for 18 +/-2 seconds during start ' sequencing with loss of air.

Results All of the performance requirements-(acceptance criteria) were met.

A number of problems extended the . testing periods in particular, leaking steam supply valves which prevented the steam supply line from cooling down to the required metal temperature. A field change was written to install a station air supply for piping cooldown between cold starts.

Reruns were due to turbine overspeed trips, and.a drain valve leak.

Although acceptable test results were obtained, a decision was made to replace valve trim internals of the three turbine steam supply valves, and additional testing was scheduled prior to initial criticality. This testing was performed under a Special Test Procedure (STP). Turbine performance was satisfactory, but- valve- leakage was still unacceptable. To enhance reliability of the system, after low power physics testing, the three steam supply valves were removed and replaced with new valves. The system was retested under a second STP, and performance with the new valves found to be satisfactory.

5-6 1

i

'5.'6 St-il. REACTOR COOLANT SYSTEM FLOW MEASUREMENT Abiective The objective of this. test was determination ~ of the reactor coolant i

system flow rate at hot standby,- using installed elbow-tap differential pressure instrumentation.. The determination of reactor coolant flow using calorimetric techniques is performed in ST-26, during power ascension testing.

The reactor coolant system flow measurement is described in FSAR, Section 14,_ Table 14.3-5 Sheet 14.

Discussion With plant parameters at steady state, hot standby conditions, a series of ten measurements of temperature, pressurizer pressure and d/p transmitter output voltages was made for each reactor coolant loop. The average output-voltage from each of the three d/p sensors per loop was then converted into a pressure (in.H2O), and-loop flow calculated from the relationship W = A(2gvR(d/p)'/D)k whers: W = Coolant Flow, gpm d/p = Differential Pressure, in.H 2O v = Specific Volume, ft 3/lb R =

• Radius of curvature, 51 inches D = *Inside Diameter, 31 inches

• Cold Leg Parameters

which reduces to.

W = 39005((d/p)v)k Thus, with the d/p voltages- converted into in.H 20, and the specific volume from Steam Tables, loop flow can be calculated.

5-7

5.6 ST-11. REACTOR _ COOLANT SYSTEM FLOW MEASUREMENT (Continued)

Results The RCS flow measurement procedure was modified by field change- to eliminate the prerequisite RCS flow transmitter calibration prior to any

' data collection. Flow transmitter output . measurements were taken with a digita1' voltmeter. .The transmitters were'then'taken out of service and the d/p required to produce the measured voltage determined. The change eliminated one set of RCS' loop flow calibrations, by simultaneously *

obtaining the data for. the test and the RCS flow instrumentation normalization to 1002 indicated flow.

Calculate'd values of loop flow are given in Table 1.

The acceptance criterion:

Total reactor coolant flow >/= 391,000 gpm was met without difficulty:

Heasured total flow rate = 433,992 gpm.

. TABLE 1. d/o FLOW CALCULATIONS Loop 1: Loop.2:

Fr +.14. W = 105920 gpm FT 424, W = 110440 gpm FT 415 W = 113136 gpm FT 425, W = 111843 gpm FT 416, W = 109129 gpm FT 426. W = 105964 gpm Loop 3: Loop 4:

FT 434, W = 106686 gpm FT 444, W = 106717 gpm-FT 435, W = 105992 gpm FT 445. W = 113133 gpm FT 436. W = 103566 gpm FT 446, W = 109449 gpm Loop Averages: Loop 1, W = 109395 gpm Loop 2, W = 109416 gpm Loop 3 W = 105415 gpm Loop 4, W = 109766 gpm Total Flow: W = 433992 gpm Acceptance Criterion: Total Flow >/= 391000 gpm 5-8

i 5.7-ST-35. STEAM DUMP-SYSTEM TEST Obiective The objective of this procedure was verification, with the plant in hot standby, of the following:

1. Stroke time of each condenser steam dump valve is in accordance with design criteria.

2. Atmospheric steam dump valve (ASDV) controls are operational.

3. Steam line drain valves operate within design parameters in a dynamic condition.

4. Main Steam Isolation (MSIV) bypass valves operate within design parameters in a dynamic condition.

The steam dump system test is not described in the Startup Tests, FSAR, Section 14. ST-55 was prepared because steam dump performance was not satisfactorily demonstrated during hot functional testing (HFT).

Discussion The steam dump system- tests were included in the precriticality test program to validate systems refurbished or modified following hot functional tests.

The steam dump valves failed to meet stroke time criteria during HFT the air supply tubing was modified and volume boosters were installed.

A backup air supply system was installed for the ASDVs and this portion of the procedure verified performance of the modified ASDV system.

Motors on the steam line drain valves for the main steam system (upstream of the MSIVs) were replaced to. meet safety-related system requirements; retests of the modified valves were necessary.

-Dur'ing HFT, the motors on the MSIV bypass valves were found to be undersized. To meet design criteria, gear ratios were changed, and retesting required.

5-9

+.  ?

5.7 ST-55. STEAM DUMP SYSTDi TEST (Continued _)

Results All acceptance criteria were met.

A number of field changes were included to permit changes in the order of testing, permit retesting, and change test instructions to clarify e instructions or wording.

• Three steam dump valves required retestings a faulty position indication, an excessive closing time, and a valve stuck in mid-position.

All were satisfactory following maintenance.

Circuitry problems initially prevented ASDV performance. A design change ~and modification to the procedure corrected the difficulties.

Measured opening and closing currents for steam line drain valves exceeded the " nominal" value specified. The currents were considered acceptable because they did not exceed the 130Z of nameplate value limit.

Stroke time measured on three MSIV bypass valves were less than the specified values also, closing current on one of the three exceeded the

' nominal" value. The actual measured values were evaluated and found to be acceptable.

5-10

~J' __

'L 5.8 ST-10. RTD BYPASS LOOP FLOW VERIFICATION 1

l Obiective

~

l The objective of the test was determination of RTD hot and cold bypass line flow rates, and verification of low flow alarm setpoints. The bypass line- flow rates necessary to ptovide adequate transport times, were i calculated based on loop dimensions, and provided the basis for verification '

of adequate loop flow.

The RTD loop flow verification test is described in FSAR Section 14, Table 14.2-5, Sheet 13. .

Discussion Reactor coolant temperature in the loop hot and cold legs is measured by the RTDs installed in loop bypass lines. Adequate transport times and low ,

flow alarms in the bypass loops are requirements for reactor protection and Control.

The measurements were obtained by recording total RTD bypass flow in each loop with the manifold' valves open. The hot leg RTD bypass manifold isolation valves were then closed, and cold leg flow measured. The process was then reversed to determine hot leg flow. As the hot leg isolation valve was closed, the low flow alarm point was recorded and, 'if necessary, adjusted to actuate at approximately 90% of total bypass loop flow. ,

The minimum acceptable flows, calculated using loop dimensions, were compared to measured values.

Results Initial measurements yielded-cold leg bypass loop flows that were too

. low. ' The original 0.50 in, orifices were replaced with 0.668 in. orifices and acceptable flows were obtained. All acceptance criteria were met.-

Except for some minor communications and equipment problems, the_ test ran smoothly. Table 1, RTD Bypass Flow Rates, tabulates the results.

TABLE 1. RTD BYPASS FLOW RATES Hot ten Bvoass Cold Len Bvoass L Required- Actual Required Actual Looo No. Flow. nom Flow. nom Flow, nem Flow. nom 1 76.23 151.80 47.13 85.20 2 77.06 157.41 48.29 82.59 3 77.88 149.37 48.00 85.63 l

4 76.81 153.49 51.49 91.51 5-11

'E 5.9 ST-9; PRESSURIZER SPRAY AND HEATER CAPABILITY Obiective The objective of this procedure was a demonstration of the following -

pressurizer capabilities:

1. Pressurizer spray bypass valve' position to maintain an equilibrium temperature above the spray line low temperature alarm t

setpoint.

2. Pressure response - to spray valves fully open or all heaters fully energized, falls within acceptable limits.

i-

3. opening and closing times of pressurizer PORVs meet criteria.

The pressurizer spray and' heater capability test is described in FSAR, ,

Section 14. Table 14.2-5, Sheet 12. 1 Discussion Pressurizer l spray and heater capability was determined with the plant in hot standby.

The manual- bypass flow spray valves were opened individually in small 3 increments until spray line temperature remained essentially constant. The setting established for each valve maintained temperature above the spray line low temperature alarm point.

Pressure response. to spray valves fully opened was measured by simultaneously opening both spray valves, with pressurizer level control in j manur.1, and all pressurizer heaters deenergized. Pressure, water level and temperature parameters were monitored with the plant computer and a strip chart recorder. The pressure was allowed to decrease to approximately 2000 psig.before terminating the transient.

An additional spray performance test was included to acquire operational data for individual spray _ effectiveness with an RCP shut down. ,

TheLtwo spray lines tap RCP-1A and RCP-1C (discharge piping). Each pump, in  ;

turn, was shut down, with its associated spray valve closed.. while single spray effectiveness was measured-for the other spray line. A measurement of-

- the spray effectiveness - when both spray valves were open with one RCP shut down (alternate flow path back to the shut down pump) was included.

A similar transient response was measured to determine pressure response to all heaters fully energized..With pressurizer level and pressure control in manual, all control and backup heaters were energized at full output. Again,- pressure, water level and temperature parameters were tracked. The pressure increase was terminated at 2300 psig.

Opening and closing times for the pressurizer power operated relief valves were measured by opening each valve, in turn, for a sufficient time for a pressure reduction of approximately 50 psi, before returning the

-control switch to automatic to close the valve.

5-12

y ,

5.9 ST-9. PRESSURIZER SPRAY AND HEATER CAPABILITY (Continued)

Results All acceptance criteria were successfully met.

All.- test sections were ' completed with a ' minimum of difficulty. The

- pressurizer. ' full spray = capability test was performed twice because the-i

^ pressurizer. level trace exceeded ~. the . chart span, with both runs producing essentially the same results.

-Figure l ', Pressurizer Spray . Effectiveness , and Figure 2 Pressurizer

• Heater Effectiveness, show the measured values of pressure within the high and low acceptability limits.

f TABLE 1.' PRESSURIZER SPRAY AND HEATER EFFECTIVENESS

1. Continuous spray flow valves = 1/2 turn open.

n

2. Depressurization rate (both valves) = 2.45 psi /second

3. Pressurization rate (all heaters) = 0.26 psi /second i

4. Depressurization rates for different RCP combinations ares

a. RCP C off, PCV-455B open = 1.0 psi /second

b. RCP A off, both sprays open = 2.16 psi /second

c. RCP A off, PCV-455A open = 1.4 psi /second

5. PORV stroke times-

a. PCV-456A, Open = 0.27 second Closed = 0.83 second b.'PCV-456B, Open = 0.30 second Closed = 0.83 second 5-13

.m. .,,

-9 a

ST-9' PRESSURIZER SPRAY EFFECTIVENESS

, FIGURE 1 PRESSURE, PSIA- t 2250 i }

1 2200 -

2150 -

s

'2100 -

, i 2050 -

. s

)

.2000 ' ' ' '

0 10 20 30 40 50 60 70 80 ,

TIME, SECONDS

=

HIGH LIMIT- l LOW LIMIT

• RUN 1 O RUN 2 l

5-14

'~

.n. &  !

?

ST-9 PRESSURIZER HEATER EFFECTIVENESS ,

FIGURE 2  :

t

PRESSURE, PSIA, 2340 j t

2320 -

i 2300 -

2280 -

2260 -

->f 2240 ' '

0- 20 40 60 -80 100 120 140 16 0 180 200  :

TIME, SECONDS

=

HIGH LIMIT I LOW LIMIT

• MEASURED l

L 5-15

=

III0 ST-12$ REACTOR COOLANT SYSTEM FLOW COASTDOWN Obiective The objective of this procedure was to verify that the rate at which reactor coolant flow decreases following a simultaneous trip of all four reactor coolant pumps (RCP), is less than that assumed in the safety analysis. The delay times associated with the loss of flow accident were

-also determined.

The flow coastdown measurement is described in FSAR Section 14 Table 14.2-5 Sheet 15.

Discussion

-Flow coastdown was initiated from hot standby by a cimulated loss of flow signal (jumper to ground in the Logic Bay for SSPS Train A). A ,

simulated P-8 permissive enabled protective circuitry to trip on low reactor coolant flow.

The data points for the pertinent time interval, which included trip of pumps, onset of flow coastdown, and reactor trip breaker actuation, were ,

printed from GETARS and the relative times precisely determined. The low flow trip time delay evaluation and calculation of the flow coastdown time constant were satisfactory.

Data from the three d/p sensors were obtained from each loop. These values were converted to flow fractions and evaluated using a statistical weighting process. Each loop was required to have two sensors with fewer than four unacceptable values. In each of two loops, one value was found to be unacceptable, so the quality of the data was very good.

Results A satisfactory flow coastdown test must meet the following:

1. RCPs must trip within 100 milliseconds of each other.

2. Measured Loi* Flow Trip Time Delay </= 1.0.second.

3. Measured Flow Coastdown Time Constant. TAUg > 11.77 seconds.

All acceptance criteria were met.

Test results are given in Table 1.

Figure 1 shows the Relative Flow Fraction versus Time from RCP Trip.

The- design flow coastdown at hot standby, and the measured flow coastdown are graphed.

5-16

r ' -

1 i

p f.c L

-+? 5.10 ST-12. REACTOR COOLANT SYSTEM FLOW COASTDOWN (Continued)

TABLE 1. FLOW COASTDOWN RESULTS j 1

Reactor Coolant Pumo' Trio Soread: 20 milliseconds 1 1

J Low Flow Trio Time Delay Calculations-1 1

Time interval, in seconds, from flow reaching the low flow trip setpoint until the last reactor trip breaker changes state: )

J Loon 1 Looo 2 Loop 3 Loop 4 0.01 --0.02* 0.06 0.04 l 0.12 =-0.02* 0.04 0.10 0.08 -0.02* 0.04 0.10 Maximum T1 = 0.12 seconds Sensor Delay Time,'Td, Seconds:

1000 1 Looo 2 Loop 3 Loon 4 0.374 0.388 0.402 0.503 0.338 0.398 0.422 0.436 0.353 0.343 0.412 0.428 Maximum Td = 0.503 seconds Gripper Release Time, Tg = 0.15 seconds (Assumed Maximum)

Low Flow. Trip Time Delay, LT p = (T1+T d + Tg) = 0.77 seconds  !

(Acceptance Criteria = 1.0 seconds)

The flow coastdown time constant calculation yielded the-following results:

Slope = 0.0767 seconds *1 Intercept -0.0227 Flow Coastdown Time Constant, TAUg = 13.03 seconds.

(Acceptance Criteria >/= 11.77 seconds)

• The d/p trip value is a function of the d/p minimum value. In each loop, d/p minimum was negative, and when this negative value was included, the calculated value for d/p trip was lower than, for example, the value if d/p minimum was zero. For Loop 2, the calculated value of d/p was lower than the ~ '

GETARS value when the breakers tripped; thus the negative value reported.

5-17

$ l x.-

ST-12 FLOW-- COASTDOWN, HOT STANDBY FIGURE 1 l 1

RELATIVE FLOW FRACTION 1

0.95 -

0.9 -

0.85 -

0.8 -

0.75 -

l 0.7 -

0.6 5 ~

0.6 -

0.55 -

0.5- '

O 1 2 3- 4 5 6 7 8 9- 10 SECONDS, TIME FROM FOUR RCP TRIP SAFETY ANAL., HSB l ST-12 DATA AT HSB ,

l' 5-18 L

l

.A-

• - '5,'11 ST-8. ROD POSIT 20N IND2 CATION Obiective The objective of this procedure was a demonstration under hot standby conditions, of proper operation of control rod position indication over-the entire range of travel.

The test of control rod position indication is described in FSAR, Section 14. Table 14.2-5, Sheet 11.

Discussion Proper operation of the control rod position indication system was demonstrated by withdrawing, one at a time, each shutdown bank and each control bank to specified positions. Digital Rod Position Indication (DRPI)

System. Process Computer, and Rod Control System- Group Counter step indications were recorded at each position.

The shutdown bank position was verified at 18, 210 and 228 steps; control bank position was verified in twenty-four step increments to the fully withdrawn position.

Baseline counts were taken on the source range instrumentation prior to rod withdrawals, and count rates were monitored during withdrawals and recorded at each withdrawal increment.

Results Satisfactory performance of the control rod position indicators was demonstrated. All acceptance criteria were met with one exception; verified computer indication of control rod demand position. A test exception was written against this step, and a work request written to correct the problem. The exception did not affect the test schedule because the MCB indication of both DRPI and demand position were available and performed acceptably and computer indication is not a required system.

5-19 4

1

]

i

.. j 5.12 ST-7. ROD DROP TIME MEASUREMENTS obiective The objective of this procedure was determination of the rod drop time of each Rod Cluster Control Assembly (RCCA) at hot standby under full flow conditions. Rod drop time is defined'as the time from decay of stationary '

gripper coil voltage to dashpot entry. In addition, the. test monitored the rod behavior in the dashpot (deceleration) region.  :

The rod drop _ time measurement is described in FSAR, Section 14, Table 14.2-5, Sheet 10.

Discussion In the procedure, a bank of rods is fully withdrawn, and individual rods dropped sequentially by removing the appropriate stationary gripper fuse, after first removing the moveable gripper fuse. This process is repeated for each bank.

A Visicorder was used'to record coil voltage-time characteristic curves for each rod. The time to dashpot entry. T 1 and time to bottom of dashpot, Tf , were measured directly from the characteristic curve. The shape of each rod drop trace was evaluated against the expected shape to identify any abnormalities or evidence of binding.

Baseline ' count rates were established for each of the Scurce Range detectors prior to starting the test, and an increase in count rates by a factor of five-during testing would require that the test be stopped. During testing, the count rates remained well below this limit.

Noise on the signal cables required filtering before satisfactory _

traces could be obtained. During the test sequence, nine test exceptions were required. Problems encountered were: loss of sound traces on BOB and P06 the need for repeated measurements due to . signal loss on ' four occasions; abnormal traces on B10 and P06: reversed signal polarity on H08: ,

and LO5 . and M12 drop times, which were outside the two-sigma acceptance band. Since adjacent microphone signals could be used to support rod drops where ' microphones were inoperative, the defective microphones were not replaced.

5-20

c 4- -5'12 ST-7. ROD DROP TIME MEASUREKENTS (Continued)

Resulte q All acceptance criteria were met.

The average rod drop ' time was 1.39 sec., consistent with results at similar Westinghouse plants, and well below the Technical Specification -l limit of 2.2 seconds. Results are tabulated in Table 1.

The remaining acceptance criteria were

2. All drop test traces exhibit a free fall with no abnormalities or evidence of binding, t

3. RCCA deceleration through the dashpot region should be similar for all rods dropped under the same plant conditions.

4. All RCCAs outside of the two-sigma limit exhibit all repeat '

measurements within a 0.02 sec. band for a given plant condition.

These remaining criteria were met by all rods with the exception of B10 and P06, which exhibited trace abnormalities. Reactor Engineering review and Westinghouse concurrence, determined that the trace abnormalities would have no effect on nuclear safety and on overall operation of the rod control or position indicating systems.

One rod, M08, exhibited reverse polarity but normal trace characteristics. . Investigation indicated that the lift, stationary and

-moveable gripper coils for M08 are probably reversed when compared to the other control rods , - however, the coil orientation does not affect normal operation of M08.

Two rods, LO5 and M12, were slightly outside the two-sigma average drop time limit, and measurements were repeated per the procedure on these rods to meet the 0.02-second band criterion.

Due to the length of time between performance of this test and initial criticality, the rod drop surveillance test was reperformed to satisfy Technical Specifications. All traces were found to be acceptable, including B10 and P06, however, the polarity of M08 was still reversed.

5-21

~

'J 5.12 ST-7. ROD DROP TIME MEASUREMENTS (Continued)

TABLE 1. CONTROL ROD DROP TIMES Hot standby,-Full Flow conditions Bank Rod Time to Time to Bank Rod Time to Time to '

Dashpot Bottom of Dashpot Bottom of Entry Dashpot . Entry Dashpot (secs.) (secs.) (secs.) (secs.)

SBA D02 1.40 1.88 CBA E05 1.38 1.81-

.B12 1.40 1.91 Ell 1.37 1.87 M14 1.41 1.90 L11 1.38 1.86 PO4 1.42 1.90 LO5 1.36 1.82 B04 1.38 1.84 H06 1.40 1.91 D14 1.40 1.91 F08 1.39 1.87 P12 1.38 1.84 H10 1.38 1.87 H02 1.42 1.90 K08 3.38 1.87

.s SBB G03 1.41 1.88 CBB F02 1,41 1,91 C09 1.39 1.88 B10 1.38 1.88 J13 1.39 1.87 K14 1.41 1.91 N07 1.40 1.84 P06 1.40 1.87 C07 1.39 1.88 B06 1.39 1.82 G13 1.42 1.94 F14 1,39 1.91 N09 1.41 1.89 P10 1.38 1.84 J03 1.38 1.81 K02 1.39 1.83 SBC E03 1.39 1.86 CBC H02 1.38 1.8? <

C11 1.40 1.88 B08 1.39 1.89 '

L13 1.38 1.86 H14- 1.37 1.88 N05 1.38 1.86 P08 1.41 1.92 SBD C05 1.39 1.87 F06 1.42 1.90 E13- 1.42 1.88 F10 1.37 1.83 N11 1,41 1.87 K10 1.41 1.86-LOS 1.40 1.83 K06 1.40 1.87 SBE H04 1.38 1.86 CBD D04 1.37 1.84 D08 1.42 1. 9P. M12 1.36 1.85

!- H12 1.41 1.87 j H08 1.38 1.90 D12 1.40 1.89 M04 1.37 1.85 H08 1.40 1.85 L

I 5-22 l

+- i 5.13 ' ST-6. ROD CONTROL SYSTD1 Obiective The procedure demonstrated that the Rod Control System performed the required rod control and indication functions. The test was conducted with the plant in hot standby, prior to inAtial criticality.

The rod control system test is' described in FSAR. Section 14 Table 14.2-5. Sheet 9.

Discussion 1

Proper operation of the rod control system was demonstrated by moving rod banks, initially one at a time, to verify rod motion of all rods in the bank and compare indicated rod speed with speed from speed signal voltage ,

measurements. After testing all shutdown snd control banks, overlap was checked with a minimum of rod withdrawal, by resatting the overlap program using the thumb wheel switches in the Logic Cabinet.

Results The test was conducted without difficulties, and all acceptance criteria were met. Test results show that:

1. Control bank overlap .was in accordance with the switch settings in the Logic Cabinet.

2. Measured rod speeds were within +/- 2 steps of the speed signal.

3. Individual rod positions (DRPI) were within +/- 4 steps of bank demand positions.

4. The rod control system operated as expected.

i l

I L

l 5-23 l

l

c t e i LV ,

~*-

5.14'ST-26 THERMAL POWER MEASUREMENT AND STATEPOINT DATA COLLECTION Obiective The objective of this procedure is a calorimetric determination of  !

reactor-power,' and verification of main steam and feedwater performance from

.various . primary and secondary, process data. Appendix A to ST-26 was determination of the net RCS fixed heat input. The appendix was conducted as -

a precritical testr the remainder will be carried out at the 502, 502, 75%,

902 and 100Z power test plateaus.

The procedure is described in FSAR Section 14. Table _14.2-5, Sheet 29.

Discussion With the reactor subcritical at hot standby conditions, steam generator levels were stabilized at-752, and then the generators allowed to steam, with all feedwater secured, to a level of 50%. A group trend from the Main Plant Computer System (MPCS) recorded loop Tavg SG Level (NR)', and SG Steam Outlet Pressure at 5 minute intervals. The process was repeated a second time.

From the data, the steam generator heat removal rate, and-the chemical and volume control system (CVCS) heat removal rate were determined. The sum of these is the net RCS fixed heat input.

Results Data analysis yielded the following results:

Trials (1) (2)

Steam Generator Heat Removal Rate (MWT). 16.21 16.25 CVCS Heat Removal Rate (MWT) 2.17 2.26 Net.RCS-Fixed Heat Input (MWT) 18.38 18.51 Roundoff Value to be Used in Calorimetric Calculations 18.4 MWT There were no test exceptions.

l l

l l

1 5-24 l

L

w ,

6.0 INITIAL CRITICALITY AND LOW POWER PHYSICS TESTING, (

6.1 Summary of Initial Criticality and Low Power Physics Testina Initial criticality and low power physics testing confirm the core design, fabrication of fuel assemblies and assembly of the reactor core, i adequacy of the ' physics characteristics and overall performance of plant monitoring and control systems, j The low power license was issued on May 26, 1989, and plant heatup and required precriticality tests, such as operational alignment of nuclear and process instrumentation, were initiated. The approach to critical began on June 13, 1989 and was completed in approximately 15 hours1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br />. Following criticality, the low power physics testing range was determined, and a reactivity computer checkout conducted. -

A license' condition on the low power license limited testing to an accumulated core exposure of 45 Effective Full Power Minutes - (EFPM) . To verify compliance with'this restriction, a Main Plant Computer System (MPCS) program was written, and used to track core exposure throughout the test sequence. A total of 19.2 EFPMs were used during the low power physics test program.

Following initial criticality, testing in the low power physics sequence included:

- Boron Endpoints.

- Isothermal Temperature coefficients.

- Control Rod Worth Measurements.

- Initial Flux-Distribution Measurements.

- Pseudo Ejected Rod.

, - Natural Circulation Tesi..

1 j= Problems with Detectors A and C on the moveable incore detector system L (MIDS) slowed flux mapping by forcing the use of the ' Emergency" mapping sequences. Despite this inconvenience, the coro physics evaluations were

! completed in 9 days, with excellent agreement between measured parameters L and-predicted results.

The final test in the low power physics sequence, natural circulation, was undertaken on June 22, 1989. The test proceeded normally, but before

i. the onset of natural circulation, the controi linkage on a steam dump valve l failed and the . valve went to full open. The resulting abnormal rate of cooling ultimately led to a manual reactor trip and termination of the test without meeting the acceptance criteria. The test has been rescheduled into the Power Ascension Test Program.

6-1

' - ^

., i 4 6.2 ST-16. INITIAL CRITICALITY 1 1

Obiective l The objective of this procedure was to achieve initial criticality in a controlled ' manner by successive rod withdrawals and a final dilution of q reactor coolant. Determination of the zero power testing range, and 1 verification of shutdown monitor operation were also included. )

l Initial Criticality is described in FSAR, Section 14 Table 14.2-5, I Sheet 19. l 1

Discussion The approach to criticality proceeded in a smooth, controlled fashion, as prescribed by the procedure. The reactor was initially in hot standby, prerequisites . completed, including- an estimated critical boron concentration, and test' personnel ready to track progress with Inverse Count Rate Ratio (ICRR) graphs.

Shutdown banks, SDA through SDE, were withdrawn sequentially in 114 step-increments until fully withdrawn. Control Banks, CBA through CBD, were next withdrawn in overlap until CBD reached 140 steps, the position at which boron dilution was initiated. A criticality prediction was made from the ICRR data at each rod withdrawal increment, and with initiation of boron dilution, boron sampling and ICRR data were taken at 15 minute intervals.

Criticality . was reached in a controlled manner and ' well within the-predicted conditions.

After stabilization of the reactor at 1 x 10-8 amps on the Intermediate

! Range NIs, the zero power testing range was determined, and operational y verification- of the reactivity computer completed. Reactor power was then-l increased in steps of 0.52 to determine the low power flux mapping range.

Determination of the lowest possible power level for satisfactory flux mapping was necessary because of the license condition limiting core exposure to a maximum of 45 effective full power minutes (EFPM) during. low o- -power physics testing.

i l

6-2

4 6.2 ST-16. INITIAL CRITICALITY (Continued) -;

Results

' All: acceptance criteria were met.

The estimated and measured critical conditions were in excellent- i agreement: H Estimated Actual Control Bank D Position 140 steps 137 steps RCS Boron Concentration 1136 ppm 1160 ppm Other measured results:

Zero Power Test Range- 10*7 to 10-8 amp IR--

' Reactivity Computer Agreement- Within 42 of Prediction Minimum Rx Power (Flux Haps)- Approximately II Appendix' A of the procedure was a Shutdown Monitor Test, to be conducted during dilution to critical. The system was not operating properly, and an exception was written for the test. A second exception was written to cover the 2Z and 32 power level steps in determination-of a low power flux mapping range. A limitation on the _ maximum IR current prevented measurementefat these levels: the range was established without these data points.

The following graphe describe the approach to critical:

Figure 1- ICRR vs Sequential Rod Withdrawal Figure 2- ICRR vs Time from Start of Dilution y Figure 3- RCS Boron Concentration vs Time 6-3

ST-16 APPROACH TO CRITICALITY FIGURE 1 INVERSE COUNT RATE RATIO (ICRR) 1-0.95 -

,[\

0.9 --

0.85 - - l-0.8 -

[

0.75 - -

i c 0.7 -

l' 0.65 -

A 0.6 ' ' ' ' ' ' I N i-0 1 2 3 4 5 6 7 8 9 SEQUENTIAL ROD WITHDRAWAL, NO. BANKS

=

SOURCE RANGE: N31 l SOURCE RANGE: N32 6-4

_ - . _ _ _- -_ _ - _ _ __- ___--_-__________________r

ST-16 ICRR VERSUS TIME (DILUTION)

FIGURE 2 INVERSE COUNT RATE RATIO (ICRR) 1.2 1~ :

0.8 -

= .

0.6 -

x 0.4 -

RENORMAllZATION 0.2. -

0 ' '

1000 1100 1200 1300 1400 1500 1600 1700 TIME, STARTING AT 1000

=

SOURCE RANGE: N31 1 SOURCE RANGE: N32 6-5

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

~4

}"

9  ;

1 ST-16 RCS BORON CONCENTRATION VS TIME  !

FIGURE 3 RCS BORON CONCENTRATION (PPM) 1500 -

P 1400 -

t 1300 -  !

1200 -

~

1100 ' ' ' ' '

1000 1100 1200 1300 1400 1500 1600 1700 TIME, STARTING AT 1000 6-6

l 6.3 ST-17. BORON ENDPOINT MEASUREHENTS Obiective l The objective of the boron endpoint measurements was determination of I critical RCS boron concentrations for the following - control rod endpoint configurations:

All Rods Out (ARO) U control Bank D Fully Inserted Control Banks D and C Fully inserted Control Banks D C, and B Fully Inserted Control Banks D C B, and A Fully Inserted All Rods Inserted, Less " Stuck Rod" H10 The boron endpoint measurement is described in FSAR, Section 14 . Table 14.2-5.. Sheet 20.

Discussion Critical. conditions were-established by boration or dilution for a rod configuration as close to the endpoint configuration as possible. This.

- generally means that one rod bank is a few steps out of position. The procedure requires that the reactivity change expected from the motion of '

the rod bank from'its initial position to the desired endpoint configuration-be less than 50 pcm. If not, then additional boron concentration adjustment is required. When the RCS boron concentration had stabilized, samples were taken, and appropriate rods moved to attain the desired configuration. The reactivity computer measured the necessary reactivity change, and this change was then converted into an equivalent boron concentration, and the RCS boron concentration - adjusted appropriately. The adjusted value is: the boron endpoint for the particular rod configuration.

Results

-The acceptance criteria were met.

Boron endpoints measured in this test were required to agree with corresponding predicted endpoints from the Westinghouse Nuclear Design Report (NDR) within the following limits:

1. Measured boron endpoint within +/- 50 ppm of predicted.

2. Measured ARO value within +/- 1000 pcm of predicted.

Results of the boron endpoint tests are given in Table 1.

6-7 t

.o f,

e i; 6~.3 ST-17. BORON' ENDPOINT MEASUREMENTS (Continued)

TABLE-1. BORON ENDPOINT TEST RESULTS Bank Measured Predicted Absolute Absolute Boron Worth Configuration (ppm) (ppm) Error (ppm) Error (pcm). (pcm/ ppm) ,

ARO 1173 1150 23 258.5 -11.24 CBD at 0- 1113 1092 21 239.6 -11.41 i CBD.C'at 0 1018! 998 20 225.2 -11.26 CBD.C.B at 0 917 901 16 185.8 -11.61

• CBD,C.B.A at 0 825 801 24 270.0 -11.25 ARI, Less H10 518 490 28 346.1 -12.36 ,

l i

i L

i

- l 6-8 I 1 I, 1

! 1 5

l' s' 6.4 57-18. ISOTHERMAL TEMPERATURE COEFFICIENT L Obiective

> The objective of thic procedure was determination of the Isothermal Temperature Coefficient (ITC). From this result, the value of the Moderator Temperature coefficient (MTC) was then calcu?.ated The ITC measurement is described in FSAR, Section 14. Table 14.2-5 Sheet 21.

Discussion The Isothermal Temperature Coefficient is the reactivity change per degree temperature change measured under conditions such that the fuel and moderator are at essentially the same temperature. The test, therefore, is conducted at zero power with slow temperature variations ~ to ensure isothermal conditions. The reactivity computer measures the reactivity change as Tava is slowly raised or lowered and the ratio, reactivity change /

temperature change, is the ITC. Subtracting a predicted value of the fuel temperature coefficient (FTC) from the measured ITC, yields the value of HTC.

Using the atmospheric steam dumps (ASDVs) to adjust steam flow, a cooldown of approw.imately 3,F was initiated at a rate of approximately 10 'F/ hour. Then, a heatup was produced. The chart recorder on the reac.

tivity computer was configured to record Tavg, reactivity and flux. In a few instances, manual rod motion was used to maintain flux level within the reactivity computer range. Three heatups and cooldowns were used to find an average ITC for a given condition. The FTC was then subtracted from the ITC to yield the MTC.

ST-18 was performed at three configurations:

1. All rods out (ARO)

2. Control Bank D Inserted

3. Control Banks C and D Inserted Predicted values of ITC vere adjusted to match actual measurement conditions.

An initial condition required RCS and prestariser boron concentrations to differ by less than 20 ppm, and the boron concentration was maintained essentially constant during the test sequences.

i

(

6-9 i

i y

• t 6.4 $7-18. ISOTHERMAL TEHPERATURE COEFFICIENT (Continued) t Results l The acceptance criteria, agreement within +/- 3 pcm/'F between the measured value of ITC and the predicted value (Westinghouse Nuclear Design Report) (NDR), was met. The results are tabulated in Table 1.

A positive value for MTC in the ARO configuration was anticipated,  ;

based on the NDR. The extrapolated value of HTC for the ARO condition is 40.290 pcm/'F. This is the value used to meet TS 3.1.1.3a, which requires that withdrawal limits be established in this event. Rod withdrawal limits were prepared and are included as Figure 1.

TABLE 1. PREDICTIONS AND RESULTS f

Configuration ITC (pcm/*F) MTC (pcm/*F)

Nominal / Actual PPH Boron Predicted

• Measured Etiegittgi ARO CDt206 steps 1174 -0.99 -1.432 +0.278 CD In CDt29 steps 1114 -1.78 -2.267 -0.557 CC.CD In CBt206 steps 1017 -5.70 -6.884 -5.174

• Predicted ITC was adjusted to conform to the actual test ,

conditions.

i

[

s 6-10

4 l

.' \

l l

ST-18 CONTROL BANK D OPERATING BAND  !

FIGURE 1

) ,

BANK D POSITION (STEPS) l f i 220 -

200 -

180 -

OPERATING BAND

]

160 -

EFFECTIVE FOR 2600 MWD /MTU l

140 -

1 120 -

100 -

1 80 - '

60 -

i 40 -

20 - '

0 '! ' ' ' '

O 10 20 30 40 50 60 70 80 90 100 .

POWER (% RTP)

=

+ MTC W/D LIMIT l- ROD INSERT LIMIT l

6-11

I

• 6,5 ST-19 FLUX DISTRIBUTION MEASURDiENTS AT LOW POWER Obiective Procedure ST-19 provided instructions for obtaining flux distribution data at Icw power levels (zero power physics test range) using the Moveable Incore Detector System (MIDS). l I

The procedure is described in FSAR Section 14. Table 14.2 5 Sheet 22. 1 i

Discussion l 1

Flux distribution measurements, using station operating procedures, i were made at the following rod configurations, with the reactor at steady i state at approximately 12 poweri l

1. All rods out (ARO) i l

2. Control Bank D at 0 steps

3. Control Banks at Hot Zero Power (HZP) Rod Insertion Lir?.ts

4. Pseudo Ejected Rod (During ST-21)

The first three rod configurations were established in ST-17. Boron Endpoint >

Measurements, and the fourth in ST-21. Pseudo Rod Ejection Test.

Results Acceptable pow;- distribution measurements were obtained for each of the configurations lasted. However, quadrant power tilts of about 2-32 were observed for the ARO map. Af ter evaluation, it was concluded that the tilts  ;

observed do not pose any nuclear safety or Technical Specification concerns.

For tilts between 22 and 42 at low power, the Westinghouse guidelines require no actions other than evaluation and vendor notification.

Data analysis was more complex than normal due to failure of the 'A' and 'C' incore detectors. The traverses normally made by the 'A' and 'C'

• detectors were made by other detectors using the ' Emergency' mode.

A second problem was failure of the computer incore automatic grid alignment scheme to align several traces. A visual inspection of the trace '

data showed that the grid depressions were displaced, so the necessary l corrections were inserted for the traces. Also, an incorrect scale setting of zero was noted in some traces.

l l 6-12 1

F

' 6.5 ST-19. FLUX DISTRIBUTION MEASUREHENTS AT LOW POWER (Continued)

Af ter the noted deficiencies in the computer analysis were corrected, i the calculated results for the map data based on the modifications were '

satisfactory. Values for Fry, Fq, and FdeltaH are consistent with core conditions when the maps were tkken. Table 1 lists values obtained for the control bank positions measured.

The acceptance criteria:

Reactor power distribution (by assembly) is

+/- 102 for relative power >0.9  ;

+/- 152 for relative power <0.9

, of the design value (NDR). The criteria were met for all core configurations after the corrections noted.

TABLE 1. FLUX MAP PARAMETERS AT LOW POWER ARO D in H2P RIL D12 Ejected E Control Bank D t 217 Dt0 D40 D40 '

Position C 6 228 C t 216 C 4 51 C 4 $4 B 4 228 B t 228 B 4 166 B t 169 A 4 228 A 4 228 A $ 228 A t 228 Fxy (Max) 1.91 1.90 2.16 6.07

'Fq (Max)* 2.54 2.58 3.13 7.92 FDeltaHN 1.55 1.58 1.73 4.62 Core Average Axial Offset 2.52 1.93 -29.93 -26.86 t

Max Quadrant '

Power Tilt ** 1.03 1.02 1.03 2.16

• Includes engineering and nuclear uncertainty

• • Normalized to 1.000 6-13

n .

'o - 6,6 ST-20. CON?ROL ROD WORTH MEASUREMENTS j

Obiective i

The objective of this procedure is determination of differential and integral rod worth of individual control rod banks. Data generated by ST-20 l provided necessary information to confirm adequate shutdown margin j requirements, and verify core design analysis. i Control Rod Vorth Measurements are described in FSAR, Section 14, Table ,

14.2-5. Sheet 23.

Discussion In the zero power testing range, reactivity effects due to temperature changes and xenon are very small. Control rod worths can be measured with precision by withdrawing or inserting rods or banks in small increments and ,

maintaining power in the zero power range by compensating with boron '

addition or dilution. The reactivity worth of each sinall increment of rod movement is measured by.the reactivity computer and chart recorder.

Starting with an initial condition of ARD, Control Banks CBD, CBC, CBB '

and CBA were moved to the fully inserted position by boron dilution. The dilution was then continued until Shutdown Banks SBE, SBD, and SBC were fully inserted. To reach the ' stuck rod' configuration, the most reactive rod, H10, was next fully withdrawn from CBA, and the dilution continued until the remaining Shutdown Banks, SBB and SBA were inserted. Shutdown margin requirements prevented full intertion of all control rods.

To establish a rod configuration for measuring control banks in overlap, the reactor was tripped, inserting H10, and the shutdown banks '

borated back to the fully withdrawn position. Worth of the control banks in overlap was then measured by further boration.

An optional section of the procedure, rod worth by rod swap, was conducted using CBC as the reference bank. The reference bank was then swapped with the other control banks.

J 6-14

i I

c 6.6 ST-20. CONTROL ROD VORTH MEASUREMENTS (Continued) i Results i

Acceptance criteria required that each RCCA bank measured agree with  !

predicted values from the Westinghouse NDR to within +/- 152 of the total i predicted value, or +/- 0.12 delta rho, whichever is larger. In addition. l the total' measured RCCA bank worth, less the ' stuck rod', must agree with ,

the NDR predicted value to within +/- 102. These acceptance criteria were  !

met.

]

The test was conducted without any major difficulties: all measured rod worths were well within the acceptance criteria.

A comparison of the various configurations measured with predicted values in shown in Table 1. Graphical comparisons of measured and predicted differential and integral rod worths are included as Figures 1 through 14.

Curves for SBB and SBA are not included because predicted data was not  !

available for the rod configurations where measurements were made. Figure 15  !

graphs control bank worths in overlap.

The acceptance criteria inadvertently called for a comparison between measured and predicted values of rod worth in the final configuration (ARO Less the ' Stuck Rod"). The NDR did not include a predicted value for this configuration. After consultation with Westinghouse concerning this omission, and concurrence by the representative, a Field Change was written to delete the comparison requirement for that particular configuration.

As noted in the Discussion, an optional rod swap section was included in ST-20. The measurements were taken and data analysis performed.There was no acceptance criteria for this section.

6-15

D' 6,6 ST-20. CONTROL ROD WORTH MF4SUREMENTS (Continued) h l

TABLE 1. ROD WORTH RES,g(IE i Bank Configuration Measured Predicted Relative .

Error (pcm) (pcm) I of Total D ARO 669 645 -3.7 '

C D Inserted 1052 1045 -0.7 }

B D & C Inserted 1152 1116 -3.2 A D, C & B Inserted 1061 1090 +2.7 l t

SE D, C, B & A Inserted 486 535 +9.2 SD D, C B A & SE 680 673 -1.0  !

Inserted i

SC D. C. B. A. SE & SD 1021 1010 -1.1 Inserted SB D C, B, A. SE, SD 1348 N/A N/A  !

& & SC Inserted '

SA (Less " Stuck Rod')  !

All Rods Measured Above 7469 7705 +3.1 A+B' All Shutdown Banks 3919 3913 -0.2

+C Out (Control Rode

+D in_0verlap)  ;

Average Boron Worth -0.0881 ppm /pem b

k 6-16 a

~

ST-20 CBD INTEGRAL WORTH '

FIGURE 1 INTEGRAL ROD WORTH (PCM) 700 600 k.g 500 -

f i

400 -

l

\. '

300 -

~

200 -

\  :

100 - '

' ' ' ' ' ' ' ' '= '

0 O 20 40 60 80 100 120 140 160 180 200 220 240 -

ROD' BANK POSITION STEPS WITHDRAWN

=

NDR PREDICTED VALUE l MEASURED VALUE 6-17

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

n e.

ST-20 CBD DIFFERENTIAL WORTH FIGURE 2 I

DIFFERENTIAL WORTH (PCM/ STEP) 7i j 6 -

5 - -

4 - '

r 3 -

2 -

O '

=

0 20 40 60 80 100 120 140 160 180 200 220 ROD BANK POSITION STEPS WITHDRAWN l

' =

NDR PREDICTED I MEASURED 6-18

a-a&_a- a A>as---. 4m- - - ,, ma r # ,,_J. . . _ . - . .__a m .4 - n.- A A

\

.o s * '

i ST-20 CBC INTEGRAL WORTH  ;

FIGURE 3 INTEGRAL ROD WORTH (PCM) '

L 1000 -

800 -

600 -

400 -

200 -

N, ,

g , , , , , , , , N. =

O 20 40 60 80 100 120 140 160 180 200220240 '

ROD BANK POSITION STEPS WITHDRAWN

=

NDR PREDICTED l MEASURED 6-19

.i

n i i

i ST-20 CBC DIFFERENTIAL WORTH l FIGURE 4 i 9

DIFFERENTIAL WORTH (PCM/ STEP)  !

1.1  !

t 10 -

9 -

8 -

7 -

P 6 -

5 -

4 -

3 -

2 -

1 0" ' ' ' ' ' ' ' ' ' '

O 20 40 60 80 100 120 140 160 180 200 220 '

ROD BANK POSITION STEPS WITHDRAWN 4

=

NDR PREDICTED l MEASURED 6-20

' 5 k

ST-20 CBB INTEGRAL WORTH FIGURE 5 '

INTEGRAL ROD WORTH (PCM) 1200 '

1000 -

800 -

~

600 -

400 -

200 -

0 x

' ~' = '

O 20 40 60 80 100 120 140 160 180 200220240 .

ROD BANK POSITION STEPS WITHDRAWN

=

NDR PREDICTED l MEASURED 6-21

i ST-20 CBB DIFFERENTIAL WORTH l FIGURE 6 ,

DIFFERENTIAL WORTH (PCM/ STEP) 12 11 -

10 -

9 -

8 -

N 9 7 -

6 -

5 -

4 -

3 -

2 -

i i , i i i i i i ,

0 0 20 40 60 80 100 120 140 160 180 200 220 ROD BANK POSITION STEPS WITHDRAWN

=

NDR PREDICTED MEASURED 6-22

I i

ST-20 CBA INTEGRAL WORTH FIGURE 7 INTEGRAL ROD WORTH (PCM) 1200 )

).

1 1000 -

.. i

\ >

800 -

x 600 -

~

400 -

200 -

\

0 O 20 40 60 80 100 120 140 160 180 200 220 240 ROD BANK POSITION STEPS WITHDRAWN '

i

=

NDR PREDICTED l MEASURED 6-23 I

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

i ST-20 CBA DIFFERENTIAL WORTH FIGURE 8  ;

DIFFERENTIAL WORTH (PCM/ STEP)

.12  :

11 -

10 -

[ ,

9 -

\ j 6 -

i 5 -

4 -

i

~

3 -

2 -

0 " .- -

O 20 40 60 80 100 120 140 160 180 200 220 ROD BANK POSITION STEPS WITHDRAWN

=

NDR PREDICTED l MEASURED -

6-24

1 i

I ST-20 SBE INTEGRAL WORTH l FIGURE 9 l INTEGRAL ROD WORTH (PCM) )

550  !

500 -

450 -

400 -

~

350 -

300 - '

250 -

200 -

150 -

100 -

50 -

0 =

0 20 40 60 80 100 120 140 160 180 200 220 240 ROD BANK POSITION STEPS WITHDRAWN

-~ NDR PREDICTED l MEASURED 6-25

ST-20 SBE DIFFERENTIAL WORTH  !

FIGURE 10  !

DIFFERENTIAL WORTH (PCM/ STEP) i i

4 -

i i

i

+

/

2 -

l I i

1 <

> +

0 "'/ ' ' ' ' ' ' ' ' ' ' '

0 20 40 60 80 100 120 140 160 180 200 220 1 i

ROD BANK POSITION STEPS WITHDRAWN NDR PREDICTED \

=

! MEASURED i

I 6-26 i

c- i r

I ST-20 SBD INTEGRAL WORTH FIGURE 11 i INTEGRAL ROD WORTH (PCM)  ;

700 j 650 -

600 -

550 -

,~

500 -

450 -

l 400 -

t 350 - '

300 250 -

200 -

150 -

\ ,

L 100 -

50 l ' ' ' ' ' ' ' ' ' '

0 =

0 20 40 60 80 100 120 140 160 180 200 220 240 i ROD BANK POSITION STEPS WITHDRAWN

=

NDR PREDICTED MEASURED 1

6-27 l l

i i

ST-20 SBD DIFFERENTIAL WORTH ,

FIGURE 12 i i

I DIFFERENTIAL WORTH (PCM/ STEP) 7 i s  :

6 -

,- ,4 -

5 -

l

• i 4 -

3 -

2 -

1 '

0'

/ ' ' ' ' ' ' ' ' ' ' '

O 20 40 60 80 100 120 140 160 180 200 220 ROD BANK POSITION STEPS WITHDRAWN  ;

P

=

NDR PREDICTED l MEASURED 4

6-28

'O ST-20 SBC INTEGRAL WORTH FIGURE 13 INTEGRAL ROD WORTH (PCM) 1000' 900 800 -

700 -

600 -

500 -

400 -

300 -

~

200 -

, , , i i , , .,

O .

~ 0 20 40 60 80 100 120 140 160 180 200 220 240 ROD BANK POSITION STEPS WITHDRAWN

~

NDR PREDICTED l MEASURED 6-29

.. . .~ .

i i

ST-20 SBC DIFFERENTIAL WORTH  :

FIGURE 14  :

i DIFFERENTIAL WORTH (PCM/ STEP)  ;

10 ,

[g_  !

9 -

4/

8 -

7 -

i 6 -

[

5 -

4 -

N 2

O O

20 40 60 80 100 120 140 160 180 200 220 A  ;

ROD BANK POSITION STEPS WITHDRAWN

=

NDR PREDICTED l MEASURED 6-30

L ,

i

-Q 1

ST-20 CONTROL BANK WORTH IN OVERLAP l

[ FIGURE 15 l

.q INT. WORTH (PCM) DIFF. WORTH (PCM/ STEP) 4000 N ,

N, -

14 l 3500 -

\

12 l 3000 - \ '

'T \~ -

10 2500 * . ,

8 2000 -

i 1500 -

6 1000 -

4

, 4 500 -d N, 2

0 ' '

O O 50 100 150 200250300350400450500550 RELATIVE BANK POSITION IN STEPS

=

INT. WORTH (PCM) l DlFF. WORTH (PCM/STP) l l

t 6-31 7 i

I 6.7 ST-20.1. ADDITIONAL CONTROL ROD VORTH MEASURDENTS Obiective The objective of this procedure was acquisition of additional control rod worth data for evaluating rod shadow effects. The reactor core conditions during low power physics testing are particularly suited for modeling and validation of computational techniques.

Procedure ST-20.1 is not described in the FSAR.

Discussion The integral / differential worth of control banks in the following configurations was measured using the-techniques indicated:

Initial condition, ARO.

1. Worth of CBB and CBC by dilution.

2. Rod Swap of CBA (Insert) for CBB (Withdraw).

3. Worth of CBC and CBA by boration.  ;

Results The data was ' collected as described in the procedure. There were no acceptance criteria.

}

6-32 i

l

+ 6,8 ST-21. PSEUDO ROD EJECTION TEST  !

Obiective The objective of this procedure was determination of the rod worth and core power distribution resulting from simulated ejection of the highest worth rod. The data collected verified the conservatism of the ejected rod analysis. The core power distribution required a full core flux map, taken when the rod was fully withdrawn, and utilized $7-19 Flux Distribution Measurements at Low Power.

The procedure is described in FSAR Section 14. Table 14.2 5, Sheet 24. 3 Discussion i An initial flux map was taken with the control banks at. the hot zero power rod insertion limits. The highest worth rod, D12 in Bank CBD, was then withdrawn in increments of approximately 50 stepe, with boration maintaining stable reactor power. At each incremental step, a partial core flux map was -

made, using station operating procedures. The reactivity corresponding to ,

each rod movement was measured using the reactivity computer. When D12 was fully withdrawn, a full core flux map was taken per ST-19. As the rod was I reinserted by dilution, an optional set of rod worth data was also taken.

Measured rod worth and core power distribution were compared to '

specifications detailed in the Vestinghouse Nuclear Design Report (NDR).

Results  ;

To meet the acceptance criterion, analysis of the data must yield an

' ejected rod' worth and hot channel factors that. are conservative with respect to FSAR analysis values. The acceptance criterion was met.

Results are summarized as follows:

Expected FSAR Hessured Analysis Worth of Ejected Rod, pcm 545 860 489.7

• Power Distribution. Fq, 8.50 13.00 7.92

• Additional Power distribution results are shown in ST-19 Power Distribution Hessurements at Low Power.

l The nearest moveable detector path to D12 (Ell) shows the shift in detector output as the rod is withdrawn. This observation, illustrated in

. Figure 1, is verification that the system is capable of indicating rod misalignment.

The ejected rod worth results reported are based on measurements taken during withdrawal of D12. During measurements taken for insertion (optional),

problems were encountered with the recorder trace, resulting in come questionable data. The data was analyzed, and demonstrated a rod worth consistent with the reported results, but was not used. A test exception was written to omit consideration of the questionable data.

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ST-21 PSEUDO EJECTED ROD TEST  !

FIGURE 1 l CORE POSITION (TOP = 61) 61 ,

i 56 -

51 -

46 -

41 -

36 -

I 31 -

26 -

21 -

m  :

16 -

11 -

6 -

i

. i. . , , , i 3

0.1 0.15 0.2 0.25 0.3 0.35 0.4 0.45 0.5 RELATIVE FLUX TRACE LOCATION B, E-11 l- D12 AT 228,106 AND 83 STEPS 6-34 1 -

i 8 6,9 ST-22. NATURAL CIRCULATION TEST  ;

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

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

Discussion '

Natural circulation requires residual heat in the reactor core to >

establish convection flow. ST-22 was scheduled in the low power physics test sequence, with the necessary core heat provided by maintaining reactor power at approximately 32.

The natural circulation characteristics to be determined were:

1. Time to stabilize natural circulation.

2. Reactor coolant flow distribution.

3. Depressurization rate following loss of pressurizer heaters.

4. Depressurization rate using auxiliary spray.

5. Effect of charging flow and steam flow on the subcooling margin.

6. Subcooling monitor performance.

The test was initiated by tripping all reactor coolant pumps with reactor power stabilized at approximately 32. The steam dumps were manually operated "

in the pressure control mode to maintain cold leg temperature constant.

Pressurizer pressure and level were also in manual, maintaining pressure stable and level in the range 25 to 302. The plant was expected to be stabilized under natural circulation conditions in 10 to 20 minutes.

l Results System behavior during the test is shown in the following figures:

L Figure 1. Loop 1 Thot Teold, and Tavg versus time.

l Figure 2 Loop 1 Delta T versus time.

l 1

'In Figure 1 hot leg temperature started rising and then leveled off, cold leg decreased and began to recover. In this portion of the transient, a steam dump valve went to the full open position, due to failure of the control linkage. The abnormal rate of cooling ultimately led to a manual reactor trip. Initial system behavior was as expected, but because the test was terminated, it will be rescheduled into the Power Ascension Test sequence.

l 6-35 1'

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ST-22 NATURAL CIRCULATION TEST FIGURE 1 1 TEMPERATURE, DEGREES F  :

570 '

560 -

550 - '

N 540 -

'J 530 - '

520 ' ' ' '

74 79 84 'I 89 94 RELATIVE TIME IN MINUTES -6 L- =

LOOP 1,TCOLD(WR) l LOOP 1,THOT(WR) '

• LOOP 1,TAVG  !

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l. ,

NOTE: RCP TRIP,79 REACTOR TRIP, 96 '

4 6-36

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ST-22 NATURAL: CIRCULATION TEST l FIGURE 2 l l

TEMPERATURE, DEGREES F 35 30 -

25 - '

20 -

15 -

10 -

5 -

0 _

7

-5 74 79 84 89 94 L- RELATIVE TIME IN. MINUTES L =

DE LTA-T l

j. NOTE: RCP TRIP, 79 REACTOR TRIP, 96 4 6-37

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.= m ,

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s . 7.0 INSTRUMENT CALIBRATION AND ALIGNMENT-7.1 Instrument Calibration and Alinnment Sn - ry The procedures included in the calibration and alignment test sequence

, ensure that protective systems can perform their planned safety functions if

. called upon, and that measurements made on plant parameters are valid and .

. can be used'for comparison to predicted values, t In general, procedures in this category are utilized at a number of test .

plateaus for example, the power range nuclear instrumentation setpoints are '

evaluated and reset, if necessary, at each new power level. Instrument calibration and alignment procedures reported herein include ' .

- Nuclear instrumentation. ,

- Process temperature instrumentation.

- Cross calibration of RTDs and incore thermocouples. 7 p - Plant system setpoints.

l During low power physics testing, the only instrumentation system which ,

failed to operate was the Shutdown Monitor System. Successful operation of ,

this' system was not included in the test acceptance criteria, and its

, performance will be monitored in the Power Ascension Test Program.

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• . 7.2'STI13. OPERATIONAL ALIGNMENT OF NUCLEAR INSTRUMENTATION Ohiective

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

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

Discussion Calibration of' nuclear instrumentation, including alarm settings, trip

. points, and operational ranges cannot be properly completed until the system is functioning on line, in its intended operational ranges. At each test condition, the nuclear instruments are adjusted,_ using the best available conservative information. Initially, setpoint data furnished.by Westinghouse was used, and, as higher neutron fluxes became available, actual measured data was used.

The sections of ST-13 completed and included here are:

1. Prior to core load-1.1 Operational settings of NIs (Source Range) 1.2 Verification of compliance of settings with Technical Specifications

2. Prior to Criticality (Hot Standby)-

2.1 Initial setting of Intermediate Range (IR) trip setpoints.

2.2 Initial settings for each Power Range channel.

2.3 Functional test

• of source and intermediate range NIs.

.2.4 Functional test

• of Power Range channels.

3. Initial Criticality-3.1 Source and Intermediate Range overlap data.

3.2 . Verification of loss of source range voltage when P6 interlock was satisfied.

- 3.3 Intermediate ' and Power Range overlap when Doppler heating was first observed.

• Analog Channel Operational Test (ACOT)

Results The acceptance criterion for the included sections (all data has been obtained) was satisfactorily completed.

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• . 7.3 ST-14;1 OPERATIONAL ALIGNMENT OF PROCESS TEMPERATURE INSTRUMENTATION t

. Obiective The objective of ST-14.1 was to ensure proper alignment of the delta T and Tavg instrumentation channels at all test plateaus.

Additional alignments will be performed, using ST-14.1, at the 30%, 502, 75%, 902-and 1002 power test plateaus.

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

Discussion With the plant in the hot standby condition, temperature stabilized, the TH ot and T Cold R/E Converters, and the TAyg and delta T Summing Amplifiers were aligned using station procedures. Temperatures were then measured, values converted to corresponding engineering units, and temperature and delta T values calculated. Results were checked against allowable tolerances.

ELn9111 The required data was collected at hot standby, and all acceptance criteria were met.

No instrument alignments were necessary, as all data was within specified tolerances.

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7-3

.. 7.4 ST-14'.2. RTD AND INCORE T/C CROSS-CALIBRATION Obiective Procedure ST-14.2 verified the expected resistance vs temperature characteristics of the Reactor Coolant System (RCS) Resistance Temperature

, Detectors (RTDs) during heatup to hot standby conditions. Installation correction factors for individual .RCS RTDs were determined, and data obtained for correction factors for the incore thermocouples. t The procedure is described in FSAR, Section 14. Table 14.2-5, Sheet 17.

Discussion RTD and incore thermoccuple temperature data were taken at three temperature _ plateaus, 250 *F, 350 'F, and 450 'F, and hot _ standby (557 'F).

Af ter correcting the data for test instrument bias, the RTD resistance measurements were converted to temperatures using Westinghouse calibration data..

t Two sets of incore thermocouple data from the main plant computer system were taken concurrently with RTD data when hot- standby conditions '

were reached.

Results All acceptance criteria were met.

Resistance versus temperature values of the RTDs agreed with the overall average RCS temperature, at hot standby, within approximately 0.25 'F, except for three values. The acceptance criteria limit was 1.'2*F, and the maximum deviation was 1.08'F.

The acceptance criteria also specified that preliminary installation l- correction factors be calculated, but need not be included in an instrument alignment unless the correction is >0.3 'F. The three values, noted above, j would have required instrument alignment if their finni correction factors, as determined by Westinghouse, exceeded this limit. . Final composite calibration sheets were provided by Westinghouse which eliminated the need to correct the three RTDs.

The Westinghouse RTD ' Cross Calibration Report verified RTD performance to be consistant with calibrations generated from HFT data. The test results

, provided calibration verifications the composite calibration sheets were l returned as ' final and approved for use". The calibration procedures were revised and recalibrations carried out in October, 1988.

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<7.4 ST-14.2. RTD AND INCORE T/C CROSS-CALIBRATION (Continued)

~

Additional core exit thermocouple- data, not required in the procedure, was collected from the plant computer'by Reactor Engineering during.heatup.,

At the 557 'F test plateau. . thermocouples in core locations C08 and K12 showed larger . deviations from the averaged = values than the balance of the core exit thermocouples. The data has been attached to ST-14.2 as startup test supplemental information..

Incore thermocouple data analysis was not required by ST-14.2. During Power Ascension Testing, additional incore thermocouple data will be taken i as part;.of ST-29, Core Performance Evaluation, and correction factors developed as necessary. ,

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,- t O! 27.5 ST-15. REACTOR PLANT' SYSTEMS SETPOINT VERIFICATION Obiective '

Procedure ST-15 provides verification that the initial setpoint ,

adjustments have been made prior to startup, and serves to document setpoint

. modifications made during startup testing.

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

r Discussion '

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

l

1. Safeguards

2. Reactor Coolant Pumps

3. Nuclear Instrumentation (Excore) 4.

5. DeltaT-.Tavhressure Pressurizer '

6. Cold Overpressurization Mitigation System (COMS) 7.- Pressurizer Level'

8. Charging Flow..

9. Rod Control

10. RCS Flow 11'. Feedwater Flow

12. Steam Generator Level

13. Steam Line Pressure

14. Steam Dump' System

15. Steam Generator Relief Valve Control (ASDVs)

16. Turbine Impulse Chamber Pressure Prior to criticality. ST-15 served as a detailed listing of setpoints for -

plant instrumentation to be compared to the Westinghouse -Precautions, Limitations and Setpoints (PLS) values, and Technical Specifications.

Duri1g low power physics tests, had any setpoint changes'been required, these would have been documented by this procedure. No changes were required.

Any setp & t changes made during power ascension testing will be documented by ST-25, and the final startup test prior to . the 1002 power

- plateau results review will be a verification that any changes made have

.been' incorporated inte both the plant herdware and station procedural text.

Resulte All. acceptance cr!.teria were met.

4 7-6 e

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s., , i 8.0 GENERAL PLANT TESTING-8.1 General-Plant Testina Su==ary .

~ ~ ~

Startup tests which examine components or systems not directly related '

to plant primary! and . secondary system performance, or calibration and alignment, are ' included in this section. Some of the~ tests, or portions thereof . might' reasonably be included in earlier sections, but the bulk of' -

the testing fallss into this category. #

All of the tests included have sections to'be included on various test plateaus in - the; Power Ascension Test Program.- Tests categorized as ' General Plant Testing" are - -

i

- Radiation survey.  ;

- Water che'mistry control. ' '

- Operability of ventilation system.

- Displacement of~ piping and components under transients. L

- Thermal-expansion..

Y

- Piping vibration.

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g Tq . I c 8.2 ST-41. RADIATION SURVEY p

, ..qbiective The objective of this procedure was determination of dose rate levels at l

' selected locations throughout.the plant to-validate shield performance and identify - high radiation levels. Also, survey readings were compared to ,

radiation monitor response to verify operability of radiation monitors.

This procedure is described in FSAR, Section 14. Table 14.2-5, Sheet 44.

Discussion .

Using station operating procedures, a shield survey was' conducted and radiation monitor readings recorded with the reactor at a nominal II power.

Radiation monitor readings were compared to local survey data.

Additional surveys will be conducted at the 50% and 100%_ power test plateaus.

Results Radiation dose rates measured at selected locations, identification of high radiation areas, and agreement between selected radiation monitors and local survey data were-the acceptance criteria. All acceptance' criteria were met.

The survey was completed in a timely fashion; a few points were repeated due to questionable results, but none of the results were unexpected.

Radiation monitors agreed with local survey data very accurately.

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M l8,3ST-42.WATERCHEMISTRYCONTROL Obiective The objective of this procedure was a demonstration that chemical and ,

radiochemical control and analysis - systems maintain primary and secondary water: chemistry within requirements. Samples were taken prior to criticality and at hot zero powers additional data under this procedure ~1s required at

,the 502, 502, 75% and 1002 power test plateaus.

The procedure is described in FSAR, Section 14. Table 14.2-5 Sheet 45.

Discussion The importance of proper primary and secondary water chemistry at nuclear plants has been demonstrated throughout the industry. ST-42 specifies the

.necessary primary and secondary sampling, and the plant conditions, to *

~ insure that water quality is maintained within the required limits. Station

. operating procedures determine the chemical analysis systems and the T 1 appropriate analytical procedures used. ,

/" Prior to heat ty) after initial fuel loading, and again during zero power

• Jtests. ST-42 measurements were taken. Additional ST-42 data sets will Ebe taken during power ascension testing.

i Results J

All results were within acceptable ranges, and met tne acceptance criteria with the following exceptions:

In -Mode- 5 feedwater and blowdown are not operational, and condensate, operating in the recirculation mode, could not provide

. analyzer readings because system water-chemistry exceeded the analyzer ranges. These required readings were-inappropriate for Mode 5 and were excepted by a Field Change. The acceptance criterion, requiring agreement within tolerance of analyzer readings and analysis results, was not met for condensate since thers were no. analyzer readings.

m In the . zero power test, a secondary oxygen sample could not be taken, since a continuous steady flow for 24 hrs. was a requirement. A

. Test Exception was prepared to address the problem.

J 8-3 c a :, ;)

1 8.4 ST-45. PROCESS-AND EFFLUENT RADIATION MONITORING SYSTEM Obiective d The objective of this procedure is proper operation and response of the process and effluent radiation monitoring system. The testing-demonstrates  !

that the process and' radiation monitors respond as required to the actual 1 activity within their respective systems.

During low power physics testing, data was taken for ST-45 at 12 reactor

' power. In the Power Ascension Test Program, additional data will be taken at the 50% and 100Z test plateaus. .

This procedure is described in FSAR, Sections 11.5 and 12.3.4, and Section 14 Table 14.2-5, Sheet 48.

Discussion Process and effluent' radiation monitoring utilizes a monitoring system, the -Radiation Data Management System (RDMS). The RDMS is designed to continuously monitor selected process and effluent streams wherever the potential for a significant release of radioactivity exists during normal operations and during postulated accidents. l l

With reactor power steady at it, simultaneously, a sampling of process streams and acquisition of RDMS readings was obtained. Following radiochemical analysis, the results of the analyses were compared to the RDMS displayed values.-

Results Acceptance criteria were met by the test.

The values resulting from analyses compared favorably with RDMS displayed values in all cases except two locations monitoring' letdown coolant. In the two cases, monitors ~ indicated values 100 times the values from analysis.

Calibration and geometry of the monitors was verified, and it was concluded that short lived. isotopes,- present in the samples, had decayed before radiochemical analyses were completed.

The acceptance criteria specifically accepted such values as long as a

" reasonable explanation" could be applied. Subsequent performances of this test _at higher power levels will monitor this concern.

The monitor dedicated to Waste Liquid from the Evaporators RM-RM-6514, was out of service due to a temporary modification in progress. Performance of this monitor'will be verified at subsequent tests.

8-4

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8.5 ST-51. POWER' ASCENSION DYNAMIC TEST Obiective The objective of this procedure was measurement of the dynamic response

- o' certain main steam, feedwater and. pressurizer relief systems under transient conditions.

Dynamic testing is described in' FSAR, Section 3.9(B).2, and Table 3.9(B)-1.

Discussion Displacement transducers, installed on the piping and components to be monitored, measure the behavior of the system during specified transients.

The data is recorded on GETARS for later analysis.

During the precritical test program, dynamic testing of the pressurizer relief system, individual operation of the condenser steam dump and atmospheric steam dump valves, and trip of the emergency feedwater pump Terry Turbine was conducted under ST-51.

The feedwater pump dynamic testing was performed in conjunction with ST-53, Turbine Driven Emergency Feedwater Pump Start Verification. The steam dump portion was performed in conjunction with ST-55, Steam Dump System Test.

The Power- Ascension Test Program includes further dynamic response testing, coordinated with ST-38 Unit Trip from 100% Power. The transients

.to be monitored are:

1. Main Steam System- Turbine trip, and simultaneous operation of the condenser steam dump valves.

2. Feedwater System- Closure of the feedwater containment isolation valve and trip of the steam generator main feedwater pump.

Results The acceptance criteria was a review and verification by New Hampshire Yankee Engineering that the measured stresses do not exceed code limits. All

' measurements met this criteria.

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.E 8.6 ST-52. POWER ASCENSION THERMAL EXPANSION TEST Obiective The objective of this test. was a demonstration that certain piping systems, not monitored during hot functional testing, were free to thermally expand consistent with design. These measurements confirm that associated restraints and supports allow the required thermal movement. Also, systems were remonitored where, during hot functional testing, movement inconsistent with design was found.

The-test is described in FSAR, Section 3.9.3.4.d and Section 14, Table 14.2-3, Sheet 6.

Discussion Thermal expansion data was obtained from displacement measuring transducers and by visual observation of spring hangers and snubbers.

Walkdowns were performed to identify areas of potential restraint to free movement.

The following systems were monitored:

1. Snubbers:

Auxiliary Steam Condensate Primary Component Cooling Chemical and Volume Control Diesel Generator Feedwater Main Steam--

Main Steam Drains Nitrogen Gas Reactor Coolant Residual Heat Removal Steam Generator Blowdown Spent Fuel Pool Cooling Safety Injection Service Water Waste Processing Liquid Drains

2. Spring Hangers:

Condensate Extraction Steam

,~m ~

Feedwater Heater Drains Main Steam Main Steam Drains Moisture Separator & Pahe4ter Drains / Sampling System 8-6 o

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- 8.6 ST-52. ' POWER ' ASCENSION THERMAL EXPANSION TEST (Continued) ,

3. Final Piping Movement Data:

Condensate Extraction Steam Feedwater

• Heater Drains i Moisture Separator and Reheater Drainc Main Steam Main Steam Drains Steam Seal System

Reactor Coolant Residual Heat Removal

4. Pipe Whip Restraint Gaps:

Feedwater Data was taken during heatup, in the precritical testing phase. With systems at ambient temperature (<100 *F), baseline data was taken; next, at approximately 350 'F, but prior to isolation of RHR, measurements were repeated, and finally, at hot standby (approximately 557'F), repeated again.

Adjustments were made, during the test sequence, to spring hangers which -

were not within their hot and cold settings.

During Power Ascension Testing, the following systems will be monitored.

under ST-52, at <200 'F, 557 'F (Non-critical), S0I, 50Z, 75Z and 100Z power test plateaus:

1. Main steam from MSIVs to moisture separators and turbines.

2. Feedwater system from feedwater pumps to steam generators and steam generator recirculation and wet-lay-up pump discharge.

-3. Selected condensate system.

4. Steam extraction system.

5. Heater lines to heater drain pumps suction and discharge to

-feed pumps suction.

6. _ Main steam dump drains, moisture separator drains and vents, steam seal system and the turbine crossaround piping.

7. Selected snubbers on steam generator blowdown and service water.

Results l.

Acceptance criteria for thermal movements were specified for Westinghouse (NSSS) scope, for UE&C scope, and for completion of NHY Engineering review.

The results met all acceptance criteria.

There were no test exceptions.

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At 0}I 8.7 ST-56. PIPING VIBRATION TESTING H

gj Obiective ,

The objective of this procedure is acquisition of vibration response data' for certain piping systems when the plant is. operating at 1002 power, steady-state. The systems'are:

1. Condensate System, Two Pumps Operating

2. Feedwater Pumps Operating n 3.'Feedwater Pumps Operating in the Recirculation Mode

4. Main Steam System n'

Measurements, with feedwater pumps operating in the recirculation mode, were made-in the precritical test sequence. .'

Vibration measurements are included in FSAR, Section 3.9(B).2 and Section 14, Table 14.2-3, Sheet 5.

Discussion The plant was-in the hot standby condition, with normal flow rate for i feedwater recirculation. Vibration data was obtained with the use of hand -

held vibration meters. Prior- to the test, the selected test points were located to verify accessibility, and any insulating material that would obstruct direct contact by the vibration meter removed.

Measured vibration was compared to an acceptable vibration amplitude for each coordinate considered.

Results

[

All acceptance criteria were met.

Feedwater pump _ lines were monitored three points were tested and the-m maximum vibration observed was 11.5% of the allowable vibration.

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