Text

Cycle 11, Startup Test Report Revision 1
	ML022700025
Person / Time
Site:	Harris
Issue date:	09/19/2002
From:	Caves J; Carolina Power & Light Co
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	HNP-02-124
	Download: ML022700025 (46)
	v • d • e

k CP&L A Progress Energy Company SERIAL: HNP-02-124 SEP 1 9 2002 United States Nuclear Regulatory Commission ATTENTION: Document Control Desk Washington, DC 20555 SHEARON HARRIS NUCLEAR POWER PLANT DOCKET NO. 50-400/LICENSE NO. NPF-63 CYCLE 11, STARTUP TEST REPORT REVISION 1

Dear Sir or Madam:

In accordance with Technical Specifications 6.9.1.1 for the Harris Nuclear Plant (HNP), on April 1, 2002, Carolina Power & Light Company (CP&L) submitted the startup test report for the HNP following steam generator replacement and power uprate modifications performed during Refueling Outage No. 10. The startup test report identified two remaining open testing items. On July 1, 2002, CP&L submitted a letter to the NRC providing a status of the two open testing items. The two remaining testing items are now complete. This revision also revises section 1.2 regarding the number of fuel assemblies stored in the spent fuel pool during Cycle 10 but used in the Cycle 11 core. Enclosed is the Harris Nuclear Plant Unit 1, Cycle 11 Startup Test Report, Revision 1.

If you should have any questions regarding this submittal, please do not hesitate to contact me at (919) 362-3137.

Sincerely, John R. Caves Supervisor, Licensing/Regulatory Programs Harris Nuclear Plant MGW c: Mr. J. B. Brady (NRC Senior Resident Inspector, HNP)

Mr. L. A. Reyes (NRC Regional Administrator, Region II)

Mr. R. Subbaratnam (NRC Project Manager, HNP)

Harris Nuclear Plant 5413 Shearon Harris Road New Hill, NC 27562

Harris Nuclear Plant Unit 1, Cycle 11 Startup Test Report August 30, 2002 CAROLINA POWER AND LIGHT COMANY 8/30/02 Revision I

Executive Summary The Harris Technical Specifications provide the following guidance for conditions specifically requiring a startup report and items that should be addressed in the startup report.

STARTUP REPORT 6.9.1.1 A summary report of plant startup and power escalation testing shall be submitted following: (1) receipt of an Operating License, (2) amendment to the license involving a planned increase in power level, (3) installation of fuel that has a different design or has been manufactured by a different fuel supplier, and (4) modifications that may have significantly altered the nuclear, thermal, or hydraulic performance of the unit.

The Startup Report shall address each of the tests identified in the Final Safety Analysis Report and shall include a description of the measured values of the operating conditions or characteristics obtained during the test program and a comparison of these values with design predictions and specifications. Any corrective actions that were required to obtain satisfactory operation shall also be described. Any additional specific details required in license conditions based on other commitments shall be included in this report.

Startup Reports shall be submitted within: (1) 90 days following completion of the Startup Test Program, (2) 90 days following resumption or commencement of commercial power operation, or (3) 9 months following initial criticality, whichever is earliest. If the Startup Report does not cover all three events (i.e., initial criticality, completion of Startup Test Program, and resumption or commencement of commercial operation), supplementary reports shall be submitted at least every 3 months until all three events have been completed.

Reviewing the tests described in FSAR Chapter 14, see Table 4.7.3 and determining the tests impacted by SGR/PUR generated the scope of this report.

The purpose of Revision 1 to this Startup Test Report is to make a correction in section 1.2 and address the two open testing items identified in Revision 0 of the Startup Test Report submitted on April 1, 2002. The open testing items were as follows:

"* EPT-287 Blowdown Flow Versus Differential Pressure (Delta-75 SG) [Reference 5.83]

"* CRC-864T Temporary Procedure For Steam Generator Moisture Carryover Sampling (Expiration Date 12/20/2002) [Reference 5.84]

The revised Startup Test Report impacts pages 2, 3, 5, 19, 21 and 24.

8/30/02 Revision 1

k Harris Nuclear Plant Unit 1, Cycle 11 Startup Test Report Revision 1 Page 3 of 45 Table of Contents Page 1.0 Introduction ....................................................................................................................... 5 1.1 General ................................................................................................................................ 5 1.2 Cycle Description ........................................................................................................ 5 1.3 Steam Generator Replacement ....................................................................................... 5 1.4 Power Uprate ....................................................................................................................... 5 1.5 Tavg Restoration ........................................................................................................ 6 2.0 Summ ary ............................................................................................................................ 6 3.0 Component & Initial Operation Test Summ aries ..................................................... 6 3.1 Protection System Engineered Safety Features Actuation Logic Test ........................... 6 3.2 Reactor Protection System Engineered Safety Features Actuation Response Time Test ... 7 3.3 Auxiliary Feedwater System Test .................................................................................. 8 3.4 Containment Integrated Leak Rate and Structural Integrity Test ................................... 9 3.5 Piping Thermal Expansion and Dynamic Effects Test .................................................... 9 3.6 M etal Impact M onitoring System Test ............................................................................ 9 3.7 Feedwater System Test ................................................................................................... 9 3.8 Steam Generator Primary Side FOSAR Inspection ...................................................... 10 3.9 Steam Generator Secondary Side Internal Inspection ................................................... 10 3.10 Steam Generator Baseline Test ..................................................................................... 10 3.11 Component Cooling W ater System Test ........................................................................ 12 3.12 Safety Injection Flow Balance ........................................................................................ 15 4.0 Operational and Power Ascension Test Summ aries .............................................. 15 4.1 Rod Drop Time M easurement ........................................................................................ 15 4.2 Reactor Coolant System Flow M easurement ................................................................. 15 4.3 Calibration Of Nuclear Instrumentation Test ............................................................... 16 4.4 Flux Distribution M easurement Test ............................................................................. 16 4.5 Core Perform ance Test ................................................................................................. 17 4.6 Power Coefficient M easurement Test .......................................................................... 17 4.7 Control Rod Reactivity W orth Test .............................................................................. 18 4.8 Boron Endpoint M easurement - All Rods Out Test ...................................................... 19 4.9 RTD/TC Cross Calibration Test ................................................................................... 19 4.10 Steam Generator M oisture Carryover Test .................................................................... 19 4.11 Load Swing Test ................................................................................................................ 20 4.12 Reactor Coolant System Leakrate Test ....................................................................... 20 4.13 M ain Steam and Feedwater Systems Test ................................................................... 20 4.14 Plant Perform ance Test ................................................................................................ 21 4.15 DEH Changes .................................................................................................................... 21 4.16 Steam Generator Blowdown Testing ............................................................................ 21 5.0 References ......................................................................................................................... 22 8/30/02 Revision 1

Harris Nuclear Plant Unit 1, Cycle 11 Startup Test Report Revision 1 Page 4 of 45 List of Tables and Figures Page Section 3 Table 3.6.1 Piping Thermal Expansion and Dynamic Effects, Special Emphasis Systems ..... 25 Section 4 Table 4.1.1 Control Rod Drop Times ........................................................................................ 26 Figure 4.2.1 Control Rod Drop Times ........................................................................................ 27 Table 4.3.1 Intermediate Range Detector R-Factor Determination .......................................... 28 Table 4.3.2 Power Range Detector R-Factor Determination ...................................................... 28 Figure 4.4.1 Flux Map 325 Measured vs. Calculated Powers ................................................... 29 Figure 4.4.2 Flux Map 326 Measured vs. Calculated Powers ................................................... 30 Figure 4.4.3 Flux Map 327 Measured vs. Calculated Powers ................................................... 31 Table 4.5.1 Flux Map Summary ................................................................................................. 32 Table 4.6.1 Reactivity Computer Checkout .............................................................................. 33 Table 4.6.2 Low Power Physics Test Results Summary ............................................................ 34 Table 4.7.1 Rod Worth Measurement of the Reference Bank ................................................... 35 Figure 4.7.1 Integral Worth of the Reference Bank .................................................................... 36 Figure 4.7.2 Differential Worth of the Reference Bank ............................................................ 37 Table 4.7.3 FSAR Chapter 14 Tests .......................................................................................... 38 8/30/02 Revision I

Harris Nuclear Plant Unit 1, Cycle 11 Startup Test Report Revision 1 Page 5 of 45 1.0 Introduction 1.1 General This startup report documents test results for Harris Nuclear Plant Unit 1, Cycle 11. This report primarily focuses on the results of the following evolutions:

"* Component & Initial Operation Tests

"* Operational and Power Ascension Tests These evolutions were modeled after those described in Chapter 14 of the Harris FSAR. The evolutions were modified to eliminate testing that is no longer appropriate. Examples of tests that were judged to be inappropriate include low power flux mapping and boron worth measurements. In the cases of boron worth measurement alternate testing described in ANSI 19.6.1 [Reference 5.37] was performed. Plant response data demonstrates that HNP control systems can safely and effectively operate following steam generator replacement (SGR) and power uprate (PUR). The Startup Test Program, defined by PLP-632T [Reference 5.1]

collected plant data from steady state operation and simulated transients to compare plant response with design predictions, specifications and accident analysis assumptions.

1.2 Cycle Description Cycle 11 introduces the sixth reload of Siemens High Thermal Performance (HTP) fuel.

Sixty-five (65) fresh HTP assemblies were loaded to replace 9 discharged Westinghouse LOPAR assemblies and fifty-six (56) discharged HTP assemblies. Cycle 11 also uses four (4) Siemens High Thermal Performance (HTP) once burned fuel assemblies that were stored in the spent fuel pool during Cycle 10. The specifics for the core reload design are presented in the Startup and Operations Report for Cycle 11 [Reference 5.2].

1.3 Steam Generator Replacement Harris Nuclear Plant recently replaced the original D4 (preheater style) steam generators with A75 (feed ring style) steam generators. The current steam generators share many operating characteristics with those installed at V.C. Summer Nuclear Power Plant.

1.4 Power Uprate Harris Nuclear Plant recently uprated core power from 2775 MWt (NSSS power = 2787.4 MW,) to 2900 MWt (NSSS power equals 2912.4 MWt). All references to reactor power are in percent of rated thermal power. The overall electrical output of the unit was increased 8/30/02 Revision I

Harris Nuclear Plant Unit 1, Cycle 11 Startup Test Report Revision 1 Page 6 of 45 approximately 50 megawatts, when compared to the previous cycle.

1.5 Tayg Restoration Harris Nuclear Plant restored Tavg to 588.8°F after operating at Tavg at 580.8°F during cycles 5 through 10, in order to reduce the degradation rate of the D4 steam generators. Satisfactory operation was verified after restoring Tavg. The A75 steam generators (SGs) were designed to operate with a Tavg of 588.8°F.

2.0 Summary Safe operation at the increased reactor power is supported by a combination of testing and plant observations. During power ascension safe operation within the analyzed bases was assured by a combination of conservative scaling adjustments and reduced setpoints. A rigorous test program was used to verify that impacted components were capable of meeting design bases assumptions. The design bases bound the measured plant parameters at uprated conditions.

3.0 Component & Initial Operation Test Summaries 3.1 Protection System Engineered Safety Features Actuation Logic Test There were no changes required to the actual Reactor Protection Logic due to any RFO-10 modifications. There were some changes to the utilization of certain SSPS slave relay contacts due to the removal of the feedwater preheater bypass valves, tempering lines, and large bore piping modifications [Reference .5.78, 5.79 and 5.80]. However, there were no changes to the actual SSPS logic cabinet. This logic was tested twice during the outage.

MST-10072 [Reference 5.43] and MST-I0001 [Reference 5.42] tested Train A and MST-10073 [Reference 5.44] and MST-10320 [Reference 5.45] tested Train B. There were some setpoint changes implemented into the PIC cabinets due to the new Lo-Lo Level (25%)

and Hi-Hi Level (78%) setpoints required for the replacement steam generators (RSGs).

There is an additional 5% operating margin (compared to the old SGs) from the normal operational setpoint of 57% to these setpoints. Also there were some changes implemented to the OTAT/OPAT setpoints and time constants due to RSG/PUR. This increased the margin to the turbine runback and reactor trip setpoints. An initial conservative value of AT of 62°F (versus a predicted 62.8°F nominal AT) was selected and implemented into the scaling, loop calibration procedures, and PICs. This value proved to be conservative during power ascension, particularly at the lower power levels prior to the initial calorimetric at 30%

power. This value was used until the results of EPT-156 [Reference 5.36] were incorporated into the revised scaling/calibration procedures for RCS loops B and C. EPT-156 was 8/30/02 Revision I

Harris Nuclear Plant Unit 1, Cycle 11 Startup Test Report Revision 1 Page 7 of 45 conducted at 75%, 90% and 100% power. Although the test results showed that only loop B was outside the acceptance criteria, both loop B and C were rescaled/recalibrated to minimize any initial errors at the beginning of the cycle. It should be noted that this test is performed quarterly throughout the cycle to ensure the RCS loop temperatures are maintained within a conservative band. New steam flow/feed flow mismatch setpoints were implemented as well to accommodate the increase in the 100% feed and steam flow from a nominal 4.067 mpph to 4.241 mpph.

EPT-093 [Reference 5.41] was used to establish any new control or protection setpoints that use turbine first stage pressure as the basis for the setting. Plots of first stage turbine pressure versus reactor power and a linear regression curve were generated. The predicted first stage pressure versus reactor power matched the actual curve such that no changes to instrument setpoints were required. This is also substantiated by the fact that the Tref curve, which is also based upon first stage pressure, did not require any instrument re-scaling/recalibration.

The Tref and Tavg signals are aligned within the tight specifications of the operating procedures.

Overlap testing was performed by the combination of PIC loop calibrations, surveillance testing, and SSPS testing.

3.2 Reactor Protection System Engineered Safety Features Actuation Response Time Test Qualified Barton model 764 transmitters were used to replace all of the narrow range and wide range steam generator level transmitters. The steam flow transmitters were replaced with qualified Rosemount units. The new SG level transmitters were tested per MST-40622

[Reference 5.46]. The remaining protection transmitters, which were scheduled to be time response tested during RFO-10, were tested per MST-10651 [Reference 5.47]. All of the applicable time response related surveillance tests were revised to reflect the new setpoints and time constants; i.e., for the OTAT/OPAT channels. The following time response related testing was also completed:

0 EST-300 [Reference 5.48]

0 EST-301 [Reference 5.49]

0 EST-302 [Reference 5.50]

0 EST-303 [Reference 5.51]

0 EST-304 [Reference 5.52]

S EST-305 [Reference 5.30]

S EST-306 [Reference 5.54]

S EST-307 [Reference 5.55]

S EST-308 [Reference 5.56]

0 EST-309 [Reference 5.57]

S EST-310 [Reference 5.58]

0 EST-31 1 [Reference 5.59]

8/30/02 Revision 1

Harris Nuclear Plant Unit 1, Cycle 11 Startup Test Report Revision 1 Page 8 of 45

EST-312 [Reference 5.60]

EST-313 [Reference 5.61]

EST-314 [Reference 5.62]

EST-315 [Reference 5.63]

EST-318 [Reference 5.66]

Results of these tests show that the acceptance criteria are met for the reactor protection and ESF circuits and components.

One narrow range RTD (TE-432D) was replaced during RFO-10 due to the element's time response falling outside the acceptance criteria of EST-300 [Reference 5.48]. The replacement RTD was satisfactorily time response tested prior to starting the first plateau (350°F) for EST-104 [Reference 5.67]. New calibration curves were generated for the following RTDs:

"* TE-422D

"* TE-422B2

"* TE-422B 2-S The test results of EST-104 [Reference 5.67] show that all RTDs met the acceptance criteria.

3.3 Auxiliary Feedwater System Test Increased steam pressures, piping reroute due to different azimuthal location of SG nozzle and elimination of the feedwater bypass piping, impacted the auxiliary feedwater system.

The impact on the system was demonstrated to be minimal analytically and by a combination of routine Technical Specification surveillances (EST-305 [Reference 5.30] and OST-1087

[Reference 5.32]) and an integrated test that was written to determine the turbine driven setpoint that verified minimal and maximum analyzed flow limits bound routine operation.

The integrated test EST-230 [Reference 5.70] demonstrated that a setpoint of 31 psid would deliver the minimum flow for various accident analyses, without exceeding the maximum flow assumed in the steam generator tube rupture (SGTR) analysis.

Pumps Maximum Maximum Minimum Minimum Operating Total Flow Flow Per SG Total Flow Flow Per SG TDAFW N/A N/A 414 gpm N/A TDAFW MDAFW A 1450 gpm 479 gpm 1194 gpm 404 gpm MDAFW B I I 8/30/02 Revision 1

Harris Nuclear Plant Unit 1, Cycle 11 Startup Test Report Revision 1 Page 9 of 45 3.4 Containment Integrated Leak Rate and Structural Integrity Test Removal and reinstallation of the equipment hatch barrel weld was identified as a potential adverse impact to the containment integrated leak rate. The closure weld was tested via a post-modification retest procedure contained in ESR 97-00805 [Reference 5.71]. The test verified that the reinstalled equipment hatch closure weld is leak tight.

3.5 Piping Thermal Expansion and Dynamic Effects Test The combination of steam generator replacement (SGR) and power uprate (PUR) impacts primary and secondary systems. Systems were visually observed at the revised operating conditions for adverse reactions from thermal expansion and dynamic effects. Special emphasis was placed on the systems listed in Table 3.6.1, for the stated reasons. The walkdown inspections were recorded using TMM- 117 [Reference 5.72].

3.6 Metal Impact Monitoring System Test The accelerometers that were located on the D4 steam generators were removed and new accelerometers that have the same fit and function were installed on the A75 steam generators. The preliminary alignment setpoints were established using conservative estimates. The digital metal impact monitoring system (DMIMS) baseline was established using EPT-012 [Reference 5.20] and data was collected and analyzed using EPT-023

[Reference 5.21]. The revised setpoints are recorded in OP-182 [Reference 5.82]. DMIMS has been restored to an operable condition.

3.7 Feedwater System Test The following impacted the feedwater system operation:

"* Head curve (impeller diameter) increased to overcome higher steam pressure

"* Piping dynamics due to conversion from preheater to feedring SG

"* Piping dynamics due to elimination of preheater bypass piping

"* System dynamics due to level control band on new SGs The feedwater pump was replaced during RFO-10 and a partial pump curve was verified using PPP-205 [Reference 5.38]. An integrated power ascension program, PLP-632T

[Reference 5.1] coordinated this activity. During power ascension steady state data was recorded and analyzed at various power levels. In addition, transient data was recorded and analyzed during the simulated transients created during the performance of EPT-848T

[Reference 5.33].

EPT-848T was performed on the main feedwater regulating valves (MFRVs) by injecting a

+5% SG level setpoint deviation at 30% power and 90% power. After making some initial 8/30/02 Revision 1

Harris Nuclear Plant Unit 1, Cycle 11 Startup Test Report Revision 1 Page 10 of 45 controller adjustments at the 30% power plateau, excellent steam generator water level (SGWL) control response was achieved using the MFRVs. Data was taken during the second main feedwater pump (MFP) start and minimal controller overshoot was experienced and rapid (<10 seconds) recovery to the setpoint was achieved. The test results at 90% power were excellent with minimal overshoot ranging from 0.34% to 1.5% and rapid stabilization at the setpoint.

Testing on the feedwater regulating bypass valves (FRBVs) was not completed due to oscillations on the SGWL control system prior to the test and a manual reactor trip due to a failure of the loop C FRBV. The C FRBV was repaired (new positioner) and the valves were operated in manual for the remainder of power ascension, with no further problems.

Continuation of the power ascension program and eventually full power operation was based on meeting the applicable acceptance criteria. The feedwater design bases bound the steady state, transient system response and accident analyses assumptions.

3.8 Steam Generator Primary Side FOSAR Inspection Foreign object search and retrieval was conducted on the primary side of the replacement steam generators (RSGs) prior to placing them into service. In addition to the normal debris created by the fabrication and installation prouesses it was necessary to remove some of the debris that was created by a partial decontamination of the RCS. A sponge-like blasting medium was used to clean away the corrosion product layer and perform a partial decontamination of the RCS piping. The resulting cleanliness level of the RCS piping met the established acceptance criteria (cleanliness level A) and accident analyses assumptions.

3.9 Steam Generator Secondary Side Internal Inspection Foreign object search and retrieval was conducted on the secondary side of the replacement steam generators (RSGs) prior to placing them into service. The potential objects consisted of debris created by the fabrication and installation processes. The visual inspections were documented in EPT-856T [Reference 5.22]. The digital photographs of the SG secondary surfaces clearly indicate that there are no materials that would adversely impact the SG tubes.

The resulting cleanliness level met the established acceptance criteria (cleanliness level B) and accident analyses assumptions.

3.10 Steam Generator Baseline A pre-service inspection (PSI) of the replacement steam generators (RSGs) open tubes was performed in June 2001 using eddy current testing (ECT). This inspection provides a baseline of eddy current signals against which future ECT inspections will be compared against in order to detect tube degradation.

8/30/02 Revision I

I Harris Nuclear Plant Unit 1, Cycle 11 Startup Test Report Revision I Page 11 of 45 NEI 97-06 [Reference 5.69] contains the PWR industry's "Steam Generator Program Guidelines" document. CP&L committed to meet the intent of the guidance in NEI 97-06.

PLP-651 [Reference 5.40] has extracted the pertinent information from the information from NEI 97-06 for a site SG Program. NEI 97-06, Section 3.1.3 / PLP-651, Section 5.2.3 states that tube integrity shall be assessed following each steam generator inspection.

The assessment of tube integrity from an eddy current test (ECT) inspection consists of two parts. Condition monitoring ("as-found condition) and operational assessment (addresses tube integrity until the next scheduled tube inspection) constitute the two parts. The guidance in NEI 97-06, and its referenced sub-tier EPRI documents, are focused on tube degradation during plant operation and do not detail guidance for pre-service inspection (PSI) for new tubing that has not been in service. Therefore, the following assessment will focus on the ECT inspection results from the PSI, highlighting indications that are attributable to the SG tube manufacturing process at the mill or the tube installation process at the SG fabrication facility.

The "conditioning monitoring" part of the integrity assessment is noted as "ensuring the performance criteria have been met for previous operating cycle" (NEI 97-06, Section 3.1.3).

EPRI document TR-107569-V1R5 "PWR SG Examination Guidelines: Rev. 5", states condition monitoring involves an assessment of the "as found" condition of the tubing relative to the performance criteria (Section 5.3). EPRI has an additional document, TR-107621-R1 "SG Integrity Assessment Guidelines", that provides guidance for actually performing tube assessments (Section 8). This document states "the process of condition monitoring involves the evaluation of the inspection results at the end of the operating interval to infer the state of the steam generator tubing for the most recent period of operation.

To satisfy the guideline for performing a condition monitoring assessment following ECT, and realizing this tubing has not been in operational service, this plant modification process has documented the "as-found" integrity of the new tubing based on the PSI results.

The operational assessment portion of the tube integrity assessment addresses the ability of the tubing to meet integrity performance criteria until the next scheduled tube inspection (integrity performance criteria is located in NEI 97-06, Sections 2 and 3.1.3 / PLP-651 Sections 4 and 5.2.3). The industry guidance documents for operational assessment evaluations are oriented toward tubing that has been in service, not for new tubing in RSGs.

The EPRI guidance document TR-017569-VlR5 indicates that this assessment is based in part on previous inspection results and repair criteria associated with each degradation mechanism. The new tubing does not have service-induced degradation mechanisms.

8/30/02 Revision I

Hams Nuclear Plant Unit 1, Cycle 11 Startup Test Report Revision 1 Page 12 of 45 3.11 Component Cooling Water System Test The component cooling water (CCW) system was revised extensively to support SGR, PUR and the activation of C & D spent fuel pools. The flow increase through the CCW heat exchanger resulted in flows that are 160% of the original design rating. Several other heat exchangers in the CCW system experienced similar flow increases. The new CCW pump impellers also develop substantially more head pressure, which caused several components to be re-rated at higher system pressures. Acceptable system performance was verified in EPT-847 [Reference 5.23]. The resulting system performance in each of the five system alignments met the established acceptance criteria and accident analyses assumptions.

Shutdown Cooling Component Acceptance Criteria Measured A/B Greater than or equal to 40 gpm 40140 RCP IA Thermal Barrier Greater than or equal to 5 gpm 6/5 RCP 1A Lower Cooler Greater than or equal to 155 gpm 166/164 RCP 1A Upper & Lower Oil Cooler Greater than or equal to 40 gpm 40/40 RCP 1B Thermal Barrier Greater than or equal to 5 gpm 5/5 RCP 1B Lower Cooler Greater than or equal to 155 gpm 162/160 RCP 1B Upper & Lower Oil Cooler Greater than or equal to 40 gpm 41/40 RCP IC Thermal Barrier Greater than or equal to 5 gpm 6/5.5 RCP IC Lower Cooler Greater than or equal to 155 gpm 164/160 RCP IC Upper & Lower Oil Cooler Greater than or equal to 330 gpm 375/340 Letdown Heat Exchanger Greater than or equal to 235 gpm 250/248 Seal Water Heat Exchanger Greater than or equal to 3425 gpm 3450/

Spent Fuel Pool I&4A Greater than or equal to 3425 gpm /3450 Spent Fuel Pool I&4B Greater than or equal to 282 gpm 320/

Spent Fuel Pool 2&3A Greater than or equal to 282 gpm /300 Spent Fuel Pool 2&3B Greater than or equal to 5600 gpm 6041/

RHR Heat Exchanger A Greater than or equal to 5 gpm 5.5/

RHR Pump A Cooler Greater than or equal to 5600 gpm /5965 RHR Heat Exchanger B Greater than or equal to 5 gpm /5.0 RHR Pump B Cooler Greater than 6 gpm 7/6.1 Gross Failed Fuel Detector Greater than or equal to 226 gpm 240/235 Reactor Coolant Drain Tank Greater than or equal to 250 gpm 255/250 Excess Letdown Heat Exchanger Less than or equal to 12650 gpm 11518/

CCW Heat Exchanger IA-SA Less than or equal to 12650 gpm /11467 CCW Heat Exchanger 1B-SB 8/30/02 Revision I

Harris Nuclear Plant Unit 1, Cycle 11 Startup Test Report Revision 1 Page 13 of 45 Normal Operation Component Acceptance Criteria Measured A/B RCP 1A Thermal Barrier Less than or equal to 60 gpm 53/52.5 RCP 1A Lower Cooler Less than or equal to 10 gpm 7.5/6.5 RCP 1A Upper & Lower Oil Less than or equal to 235 gpm 212/210 Cooler RCP 1B Thermal Barrier Less than or equal to 60 gpm 55/55 RCP 1B Lower Cooler Less than or equal to 10 gpm 5.75/7 RCP lB Upper & Lower Oil Less than or equal to 235 gpm 210/210 Cooler RCP IC Thermal Barrier Less than or equal to 60 gpm 53/52.5 RCP IC Lower Cooler Less than or equal to 10 gpm 7.5/7.5 RCP IC Upper & Lower Oil Less than or equal to 235 gpm 206/204 Cooler Letdown Heat Exchanger Less than or equal to 1300 gpm 1150/1140 Seal Water Heat Exchanger Less than or equal to 350 gpm 315/315 Spent Fuel Pool 1&4A Less than or equal to 4900 gpm 4475/

Spent Fuel Pool I&4B Less than or equal to 4900 gpm /4250 Spent Fuel Pool 2&3A Less than 750 gpm 410/

Spent Fuel Pool 2&3B Less than 750 gpm /385 RHR Pump A Cooler Less than or equal to 10 gpm 7.6/5.9 RHR Pump B Cooler Less than or equal to 10 gpm 5.6/7.0 Primary Sample Panel Less than or equal to 170 gpm 67.3/63.3 Gross Failed Fuel Detector Less than 12 gpm 11/10.5 Reactor Coolant Drain Tank Less than or equal to 360 gpm 300/295 Excess Letdown Heat Less than or equal to 355 gpm 330/327 Exchanger CCW Heat Exchanger 1A-SA Less than 8500 & 10600 gpm 7887 & 10501 CCW Heat Exchanger lB-SB Less than 8500 & 10600 gpm 7911 & 10537 8/30/02 Revision 1

Harris Nuclear Plant Unit 1, Cycle 11 Startup Test Report Revision4 Page 14 of 45 Recirculation @ 243.5°F Component Acceptance Criteria Measured A/B RHR Heat Exchanger A 7850 to 8050 gpm 7910/

RHR Pump A Cooler 7 to 9 gpm 8.1/

RHR Heat Exchanger B 7850 to 8050 gpm /7921 RHR Pump B Cooler 7 to 9 gpm /8.2 Recirculation @ 200'F Component Acceptance Criteria Measured RHR Heat Exchanger A Greater than or equal to 4029 gpm 6187 RHR Pump A Cooler Greater than or equal to 5 gpm 5 RHR Heat Exchanger B Greater than or equal to 3425 gpm 6187 RHR Pump B Cooler Greater than or equal to 5 gpm 5 Spent Fuel Pool 1&4A Greater than or equal to 3425 gpm 3650 Spent Fuel Pool 1&4B Greater than or equal to 3425 gpm 3650 Spent Fuel Pool 2&3A Greater than or equal to 282 gpm 310 Spent Fuel Pool 2&3B Greater than or equal to 232 gpm 310 Min / Max Flow Component Acceptance Criteria Measured A/B CCW Heat Exchanger A Greater than 57 psig 81 Discharge Pressure CCW Heat Exchanger B Greater than 57 psig 79 Discharge Pressure CCW Heat Exchanger A 12500 to 12650 gpm 12553 Outlet Flow CCW Heat Exchanger B 7850 to 7950 gpm 7919 Outlet Flow Service water system flow balance was also verified EPT-250 [Reference 5.74] and EPT-251

[Reference 5.75].

8/30/02 Revision 1

Harris Nuclear Plant Unit 1, Cycle 11 Startup Test Report Revision 1 Page 15 of 45 3.12 Safety Injection System Flow Balance The high head safety injection (HHSI) system was revised extensively to eliminate erosion of the throttle valves and eliminate a potential flow restriction inside the throttle valves

[Reference 5.81]. Installing a flow-restricting orifice in each of the twelve HHSI flow paths modified the system resistance. This allowed the throttle valves to be repositioned to a position that significantly reduced the probability of erosion and flow blockage from particles that may pass through the safety injection sump screens. The design bases flow rate was verified with plant procedure EST-206 [Reference 5.77].

4.0 Operationaland PowerAscension Test Summaries 4.1 Rod Drop Time Measurement Test Rod drop tests were performed in accordance with plant procedure EST-724 [Reference 5.3]

at hot full flow coolant conditions. Briefly, a bank is selected and pulled to the fully withdrawn position. Opening the reactor trip breakers, thus interrupting the circuit, then drops rods.

The acceptance criteria, from Technical Specifications, require that the rod drop time from the beginning of the drop to dashpot entry be no greater than 2.7 seconds at full core flow and operating temperatures. All rod drop tests were completed within the acceptance criteria.

Results of the rod drop testing are included in Table 4.1.1 and Figure 4.2.1 4.2 Reactor Coolant System Flow Measurement Test Reactor coolant system flow was measured using EST-709 [Reference 5.4]. The various RCS flows were established in the Final PCWG parameters [Reference 5.5]. The corresponding description and numerical values are as follows:

RCS Flow Description Flow (gpm)

Thermal Design (low) 277,800 FSAR Chapter 15/Technical 293,540 Specification (minimum)

Cycle 11 Measured "EST-709" 306,406 Nominal Predicted (best estimate) 306,600 Maximum (limiting) (high) 321,300 The HNP accident analyses are based on the most limiting RCS flow values (minimum or maximum). The RFO-10 (Cycle 11) measured EST-709 [Reference 5.4] flow value is bounded by the various accident analysis values.

8/30/02 Revision I

Harris Nuclear Plant Unit 1, Cycle 11 Startup Test Report Revision 1 Page 16 of 45 Based upon the results of EST-709 [Reference 5.4], nine of the reactor coolant flow protection loops required re-scaling. This re-scaling was implemented into the applicable surveillance tests and these channels were recalibrated on 1/07/02 to 1/08/02. This re-calibration resulted in acceptable reactor coolant flow indication.

4.3 Calibration of Nuclear Instrumentation Test The intermediate (IR) and power range (PR) detectors were adjusted after refueling (prior to startup), per procedure EPT-008 [Reference 5.6].

The IR adjustment factor (also referred to as the "R-factor") for Cycle 11 was calculated to be 1.1569. This pre-calculated value includes a bias multiplier of 1.1454 based on benchmark data [Reference 5.8] and a correction factor of 1.01 for restoration of Tavg from 580.8°F to 588.8'F. The Cycle 11 measured IR R-factor is determined from the N35 and N36 determined trip and rod stop setpoints between the last setpoint determination of Cycle 10 and the first setpoint determination of Cycle 11. The post-startup setpoint determination was performed under procedure EPT-009 [Reference 5.7]. These data are included in Table 4.3.1.

The actual IR trip setpoint, prior to recalibration, was calculated as N35 = 21.65% and N36 = 19.16%. The Technical Specification maximum allowable limit is 30.9%.

The PR adjustment factor (also referred to as the "R-factor") for Cycle 11 was calculated to be 1.2442. This pre-calculated value includes a bias multiplier of 1.1963, based on benchmark data [Reference 5.8] and a correction factor of 1.04 for restoration of Tavg from 580.8¶ to 588.80F. The Cycle 10 measured PR R-factor is determined from the N41, N42, N43, and N44 top and bottom HFP normalized detector currents between the last incore/excore calibration of Cycle 10 (Flux Map 322) and the first incore/excore calibration of Cycle 11 (Flux Map 327). The incore/excore calibrations were performed under procedure EST-911 [Reference 5.15]. These data are included in Table 4.3.2.

4.4 Flux Distribution Measurement Test Core power distributions for Cycle 11 are measured by processing moveable detector traces with the INPAX-W code, which is a module of the POWERTRAX core monitoring system.

"Powerdistribution maps for the power ascension flux maps are included as Figures 4.4.1 through 4.4.3.

The initial low power flux map is taken near 30% power to verify core loading is as designed.

Map @30% was taken immediately after stabilizing power near 30% (before equilibrium xenon was established) for core verification. The maximum difference between measured and calculated powers was 6.7% (location B-08), as shown in Figure 4.4.1. The Map @100% indicated that the limiting fuel assembly (L-09) had an F-dh (peak pin) fraction of limit of 0.918, see Figure 4.4.3. The following flux maps passed acceptance criteria 8/30/02 Revision I

Harris Nuclear Plant Unit 1, Cycle 11 Startup Test Report Revision 1 Page 17 of 45 contained in FMP-200 [Reference 5.13].

Map 325 @ 30% (verifying that the core was loaded as designed)

Map 326 @ 75%

Map 327 @ 100%

The core operating limits report (COLR) [Reference 5.11] requires a minimum of 44 measured traces for the core verification flux map; following the core verification flux map, all flux maps require a minimum of 38 measured traces.

4.5 Core Performance Test The flux maps following core loading verification are taken to verify compliance with Technical Specification requirements and limits on hot channel factors, quadrant power tilts, and to establish allowed power limits for successive power ascension. The following flux maps were taken near 75% and 100% power, respectively.

"* Map 326 @75%

"* Map 327 @100%

All flux maps allowed full power operation with no additional intermediate power level maps other than those required per PLP-632T [Reference 5.1]. Table 4.5.1 includes pertinent statistics for evaluating map quality and core parameters, which must be monitored.

The flux maps allowed power ascension and then full power operation based on meeting the applicable acceptance criteria.

4.6 Power Coefficient Measurement Test The RMAS reactivity computer is set up before LPPT by procedure EPT-026

[Reference 5.9]. Comparing period measurements to the startup rate indicated by the computer performed following initial criticality, checkouts the reactivity computer. The six group constants input to the reactivity computers were provided by SPC and are listed in Table 4.6.1.

The reactivity computer checkout requires that the positive and negative reactivity insertion period checks agree within 5%. Results of the reactivity computer checkout are included in Table 4.6.1. The reactivity computer acceptance criteria for Cycle 11 were met.

The isothermal temperature coefficient (ITC) is measured at ARO, HZP to verify that Technical Specification requirements limiting the ARO moderator temperature coefficient (MTC) to less than or equal to +5 pcm/°F at HZP. Should the MTC exceed the acceptance 8/30/02 Revision 1

Harris Nuclear Plant Unit 1, Cycle 11 Startup Test Report Revision 1 Page 18 of 45 criteria, rod withdrawal limits for startup and power ascension must be established. The MTC is derived from the measured ITC using the equation below, where the doppler temperature coefficient (DTC) is -1.54 pcm/°F [Reference 5.2].

ITC = MTC + DTC The low power physics testing (LPPT) is performed under a single test procedure (EST-923

[Reference 5.10]). EST-923 covers:

Initial criticality

Reactivity computer period checks

Test band determination (point of adding heat determination)

ARO boron endpoint

Temperature coefficient determination

Rod swap Results for Cycle 11 LPPT and the corresponding acceptance criteria are listed in Table 4.6.2.

This table (Table 4.6.2) also contains test results from sections 4.7 and 4.8.

4.7 Control Rod Reactivity Worth Test The worths of the control and shutdown banks are measured using the rod swap technique.

The reference bank (for Cycle 11, control bank B) was measured via boron swap. The remaining banks were measured fully inserted in the presence of the reference bank in a critical configuration.

The review criteria for the rod worths are as follows:

1. The absolute value of the percent difference between measured and predicted integral worth of the reference bank is less than 10%.

2. For all banks other than the reference banks, the absolute value of the percent difference between measured and predicted worths is less than 15% or the absolute value of the reactivity difference between measured and predicted worths is less than 100 pcm, whichever is greater.

The acceptance criteria require that the sum of the measured worths be between 90% and 110% of the sum of the predicted worths.

Results for Cycle 11 LPPT and the corresponding acceptance criteria are listed in Table 4.6.2.

Table 4.7.1 presents the integral and differential worth of the reference bank. Figures 4.7.1 and 4.7.2 graphically compare the predicted and measured integral and differential rod worths for the reference bank.

8/30/02 Revision I

Harris Nuclear Plant Unit 1, Cycle 11 Startup Test Report Revision 1 Page 19 of 45 4.8 Boron Endpoint Measurement - All Rods Out Test The boron endpoint is measured at ARO and again with the reference bank (control bank B) inserted. The acceptance criteria for boron endpoint measurement require the ARO endpoint to be within 50 ppm of the predicted value.

Results for Cycle 11 LPPT and the corresponding acceptance criteria are listed in Table 4.6.2.

4.9 RTDITC Cross Calibration Test EST-104 [Reference 5.67] is performed at three temperature plateaus, 350'F, 450'F, and approx 548'F. The data at the 350'F plateau was acceptable for all of the RTDs. However, at the 450'F plateau, TE-422D and TE-422B2 both showed low out of spec readings. Low insulation resistance was found to be the cause of the problem. A self-heating test was performed to dry the insulation. Follow-up insulation resistance testing at the 450'F, and approx 548°F plateaus were found to be acceptable.

4.10 Steam Generator Moisture Carryover Test CRC-864T [Reference 5.84], a steam generator moisture carryover test using lithium was performed on 5/23/02. At the time the test was performed, the plant had been operating for greater than 136 effective full power days. This allowed sufficient time for nominal steam generator tube fouling to occur before performance of the test. The moisture carryover limit that was established for the replacement steam generators was 0.10%. The test results indicated that all steam generators met or exceeded the specified requirement. The average measured moisture carryover for steam generators A, B, and C was found to be 0.0281, 0.0339, and 0.0032% respectively.

8/30/02 Revision 1

Harris Nuclear Plant Unit 1, Cycle 11 Startup Test Report Revision 1 Page 20 of 45 4.11 Load Swing Test A planned load swing test EPT-849T [Reference 5.34] was eliminated from scope.

Acceptable plant performance during a load swing is the result of several control systems (rod control, steam dumps, steam generator level, feedwater flow, etc.) working in an integrated manner. SGR/PUR impacted only the steam generator level and feedwater flow control systems. SGR/PUR setpoints established additional margin between the nominal and the expected value during a load swing. Simulator testing demonstrated that if the feedwater regulating valves are tuned to produce acceptable results during EPT-848T [Reference 5.33]

then moderate (<25%) load swing tests were not challenging. After reviewing the level swing testing EPT-848T [Reference 5.33] results Westinghouse confirmed that the planned load swing test EPT-849T [Reference 5.34] would generate minimal additional data.

4.12 Reactor Coolant System Leakrate Test The RCS was severed at the connections to the original SGs (D4 SGs) and narrow groove welds were utilized to reconnect the RSGs (A75 SGs). ASME Section XI code case N416-1 was invoked to substitute a normal in-service visual inspection in lieu of an ASME Section 111 hydrostatic test. The visual inspections of the RCS piping and components were documented in EST-201 [Reference 5.39]. The visual inspection was further confirmed by OST-1026 [Reference 5.19] leakage results during power ascension and subsequent operation.

4.13 Main Steam and Feedwater Systems Test The integrated power ascension program was coordinated by PLP-632T [Reference 5.1].

During power ascension steady state data was recorded and analyzed at various power levels.

In addition, data was recorded and analyzed during the simulated transients created during the performance of EPT-848T [Reference 5.33].

EPT-848T [Reference 5.33] was completed on the main feedwater regulating valves (MIFRVs) by injecting a +5% setpoint deviation at 30% power and 90% power. Also the FRV control response was checked during the second main feedwater pump (MFP) start. After making some initial controller adjustments at the 30% power plateau, excellent steam generator water level (SGWL) control response was achieved at the test plateaus and throughout power ascension.

Testing on the feedwater regulating bypass valves (FRBVs) was not completed due to oscillations on the SGWL control system and a manual reactor trip due to a failure of the loop C FRBV valve. During the subsequent restart the FRBV worked satisfactorily in the manual mode.

8/30/02 Revision I

Harris Nuclear Plant Unit 1, Cycle 11 Startup Test Report Revision 1 Page 21 of 45 The MFRVs and FRBVs were stroke tested and the data recorded in EPT-120 [Reference 5.76]. Design bases assumptions, with respect to feedwater isolation, were validated.

EPT-852T [Reference 5.68] was completed with acceptable results. These loops were calibrated per the input of ESR 00-00262, which was based upon a "best estimate" calculation of the steam flow differential pressure across the new steam generator outlet restrictor orifice and the steam flow piping. Data was taken at power plateaus of 30%, 50%,

75%, 90%, and 100% power. The steam flow transmitter differential pressures (D/Ps) and the steam flow proportionality constant resulted in the steam flow instrumentation providing acceptable results without any instrument rescaling or calibrations required.

Continuation of the power ascension power and eventually full power operation was based on meeting the applicable acceptance criteria. The design bases bound the main steam/feedwater steady state and transient system response and accident analyses assumptions.

4.14 Plant Performance Test The integrated power ascension program was coordinated by PLP-632T [Reference 5.1].

During power ascension steady state data wa:_ recorded and analyzed at various power levels.

In addition, data was recorded and analyzed during the simulated transients created during the performance of EPT-848T [Reference 5.33]. The plant performance baseline data was reviewed to ensure that the plant performance met design bases assumptions.

4.15 DEH Changes The digital electric hydraulic (DEH) system was modified to increase turbine blade reliability at low power levels. To accomplish this objective the arc of admission was increased from 180 degrees to 270 degrees. The DEH system was also tuned to increase controllability at higher steam pressures and higher power levels. The changes to the DEH system were verified to be acceptable in CM-C004 [Reference 5.27], ORT-8001 [Reference 5.26], and SCP-006 [Reference 5.53].

- 4.16 Steam Generator Blowdown Testing The steam generator blowdown system was revised to accommodate differences in the replacement steam generators. The changes include a maximum flow limit for sustained operation of 4.3 E4 ibm/hour. This limit is commercial in nature and is related to erosion of the steam generator blowdown channel and nozzle. EPT-287 [Reference 5.83] was intended to verify the alarm setpoint calculation with actual plant data. The test results were indeterminate and the calculation continues to be the basis for the alarm setpoint.

8/30/02 Revision 1

Harris Nuclear Plant Unit 1, Cycle 11 Startup Test Report Revision 1 Page 22 of 45 5.0 References 5.1 PLP-632T "Power Ascension Testing Program After R10 Refueling Outage (Steam Generator & Power Upgrade Modifications" 5.2 Startup and Operations Report for Cycle 11 "SOR" 5.3 EST-724 "Shutdown and Control Rod Drop Test Using Computer" 5.4 EST-709 "Reactor Coolant System Flow Determination By Calorimetric" 5.5 CQL-98-030 "Final PCWG Parameters for the SGR/Uprating Analysis and Licensing Project" 5.6 EPT-008 "Intermediate and Power Range Detector Setpoint Determination" 5.7 EPT-009 "Intermediate Detector Setpoint Determination" 5.8 HNP-F/NFSA-0010 "HNP BOC NI Adjustment" 5.9 EPT-026 "RMAS Setup and Operation" 5.10 EST-923 "Initial Criticality and Low Power Physics Testing" 5.11 PLP-106 "Core Operating Limits Report" 5.12 FMP-201 "Incore Flux Mapping Using POWERTRAX" 5.13 FMP-200 "Full Core Flux Map Review Checklist (POWERTRAX Version)"

5.14 EST-910 "Hot Channel Factors Test (POWERTRAX Version)"

5.15 EST-911 "Incore/Excore Detector Calibration Using POWERTRAX" 5.16 EST-216 "Steam Generator Tube Indication Tracking and Reporting Procedure" 5.17 EPT-244T "ABB/Combustion Engineering XS10370016, Temporary Procedure for Tube Plugging and Repair of Steam Generator Tubes (Expires 12/31/2001)"

5.18 ISI-202 "Safety-Related Component Support (Hangers and Snubbers) Examination and Testing Program" 5.19 OST-1026 "Reactor Coolant System Leakage Evaluation, Computer Calculation, Daily Interval, Modes 1-2-3-4" 5.20 EPT-012 "Digital Metal Impact Monitoring System (DIMIMS) Baseline Data Procedure" 5.21 EPT-023 "Digital Metal Impact Monitoring System Data Analysis Procedure" 5.22 EPT-856T "Steam Generator Secondary Side Internal Inspection" 5.23 EPT-847 "CCW Pump Performance Test and System Flow Balance" 5.24 EPT-243T "ABB/Combustion Engineering XS10370016, Temporary Procedure for Data Acquisition for Steam Generator Tube Examinations (Expires 12/31/2001)"

5.25 EPT-242T "ABB/Combustion Engineering XS10370016, Temporary Procedure for Eddy Current Examinations of Steam Generator Tubes (Expires 12/31/2001)"

5.26 ORT-8001 "DEH Computer System Dynamic Simulation Test" 5.27 CM-C004 "DEH Computer Reload and Restart" 5.28 OST-1825 "Safety Injection: ESF Response Time, Train A 18 Month Interval on a Staggered Test Basis Mode 5-6" 5.29 OST- 1853 "Feedwater Isolation ESF Response Time Trains A and B, 18 Month Interval Modes 3, 5, 6" 8/30102 Revision I

Harris Nuclear Plant Unit 1, Cycle 11 Startup Test Report Revision 1 Page 23 of 45 5.30 EST-305 "Engineered Safety Features Response Time Evaluation Motor Driven Auxiliary Feedwater Pumps" 5.31 EPT-196 "Motor Driven Auxiliary feedwater Pumps Performance Test" 5.32 OST-1087 "Motor Driven Auxiliary Feedwater Pumps Full Flow Test Quarterly Interval" 5.33 EPT-848T "S/G Level Swings" 5.34 EPT-849T "Load/Swing Reject Test" 5.35 EST-919 "Incore Versus Excore Axial Flux Difference Comparison (POWERTRAX Version)"

5.36 EPT-156 "Reactor Coolant System (RCS) AT Scaling at 100% Reactor Power" 5.37 ANSI 19.6.1 "Reload Startup Physics Test for Pressurized Water Reactors" 5.38 PPP-205 "Main Feed Pump Performance Test" 5.39 EST-201 "ASME System Pressure Tests" 5.40 PLP-651 "Steam Generator Program" 5.41 EPT-093 "Turbine First Stage Pressure Data" 5.42 MST-I0001 "Train A Solid State Protection System Actuation Logic & Master Relay Test" 5.43 MST-10072 "Train A 18 Month Manual Reactor Trip Solid State Protection System Actuation Logic & Master Relay Test" 5.44 MST-10073 "Train B 18 Month Manual Reactor Trip Solid State Protection System Actuation Logic & Master Relay Test" 5.45 MST-40320 "Train B Solid State Protection System Actuation Logic & Master Relay Test" 5.46 MST-10622 "Bench Transmitter Response Time Test" 5.47 MST-10651 "Transmitter Noise Analysis Time Response Test" 5.48 EST-300 "Reactor Trip response Time Evaluation" 5.49 EST-301 "Engineered Safety Features Response Time Evaluation Safety Injection" 5.50 EST-302 "Engineered Safety Features Response Time Evaluation Feedwater Isolation" 5.51 EST-303 "Engineered Safety Features Response Time Evaluation Containment Phase "A" Isolation" 5.52 EST-304 "Engineered Safety Features Response Time Evaluation Containment Ventilation Isolation" 5.53 SCP-006 "Throttle Valve and Governor Valve Calibration Procedure" 5.54 EST-306 "Engineered Safety Features Response Time Evaluation Emergency Service Water Pumps" 5.55 EST-307 "Engineered Safety Features Response Time Evaluation Containment Fan Coolers" 5.56 EST-308 "Engineered Safety Features Response Time Evaluation Steamline Isolation" 5.57 EST-309 "Engineered Safety Features Response Time Evaluation Containment Spray" 5.58 EST-310 "Engineered Safety Features Response Time Evaluation Phase-B Isolation" 5.59 EST-311 "Engineered Safety Features Response Time Evaluation Turbine Trip" 5.60 EST-312 "Engineered Safety Features Response Time Evaluation Turbine Driven AFW Pump" 8/30/02 Revision I

Harris Nuclear Plant Unit 1, Cycle 11 Startup Test Report Revision 1 Page 24 of 45 5.61 EST-313 "Engineered Safety Features Response Time Evaluation Switchover to Recirculation Sumps with SI" 5.62 EST-314 "Engineered Safety Features Response Time Evaluation Switchover to Recirculation Sumps with Containment Spray" 5.63 EST-315 "Engineered Safety Features Response Time Evaluation Containment Purge Isolation" 5.64 EST-316 "Emergency Sequencing System IA-SA Response Time Test" 5.65 EST-317 "Emergency Sequencing System lB-SB Response Time Test" 5.66 EST-318 "Engineered Safety Features Response Time Evaluation Isolate AFW to the Affected SG" 5.67 EST-104 "Incore Thermocouple and RTD Cross Calibration Data Compilation" 5.68 EPT-852T "Temporary Procedure for Calibration of Steam Flow Instruments" 5.69 NEI 97-06 "Steam Generator Program Guidelines" 5.70 EST-230 "Determination of Turbine Driven Auxiliary Feedwater Pump Differential Pressure Controller Setpoint" 5.71 ESR 97-00805 "SGR Containment Modifications" 5.72 TMM-1 17 "System Walkdowns and Observations" 5.73 OST-1204 "Power Range Heat Balance, Manual Calculation, Daily Interval, Mode 1 (Above 15% Power)"

5.74 EPT-250 "A Train ESW Flow Verification/Balance" 5.75 EPT-251 "B Train ESW Flow Verification/Balance" 5.76 EPT-120 "Stroke Timing of the Main and Bypass Feedwater Regulating Valves" 5.77 EST-206 "ECCS Flow Balance" 5.78 ESR 99-00468 "RSG Blowdown and Wet Layout System setpoints" 5.79 ESR 98-00537 "Tempering Lines Modification" 5.80 ESR 97-00807 "SGR Large Bore Modifications" 5.81 ESR 99-00407 "ECCS Flow Balancing Orifice Installation" 5.82 OP-182 "Digital Metal Impact Monitoring System" 5.83 EPT-287 "Blowdown Flow Versus Differential Pressure (Delta-75 SG)"

5.84 CRC-864T Temporary Procedure For Steam Generator Moisture Carryover Sampling (Expiration Date 12/20/2002) 8/30/02 Revision 1

Harris Nuclear Plant Unit 1, Cycle 11 Startup Test Report Revision 1 Page 25 of 45 Table 3.6.1 Piping Thermal Expansion and Dynamic Effects Special Emphasis Systems System Reason for Special Emphasis Reactor Coolant System The D4 steam generators were removed from the system and A75 steam generators installed. The replacement steam generators (A75) have a higher center of gravity and are predicted to have a slightly greater RCS flow.

Feedwater System The feedwater piping was rerouted from the bottom of the (Inside Containment) steam generators (D4 - preheater style) to the top of the replacement steam generators (A75 - feedring style). The preheater bypass flow path was eliminated, increasing the feedwater flow through the main feedwater piping by 30% (prior to PUR). PUR further increased feedwater flow by an additional 4.5%.

Feedwater System Feedwater flow has increased by 4.5%. This change (Outside Containment) impacts the feedwater system and various support systems (condensate, heater drains, etc.). The preheater bypass flow path and tempering line flow path was eliminated.

Steam System Steam flow has increased by 4.5%. This change impacts the steam system and various support systems (moisture separators, heater drains, etc.).

Component Cooling Water In conjunction with SGR, PUR, and activation of the C & D spent fuel pools Harris upgraded the CCW system to restore operating margin. Significant flow increases were experienced at the CCW heat exchangers (160% of original design flow rating) and similar increases were observed throughout the system. Larger CCW pump impellers increased the flow and pressure transient response associated with pump starts and other system transients.

8/30/02 Revision 1

Harris Nuclear Plant Unit 1, Cycle 11 Startup Test Report Revision 1 Page 26 of 45 Table 4.1.1 Control Rod Drop Times' Control Banks Shutdown Banks Rod Core Time to Time to Rod Core Time to Time to Bank Location Dashpot Bottom of Bank Location Dashpot Bottom of Entry (sec) Dashpot Entry2 Dashpot (sec) (sec) (sec)

CBA F-02 1.52 2.02 SBA G-03 2.03 B-10 1.60 2.05 C-09 2.02 K-14 1.66 2.09 J-13 1.96 P-06 1.49 1.92 N-07 1.99 K-02 1.56 1.97 J-03 1.98 B-06 1.80 2.25 C-07 2.08 F-14 1.54 1.93 G-13 1.98 P-10 1.53 1.90 N-09 1.98 CBB F-04 1.54 1.92 SBB E-05 2.01 D-10 1.53 1.88 E-I 1 2.02 K-12 1.50 1.94 L-11 1.98 M-06 1.54 1.88 L-05 1.98 K-04 1.51 1.89 G-07 1.97 D-06 1.54 1.94 G-09 1.98 F-12 1.52 1.89 J-09 1.97 M-10 1.51 1.86 J-07 1.96 CBC D-04 1.52 1.94 SBC E-03 1.90 D-12 1.50 1.94 C-11 1.91 M-12 1.48 1.93 L-13 1.94 M-04 1.53 1.96 N-05 1.90 H-06 1.53 1.93 F-08 1.49 1.85 H-10 1.52 1.90 K-08 1.51 1.90 CBD H-02 1.54 1.95 B-08 1.65 2.08 H-14 1.52 1.93 P-08 1.56 2.00 F-06 1.55 1.94 F-10 1.54 1.91 K-10 1.54 1.91 K-06 1.53 1.90 2

2 Measured data obtained from Cycle 11 (RFO 10) performance of EST-724 [Reference 5.3].

Dashpot entry times were not recorded for Shutdown Banks. All rod bottom data are within TS Limit of 2.7 seconds for dashpot entry.

8/30/02 Revision 1

Harris Nuclear Plant Unit 1, Cycle 1I Startup Test Report Revision 1 Page 27 of 45 Figure 4.2.1 Control Rod Drop Times Core Location oL=i- VCO Ll*°*OL DL-,D* I,*- E,!3O-S -5062o =_*LIL 6-,o-?, 2*L.*.

3.0 2.9 2.8 2.7 2.6 2.5 2.4 2.3 2.2 2.1 2.0 1.9 1.8 1.7 1.6 E 1.5

0. 1.4 0

1.3 1.2 1.1 1.0 0.9 0.8 0.7 0.6 0.5 0.4 0.3 0.2 0.1 0.0 CBA CBA CBB CBB CBC CBC CBD CBD SBA SBA SBB SBB SBC Rod Bank 8/30/02 Revision I

I Harris Nuclear Plant Unit 1, Cycle 11 Startup Test Report Revision 1 Page 28 of 45 Table 4.3.1 Intermediate Range Detector R-factor Determination N35 N36 Cycle Trip Rod Stop Trip Rod Stop Average 101 Startup 79 6.10E-5 4.88E-5 5.32E-5 4.26E-5 112 Startup 82 8.16E-5 6.53E-5 8.38E-5 6.70E-5 Cll R-value (Cll/ClO) 1.338 1.338 1.575 1.573 1.457 1 Cycle 10 data obtained from Startup 79 performance of Reference 5.7 (5/25/00) 2 Cycle 11 data obtained from Startup 82 performance of Reference 5.7 (1/11/02)

Table 4.3.2 Power Range Detector R-factor Determination Detector Cycle 10' Cycle 111 C9 R-value (Flux Map 322) (Flux Map 327) (ClI / CIO)

N41 top 137.2 159.2 bottom 155.6 184.2 sum 292.8 343.4 1.1728 N42 top 149.2 177.4 bottom 171.6 208.2 sum 320.8 385.6 1.2020 N43 top 169.4 201.9 bottom 189.5 229.2 sum 358.9 431.1 1.2012 N44 top 136.5 160.8 . .....

bottom 164.3 199.2 sum 300.8 360.0 1.1968 Average NomeInn_____ _ 1.1932 1 Power Range Data taken from respective performance of EST-9 11 [Reference 5.15].

8/30/02 Revision I

Harris Nuclear Plant Unit 1, Cycle 11 Startup Test Report Revision 1 Page 29 of 45 Figure 4.4.1 Flux Map 325 Measured vs. Calculated Powers P N M L K . H G F E D C B A R

I .2971 .3151 .2981 1 I 2991 .3231 .3021 I .71 2.51 1 31

_______ I I .3861 .9981 1 2041 .9431 1.2031 1 0031 .3921 Measured Power 2 1 .3911 1 0051 1.2121 .9621 1 2171 1.0081 .3911 Calculated Power 1 1.31 71 .71 2.01 1 21 .51 -. 31 Percent Difference I .4671 1 1881 1 2381 1.2401 1.1441 1.2241 1.2471 1.2211 .4731 3 .473 1 2141 1 2471 1.2311 1 1701 1.2351 1.2481 1 2151 .4751 I 1 31 2 21 71 2 31 .91

-. 71I.........I.....-.....I .1! I _.51 .41 1 .4631 1 2751 1 2031 1.0651 1 0881 1.2201 1.0821 1 0771 1 2721 1.3211 .4781 4 I .4731 1 3051 1 2501 1.0761 1 0901 1.2491 1.0871 1.0711 1.2501 1.3091 .4751 I 2.21 2 41 3.91 1 01 21 2.41 .51 -. 61 -1.71 -. 91 -. 61 I .3871 1 1811 1.2361 1.0281 1 1021 1.2761 1.1821 1.2561 1 0881 1 0791 1.2671 1.2151 .3811 5 I 1 .3901.811 2.512111 1.2471 .91 1.0261 -. 21 1 0961 - 51 1-1.412581 1.1681-1.21 1 -1.012441 1 -1 069171 1-40291 -1.3.1 1.2161 61 1.2531 .11 .3911 2.61

________ _______ i--______I________ ....

I...............--.......... --...........

1 1.0101 1.2511 1.0801 1 1011 8851 1.0771 1.1161 1 0521 .8771 1.1211 1.0881 1.2441 .9871

.8731 1.0771 1.2481 1.0061 6 I 1.0051 1 -. 51 1.2441 -2.01 1 -30671

-. 61 1.0691 11 -1.51.8721 1.0511-2.41 1-4.81 0621 1.0401

-1.11 -. 51 1.0981

-2.11 -1.01 .31 1.91 I.3021 1.2281 1.2621 1.1061 1 2701 1.0581 1 1481 1.0001 1.1101 1.0651 1 2791 1.1021 1.2181 1.1761 2891 I .3001 1.2131 1.2321 1 0851 1.2421 1.039i 1 1301 9841 1 1301 1.0511 1.2591 1.0901 1.2321 1.2121 3001 3 81 1 -. 71 -1.21 I I -- I

-2 41 -1 ___91 -2 21 -1.81 -1 61 -1 61I 1.81I -1 31l.........I-...........I I

-1.61 -1.11 1.11 II 3.11 I .3151 .9681 1.1881 1.2701 1.1881 1 0811 1.0031 1.1921 1.0011 1 1121 1 1891 1.2511 1.1461 .9021 .3071 8 1 .3221 .9601 1.1681 1.2471 1.1651 1 0611 .9841 1.1701 .9841 1 0611 1 1671 1.2481 1.1701 .9621 .3231

-1.71 -1.81 -1 91 -1 91 -1 91 -1.81 -1.71 -4 61 -1.91 -. 21 2.11 6.71 5.21 I 2.21 - 01 -- ________.____ ________ I I .3001 1.2211 1.2461 1.1061 1 2801 1 0691 1.1531 1 0021 1 1511 1 0631 1.2541 1.0811 1.2071 1.1671 .2891 9 I .2991 1.2091 1.2291 1 0881 1 2571 1 0501 1.1291 .9831 1.1301 1 0391 1.2431 1.0861 1.2341 1 2151 .3011 1 -. 31 -1.01 -1.41 -1 61 -1 81__________ -1 81 -2.11 -1.91 -1.81 -2 31 _________-. 91 .51 2.21 4.11 4.21

__________I I 1.0141 1.2611 1 0891 1 1101 .8811 1 0661 1.0771 1.0611 .8731 1.0651 1.0581 1.2191 .9671 10 1 0031 1 2451 1 0751 1.0951 .8721 1 0391 1.0601 1.0501 .8721 1 0681 1 0691 1.2451 1.0061 I -1.11 -1 31 -1.31 -1.41 -1 01 -2 51 -1.61 -1.01 -. 1l .31 1.01 2.11 4.01

_______ I I .................... I -- ............ I..........l......

I I.....

I....-............

1 3941 1.2241 1.2621 1.0361 1 0761 1.2551 1.1741 1.2601 1.0971 1 0101 1.2321 1.1901 .3811 I .3901 1.2121 1 2491 1 0261 1 0671 1.2421 1.1661 1 2571 1.0951 1 0261 1.2471 1.2111 .3911 11 I -1.01 -1.01 -101 -1 01 1.01 -. 71 -. 21 - 21 1 61 1.21 1.81 2.61 I .4761 1 3111 1 2511 1 0691 1.0821 1.2411 1.0761 1 0571 1.2301 1.2961 .4661 12 I .4731 1.3051 1 2461 1 0681 1.0841 1.2461 1 0881 1 0751 1 2491 1.3051 -4731 I -. 61I - sI __-- - ___ 41 - 11II .21 .41 1.111 1.71I 1.51 .71 1.51I

-- I I .4741 1.2081 1 2371 1 2181 1.1471 1.1941 1.1941 1.1791 .4651 13 I .4731 1.2101 1 2441 1 2321 1.1681 1.2281 1.2441 1.2121 .4731 1 71 1 11 I I-. 21l.........I..........I....................-

I .21 61 1.81 2.81 I 4 21 2 81 I 3861 9921 1 1891 .9301 1.1611 .9641 3781 14 I 3901 1 0051 1 2131 .9601 1:2091 1.0031 3901 II ________

o01 1 31 2.01 3.21 4 11 4 01 3 21

........... I I .2891 .3121 .2881 15 I .3011 .3231 2991 I 4 21 3 51 3 81 8/30/02 Revision I

Harris Nuclear Plant Unit 1, Cycle 11 Startup Test Report Revision 1 Page 30 of 45 Figure 4.4.2 Flux Map 326 Measured vs. Calculated Powers R P N M L K .7 H G P K D C B A I .3061 318j .3091 1 I .3151 3471 3171 1 2.91 9 11 2.61

_ __ _ I I I I I 3831 9671 1.1821 9961 1.1891 .9761 .3901 Measured Power 2 I .3911 .9801 1.1961 1 0141 1 2011 .9831 .3921 Calculated Power I 2.11 1.31 1.21 1 81 1.01 .71 .51 Percent Difference 1 .71 1.1341 1 1931 1.2071 1.1691 1 2081 1.2041 1.1621 .460!

3 1 .4681 1.1661 1.2071 1.2081 1 1651 1 2121 1.2081 1.1671 .4701 4.71 2.81 1.21 .11 - 31 .31 31 .41 2.21 I ___ I ________I

-- I I.........-I _______ ! ______ I

... I....!

I .4581 1.2151 1.1651 1.0571 1.0831 1 2131 1 0821 1.0591 1.2151 1.2311 .4631 4 I .4681 1.2451 1 2091 1 0661 1.0871 1.2381 1 0841 1 0611 1.2091 1.2481 .4701 I 2.21 2.51 3 81 .91 .41 2 11 21 .21 -. 51 1.41 1.51 I______ _______I I_______I_______I_______ I____I____

I .3851 1.1351 1.2061 1 0761 1.1371 1.2881 1 1951 1 2741 1 1111 1.0661 1.1831 1.1451 .3841 5 I .3901 1.1641 1.2061 1.0271 1.1181 1.2691 1 176! 1 2561 1.0931 1.0291 1.2111 1.1681 .3911 1 1 31 2.61 .01 -4 61 -1 71 -1.51 -1 61 -1 41 -1 61 -3.51 2.41 2.01 1.81 I -- I l-..-.--..-! I I I I I I .9721 1.1971 1 0641 1.1111 .9751 1.1221 1 1601 1 1121 9801 1.1381 1.0631 1.1921 .9601 6 I 980! 1.2051 1.0591 1 0911 9631 1.0951 1.0961 1.0851 9641 1.1201 1.0671 1.2081 .9811

-81

. .71 -. 51 -1.81 -1.21 -2.41 -5 51 -2 41 -1 61 -1.61 .41 1.31 2.21

-I-I

______ I I -- I I I I I I .3121 1 1901 1.2081 1 0891 1.2681 1.0971 1.1871 1 0561 1 1881 1.1171 1.2881 1.0961 1.1911 1.1631 .3051 7 I .3161 1 1981 1.2091 1.0831 1.2541 1.0841 1.1751 1.0271 1 1761 1.0951 1 2691 1.0871 1.2091 1.1971 .3151 1 1 31 .71 .11 -. 61 -1 11 -1 21 -1.01 -2 71 -1 01 -2 01 -1 51 -. 81 1.51 2.91 3.31 I -- I !-..........-. - I I ......-l I t _______________

1 .3351 1.0031 1.16411 1 2451 1.1831 1 111! 1.0481 1.2491 1 0661 1.1351 1.1941 1.2381 1.1381 .9501 .3291 8 1 .3461 1.0111 1.1621 1 2361 1.1741 1.0951 1.0271 1.2121 1 0281 1 0961 1 1761 1.2371 1.1641 1.0131 .3471 1 3 31 .81 -. 21 - 71 1.41 -2.01 -3 01 -3 61 -3 11 -1 51 -. 11 2.31 6.61 5 51

_____ ____ ____I-- _____I_____ I I t

1 3121 1 1931 1.2111 1.0971 1.2951 1.1141 1.2021 1.0641 1 2551 1 1081 1.2681 1.0801 1.1851 1.1511 .3031 9 1 3141 1.194! 1.2061 1 0851 1.2671 1.0941 1.1751 1.0271 1 1751 1 0841 1 2551 1.0831 1.2111 1.1991 .3161 1 61 .11 -. 41 -1 11 -2.21 -1 81 -2.21 -3 51 -6 41 -2 21 -1 01 .31 2.21 4.21 A 31 I --

_____I -- I .-....- ! I I______

_______ ___I_

I .9811 1.2141 1 0751 1.1311 .9701 1.1021 1 1171 1 1261 9751 1 0991 1 0571 1.1841 .9431 I 1.2051 1.0651 1 1171 .962! 1.0841 1.0951 1 0941 9621 1.0911 1.0591 1.2051 .9811 10 1 .9791

-. 21 -. 71 -. 91 -1.21 -. 81 -1.61 -2 01 -2 81 -131 - 71 .21 1.81 4.01 I -- I I I I II I I I I I I .3931 1.1731 1 2161 1 0291 1 095! 1.2591 1.1761 1.2791 1 1331 1 0331 1.2111 1.1501 .3821 11 I .3901 1.1651 1 2081 1.0271 1.0911 1.2541 1.1741 1 2671 1 1171 1 0261 1 2061 1.1641 .3911 1 -. 81 -. 71 - 71 -. 21 -. 41 -. 41 - 21 -. 91 -1 41 - 71 -. 41 1.21 2.41

___ --____I _______I I I I_______I_______ I_______ ______ I I .4761 1.2491 1.2071 1.057! 1 0771 1 2291 1 0791 1 0581 1 2121 1 2711 .4561 12 I .4681 1 2441 1.2051 1.058! 1 0821 1.2361 1 0851 1 0641 1.2081 1.244! .4681 I -1 71 -. 41 - 21 .1' *51 .61 .61 61 -. 31 -2.11 2.61 I ______ _______ I I I ____ I I .4691 1.1601 1.197! 1.1951 1.1441 1 1781 1 162! 1 1471 .4701 13 I .4681 1.1631 1.204 1 2091 1.1621 1.2061 1.2051 1 1641 .4681 I - 21 .31 .61 1.21 1.61 2 41 3 71 1 51 -. 41 I ________

I______ I I I______________ I !

I .3881 .9681 1 1761 .9831 1.1551 9471 3821 14 I .3901 .9801 1 1981 1.0111 1 1941 9781 3901 I .51 1.2 1 91 2 81 3 41 3 31 2 1!

i I I I I I.........

1 .3061 3371 .3051 15 1 .3161 .3461 .3141 1 3.31 2 71 3.0!

I -- I - I -- _

8/30/02 Revision I

Harris Nuclear Plant Unit 1, Cycle 11 Startup Test Report Revision 1 Page 31 of 45 Figure 4.4.3 Flux Map 327 Measured vs. Calculated Powers R P N M L K 7 H G F E D C B A I .3151 .3371 .3171 1 I 3221 .3611 .3251 1 2 21 7.11 2 51

!3811.9481 1.1711 1.0321 1 1721 .9511 .3851 Measured Power I

2 I .3871 .9591 1.1851 2 0501 1 1891 .9621 .3881 Calculated Power 1 1.61 1.21 1.21 1 71 1.51 1 21 .81 Percent Difference I .4461 1.1051 1.1691 1.1971 1.1541 1.1901 1.1721 1.1231 .4491 3 I .4611 1.1311 1.1811 1.1971 1 1651 1.2001 1.1821 1.1331 .4631 I 3.41 2.41 1 0! .01 1.01 .81 -91 .91 3.11

___I____-- I I...--....-.I.

I .4521 1.1741 1.1441 1 0531 1.0871 1 2171 1.0821 1.0521 1.1861 1 1881 .4551 4 I .4611 1 2031 1.1821 1.0611 1 087! 1 2331 1.0851 1 0561 1.1821 1.2071 .4631 I 2.01 2 51 3 31 .81 01 1.31 .31 .41 -. 31 1.61 1.81 1 .3821 1.1041 1.1711 1.0451 1.1491 1 2991 1 2011 1.2831 1 1341 1.0731 1.1601 1.1081 .3701 5 1 .3871 1 1291 1 1791 1.0301 1.1381 1 2771 1.1841 2651 1.1131 1 0331 1.1851 1.1341 3881 1 31 2.31 .71 -1.41 -1.01 -1 71 -1.41 -1.41 -1 91 -3.71 2.21 2.31 4.91

_____I__ IIII.-........!.........! I....... ! --... I I.........I I 1 .9531 1.1721 1.0571 1.1291 1.0511 1 1511 1.1651 1 1441 1 0651 1.1621 1.0601 1.1651 .9381 6s ! .9581 1.1781 1.0541 1.1111 1.0371 1 1271 1.1181 1 1171 1 0381 1 140! 1.0621 1.1821 .960j

-1.61 -1.31 ____I___l-2 11I______ -2.41

-4 01!-.....-..-!-_______I -2 51 -1.91 .21 1.51 2.31

____ 51I__

.51 -. 31 - !- . I 1 --......

3211 1.1811 1.2011 1.0901 1.2771 1.1311 1.2181 1.0781 1 2211 1 1551 1.2991 1.0951 1 1821 1.1581 .3151 7 1 3231 1.1851 1.1971 1.0831 1 2631 1.1161 1.2021 1.0531 1.2021 1 1271 1.2771 1.0871 1.1971 1.1851 .3231

-1 31I _-1.31 -2 31 -1 61 -2 4II -1.7! I -. 71 1.31 2.31 2 51 1 .61 I .31-- I -. 31I -. 61 -1.11 _ I______ I I ____I __

1 3521 1.0431 1.1661 1 2391 1.1901 1.1321 1.0711 1 2591 1 0831 1 1621 1.2031 1.2341 1.1421 .99o! 3471 8 .3601 1.0471 1.1621 1 2301 1 1811 1.1171 1 0521 1.2311 1 0531 1.1181 1.1831 1.2321 1.1641 1.0491 .3611 1 2.31 .- 1 -. 31 -. 71 - 8! -1.3! -1 81 -2 21 -2 81 -3.81 -1.71 - 21 1.91 5.31 4.01 I I I !I.......... l..........I-..........I--........lI . I.......... ! l.I.......I.

.3201 1.1811 1 1991 1.0961 1.2991 1.1441 1 2231 1.0761 1.2411 1 1401 1.2781 1 0821 1.1781 1.1481 3151 9 1 .3221 1 1811 1 1941 1.0851 1.2751 1.1251 1.2011 1.0521 1.2021 1 1161 1 2641 1.0841 1.1991 1.1871 .3241 1 .61 .0! 1.0! -1.81 -1.71 -1.81 -2.21 -3.11 -2 11 -1.11 .21 1.81 3.41 2.91

______I I ___ --___ I I _I____I I 1 .9591 1.1841 1.0691 1.1521 1.0481 1 1331 1.1331 1 1441 1.0481 1.1191 1.0511 1.1601 .9311 10 1 .9571 1.1781 1.0591 1.1361 1.0361 1.1151 1.1171 1 1251 1 0361 1 1111 1.0541 1.1791 .9591 1 -. 21 -. 51 -- 91 -1.41 -1.11 -1.61 -1 41 -1 71 -1 11 - 71 .31 1.61 3 01

.3881 1.1321 1.1871 1 0411 1.1201 1.2711 1 1881 1 2831 1 1451 1.0351 1.1741 1.1121 .3801 11 1 .3861 1 1301 1 1811 1 0291 1.110! 1 2621 1 1811 1 2751 1 1361 1 0291 1.1791 1.1291 .3871 1 -. 51 -. 21 - 51 -1.21 -. 91 -. 71 - 61 -. 61 -. 81 -. 61 .41 1.51 1.81 1 .4571 1 2031 1.1821 1.0551 1 0801 1.2281 1 0811 1.053! 1.1761 1.1991 .4481 12 1 4611 1.2021 1.1781 1.0531 1 0821 1 2301 1 0851 1.0591 1.1811 1.2021 .4611 I .91 -. 11 -. 31 -. 21 21 21 41 .61 -41 -31 2.91 I I !..........l.-................-

I ......... I I-......... I I .4621 1.1271 1 1741 1.1881 1.1501 1.1761 1.1491 1.1131 .4581 13 j .4611 1.1281 1 1771 1.1961 1.1611 1.1931 1 1781 1.1291 .4611 1 -. 21 .11 31 -71 1 01 1 41 2 51 1.41 .71 I I !....................... 1 ! ......... l....I .....

1 3861 .9511 1.1701 1 0301 1 1571 .9351 .3801 14 1 .3861 .9581 1.185! 1 0471 1 1811 9571 .3871 I -01 .71 1.31 1 71 2 11 2.41 1.81 I I....I.......I.......i..

1 .3161 3551 3161 15 1 .3231 3601 3221 I 2 2! 1 41 1 91 8/30/02 Revision I

Ip Iz Harris Nuclear Plant Unit 1, Cycle 11 Startup Test Report Revision 1 Page 32 of 45 Table 4.5.1 Flux Map Summaryl Map # Burnup Date Time Power D-bank Boron (EFPD) (%) (steps) (ppm) 325 0.09 01/03/2002 16:30:00 26.9 150 1851 326 0.87 01/05/2002 11:00:00 74.5 176 1550 327 2.15 01/06/2002 23:00:00 98.2 206 1410 Map # RMS max FAh Fraction to max FQ Fraction to Axial Power2 Limit, F*h Limit, FQ Offset (M-P) (FLFH) (FLFQ) (%)

325 1.97% 1.672 0.802 2.465 0.511 4.284 326 2.05% 1.597 0.883 2.236 0.692 -0.444 327 1.69% 1.533 0.918 2.145 0.874 2.230 Map # Thimbles Thimbles Quadrant Power Tilt Ratio Used Required NW NE SW SE 325 44 38 1.003 1.004 1.006 0.988 326 43 38 0.999 1.001 1.003 0.996 327 43 38 0.999 1.001 1.003 0.996 1

Flux Map summary data taken from respective INPAX runs [Reference [POWERTRAX]].

2 RMS Power is a figure of merit for how well the core power distribution is predicted.

8/30/02 Revision I

Harris Nuclear Plant Unit 1, Cycle 11 Startup Test Report Revision 1 Page 33 of 45 Table 4.6.1 Reactivity Computer Checkout Input Parameters to the Reactivity Computer' Group f, I* ,i 1 0.000213 0.000206 0.0128 2 0.001334 0.001294 0.0317 3 0.001210 0.001174 0.1208 4 0.002610 0.002531 0.3210 5 0.000960 0.000931 1.4025 6 0.000233 0.000226 3.8751 ZP, 0.006559 0.006363 Prompt Neutron Lifetime = 17.89 Importance Factor -I = 0.97 Delayed Neutron Fraction - 4 3 efr = 0.006363 2

Positive Insertion Period Check Collection # Atime Period Calculated Measured % Difference (sec) (sec) reactivity reactivity 1 53.4 77.1 70.08 69.50 0.83%

2 55.6 80.3 67.99 68.68 -1.00%

3 57.2 82.6 66.55 67.23 -1.02%

4 61.5 88.7 63.03 64.20 -1.83%

5 70.8 102.1 56.46 58.12 -2.87%

average 1_ 64.82 65.55 -1.18%

2 Negative Insertion Period Check Collection # Atime period Calculated Measured % Difference (sec) (sec) reactivity reactivity 1_-145.4 209.8 -44.12 -43.94 0.41%

average -44.12 -43.94 0.41%

2 Reactivity Computer inputs from Reference [SOR], Table 5.1-lb 2

Measured data from Cycle 11 (RFO 10) performance of EST-923, Reference 5.10.

8/3 0/02 Revision 1

Harris Nuclear Plant Unit 1, Cycle 11 Startup Test Report Revision 1 Page 34 of 45 Table 4.6.2 Low Power Physics Test Results Summary

,p pqmt C1,nfigkurationha ~:esird iference iPeitd <' Acpac~

ARO 2090 2105 16 _50 CBB-in 1924 1934 10 +/-4_50

_____________ :~>~,ntr6VRod Wyorths (pcm~

Bank~ ~ 's' Di e renrdcePditci-, cei~c Vic 2 , Wiff I

CBB 1063 1042 21 1.98 +/- 10%

SBA 952 981 -29 -3.05 +/- 15%

SBB 866 887 -21 -2.44 +/- 15%

SBC 292 306 -14 -4.83 +/- 100 CBA 509 526 -17 -3.42 +/-100 CBC 916 952 -36 -3.95 +/-15%

CBD 879 1001 -122 -13.9 +/-15%

Sum of worths 5477 5696 /,= 0.962 0.9 _ I/p

1.1 a' ~I[IZP Temperature Coefficient (cilF RC@S 2 070 aaMeasuredl )Prpdifd 1 ifeeceAcc~epiance. ,

ITC -5.03 -4.94 0.09 Difference +/- 2 MTC -3.49 -3.40 0.09 Measured _<+5

-Diff~rential oron Wort (Pcr/p)>

,cqonfigdration, MesrdPeitd i %Dferend ~ D>cetance r CBB going in 3 -6.40 -6.09 4.8% +10%

and predicted data obtained from Cycle 11 (RFO 10) perform ance of EST-923, S Reference 5.10 2 % Difference = [(Measured - Predicted) / Measured]

100.0 3 DBWcBB going in = (WorthceB) / (Boron EndpointB.-i - Boron EndpointARo) 8/30/02 Revision 1

Harris Nuclear Plant Unit 1, Cycle 11 Startup Test Report Revision I Page 35 of 45 Table 4.7.1 Rod Worth Measurement of the Reference Bank Initial Final Average Integral Differential Step Step Step Worth Worth (pcm) (pcm) 70 0 35.0 1063.45 0.73 76 70 73.0 1012.23 2.27 85 76 80.5 999.44 2.11 101 85 93.0 980.61 3.64 111 93 102.0 950.74 3.94 111 101 106.0 919.24 4.31 121 111 116.0 876.22 4.66 128 121 124.5 829.78 4.95 136 128 132.0 794.06 5.84 146 136 141.0 744.98 6.09 154 146 150.0 688.65 6.83 162 154 158.0 628.56 6.82 169 162 165.5 577.71 7.92 176 169 172.5 523.34 8.60 183 176 179.5 462.60 9.63 190 183 186.5 396.24 9.69 194 190 192.0 325.84 11.47 200 194 197.0 283.71 11.09 206 200 203.0 213.19 12.51 212 206 209.0 132.58 9.85 225 212 218.5 78.40 4.17 225 225 225.0 0.00 4.17 Measured data from Cycle 11 (RFO 10) performance of EST-923, Reference 5.10 8/30102 Revision 1

Harris Nuclear Plant Unit 1, Cycle 11 Startup Test Report Revision 1 Page 36 of 45 Figure 4.7.1 Integral Worth of the Reference Bank 1100 1000 900 800 700 i

600 0

0 500 C) 400 300 200

-- Predicted 100

-UMeasured - _

0 0 25 50 75 100 125 150 175 200 225 Reference Bank Position (steps)

Measured data from Cycle 11 (RFO 10) performance of EST-923 [Reference 5.10]

8/30/02 Revision 1

Harris Nuclear Plant Unit 1, Cycle 11 Startup Test Report Revision 1 Page 37 of 45 Figure 4.7.2 Differential Worth of the Reference Bank 13 12 11 10 9

0 E

8 0 7 0

6 r

Q) 5 1i 4

3 2

-- s Predicted 1 S-r, -*-Measured 0

0 25 50 75 100 125 150 175 200 225 Reference Bank Position (steps)

Predicted data taken from SOR [Reference 5.2], Figure 5.3-1 Table contains measured data from Cycle 11 (RFO 10) performance of EST-923 [Reference 5.10].

8130/02 Revision I

Harris Nuclear Plant Unit 1, Cycle 11 Startup Test Report Revision 1 Page 38 of 45 Table 4.7.3 FSAR Chapter 14 Tests Test Summary Explanation Description Communications System Not impacted by SGR/PUR, system performance is monitored during routine operation.

Annunciator System Not impacted by SGR/PUR, system performance is monitored during routine operation, maintenance, and surveillance tests.

Reactor Protection System Minimal impact on system from SGR/PUR, Engineered Safety Features system performance discussed in sections Actuation Logic 3.1 and 3.2.

Reactor Protection System Minimal impact on system from SGR/PUR, Engineered Safety Features system performance discussed in sections Actuation Response Time Test 3.1 and 3.2.

Piping Vibration Minimal impact on system from SGR/PUR, system performance discussed in section 3.5.

Metal Impact Monitoring Minimal impact on system from SGR/PUR, system performance discussed in section 3.6.

Radiation Monitoring System Not impacted by SGR/PUR, system performance is monitored during routine operation.

Excore Nuclear Instrumentation Minimal impact on system from SGR/PUR, (NIS) system performance discussed in section 4.3.

Emergency Diesel Not impacted by SGR/PUR, system performance is monitored during routine operation and surveillance tests.

Fire Protection System Not impacted by SGR/PUR, system performance is monitored during routine operation and surveillance tests.

Normal Service Water Not impacted by SGRIPUR, system performance is monitored during routine operation.

8130/02 Revision 1

4 Harris Nuclear Plant Unit 1, Cycle 11 Startup Test Report Revision 1 Page 39 of 45 7- I Emergency Service Water Minimal impact on system from SGR/PUR, system performance discussed in section 3.11.

Compressed and Instrument Air Not impacted by SGR/PUR, system Systems performance is monitored during routine operation.

Reactor Coolant System Minimal impact on system from SGR/PUR, Hydrostatic Test system performance discussed in section 4.12.

RTD/TC Cross Calibration Test Minimal impact on system from SGR/PUR, system performance discussed in section 4.9.

Pressurizer Relief Tank (PRT) Test System operation was reviewed and determined to be acceptable by analytical methods.

Safety Injection System System performance discussed in section Performance Test 3.12.

High-Head Safety Injection System Not impacted by SGR/PUR, system Check Valve Test performance is monitored during routine startup and surveillance tests.

Safety Injection (SI) Accumulator Not impacted by SGR/PUR, system Test performance is monitored during routine operation startup and surveillance tests.

Residual Heat Removal System Not impacted by SGR/PUR, system Cold Test performance is monitored during routine startup and surveillance tests.

Residual Heat Removal System Not impacted by SGR/PUR, system Hot Test performance is monitored during routine startup and surveillance tests. System operation was reviewed and determined to be acceptable by analytical methods.

Containment Spray System Test Not impacted by SGR/PUR, system performance is monitored during routine surveillance tests. System operation was reviewed and determined to be acceptable by analytical methods.

Chemical and Volume Control Not impacted by SGR/PUR, system Cold Test performance is monitored during routine operation and surveillance tests.

8/30/02 Revision I

dP, Harris Nuclear Plant Unit 1, Cycle 11 Startup Test Report Revision 1 Page 40 of 45 I

Chemical and Volume Control Hot Not impacted by SGR/PUR, system Test performance is monitored during routine operation and surveillance tests. System operation was reviewed and determined to be accentable by analytical methods.

Auxiliary Feedwater System Test Minimal impact on system from SGR/PUR, system performance discussed in section 3.3.

Fuel Handling Equipment System Not impacted by SGR/PUR, system Test performance is monitored during fuel transfers and surveillance tests.

Fuel Pool Cooling and Cleanup Not impacted by SGR/PUR, system System Test performance is monitored during normal operation.

Component Cooling Water Impact on system from SGR/PUR, system performance discussed in section 3.11.

Gaseous Waste Processing System Not impacted by SGR/PUR, system Test performance is monitored during routine operation and surveillance tests. System operation was reviewed and determined to be acceptable by analytical methods.

Solid Waste Processing Test Not impacted by SGR/PUR, system performance is monitored during routine operation.

Liquid Waste Processing System Not impacted by SGR/PUR, system Test performance is monitored during routine operation.

Containment Isolation Test Not impacted by SGR/PUR, system performance is monitored during routine startup and surveillance tests.

Containment Integrated Leak Rate Minimal impact on system from SGR/PUR, Test and Structural Integrity Test system performance discussed in section 3.4.

Reactor Coolant System Hot Not impacted by SGR/PUR, system Functional Test performance is monitored during routine operation. System operation was reviewed and determined to be acceptable by analytical methods.

Piping Thermal Expansion and Impact on system from SGR/PUR, system Dynamic Effects Test performance discussed in section 3.5 Pressurizer Pressure and Level System operation was reviewed and Control Test determined to be acceptable by analytical methods.

8/30/02 Revision 1

Harris Nuclear Plant Unit 1, Cycle 11 Startup Test Report Revision 1 Page 41 of 45 Main Steam System Test Impact on system from SGR/PUR, system performance discussed in section 3.13.

Feedwater System Test Impact on system from SGR/PUR, system performance discussed in section 3.13.

Condensate System Test Impact on system from SGR/PUR, system performance discussed in section 3.13.

Turbine Generator Test Impact on system from SGR/PUR, system performance discussed in section 3.14.

Circulating Water System Test System operation was reviewed and determined to be acceptable by performance testing, discussed in section 4.14.

Condenser Vacuum and System operation was reviewed and Condensate Makeup System determined to be acceptable by performance testing, discussed in section 4.14.

Waste Processing Computer Test Not impacted by SGR/PUR, system performance is monitored during routine operation.

Containment Ventilation and Not impacted by SGR/PUR, system Cooling, Primary Shield and performance is monitored during routine Reactor Supports Cooling System startup and surveillance tests.

Test Plant HVAC Test Not impacted by SGR/PUR, system performance is monitored during routine operation.

Engineered Safety Features System operation was reviewed and Integrated Test determined to be acceptable by testing described in sections 3.1 and 3.2.

Process Computer Test Not impacted by SGR/PUR, system performance is monitored during routine operation.

Boron Recycle Test Not impacted by SGR/PUR, system performance is monitored during routine operation.

Refueling Water Storage Tank Not impacted by SGR/PUR, system Test performance is monitored during routine operation. System operation was reviewed and determined to be acceptable by analytical methods.

Primary Makeup Water System Not impacted by SGR/PUR, system Test performance is monitored during routine operation.

8/30/02 Revision 1

4' -

Harris Nuclear Plant Unit 1, Cycle 11 Startup Test Report Revision 1 Page 42 of 45 Rod Control System Test Not impacted by SGR/PUR, system performance is monitored during routine operation.

Passive Safety Injection System Not impacted by SGR/PUR, system Check Valve Test performance is monitored during routine startup and surveillance tests. System operation was reviewed and determined to be acceptable by analytical methods.

Containment Recirculation Sump System operation was reviewed and Test determined to be acceptable by analytical methods.

Containment Vacuum Relief Test System operation was reviewed and determined to be acceptable by analytical methods.

Combustible Gas Control System System operation was reviewed and In Containment Test determined to be acceptable by analytical methods.

Gross Failed Fuel Detection System operation was reviewed and System Test determined to be acceptable by analytical methods.

Essential Services Chilled Water System operation was reviewed and System Test determined to be acceptable by analytical methods.

Stud Tensioner Hoist Load Test Not impacted by SGR/PUR, system performance is monitored during routine operation.

Polar Crane Test Summary System was revised extensively for SGR and restored to pre-existing configuration prior to startup. System operation was reviewed and determined to be acceptable by analytical methods.

Feedwater Heater Drain, Level and Not impacted by SGR/PUR, system Bypass Control Systems Test performance is monitored during routine operation. System operation was reviewed and determined to be acceptable by analytical methods.

Seismic Instrumentation Test Not impacted by SGR/PUR, system performance is monitored during routine operation.

8/30/02 Revision I

.- - I-Harris Nuclear Plant Unit 1, Cycle 11 Startup Test Report Revision 1 Page 43 of 45 Extraction Steam System Test Not impacted by SGR/PUR, system performance is monitored during routine operation. System operation was reviewed and determined to be acceptable by analytical methods.

Primary Sampling System Test Not impacted by SGR/PUR, system performance is monitored during routine operation.

Secondary Sampling System Test Not impacted by SGR/PUR, system performance is monitored during routine operation.

Loss of Instrument Air Test System operation was reviewed and determined to be acceptable by analytical methods.

Containment Building Hot Not impacted by SGR/PUR, system Penetration Test performance is monitored during routine operation.

Simulated Loss of On-Site Power System operation was reviewed and Test determined to be acceptable by analytical methods.

AC Distribution System Optimum System operation was reviewed and Operating Voltage Test determined to be acceptable by analytical methods.

Auxiliary Feedwater Turbine System operation was reviewed and Pump Two-Hour Run Test determined to be acceptable by analytical methods.

Power Ascension Test The power ascension program described in PLP-632T controlled post SGR/PUR testing.

Moveable Incore Detector Test Minimal impact on system from SGR/PUR, system performance discussed in section 4.4.

Rod Control and Position Not impacted by SGR/PUR, system Indication System Test performance is monitored during routine startup and surveillance tests.

Rod Drive Mechanism Timing Minimal impact on system from SGR/PUR, Test system performance discussed in section 4.1.

Rod Drop Time Measurement Test Minimal impact on system from SGR/PUR, system performance discussed in section 4.1.

8/30/02 Revision I

Harris Nuclear Plant Unit 1, Cycle 11 Startup Test Report Revision I Page 44 of 45 1 -, -

Reactor Coolant System Flow Minimal impact on system from SGR/PUR, Measurement Test system performance discussed in section 4.2.

Reactor Coolant System Flow System operation was reviewed and Coastdown Test determined to be acceptable by analytical methods.

Calibration of Nuclear Minimal impact on system from SGR/PUR, Instrumentation Test system performance discussed in section 4.3.

Rod Control System Test System operation was reviewed and determined to be acceptable by analytical methods.

Flux Distribution Measurement Minimal impact on system from SGR/PUR, Test system performance discussed in section 4.4.

Core Performance Test Minimal impact on system from SGR/PUR, system performance discussed in section 4.5.

Power Coefficient Measurement Minimal impact on system from SGRIPUR, Test system performance discussed in section 4.6.

Control Rod Reactivity Worth Test Minimal impact on system from SGR/PUR, system performance discussed in section 4.7.

Boron Reactivity Worth Test Minimal impact on system from SGR/PUR, system performance discussed in section 4.8.

Automatic Rod Control Test System operation was reviewed and determined to be acceptable by analytical methods.

Steam Generator Moisture Impact on system from SGR/PUR, system Carryover Test performance discussed in section 4.10.

Load Swing Test Impact on system from SGR/PUR, system performance discussed in section 4.11.

Large Load Reduction From 75 System operation was reviewed and Percent Power Test determined to be acceptable by analytical methods.

Turbine Trip From 100 Percent System operation was reviewed and Power Test determined to be acceptable by analytical methods.

Remote Shutdown Test System operation was reviewed and determined to be acceptable by analytical methods.

8130/02 Revision 1

Harris Nuclear Plant Unit 1, Cycle 11 Startup Test Report Revision I Page 45 of 45 Loss of Offsite Power Test System operation was reviewed and determined to be acceptable by analytical methods.

Pressurizer Heaters and Spray System operation was reviewed and Valves Capability Test determined to be acceptable by analytical methods.

Gross Failed Fuel Detection System operation was reviewed and System Test determined to be acceptable by analytical methods.

Pressurizer Continuous Spray Flow Not impacted by SGR/PUR, system Verification Test performance is monitored during routine operation.

Reactor Coolant System Leakrate Minimal impact on system from SGR/PUR, Test system performance discussed in section 4.12.

Main Steam and Feedwater Impact on system from SGR/PUR, system Systems Test performance discussed in section 4.13.

Shield Survey Test Not impacted by SGR/PUR, system performance is monitored during routine operation. System operation was reviewed and determined to be acceptable by analytical methods.

Loss of Feedwater Heater(s) Test System operation was reviewed and determined to be acceptable by analytical methods.

Main Steam Isolation Valve Test Not impacted by SGR/PUR, system performance is monitored during routine startup and surveillance tests.

Steam Generator Test for Water Hammer was a concern for D4 SGs; Condensation Water Hammer the feedring design of the A75 SGs eliminates this concern. System operation was reviewed and determined to be acceptable by analytical methods.

Steam Turbine-Driven and Motor- System operation was reviewed and Driven Auxiliary Feedwater determined to be acceptable by analytical Pumps Endurance Test methods.

Resistance Temperature Detector Test is no longer applicable with current (RTD) Bypass Flow Verification RTD configuration.

Test Secondary Sampling System Test Not impacted by SGR/PUR, system performance is monitored during routine operation.

8/30/02 Revision I

HNP-02-124, Cycle 11, Startup Test Report Revision 1

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HNP-02-124, Cycle 11, Startup Test Report Revision 1

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