JAFP-89-0140, Cycle 9 Startup Testing Rept

Cycle 9 Startup Testing Rept

ML20235Q097
Person / Time
Site:	FitzPatrick
Issue date:	02/21/1989
From:	Fernandez W POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
JAFP-89-0140, JAFP-89-140, NUDOCS 8903020497
Download: ML20235Q097 (7)

Text

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James A. FitzPatrick Cycle 9 Start-up Testing Report Cycle 9 on line operations commenced November 24, 1988, ending a 89 day refueling outage. The startup testing program commenced l November 20, 1988 and was completed February 10, 1989. The test program was conducted in accordance with Reactor Analyst Procedure 7.3.30 " Refuel Startup Program".

Core Loading and Verification:

A full core offload and reload was performed during this-outage in which three hundred and seventy six bundles were relocated, one hundred eighty four bundles were discharged, and one hundred eighty four n..w bundles were loaded. The cycle 9 fuel bundle breakdown is as follows:

1. GE7 fuel, 2.99% enrichment, 188 bundles, Reload 6

2. GE8 fuel, 3.19% enrichment, 184 bundles, Reload 7

3. Westinghouse LTA's, 2.99% enrichment, 4 bundles, Reload 7

4. GE8 fuel, 3.39% enrichment, 32 bundles, Reload 8

5. GE8 fuel, 3.36% enrichment, 152 bundles, Reload 8 The final core loading was verified in accordance with' RAP-7.2.4

" Reactor Fuel Verification" using an underwater television camera and video recorder. The videotape was independently checked by QA personnel and documented by QA Surveillance Report Audit 1283.

Control Rod Drive Tests:

Twenty five control blades were replaced during the outage and four control blades were shuffled to new locations. Prior to start-up, Surveillance test (ST-20K) " Control Rod Exercise / Venting" was performed on all 137 control rods to demonstrate that each rod la coupled to its drive mechanism, and to check that the normal operating drive speed of each control rod satisfies a travel timing test. Sventy control rods required a timing adjustment.

Control tod scram time testing was performed on all 137 control rod drives prior to reaching 40% rated core thermal power.

The results were as follows:

Notch Tech Spec . Limit Actual for 137 rods 46 .338 sec .307 sec 38 .923 see .727 sec 24 1.992 sec 1.481 sec 04 3.554 sec 2.606 see The average of the scram insertion times of the three fastest operable control rods for all groups of four control rods in a two by two array were less than the maximum times allowed by the Technical Specifications.

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I Shutdown Margin Test:

Initial Criticality for cycle 9 was achieved on November 20,

'1988. Shutdown margin was demonstrated using the insequence i' critical method, which showed the core to have a shutdown margin of 1.819% delta k/k. The predicted shutdown margin was 1.42% delta k/k, indicating a difference of 0.004 delta k between the critical eigenvalue used as a basis for the prediction in the Cycle Managment Report and the actual critical eigenvalue demonstrated by the insequence critical test.

1 Control Rod Sequence:

Sequence A2 was loacae into the RWM program in accordance with the requirements of Banked 'osition Withdrawal Sequence and Reduced Notch Worth Procedure. Prior to startup, functional tests of the RWM and the RSCS were performed to demonstrate their operability.

SRM Performance Check:

SRM Functional Testing and SRM Range Channel checks were performed prior to startup to demonstrate operability of the SRM monitoring instrumentation. During reactor startup, an SRM/IRM Overlap check was performed to demonstrate that each IRM was on scale before any SRM exceeded the rod block setpoint.

Reactivity Anomaly Check:

A comparison of the calculated and actual control rod density was performed at 100% rated core thermal power and 95.7% rated core flow. The actual rod inventory was 471 notches inserted which showed close agreement to the predicted notch inventory of 464 notches, where 1% delta k/k is equivalent to 260 notches.

Power Distribution Measurements:

Core power distribution was monitored throughout the power ascension using the Traversing Incore Probe System (TIP) and the Local Power R?nge Monitors (LPRMS). LPRM calibrations were performed at 30%, 75%, 90%, and 100% of rated core thermal power.

Fuel thermal limits were maintained uithin Technical Specification limits.

Core Power Symmetry:

Core power symmetry was checked at 50%, 75%, and 100% rated core thermal power. In all cases, the maximum percent difference in power level of symmetrically located fuel bundles was found to be less than 10%. The actual values calculated are shown below.

Test Plateau Maximum % Difference l

50% 5.95%

75% 5.56%

100% 5.20%

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Manual Heat' Balance:

A manual heat balance was performed per RAP 7.3.3 at 25%, 50%,

75%, and 100%. rated core thermal power and in each case thermal power was found to be within 50 MW (2% of rated) of the plant computer thermal power calculation. The values are tabulated below.

Test Plateau Hand Calc. Computer Calc 25% 689.6 MW 723.5 MW 50% 1210.6 MW 1218.7 MW 75% 1850.8 MW 1839.4 MW 100% 2356.5 MW 2352.4 MW LPRM Response Test:

During scram time testing, when control rod insertions and withdrawals were performed, a LPRM response test was conducted on all 124 incore fission detectors. This test verified that each operable detector in connected to the appropriate flux amplifier.

Five LPRM assemblies were replaced during.the outage. Two LPRM detectors were placed in bypass because they read downscale at 35%

rated power.

Plant Computer Checkout:

The computer databank for Cycle 9 operations was installed and verified to be correct per Reactor Analyst Procedure 7.3.17 " Core Monitoring Software and Database Changes". "' h e plant computer calculations were compared to offline core performance programs at 25%, 50%, 75%, and 100% rated thermal power. The results showed close agreement in location and magnitude of all thermal limits.

Core Flow Evaluation

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- Core Flow Indication Calibrations were performed at 75% and 100% rated power per Reactor Analyst Procedure 7.3.7. The indicated flow matched the calculated flow within less than .5%,

which is well within the !2.5% tolerance assumed for the statistical uncertainty in the Licensing Topical Report.

Determination of Rated Drive Flow:

A rated drive flow calculation was performed at 100% raged power, and the results showed that a drivg flow of 33.0 x 10 lb/hr produces the rated core flow of 77.0 x 10 lb/hr. The origina1 6 design value for rated drive flow (with 7x7 fuel) was 34.2 x 10 lb/hr.

TIP Axial Alignment -and System Checkout:

Prior to startup, the core top and bottom limits for each LPRM string were set by hand cranking the TIP probes to the top of each instrument tube. These liuits were checked per Reactor Analyst Procedure 7.3.14 "TIP SYSTEM" at full power by checking the location of spacer dips on the flux traces.

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• 4 TIP Reading Uncertainty:

The standard deviation between symmetrically located TIP strings was determined from BASE distributions obtained at 50, 75 and 100 percent power. -The resulting TIP reading uncertainties were calculated to be 2.4% at 50% power, 2.8% at 75% power, and 3.7% at 100% power. These values are well within the 7.1% TIP reading uncertainty assumed in the Licensing Topical Report.

Core Thermal Hydraulic Stability:

Data was acquired from the APRMs and LPRM detectors at 25 and 75 percent power in accordance with ST-5S, " Neutron Instrumentation Noise: Monitoring". This information will serve as baseline data for the operating cycle when Technical Specifications require performance of ST-SS.

Other Start-up. tests performed to satisfy Technical Specification Requirements included:

1. Chemical and Radiochemical tests - Performed per PSP-1,

" Reactor Water Sampling and Analysis", and PSP-16, " Guidelines for Start-up, Shutdown, and Scram" which ensures Technical Specification requirements with regard to reactor water chemistry are met.

2. Reactor Vessel Heatup - Performed in accordance with ST-26J, "Heatup and Cooldown Temperature Checks". The reactor vessel heatup was performed in accordance with the requirements of ST-26J which requires reactor coolant system pressure and temperature be at or to the right of curve C shown in Figure 3.6-1, and the maximum temperature change during any one hour equal to or less than 100 degrees fahrenheit.

3. IRM Performance - Performed ST-5C, "IRM-APRM Instrument Range Overlap Check" which demonstrated that each APRM channel was on scale before any IRM exceeded the rod block setpoint.

4. Safety Relief Valves - Performed ST-22B, " Manual Safety Relief Valve Operation and Valve Monitoring System Functional Test". .

The acceptance criteria of ST-22B were satisfied which I demonstrated (1) that each safety relief valve opens and closes fully through operation of control switches on 09-40 control room panel and the remote 02 ADS-071 panel, (2) the valve monitoring system operated satisfactorily to indicate valve position, (3) opening of each safety relief valve was verified by observing a ten percent or greater closure of the turbine bypass valves.

5. Main Steam Isolation valves - performed ST-1B, "MSIV Fast Closure" which demonstrated that all MGIV's close within the Technical Specification and IST stroke time of 3 to 5 seconds.

6. RCIC System - Performed ST-24A, "RCIC Pump and Valve Operability Test" which verified RCIC pump, motor, and valve j operability.

.A A Simulated Automatic Actuation test was performed in accordance with F-ST-24E which demonstrated the ability of  !

the RCIC system to deliver a flow' rate of 400 gpm.

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7. HPCI System - Performed ST-4B, "HPCI Pump and MOV Operability Tests" to verify operabil. of the MPCI turbine and pump assembly, and associated motor operated valves. A simulated Automatic Actuation Test was performed in accordance with F-ST-4A which demonstrated the ability of the HPCI system to deliver a flow rate of 4250 gpm.

The following start-up tests described in Chapter 13.7 of the FSAR were not performed due to one or more of the following reasons:

(1) Performance of the test challenges reactor protection and-safety systems.

(2) The test places the plant in a degraded condition.

(3) The test measures system dependent parameters rather than fuel dependent parameters and no modifications were performed which would alter system response.

(4) The test measures parameters which needed to be established or verified during the initial plant startup_before the plant had any operating history.

Turbine and Generator trip tests: the purpose of this test was to demonstrate the response of the reactor and its control systems to protective trips in the turbine and generator.

Loss of turbine generator and loss of offsite power: The purpose of this test was.to determine reactor transient performance during the loss of the main generator and all off site power.

Pressure Regulator test: the main purpose of this test was to determine the optimum setting for the pressure control loop by analysis of transients induced in the reactor pressure control system by means of the pressure regulators.

Feedwater level control: The main purpose of this test was to i adjust the feedwater control system for acceptable reactor water level control.

Recirculation pump trip test: The purpose of this test was to evaluate the recirculation flow and reactor power level transients following a single and then dual pump trip.

Shutdown from outside the control room: This test demonstrates that the reactor can be brought from a steady state power level to a point where reactor cooldown is initiated using controls outside the control room.

RHR steam condensing mode demonstration: This test demonstrates the RHR system is capable of operating in the steam condensing mode.

System Expansion test: The purpose of this cist is to verify that the reactor drvwell piping system is free and unrestrained with regard.to thermal expansion.

Turbine Bypass Valve Measurement test: The purpose of this test is j

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to demonstrate the ability of the pressure regulator to minimize the reactor pressure disturbance during an abrupt change in steam j flow by tripping open and closing a turbine bypass valve.

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Flux Response to Control Rods: The purpose of this test is to demonstrate the stability of the core local power / reactivity

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feedback mechanism to small perturbations caused by rod movement.

Selected Process Temperatures: The purpose of this test was to establish the minimum recirculation pump speed that ensures adequate nixing in the lower vessel plenum, and to assure that the measured bottom head drain temperature corresponds to the bottom head coolant temperature during normal operation. I Vibration Measurements: This test performed vibration measurements on various reactor components to demonstrate the mechanical integrity of the system to flow induced vibrations.

Flow Control: The purpose of this test was to demonstrate a stable plant response to recirculation flow control changes.

Radiation Measurements: This test determined pre-operational background radiation levels in the plant environs to assure protection of plant personnel during plant operation.

Recirculation MG Set Speed Control: The purpose of this test was  ;

to determine the speed control characteristics of the MG se s, obtain acceptable speed control system performance, and deterr.ne maximum allowable pump speed.

Reactor Water Cleanup System: The purpose of this test was to demonstrate specific aspects of the mechanical operability of the RWCU system.

Reactor Water Level: The purpose of this test was to verify the calibration and agreement of the GEMAC and YARWAY water level instrumentation under various conditions. The instrumentation is presently calibrated in accordance with Technical Specifications.

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'... Jim *s A. FitzPatrick Nuclear Powsr Plant

, P.O. Box 41

, Lycoming New York 13093 l

• 315 342-3840

1. > NewYorkPower d

(# Authority ifa'n'et'o"e, "

February 21, 1989 JAFP-89-0140 U n i t e ('. States Nuclear Regulatory Commission Document Control Desk Washiagton, D.C. 20555

Subject:

JAMES A. FITZPATRICK NUCLEAR POWER PLANT DOCKET No. 50-333 CYCLE 9 START-UP TESTING REPORT Gentlemen:

Enclosed you will find the Cycle 9 Start-up Testing Report for the

.rames A. FitzPatrick Nuclear Power Plant, which is submitted to you in accordance with the reporting requirements of section 6.9.A.1 of the Plant Technical Specifications.

We trust you will find this information satisfactory. However, should you desire wore information, please contact Mr. David Burch at (315) 349-6311.

Very Truly Yours,

, ,/) -

4,LI TL AM FERNANDEZ WF:vw lg CC: U.S. Nuclear Regulatory Commission Region 1 NRC Resident Inspector WP0 for RMS Headquarters Distribution G. Rorke D. Burch Document Control Center

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