JAFP-87-0441, Forwards Cycle 8 Startup Test Rept,Per Reg Guide 10.1 & Tech Spec 6.9.A.1

Forwards Cycle 8 Startup Test Rept,Per Reg Guide 10.1 & Tech Spec 6.9.A.1

ML20214S724
Person / Time
Site:	FitzPatrick
Issue date:	05/29/1987
From:	Radford Converse POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK
To:	NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM)
References
RTR-REGGD-10.001 JAFP-87-0441, JAFP-87-441, NUDOCS 8706090387
Download: ML20214S724 (6)

Text

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Jzmas A. Fit:Pstriels l

Nuclett Powst Pl:nt l PO Bos 41 ,

( Lycoming. New York 13093 f 31$ 342.384o I l i l Radford J. Converse  ;

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! #D NewYorkPbwer tv Authority

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May 29, 1987 JAFP-87-0441 i

United States Nuclear Regulatory Commission  !

Document Control Desk Washington, DC 20555 i

Subject:

James A. FitzPatrick Nuc1 ar Power Plant  !

Docket No. 50-333

. Cycle 8 Start-Up Testing Report Centlement Enclosed you will find the Cycle 8 Start-up Testing Report for the James A. >

FitzPatrick Nucicar Power Plant, which is submitted to you in accordance with Regulatory Guide 10.1 and Plant Technical Specifications paragraph 6.9.A.1, We trust you will find this information satisfactory. Ilowever, should you desire additionni information, please do not hesitate to contact Mr. David Burch at (315) 349-6311.

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RADFORD J. CONVERSE  :

RJC admh Enclosure  !

i CC: U.S. Nuclear Regulatory Commission Region 1. Attention: W.T. Russell, Regional Administrator NRC Resident Inspector .

WPO for RMS lleadquarters Distribution G. Rorke W. Fernandez l D. Burch R. Cliane Document Control Center pa87060 60387 03000,22870529 PDN 4y ,1 ,

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I i James A. FitzPaerick Cycle 8 Start-up Testing Report

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1 j Cycle 8 on-line operations commenced April 30, 1987, ending a 105 day ,

! refueling outage. The start-up testing program was conducted from March 24, 1987 i through May 15, 1987 utilizing Reactor Analyst Procedure (RAP) 7.1.17. " Refueling j Start-up Program". i 1

Core Loading and Verification l Refueliig activities this outaga included a full core offload and reload. l During the riload three hundred and seventy-two irradiated assemblies and one ,

hundred and eighty-eight new fuel assemblies were loaded. Of these new j assemblies, tour were Westinghouse Lead Test Assemblies of the QUAD + design and f I the remaindet were General Electric GE8 fuel design. The finni loading map is -

attached as Figure 1.  ;

! Inverse multiplication plots were used throughout the fuel loading process j to monitor the subcriticality of the core. The final core loading was verified I in accordance with RAP-7.2.4, " Reactor Core Fuel Verification", using an  :

j underwater television camera and recorder. This has been documented by QA Audit l 1 number 1170. '

i f Control Rod Drive Testat L l i i Prior to start-up, surveillance tests F-ST-23B, " Control Rod Coupling i Integrity Test" and F-ST-20K, "CRD Exercise / Venting Timing Test", were  :

satisfactorily completed for all 137 control rods. Twenty control blades were l repinced during the outage.

Control rod scram time testing was conducted in accordance with RAP-7.3.10, i j " Control Rod Scram Time Evaluation", prior to reaching 40% rated core thermal  ;

j power. The results were as follows: r

\l Notch Tech Spec Average of

! Observed Limit (acc) 137 rods (ace) i i  !

j 46 .338 .321 i 38 .923 .741 l j 24 1.992 1.499 I

04 3.554 2.633 1

j The average time to notch 38 required the alternata method of determining r i the Minimum Crittent Power Itatio (MCPR) limit as outlined in Technical Specifica- i j tion 3.1.B.2. No adjustment to the MCPR limit was required at this point in the l' j cycle.

The average of the scram insertion times of the three fastest operable i control roda for all groups of four control rodn in a two by two array wore less  !

than the maximum times allowed by the Technical Specittentions. l 1 ,

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l Shutdown Margin Test:

Initial criticality for this fuel cycle was achieved on April 22, 1987.

At this time, shutdown margin was demonstrated using the insequence critical method. This test demonstrated a shutdown margin of 1.511% AK.

Reactivity Anomaly Check:

A comparison of the expected and actual control rod density was performed at 100% rated core thermal power and 99.3% rated core flow, in accordance with RAP-7.3.6, " Reactivity Follow". The actual control rod inventory was 456 notches inserted. This showed good agreement with the predicted inventory of 432 notches inserted. (2250 notches is equivalent to 11% AK)

Power Distribution Mensurements:

Core power distribution was monitored throughout the power ascension testing with the Traversing Incore Probo (TIP) System and Local Power Range Monitors (LPRM). LPRM calibrations were performed at 35%, 50% 75%, and 100% rated core thermal power. Reactor thermal limits were maintained within Technical Specification Limits.

Coro Power Symmetry:

Core power symmetry was checked at 50%, 75%, and 100% rated core thermal power. The maximum percent difference in power level of symmetrically located fuel bundles was found to be less than 10%.

LPRM Responne Tent During scram time testing, when control rod insertions and withdrawals were being modo, a LPRM renponse test was conducted on all 124 LPRM detectors. This verified neutron responne and demonstrated that the detectors were connected to their proper flux amplifiers. Twenty LPRM nonemblien were repInced during the outage.

, Prenmura Regulator T nt KAP-7.3.18. "Prennure Regulator Tents", was performed at 25% rated core thermal power. The doeny ratio of the responne to changes in the preneure regulator was verified to be less than 1.0. The take-over capability of the back-up pressure regulator was demonstrated during simulated failures of the controlling regulator.

Procann Computar Checkout:

The procons computer databank for Cycle H operation van installed and satisfactorily checked out over the time period of April 16-17, 1987. This work van performed by site pernonnel with the assistance of pornonnel from the General Electric Nuclear Data Hanks group in San June, CA.

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l Core Performance calculations were made at 35%, 50%, 75% and 100% rated core thermal power using manual, process computer, and the off-line Back-Up Core i Limits Evaluation methods. In each case, the power levels were in agreement.

l Values and core locations for Maximum Fraction of Limiting Critical Power Ratio I (MFLCPR), Maximum Fraction of Limiting Power Density (MFLPD), and Maximum Average l Planar linear heat generation ratio to limits (MAPRAT) determined by BUCLE agreed j closely with the plant process computer calculations.

TIP Axial Alignment and System Checkout:

Since some TIP tubing was replaced, tests to verify the proper alignment and operation of the TIP System were performed in accordance with F-IMP-7.26 and RAP-7.3.14. Some adjustments were required to make the system operate correctly for the OD-1 Whole Core LPRM Calibration and Base Distribution computer program.

TIP System Uncertainty Test:

The TIP Reading uncertainty was determined to be 5.776% from a comparison of symmetric LPRM string Base distributions taken at 100% rated core thermal power.

Using the annumed uncertainty values of 3.4% for LPRM extrapolation and 3.7% for TIP Software modeling, the total TIP System uncertainty was determined to be 7.66%. This is below the 8.7% value assumed in the Licensing Topical Report and is, therefore, satisfactory.

Core Flow Evaluation:

RAP-7.3.7, " Core Flow Evaluation and Indication Calibration", was performed at 85% rated core thermal power with results showing a .92% difference between indicated flow and calculated. This was within the 22.5% required by the procedure and assumed for the statistical uncertainty in the Licensing Topical Report.

At 100% rated core thermal power and 100% rated core flow by indication, another evaluation was performed. Again, the results were within the 12.5%

criterion. Closer inspection of the data revealed, however, that the output from

) the flow summer 02-3-SUH-/5D was reading low. Instrument and Control Department corrected the flow nummer problem and adjusted the output of the proportional amplifiera 02-3-PA-76A and 768 to make the final indicated flow read closer to the calculated value.

A third core flow evaluation was performed at 100% rated core thermal power and 98% rated core flow by indication. Results showed the difference between indicated and calculated to be 1.65%. One final adjustment was made to lower the indicated flow to match the calculated value of 97% rated core flow.

.Dotermination of Hated Driva Flowl In conjunction with the core flow avaluations, data was obtained to determino ['ated recirculation drive flow. This flow was calculated to be 33.34 x 10 lb/hr.

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Core thermal liydraulic Stability: j li 1 Data from the APRMs and the A and C level detectors of one LPRM in each ,

r octant was acquired in accordance with F-ST-5S. " Neutron Instrumentation Noise  ;

j Monitoring", at 35% and 75% rated core thermal power. This information will j serve as baseline data for the operating cycle it, and when Technical  ;

i Specifications require performance of F-ST-55. l

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other Tests

!

l Y Various surveillance tests were satisfactorily completed as required by {

i' RAP-7.1.17 and F-0P-65. i l

Because one of the low pressure turbines was replaced during the outage. l i extensive testing was performed on the turbine generator. This included a I

! torsional test, overspeed tests, and vibration tests. All were completed l satisfactorily, l

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