JAFP-90-0697, Cycle 10 Startup Testing Rept
| ML20065D612 | |
| Person / Time | |
|---|---|
| Site: | FitzPatrick  |
| Issue date: | 09/17/1990 |
| From: | Fernandez W POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| JAFP-90-0697, JAFP-90-697, NUDOCS 9009260005 | |
| Download: ML20065D612 (9) | |
Text
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- -. Nuoiser Power Mont u p.a e u !
. Lycoming, New York 13093 i
.' 315 542-36W ;
William Femander ll !
Resident Manager September 17, 1990 l JAFT-90-0697 l l
United States Nuclear Regulatory Commission
- Document Control Desk l Washington, D.C. 20555
{
Subject:
JAMES A. FITZPATRICK NUCLEAR POWER PLANT DOCKET No. 50-333 CYCLE 10 START-UP TESTING REPORT Gentlemen:
Enclosed you will find the Cycle 10 start-up Testing Report for the t James 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. :t i We trust you will find this information satisfactory. However, .
l should you desire more information, please contact Mr. Desid Burch '
at'(315) 349-6311.
Very.Truly Yours, ILLIkM, FERN DEZ WFf T:dmh
. CC: U.S. Nuclear Regulatory Commission I Region 1 NRC Resident Inspector .
WPO for RMS Headquarters Distribution G. Rorke '
D. Burch Document Control Center j I
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James A.'FitzPatrick Cycle 10-Start-up Testing Report
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- i. Cycle.10 on-line operations commenced June 24, 1990, ending a i 85 day refueling outage. The startup testing program commenced !
June 19,.1990 and was completed August 7, 1990. The test l program.was conducted in accordance with Reactor Analyst )
Procedure 7.3.30 " Cycle Startup Reactor Physics Test Program".
Listed below is a summary of the tests performed in accordance 1 with RAP-7.3.30. i j
- 1. Core Loadina and Verification:
A fuel shuffle was performed during this outage in which four hundred and-twelve bundles were relocated, one hundred forty-eight bundles were discharged, and one. .
hundred forty-eight new bundles were loaded. The cycle 10 fuel bundle breakdown is as follows:
- 1. Gu? fuel, 2.99% enrichment, 40 bundles, Reload 6 I
- l. 2. Westinghouse LTA's, 2.99% enrichment, 4 bundles, j Reload 7 l
- 3. GE-8 fuel, 3.19%_ enrichment, 184 bundles, Reload 7
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1 ll 4. GE-8 i el, 3.39% enrichment, 32 bundles, Reload 8 i f 5.. GE-8 fuel, 3.36% enrichment, 152 bundles, Reload 8 l 6. GE-10 fuel, 3.22% enrichment, 56 bundles, Reload 9 y l
- 7. GE-10 fuel, 3.24% enrichment, 88 bundles, Reload 9 ;
- 8. GE-11 fuel, 3.02% enrichment, 4 bundles, Reload 9 ;
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1 The final core loading was verified in accordance with i RAP-7.2.4 " Reactor Fuel Verification" using an undertater :
L television camera and video recorder. The videotape was l
independently checked by QA personnel and documented by QA l Surveillance Report # 1406.
- 2. Control Rod Drive Tests:
1 L , Eighteen control blades were replaced during the outage with General Electric Duralife 230 control blades. A total.of sixty three of the one hundred.and thirty seven original equipment control-blades have'now been-replaced,
- t Prior to start-up, Surveillance test-ST-20K, " Control Rod l Exercise / Venting" was performed on all 137 control rods to demonstrate that each rod is coupled to its drive mechanism, and to check that~each control rod. drive :
satisfies a~ travel timing' test. ;
Control rod scram time;testingtwas' performed.on all 137 control rod drives prior to reaching-40%' rated core thermal power. i
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'O i The results ware as follows:
Notch Tech Spec Limit Average for 137 rods 46 .338 sec .295 sec 38 .923 sec .709 sec 24 1.992 sec 1.452 sec 04 3.554 sec 2.566 sec 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. l
- 3. Shutdown Marain Test:
Initial Criticality for cycle 10 was achieved on June l
'19,1990. Shutdown margin was demonstrated using the in-sequence critical method which showed the core to have a shutdown margin of 2.703% delta k/k which exceeds the- '
Technical Specification requirement of 0.38% delta k plus R where R = .41%-delta k for a total of .79% delta k.
- 4. Control' Rod Secuence:
Sequence 10A2 was loaded into the RWM program in accordance with the requirements of the Reduced Notch Worth Procedure.- Prior to stat;wp, a surveillance test of the RWM was performed to demonstrate system operability;-
- 5. SRM Performance Check! .
SRM Functional Testing was 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.
- 6. Reactivity Anomalv Check!' .
A comparison'between the predicted and actual' control rod
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i density was performed at 100% rated core thermal _ power;and -
99.1%. rated core. flow. The actual. rod inventory was 288 '
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notches; inserted which is 65 notches'less than the a predicted notch inventory of 353 notches. A reactivity
' anomaly" of 11% delta k/k is equivalent to i300. notches.
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,7 . Power Distribution Measurements:
Core power distribution was monitored throughout the power ascension using the Traversing Incore Probe System (TIP) and the Local Power Range Monitors (LPRMS). LPRM calibrations were performed at 25%, 50%, 75%, and 100% of rated core thermal power. ruel thermal limits were maintained within Technical Specification lisaits.
- 8. Core Power Symmetry Calculations:
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 Average % Difference 50% 8.96% 1.84%
75%. 5.73% 2.60%
100% 8.49% 2.45%
- 9. Manual Heat Balancet 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 Calculation Computer calc 25%
573.7 MW -623.4' MW 50% 1213.1 MW 1228.2 MW 75% 1723.4 MW 1714.9 MW:
During. scram time-testing, when control rod.insertivns and withdrawals 4were performed, LTs i response testing was conducted on all operable detevcors.' This test verified that each. operable detector is connected to the appropriate flux amplifier. Six LPRM assemblies were replaced during the outage. In addition, a TIP response test was conducted to verify that each TIP tube is connected to the appropriate LPRM assembly.
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- 11. Plant Comouter Checkout The computer databank for Cycle 10 operations was= l installed and verified to be correct per Reactor Analyst !
Procedure 7.3.17 " Core' Monitoring Software and Database i
,' Changes". The plant computer calculations were compared 4 with the:offline core performance programs at 25%, 50%, j 75%, and 100% rated thermal power. The results showed. ,
close agreement in location and magnitude of all thermal I limits.- l 12.: Core Flow Evaluation 1 Core Flow Indication Calibrations were. performed-at 75%
and 100% rated power per Reactor Analyst Procedure 7.3.7. l In both evaluations the indicated flow matched the !
calculated flow within less than .5%, which is well within l the 12.5% tolerance assumed for the statistical I I
L uncertainty in the Licensing Topical Report.
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- 13. Determination of Rated Drive Fl'ow ,
l A rated drive flow, calculation was' performed ats100% -!
power, and-the results'show that a' drive flow of 32.56 x i 10 6 lb/hr produces the rated core flow of 77.0 x 10 6 4
lb/hr. The original design value for rated
- drive flow was 34.2 x 10 6 lb/hr.' y . .
[ 14. . TIP 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 LPRM instrument tube. These limits wara' j
' checked per Reactor Analyst Procedure 7.3.14 "TIP SYSTEM"' ;
-at full power by checking the location of-spacer dips on 1
- the-flux traces, t- 15. TIP Readina Uncertainty.
I The standard deviction between symmetricallyflocated TIP )
strings was determined from BASE distributions.obtained at ,
25, 50, 75 and 100 percent power. The resulting TIP ;
reading uncertainties were calculated to be 3.2% at-25%
power, 2.0%'at 50% power, 3.1% at 75% power, and'2.6% at 100% power. These values are well within the 7.1% TIP ,
reading uncertainty assumed in the Licensing Topical 't Report.
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- 16. CRIA Thermal Hydraulic Stability Data was acquired frca the APRMs and LPRM-detectors at 25 and 75 percent power in accordance with ST-SS, " Neutron Instrumentation Noise Monitoring". This information will serve as baseline data for the operating cycle when Technical Specifications require performance of ST-5S.
II. Other Start-up tests performed to satisfy Technical Specification Requirements included the following:
- 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 Heatuo - Performed in accordance with ST-26J, "Heatup and Cooldown Temperature Checks". The reactor vessel heatup was performed in accordance with the L requirements of ST-26J which requires reactor coolant ,
L system pressure and temperature be at or to the right of !
,. curve C shown in Figure 3.6-1, and the maximum temperature L change during any one hour equal to or less than 100 degrees fahrenheit, i l
- 3. IRM Performance - Performed ST-5C, "IRM-APRM Instrument Range Overlap Check" which demonstrated thatieach APRM
,t channel was on scale before any IRM exceeded the high IRM L rod block setpoint.
- 4. Safety Relief Vglygg - Performed ST-22B, " Manual Safety Relief) Valve OperationLand Valve l Monitoring System Punctional Test". The acceptance criteria of ST-22B were
, satisfied which demonstrated (1) that each safety' relief valve opens and closes fully through' operation of control switches lon 09-1 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 MSIV's close within the Technical Specification and IST stroke time of 3 to 5 seconds.
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- 6. ROIC System . Performed ST-24A, "RCIC Pump and Valve j Operability Test" which verified RCIC pump, motor, and valve operability. A Simulated Automatic Actuation test j was performed in accordance with F-ST-24E which ,
demonstrated the ability of the RCIC system to deliver a 1 flow rate of 400 gpm. ;
i
- 7. HPCI System - Performed ST-4B, "HPCI Pump and MOV Operability Tests".to verify operability of the.HPCI
. turbine and pump assembly, and associated motor operated' valves. A simulated Automatic Actuation Test was performed in accc.cance with F-ST-4A which demonstrated
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the ability of the HPCI system to deliver a flow rate of 4250 gpm. In addition, Preoperational Test POT-23E "HPCI I INJECTION TO REACTOR VESSEL" was conducted which verified l the HPCI system not all applicable system design requirements. These included (1) demonstrating the HPCI pump discharge flow rate equals 4250 gpm in less than 30 seconds from automatic initiation at rated reactor pressure, (2) the HPCI turbine does not trip or isolate I during the test, and (3) the decay ratio of any HPCI I system ~related variable is not greater than 0.25. 'j III. Some of the start-up tests performed during the initialfcycle startup were not performed due to the reasons specified.below. ;
(A) Performance of the test challenges the reactor protection 4 and safety systems of the plant and/or places the plant:in a degraded condition.
t
- 1. Turbine Trio and Generator Load Reiection Test: the purpose of this test is to demonstrate the response of the reactor and its control systems to protective trips in the turbine and generator. The turbine stop .
valves are closed, and the main generator breaker tripped in such a way that a load imbalance trip occurs.
- 2. Simultaneous closure of_All MSIV'S: The purpose'of this test'is to (1) functionally check the'MSIV's for t proper operation, (2) _to determine the resultant reactor transient behavior, (3) determine-valve closure tian,.and'(4) determine the maximum power at which a s. gle valve closure can occur without causing a teactor scram. , ;
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- 3. Loss of Turbine Generator and Offsite PoWarl The purpose of this test is to determine reactor ;
transient performance during the loss _of the main generator and all off site power. 1
- 4. Shutdown From Outside the Control Ro2R1 This test demonstrated that using controls _ located outaide the I control room the reactor can be scrammed and MSIV's I closed, and that operators can control vessel water- l level and pressure such.that a reactor cooldown is !
initiated, j
- 5. Recirculation Pumn Trio Test: The purpose of this 'l test is to evaluate the recirculation flow and l reactor power level transients following a single and '{
then dual pump trip.
(B) The test measures parameters which needed to be.
established or. verified during the initial plant-startup before the plant had any operating history. l1 i
- 1. System ExDansion Testi The purpose of this test is l' to verify that the reactor drywell piping system is free and unrestrained with regard to thermal expansion.
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- 2. Turbine BvDass Valve Measurement Testi. The purpose of this test is to demonstrate the ability.of the R pressure regulator to minimize the reactor pressure disturbance during an abrupt change in steam flow by q
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tripping open and closing a turbine bypass valve. 1 l
- 3. Selected Process TemDeratures: The purpose of this test was to establish the minimum recirculation pump speed that ensures adequate mixing in the lower vessel plenum, and to assure that.the measured bottom o head drain temperature corresponds to the bottom head coolant temperature during normal operation. ;
a 4.- Vibration Measurements: This. test performed i vibration measurements on'various reactor components' to demonstrate the mechanical integrity of the' system-L to ' flow induced vibrations'.
- 5. Radiatien Measurements: This test determined pre-
, operational background radiation levels in the plant environs to. assure protection-of plant personnel' -
c during plant operation, e
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. ?ty . . .' i Recirculation MG Set 1Soeed' Control:
4' :6. lThe purpose of 1
/i ~ this test-was,to. determine the-speed control- _ i H : characteristics of the:MG sets'obtain acceptable.- , j w: ,
speedicontrol-system performanc2,-and' determine-g%; maximum allowable pump speed. 4 s
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- 7. Flux Resoonse to-Control Rods: L The purpose of.this
. test is to demonstrate the stability of the. core 3
'* local power / reactivity feedback mechanism to small t i
, perturbations caused by rod movement.
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- 8. -RRR' Steam'Condensina Mode Demonstration: This test' 2i demonstrates.the RHR systet is capable of_ removing. ;
1, decay heat from the, reactor by operating:in the Steam ?
Condensing Mode. 1 i >
0, it 9.- Feedwater Systgmi This test-(1) adjusted _the.
feedwater. control system for-acceptable 1 reactor water i.'
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1evel1 control, (2) demonstrated r. table reactor 1 f'
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- rasponse toLsubcooling changes, (3) demonstrated
' capability-of the automatic rerdrouletion. flow i runback feature to prevent low wh'or: level scram' i
( following the trip of one feed pr x ,.and<(4) wP demonstrated reactor response to'lo.'
- Specifications.
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g" , 13. Pressuref R$aulator Test: theimain" purpose.of this'.
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test was to determine ~the optimum setting for.the h pressure control loop by analysis _of transients d
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induced in the reactor. pressure control _ system by g means of'the' pressure regulators.
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