ML19241B289

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Cycle III Startup Test Rept
ML19241B289
Person / Time
Site: FitzPatrick Constellation icon.png
Issue date: 04/18/1979
From:
POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK
To:
Shared Package
ML19241B288 List:
References
NUDOCS 7907130439
Download: ML19241B289 (7)


Text

.

CYCLE 3 STARTUP TEST REPORT Cycle 3 operations commenced December 5, 1978 with the withdrawal of the first control rod. The startup test program was conducted from December 3, 1978 through February 23, 1979, in accordance with Reactcr Analyst Procedure (RAP) 7.1.17, titled Refuel Startup Program Revision 1.

hhen reference is made to values of core thermal power and core flow, these are nominal values rather than exact percentages.

COPTROL ROD DRIVE TESTS Control rod drive coupling checks were satisfactorily completed foi all rods on December 11, 1978. In addition, the insert and witndrawal times for all rods were checked and adjusted as required.

Prior to reaching 40'e rated core thermal power, contro.' rod scram time testing was conducted in accordance with RAP 7.3.10 titled Control Rod Scram Time Evaluation Revision 2. This test requires that each control rod be scranmed from position 48 (full out) with reactor pressure >950 psig. The results of these tests are tabulatad below.

Results:

Average of Control Rod Insertion Technical Specification 137 Rods (Percent) (Seconds) (Seconds) 5 0.375 0.28 20 0.900 0.74 50 2.000 1.53 90 3.500 2.62 Since all times for each amount of insertion for individual rods were loss than the maxir""a allowed by the technical specificatior.s ,

calcu;.n icas were not ud for 2 X 2 arrays.

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9 SillTTDOWN MARGIN DEMONSTRATION A shutdown margin (SDM) demonstration was performed December 5,1978 in accordan:e with Reactor Analyst Procedure 7.3.9 Revision 1. The required SDM was 0.38% A k + R + temperature defect. The fuel vendor has calculated that the most reactive point is at the e. ginning of the cycle, hence the value of R is 0.0. After calc lating the temperature defect, the required SDM was determined to be 0.47*6 A k. Sufficient control rods were withdra.m to demonstrate a SDM of 0.57*6 A k.

INSEQUENCE CRITICAL _

In order to climinate one control rod sequence exchange during Cycle 3, the startup was conducted in sequence B rather than sequence A.

It was necessary to withdraw five control rods (one of which was a low worth peripheral rod) more than estimated. T..ts difference can be attributed to criticality occurring at a reactor water temperature of 174 F rather than 6S F, and a critical eigenvalue of 1.0093 instead of 1.006.

REACTIVITY ANOMALY CIIECK .

A comparison of the expected and actual control rod density was perfccmed at 100% core thermal power (CTP) and 100% rated core flow.

The control rod isventory was 282 notches which was in close agreement with the predicted value of 315 notches. The !1" reactivity boundaries were 35 to 595 notches.

POWER DISTPIBUTION MEASUREMENTS Core power distribution was monitored throughout the startup using the process computer. Following significant changes in control rod pattern and po..er level, a complete r ,wer dist ribution measurement was performed using the Traversing in-core Probe (TIP) system. Core. parameters were maintained within technical specification limits.

  • IIP REPRODUCIBILITY ,[

Successive TIP plots taken on each machine indicate a maximum difference of approximately 3. % This occurred adjacent to a fuel bundle ; pacer. In general, the plo*s w e re w i t hi n ."', o f eac h o t h e r .

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CORE POWER SYMMETRY Core power symmetry was checked at 25%, 50%, 75%, and 100's CTP.

Mirror symmetric fuel assemblics checked at 100% CTP and 100% core flow using the process computer indicate a maximum difference of 4's for peri-pheral assemblies and less than it for interior, high power assemblies.

Total TIP uncertainty is =5%, well below the 8.7% assumed by the vendor in the statistical analysis performed for the licensing topical report (NEDE-24011-P-A) for the reload fuel application. The highest individual asymmetries were observed in locations where the rod pattern was not rotationally symmetric. If the pattern had been rotationally symmetric, the uncertainty would have been cven lower.

CORE LOADING A copy of the final core loading is attached as Figure 1. Irradiated fuel returned to .he core is designated EA or LJ5 or LJ6. There were 136 new feel asr <:mblies, designated LJB, (100 2.83 w/o U-235, 36 2.65 w/o U-235)

Ic. Md during the refueling out.'ge. The new fuel was of the 8 X 8 R dec.ign with an active fuel lengta of 150 inches and received 100 mil channels.

Figure 2 shows the approximate irradiated bundle average exposure fol' awing refueling. A blank indicates a new fuel assembly.

All new fuel assemblies contain burnable poison in the form of GdO3.

The concentration and location is proprietary to the fuel vendor.

Figure 3 shows the rod sequence control system (RSCS) designations for the A and B rod withdrawal sequences.

During the refueling cperation, each fuel nove was checked by an individual, other than the

  • operator performing the move, and verified independently by a third individual. Two lir.es of communication were established between the refuel bridge and the control room. Following leading, a core verification was conducted (and video-taped and examined later by quality assorance personnel) to verify the correct placement and orientation of each assembly.

ADDITIONAL TESTS [

1. Tests were performed in accordance with RAP 7.1.17 and F-ST-5C to ver ! ^ that there is approximately one decade omrlap between the sou range acnitor and inter.nediate range monitor (IE!) systems, and een the IRM and average power range monitor (APDI) systems.
2. Reactoi more isolation cooling and high pressure coolant injection flow rate tests were performed in accordance with F-ST-24C and F-ST-4B and demonstr tted compliance with the technical specifications.

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3. Both the rod worth minimizer and rod sequence control system functior:d properly during the startup test program.
4. Since some TIP tubing were replaced, the TIP alignment and logic limits were chec}-1 and adjusted as required.
5. New computer software was installed during the outage. Extensive testing was performed prior to and during the startup in accordance with the vendor's recommendations.
v. The APRM system was calibrated :o core thermal power and satisfactorily tracked power changes.
7. Heat balances were calculated manua'11y .ind used to verify the process computer calculations.
8. Process computer calculations of fuel assembly parameters, maximun average planar linear heat generation rate, minimum critical power ratio, and maximum fraction limiting power density compared satisfactorily with results obtained using off-line computer calculations.
9. RAP 7.3.13, " Pressure Regulator Tests," was performed satisfactorily when it was verified that an induced pressure transient of 10 psi was controlled by the electro-hydraulic control system pressure regulator. In addition, transfer from the primary to back-up pressure regulator was demonstrated following a simulated failure of the primary regulator.
10. RAP 7.3.7, " Core Clow Evaluation and Indication Calibration," was performed satisfactorily at 75% CTP. However, when performing the test at 100$ CTP on February 2, 1979, operations and instrumentation arl control personnel noted a step increase in the indicated core flow.

The absolute value of the step increase was approximately 1 1/2 to 30 of indicated flow'and this step increase resulted in a total indicated core flow of between 101.5% and 103% of rated flow. Operations personnel immediately reduced the recirculation pump speed until the indicated core flow was below 100$ of rated flow.

Investigation revealed that the square root computation module for one of 20 jet pumps had failed at some earlier time. The first indication of this failure, however, was the step change in the output of the square root module which occurred when technicians " disturbed" the module oy attaching test equipment to it This disturbance caused the module output to return to its proper value, that is, the module began perfonaing its square root func t io n .

The square root computation module was replaced with an identical unit from inventory and the Core Flow Evaluation test was completed without further dif ficul ty. This event was previous 1) reported to the Nuclear "egulatory Co=ission in LER-79-10.

The test was repeated later at 1000 CTP and 100$ flow with satisfactory results, f g _.,,..s" bbb

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