ML033640573
| ML033640573 | |
| Person / Time | |
|---|---|
| Site: | Seabrook  |
| Issue date: | 12/23/2003 |
| From: | Peschel J Florida Power & Light Energy Seabrook |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| NYN-03106 | |
| Download: ML033640573 (13) | |
Text
FPLEnergy Seabrook Station FPL Energy Seabrook Station P.O. Box 300 Seabrook, NH 03874 (603) 773-7000 DEC 23 mnm Docket No. 50-443 NYN-03106 United States Nuclear Regulatory Commission Attention: Document Control Desk Washington, DC 20555-0001 Seabrook Station Cycle 10 Startup Report In accordance with the requirements of Technical Specification 6.8.1.1, enclosed is the Cycle 10 Startup Report for Seabrook Station.
Should you require further information regarding this matter, please contact Mr. Paul V. Gurney, Reactor Engineering Manager, at (603) 773-7776.
Very truly yours, FPL Energy Seabrook, LLC gaes M. Peschel 44egulatory Programs Manager cc:
H. J. Miller, NRC Region I Administrator V. Nerses, NRC Project Manager, Project Directorate I-2 G.T. Dentel, NRC Senior Resident Inspector Z., (
C-an FPL Group company
ENCLOSURE TO NYN-03106
SEABROOK STATION UNIT NO. 1 STARTUP TEST REPORT CYCLE 10
INDEX 1.0 CHRONOLOGICAL
SUMMARY
2.0 CORE DESIGN
SUMMARY
3.0 LOW POWER PHYSICS TESTING
SUMMARY
(LPPT) 4.0 POWER ASCENSION TESTING
SUMMARY
(PAT) 5.0 TEST RESULTS TABLE 1 LPPT RESULTS TABLE 2 PAT FLUX MAP RESULTS TABLE 3 FULL POWER THERMAULHYDRAULIC DATA i
1.0 CHRONOLOGICAL
SUMMARY
Cycle 10 Fuel Load was completed October 19, 2003. Subsequent operation/testing milestones were completed as follows:
INITIAL CRITICALITY 10/26/03 LPPT COMPLETED 10/27/03 ON LINE 10/29/03 30% PAT COMPLETED 10/29/03 50% PAT COMPLETED 10/29/03 80% PAT COMPLETED 10/30/03 94% PAT COMPLETED 10/31/03 FULL POWER 10/31/03 1 of 9
2.0 CORE DESIGN
SUMMARY
The Cycle 10 core is designed to operate for 19,204 MWD/MTU with a coastdown to 20,512 MWD/MTU. Eighty-four (84) fresh fuel assemblies were loaded into the Cycle 10 core. Forty-four have an enrichment of 4.3 w/o and forty have an enrichment of 4.7 w/o. In addition, the top and bottom 6 inches have an enrichment of 2.6 w/o creating an axial annular blanket. By comparison, Cycle 9 utilized 84 fresh fuel assemblies, 72 with an enrichment of 3.6 w/o and the remaining 12 at 4.2 w/o, both with a similar 2.6 w/o axial annular blanket configuration.
Two mechanical designs are used in the Cycle 10 core:
- The fresh region 12 and reload region 11 fuel are of the Robust Fuel Assembly (RFA) design, which includes slightly thicker RCC guide and thimble tubes as well as a different mid-grid design.
The mid-grid design is expected to mitigate grid-to-rod fretting seen in V5H assemblies.
- The remaining reload region 10 is Vantage 5H.
All fuel utilizes ZIRLO for fuel clad, control rod guide tubes and instrument thimbles. The top and bottom grids are Inconel-718. The six low-pressure drop mid-zone and three intermediate flow mixer grids are ZIRLO with ZIRLO sleeves.
In addition, all fuel contains a Performance+ debris mitigation grid located at the bottom end plug of the fuel rod 2of9
3.0 LOW POWER PHYSICS TESTING
SUMMARY
Testing was performed in accordance with the following general sequence:
- 1. Initial Criticality: Criticality was achieved by withdrawing all shutdown and control banks and diluting to critical.
- 2. Zero Power Test Range Determination: This was determined after the point of adding heat had been demonstrated.
- 3. On-line Verification of the Reactivity Computer:
This was determined by examining the output of the Advanced Digital Reactivity Computer (ADRC) during rod withdrawal and the determination of the point of adding heat.
- 4. Boron Endpoint Measurement:
This was determined with all the Control and Shutdown banks withdrawn using the ADRC.
- 5. Rod Worth Measurement: Individual control bank and shutdown bank worths were measured using the Dynamic Rod Worth Measurement (DRWM) technique with the ADRC.
- 6. Isothermal Temperature Coefficient Measurement (ITC):
This was determined using the ADRC during a Reactor Coolant temperature change.
The Moderator Temperature Coefficient (MTC) was calculated from the ITC Data.
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4.0 POWER ASCENSION TESTING
SUMMARY
Testing was performed at specified power plateaus of 30%, 50%, 80%, 94% and 100%
Rated Thermal Power (RTP). Power changes were governed by operating procedures and fuel preconditioning guidelines.
In order to determine the core power distribution, flux mapping was performed at 30%,
50% and 100% RTP using the Fixed Incore Detector System. The resultant peaking factors were compared to Technical Specification limits to verify that the core was operating within its design limits.
Thermal-hydraulic parameters, nuclear parameters and related instrumentation were monitored throughout the Power Ascension. Data was compared to previous cycle power ascension data at each test plateau to identify calibration or system problems.
The major areas analyzed were:
- 1. Nuclear Instrumentation Indication:
Overlap data was obtained between the Intermediate Range and Power Range channels.
Secondary plant heat balance calculations were performed to verify the Nuclear Instrumentation indications.
- 2. RCS Delta-T Indication:
All RCS AT loops were initially scaled using Cycle 9 values. Data from 30%, 50%, 80%, and 94% RTP met prescribed acceptance criteria. Data was evaluated at 100% RTP and the AT loops were re-scaled.
- 3. Upper Plenum Anomaly: In early 1992, Westinghouse notified Seabrook Station that it may be susceptible to a phenomenon known as the Upper Plenum Anomaly (UPA). The UPA is primarily characterized by a periodic step change of 1IF to 20F in hot leg temperature and a corresponding change in steam flow. Cycle 10 data collected at 100% RTP identified the presence of UPA.
- 4. RCS Temperatures: Data was obtained for the Narrow Range Loop temperatures.
Evaluations for Delta-T (OF), TAVG Deviation Alarm Setpoint and TAVG / TREF Indication were performed.
Evaluations for deviations between redundant channels on individual steam generators were performed.
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4.0 POWER ASCENSION TESTING
SUMMARY
(Continued)
- 6. Steam Generator Pressures: Data was obtained for the steam generator pressures.
Evaluations for deviations between redundant channels on individual steam generators were performed.
- 7. Turbine Impulse Pressure (TREF): The initial scaling of impulse pressure was left the same as Cycle 9. The full power value was slightly higher than the current scaling value and is within the allowed deviation criteria.
- 8. Incore/Excore Calibration:
Scaling factors were calculated from flux map data using the single point calibration methodology. The nuclear instrumentation power range channels were re-scaled at 50% and 100% RTP.
- 9. RCS Flow: The RCS flow was measured at the 94% RTP plateau using elbow tap measurements to minimize the effects of observed hot leg streaming.
The power ascension test program is essentially unchanged from Cycle 9.
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5.0 RESULTS
- 1. Low Power Physics Testing: Acceptance criteria and review criteria were met. See Table 1 for results.
- 2. Flux Mapping: No problems were identified during the flux maps at 30%, 50%, and 100% RTP. See Table 2 for results.
- 3. Full Power Thermal/Hydraulic Evaluation: No problems were encountered with the instrumentation. Data collected at 100% RTP identified the presence of UPA.
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TABLE 1 LOW POWER PHYSICS RESULTS: CYCLE 10 ITEM MEASURED PREDICTED ERROR CRITERIA BORON END POINTS:
+ 1000 pcm HZP ALL RODS OUT 1939 ppm 1936 ppm 19.6 pcm 500 pcm*
ALL RODS OUT 1TC (pcmt'F)
-1.38
-2.09 0.71
+/-2*
ALL RODS OUT MTC (pcmF) 0.37
-0.34 N/A
<+ 3.12**
CONTROL BANK ROD WORTHS:
(pcm)
A 700.7 713.6
+12.9 B
747.5 686.3
-61.2 e
C 813.5 801.4
-12.1 D
485.8 471.5
- 14.3 SA 293.5 278.5
-15.0 100 pcm, or
- SB 931.4 872.9
-58.5 15%O
- SC 424.9 403.0
-21.9 SD 412.2 396.8
-15.4 SE 459.5 452.0
- 7.5 8%*
- TOTAL 5269.0 5076.0
- 193.0
>90%
NOTE:
- Review criteria, all others are acceptance criteria.
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TABLE 2 POWER ASCENSION FLUX MAP RESULTS: CYCLE 10 ITEM MAP I MAP 2 MAP 3 DATE OF MAP 10/29/03 10/29/03 10/31/03 POWER LEVEL (%)
29.18 48.54 99.97 CONTROL BANK D
157 195 227 POSITION (steps)
RCS BORON (ppm) 1723 1649 1409 FQ 2.2505 2.2783 1.9662 FAH 1.6236 1.5792 1.5071 INCORE TILT 1.0172 1.0135 1.0093 8of9
I TABLE 3 FULL POWER THERMAL-HYDRAULIC DATA: CYCLE 10 ITEM VALUE REACTOR COOLANT TAVG 586.68 OF REACTOR COOLANT DELTA-T:
LOOP 1 58.06 OF LOOP 2 59.45 OF LOOP 3 57.37 OF LOOP 4 58.02 OF REACTOR COOLANT FLOW:
(Based on elbow tap measurements)
TOTAL 406085 GPM AUCTIONEERED HIGH TAVG 587.40 0 F TREF 587.55 °F TURBINE IMPULSE PRESSURE 669.42 PSIG STEAM GENERATOR PRESSURES: