ML17264A601
| ML17264A601 | |
| Person / Time | |
|---|---|
| Site: | Ginna  |
| Issue date: | 09/09/1996 |
| From: | Mecredy R ROCHESTER GAS & ELECTRIC CORP. |
| To: | Vissing G NRC (Affiliation Not Assigned), NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NUDOCS 9609190060 | |
| Download: ML17264A601 (8) | |
Text
~
CATEGORY REGULATORY INFORMATION DISTRIBUTION SYSTEM (RIDS)
ACCESSION NBR:9609190060 DOC.DATE: 96/09/09 NOTARIZED: NO FACIL:50-244 Robert Emmet Ginna Nuclear Plant, Unit 1, Rochester G
AUTH.NAME AUTHOR AFFILIATION MECREDY,R.C.
Rochester Gas a Electric Corp.
RECZP.NAME RECIPIENT AFFILIATION VISSING,G.
DOCKET ¹ 05000244
SUBJECT:
Forwards startup rept, describing thermal/hydraulic s nuclear testing during Cycle 26 plant startup
& power escalation after installation of replacement SGs.
DISTRIBUTION CODE:
IE26D COPIES RECEIVED:LTR ENCL SIZE:
TITLE: Startup Report/Refueling Report (per Tech Specs)
NOTES:License Exp date in accordance with 10CFR2,2.109(9/19/72).
05000244 RECIPIENT ID CODE/NAME PD1-1 PD COPIES LTTR ENCL' 1
RECIPIENT ID CODE/NAME VISSING,G.
COPIES LTTR ENCL 1
1 INTERNAL~I-NTER
~RGN EXTERNAL: NOAC Ol 1
1 1
1 1
1 NRR/DSSA/SRXB/B NRC PDR 1
1 1
1 D
N NOTE TO ALL "RIDS" RECIPIENTS:
PLEASE HELP US TO REDUCE WASTE!
CONTACT THE DOCUMENT CONTROL DESK/
ROOM OWFN SD-5(EXT. 415-2083)
TO ELIMINATE YOUR NAME FROM DISTRIBUTION LISTS FOR DOCUMENTS YOU DON'T NEED!
TOTAL NUMBER OF COPIES REQUIRED:
LTTR 7
ENCL 7
p ROCHESTER GAS AND ELECTRIC CORPORATION
+E
$'tn'lc 0
89 EAST AVENUE, ROCHESTER N. Y. 14649-0001 ROBERT@ MECREDY Vice President Cinne Nucfeer Production TEt.EPSSQNE issEnccoE7ie 546 2700 September 9,
1996 U. S. Nuclear Regulatory Commission Document Control Desk Attn:
Guy Vissing Project Directorate I-1 Washington, D.C.
20555
Subject:
Cycle 26 Startup Report R.E.
Ginna Nuclear Power Plant Docket No. 50-244
Dear Mr. Vissing:
In accordance with Ginna Station administrative procedures, the attached startup report is hereby submitted.
It is being submitted to describe the thermal/hydraulic and nuclear testing during the cycle 26 plant startup and power escalation after the installation of replacement steam generators and the 18 month cycle transition core.
PJB/tjn Very truly yours, g
Robert C. Mecredy xc:
Mr. Guy Vissing (Mail Stop 14C7)
Project Directorate I-1 Washington, D AC.
20555 U. S. Nuclear Regulatory Commission Region I 475 Allendale Road King of Prussia, PA 19406 Ginna Senior Resident Inspector 9609190060 960909 PDR ADOCK 05000244 P
PDR 6
TESTING FOR STEAM GENERATOR REPLACEMENT During the 1996 Refueling Outage, both steam generators were replaced with BWI Replacement Steam Generators (RS/Gs) of similar design.
Associated with the steam generator replacement, Reactor Coolant System (RCS) T,, was reduced from a nominal 573.5 F to 561 F.
The valve trim for the main feedwater regulating valves was also modified affecting the valves'.
As part of startup, tests were conducted to verify the design of the RS/Gs and the cumulative affects of these changes.
Specifically:
RCS FLOW VERIFICATION Station Modification Procedure SM-10034-7.21 Testing was performed to verify the RCS elbow tap and Reactor Vessel Level Indication System (RVLIS) level transmitter differential pressure outputs are appropriate for operation with the replacement steam generators.
RCS flow transmitters are calibrated such that nominal conditions at 100% power produce an ihdication of 100% flow. Nominal conditions at 100% power change slightly as a result of steam generator replacement and reduced T,, operation.
Therefore, new calibration spans were calculated and verified by the test.
RVLIS level transmitters detect reactor vessel differential pressure (from lower head to upper head) during various operating conditions.
The differential pressure will change slightly during operation with the reactor coolant pumps running since the RCS hydraulic resistance is changed by the RS/Gs.
Therefore, it was necessary to verify that this change was insignificant to RVLIS.
Test results indicate that both RCS flow indication and RVLIS are operating acceptably and that outputs are appropriate for operation.
RCS FLOW TESTING Reactor Engineering Procedure RE-20.1 In accordance with the Improved Technical Specifications RCS flow was measured on June 18, 1996 using the precision calorimetric heat balance methodology.
The calculated flow was 89,957 GPM per RCS loop which is greater than the required flow of 88,650 GPM per loop.
The required flow is based on the assumed thermal design flow of 85,000 GPM/loop and the application of a 4% uncertainty to account for instrument accuracy and hot leg streaming effects.
ADVANCED DIGITALFEEDWATER CONTROL SYSTEM (ADFCS) TUNING Station Modification Procedure SM-10034-7.22 Steam generator replacement, reduced T,,, and a change to the C for the feed regulating valves all have the potential to affect the ability to control steam generator water level.
This testing was to verify and identify required adjustments to the Advanced Digital Feedwater Control System to assure stable level control of the replacement steam generators.
Inputs and outputs for the ADFCS were monitored during normal startup and power ascension as well as during open loop testing performed during power holds at 30% and 75% power.
Analyses had identified adjustments to signals gains and setpoints for ADFCS. The testing verified these.
The major changes were:
~
A change to a constant water level program for start-up.
Previously a variable water level program had been used at low power.
~
Adjustment to the gain for wide range level feed forward signal used at low power in place of steam-feed mismatch.
Analysis had predicted a
substantially lower slope with respect to power for this signal for the RS/Gs.
~
Adjustment to the valve position demand calculation due to the resized control valves.
The testing verified the acceptability of the ADFCS. Water level response was stable throughout the range of testing.
Wide range level response with respect to power was as predicted and adequately compensated for by ADFCS adjustments.
In addition to the planned testing, steam generator level and ADFCS response were monitored during an unplanned turbine trip from 25% power caused by loss of condenser vacuum.
ADFCS and the steam generators demonstrated stable level control and acceptable shrink/swell performance throughout the transient.
Specifically, steam generator water level shrunk only 9" before condenser steam dump valves opened.
ADFCS returned water level to normal within several minutes of the trip.
l t
STARTUP PHYSICS TESTING PROGRAM Cycle 26 Startup Physics Testing was conducted during the period from June 9, 1996 to June 17, 1996.
The stated dates span from initial criticality to attainment of 100% power.
The results of physics testing showed that all measured data was within the bounds of acceptance criteria tolerances and Improved Technical Specification requirements.
The program consisted of determining the following parameters:
1.
2.
3.
4.
5.
6.
7.
All rods out (ARO) critical boron concentration ARO zero power ISO Thermal Temperature Coefficient Moderator Temperature Coefficients Control Rod worths for Banks "D", "C" and "B" Boron Endpoint (B+ C+ D+ A IN)
Core symmetry and power distribution measurements Critical Boron Concentration
- Full Power Initial criticality was achieved on 6/9/96 at 0436 hours0.00505 days <br />0.121 hours <br />7.208995e-4 weeks <br />1.65898e-4 months <br />.
The Hot Zero Power (HZP) parameters listed in 1 through 5 above were then measured.
The initial flux symmetry mapping and subsequent power escalation to 45% occurred between 6/9/96 and 6/11/96.
Flux maps at 30% and 45% power were taken during this period.
Further testing was delayed by a reactor shutdown on 6/11/96 to repair a high pressure Turbine Governor valve leak.
The reactor was returned to operation on 6/12/96.
Additional flux maps were taken at 75% power on 6/14/96 and 100% power on 6/17/96 with the flux map at 75% providing data for a full incore/excore calibration.
The following data summarizes the results of the testing program.
Critical Measurements a ~
Nuclear Heat Determination - Nuclear heat was observed at 0.407x10 amps on Channel N-44 upper and 0.495x10 amps on Channel N-44 lower.
The test limits of 0.122x10'mps (upper) and 0.148x10'mps (lower) were established to stay below the effects of nuclear heating.
Reactivity Computer Checkout - The reactor was placed on varying periods to provide a comparison of indicated reactivity with that derived from period measurements.
The following table summarizes the computer checkout.
Control Bank B Steps Withdrawn Measured Reactor Period (sec)
Reactivity (PCM)
% Difference (M-P) x 100 p
20 121.9 162 Measured (M) 11.6 40 Predicted (P) 11.5 40.5 0 87%
-1.23%
c.
Isothermal Temperature Coefficient (ITC) - Heatup and cooldown rates of approximately 10'F/hr were established to determine the isothermal temperature coefficient.
The following table summarizes the results of the average of the (5) traces taken.
All valves are given terms of pcm/'F.
CORE Configuration ARO MTC Predicted (P) 4.38 Acceptance Criteria less than
+ 5.0 MTC Measured (M) 4.29 Difference (M-P)
-0.09 NOTE:
Moderator Temperature Coefficient
= ITC - (-1.70) where
-1.70 is the predicted fuel temperature coefficient (FTC).
70% power and end of life temperature coefficients were inferred from the difference between measured and predicted values with the following results:
Calculated Predicted Acce tance Criteria 70%
Power EOL
-. 61
-23.94
-.52
-23.85 Not positive Less negative than -42.9 d.
Control Rod Worth - Control Rod Worth was measured by adjusting rod position to compensate for dilution. The following table summarizes the integral rod worth data:
Control Bank D
Predicted (P)
PCM 595 1010 1355 Measured (M)
PCM 554.5 973.5 1308
+15%
+15%
15
-6.79%
3 62%
-3.47%
Difference Acceptance M-P Criteria P
x 100 A at 157 steps SUM 280. 5 3240. 5 252.5 3088.5
+15%
+ 10%
-9.98%
-4 69%
As shown above, each individual bank met the +15% acceptance criteria and the total worth met the +10% acceptance criteria.
e.
0 Boron End Point - The boron endpoints were determined for the All Rods Out (ARO) position and the B + C + D + A inserted configuration.
The following table summarizes the boron endpoint data.
Configuration ARO B+C+D inserted + A at 154 steps Predicted 1755 ppm 1384 ppm Measured 1751 ppm 1382 ppm Difference M-P
-4 ppm 2 ppm Acceptance
+1000 pcm
(= 100 ppm)
+1000 pcm
(= 100 ppm)
Flux Symmetry and Power Distribution - The flux symmetry map was taken at 30% power.
All locations were within the acceptance criteria listed below.
In addition all Nuclear Hot Channel factors were well within the bounds of the Improved Technical Specifications and the Core Operating Limits Report.
During the subsequent power escalation flux maps were generated at 45%,
75% and 100% power with all parameters within the acceptance criteria and Improved Technical Specification/Core Operating Limits Report requirements.
Incore/Excore calibration was successfully performed at 75% power.
Acceptance Criteria:
The acceptance criteria for a flux map is that the plant Technical Specification on peaking factors be met.
As an aid in evaluating the power distribution maps, the differences between measured and predicted assembly power levels are reviewed.
General criteria for these comparisons are that the difference for assemblies with relative power greater than 0.9 be less than 10% at or below 30% Rated Thermal Power (RTP) and that deviations from predicted be less than 0.10 relative power at power levels >40% RTP.
If these differences are exceeded an evaluation will be performed to ensure that the required peaking factors will be met.
None of these differences were encountered on any of the four flux maps.
In addition, the incore flux tilt was measured at each flux map.
In all cases, it was well below the administrative limit of 2% (1.0% maximum tilt was measured).
Critical Boron Concentration - Full Power:
This parameter was measured on 7/1/96 at a core exposure of 626 MWD/MTU. The calculated concentration was 1272.7 ppm versus a predicted value of 1305 ppm.
This difference of 32.3 ppm corresponds to a reactivity equivalent of 281 pcm, well within the acceptance criteria of +1000 pcm.