ML20138Q977

Rept of Plant Startup Following Sixth Refueling Outage, 840428-850713

ML20138Q977
Person / Time
Site:	Peach Bottom
Issue date:	10/28/1985
From:	Fleischmann R PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC
To:
Shared Package
ML20138Q962	List: ML20138Q960 ML20138Q977
References
NUDOCS 8511180359
Download: ML20138Q977 (11)

Text

r PHILADELPHIA ELECTRIC COMPANY

! Peach Bottom Atomic Power Station Unit No. 2 Docket Number 50-277 k-i l-i

REPORT OF PLANT START-UP FOLIOWING SIXTH REFUELING OUTAGE APRIL 28, 1984 TO JULY 13, 1985 i

SUBMITTED TO THE UNITED STATED NUCLEAR REGULATORY COMMISSION PURSUANT TO e, FACILITY OPERATING LICENSEE DPR-44 i Preparation Directed By:

R. S. Fleischmann, II Peach Bottom Atomic Power Station R.D. #1 Delta, PA 17314 hok bOK [7 P

m 1

't Pag 2 2 TABLE OF CONTENTS Page i INTRODUCTION AND

SUMMARY

3

SUMMARY

OF TESTS

1. Control Rod Functional and Subcriticality Check ----- 4

2. Demonstration of Core Shutdown Margin --------------- 4

3. Control. Rod Drive Testing --------------------------- 4

<4 . SRM Performance ------------------------------------- 5

5. IRM Performance ------------------------------------- 5

6. LPRM Calibration ------------------------------------ 5

7. APRM Calibration ------------------------------------ 6

8. Selected Process Temperatures ----------------------- 6 i

9. Sys tem Expa ns ion and S t rain ------------------------- 7

10. Core Power Distribution ---------------------------- 8

11. Core Performance ----------------------------------- 8

12. Feedwater System ----------------------------------- 8

13. Flow Control --------------------------------------- 8

14. Recirculation System ------------------------------- 9

15. Vibration and Strain ------------------------------- 9

16. Co r e Ve r i f ica t i on ---------------------------------- 10 i

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Pcga 3 INTRODUCTION The Peach Bottom Technical Specification 6.9.1 Routine Reports requires submittal of a startup report following an outage in which certain safety related events occur. Fuel of a different design was installed during the sixth refueling outage of Unit 2 and plant modification 1278, " Removal and Replacement of Recirculation System, RHR Shutdown, Cooling, RHR Head Spray, and RWCU Piping",

was performed. This report summarizies the plant startup and power escalation testing performed to assure that no conditions or system characteristic changes have been created by the sixth refueling outage of Unit 2 which in any way diminishes the safe operation of the plant.

SUMMARY

Startup testing was performed in accordance with the FSAR Section 13.5, "Startup and Power Test Program". Measured and calculated values of operating conditions and characteristics obtained during the startup test program are compared to design predictions and spe ci fications . All Level I criterion were met in accordance with FS AR Section 13. 5. 2. 2. All Level 2 criterion were either met, or discrepancies were investigated and determined to have no ef fect ' no safety, reliability, operability, and pressure integrity of the systems tested. Corrective actions were not required to obtain satisfactory plant operation.

The Peach Bottom Unit 2 was out of service from April 28, 1984 to July 13, 1985 to accommodate a recirculation system pipe replacement, maintenance, and refueling outage. During this 432 day outage 211 bundles of the 8X8R design, 79 P8X8R, and 2 Lead Test Assemblies (LTA) were replaced with 292 bundles of the BP8X8R fuel design. The unit returned to service on July 13, 1985, and reached full power on July 29, 1985. Startup tests were performed before and during the return to power and were completed during September.

The successfully implemented startup program insures that the sixth refueling outage of Unit 2 has resulted in no conditions or system characteristics that in any way diminishes the safe operation of the plant. The tests and data referenced in this report are on file at the Peach Bottom Atomic Power Station.

Page 4 STARTUP REPORT Peach Bottom Atomic Power Station Unit No. 2

1. Control Rod Functional and Subcriticality Check Control rod functional and subcriticality check was performed in accordance with FSAR Section 13.5.2.2.(3). Each control rod was withdrawn and inserted to verify rod coupling integrity (exception noted in Item #3 below), proper rod withdrawal and insertion, and subcriticality.' Level 1 criteria was met when core shutdown margin was demonstrated with a fully loaded core on July 6, 1985. Test data is documented in ST 10.8 completed July 6, 1985.

2. Demonstration of Core Shutdown Margin Core shutdown margin was demonstrated in accordance with FSAR Section 13.5.2.2.(4). An "In-Sequence" shutdown margin of -

1.512% delta K was obtained during the initial reactor 'startup in the A sequence. This satisfies the Level I criteria that the core must be subcritical by at least 0.40% delta K with any rod fully withdrawn. Test data is documented in ST 3.8.2 completed July 6, 1985.

The design predicted core Keff was compared to the actual value at initial startup on July 6, 1985. The predicted Kaf f was 0.9979 as compared to the actual Kef f of 1.003. The dif ference between predicted and actual values was -0.544, which meets the acceptance criteria of + 14. Test data is documented in ST 3.9.

The design calculated core reactivity at rated conditions was compared to the actual core reactivity on August 1, 1985. The predicted number of control rod notches inserted at rated conditions was 650. The actual number of notches inserted was 790. The actual number of. notches meets the range of + lt delta K of 950 to 360 control rod notches inserted. Test data is documented ' in ST 3.7-2.

3. Control Rod Drive Testing CRD testing was performed in accordance with FSAR Section 13.5.2.2.(5). At cold shutdown conditions, each CRD was tested for position indication, normal insert / withdrawal times, coupling, and friction. At an average reactor pressure of 1000 psig, position indication, coupling, and scram insertion times were tested. The testing performed at cold

PHILADELPHIA ELECTRIC COMPANY Peach Bottom Atomic Power Station Unit No. 2 Docket Number 50-277 REPORT OF PLANT START-UP FOLIDWING SIXTH REFUELING OUTAGE APRIL 28, 1984 TO JULY 13, 1985 4

SUBMITTED TO THE UNITED STATED NUCLEAR REGULATORY COMMISSION PURSUANT TO FACILITY OPERATING LICENSEE DPR-44 Preparation Directed By:

R. S. Fleischmann, II Peach Bottom Atomic Power Station

, R.D. #1 Delta, PA 17314

Page 5 shutdown conditions satisfied all Level 1 and 2 criteria.

With the exception of control rod at location 30-15, all 1000 psig testing satisfied Level 1 and 2 criteria.

Control rod 30-15 was scram time tested from a position which was known to be not fully withdrawn. Neither a '48' position indication, nor a red backlight was observed when attempting to fully withdraw the rod (it is important to note that a ' rod overtravel' check was performed subsequent to blade changeout with the results indicating that coupling does exist) . It was determined, however, that the control rod was at position '47' or beyond, but not at position '48'. A conservative adjustment factor was added to the scram time for control rod 30-15 to account for its not being testing from the fully withdrawn position. The scram time that resulted meets all Technical Specification requirements. The inability to fully withdraw control rod 30-15 is probably caused by a bent spud finger in the drive to blade coupling. At this time control rod 30-15 has been declared inoperable and is blocked in the full in position. The results of a safety evaluation performed by General Electric (July, 1985) shows that the 30-15 control blade is partially coupled. Plans for withdrawal of the 30-15 blade in accordance with the safety evaluation :

results are currently being developed. I

4. SRM Performance SRM instrumentation operability was checked during performance of startup procedure GP-2._ FSAR Section 13.5.2.2.(8) criteria of a minimum count rate of 3 counts /sec. was verified to be met for all SRM's. Data is documented in GP-2 dated July 6, 1985.

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5. IRM Performance IRM performance was tested in accordance with FSAR Section 13.5.2.2.(9). All IRM scram set-points met the Level I criteria of ST 3.2.3, "IRM Functional and Calibration Check",

during performance.

6. LPRM Calibration ,

LPRM calibrations were completed in accordance with FSAR Section 13.5.2.2.(10). Calibrations were performed at 334 and 100% rated thermal power per ST 3.4.1 on 7-15-85 and 9-6-85 respectively.

Page 6

7. APRM Calibration Numerous APRM calibrations were completed during startup in accordance with FSAR Section 13.5.2.2.(11). Test data is documented in ST 3.3.2's completed from July 6, 1985 to July 29, 1985 when full power was first achieved.

t 8. Selected Process Temperatures Selected process temperatures were monitored in accordance with FSAR Section 13.5.2.2.(15). The recirculation #1 speed limiters prevented bottom head temperature stratification during two recirculation pump and one recirculation pump operations at minimum speed, and CRD cooling water at maximum flow. Maximum CRD flow tested corresponds to about .33 gpm per drive unit in accordance with FSAR Section 3.4.5.3.1.  :

The "two pumps in operation test" was performed at approximately 27% reactor power. The temperature dif ference between steam dome and bottom head drain line remained constant at 36 degrees F. The maximum allowable temperature i

difference is 145 degrees F per FSAR criteria. The

' individual pump' test was performed with the 'A'

! recirculation pump operating at minimum speed, the "B" recirculation pump of f, and all control rods inserted. The temperature difference between steam done and bottom head drain line remained a constant 20 degrees F, which is less "

than the 145 degrees F FSAR limit. These test's data are documented in Modification Acceptance Test 1278 PB2-67.

Selected process temperatures were monitored during recirculation pump trips and starts. On the ' A' recirculation pump trip maximum steam dome - bottom head drain temperature dif ference was 49 degrees F. This is within the 145 degrees F i maximum required by Tech. Specs. The maximum loop suction temperature dif ference was 11 degrees F, which is within the FSAR limits of 50 degrees F.

On the 'B' recirculation pump trip, maximum steam dome -

bottom head drain temperature dif ference was 49 degrees F.

This is within the 145 degrees F maximum required by Tech.

Spe cs . The maximum loop suction temperature dif ference was 15 degrees F, which is within the FSAR limits of 50 degrees F.

Test data is documented in MAT 1278 PB2-68.

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9. System Expansion and Strain Recirculation, RNCU, and RHR system piping expansion was inspected during Reactor Pressure Vessel Hydrostatic (RPV-HYDRO) testing and Nuclear Heatup in accordance with FSAR Section 13.5.2.2.(16). Level 1 criteria was verified by visual inspections inside the drywell. Hangers and snubbers were not bottomed out or fully extended. Piping systems were observed to be free of obstructions that could constrain free pipe thermal movement, and piping could expand during system heatup within design limits. Therefore, FSAR Level I criteria for expansion was met.

The FSAR Level 2 criteria requires that displacements of instrumented points not vary from the calculated values by more than + 50 percent or + 0.5 inches, whichever is larger.

The FSAR Level 2 criteria Tor expansion was met during RPV-HYDRO and nuclear heatup testing. For acceptance testing of the Recirc., RNCU, and RHR System piping expansions, a more specific criteria recommended by General Electric was applied in addition to the FSAR Level 2 criteria. The GE Level 1 criteria defines a range of displacements for each instrumented point corresponding to the allowable stress l limits for the piping associated with that instrumented point.

During RPV-HYDRO, piping expansion from ambient to maximum hydro temperature (approximately 195 degrees F) was measured with lanyard potentiometers at various locations. The measured values of the displacements were within the Level I criteria specified by General Electric.

During the nuclear heatup, piping expansion for approximately 150 degrees F to rated reactor temperature (approximately 530 degrees F) was measured. For eleven instrumented points on the RHR system piping, the measured displacements were slightly less than the GE Level 1 criteria. However, the piping was not overstressed in these cases, because the design analysis to establish the GE criteria did not consider the thermal gradients which existed in the piping systems during testing. Consideration of these thermal gradients make all the measured displacements acceptable, as documented in General Electric FDDR HE-2-0694, Rev. 1. Therefore, there is no ef fect on safety, reliability, operability, and pressure inte Test data is documented on MAT'grity s 1278of the piping PB2-52 system.

and 64.

Strain data was recorded on recirculation system piping at steady state conditions of 100% rated power and 100% core flow. Measured micro-strain was less than 20% of the design criteria estab11shed by General Electric. Data is documented in MAT 1278 PB2-70.

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Page 8

10. Core Power Distribution Core power symmetry and TIP reproducibility were tested in accordance with FSAR Section 13.5.2.2.(17). Two full sets of TIP traces were 6btained at approximately rated power on

+

September 9, 1985. The total TIP uncertainty was 4.30% and 4.27% for the two sets. The TIP readings are therefore within the standard deviation used to establish safety limit criteria of 8.7%, per General Electric Document NEDE-240ll Table S.2-1.

The maximum deviation between symmetrical.'.y located pairs was 15.4% and 15.5% for the two data sets respectively, which satisfies the 25% acceptance criteria for core power symmetry.

Test data is documented in the RE-27 procedure completed October 3, 1985.

11. Core Performance core performance was evaluated along the 834 and 100% load lines in accordance with FSAR Section 13.5.2.2.(18). All thermal and hydraulic limits specified in the FSAR were met.

Jet pump baseline data was obtained and is documented in MATS 1278 PB2-59 and 60.

12. Feedwater System Feedwater controller stability testing was performed in accordance with FSAR Section 13.5.2.2.(22) to demonstrate l

acceptable reactor water level control. The response of each l

reactor feedpump to changes to the master level controller of plus and minus six inches of level change was 6bserved at 324, f

784, and 994 rated power. The overall feedwater control system tested in the three element mode displayed satisfactory system response. Test data is documented in RT 15.7, completed August 22, 1985.

Both "A" and "B" recirculation pumps were tripped individually at about 90% and 100% core thermal power. The feedwater control system satisfactorily controlled the water level, avoiding a turbine trip on high water level. Test data is documented in MAT 1278 PB2-63.

! 13. Flow Control

Plant response to changes in recirculation flow was tested according to FSAR Section 13.5.2.2.(28). At 824 and 93%

5 reactor core thermal power, positive and negative step changes i

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of about 5% to 10% pump speed were introduced into the recirculation M/A transfer station. Each recirculation pump was tested individually. All decay ratios exhibited for oscillatory variables were less than 0.25, and met both FSAR l Level 1 and Level 2 criteria. Test data is documented in MATS 1278-PB2-55 and 56.

14. Recirculation System Jet' pump flow instrumentation was tested according to FSAR Section 13.5.2.2.(29). At 854 core thermal power, core flow instrumentation was calibrated according to the results of General Electric JRPM01 computer program. Core flow was ,

checked again at 100% core thermal power, and no further.

instrumentation calibration was required. Test data is documented in MAT's 1278 PB2-61 and 62.

The recirculation system was' verified to not exhibit cavitation in the operable region of the power-flow map.

Power was decreased in approximately 10% intervals by-adjusting the rod pattern until the recirculation flow liq $ter annunciator was activated. The power was reduced from 64.6%

CTP to 23.4% CTP, at which point the 20% feedwater ' interlock was activated. At each power reduction interval, jet pump

-single tap dP's, reactor recirculation pump dP's, core plate dP's and reactor recirculation drive flows were' monitored for any significant fluctuations which might indicate cavitation.

Also, vibration data obtained from lanyard potentiometers located on reactor recirculation, RWCU, and RHR system piping was compared to and successfully met both the Level 1 and Level 2 criteria specified by General Electric. Data is documented in MAT 1278 PB2-66. I Throughout all recirculation system testing, transient thermal  !

limits were met at all times. Flow instrumentation calibration resulted in the reactor jet pump total flow l recorder providing correct flow indication. l

15. Vibration and Strain In accordance with FSAR Section 13.5.2.2.(32), vibration and '

strain data were obtained for the Reactor Recirculation (RR),

Reactor Water Cleanup (RWCU), and Residual Heat Removal (RHR)

System piping. Preoperational vibration and strain data were I obtained during RR system ' A' and 'B' pump starts and trips, and steady state operation at 30, 35, 40 and 45% pump speeds.  !

Vibration data was obtained during RHR shutdown cooling at i various conditions of operation. Steady-state vibration and i l

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o Page 10 strain data were obtained at (a) minimum core flow, (b) approximately 50% of rated core flow, (c) approximately 75% of rated core flow, and (d) approximately 100% of rated core flow; all at operating temperatures.

All measured strain and vibration parameters were within design peak to peak criteria. The vibration which had the most limiting peak to peak design limit of 0.04 inches was the recirculation system discharge vibration. Vibrations up to 0.02 inches (50% of the design limit) were observed. All other vibrations and strains were less than 50% of the design criteria set by General Electric Company." Test data is documented in MATS 1278 PB2-51, 53, and 54.

16. Core Verification -

Post alteration core verification (bundle location, seating, orientation) began in accordance with Surveillance Test ST 12.10, " Core Post-Alteration Verification". The seating verification revealed five assemblies to be improperly seated, these were 25-60; 47-38: 47-40; 45-38: 45-40. The bundle at location 25-60 was lifted and reseated and was visually i verified to be seated satisfactorily. Bundles 47-38: 47-40; 45-38; 45-40 are contained in cell 46-39. "As Found" visuals show the' support piece raised approximately 3/4". The cell was disassembled, the support resented and the cell reloaded.

The cell now has an elevation of 1/4" above the surrounding bundles. This 1/4" higher elevation of the four fuel bundles compared to the rest of the core has been evaluated, and determined that the variation should have no measurable ef fect on reactor performance. This conclusion was based on the following

- The fuel assemblies are visually verified to be seated properly

- There is no inlet orifice blockage visible.

- Core bypass flowrate due to these circumstances is negligible.

- The

- A 1/ 4" guide tube is properly misalignment aligned is within with thevariation the expected drive housing.

of fuel bundle height.

Friction Pictures and Scram Timing were done on this cell after the resenting and it was found the cell functioned properly. All fuel bundles were correctly located, seated, and oriented prior to cycle 7 startup and operation.