ML19274D330

Rept of Plant Start-up Following Third Refueling Outage Sept-Oct 1978. Assures That Unit 2 Refueling Outage Caused No Conditions or Sys Characteristics to Diminish Safe Operation

ML19274D330
Person / Time
Site:	Peach Bottom
Issue date:	10/31/1978
From:	Ulrich W PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC
To:
Shared Package
ML19274D328	List: ML19274D327 ML19274D330
References
NUDOCS 7901230187
Download: ML19274D330 (12)

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I h PHILADELPHIA ELECTRIC COMPANY _

PHILADELPHIA I ~

PEACH E07f0M ATOMIC POWER STATION UNIT NO. 2 DOCKET NUMBER $0-277 I

REPORT OF PLANT START-UP FOLLOWING THIRD REFUELING OUTAGE SEPTEMBER-0CTOBER 1978 I

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, SexmEs .

THE UNITED STATES NUCLEAR REGULATORY COMIISSION PURSUANT TO FACILITY OPERATING LICENSE NO. DPR-Id4 I -

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w PHILADELPHIA ELECTRIC COMPANY Peach Ecttom Atomic Power Statior.

I Unit No. 2 Cocket Number 50-277 I

I REPCRI OF PLANT START-UP FOLLOWING THIRD REFUELING OUTAGE SEPTEMBER - OCTOBER 1978 SUBMITTED TO THE UNITED STATES NUCLEAR REGULATORY CCM.liSSION I PURSUANT TO FACILITY OPERATING LICENSE 2 DPR-44 Preparation Directed cy:

W. T. Ullrich , Superintendent I Peach Bottom Atcalc Power Station

TABLE OF CONTINTS PaJe INTRODUCTION .................................... 1 SU:01ARY OF TESTS

1. Verification of Shutdown Margin .............. 2

2. Power Distribution and Reactor Core Operating Limits ........................ 2

3. LPRM Calibration ............................. 3 4 htiactivity Anomalies ......................... 3

5. Core verification ............................ 3

6. Axial Power Distribution Comparison .......... 3

7. Cold critical Rod Pattern Prediction . . . . . . . . . 3

8. Core Power Symmetry and TIP Reproducibility Test ......................... 4

9. Centrol Rod Drive Scram Timing . . . . . . . . . . . . . . 4

10. Special Radiation Survey .................... 4

11. RhR 2A and 2D Flow Control Valve ............ 4

12. Other Startup Surveillance Tests . . . . . . . . . . . . 5

13. Control Rod Drive Pump Suction Reroute ...... 6 14 Drywell Instrument Nitrogen Isolation valve Power Supply Separation . . . . . . . . . . . . . . . . 6

15. HPCI control Oil Drain System Modificacion. . .. 6

16. Rod Sequence Control System Bypass Pressure Iransducer/ Trip Unit ....... ........ 7

17. Relay Operacility of Various CR-120A Control Relays ............................... 7

APPENDICES Page A. Feedwater Nozzles and Control Rod Crive Return Nozzle Inspections ............... .4- 1 I B. Radiation Dose Rates and Man Rem Expended During Inspection of tne Feedwater Nozzles and Control Rod Drive Return Ibzzle . . ... 3-1 I

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I INTROCUCTION The Peach Bottom Unit 2 was out of service from Septemoer 9, I to Octcher 17, 1978 to accommodate a maintenance /re:ueling outa ge . During this 39 day outage, 260 fuel bund.es of ene 1978 original (7x7) design were replaced with the new (6xd R) ruel de sign . The new fuel was manufactured by the General Flectric company and was designed to provide additional operational flexibility and fuel economy. The unit was returned to service on October 17, 1978 and reached f ull power on Octocer 26, 1978.

The Peach Bottom Technical Specification 6.9.1 Routine hecorts I requires submittal of a startup report following cany outage in which certain safety related events may occur. Installation of fuel of a different design is one of these events. This report, prepared to meet the Technical Specification requirement, I describes the startup program implemented to provide assurance that the safe operation of the plant was not diminished by the activities of the third ref ueling outage.

The startup tests identified in the F.S. A.R. were addressed and those which involve areas which were af fected by outage I activities were included and are summarized herein. Additional special tests connected with specific outage activities were also included in ene program and are discussec in, this report. Tne successful isplementation of this startup program insures that the Unit 2 rerueling outage has resulted in no conditiens or system characteristics that in any way diminisn c:.e safe operation of the Unit.

The tests and data referenced in this report are on file at tne Peach F:ottom Atomic Power Station.

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1. Verification of Shutdown Marain The Core Shutdown Margin was determined during the initial reactor criticality in the A-1 sequence.

I The actual Shutdown Fergin was 1.80% delta A/.<, as ccmparec with Technical Specification minimum value of 0.38 % delta k/x.

.? . Power Distribution and Reactor Core Ooeratina Limits Powr distribution and core performance paramaters were calculated throughout reactor startup utilizing One process computer.

All core performance parameters were within the required limi ts . The significant values of core performance are sumarized below for near rated conditions.

Da te : Novemcer 2, 1978 Iime 0809 Hours CMWT = Reactor Thermal Power = 3281 Reactor Core Elow =

100.0 x 106 L3/HR.

MAPRAT = Ratio of Actual - MAPLHGRE Technical Specification MAPLHda Limit 7 x 7 fuel less han 0.333 8 x 8 Standard Fuel & LIA2 less tnan 0.833 8 x 8 Retrofit = 0.850 MCPRRAT = Ratio of Technical Specification MCPR Limit Actual MCPR3 7 x 7 Fuel = 0.847 I 8 x 8 Standard Fuel & .IA2 8 x 8 Retrofit = 0. 93 2

= 0.932 MFLPD = Ratio of Actual Peak LHGR* (KW/ f t)

Technical Specificat1Cns Peak LEGE (KW/ft) 7 x 7 Fuel less than 0.780 I 8 x 8 Standard Fuel & LTA = 0.856 8 x 8 Retrorit Fuel = 0.847 I Notes: 1.

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MAPLHGR - Maximum Average Planar Linear Heat Generation Rate LTA = Lead Test Assembly I 3.

4 MCPR = Minimum Critical Power Ratio LHGR - Linear Heat Generation Rate I

3. LPMt Calibration An LPRM calibration was performed at a power lavel or approximately 30% of rated in accordance wit 2. Surveillance Test ST 3.4.1, "LPRM Gain Calibration," to provide accurate LPRM input for core calculations during the rise to rated pow r. Another calibration was performed at a power level of I 94% of rated to provide accurate indication at high power level.

4 Reactivity Anomalies Surveillance Test ST 3.7-2 " Reactivity Anomalies," was successfully performed. The predicted number of control ro nctrhes inserted at rated conditions was 875 with a + 1%

delta keff range of 670 to 1070 control rod notenes inserted.

The actual number of control rod notches inserted was 1012, which satisfies the + 1% delta kef f criteria.

5. Core verification The Unit 2 BCC4 post-alteraticn core verification was I performed in accordance with Fuel Handling Procedure FH-12

" Core Post-Alteration Verification." Fuel bundle PH 222 was misoriented by 180 degrees and 2 bundles were improperly seated. PH 222 was rotated to its proper orientation anc :na 2 bundles were reseated. The corrective f uel sofements were verified again.

6. Axial Power Cistribution Cemcarison Reactor Engineering Procedure RI-26-2, "PS2 - BCC4 Axial Power Distribution Comparison" was performed at the BCCu in the A-1 control rod sequence. The average deviation f rom the I actual axial power distribution is 0.1.

were satisfactory.

The test results

7. Cold critical Rod Pattern Prediction I The Cold Critical Rod Pattern Prediction Comparison (RE-25-2) was successfully performed for Unit 2. The predictea core ke ff was 1.00088 and the actual core keff was 1.008 for a difference of -0. 00712. The dif f erence is -0.712% and I satisfies the + 1% test criteria.

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3. Core Power Symmetry and TIP Recroducicility Test Core power symmetry and TIP reproducibility t ast cata was analyzed for sequence "A" conditions at 91% cower. The totai I TIP uncertainty was 6.4d%; therefore, tne test crite ria that the total TIP uncertainty not exceed 12'. is satisfied.

9. Cent rol Rod Crive Scram Timing All 185 control rods were scram timed following drive maintenance, fuel loading and replacement of reactor vessels I inte rnals. The testing was accomplished in accordance witn Surveillance Test ST 10.7, " Scram Insertion Times, Rod coupling Integrity and RPIS for Full in and Full out reses,"

I at a nominal reactor pressure of 1000 psig. Ihe reccrced times were within the maximum and average time requirements of the Technical Specifications.

10. Scecial Radiation Survey A special radiation survey was conducted enroughout the plant I at 9 0% reactor power. A measurement for neutron exposura at the high pressure turbine indicated 2.5 mR/hr neutrons, confirming earlier initial startup readings. Ine gamma dose rate at the high pressure turbine was lower than expected, possirly due to better reactor water chemistry. All other results of this survey verifies tnat all shleiding removed I during the outage was properly reinstalled and tnat no major changes in the radiation levels occurred due to the outage.

11. RHR 2A and 2D Flow Control Valves During the refueling ou. age, flow control valves were installed on the discharge of the A and D RHR pumps. These I valves serve two purposes, namely a) providing throttling capability in shutdown cooling mode of operation and c) providing additional piping system resistance to prevent pump run out wnich could occur with one RHR pump injecting into a brcken recirculation loop. ,

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The flow control valves on une 2A and 20 RER loop (C V-2677 AS D) were satisfactorily tested. The rull open mechanical stemtravel stop was set to control the maximum RHR flow between 11,500 and 12,200 GPM, in One torus-to-torus flowpath, with all other valves full open.

These valve settings ensure that the following design conditions are met:

A. Adequate flow in the open position to meet LPCI I B.

requirements.

Flow limited in the open position to prevent pump runout.

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12. Other Startuc Surveillance Tests A. Prior to Startup

1. ST 12.6-1 Primary Containment Dryw=1A to Tcrus I 2.

Bypass Area Test ST 1.4 and ST 1.5 Core Spray Logic System Functional I 3.

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Test ST 1.6 and ST 1.7 RER Logic System Functional Test ST 1.8 and ST 1. 9 Automatic Depressuriza tion System Logic System Functional Test I 5. ST 1.10 Rod Withdrawal Block Logic System Functional Test

6. ST 1.13 CAD System Functional rest I ,

7. ST 11.6-2 Diesel Generator Simulated Auto-Actuation Test

8. ST 13. 8 Excess Flow Cneck Valve Operacility Test

9. SI 1.3 PCIS Logic System Functional Test B. Beginning of Startup (Rod Withdrawal to Critical)

1. RE-12 Shutdown Margin Test

2. ST 10.8 Control Rod Withcrawal Tests

3. RE-25-2 Cold critical Rod Pattern Prediction Comparison 4 ST 12.8 Recirculation Baseline Data I C. 100 to 150 PSIG Steam Pressure

1. ST 10.4 Relief Valve Manual Actuation

2. ST 1.1 HPCI Logic System Functional fest

3. ST 1.2 RCIC Logic System Functional rest I 4 5.

ST 6. 4 Main Steam Isolation Valve closure Iiming ST 6.5 HPCI Pump, Valve, Flow and Unit Cooler Functional Test I 6.

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ST 6.11 RCIC Pump, Valve, Flow and Unit Cooler Functional Test ST 10.1 HPCI Flow Rate At 150 PSIG Steam Pressure I 8.

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ST 12. 8 Recirculation Baseline Data ST 10.2 RCIC Flow Rate At 150 PSIG Steam Pressure D. 1000 PSIG Steam Pressure

1. ST 1.1 HPCI Logic System Functional Test (quick start only)

2. ST 1.2 RCIC Logic System Functional Test (quick start only)

3. ST 12.8 Recirculation Baseline Date I

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13. CRC Suction Rercute I The CRO pump suction piping was modified to provide a I pref erred source of high quality water to the ccatrol roc drives from the condensate demineralizer outlet neader through a flow control valve. Excess flow recurr.s to tne condensate storage tank. On loss of pressure in the I condensate demineralizer header, the flow control valve automatically closes, returning to a CRD pump suction flowpath trom the condensate storage tanx.

A preoperational test was performed to ensure that adequate pressure and flow was availacle to the C2D pumps, and that I the flow control valve fails closed upon condensate header low pressure, without limiting the CRD pumps f rom delivering the pressure and flow required ror une CRO system operation.

I Upon closure of the control valve, CRC suction returns to tne condensate storage tanx.

14 Crywell Instrument Nitrcaen Isolation Valve Power Su tolv, Se ca ra tion I This modificaricn provides separate power supplies to ene drywell instrument isolation valves (AO-2969a and 29 693) to prevent total isolation of instrument nitrogen to the I drywell, on a loss of power to one power supply. Th e preoperational test demonstrated the following: a) that loss of power to each power supply resultec in the closure of only I one valve, b) with one isolation valve closed, cae normal dr pell instrument nitorgen leads can ce supplied withcut a loss of instrument nitrogen pressure, c) both valves respond properly to an isolation signal.

15. HPCI control Oil Crain System Modification The control oil system to the BPCI turoine throttle valve was modified to prevent control oil drainage during shutdown I pe riods. Two quick start tests were perf ormed to ensure satisfactory system operation.

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16. RSCS Evcass Pressure Transducer / Trio Unit This modification involved the replacement of the original turbine first stage pressure switches PS-2-3-250A and S with electronic transmitter / trip units.

The installation of these instruments will permit changing the RSCS bypass setpoint from 30% of rated power to 21A as governed by Technical Specification Amendment No. 43, dated June 22, 1978.

The instruments were installed, calibrated, and functionally tested in accordance with the approved surveillance tests for the new instrument loops.

I 17. Re la y Ooerability of Various CR-120A control Relays The program to replace the Celon contact-arm retainers on G . E. CR-120A control relays with Valox retainers, was completed during this refueling outage. This completes the commitments made by letter from J. L. Hankins, P.E., to Boyce H. Grier, NRC, dated February 10, 1978, in res pons e to I. E.

Bulletin No. 78-01.

Functional esting of saf ety-related relays wncse contact retaine3 s wera replaced was done by perrorming surveillance I tests, a.odified, if required, to include all necessary checks.

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AP P EN C IX _A FEEDWATER NOCZLE AND CCNTROL RCD CRIVE RETURN LIN2 NO22LE INSPECTION UNIT No. 2 PEAcn EOTTOM ATCMIC ?DWIR STAIICN During the 1978 maiatenance/ refueling outag2, special inspections were performed on t ae f eedwater nozzles and che control rod crive I return line nozzle as part of our program respons: to Interim Technical Report on BWR Feedwater and ControA Rod Drive Estura Line Nozzle Crackin g, UUREG-0312. The resulcs of these inspections were as follows:

(( 5CWAT ER, NO Z2 L E INSPECTIONS An external ultrascnic examination utilizing mechanized and manual techniques was performed by qualified Southwest Assearch Institute personnel on all six f eedwater nozzles. The areas examinec included the nozzle bores, inside clend radil a nc nozzle to safe end welds. In addition, a visual inspection was per:cr:ed by qualified General Electric Company personnal on all

.E six feedwater spargers. No unacceptable indications were found 5 during these inspections.

The following manpower was utilized during enese f eedwater r.ozzle inspections:

SWRI Personnel - 12

. GE Parscnnel -

2 PECO Personnel - 3 I Radiation exposures to these indivicuals are addressed in Appendix ..

CCNTFCL RCC CPIVE RETURN LINE NOZZLi IN3PECIIJ.i A liquid penetrant examination was performed by qualified Jeneral Elect-ic Ccapany personnel on the control rod drive return line I nozzle. The area examined consisted of the surf ace 3o0 a=graes arcund the CRD return nozzle enclosed by a four inct. raulus circle, nine inches into the no::le core and a six inch square area below the no::le. The initial examination revealed seven indications of cracking within the inspec lon area. Ine se indications were removed using air grinders and carcide currs.

The grcundout areas were 'c lended to the adjacent surfaces and a final liquid penetrant examination was performed with no unacceptable indications.

The following manpower was utilized during the CRD return line nozzle inspection and repair:

GE Personnel - 35 PECO Personnel - 3 Radiation exposures to these individuals are addressec in e.ppendix 2.

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I APP ENDIX a Radiation Cose Rates ind Manrem Exc-2nded vurina inscsction of tne Feedwater Nozzles and Control Rod Lrive Return No::le.

Invessel f ee dwat e ' sparger inspection during tne outage resulted in a pproximate'.2 0.450 Rem exposure divided a.nang enree individuals. Dose rates ranged from 100-300 anem/hr during the inspection. Total exposure for drywell and invessel reedwater nortle and sparger inspection totaled approxi.nately 3 :nanrem for the 17 individuals involved.

Control rod drive return not le inspection and repair resulted in approximately 12 Rem exposure to 38 different incividuals. Uose rates ranged from 600-1000 mrem /hr during the CRD return nozzle inspection and repair.

In additjon to the above exposures, approximately 1. 6 5 Rem of exposures was expended in order to ao reactor pressure vessel decontamination during the lowering of reactor water level. Inis exposure was distributed among 14 individuals.

For specir1 tooling and equipment used, refer to deach Bottom I Unit 3 S tartup Report,May 1978, Cocket No.50-27d, Facility Operating License No. DPR-56, 1976 (Appendix b).

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