ML19345A824

Rept of Plant Startup Following Fourth Refueling Outage, Mar 1980.

ML19345A824
Person / Time
Site:	Peach Bottom
Issue date:	03/31/1980
From:	PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC
To:
Shared Package
ML19345A825	List: ML19345A824 ML19345A829
References
NUDOCS 8011250299
Download: ML19345A824 (9)

Text

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PHILADELPHIA' ELECTRIC COMPANY PHILADELPHIA II PEACH BOTTOM ATOMIC POWER STATION UNIT NO. 2

- DOCKET NUMBER 50-277

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REPORT OF PLANT START-UP FOLLOWING FOURTH REFUELING OUIAGE MARCH 1980 1

l SUBMITTED TO  ;

THE UNITED STATES NUCLEAR REGUL\ TORY COMMISSION PURSUANT TO -

FACILITY OPERATING LICENSE NO. DPR-44 I

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PEACil BOTTOM ATOMIC POWER STATION l t

LNIT NO. 2 i

DOCKET NUMBER 50-277 I l REPORT OF PLANT START-UP FOLLOWING FOURTil REFUELING OUTAGE MARCil 1980 i

SUBMITTED TO Tile UNITED STATES NUCLEAR REGUIATORY COMMISSION PURSUANT TO FACILITY OPFRATING LICENSE NO. DPR-44 53 W

PBAPS )

I TABLE OF CONTENTS Page INTRODUCTION............................................... 2 I

SUMMARY

OF TESTS

1. Verif! cat i a. of Shutdown Margin ...................... 3

2. Control Rod Operability and Suberiticality Check...... 3

3. LPRM Calibration...................................... 3 4 Reactivity Anonalies.................................. 3

. 5. Core Verification..................................... 3

6. Cold Critical Rod Pattern Prediction.................. 3

7. Core Power Symnetry and TIP Reprodu'cibility Test.................................................. 4

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3. Control Rod Drive Scram Timing........................ 4

9. Special Radiation Survey.............................. 4

10. CRD Return Line Cut and Capped........................ 4

11. Analog Transmitters / Trip Units for RPS/PCIS and ECCS Functions (MOD 439).......................... 5

12. Feedwater Flow Linit on Scram (MOD 531)............... 5

13. Main Steam Relief Valve (MSRV) Discharge

'T' Quencher (NOD 537)................................ 5 14 Autonatic Isolation of RHR Sample Lines (MOD 555)..... 6

15. Automatic Isolation for Radioactive Gas Sampler (MOD 577D)............................................ 6

16. Additional Containment Isolation Valves and Logic Additions (MOD 578)................................... 6

17. Miscellaneous Containment Isolation Inprovements (MOD 577E)............................................. 7

l l PBAPS l INTRODUCTION i

l The Peach Hotton Technical Specification 6.9.1 Routine Reports requires submittal of a startup report following any outage in l

l which certain safety related events may occur. Installation of l fuel of a different design is one of these events. This report,

! prepared to neet the Technical :i p e c i f i c a t i o n requirenent, describes the startup progran inplenented to provide assurance that the safe operation of the plant was not diminished by the i activities of the fourth refueling outage.

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The Peach Botton Unit 2 was out of service from March 21, 1980, to August 9, 1980, to acconnodate a naintenance/ refueling outage.

During this 145 day outage, 144 fuel bundles of the original (7X7) design and 132 (8X8) were replaced with the new (P8X8R) fuel design. The new fuel was manufactured by the General Electric Company and was designed to provide additional

!g operational flexibility and fuel economy. The unit was returned

! 3 to service on August 9, 1980, and reached full power on August 25, 1980. Start up tests were performed before and during the return to power.

i The startup tests identified in the r' . S . A . R . were addressed and

! those which involve areas which were affected by outage j activities were included and are sunnarized herein. Additional l

special tests connected with specific outage activities were also included in the progran and are discussed in this report. The successful implementation of this startup progran insures that i the Unit 2 refueling outage has resulted in no conditions or j systen characteristics that in any way dinish the safe operation i of the Unit.

The tests and data referenced in this report are on file at the 4 Peach Botton Atomic Power Station.

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1. VERIFICATION OF SIIUTDOWN MARGIN

! The co"e shutdown margin was determined during the initial reactor criticality in the B sequence. The actual shutdown nargin was 1.94% IK/K, as compared with a Technical

! Specification mininum value of 0.38% 1K/K. This test was

! performed on August 9, 1980.

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2. CONTROL ROD OPER ARILITY AND SUBCRITICALITY CilECK Each control rod was withdrawn and inserted and sub-criticality verified per surveillance test ST 10.8. The test was completed on August 5, 1980, prior to start up.

I 3. LPRM CALIBRATION 1

LPRit calibrations were successfully performed at approximately 25% anu 1007 power levels per surveillance test ST 3.4.1.

4. REACTIVITY ANOMALIES Surveillance test ST 3.7, " Reactivity Anomalies", was I successfully performed on August 26, 1980. The predicted number of control rod notches inserted at rated conditions was 615 with a + 1% 1Keff range of 415 to 815 cont rol rod

notches inserted. The actual number of control rod notches j

inserted was 506, which satisfies the + 1% 1Keff criteria.

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5. CORE VERIFICATION Post-alteration core verification was begun on July 4, 1980 in accordance with surveillance test ST 12.10 " Core-Post-Alteration Verification". A seating verification was performed for all fuel assemblies. Several assemblies were I improperly seated. These assenblies were rescated and reverified, on July 6, 1980. All fuel bundles were in the 2

correct locations and in the proper orientation prior to beginning cycle 5 operation.

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6. COLD CRITICAL ROD PATTERN PREDICTION i

The cold critical rod pattern prediction comparison

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surveillance test ST 3.9 was successfully performed for Unit

2. The predicted core keff was 1.003 and the actual core I

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PBAPS keff was 1.00/ tor a difference of -0.004 The difference 1lm is - 0 . 4 7. and satisfies the + 1% test acceptance criteria.

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j 7. CORE POWER SYttMETRY AND TIp REPRODUCISILITY TEST i

! Core power synnutry and TIP reproducibility test data was

{ analyzed for sequence 'A' conditions on 9/4/80 (98'. i power).

l Two data sets were taken. The total TIP uncertainty was 4.4267 and 4.4397. for the two sets; therefore the test 15

!g criteria that the total TIP uncertainty not exceed 9% is

} satisfied. The naxinun deviation between synaetrically

! located pairs was 19.8% and 19.3% for the two data sets

! respectively, which satisfies the 25% acceptance criteria.

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8. C0" TROL ROD DRIVE SCRAM TIMING 1

{ All 185 control rods were scran tined satisfactorily j following all core alterations at nominal reactor pressure l of 1000 psig in accordance with surveillance test ST 10.7.

l Technical Specification requirements were all satisfied.

The test was perforned July 17, 1980.

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i 9. SPECIAL RADIATION SURVEY l

i I Radiation dose rates neasured in aLcordance with ST 7.9.1 lgg with the unit at greater then 90% rated power, evidenced i expected levels. This survey "crified that any shielding jl j

!E disturbed during the refueling outage had been properly repositioned. In addition surveys conducted in areas traversed by new off gas piping resulted in acceptable

! anhient radiation levels.

i 10 CONTROL ROD nRIVE RETURN L I ': E REMOVAL The CRD return line had been valved out of service during Cycles 3 and 4 to prevent thernal fatigue cracking on the I inner radius of the reactor vessel nozzle. To clininate any further cracking, the CRD return line uns cut and capped at the reactor vessel nozzle. The pining inside the drywell lg was renoved. The drvwell penetration was capped inside and lg blanked outside.

i l Operation in this node was satisfactorily tested on P33 in l April 1977 and further confirned by subsequent operat1.

I 3th Units ' a- for t' ree years, uith the ' ' " '

return line

! itier valved out or renoved.

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I p ilA p S I To satisfy the NRC requirement of compifance with draft NUREG-0619, a test was performed to measure the maxinua systen flow capability to the reactor vessel through the CRD's. This was done during the operational !! y d r o s t a t i c Test with both pumps running. Measured naximum flow was 212 G p :1.

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11. ANALOG T R A N S!!ITT E R S / TR I p UNITS FOR RpS, pCIS AND ECCS FUNCTIONS ( ?!O D 439)

This nodification replaces twenty-four mechanical switches, I which had been prone to setpoint drift and functional failures, with more reliable analog transnitters and histable trip units. The instruments affected are Reactor and Drywell pressure signals to the RpS, pCIS and ECCS Control Logics. Also, as part of this nodification, two electronechanical transmitters which transnit reactor level I to control roon indicators were replaced by existing e l e c t r o r. i c transmitter signals.

control logic were made.

No changes in the systen The instrument channels were calibrated and functionally tested with approved, revised surveillance tests. Unit 2, Technical Specification Anendment 68, Dated May 5, 1930 '

. governed these changes.

12. FECDWATER FLOW LIMIT O_N S C R A?! (MOD 531)

This nodification added an upper flow limit to the feedwater punp control circuit following a scran with three feed pumps operating. Void collapse during scran causes a transient low level signal to the feedwater control systen. In response to this sudden level error feedpump flow increases quickly which results in a low feedpunp suction pressure i trip of all three feedpunpa. P, y limiting the feedpump speed to 9 0 7, the low suction pressure trip is averted. The systen was checked using an approved pre operational test.

13. ?t AI N STEAM RELIEF /ALVE (MSRV) DISCilARGE 'T' OUENCHER (MOD S37)

I f5 The "Rans li c a d " discharge on the MSRV line were replaced with 'T' Quenchers. The 'T' quenchers are located near the vertical center line of the torus. The purpose of this modification is to reduce the reaction forces on the torus structure during !!SRV operation. Special instrunentation was added to the torus structure and a series of one and two

?tSRV operation tests were conducted on August 18 and 19, i 1990. l 1

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{ PBAPS The tests indicated that strens levels with the 'T' quencher

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are on the order of 1/3 of the stress with the " Rams lle a d ' .

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14. A UTO ?! A T I C ISOLATION O_F R il R SAMPLE LINES (!!O D 555)

This modification installed tuo air operated valves in each of the four sample lines. The modification autonatically isolates the Rif R sanple lines on a Group I isolation signal, l to prevent the release of radioactive water fron the R il R

! system to the Reactor Fullding.

l The valves were installed and initially checked in accordance with Construction Division Quality Control I Procedures, and functionally tested in accordance with the PORC approved Preoperational Test and Surveillance Test. ,

This modification was a result of the review conducted in i accordance with NUREG 0578 Iten 2.1.4.

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A UTO!! A T I C ISOLATION FOR RADI0 ACTIVE CAS S A?!P LER (!10D 57/D)

This modification was installed to provide an automatic j iI isolation signal to the existing radioactive gas sanple isolation valves. On a Group III isolation signal these valves isolate to prevent the release of radioactive materials under post accident conditions. '

The isolation was installed and initially checked in accordance with Construction Division Duality Control  !

Procedures, and functionally tented in accordance with the I I PORC approved Preoperational Test and Surveillance Test.

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This nodification was a result of the review conducted in accordance with NUREC 05/8 Iten 2.1.4.

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- ADDITIONAL CONTAIN!!ENT ISOLATION VALVES AND LOGIC ADDITIONS (MOD 573) 1 1

I This modification wa- Installed to add containnent isolation valves in the drywell radioactive gas sanpler and instrunent nitrogen comp re s s o r suction lines. On a Group III isolation

,g signal these valves isolate to prevent the release of radio-E ac'ive ""toria o """or " s' accid""' candi't "s-1 The aodification was installed and initially checked in j

accordance with Construction Division Quality Control

{ Procedures, and functionally tested in a c c o r ,i a n c e with the PORC approved Preoperational Test and PORC approved jI Surveillance Test. This nodification was a result of the review conducted in accordance with '!U R E G 057S Iten 2.1.4.

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17. MISCELLANEOUS C O N T A I N!!E N T ISOLATION IMPROVEMENTS (MOD 57/E)

This nodification was installed to provide the following improvements:

l l 1. Relocating the instrument nitrogen drywell supply line

} isolation valve bypass switches from the cable

! spreading room to the control room.

2. Adding indicating lights on the graphic display panel for instrument nitrogen supply line isolation valves.

3. Upgrading torus vacuun relief isolation valves to

)g j cafeguard and provide separation between redundant

p valves.

4. Test connections to facilitate the LLRT of the CAD Systen injection line check valves, j The nodification was installed and initially checked in accordance with Construction Division Quality Control ll jB Procedures, and functionally tested in accordance with the PORC approved Preoperational Tests and Surveillance Tests.

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