ML20082B939

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Rept of Plant Start-Up Following Eighth Refueling Outage on 910112-0412.W/undated Ltr
ML20082B939
Person / Time
Site: Peach Bottom Constellation icon.png
Issue date: 07/17/1991
From: Miller D
PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC
To:
NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NUDOCS 9107170182
Download: ML20082B939 (8)
v • d • e


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PIIILADELPIIIA ELECTRIC COMI%NY PIACilIKII'IUM AloMIC POWIR WIATION 2.2 R. p.1.Ikn 20H Q j , -

Delta l'enm)lvania 1731i reum wrtsta-sus nre sa or a nottawe (717) 4 %-7014 D. It Miller, Jr.

Vice hendent Docket No. 50 277 Licenso No. OPR 44 U.S. Nuclear Regulatory Commission Document Control Desk Washington, DC 20555

Subject:

Peach Bottom Atomic Power Station Unit No. 2 Roport of Plant Startup Following Eighth Rofueling Outago Gentolmen:

Attachod is the Peach Bottom Atomic Power Station Unit No. 2 Report of Plt.nl Startup Fellowing Eighth Refueling Outage. The toport is submitted persuant to toporting roqulromont G.9.1.a in Appendix A to Licenso No. DPR 44.

sincerely, to, nnf ni~6 Q QM).

DBM/DRM/AAF/MJB:lph

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t Enclosures cc:

R.A. Burricelli, Public Service Electric & Gas T.M Gerusky, Commonwealth of Pennsylvania J.J. Lyash, USNRC Sonlor Resident inspector R.L. McLean, Stato of Maryland T.T. Martin, Administrator, Region I, USNRC H.O. Schwomm, Atlantic Electric J. Urban, Dolmarva Power INPO Records Contor 9107170182 910717 PDR ADOCK 05000277

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.e PKIIADELPEIA ELECTRIC 00HPANY Peach Botten Atcutic Power Station Unit No. 2 Docket Number 50-277 REPORT OF PIANT START-UP FOIJDifIMI EICRfrH REFUELING OUIME January 12, 1991

'IO April 12, 1991 SUBMITTED TO THE UNITED STATES NUCLEAR REGUIATORY 00tMISSION PURBUhNT 70 FACILITY OPERATING LICENSEE DPR-44 i

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l Preparation Directed.By John T. Budstynski Peach Botton Atomic Power Station R.D. #1 Delta,PA 17:114

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TABLE OF CO?m3 FPS i

PAGE #

IJ mOCUCTION AND SLM4ARY...........................................

3 i

SLM4ARY OF TESTS 1.

Chemical and Radiochemical..................................... 4 2.

Radiation Measurements.........................................

4 3.

Fuel Loading...................................................

5 4.

Shutdown Margin................................................

5 l

5.

Control Rod Drivo 'Ibsting......................................

5 6.

Control Rod S(quence...........................................

6 7.

Rod Pattern Exchange...........................................

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8.

SR!4 Ibrformance................................................

6 9.

IR!4 Ibrformnce................................................

6

10. ITRI4 Calcul ation...............................................

6

11. APR!i Calculation...............................................

6

12. Process Camputer...............................................

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13. Core Power Distribution........................................

7 14. Core Ibrformance............................................... 7 l

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I!TIROEUCTIOU f

Unit 2 Peach Bottom Technical Specification Section 6.91.n Routine Reports requires submittal of a Startup Report following an outago in khich fuel of a different design was installed. 'Ihis report summarizes the plant startup i

testing performed to assure that no conditions or syst'm characteristic changes j

have been created by the Eighth Refueling Outage of Unit 2 which diminish the i

safe operation of the plant.

SUMMARY

Startup testing was perfomed in accortlance with the Final Safety Analysis Report (PSAR) Section 13.5, Et;artup and Power Test Procrar). 'Ihe startup tests identified in Section 13.5 were aadressed and those which involved areas which were affected by outage activities were included and summarized herein. Measured and calculated values of operating cortlitions arti charactoristics obtained during the Startup rest Program were campared to design predictions and specifications. Invel 1 criterion were either met, or discrepancies were investigated and determined to have no etfect on safety, reliability, operability and pressure integrity of the systers tested. Corrective actions were not required to obtain catinfactory plant operation.

Peach Bottom Unit 2 was out of service from January 12,1991 to April 12,1991 to accomm. late the Eighth Refueling Outage maintenance activities.

I During this 97 day outage:

  • 61 P8x8R-P8IRB284,and 99 P8X8R-P8DRB299 fuel bundlen were replaced with 148 GE9B-P8DNB320, 4 Gell-P9HVB307, 4 ANF9X-314B, airl 4 ABB-SVEA96-286

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fuel htndles.

  • 40 Unit 2 & Common modifications were completed.

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'Ihe unit returned to cammercial operation on April 18,199.'. and reached full power on April 28,1991. Startup testing was completed on July 2,1991.

'Ihe successfully implemented Startup Test Program insures that the Eighth Refueling Outage of Unit 2 has resulted in no conditions or system characteristics that diminish the safe operation of the plant.

The tests airl data references in the report are on file at the Ibach Bottoni Atomic Power Station.

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i STARIUP REPORT

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Peach Bottom Atomic Pcuer Station Unit No. 2 1.

.Q1m.ical and Radi.pchemical

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Chemical and Radiochemical analyses were performed in accortlance with FSAR Section 13.5.2.2.(1):

a.

Prior to Fuel load:

Chemistry Limits per G-10 (Chemistry Goals) were verified on a daily basis.

t b.

Prior to Startup:

The Shift Chemist verified that RP 7.8 (Chemistry Preparation for Reactor Startup) was performed prior to startup airl completed on April 21, 1991. Also verified that Chemit,try Limits per CH-10 were withir specification.

c.

During Startup:

Coolant chemistry was determined to meet water quality specifications and process requirements via ST 7.2.3B (Reactor Startup Chemistry i

( <100 Kibs/11r on April 21, 1991. For high steaning rates ( >100 Klbs/Hr) ST 7.2.3A (Reactor Startup Chemictry) was performed on May 5,1991. 2. Radiation Measurements Radiation Measuruments were made in accordance with FSAR Section 13.5.2.2.(2): a. Prior to Fuel load: 1 Routine surveys were taken daily throughout the protected area to assure personnel safety and to maintain Activity Buildup base data via HP 200 (Routive Survey Program). b. During Startup: Radiation was monitored to assure the protection of personnel and continuous compliance with the guidelines of 10CFR20 during plant operation above' 20% pcuer, performed on April 19, 1991, via ST 7.9.1 (Radiation Survey After Refueling). l l

Page 5 of 7 [ N t 3. Juel Imdim i Fuel loading, Control Rod Functional and Subcriticality Checks were performed in accortlance with FSAR Section 13. 5. 2. 2 (3). N el loading was completed on March 1,1991 via N-6C (Wel Movement and Core Alteration l Procedure During a Nel Handling Outage). Bundle locations, and orientation were verified via ST 12,10 (Core Post-Alteration Verification) and canpleted on March 5, 1991. Each control rod was withdrawn and inserted to verify rod coupling integrity, proper rori withdruatl and insertion, and i subcriticality. Level 1 criteria was mt when core shutdown margin was demonstrated with a fully loaded core on April 13, 1991. Control Rod Test data is documented in ST 10.8 (Control Rod Ierformanco Test) completed i April 6, 1991. I 4. Shu_t4cwn Mamin Core shutdown margin was demonstrated in accontar~e with MAR Section

13. 5. 2. 2. (4). An "In-Sequence" shutdown margi.1 of 1.41% delta K/K was l

obtained during the initial reactor startup in the A nequence. his satisfies the Invel 1 criteria that the core must be subcritical by at lease 0.38% delta F/K with any rod fully withdrawn. Test data is e docunented in ST 3.8.2 (Shutdown Margin) ::ampleted April 13,1991. h e design predicted core Keff was canpared to the meatsured value at initial startup on April 15,1991. W e predicted M ir was 1.003 as conpared to the reasured Keff of 1.005. We differenco between predicted and the measured values was -0.2%, which meets the acceptance criteria of 1%. W e test data is documented in ST 3.9 (Critical Eigenvalue Camparison) completed on April 15, 1991. 5. Control Rod Drive Testim Control Rod Drive (CRD) testing was perfomed in acconlance with FSAR Section 13. 5. 2.2. (5). In cold shutdown, each CRD was tested for position indication, normal insert / withdrawal times and coupling via ST 10.8 (Control Rod Withdraw Test) completed on April 6,1991. At rated pressure during hydrostatic testing, CRD scram insertion times were obtained vic ST 10.7 (CRD Scram Insertion Timing, " Full In" & " Full Out" Position Indication, & Rod Coupling Integrity Check For All 18S Control Rods) completed on April 9,1991. During startup, at rated reactor pressure, position in;iication via GP-2(Nornal Plant Startup), coupling via ST 10.8-1 (CRD Coupling Integrity Test), and scram insertion tines via ST 10.13 (CRD Scram Insertion Timing of Selected Control Rods) were tested. The testing performed satisfied IcVel 1 arri 2 criteria. - I i .., ~.. _

~ Page 6 of 7 j 6. Control Rod Scauence '!he conttul rod coquence was followed in accordance with FEAR Section 33.5.2.2.(6). The sequence was defined in GP-2-2 Appendix 1 (Startup Rod l W4.thdraw Sequence Instmetions) and verified for use by the Rod Worth 141nimizer (RC4) via ST-R-62A-220-2 (RWM Sequence loading Verification) on April 9,1991. The MQ4 was tested operable on April 12, 1991 via ST 10.5 (Riti Operability Check). The critical rod pattern was recorded on April 13, 1991 via ST 3.8.2 (Shutdown 14atvin). 7. Egl Pattern Excharrig Rod pattern adjustments were performed in accordance with TSAR Section 13.5.2.2.(7). Rod pattems adjuatments were guided by RE-31 (Reactor Engineering Startup/Ioad Drop Instructions) during power ascension. 7hemal limits were not exceeded. 8. SH4 Performmla Soutre Range 14cnitor (SR4) ins,trumentation operability was checkcd during perfonunce of startup procodure GP-2. FSAR Section 13. 5. 2. 2. (8) criteria of a minimu:r. count rate of 3 counts / soc. wan verified to be met for the SR4's. Ibta la documented in GP-2. 9. IR4 Performance Intermediate Rarge lionitor (IRM) performance was testod in accordance with ISAR Section 13.5.2.2. (9). The IH4 scram net-points mot the Invel 1 criteria of SI2N-60C-IR4-A404 (Intermediate Range Monitor mannel "A" CalibratioryTunctional Geck) and SI2N-60C-IRM-B401 (Internodiate Rarge Monitor mannel "B" Calibratiorg/ Functional meck) dated April 12, 1991.

10. IPR 4 Calculation local Power Range Monitor (LPR4) calibrations were conpleted in accordance with FSAR Section 13.5.2.2. (10). Calibrations were performed at 24% and 99% ratcd thermal power per ST 3.4.1 (IPR 4 Gain Calibration) on April 22, 1991 ard May 1,1991 respectively.
11. APRM Calibration thunerous Average Power Range Monitor (APRM) calibrations were completed I

during startup in accordance with FSAR Section 13.5.2.2.(11). Test data is documented in ST 3.3.2's completed frun April 16,1991 at approximately i 1.5% power to April 28, 1991 at 100% power. l. 9 ,,.--nr...,, n-e- --n-e

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12. Process Comuter he Process Camputer was tested in accordance with ISAR section 13.5.2.2.(12). NSS Monicore Program was installed in the Plant Monitoring System to support fuel thermi limit calculations of the latest fuel design. A mnual calculation was perforned via RT-R-OS9-500-2 (Checkout of the NSS Campater Calculation of Core terml Power) at approximately 100% p wer June 17, 1991.

We therm 1 limit calculations were verificd via General Electric DUCIE program with full power NSS data collected from the Plant Monitoring System on June 25,1991.

13. Core Power DistributiQD Core power symmetry and Traversing Incore Probe (TIP) reproducibility were tested in accordance with FSAR Section 13.5.2.2.(17). Two full sets of TIP traces were obtainod at approximately rated power on June 19, 1991.

We TIP readityJs were within the standard deviation ur.ed to establinn safety limit criteria of 8.7%, per General Electric Examont NEDE-24011 Table S.2-1. We maximum deviation between symmetrically located pairs satisfied the 25% acceptance criteria for core power rumnintry. Test data is documented in RE-27 (Feach Bottom 2 and 3 Core PtNur Symmetry and TIP Reproducibility Test) procedure completed on July 2,1991.

14. Core Performance Core performance was evaluated in accordance with FSAlt Section 13.5.2.2.(18). The core thermal limits were verified daily above 25%

i reactor power via the Plant Monitoring System. Se full pcwer control pattem provided by PEco Fuel Mariagement Section was verified per ST 3.7-2 (Reactor Anamalics) on April 29, 1991. l i l l l}}

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