ML20095D207

Startup Rept,Cycle 9

ML20095D207
Person / Time
Site:	Peach Bottom
Issue date:	04/20/1992
From:	PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC
To:
Shared Package
ML20095D201	List: ML20095D197 ML20095D207
References
NUDOCS 9204270006
Download: ML20095D207 (32)

Text

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PillLADELPillA El.ECTRIC COMPANY PEACll BOTTOM ATOMIC POWER STATION UNIT NO. 3 STARTUP REPORT I

CYCLE 9 SUBM11TED TO TliE UNITED STATES NUCLEAR REGUIATORY COMMISSION PURSUANT TO FACILITY OPERATING LICENSE DPR-56 DOCKF NUMBER 50 278 44 9204270006 920420 PDR ADOCK 05000278 P PDR

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. PAGE q 1.0 INTRODUCrlON/

SUMMARY

. . . . . . . ........ ................. ........... 11 1.1 Rr vrt Abstract .................,. ...... ....................... 12 1.2 S u m m a ry , . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 1-3 13 I cach flottom Plant Description . . . . . . . . . . . . . ...... ..... .... ... ...14 ,

Table 13 2 Peach Bottom Unli 3 Plant Parameters . . . . . , ....... .. .... . . . 16 2.0 TEST OBJECTIVfE DESCRIPTION, ACCEL'rANCE CRITEiRIA, AND RESUI.TS . . . . . . 21 2.1 Chernical and Radiochemical . . . . . . . . . . . . , . . . . . . . . . . . . . . . . . . . . , . . . . 2-2 2.2 Radiation Measurements ... ... ..... .. ................. . . ..24 23 Fuel lmading . . . . . . . .... .... ... .. ...... ..... .. . ..........25 2.4 Shutdcnm Margin . ...... ..... .... . ....- . ... . ......... 26 2.5 Control Rod Drive , ....... . .............. ....... ..., . . . . . 2-7 -

'6- Control Rod Sequence . . . . . . . . ....... ,. . .... , .. .. . 2-9 2.7 SRM Performance . . . . . . . .. ..... ...... . .......... . . . . . . . 2 10 2.8 IRM Performance . . . . . . . . .... . ........ ....... ... ,. . . 2 11 2.9 LPRM Calibration . . . . . . . . . . . . ......... ,....... . .... ... . . . 2 12 2.10 APRM Calibrar;on . . . . . . . . . . . . . . . . . ., ......,.... ... .... . . . 2-13 ,

i 2.11 Process Computer . . . . . . . ........................ ......... . . . . . 2- 14 2.12 . Reactor Core Isolation Cooling . .. . , ......... . .... . ..... . . . 2 15 2.13' liigh Pressure Coolant injection (NPCI) System . .... ............ . . . . . . 2 16 2.14- Core Power Distribution and TIP Uncertainty . . ........... .. , . . . . . . . 2 17 2.15 Core Performance . . . . . ... .. .. .. . ... . ... ........ . . . 2-18 .

2.16 Feedwater System . . . . . . .. .... ........ .... .. .. ... . . . . . 2 19 2.17 R elie f Valve s . . . . . . . . . . . . . . . . . , ,. .. ,,,.. .... .. ... ... . 2-21 2.18 Recirculation Flow Control System . . ... .. ....... ..,... . .... .. . 2 22 2.19 Recirculation System . . . . . . . . . . . .. . .... .. ..... . 2 23 a

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' ItCHODUCI'lON/

SUMMARY

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, 1.1 REPORT AllSTRACT This Startup Report, writtrn to comply with Technical Specification paragraph (i.4.1.a. consists of a summary of the Startup and power Escalation Testing pe formed at Peach Ilottom Unit 3. During this refueling outage, 2% new bundles of GE911 were added, while the remaining bundles were shuffled. All reused bundler, were inspected and cleaned as necessary, to remove debris. Additionally, the tiottom llead and the llottom licad Drain line wt re cleaned and inspected.

Tee report addresses the Startup Tosts identified in chapter 13.5 of the 13AR and includes a description of the mtaured values of the operatir;g conditions or characteristics obtained during the test program with a con parison of these salues to the Acceptance Criteria. Pertinent portions of the Digital I'ecdwater System (Mod IM3) testing are also addressed in this report.

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SUMMARY

Peach llottom Unit 3 was out-of senice from September 14,1991 to January 8,1992 to accommodate a refueling outage. The unit returned to scryice on January 9,1912 and reached full power operation January 20,1992.

The successfully imt.!emented Peach llottom 3 Startup Propam ensures that the eighth refueling outage of Peach Bottom 3 has sesulted in no conditions or system characteristics that diminishes the safe operation of the plant. The tests and data referenced in this report are on Glc at the Peach Ilottom Atomic Power Plant.

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. I r 1.3, PEACil BO'ITOM PLANT DESCRIPTION The Peach Hottom Atomic Power Station is a two unit nuclear power plant. The two units share a common control room, turbine operating deck, radwaste system, and other auxiliary systems. Each unit has a separate refueling fkor.

The PilAPS is located partly in Peach Bottom Township, York County, partly in Drumore Township, Lancaster County, and partly in Fulton Township, Lancaster County,in southeastern Pennsylvania, on the westerly shore of Conowingo Pond at the mouth of Rock Run Creek. The plant is about .M mi north northeast of Haltimore, Maryland, and fi3 mi west southwest of Philadelphia, Pennsylvania. Conowingo Pond is formed by the backwater of Conowingo Dam on the Susquehanna River; the dam is hicated almut 9 mi downstream from the Unit 3 reactor. The nearest communities are Delta, Pennsylvania and Cardiff, Maryland approximately 4 mi .

southwest of Unit 3.

PECO owns the 620-acre property lying within the site boundary except that the immediate area on which Unit 2 and 3 stand is owned by PECO, Public Scrvice Electric and Gas Company, Delmarva Power & Light Company, and Atlantic Electric Company as tenants in common. The adjoining area lying along ihe discharge canal in a downstream direction, was owned by the Philadelphia Electric Power Company, a wholly owned subsidiary of PECO, as part of the Federal Power Commission Project No. 405 (Conowingo). An application dated February 25,1%9, was filed with the Federal Power Commission to alter the Conowingo Project boundary so as to exclude from that project this 26 acre tract and a littoral strip ol' land about 7,tm ft in length  ;

l (upstream of the 26-acre tract) which was subject to flowage rights in favor of the Conowingo Project. On October 13,1970, the Commission issued an order approving the application.

Although not a part of the Peach llottom project, the remaining land along both sides of Conowingo Pond, fnun below lioltwood Dam to Conowingo Dam ranging up to 300 ft back from the waterline, is owned by PLCO's wholly owned subsidiaries. <

Each of the Peach Bottom units emplop a General Electric Compaay boiling water reactor (BWR) designed to operate at a rated core thermal power of 3293 MWt (1@J steam flow) with a corresponding gross electrical output of 10'u MWe. Approsianately 37 MWe are used for auxiliary power, resulting in a net electrical output of iO55 MWe. See Table 1.21 for Peach Bottom Plant Parameters.

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. The primary containment is a pressure suppression system and houses the reactor vessel, the reactor coolant recirculation systems, and other primary sptem piping. The primary containment system consists of a drywell, a pressure suppression chamber which stores a large volume of water, a connecting vent system between the drywell and the suppression 1xxil, isolation valves, vacuum breakers, containment cooling systems, and other senice equipment, The drywell is a light bulb-shaped steel pressure vessel with a spherical lower portion. ,

The drywell is enclosed in reinforced concrete for shiciding purposes.

The Architect Engineer and constructor was Bechtel Power Corporation.

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.. TAllLE 13-1 PEACil 1101 TOM UNIT 3 PLANT PARAN1ETERS Parametes Ms Rated Power (MWt) 3293 Rated Core Flow (Mlb/ht) 102.5 (1)

Reactor Dome Pressure (psia) 1020 Rated Feedwater Temperature (Deg. F.) 376.1 (2)

Total Steam Flow (Mlb/hr) 1337 Vessel Diameter (in) 251 Total Number of Jet Pumps 20 Core Operating Strategy Control Cell Core Number of Control Rods . 185 Number C Fuel Bundles 7M Fuel Type 8 X 8 (Harrier) _

Core Active Fuel Length (in) 150 Cladding Thickness (in) 3 0.032 Channel Thickness (in) 0.080 MCPR Operating Limit L27 for llP/P8XSR & GE8X8 Ell 1.29 for GE8X8NB (3)

Maximum LilGR (KW/ft) 13.4 for 13P/PMX8R & GE8X8EB 14.4 FOR GE8X8 Nil Turbine Control Valve Mode Modified Partial Are Turbine Ilypass Valve Capacity (% NilR) 25 Relief Valve Capacity (% NilR) 87.4 Number of Relief Valves 13 Recirculation Flow Control Mode Variable Speed M/G Sets 1-6

i NOTES FOR TAHlJilJd (1) Unit 3 is analyzed for increased core flow to 105'I.

(2) Unit 3 is analped for a 60 degrees F final Feedwater ternperature reduction.

(3) See Core Operating Limits Report for Peach Bottom hioad 8, Cycle 4 for specific,s.

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4 4 2.0 - Test Objectives, Description, Acceptance Criteria, and RerAttts

. The tests comprising the startup and power test program are discussed with reference to the particular test purpose, brief description, statement of acceptance criteria where applicable, and the test results.

' Where applicable, a definition of the relevant acceptance criteria for the test is given and is designated either

• Level 1

• or " Level 27 A Level 1 criterion normally relates to tbc value of a process variable assigned in the ,

design of the plant, component systems, or associated equipment. If a Level I criterion is not satisfied, the plant will be placed in a suitable hold-condition until resolution is obtained. Tests compatible wah this hold-condition may be continued. Following resolution, applicable tests must be epeated to verify that the requirements of the Level I criterion are satisfied, i

A Level 2 criterion is associated with expectations relating to the performance of systems. If a I.evel 2 crilenon is not satisfied, operating and testing plans would not necessarily be altered, Investigations of the incasuren.ents and of the analytical techniques used for the predictions would be stailed.

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. 2.'1 Chemical and Radiochemica!

l 01 iectives i

j The principal objectives of this test are (1) to maintain control of and knowledge about J ; quality of the reactor coolant chemistry, and (2) to determine that the sampling equipment, procedures, and analytic techniques are adequate to supply the data required to demonstrate that the coolant themistry meets water quality specifications and process requirements.

Secondary objectives of the test program include data to evaluate the performarce of the fuel, operation of the demineraliiers and filters, condenser integrity, operation of ik off-gas system,and calibration of certain process instruments, pescription Subsequent to fuel loading during reactor heatup and at each major power level change, samples will be taken and measurements will be made to determine the chemical and radiochemical quality of reactor water and reactor feedwater, gaseous activities leaving the air ejectors, decay times in the off pas lines, and performance of filters and demineralizers. Calibrations will be made of monitors in the stack, liquid waste system, and liquid process lines.

AcceplansrSiitsfia LevelI Water quality must be known r.nd must conform to the Water Oaality specMenions at ali <it.cs.

The aethities of gaseous and liquid effluents must be known and they mar, conform to license limitations.

Chemical factors defined in the Technical Specifications must be md.talud wiOw the limits specified.

_ Level 2 Chemical factors in the Fuel Warranty must be maintaiwJ k 2n the :,peciGed limits.

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. R esults a ., Prior to l'uel lead:

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

b. Prior to Startup:

Chemistry requirements were verified by RT 7.8 (Chemistry preparation for Reactor Startup) on January 10, 1W2, The Shift Chemist also verified that chemistry hmits were within specification per Cil 10.

c. During Startup:

Coolant chemistry was determined to meet water quality specifications and process requirements via ST-C4NS-824-3 (Reactor Startup Chemistry { <100Klbs/hr) completc

• auary 3, IW2.

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. 22 Radiations Measurements Obiectives Tu determine the background garama and neutron radiation1 :vels in the plant environs prior to operation in order to provide base data on activity buildup. Also to monitor iadiation at selected power levcis to assure the protection of personnel and continuous compliance with the guideline standards of 10CFR20 during plant operation.

Descrintion Subsequent to fuel soading, during reactor heatup and at various power levels, gamma radiation level measurements and, where appropriate, thermal and fast neutron dose rate measurements, will be maJe at significant locations throughout the plant. All potentially high radiation areas will be suncyed.

AcceNansq Criteria intl 1 The radiation doses of plant origin and occupancy times shall be controlled consistent Eth the guidelines of the standards for protection against radiation outlined in 10CFR20 NRC General Design Criteria.

Results

a. Prior to Fuel lead:

Routine surveys were taken throughout the protected area to assure personnel safety and to maintain Actisity Buildup base data via HP 200 (Routine 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.

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_-~q'i 23 Fuel Leading Obiecthu The purpose of this test is to load new fuel and shuffle existing fuel safely and efficiently.

DescriptiqD Prior to loading a control cell, control rods and control blades will be installed. Fuel loading / shuffling will be such that each bundle loaded will be neutronically coupled Ic, an operable SRM, in that quadrant.

_. Acccotance Critetig Itvel1 Tbc criteria for successful completion of this test are that (1) the core is fully loaded, and (2) the core shutdown margin demonstration has been completed.

BSsults e

Fuel loading was completed on December 8,1991 via FH 6C (Core Component Movement and Core Al;cration Procedure During a Fuel Handling Octage).15undle hxations and orientation were verified via RT R4x4970 3 (Core Post-Alteration Verification) and completed on December 9,1991.- Each Control Rod was withdrawn and inserted to verify rod coupling integrity, proper rod withdrawal and insertion si sdt, and subcriticality.

Level I criteria were met wl:cn core shutdown margin was demonstrated t.ith a fully loaded core en January 2,1992. .

Control Rod Test data is documented in ST 10.8 (Control Rod PerformanG TN) completed on December M IW1.

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• 2A Shulden.MarWt Dbituild The purpose of this test is to demonstrate that the reactor will be subcritical throughout the fuct cycle with any single control rod fully withdrawn.

Description Suberiticality was demonstrated with the 'in-sequence melbod".

Amputnce Criteria isll a) The fully loaded core must be suberitkal by at least 0.38bK throughotc the fuel cycle with any rod fully withdrawn.

.B11Mll6 An 'in-sequence shutdown margin of 1.915% delta K/K was obtained during the initial reactor startup in the A sequence. This satisfies the Level I criteria that the core must be suberitical by at least 0.38% ocita K/K with any rod fully withdrawn, Test data is documented in ST-R-032-910-3 (Shutdown Margin) completed January 2, IW2.

The design predicted core Keff was compared to the measured value at initial startup on January 2,1992. The predicted Keff was 1.00415 as compared to the measured Keff of 1.0033. The difference between predicted and measured values was 0.000SS, which meets the acceptance criteria of1' 1% The test data is documented in ST 3.9 (Critical Eigenvalue Comparison) completed lanuary 2,1992.

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. 2 fairel Rod Dflygs l DhiLCinTA To dernonstrate that the CRDS operater properly over the full range of primary coolant temperatures and pressures fiorn ambient to operating, and particularly that thermal expansion of core components does not bind or significantly slow control rod movements. Also, to determine the initial operating characteristics of the entire CRDS.

Descriptis The CRD tests performed during the startup are designed as an txtension of the preoperational CRDS tests.

Ar,v.rptance Criteria

.LGl.1 Each CRD must have a normsl withdraw speed less than or equal to 3.6 in/sec (9.14 cm/sec), indicated 1:y a full 12.ft stroke in greater than or equal to 40 sec.

Upon scramming, the average of the insertion times of all operable control rodr., exclusive of circu'.t response times, must be no greater than:

Percent insertion Time Inserted (sec) 5 0375 Scram time is measured 20 0.90 from time pilot scram 50 2.0 valve solenoids are 90 5.0 deenergized.

The average of the scram insertion times for the three fastest control rods of all groups of four control rods in a two-by-two array shall be no greater than:

Percent insertion Time Inserted (sec) 5 0398 Scram time is measured 20 0.950 from time pilot scram 50 2.120 valve solenoids are 90 5300 deenergiicd.

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Each CRD must have a normal insut or withdraw speed of 3,01 0.6 in/sce (7.62 i 1.52 cm/sec), indicated by a full 12 ft stroke in 40 to 60 see.

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- Scram timing (at rated pressure) was performed on all CRDs as per ST 10.7, completed on December 15,1991. Scram timing of selected CRDs was completed on January 14,1992.

In cold shutdown, each CRD was tested for position indication, normal insert /v.ithdrawal times and coupling via ST 10.8 (Control Rod Withdrawal Tests) r.ompleted on December 30,1991.-

At rated reactar pressure, Position Indication (GP-2 Normal Plant Startup), and Coepting Checks (ST 10.8-1 CRD Coupling Integrity Test) were performed and completed on January 16, IW2. The testing performed sati.sfied Level 1 and 2 criteria.

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To whieve irdtiAl frIliC(lity in 8 fi41C and Pflhient mannct ming the hlattup I withdt . v wquetac. Also to determine e

the cifcet on reactor pmer of mntrol tod motion at varica operating conditions.

Desttiriks -

The start up 1 control rod withdraw sequence hn been cakulated which completely sperily control rist withdraws hom Ibc all rodi.in rondition to the anled power configuration. The sequence will be uwd to attain ci id criticality Crilical rtxt putkrus will be recorded perhxtically as the reactor is heated to rated temperature.  ;

Movement of tods in a prescribed wquence is monitored by the rod worth m' rmlict, which will prevent out of- l wquence withdraw. Also, not more than to todt may be inserted out of sequence. ,

lhkuks The sequence was defined in (il' 2-3 Appendis 1 (htartup Rod Withdraw.d Sequence Instructions) ar d wrified for uw by the Rod Woith Minimiter (RWM) sia ST R 62A.220 3 (RWM Sequence leading Verification) on December 12, 16, & 211, lW1. ST-M2A 210 *. RWM Operability Ched) wn perfortned on January 1.1W2 and 5'l-R-0024tu-3 -

(Shutdown Margin) recorded the critical rod pattern on January 2,1W2.

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To dernonstrate that the operational sour 6ea, SRM lintrumentation, and rod withdrawal sequence provMe adequate information to the aperator during $1attup,  !

t DLIlipticit i Source range monitor count rate data will be taken dariru!. rej withdrawals to (ntical and compared with 51ated criteria i on signal count-to noir count ratio.  ;

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LsxcL1 There enuit be a acutron signal count.to noiw tuunt ratio of at least 2 to 10 the required operable SRM's. f Eclulb -

Source Range Monitor (SRM) instrumentation opuability was cheded during the performance of startup proced tre OP 2 The criteria of a rninimum count rate of 3 cour.ts/sce, was verified to be met for the SRM's. Data is documentrd it. GP'2 dated January 20, IW2.  :

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. 2.8 110ifuleimmte Dktim The purpisc of this test is to adjust the IRMS to obtain an optimum userlap with the SRhls and Al'RMS.

&4SIA!!EElfilida JastL1 1:.ach IRM channel mm.t le adjmted so the owelap with the $RM's and Al'RM's is assured. The IRM's met produce a serem at 120 on a full scale of 12$.

Jhalts Intctmed'. ate Rang Monitor (IRM) performance was tested. The IRM r. tram sotpoints met lxvtl I criteria of Sl3N.

NC-lRM A(ll)4CW (Interraediate Range Monitor Channel'N(*lr) Cahbratioa/l'un(ti.+nal Ched) daa rd December 2,4,17,24, and ',1,1992 (for the *N thannti), and December 3,10,17,24, and 31, lin2 (for the "ir chantici).

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! Dhiatim To cabbrate the 1.wal Power R~ngt Wemring (1.PR ht ) System.

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'Ibe 1.PRM (hannels will bc tal.brated to r,att the LPRM readings progettional to tbc neution 11us in the narrow.

narrow water gap at the chamkr eination. Calibration factornill be obtained through the use of a proccu (omputei calcula' ion that relates the Li'RM reading to average fuct aucc6!y gnwer at the (harnber height.

.61Lcrid!1r fillt.th

.Lcull With the reactor in the rod pattern and at the power inct at which the ca'ibration is to be performed, the i eter readingof each ifRM chamber will be pn.rortional to the aserage heat llus in the four adjacent fuel rads at the height ~

of he t chamber.

Reut'n local Power Range Monitor (ifMM) cabbrations were periwmed at W; eated thermal 1.ower or. January 13, Iwa and at itM9 rated theirnal power on January 2X, lW2 per ST R NIA 2M3 (LPRM (lain Calibratior).

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To calibrate the Avretage Power Range Monitor (APRM) channels.

DC1LLil'lICD A heat balance will be made at least once cach thift and af ter each major p>wer level change. Each APRM channel reading will be adjusted to twe consistent with the co'c thermal power as determined from the heat balance, r

Agrctgote Criteria Intl 1 The APRM channc!s must bc calibra.ted to read equal to or greater than the actual core thermal power.

Rentlb l Numerous Average Power Range Monitor (APRM) calibrations were completed during Power Aecmion. Ter,t data is doeurr.ented in ST RWIA 210-3 completed from January 7, Un2 at 6% rated thermal power to January 21, lW2 at  !

1(U% rated thermal power.

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'To verify the performance of the process computer un&r operating conditions. i

.pescriotion Following fuel loading, during plant heatup and the ascension to rated pow (r, the nuclear steam supply sy, tem and the ,

l balance-of plant system process variables sensed by the compr.fer as digital or analog signals will become available.

Verify that the computer i> receiving results of pciformance calculations of the nuclear steam Supply system and the balance of plant are corrett. Wrify proper operation of ell compulcr functions at rated power operating conditions.

ACCtriantdulsda Level 2 Program OD 1 and P-1 will be cortsidered operational w hen the Thermal Limits calculated by an independent method and the process computer are in the same fuel assembly and do not differ in value by more than 10 percent, and when  ;

the Li'RM cahbration factors calculated by the independent rnethod and the process computer agree to within 5 i gv:rcent. ,

lkudh l 4 A manual calculation was performed via RT R-059.90 'l(chedout of the NSS Computer Calculation of Core Thermal Power) at approtimately IW'1 power on January 23,1W2.

t The Thermal Limit calculations were verified by General !!lectric di IlUCKl_E with full-power data provided by the 1-rcwiss Computet, The ill'CLE run was performcd on February 23 Un2 with jilant data obtained on Fchruary 11, ,

1W2 while Unit 3 was operating at steady state full power conditions.

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. 112 ljfKlor Cettimldian Cwlartt.Jhitt!D Didulito To verify the operation of the RCICS at operation reactor pressure conditions.

.[h20.iL iits A controlled statt of the itCICS will be donc at a reactor preuure of IM) psig and a quid start will be done at a teactor peruure of 1,nK) osig. Verify proper operation of the ROICS and determine time to reach rat's flow. These tests may be performed with the systern in the test mode so that discharge flow will not be routed to the reactor pressure ve ucl.

AJutlantLClitniit

.Lurl.A The RCICS must have the capability la deliver ratcd flow, ru) gpm,in leu than or equal to the rated actua' ion time, 30 sec, against rated reactor prcuure.

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.llnVJu A controlled start was performed at IN) psig via ST 10.' (RCIC 110w Rate at 150 psig) on .lanuary 4, lW2; Although RCIC was unable to meet the required flow of NU ypm at IDI pdd over reactor preuure, it did meet that flow rate requirement at N) psid over reactor preuure. A IKTR50.59 review was pc rformed by Nuclear Engineering and RCIC was declared operable by PORC 924)2 on January 4, IW2. A Ouick start at rated pressure was perforrned via ST d

6.11F.3 (RCIC Pump, Valve,110w and Coolcr) on January 7,1W2.

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l Disttuto To strify the proper operation of the llPCis throughout the range of reactor preuure conditions.

llettillinil Controlled starts of the llPCIS will be done at reactor preuurn near INI, and I,ml psig during the heatup phase, and a quick start will be initiated during Phase 3. Verify proper operation of the lil'CIS, determine time to reath rated now, adjust flow (ontroller in lil'CIS for pron r now rate and adjust osersp ed trip of IIPCI turbine. These tem will be performed with the estem in the test imide so that diuharge Dow will not be routed to the scactor preuure veucl.

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Tbc time from actuating signal to required now must be leu than 25 sec with reactor preuure at 1,(nxi psig. With pump diwharge preuure at 1,220 psig the flow should be at least 5,(m r pm. The llPCI turbine must not trip oil

• during startup.

Entilb A controlled start was performed at 150 psig sia ST 10.13 (Ill'Cl llow Rate at 191 psig) on January 3, IW2, A quick start at rated pressure was performed via ST 6.51' 3 (111'C1 Pump, Vahe, l~ low and Cooler) on Jan ry 7, I w 2.  %

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. 2.14 fnte PowrLDistribution and_Tl} IUD.tcliainly 1 Dhitethtb To (1) confirm the reptmlucibility of the TIPS readings,(2) determine the core gewer dishibution in three dimensions, ano (3) determine core powr symmelt).

DmIirliOD 1he check is made with the plant at sicady. state condition by producing several TIP traces in the same !xation with each TIP machine. The traces are evaluated to determine the extent of deviations between trates from the same TIP

, machinc.

Core power didribution including power symmetry will be obtained during the power ascension program. Axi.el;wer >

traces will be obtained at cach of the TIP locations. Several TIPS's have been provided to obtain these traces. A cominun location can be traversed by each TIP chamber to permit intercalibration. The results of the complete set cf TIP traces will be evaluated to determine core power r.ymmetry.

Accentance Criteria ,

ic1d.2 i in the TIP reproducibility test, the TIP traces i, hall be reptmlucible within i 3.5 percent relative error or 10.1$ in (3.8 i mm) absolute error at each axial po itkm, whichevet i: greater.

Rcsults RE-27 (Core Power Symmetry and TIP P.cprmlucibility Test) was performed at 100% rated u.re thermal powc to demonstrate TIP reproducitility and core symmetry. The TIP readings were within the standard deviation med Ic. -

establish safety limit criteria of M'?o, per General Electric Document N!!Dil 2401) Table SSI. The maximum deviation between symmetrically imwied pairr, rea:ishcd the 259; acceptance criteria for sorv power symmetry.

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DMtctives  !

1 To evaluate the rote performance parameters of the torc flow rate, core tl.crmal pimer levci, hlatimum Fraction of Limiting Critical Power Ratio, matimum fuel rod 1.inear lleat Generation Rate (LilGR), and hiasimum Ascrage Planar IJncar lleat Ocnciation Itate (APLiluf t).

Dunintien ,

i Core power level, recirculation flow, fuct assemt.ly power, htFCP, hill PD, and h1 APit will Iw determined. Plant and  !

in-core instrumentation, and conventional heat balance rechniques will be used. This will be performed '

above 25 percent power and at varine pumping conditions.

Accstimitt.Criteriit lxycl1 Steady state reactor power shall be limited to 3,293 MWt.

hiaximum Fraction of Limiting Critical Power Ratio (hiFCP) shall not exceed 1.to during $ttady-$ tate conditions when  !

cvaluated at the operating power level, hiasimum Fraction of Limiting Power Dendly (hirl.PD) shs11 not cuecd 1.(x) hiatimum Average Planar Ratio (h!APR) shal; not exceed 1.(kl.

.Bnulla .

The core thermal limits were verified daily above 25% power via the Prwess Computer. ST-R-m2aO 3 (8teacmr Anomatics) verified the full Power control Rod Pattern provided by the PI:Co Fuel hinnagement Section was completed on January 2A February 18, and htarch 16,1W2.

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,, 2'1,6 fcsslutgLSattal Objecthes To 'demonstrate acceptable reactor water level control, evaluate and adjust feedwater controls, and demonstrate capability of automatic flow ruoback feature to prevent low water level scram following trip of one fct d, sier pump.

Dcstriptinn Reactor water level r,ct point changes of approximatelyi 6 inches will be used to evaluate and acceptably adjust the feedwater control system settings for all power and feedwater pump modes.

One of the three operating fredwater pumps will be tripped at 75 percent and full power while the automatic flow runback circuit acts to drop power to within the capacity of the remaining two pumps.

- Various additional test icquired by the rmxlification to the Feedwater level Control System (Mod 180) are described in the results section.

ActsetancsSJtitcIla Leul1 The decay ratio must be less than 1.0 for each process sariable that ethibits oscillatory response to feedwater system changes.

Incl 2 The decay ratio is crpected to be leu than or equal to 0.25 for each process variable that exhibits meillatory respom.c to fesdwater system set point changes when the plant is operating above the lower limit of the master flow contioller.

The automatic flow runback feature will prewnt a scram from low water level following a trip of one feedwater purnp.

Rcaha Feedwater Controller Stability testing was performed to demonstrate acecptable reactor water level contro.. The response of each Reactor Feedpurnp to changes to the Master level controller of positive and negative 3 and 6 inch level changes was observed.

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. ST.041211250 3 Reactor Water level Instrument Pesturbation Test Completed Sat.1/01/92, hlAT 1843 F demonstrated the operability of the start Up flypau *C" Feedpump Discharge liypau level Controller to Maintain Reac;or level, completed Sat.1/06/92, and 1/07/92.

htA1 184311 demonstrated acceptable irsponse to small signal step changes to the Alasic lxvel Controller, complcted Sat. on 1/22/92.

htAT 1843 K demonstrated satisfactory response of the Feedwater Level Control hystem during the trip of one Condensate Pur p with Reactor Power greater than M191, tompleted Sat on 1/20/92, and 2/07/92.

htAT 1843 J proved the satisfactory response of the Feedwater level Control Splem during the trip of one Ecartor feed pump with Reactor Power between 70 to 7551 and then again with Reactor Power greater than MFI, completed ,

Sat on 1/29/92.

hlAT 1843 L verified the ability of the feedwater level Control Sptem to Control Reactor 1xvel following a Hedre <

Pump trip from 95% power and 1(U% flow, wmpleted Sat on 1/21/92.

htAT 1843 ht tested the response of the Feedwater lxtel Control Sptem to a loss of its 120 VAC feeds, completed Sat on 12/28/91.

htAT 1843 P demonstrated that the Feedwater Lesel Control System can mntrol Reactor level within + / (LO inches ,

of the desired setpoint when increasing and decreasing reactor power at a rate of 10 htWe/ min, completed Sat on 1/19/92.

hlAT 1843 O demonstrates the Fault Tolerance of the Feedwater lxtel Control Systent, performed, but failed due to Reactor 1.evel dropping more than anticipated. This problem is being worked on by Engineering and the Vendor, and is not expected to I e resolved by the time this report is issued.

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• 2,17 ErlitLhba Di'itulE3 To wrify the proiwr o;wration of the dual purpose safety rctici saht s, to determine thcir capacit,, and to scrify their leak tightness following opcration Dnnittien The main . steam relicf valves will each be opened manually so that at any time only one is open. Capatity of each relief valve will be determined by the amount the bypan er control sahes chise to maintain reactor prenure. Proper rescating of each relief valve will be verified by observation of temperatures in the relici valve disci'arg: riP ing.

AEn%1nce Cli1Mia level 2 IIach relief valve is expected to have a capacity of at least NKunnilb/hr at a preuure setting of 1,tw) pag. Relici valve leakage must be low enough that the temperature measured by the thermocouples in the diuharge r.ide of the valves falh to within MT of the temperature recorded before the valve was opened, liach sahe rnust mose from fully r.losed to fully opened in 0.3 sec.

Ihlu!!$

liach Safety Itclicf Valve (SRV) was manually cycled at IM pdg Reactor l'rcuure. Test data is documented in ST OlC 440-3 (Relief Valve hianual Actuation) dated January 3, and January 6, Un2.

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, 2 18 Rgdr.pdadon flow Centrol system ]

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.011erdna j

To & .nine the plant req =use to chaeges in recirculation uuw, and to set the NG and w.4,11ow limiters. ,

.DDr'*' 40 Various process variables will be recorded while changes are introduced into the ree...ulation flow control system.

Acccptance C*ittria

.luel1 ,

The decay ratio must be less than 1.0 for each process tariabic that enhibits oscillatory response to Dow control changes.

1&YII 2 The decay ratio is expected to be less than or equal to 0.25 for cach process variable that exhibits oscillatory response to Dow changes when the plant is operating above the lower limit setting of the n. aster now controller.

?  :

EEllllll Various tests were performed as a result of the Digital Feedwater Control System that was installed under Mod 1843, ,

see Section 2.16 Feedwater System for a list of those tests. The Recirc pump NM speed !; miters were set on January 1,1W2 (ItT 3.12). The Mr~v speed limiters were set as per RT 3.13 on January 17,1W2. The liigh Speed Stops were s,et as per RT 3.14 on January 28,1W2. ,

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,Rrtirculation Sag Dblecthts  !

To calibrate the jet pump flow instrumentation.

{

v Descrir, tion j The jet pump instrumentation will be calsbrated to read total core flow.

Accerdance Criteria i

Ind.1 None - ,

1 1s.rd.2  ;

i flow 1.~.strumentation has been calibrated such that the reactor jet pump total flow recorder provides correct flow indication.

IMults l Jet Pump Operability was checked during the performance of partup Psoredure GP 2, and documented in ST 0 02F-

$50-3 (Jet Pump Operability). - Jet pump calibration was Scrified at 100% Core Thermal Power in RT.it 02F.250-3 (Core Flow Calibration Verification U/3) dated .19uary 27,1W2.

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