ML20086C614

Cycle 6 Final Power Ascension Test Rept,910220-0806

ML20086C614
Person / Time
Site:	Browns Ferry
Issue date:	10/30/1991
From:	TENNESSEE VALLEY AUTHORITY
To:
Shared Package
ML20086C612	List: ML20086C608 ML20086C614
References
NUDOCS 9111250022
Download: ML20086C614 (33)

Text

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Browns Ferry Unit 2 Cycle 6 Final Power Ascension Test Report 20 February 1991 Through 06 August 1991 l

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$$$f l.,?,bli'_ c, e cm.

l yyggg SectipD BE9 Startup Chronology 2

Itser Ascension Histcgram 5

PATP 'Ibst Status 6

Test Deficiency Sumary 8

Response to IBC inspection Report 50-2r./91.'6 14 Conclusion 16 APFC dix A.

Descriptio" Gf Ptuer Ascension Testirg 17 B.

Sumcary of Open Test Deficiencies 25 C.

Surstary of Results for FEAR Section 26 13.10 'Ibstiry Ebge 1 of 32

1 BwI_Apomsbn ibst Prmram startup _QuT>nolcrty

'Ihe Ituer Ascension Test Program was divided into three test phases as defined in the Brwns Ferry IT2AR Clapter 13.10, Refueling 7bst Program.

This 01ronology details the testiry ard mjor plant activities that occurrtd durire the irdicated test phase.

Hyg.e I (Open Vessel)

Fuel Imd (2-TI-147A) commenccd on 20 February and completed on 6 March 1991.

Final core loading verification was canpleted on 7 Ruth.

Reactor vessel assembly started on 8 Ruch and the Drywell Head was set and tensioned in order to support the Integrated Inak T1te 'Ibst on 14 R1rrh. After the IUTI' was completed on 19 March the D ywell llead was removed to support the fiml Reactor liead tensionirn, insulation and pipire installation which was cmpleted on 4 April.

The Reactor Vessel liYDRO commenced on 6 April and finished successfully on 9 April. During the flydro, open loop testify of the Recirrulation Flow control system (2 'I'I-132) was performed to check the responsiveness of each Recirculation loop. Final assambly of the reactor was completcd on 15 April when the reactor cavity chield blocks were installed.

Control Rod Drive friction testing (TI-20) started on 4 May and completed on 17 May. The Feedwater startup level control valve was testcd on 14 May, per 2-TI-131, to assess its performnce before foodwater was aligned to the vessel.

Initial (Ambient)

'Ih e r m l D:pansion walkdown (2-TI-190) began on 14 May and completed on 18 May.

PIT. cess comp 2ter data base verification (0-TI-135) started on 15 May ard finished on 15 May. 'Ihe reactor vessel level referenoe legs ard cordensity chambers were insprted, por 2-TI-149, in conjunction with the drywell therml exynnsion walkdwns on 21 May.

The completion of 2-TI-149 mrkcd the end of open vessel testirg for the PATP.

Ihase II (Heakitp to 55% Reactor Pruer)

Followirg completion of all prerequisites, plant startup commenced at 14:58 on 23 May 1991.

Initial criticality was achieved at 00:25 on 24 1

Riy and a verification of plant rhutdwn mart 3 n (2-TI-115) was performcd.

Follwiro che last traini.9 critical, a brief mintenance outage was taken to Irpair the Containment ifydIDgen analyzer ard to complete alignment of the Safety Relief Valve Acoustic Monitor.

Page 2 of 32 1

1

, Reactor Startup comnenced at 22:15 on 01 Jurts. At 01:30 on 2 June startup was tenporarily haltM due to a reactor water level excursion (to 50 inches) that occurred when 2-KV-3-53 (Startup Invel Control Valve) was unisolated in preparation for goirs into larg cycle cleanup.

Startup rocamnenood at 0649 on 3 June ard heatup continued to 150 psig for therm 1 expansion walkdcuns (2-TI-190). Eermi expansion walkdcuns were completed, for 150 psig, on 5 June.

'It.2 plant reached an all rods in cordition at 18:53 on 5 June due to an incident that resulted in the loss of primry containnent. A decision was mde tc, replace the 2B Recirrulation Punp seal prior to restart.

After seal replacraent, the plant restarted and was critical at 12:24 on 12 June 1eatup continued to 250 poig for continuation of therm 1 expansion vu!.dcuns and SRV testing.

In order to resolve therm 1 expansion interferences identified during the walkdcuns, the plant reduccd pressure to approximtely 50 poig on 17 June. On 20 June with the therm 1 expansion interferences resolved, the plant increascd pressure to approximtely 250 poig in order to por Nm the HPCI operability surveillance.

21s was required due to mintenance wor't that had been performed on the 73-16 valve which was previously leaking through ard mintaining the turbine casing hot, ard to ccanplete SRV testirg.

On 21 June reactor pressure was increased to 890 psig to perfom the rated therm 1 expansicn walkdowns. On 22 June reactor pressure was increased to 920 poig to perfonn RCIC (2-TI-188) ard HFCI (2-TI-189) tunirg ard surveillances. AT 02:28 on 24 June a shutdcun ccxmenced due to inability to establich drywell to torus pressure differential, me plant was critical at 04:02 on 25 June with heatup to 920 psig in prtgress.

At 05:00 on 26 June at approximtely 16% therml pcuer, in RUN Mode, DiC testirg (2-TI-130) ccumenced followed by Main Wrbine roll ard testing.

AT 20:41 on 27 June the generator was synctuunized to the grid and loaded to abcut 78 MWe.

At 04:52 on 28 June power was increased to about 160 MWe for generator soakirg. We decision was mde to bring the plant t.o hot stardby behird the MSIV's for turbine balancing..

At 13:20 the Mode Switch was placcd in Startup/ Hot Standby ard turbine balancity conuncxd.

While in hot standby, RCIC was placed in service in the CST to CST mode to control reactor pressure.

After approximtely 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> of operation, a RHR punp was started for surveillance testing. Wrus water tenperature started increasirg. At 0248 on 29 June a mnual Scram was inserted because the Suppression Ptol hmperature exceeded 110 degrees.

'Ihe turbine balance was conple:.ted ard the reactor taken critical again at 03:23 on 30 June.

At 08:10 on 1 July the plant was back on-line at about 165 MWe. On 4 July reactor pcNer was increased to about 35% to support CRD scram time testing (TI-20), IMM/AIPM adjustments, Process Computer testing (0-TI-135) awl Feedwater pump testirg (2-TI-131).

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Ihase III (55% to 100% Reactor Power)

Following roccipt of NRC and plant Maragesent authorization to exceed 55% power on 16 July, pwer was increased to the 80% plateau in preparation for further Ixwer ascension testing.

On 18 July, the precoMitioning soak was completed durirg which tire 2-TI-137, Core lower Distribution was performed.

Ibser was then reduced to approxim tely 55% for perfomance of 2 TI-131, Feodwater Master IcVel Controller Tuning.

Ptwar was then raised to 62% for performnce of 2-TI-130, DIC Tunirg.

2-TI-132, Recirculation Flow Control System Tunirg was perfomed at 40% speed (40% pwer), 55% cpeed (50% power) and 75% speed (60% power).

With the satisfactory completion of testirg at the 80% plateau, pwer was increased to greater than 90% to be within the 100% power test window. 'Ihe plant exceeded 90% reactor power at 2250 hours0.026 days <br />0.625 hours <br />0.00372 weeks <br />8.56125e-4 months <br /> on 25 July. On 27 July the preconditioning ranp to approximtely 95%

pwer and subsequent 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> soak were completed.

Mter ;xtrformnce of 2-TI-137, Core Performance, 2-TI-174, Recitrulation loop Flow Calibration, 2-TI-82, Drywell Atmosphere Cooling System, ard 2-TI-149, Reactor Water Invel Measurcrents, 2-TI-131, Teodwater Ibster invel Controller Tuning ard 2-TI-130, DiC

%2nirg were performcd to demonstrate stability ard response near rated pcuer corditions.

2-tie 132, Recirculation Pump Rurback to 65% pump speed was performed in order to evaluate plant response in preparation for 2-TI-191, Food Punp Trip.

'Ihe Feed Pump Trip (2-TI-191) was performed at 94% power ard did not result in a Recirculation runback due to the fact that the remainity two (2) Feed Punpo were capable of maintainirg water level.

At approximtely 90% power, 2-TI-188, RCIC Vessel Injection, was perforned. Power was reduced to approximtely 72% and 2-TI-189, 1HCI Vessel Injection was performed. On 2 August power was reduced to approximtely 42% in preparation for 2-TI-193, Main Turbine Trip, and 2-TI-180, Backup Contrul Panel Testing. As

expected, 2-TI-193, Main nutine Trip, resulted in a Reactor Scram.

On 4 August at 1900 hours0.022 days <br />0.528 hours <br />0.00314 weeks <br />7.2295e-4 months <br /> the plant restarted. On 5 August 2-TI-188, RCIC ard 2-TI-189, HPCI 150 psig Operability Dte nstrations were performcd.

'Ihis concluded all testirg defined within the scope of the Power Ascension Test Program.

A final Special Test, ST 91-01, Tuttine/ Generator 'Ibrsional Test -was perfomed at approximtely 20%

power.

At the conclusion of the Special 'Ibst all testing requirements for the restart of Brtuns Ferry Unit 2 were completed.

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l 1%TP_Tes_t Status Ebtes J1G

'Ibtal Open Erocedure PrTrrdure Titig Perfomcd Arvroval 192n 113 0-TI-20 Control Rod Drive 04 May thnt 09 Aug 1991 16 0

System Testiry 07 Jul 1991 2-TI-82 Drywell Atrospheric 21 Jun thn1 02 Aug 1991 2

0 Coolirg System 26 Jul 1991 2-TI-115 Full Core Shutdown 23 May thru 30 May 1991 1

0 ManJ n 24 May 1991 i

2-TI-130 Main Steam Pressure 26 Jun thru 09 Aug 1991 1

0 Control 28 Jul 1991 2-TI-131 Fcetaater level 14 May thn1 22 Aug 1991 4

0 Control System 27 Jul 1991 2-TI-132 Recirculation Flow 09 Apr thn1 15 Aug 1991 2

0 Control 29 Jul 1991 0-TI-135 Process Camputer 15 May thn1 15 Aug 1991 1

0 and Core 28 Jul 1991 Perior:ance 2-TI-136 APRM Calibration 03 Jun thru 12 Jul 1991 0

0 25 Jun 1991 0-TI-137 Core Power 18 Jul thnt 09 Aug 1991 2

0 Distribution 24 Jul 1991 2-TI-147A Fuel loading After 20 Feb thnt 18 Ipr 1991 0

0 A Complete Core 06 Mar 1991 Unlmd 2-TI-149 Reactor Water Irvel 21 May thru 15 Aug 1991 6

0 Measurements 25 Jun 1991 2-TI-174 Recirculation 26 Jul thru 22 Aug 1991 10 0

System Flow 28 Jul 1991 Calibration 2-TI-180 Bac} up Control 02 Aug 1991 09 Aug 1991 1

0 Panel TestirrJ 2-TI-183 Reactor Water 22 Jun 1991 27 Jun 1991 0

0 Cleanup System Page 6 of 32 l

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Dates J7G

'Ibtal Open Prrcedure PrTodare Title Perforrrd Amruval M

M 2-TI-186 Control Ro:1 Drive 02 Aug 1991 09 Aug 1991 0

0 System 2 4I-188 Reactor Core 22 Jun thru 15 Aug 1991 0

0 Isolation Cooling 05 Atg 1991 System 2-TI-189 High Pressure 23 Jun thru 15 Aug 1991 4

0 coolant Injection 05 Aug 1991 System 2-TI-190 System 'Ihcrmal 14 May thru 22 Aug 1991 8

2 Dqnnsion (Partial) 22 Jun 1991 2-TI-191 Feedwater Pump Trip 29 Jul 1991 13 Aug 1991 1

0 Testirg 2-TI-193

'Iurbine Trip 02 Aug 1991 09 Atg 1991 0

0 Hir-201 Piping Vibration 08 lbr thru N/A 0

0 Qualification 01 Aug 1991 Testing Page 7 of 32

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__--_m-Power Ascensten Program Unit 2 Cycle 6 Test Deficiency Sutenary TE$1 PROC # 108 f est Plateau TE$f EXCEPflDN DESCRIPilDN CWEN/ CLOSED REMARK $

i 0 11 020 01 open vesset CR0 58 23 withdraw tine outside level 2 Closed timing valve adjusted and rod criteria satisfactority retested 5/12 1

02 Open vessst CRD $819 inssrt tine outside level 2 Closed timing valve adjusted and rod criteria satisfactority retested 5/1b 03 open vessel CRD 46-51 f alted the f riction test criteria Closed Control rod stroked and retested satisf actority on 5/14 04 open vesset CRD 46 39 f alted the f riction test celteria Clo6ed Control rod was stroked and retested satisf actorily 5/14 05 Open vessel CRD 42 33 futt core display "ones" digit did Closed WR ccZr466 was written and retest not indicate "0" for positions 00, 10, 20, performed satisfactorily 5/16 30,& 40.

06 Open Vessel CRD 22 55 futt core display "ones" digit did Closed Light buld replaced under 91-31995 00 not indicate "0" or "3".

No green and retested satisf actority 5/15.

background at full in overtravet.

07 Open vessel CRD 14 $1 and 06 39 display "ones" digit did Closed WR C070762 generated and display not indicatt *0".

retested 5/16 08 Open vesset CR018 35 was dif ficult to time and had Closed WO 91 32030-02 repaired disec<rected erratic D/P trace, power lead and rod retested satisfactority 5/17.

09 Open Vesset CRD 10 47 did hot weet the acceptance Closed control rod was stroked and retested criteria for friction testing.

satisf actority 5/13.

- 10 Oper vessel CRD 10 31 has no RPl$ indication on either Closed WR C04475? was written and the rods futi core or 4 rod display at rod positions retested satisfactority 5/16 05,15,25,35, or 45.

11 Open Vessel CRD 26 31 falted the criteria for friction

. Closed Rod will be replaced next outage testing and settle testing.

(WR CO23553) 12 Open vesset Unable to obtain friction traces for rod Closed Wo 91 33563 was written to replace the j

58-31 due to fatture of vent valve 2 85 614.

vent volve during the neat outage.

13.Open Vessel Did not get the "0" in the tens digit field Closed Lightbulb was replaced under when in position 00 through 09 for rod Wo 91 33563 00 and retested on 5/15/91 30 03.

Open Vessel Did not get the green background when rod closed Wo 91-31966 00 38 51 was in the fuit in overtravel position.

Page 8 of 32

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1t51 EROC 81D# Test Plateau f(Si EXCEPTION DEstRIP110N OPEN/ CLOSED RLMARKS

• 0 11 020 13 bren Vessel Old not get the green background when rod Closed wo 9131%8 00 46 19 was in the fv!! in overtravel position.

Oren vessel Did not get the green teckgrourd when rod Closed WO 91-319 @ 00 46-23 was in the full in overtravel position.

14 0 $5%

Flange ball check valve (k>es not move Closed WR C 022775 e trg scram 15 0 55%

only received 23 notches during scram Closed WR C 026264 16 0 - 55%

Several LPRH's were in BYPASS when this Closed in e;,ch case, there was only one I.fRM in procedure was perforred and could not be a string which was in BYFASS. This response checked as the control rods were allewed evaluating the Bypassed LFRM Ly

moved, monitoring the other LPRM's in that string.

0 fi 135 01 0 $$%

00 8 and panel 9 14 LPRM indicators do not Closed WR C064852 agree within 3 units 2 fi 082 01 0 - 55%

All drywell cooling f ans were in service Closed This data was preliminary only. Test when preliminary data was collected.

was reperformed at higher power with Procedure req; ires two f ans to be the correct drywell cooling tircup.

secured.

02 0 - 55%

the 00 3 obtaired for this test was Closed this data was preliminary only. Test inacurate d>e to the low power level when was reperf ors;rd at higher power and test performed. Power was estimated using values for 00 3 were satisf.ctory furbine Bypess valve capacity 2 11 115 01 0 55'4 Documents that reactor period was calculated Closed This TD documentsd that SRM courits were f rom SRM recorders vice IRM'S.

used to calculate reactor period vice IRM's dLrs to proximity to SRM sc ram setpoint when criticality was achieved.

2 11 130 01 0 - 55%

unable to evaluate system response using (HC Closed harrow Range Reactor pressure Was used sensed pressure due to problem with test to evaluate system performance equiprent.

2 11 131 01 oren vesset 01d not reach flows as stated in section 7.1 Closed ilow values are used to be consistent of procedure, with existing cond. demin, lineup and not to be a specific requirement of the proc edur e.

A note will be added to the procedure to clarify this.

Page 9 of 32

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TEST FRCC # TD# 1est Plateau itsi (KCEPf!DN DESCRIPTION OrtN/ClosID REMARLS a

• 2 11 131 02 0 55%

Reactor feedtup turbines did not meet their Closed Reactor f eedptros were adjusted ard the level 2 acceptance criteria for system test was reperforw d.

1D 04 docuwnts performance the results.

03 55 100%

unable to retrieve data from Transient Closed Affected section of test was Recorcer due to a software probten reperf orned 04 55 - 100%

Reactnr feedptsp perfornonce does not neet closed l'erfornerce of overall f eedwater sntem tevel 2 criteria.

was evaluated ty San Jose Engineering as satisf actory 2 11 132-01 Open vessel the *Aa Recirc MG is not as responsive as Closed the system was adjusted and retested in the "B" Recire MG master nenual. Iloth RRMG rerforned satisfactorty.

02 Open vessel there is noise on the output of teth Rectre Closed the reise has not teen otmerved af ter MG speed ccotrollers peintenarce was perf orned.

2 11 137 ' 01 55 100%

tip channels 41 and % and 43 and 38 were Closed CAga BfP 910176 electrically reversed to the process conput er.

02 $$ 100%

f ailed to obtain all conputer edits that Closed These edits were not required in order were required by the procedJre due to to at4 port analysis.

operator Inistrderstarding 2 TI-149 01 0 $$1 Narrow Range urconvensated Levet did not closed f ollowtp data taken and levels were agree within 3 inches.

within 3 itch tolerance.

02 0 - $5%

Initlet set of date taken at 700 psig did closed Evaluation of plant corditions irdicated ret agree with 10 inches of calculated that level was not steady when the data level.

was obtained. f ollowup data was taken ard net acceptance criterla, 03 0 - 55%

initial set of data taken at 10 15% did Closed Evaluation of plant corditions indicated not agree with 10 inches of calculated that actual pressure was 923 psig vice level.

1010 psig which was assuned for the calculated power. When this was used level agreement was satisfactory.

04 0 55%

tenperatures taken on water level conden6ing C!osed Data has been evaluated as acceptable by pots did not neet the level 2 acceptarce Nuclear t.ngineering.

criteria.

05 0 - 55%

Initial set of data taken at 20 25% did Clot.ed Evaluation of plant corditions irdicated not agree with 10 inches of calculated that actual pressure was 923 psig vice level.

1010 psig which was assumed for the calculated power. When this was used level agreement was satisfactory, j

Page 10 of 32

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Test fact 0 108 Test Pl.; eau Itti EXCEP110N DESCRIPIlON CPEN/ CLOSED dtE @kK$

e 2 T!*149 06 $$ 100%

Recirc flow conparisons were not performed Closed Data was retrieved from various power We to problems with the calibration of Core re&ctions ord the last startup which riow Instrwentations allowed conpletion of this test.

2 11 174 01 55 100%

Indicated Core Flow does not agree with closed Anplifier gain was adjusted per WR C caicutated Core fIow.

023600 and i! 174 reperior= d.

10 03

&cwents subsequent f at ture.

02 55 - 100%

Rectre fergerature on TR 68 2 did not agree Closed Problem corrected ty W C 0920$3.

with the value found in (0 3 option 2.

(also see 1D 08) 03 55 100%

Indicated Core Flow does not agree with Closed Anplifier gain was adjusted ter WO calculated Core flow.

9136935 00 and 11 174 reperformed, TD 05 decwents stbsequent f atture.

04 55 100%

Cain Adjustmmt f actor (GAF) for the loop Closed Applifier gain was adjusted per Wo flow proprtional amplyf ter does not meet 9156935-00 and 11 174 reperf ormed, to the acceptance criteria recpired for in this 06 doewents stbsequent Failure.

procedure.

05 55 100%

Indicated Cora Flow does not agree with closed Amplifier gain was adjusted ter W calculated Core flow.

9136935 00 and TI 174 reperformed.10 09 was written to dacwent subsequent f ailure.

06 55 100%

GAFs for the toop flow prepartional closed Anplifier gain was adjusted r.er Wo w ptyffer do not agree within the acceptance 9136935 00 and 11 174 reperformed. 10 criteria of the procedure 10 was generated to document f ailure to meet criteria.

07 55 100%

.let Pumps 11 & 12 aruf 15 & 16 did not pass Closed Problem appears to be We to system the noute plugging criterla reise 08 55 100%

The recirc ptre discharge tenperatures are Closed WR C 092083 written to troubleshoot and ret within 5 degrees of each other.

Investigate 09 55 100%

helther CD+3 core flow nor 2 FR 68-50 agree Closed systems Engineering has reviewed the data within the timits of this tett procedure and determined that the results are within the instrwentation occuracles.

10 55 - 100%

CAfs for the loop flow proportional Closed Systems Engineering han ev$luated the data anplifier does not meet the acceptance and determined that the results are within criteria of this proce &re

<he instrumentation accuracies, Sin were revised with new gain value to ensure mre conservative flow signal.

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2 11-180 1 55 100%

White attenpting to lower flow of RCic in closed Emergency WR written (C -092366) and the

(

automatic roted large swings in system flow.

systm tuned to the vessel following conpletion of the Backup contro! Panet Denons t r at ion.

Page 11 of 32

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i TEST I' ROC # 1D# Test Plateau TE$i EXCEPfl0N DESCRIPfl0N CPE N/ CLOSED REMARKS

'2 11 189-01 0 + $$%

HPCl Turbine Exhaust Pressure was reading Closed Evaluated by NE as acceptable j

higher than expected; 02 0 55%

During the initlet starte of the HPCI Closed surveitlance instructions have turbine frcn cold conditions, a Jack rabbit been modified to monitor system.

start occurred irdicating the reed to perform balancing chanter adjustments.

03 55 100%

HPCI turbine trimed during automatic Closed CAQR BfP 910179 initiation to the vessel due to low suction pr e',sure. Stand Seat tehk Head Gasket also ruptured during resulting transient. (Level 1 violation) 04 55 100%

the Decay Ratio for the Reactor vesset closed GE Engineering has evaluated the system injection at low flows is greater than 0.25 perfortrence during vessel injection ned (0.6).

fird that the controller performance is i

satisfactory.

(

2 11 190

-01 Open vesset several locaticris in the Rx Bultd and Open NE has evolunted each probteam as not -

Drywell had potential interference effecting first therpel Cycle. Points probtens which could ef fect thermal growth, will be re inspected after first cycle.

02 oren vessel Main steam stutter was hard to rotate and Cl osed Srdter was inspected by the system the rear paddle is hard to move.

Engineer and Civil Engineering ord determined that this condition does not affect snubber operability.

03 0 55%

RWCU pipe insulation in contact with floor closed this deficiency was previounty gratint identifled ard is acceptabte as Ione an insulation is not defornal nere than.

1/2 inch 0

$5%.

RWCU pipe insulation in contact with HPCI Closed insulation notched under WO 91 34135 Line 04 0 55%

sucport tug for MSL C is in contact with a Closed Work perforend urrier WP ?O94 91 Dryweti Terperature Bracket

+

05 0 55%

test Connection for HPC) flow etenent is in Closed Work performed under VD 2095 91 contact with s@ port 47B465076 06 0 55%

- Mst B sn@ ber 2 47B400$00097 in contact with closed Work performed urder WO 91 34246 00

- Kickpl_at e.

0 55%

Mst B sn@ter 2 478400s00102 in contact with Closed Work performed urder WP 2096 91 (DCN Drywett floor st ut.

F16883)

Page 12 of 32 i

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Tist* PROC # TDW Test Plateau TEST EXCEPil0N DESCRIPTION CPfW/ CLOSED REMA2KS

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0 2 TI 190 06 0 55%

Mst C sntiter 2 476400s00106 restrained by Closed work performed under WO 91 34248 01 structural fee 07 0 55%

HPCI flow etenent test conrection was in Closed work performed urder WP 2097-91 (DtN contact with soport 47B464$076 F16895) 08 0 55%-

RClc pipe is in contact with penetration Open insulation has been adjusted bowever to sleeve In steam vault f toor ard a hanger adjustments still need to te nude rod on top of torus.

urxler WR Ca?0832.

I 2 11 191 01 55 100%

1he Recire system did not rurbeck when one Closed WR CO23598, WR C043378 reactor feed;wp was tripped at 95% Reactor Power.

Total TD's 59 i

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e Page 13 of 32

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Response to IEC Ir*pection Report 50-260/91-26 Durirn a review of test deficiencies (70s) recorded during performnce of the Technical Instructions (tis) comprising the Ibwer Ascension Test Program (PATP), imC inspectors noted several TDs that required further evaluation.

The follwing describes the results of the evaluations for these 1Ds.

a a.

2-TI-131 Feedwater Irvel Control System The transient response of the fecdpumps to manual flow step changes failed to moet the Invel 2 acceptance criteria. General Electric Ergineeriry has reviewed the data provided for the Reactor Water IcVel Control System and determined that the system is performing satisfactorily.

Although the Reactor Feedpump Turbines are not meeting their design criteria, the overall system perfomance is sufficient to meet plant cperational requirements.

This was demnstratcd by both the Feedpump Trip Test (2-TI-191) and the manual runback demonstration perfomcd as part of the Recirculation Flw Control System Testirg (2-TI-132). This TD is considered closed.

b.

2-TI-174 Recirculation System Flow Calibration (1) 1he calculated core flow failed to met the Invel 2 acceptance criteria when it did not agree within 1.0 Mlb/hr of the core flow indicated by OD-3 and 2-FR-68-50.

Systems Engineering has reviewed the accuracles of the core flow instrument loop and have determined that the differences between core flow indications and the calculated value are acceptable ard are within the accuracy e>prted from the instnrnentation.

This TD is considered closed.

2-TI-174 will be ruvised to include acceptance criteria which more accurately reflect the core flw instrument loop accuracies.

(2) 7he calculated GAFs for the loop proportional amplifiers were not between 0.99 ard 1.01 and therefore failed the invel 2 acceptance criteria.

Systems DvJineering has reviewed the test data ard have determined that the results are within the recirc loop instrumentation accuracies.

Surveillance Instructions (sis) have been revised to include new gain values in order to ensure a mre conservative flow signal.

A work request (WR) has been initiated to perform these sis.

Based on these actions, this 1D is considered closed.

Page 14 of 32

"c.

2-TI-189 liigh Pressure Coolant Injection System (1)

Durirg the initial startup of the IIICI turbine frun a cold con:11 tion, the firCI Turbine stop Valve exhibited rapid opening then closing.

The Balance Chanber pruccure was adjusted in acco111ance with GE SIL #252.

The retest perfortned a cold quick start of the IIICI System in the CST to CST rode of operation ard the valve functiontxi properly.

'Ihis TD is considered closed.

(2)

During vessel injection, the decay ratio for a flow step charge from 3000 gpn to 2500 gpm was 0.60 khich resulttd in a level 2 acceptance criteria failure (< 0.25).

General Electric Ergineering evaluated the data and determined that the IIPCI cystem is performing satisfactorily.

System operation durirg low flow step charges necd only denonstrato stable perfcrmnce (decay ratio less than 1.0). 1herefore, the actual measured decay ratio of 0.60 does not affect the ability of the system to perform its design function. 'Ihin TD is consirlerad closed.

Page 15 of 32

4 Conclusion All testing defincd within the BFN2 Power Arm nsion Test Program (PATP) has been satisfactorily completed and all open test deficiencies reviewed to ensure they will not adversely impact continue 2 power operation. All open items have been assigned trackirg numbers and will be trackod by the Trackhg and Reporting of Open Items

(.TOU system.

This will ensure the closure of the open items as p.),r1 o',rprs tons pennit.

All P;iN 't Wls ' met their performnoe criteria or were dispositioncd sat.infu;f:cfily, with the exception of two Test Deficiencies (1D-01,

08) :sg4r$t 2 TI-190 (System 7hennal I W ion).

The open Test Deficlo)Mim have no irgoct on plant operation aid are derrribed in the 3 tut'rt!x nual en the test results, it is concluaed that the plant is performin;; natisfactorily ard that thezu are no known deficiencien that r:ouM bo detrirrental to continued plant operation.

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APID OIX A Descrintion of Pwer Ascension Testirn ILtJTJ 1his enclosure provides a brief description of each test perfomed durity the Unit 2 Startup.

Routine testig,

such as surveillance procedures, chamistry pmxdures, and mintenance pmcodures have not been listed.

Section 1 lists the testing which was performed (with a brief explanation of the results) during open vessel (prior to Initial Criticality).

Section 2 describes the testirg which was perfomed at the O to 55% power plateau ard Section 3 describes the testirg perfomcd during the 55% to 100% plateau.

1.0 Fhase I Test Results for Open Vessel Test Plateau 1.1 Fuel Imdirn After a Complete Core Unioad 2-TI-147A 1his procedure was used to perform a controlled loading of the complete core follwirq the Unit 2 Cycle 5 outage.

After loading, a visual verification was perfomed and documented on Vidco Tape. This procedure has been completed satisfactorily.

1.2 Pinirn Vibration Oualification Testirn IMP-201 This test provides a qualitative wm=?nt of pipirn vibration for a prescribed system ard plant cordition.

7he followirg systems were evaluated prior to startup:

EECM piping during pump auto start transients Fuel Pool Coolirg pipirg Diesel Starting Air piping No system exhibited unsatisfactory performnce after evaluation of the inspection data.

1.3 Recirculation Fjow Control System 2-TI-132 i

During the Reactor Vessel Hydro, m nual speed stcps to each Recirculation Pump were performed to evaluate system response.

Baseline data was also collected at I

various speeds.

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Page 17 of 32

4 1.4 Control Rod Drive System 'Ibstim 0-TI-020 This test was used to verify proper operation of the Contro3 Rod Drive System.

Prior to startup, Differential

Pressure, Drive F1w, Stall F1 w ard control red speeds were recorded and adjusted as necessary.

Irdication of control Rod position was also verified at this time.

1.5 Feodsator Centrol Svetem 2-TI-131 The operation of the startup level control valve was checked prior to Reactor Startup.

1.6 Themal Dmnsion 2-TI-190 This test consisted of Visual inspections performed on selected systems khich were identified by lluclear Drjineerirg.

Them inspections were performod at three different coMitions durirg the startup.

The first inspection was at ambient coMitions, prior to the initial plant heatup.

A detailed walkd wn of each system was made to ensure no interferrnoe existed khich would prevent the prcdicted novement.

During this walkd wn, baseline readirqs were chtained to monitor system novament.

1.7 Process Comoter 2-TI-135 Prior to startup and follwirg completion of the refuelirg ner.itor update program (OD-20), the nuclear steam supply system data was verified for accuracy and proper location in the computer memory.

1.8 Reactor Water icvel M3asuremnt 2M[-149 Just prior to the initia3 raactor startup, refernnce legs were backfilled ard verified by ultrasonic inspection.

Piping was also checked for adcquate slope.

Page 18 of 32

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2.0 Phase II Test Results for lleatup - 551. Power Est Plateau 2.1 R111 Core Shutdown Rsmin 2-TI-115 This procedure was used to verify that the core could be shutdown under all conditions throughout the current fuel cycle.

This was done by comparing the actual point at which the reactor achieved criticality with that predicted by analysis.

7he actual difference was 0.19%, well within the allowable limit of 1.01,.

2.2 Averace Power Rvne Monitor (APIM 2-TI-136 1his procedure was used to adjust the APitt's prior to the point khere instrumentation (e.g. Feedwater Flow) would allcu the normal mnual heat balance to be perfomed.

Core thermal power was determined either by a constant heatup rate heat balance or by a min turbine bypass valve comparison.

This procedure has been canpleted satisfactorily.

2.3 Reactor Water Invel Measunmein.s 2-TI-149 Data was collected during lleatup to rated pmssure and as power was increased to 50%.

Water level indications were consistent both with predicted values and within the d.ifferent instrument groups. Temperature readings were recorded on the reference legs and condensing chambers ard the data was evaluated by Nuclear Engineering.

2.4 Therml Exnansion 2-TI-190 The secord inspection for thermi expansion was performed at 150 psig reactor pressure (approximately mid temperature).

Specific inspection points which had been selected by engineering were measured to ensure that the actual novement agreed with that predicted by the computer model.

Several problems were identified and corrected before the heatup resumed.

A third inspection was performed with the plant at rated pressure.

Measurements were again cognred to predicted movements and any discrepancies evalustal by i

nuclear Engineering.

Only 2 test deficiencies remin open arxl the are discussed in the open Test Deficiency summary (Appendix B).

Page 19 of 32

2.5 Pinim Vibration Oualification Tertirn RfD-201

'Ihis test provides a qualitative assessnent of pipirq vibration for a prescribed system and plant condition.

me following systems were evaluated during 0 - 55%

Plateau:

Main Steam pipirg at 150 psig ard rated conditions RG pipirg at 150 psig ard rated conditions RCIC piping durirg mnual, hot and cold quick starts.

HPCI pipiJg during mnual, hot ard cold quick starts.

EECW piping durirg pump auto start trcasients Fuel Pool Coollrg pipirg Diesel Startiry Air pipi19 No system exhibited unsatisfactory performance after evaluation of the inspection data.

2.6.

Drywell Atmoschpj:19 Coolirn System 0-TI-082 A prelimirary set of data was collected to evaluate

_;ystem performance.

7bmperatures were within limits.

This test was reperfornr.d at full power conditions.

2.7 Reactor Water Cleanuo System 2-TI-183 This proccdure was used to verify that the check valve in the Reactor Water CJeanup System discharge pipirg was free to move.

This test was performed to neet the requirenents of the Restart Test Pr v am.

This procedure has been completed natisfactorily.

2.8 Bet-tor Core Isolation Coolirn System (RCIC) 2-TI-10.D

'Ihis system was also testcd at rated pressure in both the manual ard automatic modes.

Final controller settirgs have been determined ard a simulated cold injection (usirg the CST test return path) performed.

The time to rated flow was 13.96 seconds, well within the required 30 seconds.

This precedure has been completed satisfactorily.

2.9 Hicth Pressure Coolant Iniection Syster (HPCI) 2-TI-1Pd_

This system has been tested at rated pressure in both the manual and automatic nodes.

Final controller settirgs have teen detemined and a sinulated cold injection (usirg the CST test return path) performed.

This procedure has been cmpleted satisfactorily.

Page 20 of 32 i

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2.10 Pmratre Rocntlator 2 'rI-130 1he Pressure setpoint was varied to produce step changes (up to 6 psig) to monitor system perfomance with the Main Turbine Dypass Valves in contml.

Simulated failurus of individual pressure regulators was also perfomed to verify capability of the remaining regulator te adequately control-plant pressure. This prtcodure has been ocxtpletod satisfactorily.

2.11 Prvoess Comruter and Core Performnce 2 'rI-135 After reachirg approxim tely 201, power, a core therm 1 power calculated by the proocss otraputer was compartd to a detailed mnual heat balance.

The therm 1 limits for minimum critical power ratio (MCPR), maximum average planar heat generation rate (MAPIJIGR), and maxinum linear heat generation rate (IJIGR) were compared to an off-line c m puter system.

2.12 Contml Rod Drive System 7bstim 0-TI-020 After the reactor was at norml operatirq pressure, individual control rods were scramed to verify that their times were within the limits of the plant's technical specifications.

'Ihe connection of the local Power Range Monitors (IIH4) were also verified to be correct.

2.13 Fecdeter Control System 2-TI-131 Manual flow steps veru performod on the system to evaluate Reactor Fee 3 pump 7\\1rbine (RFPI) Performnce.

Initial adjustments were made to the m ster controller based on observed system performrce.

RFPI's were observed to be sluggish, however it appeared that the feodwater control system could be adjusted to compensate for this so no adjustments were made.

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3.0 Phase III Test Results for 55 - 100% Power Test Plateau 3.1 Pipim Vibration Gla_lification Testim IMF-201 We followirg systems were evaluated:

RCIC durirg vessel injection HICI durirg vessel injection No system exhibited unsatisfactory performnce after evaluation of the inspection data.

3.2 Recirculation Flw Control System 2-TI-132 Several small charges to control system output were inserted into the systcm to nonitor both transient and steady state response.

We system reponse was satisfactory.

A manual runback was perfomed to monitor the integratcd control system response to this transient.

All control systems responded as expected.

3.3 Feedwater Control System 2-TI-131 Changes in water level setpoint were inserted into the system to evaluate and optimize system perfornance. We system initially exhibited divergent behavior while coeratirg in single elemnt.

Adjustments were madG to both the Reactor Feedpump Turbine controls and the Master invel Controller.

S e system was retested and although the Reactor Feedpunp performance was still sluggish, the overall system performnce was evaluated as satisfactory by the NSSS Vendor.

3.4 Erocess Corouter and Core Performnoe 2 'rI-133 After reachirg approximtely 90% power, a core them11 power calculated by the process camputer was campared to a detailed manual heat balance.

We thennal limits for minimum critical power ratio (MCPR), maximum average planar heat generation rate (MAPUlGR), and maximum linear heat generation rate (UlGR) were compared to an off-line computer system.

All acceptcnce criteria were met.

3.5 Reactor Water level Measurements 2-TI-149 Data was recorded as power was increased to 100%.

All criteria were met.

3.6 Drywell Atmospheric Coolim System 0-TI-082 System performnce was evaluated as satisfactory at 100%

reactor power with one coolirg fan in each bank secured.

Page 22 of 32

Reaqtoj,' Core Isolation Coolim System (RCIC) 2-TI-188 3.7 o

A cold quick-start vessel injection was perfomed. Time to rated flow was 16.2 seconds. S e system was also tested at 150 psig to confim the new controller settirgs.

3.8 Hitth Pressure Coolant Iniection System (HPCI) 2-TI-189 Durirg the quick start vessel injection, the IHCI turbine tripped on low suction pressure to the IHCI Pump.

Se trip logic was modified to include a time delay and a cold quick-start was perfomed. We system operatcd nomally ard reached rated f1w in 26.9 seconds.

We system was also tested at 150 psig to confirm the new controller settirgs.

3.9 M3in Steam Pressure Control 2-TI-130 The pressure setpoint was varied to produce step charges (up to 6 psig) to monitor system performance with the turbine control valves controllirg pressure, cd both the control valves and bypass valves controlling pressure.

Simulated failures of irdividual pressure regulators was performed to verify the capability of the remainiry Itqulator to adequately control plant pressure.

All acceptance criteria were satisfiM.

3.10 Feedwater Pump Trin 2-TI-191 A sirgle reactor feedpunp was tripped at 94% reactor power.

Se two remainirg feedptmps increased their discharge flows ard restored reactor level before the feedwater runback circuit was actuated.

All control systems perfomed satisfactorily.

3.11

'Iurbine Trio 2-TI-193. CRD System 2-TI-186 A manual trip of the main turbine was performed at a power level high enough to cause a Reactor Scram.

We resultirg transient was ronitored ard all systems performed as expected. m is data has been provided for validation of the operations simulator. Data collected during this test was also used to closecut testirg required by the Restart Test Prupcun (2-TI-186).

2-TI-186 demonstrated the acceptable performance of the SDV ard the SDV instrumentation.

~1.12 Recirculation System Flow Calibration 2-TI-174 mis test was performed several times to provide data for the adjustment of the Recitu11ation Flow instrumentation.

Amplifier gains were adjusted several times ard the results were evaluated erd detemined to be acceptable.

Page 23 of 32

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3.13 Bachio Control Panel 7bstim 2-TI-180 Following the Main Turbine Trip (2-TI-193), control of plant pnssure ar i level was shiftM to the bachip s

control stations.

Safety Relief Valves ard the RCIC t,ystem were operated to lower plant tempnxature 47 degrees.

After satisfactory performance was demomtrated, control was restored to the control roca.

3.14 Core Ptuer Distribution 2 'IT-137 At least two full Traversing Incore Probe (TIP) sets were performd in order to reasure the TIP uncertainty.

Four channels were fourd to have been misconnected.

This was corrected by civging the process camputer software and the test was reperformed satisfactorily.

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APPENDlW 9 4

Power Ascension Program Unit 2 Cycle 6 Somery of Open Test Deficiencies TEST PROC # TD4 Test Plateau TEST EXCEPTION DESCRIP110N OP;:0 V SED REMARKS 2 11 190 01 OPEN VESSEL Several locations in the Rn Build and Open NC has evaluated each probtens es not Drywell had potential interference af fecting first thereat cycle. Points will problens which could af f ect thernet be re-inspected after first thernnt cycle,

growth, 08 0 55%

RCic pipe is in contact with penetration Open insulation has been notched by penentration t

sleeve in Steam Vault floor. Rod Hanger sleeve.

Further work will te rerf ornd is in contact with insulation.

tJnder WR C 030832.

Total:

2 Page 25 of 32 l

APPDIDIX C d-2-TI-147A, Fuel Imdim After a Complete Core Offload BJroose As described in the BEN FSAR Section 13.10.2.1, this test safely ard efficiently loads fuel to full core size.

Acceptance Criteria

'Ihe core must be altered to exactly reflect the final design-configuration while maintainirg subcriticality.

'Ihis criteria =was met.

2-TI-]15.

Rill Core Shutdown Maruin Purpose As described in the BFN FEAR Section 13.10.2.2, this test verifies that the reactor will remain subcritical throughout the cycle with any single control red withdrawn and all other control rods fully inserted.

Acceptance Criteria level 1'-

'Ibe Sad of the fully loaded core must be calculated to remain at least 0.38 percent dK/K, with the analytically determned strongest red fully withdrawn, at any point in the tWrg cycle.

'Ihe actual shutdown mrgin is 1.23 percent dK/K, which is well within the acceptance criteria.

Iavel 2 -

Criticality should occur within 1 1 percent dK of the predictc<3 critical red configaration.

'Ihe difference between the actual and predicted Keff was 0.19 N rcent, which was well within the acceptance criteria.

Page 26 of 32 e

TI-20.

Cc,, m' Rod Drive System Purrxre As describxl in the BFN ISAR Section 13.10.2.3, this test denonstrates that the Control Rod Drive system is operating properly and is capable of meeting its nonnal and emergency functions.

beceptance Criteria Irvel 1 -

1.

Each CRD must have a normal withdraw speed of less than or equal to 3.6 inches per second, iMicated by a full 12-foot stroke time of no less man 40 secoMs.

All CRD withdraw times met this requirement.

2.

The control rtxl scram insertion times must be within the limiting coMitions for operations specified in '1LWical Specifications.

All CRD scram times were within Technical Specifications requilunents.

Irvel 2 -

1.

Each CRD must have a naminal insert or withdraw spxxl of 3.0 i

.6 inches per second, indicated by a full 12-foot stroke time between 40 ard 60 seconds.

All CRD withdraw and insert times mot this requirement.

2.

With respect. to the control rod drive friction tests, if the differential pressure variation excoxis 15 psid for a continuous drive in, a settling test may be perforw d, in which case the differential settling pressure should not be less than 30 psid nor should it vary by more than 10 psid over a full stinke.

3 All CRD friction test results were satisfactory with the exception of CRD 26-31.

Test Deficiency 11 docunented the friction irdication and was evaluated by GE as havirs little or no affect on the operation of the rod.

3.

Proper LPRM connections shall be verified and adjustncnts made to provide proper inputs to the process conputer.

All LPRM connections were verified correct.

Page 27 of 32

0-TI-135.

Process Connuter and Core Performance I

Entree As describcd in the BFN FSAR Section 13.10.2.4, this test verifies the capability of the proocss computer to nonitor plant corditions and to evaluate core performnce parameters.

Acceptance Criteria level 2 -

The prtcess computer pmgrams P1 ard OD-6 are considered operational when:

1.

The location ud value of the MCPR as calculatal by the process computer are in the same At..ation ard within 2 pertent of the MCPR as detemined t.f _iny offline conputer system qualified; or, 2.

If the MCPR, as detennined by the process computer, is in a different location than that determined by any offline computer system qualified, the values calculated for CPR by the two methods shall agree to within 2 percent for each fuel Assembly.

The location ard value of the MCPR as calculatxd by the process computer were in the sa ne location and within -0.35 percent of the MCPR as determined by the qualified backw method.

0-TI-137, Core Iwaer distribution EutT w e As described in the BFU FSAR Section 13.10.2.5, thin test confirns the reproducibility of the TIP system readings, detennnes the core power distribution, ard checks the core power cymmetry.

Acceptance Criteria Invel 2 -

1.

The total TIP uncertainty shall be less than 9.0 percent.

This total TIP uncertainty will be obtained by averagity the total uncertainty for all data sets obtained. A minimum of two data sets is sufficient for determination of the total TIP uncertainty.

The 9.0 percent represents the limiting uncertainty for which the present MCPR safety limit is valid.

If this 9.0 percent, uncertainty is exceeded, a detailed aralysis will be made and possibly additional data sets will be taken.

Ibge 28 of 32

0-TI-137.

Core Power distributioD (continued) o i

Acceptance Criteria (continual)

Invel 2 -

2.

'Ibe gross check of the TIP signal symmetry should yield a maximum deviation between synraetrically located pairs of less than 25 perrent. If this criterion cannot be met, the cause of the asymnetry should be investigated.

'Ihis criteria was not met. TIP Channels 41, 36 and 43, 38 were misconnected (swappod).

Test Deficiency 1 aM CAQR BFP910176 were initiated to document this condition. A software services request (Bm-SSR-91-003) was generated to change the data arrays in the process computer for the TIP index positions.

All criteria were met upon reperfonence of test.

2-TI-136.

APRM Calibration Purpore As described in the BEN FM Section 13.10.2.6, this test calibrates the average power range monitor system.

Acceptance Criteria level 1 -

At least two APRMs in each RPS channel must be calibrated to read greater than or equal to the actual themal power.

'Ihis criteria was satisfied.

Level 2 -

If the level 1 criterion is satisfied, then the APBM channels are considered to be readirg accurately if they do not read more than 7 percent greater than the actual core thermal power.

This criteria was satisfied.

Page 29 of 32

e 2-TI-130.

Mairi Steam Pressure Control purTxre As described in the BFN ISAR Section 13.10.2.7, this test demonstrates:

(a) snooth pressure control durirg transients induced in the Imactor pressure control by the pressure regulators, (b) snooth pressure contrul transition between contrul valves ard bypass valves, and (c) take-over capability of the backup pressure ngulator by establishirg appropriate values for this take-over.

beceptanco criteria level 1 -

'Ibe decay ratio must be less than 1.0 for each process variable that exhibits oscillatory response to pressure regulator charges.

4

'Ihis criteria was met by all variables.

Invel 2 -

1.

In all tests except the simulated failure of the operating pressure regulator, the decay ratio is expected to be 5 0.25 for each process variable that exhibits oscillatory response to pressure regulator changes when the plant is operating above the lower limit of the master flow controller.

'Ihis criteria was met by all variables.

2.

Pressure control deadband, delay, etc.,

shall be snnll enough that steady state limit cycles, if any, shall pruduce variations in turbine steam f1w no larger than those specified in the following table.

Barcent of full Power Perrent of Rated Flow

~

90 - 100 i 0.5 70 - 90 1.5 to 0.5 70 ard below 1.5

'Ihis criteria was met by all variables.

3.

During the simulated failure of the prinuy controlling pressure regulator, peak neutron flux ard peak vessel pressure should remain below scram settings by 7.5 percent and 10 psi respectively.

This criteria was satisfied.

4.

Following each pressure setpoint change (2 to 10 psi), the time between the setpoint charge ard the occurrence of the pressure peak shall be 10 secords or less. 'Ihe peak neutron flux and peak vessel pressure should remain below scram setting by 7.5 perumt ard 10 psi, respectively for pressure setpoint changes 5 5 psid.

This criteria was satisfied.

Page 30 of 32

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4 2-TI-131.

Feedwater Invel Control System

.r <

As described in the-BEN FSAR Section 13.10.2.8,-

this tast denonstrates that the ccrnponents of the feedwater control system satisfactorily control reactor water level.

Accer>tanca Criteria i

Invel l'-

he decay ratio must be less than 1 for each process variable a

i that exhibits oscillatory response _to feedsater system changes.

'Ihis criteria was met by all variables.

Invel 2 -

1.

We decay ratio of an oscillatory control loup rode must be 5 0.25 for each process variable thr3 exhibits oscillatoly i

response to feedwater system charges where the unit is operatirg above the lower limit setting of tha masr.er flow controller,

'Ihis criteria was met by all variables.

l 2.

'Ihe transient response of each feedwater pump to a 10 percent flow demand input change, as measured by the turbine speed ard flow recorder outputs shall be as follows:

a.

Time to 10 percent of demand should be 1.1 secords and must be less than or equal to 2.2 secords.

b.

Time from 10 pcIrent to 90 percent of demand should be 1.0 seconds and must be less than or equal to 2.5 seconds.

i c.

Settling tine to within 5 percent of the final value should be 14 seconds.

'Ihis criteria was not met in all cases for the three feedwater. pumps.

Test Deficiency 4 was written to document this condition.

'Ihe test results have been evaluated by GE ard found to show that the Feedwater Control System functions adequately and is capable of perfoming its interded function.

I Page 31 of 32 i

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2-TI-132.

Recirculation Flow Control l

B1mose As described in the BFN FSAR-Section 13.10.2.9,-

this test demonstrates that the recirculation speed control system can satisfactorily perform _its function by comparing transient test results against system criteria.

Acceptance Criteria invel 1 -

The decay ratio must be less than 1.0 for each process variable that' exhibits oscillatory response to recirculation system changes.

Bis criteria was met by all variables.

Level 2 -

When the unit is operatirg above the lower limit of the master manual limiter, the decay ratio of any oscillatory control loop made must be < 0.25 for each process variable that exhibits oscillatory response to recirullation system changes.

Wis criteria was met by all variables.

2-TI-82, Drywell At2nosphere Coolinct System PurTx:se As described -in the BFN -FSAR Section 13.10.2.10, this test verifies the ability of the drywell atmosphere cooling system to maintain design conditions - in the drywell during operatire conditions.

Acceptance Criteri,a level 2 -

Re drywell cooling system shall maintain the bulk volumetric average temperature in the drywell below design values during normal operation.

'Ibe average drywell temperature was calculated _to be 130.8 F which is well within the acceptance criteria of 150 F.

Page 32 of 32

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