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ACCESSION NBR: 9208250103 DOC. DATE: 92/08/17 NOTARIZED: NOFACIL: 50-387 Susquehanna Steam Electric Stationi Unit 1i Pennsglva

AUTH. NAKE AUTHOR AFFILIATIONKEISERz H. W. Pennsylvania Pacer 5 Light Co.

REC IP. MANE REC IP IENT AFFILIATION

DOCKET05000387

NARTINiT. T. Region 1 (Post 820201)

SUBJECT: "Susquehanna SES Unit 1BCgcle 7 Startup Test Summary. "M/920817 ltr.

DZSTRZBGTZGN CODE: ZEESD CGPZES RECEZVED:LTR ~ ENCL L SZZE:TITLE: Startup Report/Refueling Report (per Tech Specs)

NOTES: LPDR 1 cd Tr anscripts. I NAXMELLzG 05000380

REC IP IENTID CODE/NANE

PD1-2 LARALEIGH'.

CQP IESLTTR ENCL„1 02 2

RECIP1'ENT'D

CODE/NAKEPD1-2

PD'QP IESLTTR ENCL

1 1

INTERNAL: ACRS

aG=F=I:L 02

5 51

1 1

AEOD/DSP/TPABNUDQCS-ABSTRACTRGN1 FILE 01

1

1

1

EXTERNAL: NRC PDR 1 1 NSIC

NOTES: 2 2

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Pennsylvania Power 8 Light CompanyTwo North Ninth. Street~Allentown, PA 18101-1179 ~ 215/774-5151

Harold W. KeiserSenior Vice President-Nuclear215/7744194

AUG I 7 1992Mr. Thomas T. MartinRegional Administrator, Region IU.S. Nuclear Regulatory Commission475 Allendale RoadKing of Prussia, PA 19406

SUSQUEEiANNA STEAM ELECTRIC STATIONUNIT I CYCLE 7 STARTUP REPORTPLA- 33 FILE R41-2A Docket No. 50-387

Dear Mr. Martin:

Attached is a copy of the Susquehanna SES Unit 1 Cycle 7 Startup Report, which is beingsubmitted to you in accordance with Technical Specifications 6.9.1.1 through 6.9.1.3. Thisreport addresses those startup tests described in our reload application dated December 11, 1991 ~

Very truly yours,-

H.. e r

Attachment

CC: NRC Document Control Desk (original)Mr. G. S. Barber - NRC Sr. Resident Inspector, SSESMr. J. J. Raleigh - NRC Project Manager, OWFN

9208250i03 9208i7PDR ADOCK 05000387P PDR

SUSQUEHANNA SES UNIT 1 CYCLE 7

STARTUP TEST SUMMARY

Prepared by: Paul Moran

Approved by:

Approved by:Manaefer Nuclear Operations

ABSTRACT

Susquehanna Unit 1Cycle 7

Startup Test Summary

Susquehanna 'Unit 1 resumed commercial operation for Cycle 7 onMay 17, 1992 following a 71 day refueling and maintenance outage.The Unit 1 Cycle 7 (hereafter referred to as SlC7) reloadincluded:

88ANF9x9228 ANF 9 x 9220 ANF 9 x 9228 ANF 9 x 9

thrice burnedtwice burnedonce burnedunirradiated fuel assemblies

The following startup tests, identified in the SlC7 ReloadLicensing Submittal, are discussed in this report:

1.2.3.4 ~

5.6.7.

Core Loading VerificationPOWERPLEX Input Deck ValidationControl Rod Functional (Insert and Withdrawal Checks)Subcritical Shutdown Margin DemonstrationIn-Sequence Critical and Shutdown Margin DemonstrationControl Rod Scram Time TestingTip Asymmetry

In addition, the startup program included core flow and LPRMcalibrations, thermal limits monitoring and baseline recirculationdata acquisition. A summary of these activities is also included inthis report.

Susquehanna Unit 1Cycle 7

Startup Test No.'Core Verification and Audit

~Per ose

The purpose of this test is to visually verify that the core isloaded per the analyzed designs.

CriteriaUpon completion of core alterations during the refueling outage,the core must be verified to conform with the reference coredesign used in the various licensing analyses. The verificationsto be performed include fuel bundle location, fuel bundleorientation, and proper seating of the fuel bundles within thecore. The verifications will be performed by the ReactorEngineering Group utilizing an underwater television camera. Theverification will be videotaped so that an independentverification may be performed. Any discrepancies discovered inthe loading will be promptly corrected and the affected bundlesshall be reverified prior to unit startup.

Results

Susquehanna took the following precautions to prevent a misloadedfuel bundle. During the total core offload, bundles to be usednext cycle were placed in the pool in the order in which theywere to be reloaded. This facilitated an orderly stripping ofbundles during the reload. After the offload was complete, aserial number verification of these bundles was performed(4/2/92) prior to reload. The core reload was performed in threeparts. First, the irradiated bundles were loaded and a partialcore verification serial number, location, orientation and heightcheck was performed on 4/18/92. Second, the first batch of newfuel, 72 bundles (3.40 wt % U-235, 10 GD5), was loaded and alocation check was performed (4/19/92) before the remaining cellscould have control rod functionals performed. Third, theremaining new fuel, 156 bundles (3.40 wt %, 9 GD4) was loaded.

The Cycle 7 final core verification consisted of two videotapedpasses over the core. During the first pass, the fuel bundleserial numbers were recorded on the videotape to verify properlocation. The second pass was performed to verify proper fuelassembly seating (assembly height check) and correct orientation.

The core tapes were independently verified to be correct by theReactor Engineering Supervisor and a representative of Quality"Control on 4/20/92. Therefore, the as-loaded core configurationis consistent with the core design Siemens Nuclear Power and PP&Lused in the evaluation of the S1C7 Reload Licensing Analyses.The S1C7 core map is included as Figure 1.

FIGURE 1

~ '~ PENNSYLVANIA POWER B LIGHT - NUCLEAR FUELS 8 SYSTEMS ENG'I NEER ING~ ~ CLASP - CORE LOADING ANO SHUFFLE PROGRAM

RUH +92- 1664 PAGE 6.03/ » /9? ~ ~

PREPARED BY/DAT

SSES UNIT-I/CYCLE-7 FULL C RE LOADING PATTERN

REVIEWED BY/DATE:

APPROVEO BY/DAtE:RECEIVED BY/DATE(SUPV REACT ENGRG)!

~), ~ II

CASE: 50 DATE STORED: 3/)I/92 TITLE: SSES UNIT ) CYC E 7 FULL CORE ~ LOADIHG PATTERN

GE-Y/GE-X: 13 15 17 19 21 23 27 29

60

58

54

52

50

X13587 X13600

X)3512 A)SQ73 *15957X)3538 X13580 A'15018 A16330 A14805

X)3SBO A15042 A 1 6314 A 15030 A16102

A11755 A ISQBI A)8302 A)5993 A)5170 A'11779

X)3577 X 13504 A)5022 A15086 A)4787 A 16394 A14782 A16290

X13579

A )5965

A)8328A 15150

A 16298

A14769

X1349 'I

A 15062

A'I4788

A)8310

A 14789

A15158

X)3597 X 13559 X13BO I

A ISQBQ A 15054 A)SO)B

518322 A14760 A 163'18

A)5090 A I B306 A)4772

A 16398 A 14738 A)6294

A 14806 A 16286 A 14836

48

46

X13540 A15028 A16282 A11763 A)5977 A14771 A)6278 A 15078 A)6274 A14730 AI6270 AI5166 A)6390

X13498 A)50)4 A16268 A15981 A)6386 A)4770 A)5116 A)4791 A)6382 A)4747 A)6262 A)5131 A)4825 A)4790

44

12

X)3528 A15997 A)6258 A15989 A)5)51 A14780

X)3195 A150)0 A14748 A)6370 A)4729 A)$216

A 16254 A 14826 A 'I 5070 A 11815

A15122 A IB366 A11808 A)6242

A 16378 A)5142 A 16374 A 15102 A 16250

A)1732 A 16238 A)4761 A16234 A)4773

40

30

28

26

21

22

20

18

A)8188 A)4737 1 16'I 82 A )1835 A 16331 1 148 18 A 16178 A 14828 A 15098 A 14739 1 15'1 38 1 14827 A 14809X 13567 A 15034

X13643 A15037 A IS IO) A11943 A15141 h 14855 A 14875

A)4851 A)6345 A)4882 A)6341 A 11934

A 'I 6337 A 14868 A 1618 I A 14856

A15089 A14901 A)5)21 A)6)93A)8189 A11911 A )d')85 A14843

XI3683 A15005 A)4873 AI6201 AI4951 A)6197

X13701 h)5053 A)6209 A)5117 A16353 hl1883 A16205 A)5)77 A16349 A11884 A)506) A15129 A)1846 A14917 A15109

X)3699 A15009 A11941 A 16221 A11927 h)5113 A'14881 A)8217 A 15133 A)6213 A)4865 A)8361 A15085 A 16357 A 11852

X 'I 3881 A)5041 A )6233 A 15097 A)6229 A 14933 A 'I 6225 A 14861 A 16365 A )4928 A 15077 A 14920 A 15988 A 11845 A 15 165

X1371S A)5013 A)4938 A18373 A)1949 AIB249 A15125 A16389 A)4874 A16245 A14952 A)6241 A)1919 A)6237 AI4Q) I

X)3881 A 15001 A)8261 A15992 A 15157 A14900 A)6257 A)1854 A)5073 A)4863 A)6381 A)5115 h'16377 A)5105 A)6253X 'I 3718 A )501 7 1 16289 A 15984 A 18389 A 14908 A 15119 A 14893 1 16385 A 'I 1935 1 16265 1 15 137 1 14853 1 14892

X)3525 A)5038 A16230 A)5091 A)8226 A14715 A)6222 A14818 A)6362 A14754 A15074 A14762 hl5985 A)4837 A15162

X13509 A)5006 A)4740 A)6218 A11753 h)51'IO A14797 A)8214 A15130 A)6210 A14817 A)6358 A15082 A16354 A)4824

X135) I A 15050 A16206 A15114 A18350 A)1799 A16202 A15174 A16346 A'14800 A)5058 A)5126 A 11838 A')4759 A 15'I06

X)3527 A15002 A14807 A IBIQB A14731 h'IS)91 A15086 A)4781 A)5118 A I 8)90 A14823 A16342 h)4798 A16338 A14716

14

)2

X13676 A15029 A16285 A14921 A)5980 A14909 A16281 A)5081 A 16277 1 11950 A 16273- A15)69XI3633 XI3710 A15025 A 15069 A)4889 A)6397 A)4902 AI6293 A)4907 A)5)61 h!4872 A)6269

A 1 6393

A 1 4844

)0 A 149?9 A )5964 A )6305 A )5996 A 15) 73 A )4899 A 16301 h 1489 I /! 1640) 4 14942 A 16'?97

X)3536 A)5045 A)63) I A 15033 A)6405 A)5153 A)63)3 A )5093 A )6309 A)4910

X)3674 X)3656 A)5049 A)6333 A)4871 A16329 A)1890 A)b375 A149)8 A1632)

X )3 702 A 15976 A 15960 A ) 5968 h)5065 h )! 9 72 h)50 7 h ) 502 'I

X)36?3 X)361? X)3635 X)37?5 X) )600 >)ll>55 x)36)3

FIGURE I (continued)

~ ~ PENNSYLVANIA POWER d LIGHT - NUCLEAR FUELS d SYSTEMS ENGINEERING~ ~ CLASP - CORE LOADING AND SHUFFLK PROGRAM

RUH 092- 1664 PAGK 7. ~ ~

03 / 'I 'I l92

PREPARED 8Y/DATEi

SSKS UNIT-1/CYCLE-7 FULL C

REVIEWED BY/DATKi

RE LOADING PATTERN APPROVED BY/DATE:RECEIVED BY/DATE

~ I 1 /4 (SUPV REACT EHGRG) i

CASE: SO DATE STORED: TITLEs SSES UNIT I CY E 7 FULL CORE LOADING PATTERN

GE-Y/GE-Xi 3'I 33 38 15 17 49 51 53 55 57

eo

58

56

54

52

50

46

14

42

X13804 X135d2 X13808 X13492 X)3582 X13d05 X)3594

A)SP)9 A15055 A)5170 A15063 A)5966 A15958 A15974 X13515

I 18319 A14767 A18323 A11795 A'I6327 A14811 A16331 A)5017 X)3561'13513A14775 A16307 A15091 A 16311 A)5151 A)6403 A15031 AI6315 A15043 X)3581

A18295 A)4743 A18391 A14794 A)6211 A)4786 A15171 A)5991 A16303 A15962 A)1756

A14841 A)8287 I)4813 A)$ 151 A14778 A)62QI A)4783 A)6395 A14798 A15067 A15023 X)3507 X13584

A)6391 A15167 A18271 A14735 A18275 A15079 I 16279 A14778 A15978 A14761 I'16283 A15027 X)3541

A14793 A14832 A)5135 A)6263 A)4750 A)6383 A)4792 A15147 A14777 A)6387 A15982 A16267 A15015 X)3499

A)6251 A15103 A16375 A)5)43 A)6379 A11822 A)5071 A14831 A)8255 A11785 Als)55 A15990 A16259 A15998 X)3533

A)1774 A18235 A)4786 A18231 A)4733 A)6243 A)4811 A16367 A15123 A)8217 A)4736 A16371 A)1719 A)5011 X)350240

38

Als)43 A)1840 A)5986 A)1745 A15075 A14757 A)6343 A14821

A11833 A18355 AIS083 *18351 A14820 A18211 A)5131 A)82)SA)6223 A14752

A11801 A15111

A)4227 A)5095 A 14231 A)5039 X)3534A14758 A182)9 A)174) A)5007 X)3518

36

34

32

30

28

2e

24

22

20

18

14

12

A)5)07 A14768 A)4839 h)5127 A15059 A)4801 A16347 A')5175 A)6203 A14802 A)8351 A15115 A16207 A15051 X13516

I )4751 A 16339 A 14803 1 16343 A)4834 I 1619 I I 'I 5'I 19 A)4784 I 15087 1 18195 1 14734 1 16199 1 148 I 2 A 15003 X 13534

A)1810 A)1830 I15139 h 11742 A15099 A)1829 Ald)79 I14819 A)6335 A)4842 I'lb)83 A14741 A)6187 A)5035 X13574

A)4876 A14858 A)5140 I')4116 A15100 A)4857 I'Ie')80 I )4887 A)8336 A 14850 Ale)84 A11948 A)6188 A 15036 X)3650I'14931 A)6340 A)4887 Alb314 Allbb2 A)8192 A)5'120 A)49D1 A)5088 A)6198 A)4Q54 A 18200 A14878 A)5004 X)3690

A IS)08 A)4926 A)1817 A15128 A15080 A)4885 A)8318 A15176 A ld204 I')1888 A)6352 A)5118 A)8208 A)5052 X13706

A)188'I A 16356 A15084 A)6360 I )1868 A16212 A15132 I )6216 A)4888 A15112 A)4Q32 A)6220 A)1915 A15008 X)3708

A15161 A14848 A15987 A)1923 A)5076 A)1931 A16364 A)4869 A16221 A14940 A)8228 A)5098 A16232 A15040 X)3692A)4912 A)6236 A)4924 A16210 A11953 A)6244 A)4877 A16368 A)5124 A)6248 A)4956 A18372 A14937 A15012 X13722

A)8252 A)5104 A)8374 A 15141 A 18380 A11870 A)5072 A11859 A14256 A14905 A)bibb A)5991 A 18240 A)5999 X)3489A14895 I 14860 I')5138 A)8281 A14938 A16381 A)4891 A)5118 A)4915 A)6388 A15983 A)4268 A15016 X)3719

A)6392 I'l5168 A)6272 A14955 A16276 A)5080 A)6280 A)4914 A)5979 A)4922 A)6284 A)5028 X)3677

A14849 A)6288 A)4879 A IS)60 A)4916 A)6292 A)4903 A)6396 A14898 A)5068 A)5024 X13713 X)3640

A ib29b A 14941 I l 6400 I 14896 A )6300 I )4906 A lb I )7 A 15995 A 16304 A 159b J A I 4930

A)4913 AlbJU8 AI5092 Aib3II A1515/ A)b404 A)50J2 AibJ Ib A)5044 XI3bJlA ib320 A 14975 A lbJ24 A 1489 l 4 16328 A 1488U A lbJ JZ A ibU48 X I Jbs 1 X I Jbl<J

A 15020 A)5056 A)5971 A)5064 A)5967 A )5959 A)5915 X 13/05

XI361b X)3658 x)3670 X)3776 X13638 XI3b)7 X13b30

Susquehanna Unit 1Cycle 7

Startup Test No. 2POWERPLEX Input Deck Validation

~Pur oee

To ensure the POWERPLEX input deck is updated correctly beforethe start of every new fuel cycle.

CriteriaPOWERPLEX is the Siemens (formerly ANF) software system designedto perform in-core monitoring of BWR cores. Core monitoring isperformed by the module, XTGBWR, a three-dimensional reactorsimulator code which calculates bundle nodal powers. ThePOWERPLEX input deck consists of all constants needed for theexecution of this code and subsequent calculation of the marginto thermal limits. These constants, must be updated prior to thestart of every new fuel cycle in order to ensure satisfactorycore monitoring of the new core configuration. The deck isupdated and validated by members of the Reactor Engineering Groupat Susquehanna.

Results

The POWERPLEX input deck was completely reviewed, all commentsresolved, verified to be correct and successfully loaded into thePOWERPLEX system prior to S1C7 startup.

Susquehanna Unit 1Cycle 7

Startup Test No. 3Control Rod Functional (Insert and Withdrawal Checks)

~Pur use

The purpose of this startup test is to assure proper control rodfunction and demonstrate that criticality will not occur due tothe withdrawal of a single rod.

CriteriaControl Rod Functionals include mobility, overtravel andsubcritical checks. These may be performed as each control cellis loaded in its final configuration.Each control rod will be cycled individually to ensure mobility.As each rod is fully withdrawn, it will be checked for overtravelby continually applying a withdrawal signal. Subcriticality willalso be verified with the rod withdrawn.

Results

Due to Shutdown Margin considerations, no control rod functionalswere allowed on fully loaded control cells until the partial coreverification was completed. No control rods overtraveled andsubcriticality was maintained as each rod was individually fullywithdrawn and reinserted.

Susquehanna Unit 1Cycle 7

Startup Test No. 4Subcritical Shutdown Margin Demonstration

~Pur ose

The purpose of this startup test is to assure at least theminimum required shutdown margin exists with the strongestworth control rod fully withdrawn.

CriteriaThe minimum required shutdown margin at BOC for Susquehanna Unit1 Cycle 7 is 0.38% ~ K/K. This test will verify at least thisamount by performance of a subcritical shutdown margindemonstration. The highest (strongest) worth control rod isfully withdrawn, then a diagonally adjacent rod is slowly notchedout verifying subcriticality at each step until the analyticallydetermined reactivity worth of the control rods at theirrespective notch position equals or slightly exceeds the requiredamount of SDM.

Results

The reactor remained subcritical with the highest worth controlrod fully withdrawn and an additional diagonally adjacent rodpulled to a notch position with a calculated worth of 1.293% aK/K. The required shutdown margin to be demonstrated wascalculated to be 0.5312% ~ K/K. This is 0.38% ~ K/K plus acorrection factor for the recirculation loop (moderator)temperature (104 degrees F) at the time of the test. Using datasupplied by Nuclear Fuels Engineering it was determined that thefollowing rods pulled to the indicated position would demonstratea shutdown margin of 1.1418% ~ K/K.

ROD POSITION TOTAL WORTH % h K/K

50-19*54-15

4812 1.293

*analytically determined strongest rod.As rods were pulled, subcritiality was verified after each notch.Subcriticality was also verified with the rods at the aboveindicated positions, thus satisfying the purpose of this startuptest. Figure 2 is a core map showing the test rod positions.

FI~ 2. CORE MAP SKSGNG TEST RX) PggITI~ ~SUBCRITICAL SHUItGNN MABQIN DENKSTPATICg

59

55

43

39—35—31

27

23

19

15 1Z

07

03

I I I I I I I

02 06 10 14 18 22 26 30 34 38 42 46 50 54 58

Btk&$ IÃ)ICKIER RX6 AT 00.

Susquehanna Unit 1Cycle 7

Startup Test No. 5In-Sequence Critical and SDM Determination

P~ur ose

The purpose of this startup test is to calculate the actualshutdown margin of the cycle 7 core and to demonstrate that noreactivity anomaly exists.

Criteria1), Shutdown Mar in

Technical Specification 3.1.1 requires an adequateshutdown margin to ensure the reactor can be madesubcritical from all operating conditions. This value,.38% ~ K/K has been determined to be the minimumrequired SDM to bring a reactor subcritical under theworst case conditions - a cold, xenon-free core at themost reactive point in the cycle with the highest worthcontrol rod unavailable for reactivity control. Atbeginning of cycle, the required SDM value must beincreased by a factor, R, if it is determined that coreshutdown margin is less at a point in the cycle otherthan the initial shutdown margin (for Cycle 7, R = 0% ~K/K). The required beginning-of-cycle SDM forSusquehanna Unit 1 Cycle 7 is 0.38% ~ K/K; the actualSDM will be calculated from data obtained during theinitial startup criticality.

2) Reactivit Anomal

Core reactivity is monitored to prevent excessivereactivity additions due to unforeseen reactivitychanges or reactivity anomalies. At BOC, a 1% a K/Kdifference between predicted and actual criticalcontrol rod positions might indicate improper coreloading or a computer code that is unreliable. Datagathered during the in-sequence critical, specificallythe Keff at the notch position of the control rod atwhich criticality occurs, is compared to predictedcritical control rod position Keff and a % reactivitydifference is calculated.

Results

The calculated SDM was 1.7455% ~ K/K and the differencebetween actual Keff and predicted Keff at criticality was0.0937% h K/K.

'Control rods were withdrawn in the B sequence until thereactor was on a stable, positive period. The notchposition at which criticality occurred was rod 22-47, notch20, step .36. A special log was initiated to record SRMcount rates and recirculation loop temperatures. Theaverage period was 152.2 seconds and the average looptemperature 134.7 degrees F which yield period andtemperature corrections of .393 x 10 ' K/K and 2.56 x 10 '

K/K, respectively.

1) Shutdown Mar inThe equation used to calculate SDM

SDM = Kcrit — Ksro — ~ p (period) — ~ p (temp)Kcrit *Ksro

Kcrit is Keff at the actual,.critical control rodposition (1.00390) and Ksro is Keff predicted with thestrongest rod out (0.98375).

The minimum required SDM for Unit 1 Cycle 7 atbeginning of cycle was 0.38% ~ K/K; the calculatedshutdown margin based on this test was 1.7455% s K/K,thus satisfying the acceptance criteria.

2) Reactivit Anomal

The reactor went critical at step 36 with Kcrit of1.00390. The equation used to calculate reactivitydifference was

Reactivity difference -„ Kcrit -1 — ~ p (period) — ~ p (temp)Kcrit

The calculated reactivity difference was 0.0937% a K/K.This satisfies + 1% ~ K/K acceptance criteria.A comparison of the predicted versus actual criticalcontrol rod patterns is included as Figure'3.

FIQJRg 3 ~ (XHPARIRN OF PREDICTED VS ACZtQL CZUTIQQ RR i%XTRRK

CC

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10 14 18 22 26 30 34 38 42

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46 50 54 58

PREDICTED CRITICAL PAT?KRN

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02 06 10 14 18 22 26 30 34 38 42 46

ACTUAL CRITICAL PAT?ZRN

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50 4 8

BLANKS INDICATE RX8 AT 00

Susquehanna Unit 1Cycle 7

Startup Test No. 6Control Rod Scram Time Testing

~Pur ose

To demonstrate the maximum scram insertion times of all rodsfollowing core alterations.

CriteriaSusquehanna Technical Specification 4.1.3.2 states thatscram insertion times of all control rods shall bedemonstrated through measurement with reactor coolantpressure greater than 950 psig prior to exceeding 40%thermal power after core alterations. For Unit 1 cycle 7approximately one-half of all control rod scram times wereto be determined by performing a black-and-white scram fromthe B sequence and using GETARS scram data. The remainingrods were to be individually scram time tested.

Results

Control rod scram times for 96 rods were obtained throughGETARS from the black-and-white scram performed 5/14/92 (Bsequence). The remaining rods were individually scram timedon 5/18/92. All scram times were within the acceptancecriteria, as shown in Table 1.

RODROD TIME T.S.

POSITION AS FOUND LIMITMAXIMUM INDIVIDUALRODSCRAM INSERTION TIMET.S. 3.1.3.2

22-27 6.23 7.0

AVERAGE SCRAM INSERTIONTIME OF OPERABLE RODST.S. 3.1.3.3

45392505

0.300.601.322.41

0.430.861.933.49

AVERAGE SCRAM INSERTIONTIME OF SLOWEST 2x2 ARRAYT.S. 3.1.3.4

45392505

0.310.651.422.61

0.450.922.053.70

TABLE 1: Results of Scram Time Testing of All Control Rods SlC7.

Susquehanna Unit 1Cycle 7

Startup Test No. 7TIP Asymmetry

~Per ose

The purpose of this test is to check core symmetry byperforming a statistical uncertainty analysis on theTraversing In-Core Probe (TIP) System. Also, by theperformance of this test, the proper operation of the TIPsystem will be assured.

CriteriaThe X test of significance will be performed with thesignificance level fixed at 1%. The test will be performedutilizing an octant symmetric rod pattern at, a power levelgreater than 75% of rated power. The startup test criteriafor symmetric TIP differences is that the X'aluecalculated shall be less than the critical X'alue. SinceSusquehanna has 19 symmetric TIP pairs, the calculated

X'aluemust be less than a critical X'alue of 36.19 (asdetermined by Siemens) . If the calculated X'alue exceedsthe critical value, the instrumentation and data processingsystem should be reviewed for any problems which maycontribute to abnormal TIP asymmetries. A seconddetermination of X'hould be then made. If the newmeasured value of X'xceeds the critical value, the fuelvendor shall be consulted and appropriate action taken toassure that a larger than anticipated TIP asymmetry does notadversely affect the safe operation of the reactor.

Results

A complete set of TIP data was obtained at the completion ofSusquehanna Unit 1 BOC7 Startup Testing Program at ratedthermal power. The nodal TIP values (Nodes 3 through 22)were summed up for each symmetric TIP pair using equation5.1 with the results summarized in Table 2. Using Equations5.2 and 5.3, the variance and X'ere calculated to be 4.15and 2.19 respectively. The X'alue of 2.19 is well withinthe 36.19 limit established by Siemens (Formerly. ANF).

Table 1Absolute Relative Difference

S etric TIP Pair

123456789

10111213141516171819

Absolute Relative Differencedm

.21-1.90-2.96

.553 ~ 7 2

-2.902 ~ 3 7

.983.69

-5.943.24

.951.18

-1.044.504.95

.48

.382.87

Equation 5.1

100 (Tml - Tm2)dm Tml + Tm2

2

22 22Note: Tml g T(k) for TIPl and Tm2 g t(k) for TIP2

K 3 K~3

where TIPl and TIP2 are symmetric TIP pairs

Equation 5 .2 (Variance)

19am~

S TIPi]2M 1 4.15

38

Equation 5 .3

2

X 36

Susquehanna Unit 1Cycle 7

Startup Program Summary

The following is a short summary of additional Reactor Engineeringactivities performed during the Startup Testing Program.

Thermal Limit Monitorin

Thermal Limits were checked throughout the startup period throughreview of the POWERPLEX core monitoring program, MONITOR, output.At no time did thermal limits exceed Technical Specification limits.TIP S stem — OD-1 Performance

A full set of TIPS was run at 38% power to update the core powerdistribution before the first core performance calculation, MONITOR,was initiated. Subsequent TIP sets were perfor'med at 60% and 100%power in conjunction with two LPRM calibrations. The LPRM currentswere updated and the LPRM GAFS found to be within the acceptablerange.

Power Distribution Com arison with Offline Monitorin

Favorable results were obtained when actual core power distributiondata was compared to SIMULATE-E/PPL core modelling code data. TheSIMULATE-E/PPL code is used by the Nuclear Fuels core managementengineer to predict power TIP response distributions throughout thecycle. This comparison is included as Figure 4.

Core Flow CalibrationA core flow calibration was performed at 98.9% core flow. Noadjustments to the jet pump and recirculation loop flowinstrumentation were required.

Recirculation Loo Baseline Data Ac uisitionRecirculation loop data was collected throughout the startup programto provide baseline data for plant performance monitoring in two loopand single loop operation. This data is used throughout the cycleduring the performance of the Technical Specification Jet PumpOperability Surveillance.

FIGURE 4

USC7CORE AVERAGE TIP QOMI AhiSON AT O.e08 ewniMTU

100

70

I „40

20

Legend0 TlP Measurement

0 SIMULATE-K/PPL

z m(xTo)

1 S I 4 S ~ 7 ~ S 10 11 Ck 1854 1S 1i 'l7 1 ~ 1$ f0 f1 RR 33 R4

Core Axial Node