ML20092C433

Startup Test Rept for Pilgrim Nuclear Power Station Cycle 11

ML20092C433
Person / Time
Site:	Pilgrim
Issue date:	09/01/1995
From:	Boulette E BOSTON EDISON CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
BECO-LTR-#95-09, BECO-LTR-#95-9, NUDOCS 9509130001
Download: ML20092C433 (21)

Text

_, .. .

t

• l

., h I Boston Edison Pi!gnm Nuclear Power Station j Rocky Hill Road i Plymouth. Massachusetts 02360 1

l E. T. Boulette, PhD I Senior Vee President - Nuclear September 1, 1995 [

BECo Ltr. #95490 1 U.S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, DC 20555 Docket No. 50-293 License No. DPR-35 Startuo Test Report Pilarim Nuclear Power Station Cycle 11 Enclosed is our Startup Test Report for Pilgrim Nuclear Power Station Cycle 11. If additional information is required please contact Mr. Robert Haladyna at (508) 830-7904 or Mr. Bruce Hagemeier at (508) 830-7808.

& ouLD E. T. Boulette, PhD ETB/ RAH /nas/ Rap 95/Startup Attachment cc: Mr. R. Eaton, Project Manager Division of Reactor Projects - 1/11 Mail Stop: 14D1 U. S. Nuclear Regulatory Commission 1 White Flint North 11555 Rockville Pike Rockville, MD 20852 U.S. Nuclear Regulatory Commission Region i 475 Allendale Road  :

King of Prussia, PA 19406 l Senior Resident inspector Pilgrim Nuclear Power Station 1100S0 9509130001 950901 i PDR ADDCK 05000293

. P .. _ . .PDR __

l

f

• STARTUP TEST REPORT a PILGRIM NUCLEAR POWER STATION

, CYCLE 11 INTRODUCTION  ;

Technical Specification 6.9.A.1 requires a summary report of plant startup and power escalation l testing be submitted to the NRC following installation of fuel that has a different design or has been manufactured by a different fuel supplier.

The Pilgrim Station Cycle 11 reload batch is based on the General Electric GE 11 fuel type. This fuel type is distinguished by a 9x9 lattice geometry and part length fuel rods. Compared to the 7x7 and 8x8 lattice geometries and full length fuel rods used in previous Pilgrim reloads the gel I reload batch for Cycle 11 constitutes a different fuel design.

General Electric has supplied all fuel loaded at Pilgrim Station since commercial operation began in 1972.

As required by Technical Specification 6.9.A.1 Pilgrim Station has provided the startup test report for Cycle 1 I within 90 days of the resumption of commercial power operation on June 6th 1995.

SUMMARY

A reload batch of 136 gel 1 fuel bundles with a bundle-average enrichment of 3.78 w/o was loaded in the Pilgrim Cycle 11 core to provide a cycle energy capability of 574 effective full-power days. This reload batch constitutes the first use of gel 1 fuel at Pilgrim Station.

As-loaded Cycle 11 core maps showing fuelloading by both bundle type and bundle serial number are presented in Figures 1 and 2. The Cycle 11 core loading is octant symmetric and is generally based on both the low-leakage and control-cell-core design principles. The Cycle 11 core design is documented in the Pilgrim Plant Design Change Package (PDC) 94-33, Reload 10 Cycle //

Core Design.

The final as-loaded core loading was verified to be consistent with the design core loading by Pilgrim personnel on May 8th 1995. Core loading verification following refueling was performed in accordance with the requirements of Station Procedure 4.5, Reactor Core Fuel Verification.

No core loading errors were identified.

Control rod coupling integrity was verified to be satisfactory consistent with the requirements of Station Procedure 9.13, ControlRodSequence andMovement Control.

Control rod scram time testing was verified to be consistent with the requirements of Technical Specifications 3.3.C.1 and 3 3.C.2. As required by Technical Specifications this testing was completed prior to exceeding 40% of rated core thermal power. Control rod scram time testing was performed in accordance with Station Procedure 9.9, ControlRodScram 7'ime Evaluation.

Page1 of20 STARTl?i. DOC A

STARTUP TEST REPORT s

PILGRIM NUCLEAR POWER STATION

, CYCLE ll Shutdown margin was demonstrated to be consistent with the requirements of Technical Specification 3.3.A.1 by both the two-rod method and the in-sequence critical method. The two-rod method was performed a total of three times on May 8th and 9th to facilitate control rod friction testing and in-vessel visual inspection. The in-sequence method was performed on June 2nd following initial criticality. These demonstrations were performed in accordance with the requirements of Station Procedures 9.16, Shutdown Afargin Check, and 9.16.1, Insequence Criticalfor Shutdown Alargin Demonstration, respectively.

Calibration ofinstrumentation important to monitoring core thermal power and core margins to thermal limits was performed as required by station procedures and Technical Specifications.

This instrumentation includes APRMs, LPRMs, TIPS and jet pump flow indicators. Calibration of this instrumentation was performed in accordance with the relevant station procedures.

Process Computer data processing checks were completed consistent with the requirements of Station Procedure 9.28, Process Conynster New Cycle Update.

Margins to thermal limits calculated by P-1 wer . compared to margins calculated by PANACEA and margins calculated by 3D-MONICORE. <-l was generally found to yield less MFLCPR margin than either 3D- MONICORE or PANACEA. P-1 was used as the official thermal limit calculation throughout the power ascension program to demonstrate compliance with Technical Specifications.

Ilot excess reactivity of the Cycle 11 core was found to be consistent with the requirements of Technical Specification 3.3.E. Ilot excess reactivity was determined in accordance with Station Procedure 9.8, Reactivity Follow.

RFO 10 oflicially ended on June 6th when Pilgrim Station went on line afler a refueling outage of 73 days. Rated power was reached on June 19th.

CORE I)ESIGN The Cycle 11 core was designed to provide 574 effective full-power days of cycle energy capability as specified by the Pilgrim Station energy utilitzation plan for Cycle 11. This cycle energy capability includes a planned power coastdown of 14 effective full-power days.

The Cycle 11 core design is based on the General Electric GE 11 advance fuel type. The GE I1 fuel type continues the basic trend of earlier advanced General Electric fuel designs by accommodating greater discharge exposures and providing more margin to thermal limits, the principle ingredients to reduced reload fuel costs and higher plant capacity factors. The GE I1 fuel type continues this basic trend through a number of key design features: a 9x9 lattice geometry, 8 part length rods, two large central water rods,10 atmospheres of helium prepressurization, high performance ferrule spacers, a high pressure drop lower tie plate and a low pressure drop upper tie plate.

Page 2 of 20 STARTUPl.IXX'

?

STARTUP TEST REPORT s PILGRIM NUCLEAR POWER STATION CYCLE ll The Cycle 11 core loading pattern is based on the low-leakage and control-cell-core design principles in use at Pilgrim since Cycle 5. The control-cell-core design principle designates selected rods in the core for reactivity control and power shaping at rated power and restricts fuel loading in the adjacent fuel cells to once- or twice-burned fuel. By avoiding rod withdrawals at power in the vicinity of fresh fuel, the control-cell-core design simplifies operation, improves fuel reliability, increases operating thermal margin and improves capacity factors.

The low-leakage design principle preferentially loads twice- and thrice-burned low-reactivity fuel on the core periphery to reduce radial neutron leakage, thereby yielding improved fuel cycle efliciency and reduced reload fuel costs. Reduced radial neutron flux also yields a reduced fast-neutron flux at the reactor pressure vessel wall, the reactor shroud and other core internals.

The Cycle 11 core design provides the cycle energy capability specified in the Pilgrim energy utilization plan for Cycle 11 with a 136 bundle reload batch of gel 1 fuel at a bundle-average enrichment of 3.78 w/o. With this reload batch the inventory of fuel in the Pilgrim Cycle 11 core is:

Number Cycle ofBundle_s Bundle Type Loaded 136 G E78 -P8 DRB 300-5 G 5.0/2G4.0-80M- 145-T 8 ,

168 G E88-P 8 DQ B 323 - 10GZ-90M-4 WR- 145-T 9 140 GE10-P811XB355-11GZ-100M-145-T 10  ;

136 GE l 1-P9 HUB 378-15G7-100T-141-T 11 Figures 1 and 2 present the as-loaded Cycle 11 core maps showing fuel loading by both bundle  !

type and bundle serial number.

The Cycle 11 core is loaded to be octant symmetric by both fuel type and, with a small allowance ,

for variance, bundle exposure.

The Cycle 11 core design is documented in Pilgrim Plant Design Change Package (PDC) 94-33, Reload 10 Cycle 11 Core Design. j

! The Cycle 11 core design meets all heensing critena specified in Revision 10 of NEDE-24011-P-

A and NEDE-24011-P-A-US, the General Electric Standard Application for Reactor Fuel l (GESTAR-il).

l l

l Page 3 of 20 STARTUPl.!XX' l

1 1TARTUP TEST REPORT PILGRIM NUCLEAR POWER STATION

. CYCLEll CORE VERIFICATION The final as-loaded Cycle i1 core loading was verified on May 8th 1995 consistent with the requirements of Station Procedure 4.5, Reactor Core Fuel Verification. Three separate criteria were verified: bmidle orientation, bundle seating and bundle location.

Bundle seating was verified by observing the channel fasteners of adjacent bundles in each fuel cell were vertically aligned.

4 Bundle orientation was verified by observing the channel fasteners of adjacent bundles in each fuel cell were oriented toward the center of the cell.

Bundle location was verified by observing bundle serial numbers in the core were consistent with bundle serial numbers in the final fuel loading plan. I Verification of the final as-loaded Cycle 11 core loading identified no core loading errors.

CONTROL ROD COUPLING INTEGRITY l

Control rod coupling integrity was verified whenever a control rod was fully withdrawn for the )

I first time following refueling. Coupling integrity was established by observing a discernible response of nuclear instrumentation during the rod withdrawal and, upon withdrawal to the full-out position, observing the rod would not reach its over-travel position.

Control rod coupling integrity is governed by Station Procedure 9.13, Control Rod Sequence and Movement Control.

CONTROL ROD SCRAM TES TING

Single rod scram time testing on all 145 control rods was successfully completed on June 9th prior to exceeding 40% of rated core thermal power as required by Technical Specification 4.3.C. l. Results of this testing are presented in Table I-A and Table I-B.

I SilUTDOWN MARGIN Shutdown margin (SDM) was demonstrated using both the local two-rod subcritical method and the in-sequence critical method. Both methods demonstrated adequate SDM although the margin i demonstrated by the local two-rod suberitical method was substantially les than predicted by design.

Page 4 of 20 STARTlTI.DO('

________________ - _ .)

STARTUP TEST REPORT PILGRIM NUCLEAR POWER STATION CYCLE ll The actual critical position observed during the in-sequence critical SDM was in excellent agreement with the estimated critical position.

Local Two-RodMethod A local SDM demonstration was iborted by Pilgrim Operations personnel on May 8th 1995 when the reactor core approached uiticality. This action was performed consistent with the requirements of Station Procedure 9.16, Shutdown Margin Ched. It is noteworthy that the intent of this procedure is to demonstrate SDM while maintaining the reactor subcritical.

An evaluation of this aborted SDM demonstration found a SDM of 0.53%Ak was demonstrated assuming the reactor to be critical when the demonstration was aborted. While 0.53%Ak meets the requircments of Technical Specifications for a 0.25%Ak minimum SDM this value is 0.96%Ak below the aign value of 1.49%Ak. The magnitude of difference between the design and demonstrated SDM was cause for concern and provoked a root cause evaluation.

The root cause evaluation determined the difference between the demonstrated and design SDM is largely a consequence of the increased uncertainty associated with a local two-rod SDM demonstration. In particular the local SDM demonstration is more sensitive to uncertainty in the control blade depletion and core exposure distribution For the SDM demonstration performed on May 8th 1995 a more accurate accounting of control blade depletion was found to reduce the predicted SDM by approximately 0.1%Ak, Accounting for exposure using P1 instead of PANACEA was found to reduce the predicted SDM by 0.25%Ak. Together these two effects account for 0.35%Ak of the observed difference between the design ar.c demonstrated SDM.

An additional 0.3%Ak of the difference between the demonstrated and design SDM was accounted for when the cold target eigenvalues used for the design SDM calculation were adjusted to reflect the cold cross section libraries actually used for the Cycle 11 design calculations.

The 0.35%Ak difference due to control blade depletion and exposure differences together with the 0.3%Ak target eigenvalue difference account for 0.65%Ak of the 1%Ak difference between the demonstrated and design SDM. The remaining difference of 0.35%Ak was attributed to the inherent uncertainty of the General Electric modeling methodology.

The results of this evaluation are documented in Pilgrim Station Problem Report 95.9270.

On May 9th 1995 the local two-rod subcritical SDM test was repeated and successfully demonstrated a SDM of 0.42%Ak without approaching criticality. This test used the same margin and object rod but positioned the margin rod at Notch 18 instead of the more conservative Notch 22 used in aborted demonstration. A less conservative temperature correction term was used as well. This temperature correction term applied to moderator temperatures less than or equal to Page 5 rf 20 STARTlTI.!XX'

STARTUP TEST REPORT PILGRIM NUCLEAR POWER STATION CYCLE 11 90 F. The first SDM demonstration had used a temperature correction term applicable for temperatures of 100 F or less.

May 9th also saw two additional local two-rod subcritical SDM demonstrations successfully performed. These SDM demonstrations were performed on the next two strongest worth control rods in the core to provide assurance SDM had in fact been demonstrated with the strongest rod fully withdrawn. This assarance was considered prudent in view of the magnitude of difference observed between the design and demonstrated SDMs. The SDMs demonstrated by these tests were 0.42%Ak and 0.43%Ak.

In-Sequence Afethad A SDM of 1.23%Ak was demonstrated on June 2nd using the in-sequence critical method of Station Procedure 9.16.l, insequence Critical For Shutdown Afargin Demonstration. The agreement between this value and the design SDM of 1.49%Ak provided independent confirmation the root cause evaluation had identified correctly the causes of the discrepancy between the local two-rod SDM and the design SDM.

Estimated CriticalPosition The estimated critical position for control rods was found to be in excellent agreement with the actual critical position. Initial criticality was estimated to occur when the 7th rod in Group 2 was

, pulled from Notch 12 to Notch 48 in an A-2 Sequence. Moderator temperature was assumed to be 180 F. This estimated critical position was based on a critical eigenvalue calculated for the core configuration observed during the aborted local two-rod SDM demonstration on May 8th. l 1

Initial criticality was actually realized when the 7th rod in Group 2 was pulled to Notch 28. ,

Control rods were withdrawn in an A-2 Sequence and the average moderator temperature was l 180 F. The reactor period was 208 seconds.

INSTRUMENT CALIBRATIONS l

APRAfs i

Average Power Range Mor.itors (APRMs) were calibrated as required during the power i ascension to maintain the APRM gain adjustment factors (AGAFs) between 0.87 and 1.00.

AGAFs are the ratio of the desired APRM reading to the actual relative reactor power reading as determined from a reactor heat balance.

1 Page 6 of 20 STAR 1UPl.rXX' .

i STARTUP TEST REPORT PILGRIM NUCLEAR POWER STATION CYCLE 11 APRMs were first calibrated in Cycle 11 on June 6th at a power level of 20% of rated.

Subsequent calibratic,ns were performed on June 8th at power levels of 22% and 38'A, on June 10th at a power level of 50%, on June Ilth at a power level of 87%, on June 14th at a power level of 96%, on June 15th at a power level of 94%, and twice on June 19th at a power level of 100%. APRMs have since been calibrated as required.

All APRM calibrations were performed consistent with the requirements of Station Procedure 9.1, APlo1 Calibration The initial APRM calibration was based on a hand heat balance. All subsequent APRM calibrations were based on reactor power values from OD-3, the Process Computer Core Thermal Power and APIM4 Calibration on-demand program.

LPIMis Local Power Range Monitcrs (LPRMs) were calibrated as required by Station Procedure 9.5, LP/MI Calibration. LPRMs are calibrated to maintain gain adjustment factors (GAFs) between 0.95 and 1.05. GAFs are the ratio between the desired LPRM console readings and the actual LPRM console readings.

LPRMs were first calibrated in Cycle 11 on June 10th. Subsequent calibrations were performed on June 22nd and August 2nd.

11Ps The Traversing incore probe (TIP) system was used as needed to update the P-1 BASE array and

calibrate LPRMs. Update of the P-1 BASE array is required following significant changes in core power distribution. Significant changes in core power distribution manifest themselves by a large number of BASE CRITs when P-1 executes and an increasing uncertainty in the P-1 calculation of thermal hmits.

Update of the P-1 BASE array is effected by execution of the Process Computer OD-1 on-demand program. Through August 1st OD-1 was executed as indicated in Table II.

All OD-l's were executed consistent with the requirements of Station Procedure 9.5.1, Operation of 11P Machinesfor Process Comjmter Updating.

Accurate axial alignment of TIP machines is required for accurate updating of the P-1 BASE array and accurate LPRM calibrations. Station Procedure 9.20,11P Arial Alignment, is used to assess the degree of alignment between TIP machines. Application of this procedure in Cycle 11 found TIP Machine A to be aligned between 4 and 5 inches too low for the 8 channels involved. 1 Page 7 of 20 l sTrnitri txic j

l i

STARTUP TEST REPORT PILGRIM NUCLEAR POWER STATION CYCLE 11 Axial alignment of Machine A was returned to specification on July 27th. Axial alignment of the remaining TIP machines was consistent with procedural requirements.

Jet Pwnps Jet pumps and recirculation drive flow were calibrated as required during the power ascension consistent with the requirements of Station Procedure 9.17, Core Flow Evaluation.

The acceptance criteria for jet pump calibration are agreement within iS% between panel C-905 indicated core flow and calculated core flow. The acceptance criteria for recirculation drive flow calibrations are agreement within 2% between the two process computer loop flows and agreement within 5% between the indicated APRM loop flows and the calculated loop flows.

Jet pump and recirculation drive flow calibrations were first performed in Cycle 11 on June lith at a core thermal power of 72% of rated. Subsequent jet pump and recirculation drive flow calibrations were performed on June 20th and July 14th at 100% power.

Jet pump flows were monitored on July 6th, July 27th, and August 3rd to demonstrate consistency with the acceptance criteria of Station Procedure 9.17.

PROCESS COMPUTER DATA PROCESSING CHECKS The P-1 Process Computer databank was updated consistent with the requirements of Station Procedure 9.28, Process Computer New Cycle Update.

A number of checks are specified by Station Procedure 9.28 to verify the new Process Computer databank is consistent with the reload core design and has been correctly loaded into the Process Computer. These checks were completed satisfactorily by August 15th.

Consistency between the oflicial databank transmittal and the reload core design was verified by a general review of the relevant core design documents against the oflicial databank transmittal.

This general review was completed on May lith.

Refere startup a number of checks were made to verify the new Process Computer databank had "o correctly loaded

1. Differences between the old and new databank identified by the Process Computer were verified to be consistent with the differences identified in the oflicial databank transmittal.

2. The bundle loading identified by the Process Computer was verified to be consistent with the bundle loading specified by the core design documentation.

Page 8 of 20 STARTL Tl D(X' l

STARTUP TEST REPORT PILGRIM NUCLEAR POWER STATION CYCLE 11

3. At least one node in each bundle was verified to be shuffled correctly into the exposure (EXF) and void history (EXVF) arrays.

4. All new LPRMs and control blades were verified to have zero exposures in their relevant arrays (CICEX and TCREX).

5. The isotopic compositions for at least one bundle in each batch were verified to be unchanged from their values in the old databank.

These checks were completed on May 22nd.

Following startup a number of additional checks were performed. These checks include-

1. Verification P-1 is calculating symmetric thermal limit and power distributions given a symmetric core loading and core control rod pattern.

2. Verification control rod positions are consistent with those indicated in the control room.

3. Verification LPRM reading . ie consistent with those indicated in the control room.

The last of these checks was completed on August 15th.

TIIERMAL HYDRAULIC LIMITS AND POWER DISTRIBUTION 1hermalLimits Calcidated by P1 The maximum fraction of limiting critical power ratio (MFLCPR), the maximum fraction of limiting power density (MFLPD) and the maximum average planar linear heat rate (MAPRAT) were monitored throughout the startup using the General Electric P1 NSS core monitoring ,

software. Margins to thermal limits were maintained as required by Technical Specifications. L

! The P1 power distribution was updated as required during the power ascension using the l traversing incore probe (TIP) system during the ascent to rated power. The core thermal power,

, rated flow and thermal limits obtained from selected updates are presented in Table II.

d Two features of Table Il are of particular note. One is the MFLCPR value of 1.015 observed on i

! June 13th at 21:32 hours. This MFLCPR resulted from executing OD-1 to clear BASE CRITICALS and update the BASE array in P-1. A P-1 executed before this OD-1 at 18:10 hours on the 13th showed MFLCPR to be 0.982. Power and flow at this time were 92% and 85%

respectively. Following execution of OD-1 the control rod pattern was adjusted to restore MFLCPR to a value less than 1.0 as required by Technical Specifications. The Pl executed at i

Page 9 of 20 STARTLT1.IXX'

STARTUP TEST REPORT PILGRIM NUCLEAR POWER STATION CYCLE 11 21:47 hours on the 13th showed a MFLCPR of 0.983. Core power and flow were 90% and 86%

respectively.

The second feature of note in Table II is the drop in MFLCPR from a value of 0.988 on the 22nd to 0.967 on the 23rd. As discussed in the section that follows this drop is the result of corrective action taken to address a large TIP asymmetry introduced by a unique TIP instrument tube for the electro-chemical potential (ECP) probe.

7hermal Limits Caletdated by PANA CEA Pl-calculated thermal limits were compared with off-line thermal limits calculated by the General Electric's PANACEA design code throughout the startup. Selected results from this comparison are presented in Tables III-A, III-B and III-C. Table III-A shows PANACEA generally underestimated the PI-calculated MFLCPR by between 0.05 to 0.09. Tables III-B and III-C show generally excellent agreement between PANACEA and P1 for MFLPD and MAPRAT.

Tip Asymmetry Part of the large difference between the P-1 calculated and PANACEA calculated MFLCPR was attributed to a significant ' IIP asymmetry between locations 28-37 and 36-29. An investigation of this TIP asymmetry revealed it to be a consequence of the unique design of the instmment tube at location 28-37. This instrument tube contains probes used to measure the electrochemical potential of reactor water. Incorporation of these probes into the instrument tube required a larger diameter and thinner wall for the outer tube sheath. As a result of this geometry difference the instrument tube at location 28-37 was surrounded by more water than would be the case with a standard tube. More water yields increased neutron thermalization and a greater LPRM reading for a given power level in adjacent bundles.

The conclusion that the TIP asymmetry between locations 28-37 and 36-29 was a consequence of the unique design of the instrument tube at location 28-37 and not a real core power asymmetry is consistent with the fact the Cycle 11 core was designed to be octant symmetric. This conclusion was confirmed by results from the General Electric 3D-MONICORE core monitoring software which was running in parallel to the oflicial NSS (P-1) core monitoring software throughout the startup. 3D Monicore showed no significant power asymmetry.

The corrective action plan developed to address this TIP asymmetry changed the P-1 data bank to effectively substitute the TIP data from the instrument tube at location 36-29 for the TIP data froru location 28-37. Due to the safety significance of MFLCPR this change was implemented only after a safety evaluation concluded this change could be effected consistent with the criteria l l of 10CFR50.59. FRN 94-44-07 documente this evaluation.

Page 10 of 20 STARTtTI.IXX' l

\ l

)

STARTUP TEST REPORT PILGRIM NUCLEAR POWER STATION CYCLE 11

{ .

The corrective action plan to address the TIP asymmetry between locations 28-37 and 36-29 was -

~

implemented on the morning of June 23rd. Following implementation of this plan MFLCPR dropped from 0.98 to 0.96 with the limiting MFLCPR of 0.96 at another core location.

ThermalLimits Calculated by 3D-Monicore 3D Monicore has been used to monitor thermal limits in parallel with P-1 since the start of Cycle

11. 3D-Monicore is GE's latest core monitoring software and is generally considered to provide a more accurate calculation of thermal limits than either PANACEA or P-1.

Table IV presents selected 3D-Monicore cases during the course of the Cycle 11 power ascension. These cases were selected to correspond as closely as possible to the times of the P-1 cases listed in Table II. A comparison of thermal limits in Tables II and IV shows 3D-Monicore generally provided 0.03 to 0.06 more MFLCPR margin than P . > hen rated power was approached. Both 3-D Monicore and P-1 provided substantial MFLPD and MAPRAT margin.

IIOT EXCESS REACTIVITY The actual control rod notch inventory (adjusted to reflect rated reactor dome pressure, rated core inlet flow rate and nominal core inlet subcooling) was verified to be consistent with the design notch inventory on June 23rd and July 17th. Table V presents both the actual and design control rod notch inventories for these dates. The acceptance criteria for this comparison is an -

actual control rod notch inventory that differs from the design notch inventory by no more than 270 notches.

Monitoring of hot excess reactivity is governed by Station Procedure 9.8, Reactivity Follow.

ADDITIONAL TESTING The gel 1 fuel loaded in Cycle 11 required no modifications to plant systems or components.

Accordingly the first reload of gel 1 fuel at Pilgrim requires no testing during startup beyond that normally performed to assure compliance with Technical Specifications. These test results have been presented in the sections above as required by Technical Specification 6.9.A.l.

i Page11of20 )

STARTUPl.IX)C i l

i 1

i STARTUP TEST REPORT PILGRIM NUCLEAR POWER STATION CYCLE 11 1

FIGURE 1: PILGRIM CYCLE 11 CORE LOADING MAP ,

j BY BUNDLE TYPE U]U$ U[ U lUU

LiTRSEJLJLJEE1EuJ o,c ,a .,Nw m c, i DJLam ao Omso myy_m Uta soJ ,LJ P 9 P O N_ LW D B N N O_LO M 5 F "3 5 M LG 9 E l ll L2SERWEOREDEOWEL IBMS - BB0J i

P9N_gNom Dam o'o mog] E P M . LE O R' lLCOWNOSOEO DRS _ DOF3BE SWOEJJ P 9 NIE MIE NIE id; P EdB 9 M M MPR

! L J $ g g ro g ro a g o E g a c$ c a c B g g E J

! ~EBMEME MBBERBENBMEBt LJEIENIES SBLe,so a a.ac P a v w mP vR.RLJ 00 JS&dRSR j

E7C #9 aPo os'No Eroo FM 9 E*1 l LJS 2df Ods UJ NUd REN NUJ

! P95%_ BPM 9 Q D D En__NWE9 i LaJLOSSLJS M UJ N REdNNE_JLJ i P9POS DO FJSRO [0td MfE9PR

! LJLd5 MS'ESU41RNRR RLJLJ E9E7M NOP9 i

LJLSMSMSMENNEJdNJLJ i P7D7P9lPRPRU9E-^9 l

LJLoJLJLgJE loJ rd GE711-P8DRIG00-5G5.02G4.0-80M-145-T GE10-P811XID55-11GZ-100M-145-T

@ GE8B-P8IX)B323-10GZ-80M-4WR-145-T Gell-P91IUIO78-15GZ-100T-141-T i

l

! Page 12 of 20 l STAR'IUPl.IXX'

1 l

1 l

STARTUP TEST REPORT '

PILGRIM NUCLEAR POWER STATION CYCLE 11 FIGURE 2 PILGRIM CYCLE 11 CORE LOADING MAP BY BUNDLE SERIAL NUMBER 01 03 05 07 09 11 13 15 17 19 21 23 25 52 #' ' #" " ## '"' #"' #"

OI O O O 01 e et 3, mo4,gh45 mio bn i

m 091 6e4 m 59 u i i 1

, LW0ie I LWO43 LYF068 lY,H4 YM990 LYZ077 YJ4977 YJ4969 YJ4963

_..O- ,0_ _

_,_D O__, ______O 0,__,. n___,0, y __ _ __O, 46 l l l I

. I LYF079 LYF017 LYF022 LYZO69 YJ5004 YJA055 YJ4974 YJ4951 YJ4986f YJ4976 LYZO29 LYF024 [YF042 LYZO53I [YZO97 YJ4953 ! JA657 LYZOO5 l YZ117 YJA658! [YZ146 42 <

t YJ4958f

, i j <

l l I l

. J .

i LWO23 !LYF016 LYF006 LYZO57 YJ4959l YJ5014 LYZ161 l LYZ157 YJ5000 YJA661 YJ4991 YJA662 40 j LYZO25 l j c' O O O Ci ,0 , Dt .D. O O D D, Cl DI O Ot 0 0 0 0 D e' O Di O O.

LWO60I LYZ129 YJ4952 YJ5007 YJA665 YJA666 LYZO21 YJ4967 YJA667 LYZ109 YJA668 38 1 LYZO61 YJA669{

1 l

. 1 LYF040 LYZ165 YJ4960 YJA675 YJ4950 LYZ113 YJ5010 YJA076 LYZ141 LYZ101 YJ4978 YJA677 LYZO65f C O O O O O D D D 0 0 0 D l D D 0 0 DI O O ID D ID' L-70 LYZ105 LYZ133 YJA681 YJ4984 LYZ005 YJA602 LYZ137 O O O LYZOO1 LYZ013 YJA683 LYZ121 YJ4962 i i l 1 ,

tYr044 LYF046 Y,996 YJAee7 YJA688 YJ4997 YJA689 YJ5015 YJA690 3, YJ5006f l LYF030 YJA691 LYZ009 D; e o o ci o 1 o

O! ci D o o e O' f0 0 0 Ol D 0 ID D' D LYF056l LYF192 YJ5005l YJA697 YJ4905 LYZO40 YJA6981 LYZ149 D D D VJA699 LYZO93 YJA700 LYZ041 YJA701 l i I ( l L ,5 m. , Y,9,5 m 707 mm. m,.

,, mmj Y,966 m.1 l Y,9,5 m ,,5 m,.

Di O 01 D Cl 0 06 O DI O O O Lm17! O!

l l

Page 13 of 20 STARTUPl.IX)C

l l

STARTUP TEST REPORT PILGRIM NUCLEAR POWER STATION 4 CYCLE 11 end FIGURE 2 PILGRIM CYCLE I1 CORE LOADING MAP BY BUNDLE SERIAL NUMBER

. (CONTINUED) 27 29 31 33 35 37 39 - 41 43 45 47 49 51 LYF115 LYF105 !LYF133 LYF143 'LYF177 LYF096 f LYF153 o o! o el e o! o o ci o ai io a:

LYF179 [o LYF132l LYF097 LYZo74l LYZ154 LYZO46l LYF139 I i i.

YJ4964 YJ4970 ! YJ4980 LYZ134 YJ4993 YJ5003 ! L YF114 LYF190 f LYF172 8

_..._D' 8_._ _._._8 (8 0' O o- 8 o ai ei o YJA650 o' io n. "o YJA651! YJA652 YJA053i YJA654 YJ49994 'LYZO90 LYF176i LYF182 46 .

e l l 1 I LYZO30 YJ4981l YJ4987 YJ4956 YJ4983 YJA656! YJ5013 LYZ070 44 LYF162 LW171 ! LYF103 i I  ;

o os o o o al a o o ci o o o. T~ oi o e o o. o Io YJ4961 LYZ147 : YJA659 LYZ118i LYZOO6 YJA660, of 42 YJ4954 LYZ162i LYZO54 LYF140i LYF164

[

1 I ll ,

I- _j l LYZO26 YJA663f YJ4992 YJA664 YJ5009 LYZ158f LYZO98 YJ5019 YJ4960 40 LYZO58 !LYF180 LYF170 ! LYF163 l j o ai o e o o. o l l o o ci o 01 !c c oi o oi to YJA670 LYZO62 VJA671 ci {o et o e o o o LYZ110 YJA072 YJ4972 ' LYZO22 YJA673; YJA674 YJ5010' YJ4955 LYZ130! YJ0793 l l l

. l i 1.

LYZ066 YJA678 YJ4979 LYZ102 ! LYZ142 YJA679 YJ5017 LY2114 YJ4957 YJA680! YJ4971 LYZ166!LYFl38 o o! o o! ,o o! o ol e of a o in o at o os YJ4905 YJA684 jo ai o al e o' o LYZ014 LYZ002 LYZ138 YJA685 LYZ006I YJ4989 o! lo LYZ122l YJA686 LYZ078 Wl106i LYF116 I l 1

. i l i i

LYZO10 32 YJA692fLYF152 YJA693 YJ5018 YJA694! YJ5000 YJA695! YJ5011 YJA696f YJ5001 LYF144 'LYF142

1  ;  ;

o el e o!, o l ci o o! o el o e' lo YJA702 LYZO42 YJA703 JA704 LY2150 LYZO94j YJA705 LYZO37 VJ4988 YJA70h YJ5012 LYF178 YF134 i i i i 1- I i LYZ018 YJA713 LYZ126 YJ4982 LYZ002 YJA711 LYZO50 YJ4973f LYZ034 YJA712 YJ5002 o ,

LYF125LYF165 f el o et o a ,

o! ei o oj o o1 jo Page 14 of 20 1

e STARTUP TEST REPORT PILGRIM NUCLEAR POWER STATION CYCLE 11 FIGURE 2 PILGRIM CYCLE !! CORE LOADING MAP BY BUNDLE SERIAL NUMBER (CONTINUED) 27 29 31 33 35 37 39 41 43 45 47 49 51

[YZ019 YJA716 [YZ127 YJ5064 [YZO93 YJA717 (YZO51 YJ5073 [YZO35 YJA718 YJ5044 LYF130 [YF166 YJAn4 tYZO43 ! LYZ151 ! YMU7 LYZ038 ! YJ5054 f

,, lvJan5 LYZO951 YJAM6 YMM8! YJ5034 LW187 lLW135 0 _

o' o. o 'S o o _o ,o o .o o e o o #o o 'o D 'o of 'o o o o iD LYZO11 YJA734 LW161 YJA735' lYJ5024 YJA736 ;YJ5042 YJA737, YJ5033 YJA738 YJ5043 LYF145 ' lLYF147 22  ; l  : l ll l ll l

_J I ll YJ5077 LYZ123 ! YJA742 LYZO15 LYZOO3 LYZ139 YJA743 YJ5055 YJA744 LYZ135 LYZ107 LYF121 o I o _.

o ._____._o. 'o_ __._,__o ,o LYZO67 YJA748 YJ5061 LYZ103 LYZ143 YJA749: YJ5023 LYZ115 YJ5089 YJA750 YJ5071 LYZ167 LYF151 l l 1 l I l YJA756 LYZ063 ! YJA757 LYZ111 YJA758 YJ5072 LYZO23 YJA759! YJA760 YJ5032! YJ5087 LYZ131 !LYF131 16 i (; j e ol o ai lo a' o al o o~ o o o o c io o; io o o 7 ei o ai o o. o LYZO27 YJA763 jYJ5048 YJA764i YJ5031 LYZ159 LYZ163 YJ5025 YJ5000 LYZO59 i LYF185 LYF175i LYF168 ll _j i j i i e . .

YJ5081 LYZ148 YJA767 LYZ11 LYZOO7 YJA768 YJ5086 LYZO99l!LYZO55 LW169  !! LYF167 o o 'o o, lo o o e' o e o o ot o YJ50631 YJ5053 O .o o o' ~o o al d LYZO31 YJ50881 YJ5005 YJA770: YJ5035 LYZ071i LYF174 LYF124 LYF112 10 i

, I I I i i YJA776 YJA777 YJA778 YJA779! YJA780 YJ5041 LYZO91 LYF189l LW149 08 o e o o l

et e, o ei o o al o o fo o! jo ai "o YJ5076 YJ5070 YJ5062 LYZ079l YJ5049 YJ5045i LYF123 LYF191 LYF173 I I 1 i LYF186 LYF137 LYF100 LYZ075 LYZ155 LYZO47 LYF150 o _ _o! ! o___. _

o lo, _> o. . _ _

o el jo o fo ai o LW122 LW110 ! LYF136 LW146 , ;LYF188 LYF1011 LYF160 i i i i

l Page 15 of 20 STARR'Pt DOC i

l

- _ _ _ _ _ _ - _ _ _ - . _ _ _ _ _ _ _ _ _ _ - - _ _ _ - - _ . _ - _ _ _ ~

STARTUP TEST REPORT PILGRIM NUCLEAR POWER STATION CYCLEll i

FIGURE 2 PFLGRIM CYCLE 11 CORE LOADING MAP BY BUNDLE SERIAL NUMBER (CONTINUED) 01 03 05 07 09 11 13 15 17 19 21 23 25

' " ** "* * '# ~ ' ' ' ' # '" '"* '#' ' '

26 * '"

LYF057 YJ5027 YJA719 YJ5051 I LYZO39 YJA720 LYZ152 YJA721! LYZO96 YJA722' LYZO44 YJA723 o lLYF013 o _ _.

o o - -. ? o -

o o. .- ovo a o c.

o lo C. o o o o Di <a LYF049 .LYF047 YJ5038f YJA729 o o oi l YJ5020 !oYJA730 YJ5039 YJA731 YJ5021' YJA732 LYF039< YJA733 22

~

LYZO121 1

i l- I i l e e e

, l $

LYF075 LYZ100 YJ5050!!LYZO88 YJA740 LYZ000fYJA739 LYZ140 LYZ004 !fLYZ016YJA741 'LYZ124 YJ5074 o- o o! e o e a o o o_ ____o- o __ _ _ _oj c o et a D. o LYF053' LYZ168 et in o. To o. o o YJ5068} YJA745 YJ5082i LYZ116 18 , YJ5022l 'YJA746 LYZ144I 'LYZ104 YJ5000: YJA747 LYZ000

{ ,

I t l l I I J l LYF000 LYZ132 YJ5084} YJ50 YJA751 YJA752 LYZO24! YJ5067 YJA753! !LYZ112 YJA754 16 '

LYZ064 YJA755!

j  !

oi o of a o1 o or o o l l ei el o al el o o, o at e o lo ai o o o LYF028! [LWO21 LYF011 LYZO60 YJ5079l YJ5020 LYZ100' LYZ160 YJ5030 YJA761 YJ5047 YJA762 LYZ028 o

i l l l l e e LWO27 LWO29 LYZO56lLYZ164 YJ5005 YJA765 LYZOO8 LYZ120 YJA766 LYZ145 YJ5078f o- 18 o o lo . ..-___o. o, o _] o o ai or fi~ o, fo~~ o fo o+ T~ or [o o LWOB8 LYF009 LYF020' LYZ072 YJ5026 YJA769 YJ5056, YJ5083 YJ5052 YJ5058 LYZO32 1  ! I I LWO51 LYF015 LYZO92 YJ5040 YJA771 YJA772 YJA773 YJA774 YJA775 e o el o a. o ni o o al f

LWOIS l LYF050 o o LYF077j YJ5036 el e YJ5046i LYZ136 YJ5059 o n YJ5069 YJ5075 o

1 ,

i 04 LYF052l LYZO48 LYZ156fLYZ076 LYF094 LWO59 LYF012{

ol o o, o ci o o, l

o. o oi Fo~~ o o o LYF038I LYF095 LYF014 LYF048 LYF058 l LYF086 LYF076 I l i:

Page 16 of 20 STARRTI.IXX'

STARTUP TEST REPORT PILGRIM NUCLEAR POWER STATION l ,

CYCLE 11 TABLE I-A SCRAM INSERTION TIMES FOR AVERAGE OF ALL RODS IN CORE (TECHNICAL SPECIFICATION 3.3.C.1)

TECHNICAL I

PERCENT INSERTED MEASURED SCRAM SPECIFICATION SCRAM FROM FULLY INSERTION TIME, INSERTION TIME, WITHDRAWN SECONDS SECONDS 10 0.49 0.55 30 0.99 1.275 50 1.50 2.00 90 2.57 3.50 TABLE I-B SCRAM INSERTION TIMES FOR AVERAGE OF THREE FASTEST RODS IN EACH GROUP OF FOUR (TECHNICAL SPECIFICATION 3.3.C.2) i TECHNICAL i PERCENT INSERTED MEASURED SCRAM SPECIFICATION SCRAM FROM FULLY INSERTION TIME, INSERTION TIME, WITHDRAWN SECONDS SECONDS 10 0.54 0.58 30 1.09 1.35 >

50 1.62 2.12 90 2.73 3.71 1

I Page 17 of 20 ,

STARTUPl. DOC

l STARTUP TEST REPORT PILGRIM NUCLEAR POWER STATION

, CYCLEll TABLEII THERMAL LIMITS CALCULATED BY P-1 FOLLOWING EXECUTION OF OD-1 i

TIME MAPRAT DATE  % CTP  % WT MFLCPR MFLPD 6-09-95 00:44 38 47 0.671 0.362 0.704 6-10-95 11:46 49 54 0.824 0.458 0.644 4

6-11-95 15:16 70 57 0.923 0.586 0.691 6-13-95 21:32 96 60 1.015 0.837 0.852 6-18-95 11:53 81 65 0.980 0.649 0.722 6-18-95 18:40 86 71 0.970 0.688 0.803 6-19-95 03:18 96 86 0.984 0.774 0.790

, 6-21-95 15:26 100 97 0.997 0.794 0.794 6-22-95 16:32 100 98 0.988 0.793 0.794 6-23-95 11:26 100 99 0.967 0.800 0.801 7-12-95 10:27 100 97 0.968 0.791 0.802

! 8-01-95 13:55 100 93 0.979 0.791 0.809

) 1 i

A TABLE III-A COMPARISON OF MFLCPRS CALCULATED BY PANACEA AND Pl l

DATE TIME  % CTP  % WT PANACEA Pl DELTA 6-16-95 10:43 94 102 0.87 0.95 0.08 6-21-95 11:59 100 98 0.90 0.99 0.09 6-27-95 9:42 99 100 0.90 0.96 0.06 7-15-95 6:27 100 97 0.90 0.97 0.07 7-22-95 7:40. 98 91 0.91 0.96 0.05 7-29 45 8:65 100 92 0.90 0.97 0.07 8-02-95 13:37 100 94 0.90 0.98 0.08 Page 18 of 20 STARTUPl.!XX'

=

STARTUP TEST REPORT PILGRIM NUCLEAR POWER STATION CYCLEll TABLE III-B COMPARISON OF MFLPDS CALCULATED BY PANACEA AND P1

.DATE TIME  % CTP  % WT PANACEA Pl DELTA 6-16-95 10:43 94 102 0.78 0.75 -0.03 ,

6-21-95 11:59 100 98 0.83 0.81 -0.02 6-27-95 9:42 99 100 0.83 0.80 -0.03 7-15-95 6:27 100 97 0.84 0.79 -0.05 '

7-22-95 7:40. 98 91 0.82 0.78 -0.04 7-29-95 8:65 100 92 0.83 0.78 -0.05 8-02-95 13:37 100 94 0.83 0.79 -0.04 i

t i

o TABLE III-C '

COMPARISON OF MAPRATS CALCULATED BY PANACEA AND P1 i

DATE TIME  % CTP  % WT PANACEA Pl DELTA 6-16-95 10:43 94 102 0.76 0.77 0.01 6-21-95 11:59 100 98 0.80 0.81 0.01 6-27-95 9:42 99 100 0.81 0.80 -0.01 7-15-95 6:27 100 97 0.82 0.80 -0.02 7-22-95 7:40. 98 91 0.80 0.79 -0.01 7-29-95 8:65 100 92 0.81 0.79 -0.02 l 8-02-95 13:37 100 94 0.81 0.80 -0.01 i

B Page 19 of 20 stranw.noc l

l 4 P

STARTUP TEST REPORT PILGRIM NUCLEAR POWER STATION  !

CYCLE ll TABLE IV  :

THERMAL L.IMITS CALCULATED BY 3D-MONICORE DATE TIME  % CTP  % WT MFLCPR MFLPD MAPRAT 6-10-95 11:59 48 53 0.762 0.444 0.614 6-11-95 14:59 70 57 0.848 0.591 0.684 6-14-95 02:59 95 100 0.943 0.815 0.821 6-18-95 11:59 81 64 0.900 0.675 0.746 6-18-95 18:59 86 72 0.911 0.701 0.744 6-19-95 05:59 99 92 0.947 0.822 0.801

• 6-21-95 14:59 100 98 0.935 0.821 0.806 6-22-95 15:59 100 99 0.928 0.818 0.795 6-23-95 11:59 100 99 0.929 0.817 0.793 7-12-95 09:59 100 96 0.931 0.811 0.795 8-01-95 07:59 100 94 0.945 0.809 0.814 I

C

, TABLE V HOT EXCESS REACTIVITY (IN EQUIVALENT NOTCHES ADJUSTED TO RATED REACTOR DOME PRESSURE, RATED CORE INLET FLOW RATE AND NOMINAL i

CORE INLET SUBCOOLING)

EXPECTED OBSERVED i DATE NOTCIIES NOTCHES DELTA NOTCHES 6-23-95 600 585 -15 7-17-95 580 586 +06 i

i i

Page 20 of 20 STAP.TUPl. DOC 1