�Entergy Entergy Nuclear Generation Co. Pilgrim Station 600 Rocky Hill Road Plymouth, MA 02360

10 CFR 50.73August 14, 2001 ENGC Ltr. 2.01.079

U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, D.C. 20555

Docket No. 50-293 License No. DPR-35

Startuo Test Ren)ort Pilarim Nuclear Power Station Cycle 14

Enclosed is the Startup Test Report for Pilgrim Nuclear Power Station Cycle 14, submitted in accordance with Pilgrim Updated Final Safety Analysis Report Section 13.5.7. If additional information is required, please contact P.M. Kahler at (508) 830-7939.

Sincerely,

°W.J. Riggs

cc: Mr. Douglas Starkey, Project Manager Office of Nuclear Reactor Regulation Mail Stop: 0-8B-1 U. S. Nuclear Regulatory Commission 1 White Flint North 11555 Rockville Pike Rockville, MD 20852

U.S. Nuclear Regulatory Commission Region 1 475 Allendale Road King of Prussia, PA 19406

Sr. NRC Resident Inspector

/201079

William J. Riggs Director, Nuclear Assessment

STARTUP TEST REPORT, PILGRIM NUCLEAR POWER STATION, CYCLE 14

TABLE OF CONTENTS

SECTION ITEM PAGE A INTRODUCTION 2 B SUMMARY 2 C GE14 BUNDLE DESIGN 3 D CYCLE 14 CORE DESIGN 4 E CORE VERIFICATION 5 F CONTROL ROD COUPLING INTEGRITY 6 G CONTROL ROD SCRAM TIME TESTING 6 H SHUTDOWN MARGIN 6 I ESTIMATED CRITICAL POSITION 6 J INSTRUMENT CALIBRATIONS 7 J.1 APRMs 7 J.2 LPRMs 7 J.3 TIPs 7 J.4 JET PUMP FLOW INSTRUMENTATION 8 K PROCESS COMPUTER DATA PROCESSING CHECKS 8 L THERMAL LIMITS 9 M HOT EXCESS REACTIVITY 9 N ADDITIONAL TESTING 9

FIGURES AND TABLES FIGURE 1, PILGRIM CYCLE 14 CORE LOADING MAP BY BUNDLE 10 TYPE FIGURE 2, CYCLE 14 CORE LOADING MAP BY BUNDLE SERIAL 11 NUMBERS TABLE I-A, SCRAM INSERTION TIMES FOR AVERAGE OF ALL 12 RODS IN CORE TABLE I-B, SCRAM INSERTION TIMES FOR AVERAGE OF THREE 12 FASTEST RODS IN EACH GROUP OF FOUR TABLE II, THERMAL LIMITS CALCULATED BY 3D MONICORE 13 FIGURE 3, ROD PATTERNS CMR VS. ACTUAL 13 TABLE III, COMPARISON OF THERMAL LIMITS IN CMR vs. 3D 14 MONICORE TABLE IV, HOT EXCESS REACTIVITY 15 FIGURE 4, CYCLE 14 REACTIVITY ANOMALY CURVE 15

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STARTUP TEST REPORT, PILGRIM NUCLEAR POWER STATION, CYCLE 14

A. INTRODUCTION

Final Safety Analysis Report (FSAR) Section 13.5.7 requires that a summary report of plant startup testing be submitted to the NRC within 90 days following the start-up of the new cycle in which a new fuel type of a different design is loaded. The Pilgrim Station Cycle 14 reload batch is based on the General Electric (GE)14 fuel type with a 10 x 10 lattice. Compared to 8 x 8 and 9 x 9 lattice geometries that were loaded in previous Pilgrim reloads, the GE14 reload batch for Cycle 14 constitutes a different fuel design. General Electric has supplied all fuel loaded at Pilgrim Station since commercial operation began in 1972.

This report satisfies the FSAR 13.5.7 start-up report requirement.

B. SUMMARY

A reload batch of 144 GE14 fuel bundles with a bundle-average enrichment of 4.12% was loaded in the Pilgrim Cycle 14 core to provide a cycle energy capability of 677 effective full-power days (EFPD). This reload batch constitutes the first use of GE14 10 x 10 fuel at Pilgrim Station.

As-loaded Cycle 14 core maps showing fuel loading by bundle type and by bundle serial number are presented in Figures 1 and 2, respectively. The Cycle 14 core loading is octant symmetric and is based on both the low-leakage and control-cell-core design principles. The Cycle 14 core design is documented in the Pilgrim Plant Design Change Package (PDC) 01-03, "Cycle 14 Core Design".

The final as-loaded core loading was verified to be consistent with the design core loading by Pilgrim personnel on May 8th 2001. Core loading verification following refueling was performed in accordance with the requirements of PNPS 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 Technical Specification SR 4.3.B.1.3 in accordance with PNPS Procedure 9.13, "Control Rod Sequence and Movement 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 on May 21, 2001 prior to exceeding 40% of rated core thermal power.

Shutdown margin (SDM) was demonstrated to be consistent with the requirements of Technical Specification 3.3.A.1 by the insequence critical method which was performed on May 17, 2001 following initial criticality. This demonstration was performed in accordance with the requirements of PNPS Procedure 9.16.1, "Insequence Critical for Shutdown Margin Demonstration". Single Rod Subcriticality tests were performed in accordance with PNPS Procedure 9.16.2, "Subcritical Demonstration" as a good practice in order to verify that the reactor will remain shutdown with the strongest rod withdrawn prior to performing the Shutdown Margin (SDM) test.

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STARTUP TEST REPORT, PILGRIM NUCLEAR POWER STATION, CYCLE 14

Calibration of instrumentation important to monitoring core thermal power and core margins to thermal limits was performed as required by PNPS procedures and Technical Specifications (TS). This instrumentation includes Average Power Range Monitors (APRMs), Local Power Range Monitors (LPRMs), Traversing Incore Probe (TIPs) and Jet Pump Flow Indicators. Calibration of this instrumentation was performed in accordance with the relevant PNPS procedures.

Process Computer data processing checks were completed consistent with the requirements of PNPS Procedure 9.34.

Margins to thermal limits calculated by 3D MONICORE were consistent with the margins calculated by Global Nuclear Fuel Inc. (GNF) using PANACEA 11 (P1 1) and documented in the Cycle Management Report (CMR).

Hot excess reactivity of the Cycle 14 core was found to be consistent with the CMR. Hot excess reactivity was determined in accordance with PNPS Procedure 9.8, "Reactivity Follow".

Refueling Outage (RFO) 13 officially ended on May 17, 2001 when Pilgrim Station went on line after a refueling outage of 28 days. Rated power was reached on May 26, 2001.

C. GE14 Bundle Design

The Cycle 14 core design is based on the GE14 advance fuel type. The GE14 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 GE14 fuel type continues these basic trends through a number of key design features: a 10 x 10 lattice geometry, 14 part length rods, two large central water rods, 10 atmospheres of helium pre-pressurization, greater plenum space, absence of the hydrogen getter, 8 high performance ferrule spacers, a high pressure drop lower tie plate with a built in debris filter and a low pressure drop upper tie plate.

Maximum lattice average enrichment is 4.57%. Including applicable manufacturing tolerances, the lattice enrichment meets the Pilgrim spent fuel pool storage rack Technical Specification 4.3.1.1 .a limit of 4.6 wt % averaged over the planar zone of highest enrichment. Maximum bundle Koo is 1.27, which meets the Ko- limit of 1.32 (TS 4.3.1.1 .a). Standby Liquid Control (SLC) SDM based on 675 PPM boron and cycle energy of 677 EFPD was determined to be 3.37% Ak at 1600 C analysis temperature. Evaluation of SLC SDM is more conservative and accurate at 1600 C than at 200 C. Therefore, design criteria of 3% Ak at 200 C as stated in TS bases 3/4.4.A and FSAR 3.8 has been met.

Maximum Peak Pellet Exposure (PPE) for both GEl1 and GE14 fuel bundles is 63.5 GWD/ST (70 GWD/MT). GE14 Bundles contain 174.5 kg of Uranium with an average 4.12 wt. % U-235. Max. Bundle Koo is reached at approximately 13 GWD/ST. Active fuel length is approximately 4" longer than GEl1. External dimensions of both GEl1 and GE14 fuel bundles are identical. GE14 is basically an improved GE12 design. GE12 has been successfully operated at Fitzpatrick, Perry and many other international nuclear plants.

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STARTUP TEST REPORT, PILGRIM NUCLEAR POWER STATION, CYCLE 14

GE14 was licensed under the GESTAR II amendment 22 process in 1998. The GE14 design consists of 92 fuel rods and two large central water rods contained in a 10 x 10 array. The two water rods encompass eight fuel rod positions. Fourteen of the fuel rods terminate just past the top of the 5th spacer and are designated as partial length fuel rods. Eight fuel rods are used as tie rods. There are 78 full-length fuel rods in the GE14 lattice. Gadolinium (Gd) rods are 6" shorter than full-length rods to accommodate additional fission gas. The rods are spaced and supported by the upper and lower tie plates and eight spacers over the length of the fuel rods. This assembly is encased in an interactive fuel channel, which has been used on the GEl0, GEl 1, GE12, and GE13 designs. Finger springs control the coolant leakage flow between the lower tie plate and the channel.

GE14 was designed for mechanical, nuclear, and thermal-hydraulic compatibility with the other GE fuel designs. The design includes many features of the GEl0, GEl 1, and GE12 fuel designs including optional PCI resistant barrier cladding, high performance spacers, partial length rods, interactive thick corner/thin wall channel, and axial and enrichment loading. The fuel rods consist of high-density ceramic uranium dioxide or urania-gadolinia fuel pellets stacked within Zircaloy-2 cladding. The cladding has an inner 3/5 mil thick zirconium liner, similar to the GEl 1 design, for mitigation of Pellet Clad Interaction. The fuel rod is evacuated and pressurized with helium to 10 atmospheres. New improvements included in the GE14 are:

0 Optimized spacer positions and part length rod (PLR) length.

* Eight High Performance Zr-2 Spacers.

* Debris filter lower tie plate as standard equipment.

* Elimination of the Hydrogen Getter

* Increased pellet density from 96.5% to 97% of theoretical

* The full-length fuel rods are on the average 4" longer than those in GEl 1 fuel.

* 14 PLRs compared to 8 PLRs in GEl 1 design.

D. CYCLE 14 CORE DESIGN

The Cycle 14 core was designed to provide 677 EFPDs of cycle energy capability. The Cycle 14 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.

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STARTUP TEST REPORT, PILGRIM NUCLEAR POWER STATION, CYCLE 14

The low-leakage design principle preferentially loads twice-burned and thrice-burned burned low-reactivity fuel on the core periphery to reduce radial neutron leakage, thereby yielding improved fuel cycle efficiency 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 Core design meets both the TS 4.3.A.1 SDM limit of 0.68% Ak (R + 0.25 = 0.43 + 0.25 = 0.68 % Ak) and the FSAR 3.6 requirement of 1% Ak SDM. An SDM demonstration on 5/17/01 proved that the FSAR and TS requirements were met and the SDM was consistent with the design SDM. The Core design achieves steady state thermal limit margin 9% for MFLCPR, 11% for MAPRAT, and 13% for MFLPD. These margins are greater than the GE/GNF standard design margins and provide adequate assurance that limits specified in the Core Operating limits will be met during Cycle 14 operation.

With this reload batch, the inventory of fuel in the Pilgrim Cycle 14 core is:

Number Cycle of Bundles Bundle Type Loaded

144 GE14-Pl ODNAB-412-16GZ-1 OOT-145-T6-3901 14 120 GEl 1 -P9DUB407-14GZ1-100T-141 -T 13 40 GEl 1 -P9DUB408-6G5.0/7G4.0-1 OOT-1 41 -T 13 64 GEl 1 -P9DUB408-6G5.0/7G4.0-1 00T-1 41 -T 12 144 GE11-P9DUB408-16GZ1-100T-141-T 12 68 GEl 1 -P9HUB378-15GZ-100T-141 -T 11

Figures 1 and 2 present the as-loaded Cycle 14 core maps showing fuel loading by bundle type and bundle serial number, respectively.

The Cycle 14 core is loaded to be octant symmetric by both fuel type and, with a small allowance for variance, bundle exposure. The Cycle 14 core design is documented in Pilgrim Plant Design Change Package (PDC) 01-03, "Cycle 14 Core Design".

The Cycle 14 core design meets all licensing criteria specified in Revision 14 of NEDE-24011 -P-A and NEDE-24011 -P-A-US, the General Electric Standard Application for Reactor Fuel (GESTAR-II).

E. CORE VERIFICATION

The final as-loaded Cycle 14 core loading was verified on May 8, 2001, consistent with the requirements of PNPS Procedure 4.5. Three separate criteria were verified: bundle orientation, bundle seating, and bundle location.

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

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STARTUP TEST REPORT, PILGRIM NUCLEAR POWER STATION, CYCLE 14

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

Bundle location was verified by observing that bundle serial numbers in the core were

consistent with bundle serial numbers in the final fuel loading plan.

Verification of the final as-loaded Cycle 14 core loading identified no core loading errors.

F. CONTROL ROD COUPLING INTEGRITY

Control Rod coupling integrity was verified whenever a Control Rod was fully withdrawn as required by TS SR 4.3.B.1.3. Coupling Integrity was established by observing that the rod would not reach its over-travel position upon withdrawal.

G. CONTROL ROD SCRAM TIME TESTING

Single rod scram time testing on all 145 Control Rods was successfully completed on May 21, 2001, prior to exceeding 40% of rated core thermal power as required by Technical Specification 4.3.C.1. Results of this testing met TS 3.3.C.1 and 3.3.C.2 limits.

Control Rod scram time testing was performed in accordance with PNPS Procedure 9.9, "Control Rod Scram Time Evaluation". Table I-A and I-B list the average scram times and the average of the 3 fastest times in 2 x 2 arrays.

H. SHUTDOWN MARGIN

SDM was demonstrated using the insequence critical method in accordance with PNPS Procedure 9.16.1, on May 17, 2001.

Minimum Cycle 14 SDM of 1.02 Ak was demonstrated on May 17, 2001. This agrees with the design SDM of 1.05% Ak. Beginning of Cycle (BOC) shutdown margin was demonstrated to be 1.45% Ak vs. the design BOC SDM of 1.78% Ak. This was within the expected range considering the differences in actual vs. design End of Cycle (EOC) 13 exposure. TS 4.3.A.1 requirement of 0.68% Ak (0.25% + R, where R is 0.43% Ak for Cycle 14) SDM and FSAR 3.6 requirement of 1% Ak SDM are therefore met.

Core Subcriticality upon withdrawal of the strongest single Control Rod was demonstrated in

accordance with PNPS Procedure 9.16.2, on May 8, 2001.

I. ESTIMATED CRITICAL POSITION

The actual critical position for Control Rods was found to be within design expectations. Initial criticality was estimated to occur when the last rod in Group 2 was pulled from Notch 12 to Notch 48 in an A-2 Sequence. Moderator temperature was assumed to be 1800 F.

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STARTUP TEST REPORT, PILGRIM NUCLEAR POWER STATION, CYCLE 14

Initial criticality was actually realized when the 10th rod in Group 3 was pulled to Notch 10. Control Rods were withdrawn in an A-2 Sequence and the average moderator temperature was 201 0 F. The reactor period was 232 seconds. These results are consistent with expectations based on higher moderator temperature and the low worth of the Group 3 rods from 00 to 08.

J. INSTRUMENT CALIBRATIONS

J.1 APRMs

APRMs were calibrated as required during the power ascension to maintain the APRM gain adjustment factors (AGAFs) between 0.87 and 1.00. AGAFs are the ratio of the actual reactor power reading as determined from a reactor heat balance to the APRM reading.

All APRM calibrations were performed consistent with the requirements of PNPS Procedure 9.1, "APRM Calibration".

The 40% core flow switch settings on the APRM Flow Control Trip Reference (FCTR) Cards were adjusted to be conservative on May 21, 2001, as required by ElA stability methodology in accordance with PNPS Procedure 9.17.1. 100% Flow Setting was not adjusted immediately after achieving rated power, as this setting was believed to be conservative and the Jet Pump and Core Flow instrumentation that is used as inputs to the 100% Flow Switch setting calculation was found to have acceptable, but higher than normal, calibration inaccuracy. After calibrating the Jet Pump instrumentation the high flow switch setting was calculated at 80% core flow conditions and set to be conservative. Subsequent testing in accordance with PNPS Procedure 9.17.1 at the 100% rated Core Flow condition has demonstrated the original BOC14 high flow switch setting to be conservative

J.2 LPRMs

LPRMs were calibrated as required by PNPS Procedure 9.5, "LPRM Calibration". LPRMs are calibrated to maintain gain adjustment factors (GAFs) between 0.95 and 1.05. GAFs are the ratio between the actual LPRM console readings and the desired LPRM console readings.

LPRMs were first calibrated in Cycle 14 on May 21, 2001. Subsequent calibrations were

performed on May 23, 2001, and July 6, 2001.

J.3 TIPs

The Traversing Incore Probe (TIP) system was used as needed to update the BASE array and calibrate LPRMs, PNPS Procedure 9.5. TIP detector "A" failed during startup and was replaced at 100% power. TIP runs were executed and TIP "A" was calibrated consistent with the requirements of PNPS Procedure 9.5.1, "Operation of Tip Machines for Process Computer Updating".

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STARTUP TEST REPORT, PILGRIM NUCLEAR POWER STATION, CYCLE 14

J.4 Jet Pump Flow Instrumentation

Jet pumps and recirculation drive flow were calibrated as required during the power ascension consistent with the requirements of PNPS Procedure 9.17.

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

Jet Pump Square Rooter calibration was checked subsequent to start-up based on observed jet pump signatures and core flow calibration results. Execution of PNPS Procedure 9.17, demonstrated acceptable accuracy of core flow measurements at 50%, 75%, and 100% power on May 23, 24 and 29, 2001, respectively. Testing results indicated that the GE14 fuel design had minimal effect upon the overall core pressure drop or the core flow to recirculation flow relationship. Rated drive flow was found to be approximately the same as that in the previous cycle. Due to the number and length of partial length rods in GE14 fuel, two phase pressure drop was expected to be lower than that of the older GEl 1 fuel. This would lead to an expectation of lower rated drive flow; however, this effect was not pronounced due to the small number of GE14 bundles in the core in the first reload of GE14.

K. PROCESS COMPUTER DATA PROCESSING CHECKS

The core monitoring system was modified to install the PANAC 11 version of 3D MONICORE. Prior to installation, PANAC 11 was benchmarked to the existing PANAC 10 software for Cycle 13 and previous cycles of operation at Pilgrim Station. Thermal limits were monitored during startup and compared to off-line results. The thermal limits comparison between actual and design performance is discussed in section L below.

The 3D MONICORE Process Computer databank was updated consistent with the requirements of PNPS Procedure 9.34.

A number of checks are specified by PNPS Procedure 9.34, 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 May 6, 2001.

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

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STARTUP TEST REPORT, PILGRIM NUCLEAR POWER STATION, CYCLE 14

L. THERMAL LIMITS

Thermal Limits Calculated by 3D MONICORE

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 3D MONICORE core monitoring software. Margins to thermal limits were maintained as required by TS. Thermal limits were reviewed to establish compliance with the TS and Core Operating Limits Report in accordance with PNPS Procedures 9.19, "Minimum Critical Power Ratio Evaluation" and 9.22, "Maximum Average Planar Linear Heat Generation Rate" at 25%, 50%, 75%, and 100% power level on May 21, 22, 24 and 25, respectively. Table II shows thermal limits calculated by 3D MONICORE at power levels from 30% to 100%.

The 3D MONICORE power distribution was updated as required during the power ascension

using the TIP system at 50%, 80%, and 100% power during the ascent to rated power.

Thermal Limits Calculated by PANACEA

3D MONICORE calculated thermal limits (PANAC1 1) were consistent with off-line thermal limits calculated by the GE's PANAC 11 design code as documented in Table Ill.

M. HOT EXCESS REACTIVITY

The actual Control Rod notch inventory (adjusted to reflect rated reactor dome pressure, rated core inlet flow rate and nominal core inlet sub-cooling) was verified to be consistent with the design notch inventory on June 5, June 26, and July 17, 2001. Table IV presents both the actual and design Control Rod notch inventories for these dates. The acceptance criterion for this comparison is an actual Control Rod notch inventory that differs from the design notch inventory by no more than ± 1 % Ak reactivity. Notches corresponding to ± 1 % Ak reactivity limits is defined by the CMR. Figure 4 shows a plot of the rated notches as a function of exposure and demonstrates that the Cycle 14 operation during start-up is within ± 1 % Ak reactivity limits in the CMR and therefore meets TS 3.3.E.

Monitoring of hot excess reactivity is governed by PNPS Procedure 9.8.

N. ADDITIONAL TESTING

The GE14 fuel loaded in Cycle 14 required no modifications to plant systems or components. Accordingly, the first reload of GE14 fuel at Pilgrim requires no testing during start-up beyond that normally performed to assure compliance with TS. These test results have been presented in the sections above as required by FSAR 13.5.7.

Page 9 of 17

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Page 11 of 17

52

50

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xis 288

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yJH 852 ya's

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A

STARTUP TEST REPORT, PILGRIM NUCLEAR POWER STATION, CYCLE 14

TABLE I-A SCRAM INSERTION TIMES FOR AVERAGE OF

ALL RODS IN CORE (TECHNICAL SPECIFICATION 3.3.C.1)

DROP OUT OF NOTCH MEASURED SCRAM TECHNICAL SPECIFICATION POSITION FROM FULLY INSERTION TIME, SCRAM INSERTION TIME,

WITHDRAWN SECONDS SECONDS

44 0.48 0.508 34 1.01 1.252 24 1.57 2.016 04 2.72 3.578

TABLE I-B SCRAM INSERTION TIMES FOR AVERAGE OF

THREE FASTEST RODS IN EACH GROUP OF FOUR (TECHNICAL SPECIFICATION 3.3.C.2)

DROP OUT OF NOTCH MEASURED SCRAM TECHNICAL SPECIFICATION POSITION FROM FULLY INSERTION TIME, SCRAM INSERTION TIME,

WITHDRAWN SECONDS SECONDS

44 0.50 0.538 34 1.07 1.327 24 1.65 2.137 04 2.87 3.793

Page 12 of 17

STARTUP TEST REPORT, PILGRIM NUCLEAR POWER STATION, CYCLE 14

THERMALTABLE II

LIMITS CALCULATED BY 3D-MONICORE

DATE TIME % CTP % WT MFLCPR MFLPD MAPRAT

5/20/01 21:00 20 38 0.467 0.475 0.521 5/21/01 06:00 31 46 0.543 0.540 0.619 5/22/01 02:00 47 61 0.574 0.603 0.694 5/22/01 09:39 50 59 0.585 0.591 0.682 5/24/01 07:54 75 61 0.770 0.710 0.820 5/24/01 19:00 80 61 0.801 0.708 0.808 5/24/01 22:57 91 76 0.791 0.700 0.803 5/25/01 02:36 100 79 0.808 0.721 0.822 5/25/01 19:00 99.5 86 0.774 0.681 0.764 5/26/01 23:00 99.5 93 0.751 0.681 0.76 5/29/01 23:00 99.4 94 0.738 0.684 0.749 5/31/01 23:00 99.6 96 0.736 0.687 0.748 6/2/01 06:00 99.6 79.1 0.824 0.769 0.884 6/3/01 14:00 99.6 83.2 0.804 0.759 0.850

Steady state rod pattern was achieved

FIGURE 3 ROD PATTERNS CMR VS. ACTUAL

BOC 14 Rod Pattern in CMR Actual Rod Pattern 511 51 47 47 43 36 10 36 43 T 40101 40 391 39 35 36 10 36 1036 35 40 8 38 8 40 311 31 27 10 36 10 36 10 27 1038 10 38 10 23 23 19 10 36 1036 19 8 408 38 8 151 15 11 36 10 - 36 11 40 10 40

7 7 3 3

2 6 1014 1822 26 303438 42 4650 2 6 1014 1822 2630 34 384246 50

Page 13 of 17

STARTUP TEST REPORT, PILGRIM NUCLEAR POWER STATION, CYCLE 14

TABLE III COMPARISON OF THERMAL LIMITS IN CMR VS. 3D MONICORE

Page 14 of 17

DATE MWD/ST % CTP % WT MFLCPR MFLPD MAPRAT EIGEN VALUE

CMR 200 100 83.3 0.786 0.758 0.816 1.013 6/3/01 226 99.4 83.2 0.804 0.759 0.850 1.016 DELTA -0.4% -0.1% +2.3% +0.13% +4.2%

CMR 600 100 89.3 0.783 0.746 0.807 1.013 6/25/01 600.3 99.6 85.5 0.793 0.765 0.866 1.011 DELTA -0.4% -4.3% +1.3% +2.55% +7.3%

CMR 1000 100 89.3 0.784 0.734 0.799 1.013 7/16/01 1000.7 99.6 81.2 0.817 0.740 0.850 1.0107 DELTA -0.4% -9.1% +4.2% +0.8% +6.4%

Actual Rod pattern was slightly different from that assumed in the Cycle Management Report in that Group 16 rods were withdrawn to 40 instead of 36 and Group 17 rods were withdrawn to 38 instead of 36. Also Group1 4 rods were withdrawn to 08 instead of 10. The above differences are considered acceptably close considering the differences in rod pattern and core flows at 600 & 1000 MWD/ST. Hot critical eigen value was 1.0116 at BOC and gradually reduced to 1.0107 vs. 1.0113 in design. Therefore, core performance is acceptably close to design upon Cycle 14 startup.

STARTUP TEST REPORT, PILGRIM NUCLEAR POWER STATION, CYCLE 14

TABLE IV HOT EXCESS REACTIVITY

(IN EQUIVALENT NOTCHES ADJUSTED TO RATED REACTOR DOME PRESSURE, RATED CORE INLET FLOW RATE AND NOMINAL CORE INLET SUBCOOLING)

EXPECTED OBSERVED +1% AK -1% AK NOTCHES NOTCHES DELTA BOUND BOUND

DATE INSERTED INSERTED NOTCHES NOTCHES NOTCHES 6/5/01 550 585 +35 810 290 6/26/01 526 593 +67 789 263 7/17/01 503 602 +99 765 240

FIGURE 4, CYCLE 14 REACTIVITY ANOMALY CURVE

CYCLE 14 REACTIVITY ANOMALY CURVE

100. 1+1% DKI

800 - . _ _.

S1 60• notches

LI

z Cl)

-- - -1% DK

rr

O0 40000

.0 0

20 '0 40 00 6C0 0 80 30 10,)00 £1200 147 )0

"-200 -e-- Rated Notches

-400 CYCLE EXPOSURE (MWD/ST)

Page 15 of 17