LR-N05-0257, General Electric Fuel (GE14) Startup Report

General Electric Fuel (GE14) Startup Report

ML051390175
Person / Time
Site:	Hope Creek
Issue date:	05/12/2005
From:	Perino C Public Service Enterprise Group
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
LR-N05-0257
Download: ML051390175 (13)

Text

PSEG Nuclear LLC

v. -P.O. Box 236, Hancocks Bridge, New Jersey 08038-0236 MAY 1 2 2005 o PSEG NuclearLLC LR-N05-0257 United States Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 GENERAL ELECTRIC FUEL (GE14) STARTUP REPORT HOPE CREEK GENERATING STATION FACILITY OPERATING LICENSE NPF-57 DOCKET NO. 50-354 PSEG Nuclear LLC hereby submits a summary report of plant startup for the Hope Creek Generating Station in accordance with the requirements of Technical Specification 6.9.1.1. This report is required since fuel manufactured by General Electric was loaded for Cycle 13. The report is included as Attachment 1.

If you have any questions or comments on this transmittal, please contact Michael Mosier at (856) 339-5434.

Sincerely, Christina L. Perino Regulatory Assurance Director Attachment 4Omo 95-2168 REV. 7/99

LR-N05-0257 2 MAY1 2 Document Control Desk 2005 C Mr. S. Collins, Administrator - Region I U. S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406 Mr. D. Collins, Project Manager- Hope Creek, Salem U. S. Nuclear Regulatory Commission Mail Stop 08C2 Washington, DC 20555 USNRC Senior Resident Inspector - Hope Creek (X24)

Mr. K. Tosch, Manager IV Bureau of Nuclear Engineering P. O. Box 415 Trenton, NJ 08625

ATTACHMENT 1 Hope Creek Generating Station Cycle 13 Startup Report April 2005

Table of Contents

1.0 INTRODUCTION

.......................................................... 3 2.0 CONTROL ROD DRIVE SYSTEM .......................................................... 3 2.1 CONTROLROD SCRAM TIME ........................................................... 3 3.0 FULL CORE SHUTDOWN MARGIN .......................................................... 4 3.1 IN-SEQUENCE CRITICALS .......................................................... 4 3.2 SHUTDOWN MARGIN DEMONSTRATION ................................. ...................................... 5 3.3 CORE COLD REAcrIvITY ANOMALY EVALUATION ...................................................................... 5 4.0 CORE PERFORMANCE .......................................................... 5 4.1 CORE HOT REACTIVITY ANOMALY EVALUATION ...................................................................... 5 4.2 THERMALLiMiTs ....................................................................... 6 4.3 CORETHERMALHYDRAULIC EVALUATION . ...................................................................... 6 4.3.1 Core Support Plate Pressure Drop Comparison ................................................................. 7 4.3.2 Cycle 13 and 12 Measured Core Support Plate Pressure Drop Comparison ........... .......... 8 4.3.3 Cycle 13 and 12 Recirculation Pump Data and Core Flow Comparison ............ ............... 9 5.0 NSSS PROCESS COMPUTER ..................... ..................................... 11

6.0 REFERENCES

...................................................................................................... 11 List of Tables TABLE 1. INDIVIDUAL SCRAM TIM .. 3 TABLE 2. AVERAGE SCRAM TIMES .4 TABLE 3. ARRAY AVERAGE SCRAM TIMES .4 TABLE 4. IN-SEQUENCE CRITICAL RESULTS .5 TABLES. CMISTiiERMALLIMITS .6 TABLE 6. CYCLE MANAGEMENT REPORT PREDICTIONS TO CMS TlERMAL LIMITS COMPARISON. 6 TABLE 7. CORE SUPPORT PLATE PRESSURE DROP COMPARISON .7 List of Figures FIGURE 1. CALCULATED VERSUS MEASURED CORE SUPPORT PLATE PRESSURE DROP . 8 FIGURE 2. CYCLE 13 AND CYCLE 12 MEASURED CORE SUPPORT PLATE PRESSURE DROP ..................... 8 FIGURE3. CYCLE 13 AND 12 RECIRCULATIONPUMPA FLOWVERSUS PUMPPSPEED...............................9 FIGURE 4. CYCLE 13 AND 12 RECIRCULATION PUMP B FLOW VERSUS PUMP SPEED .9 FIGURE 5. CYCLE 13 AND 12 RECIRCULATION PUMp A HEAD VERSUS PUMP FLOW .1 0 FIGURE 6. CYCLE 13 AND 12 RECIRCULATION PUMP HEAD ] VERSUS PumP FLOW .10 FIGURE 7. CYCLE 13 AND 12 CORE FLOW VERSUS RECIRCULATION PUMP FLOW .1 0 Page 2 of 11

1.0 Introduction Hope Creek Generating Station transitioned from the Westinghouse BWR SVEA 96+

fuel design to the General Electric GE14 fuel design in Cycle 13. The GE14 fuel is a lOxI Odesign with two large central water rods, consisting of 92 fuel rods. The SVEA 96+ fuel is a lOxlO water cross design consisting of 96 fuel rods. Hope Creek Technical Specification 6.9.1.1 requires a submittal of a startup report following the installation of fuel that has a different design, or has been manufactured by a different fuel supplier.

This startup report will address each of the initial startup tests identified in the Final Safety Analysis Report that could be impacted by the introduction of a new fuel design.

The fuel transition project was performed over a two-year period. During Hope Creek's twelfth refueling outage (RF12), that began on 10/24/2004 and was completed on 01/26/2005, 164 GE14 fuel bundles were loaded. Additionally, the core monitoring system (CMS) was replaced during RF12. The following sections provide a description of the test results for those initial startup tests described in the Hope Creek FSAR that were affected by the introduction of the GE14 fuel design (Reference 6.1).

2.0 Control Rod Drive System The description of the initial startup testing for the control rod drive system is provided in the Hope Creek FSAR section 14.2.12.3.5. The operability of the control rod system may be impacted by the introduction of a new fuel design. The new fuel design could cause additional friction on control rod movement, which may impact the scram speeds.

2.1 Control Rod Scram Time The control rod drive (CRD) scram times were measured in accordance with procedure HC.RE-ST.BF-0001(Q)," Control Rod Scram Time Surveillance". The objective of this test was to verify that the CRD scram times meet all Technical Specification acceptance criteria. The measured scram times were compared against acceptance criteria for the purpose of determining control rod drive system performance. The acceptance criteria for the individual scram time to notch position 05, core average scram times to notch positions 45, 39, 25, and 05, and 2x2 array average scram times to notch positions 45, 39, 25, and 05, are given in the Hope Creek Technical Specifications 3.1.3.2, 3.1.3.3, and 3.1.3.4 respectively. A summary of results from the test is provided in Tables 1, 2 and 3.

The results indicate that the measured scram times are faster than the acceptance criteria which demonstrates that the introduction of the GE14 fuel design did not have an adverse effect on control rod drive system performance.

Table 1. Individual Scram Time Notch Most Limiting Scram Acceptance criteria Position Insertion Time to Notch (Seconds) 05 (Seconds) 05 2.652 < 7.00 Page 3 of II

Table 2. Average Scram Times Notch Measured Core Average Acceptance criteria Position Scram Time (Seconds) (Seconds) 45 0.268 < 0.43 39 0.571 < 0.86 25 1.304 < 1.93 05 2.414 < 3.49 Table 3. Array Average Scram Times Notch Most Limiting 2x2 Acceptance criteria Position Scram Time (Seconds) (Seconds) 45 0.290 < 0.45 39 0.628 < 0.92 25 1.443 < 2.05 05 2.652 < 3.70 3.0 Full Core Shutdown Margin The description of the initial startup testing for the full core shutdown margin demonstration is provided in the Hope Creek FSAR section 14.2.12.3.4. The core neutronic characteristics, and the ability of the vendor design tools to accurately model the core in cold conditions may be impacted by the introduction of a new fuel design.

The Cycle 13 startup testing demonstrated that the shutdown margin was greater than 0.38% Ak/k, and the cold reactivity anomaly was within +/-1.0% Ak/k 3.1 In-Sequence Criticals The in-sequence critical was performed, by withdrawing the control rods in a Banked Position Withdrawal Sequence (BPWS), until criticality was achieved as part of the shutdown margin demonstration that was accomplished in accordance with procedure HC.RE-ST.ZZ-0007(Q), "Shutdown Margin Surveillance". The objective of the test was to evaluate the vendor's PANACI I methods used in the design and licensing of Cycle

13. The in-sequence critical test was performed on 01/18/2005 at a temperature of 164 0 F.

The results for the critical control rod configuration are shown in Table 4. The BOC13 cold target keff is also provided in Table 4. The results show that the difference between the BOC13 cold target keff that was established by the vendor's methods, and the cold critical keff calculated during the test is within the expected range observed from previous Hope Creek in-sequence critical calculations. The differences are acceptable, and are within the data used to establish the Cycle 13 shutdown margin design criteria.

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Table 4. In-sequence Critical Results Measurement PANACI I BOC Cold Target kfr 1.00100 In-sequence Critical kff 1.00126 3.2 Shutdown Margin Demonstration The core shutdown margin (SDM) was demonstrated in accordance with procedure HC.RE-ST.ZZ-0007(Q), "Shutdown Margin Surveillance". The objective of the test was to demonstrate that the core would remain subcritical by at least 0.38% Ak/k throughout the cycle at cold xenon free conditions, with the strongest worth control rod withdrawn.

The core SDM was demonstrated during the first in-sequence critical. The demonstrated SDM for Cycle 13 was 1.56% Ak/k, which meets the Technical Specification minimum requirement of 0.38% Ak/k.

3.3 Core Cold Reactivity Anomaly Evaluation The core reactivity anomaly was evaluated in accordance with procedure HC.RE-ST.ZZ-0005(Q),"Reactivity Anomaly Surveillance". The objective of the test was to demonstrate that the core reactivity is within +1.0% Ak/k of the predicted core reactivity.

The reactivity anomaly test was performed at cold conditions during the SDM demonstration. The predicted SDM at BOC was 1.65% Ak/k and the demonstrated SDM was 1.69% Ak/k, resulting in a difference of -0.04% Ak/k. The result from the test was within the Technical Specification requirement of 4-1.0% Ak/k.

4.0 Core Performance The description of the initial startup testing to evaluate the core performance, with respect to thermal limits, is provided in the Hope Creek FSAR section 14.2.12.3.16. The objective of the test is to calculate the principal thermal and hydraulic parameters associated with core behavior. The initial test evaluated the thermal limits at various power levels, and compared the thermal limits at rated power to the predicted values in the Cycle Management Report (Reference 6.2). The core performance tests and evaluations performed during the Cycle 13 power ascension were the hot reactivity anomaly evaluation, thermal limits evaluation and core thermal hydraulics evaluation.

4.1 Core Hot Reactivity Anomaly Evaluation The core reactivity anomaly was evaluated in accordance with procedure HC.RE-ST.ZZ-0005(Q),"Reactivity Anomaly Surveillance". The objective of the test was to demonstrate that the core reactivity is within +1.0% Ak/k of the predicted core reactivity.

The hot reactivity anomaly test was performed at 100% power equilibrium conditions at a cycle exposure of 308 Mwd/Mtu. The predicted keff was 1.0055, and the monitored ke.f from the CMS was 1.0079, resulting in a difference of -0.24% Ak/k. The result from the test was within the Technical Specification requirement of +1.0% Ak/k.

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4.2 Thermal Limits The thermal limits, given in Table 5, were obtained from the core monitoring system (CMS) during the BOC power ascension. The thermal limits were of an acceptable magnitude at each power and flow condition, and trended as expected for the actual power, flow and control rod pattern conditions experienced during the startup.

Table 5. CMS Thermal Limits Date/Time Power (%) Flow (%) MFLCPR MFLPD MAPRAT 1/27/200517:01 22.0 44.3 0.339 0.196 0.226 1/29/2005 03:02 30.3 45.2 0.446 0.275 0.321 1/29/2005 06:01 40.3 54.3 0.505 0.324 0.376 2/01/2005 00:01 50.4 65.9 0.549 0.433 0.514 2/01/2005 12:01 66.2 64.4 0.715 0.460 0.547 2/04/2005 21:31 75.0 79.3 0.677 0.521 0.603 2108/2005 07:32 90.3 90.9 0.773 0.646 0.742 2/12/2005 15:32 100.0 98.4 0.860 0.648 0.738 The thermal limits at full power conditions were compared against the predicted values from the Cycle Management Report (CMR) as shown in Table 6. The CMS data was obtained from an edit generated on 02/22/2005 22:02, at a cycle exposure of 500.1 Mwd/Stu. The differences are acceptable, and are within the Cycle 13 design margin criteria specified by PSEG.

Table 6. Cycle Management Report Predictions to CMS Thermal Limits Comparison CMR CMS ExposureMwd/St o500 500.1 MFLCPR 0.871 0.859 MFLPD 0.771 0.713 MAPRATj 0.670 0.624 4.3 Core Thermal Hydraulic Evaluation The introduction of the GE14 fuel design into the Hope Creek core has the potential to affect the thermal-hydraulic performance of the core. One of the vendor's thermal-hydraulic design bases is that the GE14 reload fuel shall be hydraulically compatible with the resident SVEA96+ fuel. The basis being that by ensuring hydraulic compatibility of the loaded fuel assemblies, the core thermal-hydraulic performance will remain unchanged by the introduction of the new fuel design.

The core thermal-hydraulic performance evaluation is comprised of the following activities:

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• A comparison of the measured and calculated Core Support Plate Pressure Drop.

• A comparison of the measured Core Support Plate Pressure Drop between Cycle 13 and Cycle 12 startup.

• A comparison of recirculation system loop data that was recorded during the startups of Cycle 12 and 13.

4.3.1 Core Support Plate Pressure Drop Comparison Steady state thermal-hydraulic calculations were performed with the computer code FJBWR2, and the CMS (3DMONICORE) at various operating conditions during the Cycle 13 startup (Reference 6.3). The FIBWR2 code and the vendor's thermal-hydraulic code were the design tools used to ensure hydraulic compatibility in the design of the Hope Creek Cycle 13 GE14 fuel assembly. A good comparison between the calculated and measured core support plate pressure drops provides evidence that the fuel assemblies are hydraulically compatible.

The operating conditions, measured data and calculated results are presented in Table 7.

The results show good agreement between the measured and calculated pressure drops.

FIB3WR2 and CMS calculated core support plate pressure drops are within 0.6 psid and 0.5 psid of measured data, respectively. Figure 1 presents a graphical representation of the results presented in Table 7.

Table 7. Core Support Plate Pressure Drop Comparison Power Flow Measured FIBWR2 3DMONICORE

_(04 of Rated)of Rated)

(% (psid) (psid) (psid) 16.00 42.51 1.39 1.81 1.67 20.01 42.84 1.42 1.83 1.67 23.93 44.00 1.47 1.96 1.79 35.85 45.49 1.79 2.30 2.12 41.68 60.53 4.14 4.73 4.59 45.43 65.94 5.22 5.79 5.65 47.78 55.02 3.44 3.95 3.82 50.41 65.82 5.36 5.89 5.72 59.63 64.99 5.52 6.02 5.83 66.32 64.27 5.60 6.09 5.86 66.79 63.95 5.54 6.04 5.79 67.26 80.38 9.00 9.41 9.14 82.03 83.00 10.40 10.88 10.48 89.54 86.08 11.58 12.02 11.52 90.07 88.44 12.25 12.67 12.24 99.97 97.31 15.02 15.51 15.02 100.02 99.38 I1559 16.07 15.59 Page 7 of 11

Figure 1. Calculated versus Measured Core Support Plate Pressure Drop 20

_ 15 W ID U

5 0

0 5 10 15 20 Plant Measured (psid) 4.3.2 Cycle 13 and 12 Measured Core Support Plate Pressure Drop Comparison The measured core support plate pressure drops, obtained during the Cycle 13 and 12 startups, are shown in Figure 2. The measured data provide further evidence that the thermal-hydraulic performance of the Hope Creek core has not been affected by the introduction of the GE14 fuel design. The excellent comparison is indicative of the hydraulic compatibility of the two fuel designs, GE14 and SVEA-96+.

Figure 2. Cycle 13 and Cycle 12 Measured Core Support Plate Pressure Drop

=. 16 CW.

r. 14 * -

0 1 12 L.

v) l

. 10

6) A'.4 A

4, 8 A

C 6 t L4A I Icck1

• AA I A Cycle 13 cr 4 *A £ Acc1 C. 2 U;)

0 40 50 60 70 80 90 100 110 Total Core Flowv (Allbm/l1r)

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4.3.3 Cycle 13 and 12 Recircuilation Pump Data and Core Flow Comparison During reactor startup, data is recorded at various pump speeds in accordance with procedure HC.OP-FT.BB-0001 (Q), "Jet Pump Data Collection". The Cycle 13 and 12 data are provided in Figures 3 through 6, and shows that no anomalous behavior of the recirculation pumps. The introduction of GE14 fuel assembly has not affected the recirculation pump performance. This indicates that the overall hydraulic resistance of the core has not changed, which is the result of having hydraulically compatible fuel loaded in the core.

Figure 3. Cycle 13 and 12 Recirculation Pump A Flow versus Pump Speed 50 I- 45 40 0,,

.2 35 30-2 25 400 600 S00 Joao 1200 1400 1600 Recirculation Pump A Speed (rpm)

Figure 4. Cycle 13 and 12 Recirculation Pump B Flow versus Pump Speed So 45 0

40 35 C'.

30 E 25 0.

20

'.1 15 -

I- 10 5 -__

400 600 800 1000 1200 1400 1600 Recirculation Pump B Speed (rpm)

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Figure 5. Cycle 13 and 12 Recirculation Pump A Head versus Pump Flow 200 .--

G, 180

- 160-v140-

< I4O~

-< 120 -

E 1o0-80-

-- Cycle 12

° 60-e 40- - Cycle 13 4 20-0 0 5 10 15 20 25 30 35 40 45 50 Recirculation Pump A Flow (KGPAI)

Figure 6. Cycle 13 and 12 Recirculation Pump B Head versus Pump Flow 200 -

180 l

-ac 160 -

V 140 -

120 -

E 100 -

so 0 l--Cycle 12 60 -

S e 40 - -'Cycle 13

4. 20 -

0-0 5 10 15 20 25 30 35 40 45 50 Recirculation Pump B Flow (KGPM)

Figure 7. Cycle 13 and 12 Core Flow versus Recirculation Pump Flow 110 I 100 i: 90 so

-_ 70 C 60

'S 50 e 40 - Cycle 12 6 30

- Cycle 13 20 10 0

0 102030 40 50 60 70 s0 90 100 Total Recirculation Pump Flow (KGPiN1)

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5.0 NSSS Process Computer The description of the initial startup testing to evaluate the performance of the process computer tinder plant operating conditions is provided in the Hope Creek FSAR section 14.2.12.3.11. The CMS thermal limit results (thermal limit performance discussed in Section 4.2 of this report) were tested during the Cycle 13 BOC power ascension.

6.0 References 6.1 NFS-0245, HCGS Cycle 13 Evaluation of the UFSAR Chapter 14 Initial Cycle Startup Test.

6.2 NFS-0242, Cycle Management Report For Hope Creek Generating Station Cycle 13.

6.3 HCG.6-0004, Cycle 13 Startup Thermal Hydraulic Analysis.

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