ML20046A661

Cycle 6 Startup Rept.

ML20046A661
Person / Time
Site:	Diablo Canyon
Issue date:	07/21/1993
From:	PACIFIC GAS & ELECTRIC CO.
To:
Shared Package
ML20046A660	List: ML20046A659 ML20046A661
References
NUDOCS 9307290181
Download: ML20046A661 (7)

Text

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PG&E Letter No. DCL-93-181 4

ENCL.05URE STARTUP REPORT DIABLO CANYON POWER PLANT UNIT 2 CYCLE 6

SUMMARY

The Diablo Canyon Power Plant (DCPP) Unit 2 Cycle 6 (2C6) startup program extended from initial cycle criticality through the completion of power ascension testing at 100 percent rated thermal power (RTP). Significant milestones included:

e Sixth Cycle Criticality: April 29, 1993

• Parallel to Grid: May 1, 1993

• Full Power Flux Map: May 13, 1993

• NIS Calibration Completed: May 20, 1993 ,

The latest reload, during the Unit 2 fifth refueling outage (2RS), completed the transition to a Vantage 5 core. This reload consisted of 88 fresh and 105 partially burned Vantage 5 fuel assemblies.

Startup testing was divised into three phases: pre-critical testing (performed '

during Mode 3); low-pov:er physics testing; and power ascension testing. The test results obtained during startup testing agreed well with predicted results. The completion of the transition to Vantage 5 fuel did not have any noticeable effects on the nuclear, thermal, or hydraulic performance of the pl ant.

Unit 2 Cycle 6 Core The DCPP 2C6 core consists of 193 Westinghouse Vantage 5 fuel assemblies.

Of these 193 asnmblies,105 had beer, previously irradiated in Cycles 4 and 5.

The remaining 88 fuel assemblies were split into two enrichment regions:

4.0 weight percent (assemblies U01 to U32), and 4.4 weight percent (assemblies U33H to V88H). The new core included a total of 9,056 fuel rods with integral fuel burnable absorber (IFBA) coated pellets.

The fuel assemblies are arranged in an octant-symmetric pattern. The core was designed with low-power assemblies in the corners ta reduce the fast neutron flux at the reactor vessel wall.

Physical loading of each assembly into its proper location was visually verified by underwater camera inspection.

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RESUITS OF IESTING A. Suberitical Testina Rod drop time measurements showed no discernible increases in rod drop times. All measured drop times were well within the Technical Specification limit of 2.7 seconds.

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B. Low Power Phvsics Testin_q

1. Objective The objective of the low power physics testing program was to verify that the nuclear design characteristics of the core were consistent with the Westinghouse design calculations.

2. Methodoloav Testing was performed in the following general sequence:

a. Approach to and achievement of initial cycle criticality.

The reactor coolant system (RCS) boron concentration was reduced to 200 ppm boron above the estimated critical boron concentration. Then, control rods were withdrawn until Control Bank D was at 170 steps. The final approach to criticality was achieved through dilution.

b. Determination of the zero-power physics test range,

c. Reactivity computer check.

d. Measurement of all-rods-out (AR0) boron concentration.

e. Measurement of AR0 isothermal temperature coefficient (ITC).

ITC is determined based on the reactivity change corresponding to a change in RCS temperature. 'The moderator temperature coefficient (MTC) is then inferred from the measured value of ITC.

f. Measurement of rod worth utilizing the " Pod Swap Method."

3. Results low power physics test results are summarized in Table 1. All acceptance criteria and review criteria were met-.

C. Power Ascension Testina

1. Objective The main objectives of power ascension testing were to monitor core power distribution at various power levels and to recalibrate instrumentation for~ power operation. This ensured that normal, full 6151S/85K

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, i power operation could be reached and sustained while meeting all-design requirements. ,

2. Methodoloav j Testing was performed at specific power plater.us. For 2C6, the ,

primary test plateaus were at 30, 50, 90, and 100 percent RTP. Power.

level changes were governed by operating procedures and fuel ,

preconditioning guidelines specified by Westinghouse.

To determine steady-state core power distribution, flux-maps were-performed using moveable incore detectors. Measured peaking factors obtained from the flux maps were then compared to Technical Specification limits to determine limitations on power ,

ascension. ,

Thermal-hydraulic parameters, nuclear parameters, and associated '

instrumentation were monitored tr.roughout power ascension. The major ,

areas addressed were: ,

a. Steam and feedwater flows - These flow rates were measured and correlated to verify the calibration of the associated  ;

transmitters. {

b. RCS temperature - RCS temperature channels were monitored  :

to verify alignment of the RCS temperature j instrumentation. j

c. RCS flow - RCS flow was calculated by performing a primary  ;

side heat balance in conjunction with a secondary side  !

heat balance. l 1

d. Neutron flux and axial flux distribution - The power range l nuclear instruments were monitored and evaluated for l calibration.

All instrumentation requiring recalibration was identified and actions were taken to recalibrate them.

3. Results During the 2C6 power ascension, core performance was acceptable.

Flux maps were performed at 30, 49, 92, and 100 percent RTP. The results of these maps are summarized in Talile 2. j The initial flux map at 30 percent power indicated that all core parameters met the associated acceptance and review criteria.

The incore/excore calibration was performed at 49 percent RTP based on the results of the second flux map. Once the. ,

incore/excore calibration was performed, all of the core '

performance indicators were within the review criteria.

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Power ascension. continued to 75 percent RTP. 'RCS temperature, RCS flow, steam generator pressure, and nuclear instrumentation were evaluated and found to be satisfactory.

Power ascension continued to the 90 percent plateau. . Power was stabilized at 93 percent RTP, where testing was performed. . A full >

core flux map was taken at 92 percent, and RCS. flow and temperature data were taken at 93 percent RTP. .The axial flux difference (AFD) calculated by the Diablo Canyon INCORE Sequence was compared to the AFD indicated by the power range detectors.

The AFD comparison did not meet the review criteria. As a result,.

performance of an incore/excore calibration was- recommended.

Since a multi-point incore/excore calibration was required at approximately 100 percent RTP, power ascension continued to the-100 percent plateau. Initial indications were that flow was down slightly in RCS loops 1 and 3, but that total flow met Technical Specification requirements.

Power ascensica continued to full power. RCS flow, RCS temperature, feedwater flow, steam flow, and NIS calibration data were taken to support sustained full power operation. Results of the power ascension testing are summarized in Table 2 and Table 3. '

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l-I TABLE 1 1

RESULTS OF PHYSICS TESTING - UNIT 2 CYCLE 6 l

VALUES DIFFERENCE CRITERIA STP ITEM UNITS j TEST PRED ABS  % REVIEW . ACCEPT R-30 RCS Boron at criticality ppm 1581 1566 - 120 N/A 1000 pcm (CB-D at 170 steps) pcm 1

( R-6 ARO Boron Endpoint ppm 1603 1588 + 15 N/A .150 R-6 CB-D in Boron Endpoint ppm 1458 1446 + 12 150 N/A l R-7A ITC (ARO) pcm/'F -0.24 -0.28 +0.04 2 N/A MTC (ARO) pcm/*F +1.38 +1.35 +0.03 N/A +5 ,

t R-6 Boron Worth pcm/ ppm -7.97 -B.06 +1.13 N/A 115%

R-31 Centrol Bank D Worth pcm 1156 1148 +0.7 110% 115%

R-31 " Swap Worth" SD-A pcm 258 274 -5.8 1 15% 30%

SD-B pcm 990 1005 -1.5 15% f30%  :

SD-C pcm 438 474 -7.6 1 15% 30%

SB-D pcm 434 474 -8.4 1 15% 230%

CB-A pcm 234 251 -6.8 115% 130%

CB-B pcm 712 759 -6.2 15% 130%

CB-C pcm 739 769 -3.9 15% 130% ,

Total pcm 4961 5154 -193 -3.7 +10% -10%

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. TABLE 2 POWER ASCENSION FLUX MAP RESULTS - UNIT 2 CYCLE 6 Date of Map 5/4/93 5/6/93 5/7/93 5/13/93 Power Level (%) 30 49 92 100 Bank D Position (Steps) 205 209 212 215 RCS Boron (ppm B) 1344 1270 1146 1126 INCORE Tilt 1.016 1.016 1.013 1.010 F",, 1.523 1.508 1.453 1.437 F" 1.908 1.828 1.816 1.812 F 1.230 1.204 1.190 1.199 Z

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. TABLE 3 FULL POWER THERMAL-HYDRAULIC DATA: UNIT 2 CYCLE 6 i i

Reactor Coolant System Average Temperature Values:

Loop 1: 570.9 F Loop'2: 574.2 F Loop 3: 570.9 F i Loop 4: 572.2 F Average: 572.0 F i

Reactor Coolant System Core AT Values:

Loop 1: 61.5 F  !

Loop 2: 66.0 F l Loop 3: 61.6 F Loop 4: 64.0 F Average: 63.3 F Reactor Coolant System Flowrates:

1 Loop 1: 90,816 gpm  !

Loop 2: 90,252 gpm i Loop 3: 92,433 gpm '

Loop 4: 92,275 gpm Total: 365,776 gpm 6151S/85K 4 .