ML20215C641

Cycle 2 Startup Test Rept

ML20215C641
Person / Time
Site:	LaSalle
Issue date:	12/02/1986
From:	COMMONWEALTH EDISON CO.
To:
Shared Package
ML20215C632	List: ML20215C628 ML20215C641
References
NUDOCS 8612150298
Download: ML20215C641 (13)

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e LASALLE COUNTY NUCLEAR POWER STATION UNIT 1 CYCLE 2 STARTUP TEST REPORT 96121bO296 0612023 ADOCK 0500 PDR P

TABLE OF CONTENTS Station Test No. Procedure No. Title Pt et

1. LTP-1700-1 Core Verification 1

2. LTS-1100-14 Shutdown Margin Subcritical Demonstration 2

3. LTS-1100-1 Shutdown Margin Test 3-4

4. LTS-1100-2 Checking for Reactivity Anomalies 5-6

5. LTS-1100-4 Scram Insertion Times 7-8

6. LTP-1600-17 Core Power Distribution Symmetry Analysis 9-11

I LTP-1700-1, CORE VERIFICATION PURPOSE The purpose of this test is to visually verify that the core is loaded as intended for Cycle 2 operation.

CRITERIA The as-loaded core must conform to the cycle core design used in the reload licensing analysis by the Core Management Organization (General Electric). At least one member of the Commonwealth Edison Company audit staff must be present for the core verification, and an observer from an independent audit company must have been notified (but need not be present). Any discrepancies discovered in the loading will be promptly corrected and the affected areas reverified to ensure proper core loading prior to unit startup.

Conformance to the cycle core design will be documented by a permanent core serial number map signed by the audit participants.

RESULTS AND DISCUSSION The Cycle 2 core verification consisted of a core height check performed by the fuel handlers and two videotaped passes of the core by the nuclear group. The height check verifies the proper seating of the assembly in the fuel eupport piece while the videotaped scans verify proper assembly orientation, location, and seating. On May 6, 1986, the core was verified as being properly loaded and consistent with the General Electric Cycle 2 core design used in the reload licensing analysis.

A serial number inventory was also performed on the Unit i fuel pool on May 6, 1986, to verify that the fuel pool contained the proper assemblies. The fuel pool contained no assemblies which should have been loaded in the Unit i vessel.

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. 1 LTS-1100-14, SHUTDOWN MARGIN SUBCRITICAL DEMONSTRATION PURPOSE The purpose of this test is to demonstrate, using the adjacent rod subcritical method, that the core loading has been limited such that the reactor will be subcritical throughout the operating cycle with the strongest control rod in the full-out position (position

48) and all other rods fully inserted.

CRITERIA If a shutdown margin (SDM) of 0.741% AK/K (0.38% AK/K + R) cannot be demonstrated with the strongest control rod fully withdrawn, the core loading must be altered to meet this margin. R is the difference between the core's beginning-of-cycle reactivity and the peak reactivity for the cycle. The R value for Cycle 2 is 0.361% ,

AK/K, with peak reactivity occurring at 6,000 MWD /ST into the cycle. (

RESULTS AND DISCUSSION On September 17, 1986, the local SDM demonstration was successfully performed using diagonally adjacent control rods 50-27 (strongest control rod at beginning-of-cycle) and 54-23. General Electric (GE) provided, in the Cycle Startup Package, rod worth information (for control rod 50-27 and diagonally adjacent rods 54-23 and 46-

23) and moderator temperature reactivity corrections to support

, this test. Using the GE supplied information, it was cetermined i that with control rod 50-27 at position 48, rod 54-23 at positon 24, a moderator temperature of 150*F, and the reactor subcritical, a SDM of 1.259% AK/K was demonstrated. The SDM demonstrated l

exceeded the 0.741% AK/K required to satisfy Technical I

Specification 3.1.1, and maintained sufficient margin to the GE calculated SDM for the core at beginning-of-cycle (2.07% AK/K).

[

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LTS-1100-1, SHUTDOWN MARGIN TEST PURPOSE The purpose of this test is to demonstrate, from a normal in-sequence critical, that the core loading has been limited such that the reactor will be subcritical throughout the operating cycle with the strongest control rod in the full-out position (position 48) and all other rods fully inserted.

CRITERIA If a shutdown margin (SDM) of 0.741% AK/K (0.38% AK/K + R) cannot be demonstrated with the strongest control rod fully withdrawn, the core loading must be altered to meet this margin. R is the difference between the core's beginning-of-cycle reactivity and the peak reactivity for the cycle. The R value for Cycle 2 is 0.361%

AK/K, with peak reactivity occurring at 6,000 MWD /ST into the cycle.

RESULTS AND DISCUSSION The beginning-of-cycle SDN was successfully determined from the initial critical data. The initial Cycle 2 critical occurred on September 17, 1986, on control rod 34-23 at position 16, using an A-2 sequence. The moderator temperature was 150*F and the reactor period was 328 seconds. Using rod worth information, moderator temperature reactivity corrections, and period reactivity corrections supplied by General Electric (in the Cycle Startup Package), the beginning-of-cycle SDM was determined to be 2.78%

AK/K (see Table 1). The SDM demonstrated exceeded the 0.741% AK/K required to satisfy Technical Specification 3.1.1.

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TABLE 1 SHUTDOWN MARGIN CALCULATION ITEM AK/K Critical Rod = 34-23 at position 16 Worth of Strongest Rod = 0.03088 * (1)

Worth of Control Rods Withdrawn to Obtain Criticality:

24 Group 1 rods at position 48 = 0.04236 * (2) 19 Group 2 rods at position 48 = 0.01775 * (3) 1 Group 2 rod at position 16 = 0.00027 * (4)

Temperature Correction for Tm = 150*F =-0.00150 * (5)

Period Correction for Period = 328 seconds = 0.00020 * (6)

Shutdown Margin Keff:

SDM Keff = 1.0000 + (1) - (2) - (3) - (4) - (5) + (6)

= 0.9722 SDM = (1.0000 - (SDM Keff))

• 100 = 2.78% AK/K "LaSalle Unit 1 Cycle 2 Startup Package", supplied by General Electric Company.

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a LTS-1100-2, CHECKING FOR REACTIVITY ANOMALIES PURPOSE The purpose of this test is to compare the actual and predicted critical rod configurations to detect any unexpected reactivity trends.

CRITERIA In accordance with Technical Specification 3.1.2, the reactivity equivalence of the difference between the actual control rod density and the predicted control rod density shall not exceed 1%

&K/K. If the difference does exceed 1% AK/K, the Core Management Engineers (General Electric Company and Commonwealth Edison Company) will be promptly notified to investigate the anomaly. The cause of the anomaly must be determined, explained, and corrected for continued operation of the unit.

RESULTS AND DISCUSSION Two reactivity anomaly calculations were successfully performed during the Unit 1 Cycle 2 Startup Test Program, one from the initial critical and the second from steady-state, equilibrium conditions at approximately full power.

The initial critical occurred on September 17, 1986, on control rod 34-23 at position 16, using an A-2 sequence. The moderator temperature was 150*F and the reactor period was 328 seconds.

Using rod worth information, moderator temperature reactivity corrections, and period reactivity corrections supplied by General Electric (in the Cycle Startup Package), the actual critical was determined to be within 0.708% AK/K of the predicted critical (see Table 1). The anomaly determined is within the 1% AK/K allowed by l Technical Specification 3.1.2.

The anomaly calculation at power was successfully performed on October 9, 1986, with the unit at 89.7% power and a cycle exposure of 185 MWD /ST. Using the Cycle 2 reactivity anomaly curve, the control rod notch correction expression, and the reactivity normalization coefficients supplied by General Electric (in the Cycle Startup Package), the reactivity anomaly was determined to be 0.116% AK/K. The anomaly determined is within the 1% AK/K allowed by Technical Specification 3.1.2.

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l TABLE 1 INITIAL CRITICALITY COMPARISON CALCULATIONS ITEM AN/K Keff with all rods in at 68'F = 0.94840

• Reactivity inserted by 24 group 1 rods at position 48 = 0.04236

• Reactivity inserted by 19 group 2 rods at position 48 = 0.01775

• Reactivity inserted by 1 group 2 rod at position 16 = 0.00027

• Predicted Keff at actual critical rod pattern (68'F) = 1.00878 Reactivity associated with the measured reactor period (period correction for 328 second period) = 0.00020

• Reactivity associated with moderator temperature (150*F actual, 68'F predicted) = 0.00150

• Reactivity Anomaly = [(predicted Keff - 1) - (period correction) - (temperature correction)]

• 100% = 0.708% AK/K "LaSalle Unit 1 Cycle 2 Startup Package", supplied by General Electric Company.

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1 LTS-1100-4, SCRAM INSERTION TIMES PURPOSE The purpose of this test is to demonstrate that the control rod scram insertion times are within the operating limits set forth by the Technical Specifications (3.1.3.2, 3.1.3.3, 3.1.3.4).

CRITERIA The maximum scram insertion time of each control rod from the fully withdrawn position (48) to notch position 05, based on de-energization of the scram pilot valve solenoids as time zero, shall not exceed 7.0 seconds.

The average scram insertion time of all operable control rods from the fully withdrawn position (48), based on de-energization of the scram pilot valve solenoids as time zero, shall not exceed any of the following:

Position Inserted From Average Scram Insertion Fully Withdrawn Time (Seconds) 45 0.43 39 0.86 25 1.93 05 3.49 The average scram insertion time, from the fully withdrawn position (48), for the three fastest control rods in each group of four control rods arranged in a two-by-two array, based on de-energization of the scram pilot valve solenoids as time zero, shall not exceed any of the following:

Position Inserted From Average Scram Insertion Fully Withdrawn Time (Seconds) 45 0.45 39 0.92 25 2.05 05 3.70 RESULTS AND DISCUSSION Scram testing was successfully performed between September 19-23, 1986. All control rod scram timing acceptance criteria were met during this test (see Table 1).

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TABLE 1 CONTROL ROD DRIVE (CRD) SCRAM RESULTS Maximum Average Average Scram Times Scram Times in a Position of all CRDs (secs.) Two-by-Two Array (secs.)

45 0.315 0.336 39 0.615 0.637 25 1.338 1.381 05 2.449 2.523 Maximum 90% scram time (position 05): CRD 18-47, 2.936 secs.

' Cave (position 39) for Minimum Critical Power Ratio determination: 0.615 seconds.

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LTP-1600-17, CORE POWER DI.9TRIBUTION SYMMETRY ANALYSIS l

PURPOSE  :

The purpose of this test is to verify the core power symmetry and the reproducibility of the TIP readings.

CRITERIA The total TIP uncertainty (including random noise and geometric uncertainties) obtained by averaging the uncertainties for all data sets must be less than 8.7%

The gross check of the TIP signal symmetry should yield a maximum deviation between symmetrically located pairs of less than 25%.

RESULTS AND DISCUSSION Core power symmetry calculations were performed based upon data obtained from two full core TIP nets (DD-1) and individual TIP traverses (DD-2) of the reference channel by each TIP machine. The

, TIP sets (OD-1 and OD-2) were initially performed on October.8, i

1986, at 90% power, and then repeated on October 9, 1986, at 88%

power. The average total TIP uncertainty from the two data sets l was 3.623%, satisfying the criteria of the test (less than 8.7%).

! The random noise uncertainty and geometrical uncertainty were determined to be 0.984% and 3.487%, respectively.

l Table 1 lists the symmetrical TIP pairs, their core locations, and their respective average deviations. The maximum deviation between symmetrical TIP pairs was 9.91% for TIP pair 26-42, satisfying the criteria of the test (less than 25%).

The method used to obtain the uncertainties consisted of l

calculating the average of the nodal BASE ratio of TIP pairs by:

i

~12 n -

R= ,[n j=r pEg Ry ,

where Ri] = the BASE ratio for the ith node of TIP pair ],

I n = number of TIP pairs = 19.

Next, the standard deviation (expressed as a percentage) of these ratios is calculated by the following equation:

- 22 n -

YZ Z (Ris - R)2 -

! g)= 0 2=t 4 100 l _ (18n- 1) _

The total TIP uncertainty ((Itotal %) iscalculatedbydividingq(%)

by 8 because the uncertainty in one TIP reading is the desired parameter, but the measured uncertainty is the ratio of two TIP readings.

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.n _ _ . _ - _ _ _ . _ _ - - . -_--_ _ _ - . - _ ~ . _ . , . - .__

To calculate random noise ttncertainty, the average BASE reading at each node for nodes 5 through 22 is calculated by:

Mr T -

3 BASE (K) = [ BASE (N,M K)

"T *

• n , a=,

~ "

where NT = number of runs per machine.

MT = number of machines.

BASE (K) = average reading at nodal level K.

K = 5 through 22.

The random noise component of the total TIP uncertainty is cerived from the average of the nodal variances:

• 8/2

~8ASE (N,M,K)- BASE (KI (K) g (4.): M:5 Msl N:1 - - g 100 13 (MT*MT-1)

Finally, the TIP geometric uncertainty can be calculated by:

'h 9 .= @arna. - %*

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l TABLE 1 TIP SIGNAL SYMMETRY RESULTS All numbers shown are averages from two OD-1 data sets (from 10-8-86 at 90% power and 10-9-86 at 88% power).

Symmetrical TIP Pair Absolute Percen't Numbers (Core Location) Difference TIP Pair a b of BASE # Deviation

• 1 (16-09) 6 (08-17) 7.31 8.02 2 (24-09) 13 (08-25) 3.27 3.29 3 (32-09) 20 (08-33) 7.10 6.72 4 (40-09) 27 (08-41) 1.62 1.64 5 (48-09) 34 (08-49) 1.92 2.98 8 (24-17) 14 (16-25) 1.49 1.41 9 (32-17) 21 (16-33) 2.36 2.06 10 (40-17) 28 (16-41) 7.60 6.95 11 (48-17) 35 (16-49) 1.97 1.92 12 (56-17) 40 (16-57) 1.22 2.04 16 (32-25) 22 (24-33) 5.68 5.42 17 (40-25) 29 (24-41) 6.45 5.58 18 (48-25) 36 (24-49) 0.72 1.36 19 (56-25) 41 (24-57) 6.19 7.75 24 (40-33) 30 (32-41) 4.84 4.62 25 (48-33) 37 (32-49) 0.72 1.19 26 (56-33) 42 (32-57) 7.39 9.70 32 (48-41) 38 (40-49) O.60 0.78 33 (56-41) 43 (40-57) 1.12 1.50 Average Deviation = 3.94%

1. - where : Absolute Difference of BASE = BASEa - BASEb and BASEj =

BASEi(K)

K:5

• - where : X Deviation = BASEa - BASE 6

• 100 0.5(BASE. + BASEg)

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