ML20134N787

Quad-Cities Nuclear Power Station Unit 2,Cycle 8,Startup Test Results

ML20134N787
Person / Time
Site:	Quad Cities
Issue date:	08/28/1985
From:	Wojnarowski J COMMONWEALTH EDISON CO.
To:	Harold Denton Office of Nuclear Reactor Regulation
References
0334K, 334K, NUDOCS 8509050356
Download: ML20134N787 (11)

Text

. 7,r QUAD-CITIES NUCLEAR POWER STATION UNIT 2 CYCLE 8 STARTUP TEST RESULTS-y-

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TABLE OF CONTENTS l

Test No.

Title Page l

l 1

Scram Timing i

2 Shutdown Margin 2

l 3

Initial Critical 2

Il' 4

TIP Reproducibility and Core Power Symmetry 3

1.

' Control Rod Scram Timing Purpose.

The purpose of this test is to demonstrate the scram capability of all of the operable control rods in compliance with Technical Specifications 4.3.C.1 and 4.3.C.2.

Criteria A.

' The average scram insertion time, based on the de-energization of the scram pilot valve solenolds as time zero, of all operable control rods during reactor power operation shall be no greater than:

1 INSERTED FROM AVG. SCRAM INSERTION FULLY WITHORAWN TIMES (sec) 5 0.375 l

20 0.900 50 2.000 l

90 3.500 l

The average of the scram insertion times for the-three fastest control rods of all groups of four rods in a two by two array shall be no greater than:

% INSERTED FROM AVG. SCRAM INSERTION FULLY WITHDRAWN TIMES (sec) 5 0.398 l

20 0.954 l

50 2.120 l

90 3.800 i

If these times cannot be met, the reactor shall not be made supercritical; if operating, the reactor shall be shutdown immediately upon determination that average scram time is deficient.

B.

The maximum insertion time for 90% insertion of any operable control i

l rod shall not exceed 7.00 seconds.

If this requirement cannot be met, the deficient control rods shall be considered inoperable, fully inserted into the core, and electrically disarmed.

l Results and Discussion All 177 control rods were scram tested at cold reactor conditions and 176 l

at hot reactor conditions.

The results are presented in Table'l.

The maximum 90% insertion time.was 3.17 seconds for~ control rod F-7 (22-27).

Control Rod H-8 (30-31) was satisfactorily tested under cold conditions but l

was;lnserted and taken out-of-service while awaiting replacement during the l-next convenient outage. This rod did not have an acceptable withdrawal-rate for the full distance of travel.

l Since Control Rod H-8 will be inoperable until replacment both criterla A and B were met.

l 0034H

u Table-1.

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Control Rod Scram Results t-i~

NUM8ER REACTOR AVERAGE TIMES FOR % INSERTED, SEC

- 0F RODS-- CONDITIONS

-5%

20%

50%

90%

177~

. Cold 0.27 0.49 0.94 1.62 176 Hot' O.31 0.68 1.45 2.54 i

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2.

Shutdown Margi Demonstration and Control Rod Functional Checks Purpose The purpose of this test is to demonstrate for this core loading in the most reactive condition during the operating cycle, that the reactor is subcritical with the strongest control rod full out and all other ods fully inserted.

Criteria If a shutdown margin of 0.773% AK (-0.25% + R + 8 C settling penalty) 4 cannot be demonstrated with the strongest control rod fully withdrawn, the core loading must be altered to achieve this margin. The core reactivity has been calculated to be at a maximum 5000 mwd /t into the cycle and R is The control rod 8 C settling penalty for Unit Two given as 0.4737. OK.

4 is 0.05% A.

K Results and Discussion On May 1, 1985, control rod J-2 (the rod which was calculated by General Electric to t;e of the highest worth) was fully withdrawn to demonstrate that the reactor would remain subcritical with the strongest rod full out.

This maneuver was performed to allow cold control rod testing prior to the shutdown margin demonstration.

Control Rod functional subcritical checks were performed as part of the l

cold scram timing and control rod friction testing. No unexpected reactivity insertions were observed when any of the 177 control rods were withdrawn.

General Electric provided rod worth information for the two strongest diagonally adjacent rods F-13 and H-13 with rod G-14 full out.

This method provided an adequate reactivity insertion to demonstrate the desired shutdown margin. On June 5, 1984, a diagonally adjacent shutdown margin demonstration was successfully performed. Using the G.E. supplied rod worth for G-14 (the strongest rod) and diagonally adjacent rods F-13 and H-13, it was determined that with G-14 at position 48, and F-13 at position 36, a moderator temperature of 160*F, and the reactor subtritical, a shutdown margin of 1.26% AK was demonstrated.

The G.E. calculated shutdown margin with G-14 withdrawn and 68'F reactor water temperature was 2.666% AK at the beginning of cycle 8.

At approximately 5000 MHd/t into cycle 8 a minimum calculated shutdown margin of 2.193% AK will occur with C-5 fully withdrawn. Note that the minimum shutdown margin shifts from rod G-14 at beginning of cycle to rod C-5 at 5000 mwd /t. 0034H

G.E.'s ability to determine rod worth was demonstrated by the accuracy _

of their in-sequence criticality prediction.

The oK difference between the expected critical rod pattern and the actual critical rod pattern was determintd to be 0.367% oK.

This initial critical demonstrated that the actual shutdown margin at the beginning of cycle 8 was 2.299% oK and that 1.826% AK at 5000 Mild /t into cycle 8.

3.

Initial Critical Prediction Purpose The purpose of this test is to demonstrate General Electric's ability to calculate control rod worths and shatdown margin by predicting the insequence critical.

Criteria General Electric's prediction for the critical rod pattern must agree within 1% AK to actual rod pattern. A discrepancy greater than 1% oK In the non-conservative direction will be cause for an On-Site Review and l

investigation by Nuclear Fuel Services.

Results and Discussion i

l On June 5,1985, at 2307 hours0.0267 days <br />0.641 hours <br />0.00381 weeks <br />8.778135e-4 months <br /> the reactor was brought critical with a l

reactor water temperature at the time of criticality of 173*F.

The oK difference between the expected critical rod pattern at 68'F and the actual l

critical rod pattern at 68'F was -0.00153 from rod worth tables supplied l

by General Clectric.

The temperature effect was -0.0018 oK from General Electric-supplied corrections.

The excess reactivity yleiding the 196 second positive period was 0.00034 oK.

These reactivities sum to give

-0.00367 AK difference (-0.37% AK) between the expected critical rod pattern and the actual rod pattern.

This is within the 1% oK required in the criteria of this test, and General Electric's ability to predict control rod worths is, therefore, successfully demonstrated.

4.

Core Power Distribution Symmetry Analysis Purpose The purpose of this test was to determine the magnitude of indicated core power distribution asymmetries using data (TIP traces and 00-1) collected in conjunction with the P-1 update.

Criteria A.

The total TIP uncertainty (including random noise and geometric uncertainties obtained by averaging the uncertainties for all data sets) must be less than 91.

B.

The gross check of TIP signal symmetry should yleid a maximum deviation between symmetrically located pairs of less than 25%. 0034H L

V Results and Discussion Core power symmetry calculations were performed based upon computer program 00-1 data runs on July 17, 1985, at 99.7% power, and June 20, 1985,'at 99.0% power.

The average total TIP uncertainty from the two TIP sets was 3.922%. The random noise uncertainty was 1.085%. This yleids a geometrical uncertainty of 2.837%.

The total TIP uncertainty was well within the 9% limit.

Table 2 lists the symmetrical TIP pairs and their respective average deviations.

Figure I snows the core location of the TIP pairs and the average TIP readings.. The maximum deviation between symmetrical TIP pairs was 13.7% for pair 11-34. Thus, the second criterion, mentloried above, was also met.

The method used to obtain the uncertaintles consisted of calculating the average of the nodal ratio of TIP pairs by:

"n 22

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Rlj R - 18n j=1 1-5 where Rij is the ratto for the Ith node of TIP pair j, there being n such pairs, where n=18.

Next the standard deviation of the ratios is calculated by:

n 22

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(Rij - R)2

~1/2 o.

j.i 15 R

(18n - 1)

'R is multiplied by 100 to express R as a percentage of the ideal value of OR of 1.0.

% oR = 'R x 100

,The total TIP uncertainty is calculated by dividing % og by T in order ta account for data being taken at 3 inch intervals and analyzed.on a 6 inch nodal basis.

In order to calculate random noise uncertainty the average reading at each node for nodes 5 through 22 15 calculated by~:

'MT NT 1

E E

BASE (N, M. K)

BASE (K) - NT MT M-1 N-1 where NT = number of runs per machine - 4 MT - number of machines. 5 BASE (K) - average reading at nodal level K, K = 5 through 22 U t 0034H L

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The random nolse is derived'from the average'of the nodal variances by:

22 MT'.

.NT

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BASE (N. M. K) - BASE (K) 10 noise -

K-5 M-1 N : BASE -(K) x 100 18 (NT x MT -1)

Finally the TIP, geometric uncertainty _can be calculated by:

% 0 geometric'- (% 0 total 2-%o not'se2>1/2 R t, b

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CORE-SYMMETRYL Based on 00-l's'Trom 06-20-85li99.0% power), and 07-20-85.(99.7% power) j.

All Numbers Shown Are Averages Foq The Two Data Sets

i y-i SYMMETRICAL TIP T=

Ta - T',

% - 100 x T/ Ta + Tb h

i PAIR NUMBERS ABSOLUTE DIFFERENCE <

% DEVIATION 2

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12.136 2

12 7.165 7.148 3

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-1.318 1.255 4 -26 7.023 7.096 5

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O.862 1.809 4.164 4.266 1.,

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6.383 6.333 11 34 10.978 11.474 1.161 1.011 15 21 1

16 28 2.568 2.568 17 35 4.286 4.094 18 39 1.292 1.727 23 29 2.074 1.955 2.025 2.119 24 36 25 40 0.694 0.785 31 37 4.559 4.733 l

32 41 0.650 1.241 22 Average Deviation.

T E T (K)

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Based on OD-l's From 06-20-85 (99.0% power) 07-20_85 (99.7% pover)

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) One First National Plata. Chicago, Ilhnois

( j'~7 Addr:ss Reply to: Post 6ttica Box 767 x

Chicago, litinois 606%

August 28, 1985 i

Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission

Washington, DC 20555

Subject:

Quad Cities Station Unit 2 Summary Startup Test Report - Cycle 8 NRC Docket =No. 50-265 -

s

Dear Mr. Denton:

Enclosed for-your information and use is the Quad Cities Station Unit 2, Cycle 8 Startup Test Report Summary. This report is submitted in accordance with previous requests from the NRC Staff and our Technical Specifications.

Please address any questions concerning this matter-to this office.

One signed original and forty (40) copies'of this letter and the enclosure are provided for your use.

Very truly yours, s

di

.. R. Wojnarowski Nuclear Licensing Administrator 1m Enclosure cc:. R. Bevan - NRR NRC Resident Inspector - Quad Cities-4 s

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0334K L

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