ML20238E007

Forwards Unit 2 Startup Test Rept Summary,Per Tech Spec Section 6.6.A.1

ML20238E007
Person / Time
Site:	LaSalle
Issue date:	09/08/1987
From:	Allen C COMMONWEALTH EDISON CO.
To:	NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM)
References
NUDOCS 8709140050
Download: ML20238E007 (13)

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CN. Commonwealth' Edison -

- 1 / Ons First Nrtional Plaza Chicago, Illinois j

,V Address Reply to: Post Office box 767 Chicago, Ulinois 60690 0767 September 8, 1987-U.S. Nuclear Regulatory. Commission Attn:. Document Control Desk Washington, DC 20555 1

Subject:

LaSalle County Station Unit 2 Startup Test Report Summary i

NRC Docket No. 50-374

Dear Sir:

Enclosed for your inforatation and use is the LaSalle County Station Unit 2 Cycle 2 Startup Test Report Summary. This report is submitted in accordance with Technical Specification NPF-18 section 6.6.A.l.

1 LaSalle Unit 2 Cycle 2 began commercial operation on June 16, 1987 following a refueling and maintenance outage. The Unit 2 Cycle 2 core loading consisted of 224 fresh GE7B (8x8 prepressurized barrier) bundles and 540 reload bundles. The core loading represents a transition from

" conventional loading" to " Control Cell Core".

A Fine Motion Control Rod Drive (FMCRD) was installed during the outage at core location 02-43..

Support of this demonstration project required Technical Specification changes to include special Shutdown Margin-allowances for the FMCRD.

The startup test program was satisfactorily completed on July 11, 1987. All test data was reviewed in accordance with the applicable test j

procedures, and exceptions to any results were evaluated to verify

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compliance with Technical Specification limits and to ensure the I

acceptability of subsequent' test results.

A startup test report is required to be submitted to the Nuclear

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Regulatory Commission (NRC) within 90 days following resumption of commercial power operation.

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i 8709140050 870908 t

PDR ADOCK 05000374, P

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A l _ September 85 1987 US NRC Attached are the evaluation results.from the~f'ollowing tests:

-Core verification-

-Shutdown Margin Subcritical Demonstration

-Shutdown Margin Test (In-sequence. critical)

-Reactivity-Anomaly Calculation (Critical'and Full Power)

-Scram Insertion Times

-Core Power Distribution Symmetry Analysis If you have any additional questions'concerning this matter'please contact this office.

t y yours, 1

C. M. Allen Nuclear Licensing Administrator Attachments cc: Regional Administrator - RIII NRC Resident Inspector - LSCS Paul Shemaski - NRR 3555K 1

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LTS-1100-1, SHUTDOWN MARGIN TEST PURPOSE The purpose of this test is to demonstrate, from a normal in-

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sequence critical, that the core loading has been limited such that the reactor will be.subcritical throughout the operating cycle with-the Fine Motion Control Rod Drive (FMCRD) 02-43 and the strongest other control rod in the full-out position (position 48) and all other rods fully inserted.

CRITERIA If a shutdown margin (SDM) of 1.232% AK/K (0.38% AK/K + R) cannot be demonstrated with the strongest control rod and 02-43 fully

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j 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 0.852% AK/K, with peak reactivity occurring at 6,000 MWD /ST into the cycle.

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RESULTS AND DISCUSSION J

The beginning-of-cycle S'A was successfully determined from the c

initial critical data.

The initial Cycle 2 critical occurred on June 16, 1987, on Jontrol rod 30-11 at position.04, using an A-2 sequence.

The moderator temperature was 132*F and the reactor period was 192 seconds.

Using ro(. orth 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 2.81%

A K/K (see Table 1).

The SDM demonstrated exceeded the 1.232% AK/K required to satisfy Technical Specifications 3.1.1 and 3.10.9.

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TABLE 1 SHUTDOWN MARGIN CALCULATION Critica1' Rod = 30-11'8 04

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Worth of Strongest Rod = 0.02468 A K/K (1),

Worth of Control Rods Withdrawn to Obtain Criticality:

1 24 Group 1 rods at 48 = 0.03470 oK/K (2) 20 Group 2 rods at 48 = 0. 01807 AK/K (3) l 13 Group 3' rods at 04 = 0.00136 4K/K (4)

Temperature Correction = -0.0010 4K/K (5) for Tm = 132*F Period Correction = 0.00035 /.K/K (6) for Period = 192 seconds Shutdown Margin Keff:

(5) + (6)

(4)

SDM Keff = 1.0000 + (1) - (2) - (3)

= 0.9719 g

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.l SDM = (1.000 - (SDit Keff)) *.100 = 2. 01% AK/K i

l The " Worth of Strongest Roda accounts for the FMCRD and the

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strongest other rod.

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LTS-1100-2, CHECKING FOR REACTIVITY ANOMALIES d

PURPOSE i

The' purpose'of this test is to compare the actual and predicted critical rod configurations.to detect any unexpected reactivity effects in.the reactor core.

CRITERIA In accordance with Technical Specification 3.1.2, the reactivity-1 equivalence of the difference between the actual control rod-

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density and the predicted control rod density shall not exceed 1%

g AK/K.

If the difference does exceed 1% A h/K, the Core Management 1

Engineers (General Electric Company and Commonwealth Edison l

Company) will be promptly notified to investigate the anomaly.

The i

cause of the anomaly-must'be determined, explained, and corrected for continued operation of the unit..

RESULTS AND DISCUSSION i

Two reactivity anomaly calculations were successfully performed i

during the Unit 2 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 June 16, 1987,-with control rod 30-11 at position 04, using an A-2 sequence.

The moderator i

temperature was 132*F and the reactor period was 192 seconds.

Using rod worth information, moderator temperature reactivity 1

corrections, and period reactivity corrections supplied by General j

Electric (in the Cycle Startup Package), the actual critical was determined to be within 0.143% AK/K of the predicted critical (see Table 2).

The difference determined is within the 1% AK/K allowed

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by Technical Specification 3.1.2.

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The reactivity anomaly calculation for power operation was

.j performed on July 6, 1987 with Unit 2 at 92.5 % power at a cycle

.j exposure of 192 MWD /ST, at equilibrium cond.t

'ns. The predicted-

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notch inventory from the vendor supplied dat. <as 520 notches.

The' j

actual notch inventory, corrected for' power and flow off rated, was 602 notches.

Using'the notch worth provided by the vendor,'the resulting anomaly was -0.16 % A K/K.

This value is within the 1%

i AK/K criteria of Technical Specification 3.1.2.

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1 TABLE 2 i

i INITIAL CRITICALITY COMPARISON CALCULATIONS 1

ITEM 4 K/K Keff with all rods in at 68 F

= 0.94865

• Reactivity inserted by 24 group 1 rods at position 48

= 0.03470 m i

Reactivity inserted by 20 group 2 rods at position 48

= 0.01807

• Reactivity. inserted by 13 group 3 rods at position 04

= 0.00136

• Predicted Keff at actual critical rod pattern (68'F)

= 1.00278 i

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Reactivity associated with the measured reactor period (period correction for 192 second period)

= 0.00035

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Reactivity associated with moderator temperature

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(132'F actual, 68'F predicted)

= 0.00100

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

• 100%

= 0.143%4 K/K 1

"LaSalle Unit 2 Cycle 2 Startup Package", supplied by General j

l Electric Company.

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l LTS-1100-4, SCRAM INSERTION. TIMES

-1 PURPOSE

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l The purpose of thir test is to demonstrate that the control rod

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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).

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CRITERIA The maximum scram insertion time of each control rod from ine fully f

withdrawn position-(48) to notch position 05, based on de-energization of the scram pilot valve solenoids as time zero, shall j

not exceed 7.0 seconds.

i The average scram insertion time of all operable control rods from the fully withdrawn position (48), based on de-energization of the 1

scram pilot valve solenoids as time zero, shall not exceed any of i

the followings j

j Position Inserted From Average Scram Insertion

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Fully Withdrawn Time (Seconds)

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j 45 0.43 39 0.86 1

'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 followings i

Position Inserted From Average Scram Insertion j

Fully Withdrawn Time (Seconds) 45 0.45 39 0.92 25 2.05 f

05 3.70 RESULTS AND DISCUSSION j

Scram testing was successfully performed between June 18, 1987 and June 19, 1987.

All control rod scram timing acceptance criteria.

vere met during this test (see table).

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LTS 1100-4 CONTROL ROD DRIVE'(CRD) SCRAM RESULTS Maximus Average Average Scram Tilaes Scram Times in a Position

'of all CRDs (secs.)

Two-by-Two Array (secs.)

45-0.325 0.349 39 0.628 0.669 25

!.355 1.437 05 2.471 2.613 Maximum 90% scram time (position.05): CRD 18-23, 2.976' secs.

ave (position 39) for Minimum Critical Power Ratio determination: 0.628 seconds.

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1 LTS-1100-14, SHUTDOWN MARGIN SUBCRITICAL DEMONSTRATION I

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 suberitical throughout the operating cycle with the strongest control rod in the full-out position (position

48) and all other rods fully inserted.

However, for Unit 2 Cycle 2 the shutdown margin (SDM) requirement must be met with an increased allowance for the fully withdrawn worth of the fine motion control rod 02-43.

CRITERIA If a SDM of 1.232% AK/K (0.38% AK/K + R) cannot.be demonstrated l

with the strongest control rod fully withdrawn, the core-loading must be altered to meet this margin.

R is the difference between j

the core's beginning-of-cycle. reactivity and the peak' reactivity for the cycle.

The R value for' Cycle 2 is 0.852% AK/K, with peak i

reactivity occurring at'6,000 MWD /ST into the cycle.

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RESULTS AND DISCUSSION l

l On June 16, 1987, the local SDM demonstration was successfully performed using control rods 58-43, 54-43 and 50-39.

For' Unit 2 Cycle 2 the SDM was determined with an increased allowance for the withdrawn worth of the fine motion control rod 02-43.

With control.

rod 58-43 (symmetric to 02-43) assumed fully withdrawn, the strongest control rod is 54-43.

Control rod 50-39 is diagonally

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adjacent to 54-43.

General Electric (GE) provided, in the Cycle i

Startup Package, rod worth information'(for control rods 58-43 and 54-43 and diagonally adjacent rods 50-39 and 50-47) and'aoderator

' temperature reactivity corrections to support this test.

Using the GE supplied information, it was' determined that with control rods 58-43 and 54-43 at position 48,' rod 50-39 at position 16, a moderator temperature of 129'F, and the reactor subcritical, a SDM of 1.372%liK/K was demonstrated.. The SDM demonstrated' exceeded the 1.232% AK/K required to. satisfy Technical Specification 3.1.1',

and:

maintained sufficient' margin to the GE calculated SDM for, the core at beginning-of-cycle to' avoid criticality during the test.

( Calculated SDM at BOC =.2.667%a K/K).

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LTP-1600-17, CORE POWER DISTRIBUTION SYMMETRY ANALYSIS f.

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 f

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 j

obtained from two full core TIP sets (DD-1) and individual TIP 1

traverses (OD-2) of the reference channel by each TIP machine.

The initial TIP set was performed on July 6, 1987 at 92.5% power, and l

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then repeated on July 7, 1987 (also 92.5% power).

The average

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J total TIP uncertainty from the two data sets was 5.74%, satisfying I

the criteria of the test (less than 8.7%)

The random noise uncertainty and geometrical uncertainty were determined to be 0.71%

1 and 5.70%, respectively.

I Table 1 lists the symmetrical TIP pairs, their core locations, and their respective average deviations.

The maximum deviation between i

symmetrical TIP pairs was 15.91% for TIP pair 11-35, satisfying the I

criteria of the test (less than 25%).

An increase in TIP pair deviation was noticed for pairs 2-13, 3-20, j

5-34, and 11-35, over the expected results. The increased

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asymmetry was correlated to the LPRM strings which were replaced J

during the maintenance outage.

A total of 5 LPRM strings were l

replaced during the outage.

However, one of the 73 placement strings was installed on the symmetry axis and as such, has no j

symmetric counterpart.

This asymmetry is being evaluated by the LPRM vendor.

Since the local and total asymmetry results are i

within the criteria of 25% (local) and 8.7% (total), there is no 1mmediate concern for the validity of the power distribution calculations driven by the replacement LPRMs.

A discussion of the calculational methodology is provided below.

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The method'used to obtain the uncertainties. consisted of-j l:

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

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~. s.

l-g.

s l$n j'

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l vhere Ri] = the BASE ratio.for the ith node of TIP pair 3, l

l n = number of TIP pairs = 19.'

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i Next, the standard deviation'(expressed as a percentage) of these..

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ratios is calculated by the following e ation:

( f '.. - f 0;(1.)=

x.s u, x too a

a (I 8 n.-l)

The total TIP uncertainty (%) is calculated by dividing 6g'(%) ' by6

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because the uncertainty in one TIP reading is the desired parameter, but the measured' uncertainty is the ratio of twa TIP readings.

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

%r p

BASE (K) prmr BASE (NAk)

Mi 4:n where NT = number'of i~uns per machine.

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MT = number of machines.

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

Kr 5 through 22.

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

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. I/z

. BhsE(N,n%K)- BAS EG) l h0 845E(k)

.)( l00 g,,

m, u

l B(urx mr-t)

Finally, the TIP geometric uncertainty can be calculated by:

. i/2,

{,m:

TI - 5 I

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TABLE 1 TIP SIGNAL SYMMETRY RESULTS All numbers shown are averages from two OD-1 data sets (from 7-6-87 and 7-7-87 at 92.5% povar, equilibrium conditions),

i Symmetrical TIP Pair Absolute Percent Numbers (Core Location')

Difference TIP Pair i

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of BASE #

Deviation

• l 1 (16-09) 6 (08-17) 3.86 4.58 2 (24-09) 13 (08-25) 13.88 14.64 3 (32-09) 20 (08-33) 11.28 11.22 4 (40-09) 27 (08-41) 2.54 2.76

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5 (48-09) 34 (08-49) 8.87 13.84 l

8 (24-17) 14 (16-25) 7.81 8.02 9 (32-i7) 21 (16-33) 2.78 2.54 10 (40-17) 28 (16-41) 3.46 3.53 11 (48-17) 35 (16-49) 14.76 15.91 12 (56-17) 40 (16-57) 2.81 4.93 16 (32-25) 22 (24-33) 4.36 4.03 17 (40-25) 29 (24-41) 5.83 5.24 18 (48-25) 36 (24-49) 4.98 4.99 19 (56-25) 41 (24-57) 2.19 2.71 24 (40-33) 30 (32-41) 3.49 3.33 25 (48-33) 37 (32-49) 10.40 9.65 1

26 (56-33) 42 (32-57) 1.09 1.33 32 (48-41) 38 (43-49) 2.44 2.62 l

33 (56-41) 43 (40-57) 1.36 1.90

1. - where : Absolute Difference of BASE =

BASE - BASE q

and BASE;

• kBASE;(K)

• - where : % Deviation =

BASEa - B ASE b

• 100 3.5(EA5E(+ BASE) b 1

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LTP-1700-1, CORE VERIFICATION PURPOSE 1

The purpose of this test is to visually verify that the core is 4

loaded as intended for Cycle 2 operation.

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CRITERIA i

f Theas-loadedcoremustconformtothecyclecoredesignusedby$

the Core Management Organization (General Electric) in the reload, licensing analysie.

The core verification must be observed by member of the Commonwealth Edison Company audit staff.

Any 7'

discrepancies discovered in the loading will be promptly corrected g and the affected areas reverified to ensure proper core loading -

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[' {y Conformance to the cycle core design will be documented by a. $

permanent core serial number map signed by the audit participants.

RESULTS AND DISCUSSION j

The Cycle 2 core verification consisted of a core height check performed by the fuel handlers and two videotaped passes of the f

core by the nuclear group.

The height check verifies the proper

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seating of the assembly in the fuel support piece while the videotaped scans verify proper assembly orientation, location, and seating.

Bundle serial numbers and orientations were recorded f

during the videotaped scan, for ccmparison to the appropriate tag 1

boards and Cycle Management documentation.

On May 2, 1987, the f

core was verified as being properly loaded and consistent with the General Electric Cycle 2 core design used in the reload licensing analysis.

On May 3, 1987, the videotapes were reviewed by the Lead

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Nuclear Engineer to reverify all bundle ID's, orientation, and j

seating, j

One exception exists to the Unit 2 Cycle 2 Cycle Management Report Loading pattern.

At core location 35-02, bundle LJC125 was loaded l.

in place of bundle LJC175 as a substitute bundle for Cycle 2.

l Bundle LJC175 was determined to contain a leaking fuel rod during the sipping program conducted prior to fuel load.

The substitute assembly LJC125 was specifded by General Electric, and approval to

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load Unit 2. Cycle 2 with the substitute bundle was obtained from i

l Nuclear Fuel Services.

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