ML20154C111

Forwards Facility Startup Test Rept Summary.Evaluation Results from Core Verification,Shutdown Margin Subcritical Demonstration,Shutdown Margin Test,Scram Insertion Times & Core Power Distribution Symmetry Analysis Encl

ML20154C111
Person / Time
Site:	LaSalle
Issue date:	09/08/1988
From:	Allen C COMMONWEALTH EDISON CO.
To:	Lieberman J NRC OFFICE OF ENFORCEMENT (OE)
References
5075K, NUDOCS 8809140238
Download: ML20154C111 (12)

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/ ^ N.> One Commonwealth Edison

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First National Plaza, Chicago. lilinois 7 Address Reply tv. Post Omce Box 767

/ CNeago, Illinois 60690 0767 September 8, 1988 Mr. James Lieberman, Director Office of Enforcement U.S. Nuclear Regulatory Comission Washington, DC 20555 Subjects LaSalle County Station Unit 1 Startup Test Report Sumary NRC Docket No. 50-373 References (1): C.M. Allen letter to USNRC dated January 18, 1988 transmitting Reload LicenslLg Package for Unit 1.

, (2): NEDE-24011-P-A "General Electric Standard Application for Reactor Fuel", Revision 8.

3 Dear Sirs Enclosed for your information and use is LaSalle County Station Unit 1 Cycle 3 Scartup Test Report Summary. This report is submitted in accordance with Technical Specification NPF-11, Section 6.6.A.1.

LaSalle Unit 1 Cycle 3 began comercial operation on June 16, 1987 following a refueling and maintenance outage. The Unit 1 Cycle 2 core loading consisted of 224 fresh GE 8x8EB bundles and 540 relaod bundles. The new fuel has an option for multiple lattice types (i.e., axial soned guidelines).

The startup test program was satisfactorily completed on August 6, 1987. All test data was reviewed in accordance with the applicable test procedures, and exceptions to any results were evaluated to verify compliance with Technical Specification limits and to ensure the acceptability of subsequent test results.

1. startup test report is required to be submitted to the Nuclear Regulatory Comission (NRC) within 90 days following resureption of comercial power operation.

8009140230 000900 PDR ADOCK 05000373 /

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P PNU t I g

US NRC September 8, 1988 s

Attached are the evaluation results from the following tests:

- Core Verification

- Shutdown Margin Suberitical 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.

Very truly yours,

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f C. M. Allen Nuclear Licensing Administrator 1m Attachments i

ces Regional Administrator - RIII HRC Resident Inspector - LSCS i Paul Shemanski - NRR l i

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LTP-1700-1, CORE VERIFICATION i

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. The purpose of this test is to visually verify that the core is l loaded as intended for Cycle 3 operation. l CRITERIA

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The as-loaded core sust confore to the cycle core design used by the Core Nanagement Organization (General Electric) in the reload t licensing analysis. The core verification must be tsbeerved by a i seeber of the Consonwealth Edison Company audit staff. Any l discrepancies discovered in the loading vill be promptly corrected and the affected areas reverified to ensure proper core loading  ;

prior to unit startup.

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Conformance to the cycle core design vill be documented by a  !

, permanent core serial number esp signed by the audit participants.

l f RESULTS AMD DISCUSSION

! The Unit 1 Cycle 3 core verification corisisted of a core height I t 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 support piece while the l videotaped scans verify proper assembly orientation, location, and i seating. Bundle serial numbers and orientations were recorded L during the videotaped scan, for comparison to the appropriate tag i j boards and Cycle Management documentation. On June 8, 1988, the L core was verified as being properly loaded and consistent with the 7 i General Electric Cycle 3 Cycle Nanagement Report. On June 9, 1988, l the videotapes were reviewed by the Lead Nuclear Engineer to  !

reverify all bundle ID's, orientation, and seating.  !

The core loading t!iffered from the Reference Core Loading Pattern

(transeitted to the Nuclear Regulatory Coseission as Attachment F to Reference 1) assuwd in the reload licensing analysis in that

the coro loading did not utilize twenty (20) SCRB176 fuel assemblies. These 20 assemblies were replaced with 20 SCRB219 fuel i assemblies in accordance with General Electric procedures. General l Electric re-examined the six parameters specified in Section 3.4.3 ,

!. of Reference 2. General Electric determined that only one r l parameter, cold shutdown eargin, would be affected by the bundle l substitutions. Since cold shutdown eargin was recalculated for the l j Statiot. Use Loading Plan (i.e., the as loaded core) and found to be [

within aceitptable sargins, the re had license analysis 1* not ,

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1 LTS-1100-14, SHUTDOWN MARGIN (SDM) SUBCRITICAL DLHONSTRATION PURPOSE The purpose of this test in to demonstrate, using the adjacent rod suberitical 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.

CRITERIA If a SDM of 0.709% AX/K (0.38% 4K/X + R) cannot be demonstrated with the strongest control rod fully withdrawn, the core loading must be altered to meet this margin. R is the reactivity difference between the core's beginning-of-cycle SDM and the minimus SDM for the cycle.

The R value for Cycle 3 is Q 329% AX/K, with the minimum SDM occurring at 5,000 MWD /ST into the cycle.

RESULTS AND DISCUSSION On July 4,1988, the local SDM demonstration was successfully l performed using control rods 18-55 and 22-51. Control rod 22-51 is diagonally adjacent to 18-55, the strongest rod at beginning-of-cycle. General Electric (GE) provided, in the Cycle Startup Package, rod worth information (for control rods 18-55 and diagonally adjacent rods 22-51 and 14-51) and moderator temperature reactivity corrections to support this test. Using the GE supplied information, it was determined that with control rod 18-55 at position 48 and rod 22-51 at position 16, a moderator temperature of 158'F, and the reactor subcritical, a SDM of 0.752% AX/K was demonstrated. The SDM demonstrated exceeded the 0.709% AX/X required to satisfy the test criteria, and maintained sufficient margin to the GE calculated SDM for the core at beginning-of-cycle (1.598% AK/K) to avoid criticality during the test.

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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 suberitical 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 0.709% 6K/K (0.38% 4K/K + R) cannot be demonstrated with the strongest control rod fully withdrawn, the core loading sust be altered to meet this margin. R is the difference between the core's beginning-of-cycle SDM and the minious SDM for the cycle. The R value for Cycle 3 is 0.329% AK/K, with einimus SDN occurring at 5,000 MWD /ST into the cycle.

RESULTS AND DISCUSSION The beginning-of-cycle SDM was successfully determined from the initial critical data. The initial Cycle 3 critical occurred on July 4, 1988, on control rod 34-55 at position 18, using an A-2 sequence. The moderator temperature was 155'T and the reactor period ves 87 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 1.228%

A K/K (see Table 1). The SDM demonstrated exceeded the 0.709% AK/K required to satisfy Technical Specification 3.1.1.

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TABLE 1 SHUTDOWN MARGIN CALCULATION Critical Rod = 34-55 0 18 Worth of Strongest Rod = 0.02756 AX/K (1)

Worth of Control Rods Withdrawn to Obtain Criticality 24 Group i rods at 48 = 0.03626 AX/K (2) 4 Group 2 rods at 48 = 0.00527 AX/X (3) 1 Group 2 rod at 18 = 0.00096 AX/X (4)

Teeperature Correction = -0.0020 AX/K (5) for To = 155*F Period Correction = 0.00065 AX/K (6) for Period = 87 seconds Shutdown Margin Xeff SDN Xeff = 1.0000 + (1) - (2) - (3) - (4) - (5) + (6)

= 0.94772 eX/C SDN = (1.000 - (SDM Xeff)) e 100 = 1.228% 4X/X

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1 LTS-1100-2, CHECXING FOR REACTIVITY ANOMALIES

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PURPOSE The purpose of this test is to compare the actual and predicted )

critical rod configurations to detect any unexpected reactivity l effec +1 in the reactor core.

1 CRITERIA i l

In accordance with Technical Specifiestion 3.1.2, the reactivity l equivalence of the difference between the actual control rod density and the predicted control rod density shall not exceed 1X

AX/K. If the difference does exceed 11 4X/X, the Core Management

Engineers (General Electric Company and Commonwealth Edison Company) vill be promptly notified to investigate the anomaly. The cause of the anomaly sust 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 3 Startup Test Program, one from the initial critical and the second froe steady-state, equilibrium conditions at approximately 87 percent of full power.

l The initial critical occurred on July 4, 1988, with control rod I 34-55 at position 18, using an A-2 sequence. The moderator l temperature was 155'F and the reactor period was 87 seconds. Using

, rod worth information, moderator tempercture reactivity corrections, and period reactivity corrections supplied by General ,

Electric (in the Cycle Startup Fackage), the actual critical was  !

I determined to be within -0.370% oM/K of the predicted critical (see i Table 2). The difference determined is within the 11 AX/K criteria l l of Technical Specification 3.1.2.

l The reactivity anomaly calculation for power operation was i performed on July 18, 1944 with Unit i at 66.91 power at a cycle '

exposure of 120 MWD /ST, at equilibrius conditions. The predicted '

notch inventory from the vendor supplied data was 990 notches. The actual notch inventory, corrected for power and flow values which were less than rated, was 1060 notches. Using the notch worth provided by the vendor, the resulting anosely was -0.141 4X/K.

I This value is within the 11 AX/X criteria of Technical Specification 3.1.2.

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TABLE 2 INITIAL CRITICALITY COMPARISON CALCULATIONS IIEH, A K/K Keff with all rods in at 64*F = 0.95646 e Reactivity inserted by 24 group 1 rods at position 48 = 0.03626 e Reactivity inserted by 4 group 2 rods at position 48 = 0.00527 e Reactivity inserted by 1 group 2 rod at position 18 = 0.00096 e Predicted Keff at actual critical rod pattern (68*F) = 0.99495 Reactivity associated with the measured reactor period (period correction for 87 second period) = 0.00065 e Reactivity associated with moderator temperature (15b'F actual, 68'F predicted) = 0.002 e Reactivity Anomaly = I(predicted Keff - 1) - (period correction) - (temperature correction)) a 100% s-0. 370% AK/K e 'LaSalle Unit 1 Cycle 3 Startup Package', supplied by General i Electric Company.

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LTS-1100-4, SCRAM INSERTION TIMES PURPOSE The purpose of this test is to demonstrate that the control rod scrae 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 scrau 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 scran 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 screa 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 scran pilot valve solenoids as time zero, shall not exceed any of the following Position Inserted From Average Screa Innertion Fully Withdrawn Time (Seconds) 45 0.45 39 0.92 25 2.05 l 05 3.70 t

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V RESULTS AND DIbCUSSION Scram testing was successfully performed between July 8,1988 and July 9, 1988. All control rod scram timing acceptance criteria were set during this test. The results of the test are given below.

Maximus Average Average Screa Times Ecrea Times in a Position of all CRDs (secs.) Two-by-Two Array (secs.)

45 0.324 0.339 39 0.621 0.645 25 1.331 1.381 05 2.418 2.525 Maximum 90% scram time (position 05): CRD 18-51, 2.656 secs.

1(ave (position 39) for Minimum Critical Power Ratio determination: 0.621 seconds.

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4 LTP-1600-17, CORE POWER DISTRIBUTION SYMMETRY ANALYSIS PURPOSE i

The purpose of this test is to verify the core power symmetry and '

the reproducibility of the TIP readings.

CRITERIA l t

The total TIP uncertainty obtained by averaging the uncertainties l for all data sets must be less than 8.7% '

The groes check of the TIP signal syneetry should yield a maxieue  !

deviation between syneetrically locateo pairs of less than 25%.

RESULTS AMD DISCUSSIOM Core power syssetry calculations were performed based upon data (

obtained from two full core TIP sets (0D-1). The initial TIP set was performed on August 2, 1988 at 93.0% power, and then repeated on August 3, 1988 at 92% power. The average total TIP uncertainty from the two data sets was 3.487%, satisfying the criteria of the test (less than 8.7%). The average standard deviation was 4.932%.

Table 3 lists the syneetrical TIP pairs, their core locations, and their respective average deviations. The maxieue deviation between syneetrical TIP pairs was 8.32% for TIP pair 33-43, satisfying the criteria of the test (less than 254).

A discussion of the calculational methodoingy is provided below.

The method used to obtain the uncertainties consisted of calculating the averagt of the nodal BASE ratio of TIP pairs by:

n n ,

R= isn u{P {R Q where RiJ = the BASE ratio for the ith node of TIP pair J, l n = number of TIP pairs = 19.

Next, the standard deviation (expressed as a percentage) of these l ratios is calculatcd by the following equations  :

- J A- S l

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TR

• gg n l00 (I bn = l) ~ \

The total TIP uncertainty (X) is calculated by dividing Fg(4) by d because the uncertainty in one TIP reading is the desired parameter, but the seasured uncertainty is the ratio of two TIP l readings.  ;

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TABLE 3 TIP SIGNAL SYMMETRY RESULTS All numbers shown are averages from two 0D-1 data mets (from 8-2-88 and 8-3-88 at 93% and 921 power, respectively. -

Syssetrical TIP Pair Absolute Percent Mumhers (Core Location) Difference TIP Pair a 5 of BASE # Deviations 1 (16-09) 6 (08-17) 1.14 1.40 2 (24-09) 13 (08-25) 3.08 2.90 3 (32-09) 20 (08-33) 6.73 5.61 4 (40-09) 27 (08-41) 4.71 4.54 5 (48-09) 34 (08-49) 2.19 3.43 8 (24-17) 14 (16-25) 2.41 1.95 9 (32-17) 21 (16-33) 0.07 1.31 10 (40-17) 28 (16-41) 2.17 1.73 11 (48-17) 35 (16-49) 4.04 3.99 12 (56-17) 40 (16-57) 0.20 1.36 16 432-25) 22 (24-33) 3.32 2.83 17 (40-25) 29 (24-41) 1.49 1.50 it (48-25) 36 (24-49) 8.14 6.84 19 (56-25) 41 (24-57) 5.40 6.45 24 (40-33) 30 (32-41) 4.11 3.55 25 (48-33) 37 (32-49) 8.42 7.21 26 (56-33) 42 (32-57) 3.91 3.82 32 (48-41) 38 (40-49) 6. % 5.82 33 (56-41) 43 (40-57) 6.31 8.32

1. - where : Absolute Difference of M = , E R , - B H ,.

and H Rg=ft(BASE (X)

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e - where : 1 Deviation = EH. - EH b Ietoo 0.5(BA5( + BASE,1,,

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