ML20034A306

Cycle 4 Startup Test Rept

ML20034A306
Person / Time
Site:	LaSalle
Issue date:	04/11/1990
From:	Morgan W COMMONWEALTH EDISON CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NUDOCS 9004230019
Download: ML20034A306 (16)

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Downers Grov6, Illinois 60515 April 11, 1990 1

U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555

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Subject:

LaSalle County Station Unit 1 Startup Test Report Sununary HRC_Dz ket No. 50-373

Dear Sir:

Enclosed for your information and use'is the LaSalle County Station.

Unit 1 Cycle 4'Startup Test Report Summary.

This report is~ submitted in-i accordance with Technical Specification NFP-18, Section 6'.6.A.1.

LaSalle Unit 1 Cycle 4 began commercial operation on January 10,.

1990 following a refueling and maintenance-outage. The' Unit l' Cycle 4 core loading consisted of 172 f resh CE 8x8 NB (CE9B) Fuel Bundles and 592 Reload Bundles. The new Fuel (CE 9B) has an option for. multiple lattice types (i.e., axial zone enrichment and gadolinia).

The startup test program was satisfactorily completed on; February 17, 1990. 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, b

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 9004230019 900411 PDR ADOCK 05000373 p

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-.,,,x U.S. NRC -April 11 -1990 If you have any questions concerning this matter, please contact this office.

Very truly yours, W

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. Morgan 1

Nuclear Licensing Administrator

- 1 cci A.Bert Davis-Regional Administrator, RIII NRC Resident Inspector, LSCS R. Pulsifer - Project Manager, NRC i

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ATTACHMENT A

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Summary of Unit 1 Cycle 4 Startup Test Program 4

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.E L41A113 Unit i Cvele 4 Startuo Test Report 1

SUMMARY

LaSalle Unit 1 Cycle 4 began commercial operation on January 10, 1990 following a refueling and maintenance outage.

The Unit 1 Cycle 4 core j-loading consisted of 172 fresh GE8x8NB-(GE98) fuel bundles and 592 reload bundles. The new fuel (GE9B) has an option for multiple lattice-types (i.e.,

axial zoned enrichment and gadolinia).

The expected changes in the' operating.

characteristics of the new fuel design are described in LOSR-69-072, which-5 evaluated the incorporation of this fuel cesign.

All applicable test results (neutron instrument calibration, computer monitoring results) indicate expected core performance with the new fuel design.'

t A comprehensive startup testing program was performed during startup and power'

'I ascension.

The startup program included i

- local and in-sequence shutdown margin tests.

- reactivity anomaly calculations at initial cr,itical and full power.-

- nuclear instrument performance verifications (SRM, IRM,.APRM response and overlap checks).

- instrument calibrations (LPRM, APRM, TIPS, core-flow).

- control rod drive friction and full' core scram timing.

- LPRM responses to control rod movement.

- process computer verification, comparison to off-line calculation.

- recirculation system performance data.

- baseline stability data acquisition.

The startup test program was satisfactorily completed on January 25, 1990 with l

the exception of the Recirculation System Performance test,_which was ccmpleted on February 17, 1990.

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 to ensure the.

acceptability of subsequent test results.

A startup test report must be submitted to the Nuclear Regulatory Commission (NRC) within 90 days following resumption of commercial power operation ~(in accordance with Technical Specification 6.6.A.1).

The startup test report presented in this on-site review (Attachment B).contains results (evaluations) from the following test:

- Core Verification

- Shutdown Margin Subcritical Demonstration

- Shutdown Margin Test (In-sequence critical)

- Reactivity Anomaly Calculation 1 Critical and Full Power)

- Scram Insertion Times-i

- Core Power Distribution Symmetry Analysis A full evaluation of the startup test program is included with the evaluation of LTP-1600-37 (On-Site Reviev 90-05), Unit Startup Test Program.

Data from each startup is available at LaSalle Station.

ELNDINGS AND RECOMMENDATIONS Based upon the preceding discussion and the review of the startup test report, On-Site Review recommends submittal of the "LaSalle County Nuclear Power Station Unit 1 Cycle 4 Startup Test Report' (Attachments B and C) to the HRC in accordance with Technical Specification 6.6.A.l.-

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LaSalle County Nuclear Power Station I

' Unit 1 Cycle 4 Startup Test Report'-

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LTP-1700-a, CORE-YERIFICATION j

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

loaded as intended for Unit 1 Cycle 4 operation.

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CRITERIA

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The as-loaded core _must conform to the cycle core design used by

+he Core Management Organization (General Electric) in the reload' licensing analysis.

The core verification must be observed by a t

member of the Commonwealth Edison Company audit staff.- Any.

j discrepancies' discovered in the locding vill be promptly corrected.

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and the affected areas reverified.to ensure proper core loading-r

. prior to unit startup.

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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 Unit 1 Cycle 4 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 l-proper seating of the assembly in the fuel' support piece while the-(

videotaped scans verify proper. assembly orientation, location,' and l,,

seating.

Bundle serial numbers and orientations were recorded.

during the videotaped scans, for comparison to-the appropriate tag l

l boards ~and Cycle Management documentation.. On December 5, 1989, the core was verified as being-properly. loaded-and consistent with the General Electric Cycle 4 Cycle Management-Report and the Final.

Station Use Loading Plan.. On December 6, 1989,-the videotapes were-reviewed by the Lead Nuclear Engineer to reverify'all bundle ID's, orientation, and seating.

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The core loading differed from the Reference Core Loading Pattern assumed in the reload licensing. analysis-(Reference.1) in that the core loading did not utilize one (1) BP8CRB299L; fuel: assembly-which had been scheduled for use during Cycle 4.-

This change was j

reviewed and.found to be acceptable against the requirements of Reference 2.

'The change was required as a result of the lenker i

fuel assembly which was identified during the Unit 1 third refuel outage.

The lenker assembly was not loaded.- The core pattern was shuffled and.this assembly was replaced with a BCRB219 fuel l

assembly in accordance with General Electric procedures.

General

-Electric re-examined the parameters specified in Section 3.4.2 of L

Reference 2.

They ceturnined that'only one parameter, cold shutdown margin, would be affected by the bundle substitutions.-

Since celd shutdown margin-was recalculated for the Station Use Loading Plan (i.e., the as loaded core).and found to be within 7

acceptable margins, the reload license analysis is not affected.

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L, 'TS-1100-14, SHUTDOWN MARGIN (SDML SUBCRITICAL-DEMONSTRATION l

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PURPDSE j

The purpose of this test is to demonstrate, using the adjacent rod l.'

subcritical method, that the core loading has been~ 1imited such that the reactor will be subcritical throughout the operating cycle _with-l the strongest control rod in the full-out position (position 48)~and L

all other rods fully inserted.-

CRITERIA' If a SDM of 0.38% A K/K (0.38% AK/K + R) cannot be demanstrated with-the strongest control rod fully withdrawn,Lthe-core loading must be L

altered to' meet this margin.. R is the reactivity difference between the core's beginning-of-cycle SDM and the minimum SDM for the cycle.

The R value for Cycle 4 is 0.0% A K/K, with the: minimum SDH occurring at 0.0 MWD /ST into the cycle.

RESULTS AND DISCUSSION On January 4, 1990, the local:SDM demonstration was successfully1,

l performed using control rods-38-55 and'42-51. ' Control rod 42-51.is-(

diagonally adjacent to 38-55, the strongest' rod at beginning-of-'

cycle.

General Electric (GE) provided, fin the Cycle Startup Package, red worth information (for control rods 38-55 and diagonally adjacent

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rods 42-51 and 34-51) and moderator temperature reactivity corrections to support this test..Using the GE supplied-information, it was determined that with_ control rod 38-55 at position 48 and-rod 42-51 at position 08, a moderator-temperature of'135'F,:and-the-reactor suberitical, a SDM of 0.502% A K/K was demonstrated.= The SDM-l demonstrated exceeded the 0.38% AK/K required to satisfy theitest I

criteria, and maintained sufficient margin 1to the GE_ calculated SDM for the core at beginning-of-cycle -(1.684X A'K/K) ~to avoid criticality during the test.

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LTS-1100-1, SHUTDOWN MARGIN TEST i

PURPOSE s

The' purpose of this teet is *o demonstrate, from a normal'in.

sequence critical, that the core loading has been111 sited such that -

.the reactor will be suberitical throughout the operating cycle with

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the strongest control rod-in the full-out position (position 481.

and all other rods fully inserted.

CRITERIA If a shutdown margin (SDM) of.38% dX/K (0.38% 4K/K'+ R)'cannot be demonstrated with the strongest control rod fully withdrawn, the

'3 core loading must be altered'to meet this margin.. R is-the reactivity difference between the core's=beginning-of-cycle ~SDN and-((

the minimus SDN for the cycle.

The R value for Cycle 4 is 0.0%

bE/K, i.e., the minimus SDM occurs at the beginning. of cycle.

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RESULTS AMD DISCUSSION r

The beginning-of-cycle SDN was successfully determined from the.

b initial critical data.

The initial Cycle 4' critical occurred.on January 4, 1990, on control rod 18-15 at position 16,.'using an A-2 sequence.

The moderator temp 6cature was 141 'F and the reactor period was 290 secands.

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 SDN was determined to be-1.717%

AK/K (see Table 1).

The SDM demonstrated exceeded the.38% AK/K required to' satisfy Technical Specification 3.1.1.

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TABLE 1 l

SHUTDOWN MARGIN CALCULATIO!!

n Critical Rod = 19-15 0 16 Worth of Strongest' Rod.

= 0.02839 K/K (1)

Worth of Control Rods Withdrawn to Obtain Criticality:

24 Group 1 rods st 48 = 0.03525 K/K (2) 7 Group 2 rods at 48 = 0.01198 K/K (3)

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1 Group 2 rod at.16L

= 0.00024 K/K.

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Temperature Correction.= -0.0017 K/K (5) for Tm = 141 F.

Period Correction = 0.00021 'K/K (6)

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- for Period = 290 seconds ke.

Shutdown Margin Keff:

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

(3) - (4)

(5) +-(6)-

= 0.98283 K/K SDM = (1.000 - (SDH Keff): a 100 s-l.717% K/K

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L.TS-1100-2,. CHECKING FOR REACTIVITY ANOMALIES-

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

4 CRITERIA In accordance'with Technical Specification 3.1, 2, 'the reactivity equivalence of. the-.differenceL between..the.' actual' control rod:

density and the! predicted control rod density shall not exceed-'1%J AK/K.

.If the difference does exceed'1% AK/K, - the Core ' Management Engineers (General Electric. Company and Commonwealth Edison s

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.

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RESULTS AND DISCUSSION Two reactivity enomaly -calculations: _ vere successfull'y' performed during.the Unit 1 Cycle 4.Startup Test Program, one =from the initial critical and 'the second from steady-state, equilibrium conditions at approximately 97' percent of full power.;

The initial critical: occurred on Jannuary 5, 1990, with control rod' 18-15 at-position 16, using. an :A-2 sequence.

The moderator temperature was 141* F and the__ reactor period was 290 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 0.033% A K/K of the predicted critical' (see Table 2).

The difference determined is within the 1% AK/K criteria of Technical Specification 3.l'.2.-

The reactivity anomaly calculation for power-operation-was performed using data from January 22,~ 1990 with Unit 1 at 96.7%~

= quilibrium power at a cycle exposure of 189.5 MWD /ST, at e

conditions.

The predicted notch inventory from the vendor supplied data was 540 notches.

The actual notch inventory'was 446 notches.

Using the notch worth provided by the vendor,.the resulting anomaly-was 0.18% A K/K.

This value is within the 1%

A K/K criteria of Technical Specification 3.1.2.

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TABLE 2~

INITIAL CRITICALITY COMPARISON CALCULATIONS K/K-11]2L Keff with all rods in at 68'T

~= 0.95477-e F

Reactivity inserted by 24 group'l rods at position ~48

= 0.03525 e' j

Reactivity inserted by 7 group 2 rods at position 48

= 0.01198 e Reactivity inserted by 1 group 2 rod at position 16

-= 0.00024 u Predicted Keff at actual critical rod' pattern (68'F)

= 1.00224 s

Reactivity associated with the measured reacto.

period (period correction for 290 second period)

= 0.00021La j

Reactivity associated with moderator temperature-(141*F actual, 68'F predicted)

= -0.0017 e 1

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Reactivity Anomaly =-((predicted Keff - 1) - (period correctioni + (temperature correction)) n 100%i

' = ~ 0. 033% A K/K

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"LaSalle Unit 1 Cycle 4 Startup Package *, supplied by General; i

u-Electric Company.

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LTS-1100-4, SCRAN INSERTION TIMES 1

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The purpose of this test is to demonstrate that the~ control rod j

scram insertion times are within the operating limits set forth by.

the Technical Specifications (3.1.3.2, 3.1.3.3, 3.1.3.41 m

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.2ero, shall not exceed 7.0 seconds.

The average scram insertion time of a11' operable control rods;fros' i

the fully withdrawn position'(48), based on'de-energization of:the scram pilot valve solenoids as time zero,Jshall not'_ exceed any of the following:

Position Inserted From Average Scram Insertion Ful1Y Withdrawn Time (Seconds)

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45 0.43-

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(j 39 0.86 25 1.93 05 3.49 The average. scram insertion time, from.the fully withdrawn position.

1 (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 l

not exceed any of the following Position Inserted From Average Scram Insertion-'

Fully Withdrawn Time (Seconds) 45 0.45 39 0.92L 25 2.'05 -

05 3.70 I

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RSSULTS AND DISCUSSION i-Scram testing was successfully performed between January 10; 1990 1

and January 11, 1990.

All control rod scram timing acceptance criteria were met during this test.

The results of the test are given below.

l Haximus Average Average Scran Times Screa Times in a Position of all CRDs (secs.)

Two-by-Two Array ( secs. )

• 45 0.324 0.339 1

39 0.625 O.653 25 1.352 L.459 05 2.467 2.675 I i Maximum 90% scram time (position 05),CRD 42-11, 2.848 secs.

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T' ave (position 39) for Minimus Critical Power Ratio Limit' l

determination: O.625 seconds.

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I LTP-1600-17, CORE POWER DISTRIBUTION SYMMETRY ANALYSIS

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

l the reproducibility of the TIP readings.

l CRITERIA The total TJP uncertainty obtained by_ averaging the uncertainties l

for all data sets must be less then 8.7%

The gross check of the TIP eignal mytmetry should yield a anximum' l

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 sets (0D-1).

The initial TIP' set was performed on January 18, 1990 at 86.1% power and the second on-

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January 18, 1990 at 86.1% power..The average total TIP-uncertainty..

4 from the two data sets was 3.586%, satisfying the criteria of the; r

test (less then.8.7%I.- The average standard deviation was 4.OSX.'

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Table 3 lists the symmetrical TIP pairs, their core locations,' and-(c_"

their respective average deviations.

The maximum deviation between-symmetrical TIP pairs was 14.88% for TIP pair 19-41, satisfying the-(g criteria of the test (less than 25%).

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i The results of the Randon'Neise Uncertainty and Geometric _ Noise t

l vstre 0.859% and 3.96X, respectively. -

i A discussion of the calculational methodology-1s provided below.-

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

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

n = number of TIP pairs = 19.

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

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0%-R) q=

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(1% n - 1) l The total TIP uncertainty (%) is calculated by dividing Q (%) by E 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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l-TIP SIGNAL SYMMETRY RESULTS 1

All numbers shown are averages from t'vo OD-1 data sets (from i

1-18-90 and 1-18-90 at 86.1% and 86.1% power, respectively.

l Symmetrical TIP Pair Absolute Percent i

l' Numbers (Core Location)

Difference TIP Pair i

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of BASE #

Deviations l'(16-09) 6 (08-17) 2.82 2.61 i

j 2 (24-09) 13 (08-25)-

5.67-4.89 I

3 (32-09) 20-(08-33) 5.95 5.32 4-(40-09)

'27 (08-41) 2.92-

'2.57 5 (48-09) 34 (08-49) 0.62 0.79 i

8 (24-17) 14 (16-25) 4.12 3.43 r

9 (32-17) 21 (16-33) 8.15-6.21 10 (40 171 28 (16-41) 1.63 1,46 11 (48-17) 35 (16-s9) 5.43 4.94

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12 (56-17) 40 (16-57).

2.72 3.88.

li 16 (32-25) 22 (24-33)

'5.41 4.97 17 (40-25) 35 (24-41) 0.93 0.77 18 (48-25)-

36 (24-49) 1.50 1.25 19 (56-25) 41 (24-57)'

13.86 14.88'

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24 (40-33) 30 (32-41) 7.10 6.23 25 (48-331 37 (32-49) 4.69 3.71 26 (56-33) 42 (32-57) 3.06 3.45 32 (48-41) 38 (40-49) 0.32-0.29 33 (56-41) 43 (40-57) 4.77 5.30

1. - where Absolute Difference of BH E =

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=h BASE;(K) and BASEt a - where

% Deviation =

BASE - BASE e 100 0.5(BASE + BASE )

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.j s./f ATTACHMENT C List of References 1.

Letter from G. J. Diederich to the Vice president of BWR Operations, Superintendent of Of f-Site Reviews, and~ the. Assistant View President of Quality Programs and Assessment, dated' December 8, 198?, transmitting LaSalle On-Site Review 89-072.

2.

NEDE-24811-p-A,_" General Electric Standard Application for Reactor-Fuel", Revision 9.

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