ML20244B755
| ML20244B755 | |
| Person / Time | |
|---|---|
| Site: | LaSalle  |
| Issue date: | 06/06/1989 |
| From: | Morgan W COMMONWEALTH EDISON CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| 0156T, 156T, NUDOCS 8906130265 | |
| Download: ML20244B755 (16) | |
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Commonwealth Edison s
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72 West Adams Street, Chicago, Illinois
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V Address Reply to: Post DificiBo'x 767.
g Chicago, Illinois 60690 - 0767 June 6, 1989 U.S. Nuclear Regulatory Commission ATTH: Document Control Desk Washington, DC 20555 Subject LaSalle County Station Unit 2 Startup Test Report Summary HB.C_.DDak.tLtJin.a itQ-374 Dear Sirs Enclosed for your information and use is the LaSalle County Station Unit 2 Cycle 3 Startup Test Report Summary. This report is submitted in accordance with Technical Specification NPF-18, Section 6.6.A.I.
LaSalle Unit 2 Cycle 3 began commercial operation on February 10, 1989, following a refueling and maintenance outage. The Unit 2 Cycle 3 core loading consisted of 240 fresh GE 8x8 EB fuel bundles and 524 reload bundles.
The new fuel has an option for multiple lattice types (i.e., axial zoned gadolinia).
The startup test program was satisfactorily completed on March 18, 1989.
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.
Attached are the evaluation results from the following tests:
- Core Verification
- Shutdown Margin Suberitical Demonstration
- Shutdown Margin Test (In-Sequence Critical) 1
- Reactivity Anomaly Calculation (Critical and Full Power) l
- Scram Insertion Times
- Core Power Distribution Symmetry Analysis I
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8906130265 890606 fobf FDR ADOCK 05000374 i
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U.S. NRC June 6, 1989
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If you have any additional questions concerning this. matter, please q
contact this office.
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Very.truly yours,
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l W. E. McYgan Nuclear Licensing Administrator i
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Enclosure I
I cc Regional Administrator - Region III NRC Resident Inspector - LSCS P.C. Shemanski - Project Manager, NRR
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ATTACHMENT A Summary of Unit 2 Cycle 3 Startup Test Program I
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I LaSallo Unit-2-Cvele 3 Startuo' Test Reoort' r
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SUMMARY
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9 LaSalle Unit 2 Cycle 3 began commercial operation en February 10,11989 I
following a refueling and maintenance outage.
The Unit 2' Cycle 3 core loading consisted of 240. fresh GE8X8EB fuel bundles and'524 reload b'undles.
The new fuel has an option for multiple lattice types (i.e.',,
.l axial zoned gadolinia).-
A comprehensive startup testing program <as performed during startup and power ascension. The startup program included:
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-local and in-sequence shutdown margLn tests.
-reactivity anomaly calculations at Lnitial critical and full power;
-nuclear _ instrument performance veri fications ~(SRM. IRM, APRM response and overlap checks).
-instrument calibrations (LPRM, APRM. TIPS, core' flow).
-control red drive friction and full core scram timing.
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-LPRM responses to control rod movement.
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process computer verification, comparisen to off-line calculation.
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-recirculation system performance data, i
-thermal hydraulic stability verification.
't The startup test program was natisfactornly comp 1st'ed on March l 18, 1989.
L All test data was reviewed in accordance with the applicable test; procedures, and exceptions to any resulta were evaluated to verify compliance with Technical Specification.imita and to ensure-the neceptability of subsequent test results,
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A startup test report.must be submitted to the Nuclear Reguintory Commission (HRC) within 90 days following resumption'of commercial power l
i operation (in accordance with Technical Specification 6.6.A.1).
The startup test report presented in this en site review (Attachment B) contains results (evaluations) from the'folicving tests:
-Core Verification
-Shutdown Margin Subcritical Demonstration
-Shutdown Margin Test (In-sequer ce critical)
-Reactivity Anomaly calculation (Critical and Full Power) 3
-Scram Insertion Times l
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-Core Power Distribution Symmetr y Analysis
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1 A full evaluation of the startup test pregram is included with the 1
evaluation of LaSalle Special Test LST-89-006, LaSalle Unit 2 Cycle:3 l
Startup Test-Program.
Data from each startup test is available at 4
LaSalle Station.
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FINDING AND RECOMMENDATIONS si' i
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. Based upon the preceeding discussion and the review of,the startup test-report, On-Site Review recomraende submitt al of the 'LaSalle County, j-Huelear Pcver Station Unit 2 Cycle 3 Startup Test Report" (Attachments i
B and C) to the NRC in accordance with To:hnical Specification 6 6.A.1.
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ATTACENENT B LaSalle County Nuclear Power Station Unit 2 Cycle 3 Startup Test Report d
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i LTP-1700-1, CORE VERIF2 CAT 1]H PURPOSE The purpose of this test is to visually verify that the core is loaded as intended for Unit 2 Cycle 3 operation.
CRITERIA The as-loaded core must conform to the cycle core design used by {
the Core Management Organ 12ntion (General Electric) in the reload!
j licensing analysis.
The core verification must be observed by a member of the Commonwealth Fdison Company audit staff.
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.
I RESULT 5 AND DISCUSSION The Unit 2 Cycle 3 core verification consisted of a core height check performed by the fuel sandlers and two videotaped pusees of the core by the nuclear group The height check verifies the proper seating of the assembly in the fuci support piece while the videotaped scans verify proper assembly orientation, location, and seating.
Bundle serial numborn and orientations were recorded during the videotaped scans, for comparison to the appropriate tag boards and Cycle Management documentation.
On January 12, 1989, the core was verified as beir.g properly loaded and censistent with the General Electric Cycle 3 Cycle Management Report and the Final Station Use Loading Plan.
Or January 14, 1989, the videotapes were reviewed by the Lead Nuclear Engineer to reverify all bundle ID's, orientation, and seating.
The core leading differed frcm the Reference Core Leading Pattern assumed in the reload licensi ng analysis (Reference 1) in that the core loading did not utilize four (4) BP8CRB299L fuel assemblies which had been scheduled for ase during Cycle 3.
This change was reported to the Nuclear Regul atory Commission in Reference 2.
The change was required as a reau Lt of the leaker fuel assembly which wcs identified during the Uni; 2 second refuel outage.
The lenker assembly and its three symretric aseemblics were not 1caded.
These-four assemblies were replaced with four 8CRB219 fuel assemblies in accordance with General Electric procedures.
General Electric re-examined the parameters specified in Section 3.4.2 of Reference 3.
i They determined that only one parameter, cold shutdown margin, would be affected by the bund]e substitutions.
Since cold shutdown margin was recalculated for tre Station Use Leading Plan (i.e., thel am leaded core) and found to t e within acceptable margins, the reload license analysis is not affected.
In addition, General Electric made some other minor core loading changes which resulted in a difference betwe6n the as leaded core and Reference Core Loading Pattern.
These changes were made to improve core symmetry and to avoid having, in a particular control cell, sore than one fuel bundle which saw duty in i vater face location during the previous cycle.
These changes were also done in accordance with
i, LTS-1100-14, SHUTDOWN MARGIN (SDM) 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 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 1.058% 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 reactivity difference between the core's beginning-of-cycle SDM and the minimum SDM for the cycle.
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The R value for Cycle 3 is 0.678% 4K/K, with the minimum SDM occurring at 6,780.3 MWD /ST into the cycle.
RESULTS AND DISCUSSION On February 7, 1989, the local SDM demonstration was successfully performed using control rods 14-15 and 18-19.
Control rod 18-19 is diagonally ad]acent to 14-15, the strongest rod at beginning-of-cycle.
General Electric'(GE) provided, in the Cycle Startup Package, rod worth information (for control rods 14-15 and diagonally adjacent rods 18-19 and 10-19) and moderator temperature reactivity corrections to' support this test.
Using the GE supplied information, it was determined that with control rod 14-15 at position 48 and rod 18-19 at position 16, a moderator temperature of 149*F, and the reactor suberitical, a SDM of 1.137% AK/K was demonstrated.
The SDM demonstrated exceeded the 1.058% 4K/K required to satisfy the test criteria, and maintained sufficient margin to the GE calculated SDM for the core at beginning-of-cycle (2.307% 4K/K) to avoid criticality during the test.
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LTS-1100-1, SHUTDOWN MARGIN TEST-i PURPOSE The purpose cf 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
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If a' shutdown margin.(SDH) of '1. 058% 4R/K (0. 38%' oK/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 reactivity difference between the core's beginning-of-cycle SDN and the minimum SDH for the cycle.
The R value for. Cycle 3'is'O.678%
AK/K, with minimum SDH occurring at 6780.3 HWD/ST into the cycle.
RESULTS AND DISCUSSION l
The beginning-of-cycle SDH was successfully ' determined from the-initial critical data.
The initial Cycle 3 critical occurred on February 8, 1989, on control rod 26-39 at position 12,'using'an A-2 I
sequence.
The moderator temperature; was 153*F and the reactor j
period was 187 seconds.
Using rod worth information, moderator J
temperature reactivity corrections,'
and period reactivity corrections supplied by General Electric (in the ' Cycle. Startup Package),- the beginning-of-cycle SDH was determined to be 2.314%.
AK/K (see Table 1).
The SDH demonstrated exceeded the'1.058% AK/K j
required to satisfy Technical Specification 3.1.1.
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4 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 red in the full-out position (position 48) and all other rods fully inserted.
CRITERIA If a shutdown margin (SDM) of 1.058% 4K/K (0.38% oK/K + R) cannot be demonstrated with the strongest control rod fully withdrawn, the core leading must be 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 3 is 0.678%
AK/K, with minimum SDM occurring at 6780.3 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 February 8, 1989, on control rod 26-39 at position 12, using an A-2 sequence.
The moderator temperature was 153*F and the reactor period was 187 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 2.314%
AK/K (see Table 1).
The SDM demonstrated exceeded the 1.058% AK/K i
required to satisfy Technical Specification 3.1.1.
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TABLE 1-SNUTDOWN MARGIN CALCULATION
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. Critical Rod = 26-39 6 12 Worth.cf Strongest Rod-
= 0. 02194 A K/K (1)
Worth of Control Rods Withdrawn to Obtain Criticality:
24 Group i rods at 48 = 0.03398 aK/K (2) 16 Group 2 rods at 48 = 0.01287 4K/K (3) 1 Group 2 rod at 12
= 0.00027 AK/K (4)
Temperature Correction = -0.0017 AK/K (5) j for Ta = 153 F.
Period Correction = 0.00034AK/K (6) for Period = 187 seconds l
Shutdown Margin Keff:
l SDM Keff = 1.0000 + (1) - (2) - (3)
(5)
(6)
(4)
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= 0.97686J1K/K SDH = (1.000 - (SDM Keff)) e 100 = 2.314% dK/K l
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1 LTS-1100-2, CHECKING FOR REACTIVITY ANOMALIES i
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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 j
effecte in the reactor core. -
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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 der.sity shall not exceed 1%
AK/K.
If the difference does exceed 1% M/K, the Core Management Engineers (General Electric Company and Commonwealth Edison Company) will be promptly notificd to investigate.the anomaly.
The cause of the anomaly must be determined, explained, and corrected for. continued operation of the unit.
i' RESULTS AND DISCUSSION Two reactivity anomaly calculations were successfully performed during the Unit 2 Cycle 3 Startup Test Program, one from the initial critical and the second from steady-state, equilibrium conditions at approximately 97 percent'of full power.
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The initial critical occurred on February 8, 1989, with control rod I
26-39 at. position 12, using an A-2 sequence.
The moderator i'
temperature was 153 *F and the reactor period was 187 seconds.
Using red worth information, moderator temperature reactivity j
corrections, and period reactivity corrections supplied by General l
Electric (in the Cycle Startu'p Package), the actual critical was determined to be within 0.00 $,K/K of the predicted critical (see Table 2)..
The dif.ference determined is within the 1% 4K/K criteria of Technical Specification 3.1.2.
The reactivity anomaly calculation for power operation was performed using data from March 1, 1989 with Unit 2 at 97.1% power at a cycle exposure of 279 ?!WD/ST, at equilibrium conditions.
The predicted notch inventory from the vender supplied data was 490 notches.
The actual notch inventory, corrected for. power and flow values which were less than rated, was 421 notches.
Using the notch worth profided by the vendor, the resulting ancmaly was 0.12%
AK/K.
This value is within the 1% nK/K criteris of Technical Specification 3.1.2.
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..i TABLE 2 INITIAL CRITICALITY COMPARISON CALCULATIONS
- Ilgt, e.1K/K
. Keff with all rods'in at 68 F
= 0.95499
- Reactivity inserted by 24 group 1 rods at position 48
= 0.03398
- Reactivity inserted by 16 group 2 rods at position 48
= 0.01287
- Reactivity inserted by ll group 2 rod at position 12
= 0.00027 *
' Predicted Keff at actual' critical rod pattern (68 F)
= 1.00211 Reactivity associated with the measured reactor period (period correction for 187 second period)
= 0.00034
- Reactivity associated with moderator temperature (153*F actual, 68*F predicted)
= -0.0017
- Reactivity Anomaly = [(predicted Keff - 1) - (period-correction) + (temperature correction))
- 100%
= 0.007% oK/K i
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'LaSalle Unit 2 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 scram insertion. times are within the operating limits set fcrth 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 scraa 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
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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:
Positicn Inserted From Average Scram Insertion Fully Withdrawn Time (Seconds)'
45
.0.45 39 0.92.
25 2.05 05 3.70 1
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RESULTS AND DISCUSSION Scram testing was successfully performed between February 10, 1989 and February 11, 1989.
All control rod scram timing acceptance s
criteria were met during this test.
The results of the test are given below.
Maximum Average Average Scram Times Scram Times in a Position of all CRDs (secc.)
Two-by-Two Array (secs.)'
45 0.325 0.355
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39 0.632 0.675 l
25 1.372 1.477 J
05 2.503 2.640 Maximum 90% scram time (position 05): CRD 22-39, 2.752 secs.
1[ ave (position 39) for Minimum Critical Power Ratio Limit determination: 0.632 seconds.
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LTP-1600-17, CORE POWER DISTRIBUTION SYlmETRY ANALYSIS 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 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 Jess-than 25%.
RESULTS AND DISCUFSION Core power, symmetry calculations were performed based upon data obtained from three full core TIP sets (OD-1).: 4'he initial TIP' set
.was performed on February 24, 1989 at 99% power, the second on March 6, 1989 at 87% power, and the third on March 7, 1989 at 87%
power.
The average total TIP uncertainty.from the three dr.ta sets was 3.278%,. satisfying the criteria of the test (less than 8.7%).
The average standard deviation was 3.27%.-
Table 3 lists'the symmetrical TIP pairs, their core locations, and their respective average deviations.
The maximum deviation between symmetrical TIP pairs was 9.03% for TIP pair 5-34, satisfying the criteria of the test (less than 25%).
A discussion of the calculation 1' methodology is provided melow.
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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.w where R13 = the BASE ratio for the ith node of TIP. pair J, n = number of TIP pairs = 19.
l Next, the standard deviation (expressed as.a percentage) of these ratios is calculated by the following equation:
q (%). f f (Ry - R)g-vw a
, soo uf,pl (isn-1)
The total TIP uncertainty (%) is calculated by dividing q (%) by 8
.because the uncertainty in one TIP reading is the desired parameter, but the measured uncertainty is the ratio of two TIP readings.
TABLE 3
'TIP SIGNAL SYMMETRY RESULTS All numbers shown are averages from three OD-1-data sets (from 2-24-89,.3-6-89 gnd 3-7-89 at 99%, 87% and 87% power, respectively).
Symmetrical TIP Pair Absolute Percent Egpbers (Core Location)
Difference-TIP Pair; a
b of BASE #
Deviations 1 (16-09) 6 (08-17)-
1.25 1.43-2 (24-09) 13 (08-25) 2.04 1.96 3-(32-09) 20 (08-33) 3.54 3.31 4 (40-09) 27 (08-41) 4.83 4.77 5 (48-09) 34 (08-49) 5.38 9.03 8 (24-17) 14 (16-25) 5.40
-4.45 9 (32-17)
'21 (16-33)-
0.30 0.26 10 (40-17) 28 (16-41) 1.31 1.08 11 (48-17) 35 (16-49) 0.90 0.89 12 (56-17) 40 (16-57) 1.18 2.15 16 (32-25) 27 (24-33) 8.41 7.37 17 (40-25) 29 (24-41) 3.44 3.03 18 (48-25) 36 (24-49) 5.74-5.12 19 (56-25) 41 (24-57) 2.03 2.46 24 (40-33) 30 (32-41) 0.46 0.~40 25 (48-33) 37 (32-49) 3.95 3.45 26 (56-33) 42 (32-57) 1.84 2.17 32 (48-41) 38 (40-49) 4.18 3.74 33 (56-41) 43 (40-57) 3.84 4.97
- - where : Absolute Difference of IIEI =
EIEI, - BASE b and EAIE. = 7 j(BASE (K)
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where : % Deviation =
IIET. - BIEI6 100
.0.5( K + E H E )g_
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ATTACHMENT C
.1 List of References
- 1. CECO letter, C. M. Allen to Office of Nuclear Reactor Regulation,
'LaSalle County Station Unit 2 Proposed. Amendment to Technical Specification'for Facility Operating License NPF-18, Reload Licensing Package for Cycle 3,* dated September 18, 1988.
- 2. CECO letter, W. E. Morgan to Office of Nuclear Reactor Regulation,
'LaSalle County Station Unit 2 Supplement to Proposed Amendment to Technical Specification for Facility Operating License NPF-18, Reload Licensing Package for Cycle 3,' dated January 6, 1989.
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- 3. NEDE-24011-P-A, ' General Electric Standard Application for Reactor H
Fuel," Revision 9, dated September 1988.
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