05000456/LER-1997-002-07, :on 970421,Unit 1 SGs Declared Category C-3,per TS Table 4.4.2,during A1R06 Refuel Outage.Caused by Mfg Process of hard-rolling SG Tubes Into Tubesheet.Repairable Indications Identified During Insp Repaired
| ML20141H398 | |
| Person / Time | |
|---|---|
| Site: | Braidwood  |
| Issue date: | 05/21/1997 |
| From: | Alexander L COMMONWEALTH EDISON CO. |
| To: | |
| Shared Package | |
| ML20141H388 | List: |
| References | |
| GL-95-05, GL-95-5, LER-97-002-07, LER-97-2-7, NUDOCS 9705230294 | |
| Download: ML20141H398 (7) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(ii) |
| 4561997002R07 - NRC Website | |
text
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NRC l'ORM 366 U S. NL' CLEAR REGULATORY COMMISSION APPROVED llY OM15 NO. 3150-0164 (4-95)
EXPIRES 04/30NN d
ESTIMATED !!URDEN PER RESPONSE TO COMPl Y % i fil Tills INFORMATION COLLEC110N REQUEST: 30 0 llRS. RI POR11 D LESSONS LEARNED ARE INFORPORATED INTO Tile I ICENSING LICENSEE EVENT REPORT (LER)
PROCESS AND FED llACK TO INDUSTRY FORWARD COMMEN 1S REOARDING llURDEN ESTIMATE TO Ti!E WTORMATION AND RECORDS MANAGEMENT llRANCil(MNBB 7714), U S NUCI E tR REGULATORY COMMISSION, WASillNGTON, DC 20555-0001. AND TO Tile PAPERWORK REDUCTION PROJECT FACILITY NAME (1):
DOCKET NUMilER 05000456 l
PAGE m Braidwood Unit 1 (2) 1 of 7 TITLE (4)
Umt i Steam Generators declared Categorv C.3 per Tech Spec Table 4 4-2 during the Al R06 Refuel Outage EVENT DATE(5)
I ER NUMilER (6) l REPORT DATE(7) l OTiiER FACll.ITIES IN) OI A FD (N)
MUMlH DAY YEAR YEAR bEQUENHAL RE VISION hK W rH DAY
) EAR FACILITY NAME IX)CKl:1 NUMilli NUMBER NUMBER 4
21 97 97 02 00 5
21 9 'l FACILITY NAME DOCKEl NUMlllR oggy Defueled Tills REPORT IS SUBMITTED PURSUANT TO Tile REQUIREMENTS OF 10 CFR g: (Check oHe or more)
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NRC rorm 366M
'x I,1CENNEE CONTArt i OR Tills i ER (12)
NAME TELErilONE NUMBER (include Area Cmie)
L. Alexander, Site Engineering (815) 458-2801 Extension 2251 COMPI.El E ONE I INE FOR EACII COMPONENT Fall.t'RE DENCRIllED IN Till% REPORT (13)
CAUSE
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DATE (IM j
ABSTRACT (12mit to 1400 spaces, a c.. approximately filteen smgle-space typewntien luws 16)
J l
A Steam Generator (SG) Tube Inservice Inspection performed on Braidwood Unit 1
)
(Westinghouse Model D-4 Steam Generators) revealed SGs lA, 1B, 1C, and ID as Category C-3 with greater than one percent of the inservice tubes being defective.
Per Technical Specification 4.4.5.2, a SG will be classified in Category C-3 if more than 10% of the total tubes inspected are degraded or more than lv of the total tubes inspected are defective. The primary cause of this event is circumferential Outside Diameter Stress Corrosion Cracking (ODSCC) at the top-of-tubesheet roll transition region and axial ODSCC at the tube-to-Tube support Plate intersections.
Corrective actions include plugging or sleeving the repairable indications, the continuation of methoxypropylamine injection with review of other alternate amines, and the continuation of the elevated hydrazine program.
Two tubes were removed from the 1A Steam Generator to evaluate the axial indications at the tube-to-Tube Support Plate as is required by Generic Letter 95-05.
Sixteen tubes, with the largest indications, were insitu pressure and leak tested which verified structural and leakage integrity of the defective tubes.
Circumferential indications were also identified at the top of the tubesheet in SG tubes that were not inservice.
These tubes were unplugged to support inspections of the locks installed during AlR05.
Forty-nine of the eighty-five SG tubes unplugged contained circumferential indications at the top of the tubesheet.
A Westinghouse Laser Welded Sleeve was installed in 1004 of the Locked tubes.
Testing was performed that demonstrated that these tubes would have performed their function during an accident.
This is the fifth Braidwood Unit 1 inspection with SGs being classified as Category C-l 3.
The SGs have not exceeded their design plugging levels.
.-.- --.....-..--.__ _ ~.~_~~--._.-
,. -..,. - ~ - - ~ ~ - -,. -
I NetC FORM 366A U.S. NUCLEAR REGULATORY CUMMISSION APPROVED BY OMH NO.3150 0104
- (4-95)
EX PAREN 04/30/9N ESTIMATED DURDEN PER RESPONSE TO COMPLY WITil Tills INFORMATION COLI.FCTION REQUEST: 50 0 llRS. REPORTED a
LICENSEE EVENT REPORT (LER)
LESSONS LEARNED ARE INCORPORATED INTO Tile IJCENSINO PROCESS AND FED BACK TO INDUSTRY. FORWARD COMMLNTS TEXT CONTINUATION REGARDING BURDEN ESTIMATE TO Tile INFORMATION AND i
RECORDS MANAGEMENT BRANCil(T-6 F33), U.S. NUCl. EAR REGULATORY COMMISSION, WASil!NGTON. DC 205$5 000E AND TO TIIE PAPERWORK REDUCTION PROJECT FACILITY NAME(1)
DOCKET NUMBER (2)
LER NUMBER (6)
PAGE(3)
)LAR hLyULNTIAl, REVislON NUMBF R NUMf4FR j
Braidwood Unit 1 05000456 97-002 00 2 of 7 (If more space is required, use additional copies of NRC Form 366A)(17) l A.
PLANT CONDITIONS PRIOR TO EVENT:
Unit (s): 1 Event Date: 4/21/97 Event Time:
0715 Hours l
Reactor Mode (s):Defueled Power Level (s): 0%
RCS [AB) Temp./ Press.
73"E / 0 B.
DESCRIPTION OF EVENT
l There were no systems or components inoperable at the beginning of this event j
that contributed to the severity of the event.
On April 1, 1997, a Steam Generator (SG) Tube Inservice Inspection o initiated on Braidwood Unit 1 in accordance with Technical Specification Surveillance Requirement (TSSR) 4.4.5.0.
Technical Specification Surveillance Requireme.,t 4.4.5.2 requires that the results of each sample inspection be classified inte l
one of three categories. A SG will be classified in Category C-3 if more than
)
10% of the total tubes inspected are c% graded or more than 1% of the inspected tubes are defective. A SG tube is considered degraded if it has an imperfection of greater than or equal to 20% of the nominal tube wall thickness. A SG tube is j
considered defective if it has an imperfection of greater than or equal to 40? of the nominal tube wall thickness or exceeds the 3.0 Volt Interim Plugging Criteria.
i An initial sample size of 100% of all available tubes was selected. The eddy i
current data was analyzed using the EddyNet 95 software. The initial sample i
inspection resulted in the following SGs being classified into Category C-3 based j
on the following reasons:
1A On April 23, 1997, at 07:30 hours, greater than 14 of the 3986
+
inservice tubes being defective.
IB On April 21, 1997, at 07:15 hours, greater than 19 of the 4154 i
inservice tubes being defective, i
IC On April 27, 1997, at 22:30 hours, greater than 14 of the 3662 inservice tubes being defective.
1D On April 23, 1997, at 21:00 hours, greater than 14 of the 4153 inservice tubes being defective.
Notification per Technical Specification 4.4.5.2 pursuant to 10 CFR 50.72 (b) (2) (1) for steam generators being classified in Category C-3 was initiated within four hours of the affected SGs being evaluated as C-3.
Table 1 provides, by SG, the number of tubes plugged during this outage, the number of tubes sleeved during this outage as well as the number of tubes plugged and sleeved in previous outages.
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l 1
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(4 95)
U.S. NUCLEAR RFEULATORY LUMMISSION APPROVED IlY Omit h0. 3150 0104EXPIRES O4/30/98 ESTIMATED BURDEN PER RESPONSE TO COMPLY WITil Tills INFORMATION COLLECTION REQUEST: 50.0 llRS REIORTED LESSONS LEARNED ARE INCORPORATED INTO Tile LICENSING LICENSEE EVENT REPORT (LER)
PROCESS AND FED BACK TO INDUSTRY. FORWARD COMMENTS TEXT CONTINUATION REGARDING BURDEN ESTIMATE TO Tile INFORMATION AND RECORDS MANAGEMENT BRANCll(T 6 F33), U S. NUCLEAR REGULATORY COMMISSION, WASil!NGTON, DC 20555-0001, AND TO Tile PAPERWORK REDUCTION PROJECT FACILITY NAME(1)
DOCKET NUMilER (2)
LER NUMilER (6)
PAG E (3)
W:
=
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Braidwood Unit 1 05000456 97 002 00 3 of 7 (If more space is required, use additional copics of NRC Form 366A)(17)
B.
DESCRIPTION OF EVENT (continued):
Table 1 Braidwood Unit 1 Sixth Refuel Outage (A1R06) SG Tube Plugging Results 1 A SG 1BSG 1C SG 1D SG TOTALS Total Tubes 4578 4578 4578 4578 18312 Previously Plugged Tubes 592 424 916 42,5 2357 Previously Sleeved Tubes (Hot Leg Roll Transition) 181 0
445 271 897 Total inservice Tubes Inspected 3986 4154 3662 4153 15955 Total Yubes inspected at Hot Leg Roll Transition 3805 4154 3217 3882 15058 Locked-Tube Model TSP Pluggable (IPC) 27 7
39 31 104 Free-Span Model TSP Pluggable (IPC) 2 0
0 0
2 TSP Pluggable Indications Excluded from IPC 14 2
7 9
32 Anti-Vibration Bar Wear Pluggable 0
0 0
0 0
New Tubes Plugged due to TSP Locking 0
0 3
1 4
Cold Leg Top-of-Tubesheet Indications Plugged Single Circumferential (SCI) 8 6
4 4
22 Multiple Circumferential (MCl) 5 0
5 1
11 Volumetric (VOL) 1 1
0 0
2 U-Bend AxialIndications 2
0 0
0 2
U-Bend Circumferential Indications 0
1 0
0 1
Volumetric Indications at TSP 1
2 0
0 3
Admin - LWS Weld Indication 0
0 1
0 1
Hot Leg Top-of-Tubesheet Indications Plugged Single Circumferential (SCI) 5 4
5 3
17 Multiple Circumferential (MCl) 0 1
0 0
1 l
Hot Leg Top-of-Tubesheet Indications Sleeved Single Circumferential (SCI) 49 33 28 102 212 Multiple Circumferential (MCl) 8 1
13 27 49 Single Axial (SAI) 0 1
0 1
2 Multiple Axial (MAI) 2 0
0 1
3 Mixed Mode Indication (MMI) 0 1
0 0
1 Volumetric (VOL) 0 2
0 0
2 Locked Tubes Unplugged / Sleeved /Replugged 21 21 22 21 85 Tubes Unplugged /Replugged 1 CL only 2
4 0
7 A1R06 Total Tubes Plugged 65 24 64 49 202 A1R06 Total Tubes Sleeved 59 38 41 131 269 Previously Sleeved Tubes Now Plugged 4
0 9
7 20 Restart Total Tubes Available 3921 4130 3598 4104 15753 Total Tubes Plugged 657 448 980 474 2559 Total inservice Tubes Sleeved,
236 38 477 395 1146 Total Equivalent Plugged 670.7 450.2 1007.7 497.0 2625.6 Percentage of Tubes Plugged 14.7 %
9.8%
22.0 %
10.9%
14.3 %U.S. NUCLEAR REGULATORY COMMISSION APPROVED HY OMH NO 3150-0104 (4-95)
'a EXPIRES O4/30/98 8
ESTIMATED BURDEN PER RESPONSE TO COMPLY WTI'll Tills INFORMATION COLLECTION REQUEST: 50.0 ilRS. REPORTl;D LESSONS LEARNED ARE INCORPORATED INTO Tile LICENSING LICENSEE EVENT REPORT (LER) i PROCESS AND FED BACK TOINDUSTRY. FORWARD COhlMIXl A TEXT CONTINUATION REGARDING BURDEN ESTIMATE TO Tile INFORMATION AND RECORDS MANAGEMENT BRANCil(T 6 F33). U.S. NUCLEAR REGULATORY COMMISSION. WASilINGTON, DC 20555-000l. AND TO Tile PAPERWORK REDUCTION PROJECT FACILITY NAME(1)
DOCKET NUMBER (2)
LER NUMBER (6)
PAGE(h RAK btQUE NUAL kE VISION NUMBFR NUMM R Braidwood Unit 1 05000456 97 002 00 4 of 7 (If snore space is required, use additional copies of NRC Form 366A)(17) l Circumferential indications were also identified at the top of the tubesheet in SG tubes that were not inservice.
These tubes were unplugged to support l
inspections of the locks installed during AlR05.
One hundred percent of the Locked Tubes were inspected at the locked intersection and at the top of the i-l tubesheet roll transition region.
Forty-nine of the eighty-five SG tubes i
i unplugged contained circumferential indications at the top of the tubesheet. No j
indications were identified at the locked intersection.
Insitu testing was conducted during the AlR06 outage on a sample of the largest indications. This testing simulated an axial load on the defective indication.
This load was greater than that which would have been seen by these Locked SG tubes during an l
accident The Locked SG tube was still able to withstand the accident axial load, even with the circumferential indication in the tube, demonstrating that the Locked SG tube would have performed as designed during an accident. A i
Westinghouse Laser Welded Sleeve was installed in 100% of the Locked SG tubes to ensure the tube will perform as designed during an accident for the upcoming l
Cycle 7.
All of the Locked Tubes were remo"ed from service by a plug.
This event is being reported pursuant to 10cFR50.73 (a) (2) (ii), which requires a 30-day written report.
l t
C.
CAUSE OF EVENT
During the Braidwood Unit 1 sixth refuel outage SG Tube inservice inspection, axial and circumferential indications were identified at the SG top-of-tubesheet roll transition region. Axial indications were also detected within the tube-to-Tube Support Plate intersection. These indications resulted in 202 tubes being t
removed from service by plugging and 269 tubes being sleeved using the l
Westinghouse Laser Welded Sleeve. The top-of-tubesheet indications and indications at the Tube Support Plates are a result of Outside Diameter Stress l
Corrosion Cracking (ODSCC).
The cause of the top-of-tubesheet indications is related to the manufacturing process of hard-rolling the SG tubes into the tubesheet. This process caused residual stresses in the SG tubes, thereby l
creating an envi ronment for stress corrosion cracking.
The cause of the i
indications within the Tube Support Plates is the environment (temperature, stress) within the tube-to-Tube Support Plate crevice. Axial and circumferential indications at the SG top-of-tubesheet roll transition region along with axial indications within the tube-to-Tube Support Plate intersection are typical l
degradation mechanisms observed in this model of Steam Generators.
The root cause of the Locked tubes forming circumferential indications at the top of the tubesheet is documented in a letter to the NRC, Additional Information Pertaining to Braidwood Unit 1 Locked Tube Circumferential Indications, dated April 30, 1997.
This report concluded that the process which installed the Locks at the Tube Support Plates caused additional residual stress at the top of the tubesheet roll transition region. This increased residual stress is sufficient j
to increase the stresses at the roll transition region to the point that ODSCC would be expected to occur at a faster rate than an inservice tube.
This
{
increased stress is sufficient to overcome the anticipated reduced rate of ODSCC as a result of the tube being taken out of service.
This report also concluded that the integrity of these Locked tubes was maintained throughout Braidwood Unit 1 Cycle 6.
i y
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- - - --U.S. NUCLEAR REGULATORY CON 1 MISSION APPROVED IW ON1H NO,3140101 (4-95)
EXPIRES 04/30/9N i
ESTIMATED BURDEN PER RESPONSE TO COMPLY WITil Til!$
INFORMATION COLLECTION REQUEST: 50 0 IIRS. REPORTED l
OCENSEE EVENT REPORT (LER)
LESSONS LEARNED ARE INCORPORATED INTO Tile LICENSING l-PROCESS AND FED BACK TO INDUSTRY. FORWARD COMMEN'ls TEXT CONTINUATION REGARDING BURDEN ESTIMATE TO TIIE INFORMATION AND RECORDS MANAGEMENT BR ANCll(T 6 F33), U.S. NUCLIMR REOULATORY COMMISSION, WASHINGTON, DC 20$55-0001, AND TO
+
Tile PAPERWORK REDUCTION PROJECT FACILITY NAME(1)
DOCKET NUMBER (2)
LER NUMBER (6)
PAG E (3)
RAR SEQUEKilAL REVISKJN i
NIWMR NIMMR Braidwood Unit 1 05000456 97 002 00 5 of 7 (If more space is required, use additional copies of NRC Form 366A)(17)
D.
ASSESt4 TENT OF SAFETY CONSEQUENCES:
Sixteen tubes, with the largest indications, were insitu pressure and leak I
tested. None of the tubes burst at tested pressure demonstrating the structural integrity of the SGs prior to this inspection.
None of the SG tubes leaked at tested pressure.
The maximum calculated leakage, calculated per the requirements of Generic Letter i
The accident leakage, using the actual indications identified during the AIR 06 outage, is estimated to be approximately 12 gpm.
j Although this is higher than the predicted leakage from the distribution of indications from the AlROS outage (6.99 gpm), it is within the site allowable leakage of 26.8 gpm.
Therefore, the accident dose rate is below a small l
fraction of the 10CFR100 limits.
l A safety assessment of having circumferential indications in the Locked tubes during Cycle 6 is documented in an April 30, 1997 letter to the NRC.
Insitu j
testing was performed on a sample of the largest indications in the Locked tubes.
l This testing demonstrated that the Locked tubes would have performed their l
function during an accident even with the circumferential indications at the top I
of the tubesheet.
The Insitu testing applied a conservative axial load to the circumferential indication, and the Locked tube remained intact.
With the additional 202 tubes plugged during this inspection and accounting for the 17.211 sleeves to plug ratio, Braidwocd Unit I has 14.3e of the total SG tubes plugged with the maximum single loop 22.0% plugged for the 1C SG loop.
These plugging levels are within the acceptable range (maximum of 24% of the total SG tubes plugged with an single loop maximum of 30s) to ensure the RCS total flowrate in Technical Specification 3.2.3 is achieved.
(
E.
CORRECTIVE ACTIONS
All repairable indications identified during this inspection have been repaired
[
using the Westinghouse Laser Welded Sleeve or have been removed from service by l
plugging.
Stabilizers were installed in the tubes that were plugged due to l
circumferential indications at the top-of-tubesheet so that further degradation of the tubes can not result in a double ended break.
The Westinghouse Laser l
Welded Sleeving process and the mechanical SG tube plugging process used at Braidwood are approved methods of SG tube repair per the Braidwood Technical l
Specifications.
In addition to the sleeving and plugging, sixteen tubes, with the largest indications, were insitu pressure and leak tested. None of the SG tubes burst during insitu testing, therefore, the tests proved the structural integrity of the tubes. None of the SG tubes leaked during insitu testing.
Two tubes were removed from the 1A SG per the requirements of Generic letter 95-05.
Testing will be conducted on these tubes to support the IPC Tube Pull Database.
Information required by Generic Letter 95-05 will be included in the required 90 Day Report.
I Westinghouse Laser Welded Sleeves were installed in all of the Locked Tubes to ensure the integrity of the top-of-tubesheet roll transition region.
Since this
. sleeve is being used for a purpose other than what it was designed, the qualification documentation of this sleeve was reviewed to verify the sleeve can withstand the axial load that the SG tube will see during an accident.
A review
...U.S. NUCl. EAR REGULATORY COMMISSION APPROVED llY OMil No. 3150-0104 EN PIRES O4/30%
(4-95) 0 ESTIMATED llCRDEN PER RESPONSE TO COMPLY WITil1111s INFORMATION COLLECTION REQUEST: 50.0 liRS REPORTI D L'ICENSEE EVENT REPORT (LER)
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TIIE PAPFRWORK REDUCTION PROJECT FACILITY NAME(1)
DOCKET NUMllER (2) 1.ER NUMllER (6)
PAG E (3) l 1 EAR htQUI N'H AL REVlhP A M ikHO R NUMBf R Braidwood Unit 1 05000456 97 002 00 6 of 7 (If more space is required, use additional copies of NRC Form 366A)(17) 1 of the qualification was sent to the NRC in a letter from John B.
- Hosmer, Information Related to the Braidwood 1 AIR 06 Steam Generator Inspection and the 3 Volt Interim Plugging Criteria Renewal Technical Specificution Amendment, dated April 29,1997.
The Braidwood Unit 1 Steam Generators will be replaced during the A1R07 refueling outage (Fall of 1998).
The new Steam Generators will include a different tube material (Inconel 690) and the tubes will be hydraulically expanded into the tubesheet instead of hard-rolled.
The Inconel 690, along with the hydraulic expansion of the tubes, will reduce the potential for stress corrosion cracking at the roll transition region.
The new Steam Generators will not have the same Tube Support Plate design.
The new Tube Support Plate design should reduce the ability to create an environment in this region that is detrimental to the Stean Generator tubing.
Comed will continue its efforts with the industry to understand the root cause of the tube degradation and take appropriate correct actions to mitigate future degradation.
Braidwood Station implemented the following programs to mitigate the corrosive environment in the SGs which lead to ODSCC:
- - Use of advanced amines, such as metroxypropylamine (MPA), for secondary pil control to reduce the amount of co,rosion products which enter the SG.
- - Compliance with the EPRI Secondary Chemistry Guidelines.
- - Maintain hotwell dissolved oxygen concentrations less than 3 ppb.
- - Use of high hydrazine concentrations for maintaining reducing conditions in the SGs and passivation of piping systems and components.
Analyses of samples of sludge from previous outages determined that Bzaidwood Unit 1 does not have significant levels of lead or copper in the sludge.
Lead and copper have been identified as contributors to the formation of ODSCC.
F.
PREVIOUS OCCURRENCES
Occurrences of ODSCC indications have been seen at Braidwood Unit 1 in previous
- outages, as documented by LER 94-007-00, LER 95-003, LER 95-015 and LER 96-012 for the AIR 04, A1MOS, A1R05, and A1P02 outages, respectively.
The number of tubes showing indications of ODSCC has been increasing each outage.
This is the fifth inspection on Braidwood Unit 1 that resulted in being classified into Category C-3.
A search of previous LERs identified only the previous Unit 1 outages as previous occurrences of SG tube degradation.
LER NUMBER TITLE 94-007-00 Steam Generator Tubes Degraded due to Axial Outer Diameter Stress Corrosion Cracking.
95-003-00
A Steam Generator Inservice Inspection revealed SGs 1A, 1C, and ID as Category C-3.
95-015-00 Degradation of Steam Generator Tubes Exceeds Technical Specification Limit.
_ _ _ _ - _ _ _ - - _ _ - _ _. _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _,U.S. NUCLEAR REGUIATORY COMMISSION APPROVED BY OslH NO. 3150-0104 EXPIRES 04/30/9N (4-95)
ESTIMATED DURDEN PER RESPONSE TO COMPLY WITil TIIls INFORMATION COLLECTION REQUEST: 50 0 llRS. REPORTED LESSONS LEARNED ARE INCORPORATED INTO TliF UCENNIMj LICENSEE EVENT REPORT (LER)
PROCESS AND FED UACK TO INDUSTRY FORWA% COW !! VI S TEXT CONTINUATION REGARDING BURDEN EFTIMATE TO Tile INFORMATION - 9 RECORDS MANAGEMENT DRANCll(T-6 F33) U.S. Nt1CILU REGULATORY COMMISSION, WASlilNOTON. DC 20$$$-0u01. \\ND TO TIIE PAPERWORK REDLICTION PROJECT FACILITY NAME(1)
IXX?KFT NUMBEP (2)
LER NUMilER (6)
PAGE O) iLAR
$hyULNTIAL REVmlON mRIMF4FR NtfMRFR Braidwood Unit 1 05000456 97 002 00 7 of 7 l
(If more space is required, use additional copies of NRC Form 366A)(17)
Unit 1 Steam Generators declared Category C-3 per Tech. Spec.
96-012-00 4.4.5.2.e during the A1P02 Midcycle Outage G.
CCNPORMNT FAILURE DATA:
l MANUFACTURER - ---NOMENCLATURF MODEL MFG. PART NO.
Since no component failure occurred, this section is not applicable.
2