02-17-2009 | In July 2007, it was identified that a 1987 calculation, detailing available design margin associated with the Auxiliary Feedwater ( AF) tunnel access covers, did not account for forces from a Main Steam (MS) line High Energy Line Break ( HELB). The calculation scope was limited to loading expected during a station-flooding event. The load from an MS HELB is greater than the load from a station flood. Also, the identified margin for installed concrete expansion anchors was less than the design standard, which requires a Factor of Safety (FOS) greater than or equal to 4.0, yet no actions were taken to recover the desired design margin.
An Operability Evaluation was performed which supported operability of the covers using a FOS of 1.0. However, on December 16, 2008, it was determined that use of a FOS of 1.0 for operability could not be supported and the AF tunnel flood seal opening covers were considered to have been inoperable from original construction until temporary modifications were installed in the summer of 2008 to regain margin.
The cause of this condition was an inadequately designed component during original construction. An investigation was conducted but could not determine why the 1987 calculation did not consider HELB forces nor document the design standard non-compliance into a Corrective Action Program. Completed corrective actions include installation of temporary and permanent modifications to regain margin and restore compliance with design standards, respectively.
There were no actual safety consequences impacting plant or public safety as a result of this event. This event is being reported pursuant to 10 CFR 50.73(a)(2)(i)(B), as a condition prohibited by Technical Specifications 3.6.3 Condition C.
NRC FORM 366 (9-2007) PRINTED ON RECYCLED PAPER |
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LER-2008-001, Technical Specification Non-Compliance Due to Inadequate Design of Auxiliary Feedwater (AF) Tunnel Access Covers Causing AF Valves Within the Tunnel to be InoperableDocket Number |
Event date: |
12-16-2008 |
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Report date: |
02-17-2009 |
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Reporting criterion: |
10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications |
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4562008001R00 - NRC Website |
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A. Plant Operating Conditions Before The Event:
Event Date: December 16, 2008 Event Time: 15:00 Unit 1 and Unit 2 were in Mode 1 — Power Operations Unit 1 Reactor Coolant System (RC) [AB]: Normal operating temperature and pressure Unit 2 Reactor Coolant System (RC) [AB]: Normal operating temperature and pressure
Background:
The Auxiliary Feedwater (AF) [BA] tunnel is located directly below the Main Steam (MS) [SB] tunnel and houses AF piping and the AF Pump Discharge Header to Steam Generator Isolation Valves (i.e., AF013). The AF013 valves are normally open valves that provide an AF flow path to the Steam Generators (SG), but are also considered Containment [NH] isolation valves. Access to the AF tunnels is provided by openings in the floors of the MS Isolation and Safety Valve Rooms. These openings have access covers. There is one AF tunnel on each unit, each with four access covers. Since the AF013 valves are not environmentally qualified, the purpose of the access covers is, in part, to provide environmental protection.
In October 1987, a calculation was performed to evaluate the adequacy of the subject AF tunnel access covers' as-built condition. This calculation only considered flooding loads on the Concrete Expansion Anchors (CEA) used to support the AF tunnel access covers. The purpose of this calculation was to determine the Factor of Safety (FOS) of the Ultimate Strength of the flood covers installed over the AF tunnels. The FOS is the ratio of the ultimate load capacity of the CEA to its anticipated design load. It is noted that the Sargent & Lundy construction design standards utilized for Braidwood station (ref. SDS-E11.0) required a FOS of at least 4.0 for all CEAs.
This calculation identified that some of the CEAs supporting the covers had a FOS less than 2.0, but greater than 1.0. The installed CEAs did not meet the 300% margin (i.e., FOS of 4.0) required by the design standard. No apparent actions were taken at the time to address the low margin identified or for the non-compliance with the design standard of a FOS of 4.0.
Description of Event:
There were no structures, systems, or components inoperable at the beginning of the event that contributed to the severity of the event.
On April 4, 2007, an Engineer at Byron (Braidwood's sister plant), after reviewing the 1987 calculation, recognized a potential discrepancy in that the AF tunnel access covers were designed for flood loads but not High Energy Line Break (HELB) loads. At this time, it was not clear if the cover was required to be designed for HELB loads.
Between April and July 2007, investigations continued into the design basis of the AF tunnel access covers to determine what plant events the covers should be designed to withstand. On July 26, 2007, it was concluded that the AF tunnel covers were required to withstand a station flood and a HELB from the MS Lines above the AF tunnel. The differential pressure exerted on the AF tunnel access covers during the HELBIs higher than that experienced during a station flood. As a result, the FOS for the installed CEAs were actually lower than previously calculated in 1987 and an operability evaluation was performed.
The operability evaluation concluded the AF tunnel access covers were sufficiently designed to support operability of AF013 valves. The FOS's were reduced but still greater than 1.0 (i.e., use of a FOS of 1.0 was deemed to be acceptable for operability). NRC documents providing FOS guidance (Inspection and Enforcement (1E) Bulletin 79-02 and Supplement 1 to IE 79-02) were determined to be not applicable to this design application. The AF013 valves in the tunnel were considered operable and actions were initiated to perform permanent modifications to restore the CEAs to the desired design FOS of 4.0 as specified in the design standard.
In June of 2008, two discrepancies were discovered with the 2007 Operability Evaluation. An NRC Inspector discovered the evaluation did not include a Dynamic Loading Factor (DLF) to account for the energy imparted to the cover as a result of the sudden pressurization of the MS tunnel during a HELB. Additionally, it was self identified that the MS tunnel ventilation blow-out panels were not installed per design. The existing MS HELB analysis was performed with the assumption that panels in the MS Isolation Valve rooms would blow out during the HELB to relieve room pressure. Inspections of the panels led to a conclusion that they would not blow out to relieve pressure as expected. This would result in an increase in room pressure. Therefore, the decision was made to install temporary modifications to modify the blow-out panels in order to provide additional pressure relief paths and to install strongbacks on the AF tunnel access covers to regain some margin pending the installation of the permanent modifications to provide a FOS greater than 4.0. The temporary modifications were installed by July of 2008.
In this timeframe, a complex analysis was begun to assess operability of the AF013 valves from a historical perspective. The conclusion was the access covers would perform their functions and protect the operability of the AF013 valves. This was based on the FOS continuing to be greater than 1.0. This conclusion was also based on an unverified assumption that each access cover had four CEAs. This assumption was to be verified the next time the access covers were removed.
In the Fall 2008 Byron Unit 2 refuel outage, the access covers were removed and it was determined that some of the access covers had only three CEAs. During installation of the permanent modifications, the same configuration was found to exist at Braidwood for a single access cover. A subsequent evaluation recommended not performing a re-analysis given the already low margin condition that existed, assuming the absence of the temporary modifications. This recommendation came after it was determined that use of a FOS of 1.0 to demonstrate operability was not supported. Consequently, on December 16, 2008, it was determined that it was possible that one or more of the AF tunnel covers may have been insufficiently designed to withstand a design basis HELB outside of containment for a time period covering date of original construction until temporary modifications were installed in the summer of 2008. This would have impacted the AF013 valves ability to perform their containment isolation function (i.e., to close), and would therefore render them inoperable during the applicable design basis event (MS Line Break outside of containment).
This potential historical condition is reportable to the NRC in accordance with 10 CFR 50.73(a)(2)(i)(B), as a condition prohibited by Technical Specifications 3.6.3 Condition C.
C. Cause of Event
The cause of this condition was an inadequately designed component during original construction.
An investigation could not determine why the 1987 calculation did not consider HELB forces nor identify the FOS less than 4.0 non-compliance into the Corrective Action Program.
D. Safety Consequences:
There were no actual consequences from the condition since an HELB did not occur while the AF tunnel access covers had minimal structural margin.
To assess the potential consequences of AF tunnel access cover failures, the function of the equipment protected by the access covers (i.e., AF013s) must be considered. From a review of the Updated Final Safety Analysis Report (UFSAR) and Technical Specifications, the AF013s have the following design functions:
1. Containment Isolation valves, 2. Maintain an AF flowpath to the SGs for heat removal, 3. Isolation of AF to the SGs during a SG Tube Rupture, and 4. Isolation of AF for secondary side piping breaks on the SGs.
For the first three functions the AF013s would not be affected because the initiating events are inside containment and do not result in pressurization of the AF tunnel access covers. Therefore, there was no challenge to the ability of the AF013 valves from performing these design functions.
Isolation of AF for secondary side breaks is desired primarily to limit mass and energy releases from the SGs and minimize reactivity effects of excessive RCS cooldown. For secondary side breaks inside containment, the rationale for the first three functions also applies (i.e., a secondary side break in containment will not result in pressurization of the AF tunnel covers).
For secondary side breaks outside containment, continued release of energy could damage safety related components in the MS Isolation and Safety Valve room. However these components have either already performed their safety function (MS Isolation Valve closure, detection of low steam line pressure) or are not needed for accident diagnosis or mitigation in conjunction with a steam line break outside containment (MS line radiation monitors).
Isolation of AF flow is not credited in the UFSAR for steam line breaks. Essentially, the event is terminated when boron reaches the core within the first few minutes of the accident. For Main Feedwater (FW) [SJ] line breaks, the UFSAR does assume isolation of AF flow at 20 minutes after the event; however, the energy release from an FW line break is substantially less than that from an MS line break. Therefore, the AF tunnel covers are designed not to fail and the AF013s will remain available for their isolation function.
Based on the above, the low structural margin of the AF tunnel access covers on the AF013s did not have adverse actual or potential consequences on the AF013 valves' ability to perform their containment isolation valve function.
There were no safety system functional failures as a result of this event.
E. Corrective Actions:
Corrective actions include:
1. Completed installation of temporary modifications in the summer of 2008 to regain some design margin.
2. Completed installation of permanent modifications (final Braidwood modification, for both Units 1 and 2, completed on 1/2/09) to restore compliance with design standards (i.e., FOS greater than or equal to 4.0).
3. Communications of the event to selected personnel.
4. Completed training request for engineering personnel for the lessons learned from this event.
5. Enhancements to the Operational and Technical Decision Making (OTDM) process. Specifically, utilize the OTDM process to document issues like low margin (operational risk exposure) that are beyond the scope of an Operability Evaluation.
6. Review existing station practices associated with recurring review of open Operability Evaluations to identify opportunities to incorporate lessons learned from this event, including confirmation of appropriate prioritization of corrective actions.
F. Previous Occurrences:
There have been no similar Licensee Event Report events at Braidwood Station in the last three years.
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G. Component Failure Data:
Manufacturer Nomenclature Model Mfd. Part Number N/A N/A N/A N/A
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05000413/LER-2008-001 | | 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition 10 CFR 50.73(a)(2)(v)(B), Loss of Safety Function - Remove Residual Heat 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability | 05000334/LER-2008-001 | Control Room Envelope Air Intake During Normal Operation Higher Than Assumed In Design Basis Accident Dose Calculations | 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000315/LER-2008-001 | Unit 1 Manual Reactor Trip due to Main Turbine Vibrations | 10 CFR 50.73(a)(2)(iv)(A), System Actuation | 05000272/LER-2008-001 | Inadvertent Start of an Emergency Diesel Generator During Testing | 10 CFR 50.73(a)(2)(iv)(A), System Actuation | 05000251/LER-2008-001 | Manual Reactor Trip due to High Level in the 4A Steam Generator | 10 CFR 50.73(a)(2)(iv)(A), System Actuation | 05000247/LER-2008-001 | Manual Reactor Trip Due to Decreasing Steam Generator Levels Caused by Loss of Feedwater Flow as a Result of Feedwater Pump Speed Control Malfunction | 10 CFR 50.73(a)(2)(iv)(A), System Actuation 10 CFR 50.73(a)(2)(v), Loss of Safety Function | 05000530/LER-2008-001 | Manual Reactor Trip when Removing a Degraded CEDM MG Set from Service | 10 CFR 50.73(a)(2)(v), Loss of Safety Function | 05000457/LER-2008-001 | 2A Essential Service Water Train Inoperable due to Strainer Fouling from Bryozoa Deposition and Growth | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000454/LER-2008-001 | Technical Specification Non-Compliance of Containment Sump Monitor Due to Improper Installation During Oriainal construction | | 05000440/LER-2008-001 | ' Condition Prohibited by Technical Specifications Due to Unrecognized Reactor Core Isolation Cooling Inoperability | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000483/LER-2008-001 | Containment Cooler Inoperability | 10 CFR 50.73(a)(2)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000412/LER-2008-001 | Unplanned Actuation of the Auxiliary Feedwater System During Plant Startup | 10 CFR 50.73(a)(2)(iv)(A), System Actuation | 05000374/LER-2008-001 | High Pressure Core Spray System Declared Inoperable Due to Failed Room Ventilation Supply Fan | 10 CFR 50.73(a)(2)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | 05000323/LER-2008-001 | Reactor Trip Due to Main Electrical Transformer Failure | 10 CFR 50.73(a)(2)(iv)(A), System Actuation 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000305/LER-2008-001 | Pressurizer PORV and Reactor Coolant System Vent Valves Appendix R Spurious Operation Concern | | 05000271/LER-2008-001 | Crane Travel Limit Stops not Installed as Required by Technical Specifications due to an Inadequate Procedure | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000269/LER-2008-001 | Gas Void Found in High Pressure Injection System Suction Piping | | 05000263/LER-2008-001 | | | 05000261/LER-2008-001 | Appendix R Pathway Impassable due to Lock Configuration | 10 CFR 50.73(a)(2)(v)(A), Loss of Safety Function - Shutdown the Reactor | 05000361/LER-2008-001 | Valid actuation of Emergency Feedwater system following Main Feedwater pump trip | 10 CFR 50.73(a)(2)(iv)(A), System Actuation | 05000335/LER-2008-001 | Unattended Ammunition Discovered Inside Protected Area | | 05000456/LER-2008-001 | Technical Specification Non-Compliance Due to Inadequate Design of Auxiliary Feedwater (AF) Tunnel Access Covers Causing AF Valves Within the Tunnel to be Inoperable | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000423/LER-2008-001 | Postulated Fire Scenario Results in Unanalyzed Condition - Pressurizer Overfill | 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition | 05000369/LER-2008-001 | Potential Failure of Containment Isolation Valves (CIV) to Remain Fully Closed and Ino erable loner than allowed bCTechnical S ecification 3.6.3. | | 05000220/LER-2008-002 | Manual Reactor Scram Due to Loss of Reactor Pressure Control | 10 CFR 50.73(a)(2)(iv)(A), System Actuation | 05000261/LER-2008-002 | Manual Reactor Trip due to High Turbine Vibrations | 10 CFR 50.73(a)(2)(iv)(A), System Actuation | 05000270/LER-2008-002 | Main Steam Relief Valves Exceeded Lift Setpoint Acceptance Band 050002 | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000305/LER-2008-002 | 5 . Blocked Open Steam Exclusion Door Results in Postulated Inoperability of Safety Systems | 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition | 05000336/LER-2008-002 | Unplanned LCO Entry - Three Charging Pumps Aligned for Injection With the Reactor Coolant System Temperature Less than 300 Degrees F. | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000354/LER-2008-002 | BLOWN lE INVERTER MAIN FUSE WITH ONE EMERGENCY DIESEL GENERATOR INOPERABLE CAUSES LOSS OF CONTROL ROOM EMERGENCY FILTRATION LOSS OF SAFETY FUNCTION | | 05000395/LER-2008-002 | Control Room Normal and Emergency Air Handling Systems Inoperable Due to Pressure Boundary Breach | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000382/LER-2008-002 | Inoperable Steam Generator Narrow Range Level Channels | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000369/LER-2008-002 | 'Energy® BRUCE H HAMILTON Vice President McGuire Nuclear Station Duke Energy Corporation MGOIVP 112700 Hagers Ferry Road Huntersville, NC 28078 704-875-5333 704-875-4809 fax bhhamilton@duke-energy.corn August 21, 2008
U.S. Nuclear Regulatory Commission
ATTENTION: Document Control Desk
Washington, D.C. 20555
Subject: Duke Energy Carolinas, LLC
McGuire Nuclear Station, Unit 1
Docket No. 50-369
Licensee Event Report 369/2008-02, Revision 0
Problem Investigation Process No.: M-08-03862
Pursuant to 10 CFR 50.73 Sections (a)(1) and (d), attached
is Licensee Event Report (LER) 369/2008-02, Revision 0,
regarding the Unit 1 Reactor trip on June 26, 2008 due to
the 1B Reactor Coolant Pump Motor trip which was caused by
a failed surge capacitor.
This report is being submitted in accordance with 10 CFR
50.73 (a)(2)(iv)(A). This event is considered to be of no
significance with respect to the health and safety of the
public. There are no regulatory commitments contained in
this LER.
If questions arise regarding this LER, contact Lee A. Hentz
at 704-875-4187.
Very truly yours,
Bruce H. Hamilton
Attachment
www. duke-energy. corn U.S. Nuclear Regulatory Commission
August 21, 2008
Page 2
cc: L. A. Reyes, Regional Administrator
U.S. Nuclear Regulatory Commission, Region II
Sam Nunn Atlanta Federal Center
61 Forsyth Street, SW, Suite 23T85
Atlanta, GA 30303
J. F. Stang, Jr. (Addressee Only)
Senior Project Manager (McGuire)
U.S. Nuclear Regulatory Commission
Mail Stop O-8G9A
Washington, DC 20555
J. B. Brady
Senior Resident Inspector
U.S. Nuclear Regulatory Commission
McGuire Nuclear Station
B. 0. Hall, Section Chief
Radiation Protection Section
1645 Mail Service Center
Raleigh, NC 27699
NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVEDBYOMB:NO.3150-0104 EXPIRES: 08/31/2010 (9-2007) Estimated burden per response to comply with this mandatory collection request: 50 hours. Reported lessons learned are incorporated into the licensing process and fed back to industry. Send ' LICENSEE EVENT REPORT (LER) comments regarding burden estimate to the Records and FOIA/Privacy Service Branch (T-5 F52), U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, or by internet e-mail to infocollects@nrc.gov, and to the Desk Officer, Office of Information and Regulatory Affairs, NEOB-10202, (3150-0104), Office of Management and Budget, Washington, DC 20503. If a (See reverse for required number of means used to impose an information collection does not display a currently valid OMB control digits/characters for each block) number, the NRC may not conduct or sponsor, and a person is not required to respond to,' the information collection. 1, FACILITY NAME 2. DOCKET NUMBER I 3. PAGE 05000- ' 1 8McGuire Nuclear Station, Unit 1 _ 0369' OF .4, TITLE . . Unit 1 Reactor Trip due to the 1B Reactor Coolant Pump Motor Trip which was caused by a
failed Surge Capacitor | 10 CFR 50.73(a)(2)(iv)(A), System Actuation | 05000440/LER-2008-002 | Inoperable Emergency Closed Cooling System Results In Condition Prohibited By Technical Specifications | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000454/LER-2008-002 | Technical Specification Non-Compliance Due to Inadequate Design of Auxiliary Feedwater (AF) Tunnel Access Covers Causing AF Valves Within the Tunnel to be Inoperable | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000298/LER-2008-002 | Technical Specification Prohibited Condition Due to Safety Relief Valve Test Failure | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000237/LER-2008-002 | Unit 2 High Pressure Coolant Injection System Declared Inoperable | 10 CFR 50.73(a)(2)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | 05000272/LER-2008-002 | | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000361/LER-2008-002 | Disturbance on the Pacific DC Intertie cause offsite power frequency to dip below operability limits | 10 CFR 50.73(a)(2)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | 05000362/LER-2008-003 | Missed TS completion time results in TS Violation | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000220/LER-2008-003 | Power Supplies for Drywell Pressure Indication not Qualified for Required Post-Accident Operating Duration | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000529/LER-2008-003 | Technical Specification - Limiting Condition for Operation 3.0.3 for Greater Than 1 Hour | 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000293/LER-2008-003 | | | 05000263/LER-2008-003 | | | 05000423/LER-2008-003 | Automatic Reactor Trip During Shutdown for Refueling Outage 3R12 ., | 10 CFR 50.73(a)(2)(iv)(A), System Actuation | 05000251/LER-2008-003 | . Class 1 Weld Leak Due to Fatigue and Completion of Technical Specification Required Shutdown | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications 10 CFR 50.73(a)(2)(ii)(A), Seriously Degraded 10 CFR 50.73(a)(2)(i)(A), Completion of TS Shutdown | 05000282/LER-2008-003 | Loss of AFW Safety Function and Condition Prohibited by Technical Specifications Due to Mispositioned Isolation Valve | 10 CFR 50.73(a)(2)(v)(B), Loss of Safety Function - Remove Residual Heat 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000286/LER-2008-003 | Automatic Actuation of Emergency Diesel Generator 33 During Surveillance Testing Caused by .Inadvertent Actuation of the Undervoltage Sensing Circuit on 480 Volt AC Safeguards Bus 5A | 10 CFR 50.73(a)(2)(iv)(A), System Actuation 10 CFR 50.73(a)(2)(v), Loss of Safety Function | 05000361/LER-2008-003 | Disturbance on the Pacific DC Intertie cause offsite power frequency to dip below operability limits | 10 CFR 50.73(a)(2)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | 05000305/LER-2008-003 | Door Bottom Seal Failure Results in Inoperability of Control Room Ventilation System | 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition 10 CFR 50.73(a)(2)(v)(B), Loss of Safety Function - Remove Residual Heat |
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