On October 12, 2006, at 2236 hours0.0259 days <br />0.621 hours <br />0.0037 weeks <br />8.50798e-4 months <br />, the Unit 1 Standby Liquid Control ( SLC) system was declared inoperable due to a through-wall leak in the SLC tank. Although the SLC tank is exposed to atmospheric pressure, it is an ASME Code Section XI Class 2 boundary. The SLC tank leak was originally identified on May 27, 2004, but the ASME Code Class of the SLC tank was not identified, and the SLC tank was incorrectly determined to be operable at that time.
On October 13, 2006, the NRC approved a Notice of Enforcement Discretion to extend the Allowed Outage Time for the inoperable SLC system. On October 15, 2006, following draining, repair, refilling and testing of the tank, the SLC system was declared operable.
The SLC tank leak was due to the use of grout material containing leachable halogens during original installation of the SLC tank supports, resulting in an environment that supported stress corrosion cracking in the stainless steel tank where it was in contact with the grout.
The safety significance of the event was minimal. The leak was extremely small and did not affect the capability of the SLC tank to perform its safety function.
Corrective actions include development of issue review guidance, redesign of the SLC tank supports, and review of systems for the use of grout. |
FACILITY NAME (1) DOCKET NUMBER (2) LER NUMBER (6) PAGE (3) Quad Cities Nuclear Power Station Unit 1 05000254 NUMBER NUMBER (If more space is required, use additional copies of NRC Form 366A)(17)
PLANT ANDSYSTEM IDENTIFICATION
General Electric - Boiling Water Reactor, 2957 Megawatts Thermal Rated Core Power Energy Industry Identification System (EIIS) codes are identified in the text as [XX].
EVENT IDENTIFICATION
Through-wall Leak in Standby Liquid Control Tank Due to the Original Construction Use of Grout with Leachable Halogens
CONDITION PRIOR TO EVENT
Event Time: 2236 hours0.0259 days <br />0.621 hours <br />0.0037 weeks <br />8.50798e-4 months <br /> Unit: 1 Event Date: October 12, 2006 Reactor Mode: 1 Mode Name: Power Operation Power Level: 100% B.D DESCRIPTION OF EVENT Standby Liquid Control (SLC) [BR] system inoperable due to a through-wall leak in the SLC tank [TK]. Although the SLC tank is exposed to atmospheric pressure, it is an ASME Code Section XI Class 2 boundary. Quad Cities. Unit 1 entered Technical Specification (TS) Limiting Condition for Operation (LCO) Section 3.1.7, Condition B, for both SLC subsystems and declared the SLC system inoperable. This action was in response to Engineering identifying the ASME Code Class of the SLC tank.
The Unit 1 SLC tank through-wall leak was originally identified on May 27, 2004, documented in an Issue Report by the SLC System Engineer, and reviewed by System Engineering supervision, Design Engineering Mechanical supervision, and the Operations Shift Manager on May 28, 2004. The ASME Code Class of the SLC tank was not identified during the origination or review of the Issue Report. Consequently, the Unit 1 SLC tank was incorrectly determined to be operable by the Operations Shift Manager on May 28, 2004. The Issue Report generated on October 12, 2006, was generated as a result of discussions with a SLC system engineer from another plant concerning repair options for the through-wall leak and the information that the SLC tank was a Code boundary.
In parallel with maintenance activities, including further analysis of the through wall leak, Quad Cities Station requested a Notice of Enforcement Discretion (NOED) on October 13, 2006. The NOED would extend the time that Unit 1 could be operated with the SLC system inoperable for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. At 0636 hours0.00736 days <br />0.177 hours <br />0.00105 weeks <br />2.41998e-4 months <br /> on October 13, 2006, Quad Cities Unit 1 entered TS LCO Section 3.1.7, Condition C, which requires the unit to be in hot shutdown in 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. At 1034 hours0.012 days <br />0.287 hours <br />0.00171 weeks <br />3.93437e-4 months <br />, power was decreased on Quad Cities Unit 1 in preparation for shutting down the unit. At 1046 hours0.0121 days <br />0.291 hours <br />0.00173 weeks <br />3.98003e-4 months <br />, an ENS notification was made in accordance with 10 CFR 50.72(b)(2)(i) due to initiation of a reactor shutdown required by Technical Specifications. At 1138 hours0.0132 days <br />0.316 hours <br />0.00188 weeks <br />4.33009e-4 months <br />, the NRC approved the NOED request, and Unit 1 was returned to full power. At 2312 hours0.0268 days <br />0.642 hours <br />0.00382 weeks <br />8.79716e-4 months <br /> on FACILITY NAME (1) DOCKET NUMBER (2) LER NUMBER (6) PAGE (3) Quad Cities Nuclear Power Station Unit 1 05000254 NUMBER NUMBER 2006 � (If more space is required, use additional copies of NRC Form 366A)(17) October 13, 2006, the station began draining the SLC tank to facilitate repair of the through-wall leak. At 0204 hours0.00236 days <br />0.0567 hours <br />3.373016e-4 weeks <br />7.7622e-5 months <br /> on October 14, 2006, the SLC tank was drained.
At 0413 hours0.00478 days <br />0.115 hours <br />6.828704e-4 weeks <br />1.571465e-4 months <br /> on October 15, 2006, the SLC tank was refilled, and at 1122 hours0.013 days <br />0.312 hours <br />0.00186 weeks <br />4.26921e-4 months <br />, following successful testing, the SLC system was declared operable.
C. CAUSE OF EVENT
The programmatic root cause for the incorrect operability determination of the Unit 1 SLC Tank was an incomplete application of technical rigor resulting in incorrect assumptions regarding the ASME Code applicability to the Ul SLC tank.
These incorrect assumptions were not adequately challenged during the condition identification and review process.
The technical root cause of the Unit 1 SLC tank leak is the use of grout material containing leachable halogens during original installation of SLC tank supports during original plant construction. When the grout is wetted, Stress Corrosion Cracking can develop at the grout/tank interface. In this case, the source of moisture was occasional condensation from above when the SLC tank lid was removed.
D.� SAFETY ANALYSIS The safety significance of this event was minimal. During the time frame from development of the through-wall leak in 2004 until draining of the SLC tank on October 13, 2006, the SLC tank was fully functional and capable of providing the required boron solution during an accident. Although the leak was through-wall, the leakage was extremely small and was identifiable only by the formation over time of boron crystals. There was no effect on the level of the tank or the concentration of the boron solution.
During the 29 hours3.356481e-4 days <br />0.00806 hours <br />4.794974e-5 weeks <br />1.10345e-5 months <br /> that the SLC tank was drained, all control rods were operable, and no performance issues existed that could impact the scram function of any individual control rod. The Alternate. Rod Insertion (ARI) system was available as a separate means for reactor shutdown in the event that the normal scram path could not be initiated by the Reactor Protection System (RPS). The ARI system is diverse and independent from RPS.
Additionally, as compensatory actions during the NOED timeframe, both Automatic Transient Without Scram Recirculation Pump Trip Systems and the RPS were protected, mitigating the need for SLC; and production risk activities were prohibited, minimizing the likelihood of initiation events (i.e., plant transients). Also, Boric Acid was staged for alternate injection if required during the repair.
This LER is being submitted in accordance with 10 CFR 50.73(a)(2)(i)(B), which requires reporting of any operation or condition which was prohibited by the plant's Technical specifications, and Part 50.73(a)(2)(vii), which requires the reporting of any event where a single condition caused two independent trains to become inoperable in a single system designed to shut down the reactor and maintain in a safe shutdown condition.
FACILITY NAME (1) DOCKET NUMBER (2) LER NUMBER (6) PAGE (3) Quad Cities Nuclear Power Station Unit 1 05000254 NUMBER NUMBER � (If more space is required, use additional copies of NRC Form 366A)(17)
E. CORRECTIVE ACTIONS
Guidance will be developed and implemented concerning the initial review process for Issue Reports with potential system operability challenges.
The existing Unit 1 and Unit 2 SLC tank supports will be redesigned to remove the Stress Corrosion Cracking environment.
A review of systems will be made to identify instances of stainless steel in contact with grout.
F. PREVIOUS OCCURRENCES
No previous incidents involving through-wall leakage on a SLC system Code boundary were identified. There were 28 instances of SLC system leaks documented since October 2004, but they were from mechanical connections, and did not cause the system to be inoperable.
A Finding involving inadequate documentation for the basis of operability determinations performed by the Operation Shift Managers was identified in NRC Inspection Report 254(265)/2006003. Corrective actions included presenting the issue as a lessons-learned to the Shift Managers. Because this occurred after the 2004 determination of operability for the SLC tank, it did not affect this event.
No instances of an incorrect prompt operability determination were identified.
G. COMPONENT FAILURE DATA
The SLC tank is a 5,000 gallon non-insulated tank manufactured by Bethlehem Steel from Type 304 stainless steel and vented to atmosphere.
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05000305/LER-2006-010 | | | 05000456/LER-2006-001 | Unit 1 Reactor Coolant System Pressure Boundary Leakage Due To Inter-Granular Stress Corrosion Cracking of a Pressurizer Heater Sleeve | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications 10 CFR 50.73(a)(2)(ii)(A), Seriously Degraded | 05000454/LER-2006-001 | Technical Specification Required Action Completion Time Exceeded for Inoperable Containment Isolation Valves Due to Untimely Operability Determination | | 05000423/LER-2006-001 | Loss Of Safety Function Of The Control Room Emergency Ventilation System | 10 CFR 50.73(a)(2)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident | 05000369/LER-2006-001 | Ice Condenser and Floor Cooling System Containment Isolation Valve inoperable longer than allowed by Technical Specification 3.6.3. | | 05000353/LER-2006-001 | HPCI Ramp Generator Signal Converter Failure | 10 CFR 50.73(a)(2)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident 10 CFR 50.73(a)(2)(v), Loss of Safety Function | 05000352/LER-2006-001 | Loss Of One Offsite Circuit Due To Invalid Actuation Of Fire Suppression System | 10 CFR 50.73(a)(2)(iv)(A), System Actuation | 05000336/LER-2006-001 | | 10 CFR 50.73(a)(2)(v)(A), Loss of Safety Function - Shutdown the Reactor | 05000316/LER-2006-001 | Failure to Comply with Technical Specification 3.6.2, Containment Air Locks | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000315/LER-2006-001 | Plant Shutdown Required by Technical Specification Action 3.6.5.B.1 | | 05000293/LER-2006-001 | | 10 CFR 50.73(a)(2)(iv), System Actuation | 05000289/LER-2006-001 | | | 05000287/LER-2006-001 | Actuation of Emergency Generator due to Spurious Transformer Lockout | 10 CFR 50.73(a)(2)(iv)(A), System Actuation | 05000251/LER-2006-001 | Turkey Point Unit 4 05000251 1 OF 6 | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000247/LER-2006-001 | Manual Reactor Trip Due to Multiple Dropped Control Rods Caused by Loss of Control Rod Power Due to Personnel Error | 10 CFR 50.73(a)(2)(iv)(A), System Actuation | 05000440/LER-2006-001 | Incorrect Wiring in the Remote Shutdown Panel Results in a Fire Protection Program Violation | | 05000413/LER-2006-001 | | 10 CFR 50.73(a)(2)(iv)(A), System Actuation | 05000368/LER-2006-001 | Completion of a Plant Shutdown Required by Technical Specifications Due to Loss of Motive Power to Certain Containment Isolation Valves as a Result of a Phase to Ground Short Circuit in a Motor Control Cubicle | 10 CFR 50.73(a)(2)(i)(A), Completion of TS Shutdown | 05000306/LER-2006-001 | | 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(i)(A), Completion of TS Shutdown | 05000298/LER-2006-001 | Cooper Nuclear Station 05000298 1 of 4 | 10 CFR 50.73(a)(2)(iv)(A), System Actuation | 05000286/LER-2006-001 | I | 10 CFR 50.73(a)(2)(iv)(A), System Actuation | 05000282/LER-2006-001 | | 10 CFR 50.73(a)(2)(iv)(A), System Actuation 10 CFR 50.73(a)(2)(v), Loss of Safety Function | 05000266/LER-2006-001 | | 10 CFR 50.73(a)(2)(v), Loss of Safety Function | 05000261/LER-2006-001 | Manual Reactor Trip Due to Failure of a Turbine Governor Valve Electro-Hydraulic Control Card | 10 CFR 50.73(a)(2)(iv)(A), System Actuation | 05000255/LER-2006-001 | | 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition | 05000461/LER-2006-002 | Turbine Bypass Function Lost Due to Circuit Card Maintenance Frequency | | 05000458/LER-2006-002 | Loss of Safety Function of High Pressure Core Spray Due to Manual Deactivation | | 05000456/LER-2006-002 | Units 1 and 2 Entry into Limiting Condition for Operation 3.0.3 due to Main Control Room Ventilation Envelope Low Pressure | | 05000443/LER-2006-002 | Noncompliance with the Requirements of Technical Specification 6.8.1.2.a | | 05000387/LER-2006-002 | DMissed Technical Specification surveillance requirement | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000362/LER-2006-002 | Unit 3 Shutdown to Inspect Safety Injection Tank Spiral Wound Gaskets | | 05000336/LER-2006-002 | Manual Reactor Trip Due To Trip Of Both Feed Pumps Following A Loss Of Instrument Air | 10 CFR 50.73(a)(2)(iv)(A), System Actuation | 05000316/LER-2006-002 | | 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000315/LER-2006-002 | Failure to Comply with Technical Specification Requirement 3.6.13, Divider Barrier Integrity | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000293/LER-2006-002 | | | 05000289/LER-2006-002 | | | 05000251/LER-2006-002 | Intermediate Range High Flux Trip Setpoint Exceeded Technical Specification Allowable Value | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000440/LER-2006-002 | Scaffold Built in the Containment Pool Swell Region | 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition | 05000413/LER-2006-002 | Safe Shutdown Potentially Challenged by an External Flooding Event and Inadequate Design and Configuration Control | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000388/LER-2006-002 | Missed Technical Specification LCO 3.8.1 Entry for Unit 2 During Unit 1 ESS Bus Testing | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000348/LER-2006-002 | Main Steam Isolation Valve Failure to Close | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000305/LER-2006-002 | | 10 CFR 50.73(a)(2)(v)(C), Loss of Safety Function - Release of Radioactive Material 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000301/LER-2006-002 | | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000286/LER-2006-002 | 450 Broadway, GSB P.O. Box 249 Buchanan, N.Y. 10511-0249Entergy Tel (914) 734-6700 Fred Dacimo Site Vice President Administration September 13, 2006 Indian Point Unit No. 3 Docket No. 50-286 N L-06-084 Document Control Desk U.S. Nuclear Regulatory Commission Mail Stop O-P1-17 Washington, DC 20555-0001 Subject:L Licensee Event Report # 2006-002-00, "Manual Reactor Trip as a Result of Arcing Under the Main Generator Between Scaffolding and Phase A&B of the Isophase Bus Housing" Dear Sir: The attached Licensee Event Report (LER) 2006-002-00 is the follow-up written report submitted in accordance with 10 CFR 50.73. This event is of the type defined in 10 CFR 50.73(a)(2)(iv)(A) for an event recorded in the Entergy corrective action process as Condition Report CR-IP3-2006-02255. There are no commitments contained in this letter. Should you or your staff have any questions regarding this matter, please contact Mr. Patric W. Conroy, Manager, Licensing, Indian Point Energy Center at (914) 734-6668. Fred R. Dacimo Site Vice President Indian Point Energy Center Docket No. 50-286 NL-06-084 Page 2 of 2 Attachment: LER-2006-002-00 CC: Mr. Samuel J. Collins Regional Administrator — Region I U.S. Nuclear Regulatory Commission U.S. Nuclear Regulatory Commission Resident Inspector's Office Resident Inspector Indian Point Unit 3 Mr. Paul Eddy State of New York Public Service Commission INPO Record Center NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB NO. 3150-0104 EXPIRES: 06/30/2007
(6-2004)
. Estimated burden per response to comply with this mandatory collection request: 50 hours.DReported lessons learned are incorporated into the licensing process and fed back to industry. Send 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 internetLICENSEE EVENT REPORT (LER) 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 means used to impose an information collection does not display a currently valid OMB control 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
INDIAN POINT 3 05000-286 1 OF 6
4.TITLE: Manual Reactor Trip as a Result of Arcing Under the Main Generator Between
Scaffolding and Phase A&B of the Iso-phase Bus Housing | 10 CFR 50.73(a)(2)(iv)(A), System Actuation | 05000282/LER-2006-002 | | 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000269/LER-2006-002 | High Energy Line Breaks Outside Licensing Basis May Result in Loss of Safety Function | | 05000263/LER-2006-002 | | | 05000255/LER-2006-002 | | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications | 05000254/LER-2006-002 | Quad Cities Nuclear Power Station Unit 1 05000254 1 of 3 | 10 CFR 50.73(a)(2)(iv)(A), System Actuation | 05000483/LER-2006-003 | Unexpected Inoperability of the Emergency Exhaust System due to Inoperable Pressure Boundary | 10 CFR 50.73(a)(2)(v)(C), Loss of Safety Function - Release of Radioactive Material 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications |
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