05000286/LER-2016-001, Regarding Safety System Functional Failure Due to an Inoperable Containment Caused by a Flaw on the 31 Fan Cooler Unit Service Water Return Coil Line Affecting Containment Integrity
| ML17003A007 | |
| Person / Time | |
|---|---|
| Site: | Indian Point  |
| Issue date: | 12/21/2016 |
| From: | Vitale A Entergy Nuclear Operations, Indian Point |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| NL-16-139 LER 16-001-00 | |
| Download: ML17003A007 (9) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(ii)(A), Seriously Degraded 10 CFR 50.73(a)(2)(viii)(A) 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition 10 CFR 50.73(a)(2)(viii)(B) 10 CFR 50.73(a)(2)(ix)(A) 10 CFR 50.73(a)(2)(iv)(A), System Actuation 10 CFR 50.73(a)(2)(x) 10 CFR 50.73(a)(2)(v)(A), Loss of Safety Function - Shutdown the Reactor 10 CFR 50.73(a)(2)(v)(B), Loss of Safety Function - Remove Residual Heat 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(i)(A), Completion of TS Shutdown 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(i) |
| 2862016001R00 - NRC Website | |
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Indian Point Energy Center 450 Broadway, GSB
-===- Entergy P.O. Box249 Buchanan, N.Y. 10511-0249 Tel (914) 254-6700 NL-16-139 December 21, 2017 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Mail Stop O-P1-17 Washington, D.C. 20555-0001 Anthony J. Vitale Site Vice President
SUBJECT:
Licensee Event Report# 2016-001-00 "Safety System Functional Failure Due to an Inoperable Containment Caused by a Flaw on the 31 Fan Cooler Unit Service Water Return Coil Line Affecting Containment Integrity" Indian Point Unit No. 3 Docket No. 50-286 DPR-64
Dear Sir or Madam:
Pursuant to 10 CFR 50.73(a)(1}, Entergy Nuclear Operations Inc. (ENO) hereby provides Licensee Event Report (LER) 2016-001-00. The attached LER identifies an event where there was a safety system functional failure due to an inoperable Containment as a result of a through wall Service Water (SW) piping leak at the 31 Containment Fan Cooler Unit at a weld in a branch line from coil 3 feeding a SW return line header. This condition is reportable under 10 CFR 50. 73(a)(2)(v). This condition was recorded in the Entergy Corrective Action Program as Condition Report CR-IP3-2016-03607.
NL-16-139 Docket No. 50-286 Page 2 of 2 There are no new commitments identified in this letter. Should you have any questions regarding this submittal, please contact Mr. Robert Walpole, Manager, Licensing at (914) 254-6710.
Sincerely, AJV/cbr cc:
Mr. Daniel H. Dorman, Regional Administrator, NRC Region I NRC Resident Inspector's Office Ms. Bridget Frymire, New York State Public Service Commission
-.-~.*.
NRC FORM 366 (11-2015)
U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB: NO. 3150-0104 EXPIRES: 10/31/2018 I
LICENSEE EVENT REPORT (LER)
, the NRG may not conduct or sponsor, and a person is not required to respond to, the information (See Page 2 for required number of digits/characters for each block) nnllnn.:nn
- 1. FACILITY NAME
- 12. DOCKET NUMBER r* PAGE Indian Point 3 05000-286 1 OF 7
- 4. TITLE: Safety System Functional Failure Due to an Inoperable Containment Caused by a Flaw on the 31 Fan Cooler Unit Service Water Return Coil Line Affecting Containment Integrity
- 5. EVENT DATE
- 6. LER NUMBER
- 7. REPORT DATE
- 8. OTHER FACILITIES INVOLVED YEAR I SEQUENTIAL I REV MONTH NUMBER NO.
YEAR FACILITY NAME DOCKET NUMBER MONTH DAY YEAR DAY FACILITY NAME DOCKET NUMBER 11 03 2016 2016 001 00 12 21 2017
- 9. OPERATING MODE
- 11. THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR §: (Check all that aoolv)
D 20.2201 (b)
D 20.2203(a)(3)(i)
D 50.73(a)(2)(ii)(A)
D 50.73(a)(2)(viii)(A) 1 D 20.2201 (ct)
D 20.2203(a)(3)(ii)
D 50.73(a)(2)(ii)(B)
D 50.73(a)(2)(viii)(B)
D 20.2203(a)(1 l D 20.2203(a)(4)
D so.73(a)(2)(iii)
D 50.73(a)(2)(ix)(A)
[] 20.2203(a)(2)(i)
D 50.36(c)(1 )(i)(A)
D 50.73(a)(2)(iv)(A)
D 50.73(a)(2)(x)
- 10. POWER LEVEL D 20.2203(a)(2)(ii)
D 50.36(c)(1 )(ii)(A)
D 50.73(a)(2)(v)(A)
D 73.71 (a)(4)
D 20.2203(a)(2)(iii)
D so.36(c)(2)
D 50.73(a)(2)(v)(B)
D 73.71 (a)(5) 100%
D 20.2203(a)(2)(iv)
D so.46(a)(3)(ii)
~ 50.73(a)(2)(v)(C)
D 73.77(a)(1)
D 20.2203(a)(2)(v)
D 50.73(a)(2)(i)(A)
D 50.73(a)(2)(v)(D)
D 73.77(a)(2)(i)
D 20.2203(a)(2)(vi)
~ 50.73(a)(2)(i)(B)
D 50.73(a)(2)(vii)
-. D 73.77(a)(2)(Hl.**
D 50.73(a)(2)(i)(C) 00THER Specify in Abstract below or in YEAR I
SEQUENTIAL I
NUMBER 2016 001 REV NO.
00 The piping for the 3 inch cooling coil return line is 904L stainless steel (SS) with a nominal pipe wall thickness for Schedule 40 pipe (0.216 inches).
Leak rate estimate was 0.16 gpm.
The leak is at a 3 inch butt-welded joint between a 904L stainless steel elbow {PSF} and pipe {PSP} located on approximately the 76 foot elevation in containment.
The piping leak is in a moderate energy ASME ISI Code Class 3, nuclear safety related piping system.
904L SS material is susceptible to the development of corrosion pits.
Pin hole leaks and weld defects in this piping have previously occurred and have been evaluated.
The evaluation concluded the 904L piping does not have a general corrosion problem.
Current analysis for SW pipe failures are postulated to be limited to small through-wall leakage flaws as opposed to guillotine breaks.
There is no evidence of leakage at any other location on this weld or elsewhere on the piping adjacent to it.
Code Class 3 piping systems are addressed in ASME Code Case N-513-3.
This Code Case provides the requirements for demonstrating structural integrity and therefore operability of a flawed pipe section.
Characterization of the weld condition was performed by conducting an ultrasonic examination (UT) on November 4, 2016.
The weld was examined circumferentially in a ~ inch by *1 inch grid pattern and by using a bulls-eye grid pattern in ~ inch increments.
These UT, (Non Destructive Examinations) NDE results were documented in an NDE report.
However, due to physical obstructions presented by the FCU enclosure, two circumferential grid rows on the backside of the weld could not be reached.
Due to the leak location on the bottom side of the pipe on the side opposite from the FCU enclosure, the bulls-eye grid was not obstructed, and all required readings were taken.
As a result of limited access, the UT examination of weld B297 resulted in completion of only approximately 70 percent of the circumference of the weld.
The remaining 30 percent (approximately 3 inches) was unable to be inspected due to space constraints between the weld and the adjacent FCU plenum wall.
ASME Code Case N-513-3. requires the flaw geometry to be characterized by volumetric inspection methods or by physical measurement.
It mandates that the full pipe circumference at the flaw location be inspected to characterize the length and depth of all flaws in the pipe section.
Since a full volumetric examination could not be completed, the Code Case requirements could not be met.
An immediate on-line weld repair of the defect was not considered feasible due to restrictions preventing 360 degree access, time required for work prep, and the potential for excessive sump filtration loading.
As such, an NRC Relief Request to deviate from the 10CFR50.55a ASME Code requirements, specificaily full compliance with ASME CC-N-513-3 was required.
Therefore, pursuant to 10CFR50.55a(z) (2) Entergy requested relief by letter NL-16-133 (Request IP3-ISI-RR-10, Alternative to the Full Circumferential Inspection Requirement of Code Case N-513-3), dated November 7, 2016.
The NRC approved the relief request which concluded the inability to obtain full circumference readings does not adversely impact the ability to fully characterize the weld condition vs the code case requirements.
NRG FORM 366 (11-2015)
The approval of the relief request allowed Entergy to re-establish SW to the FCU to verify that leakage limits are met using a qualified clamp over the pinhole leak.
The leakage will be inspected daily in accordance with the Code Case requirements.
SW piping to the 31 FCU must be isolated when an allowable leakage is exceeded.
The qualified clamp is an engineered clam-shell type clamp comprised on an outer metal jacket and rubber gasket.
The clamp is classified as a temporary modification.
The clamp will withstand post-accident containment pressure, temperature and environment, is located away from potential missiles and.pipe whip effects, is dedicated for safety-related use, and is rigidly attached to Seismic Category 1 piping.
The clamp over the defect will re~urn the system to its original containment integrity configuration and allow the 31 FCU to remain operable.
The defect in the SW return pipe from the 31 FCU was evaluated with respect to TS 5.5.15 (Containment Leak Rate Testing Program), and the Appendix J Leakage program.
TS 5.5.15 requires that the SW in-leakage into containment must be limited to less than 0.36 gpm per FCU when pressurized equal to or greater than l.lPa.
This limit protects the internal recirculation pumps from flooding during the 12 month period of post-accident recirculation.
TS 5.5.15 also implements the leakage rate testing of the containment as required by 10CFR50, Appendix J.
The maximum leakage to assure that the post-accident containment leakage remains within allowable limits is 0.023 gpm.
This limit is based upon an evaluation to calculate the amount of SW which can leak through this pinhole under normal system operating conditions to ensure that 10CFR50 appendix J containment leakage limits are not exceeded under any mode of operation including accident conditions.
The leak does not impinge upon any safety related equipment.
As a result, no damage from the leakage.is expected to occur.
Based on UT measured readings from the NDE Report, a new calculation was generated (IP-CALC-16-00079 FCU 31 Leak) per the ASME CC-N-513-3 requirements.
The minimum required thickness for the elbow containing the weld B297 is 0.073 inches.
The minimum measured thickness was 0.117 inches.
The maximum allowable axial flaw size is 4.11 inches and the maximum allowable circumferential flaw size is 3.65 inches.
The existing flaw is characterized as approximately 0.50 inches by 0.50 inches, and the uninspected arc length (approximately 3 inches) of the pipe circumference is less than the allowable circumferential flaw length.
Therefore, if the entirety of the uninspected portion of the pipe were to be considered a flaw, the pipe would still retain its structural integrity as evaluated in the new calculation.
The pinhole flaw is opposite the uninspected portion and the flaw sizes of the two areas are independent and not additive.
Based on this information, the pipe is structurally adequate for service consistent with the requirements of ASME Code Case N-513-3.
The remaining service life was calculated to be 3.3 years, which is beyond the next scheduled refueling outage in the spring 2017 when a permanent repair will be made.
An extent of condition review determined the Code Case requires five similar and susceptible locations in the SW system to be volumetrically examined.
NDEs were performed on November 5, 2016, on five 31 FCU SW pipe welds and recorded in NDE reports.
All five weld locations were found to be structurally acceptable and documented in CR-IP3-2016-03607, corrective action (CA-6).
The additional inspections confirm the integrity of the SW piping inspected since all UT data measurements were above the 87.5 percent of pipe nominal wall thickness.
Also, there was no evidence of additional leakage at any other place in the 31 FCU 3 inch return line or at any other location in the other four FCU return lines.
Unit 2 does not apply as it does not have similar 904L SS FCU supply or *return lines.
NRG FORM 366 (11-2015)
CAUSE OF EVENT
The direct cause was a SW leak associated with the 31 FCU at weld B297 in branch line C from coil 3 feeding the 10 inch SW return line 12b.
The leak was from a through-wall pinhole flaw at a butt-welded joint between a 904L SS elbow and pipe in containment.
The likely degradation mechanism leading to the leak was corrosion.
The pipe with the flaw resulted in containment out leakage in excess of 10CFR50, Appendix J limits.
The root cause is indeterminate.
The specific cause for the pipe joint defect requires the component to be removed and a metallurgical failure analysis performed.
CORRECTIVE ACTIONS
00 The following corrective actions have been or will be performed under the Corrective Action Program (CAP) to address the causes of this event:
- A leak-limiting engineered clam-shell type clamp was applied to the pipe flaw.
- The clamp is being monitored daily by a special operator log for any signs of increased leakage.
The maximum allowed leakage rate past the clamp is 0.023 gpm.
- UT monitoring will be performed every 90 days until the pipe is repaired.
- The pipe/elbow will be replaced in the next refueling outage (RO) in the spring 2017.
- The removed pipe/elbow will be inspected and a metallurgical analysis performed by an independent vendor to determine the specific cause.
- A volumetric inspection of a sample of 3 inch 904L SS butt-welds at the 32, 33, 34, and 35 FCUs will be performed in the spring 2017 RO.
- The Generic Letter 89-13 Program will be revised to include a requirement to conduct a definitive number of 904L weld volumetric inspections each pre-outage interval.
- This LER will be updated after engineering review of the metallurgical analysis and revision as necessary of the cause analysis.
EVENT ANALYSIS
The event is reportable under 10 CFR 50.73(a) (2) (v) (C).
The licensee shall report any event or condition that could have prevented the fulfillment of the safety function of structures or systems that are needed to (C) Control the release of radioactive material.
This condition meets the reporting criteria because TS 3.6.1 Containment Operability was not met.
NRG FORM 366 (11-2015)
The pipe flaw leakage was approximately 0.16 gpm which was greater than the calculated 10 CFR 50, Appendix J allowable leak rate of 0.023 gpm.
TS 3.6.1 (Containment) requires the containment to be operable in Modes 1-4.
TS Surveillance Requirement (SR) 3.6.1.1 requires visual examinations and leakage rate testing in accordance with the containment Leakage Rate Testing Program specified in TS 5.5.15.
SR 3.6.1.l leakage rate requirements comply with 10 CFR 50, Appendix J, Option B.
As SW is required in an accident, the SW to the FCU would not be isolated in DBA and the piping credited as a closed system inside containment for containment integrity.
Consequently, defects discovered within the FCU SW piping may adversely affect containment integrity and the ability to control releases of radioactive material.
The condition also meets the reporting criteria of 10 CFR 50. 73 (a) (2) (i) (B).
The licensee shall report any operation or condition which was prohibited by the plant's TS.
During the previous period of operation for an unknown period of time the SW pipe contained a through wall leak that did not meet code requirements.
This previously unrecognized condition required entry into TS 3.7.9 and corrective actions implemented to return the pipe to operable.
Failure to comply with the TS LCO and perform required actions is a TS prohibited condition.
PAST SIMILAR EVENTS A review of the past three years of Licensee Event Reports (LERs) for events that involved containment integrity due to flawed piping credited as a closed system inside containment.
No applicable LERs were identified. There was one LER, LER-2014-002 reporting a Technical Specification prohibited condition for a flaw discovered on a SW pipe connected to the Component Cooling Water Heat Exchanger.
This LER is not similar as the impacted piping is outside containment and not credited as a closed system for containment integrity.
SAFETY SIGNIFICANCE
This condition had no effect on the health and safety of the public.
There were no actual safety consequences for the event because there were no accidents or events during the degraded condition.
There were no significant potential safety consequences of this event.
The leakage from the affected SW pipe was within the capability of the SW system to provide adeqiiate SW flow to SW loads.
The degraded piping was on the discharge of the FCU therefore any failure would not prevent the SW cooling function.
Current analysis for SW pipe failures are postulated to be limited to small through-wall leakage flaws as SW is defined as a moderate energy fluid system.
The SW leak would eventually drain to the containment sump.
The containment sumps have pumps with sufficient capacity to remove excessive leakage and instrumentation to alert operators to a degraded condition.
NRC FORM 366 (11-2015)
The containment consists of the concrete reactor building, its steel liner, and the penetrations through the structure.
The containment building is designed to contain radioactive material that might be released from the reactor following a design basis accident (DBA).
The containment building steel liner and its penetrations establish the leakage limiting boundary of the containment.
Maintaining the containment operable limits the leakage of fission product radioactivity from the containment to the environment.
The DBA analysis assumes that the containment is operable such that, for the DBAs involving release of fission product radioactivity, release to the environment is controlled by the rate of containment leakage.
The containment was designed with an allowable leakage rate of 0.1 percent of containment air weight per day.
Containment isolation valves form a part of the containment pressure boundary.
Two barriers in series are provided for each penetration so that no single credible failure or malfunction of an active component can result in a loss of isolation or leakage that exceeds limits assumed in the
safety analysis
One of these barriers may be a closed system such as the SW piping for the FCUs.
The only time containment integrity can be affected is post accident when the FCUs safety function is being performed and SW pressure for the FCU cooling piping and coils fall below peak accident pressure.
Mitigation of radiation release by the degraded SW pipe pathway can be by use of radiation monitors R-16A and R-16B which monitor containment fan cooling water for radiation indicative of a leak from the containment atmosphere into the cooling water. If radiation is detected, each FCU heat exchanger can be individually sampled to determine the leaking unit.
The SW for the 31 FCU can be isolated to prevent radioactive effluent releases.
During the time the FCU SW piping was degraded there was no leakage out of containment.
A risk assessment was performed to determine the overall probability of a core damage event which could cause a loss 'of containment integrity due to a SW to FCU leak assuming it would take 5 days to detect a SW leak to a FCU.
The risk result was 7.8E-8 which is considered negligible in terms of both core damage and large early release.
NRG FORM 366 (11-2015)