05000286/LER-2018-002, Manual Reactor Shutdown Due to Weld Leak in Safety Injection System Tank

Manual Reactor Shutdown Due to Weld Leak in Safety Injection System Tank

ML18313A129
Person / Time
Site:	Indian Point
Issue date:	11/05/2018
From:	Vitale A Entergy Nuclear Northeast
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
NL-18-081 LER 2018-002-00
Download: ML18313A129 (5)

LER-2018-002, Manual Reactor Shutdown Due to Weld Leak in Safety Injection System Tank

Event date:
Report date:
Reporting criterion:	10 CFR 50.73(a)(1), Submit an LER, Invalid Actuation 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)(iii) 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)
2862018002R00 - NRC Website

text

NL-18-081 November 5, 2018 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Mail Stop O-P1-17 Washington, D.C. 20555-0001 Entergy Nuclear Northeast Indian Point Energy Center 450 Broadway, GSB P.O. Box249 Buchanan, N.Y. 10511-0249 Tel (914) 254-6700 Anthony J Vitale Site Vice President SUBJECT: Licensee Event Report# 2018-002-00, "Manual Reactor Shutdown Due to Weld Leak on Safety Injection System Tank" 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) 2018-002-00. The attached LER identifies an event where the reactor was manually shutdown due to a leak from a thermowell weld on the Safety Injection system Boron Injection Tank, which is reportable under 10 CFR

50. 73(a)(2)(iv)(A). This event was recorded in the Entergy Corrective Action Program as Condition Report CR-IP3-2018-02638.

There are no commitments made or revised in this letter. Should you have any questions regarding this submittal, please contact Mr. Robert Walpole, Manager, Regulatory Assurance at (914) 254-6710.

/

AJV/gd cc:

Mr. David Lew, Regional Administrator, NRC Region I NRC Resident Inspector's Office Ms. Bridget Frymire, New York State Public Service Commission

NRCFORM366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY 0MB: NO. 3150-0104 EXPIRES: 03/31/2020 (04-2018)

, the NRC may not conduct or sponsor, and a oerson is not required to resoond to the information collection.

~.Page Indian Point Unit 3 05000-286 1 OF4

4. Title Manual Reactor Shutdown Due to Weld Leak in Safety Injection System Tank

5. Event Date

6. LER Number

7. Report Date

8. Other Facilities Involved Sequential Rev Facility Name Docket Number Month Day Year Year Month Day Year Number No.

05000 11 2018 Facility Name Docket Number 9

7 2018 2018 002 00 5

05000

9. Operating Mode

11. This Repart is Submitted Pursuant to the Requirements of 10 CFR §: (Check all that af)fJ/V) 1 D 20.2201(b)

D 20.2203(a)(3)(i)

D 50.73(a)(2)(ii)(A)

D 50.73(a)(2)(viii)(A)

D 20.2201(d)

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)

D 20.2203(a)(4)

D 50.73(a)(2)(iii)

D 50.73(a)(2)(ix)(A)

D 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 50.36(c)(2)

D 50.73(a)(2)(v)(B)

D 73.71 (a)(5) 100 D 20.2203(a)(2)(iv)

D 50.46(a)(3)(ii)

D 50.73(a)(2)(v)(C)

D 73.77(a)(1)

D 20.2203(a)(2)(v)

~ 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)

D 50.73(a)(2)(i)(B)

D 50.73(a)(2)(vii)

D 73.77(a)(2)(ii)

D 50.73(a)(2)(i)(C)

D Other (Specify in Abstract below or in NRC Form 366A

12. Licensee Contact for this LER Licensee Contact relephone Number (Include Area Code)

!Nelson Azevedo, Suoervisor, Code Proarams Engineering

[(914) 254-6775

13. Complete One Line for each Companent Failure Described in this Report

Cause

System Component Manufacturer Reportable To ICES

Cause

System Component Manufacturer Reportable To ICES 8

BQ nw 0015

~

14. Supplemental Report Expected Month Day Year D Yes (If yes, complete 15. Expected Submission Date) ~ No

15. Expected Submission Date Abstract (Limit to 1400 spaces, i.e., approximately 14 single-spaced typewritten lines)

On September 7, 2018 during a walkdown, an operator identified boron accumulation on a welded joint to a thermowell located on the Boron Injection Tank, which is a retired-in-place component in the safety injection system. Since the flaw could not be characterized by any available Non-Destructive Examination techniques due to the configuration of the weld, it was conservatively assumed that the weld was structurally unacceptable. This was postulated to result in ejection of the thermowell and a subsequent potential leakage rate that would render the safety injection pumps incapable of performing their functions during applicable design basis events. Additionally, the American Society of Mechanical Engineers (ASME)

Section XI Code does not allow through-weld leakage in any Class 2 components. Therefore, at 1803 hours0.0209 days <br />0.501 hours <br />0.00298 weeks <br />6.860415e-4 months <br />, with reactor power at 100 percent, shutdown of the plant was commenced in accordance with Technical Specification 3.0.3. The reactor was manually tripped at 2129 hours0.0246 days <br />0.591 hours <br />0.00352 weeks <br />8.100845e-4 months <br /> in accordance with normal shutdown procedures.

The Auxiliary Feedwater System was manually initiated to provide feedwater flow to the steam generators in preparation for reactor trip, and all control rods fully inserted and all required safety systems functioned properly. The plant was stabilized in hot standby with decay heat being removed by the main condenser. A plant modification removed and plugged the hermowell and a second identical thermowell on the tank.

tThis event had no effect on the public health and safety. The event was reported to the Nuclear Regulatory Commission (NRC) on September 7, 2018 under 10CFR50.72(b)2(i), 10CFR50.72(b)(3)(ii)(A), and 10CFR50.72(b)(3)(v)(D).

NRC FORM 366 (04-2018)

Note: The Energy Industry Identification System Codes are identified within the brackets { }.

YEAR 2018 SEQUENTIAL NUMBER

- 002 On September 7, 2018 with reactor power at 100 percent, at approximately 0900 hours0.0104 days <br />0.25 hours <br />0.00149 weeks <br />3.4245e-4 months <br /> during operator rounds, REV NO.

- 00 an operator identified boron accumulation on the thermowell {TW} for temperature indicator {Tl} Tl-917 on the Boron Injection Tank (BIT) {TK}.. A leak was apparent based on the dried boron that had accumulated on the outside of the vessel at the thermowell penetration. The operator promptly initiated CR-IP3-2018-02638. During the course of the operability evaluation, it was determined that the flaw could not be characterized by available Non-Destructive Examination (NDE) techniques due to the configuration of the weld, and complete failure of the weld was conservatively assumed as a bounding condition. Additionally, IWC-3000 of the American Society of Mechanical Engineers (ASME)Section XI Code does not allow through-weld flaws in Class 2 components. Due to the relatively high operating pressure of 1500 psi, there are no applicable NRC approved ASME code cases which could be used to restore operability without an NRC approved relief request. Once this was identified the BIT and safety injection (SI) {BQ} header were declared Inoperable. Technical Specification 3.0.3 was entered at 1711 hours0.0198 days <br />0.475 hours <br />0.00283 weeks <br />6.510355e-4 months <br /> because Condition A of Technical Specification 3.5.2 could not be met - one or more Emergency Core Cooling System (ECCS) {BQ} trains were inoperable but only one SI pump {P} was operable, and Condition A requires two operable SI pumps. Plant shutdown was commenced at 1803 hours0.0209 days <br />0.501 hours <br />0.00298 weeks <br />6.860415e-4 months <br />, the Auxiliary Feedwater (AFW) system {BA} was manually initiated in preparation for manual reactor trip, and manual reactor shutdown occurred at 2129 hours0.0246 days <br />0.591 hours <br />0.00352 weeks <br />8.100845e-4 months <br /> at approximately 30 percent reactor power. All control rods {AA} fully inserted and all required safety systems functioned properly. The plant was stabilized in hot standby with decay heat being removed by the main condenser {SG}. There was no radiation release. The emergency diesel generators {EK, DG} did not start, as offsite power remained available and stable. Plant cooldown was initiated at 0035 hours4.050926e-4 days <br />0.00972 hours <br />5.787037e-5 weeks <br />1.33175e-5 months <br /> on September 8, and Mode 4 (hot shutdown) was entered at 0405 hours0.00469 days <br />0.113 hours <br />6.696429e-4 weeks <br />1.541025e-4 months <br />. The following notifications were made to the NRC under Event Notification 53589 at 2056 hour0.0238 days <br />0.571 hours <br />0.0034 weeks <br />7.82308e-4 months <br />s: 4-hour per 1 OCFR50.72(b)2(i) for the initiation of a plant shutdown required by the Technical Specifications; 8-hour per 10CFR50.72(b)(3)(ii)(A) for an event or condition that resulted in the plant being in a seriously degraded condition; and 8-hour per 10CFR50.72(b)(3)(v)(D) for a safety system functional failure. Indian Point Unit 2 {IP2) was unaffected and remained at 100 percent power.

The original purpose of the BIT was to provide for the injection of highly concentrated boric acid into the reactor coolant system (RCS) {AB} cold legs following actuation of a safety injection signal to mitigate the reactivity addition resulting from a Main Steam Line Break (MSLB). The BIT was originally designed to contain a high concentration boric acid solution and was equipped with heaters, with fluid temperature sensed by resistance temperature detectors (RTDs) in the thermowells, to maintain the temperature of the fluid at or above the Technical Specification limit of 145 degrees Fahrenheit (F) to keep the boric acid from precipitating out of solution.

The design temperature of the BIT is 300F. The BIT is now retired in place, but has a passive safety function to maintain pressure boundary integrity as it is still part of the SI system.

The BIT is pressurized quarterly during the performance of quarterly SI pump surveillance tests. Additionally, an ASME Section XI required pressure test is performed three times every 10 years. During performance of this test, the BIT is pressurized to between 1450-1550 pounds per square inch (psi) (the design pressure of the vessel is 1750 psi) and held at that pressure for four hours before performing a visual examination to check for leakage.

This test was last performed in March 2017 with no leakage identified.

The BIT is in the flowpath of the SI system as it is in the discharge path to the RCS for 32 SI Pump and 33 SI Pump. Since characterization of the through-wall flaw was not practical, it was conservatively assumed that complete failure of the thermowell occurs during the performance of the required safety functions of the SI system.

In other words, when the SI system is required to operate during applicable design basis accidents, it was postulated that the weld fails and the thermowell is ejected from the BIT, creating a hole sufficient in size to result in a maximum leakage rate that would render the SI pumps incapable of performina their desian basis functions Page 2 of 4 (04-2018)

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1. FACILITY NAME

2. DOCKETNUMBER

3. LERNUMBER YEAR Indian Point Unit 3 05000-286 2018 SEQUENTIAL NUMBER

- 002 REV NO.

- 00 during the injection and recirculation phases. Therefore, the BIT flowpath was considered to be inoperable, which rendered two ECCS trains inoperable. Technical Specification 3.0.3 was entered at 1711 hours0.0198 days <br />0.475 hours <br />0.00283 weeks <br />6.510355e-4 months <br /> because Condition A of Technical Specification 3.5.2 could not be met - one or more ECCS trains were inoperable but only one SI pump was operable, and Condition A requires two operable SI pumps. The 32 SI Pump can inject via two flowpaths, but must be aligned to both the BIT flowpath and the non-BIT flowpath to be considered operable.

Subsequent to the event, the thermowells were examined by surface examination. In addition, a failure analysis consisting of hardness testing, metallurgical examination, scanning electron microscopy, finite element analysis, and fracture mechanics analysis were performed. The thermowell is attached to the inside surface of the BIT by a partial penetration weld. The failure analysis identified three distinct cracking mechanisms. Voids/cracks were determined to be consistent with solidification cracks in the weld metal, due to the highly austenitic weld material.

Fine branched cracks (Transgranular Stress Corrosion Cracking or TGSCC) in the weld material are believed to be the mechanism which propagated solidification cracks through the weld material resulting in through-wall leakage. lntergranular attack (lntergranular Stress Corrosion Cracking or IGSCC) on the nozzle material is likely the result of sensitization caused by the welding process and is not believed to be a significant contributing factor to the crack propagation.

The fracture mechanics and finite element analysis performed for the identified flaws confirmed that the nozzle and the associated weld remained structurally stable under the stresses associated with design basis loading. As such, the observed leakage would not have increased in an unstable manner as a result of the design basis loading conditions.

Based on these results, it was concluded that there was no loss of safety function of the SI system as a result of these flaws, as opposed to the initial 8-hour notification per 10 CFR 50. 72(b)(3)(v)(D) for a safety system functional failure.

CAUSE OF EVENT

The direct cause of the through-wall leakage on the weld was solidification cracking of the original fabrication weld due to inadequate filler metal used during construction of the vessel. This cracking then propagated through the weld thickness by stress corrosion cracking. The original filler metal used by the manufacturer to attach the thermowell to the BIT was a stainless steel filler metal which was not adequate for joining the Alloy 600 thermowell to the carbon steel BIT and ultimately led to cracking of the weld.

Contributing causes are TGSCC of the original fabrication weld, which occurred due to the solidification cracking that existed, and IGSCC of the original lnconel penetration nozzle due to sensitization of the lnconel material during original construction. Either or both could have provided a growth mechanism by which crack growth of the existing solidification cracks could have occurred and led to through-wall leakage.

An extent of condition evaluation was performed for dissimilar metal lnconel welds in ISi Class 2 or 3 vessels or heat exchangers, and, for partial penetration welds at a nozzle or penetration welded from the inside of the vessel or heat exchanger. *Other than on the IP3 BIT, the review did not identify any similar configurations at either IP2 or IP3. Note that IP2 does not have a BIT.

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3. LER NUMBER YEAR Indian Point Unit 3 05000-286 2018 SEQUENTIAL NUMBER

- 002 The following corrective actions have been or will be performed under the Entergy Corrective Action Program to address the causes of this event.

A second thermowell (TW-918) {lW} was also liquid penetrant tested and similar flaws were found, but there was no leakage Both thermowells were removed and permanently plugged (EC-79305) with a welded plug design The removed parts were shipped to a radioactive materials testing lab for failure analysis

EVENT ANALYSIS

REV NO.

- 00 The event was reported under 10 CFR 50. 72(b )(2)(i) for the initiation of the plant shutdown, and is reportable under 10 CFR 50. 73(a}(2)(i)(A}, "the licensee shall report the completion of any nuclear plant shutdown required by the plant's Technical Specifications." This event was also reported under 10 CFR 50.72(b)(3)(v)(D) as an event or condition that could have prevented the fulfillment of the safety function of a system that is needed to mitigate the consequences of an accident. Since the failure analysis has demonstrated that the weld would have remained structurally intact and would not have leaked sufficiently to prevent fulfillment of the safety functions of the SI system, this event is not reportable under 10 CFR 50. 73(a)(2)(v)(D), "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 mitigate the consequences of an accident." This event was further reported under 10CFR50.72(b)(3)(ii)(A) "for any event that results in the condition of the nuclear power plant, including its principal safety barriers, being seriously degraded." Again. since the failure analysis has demonstrated that the weld would have remained structurally intact and would not have leaked sufficiently to prevent fulfillment of the safety functions of the Safety Injection system, this event is not reportable under 10 CFR 50. 73(a)(2)(ii)(A}, the licensee shall report any event or condition that resulted in the condition of the nuclear power plant, including its principal safety barriers, being seriously degraded.

PAST SIMILAR EVENTS A review was performed of the past five years for IP2 and IP3 Licensee Event Reports (LERs) for original design weld flaws. There were no past similar events.

SAFETY SIGNIFICANCE

This event had no effect on the health and safety of the public. There were no actual safety consequences for the event because it was an uncomplicated manual reactor trip with no other transients or accidents, and the required primary safety systems performed as designed.

For this event, all control rods inserted as required upon initiation of the reactor trip. The RCS pressure remained below the setpoint for pressurizer power operated relief valve (PORV) {AB, RV} and code safety valve {AB, RV}

operation, and above the setpoint for automatic SI actuation. Following the reactor trip, the plant was stabilized in hot standby with decay heat being removed by the main condenser. Page 4 of 4