05000249/LER-2017-001

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LER-2017-001, Unit 3 Standby Liquid Control System Inoperable Due to a Manufacturing Defect Causing a Piping Leak
Dresden Nuclear Power Station, Unit 3
Event date:
Report date:
Reporting criterion: 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications

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)(A), Loss of Safety Function - Shutdown the Reactor
LER closed by
IR 05000237/2017004 (22 January 2018)
2492017001R01 - NRC Website
LER 17-001-01 for Dresden Nuclear Power Station, Unit 3 Regarding Unit 3 Standby Liquid Control System Inoperable Due to a Manufacturing Defect Causing a Piping Leak
ML17363A214
Person / Time
Site: Dresden Exelon icon.png
Issue date: 12/27/2017
From: Karaba P J
Exelon Generation Co
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
17-0048
Download: ML17363A214 (5)


comments regarding burden estimate to the Information Services Branch (T-2 F43), U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, or by e-mail to 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.

3. LER NUMBER

17 - 01 001

PLANT AND SYSTEM IDENTIFICATION

Dresden Nuclear Power Station (DNPS), Unit 3, is a General Electric Company Boiling Water Reactor with a licensed maximum power level of 2957 megawatts thermal. The Energy Industry Identification System codes used in the text are identified as [XX].

A. Plant Conditions Prior to Event:

Unit: 03 Reactor Mode: 1 Event Date: 09/12/17 Event Time: 1131 CDT Mode Name: Power Operation Power Level: 100 percent

B. Description of Event:

On September 10, 2017, during Equipment Operator (EO) rounds, the EO found crystalized boron on Dresden, Unit 3 Standby Liquid Control System (SLC) [BR] discharge piping [PSF]. There was no active leak at the time of discovery and the source of the boron crystals was unknown. Work activities began to determine the source of the boron deposits.

On September 12, 2017, the Division 1 SLC pump [P] was started to pressurize the system to the normal In-Service Testing test pressure as directed by station procedures. With the system pressurized, a leak of approximately one drop per minute was identified on the common discharge line of the SLC pumps. The leak was characterized as a through wall leak from an American Society of Mechanical Engineers (ASME), Class 2 pressure boundary of the SLC system. This leak was treated as a structural integrity issue; therefore, the affected piping was isolated in accordance with station procedures. Isolating the failed piping resulted in both divisions of SLC being declared Inoperable. This action led to entering Technical Specification (TS) Limiting Condition for Operation (LCO) 3.1.7, "Standby Liquid Control (SLC) System," Condition B, "Two SLC subsystems inoperable," and Required Action B.1, "Restore one SLC subsystem to OPERABLE status," with a Completion Time of eight hours.

DNPS requested a Notice of Enforcement Discretion (NOED) to exceed the TS

8 hours
9.259259e-5 days
0.00222 hours
1.322751e-5 weeks
3.044e-6 months

Completion Time to complete the pipe repair and replace the pipe. The NRC verbally granted the NOED on September 12, 2017 at

1746 hours
0.0202 days
0.485 hours
0.00289 weeks
6.64353e-4 months

. At

2035 hours
0.0236 days
0.565 hours
0.00336 weeks
7.743175e-4 months

, the failed piping was replaced in accordance with station work instructions, thereby restoring the Unit 3 SLC system to Operable status within the time allowed by the NOED.

The piping through-wall flaw did not meet ASME Code structural integrity requirements.

Therefore, Dresden determined that since there was boron buildup identified on September 10 and later confirmed leakage on September 12, the system was inoperable for longer than allowed by TS and is reportable under 10 CFR 50.73(a)(2)(i)(B) for a condition prohibited by TS.

comments regarding burden estimate to the Information Services Branch (T-2 F43), U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, or by e-mail to 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.

3. LER NUMBER

17 - 01 001 Subsequently, the pipe flaw was analyzed and an evaluation concluded that the leak flaw met the ASME Code, Section XI structural margin, considering Service Level D structural factor in the evaluation, as required by the NRC Inspection Manual. Even though the flaw was through- wall, which exceeded the ASME Code, Section XI allowable flaw depth of 75% of wall thickness, the safety function of the component was not compromised when the leak was identified. Based on meeting structural integrity criteria this condition is not reportable under 10 CFR 50.73(a)(2)(v)(A) and 10 CFR 50.73(a)(2)(v)(D) for an event or condition that could have prevented the fulfillment of a safety function.

C. Cause of Event:

The determined root cause is a manufacturing defect which evolved into a through wall leak.

Metallurgical evaluations show that the leak area coincided with a cluster of large, closely spaced inclusions from the manufacturing process. The through wall leakage occurred when the closely spaced inclusions connected due to service induced stresses to form a through wall leak path. There was no evidence that corrosion or stress corrosion cracking contributed to defect growth in the evaluated sample. Based on the qualitative chemistry evaluations of the inclusions, the defects were characterized as oxide inclusions with elevated levels of silicon, manganese and aluminum in comparison to the base metal composition.

D. Safety Analysis:

The SLC System is designed to provide the capability of bringing the reactor, at any time in a fuel cycle, from full power and minimum control rod inventory, which is at the peak of the xenon transient, to a subcritical condition with the reactor in the most reactive, xenon free state without taking credit for control rod movement. The SLC System is also used to maintain suppression pool pH at or above 7 following a Loss of Coolant Accident (LOCA) involving significant fission product releases. Maintaining suppression pool pH levels at or above 7 following an accident ensures that iodine will be retained in the suppression pool water.

The SLC System consists of a bordn solution storage tank, two positive displacement pumps, two explosive valves that are provided in parallel for redundancy, and associated piping and valves used to transfer borated water from the storage tank to the reactor pressure vessel.

The borated solution is discharged near the bottom of the core shroud, where it then mixes with the cooling water rising through the core.

comments regarding burden estimate to the Information Services Branch (T-2 F43), U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, or by e-mail to 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.

3. LER NUMBER

17 - 01 001 An engineering evaluation was performed on the pipe flaw. The evaluation determined that the safety function of the component was not compromised when the leak was identified. The pipe flaw would not have grown beyond the allowable flaw size during SLC System operation. The required flow rate for SLC System to inject into the Reactor Pressure Vessel is 40 gallons per minute minimum. The leakage through the flaw is relatively small (i.e., approximately 1 drop per minute) and would not jeopardize the SLC system from performing its safety function.

Therefore, the leaking component had adequate structural margin when the leak was identified and the resulting leakage would not have prevented the SLC system from performing its intended safety function.

Since the SLC subsystems were available to perform their safety functions, the overall safety significance of this event was minimal. Normal means of reactivity control were maintained during this event.

The engineering analysis demonstrates this event did not constitute a Safety System Functional Failure (SSFF) (Reference NEI 99-02, Regulatory Assessment Performance Indicator Guideline, Section 2.2, "Mitigating Systems Cornerstone, Safety System Functional Failures, Clarifying Notes, Engineering Analyses"). As such, this event will not be reported in the NRC Performance Indicator (PI) for SSFF since an engineering analysis was performed which determined that the system could perform its safety function during this event.

E. Corrective Actions:

The failed piping was replaced in accordance with station procedures and work instructions on September 12, 2017. Additionally, corrective actions include performing extent of condition piping inspections with piping insulation removed.

F. Previous Occurrences:

A search of similar events from the past 15 years was performed. One event was identified:

On January 18, 2007, Dresden, Unit 2 identified a through wall linear crack at the SLC Tank temperature switch well. However, the degradation mechanism (i.e., transgranular stress corrosion cracking) is not the same as this event (i.e., latent forging defect).

G. Component Failure Data:

The piping with the leak was a 1-1/2 inch pipe tee, ASME A/SA-182-Grade stainless steel. The pipe was in service for over 50 years.