05000249/LER-1917-001, Regarding Standby Liquid Control System Inoperable Due to a Manufacturing Defect Causing a Piping Leak

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Regarding Standby Liquid Control System Inoperable Due to a Manufacturing Defect Causing a Piping Leak
ML17319A090
Person / Time
Site: Dresden Constellation icon.png
Issue date: 11/10/2017
From: Karaba P
Exelon Generation Co
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
17-0045 LER 17-001-00
Download: ML17319A090 (4)


LER-1917-001, Regarding Standby Liquid Control System Inoperable Due to a Manufacturing Defect Causing a Piping Leak
Event date:
Report date:
Reporting criterion: 10 CFR 50.73(a)(2)(v), Loss of Safety Function

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)(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
2491917001R00 - NRC Website

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751 Exelon Generation SVPL TR # 17-0045 November 10, 2017 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Dresden Nuclear Power Station, Unit 3 Renewed Facility Operating License Nos. DPR-25 NRG Docket No. 50-249 Dresden Nuclear Power Station 6500 North Dresden Road Morris, IL 60450 10 CFR 50.73

Subject:

Licensee Event Report 249/2017-001-00, Unit 3 Standby Liquid Control System Inoperable Due to a Manufacturing Defect Causing a Piping Leak Enclosed is Licensee Event Report 249/2017-001-00, "Unit 3 Standby Liquid Control System Inoperable Due to a Manufacturing Defect Causing a Piping Leak." This report describes events which are being reported in accordance with 10 CFR 50.73(a}(2}(v)(A), "Any event or condition that could have prevented the fulfillment of the safety function of... systems that are needed to shut down the reactor and maintain it in a safe shutdown condition," and in accordance with 10 CFR 50.73(a)(2)(v)(D), "Any event or condition that could have prevented the fulfillment of the safety function of... systems that are needed to mitigate the consequences of an accident."

There are no regulatory commitments contained in this submittal.

Should you have any questions concerning this letter, please contact Mr. Bruce Franzen at (815} 416-2800.

Respectfully, Peter J Karaba Site Vice President Dresden Nuclear Power Station Enclosure Licensee Event Report 249/2017-001-00 cc:

Regional Administrator-NRG Region Ill NRG Senior Resident Inspector-Dresden Nuclear Power Station

NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY 0MB: NO. 3150-0104 EXPIRES: 03/31/2020 (04-2017)

, the http://www. nrc.i:iov/readinQ-rm/doc-collections/nurei:is/staff/sr1022/r3/)

NRG may not conduct or sponsor, and a person is not required to respond to, the information collection.

3. PAGE Dresden Nuclear Power Station, Unit 3 05000249 1 OF 3
4. TITLE Unit 3 Standby Liquid Control System Inoperable Due to a Manufacturing Defect Causing a Piping Leak
5. EVENT DATE
6. LER NUMBER
7. REPORT DATE
8. OTHER FACILITIES INVOLVED YEAR I SEQUENTIAL I REV FACILITY NAME DOCKET NUMBER MONTH DAY YEAR MONTH DAY YEAR N/A N/A NUMBER NO.

J FACILITY NAME DOCKET NUMBER 09 12 2017 2017 - 001

- 00 11 10 2017 N/A N/A
9. OPERATING MODE
11. THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR §: (Check all that apply)

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) 1 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)

[gl 50.73(a)(2)(v)(A)

D 13.11(a)(4>

D 20.2203(a)(2>(m>

D 5o.36(c>(2>

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

D 13.11(a)(5>

D 20.2203(a)(2)(iv)

D 5o.46(a)(3)(ii)

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

D 13.11(a)(1>

100 D 20.2203(a)(2)(v)

D 50.73(a)(2)(i)(A)

[gl 50. 73(a)(2)(v)(D)

D 13_11(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)

!ft i~ *)")if::,=;~,~:; *r1~:,~rr?

D 50. 73(a)(2)(i)(C)

D OTHER Specify in Abstract below or in C.

Cause of Event

D.

The cause of these events is under investigation and will be documented in a supplemental report.

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 boron 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.

Additional impacts to the safety analysis will be identified during the root cause investigation and will be documented in a supplemental report.

E.

Corrective Actions

Corrective actions will be developed during the ongoing root cause investigation and will be documented in a supplemental report.

F.

Previous Occurrences

Previous occurrenc'es will be identified through the root cause investigation and will be documented in a supplemental report.

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. Page of _3_