PLA-7901, Condition Prohibited by Technical Specifications Due to Drift of Reactor Pressure Switch

Condition Prohibited by Technical Specifications Due to Drift of Reactor Pressure Switch

ML20281A825
Person / Time
Site:	Susquehanna
Issue date:	10/07/2020
From:	Cimorelli K Susquehanna, Talen Energy
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
PLA-7901 LER 2020~002-00
Download: ML20281A825 (4)

Text

Kevin Cimorelli Site Vice President October 7, 2020 Attn: Document Control Desk U. S. Nuclear Regulatory Commission Washington, DC 20555~0001 Susquehanna Nuclear, LLC 769 Salem Boulevard Berwick, PA 18603 Tel. 570.542.3795 Fax 570.542.1504 Kevin.Cimorelli@TalenEnergy.com SUSQUEHANNA STEAM ELECTRIC STATION LICENSEE EVENT REPORT 50-388/2020~002-00 UNIT 2 LICENSE NO. NPF-22 PLA-7901 TALEN~

ENERGY 10 CFR 50.73 Docket No. 50-388 Attached is Licensee Event Report (LER) 50~388/2020~002~00. The LER reports an event involving drifting of Reactor Pressure Steam Dome-Low permissive switches (Microswitch 2).

This event was determined to be reportable as a condition prohibited by Technical Specifications in accordance with 10 CFR 50.73(a)(2)(i)(B) and a condition that could have prevented fulfillment of a safety function in accordance with 10 CFR 50.73(a)(2)(v)(D).

There were no actual consequences to the health and safety of the public as a result of this event.

This letter contains no new or revised regulatory commitments.

Attachment:

LER 50~388/2020-002-00 Copy:

NRC Region I Mr. C. Highley, NRC Sr. Resident Inspector Ms. S. Goetz, NRC Project Manager Mr. M. Shields, PADEP/BRP

NRC FORM366 (08-2020)

U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB: NO. 3150-0104 EXPIRES: 08/31/2023 Estimated burden per response to comply with this mandatory collection request: 80 hours9.259259e-4 days <br />0.0222 hours <br />1.322751e-4 weeks <br />3.044e-5 months <br />. Reported lessons learned are incorporated into the licensing process and fed back to industry. Send LICENSEE EVENT REPORT (LER) comments regarding burden estimate to the FOIA, Library, and Information Collections Branch (T-6 A10M), U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, or by e-mail to (See Page 3 for required number of digits/characters for each block) lnfocollects.Resource@nrc.gov, and the OMB reviewer at: OMB Office of Information and Regulatory (See NUREG-1 022, R.3 for instruction and guidance for completing this Affairs, (3150-0104), A~n: Desk ail: oira submission@omb.e.op.gov. The ~RC may not conduct or f

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1 taff/ 102213{) sponsor, and a person IS not reqUired to respond to, a collection of 1nformat1on unless the document orm ps.

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requesting or requiring the collection displays a currently valid OMB control number.

1. Facility Name

2. Docket Number Susquehanna Steam Electric Station Unit 2 05000388 1

3. Page 1 of 3

4. Title Condition Prohibited by Technical Specifications Due to Drift of Reactor Pressure Switch

5. Event Date Month Day Year 08 10 2020

9. Operating Mode Year

6. LER Number Sequential Number Rev No.

2020 -

002 00 1

7. Report Date Month Day Year Facility Name 10 07 2020 Facility Name 1 0. Power Level

8. Other Facilities Involved 100 Docket Number 05000 Docket Number 05000

11. This Report is Submitted Pursuant to the Requirements of 10 CFR §:

(Check all that apply)

~10"(3FRz:R~'ij:g(}*~ ; D 20.2203(a)(2)(vi)

D 50.36(c)(2)

D 50.73(a)(2)(iv)(A)

D 50.73(a)(2)(x)

D 20.2201(b)

D 20.2203(a)(3)(i)

D 50.46(a)(3)(ii)

D 50.73(a)(2)(v)(A)

} **.* fQS

Cfi'Bt~'c(rf*1$"+::

D 20.2201 (d)

D 20.2203(a)(3)(ii)

D 50.69(g)

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

D 73.71 (a)(4)

D 20.2203(a)(1)

D 20.2203(a)(4)

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

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

D 73.71(a)(5)

D 20.2203(a)(2)(i)

.i:<f*~ffiRz)~a(t.2f~ ~; [gJ 50.73(a)(2)(i)(B)

[gJ 50.73(a)(2)(v)(D)

D 73.77(a)(1)(i)

D 20.2203(a)(2)(ii)

D 21.2(c)

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

D 50.73(a)(2)(vii)

D 73.77(a)(2)(i)

D 20.2203(a)(2)(iii)

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

D 73.77(a)(2)(ii)

D 20.2203(a)(2)(iv)

D 50.36(c)(1)(i)(A)

D 50.73(a)(2)(ii)(B)

D 50.73(a)(2)(viii)(B)

D 20.2203(a)(2)(v)

D 50.36(c)(1)(ii)(A)

D 50.73(a)(2)(iii)

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

D Other (Specify here, in Abstract, or in NRC 366A).

12. Licensee Contact for this LER Licensee Contact C. E. Manges, Jr, Senior Engineer-Nuclear Regulatory Affairs Phone Number (Include Area Code) 570-542-3089

13. Complete One Line for each Component Failure Described in this Report Cause I

System Component I Manufacturer I Reportable to IRIS I Cause I

System I Component I Manufacturer I Reportable to IRIS I

I I

I I

I I

I

14. Supplemental Report Expected 11-_M.:.:.o.:.:.n..:::.th:__l--I_D~a~y-l-1-:Y:-::e-::-a::-r -I I,

15. Expected Submission Date I

D No 1 [gj Yes (If yes, complete 15. Expected Submission Date)

I 12 I

18 2020

16. Abstract (Limit to 1560 spaces, i.e., approximately 15 single-spaced typewritten lines)

On August 10, 2020, the Unit 2 "D" Reactor Steam Dome Pressure-Low permissive pressure switch, Microswitch 2, was found outside of the Technical Specification (TS) 3.3.5.1 allowable value. The switch drifted outside of the upper allowable value which is intended to ensure that the reactor dome pressure has fallen to a value below the Core Spray and Residual Heat Removal (RHR)/Low Pressure Coolant Injection (LPCI) maximum design pressures to preclude over-pressurization of the low pressure systems prior to low pressure injection initiation.

Based on the information available, the condition existed for longer than allowed by TS 3.3.5.1 and TS 3.5.1. As such, this is a condition prohibited by TS and is being reported in accordance with 10 CFR

50. 73(a)(2)(i)(B). In addition, since the "C" channel was surveillance tested just prior to identification of the drift of the "D" channel, redundant channels were inoperable at the same time impacting both Core Spray and LPCI functions; therefore, this is also considered a condition that could have prevented fulfillment of a safety function (10 CFR 50.73(a)(2)(v)(D)). The cause of the event is under investigation and will be provided under a supplement to this LER along with associated corrective actions.

There were no actual consequences to the health and safety of the public as a result of this event.

U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB: NO. 3150-0104 EXPIRES: 08/31/2023 Estimated burden per response to comply with this mandatory collection request: 80 hours9.259259e-4 days <br />0.0222 hours <br />1.322751e-4 weeks <br />3.044e-5 months <br />. Reported lessons learned are incorporated into the licensing process and fed back to industry. Send comments Ll C ENS E E EVENT REP 0 RT ( L E R) regarding burden estimate to the FOIA, Library, and Information Collections Branch (T-6 A 1OM), U.S.

CONTINUATION SHEET Nuclear Regulatory Commission, Washington, DC 20555-0001, or by e-mail to lnfocollects.Resource@nrc.gov, and the OMB reviewer at: OMB Office of Information and Regulatory Affairs, (3150-0104), Attn: Desk ail: oira submission@omb.eop.gov. The NRC may not conduct or (See NUREG-1 022, R.3 for instruction and guidance for completing this form sponsor, and a person is not required to respond to, a collection of information unless the document https:/lwww.nrc.gov/readinq-nn/doc-collections/nureqs/staff/sr1 022/r3D requesting or requiring the collection displays a currently valid OMB control number.

1. FACILITY NAME

2. DOCKET NUMBER

3. LERNUMBER Susquehanna Steam Electric Station Unit 2 05000-388 NARRATIVE CONDITIONS PRIOR TO EVENT Unit 1 - Mode 1, approximately 100 percent Rated Thermal Power Unit 2 - Mode 1, approximately 100 percent Rated Thermal Power YEAR 2020 SEQUENTIAL NUMBER 002 There were no structures, systems, or components that were inoperable at the start of the event that contributed to the event.

EVENT DESCRIPTION REV NO.

00 Prior to September 2017, Susquehanna Steam Electric Station (SSES) had been utilizing International Telephone and Telegraph (ITT)-Barton 288A pressure switches in the Reactor Steam Dome Pressure-Low channels [EllS System/Component Identifier: JE/PS] that provide the injection permissive for the Core Spray system [EllS System Identifier: BM] (Technical Specification (TS) 3.3.5.1, Function 1 d) and the Residual Heat Removal (RHR)/Low Pressure Coolant Injection system (LPCI) [EllS System Identifier:

BO] (TS 3.3.5.1, Function 2d). Due to instrument drift concerns, all eight obsolete ITT-Barton 288A pressure switches were replaced with General Electric (GE) recommended Cameron-Barton 288A pressure switches between September 6, 2017 and November 15, 2017. Following replacement, instrument drift continued to be an issue, and additional corrective action was determined necessary to resolve the drift concerns. The additional corrective action included procuring and installing Cameron-Barton 288A instruments that had been modified to remove an over-range condition and the movement assembly/associated linkages that were determined to be affecting instrument drift. The first of these modified instruments was installed and calibrated on July 8, 2020 in the Unit 2 "D" channel (pressure switch for PS-B21-2N021 D).

On August 10,2020, the new PS-B21-2N021D switch, Microswitch 2, was again tested and found to be outside of TS acceptance criteria. The switch drifted outside of the upper allowable value which is intended to ensure that the reactor dome pressure has fallen to a value below the Core Spray and RHR/LPCI maximum design pressures to preclude over-pressurization of the low pressure systems prior to low pressure injection initiation.

Based on the information available, the condition existed for longer than allowed by TS 3.3.5.1 and TS 3.5.1. As such, this is a condition prohibited by TS and is being reported in accordance with 10 CFR 50.73(a)(2)(i)(B). In addition, since the "C" channel (PS-B21-2N021 D) was surveillance tested just prior to identification of the drift of the "D" channel, redundant channels were inoperable at the same time impacting both Core Spray and LPCI functions; therefore, this is also considered a condition that could have prevented fulfillment of a safety function (10 CFR 50.73(a)(2)(v)(D)).

CAUSE OF EVENT The cause of the event is under review and will be provided in a supplement to this LER.

NRC FORM 366A (08-2020)

Page _6. of.l_

NRC FORM 366A (08-2020)

U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB: NO. 3150*0104 EXPIRES: 08/31/2023 Estimated burden per response to comply with this mandatory collection request: 80 hours9.259259e-4 days <br />0.0222 hours <br />1.322751e-4 weeks <br />3.044e-5 months <br />. Reported lessons learned are Incorporated into the licensing process and fed back to industry. Send comments Ll C ENS E E EVENT REP 0 RT { L E R) regarding burden estimate to the FOIA, Library, and Information Collections Branch (T-6 A 10M), U.S.

CONTINUATION SHEET Nuclear Regulatory Commission, Washington, DC 20555-0001, or by e-mail to lnfocollects.Resource@nrc.gov, and the OMB reviewer at: OMB Office of Information and Regulatory Affairs, (3150-0104), Attn: Desk ail: oira submission@omb.eop.gov. The NRC may not conduct or (See NUREG-1 022, R.3 for instruction and guidance for completing this form sponsor, and a person is not required to respond to, a collection of information unless the document https:/lwww.nrc.gov/reading-nn/doo-collections/nuregs/staff/sr1022/r3D requesting or requiring the collection displays a currently valid OMB control number.

1. FACILITY NAME

2. DOCKET NUMBER

3. LERNUMBER Susquehanna Steam Electric Station Unit 2 05000-388 NARRATIVE ANALYSIS/SAFETY SIGNIFICANCE YEAR 2020 SEQUENTIAL NUMBER 002 REV NO.

00 All components in the LPCI and Core Spray would have been able to withstand a pressure of 456.2 psig (434.2 psig with water head removed which allows direct comparison toTS value) which was the maximum recorded pressure for instrument PS-B21-2N021 D. Therefore, Core Spray and RHR would have been able to perform its safety function and the condition described herein did not result in a safety system functional failure. Accordingly, this event will not be counted as a safety system functional failure in the Reactor Oversight Process Performance Indicators. There were no actual consequences to the health and safety of the public as a result of this event.

CORRECTIVE ACTIONS The cause of the event is under review and will be provided in a supplement to this LER.

COMPONENT FAILURE INFORMATION Component failure information will be provided in a supplement to this LER.

PREVIOUS OCCURRENCES LER 50-387(388)/2018-005-01, "Condition Prohibited by Technical Specifications Due to Drift of Reactor Pressure Switches", dated December 16, 2019.

LER 50-388/2017-010-01, "Condition Prohibited by Technical Specifications Due to Drift of Reactor Pressure Switches", dated December 16, 2019.

LER 50-388(387)/2015-001-01, "Condition Prohibited by Technical Specifications Due to Drift of Reactor Pressure Steam Dome-Low Switches", dated February 10, 2016.

NRC FORM 366A (08-2020)

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