05000285/LER-2013-014-02, Regarding Unqualified Components Used in Safety System Control Circuit

Regarding Unqualified Components Used in Safety System Control Circuit

ML15029A717
Person / Time
Site:	Fort Calhoun
Issue date:	01/29/2015
From:	Cortopassi L Omaha Public Power District
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
LIC-15-0019 LER 13-014-02
Download: ML15029A717 (4)

LER-2013-014, Regarding Unqualified Components Used in Safety System Control Circuit

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)(B), Loss of Safety Function - Remove Residual Heat 10 CFR 50.73(a)(2)(i) 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 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)(i)(A), Completion of TS Shutdown 10 CFR 50.73(a)(2)(v), Loss of Safety Function
2852013014R02 - NRC Website

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Diiiiii Omaha Public Power District 444 South 16111 Street Mall Omaha, NE 68102-2247 LIC-15-0019 January 29, 2015 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001 Fort Calhoun Station, Unit No.1 Renewed Facility Operating License No. DPR-40 NRC Docket No. 50-285 10 CFR 50.73 Reference

1) Letter from OPPD (L. P. Cortopassi) to NRC (Document Control Desk) dated December 18, 2013 (LIC-13-0178)

2) Letter from OPPD (L. P. Cortopassi) to NRC (Document Control Desk) dated May 2, 2014 (LIC-14-0044)

Subject:

Licensee Event Report 2013-014, Revision 2, for the Fort Calhoun Station Please find attached Licensee Event Report 2013-014, Revision 2. This report is being submitted pursuant to 10 CFR 50.73(a)(2)(i)(B) and 10 CFR 50.73(a)(2)(v)(B). The revision is being submitted to correct the reporting criteria for this event and minor editorial corrections. There are no new commitments being made in this letter.

If you should have any questions, please contact Terrence W. Simpkin, Manager, Site Regulatory Assurance, at (402) 533-6263.

sfd:-

Louis P. Cortopassi Site Vice President and CNO LPC/epm Attachment c:

M. L. Dapas, NRC Regional Administrator, Region IV C. F. Lyon, NRC Senior Project Manager S.M. Schneider, NRC Senior Resident Inspector NRC FORM 366 (02-2014)

NRC FORM 366 (02-2014)

NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION (02-2014)

LICENSEE EVENT REPORT (LER)

(See Page 2 for required number of digits/characters for each block)

APPROVED BY OMB: NO. 3150-0104 EXPIRES: 01/31/2017 Estimated burden per response to comply with this mandatory collection request: 80 hou rs.

Reported lessons learned are incorporated into the licensing process and f ed back t o industry.

Send comments regarding burden estimate to the FOIA, Privacy a nd Information Collections Branch (T-5 F53), U.S. Nuclear Re gulatory Commission, Washington, DC 205 55-0001, or by internet e-mail to Infocollects.Resource@nrc.gov, and to the Desk Officer, Office of Information and Regulatory Affairs, NEOB-10202, (3150-0104), Office of Management and Budget, Washington, DC 20503. If a means 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.

1. FACILITY NAME Fort Calhoun Station

2. DOCKET NUMBER 05000285

3. PAGE 1 OF 3

4. TITLE Unqualified Components used in Safety System Control Circuit

5. EVENT DATE

6. LER NUMBER

7. REPORT DATE

8. OTHER FACILITIES INVOLVED MONTH DAY YEAR YEAR SEQUENTIAL NUMBER REV NO.

MONTH DAY YEAR FACILITY NAME DOCKET NUMBER 05000 10 18 2012 2013 014 - 02 01 29 2015 FACILITY NAME DOCKET NUMBER 05000

9. OPERATING MODE

11. THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR §: (Check all that apply) 5 20.2201(b) 20.2203(a)(3)(i) 50.73(a)(2)(i)(C) 50.73(a)(2)(vii) 20.2201(d) 20.2203(a)(3)(ii) 50.73(a)(2)(ii)(A) 50.73(a)(2)(viii)(A) 20.2203(a)(1) 20.2203(a)(4) 50.73(a)(2)(ii)(B) 50.73(a)(2)(viii)(B) 20.2203(a)(2)(i) 50.36(c)(1)(i)(A) 50.73(a)(2)(iii) 50.73(a)(2)(ix)(A)

10. POWER LEVEL 20.2203(a)(2)(ii) 50.36(c)(1)(ii)(A) 50.73(a)(2)(iv)(A) 50.73(a)(2)(x) 0 20.2203(a)(2)(iii) 50.36(c)(2) 50.73(a)(2)(v)(A) 73.71(a)(4) 20.2203(a)(2)(iv) 50.46(a)(3)(ii) 50.73(a)(2)(v)(B) 73.71(a)(5) 20.2203(a)(2)(v) 50.73(a)(2)(i)(A) 50.73(a)(2)(v)(C)

OTHER 20.2203(a)(2)(vi) 50.73(a)(2)(i)(B) 50.73(a)(2)(v)(D)

Specify in Abstract below or in NRC FORM 366 (02-2014)

BACKGROUND Fort Calhoun Station (FCS) is a two-loop reactor coolant system of Combustion Engineering design.

FCS Technical Specification (TS) Section 2.5 states, in part that, two auxiliary feedwater (AFW) trains shall be OPERABLE when Tcold is above 300°F and with one AFW train inoperable for reasons other than Condition A, inoperable steam supply, restore the AFW train to OPERABLE status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

FCS Updated Safety Analysis Report (USAR) Section 9.4.1 states, in part that, FCS has two safety class auxiliary feed pumps, each capable of meeting system requirements and with diverse power sources; one electric motor driven and the other steam turbine driven.

EVENT DESCRIPTION

In October of 2012, condition report (CR) 2012-15755 identified that an inadvertent closure of the steam driven auxiliary feedwater recirculation valve, FCV-1369, could cause steam driven auxiliary feedwater pump, FW-10, damage. FCV-1369 is not a safety related component (CQE). Although FCV-1369 is a fail-open valve, a postulated closure of this valve coincident with a demand closure of HCV-1107B, Steam Generator RC-2A Auxiliary Feedwater Inlet Valve and HVC-1108B, Steam Generator RC-2B Auxiliary Feedwater Inlet Valve, could result in damage to FW-10 due to dead heading. The station was shutdown in MODE 5 when this condition was discovered. This issue was documented in CR 2013-18752.

A review determined that the components in question, although procured as CQE, were not maintained as safety related components. Although the condition only applies to FW-10, during the last operating cycle the motor-driven AFW pump (FW-6) had been taken out of service for testing.

This report is being submitted pursuant to 10 CFR 50.73(a)(2)(i)(B) and 10 CFR 50.73(a)(2)(v)(B).

CONCLUSION The apparent cause was determined to be that the processes used for design and modification of systems, structures, and components do not use explicit and systematic failure mode analysis when adding, deleting, or modifying safety-related system design allowing credible failure modes to not be considered during design and modification.

CORRECTIVE ACTIONS

Immediate Corrective Actions FCV-1369 has been opened by isolating operating air to the valve to ensure the operability of FW-10 for the current operating cycle.

NRC FORM 366 (02-2014)

Long Term Corrective Actions Appropriate components associated with FCV-1369 will be upgraded or otherwise evaluated to resolve the issue.

To correct the apparent cause of the issue FCS will upgrade design control procedures by implementing the appropriate procedures from Exelon as part of the integration process.

SAFETY SIGNIFICANCE

The failure of FCV-1369 to open could result in overheating of its associated pump (FW-10).

However, with the exception of some maintenance and testing an independent safety related pump (FW-6) was available. In addition, a non-safety related diesel powered AFW pump is available to supply feedwater to the steam generators.

SAFETY SYSTEM FUNCTIONAL FAILURE This does represent a safety system functional failure in accordance with NEI 99-02, revision 7.

PREVIOUS EVENTS LER 2006-002