IR 05000348/2016007

U.S. Nuclear Regulatory Commission Evaluation of Changes, Tests, and Experiments and Permanent Plant Modifications Inspection Report 05000348/2016007 and 05000364/2016007, January 11, 2016, Through February 4, 2016

ML16078A370
Person / Time
Site:	Farley
Issue date:	03/18/2016
From:	Bartley J NRC/RGN-II/DRS/EB1
To:	Gayheart C Southern Nuclear Operating Co
References
IR 2016007
Download: ML16078A370 (14)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION rch 18, 2016

SUBJECT:

FARLEY NUCLEAR PLANT - U.S. NUCLEAR REGUALTORY COMMISSION EVALUATION OF CHANGES, TESTS, AND EXPERIMENTS AND PERMANENT PLANT MODIFICATIONS INSPECTION REPORT 05000348/2016007 AND 05000364/2016007

Dear Mrs. Gayheart:

On February 4, 2016, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at your Farley Nuclear Plant, Units 1 and 2, and discussed the results of this inspection with Mr. R. Hruby, Jr. and other members of your staff. Inspectors documented the results of this inspection in the enclosed inspection report (IR).

NRC inspectors documented one finding of very low safety significance (Green) in this report.

The finding involved a violation of NRC requirements. The NRC is treating this violation as a non-cited violation (NCV) consistent with Section 2.3.2.a of the Enforcement Policy.

If you contest the violation or significance of this NCV, you should provide a response within 30 days of the date of this inspection report, with the basis for your denial, to the U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington DC 20555-0001; with copies to the Regional Administrator, Region II; the Director, Office of Enforcement, U.S.

Nuclear Regulatory Commission, Washington, DC 20555-0001; and the NRC resident inspector at the Farley Nuclear Plant.

In accordance with Title 10 of the Code of Federal Regulations (10 CFR) 2.390, Public inspections, exemptions, requests for withholding of the NRCs "Agency Rules of Practice and Procedure," a copy of this letter, and its Enclosure, will be available electronically for public inspection in the NRC Public Document Room, or from the Publicly Available Records (PARS)

component of NRC's Agencywide Documents Access and Management System (ADAMS); accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).

Sincerely,

/RA/

Jonathan H. Bartley, Chief Engineering Branch 1 Division of Reactor Safety Docket Nos. 50-348 and 50-364 License Nos. NPF-2 and NPF-8

Enclosure:

NRC IR 05000348 and 364/2016007 w/Attachment: Supplementary Information

REGION II==

Docket Nos: 50-348 and 50-364 License Nos: NPF-2 and NPF-8 Report Nos: 05000348/2016007 and 05000364/2016007 Licensee: Southern Nuclear Operating Company, Inc.

Facility: Farley Nuclear Plant, Units 1 and 2 Location: Columbia, AL Dates: January 11, 2016, through February 4, 2016 Inspectors: Robert N. Patterson, Acting Senior Reactor Inspector (Team Leader)

Sandra Herrick, Reactor Inspector Teh-Chiun Su, Reactor Inspector Approved by: Jonathan H. Bartley, Chief Engineering Branch 1 Division of Reactor Safety Enclosure

SUMMARY

Inspection Report (IR) 05000348/2016007 and 05000364/2016007; 1/11/2016 - 2/4/2016;

Farley Nuclear Plant, Units 1 and 2; Evaluations of Changes, Tests, and Experiments and Permanent Plant Modifications This report covers a 2-week onsite inspection by one senior reactor inspector (acting) and two reactor inspectors. One Green non-cited violation (NCV) was identified. The significance of inspection findings is indicated by their color (Green, White, Yellow, Red) using the NRC Inspection Manual Chapter (IMC) 0609, Significance Determination Process, dated April 29, 2015. All violations of NRC requirements are dispositioned in accordance with the NRCs Enforcement Policy, dated February 4, 2015. The NRC's program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG-1649, Reactor Oversight Process, Revision 5, dated February 2014.

Cornerstone: Mitigating Systems

Green.

The NRC identified a non-cited violation (NCV) of 10 CFR Part 50, Appendix B,

Criterion III, Design Control, for failure to verify design assumptions associated with the operation of the atmospheric relief valves (ARVs) following a steam generator tube rupture (SGTR). The licensee failed to verify that all credited methods of ARV operation as specified in procedure FNP-1-EEP-3, Steam Generator Tube Rupture, Rev. 27 could be performed within the FSAR specified time limit of 30 minutes. Upon identification of the issue, the licensee initiated Technical Evaluation 952125 and conducted two simulated scenarios using the two credited means of operating the ARVs following a SGTR. The licensee was able to show that the actions could be performed within the specified time, although the time results were marginal and did not account for operator error or repeatability. This issue has been entered into the licensees corrective action program as CR 10193323.

The performance deficiency was more than minor because it was associated with the Design Control attribute of the Mitigating Systems Cornerstone and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences.

The finding was not greater than green because it affected the design or qualification of a mitigating structure, system, or component (SSC), but the SSC maintained its operability or functionality as documented in CR 10193323. This finding was not assigned a cross-cutting aspect because the issue did not reflect current licensee performance. (Section 1R17.b)

.

REPORT DETAILS

REACTOR SAFETY

Cornerstones: Initiating Events, Mitigating Systems, and Barrier Integrity

1R17 Evaluations of Changes, Tests, Experiments and Permanent Plant Modifications

a. Inspection Scope

Evaluations of Changes, Tests, and Experiments: The team reviewed six safety evaluations performed pursuant to Title 10 of the Code of Federal Regulations (10 CFR) 50.59, Changes, tests, and experiments, to determine if the evaluations were adequate, and that prior NRC approval was obtained as appropriate. The team also reviewed 16 screenings where licensee personnel had determined that a 10 CFR 50.59 evaluation was not necessary. The team reviewed these documents to determine if:

• the changes, tests, or experiments performed were evaluated in accordance with 10 CFR 50.59, and that sufficient documentation existed to confirm that a license amendment was not required

• the safety issues requiring the changes, tests, or experiments were resolved

• the licensee conclusions for evaluations of changes, tests, or experiments were correct and consistent with 10 CFR 50.59

• the design and licensing basis documentation used to support the change was updated to reflect the change The team used, in part, Nuclear Energy Institute (NEI) 96-07, Guidelines for 10 CFR 50.59 Implementation, Rev. 1, to determine acceptability of the completed evaluations and screenings. The NEI document was endorsed by the NRC in Regulatory Guide 1.187, Guidance for Implementation of 10 CFR 50.59, Changes, Tests, and Experiments, dated November 2000.

Permanent Plant Modifications: The team reviewed six permanent plant modifications that had been installed in the plant during the last 3 years. The modifications reviewed are listed below:

• SNC86713 TDAFWP UPS Replacement

• SNC87850 2C EDG Governor Replacement

• SNC459692 NFPA 805: U1 Alternate Air Supply to MDAFWP Control Valve

• SNC524957 Addition of Auto Isolation Valves For RWST TO RWPP

• SNC582588 UNIT 2 RCP SDS Generation III Replacement

• SNC731742 UNIT 1 Low Idle Set-point Change For TDAFWP The modifications were selected based upon risk significance, safety significance, and complexity. The team reviewed the modifications selected to determine if:

• the supporting design and licensing basis documentation was updated

• the changes were in accordance with the specified design requirements

• the procedures and training plans affected by the modification had been adequately updated

• the test documentation, as required by the applicable test programs, had been updated

• post-modification testing adequately verified system operability and/or functionality The team also used applicable industry standards to evaluate acceptability of the modifications and performed walkdowns of accessible portions of the modifications.

Documents reviewed are listed in the Attachment.

b. Findings

Failure to Verify Design Assumptions Associated With the Operation of the Atmospheric Relief Valves (ARVs)

Introduction:

An NRC-identified Green non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, was identified for the licensees failure to verify a design assumption associated with the operation of the atmospheric relief valves (ARVs) following a steam generator tube rupture (SGTR).

Specifically, the licensee failed to verify that a SGTR could be mitigated with the credited methods of ARV operation within the FSAR time limit of 30 minutes.

Description:

During review of modification package SNC459692, NFPA 805: U1 Alternate Air Supply to MDAFWP Control, the team identified that the licensee had not verified that a steam generator tube rupture could be mitigated using the credited methods of operating the ARVs.

FSAR Section 15.4.3.2.1 specified The operator identifies the accident type and terminates break flow to the affected steam generator within 30 min of accident initiation.

The team reviewed licensing and design bases documentation provided by the licensee and concluded that the two credited means for operating the ARVs include motive force provided by the emergency air compressors (EACs), or the local hand wheels. During further review, it was identified that the licensee had never verified that break flow could be terminated via either of these two methods within 30 minutes. Procedure FNP-1-EEP-3, Steam Generator Tube Rupture, Rev. 27, directed operators to use either method to terminate the break flow from the RCS. The licensee had Job Performance Measures (JPMs) for operating the ARVs using the EAC and the local hand wheels; however, these JPMs did not demonstrate that the actions could be accomplished within required time frame. In addition, these actions were not incorporated into the sites time critical operator actions program.

In response to the teams questioning, the licensee initiated corrective actions to further evaluate the sites ability to respond to a SGTR event and provide the team with reasonable assurance of operability. To support operability, the licensee conducted two simulated scenarios using the two credited means of operating the ARVs following a SGTR. The licensee was able to show that the actions could be performed within the specified time. Additionally, on March 3, 2016, the licensee initiated Technical Evaluation 952125 to revise the Time Critical Operator Action Program (NMP-ES-014-001) to update the Operator Response Time Initial Validation Sheet to include actions associated with the EAC and ARV hand wheel operation.

Analysis:

The licensees failure to verify that all credited methods of ARV operation as specified in procedure FNP-1-EEP-3, Steam Generator Tube Rupture, Rev.

27, could be performed within the FSAR time limit of 30 minutes was a performance deficiency. The performance deficiency was more than minor because it was associated with the Design Control attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective in that failure to verify that a SGTR could be terminated within 30 minutes using the credited methods of operating the ARVs adversely affected the capability to mitigate a STGR event. The team used IMC 0609, Att. 4, Initial Characterization of Findings, issued June 19, 2012, for Mitigating Systems, and IMC 0609, App. A, The Significance Determination Process (SDP) for Findings At-Power, issued June 19, 2012, and determined the finding to be of very low safety significance (Green) because the finding was a deficiency affecting the design or qualification of a mitigating structure, system, or component (SSC), and the SSC maintained its operability or functionality. This finding was not assigned a cross-cutting aspect because the issue did not reflect current licensee performance.

Enforcement:

10 CFR Part 50, Appendix B, Criterion III, Design Control, required, in part, that design control measures shall provide for verifying or checking the adequacy of design. Licensee procedure NMP-ES-084, Design Control/Configuration Management Processes, implemented the requirements of Section 3, Design Control in the SNC-1, Quality Assurance Topical Report (QATR). Section 3 of the licensees QATR implemented the requirements of Criterion III and specified that design verification procedures are established and implemented to assure that an appropriate verification method is used, the appropriate design parameters to be verified are chosen, the acceptance criteria are identified, and the verification is satisfactorily accomplished and documented. Contrary to the above, since the operating licenses were issued in December, 1977, for Unit 1 and July, 1981, for Unit 2, the licensee failed to establish design control measures to provide for verifying or checking verify design assumptions associated with the operation of the ARVs following a steam generator tube rupture event. Specifically, the licensee failed to verify that all credited methods of ARV operation as specified in procedure FNP-1-EEP-3, Steam Generator Tube Rupture, Rev. 27, could be performed within the FSAR time limit of 30 minutes. Because this violation was of very low safety significance (Green), and was entered into the licensees corrective action program, this violation is being treated as an NCV, consistent with Section 2.3.2 of the NRC Enforcement Policy. The violation was entered into the licensees corrective action program as CR 10193323. (NCV 05000348/ 2016007-01 and 05000364/2016007-01, Failure to Verify Design Assumptions Associated with the Operation of the Atmospheric Relief Valves)

4OA6 Meetings, Including Exit

On March 15, 2016, the team presented the inspection results to Ms. Gayheart and other members of the licensees staff. The team verified that no proprietary information was retained by the inspectors, or documented in this report.

ATTACHMENT:

SUPPLEMENTARY INFORMATION

KEY POINTS OF CONTACT

Licensee personnel

R. Hruby, Engineering Director

J. Andrews, Maintenance Director

R. Still, RP Superintendent of Operations

K. Baity, Site Design Manager

B. Taylor, Regulatory Affairs Manager

J. McCory, Design Supervisor

J. Collier, Licensing Engineer

J. Simmons, Design Engineer

NRC personnel

S. Sandal, Chief, Division of Reactor Projects

P. Niebaum, Senior Resident Inspector, Division of Reactor Projects

K. Miller, Resident Inspector, Division of Reactor Projects