ML120860447

Davis Besse, Unit 1 - Issuance of Amendment Modification of the Special Visual Inspection Requirement of TS 5.5.8 Steam Generator Program (TAC No. ME6396)

ML120860447
Person / Time
Site:	Davis Besse
Issue date:	04/18/2012
From:	Michael Mahoney Plant Licensing Branch III
To:	Allen B S First Energy Services
Mahoney, M NRR/DORL/LPLIII- 2 415-3867
References
TAC ME6296
Download: ML120860447 (12)

Text

UNITED NUCLEAR REGULATORY WASHINGTON, D.C. 20555-6001 April 18, 2012 Mr. Barry S. Allen Site Vice President FirstEnergy Nuclear Operating Company Davis-Besse Nuclear Power Station Mail Stop A-DB-3080 5501 North State Route 2 Oak Harbor, OH 43449-9760 DAVIS-BESSE NUCLEAR POWER STATION, UNIT NO.1-ISSUANCE OF AMENDMENT RE: MODIFICATION OF THE SPECIAL VISUAL INSPECTION REQUIREMENT OF TECHNICAL SPECIFICATION 5.5.8, "STEAM GENERATOR (SG) PROGRAM" (TAC NO. ME6296)

Dear Mr. Allen:

The U.S. Nuclear Regulatory Commission (the Commission) has issued the enclosed Amendment No. 285 to Facility Operating License No. NPF-3 for the Davis-Besse Nuclear Power Station, Unit No.1 (DBNPS). The amendment revises the technical specifications (TSs) in response to your application dated May 20,2011 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML 11144A289), as supplemented by letter dated February 7,2012 (ADAMS Accession No. ML 12039A199).

This amendment revises the TS 5.5.8.g requirement for a special visual inspection of the secured internal auxiliary feedwater header (AFWH) for the steam generators (SGs) to be condition-based instead of frequency-based.

Specifically, TS 5.5.8.g requires visual inspection of the secured internal AFWH, header to shroud attachment welds, and external header thermal sleeves of the SGs at DBNPS during the third period of each 10-year inservice inspection interval (lSI). With the proposed change, if eddy current inspections (required by TS 5.5.8.d.5) identify any SG peripheral tube to secured internal AFWH gaps less than % inches or there is evidence that the header is degrading or has moved, then the TS 5.5.8.g visual inspections shall be performed on the affected SG.

B. Allen A copy of the Safety Evaluation is also enclosed.

The Notice of Issuance will be included in the Commission's next biweekly Federal Register notice. Docket No. 50-346

Enclosures:

1. Amendment No. 285 to NPF-3 2. Safety Evaluation cc w/encls: Distribution via Listserv UNITED NUCLEAR REGULATORY WASHINGTON, D.C. 20555"()001 FIRSTENERGY NUCLEAR OPERATING COMPANY AND FIRSTENERGY NUCLEAR GENERATION CORP. DOCKET NO. 50-346 DAVIS-BESSE NUCLEAR POWER STATION, UNIT NO.1 AMENDMENT TO FACILITY OPERATING LICENSE NPF-3 Amendment No. 285 Facility Operating License No. NPF-3 The U.S. Nuclear Regulatory Commission (the Commission) has found that: The application for amendment filed by FirstEnergy Nuclear Operating Company (FENOC) et al. (the licensee), dated May 20,2011 as supplemented by letter dated February 7, 2012, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; There IS reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (Ii) that such activities will be conducted in compliance with the Commission's regulations; The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment and paragraph 2.C.(2) of Facility Operating License No. NPF-3 is hereby amended to read as follows:

-2 Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 285, are hereby incorporated in the license. FENOC shall operate the facility in accordance with the Technical Specifications. This license amendment is effective as of its date of issuance and shall be implemented within 30 days of the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION Ja b I. Zim man, Chief PIt Licensing Branch 111-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications and Facility Operating License Date of Issuance: .April 18, 2012 ATTACHMENT TO LICENSE AMENDMENT NO. FACILITY OPERATING LICENSE NO. DOCKET NO. Replace the following pages of the Facility Operating License (FOL) and Appendix A Technical Specifications (TSs) with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change. Remove Insert FOL NPF-3 FOL NPF-3 Page 4 Page 4 TSs TSs 5.5-8 5,5-8

-4 2.C. This license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations in 10 CFR Chapter I: Part 20, Section 30.34 of Part 30, Section 40.41 of Part 40, Sections 50.54 and 50.59 of Part 50, and Section 70.32 of Part 70; and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below: Maximum Power Level FENOC is authorized to operate the facility at steady state reactor core power levels not in excess of 2817 megawatts (thermal).

Prior to attaining the power level, Toledo Edison Company shall comply with the conditions identified in Paragraph (3) (0) below and complete the preoperational tests, startup tests and other items identified in Attachment 2 to this license in the sequence specified.

Attachment 2 is an integral part of this license. Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 285, are hereby incorporated in the license. FENOC shall operate the facility in accordance with the Technical Specifications. Additional Conditions The matters specified in the following conditions shall be completed to the satisfaction of the Commission within the stated time periods following the issuance of the license or within the operational restrictions indicated.

The removal of these conditions shall be made by an amendment to the license supported by a favorable evaluation by the Commission: FENOC shall not operate the reactor in operational Modes 1 and 2 with less than three reactor coolant pumps in operation. (b) Deleted per Amendment 6 (c) Deleted per Amendment 5 Amendment No. 285 Programs and Manuals 5.5 Programs and Manuals 5.5.8 Steam Generator (SG) Program (continued) Special visual inspections:

If the inspections required by TS 5.5.8.d.5 identify any peripheral tube to secured internal auxiliary feedwater header gaps less than 1/4 inch, or there is evidence that the header is degrading or has moved, visual inspections of the secured internal auxiliary feedwater header, header to shroud attachment welds, and external header thermal sleeves shall be performed on the affected SG through the auxiliary feedwater injection penetrations.

5.5.9 Secondary

Water Chemistry Program This program provides controls for monitoring secondary water chemistry to inhibit SG tube degradation.

The program shall include: Identification of a sampling schedule for the critical variables and control pOints for these variables; Identification of the procedures used to measure the values of the critical variables; Identification of process sampling points; Procedures for the recording and management of data; Procedures defining corrective actions for all off control point chemistry conditions; and A procedure identifying the authority responsible for the interpretation of the data and the sequence and timing of administrative events, which is required to initiate corrective action. 5.5.10 Ventilation Filter Testing Program (VFTP) A program shall be established to implement the following required testing of safety related filter ventilation systems in accordance with Regulatory Guide 1.52, Revision 2, ANSIIASME N510-1980, and ASTM D 3803-1989. Demonstrate for each of the safety related systems that an inplace test of the high efficiency particulate air (HEPA) filters shows a penetration and system bypass < 1.0% when tested in accordance with Regulatory Guide 1.52, Revision 2, and ANSIIASME N510-1980 at the system flowrate specified below. Safety Related Ventilation Flowrate (cfm) Station Emergency Ventilation System (EVS) <:: 7200 and s 8800 Control Room Emergency Ventilation System <:: 2970 and s 3630 (CREVS) 5.5-8 Amendment 285 UNITED NUCLEAR REGULATORY WASHINGTON, D.C. 20555*0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 285 TO FACILITY OPERATING LICENSE NO. NPF-3 FIRSTENERGY NUCLEAR OPERATING COMPANY FIRSTENERGY NUCLEAR GENERATION CORP. DAVIS-BESSE NUCLEAR POWER STATION, UNIT NO.1 DOCKET NO. 50-346

1.0 INTRODUCTION

By letter to the Nuclear Regulatory Commission (NRC, the Commission) dated May 20, 2011 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML 11144A289), as supplemented by letter dated February 7,2012 (ADAMS Accession No. ML 12039A 199) FirstEnergy Nuclear Operating Company, et al. (the licensee) requested changes to the Technical Specifications (TSs) for the Davis-Besse Nuclear Power Station, Unit No. 1 (DBNPS). The proposed change would revise TS 5.5.8.g to perform the special visual inspections based on a condition rather than a specific frequency.

Specifically, TS 5.5.8.g requires visual inspection of the secured internal auxiliary feedwater header (AFWH), header to shroud attachment welds, and external header thermal sleeves of the steam generators (SGs) at DBNPS during the third period of each 10 year inservice inspection interval (lSI). With the proposed change, if eddy current inspections (required by TS 5.5.8.d.5) identify any SG peripheral tube to secured internal AFWH gap less than % inches or there is evidence that the header is degrading or has moved, then the TS 5.5.8.g visual inspections shall be performed on the affected SG. The February 7, 2012 supplement provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the NRC staff's initial proposed finding of no significant hazards consideration determination.

2.0 BACKGROUND

DBNPS has two Babcock and Wilcox once-through SGs, designated as 1-B and 2-A Each SG has approximately 15,500 Alloy 600 tubes in the mill-annealed condition.

The tubes have a nominal outside diameter of 0.625 inches and a nominal wall-thickness of 0.037 inches. Both SG 1-B and 2-A contain tubes with sleeves and tubes with shop re-rolls.

-2 During a refueling outage in 1982, eddy current inspections revealed indications of tube damage in some of the SG peripheral tubes. Follow-up visual inspections in the secondary side of the SGs confirmed that the internal AFWH in both SGs had been dislodged and severely deformed, resulting in damage to some of the peripheral tubes in both SGs. As a result of the extensive damage to the internal AFWHs, the licensee secured and abandoned-in-place the internal AFWH in each SG. On September 30, 1983 (ADAMS Accession No. ML021160474), the NRC issued a license amendment that required, in part, the licensee to perform a special visual inspection on the secured internal AFWH, the shroud attachment welds, and the external header thermal sleeves.

3.0 REGULATORY EVALUATION

The tubes within a SG function as an integral part of the reactor coolant pressure boundary (RCPB) and, in addition, serve to isolate radiological fission products in the primary coolant from the secondary coolant and the environment.

For the purposes of this safety evaluation, tube integrity means that the tubes are capable of performing these functions in accordance with the plant design and licensing basis. Title 10 of the Code of Federal Regulations (10 CFR) establishes the fundamental regulatory requirements with respect to the integrity of the SG tubing. Specifically, the General Design Criteria (GDC) in Appendix A to 10 CFR Part 50 states that the RCPB shall have "an extremely low probability of abnormalleakage

...and of gross rupture" (GDC-14), "shall be designed with sufficient margin" (GDC-15 and -31), shall be of "the highest quality standards practical" (GDC-30), and shall be designed to permit "periodic inspection and testing ... to assess ... structural and leaktight integrity" (GDC-32).

To this end, 10 CFR 50.55a specifies that components which are part of the RCPB must meet the requirements for Class 1 components in Section III of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code). Requirements of 10 CFR 50.55a state, in part, that throughout the service life of a pressurized water reactor (PWR) facility like DBNPS, ASME Code Class 1 components meet the requirements, including design and access provisions and pre-service examination requirements, in Section XI, "Rules for Inservice Inspection of Nuclear Power Plant Components," of the ASME Code.Section XI requirements pertaining to inservice inspection of SG tubing are augmented by additional SG tube surveillance requirements in the TSs. As part of the plant licensing basis, applicants for PWR licenses are required to analyze the consequences of postulated design-basis accidents such as an SG tube rupture and main steamline break. These analyses consider the primary-to-secondary leakage through the tubing which may occur during these events and must show that the offsite radiological consequences do not exceed the applicable limits of the 10 CFR Part 100.11 guidelines for offsite doses (or 10 CFR 50.67, as appropriate), GDC-19 criteria for control room operator doses, or some fraction thereof as appropriate to the accident, or the NRC-approved licensing basis. It states in 10 CFR 50.36(b), in part, that the TSs will be derived from the analyses and evaluation included in the safety analysis report, and amendments thereto, submitted pursuant to 10 CFR 50.34. It also states in 10 CFR 50.36(c)(5) that TSs shall contain administrative controls "relating to organization and management, procedures, recordkeeping, review and audit, and reporting necessary to assure operation of the facility in a safe manner." As discussed below, TS 5.5.8 establishes a SG program to ensure that tube integrity is maintained to assure safe operation of DBNPS. TS 5.5.8 for DBNPS requires that a SG Program be established and implemented to ensure that SG tube integrity is maintained.

The DBNPS TSs are modeled after Technical Specification Task Force Traveler 449 (TSTF-449), "Steam Generator Tube Integrity," Revision 4. Tube integrity is maintained by meeting specified performance criteria for structural and leakage integrity consistent with the plant design and licensing bases. TS 5.5.8 requires a condition monitoring assessment be performed during each outage during which the SG tubes are inspected, plugged or repaired to confirm that the performance criteria are being met. TS 5.5.8 also includes provisions regarding the scope, frequency, and methods of SG tube inspections.

4.0 TECHNICAL

EVALUATION The licensee proposed to change the TS 5.5.8.g requirement for a special visual inspection of the secured internal AFWH to be condition-based instead of frequency-based.

The licensee is currently required to perform the eddy current inspections required by TS 5.5.8.d.5 on the peripheral SG tubes during every refueling outage. These inspections are intended to detect degradation of inservice tubing due to interaction with the secured internal AFWH. The licensee has proposed that if the eddy current inspections identify a gap of less than X inch between any SG peripheral tube and the secured internal AFWH, or there is evidence that the header is degrading or has moved, then the visual inspections of the secured internal AFWH, the header-to-shroud attachment welds, and the external header thermal sleeves shall be performed on the affected SG, through the auxiliary freshwater injection penetrations.

This would replace the current requirement to do these inspections during the third period of each 10-year lSI. The licensee stated that modifying this requirement is predicated on historically acceptable visual and eddy current inspection results for both SGs and an evaluation of changes to operational conditions.

After performing a baseline special visual inspection in 1983, the licensee performed the special visual inspection of the internal AFWH during the next two refueling outages and during the third period of each 10-year lSI interval.

The special visual inspection program included inspecting the internal AFWH and the internal AFWH-to-shroud welds for degradation, and inspecting the external header thermal sleeves for indications of cracking.

In addition to these visual inspections, the licensee has also performed eddy current inspections on a minimum of 150 peripheral tubes in each SG, during every refueling outage since the external AFWHs were installed.

The eddy current inspection technique used by the licensee was developed and qualified with the specific purpose of detecting if the internal AFWH had moved to within X inch of any of the inspected peripheral tubes. To date, several special visual inspections have been performed on SGs 1-B and 2-A. Four small linear indications were found on one of the thermal sleeves in SG 1-B in 1984. The licensee reported that the indications appeared to be a result of the visual inspection process (Le., disassembly and reassembly).

The sleeve was reworked and no other indications have been reported as a result of the special visual inspections in SG 1-8. For SG 1-8, the last special visual inspection was performed in 2010, and no other special visual inspections are scheduled before the planned replacement of the SG in 2014. No indications have ever been reported as a result of the special visual inspections in SG 2-A The last special visual inspection of SG 2-A was performed in 1998 and only one special visual inspection is scheduled before the planned replacement of the SG in 2014. The eddy current inspections have been performed on the peripheral tubes of both SGs during every refueling outage since 1983, and no indications of internal AFWH movement have been reported.

In support of a turbine header pressure modification in 2007, SGs 1-8 and 2-A were evaluated by the licensee for the expected changes in operational conditions.

The licensee's evaluations concluded that the increase in header pressure from 870 pounds per square inch gauge (psig) to 880 psig resulted in a negligible effect on flow-induced vibration and would not significantly impact the SGs. In support of a reactor power uprate in 2008, SGs 1-8 and 2-A were further evaluated for the expected changes in operational conditions.

The licensee's evaluations concluded that the approximate

1.6 percent

reactor power increase resulted in a negligible effect on flow-induced vibration and would not significantly impact the SGs. No other modifications with potential to affect the SGs, the secured internal AFWHs, or the external header thermal sleeves, were identified by the licensee.

The licensee performed special visual inspections of steam generator 1-8 in 2010, and eddy current inspections of SGs 1-8 and 2-A in 2010, and concluded that the acceptable findings of these inspections validated both the header pressure change and power uprate evaluations.

The NRC staff concludes that the historical visual and eddy current inspection results demonstrate that the internal AFWH is adequately secured. In addition, the NRC staff concludes that the requirement to inspect the peripheral SG tubes by eddy current techniques and the proposed requirement to perform visual inspections of the internal AFWH, the internal AFWH-to-shroud welds, and the external header thermal sleeves if there is evidence of less than a }'4 inch gap between any SG peripheral tube and the secured internal AFWH, or there is evidence that the header is degrading or has moved, provides reasonable assurance that the safety functions of the SG will not be compromised.

5.0 STATE CONSULTATION

In accordance with the Commission's regulations, the Ohio State official was notified of the proposed issuance of the amendment.

The State official had no comments.

6.0 ENVIRONMENTAL

CONSIDERATION This amendment changes a requirement with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20. The NRC staff has determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluent that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure.

The Commission has previously issued a proposed finding that the amendment involves no significant hazards consideration and there has been no public comment on such finding (76 FR 58306; September 20,2011).

Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9).

Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendment.

7.0 CONCLUSION

Based on the above considerations, the NRC staff finds the licensee's proposed change to switch from frequency-based special visual inspections to condition-based special visual inspections in TS 5.5.8.g acceptable.

The NRC staff has concluded, based on the considerations discussed above, that: (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public. Principal Contributor:

AJohnson, NRR Date: April 18, 2012 B. Allen A copy of the Safety Evaluation is also enclosed.

The Notice of Issuance will be included in the Commission's next biweekly Federal Register notice. Sincerely, I RAJ Michael Mahoney, Project Manager Plant Licensing Branch 111-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-346

Enclosures:

1. Amendment No. 285 to NPF-3 2. Safety Evaluation cc w/encls: Distribution via Listserv DISTRIBUTION:

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