L-14-297, Reply to Request for Additional Information for the Review of the Davis-Besse Nuclear Power Station, Unit No. 1, License Renewal Application (TAC No. ME4640) and License Renewal Application Amendment No. 53

Reply to Request for Additional Information for the Review of the Davis-Besse Nuclear Power Station, Unit No. 1, License Renewal Application (TAC No. ME4640) and License Renewal Application Amendment No. 53

ML14259A067
Person / Time
Site:	Davis Besse
Issue date:	09/16/2014
From:	Lieb R FirstEnergy Nuclear Operating Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
L-14-297, TAC ME4640
Download: ML14259A067 (17)

Text

FENOC' ~

5501 North State Route 2 Oak Harbor, Ohio 43449 FirstEnergy Nuclear Operating Company Raymond A Lieb 419-321-7676 Vice President, Nuclear Fax: 419-321-7582 September 16, 2014 L-14-297 10 CFR 54 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001

SUBJECT:

Davis-Besse Nuclear Power Station, Unit No. 1 Docket No. 50-346, License Number NPF-3 Reply to Request for Additional Information for the Review of the Davis-Besse Nuclear Power Station, Unit No. 1, License Renewal Application (TAC No. ME4640) and License Renewal Application Amendment No. 53 By letter dated August 27, 2010 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML102450565), FirstEnergy Nuclear Operating Company (FENOC) submitted an application pursuant to Title 10 of the Code of Federal Regulations, Part 54 for renewal of Operating License NPF-3 for the Davis-Besse Nuclear Power Station, Unit No. 1 (Davis-Besse). By letter dated August 19, 2014 (ML14218A145), the Nuclear Regulatory Commission (NRC) requested additional information to complete its review of the License Renewal Application (LRA).

The Attachment provides the FENOC reply to the NRC request for additional information. The NRC request is shown in bold text followed by the FENOC response.

The Enclosure provides Amendment No. 53 to the Davis-Besse LRA.

Davis-Besse Nuclear Power Station, Unit No. 1 L-14-297 Page 2 There are no regulatory commitments contained in this letter. If there are any questions or if additional information is required, please contact Mr. Clifford I. Custer, Fleet License Renewal Project Manager, at 724-682-7139.

I declare under J?enalty of perjury that the foregoing is true and correct. Executed on September 10, 2014.

s~a$

Raymond A. Lieb

Attachment:

Reply to Requests for Additional Information for the Review of the Davis-Besse Nuclear Power Station, Unit No. 1 (Davis-Besse), License Renewal Application (LRA), Sections 2.5 and 4.3.2

Enclosure:

Amendment No. 53 to the Davis-Besse License Renewal Application cc: NRC DLR Project Manager NRC Region Ill Administrator cc: w/o Attachment or Enclosure NRC DLR Director NRR DORL Project Manager NRC Resident Inspector Utility Radiological Safety Board

Attachment L-14-297 Reply to Requests for Additional Information for the Review of the Davis-Besse Nuclear Power Station, Unit No. 1 (Davis-Besse),

License Renewal Application (LRA),

Sections 2.5 and 4.3.2 Page 1 of 6 Section 2.5 Question RAI 2.5.6.2a

Background:

In the license renewal application (LRA) Amendment No. 50 - Annual Update, the applicant provided a revision to LRA Section 2.5.6.2, "Station blackout Recovery Path Evaluation Boundaries" and Figure 2.5-1, "Davis-Besse Station Blackout Recovery Path," to add a new breaker 81-B-65 to the switchyard components that are in-scope of license renewal. The Davis-Besse Updated Safety Analysis Report (USAR) Section 8.3.1.1.4.2, "Alternate AC Source - Station Blackout Diesel Generator," describes a Station Blackout Diesel Generator (SBODG), which is capable of supplying power to either of the Station's essential 4.16kV buses (D1 or C1) through nonessential Bus D2 for coping with a Station Blackout (SBO) event. USAR Figure 8.3-1, which depicts Davis-Besse alternating current electrical system one-line diagram, shows the SBODG connected to Bus D2 through Bus D3. Code of Federal Regulations Part 10 (10 CFR) 54.4(a)(3) requires that all systems, structures, and components (SSCs) relied on in safety analyses or plant evaluations to perform a function that demonstrates compliance with NRC regulations for SBO (10 CFR 50.63) be included within the scope of license renewal. LRA Table 2.2-1, "License Renewal Scoping Results for Mechanical Systems," shows the SBODG system in-scope of license renewal.

Issue:

SBODG is not mentioned in LRA Section 2.5, "Scoping and Screening Results:

Electrical and Instrumentation and Controls Systems," and Section 3.6, "Aging Management of Electrical and Instrumentation and Control Systems." In addition, Figure 2.5-1 does not show the SBODG.

Request:

Provide updated LRA Section 2.5 including Figure 2.5-1 and Section 3.6 that include the scoping and screening and aging management review of SBODG system electrical components in accordance with 10 CFR 54.4(a)(3).

Attachment L-14-297 Page 2of6 RESPONSE RAI 2.5.6.2a Figure 2.5-1 is revised to depict the onsite power recovery equipment (emergency diesel generators and the station blackout diesel generator).

LRA Section 2.3.3.12, "Emergency Diesel Generators System," describes the onsite power system meeting the requirements under 10 CFR 54.4(a)(1) (safety-related systems) and LRA Section 2.3.3.30, "Station Blackout Diesel Generator System," for the equipment that is required to cope with an SBO (e.g., alternate ac power sources) meeting the requirements under 10 CFR 54.4(a)(3). Figure 2.5-1 continues to display the plant system portion of the offsite power system that is used to connect the plant to the offsite power source meeting the requirements under 10 CFR 54.4(a)(3). The electrical distribution equipment out to the first circuit breaker with the offsite distribution system (i.e., equipment in the switchyard) is shown.

LRA Section 2.5.6.2 is revised to state that the cable and connectors from the emergency and SBO diesel generators provide connection to the onsite power system, meeting the requirements of 10 CFR 54.4.

As stated in LRA Section 2.5.6.1, "System Evaluation Boundaries," electrical and instrumentation and control (l&C) component types within the boundaries of in-scope mechanical systems are also included within the electrical and l&C evaluation boundaries. Adding the emergency diesel generators and station blackout diesel generator to Figure 2.5-1 did not alter the aging management review since no new component types were added. The electrical cables and connectors continue to be in scope for license renewal and managed for age-related degradation.

LRA Section 3.6 provides the results of the aging management reviews for the electrical components and commodities. The system electrical components for the onsite power recovery equipment (i.e., emergency diesel generators and the station blackout diesel generator) are already included in the commodity groups in this section. Therefore no changes are required to LRA Section 3.6.

See the Enclosure to this letter for the revision to the Davis-Besse LRA.

Section 4.3.2 Question RAI 4.3.2.2.6.1-1 (LRA Update follow-up)

Background:

On June 23, 2014, the applicant submitted an LRA update to comply with the LRA update requirements in 10 CFR 54.21 (b). As part of this submittal, the applicant

Attachment L-14-297 Page 3of6 identified that the original once-through steam generators (OTSGs) in the Davis-Besse Nuclear Plant were replaced in the Cycle 18 (Spring 2014) refueling outage. Based on this plant modification, the applicant amended LRA Section 4.3.2.2.6 to propose changes to the metal fatigue time-limited aging analyses (TLAA) bases for the OTSGs at the facility. The applicant also amended LRA AMR Table 3.1.2-4 to include updated AMR items that credit a TLAA as the basis for managing fatigue-induced cracking (i.e., "cracking - fatigue") in the following replacement OTSG components:

• pressure boundary bolts

• primary manways and inspection opening covers

• primary side tubes

• primary side tube plugs

• primary side upper and lower heads

• primary side inlet and outlet nozzles

• primary side upper and lower tubesheets

• primary side tube-to-tubesheet welds

• secondary side auxiliary feedwater headers, risers, and nozzles

• secondary side shrouds and shroud support rings and lugs

• secondary side manways and handhole covers

• secondary side main feedwater headers, risers, support plates, and gussets

• secondary side main feedwater nozzles, main feedwater nozzle thermal sleeves, and auxiliary feedwater nozzle thermal sleeves

• secondary side steam outlet nozzles, vent nozzles, drain nozzles, and level sensing nozzles

• secondary side shells

• secondary side tube support plates

• secondary side tube support plate spacers

• secondary side tube support rods (tie rods)

• base support stools and base support platforms Issue:

The amended version of LRA Section 4.3.2.2.6.1 in the letter of June 23, 2014, states that cumulative usage factors (CUFs) were calculated for the limiting primary and secondary side steam generator locations. The amended TLAA basis

Attachment L-14-297 Page 4of6 for replacement OTSG components in LRA Section 4.3.2.2.6.1 does not: (1) reflect that these replacement OTSG components are the limiting steam generator locations that were analyzed in accordance with an updated ASME Code Section Ill fatigue analysis (i.e., CUF analysis) for the current licensing basis, or (b) identify the basis for accepting the fatigue analysis for each of these components in accordance with 10 CFR 54.21(c)(1)(i), (ii), or (iii).

Request:

1. Identify all replacement OTSG components that were required to be analyzed in accordance with an ASME Code Section metal fatigue analysis (i.e., CUF analysis).

2. For each replacement OTSG component that has been analyzed in accordance with an updated CUF analysis, provide a comparison of the CUF analysis for the component to the six criteria for defining a TLAA in 10 CFR 54.3(a) and justify why the updated CUF analysis for the component would not need to be identified as a TLAA in accordance with the requirement in 10 CFR 54.21(c)(1).

3. For each replacement OTSG component that was analyzed in accordance with a metal fatigue analysis conforming to the definition of a TLAA in 10 CFR 54.3(a), provide justification for accepting the metal fatigue analysis in accordance with the requirements in 10 CFR 54.21(c)(1)(i), (ii), or (iii).

RESPONSE RAI 4.3.2.2.6.1-1 (LRA Update follow-up)

1. Replacement steam generator components that credit a TLAA as the basis for managing fatigue cracking in LRA Table 3.1.2-4, "Aging management Review Results - Steam Generators," were analyzed for metal fatigue in accordance with ASME Code Section Ill, 2001 Edition with 2003 Addenda. The fatigue analysis of the replacement steam generators is evaluated as a TLAA in LRA Section 4.3.2.2.6.1, "OTSGs Fatigue."

For clarification, LRA Section 4.3.2.2.6.3, previously revised by letter dated June 23, 2014(ML141758381 ), is not needed as a separate section and remains as "Not used." Unlike the original steam generators, where by modification the auxiliary feedwater header was relocated to the outside of the steam generator and the modification was evaluated separately, the replacement steam generator auxiliary feedwater header is also located on the outside of the steam generator but is part of the original design, and therefore the header is included with the evaluation in LRA Section 4.3.2.2.6.1.

2. The CUF analysis for the subject steam generator components meet the six criteria for defining a TLAA in accordance with 10 CFR 54.3(a). As provided in LRA Section

Attachment L-14-297 Page 5of6 4.3.2.2.6.1, the effects of fatigue on the steam generators will be managed for the period of extended operation by the Fatigue Monitoring Program in accordance with 10 CFR 54.21 (c)(1 )(iii).

3. The cumulative usage factors for the steam generators components were calculated based on design transients, and are all less than 1.0. The design transients used in the fatigue analyses for the subject steam generator components are included in LRA Table 4.3-1, "60 year Projected Cycles." As provided in LRA Section 4.3.2.2.6.1, the number of occurrences of design transients is tracked by the Fatigue Monitoring Program to ensure that action is taken before the design cycles are reached. As such, the effects of fatigue on the steam generators will be managed for the period of extended operation by the Fatigue Monitoring Program in accordance with 10 CFR 54.21 (c)(1 )(iii).

Question RAI 4.3.2.2.6.4-1 (LRA Update follow-up)

Background:

In the LRA update letter dated June 23, 2014, the applicant identified that the OTSGs in the plant design were replaced in the Spring 2014 refueling outage and that the previous flow-induced vibration analysis for original OTSG tube and tube stabilizers did not apply to these replacement OTSGs. Therefore, in the letter of June 23, 2013, the applicant proposed to delete LRA Section 4.3.2.2.6.4 from the scope of the LRA.

Issue:

It is not evident to the staff why the tubes and or other components in the replacement OTSG would not have been required to be analyzed with a flow-induced vibration analysis, similar to the manner that the tubes and tube stabilizers in the original OTSGs were analyzed for flow-induced vibrations, or why such a flow-induced vibration analysis would not need to be identified as a TLAA for the replacement OTSGs or specific subcomponents in the replacement OTSGs.

Request:

Clarify whether the design code or codes for the replacement OTSGs, or specific components in the replacement OTSGs, required performance of a flow-induced vibration analysis for the components. If the design code for the replacement OTSGs, or specific components in the replacement OTSGs, did require performance of a flow-induced vibration analysis, justify why the applicable flaw-induced vibration analysis would not need to be identified as a TLAA, when

Attachment L-14-297 Page 6of6 compared to the six criteria in 10 CFR 54.3(a) for defining a specific analysis as a TLAA.

RESPONSE RAI 4.3.2.2.6.4-1 (LRA Update follow-up)

The Davis-Besse design specification for the replacement steam generators required an analysis for flow-induced and turbulence-induced vibration throughout the tube bundle, including tube plugs and stabilizers, to show that fatigue failures, excess tube fretting and tube wear, or wear of other steam generator internals, will not occur.

The replacement steam generators were qualified for 40-years of service and were installed in the spring of 2014. The steam generators will experience approximately 23 years of operation by the end of the period of extended operation. Since the in-service time of the steam generators will not exceed their qualified life of 40-years, the flow-induced and turbulence-induced vibration analysis does not meet the six criteria in 10 CFR 54.3(a). Therefore, the analysis is not a TLAA requiring evaluation in accordance with 10 CFR 54.21 (c).

Enclosure Davis-Besse Nuclear Power Station, Unit No. 1 (Davis-Besse)

Letter L-14-297 Amendment No. 53 to the Davis-Besse License Renewal Application Page 1 of 9 License Renewal Application Sections Affected Section 2.5.6.2 Figure 2.5-1 Table 3.1.2-4 Table 3.3.2-14 The Enclosure identifies the change to the License Renewal Application (LRA) by Affected LRA Section, LRA Page No., and Affected Paragraph and Sentence. The count for the affected paragraph, sentence, bullet, etc. starts at the beginning of the affected Section or at the top of the affected page, as appropriate. Below each section the reason for the change is identified, and the sentence affected is printed in italics with deleted text lined out and added text underlined.

Enclosure L-14-297 Page 2 of 9 Affected LRA Section LRA Page No. Affected Paragraph and Sentence 2.5.6.2 2.5-7 &2.5-8 Paragraphs 1 and 2 Based on the response to request for additional information (RAI) 2.5.6.2a, and the depiction of the onsite power recovery equipment (emergency diesel generators and station blackout diesel generator) on Figure 2.5-1, "Davis-Besse Station Blackout Recovery Path," paragraphs 1 and 2 of LRA Section 2.5.6.2, "Station Blackout Recovery Path Evaluation Boundaries," are replaced in their entirety to read as follows:

The License Renewal Rule, 10 CFR 54.4(a)(3), requires that plant svstems, structures. and components relied on for compliance with the NRG regulation on station blackout. 10 CFR 50. 63. be included in the scope of license renewal. The cable and connectors from the Emergency Diesel Generators provide connection to the onsite power system. meeting the requirements under 10 CFR 54.4(a)(1)

(safety-related systems). The Station Blackout Diesel Generator System cable and connectors are part of the equipment that is required to cope with an SBO (e.g., alternate ac power sources), meeting the requirements under 10 CFR 54.4(a)(3).

The following paragraphs describe the station blackout license renewal offsite power recoverv paths for Davis-Besse. USAR Sections 8.1.1 and 8.2 provide a detailed description of the offsite power system and offsite power pathways for Davis-Besse. USAR Figure 8.2-2 provides a simplified single-line diagram showing the switchyard configuration. USAR Figure 8.3-1 provides a single-line diagram of the AC Electrical System.

Enclosure L-14-297 Page 3of9 Affected LRA Section LRA Page No. Affected Paragraph and Sentence Figure 2.5-1 2.5-9 Entire Figure Based on the response to RAI 2.5.6.2a, LRA Figure 2.5-1, "Davis-Besse Station Blackout Recovery Path," is replaced in its entirety as follows:

[See LRA Figure 2.5-1 on page 4 of 9]

Enclosure L-14-297 Page 4 of 9 Figure 2.5-1 Davis-Besse Station Blackout Recovery Path 4.16KV ESSENlIAL BUS 'DJ' 1

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Enclosure L-14-297 Page 5 of 9 Affected LRA Section LRA Page No. Affected Paragraph and Sentence Table 3.1.2-4 3.1-168 & 169 Rows 27 "Notes" 3.1-178 Row 78 - "Aging Management Program" Errata: During review of the LRA Annual Update, it was identified that the Notes for Rows 27-31 were listed incorrectly, and that an incorrect Aging Management Program was listed for Row 78 in LRA Table 3.1.2-4, "Aging Management Review Results - Steam Generators." These rows, previously revised by FENOC letter dated June 23, 2014(ML141758381), are revised to read as follows:

Table 3.1.2-4 Aging Management Review Results - Steam Generators NU REG-Aging Effect Aging Row Component Intended 1801, Table 1 Material Environment Requiring Management Notes No. Type Function(s) Volume 2 Item Management Program Item Borated Primary Side; Pressure Nickel reactor Cracking -

c 27 Alloy TLM IV.02-15 3.1.1-06 .()4()4 Tube Plug boundary coolant Fatigue 690TT 0101 (Internal) c

{)4.().1, Borated Nickel ()4()2, Primary Side; Pressure reactor Cracking - Flaw lnservice 28 Alloy IV.C2-26 3.1.1-62 {)400 Tube Plug boundary coolant Growth Inspection 690TT 0101 (Internal) 0102 0103 Borated Nickel Cracking - A Primary Side; Pressure reactor PWR Water 29 Alloy PWSCC, IV.02-12 3.1.1-73 .()4()4 Tube Plug boundary coolant Chemistry 690 TT SCC/IGA 0101 (Internal)

Enclosure L-14-297 Page 6 of 9 Table 3.1.2-4 Aging Management Review Results - Steam Generators NU REG-Aging Effect Aging Row Component Intended 1801, Table 1 Material Environment Requiring Management Notes No. Type Function(s) Volume 2 Item Management Program Item Borated Nickel Cracking - A Primary Side; Pressure reactor Steam Generator 30 Alloy PWSCC, IV.02-12 3.1.1-73 {)4(#

Tube Plug boundary coolant Tu be Integrity 690TT SCC/IGA 0101 (Internal)

Borated Primary Side; Pressure Nickel reactor PWR Water c

31 Alloy Loss of Material IV.C2-15 3.1.1-83 {)4(#

Tube Plug boundary coolant Chemistry 690 TT 0101 (Internal)

Secondary Side; MFW Nozzle, MFW lnsor.tico Nozzle Pressure Nickel Treated water l-nspoction A 78 Thermal Loss of Material Vlll.B1-1 3.4.1-37 boundary Alloy 690 (Internal) One-Time 0101 Sleeve and Inspection AFW Nozzle Thermal Sleeve

Enclosure L-14-297 Page 7 of 9 Affected LRA Section LRA Page No. Affected Paragraph and Sentence Table 3.3.1 3.3-88 Row 3.3.1-58, "Discussion" column Errata: Based on the correction to Row 119 (see Enclosure page 9 of 9) of LRA Table 3.3.2-14, "Aging Management Review Results for the Fire Protection System," LRA Table 3.3.1, "Summary of Aging Management Programs for Auxiliary Systems Evaluated in Chapter VII of NUREG-1801," is revised to read as follows:

Table 3.3.1 Summary of Aging Management Programs for Auxiliary Systems Evaluated in Chapter VII of NUREG-1801 Further Item Aging Aging Management Component/Commodity Evaluation Discussion Number Effect/Mechanism Programs Recommended 3.3.1-58 Steel external surfaces exposed Loss of material due External Surfaces No Consistent with NUREG-1801.

to air - indoor uncontrolled to general corrosion Monitoring Loss of material in steel (including (external), air - outdoor gray cast iron) external surfaces (external), and condensation that are exposed to air-indoor (external) uncontrolled (external), air-outdoor (external), and condensation (external) is managed by the External Surfaces Monitoring Programi.

exceg_t for the external surfaces of the fire water storage tank

[DB-T81 l 1 which is managed b'{

the Fire Water Program.

This item is also applied to steel internal surfaces that are exposed to air-indoor uncontrolled (internal) or air-outdoor (internal)

Enclosure L-14-297 Page 8 of 9 Table 3.3.1 Summary of Aging Management Programs for Auxiliary Systems Evaluated in Chapter VII of NUREG-1801 Further Item Aging Aging Management Component/Commodity Evaluation Discussion Number Effect/Mechanism Programs Recommended where it was determined that the internal environment is the same as the external environment.

Enclosure L-14-297 Page 9 of 9 Affected LRA Section LRA Page No. Affected Paragraph and Sentence Table 3.3.2-14 3.3-328 Row 119 - "Aging Management Program",

"NUREG-1801, Volume 2 Item", "Table 1 Item", and "Notes" Errata: It was identified that, contrary to NRC License Renewal Interim Staff Guidance LR-ISG-2012-02, "Aging Management of Internal Surfaces, Fire Water Systems, Atmospheric Storage Tanks, and Corrosion Under Insulation," the Fire Water Program was not credited for managing the external surfaces of the fire water storage tank. Row 119 of LRA Table 3.3.2-14, "Aging Management Review Results for the Fire Protection System,"

previously revised by FENOC letter dated September 16, 2011 (ML11264A059}, is revised to credit the Fire Water Program and align the row information with the guidance in the ISG. Note that row 118 of Table 3.3.2-14, previously revised by FENOC letter dated February 19, 2014(ML14055A067), remains as "Not Used."

Row 119 of LRA Table 3.3.2-14 is revised to read as follows:

Table 3.3.2-14 Aging Management Review Results - Fire Protection System NU REG-Aging Effect Aging Row Component Intended 1801, Table 1 Material Environment Requiring Management Notes No. Type Function{s) Volume 2 Item Management Program Item Tank- Fire Ex:temat £(;1ffaees W:-1--9 3.3.1 58 Water Storage Pressure Air-outdoor Loss of lJ.4eAitefiAfj- VII. G.A-412 3.3.1-136 119 Steel A Tank boundary (External) material (LR-ISG- [LR-/SG-Fire Water (DB-T81) 2012-021 2012-021