NRC-15-0001, Response to NRC Request for Additional Information for the Review License Renewal Application - Set 7

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Response to NRC Request for Additional Information for the Review License Renewal Application - Set 7
ML15005A508
Person / Time
Site: Fermi DTE Energy icon.png
Issue date: 01/05/2015
From: Kaminskas V
DTE Energy
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
NRC-15-0001, TAC MF4222
Download: ML15005A508 (29)
v • d • e


Text

viO A. Kainska s Ste vc rsdn 6400 N. Di Higway

Newpot, M

416 Tel:S 735661 Fax 734472 Email: k m

sr 10 CFR 54 January 5, 2015 NRC-15-0001 U. S. Nuclear Regulatory Commission Attention: Document Control Desk Washington D C 20555-0001

References:

1) Fermi 2 NRC Docket No. 50-341 NRC License No. NPF-43
2) DTE Electric Company Letter to NRC, "Fermi 2 License Renewal Application," NRC-14-0028, dated April 24, 2014 (ML14121A554)
3) NRC Letter, "Requests for Additional Information for the Review of the Fermi 2 License Renewal Application - Set 7 (TAC No.

MF4222)," dated December 4, 2014 (ML14323A880)

Subject:

Response to NRC Request for Additional Information for the Review of the Fermi 2 License Renewal Application - Set 7 In Reference 2, DTE Electric Company (DTE) submitted the License Renewal Application (LRA) for Fermi 2. In Reference 3, NRC staff requested additional infonnation regarding the Fermi 2 LRA. The Enclosure to this letter provides the DTE response to the request for additional information.

No new commitments are being made in this submittal.

Should you have any questions or require additional information, please contact Lynne Goodman at 734-586-1205.

USNRC NRC-15-0001 Page 2 I declare under penalty of perjury that the foregoing is true and correct.

Executed on Janu 5, 2015 Vito A. Kaminskas Site Vice President Nuclear Generation

Enclosure:

DTE Response to NRC Request for Additional Information for the Review of the Fermi 2 License Renewal Application - Set 7 cc: NRC Project Manager NRC License Renewal Project Manager NRC Resident Office Reactor Projects Chief, Branch 5, Region III Regional Administrator, Region III Michigan Public Service Commission, Regulated Energy Division (kindschl@michigan.gov)

Enclosure to NRC-15-0001 Fermi 2 NRC Docket No. 50-341 Operating License No. NPF-43 DTE Response to NRC Request for Additional Information for the Review of the Fermi 2 License Renewal Application - Set 7

Enclosure to NRC-15-0001 Page 1 RAI B.1.8-1

Background

GALL Report AMP XL M7, "BWR Stress Corrosion Cracking," states that the program to manage intergranular stress corrosion cracking (IGSCC) in BWR coolant pressure boundary piping is delineated in NUREG-0313, Revision 2 and NRC Generic Letter (GL) 88-01 with its Supplement ] The "detection of aging effects" program element of GALL Report AMP XI M7 also states that modifications of the extent and schedule of inspection in NRC GL 88-01 are allowed in accordance with the inspection guidance in staff-approved BWRVIP-75-A.

License Renewal Application (LRA) Section B. 1.8 states that the applicant's BWR Stress Corrosion Cracking Program is consistent with GALL Report AMP XL M7. LRA Section B.1.8 also states that the scheduled volumetric examinations of the applicant's program provide timely detection of IGSCC and leakage of coolant in accordance with the methods, inspection guidelines, and flaw evaluation specified in NUREG-0313, Revision 2; NRC GL 88-01 and its Supplement 1; BWRVIP-75-A; ASME Code; and other requirements of 10 CFR 50.55a with the NRC-approved alternatives.

During the audit, the staff noted that the applicant implemented risk-informed inservice inspection for the current (third) inservice inspection interval. The staff also noted that GL 88-01 Category A (resistant material) welds are subsumed in the applicant's risk-informed inservice inspection.

Issue The staff noted that the LRA and program evaluation report do not describe what percentage of the Category A welds, which are subsumed in the risk-informed inservice inspection, are inspected by the applicant. It is unclear to the staff whether the percentage of Category A welds, which the applicant's program inspects, is consistent with the guidance provided in GL 88-01 and BWRVIP-75-A. The staff finds that additional information is necessary to confirm the consistency of the applicant's program with GALL Report AMP XI M7.

Request

1. Provide the percentage of Category A welds that the BWR Stress Corrosion Cracking Program will inspect during the period of extended operation.
2. If the extent of the inspection for Category A welds is different from the guidance in GL 88-01 and BWR VIP-75-A, provide justification for why the program is adequate to manage the aging effect of IGSCC for Category A welds.

Enclosure to NRC-15-0001 Page 2

Response

1. The Fermi 2 BWR Stress Corrosion Cracking Program selects the Category A intergranular stress corrosion cracking (IGSCC) welds at a 10% selection rate consistent with BWRVIP-75A recommendations. BWRVIP-75A paragraph 3.1.1 states, "Welds that are classified as Category A and are classified as Category B-J per Section XI will be examined at a rate of 10% over ten years. Fifty percent (50%) of these examinations must be completed during the first six years of the ten-year interval." Fermi 2 has 38 Generic Letter (GL) 88-01 Category A piping welds and four of those are selected for inspection during each ISI interval meeting the 10% selection rate. The welds are scheduled such that two are completed during the first period, and one in each of the following two periods of the 10-year interval. This meets the additional stipulation that 50% are examined in the first 6 years.

The selection rate and scheduling pattern will be maintained through the period of extended operation.

2. The BWR Stress Corrosion Cracking Program is consistent with the BWRVIP-75-A provisions as specified by the NRC in AMP XI.M7 and as stated in License Renewal Application (LRA) Section B.1.8.

LRA Revisions:

None.

Enclosure to NRC-15-0001 Page 3 R AIB.8-2

Background:

LRA Section B. 1.8 states that the applicant's B WR Stress Corrosion Cracking Program is consistent with GALL Report AMP XI M7, "BWR Stress Corrosion Cracking." In its review of the applicant's program and related information, the staff noted that the following references indicate that the applicant's condensate and feedwater systems include 24 Category D welds per GL 88-01 (i.e., welds with materials non-resistant to intergranular stress corrosion cracking and with no stress improvement process).

  • Table 3.4.2-3-2, "Condensate System, Nonsafety-Related Components Affecting Safety-Related Systems" Table 3.4.2-2, "Feedwater and Standby Feedwater System "

Table 3.4.2-3-3, "Feedwater and Standby Feedwater System, Nonsafety-Related Components Affecting Safety-Related Systems" Issue:

The staff noted that the AMR tables for the condensate and feedwater systems in the LRA do not include any AMR items to manage IGSCC for the Category D welds that were identified in the 1992 communications between the Detroit Edison Company and NRC. The staff cannot determine the adequacy of the applicant's program and AMR results without additional information to justify the omission of relevant AMR items.

The staff also noted that the 1992 communications indicate that these Category D welds are located outboard of the containment isolation valves and at least 10 percent of these welds should be inspected during each refueling outage as part of the applicant's inservice inspection.

The staff further noted that the extent and frequency of the application's inspections are different from the inspection guidance provided in GL 88-01 and BWRVIP-75-A. For example, BWRVIP-75-A states that in the case of the implementation of hydrogen water chemistry 100 percent of Category D welds should be inspected every 10 years and at least 50 percent of these welds

Enclosure to NRC-15-0001 Page 4 should be inspected in the first 6 years. However, the LRA does not identify this difference as a program exception.

Request:

1. Explain whether or not the condensate and feedwater systems include Category D welds.

Provide adequate justification for why the LRA AMR tables for condensate and feedwater systems do not include AMR items to manage IGSCC for the Category D welds.

2. Clarify why the LRA does not identify the inspection extent and frequency for Category D welds as a program exception to GALL ReportAMPXI.M7. In addition, provide technical justification for why the inspection extent and frequency for Category D welds are acceptable for adequate aging management. As part of the response, discuss whether the plant-specific operating experience, including inspection results, justify the adequacy of aging management for Category D welds.

Response

Generic Letter (GL) 88-01 (Position on IGSCC in BWR Austenitic Stainless Steel Piping) and NUREG-0313, Rev. 2 (Technical Report on Material Selection and Processing Guidelines for BWR Coolant Pressure Boundary Piping) are applicable to susceptible piping that contains reactor coolant. The term. reactor coolant pressure boundary is a regulatory defined term in 10 CFR 50.2, that includes the reactor coolant system up to and including the outermost containment isolation valve in the main steam and feedwater piping for boiling water reactors and reactor coolant system connected piping up to the outer containment isolation valve in system piping which penetrates primary reactor containment.

In letter NRC-92-0090 dated July 29, 1992, Fermi 2 proposed deleting the 24 Category D welds since they do not contain reactor coolant as defined in NRC regulations, and the welds were not likely susceptible to IGSCC based on the operating environment. The NRC did not completely agree with that position because the water was the same water that would enter the reactor pressure vessel (RPV). Instead, in the response dated December 18, 1992 (TAC No. M84177) the NRC stated, "These welds may be susceptible to IGSCC. Therefore, the 24 welds should not be removed from the Fermi-2 ISI-NDE Program." The correspondence also stated, "the staff has determined that the same staff position delineated in GL 88-01, Supplement 1 for reactor water cleanup (RWCU) piping outboard of the containment isolation valves should be applied to those 24 welds." GL 88-01, Supplement 1 was issued to discuss some unnecessary hardships created by GL 88-01. One specific hardship was inspection of susceptible nonsafety-related piping outside the containment isolation valves. Therefore, the NRC correspondence concluded that the staff determined that "an inspection of the subject piping on a sampling basis of at least 10 percent of the weld population should be performed during each refueling outage."

Enclosure to NRC-15-0001 Page 5 Fermi 2 subsequently established an augmented inspection requirement to examine 3 of the nonsafety-related feedwater condensate welds during each refueling outage. No IGSCC has been found in the inspections performed to date.

1. The condensate and feedwater systems do include Category D welds. The License Renewal Application (LRA) did not discuss the nonsafety-related Category D welds because AMP XLM7 (BWR Stress Corrosion Cracking) indicates that its purpose is "to manage intergranular stress corrosion cracking (IGSCC) in boiling water reactor (BWR) coolant pressure boundary piping made of stainless steel (SS) and nickel-based alloy components as delineated in NUREG-0313, Rev. 2, and Nuclear Regulatory Commission (NRC) Generic Letter (GL) 88-01 and its Supplement 1." As stated previously these nonsafety-related welds are not in the reactor coolant pressure boundary as defined in 10 CFR 50.2.

Based on this request, DTE will add line items to LRA Tables 3.4.2-2 and 3.4.2-3-2.

Additionally, the BWR Stress Corrosion Cracking Program will be added to LRA Sections 3.4.2.1.2 and 3.4.2.1.3. The LRA revisions are indicated below.

2. The LRA did not consider the inspection extent and frequency for Category D welds an exception because NUREG-1801 Section XI.M7 specifically references GL 88-01, Supplement 1. The extent and frequency of the nonsafety-related feedwater and condensate piping examinations that are performed are based on the NRC determination that examination of 10% of these welds should be performed each outage. No IGSCC has been found in the inspections performed to date. Therefore, cracking due to IGSCC in the nonsafety-related feedwater and condensate piping is being managed adequately as recommended by the NRC.

Since the determination was made in the December 18, 1992 NRC letter to apply the position in GL 88-01, Supplement 1 for RWCU piping to the Fermi 2 feedwater and condensate Category D welds, rather than directly in the GL 88-01, Supplement 1, the LRA will be revised in Appendices A and B to state that this is an exception. The LRA revisions are indicated below.

LRA Revisions:

LRA Sections 3.4.2.1.2, 3.4.2.1.3, A.1.8, and B.1.8, and LRA Tables 3.4.2-2, 3.4.2-3-2, and B-3 are revised as shown on the following pages. Additions are shown in underline and deletions are shown in strike-through.

Enclosure to NRC-15-0001 Page 6 3.4.2.1.2 Feedwater and Standby Feedwater System Materials Feedwater and standby feedwater system components are constructed of the following materials.

Carbon steel

" Copper alloy > 15% zinc (inhibited)

" Glass Inccneli Stainless steel Aging Management Programs The following aging management programs manage the aging effects for the feedwater and standby feedwater system components.

Bolting Integrity Buried and Underground Piping BWR Stress Corrosion Cracking External Surfaces Monitoring

" Flow-Accelerated Corrosion

" Oil Analysis

" One-Time Inspection

" Water Chemistry Control - BWR

Enclosure to NRC-15-0001 Page 7 3.4.2.1.3 Miscellaneous Steam and Power Conversion Systems in Scope for 10 CFR 54.4(a)(2)

The following lists encompass materials, environments, aging effects requiring management, and aging management programs for the series 3.4.2-3-xx tables.

Nonsafety-related components affecting safety-related systems are constructed of the following materials.

o Aluminum

" Carbon steel o Copper alloy > 15% zinc or > 8% aluminum

" Elastomer

" Glass Inconel

" Nickel alloy

" Plastic

" Stainless steel Aging Management Programs The following aging management programs manage the effects of aging on nonsafety-related components affecting safety-related systems.

Bolting Integrity B\\ R _Stress Corrosion Crackig

  • External Surfaces Monitoring
  • Flow-Accelerated Corrosion
  • Internal Surfaces in Miscellaneous Piping and Ducting Components Oil Analysis
  • One-Time Inspection
  • Periodic Surveillance and Preventive Maintenance

" Selective Leaching

" Water Chemistry Control - BWR

Enclosure to NRC-15-0001 Page 8 Table 3.4.2-2 Feedwater and Standby Feedwater System Summary of Aging Management Evaluation Table 3.4.2-2 Feedwater and Standby Feedwater System Aging Effect Aging Component Intended Requiring Management NUREG-Table 1 Type Function Material Environment Management Programs 1801 Item Item Notes Valve body Pressure Stainless Lube oil (int)

Loss of material Oil Analysis VIII.D2.SP-3.4.1-44 A, 402 boundary steel 95 Valve body Pressure Stainless Treated water Loss of material Water VIll.D2.SP-3.4.1-16 A, 401 boundary steel (int)

Chemistry 87 Control -

BWR V\\eAd Pressure Stainless Treated water Cracking BWR Stress IV.A1.R-68 3.1.1-97

(.ozztt boundary steel

>_140F int Corrosion safe end and inconel Crackinq safe end to r Eeedwater heaters o 6

Enclosure to NRC-15-0001 Page 9 Table 3.4.2-3-2 Condensate System Nonsafety-Related Components Affecting Safety-Related Systems Summary of Aging Management Evaluation Table 3.4.2-3-2: Condensate System, Nonsafety-Related Components Affecting Safety-Related Systems Aging Effect Aging Component Intended Requiring Management NUREG-Table 1 Type Function Material Environment Management Programs 1801 Item Item Notes Valve body Pressure Stainless Treated water Loss of material Water VIII.E.SP-87 3.4.1-16 A, 401 boundary steel

> 140°F (int)

Chemistry Control -

BWR Valve body Pressure Stainless Waste water Loss of material Periodic VII.E5.AP-3.3.1-95 E

boundary steel (int)

Surveillance 278 and Preventive Maintenance Welds Pressure Stainless Treated water Crackinrg BWR Stress IV.A1.R-68 3.1.1-97 D

(nogleto boundary

stecl,

>_ 110 F_(int)

Corroion safe end and inco nel Craciing safe end to Feedwater heaters Nos.

Enclosure to NRC-15-0001 Page 10 A.1.8 BWR Stress Corrosion Cracking Program The BWR Stress Corrosion Cracking Program manages intergranular stress corrosion cracking (IGSCC) in stainless steel or nickel alloy reactor coolant pressure boundary piping and piping welds 4 inches or larger in nominal diameter containing reactor coolant at a temperature above 93°C (200 F) during power operation, regardless of code classification.

Scheduled volumetric examinations provide timely detection of IGSCC and leakage of coolant in accordance with the methods, inspection guidelines, and flaw evaluation criteria delineated in the ASME Code; NUREG-0313, Rev. 2; NRC GL 88-01 and its Supplement 1; NRC-approved BWRVIP-75-A; and other requirements specified per 10 CFR 50.55a with NRC-approved alternatives._Ten ercent ofVthe feedater anc condensate s stemsCatec ryD welds are ispecgd edeach rtefueNqutag The program includes preventive measures such as induction heating stress improvement, solution annealing, and mechanical stress improvement process to minimize stress corrosion cracking.

Enclosure to NRC-15-0001 Page 11 Table B-3 Fermi 2 Program Consistency with NUREG-1801 NUREG-1801 Comparison Consistent Programs with Program with NUREG-Programs with Exception to Name 1801 Enhancement NUREG-1801 Plant-Specific BWR Stress Corrosion Cracking

Enclosure to NRC-15-0001 Page 12 B.1.8 BWR STRESS CORROSION CRACKING Program Description The BWR Stress Corrosion Cracking Program manages IGSCC in nickel alloy, stainless steel, and cast austenitic stainless steel (CASS) reactor coolant pressure boundary piping and piping welds 4 inches or larger in nominal diameter containing reactor coolant at a temperature above 93°C (200°F) during power operation, regardless of code classification.

Scheduled volumetric examinations provide timely detection of IGSCC and leakage of coolant in accordance with the methods, inspection guidelines, and flaw evaluation criteria delineated in the ASME Code; NUREG-0313, Rev. 2, Technical Report on Material Selection and Processing Guidelines for BWR Coolant Pressure Boundary Piping; NRC GL 88-01 and its Supplement 1, NRC Position on Intergranular Stress Corrosion Cracking (IGSCC) in BWR Austenitic Stainless Steel Piping; NRC-approved BWRVIP-75-A; and other requirements specified per 10 CFR 50.55a with NRC-approved alternatives.

The program includes preventive measures such as induction heating stress improvement, solution annealing, and mechanical stress improvement process to minimize stress corrosion cracking.

NUREG-1801 Consistency The BWR Stress Corrosion Cracking Program is consistent with the program described in NUREG-1801,Section XI.M7, BWR Stress Corrosion Cracking th olne excp_ on.

Exceptions to NUREG-1801 None The j\\ R_ Stress Corrosion Crackin gPr ram is consistent withe nrooram described inU RE 1801.Section XIl. 7, BWR Stress Corrosion Crackin, with the fojjowini er~e tiorc n.

EleetAf Eggption Dj.eDtection of AgjgEjle~a The inspecton frequency for the

5. vlonotri and _Trerdjn feedwater and condensate system cateor jD welds is 10% each refuejno outage)

Ecejotjor gjote 1_ The section freouencko jt ecot 10% of je _CategorD feedvswter no condensate weid opulaton each rrefuejjnqgtage was aorioyedin an NRC letter dated Decenber 18, 1992 (TAC No. M84177) usina the same staff osion

Enclosure to NRC-15-0001 Page 13 oeineated 'nL 88011 ypemert '(orreagtor eater ;ean y 2-pjncg t

O tee -ortirnent Esolaton vala;es.

Enhancements None

Enclosure to NRC-15-0001 Page 14 RAI B.1.8-3 Backg ound:

During the AMP audit, the staff noted that the applicant's Condition Assessment Resolution Document (CARD) 13-23127 addresses the Institute of Nuclear Power Operations Event Report (IER) 13-17, "Main Condenser Cooling Water Inleakage." CARD 13-23127 states that, since January 2011, events have been reported in which condenser cooling water inleakage resulted in scrams, a forced shutdown, and outage extensions. CARD 13-23127 also states that IER 13-17 indicates that the condenser inleakage events caused the introduction of sodium, chloride, sulfate, and other contaminants into the reactor coolant system and contributed to out-of-specification reactor water chemistry, requiring operations personnel to enter abnormal operating procedures more frequently.

The staff noted that CARD 11-21607 further states that during the startup on February 10, 2011, the applicant's plant was shut down due to main condenser tube inleakage and associated water chemistry excursions. CARD 11-21607 also indicates that inspections of all condenser water boxes identified the ejection of tube plugs from receptive condenser tubes.

The staff noted that another condition assessment document of the applicant (CARD 08-26361) indicates that at the applicant's plant, condenser cooling water inleakage occurred at an estimate inleakage rate of 40 - 50 gallons per day. This CARD also states that pressure testing and inspections identified the leaking condenser tube and the tube was plugged along with several other tubes.

Issue:

The ingress of chloride, sulfate, and other contaminants into the reactor coolant system due to main condenser inleakage can promote IGSCC in BWR piping and piping welds. However, LRA Section B. 1.8 and onsite program evaluation report for the applicant's BWR Stress Corrosion Cracking Program do not clearly address the potential impact of condenser cooling water inleakage on the effectiveness of the applicant's program. The staff finds that additional information is necessary to confirm that the applicant's assessment of the operating experience regarding condenser inleakage ensures the effectiveness ofthe applicant's program.

Request:

1. Clarify whether the main condenser inleakage and associated water chemistry excursions contributed to IGSCC in the piping and piping welds that are within the program scope of GALL Report AMP XI.M7. As part of the response, explain whether previous occurrences of IGSCC, if any, were attributed to water chemistry control issues.

Enclosure to NRC-15-0001 Page 15

2. Discuss the assessment of industry and plant-specific operating experience regarding condenser cooling water inleakage and provide adequate justification for why there is no need to enhance the BWR Stress Corrosion Program.

Response

1. Chemistry excursions have not contributed to initiation of IGSCC since no cracking has been detected to date in any piping within the scope of the Fermi 2 BWR Stress Corrosion Cracking Program that is consistent with the NUREG-1801 Section XI.M7 BWR Stress Corrosion Cracking Program.
2. Multiple factors are necessary for initiation of IGSCC (susceptible material, residual stress, and corrosive environment). These factors are assumed to be present. Fermi 2 proactively mitigated high inside diameter (ID) tensile stress prior to two years of service in accordance with GL 88-01 and NUREG-0313 for all susceptible welds within the reactor coolant pressure boundary. Site water chemistry monitoring procedures and processes manage water chemistry in accordance with EPRI guidance, which is based on industry-wide operating experience that has shown that guidance is effective. If a parameter is out of specification, it is addressed and corrected in a timely manner. This is evidenced by the action taken in 2014 to promptly take action when tube leakage was identified, including lowering reactor power for condenser tube maintenance. Hydrogen Water Chemistry and Noble Metal Chemical addition are part of the site plan to mitigate the environmental conditions. Additionally, all reactor coolant pressure boundary materials are GL 88-01 Category A or B (resistant or mitigated). A sample of 10% of feedwater and condensate welds which may be potentially susceptible to IGSCC are also included as described in the response to RAI B.1.8-2. The BWRVIP-75 inspection sample sizes of 10% for Category A and 25% for Category B are acceptable to detect the onset of degradation (cracking). Inspections are performed routinely each outage, period, and interval to detect the start of cracking if it were to occur. No changes to inspection population or inspection frequency are necessary based on assessment of industry and plant-specific operating experience regarding condenser cooling water inleakage.

LRA Revisions:

None.

Enclosure to NRC-15-0001 Page 16 RAI B.1.38-1

Background:

LRA Section B.1.38 for the Reactor Vessel Surveillance Program indicates that the applicant participates in the BWR VIP Integrated Surveillance Program which is described in BWRVIP-86, Revision 1, "Updated BWR Integrated Surveillance Program (ISP) Implementation Plan,"

September 2008 (ADAMS Accession Number ML090300555). The staff noted the following reference also addresses technical information related to ISP surveillance materials for the applicant's reactor vessel.

Tables 4-5 and 4-6 of GE Report NEDO-33133, Revision 0, "Pressure-Temperature Curves for DTE Energy Fermi Unit 2," February 2005 (ADAMS Accession Number ML050870587)

LRA Section B.1.38 identifies a program exception to the "detection of aging effects" program element of GALL Report AMP XI M31, "Reactor Vessel Surveillance." The program exception states that GALL Report AMP XI M31 recommends that the program shall have at least one capsule with projected neutron fluence equal to or exceeding the 60-year peak reactor vessel wall neutron fluence prior to the end of the period of extended operation. The program exception also states that a capsule meeting this qualification is not expected to be obtained prior to the end of the period of extended operation.

In its review of the program exception, the staff noted that the BWRVIP ISP includes a surveillance weld material which represents applicant's target vessel weld material (heat number 13253/12008). The staff also noted that the BWRVIP ISP includes a surveillance plate material which represents applicant's reactor vessel plate materials (heat numbers C4554-1 and C4568-2). The stafffurther noted that each of these surveillance materials was irradiated or is being irradiated in one of the host reactor vessels, which are diferent from the applicant's reactor vessel, as planned in the ISP.

In addition, the staff noted that the BWRVIP ISP tested the surveillance weld material at fluence levels greater than 1.43x1018 n/cm2 (E > 1 MeV), as described in the following references:

Table 5-1 of BWRVIP-111NP, Revision 1, "Testing and Evaluation of BWR Supplemental Surveillance Program Capsules E, F, and I," August 2010 (ADAMS Accession Number ML080780267) e Table 5-2 ofBWR VIP-87NP, Revision 1, "Testing and Evaluation ofBWR Supplemental Surveillance Program Capsules D, G, and H, "September 2007 (ADAMS Accession Number ML080770344)

LRA Section 4.2.1 states that the applicant's peak reactor vessel wall fluence for 60 years of operation is 1.43xJ18 n/cm2 (E > 1 Me V) indicating that the fluences of the tested surveillance

Enclosure to NRC-15-0001 Page 17 weld material are between one and two times the applicant's peak reactor vessel wall fluence for 52 EFPYs (60 years of operation).

In its review of the program exception, the staff also noted that the BWRVIP ISP has a plan to test the representative surveillance plate material for the applicant's reactor vessel prior to the end of the period of extended operation at an estimated fluence between one and two times the applicant's peak reactor vessel wall fluence for 52 EFPYs, consistent with the GALL Report.

Issue:

As described above, the fluences (E > 1 Me V) of the ISP surveillance weld and plate materials, which represent the applicant's reactor vessel materials, range between one and two times the peak reactor vessel wall fluence for 52 EFPYs (60 years of operation). However, the program exception identified in the LRA states that the applicant's program does not include a surveillance capsule which meets the fluence range specified in the GALL Report for the period of extended operation. The stafffinds that additional clarification is necessary to resolve this apparent inconsistency between the program exception and the ISP surveillance capsule withdrawal schedule for the applicant's reactor vessel.

Request:

Clarify whether the applicant's program includes a surveillance capsule which meets the fluence range specified in the GALL Report.for the period of extended operation. As part of the response, clarify whether the capsule withdrawal schedule and associated fluences of the ISP for the applicant's reactor vessel have been changed or updated in such a manner that the program exception needs to be identified.

Response

The intent of the exception was not to circumvent participation in a reactor vessel surveillance program as recommended by NUREG-1801. Rather, the exception is in regards to the fact that a capsule with the representative surveillance materials is not physically located within the Fenni 2 reactor vessel. Fermi 2 License Amendment No. 152 approved the participation of the plant in the Boiling Water Reactor Vessel Internals Project (BWRVIP) Integrated Surveillance Program (ISP). Through the ISP, Fermi 2 representative surveillance materials with the appropriate peak fluences are tested in accordance with 10 CFR 50 Appendix H.

Since Fermi 2 maintains participation in the BWRVIP ISP consistent with provisions of NUREG-1801 Section XI.M31, the License Renewal Application (LRA) will be revised as indicated below to remove this particular exception.

Enclosure to NRC-15-0001 Page 18 LRA Revisions:

LRA Section B.1.38 and LRA Table B-3 are revised as shown on the following pages. Additions are shown in underline and deletions are shown in strike-through.

Enclosure to NRC-15-0001 Page 19 Table B-3 Fermi 2 Program Consistency with NUREG-1801 NUREG-1801 Comparison Consistent Programs with Program with NUREG-Programs with Exception to Name 1801 Enhancement NUREG-1801 Plant-Specific Reactor Vessel X

X Surveillance

Enclosure to NRC-15-0001 Page 20 B1.38 REACTOR VESSEL SURVEILLANCE NUREG-1801 Consistency The Reactor Vessel Surveillance Program, with enhancement, is consistent with the program described in NUREG-1801,Section XI.M31, Reactor Vessel Surveillance-v4ta Exceptions to NUREG-1801 Non e The-R c-tr-Vessel SUvreilenee Ggfam-(

th ehaneement)- is-ntt-wit+/-The program described in NUREG 1801, Section XEM31, with the fo!!owing exception.

EleeentAffes e Excepi~itionpm~a Detection of ^Aging EffectN4nrte reactor vessel survelliance program shal! have-at least one capsule with p~ejeceed-eten4ueneequeMGe-oF exeeedi-h-0-eGFk-rat~

vessc! wall neutron fluence prior to the e

aio the psednoetexten is quificaion note-p e id-e-h obtained prior to the end of the period ELsien-NeTT t

1. 4 ettedtc Fbu r 0 203 F

L--as-issued Lce men m Ilo 152 approing paticipatiin in the D\\A/D\\/

D afety Eahion Depor /SER\\

(TAOC Mo MC. E 90' dated- "Oto 9 -

ih04 e Etaf ap proed BWr\\ADE 1\\--

Drcniecn n,"pedn D\\APD Interaedr SRmeillanec rrrr ID imnementation Pln" The MRC staffc eluded that the ISD and P(E\\

Gntinu toL4 dequ address ho ret f

dix H to 10 CFPr BWR censeesdtr'

-ft end-eaehei liene. -BWRVI-P-86,Revision 1, Sction 5, includes provisions to applythe embrittlement evaluation decrbed in RG 1.9, Revision 2, for evaluating materials and ca~eul-ating-an -adju-sted referce temperature. The use of R~G 1.119, Roviin 2.prjet heembrittemontovaluati o is also dosrmbd in Ea m;t-fNR'E '301 Y' 3 Th -is-e e is justified b~c"t t.e rovisin-set fth in PG 1.9, Revision 2, are acceptable for embrittlement

Enclosure to NRC-15-0001 Page 21 RAI B. 1.38-2

Background:

LRA Section.B. 1.38 describes a program enhancement to the "monitoring and trending" program element of GALL Report AMP XI. M31, "Reactor Vessel Surveillance." The program enhancement states that the applicant will revise the program procedures to ensure that new fluence projections through the period of extended operation and the latest vessel beltline adjusted reference temperature (ART) tables are provided to the BWRVIP prior to the period of extended operation.

Issue:

The staff noted that the applicant's Reactor Vessel Surveillance Program is an existing program and upon receipt of a renewed license the applicant's program should continue to provide adequate fracture toughness and dosimetry data throughout the license renewal term. However the LRA states that the applicant's enhancement regarding data sharing of new fluence projections and associated ART tables will be implemented prior to the period of extended operation, but not within a specific time period upon receipt of the renewed license.

Request:

Provide adequate justification for why the program enhancement regarding data sharing of new fluence projections and associated ART tables will not be implemented within a specific time period upon receipt of renewed license.

Response

The Fermi 2 Surveillance Program has been integrated into the BWRVIP Integrated Surveillance program which follows the provisions of BWRVIP-86-1A "Updated BWR Integrated Surveillance Program Implementation Plan" and the latest revision of BWRVIP-135, "BWRVIP ISP Data Source Book and Plant Evaluations."

BWRVIP-135 Revision 2, Section 3 Item 10 (Licensee Responsibilities Regarding Information Exchange) states:

All plants are responsible to notify the BWRVIP of any changes in:

a. Fluence projections for RPV (I.D. and %T)
b. Fluence values for any previously withdrawn capsules (due to recalculated fluence)
c. Latest vessel beltline ART tables
d. Placement and location of all capsules This guidance is already in place to ensure that the transmittal of new fluence projections and associated ART tables in support of license renewal are shared with the BWRVIP. The

Enclosure to NRC-15-0001 Page 22 requirements in BWRVIP-86 or BWRVIP-135 do not indicate a specific time period in which new fluence projections and ART tables must be provided to the BWRVIP. The intent of the enhancement committing that the identified information would be provided prior to the period of extended operation is to formalize the need for information exchange, specifically for license renewal. Fermi 2 has had no previous issues in promptly submitting changes affecting the requested information to the BWRVIP. For example, revised ART tables due to the Measurement Uncertainty Recapture/Thermal Power Optimization (MUR/TPO) project implemented in early 2014 were submitted to the BWRVIP in March 2013 and January 2014. Therefore, no additional enhancement is necessary.

LRA Revisions:

None.

Enclosure to NRC-15-0001 Page 23 RAI 33.2173-1

Background:

LRA Table 3.3.2-17-3 states that copper alloy > 15% Zn or > 8% Al sight glasses exposed internally to waste water will be managed for loss of material by the Internal Surfaces in Miscellaneous Piping and Ducting Components Program and references GALL Report item VII E5.AP-272. GALL Report item VII E5.AP-272 recommends that for copper alloys exposed to waste water the loss of material due to pitting, crevice, and microbiologically-influenced corrosion should be managed by AMP XIM38, "Inspection of Internal Surfaces in Miscellaneous Piping and Ducting Components. " GALL Report AMP XI M38 does not address selective leaching.

Issue:

GALL Report AMP XIM33, "Selective Leaching," states that components constructed of copper alloys > 15% Zn or > 8% Al exposed to raw water, closed cooling water, treated water, or ground water may be susceptible to selective leaching. GALL Report Section IXD states that waste water may contain "treated water that is not monitored by a chemistry program."

Therefore, the copper alloy > 15% Zn or > 8% Al sight glasses exposed internally to waste water identified in the LRA corresponds to a material and environment combination which may be susceptible to selective leaching. Although, the LRA states in table 3.3.2-17-3 that this component will be managed for loss of material by AMP XIM38, this AMP does not address selective leaching. Therefore, it is not clear to the staff that selective leaching will be managed in a manner consistent with the GALL Report for these components.

Request:

Provide justification as to why this component is not susceptible to selective leaching; or state how selective leaching will be managed.

Response

The subject component, D 1R430, Condensate Collection Tank D110OA002 Level Indicator, is susceptible to selective leaching. A line item will be added to License Renewal Application (LRA) Table 3.3.2-17-3 for copper alloy > 15% Zn or > 8% Al sight glasses exposed internally to waste water crediting the Selective Leaching Program to manage loss of material due to selective leaching.

LRA Revisions:

LRA Table 3.3.2-17-3 is revised as shown on the following page. Additions are shown in underline and deletions are shown in strike-through.

Enclosure to NRC-15-0001 Page 24 Table 3.3.2-17-3 Process Radiation Monitoring System Nonsafety-Related Components Affecting Safety-Related Systems Summary of Aging Management Evaluation Table 3.3.2-17-3: Process Radiation Monitoring System, Nonsafety-Related Components Affecting Safety-Related Systems Aging Effect Aging Component Intended Requiring Management NUREG-Table 1 Type Function Material Environment Management Programs 1801 Item Item Notes Sight glass Pressure Copper Waste water Loss of material Internal VII.E5.AP-3.3.1-95 C

boundary alloy (int)

Surfaces in 272

> 15%

Miscellaneous Zn or Piping and

> 8% AI Ducting Components Sight glass Pressure Copper Waste water Loss of material Selective 3.3.172 A boundar i

( AinA Leachinq

> 15%%

.~n or

> 8% AI Sight glass Pressure Glass Condensation None None VII.J.AP-97 3.3.1-A boundarv ext 117

Enclosure to NRC-15-0001 Page 25 RAI 4.2.1-1

Background:

LRA Section 4.2.1 describes the applicant's time-limited aging analysis on reactor vessel fluence calculations. Specifically, LRA Section 4.2.1 states that the peak reactor vessel wall neutron fluence projected for 52 EFPY is 1.43x1018 n/cm2 (E > 1 Me V).

In comparison, Section 4.3.2.8.2, "Extended Power Uprate Analysis, " of the applicant's UFSAR indicates that a reactor vessel fluence evaluation was performed in support of an extended power uprate (EPU) to 120 percent to original licensed power. In addition, UFSAR Table 4.3-2 describes pre-EPU and EPUfluences for 32 EFPY (i.e., original license term).

Issue:

The LRA does not clearly address whether the fluence calculations are based on the operating conditions of a potential EPU described in UFSAR Section 4.3.2.8.2. It is unclear to the staff which operating power levels are used in the reactor vessel neutron fluence.

Request:

Clarify whether the neutron fluences described in the LRA are based on the operating conditions of a potential EPU described in UFSAR Section 4.3.2.8.2. As part of the response, clarify the operating power levels, on which the fluence calculations are based.

Response

The Updated Final Safety Analysis Report (UFSAR) Section 4.3.2.8.2 states that, "The 32-EFPY fluence used in developing the P-T curves detailed in UFSAR Section 5.2.4 is conservatively based upon operation at 3430 MWt for 12.04 EFPY and 3952 MWt for 19.96 EFPY."

3430 MWt is the initial power uprate and 3952 MWt is the Extended Power Uprate (EPU) power level (120% of original licensed power). Although, the EPU power increase was never pursued, EPU power was conservatively assumed for the balance of the original operating period.

Also described in the UFSAR Section 4.3.2.8.2 for the N16 Water Level Instrument (WLI), "The fluence determined for the WLI nozzles is based upon operation at 3293 MWt for 3.4-EFPY, 3430 MWt for 16.38-EFPY, and 3486 MWt for 12.22-EFPY. TPO is expected to be implemented in Cycle 17, or after RF16." 3293 MWt is the original licensed power level, 3430 MWt is the initial power uprate and 3486 MWt is the Measurement Uncertainty Recapture/Thermal Power Optimization (MUR/TPO) power level. The N16 water level instrumentation nozzles fluence evaluation removed the conservatism associated with EPU and deternined fluence based on a realistic power history; including power increase to 3486 MWt for MUR/TPO beginning in Cycle 17. MUR/TPO power level increase was approved by the NRC for Cycle 17.

Enclosure to NRC-15-0001 Page 26 As stated in License Renewal Application (LRA) Section 4.2.1 Reactor Vessel Fluence, "The reactor vessel fluence is calculated using a higher power level beginning with cycle 17, when the reactor power increased due to the MUR/TPO uprate. The peak neutron fluence projected for 52 EFPY is 1.43E+18 n/cm at the vessel inner surface." The Operating Power Level used to determine the neutron fluence projection for 52 EFPY is 3486 MWt beginning in Cycle 17 and continuing through the period of extended operation. MUR/TPO power increase was approved by the NRC beginning in Cycle 17 to increase licensed power from 3430 MWt to 3486 MWt.

The 52-EFPY fluence is not based on the use of EPU power level projections.

LRA Revisions:

None.

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