ML14323A880
| ML14323A880 | |
| Person / Time | |
|---|---|
| Site: | Fermi  |
| Issue date: | 12/04/2014 |
| From: | Melendez-Colon D License Renewal Projects Branch 1 |
| To: | Kaminskas V DTE Electric Company |
| Melendez-Colon D, DLR/RPB1, 301-415-3301 | |
| References | |
| TAC MF4222 | |
| Download: ML14323A880 (10) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 December 4, 2014 Mr. Vito Kaminskas Site Vice President - Nuclear Generation DTE Electric Company Fermi 2 - 280 OBA 6400 North Dixie Highway Newport, MI 48166
SUBJECT:
REQUESTS FOR ADDITIONAL INFORMATION FOR THE REVIEW OF THE FERMI 2 LICENSE RENEWAL APPLICATION - SET 7 (TAC NO. MF4222)
Dear Mr. Kaminskas:
By letter dated April 24, 2014, DTE Electric Company submitted an application pursuant to Title 10 of the Code of Federal Regulations (CFR) Part 54, to renew the operating license NPF-43 for Fermi 2, for review by the U.S. Nuclear Regulatory Commission (NRC or the staff). The staff is reviewing the information contained in the license renewal application and has identified, in the enclosure, areas where additional information is needed to complete the review.
These requests for additional information were discussed with Ms. Lynne Goodman, and a mutually agreeable date for the response is within 30 days from the date of this letter. If you have any questions, please contact me at 301-415-3301 or e-mail Daneira.Melendez-Colon@nrc.gov.
Sincerely,
/RA/
Daneira Meléndez-Colón, Project Manager Projects Branch 1 Division of License Renewal Office of Nuclear Reactor Regulation Docket No. 50-341
Enclosure:
Requests for Additional Information cc w/encl: ListServ
- Concurred via e-mail OFFICE LA:RPB1:DLR* PM:RPB1:DLR PM:RPB1:DLR BC:RPB1:DLR PM:RPB1:DLR NAME YEdmonds DMelendez-Colon ESayoc YDiaz-Sanabria DMelendez-Colon DATE 11/24/14 11/25/14 11/24/14 12/4/14 12/4/14
SUBJECT:
REQUESTS FOR ADDITIONAL INFORMATION FOR THE REVIEW OF THE FERMI 2, LICENSE RENEWAL APPLICATION - SET 7 (TAC NO. MF4222)
DISTRIBUTION:
E-MAIL:
PUBLIC RidsNrrDlr Resource RidsNrrDlrRpb1 Resource RidsNrrDlrRpb2 Resource RidsNrrDlrRerb Resource RidsNrrDlrRarb Resource RidsNrrDlrRasb Resource RidsNrrDlrRsrg Resource RidsNrrPMFermi2 Resource D. Melendez-Colon Y. Diaz-Sanabria M. Wentzel B. Wittick D. McIntyre, OPA C. Kanatas, OGC D. Roth, OGC M. Kunowski, RIII B. Kemker, RIII V. Mitlyng, RIII P. Chandrathil, RIII A. Stone, RIII S. Sheldon, RIII
ENCLOSURE FERMI 2 LICENSE RENEWAL APPLICATION REQUESTS FOR ADDITIONAL INFORMATION SET 7 (TAC NO. MF4222)
RAI B.1.8-1
Background:
GALL Report AMP XI.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 1. 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 applicants BWR Stress Corrosion Cracking Program is consistent with GALL Report AMP XI.M7. LRA Section B.1.8 also states that the scheduled volumetric examinations of the applicants 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 applicants 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 applicants 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 applicants 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 BWRVIP-75-A, provide justification for why the program is adequate to manage the aging effect of IGSCC for Category A welds.
RAI B.1.8-2
Background:
LRA Section B.1.8 states that the applicants BWR Stress Corrosion Cracking Program is consistent with GALL Report AMP XI.M7, BWR Stress Corrosion Cracking. In its review of the applicants program and related information, the staff noted that the following references indicate that the applicants 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).
Letter from the Detroit Edison Company to the NRC (NRC-92-0090), Fermi 2 Response to GL 88-01, Supplement 1, NRC Position on Intergranular Stress Corrosion Cracking (IGSCC) in BWR Austenitic Stainless Steel Piping, July 29, 1992 Letter from the NRC to the Detroit Edison Company, Fermi-2 Removal of 24 Condensate and Feedwater System Welds from the Inservice Inspection Nondestructive Examination (ISI-NDE) Program (TAC No. M84177), December 18, 1992 The staff also noted that the following tables of the LRA describe the applicants aging management review (AMR) items for the condensate and feedwater systems.
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 applicants 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 applicants inservice inspection.
The staff further noted that the extent and frequency of the applications 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 should be inspected in the first 6 years. However, the LRA does not identify this difference as a program exception.
Request:
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 Report AMP XI.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.
RAI B.1.8-3
Background:
During the AMP audit, the staff noted that the applicants 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 applicants 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 applicants 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 applicants BWR Stress Corrosion Cracking Program do not clearly address the potential impact of condenser cooling water inleakage on the effectiveness of the applicants program. The staff finds that additional information is necessary to confirm that the applicants assessment of the operating experience regarding condenser inleakage ensures the effectiveness of the applicants 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.
- 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.
RAI B.1.38-1
Background:
LRA Section B.1.38 for the Reactor Vessel Surveillance Program indicates that the applicant participates in the BWRVIP 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 applicants 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 applicants target vessel weld material (heat number 13253/12008). The staff also noted that the BWRVIP ISP includes a surveillance plate material which represents applicants reactor vessel plate materials (heat numbers C4554-1 and C4568-2). The staff further noted that each of these surveillance materials was irradiated or is being irradiated in one of the host reactor vessels, which are different from the applicants 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)
Table 5-2 of BWRVIP-87NP, Revision 1, Testing and Evaluation of BWR Supplemental Surveillance Program Capsules D, G, and H, September 2007 (ADAMS Accession Number ML080770344)
LRA Section 4.2.1 states that the applicants peak reactor vessel wall fluence for 60 years of operation is 1.43x1018 n/cm2 (E > 1 MeV) indicating that the fluences of the tested surveillance weld material are between one and two times the applicants 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 applicants reactor vessel prior to the end of the period of extended operation at an estimated fluence between one and two times the applicants peak reactor vessel wall fluence for 52 EFPYs, consistent with the GALL Report.
Issue:
As described above, the fluences (E > 1 MeV) of the ISP surveillance weld and plate materials, which represent the applicants 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 applicants program does not include a surveillance capsule which meets the fluence range specified in the GALL Report for the period of extended operation. The staff finds that additional clarification is necessary to resolve this apparent inconsistency between the program exception and the ISP surveillance capsule withdrawal schedule for the applicants reactor vessel.
Request:
Clarify whether the applicants 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 applicants reactor vessel have been changed or updated in such a manner that the program exception needs to be identified.
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 applicants Reactor Vessel Surveillance Program is an existing program and upon receipt of a renewed license the applicants program should continue to provide adequate fracture toughness and dosimetry data throughout the license renewal term.
However, the LRA states that the applicants 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.
RAI 3.3.2.17.3-1
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 XI.M38, Inspection of Internal Surfaces in Miscellaneous Piping and Ducting Components. GALL Report AMP XI.M38 does not address selective leaching.
Issue:
GALL Report AMP XI.M33, 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 IX.D 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 XI.M38, 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.
RAI 4.2.1-1
Background:
LRA Section 4.2.1 describes the applicants 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 MeV).
In comparison, Section 4.3.2.8.2, Extended Power Uprate Analysis, of the applicants 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 EPU fluences 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.