NL-08-084, Official Exhibit - ENT000194-00-BD01 - Reply to Request for Additional Information Regarding License Renewal Application - Time-Limited Aging Analyses and Boraflex
| ML12338A529 | |
| Person / Time | |
|---|---|
| Site: | Indian Point  |
| Issue date: | 05/16/2008 |
| From: | Dacimo F Entergy Nuclear Northeast |
| To: | Atomic Safety and Licensing Board Panel, Document Control Desk, Office of Nuclear Reactor Regulation |
| SECY RAS | |
| References | |
| RAS 22112, 50-247-LR, 50-286-LR, ASLBP 07-858-03-LR-BD01, NL-08-084 | |
| Download: ML12338A529 (8) | |
Text
United States Nuclear Regulatory Commission Official Hearing Exhibit Entergy Nuclear Operations, Inc.
In the Matter of:
(Indian Point Nuclear Generating Units 2 and 3) c,,,t..p.'R REGV<.q" ASLBP #: 07-858-03-LR-BD01 l~\
Docket #: 05000247 l 05000286
" 0~
Exhibit #: ENT000194-00-BD01 Identified: 10/15/2012 I-
'" . ~ Admitted: 10/15/2012 Withdrawn: ENT000194
~
........1-: &
i Rejected: Stricken: Submitted: March 29, 2012
" ****. " Other:
( Entergy Nuclear Northeast
\ Indian Point Energy Center
~Entergx 450 Broadway, GSB P.O. Box 249 Buchanan, NY 10511-0249 Tel (914) 788"2055 Fred R. Dacimo Vice President License Renewal May 16, 2008 Re: Indian Point Units 2 & 3 Docket Nos. 50-247 & 50-286 NL-08-084 U.s. Nuclear Regulatory Commission A TIN: Document Control Desk Washington, DC 20555-0001
SUBJECT:
Reply to Request for Additional Information Regarding License Renewal Application -
Time-Limited Aging Analyses and Boraflex
Reference:
NRC letter dated April 18, 2008; "Request for Additional Information for the Review of the Indian Point Nuclear Generating Unit Nos. 2 and 3, License Renewal Application - Time-Limited Aging Analyses and Boraflex"
Dear Sir or Madam:
Entergy Nuclear Operations, Inc is providing, in Attachment I, the additional information requested in the referenced letter pertaining to NRC review of the License Renewal Application for Indian Point 2 and Indian Point 3. The additional information provided in this transmittal addresses staff questions for Time-Limited Aging Analyses and Boraflex.
There are no new commitments identified in this submittal. If you have any questions or require additional information, please contact Mr. R. Walpole, Manager, Licensing at (914) 734-6710.
I declare under penalty of perjury that the foregoing is true and correct. Executed on S - IlsJ- 0"6.
Sincerely, v(Jd£Wk!i!~ P Fred R. Dac~~o - - I Vice President License Renewal
NL-08-084 Docket Nos. 50-247 & 50-286 Page 2 of 2
Attachment:
- 1. Reply to NRC Request for Additional Information Regarding License Renewal Application - Time-Limited Aging Analyses and Boraflex cc: . Mr. Bo M. Pham, NRC Environmental Project Manager Ms. Kimberly Green, NRC Safety Project Manager Mr. John P. Boska, NRC NRR Senior Project Manager Mr. Samuel J. Collins, Regional Administrator, NRC Region I Mr. Sherwin E. Turk, NRC Office of General Counsel, Special Counsel IPEC NRC SeniQr Resident Inspectors Office Mr. Paul D. Tonko, President, NYSERDA Mr. Paul Eddy, New York State Dept. of Public Service
ATTACHMENT I TO NL-08-084 REPLY TO NRC REQUEST FOR ADDITIONAL INFORMATION REGARDING LICENSE RENEWAL APPLICATION Time-Limited Aging Analyses and Boraflex ENTERGY NUCLEAR OPERATIONS, INC INDIAN POINT NUCLEAR GENERATING UNIT NOS. 2 and 3 DOCKETS 50-247 and 50-286
NL-08-84 Attachment I Docket Nos. 50-247 & 50-286 Page 1 of 5 INDIAN POINT NUCLEAR GENERATING UNIT NOS. 2 AND 3 LICENSE RENEWAL APPLICATION (LRA)
REQUESTS FOR ADDITIONAL INFORMATION (RAI)
REGARDING TIME-LIMITED AGING ANALYSES AND BORAFLEX Time-Limited Aging Analyses RAI 4.3.1.8-1 License renewal application (LRA) Section 4.3.1 states "[c]urrent design basis fatigue evaluations calculate cumulative usage factors (CUFs) for components or sUb-components based on design transient cycles." For CUF values listed in LRA Tables 4.3-13 and 4.3-14, please describe the details of how various environmental effects are factored into the calculation of the CUF using Fen values.
Response to RAI 4.3.1.8-1 For CUF values listed in LRA Tables 4.3-13 and 4.3-14, the Fen values were determined as described below.
NUREG-1801 calls for using formulas provided in NUREG/CR-5704 for austenitic stainless steel and NUREG/CR-6583 for carbon steel and low-alloy steel to calculate environmentally assisted fatigue correction factors (Fen). For IPEC, none of the locations identified in Tables 4.3-13 and 4.3-14 (NUREG/CR-6260 locations) are made of carbon steel, so calculation of Fen for carbon steel was not required.
The environmentally assisted fatigue correction factor (Fen) for low alloy steel was calculated as follows.
Fen = exp(0.929-0.00124T-0.101 S"T"O"e"} based on NUREG/CR-6583, Eq. 6.5b T = 25°C* Reference temperature for original fatigue curves S" = S (0 < S (Sulfur) s 0.015 wt.%)
S" = 0.015 (S ~ 0.015 wt. %) NUREG/CR-6583, Eq. 5.5a T" = 0 (T (Temperature) < 150°C)
T" = T-150 =
(T 150-350°C) NUREG/CR-6583, Eq. 5.5b 0" = 0 (DO (Dissolved Oxygen) < 0.05ppm}
0" = In(DOI0.04} (0.05 ppm s DO s 0.5 ppm) 0" = =
In(12.5} 2.53 (00 > 0.5 ppm) NUREG/CR-6583, Eq. 5.5c E
." = 0 (t (strain rate) > 1%/s}
,
- E
= In(t} (0.001 s t s 1%/s)
E = In(O.001} (t < 0.001 %/s) NUREG/CR-6583, Eq. 5.5d There are four low alloy steel subcomponents for each unit in Tables 4.3-13 and 4.3-14 for the NUREG-6260 locations at lPEC. The Fen was calculated for each location on each unit as shown below.
NL-OS-S4 Attachment I Docket Nos. 50-247 & 50-286 Page 2 of 5 IP2 T(reference temperature °C) =25 Reference temperature for original fatigue curves O*(RCS) = 0.0 ReS dissolved oxygen is s 50 ppb Since O'(RCS) equals 0.0, S*, T*, and E' terms are eliminated.
= exp(0.929-(0.00124)(T))
Fen (bottom head to shell) = exp(0.929-(0.00124)(25)) = 2.45 Fen (inlet nozzles) = exp(0.929-(0.00124)(25)) = 2.45 Fen (outlet nozzles) = exp(0.929-(0.00124)(25)) = 2.45 Fen (surge line nozzles) = exp(0.929-(0.00124)(25)) = 2.45 IP3 T(reference temperature °C) = 25 Reference temperature for original fatigue curves Since O'(RCS) equals 0.0, S*, T*, and E' terms are eliminated.
Fen = exp(0.929-(0.00124)(T))
Fen (bottom head to shell) = exp(0.929-(0.00124)(25)) = 2.45 Fen (inlet nozzles) = exp(0.929-(0.00124)(25)) = 2.45 Fen (outlet nozzles) = exp(0.929-(0.00124)(25)) = 2.45 Fen (surge line nozzles) = exp(0.929-(0.00124)(25)) = 2.45 The environmentally assisted fatigue correction factor (Fen) for austenitic stainless steel was calculated as follows.
Fen = exp(0.935-T'O'e') NUREG/CR-5704, Eq. 13 T' = 0 (T < 200QC)
T' = 1 (T ~ 200QC) NUREG/CR"5704, Eq. 8a 0' = 0.260 (DO < 0.05 ppm) 0' = 0.172 (DO ~ 0.05 ppm) NUREG/CR-5704, Eq. 8b
£' = 0 (£ > O.4%/s)
£' = In(£ 10.4) (0.0004 s £ s O.4%/s)
£' = In(0.0004/0.4) (£ < 0.0004%/s) NUREG/CR-5704, Eq. 8c There are four stainless steel subcomponents for each unit in Tables 4.3-13 and 4.3-14 for the NUREG-6260 locations at IPEC. The Fen will be calculated for each location on each unit as shown below.
IP2 T' (Surge line) = 1.0 (T ~ 200QC)
T' (Charging nozzle) = 1.0 (T ~ 200QC)
T' (SI nozzle) = 1.0 (T ~ 200QC)
T' (RHR piping) = 1.0 (T ~ 200QC)
i, I
NL-08-84 Attachment I Docket Nos. 50-247& 50-286 Page 3 of 5 0' (All) = 0.260 RCS dissolved oxygen is s 50 ppb t ' (all) = -6.91 Assume bounding strain rate Fen (Surge line) = exp(0.935-(1.0)(0.260)(-6.91)) = 15.35 Fen (Charging nozzle) = exp(0.935-(1.0)(0.260)(-6.91)) = 15.35 Fen (SI nozzle) = exp(O.935-(1 .0)(0.260)(-6.91)) = 15.35 Fen (RHR piping) = exp(0.935-(1.0)(0.260)(-6.91)) = 15.35 IP3 T' (Surge line) = 1.0 (T ~ 200QC)
T' (Charging nozzle) = 1.0 (T ~ 200QC)
T' (SI nozzle) = 1.0 (T ~ 200QC)
T' (RHR piping) = 1.0 (T ~ 200QC) 0' (All) = 0.260 RCS dissolved oxygen is s 50 ppb t '(all) = -6.91 Assume bounding strain rate Fen (Surge line) = exp(0.935-(1.0)(0.260)(-6.91)) = 15.35 Fen (Charging nozzle) = exp(0.935-(1.0)(0.260)(-6.91)) = 15.35 Fen (SI nozzle) = exp(0.935-(1.0)(0.260)( -6.91)) = 15.35 Fen (RHR piping) = exp(0.935-(1.0)(O.260)(-6.91)) = 15.35 RAI 4.3.1.8-2 Standard Review Plan for Review of License Renewal Applications for NuClear Power Plants (SRP-LR) Section 4.3.2.1.1.3 provides the basis for the staff acceptance of an aging management program to address environmental fatigue. It states, "[t]he staff has evaluated a program for monitoring and tracking the number of critical thermal and pressure transients for the selected reactor coolant system components. The staff has determined that this program is an acceptable aging management program to address metal fatigue of the reactor coolant system components according to 10 CFR 54.21 (c)(1 )(iii)." The staff is unable to determine if the Fatigue Monitoring Program of Indian Point 2 and Indian Point 3 contain sufficient details to satisfy this criterion. Please provide adequate details of the Fatigue Monitoring Program such that the staff can make a determination based on the criterion set forth in SRP-LR Section 4.3.2.1.1.3. Also, please explain in detail the corrective actions and the frequency that such actions will be taken so the acceptance criteria will not be exceeded in the period of extended operation.
Response to RAI 4.3.1.8-2 The IPEC Fatigue Monitoring Program was compared to the program described in NUREG-1801 (GALL),Section X.M1, Metal Fatigue of Reactor Coolant Pressure Boundary. The program description in the GALL report is directly applicable to the IPEG units. As indicated in LRA Section B.1.12, during the period of extended operation the IPEC program will be consistent with the GALL program with one exception. The exception to GALL is that rather than performing periodic updates of CUF calculations, IPEC periodically assesses the number of transient cycles compared to calculation assumptions and updates the CUF calculations, if
,I
, I
, I NL-OS-S4 Attachment I Docket Nos. 50-247 & 50-286 Page 4 of 5 necessary. Based on this comparison including evaluation of the exception, the approvals set forth in the GALL report apply to the IPEC Fatigue Monitoring Program.
The program description was modified per letter NL-08-21, Indian Point Nuclear Generating Units Nos. 2 & 3 License Renewal Application Amendment 2, dated January 22, 2008. This letter commits to complete CUF calculations for all areas identified in NUREG-6260 (LRA Table
.4.3-13 for I P2 and LRA Table 4.3-14 for IP3), incorporating the effect of the reactor coolant environment, for IP2 and IP3. Once these CUF calculations are complete (at least 2 years prior to the period of extended operation), IPEC will ensure that the cycles analyzed in the new or updated calculations are included in the Fatigue Monitoring Program. IPEC will continue to manage the effects of fatigue throughout the period of extended operation by monitoring cycles incurred and assuring they do not exceed the analyzed numbers of cycles.
As required by IPEC technical specifications, the Fatigue Monitoring Program tracks actual plant transients and evaluates these against the design transients. The plant transient counts are updated at least once each operating cycle. This frequency is acceptable since the evaluation during each update determines if the number of design transients could be exceeded prior to the next update. The Fatigue Monitoring Program ensures that the numbers of transient cycles experienced by the plant remain within the analyzed numbers of cycles and hence, the, componentCUF calculations remain valid. '
The program requires corrective actions before exceeding the analyzed number of transient cycles. The corrective actions are implemented in accordance with the IPEC corrective action program. IPEC may perform further reanalysis if cycle counts approach analyied numbers.
These calculation updates will be governed by Entergy's 10 CFR 50 Appendix B Quality Assurance (QA) program and include design input verification and independent reviews ensuring that valid assumptions, transients, cycles, external loadings, analysis methods, and environmental fatigue life correction factors will be used in the fatigue analyses. Repair or replacement of the affected component(s), if necessary, will be done prior to exceeding the allowable CUF in accordance with' established plant procedures governing repair and replacement activities. These established procedures are governed by Entergy's 10 CFR 50 Appendix B QA program and meet the applicable repair or replacement requirements of the ASME Code Section XI.
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NL-08-84 Attachment I Docket Nos. 50-247 & 50-286 Page 5 of 5 RAI 3.0.3.2.3-1 Indian Point 2 Updated Final Safety Analysis Report, Revision 20, dated 2006, Section 14.2.1 on page 55 of 218, states in part that:
"Northeast Technology Corporation report NET-173-01 and NET-173-02 are based on conservative projections of amount of boraflex absorber panel degradation assumed in each sub-region. These projections are valid through the end of the year 2006."
Please confirm that the Boraflex neutron absorber panels in the Indian Point Unit 2 spent fuel pool have been re-evaluated for service through the end of the current licensing period. Also, please discuss the plans for updating the Boraflex analysis during the period of extended operation.
Response to RAI3.0.3.2.3-1 Boron-10 areal density gage for evaluating racks (BADGER) testing was performed in February 2000, July 2003, and again in July 2006. Using the latest test data and RACKLIFE code projections, the Boraflex neutron absorber panels in the Indian Point Unit 2 spent fuel pool will meet the Technical Specification requirements through the end of the current licensing period.
The next BADGER test will be performed prior to the end of calendar year 2009. As requiredby the Boraflex Monitoring Program (LRA Section B.1.3), periodic BADGER testing and RACKLIFE code projections will continue through the period of extended operation to confirm acceptable Boraflex condition.
The referenced section of the Indian Point 2 Updated Final Safety Analysis Report will be updated in the next revision.