DCL-10-131, Response to NRC Letter Dated September 23, 2010, Request for Additional Information (Set 25) for the Diablo Canyon License Renewal Application
| ML102950069 | |
| Person / Time | |
|---|---|
| Site: | Diablo Canyon  |
| Issue date: | 10/21/2010 |
| From: | Becker J Pacific Gas & Electric Co |
| To: | Office of Nuclear Reactor Regulation, Document Control Desk |
| References | |
| PG&E Letter DCL-10-131 | |
| Download: ML102950069 (21) | |
Text
Pacific Gas and Electric Company James R. Becker Diablo Canyon Power Plant Site Vice President Mail Code 104/ 5/601
- p. O. Box 56 Avila Beach, CA 93424 805.545.3462 Internal: 691.3462 October 21,2010 Fax: 805.545.6445 PG&E Letter DCL-10-131 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20852 Docket No. 50-275, OL-DPR-80 Docket No. 50-323, OL-DPR-82 Diablo Canyon Units 1 and 2 Response to NRC Letter dated September 23, 2010, Request for Additional Information (Set 25) for the Diablo Canyon License Renewal Application
Dear Commissioners and Staff:
By letter dated November 23, 2009, Pacific Gas and Electric Company (PG&E) submitted an application to the U.S. Nuclear Regulatory Commission (NRC) for the renewal of Facility Operating Licenses DPR-80 and DPR-82, for Diablo Canyon Power Plant (DCPP) Units 1 and 2, respectively. The application included the license renewal application (LRA), and Applicant's Environmental Report-Operating License Renewal Stage.
By letter dated September 23, 2010, the NRC staff requested additional information needed to continue their review of the DCPP LRA.
PG&E's response to the request for additional information is included in Enclosure 1. Enclosure 2 is a summary of errata changes. LRA Amendment 21 resulting from the responses and errata is included in Enclosure 3 showing the changed pages with line-inlline-out annotations.
PG&E revises a commitment in revised LRA Table A4-1, License Renewal Commitments, shown in Enclosure 3.
If you have any questions regarding this response, please contact Mr. Terence L.
Grebel, License Renewal Project Manager, at (805) 545-4160.
A member of the STARS (Strategic Teaming and Resource Sharing) Alliance Callaway . Comanche Peak. Diablo Canyon. Palo Verde. San Onofre. South Te x as Project. Wolf Creek
Document Control Desk PG&E Letter DCL-10-131 October 21, 2010 Page 2 I declare under penalty of perjury that the foregoing is true and correct.
Executed on October 21, 2010.
Sinc rely,
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Site Vice Presi ent tlg/50344198 Enclosure cc: Diablo Distribution cc/enc: Elmo E. Collins, NRC Region IV Regional Administrator Nathanial Ferrer, NRC Project Manager, License Renewal Kimberly J. Green, NRC Project Manager, License Renewal Michael S. Peck, NRC Senior Resident Inspector Fred Lyon, NRC Project Manager, Office of Nuclear Reactor Regulation
Enclosure 1 PG&E Letter DCL-10-131 Page 1 of 4 PG&E Response to NRC Letter dated September 23, 2010 Request for Additional Information (Set 25) for the Diablo Canyon License Renewal Application RAI 4.1-3 License Renewal Application (LRA) Section 4.3.2.5 states that the cumulative usage factor (CUF) calculations for the steam generator (SG) tubes do not need to be identified as a time limited aging analysis (TLAA) because the applicant is required to perform inservice inspections (ISI) of the SG tubes to comply with Technical Specification, and that as a result of this activity, the CUF calculations for the SG tubes do not serve a relevant safety basis.
The staff noted that the safety basis for performing the ISIs of the SG tubes is based on 10 CFR 50.55a and Technical Specification ISI requirements. In contrast, the safety basis for performing the CUF calculation for the SG tubes is based on ASME Section III requirements. FSAR Table 5.2-2 indicates that ASME Section III 1965 inclusive of Winter 1965 Addenda are the applicable design codes. Therefore the applicant is required to comply with 10 CFR 50.55a requirements in accordance with ASME Section III.
Provide and justify a basis for concluding that the CUF calculation for the SG tubes does not serve a safety basis and does not need to be identified as a TLAA when considering that this analysis was performed to comply with ASME Section III design requirements for the SG tubes.
PG&E Response to RAI 4.1-3 As discussed in License Renewal Application (LRA) Section 4.3.2.5, the steam generators (SGs) were replaced during Unit 1 Refueling Outage 15 (1R15, Spring 2009) and the Unit 2 SGs were replaced during Unit 2 Refueling Outage 14 (2R14, Spring 2008). The replacement SG fatigue analysis for the tubes was for 50 years of operation; therefore, the fatigue analysis of the tubes is valid through the period of extended operation in accordance with 10 CFR 54.21(c)(1)(i). See the revised LRA Sections 3.1.2.2.1 and 4.3.2.5 and Tables 3.1.1, 3.1.2-4, and 4.3-7 in Enclosure 3.
Enclosure 1 PG&E Letter DCL-10-131 Page 2 of 4 RAI 4.1-4
Background:
LRA Section 4.3.2.10 states that the reactor coolant pumps (RCPSs) are Westinghouse Model 93A fabricated from SA351 CF8 cast stainless steel. The applicant further stated that the casings are required to be inspected per ASME Section XI, Table IWB-2500-1. The applicant further indicated that ASME Code Case N-481 allows the replacement of the volumetric examination of the primary loop pump casing with a fracture mechanics based integrity evaluation supplemented by specific visual inspections. The applicant stated that the generic faulted screening loads in WCAP-13045 were not bounding. The applicant further stated that a plant-specific Westinghouse analysis, WCAP-13895, re-performed portions of the WCAP-13045 that were not bounding for the applicants design. The applicant stated that this analysis is not a TLAA because (1) the referenced material was fully aged with fracture toughness properties at the saturation levels and (2) fatigue crack growth analysis was not considered in the WCAP-13895 because Code Case N-481 does not require a fatigue crack growth analysis.
ASME Code Case N-481 is endorsed in Regulatory Guide 1.147, as referenced for acceptability in 10 CFR 50.55a. The staff noted that to use the alternative inspection criteria provision (d) of the Code Case mandated that an applicant would need to perform an evaluation of their RCP casing that addresses all following technical considerations:
(1) evaluates material properties, including fracture toughness values; (2) performs a stress analysis of the pump casing; (3) includes a review of the operating history of the pump casing; (4) selects locations for postulating flaws; (5) postulates the occurrence of a one-quarter thickness flaw with an aspect ratio of 6:1; (6) establishes the stability of the selected flaw under governing stress conditions; and (7) considers thermal aging embrittlement and any other processes that may degrade the properties of the pump casings during service.
The staff noted that provision (e) of the Code Case mandated that a report of the evaluation be submitted to the NRC for review.
Issue: The staff has confirmed that the plant-specific analysis, WCAP-13895, does not include a fatigue flaw growth analysis of the pump casings. The staff noted that the lack of a plant-specific fatigue flaw growth analysis in WCAP-13895 does not invalidate the applicability of the generic fatigue flaw analysis in WCAP-13045 to the applicants current licensing basis (CLB) because the scope of WCAP-13895 identified that generic portions of WCAP-13045 not replaced by the plant-specific report were still applicable.
Therefore, the lack of a plant-specific fatigue flaw growth analysis in WCAP-13895 does not constitute a valid basis for concluding that there are not any TLAAs associated with
Enclosure 1 PG&E Letter DCL-10-131 Page 3 of 4 use of Code Case N-481 because the generic fatigue flaw growth analysis in WCAP-13045 may still be applicable to the applicants CLB and may need to be identified as a TLAA for the LRA.
Request: Justify your basis for not identifying the generic fatigue flaw growth analysis in Chapter 9.0 of WCAP-13045 as a TLAA.
PG&E Response to RAI 4.1-4 The fatigue flaw growth analysis for reactor coolant pumps (RCPs) is a time limited aging analysis (TLAA). The design basis number of transients used in the fatigue crack growth analysis for Model 93A RCPs in WCAP-13045 will be managed for the period of extended operation by the Diablo Canyon Power Plant Metal Fatigue of Reactor Coolant Pressure Boundary Program. Action limits will permit completion of corrective actions before the design basis number of events is exceeded. The TLAA will be managed for the period of extended operation, in accordance with 10 CFR 54.21(c)(1)(iii). See revised License Renewal Application Section 4.3.2.10, Table 4.1-1, and Appendices A4-1, A3.2.1.7 and A3.2.1.8 in Enclosure 3.
Enclosure 1 PG&E Letter DCL-10-131 Page 4 of 4 RAI 4.1-5
Background:
10 CFR 54.21(c)(2) requires an applicant to provide a list of all exemptions that have been granted pursuant to 10 CFR 50.12 and that are based on a TLAA. LRA Section 4.8 provides the applicants list of exemptions that need to be identified in accordance with10 CFR 54.21(c)(2). The applicant identified leak-before-break (LBB) as the only exemption that was granted based on a TLAA.
By letter dated May 3, 1999, the staff issued Pacific Gas and Electric Company (PG&E) an exemption under 10 CFR 50.12 granting PG&E the right to apply ASME Code Case N-514 as the basis for establishing the low temperature over-pressurization protection (LTOP) system pressure lift and arming temperature set points for the credited power operated relief valves (PORVs). It also granted the use of the Code Case as a basis for setting the LTOP system pressure lift set points for the relief valves to a pressure value that is equivalent to 110 percent of the limiting pressure established in the approved P-T limits curve for the systems temperature enable set point. The staff noted that the exemption granting the use of the Code Case also permitted the applicant to set the arming temperature for the LTOP system in accordance with the Code Case N-514 arming temperature setpoint methodology.
LRA Section 4.2.3 identifies that the P-T limits for Units 1 and 2 are TLAAs for the LRA.
The staff noted that the LTOP system set points and P-T limits are currently updated according to NRC-approved Pressure Temperature Limits Report (PTLR) and that the current version is PTLR-1, Revision 9.
Issue: The staff noted that granting this exemption and the establishment of the LTOP system pressure setpoint was a function of the limiting pressure value established in the P-T limit curves for the LTOP system enable temperature. The staffs position is that, if this exemption remains in effect for the CLB, the exemption may need to be identified as an exemption for the LRA that meets the requirements in 10 CFR 54.21(c)(2) because granting the exemption under 10 CFR 50.12 was based on a value in the approved P-T limits and the P-T limits for the facilities have been identified as a TLAAs for the LRA.
Request: Clarify whether the exemption on Code Case N-514 remains in effect for the CLB. If the exemption on Code Case N-514 is still in effect, provide your basis for not identifying this exemption in accordance with 10 CFR 54.21(c)(2).
PG&E Response to RAI 4.1-5 The 10 CFR 50.12 exemption which allows the use of ASME Code Case N-514 remains in effect. The License Renewal Application (LRA) has been revised to identify this exemption in accordance with 10 CFR 54.21(c)(2). See revised LRA Sections 4.1.4, 4.2.4, and 4.8 in Enclosure 3.
Enclosure 2 PG&E Letter DCL-10-131 Page 1 of 1 PG&E Errata Below is a table of errata identified in the license renewal application. The associated changed pages are included in Enclosure 3.
Errata No Section or Table LRA Revision 21 Appendix A1.15 Change the number of standby capsules that will remain inside the vessel throughout the vessel lifetime from "five" to "four."
PG&E Letter DCL-10-131 Page 1 of 14 LRA Amendment 21 LRA Section RAI Section 3.1.2.2.1 4.1-3 Section 4.1.4 4.1-5 Section 4.2.4 4.1-5 Section 4.3.2.5 4.1-3 Section 4.3.2.10 4.1-4 Section 4.8 4.1-5 Table 3.1.1 4.1-3 Table 3.1.2-4 4.1-3 Table 4.1-1 4.1-4 Table 4.3-7 4.1-3 Appendix A1.15 Errata 21 Appendix A3.2.1.7 4.1-4 Appendix A3.2.1.8 4.1-4 Appendix A4-1 4.1-4 Section 3.1 PG&E Letter DCL-10-131 AGING MANAGEMENT OF REACTOR VESSEL, Page 2 of 14 INTERNALS, AND REACTOR COOLANT SYSTEM Section 3.1.2.2.1 Cumulative Fatigue Damage
[3.1.1.06] Cumulative fatigue damage of steam generator tubes is not a TLAA as defined in 10 CFR 54.3. See Section 4.3.2.5 describes the disposition of TLAAs for steam generator tubes.
Section 4 PG&E Letter DCL-10-131 TIME-LIMITED AGING ANALYSIS Page 3 of 14 Section 4.1.4 Identification of Exemptions The License Renewal Rule requires that an application for a renewed license include a list of plant-specific exemptions granted pursuant to 10 CFR 50.12 and in effect that are based on time-limited aging analyses as defined in §54.3. The applicant shall provide an evaluation that justifies the continuation of these exemptions for the period of extended operation consistent with 10 CFR 54.21(c)(2).
A search of docketed correspondence, the operating license, and the FSAR Update identified and listed all exemptions in effect. Each exemption in effect was then evaluated to determine whether it involved a TLAA as defined in 10 CFR 54.3.
The search found 14 10 CFR 50.12 exemptions currently in effect for DCPP. Of those, only one two exemptions, the use of the leak-before-break (LBB) evaluation of reactor coolant system piping for DCPP, Units 1 and 2, and the use of ASME Code Case N-514 to establish the LTOP setpoints for Units 1 and 2 is are based in part on a time-limited aging analysis. The LBB analysis is described in Section 4.3.2.12. The use of ASME Code Case N-514 is described in Section 4.2.4.
Section 4 PG&E Letter DCL-10-131 TIME-LIMITED AGING ANALYSIS Page 4 of 14 Section 4.2.4 Pressure - Temperature Limits Low-temperature overpressure protection (LTOP) is provided by the cold overpressurization mitigation system (COMS). The temperature setpoints is are determined by the calculation of the P-T limit curves and in accordance with ASME Code Case N-514. Any changes to the RCS P-T limit curves also require an evaluation of the LTOP enable temperature setpoint and the power operated relief valve (PORV) pressure setpoint, and supporting safety analyses.
Section 4 PG&E Letter DCL-10-131 TIME-LIMITED AGING ANALYSIS Page 5 of 14 Section 4.3.2.5 Steam Generator ASME Section III Class 1, Class 2 Secondary Side, and Feedwater Nozzle Fatigue Analyses and Fatigue Qualification Tests Summary Description Replacement steam generators (RSGs) are installed at DCPP. Unit 2 RSGs were installed in the spring 2008 refueling outage. Unit 1 RSGs were installed in the spring 2009 refueling outage. The design life of the RSGs is a minimum of 50 years, ending in 2058 and 2059, respectively. The period of extended operation will expire in 2044 and 2045, prior to the end of the design life of the RSGs.
The RSGs are designed and fabricated to the requirements of ASME Code Section III, 1998 Edition, with Addenda through 2000. The design specification classifies the primary side of each RSG as ASME Code Class 1, and the secondary side of each RSG as ASME Code Class 2. However, the entire pressure boundary of the component is designed and constructed in accordance with ASME Code Section III, Class 1 requirements.
Analysis The applicable fatigue analyses and fatigue qualification tests of the RSGs are TLAAs.
Fatigue usage factors in the steam generator components depend on effects of transient events specified in the design specification. The Unit 1 and 2 RSG analyses and qualification tests use a 50-year design basis numbers of transients, and are therefore valid beyond the period of extended operation.
Steam Generator Tube Code Fatigue Analysis (Not a TLAA)
The design of the RSGs includes a Code fatigue analysis of the steam generator tubes.
This analysis is not a TLAA because the code fatigue analysis is not used to support the safety determination, in accordance with 10 CFR 54.3(a) Criterion 4.
The RSGs are inspected in accordance with DCPP TS 3.4.17, 5.5.9, and 5.6.10. Tube degradation will be detected during scheduled inservice steam generator (SG) tube examinations to ensure that SG tube integrity is maintained. The SG Program includes provisions for condition monitoring assessments, performance criteria for SG tube integrity, SG tube repair criteria, and SG tube inspections.
Section 4 PG&E Letter DCL-10-131 TIME-LIMITED AGING ANALYSIS Page 6 of 14 4.3.2.10 Absence of TLAA for Thermal Embrittlement of Cast Austenitic Stainless Steel Reactor Coolant Pumps WCAP-13045: Fatigue Crack Growth Analysis Westinghouse did not address the applicability of the WCAP-13045 fatigue crack growth analysis in the DCPP plant-specific analysis, WCAP-13895, because Code Case N-481 does not require a fatigue crack growth analysis. On the basis that a fatigue crack growth analysis is not included in the Code Case N-481 implementing reference WCAP-13895, it is concluded that there is no TLAA, in accordance with 10 CFR 54.3(a),
Criteria 4 and 5.
In the stability analyses cracks are postulated at various locations in the pump casing.
These postulated cracks would be subject to the various conditions the pump casing experiences. A fatigue crack growth evaluation was performed for the high stress outlet nozzle crotch regions of the Model 93A pump casing to ensure reasonably sized flaws would not propagate beyond the postulated flaw sizes specified by Code Case N-481 and shown to be stable. The analysis deemed that fatigue crack growth was small. The reasonably-sized flaws used in the analysis are equal to and in excess of the maximum allowable crack depth of 0.3 inch given in Table IWB 3518-2 of Section XI of the ASME Code (1989 Edition). The fatigue crack growth analysis loadings included deadweight, pressure, and normal and upset loadings that were based on 40-year design life.
Disposition: Aging Management, 10CFR54.21(c)(1)(iii)
The design basis number of transients used in the fatigue crack growth analysis for Model 93A RCPs in WCAP-13045 will be managed for the period of extended operation by the DCPP Metal Fatigue of Reactor Coolant Pressure Boundary program, which is summarized in LRA Sections 4.3.1 and B3.1. Action limits will permit completion of corrective actions before the design basis number of events is exceeded. These effects will be managed for the period of extended operation, in accordance with 10 CFR 54.21(c)(1)(iii).
Section 4 PG&E Letter DCL-10-131 TIME-LIMITED AGING ANALYSIS Page 7 of 14 4.8 TLAAS SUPPORTING 10 CFR 50.12 EXEMPTIONS Pursuant to 10 CFR 54.21(c)(2), this section identifies the plant-specific 10 CFR 50.12 exemptions for DCPP that are currently in effect; identifies those based on a TLAA; and evaluates the TLAA-based exemptions for the period of extended operation. The NUREG-1800 requires a review of these exemptions to determine whether any are supported by TLAAs.
From a search of the current licensing basis, fifteen 10 CFR 50.12 exemptions were identified. Fourteen 10 CFR 50.12 exemptions are currently in effect for DCPP. Of those, only one two exemptions, the use of the Leak-Before-Break (LBB) Evaluation of Reactor Coolant System Piping for DCPP, Unit No. 1 (TAC No. M83283) and Unit No. 2 (TAC No. M83284) dated March 3, 1993, and the use of ASME Code Case N-514 to establish the LTOP system setpoints for Units 1 and 2, dated May 3, 1999;is are based in part on a time-limited aging analysis. The LBB analysis is described in Section 4.3.2.12. The use of ASME Code Case N-514 is described in Section 4.2.4.
The disposition of this exemption is dependent on the disposition of the LBB analysis presented in Section 4.3.2.12.
Section 3.1 PG&E Letter DCL-10-131 AGING MANAGEMENT OF REACTOR VESSEL, Page 8 of 14 INTERNALS, AND REACTOR COOLANT SYSTEM Table 3.1.1 Summary of Aging Management Evaluations in Chapter VII of NUREG-180 for Reactor Vessel, Internals, and Reactor Coolant System Aging Effect / Aging Management Further Evaluation Item Number Component Type Discussion Mechanism Program Recommended Fatigue of metal components is a Nickel Alloy tubes and TLAA.Cumulative sleeves in a reactor TLAA, evaluated in fatigue damage of Cumulative fatigue 3.1.1.06 coolant and secondary accordance with Yes, TLAA steam generator tubes damage feedwater/steam 10 CFR 54.21(c) is not a TLAA as defined environment in 10 CFR 54.3. See further evaluation in Section 3.1.2.2.1.
Section 3.1 PG&E Letter DCL-10-131 AGING MANAGEMENT OF REACTOR VESSEL, Page 9 of 14 INTERNALS, AND REACTOR COOLANT SYSTEM Table 3.1.2-4 Reactor Vessel, Internals, and Reactor Coolant System - Summary of Aging Management Evaluation -
Steam Generators Aging Effect Component Intended Aging Management NUREG-1801 Table 1 Material Environment Requiring Notes Type Function Program Vol. 2 Item Item Management Time-Limited Aging Reactor Coolant Cumulative Analysis evaluated for SG Tubes TH, PB Nickel Alloy IV.D1-21 3.1.1.06 A (Int) fatigue damage the period of extended operation Time-Limited Aging Secondary Water Cumulative Analysis evaluated for SG Tubes TH, PB Nickel Alloy IV.D1-21 3.1.1.06 A (Ext) fatigue damage the period of extended operation Section 4 PG&E Letter DCL-10-131 TIME-LIMITED AGING ANALYSIS Page 10 of 14 Table 4.1-1 List of TLAAs TLAA Disposition Description Section Category Category(a)
- 2. Metal Fatigue Analysis NA 4.3 Absence of a TLAA for Thermal Embrittlement of Cast Austenitic Stainless Steel (CASS) Reactor NAiii 4.3.2.10 Coolant Pumps Section 4 PG&E Letter DCL-10-131 TIME-LIMITED AGING ANALYSIS Page 11 of 14 Table 4.3-7 DCPP Units 1 and 2 Steam Generator Cumulative Fatigue Usage Component Unit 1 & 2 Cumulative Usage Factor Tubes Welds 0.89 Appendix A PG&E Letter DCL-10-131 Final Safety Analysis Report Supplement Page 12 of 14 A1.15 Reactor Vessel Surveillance The Reactor Vessel Surveillance program manages loss of fracture toughness due to neutron embrittlement in reactor materials exposed to neutron fluence exceeding 1.0E17 n/cm2 (E>1.0 MeV). The program is consistent with ASTM E 185-70 and ASTM E 185-73 for Units 1 and 2, respectively. Capsules are periodically removed during the course of plant operating life. Neutron embrittlement is evaluated through surveillance capsule testing and evaluation, ex-vessel neutron fluence calculations, and monitoring of reactor vessel neutron fluence. The testing program and reporting conform to requirements of 10 CFR 50 Appendix H, Reactor Vessel Material Surveillance Program Requirements.
Data resulting from the program is used to:
Determine pressure-temperature limits, minimum temperature requirements, and end-of-life Charpy upper-shelf energy (CV USE) in accordance with the requirements of 10 CFR 50 Appendix G, Fracture Toughness Requirements; and, Determine end-of-life RTPTS values in accordance with 10 CFR 50.61, Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock.
The Reactor Vessel Surveillance program provides guidance for removal and testing or storage of material specimen capsules. All capsules that have been withdrawn and tested were stored.
For Unit 1, the last capsule is expected to be withdrawn during the current operating term after it has accumulated a fluence equivalent to 60 years of operation. The remaining four five standby capsules have low lead factors, will remain inside the vessel throughout the vessel lifetime, and will be available for future testing.
There are no capsules remaining in the Unit 2 vessel. All capsules were removed because high lead factors produced exposures comparable to the fluences expected at the end of the period of extended operation.
DCPP Units 1 and 2 currently use ex-vessel monitoring dosimetry, which consists of four gradient chains with activation foils outside the reactor vessel, which will be used to monitor the neutron fluence environment within the beltline region.
Appendix A PG&E Letter DCL-10-131 Final Safety Analysis Report Supplement Page 13 of 14 A3.2.1.7 Thermal Embrittlement of Cast Austenitic Stainless Steel Reactor Coolant Pumps A fatigue crack growth evaluation was performed for the high stress outlet nozzle crotch regions of the Model 93A pump casing. The transients used in the fatigue crack growth analysis will be managed for the period of extended operation by the DCPP Metal Fatigue of Reactor Coolant Pressure Boundary program, which is summarized in Section A2.1. The TLAA will be managed for the period of extended operation, in accordance with 10 CFR 54.21(c)(1)(iii).
A3.2.1.7 8 TLAAs in Fatigue Crack Growth Assessments and Fracture Mechanics Stability Analyses for Leak-Before-Break Elimination of Dynamic Effects of Primary Loop Piping Failures Appendix A PG&E Letter DCL-10-131 Final Safety Analysis Report Supplement Page 14 of 14 Table A4-1 License Renewal Commitments Item # Commitment LRA Implementation Section Schedule 38 The actual plant transient cycles related to the SWOL and Model 93A Reactor 4.3 Prior to January Coolant Pumps fatigue crack growth analyses will be included in the existing 31,2011 plant transient monitoring program by January 31, 2011 to ensure that the actual plant transients do not exceed the SWOL fatigue analysis limits.