ML101320262

Supplemental Response to Request for Additional Information B3.1-4 Regarding Analyzed Design Basis Transients, and License Renewal Application (LRA) Amendment No. 14

ML101320262
Person / Time
Site:	Palo Verde
Issue date:	04/28/2010
From:	Hesser J Arizona Public Service Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
102-06175-JHH/GAM
Download: ML101320262 (83)

Text

{{#Wiki_filter:1, A M A subsidiaty of Pinnacle West Capital Corporation John H. Hesser Mail Station 7605 Palo Verde Nuclear Vice President Tel. 623-393-5553 P. 0. 'Box 52034 Generating Station Nuclear Engineering Fax 623-393-6077 Phoenix, Arizona 85072-2034 102-06175-JHH/GAM April 28, 2010 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001

Subject:

Palo Verde Nuclear Generating Station (PVNGS) Units 1, 2, and 3 Docket Nos. STN 50-528, 50-529 and 50-530 Supplemental Response to Request for Additional Information (RAI) B3.1-4 Regarding Analyzed Design Basis Transients, and License Renewal Application (LRA) Amendment No. 14 By letter dated December 29, 2009, the NRC issued a request for additional information (RAI) related to the PVNGS license renewal application (LRA). Arizona,Public Service Company (APS) submitted a response to the RAI in letter no. 102-06134, dated February 19, 2010. A supplement to the APS response to RAI B3.1-4 is provided in to replace the response provided in the February 19, 2010, submittal. This supplemental RAI response, along with associated LRA Section 4.3.1 changes in Amendment No. 14 in Enclosure 2, is intended to clarify and correct LRA Section 4.3.1 as discussed with the NRC staff. Markup pages showing the changes to the text portions of LRA Section 4.3.1 are provided in Enclosure 3. Conforming changes to other affected LRA sections to reflect these Section 4.3.1 changes will be submitted by May 28, 2010. Commitment No. 39 in LRA Table A4-1 is being revised as shown in Enclosure 2. In addition, a new Commitment No. 55 is being added to Table A4-1 as shown in for the following: The transient in UFSAR Table 3.9-1 Sheet No. 9 Item No. I.E.1.b, and Sheet No. 18, Item No. IIl.A.1.f, "Standby to SI hot leg injection check valve stroke test tostandby (using the HPSI pump)," will be added to the cycle counting surveillance procedure 73ST-9RC02 by August 25, 2010. A member of the STARS (Strategic Teaming and Resource Sharing) Alliance Callaway Comanche Peak

• Diablo Canyon
• Palo Verde
• San Onofre
• South Texas ° Wolf Creek

.4

ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Supplemental Response to Request for Additional Information (RAI) B3.1-4 Regarding Analyzed Design Basis Transients, and License Renewal Application (LRA) Amendment No. 14 Page 2 Should you need further information regarding this submittal, please contact Russell A. Stroud, Licensing Section Leader, at (623) 393-5111. I declare under penalty of perjury that the foregoing is true and correct. Executed on (date) JHH/RAS/GAM

Enclosures:

1.

Supplemental Response to Request for Additional Information (RAI) B3.1-4 Regarding Analyzed Design Basis Transients

2.

Palo Verde Nuclear Generating Station License Renewal Application Amendment No. 14

3.

Markup Showing Changes to LRA Section 4.3.1 cc: E. E. Collins Jr. J. R. Hall L. K. Gibson R. I. Treadway L. M. Regner G. A. Pick NRC Region IV Regional Administrator NRC NRR Project Manager NRC NRR Project Manager NRC Senior Resident Inspector for PVNGS NRC License Renewal Project Manager NRC Region IV (electronic)

ENCLOSURE 1 Supplemental Response to Request for Additional Information (RAI) B3.1-4 Regarding Analyzed Design Basis Transients

Enclosure I Supplemental Response to Request for Additional Information (RAI) B3.1-4 Regarding Analyzed Design Basis Transients NRC RAI B3.1-4 (From NRC letter to APS dated December 29, 2009) Backgqround: The Detection of Aging Effects element of the Metal Fatigue of Reactor Coolant Pressure Boundary AMP in the GALL Report states that the program provides for periodic update of the fatigue usage calculations. Issue: The Program Description element states that the LRA, Section 4.3, AMP monitors and tracks the number of critical thermal and pressure transients for the selected reactor coolant system components. Subsection 4.3.1.4, "Present and Projected Status of Monitored Locations," of the LRA states that a composite worst-case (composite-unit) envelope of operating transients was created including only the highest accumulation of each transient experienced among the three units from 1985 through 2005. However, the applicant did not provide individual plant data for each unit that was used to develop the composite-unit envelope. Request: Provide the accumulation of transients for each of the three units that were used to develop the composite-unit envelope for the period from 1985 to 2005. APS Supplemental Response to RAI B3.1-4 Section 4.3.1 of the License Renewal Application (LRA) has been revised and supplements the prior response submitted in APS letter no. 102-06134, dated February 19, 2010. The revision is provided in LRA Amendment No. 14 in Enclosure 2. This revision incorporates the following changes:

1. Revised Table 4.3-2 to clearly show correspondence between the LRA and the Updated Final Safety Analysis Report (UFSAR).

2. Revised Table 4.3-2 to identify transients that are tracked, provided justification for those transients that are not tracked, and clarified UFSAR limits.

3. Changed "global" monitoring to "cycle counting."

4. Simplified the transient projection process and clarified that it is not intended to be used for action.

5. Revised the discussion on how the transient count data was recovered.

1

Enclosure I Supplemental Response to Request for Additional Information (RAI) B3.1-4 Regarding Analyzed Design Basis Transients

6. Revised Table 4.3-3 to be consistent with the UFSAR transients, replaced the worst case unit with actual totals for all three units, and provided the new simplified projections.

7. Revised the location-specific monitoring points. Table 4.3-4 now identifies the NUREG/CR-6260 locations and the pressurizer spray nozzle location.

8. Incorporated miscellaneous clarifications and editorial changes.

2

ENCLOSURE 2 Palo Verde Nuclear Generating Station License Renewal Application Amendment No. 14 LRA Section Page Nos. RAI No. 4.3 4.3-1 through 4.3-42 B3.1-4 Table A4-1, Item 39 A-54 B3.1-4 Table A4-1, Item 55 A-59 B3.1-4

Section 4 TIME-LIMITED AGING ANALYSES 4.3 METAL FATIGUE ANALYSIS This section addresses design of mechanical system components supported by fatigue analyses; and also of components whose design depends on an assumed number of load cycles without a calculated fatigue usage factor. Section 4.6 describes fatigue in the containment vessel. Section 4.7.4, describes corrosion and fatigue crack growth and stability in the primary coolant nozzles. Fatigue analyses are required for piping, vessels, and heat exchangers designed to the American Society of Mechanical Engineers Boiler and Pressure Vessel Code, Section III, Rules for Construction of Nuclear Power Plant Components, Division 1, "Metal Components," Subsection NB, "Requirements for Class 1 Components" (ASME III Class 1).1 Fatigue analyses may also be invoked for Class 1 pump and valve pressure boundaries. Fatigue analyses are required for portions of the reactor pressure vessel internals designed to American Society of Mechanical Engineers Boiler and Pressure Vessel Code, Section III, Rules for Construction of Nuclear Power Plant Components, Division 1, "Metal Components," Subsection NG, "Core Support Structures." The design of piping and vessels to certain other codes and code sections, including ASME III Class 2 and 3, ANSI-ASME B31.1, and ASME VIII Division 2, may assume a stated number of full-range thermal and displacement cycles. Section 4.3 also describes fatigue analyses and evaluations of a limited number of other non-Class 1 components that were evaluated to these and similar rules. Basis of Fatigue Analyses ASME III Class 1 design specifications define a set of static and transient load conditions for which components are to be designed. Although original design specifications commonly state that the transient conditions are for a 40-year design life, the fatigue analyses themselves are based on the specified number of occurrences of each transient rather than on this lifetime. The design number of occurrences of each transient for use in the fatigue analyses was specified to be larger than the number of occurrences expected during the 40-year licensed life of the plant, based on engineering experience and judgment. This provides an allowance for future changes in design or operation that may affect system design transients. Titles are from the 1971 edition of the code, as used for the reactor vessel. Later editions reorganized the Section III material and removed the Division 1 title, so that this subsection became "Division 1 - Subsection NB, Class 1 Components". Palo Verde Nuclear Generating Station Amendment 14 Page 4.3-1 License Renewal Application

Section 4 TIME-LIMITED AGING ANALYSES Operating experience at PVNGS and at other similar units has demonstrated that the assumed frequencies of design transients, and therefore the number of transient cycles assumed for a 40-year life, were conservative; and that with few exceptions the design numbers are not expectedto be exceeded within a 60-year life. The exceptions are of two kinds. First, the NRC, industry, and specific plants, including PVNGS, have identified some transient loads on some components that were not foreseen in the original design process; for example thermally stratified flow in the pressurizer surge line and feedwater system, and Combustion Engineering Infobulletin 88-09. These cases have required evaluations to assess their significance and some have required revision to design specifications and analyses. Second, plant and industry operating experience has identified a few cases where cycles were being accumulated more rapidly than originally anticipated. At PVNGS, these were principally due to first-of-a-kind startup and shutdown cycles during the early plant life. Fatigue cycles are currently tracked in a PVNGS surveillance test procedure, 73ST-9RC02 "Reactor Coolant System Transient and Operational Cycles," as required by PVNGS Technical Specification 5.5.5 "Component Cyclic or Transient Limit." In the text of this discussion the activities governed by 73ST-9RC02 will be referred to as the "current fatigue monitoring program." The enhanced Metal Fatigue of Reactor Coolant Pressure Boundary program (See LRA Appenbix B section B3.1) will continue to track events to ensure that appropriate re-evaluation or other corrective action is initiated if an action limit is reached. Action limits will permit completion of corrective actions before the design basis number of events is exceeded. See Section 4.3.1.5. The Industry Operating Experience Review (OE) program ensures that industry experience is evaluated and incorporated in plant analyses and procedures. The OE program includes review of experience that may indicate concerns with fatigue effects. Any necessary evaluations are conducted under the plant corrective action program. The OE program has remained responsive to both industry and plant-specific emerging issues and concerns. 4.3.1 Enhanced Fatigue Aging Management Program (B3.1) The current fatigue monitoring program is a cycle counting program with one location-specific cumulative usage factor (CUF) calculation (Pressurizer Spray Nozzle). No later than two years prior to the period of extended operation, the current fatigue 'monitoring program governed by 73ST-9RC02 will be enhanced to include additional location-specific CUF calculations and an automated and computerized management software program for cycle counting and fatigue usage factor tracking. The automated and computerized software program will be used to supplement manual counting. The enhanced Metal Fatigue of Reactor Coolant Pressure Boundary program (B3.1) will use cycle counting (CC), cycle based fatigue CUF calculations (CBF) and stress based fatigue CUF calculations (SBF) (see methods discussion below) to monitor fatigue. FatiguePro will Palo Verde Nuclear Generating Station Amendment 14 Page 4.3-2 License Renewal Application

Section 4 TIME-LIMITED AGING ANALYSES be used for cycle counting and cycle-based fatigue (CBF) monitoring methods. FatiguePro is an EPRI licensed product. APS commits to the use of a fatigue monitoring software program that incorporates a three-dimensional, six-element stress tensor method meeting ASME III NB-3200 requirements for stress-based fatigue monitoring (SBF). APS also commits to the implementation of this method for SBF monitoring at least two years prior to the period of extended operation. The enhanced Metal Fatigue of Reactor Coolant Pressure Boundary program (B3.1) will monitor plant transients as required by PVNGS Technical Specification 5.5.5. Cumulative usage factors (CUFs) will be calculated for a subset of ASME III Class 1 reactor coolant pressure boundary vessel and piping locations, and component locations with Class 1 analyses. The enhanced Metal Fatigue of Reactor Coolant Pressure Boundary program (B3.1) will provide action limits on cycles and on CUF that will initiate corrective actions before the licensing basis limits on fatigue effects at any location are exceeded. Scope The scope of the enhanced Metal Fatigue of Reactor Coolant Pressure Boundary program (B3.1) will include all ASME Section III Class 1 components and components with Class 1 fatigue analysis and will monitor their fatigue usage by a combination of cycle counting and CUF monitoring. Methods The Cycle Counting (CC) monitoring method in Table 4.3-4 means that the enhanced Metal Fatigue of Reactor Coolant Pressure Boundary program (B3.1) will not periodically calculate accumulated fatigue usage at the location. However, transient event cycles affecting the location (e.g. plant heatup and plant cooldown) will be counted and tracked to ensure that the numbers of transient events assumed by the design basis calculations will not be exceeded. Cycle counting is the method used by the current fatigue monitoring program for all monitored components except the pressurizer spray nozzle. Cycle counting ensures fatigue usage does not exceed 1.0. It is employed as the preferred method of monitoring due to its simplicity. Cycle-based fatigue (CBF) monitoring will consist of (a) automated cycle counting; supported as needed by manual data entry for infrequent events, and (b) CUF computation based on the counted cycles. The component CUF contributions due to each cycle are determined from the component Class I fatigue analysis. CBF is a more complex and resource intensive method than CC because it goes beyond counting of cycles to evaluate the CUF contributions of each cycle. Three CBF methods will be used, Per-Cycle CBF (CBF-C), Per-Cycle CBF with partial cycles (CBF-PC), and Event-Pairing CBF (CBF-EP). The CBF-C and CBF-PC methods will compute fatigue usage for a component by determining a location-specific fatigue usage increment for each counted event, and then adding up those increments for all events in the cycle record. CBF-PC will be used for some Palo Verde Nuclear Generating Station Amendment 14 Page 4.3-3 License Renewal Application

Section 4 TIME-LIMITED AGING ANALYSES components, where the fatigue severity of individual plant events can be scaled using partial-cycle curves. A partial-cycle curve assigns a fractional severity to a cycle, as compared to a full design cycle, based upon significant characteristics of that event, such as temperature difference or heatup rate. CBF-EP is derived from the application of Miner's rule for combining fatigue effects, under the guidance of ASME III, NB-3222.4. This method will use an event-pairing table which assumes that the effect of pairs of monitored events is equal to the effect of similar pairs of design basis events. Stress-based fatigue (SBF) monitoring will compute a "real time" stress history for a given component from actual temperature, pressure, and flow histories. SBF monitoring uses data collected from existing plant instruments to calculate local pressure and temperature, and the corresponding stress history at the critical location in the component. The stress history is analyzed to identify stress cycles, and then a CUF is computed using the formulas defined in ASME Code Section III sub-article NB-3200. SBF is the most complex and resource intensive method of fatigue monitoring, but it is the most accurate method and requires fewer conservative assumptions than CC or CBF methods. Corrective Action Limits The enhanced Metal Fatigue of Reactor Coolant Pressure Boundary program (B3.1) corrective actions will be initiated whenever an action limit is reached, for either the number of transient cycles or calculated fatigue usage factor. In the current fatigue monitoring program cycle action limits are set at 90% of the allowed cycles for each transient, and a CUF action limit of 0.65 is set for the pressurizer spray nozzle. In the enhanced Metal Fatigue of Reactor Coolant Pressure Boundary program (B3.1) corrective action limits will be set to permit completion of corrective actions before the design basis number of events is exceeded, or before the cumulative usage factor exceeds the code limit of 1.0. See Section 4.3.1.5 for the description of these actions and action limits, for the basis for the margins between the fatigue usage factor action limits and the code usage factor limit of 1.0, and for the basis for the margins between the cycle count action limits and the design basis cycle count assumptions. Analytical Margins Fatigue analyses incorporate several conservative assumptions and methods. These ensure that usage factors predicted by the design calculation will exceed (or "bound") the usage factors actually accumulated by the components. These conservatisms are discussed below. Fatigue Desiqn Curve with Margin for Uncertainties and Moderate Environmental Effects: The ASME Section III fatigue S-N curves (allowable alternating stress intensity versus number of cycles) are based on regression analysis of a large number of fatigue data points for samples strain-cycled in air, with adjustments for the elastic modulus and departure from zero mean stress for elastic cycling, less a design margin for uncertainties, including modest Palo Verde Nuclear Generating Station Amendment 14 Page 4.3-4 License Renewal Application

Section 4 TIME-LIMITED AGING ANALYSES environmental effects (ASME III - 1965, Par. N-415). The design margin is a factor of 2 on stress or a factor of 20 on cycles, whichever produced the lower, more conservative allowable for the data set. Bounding Parameters for Transients: Fatigue analyses assume a given number of cycles of each of a set of transient events. Actual event cycles are seldom as severe as those considered in the analysis; the resulting stress ranges are lower, and the contributions to cumulative usage factor are therefore lower. Use of Stress Based Fatigue: Since an automated six-element stress tensor fatigue calculation will calculate stresses from the actual event severity, usage factors reported by the software program at locations for which the stress-based method is used will be more realistic than values predicted by the code analysis for the same number of cycles, or which would be determined by cycle-count monitoring. Actual Number of Event Cycles versus Design Number of Cycles: The analytical limit for a fatigue analysis is a cumulative usage factor at any location of 1.0. The design CUF is the sum of all contributing partial usage factors resulting from the total of all design basis events at their design number of occurrences. Therefore, even if the analysis showed a calculated usage factor of 1.0 for a location, and even if the design basis number of cycles were reached for one event, the fact that all contributing cycle types will not simultaneously arrive at their assumed limit indicates that some fatigue margin would remain. Action limits in the enhanced Metal Fatigue of Reactor Coolant Pressure Boundary program (B3.1) will be set below the cycle count assumed by the analysis to ensure that actual component usage remains bounded by the assumptions used in the design calculations, or that appropriate reevaluation or other corrective action is initiated if an action limit is reached. Action limits will permit completion of corrective actions before the design basis number of events is exceeded. Therefore, the enhanced Metal Fatigue of Reactor Coolant Pressure Boundary program (B3.1) will ensure that there is ample margin to the CUF limit of 1.0. 4.3.1.1 Licensing Basis of the PVNGS Component Cyclic or Transient Limit Program The "Component Cyclic or Transient Limit" program is required by Technical Specification 5.5.5, which states: "This program provides controls to track the UFSAR Section 3.9.1.1 cyclic and transient occurrences to ensure that components are maintained within the design limits." UFSAR Section 3.9.1.1 includes, by reference, information and transient definitions from several UFSAR sections and tables, which represent conservative estimates for design purposes (see Table 4.3-1). The UFSAR states that this information accounts for all expected transients, and that the number and severity of the design transients exceeds those which may be anticipated during the 40-year life of the plant. These transients are listed in Table 4.3-2. Palo Verde Nuclear Generating Station Amendment 14 Page 4.3-5 License Renewal Application

Section 4 TIME-LIMITED AGING ANALYSES Table 4.3 PVNGS Unit 1, 2, and 3 Licensing and Design Basis Transient Citations frnm I I1.4ZA/P.4 0 1 1 -ection j.t.j.z uperaing basis -artnguaKe **t-) uycies Table 3.9.1-1 ASME III Class I Components by the NSSS Vendor (CE) Table 3.9-1 ASME III Class 1 Piping Not by the NSSS Vendor (CE) Section 3.9.3 ASME III Class 2 and 3 Components Section 5.4.1 Reactor Coolant Pumps Section 5.4.2 Steam Generators Section 5.4.3 Reactor Coolant Piping Section 5.4.10 Pressurizer 4.3.1.2 Enhanced PVNGS Fatigue Management Program (B3.1) The enhanced Metal Fatigue of Reactor Coolant Pressure Boundary program (B3.1) differs from the current fatigue monitoring program in the following two respects:

1. The current fatigue monitoring program is a manual cycle counting program. The enhanced Metal Fatigue of Reactor Coolant Pressure Boundary program (B3.1) will include an automated and computerized software program to support safe operation of PVNGS for the period of extended operation.

2. The current fatigue monitoring program is cycle based and includes only one specific location for CUF monitoring (pressurizer spray nozzle). The enhanced Metal Fatigue of Reactor Coolant Pressure Boundary program (B3.1) will continue to count cycles

-and will also monitor CUF values as specified in Table 4.3-4 for bounding locations subject to environmentally assisted fatigue (locations identified through implementation of NUREG/CR-6260). Usage factor monitoring will include environmental effects at NUREG/CR-6260 locations. Palo Verde Nuclear Generating Station License Renewal Application Amendment 14 Page 4.3-6

Section 4 TIME-LIMITED AGING ANALYSES Table 4.3 PVNGS Units 1, 2, and 3 Licensing Basis Transients 1 2 3 4 5 6 8 Row TransientTitle Limiting UFSAR UFSAR Other UFSAR Notes No. (Shaded-items Value Table Table 3.9-1 UFSAR

Category,

-are not counted) 3.9.1-1 (Sheet No. Reference -Item No.) 1 Reactor Coolant System 500 (The Sheet 1 1-1.A.l.a Normal (excluding pressurizer) heatup reactor 13-11.B.1.a from 70F to hot standby vessel studs 21-Note

• conditions at a rate of shall be

<100F/hr. limited to 250 occurrences 2 Reactor Coolant System 500 (The Sheet 1 1-I.A. 1.d Normal (excluding pressurizer) reactor 13-11.B.13 .a cooldown from hot standby vessel studs 21-Note* conditions to 70F at a rate of shall be <100F/hr limited to 250 occurrences) 3 5%/minute power ramp 15000 Sheet 1 1-I.A.l.b Normal To reach this number the plant would increase, from '15% to 100% 13-11.B.1.b have to experience a power increase on power the average of once every 31.5 hours of operation in 60 years with a 90% capacity factor. Since the PVNGS units operate as base loaded plants this is not credible. This item is not counted. 4 5%/minute power ramp 15000 Sheet 1 1-I.A.l.c Normal To reach this number the plant would decrease, from 100% to 15% 13-11.B.11.b have to experience a power decrease power on the average of once every 31.5 hours of operation in 60 years with a 90% capacity factor. Since the PVNGS units operate as base loaded plants this is not credible. This item is not counted. Palo Verde Nuclear Generating Station License Renewal Application Amendment 14 Page 4.3-7

Section 4 TIME-LIMITED AGING ANALYSES fU-/o 0UI

• t*,. 10%Utpr,

.from 90% to 100% powoejr 6-I.C.l.a 15-11.D.1.a I .1 6 4 I' I 6 110% power step decrease, from 100% to 90% power I 2000 6-1.C.l.b 15-11.D.1.a Normal 7 8 1.E+06 Sheet 1 6-I.C.l.c Normal Per UFSAR Table 3.9.1-1 "This 7-1.C.1.d condition is selected based on 1 million cycles approximating an infinite number of cycles so that the limiting stress is the endurance limit." Therefore the transient does not impact fatigue usage. This item is not counted. Startup of one reactor coolant pump at hot standby conditions 1000 1 -I.A.1.e Normal 9 Coastdown of one reactor 1000 1-I.A.1.f Normal coolant pump at hot standby conditions 10 Adding 40F feedwater at 875 15000 5.4.2.1.C Normal gpm to the steam generator through the downcomer feedwater nozzle when at hot standby conditions 11 Pressurizer heatup from 70F 500 Sheet 1 1-I.A.1.a Normal to 653F at a rate of <200F/hr Palo Verde Nuclear Generating Station License Renewal Application Amendment 14 Page 4.3-8

Section 4 TIME-LIMITED AGING ANALYSES Pressurizer cooldown from 653F to 70F at a rate of <200F/hr 13 Shift from normal to maximum 1000 8-I.D.1.a Normal purification flow at 100% 11-11.A.1.a power 14 Standby to Sl cold leg 160 9-I.E.1.a Normal injection check valve stroke 18-I11.A.1.e test to standby (using charging pumps) 15 High-pressure safety injection 40 11-II.A.1.b Normal header check valve test 16 Turbine roll test at hot standby 10 7-I.C.l.e Normal 17 Initiation of auxiliary spray 500 12-11.A.1.c Normal This transient is tracked by pressurizer during cooldown cooldown events. 18 Startup of SDC system from 500 18-111.A.1.b Normal standby to shutdown cooling (RCS >200F) to shutdown cooling (RCS <200F) to standby Palo Verde Nuclear Generating Station License Renewal Application Amendment 14 Page 4.3-9

Section 4 TIME-LIMITED AGING ANALYSES Startup of safety injection system from standby to injection to short term recirculation to long term recirculation to shutdown cooling to standby Normal 20 Standby to LPSI pump test to 500 18-111.A.1.d Normal standby 21 Standby to HPSI pump test to 500 18-111.A.1.c Normal standby 22 Adding 40F feedwater at 875 500 5.4.2.1.D Normal gpm to the steam generator through the downcomer feedwater nozzle during loading conditions 23 Adding 1 OOF feedwater at 875 500 5.4.2.1.E Normal gpm to the steam generator through the downcomer feedwater nozzle during loading conditions 24 Pressure transients of 85 psi 4000 5.4.2.1.H Normal across the primary divider plate in either direction caused by starting and stopping reactor coolant pumps Palo Verde Nuclear Generating Station License Renewal Application Amendment 14 Page 4.3-10

Section 4 TIME-LIMITED AGING ANALYSES 01 A 4 '1) 25 26 30 9-I.E.1.b 18-II.A.1.f ormal This is conducted during refueling outages. It is not currently being counted because it was recently identified and added to UFSAR Table 3.9-1. It will be added to the cycle counting surveillance procedure 73ST-9RC02 by August 25, 2010, as shown in Commitment No. 55 in Table A4-1. 1.-I I. 9 9 80 LUW-IUW VUlUlII t, UIILIUI tank/charging pump suction diversion to RWT 9-1.D.2.d 12-11.A.2.b Upset 27 Pressurizer level control, 100 9-l.D.2.e Upset failure to full open 28 C asido of one reactor 10 2-I.A.2.a Upset This transient is not counted in coolant pump at 100%/ power surveillance procedure 73-ST-9RC02, (no reactor trip) but its affect would be assessed in the corrective action program since it would be an Anticipated Transient Without Scram (ATWS) event. 29 Spurious reactor trips 50 (RCS) 3-I.A.2.h Upset (including operator error) at 240 (CVCS) 13-11.B.2.d 100% power 30 Loss of reactor coolant system 40 13-11.B.2.a Upset flow 31 Arbitrary load rejection, from 40 7-I.C.2.a Upset 100% to 15% power 16.1l.D.2.a Palo Verde Nuclear Generating Station License Renewal Application Amendment 14 Page 4.3-11

Section 4 TIME-LIMITED AGING ANALYSES OBE Condition - Full-load cycles about a mean value of zero and with an amplitude equal to the maximum response produced during the entire OBE event 33 Inadvertent control element 40 2-l.A.2.c Upset assembly drop, at 100% 13-1l.B.2.b power 34 Inadvertent control element 40 2-I.A.2.d Upset assembly withdrawal from 0% 13-11.B.2.c power 35 Loss of charging and recovery 200 8-1.D.2.b Upset at 100% power 12-11.A.2.d 36 Loss of letdown and recovery 300 8-1.D.2.a Upset at 100% power 12-I1.A.2.a 37 Charging cycles (on/off) 800 9-I.D.2.f Upset during an extended loss of 12-11.A.2.e letdown 38 Depressurization by spurious 40 2-I.A.2.e Upset actuation of pressurizer spray 13-11.B.2.e control valve(s) at 100% power (normal and auxiliary spray valves are considered) 39 Partial loss of condenser 40 5-1.B.2.e Upset cooling at 100% power 15-11.C.2.b Palo Verde Nuclear Generating Station License Renewal Application Amendment 14 Page 4.3-12

Section 4 TIME-LIMITED AGING ANALYSES r"X,;Ufb IUUUWdLf1 Il1W UUV LU control system malfunction at 100% power 1-Il.D.2.U 15-11.C.2.c 41 Turbine trip without 40 (RCS) 7.I.C.2.d Upset accompanying reactor trip at 120 (CVCS) 16.1l.D.2.b 100% power 42 Inadvertent actuation of one 5 (RCS) 7.l.C.2.c Upset main steam line isolation valve 40 (CVCS) 16.1l.D.2.c at 100% power 43 Inadvertent actuation of one 40 7.I.C.2.b Upset turbine bypass valve or 16.1l.D.2.d atmospheric dump valve at 100% power 44 Seismic Event up to & 2 10-I.F.2.a Upset including One-Half of the Safe 17-Il.E.2.a Shutdown Earthquake, at 100% Power 45 Inadvertent isolation of one 5 4-I.B.2.c-Upset main feedwater heater at 100% power 46 Loss of Feedwater Flow (to 85 15.1l.C.2.a Upset S/G) 47 Inadvertent initiation of 5 8-I.D.2.c Upset auxiliary spray at 100% power 12-11.A.2.c Palo Verde Nuclear Generating Station License Renewal Application Amendment 14 Page 4.3-13

Section 4 TIME-LIMITED AGING ANALYSES e 4.3 PVNGS Units 1. 2. and 3 Licensi, Oyb*LIII I 1'dIK UU", LU IUPJLUIV UI largest instrument or sampling connection at 100% power 49 Inadvertent closure of one 40 4-1.B.2.a Upset main feedwater valve at 100% power 50 Inadvertent trip of one main 40 4-I.B.2.b Upset feedwater or one main condensate pump at 100% power 51 inadvertent closure of all main 5 5-1.B.2.f Upset feedwater valves (due to loss of pressure in compressed air system) at 100% power 52 !Sta~upofne reatoFcoolant 10 2-I.A.2.b Upset The plant design will not allow the piump t 50O/*,power reactor to be critical without all four reactor coolant pumps operating. This item is not counted. 53 Loss of an electrical bus 40 2-I.A.2.g Upset supplying two reactor coolant pumps at 100% power 54 Standby to spurious startup of 40 10-I.E.2.a Upset a normally secured 19-111.A.2.a pump/spurious stopping of a normally running pump/spurious valve opening/spurious valve closure Palo Verde Nuclear Generating Station License Renewal Application Amendment 14 Page 4.3-14

Section 4 TIME-LIMITED AGING ANALYSES Rx Trips, Turbine - Loss of RCS Flow 56 Adding 40F feedwater at 1750 280 5.4.2.1.G Upset gpm to the steam generator through the downcomer feedwater nozzles with the flow initiated 30 seconds after a loss of normal feedwater 57 Pressurization by spurious 10 2-1.A.2.f Upset actuation of all pressurizer 14-11.B.3.c Emergency heaters at 100% power 58 Depressurization due to 10 7-I.C.3.a Emergency Emergency and faulted events are not inadvertent actuation of one fatigue cycles in Class 1 fatigue secondary safety valve at analyses. Item is counted, but is not 100% power required to be counted. 59 Loss of offsite and onsite ac

• 5 11-1.F.3.a Emergency Emergency and faulted events are not power, with retention of onsite 17-11.E.3.a fatigue cycles in Class 1 fatigue emergency ac and dc power analyses. Item is counted, but is not at 100% power required to be counted.

60 Depressurization of the SIS, 5 19-II.A.3.a Emergency Emergency and faulted events are not CSS, SCS by full opening of a fatigue cycles in Class 1 fatigue safety or relief valve without analyses. Item is counted, but is not reseating required to be counted. 61 Depressrization dueto 1 3-1.A.3.a Emergency Emergency and faulted events are not inadeertent actuation ofone fatigue cycles in Class 1 fatigue pressurzersafety va!ve at= analyses. Item is not required to be 100%0/ ow,,&ýe.I I I I I counted. Palo Verde Nuclear Generating Station License Renewal Application Amendment 14 Page 4.3-15

Section 4 TIME-LIMITED AGING ANALYSES Adding 4)1-teedwater at 11 gpm to the steam generator through the downcomer feedwater nozzles during a steam line break 1 MSLB event with 7 feedwater addition cycles Emergency and faulted events are not fatigue cycles in Class 1 fatigue analyses. Item is counted, but is not required to be counted. 65 Single reactor coolant pump 1 3-I.A.4.b Faulted Emergency and faulted events are not shaft seizure at 100% power fatigue cycles in Class 1 fatigue analyses. Item is counted, but is not required to be counted. 66 Major loss of coolant incident 1 3-I.A.4.c Faulted Emergency and faulted events are not (system operating mode fatigue cycles in Class 1 fatigue dependent upon design analyses. Item is counted, but is not application for worst case required to be counted. conditions) 67 Single reactor coolant pump 1 4-I.A.4.d Faulted Emergency and faulted events are not sheared shaft at 100% power fatigue cycles in Class 1 fatigue analyses. Item is counted, but is not required to be counted. 68 Class 2 line break 1 12-11.A.4.a Faulted Emergency and faulted events are not fatigue cycles in Class 1 fatigue analyses. Item is counted, but is not required to be counted. Palo Verde Nuclear Generating Station License Renewal Application Amendment 14 Page 4.3-16

Section 4 TIME-LIMITED AGING ANALYSES Seismic event up to and including the safe shutdown earthquake (system operating mode dependent upon design application for worst case conditions) 11-I.F.4.a 17-Il.E.4.a Emergency and faulted events are not fatigue cycles in Class 1 fatigue analyses. Item is counted, but is not required to be counted. 70 Major rupture of the safety 1 19-II.A.4.a Faulted Emergency and faulted events are not injection system at the highest* fatigue cycles in Class 1 fatigue sysýýtem pressure encountered analyses. Item is not required to be during a normal operating counted. mode; namely, rupture during the first phase of the preoperational hydrostatic tes~t 71 The concurrent loading 1 Sheet 2 Faulted Emergency and faulted events are not produced by normal operation fatigue cycles in Class 1 fatigue at full power, plus the design analyses. Item is counted, but is not basis earthquake, plus loss-of-required to be counted. coolant accident (pipe rupture) are used to determine the faulted plant loading condition. 72 Major rupture in the main 1 5-l.B.4.a Faulted Emergency and faulted events are not feedwater piping (system 15-11.C.3.a Emergency fatigue cycles in Class 1 fatigue operating mode dependent analyses. Item is counted, but is not upon design application for required to be counted. worst case conditions) Palo Verde Nuclear Generating Station License Renewal Application Amendment 14 Page 4.3-17

Section 4 TIME-LIMITED AGING ANALYSES Major rupture in the auxiliary feedwater piping (system operating mode dependent upon design application for worst case conditions) 6-l.B.4.b 15-11.C.3.b Faulted Emergency Emergency and faulted events are not fatigue cycles in Class 1 fatigue analyses. Item is counted, but is not required to be counted. 74 Major rupture in the main 1 8-1.C.4.a Faulted Emergency and faulted events are not steam piping (system 16.1l.D.3.a Emergency fatigue cycles in Class 1 fatigue operating mode dependent analyses. Item is counted, but is not upon design application for required to be counted. worst case conditions) 75 Control element assembly 1 3-I.A.4.a Faulted Emergency and faulted events are not ejection at 0% power 14-11.B.3.b Emergency fatigue cycles in Class 1 fatigue analyses. Item is counted, but is not required to be counted. 76 L ossoecondryessure Not Credible Sheet 2 Faulted This item is not counted because per One cycle of a postulated loss UFSAR Table 3.9.1-1 "These are not of secondary pressure due to considered credible events in forming a complete double ended the design basis of the reactor coolant severance of one steam system. However, they are included to generator or feedwater nozzle, demonstrate that the reactor coolant but not simultaneously. Thesei system components will not fail are not considered credible structurally in the unlikely event that one eventsin forming the design of these events occur." 'b'asis of the reactor coolant system However they are included to demonstrate that ~the reactor coolant systeri

comon nents will not fail structurally in the unlikely event that one of these events

'occur. Palo Verde Nuclear Generating Station License Renewal Application Amendment 14 Page 4.3-18

Section 4 TIME-LIMITED AGING ANALYSES I I Reactor Coolant System hydrostatic test 4-l.A.b.a 14-ll.B.5.a ýst 78 Secondary system hydrostatic 10 6-I.B.5.a 5.4.2.1.A Test test 79 Reactor Coolant System leak 200 Sheet 3 4-I.A.5.a Test test 14-11.1B.5.b 80 Secondary system leak test 200 6-I.B.5.b 5.4.2.1.B Test 81 CVCS System Hydrostatic 40 13-I1.A.5.a Test Test 82 Standby to preoperational 10 20-111.A.5.a Test hydrostatic test to standby 83 Standby to inservice 10 20-11I.A.5.b Test hydrostatic test to standby Palo Verde Nuclear Generating Station License Renewal Application Amendment 14 Page 4.3-19

Section 4 TIME-LIMITED AGING ANALYSES 4.3.1.3 Seismic History Design analyses that compare seismic stresses against allowable stresses, in the absence of any consideration of the number of cycles or of fatigue effects, are not TLAAs. However, design of structures, systems, and components may include seismic loads in fatigue analyses, or may assume a stated number of seismic load cycles for purposes of establishing an allowable stress or stress range. Significant earthquakes at the site can therefore increase the accumulated fatigue usage factor. The site seismic history can therefore affect the disposition of TLAAs. However, no significant earthquakes have occurred at PVNGS since construction. For design purposes the PVNGS safe shutdown earthquake (SSE) and operating basis earthquake (OBE) are defined as 0.20 g and 0.10 g ground motion, respectively. Analyses of Seismic Category I structures used a conservative design basis 0.25 g SSE and 0.13 g OBE [UFSAR 3.7]. For the purposes of evaluating actual events at PVNGS, an SSE is defined as one with a modified-Mercalli intensity level 8 (ground motion of 0.15 to 0.33 g or above); and an OBE is defined as one with a modified-Mercalli intensity level 7 (ground motion of 0.072 to 0.15 g). No SSE or OBE has occurred to date. The site has recorded seven minor earthquakes as of 2008, some of these not strong enough to qualify as recordable "earthquake events." The strongest had a ground motion of only 0.015 g, or about 12% of the acceleration, and therefore the applied loads, of a design basis 0.13 g OBE. 4.3.1.4 Present and Projected Status of Monitored Locations Summary Description The current fatigue monitoring program transient cycle count procedure, 73ST-9RC02, recorded accumulated transient events for the 9 transients listed in Appendix J of the procedure since the Unit 1 startup in 1985. This transient list did not include every transient in UFSAR Section 3.9.1.1 because, prior to implementation of Improved Technical Specifications in 1998, the Technical Specifications required monitoring only transients that are now in UFSAR Section 3.9.1.1.1. In 1995 (after 10 years of Unit 1 operation), the cycle count procedure was revised to include the 48 remaining UFSAR transients listed in Appendix K of the procedure. In the 1995 record of the revised procedure, accumulation for all transient events not counted to date was assumed at 25% of the limiting value for the 40-year design. After the 1995 revision of the cycle count procedure, transients were recorded and were added to the 25% accumulation assumed in 1995. APS Fatigue Cycle Count Verification The goal of the APS fatigue cycle count verification was to reduce the uncertainty created by the 25% accumulation assumed in 1995. Palo Verde Nuclear Generating Station Amendment 14 Page 4.3-20 License Renewal Application

Section 4 TIME-LIMITED AGING ANALYSES Scope The scope of the cycle count verification included all transients required to be monitored by PVNGS Technical Specifications 5.5.5. Recount Method Several APS employees and contractor personnel were designated based on their long-term familiarity with PVNGS to perform document reviews. The reviewers examined the microfilmed control room logs, NRC Monthly Operating Reports and LERs for the period prior to January 1996 for all three PVNGS units. The personal recollections and records of unit personnel were used to supplement the record review, and a best-source total was determined for each monitored transient. The best-source total was added to the actual count of events following 1995 to obtain a best-source total as of the end of 2005. These best-source totals as of the end of 2005 are reported in Table 4.3-3, Columns 4, 5, and 6. Transient Projections Transient totals were projected to the end of the PEO for information only. The projections predict that 10 CFR 54.21(c)(1)(iii) aging management will be successful and that in most cases future corrective actions will not be necessary. However, the projections are not intended to justify 10 CFR 54.21(c)(1)(i) validations, or to provide revised design bases for 10 CFR 54.21(c)(1)(ii) analysis revisions. The projections are based on a linear extrapolation as follows:

1. The shortest period of operation as of the end of 2005 was 18 years in Unit 3 and the longest was 20 years in Unit 1, so a scaling factor of 3.33 (60 years extended life divided by 18 years shortest operation) was used to project totals to the end of the PEO. In a few special cases a scaling factor of 6.66 was used when the available data covered a ten year period. These exceptions are noted and explained in Table 4.3-3.

2. The highest total accumulation for an event was selected without regard to which unit it occurred in.

3. A highest unit 60 year projection (column 7) as of the end of the PEO was obtained by multiplying the highest total accumulation for each event by the scaling factor.

Example: Event #1 RCS Heatup highest unit total was in Unit 2 (64 heatups). 64 X 3.33=213. The highest unit 60 year projection (column 7) is 213. Palo Verde Nuclear Generating Station Amendment 14 Page 4.3-21 License Renewal Application

Section 4 TIME-LIMITED AGING ANALYSES It should be noted that only a few events such as recurring test events lend themselves to projection with well defined assumptions, so the projections presented in Table 4.3-3 are only best-estimates. Early plant history involved a number of first-of-a-kind issues that may make the projections artificially high, and end of life issues may make the projections artificially low. Therefore, consistent with aging management, no attempt has been made to reanalyze or implement other corrective actions based on these projections. Corrective actions will be triggered by the action limits that will be established in the enhanced Metal Fatigue of Reactor Coolant Pressure Boundary program (B3.1) Palo Verde Nuclear Generating Station License Renewal Application Amendment 14 Page 4.3-22

Section 4 TIME-LIMITED AGING ANALYSES Table 4.3-3, PVNGS Units 1, 2, and 3 Fa Count and Proections 1 2 3 4-5 ,6 7 8 Rbw" Transient Title Limiting -,Unit 1 Unit 2-Unit 3 Highest Notes No. -(Shaded items Value Accumulation Accumulation Accumulation Unit 60yr are not counted) as of January as of January as of January Projection 2006 2006 ,2006 (Highest Unit Total Reactor Coolant System 500 (The 62 64 59 213 (excluding pressurizer) heatup reactor vessel from 70F to hot standby studs shall be conditions at a rate of <100F/hr. limited to 250 occurrences) 2 Reactor Coolant System 500 (The 61 63 58 212 (excluding pressurizer) cooldown reactor vessel from hot standby conditions to studs shall be 70F at a rate of <10OF/hr limited to 250 occurrences) 3 5%/minute power ramp,," 15000 Not Counted Not Counted Not Counted Not This item is not counted. To reach increase,from 15% to i00% Counted this number the plant would have power to experience a power increase on the average of once every 31.5 hours of operation in 60 years with a 90% capacity factor. Since the PVNGS units operate as base loaded plants this is not credible. 4 5%/minute power ramp 15000 Not Counted Not Counted Not Counted Not This item is not counted. To reach this decrease,from 100% to 15% Counted number the plant would have to power experience a power decrease on the average of once every 31.5 hours of operation in 60 years with a 90% capacity factor. Since the PVNGS units operate as base loaded plants this is not credible. 5 10% power step increase, from 2000 264 248 206 879 90% to 100% power I I I I Palo Verde Nuclear Generating Station License Renewal Application Amendment 14 Page 4.3-23

Section 4 TIME-LIMITED AGING ANALYSES 10% power step decrease, from 100% to 90% power 7 Normal cyclic variatios at100 % 1.E+06 Not Counted Not Counted Not Counted Not This item is not counted. Per UFSAR Table power; +/-80 psi, +/-10F Counted 3.9.1-1 "This condition is selected based on 1 million cycles approximating an infinite number of cycles so that the limiting stress is the endurance limit." Therefore the .transient does not impact fatigue usage. 8 Startup of one reactor coolant 1000 273 281 275 936 pump at hot standby conditions 9 Coastdown of one reactor 1000 269 275 268 916 coolant pump at hot standby conditions 10 Adding 40F feedwater at 875 15000 0* 0* 0* 13*

• Note that per UFSAR 5.4.2.1 this is a SG gpm to the steam generator transient. SGs were replaced in the fall through the downcomer outages of 2003, 2005 and 2007 for U2, U1 feedwater nozzle when at hot and U3 respectively resetting this event to standby conditions zero. Therefore, the U1 and U3 totals are reported as zero. Two events were counted in U2 between 1995 and 2005. Although both occurred prior to SGR they were assumed to, apply to the replacement SG's to calculate a projection for conservatism.

Since the accumulation period was 10 years versus 20 years the scaling factor was doubled to 6.66 Palo Verde Nuclear Generating Station License Renewal Application Amendment 14 Page 4.3-24

Section 4 TIME-LIMITED AGING ANALYSES Pressurizer heatup from 70F to 653F at a rate of <200F/hr 12 Pressurizer cooldown from 653F 500 85 82 76 285 to 70F at a rate of <20OF/hr 13 Shift from normal to maximum 1000 250 250 250 833 purification flow at 100% power 14 Standby to SI cold leg injection 160 0 0 0 0 PVNGS has never done this ASME Section check valve stroke test to XI test under hot conditions and has no standby (using charging pumps) plans to do it at temperature. 15 High-pressure safety injection 40 0 0 0 0 PVNGS has never done this ASME Section header check valve test Xl test under hot conditions and has no plans to do it at temperature. 16 Turbine roll test at hot standby 10 3 3 2 10 Recount activities did not identify this test in U2 logs. However, since it is part of initial plant testing a value of 3 was assumed to equal the highest actual count. 17 Initiation of auxiliary spray 500 85 82 76 285 This transient is tracked by pressurizer during cooldown cooldown events. 18 Startup of SDC system from 500 136 148 145 493 standby to shutdown cooling (RCS >200F) to shutdown cooling (RCS <200F) to standby Palo Verde Nuclear Generating Station License Renewal Application Amendment 14 Page 4.3-25

Section 4 TIME-LIMITED AGING ANALYSES Startup of safety injection system from standby to injection to short term recirculation to long term recirculation to shutdown cooling to standby 20 Standby to LPSI pump test to 500

239, 228 252 839 The original assumption of a monthly run standby has shown to be an underestimation. This surveillance is run 10% more frequently than required to ensure test intervals are not exceeded, and it is conducted for post maintenance testing. Due to the predictable nature of this transient it is probable that the cycle counting action limit will be reached prior to the end of the PEO.

21 Standby to HPSI pump test to 500 246 222 243 819 The original assumption of a monthly run standby has shown to be an underestimation. This surveillance is run 10% more frequently than required to ensure test intervals are not exceeded, and it is conducted for post maintenance testing. Due to the predictable nature of this transient it is probable that the cycle counting action limit will be reached prior to the end of the PEO. Palo Verde Nuclear Generating Station License Renewal Application Amendment 14 Page 4.3-26

Section 4 TIME-LIMITED AGING ANALYSES Pressure transients of 85 psi across the primary divider plate in either direction caused by starting and stopping reactor coolant pumps

• Note that per UFSAR 5.4.2.1 this is a SG transient. SGs were replaced in the fall outages of 2003, 2005 and 2007 for U2, U1 and U3 respectively resetting this event to zero. Therefore, the U1 and U3 totals are reported as zero.
  • The U2 total is based on assuming all RCP starts and stops reported between '95

- '05 in transients 8 & 9 apply to the RSG. The sum (56) was multiplied by 6 assuming all 4 RCPs experienced a start and stop for each Mode 3 start and stop plus 2 pump start/stop cycles for sweeps. This was multiplied by 6.66 to account for the 336 being accumulated in a 10 year period versus 20 years. a a a a a Palo Verde Nuclear Generating Station License Renewal Application Amendment 14 Page 4.3-27

Section 4 TIME-LIMITED AGING ANALYSES ble 4.3-3. PVNGS Units 1. 2. I 30 16 17 16 57 This transient is conducted during refueling outages. It is not currently being counted because it was recently identified and added to UFSAR Table 3.9-1. It will be counted when added to the cycle counting surveillance procedure 73ST-9RC02. Totals are estimated and reflect one performance per refueling through the end of 2005 plus a margin of five events. 26 I I I I _____ Low-low volume control tank/charging pump suction diversion to RWT 80 20 20 20 67 I 1 4 I I-I 27 28 29 Pressurizer level control, failure to full open 100 25 25 25 83 10 0 0 0 0 T This transient is not counted in surveillance procedure 73-ST-9RC02, but its affect would be assessed in the corrective action program since it would be an Anticipated Transient Without Scram (ATWS) event. Spurious reactor trips {inciuding operator error) at 100% power 50 (RCS) 240 (CVCS) 33 32 26 110 30 Loss of reactor coolant system 40 13 flow Palo Verde Nuclear Generating Station License Renewal Application Amendment 14 Page 4.3-28

Section 4 TIME-LIMITED AGING ANALYSES 31 Arbitrary load rejection, from 100% to 15% power 40 6 14 4/ 32 OBE Condition - Full-load cycles 200 0 0 0 0 about a mean value of zero and with an amplitude equal to the maximum response produced during the entire OBE event 33 Inadvertent control element 40 3 2 5 17 assembly drop, at 100% power 34 Inadvertent control element 40 0 0 0 0 assembly withdrawal from 0% power 35 Loss of charging and recovery at 200 7 0 2 23 100% power 36 Loss of letdown and recovery at 300 17 18 10 60 100% power 37 Charging cycles (on/off) during 800 64 1 2 213 an extended loss of letdown 38 Depressurization by spurious 40 0 1 1 3 actuation of pressurizer spray control valve(s) at 100% power (normal and auxiliary spray valves are considered) Palo Verde Nuclear Generating Station License Renewal Application Amendment 14 Page 4.3-29

Section 4 TIME-LIMITED AGING ANALYSES A4 DQ/A 39 Partial loss of condenser cooling at 100% power 40 1 0 0 3 40 Excess feedwater flow due to 40 2 0 1 7 control system malfunction at 100% power 41 Turbine trip without 40 (RCS) 14 7 6 47 accompanying reactor trip at 120 (CVCS) 100% power 42 Inadvertent actuation of one 5 (RCS) 0 1 1 3 main steam line isolation valve at 40 (CVCS) 100% power 43 Inadvertent actuation of one 40 2 1 0 7 turbine bypass valve or atmospheric dump valve at 100% power 44 Seismic Event up to & including 2 0 0 0 0 One-Half of the Safe Shutdown Earthquake, at 100% Power 45 Inadvertent isolation of one 5 0 0 0 0 main feedwater heater at 100% power 46 Loss of Feedwater Flow (to S/G) 85 9 8 11 37 47 Inadvertent initiation of auxiliary 5 0 0 1 3 spray at 100% power Palo Verde Nuclear Generating Station License Renewal Application -Amendment 14 Page 4.3-30

Section 4 TIME-LIMITED AGING ANALYSES System leak due to rupture ot largest instrument or sampling connection at 100% power 49 Inadvertent closure of one main 40 1 0 0 3 feedwater valve at 100% power 50 Inadvertent trip of one main 40 7 8 11 37 feedwater or one main condensate pump at 100% power 51 Inadvertent closure of all main 5 1 0 0 3 feedwater valves (due to loss of pressure in compressed air system) at 100% power 52 Startupof oonereactor coolant J> 10 0 0 0 0 The plant design will not allow the pump at 50% power reactor to be critical without all four reactor coolant pumps operating. This item is not counted. 53 Loss of an electrical bus 40 2 4 4 13 supplying two reactor coolant pumps at 100% power 54 Standby to spurious startup of a 40 0 1 0 3 normally secured pump/spurious stopping of a normally running pump/spurious valve opening/spurious valve closure Palo Verde Nuclear Generating Station License Renewal Application Amendment 14 Page 4.3-31

Section 4 TIME-LIMITED AGING ANALYSES Rx Trips, Turbine Trips and Loss of RCS Flow -4 I I I I-I 56 Adding 40F feedwater at 1750 gpm to the steam generator through the downcomer feedwater nozzles with the flow initiated 30 seconds after a loss of normal feedwater 280 0* 0* 0* 13*

• Note that per UFSAR 5.4.2.1 this is a SG transient. SGs were replaced in the fall outages of 2003, 2005 and 2007 for U2, U1, and U3, respectively, resetting this event to zero. Therefore, the U1 and U3 totals are reported as zero. Two events were counted in U2 between 1995 and 2005. Although both occurred prior to SGR they were assumed to apply to the replacement SG's to calculate a projection for conservatism.

Since the accumulation period was 10 years versus 20 years the scaling factor was doubled to 6.66 57 Pressurization by spurious 10 2 2 2 7 actuation of all pressurizer heaters at 100% power 58 Depressurization due to 10 5 2 0 17 Item is not required to be counted. inadvertent actuation of one Emergency and faulted events are not secondary safety valve at 100% fatigue cycles in Class 1 fatigue analyses. power 59 Loss of offsite and onsite ac 5 1 2 2 7 Item is not required to be counted. power, with retention of onsite Emergency and faulted events are not emergency ac and dc power at fatigue cycles in Class 1 fatigue analyses. 100% power Palo Verde Nuclear Generating Station License Renewal Application Amendment 14 Page 4.3-32

Section 4 TIME-LIMITED AGING ANALYSES Depressurization of the SIS, CSS, SCS by full opening of a safety or relief valve without reseating Item is not required to be counted. Emergency and faulted events are not fatigue cycles in Class 1 fatigue analyses. 61 62 63 64 1 0 0 0 0 Item is not required to be counted. Emergency and faulted events are not fatigue cycles in Class 1 fatigue analyses. 5 0 0 0 0 Item is not required to be counted. Emergency and faulted events are not fatigue cycles in Class 1 fatigue analyses. 5 1 0 0 3 Item is not required to be counted. Emergency and faulted events are not fatigue cycles in Class 1 fatigue analyses. Adding 40F feedwater at 1750 gpm to the steam generator through the downcomer feedwater nozzles during a steam line break 1 MSLB event with 7 feedwater addition cycles 0* 0 0* 0

• Note that per UFSAR 5.4.2.1 this is a SG transient. Since the U1 and U3 SGs were replaced after January 2005 the total is reported as zero here. Item is not required to be counted.

Emergency and faulted events are not fatigue cycles in Class 1 fatigue analyses. 65 Single reactor coolant pump 1 0 0 0 0 Item is not required to be counted. shaft seizure at 100% power Emergency and faulted events are not I L _I _fufatigue cycles in Class 1 fatigue analyses. Palo Verde Nuclear Generating Station License Renewal Application Amendment 14 Page 4.3-33

Section 4 TIME-LIMITED AGING ANALYSES Major loss of coolant incident (system operating mode dependent upon design application for worst case conditions) Item is not required to be counted. Emergency and faulted events are not fatigue cycles in Class 1 fatigue analyses. 67 Single reactor coolant pump 1 0 0 0 0 Item is not required to be counted. sheared shaft at 100% power Emergency and faulted events are not fatigue cycles in Class 1 fatigue analyses. 68 Class 2 line break 1 0 0 0 0 Item is not required to be counted. Emergency and faulted events are not fatigue cycles in Class 1 fatigue analyses. 69 Seismic event up to and including the safe shutdown earthquake (system operating mode dependent upon design application for worst case. conditions) 1 0 0 0 0 Item is not required to be counted. Emergency and faulted events are not fatigue cycles in Class 1 fatigue analyses. 70 1 0 0 0 0 Item is not required to be counted. Emergency and faulted events are not fatigue cycles in Class 1 fatigue analyses. Palo Verde Nuclear Generating Station License Renewal Application Amendment 14 Page 4.3-34

Section 4 TIME-LIMITED AGING ANALYSES 71 The concurrent loading produced by normal operation at full power, plus the design basis earthquake, plus loss-of-coolant accident (pipe rupture) are used to determine the faulted plant loading condition. 1 0 0 0 0 Item is not required to be counted. Emergency and faulted events are not fatigue cycles in Class 1 fatigue analyses. 72 Major rupture in the main 1 0 0 0 0 Item is not required to be counted. feedwater piping (system Emergency and faulted events are not operating mode dependent upon fatigue cycles in Class 1 fatigue analyses. design application for worst case conditions) 73 Major rupture in the auxiliary 1 0 0 0 0 Item is not required to be counted. feedwater piping (system Emergency and faulted events are not operating mode dependent upon fatigue cycles in Class I fatigue analyses. design application for worst case conditions) 74 Major rupture in the main steam 1 0 0 0 0 Item is not required to be counted. piping (system operating mode Emergency and faulted events are not dependent upon design fatigue cycles in Class 1 fatigue analyses. application for worst case conditions) 75 Control element assembly 1 0 0 0 0 Item is not required to be counted. ejection at 0% power Emergency and faulted events are not fatigue cycles in Class 1 fatigue analyses. Palo Verde Nuclear Generating Station License Renewal Application Amendment 14 Page 4.3-35

Section 4 TIME-LIMITED AGING ANALYSES A `' 0 01/AI, 0--f ý Not Credible ! Not Counte Not Counted* I Not Counted* Not Counted This item is not counted because per UFSAR Table 3.9.1-1 "These are not considered credible events in forming the design basis of the reactor coolant system. However, they are included to demonstrate that the reactor coolant system components will not fail structurally in the unlikely event that one of these events occur." 77 b i Reactor Coolant System hydrostatic test 10 1 1 1 3 78 Secondary system hydrostatic 10 1 1 3 The U2 RSG was subject to one preservice test hydrostatic test prior to receipt. The associated piping experienced one hydrostatic test during original construction. The leak test following replacement was done at normal operating pressure. The U1 and U3 reflect the same sequence of events. 79 Reactor Coolant System leak test 200 5 4 2 17 Palo Verde Nuclear Generating Station License Renewal Application Amendment 14 Page 4.3-36

Section 4 TIME-LIMITED AGING ANALYSES 80 Secondary system leak test 200 50 50 50 167 81 CVCS System Hydrostatic Test 40 1 1 1 3 82 Standby to preoperational 10 2 2 2 7 hydrostatic test to standby 83 Standby to inservice hydrostatic 10 2 2 2 7 test to standby Palo Verde Nuclear Generating Station License Renewal Application Amendment 14 Page 4.3-37

Section 4 TIME-LIMITED AGING ANALYSES 4.3.1.5 Enhanced Metal Fatigue of Reactor Coolant Pressure Boundary program (B3.1) Scope, Action Limits, and Corrective Actions Scope The scope of the enhanced Metal Fatigue of Reactor Coolant Pressure Boundary program (B3.1) will include all ASME Section III Class 1 components and components with Class 1 fatigue analysis and will monitor their fatigue usage by a combination of cycle counting and CUF monitoring as noted in Table 4.3-4. Method The "Fatigue Management Method" column of Table 4.3-4 indicates the method the enhanced Metal Fatigue of Reactor Coolant Pressure Boundary program (B3.1) will use to track fatigue usage for each component. These are stress-based fatigue (SBF), cycle-based fatigue (CBF-C - per cycle, CBF-PC.- per cycle with partial cycles, or CBF-EP - event pairing), and CC. The CC method will be used for components whose cumulative usage can be shown to be satisfactory with this highly conservative monitoring approach. Most of the NUREG/CR-6260 locations and the pressurizer spray nozzle require more sophisticated CBF or SBF algorithms to periodically calculate accumulated fatigue usage and demonstrate that component usage remains less than one. Transient event cycles that are required to be monitored by PVNGS Technical Specifications 5.5.5 will continue to be tracked to ensure that the numbers of transient events assumed by the design basis calculations will not be exceeded. This cycle counting monitoring method will demonstrate design basis compliance for the components using CC monitoring. See Table 4.3-2 for the list of tracked transients. Corrective Action Limits and Corrective Actions The PVNGS current fatigue monitoring program is based on cycle counting with one location tracked by a CUF calculated using CBF-PC (pressurizer spray nozzle), and it incorporates a cycle based action limit of 90% of the design event occurrences and a CUF based action limit of 0.65 for the pressurizer spray nozzle usage. The current fatigue monitoring program requires this evaluation at least once per fuel cycle. The current action limits are established to allow action to be taken in time to prevent exceeding the maximum number of allowed cycles or a pressurizer spray nozzle CUF of 1.0, as applicable, and should provide at least one fuel cycle of warning. During the period of extended operation, projections indicate that certain allowable cycles and fatigue limits may be approached. Therefore, specific and targeted action limits will be necessary to ensure actual fatigue limits are not exceeded. Those action limits have not yet been developed. As. the transition to the enhanced Metal Fatigue of Reactor Coolant Pressure Boundary program (B3.1) and FatiguePro is implemented, there are certain embedded administrative tools in FatiguePro that will allow for specification of action limits based on projected fatigue usage at specific locations that account for actual cumulative fatigue. The action limits can be based on the time required to implement expected or projected mitigating actions (such as component replacements or revisions to ASME Code Fatigue Analyses of Record) prior to actual fatigue limits being exceeded. Palo Verde Nuclear Generating Station Amendment 14 Page 4.3-38 License Renewal Application

Section 4 TIME-LIMITED AGING ANALYSES Action Limit Margins The enhanced Metal Fatigue of Reactor Coolant Pressure Boundary program (B3.1) corrective action limits will ensure that corrective actions are taken before the design limits are exceeded. Corrective action limits will ensure that appropriate reevaluation or other corrective actions are initiated while sufficient margin remains to allow at least one occurrence of the worst case (highest fatigue usage per cycle) low probability transient that is included in design specifications, without exceeding the code limit CUF of 1.0. For NUREG/CR-6260 locations, CUF calculation will be done using the appropriate Fen environmental factor. Cycle Count Action Limits and Corrective Actions Cycle Counting monitoring (CC) action limits for the enhanced Metal Fatigue of Reactor Coolant Pressure Boundary program (B3.1) will be established based on the design-specified number of cycles. Since sufficient margin must be maintained to accommodate any design transient regardless of probability, the enhanced Metal Fatigue of Reactor Coolant Pressure Boundary program (B3.1) corrective actions will be taken before the remaining number of allowable occurrences for any specified transient becomes less than one. Corrective actions will be required when the cycle count for any of the significant contributors to usage factor is projected to reach the action limit defined in the enhanced Metal Fatigue of Reactor Coolant Pressure Boundary program (B3.1) before the end of the next fuel cycle. Cycle Counting Corrective Actions to be incorporated into the enhanced Metal Fatigue of Reactor Coolant Pressure Boundary program (B3.1): If a cycle count action limit is reached, corrective actions will be performed as necessary:

1. Review of fatigue usage calculations.

To determine whether the transient in question contributes significantly to CUF. To identify the components and analyses affected by the transient in question. To ensure that the analytical bases of the high-energy line break (HELB) locations are maintained. To ensure that the analytical bases of the fatigue crack growth and stability analysis in support of relief from ASME Section XI flaw removal[ and inspection requirements for hot leg small-bore half nozzle repairs are maintained.

2. Evaluation of remaining margins on CUF based on cycle-based or stress-based CUF calculations using the PVNGS fatigue management software.

3. Redefinition of the specified number of cycles (e.g., by reducing specified numbers of cycles for other transients and using the margin to increase the allowed number of cycles for the transient that is approaching its specified number of cycles).

4. Redefinition of the transient to remove conservatism in predicting the range of pressure and temperature values for the transient.

Palo Verde Nuclear Generating Station Amendment 14 Page 4.3-39 License Renewal Application

Section 4 TIME-LIMITED AGING ANALYSES These actions are designed to determine how close the usage is to 1.0, and from those determinations, set new action limits. Further actions for cumulative fatigue usage action limits may be invoked if good engineering judgment determines that is necessary. Cumulative Fatigue Usage Action Limits and Corrective Actions The enhanced Metal Fatigue of Reactor Coolant Pressure Boundary program (B3.1) will use an automated three-dimensional, six-element stress tensor, stress-based fatigue management software module (the SBF module, meeting ASME III NB-3200 requirements) to continually monitor cumulative usage factor (CUF) at the stress-based fatigue monitoring locations, and cycle-based CUFs will be calculated periodically. The CUF action limits for the enhanced Metal Fatigue of Reactor Coolant Pressure Boundary program (B3.1) will be established to provide two to three fuel cycles of warning prior to exceeding a CUF of 1.0. CUF Action Limit Margins: To provide adequate time for corrective actions and adequate margin to permit continued operation, corrective actions for the enhanced Metal Fatigue of Reactor Coolant Pressure Boundary program (B3.1) will be required when calculated CUF (from cycle based or stress based monitoring) for any monitored location is projected to reach 1.0 within the next 2 or 3 fuel cycles. In order to assure sufficient margin to accommodate occurrence of a low probability transient, corrective actions must also be taken while there is still sufficient margin to accommodate at least one occurrence of the worst case (highest fatigue usage per cycle) design transient event. Action limits will be established to permit completion of corrective actions before the usage factor reaches 1.0. For PVNGS locations identified in NUREG/CR-6260 and described in Section 4.3.4, "Effects of the Reactor Coolant System Environment on Fatigue Life of Piping and Components (Generic Safety Issue 190)," this action limit will be based on accrued fatigue usage calculated with the Fen factors required for including effects of the reactor coolant environment. For example, if inadvertent RCS depressurization, when adjusted for the environmental effects of the reactor coolant system at a NUREG/CR-6260 location, causes 20% of the total allowable fatigue usage, corrective action for that location would be required before calculated usage (including the environmental effects factor, Fen) reached 0.8. CUF Corrective Actions: If a CUF action limit is reached, corrective actions will be performed as necessary:

1. Determine whether the scope of the enhanced Metal Fatigue of Reactor Coolant Pressure Boundary program (B3.1) must be enlarged to include additional affected reactor coolant pressure boundary locations. This determination will ensure that other locations do not approach design limits without an appropriate action.

2. Enhance fatigue monitoring to confirm continued conformance to the code limit.

3. Repair/modify the component.

4. Replace the component.

5. Perform a more rigorous analysis of the component to demonstrate that the design code limit will not be exceeded.

Palo Verde Nuclear Generating Station Amendment 14 Page 4.3-40 License Renewal Application

Section 4 TIME-LIMITED AGING ANALYSES

6. Modify plant operating practices to reduce the fatigue usage accumulation rate.

7. Perform a flaw tolerance evaluation and impose component-specific inspections, under ASME Section XI Appendices A or C (or their successors), and obtain required approvals by the NRC.

Palo Verde Nuclear Generating Station License Renewal Application Amendment 14 Page 4.3-41

Section 4 TIME-LIMITED AGING ANALYSES Table 4.3 Summary of Fatigue Usage from Class I Analyses, and Method of Management by the Metal Fatigue of Reactor Coolant Pressure Boundary Program 3 Maximum 5 Design Basis 4 Fatigue 1 2 CUF Reason For Management Number Component (40 year Monitoring Method Analysis unless (See Section 4.3.1 for otherwise description) noted) 1 RPV Inlet 0.07308 NUREG/CR-6260 CBF-C Nozzle 2 RPV Outlet 0.309574 NUREG/CR-6260 CBF-C Nozzle 3 RPV Wall and 0.0012 NUREG/CR-6260 CC Bottom Head Juncture 4 Surge Line 0.937 SBF (Elbow) NUREG/CR-6260 5 Charging Inlet 0.9205 CBF-EP Nozzle NUREG/CR-6260 6 Shutdown 0.1118 NUREG/CR-6260 CBF-EP Cooling Line Elbow 7 Safety 0.3409 NUREG/CR-6260 CBF-PC Injection Nozzles (Loop I and Loop 2) 8 Pressurizer 0.9923 High CUF CBF-PC Spray Nozzle 9 All other cc locations (Locations not subject to specifically called out fatigue in this table will be monitoring monitored by counting design transients.) Palo Verde Nuclear Generating Station License Renewal Application Amendment 14 Page 4.3-42

Palo Verde Nuclear Generating Station License Renewal Application Amendment No. 14 LRA Table A4-1 Commitment 39, page A-54, is revised to read as follows (deleted text is struck out, new text is underlined): I he current tatique monitoring program is a cycle counting program with one location-specific cumulative usaae factor (CUF) calculation (Pressurizer Spray Nozzle). No later than two years prior to the period of extended operation, the current fatigue monitoring program governed by 73ST-9RC02 will be enhanced to include additional location-specific CUF calculations and an automated and computerized management software program for cycle counting and fatigue usage factor tracking. The automated and computerized software program will be used to supplement manual counting. The enhanced Metal Fatigue of Reactor Coolant Pressure Boundary program (B3.1) will use cycle counting (CC), cycle based fatigue CUF calculations (CBF) and stress based fatigue CUF calculations (SBF) to monitor fatigue. FatiguePro will be used for cycle counting and cycle-based fatigue (CBF) monitoring methods. FatiquePro is an EPRI licensed product. APS commits to the use of a fatigue monitoring software program that incorporates a three-dimensional, six-element stress tensor method meeting ASME III NB-3200 requirements for stress-based fatigue monitoring (SBF). APS also commits to the implementation of this method for SBF monitoring at least two years prior to the period of extended operation. The enhanced Metal Fatigue of Reactor Coolant Pressure Boundary program (B3.1) will monitor plant transients as required by PVNGS Technical Specification 5.5.5. Cumulative usage factors (CUFs) will be calculated for a subset of ASME III Class 1 reactor coolant oressure boundary vessel and 4.3.1 Fatigue Aging Management Program A2.1 B3.1 Metal Fatigue of Reactor Coolant Pressure Boundary A3.2 Metal Fatigue Analysis No later than two years prior to the period of extended operation1. DiDing locations, and Class 2 steam aenerator locations with Class 1 analyses. The enhanced Metal Fatiaue of Reactor Coolant Pressure Boundarv Drooram

Palo Verde Nuclear Generating Station License Renewal Application Amendment No. 14 (B3.1) will orovide action'limits on cycles and on CUF that will initiate corrective actions before the licensing basis limits on fatigue effects at any location are exceeded. - (1) The existing Metal Fatigue of Reactor Coolant Pressure Bounda.' program \\ "/ will be enhan ced to prov ide gui. de ines andt r e t trackinq eotn I Trn inpnrT'ý'CI 1 iag*lR fP* I*TP*P!P P* nm nnnn,' v i" 'ifn I E)ftwa re, to m aiRtain the fatigue usage faGtOF of these less thaR 1.0. The eRhaRGed pFegFam will iRGl6ide traGkiRg G GUmulative usage, GE)URtiRg of tFaR6i8Rt GyGles, maRual F8GGFd*Rg Gf BeIG tFaRsieRtG, review ef plaRt GyGle data, aRd Feview of the resulteRg Usage faG data. (2) The Metal Fatigue E)f ReaGtE)F GeelaRt PFesswre BE)URdaFy pFE)gram will 8RhaRGed tG !RGlude a GGmputeFized PFGgFaFR te tFaGk aRd maRage bGth Gy GGURtiRg aRd fatigue usage faGteF. FatiguePreo will be used fOF GyGle GGURtiRg aRd GyGle ba6ed fatigue (GBF=) MeR!tE)FiRg MethGGIG. FatigueRre-s an FE IiGensed pFed6IGt. A fatigue MOROWFORg seftwaFe pFogram that iRGE)rpoFate thFee dnmeRsiGRal, 6ix-elemeRt FRGG181 Me8tiRg ASME III NIR-32 Q FeqUiFemeRtS Well be used f9F r*ess 4 )ROWFORg (94F+. (3) The Metal Fatigue of ReaGtGF C006At Press, re B96IRdary PFGgFam well enhaRGed tO iRG'Ljde additiGRal GlasS 1 IE)Gat*E)Rr:, with high GaIGUlated GUmulative usage faGtOFS, Glass 1 GOMPOReRtS fGF WhiGh tFaRGIeF 41RGtieRs 4ave be-e-A.. developed feF 64866 based ME)RwtGF*Rg, aRd Class 2 POFtiORF, ef thB steam geReFaters with a Class 1 analysis aRd high GaIGUlated GUFAulative usage faGtOFS. The 6peGifiG IGGatieRs aFe 166ted iR Table 4.3 4, "Summapy-qf r=ý+w i eý 1 inn ^ frný rlpoý 1 Ann! QzQ nRd Methed Af M;;P;;c4P eRt bv the %I I I ýI Metal Fatigue E)f ReaGtGF Cc)G!aRt PFer,6WFe beunawy wregFam-.- I 1 (4) The Metal Fatigue c)f-ReaGtGF Geolant PFessuFe BOURdaFy prGgram wlil - I I

Palo Verde Nuclear Generating Station License Renewal Application Amendment No. 14 ennancoaI witn aciawonaia cycle count ana1 ;atigue usage action "FARS inciuain~g appropriate correctivo actions to be invoked if a component approaches a c'cle count action limit or a fatigue usage action limit. Action limits shall be chosen with the intent that they will permrit completioR Of rr*cO-tlve actiR6o befere the design limits are exceeded. (RCTSAI 3246934) LRA Table A4-1 Commitment No. 55, page A-59, is being added as follows (new text is underlined): The transient in UFSAR Table 3.9-1 Sheet No. 9 Item No. I.E.1.b, and Sheet No. 18, Item No. III.A.1.f, "Standby to SI hot lea iniection check valve stroke test to standby (usingq the HPSI pump)," will be added to the cycle counting surveillance procedure 73ST-9RC02 by Augqust 25, 2010. 4.3.1 Fatigque Aging Management Program (Table 4.3-2, Row No. 25 (RCTSAI 34R9fl24~i I I

ENCLOSURE 3 Markup Showing Changes to LRA Section 4.3.1

Section 4 TIME-LIMITED AGING ANALYSES 4.3 METAL FATIGUE ANALYSIS This section addresses design of mechanical system components supported by fatigue analyses; and also of components whose design depends on an assumed number of load cycles without a calculated fatigue usage factor. Section 4.6, ",,RtainRmnt incrF Plate, Equipm..t Hatch anRd Por..onel Air

ILocks, PeRetratins,, and PolaF Grane Brackets-s,"4 6 describes fatigue in the containment vessel.

S rctiot.n 4.7.4, "Fatigue CQrack Grot' h and, Fracture MechaRics Stability Analyscs of Half Nezzlo Repairs to Alloy 600 Material in Rcactor Coolant Hot Legs; Absence of a TLAP. forF Suor,* ting C;orrosionR Analyses,"Section 4.7.4. describes corrosion and fatigue crack growth and stability in the primary coolant nozzles. Fatigue analyses are required for piping, vessels, and heat exchangers designed to the American Society of Mechanical Engineers Boiler and Pressure Vessel Code, Section III, Rules for Construction of Nuclear Power Plant Components, Division 1, "Metal Components," Subsection NB, "Requirements for Class 1 Components" (ASME III Class 1).1 Fatigue analyses may also be invoked for Class 1 pump and valve pressure boundaries. Fatigue analyses are required for portions of the reactor pressure vessel internals designed to American Society of Mechanical Engineers Boiler and Pressure Vessel Code, Section III, Rules for Construction of Nuclear Power Plant Components, Division 1, "Metal Components," Subsection NG, "Core Support Structures." The design of piping and vessels to certain other codes and code sections, including ASME III Class 2 and 3, ANSI-ASME B31.1, and ASME VIII Division 2, may assume a stated number of full-range thermal and displacement cycles. Section 4.3 also describes fatigue analyses and evaluations of a limited number of other non-Class 1 components that were evaluated to these and similar rules. Basis of Fatigue Analyses ASME III Class 1 design specifications define a set of static and transient load conditions for which components are to be designed. Although original design specifications commonly state that the transient conditions are for a 40-year design life, the fatigue analyses themselves are based on the specified number of occurrences of each transient rather than on this lifetime. The design number of occurrences of each transient for use in the fatigue analyses was specified to be larger than the number of occurrences expected during the Titles are from the 1971 edition of the code, as used for the reactor vessel. Later editions reorganized the Section III material and removed the Division 1 title, so that this subsection became "Division 1 - Subsection NB, Class 1 Components". Palo Verde Nuclear Generating Station Page 4.3-1 License Renewal Application

Section 4 TIME-LIMITED AGING ANALYSES 40-year licensed life of the plant, based on engineering experience and judgment. This provides an allowance for future changes in design or operation that may affect system design transients. Operating experience at PVNGS and at other similar units has demonstrated that the assumed frequencies of design transients, and therefore the number of transient cycles assumed for a 40-year life, were conservative; and that with few exceptions the design numbers are not expected to be exceeded within a 60-year life. The exceptions are of two kinds. First, the NRC, industry, and specific plants, including PVNGS, have identified some transient loads on some components that were not foreseen in the original design process; for example thermally stratified flow in the pressurizer surge line and feedwater system, and Combustion Engineering Infobulletin 88-09, "Error! Reference source not found.." These cases have required evaluations to assess their significance and some have required revision to design specifications and analyses. Second, plant and industry operating experience has identified a few cases where cycles were being accumulated more rapidly than originally anticipated. At PVNGS, these were principally due to first-of-a-kind startup and shutdown cycles during the early plant life. The enhanccdl Metal Fatigue of Roactor Coo-lant Preessure Boundar' program wl Fatigue cycles are currently tracked in a PVNGS surveillance test procedure, 73ST-9RC02 "Reactor Coolant System Transient and Operational Cycles" as required by PVNGS Technical Specification 5.5.5 "Component Cyclic or Transient Limit."T In the text of this discussion the activities governed by 73ST-9RC02 will be referred to as the "current fatigue monitoring program". The enhanced Metal Fatigue of Reactor Coolant Pressure Boundary program (See LRA Appenbix B section B3.1) will continue to track events to ensure that appropriate re-evaluation or other corrective action is initiated if an action limit is reached. Action limits will permit completion of corrective actions before the design basis number of events is exceeded. See "Error! Reference source not found." in Section 4.3.1.5. The Industry Operating Experience Review (OE) program ensures that industry experience is evaluated and incorporated in plant analyses and procedures. The OE program includes review of experience that may indicate concerns with fatigue effects. Any necessary evaluations are conducted under the plant corrective action program. The OE program has remained responsive to both industry and plant-specific emerging issues and concerns. 4.3.1 Enhanced Fatigue Aging Management Program (B3.1) The current fatigue monitoring program is a cycle counting program with one location-specific cumulative usage factor (CUF) calculation (Pressurizer Spray Nozzle). No later than two years prior to the period of extended operation, the Metal Fatigue of Reactor Coolant Pre..ure Bo....r,'cu-.,nt fatigue monitoring program governed by 73ST-9RC02 will be enhanced to include additional location-specific CUF calculations and an automated and computerized management software program for cycle counting and fatigue usage Palo Verde Nuclear Generating Station Page 4.3-2 License Renewal Application

Section 4 TIME-LIMITED AGING ANALYSES factor tracking and management program.. The automated and computerized software program will be used to supplement manual counting. The enhanced Metal Fatigue of Reactor Coolant Pressure Boundary program (B3.1) will use cycle counting (CC), cycle based fatigue CUF calculations (CBF) and stress based fatigue CUF calculations (SBF) (see methods discussion below) to monitor fatigue. FatiguePro will be used for cycle counting and cycle-based fatigue (CBF) monitoring methods. FatiguePro is an EPRI licensed product. Fer stress based fatigue monitoring (SBF), APS commits to the use of a fatigue monitoring software program that incorporates a three-dimensional, six-element miodelstress tensor method meeting ASME III NB-3200 requirements,-and for stress-based fatique monitoring (SBF), APS also commits to the implementation of this method for SBF monitoring at least two years prior to the period of extended operation. The enhanced Metal Fatigue of Reactor Coolant Pressure Boundary program LB3.1)will monitor plant transients and cumulaticas reguired by PVNGS Technical Specification 5.5.5. Cumulative usage factors (CUFs) will be calculated for a subset of ASME III Class 1 reactor coolant pressure boundary vessel and piping locations, and Class 2 steam generator locations with Class 1 analyses, to ensure that reevaluation or other corrcctivc. The enhanced Metal Fatigue of Reactor Coolant Pressure Boundary program (B3.1) will provide action is initiated if an a*ction limit is rcached. Action limits on cycles and on CUF that will permit completion otinitiate corrective actions before the licensing basis limits on fatigue effects, in all locations, at any location are exceeded. Scope -The PV4GSMetalscope of the enhanced Metal Fatigue of Reactor Coolant Pressure Boundary program will monitor the (B3.1) will include all ASME Section III Class I components and piping isted'Class 2 portions of the steam generators with a Class 1 analysis and will monitor their fatigue usage by a combination of cycle counting and CUF monitoring ia R Tb 1.3. -. Methods The "GlebalCycle Counting" (CC) monitoring method in Tabe -Table 4.3-4. means that the fatigue managementenhanced Metal Fatique of Reactor Coolant Pressure Boundary program (B3.1) will not periodically calculate accumulated fatigue usage at the location. However, transient event cycles affecting the location (e.g. plant heatup and plant cooldown) will be counted and tracked to ensure that the numbers of transient events assumed by the design basis calculations will not be exceeded. "Global Replaceable" applies to bolting with pFred'icted lifetime usage factFrs greater than 1.0, and Which WIll therefore be replaced a requ44ed Cycle counting is the method used by the current fatigue monitoring program for all monitored components except the pressurizer spray nozzle. Cycle counting ensures fatigue usage does not exceed 1.0. It is employed as the preferred method of monitoring due to its simplicity. Palo Verde Nuclear Generating Station Page 4.3-3 License Renewal Application

Section 4 TIME-LIMITED AGING ANALYSES nputo a "real tffiMe" stFress histoFy foFr a g; S-tress b--hased fatigue (SBF=) moni~tGorng Will component from actual temperature, prcssl. -GGF mv, C3 H VW C)ttýpr_ lvz). C, hilgh fatigue components where a. more retined approach is necessar,' to show long teFrm Structural acceptability. SBF' monitorig depends On "global to l" crrelation or "tRnsf** " func~tions6 which calculato local tranAsiont pressuros and temperatures fFrom data collected by the lim~ited-number of plant instru1ments, and from themn, local stresses and fatigue usage-. Cycle-based fatigue (CBF) monitoring will consist of (a) automated cycle counting; supported as needed by manual data entry for infrequent events, and (b) CUF computation based on the counted cycles. it is intended fo,,r co.mpoents where long term structur acceptability can readily be shown based on cyclo_ coun-Its alone. Three CBF methods will beThe component CUF contributions due to each cycle are determined from the component Class I fatigue analysis. CBF is a more complex and resource intensive method than CC because it goes beyond counting of cycles to evaluate the CUF contributions of each cycle. Three CBF methods will be used, Per-Cycle CBF (CBF-C), Per-Cycle CBF with partial cycles (CBF-PC), and Event-Pairing CBF (CBF-EP). The CBF-C and CBF-PC methods will compute fatigue usage for a component by determining a location-specific fatigue usage increment for each counted event, and then adding up those increments for all events in the cycle record. CBF-PC will be used for some components, where the fatigue severity of individual plant events can be scaled using partial-cycle curves. A partial-cycle curve assigns a fractional severity to a cycle, as compared to a full design cycle, based upon significant characteristics of that event, such as temperature difference or heatup rate. CBF-EP is derived from the application of Miner's rule for combining fatigue effects, under the guidance of ASME III, NB-3222.4. This method will use an event-pairing table which assumes that the effect of pairs of monitored events is equal to the effect of similar pairs of design basis events. Stress-based fatigue (SBF) monitoring will compute a "real time" stress history for a given component from actual temperature, pressure, and flow histories. SBF monitoring uses data collected from existinq plant instruments to calculate local pressure and temperature, and the corresponding stress history at the critical location in the component. The stress history is analyzed to identify stress cycles, and then a CUF is computed using the formulas defined in ASME Code SectionIll sub-article NB-3200. SBF is the most complex and resource intensive method of fatigue monitoring, but it is the most accurate method and requires fewer conservative assumptions than CC or CBF methods. Corrective Action Limits Corrective actions will be initiated whenever a eyle Ic ount or fatigue usage action limit is reaGched. A.ctieo lmits will4! The enhanced Metal Fatigue of Reactor Coolant Pressure Boundary program (B3.1) corrective actions will be initiated whenever an action limit is Palo Verde Nuclear Generating Station Page 4.3-4 License Renewal Application

Section 4 TIME-LIMITED AGING ANALYSES reached, for either the number of transient cycles or calculated fatigue usage factor. In the current fatique monitoring program cycle action limits are set at 90% of the allowed cycles for each transient, and a CUF action limit of 0.65 is set for the pressurizer spray nozzle. In the enhanced Metal Fatigue of Reactor Coolant Pressure Boundary program (B3.1) corrective action limits will be set to permit completion of corrective actions before the design basis number of events is exceeded, or before the cumulative usage factor exceeds the code limit of 1.0. See Section 4.3.1.5 for the description of these actions and action limits, for the basis for the margins between the fatigue usage factor action limits and the code usage factor limit of 1.0, and for the basis for the margins between the cycle count action limits and the design basis cycle count assumptions. Analytical Margins Fatigue analyses incorporate several conservative assumptions and methods. These ensure that usage factors predicted by the design calculation will exceed (or "bound") the usage factors actually accumulated by the components.-. These conservatisms are discussed below. Fatigque Desiqn Curve with Margin for Uncertainties and Moderate Environmental Effects: The ASME Section III fatigue S-N curves (allowable alternating stress intensity versus number of cycles) are based on regression analysis of a large number of fatigue data points for samples strain-cycled in air, with adjustments for the elastic modulus and departure from zero mean stress for elastic cycling, less a design margin for uncertainties, including modest environmental effects (ASME III - 1965, Par. N-415). The design margin is a factor of 2 on stress or a factor of 20 on cycles, whichever produced the lower, more conservative allowable for the data set. Bounding Parameters for Transients: Fatigue analyses assume a given number of cycles of each of a set of transient events, each transient evet* is dcfiRed by liiti*ng pressure and temperatur. transients and other. lad conditions.. Actual event cycles are seldom as severe as those considered in the analysis; the resulting stress ranges are lower, and the contributions to cumulative usage factor are therefore lower. Use of Stress Based Fatigue: Since an automated six-element stress based tensor fatigue calculation will calculate stresses from the actual event severity, usage factors reported by the software program at locations for which the stress-based method is used will be more realistic than values predicted by the code analysis for the same number of cycles, or which would be determined by cycle-count monitoring. "The stress based algo.ithms will accurately calculate the actual fatigue effeGts. The automated six element Stresz monGitGorig 5oehware will use the sam~e metho-ds -AS an ASME III code analysis to ca cUlates a three dimensional, SiX com;ponent state of stress at critical locations monitored by the SBF= methodology.- Palo Verde Nuclear Generating Station Page 4.3-5 License Renewal Application

Section 4 TIME-LIMITED AGING ANALYSES Actual Number of Event Cycles versus Design Number of Cycles: The analytical limit for a fatigue analysis is a cumulative usage factor at any location of 1.0, Galculated . The design CUF is the sum of all contributing partial usage factors fejresultinq from the total of all design basis events at their design number of cyles, of each of the design basiG Gcycli e-Rt&7occurrences. Therefore, even if the analysis showed a calculated usage factor at4heof 1.0-limit for a location, and even if the design basis number of cycles were reached for one event-ef--a-set, the fact that all contributing cycle types will not simultaneously arrive at their assumed limit indicates that some fatigue margin would remain to the 1.0 limit. For locations for whichAction limits in the enhanced Metal Fatigue of Reactor Coolant Pressure Boundary program maintains a c'urrent estimate of fatigue usage factor based on cyclocuWRnting, action limits are(B3.1) will be set below the cycle count assumed by the analysis to ensure that actual plant experiencecomponent usaqe remains bounded by the assumptions used in the design calculations, or that appropriate reevaluation or other corrective action is initiated if an action limit is reached. Action limits will permit completion of corrective actions before the design basis number of events is exceeded. Therefore, the enhanced Metal Fatique of Reactor Coolant Pressure Boundary program (B3.1) will ensure that there is ample margin to the cumulative usage fa*t*r analyticC-UF limit of 1.0. 4.3.1.1 Licensing and-Desig*-Basis of the PVNGS Component Cyclic aP41or Transient Limit Program The "Component Cyclic or Transient Limit" program is required by Technical Specification 5.5.5 which states: "This program provides controls to track the UFSAR Section 3.9.1.1 cyclic and transient occurrences to ensure that components are maintained within the design limits." UFSAR Section 3.9.1.1 includes, by reference, information and transient definitions from several UFSAR sections and tables, which represent conservative estimates for design purposes listed--in(see Table 4.3-1Table--4-.31)4 The F-SARUFSAR states that this information accounts for all expected transients, and that the number and severity of the design transients exceeds those which may be anticipated during the 40-year life of the plant. Table 4-.3 PVNGS Unit 1, 2, and 3 Licensing and Design Basis Transient Citations from UFSAR 3.9.1.1 or Tablecto Applicable Scope of Transient Data Section 3.7.3.2 Operating Basis Earthquake (OBE) Cycles Table 3.9.1-1 ASME III Class 1 Components by the NSSS Vendor (CE) Table 3.9-1 ASME III Class 1 Piping Not by the NSSS Vendor (CE) Section 3.9.3 ASME III Class 2 and 3 Components" Section 5.4.1 Reactor Coolant Pumps Palo Verde Nuclear Generating Station Page 4.3-6 License Renewal Application

Section 4 TIME-LIMITED AGING ANALYSES Table 4-.3 PVNGS Unit 1, 2, and 3 Licensing and Design Basis Transient Citations from IJIFAR P.39 1 1 oecuon o..z_ oeam ueneraLUrs Section 5.4.3 Reactor Coolant Piping Section 5.4.10 Pressurizer 4.3.1.2 Enhanced PVNGS Fatigue Management Program (B3.1) The enhanced fatigue managoment program. (Metal Fatigue of Reactor Coolant Pressure Boundary program (B3.1) differs from the current fatigue monitoring progqram in the followinq two respects:

1) The current fatigue monitoring program is a manual cycle counting program. The enhanced Metal Fatigue of Reactor Coolant Pressure Boundary program (B3.1) will include an automated and computerized software program to support safe operation of PVNGS for the period of extended operation. The, enhanced program

2) The current fatigue monitoring program is cycle based and includes only one specific location for CUF monitoring (pressurizer spray nozzle). The enhanced Metal Fatigue of Reactor Coolant Pressure Boundary program (B3.1) will continue to count cycles and will also monitor fatigue effects for a subset of ASME 111 Class 1 react.or coolant pressure boundar,' vessel and *.i**gCUF values as specified in Table 4.3-4 for bounding locations, and Class 2 sta g.ncrator seconday,'

side subject to environmentally assisted fatigue (locations with Class 1 analyses. Table 4.3 2 lists those plant transientS that formq the basis for the cyclic duty for which cOMPOnontS weFe de-EiOg identified through implementation of NUREG/CR-6260). Usage factor monitoring will include environmental effects at NUREG/CR-6260 locations. if the limiting value for the transient is not stated in the UFSAR; the limitin~g valueis determined by the limiting number of transients fromn design specifications of affected systems and componRents, unless otherwise noted-. INSERT TABLE 4.3-2 HERE Palo Verde Nuclear Generating Station License Renewal Application Page 4.3-7

Section 4 TIME-LIMITED AGING ANALYSES "44 Table 4.3 PVNGS Unit 4, 2ý, and 3 L enfsiflg and' %.gn.B3sIs T~se I'/

2.

PlantC,*, ldvwn. 100

7. Normal Plant VariationR 40W y0

.4,R,09 f~oo 800 500 + I~~.

• ÷ d O/_ D,,,-A-,,r 11,000

[m I.. I::1 ' 1 80 Divo*,rion*, to", RWT/{ Pump Casing-Studs I__N&___ 2-5 4U0 4-60 _[40 144 40 1 500.800 of~Evcnts>~

23. Reactor Coolant Pump CoAAt~dow~n. ;at 100%0 Power

= 4-0 _

21. Reactor TriD 5Wý40

,.4o__40 4-0 fr-om 100 to IS% PowerP) 40 40 240 200 ,sseR* byD,,*^ 40 40 ,ssembly Wthdr;awal 40 40 ",,200 200 y W*,,o Woo 8oo t I 11-1._ _4.4nnot n 40 IRA-:- 0 A..- all_ 1__ 40 1 40 40 40 Palo Verde Nuclear Generating Station License Renewal Application Page 4.3-8

Section 4 TIME-LIMITED AGING ANALYSES Table 4J.3 2 RV.'S Unit 1, 2ý, andi 3 LJceRnsg aRC '-iiusign Ras~s Tmnsicnrs .nadvc.t.nt Actuation of Mai Steam nc seolation -Valve Ib4u 1 44v- _l __ IP* l

• JL l

.s~. UDCflIfl~ une ~~mos~neric uump vaivc or ~icam ~vpass Valve at 100% -Poer

• S 40 40

39. Se,6.,r Ev.nt Up t ad,cluding One Half of the Safe Shutdown Eahguake, at 100% Power 10.'Initiation of Safety Injection 4-0 4-0

41. Inad'ertent Isolation of Feedwater Heater 5

5

12. Loss o~f Feedwater Flow to) Stcam G8nerP.atorFS 85 423 Loss6 of Reactor-Coolan PumFP Seal Coolant NI 4-0

44. Loss of Reactor CGoolnt Pump Scl lnjcctien 40

45. Inadver-tent Auxiliary Spray at 1002% Power 5

5

46. Systom Leak duo to Rupture of Inl÷trumnt LinoR or Sampling 40 40 Ge~ne~tiop

17. Inadvertent Main Feedwater Isolation Valve Closure 40 40 at 100% Power- (One MFIY 18..Inadvertent...........or.Conden.at Pump Trip 40 40 at 100% PowL-r

49. Main Feed'ater Isolatfion Val'e Closures due to Loss of Amr at 100% Power

51. Startup Of One Reactor Coolant Plump at 50% PoweR 4

40

52. Lo.,of~

Elcti~ca.UsPplYing two Reactor Coolant 40 40 Pumps at 10094 PowAMer

53.

n .adver-t,-t Closure of All Main Fc1dwater isolation Valves a t IQ0Q%_ Pow_;er

54. Spurious Startup Or Shutdmown of SI PumRp, Or Spurious400 Opening or-Closing of SI Isolation Valve 55 PimrySide Hydr-ostatic Test, 3115-psia-, 10 00 4-0 4-0 56m Seondarny Side Hydron-static-Test 4-0

-140

57. Primary Side Leak Test, 200 2-00 2250 psia, 10 100 F

58. Seco~nday Side L eak Test, 820 psia to Design Pressure 200 200

59. CVCS System Hydrostatic Test 40 40-

60. Low Pressure Safety Injection PUm Ts;,

1 560

61. High Pressure Safety Injection Pump Tesý 1

&00 Palo Verde Nuclear Generating Station License Renewal Application Page 4.3-9

Section 4 TIME-LIMITED AGING ANALYSES 4.3.1.3 Seismic History Design analyses that compare seismic stresses against stress al*owa;-blesallowable stresses, in the absence of any consideration of the number of cycles or of fatigue effects, are not TLAAs. However, design of structures, systems, and components may include seismic loads in fatigue analyses, or may assume a stated number of seismic load cycles for purposes of establishing an allowable stress or stress range. Significant earthquakes at the site can therefore increase the accumulated fatigue usage factorj-e*. The site seismic history can rFledeaffect the a.a... eudisposition of TLAAs. However, no significant eatquake load cycles assumed, by the dcsign, to be allowed fo-r tho-ropmaining opcrating tfeearthquakes have occurred at PVNGS since construction. The site sci.mic h*str,'¥ cGa thereby affcct the disposition of TLAA-.

Howover, Iinc construction no) Significant earthquakcs have occurred at PVNGS to date.-

For design purposes the PVNGS safe shutdown earthquake (SSE) and operating basis earthquake (OBE) are defined as 0.20 g and 0.10 g ground motion, respectively. Analyses of Seismic Category I structures used a conservative design basis 0.25 g SSE and 0.13 g OBE [UFSAR 3.7]. For the purposes of evaluating actual events at PVNGS, an SSE is defined as one with a modified-Mercalli intensity level 8 (ground motion of 0.15 to 0.33 g or above); and an OBE is defined as one with a modified-Mercalli intensity level 7 (ground motion of 0.072 to 0.15 g). No SSE or OBE has occurred to date. The site has recorded seven minor earthquakes as of 2008, some of these not strong enough to qualify as recordable "earthquake events." The strongest had a ground motion of only 0.015 g, or about 12% of the acceleration, and therefore the applied loads, of a design basis 0.13 g OBE. 4.3.1.4 Present and Projected Status of Monitored Locations Summary Description The current fatigue -maRagementmonitorinq program transient cycle count procedure, 73ST-9RC02, recorded accumulated transient events for the 9 transients, listed in Appendix J of the procedure since the Unit 1 startup in 1985. This transient list did not include every transient in the-F&SAR.UFSAR Section 3.9.1.1

because, prior to implementation of Improved Technical Specifications in 1998, the Technical Specifications required monitoring only transients that are now in UFSAR Section 3.9.1.1.1.

The-efere, n-In__1995 (after 10 years of Unit 1 operation), the cycle count procedure was revised to include the 48 remaining F-SARUFSAR transients listed in Appendix K of the procedure. In the 1995 record of the revised procedure, accumulation for all transient events not counted to date was assumed at 25% of the limiting value for the 40-year design. After the 1995 revision of the cycle count procedure, transients were recorded on a casc by cas basis and were added to the 25% accumulation assumed in 1995. Palo Verde Nuclear Generating Station Page 4.3-10 License Renewal Application

Section 4 TIME-LIMITED AGING ANALYSES APS Fatigue Cycle Count Verification The goal of the APS fatigue cycle count verification was to reduce the uncertainty created by the 25% accumulation assumed in 1995. Scope Transients adding significGant fatigue to components weoe included in the APS transient ,ecount. TraRnsie*tS Rno conRt*rib*lt*g sig*n*ic*atly to fatigue Wer8R nt inGcydd in the APS transient r.eount. The tran..ie*nts nt included in the recount are retained in the composite WorSt case unit accumFulation, incl61udig the 250% accumnulation assumed in 19915. The scope of the cycle count verification included all transients required to be monitored by PVNGS Technical Specifications 5.5.5. Recount Method Unit 1 was the prototype Combustion EnRgineering System 80 plant. D)ue to a lack Of operating experIene early Un~it 1 operation included tests and events that did no gInerally occur as frequently in subsequent units. A cycle count reer-d fromF Un~it 1 should there-fore be a conserwative estimate for Unit 2 and Unit 3. However, Unit 1 had a 460 day outage, w.ith Unit 2 Funning, but dul~ng which Unit 2 expe~ienced mnany startup shutdown tfansients. Therefore, APS has created a compesite worst ease (Gomposite u.it) enVelope including only the highest acuuainof each transient experienced amonGg the three units from 1985 through 2005. APS performed a best efforFt retrieval of the transfient counRt data recorded from 1985 thSrough 1995 (the "PS transient recourt"). Sources foe this effort iencuded (1) NRC -term mation Reports for all three urits, (2) unit 1 cnroli room logs freom 1985 throdgh 1995, (3) inritl control room legs from 1986 through 1995, (4) Unist 3 control roo logs feton 1987-thugh 1995, and (45) interiews with plant personnelw. The result ef this data retrieval is the "worst case APS transient freahut from 1985 through 1995." The 250% accumulation assumed in 1995 was subtracted from the totals recorded through 2005 in the cycle co)unt procedure to obtain the accumnulation from 1996 through 2005 foF each transient. This accumuliation from 1996 throeugh 2005 was then added to-the worst case APS transient recount frome 1985 thfouogh 1995 to obtain the ceompsiue worst case unit accumu~lation of cycleS from 1985 to 2005, for each transiet Several APS employees and contractor personnel were designated based on their long-term familiarity with PVNGS to perform document reviews. The reviewers examined the microfilmed control room logs. NRC Monthly Operating Reports and LERs for the period prior to January 1996 for all three PVNGS units. The personal recollections and records of unit personnel were used to supplement the record review, and a best-source total was determined for each monitored transient. The best-source total was added to the actual count of events following 1995 to obtain a best-source Palo Verde Nuclear Generating Station Page 4.3-11 License Renewal Application

Section 4 TIME-LIMITED AGING ANALYSES total as of the end of 2005. These best-source totals as of the end of 2005 are reported in Table 4.3-3. Transient Projections A yoarly accumulation rate must be calculatod in order to accurately project transient accu1mulation through the poriod of extended oper-ation. The yearly accumu~ation rate waScalculated by dividing the comnposite unit accu1mulation frFom 1985 through 2005 by the leastnumber of years of operation up to 2005 (Unit 3, operating pwerid of 18 years). T-his resulted in the WOrst case accumnulation of cycles over the least amo)unt Of time. This accumulation rate was then multiplied by 22 (18+22-40) and added to the composite un~it 2005 accumnulationR to calcu1late the projected accumu~lation at 40 years of o~peratin Similarly, the accumnulation rate was multiplied by 42 (18+42-60) and added to the composite unit 2005 accumul61ation to calculate the projected accumunlationR at 60 years oA Transicnts not hncluded in thee FSAR Some transients w.~hich are required by the fatigue management pro~gram to Pcuatl calcu1late fatigue usage are not required to be monRitored by the PVNGS ESAR, n were therefere not separately cOunted in the procedure through 2005. These tranlsients were therefoem included in the cycle count verificationR. Hoe'ever, there 'IS no accumulation record of these transient events from 1996 through 2005. APS has therefere deterMined accumuflation data fro~m. 198-5 thro-ugh 2005, the _accurwmulation rate was then ca*,*, e B

• V.I*

aGcc~umu=abEn By.. !,,e,ea.,.u,,b. e, years EA-* opemlra.n u1p te. 2005 (Unit 3, operating time of 18 years). Tr"a n s-ien.ts Mxilt-h -a to, da,,te a G G u cmIu. I fn of e ro The yearly accumun~lation rate for transients which to date have no accumulation was deteFFined by dividing the design basiS RUFnber of transient events by 10 years. This resul1ted in the eriginal expected annual accumulatfion rate of transients, excGept that no transients have occGurred to date. Therefore, the accum~ulationR rate was determ~ined-1by m:Ultiplying the original expected accumulation rate by the percentage Of years left in the denE~mR ass(22/40). Tr~an 4cnts not expected to occGuF No y* rly accumnulat*oRn ate was Gcalulated for transients which are not expected to eseUr. For these. transient events at least one event was-assumed to eccur during the period e~f extended operation. Palo Verde Nuclear Generating Station License Renewal Application Page 4.3-12

Section 4 TIME-LIMITED AGING ANALYSES Table 4.3 3 APS Fatigue cyrle Count VerXfiation (Cornpesite or-St Case L439), an~d Pre~ection#9_ Palo Verde Nuclear Generating Station License Renewal Application Page 4.3-13

Section 4 TIME-LIMITED AGING ANALYSES 4aQ~e 4.4 4 -:=ýt-'z -ai'gUe t6Yce GUrnf Ver1T4Gar9Gn (I&QR4pos8r VVGr&F (-ia~ 6nRty, ana t-"roeciln

13. Sh tt tram Normal to MimmPurFification Flow at 100% Po2wor 41700 250 4-gg89 934

14. Low Lo Vo lu*,o Co-ntRo Tank' Charging Pump 80 20 2-0 NG 20"

444 4-67-SuJction Di'ermciOn to RWT

15. Prossure Lovl e ntre!,

400 2C5 2-NG 2-5--4 6 48.4 -eFeae!-Ur t__Opon

16. Unbolting! Blting o 2-5 NR NR 4-9 41-"

4 4344 641-IRC Pump Casing Studs

17. TensRoning/

Dct*nsioning of RV 5G0 NR NR 7-1 0,94 38 57-1, Head4

18. Safety 'n.ction Chock 4160 NR NR 0

0 4-4 Valv~e Tz-e,

10. High Pressure Safety Injection Header Chec 40 NR NR 0

0 18 4 Valve Test__

20. Turbine Roll Test at H0t NR 4

4 Standby Palo Verde Nuclear Generating Station License Renewal Application Page 4.3-14

Section 4 TIME-LIMITED AGING ANALYSES Table 4.3-3 ARS Fatigue G.!ei Count Verifiat*ion (Composite Worst-Case Unit% and Prolectionst Palo Verde Nuclear Generating Station License Renewal Application Page 4.3-15

Section 4 TIME-LIMITED AGING ANALYSES ý -4 ajie if0

4.

-4 AW-arau 6Vce 60ouR Ver:TFc.M1ofl 160oRnOO611 VVowFS ý7a89 UR4. anfG.WF'rv;QGFuvnP CEA Withdrawal

30. Lo..

of Chaging 2-00 2-5 2"7-5 7" 04 -I' and Re~oery

31. Less

.f L.td. wn 243 24-1-7 20 4-- 45 67 an~d RennveW

32. Extended Loss 800 NR 64*

21 36 44-7-2 of-Letdown

33. Deprcssurization by Spurious Actuation of Prcssur-Izr Spray 4

4 Contro~l Valvo a 100% Power (Main &8 Aux.-S-Kay_

31. Partia Loss of CondCnser Cooling 40 40 414 NW 414*-

0644 25 3-at 100% Power-

35. EXGGSS Fccdwatcr-40 4-9 40 2

2 457 at 100% Powcr

36. Turbinc Tri p-1401 64 Wi~tha-0 40 4-0 4-5 4-9 49 4-.00 4

6 Reaeto~Tp Palo Verde Nuclear Generating Station License Renewal Application Page 4.3-16

Section 4 TIME-LIMITED AGING ANALYSES Table 4.3-3 APS Fatigue Gycle Count Ver/fication (Composite Wo4r-St-Case Unit), and PRojection s,4 inadve~tent Actuation of Main Steam Line Isolation VaWe 2 415 445 0.06/0.28 3912

38. Opening One AD1 OF Steam Bypa&&4040 4

4- -045-Valveat 400%-Pewe

39. SoismicG Event up to and Including Onoe Half of the S.

2 NR 4R NCr NCG 4N NG Sh4u44GWR Ea~thquake, at 400%-Pewef_

40. Initiation of Safety 4-0 NR 7-4 7-02 073-U6{-5

-46* InjeGtR

41. inadvertent Isoation ef 5

4 4 0 O 0-.07-{4 2-FW]4-eate

42. Losz6 of Foodwator Flw to Steam4 85 24 22-4 4142-0-.72 2-9 44 Palo Verde Nuclear Generating Station License Renewal Application Page 4.3-17

Section 4 TIME-LIMITED AGING ANALYSES Table 4.3 3 APS Fagtice CVcle COUnt VerFicatiOn (COrnBOSite WorSt-Case Unit), and ProiectionS 4 I / 4 Less-eof GGQeaR NR NR NG NG NG NG

11. r L OsRPoa 4.0 NR NR NG NG NG NG NG

5. In," adcr*tecnt Auxilia'y SpFa..at 5

4-2 0 1-4 400%-Pewe-_ 416. System Loak due to Ruptu o f n.trumnt. 40 4-0 40 0 0 ,55') 4 UR F Samlip-§

17. In.,id*veFtent MFIl ea..

40 4G 46 0 3 4 100% Powec

48. inadv'crtent FW or Condcnsate Pump 44 44 4-4 44

-1 0,64 2-5 Tripat 100% Power I I_11

49. MFIV cosU'rcs d-c to Lss ofA 5

4 4 1-0,D 4 400%-PGwef Palo Verde Nuclear Generating Station License Renewal Application Page 4.3-18

Section 4 TIME-LIMITED AGING ANALYSES ¢ L B upu I i I* Table '4.3 3 /U-'. Favtgue Gugvle C~ount Vermcathqn 4Ge~nposme Wbrst Case UP4~' and P-FOiectGns F*F=7 V ueffessurlz MSS3I-at 44OO%4zwef 40 2 5 G-.2-8 51 tatp of onc Reactre Coolant 4-0 NR N 4R NG N 0-.44(4* 4 6 Pump at 50% Po'wcr____________

52. Loss of Elcctrical Bus Supplying two 49 4-0 44 2

6 0,33 44 2-RCPset 440O-Powef_

53. Inadveot*nt Closuro of all MF!Vs at 5

NR NR NG NCG a) NG NG 400%4Pe's-c_ 54.~ Spurious Startup! Shutdown of-4 SI Pump or-Spurious 40 4-0 1-0 1-1 0_.G6 34 Opening! Clo i of SI Isolation Valve

55. PriayFd;

'*.....t. 4-NR 4 NC -9) } 2 2 3125 ps___ Palo Verde Nuclear Generating Station License Renewal Application Page 4.3-19

Section 4 TIME-LIMITED AGING ANALYSES Table 4.3 3 ARS Fatigue cycle Count Veification (Composite WorFst Case UnPt) and PFOjeGUGROns_' Palo Verde Nuclear Generating Station License Renewal Application Page 4.3-20

Section 4 TIME-LIMITED AGING ANALYSES Basis for Reduced Cyclc Counts Tra.psiont 18, "Safety InRjection Check Valve Test"; The limiting rumberp f 160 eventS ir the JF=SAR originated ftrom theobuto Enginern gener al s;pecification. Combustion EnRgineering plants su1bsoquently petitionRed the NRC (in the early 1980's) for perm~ission to not perfor* this quarterly test because of the signifiGa*t fatigue whiGh wou,'ld result from inserting cld Safety Inje** ti water. The quarterly test was neVe* pe,,Fomed and never incorporated in the procedure. The check valve test is performed during a stage of statu at normal heatup pressure and tempeature, resulting 4i no significant fatigue acc*umu.ation This transiet event is therefore ne. t pe.formed as originally characterized and analyzed and need not be traecrkdedt. reTransient 26, "Lss of Lad": The projected number of events may ot be Freached because the loses of load transient is ony sFRignificnt to fatigue when it causes a turbine trip. This is avoided by performing turbine runback. If co rrective actines to reanalyze compnents become Reccssa~y, the" ma" include a reiso of the definition of this transient event. TrFansient 37, "Inadvertent Actuation of MSIV":- ~imiting numbers in the UJF=S-;AR _arc 5 events fro)m 100% power and 40 events from an unspecified power level. Only onRe event for each unit has been recorded at 1990% power. PE~giaTransient totals were projected to the end of the PEO for information only. The proiections predict that 10 CFR 54.21 (c)(1 )(iii) aging management will be successful and that in most cases future corrective actions will not be necessary. However, the projections are not intended to justify 10 CFR 54.21 (c)(1)(i) validations, or to provide revised design bases for 10 CFR 54.21(c)(1)(ii) analysis revisions. The proiections are based on a linear extrapolation as follows: 1.The shortest period of operation as of the end of 2005 was 18 years in Unit 3 and the longest was 20 years in Unit 1, so a scaling factor of 3.33 (60 years extended life divided by 18 years shortest operation) was used to project totals to the end of the PEO. In a few special cases a scaling factor of 6.66 was used when the available data covered a ten year period. These exceptions are noted and explained in Table 4.3-3.

2. The highest total accumulation for an event was selected without regard to which unit it occurred in.

3. A highest unit 60 year proiection (column7) as of the end of the PEO was obtained by multiplying the highest total accumulation for each event by the scaling factor.

Example: Event #1 RCS Heatup highest unit total was in Unit 2 (64 heatups). 64 X 3.33=213. The highest unit 60 year projection (column 7) is 213. Palo Verde Nuclear Generating Station License Renewal Application Page 4.3-21

Section 4 TIME-LIMITED AGING ANALYSES It should be noted that only a few events such as recurring test events lend themselves to projection with well defined assumptions, so the projections presented in Table 4.3-3 are only best-estimates. Early plant history involved a number of first-of-a-kind issues that may make the projections artificially high, and end of life issues may make the proiections artificially low. Therefore, consistent with aging management, no attempt has been made to reanalyze or implement other corrective actions based on these projections. Corrective actions will be triggered by the action limits that will be established in the enhanced Metal Fatigue of Reactor Coolant Pressure Boundary program (B3.11) INSERT TABLE 4.3-3 HERE Palo Verde Nuclear Generating Station License Renewal Application Page 4.3-22

Section 4 TIME-LIMITED AGING ANALYSES 4.3.1.5 Enhanced Metal Fatigue of Reactor Coolant Pressure Boundary program (133.1) Scope, Action Limits, and Corrective Actions Scope The scope of the enhanced Metal Fatigue of Reactor Coolant Pressure Boundary program (B3.1) will include a bounding set of loc.ations within ex,.tinaLl ASME Section-III Class 1 vessel and pip.ig fatigue analyses. This set includes the NUREGlI R 6260 sample locGationRs. The sco~pe of the bounding set of monitored locations is Sufficientt enSUre that fatigue inR an" other Iocations of,. concern, n,-t irnIuded in the set, is withiR th came 6ystem and subjcct to the samne transients, or within a system affectcd byth same transients. Tablhe 4.3-- 1 reflects the-scope of the enhanced PVNGS fatigue mnanagement program. The enha*ned prgram will inlude (1) Class 1 Iloations with high calcu*latd cu.mulatie* usage factors, (2)/components listed in NUREG!CR 6260, (3) Class 1 components for which partial cycle equations have been developed for stress based moniRtOorig, and (4)-and Class 2 portions of the steam generators with a Class 1 analysis and high calculated cumulative usage factorswill monitor their fatigue usage by a combination of cycle counting and CUF monitoring as noted in Table 4.3-4. Method The "Fatigue Management Method" column of Table 4.3-4 indicates the method the automated softw-reenhanced Metal Fatigue of Reactor Coolant Pressure Boundary program (B3.1) will use to track fatigue usage for each component. These are stress-based fatigue (SBF), cycle-based fatigue (CBF-C - per cycle, CBF-PC - per cycle with partial cycles, or CBF-EP - event pairing), and,,4eba,-p..,C. The LgqýedaCC method will oGly-be used for components wi,,th low ca....ilcuated design basis fatig-u..whose cumulative usage values, for whichcan be shown to be satisfactory with this highly conservative monitorinq approach. Locations with high end of life accumulated usage, such as most of the fatigue management program does netNUREG/CR-6260 locations, require more sophisticated CBF or SBF algorithms to periodically calculate accumulated fatigue usage to date. However, transientand demonstrate that component usage remains less than one. Transient event cycles that have significant fatigue effectsare required to be monitored by PVNGS Technical Specifications 5.5.5 will continue to be-Eeunted and tracked to ensure that the numbers of transient events assumed by the design basis calculations will not be exceeded. This "global" coveragecycle counting monitoring method will therefore suffice to demonstrate design basis compliance for the components using CC monitoring. See Table 4.3-2 for the list of tracked transients. Corrective Action Limits and Corrective Actions The PVNGS current fatigue managernetmonitoring program Gt-reP vyis based on cycle counting with one location tracked by a CUF calculated using CBF-PC (Pressurizer spray nozzle), and it incorporates a cycle based action limits that provide for eva!uatiOn and cycle count tracking Of critical thearmial and pres~sure transients to Verify' that the ASME Code CUE limit of 1.0 and Gthe-g90% of the design limits will not be exceeded. T-heevent occurrences and a CUF based action limit of 0.65 for the pressurizer spray nozzle usage. The current fatigue monitoring program requires this evaluation at least once per fuel cycle. Action limits arc based ,o a fixed percentage Of allowed Palo Verde Nuclear Generating Station Page 4.3-23 License Renewal Application - Revision C

Section 4 TIME-LIMITED AGING ANALYSES cycles for components monitored by a maximum number of defined transients. The current action limits are established to allow action to be taken in time to prevent exceeding the maximum number of allowed cycles or a pressurizer spray nozzle CUF of 1.0, as applicable, and should provide at least one fuel cycle of warning. The enhanred prog*am specificS Gorrective actions e tobe implemeRnted to e re-hat appropriate reevaluation Or other corrective action) is initiated if an action limit is reached. Durinq the period of extended operation, projections indicate that certain allowable cycles and fatigue limits may be approached. Therefore specific and targeted action limits will be necessary to ensure actual fatigue limits are not exceeded. Those action limits have not yet been developed. As the transition to the enhanced Metal Fatigue of Reactor Coolant Pressure Boundary program (B3.1) and FatiquePro © is implemented, there are certain embedded administrative tools in FatiquePro © that will allow for specification of action limits based on proiected fatigue usage at specific locations that account for actual cumulative fatigue. The action limits can be based on the time required to implement expected or projected mitigating actions (such as component replacements or revisions to ASME Code Fatigue Analyses of Record) prior to actual fatigue limits being exceeded. Action Limit Margins Corrective aGctio limits mustThe enhanced Metal Fatigue of Reactor Coolant Pressure Boundary program (B3.1) corrective action limits will ensure that corrective actions are taken before the design limits are exceeded. Corrective action limits must thereferewill ensure that appropriate reevaluation or other corrective actions are initiated while sufficient margin remains to allow at least one occurrence of the worst case (highest fatigue usage per cycle) low probability transient that is included in design specifications, without exceeding the code limit CUF of 1.0. For NUREG/CR-6260 locations, CUF calculation will be done using the appropriate Fen environmental factor. Cycle Count Action Limits and Corrective Actions F-er-Cycle Based F=tigue Counting monitoring (GB+,FCC) action limits have beeffor the enhanced Metal Fatique of Reactor Coolant Pressure Boundary program (B3.1) will be established based on the design-specified number of cycles. Usage factors in locaPtions monitored by this; mothod -are Mos-1t a-ffected by transient events; which a;re of -hw p-**,a*liy, aRn cycle countig oGf noso eVents 1s nRcror SUFFIicent !Ao arconun Tor the fatigue accumnulation in them. Cycle CouRt Ac^tion L-imit Margins: I'n orde*_r o .assure Since sufficient margin must be maintained to accommodate eGGU-r-eaeany design transient regardless of probability, the enhanced Metal Fatigue of a low probability t-asiet;,*Reactor Coolant Pressure Boundary program (B3.1) corrective actions inw-stwill be taken before the remaining number of allowable occurrences for any specified transient, including the l*ow probability, higher usage factor eve-nts, becomes less than one. Othe...ev. counted by cycle based monitorFing contribute less per event to usage factorF, but occur moRe frequeRtly. To account for both cases, corrective Corrective actions a-ewill be required when the cycle count for any of the significant contributors to usage factor is projected to reach the action limit defined in the pFegr ffenhanced Metal Fatigue of Reactor Coolant Pressure Boundary program (B3.1) before the end of the next fuel cycle. Palo Verde Nuclear Generating Station License Renewal Application - Revision C Page 4.3-24

Section 4 TIME-LIMITED AGING ANALYSES For example, i* Table 4.3 3 the specified umber of "Inadve-te* t Auxilia*y SpFay ft 100-%, Power" transient eventS is (5) Se corrective action would be required when 80% (4) of the specificd cyc!es have occurred. Cycle CeuPtCounting Corrective Actions-to be incorporated into the enhanced Metal Fatigue of Reactor Coolant Pressure Boundary program (B3.1): If a cycle count action limit is reached, aGeeptable corrective actions twIwdewill be performed as necessary:

1) Review of fatigue usage calculations.

To determine whether the transient in question contributes significantly to CUF. To identify the components and analyses affected by the transient in question. To ensure that the analytical bases of the high-energy line break (HELB) locations are maintained. To ensure that the analytical bases of athe fatigue crack growth and stability analysis in support of relief from ASME Section XI flaw removal and inspection requirements for hot leg small-bore half nozzle repairs are maintained.

2) Evaluation of remaining margins on CUF based on cycle-based or stress-based CUF calculations using the PVNGS fatigue management preqraRmsoftware.

3) Redefinition of the specified number of cycles (e.g., by reducing specified numbers of cycles for other transients and using the margin to increase the allowed number of cycles for the transient that is approaching its specified number of cycles).

4) Redefinition of the transient to remove conservatism in predicting the range of pressure and temperature values for the transient.

Since the CBF= actieR limits arc based On a somewhat arbitrary cycle coun~t that does not aGcuratoly indicate approach to the CUE - 1.0 fatigue limit, theseThese preliminary actions are designed to determine how close the a...,aGh-usae is to the-I.0 4im#i, and from those determinations, set new action limits. If the CUF has approached 1.0 then fuihep Further actions for cumulative fatigue usage action limits may be invoked if good engineering iudgment determines that is necessary. Cumulative Fatigue Usage Action Limits and Corrective Actions T-h4eThe enhanced Metal Fatigue of Reactor Coolant Pressure Boundary program (B3.1) will use an automated three-dimensional, six-element stress tensor, stress-based fatigue management proir, software module (the SBF module, meeting ASME III NB-3200 requirements) wilto continually monitor cumulative usage factor (CUF) at the stress-based fatigue monitoring locations, and cycle-based CUFs will be calculated periodically. The CUF action limits for the enhanced Metal Fatigue of Reactor Coolant Pressure Boundary program (B3.1) will be revisedestablished to provide two to three fuel cycles of warning prior to exceeding a CUF of 1.0. Palo Verde Nuclear Generating Station Page 4.3-25 License Renewal Application - Revision C

Section 4 TIME-LIMITED AGING ANALYSES CUF Action Limit Margins: To provide adequate time for corrective actions and adequate margin to permit continued operation, corrective actions for the enhanced Metal Fatigue of Reactor Coolant Pressure Boundary program (B3.1) will be required when calculated CUF (from cycle based or stress based monitoring) for any monitored location is projected to, reach 1.0 within the next 2 or 3 fuel cycles. In order to assure sufficient margin to accommodate occurrence of a low probability transient, corrective actions must also be taken while there is still sufficient margin to accommodate at least one occurrence of the worst case (highest fatigue usage per cycle) design transient event. Action limits will be established to permit completion of corrective actions before either-:the usage factor "iPAA-efreaches 1.0 or the design basis number of events, as applicable, is exceeded. For PVNGS locations identified in NUREG/CR-6260 and described in Section 43 3.4, "Effects of the Reactor Coolant System Environment on Fatigue Life of Piping and Components (Generic Safety Issue 190)," this action limit lswill be based on accrued fatigue usage calculated with the Fen factors required for including effects of the reactor coolant environment. For example, if inadvertent RCS depressurization, when adjusted for the environmental effects of the reactor coolant system at a NUREG/CR-6260 location, causes 20% of the total allowable fatigue usage, corrective action for that location would be required before calculated usage (including the environmental effects factor, Fen) reached 0.8. CUF Corrective Actions: If a CUF action limit is reached, aGGeptable-corrective actions

• iekudewill be performed as necessary:

1) Determine whether the scope of the menitcringenhanced Metal Fatigue of Reactor Coolant Pressure Boundary program (B3.1) must be enlarged to include additional affected reactor coolant pressure boundary locations.

This determination will ensure that other locations do not approach design limits without an appropriate action.

2) Enhance fatigue monitoring to confirm continued conformance to the code limit.

3) Repair/modify the component.

4) Replace the component.

5) Perform a more rigorous analysis of the component to demonstrate that the design code limit will not be exceeded.

6) Modify plant operating practices to reduce the fatigue usage accumulation rate.

7) Perform a flaw tolerance evaluation and impose component-specific inspections, under ASME Section XI Appendices Aor C (or their successors), and obtain required approvals by the NRC.

Palo Verde Nuclear Generating Station Page 4.3-26 License Renewal Application - Revision C

Section 4 TIME-LIMITED AGING ANALYSES Table 4.3 4 Summary of Fatigue Usage from CPass 4 Analyses, and M~ethed ot M~anagement by the M4etal Fatiguo of Reactqr-Coolant Presr 'Bo "ndx Palo Verde Nuclear Generating Station License Renewal Application - Revision C Page 4.3-27

Section 4 TIME-LIMITED AGING ANALYSES T";-hl1 4 2-4 Summar..y of Fat.gu. Usage from Glass i Analyso, and Method o Management by the M4etal Fat~ige of Reactor Coolant PrFeSsUre BOUndar Palo Verde Nuclear Generating Station License Renewal Application - Revision C Page 4.3-28

Section 4 TIME-LIMITED AGING ANALYSES -IQpLR a -i-a iiýýn ^ jiý n ý m L, L, r- /QQ.Q =Ra PAQFRGG 0 1.y M Jýý I M'lanageme nt bv the Metal -a tioue of Rea Itnr G-A-1anpt l-rsai4r ~ou.ncdar v -1 Weld U1&3 - 0.1881 I2 I.q-- 0.7921 Bounding lccatien

Global, OUR-ndOd DrwnrArcmor FW

77. RUSG. Tubeshect_

Global, Boundod Hot Side WU1&3 - 0.0657-0 Boun~ding loctio by__________ _Nez-ze& Global, Boundcd Cold Side Bounding locbtion I o-n omr izeF

78. RSG Tu,,*bcset. to Shell (Stub Bre)Jnto

.Ho SUP .10 Bo)unding location Global U2-9,996 Co*d.,, e 1U1&3-,--I h,-,.-8i,,

79. RSG Economizer Cylindcr at Tubeshcet Cold Side (Uni 1_1 U132 - 0.0 1 07-5 Bounding loaion Global UI&3 - 0.00n*(

80. RSG Sccondary Shcll U1_2

-0.0099-Bounding location

Geba, 81.RS D

enomr Blowdown U 1&3 - 0.197 Bounding location Global Nozzle U2--0,273

82. Ra-Rcci-rulati;n Nozzle U4&

l3 Q_99 BoundingloainGlbal RSG Steam Outlet Nozzle I2c-Bo un.....d.ing location Glo U2-- 00176 7-

84.

RD_-G S ndary Manway Bn* ,2 B-ding lcation Globa pad U2-0Q440 .. RG Spccondar; Manwa. U.&3 - 0.6.. Bounding lccation Studs U-e0.77!4 Replaeeable

86. RSG Wcldcd Sccondar, W1_013________

Handhces (Ulnit II & 3 Only) U = Bounding loctio..... 8.RRG Wcldcdi Seccndary 9oal Handholc St, ds (Unit 1 & 3 U1 R3 - 0.4124 Bounding location Rl

88. RSG Stub Barrel Sccondar

1. &-

0 A4O U2 - .955High CUPF GBF--E-P-HaRdhole U2 -04.55 8-9.,S Su Br-rcl Sccndary Boudin locatio Global Handholc Studs U2-245 g')aiR1;p;l

90. RSG U per-Support Lugs 11m2---.0-5 Bounding location Glebal

_U~2-4.464 I I Palo Verde Nuclear Generating Station Page 4.3-29 License Renewal Application - Revision C

Section 4 TIME-LIMITED AGING ANALYSES Table 4_.24 S,,,mary of Fatigue Usage from Class " Analys8es, and,,ethod of M~anagomoent by the Metal Fatiue of Reactor-Coolant Prosisuro Boundar Dr^ INSERT TABLE 4.3-4 HERE STra*nRc-;!ient rnt -*,*)Rtd ir the APS fat*gUc cGyle Cpu nt Vrific;ation Arp*e*d*r "NS " 4 Transientt wpatcs counted bnthcycl count proe dure sin ia plant stahfup, therefero RE cyres't w '-dd ass ed. The "Cemposite Werst Case Unit Accumulation" is the sam e as the !198E O005" procedure count. 44 nie n.o....... u,.-t e-A-orded in the 73ST -9RCQ2 procedure ae m.arked as .NR." 42 T-ransients 5Fan 6-- wore ne~fAt counted separately in the cycle coun~t precedure; onRly 10'% poWer inrGeases were Fecor~dod in the precedu'_re. Duo te an incemplete transient description, th proce~dure only included pewer changes between 901% and 1005% pewer. ~ Tansentwas not separately counted in the cycle ceunt precGedure, therefere the APS reonticlddall occrrencesmrAF from 108R5-2005. The "Accumul, Ation Ratea" mwas c,;t;alulatd by taking the APS roceunt numbe~hPr and-diVfiding by the leaRst number Of years in eperatien Up to 19952005 (Unit 3 operating period of 18 years) to dete~rm~ine the woerst c-ase number of event experienced per year. The "Composite Wo.1rst C-ase U nit cumlto"was calculated by mul1tiplying the calculated "Acc;umulation Rate" by I0 years (1095 to 2005) and adding the resul to the APS recount. 4TMransient event does no~t Gontribute significantly to fatigue and is not coun~ted by the Fatigue Managemon9t Program APS reco-unt. The "Composite Worst Case Unimt Accumulation" includes the 25% accumuA'_lationR _ass1.umed in 1905. 4" The "Comporsito Worst Case Un~it cumlto"fo-r TransienPt 16, "Un~bolting/Bolting of RC Pump Casing Studs," is; a censeRmative estimate for a worFst case stud, extracted by reviewo maintenance work orders, for the APS fatigue cycle count ve(rificationR. Palo Verde Nuclear Generating Station License Renewal Application - Revision C Page 4.3-30

Section 4 TIME-LIMITED AGING ANALYSES ,.vThe APS fatiguo cyclc count v eifiation rosulted On highcr than,Xpcctod pr*oGctcd values for Transients 16, 17, 24, 26, 36, 40, and50. 60. and 61 These tr*asient Will Fe*ui*c ,e evaluation or o-ther correcotivo cin when action limits are reachoed. 4'TrFansient 18, 'Saft Injection C~heck Valve Test" is not counted Gpecifically becauseth check valve test is Pe+FMod during a stage of startup at normal heatup prssuro and teperatngre reSUlting in no significant fatigue accuhulation. ~ Transintosnt expected to) occur; thereforFe no "Accum~ulation Rae v -alu!clulated for thi transient. However, at least onRe occGurrence was assumned to cu during the perio~d Of Pext~ended Transiet has no) to date accumulation through 2005. The "Accumnulation Rate" was determined by dividing the design basis number Of transife~nt events; by 40 years and multiplying the result by the percentage of years left in the design basis (22140). U0-F=S.A.R nu~mb-ers; of 5 evenpts from 10%power; 10 evenpts; from _an unspecified power level. 4TrFansient 39, "Seismnic Event up to and including One Half of the SaeShutddown Earthquake, a t 100 Poem-r" is A.Ot coGunRte d speci 0fically b ecGau se it is inc lu11ded in th e co)UnRt for-t r_ Ansien t _27 "Operating Basis Ea~thguake--" 21 Transi.ent_12 "Loss of Feedwater Flew (to S!G)" is net counted spBecifically by theFaie Manaqenme. t Pro"am seoftwae because it is included in;the su..m. the counts for tFansients 47, 48, and 49. Transe*nt 53, "lnadvFt+nt CIosure of all MFIVI at 10004 PR*oer" is Rnot c-unted specifi*all, because it is a duplic.ate of transient19, ,"MFV Clsu d Loss, of Air at 100-% Powe 24-Tr.ansienrts 60 and 61, Ol and-HPI-Pump Tests" are not listed -as 'i*nPn..;g aRd Desig4n Basis Transients. These are qua~terly tests that add significant fatigue to the pumps and4 compon(ents upstream of the isolation valves. 4 Tras~ien.t 32,2 "Exte-nded Loss of Letdown" was added to the 73ST 9RC;02 procedure in1908. At that point, 200 cYcles we.re asumed.for. Uit 3 only (25% of desig), and 0 cycles for Units 1 and 2. The actual data reoddfro-m 10951985 2005 include 614 cycles Of this transiet for Unit 1, 0:1 cycles for Unit 2, and 2 cyclPes; for. Unit 3. The 73ST 9RC02 Wor)st Case (1985 2005) column ; ignErEo the 200 assumed for UInit 3. The Worst Case Composite Unit Accumulatien fr Transient 32, "Fxtende1d.* Ls ,,,Of Le,,tdowVn," is the 61 counted cycles from the Unit 1 xO1,;,;.n data. Transient 12, "Pressur~fizer Cooldown4." Palo Verde Nuclear Generating Station License Renewal Application - Revision C Page 4.3-31}}