ML102360335

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Response to July 21, 2010, Request for Additional Information Regarding Metal Fatigue for the Review of the License Renewal Application, and License Renewal Application Amendment No. 22
ML102360335
Person / Time
Site: Palo Verde  Arizona Public Service icon.png
Issue date: 08/12/2010
From: Hesser J
Arizona Public Service Co
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
102-06234-JHH/GAM
Download: ML102360335 (41)
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Text

L A M A subsidiary of Pinnacle West CapitalCorporation John H. Hesser Mail Station 7605 Palo Verde Nuclear Vice President Tel: 623-393-5553 PO Box 52034 Generating Station Nuclear Engineering Fax: 623-393-6077 Phoenix, Arizona 85072-2034 102-06234-JHH/GAM August 12, 2010 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001

Dear Sirs:

Subject:

Palo Verde Nuclear Generating Station (PVNGS)

Units 1, 2, and 3 Docket Nos. STN 50-528, 50-529 and 50-530 Response to July 21, 2010, Request for Additional Information Regarding Metal Fatigue for the Review of the PVNGS License Renewal Application, and License Renewal Application Amendment No. 22 By letter dated July 21, 2010, the Nuclear Regulatory Commission (NRC) issued a request for additional information (RAI) regarding metal fatigue for the review of the PVNGS license renewal application. Enclosure 1 contains Arizona Public Service Company's (APS's) response to the RAI. contains PVNGS LRA Amendment No. 22 to reflect changes made as a result of the RAI response.

APS makes no commitments in this letter. Should you need further information regarding this submittal, please contact Russell A. Stroud, Licensing Section Leader, at (623) 393-5111.

A member of the STARS (Strategic Teaming and Resource Sharing) Alliance Callaway

  • Comanche Peak
  • Diablo Canyon
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  • South Texas - Wolf Creek bjL~z

ATTN: Document Control Desk U.S, Nuclear Regulatory Commission Response to July 21, 2010, Request for Additional Information Regarding Metal Fatigue for the Review of the PVNGS License Renewal Application, and License Renewal Application Amendment No. 22 Page 2 I declare under penalty of perjury that the foregoing is true and correct.

Executed on /IgZ/ Io (date)

JHH/RAS/GAM

Enclosures:

1. Response to July 21, 2010, Request for Additional Information Regarding Metal Fatigue for the Review of the PVNGS License Renewal Application
2. Palo Verde Nuclear Generating Station License Renewal Application Amendment No. 22 cc: E. E. Collins Jr. NRC Region IV Regional Administrator J. R. Hall NRC NRR Senior Project Manager L. K. Gibson NRC NRR Project Manager J. H. Bashore NRC Senior Resident Inspector (acting) for PVNGS L. M. Regner NRC License Renewal Project Manager G. A. Pick NRC Region IV (electronic)

ENCLOSURE 1 Response to July 21, 2010, Request for Additional Information Regarding Metal Fatigue for the Review of the PVNGS License Renewal Application RAI 4.3-7 RAI 4.3-8 RAI 4.3-9 RAI 4.3-10 RAI 4.3-11 RAI 4.3-12 RAI 4.3-13 RAI 4.3-14 RAI 4.3-15 RAI 4.3-16 RAI 4.3-17 RAI 4.3-18

Enclosure 1 Response to July 21, 2010, Request for Additional Information Regarding Metal Fatigue for the Review of the PVNGS License Renewal Application NRC RAI 4.3-7 Backgqround:

In license renewal application (LRA) Amendment 14, the applicant amended LRA Table 4.3-4. In the table, the applicant credits the following enhanced fatigue aging management program (AMP) monitoring bases for American Society of Mechanical Engineers (ASME) Code Class 1 components:

" Stressed based fatigue (SBF) monitoring as the 10 CFR 54.21 (c)(1)(iii) aging management monitoring basis for the pressurizer surge line elbow, which is the limiting environmentally-assisted fatigue location (i.e., limiting NUREG/CR-6260 location).

" Cycle based fatigue-partial cycle (CBF-PC) monitoring as the 10 CFR 54.21 (c)(1)(iii) aging management monitoring basis for the pressurizer spray nozzles, which are the limiting non-environmental CUF components for the current fatigue aging management program (with a limiting design basis cumulative usage factor (CUF) value of 0.9923).

Issue:

The staff noted that under the amended basis in LRA Table 4.3-4, as given in LRA Amendment 14, the applicant currently credits SBF monitoring only for 10 CFR 54.21 (c)(1)(iii) management of the pressurizer surge line elbow, which according to the LRA is the limiting ASME Code Class 1 location for environmentally-assisted fatigue. For-the current fatigue AMP, the pressurizer spray nozzles are the limiting ASME Code Class 1 component (with a limiting design basis CUF of 0.9923).

The updated table does not credit SBF for this limiting component.

Request:

Justify your basis for not evaluating-the pressurizer spray nozzles for environmentally assisted fatigue, when considering that the pressurizer spray nozzle has a limiting design basis CUF of 0.9923.

APS Response to RAI 4.3-7 The environmental factor (Fen) for stainless steel is a function of three factors:

temperature (T), dissolved oxygen (DO), and strain rate (edot). While the two locations (pressurizer surge line elbow CUF of 0.937 and pressurizer spray nozzle CUF of 0.9923 as shown in LRA Table 4.3-4) experience similar T and DO histories, the spray nozzle fatigue analysis is dominated by transients with large, sudden temperature shocks, which gives rise to a high effective edot and a lower Fen (typically < 5-8). By contrast, the surge line elbow experiences a mix of rapid (insurge/outsurge) and slow (heat-1

Enclosure 1 Response to July 21, 2010, Request for Additional Information Regarding Metal Fatigue for the Review of the PVNGS License Renewal Application up/cool-down, stratification) transients, giving a mix of low and high edots and thus a higher Fen (generally > 10-13).

Assuming the original CUFs of 0.9923 for the pressurizer spray nozzle and 0.937 for the surge line elbow (as shown in LRA Table 4.3-4) as a fair comparison, the disparity in Fen reverses the direction of bounding, as shown below:

(0.9923

  • 8) = 7.94 < 9.37 = (0.937
  • 10)

There are additional differences which validate the conclusion that the surge line elbow location is an adequate sentinel location for monitoring.

1. The spray nozzle transients are defined in the analysis of record very conservatively, both in severity (as large, bounding step changes) and in frequency (many more cycles assumed than have been experienced to date). By contrast, the surge elbow transients were constructed to reflect a delta temperature and frequency spectrum derived from actual plant operating data.
2. The surge line elbow analysis includes effects from stratification mechanisms (which are known to occur frequently at that location), while it is known that the spray nozzle does not experience stratification effects.

Therefore, the surge line elbow is the proper choice for the monitored (sentinel) location.

NRC RAI 4.3-8 Backgqround:

In LRA Amendment 14, dated April 28, 2010, the applicant updated LRA Tables 4.3-2 and 4.3-3. In the updated LRA Table 4.3-2, the applicant lists Transient 17, "Initiation of Auxiliary Spray," as an applicable normal operating condition transient. In the updated LRA Table 4.3-3, the applicant stated that the tracking of Transient 17 will be correlated to the tracking of pressurizer cooldown events, which is listed in these updated tables as Transient 12, "Pressurizer cooldown from 563 degrees F to 70 OF at a rate of 200°F/hr."

Issue:

It is not clear to the staff whether Transient 17 is referring to an initiation of the pressurizer spray system or an initiation of the containment spray system. It is also not clear to the staff why it is valid to correlate the tracking of Transient 17 to the tracking of Transient 12.

2

Enclosure 1 Response to July 21, 2010, Request for Additional Information Regarding Metal Fatigue for the Review of the PVNGS License Renewal Application Request:

Clarify whether Transient 17 is referring to an initiation of the pressurizer spray system or an initiation of the containment spray system. Provide your basis for why it is valid to correlate the tracking of Transient 17 to the tracking of Transient 12.

APS Response to RAI 4.3-8 Transient 17 in LRA Table 4.3-2 and 4.3-3, "Initiation of auxiliary spray during cooldown," refers to the initiation of auxiliary pressurizer spray during pressurizer cooldown, and not containment spray.

Normal pressurizer spray is provided by reactor coolant pump differential pressure.

During plant cooldown and depressurization the reactor coolant pumps must be secured prior to full depressurization in order to prevent pump cavitation. When this occurs, main pressurizer spray becomes unavailable, and auxiliary pressurizer spray is used to complete pressurizer cooldown. Therefore, there is a correlation between the number of pressurizer cooldowns (Transient 12) and the initiation of auxiliary spray during cooldown (Transient 17). This is consistent with the transient limits for pressurizer cooldown events and auxiliary spray actuations for cooldown both being equal (500 events).

NRC RAI 4.3-9 Backgqround:

LRA Amendment 14, dated April 28, 2010, the applicant updated LRA Table 4.3-3. In this table, the applicant provides its updated counting and 60-year projections for Transient 25, "Standby to SI hot leg injection check valve stroke test to standby (using the HPSI pump)." The applicant stated that the transient is conducted during refueling outages, and that the transient is not currently being counted because it was recently identified and added to the updated final safety analysis report Table 3.9-1. The applicant also. stated that the transient will be counted when it is added to the scope of the transient cycle counting procedure.

Issue:

The applicant identified 16 occurrences of this transient for Units 1 and 3, and 17 occurrences for Unit 2, inclusive of December 31, 2005. The staff noted this transient is projected to occur 57 times through the end of the period of extended operation. The staff has noted a disconnect in the recording of occurrences for this transient going forward from January 1, 2006, and the time when the transient will be accounted in a future revision of the transient cycle counting procedure.

3

Enclosure 1 Response to July 21, 2010, Request for Additional Information Regarding Metal Fatigue for the Review of the PVNGS License Renewal Application Request:

Clarify if the transient cycle counting procedure has been updated to include Transient 25 and if not, when the procedure will be updated. Explain how all occurrences of Transient 25 are considered.

APS Response to RAI 4.3-9 Amendment No. 14 to the PVNGS license renewal application submitted in letter no.

102-06175, dated April 28, 2010, included the following Commitment No. 55 in LRA Table A4-1:

"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 leg injection check valve stroke test to standby (using the HPSI pump),' will be added to the cycle counting surveillance procedure 73ST-9RC02 by August 25, 2010."

This procedure change is on schedule to complete on or before August 25, 2010.

LRA Table 4.3-3 was revised in Amendment No. 14 (APS letter no. 102-06175, dated April 28, 2010) to include the total number of occurrences for the period 1985 through 2005 based on plant refueling history. The total number of actual occurrences subsequent to 2005 will be counted.

NRC RAI 4.3-10

Background:

In LRA Amendment 14, dated April 28, 2010, the applicant updated LRA Table 4.3-3. In this table, the applicant provides its updated counting and 60-year projections for Transient 79, "Reactor coolant system leak test."

Issue:

For Transient 79, "Reactor coolant system leak test," the applicant stated that its recent recount indicated that the transient occurred 5 times for Unit 1, 4 times for Unit 2, and 2 times for Unit 3 through the end of December 2005. It is not clear whether this transient represents the system leak test for the reactor coolant pressure boundary, mandated by ASME Code Section Xl, Table IWB-2500-1, Examination Category B-P and 10 CFR 50.55a. The staff noted that this requires the applicant to pressurize its reactor coolant pressure boundary once every refueling outage to the normal operating pressure for the system and to perform a visual VT-2 examination of the system's components for evidence of reactor coolant leakage. The staff has noted that PVNGS has been operating for about 22 to 24 years of licensed operation. Thus, based on the time from 4

Enclosure 1 Response to July 21, 2010, Request for Additional Information Regarding Metal Fatigue for the Review of the PVNGS License Renewal Application initial operation, the staff estimates that the reactor coolant system leak test would have occurred approximately 14 to 16 times since initial operation of the units.

Request:

Clarify whether Transient 79 is different than the system leak test that is required by ASME Code Section XI. If the Transient 79 and the ASME Code Section XI system leak are different, clarify how the ASME Code Section XI system leak test is accounted for. If these two are not different, justify the occurrences of Transient 79, as described above, considering that it is required to perform this system leak test on a frequency of once every refueling outage.

APS Response to RAI 4.3-10 Transient 79, "Reactor Coolant System Leak Test," test condition is listed in UFSAR Table 3.9.1-1 and does represent the ASME Section XI test. However, when the test is performed as part of a normal plant start-up, the fatigue effects are accounted for by counting the normal plant heat-up transient. This is a valid approach to monitoring the fatigue effects on the components since they are separate transients in the fatigue calculations.

During the review of design documents, it was noted that the pressure and temperature ranges provided in the Combustion Engineering General Specification and utilized as input to the ASME Code analyses are inconsistent with the description of the reactor coolant system leak test condition in UFSAR Table 3.9.1-1. Palo Verde Action Request (PVAR) No. 3514767 has been initiated to reconcile this difference and to update the UFSAR as necessary. The values presented below reflect the transient definitions used in the analyses of record.

Transient 79 is applicable to the Combustion Engineering Nuclear Steam Supply System (NSSS) scope of supply for the reactor coolant system piping and components.

The associated ASME Code Analyses determine the cumulative fatigue resulting from the specified transients in UFSAR Table 3;9.1-1, including 500 heatup and cooldown cycles in which the pressure ranges from 15 psia and 800 F to 2250 psia and 5650 F and back to 15 psia and 800 F. Different from these 500 heatup and cooldown cycles, the analyses consider 200 additional transients in which the pressure cycles from 400 psia and 1600 F to 2250 psia and 4000 F and back to 400 psia and 1600 F. The fatigue effects for these two transients are determined as separate events and are thus additive to each other in the analyses.

In actual operating practice, the ASME Section XI leak test is performed at 2250 psia normal operating pressure and 5650 F normal operating temperature in Mode 3 hot standby as part of the normal plant heat up. The reactor coolant system pressure and temperature is not reduced as part of or following the leak test. As such, no actual fatigue effects occur as a result of the test since the reactor coolant system pressure and temperature are not cycled. The fatigue effects are due solely to the plant heatup.

5

Enclosure 1 Response to July 21, 2010, Request for Additional Information Regarding Metal Fatigue for the Review of the PVNGS License Renewal Application Accordingly, the fatigue monitoring program would record the plant evolution solely as a plant heatup. Even if the reactor coolant system pressure and temperature were to be reduced as a result of the test so as to affect plant repairs, the evolution would be recorded as a heatup and cooldown cycle since the transient profile would be better represented by the heatup and cooldown profile. If the evolution were to be recorded as both a plant heatup and cooldown and a reactor coolant system leak test, the fatigue monitoring program would have accounted for the fatigue effects of two separate occurrences cycling to 2250 psia and normal operating temperature and back to cold conditions, when in fact the second transient cycle and the associated fatigue effects did not actually occur. Therefore, for the current license and during the period of extended operation, fatigue effects are appropriately accounted for by recording an ASME Section X1 leak test performed concurrently with a plant heatup solely as a plant heatup and not as a separate leak test transient. The Transient 79 events counted to-date are reactor coolant system leak tests where the units were cycled to 2250 psia and back.

NRC RAI 4.3-11

Background:

On April 28 and May 27, 2010, the applicant submitted LRA Amendments 14 and 16, respectively. The amendments include an updated LRA Section 4.3.1.5, "Cycle Count Action Limits and Corrective Actions Subsection," which states the following:

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 the enhanced Metal Fatigue of Reactor Coolant Pressure Boundary Program (B3.1) before the end of the next fuel cycle.

Issue:

The staff noted that, according to the second sentence of the quoted material, the applicant will require cycle counting corrective actions only for those design basis transients which the applicant considers to be significant contributors to fatigue usage.

Request:

Clarify the definition of the term "significant contributors to usage factor" and how this is associated with the corrective action limits in the enhanced Metal Fatigue of Reactor Coolant Pressure Boundary Program.

6

Enclosure 1 Response to July 21, 2010, Request for Additional Information Regarding Metal Fatigue for the Review of the PVNGS License Renewal Application APS Response to RAI 4.3-11 The significant contributors to usage factor for the license renewal application are all of the transients listed in UFSAR Tables 3.9.1-1 and 3.9-1, as shown in LRA Table 4.3-2.

Each transient listed in LRA Table 4.3-2 is considered a significant contributor to fatigue, and each will have one or more appropriate corrective action limits associated with it (e.g., plant heat-up [Transient 1] has two limits based on what component is being considered). These cycle counting corrective action limits will reflect the UFSAR transient limits and assumptions made in the analyses of record. This includes the transients that are tracked by counting as well as those that are tracked by accounting for the occurrences such that components are maintained within the design limits.

NRC RAI 4.3-12 Backqround:

On LRA page 4.3-40 and 4.3-41, of Amendment 14, the applicant discusses its action limit for the CUF monitoring techniques and provides 7 corrective actions that may be used when a cumulative usage factor (CUF) action limit is reached.

Issue:

Corrective Action (1) on LRA page 4.3-40 states "Determine whether the scope of the enhanced fatigue management program must be enlarged to include additional effected reactor coolant pressure boundary locations." In regard to this corrective action, the staff noted that the applicant indicates that the corrective action is only applicable to reactor coolant pressure boundary components. However, in its review of LRA Section 4.3.2, the staff confirmed that the time limited aging analysis (TLAA) includes the CUF results for some ASME Code Class 2 components that were analyzed to ASME Section III CUF requirements for Code Class 1 components. As a result, the staff noted that the action in Corrective Action (1) may also be applicable to those ASME Code Class 2 components that were analyzed to ASME Section III CUF requirements for Code Class 1 components.

Request:

Clarify if the scope of Correction Action (1) on CUF monitoring includes all components with ASME Section III CUF calculations for Code Class 1 components and ASME Code Class 2 components that were analyzed to ASME Section III CUF requirements for Code Class 1 components. If the scope of Correction Action (1) does not include all components, justify why they are not within the scope.

7

Enclosure 1 Response to July 21, 2010, Request for Additional Information Regarding Metal Fatigue for the Review of the PVNGS License Renewal Application APS Response to RAI 4.3-12 The scope of the enhanced Metal Fatigue of Reactor Coolant Pressure Boundary program will include the ASME Section III Class 1 components and components with Class 1 fatigue analysis, which includes ASME Code Class 2 components that were analyzed to ASME Section III CUF requirements for Code Class 1 components. LRA Table 4.3-4 identifies the methods that will be used to monitor components within the scope of the enhanced Metal Fatigue of Reactor Coolant Pressure Boundary program.

Those locations monitored by CUF monitoring (CBF, CBF-PC, CBF-EP and SBF) will be subject to CUF based corrective actions. Locations monitored by cycle counting (CC) will be subject to cycle based corrective actions discussed under the heading "Cycle Count Action Limits and Corrective Actions" in LRA Section 4.3.1.5.

NRC RAI 4.3-13 Backgqround:

The LRA includes Table 1 aging management review (AMR) items on management of cumulative fatigue damage in mechanical systems in LRA Tables 3.1.1 (reactor coolant system), 3.2.1 (engineered safety feature systems), 3.3.1 (auxiliary systems), and 3.4.1 (steam and power conversion systems). The applicant provided its further evaluation discussions on how the applicant would manage cumulative fatigue damage in components addressed in the applicable AMR items in LRA Sections 3.1.2.2.1, 3.2.2.2.1, 3.3.2.2:1, and 3.4.2.2.1.

LRA Sections 3.1.2.2.1 and 4.3.2.5 identify that the CUFs analyses for the recirculating steam generator tubes do not need to be identified as a TLAA because the analyses are not being credited to manage either cumulative fatigue damage or cracking that could be induced in the tubes by a mechanism of fatigue. In LRA Amendment 16, the applicant makes the following statement to support its conclusion that the CUF calculations for the recirculating steam generator tubes do not need to be identified as TLAAs for the LRA:

The design of the PVNGS steam generators includes a code fatigue analysis of the steam generator tubes, as indicated in Table 4.3-8. This analysis would be a TLAA if the safety determination depended upon it. However the design report indicates a zero fatigue usage factor, and a code fatigue analysis has historically not proved sufficient to support the safety determination for steam generator tubes, which depends on a separate tube inspection program.

The various tube degradation mechanisms not anticipated in the original design have required stringent periodic inspection programs in order to ensure adequate steam generator tube integrity. The steam generator tubes are, in effect, (1) no longer qualified for a licensed design life (10 CFR 54.3(a) Criterion 3), and the (2) the fatigue analysis is therefore no longer the basis of the safety determination; in 8

Enclosure 1 Response to July 21, 2010, Request for Additional Information Regarding Metal Fatigue for the Review of the PVNGS License Renewal Application this case that the tubes will maintain their pressure boundary function between primary and secondary systems (Criterion 5).

Issue:

The staff noted that a CUF calculation of the replacement recirculating steam generator tubes was performed because the tubes are considered ASME Code Class 1 components that are designed to ASME Section II1.The staff noted that the various degradation mechanisms discussed in the second paragraph of the quoted paragraph appear to make reference to steam generator tube cracking induced either by stress corrosion cracking (SCC) or any other mechanisms. Cracking induced by these mechanisms has no relationship to cracking induced by high cycle or low cycle fatigue mechanisms. The staff noted that cracking of steam generator tubes has been induced either by intergranular SCC, primary water SCC, outside diameter SCC or intergranular attack mechanisms, and that the in-service inspections (ISI) of the tubes required by plant technical specifications have largely been implemented to detect cracking induced by these mechanisms. The staff also noted that these mechanisms do not have a relationship to the use of CUF calculations to qualify the tubes for cracking by a fatigue mechanism and do not constitute a valid basis for concluding the CUF values do not qualify the tubes for fatigue-induced cracking during their design life. It is not clear to the staff the basis for the CUF value of the recirculating steam generator tubes be equated to a value of zero.

In LRA Section 3.1.2.2.1, the applicant states that the pressurizer support skirts'and attachment welds had been designed to ASME Section III requirements and had received an applicable ASME Section III CUF analysis. The staff determined that neither LRA Table 3.1.2-2 nor LRA Table 3.1.2-3, include any applicable line items on management of cumulative fatigue damage in the pressurizer support skirts and attachment welds, as aligned to any of the AMRs on cumulative fatigue damage.

The staff noted that the Summary Description in LRA Section 4.3.5 states that implicit fatigue analyses discussed in the section are applicable to all ASME Code Class 2 and 3 and ANSI B31.1 piping, piping components, and piping elements. The staff noted that it is not clear whether the LRA includes all corresponding AMR items for applicable ASME Code Class 2 and 3 or ANSI B31.1 piping, piping components, and piping elements scoped in for license renewal. The staff also noted that this includes those components in the Engineered Safety Features Systems (LRA Section 3.2), Auxiliary Systems (LRA Section 3.3) and the Steam and Power Conversion Systems (LRA Section 3.4).

Request-

1) Justify your basis for concluding that the CUF calculation for the replacement recirculating steam generator tubes do not need to be identified as a TLAA, when considering the use of SCC mechanisms and TS examinations does not appear 9

Enclosure 1 Response to July 21, 2010, Request for Additional Information Regarding Metal Fatigue for the Review of the PVNGS License Renewal Application to be a valid basis for concluding that the CUF calculations would not qualify the tubes for metal fatigue during the remainder of the licensed life of the tubes.

2) Justify your basis for omitting applicable AMR items on cumulative fatigue damage of the pressurizer support skirts and pressurizer attachment weld components in either LRA Table 3.1.2-2 or LRA Table 3.1.2-3.
3) Clarify if the LRA includes all applicable AMR items with an aging effect of cumulative fatigue damage for those components scoped into license renewal. If not, justify why the' LRA does not include all corresponding AMR items on cumulative fatigue damage for applicable ASME Code Class 2 and 3 or ANSI B31.1 piping, piping components, and piping elements scoped in for license renewal. Identify all component types that are within the scope of the implicit fatigue analyses for ASME Code Class 2 and 3 components and B31.1 components in LRA Section 4.3.5 and hence should be within the scope of applicable component-specific AMR items on cumulative fatigue damage.

APS Response to RAI 4.3-13 Response 1:

The steam generator (SG) tube CUF value was taken from the Ansaldo Camozzi Design Reports PX-RPD-00-AESCO01 (Units 1 and 3) and PV-RPD-00-000003 (Unit 2).

The statement that the SG tube CUF is zero was included in the design reports. It is based on the cyclic stress range being below the endurance limit.

In order to resolve the discussion regarding the TLAA status of the SG tube fatigue analysis, APS has revised the LRA in Amendment No. 22 in Enclosure 2 to identify the SG tube fatigue analysis as a TLAA. Since the design reports note a CUF of zero for the replacement SG tubes, the disposition for the SG tubes is validation as permitted by 10 CFR 54.21 (c)(1)(i). As shown in LRA Amendment No. 22 in Enclosure 2, cumulative fatigue damage of the steam generator tubes has been added to LRA Tables 3.1.1 and 3.1.2-4 as a TLAA requiring evaluation for the period of extended operation consistent with GALL line IV.D1-21.

Response 2:

Further Evaluation, LRA Section 3.1.2.2.1, currently notes that LRA Section 4.3.2.4 describes the evaluation of the TLAA for pressurizer support skirt and attachment weld.

As shown in LRA Amendment No. 22 in Enclosure 2, cumulative fatigue damage of the pressurizer support skirt and attachment weld has been added to LRA Table 3.1.2-3 as a TLAA requiring evaluation for the period of extended operation consistent with GALL line IV.C2-10.

10

Enclosure 1 Response to July 21, 2010, Request for Additional Information Regarding Metal Fatigue for the Review of the PVNGS License Renewal Application Response 3:

Further Evaluation, LRA Sections 3.2.2.2.1 (Engineered Safety Features Systems),

3.3.2.2.1 (Auxiliary Systems), and 3.4.2.2.1 (Steam and Power Conversion Systems) currently note that LRA Section 4.3.5 describes the evaluation of the TLAA for piping and piping components designed to ASME Class 2, Class 3, and ANSI B31.1 requirements. LRA Section 4.3.5 describes the systematic survey of all plant piping systems to identify components that might be subject to thermal fatigue effects. As shown in LRA Amendment No. 22 in Enclosure 2, cumulative fatigue damage of the ASME Class 2, Class 3, and ANSI B31.1 piping and piping components has been added to the LRA tables noted below as a TLAA requiring evaluation for the period of extended operation consistent with GALL lines as noted below.

Engineered Safety Features Systems Table 3.2.2-4 Safety Injection and Shutdown Cooling (Note 1)

Auxiliary Systems Table 3.3.2-8 Nuclear Sampling (VII.E-16)

Table 3.3.2-10 Chemical and Volume Control (Note 1)

Table 3.3.2-21 Diesel Generator (VII.El-8)

Table 3.3.2-30 Miscellaneous Auxiliary Systems: only applicable to Auxiliary Steam and Secondary Chemical Control (VIII B1-10)

Steam and Power Conversion Systems Table 3.4.2-1 Main Steam (Note 1)

Table 3.4.2-3 Auxiliary Feedwater (Note 1)

Note 1 - No additions are required, this LRA table includes an AMR item(s) with an aging effect of cumulative fatigue damage.

NRC RAI 4.3-14 Backqround:

LRA Section 4.3.2.1 provides a CUF value of 0.823 for the reactor pressure vessel (RPV) studs and a CUF value of 0.954 for the RPV lugs. LRA Section 4.3.2.1 also states that the RPV studs are the more limiting component because they will experience more severe stresses during each transient event, even though they are limited to a lower design limit on the number of allowable heatup and cooldown events.

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Enclosure I Response to July 21, 2010, Request for Additional Information Regarding Metal Fatigue for the Review of the PVNGS License Renewal Application Issue:

In updated LRA Table 4.3-4, in LRA Amendment 16, the applicant states that both of these component locations will be monitored using only cycle-based monitoring methods. The applicant's cycle-based monitoring methods do not include automatic periodic updates of CUF calculations. It is not clear to the staff if only cycle-based counting of the RPV studs will be performed, even though the RPV lugs have an existing CUF of 0.954.

Request:

Clarify whether or not the RPV stud/RPV lug limiting component discussion in LRA Section 4.3.2.1 is being made to clarify that cycle-based monitoring will be performed on the RPV studs. The staff requests the following additional actions if the discussion is being made to justify that cycle-based monitoring will only be performed in the RPV studs: (1) summarize the transients that were used for the CUF calculations for the RPV studs and RPV bottom head lugs, and for each of the transients analyzed, clarify the quantitative contribution to fatigue usage.

APS Response to RAI 4.3-14 Both the RPV studs and RPV external bottom head support lugs will be monitored by cycle counting and appropriate corrective action limits will be applied to both components.

NRC RAI 4.3-15

Background:

In LRA Amendment 16, the applicant updated LRA Section 4.3.2.8, Absence of Supplemental Fatigue Analysis TLAAs in Response to Bulletin 88-08 for Intermittent Thermal Cycles due to Thermal-Cycle Interface Valve Leaks and Similar Cyclic Phenomena.

Issue:

The U.S. Nuclear Regulatory Commission (NRC or the staff) Bulletin 88-08 referenced in LRA Section 4.3.2.8 recommended that a high cycle fatigue analysis be performed for the auxiliary pressurizer spray systems. LRA Section 4.3.2.8 states that Arizona Public Service (APS) performed a "supplemental bounding thermal gradient stress analysis to determine the effect of low cycle fatigue," and that the analysis did not evaluate the effects of high cycle fatigue on these lines, as recommended in Bulletin 88-08. The staff confirmed that the APS response to NRC Bulletin 88-08, dated October 3, 1988, did not commit to the performance of a high cycle fatigue analysis. It is not clear to the staff if the low-cycle fatigue analysis that was performed included any cycle based fatigue flaw 12

Enclosure I Response to July 21, 2010, Request for Additional Information Regarding Metal Fatigue for the Review of the PVNGS License Renewal Application growth or cycle based fracture mechanics analysis and thus, should be identified as a TLAA for the LRA. LRA Section 4.3.2.7, Subsection "Flow Stratification Thermal Gradient in the Auxiliary Spray Line and Tee" states that "the analysis of the thermal gradient demonstrated that the cumulative fatigue usage factor, including the effects of this thermal gradient, meets ASME Section III Subsection NB-3600 for a 40-year plant life." Based on this statement in LRA Section 4.3.2.7 it appears that this analysis meets the definition of a TLAA in accordance with 10 CFR 54.3(a).

Request:

Identify the low cycle fatigue analysis that is being referred to in LRA Sections 4.3.2.7 and 4.3.2.8 and clarify whether the low-cycle fatigue analysis on the auxiliary pressurizer spray systems included an applicable, implicit fatigue analysis, cycle-based fatigue flaw growth or cycle-based fracture mechanics analysis. Justify why the low-cycle fatigue analysis would not need to be identified as a TLAA if it is determined that analysis does include a cycle dependent analysis.

APS Response to RAI 4.3-15 The low cycle fatigue analysis that is being referred to in LRA Sections 4.3.2.7 and 4.3.2.8 is documented in PVNGS Calculation 13-MC-ZZ-0643, including Engineering Document Change (EDC) 2009-00582. This analysis considered deadweight, thermal (including thermal stratification), seismic and LOCA load cases to determine their effects on the existing fatigue cycle stress ranges. It was determined that all support loads remained the same. All stresses decreased or remained the same. The nozzle load at the pressurizer was reviewed and had a minimal increase in moment, and was therefore determined to be acceptable. The calculation did not contain implicit fatigue analysis, cycle-based fatigue flaw growth or cycle-based fracture mechanics analysis and is therefore not a TLAA. The analysis concluded that although it is important that the thermal stratification be documented, its effects do not negatively impact the pressurizer auxiliary spray system or the stress ranges of the fatigue analysis of record, and therefore the original analysis of record remains valid.

The-low cycle fatigue analysis referred to in LRA Sections 4.3.2.7 and 4.3.2.8 that was performed in response to NRC Bulletin 88-08 does not include cycle based assumptions and is, therefore, not a TLAA.

NRC RAI 4.3-16 BackQround:

In LRA Amendment 16, the applicant amended LRA Section 4.3.2.10, "Class 1 Fatigue Analyses for Regenerative and Letdown Heat Exchangers."

13

Enclosure 1 Response to July 21, 2010, Request for Additional Information Regarding Metal Fatigue for the Review of the PVNGS License Renewal Application Issue:

LRA Section 4.3.2.10 states that the regenerative and letdown heat exchanger fatigue analyses were performed with transients specified in the Combustion Engineering (CE) general specification for System 80 plants. It further states that the original assessment that fatigue in the regenerative and letdown heat exchangers was bounded by the fatigue of the charging nozzle is still valid. However, LRA Section 4.3.2.10 does not identify the current design basis CUF values for the regenerative heat exchangers and letdown heat exchangers or which transients were evaluated in the System 80 CUF calculations for these heat exchangers.

Request:

Provide the current design basis CUF values for the regenerative heat exchangers and letdown heat exchangers and identify the transients that were evaluated in the CUF calculations of these heat exchangers, and the design basis limits for the transients analyzed in these calculations.

APS Response to RAI 4.3-16 Regenerative Heat Exchangers:

The components analyzed and their CUF are shown below:

Component CUF Tubeside Inlet Region 0.482 (TubeSheet/Channel/Shell) Junction)

Shell Side Outlet Nozzle (Nozzle Reinforcement /Shell Junction)

Letdown Heat Exchangers:

The components analyzed and their CUF are shown below:

Component CUF Tube Sheet 0.631 Flat Head 0.132 Nozzle/Shell Intersection 0.986 Flange/Shell Intersection 0.934 Bolts for Head to Flange 0.039 The transients and design basis limits included in the regenerative and letdown heat exchanger fatigue analyses are stated in UFSAR Table 9.3.4-1 and UFSAR Section 3.9.1.1. The design reports included the following applicable transients:

14

Enclosure 1 Response to July 21, 2010, Request for Additional Information Regarding Metal Fatigue for the Review of the PVNGS License Renewal Application

.Letdown Heat Exchanger LRA Tabl 43-2 RA-Table 4.3-2 ' Number .. I ofI Transients,',

..Assumed in . Design, . . Transient Title Item Number Report (Note 1,)

3060 for the total of Atransients -1,l13,.31Y,36, 39,t

ý'41, 42,:55, 77*7and 79 which!' Reactor Coolant System (excluding pressurizer) have the same usagei< . heatup from 70F to hot standby conditions at a irmpact. Thisexceeds the rate of < 1O0F/hr

12730)sp edified in the UFSAR-<' 'K'

.790.for,-the total of . .

tran,,ieis,34 :and 46 *i Reactor Coolant System (excluding pressurizer) 2 which im~pact.have the same usage cooldown from hot standby conditions to 70F at a This exceeds

6 2 5 ,sp e i -i...i..... the rate of <IOOF/hr 34,000 for the total of 3 transients 3, 4, 5 and 6 5%/minute power ramp increase, from 15% to which. have the same usage 100% power impact 34,000 for the total of transients 3, 4, 5 and 6 5%/minute power ramp decrease, from 100% to which have the same usage 15% power impact 34,000 for the total of transients 3, 4, 5 and 6 10% power step increase, from 90% to 100%

which have the same usage power impact 34,000 for the total of transients 3, 4; 5 and 6 10% power step decrease, from 100% to 90%

which have the same usage power impact 43060 for the total of0' '

transients 1, 13, .31, 361 9'39,

'41 42,.,55'.77 79 which,,,,Shift from normal to maximum purification flow at

,and 13 ha've th.e same. usage1 impact. IThis exceeds the 100% power 627301'spcified in thet UFSAR.~

601 for the total of Startup of SDC system from standby to shutdown 18 transients 18, 27, and 72 cooling (RCS > 200F) to shutdown cooling (RCS which have the same usage < 200F) to standby impact 2 foithe total 6oftransiebts..

S2.48:and 74 wl~*hich:*

h fiave: .

26,I 4 heIsan hPav'; .e Low-low volume control tank/charging pump

26. 't: a. 2 suction diversion to RWT soecif iedin' the UFSAR.

601 for the total of 27 transients 18, 27, and 72 Pressurizer level control, failure to full open which have the same usage impact 15

Enclosure 1 Response to July 21, 2010, Request for Additional Information Regarding Metal Fatigue for the Review of the PVNGS License Renewal Application Letdown Heat Exchanger

  • *;:.TIi~l
  • """i:
  • N' 4;*2

" r'A'"r  : *! umbrn 0fran~iients-sumedin'Design  :. *:!

,A-ter" ui be' s u . ,, Trans ient'Title

  • . *!.,.,.
  • . '*-; *S.* Report,(Note 1),

31 h ave the saeusq Arbitrarý load rejection, from 100% to 15% power Inadvertent control element assembly withdrawal 34 from 0% power 35 Loss of charging and recovery at 100% power 36 Loss of letdown and recovery at 100% power 39 Partial loss of condenser cooling at 100% power Turbine trip without accompanying reactor trip at 41 100% power Inadvertent actuation of one main steam line 42 isolation valve at 100% power 46 Loss of Feedwater Flow (to S/G) 16

Enclosure I Response to July 21, 2010, Request for Additional Information Regarding Metal Fatigue for the Review of the PVNGS License Renewal Application Letdlown Heat Exchanger L Table 4.3-2.. Number of Transients

,"Assumed in Design, Transient-Title Item Numberi;~ ...~~ li*:,.Reort,(Note

~~k R

...';"*;. Not-1 11)* .

2fo the *total oftra'nsients 4;,6sa m sagnd 7

  • a~che* System leak due to rupture of largest instrument 48 thesame uage.impact.

4his8is Ile*s'than s the121 or sampling connection at 100% power

!specified in the UFSAR-

'3060 *for-the*total of. ' .

'transients i 13,31 '36,"39, 41, 42,-55,ý,77,and 79,which 55 Ihv&,the**eame usig` ` ' ` Rx Trips, Turbine Trips and Loss of RCS Flow impaý.c~t ,This exc~e'ds-the

,UFSAR.KhKV 601.for 721for the total of trients ,and72 Major rupture in the main feedwater piping 72 transients which have18, the27, and usage same 72 (system operating application for worstmode conditions)upon design casedependent impact 2 for th~etotal of transients'

  • 26, 48andh74 :whiclchave:!j Major rupture in the main steam piping (system 74 the same usage impacft., operating mode dependent upon design

.Thisiiesstthian th'e 1,2,1 application for worst case conditions) specified in~th&UFSAR!. _____________________

f360'for. the t6tal'of .-,

transierntsi 1,1331 36 *39, 4;1 455,77 42 and'79 which, 77 ýhýave :the ,same uisage'," Reactor Coolant System hydrostatic test impact..This exceeedsthe 2730 specified iný thea '

U~FSAR.

3060 for the total of 9 i41-, 4'2, 55;77'and 79 which 79 have the same usage Reactor Coolant System leak test.

impact. This exceeds the' 2730 specified in'the Note 1: The fatigue analysis grouped the applicable transients into six cases based on fatigue usage. The total number of transients for the group equals the sum of the individual transient allowed totals. For example: Group one consists of transients 3, 4, 5 and 6.

The sum of the respective allowed totals (34,000) is obtained from the UFSAR allowed limits for these transients:

15,000 + 15,000 + 2,000 + 2,000 = 34,000 Deviations from this methodology are noted above by shading. In some cases more transients than' specified in the UFSAR were assumed, and in some cases less were assumed.

17

Enclosure I Response to July 21, 2010, Request for Additional Information Regarding Metal Fatigue for the Review of the PVNGS License Renewal Application Re'generative ýHeat.Exchangier.,

Number of Transients Tumbe4 Assumed in Design Transient Title Number * .*  : Report (Note 2)

Reactor Coolant System (excluding 171* wtihich 'exceeds,.the-500,-ýý:':*. pressurizer) heatup from 70F to hot specified inIthe-ý FSAR standby conditions at a rate of

____ ____ ______< ____ <IOOF/hr which*"e:xc: s.Reactor Coolant System (excluding 2710whichexcd the 500 pressurizer) cooldown from hot 2s;pecified inthe UFSAR standby conditions to 70F at a rate of <100F/hr 5%/minute power ramp increase, 3 15,000 from 15% to 100% power 4 15%/minute 415,000 from 100% power to 15%ramp powerdecrease, 10% power step increase, from 52,000 90% to 100% power 10% power step decrease, from 6 2,000 100% to 90% power Shift from normal to maximum purification flow at 100% power Low-low volume control 26 80 tank/charging pump suction diversion to RWT Inadvertent control element 34 40 assembly withdrawal from 0%

power 351i00 Which is less than the 200 Loss of charging and recovery at 35specified ihnthe UFSAR 100% power 36 950.whi6,e'xceeds the 300: Loss of letdown and recovery at 36______________ pýecified in the UFSAR. 100% power 4401which is less than the,'85 LFo 46  ::speciffid~in thebUFSAR Loss of Feedwater Flow (to S/G)

26,1 which is less than the 480,: Rx Trips, Turbine Trips and Loss of spedified inthel UFSAR RCS Flow Major rupture in the main steam 74 1piping (system operating mode dependent upon design application for worst case conditions)

Note 2: Shading indicates design analysis transient assumptions that differ from those specified in the UFSAR.

During the review of the analyses of record for the regenerative heat exchanger and letdown heat exchanger it was noted that the analyst performed bounding fatigue analyses assuming higher cycles of the significant design transients (e.g. loss of let down), and assuming a lower number of the less significant transients (e.g. loss of charging) than stated in the UFSAR for several transients. The review did not identify any transient limits that are challenged by current operating history. This inconsistency 18

Enclosure 1 Response to July 21, 2010, Request for Additional Information Regarding Metal Fatigue for the Review of the PVNGS License Renewal Application between the transient assumptions in the UFSAR and those in the analysis is being tracked for evaluation and resolution (PVNGS CRAI 3494095).

NRC RAI 4.3-17

Background:

In LRA Amendment 16, the applicant amended Section 4.3.3, "Fatigue and Cycle-Based TLAAs of ASME Ill Subsection NG Reactor Pressure Vessel Internals." LRA Section 4.3.3 identifies that some of the reactor vessel internal (RVI) components were designed to the 1974 Edition of the ASME Code Section III, Subsection NG, or to more recent endorsed versions of the ASME Code Section III. The applicant identifies that the design codes required CUF calculations for these ASME Code Section III NG components. The applicant identifies that these analyses are TLAAs for the LRA.

Issue:

Materials Reliability Program Report MRP-227 identifies that the following CE RVI components are considered to be Code Class 1 components: (1) guide lugs and guide lug inserts and bolts, (2) fuel alignment pins, and (3) RVI components in the upper flange assembly. The assessment in LRA Section 4.3.3 does not identify which of the RVI components were designed to ASME Section NG requirements and were required to have a CUF calculation.

Request:

Identify which RVI components are designed to ASME Section III NG requirements, and of these, which RVI components were required to have a CUF design calculation. For those RVI components that were required to have been analyzed with a CUF calculations, identify what the design basis CUF is for the given RVI component, and identify the transients that were analyzed along with their design basis limits on cycle occurrences. Justify the use of cycle-based monitoring if the existing design basis CUF value for any RVI component is high, for example in excess of 0.9.

APS Response to RAI 4.3-17 As described in LRA Section 4.3.3, the PVNGS reactor vessel internals were designed and fabricated to the 1974 Edition of the ASME Code Section III, Subsection NG. As described in UFSAR Section 3.9.5, the reactor pressure vessel internals consist of the core support barrel assembly including the core support barrel, lower support structure, instrumentation assembly, core shroud assembly; and the upper guide structure assembly including the upper guide structure support barrel, control element assembly shroud assembly, and hold down ring.

19

Enclosure 1 Response to July 21, 2010, Request for Additional Information Regarding Metal Fatigue for the Review of the PVNGS License Renewal Application APS performed a review of the ASME Subsection NG design reports and addenda, which included calculated usage factors for the components listed in the table below. A review of the UFSAR did not reveal any additional RVI components other than those analyzed and listed in the table below.

Fatigue Usage Factor Component Ul U2 U3 Core Support Barrel (CSB) 0.475 0.462 0.411 Lower Support Structure 0.074 0.078 0.075 Upper Guide Structure (UGS) Flanges and 0.258 0.31 0.125 Cylinders Upper Guide Structure Tubesheet Region CEA Guide Tube 0 0 0 Guide Tube to UGS 0 0 0 Support Plate Weld UGS Support Plate 0.1178 0.1178 0.1611 Fuel Alignment Plate (FAP) 0.7207 0.7207 0.9176 Guide Tube Extension to 0 0 0 FAP Weld Te009 Ne 00 CEA Guide Tube 0.0029 None 0.0088 The transient cycles from the replacement steam generator design reports shown in the table below were the input data for the RVI analysis. The report addenda for power uprate and steam generator replacement concluded that all code and specification requirements were satisfied. The design inputs in these references include a 40-year design life. The replacement steam generator transients were used to determine the limiting values for the enhanced Metal Fatigue of Reactor Coolant Pressure Boundary program.

20

Enclosure 1 Response to July 21, 2010, Request for Additional Information Regarding Metal Fatigue for the Review of the PVNGS License Renewal Application Limiting Transient(') Number of Events

1. Reactor coolant system (excluding 500 (The pressurizer) heatup from 70 OF to hot studsshl standby conditions at a rate of lted to 25

<100 °F/hr limited to 250 occurrences)

2. Reactor coolant system (excluding 500 (The pressurizer) cooldown from hot standby studs shall be conditions to 70 OF at a rate of limited to 250

<100 0F/hr occurrences)

3. 5%/minute power ramp increase, from 15,000 15% to 100% power
4. 5%/minute power ramp decrease, from 15,000 100% to 15% power
5. 10% power step increase, from 90% to 100% power
6. 10% power step decrease, from 100% 2,000 to 90% power
7. Normal cyclic variations at 100% power; 1.0E+06

+/-80 psi, +/-10 OF

8. Startup of one reactor coolant pump at 1,000 hot standby conditions
9. Coastdown of one reactor coolant pump 1,000 at hot standby conditions
10. Adding 40 oF feedwater at 875 gpm to the steam generator through the 15,000 downcomer feedwater nozzle when at hot standby conditions
22. Adding 40 OF feedwater at 875 gpm to the steam generator through the downcomer feedwater nozzle during50 loading conditions.

1 Transients not included in the analysis are not included. Transient numbering and descriptions are consistent with those in LRA Amendment 14, Table 4.3-2.

21

Enclosure 1 Response to July 21, 2010, Request for Additional Information Regarding Metal Fatigue for the Review of the PVNGS License Renewal Application Limiting Transient(1 ) Number of Events

23. Adding 100 'F feedwater at 875 gpm to the steam generator through the downcomer feedwater nozzle during loading conditions.
24. Pressure transients of 85 psi across the primary divider plate in either 4000 direction caused by starting and stopping reactor coolant pumps
29. Spurious reactor trips (including 50 (RCS),

operator error) at 100% power 240 (CVCS)

30. Loss of reactor coolant system flow 40
31. Arbitrary load rejection, from 100% 40 to 15% Power
32. OBE Condition - Full-load cycles about a mean value of zero and with an 200 amplitude equal to the maximum response during the entire OBE event
56. Adding 40 'F feedwater at 1750 gpm to the steam generator through the downcomer feedwater nozzles with the 280 flow initiated 30 seconds after a loss of normal feedwater
66. Major loss of coolant incident (system operating mode dependent upon design application for worst-case conditions)
69. Seismic event up to and including the safe shutdown earthquake (system operating mode dependent upon 1 design application for worst case conditions)
71. The concurrent loading produced by normal operation and full power, plus the design basis earthquake, plus loss- 1 of-coolant accident (pipe rupture) are used to determine the faulted plant loading condition 22

Enclosure 1 Response to July 21, 2010, Request for Additional Information Regarding Metal Fatigue for the Review of the PVNGS License Renewal Application Limiting Transient(') Number of Events

72. Major rupture in main feedwater piping (system operating mode dependent upon design application forworst case conditions)
76. Loss of Secondary Pressure: One cycle of a postulated loss of secondary pressure due to a complete double ended severance of one steam generator or feedwater nozzle, but not simultaneously. These are not considered credible events in forming Not Credible 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 occurs.
77. Reactor Coolant System hydrostatic 10 test
78. Secondary system hydrostatic test 10
79. Reactor Coolant System leak test 200
80. Secondary system leak test 200 Since the Subsection NG fatigue usage factors depend on effects of normal, upset, and emergency transient events, the increase in operating life to 60 years will not have a significant effect on these fatigue usage factors if the number of design basis transient cycles remains within the number assumed by the 40-year analyses. Monitoring the transient counts to ensure they remain less than their 40-year values will ensure that the CUF values remain less than their design basis CUFs. Because any design basis CUF less than 1.0 is an acceptable result, no additional action is required to be taken for components with CUFs close to, but less than, 1.0.

As demonstrated in LRA Table 4.3-3, the specified set of primary coolant design basis .

transient events should not be exceeded during the 60-year period of extended operation. Therefore, transient cycle counting under the enhanced Metal Fatigue of Reactor Coolant Pressure Boundary program will ensure that appropriate corrective action will be initiated if an action limit is reached for any transient cycle. Action limits will be established to permit completion of corrective actions before the design basis number of events is exceeded. Subsection NG fatigue in the reactor vessel internals 23

Enclosure 1 Response to July 21, 2010, Request for Additional Information Regarding Metal Fatigue for the Review of the PVNGS License Renewal Application will therefore be adequately managed for the period of extended operation, in accordance with 10 CFR 54.21 (c)(1)(iii).

NRC RAI 4.3-18

Background:

In LRA Amendment 16, the applicant amended LRA Section 4.3.5, "Assumed Thermal Cycle Count for Allowable Secondary Stress Range Reduction Factor in ANSI B31.1 and AMSE III Class 2 and 3 Piping." In LRA Section 4.3.5, the applicant identified all implicit fatigue analyses for ANSI B31.1 and ASME Class 2 and 3 piping components will remain valid for the period of extended operation except for the implicit fatigue analysis of reactor coolant system hot leg sampling lines and the recirculating steam generator downcomer and feedwater recirculation lines.

Issue:

The implicit fatigue analysis table provided for the RCS hot leg sampling lines includes a column, "Max. Calculated Stress Range per Eq. (11) (psi).". However, the column does not identify the source document for the referenced equation 11. Similarly, the implicit fatigue analysis table provided for the RSG DC and FW recirculation lines includes a column, "Max. Calculated Stress Range per Eq. (10) (psi)." However, the column does not identify the source document for the referenced equation 10.

In the assessment of the recirculating steam generator downcomer and feedwater recirculation lines, the applicant discussed two different analyses; the original implicit fatigue analysis and an updated pipe break analysis. LRA Section 4.3.5 does not clarify whether the pipe break analysis has a relationship to the original implicit fatigue analysis for these lines. It is also not clear whether both analyses are relied upon for the CLB or whether the pipe break analysis is a replacement for the original implicit fatigue analysis. It is not clear to the staff which of the analyses is the current analysis of record for the CLB and thus needs to be assessed as a TLAA for these lines.

Request:

a) Identify the source documents for the stated equation references.

b) Clarify which of the implicit fatigue analyses discussed in LRA Section 4.3.5 for the recirculating steam generator downcomer and feedwater recirculation lines is the analysis of record for these lines (i.e., the original analysis, the pipe break analysis, or both analyses).

24

Enclosure 1 Response to July 21, 2010, Request for Additional Information Regarding Metal Fatigue for the Review of the PVNGS License Renewal Application APS Response to RAI 4.3-18

Response

a) The stated equation references are those listed in ASME Section III, Subsection NC-3600 paragraph NC 3652.3 for the Class 2 piping and Subsection ND-3600 paragraph ND 3652.3 for the Class 3 piping.

b) The original analysis, calculation 13-MC-SG-506, is the code analysis of record for the downcomer, feedwater and recirculation lines. The pipe break analysis is a part of this calculation.

None of ANSI B31.1 or the ASME III Subsections NC and ND for Class 2 and 3 piping invokes fatigue analyses. The implicit fatigue analysis refers to the methodology prescribed in subsection NC and ND. To account for thermal cycling, piping in the scope of license renewal that is designed to these codes requires the application of a stress range reduction factor (SRRF) to the allowable stress range for secondary stresses (expansion and displacement). If the number of full-range thermal cycles is expected to exceed 7,000, ANSI B31.1, and ASME III Subsections NC and ND for Class 2 and 3 piping, require the application of a stress range reduction factor to the allowable stress range SA for expansion stresses (secondary stresses). These piping analyses are TLAAs because they are part of the current licensing basis, are used to support safety determinations, and depend on an assumed number of thermal cycles that can be linked to plant life.

PVNGS UFSAR Section 3.6 discusses high energy line break analyses, and the SG downcomer recirculation line (line SG-008) is identified in UFSAR Table 3.6-1 as a high energy line requiring a pipe break analysis, and discussed in UFSAR paragraph 3.6.2.1.1.2 B.

25

ENCLOSURE 2 Palo Verde Nuclear Generating Station License Renewal Application Amendment No. 22 LRA Section Affected Page Nos. RAI No.

2.1.6 2.1-26 4.3-18 3.1.2.1.3 3.1-5 4.3-13 3.3.2.1.8 3.3-11 4.3-13 3.3.2.1.21 3.3-25 4.3-13 3.3.2.1.30 3.3-34 4.3-13 Table 3.1.1 3.1-16 4.3-13 Table 3.1.2-3 3.1-84 4.3-13 Table 3.1.2-4 3.1-92 4.3-13 Table 3.3.2-8 3.3-119 4.3-13 Table 3.3.2-21 3.3-194 4.3-13 Table 3.3.2-30 3.3-236 4.3-13 4.3.2.5 4.3-63, 64, 65, 66, 67 4.3-13

Palo Verde Nuclear Generating Station License Renewal Application Amendment 22 Source: Response to RAI 4.3-18 Section 2.1.6, Generic Safety Issues (page 2.1-26) is revised as follows (new text underlined)

1. GSI-156.6.1, Pipe Break Effects on Systems and Components This GSI involves assumed high energy line breaks in which the effects of the resulting pipe break prevent the operation of systems required to mitigate the effects of the break. The aspects of pipe breaks that are associated with degradation are addressed in the aging management review tables associated with mechanical systems in Chapter 3.0. TLAA evaluations of high energy line breaks are presented in Section 4.3.2.14, High Energy Line Break Postulation Based on Fatigue Cumulative Usage Factor, and Section 4.3.5, Assumed Thermal Cycle Count for Allowable Secondary Stress Range Reduction Factor in ANSI B31.1 and ASME III Class 2 and 3 Piping.

Palo Verde Nuclear Generating Station License Renewal Application Amendment 22 Source: Response to RAI 4.3 Responses 2 and 3 Section 3.1.2.1.3, Pressurizer (page 3.1-5) is revised as follows (new text underlined)

Aging Effects. Requiring Management The following pressurizer aging effects require management:

  • Cracking
  • Cumulative fatique damage
  • Loss of material
  • Loss of preload Section 3.3.2.1.8, Nuclear Sampling System (page 3.3-11) is revised as follows (new text underlined)

Aging Effects Requiring Management The following nuclear sampling system aging effects require management:

  • Cracking
  • Cumulative fatigue damagle
  • Loss of material
  • Loss of preload
  • Reduction of heat transfer Section 3.3.2.1.21, Diesel Generator System (page 3.3-25) is revised as follows (new text underlined)

Aging Effects Requiring Management The following diesel generator system aging effects require management:

  • Cracking
  • Cumulative fatique damane
  • Hardening and loss of strength
  • Loss of material
  • Loss of preload
  • Reduction of heat transfer

Section 3.3.2.1.30, Miscellaneous Auxiliary Systems (page 3.3-34) is revised as follows (new text underlined)

Aging Effects Requiring Management The following miscellaneous auxiliary systems in-scope ONLY based on Criterion 10 CFR 54.4(a)(2) aging effects require management:

  • Cracking
  • Cumulative fatique damage
  • Loss of material
  • Loss of preload
  • Wall thinning

Palo Verde Nuclear Generating Station License Renewal Application Amendment 22 Source: Response to RAI 4.3 Response 1 Table 3.1.1, Summary of Aging Management Evaluations in Chaper IV of NUREG-1801 for Reactor Vessel, Internals, and Reactor Coolant System, (page 3.1-16) is revised as follows (deleted text shown with strikethrough)

Table 3.1.1 Summary of Aging Management Evaluationsin ChapterVII of NUREG-1801 for Auxiliary Systems (Continued)

ItemTpe ComonetAingEffct Mecanim AingManagement Further Dsuso Numberm C p n p Aging Efc .... .Program . _ Evaluation t*imber ;;!<*j  ! .... . ."'Recommended 3.1.1.06 Nickel Alloy Tubes and ICumulative Fatigue Damage ITLAA, evaluated in Yes, TLAA Cumulative fatigue daimage sleeves in a reactor coolant and secondary j accordance with 10 CFR

,54.21(c) of steam generator tubes is iet a TLAA as defined in 10 feedwater/steam CFR 54.3. See further environment _ J jevaluation 3.1.2.2.1.

Palo Verde Nuclear Generating Station License Renewal Application Amendment 22 Source: Response to RAI 4.3 Responses 1. 2. and 3 Table 3.1.2-3, Reactor Vessel, Internals, and Reactor Coolant System - Summary of Aging Management Evaluation - Pressurizer, (page 3.1-84) is revised as follows (new text underlined)

Table 3.1.2-3 Reactor Vessel, Internals, and Reactor Coolant System - Summary of Aging Management Evaluation Pressurizer Componenty ntended Materal Environment Aging Effect 1 Aging Management NUREG- Table1 Item <,Notes Typ FucinRquiring Poga 10 Vl

_______

__________ ________ Manag~ementZj 7 2 Itemi~J iPZR Support SS Carbon Steel Borated Water Cumulative Time Limited Aging IV.C2-10 31.07 A ISkirt Leakage (Ext) Fatigue Damage Analysis evaluated for SLeaka' Fati.uea the period of extended

__~°____ ____I ________Ioperation ____ _i Table 3.1.2-4, Reactor Vessel, Internals, and Reactor Coolant System - Summary of Aging Management Evaluation - Steam Generator, (page 3.1-92)is revised as follows (new text underlined)

Table 3.1.2-4 Reactor Vessel, Internals, and Reactor Coolant System - Summary of Aging Management Evaluation - Steam Generator Component nltended 1-:Material. Environment Aging Effct. Aging Management NUREG- Table 1 Item Notes Type Fu.nction i . .quiring Program 1801 Vol.

_____

___ ____ ______ Management~ Item 2______ ___

jSG Tubes HT PB Nickel Alloy Reactor Co()oant lCumulative Time Limited Aging iIV.D1-21 3.11.06 IA land Second ay Fatigue Damage iAnalysis evaluated for IWater (Int) Mthe period of extended

_____ operation ____

Palo Verde Nuclear Generating Station License Renewal Application Amendment 22 Source: Response to RAI 4.3 Responses 1, 2. and 3 Table 3.3.2-8, Auxiliary Systems - Summary of Aging Management Evaluation - Nuclear Sampling System, (page 3.3-119) is revised as follows (new text underlined)

Table 3.3.2-8 Auxiliary Systems - Summary of Agin. Management Evaluation - Nuclear Sampling System

-Comrponent I Intended Material f EnvirOnment-, v-Aging Effect- fAging Management NUREG- Table1 ItemT Notes'.

Type* Function... . Requiring-J Program 1801 Vol. I .'

PB Stainless Treated Borated Cumulative Time Limited Aging VII.El-16 3.3.1.02 A Steel Water (Int) Fatigue Damaqe Analysis evaluated for the period of extended operation _ __ ____

Table 3.3.2-21, Auxiliary Systems - Summary of Aging Management Evaluation - Diesel Generator System, (page 3.3-194) is revised as follows (new text underlined)

Table 3.3.2-21 Auxiliary Systems - Summary of Aginq Management Evaluation - Diesel GeneratorSystem

[ Component *; *ntended Material Environment Aging Effect Aging Management ". NUREG6- Table 1,Item -. Notes K type Function Requiring Program, 1801- Vol.i.

P._BB Carbon Steel Diesel Exhaust Cumulative Time Limited Aging VII.E1-18 3.3.1.02 t(Int) Fatique Damage Analysis evaluated for the period of extended

_ _operation

Palo Verde Nuclear Generating Station License Renewal Application Amendment 22 Source: Response to RAI 4.3 Responses 1, 2. and 3 Table 3.3.2-30, Auxiliary Systems - Summary of Aging Management Evaluation - Miscellaneous Auxiliary Systems, (page 3.3-236) is revised as follows (new text underlined)

Table 3.3.2-30 Auxiliary Systems - Summary of Aging Management Evaluation - MiscellaneousAuxiliarSystems Component; Intended Materia Envii~onmient Aging Effect Agi ,ng Managemlent NIJREiG- Table 1 ItemV Notes Type.. Function . Requiring Program P: 1801 Vol.

~Management 2 Item Pipingi PB Carbon Steel Secondary Water 'Cumulative Time Limited Aging 1VIII.B1-1 a 13.4.1.01 'A (Int) Fatigue Damage lAnalysis evaluated for the period of extended

_ t _ _ operation _ _ _

Section 4 TIME-LIMITED AGING ANALYSES 4.3.2.5 Steam Generator ASME III Class 1, Class 2 Secondary Side, and Feedwater Nozzle Fatigue Analyses Summary The PVNGS replacement steam generators (RSGs) are designed to ASME Ill, Subsection NB (Class 1) and NC (Class 2), 1989 Edition with no addendum. The design reports included design for a concurrent power uprate. The results of the fatigue analyses from these design reports are presented in Table 4.3-8.

Analysis Pressure-retaining and support components of the primary coolant side of the steam generators are subject to an ASME Boiler and Pressure Vessel Code, Division 1,Section III fatigue analysis. Although the secondary side is Class 2, all pressure retaining parts of the steam generators satisfy the Class 1 criteria, including a Division 1,Section III fatigue analysis.

The replacement steam generators were evaluated for a spectrum of design basis transients sufficient for a 40-year operating life, from date of installation.

Effect of Combustion EngineeringInfobulletin 88-09 The CE Owner's Group review of Combustion Engineering Infobulletin 88-09, "Nonconservative Calculation of Cumulative Fatigue Usage," did not identify any effects on the fatigue analysis of the original steam generators, and all of the PVNGS steam generators have been replaced.

Steam GeneratorTube Code FatigueAnalysis Not-a-TLA The design of the PVNGS steam generators includes a code fatigue analysis of the steam generator tubes, as indicated in Table 4.3-8. However, the cyclic stress range for the steam generator tubes is less than the endurance limit allowing an infinite number of cycles, so the CUF was determined to be zero. Since the steam -generator tube CUF is zero, the analysis of record will remain valid through the period of extended operation for all three PVNGS units. This analysis would be a TLAA if the safoty deter,,miation deprnded upon it. Howeer .. e dgn I. roport indi*ates a zro fatigue usago factor, and -a code fatigue analysis has historFically not proved1 suffic~ient to support the safety deteFrmination for steam genreatorF tubes, which depends On a separate tuboipecto pregram.

The various tube degradation mechanismsG not anticipated in the original design havýe r.eired stringen periqir;-odi iRrpectin prr rgam in r~dor t* ensuror adequate stear gee*ratorF tube integFrity. The teamF generator tubes are, in effect, (1) nRo longr qualifi*d for a licensed design life (10 CFIR 54.3(a) Criterion 3), and the (2) the fatigue analysis i therefore no longer the basis of the safety deteFrmination; in this caso that the tubes wl Palo Verde Nuclear Generating Station Amendment 22 Page 4.3-63 License Renewal Application

Section 4 TIME-LIMITED AGING ANALYSES iwoon primar-y F L*

mnaintain inor prsur ounoar' Trun ctonR ana 88oconaar; systems Thereforo, oven inisaltos(such as PVNGS) with oxcollont matorial and chemistr,'

control, Or on this case, new . team generators, the safety de8termiation for. integrity ot steam geoReator tubes now depondreonmanaging aging offocts by a periodic inspection program rather than On the fatigue aRalySis, and tho codo fatigue analysis of the tuber, i thereforo nt "aTAA^ However, the Steam Generatr Tube Itgrit'Fy Program (62.1 .8) will be Fredltod for. maintaiRi*g tub* integrity fEor tube dg*radatiOn MeGhanisms.

Table 4.3 PVNGS Steam GeneratorUprated Fatigue Comparison 1 7'

  • ,,,

. ....:..,-,

.*.... ,. . . .* . . . . . -Maxim'um

..- .*.. .. .... (R SD~si'n Baýsis 8 s U ra 'CUFý te d ) ,. ...

Component R sUp Lj Cu~rrent. Current "Desigh, IU1 /3 Design, U2 Support Skirt 0.08331 0.155 0.75104 NR(2)(3)

Support Skirt Access Opening Region Primary Head Hot Side 0.02895 0.0309 Cold Side 0.08502 0.0352 Primary Inlet Nozzle 0.04857 0.04634 Primary Outlet Nozzle 0.01683 0.01683 Primary Nozzle Dam Retaining Rings, Inlet and Outlet 0.0 0.0 Primary Manway Pad 0.02747 0.037 Cover 0.03494 0.019 Primary Manway Studs Hot Side (No TLAA, replaced every six years) 6.53 6.33 Cold Side (No TLAA, replaced every six years(4) in 4.011 4.67 Unit 2, nine years in Unit 1 and 3)

Primary Divider Plate 0.03 0.06 Tubes 0 0 Tube to Tubesheet Weld 0.18816 0.792 Tubesheet Hot Side 0.06570 0.928 Cold Side 0.39410 0.507 Tubesheet to Shell (Stub Barrel) Junction Hot Side 0.10059 0.064 Cold Side 0.99876 0.996 Economizer Cylinder (at the Tubesheet Cold Side) 0.01075 NR Palo Verde Nuclear Generating Station Amendment 22 Page 4.3-64 License Renewal Application

Section 4 TIME-LIMITED AGING ANALYSES Table 4.3 PVNGS Steam GeneratorUprated Fatigue Comparison 1

.MaximiunDesign Basis ICFUF Current Currenf

.. .i.....U.. : D es in,-U2 Secondary Shell 0.00773 0.00899 Secondary Shell Instrument Nozzle Holes and Nozzles NB-3222.4(d) NB-3222.4(d)

Exemption Exemption Small Nozzles NB-3222.4(d) NB-3222.4(d)

Exemption Exemption Economizer Feedwater Nozzle 0.90970 0.981 Downcomer Blowdown Nozzle 0.197 0.273 Downcomer Feedwater Nozzle 0.983 0.996 Downcomer Feedwater Piping Assembly 0.106 0.125 Recirculation Nozzle 0.099 0.114 Steam Nozzle 0.169 0.1767 Secondary Manway Pad 0.129 0.140 Secondary Manway Studs 0.618 0.7714 Secondary Handholes Welded on Lower Shell 0.113 NR Studs for Secondary Handhole Welded on Lower Shell 0.424 NR Secondary Stub Barrel Handhole 0.940 0.955 Secondary Stub Barrel Handhole Studs (No TLAA, replaced every 29 years in Units 1 and 3, every 18 1.35 2.15 years in Unit 2)

Upper Support Lugs 0.405 0.161 Feedwater Distribution Box 0.99201 0.988 The analyses are for a 40-year component life. The Unit 2 replacement steam generators were installed at about operating year 18, the Unit 1 and 3 replacements at or after operating year 20.

The analyses therefore qualify the replacement steam generators for a nominal 60-year plant life in Units 1 and 3, and 58 years in Unit 2.

Effects of the opening on the stress analysis were evaluated by evaluating stress concentration factors but no fatigue usage was calculated for the Unit 2 opening.

3 Not reported.

The Unit 2 design report does not distinguish between the hot- and cold-side studs, does not state the separate, lower 4.67 CUF for the cold side, and states a six-year replacement interval for both. However the supporting design analysis reports the 4.67 CUF for the cold side, and therefore states a Unit 2 cold-side stud replacement interval of eight years.

Palo Verde Nuclear Generating Station Amendment 22 Page 4.3-65 License Renewal Application

Section 4 TIME-LIMITED AGING ANALYSES The high usage factors calculated for the primary manway and secondary handhole studs require that these studs be periodically replaced. The fatigue analysis determines the replacement interval but is not otherwise the basis for a safety determination that depends on the licensed life, and the fatigue analysis is therefore not a TLAA for these studs.

Although the replacement steam generator designs are essentially identical, the Unit 2 code analysis was performed first, under separate contract. The calculated CUFs therefore differ to some extent. The results are identical or comparable where comparable methods were used. However:

  • The code requires only that the calculated CUF be less than 1.0. In some cases a simple analysis achieved this, and no finer analysis was applied to further reduce the result; though this may have been done for the other unit or units. Compare, for example, the CUFs for the tube sheet hot side and for the tube-to-tube sheet welds.
  • The code does not specify all locations which must be analyzed, leaving many of the detailed choices to the experience and skill of the analyst. For example, the Unit 2 analyst did not elect to perform a fatigue analysis at the support skirt opening or in the economizer cylinder near the tubesheet; the Unit 1 and 3 analyst did so.

With power uprate and replacement steam generators the worst-case usage factors calculated for the specified set of design basis transients exceed 0.9 in several other steam generator components. However, except for the steam generator tubes (which are subject to periodic inspection), fatigue usage factors in the steam generator components do not depend on flow-induced vibration or other effects that are time-dependent at steady-state conditions, but depend only on effects of operational and upset transient events. The enhanced Metal Fatigue of Reactor Coolant Pressure Boundary program (B3.1) will track these events to ensure that appropriate reevaluation or other corrective action will be initiated if an action limit is reached. Action limits will be established to permit completion of corrective actions before the design basis number of events is exceeded.

The Unit 1 and 3 replacement steam generators are also analyzed for a period sufficient to cover their installed life, to the end of the extended period of operation, and the Unit 2 replacement steam generators are analyzed for a period sufficient to cover all but two years of their installed life, including the period of extended operation.

The Unit 2 RSG tube-to-tubesheet welds and the hot sides of the tubesheets; and the cold side of the tubesheets and the feedwater distribution boxes in all three units, have high calculated CUFs but will be monitored using the cycle counting method as shown in Table 4.3-4. This will prompt actions that address the high-CUF locations when a cycle count action limit is approached.

Palo Verde Nuclear Generating Station Amendment 22 Page 4.3-66 License Renewal Application

Section 4 TIME-LIMITED AGING ANALYSES Disposition: Validation, 10 CFR 54.21(c)(1)(i), and Aging Management, 10 CFR 54.21 (c)(1)(iii)

The Unit 1 and 3 steam generators were replaced after the 2 0 th year of operation and have fatigue analyses extending beyond the period of extended operation (PEO). The fatigue analyses of the Unit 2 replacement steam generators are for a period sufficient to cover all but about two years of their expected 42-year installed life, including the period of extended operation.. Since the steam generator tube CUF is zero, the Unit 2 steam generator tube fatigue analysis of record also remains valid through the period of extended operation.

However, PVNGS has chosen to apply aging management to all the Unit 1, 2 and 3 steam generators to achieVeo uniformity in -ging management practices. The enhanced Metal Fatigue of Reactor Coolant Pressure Boundary program (B3.1) will track events to ensure that appropriate reevaluation or other corrective action will be initiated if an action limit is reached. Action limits will permit completion of corrective actions before the design basis number of events is exceeded, and before the cumulative usage factor exceeds the code limit of 1.0. Although tho steam g...rato tub* fatigue analysis iG not considero, a-dTL.fo.

T, tho roaso. . .tatod above steam gonorator tubo fatiguo will be managed by the Steam Generator Tube

" !tegrity program (B2.1.8). Effects of fatigue in the replacement steam generator pressure boundaries with Class 1 analyses, with the exception of the steam generator tubes, will thereby be managed for the period of extended operation, in accordance with 10 CFR 54.21(c)(1)(iii). The steam -generatortube CUF has been validated for Unit 1, 2, and Unit 3 consistent with 10 CFR 54.21(c)(1)(i) with a calculated CUF of 0.0.

Therefore the fatigue of the steam generator tubes is satisfactory for the period of extended operation.

The enhanced Metal Fatigue of Reactor Coolant Pressure Boundary program (B3.1) is described in Section 4.3.1; and is summarized in Appendix B, Section B3.1. See Table 4.3-4 for details of the program, and Section 4.3.1.5 for a discussion of its action limits and corrective actions. The Steam Generator Tube Integrity program is described in Appendix B, Section B2.1.8.

4.3.2.6 ASME III Class 1 Valves Summary Description PVNGS Class 1 valves (pressurizer safety, control, motor- and air-operated, manual, check, and solenoid) are designed to ASME IlI, Subsection NB, 1974 Edition with multiple addenda, the 1977 Edition with Winter 1977 addendum, and the 1989 Edition no addenda

[UFSAR Table 5.2-1]. ASME Section III requires a fatigue analysis only for Class 1 valves with inlets greater than four inches nominal. At PVNGS, specifications for some Class 1 valves with inlets four inches or less also require a fatigue analysis.

Palo Verde Nuclear Generating Station Amendment 22 Page 4.3-67 License Renewal Application

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