ML101540063

Responses to April 28, 2010, Request for Additional Information and Follow-Up Rais, for Review of License Renewal Application, and License Renewal Application Amendment No. 15

ML101540063
Person / Time
Site:	Palo Verde
Issue date:	05/21/2010
From:	Hesser J Arizona Public Service Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
102-06194-JHH/GAM
Download: ML101540063 (78)

Text

AM A subsidiaryof 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 102-06194-JHH/GAM May 21, 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 Responses to April 28, 2010, Request for Additional Information (RAI) and Follow-up RAIs, for the Review of the PVNGS License Renewal Application, and License Renewal Application Amendment No. 15 By letter dated April 28,,2010, the Nuclear Regulatory; Commission staff issued a request for additional information (RAI) regarding stainless steel, containment tendons, concrete examination, supports, and bolting, related to the PVNGS license renewal application (LRA). Enclosure 1 contains Arizona Public Service Company's (APS's) response to the. RAI.

In addition, the NRC staff has requested follow-up information for previous RAI responses regarding pressurizers, underground cables, and structures. Enclosure 2 contains APS's responses to the follow-up RAIs. Enclosure 3 contains LRA Amendment No. 15 to reflect changes made as a result of the RAI and follow-up RAI responses. also contains changes to LRA Section 4.5, Table 4.5-1, and Figures 4.5-1,

-2, -5, and -6 to reflect the results of the Unit 1 25-year tendon surveillance data regression analysis. This will complete the commitment in Item No. 53 in Table A4-1 made in APS Response to RAI 4.5-1 in letter no. 102-06160, dated April 1, 2010.

The commitment in Item No. 23 in LRA Table A4-1 has been revised, as shown in LRA Amendment No. 15 in Enclosure 3, to comply with applicable NRC Orders and implement applicable (1) Bulletins and Generic Letters, and (2) staff-accepted industry guidelines, for the pressurizer spray heads. This change is being made to be consistent with April the pressurizer spray head changes in APS letter no. 102-06160, dated 1, 2010. Lig A member of the STARS (Strategic Teaming and Resource Sharing) Alliance Callaway

• Comanche Peak

• Diablo Canyon

• Palo Verde - San Onofre

• South Texas

• Wolf Creek A a A0

ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Responses to April 28, 2010, Request for Additional Information (RAI) and Follow-up RAIs, for the Review of the PVNGS License Renewal Application, and License Renewal Application Amendment No. 15 Page 2 Changes to commitments are shown in the LRA Table A4-1 changes in Enclosure 3.

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.

Executedon *-'Z.-to (date)

Sincerely, JHH/RAS/GAM

Enclosures:

1. Response to April 28, 2010, Request for Additional Information Regarding Stainless Steel, Containment Tendons, Concrete Examination, Supports, and Bolting, for the Review of the PVNGS License Renewal Application

2. Responses to Follow-up Requests for Additional Information'Regarding Regarding Pressurizers, Underground Cables, and Structures for the Review of the PVNGS License Renewal Application

3. Palo Verde Nuclear Generating Station License Renewal Application Amendment No. 15

4. Construction Technology Laboratories (CTL) Report, October 18-20, 2005 Site Visit, Examination of Spent Fuel Pool Concrete Walls in Unit 1, CTLGroup Project No: 059084 cc: E. E. Collins Jr. NRC Region IV Regional Administrator J. R. Hall NRC NRR Project Manager L. K. Gibson NRC NRR Project Manager R. I. Treadway NRC Senior Resident Inspector for PVNGS L. M. Regner NRC License Renewal Project Manager G. A. Pick NRC Region IV (electronic) i i I

ENCLOSURE 1 Response to April 28, 2010, Request for Additional Information Regarding Stainless Steel, Containment Tendons, Concrete Examination, Supports, and Bolting, for the Review of the PVNGS License Renewal Application K RAI 3.3.2.2.3-1 RAI 4.5-3 I RAI B2.1.28,3 RAI 3.5.2.3-4 RAI 3.3.2-2

Enclosure 1 Response to April 28, 2H010, Request for Additional Information Regarding Stainless Steel`, Cdntailnment Tendons, Conrcfete Examination, Supports, and Bolting, for the Review of the PVNGS LicenseRenewal Application NRC RAI 3.3.2.2.3-! " '

Background

The staff reviewed Section 3.3.2.2.3."2 of the LRA against the criteria in the Standard Review Plan-License Renewal (SRP-LR), Section 3.3.2.2.3.2, which states that stress corrosion'cracking' (SCC) could occur'in stainless steel and stainless clad'sieei heat"exch'anger 6or'ipn'ents exposed to treated water greater than 60: C (14"0 F). The SRP-LR'rec6mmends further evaluation bf a plant-specific aging managementf program tenesure this aging effect is adequately managed :and notes that acceptance criteria are described in Appendix A.1 of the SRP-LR.

Issue The staff noted that, contrary to the applicant's statement, the related item applies to both boiling water reactors and.pressurized water reactors. Furthermore, in reviewing LRA Table 3.3.2-30,'the staff noted that the sample cooler heat exchanger is listed with anr*aging mechanism'of cracking for stainless steel exposed to secondary water. , Accordi'ng to LRA Table 3W.0-1, "Mechanical Environments,"

secondary water is described as treated water; however, the item description in Table 3.3.2-30 did not indicate th'e temperature. As such, it was unclear to the staff that this item is not applicable to PVNGS, as stated by the applicant.

Request Provide justification as to why SRP-LR Section 3.3.2.2.3.2 is not applicable or, if it is applicable, 'explain how aging will'be managed. '

APS Response to RAI 3.3.2.2.3-1 SRP-LR Section 3.3.2.2.3.2 is not applicable to the stainless steel sample cooler that is in a secondary water environment in LRA Table 3.3.2-30. SRP-LR Section 3.3.2.2.3.2 is referenced by SRP Table 3.3.1, item 5,'which is based on the following NUREG-1 801 linesý that are applicable to BWR lines for treated water:.

" VII.E3-3 Reactor Water Cleanup System (BWR) ... ..

" VII.E37-19 Reactor Water Cleanup System (BWR)

LRA Table 3.3.2-30 is for miscellaneous Auxiliary Systems within the scope of license renewal based only on criterion 10 CFR 54.4(a)(2).- The stainless'steel sample cooler in a secondary water environment in LRA Table 3.3.2-30 is in the secondary chemical control system. The stainless steel sample cooler that is exposed to secondary water will experience cracking and loss of materials aging effects that are managed by the Water Chemistry aging management program (AMP) and One-Time Inspection AMP consistent with NUREG-1 801 line VIII.F-3 (cracking) and NUREG-1801 line VIII.F-27 1

Enclosure 1 Response to April 28, 2010. Request for Additional Information Regarding stainless Steel, Containment Tendon'S, ConcreteExamintionhSupports, and S

- Bolting, for theReviewIof the PVNGS Licen~s Renewal Application (loss of material). The sample cooler in the secondary chemical contro! system has been evaluated consistent with other secondary PWR systems noted in, NUREG-1801 Chapter VIII for Steam and Power Conversions Systems. There are no PWR system aging management review (AMR) lines in NUREG-1801 Chapter VII (Auxiliary Systems) for stainless steel heat exchangers in secondary or, treated-wate~r.

LRA Table 3.0-1 Mechanical.Environments section defines.s~condary water as steam generator secondary systems water (incl6ding 6ondensate, feedwater, and steam) that is treated and monitored for quality under the. Water Chemistry Aging Management Program and controlled for protection of steam generators.'

NRC RAI 4.5-3

Background

In Section 4.5 of the LRA, the applicantt provided Table 4.5-"1 7Tendon:Regression Analysis Input Data. The table contains lift-off forces ofjindividual .tend6ns used in the regression analysis. The review of the, table identified. the.following anomalies:

a) Only the "shop end" force is providedfor ,6tndons H21-04 (3rd year surveillance),

V07 and VOi5. . . - .,

b) The lift-off force for tendon H21-04 was measured in the 3rd year surveillance and again in the 5th year.

~

"don.~ ~ forc thf-fava c) The Unit 3 dome horizontal tendon lift-off average forces are g ater the wall horizontal lift-off average forces. In some cases: they are greater by 'nearly 100 kilo-pounds-force (i.e., H13-45, H32-42, and H21-43).,

Issue The anomalies in items (a) and (c) could influence theslopeofregression analysis...

conducted to demonstrate that the predicted prestress forces&remain above their respective minimum required values (MRVs) through the period of extended operation. In particular, the staff is concerned with the influence of the higher Unit 3 dome horizontal tendon forces on the combined wall "and dome trend line (Figure 4.5-3), which shows the least margin. Ite'm (b), may not satisfy' IWL-2521i which requires that tendons examined during an inspection be selected on a. random basis from a sample of tendons that have not been examined in earlier examinations.

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Enclosure 1 Response to April 28,201,0, Request fo Additional Information Regarding S...tainless Steel, Containment 'end{ons, Concrete Examination' Supports, and Bolting, for the Review of the'PVNGS Licerise'ReneWal Application Request The app icant is requested to explain the ano'nalieS and 'confirm that they have' no impact' n its conclusion that' reg'rei6r !analysisrtrend lines indicate that tendon prestress will'remain alove .the ,MRV tlhrough the period of extended operation.

APS Response RAI 4.5-3 Response (a)

For the'regression'analyses of tendor'ipredsiress forces, Palo Verde uses the average of the shop and field .end lift-offs,' or, the sin~gievalue if only one end was lifted. The use of the mean of the shoýpl and field. nds r"athe'rihan including' both lift-offs in the regression was'usedto approximatethe method ýS4el at other nuclear planrts, either where a single end is lift*edor where'the averagfeis dused ifbboth ends are lifted:" In'any case, the effect on the regression analysislof using-the mean of the shop andfield ends rather than the individual end lift-offs is negligible for any difference observed to date.

A review of LRA Table 4.5-1 finds that a typical field-to-shop end difference is about 50 kips or less, with a few over 100 kips. The two most significant differences are the Unit 1, 3rd-year H21-42 with a difference of 177 kips, and the Unit 3, 3rd -year V1 6 with a difference of 171 kips. However, because V1 6 has a lower mean liftoff the effect on it is larger. ,

Since a linear regression (or in this case, a log-linear regression) is a straight line that minimizes the sum of the squares of the differences between the data points and the regression line, the effect of using a mean of two values instead of the two separate valuescan, be approximated by comparing the root mean square (RMS) of the paired value' to th'eir mean. For the Unit 31 3rd-year V1 6, the difference between the mean and the RMS of the paired values is-only 2.49 kips. -..

Even in this Worst case, this difference is"much less than the precision of a typical -

tendon lift-off. The lift-off precision islirited by the preci'sion of the pressure gauge used for the hydraulic jack. 'The best obtainable pressure gauge can usually be read to about +/-50 psig, which, for a typical tendon jack piston diameter, is equivalent to a lift-off precision of about +/-6 kips. The.jack-gauge calibration accuracy will of course introduce additional error.

There is therefore no'sig'nificant effect on the regression analysis results from differences between shop"and field ends up' to about 200 kips, when using the mean of the two ends instead of separate values in the regression analysis.

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Enclosure 1

-Response to April 28, 2010, Request for Additional Information Regarding Stainless Steel, ContainmentTendoAn, ConcreteEXaminationrASupports, and

' Bolting, for the Review %of the PVN'GS- License Renewal Application In the particular cases mentioned,

" The Unit 3 3rd.-year H21-04 lift-off was limited to the shop end because the lift-off of adjacent tendon H21-05 was below the Lredicted force,1 and this tendon and H21-06 were therefore lifted, in addition to those in the oiginally-specified sample, only to confirm the adjacent-tendon lift-off acceptance criteria. See the further discussion under Response (b).

• The Unit 3, 1st-year V07 and V15 lift-offs were limited to the shop ends, because the 1392 kip lift-off of the shop end of adjacent tendon V09 did not exceed the 1399 kip calculated, tendon-and-end-specific predicted force, apparently due to the V09 tendonfuntwisting fouir to six turnsduring tke surveillance. ,The shop ends of V07 and V1 5 were therefore lifte.l tio confirm the ladja 6ent-t16edn lift-off acceptance criteria. *Because the unt6wisting of Cv09 reduced the ,measured V09 shop end lift-off, the V09*data were not considered reliable for the, regression analysis, and were not used.. TheYIO9' shop end was rel-shimmed and successfully retested.

Response (b) . , ,,.

As stated in LRA Section 4.5, page 4,5-2,. ,,.

Priorto September of 1996 the'tendon examinations were governed by Regulatory Guide 1.35.... Under License Amendment 151 ... the program has been governed by ASMEXI Subsection IWL - 1992, instead of Regulatory Guide 1.35.. ...

The 1991 Unit 3 5th-year surveillance was therefore under Regulatory Guide 1.35"rather than ASME XI Subsection IWL. Ho[wever~similar sample requirements applied. The purpose of the requirementthat tendons not be repeatedly examined (except the "common" tendons), is to ensure an adequaie'random sample, and to eliminate effects on the surveillance data and conclusions of exercising any but the "common" tendons.

These intentions are adequately addressed ,by selecting a rahdom sample of adequate size, absent any repetitions, as spe~cified by the program then in'fore., This has ,been the practice at Palo Verde. ,

The 189-190, Uni 3, rd-yea6 ' ':.i-The 1989-1990, Unit 3, 3year surveillance required lift-off of only'six horizontal tendons including H21-05. H21-04 and H21-06 were added to the sample as "adjacent" tendons because the H21-05,lift-off was below 90 percent of the predicted force.

H21-05 was examined further, and no apparent cause being found, was re-tensioned to an acceptable value. H21-06 w'as also used for the sample Wire, because H21-05 had been shortened at initial installation and a sample wire could rnot be withdrawn (although The Palo Verde surveillance procedure and test records use different terms inherited from the Regulatory Guide 1.35 history. IWL-3000 uses "90% of the predicted force" for this action limit.

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Rsoeo Enclosure 1

....Respons to April 2t010, Request forAdditional Information Regarding Stainless Steel Containmient Tendons, C6ncrete Examiatin, Supports, and Bolting, for the RevieW of the PVNGS License Renewal Application

.

the tendon could be lifted off, asoshownby the dpta). This surveillance therefore examined the required horizontal samtple"of'six'tendons, plus the two adjacent required by the failure of H21-05 to lift-off above 90 perceht of the predicted force. These adjacent tendons wo*ld alto'havebeen iexarmiined under ASME XI Subsection IWL.

The repetition of the lift-off of Unit 3 H21-04,at the 5th-year surveillance was a coicmidence, due to this tendon having been placed in the original random sample for this surveillance consistent with IWL-2521.

Response (b)

Since dome tendons are shorter than"Vvall tendons, and since both are prestressed to comparable loads, it § expected that' the' domYe tenrdonsw*ill eilax less and therefore maintain a higher percentage of initial prestress. Other factors, however, such as initial prestress and prestressing order, will also have significant effects. For example, the first tendons stressed will relax more because of compression of the structure by the loads added by those that follow.

The original design and licensing basis of these and other, similar, Bechtel-designed inverted-U-and-hoop hemispherical-dome pr'stressedcontainments concluded that a single; horzontal hoop tendon group, composedof.both the wall and dome hoop tendons, was s§ufficient botfih for sample determination and for evaluation of prestressing trends. The design and inspection program are described in LRA Section 4.5 and in UFSAR Section 3.8.1". UFSAR Section 3.88.1.72 cites Bechtel Topical Report BO-OP-A Section 9.3.: BO-TOP-5A Section 9.3ý.'2.2 is the original basis for the use of only a single horizontal (or hoop) tendon group for both the dome and'cylinder tendons for this containment.

The influence of the higher Unit 3 dome' hoop tendon forces on the Unit 3 horizontal tendon regression analysis (LRA Figure,4.5,-3) will be assessed, ,in accordance with code requirements, by a revision of the regression"analysi& following each surveillance.

Thiswill eitherc0nfirm that the combined wall and dmrhe trend line indicates" atisfactory future performance, or will initiate necessary, corrective actions.

APS has also calculated separate regression lines for Unit 1"and Unit 3. Only four dome hoops' have been lifted in each'of Units"1 and 3. Due'to the limited data sets, these regressions have minimallplredictiv6 value.'

As shown in 'LRkA Table 4:5-1, only a single set of post-tensioned lift-offs has been done for Unit 2 (including no dome hoop tendons),and there aire,therefore, insufficient data for a meaningful Unit 2 regression analysis. The Unit 2 data were included in the unit common regressions.

APS believes that inclusion of data points for a given group7 in the regression analysis is the correct interpretation of ASME XI Subsection IWL and of the regulatory guidance.

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Enclosure 1 Response to April 28, 2010, Request forAdditional Information Regarding Stainless Steel, Containment Tendons, Concrete Exanination, Supports, and Bolting, for the Review of the PVNGS License Renewal Application In any case, elimination of any of these data would have a negligible effect on the regression analyses, particularly compared to the effects of theUnit 1, 25-year" surveillance data'on the Unit 1' and common reg ressions. The Unit 1,-25-year data*

have more significant effects on the regres'.s~n an~lyses, oth lbecause they are later, and because all of the 25-yeari data points dire below thie previously'-caldiiiaated Unit T regression lines. This slightly depresses the slope of.the Unit 1 and common regression lines but the revised regression analyses still do not fall b1low the' respective MRVs before 60 years..

If any of the regression lines were to fall below the MRV before 60 years, the LRA Appendix B3.3 Concrete Containment Prestress Program would still ensure that the';

safety function of the containment vessel is maintained, and if indicated by the surveillance results, tendons would be reotensioned, repaired, or replaced.  :

NRC RAI B2.1.28-3

Background

In response to RAI B2.1.28-2, dated February 19 2010 the apphcant stated that the frequency of inspection for the ASME .Section Xl, SubsectionjVVIL Aging Management Program (AMP) is consistent with ASME Section Xl Subsectipn IWL,;'laragra6h IWL-2421, Sites with Multiple Plants. This paragraph allows1nspecton intervals of ever y ten years, staggered, so that at least one.unit.is inspected every five years.,Therefore, the applicant concluded thatdthe exigting program isco*nsiste"nt with ASME .Section xi Subsection IWL and the Generic Aging Lessons Learned'(GALL) Re'port, and no`,..

exception is required. .,,

Issue ...

ASME*Section I11,Subsection IWL,`Subartice IAIL-2421 allowsthe inspection frequency to increase to ten years for examinations required by IWL-2524 and IWL-2525. However, IWL-2524 and WL-2525 requirements are for. examination of tendon anchorage areas, and corrosion protection medium and free water in the post-tensioning system, respectively. IWL-2410(a) Inservice Inspection Schedule of Concrete provides the inspedctib' periodicity of concrete surface exams and states, "Concrete shall be examined in accordance with'.IWL-2510 at 1,'3, and 5years following the completion of the containment Structural Integrity Test CC-6000' and every 5 years thereafter." IWL-2510 Surface Examination Requirements provides the requirements for concrete surface 'examinations.

Request .

Provide the basis for containment concrete surface visual examination inspection frequency often years.

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Enclosure 1 Response to.Apri I28,. 2010, Request for Additional Information Regarding stainless steel, iendons, Concrete Examintion- Supports, and

-ontainm:ent Bolting, for the Review of the PVNGS License Renewal Application APS Response to.RAl B2.1.28-3 During the current interval, Relief Request No. RR-L3,"approved by the NRC in a letter to APS dated October 6, 2000, (ADAMS Accession No. ML003758134) allows a frequency of 0*years for performing Section .Xl, Subsection IWL inspections of the concrete containment ~exterior surfaces. Subsequent intervals wil be in accordance with the requirements of ASME Section Xl, Subsection IWL (5 year frequency),

supplemented with thei al*plicalble requ~ireme"rnts 'of 10 CFR 50.55a, unless a relief request is approved to alter these frequencies.

NRC RAI 3.5.2.3 Background For the component type, "supports ASME 2 and 3," LRA Table 3.5.2-14 credits the ASME Section XI, Subsection IWF Program (B2.1.29) to manage loss of material for carbon and stainless steel components in'l) aeraWwater environment, and.2) a fuel oil'environment.

The'LRA states.that thes;e. environments are not in NUREG- 1801 for carbon'arid stainless steel components. ..

Issue The components in the environments identified above may have limited accessibility. It is not clear to the staff how the ASME Section Xl Subsection IWF Program will inspect the components to ensure the aging effect is being managed.

Request.,

Explain how' the ASME Section Xi Subsetion IWF Program will manage the effect of' aging on carboh and stainless steel compbnents in 1);a raw water environment, and 2) a fuel oil environment that have limited acd'essibility.

APS Response to RAI 3.5.2.3-4 ASME 2 and 3 Supports in a Raw Water Environment Palo Verde ASME Class 2 and 3 supports in a raw water environment are ASME Class 3 stainless steel supportslocate2 1n"the essential spray ponds. AgingofASME Class 3 stainless steel supports located in the essentiai slray ponds are managed by the ASME Section Xl Subsection IWF (B2i1.29) AMP.' Duringý intervals 1 arid 2, ASME Class 3 supports in the raw water. ehvironment of th6 essential spray ponds were

• examined using remote camer'as. During the third interval (current interval), only 10%

of the Essential Spray Pond ASME Class 3 supports are examined. No essential spray pond ASME Class 3 stainless steel supports in a raw water environment were selected 7

Enclosure 1 Response to April 28, 2010, Request for Additional Information Regarding Stainless Steel, Containment Tendons, C6ncrete Examination, Supports, and

'Bolting, for the Review ofthePVNGS License Renewal Application for examination as part of the 10% sample population during the third interval because.

they were not the most susceptible to corrosion and there was no prior history of component support member deformation..

Submerged supports for the essential spray ponds are made of stain`1esslsteei."There" are no carbon steel supports in- the raw water environment of the essential spray ponds.

LRA Table 3.5.2-14 has been revised,'as shown inrAmendmeht No. 15'in Enclosure 3, to delete carbon steel ASME Class 2 and 3 supports in a raw water environment.

- 1"` 7 ASME 2 and 3 Supports in a Fuel Oil Environmehnt.

Palo Verde ASME Class 2 and 3 supports in a fuel oil environment areASME Class 3 supports located in the diesel fuel oil storage tanks for support of the diesel fuel oil transfer pump. Supports for the diesel fuel oil transfer pump are within the scope of the ASME Section XI Subsection IWF (B2.1.29) AMP, but exempt from examination requirements. , , ,

Supports for the diesel fuel oil transfer pump are rna.de of carbon steel. However; the' LRA incorrectly listed stainless steel sui ports in a diesel fuel 'oil envirbnmeht.' LRA'Table 3.5.2-14 has been revised, as shown in Amendnent No. 15 in Enclosure 3, todelete stainless steel mechanical equipment ASME Class 2 and 3 supports in a fuel oil environment.

NRC RAI 3.3.2-2 '

Backgqround In the LRA, Tables 3.3.2.2 Spent Fuel Pool Cooling and Cleanup System, 3.3.2.7 Essential Spray Pond System,, and 3.3.2.20 Diesel.Fuel Oil Storage and Transfer System, there are aging management review (,,MR)"result lines for stainless steel or carbon steel closure bolting in environmeiits of treated borated water, raw water, and fuel oil, respectively. The AMR results credit the Bolting' Integrity Aging Management Program (AMP) with managing the aging effect-loss of preload for bolts in these wet environments.

Issue The LRA does not provide sufficient information for the staff to understand how the Bolting Integrity AMP can effectiv.ely manage loss of p reload for boltsin wet' environments where simp!e visual ,techniques during normal system walk downs could detect sign s of closure bolting loosening, such as indications of leakage around a flange or gasket which would Otherwise be noticed in'a dry environment.

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Enclosure 1 StaiRdespOnse to Apil 28r,2010, Request for Additional Information Regarding SStarles s Steet, Containment Tendons, Concrete Examination, Supports, and Bofting, for ihe Review of the PVNGS License Renewal Application Request a) Explain what features or activities of the Bolting Integrity AMP will manage the aging effect of loss of preload for closure bolting that is in a wet environment and why-the credited features or activities are adequate to ensure that loss of the subject bolts' intended fuhction(sd) Oes.not occurrduring the period of extended operation.

b) 'Clarify whether there are any indirect indicators that might be indicative of loss of bolting preload for closure bolting in each of these wet environments and their associated systers.

APS Reps onse to RAI 3.3"i.22 . .

Response (a) j. .

The LRA B2. .7 Bolting Integrity program manages loss of preload for pressure retaining bolting and includes prel6ad control, selectio hof bolting material, use of lubricants/sealants consistent witlh EPRI gbOddb6Itihig: practices, bnd performance of periodic inspections for indication of aging effects. In the case of submerged bolting in the spent fuel pool cooling and cleanup system and the diesel fuel oil storage and transfer system, the LRA B2.1.1 (XI.M1) ASME Section XI, Subsections IWB, IWC, and IWD aging management program manages the inspection of safety-related bolting and supplements the Bolting Integrity Program.

For safety-related pressure retaining bolting covered by LRA B2.1.1 (XI.M1), ASME Section XI Inservice Inspection, Subsections IWB, IWC and IWD, the inspection is conducted to detect loss of material due to corrosion and evidence of leakage. System pressure testing is performed to detect evidence of leakage. The extent and schedule of ASME Section XI, Subsections IWB, IWC, and IWD inspections combined with periodic system walkdowns, assure detection of leakage before the leakage becomes excessive. Any unusual indications of system performance (pressure drop over a holding time, pump/piping vibration, noise, etc.) observed during periodic system walkdowns and pressure testing would also be identified and entered into the corrective action program.

An exception to the LRA B2.1.7 Bolting Integrity Program identifies that loss of preload is not a parameter of inspection for the PVNGS Bolting Integrity Program. The discussion of bolt preload in EPRI NP-5769, Vol. 2, Section 10, indicates that job inspection torque is non-conservative since for a given fastener tension more torque is required to restart the installed bolts. The techniques for measuring the amount of bolt tension in an assembled joint are both difficult and unreliable. EPRI NP-5769, Vol. 2, Section 10 suggests that inspection of preload is usually unnecessary if the installation method has been carefully followed. Torque values are provided in procedures, if not provided by the vendor instructions, design documents or specifications. The torque values provided in procedures are based on the industrial experience that includes the 9

Enclosure 1 Response to April 28, 2010, Request for Additional Information Regarding Stainless Steel:,Containment Tendons, Concrete Examination, Supports, and Bolting, for the Review of the PVNIGS License Renewal Application consideration of the expected relaxation of the fasteners over the life of the joint and..,

gasket stress in the application of pressure closure bolting.

Resbonse (b)

Loss of preload'is not a parameter of inspection as indica'ted above* The aging1 management of loss of preload does not rely onthe indicators'bf loss of preload forC inspection. It relies on proper installation torque, and techniques to manage the loss of preload instead. -

Inability to successfully complete an ASME Section Xl pressure test'or observation of unusual indications of system performance (pump/piping vibration, noise, etc.) during, periodic system walkdowns would be used as indicators to reveal a'pctential loss of-preload or other aging effects for closure bolting. Failure to meet test acceptance criteria for an ASME Section XI pressure test would require corrective actions consistent with the ASME Section XI Inservice Inspection Subsections IWB IWC and IWD LRA B2.1.1 AMP. Observation of unusual indications of system performance would be identified and entered into the corrective action program.,. .

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ENCLOSURE 2, Responses to Follow-up Requests for Additional Information Regarding Regarding Pressurizers, Underground Cables,

.- :andStructures forthe Review of the PVNGS License Renewal Application Follow-up RAI Follow-up 4.3.2.4-3 RAI 4.3.2.15-1 Follow-up RAI 4.7.5-1 Follow-up RAI 4.7.8-1(b)

.1!!I .Followmup RAI B2.1.26-1

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.....

......

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.....

..........

.Follow-up 2 -.

".PRAI B2.1.32-1 '

.!.............................

Follow-up RAI B2*1.32-2 Follow-up RAI B2.1.32-3

Enclosure 2 Responses to Follow-up Requests for Additional Information Regarding Regarding Pressurizers, Underground Cables, and Structures for the Review of the PVNGS License Renewal Application NRC Follow-up RAI 4.3.2.4-3 Identify the impact of the power uprate and steam generator replacement on design reports for the pressurizer.

APS Response to Follow-up RAI 4.3.2'.4 -3*

(This response supplements the RAI 4.3.2.4-3(a) response provided in APS letter no. 102-06139, dated March 1, 2010)

The originailresponse to RAI 4.3.2.4-3(a) states that the Westinghouse design report addendum for theipressurizer confirms that thepower uprate and steam generator replacement modifications have no effecton the pressurizer design reports for any of the three units. This' bonclusioný applies to.the severity of the design basis transient events, and is unaffected bythe numberof occurreI ces~ofeach transient event assumed by the analyses, and is also unaffected by the design life. In conclusion, the power uprate and steam generator replacement modifications have no effect on the pressurizer design reports through the. period of extended operation.

NRC Follow-up RAI 4.3.2.15-1 APS Response to RAI 4.3.2.15-1 mentioned three cracks (0.5 x 39 inches, 1.0 x 34 inches, and 0.35 x 45.5 inches),:an 8 x ;18 inch long crack, and a 1 x 8 inch long crack for the fatigue crack growth calculations.

(1) Discuss whether these cracks will be acceptable at the end of 60 year per SRP 3.6.3.

(2) The purpose of performing fatigue crack growth calculation is to demonstrate that fatigue is not an active degradation mechanism in the LBB piping as specified in SRP 3.6.3. The applicant mentioned that the crack of 1 by 8 inches long would not grow through the pipe wall for more than 400 years. Describe the methodology (e.g., input parameters) to show that how 400 years were derived. Discuss how many year the other four cracks (other than the 1 x 8 inch crack) would grow through the pipe wall.

(3) The applicant stated that for three cracks (0.5 x 39 inches, 1.0 x 34 inches, and 0.35 x 45.5 inches), the evaluation calculated that through-wall leaks would occur at 21, 4, and 38 years, respectively. Discuss if these results are related to fatigue crack growth calculations.

(4) The applicant needs to conclude that fatigue is not an active degradation mechanism during the period of extended operation and that the original LBB evaluation is valid for the extended period of operation.

1

Enclosure2 Responses toQ Foll6w`i-up Requests forAdditional.Information Regarding

,Regarding Pressurizers, Underg'round Cables, and Structures

. rfthe Review of the PVNGS License Renewal Application (5) The applicant stated that"...The Metal Fatigue'aging manragerment program is not implemented to .monitor the transient cycles to confirm that the transient cycles used in the fatigue crack growth analyses for the LBB piping exceed the actual transient cycles becauseethe existing LBbfaiigue craick growth eValuation is valid for the period of extended'oper'atjon. .'.-'. The staff beli6ves that the AMP B33. 1, Metal Fatigue of Reactor Coolant Pressrii'e Bo6nd'ary," needs tb be'implehente'd to' verify that the' transieht cycles used ihnthe fatigue crack growth analyses for'the. LBB piping do not exceed the actual transient cycles. Although the applicant may have Used conservative transient'cycles in its fatigue crack growth calculations, the applicant still needs to verify and confirm that the transient cycles u'sed in the-fatiglue c ack'growth calculationr exceed'and bound the actual operating transient bycles. Therefore".theapplicant needs to implement the, metal fatigue aging management program for LBB piping, or-justify why the metal' fatigue aging management program should not be implemented on the LBB piping.

(6) The applicant stated that "...The fracture mechanics stability analysis is not time-dependent and therefore remains applicable for the period of extended operation..."

Explain why the fracture mechanics stability analysis is not "time-dependent.

APS Response to Follow-up RAI 4.3.2.15-1 (This response supplements the' RAI 4.3.2.15-1iresponse provided in APS letter no.' 102-06139 date'dMarch 1,2010)

The postulated semi-eliptical cracks (three crac6ks 0.5 x 39 i*rches 1.0 x 34 inches, and 0.35,x 45.5 inches for the stability analysis; ahd Mtwcracks 8 x 18 inches and 1 x 8 inches: for the fatigue crack growth calculations) would b'e accelptable at the end of 60 years..

Thebpostulated flaws assumed in the CE analysis (O15 x 39 inches, 1.0 x 34 inches, and 0.35 x 45.5 inches) dedmonstrated that the cracks preferentially grow in the-radial direction With verysmall circumferential 'extensi(n from the initial length. These flaws were shown to be stable even under the Safe Shutdown Earthquake which is the most severe loading and will also be detectable by the normal leakage monitoring system of the plant. Similarly; it was shown that'cracks (1'incl deep and from:8 to 18 inches in length), smaller than those described above but larger than the ASME Section III initial inspection acceptance criteria, required several design life times to grow through the -

pipe wall. Based on the above, the criteria required to demonstrate that a leak-before-break condition exists for the RCS main loop pipin g 1l6as been satisfied and is app~licable for the current license and for the extended period of operation (i.e., a total of 60 years) as further explained in responses to requests for clarification (2), (3) and (5)'below.

2

Enclosure 2 Responses to Follow-up Requests for Additional Information Regarding Regarding PNessurizers, Underground Cables,and Structures for the Revýiew "ofthe PVYNGS License Renewal Application Response (2)

A Linear Elastic Fracture Mechanics (LE.FM) analysis' was performed to determine the crack growth of the various postulated semi-eiliptical shaped inner surface cracks in the RCS main loopý,piping. The.method of analysis is based on ttb-Section Xl, AOp6*ndixA subsurface flaw evaluation procedure, where thgefatigue crack, growth rate'da/dNof the .

material is characterized in terms" f th'eý argofappliedlstress inten'sitV factor AK (pages 5 thruT7 of CE R6porthLD-83-;108). "

This characterization is generally of the form*da/dN -C, (AKi )n which has been determrined experimentally. The material co*'stant'sfor carbon steei faiigue cracK :.'

growth in a water environment utilized by CE in the ir evabluatidn are as follows:

n = 3.726 C= 3.795 x 10-10 The fatigue crack growth curve used inithe'evaluatioin'also'inclided upper-bound data to envelope the Section Xl curve as illustrated in Figure 1 of the CE report.""

From this method of stress intensity factor determriination; the AKli level'is calculated based on the crack sizepand loading conditions asdefined in Table 1. Using a stepping procedure for the number of cycles of loading in a,given time period, depth and length crack growth rates are calculated, and the correspondihg change in cra1ck size is determined as well as the time required to penetrate the entire pipe wall and produce a leak. The start-up shutdown transient was found to be the greatest contributor to the usage factor for.the main loop piping. A cyclic *stress of 18 ks.i,. conservatively enveloping the ,. , start-up shutd:down stress,

. , 1,,  :'was

.. !. .applied

, tp

,. thehy'pothetical

• -' p . e -, flaws one, inch, deep and from 8 to 18 inches long inthe circumferential direction in both the 42. inch diameter hot leg and 30 inch diameter cold leg piping. Figure 5shows-that the number of start-up shutdown cycles necessary to cause a one inch deep crack from 8 to 18 inches long to grow through the pipe wall ,and leak is at least 3000 to 8000 cycles, which is significantly greater than the .40 year design value of 500 cycles'and projected Value of 213.cycles in 60 yeIars. Thus )it can ýbe said ,that the part-through-wall cracks will not propagate through the entire pie p awall formore than,400 years.

For flaws,1 inch deep, larger than 18 inches, see Response (3) below.

Response 3)"

Yes, these results are related to fatigue crack growth calculations.

The procedure used for the"postulated cracks that will penetrate~the pressure boundary is similar to the one described above except that the stress intensity factor, Ki, was evaluated using the finite element method since the ASME Section XI flaw evaluation procedure does not extend to through wall cracks (pages 6 and 7 of CE Report LD .3

,Enclosure2 Responses to,, F6IIow-up'Reque'sts for Additional Information, Regarding Regarding Pressurizers. Underground Cables, and Structures for the Review of the PVNGS License Renewal Application 108).' Fo the range 6f flaw sizes§ and shapes assumed the resulting crack growth rates were calculated for the design basis load transients and corresponding frequencies of occurrence given in Table 1 of the CE report. Therefore, the results are related to.

bounding fatigue crack evaluations, meant to demonstrate that cracks, significantly larger than the ASME III inspection acceptance criteria, in the RCS main loop piping will remain stable afte propagatig throuh ral" the entre pipe Res§onsd (4)

Palo Verde was reviewed 'to NUREGT0800"Standard Review Plin Section 3.6.3 Revision 0.' Within the context of Sad~rd Review Plan Section 3.6.3 demonstration that fatigue Js not an active degradation mechanism means that:

a. Fatigue crack growth analyses for the'proposed licensed operating period are acceptable (in this case, including thep'lriod of extended operation).

The'acceptability 6f the 'LB"B'fatigue. crack groWth analyses for the licensed oper aiing period iick'iding"'the','period 6f.extended operation'is addressed by this" section of the LRA, as further explained in responses'to requests for

• clarificatiuon (1),'.(? a6i6d (3) abdve.n and (5)"below*

b. CIack stability' ahalyses are acceptable;ýand if time dependent, are acceptable for the proposed licensed operating period (in this case, including' the period of extended operation).

The acceptability of.the LBB crack stability analyses, and the fact that they are hot time-cdependent,' are addressed by thig section of the LRA, as further explair'&d in the response to requestis for clarification.'(1) and (3) above and (6) below.

C. M.aterial fraciure't60ghne~s paramet'ers, Jncluding effects of long-term thermal aaging,.are acceptable...,.

The LBB crack stability analyses include appropriate material toughness parameters, which are not time-dependent, as described in this section of the LRA ahd as further'explained' in the respQnseIo' r6.quest for clarification (6) below.

d. There are no indications of pressurized water stress corrosion cracking, erosion, erosion-corrosion, water hammer, creep, other cracking, leakage, or other evidences of actual or incipient fatigue or failures that would indicate that the LBB analysis are invalid, within the scope of the piping exempted from'MEB-3.1 crack postulation by the LBB analysis.,,

4

Enclosure 2 Responses to Follow-up Requests for Additional Information Regarding Regarding Pressurizers, Un derground Cables, and Structures for the Review of the PVNGS License Renewal Application No such indications or failures have been observed in the scope of Palo Verde, Qnit 1,2, or3LBB piping. ,.

Response (5)

The LBB analysis performed by CE for the RCS main loop piping consisted of two bounding evaluations that in essence remove the time dependOnc'y"from the projected' crack growth evaluation. In the first evaluation cracks larger than allowed by the ASME Section III initial inspection acceptance criteria are determined to take 3000 to 8000 ':

cycles, of the most significant contributor tofatigue usage,factor (start-up and shutdown), to be able to propagate through thi entire. pipeWall (Figure 1). The40 year design value for this transient is 500 cycles, an'd thei,'rojected value for 60 years is 213" cycles. In the second evaluation, 'cracks significantly la'rger thaei the ASME allobwable are demonstrated to remain stable after leaking under the most severe loading which is the Safe Shutdown Earthquake. Therefoe.,.given the results of the above sensitivity evaluations, it is confirmed that the' numberiof transient cycles accumulated oveira 60 year period will not affect the results of the LBB eyaluations and as such the metal fatigue aging management program should not ben'implemented to track the cycles on the LBB piping as it relates to the CE. evaluation'.. Hoever, the'RCS main loop, pIiping is part of the overall ASME Class 1 piping mnietal fatigue 'man0agement program Where transient cycles are monitored in accordance with the Technical Specification 5.5.5 Component Cyclic or Transient Limit Programto ensure components, are maintained within design limits.

Response (6)

Two crack stability criteria have been used to assess the likelihood!that a crack with opening stress intensity K,or a J-inhegral at the tip would remain sltable. Both'nmethods are independent of time. The first involves the use of linear elastic fracture mechanics fracture toughness Kjc which is a measure of the material's resistance to fracture. A K, value below which there. is no crack extension is Kc. As a practical consideration,' Kjc is a measure of the stress intensity at Which fracture takes place. Its value has been empirically correlated to the material's Charpy V-Notch test Value using the Rolfe-Novak-Barsom correlation that is independent of time. The second method involves the elastic, plastic crack instabiity theory using the J-integral crac'k tip parameter T,.the tearing modulus, when thee volume of plastically deformed material is appreciable. As before, a TApplied Value below which theere is no' crack extension is TMaterial. TMaterial is only a function of the J-integral, modulus of elasticity E, and the material's yield stress Sy and not time.

NRC Follow-up RAI 4.7.5-1

1. Provide confirmation that the evaluation in Reference 12 of WCAP 15973-P-A applies to the Palo Verde pressurizer nozzle repairs.

5

Enclosure 2 Responses to Follow-up, Requests for'Additional Information Regarding

. Regarding Pressur!zers,,Underground Cables, and Structures for the Review of the PVNGS License Renewal Application

2. Provide confirmation that the plant-specific corrosion rate using times (in-percentage oftotal plant life) at normal operating,. intermediate temperature, startup, and cold shutdown'oniditito s based on the plant-specific operating experience arid the assumed conditions for the extended period of operatioh will remain within the projected overall cgrrosion rate of WCAP-15973-P-A, Rev. Q..

APS Response to Foll6oW-Ud RAI 4.7.5-1 (This response supplements the RAI 4.7.5-1 supplemental response provided in APS letter no. 102-06160, dated April 1, 2010)

Re'poh'se 1 " ' ",. i ." . ," . .' t  :

• Arizona Publc Serice Company (APS).hs reviewed Reference 12 of WCAP 15973-P-A and confifnmbd that it considers"lplant-Specific Palo Verde data-ahd it applies to the Palo Verde pressurizer nozzle repairs.

Response 2 APS has assessed the plant-specifi6 corirb'sion rate, as §pecified in Section 2.3.4 of WCAP-1 5973-P-A, Rev. 0, with plant specific operating data for normal operation, start-up, and shutdown conditions after implementation of the pressurizer heater sleeve repairs, based on the following:

Unit 1 - From implementatibn of repairs to Unit i Cycle 15 Short Notice Outage

- Normal operation: 88.92o%.

- Startup: 0.95% -,

- Shutdown: 10.13%"

Unit 2 - From implementation of repairs to end of Unit 2.Refueling 15

- Normal operation: 89.14% "

- Startup: 0.99% ,. . .. "

Shutdown-: 9.87%... ..... .. .

Unit 3 - From implementation of repairs to end of Unit 3 Refueling 15

- Normal operation: 89.74%

- Startup: 1.00%

- Shutdown: 9.26%

The operating data is generally consistent with the a suiintions of the WCAP. The.

resulting assessment estimated repair lifetimes of,220' years for'Unit 1; 221.6 years for Unit 2, and 230.2 years for Unit 3, which are within the&projected overall 'corrosion rate of WCAP-1 5973-P-A, Rev. 0.

6

Enclosure 2 Responses to Follow-up Requests for Additional Information Regarding Regarding Pressuri;ersrUnergroUnd Cablesnand Structures for the Review of the PVNGO License Renewal Application In addition, APS made the following commitment in letter no. 102-05324, dated August 16,2005: ,

APS commits to continue to track the time at cold shutdown conditions against the assumptions made in the corrosionanalysis to assure that the allowable bore diameter is not exceeded over thelife oJ6the pant., if the analv'sisassumpti.ns are exceeded, APS shall provide a revised an'alysis t6 thei NRC and provide a discussion on whether volumetric inspection of the area is required."

NRC Follow-up RAI 4.7.8-1(b).

Describe how the Structures Monitoring Program will be used to monitor settlement of the Fire Water Pump House, the Transformer foundations, and the Station Blackout Generator Structures during the period of extended operation. If settlement of these structures is not in the scope of this TLAA, provde a technical basis for their exclusion.

APS Response to Follow-up RAI 4.7.8-1(b)

(This response supplements the RAI 4.7.8-1(b) response provided in APS letter no. 102-06160, dated April 1,201,0,) 7 The Fire Water Pump House, the Transformer Foundations, and the Station Blackout Generator are not within the scope of the TLAA for Settlemient Monitoring des'ribed in UFSAR Section. ", 2.5.4.13.

- ' Because k,i*' the, .* - 'rWater

' ' Fire * "il' Pump

- , ; ý"*House,

,t : ' e., the: Transformer

ý"* . ' - 1ý Foundations, and the Station Blackout Generator are not contained or incorporated by reference in UFSAR Section 2.5.4.13, they do not meet criterion (6) Of the TLAA definition in 10 CFR 54.3(a). Therefore, the settlement monitoring consistent'with UFSAR Section 2.5.4.13 for these structures would not be a TLAA.

The B2.1.32 Structures Monitoring Program*will monitor the Fire Water Pump House, the Transformer Foundations, and the Station Blackout Generator structuires on'a 10-year frequency to detect degradation due to settlement. Justification for the 10-year frequency is provided in response to Follow-up RAI B2.1.32-1 in this Enclosure.

NRC Follow-up RAI B2.1.26-1 Provide clarification regarding the inspection of manholes associated with'Inaccessible Medium Voltage Cables Not Subject to 10 CFR 50.49 Environmental Qualification Requirements (g2.1.26. ,)!.E3), including the water level that will require manhole inspections and the criteria for moving inspections tp a two-year interval.

7

Enclosure 2 Responses to Follow-up Requests for"Additional Infonnation Regarding

.. Regarding Pressurizers, Underground Cables, and Structures "fortlhe Reviewof theoPVNGS License Renewal Application APS Reshonse to Fol ou RAI B2.126:1 (This response supplements the RAI B2.1.26-1, Response 2 provided in APS letter no. 102-06134, dated February 19, 2010)

During ,manhoie'walkdow!nWin 2009, 6ne statioh'blackout cable manhole was found to contain water s'ubm&ging the cables. Subs*qduent-insioection of a '6onnected manhole found additional watet, but no-additioIal s'gubmnerged cables.' A'revievwof the history of these manholes found recurring instances'of water intrusion. The'manhole found to contain (wate'r with subrerged'Cablies has hid a-seal: replaced, lid raised'above grade, and the ground surface reworked to route Water away from the manhole. 'As a result of this operating experience, APS moved these manholes to the most frequent inspection interval.

The 'preventive naintenan'e: (PM) pro6gram grouPs the manholes into three frequencies of inspections based on their history of w'ater intrusion: twoNeeks, six months, and two years. The two-week PM task requires inspection of manholes if it has rained 0.3 inches or more within a424-hoUr period since the last time the PM'was pe'rform-ed.

(Note:. as a result of a typographical error, the February 19, 2010, RAI B2.1.26-1, response incorrectly slpe!'ified`3 inches instead of 0:3 inches.) The manholes grouped into this'PM task aref also inspected dn o a s ix-monthhfrequency to ensuIre they are regularly inspected even durng dry peiods. Underthe PM program, a manholewill not be moved to the two-year, reduced frequenCy inspection interval until it has been found dry for two years.

The Operating Experience discussion in LRA1B2.1.26 hs been revised, as shown in A d tNo.15 'in Enclosure 2 torefle~t this clarification.

LRA~ ~ ~ ~ ~ A. mi.dn tor16t NRC Follow-up RAI B2.1.32-1

Background

By letter dated April 1, 2010, the applicant supplemented it's response to RAI B2.1.32-1 regarding the Structures Monitoring Program inspection !interval.

Issue In the supplement, the applicant enhanced the Structures Monitoring Program to inspect structures within the scope of license renewal at intervals not to exceed 10 years. The spray pond inspections and settlement monitoring surveillances both remained at five years. The applicant did not include structures exposed to a natural environment or, structures inside' primary containment within the five year inspection interval as suggested in ACI 349.3R. The supplement includeda brief explanation of why the applicant believed a ten year inspection, frequency was adequate for structures exposed 8

Enclosure 2 Responses to Follow7up Requests for Additional InformationRegarding Regarding Pressurizers, Underground Cables, and Structures for the Re'view'of the PVNGS License Renewal Application to a natural environment; however, the' explanation 'did not consider environmental factors and no discussion was provided for structures inside primary containment,.

Reýuest .

Align the Structures Monitoring Program.inspection interval with the guidance in ACl 349.3R, or provide a detailed techniqal justification for a Iqngerinspectipho interval for structures inside primary containment orexikposed to a natural enyironr1ent. An.'

adequate exp anation needs t6addressall environmental factors to which a structurer may be exposed, as well as relevant operating experience (eg., hum idity, high winds, temperature fluctuations, cracking, spalli.ng, ,ec',*)..,

APS Responseto Follow-up RAI B2.1.32-1 (This response replaces theRAI B2.1.32-.1 supplemental, response provided in APS letter no. 102-06160, dated Op'i 120.

APS will align the Structures MOnitoring ,Program inspection frequency to comply with' the guidance in ACI 349.3R for the, safety-related...category i stihuctures." For structures within the scope of license renewal, prior to the Operiod of, extended operation the Structures Monitoring Program will be enhanced to esiablish*the freq .uenc'o")finspection for each unitat a 5 year interval. and an inspection frequency of 10 years for the.

exterior surfaces of nonsafety-related ,structures, below-grade structures, and structures within a controlled' interior environment.

All of these inspection frequencies are consistent with ACl 349.3R-96, except the 10-year frequency for inspection ofthe',exterior surfacs6sof n*nsafety-related structures.

The justification for the,10-year frequenlcy is based'on a Maintenance Rule r'isk' assessment, non-aggressive environmental factors, and operating experience, as described below. The exterior surfaces of the following nonsafety-related structures within the scope of license renewal will be inspected on a 10-year frequency:

• Fire Pump House (Yard Structures)

Radwaste Building

• Station Blackout"Generator Structures

• Turbine Building

• Non-Safety Related Tank Foundations and Shells

• Non-Safety Related Transformer Foundations and Electrical Structures Maintenance Rule I ' ..

The non-safety related structures have been determined-to be low risk in compliance with 10 CFR 50.65 Maintenance Rule- risk significance assessment and performance monitoring criteria in.accordance with NUMARC 93-01; 9

Enclosure 2 Responses ,to Follow-up Requests for Additional Information Regarding

.Regarding Pressurizers,. Underground Cables, and Structures

.fo6r the Review of the PVNGS License Renewal Application Environmental Factors In additio;n to the l'ow rislk significance of the nonsatety-related structures, these structures are'ex*sed tora relatively benign envir'oment. Palo Verde is located in~a very arid area with low humridity pr'evailing"throughlout'most ofthe year. Rainfalliis very.

I1w (averaging just 7 inches per year), arid the water tabl, is well below the lowest structures. Therefore, degradation resulting from high humiditV, rainf*.l, and"gro*nd water is not a concern. Freezes occur, but these are infrequent, light, and -of relatively short duration. Also, the freezes that do occur are generally associated.with clear and dry conditions rather than humid or wet conditions that can degrade structures more.

readily therefore degr'adation due to freeze/thaw cycles ris not a concern. Palo Verde is in the negligible weathering region as sho*wn, on Figure 1 of ASTM C 33.' In addition to low humidity and rainfall,;the atm0Si'l~lere at Palo Verle is free of corrosive vapors.

Therefore, atmospherically induced degradation is not a concern for the exterior surfaces of non-safety related structures Within the scope of license renewal.

Operating- Experience ~.

With exception of prior Turbine Building roof leakage described in the Operating Experience section in LRA B211 32,' there haVe beehnnoadverse conditions noted during the Structures M6nitoring Program inspections of the exterior surfaces of non-safety related structures withi*r the scope oft icense'renewal that resulted in a loss of structural integrity'or a sheltdr/p0rotect intended function. Items identified during the periodic inspections that require c~rrect'i actions are alsobinspected in the other two units. After each periodic inspection of a structure, a reassessment of the inspection frequency is required to be performed based on the results, of the inspection. Again, the inspection frequen6y is deterrmiined commenrsurate with the safety significance of the structure and ts condition, and future inspectioni results m'ay result in more frequent monitoring. '".

LRA Sections A1.32 and B2.1.32, and Item No. 34 in Table A4-1, have been revised, as shown in LRA Amendment No. 15 in Enclosure 3, to specify the ifnspection frequency enhancements to the Structures Monitoring Program described above.

NRC Follow-up RAI B2.1.32-2 ' .. " "7

Background:

By letter dated February 19, 2010, the applicant responded to RAl B2.1.32-2 regarding the Structures Monitoring Program acceptance criteria'"

10

Enclosure 2 S .Responses to Follow-up Requests for Additional ,Information, Regarding Regarding Pressurizers, Underground Cables, and Structures for the Review of the PVNGS License Renewal Application Issue:

In response to RAI B2.1 .32-2, the applicant provided a description of minor, adverse, and critical deficiency categorizations; however, the'.descriptions are vague and qualitative and do not incorporate th6e 'qu'an'titative-ac'ce"ptance' criteria di'cussed 'in'ACI 349.3R. The category descriptions ieave muc of e rating to the dgment of the inspector or program owner.

Request:

Explain how the quantitative requirements of ACl 349:3R are incorporated into the, Structures Monitoring Program acceptance ciriferia. lhclude'reference8 to thetsite documents or procedures which contain the accep'tanceý criteria...

APS Response to Follow -u RAI B2'.1.322i (This response supplements the RAI B2.1.32-2 response provided in APS letter no. 102-06134, dated February 19, 2010)

Prior to the period of extended oper'ati.on'*, "theb StrcturesMonit~ring Program vI ii.be ilý enhanced to quantify the ac6eptance criteria .°andcritical parameters for monitoring degradation, and to provide guidance for identifying unacceptable'conditions requiring further technical evaluation or correctiye'. actioni. The Struciures Monitoring Program Will also be enhanced to incorporate applicableindustry codes, standards, and guidelines" (e.g., ACI'349.3R-96, ANSI/ASCE 11-90, etc.) foracceptance criteria.

LRA Sections A1.32 and'B2.1.32, and Item Nb 34 in Tb Ie A4 have been revised, as shown in LRA Amendment No.15 in Enclosure 3 to specify an enhancement to the' Structures Monitoring Program to include quantitative acceptance criteria.

NRC Follow-up RAI B2.1.32-3

Background:

By letter dated February 19, 2010, the applicant responded to RAI B2.1 .32-3 regarding leakage from the Unit 1 spent fuel pool. .

Issue:

In response to RAI B2.1 .32-3, the applicant provided a discussion of an event in 2005, which led to borated water migrating through the concrete walls of the spent fuel pool.

The staff requires additional information about the leakage to ensure aging is being properly managed.

11

Responses......to Follow-up

. ... Requests for Additio nclosure sRe 2 E'cMoti Re onal'Inormation Regadn

.. .,egardin gaPressurizers, Underground-Cables, and Structures for thoe"Reviewof the PVNGS License Renewal Application Request:

1. Provide cppyCof thd h Constructi6n'Technology Laboratories (CTL)report which found the 2005 leakage' did not have an adverse effect on the SFP concrete.

2. Exlgih whether o.r ot the'leakage through the concrete has stopped completely. If the leakao I-as s'toppede.lprb~vide the ba-sis for that conclusion., Explain how the possibility of leakkge 'throdgh 6c6'!n,6rete:! in ina'6e;siblb area~s has bee6 addressed.

3. ..How !orig. did tlie 2005. leakagd last and.'how. much'water leaked through the con6rete during the event?'

4. Provid [e th current am_6ut of dria iagfromthe tell-tale RD drain valves in each Unmit.

5. Havelthere been any' indications of 'leakage through the SFP concrete in Unit 2 or 3?

APS Response to Follow-up RAI B2.1.32-3 (This response supplements the RAI B2.1.32-3 response provided in APS letter no. 102-06134, dated February 19, 2010) .

.Response.1 Construction T.echrnilogy Labloratotie's (CTL) Report, October 18M20, 2005 Site, Visit, ExZminatioki of Spent Fuel A PobiConcreteWalls in Unit 1, CTL Grbup Project No:

059084 "is provided in`En;closu're"4 to"this letter.

Response 2 Leakage through the concr'ete in the areas tnat were initially identified as showing leakage has stopped completely. This is based on the fact that there are no longer wetted areas visible on the concrete surfaces either in the originally identified wetted concrete areas or other areas where the concrete structure that surrounds the spent fuel pool is visible.

As it suggests, leakage through concrete in inaccessible areas does not readily lend itself to visual examination. Palo Verde drains and monitors the drainage from the leak collection and detection system through the tell-tale radioactive waste drain (RD) valves on a daily basis. In addition, in,2006 and 2007 shallow aquifer wells were installed outside theProtected Area downgradient of each unit. These wells are periodically sampled as part of the groundwater monitoring program to detect groundwater impacts, such as radioactivity. To date, no activity has been detected in these wells. The Palo Verde corrective action program would be used to address degraded or non-conforming conditions.

12

Enclosure 2 Responses to Follow-up Requests for Additional Information Regarding Regarding Pressurizers, Uhderground Cables, and Structures for the Review of the PVNGS License Renewal Application Response 3 According to information provided in the, station assessment performed to evaluate the 2005 leakage, draining the Unit 1 leakage collectiohand detection systemnvia the tell-tale drain lines had not been performed for approxifiately ,f*ur imlonths due to limitied flow in the downstream lines. Once thesedrain, lines were clbar~d ,and the tel!'tale drain*

lines opened, the driiving force behind* the leakage was,'stopped.,lit'was determined that at the plant east exterior wall of the,fuel building, le-sss than- eight ounces' f fluid .was released to the environment on the wall and pavbed ground su:rface. At the plant south side of the fuel pool wall (which is inside the fuel building), iess t.an eight ounces of" fluid was released but contained inside the building. A survey of the exterior Wall where the 2005 leakage was located did not indicate anydirect contamination or smeara6le contamination, as detected on a frisker.

Response 4 . .. . ..

The amount drained daily is recorded by the plant operators'aind anomalies'evaiuated on by the system engineer.

For the period of March 1, 2009, to May 5, 2010, drainage in Unit 1 was less than * -

approximately 51 quarts, Unit 2 was less than approximatelv 1,37 quarts and Unit 3 was less than approximately 78,quarts. Under curreht guidance, any ýamount dra'nedfrom-a tell-tale line that is greater than'0 quart but lessý than 0.5 quart is recorded as 0.5 quart.

Response 15  : ..  ;.

There have not been any indications .of leakage through the SFP concrete in Unit 2 or 3.

13

ENCLOSURE 3 Palo Verde Nuclear Generating Station License Rengwal Application Amendmen;t No.15 LRA Section . Page Nos. RAI No.

Table 3.5.2-14 ,3.5-132, 135 i-3.5.2.3-4 4.5 4.5-4 Completed commitment 53 Table 4.5-1 4.5-5 to 4.5-10 - Completed commitment 53 Figures 4.5-1, -2, -5, -6 4.5-11,12, 15, 16,-1.7 Completed commitment 53 A1*32"  : '":i.A-18 *.A-j- ' 1* Foilow-up

• . B2.1.32-1 and Al . - B2.1.32-2 Table A4-1, Item No`.;'ý' .. ...

A-50 Add pressurizer spray heads 23 Table A4-1, Item No., - A52 Follow-up B2.1.32-1 and 34 B2.1.327-2 Table A4-1 ,5Item No. A-59 Gompleted commitment 53 B2.1.26* B'i76, 77 Follow-up B2.A..26-1 B2.1.32" 2B92 B- 2 93, 94 94Follow-up B2.1.32-1 and

______________________________B2.1.32-2 B2 1 .2 -

The complete Appendix A and B,aging management program'sections are provided for reviewer conveniennce when there is any change to the sections.

Palo Verde Nuclear Generating Station License Renewal Application Amendment No. 15 Source: Response to RAI 3.5.2.3-4 Table 3.5.2-14, Containments, Structures, and Component Supports,- Summary of Aging Management Evaluation -

Supports (pages 3.5-132 and 3.5-135), is revised as follows (deleted text shown with strike through)

Table 3.5.2-14 Containments, Structures, and Component Supports - Summary of Aging, Managemnent Evaluation - Supports (Continued)

Component, JIntended Material, Environment'. Aging Effectt Aging Management' i, NUREG. Table 1*Ite Notes Type Function Requiring Program

______

___ _____

___ _ _____I

_____ _____ Manaement2 Item _ __ _ _ _ __ _

SUPPe~te GaFben Steel I Raw w4tbF (ýý Less of matorbia None NOne ASME Scction X ASME 2 & 3 (ub2.4.2n I)

Table 3.5.2-14 Containments, Structures, and Component Supports Summary of Aging Manage,6ent Evaluation - Supports (Continued)

Section 4 TIME-LIMITED AGING ANALYSES

• * . .,I ,.*.. .., '

• The tendon liftoff acceptance criteria of the ti6don inspection procedures are consistent'with those of IWL-3221.1, 1992 Edition.

Surveillance Results A regression a'nialysis was-perform'ed on Unit I'and 3 'horizontal and vertical tendon'data including the 25-year s'kv'eillanlefand for Unit2, and the 15-vear'surveillance'for Unit 31:5'y64 guR'bilahnes for1nits 4 and 3, an&OR the 20 year survillance data for Uni~t 2. The Unit 1,.25 year surveillance was not Gomplete in ie.ob included in this analysis.

The average value. between the shopand field epd liftoffs.of,.eachtendon was used to obtain a trend line indicating loss of prestress. This trend line was projected to 60 years. The loss curves. were compared to the predicted loss. of .force.- The projected..loss lines were recalculated using Reg Guide 1.35.1 methods to determinle Ioss of prestress as a percent of the original prestress value for high-loss, median-loss, and low-loss cases. The projected loss lines are consistent with the original design calculation,,.which calculated. loss of prestress. for each 'tendon in.the surveillance .sample;_.. using' methods. consistent with Proposed Reg Guide c:1.35.1, Rev..O, April 1979.

The regressioný analysis trend -lines indicate*that ,tendon prestress: will: remain above the minimum required value (MRV,) throug'h the. end of theperiod of extehnded operation.

Table 4.5-1, summarzes input. data. --In the.table, common, tendons-(those-surveyed at each inspection) are marked in boldface., Figures 4.5-1 through, 4.5-6 summarize the results of these regressions, . -

Disposition: Revision, 10.CFR"54.21(c)(1)(ii); andAging Management, 10 CFR-54.21(c)(1)(iii).

• Revision - ,' . ' -

The condition of the PVNGS containment prestressing system meets criteria for revision of the predicted loss of prestress for the. period of extended' operation as described in NUREG-1800, Section 4.5.3.11.2' (1) The' lift-,off trend lines Were'calculated by regression of individual tendon lift-off data, including results of the 20085,' Unit 12 25g-year.surveillance.

These calculations "are% therefore consistent 'with :NRC. Information Notice 99-10, . (2) The, regression ýnalysis!of surveillancei lift-off data extends the trend lines for both the 'vertical and horizontal cylinder tendons Ito 60 years..- (3) The trend lines for all tendon groups remain- above their minimum required values for the period of extended operation. . ' ..

The current regression analysis of the vertical and horizontal. cylinder tendons therefore revises the predicted loss of prestress. f6 r the period.: of extended oleration, and Palo Verde Nuclear Generating Station " Page 4.5-4 License Renewal Application Amendment 15

Section 4 TIME-LIMITED AGING ANALYSES demonstrates that loss of prestress is expected to remain within acceptable values for the period of, extended operation, in. accordance with 10 CFR 54.21:(c)(1)(ii).

Aging Management The existing Concrete Containment Tendon Prestress program '(Section B3.3) will be continued for the period of extended operation. The average tendon prestress .in each of the Vertical *and hoop tendon groups will thereby be maintained above t design, basis minimum required value (MRV) for.the Reriod of extended operation, in accordance with 10 CFR 54.21 (c)(1)(iii). .. .. , S .

-Table4.5-1 Tendon'RegressionAnalysis input Data For PVNGS Units' ;,2, and 3 Year ~

' ~ I Time at Yer Tendon Force ~ ,<Teosic-n

~Unit (1) (2 Date <~ITension.

'T Kips ~ ~ Date~, (er'

,j~~r Horizontal CylindWerWall anI Dome (HqoL )endonls 1 1 H13-07 Field ýý1393. 01,/27/84, 02/16/82 1 .94 1 1 H13-07 Shop 1366 01/27/84 02/16/82 1.94 1 1 H13-21, Field:' 1497 03/08/84, 01/15/82t 2.14 1 1 H13-21 .Shop 1446'ý 03/01/84 01/15/82iý 2.12 1 1 H21-37 Field 1471 02/13/84 01/11/82 2.09 1 1 H21-37 Sh_§' 1453 02/28/84 01/11/82 2.13 1 1 H32-16' Field ' 1387 02/1:0/84 06/01/81 2.69 1 1 H32-16 Shop 1370 03/07/84 06/01/81 '2.77 1 ,! H32-30 Field_, 1502, 02/13/84 05/27/81- !2.72 1 1 H32-30 Shop 1492 03/07/84 05/27/81: 2.78 1 1 H21-44 Field. 1508 02/14/84 05/18/81 2.74 1 1 H21-44 Shop 1517 02/28/84 05/18/81 2.78' 1 3 H13-25 Field 1486 01/02/86 01/14/82 3.97 1 3 H13-25 Shop, 1410" '01/03/86 '01/14/82 3.97 1 3.. H21-11. Field ,.1451 01/12/86 '02/12/82 3.92 1

• 31,. H2111 Shop _1440 01/06/86 02/12/82 3.90

.1, 3.' H32-09 Field 1464 01,/05M86 02/16/82 3.89 1 3 H32-09, Shop 1483 01/02/86 _02/16/82 3.88 1ý 3 H32-30 Field ,1516 01/09/86 05/27/81 4.62.

1 3 H32-30 Shop - 1487 01/02/86 05/27/81 4.60

1. 3 H32-33 Field 1505 01/09/86 01/12/82 4.04 1: 3- H32-33 Shop 1402 01/02/86 01/12/82 4.04 1 3 H2"1-42, :Field& , 1645 01/13/86 05/19/81 4.65 1 . 3 H21-42 Shop :1468 -01/03/86 05/19/81 4.63 Palo.Verde Nuclear Generating Station ,Page 4.5-5 License Renewal Application Amendment 15

Section 4 TIME-LIMITED AGING. ANALYSES Table 4.5-1 Tendon Regression Analysis Input-Data For PVNGS Units 1,.2, and.3...

'Year Tendon Force, TensTieoa

~4~i 1) (2), End Di.ps ate Date Tension (Years) 1.__ 3/45 H13-19 Field :1420 03/06/88 02/11/82 6.06 1:- 5,- 1413-19 ý,Shfop 13864 , 04/26/88 02/11/82 6.20 1'. 5 H21-03 Field,: A1419 04/19/88&- 02/17/82 6:17 1' 5' H21-'03T. Shop 1507- 04/26/88 02/17/82 6.19 1 5 H21-28. Field 11419 04/15/88  :'05/27/81 6.89 1 5 H21.28 j Shop! 168 04/15/88 0,5/27/81 6.89 1 5, H32-23 -Field .1458'-: .04/19/88, 01/15/82 6.26

1 5.. H32-23 SioL§ 1442 03/26/88 01/15/82 6.19, 5 U1 H3230 Field' _1493 ,04/19/88 05/27/81 6.90 1 5; H32-30. Shop . 1479 03/25/88 0.5/27/81: 6.83, 6

1 ' 5  : "H32-44 -Field :1557 .104/8/88, 05/19/81 j 6.92 1, 5 "H32-44'-- -..-

S1hop - 1,486 03/2688 05/19/81 6.85 1! 10 H1308 Field 1389 08/27/92 06/30/81 11.16 1 10 H13-08 H j Shop 1380 '08/25/92 ,06/30/81 11.15 i 10 .H,-H32130 Field 1483, 08/12/92, 05/27/81 11.21 1I 10 ..- H32-30` Shop' 1465 08/14/92 05/27/81 11.22

1. 10, H32-4 Field., 1466 . 08/11/92 01/11/82 ' 10.58 1 10., H32-41 , Shop :1413 0/11/82 108/14/92, 10.59 1 15 H21-06 1Field 483 .05/15/98. 06/30/81 16.87 I 1'5 H21-06,. Shop 1,303 06/11/98 06/30/81 16.95

.1 15 H32-15 Field '.1442 0~

06/02i98 02/12/82 16.30 1 15 ,- H32-15 Shopý! 1:367- :-06/02/98- 102/12/82- 16.30 1 15, H32-30 Field: 1,463-: 05/15/98 05/27/81 16.97 i 1 15. H32-30 Shop ,-463 06/02/98 ,05/27/81 17.02 H13-114 Field 6

_25 09/13/08 0625/81 27,.22 1" 256 H3-14, H' Shop 11399 09/,16/08. 0'6/,25/8'1 273 I ,25 H32-26, Field 1414' 09/19/08 05/28181' 27.31 1 25: H32-26 hop. 1333 09/13/08 q05/28/81 27.30 1 25 H32-30 Field 1413 09/1 9/08 ;05/27/81 -,- 27.32 ,

1 25. H3240 ShoO. :1.424 09/.13/08 05/27/81 27.30' 2 20' H21-40 -Field 1368 07/08/05 05/03/82 23.18 2 20 H21-40. Shop 1:402 07/07/05. 05/03/82 -23.18 2 20 H32112 Field 1,301 '.,08/17/05 05/03/82 23.29 2 20 H32-12 Shop 1:351 07/01/05 06/15/82 23.04 2 20 H32-30 Field. 11358 08/17/05, 061/15/82 23.17 2 20 H32-30 Shop 1297 07/01/05: :05/20/82 23.12 Palo Verde Nuclear Generating Station Page 4.5-6 License Renewal Application Amendment 15

Section 4 TIME-LIMITED AGING ANALYSES Table 4.5-1 Tendon Regression Analysis Input Data For PVNGS Units"1,'2,' and 3' unit.

Year

)

Tendon *En e.

Tension Time at

.,c Kips Date Date, 4 eso (Years) 3 1 H13-..0 Field. --..1471 .1017!87 03/28/84 3.55 3 1-: H13-10 Shop Y1415 10/13/87 .*_03/28/84.' 3.54 3 1 ..- H13-36. ,Fieid '1401- 10/17187T. 10/21/83 .. 399 3 1' H13-36 Shop; 1392. 10G/13/87.1 -1.0/21/83 3.98 3,- 1. H13-44 Field '1490 .. 10/15/97 08/12/83 4.18 3 1 H13-44 Shop 1457 1ý0717/97 08/12/83 4.18, 3 1 .H21-07 Field 1422. 10/1!9/87 "04/09/84 3.53

3. 1 H21-07, "Shop 1414 . 10/14/87-. .04/09/84 . 3.5.1

3 1 H32-13 Field, 1433. 10/19/87.o04/05/84 3.53 3 1 . H32-13 Shop' 1".!466- 1 i'0.!6/87. 04/05/84 3.53 3 .1 H32-2! Field 1453 10/19,87'W""04/03/84 ,3.541 3 1 .. H32 -Shop.: ._1448 10/16/87 04/03/84 ,3.53:

3 3 H13-36 Field 1375- 12/20i89 10/21/83 6.17, 3 3 H13-36 Shop 1363 01/22/90 I'.10/21/83 6.26' 3 3 H13-45 Field 1504 12/20/89:.. 10/27/83 ' 6.15' 3 3 H13-45 Shop;,,. 1498 01/22/90 '10/27/83 .6.24 3 3 - H21-'04-- Shop 1403 01/24/90 '03/30/84 5.82 3 3'-' H21-.05 Field  !:*1392 1.2/14/89 04/10/84 5.68 3 3 . H21-05" Shop' 1349 '01/22/90". 04/10/84 .5.79

`3 3 H21-:06. Field .1434 .03/23/90 03/29/84. . 5.981 3- 3 H21-'06' Shop 1352 .. '01/24/90 03/29/84 5.82 3 3 H21-09 Field'- 1431 112/14189 04/09/84 . 5.68 3 3 H21-09 ._Shopj. 1387 101/22/90 04/09/84. 5.79 3 , 3 H32-18 Field. 1453-.. 12/13/89' 03/11/84 5.76 3 3. H32J18, "Shop., 1483 t12/19/89 03/11/84 5.77 3 - 3. H32-29 Field:.- 14.10 .12/14/89 1/1!01/83 6.12

.3 3 H32-29 Shop, 1444- 12/19/89 11/01/83 6.13 3: 5 H13-09 Field _' .1469 08/08/91 04/09/84 7.33 3 5, H13-'09. ' -Shopi 1361 '"'08/09/91"I 04/09/84 7.33

.5-e H13116' j .3 Fld 1479., 08/20/91 03/27/84 7.40

• 3, 5 H13-16 Shop 1354 08/19/91, 03/27/84 7.39

3. 5 H13-36 Field 1,394 .08/05/91 10/21/83 7.79

• 3 5 -H13-36 Shop. 1368 08/09/91 10/21/83 -. 7.80 3 :H21-04*

-5 Field 1314 P.07/19/91. 03/30/84 7.30 3 5 H21-04 " Shop 1394 i 108/09/91' 03/30/84 7.36 3 5 " H21-25 'Field 1396 07/19/91 ':03/30/84 ,_7.30 Palo Verde Nuclear Generating Station Page 4.5-7 License Renewal Application

'Amendment 15

Section 4 TIME-LIMITED AGING ANALYSES Table 4.5-1 Tendon Regression An'alysislinput Data ForPVNGS Units 1,, 2, and3 3 - . _

P Ir*IcTimeatv

-,Year, Tendon 'Di '

'Unit. (1

• • (2)2)-.n End

.. . o

, Kips 'Date Tni

-(Years) Tension Date shop 1570- 08/14/91 03/30/84 7.37 31 H32-42:,P Fie.ld- -.T100- 07/22/91 10/20/83 7:75 3 5 H32-42 .Shop'! 1.468 08/05/91 - .10/20/83- 7.79 3 10 .H1'3-2 Field . :1458- - i/06l96 03/26/84' 12.70

3. 1 '13--24 .Shpp '- 1314 '112/06/96 03/26/84 1-2.70 3* 10 H13-36 " . 371' 1211 /96 13.14 1':10/21/83 S 10 H13.'36 Shop 1,342 12/11/96 -: 10/21/83 -- 13.14 3 10 H21-10: 'ield-; 4313 '1/13/96 03/28/84- 12.63 3 10 H21-10" Shop 1A324- .12/11/96 03/28/84 12.71 3, 5 :H13ý36 -Field--- 1i315 C08/19/02 10/21/83. 18.83 Sop 133 08/16/02 10/21/83. 18.82 3 15 H21-:22 Fi.eld. 1345 '08*16/02 03/26/84 - 18.39 3 1 :H21422 Sh6p. -1,317 :08/16/02 03/26/84, 18.39

'iS21*43-.ield -- 1456 08/16/02 10/27/83 -18.80 H21-43.Shop Sb'p15 :1408-' 08/16/02 1-10/27/83 1.8.80

______~~~~ Infted-U'V~rticITnos____

n_

I . 1., V32 FieId z1462 011/09/84 12/1/1981 - 2.11 1 1', V32= S'hop- 1338 -01/09/84. 12/1/1981 2.11 1 " V43-,. ':Field' 1517 01/116/84 -10/2/1981 -2.29 1 "

1.. .V43 Shop 1391 01/17184 10/2/1981' 2.29 1  !. 1- V62 Field 1468 ' 01/12/84 11/25/1981 2.13 1- 1 V V62 -,Shop 1453-- 01/12/84 11/25/1981 -- 2.13

.1 ,.1 ,V75 Field 1468 '-:01/10/84 '11/24/1981 2.1-3 1 1 V75 -,Shop 1438- 01/10/84 1-1/24/1981 2.13

1. .3 .V02 'Field 1440 "02/16/86- 10/19/1981 - 4.33 1 3 ': V02&. Shop; -:1495-!- :02/15/86 10/19/-1981 - 4.33 1 .3. V18 Field~ 1f355 02/1.2/86 10/7/1981 4.35 1 3 " V18 *Shop' 1473, 02/09/86 10/7/1981 . 4.34' 1 3 V55. 'Field '14:62- 02/13/86 10/9/1981 4.35 1 3 V55 Shop 1445 A02/110/86 10/9/1981 -4.34 1 3 V75 Field .- 14:71 02/13/86 11/24/1981 , -4.22 1 3 V75 . Shopi 1474 02/07/86 -11/24/1981- 4.21 1 5 ,V,1 Field 1.477 02/26/88 -10/14/1981 - 6.37 1 5 V1.1 ,Shop, 1.452 "'02/26/88 10/14/1981 6.37

- 1 5- V36, Field 1'515 .02/26/88- 12[1/1981 6.24 1 5 " V36 Shop*. 11363 - 02/29/88- -12/1/1981 - 6.25 Palo Verde Nuclear Generating Station .Page 4.5-8 License Renewal Application Amendment 15

Section 4 TIME-LIMITED AGING ANALYSES Table 4.5-1 Tendon Regression Analysis Input Data For PVNGS Uinit 1,2, and3 .' ..

Unt. Year Tendon .

.~ .I ~ -~Time at

'Force, Date - Tension.

_ _un,_ . _ _ -_ _ (Years)-,

1 -5 V75 -Field -1454 02/26/88 11/24/1981- 6.26)

.1 5 --. V75 -Shoo; -. 1468 02/29/88 -11/24/1981 - 6.26

_1 5. V86 Field-, 1474 02/29/88 7-1/f011981 6.64 1 5 V86 Shop 1492-- 03/01/88.I 7/10/1981.- 6.64 1-, 10' V40 Field- 1533- -09/*1892 -'11/23/1981 10.82 1 0 V40 Shop 1538.,i 09/18/92_ 1/23/1981 -10.82

1. 10 V53 Field .14-16--, 0ý9/15/92. 115198 -10.86 1 10, V53  :. hopp -1390-' 09/17/92 - 11/5;1981 - 10.87

-1 10 -V75 Field .1442 0902/92 -11/24/198.1 10.77

- 10 V7-5 Shp: -461 '09,14/i92* 11/24./1981 -10.81 1 -1.5 V37 Field- 1434.' -07/24/98 -.11,/30/1981 16.65 1 -15 V37 Shop 186. 07/24/98 .1-1/30/1981 16.65 1 15 V72 Field -1402-1 07!17/.98 - 12/2,1981-- 16.62 1 1.5 V72 Shop_ 1390

• 07/22/98 12/2/1981 16.64 1 -15 . V75 ---- Field. ,4487 07/21/98-' 1,1241981 16.65

-. -15 .V75 Shop_ !409 07W/116/98 [-11/24/1981 16.64 1' . V8' Field .,1385', .,02/22*/088MZ 1,1/09/1981 26.29.

1 - 25 V8 .. _ShOp L-- 1 4 05 02/22/08 ý6.29 1 VS 2. Fild 132 0/2508 10/1 4/1981 -26.37 111ý 25> V

-

S1 h __ 1403.,' 042/25/08 10141 981 3/42~7 -

_ 25 I V . Field- - 1421- 02/25/08, 'J1124/1981 26:25, 17 S2hVp:: ;__i! 42,2 - '*f02/22/08 14/24/1'981 2-24:

2 20- V26-, "'Feid - 1380 Q9/08/05 8/13/1982 _ 23.07 2 --*20 V26 - Shop' -1298 09/08/O5: 8/13/1982-, 23.07

,2 20' V67, Field 1:338 '08/31/05 8/16/1982 23.04 2 20 V67 ',.ShopL -1479 1'08/30/05 - 8/16/1982 23.04

2 20 . V75 Field 14,30 09/01/05 8/16/1982 23.02

- 2 - 20- V75 -Shop- :140. 09/15/05 8/24/1982. 23.06 3 T-. V07 -Shp .1,520; `.11/03/87. 12/19/1983 3.87 3 1 V1 ,iS .Shop .1491 . 11/03/87- 1/17/1984 3.79 3 - - V16 "Field 1,558 "1'0/29/87, 2/3/1984 3.73 3 1.. V16 S.hop -14.04 -10/26/87: 213/1984 3.73 T1-. V20 -Field -, 1486. 12/,10/87 2/3/1984 3.85 3 -1 - V20 Shop 1574* 12/10/87 2/3/1984 3.85 3 1  : ',:V28--' ..-:Field. 1.485 1 12/03/87' 2/9/1984 - 3.81 3 .1 V28,.: Shop' 1421 12/03/87' 2/9/1984 3.81 Palo Verde Nuclear Generating Station Page 4.5-9 License Renewal Application Amendment 15

Section 4 TIME-LIMITED AGING ANALYSES Table 4.5-1 Tendon Regression Analysis Input Data For PVNGS Units 1, 2, and 3

-Year Tendon Fbrce, Time at Date enslion: o Tesi unit .0) (2)- n Kips, Date:-. i e rs) 3 1 V49 Field' 15.27 12/04/87 1/20/1984 3.87 3 1 V49 "Shop 1353 1:.0/24/87 1/20/1984 3.76 3 ' 3 V16 Field 1,550 "11/01/89 2/3/1984. 5.74 3 3 V16 . Shol5 1,379 10/30/89 2/3/1984 5.74 33 3 V39 'Field, 1471 11/01/89 12/20/1983 5.87 3 3 V39 Shop 1:398 11/01/89 12/20/1983 5.87 3 13 V59, Field 1485 10/31/89 1/16/1984 5.79 3- 3 V59 Shop 1.387 11/01/89 1/16/1984 5.79 3 -. :'3 V66 Field '1481 10/31/89 2/7/1984 5.73 3 3 V66 Shop 1435 11/02/89 2/7/1984 5.74 3 , 5 V16 Field !1543 09/11/91 2/3/1984 7.60 3 5 V16 Shop 1410 09/13/91 2/3/1984 7.61 3 5 V33; Field 1598 09/18/91 2/16/1984 7.59 3 5 V33 Shop 1456 09/18/91 2/16/1984 7.59

3. ' 5 V48 Field .1445 09/11/91 12/22/1983 7.72 3 5 V48 Shop .1,293 09/12/91 12/22/1983 7.72

'3 5 V71 Field 11558 09/10/91 2/9/1984 7.58 33 .5 V71 Shop 1468 09/12/91 2/9/1984 7.59 3i '-V10 V13 Field 11507 08/01/96* 2/6/1984 12.48 3 10 V13! Shop 1405 08/14/96 2/6/1984 12.52 3 :10 V16 Field 1,524 08/02/96 2/3/1984 12.50 3 10 V16 Shop 1366 07/31/96 2/3/1984 12.49 3 .10 V82 Field 1478 08/02/96 12/22/1983 12.61 3 ,10 V82, Shop '1394 08/02/96 12/22/1983 12.61 3 15 V16 Field 1'517 07/11/02 2/3/1984 18.43 3 15 V16, Shop .11,367 07/11/02 2/3/1984 18.43 3 '15 V41 Field :1542 07/11/02 1/24/1984 18.46 3 , 15 V41 Shop 1419 07/16/02 1/24/1984 18.48 3 15 V57 Fieid 1520 07/16/02 2/7/1984 18.44 3 _15 V57 Shop 1378 07/10/02 2/7/1984 18.42 1 Nominal, post- SIT (Structural Integritv Test).

2 Boldface numbers are "common" tendons, examined at each tendon prestress surveillance.

Palo Verde Nuclear Generating Station Page 4.5-10

!License Renewal Application Amendment 15

Section 4 TIME-UMITED AGING ANALYSES 1600 A - -

-egression Apa!ysis Predicted Prestress y -48 703Log(t) + 1486.7 A  ; -.0 1195 1560 a

a A

Reg Guide 1.35.1 PredictedPrestress

a. 14-do -A- y = .44.772Log(t) +1422.2 ChartLeciend L0 '

A. asured Tendon Prestre~s U) 1300 4 -- _____

'-.1< -Mii'mum ir qi,iired Valu Wall Minimum Required.Value *MRV) =:1248 uip (MRV)

- - - -- --- - --- -- -----. 4 1200 r-"-RG 1;.35 1 Predicted Prestress Dome Minimum Required Value (MRV) = 1167 ki

- - -- - -- - - - -

- Regression -Analysis Predicted "Prestress

-. 1100

-.0 2'

. , I I

• ', 10 "" 40'. 60-! -100 I

-4 Iog(Time), Tim'e (Years)

- I Figure 4.5-1: Regression Analysis of Unit I Horizontal Tendon Lift-off Data Through the 425-Year Surveillance; With Reg Guide 1.35.1 PredictedPrestress (See Notes)

Palo Verde Nuclear Generating Station Page 4.5-11 License Renewal Application Amendment 15

Section 4 TIME-LIMITED AGING ANALYSES 1600 Regression Analysis Predicted Prestress A y = -42.233Log(t) + 1475.7 Fe = 0.1656 1500

.1400 RG,1.35.1 PredictedPrestress y = -24.08Log(t) + 1426.5 0.

. - ChartLeqend' 1300 0)

LL A Measured Tendon Prestress

'0 t-1200 -Regression Analysis Predicted P re s ire s s - . .; : .,., -

- - - RG 1.35.1 Predicted Prestress 1100

.. Minimum Required Value (MRV)

Minimum Required Value (MRV) = 1016 kip F --

- -- I - -- ---- ý- - -- -- -- - - -- -- - -- - ---

1000 0 2

'-'1 ~ 10 40 60 100 Time (Years - log scale)

Figure 4.5-2: Regression Analysis of Unit I Vertical Tendon Lift-off Data Through the 425-Year Surveillance; With Reg Guide 1.35.1 PredictedPrestress (See Notes)

Palo Verde Nuclear Generating Station Page 4.5-12 License Renewal Application Amendment 15

Section 4 TIME-LIMITED AGING ANALYSES 1600 A

Regression Analysis Predicted Prestress y = -85.029Log(t) + 1495.4 A 2 A

R = 0.2297 1500 A

RG.1.35,1 Predicted Prestress...

EL 1400 y = -55.586Log(t) + 1437.7 -

A 0.

U.

0 U.

A Measured Tendon Prestress 1300 Wall Minimum Required Value (MRV) = 1248 kip

....- MRV 1200

- -- RG1].35.1 Predicted Prestress Dome Minimum Required Value (MRV) = 1167 kip

-Regression Analysis Predicted

%>.Prestress':"

1100 0 2 1 10 40 60 100 Time (Years in log scale)

Figure 4.5-5: Joint Regression Analysis of Unit 1, 2, and 3 Horizontal Tendon Lift-off Data Through the Unit 12 250-Year Surveillance; With Reg Guide 1.35.1 PredictedPrestress (See Notes)

Palo Verde Nuclear Generating Station Page 4.5-15 License Renewal Application Amendment 15

Section 4 TIME-LIMITED AGING ANALYSES-1600 ___- " ".._ _ _"°. __ __,_ -- , -______-.

Regression Analysis Predicted Pyestress 1500~ :" ~~ ~ ,*" ~ ~

'.{- ~ . ~~R yY -56.041Ldg(t) 0.1813+ 1497.6

)-

1500-1400 -A ARG 1.35.1 Predicted Prestress y -26.678Log(t) + 1441.4 C1300 Chart Le-end 0

. Measured Tendon Prestess..

a 1200

."".- Minimum Required Value (MRV) 1100 ' " -.. - -*-RG 1.35.i Predicted Prestress Minimum Required Value tMRV) 1016 kip .. Regression Analysis Predicted 1000 "" ". "" Prestress" -

" .___ _____,___._

0 12 1 10 40 60 100 "

Time (Years in log scale)

Figure 4.5-6: Joint Regression Analysis of Unit 1, 2, and 3 Vertical Tendon Lift-off Data Through the Unit 1-2 250-Year Surveillance; With Reg Guide 1.35.1 PredictedPrestress (See Notes)

Palo Verde Nuclear Generating Station "Page 4.5-16 License Renewal Application Amendment 15

Section 4 TIME-LIMITED AGING ANALYSES Notes to Figures:

a) Since only a single, 20-year inspection has been performed on the Unit 2 tendons, insufficient data were available fo.r a regression analysis of Unit 2 tendons alone. The Unit 2 20-year data were included in the jointregression analysis of all three units.

b) "Trend line" and "minimumýrequired value" (MRV) are equivalent to the same terms in NUREG-1800 and NUREG-1801 Chapter 4,-a6d X.S1.- i*

c) The time scale and variable t are actual years from original tensioning date.

d) The original predicted loss of prestress values used by the surveillance program were calculated separately for each tendon in the surveillance sample, using methods consistent with Proposed'Reg Guide 1.35.1 Rev. 0. These lines were calculated "per wire" because they are used to calculate a predicted force per tendon, based on the actual number of wires, in order to compare-the predicted. fo.rce.wit.h the surveillance lift-off force. The new predicted force lines ("predicted losses") shown on these plots are average values, recalculated.us ing Reg Guide 1.35.1 methods. The values plotted here assume 186 wires per tendon. Some tendons have fewer due to failure to meet acceptance criteria at installation, or due to removal for surveillance testing.

e) The dome and cylinder horizontal hoop tendons are sampled and tested as a single group.: The predicted force lines were developed from data for the entire horizontal population. The 'regression lines were similarly developed from surveillance data of a sample that includes both dome and hoop tendons. However the dome tendons have a lower MRV. Bothl MRVs are shown on the horizontal tendon plots.

f) The Unit 1,25 year SurFeillanco waS not com~ploto in time ýtO be included in this analysis.

g) Each of the trend lines is. a regression analysis of the entire set (or stated subset) of individual tendon lift-off data for each of the tendon groups, "'The trend lines were calculated from the actual tendon Iift-off forces, regardless of the number of effective wires -per tendon.

Palo Verde Nuclear Generating Station Page 4.5-17 License Renewal Application Amendment 15

Appendix B AGING MANAGEMENT PROGRAMS A1.32 STRUCTURES MONITORING PROGRAM The Structures Monitoring Program manages the cracking, loss of material, and change in material properties by monitoring the condition of structures and structural. supports that are within the scope of license renewal. The Structures Monitoring Program implements the requirements of 10 CFR 50.65 and is consistent with .the guidance of NUMARC 93-01, Revision 2"and Regulatory Guide 1.160, Revision 2. ,

The Structures Monitoring Program provides inspection- guidelines, for concrete elements, structural steel, masonry walls, structural features (e.gi, caulking, sealants, roofs, etc.),

structural supports, and miscellaneous components such as doors. The Structures Monitoring Program includes all masonry wallsand Water*control structures within the scope of license renewal.: The, Structures :,Monitoring Program also monitors settlement for each major structure and; inspects supports for equipment, piping, conduit, cable tray, HVAC, and instrument components.

Prior to the period of extended operation:

The Structures Mqnitorinig Program; will.!: be enhanced to specify AC 349.3R-96 as the reference for qualification: of persbnnel to ins pect structuresý under the Structures Monitoring Program.

For structures withi'n the scope of' license- re'newal Ttlhe Structures.Monitoring Program will be enhanced to establish the freqgenc,' of irnspection for. each unit at a 5 year interval, with the exception of exterior surfaces of the following nonsafety-related structures, below-grade structures, and structures within a controlled interior environment, which will be inspected at an interval of 10 years:" ;pr*ctupro,,

np t within tho *cp of liconso renewal within A 4110

• Fire Pump House (Yard Structuresi

• Radwaste Building .

" Station Blackout Generator Structures

• Turbine Buildinq . -

• Non-Safety Reilted Tank Foundations and;Shells.

• Non-Safety Related Transformer Foundations and:Electrical Structures The Structures Monitoring Program will be enhanced to guantify the acceptance criteria and critical parameters for monitoring degradation, *and to provide -cquidance for identifying unacceptable conditions requiring further technical evaluation or corrective action.

Procedures will also be enhanced to incorborate applicable indu.stry codes, standards and guidelines for acceptance criteria. defi-e the specific critoria for .8t9g.riZing dofiGio*nGRS for conrGete inspections.

Palo Verde Nuclear Generating Station Page A-18 License Renewal Application Amendment 15

Appendix B AGING. MANAGEMENT PROGRtAMS Table A4-1 License Renewal Commitments Item Comnmitment-  ;,4R Section Implementation No: ~.$chedule.

23 APS will: A1.21 - Not less than 24 A. Reactor Coolant System Nickel Alloy Pressure Boundary Components B2.1.21 months prior to the Implernment applicable (1) NRC Orders, Bulletins and .Generic Letters Reactor Coolant period of extended associated with nickel alloys and (2) staff-accepted industry guidelines, (3) System Supplement operation1 .

participate in the industry initiatives, such as owners group programs and the EPRI Materjals Reliability Program, for managing aging effects associated with 3.1.2.2.16.2 nickel alloys- (4) upon completion of these programs, but not less than 24 - ' Pressurizer spray months before entering the period of extended operation, APS:will submit an' head-crackinq inspection: plan for reactor coolant system .niclkel alloy pressure boundary components to theNRC for review and approval 7and B. Rea6to-r Vessel Inte rn als (1) Participate in the industry programs for investigating and managing aging effects on reactor internals;, (2) evaluate and -implem-ent the results of the industry programs as applicable to the reactor.i nternals; and"(3) upon completion of these programs, but not less thean 24 months before enterinq the period of extended operation, APS will submit an inspection plan for reactor -

internals to the NRC for review and approval.

C. Pressurizer Spray Heads ComplV With~applicable NRC Orders and implement applicable (1)-Bulletins and.,

Generic Letters, and (2) -staff-accepted industry quilelines. "

(RCTSA"s 3246912 [Ul]; 3247274 [U2],"3247276 [U3]) -[

Palo Verde Nuclear Generating Station Page A-50 License Renewal Application Amendment 15

Appendix B AGING MANAGEMENT PROGRAMS Thhle A4-1 License Renewal Commitments Itýe - Comrfitrne'nf LRA Section' Implementation-Schedule 34 Existing Structures Monitoring Program is credited for license renewal, AND A1.32 Prior to the period of Prior to the period of extended operation: B2.1.32 extended operation'.

" The Structures Monitoring Program will be enhanced to specify ACI 349.3R-96 as Structures the reference for qualification of personnel to inspect structures under the Monitoring Structures Monitoring Program. Program

• For structures within the scope of license renewal. _Tthe Structures Monitoring Program will be enhanced to establish the frequency of inspection for each unit at a 5 year interval, with the exception of exterior surfaces of the following nonsafetv-related structures, below-grade structures, and structures within a controlled interior environment, which will be inspected at an interval of 10 years: inspeEt 1 t *LI_ *._ L Gtlruciuroct IoWiRtiAnRn tno o pn or !iicn no re*nAt;al wnnin. n a it) yers peFriu

" r-ire rump nouse Taru OtruCturesj

" Radwaste Buildinq

• Station Blackout Generator Structures -

0 Turbine Building

• Non-Safety Related Tank Foundations and Shells

• Non-Safety Related Transformer Foundations and Electrical Structures

.T TheStructuLr'es Monitoring Prograri-will be enhanced tdcquahtify the acceptance

'" critbeia and critical parameters for monitoring degradation, and to provide guidance for identifying unacceptable conditions requiring further technical

evaluation or~corrective action. Procedures will also be enhanced to incorporate

• applicable industry codes, standards and guidelines for acceptance criteria. defiRe

.Ve-Aaeoii ~eafFntpoAgdf~R~H G ERrt

" (RCTSAI 3246927)

Palo Verde Nuclear Generating Station PageA-52 '

License Renewal Application Amendment 15

Appendix B AGING MANAGEMENT PROGRAMS Table A4-1. License Renewal Commitments Ite 7 omitmntLRA Section Implementto No!. Schedule 53 The changes to LRA.Table 4.5:1 shown in the response to..-RAI 4.5-1, and Response to RAI 4.5-1 512RI10 changes to LRAFigures4.5-1, -2, -5, and -6to reflect the results of the - in APS letter no. Completed Unit 1,25-year tendbn surveiilance data regression analysiswill be .102-06160, dated submitted tothe NRC inan LRA amendment by May 28, 2010.. 4/1/10.

Completed 3 3 .

___(RCTSAI 3429933) .- __________ _________

• ,,. %*

a.

........................

Palo Verde Nuclear Generating Station - .Page A.,59.

License Renewal Application Amendment 15

.? , , ,Appendix B AGING MANAGEMENT PROGRAMS B2.1.26 Inaccessible Medium Voltage Cables Not Subject to 10 CFR 50.49

.Environmental Qualification Requirements Program Description The Inaccessible Medium Voltage Cables Not Subject to 10 CFR 50.49 EQ Requirements program manages localized ,damage,anrd'breakdown. qf insulation leading to electrical failure in inaccessible medium voltage cables exposed to:.adverse localized environments caused by :significant- moisture, (moisture ,that,.lasts more than a few days) simultaneously with significant voltage (energized greater than 25% of. the time) to ensure.that -inaccessible medium voltage cables not subject to. the environmental qualification' (EQ) requirementsof 10 CFR 50249 and within the scoe oY lcense renewal are capable*of performing their itended functiion'. ' AK" -

All cable' manholes "that 6dntain" in'ýscop'6 non'EQ*1naccessible mediurmi voltage cables will be inspected for water collection. The collected water will be removed as ."required. This inspection andwater, removal, will Ibe,.peprformed based,,.on actual plant experience but at least every two years... ' . .

All in-scope non-EQ naccessible me-dium voltage,6ablesrouted thr6iUgh manholes Will be tested to provide an. indication of the.'condukctorrinsuiation _conditior. A polarization index test as described in EPRI.TR-1.0334-1-2 oother ttien'gthat is state-of-the-art at the time of the testing willbe performed at least .once every ten years The first test w be competed the period of extended bhefore~ inc" tp~rt Willn* e.ri'ft The acceptance'cnrteria foreaCh testwi be defined for the specific type of test performed and the specdfic cable tsted. Peridic inspections of cable manholes for the'accumulation of water will' minimize cabl eposue 'to water.' Corrective actions Tfr donditions that are adverseto quality are, perfformed ina,ccdrdan6e'with the corrective action program as part of the QA program- The obrrective action ipogram proides reasonable-assurance that deficiencies" adverse -to q'uality-are either nprorimptly-corrected or are, evaluated to be acceptable. - . .-

Procedures will implement the aging: management--program -for- testing .ofthe medium voltage cables,-not- subject' to;-1 0 CFR 50.49 'EQ requiremerits:an'd the periodic inspections and removal of water from- the cable manholes containing in-scope medium -voltage cables not subject to 10 CFR 50.49 EQ requirements. ...

- -

NUREG-1801 Consistency .* ,.

The Inaccessible Medium Voltage Cables Not Subject to 10 CFR 50.49 EQ.:Requirements program is a new program that, when implemented, will .be consistent with NUREG-1801,Section XI.E3, "Inaccessible Medium V61tage Cables Not Sub.ject to 10 CFR 50.49 Environmental Qualification Requirements"'.

Palo Verde Nuclear Generation Station " Page B-76 License Renewal Application Amendment 15

. Appendix B AGING MANAGEMENT PROGRAMS Exceptions to NUREG-1801 None i 4' r'. .

Enhancements None Operating Experience. - 0 Indust-y operating experience has "sh6wn' that' crbss linked .pdlyethylene ýor-* high molecular weight polyethylene insulati nr rMaterial;,, exposed ýto sig'ificant moistu.e and voltage, ýare most susceptible t0lWater tree ýf~riration. Formation' and growthý of,.watei trees varies directly With operating voltage..

PVNGS has not. expe rienced a failure of. any.. inaccessible. medium voltage .cabl'es.

PVNGS has experienced cases where medium violtage cable splices .have been subjected to water intrusion resulting in low megger readings. In all cases the splices were reworked. In addition,:in one case the splice was moved.to a-manhole less subject to water intrusion. '".  : . , -.

During manhole walkdowns in"2009,'one stdtiri 'blacko~t cablehnianhble vIWas found to contain water submerging the cables. Subsequent inspection of a connected marihol&e*

found additional water, but no additional submerged cables. A review of the history of these awnd n manholesfownid .ecurringinstariice ;s oWtewiritrus'n. ih W mhcoaine fwuned to ,as eplae;V4Fipdr&6hang to.soed'aboe gr and the manho is addeodredAto rLteo.wti:aaYmghhAlf. ihmany tiHnoAa tlhesru IS a rainaccu~~~~~~~~~'u~lation of gro66tor th~h3in; 6i 1hdrad&r onh 4Ii beon fourd dn,qfpri tWo P Maoholouinspectio metea. f maximum fivo year ,yept dmabhole fouad t.

cotinae \vtsuibmerged Qab~les ýhaý had a,seol'repla-ed' id, raisý!ed aoeOae, and the ground surface reworked to route w~i~~,ate' ay froM the.manhole.'. As.'a result'of~

this operating -experiente, APS moved these 'nnae*(M oqa ,ro mosthfrequmnh t inspection interval. The preventioe mainte'nance(PM)'br h wrin the

'manholes into three freqluencies of inspections based'on their historvobf waterintrUsibri: two weeks. six months, and two years. The two-week P'M task requirem inspection of rrfmnholes if it has rained 0.3 inches or more within a 24-hour period since the last time the PM was performed. The manholes grouped into this P.M task are also inspected on a six-month frequencyvto ensure they are.always i.nspected even during.dry periods. Under the PM program, a manrhole will .not be moved. to th6 two'vear reduced frequency inspection interval until it has been found dry for two years. , .

Industry and plant-specific operating experience will be evaluated in the development and implementation of this program.

Conclusion .

.The implementation 'of Iriacd*essible' Medium Voltage Cables Not Subject to

'10 CFR 50.49 EQ Requireime'nts program will provide reasonable assurance that aging effects will be managed so that the intended fudnctions of thie*inaccessible medium voltage cables within the scope of license renewal are maintained during the period of extended operation.

Palo Verde Nuclear Generation Station Page B-77 License Renewal Application Amendment 15

. .Appendix B AGING MANAGEMENT PROGRAMS

,B2.1.32:, '- Structures Monitoring Program, Program Description The Structures Monitoring Program manages cracking, loss of material, and change in material properties by monitoring the condition of structures and structural supports that are within the scope of license renewal. The program implements the requirements of 10 CFR 50.65 (Maintenance Rule) and is consistent with the guidance of NUMARO 93-01, Rev 2 and Regulatory Guide- 1.160, Rev. 2. The Structures Monitoring Program provides inspection guidelines', -,nd.;'waikdown"'WcFieckists,-f6r cbncrete :elements, structural steel, masonry walls, structural features (e.g. caulking, sealants, roofs, :efc.), structural sLipports, and miscellaneous components such as doors. The scope of .the Structures Monitoring Program includes all masonry walls and water-control §tructhres within the scope of license renewal.. The program also monitors, settlement for each major structure and inspects su p rtS for equipmentf ping .c6ridiit cable tray'. HV d

." instrument cdmponents. The scpeorfo equipment, pihg otray, HAC and scope of the Structures Mont onn'Program does not include the inspection of the supports specifically inspected per the requirements of the ASME Section XI In-Service Irispection Program. Though coatings may. have been applied to the external-surfaces of.structural members, no crýeditwas taken, for these coatings in the determination of aging effects for the underlying materials;, The Structures :Mdnitoringoprgram evaluates the condition of the coatings as an' indication-of the condition of the underlying materials.

Periodic ihSriecti6ris' r6qu'iredbythke' Str'uctures Monito'ring Program are performed and documented per plant procedures. Initial baseline inspections under tthe Structures Monitoring Program were performed from June 1994.to June 1996. Each of the spray ponds is inspected every five years, and settlement-moiritoring surveillance is performed for each major structure, consistent with UFSAR 2.5.4.13, every five" years. Prior to the period of extended operation, for structures withirh "the;cp -icei~e renewal, Tthe Structures Monitoring Program will be enhanced to. establish the frequency of inspection for each unit at a 5 year interval, with the exception of-exterior surfaces of the followihiq nonsafety-related structures, below-,rade structures, 6nd" structures'withii a 'controlled interio'r environment, which will be inspected at aninterval of 10 years. incpoc ... t..

cuc. Within the*c.... f licence Frenwal within a 10 years poriod Fire Pump House (Yard Stru6tures '

• ; Radwaste Buildinq .. .

• Station Blackout Genierator"Structures - ...

Turbihe Building .. .

• Non-Safety Related Tank Foundations and Shells .

• Non-Safety Related Transformer Foundations and Electrical "Structures Palo Verde Nuclear Generating Station . . ' Page B-92 License Renewal Application Amendment 15

Appendix B AGING MANAGEMENT PROGRAMS NUREG-1801 Consistency The Structures Monitoring Program is an existing program tlat, following enhancement, will be consistent with NUREG-1801, Section Xl.S6, "Structures Monitoring Program".

Exceptions to NUREG-1801 None . ",  :;c.,c . ,

Enhancements -

Prior to the'period of extended operation, tte following eppha.?cerents wi!] beimplemented in the following program elements:. . . .. ' -.

Detection, of Aging Effects - Element , .

The Structures Monitoring Progr6am ,Vi.[. be ehha nced to specify AC.I 349.3R-96 as the referende for qualification of persorin1e't6' insp;bt st'ru'cturees:Uhder the Structures Monitoring Programn. ~ r~-

For structures witin tle' scope of licerise renwa,"Tthe Siru&tires Monitdring Program will be enhanced to establish the frdquencV of inspc'rticn-'for eci'h:uriit" t a 51v/ear interval, with the exception of exterior surfaces of theb foll6win6 no'nsafety-r6lated 'structures, below-grade structures, and structures within a cont6'lled interior ehkirýnment Vhich wir[ be irspected at an interval of 10 years: inspcctF.,Gt .... .. itnhFpo Ren.Wal Within a. 10 yearG pFe~d 0 Fire Pump House (YardStructures) , -, "

0 Radwaste Building .

• Station Blackout Generator Structures "

STurbine Building, ..

S ,Non-Safety RelatedfTank Frundations:and Shells -

'.Non-Safety Related Transformer Foundations and Electrical Structures Acceptance Criteria- Element 6 The. Structures Monitoring Program will be enhanced to guantify the acceptance criteria and critical parameters for monitoringq degradation';: and -.to provide guidance for identifying unacceptable conditions requiring further technical evaluation or corrective' action.

Procedures will also be enhanced to incorporate applicable industry codes, standards and guidelines for acceptance criteria. dofino the.pecific criteria for catogor:izing deficiencie. fo Palo Verde Nuclear Generating Station Page B-93 License Renewal Application Amendment 15

Appendix B AGING MANAGEMENT PROGRAMS Operating Experience Miscellaneous openings and gaps in barriers that-may impact the environmental equipment qualifications at PVNGS were reviewed and all identified deficiencies were corrected in accordance with NRC Information' Notice -IN"95-52 "Barrier and Seals between Harsh Environments".

NRC Information Notice IN 2002-12 "Submerged Safety-Related Electrical Cables" identified several failures and weaknesses associated with protracted submergence in water of electrical cables that feed safety-related equipment. Significant amounts of water have been found ihnvarious manholes an'd the' entryts from an. unknown source. 'The intrusionof water into the manholes is being effectively pontrdolied through a pumping program.

NRC Informat ion Notice IN. 2003-08""Potehtiai Flooding through Jnsealed Concrete Floor Cracks" identified failures involving, flooding of rooms containing safety-related panels and equipment as a result of fire water seepage through ,unsealecd concrete floor cracks. No through cracking has been identified at PVNGS and the program has been revised to provide guidance for the identification of through wall cracks in flood barriers in the future.

NRC Information Notice IN 2005-11, "Internal Flooding/Spray-Down of Safety-Related Equipment Due to Unsealed Equipment Hatch Floor Plugs and/or Blocked Floor Drains" identified the possibility of flooding safety-related equipment as a result of.(1) equipment hatch floor plugs that are not water tight and (2) blockage of equipment floor drain systems that are credited to mitigate the effects of flooding. All hatches/plugs that are credited as flood barriers are water tight. Instructions were developed to provide removal and reinstallation instructions for hatches and plugs to maintain the required seals.

Adverse and critical conditions were found on the roof of Unit l's Turbine Building. These conditions included punctured membrane and rigid insulation, deteriorated tar patches with mesh reinforcement exposed, damaged flashing exposing the roof membrane seal, raised blisters/raised areas in the membrane, several long areas of damaged flashing, and large cracks through the roof membrane into the rigid insulation. The large cracks and large blister/raised areas in the roof membrane are significant leakage paths and classify the condition of the roof at Elevation 240' as critical. A previously issued CRDR addressed the concern that in inclement weather the Turbine building had experienced consistent and dependable flooding, which had caused equipment failure. To address these concerns Unit l's Turbine Building roof was replaced. Unit 2 and 3's Turbine Building roofs have been previously replaced.

Conclusion The continued implementation of the Structures Monitoring Program provides reasonable assurance that aging effects will be managed such that the systems and components within the scope of this program will continue to perform their intended functions consistent with the current licensing basis for the period of extended operation.

Palo Verde Nuclear Generating Station Page B-94 License Renewal Application Amendment 15 r

ENCLOSURE 4 ConStruction Technology .Laboratores,(CTL) Report, October 18-20, 2005 Site Visit, Examination of Spent FuelPoolConcrete,Walls inB Unitr1,.;,

CTLGrOup Project No:.059084#

It I

CT GROUP Building Knowledge. DOlivering Resultsý CONSTRUCTION

.. TECHNOLOGY LABORATORIES.

ENGINEERS & CONSTRUCTION November 18,2005. TECHNOLOGY CONSULTANTS www.CTLGroup.com via email: wjones04@apsc.com Mr. Warren Jones Arizona Public Service, Company .

Palo Verde Nuclear Generating Station 5801 South Wintersberg Road Tonopah, AZ 85354-7529 Tel. (623) 393-5000 October 18-20, 2005 Site Visit Examination of Spent Fuel Pool Concrete Walls in Unit I CTLGroup Project No: 059084,

Dear Mr. Jones:

As authorized by the Arizona Public Service Company (APS), CTLGroup Inc. conducted an evaluation of the spent fuel.pool walls at;the above referenced project. Messrs George Seegebrecht, Dr:.! Mohamad Nagi and Jerry Harapo of CTLGroup conducted visualobservations and nondestructive fieldctesting on Octoberl,8-20, 2005.,

It was reported that the spent fuel pool walls are 5-ft thick, stainless steel-lined reinforced concrete. The main area of the wall that has been evaluated is at the 100 ft elevation. The owner has noticed water leakage at a previous date near the middle of this area.

The objective of the site visit was to visually and nondestructively examine spent fuel pool wall surfaces to evaluate its current condition for.,any signs of distress, such as cracking or.poor consolidation'. In addition to visualobservations made of the wall interior surfaces, the following nondestructive techniqueswereautilized: .

• Ground Penetrating Radar (GPR)

• ..,.Impulse,Response (IR)

Ultrasonic Pulse Velocity (UPV)

On the last day of our site visit, we m't with APS personnel including Palo Verde Director of Nuclear Engineering to discuss our observations, preliminary test results, and general recommendations. ,

FIELD OBSERVATIONS The field observations were limited to wall sections of Unit 1. The visual observations were limited to cracking and other surface features. Photographs of typical concrete surface Main Office 5400 Old Orchard Road Skokie, Illinois 60077-1030 Phone 847-965-7500 Fax 847-965-6541 Noitheast Office 5565 Sterrett Place6, Suite 3:12 Columbia, Maryland 21044-2685 Phone 410-997-0400 Fax 410-997-8480

Mr. Warren Jones, Palo Verde Nuclear Generating Station Page 2 of 7 CTLGroup Project No. 059084 November 18, 2005 conditions are presented in Figures 1-14 in-Appendix A. Whenever possible, crack widths were measured and recorded on the wall surfaces. In general, the crack widths in the vicinity of the reported leakage ranged from 0.010 to 0.016 inches (see Figure 3 in Appendix A). Cracking patterns were predominately vertical but numerous horizontal cracks were also common.,

Based on our experience, the observed cracking patterns may be indicative of the following causes:

. Crazing, which commonly occurs shortly after form removal

• Thermal, which is not uncommon for such massive concrete (5 fttb 8 ft thick) .

• Drying shrinkage

" Alkali-silica reaction (ASR) (in isolated areas)

All concrete surfaces appear to have been coated with a clear surface coating, possibly polyurethane or similar. -

The initially discussed visual observations of Units 2 and 3 were not codnducted due'to time constraints during the visit. If desired, visual observations and nondestructive testing can be conducted for comparison purposes at a later date.

NONDESTRUCTIVE TESTING Nondestructive testing(NDT) was utilized to asseýss concrete integrity in'"1ocations as requelsted by Palo Verde Nuclear Generating Station pers6nnel.!, NDT methods sUch as Ultra-Sonic&Pulse Velocity (UPV), Impulse Radar and lmpulse Response testing Were employed-to assess.

concrete integrity. Severity of concrete cracking in locations adjacent to recent reported leakage, were assessed using UPV. Radar"andimpdIse resp~onse were utilized.t6determine if any significant subsurface cracks, delarminatiobn or voiding was detectable in the test areag. We collected field data in five (5) wall areas from the 6ut0s~ide face'of the sp'erit fuel p0o1o 'of Unit 1.

These five areas are described as follows:

• Area One:~ This area at~elevation 100ft. measured"approximately 20'ft. wide by eight:

ft. higllstartibg from th6 west edge"of the spent fuel pool southern exterior wall. A 2-ft. test grid was established. the test grid was started 1 ft. from the west wall edge and 2 ft. above the floor elevation.

• Area Two: Area two also at elevation 100 ft. measured 14 ft. wide by 16 ft. high. This area is adjacent to area 1 beginning 23 ft. from the west edge of the spent fuel pool southern exterior wall and 1 ft. above the floor elevation a 2-ft. test grid was established in area 2. The 1-ft. test grid was chosen in this location since it included one-of the reported leak locations.

• Area Three: This area at elevation 100 ft. measures 22 ft. by 4 ft. high-and begins 37 ft.

east from the west edge of the spent fuel pool southern exterior wall. A 1-ft.

test grid was established in this area.

C GROUP 8dilding Knov*edge.Devering Results.. vww.CTLGro0Lp.co01T

Mr. Warren Jones,. Palo Verde Nuclear Generating Station . Page 3 of 7, CTLGroup Project No. 059084 November 18, 2005

.. Area Four: .,A test area was examine~d at elevation 120 ,.-of spent the fuel poo!

southern exterior wall.. A2- foot test grid. was established and tested using ImpulseResponse.,The area tested measured 8 ft. wide by 6 ft. high.

Area Five: An exterior location at ground level as shown in the attached photograph

,.' , (Figure. 14, Appendix A) was selected by.the APS staff. This location is the secodl r'ep,orted leak location. A 2-ft. test grid was established in order to conduct !impulse Response testing., Hammer sounding was 'also, conducted.

The area measured 13,ift.,wide by 7 ft. high.

Schematic sketch of the tested ar6as is slo\n in Figure 1.

,.,, .2_9F ,I .

ARE'AONE ,,,,ARA,"woO'-',.. AREATHREE Elevation 100 ft Elevation 100 ft Elevation 100 ft AREA FOUR AREAFIVE Elevation 120 ft Exterior Location Figure 1. Schematic Sketch of Tested Areas.

The following NDT test methods were used in the evaluation:

1. Ground Penetrating Radar (GPR): GPR testing was performed on Area One and Area Two to identify the presence and location of reinforcing steel in the'waills. Walls were scanned vertically and horizontally at.l,-ft intervals. A description of the GPR method is presented in Appendix B. .,

UP wRO

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2. Impulse Response Testing; This NDT method was used to evaluate the concrete integrity of the wall st 'ructure, especially in th6-reas wher-e th'ei cra6ks alre, pres6ht and.predict the presehce of iowý-density con'nrete' t'ypical 'fhoneycombing'0'oor con'olidation. The nondestructive Impulse Response (IR) test methodlis also desciribe'd in Appendix B of this report.

For the purpose"Of this evaluation, average: mobility and' mobility slope parameter were evaluated. The presence of significant cracking or honeycombinig irn the concrete will increase the mobility over the tested frequency range; The slobpebfthle mobility plot between 100 and 800 Hz was derived. In the event of poor consolida tion, honeycombinig, or the presence of microcracks in the concrete, the value of the mobility slope increases.

Generally, results obtained from IR testing demonstrate the relative, condition of the concrete and will vary due to a wide range of conditions. Mobility will vary according to the structural thickness and restraint of a member along with variations in concrete properties and significant cracking and delaminations. Mobility slope will not vary significantly with structural variations, but will vary with changes in concrete properties and micro-cracking.

Due to this variation,,it is not possible to set absolute threshold valuesfor either mobility or mobility slope that would indicate the presence of damage. Such threshold values would vary from structure to structure and even element to:element within a single structure. Measured values are generally shown on a contour:plot to highlight any unusual patterns or variations that would indicate damage. In addition,- measured values can be calibrated, alongside other methods, including visual inspection and coring:, to provide a standard for.the measured Values.

Impulse response testing was conductedon alL§selected tareas mentioned ab6ve. A grid, approximately 1-ft by, 1-ft or 2-ft by 2-ft was laid' out on each tested areas and each point was tested.  :, ,

3. Ultrasonic Pulse Velocity (UPV): Due to the lack of access to the opposite side of the wall the indirect method was used to estimate the strength of the concrete and predict the crack depth.

The crack depth measurement is based on the concept that the travel' time of a pulse encountering an air-filled crack will be greater than through similar concrete without any cracks.

This is due to the fact that the pulse willbe diffracted around the crack tip. In order to estimate the crack depth, one transducer is placed on one side of the crack and the other is placed at equal distance on the other side:of the crack.

The depth of crack is calculated in accordance with the following formula:

2 2 2 h = (x/T 2)( Ti -T2 21/)

Where: .

h istheccrack depth'.......... .

x is the distahce 6f~the-trbasducer'fr*m -the crack (both transdbcer. and receiver are placed equidistant form the crack),

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Mr.. Warren Jones, Palo Verde Nuclear Generating Station Page 5 of 7, CTLGroup Project No. 059084 November 18,' 2d05

,T is the transit time around the crack and T 2 is the transit time along the surface of the same'concrete Without any'crack."

Surface path length for.T, and.T2 4stould be equal.. . Transit timeiS measured in microseconds using a y-meter" conniectedto transducers.

More details aboiut test procedures are pre'sented in Appendix B.

ND' RESULTS Ground P1enetrating Radar (GPR): GPR was used to evaluate the'presence andlocation of reinforcing steel bars in the'wall and estimate the* concrete'cover.-.Typical Radar scans are shown in Figures 1 and 2 in Appendix C. Each figure is an array of. closely spaced waveforms where the x-axis represeotsfhe antenna path on the'wall and the Y-axis is pulse transit time (nan6seconds). Generally, for hbormal cncrete, 1 nan6§sc6nd is equivalent to 2 in. of concrete cover. 'Based onthis ;criterion'jtheOaverage conccr~te cover for both horizontal and vertidal reinfoýcihrg s~teel *nd rebar spalcing are showh in'Tbld 1.

"TABLE 1. Mepsured Concrete Cover and Bar Spacing A r-ea0hne Area'Two Vertical Steel Min. Cover (in) 0.72 1.60 Max. Co.ve* .t , 1*..

.1i8 1:83

- * . Average'Cover (in..) :0.98 . 1.38 Average , paci hgi.'): 6.0" .. 6.0 Horizontal Steel Min.ý Cover (in.) 1.44 1.37 "Max. Cover . . ; 1.70, 1(in.)y. 2.29 Average Cover (in') " 1.57.."' 1.86 -

__

_ Average Spacing (in..). 6.0 . .6.0 Impulse Response (IR): Asmentioned above,,for the purpose:.ofthis evaluation, two principal.

parameters, average mobility and mobility slope were used.t The collected ,data are presented in color-coded contour maps presented in; Figures 3 through; 7,in Appendix C.. The average ,

mobility values calculated from the tested areas from.0.0 to. 5.0 .ad mobility slope ranges from 0 to 3.5. Low mobility and low mobility slope indicate higher stiffness and good quality concrete.

The variations of average mobility and mobility slope, in this eva.luation are indicative of normal variation due to variations in structural configurations, concrete density and the cracking observed in the visual inspection. ,No evidence, of significant ;hidden, damage was observed.

Area One. The contour plots foraverage mobility, and mobility, slopevalues for Area One are shown in Figures 3.a and 3.b in Appendix C. As shown in Figure 3.a,.the average mobility varies from 0.0 to 5.0. The highest mobility values measured around the upper right corner of the area is due to the pipe opening in the wall (approximately 8 in. in diameter). The cold joint

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Mr. Warren Jones, Palo Verde Nuclear Generating Station Page 6 of 7 CTLGr.up Project No. 059084 NoVeimber 18', 2005 observe in the lower part of the wall did not seem to affect the stiffness of the wall (lower.

mobility values were calculated ihnthe lower pait of the area. The mrobility slopes values (Figure 3.b) are low, indicating good quality concrete.

Area Two. As reported, the lower part of this area was considered the most critical, since water was leaking through cracks present at this part of the wall.* However, as shown in Figure 4.a in Appendix C, the average mobility in the lower part of the area ;was-relativei _166 7indidaiting that; the presence of the cracks did not affect the stiffness or the structural integrity of the wall at this area. Lower mobility slopes was also calculated (Figure 4.b in Appeindix C) at this part 6f the wall indicating the quality of concrete with, no evidence of hidden damage or poor consolidation at this area. The high mobility and mobility slope shown in the upper part of the area are due to the presence of rectangular steelpates bonded to the wall at these locations. It should be indicated that the clear areas (white) in the figures are untested. locations dueto lack,6f access.

Area Three. Average mobility and mobility slopes are. shown in .Figures 5.a and 5.b in Appendix C. Again as in Area Two, the high rnobilfty(red areas)!a're due the presence of steel plates bonded to the wall and white ar6as represent the untested locations due to lack Of access.

Generally the average mobility and mobility slopes Yýere alsolo'w in this'area indicating of higher stiffness and good quality concrete "

Area Four. Average mobility and mobility slope values for this ar~ea are shown in Figure 6.a and 6.b in Appendix.C. As'shown in both figures, average mobility and mobility slope are low.

There is no indication of hidden damage 6r defects in the wall at this area.

Area Five. This area Was selected to be tested due to-the presence of vertical cold joint near the edge of wall. The. contour plots fo& average mobility and tm" obility slope values are present in Figures 7.a and 7.b in Appendix C. The low average mobility values-obtained at this area indicate that the presence of cold joint did not influenced the stiffness or integrity of the wall at this area. Lower mobility slope values are also an indication of good Ouality concrete. .

Ultrasonic Pulse Velocity (UPV): Due to the presence of a glossy surface coating on all wall surfaces, stable UPV transmission were difficult to obtain. 'However, three locations in the tested areas were found to be reliable and indicated estimated compressive strength of approximately 4500 to 5500-psi: Further confirmation-of in-place compressive strength would require destructive sampling and compressive strength testing in accordance with ASTM C 42.

The UPV method utilized to estimate the crack depths in Area Two near the reported leakages, found cracks extended to at least 8 in. ftornithe concrete wall surface. Since water transmission has been reported through the wa[I thickness, it is Obvious that Crack propagation is greater than 8-in* However; the crack is most:likelyfilled with water or boric acid crystals asseen' deposited' on the wall surface. Such deposits permit UPV signal continuity indicating the crack depth recorded.

".,CONCLUSIONS AND RECOMMENDATIONS The overall condition of the concrete is good:-- The visual observations of the exposed concrete surface found that the overall surface integrity is very good, in spite of minor cracking exhibited in the forms of craze cracking- hormallyrexpected drying shrinkage cracking, as well as possible thermal andASR related'cracking'., , . -

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Mr,: Warren Jones" Palo Verde Nuclear Generating Station "7 CTLGroup Project No. 059084 November' 18, 2005 Isolated areas of concrete cracking exhibit patterns similar to concretes experiencing expansion due to ASR. -Confirmation of the cause of the cracking patterns potentially being related to ASR cannot be completely identified simply by visual observations. Confirmation of the potential of this type of cracking would require examination of core samples. Reportedly, a section of the steam generator bio-shield wall may have been removed at an earlier date. If such a section is available, sampling could be conducted on this section to avoid destructively sampling the spent fuel pool wall. If you decide to allow destructive testing, information on the original mix design and the results of aggregate testing would be helpful in the determination of the possible cause of cracking.

Nondestructive testing did not locate majorlcracking, delamination, or-honeycombing. Although previous leakage through the spent fuel pool wall indicates through-wall cracking, NDT indicates that the water-borne deposits may be present within the crack. This assumption is based upon UPV data'indicating crack'depths of approximately 8 inches. At greater depths, deposits within the crack may be sufficient to facilitate signal transmission through the crack.

The future performance of the concrete in all areas tested is expected to be unimpaired as long as the leakage collection system is once again operated properly. Both visual inspection and NDT results indicate the concrete to be in good condition.

Typical crack repair approach would include epoxy or pqlyurethane injection. However, the lack of continued water leakage and the relatively low frequency of cracking do not appear to warrant an immediate need for repair. Furthermore, the potential that injected crack repair material could enter and compromise the leakage co lection system makes crack injection less desirable unless absolutely necessary. ,It.is*suggested that, as lor.g~as the leakage collection system is operational, and no further moisture is found to pass through the'existing cracks, repair may not be necessary. It is suggested to monitor the widths of the cracks and note any moisture movement at the cracked surfaces. If any significant crack movement or leakage is detected in the future, please contact us to discuss possible options to further examine the structure.

Please call us if you have any questions.

Very truly yours, George W. Seegebrecht, PE Mohamad A. Nagi, Ph.D.

Senior Evaluation Engineer NDT Specialist GseeqebreechtACTLGroup.com MNagikCTLGroup.com 847-972-3232 847-972-3266 Tig GROUP

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Mr .-WarrenJqnes:,. Palo Verde Nuclear Generating Station Page A-2 of 7, CTLGroup.Project,No,-059084 .: ,,. November 18, 2005 Figure 1 Initial point of leakage reported by:

P/NGS personnel., No leakage was evidient at the time of the visit since the

.leakage collection system piping had been on a regular drainage schedule..

-v Figure 2 Typical appearance of the spent fuel pool

,outside face in the proximity of the reported leakage.

Figure 3 Crack widths in the vicinity of'one of the reported leakage ranged from 0.01 to 0.016 inches'in general.

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."C 7 Mr. Warren Jones, Palo Verde Nuclear Generating Station Page A-3 of 7 CTLGr6Lup Project No. 059084 November 18, '2005 Figure 4 Cirackin'g patterns were Dredrminat6ly-vertica l b(it*rmurnerbus horizontal cracks.

wiere also com-mdn.,Cracking appears to

be

,typical ofcrazing, wkhich occur very arl frmremovalas well as thermal related cracking for'such massive concrete (5 ft to 8 ft thick). Also as expected cracking appears to be exhibited due to normal drying shrinkage.

All concrete surfaces "appearedOto have been coated with a clear surface coating,;

possibly polyurethane or similar.

"Figure 5 C!oser view Of.crackifig as.sho6wn in Figu'r&4' FigiJre6: " "

Impulse Response testing beginning at the west end of the south wall of the spent fuel pool wall area.

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CTLGrodP Project No.,059084 November 18, 2005 Figure 7 Ultra-Sonic Pulse Velocity being used to estimate crack depths at various cracks of the south spent fuel pool wall.

Figure 8

.Typical cracks widths observed.

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Figure 9 Outside of South spent fuel pool wall looking.west:. , , *! " ." ". 'L-*  :

Note test grid spacing established-on vriedthe

ýpent fuel-pool wall. Grid spacing from one (T) toýtwo (2) feetdepending on2' conditions enco0untered."..

FigurelO Impulse response testing in progress at.

upper elevations of the spent fuel pool 1 walls.

Figure 11 .

View of adjacent sump"room exterior piping of the leakage collections system.

White deposits at the lower portion of the wall is reported boron deposits

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Closer view of leakage collection system Figure 13 One foot grid spacingwas used for the Impulse response testing in the immediate vicinity of the reported leakage.

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Mr. Warren Jones, Palo Verde Nuclear Generating Station PageA-7of7 CTLGroup Project No. 059084 November 18,'2005 Figure 14 -

Exterior wall '(second leakage area),

selected byAPS staff for nondestructive.

testing: IR on a 2-ft by 2-ft grid and hammer sounding. Tested area (Area Five) was, 13-ft wide by ft high.:

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Mr. Warren Jones, Palo Verde Nuclear Generating Station Page B-2 of 4 CTLGroup Project No. 059084 .:N6vember18, 2005 THE IMPULSE RESPONSE TEST The Impulse Response (IR) test method is a nondestructive, stress wave test, used extensively in the evaluation of machined metallic components in the aircraft industry. Its application to concrete structures in Civil Engineering is less well known, and the method has received far

.less publicity than the recently developed Impact-Echo (I-E) test. Both methods are described in the American Concrete Institute Report ACI 228.2R-98, "Nondestructive Test Methods for-Evaluationof Concrete in Structures".

The IR method (also referred to in earlier literature as the Transient Dynamic Response or Sonic Mobility method) is a direct descendant of the Forced Vibration method-for evaluating the integrity of concrete drilled shafts, developed in France in the 1960's. The basic theory of dynamic mobility developed at that time has not changed; however, its range of applications to different structural elements has increased to incorporate the following problems:

• voiding beneath concrete highway, spillway and floor slabs.

• delamination of concrete around steel reinforcement in slabs, walls and large structures such as dams, chimney stacks and silos.

• low density concrete (honeycombing) and cracking in concrete elements

• the depth of ASR attack in drilled shafts used as pylon foundations 0 debonding of asphalt and concrete overlay's to concrete substrates

• the degree of stress transfer th*ough. 1ad tiransfer!systems across joints in concrete slabs).

IR TESTING EQUIPMENT The method uses a low strain impact to send a stress wave through the tested element. The impactor is usually a 1-kg sledgehammer with a built-in load cell in the hammerhead. The maximum compressive stress at the impact point in concrete is directly related to the elastic properties of the hammer tip. Typical stress levels range from 5. MPa for hard rubber tips to more than 50 MPa for aluminum tips. The response to the input stress is normally measured using a velocity transducer (geophone). This receiver is preferred because of its stability at low frequencies and its robust performance in practice. Both the hammer and the geophone are linked to a portable field computer for data acquisition and storage.

METHOD DESCRIPTION When testing plate-like structures, the Impact-Echo method uses the reflected stress wave from the base of the concrete element or from some anomaly within that element (requiring a frequency range normally between 10 and 50 kHz). The IR test uses a compressive stress impact approximately 100 times that of the I-E test. This greater stress input means that the plate'responds to the IR hammer impact in a bending mode over a verymuch lower frequency range (0-1 kHz for plate structures), as opposed to the reflective mode of the I-E test.

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Mr. Warren Jones, Palo.Verde Nuclear Generating Statibn .. , Page B1.3of4 CTLGroup Project No. 059084 November 18, 2005 Both the time records for the hammer force and the geophone velocity response are processed in the field computer using the Fast Fourier.Transform (FFT) algorithm. The resulting velocity spectrum is divided by the. force spectrumto obtain a transfer function;, referred, to as the Mobilityofithe element .under.testc The test graph of Mobility plotted against frequency over,.the 0-1kHzirange*contains information on the condition and the integrity:of the concrete in the tested elements; obtained from the~following-measureid parameters;r Dynamic Stiffness: The slope of the portion'of the Mobility:: plot belbw.0.1 kHz defines the compliance or flexibility of the area around the test point for a normalized force.

input. The inverse of the compliance is the dynamic stiffness of the structural element attthe test point.'Thisean be expressed as: ,

Stiffness f [concretequality,element thioknes,; elementsupport condition]:

, . . . I; - . I .

o Mobility and Damping: The element's response tothe impact-generated elastic wave

,will be damped by the eiement'ssitrinsicrigidity (body damping). The meanmobility value over the 0.141 kHz rangeis .diredtly~related to the density and the thickness of a plate eiement,.for-example. A reduction in :p!ate' thickness corresponds to an increase in, mean mobility. As ahexamplej when, total debonding of an upper layer is present.,

the mean mobility reflects the thickness of the upper, debonded layer,(in other words, the slab becomes more mobile)., Also, any cracking or honeycombing in the concrete will reduce the-damping and hence the stability of'the mobility plots over the tested frequency range. ý.. . , .

o Peak/Mean Mobility Ratio: When debonding or delamination is present within a structural element, or when there is loss of support beneath a concrete slab on grade, the response behavior of the uppermost layer controls the IR result. In addition to the increase in mean mobility between 0.1 and 1 kHz, the dynamic stiffness decreases greatly.The-peak mobility below*O!AJFkHz becdimes:alpreciably higher than the mean mobility from 0.1-1 kHz. The ratio of this peak to mean mobility is an indicator of the presence, and degree of either debondipg within the element.or voiding/loss of support beneath a,slab on grade. , , . .- , . .

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Mr. Warren Jones, Palo Verde Nuclear Generating Station Page B-4 of 4 CTLGroup Project No. 059084 C. No-6ember'13, 2005 THE IMPULSE RADAR TEST (GROUND PENETRATING RADAR)

The impulse radar technique employs high-frequency electromagnetic energy waves for, rapidly, and, continuously assessing a variety of characteristics of concr.ete structures., The principle of operation, is based on reflection of electromagnetic waves from varyingodielectricconstant boundaries in the material being probed. The impulse radar equipment is self-contained,,

compact, and portable. The system consists of the main radar unit, antenna and transducer cable. All data is stored in the main radar unit,:by meansof a computer hard drive (ACI 228, 1998). ... .. ,. ,

A single or double contacting transducer (antenna) transmits and receives radar signals. High frequency, short pulse electromagnetic energy is transmitted into the element under test (concrete, sub-base). Each transmitted'pulse-travels through the material, and ,is partially .

reflected when it encounters a change in dielectric constant. The receiving section of the transducer detects reflected pulses-. *The location :and ;depthýof the-dielectric constant boundary is evaluated by noting' the transit time from-start of pulse .to reception.of reflected pulse.

Boundary depth is proportional to transit time:Srnce-concrpte to air; water, and/or backfill interfaces are electronically detected by.the ;instrument as dielectric constant.boundaries; the impulse radar method, is capable .of..assessing avariety of reinforced concrete, masonry and environmental characteristics: .. ,

Impulse'radar has been successfulty used, toevaluate locations of embedded reinforcing, to distinguish between grouted and ungrouted cells in masonry block walls, to locate, high moisture and chloride concentration in bridge decks, embedded foreign objects (clay balls) in concrete pavements, alignment of dowellbars and the-,consolidation ofthe concrete around dowel bars in concrete pavements. .-.  :.  ; , ;

Y' -ETM HDOO-ULTRASONIC PULSE VELOCiTY.-,TESTMETHODOLOGY The Ultrasonic pulse vblocity'(U PV)' test i*sa.'f4cl'hiniqe'e mploys sonic energy pulses'to evaluate the consistency and quality of concrete. In the direct transmission mode/the UPV test uses an ultrasonic pulse to generate a compression wave through the concrete element being tested.

The pulse is received by a transducer on the opposite side of the element under test. The pulse is rapidly attenuated in the concrete and maximum transmission path lengths in good concrete are of the order of 5 feet.

The Ultrasonic Pulse Velocity is calculated by dividing the pulse path length by the pulse travel time. UPV is a function of both the concrete dynamic elastic modulus and its density, and not directly of the concrete strength. However, correlation of measured concrete strengths and UPV can be established when the compression strengths of cores taken from the structure are measured. The test and its recommended use are described in ASTM Standard C-567.

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