ULNRC-05860, Amendment to Application for Renewed Operating License

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Amendment to Application for Renewed Operating License
ML12128A150
Person / Time
Site: Callaway Ameren icon.png
Issue date: 05/03/2012
From: Kanuckel L
Ameren Missouri, Union Electric Co
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
ULNRC-05860
Download: ML12128A150 (37)
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Text

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'WAmeren MISSOURI Callaway Plant May 3, 20I2 ULNRC-05860 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-000 I IO CFR2.IOI IO CFR 2.I09(b)

IO CFR 50.4 IO CFR 50.30 IO CFR 51.53(c)

IO CFR 54 Ladies and Gentlemen:

DOCKET NUMBER 50-483 CALLAWAY PLANT UNIT 1 UNION ELECTRIC CO.

FACILITY OPERATING LICENSE NPF-30 AMENDMENT 2 TO APPLICATION FOR RENEWED OPERATING LICENSE

References:

1. ULNRC-05830 dated December IS, 20II
2. ULNRC-05856 dated April 25, 20I2 By Reference I, Union Electric Company (Ameren Missouri) submitted a license renewal application (LRA) for Callaway Plant Unit I . Reference 2 transmitted Amendment I to the Callaway LRA, and the purpose of this letter is to provide Amendment 2 to the Callaway LRA.

The changes being made by Amendment 2 are contained in Enclosure I and Enclosure 2. Enclosure 1 identifies Callaway time-limited aging analyses (TLAA) changes associated with reactor vessel underclad cracking analysis, limiting locations for environmental assisted fatigue, and other TLAA changes. Enclosure 2 identifies Callaway aging management review (AMR) changes that are being made as follows:

  • In LRA Table 3.2.2-5, correct the identified valve material to indicate "carbon steel" for the high pressure coolant injection stainless steel valve in an "atmosphere/weather" environment (external) and the associated internal steam environment AMR lines.
:::::::::::::::::::::::::::::::::::::::::::::::::::::::::::::::::::::::::::::::::::::::::::::;::::::::::::::::::: PO Box 620 Fulton, MO 65251 AmerenMissouri.com

ULNRC-05860 May 3, 2012 Page2

  • In LRA Table 3.3.2-5, correct the identified environment from "plant indoor air" to "atmosphere/weather" for the ductile iron valve AMR line in the service water system. In addition, carbon steel piping in an "atmosphere/weather" environment for the service water system will also be added to the scope oflicense renewal. Reference Section 3.3.2.1.5 of the LRA.

It should be noted that changes to one commitment (Item #37) are reflected in Table A4-1 (within ).

If you have any questions on LRA Amendment 2, please contact me at (573) 823-9286 or Ms. Sarah Kovaleski at (314) 225-1134.

I declare under penalty of perjury that the foregoing is true and correct.

Sincerely, Executed on: ft1 o.y 3 1 2 0 12-Les H. Kanuckel Manager, Engineering Design DS/SGK/nls

Enclosures:

1. Callaway Plant Unit 1 License Renewal Application Amendment No. 2 Time-Limited Aging Analyses Changes
2. AMR Changes for Callaway Plant Unit !License Renewal Application Amendment No.2

ULNRC-05860 May 3, 2012 Page3 cc: U.S. Nuclear Regulatory Commission (Original)

Attn: Document Control Desk Washington, DC 20555-0001 Mr. Elmo E. Collins Regional Administrator U. S. Nuclear Regulatory Commission Region IV 1600 East Lamar Boulevard Arlington, TX 76011-4511 Senior Resident Inspector Callaway Resident Office U.S. Nuclear Regulatory Commission 8201 NRC Road Steedman, MO 65077 Mr. Brian Harris Safety Project Manager Office ofNuclear Reactor Regulation U.S. Nuclear Regulatory Commission Mail Stop 0-11D19 Washington, DC 20555-0001 Mr. Kaly Kalyanam Senior Project Manager, Callaway Plant Office ofNuclear Reactor Regulation U. S. Nuclear Regulatory Commission Mail Stop 0-8G14 Washington, DC 20555-2738

ULNRC-05860 May3, 2012 Page4 .

Index and send hardcopy to QA File A160.0761 Hardcopy:

Certrec Corporation 4150 International Plaza Suite 820 Fort Worth, TX 76109 (Certrec receives ALL attachments as long as they are non-safeguards and may be publicly disclosed.)

Electronic distribution for the following can be made via Tech Spec ULNRC Distribution:

  • A. C. Heflin F.M.Diya C. 0. Reasoner ill D. W. Neterer L. H. Graessle J. S. Geyer S. A. Maglio R. Holmes-Bobo NSRB Secretary S. G. Kovaleski T. B. Elwood G. G. Yates E. Blocher (STARS PAM COB)

Mr. Bill Muilenburg (WCNOC)

Mr. Tim Hope (Luminant Power)

Mr. Ron Barnes (APS)

Mr. Tom Baldwin (PG&E)

Mr. Mike Murray (STPNOC)

Ms. Linda Conklin (SCE)

Mr. John O'Neill (Pillsbury Winthrop Shaw Pittman LLP)

Missouri Public Service Commission Mr. Dru Buntin (DNR) to ULNRC-05860 Page 1 of29 ENCLOSURE 1 Callaway Plant Unit 1 License Renewal Application Amendment No. 2 Time-Limited Aging Analyses Changes

  • Further Evaluation 3.1.2.2.5 Crack Growth due to Cyclic Loading
  • Table 4.1-1 List of TLAAs (TLAA Category 5 and 6)
  • Table 4.1-2 Review of Analyses Listed in NUREG-1800 Tables 4.1-2 and 4.1-3
  • Section 4.3.4 Effects of the Reactor Coolant System Environment on Fatigue Life of Piping and Components (Generic Safety Issue 190)
  • Section 4.6 Containment Liner Plate, Metal Containments, and Penetrations Fatigue Analyses
  • Section 4.7.21n-8ervlce Flaw Analyses that Demonstrate Structural Integrity for 40 years
  • Section 4.7.4 Reactor Vessel Underclad Cracking Analysis
  • 4.8 References
  • Appendix A3.2.3 Effects of the Reactor Coolant System Environment on Fatigue Life of Piping and Components (Generic Safety Issue 190)
  • Appendix A3.6.4 Reactor Vessel Underclad Cracking Analysis
  • Table A4-1 Item 37 to ULNRC-05860 Page 2 of29 Callaway LRA Amendment 2 Time Limited Aging Analysis Changes Affected Pages LRASection Page Nos 3.1.2.2.5 3.1-10 Table 3.1.2-1 3.1-61 and 66 Table 4.1-1 4.1-5 Table 4.1-2 4.1-6 and 7 4.3.4 4.3-31-37 4.6 4.6-1 4.7.2 4.7-4 4.7.4 4.7-8 and9 4.8 4.8-2 A3.2.3 A-27 A3.6.4 A-34 and35 TableA4-1 A-49 to ULNRC-05860 Page 3 of29 Callaway Plant License Renewal Application Amendmentl Revision to further evaluation to revise disposition of TLAA for underclad cracking.

Section 3.1.2.2.5 (page 3.1-1 0) is revised as follows (deleted text shown with strike through):

3.1.2.2.5 Crack Growth due to Cyclic Loading An analysis of crack growth of underclad flaws in reactor vessel forgings due to cyclic loading to qualify them for the current licensed operating period would be a TLAA. This phenameAOA has been a~EiFesseEi iA the Callaway vessel by weiEi GlaEiEiing prosesses EiesigneEi to avoiEi these EiefeGts.

No ~:~nEieFGiaEi GFaGks have been Eietestea ar analyzes for the Calla\\'ay vessel, in tRe absenoe of whiGh theFe aFe no TLMs. Section 4. 7.4 describes the absenoe of a TLAA for underclad cracking.

to ULNRC-05860 Page4 of29 CaRaway Plant License Renewal Application Amendment2 Revision to Table 3.1.2-1 to add additional TLAA lines for crack growth due to cyclic loading.

Table 3.1.2-1 (pages 3.1-61 and 66) is revised as follows (new text shown under1ined):

Table 3.1 .2-1 Reactor Vessel, Internals, and Reactor Coolant System- Summary of Aging Management Evaluation- Reactor Component Intended Material Environment Aging Effect Aging Management NUREG-1801 Table1 Notes Type Function Requiring Program Item Item Management RV Inlet and PB Carbon Reactor Coolant Crack growth due Time-Limited Aging IV.A2.R-85 13.1.1.018 A Outlet Nozzles Steel with (lnt) to C!iclic loading Anal!isis evaluated for Stainless the geriod of extended Steel operation Cladding RV Ugger, PB Carbon Reactor Coolant Crack growth due Time-Limited Aging IV.A2.R-85 3.1.1.018 A Intermediate, Steel with (lnt) to Cl£clic loading Anal!isis evaluated for Lower Shell Stainless the geriod of extended and Welds Steel operation Cladding I I I to ULNRC-05860 Page 5 of29 Callaway Plant License Renewal Application Amendment2 Revision to Table 4.1-1 to revise titles and dispositions.

Table 4.1-1 (page 4.1-5) is revised as follows (deleted text shown with strikethrough, new text underlined):

Table 4.1-1 List of TLAAs TLAA .**. Disposition Description Section Category Categoryl11

1. Reactor Vessel Neutron Embrlttlement Analysis N/A 4.2 Neutron Fluence Values ii 4.2.1 Charpy Upper-Shelf Energy ii 4.2.2 Pressurized Thermal Shock ii 4.2.3 Pressure-Temperature (P-T) Limits iii 4.2.4 Low Temperature Overpressure Protection iii 4.2.5 2 Metal Fatigue N/A 4.3 Fatigue Monitoring Program N/A 4.3.1 ASME Section Ill Class I Fatigue Analysis of Vessels, iii 4.3.2 Piping and Components Reactor Coolant Pump Thermal Barrier Flange ill 4.3.2.1 Pressurizer lnsurge-Outsurge Transients iii 4.3.2.2 Steam Generator ASME Section Ill Class 1, Class 2 Secondary Side, and Feedwater Nozzle Fatigue i 4.3.2.3 Analyses NRC Bulletin 88-11 Revised Fatigue Analysis of the Pressurizer Surge Line for Thermal Cycling and iii 4.3.2.4 Stratification ASME Section Ill Subsection NG Fatigue Analysis of iii 4.3.3 Reactor Pressure Vessel Internals Effects of the Reactor Coolant System Environment on Fatigue Life of Piping and Components (Generic iii 4.3.4 Safety Issue 190)

Assumed Thermal Cycle Count for Allowable Secondary Stress Range Reduction Factor in i 4.3.5 ANSI B31 .1 and ASME Section Ill Class 2 and 3 Piping to ULNRC-05860 Page 6 of29 Table 4.1-1 List of TLAAs TLAA

.. , Disposition Description Section Category Category'11 Fatigue Design of Spent Fuel Pool Liner and Racks for i 4.3.6 Seismic Events Fatigue Design and Analysis of Class 1E Electrical i 4.3.7 Raceway Support Angle Fittings for Seismic Events Fatigue Analyses of Class 2 Heat Exchangers ii, iii 4.3.8 Environmental Qualification (EQ) of Electric

3. Ill 4.4 Equipment
4. Concrete Containment Tendon Prestress I, II 4.5 Containment Liner Plate, Metal Containments, and
5. N/A 4.6 Penetrations Fatigue Analyses Design Cycles for the Main Steam Line and Feedwater i, ii 4.6.1 Penetrations Fatigue Waiver Evaluations for the AccessEqijipmeRt i 4.6.2 Hatches and Leak Chase Channels
6. Other Plant..Speclflc Time-Limited Aging Analyses N/A 4.7 Containment Polar Crane, Fuel Building Cask Handling Crane, Spent Fuel Pool Bridge Crane, and i 4.7.1 Refueling Machine CMAA 70 Load Cycle Limits In-service Flaw Analyses that Demonstrate Structural i 4.7.2 Integrity for 40 years Corrosion Analysis of the Reactor Vessel Cladding i 4.7.3 Indications AbseRGe sf a TbAA. fer Reactor Vessel Underclad Cracking Analyses NIAl 4.7.4 Reactor Coolant Pump Flywheel Fatigue Crack Growth i 4.7.5 Analysis High Energy Line Break Postulation Based on Fatigue iii 4.7.6 Cumulative Usage Factors Fatigue Crack Growth Assessment in Support of a Fracture Mechanics Analysis for the Leak-Before-i 4.7.7 Break (LBB) Elimination of Dynamic Effects of Piping Failures Replacement Class 3 Buried Piping i 4.7.8 Replacement Steam Generator Tube Wear i 4.7.9 to ULNRC-05860 Page 7 of29 Callaway Plant License Renewal Application Amendmentl Revision to Table 4.1-2 to revise applicability. to Callaway Plant.

Table 4.1-2 (pages 4.1-6 and 7) is revised as follows (deleted text shown with strikethrough, new text underlined):

NUREG-1800 Examples Applicability to Callaway ,. Section NUREG-1800, Table 4.1 Potential TLAAs Reactor Vessel Neutron Embrittlement Yes 4.2 Metal Fatigue Yes 4.3 Environmental Qualification (EQ) of Electric Yes 4.4 Equipment Concrete Containment Tendon Prestress Yes 4.5 In-Service Local Metal Containment No - No exglicit basis based on Qlant life 4-.+.-3:

Corrosion Analyses a12121ies.¥es NUREG-1800, Table 4.1 Additional Examples of Plant.Speclflc TLAAs lntergranular Separation in the Heat-Affected t>la t>lai=IAZ aRalyses were i9eAtifie9 Zone (HAZ) of Reactor Vessel Low-Alloy 4.7.4 wilhiR the Gb8. Yes Steel Under Austenitic SS Cladding Low-Temperature Overpressure (LTOP)

Yes 4.2.5 Analyses Fatigue Analysis for the Main Steam Supply Lines to the Turbine-Driven Auxiliary Yes 4.3.5 Feedwater Pumps Fatigue Analysis for the Reactor Coolant Yes 4.7.5 Pump Flywheel Fatigue Analysis of Polar Crane Yes 4.7.1 Flow-Induced Vibration Endurance Limit for No-No explicit basis based on plant life 4.3.3 the Reactor Vessellntemals applies.

Transient Cycle Count Assumptions for the Yes 4.3.3 Reactor Vessel Internals Ductility Reduction of Fracture Toughness No-No explicit basis based on plant life 4.3.3 for the Reactor Vessel Internals applies.

Leak Before Break Yes 4.7.7 to ULNRC-05860 Page 8 of29

'~ .<"* '~

NUREG-1800 Examples Applicability to Callaway Section Fatigue Analysis for the Containment Liner No - No fatigue or cycle-based analysis 4.6.0 Plate supports design of the liner.

Containment Penetration Pressurization Yes 4.6.2 Cycles No No eMplicit basis based OR plaRt life Metal Corrosion Allowance 4.7.3-applies. Yes High-Energy Line-Break Postulation Based Yes 4.7.6 on Fatigue Cumulative Usage Factor In-Service Flaw Growth Analyses that Yes 4.7.2 Demonstrate Structure Stability for 40 Years

Enclosure 1 to ULNRC-05860 Page 9 of29 Callaway Plant License Renewal Application Amendment 2 Revision to Section 4.3.4.

Section 4.3.4 (pages 4.3-31 through 37) is revised as follows (deleted text shown with strikethrough, new text underlined):

4.3.4 Effects of the Reactor Coolant System Environment on Fatigue Life of Piping and Components (Generic Safety Issue 190)

The NRC concluded that effects of the reactor coolant environment might need to be included in the calculated fatigue life of components, and opened three generic safety issues to address this question, all finally closed to a single Generic Safety Issue 190. Subsequent research and studies refined the methods, which no longer use the interim fatigue curves of NUREG/CR-5999 but calculate an environmental fatigue effect multiplier F8 ," which depends on material type, temperature, strain rate, and dissolved oxygen; and for carbon and low-alloy steel, sulfur content.

NUREG-1800, Standard Review Plan for Review of License Renewal Applications for Nuclear Power Plants states that "The applicant's consideration of the effects of coolant environment on component fatigue life for license renewal is an area of review," noting the staff recommendation

" ... that the samples In NUREG/CR-6260 should be evaluated considering environmental effects for license renewal."

The GSI-190 review requirements are therefore imposed by the Standard Review Plan and do not depend on the individual plant licensing basis. Callaway addressed GSI-190 review requirements by assessing the environmental effect on fatigue at the NUREG/CR-6260 locations for the newer-vintage Westinghouse Plant.

NUREG/CR-6260 identifies seven sample locations for newer vintage Westinghouse plants which need to consider the effects of reactor coolant environment on component fatigue life for license renewal:

1. Reactor Vessel Lower Head to Shell Juncture
2. Reactor Vessel Primary Coolant Inlet Nozzle
3. Reactor Vessel Primary Coolant Outlet Nozzle
4. Hot Leg Surge Nozzle
5. Charging Nozzles
6. Safety Injection Nozzles
7. Residual Heat Removal Line Inlet Transition Table 4.3-6, Summary of Fatigue Usage Factors at NUREG/CR 6260 Sample Locations is a summary of environmentally-assisted fatigue of the NUREG/CR-6260 locations. The Fan relationships are calculated from NUREG/CR-6583 for carbon and low-alloy steels and from NUREG/CR-5704 for stainless steels, as appropriate for the material at each of these locations.

The NUREG/CR-6260 locations in Table 4.3-6, Summary of Fatigue Usage Factors at NUREG/CR 6260 Sample Locations with an EAF CUF below 1.0, when using the design basis

Enclosure 1 to ULNRC-05860 Page 10 of29 CUF and the maximum Fan. require no further analysis. Three of the NUREG/CR-6260 locations, (1) RPV lower head to shell juncture, (2) RPV inlet nozzle, and (3) RPV outlet nozzles meet this criterion (Reference 4 ). All three locations are low alloy steel locations.

The maximum Fan for low alloy steel assumes the dissolved oxygen level to be less than 0.05 ppm, which corresponds to a low oxygen environment. This is consistent with the Callaway primary chemistry program, which maintains RCSG hydrogen level at 25 to 50 cc/kg. A minimum hydrogen concentration will ensure the RCS is free of oxygen. Sulfur content is assumed to be at the maximum concentration in the NUREG.

The remaining NUREG/CR-6260 locations were reevaluated with a refined fatigue analysis using NB-3200 methods in a 3-D finite element analysis model using the design number of transients to reduce the CUF values. After reanalysis the RHR inlet transition was the only location to pass the EAF CUF criterion of 1.0 (Reference 5).

Two options are available to further reduce the EAF CUFs for the charging system nozzles, safety injection nozzles, and hot leg surge line nozzle: (1) calculate a strain rate dependent Fan; and (2) calculate CUF based on the 60 year projected numbers of transient events or both.

Revision of F., Based on Strain-Rate The strain-rate dependent Fan values are calculated for the significant load set pairs in the fatigue analyses. Load set pairs that produce no significant stress range or fatigue contribution were assigned the maximum Fan for the material. The integrated strain rate method described in MRP-47, Guidelines for Addressing Fatigue Environmental Effects in a License Renewal Application, was used to calculate Fan values for individual load pairs that produce significant stress ranges. Dissolved oxygen of less than 0.05 ppm is assumed, which corresponds to a low oxygen environment. This is conservative since lower dissolved oxygen concentrations yield higher Fen values for stainless steel. Sulfur content is only applicable to low alloy steel locations.

Revision of CUF Based on 60-Year Projections of Transients However, multiplying the revised CUF by the weighted average Fan value computed above still results in EAF CUFs greater than 1.0 for the charging system nozzles, safety injection nozzles, and hot leg surge line nozzle after conservatism has been removed. In order to demonstrate that monitoring fatigue in these locations is a sufficient form of aging management, the EAF CUF was calculated based on the numbers of transients projected to 60 years in Table 4.3-2, Transient Accumulations and Projections. If the transient is not projected, then the full number of design basis events is used. There were two transients that were not analyzed at the 60 year projection. The two exceptions are the "inadvertent safety injection" and "loss of power" events, which were analyzed-at as the events to-date. The only EAF CUF calculation signifisantly that could be affected by the use of the "inadvertent safety injection" and "loss of power" transients to-date is associated with the safety injection nozzle, e.g. affected greater than the order of magnitude. This is addressed below.

The projected normal and alternate charging nozzles EAF CUFs are 0.57 and 0.53 based on SBF usage factors of 0.092 and 0.078, and Fan of 6.22 and 6.75 (Reference 6). The SBF usage factors were generated with computer software that was benchmarked against NB-3200 methods consistent with RIS 2008-30 as discussed in Section 4.3.1.1, Fatigue Monitoring Methods.

Enclosure 1 to ULNRC-05860 Page 11 of29 The projected safety injection nozzle EAF CUF is 0.74 based on the usage factor of 0.11 and Fen of 6.5 (Reference 7). Even though this location is analyzed for the numbers of "inadvertent safety injection" and "loss of power events to-date, it is monitored with CBF; therefore EAF CUF will be updated as additional events occur.

The projected hot leg surge line nozzle EAF CUF is 0. 765 based on the usage factor of 0.076 and Fen of 10.10 (Reference 8 ).

All of the locations specified in NUREG/CR-6260 for newer vintage Westinghouse plants listed in Table 4.3-6, Summary of Fatigue Usage Factors at NUREGICR 6260 Sample Locations will be monitored by the Fatigue Monitoring program, described in Appendix 83.1 . Most of the locations will be monitored using CBF or SBF. The hot leg surge nozzle will be monitored by incorporating the 60 year cycle projections into the cycle counting action limits to ensure that the results for the hot leg surge nozzle presented in this section are not exceeded. Therefore, the effects of the reactor coolant environment on fatigue usage factors will be managed for the period of extended operation. These TLAAs are dispositioned in accordance with 10 CFR 54.21 (c)(1 )(iii).

Disposition: Aging Management, 10 CFR 54.21(c)(1)(111)

Evaluation of Limiting Locations for Environmental Assisttld Fatigue In order to assure that the limiting plant-specific EAF locations are identified, Callaway performed a systematic review of all wetted, RCPB components with a Class 1 fatigue analysis

[Ref. 171. This was done either to show that the NUREG/CR-6260 locations are bounding or to incorporate EAF into the licensing basis for those more limiting components. The screening used EPRI Technical Report 1024995 "Environmentally-Assisted Fatigue Screening, Process and Technical Basis for Identifying EAF Limiting Locations," [Ref. 181. Tl:le fiFSt step in the ssreening was te apply tl:le maxiFRI:JFR F8A to all nen NUREGICR 626() lesatiens ~:~sing NURI!Gf.CR 6983 fer sarl:len anc:t lew alley steels, NUREGf.CR 97Q4 fer a1:1stenitis stainless steels, anc:t NUREG!CR 69()9 fer niskel alleys. Fer tl=lese lesatiens witl=l an EAF CUF less tt=1an 1.Q, ne t"Yrtl=ler werk is req~:~irec:t. Tl:lese lesatiens witt:! an EAF CUF greater than 1.Q were categerizee 9asec:t en the strain rate ef tl=le 9emiRant transient. Tl=le strain rate slassifisatien was 9etermines witl=l a q~:~alitative assessment !:lases en experiense ans net a q~:~antitative stress analysis fer tl=le strain rate slassifisatien . Tl=le strain rate slassifisatien, an ass1:1mee lew 9issel*.'e9 e*ygen envirenment, tl:le maMiFAI:JFR ft~:~istmetal temperat~:~re, and tl:le maximum sensentratien fer s1:1lfur centent were 1:1ses te salc~:~late tl:le Fett val1:1e eases en tl:le same NUREGs ~:~see in tl=le initialssreening. Tl:le F1111 and tl:le eesign l:lasis CUFs *.vere used te isentify tl=te limiting EAF CUF fer eacl=t material type in eacl=t system inclusing tl:lese systems net censic:terec:t iR tAe NUREGICR 626() eval1:1atien (e.g., steam generator primary siee an9 press~:~Fizer, press~:~rizer spray line, etc.). Tal:lle 4.3 7, PFelim!Rary lfleRtifisat!eR efAdditieRal SeRURel l:.esatieRs for EiAF isentifies tl=te lesatiens, in aesitien te tl=te NUREGICR 626() lecaliens, wl=tish were c:teterminec:t tel:le sanc:tieate sentlnellecatiens. The res~:~lts presented in Tal:lle 4.3 7, PFelimiRary lfleR#fiGatieR efAeeitieRal S&RtiRel LesatieRB for eAF are preliminary and se net represent tl=te final list ef l:le~:~neing EAF lecatiens. Prier te tl:le paries ef e}(fense9 eperatien Calla*n ay \*:ills1:11:lmit te tl=te NRC fer appreval a finalizes list ef l:le~:~n9iA9 EAF lecatiens wl=ticl:l will l:le menitere9 fer EAF *.viti:! tl:le Fati91:1e MenitoFing pregram. Tl=le s1:1ppertin9 P" calc~:~latiens will l:le perfermes 'NitA NUREG!CR 69()9 er NUREG/GR 6983 fer samen ane lew alley steels, to ULNRC-05860 Page 12 of 29 NUREGICR 6909 ar NUReG/CR 5704 for al:lsteRitiG staiRiess steels, aRel NUReG/CR e9Q9 far Rickel alloys.

The CUF for wetted. RCPB locations were categorized based on the strain-rate of the dominant transient. The strain-rate classification was determined with a qualitative assessment based on experience and not a quantitative stress analysis. The estimated strain rate was used to calculate an estimated F!l.!l. The estimated FO!l value was also calculated assuming a low DO environment: the maximum fluid/metal temperature; and the maximum sulfur concentration. and is based on the methods in NUREG/CR-5704 for austenitic stainless steels. NUREG/CR-6583 for carbon and low alloy steels. and NUREG/CR-6909 for Ni-Cr-Fe steels. This estimated F~

was then averaged with the maximum FM for that material tvpe to calculate the average F!!!!A The average F!l!l and the design basis CUFs were used to calculate the estimated EAF CUF.

These estimated EAF CUFs were then organized according to their system, thermal zone. and material tvpe. A thermal zone is defined as a collection of piping and/or vessel components which undergo essentially the same group of thermal and pressure transients during plant operations. The maximum EAF CUF for each thermal zone and material was selected as a sentinel location. In addition, if the next highest EAF CUF with the same thermal zone and material is within 50% of the maximum. additional locations were identified as sentinel locations.

This initial list was reduced further using EPRI Technical Report 1024995 [Ref. 181.

  • One Thermal Zone can bound another Thermal Zone in a System:

Both the CUF and Fgn values for one sentinel location in one thermal zone are each higher than the CUF and F~m values for the sentinel locations in other thermal zones.

  • One material in a Thermal Zone can bound other materials in the same Thermal Zone:

This circumstance could be achieved if within the same thermal zone. both the CUF and f&n values for one sentinel location composed of one material are each higher than the CUF and F!!l values for the sentinel locations composed for all other materials.

  • One material in a Thermal Zone can bound other materials in another Thermal Zone:

This circumstance combines the guidelines of the two listed above and must satisfy both criteria listed.

  • A location with EAF CUF < 1.0 may be removed from the sentinel location list:

If the sentinel location EAF CUF for the projected number of design cycles is low (e.g ..

EAF CUF < 0.25). that sentinel location may be removed from the final list due to the small likelihood that it will be the leading sentinel location in a system. If, however. the sentinel location EAF CUF for the projected number of design cycles is fairly high (e.g .. EAF CUF

> 0.8). the possibilitv exists that it could remain the sentinel location for its group and should be included in the monitoring program that ensures that it does not exceed a value of 1.0.

Table 4.3-7 identifies the final locations. including the NUREG/CR-6260 locations. that will be used as sentinel locations during the period of extended operation to manage the EAF aging mechanism. Those non-NUREG/CR-6260 locations with an EAF CUF greater than 1.0 will be evaluated further using the same methods as those used to remove conservatisms for the NUREG/CR-6260 locations described above. The results of these final analyses will be incorporated into the Fatigue Monitoring program by either counting the transients assumed or incorporate the stress intensities into a CBF ability of the program. As an alternative. the Fatigue Monitoring program may implement SBFs of certain locations in order to ensure the to ULNRC-05860 Page 13 of 29 component does not exceed an EAF CUF of 1.0. Any use of SBF will be implemented in compliance with RIS 2008-30. Therefore. the effects of the reactor coolant environment on the non-NUREG/CR-6260 locations will be managed for the period of extended operation. These TLAAs are dispositioned in accordance with 10 CFR 54.21Ccl(1)(iii).

Disposition: Aging Management, 10 CFR 54.21(c)(1)(1ii) to ULNRC-05860 Page 14 of29 Table 4.3-6 Summary of Fatigue Usage Factors at NUREGICR 6260 Sample Locations Location Material CUF . Fen EAFCUF CUF-Fen Basis SA533, Design basis CUF RPV Bottom Head to Shell Junction Grade B, Class 1, 0.0070 2.45 0.01715 Low Alloy Steel NUREG/CR-6583 maximum Fen SA 508, Class 2, Design basis CUF RPV Inlet Nozzle 0.0795 2.45 0.195 Low Alloy Steel NUREG/CR-6583 maximum F..,.

SA 508, Class 2, Design basis CUF RPV Outlet Nozzle 0.1078 2.45 0.264 Low Alloy Steel NUREG/CR-6583 maximum Fen CUF re-evaluated with NB-3200 methods based on 60 year cycle SA 182, Type 316, projections, Hot Leg Surge Line Nozzle 0.07572 10.097 0.7646 Stainless Steel NUREG/CR-5704 strain-rate dependent F...,

CUF re-evaluated with SBF and Charging System Nozzle SA 182 Type 316, 60 year cycle projections, 0.0919 I 0.0782 6.22/6.75 0.5715/0.5273

[Normal and Altemate] Stainless Steel NUREG/CR-5704 strain-rate dependent F..,

CUF re-evaluated with NB-3200 methods based on 60 year cycle Safety Injection Nozzle [Boron Injection SA 182 Type 316, projections, 0.1135 6.495 0.7374 Header nozzles] Stainless Steel NUREG/CR-5704 strain-rate deDendentFen CUF re-evaluated with NB-3200 Residual Heat Removal Inlet Nozzle SA 182 Type 316, methods based on design cycles, 0.0234 15.35 0.3591

[RHR nozzle-hot-leg) Stainless Steel NUREG/CR-5704 maximum F -

to ULNRC-05860 Page 15 of 29 Table 4.3-7: Sentinel Locations for EAF Monitoring Thermal NUREG Design Avg. Est System Material Coml!2nent Zone /CR-6260 CUF &n EAFCUF

1. RPV Outlet Nozzle y_ 0.1078 2.455 0.265 RPV Nozzles LAS
2. RPV Inlet Nozzle y_ 0.0795 2.455 0.195 Reactor RPV U1212er Pressure Head ss 3. CETNA U1212er Nozzle Housing N 0.37 13.117 4.853 Vessel LAS 4. Bottom Head-to-Shell Junction y_ 0.007 2.455 0.017 RPV Bottom Head Ni-Cr-
5. Bottom Head Instrument Tubes N 0.3184 4.093 1.303 Fe ss 6. Pressurizer Heater Penetration N 0.562 13.117 7.372
7. Pressurizer Shell at Su12gort Lug N 0.992 2.455 2.435 PZR Lower Head 8. Pressurizer Surge Nozzle N 0.963 2.455 2.364 LAS
9. Pressurizer Lower Head/Su12120rt N 0.734 2.455 1.802 Pressurizer Skirt PZR S12ray ss 10. Pressurizer S12ray Nozzle N 0.411 9.013 3.704
11. Safe!)£ and Relief Valve Pi12ing N 0.975 11.486 11.199 PZR SRV/PORV ss 12. Power 012erated Relief Valve Solenoid N 0.68 11.486 7.811 Surge Pi12ing Surge Line ss 13. Hot Leg Surge Nozzle y_ 0.3 11.486 3.446

- -* --- -- - - -- ---- -~ -- ~

Enclosure 1 to ULNRC-05860 Page 16 of29 Table 4.3-7: Sentinel Locations for EAF Monitoring Thermal NUREG Design Avg. Est Sl£Stem Material Component ~

Zone "

. /CR-6260 CUF Em EAFCUF

14. Normal Charging Nozzles, Loo12 1 y_ 0.90 7.240 6.516 Charging ss eves 15. Alternate Charging Nozzles, Loo12 4 y_ 0.90 7.240 6.516 Auxilia!Y S...l2r~

ss 16. Auxilia!Y S12ray Pi12ing N 0.72 5.970 4.298 RCS Cold 17. RCP Casing/Discharge Nozzle RCS Leg ss Junction N 0.915 9.628 8.810 RHR Inlet RHR (Suction) ss 18. RHR Nozzles, Hot Leg Loo12s 1 & 4 y_ 0.81 10.350 8.384 BIT ss 19. BIT Nozzles {All Loo12s} y_ 0.999 7.811 7.803 Sl Accumulator ss 20. Accumulator Nozzles (All Loo12s} N 0.95 7.811 7.420 SIS ss 21. Hot L~ SIS Nozzles, Loo12s 2 & 3 N QJ. 10.350 1.035 Steam 22. RSG Tubesheet {Continuous Tubesheet LAS N 0.428 2.455 1.051 Generator Rea ion)

Tast-e 4.3 7 PFe!iHiiRary !eieRtifi6ati9R efAEIEtttieRa! SeRtiRe! LesatieRS fer eAF R8G81111R8Rded Candidate Tha.......,.4!1ll7ftn.a ,... C!:a.n+in.al I "-"'"!!.+iftn.e!>

RPV Neule DD\/ t .... ln.+ "-1-.,,1_ DD\J ~~~~~~~- l:"ll:n _ I AC RPV Nenle DD\1 r"\oo+l ..... + 1\J ..... ,.,.J,.., DD\/ ft..J.- .... 1- _ .t:"').t:n I AC Reaeter 1 RPV Upper loleael 1 RPV Cere EMit Tl=lerR'Ieeeyple Nei!!i!!le AsseR'II:IIy Upper Nenle loleysiRg I RPV Upper loleael SS I PressYre~ \Jessel

  • ......... ' RPV Upper loleael RPV Vessel FlaRge I RPV Upper loleael lAS RPV 8etteR'I loleael ~2~9 RPV 8ettem loleael DD\/ D-.++-.- Un ....... +- C).,.-11 J, * .,..,..,,.,._

~

i<

(

II I-Ii I ~

It I~

Enclosure 1 to ULNRC-05860 Page 19 of29 Callaway Plant License Renewal Appllcation Amendment2 Revision to Section 4.6.

Section 4.6 (page 4.6-1) is revised as follows (deleted text shown with strikethrough, new text undertined):

4.6 CONTAINMENT LINER PLATE, METAL CONTAINMENTS, AND PENETRATIONS FATIGUE ANALYSES The Callaway prestressed concrete containment vessel is designed to Bechtel Topical Report BC-TOP-5-A, Revision 3. It is poured against a steel membrane liner designed to BC-TOP-1 Revision 1. No credit is taken for the liner for the pressure design of the containment vessel, but the liner and penetrations ensure the vessel is leak-tight.

The Callaway containment liner and e#lef:-metal containment (MC) components, e.g.

containment penetrations, were designed to stress limit criteria of BC-TOP-1 Revision 1 (Part I and Part II respectively), independent of the number of load cycles, and require no fatigue analyses with the exception of the main steam and feedwater penetrations, the containment access hatches, and the leak chases. For the MC containment penetrations. BC TOP-1 . Part II provides guidance on how to satisfy the ASME Section Ill. Division I NE requirements.

Enclosure 1 to ULNRC-05860 Page 20 of29 CaUaway Plant License Renewal Appllcation Amendment2 Revision to Section 4.7.2.

Section 4.7.2 (page 4.7-4) is revised as follows (deleted text shown with strikethrough, new text underlined):

4. 7.2 In-service Flaw Analyses that Demonstrate Structural Integrity for 40 years In-service flaw growth is Identified in NUREG-1800 as a potential TLAA. Flaws of such size that they cannot be dispositioned through comparison with the Code tables must be analyzed.

These analyses depend on a specified number of operating events or years, and thus may be TLAAs.

A search of the CLB did not identify any flaws evaluated for the remaining life of the plant other than those identified below.

  • Cold Leg Elbow-to-Safe End Weld Flaw Indications During the Refuel 13 (Spring 2004 ), two flaw indications were identified in the cold leg elbow-to-safe end weld. The weld and base metal material for the subject weld is stainless steel. The safe end is forged stainless steel (SA-182 F316). The weld is a stainless steel weld (ER308). The elbow is statically cast stainless steel (SA-351 CF8A, which is the same as wrought Type 304).

Flaw Indication #1: The flaw was an embedded flaw that was found acceptable in accordance with IWB-3500 (Acceptance Standards), but was conservatively treated as inside diameter surface breaking flaw (Reference 10). The depth of flaw #1 is 0.49 in.

Including inspection uncertainty. This represents 21.1 percent of the local pipe wall.

The flaw length is 4. 75 in. (5.1 percent of circumference based on nominal diameter).

Flaw Indication #2: The flaw was an inside diameter surface breaking flaw that was found to be greater than the size allowed by IWB-3500 (Reference 10). The depth of flaw #2 was found to be 0.94 in. including inspection uncertainty. This represents 40.5 percent of the local pipe wall. The length of the flaw was determined to be 2.625 in. (2.8 percent of circumference based on nominal diameter).

The root cause evaluation determined that these flaws were formed during initial plant construction. Low cycle fatigue, such as that experienced during pressurization, heatup and cooldown, caused flaw #2 to break through. Subsequent volumetric examinations did not identify any degFadatieR propagation in either of the flaws. These flaws will continue to be inspected through the lSI program at regular 10-year intervals after the two remaining followup inspections.

The continued operation with these flaws in place was justified in accordance with IWB-3640, which is supported by a flaw evaluation. The evaluation concluded that wide margin exists for both flaws, which allows further services throughout the remainder of

Enclosure 1 to ULNRC-05860 Page 21 of 29 the plant design life, as long as the same plant operation conditions as those considered in the analysis are maintained.

The evaluation of the two flaws considers two possible modes of failure. A fatigue crack growth analysis is used to demonstrate that the crack will satisfy the IWB-3640 requirements for the remainder of the plant life. A fracture mechanic analysis was also performed to predict crack instability.

Fracture Mechanics Operation with a crack is acceptable with respect to unstable ductile tearing mechanism if the applied J-integral remains below the J 1c fracture toughness. The only aging mechanism that affects the criteria is thermal aging. The forged safe end material is not subject to thermal aging. The gas tungsten arc welds are subject to thermal aging, but the effects are considered negligible. The fracture mechanics analysis does not consider aging effects and is not a TLAA, by 10 CFR 54.3(a), Criterion 2.

Fatigue Crack Growth The analysis procedure involves postulating an initial flaw at start of life and predicting the flaw growth due to an imposed series of loading transients. The incremental growth is then added to the original crack size, and the analysis proceeds to the next cycle or transient. The procedure is continued in this manner until all of the analytical transients known to occur have been analyzed. The transients considered were distributed evenly over the plant design life, with the exception of the preoperational tests, which are considered first. The design numbers of transients assumed to occur over the plant life are consistent with those of FSAR Table 3.9(N)-1 SP. As long as the plant design basis numbers of transients are maintained the same as those considered in the analysis, regardless of whether the transients occur over a 40 or 60-year plant life, the analysis and conclusions will remain valid.

This fatigue growth analysis does not consider intergranular stress corrosion cracking as a credible aging mechanism. This is based on stress corrosion cracking having been observed to occur in stainless steel in operating BWR piping systems, but not in PWR plants due to hydrogen overpressure. While the RPV inlet and outlet nozzles are Alloy 600 and the nozzle-to-safe end welds are Alloy 182, the cracks were Identified in the safe end-to-elbow region which is a gas tungsten-arc process, with a root pass TIG weld and does not contain any susceptible material. The analyses are only applicable in this region, and primary water stress corrosion cracking (PWSCC) was not considered a viable mechanism because expert and industry experience indicate stainless steels have been shown to be very resistant to PWSCC.

The projected transient accumulations in Table 4.3-2, Transient Accumulations and Projections show that the numbers of transient cycles are expected to remain within the assumed numbers and therefore the analyses are valid through the period of extended operation. This TLAA is dispositioned in accordance with 10 CFR 54.21 (c)(1 )(i).

Disposition: Validation, 10 CFR 54.21(c)(1)(1)

Enclosure 1 to ULNRC-05860 Page 22 of29 Callaway Plant License Renewal Application Amendment2 Revision to Section 4.7.4.

Section 4.7.4 (pages 4.7-8 and 9) is revised as follows (deleted text shown with strikethrough, new text underlined):

4.7.4 AbaanGe ef a Tlft.A fer Reactor Vassel Underclad Cracking Analyses NUREG-1800 Identifies "lntergranular separation in the heat-affected zone (HAZ) of reactor vessel low-alloy steel under austenitic SS cladding" as a potential TLAA. Ne sush sFasks have been EtisGO'IeFeEI, nor therefoFe, analyzeEt at Calla*Nay. In the a9sonse of any analyses no TbMs eMist. This phenomenon has been addressed in the Callaway vessel by weld cladding processes designed to avoid these defects, consistent with Regulatory Guide 1.43. In addition, WCAP-15338-A has demonstrated that the vessel integrity is maintained in the presence of underclad cracks. No such cracks have been discovered at Callaway.

Regulatory Guide 1.43 states that underclad cracking has been reported only in forgings and plate material of SA-508 Class 2 when clad using "high-heat-input" processes such as the submerged-arc wide-strip and the submerged-arc 6-wire processes. Cracking was not observed in SA-508 Class 2 materials clad by "low-heat-input" processes controlled to minimize heating of the base metal. Further, cracking was not observed in clad SA-533 Grade B Class 1 plate material, regardless of the welding process used.

Callaway VtJsstll Matsrlal SubjtJCt to Und11rclad Cracking The vessel shell and head plates are vacuum treated SA-533, Grade A, B, or C, Class 1 or 2 (Grade A orB, Class 1 for beltline plates). Only the vessel nozzles and flanges are SA-508 Class 2 forgings. The cladding is stainless steel weld metal, Analysis A-8; and Ni-Cr-Fe Weld Metal, F-Number 43.

The Callaway lSI pFegFaFR eMamines flanges unEier I'PJ8 Table :l5QQ 1 Category 8 A using GaEta Case N 623 , anEI e><aFRines R¥ nozzles unEter Categery 8 [) l:lsing CoEte Case N 648 1. A review ef inser:vlse inspestien reperts feunEI ne r:eserEI ef inEiisatiens of l:lnEieFGiaEI sFasking in the RV nozzles or flanges .

Qualification of Clad Wtllding ProctJSstls to A void Undtlrclad Cracking Although the Callaway vessel contains these SA-508 forgings clad by high-heat-input processes, freedom from underclad cracking is assured by special evaluation of the procedure qualification for cladding applied on low alloy steel (SA-508, Class 2).

This special evaluation is documented in FSAR SP, Appendix 3A and determined that Callaway meets the requirement of Regulatory Guide 1.43 by requiring qualification of any "high heat input" processes, such as the submerged arc wide strip welding process and the submerged-arc 6-wire process used on ASME SA-508, Class 2, material, with a performance test as described in Regulatory Position C.2 of the guide. No qualifiCations are required by the regulatory guide

Enclosure 1 to ULNRC-05860 Page 23 of29 for ASME SA-533 material and equivalent chemistry for forging grade ASME SA-508, Class 3, material.

The fabricator monitors and records the weld parameters to verify agreement with the parameters established by the procedure qualification as stated in Regulatory Position C.3.

Stainless steel weld cladding of low-alloy steel components is not employed on components outside the NSSS.

Applicability of Wt~stlnghousll Own11rs Group Gllnllrlc 60- Y11ar Flaw Growth Analysis Westinghouse prepared a topical report on underclad cracking. WCAP-15338-A (Ref. 191. which included fatigue crack growth analyses and ASME Section XI allowable flaw size evaluations for typical Westinghouse vessels, and found that the expected maximum flaw predicted by the crack growth analysis is less than the Section XI allowable flaw size. These WCAP-15338-A analyses assumed 1.5 times the numbers of cyclic and transient loads assumed for the original 40 year life. and demonstrated that these effects are acceptable for a 60 year life.

The NRC safety evaluation of this topical report determined that it might be incorporated by reference in a license renewal application, provided that the analysis is applicable to the applicant's plant. The licensee must demonstrate that the vessel will withstand growth of underclad cracks for a 60 year life by (1) verifying that the design cycles and transients assumed in WCAP-15338-A bound the cycles for 60 years of operation. and (2) providing a description of the programs and activities for managing the effects of aging and the evaluation of TLAAs for the period of extended operation . However, no l:IRdeFGiad srasks have &Jeen dissevered and this analysis is rull invoked in the Calla..,*Jay CbQ, therefore it is net a TlM 9y 1Q C~R 54 .3(a) sFiterien 9.

For Callaway (1) the numbers of transient cycles assumed in the WCAP-15338-A bound the projected cycles in 60 years presented in Table 4.3-2. (2) LRA Appendix A3. the FSAR supplement. provides a description of the evaluation of TLAAs for the period of extended operation.

In conclusion. at Callaway. the WCAP-15338-A addresses the aging mechanism of Underclad Cracking. WCAP-15338-A is a TLAA. which is dispositioned in accordance with 10 CFR 54.21 (c)(1 )(i).

Disposition: Validation. 10 CFR 54.21(c)(1)(1) to ULNRC-05860 Page 24 of29 Callaway Plant License Renewal Application Amendment2 Revision to Section 4.8 to identify new references.

Section 4.8 (page 4.8-2) is revised as follows (deleted text shown with strike through, new text underlined):

4.8 REFERENCES

1. Westinghouse Report WCAP-15400-NP. Analysis of Capsule X from the Ameren UE Callaway Unit 1 Reactor Vessel Surveillance Program. Rev. 0. June 2000. Westinghouse Non-Proprietary Class 3.
2. Callaway PTLR. "Callaway Plant Pressure and Temperature Limits Report." Rev. 5.

Released 11. December 2006.

3. Westinghouse Report. WCAP-17168-NP. Callaway Unit 1 Time-Limited Aging Analysis on Reactor Vessel Integrity. Rev. 0. September 2010. Westinghouse Non-Proprietary Class 3.
4. SIA Calculation 0900694.301. "Environmentally-Assisted Fatigue (EAF) for Callaway."

Rev. 0. Structural Integrity Associates, Inc. San Jose, California. 19 August 2010.

5. SIA Calculation 0901271.315. "Residual Heat Removal (RHR) Inlet Nozzle Environmentally-Assisted Fatigue Analysis Calculation." Rev. 0. Structural Integrity Associates, Inc. San Jose, California. 11 August 201 0.
6. SIA Calculation 0901271 .332. "Charging Nozzle Environmentally-Assisted Fatigue (EAF)

Analysis Using 60-Year of Operation Using Stress Based-Fatigue (SBF) Results from the Baseline Evaluation." Rev. 0. Structural Integrity Associates, Inc. San Jose, California.

27 October 2011.

7. SIA Calculation 0901271.331. "Safety Injection (BIT) Nozzle Environmentally-Assisted Fatigue (EAF) Analysis Using 60-Year Projected Numbers of Cycles." Rev 0. Structural Integrity Associates, Inc. San Jose, California. 16 September 2011 ..
8. SIA Calculation 0901271.330. "Hot Leg Surge Nozzle Environmentally-Assisted Fatigue (EAF) Analysis Using 60-Year Projected Numbers of Cycles." Rev. 0. Structural Integrity Associates, Inc. San Jose, California. 15 September 2011.
9. Precision Surveillance Corporation Document No. CA-N 1042-500. Final Report of the 25th Year IWL Inspection. Rev. 0. 16 September 2010. Supplemented by Callaway CAR 201009644.
10. Ameren Missouri Letter ULNRC-51 00. "Docket Number 50-483, Union Electric Company Callaway Plant, Transmittal of lnservice Inspection Summary Report for Refuel 13, and WCAP-16280-P, 'Flaw Evaluation Handbook For Callaway Unit 1 Reactor Vessel Inlet to ULNRC-05860 Page 25 of29 Nozzle Safe-End Weld Region,' May 2004." 13 December 2004. (ADAMS Accession No ML043650441 ).
11. Ameren Missouri Calculation BB-183. "Evaluation of Reactor Vessel Cladding Indication Inside Bottom Head During Refuel 13." Rev. 1.
12. Westinghouse Topical Report WCAP-15666-A. Extension of Reactor Coolant Pump Motor Flywheel Examination. Rev. 1. October 2003.
13. Ameren Missouri Letter ULNRC-05553. Graessle, Luke H. "Docket Number 50-483 Callaway Plant Unit 1 Union Electric Co. Facility Operating License NPF-30 Follow-Up Information Regarding 10 CFR 50.55a Request: Proposed Alternative to ASME Section XI Requirements for Replacement of Class 3 Buried Piping (TAC No. MD6792)." Fulton, MO.

9 October 2008. (ADAMS Accession No ML082900027).

14. Ameren Missouri Letter ULNRC-05542. Graessle, Luke H. "Docket Number 50-483 Callaway Plant Unit 1 Union Electric Co. Facility Operating License NPF-30 Additional Information Regarding 10 CFR 50.55a Request: Proposed Alternative to ASME Section XI Requirements for Replacement of Class 3 Buried Piping (TAC No MD6792)." Fulton, MO.

15 September 2008. (ADAMS Accession No ML082630798).

15. SIA Report FP-CALL-310. Benchmarking of Charging Nozzle Stress-Based Fatigue. Rev.
0. San Jose, California: Structural Integrity Associates. 22 June 2011.
16. SIA Report FP CALL 304. Baseline Analysis of Callaway Plant Cycles and Fatigue Usage-Startup through 1/31/2011. Rev. 1. San Jose, California: Structural Integrity Associates. 13 October 2011.
17. SIA Report FP-CALL-307. "Environmentally-Assisted Fatigue Screening." Rev. 2. San Jose. California: Structural Integrity Associates. 30 Apri12012.
18. EPRI Technical Report 1024995. "Environmentally-Assisted Fatigue Screening. Process and Technical Basis for Identifying EAF Limiting Locations."
19. WOG Topical Report WCAP-15338-A. A Review of Cracking Associated with Weld Deposited Cladding in Operating PWR Plants. Westinghouse Electric Company LLC.

October 2002.

to ULNRC-05860 Page 26 of29 Callaway Plant License Renewal Appllcation Amendment2 Revision to Section A3.2.3 to show completion of commitment.

Section A3.2.3 (page A-27) is revised as follows (deleted text shown with strike through, new text undertined):

A3.2.3 Effects of the Reactor Coolant System Environment on Fatigue Life of Piping and Components (Generic Safety Issue 190)

All of the locations specified in NUREG/CR-6260 for newer vintage Westinghouse plants will be monitored by the Fatigue Monitoring program, described in Section A2.1 . If any of the analyzed CUF values for these locations exceeds the fatigue design limit, the analyses may be revised using actual plant transients experienced. Callaway wimtas completed an evaluation fGFto identifv any additional plant-specific bounding EAF locations prier te the periee ef el<teRded eperatieRs . The supporting environmental factors, Fan. calculations will be performed with NUREG/CR-6909 or NUREG/CR-6583 for carbon and low alloy steels, NUREG/CR-6909 or NUREG/CR-5704 for austenitic stainless steels, and NUREG/CR-6909 for nickel alloys.

Therefere the effects of the reactor coolant environment on fatigue usage factors in the remaiRiRgNUREG/CR-6260 and plant-specific bounding EAF locations will be managed for the period of extended operation. These TLAAs are dispositioned in accordance with 10 CFR 54.21 (c)(1 )(Iii).

Enclosure 1 to ULNRC-05860 Page 27 of29 Callaway Plant License Renewal Appllcation Amendmentl Revision to Section A3.6.4, A3.6.5, A3.6.6, A3.6. 7, A3.6.8. and A3.6.9 to incorporate new Section A3.6.4 and renumber remaining sections.

Sections A3.2.4, 5, 6, 7, 8, and 9 (pages A-34 and 35) is revised as follows (deleted text shown with strike through, new text underlined):

A3.6.4 Reactor Vessel Underclad Cracking Analyses Reactor Vessel Underclad Cracking been addressed in the Callaway vessel by weld cladding processes designed to avoid these defects. consistent with Regulatory Guide 1.43. In addition WCAP-15338-A found that the maximum flaw predicted by the crack growth analysis is less than the Section XI allowable flaw size. These WCAP-15338-A analyses assumed 1.5 times the numbers of cyclic and transient loads assumed for the original 40 year life and bound the numbers of cycles projected in 60 years. This TLAA is disposition in accordance with 10 CFR 54.21 (c)(1 )(i).

A3.6.4.§. Reactor Coolant Pump Flywheel Fatigue Crack Growth Analysis Fatigue in the reactor coolant pump flywheels is supported by a fatigue crack growth analysis which demonstrates that 6,000 start-stop cycles (over an assumed 60 year life) will produce an acceptable extension of the crack. The evaluation is based on the 60-year operating period, therefore the TLAA extends to the end of the period of extended operation and the TLAA is dispositioned in accordance with 10 CFR 54.21 (c)(1 )(i).

A3.6.i§ High Energy Line Break Postulation Based on Fatigue Cumulative Usage Factors The selection of ASME Ill, Class 1 piping HELB locations depends on usage factors, which will remain valid as long as the assumed numbers of cycles are not exceeded. The Fatigue Monitoring program, summarized in Appendix B, Section A2.1, ensures that the analytical bases of the HELB locations are maintained or that a HELB analysis for the new locations with a CUF greater than 0.1 is performed. These TLAAs are dispositioned in accordance with 10 CFR 54.21(c)(1)(iil).

A3.6.GZ Fatigue Crack Growth Assessment In Support of a Fracture Mechanics Analysis for the Leak-Before-Break (LBB) Elimination of Dynamic Effects of Piping Failures Reactor Coolant Loops The fatigue crack growth analysis associated with the leak-before-break analyses depend on design transient cycle assumptions, and will remain valid as long as the assumed numbers of cycles are not exceeded. The projected transient accumulations show that the numbers of transient cycles are expected to remain within the assumed numbers and therefore the analyses

Enclosure 1 to ULNRC-05860 Page 28 of29 will remain valid for the period of extended operation. Therefore, these TLAAs are dispositioned in accordance with 10 CFR 54.21 (c)(1 )(i).

Accumulator Injection and Residual Heat Removal Lines These analyses are based on assumed 40 year design transients. The projected transient accumulations are expected to remain within the assumed numbers and therefore the analyses will remain valid for the period of extended operation. Therefore, these TLAAs are dispositioned in accordance with 10 CFR 54.21 (c)(1 )(i).

A3.6.-1§ Replacement Class 3 Burled Piping The replacement of buried Essential Service Water (ESW) piping with high-density polyethylene (HOPE) material began in 2008 with a service life of 40 years, which extends beyond the period of extended operation. Therefore the design of buried HOPE ESW piping will remain valid for the period of extended operation, and the TLAA is dispositioned in accordance with 10 CFR 54.21 (c)(1 )(i).

A3.6.8! Replacement Steam Generator Tube Wear The replacement steam generator tube wear analysis determined the maximum wear for a 45-year design life. The 45-year design life of the replacement steam generator tubes extends beyond the period of extended operation. Therefore, the design of the replacement steam generator tubes is valid through the period of extended operation and the TLAA is dis positioned in accordance with 10 CFR 54.21 (c)( 1)(i).

to ULNRC-05860 Page 29 of29 CaRaway Plant License Renewal Application Amendment2 Revision to Section A4, Table A4-1 to revise Fatigue Monitoring commitments.

Sections A4, Table A4-1, Item 37 (page A-49) is revised as follows (new shown text under1ined):

A4 LICENSE RENEWAL COMMITMENTS Table A4-1 identifies proposed actions committed to by Ameren Missouri for the Callaway Plant Unit 1 in its License Renewal Application. These and other actions are proposed regulatory commitments. This list will be revised, as necessary, in subsequent amendments to reflect changes resulting from NRC questions and Ameren Missouri responses. Ameren Missouri will utilize the commitment tracking system to track regulatory commitments.

Table A4-1 Ucense Renewal Commitments Item# Commitment LRA Implementation

~

Section Schedule 37 Complete an evaluation to determine if there are any additional plant-specific bounding EAF 4.3.2.2 Prior to the period of locations. The supporting environmental factors, F(en), calculations will be performed with 4.3.4 extended operation NUREG/CR-6909 or NUREG/CR-6583 for carbon and low alloy steels, NUREG/CR-6909 or NUREG/CR-5704 for austenitic stainless steels, and NUREG/CR-6909 for nickel alloys.

(Completed Amendment 2)

In order to determine if the pressurizer contains a limiting EAF location, the fatigue analyses will be revised to incorporate the affect effect of insurge-outsurge transients on the pressurizer lower head, surge nozzle, and heater well nozzles at plant specific conditions.

<Completed Amendment 2)

Those non-NUREG/CR-6260 locations with an EAF CUF greater than 1.0 will be further evaluated using same methods as those used for NUREG/CR-6260 locations to remove conservatisms from the prelimina~ EAF CUF. The results of these final anal)lses will be incomorated into the Fatigue Monitoring program b)l either counting the transients assumed or inco!:PQrate the stress intensities into a CBF abili!)l of the program. As an alternative, the Fatigue Monitoring prQgram will implement SBFs of certain locations in order to ensure the component does not exceed an EAF CUF of 1.0. An)l use of SBF will be implemented in compliance with RIS 2008-30.

The pressurizer contains a limiting EAF location. The fatigue anal)lses will be revised to incoroorate the effect of insurge-outsurge transients in the ~ressurizer lower head. I to ULNRC-05860 Page 1 of4 ENCLOSURE2 AMR Changes for Callaway Plant Unit 1 License Renewal Application Amendment No. 2 to ULNRC-05860 Page 2 of4 Callaway Plant License Renewal Application Amendment2 Revision to Table 3.2.2-5 to delete the aging evaluation lines of stainless steel valve with intended function of LBS.

Table 3.2.2-5 (pages 3.2-64 and 65) are revised as foUows (deleted text shown in strikethrough):

Table 3.2.2-5 Engineered Safety Features - Summary of Aging Management Evaluation - High Pressure Coolant Injection System (Continued.

Component Intended I Material Type Function Environment Aging Effect Requiring Aging Management NUREG-1801 Table 1 Program I Item Item Notes Manag_ement I

Valve b8S StaiR Steel less Atmespl:!ere!

'A!eatAer (~)

I CraskiR~ ~emal S~rfases MeRiteriRg ef V.D1.eP 1Q3 j3.2.1.QQ7 JA Mesl:!aRisal

' ' " ' - - - - - - -* - I D'l~

Valve kBS StaiRiess ~.tmespl:!ere!

1bess ef material ElEtemal S~rfases V.D1.eP 1Q7 3.2.1.QQ4 lA Steel WeatAer(~) MeRiteriRg ef Mesl:!aRisal I ~I'V'n<>nt .. / D'l '1. 'li \

Valve lb8S CraskiRg 'Nater Cl:!emistry IV111.81.SP 98 3.4.1.011 lA (82.1.2) aREI ORe Time

  • ID'l~

Valve bess ef material 'liJater Cl:!emistry VIII.81 .SP 155 13.4.1.019 lA

'!:BS i

(82.1.2) aREI ORe Time l l"',r_.".aMi"'"' /D") -1 <1 Q\

to ULNRC-05860 Page 3 of4 Callaway Plant License Renewal Appllcation Amendment2 Revision to Section 3.3.2.1.5 to add Atmosphere/Weather as an environment in the Service Water System.

Section 3.3.2.1.5 (page 3.3-8) is revised as follows (deleted text shown in strikethrough and new text shown underllned):

3.3.2.1.5 Service Water System Environment The service water system components are exposed to the following environments:

  • Atmosphere/Weather
  • Borated Water Leakage
  • Buried
  • Plant Indoor Air
  • RawWater to ULNRC-05860 Page 4 of4 Callaway Plant License Renewal Application Amendment2 Revision to Table 3.3.2-5 to add Carbon Steel Piping in Atmosphere/Weather and change the external environment of the Ductile Iron Valve to Atmosphere Weather.

Table 3.3.2-5 (page 3.3-100 and 3.3-103) is revised as follows (deleted text shown in strikethrough and new text shown underlined):

3'*~-~-0 AUXIIIBry \:>ysrems - \:>Ummarv or Aama Manaaemenr t::vatuauon - ;::;ervtce warer .sysrem Component Intended Material Environment Aging Effect Aging Management NUREG-1801 Table11tem Notes Type Function Requiring Program Item Manaaement . -* I Piping PB Carbon Steel Atmosphere/ Loss of material External Surfaces IVII.I.A-78 ,3.3.1.078 8 Weather {Ext} Monitoring of Mechanical Comoonents782.1.21)

Valve PB Ductile Iron PlaRl IRdaaF Loss of material Extern8J Surfaces ¥11.1.,6, 77VII.I.A- 3.3.1 .078 A AA Monitoring of Mechanical 78

~tmosph Components (82.1.21 )

ere/ Weather ILExt1 I I

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