L-2018-166, WCAP-15355-NP, Revision 0, a Demonstration of Applicability of ASME Code Case N-481 to the Primary Loop Pump Casings

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WCAP-15355-NP, Revision 0, a Demonstration of Applicability of ASME Code Case N-481 to the Primary Loop Pump Casings
ML18296A025
Person / Time
Site: Turkey Point  
Issue date: 10/31/2018
From: Abbott S
Westinghouse, Westinghouse
To:
Office of Nuclear Reactor Regulation
References
L-2018-166 WCAP-15355-NP, Rev 0
Download: ML18296A025 (36)
v • d • e


Text

Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 FPL Response to NRC RAI No. 4.7.5-1 L-2018-166 Attachment 7 Enclosure 2 Page 1 of 36 WCAP-15355-NP Revision 0, "A Demonstration of Applicability of ASME Code Case N-481 to the Primary Loop Pump Casings of the Turkey Point Units 3 and 4" October 2018

L-2018-166 Attachment 7 Enclosure 2 Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 Withheld from Public Disclosure Under 1 O CFR 2.390 Page 2 of 36 WCAP-15355-N P Revision O Westinghouse Non-Proprietary Class 3 A Demonstration of Applicability of ASME Code Case N-481 to the Primary Loop Pump Casings of the Turkey Point Units 3 and 4 October 2018

~-----

Westirtghoose

  • _. This record was final approved on 10/11/2018 9:13:12 AM. (This statement was added by the PRIME system upon its validation)

Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 L-2018-166 Attachment 7 Enclosure 2 Withheld from Public Disclosure Under 10 CFR 2.390 Page 3 of 36 Westinghouse Non-Proprietary Class 3 WCAP-15355-NP Revision 0 A Demonstration of Applicability of ASME Code Case N-481 to the Primary Loop Pump Casings of the Turkey Point Units 3 and 4 October 2018 Preparer:

Stephan L. Abbott*

Structural Design & Analysis III Preparer:

George J. Demetri*

Structural Design & Analysis III Approved:

Lynn A. Patterson*, Manager Structural Design & Analysis III This non-proprietary version, WCAP-15355-NP, was prepared in October 2018. The various references as included herein are the same references as in the originally prepared document, WCAP-15355, dated January 2000 and do not coincide with current references.

  • Electronically approved records are authenticated in the electronic document management system.

Westinghouse Electric Company LLC 1000 Westinghouse Drive Cranberry Township, PA 16066, USA

© 2018 Westinghouse Electric Company LLC All Rights Reserved

      • This record was final approved on 10/11/2018 9:13:12 AM. (This statement was added by he PR ME system upon its validation)

Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 L-2018-166 Attachment 7 Enclosure 2 Withheld from Public Disclosure Under 10 CFR 2.390 Page 4 of 36 iii FOREWORD This document contains Westinghouse Electric Company LLC proprietary information and data which has been identified by brac:kets. Codingi-1 associated with the brac:kets sets forth the basis on which the information is considered proprietary. These codes are listed with their meanings in WCAP-7211 (Reference 15).

The proprietary information and data contained in this report were obtained at considerable Westinghouse expense and its release could seriously affect our competitive position. This information is to be withheld from public discloswe in accordance with the Rules of Practice 10CFR2.790 and the information presented herein be safeguarded in accordance with 10CFR2.903. Withholding of this information does not adversely affect the public interest.

This information has been provided for your internal use only and should not be released to persons or mganizations outside the Directorate of Regulation and the ACRS without the express written approval of Westinghouse Electric Company LLC. Should it become necessary to release this information to such persons as part of the review procedure, please contact Westinghouse Electric Company LLC, which will make the necessary arrangements required to protect the Company's proprietary interests.

  • - This record was final approved on 10/11/2018 9:13:12 AM. (This statement was added by the PRIME system upon its valida ion)

Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 L-2018-166 Attachment 7 Enclosure 2 Withheld from Public Disclosure Under 10 CFR 2.390 Page 5 of 36 iv TABLE OF CONTENTS LIST OF TABLES............................................ -......................................................................................... V LIST OF FIGURES..................................................................................................................................... vi 1

BACKGROUND AND OBJECTIVE........................................................................................ 1-1

1.1 BACKGROUND

........................................................................................................... 1-1 1.2 OBJECilVE.................................................................................................................... 1-2 2

DESCRIPTION OF THE PRIMARY LOOP PUMP CASINGS OF TURKEY POINT UNITS 3 AND 4........................................................................................... 2-1 3

LOADS ON THE PUMP CASING NOZZLES....................................................................... 3-1 4

MATERIALCiiARACTERIZATION...................................................................................... 4-1 4.1 TENSILE PROPERI'IES................................................................................................ 4-1 4.2 FRACTURE TOUGHNESS PROPERTIES AND CRITERIA FOR THE PUMP CASINGS................................................................................................. 4-l

  • 5 STABILITY EVALUATIONS..................................................................................................... 5-1 5.1 SELECTION OF LOCATIONS FOR POSTULATED QUARTER TliICI<NESS CRACKS................................................................................................. 5-1 5.2 FLAW LOCATIONS FOR THE MODEL 93 PUMP CASING................................. 5-1 5.3 THE FINITE ELEMENT STRESS ANALYSIS MODELS OF THE PUMP CASINGS........................................................................................................... 5-1 5.4 STABILITY ANALYSIS................................................................................................ 5-2 6

FATIGUE CRACK GROWTI:I ASSESSMENT....................................................................... 6-1

6.1 INTRODUCTION

......................................................................................................... 6-1 6.2 DISCUSSION AND CONCLUSIONS........................................................................ 6-1 7

OPERATION AND STABILITY OF THE REACTOR COOLANT SYSTEM...................... 7-1 7.1 STRESSCORROSIONCRACKING........................................................................... 7-1 7.2 WATERHAMMER....................................................................................................... 7-2 7.3 LOW CYCLE AND lilGH CYCLE FATIGUE........................................................... 7-2 8

DISCUSSION AND CONCLUSIONS..................................................................................... 8-1 9

REFERENCES............................................................................................................................. 9-1 APPENDIX A............................................................................................... *......................................... A-1

  • - This record was final approved on 10/11/2018 9:13:12 AM. (This statement was added by the PRIME system upon its valida ion)

Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 V

Table3-l Table3-2 Table4-1 Table4-2 Table4-3 TableS-1 TableS-2 Table6-1 L-2018-166 Attachment 7 Enclosure 2 Withheld from Public Disclosure Under 1 O CFR 2.390 Page 6 of 36 UST OF TABLES Comparison of the Normal Loads for the Pump Casing Nozzles of Turkey Point Units 3 and 4 with the Screening Level C Normal Loads............... 3-3 Comparison of the Faulted Loads for the Pump Casing Nozzles of Turkey Point Units 3 and 4 with the Level A Faulted Screening Loads............... 3-3 Lower Bound Mechanical Properties for Turkey Point Units 3 and 4 Materials at Operating Teinperature.......................................................................... 4-2 Chemical Content & KCU Values for Turkey Point Unit 3..................................... 4-3 Chemical Content & KCU Values for Turkey Point Unit 4.................. """************** 4-4 Dimensions and Equivalent Stresses Associated with the Postulated Flaws in the Model 93 Pump Casings for the Turkey Point Units 3 and 4"Plants.............................................................................. 5-4 Stability Results for the Model 93 Pump Casings of Turkey Point Units 3 and 4 Plants.............................................................................. 5-5 Fatigue Crack Growth for Postulated Flaws in the Outlet Nozzle Kunckle Region of the Model 93 Pump Casings........................................ 6-2

      • This record was final approved on 10/11/2018 9:13: 12 AM. (This statement was added by the PRIME system upon its valida ion)

Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 Figure2-1 Figure4-1 Figure4-2 FigureS-1 L-2018-166 Attachment 7 Enclosure 2 Withheld from Public Disclosure Under 10 CFR 2.390 Page 7 of 36 vi UST OF FIGURES Dimensional Sketch of a Typical Model 93 Pump Casing with the Weld Seams Identified........................................................................................... 2-2 Representative Lower Bound True Stress -True Strain Curve for SA351 CFS at 55G°F for Turkey Point Units 3 and 4................................................. 4-5 Representative Lower Bound True Stress - True Strain Curve for SA351 CFS at 598°F for Turkey Point Units 3 and 4.......................................... -..... 4-6 Location of Flaws Postulated in the Pump Casing.................................................. 5-6

  • - This record was final approved on 10/11/2018 9: 13: 12 AM. (This statement was added by the PRIME system upon its valida ion)

L-2018-166 Attachment 7 Enclosure 2 Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 Withheld from Public Disclosure Under 10 CFR 2.390 Page 8 of 36 1-1 1

BACKGROUND AND OBJECTIVE

1.1 BACKGROUND

Periodic volumetric inspections of the welds of the primary loop pump casings of commercial nuclear power plants are required by Section XI of the ASME Boiler and Pressure Vessel Code (see Table lWB-2500-1, Examination Categories). These inspections are quite costly in terms of both dollars and radiation exposure (man-rem). To perform a volumetric inspection, complete disassembly of the pump is required. A lowering of the primary coolant water level most likely would be necessary which would, in tum, necessitate a complete core unload. Even then the volumetric inspection is very difficult. The pump casings are inspected twice prior to placing in service. When fabricated the castings are radiographed and liquid penetrant tested. After assembly, the welds are again radiographed and liquid penetrant tested. This in-shop examination is required per Section ill of the American Society of Mechanical Engineer's (ASME) Boiler and Pressure Vessel Code. The pre-service inspection criteria are the same as the in-service inspection criteria. Since no significant mechanisms exist for crack initiation and propagation, these criteria requiring that all welded surfaces be volumetrically and surface examined may not be warranted. In recognition of these facts the ASME Code body approved Code Case N-481 which provides an alternative to the volumetric inspection requirement (Reference 1 ').

The ASME Code Case, N-481 (Altemate Examination Requirements for Cast Austenitic Pump Casings), allows the replacement of volumetric examinations of primary loop pump casing welds with fracture mechanics based integrity evaluations (Item (d) of the code case) supplemented by specific visual inspections. It also :requires that a report of the evaluation be submitted to the regulatory and enfmcement authorities having jurisdiction at the plant site for review (Item (e) of the code case). A copy of the code case is given in Appendix A.

Following approval of Code Case N-481 by the ASME, the Westinghouse Owners Group sponsored the analyses required by the cod~ case which are applicable to the various primary loop pump casing models found in Westinghouse design nuclear steam supply systems. This work is documented in WCAP-13045 (Reference 2). Specifically, stress analyses for loadings on the pump casings were performed to support the fracture mechanics analyses for postulated flaws. Compliance to Item (d) of ASME Code Case N-481 was demonstrated on a generic basis.

However, a plant specific evaluation to demonstrate safety and serviceability is :required by Code Case N-481. Since there is a variety of pump casing models, loads and materials as discussed in WCAP-13045, it was not feasible to qualify each plant of Westinghouse design specifically to the requirements of the code case. Rather, enveloping or bounding criteria were set up whereby a specific utility, in most cases, needs only to show that the primary loop pump casings fall under the umbrella established by the analyses. The U.S. Nuclear Regulatory

'See Section 9.0 for a listing of referenc:es.

      • This record was final approved on 10/11/2018 9:13:12 AM. (This statement was added by the PRIME system upon its valida ion)

Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 1-2 L-2018-166 Attachment 7 Enclosure 2 Withheld from Public Disclosure Under 10 CFR 2.390 Page 9 of 36 Commission (U.S. NRC) has approved ASME Code Case N-481 in Revision 9 of Regulatory Guide 1.147, "Inservice Inspection Code Case Acceptability ASME Section XI Division I," dated April 1992.

1.2 OBJECTIVE It is the objective of this report to qualify the primary loop pump casings of the Turkey Point Units 3 and 4 to Item (d) of A5ME Code Case N-481 (Reference 1). H this report is supplemented by the visual inspections specified in the code case (Items a, b, and c),

compliance to the code case will be accomplished.

It is, also, the objective of this report to show the integrity of the primary loop pump casings to the ASME Code Case N-481 of the Turkey Point Units 3 and 4 for 40 year and 60 year plant life (as a part of the License Renewal Program).

-This record was final approved on 10/11/2018 9:13:12 AM. (This statement was added by the PRIME system upon its valida ion)

Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 L-2018-166 Attachment 7 Enclosure 2 Withheld from Public Disclosure Under 10 CFR 2.390 Page 1 O of 36 2

DESCRIYI'ION OF THE PRIMARY LOOP PUMP CASINGS OF TURKEY POINT UNITS 3 AND 4 The primary loop pump casings of Tmkey Point Units 3 and 4 are Westinghouse Model 93 design. The pump casings are fabricated from SA351 CFS cast stainless steel. A sketch of a typical pump casing of this type along with the weld locations is shown in Figure 2-1. This figure also contains typical dimensions.

2-1

  • -This record was final approved on 10/11/2018 9:13:12 AM. (This statement was added by the PRIME system upon its valida ion)

Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251

-:Weld Seam L-2018-166 Attachment 7 Enclosure 2 Withheld from Public Disclosure Under 10 CFR 2.390 Page 11 of 36

~1.21*-1 l~blAM,

,, - - - - Range Outer Quarter

/

-,-1-~--

Inna, Ouarte<

70.376" 68.2&* I

--t-1--- Nozzla lM8' au.....

Nozzle Outer Quarter Figure 2-1 DimensionaJ Sketch of a Typical Mode) 93 Pump Casing with the Weld Seams Identified

  • -This record was final approved on 10/11/2018 9:13:12 AM. (This statement was added by the PRIME system upon its valida ion)

Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 L-2018-166 Attachment 7 Enclosure 2 Withheld from Public Disclosure Under 10 CFR 2.390 Page 12 of 36 3-1 3

LOADS ON THE PUMP CASING NOZZLES In WCAP* 13045 enveloping axial force and moment loadings on the inlet and outlet pump casing nozzles were applied in the tluee--dimensional finite element analyses of the WOG plant pump casings (Reference 2).

Normal Loads The normal operating loads are calculated by the following equations:

F=Fow+FTH+Fp M., = <Mv)ow + (M...>m + (M...),

Mz: (MJDW + ~

+ (MJp The subsaipts of the above equations,:epresent the following loading cases:

DW deadweight TH =

normal therm;µ expansion P

=

load due to intemal pressure This method of combining loads is often referred as the al&ebraic sum method.

Faulted Loads The faulted loads are calculated by the absolute sum of loading components. The absolute summation of loads are shown in the following equations:

F =!Fowl+ l F,,J + IF,1 + IFss,;I M., = l<Mv).,.,I + l(My)THI + l(My)pl + l(My),i;J Mz = l(MJ.,.,I + l(MJTHI + l(MJ~ + l(MJ.I where subscript SSE means Safe Shutdown F.arthquake.

(3-1)

(3-2)

(3-3)

(3-4)

(3-5)

(3-6)

The bending moments for the desired loading combinations are calculated by the following equation:

(3-7) where M

=

bending moment for required loading My

=

Y component of bending moment

      • This record was final approved on 10/11/2018 9:13:12 AM. (This statement was added by the PRIME system upon its valida ion)

L-2018-166 Attachment 7 Enclosure 2 Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 Withheld from Public Disclosure Under 1 O CFR 2.390 Page 13 of 36 3-2

~ =

Z component of bending moment F

=

axial fmce NOTE: X axis is along the center line of pipe Summacy and Comparison of Loads The faulted nozzles loads (i.e., the normal plus safe shutdown earthquake nozzle loads) for the Turkey Point Units 3 and 4 pump casings are compared with the screening (i.e., enveloping) faulted loads. The normal loads are compared with screening no:nnal loads for evaluating the loss-of-load condition. The nonnal and faulted loads utilized here are based on the latest primary loop piping analyses on record and also shown in Reference 3.

In Table 3-1 the normal operating loads obtained, as mentioned above, for Turkey Point Units 3 and 4 are compared with the Level C screening nozzle loads (see Table 6-2 of WCAP-13045) which were used for evaluating the loss-of-load upset condition. The Turkey Point Units 3 and 4 normal moment at the inlet nozzle is seen to be bounded by the corresponding Level C screening moment. The Turkey Point Units 3 and 4 normal force at the inlet nozzle is not bounded by the corresponding Level C force. The Tur.key Point Units 3 and 4 normal moment at the outlet nozzle is seen to be bounded by the conesponding level screening moment. The Turkey Point Units 3 and 4 normal force at outlet nozzle is not bounded by the screening loads.

Also for normal case pressure of 2750 psia for Turkey Point Units 3 and 4 is higher than the screening pressure of 2635 psig used in Reference 2 Additional analyses were performed for the normal conditions as shown in Section 5.0.

The Turkey Point Units 3 and 4 faulted loads detemuned, as mentioned above, are compared in Table 3-2 to the Level A screening loads as defined in WCAP-13045 (see Table 6-2 of WCAP-13045). The Turkey Point Units 3 and 4 plant faulted force and moments at the inlet and outlet nozzles are bounded by the faulted screening loads. No additional analysis is required for the Level A loading case. Analysis performed in Reference 2, is conservatively applicable for the Turkey Point Units 3 and 4.

Temperature and Pressure Faulted Case Temperature = 547°F (Use 550° conservative)

Pressuie = 2250 psia Loss-of-Load CWL) Case (Reference 4)

Temperature = 598°F Pressure = 2750 psia

  • - This record was final approved on 10/11/2018 9: 13: 12 AM. (This statement was added by the PRIME system upon its valida ion)

L-2018-166 Attachment 7 Enclosure 2 Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 Withheld from Public Disclosure Under 10 CFR 2.390 Page 14 of 36 3-3 Table 3-1 Comparison of the Normal Loads for the Pump Casing Nozzles of Turkey Point Units 3 and 4 with the Screening Level C Normal Loads Inlet Nozzle Outlet Nozzle Temperature Force Moment Force Moment Load t°F)

(kips)

(In-kips)

(kips)

(in-kips)

Turkey Point Units 3 598 2202 8794 1664 6862 and4 Screening Level C 590 1900 23000 1400 8000 Note: As explained in WCAP-13045, the enveloping stresses were determined for the loss-of-load transient.

This was conservatively asswned as the limiting Level C transient.

Table 3-2 Comparison of the Faulted Loads for the Pump Casing Nozzles of Turkey Point Units 3 and 4 with the Level A Faulted Screening Loads Inlet Nozzle Outlet Nozzle Temperatun Force Moment Force Moment Load t°F)

(kips)

(in-kips)

Ckips)

(in-kips)

Turkey Point Units 3 550 182D 10705 1425 9(,0()

and4 Screening Level A S50 2000 40000 1800 20000

  • -This record was final approved on 10/11/2018 9:13:12 AM. (This statement was added by the PRIME system upon its valida ion)

Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 L-2018-166 Attachment 7 Enclosure 2 Withheld from Public Disclosure Under 10 CFR 2.390 Page 15 of 36 4-1 4

MATERIAL CHARACTERIZATION 4.1 TENSILE PROPERTIES The ASME Code material tensile properties were conservatively used for Turkey Point Units 3 and 4 to establish the tensile properties for the fracture mechanics analyses.

For the Turkey Point Units 3 and 4, the properties at 550°F and 598°F were required for the analyses. The lower bound properties at S50°F and 598°F were established from the tensile properties of the Section m 1989 ASME Boiler and Pressure Vessel Code (Reference 5). Code tensile properties at 550°F and S98°F were obtained by interpolating between the 500°F and 600°F tensile properties.

The lower bound yield strengths and ultimate strengths at operating temperatwes are given in Table 4-1. Modulus of elasticity values obtained from Reference 5 at 550°F and 598°F are also shown in Table 4-1. Poisson's Ratio used is 0.30.

For fracture evaluations the true stress-true strain curves for SA351 CF8 at the temperature of interest must be available. These curves were obtained from the Westinghouse tensile property database by using the information shown in Table 4-1. The lower bomtd true stress-true strain curves are given in Figures 4-1 and 4-2.

4.2 FRACTURE TOUGHNESS PROPERTIES AND CRITERIA FOR THE PUMP CASINGS The Turkey Point Units 3 and 4 pump casings are fabricated from SA351 CFS. This material has a 304 stainless steel chemistry and is not extremely susceptible to thermal aging degradation.

Values for the chemistry of each heat of material used in fabricating the Turkey Point Units 3 and 4 pump casings are taken £tom the certified. material test reports (CMTRs}. Predictions for the fully aged fracture toughness values are based on the material chemistry content Appendix A of Reference 2 contains the basis for determining the fracture toughness values for the generic evaluation. The Turkey Point Units 3 and 4 plant-specific values are used in determining the validity of these bounding values.

As described in Appendix A of Reference 2, the full-service Charpy U-notch impact energy of each heat of material is calculated based on the chemistry content These values are compared to a benchmark value of [

._ This record was final approved on 10/11/2018 9:13:12 AM. (This statement was added by the PRIME system upon its valida ion)

L-2018-166 Attachment 7 Enclosure 2 Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 Withheld from Public Disclosure Under 10 CFR 2.390 Page 16 of 36 4-2

.]......

[.

Table 4-1 Lower Bound Mechanical Properties for Turkey Point Units 3 and 4 Materials at Operating Temperature Temperature (°F)

Yield Sbength (psi)

Ultimate Strength (psi) 550 18800 598 18224 Modulus of Elasticity: E = 25.55E6 psi at SSD°F; E = 25.31E6 psi at S98°F.

Poisson's Ratio= 0.30 59300 59300

  • Note, the U.S. Nuclear Regulatmy Commission has found acceptable values J_ up to 3000 in-lb /in2 as indicated in Appendix A of Reference 2.
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L-2018-166 Attachment 7 Enclosure 2 Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 Withheld from Public Disclosure Under 1 O CFR 2.390 Page 17 of 36 4-3 Table 4-2 Chemical Content&: KCU Values for Turkey Poiqt Unit 3 a,c,e

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L-2018-166 Attachment 7 Enclosure 2 Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 Withheld from Public Disclosure Under 10 CFR 2.390 Page 18 of 36 4-4 Table 4-3 Chemical Content&: KCU Values for Turkey Point Unit 4 a,c,e

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Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 L-2018-166 Attachment 7 Enclosure 2 Withheld from Public Disclosure Under 1 O CFR 2.390 Page 19 of 36 Figure 4-1 Representative Lower Bound True Stress -True Strain Carve for SA351 CFS at 550°F for Turkey Point Units 3 and 4 4-5 a,c,e

      • This record was final approved on 10/11/2018 9:13:12 AM. (This statement was added by the PRIME system upon its valida ion)

Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 L-2018-166 Attachment 7 Enclosure 2 Withheld from Public Disclosure Under 10 CFR 2.390 Page 20 of 36 Figure 4-2 Representative Lower Bound True Stress - True Strain Curve fo~ SA351 CFS at 598°F for Turkey Point Units 3 and 4 a,c,e

  • - This record was final approved on 10/11/2018 9:13:12 AM. (This statement was added by the PRIME system upon its valida ion)

Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 L-2018-166 Attachment 7 Enclosure 2 Withheld from Public Disclosure Under 10 CFR 2.390 Page 21 of 36 5-1 5

STABILITY EVALUATIONS 5.1 SELECTION OF LOCATIONS FOR POSTULATED QUARTER THICKNESS CRACKS In this section selection of flaw locations for evaluation per ASME Code Case N--481 are desaibed. Three criteria for flaw location selections are applied as follows:

1.

for each weld, a flaw will be located in the highest stressed region; 2..

flaws will be located in regions of significant stress concentrations;

3.

flaws will be.located in welds not affected by discontinuities such as nozzles.

The selection of quarter thickness flaw locations and related information are given below for the finite element model.

5.2 FLAW LOCATIONS FOR THE MODEL 93 PUMP CASING Seven locations were selected for postulating quarter thickness flaws in the Model 93 pump casing. The flaws so selected have a 6 to 1 aspect ratio as required by ASME Code Case N-481 with one exception. The exception is the flaw selected at the outlet nozzle knuckle. An aspect ratio is not defined but the crack front curvature is representative of the aack front curvature for a crack having a 6 to 1 aspect ratio. Also for the outlet nozzle knuckle, the depth of the aaclc is taken as one-fourth the nominal casing wall thickness, not one-fourth the distance from the nozzle knuckle to the nozzle crotch.

The seven flaws are identified in Figure 5-1 of this report and the detailed description is giVen in Table 9-1 of Reference 2. Three of the postulated flaws are on the outside surface. Three locations were selected based on high stresses in weld regions. Four were selected based on.

high stress ~ncentrations and one selection was a nominally stressed weld location. One location (5-93) was selected for two :reasons - high stresses in a weld and highest stressed location in the pump casing.

5.3 mE FINITE ELEMENT STRESS ANALYSIS MODELS OF THE PUMP CASINGS Detailed stress analyses for Model 93 were per£ormed in Reference 2.. A large tluee-dimensional (3D) finite element model, containing the inlet and outlet nozzles, was developed for the pump casing.

Details of the finite element model are given in Figures 7-1 through 7-4 of Reference 2. For the complete 30 model, 30 isoparametric brick elements (20 nodes) and 3D isoparametric wedge elements (15 nodes) were used. Pipe extensions were made to the nozzles to allow continuous remote loadings of the nozzles.

  • - This record was final approved on 10/11/2018 9:13:12 AM. (This statement was added by the PRIME system upon its valida ion)

L-2018-166 Attachment 7 Enclosure 2 Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 Withheld from Public Disclosure Under 10 CFR 2.390 Page 22 of 36 5-2 For the Model 93 pump casing there are 1643 elements and 8729 individual nodes. Exterior details of the model are shown in Figures 7-1 and 7-2 of Reference 2. The coordinate directions are noted on Figure 7-1 (Reference 2). The support lugs are fully developed. The head is attached to the casing by a ring of small elemen~ simulating the ring of bolts as indicated in Figure 7-1 (Reference 2). The pump shaft hole was introduced to add flexibility to the head. An interior half view containing the outlet nozzle is shown in Figure 7-3 (Reference 2). Various sections of the outlet nozzle region are shown in Figure 7-4 (Reference 2).

Using the finite element stress analysis results from Reference 2, plant specific through-wall stresses for Turkey Point Units 3 and 4 at flaw locations were calculated for the normal loading condition.

S.4 STABILITY ANALYSIS In Section 3.0, it was determined that the normal fmces at the inlet and outlet nozzles of the Turkey Point Units 3 and 4 are not bounded by the respective screening loads of WCAP-13045.

Consequently, in this section, an analysis is conducted to determine whether enough margin is available to allow the Turkey Point Units 3 and 4 pump casings to meet the stability criteria.

As explained in Section 10.0 of WCAP-13045, a postulated flaw is stable if either:

1.

J"l'Plled < Jr. or

2.

HJ"l'Plled:.!:Jirthen T"l'Pllod <T-,andJIJ>llfei<Jmu The limiting material toughness values are listed in Section 4.0 of this report. They are:

In WCAP-13045, Jqpll,d and Tappllld values are calculated using the screening loads and the minimum ASME Code mechanical properties for SA351 CFS stainless steel. To determine if the flaw stability criteria is met for the Turkey Point Units 3 and 4 pump casings for the normal conditions, these parameters can be.recalculated using the Turkey Point Units 3 and 4 loads and material properties. This evaluation is conducted at the critical locations desmbed in WCAP-13045. These locations were 1-93, 2-93, 3-93, 4-93, 5-93, 6-93, and 7-93 and are shown in FigureS-1.

While there are extensive solutions for surface flaws in structures (e.g., Reference 6) assuming linear elastic behavior, there is no procedure comparable to that of the EPRI fracture mechanics handbook (Reference 7) available for elastic-plastic considerations. There are stresses in pump casings well in excess of yield stress; thus EPFM procedures are necessary. Such solutions are developed in three steps as discussed below.

This record was final approved on 10/11/2018 9: 13: 12 AM. (This statement was added by the PRIME system upon its valida ion)

Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 L-2018-166 Attachment 7 Enclosure 2 Withheld from Public Disclosure Under 10 CFR 2.390 Page 23 of 36 5-3 As a first step, a linear elastic fracture mechanics (LEFM) solution is obtained for a quarter thickness flaw having a six-to-one aspect ratio using the methodology of References 4 or other available solutions, as appropriate. For all but the nozzle knuckle the through-wall stress distribution is used and single curvature is accounted for (Reference 6). Actually this reduces the analysis to that of a cylinder. For the nozzle knuckle, the LEFM solution of Reference 8 applies. The knuclde-to-crokh stress distnbution is used, As a second step, the stress intensity factor, Ki, so obtained as above, is then evaluated for elastic-plastic behavior in the following manner. An LEFM solution is obtained for a cylinder with the same dimensions as in the first step with a quarter thickness internal surface continuous cin:wnferential flaw subjected to a constant stress loading. By interpolation, the constant stress level is then determined. which produces the same K1 as in the first step. The stress so determined. is called the equivalent stress. As a final step, J'l'P and T'l'P are found using the material curves (Figures 4-1 and 4-2) and properties given in Table 4-1 in conjunction with an EPFM model of the cylinder of the second step. That is, the cylinder is subjected to the equivalent stress using the EPFM solutions developed in Reference 7.

The temperature and dimensions associated with the postulated cracks are summarized in Table 5-1 for Turkey Point Units 3 and 4. The equivalent stresses shown in Table 5-1 for loading level Care associated with the Turkey Point Units 3 and 4 loading conditions. The equivalent stresses shown in Table 5-1 for loading Case A (faulted condition) are taken from Reference 2 and they are conservative for Turkey Point Units 3 and 4.

Table 5-2 shows stability results for Turkey Point Units 3 and 4. As shown in Table 5-2 all the stability criteria are met Therefore, it can be concluded. that flaws postulated in the Turkey Point Units 3 and 4 pump casing per Code Case N-481, when subject to the normal and faulted loadings are determined to be stable.

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Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 L-2018-166 Attachment 7 Enclosure 2 Withheld from Public Disclosure Under 1 O CFR 2.390

  • Page 24 of 36 Table S-1 Dimensions and Equivalent Stres~ Associated with the Postulated Flaws in a,c,e I Normal loads 2 Faulted Load the Model 93 Pump Casings for the Turkey Point Units 3 and 4 Plants
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Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 L-2018-166 Attachment 7 Enclosure 2 Withheld from Public Disclosure Under 10 CFR 2.390 Page 25 of 36 5-5 Table S-2. Stability Results for the Moc:lel 93 Pump Casings of Turkey Point Units 3 and 4 Plants a,c,e 1 NotApplicable,J.,,. <J"'

  • -This record was final approved on 10/11/2018 9:13:12 AM. (This statement was added by the PRIME system upon its valida ion)*

Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 5-6 L-2018-166 Attachment 7 Enclosure 2 Withheld from Public Disclosure Under 10 CFR 2.390 Page 26 of 36 a,c,e Figme S-1 Location of Flaws Postulated in the Pump Casing

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Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 L-2018-166 Attachment 7 Enclosure 2 Withheld from Public Disclosure Under 1 O CFR 2.390 Page 27 of 36 6-1 6

FATIGUE CRACK GROWTH ASSESSMENT

6.1 INTRODUCTION

In the stability analyses presented in the Section 5, cracks are postulated at various locations in the pump casings. Such postulated craclcs would be subject to the various cyclic conditions the pump casings experience. Thus, the sensitivity to cyclic loadings of postulated aaclcs in the pump casings was evaluated as a generic fatigue crack growth analysis for pump casing Model 93 in Section 12 of Reference 2.

The highest stressed location was chosen for the fatigue aack growth. This region is at Flaw 5-93. The postulated flaws are at the outlet nozzle knuckle in the plane of the weld. The stress contours for Level A loads are given in Figure 8-8 of Reference 2 and typify this location.

The generic transients considered for the fatigue crack growth are given in Table 12-2 of Reference 2. Recently Westinghouse perfonned an investigation (Reference 9) to predict the bansients and cycles for the 60 year plant life of Turkey Point Units 3 and 4 plants. This was done by reviewing the actual plant operating transient severity and the frequency of occurrences..Reference 10 shows the transients for the Turkey Point Units 3 and 4 uprating conditions. By reviewing the information documented in References 9 and 10, it is concluded that the typical design transients and cycles used for the Fatigue Crack Growth evaluation in Reference 2 can also be applied for the Turkey Point Units 3 and 4 pump casings for the 60 year plant life. The main conclusion derived from Reference 9 was that the number of cycles predicted for the 60 year plant life based on the operating experience are still less than the design basis cycles and the 40 year design basis transients bound the 60 year plant life transients. The significant design basis transients which will effect the Fatigue Crack Growth evaluation remain the same. Although there are some minor additional transients but they will have insignificant impact on the Fatigue Crack Growth evaluation performed for the Model 93 pump casing as documented in Reference 2. It is therefore, concluded that the Fatigue Crack Grown evaluation performed in Reference 2 for 40 year plant life is also applicable for the 60 year plant life of the Turkey Point Units 3 and 4 plants.

The fatigue crack growth results taken from Table 12-2 of Reference 2 are given in Table 6-1.

The maximum acceptable flaw size (0.30 in.) is seen to inaease by less than [

1*~

6.2 DISCUSSION AND CONCLUSIONS The highest stressed location in the Models 93 pump casing has been evaluated for fatigue crack growth. The crack growth observed are bounding for other less severely stressed locations. For 40 and 60 year plant life, postulated aack depths initially well in excess of the maximum ASME code allowable remain well less than the flaw sizes shown to be stable in Section 5.

  • -This record was final approved on 10/11/2018 9:13:12 AM. (This statement was added by the PRIME system upon its valida ion)

Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 6-2 L-2018-166 Attachment 7 Enclosure* 2 Withheld from Public Disclosure Under 10 CFR 2.390 Page 28 of 36 It is concluded that any reasonably sized flaws in the pump casings will exlul>it only minimal aack extension during service life of 60 years, such flaws remaining well below the flaw sizes shown to be stable.

Table 6-1 Fatigue Crack Growth for Postulated Flaws in the Outlet Nozzle Knuckle Region of the Model 93 Pump Casings a,c,e

'The maximum acceptable depth of a flaw per Table IWB 3518-2 of Section XI of lhe ASME Code (1989 Edition).

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Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 L-2018-166 Attachment 7 Enclosure 2 Withheld from Public Disclosure Under 1 O CFR 2.390 Page 29 of 36 7

OPERATION AND STABILITY OF fflE REACTOR COOLANT SYSTEM 7.1 STRESS CORROSION CRACKING 7-1 The Westinghouse reactor coolant system primary loops have an operating history that demonstrates the inherent operating stability characteristics of the design. This includes a low susceptibility to aacking failure from the effects of corrosion (e.g., intergranular stress corrosion aacking (IGSCC)). This operating histoi:y totals over 900 reactor-years, including five plants each having over 20 years of operation and 15 other plants each with over 15 years of operation.

In 1978, the United States Nuclear Regulatoi:y Commission (USNRC) formed the second Pipe Crack Study Group. (The fust Pipe Crack Study Group [PCSGJ established in 1975 addressed cracking in boiling water reactors only.) One of the objectives of the second PCSG was to include a review of the potential for stress corrosion cracking in Pressurized Water Reactors (PWR's). The results of the study performed by the PCSG were presented in NUREG-0531 (Reference 11) entitled "Investigation and Evaluation of Stress Corrosion Cracking in Piping of Light Water Reactor Plants." In that report the PCSG stated:

'The PCSG has determined that the potential for stress-corrosion aacking in PWR primary system piping is extremely low because the ingredients that produce IGSCC are not all present. The use of hydrazine additives and a hydrogen overpressure limit the oxygen in the coolant to very low levels. Other impurities that might cause stress-corrosion aacldng, such as halides or caustic, are also rigidly controlled. Only for brief periods during reactor shutdown when the coolant is exposed to the air and during the subsequent startup are conditions even marginally capable of producing stress-com,sjon aacking in the primary systems of PWRs. Operating experience in PWRs supports this determination. To date, no stress co.rrosion aacking has been reported in the primary piping or safe ends of any PWR."

During 1979, several instances of aacking in PWR feedwater piping led to the establishment of the thud PCSG. The investigations of the PCSG reported in NUREG-0691 (Reference 12) further confirmed that no occurrences of IGSCC have been reported for PWR primary coolant systems.

As stated above, for the Westinghouse plants there is no history of cracking failure in the reactor coolant system loop. The discussion below further qualifies the PCSG's findings.

For stress corrosion aacking (SCC) to occur in piping, the following three conditions must exist simultaneously: high tensile stresses, susceptible material, and a corrosive environment. Since some residual stresses and some degree of material susceptibility exist in any stainless steel piping, the potential for stress COITOSion is minimized by properly selecting a material immune to SCC as well as preventing the occurrence of a conosive environment. The material specifications consider compatibility with the system's operating environment (both internal

  • - This record was final approved on 10/11/2018 9:13:12 AM. (This statement was added by the PRIME system upon its valida ion)

L-2018-166 Attachment 7 Enclosure 2 Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 Withheld from Public Disclosure Under 10 CFR 2.390 Page 30 of 36 7-2 and external) as well as other material in the system, applicable ASME Code rules, fracture toughness, welding, fabrication, and processing.

The elements of a water environment known to inaease the suscepb'bility of austenitic stainless steel to stress corrosion are: oxygen, fluorides, chlorides, hydroxides, hydrogen peroxide, and reduced forms of sulfur (e.g., sulfides, sulfites, and thionates). Strict pipe cleaning standanis prior to operation and careful control of water chemistry during plant operation are used to prevent the oa:unence of a corrosive environment. Prior to being put into service, the piping is cleaned internally and externally. During flushes and preoperational testing, water chemistry is controlled in accordanc:e with written specifications. Requirements on chlorides, O.uorides, conductivity, and Ph are included in the aCt"eptance aiteria for the piping.

During plant operation,. the reactor coolant water chemistry is monitored and maintained within very specific: limits. Contaminant concentrations are kept below the thresholds lcnown to be conducive to stress corrosion. aacking with the major water chemistry control standards being included in the plant operating procedures as a condition for plant operation. For example, during normal power operation, oxygen concentration in the RCS is expected to be in the ppb range by controlling charging Oow chemistry and maintaining hydrogen in the reactor coolant at specified concentrations. Halogen concentrations are also stringently controlled by maintaining concentrations of chlorides and fluorides within the specified limits. Thus during plant operation, the likelihood of stress corrosion cracking is minimized.

7.2 WATER HAMMER Overall, there is a low potential for water hammer in the RCS since it is designed and operated to preclude the voiding condition in normally filled Jines. The reactor coolant system,.

including piping and primary components, is designed for normal, upset, emergency, and faulted condition transients. The design requimnents are conservative relative to both the number of transients and their severity. Relief valve actuation and the associated hydraulic transients following valve opening are considered in the system design. Other valve and pump actuations are relatively slow transients with no significant effect on the system dynamic loads.

To ensure dynamic system stability, reactor coolant parameters are stringently controlled.

Temperature during normal operation is maintained within a ruurow range by control rod position; pressme is controlled by pressum.er heaters and pressurizer spray also within a narrow range for steady-state conditions. The Oow characteristics of the system remain constant during a fuel cycle because the only governing parameters, namely system resistance and the reactor coolant pump characteristics, are controlled in the design process. Additionally, Westinghouse has instrumented typical reactor coolant systems to verify the flow and vibration characteristics of the system. Preoperational testing and operating experience have verified the Westinghouse approach. The operating transients of the RCS primary piping are such that no significant water hammer can occur.

7.3 LOW CYCLE AND HIGH CYCLE FATIGUE Low cycle fatigue considerations are accounted for in the design of the piping system through the fatigue usage factor evaluation to show compliance with the rules of Section m of the ASME

-This record was final approved on 10/11/2018 9:13:12 AM. (This statement was added by the PRIME system upon its valida ion)

Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 L-2018-166 Attachment 7 Enclosure 2 Withheld from Public Disclosure Under 10 CFR 2.390 Page 31 of 36 Code. A further assessment of the low cycle fatigue loadings was carried out as part of this study in the form of a fatigue aack growth assessment, as discussed in Section 6.0.

7-3 High cycle fatigue loads in the system would result primarily from pump vibrations. These are minimized by restrictions placed on shaft vibrations during hot functional testing and operation. During operation, an alarm signals the exceedance of the vibration limits. Field measurements have been made on a number of plants during hot functional testing. Stresses in the elbow below the :reactor coolant pump resulting from system vibration have been found to be very small, between 2 and 3 ksi at the highest. These stresses are well below the fatigue endurance limit for the material and would also result in an applied stress intensity factor below the threshold for fatigue aack growth.

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Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 L-2018-166 Attachment 7 Enclosure 2 Withheld from Public Disclosure Under 1 O CFR 2.390 Page 32 of 36 8-1 8

DISCUSSION AND CONCLUSIONS This report provides an assessment of the primary loop pump casings of Turkey Point Units 3 and 4 to the conditions of Item (d) of ASME Code Case N-481 (see Appendix A).

This evaluation considers actual Turkey Point Units 3 and 4 fracture toughness values. Thus Item (d) (1) is satisfied.

Stress analyses of a representative primary loop pump casing are presented in WCAP-13045.

(See also a description in Section 5.2 of this report.) This satisfies Item (d) (2).

The operating history of Westinghouse design primary loop pumps is reviewed in Section 7.

This satisfies Item (d) (3). Flaws are postulated in the pump casings as described in Section 5.1 (also see Section 9.0 ofWCAP-13045) satisfying Item (d) (4). One-quarter thickness reference flaws with a six-to-one aspect ratio are postulated consistent with Item (d) (5).

Comparisons of the loads of the Tur.key Point Units 3 and 4 pump casings with the screening loads of WCAP-13045 are presented in this report. The stability of the flaws postulated in the Turkey Point Units 3 and 4 primaiy loop pump casings are established by evaluating the resulting Jappllad and T ~

against the fracture toughness values noted in the discussion of Item (d) (1) (See Section 5.0). This satisfies Item (d) (6).

The preservice fracture toughness of cast stainless steels is very high. Thermal aging causes a reduction in the toughness. The effect of thermal aging has been evaluated in Section 4.0 of this report and Appendix A of WCAP-13045. No other mechanism is known to degrade the properties of the pump casings during service. Item (d) (7) is so satisfied.

It is concluded that the primary loop pump casings ofTurkey Point Units 3 and 4 are in compliance with Item (d) of ASME Code Case N-481 for both 40 year and 60 year plant life.

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Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 L-2018-166 Attachment 7 Enclosure 2 Withheld from Public Disclosure Under 10 CFR 2.390 Page 33 of 36 9-1 9

REFERENCES

1.

Case N-481: "Alternate Examination Requirements for Cast Austenitic Pump Casings,"

Section XI, Division 1, Cases of ASME Boiler and Pressure Vessel Code, Approval Date:

March 5, 1990.

2.

F. J. Witt and J. F. Petsche, Compliance to ASME Code Case N-481 of the Primary Loop Pump Casings of Westinghouse Type Nuclear Steam Supply Systems, WCAP-13045, September 1991 (Westinghouse Proprietary aass 2).

3.

WCAP-14237, December 1994, "Technical Justification for Eliminating Large Primaiy Loop Pipe Rupture as the Structural Design Basis for the Turkey Point Units 3 and 4 Nuclear Power Plants," (Westinghouse Proprietary Class 2).

4.

Equipment Specification 676335 Revision 1 dated October 9, 1967, Florida Power and Light Reactor Coolant Controlled Leakage Pump.

5.

ASME Boiler and Pressure Vessel Code Section III, "Rules for Construction of Nuclear Power Plant Components; Division 1-Appendices." 1989 Edition, July 1, 1989.

6.

Raju, I. S. and Newman, J. C., "Stress Intensity Factor Influence Coefficients for Jntemal and Extemal Surface Cracks in Cylindrical Vessels," in Aspects of Fracture Mechanics in Pressure Vessels and Piping, ASME publication PVP. Vol. 58, 1982

7.

Kumar, V., German, M. D. and Shih, C. P., "An Engineering Approach for Elastic-Plastic Fracture Analysis," EPRI Report NP-1931, Project 1237-1, Eledric: Power Research Institute, July 1981.

8.

EPRI-NP-719-SR, August 1978 and Errata for this report issued April 14, 1980.

9.

WCAP-15370, "Turkey Point Units 3 and 4 Design Basis Transient Evaluation for License Renewal," January 2000.

10.

WCAP-14291, Turkey Point Units 3 and 4 Uprating Engineering Report,"

December 1995.

11.

Investigation and Evaluation of Stress-Corrosion Cracking in Piping of Light Water Reactor Plants, NUREG-0531, U.S. Nuclear Regulatory Commission, February 1979.

12.

Investigation and Evaluation of Cracking Ind.dents in Piping in Pressurized Water Reactors, NUREG-0691, U.S. Nuclear Regulatory Commission, September 1980.

13.

O. K. Chopra, "Estunation of Fracture Toughness of Cast Stainless Steels During Thermal Aging in LWR Systems", NUREG-CR-4513, Revision 1, U. S. Nuclear Regulatory Commission, Washington, DC, August 1994.

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Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 9-2 L-2018-166 Attachment 7 Enclosure 2 Withheld from Public Disclosure Under 1 O CFR 2.390 Page 34 of 36 9

REFERENCES(cont)

14.

"Flaw Evaluation¢ Thermally aged Cast Stainless Steel in Light-Water Reactor Applications," Lee, S.; Kuo, P.T.; Wichman, K.; Chopra,0.; Published in International Joumal of Pressure Vessel and Piping, June 1997.

15. WCAP-7211, Revision 3, "Eneigy Systems Business Unit Policy and Procedures for Management, Oassification, and Release of Information," March, 1994.
  • -This record was final approved on 10/11/2018 9:13:12 AM. (This statement was added by the PRIME system upon its valida ion)

Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 L-2018-166 Attachment 7 Enclosure 2 Withheld from Public Disclosure Under 10 CFR 2.390 Page 35 of 36 A-1 APPENDIXA CASES OF ASME BOILER AND PRESSURE VESSEL CODE CaseN-481 Approval Date: Maiclt 5, 1990 See Numerical Index for expiration and any reajjirmatian dates.

Alternate Examination Requiremenls for Cast Austenitic Pump CasingsSection XI, Division 1 Inquiry:

Reply:

When conducting examination of cast austenitic pump casings in accon:lance with Section XI, Division 1, what examinations may be performed in lieu of the volumetric examinations specified in Table IWB-2500-1, Examination Category B-L-1, Item 812.10:

It is the opinion of the Committee that the following iequirements shall be met in lieu of performing the volumetric examination specified in Table IWB-2500-1, Examination Category B-L-1, Item B12.10:

(a)

Perform a VT-2 visual examination of the exterior of all pumps during the hydrostatic pi:essure test required by Table IWB-250~ 1, Category B-P.

(b)

Perform a VT-1 visual examination of the external surfai:es of the weld of one pump casing.

(c)

Perform a vr-3 visual examination of the intemal surfaces whenever a pump is disassembled for maintenance.

(d)

Perform an evaluation to demonstrate ~

safety and serviceability of the pump casing. The evaluation shall include the following:

(1) evaluating material properties, including fracture toughness values; (2) performing a stress analysis of the pump casing; (3) reviewing the operating history of the pump; (4) selecting locations for postulating flaws; (5) postulating one-quarter thickness reference flaw with a length six times its depth;

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Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 A-2 L-2018-166 Attachment 7 Enclosure 2 Withheld from Public Disclosure Under 10 CFR 2.390 Page 36 of 36 (6) establishing the stability of the selected Daw under the governing stress conditions; (7) considering thermal aging embrittlement and any other processes that may degrade the properties of the pump casing during service.

(e)

A report of this evaluation shall be submitted to the regulatory and enforcement authorities having jurisdiction at the plant site for review.

      • This record was final approved on 10/11/2018 9:13:12 AM. (This statement was added by the PRIME system upon its valida ion)
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