ML20092K987

Rev 3 to Shroud Stabilizer Hardware

ML20092K987
Person / Time
Site:	Peach Bottom
Issue date:	06/13/1995
From:	GENERAL ELECTRIC CO.
To:
Shared Package
ML20092K955	List: ML20092K980 ML20092K987
References
25A5579, 25A5579-R03, 25A5579-R3, NUDOCS 9509260229
Download: ML20092K987 (25)

Text

(, ' ' .

) gg 25A5579 SH NO.1 EIS IDENT: ' SHROUD STABILIZER HARDWARE

~

REVISION STATUS SHEET ,

DOCUMENT TITLE SHROUD STABILIZER HARDWARE ,

LEGEND OR DESCRIPTION OF GROUPS TYPE: DESIGN SPECIFICATION FMF: PEACH BOTTOM 2 AND 3 MPL NO: PRODUCT

SUMMARY

SEC. 7 l - DENOTE CHANGE THIS ITEM IS OR CONTAINS A SAFETYRELATED ITEM YES @ NO O EQUIP CLASS CODE C REVISION l A RM-01386 9/12/94 1 _ITROVATO 9/24/94 RJA CONTROL ISSUE RM-01502 CHK BY: ITROVATO 2 TL TROVATO 11/17/94 RTA CN01859 CHK BY: IL TROVATO 3 L FRENCH RfA l CN02716 l

l CHK BY: L FRENCH i

PRINTS TO l MADE BY APPROVALS GENERAL ELFLIKIC COMPANY -

l 175 CURTNER AVENUE J.L. TROVATO 8-22-94 M.O. LENZ 9-12-94 SAN _ JOSE CALIFORNIA 95125 CHKD BY: ISSUED 9-12-94 1.L.TROVATO 9-12-94 R.J. AHMANN CONT ON SHEET 2 MS WORD (3/28/94) DISK =

9509260229 950915 i

PDR ADOCK 05000277 p PDR ,

f 25A5579 SH NO. 2

l. SCOPE ,

1.1 This document defines the design and performance require nents for stabilizers for the core shroud which will functionally replace welds H1 through H7. The assumption of full 360 degree through wall cracking at the H8 weld shall be considered in the design analyses. A sketch of the welds and their nomenclature is given in Figure 1. All ASME Code requirements are given in the document of Paragraph 2.1.1.g. This specification herein contains those requirements that are not ASME Code requirements.

2. APPLICABIl. DOCUMENTS 2.1 General Electric Documents. The following documents fonn a part of the specification to the extent specified herein.

2.1.1 Supoorting Documents

a. Arc Welding of Austenitic Stainless Steel P50YP102

b. Sensitization Tests for Austenitic Stainless Steel, Modified ASTM -

E50YP13 A262 Practice E

c. Determination of Carbide Precipitation in Wrought Austenitic E50YP20 Stainless Steel (Modified ASTM A262 Practice A)

d. Examination for Intergranular Surface Attack E50YP11

e. Age Hardening of NI-CR-FE Alloy X750 P10JYP2 l

f. Liquid Penetrant Examination E50YP22 Shroud Stabilizers 25A5580 g.,

h. Reactor' Vessel Thermal Cycles 729E762 ..

i. Seismic Analysis of Peach Bottom 2 Reactor Vessel and Internals 383HA691

j. Peach Bottom 2,3 Power Rerate Analysis NEDC 32230P _j

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2.1.2 Suoptemental Documents. Documents under the following identities are to be used with ,

this specification *

a. Reactor Components . 3~83HA715

b. Essential Components 22A3041 2.2 Codes and Standards. The following documents of the latest issue (or specified issue) form a part of this specification to the extent specified herein.

2.2.1 American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV)

Code

a.Section III, Appendices,1989 Edition.

b.Section IX, Welding and Brazing Qualifications,1989 Edition.

c.Section III, Subsection NG,1989 Edition.

d.Section XI, Rules for Inservice Inspection,1980 Edition, Wmter 1981 Addenda.

2.2.2 American Society for Testing and Materials (ASTM)

a. ASTM A-182, Specification for Forged or Rolled Alloy-Steel Pipe Flanges, Forged Fittings, and Valves and Parts for High-Temperature.

b. ASTM A-240, Specification for Heat-Resisting Chromium and Chromium-Nickel Stainless Steel Plate, Sheet, and Strip for Pressure Vessels.

c. ASTM A-262, Detecting Susceptibility to Intergranular Attack in Stainless Steel.

d. ASTM A-479, Specification for Stainless and Heat-Resisting Steel Bars and Shapes for Use in Boilers and Other Pressure Vessels.

v

e. ASTM A-480, Specification for General Requirements for Flat Rolled Stainless and Heat-Resisting Steel Plate, Sheet, and Suip.

f. ASTM B-637, Specification for Precipitation Hardening Nickel Alloy Bars, Forgings, and .

Forging Stock for High-Temperature Senice.

2.3 PECO Enerev Documents i

a. UFSAR, Peach Bottom Units 2 and 3. l l

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& MNrusuarEr.: gy \*N1"

3. GENERAL DESCRIPTION 3.1 The purpose of the shroud stabilizers is to stmcturally replace welds H1 through H7. Welds .

H1 through H6 are all of the circumferential~ welds in the shroud, as well as the (H7) bimetallic attachment weld of the shroud to the shroud support. cylinder. These welds were required to both vertically and honzontally support the core top guide, core support plate, and siiroud head; and to prevent core flow bypass into the downcomer region. The core top guide and core support plate horizontally support the fuel assemblies and maintain the correct fuel channel spacing to permit control rod insertion. The design analyses shall consider full 360 degree through wall cracking at the H8 weld. The H8 weld connects the shroud support plate to the shroud support cylinder.

4. REQUIREMENTS 4.1 Code 4.1.1 The shroud stabilizer components are not classified as ASME Section III Code components.

However, material properties shall be obtained from the document in Paragraph 2.2.1.a, and l

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welding qualification shall be performed in accordance with the document in Paragraph 2.2.1.b.

The nomenclature for stress intensity used in this document is the same as that used in the documcat of Paragraph 2.2.1.c.

4.2 Structural Criteria ,

4.2.1 All structural analysis shall be performed in accordance with the criteria given in the Peach i Bottom UFSAR. All of the load combinations given in Paragraph 4.3.5 shall be shown to satisfy the primary stress limits given in Tables C.5.2 and C.5.6 of the Peach Bottom UFSAR, with values '

of SFmin as defined in Paragraph 4.3.6. The appropriate SFmin values have been incorporated into the allowable stress intensity values given in Paragraphs 4.2.1.1 and 4.2.1.2.

4 1 4.2.1.1 The primary stresses (Pm, P1, and Pl + Pb) in the existing shroud, during Normal and l Upset events, shall be shown to be less than Sm,1@Sm, and 1.5Sm respectively. During Emergency events, the allowable stresses are increased'by a factor of 1.5 times the values for Normal and Upset events. During Faulted events, the allowable stresses are increased by a factor ,

of 2.0 times the values for Normaland Upset evenu.

4.2.1.2 The stresses (Pm, Pm + Pb, and Pm + Pb + Q) inithe repair hardware, during Normal and Upset events, shall be shown to be less than Sm,1.5Sm, and 3.0Sm respectively. During Emergency events, the allowable primary stresses are inc'reased by a factor of 1.5 times the values for Normal and Upset events. During Faulted events, the allowable primary stresses are increased by a factor of 2.0 times the values for Normal and Upset events. Secondary stresses are notlimited during Emergency and Faulted events.

1 1

I O GENuclearEnergy 'N 'l" 4.2.2 The values of Sm and Sy as well as any other required material property shall be obtained ,

from the document in Paragraph 2.2.1.a (ASME Code,Section III Appendices), except for alloy -

X-750. The values of Sm and Sy for alloy X-750 at , operating temperature are 47,500 psi and 92,300 psi respectively. These values must be verified from the Certified Material Test Reports (CMTR's). The value of Sm must be determined using the method of Appendix III from the document of paragraph 2.2.1.a. If Certified Material Test Reports (CMTR's) are available, the value of Sm for XM-19, or for stainless steel may be determined using the method in Appendix III ,

of the document in Paragraph 2.2.1.a.

4.2.3 The maximum permanent deflection of any point on the shroud adjacent to either the H2 or the H3 weld shall be less than 2.1 inches divided by SFmin, during all of the load combinations specified in Paragraph 4.3.5. The maximum permanent deflection of any point on the shroud adjacent to either the H5 or H6 weld shall be less than 0.75 inch divided by SFmin, dudng ad of the load combinations specified in Paragraph 4.3.5. The maximum transient clastic deflection during the seismic event adjacent to either the H5 or H6 weld shall be less than 1.68 inch divided by SFmin specified in Paragraph 4.3.6. The allowable deflections are based on test data, and on Tables C.5.1 and C.5.5 of the Peach Bottom UFSAR.

4.3 Desien Requirements 4.3.1 General. The shroud repair hardware shall be designed to horizontally support the top guide, core support plate, the fuel assemblies and the shroud head. The shroud repair shall be designed to prevent upward displacement of the shroud. The shroud repair shall be designed for a life equal to the remaining design life of the plant plus possible life extension. The shroud repair shall be removable.

4.3.2 Soring Preload 4.3.2.1 Ingallation Preload. All of the springs shall be installed with a preload due to bending deflection greater than the deflection resulting from the limiting design upset condition, exclusive ofseismic events. The required installation spring bending preload is 0.05 inch for the upper springs and 0.01 inch for the lower springs.

'4.3.2.2 Preload Relaxation. The design shall consider an End-of-Life preload relaxation of 5%

for the upper springs near the H2 and H3 welds and a relaxation of 5% for the lower springs near ,

the H5 and H6 welds. Potential axial preload relaxation due to a reduction in shroud stiffness  !

resulting from assumed cracking at the H2, H3, H5, and H6 welds shall be considered.

l 4.3.3 Emironmental Conditions 4.3.3.1 Temocrature. The design temperature for the repair hardware is 550 degrees F. The operating temperature is 527 degrees F. Operating temperature shall be used for emergency and faulted evaluations.

ng 2 579 SH No. 6 U/ @8[gggggy, 4.3.3.2 Radiation. The maximum neutron radiation level (flux) at the shroud stabilizers in the ~

shroud vessel annulus is 4.8E10 neutrons /cm2/sec. This will not affect the properties of the stabilizer materials over the remaining life of the plant. .

4.3.4 Physical Interfaces 4.3.4.1 The shroud repair hardware shall restrain the shroud during all of the load combinations in Paragraph 4.3.5. The allowable permanent motion is dependent on the safety significance of the portion of the shroud under consideration. The allowable permanent motion for those portions of the shroud, which affect control rod insertion, is given in Paragraph 4.2.3. For the remaining portion of the sh:oud below H3, the allowable permanent motion is determined such that the reflooding of the inside of the shroud up to two thirds of core height is assured. For the portion of the shroud above H2, the allowable motion is 2.6 inches, which assures that the core spray lines are not impacted by the shroud.

4.3.4.2 The shroud repair hardware must provide features which facilitate handling during installation. The upper and lower springs shall be movable without removing the tie rod and without welding, .in order to permit inspection of the reactor pressure vessel with GERIS 2000.

4.3.4.3 The shroud repair hardware shall be designed and installed such that removal ofjet pump inlet mixers can be performed without removalof any of the repair hardware.

4.3.4.4 All parts shall be captured and held in place with a method that will last for the design life given in Paragraph 4.3.1.

4.3.5 Load Combinations. The load combinations that the shroud and shroud repair shall be analyzed for are from the Peach Bottom UFSAR. The limiting Upset event is a Design Basis Earthquake (DBE), plus Normal pressure differences, plus dead weight. The Emergency 1 event is a Maximum Credible Earthquake (MCE), plus Normal pressure differences, plus dead weight.

The Emergency 2 event is a main steam line LOCA, plus dead weight. The Faulted event is a Maximum Credible Earthquake (MCE), plus a main steam line LOCA, plus dead weight.

4.3.5.1 The pressure differences for these events are given in the table below. The pressure inside the shroud is higher than that outside of the shroud, and the pressure is higher below the core plate than above the core plate. These values include Power Rerate conditions based on

• 110% core flow, and 110% original power.

Comoonent' Normal Pressure LOCA Pressure Shroud Support Plate 33.03 psi 51.0 psi Core Plate 23.67 psi 28.0 psi Upper Shroud 9.35 psi 30.0 psi Shroud Head 9.41 psi 31.0 psi

4 sa xo. 7 U GE' h % L^55793 4.3.5.2 A new seismic analysis based on the documents in Paragraph 2.3 and 2.1.1.i shall be ,

performed which includes the shroud stabilizers. The shroud stabilizers shall function for the entire continuum from an uncracked shroud to a fully cracked shroud. Therefore, multiple conditions must be analyzed, for both the DBE and the MCE events. As a minimum, the following shroud conditions shall be analyzed an uncracked shroud with the installed stabilizers, a shroud with a through wall 360 degree crack at the H7 weld with the installed stabilizers, and a shroud with a through wall 360 degree crack at the H6 weld with the installed stabilizers. The limiting seismic loads on the stablizer are given in the table below:

Comoonent DEI hiCE Upper Spring 17,110 lb. 31,390 lb.

Lower Spring 33,440 lb. 92,900 lb.

Set of 4 Tie Rods (each) 75,640 lb. 117,600 lb.

4.3.5.3 Two steady state thermal conditions shall be evaluated. The first is Normal operation with the shroud at 539 degrees F, and the stabilizer assembly at 527 degrees F. The second condition is an Upset transient (scram with loss of feedwater pumps) with the shroud at 433 degrees F, and the stabilizer at 300 degrees F. The number of events is defined by 729E762 (document 2.1.1.h).

4.3.5.4 During the recirculation line LOCA event, there is a force applied to the shroud of 169,000 lbs, with a moment of 13.0E6 in-lb acting at the base of the shroud. This is due to asymmetric pressures in the annulus between the shroud and the RPV. This force exists for a sufficient time to be treated as a static force.

4.3.6 Reouired Safety Factors. The minimum safety factors (SFmin) shall be 2.25 for Normal and Upset events,1.5 for Emergency events, and 1.125 for Faulted events. These are based on Table C.5.5 of the Peach Bottom UFSAR.

4.4 Materials. ASTM specification material is acceptable for the Shroud Repair. CMTRs are ,

required for all mateda!.

4.4.1 The springs shall be made of nickel-chrome-iron alloy X-750 (UNS N07750). The cobalt content shall be limited to a maximum of 0.10%. Alloy X-750 shall be purchased per ASTM B-637 , ,

and age hardened per P10JYP2. Alloy X-750 material shall be tested per E50YP11. In lieu of testing per E50YP11, all finished components may incorporate the removal, after solution heat treatment, of a minimum of 0.030 inches of material from all surfaces of the original raw material form.

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25A5 79 SH NO. 8 L U,, GE%dserEnergy a 4.4.2 The tie rods may be made of either 304, 304L, 316, or 316L material with a maximum carbon content of 0.02%, and annealed at.1900 to 2100 degrees F followed by quenching in circulating water to a temperature below 400 degrees F. The tie rod material shall be tested per E50YP11 and E50YP20. The maximum hardness shall be RB90 for 304 and 304L. The maximum hardness shall be RB92 for 316 and 316L. XM 19 with a maximum carbon content of 0.04% may also be used for fabrication of the tie rods. XM-19 shall be annealed at 2,000 50 degrees F, followed by rapid cooling, and shall be tested per E50YP13, or per ASTM A-262 Practice E.

4.4.3 Other parts shall be made of any of the matedals listed in Paragraph 4.4. The filler matedal for any required weld buildups on 300 sedes stainless steel shall be Type 308L per P50YP102. All assembly welds shall satisfy P50YP102.

4.5 Leakage Due to Repair. Zero leakage is not required. However, the design shall control the normal operating condition leakage to preventcavitation of thejet pumps. The leakage after any required load combination shall be limited such that core flooding to 2/3 the height of the core is assured.

4.6 Inspections. Liquid penetrant examination shall be performed on all final machined surfaces of all stabilizer components, and on all structural welds in accordance with the requirements E50YP22A.

4.7 Fabrication 4.7.1 Welder and Weld Procedure Oualification. Welders and weld procedures shall be quahfied per the document in Paragraph 2.2.1.b. Welder qualifications shall include limited access similar to the actual welds to be completed.

4.7.2 Root Pass. The root pass of all full penetration single sided stainless steel weldedjoints shall be made by the GTAW process. Protective gas back-purging is required for all full penetration single sided weldedjoints until a minimum of 3/16 inch of weld thickness is completed.

4.7.3 Weld Surface Finish. All welds shall have the final outer surface suitable for liquid penetrant examination. The final surface shall meet the hardness requirements of Paragraph 4.4. ,

5. QUALITY ASSURANCE 5.1 The shroud repair hardware components are Safety Related as referenced in Paragraph 2.1.2.b, and design, fabrication, and installation activities shall be controlled per a QA Program which satisfies 10CFR50 Appendix B,in order to assure safe and reliable components.

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(f GENuedearfra vy 25A5580 REV.4 SH NO.1 EIS IDENT: SHROUD STABILIZERS ,

REVISION STATUS SHEET "

i DOC TITLE SHROUD STABILIZERS -

LEGEND OR DESCRIPTION OF GROUPS 7YPE: CODE DESIGN SPECIFICATION FMF: PEACH BOTTOM 2 AND 3  :

MPL NO: PRODUCT

SUMMARY

SEC. 7

• THIS ITEM IS OR CONTAINS A SAFETY RELATED ITEM YES DE NO O EQUIP CLASS CODE P_

REVISION C t A RM-01386 9/12/94 -

1 JTROVATO 9/24/94 RJA CONTROL ISSUE RM-01502 CHK BY:J TROVATO 2- JL TROVOTO RJA CN01859 .

CHK BY: JL TROVATO 3 L. FRENCH 6/13/95 RJA CN02716

' ~ ~~'

CHK BY: L. FRENCH i 4 M.BURT MN19 N RJA CN02808 ENGR: M.O.LENZ l I

PRINTS TO MADE BY APPROVALS GENERAL ELECTRIC COMPANY 175 CURTNER AVENUE J.L. TROVATO 8-22-94 M.O. LENZ 9-12 94 SANJOSE, CALIFORNIA 95125 CHK BY ISSUED 9-12-94 J.L. TROVATO 9-12 94 R.J. AHMANN CONT ON SHEET 2 MS-WORD

E baW 25A5580 SH NO. 2 REV.4

1. SCOPE ,

1.1 This document defines the ASME Code design requirements for the analysis of the -

reactor pressure vessel (RPV) as a result of the installation of the shroud stabilizers. The shroud stabilizers function to structurally replace the horizontal shroud welds H1 through '

H7, and will add new points of application for forces applied to the RPV. In addition, the analysis shall consider the assumption of full 360 degree through wall cracking at the H8 weld between the shroud support plate and the shroud support cylinder.

i

2. APPLICABLE DOCUMENTS '

2.1 General Electric Documents. The following documents form a part of this specification to the extent specified herein. ,

2.1.1. Supporting Documents 1

a. Reactor Vessel- Power Rerate 25A5341 Rev. O  !

b. Reactor Pressure Vessel, Purchase Specification 1A1111 Rev. 9

c. Reactor Vessel, Purchase Part  ;- 886D499 P2

i. 1 Sheet No. Revision No. I 1 11 2 8 3 3 4 6 5 3 6 4 7 6 8 0

d. Reactor Thermal Cycles 729E762 Rev. O

e. Nozzle Thermal Cycles 135B9990 Sheet No.

• Revision No.

1 1 2-8 0

f. Vessel Flange Bolting 885D911 Rev.2

g. Nozzle End Preparation 107C5305 Rev.2

h. Standard Requirements For Core Structure 21 A3319 Rev.1

O GENuclearEnergy 25as580 REV.4 su so. g

i. " Fatigue Evaluation of the Peach Bottom II and III Reacroe Vessels", G.E. Report No. GE- -

NE-523-61-0493, dated May 1993.

2.1.2 Supolemental Documents. Documents under the following identities are to be used with this specification:

a. Shroud Stabilizer Hardware Design Specification 25A5579 2.2 Codes and Standards. The following documents of the specified issue form a part of this specification to the extent specified herein.

2.2.1 American Society of Mechanical Encineers (ASME) Boiler and Pressure Vessel Code

a.Section III,1965 Edition and Addenda through Winter 1965

b.Section XI,1980 Edition and Addenda through Winter 1981 2.2.2 Other Documents

a. UFSAR, Peach Bottom 2 and 3

b. Shroud Support VPF 1896-064-7

c. Design Certification VPF 1896-142-1

d. Design Stress Report VPF 1896-146-1

e. Final Design Report VPF 1896148-2  ;

3. GENERAL DEFINITION j i

3.1 The purpose of the shroud stabilizers is to structumily replace all of the horizontal welds (H1 through H7) in the shroud. These welds were required to both horizontally and vertically support the core top guide, core support plate, and shroud head, and to prevent '

core bypass flow to the downcomer region. The core top guide and core support plate horizontally support the fuel assemblies and maintain the correct fuel channel spacing to permit control rod insertion, as well as having other structural functions. The H8 weld connects the shroud support plate with the shroud support cylinder. The analysis of the RPV _ ,

shall consider full 360 degree through wall cracking at the H8 weld.

3.2 All of the non ASME Code requirements for the shroud stabilizers are defined in the Document of Paragraph 2.1.2.a. The ASME Code requirements are defined herein.

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.rN U GENuclearEnergy 2 sass 80 REV.4 su no. 4

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4. REQUIREMENTS ,

4.1 The shroud stabilizer constniction shall be performed in accordance with a Section XI -

Replacement Program per the requirements of Article IWA-7000. The core shroud was not supplied as a ASME Code component. However,Section XI requires In Service Inspection (ISI) of the Core Support Structures. The r'equired Replacement Program is different than most Replacement Programs, because the stabilizers are not a direct replacement. Instead, the structural functions of the shroud horizontal welds are replaced by new components. Any defects found in the shroud horizontal welds are acceptable after the installation of the stabilizers.

1 4.2 The shroud stabilizers shall be constructed to the original Owners Requirements 1 (document of Paragraph 2.1.1.h) for the shroud, as there was no Code of Construction. l 4.3 The shroud stabilizers change the points of application of the forces applied to the reactor pressure vessel from the core shroud. These new forces shall be analyzed in  ;

accordance with the original Code of Construction (document in Paragraph 2.2.1.a).

4.4 The new forces and their points of application are defined in Figure 1, and in Table 1.

The values given in Figure 1, and in Table 1 shall be combined with the forces defined in the Design Specification (documents of Paragraphs 2.1.1.a through 2.1.1.e).

4.5 ' The original purchase specification for the reactor pressure vessel (document of  ;

Paragraph 2.1.1.b) specified that the boundary of jurisdiction of Section III of the ASME Code (document of Paragraph 2.2.1.a) shall include all attachments to the pressure boundary parts, but does not include the components that are welded to the attachments. Thus, the jurisdiction of the original Code of Construction included all weld build up pads used to attach internal components to the reactor pressure vessel, but did not include the shroud support within the bounday of Codejurisdiction. The bounday of ASME Codejurisdiction is shown in Figure 2.

4.6 The analysis required by this Design Specification shall be Certified.

5.0 PROFESSIONAL ENGINEER CERTIFICATION To the best of my knowledge and belief, this Design Specification satisfies the requirements of the ASME Boiler and Pressure Vessel Code 1965 Edition with Addenda through Winter 1965.

Signature:k 8* J g VESS/ Date: June 19, 1995 ~I No.22212 m E3 2221gd ca m ornia License Number: State:

• F CHA Cf..

jq O GENuedeerEnergy 25A5580 REV.4 SH NO. 5 ADDITIONAL DESIGN MECHANICAL LOADS Force DBE + Normal Pressure MCE + Normal Pressure MCE+LOCA F1 33,400 lbs 92,900 lbs 92,900 lbs F2 17,110 lbs 31,390 lbs 31,390 lbs F3 172,910 lbs 218,500 lbs 372,650 lbs F1, F2, and F 3are discrete loads applied over a small area. At any one point in time, F1 and F2are each applied to one location. At any one point in time, F3is applied to 4 locations 90*

apart for the installation of four shroud stabilizer assemblies. DBE is a Design Basis Earthquake (OBE). MCE is a Maximum Credible Earthquake (SSE).

For Normal Operation (d-pressure plus thermal loads) without consideration of seismic loads, F3 = 94,710 lbs.

For the limiting Upset thermal transient without consideration of seismic loads, F, = 249,985 lbs.

For a Main Steam Line LOCA without seismic loads, F, = 248,500 lbs. This must be addressed with the same allowables as the MCE + Normal Pressure load case.

The number of thennal cycles is defined by documents 2.1.1.d and 2.1.1.i.

The stress intensities shall meet the stress allowables of the ASME Code,Section III, for the load combinations defined by the Peach Bottom UFSAR. The original Code of Construction did not include Faulted load combinations. Faulted load combinations shall meet the stress allowables as defined by the Peach Bottom UFSAR for the reactor pressure vessel.

TABLE 1 i

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APPLICATION OF DESIGN MECHANICAL LOADS i FIGURE 1

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REV.4 FINAL

.e NON-CODE INTERNALLY - SECTION III CIASS A SUPPORTED STRUCTURE _

l REQUIRED BOUNDARY l OFJURISDICATION l l

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_ 581 SH NO.1 EIS IDENT:

REVISION STATUS SHEET DOC TITLE STABILIZER INSTALLATION LEGEND OR DESCRIPTION OF GROUPS TYPE: INSTALLATION SPEC FMF. PEACHBOTTOM 2,3

"" 8#*

MPL NO: PRODUCT

SUMMARY

SEC. 7 THIS ITEM IS OR CONTAINS A SAFE'IY RELATED ITEM YES E NO O EOUIP CLASS CODE P_

REVISION C

0 RM-01490 SEP 24 1994 RJA 1 C.F. SERMON 5/17/95 RJA CN02668 Added Sections 6.6 and 6.10 Changed section 6.26 2 M.BURT .393191995 RJA CN02808 ENGR: M.O.LENZ PRINTS TO MADE BY APPROVALS GENERAL ELECTRIC COMPANY 175 CURTNER AVENUE D.B. DRENDEL 8/29/94 D.B. DRENDEL 9/23/94 SAN JOSE, CALIFORNIA 95125 CHK BY ISSUED j

J.L. TROVATO 9/23/94 RJ. AHMANN SEP 241994 CONT ON SHEET 2 MS-WORD

GMW&a 25A5581 REV.2 SH NO. 2

1. SCOPE 1.1 Purnose. This specification provides the engineering requirements for installing stabilizers which replace the H1 through H7 horizontal shroud welds in the Peachbottom reactor assembly.

1.2 If any conflict exists between this document and any other document referenced herein, this document shall govern.

1.3 This document, along with the reactor modification and installation drawing, defines all the engineering requirements for installation of the shroud stabilizers.

1.4 As used herein, the term " Installer" refers to the company or personnel contracted by the Plant Owner to install the shroud stabilizers.

2. APPLICABLE DOCUMENTS 2.1 General Electric Documents. The following documents form a part of this specification to the extent specified herein.

2.1.1 Suncorting Documents

a. 25A5579, Shroud Stabilizers

b. 105E1455, Reactor (Modification & Installation)

c. 21 A2040, Cleaning and Cleanliness Control

d. D50YP5, Nickel-Graphite Thread Lubricant

e. Il2D6355, Contact, Lower

f. 112D6360, Lower Stabilizer (lower contact assembly)

g. I12D6357, Contact, Upper

h. I12D6347, Upper Stabilizer Assembly (upper spring assembly) ,

i. Il2D6349, Tie Rod Assembly

j. I12D6351, Spring, Lower

k. 'l12D6331, Ring, Mid Support

1. I12D6358, Tie Rod-Spring Assembly

m. I12D6356, Support, Mid

gg{g 25A5581 SH NO. 3 REV.2

n. 112D6359, Mid Support Assembly

o. I12D6490, Support, Lower

p. I12D6495, Toggle Bolt Assembly

q. I12D6493, Washer, Toggle Bolt-

r. I12D6494 Nut, Toggle Bolt

s. I12D6505, Pin, Clevis

t. I12D6348, Stabilizer Support Assembly

u. Il2D6752, Upper Suppor: Spacer

v. I12D6777, Nut, Tie Rod 2.1.2 Supolemental Documents

a. NEDC-31735P GE BWR Operator's Manual- Materials and Processes 2.2 Codes and Standards. The following codes and standards of the latest issue (or specified issue) form a part of this specification to the extent specified herein. l 2.2.1 American Societv of Mechanical Engineers (ASME) Boiler and Pressure Vessel  !

Code l

None

3. DESCRIPTION '

3.1 The purpose of the stabilizer installation is to structurally replace horizontal girth welds H1 through H7 in the shroud; weld designations and the design requirements for the stabilizers are defined in the 2.1.1.a design specification. The installation of the shroud stabilizers involves electric discharge machining (EDM) of some slots and holes in the existing structure, assembling the stabilizer hardware in the reactor, and preloading the threaded fasteners. No structural welding or defect removal by machining are involved.

4. RESPONSIBILITIES 4.1 The Installer shall accept full responsibility for his work. The Installer shall comply with the requirements of this document and the supporting documents listed herein.

4.2 The Installer shall take the responsibility for coordination of his work with the work of others including the coordination of work planning and radiation monitoring with the Plant Owner.

l gggpgW 25A5581 REV.2 SH NO. 4 4.3 The Installer shall be responsible for providing all specialized handling, alignment, and installation equipment, as may be necessary to perform this work, except as otherwise agreed to by the Plant Owner.

4.4 The Installer, except as otherwise agreed to by the Plant Owner, shall be responsible for machining as specified and limited by the applicable modification drawing.

4.5 The Installer shall supply adequately qualified personnel for supenision and for performing the tasks required to complete the stabilizer installation.

5 REQUIREMENTS 5.1 General 5.1.1 During installation, the Installer, except as otherwise agreed to by the Plant Owner, shall complete data sheets and quality control check sheets as required by the specifications and instructions listed in this document. The Installer shall also keep log notes, records, etc., for future reference. Video tapes shall be taken of the completed repair. Tabular data entries designated for as-built measurements on the installation drawing shall be recorded.

5.1.2 Procedures and installation equipment shall be developed and designed to minimize the potential ofloose parts within the Reactor Pressure Vessel (RPV).

5.1.3 Following completion of the installation of the stabilizers, verification, inspection and signoff shall be performed to ensure that all objects have been removed from the RPV.

5.1.4 All uncontaminated tools shall be stored in an uncontaminated controlled area and brought to the work area only as needed for fit-up and installation.

5.1.5 Refer to Paragraph 2.1.2.a for miscellaneous consumables approved for use in the reactor vessel.

5.2 EnscLnnel Safety 5.21 Radiation Control 5.2.i.1 All work shall be done with the concurrence of and per the instructions of the authoiind site Health Physics Personnel. At no time shall their requirements for dosimmer monitors, protective clothing or devices, time limits, exposure limits, etc., be violated.

5.2.1.2 Machining on contaminated surfaces, as required, shall be done in accordance with Health Physics and Safety Personnel requirements.

5.2.1.3 Radiation control practices shall be used to reduce exposure to workers to levels which are as low as reasonably achievable (ALARA).

ggggy 25A5581 SH NO. 5 REV.2

. 5.2.2 Safetv Precautions 5.2.2.1 Concern for personnel safety shall govern all work operations. All personnel working in hazardous locations shall be under constant surveillance by other personnel.

All electric equipment shall be grounded or double insulated. Welding cables and leads shall be in good condition. ,

5.2.2.2 All work areas shall be kept neat and orderly. Protective measures and devices shall be used to keep all tools, equipment, and materials from inadvertently dropping into the RPV.

5.2.2.3 Care shall be exercised to keep contamination of articles which must enter and leave contamination zones to a minimum. In all cases, site radiation control requirements shall be met.

5.3 Cleaning and Cleanliness Control 5 3.1 During this stabilizer installation program, cleaning and cleanliness control shall be in accordance with the document listed in paragraph 2.1.1.c. In addition, no graphite lead pencils are allowed to contact stainless steel and nickel alloys.

5.4 Prereauisites 5.4.1 Jet Pumn Throat Covers. Prior to the shroud stabilizer installationjet pump throat covers shall be installed as required.

5.4.2 Reactor Temperature. The reactor water temperature shall be less than 100"F, however the RHR shutdown cooling flow must be off whenever the installation activity in progress involves critical remote underwater handling in the annulus area.

6. INSTALLATION REQUIREMENTS 6.1 The installation sequence described below is not itself mandatory, so long as all specified installation requirements are accomplished. To assist in evaluating alternative sequences, the intent of some requirements, which are not self evident, are summarized in the step description.

6.2 Shroud head bolt (SHB) lug sets which straddle the 45,135, 225, and 315 degree azimuths, on the shroud, are specified, on the 105E1455 Modification and Installation drawing, for locating the stabilizer support installation and for machining shroud head flange slots. These SHB lugs shall be determined and independently verified as a prerequisite to any physical work at each of the four installation locations. Prior to removing the shroud head (SHBs may be unlatched), a common scribe line shall be

1 (s

) GENuclearEneryy y581 ,

sn NO.6 made on both the shroud and shroud head at each of the four installation locations, in accordance with the 105E1455 Modification and Installation drawing. This scribe line will then become the datum for locating the slots in the shroud head flange and installing the stabilizer support assemblies on the shroud flange.

6.3 Go-gage checks shall be performed on: the shroud flange and steam dam width for fit-up with the upper support (also checks for possible prior damage to the steam dam),

and the jet pump restrainer bracket to RPV inside diameter clearance, 5.4 inch minimum, to allow passage of the lower spring (temporarily ignoring the jet pump restrainer bracket guide plates).

6.4 Install protective shielding for the feedwater sparger and core spray line.

NOTE: The below step is a contingency, which will only be performed if there is insufficient clearance to complete the installation.

6.5 Machine (EDM) the jet pump restrainer brackets, if required, as shown on the 105E1455 Modification and Installation drawing. EDM swarf shall be captured to the maximum extent practical.

6.6 Measure and record the outside-to-outside distance between the SHB lug sets, as shown on the 105E1455 Modification and Installation drawing.

6.7 Measure and record the annulus width at the top guide support ring and at the core support ring elevations as shown on the 105E1455 Modification and Installation drawing.

Examine the RPV and shroud contact areas to assure that there are no abrupt discontinuities; if so, EDM spotface these areas flush. The vessel and shroud contact locations of the final stabilizer parts shall be simulated in taking these measurements.

CAUTION: Several piece parts are to be machined based on in-reactor measurements at a specific reactor azimuth. These parts shall then be designated by specific serial number, as recorded on the as-built data table on drawing 105E1455, for that specific azimuth.

6.8 Based on the in-reactor measurements, machine the RPV contact surface of the lower ,

contact, drawing Il2D6355, as shown on the 105E1455 Modification and Installation  :

drawing. Assemble the lower contact as shown on the lower stabilizer assembly, drawing ll2D6360. i 6.9 11ased on the in-reactor measurements, machine the RPV contact surface of the upper contact, drawing 112D6357, as shown on the 105E1455 Modification and Installation drawing. Assemble the upper contact as shown on the upper stabilizer assembly, drawing 112D6347.

I

em \

ggh{- 25A5581 SH NO. 7 I REV.2

, 6.10 Based on the in-reactor measurements of the outside-tocutside distance between j the SHB lug sets, and the measurement of the outside dimension of the stabilizer i support, machine the contact surface of the upper support spacer,112D6752, in ,

accordance with the 105E1455 Modification and Installation drawing. Assemble the upper support spacer,112D6752 on the stabilizer support assembly 112D6348.

6.11 Working in the equipment pool, locate the proper datum on the shroud head flange l as shown on the 105E1455 Modification and Installation drawing. Machine (EDM) slots  ;

in the shroud head flange as specified on the 105E1455 modification and installation drawing.

6.12 In accordance with the 105E1455 Modification and Installation drawing, machine [

(EDM) two holes in shroud support plate. EDM swarf shall be captured to the maximum extent practical.

6.13 Hone the holes in the shroud support plate. To assure the removal of microfissures from the EDM hole in the shroud support plate, the hone operation shall remove a  !

minimum of 0.005 inch from the inside surface of the hole while meeting the final hole size requirement on the 105E1455 Modification and Installation drawing.

6.14 Install lower support,112D6490, over the two shroud support plate holes using two toggle bolt assemblies,112D6495, and two toggle bolt washers,112D6493, and two toggle  ;

bolt nuts, ll2D6493, as shown on the 112D6494 Modification and Installation drawing. l Lubricant (D50YP5B) shall be applied to the threaded smfaces. Tension the two toggle  !

bolts to the specified load, and tighten the toggle bolt nuts. Inspect to verify the  !

installation of the lower support. Crimp the toggle bolt nuts, and inspect for proper '

crimping of the retainers.

6.15. Install the clevis pin,112D6505, in the mating hole of the lower support in accordance with the requirements of the 105E1455 Modification and Installation drawing.  ;

6.16 Complete the tie rod-spring assembly. Assemble the tie rod, assembly drawing ,

112D6349, with the lower spring, drawing 112D6351 (Drill pin hole and install lock pin.), and the lower stabilizer, drawing 112D6360 as shown on assembly drawing

. 112D6358. Lubricant (D50YP5B) shall be applied to the threaded surfaces.

6.17 Temporarily protect the exposed tie rod thread from damage. l CAUTION: Maneuvering of the tie rod-spring assembly must be done with extreme care to avoid damaging reactor  !

hardware such as thejet pump sensing lines.

6.18 Install the tie rod-spring assembly,112D6358, in accordance with the requirements of the 105E1455 Modification and Installation drawing. Maneuver lower spring clevis over clevis pin and support vertically. -

I l

ggM 2 581 SH NO. 8 6.19 Position the stabilizer support assembly,112D6348, over the tie rod. Lower the stabilizer support assembly over the steam dam and locate properly on shroud flange in accordance with the requirements of the 105E1455 Modification and Installation drawing.

6.20 Rotate and position the lower stabilizer assembly, ll2D6360, as shown on the 105E1455 modification and installation drawing. Verify that the' lower stabilizer assembly latch is engaged in the tie rod slot.

6.21 While forcing the upper end of the tie rod radially inward, taking up clearance (0.25 inch diametrical) in the support block's clearance hole, measure the radial gap from the tie rod to the vessel wall at the mid support elevation. The tie rod itself should not be bowed while taking this measurement. The vessel contact locations of the mid support shall be simulated in taking these measurements. Based on this in-reactor measurement, machine the contact surfaces of the mid support, drawing ll2D6356, in accordance with the requirements of the 105E1455 Modification and Installation drawing. Complete the mid support assembly as shown on the mid support assembly drawing,112D6359.

6.22 Remove the temporary thread protection from the tie rod. Install the tie rod nut and torque in accordance with the requirements of the 105E1455 Modification and Installation drawing; continue to force the upper end of the tie rod radially imvard during tensioning. Verify that the tie rod nut is properly locked by its retainers.

Lubricant (D50YP5B) shall be applied to the nut threaded surfaces.

6.23 Install the mid support in accordance with the requirements of the 105E1455 Modification and Installation drawing. Verify that the mid support latch is engaged in the tie rod slot.

6.24 Install upper stabilizer (spring) assembly, ll2D6347, in accordance with the requirements of the reactor modification drawing. Lubricant (D50YP5B) shall be applied to the 0.50 inch slot areas and thejacking bolt (threaded and moving surfaces). .

Engage with stabilizer support assembly and adjust the jacking bolt as specified on the l 105E1455 Modification and Installation drawing to preload the upper spring. Check

. that the spring retainers are properly engaged to lock thejacking bolt.

6.25 Reinove the protective shielding for the feedwater sparger and core spray line.

6.26 Repeat steps 6.3 through 6.25 for the installation of stabilizer hardware at the remaining azimuth locations. Step 6.11 is envisioned as an independent parallel activity.

O l

1 gg s @hbLy)/ 2

,. i

7. 'IXAMINATION AND TESTING 7.1 Visual Examination. Visually examine the stabilizer installation preparations to verify that all of the required holes have been machined in the proper locations and that all debris has been removed from the area. Visually examine the installed stabilizers to.

verify compliance with the 105E1455 Modification and Installation drawing. To minimize inspection time, personnel exposure, and tooling requirements, installation '

requirements, as indicated on the installation drawing, may be verified by tool design, process control and mockup qualification testing.

8. RECORDS AND SUBMITTAIS -

8.1 Prior to implementation of this stabilizer installation program, the following procedures shall be submitted by the Installer and approved by the Owner,

a. Installation and inspection procedures including sequence data sheets, '

measurement data sheets, quality control check sheets, drawings, sketches, instructions, etc. ,

b. Cleaning and cleanliness control procedures.

t

c. Machining procedures as applicable.

d. As-built drawing (data required by 105E1455).

8.2 After implementation of this stabilizer installation program, all recorded data records, photographs, video tapes, logs, etc., shall be submitted by the Installer to the Owner for file and information within 30 days. The 105E1455 modification and ,

installation drawing shall be updated to incorporate the in-reactor as-built measurements,

- and the as-built measurements with corresponding serial numbers of the parts machined  ;

as part of the installation process. One copy shall be submitted to GENE within 30 days. >

9. DEVIATIONS AND SUBSTITUTIONS i

9.1 All deviations, as a result of damaged equipment, nonconforming conditions, or any , :

proposal by the Installer for substitutions, modifications, or relaxation of the specified i materials, procedures or design shall be submitted to'the Owner for consideration and ,

approval.

6

GENE-771-59-0994 Rev.4 DRF B13-01732 JUNE,1995 f

DESIGN REPORT FOR THE INSTALLATION OF STABILIZERS ON THE PEACH BOTTOM 2,3 CORESHROUD GE NUCLEAR ENERGY Prepared by: ,

M.O. Lenz /

Project Engineer Verified by: M " -

N.A. McClean Principal Engineer Approved by:

L.f Tschantre Project Manager ft7v -

k l'

l GENE-771-59-0994 Rev.4 DRF B13-01732 I JUNE,1995 IMPORTA.NT NOTICE REGARDING CONTENTS OF THIS REPORT l Please Read Carefully i

The only undertakings of GE Nuclear Energy (GENE) respecting information in this document are contained Contract No. PB263848 between PECO Energy and GENE, as identified in the purchase order for this report (PECO Purchase Order No. PB263848) and nothing contained in this document shall be construed as changing the contract. The use of this information other than as provided therein of for any purpose other than for which'it is intended, is not authorized; and i with respect to any unauthorized use, GENE makes no representation or warranty, and assumes no liability as to the completeness, accuracy or usefulness of the information contained in this document, or that its use may not infringe upon privately owned rights.

Changes made to the previous revision of this report are indicated by " bars" drawn in the left margin.

6 I

E i

GENE-771-59-0994 Rev.4 -

DRF B13-01732 JUNE,1995 i

Table of Contents Page

1.0 DESCRIPTION

OF CHANGE 1

'1.1 General 1 1.2 Design - 2 1.3 ~ Materials . 4 1.4 System Evaluation 5 1.4.1 Steam Separation System 6

~

1.4.2 Jet Pumps 6 1.4.3 Anticipated Operational Transients 6 1.4.4 Emergency Core Cooling System 7 1.4.5 Fuel Cycle Length 7 1.4.6 Leakage through Shroud Cracks and Support Plate 7 1.4.7 Conclusions 8 1.5 Seismic Analysis 9 i

1.6 Design Evaluation 11 1.6.1 Load Combinations 12 l.6.2 Results 13 i

2.0 - REASON FOR CHANGE 16 3.0 DESIGN AND LICENSING DOCUMENTATION REVIEW 16 4.0 REFEREt,CES 17 l

I d

GENE-771-59-0994 Rev.4 DRF B12-01732 JUNE,1995 1.0 DESCRIPY10N OF CHANGE An in-vessel visual inspection (IVVI) of the horizontal welds in the shroud will be performed at Peach Bottom Unit 2. Based on the results of the weld inspection, shroud stabilizer assemblies, which function to replace cracked horizontal welds in the shroud, may be installed in Peach Bottom Unit 2 during the next refueling outage. In the event that the installation of the stabilizer assemblies is not required for Unit 2, the assemblies will be retained as a contingency for possible later installation, or for installation at Peach Bottom Unit 3. If the stabilizers are installed at Unit 3, replacement of the core spray line repair brackets should be considered, as discussed in Section 1.6.2.

1.1 General Welds H1 through H7 of the core shroud will be structurally replaced by a set of 4 stabilizer assemblies. Figure I shows a stabilizer assembly. Each stabiliza assembly attaches to the top of the shroud and to the shroud support plate.

Radially acting stabilizer springs are used to maintain the alignment of the core shroud to the reactor pressure vessel (RPV) during seismic events. The set of stabilizers replace the structural functions of the shroud welds which are postulated to contain cracks. Each stabilizer assembly consists of a tie rod, an upper spring, a lower spring, an upper bracket, a midspan tie rod support, a lower anchor assembly, and other minor parts. The tie rod provides the vertical load carrying ability from the upper bracket to the RPV shroud support plate attachment, as well as support for the springs. The vertical locations of the radial springs were chosen to provide the maximum support for the shroud and fuel assemblies. The upper spring provides radial load carrying ability from the shroud, at the top guide elevation, to the RPV. The lower spring provides radial load carrying ability from the shroud, at the core support plate elevation, to the RPV. The upper bracket provides an attachment feature to the top of the shroud as well as restraint of the upper shroud welds. The midspan tie rod support functions to increase the natural frequencies of the stabilizer assembly.

There are seven horizontal welds (Eigure 2) in the Peach Bottom 2,3 shroud. These welds are identified as H1 through H7, with H1 being the uppermost weld and H7 being the attachment of the shroud to the shroud support cylinder. Each cylindrical section of the shroud is prevented from unacceptable motion by the stabilizers. The motion of the sections above H1, between H1 and H2, and between H2 and H3 are restrained by the upper bracket. The upper bra:ket contacts the shroud and is radially supported by the upper spring which contacts the RPV. There is a stop 1

GENE-771-59-0994 Rev.4 DRF B13-01732 JUNE,1995 on the upper bracket which prevents unacceptable motion of the shroud section between the H3 and H4 welds. The lower spring contacts the shroud such that it prevents unacceptable motion of the section between the H4 and H5 welds, as well as the section between the H5 and H6 welds.

There is also a feature on the lower spring which prevent unacceptable motion of the shroud section between the H6 and H7 welds.

The H8 weld (Figure 2) connects the shroud support shelf to the shroud support cylinder. The assumption of cracking in the H8 weld was also considered in the seismic analysis, the analysis of the stresses in the shroud support plate, and in the systems performance evaluation with the assumption ofleakage through the weld cracks.

The primary forces applied to the stabilizers are from seismic events, LOCA differential pressure loads, and differential thermal expansion. The installation of the stabilizer assemblies and the assumption of cracks in the shroud change the seismic response of the reactor internals. Thus, it was necessary to modify the seismic analysis of the reactor to include the assumed cracks and the stabilizers. This dynamic analysis was performed in a. iterative manner to determine the appropriate values of the spring constants for the upper and lower springs, as well as the number of stabilizer assemblies requ' - - was determined that four stabilizer assemblies would be acceptable. Each assembly s an upper s@ith a spring rate constant of 20,000 pounds per inch, and a lower spring with a spring rate constant of 150,000 pounds per inch. The midspan tie rod support increases the natural frequency of the rod to prevent unacceptable vibration.

1.2 Design Significant cracking adjacent to the horizontal shroud welds has been observed at several BWRs.

If significant cracking is discovered in the horizontal shroud welds at Peach Bottom Unit 2, or at Unit 3, a repair will be implemented. The repair will structurally replace all of the horizontal girth welds in the shroud with a set of shroud stabilizers.

The core shroud is a safety related component. It provides horizontal support for the fuel

~

assemblies, control rods, and incore instrumentation. It provides vertical support for the peripheral fuel assemblies, top guide, and core support plate. It also provides a floodable volume inside the reactor pressure vessel (RPV), which is necessary in the event of a Loss of Coolant Accident (LOCA). Note, however, that welds H.1 through H3 are above the floodable volume.

In addition, the shroud supports the core spray spargers and the core spray lines.

The stabilizers were designed to the structural criteria specified in the original Peach Bottom 2,3 UFSAR. The ASME Code, Section Ill, Subsection NG was used as an analysis guide. The 2

GENE-771-59-0994 Rev.4 DRF B13-01732  !

. TUNE,1995  !

i i

stabilizers were designed to the requirements of References 4.1 and 4.2. Reference 4.1 is the design specification for the stabilizers, exclusive of ASME Code aspects. Reference 4.2 contains the ASME Code requirements for the interfaces of the stabilizers with the RPV, since the installation of the stabilizers will result in new loads being applied to the RPV. All of the loads l and load combinations specified in the UFSAR, that are relevant to the core shroud, are included in the design specifications. The increased operating pressures due to Power Rerate and increased core flow are also included in the design specifications.

The stabilizers are installed with a small vertical mechanical preload, which assures that all components are tight and which provides approximately 3300 pounds of axial load on the 3.5 inch diameter tie rods. The upper support, upper spring, lower spring, and lower anchor assembly are fabricated from alloy X-750, which has a smaller coefficient of thermal expansion than does the 304 stainless steel shroud. Additionally, the tie rods are fabricated from XM-19 stainless steel, which also has a smaller coefficient of thermal expansion than does the shroud.

Thus the stabilizer assemblies are thermally preloaded when the reactor is at operating conditions. The spring constant of the stabilizers in the vertical direction was designed to yield a total vertical preload at operating conditions of greater than the net upward applied loads on the shroud. The vertical loads are largely a result of the differential pressures across the shroud head and across the core support plate. Thus, if any or all of the H1 through H7 welds were completely cracked, the stabilizers will vertically restrain the shroud such that no vertical displacement will occur during Normal operation, which minimizes potential leakage through the assumed cracks.

The upper and lower springs are installed with a small radial preload such that they provide radial support for the shroud. During Normal operation, the shroud and springs radially expand due to thermal growth slightly more than the RPV due to both thermal and pressure, which increases the radial preload and assures that the springs provide linear support for the shroud during normal operation.

l -

The vertical location of the upper and lower springs were chosen to provide the maximum horizontal support for the fuel assemblies. The upper springs are at the top guide elesation and i the lower springs are at the core support plate elevation. All of the horizontal support for the fuel assemblies is provided by the top guide and the core support plate.

The stabilizer assemblies are designed and fabricated as safety related components, as is the requirement for the shroud. The installation of the stabilizer assemblies will replace the functions of welds H1 through H7.

3

GENE-771-59-0994 Rev.4 DRF B13-01732 JUNE,1995 At the top, each stabilizer assembly fits through two slots, which are machined through the non-safety related shroud head and steam separator assembly. The stabilizer upper bracket contacts the top surface and the inside surface of the shroud flange. It then extends downward to below weld H3. It supports the upper spring, and has a hole through which the tie rod passes. The tie rod is held against the upper bracket with a nut. The tie rod extends downward approximately 165 inches to the lower spring. At the middle of the tie rod there is a support between the tie rod and the RPV. The support is installed such that there is a horizontal preload between the tie rod and the RPV. The midspan support functions to minimize the potential for vibration of the stabilizer assembly. At the bottom, the tie rod threads into the lower spring. The lower spring has a clevis at the bottom, which is attached to the pin in the lower support. The lower support is bolted to the shroud support plate with two toggle bolts.

All pieces of the stabilizer assemblies are lacked in place with mechanical devices. Loose pieces can not occur without the failure of a locking device. The stresses in the stabilizer components during Normal plant operation are less than the Normal event allowable stresses. The stabilizers are fabricated from stress corrosion resistant material. Therefore, it is very unlikely that a stabilizer component will fail.

The fast flux levels 'at the stabilizers are low compared with the values which could result in the degradation of material properties. After 20 years of operation, the maximum fast fluence at the 2

stabilizers will be approximately 3.0E19 n/cm , which is well below the value to cause damage to stainless steel.

1.3 Materials The stabilizers are fabricated entirely from the type 316 stainless steel, XM-19 stainless steel, and alloy X-750. There is no welding required.

The upper and lower springs, upper nut, upper bracket, and lower support are fabricated from Alloy X-750 (Ni-Cr-Fe) material that has been heat treated at 1975 25'F followed by an air ~

cool and then age hardened to increase its strength. The annealing and age hardening processes used are essentially the sr.me as those used on the improved jet pump beams for maximum resistance to IGSCC initiation. As a control for intergranular attack (IGA), a minimum of 0.030 inches of material is removed after the last high temperature annealing operation. This material  !

is certified to ASTM B637, Grade UNS N07750. Alloy X-750 was chosen because high strength was required, and it has a coefficient of thermal expansion that is less than that of the shroud.

4

GENE-771-59-0994 Rev.4 DRF B13-01732 JUNE,1995 L ,

Alloy X-750 is resistant to IGSCC at the stress levels the components will experience during operation, which is less sustained tensile stress than the jet pump beams experience. Some smaller components are also fabricated from alloy X-750. The cobalt content of all X-750 material was limited to a maximum of 0.09 %.

The tie rods are fabricated from XM-19 stainless steel with a maximum carbon content of 0.04%.

The XM-19 rods were solution annealed at 2000 50 F, and tested for evidence of sensitization and IGA. XM-19 stainless steel was chosen for the fabrication of the tie rods because of its higher strength over that of 300 series stainless steel.

Other components are fabricated from type 316 stainless steel material with a maximum carbon content of 0.02%. The material was annealed at 1900 to 2100 degrees F followed by quenching in circulating water to a temperature below 400 degrees F. All material was tested for evidence of sensitization and IGA.

1.4 System Performance Evaluation The installation of the four shroud stabilizer assemblies at Peach Bottom Unit 2 or 3 requires the machining of four pairs of holes through the shroud head flange. Each of these holes will have some clearance after the installation of the upper support, which will allow a small amount of leakage flow to bypass the steam separation system. In addition, four pairs of holes will be machined in the shroud support plate, which will also allow a small amount of core flow leakage through the clearance between the holes and the mating bolts. There are seven horizontal welds in the shroud, with two welds below the core plate, and five welds above the core plate. These welds may develop cracks, which will present additional small leakage flow paths for the core flow. In addition, potentialleakage through cracks in the H8 weld has been considered. Finally, leakage past the replacement access hole covers in the shroud support shelf has been predicted to be 0.07% of core flow. This section summarizes the performance impact of the total leakage flow for up to 110% of rated power, and 87% to 110% of rated core flow (i.e., potential future power rerate conditions).

Leakage flow past the replacement access hole covers is 0.07% of core flow. Leakage from the machined holes in the shroud head flange is assumed to be two-phase fluid at the core exit quality. The maximum leakage through these eight holes is predicted to be equal to 0.10% of core flow. The steam portion of this leakage will contribute to increasing the total carryunder from the steam separators. The cumulative leakage flow from the shroud support plate holes and 5

GENE-771-59-0994 Rev.4 DRF B13-01732 JUNE,1995 from the assumed weld cracks is conservatively estimated to be 0.21% of core flow. The impact

- of the total leakage on the performance of the steam separation system, the performance of the jet pumps, the fuel thermal margin, emergency core cooling system (ECCS) performance, and the length of the fuel cycle have been evaluated as summarized in the following subsections.

i 1.4.1 Steam Separation System

' The shroud leakage flow includes steam flow, which effectively increases the total carryunder in the downcomer by a maximum of about 0.02% at 110% rated power, and 87% to 110% rated core flow. The total leakage flow also has the effect of slightly decreasing the flow per separator, and slightly increasing the separator inlet quality. The carryunder from the separators is based on the applicable separator test data at the lower limit of the operating water level range. The combined effective carryunder from the separators and from the shroud leakage at 87% and 110% rated core flow is about 0.28% and 0.25% respectively. The specified design value corresponding to 0.25% is slightly exceeded, but is acceptable. The impact of the increased carryunder is evaluated in the following subsections. The carryover from the separators remains within the design limits, so that moisture from the steam dryer meets the plant performance requirement ofless than 0.1%.

1.4.2 Jet Pumps The increased total carryunder will decrease the subcooling of the flow in the downcomer. This in tum reduces the operating margin to jet pump. cavitation. Any increase in flow resistance in the annulus due to the installation of the shroud stabilizer assemblies will be insignificant. The results show that, although that the jet pump cavitation operating margin decreases slightly, it will remain adequate. The change in cavitation operating margin is not significant enough to require a change to a cavitation set point, or to the power flow map.

1.4.3 Anticipated Operational Transients The computer code used to evaluate performance under plant transients and to calculate fuel thermal margin includes carryunder as one of the inputs. There will be a slight increase in carryunder due to shroud head installation hole leakage. The increase in carryunder results in an insignificant decrease in subcooling in the downcomer region, with a corresponding decrease in core inlet subcooling. The leakage flow rate via the shroud support plate holes and the shroud weld cracks tends to (insignificantly) reduce the coolant flow rate into the core region. This core flow reduction will slightly decrease the core inlet subcooling. However, the total core flow reduction, and the core inlet subcooling changes are both within the sensitivity range of the ,

i 6

5

l GENE-771-59-0994 Rev.4 DRF B13-01732 JUNE,1995 critical power ratio (CPR) calculations. Also, the increased canyunder fraction will tend to i reduce (make milder) the response from pressurization transients. Therefore, the fuel thermal limits will not be affected.

1.4.4 Emergency Core Cooling System The effect of the shroud and agcess hole cover leakage flows is to decrease the time to core uncovery slightly, and also to increase the time that the core is uncovered. An evaluation of the ECCS performance, conservatively based on the cumulative effect ofleakage at the shroud head, at the shroud support plate holes, for five full (360') shroud weld cracks, and at the access hole covers, was performed. The combined effects have been determined to increase the peak cladding temperature (PCT) for the limiting LOCA event by less than 20'F. This increase is sufficiently small to be judged insignificant, and thus, the licensing basis PCT for the Normal operation condition with no shroud leakage is applicable. In addition, the licensing basis PCT was calculated assuming a 20% reduction in the ECCS flow rates. The effect of the ECCS flow rate reduction bounds the effects of the assumed shroud leakage. The sequence of events remains essentially unchanged for the LOCA events with the shroud leakage.

. The conservative ECCS flow rates in the current LOCA analysis significantly bound all potential flow losses resulting from the installation of the shroud stabilizers, and thus the current ECCS-LOCA analysis remains valid. However, even if the current ECCS-LOCA analysis did not use reduced ECCS flow rates, the calculated PCT (1690 F + 20 F) would remain well below the regulatory PCT limit (2200 F), the change in PCT would be less than the 50 F requirement for NRC notification, and a floodable volume to 2/3 core height will be maintained.

1.4.5 Fuel Cycle Length The cumulative leakage flow results in an increase in the core inlet enthalpy by about 0.1 Btu /lb, compared with the no leakage condition. The combined impact of the reduced core inlet subcooling and the reduced core flow due to the leakage results in a minor effect (approximately 0.8 days) on fuel cycle length, and is considered negligible.

l.4.6 Leakage through Shroud Cracks and Shroud Support Plate Holes The amount ofleakage that would pass through an assumed 360' through wall crack at operating conditions is limited by the tie rods, which hold the crack surfaces in contact. If it is conservatively estimated that the crack can be modeled as a 0.001 inch wide smooth slit through the shroud, then the leakage is approximately 145 GPM for all seven welds combined. If the H8 7

GENE-771-59-0994 R::v.4 '

DRF B13-01732  :

JUNE,1995 l I

weld is assumed to be cracked, the combined leakage for all eight welds is approximately 170 GPM. All of the leakage below the core plate would be water of the same temperature as the water in the downcomer. All of the leakage between the top guide and the core plate would be i water. Only the leakage from H1 and H2 would include any steam. The leakage through the eight machined mounting holes in the shroud support plate will not be more than 525 GPM, calculated at rated power and Increased Core Flow.

Summary of Enhoe Flows at Ucrated Power and ICF '

Pressure difference (psi) f Shroud Head 9.41 Upper Shroud 9.35  :

Lower Shroud 33.0 i Leakage Flow (gpm) 1 Repair Holes in Shroud Head 1105 Weld Cracks 170  !

Repair Holes in Support Plate 525 Access Hole Covers 215  :

Leakage to Core Mass Flow (%) l Repair Holes in Shroud Head 0.10  !

Weld Cracks 0.04  :

Repair Holes in Support Plate 0.17 Access Hole Covers 0.07 l I

1.4.7 Conclusions The stabilizer assemblies will not significantly affect the flow within the downcomer and will not adversely affect the performance of any reactor internal. The water inventory in the downcomer 1 with the stabilizers installed exceeds the volume used in the existing analyses.

The cumulative effects ofleakage at the shroud head flange, shroud support plate holes, assumed sitroud weld cracks, and access hole covers have been evaluated for degradation in steam separation system performance, jet pump performance, fuel thermal margin, ECCS system perfomiance, and for changes in fuel cycle length. The leakage flows are sufficiently small so that steam separation system performance, jet pump performance, fuel thermal margin, and fuel 8

I GENE-771-59 0994 Rev.4 DRF B13-01732 JUNE,1995 cycle length are not significantly affected. Also, the effect on ECCS performance is sufficiently small to be judged insignificant, and thus, the licensing basis PCT for Normal operation with no shroud leakage is applicable.

1.5 Seismic Analysis A seismic analysis of Peach Bottom Unit 2, and 3 has been performed (Reference 4.8). The new seismic model incorporated the installation of the shroud stabilizers. Included in the analysis were the determination of the natural frequencies and mode shapes of the RPV building system for the E-W and N-S direction, computation of the dynamic response to the DBE and MCE time histories, and the calculation of the shears, moments, and displacements. The building model was the 1970 model~specified by the architect engineers (Bechtel). The Design Earthquake (DBE) and Maximum Credible Earthquake (MCE) time histories used are based on response spectra in Figures C.3.1 and C.3.2 of the Peach Bottom 2,3 Updated Final Safety Analysis Report (Reference 4.3). Peak ground accelerations for the DBE and MCE are 0.05g and 0.12g respectively (Reference 4.3)

The model was qualified by achieving excellent agreement with frequencies and mode shapes predicted by the previous GE model of Reference 4.5. The synthetic time history was generated from the design response spectra.

The computer program SAP 4G07 (Reference 4.4) was used in this study. Transient response i analysis with modal superposition method (NDYN=12) was adapted. A solution time step of l 0.002 second was chosen, according to the user's manual for this linear analysis.

l The vertical seismic effect was combined directly with the horizontal seismic load (either North-South or East-West). Forces and moments due to vertical loading were calculated as a ,

multiplier of the dead weight. This is based on the approach discussed in the UFSAR (Reference 4.3).

l The springs for the stabilizer supports and tie-rods are included in the model. Each stabilizer assembly has a radial spring at both the top guide and at the core support plate elevation. Each radial spring can take only compressive loads. The axial spring constant for the tie-rods was calculated based on the four tie-rods rotating about the shroud neutral axis 9

GENE-771-59-0994 Rev.4 DRF B13-01732 JUNE,1995 In order to determine that the installation of shroud stabilizers does not adversely affect the existing structural integrity with the assumption that no defective welds are present in the shroud, analyses for the uncracked case were performed with the shroud stabilizers in place. The results in this study were compared with those given in Reference 4.5 to insure that all the RPV and internals loads and displacements are within the allowable limits.

An enveloping combination of cracked /uncracked welds has been determined to define the worst situation for the core plate and top guide displacements, in order to insure control rod insertion and plant shut-down. Each cracked weld was postulated to have a 360 degree through wall crack. Seven individual cracked welds as well as several combinations of two cracked welds have been evaluated. Also, an all cracked combination has been evaluated. Each 360 degree through wall crack has been modeled as a hinge or a roller to determine the limiting boundary conditions. The specific load combination thrt is being analyzed also influences the choice of the boundary condition of hinge or roller. For example, during Normal operation differential thermal tightening of the tie rods prevents upward motion of the shroud. Thus, the tie rods hold the crack surfaces in contact, which assures that a hinge boundary condition is conservative.

During a main steam line LOCA, the tie rods will momentarily stretch such that the crack surfaces in the shroud may part. This condition can only exist for a few seconds while the shroud head pressure drop is sufficient to stretch the tie rods. During this short period of time, the assumption of a roller is appropriate. The seismic analysis conservatively assumes that the roller assumption applies during the entire seismic time history.

For the LOCA case with all welds cracked, the uppermost weld (Hl) is postulated to be a roller and all the other cracks are postulated as hinges. In this case, to prevent modeling instability, additional torsional and translational springs with small spring rates were inserted at the cracks to stabilize the mathematical model for the all cracked case. In reality, these small torsional springs represent certain degree of resistance from the structure. It was found, through a sensitivity study, that the analysis results were not affected by the small spring rates that were chosen.

The seismic time history analysis considered a total of 21 load cases. They represent ,

combinations of two earthquake events (DBE and MCE), two horizontal directions (North-South and East-West), several cracked weld assumptions, and the two Peach Bottom units.

The maximum deflection of any part of the shroud that is not directly supported by either the upper or lower radial springs is limited to approximately 1.0 inch by mechanical limit stops.

These stops do not perform any function unless a section of the shroud, for example the section between the H4 and H5 welds, becomes loose and a combined LOCA plus seismic event occurs.

10

GENE-771-59-0994 Rev.4 DRF B13-01732 JUNE,1995 If this unlikely scenario were to occur, the stops will limit the horizontal displacement of the '

shroud section to approximately 1.0 inch, which is equal to one half of the shroud wall thickness.

These stops do not invalidate the linear seismic analysis. A displacement equal to one half of the shroud wall thickness will not result in post event leakages that prevent core cooling, because the shroud sections still overlap each other by 'one half (1.0 inch) of the shroud wa'll thickness.

The original seismic model of the NSSS was revised to incorporate the current core reload and the installation of the shroud stabilizers. This case models the current configuration with an uncracked shroud. The seismic response of this model was then determined using a ground motion input time history equivalent to the Design Response Spectra. A comparison of the results with those from the original seismic analysis showed a close comparison with all resulting loads in NSSS components, which were well below the allowable values. The updated model was then modified to reflect postulated 360 degree through wall cracking at the H1 through H7 shroud welds. Various combinations of postulated cracking locations were investigated in order to identify the bounding combinations and locations. The shroud stabilizers are intended to relieve the shroud seismic overturning moment, and transfer it to the stabilizers and to the RPV.

The modeling and the resulting seismic loads in the shroud reflect this design concept, the bending moments in the cracked and stabilized shroud being only 2% of those from the original seismic analysis reported in the UFSAR.

Potential cracking in the H8 weld between the shroud support plate and the shroud support cylinder was also evaluated. Assumed cracking at the H8 weld was determined to have no significant effect on the properties of the seismic model, and thus had no significant effect on the results of the seismic analysis.

1.6 Design Evaluation i

The results of the structural evaluations per References 4.1 and 4.2 are documented in References 4.6 and 4.7. Reference 4.6 addresses the ASME Code,Section III, RPV and Reference 4.7 i addresses the non-code shroud stabilizers, shroud, and shroud support shelf. The stabilizers and i affected shroud and RPV components are shown to satisfy the UFSAR structural requirements l using the UFSAR load combinations. I The displacements of the core support plate and the top guide are limited to the allowable displacements given in P derence,4.1 for all load combinations.

1I i

e I GENE-771-59-0994 Rev.4 p

DRF B13-01732 JUNE,1995 1.6.1 Load Combinations The UFSAR requires evaluation of Normal operating loads as well as DBE, MCE, main steam line LOCA, and recirculation line LOCA. The following load combination and their classification were considered:

Normal: Normal operating pressure differences, temperatures, and weight Upset 1: Normal operating weight, temperature and pressures, plus DBE Upset 2: Limiting thermal condition, plus Normal pressure differences and weight Emergency 1: Normal operating weight, temperature and pressures plus MCE Emergency 2: Weight plus main steam line LOCA pressure differences Faulted 1: Weight, plus MCE, plus main steam line LOCA pressure differences Faulted 2: Weight, plus MCE, plus recirculation line LOCA pressure differences The values of the individual loads on the shroud stabilizer springs and tie-rod were obtained from the design specifications (References 4.1 and 4.2), which included the seismic analysis of the degraded shroud with stabilizers. The seismic loads in the design specifications are the limiting values of the various assumed cracking conditions.

In general, the limiting loads in the tie rods occur with different assumed shroud cracks than for the limiting radial loads in the upper and lower springs. It was conservatively assumed that the worst loads on each component in each direction would be combined. The limiting loads in the tie rods occur when it is assumed that there is a 360 degree through wall crack in weld H7 and I that crack behaves as a hinge (shear transfer but no moment transfer). If the H7 is assumed to behave as a roller (no shear and no moment transfer), then the loads are reduced in the tie rods.

The limiting loads in the radial direction on the upper springs occur when it is assumed that all of the horizontal girth welds in the shroud have 360. degree through wall cracks, with each of these j joints rnodeled as hinges. The limiting radial loads on the lower spring occur with only the H7 cracked, and the joint modeled as a roller. Since the crack surfaces are held in intimate contact, ,

thejagged IGSCC cracks can transmit shear, which is the hinge condition. During a main steam 12 i- i

i 1

l GENE-771-59-0994 Rev.4 l DRF B13-01732 JUNE,1995 line break the upward pressure forces are sufficient, for a short period of time, to stretch the tie rods and allow the crack surfaces to separate. For this situation, the roller case was assumed at the joints.

1.6.2 Results The potential for flow induced vibration has been evaluated by calculating the lowest natural frequency of the shroud stabilizer assemblies, and conservatively calculating the highest vortex shedding frequency due to the water flow across the stabilizer tie rods, in the downcomer (RPV-shroud annulus). The lowest natural frequency of the assembly is 45.5 Hertz and the maximum vortex shedding frequency is 5.2 Hertz.

There will be essentially no fatigue usage for any of the stabilizer components.

References 4.6 and 4.7 document that all structural limits have been satisfied for the shroud, shroud support plate, shroud stabilizers, and the reactor pressure vessel. The analysis of the shroud support plate considered cracking at the H8 weld. The predicted worst case transient deflections of the core plate is 0.87 inch for a load combination of a MCE with the H7 weld cracked plus a main steam pipe LOCA. The allowable transient displacement for this Faulted event is 1,49 inch. All calculated stress intensities in the lower radial spring meet the UFSAR allowables. The predicted worst case transient displacement of the top guide is 2.20 inches for a MCE with an all cracked shroud. The allowable permanent deflection of the top guide for this Emergency event is 1.40 inch. The 2.20 inch displacement is a momentary value which occurs during the MCE, and is not permanent. The stresse i H % + e er radial spring meet the UFSAR allowables for an Emergency event with sufficiera margm such that the spring remains elastic, and the permanent displacement is zero. The predicted deflections of both the top guide and the core plate for all load combinations are within the allowables defined in the design specification, which are based on test results and the criteria of th: UFSAR.

The Peach Bottom 2,3 core spray lines were analyzed for the increase in seismic loads and anchor movements which may result from the assumption of a worst case cracked shroud, with the installation of the shroud stabilizers. The non-ASME Code, Safety Related, core spray lines were conservatively analyzed using the rules for Class I piping in the ASME Code Section III, NB-3600, as an analysis guide. It was determined that the Peach Bottom core spray lines as originally designed meet the ASME Code allowables for stress and fatigue usage for Normal /

Upset, Emergency, and Faulted conditions, and thus remain functional with the assumption of a 13

lf GENE-771-59-0994 Rev.4 DRF B13-01732 JUNE,1995 l

cracked shroud, and the installation of the shroud stabilizers. The allowable stress limits are in accordance with Tables C.S.2 and C.5.5 of the Peach Bottom UFSAR.

0 Each core spray line attaches to the reactor pressure vessel at the 120 or the 240 core spray nozzle. Each line branches into a 6.0" schd. 40 horizontal header, with two vertical downcomers that are welded to elbows in the upper shroud. The increase in loads on the core spray lines are largely a result of the increase in movement of the assumed worst case cracked shroud dunng a seismic event, or loss of cooling accident (LOCA). The shroud displacements are documented in

- Reference 4.8. The core spray lines with the mounting brackets, and appropriate boundary conditions were modeled with the ANSYS Revision 4 finite element computer program.

The load combinations addressed were as follows:

Service Levd Load Combination Stress Tyne N /U Weight + DEin ni, Primary '

N/U Thermal + DE ine ni,+ DExncno,uo,,,,ni Primary + Secondary 3

E Weight + MCEineni, Primary l

F Weight + MCE ineni, Primary l

l 1

All calculated stresses in the core spray piping met the ASME Code stress allowables. The maximum 40 year fatigue usage was calculated to be less than the Code allowable of 1.0. The i Normal / Upset condition for primary plus secondary stress included the existing thermal stresses due to thermal expansion and thermal anchor movements, as well as seismic inertia and seismic

. anchor movements.

It should be noted that , per NB-3217-2, the stresses resulting from piping anchor point motions are considered to be secondary, and are not required to be evaluated for Emergency and Faulted conditions (one time load application), in accordance with the requirements of NB-3600.

However, as a functional check on the core spray lines during the anchor movement resulting from a Faulted Condition main steam line LOCA combined with a maximum credible earthquake (MCE) event, the maximum strain in the piping was calculated. This was found to be less than 1.0%, which is well below the minimum allowable strain at failure of 25% for ASTM A-312 14 l

c GENE-771-59-0994 Rev.4 DRF B13-01732 JUNE,1995 conditions (one time load application), in accordance with the requirements of NB-3600.

However, as a functional check on the core spray lines during the anchor movement resulting from a Faulted Condition main steam line LOCA combined with a maximum credible earthquake (MCE) event, the maximum strain in the piping was calculated. This was found to be less than 1.0%, which is well below the rninimum allowable strain at failure of 25% for ASTM A-312 TP304 stainless steel piping, thus demonstrating the functionality of the core spray line after a Faulted event. The displacement of the shroud during a recirculation line LOCA was not found to be significant.

Repair brackets were installed on the core spray lines at Peach Bottom Unit 3 in 1985 (Reference 4.12) after the discovery of cracking in the 240' core spray line header near the connection of the pipe with the tee-box. The brackets are welded to the tee-box and to each header pipe, with one bracket mounted above the tee-box, and one mounted below. The core spray lines and repair brackets were also modeled and analyzed with the assumption of a 360 degree crack in each header pipe. The core spray piping was determined to meet the allowables for stress and fatigue usage in accordance with the requirements of NB-3600. The repair brackets and attachment welds to the piping were analyzed in accordance with the requirements of NB-3200. The stresses in the repair brackets and attachment welds were determined to meet ASME Code primary stress allowables for all load cases l Although the ASME Code Primary and Primary plus Secondary stress criteria is satisfied, the fatigue usage in a local area of the bracket to pipe weld for one of the two repair brackets was calculated to be 2.0, which is in excess of the Code allowable of 1.0. The fatigue usage for the second repair bracket was found to be acceptable. The main contribution to the calculated fatigue usage was due to the relative displacement of the shroud during 10 full amplitude cycles of a design earthquake (DE) event. A stress concentration factor of 4.0 was assumed for the fillet welds between the repair brackets and the core spray piping.

The ant. lysis 'of the core spray line header tee-box repair brackets described above does not support the plant design basis of 50 DE earthquake cycles, as indicated in Section C.5.3.6 of the ,

Peach Bottom UFSAR, with the core shroud horizontal welds postulated to be fully cracked and the core shroud stabilizers installed (Modification P00435). Therefore, PECO Modification P00541 has been initiated to provide a revised analysis which demonstrates that the repair brackets are acceptable with Modification P00435 installed, or to provide replacement brackets that meet the design requirements. Modification P00541 will be installed prior to, or during the implementation of the core shroud stabilizer modification.

15

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GENE-771-59-0994 Rev.4 DRF B13-01732  !

JUNE,1995 l i

2.0 REASON FOR CHANGE -

t Cracks have been observed in the core shrouds of several BWRs. The NRC has issued a generic l letter, Reference 4.9, which required inspection or repair. This report discusses the i,nstallation of  ;

shroud stabilizers that will functionally replace horizontal welds H1 through H7 in the Peach -  !

Bottom 2 and 3 shroud. At this time it is not known if there are any actual cracks in the shroud. }

This report is based on the assumption that all of the horizontal girth welds (H1-H7) have significant cracking. However, there is no degradation of function if the stabilizers are installed in the absence of crac! s in We shroud welds. l l

3.0 DESIGN AND LICENSING DOCUMENTATION REVIEW -

The Peach Bottom 2 and 3 UFSAR (Reference 4.3) was reviewed. The results of that review are as follows. The numbers in ( ) are the paragraph numbers from which the information was extracted, f (3.3.4.1.1) Gives a brief description of the shroud .

(3.3.5.1.2) Defines the pressure differences for Normal operation (3.3.5.2) Defines recirculation line LOCA pressure differences l.

(3.3.5.3) Defines main steam line LOCA pressure differences i (Table 4.2.1) Defines RPV material (Table C.5.9) Defines the damping values to be used in seismic analysis (Figures.C.3.1, C.3.2) Seismic response spectra (C.5.3.2.3) Internals seismic analysis methodology (Table C.5.5) Defines the required load combinations and required safety factors (Table C.5.1) Defines the allowable deflections )

(Table C.5.2) Defines stress limits )

I 1

16 1

. - _ _ _ . _ _ _ _ _ __ _ \

iT

GENE-771-59-0994 Rev.4 4

. DRF B13-01732

JUNE,1995 E ,

(Table C.S.6) Defines load combinations and stress limits i-

~ (C.5.3.2) States that ASME Section III was used as a guide in the design of the reactor intemals.

(Appendix K) RPV summary stress report

4.0 REFERENCES

4.1 25A5579 Rev. 3 " Shroud Stabilizer Hardware; Design Specification".

4.2 25A5580 Rev. 4" Shroud Stabilizer; Code Design Specification".

4.3 Peach Bottom Units 2 and 3 UFSAR, Rev.12.

l l

4.4 SAP 4G07 Computer Program.

4.5 383HA691 Rev. 0;" Seismic Analysis of Peach Bottom 2 Reactor Vessel and Intemals".

4.6 25A5607 Rev. 4; " Reactor Pressure Vessel; Code Stress Report".

4.7 GENE-771-58-0994 Rev. 4;" Shroud and Shroud Stabilizer Hardware Stress Analysis".

4.8 GENE-771-60-0994 Rev 2;" Peach Bottom Seismic Analysis, Shroud Mechanical Repair Program" 4.9 NRC Generic Letter 94-03, July 25,1994, "Intergranular Stress Corrosion Cracking of Core Shrouds in Boiling Water Reactors".

4.10 10CFR50.59 Review, Rev.0, for Mod P00435; " Installation of Stabilizers on Core .-

Shroud In Peach Bottom Atomic Power Station Units 2 & 3".

4.11 105E1455 G001; Reactor Modification Drawing l 4.12 797E551 G001; Core Spray Line Repair (Unit 3) 17

GENE-771oS9-0994 Rev.4 DRF B13-01732  ;

JUNE,1995  ;

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DRF B13 01732 REV 0 GE RESPONSES-TO NRC QUESTIONS i

FOR PEACH BOTTOM SHROUD REPAIR Jtme 1995 I

s Peach Bottom Shroud Repair NRC Questions and Responses Page1 l

1  ;

DRF B13 01733 REV U i

CONTENTS  ;

Tide Page -  !

l

1. Introduction 3 l

2. Description of the Analytical Model .

3  !

2.1 Component Stiffness 4  ;

12 Pressure Loads 4  !

i 2.3 Component Weights . 4 2.4 Mechanical Preload 5

3. Load Calculations 5 {

3.1 Load Case 1 -Stabilizer Assembly 5 3.2 Load Case 2 - Shroud Head Installation 5-3.3 Load Case 3 - Normal Operation with Uncracked Shroud 5 3.4 Load Case 4 -Thermal Transient Event with Uncracked 6 Shroud 3.5 Load Case 5 - Normal Operation with Cracked Shroud 6 Seismic Analysis 11 GE Inspection Recommendations Following the Installation of the 25 [

Shroud Stabilizers at Peach Bottom 2,3. {

Materials 26 i Core Plate Wedge issue 40 f

i ATTACHMENT A GE Document 383HA691, Rev. O, Peach 23 Bottom Units 2 & 3 and Internals ATTACHMENT B GE Drawing 886D499, Sheer 7 of 8, Revision 8 24 >

I Peach Bottom Shroud Repair NRC Questions and Responses Page 2

DRF B13-0192 REV O SHROUD SEPARATION AND THERhfAL UPSET CONDITIONS

1. Introduction The purpose of this document is to describe the analyses performed which

• evaluate the axialloading between the Peach Bottom shroud and the shroud repair stabilizers. The specific concerns to be addressed here are that of the potential of shroud separation during normal operating conditions with a shroud having all of the horizontal welds completely cracked and that of the potential for loss of preload due to plastic deformation of the stabilizers during a thermal transient event.

The method used to evaluate the axialloads in the stabilizer system was to determine the axial stiffness of the various components of the stabilizers and the axial stiffness of the various shells and rings of the shroud and to combine these using a finite element model with the effects of the appropriate temperatures, pressure differences and component weights. The resulting load balance of stabilizer tension and shroud compression demonstrates that the stabilizer tie rods maintain compressive loads at all of the shroud horizontal welds for the cracked shroud condition thereby assuring that the shroud rings and shells are held together and no separation occurs at the cracked horizontal welds. Additionally, the stabilizer components are shown to remain elastic and thus no loss of preload will occur for the axial loads in the stabilizer components which result from the uncracked shroud during a thermal transient event.

2. Description of the Analytical biodel This section presents a descripdon of the analytical model and the inputs to the model which produce the axial load distributions in the stabilizer assemblies and shroud for the cases of normal operation with normal temperature and pressures and the case cf a thermal transient with the corresponding temperature and pressures. The model is schematically shown in Figure 1. The model consists of a system of collinear two dimensional beams, each beam representing the length and stiffness of a major component of the stabilizer assembly and shroud. Direct inputs to the beam elements of Modulus of Elasticity, coefficient of thermal expansion, temperature, distributed axial loading and point loads appropriately addressed the effects of component stiffness, temperature ditTerences, pressure differences, component weights and mechanical preload of the tie rods. The ASIST computer program was used to solve the system of beam elements and produce the resulting load distributions.

Peach Bottom Shroud Repair NRC Questions and Responses Page 3

L DRF B13-01732 l

REV 0 2.1 Component Stiffness The axial stiffness of each of the major components of the stabilizer and shroud system was determined by various methods depending upon the complexity of the -

individual structure. The stiffness was input to the beam elements of the ASIST

• model by means of an effective area that combined with the component's actual Modulus of Elasticity and length would produce the correct stifTness.

2.1.1 Stabilizer Axial Stiffness:

The axial stiffness of the lower stabilizer spring assembly and the upper support assembly were determined by finite element analysis. The tie rod axial stiffness was determined by a simple hand calculation of a beam in tension. The values used are shown in Figure 2.

2.1.2 Shroud Axial Stiffness:

The axial stiffness of the shroud was determined by a combination of finite element analysis and theoretical ring stiffness calculations. The axial stiffness of

.the shroud shell members, shroud support ring and struts were determined by ,

finite element analysis. The resuldng axial stiffnesses are shown in Figure 2.

The axial stiffness of the shroud shell to ring welded connection was determined using the method of analysis and equations from Section III, ASME B&PV Code, 1965 Edition, Appendix I, Article 1-2. This method was used for the ring axial stiffness between welds H2-H3 and H5-H6. The stiffness was determined for both the case ofintact welds and fully cracked welds. The fully cracked welds were assumed to be through wall at the top of the weld fillet as schematically shown in Figure 3. Values of the axial stiffnesses used for the cracked and uncracked condition are shown in Figure 2.

2.2 Pressure Loads The axial forces acting on the shroud due to reactor internal pressure differences .

were input to the ASIST model as point loads at the appropriate node location.

Figure 4 shows the various pressure induced forces that were calculated. The force values are tabulated for both the normal operation and upset event pressure differences in Table 1.

2.3 ComponentWeights The weights of the various components that provide axialloads on the shroud including the shroud weight itself were corrected for the efTects of buoyancy and input to the ASIST model as either point loads or distributed axial loads along the beam elements. The values of the weights of the various components is summarized in Table 2.

Peach Bottom Shroud Repair NRC Questions and Responses Page 4

DRF BI3-01732 REV 0 2.4 MechanicalPreload The mechanical preload that the tie rods exert on the shroud due to the initial installation torque of the tie rod nuts is directly accounted for in the ASIST model. -

The temperature value input to the tie rod beam member was adjusted to produce

• a length difference between the stabilizer assembly and the shroud at installation.

The resulting length difference stretched the tie rod member the appropriate amount to produce the initial tie rod tension of 3346 Lb per tie rod.

3. Load Calculations

. A sequence of five load cases was made with the ASIST model to produce the loads ofinterest. The sequence begins with the initialinstallation of the stabilizer assemblies onto a shroud that has all welds intact. Using the larger shroud stiffness for uncracked shroud welds will produce the maximum tie rod loads and stresses for the thermal transient condition. The shroud welds are then assumed to be fully cracked which reduces the shroud stiffness which produces a reduction in mechanical preload due to the stiffness change and produces the condition most likely to have shroud separation. The load values presented for each load case are the tie rod tension and the compressive load at shroud weld H6.

3.1 Load Case 1 - Stabilizer Assembly The first load case is for the stabilizer assembled onto an uncracked shroud in the cold condition. Uncracked shreud stiffnesses are used, pressure forces are zero, all weights except the shroud head are applied and all temperatures are set to 70 F. The tie rod temperature is then artificially reduced which creates tension in the tie rod system equivalent to the tension produced by the initial torque of the tie rod nut during assembly. The tie rod tension value is 3346 Lb and the compressive load at shroud weld H6 is 156,057 Lb.

3.2 Load Case 2 - Shroud Head Installation This load case is identical to Load Case 1 except the shroud head weight is included as downward force. The net effect is to increase the compressive load at

. H6 to 277,657 Lb and reduce the initial tension in the tie rod to 962 Lh.

3.3 Load Case 3 - Normal Operation with Uncracked Shroud This load case considers the uncracked shroud and stabilizer system in the hot operating condition with normal temperatures and pressures. Uncracked shroud stiffnesses are used, normal operation pressure forces from Table 1 and weights from Table 2 are input to the ASIST model. The temperature of the stabilizer components and shroud support members below H5 was set to 527 *F and the temperature of the shroud members above H5 was input as 539 F. The resulting l l

Peach Bottom Shroud Repair NRC Questions and Responses Page 5 q l

l

DRF Bl3-01732 REV 0 tie rod load was 94,708 Lb and the compression load at shroud weld H6 was -

65,564 Lb.

l 3.4 Load Case 4 -Thermal Transient Event with Uncracked Shroud - l l This load case considers the uncracked shroud and stabilizer system in the l j operating condition with upset pressures and temperatures from a thermal i transient event combined to produce large tie' rod tension in the stabilizer system.  !

Uncracked shroud stiffnesses, upset pressure forces from Table 1, and weights from Table 2 were input to the ASIST model. The temperature of the stabilizer l' components was set to 300 *F and all members of the shroud and shroud support were input as 433 *F. The resulting tie rod load was 249,985 lb and the compression load at shroud weld H6 was 500,112 Lb.

Each component of the stabilizer / shroud system was evaluated for stresses  !

resulting from the 249,985 Lb axialload. The stress in each of the components was shown to be less than the material yield strength at the appropriate temperature thus assuring an clastic system with no loss of preload resulting from the thermal transient event. l i

3.5 Load Case 5 -Normal Operation with Cracked Shroud i This case is identical to Load Case 3 except the cracked shroud stiffnesses are  ;

used. The significant reduction in the stiffness of the system reduces the tie rod ,

tension and compressive load at the shroud H6 weld, however, the shroud is maintained in compression by the tie rod stabilizers and no shroud separation occurs. The resulting tie rod tension was 79,343 Lb and the compressive load at '

shroud weld H6 was 4081 Lb.

i e

P l

. 3 I

t Peach Bottom Shroud Repair NRC Questions and Responses Page 6

, . . -.. - . ., . . ~ . . . .... . - . - . . ._ _. . . . . . ,~.

l 1

E DRF B13-01732 REV O I

Table 1.

Pressure Forces Inputs to ASIST Model l 1

Location - Normal AP, Psi Transient AP, Psi

~

Lower Shroud 33.03- 35.68 Upper Shroud 9.35 .14.03 l Core Plate 23.68 , 26.32 i i

Force Location Normal AP Force Transient AP (Figure 4) (Lb) Force i (Lb) l' F1 109,150 117,910 ,

F2 96,430 104,160  ;

F3 423,910 457,920 F4 12,180 12,190 i F5 130,140 130,280  !

F6 39,630 59,460  !

F7 12,570 18,870 ,

F8 6,260 9,390 i Shroud Head 336,310 504,630  ;

}

Table 2 ,

Weight Inputs to ASIST Model Weight in Cold Weight in Hot )

Component Water Water  ;

(Lb) (Lb) '

Shroud Head 131,136 136,042 Shroud 99,761 103,494 Core Spray 1,751 1,817 r

Top Guide 18,118 18,796 i

Core Plate 27,182 28,199 Peripheral Fuel Supports 231 240 Peripheral Fuel Bundles 13,272 13,728 l l

I e

Peach Bottom Shroud Repair NRC Questions and Responses Page 7 -

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. . , . . . .- - .-. , . . . . . . . . . ~ . . . . ~ . . ~ . . . .

i

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' DRF B1341732 REV 0 ,

ASIST Model ,

a Peach Botton Shroud / Stabilizer ..

i 6 7 .8 j

d .-

9"j s

e .,  !

5 10 w 11 m:r }

l, i

i l=

i 13  :

3 14  ;

j =g= 15 .

[' '

16- l g, f i

.l i 17 i

Figure 1. Peach Bottom Shroud / Stabilizer ASIST Model Definition l 1

i i

Peach Bottom Shroud Repair NRC Questions and Responses Page 8 i.

T y -..-an- m -

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!/ DRF B1341732 ,

REV O ASIST Model.

Peach Botton Shroud / Stabilizer .

6 7'8  :

Kg = 995,000 kip /in CJ Kg = 3,700 kip /in Kp, = 984,400 kip /m l (4 Places) ,

[

5 10

• 11 K., = t33,700 k'p/in (uncrackee K,, = 6,400 kip /in (Cracked Kg = L500 kip /m Km ., = 254.100 kip /in i (4 Places)

K,, = 619,800 kip /in (Uncrackes i 12 K,, = 355,900 kip /m (Crackes 3

Kg = L300 kip /in bu = 576,000 kip /m u

(4 Places) 3 , 14

- 45= 1892700 kip /in j#ay 15

/

{g K,., = 14L900 kip /in Kg = L835,200 kip /in K,, = 105,000 kip /in

- 1 a 17 1

Figure 2. Peach Bottom Shroud / Stabilizer ASIST Model Axial Stiffness Values l

Peach Bottom Shroud Repair NRC Questions and Responses Page 9

]

DRF B13-01732 REV 0 a

CRACKED F8 -

-l' WELD H2 _f 1 POINT OF RING ROTATION F7 F6 _

p  !

a fhtg . 3?k l v=5:p mm e

• O;:.' . ; b .h'*{ l,

t.es.g;yl' 8  ?

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/ Y I TOP GUIDE RING FP 11 F3 l

l s

t i CRACKED WELD 10 F1 i LOWER SHELL I. ,

I l

FIGURE 3 RING ROTATION ABOUT TOP OF FILLET f l

(NOT TO SCALE)

Peach Bottom Shroud Repair NRC Questions and Responses Page 10

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. DRF B13-01732  ;

REV 0 )

g Seismic Analysis -

-Introduedon '

i- _

jThe information provided below are a selection of Seismic related questions asked by the NRC about the shroud repair installation. These seismic questions with l their appropriate responses to the NRC were selected from the NRC questions and

- GE prepared response' s for Nine Mile Pt.1, Qdad Cities, Hatch 1, and Pilgrim shroud repair projects.  ;

Seismic Analysis / Methods, Input, & Results

' The seismic analysis withs' hroud repair hardware installed, and with numerous i combinations of cracks at welds H1 through H8, was performed by time-history  :

method. The peak responses from these seismic analysis were used to obtain loads and deflections in the shroud, the shroud repair hardware and the RPV & RPV internals for use in subsequent stress analysis. The NRC questions address the subjects which are listed below-

.i

1. Seismic analysis methodology including request for copies of Criteria & ,

computer program User's Manual. l

2. Seismic Model details. 1 au3 3. Seismic Input detalk

& 4. Maximum gaps calculations at cracks with seismic loads included.

5. _ Seismic analysis results details / explanations.

Following is the gwuping ofquestions into the above RVE topics.

t Tonle # 1 >

Seismic analysis methodology including request for copies of Criteria & computer program User's Manual.

Ouesdon # 1 Pmvide Reference 2 of GENE-771-60-0994 (GE document NEDO-10909,Rev. 7, December 1979, SAP 4G07 Users Manual) which discusses the " standard, strain-energy weighted modaldamping ratios. " \

~ Besnonse # 1 The reference document is a Users Manual for computer program used by GE

)

in performing the seismic analysis. This document is available in NRC offices because it was provided as part of comed's March 3,1995 submittal. -

Therefore, it is not submitted as part of Peach Bottom submittal. See page .

VIII.L-4 of the manual for the related details applicable to this question.

n Peach Bottom Shroud Repair NRC Questions and Responses PageIi

DRF B 13-01732 REV 0

'm j

Ouestion # 2 It is stated that the component stress evaluations are determinedfrom the dynamic analysis of the horhontal seismic loads in combination with the vertical seismic loads.

How were the horhontal and vertscal loads combined (i.e., by absolute summation or by -

another method), and how were they combined according to the originalplant design basis t Similarly, how were the SSE and LOCA dynamic loads combinedfor the two . ,

faulted service conditions, and how were they combined in the original design of the core shroud? >

Resoonte # 2 There are three possible collinear contributions for each seismic load. One for each of the two horizontal components of seismic excitation and one for the vertical. For the Peach Bottom shroud repair hardware design, the maximum contribution from either of the two horizontal seismic analyses was combined absolute sum with the corresponding peak vertical contribution for each seismic load "See GENE-771-60-0994". This method for combining peak collinear contributions due the three orthogonal spatial components of excitation is identical to the original (and existing) licensing design basis for Peach Bottom.

Also, commensurate with the primary structure being vertically rigid with  ;

respect to the vertical seismic free-field input motion, the Peach Bottom

" seismic licensing design basis necessitates only horizontal dynamic analyses.

- Consequently,in conjunction with existing geometric symmetryin the primary structure, the peak horizontal components were obtained from separate N/S and E/W horizontal analyses and the peak vertical contributions were from static dead weight analyses.

The peak collinear contributions due to the SSE and LOCA dynamic loads were combined by absolute sum for the Peach Bottom shroud repair hardware design. It is not clear how they were combined in the original design of the i core shroud. However, the absolute sum methodology presently used is the j most conservative method employed for nuclear application.  !

l Topic # 2 Provide seismic model details such as core data, rotational spring constants, including their derivations, how a hinge or roller condition at cracks is determined as well as modeled. l l

Ouestion # 1 '

Provide the differences between the original core configuration and the Cycles 11 (Unit 2) - :

& 10 (Unit 3) core configuration l Peach Bottom Shroud Repair NRC Questions and Responses Page 12

DRF B13 01733 4 REV 0

r. . R*=nonse # 1 w Both the original and Cycles 10 & 11 core configurations contain 764 fuel bundles in each unit. The differences between the two core configurations are summarized below. ,

Oricinal Cycle 11 Core Cycle 10 Core .

CDIs

' Unit 2 Unit 3 Young's Modulus (ks0 1.67E6 1.59E6 1.59 E6 Poisson's Ratio 0.30 0.41 0.41 Total Shear Area (ft2) 4.48 3.99 4.30 Total Moment ofInertia 0.285 0.268 0.286 (ft4) -

Total Fuel + Water Mass (k- 16.4 15.22 15.58 sec2/ft)

Both the stiffness and mass of the Cycles 10 and 11 cores are smaller than those of the original core. The revised Cycles 10 and 11 core properties were included in the seismic analyses used for the design of the shroud repair hardware

,.74 Future core configurations are not expected to be significantly different from W ' the Cycles 10 & 11 core configurations and, hence the seismic loads in the restraint hardware and the RPV/ internals are not expected to be sensitive to future changes in the core configurations.

Quesdon # 2 Provide Reference 1 (GEDocument 383HA691, Rev. O, Peach Bottom Units 2 &3 -

RPVandInternals Seismic Analysh, April 6,1972).

Resnonse # 2 The reference is provided as Attachment A.

Quesdon # 3 Pwvide the detailed calculation of the rotary spring constant K4for both an uncracked ,

and cracked weld H8, and indicate how it is introduced in the analysis.

Resoonse # 3  :

The Peach Bottom Core shroud is physically supported by a shroud support plate and six vertical shroud support legs. The shroud support plate is an  ;

annular plate with its outer edge welded to the vessel wall and its inner edge  :

welded to the shroud support ring. The plate also serves as a support for the l jet pumps. The shroud support legs are steel bars of rectangular cross-section.  !

~

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Page 13 l Peach Bottom Shroud Repair NRC Questions and Responses l

l w <

DRF B13-01732 REV 0 s The vertical legs are connected in uniform circumferential spacing between W the bottom of the shroud support ring and the vessel bottom head.

Considering the moment resistance characteristics of the design configuration -

and the fact that the primary structure seismic model is a mathematical beam

• element model, the shroud support hardware is modeled as a rotational spring (K4). Both the shroud support plate and the support legs contribute to.the total stiffness of the K4 spring. The rotational stiffness of the two parts are computed separately, and the total K4 stiffness is the summation of the two.

For the shroud support plate, the rotational stiffness is derived using the appropriate moment-rotation resistance equation representing a fixed-support annual ring by applying a moment to a central rigid circular block. The reduction in plate area due to thejet pump holes are also taken into l consideration in computing the equivalent plate stiffness. For the shroud support legs, the axial stiffness of the individual legs is computed first. The rotational stiffness is derived by a moment-rotation relationship considering stiffness contributions from support legs placed at proper geometrical locations around the bottom of the shroud support ring. For a total rotational stiffness K4 of Peach Bottom, the rotational stiffness contributions from the support plate and support legs are, respectively, about one-tenth ( 10 % ) and

_ nine-tenth ( 90 % ) of the total stiffness value.

1. The H8 weld is the weld that connects the shroud support plate to the shroud support ring. When H8 has a full 360 degree through-wall crack, the worst case will resultin no rotational coupling between the support plate and the shroud. However, since the support plate only contributes about 10 % of the total rotational stiffness for K4 spring, there is stili at least 90 % of the rotational stiffness remaining to resist the seismic responses. As compared to the other crack cases, which assume full 360 degree through-wall cracks in the shroud wall, the H8 crack is not a controlling case in the shroud safety evaluation nor in the shroud repair hardware design. The reduction in the stiffness of the shroud support by approximately 3 % is a much less severe case than either the hinge or roller cases assumed in the design of the shroud repair hardware and thus is adequately enveloped by the various parametric analysis performed.

Ouestion # 4 -

Clanfy the basis and conditions under which the assumption of hingas or rollers is chosen to model the shroud. Provide detailed diagrams of the hinge and roller models of the shroud structure, showing the attachments to the RPV. For each loading combination and crack configuration state the model which was used in the analysis. -

~

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DRF Bl3-01732 REV 0

- Resoonse # 4 i

• 1 hinge was chosen to model a 360 degree through wall circumferential crack in the shroud when contact is maintained under a compressive load, based on mechanical interference at the crack resisting lateral motion. A roller was -

chosen to model a crack when sufficient pressure and seismic loads exist to

• overcome the mechanical and thermal preloads and dead weight such that the full crack surface may open up, losing mechanical interference. This only occurs for the main steam line LOCA in co'mbination with a MCE. Note that the results of the parametric study were used to determine the location of the crack that produced the largest response for the shroud and the shroud stabilizer hardware.

The detailed seismic modelis shown in Figure 1 of the GE Document GENE 771-60-0994, Revision 2," Peach Bottom Units 2 & 3 Seismic Analysis. Two coincident nodes are modeled at each shroud circumferential weld location; one attached to the upper beam element, and one attached to the lower beam element. If there is no crack, the nodes are coupled in both translation and rotation. If there is a crack which is modeled as a hinge, the nodes are coupled in translation only, with no moments transmitted through the nodes. If there is a crack which is modeled as a roller, the nodes are uncoupled, with no shear or moment transmitted through the nodes.

] For the design basis load combinations of DBE plus normal condition, and the MCE plus normal condition, assumed weld cracks are always modeled as hinges, since for the normal operating conditions a preload is always maintained. For the beyond design basis load combination of MCE plus recirculation line LOCA condition, the assumed weld cracks are always modeled as hinges, since for this condition a preload is always maintained. For the beyond design basis load combination of MCE plus main steam line LOCA condition, with a single crack (critical bounding location assumed), the crack is modeled as a roller, since the uplift forces overcome the preload. For the MCE plus main steam line LOCA condition, with multiple assumed cracks, all cracks are modeled as hinges except for the topmost crack which is modeled as a roller. In this case, the uplift forces overcome the preload for the topmost crack, while the deadweight of the lower portion of the shroud is sufficient to keep the remaining cracks closed (i.e. pinned condition).

Tople # 3 Seismic model input details Ouestion # 1 -

Descnbe how uncertainties were accountedfor in the structural model and the time-history input.

c- . .

- Peach Bottom Shroud Repair NRC Questions and Responses Page 15

1 DRF B13-01732 REV 0 g Resoonse # 1 The purpose in addressing the effects of parameter variations is to account for uncertainties in the calculated natural frequencies due to uncertainties in such parameters as material and soil properties, structural damping and soil -

damping, soil-structure interaction techniques, geometrical dimensions and

• approximations inherent to dynamic modeling and analysis. The concern is .

the effect the parameter variations will have on response of the RPV and internals.

The current NRC acceptance criteria for the consideration of the effects of parameter variations are'provided in Subsections 11.9 and 11.5 of Standard Review Plan (SRP), Section 3.7.2, " Seismic System Analysis", Rev. 2 - August 1989. Per the SRP, one acceptable approach is to smooth the computed floor response spectra obtained from the primag structure time histoy analyses and broaden the spectral peaks that are coincident with the structural natural frequencies for subsequent use in decoupled seconday system analyses. Also, from Subsection II.5 of SRP 3.7.2, it is acceptable to use a " single" artificial input time history (without expanding or contracting the time histoy time step) in the primary structure seismic analysis. However, since the original design basis seismic analysis in UFSAR did not require applying these criteria, the current analysis does not specifically document compliance with these criteria even though, the intent of Subsection II.5 of SRP 3.7.2 is met.

p

• For the Peach Bottom shroud repair, a coupled model time histoy analysis was i

performed in which the reactor and internals are modeled in detail in the primag structure model. Keyinput parameters such as the rotational stiffness of the tie rod assemblies along with the various crack locations andjoint restraint configurations were parametrically evaluated by performing analysis of bounding conditions. Conservative peak dynamic loads were then calculated in the primag structure analysis; without the need for a separate seconday system analysis procedure.

Two separate time histories ( DBE and MCE ) were used in the primag structure time history analyses for the seismic design adequacy evaluation of the shroud repair hardware. Time histoy analyses were performed using both of these synthetic time histories. Synthetic acceleration time-histories were constructed for both DBE and MCE in accordance with the guidelines in the USNRC Standard Review Plan ( NUREG-0800 ), such that the response spectra corresponding to the synthetic time-history envelopes the spectra of Figures C.3.1 & C.S.2 of Peach Bottom UFSAR. The UFSAR spectra ( Target spectra )

and the spectra corresponding to the synthetic time-histog used in the current analysis are shown in Figure 2 of the seismic report GENE-771-60-0994, Rev. 2.

Finally,it is noted that only the peak bounding dynamic loads from the Peach

- Bottom time history analyses were applied to perform the static stress analyses Peach Bottom Shroud Repair NRC Questions and Responses Page 16

DRF B13 01732 i REV 0

• m of the shroud and shroud repair hardware, and thus no secondary subsystem "7;. analysis was required. j

~

Tonic # 4 -

I Provide Maximum gaps calculations at cracks with seismic loads included.

Ouesdon # 1 )

Provide the magnitude of the maximum separation ofeach cracked weld during a MSLB  ;

plus verticalseismic loading.

Resnonse # 1 The magnitude of the maximum separation at a cracked weld during a MSLB plus the Maximum Credible Earthquake considering vertical seismic excitation will be 0.30 inches.

l Tonic # 5 t Provide seismic analysis results details such as a few mode shapes & modal dampings for uncracked, the hinged and the roller models; RPV loads etc.

Ouesdon # 1

=I=A ~ Pmuide a detailed discussion of the difference between the seismic axial and bending loads in the original skirt stress analysis report and the present analysis.

Resoonse # 1 The bending moments in the original analysis and present analysis are of the  ;

same nature, but have different numerical values as noted in Section 4.9 of the  ;

present stress report. The original bending moments were calculated based on l' the seismic horizontal design acceleradon coefficient of 0.6 g( in DBE ) and '

O.12 g (in MCE ) and taking the summation of the moments of the horizontal seismic loads located at the center of gravity (C.G.'s) of each load causing overturning moments. This was given in GE Drawing 886D499, sheet 7, [

Revision 8, table 10 for the original analysis. This drawing is being provided as  ;

Attachment B.

- 1 r

In the present analysis the bending moments are taken directly from the dynamic analysis of the seismic model, the details are available in GENE-771-60-0994, Revision 2, which is included in this submittal. This analysis was performed using time-history methodology. l

?

- i The original seismic axial load was based on a vertical acceleration coefficient

(=0.08g DBE & 0.16g MCE ) multiplied by the total downward load. This axial  !

load has not been changed in the present analysis.

Peach Bottom Shroud Repair NRC Questions and Responses Page1"/

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DRF 813-01732 REV 0

% Ouesdon # 2

..w ~ Pmvide the detailed calculation showing that the shroud legs have been checkedfor buckling with H8 cracked under preload and venical and hornantal seismic loading.

Response # 2 The original compressive stress in the shroud support legs using conservative .

design ( DBE ) loads is 12.1 ksi, which is below the allowable of 14 ksi (= 0.4 Sy) thus meeting tb buckling stress limits. The new compressive stress in the shroud supportlegs for MCE is 22.9 ksi, which is below the allowable of 28 ksi (

= 0.8 Sy ). Since the revised seismic moments from the seismic time history analyses on the shroud supportlegs are lower than those used in the original analysis, the existing buckling stability evaluation documented in the original stress report is still valid. The new loads are from the case of H8 uncracked but are the same for the H8 cracked case.

Quesdon # 3 Pmvide the number of modes, thefirstfive modal shapes and the corresponding modal damping values, for the uncracked, the hinged and the roller models.

Resnonse # 3 The seismic analysis included all modes with modal frequencies less than 33.

Hertz. Shown in the following tables are the number of modes, the first five modal frequencies and composite modal damping ratios along with the identification of the location of the predominate mass by mode. Provided in Figures 1,2, and 3 below are the first five mode shapes for the uncracked model, a representative hinged model (weld H7 cracked as a hinge), and a representathe roller model (weld H7 cracked as a roller). Because of similarity in modal properties between the East-West and North-South direction, it is sufficient to show only the modal properties in the North-South direction.

Note that for most of the internal components the N-S model produces the bounding responses.

~

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Peach Bottom Shroud Repair NRC Questions and Responses Page 18

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L , DRF B13 01733 REV.0 ,

4_ y H3 Hinge H3 Roller i Uncracked (22 modes) (22 modes) (23 modes)  !

Freq. Damping Freq. Damping Freq. Damping l (Hz) ( % erit.) (Hz) (% crit.) (Hz) (% crit.) ,

3.64 2.00 (2) 1.66 1.12 (1) 1.08 2.04 (1) .!

4.34 6.07 (4) 3.64 2.00' (1) 3.02 2.18 ' (1)  !

5.25- 3.48 (3) 4.87 6.65 (4) 3.64 5.02 (2) i 5.89 2.06 (2) 5.25 3.49 '(3) 5.24 3.50 (3) .

I

.7.17 1.74 '(1). 5.89 2.02 (2) 5.54 6.63 - (4)

Predominant Resoonse Location Legend l (1) Shroud, Fuel l (2) Reactor Building Roof Stnicture (3) Control Rod Drives l (4). Fuel  :

i I

l Y

'i i

b I

i

. j l

. i A -j

c. .

Peach Bottom Shroud Repair NRC Questions and Responses Page 19

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, . ;-- . 1._ ._ _ . . , . _ _ . _ . , _ _ _ _ . _ _ .

m, - --

DRF B1341732 REV 0 Figure 1 Mode Shapes t'or Uncracked Case 9

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Peach Bottom Shroud Repair NRC Questions and Responses Page 20

4[', a DRF BI3-01732 REV O

i. - ..

Figure 2 Mode Shapes for HS Hinge Crack Case -

L I:

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+

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'I Peach Bottom Shroud Repair NRC Questions and Responses Page 21 {

- + - * + =+c - _ _ _ ___ _ _ ___

DRF D13-01732 REV O

. tes

.. Figure 3 Mode Shapes for H3 Roller Crack Case b

_. i Peach Bottom Shroud Repair NRC Questions and Responses Page 22

DRF B13-01732 REV 0 A

ATTACHhiENTA ,

GE Document 383HA691, Rev. O, Peach Bottom Units 2 & 3 - RPV and Internals Seismic Analysis, April 6,1972.

m

.,y Peach Bottom Shroud Repair NRC Questions and Responses Page 23 j l

l u

DRF B13-01732 REV 0 10 4

l ATTACHMENTB .

GE Drawing 886D499, Sheet 7 of 8 Revision 8

.a w

(.e Peach Bottom Shroud Repair NRC Questions and Responses Page 24

2 DRF B13-01732 REV O- ,

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't

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- GENERAL ELECTRIC INSPECTION RECOMMENDATIONS  :

a=m> .

FOLLOWING THE INSTALLATION OF THE SHROUD STABIT HERS  ;

AT PEACH BOTTOM 2, S -

GE Nuclear Energy recommends that the following in vessel inspections be ,

. performed after the installation of the shroud stabilizer assemblies at Peach Bottom Unit 2 or Unit 3. The complete inspection recommendations are not yet finalized, and will depend in part on the criteria established by the BWRVIP.

Perform a VT-3 visual inspection of each shroud stabilizer assembly and  ;

interfacing areas on the shroud and vessel after one operating cycle, and once per interval thereafter. This inspection should include verification that the crimp nuts on the toggle bolts, and other locking devices are in place. The inspection should include the interface locations between the stabilizers and the shroud and reactor ,

pressure vessel. The anchor points at the shroud support plate, lower spring and ,

clevis pin should be included to verify structural rigidity. It is not considered f necessary to mechanically check the tie rod nut torque.

FJ Perform an enhanced VT-1 visualinspection or UT inspection of each of the four >

vertical shroud welds at the intersection with the H4 weld once even two fuel cycles or after even three years, whichever is longer. A minimum length of eight  !

inches should be inspected for each weld. It is not considered necessay to inspect  !

the remaining vertical welds which do not intersect the H4 weld. ,

i Follow the BWRVIP plan for the inspection of the H9 weld between the shroud support plate and the vessel, and any recommendations for inspection of the ,

. shroud support plate segment welds. An analysis for maximum allowable flaws in the H9 weld would help determine inspection criteria.  ;

i There is no inspection required for the top guide support ring (between the H2 and H3 welds), and the core plate support ring (between the H5 and H6 welds) as these are one piece rings without any welded ring segments. l There is no recommended inspection for the H8 weld between the shroud 1 support plate and the shroud support cylinder as it has been shown by analysis that i

' the structural integrity of the H8 weld is not required by the shroud stabilizers. ,

t

.6 i

Peach Bottom Shroud Repair NRC Questions and Responses Page 25 ,

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- . . , - - - - - , - . .a --, -- - - - . _ . - - - - - - - - . - - - _ _ _ - - - - -

DRF B13-01733 REV O p Materials w

Introduction The information provided below are a selection of material related questions ~

asked by the NRC about the shroud repair installation. These material questions with their appropriate responses to the NRC were selected from the NRC questions and GE prepared responses for Nine Mile Pt.1, Quad Cities and Peachbottom shroud repair projects.

Shroud Renair Materials The shroud is fabricated from 304 series stainless steel. The shroud repair hardware is fabricated from three (3) types of materials which are listed in the Peachbottom fabrication specification 25A5601 Rev 1. These repair hardware materials are;

'l. The X-750 is a nickel-chrome-iron (Ni-Cr-Fe) alloy shall be in accordance with ASTM B-637, UNS N07750, Type 3 with additional requirements as defined in paragraph 3.2.1 of the fabrication specification.

2. The austenitic 300 stainless steel shall be in accordance with ASTM A-479, A-182 or A-240 type 304,304L,316 or 316L v .h a maximum carbon W content of 0.020 percent with additional requirements as defined in paragraph 3.2.2 of the fabrication specification..

3. The XM-19 stainless steel shall be in accordance with ASTM A-479, A-182 or A-240 with a maximum carbon content of 0.040 percent with additional requirements as defined in paragraph 3.2.3 of the fabrication specification.

Subjects Addressed By NRC Material O_uestions The NRC questions address the follow subjects and are listed below.

Details of heat treatment

• l Sensitization testing Cold work during fabrication Lubricant used during installation Use of XM-19 material for tie rods Certified Material Test Reports (CMTR)

Peach Bottom Shroud Repair NRC Questions and Responses Page 26

DRF Bl3-01732 i' REV 0 Selected NRC Ouestions prc_1

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Listed below are a selection of material questions asked by the NRC during the review of the documentation submitted for Nine Mile Pt.1, Quad Cities and -

Peachbottom shroud repair projects for additional information. GE reviewed

• each NRC question and submitted the appropriate response for each material question. .

Ouestion 1 Please provide the heat treatment details such as time, temperature and cooling rate that are specifiedfor the alloy X-750 components.

Resnonse 1 Heat treatment for the X-750 consists of 1975* 25* for 60 to 70 minutes, followed by forced-air cooling. Age hardening is done at 1300 15 F for 20 to 21 hours2.430556e-4 days <br />0.00583 hours <br />3.472222e-5 weeks <br />7.9905e-6 months <br />, followed by air cooling.

Ouestion 2 Is solution annealed condition specified as thefinal matenal conditionfor all

' components made of XM-19 and 316 or 316L austenitic stainless steel materials ? l Resoonse 2

'Q 5olution heat treatment is required for all 300-series and for XM-19 parts. This is done before machining and trimming. Threads on the XM-19 tierods are not  !

annealed after machining. ,

Ouestion 3 What is the material condition ofXM-19 that were previously used in the BWR \

environment? ,

' Resoonse 3 i This material was selected for the shroud-reinforcement-assembly tie rods because ofits strength, corrosion resistance in BWR environments, and other engineering properties.

As purchased, this material is delivered in the solution heat treated condition. l This involves heating the item into the temperature range, approximately l 2000* F for UNS S20910, where the Cr,,C, carbides break dmm and the component atoms disperse into the matrix. This heating is followed by cooling at a rate rapid enough to maintain the dispersion in the matrix.

For some materials, and for small parts, the rapid cooling is most conveniently achieved by quenching in a water bath. But for the stock for shroud-restraint tie rods, which are approximately 3-1/2 inches in diameter by 14 feet long,  !

attempts at water quenching will introduce severe local distortions because the l e

Peach Bottom Shroud Repair NRC Questions and Responses Page 27 l

DRF B13-01732 REV 0 g entire length of such an object rarely,if ever, reaches the same temperatures at the same times. Therefore, air cooling was selected to reduce the likelihood of problems due to nonuniform temperature distribution.

For this specific alloy, air cooling is sufliciently rapid to avoid reprecipitation of ~

carbides, as has been verified by a number oflaboratory tests and experiments.

For example, data published by EPRI's M.J. Povich and D. E. Broeker in the October 1979 issue of Materials Performanie (National Association of Corrosion Engineers;"The Stress Corrosion Cracking of Austenitic Stainless Steel Alloys in High Temperature Air Saturated Water") show that UNS S20910 is as resistant to corrosive attack in the cold-worked and/or furnace-sensitized condition as in the fully solution-heat-treated condition. Therefore, for parts of the size, shape, and basic material of the tie rods, more is to he lost than gained from water cooling after solution heat treatment. Air cooling is sufficiently rapid to preserve the corrosion-resistant properties of the material without the risk of unacceptable loss of straightness.

This, and other similar data in the literature, also supports the decision to place the completed tie rods into service in the as-machined condition, without incurring the schedule delays and added expenses that would have resulted from an unnecessary post-machining heat treatment cycle.

m Ouestion 4 W in thefabdcation specification, GE stated that the successful completion of the '

sensitization testing shall be accepted as evidence of the comet solution heat treatment sf time and temperature charts are not available. The staffconsiders that in a good quality assurance program, accurate records ofitem, temperature and cooling rate are necessary to be maintained as evidences that proper heat treatment has been perfonned Therefore, the use ofsensiti:ation test results as a substitutefor the proper heat treatment documentation is not acceptable.

Resnonse 4 i The purpose of the requirements on heat treatment of stainless steel core l shroud repair materials is to provide material that is not sensitized.  !

Accordingly, sensitization testing of the material after heat treatment is an accurate indicator that the heat treatment was effective. The attributes of the final material condition is considered as evidence that the engineering requirements of material performance have been met. Therefore, sensitization -

testing is considered an adequate alternative to detailed heat treatment records in assuring that proper heat treatment has been performed. Also, complete reliance on heat treatment records can be misleading. GE has indicated that j isolated cases have occurred in which heat treatment details were recorded. -i even to the extent of using embedded thermocouples, but that the subject material failed to pass a sensitization test.

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i Peach Bottom Shroud Repair NRC Questions and Responses Page 28 J

DRF B13 01732 RD!O m ASME NCA-3800 was followed to procure the core shroud repair material.

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With respect to material test reports, NCA-3860 does not require that detailed time / temperature records for heat treatment be recorded but that specific time and/or temperature parameters be recorded if such values are specified -

in the underlying Section II material specification. For the austenitic stainless ~

steels used for the shroud restraints, the only stated requirement is a minimum temperature of1900 degrees F (followed by rapid cooling).

The fabrication specification adds additional requirements for heat treating stainless steel that is over and above the ASME code. When material is ordered in heat lots from a primary metter,it is possible to get such detailed records of heat treatment. However,in the current environment of performing internals repairs, materials are procured in small quantities, often from a third party supplier out of a warehouse inventog. In these cases, detailed records showing complete conformance to the additional requirements are not always retrievable. It is in these cases where, as an alternative to detailed heat treatment records, the attributes of the final material condition is considered as evidence that the engineering requirements for materials performance have been met.

In summary, the material ordering requirements are appropriate for the intended use and are in conformance with applicable codes.

Ea

- Ouestion 5.

GeneralElectric stated that a minimum of 0.030 inches of austenitic 300 series stainless steel materials will be removed after high temperature annealing as a control of intergranular attack (IGA). Will this process or any other process be applied to machined threads made ofXM-19 and Alloy X-750 after high temperature annealing to ensure there is no IGA.

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Please provide test data to support that the removal of 0.030 inches surface matenal would effectively eliminate the IGA effect resultingfrom the high temperature annealing.

Resoonse 5.

Material removal processes are not used on the XM-19 tie rod threads after the final machining. The minimum of 0.030-inches to be removed after annealing or high temperature heat treatment of 300 series stainless steel, XM-19 and Alloy X-750 is intended to address parts which may be in an air atmosphere v furnace for several hours. At temperatures of 1800-2000 degrees F,in air for several hours a thick surface oxide layer and grain boundag oxidation (IGA) may occur.

' The local solution annealing of the threads is a very short cycle with very short heat up and cool down times. Considering the outstanding corrosion and j oxidation resistance of XM 19,it is expected that the oxidation of the surface Peach Bottom Shroud Repair NRC Questions and Responses Page 29

DRF B13-01732 REM 0 A will not be of a significant depth, and IGA is not expected during this very w short process. Therefore, no material removal processes are used on the XM-19 tie rod threads after the local induction annealing.

Re-solution anneal or additional material removal processes are not performed ~

on X-750 parts after machining. Typical UNS N07750 (X-750) samples were selected at random from their actual production runs, cross-sectioned, and the microhardness measured as a function of depth into the metal from the polished surface. The sequence of manufacture was:

Samole "A" Samole "B" Machined Age Hardened Polished Machined Age Hardened Polished Penetrant Examined Penetrant Examined Dimensions Checked Dimensions Checked ,

In both specimens, the hardness at and near the polished surfaces was identical to that of the unaffected interior: Rc 36 for "A" and Rc 35 for "B". All readings remained within a plus or minus 3 point tolerance band, which is uniform and consistent for 100-gram Knoop readings. The surface showed no evidence of M work hardening or cold work. Additionally, samples representing turning and D milling operations on X-750 after age hardening showed no significant increase of surface hardness compared to internal bulk hardness (38Rc max surface verses 34Rc interior). Therefore, no re-solution annealis required for X-750 parts after machining.

Over the past 25 to 30 years GE has implemented a metallographic receipt inspection requirement for heats of stainless steel. The receipt inspection requirements consist of a destructive metallographic examination of the cross section from each heat to determine the depth ofIGA that may have occurred due to high temperature annealing atmosphere or due to an overaggressive acid pickling process. The acid pickling process is the primary cause ofIGA.

The results of the numerous tests performed have shown that less than one sample per year have shown IGA deeper than 0.001 inch and in those cases the depth of the IGA was less than 0.003 inches.

The criteria for removal of 0.030 inches of material was originally established many years ago to be a conservative bounding limit to ensure that any IGA induced by any process, especially due to overaggressive pickling, would be removed. This criteria was established based on engineeringjudgment p.ior -

to the results of the material receipt inspection testing as described above. The requirement for removal of 0.030 inches from the affected surfaces has been Peach Bottom Shroud Repair NRC Questions and Responses Page 30

DRF B13 01732 REV 0 g

,s implemented for this project to provide an order of magnitude of the margin over the maximum depth ofIGA tha: has been observed.

Ouestion 6 -

Provide details ofyour controls in the practices of machining, grinding and threading to '

minimize the effect ofcold work, such as amount of material to be removed in each pass, application ofcoolant and sharpness of the tool..

Resoonse 6 The amount of material removed in a single pass depends on the part and the particular machine doirig the work. Parts are generally rough machined to within 0.10" of final size and skim passes are used to achieve the final dimensions. Tests show that Inconel X-750 surfaces are unaffected by machining and do not require heat treatment. The application of coolant and sharpness of the tool is considered adequate provided a surface finish of 125 root mean square or better is obtained.

O_uestion 7 Please provide details ofyour controls in the practices of machining, grinding and threading to minimi:e the efect ofcold work, such as amount of materials to be removed in each pass, application ofcoolant and sharpness of the tool.

- Resoonse 7

, Each item thatis manufactured has its own specific requirements when it comes to "how much" material is removed per pass and which machine is doing the work. Generally speaking, parts are " rough machined" down to within .100" of final dimensions. Then the final clean up (about .010") pass skims off the required amount of material to achieve the required size and surface finish. If a tool is dull, then the 125 rms surface finish would not be produced as required on all drawings. A dull tool produces a smeared or torn surface appearance which is the primary method of monitoring the adequacy of the tooling and the process in general.

Thejudgment and experience of the machinistis relied upon to determine how much material can be safely removed per cut or per pass. Written documents could not possibly address all possible eventualities of workpiece size, shape, and material or machine type and capacity or dimensions, tolerances, and surface finish necessary. Vendor in process control sheets or travelers are used to control the flow of material in the shop. While in process, machining is seldom,if ever, controlled by fixed documents. The end results are carefully specified on the drawings.

In addition, the fabrication specification states " Machined components that are not solution annealed after machining shall have metallographic and microhardness evaluation on test samples. Samples shall be provided from the v same material, same fabrication shop and using the same process variables."

Peach Bottom Shroud Repair NRC Questions and Responses Page 31

.1 DRF B13-01732 REV O The purpose of these evaluations is to verify the materials' surface conditions i

"**" have very shallow cold work depth. Control of the cold work depth will l minimize the materials susceptibility to IGSCC.

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The coolant used during the machining process is Trim-Sol. A stream of Trim- ~

Sol is applied directly to the cutting tool where it makes contact with the part.

Afterward, the component is washed with acetone and followed by a demineralized water wash prior to any other operation.

Ouestinn 8 In yourfabrication specification it is stated that ,ifgdnding is notfollowed by solution heat treatment or machining, the grind surface shall be polished to an PS15 32Jinish or better using successivelyfinergrit abrasive. Please provide test data to support the above stated polishing to be effective in eliminating the unacceptable surface cold work resulting from grinding l Resnonse 8 The fabrication specification was based upon several evolutions of shroud-reinforcement documents and has retained several contingency options that at earlier times were thought to be prudent to have available "just in case".

Shaping metal by heavy grinding was one of these options. Experience now shows that such a technique is not necessary and, in fact, has not been used.  :

ds While experiments have been conducted to verify the metallurgical acceptability of the manufacturing operations that we do use or intend to use, no specific tests were made of' heavy grinding with and without subsequent polishing'. Thus there are no test data available to be provided on this specific phenomenon Ouestion 9  !

In thefabrication specification it is stated that electropolishing can be specified to remove any cold-worked surface by using mixed phosphoric / sulfuric acid. Please describe the ,

details of the electropolishing process and its controlling parameters, as well as how this process was qualified. Is there any test requirement after electropolishing to ensure that j there is not pitting or IGA on the electropolished surface?

Response 9 .

Electropolishing to remove surface material for elimination of mechanical-  ;

working effects, refinement of surface texture, fine adjustments to dimensional l

attributes, or similar reasons; is permitted under the following mandatory requirements:

a) All work shall be performed in accordance with written and approved j procedures, Such documents shall control, as a minimum: basic '

J

. electrolyte composition with limits, maximum permissible bath Peach Bottom Shroud Repair NRC Questions and Responses Page 32

DkF B13-01733 REV 0 g contaminants by type and level, immersion temperatures with limits, current polarity and maximum amperes /cm' density, precleaning and postcleaning requirements, and limitations to preclude insufficient or '

excessive material removal.

b) The electrolyte shall be based upon the mixed phosphoric / sulfuric acid system, with appropriate buffers and/o.r intensifiers permitted.

c) Demonstration samples shall be prepared and submitted for information in conjunction with the written procedure (subparagraph (a) above). These shall clearly show the alteration in surface texture and appearance after electropolishing and compared with before electropolishing, and shall include polished and etched metallographic cross sections of before and after specimens to illustrate typical internal structures.

O_uestion 10 Please discuss the mitigation methods'that you plan to apply to the machined threads such as re-solution annealing to minimi:e the cold work effect. Please also describe how the methods were qualified and the details ofcontrolsfor application.

Resoonse 10 aans The XM-19 material used for the tierods is procured in the solution annealed W condition. The machined threads on the tierods are not solution annealed after the machining of the threads.

General Electric has been specifying and using XM-19 components in commercial power reactors since the mid-1970s, especially for parts which must be stainless steel and also must be nitrided. The nitriding process involves holding the parts in a chlorine-rich nitrogen environment at 1060" to 1100*F for 16 to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Only XM-19 is able to withstand this treatment without becoming sensitized to Intergranular Stress Corrosion Cracking (IGSCC).

Types 304,304L,316, and 316L quickly fail after exposure to such thermal conditions, but not XM-19. Therefore it is not necessary to remove or anneal any "IGSCG-prone" surface layers after machining, because the material is inherently resistant to IGSCC, even in the as-machined condition.

GE has access to test results which show that XM 19, in BWR environments,is ,

more corrosion and crack resistant in tight crevice situations than 316L or other 300-series stainless. To cite an example; a set of tensile specimens, each bearing a sleeve to create 'a crevice at the specimen surface, was loaded to 120% of yield in 1981 and that load maintained for the duration of the test. .

Today,14 years later, none of the XM-19 specimens has failed or even cracked.

Thus, concerns about the durability of XM-19 in threadedjoints because of m

" crevices" are not valid.

Feach Bottom Shroud Repair NRC Questions and Responses Page 33 j i

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DRF Bl3-01733 REV 0 gg All in all, XM-19 is an excellent material for use in BWR applications. It may

- safely be used in the as-machined condition; it may safely be used in threaded and other highly-creviced geometries, and is superior to 316,316L, and other alloys for the intended application. Its only drawback is cost and availability-it -

is not as commonly found in supplier's inventories as the more traditional ~

materials.

Ouestion 11 Please provide details ofyour controls in the practices of machining, gnnding and threading to minimize the effect ofcold work, such as amount of materials to be removed in each pass, application ofcoolant and sharpness of the tool.

Resoonse 11 Each manufactured piece has its own specific requirements when it comes to "how much" material is removed per pass and which machine is doing the work. Generally speaking, parts are " rough machined" down to within .100" of final dimensions. Then the final clean up (about .010") pass skims off this rough surface to achieve the required size and surface finish. If a tool is dull, then the 125 rms surface finish which is required on all drawings would not be produced. A dull tool produces a smeared or torn surface appearance which is the primary method of monitoring the adequacy of the tooling and the

2. _ machining process in general. A number of tests on various parts in various N heat treated conditions has demonstrated that the most severe operation in terms of surface cold work is generation of the stub Acme threads on the tie rods which serve as a " worst case" bounding condition.

Thejudgment and experience of the machinistis relied upon to determine how much material can be safely removed per cut or per pass. Written documents could not possibly address all possible eventualities of workpiece size, shape, and material or machine type and capacity or dimensions, tolerances, and surface finish necessary to produce. Vendor inprocess control sheets or travelers are used to control the flow of material in the shop. While in process, machining is seldom, if ever, controlled by fixed documents, the end results are carefully specified on the drawings.

OuestiorLJ2 Identify the lubricants that would be used on the machined threads during installation.

What are the controls of the content ofchloride,s, sulfides, halogens and other elements that are known to promote stress corrosion cracking in stainless steel and high nickel alloy?

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Resoonse 12 The lubricant used for machined threads during installation is thread lubricant D50YP5B. The lubricant is referenced on GE reactor modification drawing t~~ parts list.

Peach Bottom Shroud Repair NRC Questions and Responses Page 34

1 DRF Bi3 01732 REV.O A

GE Specification D50YP12 provides the controls of the content of halogens, nitrate, sulfur and other elements in lubricant D50YP5B. The specification defines the requirements for limiting impurities in lubricants that are in ,

crevices exposed to BWR primary reactor coolar.t at temperatures above 165"F.

. Impuritylimits are as follows: -

1. The maximum allowable level of halogens, when both sulfur and nitrates  !

are less than 1 ppm,is 450 ppin.

2. The maximum allowable level of sulfur, when both halogens and nitrates are less than 1 ppm,is 630 ppm. ,

3. The maximum allowable level of nitrate, when both total halogens and total sulfur are less than 1 ppm, is 820 ppm.

4. Allowable simultaneous levels of halogens, sulfur and nitrates in combination can be determined from the nomograph in Addendum A in the following manner when the values of any two of the impurities are known, or are assumed. Place a straightedge through the two known or assumed values on their scales and read the third value from its scale. The ,

third value must not be exceeded. -

.aias

.g 5. Allowable combined levels of halogens, sulfur and nitrates can also be determined using the formula below: l com Halocens

+ oom Sulfur + com Nitrates <13.2 35.453 48.096 ~ 62.004

6. The maximum allowable level of any single low melting point metal is 200 i

ppm.

7. The maximum allowable level of all low melting point metals in combination is 500 ppm.

Quesdon 13 The stafrealizes that the repair assemblies may be inspected by a combination ofvisual and ultrasonic examinations. However, the staff has some concern regarding the  ;

reliability ofsuch inspection to identsfy the potential degradation in the threadaljoints and areas ofcrevices and stress concentrations, which have limited accessfor inspection. l Please provide a discussion and/orpwpose an alternative inspection such as '

disassembling the threadedjointsfor inspection to ensure that the areas mentioned above ~

in the repair assemblies will be adequately inspectedfor early detection ofpotential degradation.

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3 DRF B13-01732

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.g g Response 13 The tendency toward stress corrosion cracking is promoted by material type, condition, local water chemistry, applied loads, residual stresses, etc. In the i case of stainless steel threaded fasteners, crevices, surface condition (surface -

i cold work) and sustained tensile stress are of specific concern. For the shroud

• l restraint hardware several factors mitigate the concern for potential stress corrosion cracking.  !

Material-XM-19 y

In the middle 1970's in the interest ofimprovmg the margin of control rod drive (CRD) performance, GE implemented the use of XM-19 for piston ':

and index tubes in place ofType 304 stainless steel. The logic was that as a  :

low carbon, high chromium, mildly stabilized (Nb,V) austenitic alloy XM-19 would offer a higher margin of resistance to intergranular stress corrosion  ;

cracking (IGSCC) in the nitrided condition than Type 304. Nitriding i involves heating the material to approximately 1100 degrees F for several l hours and results in furnace sensitization of 300 series stainless steels, but j has no' detectable effect on XM-19. As a side benefit, XM-19 has a .

c significantly higher strength than Type 304 so equivalent components are  ;

stressed to a lower fraction of yield stress in service. Since the late 1970's all l control rod drives manufactured by GE have contained XM-19 piston and i

^ index tubes. This includes all BWR-6s (more than 1500 drives) plus several 'l

- other BWR 4/5's under construction at the time; as well as, all replacement ,

drives manufactured since. In total there are easily more than 2000 such j control rod drives in service.

By the nature of the CRD design there are numerous crevices including j threadedjoints exposed to the reactor environment. On the average 10 to i 20 per cent of the drives at a given plant are refurbished each outage. i During this work the drives are disassembled giving ample opportunity for examination and detection of problems. To date no instances of intergranular attack or IGSCC of nitrided XM-19 have been reported.

XM-19 shroud repair hardware material is provided solution annealed at 2000

• 50 degrees F after completion of final reduction, sizing, and forming i

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operations. All XM-19 material has had sensitization testing performed for each heat and each heat treat lot. The sensitization requirements exceed

.the requirements ASTM A-262.

Material-X-750 X-750 shroud repair hardware material is provided solution annealed at -

. 1975 25 degrees F after completion of final reduction, sizing, and forming

.qs operations. In addition, this material is age hardened at 1300 15 degrees.

Peach Bottom Shroud Repair NRC Questions and Responses Page 36

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DRF Bl3-01732 1 REV 0' l

1. F. 'All X-750 has~ Intergranular attack (igd) testing performed after j-owl annealing for each heat and each heat treat lot.

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It should be noted that there are no welds in the shroud restraint design, so + t there are no weld residual stresses. Also, there is no grinding in the shroud ,

• ~t

- restraint design, so there are no grin' ding induced residual stresses or cold work. As a consequence, the threaded fasteners in the restraint design i experience a relatively low level of sustained tensile stress compared to welded crevicejoints.

Finally, machined components that are not solution annealed after l machining shall have metallographic and microhardness evaluation on test j samples. Samples shall be provided from the same material, same  ;

fabrication shop and itsing the same process variables as the components - l which are being fabricated for the repair. The purpose of these evaluations '

is to verify the material's surface condition have very shallow cold work depth. Control of the cold work depth will minimize the materials j susceptibility to IGSCC initiation.  :

- Based on the material properties and fabrication processes described above, future disassembly of threaded fasteners, crevices, and stress concentrations areas for the express purpose ofinspection is not intended.  ;

iN-However,if these areas require disassemblyin the future for other reasons,  !

a visual inspection of the threaded, creviced, and stress concentration areas -

will be performed prior to reassembly. The detailed plans for future  !

inservice inspection of the installed core shroud repair components have ,

not yet been finalized. Quad Cities will submit these plans to the NRC staff '  :

at least ninety days prior to the first refueling outage following the outage  ;

in which the shroud repair components are installed."

i I

Ouesdon 14 Please discuss the reasons that GE selects XM 19 matenalfor the tie rods instead of l austenitic 304 or 316 stainless steel (low carbon content). The 304 or 316 stainless steel l has extensive service experience in the BWR envinnment. It should be noted that the i acceptableyield strength ofXM 19 matenal is limited to 90 ksi. Since there is limited  !

. service exponence ofXM-19 materialin the BWR environment, the staffrecommends that l an accelerated stress corrosion testing of a mock up simulating the XM-19 tie rod thread l joint in a BWR entnronment should be performed to ensure there is no development of l unexpected degradation. )

• ' Resnonse 14 .  :

XM-19 has experienced no known failures or other problems in approximately twenty years of BWR service. This is considered to be adequate confirmation l that the material is acceptable for use.

i Peach Bottom Shroud Repair NRC Questions and Responses Page 37 1

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DRF B13-01733 REV 0 pJ XM-19 was extensively studied and tested in the mid-1970s. Results of these tests were published in Document NEDE-21653, of which the NRC received copies in 1977 This document contains all of the applicable test information and will be provided in a separate submittal. XM-19 has experienced no -

known failures or other problems in approximately twenty years of BWR '

service. This is considered to be adequate confirmation that the material is acceptable for use. ,

Ooerating Exoerience With XM-19.

In the middle 1970's in the interest ofimproving the margin of control rod drive (CRD) performance, GE developed and implemented the use of XM-19 for piston and index tubes in place of Type 304 stainless steel. The logic was that as a low carbon, high chromium, mildly stabilized (Nb,V), austenitic alloy, XM-19 would offer a higher margin of resistance to intergranular stress corrosion cracking (IGSCC) in the nitrided condition than Type 304.

Nitriding involves heating to -1100 degrees F for several hours and results in furnace sensitization of 300 series stainless steels. As a side benefit, XM-19 has a significantly higher strength than Type 304 so equivalent components are stressed to a lower fraction of yield stress in service. Since the late 1970's all control rod drives manufactured by GE have contained XM-19 piston and

~

index tubes. This includes all BWR-6s (more than 1500 drives) plus several MF other BWR-4/5 s under construction at the time as well as all replacement drives manufactured since. In total there are easily more than 2000 such control rod drives in service.

By the nature of the CRD design there are numerous crevices including threadedjoints exposed to the reactor environment. On the average 10 to 20 per cent of the drives at a given plant are refurbished each outag'e. During this work the drives are disassembled giving ample opportunity for examination and detection of problems. To date no instances ofintergranular attack or IGSCC of nitrided XM-19 have been reported.

O_uestion 15 Please identify the ASME Code Matedal Specification Nos. that are specifledfor the procurement of the XM-19, 316 or 316L austenitic stainless steel and alloy X-750.

Resoonse 15 All of the materialr are purchased to American Society for Testing Materials (ASTM) specifications: XM-19 = A182, A240, A336, A412, or A479. 316 or 316L = A182, A240, A336, or A479. X-750 = B637. The materials are identified -

in the GE fabrication specification.

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h DRF B13-01732 REV O

f. Quesdon 16 I ** ~ Are Certifud Matedal Test Report (CMTR) and heat treatment records availablefor all

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pmcured matedals?

Response 16

• All materials procured for these assemblies are purchased with CMTRs. For the most part, and in keeping with the ASTM specifications, heat treatment information on CMTRs is limited.to a statement that the material is "in the solution annealed condition", and the holding temperature is given. For the most part the mills do not include furnace charts or other specific details of L the process to the suppliers from whom we buy the material. However, we require confirmatory testing, either to ASTM procedures or to GE equivalents of those procedures (such as E50YP20), which assures that the material was adequately and properly heat treated.

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Peach Bottom Shroud Repair NRC Questions and Responses Page 39 L _ - - _ - _ _ _ _ - _ _ _ - _ - _ - _ _ _ _ _ _ - - . - - _ _ _

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DRF B13-01733 REV 0

-g CORE PLATE WEDGE ISSUE

1. Introduedon The purpose of this documentis to describe the analysis perfonned to determine

• if core plate wedges were needed for the Peach Bottom shroud repair. The specific concern to be addressed was that of the horizontal motion of the core plate relative to the shroud during a.scismic event. That motion when cornbined with the shroud horizontal seismic motion at the core plate elevation must meet the deformation limits specified in the design specification.

The method used to evaluate the horizontal motion of the core plate as a whole was to compare the friction force provided by the initial preload clamping force of the core plate studs to the horizontal seismic loads. Conservative localized deflections were assumed to occur between the studs and corresponding core plate holes and due to localized bending of the core plate studs. The total deflection of the core plate resulting from the shroud displacement plus the localized displacement of Ae core plate reladve to the shroud was shown to be within the allowable displacements for all load combinations. , ;-

2. Analysis a The configuration wie restraint hardware installed but no shroud cracking is non-e limiting since the resulting shroud seismic loading is unchanged from the original design basis and the restraint hardware sees very low loads. Therefore, the case of seismic loading with stabilizers installed and a cracked shrcvd is evaluated.

2.1 Gross Core Plate Deflection The evaluation determined the net uplift force on the core plate due the combined effects of pressure differences, component weights and vertical seismic accelerations. The values are summarized in Table 1. The Peach Bottom core plate is installed with thirty-four 2% inch diameter studs with each stud preloaded to approximately 43,500 pounds. Considering relaxation in the nuds, the resulting clamping force on the core plate at operating conditions is approximately 1,257,000 pounds. Combining the stud preload with the operational and seismic uplift forces produces the net clamping forces reported in Table 1. Two values of coeflicient of friction were considered in the analysis of the horizontal friction force maintained between the core plate and shroud to resist the horizontal seismic loading; 0.2 as specified by the VIP / BWROG and'0.5 which has been widely used in reactor internals design. The more conservative friction values for the 0.2 friction coeflicient are given is Table 1. Also shown in -

Table 1 are the horizontal seismic loads on the core plate which result from the fuel and control rod guide tubes. A comparison of the horizontal seismic loads on y the core plate to the friction force available to resist the seismic loading shows the Peach Bottom Shroud Repair NRC Questions and Responses Page 40

DRF B13 01733 REVJ l

I available friction force to be the significantly larger of the two. Therefore, no p] '

gross sliding or relative motion of the core plate relative to the shroud will occur due to the horizontal seismic shear loads.

2.2 Localized Core Plate Deflection , ,

The seismic shear load that is transferred through the core plate into the shroud is not transferred uniformly around the circumference. For an uncracked shroud there is a tendency for more load transfer to occur at 90 degrees from the seismic line of action where the shroud is stiffer. For a cracked shroud with restraints installed, the lower spring provides a hard spot on the shroud at 45 degrees from the seismic line of action and a larger percentage of the seismic load would be transferred there. These non-uniform load paths may cause the friction between the core plate and the shroud to be overcome producing local sliding of the core plate relative to the shroud. The magnitude of the relative sliding will be limited to the sum of the gaps between the stud and the stud hole in the core plate plus the gap between the stud and its hole in the shroud. After these gaps are closed the stud acts as a shear pin with a horizontal stiffness of approximately 174 kip / inch. The total nominal radial gap is 5/16 inch. With stiffness of 174 kip / inch, the relative elastic deformation due to stud bending would be about 0.025 inch if only 17 of the 34 studs are conservatively assumed to resist the 74 kip seismic shear load for the Emergency case. The largest local horizontal motion of a the core plate relative to the shroud could therefore be 0.34 inch. From the Peach

- Bottom seismic results for the shroud repair that appear in GENE-771-60-0994, the -

gross clastic horizontal seismic motions of the shroud at the core plate elevation are 0.32 inch,0.67 inch and 0.88 inch for N+U, E and F load combinations respectively. Adding the 0.34 inch local deformation from above to these gross shroud motions results in values of 0.66,1.01, and 1.22 inch for N+U, E and F load combinations which are less than the allowable elastic displacement values from the Design Specification of 0.75,1.12, and 1.49 inch respectively.

3. Conclusions 3.1 No gross sliding of the core plate as a unit relative to the shroud will occur for any of the load combinations in the Peach Bottom Design Specification for the shroud repair project.

3.2 Local sliding of the core plate relative to the shroud may occur near one or more core plate studs. The magnitude of the local sliding is limited to the gap between the stud and the stud holes in the core plate and shroud plus the elastic deformation of the stud in shear and bending. This resultant local sliding, when added to the gross shroud seismic displacement, is less than the allowable elastic displacements for all load combinations.

- m.

O Peach Bottom Shroud Repair NRC Questions and Responses Page 41

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3.3 Wedges between the core plate and shroud are not required.  ;

.l Table 1 -

{

. -l Uplift- Net Core .l Load- Force on Plate Available Horizontal  ;

Condition Core Plate Clamping Friction Seismic l (kip) Force Load Load j (kip) (kip) (kip) i Norm + Upset 258: 999 200 42 l Emergency ,260 997 199 74

-l Faulted 341 916 183 69  !

l i

f I

l f

i r-a i

.W -

I i

I i

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-l l~

l i

. i

-l  ;

t,

---

l i

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'1. SCOPE ,

1.1 This document is the ASME Code Secdon III Paragraph N-142 Streu Report for the Reactor Pressure Vessel. This analysis addresses the new loads applied to the vessel as a result of the installation of the shroud stabilizers, which function to replace horizontal welds H1 through H7 and cracking of weld H8 in the core shroud. -

2. APPLICABLEDOCUMENTS 2.1 General Electric Documents. The following documents form a part of this stress report to the extent specified herein.

2.1.1 Suoporting Documents

a. Code Design Specification 25A5580 Rev. 4

b. Shroud Repair Hardware Design Specification 25A5579 Rev. 3 2.1.2 Supplemental Documents. Documenu under the following identities are to be usedwith this stress report.

None

. 2.2 Codes and Standards. The follo#ng documents of the specified issue form a part of this specification to the extent specified herein.

2.2.1 - American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code

. a.Section III,1965 Edition and Addenda through Wmter 1965 .

2.2.2 Other Documents

a. General Electric Drawing 886D449 P2, Sht.1 Rev. Il

b. Ceneral Electric Drawing SS6D449 P2, Sht. 7, Rev. 8, " Vessel Loadings" .

c. Babcockand Wilcox Report dated May 16,1973," Stress Report for Peach Bottom Unit 2 Reactor Vessel" (VPF #1896-1461)

(1) Report No. 20 "ShcIl Analysis", Rev. 0 (2) Report No.11 " Shroud Support System Analysis", Rev. 2 (3) Report No.10 " Brackets". Rev. O.

JUN 21 '95 17: 49 408 925 4533 PAGE.003

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d. Babcock and Wilcox Report dated September 5,1973, " Stress Report for Peach Bottom Unit 3 Reactor Venel" (#610 0146-51/52)

(1) Report No. 20 "Shell Analysis", Rev. 0 (2) Report No.11 " Shroud Support 6ystem Analysis", Rev. 2 (3) Report No.10 "Brackeu", Rev. O.

(4) Report No. 8, " Support Skirt Analysis", Rev. O

e. " Theory of Plates and Shells", by S. Timoshenko,2nd Edition

f. "Roark's Formulas for Stress and Strain", by W.C. Young, 6th Edition  ;

g. General Electric Drawing No.112D6490, Rev. O, " Detail Support, Lower."

h. " Reactor Preuure Vessel Power Rerate Stress Report Reconciliation for Peach Bottom Nuclear Power Plant Units 2 and $", G.E. Report No. NEDC 32166, Class II, datedJanuary 1993

1. " Fatigue Evaluadon of the Peach Bottom 11 and III Reactor Vessels", G.E. Report No. GE-NE-523-610493, dated May 1993.

J. General Electric Drawing No. 729E762, Rev. O, " Reactor Thermal Cycles." -

k. General Electric Drawing No.VPF # 189M4 7, " Shroud Support."

3. GENERAL DESCRIPTION 3.1 The purpose of the shroud stabilizers is to stmeturally replace all of the horizontal girdi welds in the core shroud and shroud support. These welds were required to both horizontally and vertically support the core top guide, core support plate, and shroud head, and to prevent core bypass flow to the downcomer region. The core top guide and core support plate horizontally support the fuel assemblies and maintain the correct fuel channel spacing to permit control rodinsertion. ,

3.2 The design requirements for the shroud stabilizers were separated into two documents.

The first document addressed those requirements that were not under thejurisdiction of the '

ASME Code (Paragraph 2.1.1.b). The second document addressed those requirements that were under thejutisdiction of the ASME Code (Paragraph 2.1.1.2). l

' ~

3.3 This Stress Report documents the acceptability of the stmeturalintegrity tequirements of the Code Design Specification defined in Paragraph 2.1.1.a.

JUN 21 '95 17: 49 408 925 4533 PAGE.004

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4. ANALYSIS 4.1 The Design Specification (2.1.1.a) defines three new design mechanical loads on the reactor pressure vessel. These loads and their point of application are shown in Figure 1 and Table 1. These loads are separated by a distance of greater than 2.5 & = 70* (2.2.1.a) and therefore, can be treated as separate forces. Each of F1. F2, and F3 are addressed below.

4.2 The force F1 (e 92,900 lbs) is applied to the reactor pressure vessel (RPV) shell 72 inches l above the shroud support plate. It is a local force applied in the radial direction by the shroud repair during a maximum credib.le earthquake (MCE). At this elevation the RPV shell is 6.125 inches thick minimum (2.2.2.c(1)).

4.2.1 Compute stresses induced in RPV due to F, = 92.9 kips applied at approximately 72 inches above the support plate during MCE:

Use theory of plate and shells by S. Timoshenko (2.2.2.e, pg. 471) ,

. R, = 125.5" Inside R of RPV h = 6.125" Thickness of RPV exclusive of cladding a = 125.5 + 6.125/2 = 128.563" mean radius ,

e 31/4 2 ,

g ,

3(1 -v )

ah11 r s i

v = 0.29 Poisson's ratio (2.2.2.c(1))

r 2 TI/4 -

=

R1-039 ) = 0.046 (128.5631 x 6.1252 ,

M = P/45 and P = F,/2f where 1is contact width oflower contact plate,5 in.

0 M" = (2)(5)(4S)

= 50.S in - k lin JUN 21 '95 17: 49 428 925 4533 PAGE.005

~~ El '95 '02349PM NIL SVC/ PROS DEPT 408 925 4533 P.e .5 l ggh 25A5607 SH NO. 5 Wman y F, (

From paragraph 2.2.2.e page 474 deflection under load is 2

W- PF 0 9.29 x 128.563 x 0.046 = 0.02"; since E = 28.7 x 10' ksi (2.2.2.c(1) ,

2Eh 3 2 x 28.7 x 10 x 6325 o, = 6Mm/h' = S.0S ksi o, = EW,/a + 6 v Mu/h' = 6.81 ksi e, = 8.08 ksi e, = 6.81 ksi 4.2.2 The maximum value of P, stress intensity due to this load is negligible and the maximum ,

value of Pb stress intensity due to this load is 8.08 ksi. These stress intensities occur directly under the point ofload application.

4.2.3 The existing primary membrane stress intensities in the shell per the original Stress Report (Paragraph 2.2.2.c(1), Page B-9-20) are 26.8 ksi (PI) and 28.2 ksi (P,+ Pb).

4.2.4 The new value of P, is same as originalvalue of 26.8 ksi. The new value of P1 + Pb can be conservatively calculated as 28.2 + 8.08 = 36.28 ksi.

4.2.5 The allowable value of primary membrane Pm stress intensity is Sm,which equals 26.7 ksi and the allowable value of primary local (P,) plus primary bending (P, + Pb) stress intensity is .

3 S, = 80.0 ksi for faulted and 2.25 S = 60 ksi for emergency conditions. "Iherefore, both faulted and einergency conditions allowables are met.

4.2.6 Primary stress intensity (Pb) for normal / upset condition F, = 33.4 kips = 33.4/92.9 x 8.08 = 2.91 ksi.'and priman local stress intensity (P,) is negligible.

4.2.6.1 The existing P, = 26.8 ksi, and (P, + Pb) = 28.2 ksi. The new P, = 26.8 ksi while new (P,+ Pb) = 28.2 + 2.91 = 31.11 ksi.< 40 ksi (1.5Sm). And since the increase in the stress -

intensity is ven small ( 2.91 ksi compared to the allevable of SSm - 80 ksi for primary plus JUN 21 '95 17:50 408 925 4533 PAGE.006

. Jt.N 2a .2 d2:49PM NJCL. SVC/PROJ MT 408 925 4533 p7_,- >

gg 607 SH NO. 6 secondan stress intensity ), the power rerate analysis ( 2.2.2.h ) & fatigue evaluadon for revised

. _ .. . .. thermal cycles ( 2.2.2.1 ) is unaffected.

4.3 The force F, is applied to the reactor pressure vessel (RPV) shell 244 inches above the shroud support plate. It is a local force applied in the radial direction by the shroud repair during a MCE. At this elevation the RPV shell is 6.125 inches thick minimum (Paragraph 2.2.2.c(1).

4.3.1 Stresses in RPV due to F, = 31.39 kips applied at approximately 244 inches above the shroud support plate during MCE, can be obtained by scaling from values obtained for F, = 92.9 kips with 't = 4 inches contact width.

e, - 3.42 ksi e, - 2.88 ksi 4.3.2 The maximum value of P, Stress intensity due to this load is negligible and the maximum value of Pb is 3.42 ksi. These stress intensities occur directly under the point of load application.

4.3.3 The existing primag membrane stress intensities in the shell per the original Stress ,

Report (Page B 9 20 of 2.2.2.c(1)) is 26.8 ksi, (P,) and 28.2 ksi (P,+ P,).

4.3.4 The new value of Plis conservatively same as existing value of 26.8 ksi. The new value of

. P! + Pb can be conservatively calculated as 3.42 + 28.2 = 31.64 ksi.

4.3.5 The faulted allowable value of priman membrane stress intensity is Sm, which equals 26.7 kai and the allowable value of primary local (P ) and the primay plus bending (P,+ P,)

stress intensity is 3 Sm = 80.0 ksi for faulted and 2.25 S. - 60 ksi for emergency condition.

Therefore, both faulted and emergency conditions allowables are met.

4.3.6 Since the faulted stress intensities (P,) and (Pi + P,) are below upset condition allowable of 40 ksi., the primary stress intensity for normal / upset condition F, = 17,100 lbs is satisfled by inspection as the F, is lower than F, of MCE condition. And since the increase in the stress intensity is very small ( (17.1/31.39) 3.42 =1.87 ksi compared to the allowable of SSm = 80 ksi for priman plus secondaty stress intensity ), the power rerate analysis ( 2.2.2.h ) & fatigue evaluation for revised thermal cycles ( 2.2.2.i) is unaffected.

4.4 The force F, is applied to vertical plate at 4.25 (2.2.2.g) inches from the inside surface of '

the RPV shell (This results in moment arm of 4.25 + 6.125/2 = 7.3 at RPV shell center line).

Since complet'e load & moment is transfered at shell center line, the analysis in 4.4.1 through 4.4.5 is valid for both weld H8 cracked k uncracked cases. The value of F3 is 372,650 pounds for maximum MCE, and 248,500 pounds for emergency and 172,910 pounds for DBE plus normal pressure & 249,500 pounds for Upset thermal condition. The effects of F, on shell are JUN 21 '95 17:50 400 925 4533 PAGE.007

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addressed in 4.4.1 thru 4.4.5 and on baffle plate junction with shell in Section 4.4.6 (H8

' uncracked) and Section 4.4.7 (H8 cracked).

4.4.1 Apply F, in any condition as vertical load and it will transfer as axial load V = F, lbs and moment of 7.31 F, k in. This load V-F, kips. and moment 7.31 F, k-in. will be assumed to be resisted by the width of RPV shell equal to the width (b = 13.5"), of the horizontal plate of stabilizer lower support (paragraph 2.2.2.g),.

4.4.2 Using analysis methods for edge loads for m (para. I233 of 2.2.1.a) and direct '

membrane stress as P/t, the stresses in shell are as follows:

e, = 6 mo /12 , y f ,;

e, =

Rm + 6v "2 t where m, = End moment = 7.31 F/13.5 k in/in; t - Thickness of shell = 6.125"; -

P = F/13.5 kips /in; -

E = Young's Modulus = 28.7 x 10' ksi; Ps, = Vessel Mean Radius = 128.5625 in.;

v = Poisson's ratio = 0.29; W, = Deflection at edge (calculated below).

Using para. I 232(2) of 2.2.1.2, the limiting value of W, = m/2p'D, where D = E t'/12 (1-v'), p = j 4 3b" ) and substituting values of D, p in terms of E, t, Ps,, the expression for e, can be 2

) Rkt

1. 8 v+ "

simpilfted as e, = . And with v = 0.29 e, = (0'84)'

2 l

Further, since t = 6.125", the final ei- 0.094 F, ksi, e, = 0.073 F, ksi.

I JUN 21 '95 17:50 408 925 4533 PAGE.008 l

l

AN 21 's s ov a toct m g w a s v m m ax y; -

,./ %

y) gg'gg 607 SH NO.T These e,,a, stresses will be tued to calculate the stress intensity by principle stress difference formulas. Since shear stress !s zero, the principle strenes are e, = o, ; e,- o,. Primary stress -

intensity is maximum of e i, e, or e,- e,.

4.4.5 Primary local membrane plus bending (P + P,) stress intensity for faulted conditinne F, = 372.65 kips are a follows:

4A.S.1 e, = u,- 0.094 x 372.65 = 35.03 ksi o, = o,= 0.073 x 372.65 = 27.2 ksi Thus the maximum primary stress intensity (P,+ Pb) = 35.05 ksi 4.4.3.2 From page B-9-20 of original stress report (2.2.2.c(1)) the existing maximum primary local membmne stress intensity is 26.8 ksi and (Pt + P,) is 28.2 kai. And as the major stresses due to F, are P., i.e: while P, = 0.007 F, = 2.61 ksi, P, is 0.087 F, = S2.42 kai out of a total P! + P, of 35.03 kai. And conservatively the new values will be P, = 26.8 + 2.61 = 29.41 ksi and P,+ Pb = 28.2 + 35.03 - 63.25 ksi However, the maximum P, , P,+ P, values from page B 920 of the existing stress report at location of F, (i.e. elem. #20 of Seal Shell model on page B 21 of 2.2.2.c(1)), are 17.86 ksi and 19.81 ksi. Thus these values could be used if required.

The allowable P, and P + P, stress intensity is 1.55, = 40 ksi in the original stress report.

However, this is a faulted event and per 2.2.1.a the allowable for faulted conditions is SS, =

80 ksi.

4 A.4 Primary stress intensity (Pe + P,) and (P,) for emergency conditions F, = 248,500 lbs are as follows:

4.4.4.1 The primary stress intensity value from 4.4.5.2 for F, - 372,650 lbs can be used to get the a, as follows:

o, =- 248500 x 35.03 and c, = 248500 x2.61 372650 372650

= 23.36 ksi (Pa + P,) = 1.74 ksi (P,)

4.4 A.2 Using same existing maximum primary stress intensity of 4A.S.2 (for faulted condition) of P, = 26.8 ksi and P, + P,=28.2, ksi., the new values are: _

JUN 21 '95 17:50 408 925 4533 PAGE.009 l

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P, + P, = 28.2 + 23.36 and Pi = 26.8 + 1.74 <

=  !

51.56 ksi < 60 ksi (2.1.1.a) = 28.54 ksi < 2.25 S. - 60 ksi. (2.1.1.a) 4.4.5 Normal / upset conditions evaluadons required for priman, primary plus secondary, and peak stress intensities per 2.2.1.a are shown in this section.

4.4.5.1 Primary stren intensity evaluation is required for F, = 172,910 pounds which will give 1 2910 (P, + P, ) value of x 35.03 = 16.25 ksi . and P, = x 2.61 = 1.21 krf -

372650 372650 ,

s 4.4.5.2 The existing primary stress intensity at this location for operating condidon is P1 =  ;

17.86 ksi. and P, + P = 19.81 ksi. (page B 91 of 2.2.2.c(1)). Thus the new value of Pr + P, at this location is '

Pr + P, = 19.81 + 16.25 and P, = 17.86 + 1.21 j

= 36.06 ksi < 1.5 S. = 40 ksi = 19.07 ksi < 1.05, = 26.7 kai 4.4.5.3 The primary plus secondary stress intensity for upset condition load F, is required for two (2) sets ofloading cycles as follows (at RPV shell)-

F, = 172,910 lbs for 482 cycles normal k excessive heatrup / cooldown transienu and F, = 249,985 lbs for 30 cycles ofloss of feedwater pump & HPCI/RCIC injection transients  !

172 910

• ~

4.4.5.4 Primary plus secondary stress range for 482 cycles is S. = 372,650 The existing value of same primary plus secondary stress intensity range is 35.0 ksi (page C9 21 of 2.2.2.c(1)). Thus the new value of S, - 35.0 + 16.25 = 51.25 ksi < 3 S, = 80 ksi. ,

4.4.5.5 Primary plus secondary stren intensity range for 30 cycles is S, = 249* 985

• 372,650 ksi. The existing value of the same primary plus secondary stren intensity range is 35.0 ksi (page G9-21 of 2.2.2.c(1)). Thus the new value of S, = 35.0 + 23.5 = 58.5 ksi < 3 S, = 80 ksi..

4.4.5.6 Fatigue, i.e., peak stress intensity range, evaluation for all 512 ( = 482 + 30 ) cycles per 2.2.2.1 is calculated ,using peak stress intensity range of 30 cycles F, since it is highest, is as follows: -

S, = K, Smce S, < 3 S , K, = 1.0 for all 512 cycles.

JUN 21,'95 17:51 408 925 4533 PAGE.010 e + n _ _ - -. -- - ,

^

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)' gg 2M5607 SH NO.10 And there is no stress concentration factor per section C 8 of 2.2.2.c(1)

\

I S, = 27.0 (existing S , page C 101 of 2.2.2.c (1)) + 23.5/2 = 38.75 ksi. -

N,, - 8400 (Figure N 415(A) of 2.2.1.a) t Usage Factor = UF = 512/8400 = 0.06 < < 1.0.

Thus the power rerate analysis ( 2.2.2.h ) and fatigue evaluadon for the revised thermal cycles

( 2.2.2.1 ) is unaffected.

4.4.6 Evaluation of RPV Shell and B2Mie Plate function ( Weld HB Uncracked i 4.4.6.1 For Faulted Cendition F, Due to the support afforded by jet pump nozzles to the baffle plate, the load F, will be essendally distributed over a rectangular plate between RPV shell and jet pump nozzle hole circle with the width equal to the width of the lower support plate as shown below:.

< a >

,,,,,,,,,,,,yRPV She\\

n load "q" b,simpiy Supported '

n r

Simply Supported q y M

Simply Supported [ Jet Pump Nozzlei (Hole Circle s where l

a= width of horizontallower support plate (2.2.2.g) = 13.5";

b= Distance (radial) between shellinside radius (-125.5") andjet pump nozzle hole circle radius = 226.5/2 - 113.25" (per 2.2.2.k) = 12.25";

q= Distributed load = F3/( 13.5 x 12.5 ) = 372.65/( 12.25 x 13.5 ) = 2.253 ksi; Using formulu for middle of fixed edge moments (for uniformally loaded plate with one edge fixed, other three edges simple supported) from Timoshenko (2.2.2.e, page 241), the moment ,

JUN 21 '95 17:51 400 925 4533 PAGE.0!!

I l

^J g 21"'h 3 :50PM NJCl. SVC/ pro [ DEPT 408 925 4533 P.. 5 f gg 607 SH NO 11 My = d,ql' (symbols per 2.2.2.3). Further, since b/a = 0.907 d, = 0.0914 from Table 52 of 2.2.2.e. And since 1 = 12.25" (smaller of a = 13.5" or b = 12.25"),

My = (0.0914 ) ( 2.253) (12.25)'

= 30.9 in. kips /in.

Further the bending stress e,= 6 My/t' and with t = 2.0625" (thickness of baffle plate), the bending stress value is e, = 30.9

• 6/(2.J625)'

= 43.58 ksi And shear stress t = F,/ Area = F,/ Perimeter * 't' t = 372.65/2 (12.25 + 13.5)(2.0625)

= 3.51 ksi Principal stress e, = 43.58/2 + ( (43.58/2)' + (3.51)' ] = 43.86 ksi e, = -0.28 ksi The maximum stress intensity - e, e, - 43.86 - (-0.28) = 44.14 ksi .

The maximum primary membrane plus bending.stren intensity at this location from the existing stress analysis (paragraph 2.2.2.c(2), page B-16 6) is 16.7 ksi. Therefore, the new ,

maximum primary membrane plus bending (P, + P,) stress intensity is = 16.7 + 44.14 - 61.84 kai i I

which is less than faulted allowable of 3 Sm = 80 kai (pages B 17-1, B 17-5 of 2.2.2.c(2)).

However, since this junction is also of inconel material and therefore, the 3 Sm of inconel material = 69.9 ksi, being lower than 3 Sm= 80 ksi of carbon steel, the vessel material, will be  !

used as allowable.  !

1 Since allowable for ( P1 ) is same as for ( P! + Pb ) and calculated value of P1 is always lower j than the calculated value of ( P! + Pb ) in all conditions, P1 evaluation is not documented  ;

separately for any cues.

4.4.6.2 Primary local membra'ne plus bending stress intensity (P1 + Pb) for the Emergency condition F3 = 248.5 kips is calculated by ratioing down the faulted condition value (Pl + Pb)

• for loading of 372.65 kips. Thus the new value is ( P1 + Pb ) = 46.14 ksi which is less than the alowable of 2.25 Sm = 52 ksi for this condition (paragraph 2.1.1.a).

4.4.6.3 Primary stress intensity evalution for upset conditions is required for F3 = 172,910 .

pounds which will give ( P1 +Pb ) value of 172,910/372,650 x 44.14 - 20.48 ksi.

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l JUN 21 '95 17:51 408 925 4533 PAGE.012 l

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6BMPM nit. SVC/PRoJ DEPr 4W 925 4533 25 L) GENuclearEingy

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The existing priman stress intensity for operating conditions is 11.5 kai (page B-16-8 of 2.2.c(2)). Thus the new value of Pl+Pb at this location is Pl+Pb = 11.5 + 20.48

= 31.98 ksi < 1.5 Sm = 35 ksi 4.4.6.4 The primary plus secondary stress intensity range for upset condition F3 is required for three (3) sets of loading cycles, (to account modified thermal cycles of 2.2.2.1 Jas follows (at Junction of baffle plate and shell):

F3 = 172,910 lbs for 226 cycles of normal & excessive heatup / cooldown transients and FS = 249,985 lbs for 30 cycles ofloss of feedwater pump & HPCI/RCIC transients and F3 - 172,910 78,200 = 94,710 lbs for 256 cycles of cooldown transient without DBE 4.4.6.5 The primary plus secondary stress intensity range for 226 cycles is Sn =

172,910/372,650 x 44.14 = 20.48 ksi. The existing value of the same primuy plus secondary stress intensity range is 51.6 ksi (page B-1615 of 2.2.2.c(2) range of all cases except cue VI See Table 2 for details). Thus the new value of Sn = 51.6 + 20.48 = 72.08 ksi > 3 Sm = 70 ksi. Thus simplified elastic plastic analysis will be required.

4.4.6.6 The primary plus seconday stress intensity range for 30 cycles is Sn - ( 249,985 /

372,650 ) x 44.14 = 29.61 ksi. The existing value of the same primary plus secondary stress intensity range is 72.0 ksi (per B 71 of 2.2.2.c(2)only cases III and VI are part of this transient, Table 4 for details). Thus the new value of Sn= 72,0 + 29.61 = 101.61 ksi. which is greater than SSm = 70 ksi. Thus simplified elastic-plastic analysis will be required.

4.4.6.7 The primary plus secondary stress intensity range for 256 cycles is Sn = 94,710/372,650 x 44.14 = 11.22 ksi. The existing value of the same primary plus secondary stress intensity range is 31.3 ksi (page B 1615 of 2.2.2.c(2) maximum of case IV either at bottom or at top).

Thus the new value of Sn = 31.3 + 11.22 = 42.52 ksi < 3 Sm = 70 ksi. Thus no simplified clastic-plastic analysis will be required.

4.4.6.8 The peak stress intensity existing range, for 226 cycles is 66.3 ksi (Table 3 stress '

concentration factor, pg. B-17-2. for details). Thus the new Sp = 66.3 + (20.48)(1.64) = 99.89 '

ksi where 1.64 is the bending stress concentration factor per page B 17-2 of 2.2.2.c(2).

The fatigue evaluation for 226 cycles F3 is u follows: ( Since Sn = 72.08 ksi> 3 Sm = 70 ksi and material parameters of m = 2, n- 0.2 (per pg. B-17 5 of 2.2.2.c(2) Sm < Sn < m 3 S.) -

JUN 21 '95 17:52 408 925 4533 PAGE.013

Jtn 21 % a2:51PM nit. S W PROJ DEPT 408 925 4533 g j

~

il CEh% $67 5" " 2 8' Ke - 1.0 + [(1-n)/(m-1)n] [S,/3 S.- 1] (per 2.2.1.a)

= 1.0 + ( 0.8/(1x0.2) ] ((72'.08/70) - 1] = 1.12 Sa = Ke . Sp/2. Since Sn < 3 Sm Sa = (99.89 ) ( 1.12 )/2 - 55.94 ksi  ;

Natl = 3,000 (Figure N-416(A) of 2.2.1.a)

Usage Factor = UF1 = 226/3,000 = 0.075 4.4.6.9 The peak stress intensity evaluation for 30 cycles F3 is as fo!!aws:(Since Sn = 101.61 ksi

> S Sm = '70 kai and material parameters of m = 2, n= 0.2 (per pg. B-17-5 of 2.2.2.c(2) Sm < Sn <

m 5 S.,). Thus Ke = 1.0 + ((1-n)/(m-1)n]-(S,/5 S,- 1] (per 2.2.1.a)

= 1.0 + [ 0.8/(1x0.2) ] [(101.61/70) - 1) = 2.81 The fatigue evalution for these 30 cycles is as follows: The existing peak streu intensity per l page B 17-3 of 2.2.2.c(2) is 146.8 kai (Table 5 for details).

The extra peak stress intensity for this F3 load is (29.61) (Bending streu concentration factor of 1.64 per page B 17 2 of 2.2.2.c(2)) = 48.56 ksi. The new (total) peak stren intensity Sp =

146.8 + 48.56 = 195.56 ksi. And with Ke = 2.81, the alternating stren Sa = 2.81/2 x 195.36 -

274.48 ksi. The allowable cycles at this Sa level per Figure N 415(A) of 2.2.1.a are 45. Thus usage factor UF2 = 30/45 = 0.67.

4.4.6.10 The peak stren intensity evaluation for 256 cycles F3 is as follows:(Since Sn = 42.52 ksi

< 5 Sm = 70 ksi, Ke= 1.0 per 2.2.1.2)

The fatigue evalution for these 256 cycles is as follows: The existing peak stress intensity per page B 17-3 of 2.2.2.c(2) , case IV, is 51.3 kai.

The extra peak stress intensity for this FS load is (11.22) (Bending stress concentration factor of 1.64 per page B-17 2 of 2.2.2.c(2)) = 18.4 ksi. The new (total) peak stress intensity Sp = 31.3

+ 18.4 = 49.7 ksi. And with Ke = 1.0, the alternating stress Sa = 49.7/2 = 24.85 ksi. The allowable cycles at this Sa level per Figure N-415(A) of 2.2.1.a are 40,000. Thus usage factor UF3 = 256/40,000 = 0.006.

4'4.6.11 The cumulative usage factor (revised) is as follows:

UF = UF1 + UF2 + UF3

~

The cumulative UF = 0.075 + 0.67 + 0.006 = 0.751 < 1.0 belowlimits of 2.2.1.a JUN 21 '95 17:52 408 925 4533 PAGE.014

J 21 ' n 02:51PM NJCL, SVC/PRoJ DEPT 408 925 4533 . pg5 6*' 14 GENucieerEneryy y4 '" "

Thus the power rerate analysis ( 2.2.2.h ) and fatigue evalution for revised thermal cycles

- ( 2.2.2.i.) is unaffected.

4.4.7 Evaluation of RPV Shell and Baffle PlateJunction( Weld H8 Cracked )

4.4.7.1 FEM for Weld H8 Cracked Condition A one quarter finite element model of the baffle plate using computer program ANSYS was analyzed for this condition. Tlie edge of plate at RPVjunction was fixed while the junction at shroud was free representing severed H8 weld condition. At lines of symmetry 1. e. , O degree and 90 degree, proper boundary conditions were used. First unit load case of 449.3 kips vertical load applied as pressure on very small areas ( representative cue of tie rod load ) was analyzed. Second unit load case of uniform pressure of 1,000 psi was also analyzed representing AP condition. The results of these unit loading cases are summarized below:

Unit AP of1,000 psi Pl - 0.5 ksi ( 0 0005 ksi for AP = 1 psi ) and P1 + Pb = 442 ksi (0.442 ksi for AP = 1 psi)

P1 + Pb + Q = 442 ksi ( = 0.442 ksi for oP = 1 psi )

Unit tie rod load of 449.3 kips Pl = 12.6 ksi ( 0.028 k:i for 1 kip load ) and P1 + Pb = 47.5 ksi ( 0.106 ksi for 1 kip load )

Pl + Pb + Q = 106.8 ksi ( 0.238 ksi for 1 kip load )

The oP values for M. S. LOCA and Nornal conditions are = 51 psi Sc 33.03 psi per 2.1.1.b while, the oP value for Upset thermal condition is = 35.68 psi per 2.1.1.b.

4.4.7.2 Evaluation for Faulted Condition Primary local membrane ( Pl ) stress intensity for faulted conditions F3 = 372.65 kips is = 0.028 x 372.65 + 0.0005 x 51 = 10.46 ksi which is less than faulted allowable of SSm = 70 ksi. for Inconel ( baffle plate material ) as well as 35m = 80 ksi, for carbon steel ( RPV shell materia! ).

Since allowable for ( Pl ) is same as for ( P1 + Pb ) and calculated value of P1 is always lower than the calculated value of ( P1 + Pb ) in all conditions, P1 evaluation is not documented for remaining cases.

Primary local membrane plus bending (Pt + Pb) stress intensity for faulted conditions F3 =

372.65 kips is = 0.106 x 372.65 + 0.442 x 51 = 62.04 ksi which is less than faulted allowable of 35m = 70 ksi. for inconel ( baffle plate material ) as well as SSm = 80 ksi. for carbon steel ( RPV shell material ).

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JUN 21 '95 17:52 408 S25 4533 PAGE.015 l l

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'Ju(2'1?'si lsais2Priintisvc/ Prof DEPT 308 925 2533 - ~~M^~~~

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4.4.7.3 Evaluation for Emergency Condition .

i

- Primary local membrane plus bending stress intensity (P1 + Pb) for the Emergency condition ,  !

load of 248.5 kips is = 0.106 x 248.5 + 0.442 x 51 = 48.88 ksi which is less than 2.25Sm = 52.5 kal.  ;

for inconel ( baffle plate material ).u we!! as 2.25Sm = 60 ksi. for carbon steel ( RPV shell material ). [

4.4.7.4 Evaluation for Normal / Upset Conditions j 4.4.7.4.1 Evaluation for Primary Stren Intensity ,

Primary local membrane plus bending stress intensity (P1'+ Pb) for the Normal / Upset  !

condition load of 172.91 kips is = 0.106 x 172.91 + 0.442 x 35.03 - 32.93 ksi which is less than 1 1.5Sm = $5 ksi, for inconel ( baHle plate material ).as well as 1.55m = 40 ksi. for carbon steel l

(RPV shell material ).

4.4.7.4.2 Evaluation for Primary plus Secondary Stress Intensity The primary plus secondary stress intensity range Sn for upset condition loads of 172.91 kips ,  !

94.71 kips and 249.985 kips is calculated for all loading cycles per 2.2.2.i atjunction of baffle j i

plate and RPV shell. ,

The highest primary plus secondary stress intensity range for 226 cycles of 172.91 kips is Sn = ,

0.238 x 172.91 + 0.442 x 33.03 = 55.75 ksi, while Sn for 30 cycles of 249.985 kips is = 0.238 x l 249.985 + 0.442 x 35.68 = 75.27ksi, and for remaining 256 cycles is = 0.238 x 94.71 + 0.442 x

$5.03 =57.14 kai. All of these Sn values are less than < SSm = 70 ksi ( conservatively, for baffle plate material ), except for 30 cycles of 249.985 kips load. Thus a simplified elastic-plastic analysis will be required for the 30 cycle loading case.

4.4.7.4.3 Evaluation for Fatigue Since the Sn values in the H8 cracked case are lower than the Sn values for the corresponding i cases of 226 cycles,30 cfcles , #c 256 cycles under H8 uncracked case,- the fatigue, and peak j stress intensity range, for all cycles for H8 cracked is enveloped by the similar calculations for H8 uncracked case. Thus the maximum cummulative fatigue usage factor for H8 cracked condition, is conservatively, also equal to 0.751 < 1.0.

4.5 Evaluation for Peach Bottom Unit 2 for F,, F,, F, and their effects on all code requirements is satisfied as documented in sections 4.1 through 4.4. The original stress report for Peach Bottom Unit 3 (2.2.2.d) states that stress reports 20 (2.2.2.d(1)) and stress report 11 (2.2.2.d(2)) for Unit 3 are exact duplicates of same reports for Unit 2. Hence F,, F,, F, assessments for Unit 3 is same as shown above for Unit 2 and thus meets all the code (2.2.1.a) requirements. It should be noted that seismic shears and overturning moments for shroud 1 support used in the analysis (page B 11-A of 2.2.2.c(2)) are higher than those required by G.E. -

Drawing Design (2.2.2.b).

JUN 21 '95 17:52 400 925 4533 PAGE.016

{

n - -

c > . ggg Q gg 607 sH No.16 4.6 In accordance with power rerate analysis / reconciliation documentation (2.2.2.h) and fatigue evaluation of Peach Bottom II and III Reactor Vessels for revised thermal cycles (2.2.2.i), there are no changes required to the original stress analysi: (2.2.2.c and 2.2.2.d) in the regions affected by loads F , F , and F,. Thus power rerate :.nalysis and fatigue evaluation for revised thermal cycles is still valid.

4.7 All of the stress intensities due to the new design mechanicalloads F,, F,, and F, satisfy the allowable stress intensities of the original Code of Construction (Paragraph 2.2.1.a).

4.8 The new seismic loads on the stabilizer bracket location (F ) is 334,000 lbs for MCE. These when conservatively converted into individual bracket loads (i.e., taken two bracket only in any one directional earthquake) result in individual bracket loads of 167 kips (for MCE). This load is less than the stabilizer bracket seismic loadings of 300 kips (conservatively DBE only) per document 2.2.2.b, Table 9. Thus the effect of F. (u a result of shroud stabilizer modificadon) on RPV is enveloped by the existing stress analysis including fatigue evaluation (2.2.2.c(3)) since page 12 of 2.2.2.c(3) states that it is using seismic loads from 2.2.2.b. Power rerate analysis (2.2.2.h) and fatigue evaluation for redsed thermal cycles ( 2.2.2.i) are not affected since the existing analysis of stabilizer brackets is unchanged and these brackets were not critical components reevaluated in either document 2.2.2.h or 2.2.2.1.

4.9 The new seismic forces and moments on the base of RPV skirt are F5 and M5. The max. of these values (MCE values) are F5 = 284.2 kips and M5 - 6611 kip-ft. These values are less than the seismic values of H, = 1083 kips and M, = 40,138 kip.ft (2.2.2.b) which are used in the original skirt stress analysis reports (page B 15 4 of 2.2.2.c(4) and 2.2.2.d(4)) for Peach Bottom

. Units 2 and 3 respectively. Thus original stress analysis of RPV skirt is still vahd. These loads are only primary loads and do not affect fatigue evaluation. The power rerate analysis documentation (2.2.2.h) and fatigue evaluation documentation for revised thermal cycles (2.2.2.i) reevaluates support skirt but since seismic loading used in power rerate is the same as the original loading (2.2.2.b), the power rerate and fatigue evaluation for revised thermal cycles documentadon is unaffected by these stabilizer modification forces and moments F5 and M5.

4.10 The new seismic overturning moments and shears at the shroud support location are M.

- 129.6 k-ft, F = 143.1 kips in MCE. These loads are lower than the seismic moment and shear of 4885 k-ft and 238 kips (conservatively DBE only) per Table 10 of 2.2.2.b. Thus effects of MS/V6 (new) on RPV is enveloped by the existing analysis since page B,.11 A of 2.2.2.c(2) states that it is using seismic loads higher than those given in GE drawing (2.2.2.b). The j shroud support leg buckling evaluation documented in the original stress report ( pages B 18 2 l through 3-18-9 of 2.2.2.c(2)) is still valid because the seismic overturning moments from the

revised analysis are lower than those used in the buckling evaluadons. Power rerate analysis i (2.2.2.h) and fatigue evaluation for revised thermal. cycles ( 2.2.2.1) are not affected since the exisdng analysis of shroud support legs is unchanged and these legs were not cridcal components reevaluated in either document 2.2.2.h or 2.2.2.1.

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JUN 21 '95 17:53 400 S25 4533 PAGE.017 i

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,eD 21 '90 . 52PM NUCL SVC/PRoJ DrPT das 925 c m p,,g, ,

i V s m sov sa so.iv GEMucinarEnergy 5.0 Based on the best of my knowledge and belief, it is hereby certified that the analpis documented in this Stress Report satisfies the requirements of ASME Boiler and Pressure Vessel Code Section III,1965 Edition with Addenda through Winter 1965 and Design -

Specification listed in Paragraph 2.1.1.a. This certification is provided as required by Paragraph N 142 of said Section III.

Signature:

/'

de/ Date: Ikb5~

I License Number: C2356? State: ciiirom;.

$y) ires on /W31//97) e . ..

, ofESSlo b@ %p C 23582 CML

1. rgy OFCAOYp+

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JUN 21 '95 17:53 400 925 4533 PAGE.018 l l

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e Jl.fi 21 02:33PM MJCL SVC/PROJ DCPT 408 925 4530 p, ig,45

/ %

25A5607 SH No. I8 g

Inh!A.1. ADDITIONA1. DESIGN MECHANICAL 1.OADS (Normal /tJpaco (Taulted)

Torce Dag + Normal Fressure Emergency MCE+LOCA temarks F, 3s,400 lbs 91.900 lbs 92.900 lbs Primary Stress F, 17,t10lbs $1,390 lbs 31.390 lbs Primary Stress F, 172.910 lbs 248.500 lbs 372.550 lbs Primary Stress 172.910 lbs(w/DBE) . . Fri plus Sec. S.I Range (226 cycles) and fatigue 94.710 lbs(w/o DBE) . . Pri plus Sec. 3.1 Range (156 cycles) ant; fatigue ,

249.9:5 lbs . . Fri plus Sec. S.! Range (30 cycles) and fatigue F, . . 334,000 lbs Seismic (Total) Tangential Lead 8 stabillaer Brache F, . . 2s4,200lbs Scismic Shear Only ORFV skirt

. 6.611,000 fWbs Selsmic Moment GRFV skirt M, .

M, . . 129.600 fWbs Scismic Moment S$hroud Support F, . . l 143.100 lbs Seismic Shear 0 Shroud Support

1) NOTES

2) F,, F,, and F, are discrete loads applied over a small area. At any one point in time, F, , F,, j F, are each applied to one location. At any one point in dme, F, is applied to 4 locations 90' apart for the installation of four shroud stabilizer assemblies. DBE is a Design Basis Earthquake (OBE). MCE is a Maximum Credible Earthquake (S.c2). The load F, shown is the maximum and applies to one tie rod 180 degrees apart from F, le F,, while remaining tie rods have loads lower than F, values shown above.

3) The stress intensides shall meet the stress allowables of the ASME Code,Section III, for the load combinations defined by the Peach Bottom UFSAR. The original Code of Construction did not include Faulted load combinations. Faulted and Emergency load combinations shall meet the stress allowables as defined by the Peach Bottom UFSAR for the reactor pressure vessel.

4) Loads F,, F,, F, to be used in the evaluations are from document 2.1.1.a. Loads F., F,, M,, M, and F, are those for cracked condidon as documented in GENE.77160-0994, Rev. 2.

JUN 21 '95 17:53 408 925 4533 PAGE.019

• JW 25 't asytt tto. 5%rWOJ DEPT 408 925 4533 607 SH No.19

'>- gh Table 2 Maximum primary plus secondary stress intensities (P + Q), ksi, for 226 cycles case is range of '

all cases on page B-1&l5 excluding case VI since it is L.O.F.W.P transistion case.

IQE BOTTOM 2.2.2.c(2)

.Cass H-L LR RH H-L LR RH Page #

Max. Primary +6.6 +2.2 +2.4 +1.8 +9.5 +0.5 B 1&l5 Case (1)

Min. Primary -1.2 8.8 -1.0 -6.0 -2.1 3.3 B-1&l5 Case (2)

+9.5 +9.4 +26.2 +0 +11.2 B-1&l5 Max./ Min of +21.0 cases III,IV,V 6.9 -31.3 9.2 8.5 -37.4 -6.3 B 1&l5

+28.5 +11.5 +11.8 +28.0 +0.3 + 11.4 Max.

Total +

Total- -8.1 -40.1 -10.2 14.5 59.5 -9.9 Min.

I l Existing +36.6 +51.6 +22.0 +42.5 +48.8 . +21.3 (Max-Min) l Range w/o Case VI 1

+51.6 -

- - - Existing Revised Absolute - -

l l Max. Range Contribution - +20A8 - - - - 172.91/372.65*

From F, 44.14 Revised - - - - Max 4, 72.08 Value l

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JUN 21 '95 17:54 400 925 4533 PAGE.020 l

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. JUN 2195 02:53PPI itCL SVC/PROJ DEPT 408 925"6533 P. 21/ 5-25A5607 SH No. 20 gg - . .

Table 3 .

Maximum primary plus secondary plus peak (P+Q+F) stress intensity in ksi for 226 cycles case  ;

is range of all cases on page B 17 5 except case VI since it is LO.F.W.P. transient case T.02 BOTTOM 2.2.2.c(2)

. Cam HL I.rR R-H Page #

+2.8 +4.1 Since " BOP" envelopes Max. (+),

Case (1) +115

" BOTTOM" only " TOP" is B-17-3 Max. Primary evaluated for peak ,

Case (2) Min. $1.4 -15.6 -1.5 stress intensities to be consistant Max (-),

with original stress report. B-17-3 Primary

+16.6 +9.4 Max (+),

Max./ min of +21.9 cases III,IV,V B 17-3

-31.3 10.1 Min (-), B-

-6.9 17-3 Max. (+) +33.5 +19.4 +13.5 ,

Min. (-) , -8.3 -46.9 -11.6 Range +41.8 +66.3 +25.1 (Max-Min)

- Existing Absolute Max -

+66.3 for 120 cycles 33.59 1.64*20.48 Contribution - -

of F, 99.S9 Existing +

Sp Revised - - -

Contri. of I

F, l

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1 JUN 21 '95 17:54 408 925 4533 PAGE.021

, ,rtn a u edsti ru L. dekw M e xd 'bu P.cva 95A5607 SH No. 21 GENuclearEnergy hy. <

Table 4 Primarv Plus Secondary Streu Intensity Range (P + 0)

Existing stress eport page 67-1 of 2.2.2.c(2) states that Case VI (which is L.0.F.W.P) is broken into two parts for down ramp i.e. VI iter 32 and VI iter 960 while up ramp is same as Case III.

Thus maximum primary plus secondary stress intensities range (ksi) of original analysis for L.0.F.W.P.& HPCI/RCIC transients is as follows.

102 l HQTTOM 2.2.2.c(2) l Chic H-L LR RH H-L LR R-H Page #

9.0 -8.3 0 8.3 B 16 3 Case III (up) -6.9 2.1 22.1 11.4 11.2 B-16-3 Case VI - 69.9 -

(max) 50.9 - -22.2 -33.5 S16-3 (min.) -19.0 -

72.0 -59.9 30.4 22.2 41.8 -

Range (Case -12.1 VI-Case III)

+72.0 - - - Existing Absolute Max -

Revised Range

. - - (249985/372 Contribution 29.61 -

650) (e )

of F,

- - - Max S, Revised Value - 101.61 -

s 408 925 4533 Fauc.exe JUN 21 '95 17:54

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,JLN 21 '95 02:53PM NJCL SVC/PRoJ DEPT dab 925 4533  !

25A5607 SH NO. 22

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Table 5 Peak Stress Intensirv (P + Q+ F)

This revised maximum primary plus secondary stress intensity for 30 cycles of this transient is 94.46 ksi which is greater than SSm = 80 ksi. Thus simplified elastic-plastic approach will be used for fatigue evaluation. Based on similar statement used for (P + Q) stress intensities, revised (P + Q+ F) stress intensities (ksi) art as follows:

IQE Doe. 2.2.2.c(2) Remark

.Catt HL LR R-H Page #

Case III (up) 6.9 -2.1 9.0 B-17 3 . Original Report evaluated only Case IV Max. - +144.7 +12.2 B 17-3 " TOP" since it enveloped Case IV Min. 68.5 0 -76.2 B-17 5 . BOTTOM" as shown on P + Q evaluation. Thisis also uue for new values as seen on previous page.

Range 61.6 +146.8 -85.2 Existing Revised Absolute - 146.8 -

Max. Range 48.56 -

• Using stress Contribution -

concentration factor ofF3 (1.64) (29.61) for bending = 1.64 page B-17-2 of 2.2.2.c(2)

Revised Sp - 195.36 -

v JUN 21 '95 17:54 400 925 4533 PAGE.023'

t r P.24/25' fr(zi *ss e2:54Pn Nuct svc/ PRO.T DEPT 408 s25 45E su so.23

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Bracket 4 RPV Shell F* _

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+ 4-425irl FIGURE 2 Vessel Skirt Fs 4  % N #

UK HH All other dimensions per 2.2.2.a FIGURE 1. APPLICATION OF DESIGN MECHANICAL LOADS JUN 21 '95 17:55 408 925 4533 PAGE.024

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'(_,) gg 25A5607 REV.4 SH NO. 24 FINAL m SECTION !!!

NON-CODE INTERNALLY " Cl. ASS A SUPPORTED STRUCTURE'

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h ANALYZED IN 4.3 t's RPV A /

SHROUD SUPPORT CYLINDER k ANALYZED IN lx 4.2 2 A

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I JUN 21 '95 17:55 400 925 4533 PAGE.025

[I- Stition Support Dh nt

, ., a 10 CFR 50.55a(t)(3) v PECO ENERGY. =?,4,gr r,!,.,

965 Chesterb,oon Boulevara Wayne, PA 19087 4691 August 17,1995 i

Docket Nos. 50-277 50 278 LIconee Nos. DPR 44 DPR 56 U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555

Subject:

Peach Bottom Atomic Power Station, Units 2 and 3 Response to Request for Additional Information Concoming the Proposed Altemative Repair Plan in Accordance with 10 CFR 50.55a(a)(3)

References:

1. Letter from G. A. Hunger, Jr. (PECO Energy Company) to U. S.

Nuclear Regulatory Commission (USNRC), dated September 16, j 1994

2. Letter from G. A. Hunger, Jr. (PECO Energy Company) to USNRC, dated September 26,1994

3. Letter from G. A. Hunger, Jr. (PECO Energy Company) to USNRC, dated February 14,1995

4. Letter from G. A. Hunger, Jr. (PECO Energy Company) to USNRC, i dated June 22,1995

5. Letter from J. W. Shea (USNRC) to G. A. Hunger, Jr. (PECO Energy Company), dated July 27,1995

Dear Sir:

in our Reference 1 letter, as supplemented by the Reference 2,3, and 4 letters, PECO Energy Company requested approval of a proposed repair plan for the Peach Bottom

- Atomic Power Station (PBAPS), Units 2 and 3 core shroud, in accordance with 10 CFR 50.55a(a)(3), in the event that such a repair is determined to be necessan/. In the Reference 5 letter, the staff requested additional information. Attachment 1 is our response to this request.

.' S

$8DA83hk !O bPDR 77 Enclosure 4

- - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ . _ _ _ _ _ _ _ _ _ . _ _ _J

- . . - . - . . ~ . - _- . - __ .. _. _ .- _ _ - - _ . - - . . _ . _ . _ . . . . _ - . _

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• - August 17,1995 .

q Page 2 Also included in this package are eight revised or new detail PBAPS, Unit 2 and 3 drawings that resect dimendmal changes to the upper support assembly.  :

Attachment 1 contains information proprietary to General Electric. General Electric requests that the Attachment 1 information be withheld from public <#aelaame in accordance with 10 CFR 2.790(a)(4). In accordance with 2.790(b)(1), an affidavit ,

supporting this request is provided in Attachment 1.

If you have any questions, please contact us, i I

very truly yours, M C. Ng G. A. Hunger, Jr.,

Director - Ucensing Attachment cc: T. T. Martin, Administrator, Region 1. USNRC W. L Schmidt, USNRC Senior Resident inspector, PBAPS r

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ATTACHMENT 1 f

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I General Electric Company j

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l AFFIDAVIT i

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. . I I, George B. Stramback, being duly sworn, depose and state as follows:

I (1) I am. Project Manager, Licensing Services, General Electric Company ("GE") and .

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'have been delegated the function of reviewing the information described in ,

paragraph (2) which is sought to be withheld, and have been authorized to apply for  :

. its withholding.

'i (2) The information sought to be withheld is contained in the GE proprietary report l GENE-B13-01732-001, GE Responses to NRC Questionsfor Peach Bottom Shroud l Repair, Revision 0, (GE Proprietary), August 1995, and those drawings listed in the  !

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Attachment which are individual parts of the overall reactor modification identified  :

, as " Reactor Modification & Installation Drawing" 10$E1455. These documents, l taken as a whole, constitutes a proprietary compilation of information, some of it  !

also independently proprietary, prepared by the General Electric Company. The j independently proprietary elements that are drawings are delineated by the GE  !

drawings, being marked as proprietary information and the independently l proprietary elements that are in reports are delineated by bars marked in the margm  !

adjacent to the specific material. -

l (3) In making this application for withholding of proprietary information of which it is  !

the owner, GE relies upon the exemption from disclosure set forth in the Freedom of l Information Act ("FOIA"),5 USC Sec 552(bX4), and the Trade Secrets Act,18 l

USC Sec.1905, and NRC regulations 10 CFR 9.17(aX4), -2.790(aX4), and  :

2.790(dX1) for " trade secrets and commercial or financial information obtained from a person and privileged or confidential" (Exemption 4). The material for which j exemption from disclosure is here sought is all " confidential commercial l information", and some portions also qualify under the narrower definition of " trade j secret", within the meanings assigned to those terms for purposes of FOIA  !

Exemption 4 in, respectively, Critical Ma== Fnerry Proiect v. Nuclear Reentatary l Commission. 975F2d871 (DC Cir.1992), and Public Citiven Health Ree* mech Groun v i

v. FDA,704F2dl280 (DC Cir.1983). i (4) Some examples of categories of information which fit into the definition of l proprietary information are: l l

a. Information that discloses a process, method, or apparatus, including supporting i data and analyses, where prevention ofits use by General Electric's competitors l

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GBS-95-9-afPBmod4. doc Affidavit Page 1 l

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without license from General ~ Electric constitutes a competitive economic j g . advantage over other companies

b. Information which, if used by a competitor, would reduce his expenditum of resources or improve his competitive position in the design, manufacture, j shipment, installation, assurance of quality, or licensing of a similar product;

c. Information which reveals cost or price information, production capacities, l '

. budget levels, or commercial strategies of General Electric, its customers, or its suppliers; ,

d. Information which reveals aspects of past, present, or future General Electric customer-funded development plans and programs, of potential commercial .j value to General Electric; i

e. - Information which discloses patentable subject matter for which it may be j desirable to obtain patent protection.

l Both the compilation as a whole and the marked independently proprietary elements incorporated in that compilation are considered proprietary for the reason described in items (4)a., (4)b. and (4)e., above.

(5) The information sought to be withheld is being submitted to NRC in confidence. l That information (both the entire body ofinformation in the form compiled in these  !

drawings, and the marked individual proprietary elements) is of a sort customarily I held in confidence by GE, and has, to the best of my knowledge, consistently been  :

held in confidence by GE, has not been publicly disclosed, and is not available in i public sources. All disclosures to third parties including any required transmittals to l NRC, have been made, or must be made, pursuant to regulatory provisions or i proprietary agreements which provide for maintenance of the information in l confidence. Its initial designation as proprietary information, and the subsequent j steps taken to prevent its unauthorized disclosure, are as set forth in paragraphs (6)  ;

and (7) following.

(6) Initial approval of proprietary treatment of a document is made by the manager of l the originating component, the person most likely to be acquainted with the value and sensitivity of the information in relation to industry knowledge. Access to such documents within GE is limited on a "need to know" basis.  ;

(7) The procedure for approval'of external release of such a document typically requires review by the staff manager, project manager, principal scientist or other equivalent i authority, by the manager of the cognizant marketing function (or his delegate), and by the Legal Operation, for technical content, competitive effect, and determination of the accuracy of the proprietary designation. Disclosures outside GE are limited to i regulatory bodies, customers, and potential customers, and their agents, suppliers, j GBS-95-9 aiPBmod4. doc Affidavit Page 2 a ,_.

W and ' licensees, and others with a legitimate need for the infonnation, and then only in . f' accordance with appropriate regulatory provisions or proprietary agreements.

(8) The information identified in paragraph (2) and the Attachment, above, is classified e as proprietary because it constitutes a confidential compilation of information,  ;

including reports and detailed design drawing results of a hardware design  ;

modification (stabilizers for the shroud horizontal welds) intended to be installed in '

a reactor to resolve the reactor pressure vessel core shroud weld cracking concern.

The development and approval of this design modification utilized systems, components, and models and computer codes that were developed at a significant l cost to GE, on the order of several hundred thousand dollars.

l The detailed results of the analytical'models, methods, and processes, including computer codes, and conclusions from these applications, represent, as a whole, an ,

integrated process or approach which GE has developed, and applied to this design modification. The development of the supporting processes was at a significant additional cost to GE, in excess of a million dollars, over and above the large cost of i developing the underlying individual proprietary reports and drawings information. ,

(9) Public disclosure of the information sought to be withheld is likely to cause i

substantial harm to GE's competitive position and foreclose or reduce the availability of profit-making oppoitunities. The information is part of GE's comprehensive i BWR technology base, and its commercial value extends beyond the original  :

development cost. The value of the technology base goes beyond the extensive  ;

physical database aad analytical methodology and includes development of the i expertise to determine and apply the appropriate evaluation process. In addition, the  ;

technology base includes the value derived from providing analyses done with  ;

NRC-approved methods.  ;

GE's competitive advantage will be lost ifits competitors are able to use the results l of the GE experience to avoid fruitless avenues, or to normalize or verify their own j process, or to claim an equivalent understanding by demonstrating that they can i arrive at the same or similar conclusions. l While some of the underlying analyses, and some of the gross structure of the  ;

process, may at various times have been publicly revealed, enough of both the l analyses and the detailed structural framework of the process have been held in , l confidence that this information, in this compiled form, continues to have great i competitive value to GE. This value would be lost if the information as a whole, in -

the context and level of detail provided in the subject GE drawings, were to be disclosed to the public. Making such information available to competitors without  :

their having been required to undertake a similar expenditure would unfairly provide l competitors with a windfall, and deprive GE of the opportunity to exercise its  ;

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__ _ _ _ _ _ _ - __ - _ , . . , . , - ~ . , . . ,,

competitive advantage to seek an adequate' retum on its large investment in

- developing its analytical process.

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STATE OF CALIFORNIA )  !

) ss: )

COUNTY OF SANTA CLARA ) l l

l George B. Stramback, being duly sworn, deposes and says: j That he has read the foregoing affidavit and the matters stated therein are true and correct to the best of his knowledge, information, and belief.

Executed at San Jose, California, this /6 day of 8 1995.  !

/ i c9r19 3 ,

/0 JJosh ~

C/orge B/Stramback l General Electric Company Subscribed and sworn before me this /bN day of 1995.

0 2 n0 Public, State of Califordi

. ---m JUUE A. CURT 5

- COMh4. # 974657 g i Notcry Pub 6c - Cohfomk2 SANTA CtARA COUNTY l W Comm Expres SEP 30.1996 '

GBS-95-9-afPBmod4. doc Affidavit Page 5

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Att:chmat i

Drawina Numher B. I12D6348 Rev.4 Stabilizer Support Assembly I. I12D6353 Rev.3 Support, Upper K. I12D6354 Rev.3 Suppon, Upper H. I12D6498 Rev.3 Bracket, Upper spring EE. I12D6502 Rev.3 Coupling, Top Support Bolting AAA. I12D6752 Rev.1 Spacer, Upper Support CCC. Il2D6788 Rev.0 Bolt, Top Support DDD. I12D6789 Rev.O Retainer i

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t 1 DOCUMENT PREVIOUS CURRENE REASON e .%. . NUMBER *s ^

SUBMTTAL SUBWTTAL FOR_

PO ' MDOCUMENT DESCSBPTION - #

REVISOtt #REVISON t  ; CHANGE s w A' s

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4 1, .

. g g) .

I REPAIR HARDWARE, DESIGN SPEC. 25A5579 3  ! 3  ! 1 ISTABI'32ER CODE, DESIGN SPEC. 25A5580 4 i 4  ! t

~ FABRICATION SPEC. I 25A5601 1 1 1 i t CLEANING AND CLEANLINESS CONTROL 21A2040 1 i 1 i 1 INSTALLATON SPECIFiCA"lON 25A5581 2 2 1 REACTOR VERL STRESS REPORT 25A5607 4 4 1 SHROUD & REPAIR HARDWARE =iMt;SS ANALYSIS 771 58 0994 4 4 1 37ABluZER INSTALLATON , DESIGN REPORT 771-59-0994 4 4 1 SEISMIC ANA.YSIS 771-6CW994 - 2 2 1 FIELD DISPOSITION INSTRUCTON 02S7 71067 1 1 1 PARTS UST PL112D6347 1 1 1

_ PARTS UST PL112D6348 1 2 8 PARTS UST PL11206349 0 0 1 PARTS UST PL112D6358 0 0 1 PARTS UST PL112D6350 0 C 1 PARTS UST PL11206360 0 0 1 PARTS UST PL112D6495 0 0 1 PARTS LIST PL105E1455 2 2 1 NUT, TIE ROD 11206313 0 0 1 NUT, TOP SUPPORT 112D6321 3 3 1 BOLT, TOP SUPPORT 112D6322 0 0 1 NUT, TOP SUPPORT 112D6323 0 0 1 RETAINER 112D6324 9 1 1 SPRING, RETAINER 112D6325 1 1 1 SLEEVE. JACK 11206327 0 0 1 WASHER. JACK 112D6328 0 0 1 RING. MID SUPPORT 112D6331  ? 2 1 SCREW, h4D SUPPORT '. 112D6332 0 0 1 LATCH 11206338 0 0 1 UPPER STA81UZER 112D6347 2 2 1 STABlu2ER SUPPORT ASSEMBLY i 112D6348 3 4 8 TIE ROD ASSEMBLY I 11206349 1 1 1 ROD, TIE I 112D6350 3 3 1 SPRING, LOWER I 112D6351 1 1 1 SPRING, UPPER i 112D6352 2 2 1 SUPPORT, UPPER i 112D6353 . 2 3 8 SUPPORT i 112D63~,4 6 2 3 8 CONTACT, LOWER I 11206355 1 1 1 SUPPORT,A4D i 112D6356 1 1 1 1 CONTACT, UPPER i 11206357 1 1 1 i TIE ROD / SPRING ASSEMBLY  ! 11206358 1 1 1 MiD SUPPORT i 112D6359 1 1 1 LOWER STABluZER i 11206360 1 1 1 BOLT, TOGGLE i 112D6489 2 2 i 1 SUPPORT, LOWER 1 11206400 2 2 1 iTOGGLE I 112D6491 2 2 1 1 PIN. TOGGLE BOLT 4 11206492 1 1 1 WASHER 70GGLE BOLT I 11206493 1 1 1 NUT, TOGGLE BOLT I 112D6494 2 2 '

1 TOGGLE BOLT ASSEMBLY -

! 112D6495 1 1 i 1 BOLT. JACK i 112D6496 1 1 1 1 ISPRING, RETAINE R i 112D6497 1 1 i 1 BRACKET, UPPER SPRING i 112D6498 2 3 i 8 SCREW, TOP SUPPORT BOLTING 112D6501 1 1 i 1 COUPUNG. TOP SUPPORT BOLTING 112D6502 2 3 i 8 EXTENSION, LOWER SPRING i 112D6503 1 1 1 1 PIN i 112D6504 1 1 i 1 PIN, CLEVIS i 11206505 1 i 1 1 1 ARM, TORSION 112D5242 1 1 i 1 BOLT, TORSION ARM i 112D5243 1 1 1 1 NUT, LOCK i 112D5244 0 0 1 MODIFICATlON DRAWING i 105E1455 2 2 1 SPACER, UPPTPiSUPPORT i 11206752 0 1 8 NUT, TIE ROD i 112D6777 0 0 1 BOLT, TOP SUPPORT 112D6788 N/A 0 8 RETAINER 112D6789 N/A 0 8 GE RESPONSES TO NRC QUESTIONS (GENERIC) DRF B13 01732 0 0 1 RESPONSES TO NRC QUESTONS FOR PBAPS GENE-B13-01732-001 N/A 0 i N/A

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( NOTES:

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1. NO CHANGE SINCE LAST SUBMITTAL

2. DRAWING CORRECTIONS, NO CHANGE IN DESIGN.

3. INCORPORATED UNIT 3 SEISMIC ANALYSIS INFORMATION.

4. MINOR MODIFICATION TO IMPROVE LOAD CAPACITY, FABRICATION, OR ASSEMBLY. l S. DELETED HEATTREATMENT REQUIREMENT FOR THREADS. <

6. INCORPORATED ANALYSIS OF CORE SPRAY PIPING INSIDE THE VESSEL  !

7. SCOPE INCREASE TO ADD WELD H8 EVALUATION <

8. INCORPORATE LESSONS LEARNED IMPROVEMENTS 9 INPROCESS REVISION DUE TO PECO/GE COMENTS AND APPROVAL Gu*%ur b

e