ML20198Q009

Responds to 971028 Telcon Between PECO & NRC Re crack-like Indications Discovered in HAZ of Three Jet Pump Riser to Thermal Sleeve Welds at Unit 3.Copy of 10CFR50.59 & Summary of Associated Analysis Encl

ML20198Q009
Person / Time
Site:	Peach Bottom
Issue date:	10/30/1997
From:	Mitchell T PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NUDOCS 9711120084
Download: ML20198Q009 (40)

Text

_ _ _ _ _ - _ _

t Thoman N.Mitchell W:e hewsont 1%ch Bottorn Atorme IbwL Staten y

l PECO NUCLEAR- i1848

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A Unit of PECO Energy gitag7p 9032 l raa 717 456 4243 l

10CFR50.59

Odober 30,1997 j Docket No. 50 278 License No. DPR 56 U.S. Nuclear R*0ulatory Commission Attn

Document Control Desk Washington, DC 20555

Subject:

Peach Bottom Atomic Power Station, Unit 3 Indications In Jet rump Thermal sleeve to Elbow Riser Welds Deer NRC Official:

This letter is in response to an October 28,1997 telephone conversation between PECO Energy Company (PECO Energy) and the U. S. Nuclear Regulatory Commission (USNRC) staff concoming crack like Indications discovered in the Heat Affecled Zone (HAZ) of three (3) jet pump riser to thennal sleeve welds at Peach Bottom Atomic Power Station (PBAPS), Unit 3. As discussed l'i the October 28,1997 conversation, PECO Energy is providing a copy of the 10CFR50.59 and a summary of the associated analysis.

Unit 3 operation will be limited to stay within the conditions specified in the 10CFR50.59.

If you have any questions, please do not hesitate to contact us.

Very truly yours, 4

T. N. Mitchell Vice President '*

Peach Bottom Atomic Power Station

Enclosures:

Attachments {

cc: H. J. Miller, Administrator, Region 1. USNRC R. S. Barkley, USNRC Senior Resident inspector, PBAPS D K 78 P PDR CCN: 9714066 lllhllh!{lll$!!!

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Peach Bottom 3 Jet Pump 1 Riser-to-Thermal Sleeve Weld Cracking .

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b Overview 4

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+ Outage Inspection Plan

+ Riser Inspection Results

+ Analysis Approach

+ Limit Load Allowable Flaw Size

+ Crack Growth Considerations

+ Allowable Operating Condition

+ Future Plans

+ Conclusions t

2 l

O 3R11 Jet Pump Inspections -

1

. Orginal Outage Scape

- VT-3 on JP's in accordance with ISI program ,

- Modified VT-i on Riser Braces

- Additional Outage Euspections Driven by industry experience 1/31/97 BWRV1P Lette.

- MVT-1 Thermal Sleeve to Elbow Weld Inspection on all 10 Jet Pump Risers (RS-1) ,

- Results

- 1 indication on each of 3 separate risers

• Expanded IVVI Scope

- Per VIP-41 guidance, look at remaining locations of the same type (i.e.,

locations with same ID number). Consider effects of degradation of one location upon other locations .i Inspected all 10 thermal sleeve to riser elbow welds (RS-1)

- Pericrmed MVT-1 of all riser welds on affected JP Risers (RS 2&3) 3

Riser Inspection Results

- All indications are on the thermal sleeve

"" side of weld RS-1,in the HAZ.

\ h

- The thermal sleeve is 304 SS and all welds are GTAW (non-flux).

- lJT examination of each thermal sleeve has

( _ ,,/ ,

quantified the following crack lengths and orientations (12:00 being the top ofthe

/

\

thermal sleeve when looking into the elbow l

[ , fiom tlie vessel OD):

. D 1'* .IP I&2 Riser 11:00 to 3:00 (10.8 IN.) +/- Tol.

.JP 9&10 Riser 1:00 to 2:00 (l.7 IN.) +/- Tol.

GE Inspection Report

.IP 13&l4 Riser 10:00 to 2:00 (l2.7 IN.) +/- Tol.

4

Basis for Expanded Scope

- Generic Implications

- Inspect all 10 Jet Pump RS-1 welds

- Consider the effect that degradation of the RS-1 location has on other locations

- Part of same welded structural member (riser) ,

- Slipjoint separates other high priority welds

- Diffuser welds do not preveru disassembly / ejection ofjet pumps

Other Welds Considereu

- Diffuser shell to tailpipe (DF-2), Adapter Backing Ring (AD-3a),

Adapter to shroud support plate (AD-2)

Structurally separated from RS-1 by slipjoint .

Negligible vibration loads Low Stress /fiitigue predominantly weld residual stress Flaw tolerant locations due to low primary stresses and absense of vibratory loads Different fit-up during installation No industry evidence of cracking 6

Summary

- Performed RS-1 weld inspections based on industry experience

- Performed UT to confirm visua. indications ,

- Expanded scope per the guidance of VIP-41 l

,l 1

7

Analysis Approach l

l

+ Determine allowable flaw size based on limit load methods l

- Appropriate load combinations from BWRVIP-41 and l

l PBAPS UFSAR

- Analysis per BWRVIP-41

+ Calculate FIV loading at crack location

- Based on BWR/4 251 inch diameter baseline Dow tests

+ Determine crack length vs. operating time .

- Consider SCC and fatigue growth

- Determined acceptable loads for desired operating window 8

Crack Growth Considerations

+ Approach -

- 100% flow strain time-history data at the riser brace in the boundingjet pump in the baseline plant

- Develop PBAPS loads from baseline plant data '

- Calculate the stress time-history at the crack location

- Pair stress amplitudes to maximize peak-to-peak stresses

- Apply stress ranges to crack, using Section XI, App. C relationships for fatigue crack growth

- Include SCC growth at 5x10-5 in/hr .

- Calculate crack length vs. operating time

IGSCC Relationship =

+ NRC approved IGSCC growth rate of 5E-5 in/hr assumed

- Conservative for Unit 3, where 13 SCFM HWC injection provides  ;

a factor of 2 improvement on growth rate l

+ Crack growth calculation conducted in several steps Fatigue and SCC growths separately calculated over a small time interval (100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br />) and then added to obtain Aa Calculation repeated for next time interval after updating crack length by calculated value of Aa

- Calculations extended up to 16000 hours of operation 10 f

Allowable Operating Condition

+ Benchmark of crack growth model

- Model predicts rapid cracking at 100% flow conditions, most of which is fatigue growth

- Clearly conservative given site observations that cracks exhibit IGSCC only

+ Benchmarked model shows margin in operation at reduced flows

- Decreased flow drops vibration stresses significantly

- Stress is proportional to Flow2 11

Plan for Continued Plant Operation

+ Short Term + Long Tenn Based on the results of the -

PB will prepare to install a thermal sleeve crack repair to JP's 1&2 and analysis,"A Sacture 13&14 within the time mechanics analysis was frame bounded by the performed to establish analysis.

acceptable conditions for - Cracking in JPs 9&l0 will continued plant operation." be re-evaluated during

- Monitoring of Jet Pump 3R12. Repair will be based flows will continue per upon inspection results.

) - Coordinate with VIP Repair I guidelines in SIL No. 605, TS Surveillance Committee for generic Requirements 3.4.2.1, and repair options.

existing plant procedures.

12 I

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b 1

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Conclusions .

+ The analytical model used to calculate crack length  ;

vs. time is conservative

+ Based upon a conservative model, drive flow vs. time

. ]

determined to assure safety margins are maintained

+ PECo Energy has selected operating conditions of '

flow and time with sufficient margin relative to the allowable curve i

1 Continued Operation is Justified 13 j

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4 ATTACHMENT II e

i. . .

'n Exhibit AG'-CG-4-1, Rev. 4 Effective Date r 08-21-97. .

? PORC /- SQR / Chesterbrook Review & Approval Fona

• • • • • • • • • • • • • • • • • • • Section 1.- Description of Item *****************

i 1. Item (s) . (Procedure or Document 8, dreR riri s Ties):

f /dCM.ya.9 kettat e ble l i M P'TkevertAl be,vr- [rtAcltW 1 [ Uma T 1 EeV. d

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Docu:nonto to les superseded:

ya o a Al,lA l

, 2. Prepared By: had E b Group: PaAP5 Mc. A(pdEnt: err-4f#7 l- 3. Indioete .^, . -.-d C.g. ' - "' - de) req'd: C8M LOS PSAPS M

4. Full Cn ^ . of Activity or Change: hd SA NA*TS/ L/_M A TibJ A Jea.s us I, vere! m l L LMK A.E-T T OIA Anis-rbrJ brL At(12 6 8 2 A 99 , Eel. d .

• rk pNb dJ b.J l m,.ta t i r L n,c 6e 4 Pd 4 L<rL, ' wr' ' L L ais.; Jis

5. If e TC to being fully :i. , O'in thie'che4e, list TC8(8) here (if multiple docuisente approved on thfe -

l form, state which pecoedures incorporate which TC's):

[ ************ Section 2 -

Review and Planning of. Activity*************

j 6. Appitomble items in AG CG-4 2 toisie are entieRed and commitmente annotated . Properor's inPJele: ,,

j, Se. For Admin Procedures, AA-C 6 obechliet completed and atte.:hed . Properor's initiale: A(/A

7. R$ochenieme in piece to *- ::-- *. . Properor's 1 Training Review (LGS only):

C 5.c. A A Ii t 1510 .

l 8. Procedure Cause Codes (see beck). List the program ti.et tad this change : 8"T N,

8e. List below the program that failed to identify the change before the one in step 8

"se ,.15I Give Beeis for determination: I V V I Antt. W E t / 4 "I l % f *I n d .

9.' 10 CFR 50.50 Review . olieck appropriate item and ettech review as m:

Determination Only NO paview Req'd per LR-C-12 (AEnor Non-Technsoal Revision)

No Rev8ew Reg'd (provide beeis in descrip)_ Safety Evel(PORC req'd) USQ

Yes_._ No/ -

s -

10.10 CFR 50.54 Reviews (EP QA, Sec Plan changes or for LOS only, envwonmental impact) . ettech review f

11. Document Services instructions: Procedure use (circle one)i i- 18 lit f Normal Distribution l Expedited Distribution (24 hre) l Hold .

Contact:

)? Need by date: Other instructione: ,

Specific Effective Date if req'd: (Weekly teste performed daily should have a toonday off date)

• • • • • • • • • • • • • • Section 3 - SQR/RS Review and Approval **************

4

12. Crose-Diecipline Review. perform per Exh AG-CG-4-2. Attach addl eheets as necessary.

j SQRXD Name SitelGroup inibals Date SQRXD Name SitelGroup initiale Date Pm e A , +r W M wde cah o. d6d ,4.1.,

Gu L, rac ia oep Ur ' '

i l V l G A.o w rks /m C'.M Properor's inihets that est commenterrosoeved. P51G[ ' SQR initiale if no-croes disciphne review:

13.M Review Signature: n Date: /Q/2a/f7

^

13a. Queitty Reviewer Signature (for Quality admin eduree): A) A Date:

14. Responsible Superintendent Dir for C8) Revsew/ Approval check appropriate box:

Reviewed erd Approved _ ,., Reviewed O PORC req'dI Reviewed Only, Pl Mgr Approval reg'd (LOS only)

R8 Dir for C8) Segnature -

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Date: #./I/O[

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

I Exhibit AG-CG-4-1, Rov 4

. Page 2 of 2 PORC / SQR / Chesterbrook Review & Approval Form Item (s) From Line 1 - (Procedure or Document #, draft rev'#, Title):

} DOR gb.59 hnwu , Crx,$-r &AD %nttEM4 L R bedL btAr &mG,.0mHs $V f

• • • • • • • • • • • • • • • • • • • • • • Section 4 - PORC Review **********************

i

15. PORC Review j- Approval Recommendeu VApproval Recommended wf below rh Reviewed Only, no action taken

. Romanded

, items for Resolution. . . * ,

$4 AK h 1 ou* Jh.'(/d AL LJu Me b C 4 ' dtecan M La r eb E/Ls r ta tX .

Resolution Choc sy: Date: /O /[h 1 PORC Mtg8 bh A, M.30l9 7 PORC Chairman Signature: _O

/

18.Walkaround PORC (limited use only- see AG CO 4):

I f PORC Chairman Approval for weikaround Waikaround 8: Date:

4 initiale signifying acceptance by PORC membordelternates (eneure at least 4 individuale of which at least G '

' 2 are membere): .

PORC Chairman Approval ofitem: Date:

17. Plant Manager (as ery see A 2): i i Plant Manager: - II/k/ -

e Date: /d ,

//77

NQA/QR Appro, val

Date:

Vice President. Date:-

(for Directives only)

18. PORC Minutse (req'd for all items presented tc. PORC or PORC Walharound, whether approved or not. Provide electronic copy to PORC Sec'y . ese Exhibit AG-CO-4 2 for minutee requiremento).

'******************* Procedure Cause Codes ************************

The followine <== codes are to be used when anevsering itens 8 and 8a on first pene of form:

CTP . Commitment Trackingfoperating Experience PEP (*) . Performance Enhancement Program Program DSD Design Basie Docuenent Program PPtS(*) . Procedure Performance improvement System DCR Design Change Request Program SET - Setpoint Change Program DEC . Design Equevaient Change Program TCP(*) Temporary Changa Process LIC Ucenomg Program (e.g. Tech Spec, UFSAR) TPA. Temporary Plant Alteration Program MOD Mod Procese (Big) VEM VendWr.mnuel Program MPC Minor Physical Change Program OTH An Unheted Process list in basis (step 8a)

. (*) . When t%ie program is identded for irntiati.g a change to a procedure, cor. sideration should be given that another process should have or could have initiated sieu change earlier.

I

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^ Peach Bottom' Atomic Power Station Unit 3 1 10 CFR 50.59 Review for Jet Pump Thermal _' Sleeve Cracking

Revision 0 -

l '

j ..  :

1. Subject.

e i - During Peach Bottom Unit 3 Refueling Outage (3R11) In Vessel Visual-Inspections (IVVI) of this location were conducted per reference 9. Cracks

' were found in the . weld HAZ joining the Recirculation inlet nozzle thermal sleeve to the elbow on three. Jet Pump riser. assemblies. The cracks were i

. found on the thermal' sleeve side of.the weld on the risers associated with Jet Pumps 1 and 2 (Nozzle N2E at150 deg. Azimuth),9 and 10 (Nozzle N2A '

i- 30 deg. Azimuth), and 13 and 14 (Nozzle N2J at 300 deg. Azimuth). The -

cracks at 30 and 150 degrees are on the "B" loop of the Reactor .

(-

Recirculation system and the crack at 300 degrees is on the "A" loop of the -

Reactor Recirculation system.

7

!~ This 10 CFR 50.59 Review will address the INTERIM USE AS-IS disposition -

l of NCR 97 02899 for cracks on the Jet Pump riser elbow to thermal sleeve

, weld heat affected zone (HAZ). The INTERIM USE AS-IS disposition is valid for continued operation within the Reactor Coolant Recirculation drive flow and time constraints evaluated by GENE.

The INTERIM USE AS IS disposition allows for administratively controlling-(Ref. 20) Reactor Coolant Recirculation drive flow to a NOMINAL value of 13.85 Mlbm/hr for.each recirculation loop for a period of up to 8 months.

This was selected as an operating strategy, hereinafter known as the "specified operating condition".

Transients, outside the specified operating condition, such as single loop operation ~or excursions above nominal values are bounded by the analysis.

Extended single loop operation greater than 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />s-will be evaluated by engineering for impact on the.specified. operating conditions.

The result of operating at higher Reactor Coolant Recirculation drive flows for extended periods of time could reduce the 8 month operating period.

. This will require evaluation by engineering for impact on the specified operating condition.

10CFR50.59 Riview for NCR 97 02899 Rsv. 0 -

- U/3 JP Thermal Sleeve Crack - 4

- Page 2 of 17 f-

,1 L 11. Discussion t

Jet Pumo Confiouration The Jet Pumps are Reactor Vessel Internals and in conjunction with the F - Reactor Coolant Recirculation system are designed to provide forced I

circulation to the core for heat removal from the fuel. The Jet Pumps are located in the annulus reg!an between the core shroud and the vessel wall.

-Since the Jet Pump suction elevation is at 2/3 core height, the reactor core ,

will remain covered to this height even with a complete break of the 5

'Recirculatio1 piping as assumed in the design basis accident (DPA). During post-LOCA LPCl operation, the Residual Heat Removal system pumps take suction from the suppression pool and discharge into the core region of the

! reactor vessel through the recirculation loops (i.e. through the Jet Pumps j into the core region).- LPCI helps to restore and maintain the' coolant

'~

inventory

{' in the reactor vessel such that the core is adequately cooled to preclude fuel' clad temperature in excess of 2,200 deg. F following a design basis LOCA -

, (Ref. 4).

i Each Reactor Coolant Recirculetion loop contains ten Jet Pumps,

Recirculated coolant passes down the annulus between the Reactor Vessel wall and the Core Shroud. Approximately one third of the coolant flows from the vessel,- through the two external recirculation loops, and becomes the driving flow for.the Jet Pumps. Each of the two external recirculation loops discharge high pressure flow into an external manifold from which individual recirculation inlet lines are routed to the Jet Pump risers within the Reactor Vessel. The remaining portion of the coolant mixture in the annulus becomes the suction flow-for the Jet Pumps. This flow enters the Jet ,

Pumps at suction inlets and is accelerated by the drive flow. The drive flow and the suction flow are mixed in the Jet Pump throat section. The total flow then passes through the Jet Pump diffuser section into the area below

. the core (lower plenum), gaining sufficient head in the process to drive the

. required flow upward through the core.

-The recirculation inlet nozzle thermal sleeve is welded to the nozzle safe end

at its outer extremity and to the jet pump riser elbow at its inner extremity.

i 'The thermal sleeve is designed to provide a pressure retaining flow path for

Reactor Coolant Recirculation drive flow to the Jet Pumps. Secondarily, the

' -thermal sleeve reduces temperature variations, and thus thermal loading, on 2t t

1 ey - - , .. . . .-. , -- + .w- .- . ,- - _ . - - - r--- -_ = __ __ - ~ _ - - -__-_ - --_ '

_ , , y.

10CFR50.59 Review f$r NCR 97-02899 Rov. 0 -

U/3 JP Thermal Sleeve Crack -

Page 3 of 17

- the recirculation inlet nozzle. The thermal sleeve is not a primary pressure boundary.

The thermal sleeve is 10" schedule 40 stainless steel type 304 pipe. The i thermal sleeve tc riser elbow joint is a field weld, performed during Jet Pump installation into the Reactor Vessel. The welding process was gas-tungsten arc with type 308 filler material. During weld preparation,'the-1 thermal sleeve was counter bored for appropriate fit up to the schedule 30 riser elbow. The welds are' non flux, non creviced, full penetration butt.

welds (Ref. 8 & 17). .

Crack Descriotion/ Geometrv Based on a review by an expert metallurgist from PECON Testing and Laboratories, visual examination of all the Jet Pump indications are:

characteristic of Intergranular Stress Corrosion Cracking (IGSCC) in the heat affected zone of the austenitic stainless steel circumferential pipe weld. The cracking is away from the toe of the weld (approximately 1/8"_ to 1/4") and jagged in appearance. The crack ends were intermittent and at the same L relativo distance from the toe of the weld. ' No indication of fatigue crack growth was observed in that the crack tips did not turn and follow the toe of the weld, the cracks were jagged and not straight lined, and no crushing.

of the crack faces was observed. However, boat samples were not 4

obtained of the crack tips to rule out the possibility of fatigue cracking. ..

The initial visual examinations were performed using modified VT-1 (1 mil L

wire) standards. __ Supplemental ultrasonic examinations (UT) were performed at the crack locations and the results are listed below.

Crack on Jet Pumps 1 and 2 Riser The thermal sleeve-to-elbow weld has a crack from 329.5 deg. through 84.6 deg., looking in the direction of flow. This corresponds to a length of 10.8 +/- 0.39 inches for uncertainty, consistent with BWRVIP protocol.

The flaw is on the Thermal Sleeve side of the weld.

Crack on Jet Pumps 9 and 10 riser -

The thermal sleeve-to-elbow weld has a crack from 12.1 deg. through-30.2 deg., looking in the direct!on of flow. This corresponds to a longth of 1.7 +/- 0.34 inches for uncertainty, consistent with BWRVIP protocol. The flaw is on the Thermal Sleeve side of the wold.

10CFR50.59 Review for .

4 NCh 97-02899 Rev. 0 U/3 JP Thermal Sleeve Crack Page 4 of117 j

Crack on Jet Pumps .13 and 14 riser.

The thermal sleeve to-elbow weld has a crack from 305.5 deg. through =

81.0 deg., looking in the d>;ection of flow. This corresponds to a length of 12.7,+/ 0.34 inches for uncertainty, consistent with BWr4 VIP protocol.

The flaw is on the Thermal Sleeve side of the weld.

Root Cause lGSCC is considered to be the most likely initiator of this cracking. The I_

cracking on the Thermal Sleeve is similar to cracking identified to date in i

other Reactor Vessel Internals. Although not a creviced joint, stainless steel i type 304 materials ussd for the thermal sleeve, in conjunetion with past poor water chemistry conditions have made the joints susceptible. Records also indicate the possibility of these being cold sprung during installation thus increasing the residual stresses in the area and increasing the joints susceptibility to IGSCC.

- Code Boundarv -

4 Jet Pump components are not part of the primary pressure boundary and do not provide a core support function. Jet Pumps are Safety Related and are optionally classified as ASME Section XI components for inspection l purposes only. The Jet Pumps provide a Safety Related flow path during

LPCI injection.

=

Flaw Evaluation U The determination of structuralintegrity was performed by using standard I accepted methods for intergranular stress corrosion cracking and fatigue.

Although examination of the crack indicates that IGSCC is the sole

. contributor, fatigue loading was also considered in developing allowable flaw sizes. The source for fatigue crack' growth was determined by analytical methods to be low amplitude-high frequency vibration from the high velocity recirculation line flow.

The allowable flaw size at the elbow to thermal sleeve location was determined using standard limit load methodology presented in BWRVIP-41, i- "BWR Jet Pump Assembly inspection and Flaw Evaluation Guidelines" I 1Ref. 5). Similar methods have been previously used to evaluate other

[ Vessel components such as the core spray lines and shroud. The flaw 3 evaluation methodology used was performed consistent with ASME Section i.

s a

..,: o-'~

-10CFR50.59 Roview for ~

NCR 97 02899 Rov,0 l U/3 JP Thermal Sleeve Crack e Page 5 of 17 LXI, Appendix C requirements (Ref. 7). This evaluation includes the ASME Section XI Safety Factors of 2.77 for Normal and Upset and 1.39 for

_ Emergency and Faulted conditions. Load combinations are in accordance i

1 -

with the UFSAR and BWRVIP-41. -

, Once the allowable flaw size was determined, the acceptability of an '

observed flaw was determined by performing a crack growth analysis. This I; analysis considered both IGSCC and fct!gue !csding. The IGSCC growth was predicted using the conservative standard of 5 x 10'in/hr crack i

growth rate from each crack tip. This growth is the accepted bounding Industry standard for IGSCC in austenitic stainless steels in a BWR

, environment with normal water chemistry. This is expected to be s

[ conservative since the Thermal Sleeve to elbow is a non-creviced weld and PBAPS injects hydrogen into feedwater at a rate equating to 0.3 ppm. The l actual growth rate is expected to be on the order of 2.5 x 10 in/hr.

. Each crack was then evaluated against its susceptibility to fatigue cracking.

3-Fatigue cracking in the riser piping is primarily a result of flow induced vibration caused by the recirculation drive flew. A time history of stress amplitude vs. time for the Jet Pump risers was obtained using baseline j testing of a BWR4/251" dia. Reactor Vessel (Browns Ferry Uriit 1). During

- start-up testing at Browns Ferry, strain measurements on the Jet Pump riser i braces were obtained at varying power levels and flow conditions.

l Measurements at the riser brace were scaled to the riser crack location by.

means of modal shape factors, determined analytica!Iy. Data corresponding

" to 100% core flow at 100% power were used to evaluate the influence of .

fatigue cracking on the subject risers.

Results of this analysis concluded the N2A riser cracking is small enough that the crack growth rate will not be influenced by fatigue cracking through the next 2 year cycle of full power operation (AK is less than AK threshold).

.Therefore, crack growth is limited to IGSCC and crack size will be limited to

= 3.7 inches by the end of the 2 year cycle and is acceptable to use as-is.

For the N2E and the N2J risers, the stross intensi_ty range for the assumed loading exceeds the threshold for cusceptibility for fatigue cracking (AK is greater than AK thresheld). When applying fatigue crack growth to both thermal sleeve cracks, the lengths would exceed the limit load allowabb flaw size by_end of cycle.

To mitigate the impact of flow induced vibration on the N2E and N2J Thermal Sleeve cracks, recirculation drive flow will be limited to the

1 .. -

10CFR$0.59 Review for NCR 97-02899 Rev. 0 l U/3 JP Thermal Sleeve Crack g .

Page 6 of 17 L specified operating conditions.' The predicted end of operating condition i

- flaw sizes are listed below. j Peach Bottom Unit 3 Flaw Evaluation Summarv i Location -Current Length

• Predicted *

• Allowable Percent of

.(in.) Longth - Flaw Length Allowable (in.) - (in.) = Flaw Length JP.1/ 2 11.2 12.5 17.9 69.8 %

JP 9 /10 2.1 3.7 17.9 20.6 % '

LJP 13 /14. 13.1 14.6 17.9 81.6 %

Length was used in GE analysis and includes UT uncertainty, ,

reference 1, 2 and 3.

Flaw length predicted to occur at the end of operating period - based on the specified operating conditions JP 9 /10 is based on a 2 year normal operating cycle.  :

L9akaae Evaluation Due to the small crack opening area, any leakage through the cracks wi!I be minimal. Postulated leakage will be less than 120 GPM total for Reacter Coolant Recirculation flows and 76 GPM for the inservice LPCI flows post-LOCA as calculated using guidance provided in BWRVIP-41. This conservatively assumes the crack gro_ws to the allowable flaw length of 17.9 inches. There is no specified allowable design leakage limit for the Reactor Coolant Recirculation flows and the postulated leakage is negligible when compared to system flows. The original design allowable leakage of 3000 GPM, for Low Pressure Coolant injection (LPCl),' will not be exceeded.

Therefore, crack leakage during operations and post accident, for the specified operating conditions, will not impact any ECCS/LOCA analysis (Ref. 4).

.-~ _ - . _ _ _ . . - _ __ _.-. _ _ _ _ _ . ._ _. _ . . _ .. _ _ _ _ . _ _ _ _ _ . _ . _ _ _

7-..

10CFR50.59 Review for NCR 97-02899 Rev. 0 -

U/3 JP Thermal Sleeve Crack Page 7 of 17

-til. Determination

1. ; Does the activity or discovered condition involve a Technical Specifications change or other Facility Operating (or possession only)

License amendment?

No. -Fracture Mechanics analysis of the cracks and evaluation of potential leakage of Recirculation coolant or LPCI (post LOCA) flow into the annulus .

region of the Reactor Pressure Vessel (RPV) has confirmed the operability of the subject Jet Pumps, Reactor Coolant Recirculation system and_ the Low Pressure Ccolant injection (LPCI) mode of the Residual Heat Removal system ~ 4 for the specified operating conditions. This analysis does not necessitate a change to surveillance requirements or limiting conditions of operation of the Jet Pumps, the Reactor Coolant Recirculation system or the LPCI mode of Residual Heat Removal (RHR) system due to the specified operating conditions on Reactor Coolant Recirculation pump flow. Therefore, the continued operation of the Jet Pumps, the Reactor Coolant Recirculation

- system and the LPCI mode of Residual Heat Removal system as-is does not-require a Technical Specification change or any Operating License amendment.

2, Does the activity or discovered condition make changes to the facility as -

described in the SAR7 Yes. ' Continued operation of the subject Jat Pumps with cracking as -

described above is considered a' change to the facility as described in the SAR. :The original design and analysis of the Jet Pumps consisted of welded, slip joint and bolted connections. There is no consideration for cracking in the original Jet Pump design. Although the subject Jet Pumps are'outside of the ASME Section XI boundary, they continue to meet the structural integrity safety margins as defined by ASME Section XI.1989, Appendix C for the specified operating conditions, including all postulated crack growth.

Potential leakage paths from the floodable inner volume of the Reactor Vessel (e.g.-2/3 core height) during a Rewirculation system pipe break and-subsequent LPCI reflooding is documented in the SAR. Postulated leakago from the Jet Pump cracks during this condition has been calculated to be less than 76 GPM for the inservice LPCI loop. This additional leakage is well

.within the 3000 GPM allowance designed in the LPCI subsystem for potential leakage paths but will be considered a change to the facility as

10CFR$0.69 R: view f:r NCR 97 02899 Rev,0 U/3 JP Thermal Sleeve Crack P 9e 8 of 17 I described in the SAR. Additionally, two loops of LPCI flow through one Reactor Coolant Recirculation loop is less than the specified Recirculation flow limits evaluated for the specified operating condition. The associated piping stresws are therefore bounded by evaluated Reactor Coolant i Recirculation system operation.

) Another poter'tlal leakage path from the Jet Pump cracks, during opvations,

is inside the Reactor Vessel pressure boundary and would not have an

unacceptable effect on the system performance of the Reactor Coolant

Recirculation system. A computation was performed and has determined '

j- -that potential leakage through the cracks is insignificant when compared to i normal system flow through the riser piping. Since the leakage flow has 4

been determined to be Insignificant and contained within the Reactor Vessel

pressure boundary this leakage is not considered to be a changr to the  !

facility as described in the SAR.

i

! 3. Does the activity or discovered condition make changes to procedures as j described in the SAR? i

! No. Jet Pump operability is verified daily per Technical Specification

!_ requirements. Jet Pump dP measurements are used to determine operability snd to calculate core flow and are unaffected by cracks on the Jet Pump

! risers.

The postulated leakage from the cracks will not manifest itself as an additional uncertainty in core flow measurement during plant operations  ;

since the leakage occurs upstream of the Jet Pump flow measurement instrumentation. Furthermore, the flow blased portions of the APRM and Rod Block functions are not credited in the core reload licensing analysis.

Thus the core reload licensing analysis is unaffected.

The flow signal used by the APRM system to establish flow biased rod block and scram trip setpoints is derived from the drive flow transmitters tapped I off of the recirculation pipe venturis. The flow value is r a essed oy the APRM flow units prior to use by the APRM system. Per h.chnical Specification Surveillance SR 3.3.1.1.7, the APRM drive flow signalis

-adjusted accordingly every 31 days to correspond to the total core flow.

Therefore, the postulated leakage that may exist due to the Jet Pump ,

Thermal Sleeve cracks will not impact the accuracy of the APRM flow ,

biased setpoints since the flow signal is gained to correlate to core flow.

The procedure that implements this surveillance is ST l 60A 220-3, Drive i

• ~

10CFR$0.59 R: view f:r l

NCR 97 02899 Rev. O U/3 JP Thermal Sleeve Crack Page 9 of 17 Flow / Core Flow Correlation Check. In addition to this c .veillance, the relationship between APRM flow and core flow is conservatively checked as part of the weekly APRM gain calibration procedure and as part of GP 2 and GP 5.

Based on the above discussion the activity or discovered condition does not make changes to procedures as described in the SAR?

i 4,

Does the activity or discovered condition involve tests or experiments not described in the SAR? 1 No. Continued operation of the Jet Pumps, Reactor Coolant Recirculation system and the LPCI mode of RHR with cracks in the Jet Pump Thermal Sleeves does not involve any tests or experiments not described in the SAR.

When applying accepted crack growth rates for the specified operating conditions to the flaw sizes identified on the Jet Pump thermal sleeves the flaw size is bounded by the limit load allowable flaw size summarized in reference 1. Therefore, margin exists in the remaining thermal sleeve ligaments to assure structural integrity and systems operability during th6 specified operating conditions interval. There are no additional tests or experiments involving riant systems or equipment required for verification of this analysis.

' Since the answer to question 2 is yes, a Safety Evaluation is required for this proposed activity.

IV. Safety Evaluation A. Those accidents potentially n6gatively impacted by this change include those accidents requiring an inner volume containing the core (e.g. 2/3 core height) that can be flooded fol!owing a break in the nuclear system process barrier external to the Reactor Vessel. The Abnormal Operating Transients potentially. negatively impacted by this change are a Recirculation Pump trip, Restart of an Idle Recirculation Pump, and a Recirculation Flow Control Failure.

A 1 May the proposed activity or discovered condition increase the probability of occurrence of an accident previously evaluated in the SAR?

Iff

10CFR60.69 Review for NCR 97 02899 Rev. O U/3 JP Thermal Sleeve Crack Page 10 of 17 No. The safety design basis of a ' Jet Pump assembly is to provide a portion of the floodabla inner volume containing the core. LPCI reflooding of the

. core, post7 LOCA, through the Jet Pumps will prevent excessive fuel

- cladding temperatures ultimately, preventing undue hazard to the health and safety of the public, initiators, assumed failures and sequences for transients and accidents are not affected. The current condition of the Jet

- Pumps is not a new accident initiator. GE's review of all postulated load combinations on the Jet Purops has determined that load combinations including the design basis accident LOCA loids are bounding for all normal, dorsted, Abnormal Operational Transient, and Accident conditions, including those mentioned in "A" above.

The inner volume is defined as:

-1. The Jet Pumps from the Jet Pump Nazzles down to the Shroud support.

2. The Shroud support which forms a barrier between the outside of the shroud and the inside of the Reactor Vessel.

3. The Reactor Vessel wall below the Shroud support.

4. The Shroud up to the level of the Jet Pump Nozzles.

Note: the identifieJ cracks are not part of the inner volume.

A fracture mechanics. evaluation at the specified oportating conditions, using

. the crack lengths verified by the UT data (Ref. 2 & 3) and applying DBA loads, has validated the continued structural Integr_ity of the Jet Pump assemblies for all postulated plant conditions. Therefore, there is no increase in the probability of occurrence of an accident previously evaluated --

'in the SAR for the specified operating conditions.

A-1 May.the_ proposed activity or discovered condition increase the consequences of an accident previously evaluated in the SAR?

No._ The consequences of an accident previously evaluated in the SAR have

not been increased due to cracks identified on the Jet Pump Thermal Sleeve. The flaw sizes identified on the Thermal Sleeves with calculated crack growth for the specified operating conditions are bounded by the allowable flaw size evaluation summarized in reference 1. Tha safety

. function of the Jet Pumps is the passive function of maintaining 2/3 Core coverage, in conjunction with'other Vessel Internals, and to provide a flow path for LPCI injection following a design basis accident. This function is an accident mitigator which allows reflooding of the core in the event of a breach in the nuclear system process barrier external to the Reactor Vessel.

The bounding design basis accident is the Loss of Coolant Accident (LOCA)

10CFR50.59 R; view f:r NCR 97 02899 Rev,0 U/3 JP Yhermal Sleeve Crack Page 11 of 17 as defined in UFSAR Section 14.6.3. Thereforo, margin exists in the romaining thormal sloovo ligament to ensure structural integrity and Jot Pump operability through the specified operating conditions. No safety limit will be impacted and no barrior design limits aro compromised.

Due to the small total area open to flow at the crack locations, any leakago through the cracks during a LPCI reflood (post LOCA) will be minimal.

Leakage through the cracks, including projected crack growth at the end of the speelfled operating conditions, is calculated to be less than 76 GPM for the Inservico loop. This leakage is well within the allowable design leakage documented in the SAR for the LPCI modo of operation.

Sinco the Jet Pump structural Integrity is assured and any additional leakage after LPCI reflooding is within existing system margins the existing accident analysis and assumptions are unchanged and valid for the specified operating conditions and the identified condition will not increase any onsito or offsite radiological conditions. Therefore, thoro will be no increased consequoncos of an accident previously evaluated in the SAR.

A 3 May the proposed activity or discovered condition create the possibility of a differont type of accident than previously evaluated in the SAR?

No. The GE evaluation has supported the operability and the structural integrity of the Jet Pumps in terms of the component's ability to mitigate the consequences of an accident, as described above. Additionally the Jet Pumps are not accident initiators and no now accident initiators will be created by operating with cracks in the Jet Pump Thermal sleeves for the speciflod operating conditions. For a change to create the possibility of an accident of a different type, the chango must allow for a now fission product release path, result in a now fission product barrier failure mode, or create a new sequence of events that results in fuel cladding failures.

Since the structuralintegrity of the Jet Pump has been assured and there are no new failures modes introduced, thoro is no possibility of a different type of accident created other than thoso currently presented in the SAR.

10CFR50.59 R: view for NCR 97-02899 Rev. O U/3 JP Thermal Sleeve Crack Page 12 of 17 B. Equipment important to Safety that is potentially adversoly impacted by this chango includes the Jet Pump assemblies, LPCl injection capability through the Jot Pump, and the components comprising the Reactor Vessel <

Internals innor volume as defined in question A 1.

B-1 May the proposed activity or disenvered condition increase the probability of occurrenc6 of a malfunction of equipment important to Safoty previously evaluated in the SAR?

No. The safety function of the Jet Pumps is the passivo function of maintaining 2/3 Core coverage, in conjunction with other VesselInternals, and to provide a flow path for LPClinjection following a design basis accident. A fracture mechanics analysis has been performed to demonstrate the structuralintegrity of the Jet Pumps for the specified operating conditions. Therefore, thoro is no degradation in the ability of the Jet Pumps to perform their intendred design function during the evaluated specified operating conditions. There is no impact on any other Reactor Vesselinternals component included in the inner volume boundary which would be affected by cracks found on the Jet Pump Thermal Sleeve. All original design and seismic requirements of the Jet Pump are still met and no additionalloads have been imposed. Postulated leakage has been evaluated and system performance of LPClis determined to be within the allowable leakage limits.

Additionally, ST-O 02F 560 3 and ST O 02F-550 3 verify the operability of the Jet Pumps by satisfying Technical Specification Surveillance's 3.4.1.1, 3.4.2.2, and 3.4.1.2, during operations greater than 25% reactor thermal power. Existing Off Normal procedure, ON 100, directs operator actions if there are operating symptoms indicative of a displaced Jet Pump Mixer. If a Jet Pump failure is confirmed the unit will be shutdown in accordance with GP-3, " Normal Plant Shutdown" por Technical Spocification requirements.

The analysis assures structural integrity and oxisting procedures will monitor safety performance and reliability of the Jet Pumps. Therefore, there is no '

increase in probability of occurrence of a malfunction of equipment Irnportant to Safety for tho specified operating conditions.

J

10CFR50.59 R: view f:r NCR 97 02899 Rev. 0 4

U/3 JP Thermal Sleeve Crack

Page 13 of 17  ;

2 B 2 May the proposed activity or discovered condition increase the j consequences of a malfunction of equipment importar.t to Safety than -i

previously evaluated in the SAR? i i  !

No. The crack sizes identified in the Jet Pump Thermal Sleeve with i i conservative crack growth assumed through the specified operating conditions are bounded by the allowable flaw size evaluation performed by  ;

i GE. Therefore, margin exists in the remaining ligament to assure structural integrity and Jet Pump operability through the specified operating i

conditions. No onsite or offsite radiological conditions assumed in the SAR l will be affected.

Since the structural integrity of the Jet Pumps is assured, there are no 1 - Increases to the consequences of a rnalfunction of equipment important to L

' Safety currently evaluated in the SAR.

I B 3 May the proposed activit,' or discovered condition create the possibility of a different type of malfunction of equipment important to Safety than any I

previously described in the SAR?

No. The GE evaluation supports the operability and the structural integrity of the Jet Pumps in terms of this equipment's (Important to Safety)' ability to mitigate the consequences of an accident, as described above. Additionally,

, the Jet Pumps are not accident initiators and no new accident initiators will  !

I be created by operating with the evaluated cracks in the Jet Pump Thermal sleeves. No new failure modes of safety related system, structures, and i

components, initiation of a new limiting transient, or new sequence of

events that can lead to a radiological release are created.  ;

Since the structuralintegrity of the Jet Pump has been assured and there

are no new failure modes introduced, there are no new or different types of malfunctions of equipment important to Safety created, other than those i currently presented in the SAR.

2 C 1 Does the proposed activity or discovered condition reduce the margin of safety as defined in the basis for any Technical Specification?

4

No. There are no specific margins associated with the structural Integrity of ,

the Jet Pumps as defined in the SAR or the Technical Specifications.

1 j

4 s .'. . , _ . ..e,.,_', .mm. .. ~ - - ~~%r--, c -- r,e,-., w,.. ., r,...,.., y ,,y .-.,...ww.m..- 3 -,c-,. ~ w,. r

10CFR50.59 R3 view f:r NCR 97 02899 Rev. 0 U/3 JP Thermal Sleeve Crack Page 14 of 17 However, the analysis described in Section ll of this document establishes the Jet Pump will maintain its structural integrity with a Safety Factor greater than 2.77. This exceeds the minimum Safety Factor of 2.25 (normal / upset conditions) applied to other Vessel Internals outilned in UFSAR Table C.5.5.

Jet Pump operability will be monitored in accordance with Technical Specification Surveillance Requirements . Continued operability will assure that Jet Pumps will be able to perform the passive safety function of maintaining 2/3 core coverage and provide a LPCI flow path post LOCA.

Leakage through the cracks during LPCI Injection is calculated to be less than 76 GPM for the inservice loop. This leakage is bounded by the allowable design leakage documented in the UFSAR Section 3.3.5.2.1. I The accuracy of the APRM flow blased setpoints are not Impacted. These setpoints do not have an associated margin of safety since they are not credited in any accident analyses.

Since the core flow measurement accuracy and uncertainty are unaffected, the Ilconsing basis for the Safety Limit MCPR is unaffected, and there is no reduction in the margin of safety as described in the SAR.

Based on the above discussion the margin of safety as defined in the basis of the Technical Specifications have not been reduced.

D-1 Does this activity as proposed involve an Unreviewed Safety Question?

No. Based on the response for Sections IV parts A through C of this Safety Evaluation, continued operation of the subject Jet Pumps with the identified cracks, is acceptable and does not constitute an Unreviewed Safety Question.

E-1 ls a change to the UFSAR necessary?

Yes. The disposition of this Safety Evaluation documents that the subject

-Jet Pumps will continue to function as described in the UFSAR. The change will revise the identified leakage from the core inner volume during LPCI injection as documented in UFSAR Section 3.3.5.2.1. Documenting the cracks found in the Jet Pump thermal sleeves is beyond the level of detail described in ihe UFSAR.

L \ ^ ^

,, e 10CFR50.59 Review f:r NCR 97 02899 R:v. 0 U/3 JP Thermal Sleeve Cree, PeDe 15 of 17 l E 2 is a change to any other SAR document necessary? No.

SAR Document Review l e Unit 3 Technical Specifications 2.0, 3.2, 3.3.1, 3.4.1, 3.4.2., 3.5.1.

• Unit 3 Core Operating Limits Report.

• Unit 3 Technical Specifications Bases B2.0, B3.2, B3.3.1, B3.4.1, B3.4.2, B'o .5.1.

• Unit 3 Technical Requirements Manual 3.10, B3.10 UFSAR Sections 1.6.2.11, 3.3, 4.2, 4.3, 4.8, 6.4, 6.5, 7.5, 7.7, 7.8, +

Chapter 14, Appendices A, C, I, J, and Figure 4.2.2.

Safety Evaluation Report by the Directorate of Licensing U. S. Atomic ,

- Energy Commission in tt.e matter of Philadelphia Electric Company Peach Bottom Atomic Power Station Units 2 and 3, August 11,1972.

Safety Evaluation Report for the General Electric Company Topical Report ,

Qualification of the One Dimensional Core Transient Model for Boiling i Water Reactors, June 1980. '

Safety Evaluatior. Report by the Office of Nuclear Reactor Regulation {

supporting Amendments Nos. 65 and 64 to Facility License No. DPR 44 and DPR 56, March 26,1980.

Safety Evaluation Report by the Office of Nuclear Reactor Regulation  !

supporting Inspection and Repair or Reactor Coolant System Piping, Recirculation Safe Ends and Core Spray Spargers, Peach Bottom Atomic l

Power Station Unit 3, March 20,1986, e Safety Evaluation Report by the Office of Nuclear Reactor Regulation supporting Amendments Nos.125 and 128 to Facility License No. DPR- t 44 and DPR 56, September 24,1987, e Safety Evaluation Report for Topical Report PECO FMS 0004, Methods of Performing BWR System Transient Analysis", November 23,1988.

V. Meferences ,

1. GE letter dated 10/29/97, " Unit 3 Jet Pump Riser Cracking Evaluations".

2. GE letter Keck to Oliver dated 10/28/97 " Jet Pump Riser Inspections".

3. -EPRI letter Selby to Hinkle dated 10/27/97

• Review of Riser VT/UT Inspections".

4. NEDC-32163P Class lil, January 1993, " Peach Bottom Atomic Power Station Units 2 cnd 3 SAFER /GESTR LOCA Loss of Coolant Accident Analysis".

10CFR50.59 R: view for NCR 97 02899 R:v. O U/3 JP Thermal Sleeve Crack Page 16 of 17 5.

BWR Jet Pump Assembly inspection and Flaw Evaluation Guldelino, i

BWRVIP 41, October 1997

6. ASME Section XI, " Rules for Inservico inspection of Nuclear Power Plant Components",1980 including addenda through Winter 1981

7. ASME Section XI, " Rules for Inservice inspection of Nuclear Power Plant Components",1989, Appendix C l 8. Original Wold installation details for Jet Pump assembly (microfilm tape PB 186).

9.

GE SIL No. 605, Revision 1, " Jet Pump Riser Pipo Cracking", February 25,1997 and BWRVIP letter dated 1/31/97 (letter no.97-139)

10. BWRVIP 28, " Assessment of BWR Jet Pump Riser Elbow to Thermal Sleeve Wold Cracking", December 1996.

11. DBD P-T-18, " Reactor Vessel Internals"

12. DBD P-T-12, " Design Basis Accidents, Transients, and Events".

13. ST 0 02F 550-3, Rev.12, " Jet Pump Operability".

14. ST O 02F 560-3, Rev. O, " Daily Jet Pump Operability".

15. ST-l 60A 220 3, Rev. 7, " Drive Flow Core Flow Correlation Check".

16. ON 100, Rev. 3, " Failure of a Jot Pump".

17. Mod 1536 CBI Contract No. 873001 Drawings 55 62, N2 Inlet Safe End Replacement

18. GE SIL 330 Supplement 2, "GE BWR/6 Jet Pump Inlet Mixer Ejection", October 27,1993

19. Specification M 733, Rev. 3, " Inservice Inspection Program".

20. A/R A1117310

C T ID 'SbidC$ RdkPkCD UCLECR C$ $ $-Y TO C073332 PROE.CO2/UeE 10CFR50.59 Review for NCR 97 02898 Rev. O urs JP The,nral sleeve crack rseejaof }s'M4'Mfl 17ee 17 A n e m vein Prepared by -

Date lo M7

( PB Design Engineering, Modifloanons)

Interface Review bQ k

( PR Component Engineering)

4. Date Interface Review b. Date 0 47

( GE Nuclear services) -

M Interface Review /58 I // X Data (4b' Manager, Reloed Analysis'3nd Methods) fi7 .

Interface Review h'[IW (CB Menag4r, Ucensing)

Date #/30!T7 ^

Peer Review I ~ Dateh/MJ

/ '

( PB D'esign Engineering, Mechanical)

Approval Date /8 '

(PB % nager, Component Engineering)

• • TOTAL PMGC.992 **

2/2*d WdtG 20 46, OE DO