ML20199J698

Rev 2 to 10CFR50.59 Review for Jet Pump Thermal Sleeve Cracking

ML20199J698
Person / Time
Site:	Peach Bottom
Issue date:	11/11/1997
From:	PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC
To:
Shared Package
ML20199H840	List: ML20199H834 ML20199J698
References
NUDOCS 9711280203
Download: ML20199J698 (17)

Text

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ORIGINAL Y. HEN SW#ED IN lied Peach Bottom Atomic Power Station Unit 3 10 CFR 50.59 Review for Jet Pump Thermal Sleeve Cracking Revision 2

1. Subject Revision 2 of this 10 CFR 50.59 Review was performed to redefine the "specified operating condition". This revision will allow the flexibility to operate at two different Reactor Coolant Recirculation drive flows for specified periods of time and stay within the bounds of the GENE onalysis.

This change also affects the Reactor Coolant Recirculation system flow leakage rates and the predicted length of the crack at the end of the specUled operating period. All other aspects of the 50.59 Review are valid and unchanged.

Revision 1 of this 10 CFR 50.59 Review was required to increase the postulated flow leakage values which were incorrectly presented in Revision 0, due to a computational error found during NCR 97-02899 reviews. All other aspects of the Revision 0 version of this 50.59 Review are valid and unchanged. Post LOCA LPCI leakage is within the allowed value identified in the SAR.

During Peach Bottom Unit 3 Refue'ing 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 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 30 dog. 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.

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

for continued operation within the Reactor Coolant Recirculation drive flow and time constraints evaluated by GENE.

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10CFR50.59 Roview for NCR 97 03899 Rov.1 U/3 JP Thermal Sleeve Crack Page 2 of 17 The INTERIM USE AS IS disposition allows for administratively controlling (Ref. 20) Reactor Coolant Recirculation drive flow to a NOMINAL value of up to 15.75 Mlbm/hr for each recirculation loop for a period of up to 800 hours0.00926 days <br />0.222 hours <br />0.00132 weeks <br />3.044e-4 months <br /> and up to 13.85 Mlbm/hr for each recirculation loop for a porlod of up to 2,224 hours0.00259 days <br />0.0622 hours <br />3.703704e-4 weeks <br />8.5232e-5 months <br />. Operations at those flow ratos may happen at any timo (i.e.

raiso and lower reactor power) as long as the total hours are within the specified limits. This war solocted as an operating strategy, hereinafter known as the "specified operating condition".

Transients, outsido the speciflod operating condition, such as single loop operation or excursions abovo nominal valuos are bounded by the analysis (Ref.1, 21, 22). Extended single loop operation greator than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> will be ovaluated by engineering for impact on the specified operating conditions.

The result of operating at highor Reactor Coolant Rocirculation drive flows for extended periods of timo could reduce the operating pericd.

Additionally, operating at lower Reactor Coolant Rocirculation drivo flows could extend the operating period. This will requiro ovaluation by engincoring for impact on the specified operating conditions to assure compliance with the drivo flow /timo constraints that has been ovaluated by GENE. Drivo flow will be monitored and tracked administratively by using plant proceduros.

11. Discussion det Pumo Confiauration The Jot Pumps aro Reactor VesselInternals and in conjunction with the Reactor Coolant Recirculation system are designed to provido forced circulation to the core for heat removal from the fuel. The Jet Pumps are located in tho annulus region between the coro shroud and the vossol wall.

Since the Jot Pump suction olevation is at 2/3 coro holght, the reactor core will romain covered to this height even with a comploto break of the Rocirculation piping as assumed in the design basis accident (DBA). During post LOCA LPCI operation, the Residual Hoat Removal system pumps tako suction from the suppression pool and discharge into the core region of the reactor vossol through the recirculation loops (i.e. through the Jot Pumps i

into the coro region). LPCI helps to rostoro and maintain the coolant J

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10CFR50.59 Revi;w for NCR 97-02899 Rev.1 U/3 JP Thermal Sleeve Crack Page 3 of 17 In the reactor vessel such that the core is adequately coolod to preclude fuel clad temperaturo in excess of 2,200 dog. F following a design basis LOCA (Ref. 4).

Each Reactor Coolant Rocirculation loop contains ton Jot Pumps.

Recirculated coolant passes down the annulus betwoon the Reactor Vossal wall and the Coro Shroud. Approximately one third of the coolant flows from the vessel, through the two external recirculation loops, and becornos the driving flow for the Jet Pumps. Each of the two external recirculation loops dischargo high pressure flow into an external manifold from which individual recirculation inlet linos are muted to the Jot Pump risors within the Reactor Vossol. The remaining portion of the coolant mixturo in the annulus becomes the suction flow for the Jot Pumps. This flow ontors the Jot Pumps at suction inlets and is accolorated by tho drive flow. The drive flow and the suction flow are mixed in the Jet Pump throat section. The total flow then passos through the Jet Pump diffuser section into the area below the core (lower plonumb gaining sufficient hoad in the process to drive the required flow upward through the core.

The recirculation inlot nozzlo thermal sloove is wolded to the nozzio safo end at its outor extremity and to the jot pump riser elbow at its Inner extremity.

The thermal sloovo is designed to provido a pressure retalning flow path for Reactor Coolant Rocirculation drive flow to the Jot Pumps. Secondarily, the thermal sloovo reduces temperaturo variations, and thus thermal loading, on the recirculation inlot nozzlo. The thermal sloovo is not a primary pressure bouridary.

The thormal sloovo is 10" schedulo 40 stainless stool type 304 pipe. The thermal sleevo to riser olbow joint is a field wold, performed during Jot Pump installation into the Reactor Vossol. Tho welding process was gas-tungston arc with typo 308 fillor matorial. During wold preparation, the thermal sloove was countor bored for appropriato fit up to the schedulo 30 riser elbow. The wolds are non flux, non croviced, full penetration butt wolds (Ref. 8 & 17).

Crack Descriotion/ Geometry Based on a review by an export metallurgist from PECON Testing and i

Laboratorios, visual examination of all the Jet Pump indications are characteristic of Intergranular Stress Corrosion Crackinn (IGSCC) in the heat affected zone of the austenitic stainless stoel circumferential pipe wold. The cracking is away from the too of the weld (approximately 1/8" to 1/4") and j

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t 10t.FR50.59 R2 view for NCR 97 02899 Rev.1 U/3 JP Thermal Sleeve Crack Page 4 of 17 jagged in appearanco. The crack ends were intormittent and at the same relativo distanco from thu too of tho wold. No indication of fatiguo crack growth was observed in that the crack tips did not turn and follow the too of the wold, the cracks woro jagged and not straight lined, and no crushing of the crack faces was observed. However, boat samplos were not obtainod of the crack tips to rule out the possibility of fatiguo cracking.

The initial visual examinations woro porformed using modified VT 1 (1 mil wiro) standards. Supplomontal ultrasonic oxaminations (UT) woro performed at the crack locations cnd the results are listed below.

Crack on Jot Pumps 1 and 2 Risor The thormal sloovo to elbow wold has a crack from 329.5 dog. through 84.6 dog., looking in the direction of flow. This corresponds to a longth of 10.8 + / 0.39 inchos for uncertainty, consistent with BWRVIP protocol.

The flaw is on the Thermal Slowe side of the wold.

Crack on Jet Pumps 9 and 10 nsor The thormal sloovo to olbow wold has a crack from 12.1 dog, through 30.2 dog., looking in tho direction of flow. This corresponds to a longth of 1.7 +/ 0.34 inchos for uncertainty, consistent with BWRVIP protocol. The flaw is on the Thormal Sloovo sido of the wold.

Crack on Jot Pumps 13 and 14 riser The thormal sloove to olbow wold has a crack from 305.5 dog. through 81.0 dog., looking in the direction of flow. This corresponds to a longth of 12.7 + / 0.34 inches for uncertainty, censistant with BWRVIP protocol.

The flaw is on the Thormal Sloovo sido of the wold.

Root Cause IGSCC is considered to be the most likely initiator of this cracking. The cracking on the Thormal Sloovo is similar to cracking identified to dato in other Roactor Vossol Internals. Although not a croviced joint, stainless stool typo 304 matorials used for the thermal sloove, in conjunction with past poor water chemistry conditions havo made the joints susceptible. Records also ind!cate the possibility of those being cold sprung during installation thus increasing the residual stresses in the area and increasing the joints susceptibility to IGSCC.

Code Boundarv

10CFRSO.59 R3 view for NCR 97 02899 Rtv.1 U/3 JP Thermal Sleeve Crack Page 5 of 17 Jet Pump components are not part of the primary pressure boundary and do not provido a core support function. Jet Pumps are Safety Related and are optionally classified as ASME Section XI components for inspection purposes only. The Jot Pumps provido a Safety Related flow path during LPCI injection.

Flaw Evaluall0D The determination of structuralintegrity was performed by using standard accepted methods for intergranular stress corrosion cracking and fatiguo.

Although examination of the crack indicates that IGSCC is the solo contributor. fatigue loading was also considered in developing allowable flaw sizos. The sourco for fatiguo crack growth was dotermined by analytical metho:Is to be low amplitudo-high frequency vibration from the high voicelty rocirculation lino flow.

The allowable flaw size at the elbow to thormal sloovo location was determinod using standard limit load methodology presented in BWRVIP-41, "BWR Jot Pump Assembly inspection and Flaw Evaluation Guidelines" (Ref. 5). 3!.Tilar methods have been previously used to evaluato other Vessel components such as the coro spray lines and shroud. The flaw ovaluation methadology used was performed consistent with ASME Section XI, Appendix C requirements (Ref. 7). This ovaluation includes the ASME Section Xi Cafety Factors of 2.77 for Normal and Upset and 1.39 for Ernorgoney and Faulted conditions. Load combinations are in accordance with the UFSAR and BWRVIP-41.

Once the allowablo flaw size was determined, the accer tability of an observed flaw was datormined by performing a crack nrowth analysis. This analysis considered both IGSCC and fatigue loading. The IGSCC growth was predicted using the conservativo standard of 5 x 10in/hr crack growth rato from each crack tip. This growth is the accepted bounding industry standard for IGSCC in austenitic stainless stoele in a BWR onvironment with normal water chemistry. This !s expected to be conservativo sinco the Thermal Slaevo to albow is a non croviced wold and PBAPS injects hydrogon into foodwater at a rato equating to 0.3 ppm. The actual growth rate is expected to be on the order of 2.5 x 10 in/hr.

Each crack was then ovaluated against its susceptibility to fatiguo cracking.

Fatiguo cracking in the riser piping is primarily a result of flow induced vibration caused by the recirculation drivo flow. A time history of stress amplitudo vs. timo for the Jet Pump risors was obtained using baselino I

10CFR50.59 R: view fer NCR 97 02899 Rev.1 U/3 JP Thermal Sleeve Crack Page 6 of 17 testing of a BWR4/251" dia. Reactor Vossol (Browns Ferry Unit 1). During start-up testing at Browns Ferry, strain measuromonts on the Jet Pump riser braces woro obtained at varying power lovels and flow conditions.

Measutomonts at the risor brace woro scaled to the riser crack location by means of modal shapo factors, datormined analytically. Data corresponding to 100% coro flow at 100% power woro used to ovaluate the influence of fatiguo cracking on the subject risors.

Results of this analysis concluded the N2A riser cracking is small enough that the crack growth rate will not bo hfluenced by f atiguo cracking through the next 2 year cyclo of full power operation (AK is loss than AK threshold).

Thoroforo, crack growth is limited to IGSCC and crack sizo will be limited to 3.7 inches by the end of the 2 year cyclo and is acceptable to use-as-is.

For the N2E and the N2J risors, the stress intensity rango for the assumed loading oxceeds the threshold for susceptibility for fatiguo cracking (AK is greator than AK threshold). When applying fatiguo crack growth to both thermal sloovo cracks, the longths would excood the limit load allowable flaw sizo by and of cycle.

To mitigato the impact of flow induced vibration on the N2E and N2J Thermal Sloovo cracks, recirculation drive flow will be limited to the specified operating conditions. The prod lcted and of operating condition flaw sizes Jro listed balow.

Peach Bottom Unit 3 Flaw Eva!uation Summarv Location Current Longth*

Predictod *

• Allowable Porcent of (in.)

Longth Flaw Longth Allowablo (in.)

(in.)

Flaw Longth JP 1/ 2 11.2 12.6 17.9 70.4 %

JP 9 /10 2.1 3.7 17.9 20.7 %

JP 13 /14 13.1 14.9 17.9 83.2 %

Length was used in GE analysis and includes UT uncertainty, referenco 1, 2,3 and 21 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.

Loakago Evaluation

160FR50.59 Review for NCR 97 02899 Rev.1 U/3 JF Thermal Steen Crack Page 7 of 17 Duo to the small crack opening area, any leakago through the cracks will be minimal. Postulated loakago will be approximately 345 GPM por loop for Reactor Ccolant Rocirc station flow at 15.75 Mlbm/hr loop drive flow and approximauy 150 GPM for the inservice LPCI flows post LOCA. This assumos the crack grows to the prodleted flaw longth. Thoro is no specified allowablo design leakage limit for the Reactor Coolant Rocirculation flows and the postulated leakago is negligible when compared to system flows.

The original design allowablo leakago of 3000 GPM, for Low Pressuro Coolant injection (LPCI), will not be exceeded. Thoroforo, crack leakago during operations and post accident, for the spec;fied operating conditions, will not impact any ECCS/LOCA analysis (Raf. 4).

Ill. Determination 1.

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

Licenso amendment?

No. Fracturo Mechanics analysis of the cracks and ovaluation of potential leakago of Rocirculation coolant or LPCI (post LOCA) flow into the annulus region of the Roactor Pressuro Vessel (RPV) has confirmed the operability of the subject Jot Pumps, Reactor Coolant Rocirculation system and the Low Prossuro Coolant Injection (LPCI) modo of the Residual Heat Removal system for the specified operating conditions. This analysis does not nocessitato a chango to surveillance requirements or limiting conditions of operation of the Jot Pumps, the Roactur Coolant Rocir::ulation system or the LPCI mode of Residual Heat Removal (RHR) system due to the specified operating conditions on Roactor Coolant Rocirculation pump flow. Therefore, the continued operation of the Jet Pumps, the Reactor Coolant Rocirculation system and the LPCI modo of Residual Heat Removal system as is does not requiro a Technical Specification change or any Operating Licenso amendment.

2.

Does the activity or discovered condition make changes to the facility as described in the SAR?

Yes. Continued operation of the subject Jot Pumps with cracking as described above is considered a chango to the facility as deswibed 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

10CFR50,59 Review for NCR 97 02899 Rev.1-U/3 JP Thermal Sleeve Crack

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cracking in the original Jet Pump design. Although the subjsct Jet Pumps

- are outside of tN 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.:

2 Potential leakage paths from the floodable inner volume of the Reactor Vessel (e.g.- 2/3 com height) during a Recirculation system pipe break and -

subsertuent LPCI reflooding is documented in the SAR. Postulated leakage from tht. Jet Pump cracks during this condition has beeii calculated to be 4

appasrstely 150 GPM for the inservice LPCI loop. This additional leAege is well w.Wa the 3000 GPM allowance designed in the LPCI subsystem for

. potentla! leskoge paths but will be considered a change to the facility as described in the SAR.. Additionally, two loops of LPCI flow through one 4

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-Reacto Coolant Recirculation loop is less than tho specifMd Recirculation flow limits evaluated for the specified operating condition The associated piping stresses are therefore bounded by evaluated Reactor Coolant l

Racircula6on system operation.

Another potential leakage path from the Jet Pump cracks, during operations, is inside the Reactor Vessel pressuse boundary and would not have an unacceptable effect on the system performance of the Reactor Coolant Recirculatlan system. A camputation was performed and has determined

_that potential leakage through the cracks is insignificant when compered to normal system flow through the_risar piping. Since the leakage flow has been detennined to be insignificant and contained within the Reactor Vessel pressure boundary this leakage is not considered to be a change to the facility as described in the SAR.

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

No. Jet Pump operability is verified daily per Technical Spt cification requirements.. Jet Pump dP measurements arn used to determine operabi'ity and to calculate core flow and are unaffected by cracks on the Jat 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 ;iased portions of the APRM and y

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10CFR50.59 Review for NCR 97-02899 Rev.1

- U/3 JP Thermal Sleeve Crack :

Page 9 of 17 Rod Block functions are not credited m the core reload licensing analysis.

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Thus the core reload licensing analysis is unaffected.

The flow signal'used by the APRM system to establish flow biased rod block -

I and scram. trip setpoints is_ derived from the drive flow transmitters tapped off of the recirculation pipe venturis.: The flow value is processed by the 1

APRM flow units prior to use by the APRM system. Per Technical-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 axist due to the Jet Pump 4

i Thermal Sloeve cracks will not impact the accuracy.of the APRM flow biased setpoints since the flow signal is gained to correlate to core flow.

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The procedure that implements this surveillance is ST-l 60A-220 3, Drive Flow / Core Flow Correlation Check, in addition to this surveillance, 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?

4, Does the activity or discovered condition involve tests or experiments not described in the SAR?

No. Continued operation of the Jet Pumps, Reactor Coolant Recirculation system and the LPCI mode of RHR with cracks in the Jet Pump Thermal Cbeves does not involve any-tests or experiments not described in the SAR.

When applying accepted crack growth rates for the specified operating conditlans to the flaw sizes identified on the -Jet Pump thermal sleeves the j.

' flaw size is bounded by_ the limit load _ailowable flaw size summarized in reference 1. Therefore, margin exists in the remaining thermal sleeve ligaments tu ascure structural integrity and systems operability during the

'specified operating conditions interval. There are no additional tests or-experiments involving plant 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.

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10CFR50.59 R;vbw for NCR 97-02899 Rev.1 U/3 JP Thermal Sleeve Crack Page 10 of 17 IV. ' Safety Evaluation A. Those accidents potentially negatively impacted by this change include those accidents requiring an inner volume containing the core (e.g. 2/3 core height) that can be flooded following 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 Rocirculation 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?

No. The safety design basis of a Jet Pump assembly is to provide o portion of the floodaole inner volume containing the core. LPCI reflooding of the core, post-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. Tiic current condition of the Jet Pumps is not a new accident initiator. GE's review of all postulated load combinations on the Jet Pumps has determined that load combinations including the design basis accident LOCA loads are bounding for all normal, derated, 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 Nozzles down to the Shroud rupport.

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 identified cracks are not part of the inner volume A 'racture mechanics evaluation at the specified operating conditions, using the crack lengths verified bv the :JT data (Ref. 2 & 3) and applying DBA loads, has validated the continued structuralintegrity of the Jet Pump assemblies for all pos'ulated plant conditions. Therefore, there is no increase in the probability of occurrence of an accident previously evaluated in the SAR tor the specified operating conditiont

10CFR50.59 '"..a for NCR 97-02899 Rev 1 U/3 JP Thermal Sleeve Crack Page 11 of 17 A-2 May the proposed activity or discovered condition increase the consequences of an accident previously evaluated in the SAR?

hlo. The consequences of an accident previously evaluated in the SAR have not been increestd 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 references 1 and 21. The safet/

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 LPCIinjection 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) as defined in UFSAR Section 14.6.3. Therefore, margin exists in the remaining thermal sleeve ligament to ensure structural integrity and Jet Pump operability through the specified operating conditions. No safety limit will be impacted and no bcrrier design limits are cornpromised.

Due to the small total area open to flow at the crack locations, any leakage 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 specified operating conditions, is calculated to be approximately 150 GPM for the inservice loop. This leakage is well within the allowable design leakage documented in the SAR for the LPCI mode of operation.

Since the Jet Pump structural integrity is assured ano any additional leakage after LPCI reflooding is within existing system margins the existing accident analysis and assumptions are unchanged end valid for the specified operating conditions and the identified condition will not increase any onsite or offsite radiological cc iditions. Therefore, there will be no increased consequences of an accident previously evaluated i.1 the SAR.

A-3 May the proposed activity or discovered condit:on create the possibility of a different 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 new accident initiators will be

l 10CFR50.59 R:vtw for NCR 9T 02899 Rev.1 Ul3 JP Thermal Sleeve Crack Page 12 of 17 created by operating with cracks in the Jet Pump Thermal sleeves for the specified operating conditions. For a change to create the possibility of en accident of a different type, the change must ellow for a new fissior roduct release path, result in a new fission product barrier failure mode, or c. ate 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, there is no possibility of a differerst type of accident created other than those currently presented in the SAR.

B. Equipment Important to Safety that is potentially advsrsely impacted by thie change includes the Jet Pump assemblies, LPCIinjection capability through the Jet Pump, and the components comprising the Reactor Vessel Internals :aner volume as defined in o,uestion A-1.

B-1 May the proposed activity or discovered condition increase the probability of occurrence of a malfunction of equipment important to Safety previously evaluated in the SAR7 No. The 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 LPCIinjection following a design basis accident. A fracture mechanics analysis has been performed to demonstrate the structural integrity of the Jet Pumps for the specified operating conditions. Therefora, there is no degradation in the ability of the Jet Pumps to perform their intended design function during the evaluated specifit operating conditions. There is no impact on any other Reactor Vessel internals 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 additior.alloads have been imposed. Postulated leakage has been evaluated and system performance of LPCIis determined to be within the ellowable leakage !imits.

Additionally, ST-0-02F-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 400, 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

10CFR50.59 R2 view for NCR 97-02899 Rev.1 U/3 JP Thermal Sleeve Crack Ptae 12 of 17 4

GP 3, " Normal Plant Shutdown" per Technical Specification requirements.

The analysis assures structural integrity and existing 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 important to Safety for the specified operating conditions.

B-2 May the proposed activity or discovered condition increase the com.equences of a malfunction of equipment important to Safety than previously evaluated in the SAR?

No. The crack sizes identified in the Jet Pump Thermal Sleeve with conservative crack growth assumed through the specified operating conditions are bounded by the allowable flaw size evaluation performed by GE. Therefore, margin exists in the remaining ligament to assure structural integrity and Jet Pump operability through the specified operating conditions. No onsite or offsite radiological cond;tions assumed in the SAR will be affected.

Since the structural integrity of the Jet Pumps is assured, there are no increases to the consequences of a malfunction of equipment important to Safety currently evaluated in the SAR.

B-3 May the proposed activity or discovered condition create the possibility of a different type of malfunction of equipment important to Safety than any previously described in the SAR?

No. The GE evaluation supports the operability and the structuralintegrity of the Jet Pumps in terms of this equipment's (Important to Safety) aoility to mitigate the consequences of an accident, as described above. Additionally, the Jet Pumps are not accident initiators and no new accident initiators will be created by operatirig with the evaluated cracks in the Jet Pumo Thermal sleeves. No new failure modes of satt.:v related system, structures, and components, initiation of a new limiting transient, or new sequence cf events that can lead to a radic!ogical release are created.

Since the structural integrity of the Jet Pump has been assured and there are no new failure modes introducM, there are no new or different typos of malfunctions of equipment important to Safety created, other than those currently presented in the SAR.

l 10CFRSO.59 R;vicw for NCR 97-02899 Rev.1 U/3 JP Thermal Sleeve Crack Page 14 of 17 C-1 Does the proposed activity or discovered condition reduce the margin of safety as defined in the basis for any Technical Specification?

No. There are no specific margins associated with the structural integrity of the Jet Pumps as defined in the SAR or the Technical Specifications.

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 outlined in UFSAR Table C.5.5.

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

Leakage through the cracks during LPCI injection is calculated to be approximately 150 GPM for the inservice loop. This leakage is bounded by the allowable design leakage documented in the UFSAR Sectim 3.3.5.2.1.

The accuracy of the APRM flow-biased 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 licensing 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 nave 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.

10CFR50.59 Rr, view for

- NCR 97-02899 Rev 1

~

U/3 JP Tnermal Sleeve Crack -

Page 15 of 17 -

t-E-1'Is 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 i, 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 r

described in the UFSAR.

E 2 is a change to any other SAR document necessary?

No.

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

e Unit 3 Core Operating Limits Report.

Unit 3 Technical Specifications Bases B2.0, B3.2, B3.3.1, B3.4.1, e-B3.4.2, B3.5.1.

Unit 3 Technical Requirements Manual 3.10,83.10 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 the matter of Philadelphia Electric Company Peach Bottom Atomic Power Station Units 2 and 3, August 11,1972.

' Safety Evaluation Noort for the General Electric Company Topical Report e

Qualification of the One Dimensional Core Transient Model for Boiling _

Water Reactors, June 1980.

Safety Evaluation 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 Power Station Unit 3, March 20,1986.

Safety Evaluation Report by the Office of Nuclear Reactor Regulation supporting Amendments Nos.125 and 128 to Facility License No. DPR-44 and DPR 56,-Septamber 24,.1987.

~ Safety Evaluation Report for Topical Report PECO FMS-0004, Methods of Srforming BWR System Transient Analysis", November 23,1988.

~

'10CFR50.59 Riview for NCR 97-02899 Rev.1

-U/3 JP Thermal Sleeve Crack Page 16 of:17:

v.-

J?

[VF References 1 1.--

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

Evaluations". -

2.1

' 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 Inspect'ons".

A.

NEDC-32163P C!ars lil, January 1993, " Peach Bottom Atomic Power Station Units 2 and 3 SAFER /GESTR-LOCA Loss of Coolant Accident Analysis".

, : 5..

- BWR Jet Pump Assembly inspection and Flaw Evaluation Guideline, BWRVIP-41, October 1997 j

16.

TASME Section XI, " Rules for Inservice Inspection of Nuclear Power 1

- Plant Components",1980 including _ addenda through Winter 1981

7. -

- ASME Section XI,1* Rules for Inservice inspection of Nuclear Power Plant Components",1989, Appendix C.

18.

Original _ Weld Installation details for Jet Pump assembly (microfilm tape PB-166).

t9.

GE SIL No. 605, Revision 1, " Jet Pump Riser Pipe 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 r

Sleeve Wald Cracking",. December 1996.

1 1.-

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

12. -

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

13.

ST-O 02F-550-3, Rev.12,'" Jet Pump Operability".

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

15.-

ST-l-60A-220-5. Rev. 7, " Drive Flow Core Flow Correlation Check".

16.

ON-100, Rev. 3, " Failure of a Jet Pump"..

17.

Mod 1536 CBI Contract No.- 873001 Drawings 55 62, N2 Inlet Safe End Replacement (M-1-E-446) l

- 18.

GE SIL 330 Supplement 2 "GE BWR/6 Jet Pump Inlet Mixer Ejection", October 27,11093 19.

Specification M-733, Rev. 3, " Inservice inspection Program".

20. ; A/R A1.117310, " Implementation of Unit 3 Recirculation limits"

?21'. : GE letter dated 11/10/97, " Unit 3 Jet Pump Riser Cracking Evaluation 1 n

_ of_ Alternative Operation '.

L 22.

GE letter dated-11/11/97, " Jet Pump Riser Elbow Crack Single Loop-cOperation Evaluation".'

4

.w-2 4

m.,,.

c

_%,4

t 10CFR50.59 feview for k

^

f-NCR 97-02899 Rev.1 -

U/3 JP Thermal Sleeve Crack Page 17 of 17 -

Anorovals

_) -.

Date Ufloil Prepared by

( PB Design Engineering, Modifications)

- Peer Review _

O DN (. $

F.tu.1]L $]

( PB Compenent Engineering)

Approval W

Date II/IIlq7 (PB@anager, Component Engineering) f C

.