ML18064A739

Provides Justification for Continued Operation for Repaired Pressurizer Instrument Nozzles

ML18064A739
Person / Time
Site:	Palisades
Issue date:	04/28/1995
From:	Haas K CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.)
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NUDOCS 9505080091
Download: ML18064A739 (53)

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consumers Power l'OWERINli MICHlliAN"S l'IUlliRESS Palisades Nuclear Plant,:. 27780 Blue Star Memorial Highway, Covert, Ml 49043 April 28, 1995 Nuclear Regulatory Commission ATTN:

Document Control Desk Washington, DC 2055.5 KurtM.Haas Plant Safety and Licensing Director DOCKET 50-255 - LICENSE DPR PALISADES PLANT - JUSTIFICATION FOR CONTINUED OPERATION FOR THE REPAIRED PRESSURIZER INSTRUMENT NOZZLES On September 16, 1993, while heating UP after a refueling outage, a leak was discovered in one of the pressurizer power operated relief valve (PORV) lines near the pressurizer nozzle.

The leak was later determined to be in the weld joining the lnconel 600 safe end to stainless steel pipe.

The plant was returned to cold shutdown and the leak repaired.

On October 9, 1993, while installing insulation on the upper head of the pressurizer that had b~en removed to repair the leaking PORV line, a leak was discovered at the base of temperature element TE-0101 nozzle on the pressurizer.

On October 12, 1993, a visual inspection of the other pressurizer temperattire nozzles identified that one other temperature element noizle~ TE-0102 also showed evid~nc~ of leakage.

Both temperature el~ment nozzles were repaired by modifying the nozzles using an external weld pad as the repair technique.

As a result of these events *and repairs, information and analysis was sent to the NRC which provided justification for operation of the Palisades Plant for one cycle of operation, with the pressurizer nozzle and temperature elements in the repaired condition.

On January le. 1994, the NRC responded with a Safety Evaluation regarding the identified_:~l'.'.~~~i_ng in the Alloy 600 (lnconel) components at the plant. The SER stated that the NRC Staff concluded based on the repairs performed,*

ev.aluati.ons and inspections performed by the licensee, existing leak rate monitoring requirements and capabilities, and the fact that a failure of ~ne of the subject lines is bounded by the small break LOCA described in the Final Safety Analysis report that; 1) the repair welds in the PORV line and p.ressurizer instrument nozzles were acceptable for operation for one fuel cycle, and 2) that the pressurizer surge and spray lines were acceptable for one cycle of operation.

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2 The purpose of this letter is to provide justification for continued operation of the Palisades Plant for the life of the plant' with the two pressurizer temperature element nozzle repairs performed in 1993 and to address the acceptability for operation of the plant, related to the condition of the RORV~ Spray and Surge lines. The justifi~ation for continued operation and supporting documentation for the Temperature Elements, PORV, Spray and Surge lines is contaJned in Attachment 1 to this letter.

In summary, Attachment 1 contains the following information:

- The repaired PORV nozzle will be replaced during the next refueling outage and therefore no justification for operation beyond the one cycle granted in the subject SER is required.

- The pressurizer surge and spray nozzle safe e.nd to pipe welds have been the subject of additional review and analysis. Detailed Palisades specific fracture mechanics analysis was performed to establish times to failure and allowable flaw sizes for these safe ends.

It has been determtned that the spray nozzle safe end operating at 540°F is less susceptible to PWSCC than the surge nozzle safe end.

Both the spray and surge nozzl~ safe end-to-pipe welds wtll be inspected during the 1995 refueling outage with enhanced ultrasonic examination methods demonstrated to accurately detect and size flaws.

Results from the inspection will augment the results of the fracture mechanics analysis to provide justification for continued operation of these components.

Additionally, Mechanical Stress Improvement Process (MSIP) will be applied to the pressurizer and hot leg surge nozzle safe ends to eliminate the potential for future PWSCC initiation or propagation.

- Justification is provided for continued operation of the plant with the pressurizer temperature element nozzles repairs performed in 1993.

Analysis has been c.omp 1 eted to show t_hat potent i a 1 corrosion of the carbon steel pressurizer vessel will be minimal and not affe~t operation fo.r the life of the p 1 ant.

The NRC SER, along with detailing the work completed by the Licensee and NRC during the review of these events, expressed some concerns over the licensee's engineering analysis that determined the predicted life of existing welds in the pl ant prima.ry cool ant system.

Although not requested as a response to the SER, Attachment 2 cont a i.ns an update o.n our progress in this area and a response to each of the concerns raised in the SER.

The Palisades refueling outage is presently. scheduled to begin at the end of May and continue until the end of August 1995.

NRC approval of the justification for continued operation is requested prior to resumption of plant power operation.

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3

SUMMARY

OtCOMMITMENTS This lett~r contains no new commitments.

Kurt M Haas Plant Safety and Li~ensing Director CC Administrator, Region 111; USNRC

• NRC Resident Inspector - Palisades Attachments

ATTACHMENT I Consumers Power Company Palisades Plant Docket 50:-255 JUSTIFICATION FOR CONTINUED OPERATION FOR THE REPAIRED PRESSURIZER INSTRUMENT NOZZLES April 28, 1995 4 Pages

JUSTIFICATION FOR CONTINUED OPERATION FOR THE REPAIRED PRESSURIZER INSTRUMENT NOZZLES 1 - PORV SAFE END LEAK One of the commitments we made as a foll ow-up to the repair of the PORV safe end leak and as reiterated in the NRC's January 18, 1994 SER follows:*

The design of the pressurizer relief valve nozzle safe end and PORV line wi77 be reviewed and appropriate modifications wi77 be made during the next refueling ouiage to ensure a suitable lifetime for the pressurizer relief valve nozzles ~afe end.

This review wi71 address the material properties of the safe end and stresses imposed on the safe end by the PORV line.

The review also wi77 be coordinated with the safety-related piping verification project review of the PORV line.

CPCo Resolution The cause of the PORv* nozzle safe end failure in 1993 was evaluated by Brookhaven National Laboratory and by CPCo metallurgists.

Several factors combined to render the originally installed PORV safe end the most susceptible Palisades Alloy 600 PCS component for primary water stress corrosion cracking {PWSCC).

1 Our review determined that; {l) the leaking PORV Inconel Alloy 600 safe end had an unusually large grain size, high hardness and high yield strength.

These material characteristics are a possible result of overheating during forging of th1? material.

{2) The PORV nozzle safe end is located on the top of the pressurizer where the operating temperature is approximately 640°F which is the highest predicted temperature within the PCS.

By contrast the reactor vessel head, hot leg and cold leg temperatures are approximately 590-5950F, 585°F, and 535°F respectively.

{3) Investigations showed that the original PORV safe end to pipe weld had been fieldrepaired from the inside.

The weld root had been ground out and rewelded with no post-weld stress relief being performed. This created very high resid~al stresses on the interior surface.

{4) Leaving the repaired nozzle in place with the conditions identified above, may not have eliminated the potential for a circumferential crack reforming which could pose a undue safety concern.

As a result of Qur review we determined that the best long term solution for plant operation with regards to the PORV Alloy 600 safe end was to replace the safe end. The Alloy 600 safe end will be replaced with. a new Type 31J5 stainless steel tr~nsitton piece which is less susceptible to PWSCC.

This replacement of the existing PORV nozzle safe end will be welded to the stainless steel cladded carbon steel nozzle with Alloy 690 weld material, thus

• eliminating the potential for PWSCC at this. location.

Our review also determined that the existing stresses imposed on the existing PORV nozzle safe end to piping are within allowables.

The safety related piping reverification project has also completed review of the PORV line and determined that no.undue stresses are applied on the safe end from the PORV line.

As such no modifications to the PORV line are needed or planned to ensure the lorig term safe operation of the existing design.

Conclusion The replacement of the Inconel 600 PORV nozzle safe end will resolve the immediate operability questions regarding this pressurizer penetration. Results of our analysis show that no modifications need to be made to the PORV downstream piping.

2 - PRESSURIZER TEMPERATURE ELEMENT NOZZLE LEAKS The January 18, 1994 NRC SER determined that Pressurizer temperature element nozzle {TE-0101 andTE-0102) modifications are acceptable for one cycle of operation.

CPCo Resolution

Background

In October of 1993, an axial through wall crack was detected on two temperature element nozzles (TE-0101 and TE-0102) on the pressurizer; A pad weld repair was performed.

At that time justification was provided to permit Palisades to oper~te safely for one fuel cycle, during which time further evaluation of the situation could:occur.

Justifi~ation for Continued Operation J.ustifi.cations for continued operation beyond one cycle of operation are summarized as the following:

i)

The 1993 weld pad modification met ASME Section III for Class I 2

component fatigue.

(Ref. EA-A600-0l, 11 Service Life Analysis of Pressurizer temperature Element TE-0101 Weld Modification," Attachment 3

• to this lette.r; EA-A600-02, "Service Life Analysis of Pressurizer temperatur~ Element TE-0102 Weld Modification", Attachment 4 to this letter~*

3 ii)

CPCo and B&W Nuclear Technologies (BWNT) analyses have determined that pressurizer base metal exposed to primary coolant at existing cracks will cu1*rode at a very slow rate (EA-AG00-15, "Justification for Continued Operation of Pre$surizer TE-0101 and TE-0102 Nozzles as repa.ired during the 1993 Refueling Outage" - Attachment 5). Hence, it is Judged to be acceptable for the remaining plant life.. Industry experience and recent base-metal examinations at Entergy Operations (AN0-1) ~nd Southern California Edison (San Onofre) along with other utilities inspections of instrumentation nozzles support this conclusion. Therefore, further base-metal corrosion inspection of the repairs to pressurizer temperature element nozzles for TE-0101 and TE-0102 is not.necessary.

iii) Potential failure modes of the existing TE-0101 and TE-0102 nozzles were investigated in EA-AG00-15 (Attachment 5).

The original J-groove weld design of the temperature element nozzle was compared with the pad weld repair performed in 1993. Also, the susceptibility of the pad weld and the TE nozzle material was compared.

From these comparisons, it was concluded that:

iv)

The TE nozzle material at the weld pad is less susceptible to P\\-JSCC than that near the original J-weld.

Hence it is expected that the time for an ID initiated PWSCC to develop leaka~~ near the weld pad wil 1 be 1 onger than the ti me it took to deve 1 op the axial cracks found in the 1993 outage.

An axial though-wall crack in the nozzle body is more likely to occur before a crack through the interface between the nozzle body and the wel~ pad.

Possible axially oriented PWSCC in the nozzle body is not a safety concern as it only results in small leakage.

Such leakage through axial cracks at the TE nozzle body can be easily detected by visual inspection well before failure.

As stated in the Alloy 600 Project Plan, a visual inspection to identify leakage of the TE-0101 and TE-0102 nozzles will be performed during the 1995 refueling outage. Also as noted in the plan, the periodicity of future inspections of the TE-0101 and TE-0102 nozzle~ will be determined following review of the dah obtained as a result of the refueling outage inspections.

Conclusion Based on industry experience and analysis of corrosion of the carbon steel pressurizer shell there is no structural or safety concern for TE-0101 and TE-0102 nozzles.

If a leak does develop it will be easily identified by visual inspection before failure.

4 3 - PRESSURIZER SURGE AND SPRAY LINE OPERATION The January 18, 1994 SER concluded that based on th.e NRG review of the material presented that the Staff finds the welds on the surge and spray lines acceptable for one cycle of operation.

CPCo Resolution Both the pressurizer spray and surge nozzles have Alloy ~00 safe ends welded to stainless steel pipes.

Inspections of these welds including heat affected zone of the safe end were performed during the 199! refueling outage.

Recently, a detailed Palisades specific fracture mechanics analysis using conservative PWSCC and fatigue crack.growth data was performed for both the spray and surge nozzle safe ends.. A.non-proprietary version of the fracture mechanics cakulation, "FM Assessment of Palisades Alloy 600 Components," BWNT

• Cale 32-1238965-00, was provided to the NRC in our letter dated April 24, 1995.

The results of this analysis provides the times to failure of the safe ends with a postulated 0.010" depth PWSCC and*also the allowable flaw size for continued operation through one refueling cycle. Also, the fracture mechanics analysis provides service life curves for various flaw sizes and shapes.

The Palisades'pressurizer spray line will normally operate at 535°F -. 540°F, which is essentially the PCS cold leg temperature, due to the 60 to 70 gpm pressurizer spray flow.

Since temperature is a major contributor to PWSCC, the spray nozzle safe end is ~onsidered to be less susceptible to PWSCC than the surge nozzle safe end.

The times to failure for both the spray (operating at 540°F) and surge (operating at 640°F) nozzle safe ends exceeds 40 years.

During the 1995 refueling outage, both the spray and surge nozzle safe end-to-pipe welds will be inspected by an enhanced ultrasonic examination method capable of accurately detecting and sizing flaws.- From the f~acture mechanics analysis, the acceptable fl aw size for one operating eye le for both cases is greater than the NOE detectable flaw size.

Conclusion The inspection results from the 1995 refueling outage along with the detailed fracture mechanics analyses will provide justification for continued operation of these two locations.

Moreover, to eliminate the residual ~tress*factor contributing to PWSCC, Mechanical Stress Improvement Process (MSIP) will be applied at both the pressurizer and the hot leg surge nozzle safe

~nds~

Applicati~n of MSJP will replace the existing tensile residual stresses that could initiate and propagate PWSCC with favorable compressive stresses. Thul MSIP completely removes the potential of any future PWSCC initiation or growth at these locations.

ATTACHMENT 2 Consumers Power Company Palisades Plant Docket 50-255 RESPONSE TO NRC CONCERNS AND COMMENTS RAISED IN THE JANUARY 18, 1994 SER CONCERNING CRACKING OF 600 COMPONENTS IN THE PALISADES PRESSURIZER April 28, 1995 6 Pages

RESPONSE.TO NRC CONCERNS AND COMMENTS RAISED IN THE JANUARY 18, 1994* SER CONCERNING CRACKING OF 600 COMPONENTS IN THE PALISADES PRESSURIZER I

The NRC SER dated January 18, 1994 in section 2.2.3 "Crack growth Evaluations" listed staff concerns about the licensee's engineering ~nalysis that

.determined the predicted life of the existing welds.

The following discussion is provided tn response to the NRC concerns raised in section 2.2.3.

NRC CONCERN 2.2.3(1)

Records.of weld repairs, stress relief heat treatment, and cold work of ID surfaces of welds are not available. This information could affect the PWSCC susceptibility and crack growth rate.

CPCO RESPONSE PWSCC and fatigue crack growth calculations were performed using conservative assumptions to compensate for the lack of actual dat~. To accotint for the effect of the ID surface cold work on PWSCC initiation, the initiation time was not included in the component projected service lifetime. The initial crack depth for the service life estimations is assumed to be 0.010 11 (IO mils) at which the cold work is minimal.. The residual stress magnitude was conservatively assumed to be 125% of yield stress at operating temperature. This stress is assumed to be uniformly distributed in both the axi~l and circumferential directions in the component.

These conservative assumptions take the effect of weld repair/rework into account.

For the surge and the spray nozzle girth butt welds, alternative residual stress distribution using the gui~elines established in NUREG-0313 ("Technical Report on Material Selection and Processing Guidelines for BWR Coolant Pressure Boundary Piping," Fi"nal Report, Rev. 2, January 1988) was also calculat~d. The results show that uniform residual stress distribution is overly conservative for girth butt welds.

The PWSCC crack growth curve that was used was derived from the widely accepted Peter Scott's model (Scott, P.M., "Ail Analysis of Primary Water Stress Corrosion Cracking in PWR Steam Generators," in Proceedings, Specialists Meeting on Operating Experience with Steam Generators, Brussels, Belgium, September 1991)

. which was developed from industry data with no specific material heat treatment. Therefore, it*is believed that the crack growth model is applicable to materials which ineet the ASME minimum heat treatment requi.rements.

NRC CONCERN 2.2.3(2)

The licensee calculated that cracks can grow through wall in 20 months.

This period is relatively short when compared with the Palisades operating cycle of 15 months.

2 CPCO RESPONSE The PORV safe end projected lifetime was calculated to be 20 months (14,400 Effective Full Power Hour, EFPHs) which is highly conservative when compared with the operating life of the failed PORV nozzle safe end (approximately*85,000 EFPHs). Furthermore, the examination of the failed safe end also indicated that PWSCC initiated and propagated only along the weld root repaired portion of the interior circumference and no crack was.found in the un-repaired weld root portion of the PORV nozzle safe end-to-pipe girth butt weld.

This is strong supporting evidence that the weld root repair was one of.the major causes of the PORV nozzle safe end failure.

The 1993 weld repair is a complete Code repair which removed the affected portion of the safe end.

A new butt weld was performed without any weld root repair. Therefore, the lifetime of the current PORV safe end is much greater than the calculated lifetime.

Hence, the projected lifetime of zo months is acceptable for an operating period of 15 months.

NRC CONCERN 2.2.3(3)

The crack growth calculations may not be conservative because the magnitude and distribution of the residual stresses are not known.

CPCO RESPONSE The magnitude of residual stress was assumed in the order of yield stress for the evaluation of Alloy 600 components lifetime and inspection criteria. Uniform welding rekidual stress distribution in both the axial and circumferential directions was also assumed.

This is overly co~servative for girth butt welds when compared with a more realisti~ through wall residual stress distribution given in NUREG-0313 as shown in the fracture mechanics calculation for the pressuri~er surge and.spray nozzle safe ends.

The non-proprietary version of the "FM Assessment of Palisades Alloy 600 Components", BWNT ~ale 32-1238965-00, was provided to the NRC in our letter responding to the NRC's request for additional info~mation supporting the Palisades Alloy 600 Project Plan, dated April 24, 1995.

NRC CONCERN 2.2.3{4)

The licensee ranks susceptibility to PWSCC according to the service

-temperature in specffic 1ocatfons in the pressuri1.er.

In an internal report dated May 22~ 1991, the nozzles for the temperature elements are rated moderately susceptible. However, since the nozzle for the temperature element cracked, the validity of the licensee's sus~eptibi1ity ratings, which were used to determine the longer 1'ifetimes for the spray and surge lines, (also rated moderately susceptible) is questionable.

' (

3 CPCO RESPONSE The PWSCC susceptibility ranking scheme is based on several factors which affect the PWSCC initiation and growth rate. The ranking scheme, however, does not estimate the lifetime of the component.

For example, components from the same heat of material and eiposed to the same environment and stress condition, will have the same susceptibility rank but the component which has a greater thickness will have a greater lifetime. Therefore, in terms of wall thickness, the surge nozzle safe end {t=l") life time is expected to be 4 times that of the temperature element nozzle {t=0.25"). Similarly, the spray nozzle safe end

{t=0.404") lifetime can be about 1.6 times of that of TE nozzle.

The other factor affecting PWSCC initiation and growth rate is welding residual stress distribution. The through-wall residual hoop stress distribution along the thickness of the TE nozzle is expected to be nearly uniform {due to "J-groove" weld shrinkage). However, the through-wall axial welding residual stress distribution in a piping girth butt weld is highly non-linear {U shaped as shown in NUREG-0313) with a compressive stress field at the middle of the thickness.

It can be shown that the applied stress intensity factor of a uniform stress.

field is greater than that of U shape stress distribution field of the same peak magnitude. Therefore, in terms of weld geometry, the lifetimes of the surge and the spray nozzle safe ends are longer than that of the TE.nozzles.

NRC CONCERN 2.2.3(5)

The flaw sizes (3 mils and 39 mils) assumed in the crack growth calculations have not been shown to.be detectable by the NDE techniques used.

The 1ice.nsee's assertion regarding the flaw detectability is not based on demonstrations.

CPCO RESPONSE In the fracture mechanics assessment calculation two approaches were employed to assess the lifetime of Palisades Alloy 600 components.

The first approach is lifetime calculation based on an assumed initial crack depth.

Initi:al crack depth at time zero is conservatively assumed to be 0.010" (10 mils) which is also a typical d~pth of cold work influence in on the component interior *surface.

The second approach is to calculate the allowable crack depth such that the component can continue to operate safely till the next inspection. Currently, the acceptance criteria of a allowable crack depth for 1 refueling cycle is greater than or equal to the-detectabl~.sjze of the NDE tethniqu~. Otherwise, further assessment is required for continued operation.

4 NRC CONCERN 2.2.3(6)

The licensee cites a study relating yield strengths to time to failure for partial penetrations to conclude that the surge line nozzle wi11 have a projected life of 4 times that of the PORV nozzle. Although the surge line nozzle is of a lower strength material (51.2 ksi), the CEOG has shown that material with yield strength as low as 36 ksi has failed.

A CEOG status report on PWSCC of Inconel 600 material (Attachment 6 to the licensee Engineering Analysis EA-SC-93-087-01 included in the licensee letter dated October 27, 1993} states that the relationship between yield strength and PWSCC is unknown for hot forged products (nozzles).

CPCO RESPONSE Service lifetime of the surge nozzle has been reass.essed independently by BWNT using Peter Scott's PWSCC crack growth rate model.

The model was developed based on industry data for the stress corrosion cracking of Alloy 600 steam generator tubing. Appropriate temperature correction was incorporated in Scott's model to account for the difference in operating temperature between the different Alloy 600 components and the SG tubing.

The BWNT calculations show that the projected lifetime of the surge nozzle £afe end is still greater than four times the

• calculated l if et tme of the failed PORV nozzle safe end (same as Spray nozzle safe end at 640°F).

The statement that no relationship between yield strength and PWSCC for Alloy 600 hot forged products as used in a CPCo report, was b~sed on the available data at that time (1991).

Many recent publications indicate that low yield strength Alloy 600 hot

. forged material is less susceptible to PWSCC than the high yield strength hot forged material.

NRC CONCERN 2.2.3(7)

The licensee concludes that the surge line nozzle safe end weld is exposed to water and not steam, the environment where most PWSCC has occurred.

However, a temperature element nozzle exposed to water did crack.

CPCO RESPONSE The recent PWSCC assessment of the surge nozzle safe end does not consider the effects of water and steam environments.

Hence, a nozzle

• at the same temperature exposed to. a water verses a ~te~m environment would not be less susceptible to PWSCC.

5 The NRC SER dated January 18, 1994 in section 2.2.4 "Overall Concerns" listed*

three concerns related to the Alloy 600 issue at Palisades. These concerns are addressed below.

NRC CONCLUSION 2.2.4(1}

The 1 icensee needs to demonstrate the performance of the NDE.

The NDE performed does not ensure detection of the initial flaw size that could grow through wa11.

CPCO RESPONSE In order to develop and verify the ultrasonic examination procedures and techniques, NOE mockups were fabricated.

An acceptable demonstration of accurate flaw detection and sizing capabilities of the NOE techniques on the mockups have been completed.

The results indicate that flaws of the size that would.be allowed to exist for one operating cycle can be detected, thus assuring the integrity of Alloy 600 components at Palisades.

NRC. CONCLUSION 2.2.4(2}

The 1i.censee needs to perform a failure modes analysis on a11 affected

. lines to provide additional assurance of safety until it completes the NDE performance demonstration or makes modifications to eliminate the susceptibility to PWSCC.

CPCO RESPONSE As a part of the Palisades comprehensive Alloy 600 program, service lifetime calculations and PWSCC assessments were performed for all Palisades Alloy 600 components.

NOE performance demonstration before the 1995 refueling outage and appropriate inspections and modifications where necessary during the 1995 refueling outage will satisfy this concern.

NRC CONCLUSION 2.2.4.(3)

The repair welds in the POR.V line and instrument nozzles are acceptable for the operation of one fuel cycle.

This conclusion is based on the r-epair, the NDE performed, the existing leak rate !JJOnitoring requirements and capabilities, the licensee's evaluations, the licensee's failure modes analysis submitted on November 30, 1993, and on the fact that a failure of one of the subject lines is bounded by the sma1.1 break loss-of-coolant accident (LOCA) described in.the Final Safety Ana.1ysis Report.

6 CPCO RESPONSE The subject of this letter provi.des confirmation that the PORV line safe

• end will be replaced during the 1995 refueling outage, and provides justification for continued operation with the as-repaired pressurizer temperature element nozzles.

The NRC SER dated January 18, 1994 in section 2.2.1 "The PORV, SRV, Spray and Surge Lines", tn the last paragraph on page 7 states that "... the licensee does not have data on the actual properties of the Incone1 600 materials, and the parameters associated with the original welding process, repairs, and heat treatments during fabrication."

CPCO UPDATE CPCo has acquired a set of Alloy 600 Certified Material Test Reports which contain ASME material specification, material mechanical and chemical properties for all Palisade Alloy 600 components.

This information has been utilized in the turrent CPCo PWSCC ranking scheme (Palisades Calculati~n EA-A600-03, "Alloy 600 Primary Water Stress Corrosion Cracking Inspection Prioritization Scheme Report").

The ASME Code specified a minimum heat treatment of 1600°F but the actual heat treatment was often not recorded.

Detailed weld repair information is generally not available.

We have obtained the individual weld sign-off sheets labeled "Inconel Weld" that show which lnconel field welds have rework and the percentage weld affected by the rework.

This information has been utilized in the decision to use the MSIP process on the surge nozzle and shutdown cooling nozzle safe ends to remove the stress factor componen~ from the PWSCC equation.

The NRC SER dated January 18, 1994 in section..... states that "... the inaccessible spray nozzle ID surface and OD geometry prevent effective UT.

11 CPCO UPDATE A mockup has been designed and fabricated for developing and demonstrating the ultrasonic examination procedure and techniques for PWSCC detection at the pressurizer spray nozzle safe end.

The mockup demonstration will show that the cri ti ca 1 fl aw s i z.e can_!:>~ dete_cted by the technique.

As a part of the 10 year ISi program, the nozzle OD surface will also be examined using PT.* We are also i:nvestigating the possibility of performing remote visual inspection of the weld lb surface for evidence of weld repair during the 1995 refueling outage.

ATTACHMENT 3 Consumers Power Company Palisades Plant Docket 50-255 SERVICE LIFE ANALYSIS OF PRESSURIZER TEMPERATURE ELEMENT TE 0101 WELD MODIFICATION EA-A600-0l April 28, 1995 10 Pages

PALISADES NUCLEAR PLANT ENGINEERING ANALYSIS COVER SHEET EA-A600-01 Total Number of Sheets 10 Title Service Life AnaJ¥sis of Pressurizer Temperature Element Nozzle TE-0101 Weld Modification INITIATION AND REVIEW Calculation Status Preliminary Pending Final Superseded 0

0 Jl 0

Initiated I nit Review Method Technically Reviewed Revr Rev Appd App CPC_o Descf1>tion By DetaU Qua I d

Appd By Date Alt Cale Review Test By Date By

p. r/. H Cf/zf qi/ R&c:\\

x LJc_u)~ ~2.(-1$1-e,"

fl. -f( O,l!fJq

. 0 Original Issue J

1.0 OBJECTIVE

The objective of this analysis is to evaluate the fatigue life of the i 993 modification of the Pressurizer

• Temperature Element Nozzle TE-0101. Leakage was found in the nozzle due to Primary Water Stress Corrosion Cracking (PWSCC) in the Alloy 600 nozzle near the J-groove weld. The modified design inclu4es a weld built-up pad to form a new pressure boundary at the pressurizer exterior wall surface. The nozzle body was partially severed for thermal stress reduction. The modification was found acceptable for at least one refuel cycle (to the REFOUT95, Ref.1 ). This Engineering Analysis (EA) Will calculate the fatigue usage factor of the modified TE nozzle and extend the TE nozzle fatigue life to the end of the plant design service life. The analysis includes only an ASME Section m fatigue calculation, the assessment of PWSCC at the new weld and the corrosion of the carbon steel pressurizer by the borated primary water will be

• provided in a separate report.

gCS'JEO'

. SEP 2 3 1994 ERO* PA\\...

-~

2.0 Analysis Input 2.1 Existing Analyses EA-A600-01 Page 2 Pressure and thermal transient stresses of the original design provided fu Ref 3.3 and the number of design cycles for the plant thermal transient events provided in Ref 3. 6 and 3. 7 are the input for this calculation.

The conservative peak stress index Kor the local discontinuity factor of 5, the highest stress discontinuity factor per Para. N-415.3 of Ref 3.9, used in Reference 3.3 and 3.4 is maintained in this calculation.

2.2 Material Property Data Material properties such as thermal expansion coeficients, elastic modulus and allowable stress are taken from Ref 3. 3. Inelastic stress strain curve of SB-166, type 600 at 550 Op characterized by Ramberg-Osgood constants are taken from Ref 3.10, a material property database provided for ASME Section ID Task Group on Pipe Flaw Evaluation by Battelle.

3. 0 References 3.1 EA-SC-93-087~01, "Justification of Weld Modifications to Pressurizer Temperature Nozzles for TE-0101 and TE-0102", Rev. 0.

3.2 EA-SC-93-087-02, "Half Bead Welding Modifications to TE-0101 and TE-0102",

Rev. 0 3.3 EA-SC-93-087-03, "Structural Analysis ofTemp.erature Nozzle Weld Modifications for Consumers Power Palisades Pressurizer", Rev. 0 3.4 EA-SC-93-087-04, "Acceptability of Partial Severing ofTE-0101.Nozzle", Rev. 0 3.5 EA.. SC-93-087-05, "Evaluation of Potential Interference between TE-0102 Nozzle and the Thermowell ",Rev. 0 3.6 CE Report No. CENC 1114, "Analytical Report for Consumers Power Pressurizer",dated March 1969.

3.7 CE Report No. CENC 1214, "Addendum to the Analytical Report for Consumers Power Pressurizer",dated October.1973.

3.8 CE Drawings:

D-9417-C093-019, "Pressurizer Head Temperature Nozzle Temporary Repair",

Rev. 02 D-9417-C093-021, "Pressurizer Vessel Temperature Nozzle Temporary Repair", Rev. 01

• EA-A600-01

.Page 3

3. 9 ASME Boiler and Pressure Vessel Code,Section III, 1965 Edition with Addenda through Winter, 1966 3.10 Quasi-Static Material Property Database, Release Date 7-16-92, Revision 1. 0, provided for ASME Section ill Task Group on Pipe Flaw Evaluation by Battelle, Colum.l?us, Ohio.

4.0 Assumptions 4.1 Major assumptions:

a. Fatigue life calculation was based on the transient stresses provided in Reference 3.3 (Structural analysis for the modification by ABB/CE) for all design transient events and the associated number of cycles given in Reference 3. Conceivably.

these stresses are conservative for the new pressure boundary at the weld pad and TE body junction since the thermal transient conditions at the new pressure boundary are less severe than at the analyzed location.

b. Local peak stress index factor*of 5 used in the calculation is conservative in compared to a value of two for a typical "as-weld" discontinuity.

c. The severed nozzle section is assumed to be 0.1" thick, 1/8" long and concentric. This

.is conservative since per Reference 1 (sheet 11), the desired ligament of 0.055" was chosen with an accuracy of+/- 0.007".

d. Inelastic stress strain relationship at the partial severed nozzle wall is assumed to be the typical stress strain curve provided in Reference3. 10 for Inconel 600 material at 5 50 op.

This curve may not represent the stress strain relationship of the installed TE nozZle.

However, it is considered to be acceptable due to the conservatism in the analytical approach and other assumptions 4.2 Minor Usage factor for any load pairs with number of allowable cycles greater than 106, is consi~ered to be zero. This is a common practice in the vessel component fatigue.

analysis.

5. 0 Analysis EA-A600-0l Page 4 Welding the nozzle to the pressurizer head at both the interior surface anf exterior surface creates a potential high stress in the nozzle due to the axial differential thermal expansion during the plant thermal transients. By partially severed the nozzle, the thermal induced load can belimited. Calculation provided in Ref 3.4 was conservatively based on the material collapse load in the severed nozzle wall. By using the inelastic stress strain relationship of the material, the thermal expansion induced stress is significantly reduced.

Thus, the service life of the TE-0101 temperory repair can be extended.

S. 1 The true stress strain curve chararacterized by Ramberg-Osgood contants provided in in Ref 3.10 for typical Inconell 600 at operating temperature of S S0°F is in the following form.

where:

~

= ~

+a*(~)na Eo* ao ao E

= 0.00089

.. 0

Refference strain at 550 °F a

= 28500 0

Refference stress at 550 °F Cl := 5.307

Ramberg"".Osgood constants n := 3.9 Therefore, strain as a function of stress is: Strain( a) *~

• 0 *[.,"

0

+a* (.,"J 0

1 a := 1000, 2000.. 50000 6*104 S.4*104 4.8*104 4.2*104 3.6*104 (J

3.*104

-2.4*104 1.s*104 l.2"104 6000 0

./

/

I 0

0.01

~*,

0.02 0.03 0.04 o.os Strain( a)

e 5.2 Axial thennal expansion induced stress e EA-A600-01 Page 5 L := 4.12 m

D 0 := 1.315 m

t 1 := 0.25 in Di := D 0 - 2*t l D

2 _ D.2 0

I A 1 := 7t*

4 t 2 := 0.1 in 1 := ~

m 8

di := D 0 - 2*t 2 D

2 - d.2 0

I A2 := 7t*

4 For a t)rpical load case.

Tl Of a.1 in/in-°F T2 Of a.2 ifl/in-°F E

psi Pressurizer head thickness Nozzle OP Nozzle thickness Nozzle IP A 1 = 0.836 in2 Nozzle cross section area Nozzle remaining ligament Nozzle severed length A 2 = 0.382 in2 Nozzle severed cross section area The average temperature of the pressurizer head - 70 The thermal expansion coef of carbon steel at T 1 The average temperature of the nozzle - 70 The thermal expansion coef of alloy 600 at T2 Elastic modulus at T2

.i..... -l... ti.........

_;.i*"

. r

• e EA-A600-0l Page 6 the axial thermal expansion differential as a function of TI and T2 is o I ( T I, T2, a I, a.2) = L-( a I * ( TI ) - a.2 * ( T2)).

Since this thermal expansion is restrained by the welds at the interior and exterior wall, force F

, is induced l!Hhe nozzle. The total displacement due to force F is F*L

. (IFI) F

& 2(F,E) := -- + l*Stram -.- *-I -

1 ArE A2 F

The compatibility condition requires that f ( F, TI, T2, a I, a.2, E) : = o I (TI, T2, a I, a.2) - o 2 ( F, E) =O Force Fis solved using MATHCAD root() function with an initial guess value:

then the axial stress Sxta is :

Sxta(Tl, T2,a.l,<i2,E) := root(f(F, Tl, T2,a.l,a.2,E),F)

A1 Using the average temperatures Tl and T2 for all design transient events and the coresponding thermal expansion coefficients and the material modulus of elasticity provided in References 3.3 (page 12 of27), the thermal axial stress, Sxta, for each transient is calculated and the result is :

Load Cases Tl T2 Of Of

• .~':

Tl.

I T2i Heat Up 566 583 Cool Dn@2.865 51 10 Steady State 583 583 Loading 5%

583 591 Unloading 5%

583 577 Loading 15%

583 598 583 574 Unloading 15%

583 591 Step Inc 583 585

• Step Dec @20 583 580 Step Dec@120 583 557

• Reactor Trip 583 557 Loss of Flow 583 557 Loss of Load 583 632 SV0per@20 583 55.7

• SV Oper @ 200 583 343 St Line Rupture 572 267 Cool Dn@l.58

a. I a.2 E

10-6 10-6 106 Sxta

o.

Ksi EA-A600-0l Page 7 02 a.1.-106a.2.-106E.. 10- 6sx.

I I

I I

o.

I o 2[ (Sx)i" 103-A 1,Ei]

7.33 7.95 29.1 6.13 7.6 31.5 7.33 7.95. 29.1 7.33 7.95 29.1 7.33 7.95 29.1 7.33 7.95 29.1 7.33 7.95 29.l 7.33 7.95 29.1 7.33 7.95 29.1 7.33 7.95 29.1 7.33 7.95 29.1 7.33 7.95 29.1 7.33 7.95 29.1 7.33 7.95 29.1 7.33 7.95 29.1 7.33 7.7 30 7.33 7.6 30.2

-11.18 -0.002003 -0.002005 6.64 0.000975 0.000968

-8.98

-0.001489 -0.001487

-10.16 -0.001751 -0.001752

-8.01

-0.001293 -0.001292

- 11.09 -0.001981 -0.001982

-7.5

-0.001194 -0.001194

-10.16 -0.001751 -0.001752

-9.28 -0.001555 -0.001551

-8.51

-0.001391

-0.00139

-4.25

-0.000638 -0.000645

-4.25

-0.000638 -0.000645

-4.25

-0.000638 -0.000645

-14.54 -0.003094 -0.003093

-4.25

-0.000638 -0.000645 20.8

. 0.006725 0.006724 23.12 0.008914 0.008915

2. 3. I 2

£

7. 67b.,,...,o 7.

/c~4 r.,L J~\\;

EA-A600-0l Page 8 The stress due to differential thermal axial growth, Sxta, is superimposed on the pressure stress and radical interaction induced stress calculated for the original design (page 10 of Ref 3.3).

The same conservative stress concentration factor of 5 is assumed for axial stress due to weld local discontinuity. The results of the stress intensity and the peak stress intensity for all load cases are:

-~-:.;.. ::

Load Cases.

Heat Up Cool Dn@2.865 Steady State Loading 5%

Unloading 5%

Loading 15%

Unloading 15%

Step Inc Step Dec@20 Step Dec @120 Reactor Trip Loss of Flow Loss of Load SVOper@20 SVOper@200 St Line Rupture Cool Dn@l.58 cr xx : = ( cr x + Sx)

• 5 ax cre Ksi Ksi CJ X*

CJ 0.

CJ r.

CJ XX*

1 1

1 I

-0.88 0.51 3.46 6:69 0.55 3.17 0.08 2.42 0.89 3.75

-0.34 1.74 1.12 4.12 0.05 2.35 0.41 2.96 0.71 3.46 2.18 5.9 2.18 5.9 2.18 5.9

-2.84 -2.64 2.18 5.9 21.33 37.92 26.27 49.22 0

0 0

0 0

0 0

0 0

0 0

0 0

0 0

0 0

-60.28 50.491

-42.132

-50.418

-35.609

-57.155

. -31.878

-50.568

~44.336

-38.978

-10.335

-10.335

-10.335

-86.878

-10.335 210.674 246.933.

I cr x - cre

-60.79

-60.28 43.801 50.491

-45.302

-42.132

-52.838

-50.418

-39.359

-35.609

-58.895

-57.155

-35.998

-31.878

-52.918

-50.568

-47.296

-44.336

-42.438

-38.978

-16.235

-10.335

-16.235

- 10.335

-16.235

-10.335

-84.238

-86.878

-16.235

-10.335 172.754 210.674 197.713 246.933 20 year design cycles n

250 250 999999 7500 7500 7500 7500 7500 7500 7500 250 50 50 100 100 1

250

• Based on the calculated stress, the cumulative usage factor for stress intensity (crx-crr) and.

( crx-cr0) of all design transient events in a 20 year period is calcwated to demontrate that. design fatigue stress is not the gorvening factor for th~ TE-0101 nozzle service life.

I

EA-A600-01

. Page 9 5.3 Usage Factor for Stress Intensity (cr~-a!}

• The stress 5*<rx-<1r for all transients are sorted to form load pairs for fatigue calculation. The Smax and Smin for each load pair and the coresponding number of cycles are tabulated in the following table. The stress intensity range Sn for Ke calculation is consevatively taken as (Snw;-S~/5.

NoteAhat Ke factor is then calculated per the Code (Ref 9) if S0 is greater than three times of Sm:

Code's terms for K0 factor:

n := 0.3 m := 1.7 Sm:= 23.3 ksi 1-n (Sn

)

Ke := 1 +

• -- - I a n*(m_:_ 1) 3*Sm.

The peak stress intensity Sp with stress concentration factor of 5 Sp := Ke* ( S max - S min) a Alternating stress:

f s.

p Sa:= -a 2

The number of allowable. cycles N for each load pair is calculated from the alternating stress Sa Using the Code design fatigue curve (Fig. 1-9.2, Ref3.9) as the folowing:

Where:

log(~)

N := Nl*(N2). log(~~). o Nl S2 <Sa< Sl Nl is the allowable cycles for alternating stress S 1 N2 is the allowable cycles for alternating stress S2 The usage factor u for I1i cycles of a load pair i with the allowable cycles Ni is:

The total usage factor n* I u i :=-a N*

I U:=Luia I.

The results:

S max. S min. S n.

• 1 1

I Ke.

I S Pi C.D. @l.58"- SVOp. @20 247

-87 66.8 1

334 C.D.@1.58 - Heat Up 247

-60 61.4' 1 307 Steam L Ru'pF.'- Heat Up 211

-60 54.2 1

271 C.D.@2.865 -Heat Up 50

-60 22 1

110 C.D.@2.865 -Loading 15% 50

-51 20.2 1

101 Enveloped the remained

-10

-51 8.2 1

41 load pairs 5.4 Usage Factor for Stress Intensity (er~~

Sa.

I 167 154 136 55 51 21 EA-A600-01 Page 10

n.

I N.

I 100 271 150 349 1

506 49 13965 201 20960 52750 1* 106 5

u.

l 0.369 0.43 0.002 0.004 0.01 0.009 L U. =O 823 < 1 l

i = 0 S max. S min. S n.

Ke. S Pi Sa.

l

n.

l N.

l

u.

l I

l 1

I C.D. @I.SS - SV Op. @20 198

...:g4 56.4 1

282 141 100 271 0.369 C.D.@l.58 - Heat Up' 198

-61 51.8 259 130 150 349 0.43 Steam L Rupt. - Heat Up 173

-61 46.8 234 117 1

506 0.002 C.D.@2.865 -Heat Up 44

-61 21 1

105 53 49 13965 0.004 C.D.@2.865 - Loading 15% 44

-59 20.6 1

103 52 Enveloped the remained

-16

-59 8.6 1

43 22 201 52750 20960 1*106 0.01 0.009 load pairs 5

L u. =0.5 1

i = 0 6.0 Conclusion The accumulative fatigue usage factor for TE-0101 nozzle is less than 1 for the remaining of the plant design life. This calculation demonstrates that fatigue is not the

• main concern for the TE-0101 nozzle service life.

<l

ATTACHMENT 4 Consumers Power Company Palisades Plant Docket 50-255 SERVICE LIFE ANALYSIS OF PRESSURIZER TEMPERATURE ELEMENT TE-0102 WELD MODIFICATION EA-600-02 Apri 1 28, 1995 6 Pages

PALISADES NUCLEAR' PLANT ENGINEERING ANALYSIS COVER SHEET EA-A600-02 Total Number of Sheets.-=-6 __

Title Service Life Anaivsis of Pressurizer Temperature Element Nozzle TE-0102 Weld Modification.

INITIATION AND REVIEW Calculation Status Preliminary Pending Final Superseded 0

0 l(

0 Initiated lriit Review Method Technically Revle-d Revr Rev Descf1>tlon 1----------1Appd1-----r---...---+----...----1 App By DetaD Qual d

CPCo Ap~

0 Original Issue By f>. H- +{

f. +/01'Ntr Date Alt Cale Review Test By Date By I

1.0 OBJECTIVE

The objective of this analysis is to evaluate the fatigue life of the 1993 modification of the Pressurizer Temperature Element Nozzle TE-0102. Leakage was found in the nozzle due to Primary Water Stress Corrosion Cracking (PWSCC) in the Alloy 600 nozzle near the J-groove weld. The modified design includes a weld built-up pad to form a new pressure boundary at the pressurizer exterior wall surface The modification was found acceptable for at least one refuel cycle (to the REFOUT95, Ref. I). This Engineering Analysis (EA) will calculate the fatigue usage factor of the modified TE nozzle and extend the TE nozzle fatigue life to the end of the plant design service life. The analysis includes only an AS:ME Section III fatigue calculation, the assessment of PWSCC at the new weld and the corrosion of the carbon steel pressurizer by the borated primary water will be provided in a separate report.

RECEIVED SEP 231994 ERC

• PA1-

2.0 Analysis Input 2.1 Existing Analyses EA-A600-02 Page 2 Pre~~re and thermal transient stresses of the original design provided in Ref 3.3 and the number of design cycles for the plant thermal transient events provided in Ref 3. 6 and 3. 7 are the input for this calculation.

The conservative peak stress index Kor the local discontinuity factor of 5, the highest stress discontinuity factor per Para. N-415.3 of Ref 3.9, used in Reference 3.3 is maintained in this calculation.

2.2 Material Property Data Material properties such as thermal expansion coefficients, elastic modulus and allowable stress are taken from Ref. 3.3.

3. 0 References 3.1 EA-SC-93-087-01, "Justification of Weld Modifications to Pressurizer. Temperature Nozzles for TE-0101 and TE-0102", Rev. 0.

3.2 EA-SC-93-087-02, "Half Bead Welding Modifications to TE-0101 and TE-0102",

Rev. 0 3.3 EA-SC-93-087-03," Structural Analysis of Temperature Nozzle Weld Modifications for Consumers Power Palisades Pressurizer 11

, Rev. 0 3.4 EA-SC-93"'.087-04, "Acceptability of Partial Severing ofTE-0101 Nozzle", Rev. 0 3.5 EA-SC-93-087-05, "Evaluation of Potential Interference between TE-0102 Nozzle and the Thermowell ", Rev. 0 3.6 CE Report No. CENC 1114, "Analytical Report for Consumers Power Pressurizer",

dated March 1969.

3.7 CE-Report No. CENC 1214, "Addendum to the Analytical Report for Consumers Power Pressurizer", dated October 1973.

3.8 CE Drawings:

D-9417-C093-019, "Pressurizer Head Temperature Nozzle Temporary Repair",

Rev. 02 D-9417-C093-021, "Pressurizer Vessel Temperature Nozzle Temporary Repair", Rev. 01

3. 9 ASME Boiler and Pressure Vessel Code,Section III, 1965 Edition with Addenda through Winter, 1966

4. 0 Assumptions 4.1 ~fajor assumptions:

EA-A600-02 Page 3 Fatigue life calculation was based on the transient stresses provided in Reference 3.3 (Structural analysis for the modification by ABB/CE) for all design transient events-and the associated number of cycles given in Reference 3. Conceivably these stresses are conservative for the new pressure boundary at the weld pad and TE body junction since the thermal transient conditions at the new pressure boundary are less severe than at the analyzed location.

Local peak stress index factor of 5 used in the calculation is conservative in compared to a value of two for a typical "as-weld" discontinuity.

4.2 Minor Usage factor for any load pairs with number of allowable cycles greater than 106, is consi~ered to be zero. This is a commori practice in the vessel component fatigue analysis.

5. 0 Analysis The ASME Section ID stress and fatigue usage factor calculations for class I components provided in References 2,3 and 4. qualified the TE -102 temporary modification for at least one fuel cycle. The calculated stress in Reference 3 included stresses due to differential thermal axial growth of the nozzle and the pressurizer wall superimposing on the pressure and radical interaction induced stress calculated for the*

original design in Reference 7. A conservative stress concentration factor of 5 were assumed for the weld local discontinuity. Based on the calculated stress of the above references, the cumulative usage factor for the all design transient events in the plant remaining 20 year period is calculated to evaluate the fatigue life of the 1993 TEO 102 weld modification. The usage factors for stress inttmsit)' (ax-ar) and (ax-00) are calculated separately and the higher is considered to be the usage factor of the

.TE0102.*.

Refer to Reference ~, page _22 an~ 23, the maximum stress intensities at the outside surface of the TE nozzle are listed below.

L_oad Cases Hydro Test Leak Test Heat Up Cool Dn @2.865 Steady State.

Loading 5%

Unloading 5%

Loading 15%

Unloading 15%

Step Inc StepDec@20 Step Dec@l20 Reactor Trip Loss of Flow Loss of Load SVOper@20 SV0per@200 St Line Rupture (crx-00) (crx-Or)

-9.4 7.4

-1.5 5.9

-17.4

-17.2 4.7 12.l

-13.7

- 9.3

- 12.1

-4.8

-12.6

-7.6

-12.4

-4.6

-12

-6.7

. -13.7

. -8.3

-13.8

-9.3

- 13.2

-8.6

-8.7

- 1.8

-8.7

- 1.8

-8.7

-1.8

- 14.5

-6.5

-8.7

- 1.8 31.1 69.2 (Scrx-00) (5crx-Or) 20.3 37.1 16.2

. 29.7

-86.2

-86.

52.9 60.3

-51.0

-46.6

-31.4

-24.2

-42.8

-37.8

-30.7

-22.9

-38.8

-33:5

-46.7

-41.3

.,.50.8

-46.3

-47.6

-42.9

-15.8

-8.8

-15.8

-8.8

-15.8

-8.8

-40.4

-32.4

-15.8

'-8.8 307.9 346.1 n

EA-A600-02 Page 4 5

160 250 250 (l

1500 1500 7500 1500 1500 1500 1500 250 50 50 100 100

. \\

S. l Usage Factor for Stress Intensity (cr~-arl EA-A600-02 Page S The stress S *crx-ar for all transients are sorted to form load pairs for fatigue calculation..

The s:.X and Smin for each load pair and the coresponding number of cycles are tabulated in the following table. The stress intensity range S0 for Ke calculation is consevatively taken a't{Smax -Smin)/S. Note that Ke factor is then calculated per the Code (Ref 9) if S0 is greater than three times of Sm:

Code's terms for Kc factor:

n := 0.3 m := 1.7 Sm:= 23.3 ksi Ke factor is

• 1-n (sn* *i Ke :=.I +

• -.-- - I a n*(m-1) 3*Sm

• The peak stress intensity Sp:.

Sp:= Ke*(S max - S min)a Alternating stress:

Sp Sa:= 2D The number of allowable cycles N for each load pair is calculated from the alternating stress Sa. Using the Code design fatigue curve (Fig. 1-9.2, Reference 9) as the following:

log(~)

N *~NI (:~) log(~~) o Where:

S2 <Sa< SI NI is the 8nowable cycles for alternating stress SI N2 is the allowable cycles for alternating stress S2 The usage factor u for Dj cycles of a load pair i with the allowable cycles Ni is:

The total usage factor i := 0.. 3 n* I Ui :=-a N*

I U:=.Luia

EA-A600~02 Page 6 The results:

n S max S min Sn Ke Sp Sa n

N N

Steam L. Rup._-:- Heat Up 346.l

-86 86.4 1.787 772.2 386.l 31 0.032

• .~-:

Cool D. - Head Up 60.3

-86 29.3 146.3 73.15 250 4227 0.059 Hydro - Steady State 37.1

-46.6 16.7 83.7 41.85 5

54460 0.0001 Enveloped the remained 29.7

-46.3 15.2 76.3 38.15 53210 92930 0.573 load pairs u := 0.663 5.2 Usage Factor for Stress Intensity (cr~~}

S max S min Sn Ke Sp Sa n

N Steani L. Rup. - Heat Up 307.9 -86.2 78.82 1.425 561.6 280.8 1

67 Cool D. -Head Up 52.9 -86.2 27.82 139.l 69.55 250 5132 Hydro - Steady State 20.3

- 51 14.26 1

71.3 35.65 5

131600 Enveloped the remained 16.2 -50.8 13.4 67 33.5 53210 184300 load pairs 6.0

Conclusion:

The accumulative fatigue usage factor for TE-0102 at the new pressure boundary weld is less than 1 for half of the plant design fatigue cycles~ This calculation demonstrates that operational fatigue is not the main concern for the TE-0102 remaining service life.

n N

o.oi5 0.049 0.000 0.289 u := 0.352

ATTACHMENT 5.

Consumers Power Company Palisades Plant Docket 50-255 JUSTIFICATION FOR CONTINUED OPERATION OF PRESSURIZER TE-0101 AND TE-0102 NOZZLES AS REPAIRED DURING THE 1993 REFUELING OUTAGE EA-600-15 April 28, 1995 19 Pages

PALISADES NUCLEAR PLANT ENGINEERING ANALYSIS COVER SHEET

• EA-A600-15 Total Number of Sheets

• 19 Title Justification for~c:>ntinued Ogeration of Pressurizer TE-0101 & TE-0102 Nozzles as Regaired Durina the 1993 Refuelina Outaae INITIATION AND REVIEW Calculation Status Preliminary Pending Final Superseded 0

0

")(

0 Initiated lnit Review Method Technically Reviewed Revr Rev Appel Appel CPCo Description By Detail Qua I By

. Appd By Date Att Cale Review Test

~y I Date 0

Original Issue f 11. If~

~~

v l.. J ~v 3/30As ~'?,

P. Hoang 3130195 ScRa~**

Number of Pages:

EA 13 Attachment A 4

Attachment B 2

TOTAL 19

1.0 Objectives

P.ADES. NUCLEAR PLANT ANALYSIS CONTINUATION SHEET

• The cpwpose of this report is to provide justifications for continued operation of the pressurizer TE-0101 and TE-0102 nozzles which were repaired during the 1993 Refueling Outage. The justifications are based on a comparison of the factors contributing to PWSCC (Primary Water Stress Corrosion Cracking) of the pad weld configuration and the original J-groove weld configuration, and a susceptibility comparison of the TE nozzle base metal and the pad weld metal.

2.0

References:

2.1

• EA-SC-93-087-01," Justification of Weld Modification to Pressurizer Temperature plemeQ.t Nozzle for TE-0101 and TE-0102", Rev 0.

2.2 Letter from W.T. Russell of NRC to Rasin of NUMARC transmitting "Safety Evaluation for Potential Reactor Vessel Adaptor Tube Cracking", dated November 19, 1993.

2.3 EA-A600-0l, "Service Life Analysis of Pressurizer Temperature Element TE-0101 Weld Modification", Rev. 0.

2.4 EA-A600-02, "Service Life Analysis of Pressurizer Temperature Element TE-0102 Weld Modification", Rev. 0.

2.5 EA-A600-07, BWNr Document 51-1235057-00, " Ser-vice Life Assessment of TE Nozzle Penetrations", Rev. 0.

2.6 EA-A600-09, BWNT Document 32-1235177-00, " FM Assessment of Palisades Alloy 600 Components", Rev. 0.

2.7 Pavlichko W.R, "Evaluation of Pressurizer Corrosion Rate Due to Expo~ure to Borated Water ", sent to Anand Gangadharan on *

.October 23, 1994 2.8 Pavlichko W., "Palisades Metallurgical Failure analysis of Pressurizer (T-72, Line.CC-11) PORV Penetration Inconel Safe-End to Type 316 Stainless Steel Piping Weld", December 1, 1993 2.9 NUREG/CR-6245, "Assessment of Pressurized Water Reactor Control Rod Drive Mechanism Nozzle Cracking ", Idaho National Engineering Laboratory, EG&G Idaho, Inc.

EA-A600-15 Sheet _2_ Rev # _O __

PISADES NUCLEARPLANT ANALYSIS CONTINUATION SHEET *

3.0 Assumritions

4.0 3.1 Maj~sumptions:

3.2 4.1 It is assumed that there was no pre-existing PWSCC in the TE base metal near the pad weld. This assumption is based on an engineering judgement that the material in this area is quite far away from the J-weld and therefore, it was not subjected to any significant welding

• residual stress which could initiate PWSCC.

Minor assumptions:

None Analysis:

~

Background:

In October of 1993, an axial through wall crack was found on two temperature element nozzles (TE-0101 and TE-0102) on the pressurizer. A pad weld repair was performed at that time to permit Palisades to operate safely for one fuel cycle (Reference 2.11). As a result of the evaluations documented in Reference 2.1, the nozzle body of TE-0101 was partially severed by an EDM cut. The TE-0102 nozzle body, however, was not cut. After the modific.ation, the TE nozzle thermal fatigue calculation was reevaluated and the TE nozzles were qualified for a longer period. (Reference 2.3 and 2.4).

Service life assessment of both TE nozZies was performed in Reference 2.5. In this reference, PWSCC and fatigue crack growth analyses of a postulated axial crack in the TE nozzle body were performed. The analysis conservatively estimated a service life of 7.5 years of continuous operation. As an additional engineering justification for continued operation of the TE nozzles, an assessment of the service life of the TE nozzles subjected to PWSCC is performed by comparing the susceptibility factors of the repaired configuration (in 1993) to the original J-weld design, The comparison provides an estimation of the service life of the existing TE based on the known lifetime of the original J-weld configuration. The same comparison is also performed for the weld pad and the nozzle body to support the judgement that an axial through wall crack in the. nozzle body is more likely to happen befor:e a crack through the interface between the nozzle body and the weld pad. Thus potential leakage at TE nozzles can be detected before failure.

EA-A600-15 Sheet _3_ Rev # _o __

• c=urs

• n rlHDI e

PALISADES NUCLEAR PLANT

• ANALYSIS CONTINUATION SHEET The evaluation in section 4.0 pertains to TE-0101 which is considered to be the limiting case of the two nozzles. The service life of-TE-0102 nozzle is discussed in section 5.0 4.2 Evaluation of Potential Failure Modes of the Existing TE Nozzles:

Potential failure modes or crack sites and crack orientations are depicted in the Figure 1. Each failure mode is evaluated in the following sections:

__ _.:_--+-----i Pressurizer Weld Pad Figure 1:

Potential crack sites and orientations:

Key notes 1-Pressurizer based metal *corrosion 2-Axial cracks in TE nozzle 3-Circlimferential Cracks in TE Nozzle 4-Circumferential Cracks in weld pad 4.2.1 Fatigue Stress:

I TENozzle A fatigue stress analysis (Reference 3) was performed for the 1993 repaired

• TE-01O1 and TE-0 i 02 nozzles per the ASME Section m Code. The fatigue life of the modified TE-0101 nozzle is qualified for 20 years. Therefore, fatigue faihlre is not a main concern for the continued service of TE-0101 nozzle.

EA-A600-15 Sheet _4_ Rev # _a __

SADES NUCLEAR PLANT ANALYSIS CONTINUATION SHEET

• 4.2.2. Cof'l'!Wion of pressurizer wall exposed to primary water.'

Evaluation reports performed by CPCo (Reference 2.7) and BWNT (Reference 2.4) indicated that the corrosion rate of the carbon steel in the crevice of TE-0101 nozzle (the annulus of the penetration as shown in Figure 1) is extremely low. The reports concluded that the exposed base metal at the pressurizer TE nozzles are acceptable for more than 20 years of continued operation. Therefore, corrosion of carbon steel exposed to primary water is. not a main concern for the continued operation of TE-0101. nozzle to continue operation.

Furthermore, the recent base metal inspection of a pressurizer level tap nozzle (repaired in 1990) at Arkansas Nuclear One Unit 1 (ANO-I) indicated that there is insignificant corrosion of the pressurizer base metal (Reference 2.4). The level tap nozzle is very similar in configuration to Palisades repaired TE nozzles. For these,

.. reasons, it is concluded that corrosion inspection of pressurizer base metal is not necessary for Palisades TE nozzles.

4.2.3 PWSCC at the TE nozzle inside diameter suiface:

The nozzle inside diameter (ID) surface is exposed to the primary water. Due to the welding residual stresses, two possible orientations of PWSCC initiation are along the axial and the circumferential direction in the TE nozzle (Figure 1) The axial failure modes of the TE-0101 nozzle body are expected to be similar in nature to other J-groove welded penetrations such as reactor vessel head CRDM (Controlled Rod Drive Mechanism) nozzles. The axially oriented PWSCC in the nozzle body is not a short-term safety concern since the crack length is limited and will not result in an ejection failure for the nozzle (Reference 2.9). Maximum axial tensile welding residual stress on the nozZie ID surface is expected to be less that the residual hoop stress. This tendency is similar to CRDM J-groove weld.residual stress distribution for which hoop stress is great~r than axial stre.ss by a factor of 1.6 (Reference. 2.9): Therefore, ID initiated circumferential through wall crack is not expected to happen before axial crack. Hence, it is concluded that ID initiated PWSCC of TE nozzles poses no short-term safety concern and visual inspection for leakage is sufficient for safe operation the component.

4.2.4 PWSCC initiated at TE nozzle weld pad root area*

The TE-0101-weld pad is a structural weld and is the new pressure boundary of the nozzle. The weld pad root area may be exposed to primary water due to leakage through the EDM *cut of the nozzle EA-A600-15 Sheet _5_ Rev # _o __

• • -*n1111K&

4.3 4.3.1

' 4.3.2 PISADES NUCLEAR PLANT ANALYSIS CONTINUATION SHEET

• body (only foe TE-0101) and through the axial cracks above the J weld. Potential PWSCC initiation at the pad weld root area is shoW11;;in Figure 1 (key note number 4). The service life assessment of the weld pad is provided in section 4.4 Com12arisons of Factors Contributing to PWSCC of the Existing Pad Weld Configuration and the Original J-weld Configuration of TE-0101 Nozzle:

The service life of the TE nozzle body subjected to an 0.01" initial PWSCC at the nozzle ID is conservatively estimated to be 7.5 years as calculated in Reference 2.5: This service life is considered to be a lower bound estimation. As an alternative approach, the service life

. of the existing TE could be estimated from the known service life of the original J-weld design and comparing the susceptibilify of the two configurations.

Operating temperature and weld geometry are the two main factors contributing to PWSCC which affect the initiation time and crack growth rate. The effects of these factors on the life time of the.

current pad weld configuration and the original J-weld configuration

• of the TE-0101 nozzle* are discussed in the following sections.

Normal Ope'rating Temperature:

PWSCC is a thermally activated process where the crack initiation and. growth rate are very sensitive to temperature. For example, from data of failed steam generator tubes, the PWSCC has been first reported in the tubes on the hot-leg side, not on the cold leg side.

From the Arrhenius relation of crack growth rate and temperature, the growth rate is reduced by approximately a half for* 18 degree.

decrease in temperature. The normal temperature at the outside surface. of the pressurizer wall is somewhat less than the temperature at the inner wall. Therefore, it is expected that the repaired TE nozzl.e* is less susceptible to PWSCC than the original J-weld de.sign.

fV e/d Geometry:

Welding induced residual stress is known as a major contributor to PWSCC susceptibility. Many PWSCC failures at J-weld penetration have been reported. These failures include the axial cracks in the original J-weld design of the TE-010 \\ nozzle.

Due to the constrain of the J-groove, the weld metal shrinkage pulls the lower end of the TE nozzle radially outward. The radially outward movement imposes hoop stress and axial bending stress on the nozzle wall. Figure 2 illustrates the imposed weld shrinking EA-AS00-15 Sheet _6_ Rev # _o __

c-'-*----1f'.nm.-.....,nt

P.SADES NUCLEAR PLANT ANALYSIS CONTINUATION SHEET

• forces on the nozzle end. The weld shrinking force in J-weld is in opposite direction of weld shrinking force on a weld overlay which is in;-CQmpression as shown in Figure 3.

Fixed Boundary Tension Load due to J Weld Shrinkage Figure 2. Weld shrinking forces on nozzle at J-weld The shrinkage of. pad weld metal deposited on the TE nozzle OD surface is less restricted by the vessel wall and more free surface than the J-groove weld. The circumferential* shrinkage of the weld tend to impose a radially inward deformation on the TE nozzle in some what similar to a weld overlay. Thus, because of weld geometry, the residual stress in the TE nozzle is conceivably less than the residual stress irt the original J-weld.

Weld Overtay Free Boundary Compressive Load due to Weld Overlay Shrinkage. *

• Figure 3. Typical weld shrinking forces at a weld overlay EA-AS00-15 Sheet _7_ Rev # _o __

P.SADES NUCLEAR PLANT ANALYSIS CONTINUATION SHEET

• The original J-weld was stress relieved during the pressurizer fabrication. A stress relief on a J-weld could redistribute stresses whicb:..was built-up due to the uneven shrinkage of a multiple weld

• pass process, but it does not affect the total radial shrinkage imposed on the nozzle. Thus the fabrication stress relief does not have an impact on the comparison between J-weld and pad weld in this report.

Based on the difference of weld geometry, it is concluded that the pad weld design is less susceptible to* PWSCC than the original J-weld design. The other effects on residual stress such as the differential thermal expansion of the TE-0101 nozzle and the pressurizer wall and the bowing of the TE-0101 nozzle during welding are evaluated in the next sections.

4.3.3 Differential Thumal Expansion in Axial Direction:

The TE-0101 nozzle was partially severed by an EDM notch to reduce the axial force induced by differential thermal expansion of the nozzle and the pressurizer wall. The pressurizer wall was pre-heated to 500°F per the Code requirement before the pad welding.

The preheat reduces the difference of the average temperature of the pressurizer wall and of the nozzle body. Therefore, it is estimated that the differential temperature is in the range of less than three hundred degree. As calculated in Reference 2.3, for a differential temperature of 305°F (of the most severe transient event), the induced axial stress is 2.3 ksi. Hence the axial stress in the TE nozzle due to the differential thermal expansion during the welding is approximately in the range of 23 ksi. The shear stress in the* weld pad is approximately a half of the axial stress in the nozzle or 11. 5 ksi because the weld pad thickness is twice the nozzle thickness. For normal operating condition, (no differential temperature), the axial stress in the nozzle is -9 Ksi, hence, shear stress in the weld pad is -4.5 ksi. The above stress level is far less than yield stress which is a typical stress level for PWSCC initiation. Therefore, it is concluded that the effect of shear stress induced by differential thermal expansion of the TE-0101 nozzle and the pressurizer wall that resulted from welding is insignificant for PWSCC initiation.

The axial force does not affect the pressure boundary portion of the nozzle which is outside the two welds and therefore, there is no potential for circumferential cracking on the TE nozzle beyond the weld pad.

EA-A600-15 Sheet _8_ Rev # _o __

. P.SADES NUCLEAR PLANT ANALYSIS CONTINUATION SHEET 4.3.4 Bowing of the TE Nozzles:

The TE nozzles were reported. as being bent slightly due to pad welcijs.g process. The sequence of weld passes may cause uneven heat applied to the nozzle circumferentially. However the bowing does not affect the pressure boundary portion of the nozzle which is outside of the two welds and therefore, there is no potential for circumferential cracking on the TE-0101 nozzle outside the weld pad. The bowing, however, could create additional stress to the welding residual stress at the ji.Jncture of the TE-0101 and the weld.

pad. The bowing imposes compressive force -Non one side arid tension force +N on the other side of the nozzle OD. This creates non-uniform stress distribution along the circumference of the nozzle OD which eliminates a potential for 360 degree crack initiation along the circumference of the weld pad/nozzle interface. In the tension side of the noZzle OD, the tension force +N at the weld pad/nozzle interface could induce an additional hoop stress in the TE nozzle as depicted in figures 4 and 5. This has a negative affect on the existing TE configuration.* However, considering other positive effects discussed on sections 4.3.1 and 4.3.2, the service life of the existing TE nozzle is judged to be greater than the service life of the failed original J-weld configuration.

Bending Moment and Differential Thermal

• Expansion Force Fixed Boundary Figure 4. Bending moment and differential thermal expansion force F EA-A600-15 Sheet _9_ Rev # _o __

IBJ PISADES NUCLEAR PLANT ANALYSIS CONTINUATION SHEET Fixed Boundary

[

Forces and Moment J

on TE nozzle and Weld Pad Figure 5. Bending moment and forces at nozzle and weld pad interface 4.4 Comparisons of Factors Contributing to PWSCC in the Weld Pad and the Nozzle Base Metal of TE-0101:

Major factor contributing to PWSCC in TE-0101 nozzle base metal and pad weld metal are compared in the following sections to support the conclusion that the weld pad is less susceptible t()

PWSCC than the nozzle. Thus leakage through axial cracks in TE-0101 nozzle could happen well before the leakage through any through wall cracks in the weld pad.

4.4.1 Welding Residual Stress in Base Metal and Weld Metal:

It. is significant to note that, for a thick wall cylinder of the saine

• size of the TE nozzle (OD=l.32", ID=.875") subjected to uniform external pressure P, the maximum hoop stress at the inner surface is approximately three times the external pressure magnitude.

Therefore, it is expected that welding residual stress including the bowing effect in the pad weld is far less than stress at the ID of nozzle base metal. This tendency is also observed from many reports on cracked CRDM and J-welded instrumentation nozzle which state that the PWSCC cracks are axial in the nozzle base metal and there are no PWSCC cracks in weld metal or along the weld heat affected zone.

EA-A600-15 Sheet _1L Rev # _a __

. ~:i11

.\\it:/~

• • =mnrr llD SADES NUCLEAR PLANT ANALYSIS CONTINUATION SHEET

• Figure 6. External pressure "p" and q9op stress in a thick cyl~nder of the same size of the TE nozzle.

• - 1.315 a.---

2 b := 0.815 2

p :=-1 2

-p*2*a crh:=--

2.

2 a - b cr h = 3.247 TE outside radius TE inside radius External pulling pressure load Maximum hoop stress at b (Roark's Handbook)

Based on the difference of welding residual stress in weld pad and in the TE nozzle, it is concluded that the weld pad is less susceptible tO PWSCC than the TE nozzle base metal. A through-wall axial crack in TE-0I01 nozzle would happen well before a crack *through the weld pad and nozzle interface. Additionally, due to non-uniform stress distribution along the circumference of the weld root, there is no potential for 360 degree PWSCC initiation and growth in the weld pad/nozzle interface. Thus cracks could be detected from the outside before a complete failure of the weld pad.

EA-AS00-15 Sheet _11_ Rev # _o __

5.0 4.4.2 4.4.3 P.SADES NUC~EAR PLANT ANALYSIS CONTINUATION SHEET Component Thickness:

The nominal thickness of the weld pad is 0.5 inches which is twice the t!ll,_~kness of the TE-0101 nozzle wall thickness. This is significant since it results in longer PWSCC propagation time to leakage and also lower crack growth rate in the weld pad when compared to the TE nozzle. The crack growth rate is a function of the applied mode I stress intensity factor K1 which is in tum a function of the crack depth vs.* thickness ratio. For the same crack depth and stress condition, the crack growth rate of the TE-nozzle is higher than the growth rate of the weld pad.

Based on the component thickness, it is concluded that leakage through axial cracks in TE-0101 nozzle would happen well before

. the leakage through any through wall crack in the weld pad.

Susceptibility of Alloy 600 base metal and Alloy 182182 weld metal:

The photomicrographs of many failed sections of the 1993 Palisades PORV nozzle in Reference 2.8 show that several large and small PWSCC initiation sites on the ID surface of the Alloy 600 base metal but no crack was found in the weld metal ID surface. PWSCC propagated along the fusion line of the Alloy.1~2/82 weld metal with multiple branches into the Alloy 600 base metal but no branch penetrated the weld metal. Therefore, Alloy 182/82 weld metal is less susceptible to PWSCC than Allpy 600 material of the PORV nozzle safe-end. Although, the material of the TE nozzle is not from the same heat of the PORV safe-end material, it is expected that the same tendency is applicable for the TE nozzle base material ~hich also failed due to PWSCC.

Assessment for TE-0102:.

During the 1993 refueling outage, leakage was also found at the TE-O 102 nozzle located at the lower part of the pressurizer. Similar weld pad desiined repair was performed for the TE-0102. However, no EDM cut was performed since the thermal fatigue loading conditions at TE-0102 are less than the conditions of the TE-0101. Since the TE-0102 nozzle body is still intact, both J-weld and pad are structural weld.

Ther~fore, even if the pad weld were completely severed the J weld is

. still present to prevent the TE nozzle from any possible ejection. Thus, such a failure results in small leakage from TE-0102 nozzle. It is concluded that visual inspection is adequate to detect the leakage.

EA-A600-15 Sheet __1L Rev # _a __

g

PALISADES NUCLEAR PLANT ANALYSIS CONTINUATION SHEET 6.0

==

Conclusions:==

All poteit'ffal failure modes of the existing TE-0101 and TE-0102 nozzles were investigated. The original J-groove weld design was compared with the pad weld repair performed in 1993. Also, the susceptibility of the pad weld and the TE nozzle material was compared. From these comparisons it is concluded that:

The TE nozzle material at the weld pad is less susceptible to

.PWSCC than that near the original J-weld. Hence it is expected that the time for an ID initiated PWSCC to develop a leakage near the weld pad will be longer than the time it took to develop the axial cracks found in the 1993 outage.

An axial though wall cra<;:k in the nozzle body is more likely to happen before a crack through the interface between the nozzle body and the weld pad. Potential leakage through such axial cracks at TE nozzles can be easily detected by visual inspection before failure.

There is no structural or short-term safety concern for TE-0102 nozzle.

Jn addition, the corrosion of the pressurizer base metal exposed to borated primary water is not a concern for the remaining life of the plant.

• 7.0 Attachments:

A: EA Checklist Reviewer's Comments and Resolution Technical Review Checklist B: Safety Evaluation EA-AS00-15 Sheet __.1L Rev # _o __

,(.

ATTACHMENT A.

EA Checklist Reviewer's Comments and Resolution Technical Reviewer's Checklist I.

EA-AS00-15 Attachment A Page A1 of A4

I PALISADES NUCLEAR PLANT E GINEERING ANALYSIS CHECKLI EA -A600-15 Aflac/rrnent A

/t)ye A2 of Af-EA -

A6oo-/5 REV_0 __

Affected Revision Items Affected By This EA Yes No Required

.Identify**.-

Closeout

-~:

1.0 Other EAs 0

~

2.0 Design Documents Electrical E-38 through E-49 0

!XI 3.0 Design Documents Mechanical M239-M246, M249, M257-M261, M660, M664-M666 0

!XI 4.0 LICENSING DOCUMENTS 4.1 Final Safety Analysis Report (FSAR) 0 4.2 Technical Specifications 0

~

4.3 Standing Order 54 0

~

5.0 PROCEDURES 5.1 Administrative Procedures 0

~

5.2 Operating Procedures (SOP,EOP, ONP, etc) 0 00 5.3 Working Procedures 0

~

5.4 Tech Spec Surveillance Test Procedures 0

~

6.0 OTHER DOCUMENTS 6.1 Q-List 0

IE 6.2

• Plant Drawings 0

l:&1 3

Equipment Data Base 0

~

v.4 Spare Parts (Stock/MMS) 0

~

6.5 Fire Protection Program Report (FPPR) 0

~

6.6 Design Basis Documents 0

181 6.7 Operating Checklists 0

IS?J 6.8 SPCC/PIPP Oil and Hazardous Material Spill Prevention Plan 0

IS?J 6.9 EEO Documents 0

181 6.10 MOV Program Documents (Voltage, thrust, weak link, etc) 0 IXI 6.11 Work Instructions 0

~

6.12 Other 0

0 Do any of the following documents need to be generated as a result of this EA:

Yes No

1.

Corrective Action Document?

0 IXI Reference

2.

EEO Evaluation Sheet?

0

~

Reference

3.

Safety Evaluation?

~

0 Reference

4.

Design Basis Document Change Request?.

0

-~

Reference

5.

FSAR Change Request?

0 gj Reference

6.

Verification Test Procedure for Changes to the Design Basis?

0

~

Reference Is PRC Review of this EA Required?

0 IXI

)mpleted By

~"""" ~

J~ P. H. Holt fl/~

Date B-3 O - qs-

-.J

~

~ j /

S. C. R.AMA I *1.*.1,.. ~.tvt I Technical Reviewed By Date 3-.3D -'/£)

/

Identify Section, No, Drawing, Document etc.

@~*

.'"S

£A~A600-1s Afhdment-A

~e A3 of A4-ENGINEERING ANALYSIS REVIEW SHEET Title Justification for Continued Operation of Pressurizer TE-0101 &

EA N~r Revision N~r TE-0102 Nozzles as Repaired During the 1993 Refueling Outage EA-A600-15 0

Page 1 of 1 Item Page, Line, or

• 1' I

N~r Section N~r Cooments Respons'e or Resolution 1

General This calculation provides a qualitative discussion towards the No resolution necessary.

justification for continued operation of the TE nozzles along with results from detailed Fracture Mechanics Assessments and Fatigue Evaluations. It is understood that the comparisons provided in this ~ngineering Analysis i~ from engineering judgement and past experience with PWSCC in Alloy 600 components.

2 General

• Please incorporate the editorial changes marked up in the body of*

~l'JCoA Foo f'A "IE I>

All MA P..>:::.E 1) the calculation.

3 Pages 6 The qualitative comparison of welding residual stresses in the COMM f,..;7 IS.

lN(OI(. Po R.A1l D,

and 7 J-groove weld and the pad weld seems reasonable.

However. it g'

fOA. i11E R.E~PON.Sf.

should be noted that this original J-groove weld would have SFf PA~*~

undergone some stress relief process as a part uf the Pressurizer fabricati.on.

How would this stress relief affect the residual stresses in the J-groove weld? Also. include a discussion on the effect of such stress relief on this comparison.

~~

~~(

• ~h*k*t.... ;... WVI' Organization Date lnitia.tor

p. H.

HOtfNG-Date Techi~' w. Supervisor Date 3/30/95

~~/t~

3/!i0/1S-3/36/95" SCRamal i ngam NECO-Design Eng.

v

).

)

.TECHNICAL REVIEW CHECKLI" EA - A6oo-15' REV 0

£A -A600-IS-Alto.cl11Y1ent-A Page A 4-of It+

Thi.s checklist provides guidance for the review of engineering analyses. Answer questions Yes or No,.or NIA if they do not apply. Document all comments on a EA Review Sheet. Satisfactory resolution of comments and completion of this checklist is noted by the Jechnical Review signature at the bottom of this sheet.

1.

Have the proper input codes, standards and design principles been specified?

2.

Have the input codes, standards and design principles been properly applied?

• 3.

Are all inputs and assumptions valid and the basis for their use documented?

4.

Is Vendor information used as input addressed correctly in the analysis?.

5.

If the anaFysis argument departs from Vendor Information/Recommendations, is the departure justification documented?

6.

Are assumptions accurately described and reasonable?

7. *Are the design basis changes permitted by this EA bounded by the applicable Safety Review/Evaluation? *

8.

Are all constants, variables and formulas correct and properly applied?

9.

Have any minor (insignificant) errors been identified? If yes; Identify on the 3110 Form and justify their insignificance'.

10.

Does analysis involve welding? If Yes; verify the following

• information is accurately represented on the analysis drawing (Output document).

• Type qf Weld

• Size of Weld

• Material Being Joined

• Thickness of Material Being Joined

• Location ofWeld(s)

• Appropriate Weld Symbology

11.

Has the objective of the analysis been met?

12.

Have administrative requirements such as numbering and format been satisfied?

~

Technical Reviewer

• ~.RAMAL1NGAM Date 3/3o}'l5*

(Y,N,N/A)

N/A N/A y

N/A N/f\\

y N[A

~y y

N y

y 1 *

\\

~I ATTACHMENT B Safety Evaluation EA-A600-15 Page 81 of 82

.\\'.

PALISADES NUCLEAR PLAN*T 10CFR50.59 SAFETY REVIE.

tj~ t>/_#;;i.

PS&L Log No *~ OVI t!1l<'Y'

.)

l~em Identification: No EA A600 01;::EA-A600-02. EA ASQQ 07. EA-A600-15 Rev _o_ Title~_

SE EA-A660-01: Service Life Ana!Ysis of Pressurizer Temgerature Element TE-0101 Weld Modification REV EA-A600-02: Service!,ife Ana!Ysis of Pressurizer Temgerature Element TE-0102 Weld Modification EA-A600-07: Service Life Assessment of the TE Nozzle Penetrations

~

EA-A600-15: Justification for Continued Ogeration of Pzr. TE-0101 and TE-0102 Noizles as regaired during the 1993 Refuling outage.

Describe Issue/Change:

Code fatigue evaluations and Fracture Mechanics assessment of the TE nozzles.and assessment of base-metal corrosion due to leaking borated water have been gerformed by the four analyses.

Reason for Issue/Change: To grovide justification for continued safe ogeration till the end of licensed life with the nozzles for TE-0101 and TE-0102 as regaired *during the 1993 REF OUT.

1.

Does the item involve a change to procedures as described in the FSAR?

Yes No FSAR Sections affected None FSAR Sections reviewed 1.2, 1.4, 4.3, 4.5, 5.1, 5.9, 6.1, 6.9.1, 7.4, 7.5, 7.6, x

14.3 14.14 14.17 Tables 1-2 3-3 3-11 4-3 4-8 4-15 I

• 2.

Does the item involve a change to the facility as described or implied in the FSAR?

FSAR Sections affected None x

FSAR Sections reviewed Same as in item 1 above DBD Sections reviewed 3.3.1, 3.3.2, 3.3.3, 3.3.4, 3.3.5

3.

Does the item.involve a test or experiment not described in the FSAR?

FSAR Sections affected None FSAR Sections reviewed Same as in item 1 above x

DBD Sections reviewed

. 3.3.1, 3.3.2, 3.3.3, 3.3.4, 3.3.5

4.

Should the Technical Specifications or any of their Bases be changed in conjunction with this item?

TS Sections affected None x

TS Seetions reviewed 2, 3, 4, 5, Standing Order 54 & 62 Justify "NO" answers below if logic is not obvious:

These nozzles were modified in 199~ per SC-93-087 and a justification for continued operation for one fuel cycle was provided. *These 4 analyses provide the justification for continued operation for the TE nozzles till the end of licensed life. The results of this analysis does not affect any sections of the FSAR, Technical Specifications or the Design Basis Documents. The analysis does not involve a change to the facility or a test or experiment. The results of the analysis does not change the operations of the temperature indicators.

If any Safety Review question listed above is answered "YES," perform a written USQ Evaluation according to Section 5.3.

If all Safety Review questions listed above are answered "NO," written USQ Evaluation is not required.

However, this attachment shall accompany other review materials for the item to document that a Safety Evaluation was not rec uir,ed. _,

~z: I lf-l'l-'f5

• • '~

SC ~

I ':ii:,./ f( S" Prepared By Date

-;viewed By Date