ML042240249

Request for Relief from ASME Section XI Code Requirements for Repair of Pressurizer Nozzle Penetrations

ML042240249
Person / Time
Site:	Palisades
Issue date:	08/09/2004
From:	Domonique Malone Nuclear Management Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
Download: ML042240249 (29)

Text

of NM C Commited to Nuclear ExceII Palisades Nuclear Plant Operated by Nuclear Management Company, LLC August 9, 2004 10 CFR 50.55a U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555 Palisades Nuclear Power Plant Docket 50-255 License No. DPR-20 Request for Relief from ASME Section Xl Code Requirements for Repair of Pressurizer Nozzle Penetrations Nuclear Management Company, LLC (NMC) is performing bare metal visual inspections of the pressurizer heater sleeves during the upcoming refueling outage at the Palisades Nuclear Plant, based on recent industry operating experience regarding Alloy 600 nozzles. NMC requires the enclosed relief requests in the event a nozzle is in need of a repair at Palisades. Therefore, NMC requests relief from certain sections of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV)

Code, Section Xl, 1989 Edition, as described in the attached enclosures.

Enclosure 1 requests relief from the ASME Code, Section Xl, IWA-4120, "Rules and Requirements." NMC proposes an alternative to the specified code requirements, in accordance with 10 CFR 50.55a(a)(3)(i). The basis for the relief is provided, describing that the alternative provides an acceptable level of quality and safety.

Enclosure 2 requests relief from the ASME Code, Section Xl, IWA-3300, "Flaw Characterization," IWB-3142.4," Acceptance by Analytical Evaluation," and IWB-3420, "Characterization." NMC proposes an alternative to the specified code requirements, in accordance with 10 CFR 50.55a(a)(3)(i). The basis for the relief is provided, describing that the alternative provides an acceptable level of quality and safety.

NMC will implement a Welding Services Incorporated/Structural Integrity Associates outer diameter pad plug repair design for the Palisades Nuclear Plant if a pressurizer nozzle repair is necessary.

4Ac7 27780 Blue Star Memorial Highway

• Covert, Michigan 49043-9530 Telephone: 269.764.2000

Document Control Desk Page 2 Relief is requested for the remainder of the current ten-year inspection interval, which will conclude on or before December 12, 2006.

NMC requests approval of the proposed relief requests by October 1, 2004, to support Palisades Nuclear Plant's upcoming refueling outage.

Summary of Commitments This letter contains no new commitment and no revisions to existing commitments.

A42 Daniel J. Malone Site Vice President, Palisades Nuclear Plant Nuclear Management Company, LLC Enclosures (2)

Attachment (1)

CC Administrator, Region Ill, USNRC Project Manager, Palisades, USNRC Resident Inspector, Palisades, USNRC

ENCLOSURE I RELIEF REQUEST #1: WELD PAD AREA PRESSURIZER VESSEL PENETRATIONS ASME Code Component Affected The affected components are the Palisades Nuclear Plant pressurizer vessel heater sleeves. The Palisades Nuclear Plant has 120 pressurizer heater sleeves penetrating the bottom head. The pressurizer heater sleeves are American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV)

Code,Section III, Subsection A (Class A) components.

Applicable Code Edition and Addenda

The applicable code edition and addenda for the pressurizer vessel heater sleeve repair is the ASME B&PV Code, Section Xl, 1989 Edition with no addenda. Palisades is currently in the third ten-year inservice inspection interval.

Applicable Code Requirement

The primary ASME code section for welding on code classified components such as the pressurizer is ASME Section Xl. Article IWA-4000 contains the rules for repair/replacement activities such as welding.

The applicable code requirement for the pressurizer lower head penetrations is ASME Section Xl, IWA 4120, 'Rules and Requirements," as follows:

(a) Repairs shall be performed in accordance with the owner's design specification and the original construction code of the component or system. Later edition and addenda of the construction code or of Section III, either in their entirety or portions thereof, and code cases may be used. If repair welding cannot be performed in accordance with these requirements, the applicable alternative requirements of IWA-4500 and the following may be used:

(1) IWB-4000 for Class 1 components (2) IWC-4000 for Class 2 components; (3) IWD-4000 for Class 3 components; (4) IWE-4000 for Class MC components; or (5) IWF-4000 for component supports.

(b) The edition and addenda of Section Xl used for the repair program shall correspond with the edition and addenda identified in the inservice inspection program applicable to the inspection interval.

(c) Later editions and addenda of Section Xl, either in their entirety or portions thereof, may be used for the repair program, provided these editions and addenda of Section Xl at the time of the planned repair have been incorporated by reference in amended regulations of the regulatory authority having jurisdiction at the plant site.

Page 1 of 6

The original construction code of record for the Palisades Nuclear Plant pressurizer vessel is ASME Section 1I1,Subsection A, 1965 Edition, including addenda through winter 1965.

Reason for Request

Nuclear Management Company, LLC (NMC) is requesting relief from ASME Section Xl, 1989 Edition, IWA-4120, pursuant to 10 CFR 50.55a(a)(3)(i),

because the alternative provides an acceptable level of quality and safety.

NMC requests relief from the applicable ASME Code requirements and requests to use the rules of ASME Code Case N-638, 'Similar and Dissimilar Metal Welding Using Ambient Temperature Machine [Gas Tungsten Arc Welding]

GTAW Temper Bead Technique," with exceptions noted in the "Proposed Alternative and Basis for Use" section below.

Repair welding per the construction code, ASME Section 1I1,Article 5, would require an 11000 F minimum postweld heat treatment. ASME Section Xl provides an alternative to this construction code requirement. ASME Xl, IWA-4530, which would eliminate the high temperature postweld heat treatment, would still require a 300OF preheat and a 4500 F to 5500 F post weld heat treatment. ASME Code Case N-638 will permit the use of the ambient temperature machine GTAW temper bead technique for the repairs. This technique provides an acceptable level of quality and safety.

Proposed Alternative and Basis for Use The topic of this relief request is repair welding on the P-Number 3 ferritic base material (SA-533 grade B) of the pressurizer in conjunction with repairs to degraded nozzles. The repair approach is shown in Figure 1 and consists of the following steps:

(a) Cutting the existing Alloy 600 nozzle outboard of the J-groove weld.

(b) Replacing the removed nozzle portion with an Alloy 690 backing plug.

(c) Installing a welded pad of weld metal over the nozzle opening and backing plug.

The weld pad deposit is performed with weld metal Alloy 52/152 having chemistry essentially equivalent to the Alloy 690 replacement nozzle material.

The application of the weld pad to the vessel head is a dissimilar material combination.

Page 2 of 6

J-Groove Weld

'Exidstg Alloy 600 nozzle Alloy 690 Backing

- Welded Pad Flaure 1 Justification for Relief NMC proposes to implement Code Case N-638, with exceptions to the following specifications:

(a) The 'Reply" section of the code case states, "... and may be made by the automatic or machine GTAW temper bead technique without the specified preheat or post-weld heat treatment of the Construction Code, when it is impractical, for operational or radiological reasons, to drain the component, and without the nondestructive examination requirements..."

(b) Examination section (4.0)(b) states, "The final weld surface and the band around the area defined in para. 1.0(d) shall be examined using a surface and ultrasonic methods when the completed weld has been at ambient temperature for at least 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />. The ultrasonic examination shall be in accordance with Appendix I."

(c) Examination section (4.0)(e) states, "Ultrasonic examination acceptance criteria shall be in accordance with IWB-3000."

Code Case N-638 has been accepted and approved for use by the NRC per Regulatory Guide 1.147, "Inservice Inspection Code Case Acceptability - ASME Section Xl Division 1," Revision 13. However, the proposed repair methodology takes exception to the Code Case specifications described in the following paragraphs.

Page 3 of 6

Code Case N-638 ReDIV Section The "Reply" section of Code Case N-638 did not clearly state the application for the code case. The reply section was clarified in Code Case N-638-1. Although the NRC has not approved Code Case N-638-1, it is identical to Code Case N-638 with the exception of the editorial changes made to the reply section of the code case.

Code Case N-638 states, "... may be made by the automatic or machine GTAW temper bead technique without the specified preheat or post-weld heat treatment of the Construction Code, when it is impractical, for operational or radiological reasons, to drain the component, and without the nondestructive examination requirements..."

Code Case N-638-1 states, "...may be made by the automatic or machine GTAW temper bead technique without the specified preheat or post-weld heat treatment of the Construction Code, when it is impractical to drain the component or impractical for radiological reasons. The nondestructive examination requirements..."

This editorial change does not affect the process of similar and dissimilar metal welding using ambient temperature machine GTAW temper bead welding.

Examination Volume Code Case N-638, section 4.0(b), requires that the final weld and the band around the area defined in paragraph 1.0(d), shall be examined using surface and ultrasonic methods when the completed weld has been at ambient temperature for at least 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />. Code Case N-638, section 1.0(d), defines this inspection requirement as the area to be welded and a band around the area, of at least 1-1/2 times the component thickness, or five inches, whichever is less.

The vessel thickness at the lower head is nominally 4%-inch. Application of this code case requirement would require an examination area including the weld pad area and an area extending five inches around the weld pad area. The weld pad has square dimensions of approximately 31/4 inches on a side. Adding the examination area of five additional inches around provides a total area defined by a square of dimensions of approximately 131/4 inches x 131/4 inches. Other pressurizer penetrations would interfere with the code case defined area.

As an alternative to the code case requirement, NMC will use the examination area defined by ASME Section III, 1989 Edition, NB-5244, 'Weld Buildup Deposits at Openings for Nozzles, Branch, and Piping Connections." This portion of the code case indicates that when weld buildup deposits are made to a surface, the weld buildup deposit, the fusion zone, and the parent metal beneath the weld buildup deposit shall be ultrasonically examined to ensure freedom from lack of fusion and laminar defects. Surface examination will be performed as directed by the code case.

Page 4 of 6

Ultrasonic Accegtance Criteria Code Case N-638 specifies in the examination section, 4.0(e), to use the acceptance criteria of ASME Section Xl, IWB-3000, for the ultrasonic examination. ASME Section Xl, IWB-3000, refers to Table IWB-3410-1, for the acceptance standards to be applied for each examination category. The examination category for the pressurizer heater sleeves is category B-P, as established in the Palisades Nuclear Plant ISI program. In accordance with Table IWB-341 0-1, the only acceptance standard given for examination category B-P items is for a visual examination (VT-2) during the pressure test. There is no acceptance criterion for ultrasonic examination of Category B-P items specifically provided for in the ASME Section Xl code.

Code Case N-638, section 4.0(e), allows for additional acceptance criteria to be specified by the owner to account for differences in weld configuration. Due to the configuration of the weld pad repair, guidance is taken from ASME Section III, Article 5, which states that any crack, lack of fusion, incomplete penetration, inclusion, or cavity, which is indicated by a reflection equal to or greater than 80% of the applicable reference hole reflection, and which has a linear dimension as indicted by transducer movement exceeding % -inch for thicknesses over 2 1/4-inch, is unacceptable. These criteria are specified as: any discontinuity interpreted to be a crack or incomplete penetration will be unacceptable, regardless of discontinuity or signal amplitude (i.e., no cracks are allowed). Therefore, to meet the Code Case N-638 requirement, NMC will perform ultrasonic examination of the Category B-P repair welds performed per Code Case N-638, and will use an acceptance standard of no cracks allowed.

Implementing the repairs as proposed using the ambient temperature machine GTAW temper bead technique will provide an acceptable level of quality and safety, as required by 10 CFR 50.55a(a)(3)(i).

Duration of Proposed Alternative NMC requests approval of the proposed alternative for the remainder of the third ten-year interval of the Inservice Inspection Program for Palisades, which will conclude on or before December 12, 2006.

Precedent Southern California Edison submitted a relief request for San Onofre Nuclear Generating Station (SONGS), Units 2 and 3, dated March 22, 2004 (ADAMS Accession # ML040850464). The relief requested approval to perform repair welding using the ambient temperature machine GTAW temper bead technique Page 5 of 6

without the pre and post weld heat treatment. The SONGS relief request is similar to Palisades in that the request involved similar pressurizer locations and materials. The SONGS relief requests also involved the use of Code Case N-638-1, and also requested an alternative to the ultrasonic acceptance criteria, similar to Palisades.

Page 6 of 6

ENCLOSURE 2 RELIEF REQUEST #2: FLAW CHARACTERIZATION PRESSURIZER VESSEL PENETRATIONS ASME Code Component Affected The affected components are the Palisades Nuclear Plant pressurizer vessel heater sleeves. The Palisades Nuclear Plant has 120 pressurizer heater sleeves penetrating the bottom head. The pressurizer heater sleeves are American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV)

Code,Section III, Subsection A (Class A) components.

Applicable Code Edition and Addenda

The applicable code edition and addenda for the pressurizer vessel heater sleeve repair is the ASME B&PV Code, Section Xl, 1989 Edition with no addenda. Palisades is currently in the third ten-year inservice inspection interval.

Applicable Code Requirement

The applicable code requirement for the pressurizer vessel head penetrations is ASME Section XI. Table IWB-2500, examination category B-E, "Pressure Retaining Partial Penetration Welds in Vessels," Item B4.1 1, is applicable to the inservice examination of the pressurizer vessel lower head to penetration welds.

IWA-3300, 'Flaw Characterization," IWB-3142.4, "Acceptance by Analytical Evaluation," and IWB-3420, "Characterization," are applicable to any flaws discovered during inservice inspection. Specifically:

(a) Subarticle IWA-3300 contains a requirement for flaw characterization.

(b) Sub-subparagraph IWB-3142.4 allows for analytical evaluation to demonstrate that a component is acceptable for continued service. It also requires that components found acceptable for continued service by analytical evaluation be subsequently examined in accordance with IWB-2420(b) and (c).

(c) Paragraph IWB-3420 requires the characterization of flaws in accordance with the rules of IWA-3300.

The original construction code of record for the Palisades Nuclear Plant pressurizer vessel is ASME Section III, Subsection A, 1965 Edition, including addenda through winter 1965.

Page 1 of 5

Reason for Request

Nuclear Management Company, LLC (NMC) is requesting relief from ASME Section Xl, IWA-3300, IWB-3142.4, and IWB-3420, pursuant to 10 CFR 50.55a(a)(3)(i). The above sections would require characterization of a flaw existing in the remnant of the J-groove weld that will be left on the pressurizer vessel lower head if a heater sleeve is partially removed.

If inspection of the pressurizer vessel lower head nozzle penetrations reveals flaws affecting the J-groove attachment welds, it may be impractical to characterize these flaws by non-destructive examination (NDE) and it may be impractical to perform any successive examinations of these flaws. The original nozzle to pressure vessel lower head weld configuration is extremely difficult to ultrasonically (UT) examine due to the compound curvature and fillet radius. The configuration is not conducive to UT due to the configuration and dissimilar metal interface between the Ni-Cr-Fe weld and the low alloy steel pressure vessel lower head. Furthermore, due to limited accessibility from the pressurizer vessel lower head outer surface and the proximity of adjacent nozzle penetrations, it is impractical to scan from this surface on the pressurizer vessel lower head base material to detect flaws in the vicinity of the original weld. These conditions preclude ultrasonic coupling and control of the sound beam in order to perform flaw sizing with reasonable confidence in the measured flaw dimension.

Therefore, presently, the technology does not exist, to characterize flaw geometries that may exist therein.

NMC is proposing an alternative, as discussed below, for not performing flaw characterization, as required in the ASME Code, Section Xl. This alternative provides an acceptable level quality or safety.

Proposed Alternative and Basis for Use The alternative requirements are:

(a) Subarticle IWA-3300 contains a requirement for flaw characterization by inservice examinations. In lieu of this requirement, a conservative worst-case flaw characterization is assumed, contingent upon the discovery of a through wall leak.

This flaw is assumed to extend from the J-groove weld surface to the pressurizer vessel lower head, low alloy steel, base material interface. A flaw evaluation, based on the worst-case flaw postulation, has been performed in accordance with the criteria set forth in IWB-3600. This flaw evaluation considered the effects of fatigue crack growth and corrosion under a reactor coolant environment. It established the minimum remaining service life of the pressurizer vessel lower head.

Page 2 of 5

(b) Sub-subparagraph IWB-3142.4 allows for analytical evaluation to demonstrate that a component is acceptable for continued service. It also requires that components found acceptable for continued service by analytical evaluation be subject to successive examination. Analytical evaluation of the worst-case flaw referred to above demonstrated the acceptability of continued operation. However, because of the impracticality of performing any subsequent inspection that would be able to characterize any remaining flaw, successive examination will not be performed.

(c) Paragraph IWB-3420 requires the characterization of flaws in accordance with the rules of IWA-3300. As previously stated, a conservative worst-case flaw is assumed contingent upon discovery of a through wall leak in the J-groove weld. A flaw evaluation has been performed based on that conservative postulation of the flaw characterization.

Justification for Relief Due to the susceptibility of Alloy 600 pressurizer vessel lower head partial penetration nozzles to primary water stress corrosion cracking (PWSCC), an ambient temperature machine gas tungsten arc welding (GTAW) temper bead repair process has been developed for pressurizer vessel lower head nozzles.

The repair approach is shown in Figure 1 and consists of the following steps:

(a) Cutting the existing Alloy 600 nozzle outboard of the J-groove weld.

(b) Replacing the removed nozzle portion with an Alloy 690 backing plug.

(c) Installing a welded pad of weld metal over the nozzle opening and backing plug.

The weld pad deposit is performed with weld metal Alloy 52/152 having chemistry essentially equivalent to the Alloy 690 replacement nozzle material.

The application of the weld pad to the vessel head is a dissimilar material combination.

Page 3 of 5

- EdsV Moy 600 nozzle Aloy 690 Backing Wekied Pad Figure 1 The original J-groove weld will not be removed. Since a potential flaw in the J-groove weld cannot be sized by currently available NDE techniques, it is assumed that the "as-left" condition of the remaining J-groove weld includes degraded or cracked weld material extending through the entire J-groove weld and Alloy 182 butter material.

Westinghouse developed an evaluation based on stress corrosion crack growth in carbon and low alloy steels through the Alloy 600 remnants and associated weld metal in place after weld repairs. WCAP-15973-P, "Low-Alloy Steel Component Corrosion Analysis Supporting Small-Diameter Alloy 600/690 Nozzle Repair/Replacement Programs," Revision 1, is a bounding ASME Code Section Xl analysis for the repair of Combustion Engineering design of hot leg piping resistance temperature detector (RTD) and sampling nozzles, pressurizer instrument nozzles and pressurizer heater sleeves. WCAP-1 5973-P, Revision 1, is supported by Westinghouse calculation CN-CI-02-71, "Summary of Fatigue Crack Growth Evaluation Associated with Small Diameter Nozzles in CEOG Plants," Revision 1. These reports are applied to the repairs of the small-bore nozzles whose pressure boundaries have been breached by a PWSCC attack in the J-groove weld penetration areas. These reports provide justification for leaving a flaw in the nozzle remnant. These reports show that the ASME Section Xl flaw acceptance standards for preventing non-ductile failure would be satisfied for the remaining life of the plant. This WCAP was submitted to the Nuclear Regulatory Commission (NRC) on May 20, 2004 for review and approval.

WCAP1 5973-P shows that the preferential direction for cracking would be axial relative to the nozzle. It is postulated that a crack in the Alloy 182 weld metal would propagate by PWSCC, through the weld and butter, to the interface with the head material, where it is fully expected that such a crack would then blunt, or arrest, as it interfaces with low alloy steel.

Page 4 of 5

In order to make use of WCAP-15973-P, Revision 1, and CN-CI-02-71, Revision 1, Palisades performed a site-specific analysis to ensure that the plant is operated such that the pressure and temperature heat-up and cool-down profiles do not exceed the analyzed profile applied in CN-CI-02-71. This site-specific analysis provided justification for use of WCAP-1 5973-P and CN-CI-02-71, for leaving the remnant of a leaked small bore nozzle in place, at the Palisades Nuclear Plant. This site-specific analysis is included as .

Based on the information presented, and pursuant to 10 CFR 50.55a(a)(3)(i),

NMC requests approval on the basis that the alternative provides an acceptable level quality or safety. The evaluations discussed above provide justification for not performing flaw characterization, as required in the ASME Code, Section Xl 1989 Edition, IWA-3300, IWB-3142.4 and IWB-3420.

Duration of Proposed Alternative NMC requests approval of the proposed alternative for the remainder of the third ten-year interval of the Inservice Inspection Program for Palisades, which will conclude on or before December 12, 2006.

Precedents Florida Power and Light submitted a relaxation request to Order EA-03-009, for Turkey Point, Unit 3. This relaxation request was based on industry report, Material Reliability Project MRP-55. The NRC approved this relaxation request on March 20, 2003 (ADAMS Accession # ML030790501), concurrent with their continued review of MRP-55. The safety evaluation was contingent upon one condition related to the final review of MRP-55. This precedent applies to NMC in that Palisades is requesting approval based on industry report, WCAP-1 5973-P.

Florida Power and Light submitted relief requests for full-nozzle repair of Alloy 600 small bore nozzles without flaw removal, for Saint Lucie Nuclear Plant, Units 1 and 2. The NRC approved these requests on May 23, 2003 (ADAMS Accession # ML031470003). Reference is also made to WCAP-15973-P, Revision 00. This precedent applies to NMC in that Palisades is requesting repair of Alloy 600 small bore nozzles without flaw removal.

Arizona Public Service Company submitted relief requests for repair of Alloy 600 small bore pressurizer sleeves, for Palo Verde Nuclear Generating Station, Units 1, 2 and 3, dated June 15, 2004 (ADAMS Accession # ML041750296). The relief request references WCAP-1 5973-P, Revision 1. This precedence applies to NMC in that Palisades is requesting relief for repair of pressurizer nozzles and references WCAP-1 5973-P, Revision 1.

Page 5 of 5

ATTACHMENT 1 RELIEF REQUEST #2 PALISADES NUCLEAR PLANT ANALYSIS, EA-A600-2004-1 1, ULOW ALLOY STEEL COMPONENT CORROSION ANALYSIS SUPPORTING SMALL DIAMETER ALLOY600/690 NOZZLE REPAIR/REPLACEMENT" (ATTACHMENT 3 OF THE ANALYSIS IS ALSO INCLUDED) 15 Pages Follow

Proc No 9.11 Attachment I PALISADES NUCLEAR PLANT Revision 15 ENGINEERING ANALYSIS COVER SHEET Page 1 of I NUCLEAR EA- A600-2004-01 MANAGEMENT Total Number of Sheets.l >

COMPANY Title Low-Alloy Steel Component Corrosion Analysis Supporting Small-Diameter Alloy 600/690 Nozzle Repair/Replacement INITIATION AND REVIEW Calculation Status Preliminary Pending Final Supeseded 01 0 P 0 __ _

Initiated nit Review Method Technically Rev'r Sup v Rev Appd _ Reviewed Appd or

1. Description By Aft Detail Qual By SIDR IBy Date Calc Revw Test By Date Appd 0 Original Issue 6 ca. ?1/f c __ 4fr2 LIST OF ATTACHMENTS

1. Westinghouse Report WCAP-15973-P,Low-Alloy Steel Component Corrosion Analysis Supporting Small Diameter Alloy 600/690 Nozzle Repair/Replacement Programs, Rev. 1.

2. Westinghouse Calculation Note Number CN-CI-02-71 Rev.1, Summary Of Fatigue Crack Growth Evaluation Associated With Small Diameter Nozzles In CEOG Plants.

cPAL-OA4 48

3. Westinghouse Letter, L -. 4 I,Corrosion Analysis of the Pressurizer Side Shell Nozzle, dated Jfie25-,20e4-. A t-Oh

- , e /V > b 1 ofX)

O t>-ca Zc+-.

( JWes-+ tc Hem coAx- z6 A, i

4. 50.59 Screen.

5. EA Checklist, 3110 Form, Technical Review Checklist AUG o52204 L ERC - PAL

PALISADES NUCLEAR POWER PLANT PALISADES ANALYSIS CONTINUATION SHEET EA-A600-2004-01 Page 2_ of l__

1.0 Objectives

The objectives of this analysis are to:

(1) Provide a cover EA for the vendor reports in Attachment 1 (referred as WCAP- 15973 hereinafter) and Attachment 2 (referred as CN-CI-02-71 hereinafter). This EA is prepared in accordance with the Reference 2.1 requirements for Vendor Technical Evaluations and Reports. Section 6.2.6 of the Reference 2.1 requires this cover EA to evaluate the effects that the WCAP-15973 and CN-CI-02-71 have on the design of the plant.

WCAP-15973, performed on a generic industry level, is a bounding ASME Code Section Xl analysis for the repair of the Combustion Engineering design of hot leg piping RTD and sampling nozzles, pressurizer instrument nozzles and pressurizer heater sleeves. CN-CI-02-71 is a supporting calculation for WCAP-15973 and is considered in this EA as an integral part of WCAP-15973. These reports are applied to the repairs of the small-bore nozzles whose pressure boundaries have been breached by the PWSCC attack in the J-weld penetration areas. Generally speaking, the flaws in a nozzle remnant (J-weld included) are difficult to remove and these reports provide a justification for leaving a flaw in the nozzle remnant. The justification includes the evaluations of the effects of corrosion, stress corrosion cracking, fatigue crack growth and environmental factors. More comprehensive descriptions of the scope of each vendor reports are in the front sections of these reports.

(2) Supercede the engineering analysis of Reference 2.2. In essence, Reference 2.2 is the cover EA for Rev.0 of the Attachments I and 2 reports. The difference is the Reference 2.2 fatigue crack analysis was prepared specifically for Palisades' pressurizer temperature nozzles repair. Both of the pressurizer temperature nozzles were repaired under Specification Change No.SC-93-087 in 1993. This EA provides broader applications than the temperature nozzles. Besides that, this EA also corrected several analysis deficiencies from the previous analysis, i.e. Revision I of WCAP- 15973 and Reference 2.2. A description of the analysis deficiencies is presented in WCAP-1 5973 Executive Summary section.

(3) Close Corrective Action CA024362. The analysis deficiencies mentioned in the preceding paragraph were officially communicated to Palisades via Westinghouse Nuclear Safety Advisory Letter NSAL-04-4. The NSAL-04-4 was placed in the corrective action program to ensure the proper evaluation and actions be performed.

Noting that the original overall conclusion to leave small bore piping J-Welds in service with pre-existing flaws is unchanged. The corrective action requires the maintaining of configuration control by replacing Reference 2.2 with this cover EA.

PALISADES NUCLEAR POWER PLANT PALISADES ANALYSIS CONTINUATION SHEET EA-A600-2004-01 Page _ of I Z_

2.0 References 2.1 Administrative Procedure No. 9.11, Engineering Analysis, Revision 15.

2.2 EA-SC-93-087-06, Evaluation of Fatigue Crack Growth of Postulated Flaw at Repaired Pressurizer Temperature Nozzle, Rev.0.

2.3 DBD 2.11 Rev. 1, Pressurizer Pressure Control.

2.4 System Operating Procedure SOP-I, Rev. 54.

2.5 Combustion Engineering Specification No. 70P-01, Engineering Specification for A Pressurizer Assembly for Palisades Nuclear Plant, Vendor File Ml -L-A, Rev.3.

2.6 Combustion Engineering Owners Group report CEN-NPSD 546-P. Pressurizer Surge Line Flow Stratification Evaluation, Rev. 1-P.

2.7 EOP Supplement 1, Pressure Temperature Curves, Rev. 5 2.8 Technical Specification Amendment 189.

2.9 ASME B&PV Code, 1989 Edition 2.10 Drawing Ml-LA-5003-1, Bottom Head Forming and Welding, Rev.4.

2.11 Drawing MI-D-106, Piping Assembly Details, Rev. 9.

2.12 Drawings: Ml-LA-989, Internal Details, Rev. 7. Ml-LA-985, Nozzle Details, Rev.13 2.13 Reactor Log Book, current record of the Perpetual Log of Pressurizer Spray Cycle with High Delta Temperature, up to date.

2.14 ABB-CE Report, Pressurizer Spray Nozzle Fatigue Evaluation. dated October 1991 Cartridge/Frame (C775/0650).

2.15 EA-A-PAL-92-095-01, Pressure-Temperature Curves and LTOP Setpoint Curve for 19 Maximum Reactor Vessel Fluence of 2.192 x 10 Neutrons/cm 2 , Rev.0.

2.16 Palisades 40 Year Master Inservice Inspection Plan, Revision 10.

2.17 Drawing: Ml -LA-982, Vessel Forming and Welding, Rev. 10.

2.18 Combustion Engineering Report Palisades PCS System Description Revision 0.

2.19 Emergency Operating Procedure EOP-I Supplement 1 Revision 5.

PALISADES NUCLEAR POWER PLANT ANALYSIS CONTINUATION SHEET EA-A600-2004-01 Page 4 of I___

3.0 Acceptance Criteria 3.1 Leaving flaws in a nozzle remnant by analytical evaluation is permitted by ASME XI Para.

IWB-3132.4. Referring to IWB-3132.4, the acceptance criteria for the analytical evaluation of the flaw are given in ASME XI Para. IWB-3600 and in Regulatory Guide 1.161 for Elastic-Plastic Fracture Mechanics approach. The ASME Code acceptance criteria for the flaw evaluation are presented in Section 4.0 of the CN-CI-02-71. WCAP-15973 evaluates the corrosion of low alloy steel in the primary coolant system. The corrosion allowables are described in Section 2.4 of WCAP-15973, which was established based on ASME Section III design requirements. Compliance to the corrosion and flaw growth acceptance criteria has been demonstrated in WCAP- 15973 and are not further evaluated by this cover EA.

3.2 In order to make use of WCAP- 15973 and its supporting calculation CN-CI-02-71, Palisades must ensure that the plant is operated such that the pressure and temperature heat-up and cool-down profiles do not exceed the analyzed profile applied in CN-CI-02-71 (see Section 3.2 of CN-CI-02-71). The pressure and temperature profile applied in CN-CI-02-71 is shown in the Figure 1 below.

Transient Definitions (Temperature vs Time) 700 600 UL.

Ch500 W

2 400 e 3Q I-E 300 5 200 100 0

0 3600 7200 10800 14400 18000 2160C Time (seconds)

Figure 1 Fluid Temperature vs Time

PALISADES NUCLEAR POWER PLANT PALISADES ANALYSIS CONTINUATION SHEET EA-A600-2004-O1 Page _of IZ 4.0 Inputs 4.1 The in-surge transients described in Figure 1 are not part of Palisades design basis, except for the purpose of the evaluation of this EA. Westinghouse has not revealed the mechanics of the in-surge transients and has not incorporated such transients in design requirements.

4.2 Palisades operators are sensitive to the adverse effects on components due to a large differential temperature between the pressurizer and the PCS (PCS-PZR AT). Palisades operations of the heatup and cooldown of the PCS result in, relative to the industry's norm, a small PCS-PZR AT. The relatively small AT is achieved by a combination of the high pressurizer heater input and the use of a continuous spray flow for pressure control. A brief description of the Palisades operation in this respect is below.

To heat up the PCS from shutdown condition, all the pressurizer heaters are energized.

[SOP-1 7.1.3.c.2]. There is a total of 1500 KW (nominal) of heater capacity available. 90%

of the total heater capacity is powered by fixed input and the other 10% heater capacity is powered by variable power input [DBD 2.1 1, Section 3.2.1]. The fixed input heater capacity, which amounts to 1350 KW (nominal), stays energized through out the fuel cycle

[DBD 2.11 Section 3.3.1.4] and provides a constant high heat input to the pressurizer.

Palisades starts the Primary Coolant Pumps (PCPs) when the pressurizer is solid which provides both the driving force for the spray flow and the energy to heat up the PCS simultaneously with the pressurizer. Both factors of the high pump heat input and a continuous spray flow reduce the PCS-PZR AT. A limit of the PCS-PZR AT is included in SOP-1. Section 7.1.3 of the operating procedure requires that [Reference 2.4] when PCS is greater than 1850 F the maximum delta between the lowest cold leg temperature and PZR vapor temperature be less than 2000 F.

To cool down the PCS, SOP-I Section 7.1.4.o.5 says to "MAINTAIN maximum possible PZR heaters energized while controlling pressure with pressure control through out the collapsing the bubble." That is, spray flow must be available for pressure control. The PCPs are operated until near the end of the cool down process when the PCS is at about 150° F [SOP-I Section 7.1.4]. Like the heat-up process, the heat input from the PCP operation during the cool-down reduces the magnitude of the PCS-PZR AT.

4.3 Palisades operators recognize the potential harmful effects of a large PCS-PZR AT on the equipment. To minimize thermal transients on the spray nozzle, the operating procedures require operators recording and trending the occurrences of "when the differential temperature between the spray water and pressurizer vapor phase is greater than 2000 F".

Based on the record [References 2.13 and 2.14], the occurrences of 2000 F can be described as infrequent and largely involved with the use of aux spray under off-normal operating conditions.

PALISADES NUCLEAR POWER PLANT PALISADES ANALYSIS CONTINUATION SHEET EA-A600-2004-01 Page of (2- T 4.4 As discussed in Section 4.2 above, Palisades operates the pressurizer with 90% of total heater capacity energized all the time and offsets the heat input with a continuous spray flow. The continuous spray flow minimizes the potential of the harmful effects due to thermal stratification in the surge line and spray line as well as the effects of the in-surge transients. During normal operation in which there are small changes in pressurizer level, prior to an in-surge, the surge line is filled with out-surged fluid from the pressurizer.

When an in-surge transient occurs, the front of the flow is the fluid in the surge line that is at about the same temperature as the pressurizer. Unless there is a sustained in-surge flow, the effect of cooling the pressurizer is expected to be small.

During heat-up and cool-down processes, the spray flow rate varies over time. EA-GEJ 03 calculated spray flow rates might provide some perspective of the system operation. For a l 00F PCS-PZR AT, a flow rate of 83 gpm (nominal) is needed to offset the 90% of the heater capacity. This flow rate estimate took into consideration ambient heat loss.

The in-surge transient is unlikely to occur during heat-up when the system volume is in expansion.

Engineering Specification for the pressurizer [ Reference 2.5] sets forth the pressurizer's design requirements. The sudden cooling due to the change of surge flow temperature has been considered in several design transients. The largest surge temperature considered was a step change of about 70'F due to Unloading at 15% per minute. The designed number of occurrences of this transient is 15,000 cycles. It should be noted that the transients Loading and Unloading at 15%/o/minute are not in Palisades licensing basis for PCS system

[FSAR 4.21. This transient was apparently deemed as unrealistically conservative for the PCS design.

4.5 Palisades Pressure Temperature Limits (P-T Curves) are defined in Tech Spec 3.4 (Reference 2.8). These P-T curves set limits on the rate of heating up and cooling down of the PCS. The limits are much more restrictive than the heatup and cool down rates applied in the component design, i.e. 1000 F per hour for heat-up and 200° F per hour for cool-down

[Reference 2.5]. The current P-T curves are applicable through the plant current licensed life [Reference 2.15]. The P-T curves may need to be revised for plant life extension.

However, as the reactor is being aged with fluence, the limits on heatup and cooldown rates will be more restrictive, so the conclusions from this EA will not be affected by the future amendments of the P-T curve.

4.6 The loading conditions, design transients and cycles applied in CN-CI-02-71 were not identical to that of the Palisades specifications. Most of the loadings applied bound the Palisades design requirements. The few exceptions were the transient cycle numbers, i.e.

the occurrences of the reactor trip, the Loss of Reactor Coolant Flow and Loss of Load were less than the occurrences specified in Palisades pressurizer design specification [Reference 2.5]. However, CN-CI-02-71 has determined that these under reported transient occurrences make no significant contribution to the fatigue crack growth and has eliminated

PALISADES NUCLEAR POWER PLANT PALISADES ANALYSIS CONTINUATION SHEET EA-A600-2004-01 Page L7 of l.Z-them from consideration, therefore, the differences in these cycle numbers have no bearing on the analysis results.

4.7 Palisades current ISI Code of record is the 1989 Edition of ASME Code [Reference 2.16.].

Since CN-CI-02-71 invoked the 1992 edition of ASME III and XI, a comparison of the material properties have been performed for the pressurizer and the Hot Leg pipe material.

The pressurizer shell material is SA-533 GR.B CL.1 [References 2.10 and 2.17]. The Hot Leg Pipe is made of SA-264 [Reference 2.11], which is the specification of roll-bonded stainless steel clad with the base material of carbon or low alloy steel. The base material for the loop pipe is SA-516 GR.70. Comparison of the material properties confirmed that the stress allowables of these materials are identical in 1992 and 1989 ASME editions. No further code reconciliation is necessary for using the WCAP in Palisades's application.

4.8 For the purpose of supporting the discussion in Section 6 of this EA, a plot of pressurizer cooldown data from the 2003 refueling outage is included in Figure 2 of this EA. The source of the data is the Palisades Plant Computer down loaded to a PIS system. Referring to the upper portion of the plot, the vapor temperatures deviates from the water temperature twice during the cooldown process. Such deviations were likely indications of the occurrence of in-surge flow. The first occurrence of in-surge causes a differential temperature of about 80'F between the water phase and vapor phase. The change of temperature was fairly steady and it took several hours as the pressurizer water level rose slowly. The second occurrence was hardly noticeable in this plot; it took place at the end of bubble collapse. A Developer's note for SOP-I Section 7.1.2.p.2 relates the cause of the diverging temperatures between the water and steam space to the non-condensable vapor bubble. The evaluation of the diverging temperature is documented in C-PAL-95-0479B.

The plot also showed a step change of the vapor temperature during this occurrence.

However, close examination of the data concluded that the step change was false data. The time span of the signal was only 2 seconds, which is not credible for such a significant temperature swing. It is believed that the false signal was due to the sudden heat transfer coefficient change when the water level reached the upper temperature element. Clearly, there was no substantial in-surge flow observed.

5.0 Assumptions

5.1 Major Assumptions:

The attached Westinghouse reports assume operational transients bound the actual transients that occur during plant operation as controlled by the Technical Specifications and plant operating procedures. This engineering analysis assumes that the Westinghouse assumed transients will continue to bound these actual plant transients in the future. This assumption is appropriate because all but one of the plant actual transients are controlled by the Technical Specification P - T curves, and a license amendment would be required to change the curves. The only transient not controlled by the Technical Specification P/T curves is a sudden in-surge flow of 220TF delta T (see Section 6.5). This transient is

PALISADES NUCLEAR POWER PLANT PALISADES ANALYSIS CONTINUATION SHEET EA-A600-2004-01 Page 5 of I-controlled by plant operating procedures, rather than the Technical Specifications, and occurs when the pressurizer bubble is collapsed during cooldown. It is very unlikely that operating procedures would be revised in the future such that a sudden surge could occur in excess of 220'F delta T. Bubble collapse normally takes hours to complete due to limited charging system makeup capacity. Moreover, operating procedures require pressurizer heater operation and continuous spray flow, which would limit the in-surge delta T. Finally, all of the transients assumed in the Westinghouse reports will be described in Design Basis Document 2.04, "Primary Coolant System", to help ensure that plant operation is not changed in the future such that these Westinghouse assumed transients are no longer bounding.

5.2 Minor Assumptions:

There are a number of conservative assumptions described in WCAP- 15973 Section 2.3 Corrosion Evaluation and in CN-CI-02-071 Sections 3.0 and 6.3.3. Those are the assumptions in association with the structural analysis approach. Westinghouse reports discuss assumptions and their bases in the body of the reports. Nonetheless, there is also a very conservative assumption that neither the WCAP-1 5973 nor the CN-CI-02-71 has explicitly acknowledged. That is, the analyses assumed the in-surge transient is a local phenomenon. The in-surge flow does not mix with the fluid in the pressurizer, thus the pressure boundary material is subject to the in-surge flow temperature.

The pressurizer water phase (lower) temperature instrument is located near the lower shell to bottom head juncture. It is judged that the in-surge temperature detected by this instrument would be close to the lowest fluid temperature in contact with the shell. This is based on the surge nozzle screen assembly extending 36" above the bottom of the pressurizer ID [Reference 2.12], the upward flow momentum of the in-surge flow, and the limited mixing in the bottom dome which is a plenum occupied with 120 heaters.

Regarding the transient loadings, Westinghouse has not provided the detailed description of the mechanics and component responses to the in-surge transients described in Figure 1. At this point in time, this EA assumes the in-surge transients are applicable to the flaw evaluations of Palisades pressurizer. The analysis of this EA purports that the assumed in-surge transients, in Palisades case, bound transients during plant operation.

6.0 Analysis 6.1 WCAP- 15973 mentioned that Palisades' pressurizer lower temperature nozzle repair requires additional evaluation to accept the long-term corrosion degradation. Such an evaluation has since been completed and is documented in Attachment 3 of this cover EA.

Section 3.2 of WCAP- 15973 asked the users to evaluate the applicability to their plants of the transients depicted in Figure 1. The preceding Inputs and Assumptions sections have pointed out, in a general sense, the conservatisms involved in the generic evaluation. To

PALISADES NUCLEAR POWER PLANT PALISADE5 w ANALYSIS CONTINUATION SHEET EA-A600-2004-01 Page 9 of LZ demonstrate that Palisades is operated within the Figure 1 transients, the four pressurizer pressure-temperature profiles presented in Figure 1 are analyzed in Sections 6.2 through 6.5.

6.2 Heatup at a rate of 1000 F per hour and with a sudden cooling of the pressurizer due to 320'F cooler insurge flow.

A magnitude of 320'F PCS-PZR AT have been observed in many other PWR operations and has been used as a bounding value for surge line thermal stratification [Reference 2.6 applied a bounding value of 340'F]. During heatup, this large AT typically occurred at the startup of the Primary Coolant Pump when the PCS temperature is near shut down condition. If there is no pressurizer spray, the fluid in the surge line may remain at the ambient temperature. In the event that an in-surge occurs, the pressurizer shell would be subject to the surge line temperature. However, this magnitude of AT does not apply to Palisades. The high heat input from the pressurizer heaters and the the use of spray flow to control the heat-up rate keep the pressurizer surge line temperature close to that of the pressurizer. For a sustainned in-surge, the pressurizer would be subject to the hot leg fluid. As discussed in Section 4.2 of this EA, the PCS-PZR AT is limited to less than 200'F, well below the postulated 320'F. In addition, Palisades Tech Spec requirements (P-T Limit Curves) would not allow such a PCS-PZR AT.

Let's give an example of how the P-T Limit Curves are involved. Say, the pressurizer is at 550'F, which corresponds to a saturation pressure of 1045 psia. Per VLTOP set point of the P-T Limit Curves, the minimum PCS temperature at this pressure is around 3750 F.

Accordingly, the maximum PCS-PZR AT is 550-375 = 1750 F; a AT much smaller than the 320'F value. Regarding the heat-up rate, Figure I postulated a 1000 F per hour rate.

For Palisades' operation, the pressurizer heatup rate is limited to 600 F/hour when the Shutdown Cooling System is in service [SOP-I Section 4.4.2]. When Shutdown Cooling is secured, the limit on pressuirizer heat up rate is 100 0 F per hour. Therefore, both the heat up rate and PCS-PZR AT shown in Figure I bound Palisades operating parameters.

6.3 Cool-down at a rate of 200'F per hour until the pressurizer is at 2000 F. then cooldown at a rate of 750 F per hour when Pressuizer temperature reaches 2001F.

Pressurizer cooldown rate is largely dictated by the PCS cooldown rate, which is limited by the P-T Limit Curves and by the administratively required subcooling margin. A constant cooldown at 200° F/hour rate bounds the allowed PCS cooldown rate. When pressurizer temperature is at or below 2000 F, the PCS cooldown is near completion at the maximum temperature of 1751F. The pressurizer is in a solid condition and the Shut Down Cooling system is in service [SOP-I 7.1.4]. The PCS cool-down rate is limited to less than 40 OF per hour [SOP-1 4.4.1.c]. The conservatism of the 75'F/hour cool-down rate can be illustrated by a simplified analysis below.

PALISADES NUCLEAR POWER PLANT PALISADES ANALYSIS CONTINUATION SHEET iGW=EneW EA-A600-2004-01 Page 1Ž of ( 4UCLEAR PLMTI Assume the pressurizer cool-down is by the auxiliary spray flow driven by the maximum allowed charging flow of 80 gpm net [see SOP-1 7.1.4.o, SOP-I 5.3.1.e limits the flow to 53 gpm]. A maximum of 4800 gallons is delivered to the pressurizer in one hour. The pressurizer has a capacity of 1503.7 cubic feet [Reference 18 section 5.0], which amounts to 1503.7 x 7.48 = 11248 gallons. Conservatively ignore the latent heat in the pressurizer shell; conservatively assume the pressurizer heaters are turned off and a spray flow temperature of 70'F. The pressurizer temperature after one hour time is computed as (200'F x 11248 + 701F x 4800) / (11248 gallons + 4800 gallons) = 161IF. It thus shows a cool-down rate of 200 - 161 = 390 F in one hour of time.

Therefore, the cool-down rate of 750 F per hour bounds Palisades operating parameters.

6.4 Cool-down at 1000 F per hour with a sudden in-surge flow of 320 0 F AT coolant when pressurizer is at a temperature of about 6000 F In this scenario, an in-surge flow to the pressurizer occurs when the pressurizer temperature is near 6000 F and the PCS temperature is about 3000 F. This postulated transient is very conservative since such a large AT is not permissible by Palisades Tech Spec. At 6000 F pressurizer temperature, the saturation pressure is 1543 psia. Per VLTOP set point of the P-T Limit Curves, the PCS must be at least 4000 F. In other words, the PCS-PZR AT is limited to 2001F by Palisades Technical Specification and a 3200 F AT well bounds that of the Palisades operation. In terms of cool-down rate, a pressurizer cool-down rate exceeding 1000 F per hour is unlikely due to the restriction on the PCS cooldown rate. Palisades pressurizer cool-down rate has been well within 1000 F per hour to maintain a subcooling margin and to meet the P-T Limits. The conservatism of 100 0 F per cool-down is illustrated in the plot in Figure 2. Therefore, both the cool-down rate and PCS-PZR AT shown in Figure 1 bound Palisades operating parameters.

6.5 Cooldown at 1000 F per hour with a sudden in-surge flow of 2200 F AT coolant when pressurizer is at a temperature about 4000 F This scenario is most likely to occur during the collapsing of the pressurizer bubble, though the collapsing of the bubble normally takes hours of time due to the equipment capacity when charging the system volume. A sudden in-surge is unlikely due to bubble collapsing.

As a reference, SOP-1 Section 7.1.4.0. addresses the steps of collapsing bubble. It requires the operator to "Maintain maximum possible PZR heaters energized while controlling pressure with sprays to aid in pressure control". Nevertheless, while collapsing the bubble, the colder PCS fluid enters into the pressurizer creating a temperature transient. Noting that the pressurizer cool-down rate is administratively limited to 100F per hour [SOP- 1 7.1.4.p]

when pressurizer is in a solid condition. As to the in-surge flow temperature, Palisades operation of pressurizer high heat input and a continuous spray flow has kept the PCS-PZR AT within the 220'F value. This point is demonstrated in the experience data of Section 4.3 of this EA and can be seen in the pressure temperature profile plot in Figure 2. The in-surge

PALISADES NUCLEAR POWER PLANT PALISADES ANALYSIS CONTINUATION SHEET EA-A600-2004-01 Page {( of l2L temperature appeared to be bounded by the design basis transients, i.e. the 15% per minute LoadingfUnloading discussed in Section 4.4 of this EA. Therefore, both the cool-down rate and PCS-PZR AT shown in Figure I bound Palisades operating parameters.

7.0 Conclusions WCAP-1 5973 and its supporting calculation CN-CI-02-71 may be used by Palisades as a justification for leaving in place the remnant of a small bore nozzle that contains flaws. The small bore nozzles include the RTD and sampling nozzles on the hot leg, the pressurizer instrumentation nozzles and the pressurizer heater sleeves. This EA supercedes the Reference 2.2 EA, which supported the design modification of the pressurizer temperature nozzles. It should be noted that the acceptance of the vendor report is contingent on the methods of heating up and cooling down the PCS (see section 5.1 of this EA). All the applications of this EA need to be recorded in Attachment 4 of this EA for tracking purpose.

Based on the evaluation in this EA, it is concluded that the in-surge transients applied in the CN-CI-02-7 1 are conservative with respect to Palisades operation. CA024362 may be closed without further action.

The in-surge transients described in the Figure 1 are not part of Palisades design basis, except for the purpose of the evaluation of this EA. Westinghouse has not revealed the mechanics of the in-surge transients and has not incorporated such transients in design requirements. These transients need not to be considered as the equipment's design basis except for the compliance to the ASME XI flaw evaluation.

Page It of IC EA-A600-2004-01 Pzr Temperatures 700_ I I I I

-Hot Leg Loop 1 Temp

. KI I Egoo - Cold Le? Loop 1 Temp-Pzr Vapor Ternp

-Pzr Water Temp 3/16/2003 10:30:00 AM Day/(5) 318/2003 3.56 12: 00:00 AM PCS Press Level _____

2200 100 _ _ _ ___ - 7I w. . . . mr w gr t

__ _ l_l_l PCS Press _ ____

t50 -PZR Level 1000 _ ____II III 3/1 6/2003 10:30:00 AM 1.56 Day~s) 3/18/2003 12:00:00 AM Figure 2, 2003 Refueling Outage Pressurizer Cool-down Profile

Westinghouse Westinghouse Electric Company Nuclear Services P.O. Box 355 Pittsburgh, Pennsylvania 15230-0355 USA August 6, 2004 CPAL-04-28 Mr. James Wong Nuclear Management Company Palisades Nuclear Plant 27780 Blue Star Memorial Highway Covert, Michigan 49043 NUCLEAR MANAGEMENT COMPANY Palisades Nuclear Plant Corrosion Analysis of the Pressurizer Side Shell Nozzle

References:

1. Westinghouse letter LTR-NEM-04-469, "Westinghouse Proposal for Corrosion Analysis of Palisades Pressurizer Side Shell Nozzle", May 28, 2004.

2. WCAP-15973-9, Rev. 1, "Low-Alloy Steel Component Corrosion Analysis Supporting Small-Diameter Alloy 600/690 Nozzle Repair/Replacement", May 2004.

3. A-CEOG-9449-1242 Rev.00, "Evaluation of the Corrosion Allowance and Effective Weld to Support Small Alloy 600 Nozzle Repairs", June 13, 2000.

4. LTR-SGDA-04-217, "Evaluation of the Corrosion Allowance and Effective Weld to Support Small Alloy 600 Nozzle Repairs in the Palisades Unit", June 25, 2004.

5. Letter LTR-CI-04-40 dated 6/25/04

6. NMC Purchase Order P802445

7. Westinghouse Sales Order No. 28355

Dear Mr. Wong:

The purpose of this letter is to provide to NMC a corrosion analysis of the pressurizer side shell nozzle, as proposed in Reference 1. The proposed scope of work was authorized by NMC purchase order P802445, dated 6/08/2004.

Westinghouse WCAP-15973-P, Rev. I (Reference 2) addresses the effect of component corrosion in the event of a small bore Alloy 600 nozzle repair in CE plants, using the half-nozzle or similar repair technique. In this type repair, the annulus region between the nozzle and component is flooded with primary coolant, thereby exposing the carbon or low alloy steel component to primary coolant. The corrosion rates for carbon and low alloy steels are low when exposed to reactor coolant with boric acid concentrations typical of operating and shut down conditions. However, Reference 2 noted that there were some concerns about the amount of corrosion that could occur over extended periods of time and its potential effect on nozzle repair lifetimes.

As a result, the CEOG sponsored a task to address these corrosion concerns. The task determined for bounding cases how much corrosion is acceptable in the nozzle bores of the small diameter nozzles and heater sleeves in CE plants. The calculations supporting Reference 2 (Reference 3) determined the maximum allowable material that can be removed because of corrosion in the annulus region between the nozzles and components in accordance with ASME Code rules. The calculations determined the maximum allowable corroded hole diameter considering the reduction in the effective shear area and the maximum allowable corroded hole diameter considering the required area of reinforcement for each type nozzle in each plant.

Official record electronically approved in EDMS 2000 A BNFL Group company

Page 2 of 3 Our ref: CPAL,04-28 The limiting hole diameter was the more limiting of these two calculations. The allowable corrosion was then defined as the difference between the limiting allowable corroded hole diameter and the diameter of the original bore for each type nozzle. Reference 2 also established a corrosion rate, based on available data, for carbon and low alloy steels in primary coolant, considering operating, start-up and shut-down conditions.

Dividing the allowable corrosion by the corrosion rate provided an estimate of the repair lifetimes for the bounding case nozzles.

Reference 2 noted that the Palisades pressurizer side shell nozzle was not covered by the bounding case. For Palisades, calculations using actual thickness measurements would be required to determine acceptable bore size for the side shell nozzle. A similar situation existed for the cold leg piping Alloy 600 nozzles, although an estimated repair lifetime was not determined for these nozzles because of the lower operating temperature.

The primary water stress corrosion cracking (PWSCC) mechanism is strongly temperature dependent and Reference 2 noted that because of this temperature dependency, the cold leg nozzles are not considered limiting in terms of PWSCC initiation.

Reference 4 summarizes calculations to determine the limiting allowable corroded hole diameter, Dljn, for the Palisades pressurizer side shell and cold leg piping nozzles, using actual material thicknesses as reported by Palisades and the methods described in reference 2. For the pressurizer side shell nozzle, Dlim is 1.49 inch.

The original bore diameter was 1.330 inch, resulting in a corrosion allowance of 0.16 inch. Dividing this value by the corrosion rate described in Reference 2 (0.00306 inch per year diametral) results in an estimated repair lifetime of approximately 52.3 years.

For the cold leg Alloy 600 nozzles, Diim is 2.26 inches, using the same method. This value replaces the asterisk in Table 2 of Reference 3. An estimated repair lifetime was not calculated for this location as described above.

Reference 4 also noted that a calculation note is currently being authored to include the actual thicknesses for the pressurizer side shell nozzle and the cold leg piping nozzles along with the information for the other CE plants. This calculation note will include the limiting allowable hole diameters, D1jj for all CEOG plants, including the ones described above for Palisades.

If you have any questions or comments, please contactJohn Hall at 860-731-6688 or the undersigned at 412-374-6119.

Yours very truly, WESTINGHOUSE ELECTRIC COMPANY, LLC Stephen P. Swigart Customer Project Manager Cc: Richard Gerling, NMC Official record electronically approved in EDMS 2000 A BNFL Group company

Page 3 of 3 Our ref: CPAL-04-28 bcc: M. Miller S. Swigart R. Fagan B. A. Bell J. F. Hall E. A. Siegel D. Siska Official record electronically approved in EDMS 2000 A BNFL Group comrpany