ML061170214

NMC Responses to NRC Follow Up Questions Relating to License Renewal

ML061170214
Person / Time
Site:	Palisades
Issue date:	04/26/2006
From:	Harden P Nuclear Management Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
Download: ML061170214 (18)

Text

Palisades Nuclear Plant Operated by Nuclear Management Company, LLC April 26, 2006 10 CFR 54 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555 Palisades Nuclear Plant Docket 50-255 License No. DPR-20 NMC Responses to NRC Follow Up Questions Relating to License Renewal In a series of telephone conversations, the NRC raised follow up questions relating to NMC responses to several previous Requests for Additional Information (RAI) relating to the Application for Renewed Operating License - Palisades Nuclear Plant (LRA) dated March 22,2005. The Nuclear Management Company (NMC) responses to those questions are provided in Enclosure 1.

In addition, a verbal request was received for additional information on NMC's plans for addressing an NRC question about the potential for underclad cracking. In a letter dated March 30, 2006, NMC provided a commitment to either supplement the existing discussion on the subject in the LRA, or to provide a new discussion to describe the issue and its disposition as a TLAA. Enclosure 2 discusses how NMC is responding to this request, and provides the associated LRA changes. This information completes the commitment in the March 30, 2006 letter, but results in an additional related commitment as described below.

Please contact Mr. Robert Vincent, License Renewal.Project Manager, at 269-764-2559, if you require additional information.

Summary of Commitments This letter provides two new preliminary commitments (i.e., subject to NRC confirmation in the SER for the renewed license), withdraws superseded versions of a preliminary commitment, provides one new commitment, and closes one previous commitment contained. in NMC letter dated March 30, 2006.

The following new preliminary commitment is provided:

To verify that Corrosion Under Insulation (CUI) is not causing excessive corrosion of insulated piping and components, inspections of opportunity will be performed 27780 Blue Star Memorial Highway Covert, Michigan 49043-9530 Telephone: 269.764.2000

to assess the external surface condition when insulation is removed for maintenance or surveillance. The piping and components of interest are those within the scope of the System Monitoring Program, constructed of carbon or low alloy steel, with low normal operating temperatures in an indoor or outdoor environment such that the piping could be wetted under its insulation (e.g., from condensation or rain water) for extended periods without being detected. The System Monitoring Program will be enhanced to require a periodic review of documented under-insulation inspection results to verify that there were a sufficient number of inspection opportunities to provide a representative indication of system condition, and to assess the need for further action. If there were insufficient opportunities for inspection, insulation will be removed from additional sample locations to assess system condition under insulation. This program requirement will be implemented prior to March 24, 201 1.

The following revised preliminary commitment replaces all previous versions of this commitment appearing in NMC letters dated March 22, 2005, September 2, 2005, and January 13,2006:

NMC will revise the Alloy 600 Program to update the PWSCC corrosion rate assessments and inspection program consistent with the latest NRC requirements and industry commitments (e.g., EPRl Report 101 0087 "Materials Reliability Program: Primary System Piping System Butt Weld Inspection and Evaluation Guidelines [MRP-139]," (August 2005)). The updated program will be submitted for NRC review and approval by March 24,2008.

The following new commitment is provided, to read as follows:

Upon completion of the final technical report on the effects of potential underclad cracking at Palisades, NMC will notify NRC that the technical report for the final disposition of the issue has been completed, and the associated LRA changes submitted on April 26, 2006, have been confirmed. If the final report identifies a need for any additional LRA revisions, the revised information will be provided at that time for NRC review and approval. NMC will submit this information no later than September 1, 2006.

I declare under penalty of perjury that the foregoing is true and correct. Executed on April 26, 2006.

Paul A. Harden Site Vice President, Palisades Nuclear Plant Nuclear Management Company, LLC 27780 Blue Star Memorial Highway Covert, Michigan 49043-9530 Telephone: 269.764.2000

Enclosure CC Administrator, Region Ill, USNRC Project Manager, Palisades, USNRC Resident Inspector, Palisades, USNRC License Renewal Project Manager, Palisades, USNRC 27780 Blue Star Memorial Highway Covert, Michigan 49043-9530 Telephone: 269.764.2000

ENCLOSURE 1 NMC Responses to NRC Follow Up Questions Relating to License Renewal (9 Pages)

NMC Responses to NRC Follow Up Questions Relating to License Renewal NRC Questions Relating to the System Monitoring Program In a conference call on September 14, 2005, several draft Requests for Additional Information (RAI) relating to the general visual walkdown inspections performed under the Palisades System Monitoring Program were discussed. The questions were not issued as RAls, and NMC did not submit written responses. Subsequent to this discussion, the NRC issued revision 1 of NUREG 1801. Revision 1 included a new program, XI.M36, "External Surfaces Monitoring," which clarified NRC1s expectations for a general visual inspection program.

On April 17, 2006, the NRC indicated in a phone call that written responses were needed on four of the earlier questions relating to System Monitoring Program visual inspections. Responses to those questions are provided as follows:

Draft RAI B2.1.20-2(a)

The draft question stated "The applicant is requested to either (1) identify specific codes and standards or manufacturer's recommendations that will be applied to the SMP, or (2) as a condition of the license, clarify that this program will be updated at a later date and submitted to the NRC for review prior to the period of extended operation."

NMC Response to Draft RAI B2.1.20-2(a)

In the September 14, 2005 phone call, NMC confirmed statements in a July 25, 2005 letter that industry standards EPRl 1009743, GS-7086 and API 575 would be used as source documents to define tank testing and inspection requirements performed under the System Monitoring Program.

For the other general system inspections, consistent with industry practice and NUREG 1801 program XI.M36, the following industry standards (or later revisions) provide the bases for the system walkdowns under the System Monitoring Program:

INPO Good Practice TS-413, "Use of System Engineers", INPO 85-033, May 18, 1988.

EPRl Technical Report 1007933 "Aging Assessment Field Guide," December 2003 EPRl Technical Report 1009743 "Aging Identification and Assessment Checklist," August 27, 2004.

Draft RAI B2.1.20-2(d)

The draft question states, "The applicant is requested to consider industry guidance on corrosion under insulation (CUI) and clarify if removal of insulation at selected locations or imaging techniques may be prudent. For example, API-570 identifies common NMC Responses to NRC Follow Up Questions Relating to License Renewal locations susceptible to CUI including the extent of visual external and CUI inspections at suspect locations."

Response to Draft RAI B2.1.20-2(d)

The reviewer's concern is valid that low temperature system piping and components, under the right circumstances, could experience condensation or wetting over an extended period such that undetected corrosion could occur. Locations in the system of primary concern for this phenomenon, the Service Water System, are available for under-insulation inspections on a periodic basis when insulation is removed for NDE under the Risk Informed lnservice Inspection Program. However there may be other in-scope, insulated piping systems where this potential also exists. To address this question, the System Monitoring Program will be enhanced as follows:

To verify that Corrosion Under Insulation (CUI) is not causing excessive corrosion of insulated piping and components, inspections of opportunity will be performed to assess the external surface condition when insulation is removed for maintenance or surveillance. The piping and components of interest are those within the scope of the System Monitoring Program, constructed of carbon or low alloy steel, with low normal operating temperatures in an indoor or outdoor environment such that the piping could be wetted under its insulation (e.g., from condensation or rain water) for extended periods without being detected. The System Monitoring Program will be enhanced to require a periodic review of documented under-insulation inspection results to verify that there were a sufficient number of inspection opportunities to provide a representative indication of system condition, and to assess the need for further action. If there were insufficient opportunities for inspection, insulation will be removed from additional sample locations to assess system condition under insulation. This program requirement will be implemented prior to March 24, 201 1.

Draft RAI B2.1.20-I (a)

The draft question states, "The applicant is requested to clarify if all exposed surfaces of tanks are accessible for inspection from existing installed plant walkways, ladders and platforms or to clarify if temporary ladders and platforms are to be installed to support inspections Response to Draft RAI B2.1.20-I (a)

To confirm our statements in the September 14 call, and in a February 14 meeting, portable ladders and platforms will be used as necessary to gain access to exposed tank surfaces for visual inspections under the System Monitoring Program.

NMC Responses to NRC Follow Up Questions Relating to License Renewal Draft RAI B2.1.20-3 The draft question states, "The applicant is requested to either (I) identify the inspection criteria and qualifications of inspectors that will be applied to the SMP, or (2) as a condition of the license (or commitment), clarify that this program will be updated at a later date and submitted to NRC for review prior to the period of extended operation."

Response to Draft RAI B2.1.20-3 Persons performing system walkdowns are qualified using a formal NMC fleet-wide standard process. NMC Fleet Mentoring Guide FL-ESP-SYS-005M, "Perform System Walkdowns as Required to Accomplish Safe, Reliable and Efficient Operation of Assigned System" defines and documents the knowledge and practical demonstration requirements for persons performing system walkdowns. This qualification standard addresses the required elements of the walkdown inspections, including system and component conditions, structural component conditions, housekeeping and unrestrained items, and administrative requirements. This standard also references the Palisades plant procedure, EM-20, "Performance Monitoring Program," which provides plant-specific guidance for the walkdowns. Copies of the mentoring guide and applicable procedures are available on site for review.

This is consistent with the guidance provided in the new program XI.M36 "External Surfaces Monitoring" in NUREG 1 801, revision 1. The XI.M36 program addresses qualification under Monitoring and Trending by stating, "Visual inspection activities are performed and associated personnel are qualified in accordance with site controlled procedures and processes."

NMC Responses to NRC Follow Up Questions Relating to License Renewal NRC Question Relating to the Inspections of High Strength Bolts A follow up question was raised about inspections of high strength bolting. The ensuing discussion indicated that it would be helpful for NMC to provide specific references to various places in the LRA and RAI responses where the general criteria and frequencies for high strength bolting inspections have been discussed. References for, and brief summaries of, previously docketed information are provided as follows:

Background

Page Ill B1-3 of NUREG-1801 indicates that high strength bolting is managed by the Bolting lntegrity Program for crackinglstress corrosion cracking.

In the Palisades Bolting lntegrity Program description in LRA section B2.1.3, at the bottom of page B-25, it states:

"The program credits activities performed under three separate aging management programs for the inspection of bolting. The three aging management programs are: (I) ASME Section XI IWB, IWC, IWD, IWF lnservice lnspection Program, (2) Structural Monitoring Program, and (3) System Monitoring Program."

On page B-26 of the LRA, it further states:

"The Bolting lntegrity Program is consistent with NUREG-1 801,Section XI.Ml8, "Bolting Integrity."

Bolting within the Scope of the ASME Section XI Program As described in LRA section B2.1.2 the ASME Section XI IWB, IWC, IWD, IWF lnservice lnspection Program includes the pressure-retaining components and their supports. The Program Description on page B-17 states:

"The program includes periodic visual, surface andlor volumetric examinations and leakage tests of all Class 1, 2 and 3 pressure-retaining components, their supports and integral attachments, including welds, pump casings, valve bodies, pressure-retaining bolting, pipinglcomponent supports, and reactor head closure studs. These are identified in ASME Section XI, "Rules for lnservice lnspection of Nuclear Power Plant Components," or commitments requiring augmented inservice inspections, and are within the scope of license renewal."

The inspection criteria are established by the rules of Section XI. For high strength bolting within the scope of Section XI, the following amplification was provided in the NMC Response to Follow up Question Concerning NRC RAI B2.1.3-I (d) provided by NMC Letter Dated October 14, 2005. It states, NMC Responses to NRC Follow Up Questions Relating to License Renewal "In summary, for the Class 1 component support A490 and SA-193-B7 bolting applications described above, SCC has been conservatively determined to be applicable to only the steam generator snubber supports and the pressurizer supports. For those instances where SCC has been identified as an aging effect requiring management, NMC will perform visual inspection to identify the potential for SCC by detecting evidence of corrosion or a corrosive environment.

If the potential for cracking is found, the extent of any degradation may be identified and measured by removing the bolting for further inspection, proof testing by tension or torquing, in situ ultrasonic testing, or hammer testing. See the response to NRC RAI B2.1.3-3(a), in NMC letter dated August 12, 2005, for more detailed information."

The inspection frequency for all bolting is established by the rules of Section XI.

Inspections are scheduled within the normal ten year intervals as the code specifies.

High Strength Bolting within the Scope of the Structural Monitoring Program lnspection frequencies for all bolting within the scope of the Structural Monitoring Program are discussed in LRA Section B2.1. I 9. The LRA Section B2.1.19, on Page B-137 states:

"Initial baseline inspections under the Structural Monitoring Program were performed, as required by 10 CFR 50.65, starting in late 1996. A second complete inspection was performed in 1999 to validate the initial inspection results. Subsequent inspections follow a 10 year interval schedule that is similar to lnspection Plan B defined in the ASME Boiler & Pressure Vessel Code,Section XI, Table IWE-2412-1.

The 10 year inspection interval is divided into three 40 month periods.

Approximately one third of the items in the program scope are examined in each period and all items are examined at least once during the 10 year interval. The first interval, first period, inspections have been completed, and Palisades is currently in the first interval, second period inspection cycle. Other features may have greater inspection frequencies such as watertighttflood barrier inspections (at least once per 5 years) and below-the-waterline water-control structures (once every 5 years).

Augmented inspection is required for items that have been repaired or that exhibit significant damage or deterioration. It may also be required for items subject to aggressive environments. Items that are tagged for augmented inspection following repair or for reasons of damage 1 deterioration are examined, at a minimum, in the period immediately following the one during which the repair was performed or the deleterious condition was found. Augmented inspection may be performed on a 40 month period basis or at more closely spaced intervals as specified by the Structural Monitoring Coordinator or in plant procedures."

NMC Responses to NRC Follow Up Questions Relating to License Renewal On LRA page B-141 the Structural Monitoring Program description also describes inspection frequency by stating:

"The Palisades Structural Monitoring Program is implemented in accordance with a 10 year interval schedule that requires completion of approximately 113 of the specified examinations within each of three consecutive 40 month periods. The program requires that 100% of the items included within its scope be examined at least once (and more often for those items requiring augmented inspection) during each interval. These requirements for frequent and comprehensive examinations provide a high degree of assurance that age related deterioration of an item will be detected and corrected long before it has a significant impact on the item's intended function.

Examinations of structures and structural elements are performed by qualified personnel using techniques appropriate for the item, its environment and its intended function. Each individual selected to perform these examinations must have the following qualifications:

A civil engineering degree from an accredited university.

Familiarity with the design and performance requirements applicable to nuclear power plant structures and experience in the in-service examination and evaluation of these structures.

A minimum of 5 years experience in engineering design and I or analysis of nuclear power plant structures."

This section goes on to state that general inspections are also performed more frequently than the formal structural monitoring program inspections, as follows:

"System engineers and plant operators augment the formal examinations by noting the conditions of structures I structural elements during periodic system walk downs, and reporting observed damage I degradation to the Structural Monitoring Coordinator."

Inspection criteria for all bolting under the Structural Monitoring Program are addressed in NMC Response to NRC RAI B.2.1.3-3(a) (Letter dated August 12, 2005), which

states, "NUREG-1 801, AMP XI.Ml8, element 4, states: "Structural bolting both inside and outside containment is inspected by visual inspection. Degradation of this bolting may be detected and measured either by removing the bolt, proof test by tension or torquing, by in situ ultrasonic tests, or hammer test. If this bolting is found corroded, a closer inspection is performed to assess extent of corrosion."

Management of Palisades' structural bolting is consistent with this discussion.

The Palisades Bolting Integrity Program Basis Document states that structural bolting and fasteners, both inside and outside containment, are inspected by visual examination in accordance with the Structural Monitoring Program. The Structural Monitoring Program facilitates visual inspection of structural bolting. If NMC Responses to NRC Follow Up Questions Relating to License Renewal visual degradation is observed requiring further evaluation, degradation (e.g.

crack initiation due to cyclic loading or SCC) may be identified and measured by removing the bolting, proof test by tension or torquing, by in situ ultrasonic tests, or hammer tests.

Structural bolting is typically not subject to significant cyclic loading or thermal stress, so cracking due to fatigue is not an applicable aging effect. For stress corrosion cracking to occur, three conditions must exist: high stress, a corrosive environment, and susceptible material. The corrosive environment is initially precluded through use of proper lubricants and proper bolt installation practices.

Visual inspection for degradation will identify the potential for SCC by detecting evidence of corrosion or a corrosive environment. If the potential for cracking is found, the extent of any degradation may be identified and measured by removing the bolting for further inspection, proof testing by tension or torquing, in situ ultrasonic testing, or hammer testing."

NMC Responses to NRC Follow Up Questions Relating to License Renewal NRC Follow Up Question About NMC's 1/13/2006 Response to RAI 4.7.2-1 In a telephone call on April 20, 2006, a follow up question was received on the NMC response to RAI 4.7.2-1 in a letter dated January 13, 2006. The RAI response did not completely resolve the potential that LRA Section 4.7.2 could be erroneously interpreted as relying on a fatigue analysis to address a PWSCC growth rate issue.

NMC Response to NRC Follow Up Question About NMC's 1/13/2006 response to RAI 4.7.2-1 The ensuing discussion in the phone call indicated that the confusion occurred because the LRA did not clearly describe the purpose of Table 4.7.2-1, and the bases for the dispositions of fatigue and PWSCC in Alloy 600 as separate TLAAs, both managed by programs, were still unclear. The following clarifications to Section 4.7.2 are provided to resolve the confusion:

On LRA page 4-54, add the following note just prior to Table 4.7.2-1 :

"NOTE: Table 4.7.2-1 below is provided for information only. The table summarizes previous analyses and shows the locations to be addressed as TLAAs. The service lives listed in the table consider the combined effects of both fatigue and corrosion; as TLAAs, however, each effect is discussed and dispositioned separately on subsequent pages. The service life for PWSCC effects is managed by the Palisades Alloy 600 Program. The service life for fatigue effects is managed by the Fatigue Monitoring Program.

On LRA pages 4-60 and 4-61, the last two TLAA disposition sections are revised to read as follows:

"Disposition for Cycle-Dependent Aspects of the Bounding Fracture Mechanics Analysis of the Hot Leg, Piping RTD and Sampling Nozzles, Pressurizer Instrument Nozzles, and Pressurizer Heater Sleeves; and Disposition for Fatigue Portions of All Other Alloy 600 Fracture Mechanics Analyses: 10 CFR 54.21 (c)(l)(iii)

The Palisades plant-specific bounding fracture mechanics analysis demonstrates the validity of the cycle-dependent aspects of the generic bounding fracture mechanics analysis (WCAP-15973-P) by demonstrating that the plant-specific load and thermal events are within those assumed by the generic bounding analysis. The basis for the safety determination of the fracture mechanics evaluation calculation will therefore remain valid for fatigue effects so long as the numbers of these events do not exceed the design basis values.

The fatigue cycle count program described in Appendix B, Fatigue Monitoring Program, will ensure a reanalysis or other appropriate corrective action if a design basis primary coolant system cycle count limit is reached at any time during the extended licensed operating period.

NMC Responses to NRC Follow Up Questions Relating to License Renewal Disposition for the Corrosion Susceptibility of All Alloy 600 Heater Sleeves, Nozzles, Safe Ends, and Flanges: 10 CFR 54.21(c)(l)(iii)

The Palisades Alloy 600 Program identifies the Alloy 600 components in the primary coolant system, ranks them according to PWSCC susceptibility, and establishes a program for inspection, repairs, and mitigation. All 250 Alloy 600 heater sleeves, nozzles, safe ends, and flanges are subject to the inspection program. At all 250 locations the program requires at least an insulated VT-2 visual inspection for leakage every refueling outage. Locations which are more susceptible to PWSCC, or whose failure could result in a more significant safety hazard, are also subject to initial or periodic bare-metal VT-2, volumetric, or penetrant inspections.

NMC will revise the Alloy 600 Program to update the PWSCC corrosion rate assessments and inspection program consistent with the latest NRC requirements and industry commitments (e.g., EPRl Report 101 0087 "Materials Reliability Program: Primary System Piping System Butt Weld Inspection and Evaluation Guidelines [MRP-139]," (August 2005)). The updated program will be submitted for NRC review and approval at least three years prior to entering the period of extended operation (March 24, 2008). Note that the language of this revised commitment supersedes all previous versions of this commitment associated with the Alloy 600 Program that appeared in letters dated March 22, 2005, September 2,2005, and January 13,2006.

Supplemental Information Regarding Underclad Cracking as a Time-Limited Aging Analysis Supplemental Information Regarding Underclad Cracking as a Time-Limited Aging Analysis

Background

In a letter dated March 30, 2006, NMC responded to an NRC follow up question on why underclad cracking need not be addressed as a potential Time-Limited Aging Analysis (TLAA) for Palisades. In that letter NMC provided a commitment either to supplement the existing discussion on the subject in the LRA, or to revise the LRA to discuss the issue as a TLAA. In a recent telephone conversation, NRC requested additional detail on NMC's plans for addressing the issue.

NMC has determined that the preferred way to disposition the issue for license renewal is to classify it as a TLAA, prepare a technical report that dispositions the TLAA for the full 60-year extended operating period, and update the pertinent LRA sections to reflect this new information. To provide NRC with a more detailed discussion of the issue, NMC is also providing the LRA changes that reflect the expected technical conclusions for the issue. These LRA changes have been prepared based on preliminary information from the NSSS vendor. Upon completion of the final technical report on the effects of potential underclad cracking at Palisades, NMC will notify NRC that the technical report for the final disposition of the issue has been completed and the associated LRA changes submitted in this letter have been confirmed. If the final report identifies a need for any additional LRA revisions, the revised information will be provided at that time for NRC review and approval. NMC will submit this information no later than September 1, 2006.

LRA Revisions Related to Potential Underclad Cracking The LRA is revised to add a new LRA Section 4.7.6, to read as follows:

"4.7.6 Reactor Vessel Underclad Cracking The issue of underclad cracking in reactor pressure vessels (RPV) has been identified since 1970 when it was first discovered at a European vessel fabricator. Because of some similarities in manufacturing processes, there is a potential for this condition to exist in the Palisades reactor vessel.

Underclad cracking has occurred in the low alloy steel base metal heat-affected zone (HAZ) beneath the austenitic stainless steel weld overlay that is deposited to protect the ferritic material from corrosion. Two types of underclad cracking have been identified. Reheat cracking has occurred as a result of postweld heat treatment of austenitic stainless steel cladding applied using high-heat-input welding processes on ASME SA-508, Class 2 forgings. Cold cracking has occurred in ASME SA-508, Class 3 forgings after deposition of the second and third layers of cladding, when no pre-heating or post-heating was applied during the cladding procedure. The cold cracking was determined to be attributable to residual stresses near the yield strength in the weld metallbase metal interface after cladding deposition, combined with a crack-sensitive microstructure in the HAZ, and high levels of diffusible hydrogen in the austenitic stainless steel or Supplemental Information Regarding Underclad Cracking as a Time-Limited Aging Analysis lnconel weld metals. The hydrogen diffused into the HAZ and caused cold (hydrogen-induced) cracking as the HAZ cooled.

Analysis A generic fracture mechanics evaluation of Westinghouse plants initially demonstrated that the growth of underclad cracks during a 40-year plant life was insignificant. The evaluation was extended to 60 years, using fracture mechanics analysis based on a representative set of design transients, with the occurrences extrapolated to cover 60 years of service life. The 60-year evaluation (WCAP-15338-A) (Reference 8) showed insignificant growth of the underclad cracks, and concluded that the cracks were of no concern relative to structural integrity of the reactor vessel. The NRC reviewed and approved the evaluation (WCAP-15338-A) for application to all Westinghouse RPVs (Reference 9), and identified two plant-specific Applicant Action Items to be completed by each applicant as a condition for referencing WCAP-15338-A. These action items include verifying that the design transients and operating conditions assumed in the report are applicable to the applicant's plant, and providing an description of the issue as a TLAA to be incorporated into the FSAR.

Although Palisades is not a Westinghouse plant, the Palisades reactor vessel was fabricated using similar processes and materials as those used in some reactor vessels fabricated by Combustion Engineering for Westinghouse. It has been determined that if underclad cracking were postulated to exist in the Palisades reactor vessel, WCAP-15338-A is applicable to the issue, and provides a valid technical reference to support a 60-year service life.

As specified in the Applicant Action Items, the design cycles and transients assumed in the WCAP-15338-A analysis for a 3-loop Westinghouse plant have been determined to be very similar to and representative of the design and operation of the Palisades Nuclear Plant. A description of the issue to be incorporated into the Palisades FSAR has been provided in Appendix A, Section A4.5.6, of the LRA.

Disposition: 10 CFR 54.21 (c)(l)(ii)

Reactor vessel underclad cracking is dispositioned under 10 CFR 54.21 (c)(l)(ii),

the analysis has been projected to the end of the period of extended operation.

The LRA Section 4.7 References on page 4-65 are revised to add the following:

"8.

Westinghouse WCAP 15338-A, "A Review of Cracking Associated with Weld Deposited Cladding in Operating PWR Plants," October 2002.

9.

NRC Letter, "Revised Safety Evaluation of WCAP-15338 "A Review of Cracking Associated with Weld Deposited Cladding in Operating Supplemental Information Regarding Underclad Cracking as a Time-Limited Aging Analysis Pressurized Water Reactor (PWR) Plants," To: Roger A. Newton, WOG Chairman, From: Pao-Tsin Kuo, Program Director, Dated September 25, 2003."

LRA Section 3.1.2.2.5 is revised to replace the existing discussion in its entirety with the following:

3.1.2.2.5 Crack Growth due to Cyclic Loading "NUREG-1800 states that crack growth due cyclic loading could occur in reactor vessel shell and reactor coolant system piping and fittings. Growth of intergranular separations (underclad cracks) in low-alloy or carbon steel heat affected zone under austenitic stainless steel cladding is a time-limited aging analysis (TLAA) to be evaluated for the period of extended operation for all the SA 508-CI 2 forgings where the cladding was deposited with a high heat input welding process.

Underclad cracking in carbonllow-alloy steel, which has been clad with austenitic stainless steel using weld-overlay processes, has been identified as an aging effect requiring management and is addressed as a TLAA. An evaluation of the TLAA for underclad cracking is contained in Section 4.7.6."

LRA Appendix A is revised to add the new section A4.5.6 to read as follows:

"A4.5.6 Reactor Vessel Underclad Cracking The issue of underclad cracking in certain reactor vessels has been identified since 1970 when it was first discovered at a European vessel fabricator.

Because of some similarities in manufacturing processes, there is a potential for this condition to exist in the Palisades reactor vessel.

Underclad cracking has occurred in the low alloy steel base metal.heat-affected zone (HAZ) beneath the austenitic stainless steel weld overlay that is deposited to protect the ferritic material from corrosion. Two types of underclad cracking have been identified. Reheat cracking has occurred as a result of postweld heat treatment of austenitic stainless steel cladding applied using high-heat-input welding processes on ASME SA-508, Class 2 forgings. Cold cracking has occurred in ASME SA-508, Class 3 forgings after deposition of the second and third layers of cladding, when no pre-heating or post-heating was applied during the cladding procedure. The cold cracking was determined to be attributable to residual stresses near the yield strength in the weld metallbase metal interface after cladding deposition, combined with a crack-sensitive microstructure in the HAZ, and high levels of diffusible hydrogen in the austenitic stainless steel or lnconel weld metals. The hydrogen diffused into the HAZ and caused cold (hydrogen-induced) cracking as the HAZ cooled.

Supplemental Information Regarding Underclad Cracking as a Time-Limited Aging Analysis Analysis A generic fracture mechanics evaluation of Westinghouse plants initially demonstrated that the growth of underclad cracks during a 40-year plant life was insignificant. The evaluation was extended to 60 years, using fracture mechanics analysis based on a representative set of design transients with the occurrences extrapolated to cover 60 years of service life. The 60-year evaluation (WCAP-15338-A) showed insignificant growth of the underclad cracks, and concluded that the cracks were of no concern relative to structural integrity of the reactor vessel. The NRC reviewed and approved the evaluation (WCAP-15338-A) for application to all Westinghouse RPVs.

Although Palisades is not a Westinghouse plant, the Palisades reactor vessel was fabricated using similar processes and materials as those used in some reactor vessels fabricated by Combustion Engineering for Westinghouse. The design cycles and transients assumed in the WCAP-15338-A analysis for a 3-loop Westinghouse plant are very similar to and representative of the design and operation of the Palisades Nuclear Plant. It has been determined that WCAP-15338-A is applicable to the Palisades reactor vessel and may be used as a valid reference to support a 60-year service life.

Disposition: 10 CFR 54.21 (c)(l)(ii)

Reactor vessel underclad cracking is dispositioned under 10 CFR 54.21 (c)(l)(ii),

the analysis has been projected to the end of the period of extended operation.