ML061710302

Submittal of 4th Interval Inservice Inspection Plan for Palisades

ML061710302
Person / Time
Site:	Palisades
Issue date:	06/12/2006
From:	Harden P Nuclear Management Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
Download: ML061710302 (163)

Text

Nl~lCPalisades Nuclear Plant Committed to Nuclear ExcellencePaideNulrPan C Operated by Nuclear Management Company, LLC June 12, 2006 10 CFR 50.55a U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Palisades Nuclear Power Plant Dockets 50-255 License No. DPR-20 4 th Interval Inservice Inspection Plan Pursuant to 10 CFR 50.55a, Nuclear Management Company, LLC (NMC) is submitting the 4 th interval inservice inspection (ISI) plan for the Palisades Nuclear Plant (PNP).

The relief requests included in the plan are being submitted for Nuclear Regulatory Commission (NRC) review and approval.

The 4 th interval ISI program has been developed to the American Society of Mechanical Engineers, Boiler and Pressure Vessel Code, Section Xl, 2001 Edition with 2003 Addenda. The 4 th interval officially begins on December 13, 2006 for PNP. Enclosure 1 provides the 4 th interval ISI plan for PNP.

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

Paul A. Harden Site Vice President, Palisades Nuclear Plant Nuclear Management Company, LLC Enclosure (1) cc: Administrator, Region Ill, USNRC Project Manager, Palisades, USNRC Resident Inspector, Palisades, USNRC 27780 Blue Star Memorial Highway

• Covert, Michigan 49043-9530 Telephone: 269.764.2000

ENCLOSURE 1 4 TH INTERVAL INSERVICE INSPECTION PLAN PALISADES NUCLEAR PLANT 96 PAGES FOLLOW

PALISADES NUCLEAR PLANT 4 TH INTERVAL INSERVICE INSPECTION PLAN INSERVICE INSPECTION PROGRAM FOR Palisades Nuclear Plant Commercial Service Date: December 31, 1971 (Docket no. 50-255) 27780 Blue Star Memorial Highway Covert, Michigan 49043 INSERVICE INSPECTION PROGRAM PREPARATION AND APPROVAL Prepared By: ______-- _ Date: 6 2oo0G Inservice Inspection Program Owner Reviewed By: . Date:____

4N~r, Engineering Programs Reviewed By: \7 /CM Date: 14 Supervisor, Egieering Programs/'

Approved By: , Date:_____

Manager, En i ering Programs 1 of 96

PALISADES NUCLEAR PLANT 4 TH INTERVAL INSERVICE INSPECTION PLAN 1.0 P U R PO S E ......................................................................................................... 4 2.0 R EFER EN C ES ................................................................................................. 4 3.0 D ISC U SSIO N .................................................................................................... 5 4.0 REQ UIREM ENTS ............................................................................................ 6 4.1 ASME Section XI Code and Regulatory Requirements .................. 6 4.2 ISI Requirem ents ....................................................................................... 6 4.3 Section XI Requirements for ASME Code Class 1 Components ...... 6 4.4 Section XI Exemptions for ASME Code Class 1 Components ....... 7 4.5 Section XI Requirements for ASME Code Class 2 Components ...... 8 4.6 Section XI Exemptions for ASME Code Class 2 Components ..... 8 4.7 Section XI Requirements for ASME Code Class 3 Components .... 10 4.8 Section XI Exemptions for ASME Code Class 3 Components ........ 10 4.9 Risk-Informed Examination Requirements ...................................... 10 4.10 Inspection Schedule ................................................................................ 11 5.0 INSERVICE INSPECTION PROGRAM

SUMMARY

TABLES ................... I1 C lass 1 Exam s ................................................................................................. 14 C lass 2 Exam s ................................................................................................. 23 C lass 3 Exam s ................................................................................................ 25 S upport Exam s ................................................................................................. 26 Risk-Inform ed Exam s ..................................................................................... 27 6.0 AUGMENTED AND OWNER ELECTED EXAMINATIONS ................... 28 6.1 Technical Specification Required Augmented Examinations ......... 28 6.2 Miscellaneous Augmented Examinations ......................................... 30 6.3 Owner Elected Examinations ............................................................... 31 A ugm ented Exam s .......................................................................................... 33 Augmented Alloy 600 Exams ...................................................................... 34 7.0 CO D E CA SES .............................................................................................. 36 8.0 RELIEF REQUESTS ..................................................................................... 36 9.0 D raw ings ......................................................................................................... 37 2 of 96

PALISADES NUCLEAR PLANT 4 TH INTERVAL INSERVICE INSPECTION PLAN C lass 1 D raw ings ............................................................................................ 37 C lass 2 Draw ings ............................................................................................ 39 C lass 3 D raw ings ............................................................................................ 42 Risk-Informed Drawings ................................................................................ 45 Defense in Depth Drawings ......................................................................... 50 Augmented Drawings .................................................................................... 50 10.0 CALIBRATION BLOCKS ........................................................................... 51 11.0 R EC O R D S ...................................................................................................... 55 Appendix A Palisades Nuclear Plant Code Cases ......................................... 56 Appendix B Palisades Nuclear Plant Relief Requests ................................... 58 RELIEF REQUEST NUMBER - RR 4-1 ............................................................ 60 RELIEF REQUEST NUMBER - RR 4-2 ............................................................ 62 RELIEF REQUEST NUMBER - RR 4-3 ............................................................. 67 RELIEF REQUEST NUMBER - RR 4-4 ............................................................. 70 RELIEF REQUEST NUMBER - RR 4-5 ............................................................. 73 RELIEF REQUEST NUMBER - RR 4-6 ............................................................. 75 RELIEF REQUEST NUMBER - RR 4-7 .......................................................... 77 RELIEF REQUEST NUMBER - RR 4-8 ............................................................. 80 RELIEF REQUEST NUMBER - RR 4-9 ............................................................. 82 RELIEF REQUEST NUMBER - RR 4-10 .......................................................... 89 RELIEF REQUEST NUMBER - RR 4-11 ......................................................... 91 RELIEF REQUEST NUMBER - RR 4-12 .......................................................... 93 ATTACHMENT 1 RELIEF REQUEST REFERENCE DRAWINGS ATTACHMENT 2 EPRI REPORT Palisades Steam Generator Inlet and Outlet Nozzle Coverage Calculations 3 of 96

PALISADES NUCLEAR PLANT 4 TH INTERVAL INSERVICE INSPECTION PLAN 1.0 PURPOSE 1.1 This document establishes the implementation controls and schedule for the Inservice Inspection (ISI) Program plan for the fourth ten-year inspection interval at Consumers Energy Palisades Nuclear Plant (PNP),

Unit 1. The Nuclear Management Company (NMC) operates the plant and is the Owner as defined by IWA-9000.

1.2 This ISI Program Plan details the requirements for the examination of ASME Class 1, 2 and 3 components and component supports and the Risk Informed (RI) piping welds at the Palisades Nuclear Plant required to be examined in accordance with the 2001 Edition through 2003 Addenda of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code Section XI within the limitations and modifications required by Title 10 of the Code of Federal Regulations, Part 50.55a, Codes and Standards.

2.0 REFERENCES

2.1 Code of Federal Regulations, Title 10, Part 50, Paragraph 50.55a, "Codes and Standards" 2.2 ASME Boiler and Pressure Vessel Code,Section XI, Division 1, "Inservice Inspection of Nuclear Power Plant Components", 2001 Edition through 2003 Addenda.

2.3 Regulatory Guide 1.26, Revision 3, "Quality Group Classification and Standards for Water, Steam and Radioactive Waste Containing Components of Nuclear Power Plants" 2.4 Regulatory Guide 1.147, Revision 14, "Inservice Inspection Code Case Acceptability, ASME Section XI, Division 1" 2.5 Palisades Technical Specifications 2.6 NMC Corporate Directive 5.6, Inservice Inspection Standard 2.7 Westinghouse Owners Group WCAP-14572, Revision 1-NP-A,

'Westinghouse Owners Group Application of Risk-Informed Methods to Piping Inservice Inspection Topical Report' and WCAP-14572, Revision 1-NP-A, Supplement 1, 'Westinghouse Structural Reliability and Risk Assessment (SRRA) Model for Piping Risk-Informed Inservice Inspection" 4 of 96

PALISADES NUCLEAR PLANT 4 TH INTERVAL INSERVICE INSPECTION PLAN 2.8 Westinghouse Owners Group WCAP-1 4572, Revision 1-NP-A,

'Westinghouse Owners Group Application of Risk-Informed Methods to Piping Inservice Inspection Topical Report" and WCAP-14572, Revision 1-NP-A, Supplement 2, 'Westinghouse Structural Reliability and Risk Assessment (SRRA) Model for Piping Risk-Informed Inservice Inspection" 2.9 ASME Section XI, Appendix VIII, 2001 Edition 3.0 DISCUSSION 3.1 The Inservice Inspection (ISI) Program implements the pre-service and in-service inspection and pressure testing requirements mandated by the Code of Federal Regulations (CFR) in accordance with the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code, Section Xl.

3.2 Per 10CFR50.55a, licensees are required to update their ISI Programs for each ten-year interval of the facility operating license. The first ten-year interval begins at the date of commercial operation. The program must be updated to the inspection, testing and examination requirements of the Edition and Addenda of the ASME Codes approved for use in paragraph (b) of these regulations 12 months prior to the start of the new ten-year interval. The fourth ten-year interval at Palisades Unit 1 begins on December 13, 2006.

3.3 The Palisades nuclear power plant was built in the late 60's and was placed into commercial operation on December 31, 1971. During the first 40-month life of the plant, in order to comply with paragraphs 4.3 and 4.12 of the previous technical specifications (dated September 1, 1972) of the operating license DPR-20 for the Palisades nuclear plant, which discusses ISI requirements of class 1 components and systems, the nondestructive examinations were performed to satisfy the requirements of the ASME Section XI Code, 1971 edition including the winter 1972 addenda. In February 1976, the NRC amended paragraph 55a(g) of 10 CFR 50 to require nuclear plants to upgrade their technical specifications in the area of the ISI requirements and the functional testing of pumps and valves. By amending paragraph 55a(g) and by invoking Regulatory Guide 1.26, the NRC required nuclear plants to upgrade their systems to include not only class 1 systems, but also class 2 and class 3 systems in their ISI programs.

3.4 Palisades has implemented a Risk Informed Inservice Inspection (RI-ISI)

Program using the Westinghouse Owners Group WCAP-14572, revision 1-NP-A, 'Westinghouse Owners Group Application of Risk-Informed Methods to Piping Inservice Inspection Topical Report" and WCAP-14572, revision 1-NP-A, supplement 1, 'Westinghouse Structural Reliability and Risk Assessment (SRRA) Model for Piping Risk-Informed Inservice Inspection" 5 of 96

PALISADES NUCLEAR PLANT 4 TH INTERVAL INSERVICE INSPECTION PLAN 4.0 REQUIREMENTS 4.1 ASME Section XI Code and Regulatory Requirements The ASME Section Xl Code and Regulatory requirements in this ISI Program shall, as a minimum, comply with the following:

• Code of Federal Regulations, Title 10, Part 50.55a

• ASME Boiler and Pressure Vessel Code, Section Xl, 2001 Edition through 2003 Addenda

" Palisades Relief Requests for the fourth ten-year interval The Inservice Inspection Program and any major revisions shall be submitted to the NRC as required. In addition, requests for specific relief shall be forwarded to the NRC for authorization prior to implementation.

The program elements and the required documentation must be maintained on site for audit.

4.2 ISI Requirements The general requirements for ISI are contained in the ASME B&PV Code, Section Xl, Article IWA. Article IWA includes requirements for examination methods, personnel qualification, flaw evaluation, repair/replacement, hydrostatic and system pressure tests, and records and reports.

ISI requirements for examination and pressure testing vary depending on whether the components are ASME Code Class 1, 2, or 3.Section XI Code Cases contain alternative rules, which may be voluntarily adopted.

A complete list of ISI Code Cases adopted, or considered for future use, at Palisades are contained in Appendix A.

4.3 Section XI Requirements for ASME Code Class 1 Components The components within the ASME Code Class 1 boundary are subject to the volumetric, surface and visual examination requirements and pressure test requirements of Article IWB-2500 and Table IWB-2500-1.

Component supports must meet the examination requirements of IWF-2500 and Table IWF-2500-1. The examination categories of table IWB-2500-1 are as follows:

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PALISADES NUCLEAR PLANT 4 TH INTERVAL INSERVICE INSPECTION PLAN Examination Category Examination Area B-A Pressure retaining welds in the reactor vessel.

B-B Pressure retaining welds in vessels other than the reactor vessel.

B-D Full penetration welded nozzles in vessels.

B-F* Pressure retaining dissimilar metal welds in vessel nozzles.

B-G-1 Pressure retaining bolting greater than 2 inches in diameter.

B-G-2 Pressure retaining bolting 2 inches or less in diameter.

B-J* Pressure retaining welds in piping.

B-K Welded attachments for vessels, piping, pumps and valves.

B-L-1 Pressure retaining welds in pump casings.

B-L-2 Pump casings.

B-M-1 Pressure retaining welds in valve bodies.

B-M-2 Valve bodies.

B-N-1 Interior of the reactor vessel.

B-N-2 Welded core support structures and interior attachments to the reactor vessel.

B-N-3 Removable core support structures.

B-O Pressure retaining welds in control rod housings.

B-P All pressure retaining components.

B-Q Steam Generator tubing.

Note: *B-F and B-J welds are included in the Risk-Informed Inspection Plan 4.4 Section XI Exemptions for ASME Code Class 1 Components The following components or parts of components are exempt from the volumetric and surface examination requirements of IWB-2500:

" Components that are connected to the reactor coolant system and part of the reactor coolant pressure boundary, and that are of such a size and shape so that upon postulated rupture the resulting flow of coolant from the reactor coolant system under normal plant operating conditions is within the capacity of makeup systems that are operable from on-site emergency power.

" Piping of NPS 1, except for steam generator tubing.

• Components and their connections in piping of NPS 1 and smaller

• Reactor vessel head connections and associated piping NPS 2 and smaller, made inaccessible by control rod drive penetrations.

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PALISADES NUCLEAR PLANT 4 TH INTERVAL INSERVICE INSPECTION PLAN Supports are exempt from the examination requirements of IWF-2000 are those connected to piping and other items exempted from volumetric, surface, VT-1, or VT-3 as outlined above.

4.5 Section XI Requirements for ASME Code Class 2 Components The components within the ASME Code Class 2 boundary are subject to the volumetric, surface, and visual examination requirements and pressure test requirements of Article IWC-2500 and Table IWC-2500-1.

Component supports must meet the examination requirements of IWF-2500 and Table IWF-2500-1. The examination categories of Table IWC-2500-1 are as follows:

Examination Category Examination Area C-A Pressure retaining welds in the pressure vessels.

C-B Pressure retaining nozzle welds in vessels.

C-C Welded attachments for vessels, piping, pumps, and valves.

C-D Pressure retaining bolting greater than 2 inches in diameter.

C-F-i* Pressure retaining welds in austenitic stainless steel or high alloy steel piping.

C-F-2* Pressure retaining welds in carbon or low alloy steel piping.

C-G Pressure retaining welds in pumps and valves.

C-H All pressure retaining components.

Note: *C-F-1 and C-F-2 Welds are included in the Risk-Informed Inspection Plan 4.6 Section Xl Exemptions for ASME Code Class 2 Components The following components or parts of components are exempt from the volumetric and surface examination requirements of IWC-2500:

• Components within Residual Heat Removal (RHR), Emergency Core Cooling (ECC), and Containment Heat Removal (CHR) or portions of systems

1. Piping NPS 4 and smaller and associated vessels, pumps, and valves and their connections, for all systems with the exemption of the high pressure safety injection system.

2. Piping NPS 1 1/2 and smaller and associated vessels, pumps, and valves and their connections in high pressure safety injection systems.

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PALISADES NUCLEAR PLANT 4 TH INTERVAL INSERVICE INSPECTION PLAN

3. Vessels, piping, pumps, valves and other components and component connections of any size in statically pressurized, passive (i.e., no pumps) safety injection systems.

4. Piping and components of any size beyond the last shutoff valve in open-ended portions of systems that do not contain water during normal plant operating conditions.

Components within systems or portions of systems other than RHR, ECC, and CHR systems

1. Piping NPS 4 and smaller, for all systems with the exception of auxiliary feedwater system.

2. Piping NPS 1 1/2 and smaller in the auxiliary feedwater system.

3. Vessels, pumps, and valves and their connections in piping NPS 4 and smaller, for all systems except auxiliary feedwater system.

4. Vessels, pumps, and valves and their connections in piping NPS 1 /2 and smaller in the auxiliary feedwater system.

5. Vessels, piping, pumps, valve, other components, and component connections of any size in systems or portions of systems that operate (when the system function is required) at a pressure equal to or less than 275 psig and at a temperature equal to or less than 200-F.

6. Piping and components of any size beyond the last shutoff valve in open-ended portions of systems that do not contain water during normal plant operating conditions.

• Welds or portions of welds that are inaccessible due to being encased in concrete, buried underground, located inside a penetration, or encapsulated by guard pipe.

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PALISADES NUCLEAR PLANT 4 TH INTERVAL INSERVICE INSPECTION PLAN 4.7 Section XI Requirements for ASME Code Class 3 Components The components within the ASME Code Class 3 boundary are subject to the visual examination and pressure test requirements of Article IWD-2500 and Table IWD-2500-1. Component supports must meet the examination requirements of IWF-2500 and Table IWF-2500-1. The examination categories of Table IWD-2500-1 are as follows:

Examination Category Examination Area D-A Welded attachments for vessels, piping, pumps and valves.

D-B All pressure retaining components.

4.8 Section Xl Exemptions for ASME Code Class 3 Components The following components or parts of components are exempt from the VT-1 visual examination requirements of IWD-2500:

• Integral attachments of supports and restraints to piping, vessels, pumps and valves and their connections in piping NPS 4 and smaller.

" Integral attachments of supports and restraints to components that operate at a pressure of 275 psig or less and at a temperature of 200-F or less, in systems (or portions of systems) whose function is not required in support of reactor residual heat removal, containment heat removal, and emergency core cooling.

• Welds or portions of welds that are inaccessible due to being encased in concrete, buried underground, located inside a penetration, or encapsulated by guard pipe.

4.9 Risk-Informed Examination Requirements Palisades has implemented a Risk Informed Inservice Inspection (RI-ISI)

Program using the Westinghouse Owners Group WCAP-1 4572, revision 1-NP-A, 'Westinghouse Owners Group Application of Risk-Informed Methods to Piping Inservice Inspection Topical Reporft and WCAP-14572, revision 1-NP-A, supplement 1, 'Westinghouse Structural Reliability and Risk Assessment (SRRA) Model for Piping Risk-Informed Inservice and WCAP-14572, Supplement 2, Westinghouse Owners Group Application of Risk-Informed Methods to Piping Inservice Inspection topical Report Clarifications."

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PALISADES NUCLEAR PLANT 4 TH INTERVAL INSERVICE INSPECTION PLAN Piping Structural elements that fall under RI-ISI Category R-A are risk ranked in accordance with WCAP-14572, Rev.1 -NP-A, 'Westinghouse Owners Group Application of Risk-Informed Methods to Piping Inservice Inspection Topical report.

The RI-ISI Program element examinations are performed in accordance with Relief Request RR 4-10 4.10 Inspection Schedule Palisades will follow the inspection schedule of IWB/C/D-2412 and IWF-2410, Inspection Program B for the fourth ten-year interval. The inspection interval shall include three inspection periods.

Per IWA-2430(d) an inspection period may be decreased or extended by as much as one year to enable examinations to coincide with a plant outage provided the adjustment does not cause successive intervals to be altered by more than one (1) year from the original pattern of intervals.

Inspection Period Minimum Maximum Interval Examinations Examinations Completed, % Completed, %

1 16 50 4 th 2 501 75 3 100 100 Note: (1) If the first period completion percentage for any examination category exceeds 34%, at least 16% of the required examinations shall be performed in the second period.

5.0 INSERVICE INSPECTION PROGRAM

SUMMARY

TABLES The following Tables in this section provide a summary of the Section XI component, component support, and system pressure testing examinations and tests for the fourth inservice inspection interval at the Palisades Nuclear Plant.

The format of the Inservice Inspection Summary Tables is shown below with an explanation for each column. The detailed examination schedule for components is contained in the Fourth Interval Master Inservice Inspection Plan.

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PALISADES NUCLEAR PLANT 4 TH INTERVAL INSERVICE INSPECTION PLAN Examination Item Description Exam Required Deferral Total Number Number Number Number RR Category Number Method Exams to End Number Req. 1st 2n 3'r No/

Period Period Period Code Case (1) (2) (3) (4) (5) (6) (7) (8) (9) (10) (11) (12)

(1) Examination Category Provides the examination category and description as identified in ASME Section XI, Tables IWB-2500-1, IWC-2500-1, IWD-2500-1, IWF-2500-1 and Table 4.1-1 for Risk-Informed Examinations.

(2) Item Number Provides the Item number as identified in ASME Section XI, Tables IWB-2500-1, IWC-2500-1, IWD-2500-1, IWF-2500-1 and Table 4.1-1 for Risk-Informed Examinations.

(3) Item Number Description Provides the description as identified in ASME Section XI, Tables IWB-2500-1, IWC-2500-1, IWD-2500-1, IWF-2500-1 and Table 4.1-1 for Risk-Informed Examinations.

(4) Examination Method Provides the examination method(s), i.e., volumetric, surface, visual, required by ASME Section Xl, Tables IWB-2500-1, IWC-2500-1, IWD-2500-1, IWF-2500-1 and Table 4.1-1 for Risk-Informed Examinations.

(5) Required Exams Provides for the examinations required by ASME Section XI, Tables IWB-2500-1, IWC-2500-1, IWD-2500-1, IWF-2500-1 and Table 4.1-1 for Risk-Informed Examinations.

(6) Deferral to End Provides for whether a deferral of examination requirement required by ASME section XI, Table IWB-2500-1 is permissible.

(7) Total Number Provides the number of components within an item Number.

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PALISADES NUCLEAR PLANT 4 TH INTERVAL INSERVICE INSPECTION PLAN (8) Number Required Provides the number of components required by ASME Section Xl, Tables IWB-2500-1, IWC-2500-1, IWD-2500-1, IWF-2500-1 and Table 4.1-1 for Risk-Informed Examinations.

(9) Number 1st Period Provides the number of components selected for examination during the first period of the fourth inspection interval.

(10) Number 2 nd Period Provides the number of components selected for examination during the second period of the fourth inspection interval.

(11) Number 3 rd Period Provides the number of components selected for examination during the third period of the fourth inspection interval.

(12) Relief Request/Code Case Provides a listing of relief request and/or code cases applicable to the Section XI Item Number. If a relief request or code case is identified, see Appendix A or Appendix B as applicable.

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PALISADES NUCLEAR PLANT 4 Th INTERVAL INSERVICE INSPECTION PLAN Class 1 Exams Inservice Inspection Summary Table Exam Category Item Description Exam Required Exams Total Number Number Number Number RR No/

Number Method Exams Deferral to Active Required 1st Period 2nd 3rd Code End Period Period Case B-A B1.11 Circumferential- Volumetric All Welds Yes 3 3 - - 3 Pressure Retaining Shell Welds Welds in Reactor B13.12 Longitudinal- Shell Volumetric All Welds Yes 9 9 - - 9 Vessel Welds B13.21 Circumferential- Volumetric Acces. Yes 1 1 - - 1 Head Welds Length All Welds B1.22 Meridional- Head Volumetric Acces. Yes 6 6 - - 6 RR 4-1 Welds Length All Welds B1.30 Shell-to-Flange Volumetric Weld Yes 1 1 - - 1 Weld I I I I I Note: Welds for the reactor vessel closure head (1-B13.21, 6-B13.22 and 1-B13.40 welds) are not included in the above table since Palisades plans on replacing the reactor vessel closure head prior to these examinations being required. The new reactor vessel closure head will not contain these welds.

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PALISADES NUCLEAR PLANT 4 TH INTERVAL INSERVICE INSPECTION PLAN Class 1 Exams Inservice Inspection Summary Table Exam Item Description Exam Required Exams Total Number Number Number Number RR No/

Category Number Method Exams Deferral Active Required 1st 2nd 3rd Code to End Period Period Period Case Pressurizer (Shell-to Head B-B Welds)

Pressure B2.1 1 Circumferential Volumetric All Welds No 2 2 1 1 Retaining B2,12 Longitudinal Volumetric lft of one No 4 2 1 1 Welds in weld Vessels intersecting Other Than B2.11 welds Reactor Pressurizer (Head Welds)

Vessels B2.21 Circumferential- Head Welds Volumetric One Weld No 2 1 - 1 RR 4-2 per Head B2.22 Meridional- Head Welds Volumetric One Weld No 8 8 4 4 RR 4-2 per Head Steam Generators (Primary Side)

B2.31 Circumferential- Head Welds Volumetric One Weld No 4 1 - - 1 -

per Head B2.32 Meridional- Head Welds Volumetric One Weld No 10 1 1 -

per Head B2.40 Tube Sheet-to-Head Weld Volumetric Weld No 2 1 -1 Heat Exchangers (Primary Side)

B2.51 Circumferential- Head Welds Volumetric One Weld No 2 1 -

per Head B2.80 Tubesheet-to-Shell Welds Volumetric Welds at No 2 1 1 1 1 Each End 15 of 96

PALISADES NUCLEAR PLANT 4 TH INTERVAL INSERVICE INSPECTION PLAN Class 1 Exams Inservice Inspection Summary Table Exam Item Description Exam Required Exams Total Number Number Number Number RR No/

Category Number Method Exams Deferral to Active Required 1st Period 2nd Period 3rd Period Code End Case Reactor Vessel B-D B3.90 Nozzle-to-Vessel Volumetric All Nozzles Yes 6 6 - 6 Full Welds Penetration B3.100 Nozzle Inside Volumetric All Nozzles Yes 6 6 - - 6 Welds of Radius Section Nozzles in Pressurizer Vessels- B3.110 Nozzle-to-Vessel Volumetric All Nozzles No 6 6 2 2 2 RR 4-2 Inspection Welds Program B B3.120 Nozzle Inside Volumetric All Nozzles No 6 6 2 2 2 Radius Section Steam Generators (Primary Side)

B3.130 Nozzle-to-Vessel Volumetric All Nozzles No 6 6 2 1 3 RR 4-3 Welds B3.140 Nozzle Inside Volumetric All Nozzles No 6 6 2 1 3 -

Radius Section Heat Exchanger (Primary Side)

B3.150 Nozzle-to-Vessel Volumetric All Nozzles No 4 4 2 1 1 RR 4-4 Welds B3.160 Nozzle Inside Volumetric All Nozzles No 2 2 1 - 1 RR 4-4 Radius Section 16 of 96

PALISADES NUCLEAR PLANT 4 Th INTERVAL INSERVICE INSPECTION PLAN Class 1 Exams Inservice Inspection Summary Table Exam Item Description Exam Required Exams Exams Total Number Number Number Number RR No/

Category Number Method Deferral to Active Required 1st 2nd 3rd Code End Period Period Period Case Reactor Vessel B-G-1 B6.10 Closure Head Nuts Visual, All Nuts >2" Yes 54 54 - - 54 -

Pressure VT-1 Retaining B6.20 Closure Studs Volumetric All Studs > 2" Yes 54 54 - - 54 -

Bolting, Greater B6.40 Threads in Flange Volumetric All Stud Holes > Yes 1 1 - - 1-Than 2in. in 2"1 Diameter B6.50 Closure Washers, Visual, All Washers >2" Yes 54 54 - - 54 -

Bushings VT-1 Pumps I B6.180 Bolts and Studs Volumetric All Bolts and Yes 4 Studs >2" B6.190 Flange Surfaces, Visual, When Yes 4 when connection is VT-1 Disassembled disassembled B6.200 Nuts, Bushings and Visual, All Nuts, Bushings Yes 4 Washers VT-1 and Washers >2" 17 of 96

PALISADES NUCLEAR PLANT 4 TH INTERVAL INSERVICE INSPECTION PLAN Class 1 Exams Inservice Inspection Summary Table Exam Item Description Exam Required Exams Total Number Number 1st Number Number RR No/

Category Number Method Exams Deferral to Active Required Period 2nd Period 3rd Period Code End Case Reactor B-G-2 Vessel Pressure B7.10 Bolts, Studs Visual, All <=2" No 16 16 5 5 6 Retaining and Nuts VT-1 Bolting, 2in Pressurizer And Less B7.20 Bolts, Studs Visual, All <=2" No 2 2 1 1 in Diameter and Nuts VT-1 Steam Generator B7.30 Bolts, Studs Visual, All <=2" No 4 4 1 2 1 and Nuts VT-1 Piping B7.50 Bolts, Studs Visual, All <=2" No 5 5 2 2 1 and Nuts VT-1 Pumps B7.60 Bolts, Studs Visual, All <=2" No 8 8 3 3 2 and Nuts VT-1 Valves B7.70 Bolts, Studs Visual, All <=2" No 27 27 9 9 9 and Nuts VT-1 CRD Housings B7.80 Bolts, Studs Visual, All <=2" No 90 90 30 30 30 and Nuts VT-1 18 of 96

PALISADES NUCLEAR PLANT 4 Th INTERVAL INSERVICE INSPECTION PLAN Class 1 Exams Inservice Inspection Summary Table Exam Category Item Description Exam Required Exams Total Number Number Number Number RR No/

Number Method Exams Deferral to Active Required 1st Period 2nd Period 3rd Period Code End Case B-K Pressure Welded Vessels Attachments B10.10 Welded Surface 1 of each No 3 For Vessels, Attachments similar Piping, Pumps vessel and Valves Piping B10.20 Welded Surface 10% No 4 1 1 Attachments Pumps B10.30 Welded Surface 1 of each No 16 4 4 Attachments similar pump 19 of 96

PALISADES NUCLEAR PLANT 4 TH INTERVAL INSERVICE INSPECTION PLAN Class 1 Exams Inservice Inspection Summary Table Exam Category Item Description Exam Required Exams Exams Total Number Number Number Number RR No/

Number Method Deferral to Active Required 1st Period 2nd 3rd Code End Period Period Case B-L-1 B12.10 Pump Casing Visual, All Welds 1 Pump Yes 4 1 1 Pressure Welds VT-1 each Group Retaining Welds in Pump Casings B-L-2 B12.20 Pump Casing Visual, Internal Surfaces Yes 4 Pump Casings Internal Surfaces VT-3 if Disassembled B-M-2 B12.50 Valve Internal Visual, Internal Surfaces Yes 14 Valve Bodies Surfaces, VT-3 if Disassembled Exceeding NPS 4

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PALISADES NUCLEAR PLANT 4 TH INTERVAL INSERVICE INSPECTION PLAN Class 1 Exams Inservice Inspection Summary Table Exam Category Item Description Exam Required Exams Total Number Number Number Number RR No/

Number Method Exams Deferral Active Required 1st 2nd 3rd Code to End Period Period Period Case B-N-1 B13.10 Vessel Interior Visual, Accessible No 1 Interior of Reactor VT-3 Areas During Vessel Each Inspection Period B-N-2 B13.50 Interior Visual, Accessible Yes Welded Core Support Attachments Within VT-1 Welds Structures and Beltline Region in Interior Attachments Reactor Vessel to Reactor Vessel B-N-2 B13.60 Interior Visual, Accessible Yes Welded Core Support Attachments VT-3 Welds Structures and Beyond Beltline Interior Attachments Region in Reactor to Reactor Vessel Vessel B-N-3 B13.70 Core Support Visual, Accessible Yes Removable Core Structure in VT-3 Surfaces Support Structure Reactor Vessel 21 of 96

PALISADES NUCLEAR PLANT 4 T' INTERVAL INSERVICE INSPECTION PLAN Class 1 Exams Inservice Inspection Summary Table Exam Category Item Description Exam Required Exams Total Number Number Number Number RR No/

Number Method Exams Deferral to Active Required 1st 2nd 3rd Code End Period Period Period Case B-O B14.10 Welds in CRD Volumetric or 10% Yes 143 4 4 Pressure Retaining Housing Surface Peripheral Welds in Control Housings Rod Housings B-P B15.10 All Pressure Visual, All Each No System Leakage test required RR 4-5 Pressure Retaining Retaining VT-2 Refueling each refueling outage Components Components Outage B-Q B16.20 Steam Generator Volumetric* -

Steam Generator Tubing in U-Tube Tubing Design

• Steam Generator Tubing will be inspected in accordance with Palisades Technical Specifications 5.5.8 22 of 96

PALISADES NUCLEAR PLANT 4 Th INTERVAL INSERVICE INSPECTION PLAN Class 2 Exams Inservice Inspection Summary Table Exam Item Description Exam Required Exams Exams Total Number Number Number Number RR No/

Category Number Method Deferral to Active Required 1st Period 2nd 3rd Code End Period Period Case C-A C1.10 Shell Volumetric At Gross N/A 8 4 1 2 1 RR 4-6 Pressure Circumferential Structural &

Retaining Welds Discontinuities RR 4-7 Welds in Cl .20 Head Volumetric Head-to-Shell N/A 6 3 1 Pressure Circumferential Weld Vessels Welds C1.30 Tubesheet-to- Volumetric Tube-to-Shell N/A 4 2 1 RR 4-7 Shell Weld Weld Exam Item Description Exam Required Exams Total Number Number Number Number RR No/

Category Number Method Exams Deferral Active Required 1st 2nd 3rd Code to End Period Period Period Case C-B C2.21 Nozzle-to-Shell (Nozzle- Surface and All Nozzles N/A 16 6 2 2 2 RR 4-7 Pressure to-Head or Nozzle-to- Volumetric Under C-F Retaining Nozzle) Weld without Nozzle Welds Reinforcing Plate, >1/2" In Nominal Thickness Vessels C2.22 Nozzle Inside Radius Volumetric All Nozzles N/A 6 3 1 1 1 CC Section Under C-F N-311 23 of 96

PALISADES NUCLEAR PLANT 4 Th INTERVAL INSERVICE INSPECTION PLAN Class 2 Exams Inservice Inspection Summary Table Exam Category Item Description Exam Required Exams Exams Total Number Number Number Number RR No/

Number Method Deferral to Active Required 1st Period 2nd 3rd Code End Period Period Case C-C C3.10 Welded Surface 100% of required N/A 12 6 6 Welded Attachments to area of each Attachments for Pressure Vessels attachment Vessels, Piping, Pumps C3.20 Welded Surface 100% of required N/A 52 6 2 2 2 and Valves Attachments to area of each Piping attachment Exam Category Item Description Exam Required Exams Total Number Number Number Number RR No/

Number Method Exams Deferral to Active Required 1st 2nd 3rd Code End Period Period Period Case C-H C7.10 Pressure Retaining Visual, System No System Leakage test required All Pressure Components - VT-2 Leakage Test each inspection period Retaining System Leakage each Period Components Test 24 of 96

PALISADES NUCLEAR PLANT 4 T' INTERVAL INSERVICE INSPECTION PLAN Class 3 Exams Inservice Inspection Summary Table Exam Category Item Description Exam Required Exams Exams Total Number Number Number Number RR No/

Number Method Deferral to Active Required 1st 2nd 3rd Code End Period Period Period Case D-A D1.20 Welded Visual, 100% of N/A 49 6 2 2 2 Welded Attachments Attachments - VT-1 required area for for Vessels, Piping, Piping each attachment Pumps and Valves Exam Category Item Description Exam Required Exams Total Number Number Number Number RR No/

Number Method Exams Deferral to Active Required 1st Period 2nd 3rd Period Code End Period Case D-B D2.10 Pressure Visual, System No System Leakage test required each All Pressure Retaining VT-2 Leakage Test inspection period Retaining Components each Period Components 25 of 96

PALISADES NUCLEAR PLANT 4 TH INTERVAL INSERVICE INSPECTION PLAN Support Exams Inservice Insoection Summarv Table Exam Item Description Exam Required Exams Total Number Number Number Number RR No/

Category Number Method Exams Deferral to Active Required 1st Period 2nd Period 3rd Period Code End Case F-A F1.1OA Class 1 Piping Supports Visual, 25% N/A 100 25 9 8 8 -

Supports (One Direction Support) VT-3 F1.1OB Class 1 Piping Supports Visual, 25% N/A 23 6 2 2 2 -

(Multi Directional VT-3 Supports)

F1.1OC Class 1 Piping Supports Visual, 25% N/A 26 7 3 2 2 -

(Spring Can Supports VT-3 and Snubbers)

F-A F1.20A Class 2 Piping Supports Visual, 15% N/A 141 22 8 7 7 -

Supports (One Direction Support) VT-3 F1.20B Class 2 Piping Supports Visual, 15% N/A 95 15 5 5 5 -

(Multi Directional VT-3 Supports)

F1.20C Class 2 Piping Supports Visual, 15% N/A 129 20 7 7 6 -

(Spring Can Supports VT-3 and Snubbers)

F-A F1.30A Class 3 Piping Supports Visual, 10% N/A 141 15 5 5 5 -

Supports (One Direction Support) VT-3 F1.30B Class 3 Piping Supports Visual, 10% N/A 97 10 4 3 3 -

(Multi Directional VT-3 Supports)

F1.30C Class 3 Piping Supports Visual, 10% N/A 36 4 2 1 1 -

(Spring Can Supports VT-3 and Snubbers)

F-A F1.40A Supports Other Than Visual, 100% (One N/A 20 11 2 3 6 -

Supports Piping Supports (Class VT-3 of Multiple) 1, 2,3 and MC)

F1.40B Supports Other Than Visual, 100% (One N/A 12 6 6 -

Piping Supports (Class VT-3 of Multiple)

I______ 1_____1, 2,3 and MC) I I 1 _1 1 26 of 96

PALISADES NUCLEAR PLANT 4 TH INTERVAL INSERVICE INSPECTION PLAN Risk-Informed Exams Inservice Inspection Summary Table Exam Item Description Exam Required Exams Total Number Number Number Number RR No/

Category Number Method Exams Deferral Active Required 1st 2nd 3rd Code to End Period Period Period Case R-A R1.11 Elements Subject to Volumetric Per Relief No 2958 126 42 42 42 RR 4-8 Risk Informed Thermal Fatigue Request Piping RR-10 Examinations Submittal R1.12 Elements Subject to High Visual, Per Relief No 401 All socket welds are VT-2 Cycle Fatigue VT-2 Request inspected on a refueling outage RR-10 frequency Submittal R1.1 5 Elements Subject to Volumetric Per Relief No 11 7 3 2 2 RR 4-8 Primary Water Stress Request Corrosion Cracking RR-10 (PWSCC) Submittal R1.16 Elements Subject to Volumetric Per Relief No 101 17 6 6 5 Intergranular or Request Transgranular Stress RR-10 Corrosion Cracking Submittal (IGSCC or TGSCC)

R1.17 Elements Subject to Volumetric Per Relief No 424 Exams performed in accordance with localized Microbiologically Request Palisades MIC Program Influenced Corrosion RR-10 (MIC) or Pitting Submittal R1.18 Elements Subject to Flow Volumetric Per Relief No 11103 Exams performed in accordance with Accelerated Corrosion Request Palisades FAC Program (FAC) RR-10 I Submittal 27 of 96

PALISADES NUCLEAR PLANT 4 TH INTERVAL INSERVICE INSPECTION PLAN 6.0 AUGMENTED AND OWNER ELECTED EXAMINATIONS This section identifies augmented and owner elected inspection programs mandated within the ISI Program.

Augmented examinations are not required by ASME Section XI. However, due to the nature of the augmented requirements, these programs have been referenced within the ISI Program for tracking purposes. These augmented programs satisfy NRC requirements and other applicable site commitments.

Owner elected examinations are not required by ASME and are not regulatory commitments. These examinations have been determined to be good practice due to operating experience, engineering judgment, etc., and may be an internal site commitment.

Each of these Augmented or Owner elected inspections are scheduled within the ISI Program Master schedule of examinations.

Augmented and owner program revisions or deviations shall be governed by the referenced documents and without need to submit a revised program plan to the regulator.

6.1 Technical Specification Required Augmented Examinations

a. Technical Specification 5.5.6 Reactor Coolant Pump Flywheels Technical Specification Section 5.5.6 requires an inspection program that provides for a 100% volumetric inspection of the upper flywheels each 10 years.

b. Operating Requirements Manual Section 4.12 Augmented Inservice Inspection Program for High Energy Lines Outside of Containment The program applies to welds in piping systems or portions of systems located outside containment where protection from the consequences of postulated ruptures is not provided by a system of pipe whip restraints, jet impingement barriers, protective enclosures and/or other measures designed specifically to cope with such ruptures.

For the Palisades Plant, this specification applies to welds in the Main Steam and Main Feedwater lines located inside the Main Steam and Feedwater Penetration Rooms.

OBJECTIVE To provide assurance of the continued integrity of the piping systems over the service lifetime.

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PALISADES NUCLEAR PLANT 4 TH INTERVAL INSERVICE INSPECTION PLAN SPECIFICATION 4.12.1 For welds identified in the Operating Requirements Manual, Figure 4.12.A, (Main Steam Lines) and Figure 4.12.B, (Feedwater Lines)

a. The inspection at each weld shall be performed in accordance with the requirements of ASME Section XI Code, with the following schedule.

Fourth Inspection Interval Volumetric inspection of 1/3 of the welds at the expiration of each period of the inspection interval with a cumulative 100 percent coverage of all welds.

Note: The welds selected during each inspection period shall be distributed among the total number to be examined to provide a representative sampling of the condition of all welds.

A. Examinations that reveal unacceptable structural defects in a weld shall be extended to require inspection of an additional 1/3 of the welds. If further unacceptable defects are detected in this additional sampling, the remainder of the welds shall be examined.

B. In the event repairs of any weld is required following the examinations performed during successive inspection intervals, the inspection schedule for the repaired weld(s) shall revert back to the first inspection interval schedule.

4.12.2 For other welds (excluding those identified in figure 4.12.A and 4.12.B)

A. Welds in the Main Steam lines including the safety valve attachment welds and in the Feedwater lines shall be examined in accordance with the requirements of Subsections ISC-100 through 600 of the 1972 Winter Addendum of ASME Section Xl Code.

4.12.3 For welds in the Main Steam and Main Feedwater lines located inside the Main Steam and Feedwater Penetration rooms.

A. A visual inspection on the surfaces of the insulation at the weld locations shall be performed on a weekly basis for detection of leaks. Any detected leaks shall be investigated and evaluated.

If leakage is caused by a through wall flaw, either the plant shall be shut down or the leaking piping isolated. Repairs shall be performed prior to returning the line to service.

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PALISADES NUCLEAR PLANT 4 TH INTERVAL INSERVICE INSPECTION PLAN B. Repairs, re-examination and piping pressure tests shall be conducted in accordance with the rules of ASME Section XI Code.

Basis:

Under normal plant operating conditions, the piping materials operate under ductile conditions and within the stress limits considerably below the ultimate strength properties of the materials. Flaws, which could grow under such conditions, are generally associated with cyclic loads that fatigue the material, and lead to leakage cracks. The examinations and the frequency of inspection will provide a means for timely detection even before the flaw penetrates the wall of the piping.

6.2 Miscellaneous Augmented Examinations

a. DEFENSE IN DEPTH The Primary Coolant System (PCS) piping will continue to receive a system pressure test and Visual VT-2 examination as currently required by the Code. In addition, the welds connecting the PCS hot and cold leg loop piping to the Reactor Vessel nozzles will continue to be inspected as part of the ASME Section XI vessel inspection program. This includes twelve PCS loop welds, two per nozzle.

b. ALLOY 600 Palisades Alloy 600 Inspection Program is based on the following documents and commitments:

The reactor vessel head will be inspected in accordance with NRC Order EA-03-009,"Issuance of First Revised NRC Order (EA 009) Establishing Interim Inspection Requirements for Reactor Pressure Vessel Heads at Pressurizer Water Reactors".

Alloy 600 butt welds are categorized and inspected in accordance with MRP-139 "Material Reliability Program: Primary System Piping Butt Weld Inspection and Evaluation Guideline".

The Pressurizer Heater sleeve penetrations will be inspected in accordance with NRC Bulletin 2004-01" Other Alloy 600 penetrations (hot and cold leg penetrations) that are not covered by the above requirements shall receive a bare metal examination using the following schedule:

For pressurizer and hot leg penetrations perform a bare metal visual examination each refueling outage.

For cold leg penetrations perform a bare metal visual examination once every three refueling outages.

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PALISADES NUCLEAR PLANT 4 TH INTERVAL INSERVICE INSPECTION PLAN

c. Weld Overlays In 1994 Palisades installed two weld overlays for repair of the containment sump check valves. Palisades will perform a surface and a volumetric examination of one overlay during the fourth interval with guidance taken from EPRI TR-1 1932 for Category E.

d. ALLOY 600 COMMITMENTS Commitments from letter to NRC dated June 7, 1995 Complete inspections for PWSCC of the areas to which the mechanical stress improvement process (MSIP) is applied every other refueling outage.

The six affected welds are:

PCS-12-PSL-1H1-1 PCS-12-PSL-11H 1-2 PCS-12-PSL-11H1-7 PCS-12-PSL-11H 1-8 PCS-12-SCS-2H1-1 PCS-12-SCS-2H1 -2 Perform a PWSCC ultrasonic examination of the Pressurizer Spray Nozzle welds at the Pressurizer on a frequency of every other refueling outage. The two affected welds are PCS-4-PSS-1P1-21 and PCS-4-PSS-1 P1-20.

6.3 Owner Elected Examinations

a. Structural Integrity of Auxiliary Feedwater System Piping Palisades plan for examination for the structural integrity of the Auxiliary Feedwater System Piping associated with the Steam Generators, "reference letters RJB 34-88, dated May 18, 1988, BVV 88-032, dated July 14, 1988 and THF 88-001, dated January 28, 1988, which shows evidence of examinations already performed. (See revision 7 of the Master Plan for referenced letters).

A synopsis of those letters mentioned above consists of the examinations listed below:

1. Pipe to Elbow - Perform volumetric examinations

2. Elbow to Pipe - Perform volumetric examinations

3. Pipe to Nozzle - Perform volumetric examinations

4. Perform ultrasonic wall thinning examinations beginning at the Elbow to Pipe weld downstream of the Steam Generators. Beginning refueling outage number 14 (1999 Outage), these examinations will be 31 of 96

PALISADES NUCLEAR PLANT 4 TH INTERVAL INSERVICE INSPECTION PLAN performed under the plants Flow Accelerated Corrosion (FAC) program.

5. Perform visual examinations of the internal knuckle region, provided the Steam Generators are open for secondary side inspections.

The above examinations are to be performed once each 3 1/3 years (equivalent to once each ISI period). These examinations are to apply to both Steam Generators and are included in this plan.

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PALISADES NUCLEAR PLANT 4 Th INTERVAL INSERVICE INSPECTION PLAN Augmented Exams Inservice Inspection Summary Table Exam Item Description Exam Required Exams Total Number Number Number Number RR No/

Category Number Method Exams Deferral Active Required 1st 2nd 3rd Code to End Period Period Period Case Augmented D-I-D Defense-In-Depth UT-M All Nozzles Yes 12 12 12 examinations for the PCS welds connecting the hot and cold leg loop piping to the reactor vessel nozzles 88-08 NRC Bulletin 88-08 Thermal Volumetric Every 5- No 1 1 1 1 Stresses in Piping years Connected to Reactor Coolant System AF-SI Structural Integrity of the Aux Volumetric Each No 8 8 8 8 8 Feedwater piping associated Inspection with the Steam Generators Period ORM4.12 Main Steam and Feedwater Volumetric 1/3 of No 24 24 8 8 8 Lines located inside the Main Exams Steam and Feedwater Each Period Penetration Rooms.

Overlay Weld Overlay Repairs on Volumetric 25% per No 2 1 - 1 -

Containment Sump Check Interval RR 4-9 Valves RG 1.14 *Primary Coolant Pump Volumetric Each Yes 4 4 1 1 2 Flywheel Examination Inspection interval

• Examination required by Palisades Technical Specification 5.5.6 33 of 96

PALISADES NUCLEAR PLANT 4 TH INTERVAL INSERVICE INSPECTION PLAN Augmented Alloy 600 Exams Inservice Inspection Summary Table Exam Item Description Exam Required Exams Exams Total Number Number Number Number RR No/

Category Number Method Deferral Active Required 1 st 2nd 3rd Code to End Period Period Period Case Augmented MRP-139A Resistant Materials Volumetric Each Inspection Yes 2 2 - - 2 Alloy 600 Interval MRP-139C Non-Resistant Volumetric Each Inspection No 6 6 2 2 2 material Mitigated Interval by Stress Improvement MRP-139D Non-Resistant Volumetric Each Inspection No 2 2 2 2 2 Material Period Pressurizer and Hot Leg >=4' MRP-139E Non-Resistant Volumetric 100% Every 6 No 8 8 4 4 2 Material Years Pressurizer and Cold Leg >=4' MRP-139H Non-Resistant Volumetric Each Inspection No 3 3 3 3 3 Material if possible Period Pressurizer and Hot Leg <4' MRP-139J Non-Resistant BMV Each Refueling No 22 22 22 22 22 Material in which Pressurizer and Hot Volumetric is not Leg performed MRP-139K Non-Resistant BMV Once Every No 24 16 10 10 12 Material Three Refuelings Pressurizer and Cold Leg 34 of 96

PALISADES NUCLEAR PLANT 4 TH INTERVAL INSERVICE INSPECTION PLAN Augmented Alloy 600 Exams Inservice Inspection Summary Table Exam Item Description Exam Required Exams Total Number Number Number Number RR No/

Category Number Method Exams Deferral to Active Required 1st Period 2nd 3rd Period Code End Period Case Augmented BL-2004-01 Alloy 600 BMV Every No 120 120 120 120 120 -

Alloy 600 Pressurizer Heater Refueling Penetrations Outage AL-600 33/44" Hot Leg BMV Every No 20 20 20 20 20 -

Penetrations Refueling Outage 3

AL-600 " Cold Leg BMV Once Every No 16 16 16 16 16 -

Penetrations Three Refuelings AL-600 Alloy 600 Hot Leg BMV Every No 10 10 10 10 10 -

J-Welds Refueling Outage AL-600 Alloy 600 Cold Leg BMV Once Every No 12 12 12 12 12 -

J-Welds Three Refuelings AL-600 Pressurizer BMV Every No 2 2 2 2 2 -

Temperature Refueling Elements Outage AL-600 Steam Generator BMV Once Every No 4 4 4 4 4 -

Bowl Plugs Three Refuelings 35 of 96

PALISADES NUCLEAR PLANT 4 TH INTERVAL INSERVICE INSPECTION PLAN 7.0 CODE CASES Several Code Cases may be used during the fourth inservice inspection interval.

Those Code Cases which have been approved for use by the NRC by inclusion in Revision 14 of Regulatory Guide 1.147 and are planned to be used by Palisades during the fourth inspection interval are listed in Appendix A.

8.0 RELIEF REQUESTS Many of the requirements of ASME Section Xl Subsections IWA, B, C, D, and F cannot be practically implemented. 10CFR50.55a allows exemption to the requirements of the ASME Section Xl Code on a case-by-case basis with the approval from the Director of the Office of Nuclear Reactor Regulation. The applicant for the exemption must demonstrate that:

• The proposed alternatives would provide an acceptable level of quality and safety, or

• Compliance with the specified requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.

Many relief requests have been previously discussed pertaining to the specific section of the code they affect. All relief requests are listed in Appendix B.

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PALISADES NUCLEAR PLANT 4 TH INTERVAL INSERVICE INSPECTION PLAN 9.0 Drawings During the development of the ISI Program Plan many drawings and sketches are used to aid the process and provide an easy means by which to locate individual components, items, welds and supports. The following is a list of those drawings and sketches used in the preparation of the ISI Program Plan and schedule.

Class 1 Drawings COMPONENT IDENTIFICATION DRAWING NUMBER (S)

Reactor Pressure Vessel A-1 Pressurizer T-72 A-2 Steam Generator E-50A A-3 Steam Generator E-50B A-3 Reactor Coolant Loops A-4 Regenerative Heat Exchanger E-56A A-34 Regenerative Heat Exchanger E-56B A-34 Primary Coolant Pump P-50A A-32 Primary Coolant Pump P-50B A-32 Primary Coolant Pump P-50C A-32 Primary Coolant Pump P-50D A-32 Group 1 Valve Internals M107 Sh 2199, 2236, 2235 and 2371 Group 2 Valve Internals M107 Sh 2371, 2372, 2373 and 2374 Group 3 Valve Internals and Valve Body Welds M107 Sh 2244 Group 1 Valve Bolting M107 Sh 2199, 2236, 2235 and 2371 Group 2 Valve Bolting M107 Sh 2371, 2372, 2373 and 2374 Group 3 Valve Bolting M1 07 Sh 2244 Group 4 Valve Bolting M107 Sh 2350, 2349, 2348 Ml10 Sh 121, 1165, 997, 996, 998 and 1434 PCS-12-PSL-1 H1 M107 Sh 2057 PCS-12-SCS-2___1 M107 Sh 2244 PCS-4-PRS-1 P1 M107 Sh 2154 PCS-4-PRS-1 P2 M107 Sh 2154 PCS-4-PRS-1 P3 M107 Sh 2154 PCS-4-PSS-1P1 M107 Sh 2347 and 2348 PCS-3-PSS-1 B1 M107 Sh 2350 PCS-3-PSS-2A1 M107 Sh 2348 and 2349 PCS-2-DRL-1A1 Ml10 Sh 113 PCS-2-LDL-2B1 M110 Sh 130 ESS-6-1Sl-1A1 M107 Sh 2373 ESS-6-SIS-1 B1 M107 Sh 2374 ESS-6-SIS-2A1 M107 Sh 2372 ESS-6-SIS-2B1 M107 Sh 2371 ESS-2-SlS-1A1 Ml10 Sh 243 ESS-2-SIS-1 BI Ml10 Sh 243 ESS-2-SIS-2A1 Ml10 Sh 243 37 of 96

PALISADES NUCLEAR PLANT 4 T' INTERVAL INSERVICE INSPECTION PLAN COMPONENT IDENTIFICATION DRAWING NUMBER (S)

ESS-2-SIS-2B1 M110 Sh 243 ESS-2-LTC-1 B M107 Sh 2454 CVC-2-CHL-1Al Ml10 Sh 997 CVC-2-CHL-1A2 M110 Sh 997 CVC-2-CHL-2A1 M110 Sh 998 CVC-2-LDL-2B1 M110 Sh 121 CVC-2-LDL-2B2 M110 Sh 1665 and 1434 CVC-2-PSS-1Pl Ml10 Sh 856, 857 and 858 PCS-6-PRS-1 Al M107 Sh 2153 PCS-6-PRS-1 B1 M107 Sh 2158 PCS-6-PRS-1 Cl M107 Sh 2157 PCS-4-PRS-1 P2 M107 Sh 2154 PCS-4-PRS-1 P3 M107 Sh 2154 38 of 96

PALISADES NUCLEAR PLANT 4 TH INTERVAL INSERVICE INSPECTION PLAN Class 2 Drawings COMPONENT IDENTIFICATION DRAWING NUMBER (S)

Concentrated Boric Acid Tank B-3 SIRW Tank B-146 Iodine Removal NaOH Tank B-148 Iodine Removal NaOH Make-Up Tank B-148 Safety Injection Tanks B-5 Boric Acid Filter B-6 Steam Generator E-50A B-8 Steam Generator E-50B B-8 Regenerative Heat Exchanger E-56A B-9 Regenerative Heat Exchanger E-56B B-9 Shutdown Cooling Heat Exchanger E-60A B-i 1 Shutdown Cooling Heat Exchanger E-60B B-11 Containment Spray Pumps B-131 Charging Pumps B-131 Concentrated Boric Acid Pumps B-133 High Pressure Safety Injection Pumps B-133 Low Pressure Safety Injection Pumps B-1_33 SIRW Recirculation Pump B-149 MSS-36-MSL-1S1 M101 Sh 2764 and 3021 MSS-36-MSL-2S1 M101 Sh 2763 and 3019 MSS-8-MSV-1Si M101 Sh 3218 MSS-8-MSV-1 S2 M101 Sh 3219 MSS-8-MSV-2S1 M101 Sh 3216 MSS-8-MSV-2S2 M101 Sh 3217 FWS-18-FWL-1S1 M101 Sh 2734 FWS-18-FWL-2S1 M101 Sh 2732 FWS-6-AWS-1S1 M101 Sh 2785 FWS-6-AWS-2S1 M101 Sh 2785 ESS-24-SIS-SH1 M107 Sh 2278 ESS-18-SIS-SH1 M107 Sh 2278 ESS-24-SIS-SH2 M107 Sh 2281 ESS-18-SIS-SH2 M107 Sh 2281 ESS-14-CSS-1PA M107 Sh 2279 ESS-8-CSS-1 PA M107 Sh 2170 ESS-14-CSS-1PB M107 Sh 2282 ESS-8-CSS-1 PB M107 Sh 2170 ESS-10-CSS-1PB M107 Sh 2170 ESS-14-CSS-1 PC M107 Sh 2282 ESS-8-CSS-1 PC M107 Sh 2170 ESS-14-SDC-LPC M107 Sh 2280 ESS-14-SDC-LPD M107 Sh 2280 ESS-14-SIS-HPA M107 Sh 2279 ESS-6-SIS-HPA M107 Sh 2279 ESS-14-SIS-LPA M107 Sh 2280 39 of 96

PALISADES NUCLEAR PLANT 4 TH INTERVAL INSERVICE INSPECTION PLAN COMPONENT IDENTIFICATION DRAWING NUMBER (S)

ESS-10-SIS-LPA M107 Sh 2171 ESS-14-SIS-LPB M107 Sh 2280 ESS-1 0-SIS-LPB M107 Sh 2171 ESS-8-SIS-1A6 M107 Sh 2370 ESS-6-SIS-1A6 M107 Sh 2370 ESS-8-SIS-1 B6 M107 Sh 2370 ESS-14-SCS-2H1 M107 Sh 2280 and 2244 ESS-12-SDC-XCO M107 Sh 2172 ESS-10-SDC-XCO M107 Sh 2172 ESS-12-SIS-SDC M107 Sh 2170 ESS-12-SIS-1A5 M107 Sh 2199 ESS-12-SIS-1 B5 M107 Sh 2236 ESS-12-SIS-1C5 M107 Sh 2233 ESS-12-SIS-1D5 M107 Sh 2071 ESS-12-SIS-1Al M107 Sh 2199 ESS-12-SIS-1 B1 M107 Sh 2236 ESS-12-SIS-2A1 M107 Sh 2235 ESS-12-SIS-2B1 M107 Sh 2071 ESS-6-SIS-1Al M107 Sh 2373 ESS-6-SIS-1 B1 M107 Sh 2374 ESS-6-SIS-2A1 M107 Sh 2372 ESS-6-SIS-2B1 M107 Sh 2371 ESS-12-SIS-1LP M107 Sh 2171,2172 and 2370 ESS-10-SDC-XlB M107 Sh 2170 ESS-10-CSS-SLA M107 Sh 2172 ESS-8-CSS-SLA M107 Sh 2172 ESS-6-CSS-SLA M107 Sh 2172 ESS-8-CSS-SLA M107 Sh 2456 and 2457 ESS-8-CSS-SLB M107 Sh 2173 ESS-6-CSS-SLB M107 Sh 2173 ESS-8-CSS-SLB M107 Sh 2375 and 2376 ESS-1 0-CSS-1 P3 M107 Sh 2170 ESS-10-SDC-X1A M107 Sh 2170 ESS-10-SDC-XOA M107 Sh 2172 ESS-12-SDC-XOA M107 Sh 2172 ESS-6-SDC-RE1 M107 Sh 2201 ESS-8-SIS-HPB M107 Sh 2283 ESS-6-SIS-HPB M107 Sh 2283 ESS-6-SIS-HPC M107 Sh 2283 ESS-6-SIS-1HP M107 Sh 2245 and 2065 ESS-6-SIS-2HP M107 Sh 2072 and 2246 ESS-6-SIS-CRH M107 Sh 2200 and 2201 ESS-6-LTC-1 A M107 Sh 2455 ESS-6-LTC-1 B M107 Sh 2454 ESS-3-SIS-HPA M107 Sh 2245 ESS-3-SIS-HPB M107 Sh 2245 ESS-2-SIS-HRA M107 Sh 2248 40 of 96

PALISADES NUCLEAR PLANT 4 TH INTERVAL INSERVICE INSPECTION PLAN COMPONENT IDENTIFICATION DRAWING NUMBER (S)

ESS-2-SIS-HRB M107 Sh 2247 ESS-4-SIS-2HP M107 Sh 2246 ESS-3-SIS-3HP M107 Sh 2246 ESS-4-SIS-CRH M107 Sh 2200 ESS-4-SIS-HPB M107 Sh 2245 ESS-3-SIS-LRA M107 Sh 2200 ESS-6-BL-CRH M107 Sh 2204 SFP-8-CPL-DL1 M107 Sh 2560 and 2023 SFP-6-CPL-SL1 M107 Sh 2667 and 2465 SFP-6-SRT-RL1 M107 Sh 2204 SFP-6-SRT-SL1 M107 Sh 2021 SWS-16-ORS-RH1 M101 Sh 2745 VAS-8-CPU-RL2 B-120 VAS-8-CPU-RL1 B-120 41 of 96

PALISADES NUCLEAR PLANT 4 TH INTERVAL INSERVICE INSPECTION PLAN Class 3 Drawings COMPONENT IDENTIFICATION DRAWING NUMBER (S)

FWS-6-AWS-OLA M101 Sh 2716 FWS-6-AWS-OLB M101 Sh 2716 FWS-6-AWS-SLA M101 Sh 2761 FWS-6-AWS-1S1 M1i01 Sh 2939 FWS-6-AWS-SLC M101 Sh 2978 FWS-6-AWS-OLC M101 Sh 2983 FWS-4-AWS-2S3 M101 Sh 2977- 1 and 2 FWS-4-AWS-1S3 M101 Sh 2977- 1 and 3 FWS-2-AWS-OLC Ml10 Sh 1758- 1,2 and 3 FWS-1.5-AWS-OLA C-16 FWS-1.5-AWS-OLB C-1 6 FWS-6-CMU-SH1 C-237 FWS-12-CMU-SH3 C-236 SWS-24-CCS-RLH M101 Sh 2744 SWS-24-CSW-HCL M101 Sh 3028 SWS-24-CSW-SH1 M101 Sh 3027, 3028 and 2810 SWS-24-CSW-SH2 M101 Sh 3029 and 2811 SWS-16-CRS-RH1 M101 Sh 2784 SWS-16-CRS-SH1 M101 Sh 3028 and 2797 SWS-16-SWP-OLA M101 Sh 2810 SWS-1 6-SWP-OLC M101 Sh 2811 SWS-12-CRS-SH1 M101 Sh 2790 SWS-12-CRS-SH2 M101 Sh 2797 SWS-10-CRS-SL1 M101 Sh 2795 SWS-24-SWP-CDL M101 Sh 2811 SWS-10-CRS-SL3 M101 Sh 2790 SWS-10-CRS-SL4 M101 Sh 2797 SWS-12-CRS-RH2 M101 Sh 2784 SWS-10-CRS-RL1 M101 Sh 2781 and 2782 SWS-10-CRS-RL3 M101 Sh 2779 SWS-10-CRS-RL4 M101 Sh 2784 SWS-6-CRS-1 R1 M101 Sh 2782 SWS-8-CRS-RL2 M101 Sh 2777 SWS-6-CRS-2R2 M101 Sh 2778 SWS-6-CRS-3R1 M101 Sh 2780 SWS-6-CRS-3R2 M101 Sh 2776 42 of 96

PALISADES NUCLEAR PLANT 4 TH INTERVAL INSERVICE INSPECTION PLAN COMPONENT IDENTIFICATION DRAWING NUMBER (S)

SWS-6-CRS-4R1 Ml01 Sh 2783 SWS-6-CRS-4R2 M101 Sh 2721 SWS-6-CRS-1Si Ml01 Sh 2796 SWS-6-CRS-2S1 M101 Sh 2792 SWS-6-CRS-2S2 Ml01 Sh 2792 and 2793 SWS-6-CRS-3S1 M101 Sh 2767 SWS-6-CRS-4S1 Ml01 Sh 2791 SWS-6-CRS-4S2 M101 Sh 2774 SWS-6-EPS-RLA Ml01 Sh 3035 SWS-6-EPS-RLB Ml 01 Sh 3033 and 3032 SWS-6-EPS-SLA M101 Sh 3030 and 3031 SWS-6-EPS-SLB Ml01 Sh 3038 and 3039 SWS-6-RE-BL M101 Sh 6331 CCS-24-CPU-1 PA M101 Sh 2842 CCS-20-CHX-1 Pi M101 Sh 2740 CCS-20-CHX-RLB M101 Sh 3226 CCS-20-CPU-1 PA Ml01 Sh 2842 CCS-1 6-CPU-1 PA M101 Sh 2740 CCS-20-CPU-1 PB M101 Sh 2843 CCS-1 6-CPU-1 PB Ml01 Sh 2740 CCS-20-CPU-1 PC M101 Sh 2843 CCS-16-CPU-iPC Ml01 Sh 2740 CCS-20-SCH-1 P1 M101 Sh 3226 CCS-20-SCH-1 P2 M101 Sh 3226 CCS-20-CHX-1 PC M101 Sh 2740 CCS-18-RHC-1P1 Ml01 Sh 2842 CCS-12-SDC-RLA Ml01 Sh 2842 CCS-14-ARH-1P1 M101 Sh 2844 CCS-14-ASH-1P1 Ml01 Sh 3226 and 3227 CCS-10-ASH-IP1 M101 Sh 3227 CCS-10-ARH-1P1 Ml01 Sh 2844 CCS-10-RWS-1 P1 M101 Sh 3227 CCS-1 0-RWS-1 P2 Ml01 Sh 2844 CCS-10-CSH-1P1 Ml01 Sh 3226 CCS-1 0-CSH-1 P3 C-58 CCS-10-CSH-1P4 Ml01 Sh 2843 SFP-1 4-FPF-RL1 M107 Sh 2020 SFP-1 0-FPP-1 PA M107 Sh 2020 43 of 96

PALISADES NUCLEAR PLANT 4 TH INTERVAL INSERVICE INSPECTION PLAN COMPONENT IDENTIFICATION DRAWING NUMBER (S)

SFP-10-FPP-1PB M107 Sh 2020 SFP-8-FPP-1 PA M107 Sh 2671 SFP-8-CPL-DL1 M107 Sh 2022 and 2023 SFP-6-SRT-SL1 M107 Sh 2022 SFP-12-SFX-SL1 M107 Sh 2671 SFP-8-FPP-1 PB M107 Sh 2671 SFP-6-SRT-RL1 M107 Sh 2669 SFP-12-SFX-RL1 M107 Sh 2665 DMW-6-CST-AFS Ml01 Sh 2761 M107 Sh 2364 DMW-6-CST-AWS C-190,192 and 193 DMW-16-CST-AWS C-236 44 of 96

PALISADES NUCLEAR PLANT 4 TH INTERVAL INSERVICE INSPECTION PLAN Risk-Informed Drawings SEGMENT IDENTIFICATION I DRAWING NUMBER (S)

Auxiliary Feedwater AFW-006B C-16A AFW-010 C-196 AFW-01 1 C-1 96A AFW-012 M101 Sh. 2983, 2939 AFW-016 M101 Sh. 2938 AFW-01 8 M101 Sh. 2937 Blowdown BLD-003 M101 Sh. 6419 BLD-004 M101 Sh. 6420 BLD-005A M101 Sh. 6419 BLD-006A M101 Sh. 6420 Condensate CDS-01 3 M101 Sh. 6369, 6381, 84, 6371, 243, 240 CDS-014 1 M101 Sh. 6369 Concentrated Boric Acid CBA-012 I M107 Sh. 2409, 2415 Critical Service Water CSW-004 M101 Sh. 2811 CSW-005A M101 Sh. 2810, 3027, 2811 CSW-005B M101 Sh. 3027 CSW-006A M101 Sh. 2811, 2812, 3029 CSW-006B M101 Sh. 3029 CSW-007 M101 Sh. 3028 CSW-008 M101 Sh. 3028 CSW-009 M101 Sh. 3028 CSW-013 M101 Sh. 3035 CSW-016 M101 Sh. 2744 CSW-017 M101 Sh. 2744 CSW-021 M101 Sh. 3038 Extraction Steam Heaters and Drains HED-001 M101 Sh. 6376 HED-002 M101 Sh. 6376 HED-003 M101 Sh. 1359, 6391 HED-005 M101 Sh. 1361, 6392 HED-006 M101 Sh. 6389 45 of 96

PALISADES NUCLEAR PLANT 4 TH INTERVAL INSERVICE INSPECTION PLAN SEGMENT IDENTIFICATION DRAWING NUMBER (S)

HED-007I M101 Sh. 6390 Non-Critical Service Water NSW-001 M101 Sh. 2816, 321 NSW-004 M10 Sh. 3545, M101 Sh. 1244, 1247 NSW-005 M101 Sh. 1847, 3683, 3688 NSW-01OA M101 Sh. 2746, 2774, 2775 NSW-010B M101 Sh. 2747 NSW-010C M101 Sh. 3035 Circulating Water CWS-011 (Dilution Pump P-40B) M653 Sh. 3 (C-1)

CWS-012 (Dilution Pump P-40A) M653 Sh. 3 (C-2)

Main Steam MSS-027 M101 Sh. 6375, M3.08 MSS-028 M3.08 MSS-036 M3.08 MSS-037 M3.08 MSS-041 M101 Sh. 322 MSS-046 M101 Sh. 6405 MSS-047 M101 Sh. 6401 MSS-048 M101 Sh. 6404, 3302 MSS-049 M101 Sh. 6406, 3301 MSS-051 M101 Sh. 6405 MSS-052 M101 Sh. 6402 MSS-059 M101 Sh. 6407, 3303 MSS-060 M101 Sh. 6408, 3304 MSS-065 M101 Sh. 6413 MSS-066 M101 Sh. 6414 MSS-067 M101 Sh. 6413 MSS-069 M101 Sh. 6414 MSS-071 M101 Sh. 6375 MSS-072 M101 Sh. 6400 MSS-073 M101 Sh. 6404 MSS-074 M101 Sh. 6402 MSS-075 M101 Sh. 6401 MSS-076 M101 Sh. 6403 Low Pressure Safety Injection LPI-001 M107 Sh. 2199 LPI-001A M107 Sh. 2373 46 of 96

PALISADES NUCLEAR PLANT 4 TH INTERVAL INSERVICE INSPECTION PLAN SEGMENT IDENTIFICATION DRAWING NUMBER (S)

LPI-002 M107 Sh. 2235 LPI-002A M107 Sh. 2372 LPI-003 M107 Sh. 2236 LPI-003A M107 Sh. 2374 LPI-004 M107 Sh. 2371 LPI-004A M107 Sh. 2371 Pressurizer PZR-001 M107 Sh. 2057 PZR-002 M1-LA Sh. 985 PZR-003 Mi-LA Sh. 985 PZR-004 M1-LA Sh. 985 PZR-005 M1-LA Sh. 985 PZR-006 M107 Sh. 2157 PZR-007 M107 Sh. 2158 PZR-008 M107 Sh. 2153 PZR-009 M107 Sh. 2154 PZR-010 M107 Sh. 2154 PZR-014A M110 Sh. 1806 PZR-01 5 M107 Sh. 2348, 2347 PZR-01 6 M107 Sh. 2350 PZR-01 7 M107 Sh. 2348, 2349 PZR-01 8 M201 Sh. 1 (B-6)

PZR-019 M110 Sh. 856 PZR-020 M201 Sh. 2 Primary Coolant PCS-011 M110 Sh. 3493, 3694, 3692, 3658 PCS-012 M110 Sh. 3693, 3694 M201 Sh. 1(D-1)

M201 Sh. 1(B-1)

PCS-013 Ml10 Sh. 3495, 3696 M201 Sh. 1(D-7)

M201 Sh. 1(E-7)

PCS-014 Ml10 Sh. 3496, 3695, 3659 M201 Sh. 1(D-7)

PCS-01 5 Ml10 Sh. 113 PCS-01 6 Ml10 Sh. 113 PCS-017 M110 Sh. 130 PCS-018 M110 Sh. 130 47 of 96

PALISADES NUCLEAR PLANT 4 TH INTERVAL INSERVICE INSPECTION PLAN SEGMENT IDENTIFICATION DRAWING NUMBER (S)

PCS-01 9A M 110 Sh. 3497 Typical PCS-01 9B M110 Sh. 3497 Typical PCS-020A M110 Sh. 3497 Typical PCS-020B M1 10 Sh. 3497 Typical PCS-021A M1 10 Sh. 3497 Typical PCS-021 B M110 Sh. 3497 typical PCS-022A M1 10 Sh. 3497 Typical PCS-022B M110 Sh. 3497 typical PCS-023 Ml10 Sh. 113 PCS-026 PCS Book PCS-027 PCS Book PCS-034A M110 Sh. 1806 PCS-035 M107 Sh. 2199 PCS-036 M107 Sh. 2071 PCS-037 M107 Sh. 2236 PCS-038 M107 Sh. 2235 PCS-039 M107 Sh. 2244 PCS-040 Ml10 Sh. 997 PCS-041 Ml10 Sh. 998 PCS-042 M201 Sh. 1 PCS-043 M201 Sh. 1 SIRW and Containment Sump Suction SSS-001 M107 Sh. 2281 SSS-001A M204 Sh. 1B (D-6)

SSS-001 B M107 Sh. 2281 SSS-002 M107 Sh. 2278 SSS-002A M204 Sh. 1B (B-3)

SSS-002B M107 Sh. 2278 SSS-002C M107 Sh. 2278 SSS-002D M107 Sh. 2278 SSS-003G M107 Sh. 2282 SSS-005 M107 Sh. 2281 SSS-005A M107 Sh. 2281 SSS-006 M107 Sh. 2278 SSS-006A M107 Sh. 2278 SSS-007 M107 Sh. 2107, 2105, 2106, 2021 SSS-008 M107 Sh. 2410, 2411, 2413 SSS-008A M107 Sh. 2411 48 of 96

PALISADES NUCLEAR PLANT 4 TH INTERVAL INSERVICE INSPECTION PLAN SEGMENT IDENTIFICATION DRAWING NUMBER (S)

SSS-009 M1 10 Sh. 381 SSS-010 M107 Sh. 5083, 2064 Shutdown Cooling SDC-002B1 M107 Sh. 2280 SDC-002B2 M107 Sh. 2280 SDC-002B3 M107 Sh. 2280 SDC-005 M107 Sh. 2171, 2200 SDC-006 M107 Sh. 2171, 2200 SDC-007A2 M107 Sh. 2171 SDC-009 M107 Sh. 2170 SDC-011A1 M107 Sh. 2172 SDC-011A2 M107 Sh. 2172, Ml10 Sh. 391 SDC-011A3 M107 Sh. 2172 SDC-012A1 M107 Sh. 2173 SDC-012A2 M107 Sh. 2173, Ml10 Sh. 391 SDC-012A3 M107 Sh. 2172, 2173 SDC-020 M107 Sh. 2280 Fire Protection FPS-012A M101 Sh. 1870 FPS-012C M101 Sh. 2870 FPS-014 M101 Sh. 2761 FPS-021 M101 Sh. 1870 FPS-022 M101 Sh. 1870 49 of 96

PALISADES NUCLEAR PLANT 4 TH INTERVAL INSERVICE INSPECTION PLAN Defense in Depth Drawings COMPONENT IDENTIFICATION DRAWING NUMBER (S)

PCS-42-RCL-1 H A-4 PCS-42-RCL-2H A-4 PCS-30-RCL-1 A A-4 PCS-30-RCL-1 B A-4 PCS-30-RCL-2A A-4 PCS-30-RCL-2B A-4 Augmented Drawings COMPONENT IDENTIFICATION DRAWING NUMBER (S)

PCS-12-PSL-1 Hi M107 Sh 2057 PCS-4-PSS-1 P1 M107 Sh 2347 PCS-12-SCS-2H1 M107 Sh 2244 PCS-4-PRS-1 P1 M107 Sh 2154 ESS-6-SIS-1Al M107 Sh 2373 ESS-6-SIS-2A1 M107 Sh 2372 ESS-6-SIS-1 B1 M107 Sh 2374 ESS-6-SIS-2B1 M107 Sh 2371 PCS-2-PSS-1 P1 Ml10 Sh. 856 MSS-36-MSL-1 Si Ml01 Sh 3021 MSS-36-MSL-2S1 M101 Sh 3019 MSS-8-MSV-i Si M101 Sh 3218 MSS-6-RVR-1 S4 Ml01 Sh 3021 MSS-6-RVR-2S1 Ml01 Sh 3019 MSS-6-RVR-2S4 M101 Sh 3019 MSS-6-RVR-2S6 M101 Sh 3019 FWS-18-FWL-1S1 Ml01 Sh 2912 and 2734 FWS-18-FWL-2S1 Ml01 Sh 2914 and 2732 FWS-4-AWS-1 Si M101 Sh 2940 FWS-3-AWS-1S1 Ml01 Sh 2940 FWS-4-AWS-2S1 M101 Sh 2941 FWS-3-AWS-2S1 M101 Sh 2941 ESS-24-SIS-SH1 M107 Sh 2278 ESS-24-SIS-SH2 M107 Sh 2281 Primary Coolant Pump P-50A A-32 Primary Coolant Pump P-50B A-32 Primary Coolant Pump P-50C A-32 Primary Coolant Pump P-50D A-32 50 of 96

PALISADES NUCLEAR PLANT 4 TH INTERVAL INSERVICE INSPECTION PLAN 10.0 CALIBRATION BLOCKS The following table identifies the calibration blocks that are used for ultrasonic examination of components at the Palisades Nuclear Plant.

Block Number Drawing Number Material PO Number Drawing Number 10.75-SS-X-.937-1 -PAL C-2249-033 A376, TP316 59459 950Y198*M1Y-38 6-CS-89-WOL-X-2-PAL 0PC009 SAE1 02 26834 CPC009 5-CSCL-3-PAL C-2249-034 SA508,CL2 59459 950Y198*M1Y-6 4.75-5.8-CS-4-PAL D-2249-035 A540, Gr. B24 87321 950Y198*M1Y-21 12X12-CSCL-6-PAL C-2249-037 SA508, CL2 59459 950Y198*M1Y-3 IR-CSCL-7-PAL C-2249-038 SA508, CL2 59459 950Y198*M1Y-5 6-SS-1 20-.562-8-PAL C-2249-047 SA-312, TP304 59459 950Y198*M1Y-4 14SS-10-.250-9-PAL D-2249-057 SA-312, TP304 98538 950Y198*MIY-8 8-SS-10S-.148-10-PAL D-2249-058 SA-312, TP304 98538 950Y198*M1Y-9 6-SS-10S-.134-11-PAL D-2249-059 SA-312, TP304 98538 950Y198*M1Y-10 10-SS-20-.250-12-PAL D-2249-060 SA376, TP304 98538 950Y198*M1Y-11 12-SS-20-.250-13-PAL D-2249-061 SA376, TP304 98538 950Y198*MlY-12 12-SS-140-1.125-14A-PAL D-2249-053 SA182, TP304 93300 950Y198*M1Y-13 4-SS-1 20-.438-15-PAL D-2249-054 SA376, TP316 93300 950Y198*M1Y-14 3-SS-1 60-.438-16-PAL D-2249-052 SA376, TP316 93300 950Y198*M1Y-15 2.5-SS-1 60-.375-17-PAL D-2249-056 SA376, TP316 93300 950Y198*M1Y-16 2-SS-160-.344-18-PAL D-2249-055 SA376, TP316 93300 950Y198*M1Y-17 301D-CSCL-3.0-19-PAL D-2249-088 SA264 93300 950Y198*M1Y-18 421D-CSCL-X-4.0-20-PAL D-2249-089 SA264 93300 950Y198*M1Y-19 6-SS-120-.562-21-PAL D-2249-086 SA376, TP316 98538 950Y198*M1Y-20 51 of 96

PALISADES NUCLEAR PLANT 4 TH INTERVAL INSERVICE INSPECTION PLAN Block Number Drawing Number Material PO Number Drawing Number 4.344-CSCL-22-PAL D-2249-085 SA533, Gr.B 98538 950Y198*M1Y-22 301D-CSS-3.0-23-PAL D-2249-090 SA351, CF8M 98538 950Y198*M1Y-23 4-IN-X-.594-24-PAL D-2249-087 SB166 98538 950Y198*MlY-24 IR-CSCL-25-PAL D-2249-091 SA508, C12 98538 950Y198*M1Y-42 IR-CSCL-26-PAL D-2249-048 SA508, CL2 1001-3663 950Y198*M1Y-25 18-CS-60-.750-27-PAL D-2249-049 SA106, Gr.B 18277 950Y198*M1Y-26 2.4-SS-XX-.552-28A-PAL D-2249-094 SA376, TP316 18277 950Y198*M1 Y-27 8-SS-40S-.322-29-PAL D-2249-095 SA312, TP304 18277 950Y198*MIY-28 6-CS-40-.280-30-PAL D-2249-093 SA1 06, Gr.B 18277 950Y198*M1Y-29 6-SS-80S-.432-31 -PAL D-2249-092 SA376, TP304 18277 950Y198*M1Y-30 6-IN-1.719-32-PAL D-2249-082 SB166 18277 950Y198*M1Y-31 PL-CS-2.5-33-PAL D-2249-081 SA533, Gr.B 19277 950Y198*M1Y-32 7-CSCL-35-PAL D-2249-096 SA533, Gr.B 1001-3915 950Y198*M1 Y-33 5-CSCL-36-PAL D-2249-097 SA533, Gr.B 19277 950Y198*M1Y-34 3-SS-80S-.300-37-PAL D-2249-098 SA376, TP304 18277 950Y198*M1Y-35 3-SS-10S-.120-38-PAL D-2249-099 SA376, TP304 18277 D-2249-099 4-SS-10S-.120-39-PAL D-2249-101 SA312, TP304 18277 950Y198*MlY-36 4-SS-80-.377-40-PAL D-2249-233 SA376, TP304 18277 950Y198*MlY-37 18-CS-X-1.0-41 -PAL D-2249-224 SA106, Gr.B 47642 950Y198*MlY-43 18-CS-X-1.0-42-PAL D-2249-255 SA106, Gr.B 47642 950Y198*M1Y-44 11-CSCL-43-PAL D-2249-230 SA533, Gr.B CP10-1332 950Y198*M1Y-46 9-CSCL-44-PAL D-2249-231 SA533, Gr.B CP1O-1332 950Y198*M1Y-47 N/S-CSCL-45-PAL D-2249-228 SA508, CL2 CP 10-1332 950Y198*M1Y-45 PL-CSCL-1.0-48-PAL 950-Y-198-M1-Y SA515, Gr.70 CP111455Q 950Y198*M1Y-50 52 of 96

PALISADES NUCLEAR PLANT 4 TH INTERVAL INSERVICE INSPECTION PLAN Block Number Drawing Number Material PO Number Drawing Number 7-1.125-8-CS-49-PAL* 950-Y-198-Mi-Y SA540,Gr.B24,CL3 LP08-7492 950Y198*M1Y-51 36-CS-1.250-50-PAL 950Y198*M1Y-56 A106, Gr.B CP11-3365 950Y198*MIY-56 6-SS-160-.719-PAL** 950Y198*M1Y-54 SA312, TP316 CP11-3365 950Y198*MIY-54 256-203 70277-296-002 SA516, Gr.70 Mi-FAA-166 296-204 70277-296-002 SA516, Gr.70 Mi-FAA-166 296-202 70277-296-002 SA516, Gr.70 Mi-FAA-166 296-201 70277-296-002 SA516, GR.70 M1-FAA-166 296-101 70277-296-001 SA533, Gr.B, CL1 Mi-FAA-165 296-102 70277-296-001 SA516, Gr.70 Mi-FAA-165 296-103 70277-296-001 SA516, Gr.70 M1-FAA-165 296-104 70277-296-001 SA533, Gr.B Mi-FAA-165 296-106 70277-296-001 SA508 Mi-FAA-165 51-PAL (Overlay Block) 52-PAL (RPVCH Stud) 53-PAL (PCP Stud) 54-PAL 1243008C IN600 55-PAL 1243010C IN600 56-PAL 1243011C IN600 57-PAL 1243017E IN600 58-PAL 1243013D IN600 59-PAL (ET Std.) 1243009E IN600 C0013583 60-PAL 1243014C IN600 C0013583 61 -PAL (ET Std.) 1243015C IN600 C0013583 53 of 96

PALISADES NUCLEAR PLANT 4 TH INTERVAL INSERVICE INSPECTION PLAN Block Number Drawing Number Material PO Number Drawing Number 62-PAL 8-CS-40-.322-63-PAL MIY-Sh. 60 A106, Gr.B G217724 950Y198*M1Y-60 Alternative Calibration Blocks 4647 CB0282R0 A516, Gr.70 G0364407 4648 CB0283R0 TP304 G0364407 4649 CB0284R0 TP316 G0364407 Mock Up Blocks NZL-MKP-52-PAL Main Coolant Pump 6446 E87 A508, SA516 CSCL-53-PAL Cracked Test Specimen SA533 Clad Pressurizer Spray Nozzle (986-01)

Charging Inlet Nozzle (676-01)

Shutdown Cooling Outlet Nozzle (675-06)

Pressure & Meas. Sampling Nozzle (675-13)

• Replacement for 7-1.125-8-CS-46-PAL

• • Replacement for 6-SS-160-.791-47-PAL 54 of 96

PALISADES NUCLEAR PLANT 4 TH INTERVAL INSERVICE INSPECTION PLAN 11.0 RECORDS An ISI Summary Report including Form NIS-1 is required to be filed with the regulator and jurisdictional authorities within 90 calander days following refueling outage completion. This is considered to be closure of the generator output breaker. Items to be attached to the ISI Summary Report shall include:

• Interval, period and refueling outage number (when applicable)

" A complete list of examined items and components and component supports during the cycle;

" An abstract of examination results;

• Extent of condition noted;

• A description of the type and estimated extent of degradation, and conditions that led to the degradation;

" An evaluation of each area, and the results of the evaluation, and;

" A description of necessary corrective actions;

" The number of additional examinations performed and the results if any;

" Form NIS-1, Owner's Report for Inservice Inspection.

• Form NIS-2, Owner's Report for Repair/Replacement Activities.

" A coversheet providing the following:

1. Date of document completion,

2. Name and address of owner,

3. Name and address of plant,

4. Name and number designation of the unit,

5. Commercial service date for the unit.

55 of 96

PALISADES NUCLEAR PLANT 4 TH INTERVAL INSERVICE INSPECTION PLAN Appendix A Palisades Nuclear Plant Code Cases Code Case Code Case Title Approved By Number N-311 Alternative Examination of Outlet Nozzle on Secondary Side of Reg. Guide Steam Generators, Section Xl, Division 1 1.147, Rev 14 N-432-1 Repair Welding Using Automatic or Machine Gas Tungsten-Arc Reg. Guide Welding (GTAW) Temper Bead Technique 1.147, Rev 14 N-460 Evaluation Criteria for Temporary Acceptance of Flaws in Reg. Guide Moderate Energy Class 2 and 3 piping Section Xl, Division 1 1.147, Rev 14 N-513-2 Evaluation Criteria for Temporary Acceptance of Flaws in Relief Request Moderate Energy Class 2 and 3 piping Section Xl, Division 1 RR-4-12 N-526 Alternative Requirements for Successive Inspections of Class 1 Conditionally and 2 Vessels Acceptable per Reg. Guide 1.147, Rev 14 N-545 Alternative Requirements for Conduct of Performance Reg. Guide Demonstration Detection Test of reactor Vessel 1.147, Rev 14 N-552 Alternative Methods - Qualification for Nozzle Inside Radius Conditionally Section from the Outside Surface Acceptable per Reg. Guide 1.147, Rev 14 N-566-2 Corrective Action for Leakage Identified at Bolted Connections Reg. Guide 1.147, Rev 14 N-586 Alternative Additional Examination Requirements for Class 1, 2, Conditionally and 3 Piping, Components, and Supports Acceptable per Reg. Guide 1.147, Rev 14 N-593 Alternative Examination Requirements for Steam Generator Conditionally Nozzle-to-Vessel Welds Acceptable per Reg. Guide 1.147, Rev 14 N-597-1 Requirements for Analytical Evaluation of Pipe Wall Thinning Conditionally Acceptable per Reg. Guide 1.147, Rev 14 N-613-1 Ultrasonic Examination of Full Penetration Nozzles in Vessels, Reg. Guide Examination Category B-D, Item No's. B3.10 and B3.90, reactor 1.147, Rev 14 Nozzle-to-Vessel Welds, Fig. IWB-2500-7(a), (b), and (c)

N-624 Successive Inspections Reg. Guide 1.147, Rev 14 N-629 Use of Fracture Toughness Test Data to Establish Reference Reg. Guide Temperature for Pressure Retaining Materials 1.147, Rev 14 N-638-1 Similar and Dissimilar Metal Welding Using Ambient Temperature Conditionally Machine GTAW Temper Bead Technique Acceptable per Reg. Guide 1.147, Rev 14 N-639 Alternative Calibration Block Material Conditionally Acceptable per Reg. Guide 1.147, Rev 14 N-641 Alternative Pressure-Temperature Relationship and Low Reg. Guide Temperature Overpressure Protection System Requirements 1.147, Rev 14 N-643 Fatigue Crack Growth Rate Curves for Ferritic Steels in PWR Reg. Guide Water Environment 1.147, Rev 14 56 of 96

PALISADES NUCLEAR PLANT 4 TH INTERVAL INSERVICE INSPECTION PLAN Code Case Code Case Title Approved By Number N-648-1 Alternative Requirements for Inner Radius Examinations of Class 1 Conditionally Reactor Vessel Nozzles Acceptable per Reg. Guide 1.147, Rev 14 N-661 Alternative Requirements for Wall Thickness Restoration of Conditionally Classes 2 and 3 Carbon Steel Piping for Raw Water Service Acceptable per section Xl, Division 1 Reg. Guide 1.147, Rev 14 N-663 Alternative Requirements for Classes 1 and 2 Surface Reg. Guide Examinations 1.147, Rev 14 N-695 Qualification requirements for Dissimilar Metal Piping Welds Reg. Guide 1.147, Rev 14 57 of 96

PALISADES NUCLEAR PLANT 4 TH INTERVAL INSERVICE INSPECTION PLAN Appendix B Palisades Nuclear Plant Relief Requests This Appendix contains relief request written pursuant to the requirements of 10CFR50.55a for situations where applicable ASME Section XI requirements cannot be met.

The following guidance was employed to determine the correct 10CFR50.55a paragraph cited for Palisades Nuclear Plant relief requests.

10CFR5O.55a(a)(3)(i): Cited in relief requests when alternatives to the Section Xl requirements, which provide an acceptable level of quality and safety, are proposed. Examples are relief requests that propose alternative NDE methods and/or examination frequency.

10CFR5O.55a(a)(3)(ii): Cited in relief requests when compliance with the Section Xl requirements is deemed to be a hardship or unusual difficulty without a compensating increase in the level of quality and safety. Examples of hardship and/or unusual difficulty include, but are not limited to, excessive radiation exposure, disassembly of components solely to provide access for examinations, and development of sophisticated tooling that would result in only minimal increases in examination coverage.

10CFR50.55a(g)(6)(i): Cited in relief requests when conformance with Section Xl requirements is deemed impractical.

Examples of impractical requirements are situations where the component would have to be redesigned or replaced to enable the required inspection to be performed.

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PALISADES NUCLEAR PLANT 4 TH INTERVAL INSERVICE INSPECTION PLAN Appendix B Palisades Nuclear Plant Relief Requests The relief request contained in the following table are subject to change throughout the inspection interval.

ISI PROGRAM RELIEF REQUEST INDEX Relief Summary Revision Status Request RR 4-1 Reactor Vessel Lower Head Meridional Welds 0 Submitted RR 4-2 Pressurizer Head, Shell and Nozzle Welds 0 Submitted RR 4-3 Steam Generator Nozzle to Shell Welds 0 Submitted RR 4-4 Regenerative Heat Exchanger Welds 0 Submitted RR 4-5 Reactor Vessel Pressure Test 0 Submitted RR 4-6 Steam Generator Upper Shell to Shell Cone 0 Submitted Welds RR 4-7 Shutdown Heat Exchanger Welds 0 Submitted RR 4-8 Pressurizer Nozzle to Flange Welds 0 Submitted RR 4-9 Appendix VIII, Supplement 11 0 Submitted RR 4-10 Risk-Informed ISI 0 Submitted RR 4-11 IWA-2600 Weld Reference System 0 Submitted RR 4-12 Code Case N-513-2 0 Submitted 59 of 96

PALISADES NUCLEAR PLANT 4 TH INTERVAL INSERVICE INSPECTION PLAN RELIEF REQUEST NUMBER - RR 4-1 COMPONENT IDENTIFICATION Code Class 1 Code Reference IWB-2500 Table IWB-2500-1 Examination Category B-A Item Number B1.22 Component Description Reactor Vessel Lower Head Meridional Welds: 1-113A at 00,1-113B at 600, 1-113C at 120 0,1-113D at 1800, 1-113E at 2400 and 1-113F at 3000 Reference Drawings Combustion Engineering Drawing 232-113 and Sketch NDT-MNA-DSD9515 Sheet 17 of 43 (See Attachment 1)

In accordance with 10 CFR 50.55a(a)(3)(ii), Nuclear Management Company, LLC (NMC) requests authorization to implement examination on the accessible volumes as identified below in lieu of the 100% volumetric requirements at Palisades Nuclear Plant.

CODE REQUIREMENT Table IWB-2500-1, Category B-A. Item B 1.12 requires a volumetric examination of the accessible length of all welds. This is clarified in footnote 2 to examine essentially 100% of the weld length.

BASIS FOR RELIEF The six (6) lower head meridional welds in the reactor vessel are not fully accessible, such that. 100% of the entire length of the weld volumes can not be achieved during an ultrasonic examination from inside of the reactor vessel using remotely operated examination equipment. The limitation on the exam is caused by an internally installed flow ring, a permanent 360* attachment in the upper part of the lower head. The flow ring covers up to the top half of the meridional welds and prevents direct access for scanning. Following are the coverage's which were achieved during the past examination:

1. Meridional Weld 1-113A at 0° - 47%

2. Meridional Weld 1-113B at 60" - 53%

3. Meridional Weld 1-113C at 120" - 53%

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4. Meridional Weld 1-113D at 1800 - 47%

5. Meridional Weld 1-113E at 240* - 53%

6. Meridional Weld 1-113F at 3000 - 53%

Two drawings are provided to support this relief request. the first drawing (232-113) is a copy from the vendor file and is the best available drawing to show the reactor vessel bottom head forming and welding.

The second drawing (Sketch NTD-MNA-DSD9515. Sheet 17 of 43) is taken from the Westinghouse final report to Palisades for the June 1995 reactor vessel examination. This cross sectional drawing details the limitation created by the flow ring and the transducers which were used for the examination of these lower head welds.

Manual ultrasonic examination from the exterior of the vessel would not be feasible due to the large amount of dose required to set-up lighting, prep the examination areas, ultrasonically examine the portions of the weld which were inaccessible from the ID, and demobilize from the area. The contact dose on the lower head is 2.5R and the general dose levels in the room range from 1.5 to 2R.

The expected dose expended to complete this scope of work could easily exceed 20R. The expected benefit of obtaining this data does not outweigh the consequences of exposing personnel to this cumulative dose.

Additionally, there were no indications in the portions of the weld which were examined from the ID using the mechanized tool.

PROPOSED ALTERNATIVE EXAMINATION The accessible weld volumes of each of the 6 identified meridional welds will be ultrasonically examined from the ID using remotely operated mechanized equipment during the performance of the reactor vessel examination.

IMPLEMENTATION SCHEDULE The proposed alternative is requested for the 4 th ten year interval of the Inservice Inspection Program for Palisades Nuclear Plant.

REFERENCE By letter dated January 9, 1997, the NRC Staff previously authorized this relief to Palisades, Docket No 50-255 [TAC NO. M93628] for the 3 rd ten year inspection interval (Previously RR-1 1). Note: The previous relief request was revised by letters from the NRC on October 14,1998 and February 14, 2000

[TAC NO.93628].

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PALISADES NUCLEAR PLANT 4 TH INTERVAL INSERVICE INSPECTION PLAN RELIEF REQUEST NUMBER - RR 4-2 COMPONENT IDENTIFICATION Code Class 1 Code Reference IWB-2500 Table IWB-2500-1 Examination Category B-B and B-D Item Numbers B2.11,B2.21,B2.22 and B3.110 Component Description Pressurizer Upper Shell to Upper Head Weld 5-988 Lower Shell to Lower Head Weld 3-982 Lower Head Circumferential Weld 2-984 Meridional Head Welds:

Lower Head 1-984A through D Nozzle to Shell Welds 1-986, 3-985, 8-986, 8-986A through C Reference Drawings Combustion Engineering Drawings M1-L-A Sh. 982, M1-L-A Sh. 983, M1-L-A Sh. 985, M1-L-A Sh. 986, M1-L-A Sh. 987 (See Attachment 1)

In accordance with 10 CFR 50.55a(a)(3)(ii), Nuclear Management Company, LLC (NMC) requests authorization to implement examination on the accessible volumes as identified below in lieu of the 100% volumetric requirements at Palisades Nuclear Plant.

CODE REQUIREMENT Table IWB-2500-1 requires all of the listed welds to be volumetrically examined during each inspection interval.

BASIS FOR RELIEF FOR WELDS 5-988 & 3-982 Volumetric examination of Welds 5-988 and 3-982 as required to satisfy the examination region E-F-G-H (as referenced in Figure 1WB-2500-1(b) will be limited due to the transition slope from the shell to the heads. Scanning distances are limited by the insulation support rings located on the shell side 7 inches from the centerline of the welds.

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PALISADES NUCLEAR PLANT 4 TH INTERVAL INSERVICE INSPECTION PLAN Percentage of the volumetric examination of the region E-F-G-H as referenced in IWB-2500-1 (B) will be as follows:

1 00 scanning of region E-F-G-H will examine approximately 62% of the required volume.

2 Axial angle beam examinations with the beam direction from the head towards the shell will examine approximately 81% of the required volume

3. Axial angle beam examinations with the beam direction from the shell towards the head will examine approximately 68% of the required volume.

4. The transverse scans with two angle beam directions in both the clockwise and the counter clockwise directions will obtain approximately 92% of the required examination volume E-F-G-H.

BASIS FOR RELIEF FOR WELD 2-984 Due to the component design configuration with relation to the pressurizer heater penetrations in the lower head, weld 2-984 is totally inaccessible for any type of a volumetric or surface examination. This weld is located inside the lower support skirt and lies between the second and third rows of heater penetrations.

The location of the support skirt which is welded to the head near the edge of weld 2-984 does not allow access from the upper side of the weld. Due to the spacing of the heater penetrations, at approximately four inches apart, and the angle of each penetration through the lower head, examination from the bottom-of the weld towards the upper side is not possible. Therefore, no examinations are planned for this weld other than VT-2 system leakage tests.

BASIS FOR RELIEF FOR WELDS 1-984A through D Approximately 75% of the lower head meridional welds 1-984A through D are totally inaccessible due to the support skirt, the skirt bracket assembly and the heater penetrations.

Approximately nine to ten inches of the lower head meridional welds are accessible from the centerline of the lower shell to the lower head weld (3-982) down to the welded support skirt. This accessible area represents 25% of the total weld length for each weld.

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PALISADES NUCLEAR PLANT 4 TH INTERVAL INSERVICE INSPECTION PLAN Of this accessible 25% of total weld length, the total volumetric examination of region E-F-G-H as referenced in Figure IWB-2500-3 will be 100% for all required scans. Therefore this volume on all four lower meridional welds will be examined in lieu of the code required 100% examination of one weld length.

BASIS FOR RELIEF FOR WELD 1-986 Weld 1-986, is the 4 inch spray line nozzle to upper head weld. Based on previous examination data and a thorough review of the design drawings, it has been determined that examination of this weld is limited. The limitation is due to nozzle 7-986 in the scanning area from the head side. This limitation will result in a loss of accessible examination length of 8.7 inches or approximately 24% of the total length of the weld.

Approximately 84% of the total required examination volumes for the regions outlined in Figure IWB-2500-7(a) can be achieved with angle beam direction from the head side towards the nozzle. This takes into account the volumes that can be examined within the 8.7 inch limitation area.

Total required examination volumes that can be obtained with the 00 scan will equal approximately 54% due to the configuration of the weld and the nozzle.

Transverse scan of the total required exam volumes with two angle beams in both the clockwise and the counter clockwise directions will equal approximately 81% of the required volume.

Due to the configuration, no examinations can be performed from the nozzle side towards the head.

BASIS FOR RELIEF FOR WELD 3-985 Weld 3-985, is the 12 inch surge line nozzle to lower head weld. Based on previous examination data and a thorough review of the design drawings, it has been determined that examination of this weld is limited. The limitation is due to the design configuration and the inability to scan three locations of the weld due to insulation studs welded at the toe of the weld.

The insulation studs result in a loss of approximately 7% of the scanning surface when scanning from the head side towards the nozzle. The 93% of the accessible examination area will be volumetrically examined from the head side towards the nozzle and will result in 100% examination of regions C-D-E-F and B-C-F-G and 84% of region A-B-G-H-l as referenced in Figure IWB-2500-7(a).

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PALISADES NUCLEAR PLANT 4 TH INTERVAL INSERVICE INSPECTION PLAN The volumetric examination of region A-B-G-H-I from the nozzle side will examine approximately 34%, approximately 5% of region B-C-F-G will be examined in this direction and approximately 25% of exam volume C-D-E-F will be examined in this direction.

The total required examination volume of regions referenced in Figure IWB-2500-7(a) with the transverse scan of weld 3-985 will equal 84% with two angle beams in both the clockwise and counter clockwise directions.

BASIS FOR RELIEF FOR WELDS 8-986. 8-986A. 8-986B and 8-986C Weld 8-986 is the PORV outlet nozzle to upper head weld, weld 8-986A, B, C are the code safety nozzles to head welds. Volumetric examination of these welds will be limited due the design configuration of the head and other limitations described below. The following discussion is applicable to all four welds.

The 00 scan is limited to 10% of the total required examination volume in the attachment weld region (B-C-F-G) and the nozzle cylinder region (A-B-G-H-I). In the adjoining region (C-D-E-F) the required examination volume for the 0° scan will be 81% due to the limitations produced by insulation studs welded in the area of interest and by the interference of the adjacent nozzles.

Based on a review of the drawings and past examination data, the angle beam scans from the head side towards the nozzle will allow examination of 81% of the examination regions identified on Figure IWB-2500-7(a). This takes into account the configuration and scanning limitations caused by the insulation studs and the proximity of the other nozzles.

Transverse scans with two angle scanning in both the clockwise and counter clockwise directions will examine 81 % of the required volume in the regions referenced in Figure IWB-2500-7(a).

Volumetric examination from the nozzle side is limited in all cases and results in examination of 10% of the required volumes.

PROPOSED ALTERNATIVE EXAMINATIONS The accessible volumes as identified above will be examined in lieu of the 100%

volumetric requirements.

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PALISADES NUCLEAR PLANT 4 TH INTERVAL INSERVICE INSPECTION PLAN IMPLEMENTATION SCHEDULE The proposed alternative is requested for the 4 th ten year interval of the Inservice Inspection Program for Palisades Nuclear Plant.

REFERENCE By letter dated January 9, 1997, the NRC Staff previously authorized this relief to Palisades, Docket No 50-255 [TAC NO. M93628] for the 3 rd ten year inspection interval (Previously RR-6).

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PALISADES NUCLEAR PLANT 4 TH INTERVAL INSERVICE INSPECTION PLAN RELIEF REQUEST NUMBER - RR 4-3 COMPONENT IDENTIFICATION Code Class 1 Code Reference IWB-2500 Table IWB-2500-1 Examination Category B-D Item Number B3.130 Component Description Steam Generator Nozzle to Shell Welds, 1-104-251, 1-102-251A, 1-102-251 B, 2-104-351 2-102-351A, 2-102-351B Reference Drawings M1-FAA Sh. 44 and M1-F-AA Sh. 40 Figure IWB-2500-7(A) (See Attachment 1)

In accordance with 10 CFR 50.55a(a)(3)(ii), Nuclear Management Company, LLC (NMC) requests approval to implement examination on the accessible volumes as identified below in lieu of the 100% volumetric requirements at Palisades Nuclear Plant.

CODE REQUIREMENT Section XI, Table IWB-2500-1 requires all nozzle to vessel welds to be volumetrically examined once during each inspection interval.

BASIS FOR RELIEF For purposes of discussion, Figure 1WB-2500-7(a) (attached) will be used to describe the 4 required weld volumes. With the exception of the nozzle inner radius section, this figure is the closest configuration to our actual nozzles.

Palisades working with EPRI have developed the attached package (See ). The attached information is an excerpt from the EPRI report and is intended to identify the exam volumes within the areas of limitations. The final EPRI report will identify exam volumes within the area of limitations (attached information), exam volumes where no limitations exist and the composite exam volumes. The final composite exam volumes will be slightly higher since this will include the areas where no limitations exist. However, the code required exam volume will not be achieved and this relief request is necessary.

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PALISADES NUCLEAR PLANT 4 TH INTERVAL INSERVICE INSPECTION PLAN The attached package includes:

1) Figure defining exam volumes.

2) Inlet nozzle inner radius coverage table followed by supporting figure.

3) Inlet nozzle-to-shell weld exam volume.

4) Axial scan coverage table for inlet nozzle-to-shell weld followed by supporting figures.

5) Transverse scan (no probe skewing) coverage table for inlet nozzle-to-shell weld followed by supporting figures.

6) Transverse scan (+/-20" probe skewing) coverage table for inlet nozzle-to-shell weld followed by supporting figures.

7) Outlet nozzle inner radius coverage table followed by supporting figure.

8) Outlet nozzle-to-shell weld exam volume.

9) Axial scan coverage table for outlet nozzle-to-shell weld followed by supporting figures.

10) Transverse scan (no probe skewing) coverage table for outlet nozzle-to-shell weld followed by supporting figures.

11) Transverse scan (+/-20" probe skewing) coverage table for outlet nozzle-to-shell weld followed by supporting figures.

There are 2 acronyms used in the EPRI report for identification, they are Consumers Power - Steam Generator Project Inlet Nozzle (CP-SGPIN) and Consumers Power - Steam Generator Project Outlet Nozzles (CP-SGPON).

The computer based modeling was performed on one steam generator and this is intended to address all primary head nozzle welds in both steam generators. The steam generators are identical indesign.

The probe skew angle for the axial exams are identified as 00 and 1800 within the coverage tables. The probe skew angle for the transverse exams were modeled using a 900 and then offset using a 70" and 1100 skew to increase exam volume coverages.

In summary, the examination volumes are limited and the maximum achievable volumes within the areas of limitations are accurately identified. A relief request from the code required examination volumes is necessary. The final EPRI report is on file at Palisades.

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PALISADES NUCLEAR PLANT 4 TH INTERVAL INSERVICE INSPECTION PLAN PROPOSED ALTERNATIVE EXAMINATION All accessible weld volumes will be examined once per interval in lieu of the 100% volumetric examination requirements.

IMPLEMENTATION SCHEDULE The proposed alternative is requested for the 4 th ten year interval of the Inservice Inspection Program for Palisades Nuclear Plant.

REFERENCE By letter dated March 20,1998, the NRC Staff previously authorized this relief to Palisades, Docket No 50-255 [TAC NO. M98925] for the 3 rd ten year inspection interval (Previously RR-4).

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PALISADES NUCLEAR PLANT 4 TH INTERVAL INSERVICE INSPECTION PLAN RELIEF REQUEST NUMBER - RR 4-4 COMPONENT IDENTIFICATION Code Class 1 Code Reference IWB-2500 Table IWB-2500-1 Examination Category B-D Item Numbers B3.150 and B3.160 Component Description Regenerative Heat Exchangers E -56A and E-56B Nozzle to Shell Welds: E-56A Welds 05 and 07 E-56B, Welds 05 and 07 Reference Drawings Mi-HF Sh. 354 and M1-HF Sh. 355 (See Attachment 1)

In accordance with 10 CFR 50.55a(a)(3)(ii), Nuclear Management Company, LLC (NMC) requests approval to implement examination on the accessible volumes as identified below in lieu of the 100% volumetric requirements at Palisades Nuclear Plant.

CODE REQUIREMENT Table IWB-2500-1 requires all to vessel welds to be volumetrically examined during each inspection interval.

BASIS FOR RELIEF FOR WELDS E-56A-05 AND E-56B-05 Note: Welds E-56A-05 and E-56B-05 are identical in configuration and the same limitations apply to both welds.

The Regenerative Heat Exchangers are vertically mounted and Weld #05 is located in the lower head. The accessible area of the circumference for Weld #05 is limited to four inches of the circumference. The remainder of the weld is covered by the support pads which make it inaccessible to any kind of surface or volumetric examination.

The four inches (18% of the circumference) that is accessible can be examined as summarized below:

1. 65% of the four required volumes using a 45* angle beam from the shell side towards the nozzle.

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PALISADES NUCLEAR PLANT 4 TH INTERVAL INSERVICE INSPECTION PLAN

2. 56% of the four required volumes using a 600 angle beam from the shell side towards the nozzle.

3. 47% of the 4 required volumes using a 450 angle beam from the nozzle side towards the shell.

4. 35% of the 4 required volumes using a 600 angle beam from the nozzle side towards the shell.

5. 61% of the total required volumes can be examined with a 00 scan.

6. 69% of the total required examination volumes can be covered with transverse scans in clockwise and counter clockwise scanning directions using 2 angle beams.

PROPOSED ALTERNATIVE EXAMINATION The accessible volumes as identified above will be examined in lieu of the 100%

volumetric requirements.

BASIS FOR RELIEF FOR WELDS E-56A-07 AND E-56B-07 Note: Welds E-56A-07 and E-56B-07 are identical in design and the same limitations apply to both welds.

Upon thorough review of the referenced drawings and a review of previous examination data it has been determined that the required examination volumes for the examination regions referenced in IWB-2550-7(a) are limited due to the configuration of the nozzle and the shell.

1. It is not possible to scan the required examination regions from the nozzle side due to the limited scanning surface available. Therefore, 0%

of the required examination volumes will be obtained from the nozzle side with the beam direction towards the shell for any of the required volumes listed on Figure IWA-2500-7(a).

2. Approximately 17% of the required exam volumes of the examination regions referenced in Figure 1WB-2500-7(a) can be examined using two angle beams with direction going from the shell side towards the nozzle.

3. Approximately13% of the required exam volumes can be examined with the 0Qscan. No examinations can be performed on the nozzle or the weld due to the design configuration.

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4. Approximately 50% of the total required examination volumes can be examined with the transverse angle beam examinations in clockwise and counter clockwise scanning directions.

PROPOSED ALTERNATIVE EXAMINATIONS The accessible volumes as identified above will be examined in lieu of the 100%

volumetric requirements.

IMPLEMENTATION SCHEDULE The proposed alternative is requested for the 4 th ten year interval of the Inservice Inspection Program for Palisades Nuclear Plant.

REFERENCE By letter dated January 9, 1997, the NRC Staff previously authorized this relief to Palisades, Docket No 50-255 [TAC NO. M93628] for the 3 rd ten year inspection interval (Previously RR-5).

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PALISADES NUCLEAR PLANT 4 TH INTERVAL INSERVICE INSPECTION PLAN RELIEF REQUEST NUMBER - RR 4-5 COMPONENT IDENTIFICATION Code Class 1 Code Reference IWB-2500 Table IWB-2500-1 Examination Category B-P Item Number B15.10 Component Description Alternative Testing for Components Under the Reactor Vessel CODE REQUIREMENT Table IWB-2500-1, Examination Category B-P, requires a system leakage test of IWB-5220 shall be conducted prior to plant startup following a reactor refueling outage.

BASIS FOR RELIEF Pursuant to 10CFR50.55a(a)(3)(ii), relief is requested on the basis that the specified requirements above would result in hardship and unusual difficulty without a compensating increase in the level of quality and safety. Also, per 10CFR50.55a(a)(3)(i), the proposed alternate examination will provide an acceptable level of quality and safety.

The area under the reactor vessel is extremely hazardous when the plant is at Hot Shutdown Conditions for system leakage testing. Radiation levels are expected to be greater than 5-10 rem/hr (on contact), which is the maximum measured during cold shutdown. Assuming 2 persons at one hour per person in this area a total dose of 10 to 20 rem of dose would be received.

In addition to radiation concerns, access to the area under the reactor vessel posses various industrial hazards. Of primary concern is confined space and heat stress. Ambient air temperatures with the Primary Coolant System at full pressure and temperature are expected to be approximately 300 degrees.

Access under these conditions would require significant ventilation for cooling.

The access tube to this area is only 30 inches in diameter. This size limits the amount of ventilation possible while allowing personnel access.

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PALISADES NUCLEAR PLANT 4 TH INTERVAL INSERVICE INSPECTION PLAN PROPOSED ALTERNATE EXAMINATION Palisades Nuclear Plant shall determine leakage from piping and components in the area under the reactor vessel in accordance with paragraph IWA-5244 "Buried Components" of ASME Section XI 2001 Edition, 2003 Addenda. This requirement will be satisfied by the conducting of Palisades System Operating Procedure SPO-1, "Primary Coolant System (PCS)", which completes the PCS leak rate calculation. Plant Technical Specification 3.1.5 states, "Ifthe primary coolant system leakage exceeds 1 gpm and the source of the leakage is not identified, reduce unidentified leakage to less than 1 gpm within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, or place the reactor in hot shutdown within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in cold shutdown within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />." Technical Specification 3.1.5d places a more restrictive leakage limit of 0.6 gpm during startups. Technical Specification Table 4.2.2, Item 7 requires this leak rate determination on a daily basis. These limits are approved as documented in Palisades Facility Operating License DPR-20, through Amendment No 161 and are applicable at all times when the Primary Coolant System is greater than cold shutdown conditions.

Additionally, Palisades will perform a remote visual examination of the area under the reactor vessel once per refueling outage. This examination will document active leakage or evidence of leakage which may have occurred during the previous power cycle.

IMPLEMENTATION SCHEDULE The proposed alternative is requested for the 4 th ten year interval of the Inservice Inspection Program for Palisades Nuclear Plant.

REFERENCE By letter dated June 28, 1996, the NRC Staff previously authorized this relief to Palisades, Docket No 50-255 [TAG NO. M95051] for the 3 rd ten year inspection interval (Previously PR-02).

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PALISADES NUCLEAR PLANT 4 TH INTERVAL INSERVICE INSPECTION PLAN RELIEF REQUEST NUMBER - RR 4-6 COMPONENT IDENTIFICATION Code Class 2 Code Reference IWC-2500 Table IWC-2500-1 Examination Category C-A Item Number C1.10 Component Description Steam Generator Upper Shell to Shell Cone Welds, 1-101-221 and 2-101-221 Reference Drawings M1-F-AA Sh. 104 and M1-F-AA Sh. 74 (See Attachment 1)

In accordance with 10 CFR 50.55a(a)(3)(ii), Nuclear Management Company, LLC (NMC) requests approval to implement examination on the accessible volumes as identified below in lieu of the 100% volumetric requirements at Palisades Nuclear Plant.

CODE REQUIREMENT Table IWC-2500-1 requires a volumetric examination of welds at gross structural discontinuities which includes essentially 100% of the weld length.

BASIS FOR RELIEF Based on examination data obtained during the preservice ultrasonic examinations which were performed on the new steam generators in 1990, there are approximately 171 inches of documented limitations on the upper shell. These limitations are caused by welded patches, snubber attachments and the 18 inch feedwater nozzles. The limitations are shown on drawings M1-F-AA Sh. 104 and M1-F-AA Sh. 74.

The axial angle beam scan from shell cone with the beam direction towards the upper shell will allow approximately 77% of the required volume E-F-G-H as noted on Figure IWC-2500-1. Also, there is a 2% loss of coverage area in the required volume due to the configuration of the shell cone. This configuration causes an abrupt transition to exist in the examination area which results in a loss of contact as the exit point of the transducer travels across this point. This condition exists for the entire circumference of the weld. The 2% loss of exam volume of area E-F-G-H exists from either the shell side or the cone side. The 75 of 96

PALISADES NUCLEAR PLANT 4 TH INTERVAL INSERVICE INSPECTION PLAN total examination volume of area E-F-G-H, when scanning from the shell cone side is approximately 75%.

The examination volume with the angle beam direction going from the upper shell towards the shell cone is equal to 98% of the required volume E-F-G-H as referenced in Figure IWC-2500-1. The transverse scans of the weld are not limited.

The total examination volume of region E-F-G-H, with axial crossing beams is limited to approximately 75%.

PROPOSED ALTERNATIVE EXAMINATION The accessible weld volumes as identified above will be examined in lieu of the 100% volumetric examination requirements.

IMPLEMENTATION SCHEDULE The proposed alternative is requested for the 4 th ten year interval of the Inservice Inspection Program for Palisades Nuclear Plant.

REFERENCE By letter dated January 9, 1997, the NRC Staff previously authorized this relief to Palisades, Docket No 50-255 [TAC NO. 93628] for the 3 dten year inspection interval (Previously RR-3).

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PALISADES NUCLEAR PLANT 4 TH INTERVAL INSERVICE INSPECTION PLAN RELIEF REQUEST NUMBER - RR 4-7 COMPONENT IDENTIFICATION Code Class 2 Code Reference IWC-2500 Table IWC-2500-1 Examination Category C-A, C-B Item Numbers C1.10, C1.30 and C2.21 Component Description Shutdown Cooling Heat Exchanger E-60B Shell to Flange Weld E-60B-01 Tubesheet to Shell Weld E-60B-02 Nozzle to Shell Welds E-60B-03 & 04 Reference Drawings M1-GD Sh. 8 (See Attachment 1)

In accordance with 10 CFR 50.55a(a)(3)(ii), Nuclear Management Company, LLC (NMC) requests approval to implement examination on the accessible volumes as identified below in lieu of the 100% volumetric requirements at Palisades Nuclear Plant.

CODE REQUIREMENT Table IWC-2500-1 requires all Category C-A welds to be volumetrically examined during each inspection interval and Category C-B, Item Number C2.21 welds to be examined by surface and volumetric techniques each inspection interval.

BASIS FOR RELIEF Weld E-60B-01 is a flange to primary shell weld and is a Category C-A, Item Number C 1.10. Volumetric examination of exam volume A-B-C-D as referenced in Figure IWC-2500-1 (a ) is limited due to configuration and scanning limitations created by the flange bolting being in the area of interest and by the flange to weld distance.

Upon review of the referenced drawings and previous examination data, the following examination volumes can be achieved:

1. The 45 0 angle beam examination from the vessel side towards the flange will allow examination of approximately 91% of the required volume 77 of 96

PALISADES NUCLEAR PLANT 4 TH INTERVAL INSERVICE INSPECTION PLAN A-B-C-D. The 45 ° angle beam examination from the flange side towards the vessel will allow examination of approximately 60% of the required volume.

2 The 600 angle beam from the vessel side towards the flange will allow examination of approximately 95% of the required volume A-B-C-D. The 600 angle beam scanning from the flange side towards the vessel will examine approximately 43% of the required volume.

3 The transverse scans with two angle beam directions in both the clockwise and the counter clockwise directions will obtain 94% of the required examination volume A-B-C-D.

4 The 0° scan of the required examination volume will obtain 96% of the required examination volume A-B-C-D.

BASIS FOR RELIEF Weld E-60B-02 is the primary shell to tubesheet weld and is a Category C-A, Item Number C1.30. Volumetric examination of exam volume E-F-G-H as referenced in Figure IWC-2500-2 is limited due to the design configuration.

Upon review of the referenced drawings and previous examination data, the following examination volumes can be achieved:

1 The 60 ° angle beam examination from the shell side towards the tubesheet will allow examination of approximately 96% of the required volume E-F-G-H. The 60" angle beam examination from the tubesheet towards the shell will allow examination of approximately 61% of the required volume.

2 The 45" angle beam from the shell side towards the tubesheet will allow examination of approximately 92% of the required volume E-F-G-H. The 45* angle beam scanning from the tubesheet side towards the shell will examine approximately 32% of the required volume.

3 The transverse scans with two angle beam directions in both the clockwise and the counter clockwise directions will obtain 92% of the required examination volume E-F-G-H.

4. The 0* scan of the required examination volume will obtain 96% of the required examination volume E-F-G-H.

BASIS FOR RELIEF Welds E-60B-03 and E-60B-04 are the primary shell inlet and outlet nozzle to shell welds and are Category C-B, Item Number C2.21. Volumetric examination 78 of 96

PALISADES NUCLEAR PLANT 4 TH INTERVAL INSERVICE INSPECTION PLAN of exam volume C-D-E-F as referenced in Figure IWC-2500-4(b), is limited due to the design configuration. Upon review of the referenced drawings and previous examination data, the following examination volumes can be achieved:

1. The 450 angle beam examination from the shell side towards the nozzle will allow examination of approximately 98% of the required volume C-D-E-F. The 45 0angle beam examination from the nozzle towards the shell can not be performed due to the design configuration.

2. The 60" angle beam from the shell side towards the nozzle will allow examination of 100% of the required volume C-D-E-F. The 60" angle beam examination from the nozzle towards the shell can not be performed due to the design configuration.

3. The transverse scans with two angle beam directions in both the clockwise and the counter clockwise directions will obtain 60% of the required examination volume C-D-E-F.

4. The 00 scan of the required examination volume will obtain 33% of the required examination volume C-D-E-F.

PROPOSED ALTERNATIVE EXAMINATIONS The accessible volumes as identified above will be examined in lieu of the 100%

volumetric requirements.

IMPLEMENTATION SCHEDULE The proposed alternative is requested for the 4 th ten year interval of the Inservice Inspection Program for Palisades Nuclear Plant.

REFERENCE By letter dated January 9, 1997, the NRC Staff previously authorized this relief to Palisades, Docket No 50-255 [TAC NO. M93628] for the 3 rd ten year inspection interval (Previously RR-7).

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PALISADES NUCLEAR PLANT 4 TH INTERVAL INSERVICE INSPECTION PLAN RELIEF REQUEST NUMBER - RR 4-8 In accordance with 10 CFR 50.55a(a)(3)(i), Nuclear Management Company, LLC (NMC) requests approval to implement inside and outside diameter surface examination on the identified welds in lieu of the 100% volumetric requirements at Palisades Nuclear Plant.

COMPONENT IDENTIFICATON Code Class 1 Code Reference Risk Informed Examination Category R-A Item Number R1.11 and R1.15 Component Description Pressurizer Relief Valve Risk Informed Segment:

PZR-006 Weld 1(RV-1 041) - Nozzle/Safe End to Flange PZR-007 Weld 1(RV-1 040) - Nozzle/Safe End to Flange PZR-008 Weld 1(RV-1 039) - Nozzle/Safe End to Flange Reference Drawing M1-LA Sh. 986 (See Attachment 1)

CODE REQUIREMENT In accordance with the Palisades Risk Informed Inservice Inspection Program, Volumetric examinations are required on 100% of the weld length as identified below:

A. Category R-A, Item R1.11 "Elements Subject to Thermal Fatigue" requires 100% volumetric examination.

B. Category R-A, Item R1.15 "Elements Subject to Primary Water Stress Corrosion Cracking (PWSCC)" requires 100% volumetric examination.

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PALISADES NUCLEAR PLANT 4 TH INTERVAL INSERVICE INSPECTION PLAN BASIS FOR RELIEF The relief valve nozzle assembly (reference drawing Mi-LA Sh. 986) is 3.0 inch inside diameter. The proximity of the flange to the weld centerline and the outside diameter contours of the nozzle will result in an extremely limited examination for all axial scanning (ability to detect circumferential flaws) with ultrasonic examination. The 45 degree search unit would examine approximately 80% of the required volume. The required volumes for a 60 and 70 degree refracted longitudinal search unit would not examine the weld root area.

Since we are primarily looking for inside diameter initiated cracking, guidance is taken from NRC Revised Order EA-03-009 which allows for dye penetrant testing of the entire wetted surface of the J-groove weld and a portion of the wetted surface of the RPV head penetration nozzle base material in lieu of performing ultrasonic testing. By examining the inside and outside diameter surfaces using the liquid penetrant method in lieu of the limited ultrasonic method there is greater assurance that indications associated with PWSCC and other cracking mechanisms will be revealed.

PROPOSED ALTERNATIVE EXAMINATION The welds will be examined from the inside diameter with the Liquid Penetrant Technique (the inside diameter of these components was machined during fabrication) in lieu of the volumetric examination. Additionally, a liquid penetrant examination will be performed on the outside of the weld to ensure the quality of the weld. Performing a liquid penetrant examination of the inside and outside diameter of the weld will result in 100 percent coverage of the required area. The probability of detecting a flaw based on an expected flaw initiation at either the inside or outside diameter, this liquid penetrant examination will provide an acceptable level of quality and safety.

IMPLEMENTATION SCHEDULE The proposed alternative is requested for the 4 th ten year interval of the Inservice Inspection Program for Palisades Nuclear Plant.

REFERENCE NRC Order EA-03-009 "Issuance of First Revised NRC Order (EA-03-009)

Establishing Interim Inspection Requirements for Reactor Pressure Vessel Heads at Pressurized Water Reactors" 81 of 96

PALISADES NUCLEAR PLANT 4 TH INTERVAL INSERVICE INSPECTION PLAN RELIEF REQUEST NUMBER - RR 4-9 In accordance with 10 CFR 50.55a(a)(3)(i), Nuclear Management Company, LLC (NMC) requests approval to implement the EPRI / PDI Supplement 11 Program requirements at Palisades Nuclear Plant.

COMPONENT IDENTIFICATION Austenitic piping welds having structural overlays subject to examination using procedures, personnel and equipment qualified to ASME Code Section XI, 2001 Edition, Appendix VIII, Supplement 11 "Qualification Requirements for Full Structural Overlaid Wrought Austenitic Piping Welds" CODE REQUIREMENT The Code requirements for which relief is requested are all contained within Appendix VIII, Supplement 11. This relief is specific to the paragraphs identified below:

- Paragraph 1.1(b) The specimen set shall consist of at least three specimens having different nominal pipe diameters and overlay thicknesses. They shall include the minimum and maximum nominal pipe diameters for which the examination procedure is applicable. Pipe diameters within a range of 0.9 to 1.5 times a nominal diameter shall be considered equivalent. Ifthe procedure is applicable to pipe diameters of 24 in. or larger, the specimen set must include at least one specimen 24 in. or larger but need not include the maximum diameter.

The specimen set must include at least one specimen with overlay thickness within -0.1 in. to +0.25 in. of the maximum nominal overlay thickness for which the procedure is applicable.

• Paragraph 1.1 (d)(1) Base metal flaws. All flaws must be cracks in or near the butt weld heat-affected zone, open to the inside surface, and extending at least 75 percent through the base metal wall. Flaws may extend 100 percent through the base metal and into the overlay material; in this case, intentional overlay fabrication flaws shall not interfere with the ultrasonic detection or characterization of the cracking. Specimens containing intergranular stress corrosion cracking (IGSCC) shall be used when available.

• Paragraph 1.1 (e)(1) At least 20 percent but less than 40 percent of the flaws shall be oriented within .20 deg. of the pipe axial direction. The remainder shall be oriented circumferentially. Flaws shall not be open to any surface to which the candidate has physical or visual access. The rules of IWA-3300 shall be used to determine whether closely spaced flaws should be treated as single or multiple flaws.

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PALISADES NUCLEAR PLANT 4 TH INTERVAL INSERVICE INSPECTION PLAN

• Paragraph 1.1 (e)(2) Specimens shall be divided into base and overlay grading units. Each specimen shall contain one or both types of grading units.

• Paragraph 1.1 (e)(2)(a)(1) A base grading unit shall include at least 3 in. of the length of the overlaid weld. The base grading unit includes the outer 25 percent of the overlaid weld and base metal on both sides. The base grading unit shall not include the inner 75 percent of the overlaid weld and base metal overlay material, or base metal-to-overlay interference.

• Paragraph 1.1 (e)(2)(a)(2) When base metal cracking penetrates into the overlay material, the base grading unit shall include the overlay metal within 1 in. of the crack location. This portion of the overlay material shall not be used as part of any overlay grading unit.

• Paragraph 1.1 (e)(2)(a)(3) When a base grading unit is designed to be unflawed, at least 1 in. of unflawed overlaid weld and base metal shall exist on either side of the base grading unit. The segment of weld length used in one base grading unit shall not be used in another base grading unit. Base grading units need not be uniformly spaced around the specimen.

• Paragraph 1.1 (e)(2)(b)(1) An overlay grading unit shall include the overlay material and the base metal-to-overlay interference of at least 6 sq. in. The overlay grading unit shall be rectangular, with minimum dimensions of 2 in.

• Paragraph 1.1 (e)(2)(b)(2) An overlay grading unit designed to be unflawed shall be surrounded by unflawed overlay material and unflawed base metal-to-overlay interface for at least 1 in. around its entire perimeter. The specific area used in one overlay grading unit shall not be used in another overlay grading unit.

Overlay grading units need not be spaced uniformly about the specimen.

• Paragraph 1.1 (e)(2)(b)(3) Detection sets shall be selected from Table VIII-S2-1.

The minimum detection sample set is five flawed base grading units, ten unflawed base grading units, and ten unflawed overlay grading units. For each type of grading unit, the set shall contain at least twice as many unflawed as flawed grading units.

• Paragraph 1.1 (f)(1) The minimum number of flaws shall be ten. At least 30 percent of the flaws shall be overlay fabrication flaws. At least 40 percent of the flaws shall be cracks open to the inside surface.

• Paragraph 1.1 (f)(3) Base metal cracking used for length sizing demonstrations shall be oriented circumferentially.

• Paragraph 1.1 (f)(4) Depth sizing specimen sets shall include at least two distinct locations where cracking in the base metal extends into the overlay material by at least 0.1 in. in the through-wall direction.

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PALISADES NUCLEAR PLANT 4 TH INTERVAL INSERVICE INSPECTION PLAN Paragraph 2.2(d) For flaws in base grading units, the candidate shall estimate the length of that part of the flaw that is in the outer 25 percent of the base wall thickness.

- Paragraph 2.3 For the depth sizing test, 80 percent of the flaws shall be sized at a specific location on the surface of the specimen identified to the candidate.

For the remaining flaws, the regions of each specimen containing a flaw to be sized shall be identified to the candidate. The candidate shall determine the maximum depth of the flaw in each region.

- Paragraph 3.1 Examination procedures, equipment, and personnel are qualified for detection when the results of the performance demonstration satisfy the acceptance criteria of Table VIII-$2-1 for both detection and false calls. The criteria shall be satisfied separately by the demonstration results for base grading units and for overlay grading units.

• Paragraph 3.2(b) All extensions of base metal cracking into the overlay material by at least 0.1 in. are reported as being intrusions into the overlay material.

Pursuant to 10 CFR 50.55a(a)(3)(i) relief is requested to use the Electric Power Research Institute (EPRI) Performance Demonstration Initiative (PDI) Program for implementation of Appendix VIII, Supplement 11 requirements.

BASIS FOR RELIEF Paragraph 1.1(b) of Supplement 11 states limitations to the maximum thickness for which a procedure may be qualified. The Code states that 'The specimen set must include at least one specimen with overlay thickness within minus 0.10-inch to plus 0.25-inch of the maximum nominal overlay thickness for which the procedure is applicable." The Code requirement addresses the specimen thickness tolerance for a single specimen set, but is confusing when multiple specimen sets are used. The PDI proposed alternative states that "the specimen set shall include specimens with overlay not thicker than 0.10-inch more than the minimum thickness, nor thinner than 0.25-inch of the maximum nominal overlay thickness for which the examination procedure is applicable."

The proposed alternative provides clarification on the application of the tolerance.

The tolerance is unchanged for a single specimen set however, it clarifies the tolerance for multiple specimen sets by providing tolerances for both the minimum and maximum thicknesses. The proposed wording eliminates confusion while maintaining the intent of the overlay thickness tolerance.

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PALISADES NUCLEAR PLANT 4 TH INTERVAL INSERVICE INSPECTION PLAN Paragraph 1.1 (d)(1) requires that all base metal flaws be cracks. PDI determined that certain Supplement 11 requirements pertaining to location and size of cracks would be extremely difficult to achieve. For example, flaw implantation requires excavating a volume of base material to allow a pre-cracked coupon to be welded into this area. This process would add weld material to an area of the specimens that typically consists of only base material, and could potentially make ultrasonic examination more difficult and not representative of actual field conditions. In an effort to satisfy the requirements, PDI developed a process for fabricating flaws that exhibit crack like reflective characteristics. Instead of all flaws being cracks as required by Paragraph .1.1 (d)(1), the PDI weld overlay performance demonstrations contain at least 70 percent cracks with the remainder being fabricated flaws exhibiting cracklike reflective characteristics.

The fabricated flaws are semi-elliptical with tip widths of less than 0.002-inches.

The PDI Program alternative to clarify when real cracks, as opposed to fabricated flaws, will be used; "Flaws shall be limited to the cases where implantation of cracks produces spurious reflectors that are uncharacteristic of actual flaws."

Paragraph 1.1 (e)(1) requires that at least 20 percent but not less than 40 percent of the flaws shall be oriented within .20 degrees of the axial direction of the piping test specimen. Flaws contained in the original base metal heat-affected zone satisfy this requirement however, PDI excludes axial fabrication flaws in the weld overlay material. PDI has concluded that axial flaws in the overlay material are improbable because the overlay filler material is applied in the circumferential direction (parallel to the girth weld), therefore fabrication anomalies would also be expected to have major dimensions in the circumferential direction.

Paragraph 1.1 (e)(1) also requires that the rules of IWA-3300 shall be used to determine whether closely spaced flaws should be treated as single or multiple flaws. PDI treats each flaw as an individual flaw and not as part of a system of closely spaced flaws. PDI controls the flaws going into a test specimen set such that the flaws are free of interfering reflections from adjacent flaws. In some cases, this permits flaws to be spaced closer than what is allowed for classification as a multiple set of flaws by IWA-3300, thus, potentially making the performance demonstration more challenging.

Paragraph 1.1 (e)(2) requires that specimens be divided into base metal and overlay grading units. The PDI program adds clarification with the addition of the word fabrication and ensures flaw identification by ensuring all flaws will not be masked by other flaws with the addition of "Flaws shall not interfere with ultrasonic detection or characterization of other flaws."

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PALISADES NUCLEAR PLANT 4 TH INTERVAL INSERVICE INSPECTION PLAN Paragraph 1.1 (e)(2)(a)(1) requires that a base grading unit shall include at least 3-inches of the length of the overlaid weld, and the base grading unit includes the outer 25 percent of the overlaid weld and base metal on both sides. The PDI program reduced the criteria to 1-inch of the length of the overlaid weld and eliminated from the grading unit the need to include both sides of the weld. The proposed change permits the PDI program to continue using test specimens from the existing weld overlay program which have flaws on both sides of the welds.

These test specimens have been used successfully for testing the proficiency of personnel for over 16 years. The weld overlay qualification is designed to be a near-side (relative to the weld) examination, and it is improbable that a candidate would detect a flaw on the opposite side of the weld due to the sound attenuation and re-direction caused by the weld microstructure. However, the presence of flaws on both sides of the original weld (outside the PDI grading unit) may actually provide a more challenging examination, as candidates must determine the relevancy of these flaws, if detected.

Paragraph 1.1 (e)(2)(a)(2) requires when base metal cracking penetrates into the overlay material a portion of the base grading unit shall not be used as part of the overlay grading unit. The PDI program adjusts for the changes in Paragraph 1.1 (e)(2)(a)(2) and conservatively states that when base metal flaws penetrate into the overlay material no portion of it shall be used as part of the overlay fabrication grading unit. The PDI program also provided clarification by the addition of the term flaws for cracks and the addition of fabrication to overlay grading unit.

Paragraph 1.1 (e)(2)(a)(3) requires that for unflawed base grading units, at least 1-inch of unflawed overlaid weld and base metal shall exist on either side of the base grading unit. This is to minimize the number of false identifications of extraneous reflectors. The PDI program stipulates that unflawed overlaid weld and base metal exists on all sides of the grading unit and flawed grading units must be free of interfering reflections from adjacent flaws which addresses the same concerns as Code.

Paragraph 1.1 (e)(2)(b)(1) requires that an overlay grading unit shall include the overlay material and the base metal-to-overlay interface of at least 6 square inches. The overlay grading unit shall be rectangular, with minimum dimensions of 2-inch. The PDI program reduces the base metal-to-overlay interface to at least 1-inch (in lieu of a minimum of 2-inches) and eliminates the minimum rectangular dimension. This criterion is necessary to allow use of existing examination specimens that were fabricated in order to meet NRC Generic Letter 88-01 (Triparty Agreement, July 1984). This criterion may be more challenging than Code because of the variability associated with the shape of the grading unit.

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PALISADES NUCLEAR PLANT 4 TH INTERVAL INSERVICE INSPECTION PLAN Paragraph 1.1 (e)(2)(b)(2) requires that unflawed overlay grading units should be surrounded by unflawed material for 1-inch around its entire perimeter. The PDI program redefines the area by noting unflawed overlay fabrication grading units shall be separated by at least 1-inch of unflawed material at both ends and sufficient area on both sides to preclude interfering reflections from adjacent flaws. The relaxation in required area on the sides of the specimens, while still ensuring no interfering reflections, may be more challenging than Code because of the possibility for having a parallel flaw on the opposite side of the weld.

Paragraph 1.1 (e)(2)(b)(3) requirements are contained in the PDI program. In addition, the PDI program requires that initial procedure qualification contain three times the number of flaws required for a personal qualification. To qualify new values of essential variables, the equivalent of at least one personal qualification set is required.

Paragraph 1.1(f)(1) requirements are contained in the PDI program, with the clarification change of the term "flaws" for "cracks." In addition, the PDI program includes the requirements that sizing sets shall contain a distribution of flaw dimensions to verify sizing capabilities. The PDI program also requires that initial procedure qualification contain three times the number of flaws required for a personal qualification. To qualify new values of essential variables the equivalent of at least one personal qualification set is required.

Paragraphs 1.1 (f)(3) and 1.1 (f)(4) were clarified by the PD! program by replacing the term "cracking" with "flaws" because of the use of alternative flaw mechanisms.

Paragraph 2.2(d) was clarified by the PDI program by the addition of the terms "metal" and "fabrication." The terms provide acceptable classification of the terms they are enhancing.

Paragraph 2.3 states that, for depth sizing tests, 80 percent of the flaws shall be sized at a specific location on the surface of the specimen to the candidate. This requires detection and sizing tests to be separate. PDI revised the weld overlay program to allow sizing to be conducted either in conjunction with, or separately from, the flaw detection test. If performed in conjunction with detection, and the detected flaws do not meet the Supplement 11 range criteria, additional specimens will be presented to the candidate with the regions containing flaws identified. Each candidate will be required to determine the maximum depth of flaw in each region. For separate sizing tests, the regions of interest will also be identified and the maximum depth and length of each flaw in the region will similarly be determined. In addition, PDI stated that grading units are not applicable to sizing tests, and that each sizing region will be large enough to contain the target flaw, but small enough such that candidates will not attempt to size a different flaw. The above clarification provides a basis for implementing sizing tests in a systematic, consistent manner that meets the intent of Supplement 11.

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PALISADES NUCLEAR PLANT 4 TH INTERVAL INSERVICE INSPECTION PLAN Paragraph 3.1 of Supplement 11 state that procedures, equipment and personnel (as a complete ultrasonic system) are qualified for detection or sizing of flaws, as applicable, when certain criteria are met. The PDI program allows procedure qualification to be performed separately from personnel and equipment qualification. Historical data indicate that, if ultrasonic detection or sizing procedures are thoroughly tested, personnel and equipment using those procedures have a higher probability of successfully passing a qualification test.

In an effort to increase this passing rate, PDI has elected to perform procedure qualifications separately in order to assess and modify essential variables that may affect overall system capabilities. For a procedure to be qualified, the PDI program requires three times as many flaws to be detected (or sized) as shown in Supplement 11 for the entire ultrasonic system. The personnel and equipment are still required to meet Supplement 11 therefore, the PDI program exceeds ASME requirements for personnel, procedures, and equipment qualification.

Paragraph 3.2(b) requires that all extensions of base metal cracking into the overlay material by at least 0.10-inch are reported as being intrusions into the overlay material. The PDI program omits this criterion because of the difficulty in actually fabricating a flaw with a 0.10-inch minimum extension into the overlay, while still knowing the true state of the flaw dimensions.

However, the PDI program requires that cracks be depth-sized to the tolerance specified in Code which is 0.125-inches. Since the Code tolerance is close to the 0.10-inch value of Paragraph 3.2(b), any crack extending beyond 0.10-inch into the overlay material would be identified as such from the characterized dimensions. The reporting of an extension in the overlay material is redundant for performance demonstration testing because of the flaw sizing tolerance.

PROPOSED ALTERNATIVE EXAMINATION Palisades will utilize the PDI program for weld overlay qualifications, in lieu of Supplement 11 to Appendix VIII of the Section XI Code.

IMPLEMENTATION SCHEDULE The proposed alternative is requested for the 4 th ten year interval of the Inservice Inspection Program for Palisades Nuclear Plant.

REFERENCE By letter dated September 14, 2004, the NRC Staff previously authorized this relief to Palisades, Docket No 50-255 [TAC Nos. MC0809] for the 3 rd ten year inspection interval.

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PALISADES NUCLEAR PLANT 4 TH INTERVAL INSERVICE INSPECTION PLAN RELIEF REQUEST NUMBER - RR 4-10 COMPONENT IDENTIFICATION Code Class: 1 and 2 Code

References:

Table IWB-2500-1, Examination Category B-F and B-J Table IWC-2500-1, Examination Category C-F-1 and C-F-2

==

Description:==

ASME Class 1 and 2 Pressure Retaining Welds Systems: Various CODE REQUIREMENT ASME Code Section XI 2001 Edition with addenda through 2003 currently contain the requirements for non-destructive examination of Category B-F, B-J, C-F-1 and C-F-2 piping components.

BASIS FOR RELIEF Nuclear Management Company (NMC) has completed the development of a full scope Risk Informed Inservice Inspection (RI-ISI) Program for Palisades Nuclear Power Plant, using Westinghouse Topical Report, WCAP-14572, Westinghouse Owners Group Application of Risk-Informed Methods to Piping Inservice Inspection Topical Report," Revision I-NP-A, WCAP-14572, Supplement 1,

'Westinghouse Structural Reliability and Risk Assessment (SRRA) Model for Piping Risk-Informed Inservice Inspection, Revision 1-NP-A and WCAP-14572, Supplement 2, Westinghouse Owners Group Application of Risk-Informed Methods to Piping Inservice Inspection topical Report Clarifications."

This program was approved by the NRC on May 19, 2003, "Palisades Plant:

Risk-Informed Inservice Inspection Program" for the remainder of the third inspection interval. This approval included four deviations from the WCAP Methodology. One deviation is to perform visual VT-2 examinations as an alternative to volumetric or surface exams for those high safety significant ASME Code Class 1 and 2 socket welds of two-inch diameter of less identified in the RI-ISI Program. The second deviation involves crediting leak detection for some pipe segments that are not reactor coolant system (RCS) piping segments. The third deviation involves determining the number of inspections for some piping segments based on the ASME percentage criteria instead of the structural sampling methodology. The fourth deviation involved the method of calculating the failure frequency of containing piping with multiple sizes.

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PALISADES NUCLEAR PLANT 4 TH INTERVAL INSERVICE INSPECTION PLAN PROPOSED ALTERNATE EXAMINATION The previously approved RI-ISI Program will be substituted for Class 1 and 2 piping (Examination Categories B-F, B-J, C-F-1 and C-F-2) in accordance with 10CFR50.55a(a)(3)(i) by alternatively providing an acceptable level of quality and safety, other non-related portions of ASME Section Xl Code will be unaffected.

With the NRC approval of WCAP-14572, Supplement 2, Westinghouse Owners Group Application of Risk-Informed Methods to Piping Inservice Inspection topical Report Clarifications" the deviation for calculating the failure frequency of piping containing multiple sizes is no longer required as Palisades original treatment of piping with multiple sizes is in conformance with Supplement 2 of the WCAP.

This relief request is to align the RI-ISI Interval and the Code Year for the 4 th Interval ISI Program. 100% of the RI-ISI program weld examinations will be completed in the 4 th Inspection Interval.

IMPLEMENTATION SCHEDULE The proposed alternative is requested for the 4 th ten year interval of the Inservice Inspection Program for Palisades Unit 1 REFERENCE By letter dated May 19,2003, the NRC Staff previously authorized this relief to Palisades, Docket No 50-255 [TAC NO. MB4420] for the 3 rd ten year inspection interval.

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PALISADES NUCLEAR PLANT 4 TH INTERVAL INSERVICE INSPECTION PLAN RELIEF REQUEST NUMBER - RR 4-11 COMPONENT IDENTIFICATION ISI examination of piping, vessel and component welds In accordance with 10 CFR 50.55a(a)(3)(i), Nuclear Management Company, LLC (NMC) requests authorization from the Weld Reference System marking requirements at Palisades Nuclear Plant.

CODE REQUIREMENT Section XI of the ASME Boiler and Pressure Vessel Code, 2001 Edition with Addenda through 2003, IWA-2600 "Weld Reference System" BASIS FOR RELIEF The original construction code used at Palisades did not require that a weld reference system be established. Establishment of a weld reference system cannot be practically attained within the scope and schedule of existing outages.

Significant effort would be expended to achieve compliance with the requirements of IWA-2600. Based on this, use of the alternative reference system identified below provides an acceptable level of quality and safety.

PROPOSED ALTERNATIVE Palisades uses isometric drawings to provide a detailed identification and location of bach weld requiring examination. In addition, the following will be performed:

Surface Examinations - Where surface examination is specified,Section XI requires that 100% of the selected weld or area be examined. Unlike the performance of volumetric examination, there is no need to indicate the direction of examination to assure uniformity in reporting results. In these cases, no marks will be placed on the weld or area. The location of any accepted surface indications will be documented by the use of a map or photograph that permits accurate identification of areas on the examination surface.

Volumetric Examinations (Manual) - If a weld is accepted for continued service that contain volumetric indications accepted under the criteria IWX-3500 or IWX-3600 shall be identified to ensure the relocation of the indication, using appropriate reference marks. These reference marks may be permanently fixed on the welds or by documentation on a map or photograph of the weld or surface that permits accurate identification of areas on the examination surface (e.g.,

reference points, orientation and/or proximity to other welds) to positively identify 91 of 96

PALISADES NUCLEAR PLANT 4 TH INTERVAL INSERVICE INSPECTION PLAN the weld or area in question and the examination starting point. The starting point is determined from the instructions provided for determining the location of the zero reference point.

Volumetric Examination (Automated Vessel) - The automated vessel tool establishes its reference point using an existing zero reference point on the reactor vessel. This point allows the device to repeat examination locations without the necessity of any other reference system. The tool determines its location by the use of an electronic encoder system which provides for sufficient repeatability.

IMPLEMENTATION SCHEDULE The proposed alternative is requested for the 4 th ten year interval of the Inservice Inspection Program for Palisades Nuclear Plant.

REFERENCE None 92 of 96

PALISADES NUCLEAR PLANT 4 TH INTERVAL INSERVICE INSPECTION PLAN RELIEF REQUEST NUMBER - RR 4-12 COMPONENT IDENTIFICATION ASME Section XI, Moderate Energy Class 2 and Class 3 Piping CODE REQUIREMENT ASME Section XI 2001 Edition with Addenda through 2003.

Flaws that exceed the acceptance criteria of the above code are required to be accepted by either a repair/replacement activity or an analytical evaluation.

The applicable code requirements are as follows:

CLASS 3 IWD-3000 states, "This Article is in course of preparation. The rules of IWB-3000 may be used."

IWB-3132 provides three ways in which an Inservice Volumetric or Surface Examination may be accepted.

1. IWB-3132.1, "Acceptance by Volumetric or Surface Examination",

2. IWB-3132.2, "Acceptance by Repair/Replacement Activity", or

3. IWB-3132.3, "Acceptance by Analytical Evaluation".

IWB-3132.2 states, "A component whose volumetric or surface examination detects flaws that exceed the acceptance standards of Table IWB-341 0-1 is unacceptable for continued service until the additional examination requirements of IWB-2430 are satisfied and the component is corrected by a repair/replacement activity to the extent necessary to meet the acceptance standards of IWB-3000."

IWB-3142 provides four ways in which an inservice visual examination may be accepted.

1. IWB-3142.1 "Acceptance by Visual Examination"

2. IWB-3142.2 "Acceptance by Supplemental Examination"

3. IWB-3142.3 "Acceptance by Corrective Measures or Repair/Replacement Activity"

4. IWB-3142.4 "Acceptance by Analytical Evaluation" IWB-3142.3 states, "A component containing relevant conditions is acceptable for continued service if the relevant conditions are corrected by a repair/replacement activity or by corrective measure to the extent necessary to meet the acceptance standards of Table IWB-3410-1 ."

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PALISADES NUCLEAR PLANT 4 TH INTERVAL INSERVICE INSPECTION PLAN CLASS 2 IWC-3122 provides three ways in which an Inservice Volumetric and Surface Examinations may be accepted.

1. IWC-3122.1, "Acceptance by Examination"

2. IWC-3122.2, "Acceptance by Repair/Replacement Activity"

3. IWC-3122.3, "Acceptance by Analytical Evaluation" IWC-3122.2 states, "A component whose examination detects flaws that exceed the acceptance standards of Table IWC-3410-1 is unacceptable for continued service until the additional examination requirements of IWC-2430 are satisfied and the component is corrected by a repair/replacement activity to the extent necessary to meet the acceptance standards of IWC-3000."

IWC-3132 provides four ways in which an inservice visual examination may be accepted.

1. IWC-3132, "Acceptance"

2. IWC-3132.1, "Acceptance by Supplemental Examination"

3. IWC-3132.2, "Acceptance by Corrective Measures or Repair/Replacement Activity"

4. IWC-3132.3, "Acceptance by Analytical Evaluation" IWC-3132.2 states, "A component containing relevant conditions is acceptable for continued service if the relevant conditions are corrected by a repair/replacement activity or by corrective measures to the extent necessary to meet the acceptance standards of Table IWC-341 0-1 ."

BASIS FOR RELIEF Relief is requested from replacement or internal weld repair of wall thinning conditions resulting from various wall thinning degradation mechanisms such as erosion, corrosion, cavitation, and pitting in moderate energy Class 2 and 3 piping systems in accordance with the design specification and the original construction code. The use of Code Case N-513-2 will provide an acceptable method to evaluate flaws on a temporary basis until the next scheduled outage.

PROPOSED ALTERNATIVE The Nuclear Regulatory Commission in Regulatory Guide 1.147, "Inservice Inspection Code Case Acceptability," Revision 14, has accepted Code Case N-513-1 with the following limitations:

94 of 96

PALISADES NUCLEAR PLANT 4 TH INTERVAL INSERVICE INSPECTION PLAN 1- Specific safety factors in paragraph 4.0 must be satisfied.

2- Code Case N-513 may not be applied to:

i. Components other than pipe and tube.

ii. Leakage through a gasket iii. Threaded connections employing nonstructural seal welds for leakage prevention (through seal weld leakage is not a structural flaw; thread integrity must be maintained).

iv. Degraded socket welds Code Case N-513-1 permits flaws in Class 2 and 3 moderate energy piping on a temporary basis until the next outage if it can be demonstrated that adequate pipe integrity and leakage containment are maintained. The Code Case is currently applicable to part-through and through wall planar flaws and part-through nonplanar flaws. Service experience has shown that some piping can suffer degradation from nonplanar flaws, such as pitting and microbiological attack, where local inconsequential leakage can occur.

The Code Case can be used for nonplanar through-wall flaws but in a restrictive situation where nonplanar geometry is dominant in one plane.

Some plants have used the intent of N-513 for nonplanar leaking flaws; however, relief requests from code requirements are still required because of the stated limited scope of N-513 in section 3.0 of the Code Case. The Code Case was revised (N-513-2) to extend the application to cover all types of nonplanar flaws. The analysis procedures were expanded to address the general case of through-wall degradation. Code Case N-513-2 has broader applications and therefore has a real direct benefit for operating plants.

Code Case N-513-2 includes the incorporation of the improved flaw evaluation procedures for piping that are provided in the new Appendix C of Section XI in the 2002 Addenda.

Code Case N-513-2 addresses the limitations posed in Regulatory Guide 1.147 as follows:

1. Paragraph 4.0 was revised to incorporate references to Appendix C for acceptance and eliminated the provision that lower safety factors may be used.

2. 1.0(a) was revised to limit the application of the code case as specified in the limitation applied in Regulatory Guide 1.147.

NMC considers the proposed alternative of using Code Case N-513-2 to provide an acceptable level of quality and safety in accordance with 10 CFR 50.55a(3)(i).

95 of 96

PALISADES NUCLEAR PLANT 4 TH INTERVAL INSERVICE INSPECTION PLAN IMPLEMENTATION SCHEDULE The proposed alternative is requested for the 4 th ten year interval of the Inservice Inspection Program for Palisades Nuclear Plant.

REFERENCE By letter dated December 12, 2005 NMC submitted this relief to the NRC as a fleet relief request (L-HU-05-24) for the 3 rd ten year inspection interval.

96 of 96

PALISADES NUCLEAR PLANT 4 TH INTERVAL INSERVICE INSPECTION PLAN ATTACHMENT 1 RELIEF REQUEST DRAWINGS

REFERENCE DRAWINGS RELIEF REQUEST RR 4-1

REFERENCE DRAWINGS RELIEF REQUEST RR 4-2

REFERENCE DRAWINGS RELIEF REQUEST RR 4-3

REFERENCE DRAWINGS RELIEF REQUEST RR 4-4

REFERENCE DRAWINGS RELIEF REQUEST RR 4-6

REFERENCE DRAWINGS RELIEF REQUEST RR 4-7

REFERENCE DRAWINGS RELIEF REQUEST RR 4-8

PALISADES NUCLEAR PLANT 4 TH INTERVAL INSERVICE INSPECTION PLAN ATTACHMENT 2 EPRI REPORT Palisades Steam Generator Inlet and Outlet Nozzle Coverage Calculations

EPRI NDE CENTER Electric Power Research Institute Nondestructive Evaluation Center . Leadership in Technology Transfer September 16, 1996 Tom Fouty Consumers Power Palisades Nuclear Plant 27780 Blue Star Memorial Highway Covert, MI 49043

SUBJECT:

Summary Report on Coverage Calculations

Dear Tom,

The attached summary report describes the coverage calculations performed by the EPRI NDE Center on the Palisades steam generator inlet and outlet nozzles. The type of coverage documented for these nozzles was restricted to calculating where the ultrasound beam interrogated the examination volume.

If you have any comments please call me at (704) 547-6130.

Sincerely, Douglas E. MacDonald Principal Engineer EPRI NDE Center DM/ks Enc.

cc: Kim Kietzman Judy Ford/CP Larry Becker Frank Ammirato 1300 Harris Boulevard

• Charlotte. North Carolina 28262 Telephone: (704) 547-6100 (P.O. Box 217097. Charlotte. North Carolina 28221)

• FAX: (704) 547-6168

Palisades Steam Generator Inlet and Outlet Nozzle Coverage Calculations Douglas E. MacDonald September 5, 1996 Introduction This report summarizes EPRI NDE Center activities to calculate the coverage obtained on the Palisades steam generator inlet and outlet nozzles using the Consumers Power Company procedure No. NDT-UT-12, Revision 4, Issued 11/21/95. The type of coverage documented for these nozzles was restricted to calculating where the ultrasound beam interrogated the ASME code examination volume.

Figure 1 defines the ASME code examination volumes, both nozzle inner radius (M-N-O-P) and nozzle-to-head weld (A-B-C-D-E-F-G-H-I) for the Palisades steam generator inlet and outlet nozzles.

PalisadesSteam GeneratorInlet Nozzle Nozzle Inside Corner Region Table 1 shows the coverage calculated for a 350 probe, skewed 900 (350/900) and scanned on the outer blend radius. (The probe skew convention adopted here has 0* looking at the pipe, 90* looking circumferentially around the nozzle, and 1800 looking at the vessel).

Figure 2 shows a cross section of the Palisades steam generator inlet nozzle upon which has been plotted the rays from the 350/900 probe to the inside surface examination volume. As can be seen in Figure 2, the entire exam volume is covered by the 350/90* probe scanned on the outer blend radius.

tI rtaShell for hood] thiclwu fi nozzle inside cornet radius

-~1.

A 0

T--N

• 9 to Z,

I a n I Vn2 L

I Exmw. Va.

NI Cornew flew -

EXAMMUATION REGION pNow fill EXAMINTION VOLUME (me" M21 CO Shell ta ad s Wng epo 111144 Noub cylinder regiaws. c A-114ý

'do co~rnetram (11 Ewwi~agon r~gion. arn idumfldq taroU* puaa.f diffwr~r~g Ue acmaa wuida,* In NW-2.5 12.

(22 ba ao.a *mm may e wdfit4 9thw by dr'a measuromen On the cUrOnfre Of by meerowenu be&" on d*sn drrun.

NOZZLE IN SHELL OR HEAD

.1 Zones in Barrel Type Nozzles Joined by Full Pefetration Corner Welds)

) Figure 1

Table 1. Coverage Table: SG Inlet; 35°/90' (blend)

Probe Angle 350 Probe Skew 900 Probe Location Blend*

Percentage of Examination Volume (M-N-O-P) Covered 100%

• Contoured Wedge -

Palisodes SG Inlet Nozzle: 35)90 (blend) Coverage 82 -

81 +

80 +

79 -

78+

77+

76 +

75 +

N 74 +

73 4-72+

471+

.70-aCos I Inner Radus Exam Volume M-N-O-P 69 68 II 2

i I I I I I 2

I 2

I I 1

z6 2i i I 6 a I D0 21 22 23 24 25 26 27 28 29 30 31 32 R Oinches)

Figure 2. Palisades SG Inlet Nozzle: 35°/90* (blend) Coverage.

Figure 2 shows a cross section of the Palisades steam generator inlet nozzle upon which has been plotted the rays from the 350/900 probe to the inside surface examination volume. As can be seen in Figure 2, the entire exam volume is covered by the 350/90* probe scanned on the outer blend radius.

3 5j90 (blend) Coverage Palisades S G Inlet Nozzle:

82 81 80 -

79 4 78 -

77-

-. 76 -

75 -

N 74*

L_

73 -r 72 -

71 -

70.-

Class I Iinnr Radlu$ Exam Volume M-N-O-P I g I 2 I I 1 2 2 24 25 26 27 28 29 30 31 32 20 21 22 23 R (Inches)

Figure 2. Palisades SG Inlet Nozzle: 35°/900 (blend) Coverage.

Nozzle-to-Head Weld Region Figure 3 shows a cross section of the Palisades steam generator inlet nozzle which indicates the nozzle-to-head weld examination volume. Table 2 lists the coverage obtained from the axial scans of 450 and 60* probes scanned on the vessel head and nozzle boss. The percent coverage is listed for the the nozzle cylinder region A-B-G-H-I; as well as, the entire weld examination volume A-B-C-D-E-F-G-H-I (See Figures 1 and 3).

Table 2. Coverage Table: SG Inlet; 45°and 60' probes scanned axially on vessel head and nozzle boss.

Probe Angle 450 450 600 600 Probe Skew- 00 1800 00 1800 Probe Location Head Boss Head Boss

% Exam Vol. (C-D-E-F) Covered 100% 0% 93% 0%

% Exam Vol. (B-C-F-G) Covered 87% 0% 90% 0%

% Exam Vol. (A-B-G-H-I) Covered 53% 0% 63% 12%

Percent Total Weld Exam Volume (A-B-C-D-E-F-G-H-I) Covered 79% 0% 81% 5%

Figures 4 through 7 support the calculated axial scan coverage listed in Table 2. Figure 4 shows a cross section of the Palisades steam generator inlet nozzle upon which has been plotted the rays of the axial scan of the 45 0/0 0 (head scan) probe through the weld examination volume. Figure 5 shows the cross section with the rays of the axial scan of the 45*/180 0 (boss scan) probe. Figure 6 shows the cross section with the rays of the axial scan of the 600/00 probe through the weld examination volume. Figure 7 shows the cross section with the rays of the axial scan of the 60"/1800 probe.

Figure 8 shows the combined coverage achieved by all the axial scans of the nozzle-to-head weld.

Table 3 lists the coverage obtained from the transverse scans (no probe skewing) of 45° and 600 probes scanned on the vessel head and nozzle boss.

Figures 9 through 12 support the calculated transverse scan coverage listed in Table 3.

Examination Volume (CP-SGPIN) 85 - -

84 83 82 -

81 --

A 80 79 --

78 -- B 77 -c 76 -

75-. D 73 -m 3 72-HN-711 70 -- H F.

69 --

68 -

67 -

66 --

65 64 -

63 -

62 -

61 60 I. I .. --I- I- I-I-'

'0 21 22 23 24 25 26 27 28 29 30 31 32 33 .5 36 37 38 39 40 41 42 43 44 45 46 47 48

Coverage For Probe Angle=45 Deg; Probe Skew=O Deg (CP-SGPIN) 85 --

84 -

83 -

82 -

81 -

80-- -.

79 - - 5 78 --

77 -

76 ---

75 -

c 74 -.

Z 73 -- , .74 72--

71 --

70 - -

69 -

68 --

67 -

66 -

65 --

64 --

63 -.

62 61 --

60 2 - I - 2 2 20 21 22 23 24 25 262728 29 3031 32 33 "35 36 3738 39 4041 42 43 44 45 46 47 48

Coverage For Probe Angle=-45 Deg; Probe Skew=180 Deg (CP-SGPIN) 85 -

84 83 - -

82 -

81 -

80-79 78 77 --

76 ...

75 --

74 73 -

72 --

71 -

70 - -

69 -

68 --

67 --

66 --

65 -

64 -

63 --

62 --

61 -'-

60- 2I 2--I--t---I--I I I31 32I 15 337839I -- I 411-43I 44 45 20 21 22 23 2425 2627 28 293031 32 33* ""35 36 37 38 394041 4243 4445 46 4748

N N

0 N 0

-44 ca -

40 Figure 6

Coverage For Probe Angle=60 Dog; Probe Skew=1 80 Dog (CP-SGPIN)

• 85 -

84 -

83 82 -

81 80 -

79 78-77

• I 76 -

• I I 75 -

• U I 74 73

• U U to z

• I I -P4 72 -. 924 71 70 -

69 --

68 -

67 -

66 -

65 -

64 -

63 62 61 60 20-I---I--I 21 26I-2--23--I -- 3 40 41 4 43 44 45 46, 47 4"+,

20 21 22 23 2425 2627 28 293031 32 33 '""353637"38 39 4041 42 434445 4647 48

Comblned Coverap (CP-SGPIN) 85 -

84 -

83 82 al 80 --

79-.

78-77 ---

76 --

75 -

74 - -

73 Z -

72 -* co 71 -

60

.14 70 -

69 68 67 --

66 65 -

62 61 60.

32 3 35 3r,,- ---

20 21 22 24252627 29 30 31 42 43 44 45 46 47 48

, 40

. J9 41

Table 3. Coverage Table: SG Inlet; 45°and 600 probes scanned transversely on vessel head and nozzle boss (no probe skewing).

Probe Angle 450 450 600 600 Probe Skew 900 900 900 900 Probe Location Head Boss Head Boss

% Exam Vol. (C-D-E-F) Covered 100% 0% 92% 0%

% Exam Vol. (B-C-F-G) Covered 58% 0% 0% 0%

% Exam Vol. (A-B-G-H-I) Covered 0% 0% 0% 0%

Percent Total Weld Exam Volume (A-B-C-D-E-F-G-H-I) Covered 54% 0% 42% 0%

Figure 9 shows a cross section of the Palisades steam generator inlet nozzle upon which has been plotted the rays of the transverse scan of the 450 probe from the head through the weld examination volume with no probe skewing.

Figure 10 shows the cross section with the rays of the transverse scan of the 450 probe from the boss. Figure 11 shows the cross section with the rays of the transverse scan of the 600 probe from the head through the weld examination volume. Figure 12 shows the cross section with the rays of the transverse scan of the 600 probe from the boss. Figure 13 shows the combined coverage achieved by all the transverse scans of the nozzle-to-head weld with no probe skewing.

Table 4 lists the coverage obtained from the transverse scans (+/-200 probe skewing) of 450 and 60* probes scanned on the vessel head and nozzle boss.

Figures 14 through 17 support the calculated transverse scan coverage listed in Table 4. Figure 14 shows a cross section of the Palisades steam generator inlet nozzle upon which has been plotted the rays of the transverse scan of the 450 probe from the head through the weld examination volume with

+/-20' probe skewing. Figure 15 shows the cross section with the rays of the. transverse scan of the 450 probe from the boss. Figure 16 shows the cross section with the rays of the transverse scan of the 600 probe from the head through the weld examination volume. Figure 17 shows the cross section with the rays of the transverse scan of the 600 probe from the boss.

Figure 18 shows the combined coverage achieved by all the transverse scans of the nozzle-to-head weld with +/-20' probe skewing.

Coverage For Probe Angle=45 Deg; Probe Skew=90 Deg (CP-SGPIN) 85 -

84 -

83 82 -

81 -

D 80 - p-79 -

78 -

77 -

76 --

75 -

74 73 Z7 2 71 70 -

69 _

60 _

67 _-

66 --

65 --

64 -

63 -

62 -

61 --

60 ... I. I I---I--I---I-t-I 20 21 22 23 24 25 26 27 28 29 30 31 32 33 - '5 36 37 38 39 40 41 42 43 44 45 46 47 48

Coverage For Probe Angle=45 Deg; Probe Skew=90 Deg (CP-SGPIN) 85 -

84 -

83 -

82 -

81 -

D 80 5

79 N

78 U

77 -

I 76 -

I 75 -

74

• I U 73 I

• U I

z 72 -I

• U I I 71 - 4%~ I U 70 -

• U I 69 - I 68 -

67 -

66 -

65 64 -

63 -

62 -

61 60 20.. I..-I- --I.2-I--I-I 6-F27 28 29 I 15 36 37 3I----I---1 4-I-+/-3-4--i- [-I--

20 21 22 23 24 25 26 27 28 29 30 31 32 33 - '15 36 37 38 39 40 41 42 43 44 45 46 47 48

Coverage For Probe Angle=60 Dog; Probe Skew=90 Deg (CP-SGPIN) 85 --

84 -

83 82 81 -

80 -o 79-78 -_

iU 77 -

76 --

75-- -

74- -4 73.-

72 .

71 -

70 --

69 -

68 --

67 --

66 --

65 --

64 --

63 - -

62 61 ---

60------I---

20 21 22 23 24 25 26 27 28 29 30 31 32 33 " 35 36 37 '38 39 40 41 42 43 44 45 46 47 48

Coverage For Probe Angle=60 Dog; Probe Skew=90 Deg (CP-SGPIN) 85 -

84 -

83 -

82 -

81 80 F

79 U

78 U

77 -

• U 76 -

I g U 75 -

U

• U 74 -

73 -

• U U It z I

• U 72 -

71 - 4~%~%%%~ I. U 70 U U 69 -

68 -

67 -

66 -

65 -

64 -

63 -

62 61 60 2 - ..24 -2I -234-

-23 1..I], I 1

--- I --4 1 -'I -11 [-

I "F -I l l 20 21 22 2324 2526 2728 29 3031 32 33 '336 37 3839 40 41 42 4344 4546 4748

Combined Coverage (CP-SGPIN) 85 - -

84 --

83 -.

82 --

81 80 --

79 - D 78 --

77 --

76 --

75 -

74/

73 --

72 _/

71 - /

70 --

69 '

68-"

67 --

66 --

65 --

64 - -

63 -

62 --

61 -

60- -- I ...I--.I---I--I----t-I--I--A +F A A F ++/-,+,+/-, ,

20 21 22 23 24 25 26 27 28 29 30 31 32 33 '5 36 37 38 39 40 41 42 43 44 45 46 47 48

Table 4. Coverage Table: SG Inlet; 45'and 60' probes scanned transversely on vessel head and nozzle boss (+/-200 probe skewing).

Probe Angle 450 450 600 600 Probe Skew 700 1100 700 1100 Probe Location Head Boss Head Boss

% Exam Vol. (C-D-E-F) Covered 100% 0% 100% 0%

% Exam Vol. (B-C-F-G) Covered 78% 0% 76% 0%

% Exam Vol. (A-B-G-H-I) Covered 19% 0% 16% 0%

Percent Total Weld Exam Volume (A-B-C-D-E-F-G-H-I) Covered 64% 0% 63% 0%

PalisadesSteam GeneratorOutlet Nozzle Nozzle Inside Corner Region Table 5 shows the coverage calculated for a 350 probe, skewed 900 (35/90)

(looking circumnferentially) and scanned on the outer blend radius.

Table 5. Coverage Table: SG Outlet; 350/90° (blend)

Probe Angle 350 Probe Skew 900 Probe Location Blend*

Percentage of Examination Volume (M-N-O-P) Covered 100%

• Contoured Wedge Figure 19 shows a cross section of the Palisades steam generator outlet nozzle upon which has been plotted the rays from the 35°/90* probe to the inside surface examination volume. As can be seen in Figure 19, the entire exam volume is covered by the 350/900 probe scanned on the outer blend radius.

Coverage For Probe Angle=45 Deg; Probe Skew=70 Deg (CP-SGPIN) 85 -"

84 I 83 -

82 81 -

pl 80 -

79 - F a

78 -

U 77 -

U 76 -

I 75 -

I 74 -

I 73 z

72 -

71 -

70 -

69 -

68 -

67 -

66 -

65 -

64 -

63 62 61 60 20 2122 2324 25 2627 28 293031 32 33n35 363738 3940'41 4243 4445 4647 48

Coverage For Probe Angle=45 Deg; Probe Skew=110 Deg (CP-SGPIN) 85 84 83 82 81 80 5

79 78 a

U 77

• I 76

• I I 75 74

• U U 73

• U I F

z 72

• I U 71 ~ I I 70 I U 69 68 67 66 65 64 63 62 61 60 20I2 22--23 4 5 6 2---2I--31-2-t---5 6I783 4I4I---- I 20 21 22 23 24 25 26 27 28 29 30 31 32 33 ""35 36 37"38 39 40 41 42 43 44 45 46 47 48

Coverage For Probe Angle=60 Deg; Probe Skew=70 Deg (CP-SGPIN) 85 84 83 82 81 9

80 'V F

79 U

78 U

77 3

76 I

75 I

74

'U 73 z

72 71 70 69 68 67 66 65 64 63 62 61 60 20 21 22 23 24 25 26 27 28 29 30 31 32 33 35 36 37 38 39 40 41 42 43 44 45 46 47 48

\~ I

Coverage For Probe Angle=60 Deg; Probe Skew=110 Deg (CP-SGPIN) 85 - -

84 83 -

82- -

81 --

80--

79.-

78 --

77 - -

76 ---

75--

74-Z 73 l- -

72-- *.,

71 70 -

69 - -

68 --

67 -

66 65 -

64 --

63 --

62 61 -I-20 21 22 23 24 25 26 27 28 29 30 31 32 33 35 36 37 38 39 40 41 42 43 44 45 46 47 48 L..,

Examination Volume (CP-SGPON) 85 -

84 -

83 -

82 -

81 -

80 -

79 -

78 -

77 -

76 -

75 -

74 -

73 -

72 -

71 -

70 -

69 -

68 -

67 -

66 -

I I I I I I I I- _____________________________________________

OD 14 16 17 I 18 I 19 I 20 I 21-----

i I I I 2 I I 3I I 14 15 16 17 18 19 20 21 22 *...24 25 26 27 28 29 30 31 32

Table 6 lists the coverage obtained from the axial scans of 450 and 600 probes scanned on the vessel head and nozzle boss. The percent coverage is listed for the head adjoining region C-D-E-F, the attachment weld region B-C-F-G, and the nozzle cylinder region A-B-G-H-I; as well as, the entire weld examination volume A-B-C-D-E-F-G-H-I (See Figures 1 and 20).

Table 6. Coverage Table: SG Outlet; 450 and 600 probes scanned axially on vessel head and nozzle boss.

Probe Angle 450 450 600 600 Probe Skew 00 1800 00 1800 Probe Location Head Boss Head Boss

% Exam Vol. (C-D-E-F) Covered 100% 0% 100% 0%

% Exam Vol. (B-C-F-G) Covered 87% 0% 91% 0%

% Exam Vol. (A-B-G-H-I) Covered 48% 0% 60% 10%

Percent Total Weld Exam Volume (A-B-C-D-E-F-G-H-1) Covered 74% 0% 80% 5%

Figures 21 through 24 support the calculated axial scan coverage listed in Table 6. Figure 21 shows a cross section of the Palisades steam generator outlet nozzle upon which has been plotted the rays of the axial scan of the 450/00(head scan) probe through the weld examination volume. Figure 22 shows the cross section with the rays of the axial scan of the 450/180*(boss scan) probe. Figure 23 shows the cross section with the rays of the axial scan of the 600/0* probe through the weld examination volume. Figure 24 shows the cross section with the rays of the axial scan of the 6 0 "/18 0 0probe.

Figure 25 shows the combined coverage achieved by all the axial scans of the nozzle-to-head weld.

Table 7 lists ,th coverage obtained from the transverse scans (no probe skewing) of45 and 600 probes scanned on the vessel head and nozzle boss.

Figures 26 through 29 support the calculated transverse scan coverage listed in Table 7. Figure 26 shows a cross section of the Palisades steam generator outlet nozzle upon which has been plotted the rays of the transverse scan of the 450 probe from the head through the weld examination volume with no probe skewing. Figure 27 shows the cross section with the rays of the transverse scan of the 450 probe from the boss. Figure 28 shows the cross

Coverage For Probe Angle=45 Deg; Probe Skew=O Dog (CP-SGPON) 85 K 84 83 0

82 I I

81 I 80 U 79 I 78 77

-I 76 75 74 73 Z

72 71 70 69 68 67 66 65 64 63 62 61 I

• l I I I a I I I I I I I I I I I I I I I I I I 60 i-- I I-I-IF I I I I I I I I I I I I I I I I I I I I I I I I I I I '

14 15 16 17 18 19 20 21 22 23 24 25 26 27 28 29 ." 4 32 33 34 35 36 37 38 39 40 41 42 43 44 45 46 47 48

Coverage For Probe Angle=45 Deg; Probe Skew=180 Deg (CP-SGPON) 85 84 83 82 -

81 80 79 --

78 -

77 -

76 -

75 -,

74 --

73-72 71 -

70 --

69 --

68 --

67 66 -

65 --

64 -

63-62 --

61 --

60-----I--I I I I--+--I I I I I I I I I I I I I I I I I I I I-I I I I I 14 15 16 17 18 19 20 21 22 23 24 25 26 27 28 29 ?" '1 32 33 34 35 36 37 38 39 40 41 42 43 44 45 46 47 48

Coverage For Probe Angle=60 Dog; Probe Skew=O Deg (CP-SGPON) 85 84 83 0 82 I I

81 II 80 79 78 77 76 75 74 73 z

72 71 70 69 68 67 66 65 64 63 62 61 I II I I I I I I I I I I 1 1I 1 !I j !  ! !_-, i 1 I 1 1 1 60 1 I

I 6I I9I I I I 23 I I I I I I

' I I I 34 I3 3 3 40 I I I I I4 I I 14 15 16 17 18 19 20 21 22 23 24 25 26 27 28 29." -1 32 33 34 35 36 37 38 39 40 41 42 43 44 45 46 47 4('

Coverage For Probe Angle=60 Deg; Probe Skew=180 Deg (CP-SGPON),

85 -

84 -

83 -

82 81 80 79

• I 78

• I

• 77 I U

• 76 -

• I I 75 - I

• I

• 74 I 73 - I

• I z I I

72 -

71 I 70 -

69 -

68 -

67 -

66 -

65 -

64 -

63 -

62 -

61 -

S I I I I I I I I I I I I I I I I I I I ~ I I I I I I 60 14 15 16 17 18 19 20 21 22 23 24 25 26 27 28 29 P- "1 32 33 34 35 36 37 38 39 40 41 42 43 44 45 46 47 4P

Combined Coverage (CP-SGPON) 84 _-

83-82 80 -

79-78 -

77 76 75 Ij.

74 -

c'J 73 - -A z72 -

71 70 --

69 --

68 -

67 -

66 65 64 -

63 -

62 -

61 -

--14 6---H--I I9 12 l2i 39 4il0 41 42 43,44454 14 15 16 17 18 19 20 21 22 23 24 25 26 27 28 29 '" " 32 33 34 35 36 37 38 39 40 41 42 43 44 45 46 47 48

Table 7. Coverage Table: SG Outlet; 45°and 60* probes scanned transversely on vessel head and nozzle boss (no probe skewing).

Probe Angle 450 450 600 600 Probe Skew 900 90* 900 900 Probe Location Head Boss Head Boss

% Exam Vol. (C-D-E-F) Covered 100% 0% 60% 0%

% Exam Vol. (B-C-F-G) Covered 14% 0% 0% 0%

% Exam Vol. (A-B-G-H-I) Covered 0% 0% 0% 0%

Percent Total Weld Exam Volume (A-B-C-D-E-F-G-H-I) Covered 39% 0% 22% 0%

section with the rays of the transverse scan of the 60* probe from the head through the weld examination volume. Figure 29 shows the cross section with the rays of the transverse scan of the 600 probe from the boss. Figure 30 shows the combined coverage achieved by all the transverse scans of the nozzle-to-head weld with no probe skewing.

Table 8 lists the coverage obtained from the transverse scans (+/-200 probe skewing) of 45* and 60* probes scanned on the vessel head and nozzle boss.

Figures 31 through 34 support the calculated transverse scan coverage listed in Table 8. Figure 31 shows a cross section of the Palisades steam generator outlet nozzle upon which has been plotted the rays of the transverse scan of the 450 probe from the head through the weld examination volume with

+/-200 probe skewing. Figure 32 shows the cross section with the rays of the transverse scan of the 450 probe from the boss. Figure 33 shows the cross section with the rays of the transverse scan of the 60* probe from the head through the weld examination volume. Figure 34 shows the cross section with the rays of the transverse scan of the 60* probe from the boss.

Figure 35 shows the combined coverage achieved by all the transverse scans of the nozzle-to-head weld with +/-20' probe skewing.

Coverage For Probe Angle=45 Deg; Probe Skew=90 Deg (CP-SGPON) 85 84 -

83 82 81 80 79 78-77 -

76 -

Z75 -

74 -

73 -

72 -

71 -

70 -

69 -

68 -

67 -

66 -

65 -,', '1I I 1 ' I I I-14 15 16 17 18 19 20 21 22 24 25 26 27 28 29 30 31 32

Coverage For Probe Angle=45 Deg; Probe Skew=90 Deg (CP-SGPON) 85 -

84 83 82 -

81 80 -

79 78 -

77 -

I 76 -

I Z 75 I I

74 - I 73 - I 9 I

72 -

I 71 -

70 -

• P 69 --

68 --

67 -

66 - -

65 -65 I 15- 16 17-18II I I I .I I I I 1 I i i

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i i

I i I

i. i i " i" i 14 15 16 17 18 19 20 21 22 24 25 26 27 28 29 30 31 32

Coverage For Probe Angle=60 Dog; Probe Skew=90 Dog (CP-SGPON) 85 84 83 p 82 -

• 81 80 79 78 -

77 -

76 -

Z 75 -

74 -

73 - U

• 72 -

71 -

70 --

69 --

68 67 -

66 -,

I I I ~~~ I I I

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I I

l I

I I

l I

l I

I I 3I I

I I

I 24 25 26 27 28 29 30 31 32 14 15 16 17 18 19 20 21 22

Coverage For Probe Angle=60 Dog; Probe Skew=90 Dog (CP-SGPON) 85 -

84 83 p 82 81 -

80 -

79 78 77 76 - ,

• Z75 -

74 -

73 72 71 70 - .

69 -

68 --

67 --

66 - -

65 - I1 14 15 16 17 18 19 20 21 22 ""24 25 26 27 28 29 30 31 32

Combined Coverage (CP-SGPON) 85 F

84-83 -

82 -

81 80 79 78 77 76 Z 75 74 -

73 7 72 -

71 -

70 -

69 68 -

67 -

66 -

65 - I I I I I I I I I I I I I I I I I I I I I I I I 1 I 24 I I I 2" I 29 3 I1 3 14 15 16 17 18 19 20 21 22 24 25 26 27 28 29 30 31 32

Table 8. Coverage Table: SG Outlet; 45'and 600 probes scanned transversely on vessel head and nozzle boss (+/-20' probe skewing).

Probe Angle 450 450 600 600 Probe Skew 700 1100 70° 1100 Probe Location Head Boss Head Boss

% Exam Vol. (C-D-E-F) Covered 100% 0% 100% 0%

% Exam Vol. (B-C-F-G) Covered 80% 0% 59% 0%

% Exam Vol. (A-B-G-H-I) Covered 8% 0% 0% 0%

Percent Total Weld Exam Volume (A-B-C-D-E-F-G-H-I) Covered 54% 0% 47% 0%

Summary The 350 contoured probe gives 100% coverage for the inner radius regions of both the SG inlet and outlet nozzles (See Tables I and 5). Table 9 summarizes the coverage of the nozzle-to-head weld for the SG inlet nozzle (See Tables 2 and 4).

Table 9. Summary Coverage Table: SG Inlet Nozzle-to-Head Weld.

Probe Angles 450 & 600 450& 600 450 & 600 450& 600 Probe Skew 00 1800 700 1100 Probe Location Head Boss Head Boss Percent Coverage 79% 0% 63% 0%

Table 10 summarizes the coverage of the nozzle-to-head weld for the SG outlet nozzle (See Tables 6 and 8).

Table 10. Summary Coverage Table: SG Outlet Nozzle-to-Head Weld.

Probe Angles 450 & 600 450& 600 450 & 600 450& 600 Probe Skew 00 1800 700 1100 Probe Location Head Boss Head Boss Percent Coverage 74% 0% 47% 0%

Coverage For Probe Angle=45 Dog; Probe Skew=70 Dog (CP-SGPON) 85 -

84 -

83- p 82 -

if 81 - U 80 a

I 79 N

78 -

I 77 - U 76 - U Z75 - U U

74 -

73 -

.72 -

71 -

70 -

69 -

68 -

67 -

66 -

65 I I I I I W II- iI I i I i I i i II415 1617 1819 20 21 2224 25 262728 *293031 32

/

Coverage For Probe Angle=45 Dog; Probe Skew=110 Deg (CP-SGPON) 85 -

84 -

83.

82 81 w 80 N 3

79 I

78

• U I

77 I

• U 76 3

• I I I

• I I Z 75 3 3 I U

74

• U 1 73 3. I U

.72 I I

71 70 69 68 67 -

66 I I I I I I I I I I I I I I I 65 4-15-I 6 178I I I I 2I I I I 2 2I5I I 2I 28 2930I I 32 4 15 16 17 18 19 20 21 22 24 25 26 27 28 29 30 31 32 I

j'

Coverage For Probe Angle=60 Dog; Probe Skew=70 Deg (CP-SGPON) 85 84 83 82 81 -

80 79 -

78 77 -

76 Z 75 74 -

73 -

,72 -

71 --

70 69 68 -

67 66 658 2 2 26 8 30 31 2 14 15 16 17 18 19 20 21 22 24 25 26 21 28 29 30 31 32

Coverage For Probe Angle=60 Deg; Probe Skew=110 Dog (CP-SGPON) 85 84 83 82 81 -

80 U 79 78 77 -

76 - 3 Z 75-74 - -

73 - -

72 - -

71 -

70- .

69 68 67 -

66 65 24 5 2 2 2 2 3 3 14 15 16 17 IS 19 20 21 22 ""24 25 26 27 28 29 30 31 32

Combined Coverage (CP-SGPON) 85 -

84 -

83 -

82 -

81 - a 80 - N I

79 - I I

I 78 - I tn 77 - U cn a wI 76 - U I bc4 Z 75 -

I 74 -

73 -

72 -

71 70 -

69 68 67 66 -

I I I I I I I I I I I I I I I I I I 65 I4 18 19 20 21 22 -- 24 25 26 27 28 29 30 31 32 15 16 17