ML22201A504

IP 50003 Pressurizer Replacement Inspection

ML22201A504
Person / Time
Issue date:	02/04/2025
From:	Shawn Campbell NRC/NRR/DRO/IRIB
To:
stephen.campbell@nrc.gov
References
DC 22-017, CN 25-001
Download: ML22201A504 (1)

Text

Issue Date: 02/04/25 1

50003 NRC INSPECTION MANUAL IRIB INSPECTION PROCEDURE 50003 PRESSURIZER REPLACEMENT INSPECTION Effective Date: February 4, 2025 PROGRAM APPLICABILITY: IMC 2515 C 50003-01 INSPECTION OBJECTIVES 01.01 To verify that engineering evaluations and design changes associated with pressurizer replacement are completed in conformance with requirements in the facility license, the applicable codes and standards, licensing commitments, and the regulations.

01.02 To verify that pressurizer removal and replacement activities maintain adequate nuclear and radiological safety.

01.03 To verify that pressurizer post-installation test program is technically adequate, in conformance with requirements, and satisfactorily implemented.

01.04 To verify appropriate shutdown risk assessments, work controls, and risk management actions are implemented during the pressurizer removal and replacement activities.

50003-02 INSPECTION REQUIREMENTS 02.01 Inspection Scoping Develop a site-specific inspection plan to select and review the safety-related aspects associated with the major phases of the pressurizer replacement project (PRP).

During inspection preparation, review the processes for PRP and plans for identifying, tracking, and resolving problems related to PRP to ascertain that they are consistent with plant processes.

02.02 Design and Planning Inspections Conduct pressurizer design and planning inspections using the inspection plan by performing selective inspections, consistent with the safety significance and inspection resources, of the following areas:

a. Conduct pressurizer replacement engineering and technical support inspections using the inspection plan by performing selective inspections that will:

1. Review and verify that the modifications and design changes made to systems, structures and components (SSCs) described in the Updated Final Safety Analysis Report (UFSAR) were reviewed under Title 10 of the Code of Federal Regulations (10 CFR) Section 50.59 Changes, Tests and Experiments. Review and verify that

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50003 the new pressurizer and associated changes to and support SSCs (including temporary containment openings) either meet the original design requirements or that any discrepancies were dispositioned accordingly.

2. Review selected key procurement specifications for the pressurizer and its components to determine that they meet applicable industry codes and standards and regulatory requirements including 10 CFR 50, Appendix B, Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants, and environmental qualification requirements. Confirm the new pressurizer vessel was properly fabricated and satisfies the requirements of the American Society of Mechanical Engineers (ASME) Code,Section III, Rules for Construction of Nuclear Facility Components.

3. Verify if the licensee has confirmed that the replacement pressurizer conforms to design drawings, and that there are no fabrication deviations from design.

4. Confirm that this ASME Code Class 1 vessel was hydrostatically tested and N-stamped.

b. Review the applicable engineering design, modification, and analyses associated with pressurizer lifting and rigging including: (1) crane and rigging equipment, (2) pressurizer component drop analysis, (3) safe load paths, and (4) load lay-down areas, and (5) heavy load haul path.

c. Review radiation protection program controls, planning, and preparation in the following areas:

1. as low as reasonably achievable (ALARA) planning

2. dose estimates and dose tracking

3. exposure controls including temporary shielding

4. contamination controls

5. radioactive material management

6. radiological work plans and controls

7. emergency contingencies

8. project staffing and training plans

d. Review security program controls, planning, and preparation for the PRP activities in the following areas:

1. vital and protected area barriers that may be affected during replacement activities

2. where applicable at multi-unit sites, the controls and plans to minimize any adverse impact on the operating unit(s) and common systems

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e. Review licensees maintenance rule (10 CFR 50.65 (a)(4)) risk assessment of PRP. In addition, review risk assessment of planned modifications to ensure shutdown risk management objectives are acceptable (e.g., reduced inventory considerations, control of heavy loads, refueling operations, and availability of alternate power).

02.03 Pressurizer Removal and Replacement Inspections Conduct pressurizer removal and replacement inspections in accordance with the inspection plan by performing selective inspections, consistent with the safety significance and inspection resources, of the following areas:

a. Review the following welding and non-destructive examination (NDE) activities:

1. where applicable, special procedures for welding and NDE

2. training and qualifications for personnel performing welding and NDE

3. NDE including radiography results and work packages for selected welds

4. completion of any pre-service NDE requirements, where required

b. Review activities associated with lifting and rigging: preparations and procedures for rigging and heavy lifting including any required crane and rigging inspections, testing, equipment modifications, lay-down area preparations, and training of personnel.

c. Inspect major structural modifications, if any, that are performed to facilitate pressurizer replacement.

d. Inspect activities associated with containment access and, where applicable, creation and restoration of temporary containment opening and containment leakage testing.

e. Inspect the following activities throughout the process as appropriate:

1. Establishment of operating conditions including defueling, reactor coolant system (RCS) drain down, and system isolation and safety tagging/blocking.

2. Implementation of radiation protection controls.

3. Controls for excluding foreign material.

4. Installation, use, and removal of temporary services directly related to the activities identified in this procedure.

5. Implementation of fire prevention and mitigation plans and controls during the replacement process (ignition sources, combustible source loading, and fire watches).

f.

Review radiological safety plans for temporary storage or disposal of old pressurizer and components.

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50003 02.04 Post-installation Verification and Testing Inspections Conduct pressurizer post-installation verification and testing inspections in accordance with the inspection plan. Perform selective inspections, consistent with the safety significance and inspection resources, of the following areas:

containment integrity and leakage testing, as applicable.

the licensee's post-installation inspections and verifications program and its implementation.

the conduct of RCS leakage testing and review the test results.

calibration and testing of instrumentation affected by pressurizer replacement.

the procedures for equipment performance testing required to confirm the design and to establish baseline measurements and the conduct of testing.

preservice inspection of new welds.

Verify Pressurizer thermal and hydraulic performance. If the Pressurizer design changes include different operating levels, capacity/volumes or spray rates or heater capacity or heater locations, review and confirm that the bounding plant transients have been fully analyzed for the new design or that appropriate integrated plant testing has been completed to demonstrate the Pressurizer system performance.

50003-03 INSPECTION GUIDANCE General Guidance Pressurizer replacement at a nuclear power plant is a significant modification/maintenance activity involving many different licensee disciplines and has historically required extensive contractor support. This activity can potentially affect the power plant safety analysis, the containment structure, and plant operational characteristics.

Comprehensive Nuclear Regulatory Commission (NRC) inspection of pressurizer replacement activities involves coordination of inspections to review engineering, safety assessment, operations, maintenance, welding, NDE, instrumentation, quality assurance, radiation protection, security, and testing. It is expected that inspections will be performed primarily by regional specialists and the resident inspectors.

Because of the engineering scope of a replacement project and its safety significance, licensees have voluntarily participated in extensive interactions with the regions and technical branches in the Office of Nuclear Reactor Regulation (NRR). This interaction starts early (three or more years before scheduled replacement) in the design effort and prior to contracting a vendor for pressurizer fabrication. It is expected that these interactions will continue with future PRPs.

Because of the complexity of the PRP, technical support from NRR technical staff for engineering reviews during the inspections should be considered. Those engineering reviews should be coordinated through the NRR project manager. Due to the lead time associated with pressurizer replacement, the resident inspector should be cognizant of upcoming pressurizer replacement at the site. As such, the resident should inform regional management of upcoming

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50003 pressurizer replacement at the site, so that adequate time is available for the coordination of the necessary staff and resources to ensure adequate inspection.

All inspections of pressurizer replacement activities should be performed in accordance with this procedure and the site-specific inspection plan developed to support this procedure. The site-specific inspection plan should be developed and administered by the regional office to integrate the inspection activities of the specialist inspectors.

The site-specific inspection plan should include the applicable inspection requirements outlined in Section 02 of this procedure. The expenditure of inspection resources and the emphasis on inspection activities should be based on the scope of the replacement activities and associated modifications, the safety significance of the activities, the licensees historical performance in that area, and industry experience. Additional inspection emphasis may be considered for those aspects of the replacement project that include new or different management controls or involve new techniques. For example, there may be new administrative procedures or quality assurance programs for the project or specialized welding techniques such as the narrow gap method.

The inspection requirements in Section 02 of this procedure are general and address the basic activities that should be inspected. Additional guidance for specific inspections should be derived from inspection procedures relative to the area being examined. Those procedures include, but are not limited to, the procedures listed in Appendix A, Applicable Inspection Procedures, of this procedure.

NOTE: Some of the procedures listed applied to original construction activities, preoperational testing, and startup testing and may be used as guidance for inspection of pressurizer replacement.

Specific Guidance 03.01 Inspection Scoping The major phases of PRP are typically design and planning, removal and replacement, and post-installation verification and testing. The licensees design and planning phase may begin several years before replacement. Contracts for pressurizer procurement have been issued three or more years before replacement. The licensee may install temporary or permanent modifications to support the PRP well in advance of the PRP outage. Significant design work and outage planning may be completed a year before the outage.

An early review of the licensees PRP scope and schedule will provide advance opportunity to identify special inspection needs and plan the design and planning phase inspections. For the significant support modifications installed prior to the outage, inspection before the PRP outage is recommended. About a year before the PRP outage, the licensees outage scope and schedule should be detailed enough to develop the inspection plan for review of onsite work. Inspection planning has typically been performed by a region-based inspector with input from the resident inspector.

Also, refer to Appendix B, Temporary Containment Opening Review Guidance, of this procedure for additional guidance.

The inspection should focus on PRP activities that verify proper restoration of pressure boundaries of the RCS, secondary systems, and containment systems, exclusion of

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50003 foreign materials, and plant modifications that could affect plant risk during subsequent plant operation.

This procedure lists certain baseline inspection procedures (IPs) to be used to perform parts of the inspection effort. The scope and depth of the inspection of these parts and the inspection hours credited to the baseline inspections should be in accordance with the guidance in this procedure. Inspection planning for the baseline inspections should consider the extended schedule for the PRP and the prudent management of baseline inspection hours.

The inspector should contact the NRR project manager to obtain any specific technical input related to the inspection from NRR licensing review of any licensee PRP submittal.

03.02 Design and Planning Inspections

a. Inspection samples of permanent and temporary plant modifications for the PRP should be selected and reviewed as a part of the baseline inspection using applicable portions of IP 71111.18, Plant Modifications, and IP 71111.21M, Comprehensive Engineering Team Inspection.

Title 10 CFR 50.59 evaluations and screening for such evaluations for selected modifications related to PRP should be reviewed using applicable portions of IP 71111.21M.

The inspection should verify that (1) key pressurizer design aspects such as those listed below, and pressurizer modifications, and the designs of other related significant modifications are reviewed and approved in accordance with procedures and (2) replacement materials and components meet the appropriate design technical requirements. Those technical requirements include the applicable codes and standards, NRC requirements, and other commitments made by the licensee in the UFSAR.

The inspection should review design attributes like those outlined in IPs 71111.18, and 71111.21M.

The following list illustrates typical topics that may need to be reviewed:

the cumulative and synergistic effects, if any, of the pressurizer, related design changes, and other modifications completed during the outage on transient and accident analyses.

adherence to and reconciliation of ASME code requirements.

compliance with regulatory requirements including the incorporation of inservice inspection requirements of 10 CFR 50.55a (g).

The surge pipe connected to the bottom of the pressurizer has experienced cracking because of severe temperature transients during insurge and outsurge evolution. The inspector should verify in the field that the fit-up between the surge pipe and new pressurizer is within design tolerance before the licensee makes the weld joint.

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50003 The operating experience has shown that cracking has occurred on the spray line nozzle at the top of the pressurizer. The inspector should verify in the field the fit-up between the spray nozzle and the new pressurizer is within design tolerance before the licensee makes the weld joint.

The operating experience has shown that heater sleeves and heaters have had cracking. The inspector should verify that during the new pressurizer installation, the licensee attached the heaters properly.

Confirm that the new pressurizer is installed at the exact location and elevation as the old pressurizer such that the fit-up between the existing pipes and the new pressurizer nozzles before welding are within the design tolerance (clearance).

Confirm that the design tolerance is within the stress analyses (including various operating conditions) of the existing pipes and new pressurizer. Confirm that the fit-up between the existing pipes and the new pressurizer nozzles in the field is within the design tolerance. This is to minimize the stresses on the existing pipes and new pressurizer nozzles.

If deviations exist between the replacement pressurizer and the old pressurizer dimensionally (e.g., outside diameter, height, supporting skirt) and nozzle locations on the pressurizer, the inspector needs to verify whether the licensee has performed new stress analyses or updated stress analyses of the attached pipes. In any event, the inspector needs to verify that the new/updated/existing stress analyses of pipes attached to the replacement pressurizer are applicable after replacement pressurizer installation (i.e., the resulting stresses at the pipes and pressurizer nozzles are within the allowable stresses of the ASME Code,Section III, and the analytical model represents the replacement pressurizer).

The inspector should also verify that the resulting loads on the replacement pressurizer nozzles do not deviate excessively (more than 5 percent) from the nozzle loads in the existing stress analyses. Inspectors should contact cognizant NRR technical branch.

1. The general design for removal and reinstallation of the Pressurizer and related components. The impact of changes in mass and center of gravity of the new pressurizer on the seismic analysis for the containment structure, pipe stress analysis, and other safety systems and components

2. Procurement specifications including qualification and functional requirements if needed, the design and analysis for the creation of a temporary containment opening for pressurizer replacement and its restorationdesign reviews should be coordinated with the Structural, Civil, Geotechnical Engineering Branch (ESEB),

NRR, via the Project Manager, to confirm the restored design margin; see Appendix B of this procedure for additional information. If available, use manufacturers shop procedures, licensees purchase orders and other relevant documents.

3. No specific guidance.

4. No specific guidance.

b. The inspection should focus on the impact of load handling activities on reactor core or spent fuel and its cooling and plant support systems for the reactor unit and common systems for the other operating unit(s) at the site. Several heavy lifting evolutions are

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50003 expected. While the radiological consequences of a heavy load drop are expected to be small with the core defueled, a dropped component could result in an unwarranted radioactive release and severe damage to equipment needed for reactor or radiological safety.

The licensees plans and analysis for lifting and rigging of heavy loads are reviewed to verify that the safe load path analysis for component removal and reinstallation is technically sound. The inspection should focus on the impact of heavy load lifting operations on spent fuel and its cooling, support systems for the reactor, and common support systems for the other operating reactor unit(s) at the site.

Modifications such as reinforcement of existing structures and/or floors, construction of new structures or platforms, and changes to the crane should be reviewed to verify that the modified equipment will safely handle expected loads. The potential impact of these modifications on safety-related equipment should be assessed for applicability under 10 CFR 50.59.

The component drop analysis should be reviewed to verify, in general, that the potential offsite releases at the exclusion area boundary are within 10 CFR Part 100, Reactor site criteria, limits and that equipment to maintain safe shutdown will be unaffected.

Additional information on lifting heavy loads is available in Generic Letter 81-07, Control of Heavy Loads, NUREG-0612, Control of Heavy Loads at Nuclear Power Plants, and Regulatory Guide 1.244, Control of Heavy Loads at Nuclear Facilities.

c. Specific guidance for radiation protection inspections is provided in IP 71124.01.

The inspector should audit the licensees outage radiation protection program and confirm that radiological concerns are factored into PRP planning. When applicable, review of radiological concerns associated with the establishment and restoration of a temporary containment opening should be included.

For exposure and contamination control issues, attention should be directed to potential high dose and/or high contamination activities such as reactor coolant system piping cutting and welding or work in or around the pressurizer. Where applicable, plans for shielding installations should consider the effects of additional weight loads on equipment.

Training for the large PRP work force should be geared toward the specifics of pressurizer removal/replacement for both the radiation workers and radiation protection personnel. Review of planning for radioactive materials management should be considered due to the need for temporary storage of reusable equipment and the handling of the increased volumes of waste generated in a short time. Efficient removal of materials from the work area improves housekeeping and reduces fire hazards.

Emergency contingencies such as actions for a dropped component and the plans for closure of the containment (including temporary openings) in a high airborne condition should be reviewed.

Refer to NUREG/CR-1595, Radiological Assessment of Steam Generator Removal and Replacement Update and Revision, (Agencywide Documents Access and Management System (ADAMS) Accession No. ML19345E032) for additional information on radiological considerations.

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d. Specific guidance for security barriers and boundaries.

1. Security boundaries may be affected during the PRP. The impact of the work on the boundaries and the licensees contingency plans should be reviewed to verify conformance with the security plan.

2. No specific guidance

e. Licensee risk assessment and risk management actions for the PRP should be reviewed using IP 71111.13, Maintenance Risk Assessments and Emergent Work Control. In addition, licensee outage activities and actions to reduce shutdown risk should be reviewed in accordance with IP 71111.20, Refueling and Other Outage Activities. This is to ensure that licensees consider risk in developing outage schedules; adhere to administrative risk reduction methodologies they develop to control plant configuration; have developed mitigation strategies for losses of key safety functions; and adhere to operating license and technical specification requirements that ensure defense-in-depth.

03.03 Pressurizer Removal and Replacement Inspections

a. Use IP 71111.08, Inservice Inspection Activities, as a guide.

1. Significant cutting, machining, and welding activities are expected with Pressurizer replacement outages. Several procedures are listed in Appendix A to this procedure that contain useful guidance for welding inspections.

2. Training and qualifications of licensee and contract quality control/assurance inspectors, and NDE examiners are reviewed to verify personnel meet site and code qualification requirements and are prepared for the site-specific tasks. In the past, mockups have been used for qualifying welding procedures and training and qualifying welders, machinists, and NDE inspectors.

3. Procedures related to radiography are located in Appendix A to this procedure. The inspector should include the nondestructive examination results in the inspection report.

4. Procedures related to preservice NDE requirements are in Appendix A to this procedure.

For the selected welds review weld procedures and welder qualification records, confirm that the ASME Code required essential and supplemental essential welding variables for the welding processes used were met, and verify that preservice NDE requirements were completed with acceptable results in accordance with the ASME Code,Section III.

Also, review the non-conformance reports for each major Pressurizer replacement weld to confirm that welding deficiencies were dispositioned in accordance with ASME Code requirements.

After the existing pipes are welded to the pressurizer nozzles, the inspector should verify whether any repairs were performed on the welds. If so, the inspector should verify whether the licensee performed the repair in accordance with the ASME Code,Section III, Section XI, and the construction code. The inspector should discuss the repair in the inspection report and inform cognizant NRR technical branch. The operating

Issue Date: 02/04/25 10 50003 experience has shown that a repaired weld generates weld residual stresses, which may cause the weld to degrade in the future

b. No specific guidance.

c. The inspection should verify that modifications to walls and other structures and removal and restoration of component supports are documented.

d. A temporary containment opening may need to be established to allow for the movement of pressurizer components. The inspections should verify that the procedures and methods used to enlarge and restore the containment access ensure that the design assumptions and requirements are satisfied, and that the containment is restored to at least its required strength and integrity criteria. The inspection should also verify that the concrete removal and replacement processes do not degrade the liner to concrete interface. Post restoration testing should demonstrate that the containment has been properly restored. See Appendix B of this procedure for additional information.

e. Implementation of controls for personnel access to radiologically controlled work areas, surveillance of work activities, and procedure adherence should be verified. Reviews should be made of the implementation of ALARA, radiological exposure, contamination, and airborne contamination controls planned for cutting, welding, and other activities including contaminated interference removal. Also, implementation of any special controls for contaminated tools and waste should be reviewed. Where applicable, controls for shielding installations should consider the effects of additional weight loads on equipment.

If a temporary containment opening is established, the method of disposal of any concrete debris may be considered. While much of this material is not expected to be contaminated, consideration should be given to any activation materials including hard to detect isotopes such as H-3 and Fe-55.

Review licensees evaluations of adverse impact of temporary services (electrical power, structural supports, and systems) on plant equipment and systems important safety.

f.

If the old Pressurizer is stored on site, the storage facility should be reviewed to verify that access is properly controlled and dose rates at the perimeter are below applicable limits. For additional information, refer to Generic Letter 81-38, Storage of Low-Level Radioactive Wastes at Power Reactor Sites.

03.04 Post-installation Verification and Testing Inspections.

The inspection should review the licensees post-installation verification and testing program to verify that modifications are completed in accordance with the design; that drawings, procedures, and training have been updated as appropriate; that post-installation walkdowns and inspections are performed to ensure equipment is restored and temporary services are removed; that equipment cleanliness has been verified; that pre-service inspection of welds to establish baseline data are performed; and that deficiencies are properly dispositioned. Verify that changes in performance of the Pressurizer and in its associated parameters, such as flow rates, pressures, and temperatures are appropriately included in design documents and plant procedures. In

Issue Date: 02/04/25 11 50003 addition to reviewing the licensees program to verify restoration, the inspector may perform independent walkdowns to confirm equipment restoration.

If a temporary containment opening was established for Pressurizer replacement, refer to Appendix B of this procedure for additional guidance on testing following restoration.

Direct inspection of portions of system leakage tests and associated visual examinations and review of the test results should be performed. The inspection should verify that testing satisfies the ASME Code,Section XI, IWA-5000, and applicable regulatory requirements, that testing was conducted according to the procedure, and that results were satisfactory or properly resolved.

Section 02.04 of Inspection Procedure 93803, Safety Systems Outage Modifications Inspection, contains additional inspection guidance for testing inspections.

50003-04 INSPECTION RESOURCES The required number of direct inspection hours per plant will vary significantly based on several factors such as licensee experience, contractor experience, and the variable complexities of the given pressurizer replacement. The scope and depth of the inspections should be sufficient to provide the desired level of assurance that the licensee adequately performs activities important to safety and that the relevant codes, standards, requirements, and commitments are met. Not all items in this procedure must be done to closeout this inspection procedure. The total resources required for PRP inspection using IP 50003 is estimated to be 350 direct inspection hours. It is expected that portions of the inspections in this procedure can be credited towards satisfying various baseline inspections, as appropriate. Approximately 140 hours0.00162 days <br />0.0389 hours <br />2.314815e-4 weeks <br />5.327e-5 months <br /> at a single unit station and 110 hours0.00127 days <br />0.0306 hours <br />1.818783e-4 weeks <br />4.1855e-5 months <br /> at a multi-unit site are expected to be credited towards baseline inspections. Inspection of the establishment/restoration of a temporary containment opening will require additional inspection effort.

50003-05 REFERENCES ASME Boiler & Pressure Vessel Code,Section III, Rules for Construction of Nuclear Facility Components (See 10 CFR 50.55a, Codes and standards)

ASME Boiler & Pressure Vessel Code,Section IX, Qualification Standard for Welding, Brazing, and Fusing Qualifications ASME Boiler & Pressure Vessel Code,Section XI, Division 1, Rules for Inspection and Testing of Components of Light-Water-Cooled Plants Generic Letter 81-07, Control of Heavy Loads Generic Letter 81-38, Storage of Low-Level Radioactive Wastes at Power Reactor Sites NRC Regulatory Issue Summary 2004-16: Use of Later Editions and Addenda to ASME Code Section XI for Repair/Replacement Activities NUREG/CR-1595, Radiological Assessment of Steam Generator Removal and Replacement Update and Revision (ML19345E032)

Issue Date: 02/04/25 12 50003 NUREG-0612 Control of Heavy Loads at Nuclear Power Plants (ML070250180)

Regulatory Guide (RG) 1.136, Revision 42, Design Limits, Loading Combinations, Materials, Construction, and Testing of Concrete Containments (ML20301A167)

RG 1.244, Revision 0, Control of Heavy Loads at Nuclear Facilities (ML21006A346)

RG 1.57, Revision 2, Design Limits and Loading Combinations for Metal Primary Reactor Containment System Components (ML12325A043)

END Appendices:

A. Applicable Inspection Procedures B. Temporary Containment Opening Review Guidance : Revision History for IP 50003

Issue Date: 02/04/25 AppA-1 50003 Appendix A: Applicable Inspection Procedures Inspection Procedure No.

Inspection Procedure Title 71111.08 Inservice Inspection Activities 71111.13 Maintenance Risk Assessments and Emergent Work Control 71111.18 Plant Modifications 71111.20 Refueling and Other Outage Activities 71111.21M Comprehensive Engineering Team Inspection 71124.01 Radiological Hazard Assessment and Exposure Controls 71124.03 In-Plant Airborne Radioactivity Control and Mitigation 71124.04 Occupational Dose Assessment 71124.06 Radioactive Gaseous and Liquid Effluent Treatment 71124.08 Radioactive Solid Waste Processing and Radioactive Material Handling, Storage, and Transportation 55050 Nuclear Welding General Inspection Procedure 55100 Structural Welding General Inspection Procedure 55150 Weld Verification Checklist 57050 Visual Testing Examination 57060 Liquid Penetrant Testing Examination 57070 Magnetic Particle Testing Examination 57080 Ultrasonic Testing Examination 57090 Nondestructive Examination Procedure Radiographic Examination Procedure Review/Work Observation/Record Review 88143 Pipe Supports and Restraints 93803 Safety Systems Outage Modifications Inspection Containment Inspections**

70307 Containment Integrated Leak Rate Test - Procedure Review 70313 Containment Integrated Leak Rate Test 70323 Containment Leak Rate Test Results Evaluation

• • May only be applicable when the containment access has been enlarged.

END

Issue Date: 02/04/25 AppB-1 50003 Appendix B: Temporary Containment Opening Review Guidance A.

OBJECTIVE This appendix provides additional guidance for inspection of activities associated with establishing a temporary containment opening in the containment shell for steam generator (SG) replacement and its subsequent restoration.

B.

BACKGROUND To replace the pressurizer in containments that have small equipment hatches, some licensees have chosen to replace the pressurizer in one piece through a temporary transfer opening cut through the containment instead of disassembling the pressurizer and transporting its sections and components through the equipment hatch. Based on time and radiological considerations, this approach has generally been preferable. After completion of the pressurizer replacement, the opening is closed, the containment structure is restored to its original configuration and integrity, and an integrated leak rate test (ILRT) is performed. Cutting the temporary opening and closing it may be carried out without NRC's approval under the provisions of Title 10 of the Code of Federal Regulations (10 CFR) Section 50.59. Changes, tests and experiments, provided the process does not meet the criteria contained in 10 CFR 50.59(c)(2) or require a change in technical specifications. However, since this process involves construction activities for the most important safety-related structure in a nuclear power plant, quality assurance and quality control practices should be strictly followed, and the restoration of the containment should be inspected by the NRC.

C.

GUIDANCE General This appendix is intended to augment the existing inspection guidance in this procedure to identify regulatory concerns and licensee activities associated with cutting and closing a temporary containment opening. Some of the discussion below may overlap previous guidance and is only provided for clarity.

Nuclear Regulatory Commission inspection emphasis should be on verification that the containment is restored to its original integrity after closure of the temporary opening. Because actions related to cutting and restoring a temporary opening in a prestressed concrete containment encompass issues concerning reinforced concrete and steel containments, issues related to prestressed concrete containments are used as the basis for this appendix.

Therefore, some items discussed below may not apply to all containment designs.

Construction as used herein is an all-inclusive term that covers materials, design, fabrication, installation, examination, and testing. The licensee should prepare specifications covering all aspects of the construction of the temporary opening as discussed in the following sections.

Regulatory Concerns In a prestressed concrete containment, the components that are cut and replaced include concrete, reinforcing bars, tendon sheathing, and the steel liner. The tendons, together with the grease in the sheathings that pass through the opening area, are removed before the opening is cut. The tendons may be reused in the restoration. The concerns are as follows:

Issue Date: 02/04/25 AppB-2 50003

1. The steel liner plate (or metallic shell for steel containments) may be torch-cut and reused. During reinstallation, the gap around the liner plate for welding may not be as uniform as it would be for new plates. The uniformity of the gap is dependent upon how skillfully the plate was initially torch-cut. The strength of the replacement welds may not be the same as the uncut plate, but it should be very similar to the strength of the weld joints in the original structure. If the strength of the replacement welds will be less than the uncut plate, the reduction in weld joint strength should be evaluated as an engineering design change.

2. The creep and shrinkage of the concrete used to close the opening, which is usually high for newly poured concrete, could result in higher compressive strain in the concrete adjacent to the repaired opening because of prestressing and may cause liner buckling.

3. The creep and relaxation of reused tendons could cause the loss of prestress because of tensioning and detensioning.

4. The leak tightness of the tendon sheathing may be compromised because of withdrawing and re-inserting the tendons, resulting in grease leakage. This leakage may reduce the strength of the concrete and leave the tendon unprotected.

5. When containment boundary concrete replacement includes having the reinforcing bar splices lined up without staggering, confirm that an engineering-based evaluation has verified that the strength of the containment has not been compromised.

6. After closing the opening, the containment is to be subjected to a containment pressure test and leakage test in accordance with Subsection IWL and Subsection IWE (specifically Articles IWL-5000 and IWE-5000) of the ASME Code Section XI, as incorporated by reference in 10 CFR 50.55a and subject to the applicable conditions in 10 CFR 50.55a(b)(2)(viii) and (ix). It should be noted that, when applying IWE-5000 to Class MC pressure-retaining boundary components, the condition in 10 CFR 50.55a(b)(2)(ix)(J) requires a Type A integrated leak rate test (ILRT) in accordance with 10 CFR 50, Appendix J, Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors, following a major containment modification or repair/replacement activity (e.g., construction openings for replacement of RV head, steam generators or pressurizers). Proposed alternative(s) to these requirements may be requested by the licensee and authorized by the NRC pursuant to 10 CFR 50.55a(z).

Design Reviews Design reviews should be coordinated with the Structural, Civil, and Geotech Engineering Branch, the Office of Nuclear Reactor Regulation (NRR), via the Project Manager, to confirm the restored design margin of the containment structure.

Codes and Standards Acceptable options for industry codes and standards to be used for the design, construction, and restoration of the containment access are as follows:

(1) The codes and standards described in the FSAR, consistent with current regulations, as the basis of the facility operating license are to be utilized.

Or preferably,

Issue Date: 02/04/25 AppB-3 50003 (2) To the extent practical, the latest ASME Boiler and Pressure Vessel Code (ASME Code) edition and addenda. Practical adjustments can be made in the use of the updated ASME Code provisions to accommodate the limitations of design and construction. The quality and overall margins required in the original design are to be maintained and should be included in decisions on the extent to which adjustments are made for using the updated ASME code provisions. If the licensee chooses this option, review of its acceptability should be coordinated with the ESEB, via the Project Manager.

Analysis The licensee should perform structural analyses for the containment, which considers the effect of the opening for the applicable loads and load combinations resulting from the containment opening.

Structural analyses of the containment structure should be performed considering both the presence of the temporary opening and conditions after its repair. This analysis should incorporate the properties of the materials of the replacement components, which may affect the behavior of the containment structure locally and globally. The analyses should demonstrate that the Code allowable limits have not been exceeded in both cases. The loads and loading combinations for the restored configuration (post-opening closure) of containment should be the same as those specified in the FSAR or the applicable Standard Review Plan sections. Also, consideration should be given to the conditions arising from the existence of the temporary opening, ranging from initial cutting to closing, including potential construction loads. For the interim configuration with temporary opening of containment, no design pressure (Pa ) load associated with the design basis accident will apply in the analysis.

Design Changes The materials and replacement components used to restore the containment should be like those originally used and documented in the FSAR except that the concrete used to close the opening can be adjusted to reduce the creep and shrinkage. If different materials or components are used, they should satisfy the applicable design criteria and requirements and provide an equivalent overall safety margin as provided in the originally licensed design.

Temporary Opening Inspections Based on the regulatory concerns previously described, the areas of inspection interest can be summarized as follows:

1. Welding of the steel liner or shell since the torch cut may result in irregular edges.

The welding process, use of proper materials, the skill and training of the welder, and post-installation examination and testing should be considered. The replacement weld strength should be confirmed to be compatible with the containment design and it should be very similar to the strength of the weld joints in the original structure.

2. Proper concrete mix design and qualification and proper placement techniques should ensure that concrete in the repair area exhibits low creep and shrinkage. The certification of the concrete mix design should ensure that the design mix results in properties like that used in original construction, and that the mix is controlled to reduce creep and shrinkage. Sufficient testing should have been done prior to placement to ensure that design strength is attainable. The documentation of the certification of materials used in the concrete containment, such as aggregate,

Issue Date: 02/04/25 AppB-4 50003 cement, water, and admixtures should be reviewed to ensure they meet specified quality requirements.

Witnessing the initial concrete placement is important to verify that the concrete is being handled and tested properly. In cases where the interior form for the concrete containment is the liner plate, additional stiffeners may be needed to prevent buckling.

3. The tensioning force of the tendons, which may be reused and tensioned to a level of preload which would not cause more creep and relaxation. The prestressing sequence should also be carefully implemented.

4. The control of grease filling so that the tendons are adequately covered and there is no leakage from the sheathing that can be observed on the surface of the containment.

5. The rebar splicing around the opening and the use of the sister splices for quality control testing. Checks on the proper certification of the rebar and splice materials and worker qualifications may also be included.

6. Witnessing of the containment pressure test or the integrated leak rate test (ILRT) of the containment.

Fabrication, Installation, Examination, and Testing Fabrication, installation, examination, and testing with appropriate instrumentation are to be performed in accordance with the applicable ASME Code as stated above under Codes and Standards or the stipulations in the FSAR. The behavior of the repaired containment structure locally and globally as obtained from the containment pressure test or the ILRT is to be assessed.

Generally, the facility technical specifications require periodic examinations of the containment integrity, including inspections of the tendons. It is not expected that those examinations will be altered by the restoration of the temporary containment access. Of concern however is the potential for degradation of the containment structural integrity or leak tightness that could result from the restoration. For example, tendon degradation could occur due to moisture intrusion via a pathway created in a damaged or inadequately restored tendon duct. An assessment should be made of the licensee's measures and/or planned actions (e.g., analysis, testing, examination) to verify that containment integrity is maintained following the restoration.

Assistance in performing this assessment should be coordinated with the ESEB via the Project Manager.

Inspection and Quality Assurance Quality assurance and independent (third-party) inspection requirements for cutting and closing the temporary opening are applicable. Verification of these activities should be performed.

Training and Qualification The training and qualifications of craft and quality assurance personnel may be reviewed as there have been past problems with the use of inexperienced personnel.

END

Issue Date: 02/04/25 Att1-1 50003 : Revision History for IP 50003 Commitment Tracking Number Accession Number Issue Date Change Notice Description of Change Description of Training Required and Completion Date Comment Resolution and Closed Feedback Form Accession Number (Pre-Decisional, Non-Public Information)

N/A ML052010023 07/07/05 CN-05-018 Initial issuance of IP 50003 N/A N/A N/A ML17170A067 09/12/17 CN 17-018 Revisions to incorporate several recommended changes from ROP Feedback Form 50003-2243 N/A N/A 50003-2243 ML17254A041 N/A ML22201A504 02/04/25 CN 25-001 Researched commitments for 4 years and found none.

Minor editorial changes. This completes the 5-year review.

N/A ML22213A212