**NRC INSPECTION MANUAL**IRIB

INSPECTION PROCEDURE 50003

PRESSURIZER REPLACEMENT INSPECTION

Effective Date:  01/01/2018

PROGRAM APPLICABILITY:2515 App. C

50003-01INSPECTION OBJECTIVES

01.01To 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.

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

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

01.04To verify that shutdown risk is minimized for the Pressurizer removal and replacement activities. 

50003-02INSPECTION REQUIREMENTS

02.01Inspection 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).  A typical PRP involves three major phases: design and planning, removal and replacement, and post-installation verification and testing.  If the PRP includes the provision of a temporary containment access opening, refer to Appendix B of this procedure for additional guidance.  The inspection should focus on PRP activities that verify proper restoration of pressure boundaries of the reactor coolant system (RCS), secondary systems, and containment systems, exclusion of foreign materials, and plant modifications that could affect plant risk during subsequent plant operation.

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

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.

This procedure lists certain baseline inspection procedures 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 take into account the extended schedule for the PRP and the prudent management of baseline inspection hours.

02.02  Design and Planning Inspections. Conduct Pressurizer design and planning inspections in accordance with 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 in accordance with the inspection plan by performing selective inspections that will:

1. Verify that selected design changes and modifications to systems, structures, and components (SSCs) described in the Final Safety Analysis Report (FSAR) are reviewed in accordance with 10 CFR 50.59.  Use IP 71111.17T, IP71111.18, and IP 71111.21M as guidance.

2. Review key design aspects and modifications for the replacement Pressurizer and other modifications associated with Pressurizer replacement utilizing IP 71111.17T, IP 71111.18, and IP 71111.21M as guidance.  Where applicable, include design reviews for the provision of a temporary containment opening. 

3. 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 requirements, and environmental qualification requirements. 

4. Determine if the licensee has confirmed that the replacement Pressurizer conforms to design drawings, and that there are no fabrication deviations from design.  

5. Confirm that this Class 1 vessel was hydrostatically tested and N-stamped.

b. Review the applicable engineering design, modification, and analysis 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.  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, 

c. Review radiation protection program controls, planning, and preparation in the following areas utilizing applicable portions of the baseline inspection procedures 71121.01 and 71121.02 as guidance:

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 PRP  activities in the following areas:

1.Security considerations associated with 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.

e. Review licensee(s 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. 

02.03Pressurizer 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.

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

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

d. Inspect activities associated with containment access and, where applicable, 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, RCS draindown, 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 retired Pressurizer and components.

02.04Post-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:

a. Containment testing, as applicable.

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

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

d. Calibration and testing of instrumentation affected by Pressurizer replacement.

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

f. Preservice inspection of new welds.

g. 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.

50003-03INSPECTION 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 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 complexity of the PRP, technical support from NRR 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 his or her site.  As such, the resident should inform his or her regional management of upcoming pressurizer replacement at his or her 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.  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 licensee's 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 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 of this procedure.  Note that 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

03.01Inspection Scoping.  The major phases of PRP are typically design and planning, Pressurizer removal and replacement, and post-installation verification and testing.  The licensee's 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 licensee's 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 licensee's 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.

03.02Design 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 IP 71111.17T, IP 71111.18, and IP 71111.21M.

10 CFR 50.59 evaluations and screening for such evaluations for selected modifications related to PRP should be reviewed using IP 71111.17T, and 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 FSAR.  

The inspection should review design attributes similar to those outlined in IP 71111.17T, IP 71111.18, and IP 71111.21M.

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

-the general design for removal and reinstallation of the Pressurizer and related components 

-if needed, the design and analysis for the creation of a temporary containment opening for Pressurizer replacement and its restoration - design reviews should be coordinated with the Mechanical and Civil Engineering Branch, NRR, via the Project Manager, to confirm the restored design margin - see Appendix B of this procedure for additional information

-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

-the effect of the Pressurizer and related design changes on  transient and accident analyses

-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 code requirements 

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

-procurement specifications including qualification and functional requirements 

-confirmation that the new Pressurizer vessel was properly fabricated

b. Several heavy lifting evolutions are 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 licensee's 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 limits and 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," and NUREG-0612, "Control of Heavy Loads at Nuclear Power Plants."

c. Specific guidance for radiation protection inspections is provided in IP 71121.01 and IP 71121.02.

The inspector should audit the licensee's 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 period.  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," for additional information on radiological considerations.

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

e. Licensee risk assessment and risk management actions for the PRP should be reviewed using IP 71111.13.  In addition, licensee outage activities and actions to reduce shutdown risk should be reviewed in accordance with IP 71111.20. 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.

03.03Pressurizer Removal and Replacement Inspections

a. 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.

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.  

For the selected welds review weld procedures and welder qualification records, confirm that the Code required essential and supplemental essential welding variables for the welding processes used were met, and verify that preservice NDE requirements were completed.  Also, review the non-conformance reports for each major Pressurizer replacement weld to confirm that welding deficiencies were dispositioned in accordance with Code requirements.

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

IP71111.05AQ should be used to evaluate the licensee(s fire protection program from design, operational status, and material condition points of view. 

f. If the retired 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."

03.04Post-installation Verification and Testing Inspections.  The inspection should review the licensee's 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 addition to reviewing the licensee's 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 leakage tests and visual examinations and review of the test results should be performed.  The inspection should verify that testing satisfies ASME Code 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 contains additional inspection guidance for testing inspections.

50003-04INSPECTION RESOURCES

The required number of direct inspection hours per plant will vary significantly based on a number of 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 in order 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 hours at a single unit station and 110 hours 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.  

50003-05REFERENCES

Generic Letter 81-07, "Control of Heavy Loads" (Microfiche locations 7964/024, 7900/310, and 90191/292)

NUREG-0612 "Control of Heavy Loads at Nuclear Power Plants" (Microfiche location 6411/280)

Generic Letter 81-38, "Storage of Low-Level Radioactive Wastes at Power Reactor Sites" (Microfiche location 10848-343)

Regulatory Guide 1.136, Revision 2, "Materials, Construction, and Testing of Concrete Containments " (Microfiche location 09162-003)

NRC Regulatory Issue Summary 2004-16: Use of Later Editions and Addenda to ASME Code Section XI for Repair/Replacement Activities

END

Appendices:

A.Applicable Inspection Procedures

B.Temporary Containment Opening Review Guidance

APPENDIX A

APPLICABLE INSPECTION PROCEDURES

Inspection 

Procedure No.Inspection Procedure Title

71111.05AQFire Protection Annual/Quarterly

71111.08Inservice Inspection Activities

71111.13Maintenance Risk Assessment and Emergent Work Control

71111.17TEvaluation of Changes, Tests and Experiments

71111.18Plant Modifications

71111.20Refueling and Other Outage Activities

71111.21MDesign Bases Assurance Inspection (Teams)

71121.01Access Control to Radiologically Significant Areas

71121.02ALARA Planning and Controls

55050Nuclear Welding General Inspection Procedure

55100Structural Welding General Inspection Procedure

55150Weld Verification Checklist

57050Nondestructive Examination Procedure Visual Examination Procedure Review/Work Observation/Record Review

57060Nondestructive Examination Procedure Liquid Penetrant Examination Procedure Review/Work Observation/Record Review

57070Nondestructive Examination Procedure Magnetic Particle Examination Procedure Review/Work Observation/Record Review

57080Nondestructive Examination Procedure Ultrasonic Examination Procedure Review/Work Observation/Record Review

57090Nondestructive Examination Procedure Radiographic Examination Procedure Review/Work Observation/Record Review

70370Testing Piping Support and Restraint Systems

81064Compensatory Measures

93803Safety Systems Outage Modifications Inspection

Containment Inspections**

70307Containment Integrated Leak Rate Test ‐ Procedure Review

70313Containment Integrated Leak Rate Test

70323Containment Leak Rate Test Results Evaluation

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

END

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 Pressurizer 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.  On the basis of 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 structural integrity testing is performed.  Cutting the temporary opening and closing it, may be carried out without NRC's approval under the provisions of 10 CFR 50.59, provided the process does not meet any of the criteria in paragraph (C)(2) of the 10 CFR 50.59 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 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.

NRC inspection emphasis should be on verification that the containment is restored to its original integrity after closure of the temporary opening and that the processes used do not degrade the remaining liner to concrete interface.  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:

a. The steel liner plate 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 dependant 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.

b. 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 as a result of prestressing and may cause liner buckling.

c. The creep and relaxation of reused tendons could cause the loss of prestress as a result of tensioning and detensioning.

d. The leak tightness of the tendon sheathing may be compromised as a result 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.

e. The location of reinforcing bar splices at one section without staggering may compromise the strength of the containment.  The use of sister splice samples for testing instead of production splices may not reflect the true quality of the production splices.

f. After closing the opening, the containment is to be subjected to a structural integrity test (SIT).  The test pressure is usually 1.15 times the design pressure. This test may result in wider cracks in the concrete shell, which may allow migration of moisture to steel components embedded therein, resulting in potential corrosion.

DESIGN REVIEWS

Design reviews should be coordinated with the Mechanical and Civil Engineering  Branch, NRR, via the Project Manager, to confirm the restored design margin.  

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 final safety analysis report (FSAR), consistent with current regulations, as the basis of the facility operating license are to be utilized.

Or, and preferably:

(2) To the extent practical, the latest NRC approved for use (i.e. 10 CFR 50.55a) American Society of Mechanical Engineers 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 Mechanical and Civil Engineering Branch, NRR, 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.

Stress analyses of the containment structure are to 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 should be the same as those specified in the FSAR or the applicable Standard Review Plan (SRP) 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.

Design Changes

The materials and replacement components used to restore the containment should be the same as 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

On the basis of the regulatory concerns previously described, the areas of inspection interest can be summarized as follows:

a. 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 welds may not be the same strength as the original uncut liner, but it should be very similar to the strength of the weld joints in the original structure.

b. 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 is similar to 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.  Also, the certification of materials used in the concrete containment, such as aggregate, cement, water, and admixtures are of interest.  

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.

c. 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. 

d. 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.

e. 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.

f. Witnessing of the structural integrity test (SIT) and 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 SIT and 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 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 Civil Engineering and Geosciences Branch, NRR, 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

ATTACHMENT 1

Revision History for IP 50003

Issue Date:  09/12/17Att1 - 150003

(2(

50003Issue Date: 07/07/05

Issue Date:  09/12/17150003