ML22307A312

Evaluation of the Proposed Change

ML22307A312
Person / Time
Site:	Callaway
Issue date:	11/03/2022
From:	Ameren Missouri, Union Electric Co
To:	Office of Nuclear Reactor Regulation
Shared Package
ML22307A310	List: ML22307A312 ML22307A313 ML22307A314 ML22307A315 ULNRC-06765, License Amendment Request to Revise Technical Specification 5.5.16, Containment Leakage Rate Testing Program, for Permanent Extension of Type a and Type C Leak Rate Testing Frequencies (LDCN 2020-0004)
References
ULNRC-06765
Download: ML22307A312 (1)

Text

Enclosure Evaluation of the Proposed Change (117 pages follow this cover sheet)

Enclosure to ULNRC-06765 Page 1 of 118

Enclosure Evaluation of the Proposed Change

SUBJECT:

License Amendment Request to Revise Callaway Plant, Unit 1 Technical Specification 5.5.16, "Containment Leakage Rate Testing Program," for Permanent Extension of Type A and Type C Leak Rate Test Frequencies 1.0

SUMMARY

DESCRIPTION 2.0 DETAILED DESCRIPTION

4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory Requirements/Criteria 4.2 Precedent 4.3 No Significant Hazards Consideration 4.4 Conclusion

6.0 REFERENCES

Attachments:

1.

Evaluation of Risk Significance of Permanent ILRT Extension

2.

Proposed Technical Specification 5.5.16 Changes (Markup)

3.

Proposed Technical Specification 5.5.16 Changes (Clean Pages)

Enclosure Evaluation of the Proposed Change 2

1.0

SUMMARY

DESCRIPTION In accordance with 10 CFR 50.90, "Application for amendment of license, construction permit, or early site permit," Union Electric Company (d.b.a. Ameren Missouri) requests an amendment to Renewed Facility Operating License Number NPF-30 for Callaway Plant, Unit No. 1 (Callaway).

The proposed change revises Unit 1 Technical Specifications (TS) Chapter 5.0, Administrative Controls, Section 5.5.16, "Containment Leakage Rate Testing Program," to reflect the following:

x Increase the existing Type A integrated leakage rate test (ILRT) program test interval from 10 years to 15 years in accordance with Nuclear Energy Institute (NEI) Topical Report (TR) NEI 94-01, "Industry Guideline for Implementing Performance-Based Option of 10 CFR 50, Appendix J," Revision 3-A (Reference 2), and the conditions and limitations specified in NEI 94-01, Revision 2-A (Reference 8).

x Adopt an extension of the containment isolation valve (CIV) leakage rate testing (Type C) frequency from the 60 months currently permitted by 10 CFR 50, Appendix J, "Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors," Option B, to a 75-month frequency for Type C leakage rate testing of selected components, in accordance with NEI 94-01, Revision 3-A.

x Adopt the use of American National Standards Institute/American Nuclear Society (ANSI/ANS) 56.8-2002, Containment System Leakage Testing Requirements (Reference 37).

x Adopt a more conservative allowable test interval extension of nine months, for Type A, Type B and Type C leakage rate tests in accordance with NEI 94-01, Revision 3-A.

Specifically, the proposed change contained herein revises Callaway Unit 1 TS 5.5.16 paragraph

a. by replacing the references to Regulatory Guide (RG) 1.163, "Performance-Based Containment Leak-Test Program," (Reference 1) with a reference to NEI 94-01, Revision 3-A, and the conditions and limitations specified in NEI 94-01, Revision 2-A, dated October 2008.

These new documents will be used by Callaway to implement the performance-based leakage testing program in accordance with Option B of 10 CFR 50, Appendix J, Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors.

This License Amendment Request (LAR) also proposes the following administrative changes to TS 5.5.16, paragraph a.:

x Delete the exception from post-modification integrated leakage rate testing (ILRT) associated with steam generator (SG) replacement (SGR) since the ILRT following SGR was performed on October 14, 2014.

Enclosure Evaluation of the Proposed Change 3

x Delete the information regarding the performance of the next Callaway Type A test, as this date has already occurred and the associated Type A test has been performed.

2.0 DETAILED DESCRIPTION Callaway TS 5.5.16, "Containment Leakage Rate Testing Program," paragraph a., currently states, in part:

a.

A program shall be established to implement the leakage rate testing of the containment as required by 10 CFR 50.54(o) and 10 CFR 50, Appendix J, Option B, as modified by approved exemptions. This program shall be in accordance with the guidelines contained in Regulatory Guide 1.163, Performance-Based Containment Leak-Test Program, dated September 1995, as modified by the following exceptions:

1.

The visual examination of containment concrete surfaces intended to fulfill the requirements of 10 CFR 50, Appendix J, Option B testing, will be performed in accordance with the requirements of and frequency specified by ASME Section XI Code, Subsection IWL, except where relief has been authorized by the NRC.

2.

The visual examination of the steel liner plate inside containment intended to fulfill the requirements of 10 CFR 50, Appendix J, Option B testing, will be performed in accordance with the requirements of and frequency specified by ASME Section XI Code, Subsection IWE, except where relief has been authorized by the NRC.

3.

The unit is excepted from post-modification integrated leakage rate testing requirements associated with steam generator replacement during the Refuel 14 outage (fall of 2005).

4.

The first Type A test performed after the October 26, 1999 Type A test shall be performed no later than October 25, 2014.

The proposed changes to Callaway TS 5.5.16, paragraph a. will replace the reference to RG 1.163 with a reference to NEI TR NEI 94-01, Revisions 2-A and 3-A.

Additionally, this LAR incorporates the following administrative changes to TS 5.5.16.a.:

x Delete TS 5.5.16.a. exceptions 3 and 4. These exceptions are no longer applicable as the SGR was completed during the fall of 2005 and the associated Type A test was performed on October 10, 2014. The exception from post-modification ILRT associated with SGR was previously approved by the Nuclear Regulatory Commission (NRC) in TS

Enclosure Evaluation of the Proposed Change 4

Amendment 168. The performance of the Type A test no later than October 25, 2014, was previously approved by the NRC in TS Amendment No. 195.

The proposed change revises the Callaway TS 5.5.16, paragraph a., to read as follows (with recommended changes using strike-out for deleted text and bold-type for added text for clarification purposes):

a.

A program shall be established to implement the leakage rate testing of the containment as required by 10 CFR 50.54(o) and 10 CFR 50, Appendix J, Option B, as modified by approved exemptions. This program shall be in accordance with the guidelines contained in Regulatory Guide 1.163, Performance-Based Containment Leak-Test Program, dated September 1995, Nuclear Energy Institute (NEI) Topical Report (TR) NEI 94-01, "Industry Guideline for Implementing Performance-Based Option of 10 CFR 50, Appendix J," Revision 3-A, dated July 2012, and the conditions and limitations specified in NEI 94-01, Revision 2-A, dated October 2008, as modified by the following exceptions:

1.

The visual examination of containment concrete surfaces intended to fulfill the requirements of 10 CFR 50, Appendix J, Option B testing, will be performed in accordance with the requirements of and frequency specified by ASME Section XI Code, Subsection IWL, except where relief has been authorized by the NRC.

2.

The visual examination of the steel liner plate inside containment intended to fulfill the requirements of 10 CFR 50, Appendix J, Option B testing, will be performed in accordance with the requirements of and frequency specified by ASME Section XI Code, Subsection IWE, except where relief has been authorized by the NRC.

3.

The unit is excepted from post-modification integrated leakage rate testing requirements associated with steam generator replacement during the Refuel 14 outage (fall of 2005).

4.

The first Type A test performed after the October 26, 1999 Type A test shall be performed no later than October 25, 2014.

Therefore, the retyped ("clean") version of TS 5.5.16, paragraph a., will appear as follows:

a.

A program shall be established to implement the leakage rate testing of the containment as required by 10 CFR 50.54(o) and 10 CFR 50, Appendix J, Option B, as modified by approved exemptions. This program shall be in accordance with the guidelines contained in Nuclear Energy Institute (NEI) Topical Report (TR) NEI 94-01, "Industry Guideline for Implementing Performance-Based Option of 10 CFR 50, Appendix J," Revision 3-A, dated July 2012, and the conditions and limitations specified in NEI 94-01, Revision 2-A, dated October 2008, as modified by the following exceptions:

Enclosure Evaluation of the Proposed Change 5

1.

The visual examination of containment concrete surfaces intended to fulfill the requirements of 10 CFR 50, Appendix J, Option B testing, will be performed in accordance with the requirements of and frequency specified by ASME Section XI Code, Subsection IWL, except where relief has been authorized by the NRC.

2.

The visual examination of the steel liner plate inside containment intended to fulfill the requirements of 10 CFR 50, Appendix J, Option B testing, will be performed in accordance with the requirements of and frequency specified by ASME Section XI Code, Subsection IWE, except where relief has been authorized by the NRC.

The marked-up TS pages 5.0-19 and 5.0-20 for Callaway TS 5.5.16.a., showing these proposed changes, are provided in Attachment 2.

The clean retyped pages 5.0-19 and 5.0-20 for Callaway TS 5.5.16.a. are provided in.

contains the plant specific risk assessment conducted to support this proposed change. This risk assessment follows the guidelines of NRC RG 1.174, Revision 3 (Reference 3) and RG 1.200, Revision 3 (Reference 17). The risk assessment concludes that increasing the ILRT test frequency on a permanent basis to a one-in-fifteen-year frequency is considered to represent a small change in the Callaway risk profile.

3.0 TECHNICAL EVALUATION

3.1 Primary Containment System 3.1.1 Description of the Reactor Building The reactor building consists of a prestressed, reinforced concrete, cylindrical structure with a hemispherical dome and a conventionally reinforced concrete base slab with a central cavity and instrumentation tunnel to house the reactor vessel. A continuous peripheral tendon access gallery below the base slab is provided for the installation and inspection of the vertical post-tensioning system. The internal structures are isolated from the shell by means of an isolation gap to minimize interaction. In addition, the connections used to provide for vertical support of the structural steel floor framing at the shell allow for independent horizontal movement. The shell is separated from its surrounding structures by a minimum 3-inch isolation gap to avoid interaction. In some instances, the gap is filled with a fireproof compressible material.

The base slab, cylinder, and dome are reinforced by bonded reinforcing steel, as required by the design loading conditions. Additional reinforcing is provided at discontinuities in the structure and at major penetrations in the shell.

Enclosure Evaluation of the Proposed Change 6

The interior of the reactor building is lined with carbon steel plates welded together to form a barrier which is essentially leak tight. A post-tensioning system is used to prestress the cylindrical shell and dome.

Principal nominal dimensions of the reactor building are as follows:

x Interior diameter 140 ft x

Interior height 205 ft x

Height to spring line 135 ft x

Base slab thickness 10 ft x

Cylinder wall thickness 4 ft x

Dome thickness 3 ft x

Liner plate thickness 0.25 in.

x Internal free volume 2.5 x 106 cubic ft 3.1.2 Post-Tensioning System The tendon system is employed to post-tension the cylindrical shell and dome of the reactor building. The system uses unbonded tendons, each consisting of approximately 170 one-quarter-inch-diameter high strength steel wires and anchorage components consisting of stressing washers. The prestressing load is transferred by cold-formed button heads on the ends of the individual wires, through stressing washers, to the steel bearing plates embedded in the structure.

The ultimate strength of each tendon is approximately 1,000 tons.

The unbonded tendons are installed in tendon ducts (sheathing) and tensioned in a predetermined sequence. The ducts, which form voids through the concrete between the anchorage points, consist of galvanized, spiral-wrapped, semirigid corrugated steel tubing. They are designed to retain their shape and resist the construction loads. The inside diameter of the ducts is sufficiently large to permit the installation of the tendons with minimum difficulty.

Trumpets, which are enlarged ducts attached to the bearing plates, allow the wires to spread out at the anchorage to suit washer hole spacing and facilitate field cold formed button heading of the ends of the wires.

The tendon duct provides an enclosed space surrounding each tendon. After stressing, a petroleum-based corrosion inhibitor is pumped into the duct.

The vertical tendons consist of 86 inverted U-shaped tendons, which extend through the full height of the cylindrical wall over the dome and are anchored at the bottom of the base slab. The cylinder circumferential (hoop) tendons consist of 135 tendons anchored at three buttresses equally spaced around the outside of the reactor building. Each tendon is anchored at buttresses located 240 degrees apart. Three adjacent tendons, anchored at alternate buttresses, result in two complete hoop tendons.

Enclosure Evaluation of the Proposed Change 7

Prestressing of the hemispherical dome is achieved by a two-way pattern of the inverted U-shaped tendons and 30 hoop tendons, which start at the springline and continue up to an approximate 45-degree vertical angle from the springline.

3.1.3 Liner Plate System A carbon steel liner plate covers the entire inside surface of the reactor building (excluding penetrations). The liner is 1/4-inch thick but is thickened locally around the penetrations, large brackets, and major attachments. The liner plate, including the thickened plate, is anchored to the concrete structure. The vertical and dome liner plates are also used as forms for concrete placement. In addition to the carbon steel liner plate, the containment normal sumps have an additional 1/4" stainless steel liner plate installed over the top of the carbon steel liner plate for corrosion protection.

Attachments to the liner plate which transfer loads through the liner plate to the base slab include equipment support anchors and reinforcing steel for the support of the internal structures.

Major structural attachments to the wall which penetrate the liner plate include polar crane brackets, floor beam brackets, and pipe support brackets.

Major structural attachments to the dome include various pipe support brackets.

Miscellaneous thickened plates, which form a part of the liner plate, are provided and anchored in the concrete to provide supports. Leak chase channels and angles are also attached at seam welds where the welds are inaccessible to nondestructive examination after construction.

3.1.4 Equipment and Personnel Access Hatches and Penetration Sleeves The equipment hatch is a welded steel assembly with a double-gasketed, flanged, and bolted cover. Provision is made for leak testing of the flange gasket combination by pressurizing the space between the gaskets.

One personnel hatch and one auxiliary hatch, both of which are welded steel assemblies, are provided. Each hatch has two doors with double gaskets in series. In order to assure leaktightness, provision is made to pressurize the space between the gaskets. The doors are mechanically interlocked to ensure that one door cannot be opened unless the second door is sealed. Provisions are made for deliberately overriding the interlock by the use of special tools and procedures. Each door is equipped with quick-acting valves for equalizing the pressure across the doors. The doors are not operable unless the pressure is equalized. Pressure equalization is possible from every point at which the associated door can be operated. The valves for the two doors are properly interlocked so that only one valve can be opened at one time and only when the opposite door is closed and sealed. Each door is designed so that, with the other door open, it will withstand and seal against design and testing pressure of the containment vessel. There is visual indication outside each door showing whether the opposite door is open or closed. Provision is made outside each door for remotely closing and latching the

Enclosure Evaluation of the Proposed Change 8

opposite door so that in the event that one door is accidentally left open it can be closed by remote control. The access hatch barrels have nozzles which permit pressure testing of the hatch at any time. The hatches are protected from tornado missiles by enclosure structures or shields. A moveable missile shield is provided on the outside of the reactor building to protect the equipment hatch. The personnel hatch is enclosed within the auxiliary building. The auxiliary hatch is enclosed within an exterior tornado-resistant concrete structure.

The personnel and auxiliary access hatch barrels are designated as American Society of Mechanical Engineers (ASME)Section III, Class MC components.

The hatch penetration sleeves project into the reactor building and are used to support the hatches. These items are made from carbon steels and conform to the requirements of ASME Section III, Subsection NE.

3.1.5 Piping Penetration Sleeves Piping penetrations are divided into three general groups:

a.

Type 1: Flued head penetrations used for most high energy piping.

Examples of Type 1 penetrations are the main steam and main feedwater (MFW) lines.

b.

Type 2: Closure plate penetrations used for some high-energy, all moderate energy, and all low-energy general piping. The use of this type of penetration for high energy piping is limited to only those cases where an analysis based on combination of pressure, temperature, and line size has demonstrated the adequacy of the design.

c.

Type 3: Spare penetrations reserved for future use.

Type 1 piping penetrations consist of the following major steel items:

a.

Process Pipe: This pipe, which is made of welded or seamless carbon or stainless steel and is welded to the flued head, conforms to the requirements of ASME Section III, Subsection NC.

b.

Flued Head: This item is made from forged carbon or stainless steel and conforms to the requirements of ASME Section III, Subsection NC. It is designed to contain the full pressure of the process fluid and full reactor building pressure in parts adjoining the pipe sleeve. The connecting process pipes and the flued heads are designed and analyzed to be capable of carrying loads resulting from the failure of the process pipe.

c.

Pipe Sleeve: This steel item consists of the portion which projects into the reactor building and supports the flued head. It conforms to ASME Section III, Subsection NE, except that authorized inspection and stamping are not performed.

Enclosure Evaluation of the Proposed Change 9

Type 2 piping penetrations consist of the following major steel items:

a.

Process Pipe: This pipe, which is made of welded or seamless carbon or stainless steel and is welded to the closure plate, conforms to the applicable requirements of ASME Section III, Subsection NC.

b.

Closure Plate: This item is made from carbon or stainless-steel plate and conforms to the requirements of ASME Section III, Subsection NC.

c.

Pipe Sleeve: This steel item consists of the portion which projects into the reactor building and supports the closure plate. It conforms to ASME Section III, Subsection NE, except that authorized inspection and stamping are not performed.

Type 3 spare penetrations consist of the following major items:

a.

Solid Closure Plate of Pipe Cap: This item is made from carbon steel and conforms to the requirements of ASME Section III, Subsection NC.

b.

Pipe Sleeve: This steel item consists of the portion which projects into the reactor building. It conforms to ASME Section III, Subsection NE, except that authorized inspection and stamping are not performed.

3.1.6 Fuel Transfer Tube Penetration Sleeve The fuel transfer tube penetration is provided to transfer fuel between the refueling canal and the spent fuel pool during refueling operations of the reactor. The penetration consists of a 20-inch-diameter stainless steel pipe installed inside a 26-inch sleeve. The steel sleeve which projects into the reactor building conforms to ASME Section III, Subsection NE, except that authorized inspection and stamping are not performed. The inner pipe acts as the transfer tube. The sleeve is designed to provide integrity of the reactor building, allow for differential movement between structures, and prevent leakage through the fuel transfer tube in the event of an accident.

3.1.7 Electrical Penetration Sleeves Steel sleeves, which form a portion of the containment pressure boundary, are provided for electrical penetrations. The sleeve consists of the portion which projects out of the reactor building and supports the electrical assembly. It conforms to ASME Section III, Subsection NE, except that authorized inspection and stamping are not performed.

3.1.8 Purge Line Penetration Sleeves The steel sleeves, which are embedded in the reactor building wall concrete, are welded to the purge line piping and form a part of the ASME Section III, Class 2 purge line piping system. The

Enclosure Evaluation of the Proposed Change 10 sleeves conform to ASME Section III, Subsection NC.

3.1.9 Containment Isolation System The containment isolation system allows the normal or emergency passage of fluids through the containment boundary while preserving the ability of the boundary to minimize the release of fission products following a loss of coolant accident (LOCA) or fuel handling accident within the containment.

Containment Isolation System safety design bases criteria are as follows:

1.

The containment isolation system is protected from the effects of natural phenomena, such as earthquakes, tornadoes, hurricanes, floods, and external missiles (General Design Criteria (GDC)-2).

2.

The containment isolation system is designed to remain functional after a safe shutdown earthquake and to perform its intended function following the postulated hazards of fire, internal missiles, or pipe breaks (GDC-3 and 4).

3.

The containment isolation system is designed and fabricated to codes consistent with the quality group classification assigned by Regulatory Guide (RG) 1.26 and the seismic category assigned by RG 1.29. The power supply and control functions are in accordance with RG 1.32.

4.

Piping systems penetrating the primary reactor containment are provided with leak detection, isolation, and containment capabilities having redundancy, reliability, and performance capabilities which reflect the importance to safety of isolating these piping systems. Such piping systems are designed with a capability to periodically test the operability of the isolation valves and associated apparatus and to determine if valve leakage is within acceptable limits (GDC-54).

5.

Each line that is part of the reactor coolant pressure boundary and that penetrates the primary reactor containment is provided with containment isolation valves as follows:

a.

One locked closed isolation valve inside and one locked closed isolation valve outside the containment; or

b.

One automatic isolation valve inside and one locked closed isolation valve outside the containment; or

c.

One locked closed isolation valve inside and one automatic isolation valve outside the containment. A simple check valve is not used as the automatic isolation valve outside the containment; or

d.

One automatic isolation valve inside and one automatic isolation valve outside the

Enclosure Evaluation of the Proposed Change 11 containment. A simple check valve is not used as the automatic isolation valve outside the containment; or

e.

Some other defined bases that meet the intent of containment isolation as an alternative to a through d above.

Isolation valves outside the containment are located as close to the containment as practical and, upon loss of actuating power, automatic isolation valves are designed to take the position that provides the greater safety (GDC-55).

6.

Each line that connects directly to the containment atmosphere and penetrates the primary reactor containment is provided with containment isolation valves as follows:

a.

One locked closed isolation valve inside and one locked closed isolation valve outside the containment; or

b.

One automatic isolation valve inside and one locked closed isolation valve outside the containment; or

c.

One locked closed isolation valve inside and one automatic isolation valve outside the containment. A simple check valve is not used as the automatic isolation valve outside the containment; or

d.

One automatic isolation valve inside and one automatic isolation valve outside the containment. A simple check valve is not used as the automatic isolation valve outside the containment; or

e.

Some other defined bases that meet the intent of containment isolation, as an alternative to a through d above.

Isolation valves outside the containment are located as close to the containment as practical and, upon loss of actuating power, automatic isolation valves are designed to take the position that provides greater safety (GDC-56).

7.

Each line that penetrates the primary reactor containment and is neither part of the reactor coolant pressure boundary nor connected directly to the containment atmosphere has:

a.

At least one containment isolation valve which is either automatic, locked closed, or capable of remote manual operation; or

b.

Some other defined bases that meet the intent of containment isolation, as an alternative to a above.

Enclosure Evaluation of the Proposed Change 12 Valves are outside the containment and located as close to the containment as practical.

A simple check valve is not used as the automatic isolation valve. For a closed system, the design is commensurate with quality group B (GDC-57).

8.

The containment isolation system, in conjunction with other plant features, serves to minimize the release of fission products generated following a LOCA or fuel handling accident within the containment.

3.1.10 Major Modifications to the Containment Structure Equipment Hatch Platform and Missile Shield Modification Callaway installed Framatome-designed Model 73/19T replacement SGs during Refuel 14 in fall of 2005.

Missile Shield Permanent Modification The Equipment Hatch Missile Shield was modified such that it may be opened and closed without having to move any pieces of the new equipment hatch platform. This required that the bottom portion of the Equipment Hatch Missile Shield be cut off and thus its lower seismic restraint system modified. Furthermore, because the Missile Shield would no longer extend all the way to the bottom of the equipment hatch barrel, a new concrete 'smile piece' was poured in the bottom of the barrel to provide missile protection for the lower portion of the equipment hatch. This section of reinforced concrete also provided a flat surface for transfer of materials in and out of the containment building. Due to this addition, the existing steel filler piece for the equipment hatch barrel was eliminated.

The Equipment Hatch Missile Shield provides protection for the equipment hatch from objects that could become missiles during a windstorm (tornado). It is a large concrete door that is supported from above on a track system. This track allows it to be moved 19 degrees from its closed position (azimuth 128 degrees) to its open position (azimuth 109 degrees) thus providing access to the equipment hatch for moving items into and out of the containment building.

The Missile Shield is designed as a Safety-Related, Seismic Category I structure and must therefore resist horizontal and vertical accelerations associated with an SSE. In its closed position, the seismic restraints on the lower portion of the missile shield include side bearing lugs and front bumpers that transfer the load from the missile shield to steel beams that are embedded in the containment shell at approximate elevation 2044-4". The seismic restraints on the upper portion of the missile shield in the closed position include side bearing lugs and front bumpers that transfer the load from the missile shield to the steel beams that are embedded in the containment shell at approximate elevation 2073'-5". The dead weight of the missile shield is carried by the overhead track system. The missile shield is not currently qualified to withstand seismic loads when it is in the open position.

Enclosure Evaluation of the Proposed Change 13 The missile shield was revised to allow closure without any sections having to be removed from the new Equipment Hatch Platform. The shield had 2'-9" of concrete and steel cut off the bottom such that the shield may be closed with no disruption to the equipment hatch platform. At the location where the missile shield concrete is cut, three steel plates will be anchored to the missile shield: one on the bottom and one on the front and back faces extending approximately 9.5" up from the bottom of the missile shield. This configuration ensures adequate edge strength and ductility in the event that a tornado missile should strike the missile shield along its lower edge.

Because the lower 2'-9" was being cut off, the existing seismic restraint system for the lower portion of the missile shield was revised to accommodate the new configuration of the missile shield. The two existing lower side bearing lugs were trimmed, however, they function in the same manner as original design. The bottom pins as well as the front bumpers at the base of the missile shield were eliminated. In their place, a side pin system was added. This pin system will provide seismic restraint for the missile shield in two horizontal directions.

The pins that attach directly to the missile shield are approximately 3'-6" higher than the existing embedded W33 beams at approximate elevation 2044'-4". Therefore, a beam section made up of a W24 welded to a W24 'T' section was connected to the W33 embeds. This beam provided the anchor points for the new seismic restraint pins on the missile shield.

There was an existing rolling guide system at the base of the missile shield which served as an aid for missile shield movement activities. This system, however, would not come into contact with the missile shield once it was cut off. The rolling guide did not provide structural support for the missile shield and was eliminated.

Permanent Equipment Hatch Modifications The existing hatch barrel steel filler piece was permanently cut out and disposed of as required to facilitate the installation of a new 'smile piece' of reinforced concrete with a top elevation of 2047'-

101/2" that will be placed at the bottom of the equipment hatch barrel.

This new section of reinforced concrete will provide tornado missile protection for the steel equipment hatch. After the permanent modifications were made to the missile shield (cutting off the lower 2'-9"), the lower portion of the equipment hatch would no longer be protected by the Equipment Hatch Missile Shield and is instead protected by the equipment hatch smile piece.

Between the bottom of the permanently modified Equipment Hatch Missile Shield and the top of the permanent concrete smile piece in the equipment hatch barrel there is a 3/4" straight line gap.

This had no implications for the ability of this combined system to provide adequate missile protection to the equipment hatch.

In addition to the aforementioned function, this concrete provided a flat surface with sufficient bearing capacity for the temporary Hatch Transfer System beams that were used to move the Original SGs out of containment and the Replacement SGs into containment during the SGR Outage.

Enclosure Evaluation of the Proposed Change 14 The concrete 'smile piece' was fitted with permanent rails that connected with the outside rails on the new equipment hatch platform and the inside rails on the operating deck concrete inside the containment building. These rails are spaced at 5'-01/2" and run 15 degrees off of the centerline of the equipment hatch such that they line up with the existing embedded plates on the operating deck floor. The rails are 60# ASCE and will bolt into an embedded channel that is recessed into the smile piece concrete such that the top of the rail is also at elevation 2047'-101/2". Because the rail was recessed, there was no interference between the rail and the temporary Hatch Transfer System beams.

The Missile Shield and Equipment Hatch 'smile piece' protect the equipment hatch from tornado missiles and are therefore designed as Safety-Related, Seismic Category I structures. All new components that contribute to the missile shielding capability of the smile piece or of the missile shield will be designed as Safety-Related, Seismic Category I components.

Any components such as the rail that is embedded in the smile piece or steel plates that function as integral formwork for the smile piece are classified as non-safety related.

Equipment Hatch Platform The existing Equipment Hatch Platform was removed in its entirety and a new Equipment Hatch Platform was erected in its place. A new elevator was erected next to this platform as well as permanent stairway access. The existing pedestal crane was permanently eliminated.

The new Equipment Hatch Platform was fitted with permanent rails that may be used to transfer materials on a cart into and out of the containment building. Inside the containment building, removable rails were added to the existing embedded plates on the operating deck at elevation 2047'-6".

The new Equipment Hatch platform has a top of grating elevation of 2047'-101/2" which is the same as the new concrete 'smile piece' that was installed in the equipment hatch barrel. The new permanent rail system was recessed below the platform grating and through the 'smile piece' concrete such that the rails also have a top elevation of 2047'-101/2". Inside on the operating deck the concrete floor has an elevation of 2047'-6". Its rails, however, sit atop the concrete and therefore have a top elevation of 2047'-6".

The hatch transfer system was installed to move SGs into and out of containment and was designed to compensate for the change in elevation between the operating deck and the equipment hatch barrel/outside platform.

Addition of Removable Rails Inside Containment A new rail system was installed to facilitate the transfer of materials into and out of the containment building. These rails are ASCE 60# rails on a 5'-0 1/2" center to center and run 15° off the centerline hatch 128° azimuth continuous into the steel equipment hatch. The rails bolt on

Enclosure Evaluation of the Proposed Change 15 to the existing plates EP 124 and EP 125 that are embedded in the concrete floor at elevation 2047'-6" inside the containment building. The top elevation of the rails are 2047'-101/2" and they are secured to the embedded steel plates with bolted floating clamps 3 feet apart. Crane rails stops are provided at the ends of the rails to match the cart wheels and bumpers.

The existing section of removable grating that is routinely removed to allow equipment hatch closure was also modified. This section of grating was modified to incorporate two removable 12" x 11/4" steel plates with ASCE 60# rails such that the rail system runs continuous from the outside equipment hatch platform to the inside operating deck in containment. These short rail sections were fitted with rail splices with quick-disconnect pins such that they may be disconnected from the rail section inside the equipment hatch barrel and the rail section on the operating deck to allow expedient removal to facilitate hatch closure. The weight of each plate/rail section is approximately 275 lbs. They can therefore be easily loaded onto a cart and moved out of the way.

The removable short rail sections are stored during normal operations on elevation 2047'-6" just north of the equipment hatch on the concrete floor of the operating deck. These are tied to the equipment hatch lift and guide beams bottom bracket using 1/4" diameter stainless steel wire rope and cable clamps such that they do not become a hazard during a seismic event. The qualification of this tie-down configuration is addressed in a calculation.

Appropriate signage is posted on these removable rail sections to notify operations that these items may remain inside containment during normal plant operation.

Equipment Hatch Platform Modification The existing equipment hatch platform was removed in its entirety and replaced with a new equipment hatch platform that is sufficiently stout to carry the loads associated with SGR. The existing pedestal crane that is located on the existing equipment hatch platform was also eliminated. Due to the modification of the equipment hatch missile shield and the redesign of the equipment hatch platform, is no longer necessary to move a section of the equipment hatch platform out of the way to allow Missile Shield closure. Furthermore, with the addition of the new rail system, this crane is not required to move materials into and out of the containment building during future outages.

Equipment Hatch Platform Elevator A new elevator was erected next to the new equipment hatch platform to facilitate movement of personnel and equipment from the ground level at elevation 2000' to the top of the platform at elevation 2047'-101/2". It was installed on the plant north side of the equipment hatch platform.

This elevator is a class C-1 elevator with 5-ton (10,000 lbs.) capacity. The elevator is a rack and pinion elevator designed for rugged duty with a welded steel structural frame, wire mesh side and a gate door. The interior car dimensions are 11'-4" wide by 11'-0" long (front to rear) by 10'-0" high. It has a speed of 130 feet per minute and will travel from the ground elevation of 2000' to the top elevation of 2047'-10 1/2".

Enclosure Evaluation of the Proposed Change 16 The equipment hatch platform along with the elevator and its foundations are classified as non-safety related.

Conclusion This modification performed in support of SGR activities, had no impact on the integrity of the concrete Reactor Containment Structure or the Steel Liner system, which required post modification testing.

3.1.11 Post-Modification Integrated Leakage Rate Testing Requirements Associated with Steam Generator Replacement During Refuel 14 Outage (Fall 2005)

The NRC issued Amendment No. 168 to Facility Operating License No. NPF-30 for Callaway on September 29, 2005. (Reference 15)

The amendment consisted of changes to the TS in response to application dated September 17, 2004 (ULNRC-05056), as supplemented by the letters dated February 11 (ULNRC-05117), May 26 (ULNRC-05145), June 17 (ULNRC-05157 and ULNRC-05159), July 15 (ULNRC-05169), July 29 (ULNRC-05178), August 16 (ULNRC-05188), and September 6 (ULNRC-05192), 2005.

The amendment supported the installation of replacement SGs at Callaway during the refueling outage in the fall of 2005. The amendment changed the following affected TS: the reactor core safety limits (TS 2.1.1); reactor trip system and engineered safety feature actuation system (ESFAS) instrumentation (TS 3.3.1 and 3.3.2); reactor coolant system (RCS) limits (TS 3.4.1);

RCS loops (TS 3.4.5, 3.4.6, and 3.4.7); RCS operational leakage (TS 3.4.13); SG tube integrity (the new TS 3.4.17); main steam safety valves (MSSVs) (TS 3.7.1); SG tube surveillance program (TS 5.5.9); containment ILRT program (TS 5.5.16); and, SG tube inspection report (TS 5.6.10).

Exception to Callaway Containment ILRT Program (TS 5.5.16)

The licensee had proposed to add the following exception to the requirement in TS 5.5.16 to perform an ILRT: "The unit is excepted from post-modification integrated leakage rate testing requirements associated with steam generator replacement during Refuel 14 outage (fall 2005)."

The exception would allow the licensee to not perform an ILRT in Refuel 14 outage after the installation of the replacement SGs. The replacement SGs can be installed without the containment being cut.

During the LOCA, portions of the SGs and attached lines are relied on as a barrier against the uncontrolled release of radioactivity to the environment. The portions impacted, that are considered part of the containment boundary, are the outer shell of the SGs, the inside containment portions of lines emanating from the SG shells (i.e., the main steam lines, the MFW lines, the SG blowdown and sample lines) and the inside surface of the SG tubes.

Enclosure Evaluation of the Proposed Change 17 TS 5.5.16.a requires (1) that a program be established to implement the leakage rate testing of the containment as required by 10 CFR 50.54(o) and 10 CFR 50, Appendix J, Option B, as modified by approved exemptions and (2) that the program must be performed in accordance with NRC RG 1.163 (Reference 1) which, in turn, endorses the Nuclear Energy Institute document NEI 94-01, Revision 0 (Reference 5). NEI 94-01, Revision 0, stated that a Type A test (or ILRT) or local leak rate test (LLRT) be conducted prior to returning the containment to operation following a modification that affects containment integrity. Replacing the SGs is such a modification since, as discussed above, the replacement would affect portions of the containment boundary.

The licensee stated that performing the ILRT was unnecessary because the ASME Boiler and Pressure Vessel Code (the ASME Code) Section III/XI pressure test requirements satisfy the intent of 10 CFR Part 50, Appendix J, Option B. The NRC staff reviewed the ASME Section XI requirements and determined that the ASME Section XI surface examination, volumetric examination, and system pressure testing requirements are more stringent than the ILRT requirements of Appendix J (which are currently stated in TS 5.5.16). Although, the objective of the ILRT test is to ensure the leak tight integrity of the containment area affected by the modification, the ASME Section XI inspection and testing requirements more than fulfill the intent of the requirements of Appendix J and the provisions of NEI 94-01, Revision 0. In addition, the test pressure for the ASME Code system pressure test is significantly greater than that of the Appendix J test. Also, the replacement SGs were installed at Callaway without the containment cut. Based on this, the NRC staff concluded that the licensees proposed exception from performing a post-modification ILRT following the installation of the replacement SGs was acceptable. Based on this, the NRC staff further concluded that the containment continued to meet GDC 50 without conducting the ILRT after the installation of the replacement SGs and, therefore, the proposed change to TS 5.5.16 was acceptable.

Environmental Qualification of Equipment The licensee addressed the effect of the replacement SGs on the environmental qualification of equipment inside containment in its application. In that section on the containment pressure/temperature response associated with the SGR, the licensee stated that the new Callaway containment evaluation model was based on the NRC-approved model for Kewaunee using the GOTHIC code with most of the input data taken from the CONTEMPT LOCA and MSLB input decks. The GOTHIC model was used to produce sample results for the LOCA and main steal line break (MSLB) transients using conservative mass and energy release data that is representative of the Callaway plant. The containment temperature for the LOCA and MSLB remained less than 270oF and 352.8oF, respectively, which are less than the peak temperature listed in the Callaway FSAR. Because the peak temperatures calculated for the replacement SGs are less than the current peak temperatures for the LOCA and MSLB in the FSAR, the licensee concluded that there will be no adverse impact on the environmental qualification of equipment inside containment.

The licensee stated that it used NRC-approved methodology to determine the peak containment temperature for the LOCA and MSLB, used conservative containment conditions and the

Enclosure Evaluation of the Proposed Change 18 replacement SG parameters, and calculated peak temperatures that are less than the current peak temperature in the Callaway FSAR. Based on this, the NRC staff concluded that the environmental qualification of equipment inside containment is bounded by the current environmental qualification of such equipment in the FSAR, and therefore, the plant continues to meet 10 CFR 50.49, "Environmental qualification of electric equipment important to safety for nuclear power plants."

Conclusions In its application, the licensee addressed containment integrity for the installation of the replacement SGs and proposed a change to TS 5.5.16 to allow the plant restart after the installation without performing an ILRT. As discussed above, the NRC staff concluded that the licensee had demonstrated that containment integrity remains acceptable for plant operation and that the proposed change to TS 5.5.16 is acceptable.

3.2 Containment Overpressure on Containment Spray System (CSS) and Emergency Core Cooling System (ECCS) Performance 3.2.1 Resolution of NRC Generic Letter 2004-02 Containment accident pressure of 1.7 psi is credited for available NPSH during the Large Break LOCA (LBLOCA) phase when containment temperature is above 212°F to assure no flashing to steam occurs (at the top of the sump strainers) in the debris bed (approximately 10% of available containment pressure). In general, for a sump temperature of 212°F and higher, the NPSH available calculation assumes that the containment pressure is equal to the vapor pressure. For sump temperatures lower than 212°F, the containment pressure is assumed to be equal to atmospheric pressure of 14.7 psia, as if there is loss of containment. Overpressure credit is not needed to meet the NPSH, air release, or strainer buckling strainer performance criteria (Reference 35).

3.2.2 Containment Spray System (CSS) Residual Heat Removal (RHR) NPSH An assured water volume of 394,000 gallons is available in the RWST to ensure that, after a LOCA, sufficient water is injected for emergency core cooling and for rapidly reducing the containment pressure and temperature. In addition, this volume ensures that sufficient water is available in the containment sump to permit recirculation flow to the core and the containment and to meet the NPSH requirements of the RHR and CS pumps and assures that a sufficient water volume is available in the RWST to allow for manual switchover of the CS pumps.

In the event of an accident, the ECCS CCPs are started automatically on receipt of an SIS and are automatically aligned to take suction from the RWST during the injection phase. These high head pumps deliver flow through the boron injection header to the RCS at the prevailing RCS pressure. During the recirculation phase, suction is provided from the RHR pump discharge.

Enclosure Evaluation of the Proposed Change 19 In the event of an accident, the safety injection (SI) pumps are started automatically on receipt of an SIS; take suction from the RWST via normally open, motor-operated valves and deliver water to the RCS during the injection phase; and take suction from the containment sump via the RHR pumps during the recirculation phase.

CSS - System piping size and layout will provide adequate NPSH to the containment spray pump during all anticipated operating conditions, in accordance with RG 1.1. In calculating available NPSH, the conservative assumption has been made that the water in the containment sump after a design basis LOCA is a saturated liquid, and no credit has been taken for anticipated subcooling. That is, although NPSH = elevation head + (containment pressure - liquid vapor pressure) - suction line losses, the (containment pressure - liquid vapor pressure) term has been assumed to be zero. Calculated NPSH exceeds required NPSH by at least 10 percent. The recirculation piping penetrating the containment sumps is nearly horizontal to minimize vortexing.

In addition, a vortex breaker is provided in the inlet of the piping from the sump.

CSS Pumps:

x Available NPSH @ 3,950 gpm 22.0 ft x Required NPSH @ 3,950 gpm 16.5 ft RHR Pumps:

x Available NPSH @ 4,800 gpm 25.7 ft x Required NPSH @ 4,800 gpm 21.7 ft 3.3 Justification for the TS Change 3.3.1 Chronology of Testing Requirements of 10 CFR 50, Appendix J The testing requirements of 10 CFR 50, Appendix J, provide assurance that leakage from the containment, including systems and components that penetrate the containment, does not exceed the allowable leakage values specified in the TS. 10 CFR 50, Appendix J also ensures that periodic surveillances of reactor containment penetrations and isolation valves are performed so that proper maintenance and repairs are made during the service life of the containment and of the systems and components penetrating primary containment. The limitation on containment leakage provides assurance that the containment would perform its design function following an accident up to and including the plant DBA. Appendix J identifies three types of required tests:

1)

Type A tests, intended to measure the primary containment overall integrated leakage rate;

Enclosure Evaluation of the Proposed Change 20

2)

Type B tests, intended to detect local leaks and to measure leakage across pressure-containing or leakage-limiting boundaries (other than valves) for primary containment penetrations; and,

3)

Type C tests intended to measure containment isolation valve (CIV) leakage rates.

Types B and C tests identify the vast majority of potential containment leakage paths. Type A tests identify the overall (integrated) containment leakage rate and serve to ensure continued leakage integrity of the containment structure by evaluating those structural parts of the containment not covered by Types B and C testing.

In 1995, 10 CFR 50, Appendix J, was amended to provide a performance-based Option B for the containment leakage testing requirements. Option B requires that test intervals for Type A, Type B, and Type C testing be determined by using a performance-based approach. Performance-based test intervals are based on consideration of the operating history of the component and resulting risk from its failure. The use of the term "performance-based" in 10 CFR 50, Appendix J refers to both the performance history necessary to extend test intervals as well as to the criteria necessary to meet the requirements of Option B.

Also in 1995, RG 1.163 (Reference 1) was issued. The RG endorsed NEI 94-01, Revision 0, (Reference 5) with certain modifications and additions. Option B, in concert with RG 1.163 and NEI 94-01, Revision 0, allows licensees with a satisfactory ILRT performance history (i.e., two consecutive, successful Type A tests) to reduce the test frequency for the containment Type A (ILRT) test from three tests in 10 years to one test in 10 years. This relaxation was based on an NRC risk assessment contained in NUREG-1493, (Reference 6) and Electric Power Research Institute (EPRI) TR-104285 (Reference 7), both of which showed that the risk increase associated with extending the ILRT surveillance interval was very small. In addition to the 10-year ILRT interval, provisions for extending the test interval an additional 15 months were considered in the establishment of the intervals allowed by RG 1.163 and NEI 94-01, but that this extension of interval "should be used only in cases where refueling schedules have been changed to accommodate other factors."

In 2008, NEI 94-01, Revision 2-A (Reference 8), was issued. This document describes an acceptable approach for implementing the optional performance-based requirements of Option B to 10 CFR 50, Appendix J, subject to the limitations and conditions noted in Section 4.0 of the NRC safety evaluation (SE) report (SER) on NEI 94-01. NEI 94-01, Revision 2-A, includes provisions for extending Type A ILRT intervals to up to 15 years and incorporates the regulatory positions stated in RG 1.163 (Reference 1). The document also delineates a performance-based approach for determining Type A, Type B, and Type C containment leakage rate surveillance testing frequencies. Justification for extending test intervals is based on the performance history and risk insights.

In 2012, NEI 94-01, Revision 3-A (Reference 2), was issued. This document describes an acceptable approach for implementing the optional performance-based requirements of Option B to 10 CFR 50, Appendix J and includes provisions for extending Type A ILRT intervals to up to

Enclosure Evaluation of the Proposed Change 21 15 years. NEI 94-01 has been endorsed by RG 1.163 and NRC SERs of June 25, 2008 (Reference 9), and June 8, 2012 (Reference 10), as an acceptable methodology for complying with the provisions of Option B in 10 CFR 50, Appendix J. The regulatory positions stated in RG 1.163, as modified by References 9 and 10, are incorporated in this document. It delineates a performance-based approach for determining Type A, Type B, and Type C containment leakage rate surveillance testing frequencies. Justification for extending test intervals is based on the performance history and risk insights.

Extensions of Type B and Type C test intervals are allowed based upon completion of two consecutive periodic as-found tests where the results of each test are within a licensees allowable administrative limits. Intervals may be increased from 30 months up to a maximum of 120 months for Type B tests (except for containment airlocks) and up to a maximum of 75 months for Type C tests. If a licensee considers extended test intervals of greater than 60 months for Type B or Type C tested components, the review should include the additional considerations of as-found tests, schedule and review as described in NEI 94-01, Revision 3-A, Section 11.3.2.

The NRC has provided guidance concerning the use of test interval extensions in the deferral of ILRTs beyond the 15-year interval in NEI 94-01, Revision 2-A, NRC SER Section 3.1.1.2, which states, in part:

As noted above, Section 9.2.3, NEI TR 94-01, Revision 2, states, "Type A testing shall be performed during a period of reactor shutdown at a frequency of at least once per 15 years based on acceptable performance history." However, Section 9.1 states that the "required surveillance intervals for recommended Type A testing given in this section may be extended by up to 9 months to accommodate unforeseen emergent conditions but should not be used for routine scheduling and planning purposes." The NRC staff believes that extensions of the performance-based Type A test interval beyond the required 15 years should be infrequent and used only for compelling reasons. Therefore, if a licensee wants to use the provisions of Section 9.1 in TR NEI 94-01, Revision 2, the licensee will have to demonstrate to the NRC staff that an unforeseen emergent condition exists.

NEI 94-01, Revision 3-A, Section 10.1, Introduction, concerning the use of test interval extensions in the deferral of Type B and Type C LLRTs, based on performance, states, in part:

"Consistent with standard scheduling practices for Technical Specifications Required Surveillances, intervals of up to 120 months for the recommended surveillance frequency for Type B testing and up to 75 months for Type C testing given in this section may be extended by up to 25 percent of the test interval, not to exceed nine months.

Notes: For routine scheduling of tests at intervals over 60 months, refer to the additional requirements of Section 11.3.2.

Extensions of up to nine months (total maximum interval of 84 months for Type C tests) are permissible only for non-routine emergent conditions. This provision (nine-month extension)

Enclosure Evaluation of the Proposed Change 22 does not apply to valves that are restricted and/or limited to 30-month intervals in Section 10.2 (such as BWR MSIVs) or to valves held to the base interval (30 months) due to unsatisfactory LLRT performance."

The NRC has also provided the following concerning the extension of ILRT intervals to 15 years in NEI 94-01, Revision 3-A, NRC SER Section 4.0, Item 2:

"The basis for acceptability of extending the ILRT interval out to once per 15 years was the enhanced and robust primary containment inspection program and the local leakage rate testing of penetrations. Most of the primary containment leakage experienced has been attributed to penetration leakage and penetrations are thought to be the most likely location of most containment leakage at any time" 3.3.2 Current Callaway Primary Containment Leakage Rate Testing Program Requirements 10 CFR Part 50, Appendix J was revised, effective October 26, 1995, to allow licensees to choose containment leakage testing under either Option A, Prescriptive Requirements, or Option B, Performance-Based Requirements. On May 28, 1996, the NRC approved Amendment No.

111 to the facility operating license for Callaway (Reference 12). The amendment allowed the implementation of the recently approved Option B to 10 CFR Part 50, Appendix J. This new rule allowed for a performance-based option for determining the test frequency for containment leakage rate testing.

Currently, TS 5.5.16.a. requires that a program be established to comply with the containment leakage rate testing requirements of 10 CFR 50.54(o) and 10 CFR 50, Appendix J, Option B, as modified by approved exemptions. The program is required to be in accordance with the guidelines contained in RG 1.163. RG 1.163 endorses, with certain exceptions, NEI 94-01, Revision 0, as an acceptable method for complying with the provisions of Appendix J, Option B.

RG 1.163, Section C.1 states that licensees intending to comply with 10 CFR 50, Appendix J, Option B, should establish test intervals based upon the criteria in Section 11.0 of NEI 94-01 (Reference 5) rather than using test intervals specified in ANSI/ANS 56.8-1994. NEI 94-01, Section 11.0 refers to Section 9, which states that Type A testing shall be performed during a period of reactor shutdown at a frequency of at least once-per-ten years based on acceptable performance history. Acceptable performance history is defined as completion of two consecutive periodic Type A tests where the calculated performance leakage was less than 1.0La (where La is the maximum allowable leakage rate at design pressure). Elapsed time between the first and last tests in a series of consecutive satisfactory tests used to determine performance shall be at least 24 months.

Adoption of the Option B performance-based containment leakage rate testing program altered the frequency of measuring primary containment leakage in Types A, B, and C tests but did not alter the basic method by which Appendix J leakage testing is performed. The test frequency is based on an evaluation of the "as found" leakage history to determine a frequency for leakage

Enclosure Evaluation of the Proposed Change 23 testing, which provides assurance that leakage limits will not be exceeded. The allowed frequency for Type A testing as documented in NEI 94-01 is based, in part, upon a generic evaluation documented in NUREG-1493. The evaluation documented in NUREG-1493 included a study of the dependence or reactor accident risks on containment leak tightness for differing containment types. NUREG-1493 concluded in Section 10.1.2 that reducing the frequency of Type A tests (ILRTs) from the original three (3) tests per 10 years to one (1) test per 20 years was found to lead to an imperceptible increase in risk. The estimated increase in risk is very small because ILRTs identify only a few potential containment leakage paths that cannot be identified by Types B and C testing, and the leaks that have been found by Type A tests have been only marginally above existing requirements. Given the insensitivity of risk to containment leakage rate and the small fraction of leakage paths detected solely by Type A testing, NUREG-1493 concluded that increasing the interval between ILRTs is possible with minimal impact on public risk.

3.3.3 Callaway 10 CFR 50, Appendix J, Option B Licensing History April 5, 1995 - License Amendment No 98 The NRC approved Amendment No. 98, which deferred the requirement to perform the Type A Containment Integrated Leak Rate Test until Refuel 8 (October 1996). (Reference 13)

February 2, 1996 - License Amendment No 111 The NRC approved Amendment No. 111, which allowed the implementation of Option B to 10 CFR Part 50, Appendix J. This allowed for the implementation of a performance-based option for determining the test frequency for containment leakage rate testing in accordance with RG 1.163 and ANSI/ANS 56.8-1994. (Reference 12)

March 17, 2004 - License Amendment No. 160 The NRC approved Amendment No. 160, which revises TS 5.5.16 to add exceptions to RG 1.163 as follows:

1.

The visual examination of containment concrete surfaces intended to fulfill the requirements of 10 CFR 50, Appendix J, Option B testing, will be performed in accordance with the requirements of and frequency specified by ASME Section XI Code, Subsection IWL, except where relief has been authorized by the NRC.

2.

The visual examination of the steel liner plate inside containment intended to fulfill the requirements of 10 CFR 50, Appendix J, Option B testing, will be performed in accordance with the requirements of and frequency specified by ASME Section XI Code, Subsection IWE, except where relief has been authorized by the NRC.

(Reference 14)

Enclosure Evaluation of the Proposed Change 24 September 29, 2005 - License Amendment No. 168 The NRC approved Amendment No. 168, which changed TS 5.5.16 in support of the installation of replacement SGs during the refueling outage in the fall of 2005. This is discussed in Section 3.1.11 above. (Reference 15)

March 17, 2010 - License Amendment No. 195 The NRC approved Amendment No. 195, which revised TS 5.5.16. The revision would reflect a one-time extension of the current containment Type A leak rate test (integrated leak rate test or ILRT) interval requirement of Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Appendix J, "Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors,"

Option B, "Performance Based Requirements," from 10 years to 15 years. The change would allow the next ILRT to be performed no later than October 25, 2014. (Reference 16)

October 21, 2022 - License Amendment No. 228 The NRC approved Amendment No. 228, which revises the TSs to address the concerns of Generic Safety Issue (GSI)-191, Assessment of Debris Accumulation on PWR [Pressurized Water Reactor] Sump Performance, and respond to Generic Letter (GL) 2004-02, Potential Impact of Debris Blockage on Emergency Recirculation during Design Basis Accidents at Pressurized Water Reactors, which is an approved change to the Standard TSs into the Callaway TSs. The proposed change requested to modify the Callaway licensing basis analyses to show compliance with Title 10 of the Code of Federal Regulations (10 CFR) Section 50.46, Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors, considering the effects of debris using both deterministic and risk-informed methodologies (Reference 35).

3.3.4 Integrated Leakage Rate Testing History (ILRT)

As noted previously, Callaway TS 5.5.16 currently requires Types A, B, and C testing in accordance with RG 1.163, which endorses the methodology for complying with 10 CFR 50, Appendix J, Option B. Since the adoption of Option B, the performance leakage rates are calculated in accordance with NEI 94-01, Section 9.1.1 for Type A testing.

Table 3.3.4-1 lists the past Periodic Type A ILRT results for Callaway.

Table 3.3.4-1, Periodic Type A ILRT Results for Callaway Unit 1 95% Upper Confidence Limit (UCL)

(wt%/day)

Test Pressure Test Date 0.1038 48.6 psig (Pa = 48.1 psig)

October 2014 0.0445 47.7 psig (Pa = 48.1 psig)

October 1999 0.066 50.3 psig (Pa = 48.1 psig)

February 1990 0.044 49.7 psig (Pa = 48.1 psig)

April 1987

Enclosure Evaluation of the Proposed Change 25 Table 3.3.4-1, Periodic Type A ILRT Results for Callaway Unit 1 95% Upper Confidence Limit (UCL)

(wt%/day)

Test Pressure Test Date 0.047 1 50.05 psig (Pa = 48.1 psig)1 January 1984 Note 1:

Preoperational ILRT performed in conjunction with a structural integrity test. The containment design pressure is 60 psig.

3.3.5 Performance Leakage Rate Determination The current ILRT test interval for Callaway is ten years. Verification of this interval is presented in Table 3.3.5-1, Callaway ILRT Test Results Verification of Current Extended ILRT Interval. The acceptance criteria used for this verification is contained in NEI 94-01, Revisions 2-A and 3-A, Section 5.0, Definitions, and is as follows:

"The performance leakage rate is calculated as the sum of the Type A upper confidence limit (UCL) and as-left minimum pathway leakage rate (MNPLR) leakage rate for all Type B and Type C pathways that were in service, isolated, or not lined up in their test position (i.e.,

drained and vented to containment atmosphere) prior to performing the Type A test. In addition, leakage pathways that were isolated during performance of the test because of excessive leakage must be factored into the performance determination. The performance criterion for Type A tests is a performance leak rate of less than 1.0La."

Table 3.3.5-1, Callaway ILRT Test Results Verification of Current Extended ILRT Interval Test Date Upper 95%

Confidence Limit (wt.%/day)

(Test Pressure)

Level Corrections (Leakage Savings)

(wt.%/day)

As Left Min Pathway Penalty for Isolated Pathways (wt.%/day)

Adjusted As Left Leak Rate (wt.%/day)

ILRT Acceptance Criteria (wt.%/day)

Test Method

/ Data Analysis Techniques October 2014 0.0388 (48.8 psig)

-0.0037 0.0688 0.1038 0.15 Absolute /

ANSI/ANS 56.8-1994 Mass Point October 1999 0.0388 (47.7 psig) 0.0 0.0057 0.0445 0.15 Absolute /

ANSI/ANS 56.8-1994 Mass Point

Enclosure Evaluation of the Proposed Change 26 3.4 Plant Specific Confirmatory Analysis 3.4.1 Methodology An analysis was performed to provide a risk assessment of extending the currently allowed containment Type A ILRT interval to a permanent fifteen years for Callaway. The extension would allow for substantial cost savings as the ILRT could be deferred for additional scheduled refueling outages for Callaway. The risk assessment follows the guidelines from:

x NEI 94-01, Revision 3-A and 2-A (References 2 and 8) x NEI Interim Guidance for Performing Risk Impact Assessments in Support of One-Time Extensions for Containment Integrated Leakage Rate Test Surveillance Intervals, Revision 4 (Reference 20) x NRC regulatory guidance on the use of Probabilistic Risk Assessment (PRA) as stated in RG 1.200, Revision 3, as applied to ILRT interval extensions (Reference 17) x Risk insights in support of a request for a plants licensing basis as outlined in RG 1.174, Revision 3 (Reference 3) x Methodology used for Calvert Cliffs to estimate the likelihood and risk implications of corrosion-induced leakage of steel liners going undetected during the extended test interval (Reference 32) x Methodology used in EPRI 1018243, Revision 2-A of EPRI 1009325 (Reference 11)

Details of the Callaway risk assessment, providing an assessment of the risk associated with implementing a permanent extension of the Callaway containment Type A ILRT interval from ten years to fifteen years, is contained in Attachment 1, "Evaluation of Risk Significance of Permanent ILRT Extension," of this submittal.

Revisions to 10 CFR 50, Appendix J (Option B) allow individual plants to extend the ILRT Type A surveillance testing frequency requirement from three in ten years to at least once in ten years.

The revised Type A frequency is based on an acceptable performance history defined as two consecutive periodic Type A tests at least 24 months apart in which the calculated performance leakage rate was less than the limiting containment leakage rate of 1La.

The basis for the current 10-year test interval is provided in Section 11.0 of NEI 94-01, Revision 0, and established in 1995 during development of the performance-based Option B to Appendix J. Section 11.0 of NEI 94-01 states that NUREG-1493, Performance-Based Containment Leak Test Program, (Reference 6), provides the technical basis to support rulemaking to revise leakage rate testing requirements contained in Option B to Appendix J. The basis consisted of qualitative and quantitative assessments of the risk impact (in terms of increased public dose)

Enclosure Evaluation of the Proposed Change 27 associated with a range of extended leakage rate test intervals. To supplement the NRCs rulemaking basis, NEI undertook a similar study. The results of that study are documented in Electric Power Research Institute (EPRI) Research Project TR-104285, Risk Impact Assessment of Revised Containment Leak Rate Testing Intervals (Reference 23).

The NRC report on performance-based leak testing, NUREG-1493 (Reference 6), analyzed the effects of containment leakage on the health and safety of the public and the benefits realized from the containment leak rate testing. In that analysis, it was determined that for a representative pressurized water reactor (PWR) plant (i.e., Surry) containment isolation failures contribute less than 0.1% to the latent risks from reactor accidents. Consequently, it is desirable to show that extending the ILRT interval will not lead to a substantial increase in risk from containment isolation failures for Callaway.

NEI 94-01, Revision 3-A supports using EPRI Report No. 1009325, Revision 2-A (EPRI 1018243), Risk Impact Assessment of Extended Integrated Leak Rate Testing Intervals (Reference 11), for performing risk impact assessments in support of ILRT extensions. The guidance provided in Appendix H of EPRI Report No. 1009325, Revision 2-A, for performing risk impact assessments in support of ILRT extensions builds on the EPRI Risk Assessment methodology, EPRI TR-104285 (Reference 7). This methodology is followed to determine the appropriate risk information for use in evaluating the impact of the proposed ILRT changes.

It should be noted that containment leak-tight integrity is also verified through periodic in-service inspections conducted in accordance with the requirements of the ASME Code,Section XI. More specifically, Subsection IWE provides the rules and requirements for in-service inspection of Class MC pressure-retaining components and their integral attachments, and of metallic shell and penetration liners of Class CC pressure-retaining components and their integral attachments in light-water cooled plants. Furthermore, NRC regulations 10 CFR 50.55a(b)(2)(ix)(E) require licensees to conduct visual inspections of the accessible areas of the interior of the containment.

The associated change to NEI 94-01 requires that visual examinations be conducted during at least three other outages, and in the outage during which the ILRT is being conducted. These requirements are not changed as a result of the extended ILRT interval. However, in the case of Callaway, inspections will be performed in accordance with TS 5.5.16 paragraph a., items 1 and 2 as shown in Section 2.0 of this LAR. In addition, Appendix J, Type B local leak tests performed to verify the leak-tight integrity of containment penetration bellows, airlocks, seals, and gaskets are also not affected by the change to the Type A test frequency.

The acceptance guidelines in RG 1.174 are used to assess the acceptability of this permanent extension of the Type A test interval beyond that established during the Option B rulemaking of Appendix J. RG 1.174 defines very small changes in the risk-acceptance guidelines as increases in Core Damage Frequency (CDF) less than 10-6 per reactor year and increases in Large Early Release Frequency (LERF) less than 10-7 per reactor year. Since the Type A test does not impact CDF, the relevant criterion is the change in LERF. RG 1.174 also defines small changes in LERF as below 10-6 per reactor year. RG 1.174 discusses defense-in-depth and encourages the use of risk analysis techniques to help ensure and show that key principles, such as the defense-in-depth philosophy, are met. Therefore, the increase in the Conditional

Enclosure Evaluation of the Proposed Change 28 Containment Failure Probability (CCFP), which helps ensure the defense-in-depth philosophy is maintained, is also calculated.

Regarding CCFP, changes of up to 1.1% have been accepted by the NRC for the one-time requests for extension of ILRT intervals. In context, it is noted that a CCFP of 1/10 (10%) has been approved for application to evolutionary light water designs. Given these perspectives, a change in the CCFP of up to 1.5% is assumed to be small (Reference 2).

In addition, the total annual risk (person-rem/yr population dose) is examined to demonstrate the relative change in this parameter. While no acceptance guidelines for these additional figures of merit are published, examinations of NUREG-1493 (Reference 6) and Safety Evaluations (SEs) for one-time interval extension (summarized in Appendix G of Reference 11) indicate a range of incremental increases in population dose that have been accepted by the NRC. The range of LQFUHPHQWDOSRSXODWLRQGRVHLQFUHDVHVLVIURPWRSHUVRQ-rem/yr and/or 0.002% to 0.46% of the total accident dose (Reference 11). The total doses for the spectrum of all accidents (NUREG-1493 (Reference 6), Figure 7-2) result in health effects that are at least two orders of magnitude less than the NRC Safety Goal Risk. Given these perspectives, a small population dose is defined as an increase from the baseline interval (3 tests per 10 years) dose of

SHUVRQ-rem per year or 1% of the total baseline dose, whichever is less restrictive for the risk impact assessment of the proposed extended ILRT interval (Reference 2).

In the SE issued by the NRC letter dated June 25, 2008 (Reference 9), the NRC concluded that the methodology in EPRI TR-1009325, Revision 2 (Reference 11), was acceptable for referencing by licensees proposing to amend their TS to permanently extend the ILRT surveillance interval to 15 years, subject to the limitations and conditions noted in Section 4.0 of the SE. Table 3.4.1-1 below addresses each of the four (4) limitations and conditions from Section 4.2 of the SE for the use of EPRI 1009325, Revision 2.

Table 3.4.1-1, EPRI Report No. 1009325, Revision 2, Limitations and Conditions Limitation and Condition (From Section 4.2 of SE)

Callaway Response

1. The licensee submits documentation indicating that the technical adequacy of their PRA is consistent with the requirements of RG 1.200 relevant to the ILRT extension application.

Callaway PRA technical adequacy is addressed in Section 3.4.2 of this LAR and Attachment 1, "Evaluation of Risk Significance of Permanent ILRT Extension, Appendix A, "PRA Acceptability."

Enclosure Evaluation of the Proposed Change 29 Table 3.4.1-1, EPRI Report No. 1009325, Revision 2, Limitations and Conditions Limitation and Condition (From Section 4.2 of SE)

Callaway Response 2.a The licensee submits documentation indicating that the estimated risk increase associated with permanently extending the ILRT surveillance interval to 15 years is small, and consistent with the clarification provided in Section 3.2.4.5 of this SE.

Since the ILRT does not impact CDF, the relevant criterion is LERF. The increase in LERF resulting from a change in the Type A ILRT test interval from 3 in 10 years to 1 in 15 years is estimated as 6.46E-8/yr using the EPRI guidance; this value increases negligibly if the risk impact of corrosion-induced leakage of the steel liners occurring and going undetected during the extended test interval is included.

Therefore, the estimated change in LERF is determined to be very small using the acceptance guidelines of RG 1.174. (See, Section 7.0 of this submittal.)

2.b Specifically, a small increase in population dose should be defined as an increase in population dose of less than or equal to either 1.0 person-rem per year or 1% of the total population dose, whichever is less restrictive.

The effect resulting from changing the Type A test frequency to 1-per-15 years, measured as an increase to the total integrated plant risk for those accident sequences influenced by Type A testing is 0.025 person-rem/yr. NEI 94-01 (Reference 8) states that a small population dose is defined as DQLQFUHDVHRISHUVRQ-UHPSHU\\HDURU

1% of the total population dose, whichever is less restrictive for the risk impact assessment of the extended ILRT intervals. The results of this calculation meet these criteria. Moreover, the risk impact for the ILRT extension when compared to other severe accident risks is negligible. (See Attachment 1, Section 7.0 of this submittal.)

2.c In addition, a small increase in CCFP should be defined as a value marginally greater than that accepted in a previous one-time 15-year ILRT extension requests. This would require that the increase in CCFP be less than or equal to 1.5 percentage point.

The increase in the conditional containment failure probability from the 3 in 10-year interval to 1 in 15-year interval is 0.913%. NEI 94-01 (Reference 2) VWDWHVWKDWLQFUHDVHVLQ&&)3RI

1.5% is small. Therefore, this increase is judged to be small. (See Attachment 1, Section 7.0 of this submittal.)

3. The methodology in EPRI Report No.

1009325, Revision 2, is acceptable except for the calculation of the increase in expected population dose (per year of reactor operation). In order to make the methodology acceptable, the average leak rate accident case (accident case 3b) used by the licensees shall be 100 La instead of 35 La.

The representative containment leakage for Class 3b sequences used by Callaway is 100 La, based on the guidance provided in EPRI Report No. 1009325, Revision 2-A. (See Attachment 1, Section 4.0 of this submittal.)

Enclosure Evaluation of the Proposed Change 30 Table 3.4.1-1, EPRI Report No. 1009325, Revision 2, Limitations and Conditions Limitation and Condition (From Section 4.2 of SE)

Callaway Response

4. A licensee amendment request (LAR) is required in instances where containment over-pressure is relied upon for ECCS performance.

Per the ECCS analysis of record, Callaway does not rely on containment overpressure for ECCS injection. In the analysis performed for resolution of Generic Safety Issue (GSI) 191 for Callaway (i.e., for Callaway's response to Generic Letter 04-02), which is a separate analysis from the ECCS/accident analysis of record, a minimum necessary overpressure credit is applied to suppress flashing at the top of the strainer assembly. For LBLOCA, 1.7 psi of credit is needed for temperatures above 212°F (approximately 10% of available containment pressure). (Refer to Section 3.2.1 of this enclosure and to Section 5.2.9 of Attachment 1.)

3.4.2 Technical Adequacy of the Probabilistic Risk Assessment (PRA)

A PRA analysis is utilized in the containment Type A ILRT risk assessment to support an extension of the Callaway test interval from ten years to fifteen years.

This section provides information on the technical adequacy of the Callaway PRA Internal Events, Internal Flooding, High Winds, Fire, and Seismic PRA models.

NEI Topical Report NEI 06-09-A, Revision 0 (Reference 18), as clarified by the NRC final safety evaluation of this report (Reference 43), defines the technical attributes of a PRA model and its associated Configuration Risk Management Program (CRMP) tool required to implement this risk-informed application. Meeting these requirements satisfies RG 1.174, Revision 3 (Reference 3),

requirements for risk-informed plant-specific changes to a plant's licensing basis.

Ameren Missouri employs a multi-faceted approach to establishing and maintaining the technical adequacy and fidelity of PRA models for Callaway. This approach includes both a PRA maintenance and update process procedure and the use of self-assessments and independent peer reviews.

The Callaway PRA models are at-power models consisting of four hazard models - Internal Flooding, Fire, Seismic, and High Wind. Each hazard model has the Internal Events model as the base with hazard specific initiators added and fault tree modifications and additions made, as necessary. Each model directly addresses plant configurations during plant Modes 1, 2 and 3 of reactor operation. The models provide both CDF and LERF. All five of these PRA models were developed to comply with RG 1.200 Revision 2 (Reference 4).

Enclosure Evaluation of the Proposed Change 31 Peer Review Findings Closure Process All of the PRA models discussed in this appendix have been peer reviewed and assessed against RG 1.200 Revision 2 (Reference 4).

The review and closure of finding-level F&Os was performed by an independent assessment team using the process documented in Appendix X to NEI 05-04, NEI 07-12 and NEI 12-16, "Close-out of Facts and Observations (F&Os) (Reference 44) as accepted by the NRC in the letter dated May 3, 2017 (Reference 45). All of the reviews also met the requirements of NEI 17-07, Revision 2, Performance of PRA Peer Reviews Using the ASME/ANS PRA Standard (Reference 46).

The assessment team assessed whether each F&O was closed through application of a PRA maintenance or upgrade activity, as defined by the ASME/ANS PRA Standard, or through application of a new method. Note that, per APC 17-13,

Subject:

NRC Acceptance of Industry Guidance on Closure of PRA Peer Review Findings, dated May 8, 2017 (Reference 71), with attachment Appendix X, a new method represents a fundamentally new approach in addressing a technical aspect of PRA. The results of the peer reviews and independent assessments have been documented and are available for NRC audit.

The PRA scope and technical adequacy is met for this application as the Standard requirements for all models are met at Capability Category II (CCII) or higher. There are no open Finding F&Os against any of the models discussed in this application, and all Finding F&Os have been independently assessed and closed using the processes discussed above. The resolved findings and the basis for resolution are documented in the Callaway PRA documentation and the F&O Closure Review reports.

Scope of the Callaway PRA Models The Internal Events, Internal Flooding, Fire, High Winds, and Seismic PRA models are at-power models (i.e., they directly address plant configurations during plant Modes 1, 2 and 3 of reactor operation). The models provide both CDF and LERF.

Note that the Callaway PRA models do not incorporate the risk impacts of external events except for High Winds and Seismic. The treatment of non-modeled external risk hazards are discussed in the OEH Notebook (Reference 47) and Enclosure 4 (Reference 48), which show that all non-modeled external risk hazards screen.

Technical Adequacy of the Callaway Internal Events and Internal Flooding PRA Model Topical Report NEI 06-09-A (Reference 18) requires that the PRA be reviewed to the guidance of RG 1.200, Revision 2 (Reference 4) for a PRA that meets CCII for the supporting requirements of the Internal Events at power ASME/ANS PRA Standard (Reference 49).

Enclosure Evaluation of the Proposed Change 32 The information provided in this section demonstrates that the Callaway Internal Events PRA model (including Internal Flooding) meets the expectations for PRA scope and technical adequacy as presented in ASME/ANS RA-S-2009 (Reference 49) and RG 1.200 to fully support this ILRT extension application. The Ameren Missouri risk management process ensures that the PRA model used in this application reflects the as-built and as-operated plant for Callaway.

Related to the technical adequacy of the Internal Events model, the Internal Events discussion below describes implementation of the methodology provided in PWROG-18027-NP (Reference

54) for assessing the loss of room cooling in PRA modeling. Following, but unrelated to, implementation of the method provided in PWROG-18027-NP into the Callaway PRA, this method was chosen by the PWROG and NEI to pilot the Newly Developed Methods (NDM) peer review process established in NEI 17-07 (Reference 46). The NEI 17-07 process was successfully completed with all applicable NDM attributes met at CC I/II and no open peer review Findings against the method in PWROG-18027-NP.

In addition, an implementation peer review and associated F&O closure review have been completed using NRC-approved processes, with no open Findings identified against implementation of the method. While the NEI 17-07 process was completed successfully, it is recognized that this process was not an endorsed process until RG 1.200, Revision 3 was issued in December 2020 (Reference 17). As a result, the NRC staff may decide to independently review the method in PWROG-18027-NP for technical adequacy. The PWROG-18027-NP report contains the technical basis for the acceptability of the method and is available for NRC audit.

Peer Review Summary The Internal Events/Internal Flooding PRA was peer reviewed in April 2019. This peer review was a full-scope review of the technical elements of the Internal Events and Internal Flooding at-power PRA as documented in PWROG-19012-P (Reference 55). As a full scope review, it included those supporting requirements (SRs) specified in PWROG-19020-NP, Revision 1 (Reference 56) for implementation of the methodology for loss of room cooling modeling provided in PWROG-18027-NP (Reference 54).

An Independent Assessment of F&Os was conducted in November 2019 and documented in PWROG-19034-P (Reference 53). The scope of the assessment included all F&Os generated in the April 2019 peer review. All F&Os except for one were closed. The remaining F&O was related to implementation of the methodology provided in PWROG-18027-NP (Reference 54) for assessing the loss of room cooling in PRA modeling. Following, but unrelated to, incorporation of the method provided in PWROG-18027-NP into the Callaway PRA, this method was chosen by the PWROG and NEI to pilot the Newly Developed Methods (NDM) peer review process established in NEI 17-07 (Reference 46). Despite the Callaway assessment, and acknowledgement by the PWROG, that the method provided in PWROG-18027-NP did not necessarily meet the definition of an NDM, Callaway decided to suspend resolution of the associated F&O until the NDM peer review and closure of any F&Os were completed using the process established in NEI 17-07. Also, during the November 2019 independent assessment, two F&O resolutions were determined to be upgrades to the Internal Events/Internal Flooding

Enclosure Evaluation of the Proposed Change 33 PRA. Thus, a focused-scope peer review was required. Based on this focused scope peer review, one new Internal Events F&O was generated.

During February and March 2020, a new peer review, following the guidance in NEI 17-07, Revision 2, (Reference 46) was conducted on the method provided in PWROG-18027-NP (Reference 54) and documented in PWROG-19020-NP, Revision 1 (Reference 56). Based on the results of this review, all applicable NDM attributes are met at CC I/II and there are no open peer review Findings against the method in PWROG-18027-NP.

In June 2020, an independent assessment of F&O resolution and a focused scope peer review, completing the review of PWROG-18027-NP implementation, were conducted on the Callaway Internal Events and Fire PRA models. The focused scope peer review determined that all of the SRs that were examined, including the SR associated with the F&O related to implementation of the method in PWROG-18027-NP, satisfy CCII or higher requirements as documented in AMN#PES00031-REPT-001 (Reference 50). The independent assessment of F&Os included an assessment of all remaining open F&O Findings. The results of this review are documented in AMN#PES00031-REPT-002 (Reference 51).

There are no open peer review Findings for the Internal Events/Internal Flooding PRA model.

Technical Adequacy of Callaway High Winds PRA Model The information provided in this section demonstrates that the Callaway High Winds PRA model meets the expectations for PRA scope and technical adequacy as presented in ASME/ANS RA-Sa-2009 (Reference 49) and RG 1.200 to fully support this ILRT extension application. The Ameren Missouri risk management process ensures that the PRA model used in this application reflects the as-built and as-operated plant for Callaway.

Peer Review Summary The High Winds PRA was peer reviewed in April 2019 and documented in PWROG-19022-P (Reference 52). The scope of this work was to review the Callaway External Hazards Screening Assessment and High Winds PRA against the technical elements in Sections 6 and 7 of the ASME/ANS RA-Sa-2009 Standard and in RG 1.200.

An Independent Assessment of F&O resolution was conducted in November 2019 and documented in PWROG-19034-P (Reference 53). The scope of the assessment included all F&Os generated in the April 2019 peer review. All F&Os were closed.

There are no open peer review Findings for the Other External Hazards Screening or the High Winds PRA model.

Technical Adequacy of Callaway Seismic PRA Model The information provided in this section demonstrates that the Callaway Seismic PRA model

Enclosure Evaluation of the Proposed Change 34 meets the expectations for PRA scope and technical adequacy as presented in ASME/ANS RA-S CASE 1, Case for ASME/ANS RA-Sb-2013 (Reference 57) and RG 1.200 to fully support this ILRT extension application. The Ameren Missouri risk management process ensures that the PRA model used in this application reflects the as-built and as-operated plant for Callaway.

Peer Review Summary The Seismic PRA was peer reviewed in June 2018 and documented in PWROG-18044-P (Reference 58). This peer review was conducted against the requirements of the Code Case for ASME/ANS RA-Sb-2013 (Reference 57), as amended by the NRC on March 12, 2018 (Reference 59). The Code Case is an approved alternative to Part 5 of ASME/ANS RA-Sb-2013 Addendum B, the American Society of Mechanical Engineers (ASME) / American Nuclear Society (ANS) Probabilistic Risk Assessment (PRA) Standard.

An Independent Assessment of F&Os was conducted in March 2019. The scope of the assessment included all but two of the F&Os generated in the June 2018 peer review. All in-scope F&Os were closed as documented in PWROG-19011-P (Reference 60). Also, in the March 2019 review documented in PWROG-19011-P, three SRs were the subject of a focused-scope peer review based on the closures of associated F&Os being assessed as upgrades. As a result of that peer review, the three SRs were determined to be met at CCII.

Subsequently, another Independent Assessment of F&Os was conducted in June 2020 and documented in AMN#PES00031-REPT-002 (Reference 51). The scope of the assessment included all remaining F&Os generated in the June 2018 peer review. All F&Os were closed.

There are no open peer review Findings for the Seismic PRA model.

Technical Adequacy of Callaway Fire PRA Model The information provided in this section demonstrates that the Callaway Fire PRA model meets the expectations for PRA scope and technical adequacy as presented in ASME/ANS RA-Sa-2009 (Reference 49) and RG 1.200 to fully support this ILRT extension application. The Ameren Missouri risk management process ensures that the PRA model used in this application reflects the as-built and as-operated plant for Callaway.

The Internal Fire PRA model was developed consistent with NUREG/CR-6850 (Reference 61) and only utilizes methods previously accepted by the NRC. Callaway was approved to implement National Fire Protection Association (NFPA) Standard 805, "Performance Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants," (NFPA-805) in January 2014, and since that time, there have been numerous updates to the approved methods through the issuance of Fire PRA frequently asked questions and new or revised guidance documents. New or revised guidance is specifically addressed through the Callaway PRA maintenance and update process. The Ameren Missouri risk management process ensures that the PRA model used in this application reflects the as-built and as-operated plant for Callaway.

Enclosure Evaluation of the Proposed Change 35 It should also be noted that, as part of transition to NFPA-805, there were several committed modifications and implementation items as documented in NFPA-805 LAR Attachment S, "Plant Modifications and Items to be Completed during Implementation," which described the Callaway plant modifications necessary to implement the NFPA-805 licensing basis. All NFPA-805 LAR Attachment S items have been implemented; therefore, there are no NFPA-805 open items impacting this application.

Peer Review Summary The Fire PRA was prepared using the methodology defined in NUREG/CR-6850, "Fire PRA Methodology for Nuclear Power Facilities," to support a transition to NFPA-805. The Fire PRA was peer reviewed to ASME/ANS RA-S-2009 and RG 1.200, Revision 2 in October 2009. The review is documented in LTR-RAM-11-10-019 (Reference 62).

An Independent Assessment of F&Os was conducted in June 2019 and documented in AMN#PES00021-REPT-001 (Reference 63).

In June 2020, an independent assessment of F&Os and a focused scope peer review were conducted for the Callaway Internal Events and Fire PRA models. The focused scope peer review generated additional Fire PRA related F&Os as documented in AMN#PES00031-REPT-001 (Reference 50). The independent assessment of F&Os included an assessment of all remaining open F&O Findings. As documented in AMN#PES00031-REPT-002 (Reference 51),

all Finding F&Os were closed, including the Fire PRA Findings identified in the Focused Scope peer review.

In fulfillment of Commitment 50437 in Enclosure 4 to ULNRC-06550 (ML20304A456) and associated with closure of NFPA-805 LAR Table S-3 Implementation Item 13-805-001, a focused scope peer review was conducted in November 2020, as documented in AMN#PES00031-REPT-003 (Reference 65), for the resolution of Fire PRA Suggestion F&O FSS-B1-03, which a July 2019 F&O closure review had determined to be an upgrade, as documented in AMN#PES00021-REPT-001 (Reference 63).

As documented in AMN#PES00042-REPT-002 (Reference 64), the F&Os from this focused scope peer review were closed during an F&O closure review in February 2021. The results of this review formally closed Commitment 50437.

There are no open peer review Findings for the Fire PRA model.

3.4.3 Summary of Plant-Specific Risk Assessment Results The PRA scope and technical adequacy is met for this application as the Standard requirements for all models are met at CCII or higher. There are no open Finding F&Os against any of the models discussed in this Enclosure, and all Finding F&Os have been independently assessed and closed using the processes discussed in Attachment 1, Section 2 of this LAR. In addition, all

Enclosure Evaluation of the Proposed Change 36 of the reviews also met the requirements of NEI 17-07, Revision 2 (Reference 46).

Based on the results from Attachment 1, Section 5.2 and the sensitivity calculations presented in, Section 5.3, the following conclusions regarding the assessment of the plant risk associated with extending the Type A ILRT test frequency to 15 years:

x Regulatory Guide 1.174 (Reference 3) provides guidance for determining the risk impact of plant-specific changes to the licensing basis. Regulatory Guide 1.174 defines very small changes in risk as resulting in increases of CDF less than 1.0E-06/year and increases in LERF less than 1.0E-07/year. Regulatory Guide 1.174 defines small changes in risk as resulting in increases of CDF greater than 1.0E-6/yr and less than 1.0E-5/yr and increases in LERF greater than 1.0E-7/yr and less than 1.0E-6/yr. Since the ILRT does not impact CDF, the relevant criterion is LERF. The increase in LERF resulting from a change in the Type A ILRT test interval from 3 in 10 years to 1 in 15 years is estimated as 6.46E-8/yr using the EPRI guidance; this value increases negligibly if the risk impact of corrosion-induced leakage of the steel liners occurring and going undetected during the extended test interval is included. Therefore, the estimated change in LERF is determined to be very small using the acceptance guidelines of Regulatory Guide 1.174. The risk change resulting from a change in the Type A ILRT test interval from 3 in 10 years to 1 in 15 years bounds the 1 in 10 years to 1 in 15 years risk change.

Considering the increase in LERF resulting from a change in the Type A ILRT test interval from 1 in 10 years to 1 in 15 years is estimated as 2.69E-8/yr, the risk increase is very small using the acceptance guidelines of Regulatory Guide 1.174.

x When external event risk is included, the increase in LERF resulting from a change in the Type A ILRT test interval from 3 in 10 years to 1 in 15 years is estimated as 5.92E-7/yr using the EPRI guidance, and total LERF is 4.08E-6/yr. As such, the estimated change in LERF is determined to be small using the acceptance guidelines of Regulatory Guide 1.174 (Reference 4). The risk change resulting from a change in the Type A ILRT test interval from 3 in 10 years to 1 in 15 years bounds the 1 in 10 years to 1 in 15 years risk change. When external event risk is included, the increase in LERF resulting from a change in the Type A ILRT test interval from 1 in 10 years to 1 in 15 years is estimated as 2.46E-7/yr, and the total LERF is 3.73E-6/yr. Therefore, the risk increase is small using the acceptance guidelines of Regulatory Guide 1.174.

x The effect resulting from changing the Type A test frequency to 1-per-15 years, measured as an increase to the total integrated plant risk for those accident sequences influenced by Type A testing is 0.025 person-rem/yr. NEI 94-01 (Reference 8) states that a small populatioQGRVHLVGHILQHGDVDQLQFUHDVHRISHUVRQ-UHPSHU\\HDURURIWKH

total population dose, whichever is less restrictive for the risk impact assessment of the extended ILRT intervals. The results of this calculation meet these criteria. Moreover, the risk impact for the ILRT extension when compared to other severe accident risks is negligible.

x The increase in the conditional containment failure probability from the 3 in 10-year

Enclosure Evaluation of the Proposed Change 37 interval to 1 in 15-year interval is 0.913%. NEI 94-01 (Reference 2) states that increases LQ&&)3RILV³VPDOO' Therefore, this increase is judged to be small.

Therefore, increasing the ILRT interval to 15 years is considered to be small since it represents a small change to the Callaway risk profile.

3.4.4 Previous Assessments The NRC in NUREG-1493 (Reference 6) has previously concluded that:

x Reducing the frequency of Type A tests (ILRTs) from 3 per 10 years to 1 per 20 years was found to lead to an imperceptible increase in risk. The estimated increase in risk is very small because ILRTs identify only a few potential containment leakage paths that cannot be identified by Type B or Type C testing, and the leaks that have been found by Type A tests have been only marginally above existing requirements.

x Given the insensitivity of risk to containment leakage rate and the small fraction of leakage paths detected solely by Type A testing, increasing the interval between integrated leakage-rate tests is possible with minimal impact on public risk. The impact of relaxing the ILRT frequency beyond 1 in 20 years has not been evaluated. Beyond testing the performance of containment penetrations, ILRTs also test integrity of the containment structure.

The conclusions for Callaway confirm these general conclusions on a plant-specific basis considering the severe accidents evaluated for Callaway, the Callaway containment failure modes, and the local population surrounding Callaway.

3.4.5 RG 1.174 Revision 3 Defense in Depth Evaluation RG 1.174, Revision 3 (Reference 3), describes an approach that is acceptable for developing risk-informed applications for a licensing basis change that considers engineering issues and applies risk insights. One of the considerations included in RG 1.174 is Defense in Depth.

Defense in Depth is a safety philosophy that employs successive compensatory measures to prevent accidents or mitigate damage if a malfunction, accident, or naturally caused event occurs at a nuclear facility. The following seven considerations, as presented in RG 1.174, Revision 3, Section C.2.1.1.2, Considerations for Evaluating the Impact of the Proposed Licensing Basis Change on Defense in Depth, will serve to evaluate the proposed licensing basis change for overall impact on Defense in Depth for Callaway.

1.

Preserve a reasonable balance among the layers of defense.

A reasonable balance of the layers of defense (i.e., minimizing challenges to the plant, preventing any events from progressing to core damage, containing the radioactive source term, and emergency preparedness) helps to ensure an apportionment of the plants capabilities between limiting disturbances to the plant and mitigating their

Enclosure Evaluation of the Proposed Change 38 consequences. The term reasonable balance is not meant to imply an equal apportionment of capabilities. The NRC recognizes that aspects of a plants design or operation might cause one or more of the layers of defense to be adversely affected. For these situations, the balance between the other layers of defense becomes especially important when evaluating the impact of the proposed licensing basis change and its effect on defense in depth.

Response

Several layers of defense are in place to ensure the Callaway containment structure(s);

penetrations, isolation valves and mechanical seal systems; continue(s) to perform their intended safety function. The purpose of the proposed change is to extend the testing frequencies of the Type A ILRT from 10 years to 15 years and Type C LLRTs for selected components from 60 months to 75 months.

As shown in NUREG-1493 (Reference 6), increasing the test frequency of ILRTs up to a 20-year test interval was found to lead to an imperceptible increase in risk. The estimated increase in risk is very small because ILRTs identify only a few potential containment leakage paths that cannot be identified by Type B or Type C testing. The study also concluded that extending the frequency of Type B tests is possible with no adverse impact on risk as identified leakage through Type B mechanical penetrations are both infrequent and small. Finally, the study concluded that Types B and C tests could identify the vast majority (greater than 95 percent) of all potential leakage paths.

Several programmatic factors can also be cited as layers of defense ensuring the continued safety function of the Callaway containment pressure boundary. NEI 94-01, Revisions 2-A and 3-A, require sites adopting the 15-year extended ILRT interval to perform visual examinations of the accessible interior and exterior surfaces of the containment structure for structural degradation that may affect the containment leak-tight integrity at the frequency prescribed by the guidance or, if approved through a TS amendment, at the frequencies prescribed by ASME Section XI, which is the case for Callaway. Additionally, several measures are put in place to ensure integrity of the Types B and C tested components. NEI 94-01 limits large containment penetrations such as airlocks, purge and vent valves, BWR main steam and feedwater isolation valves, to a maximum 30-month testing interval. For those valves that meet the performance standards defined in NEI 94-01, Revision 3-A and are selected for test intervals greater than 60 months, a leakage understatement penalty is added to the MNPLR prior to the frequency being extended beyond 60-months. Finally, identification of adverse trends in the overall Types B and C leakage rate summations and available margin between the Type B and Type C leakage rate summation and its regulatory limit are required by NEI 94-01, Revision 3-A to be shown in the Callaway post-outage report(s). Therefore, the proposed change does not challenge or limit the layers of defense available to assess the ability of the Callaway containment structure to perform its safety function.

PRA Response:

Enclosure Evaluation of the Proposed Change 39 The use of the risk metrics of LERF, population dose, and conditional containment failure probability collectively ensures the balance between prevention of core damage, prevention of containment failure, and consequence mitigation is preserved. The change in LERF is very small with respect to internal events and small when including external events per RG 1.174, and the change in population dose and CCFP are small as defined in this analysis and consistent with NEI 94-01, Revision 3-A.

2.

Preserve adequate capability of design features without an overreliance on programmatic activities as compensatory measures.

Nuclear power plant licensees implement a number of programmatic activities, including programs for quality assurance, testing and inspection, maintenance, control of transient combustible material, foreign material exclusion, containment cleanliness, and training. In some cases, activities that are part of these programs are used as compensatory measures; that is, they are measures taken to compensate for some reduced functionality, availability, reliability, redundancy, or other feature of the plants design to ensure safety functions (e.g., reactor vessel inspections that provide assurance that reactor vessel failure is unlikely). NUREG-2122, Glossary of Risk-Related Terms in Support of Risk-Informed Decision Making, (Reference 19), defines safety function as those functions needed to shut down the reactor, remove the residual heat, and contain any radioactive material release.

A proposed licensing basis change might involve or require compensatory measures.

Examples include hardware (e.g., skid-mounted temporary power supplies); human actions (e.g., manual system actuation); or some combination of these measures. Such compensatory measures are often associated with temporary plant configurations. The preferred approach for accomplishing safety functions is through engineered systems.

Therefore, when the proposed licensing basis change necessitates reliance on programmatic activities as compensatory measures, the licensee should justify that this reliance is not excessive (i.e., not overly reliant). The intent of this consideration is not to preclude the use of such programs as compensatory measures but to ensure that the use of such measures does not significantly reduce the capability of the design features (e.g.,

hardware).

Response

The purpose of the proposed change is to extend the testing frequencies of the Type A ILRT from 10 years to 15 years and select Type C LLRTs from 60 months to 75 months.

Several programmatic factors were defined in the response to Question 1 above, which are required when adopting NEI 94-01, Revisions 2-A and 3-A. These factors are conservative in nature and are designed to generate corrective actions if the required testing or inspections are deemed unsatisfactory well in advance to ensure the continued safety function of the containment is maintained. The programmatic factors are designed to provide differing ways to test and/or examine the containment pressure boundary in a

Enclosure Evaluation of the Proposed Change 40 manner that verifies the Callaway containment pressure boundary will perform its intended safety function. Since the proposed change does not alter the configuration of the Callaway containment pressure boundary, continued performance of the tests and inspections associated with NEI 94-01 will only serve to ensure the continued safety function of the containment without affecting any margin of safety.

PRA Response:

The adequacy of the design feature (the containment boundary subject to Type A testing) is preserved as evidenced by the overall small change in risk associated with the Type A test frequency change.

3.

Preserve system redundancy, independence, and diversity commensurate with the expected frequency and consequences of challenges to the system, including consideration of uncertainty.

As stated in RG 1.174, Revision 3, Section C.2.1.1.1, Background, the defense-in-depth philosophy has traditionally been applied in plant design and operation to provide multiple means to accomplish safety functions. System redundancy, independence, and diversity result in high availability and reliability of the function and also help ensure that system functions are not reliant on any single feature of the design. Redundancy provides for duplicate equipment that enables the failure or unavailability of at least one set of equipment to be tolerated without loss of function. Independence of equipment implies that the redundant equipment is separate such that it does not rely on the same supports to function. This independence can sometimes be achieved by the use of physical separation or physical protection. Diversity is accomplished by having equipment that performs the same function rely on different attributes such as different principles of operation, different physical variables, different conditions of operation, or production by different manufacturers which helps reduce common-cause failure (CCF).

A proposed change might reduce the redundancy, independence, or diversity of systems.

The intent of this consideration is to ensure that the ability to provide the system function is commensurate with the risk of scenarios that could be mitigated by that function. The consideration of uncertainty, including the uncertainty inherent in the PRA, implies that the use of redundancy, independence, or diversity provides high reliability and availability and also results in the ability to tolerate failures or unanticipated events.

Response

The proposed change to extend the testing frequencies of the Type A ILRT from 10 years to 15 years and select Type C LLRTs from 60 months to 75 months does not reduce the redundancy, independence or diversity of systems. As shown in NUREG-1493, increasing the test frequency of ILRTs up to a 20-year test interval was found to lead to an imperceptible increase in risk. The estimated increase in risk is very small because ILRTs identify only a few potential containment leakage paths that cannot be identified by Type B

Enclosure Evaluation of the Proposed Change 41 or Type C testing. The study also concluded that extending the frequency of Type B tests is possible with no adverse impact on risk as identified leakage through Type B mechanical penetrations are both infrequent and small. Additionally, the study concluded that Types B and C tests could identify the vast majority (greater than 95 percent) of all potential leakage paths.

Despite the change in test interval, containment isolation diversity remains unaffected and will continue to provide the inherent isolation, as designed. In addition, NEI 94-01 Revisions 2-A and 3-A, Section 11.3.2 requires a schedule of tests be developed, for components on a test interval greater than 60 months, such that unanticipated random failures and unexpected common-mode failures are avoided. This is typically accomplished by implementing test intervals at approximately evenly distributed intervals.

Therefore, the proposed change preserves system redundancy, independence, and diversity and ensures a high reliability and availability of the containment structure to perform its safety function in the event of unanticipated events.

PRA Response:

The redundancy, independence, and diversity of the containment subject to the Type A test is preserved, commensurate with the expected frequency and consequences of challenges to the system, as evidenced by the overall small change in risk associated with the Type A test frequency change.

4.

Preserve adequate defense against potential common-cause failures (CCFs).

An important aspect of ensuring defense in depth is to guard against CCF. Multiple components may fail to function because of a single specific cause or event that could simultaneously affect several components important to risk. The cause or event may include an installation or construction deficiency, accidental human action, extreme external environment, or an unintended cascading effect from any other operation or failure within the plant. CCFs can also result from poor design, manufacturing, or maintenance practices. Defenses can prevent the occurrence of failures from the causes and events that could allow simultaneous multiple component failures. Another aspect of guarding against CCF is to ensure that an existing defense put in place to minimize the impact of CCF is not significantly reduced; however, a reduction in one defense can be compensated for by adding another.

Response

As part of the proposed change, Callaway will be required to adopt the performance-based testing standards outlined in NEI 94-01, Revisions 2-A and 3-A along with ANSI/ANS 56.8-2002. NEI 94-01, Revisions 2-A and 3-A, Section 11.3.2 requires a schedule of tests be developed, for components on test intervals greater than 60 months, such that unanticipated random failures and unexpected common-mode failures are avoided. This is typically accomplished by implementing test intervals at approximately

Enclosure Evaluation of the Proposed Change 42 evenly distributed intervals. In addition, components considered to be risk-significant from a PRA standpoint are required to be limited to a testing interval less than the maximum allowable limit of 75 months. For those components that have demonstrated satisfactory performance and have had their testing limits extended, administrative testing limits are assigned on a component-by-component basis and are used to identify potential valve or penetration degradation. Administrative limits are established at a value low enough to identify and should allow early correction in advance of total valve failure. Should a component exceed its administrative limit during testing, NEI 94-01, Revisions 2-A and 3-A require cause determinations be performed designed to reinforce achieving acceptable performance. The cause determination is designed to identify and address common-mode failure mechanisms through appropriate corrective actions. The proposed change also imposes a requirement to address margin management (i.e., margin between the current containment leakage rate and its pre-established limit). As a result, adoption of the performance-based testing standards proposed by this change ensures adequate barriers exist to preclude failure of the containment pressure boundary due to common-mode failures and therefore continues to guard against CCF.

PRA Response:

Adequate defense against CCFs is preserved. The Type A test detects problems in the containment which may or may not be the result of a CCF; such a CCF may affect failure of another portion of containment (i.e., local penetrations) due to the same phenomena.

Adequate defense against CCFs is preserved via the continued performance of the Types B and C tests and the performance of inspections. The change to the Type A test, which bounds the risk associated with containment failure modes including those involving CCFs, does not degrade adequate defense as evidenced by the overall small change in risk associated with the Type A test frequency change.

5.

Maintain multiple fission product barriers.

Fission product barriers include the physical barriers themselves (e.g., the fuel cladding, reactor coolant system pressure boundary, and containment) and any equipment relied on to protect the barriers (e.g., containment spray). In general, these barriers are designed to perform independently so that a complete failure of one barrier does not disable the next subsequent barrier. For example, one barrier, the containment, is designed to withstand a double-ended guillotine break of the largest pipe in the reactor coolant system, another barrier.

A plants licensing basis might contain events that, by their very nature, challenge multiple barriers simultaneously. Examples include interfacing-system LOCAs, SG tube rupture, or crediting containment accident pressure. Therefore, complete independence of barriers, while a goal, might not be achievable for all possible scenarios.

Enclosure Evaluation of the Proposed Change 43

Response

The purpose of the proposed change is to extend the testing frequencies of the Type A ILRT from 10 years to 15 years and select Type C LLRTs from 60 months to 75 months.

As part of the proposed change, Callaway will be required to adopt the performance-based testing standards outlined in NEI 94-01, Revisions 2-A and 3-A, along with ANSI/ANS 56.8-2002. The overall containment leakage rate calculations associated with the testing standards contain inherent conservatisms through the use of margin. Plant TS require the overall primary containment leakage rate to be less than or equal to 1.0 La.

NEI 94-01 requires the as-found (AF) Type A test leakage rate must be less than the acceptance criterion of 1.0 La given in the plant TS. Prior to entering a mode where containment integrity is required, the as-left (AL) Type A leakage rate shall not exceed 0.75 La. The AF and AL values are as determined by the appropriate testing methodology specifically described in ANSI/ANS 56.8-2002. Additionally, the combined leakage rate for all Type B and Type C tested penetrations shall be less than or equal to 0.6 La, determined on a maximum pathway basis from the as-left LLRT results prior to entering a mode where containment integrity is required. This regulatory approach results in a 25%

and 40% margin, respectively, to the 1.0 La requirements. For those local leak rate tested components that have demonstrated satisfactory performance and have had their testing limits extended, administrative testing limits are assigned on a component-by-component basis and are used to identify potential valve or penetration degradation. Administrative limits are established at a value low enough to identify and allow early correction in advance of total valve failure. Should a component exceed its administrative limit during testing, NEI 94-01, Revisions 2-A and 3-A, require cause determinations be performed designed to reinforce achieving acceptable performance. The cause determination is designed to identify and address common-mode failure mechanisms through appropriate corrective actions. Therefore, the proposed change adopts requirements with inherent conservatisms to ensure the margin to safety limit is maintained, thereby, preserving the containment fission product barrier.

PRA Response:

Multiple Fission Product barriers are maintained. The portion of the containment affected by the Type A test extension is still maintained as an independent fission product barrier, albeit with an overall small change in the reliability of the barrier.

6.

Preserve sufficient defense against human errors.

Human errors include the failure of operators to correctly and promptly perform the actions necessary to operate the plant or respond to off-normal conditions and accidents, errors committed during test and maintenance, and incorrect actions by other plant staff. Human errors can result in the degradation or failure of a system to perform its function, thereby significantly reducing the effectiveness of one of the layers of defense or one of the fission product barriers. The plant design and operation include defenses to prevent the

Enclosure Evaluation of the Proposed Change 44 occurrence of such errors and events. These defenses generally involve the use of procedures, training, and human engineering; however, other considerations (e.g.,

communication protocols) might also be important.

Response

Sufficient defense against human errors is preserved. Errors committed during testing and maintenance may be reduced by the less frequent performance of the Type A, Type B and Type C tests (less opportunity for errors to occur).

PRA Response:

Sufficient defense against human errors is preserved. The probability of a human error to operate the plant, or to respond to off-normal conditions and accidents is not significantly affected by the change to the Type A testing frequency. Errors committed during test and maintenance may be reduced by the less frequent performance of the Type A test (less opportunity for errors to occur).

7.

Continue to meet the intent of the plants design criteria.

For plants licensed under 10 CFR Part 50 or 10 CFR Part 52, the plants design criteria are set forth in the current licensing basis of the plant. The plants design criteria define minimum requirements that achieve aspects of the defense-in-depth philosophy; as a consequence, even a compromise of the intent of those design criteria can directly result in a significant reduction in the effectiveness of one or more of the layers of defense.

When evaluating the effect of the proposed licensing basis change, the licensee should demonstrate that it continues to meet the intent of the plants design criteria.

Response

The purpose of the proposed change is to extend the testing frequencies of the Type A ILRT from 10 years to 15 years and select Type C LLRTs from 60 months to 75 months.

The proposed extensions do not involve either a physical change to the plant or a change in the manner in which the plant is operated or controlled. As part of the proposed change, Callaway will be required to adopt the performance-based testing standards outlined in NEI 94-01, Revisions 2-A and 3-A along with ANSI/ANS 56.8-2002. The leakage limits imposed by plant TS remain unchanged when adopting the performance-based testing standards outlined in NEI 94-01, Revision 3-A and ANSI/ANS 56.8-2002.

Plant design limits imposed by the Updated Final Safety Analysis Report (UFSAR) also remain unchanged as a result of the proposed change. Therefore, the proposed change continues to meet the intent of the plants design criteria to ensure the integrity of the Callaway containment pressure boundary.

Enclosure Evaluation of the Proposed Change 45 PRA Response:

The intent of the plants design criteria continues to be met. The extension of the Type A test does not change the configuration of the plant or the way the plant is operated.

==

Conclusion:==

The responses to the seven Defense in Depth questions above conclude that the existing Defense in Depth has not been diminished; rather, in some instances has been increased.

Therefore, the proposed change does not comprise a reduction in safety.

3.5 Non-Risk Based Assessment Consistent with the defense-in-depth philosophy discussed in RG 1.174, Callaway has assessed other non-risk-based considerations relevant to the proposed amendment. Callaway has multiple inspection and testing programs that ensure the containment structure continues to remain capable of meeting its design functions and is designed to identify any degrading conditions that might affect that capability. These programs are discussed below.

3.5.1 Containment Building Coatings Containment Building Coatings procedure describes the responsibility and actions of Engineering, Work Control, Quality Control, and painters for the surface preparation of substrates to be painted, qualification of painters and inspection personnel, and the application and inspection of coatings and surfacers to equipment, systems, and structures for use in Containment.

This procedure implements the Protective Coatings Monitoring and Maintenance Aging Management Program by implementing required coating inspections described in Section 3.7 of this activity.

Exceptions Painting Category 3 - Small Equipment as defined in the FSAR, Engineered Safety Feature Materials and any temporary equipment which will not remain inside Containment during normal plant operation.

Coatings excluded from this procedure should be applied in accordance with procedure Field Coatings (Non-Containment Building Coatings).

Definitions x

Coating - a fluid which when applied to a surface, forms a solid continuous film or barrier

Enclosure Evaluation of the Proposed Change 46 by some physical or chemical means whether used as a prime, intermediate, or final coat.

x Coating System - one or several coating materials to be applied on a prepared surface in specific sequence (primer, intermediate, and final) with a required dry mil thickness for each coat.

x Coatings Specialist - Responsible Engineer or his designee.

x Containment Qualified Coating - A coating application to the Reactor Building interior or any equipment, structure, or component which will be permanently located inside containment, whether or not the coating is applied inside containment or a remote location and has been applied in accordance with Callaway procedure Containment Building Coatings with coating material designated by Callaway procedure Containment Building Coatings.

Responsibilities Coatings Specialist x

Reviews all changes, corrections, deletions, and additions to this procedure.

x Provides technical assistance to Work Control, Quality Control and contract personnel concerning the selection and application of coatings.

x Monitors coating application and coating performance and revise this procedure accordingly.

x Ensures verification of Nuclear Coating Specialist qualification is achieved.

x Reviews past Containment Building Coating Condition Assessments.

x Performs coating inspections and prepares the Containment Building Coating Condition Assessment.

Work Management, Planning x

Plans and processes all applicable coating work requests in accordance with this procedure and the Integrated Work Management Process Description.

Enclosure Evaluation of the Proposed Change 47 Director, Maintenance x

Ensures work performed by Ameren Missouri Maintenance is in accordance with this procedure and other applicable procedures.

x Ensures work performed by other than Ameren Missouri personnel is in accordance with this procedure and other applicable procedures.

General Ensure surface preparation of substrates and application of coatings meet the requirements of this procedure, and Attachments as applicable.

x All work NOT meeting these requirements:

o Document in accordance with Corrective Action Program.

o Repair to meet requirements.

o If work can NOT be brought into compliance with the above requirements, Engineering: Determine acceptability or corrective action.

x Engineering: Evaluate coatings applied to items which will be permanently located inside Containment and are NOT in compliance with the requirements of this procedure for impact on performance of Containment Recirculation Sumps.

x Document and track any coatings determined to be unqualified per requirements of this procedure on the Unqualified Containment Coatings Log.

Ensure coating work performed within Containment Building or on equipment or components to be installed in Containment Building is considered "Containment Qualified" except as exempt per exceptions above.

Ensure all coating work performed in Containment or on equipment, structures or components, which will be permanently located inside Containment, is inspected and documented by certified Quality Control personnel.

Ensure coating applicator reports daily on surface preparation conditions and application work during each shift for each area of work. This report shall be duly verified by Ameren Missouri and documented in the Work Request package.

Ensure coating materials used in implementing this procedure are only those approved and listed in this procedure, unless otherwise approved by Engineering. Refer to Storeroom Issue of Material, Components, and Equipment.

Enclosure Evaluation of the Proposed Change 48 Inspection of Coatings Inspect coated surfaces each refueling outage in accordance with ASTM D5163-08, Standard Guide for Establishing a Program for Condition Assessment of Coating Service Level I Coating Systems in Nuclear Power Plants (Reference 66).

Review the Callaway Plant Unqualified Containment Coatings Log and the previous two coating condition assessment reports prior to performing the coated surfaces inspection.

Ensure verification of Nuclear Coating Specialist qualification is achieved and that inspection personnel, the inspection coordinator, and the inspection results evaluator are properly qualified in accordance with ASTM D5163-08.

Ensure that inspection personnel have proper instruments and equipment needed for inspection, including, but not be limited to, flashlights, spotlights, marker pen, mirror, measuring tape, magnifier, binoculars, camera with or without wide angle lens, and self-sealing polyethylene sample bags.

Conduct a general visual inspection on all readily accessible coated surfaces during a walk-through. After a walk-through, or during the general visual inspection, carry out thorough visual inspections on previously designated areas and on areas noted as deficient during the walk-through.

x Inspect coated surfaces for any visible defects, such as blistering, cracking, flaking, peeling, rusting, and physical damage.

x Photograph defective or deficient coating surfaces.

x Document inspection activities with numerical identifiers on Reactor Building Grid Maps.

The locations of all defects and of all tests performed shall be recorded on the maps so that additional testing, recoating, and further monitoring may be performed. The maps shall also identify items/areas requiring special testing, if any.

x Document deficiencies in the Containment Coating Deficiencies Log in the Containment Building Coating Condition Assessment, as noted below:

o Blistering - Measure, record size and extent. Report if blistered portions are intact.

o Cracking - Measure the length of the crack or if extensive cracking has occurred, measure the size of the area affected. Record measurements and describe crack depth and pattern.

o Flaking/Peeling/Delamination - Measure the approximate size of the degraded coating area and note the pattern formed. Carefully test to see if lifting can be easily achieved

Enclosure Evaluation of the Proposed Change 49 beyond the obvious peeled area. Note all observations including location of failure within the coating film, whether the failure is cohesive or adhesive, etc.

o Rusting - Try to determine the source of rusting (that is, is it surface stain caused by rusting elsewhere, or is it a failure of the coating allowing the substrate to rust).

o Signs of physical damage to the coating - deficiencies noted above may indicate where coating damage has occurred.

o Conduct a thorough visual inspection on all coatings near sumps or screens associated with the ECCS.

o For coating surfaces determined to be suspect, defective, or deficient, physical tests, such as dry film thickness and adhesion may be performed when directed by the Nuclear Coating Specialist. Samples may be gathered, and the size and extent of defective patterns may be described.

Prepare the Containment Building Coating Condition Assessment.

x Ensure the assessment report prioritizes repair areas as either needing repair during the same outage or as postponed to future outages, but under surveillance in the interim period.

x Evaluate damaged, defective, or deficient coating surfaces in accordance with ASTM D5163-08 as specified below using the Containment Coating Deficiencies Log.

o Blistering - Compare any blistering found to the blistering pictorial standards of coatings defects and record size and frequency. Report if blistered portions are intact.

o Cracking - Cracking may be limited to the one layer of coating or extend through to the substrate. Determine if the cracking is isolated or is part of a pattern. Describe crack depth and pattern on the inspection report.

o Rusting - Compare with the pictorial standards to determine the degree of rusting.

o If no defects are found, mark Coating Intact, No Defects on the inspection report.

o If portions of the coating cannot be inspected, note the specific areas on the location map-inspection report, along with the reason why the inspection cannot be conducted.

x Provide written or photographic documentation, or both, of coating inspection areas, failures, and defects in the Containment Building Coating Condition Assessment.

Enclosure Evaluation of the Proposed Change 50 x

Ensure responsible Coating Specialist prepares a summary of findings and recommendations for future surveillance or repair.

Unqualified Containment Coating Log Summary of RF25 Changes The Unqualified Containment Coatings Log was updated after RF25 to account for unqualified coatings that were added, removed, replaced, or confirmed to be DBA Qualified Service Level 1 coatings.

Table 3.5.1-1, Unqualified Coatings Log - Post RF25 Update 06/20/2022 November 2015 Baseline Quantity of Unqualified Coatings in Containment (ft2) 23,778.70 ft2 Current Total Area of Unqualified Coatings in Containment (ft2) 22,105.00 ft2 Margin Gained (ft2) 1,673.70 ft2 One hydrogen mixing fan motor was replaced during RF25. Evaluation verified the replacement motor to be coated with DBA Qualified Service Level 1 coatings. This accounted for 153.8 square feet.

Added 13 square feet of unqualified coatings due to new refueling machine air compressor.

A total of 63.9 square feet of unqualified coatings was removed during Refuel 25.

Summary of RF24 Changes The Unqualified Containment Coatings Log was updated after Refuel 24 to account for unqualified coatings that were added, removed, replaced, or confirmed to be DBA Qualified Service Level 1 coatings.

Table 3.5.1-2, Unqualified Coatings Log - Post RF24 Update 01/26/2021 November 2015 Baseline Quantity of Unqualified Coatings in Containment (ft2) 23,778.70 ft2 Current Total Area of Unqualified Coatings in Containment (ft2) 22,168.90 ft2 Margin Gained (ft2) 1,609.80 ft2

Enclosure Evaluation of the Proposed Change 51 One containment cooler fan motor was replaced during RF24. Evaluation verified the replacement motor to be coated with DBA Qualified Service Level 1 coatings.

The check and isolation valves associated with the Water Hammer Mitigation Modification, were determined to not be coated with a DBA Qualified Service Level 1 Coating. These valves were installed during Refuel 23.

During the Cycle 24 review, an additional 508 square feet of unqualified coating was determined to exist on the polar crane and its control system.

3.5.2 Containment Pressure Boundary Inservice Inspection Program Introduction and Background The Containment Pressure Boundary Inservice Inspection (ISI) Program outlines the requirements for the Non-Destructive Examination of Callaways containment pressure boundary and related components as specified by 10 CFR 50.55a. These requirements include ASME Section XI, Subsection IWE, 2007 Edition with the 2008 Addenda and the applicable conditions in 10 CFR 50.55a(b)(2)(ix).

The first ten-year Containment Pressure Boundary ISI Program was initially developed in accordance with the 1992 Edition, 1992 Addenda of ASME Section XI Subsection IWE. Prior to performance of the initial inspection, Callaway submitted a relief request and received approval from the NRC to use the 1998 Edition of the ASME Section XI, Subsection IWE as supplemented by specific commitments, in lieu of the 1992 Edition, 1992 Addenda of the Code. Callaway's supplemental commitments are documented in letters of July 9, 1999, and September 10, 1999, which responded to the NRC's RAI. The supplemental commitments were adopted for Callaway's Containment Inspection Program for the first ten-year interval. The first ten-year interval commenced on September 9, 1996, and was completed on November 30, 2008.

The first interval included an Initial Inspection Period of 5 years, from September 9, 1996, to September 8, 2001; a First Inspection Period of approximately 1 year, September 9, 2001, to November 30, 2002; a Second Inspection Period of 3 years, December 1, 2002, to November 30, 2005; and, a Third Inspection Period of 3 years, December 1, 2005, to November 30, 2008.

Because the first interval covered more than 12 years, Ameren Missouri is planning to limit Intervals 2 and 3 to nine (9) years as allowed by IWA-2430 in order to revert to a cumulative 30-year total for Intervals 1 through 3.

The second interval began on December 1, 2008, and ended on November 30, 2017.

Examinations and tests during the second interval were conducted in accordance with ASME Section XI, 2001 Edition through the 2003 Addenda.

Enclosure Evaluation of the Proposed Change 52 Prior to the implementation of Section XI requirements in Interval 1, ISI of the Callaway Plant Containment Pressure Boundary was performed in accordance with 10 CFR 50, Appendix J requirements.

Basis of Containment Inspection Program-Interval 3 In the Federal Register (FR) dated June 21, 2011 (76 FR 36232), the NRC amended 10 CFR 50.55a to incorporate by reference the ASME Boiler and Pressure Vessel Code,Section XI, 2007 Edition with the 2008 Addenda. Subsection IWE of Section XI includes the Requirements for Class MC and Metallic Liners of Class CC Components of Light-Water Cooled Plants. The amended rule became effective on July 21, 2011.

As required by 10 CFR 50.55a(b)(2)(vi), Effective edition and addenda of Subsection IWE and Subsection IWL,Section XI, states, in part, that Successive 120-month interval updates must be implemented in accordance with paragraph (g)(4)(ii) of this section.

10 CFR 50.55a(g)(4)(ii) requires that Inservice examination of components and system pressure tests conducted during successive 120-month inspection intervals must comply with the requirements of the latest edition and addenda of the Code incorporated by reference in paragraph (b) of this section 12 months before the start of the 120-month inspection interval (or the optional ASME Code Cases listed in NRC Regulatory Guide 1.147, Revision 16, when using Section XI; or NRC Regulatory Guide 1.192, when using the OM Code, that are incorporated by reference in paragraphs (a)(3)(ii) and (iii) of this section), subject to the conditions listed in paragraph (b) of this section.

Based on a third interval start date of December 1, 2017, Callaway was required to implement the latest edition and addenda of the Code incorporated by reference in paragraph 10 CFR 50.55a(b) as of December 1, 2016. As noted above, the applicable version of ASME Section XI at this time was the 2007 Edition with the 2008 Addenda.

The 10 CFR 50.55a conditions applicable to the Callaway Plant Interval 3 Containment Pressure Boundary ISI Program are included in 10 CFR 50.55a(b)(2)(ix), Examination of metal containments and the liners of concrete containments, and are as follows:

10 CFR 50.55a(b)(2)(ix),Section XI condition: Metal containment examinations.

Applicants or licensees applying Subsection IWE, 2007 Edition through the latest addenda incorporated by reference in paragraph (a)(1)(ii) of this section, must satisfy the requirements of paragraphs (b)(2)(ix)(A)(2), (b)(2)(ix)(B) and (b)(2)(ix)(J) of this section.

10 CFR 50.55a(b)(2)(ix)(A) Metal containment (MC) examinations: First provision. For Class MC applications, the following apply to inaccessible areas.

10 CFR 50.55a(b)(2)(ix)(A)(2) For each inaccessible area identified for evaluation, the applicant or licensee must provide the following in the ISI Summary Report as required by IWA-6000:

Enclosure Evaluation of the Proposed Change 53 (i) A description of the type and estimated extent of degradation, and the conditions that led to the degradation; (ii) An evaluation of each area, and the result of the evaluation; and (iii) A description of necessary corrective actions.

10 CFR 50.55a(b)(2)(ix)(B) Metal containment examinations: Second provision.

When performing remotely the visual examinations required by Subsection IWE, the maximum direct examination distance specified in Table IWA-2211-1 may be extended and the minimum illumination requirements specified in Table IWA-2211-1 may be decreased provided that the conditions or indications for which the visual examination is performed can be detected at the chosen distance and illumination. [Note that the IWA Table number was corrected from IWA-2210-1 to IWA-2211-1 in a 2017 NRC FR change and the table correction is reflected here.] Note that this condition is being implemented by Callaway.

10 CFR 50.55a(b)(2)(ix)(J) Metal containment examinations: Tenth provision.

In general, a repair/replacement activity such as replacing a large containment penetration, cutting a large construction opening in the containment pressure boundary to replace steam generators, reactor vessel heads, pressurizers, or other major equipment; or other similar modification is considered a major containment modification. When applying IWE-5000 to Class MC pressure-retaining components, any major containment modification or repair/replacement must be followed by a Type A test to provide assurance of both containment structural integrity and leak-tight integrity prior to returning to service, in accordance with 10 CFR part 50, Appendix J, Option A or Option B on which the applicants or licensees Containment Leak-Rate Testing Program is based. When applying IWE-5000, if a Type A, B, or C Test is performed, the test pressure and acceptance standard for the test must be in accordance with 10 CFR part 50, Appendix J.

In addition to the 10 CFR 50.55a(b)(2)(ix) conditions, the following clarification is included in 10 CFR 50.55a(g)(6)(ii):

10 CFR 50.55a(g)(6)(ii)(B): Augmented ISI requirements: Submitting containment ISI programs.

Licensees do not have to submit to the NRC for approval of their containment inservice inspection programs which were developed to satisfy the requirements of Subsection IWE and Subsection IWL with specified conditions. The program elements and the required documentation must be maintained on site for audit.

Subsequent 10 CFR 50.55a Amendments:

Note that subsequent to the 10 CFR 50.55a amendment to incorporate the 2007 Edition with the 2008 Addenda, (76 FR 36232, discussed previously), 10 CFR 50.55a has been amended nine (9) times, with minor (or no) impact to the Interval 3 Containment Pressure Boundary ISI Program at Callaway as follows:

Enclosure Evaluation of the Proposed Change 54 (1)

Final Rule dated January 23, 2012 (77 FR 3073): Incorporated corrections to amendments to the Final Rule published June 21, 2011 (76 FR 36232) effective July 21, 2011. Specifically, this change made typographical, formatting and punctuation corrections.

(2)

Final Rule dated June 7, 2013 (78 FR 34245): Incorporated miscellaneous corrections.

(2)

Final Rule dated November 5, 2014 (79 FR 65775): Incorporated by reference the latest revisions of three RGs including RG 1.147, Revision 17, Inservice Inspection Code Case Acceptability, ASME Section XI, Division 1 and reformatted 10 CFR 50.55a.

(3)

Final Rule dated November 10, 2014 (79 FR 66598): Incorporated miscellaneous corrections including making several spelling corrections by adding hyphens to words in 10 CFR 50.55a(b)(2)(viii)(B) (pre-stressing) and 50.55a(b)(2)(ix)(J) (leak-tight).

(4)

Final Rule dated December 11, 2014 (79 FR 73461): Incorporated technical corrections, including adding three inadvertently omitted addenda to 10 CFR 50.55a(a)(1)(ii)(B) for Section XI.

(5)

Final Rule dated August 3, 2015 (80 FR 45841): Corrected Institute of Electrical and Electronics Engineers (IEEE) Standard reference.

(6)

Final Rule dated July 18, 2017 (82 FR 32934): Incorporated by reference, the 2013 Edition of Section XI and updated the conditions in 10 CFR 50.55a(b)(2). The effective date for this Final Rule was August 17, 2017. The NRC revised 50.55a(g)(4)(ii) to add an implementation period of 18 months for licensees whose ISI interval commences during the 12 through 18-month period after the publication of this Final Rule. Note that per 10 CFR 50.55a(g)(4)(iv), Callaway could optionally elect to implement the 2013 Edition of Section XI; however, based on the 3rd interval start date of December 1, 2017, the Callaway Subsection IWE Program was required to meet 10 CFR 50.55a(g)(4)(ii), which allows the inservice examination of components and system pressure tests conducted during successive 120-month inspection intervals to comply with the requirements of the latest edition and addenda of the ASME Code incorporated by reference in paragraph (b) of this section 12 months before the start of the 120-month inspection interval, which was the 2007 Edition with the 2008 Addenda.

(7)

Final Rule dated November 15, 2017 (82 FR 52823): Miscellaneous Corrections to 10 CFR 50, Parts 2, 9, 40, 50, 61, 71, 73, and 100. Correction to 10 CFR 50.55a was limited to paragraph (b)(2)(ix)(B), revising the reference to Table IWA-2210-1 to be Table IWA-2211-1. This correction is reflected in 10 CFR 50.55a(b)(2)(ix)(B) above.

(8)

Final Rule dated January 17, 2018 (83 FR 2331): Incorporated by reference the latest revisions of three RGs including RG 1.147, Revision 18, Inservice Inspection Code Case Acceptability, ASME Section XI, Division 1. Incorporation of this version of RG 1.147 is included.

Enclosure Evaluation of the Proposed Change 55 (9)

Final Rule dated January 18, 2018 (83 FR 2525): Issued a Correcting Amendment to the Final Rule published July 18, 2017. The corrections included the revisions in 10 CFR 50.55a(b)(2)(xviii) regarding implementation of Section XI Appendices VII and VIII related to certification requirements for non-destructive examination personnel and a revision to 10 CFR 50.55a(b)(3)(iv) regarding the implementation of the ASME OM Code for check valve condition monitoring. A review of these corrections concluded that they do not impact the Containment Pressure Boundary ISI Program.

The three inspection periods during the Third Containment Inspection Interval are as follows:

First Period (36 months):

December 1, 2017, to November 30, 2020 Second Period (36 months):

December 1, 2020, to November 30, 2023 Third Period (36 months):

December 1, 2023, to November 30, 2026 The three proposed inspection periods during the Fourth Containment Inspection Interval are as follows (fourth interval plan has yet to be developed):

First Period (36 months):

December 1, 2026, to November 30, 2030 Second Period (36 months):

December 1, 2030, to November 30, 2033 Third Period (36 months):

December 1, 2033, to November 30, 2036 ASME Section XI, Paragraph IWA-2430 allows the inspection interval to be increased or decreased by 12 months. This allows the inspection periods to be adjusted to coincide with Callaway refueling outages. In addition, Callaway is implementing Code Case N-765, which modifies IWA-2430(c), providing additional scheduling flexibility.

This program was developed in accordance with the requirements of the 2007 Edition with the 2008 Addenda of ASME Section XI, Subsections IWA and IWE as modified by 10 CFR 50.55a, applicable Relief Requests, Supplemental Commitments and Code Cases.

As allowed by 10 CFR 50.55a(c)(3) and NRC RG 1.147, Revision 20 certain ASME Section XI Code Cases have been determined acceptable for application to the ISI Program. Code Case N-532-5, "Repair/Replacement Activity Documentation Requirements and Inservice Inspection Summary Report Preparation and Submission Section XI, Division 1," has been adopted by Callaway and incorporated in the Containment Pressure Boundary ISI Program. The use of Code Cases or alternatives not currently identified in RG 1.147 are addressed as Relief Requests.

Primary Containment Design The Callaway Nuclear Power Plant Primary Containment Design is a metal lined, post tensioned and reinforced concrete containment structure. The Callaway Plant was granted an operating license on October 18, 1984, and commenced commercial operation on December 19, 1984.

Enclosure Evaluation of the Proposed Change 56 Callaways metallic liner received a waiver allowing construction not in accordance with Class MC component criteria. The liner is a carbon steel plate covering the entire inside surface of the reactor building (excluding penetrations). The liner is 1/4-inch thick but is thickened locally around the penetrations, large brackets, and major attachments. The liner plate, including the thickened plate, is anchored to the concrete structure. The vertical and dome liner plates were also used as forms for concrete placement. Attachments to the liner plate which transfer loads through the liner plate to the walls and base slab include equipment structural supports for internal structures and similar items. Major structural attachments to the wall, which penetrate the liner plate, include polar crane brackets, floor beam brackets, and pipe support brackets. Major structural attachments to the dome include various pipe support brackets. Miscellaneous thickened plates, which form a part of the liner plate, are provided and anchored in the concrete to provide supports. Leak chase channels and angles are also attached at seam welds where the welds are inaccessible to nondestructive examination after construction.

ASME Section XI, Subsection IWE Inservice Inspection The Subsection IWE Summary Table, Table 3.5.2-1 provides the following information:

Examination Category This column lists the examination category as identified in ASME Section XI, Table IWE-2500-l.

Only those examination categories applicable to Callaway are identified.

Item Number and Description of Components Examined These columns list the item number and description as defined in ASME Section XI, Table IWE-2500-1.

Number of Components This column lists the population of components potentially subject to examination. The number of components actually examined during the inspection interval will be based upon the Code requirements for the subject item number.

Examination Method The column lists the examination method(s) required by ASME Section XI, Table IWE-2500-1, Categories E-A, E-C and E-G.

Relief Request or Reference Number This column provides a listing of applicable relief requests and references. If a Relief Request number is identified, see the corresponding document in Section 4.4. If a Technical Approach and Position Paper is identified, see the corresponding document in Section 4.3. Otherwise, refer

Enclosure Evaluation of the Proposed Change 57 to the corresponding reference number for details (Request For Resolution (RFR), Callaway Condition Reports (CRs), Notes at the end of the table, etc.)

Enclosure Evaluation of the Proposed Change 58 Table 3.5.2-1, Subsection IWE Summary Table Examination Category Item Number Description of Components Examined Number of Components Examination Method Relief Request or Reference Number E-A Containment Surfaces E1.11 Accessible Surface Areas (not including areas embedded in concrete, covered with insulation, or wetted surfaces) 32 Zones General Visual Note 1 E1.11B Pressure Retaining Bolting 74 Connections General Visual Note 2 E1.12 Wetted Surfaces of Submerged Areas 0

VT-3 Note 6 E1.30 Moisture Barriers 0

General Visual Note 6 E-C Containment Surfaces Requiring Augmented Examination E4.11 Visible Surfaces 0

Visual, VT-1 Notes 4 and 5 E4.12 Surface Area Grid, Minimum Wall Thickness Location 0

Ultrasonic Thickness Notes 4 and 5 E-G Pressure Retaining Bolting E8.10 Bolted Connections 74 Connections Visual, VT-1 Note 3

Enclosure Evaluation of the Proposed Change 59 Notes:

1.

See Table IWE-2500-1, Notes 1(a), 1(b), and 1(c) for examination scope. Includes Zones 1 through 28 and 31 through 34.

2.

Pressure-retaining bolted connections, including bolts, studs, nuts, bushings, washers, and threads in base material and flange ligaments between fastener holes. Bolted connections need not be disassembled for performance of examinations, and bolting may remain in place under tension. A General Visual examination is required per Table IWE-2500-1.

3.

Examination shall include bolts, studs, nuts, bushings, washers, and threads in base material and flange ligaments between fastener holes. Examination may be performed with the connection assembled and bolting in place under tension, provided the connection is not disassembled during the interval. If the bolted connection is disassembled for any reason during the interval, the examination shall be performed with the connection disassembled. One VT-1 examination of the bolting in each connection must be performed, either disassembled or in place, during the interval.

4.

Containment surface areas requiring augmented examination are those identified in IWE-1240.

5.

The extent of examination shall be 100% for each inspection period until the areas examined remain essentially unchanged for the next inspection period. Such areas no longer require augmented examination in accordance with IWE-2420(d) and will then revert back to a Category E-A, Item No E1.11 classification.

6.

The Callaway containment design does not include wetted surfaces of submerged areas or moisture barrier.

Enclosure Evaluation of the Proposed Change 60 Requirements of ASME Section XI, Subsection IWE This section summarizes the requirements of ASME Section XI, Subsection IWE, which have been adopted for the Third Interval Containment Pressure Boundary ISI Program at Callaway.

In addition, this section summarizes Owner Elected examinations of Subsection IWE components. The alternative requirements presented are in accordance with ASME Section XI and 10 CFR 50.55a, as applicable.

Application of Exemption and Optional Criteria Exemption Criteria Per IWE-1220, the following components (or parts of components) are exempted from the examination requirements of IWE-2000:

x Vessels, parts, and appurtenances that are outside the boundaries of the containment as defined in the Design Specification; x

Embedded or inaccessible portions of containment vessels, parts, and appurtenances that met the requirements of the original Construction Code; x

Portions of containment vessels, parts, and appurtenances that become embedded or inaccessible as a result of vessel repair/replacement activities if the conditions of IWE-1232(a) and IWE-5220 are met; x

Piping, pumps, and valves that are part of the containment system, or which penetrate or are attached to the containment vessel. These components shall be examined in accordance with the rules of IWB or IWC, as appropriate to the classification defined by the Design Specifications.

Augmented Examinations of Subsection IWE Components Augmented Subsection IWE Components are those that meet the criteria in IWE-1241. These components shall be inspected in accordance with the guidance in IWE-1242. The components will continue to require augmented inspection until the criteria in IWE-2420(d) is met.

Currently, Callaway does not include any Subsection IWE components that require augmented examination.

Owner Elected Examinations of Subsection IWE Components Owner Elected Examinations are those that Callaway has determined to be necessary based on good engineering practice or other criteria. The following Owner Elected examinations have been scheduled for the Third IWE Interval.

x Containment Normal Sumps A and B, Zones 30 and 29, respectively, will be examined at least once during the interval. The liner plate within these sumps has been

Enclosure Evaluation of the Proposed Change 61 determined to be inaccessible and therefore exempt from examination per IWE-1220(b),

see Technical Approach and Position TAP-IWE-1 for details.

Adoption of Code Cases This Section addresses the adoption of Code Cases for the Third Containment Inservice Inspection Interval at Callaway. Code Cases adopted for ISI used during this interval are listed in Table 3.5.2-2. Code Cases for Repair/Replacement activities are not addressed in this document. In all cases, the use and adoption of Code Cases will be in accordance with ASME Section XI, IWA-2440 and 10 CFR 50.55a. The methodology for adopting Code Cases is divided into the four categories clarified below.

Adoption of Code Cases Listed for Generic Use in RG 1.147 Code Cases listed for generic use in RG 1.147, Revision 18, and later, will be adopted for use during the Third Inservice Inspection Interval by listing them in Table 4.2 of this Inservice Inspection Program. All conditions or limitations in RG 1.147 for a particular Code Case will apply or an appropriate Request for Relief will be submitted and approved by the NRC prior to use.

Adoption of Code Cases Not Listed for Generic Use in RG 1.147 Code Cases that have been approved by the Board of Nuclear Codes and Standards, but that have not been listed for generic use in RG 1.147, may be submitted in the form of a Relief Request in accordance with 10 CFR 50.55a(z).

Adoption of Code Cases Listed for Generic Use in RG 1.147 But Subsequently Annulled by ASME Section XI A Code Case may be adopted for use in accordance with RG 1.147 and subsequently be annulled by ASME Section XI (for example, due to being incorporated into a later edition or addenda). Such Code Cases, once adopted, can continue to be used through the remainder of the interval. Code Cases which have been annulled prior to adoption and which are not approved for use in the most current revision of RG 1.147 may not be used unless as an approved alternative per 10 CFR 50.55a(z).

Adoption of Code Cases Issued Subsequent to Issuing this Inservice Inspection Program Code Cases issued by ASME Section XI subsequent to issuing this Inservice Inspection Program will be proposed for use in amendments to this Program in accordance with 10 CFR 50.55a, RG 1.147, and ASME Section XI, IWA-2441.

Enclosure Evaluation of the Proposed Change 62 Table 3.5.2-2, List of Adopted Code Cases CODE CASE NO.

TITLE RG 1.147 REVISION DATE ADOPTED N-532-5 Repair/Replacement Activity Documentation Requirements and Inservice Summary Report Preparation and Submission Section XI, Division 1 18 02/16/2018 N-765 Alternative to Inspection Interval Scheduling Requirements of IWA-2430,Section XI, Division 1 18 02/16/2018 Inservice Inspection Technical Positions This section provides guidance and clarification as to how ASME Section XI Subsection IWE will be applied at Callaway where Relief Requests are not required.

Table 3.5.2-3, Inservice Inspection Technical Positions Position Rev.

Summary TAP-IWE-1 0

Criteria For Identification of Inaccessible Surfaces Technical Approach and Position Number: TAP-IWE-1 Component Identification Code Class:

Class MC Components and Metallic Liners of Class CC Components

Reference:

Table IWE-2500-1 Exam Category:

E-A Item Number:

E1.10, E1.11

==

Description:==

Accessible Surface Areas Code Requirement Table IWE-2500-1, Category E-A, Item Numbers E1.10 and E1.11 requires all accessible interior and exterior surface areas Class MC components, parts and appurtenances and metallic shell and penetration liners of Class CC components to be 100% examined during each inspection period by the General Visual examination method per Note 1.

Enclosure Evaluation of the Proposed Change 63 Position Callaway will utilize the following criteria to determine whether the surface areas will be considered accessible and thus subject to the General Visual examinations required by Table IWE-2500-1.

x IWE-1220(b) and IWE-1220(c) allow components (or parts of components such as surfaces) to be exempted from the examination requirements of IWE-2000, if it can be demonstrated that these items are embedded or inaccessible x

IWE-1232(c) states, "Surface areas of Class MC containment vessels, parts and appurtenances, and surface areas of Class CC metallic shell and penetration liners may be considered inaccessible if visual access by line of sight with adequate lighting from permanent vantage points is obstructed by permanent plant structures, equipment, or components, provided these surface areas do not require examination in accordance with the inspection plan or IWE-1240."

x IWE-2310(c) states, "Visual examination shall be performed directly or remotely, by line of sight from available viewing angles from floors, platforms, walkways, ladders, or other permanent vantage points, unless temporary access is required by the inspection plan."

x ASME has determined that removal or disassembly of plant structures, equipment, or components in order to make surface areas accessible for visual examination is not required provided the examination areas do not require examination in accordance with the inspection plan or IWE-1240, see Interpretation XI-1-10-02.

Accordingly, based on the criteria in IWE-2310(c) and Interpretation XI-1-10-02, the Callaway Containment Normal Sump A and B liner surfaces, Zones 30 and 29, respectively, are considered exempt from examination per IWE-1220(b) based on being inaccessible due to the visual obstruction of the sump pumps, associated piping, and cover plates.

Containment Inspection Relief Requests This section contains Relief Requests written in accordance with 10 CFR 50.55a(g)(5) when specific ASME Section XI requirements for ISI are considered impractical. The enclosed Relief Requests are subject to change throughout the inspection interval. If examination requirements are determined to be impractical during the course of the interval, additional or modified relief requests shall be submitted in accordance with 10 CFR 50.55a(g)(5).

Exceptions to Code required examinations may also be authorized by the NRC, as allowed by 10 CFR 50.55a(z), provided that design, fabrication, installation, testing, and inspection performed in compliance with applicable Codes and Section XI requirements would result in hardship without a compensating increase in the level of quality and safety, or provided that the proposed alternative examination will assure an acceptable level of quality and safety. Specific exceptions may also be documented in the form of Relief Requests and included in this Section, as applicable.

Enclosure Evaluation of the Proposed Change 64 Relief Requests for incomplete examinations shall be submitted in accordance with 10 CFR 50.55a(g)(5)(iv) throughout the interval as limitations are identified. Due to ongoing changes in nondestructive examination procedures, techniques and requirements, Ameren Missouri considers that submitting Relief Requests for incomplete examinations when they are evaluated will provide a more accurate representation of the limitations.

Table 3.5.2-4, Containment Inspection Relief Requests Relief Request Summary Rev.

No IWE Relief Requests have been submitted at this time.

Use of Subsequent Editions and Addenda of ASME Section XI In accordance with 10 CFR 50.55a(g)(3)(v), components (including supports) may meet the requirements set forth in subsequent editions of Codes and Addenda, or portions thereof, which are incorporated by reference in 10 CFR 50.55a(b), subject to the limitations and modifications listed therein. This Section of the Inservice Inspection Program provides for alternative requirements from approved subsequent Code editions and addenda that may be adopted during the Third Containment Inservice Inspection Interval. Specific requirements for adoption of later Editions and Addenda of ASME Section XI are included in NRC Regulatory Issue Summary (RIS) 2004-12, Clarification on Use of Later Editions and Addenda to the ASME OM Code and Section XI, dated July 28, 2004 (Reference 73). This Inservice Inspection Program will be amended for adoption of subsequent Code rules, as applicable.

Records and Reports Records and Reports will be prepared and maintained in accordance with IWA-6000, and Subsection IWE of ASME Section XI, 2007 Edition with the 2008 Addenda, as modified by Code Case N-532-5 and 10 CFR 50.55a.

3.5.3 Containment Exterior Concrete and Tendon Inspection Program Introduction and Background This Program outlines the requirements for the Non-Destructive Examination and Testing of Callaways concrete containment and post-tensioning system as specified by 10 CFR 50.55a.

These requirements include ASME Section XI, Subsection IWL, 2007 Edition with the 2008 Addenda and the applicable conditions in 10 CFR 50.55a(b)(2)(viii).

The first ten-year Containment Exterior Concrete and Tendon Inspection Program at Callaway, which included the 15th and 20th Year Inspections, was based on the 1992 Edition with 1992

Enclosure Evaluation of the Proposed Change 65 Addenda of ASME Section XI, Subsection IWL and Relief Request ULNRC 3934 (Reference 70). The requirement to implement Subsection IWL was based on a Final Rule published August 8, 1996 (61 FR 41303), amending 10 CFR 50.55a, Codes and standards. The Final Rule was effective on September 9, 1996, and required licensees to incorporate the new requirements into their ISI Plans and to complete the first period containment inspections by September 9, 2001. Based on these requirements, Callaway conducted the first ten-year inspection interval from September 9, 1996, to September 8, 2006. The First 5-year Period Inspections (Year 15) were performed between September 9, 1996, and September 8, 2001, and the Second 5-year Period Inspections (Year 20) were performed between September 9, 2001, and September 8, 2006.

Prior to the implementation of Section XI requirements in Interval 1, Inservice Inspection of the Callaway Containment Exterior Concrete and Tendons was performed in accordance with Specification C-1003(Q), which implemented the requirements of RG 1.35, Inservice Inspection of Ungrouted Tendons in Prestressed Concrete Containments, Proposed Revision 3, dated April 1979 (Reference 68), and Proposed RG 1.35.1, Determining Prestressing Forces for Inspection of Prestressed Concrete Containments, dated April 1979 (Reference 69). During this timeframe, Callaway conducted the inspections for Years 1.5, 3.5, 5.5 and 10, as discussed in FSAR Section 16.6.1.2.1, which has since been deleted. The inspection scheduling discussed in FSAR Section 16.6.1.2.1 is based on the Containment Structural Integrity Test performed in January 1984 and has been performed approximately every five years since the Year 5.5 inspection.

Note that 61 FR 41303 included a provision in 10 CFR 50.55a(g)(6)(ii)(B)(4) which states, "The requirement for the expedited examination of the containment post-tensioning system may be satisfied by the post-tensioning system examinations performed after September 9, 1996, as a result of licensee post-tensioning system programs accepted by the NRC prior to September 9, 1996." Based on this provision, Callaway performed tendon selection and examination in accordance with RG 1.35 rather than Section XI, Subsection IWL requirements for the Year 15 Inspections.

As allowed by the NRC Final Rule published September 22, 1999 (64 FR 51370), the Callaway scheduling of Section XI, Subsection IWL examinations and tests will be based on the date that the first Section XI examinations and tests were performed, June 1999, rather than the Structural Integrity Test date, January 1984. This approach allowed Callaway to maintain the previously established 5-year schedule as the plant transitioned to Section XI, Subsection IWL requirements, as discussed in 64 FR 51370, page 51384.

Note that the containment inspections continue to be performed in accordance with the requirements of Specification C-1003(Q), which has been revised to incorporate the applicable Section XI and 10 CFR 50.55a requirements.

Enclosure Evaluation of the Proposed Change 66 Basis of Containment Inspection Program - Interval 3 In the Federal Register dated June 21, 2011 (76 FR 362324), the NRC amended 10 CFR 50.55a to incorporate by reference the ASME Boiler and Pressure Vessel Code,Section XI, 2007 Edition with the 2008 Addenda. Subsection IWL of Section XI includes the requirements for inservice inspection (ISI) of Class CC Concrete Components of Light-Water Cooled Plants.

The amended rule became effective on July 21, 2011.

As required by 10 CFR 50.55a(b)(2)(vi), Effective edition and addenda of Subsection IWE and Subsection IWL,Section XI, Successive 120-month interval updates must be implemented in accordance with paragraph (g)(4)(ii) of this section.

10 CFR 50.55a(g)(4)(ii) requires that Inservice examination of components and system pressure tests conducted during successive 120-month inspection intervals must comply with the requirements of the latest edition and addenda of the ASME Code incorporated by reference in paragraph (b) of this section 12 months before the start of the 120-month inspection interval (or the optional ASME Code Cases listed in NRC Regulatory Guide 1.147, when using ASME BPV Code,Section XI, or NRC Regulatory Guide 1.192, when using the ASME OM Code, as incorporated by reference in paragraphs (a)(3)(ii) and (iii) of this section), subject to the conditions listed in paragraph (b) of this section.

Based on a start-of-third interval date of December 1, 2017, Callaway was required to implement the latest edition and addenda of the Code incorporated by reference in paragraph 10 CFR 50.55a(b) as of December 1, 2016. As noted above, this version of ASME Section XI was the 2007 Edition with the 2008 Addenda.

The 10 CFR 50.55a conditions applicable to the Callaway Interval 3 Containment Exterior Concrete and Tendon Inspection Program are included in 10 CFR 50.55a(b)(2)(viii) and are as follows:

10 CFR 50.55a(b)(2)(viii)(E):

For Class CC applications, the licensee shall evaluate the acceptability of inaccessible areas when conditions exist in accessible areas that could indicate the presence of or result in degradation to such inaccessible areas. For each inaccessible area identified, the licensee shall provide the following in the ISI Summary Report required by IWA-6000:

(1)

A description of the type and estimated extent of degradation, and the conditions that led to the degradation; (2)

An evaluation of each area, and the result of the evaluation, and; (3)

A description of necessary corrective actions.

In addition to the 10 CFR 50.55a(b)(2)(viii) conditions, the following clarification is included in 10 CFR 50.55a(g)(6)(ii):

Enclosure Evaluation of the Proposed Change 67 10 CFR 50.55a(g)(6)(ii)(B):

Licensees do not have to submit to the NRC staff for approval of their containment ISI programs, which were developed to satisfy the requirements of Subsection IWE and Subsection IWL with specified modifications and limitations. The program elements and the required documentation must be maintained on site for audit.

Subsequent 10 CFR 50.55a Amendments:

Note that subsequent to the 10 CFR 50.55a amendment to incorporate the 2007 Edition with the 2008 Addenda, (76 FR 36232, discussed previously), 10 CFR 50.55a has been amended nine (9) times, with minor impact to the Interval 3 Containment Exterior Concrete and Tendon Inspection Program at Callaway:

(1)

Final Rule dated January 23, 2012 (77 FR 14): Incorporated correcting amendments to the Final Rule published June 21, 2011 (76 FR 362324).

(2)

Final Rule dated June 7, 2013 (78 FR 34245): Incorporated miscellaneous corrections.

(3)

Final Rule dated November 5, 2014, (79 FR 65776): Incorporated by reference Regulatory Guide 1.147, Revision 17, Inservice Inspection Code Case Acceptability, ASME Section XI, Division 1 and reformatted 10CFR50.55a.

(4)

Final Rule dated December 11, 2014, (79 FR 73461): Incorporated technical corrections.

(5)

Final Rule dated August 3, 2015, (80 FR 148): Corrected Institute of Electrical and Electronics Engineers (IEEE) Standard reference.

(6)

Final Rule dated July 18, 2017, (82 FR 32934): Incorporated by reference the 2013 Edition of Section XI and updated the conditions in 10CFR50.55a(b)(2). The effective date for this Final Rule was August 17, 2017. Note that per 10 CFR 50.55a(g)(4)(iv),

Callaway could optionally elect to implement the 2013 Edition of Section XI; however, based on the third interval start date of December 1, 2017, the Callaway Subsection IWL Program was required to meet 10 CFR 50.55a(g)(4)(ii), which allows the inservice examination of components and system pressure tests conducted during successive 120-month inspection intervals to comply with the requirements of the latest edition and addenda of the ASME Code incorporated by reference in paragraph (b) of this section 12 months before the start of the 120-month inspection interval, which was the 2007 Edition with the 2008 Addenda.

(7)

Final Rule dated November 15, 2017, (82 FR 52823): Miscellaneous Corrections to 10 CFR 50, Parts 2, 9, 40, 50, 61, 71, 73, and 100. Correction to 10CFR50.55a was limited to paragraph (b)(2)(ix)(B), revising the reference to Table IWA-2210-1 to be Table IWA-2211-1.

Enclosure Evaluation of the Proposed Change 68 (8)

Final Rule dated January 17, 2018, (83 FR 2331): Incorporated by reference Regulatory Guide 1.147, Revision 18, Inservice Inspection Code Case Acceptability, ASME Section XI, Division 1. Incorporation of this version of RG 1.147 is included in Revision 4 of this document.

(9)

Final Rule dated January 18, 2018, (83 FR 2525): Issued a Correcting Amendment to the Final Rule published July 18, 2017. The corrections included the condition in 10 CFR 50.55a(b)(2)(xviii) regarding implementation of Section XI Appendices VII and VII and the condition in 10 CFR 50.55a(b)(3)(iv) regarding the implementation of the ASME OM Code for check valve monitoring. A review of these corrections concluded that they do not impact the Containment Exterior Concrete and Tendon Inspection Program.

The two inspection periods during the third containment inspection interval are as follows:

First Period (5 years):

September 9, 2016, to September 8, 2021 (35th Year Inspection)

Second Period (5 years):

September 9, 2021, to September 8, 2026 (40th Year Inspection)

The two inspection periods during the fourth containment inspection interval are as follows (fourth interval plan has yet to be developed):

First Period (5 years):

September 9, 2026, to September 8, 2031 (45th Year Inspection)

Second Period (5 years):

September 9, 2031, to September 8, 2036 (50th Year Inspection)

ASME Section XI, Subarticles IWL-2410 and IWL-2420 allow the extension of the concrete and post-tensioning system examination by 12 months, if necessary.

In addition to the requirements of Subsection IWL of the 2007 Edition, 2008 Addenda, and applicable 10 CFR 50.55a conditions, the Interval 3 Containment Exterior Concrete and Tendon Inspection Program at Callaway is also subject to applicable Subsection IWA requirements, Code Cases and Relief Requests.

Inservice Inspection Exemptions and Summary Table This section provides a summary listing of all items subject to inservice inspection, as well as those items exempt from examination.

These items are exempt from the requirements of IWL-2000:

x IWL-1220(a) tendon end anchorages that are inaccessible, subject to the requirements of IWL-2521.1.

x IWL-1220(b) portions of the concrete surface that are covered by the liner, foundation material, or backfill, or are otherwise obstructed by adjacent structures, components parts or appurtenances.

Enclosure Evaluation of the Proposed Change 69 ASME Section XI Inservice Inspections Table 3.5.3-1, Subsection IWL Inservice Inspection Summary table provides the following information:

Examination Category This column lists the examination category as identified in ASME Section XI, Table IWL-2500-1.

Only those examination categories applicable to Callaway are identified.

Item Number and Description of Components Examined These columns list the item number and descriptions as defined in ASME Section XI, Table IWL-2500-1. Only those item numbers applicable to Callaway are identified.

Component This column lists the components selected for examination during Interval 3, Inspection Years 35 and 40. The following text provides a brief explanation of the component selection process.

The total population of Subsection IWL components subject to examination is:

x Concrete Surfaces-1, x

Vertical Tendons-86, x

Horizontal Tendons-165 For concrete surfaces, no selection process is specified in Subsection IWL. The examination requirements are included in IWL-2510(a), IWL-2510(b) and IWL-2510(c), and exclude surfaces exempted per IWL-1220(a) and IWL-1220(b). Examination of concrete surfaces is required every 5 years as discussed in IWL-2410.

The requirements for selection of tendons for examination are included in IWL-2521. At Callaway, the tendon population includes two (2) horizontal tendons, 23CB and 24AC, that are completely inaccessible and are therefore considered exempt per IWL-2521.1. As discussed in IWL-2521.1(a), these tendons will be removed from the sample population after the random selection of IWL-2521 is completed if they are randomly selected. Also, if these inaccessible tendons are randomly selected, Callaway will select substitute tendons per IWL-2521(b);

however, tendons, 23CB and 24AC will not be examined per IWL-2521.1(c).

The Callaway post-tensioning system design also includes four (4) horizontal tendons, 7CB, 8AC, 14CB, and 15AC, which are accessible only for visual inspection and are therefore considered exempt per IWL-2521.1. As discussed in IWL-2521.1(a), these tendons will be removed from the sample population after the random selection of IWL-2521 is completed, if

Enclosure Evaluation of the Proposed Change 70 they are randomly selected. Also, if these inaccessible tendons are randomly selected, Callaway will select substitute tendons per IWL-2521(b) and perform tendon end anchorage examinations (IWL-2524 and IWL-2525) per IWL-2521(c).

The tendon selection process and results are as follows Vertical Tendons: Per Table IWL-2521-1, 2% of the 86 Vertical Tendons are required to be examined in the 35th and 40th Year Inspection Periods. Two percent (2%) of the 86 tendons is 1.72 tendons; however, Table IWL-2521-1 requires a minimum of three (3) tendons to be examined.

For the 35th Year Inspection period, Callaway has selected tendons V39, V65 and V81 for examination. Per IWL-2521(b), tendon V65 has been designated the common tendon. Per IWL-2523.1, tendon V39 has been selected for wire and strand sample examination and testing.

For the 40th Year Inspection period, Callaway has selected tendons V24, V59 and V65 for examination. Per IWL-2521(b), tendon V65 has been designated the common tendon. Per IWL-2523.1, tendon V24 has been selected for wire and strand sample examination and testing.

Horizontal Tendons: Per Table IWL-2521-1, 2% of the 165 Horizontal Tendons are required to be examined in the 35th and 40th Year Inspection Periods. Two percent (2%) of the 165 tendons is 3.30 tendons which is rounded up to 4 tendons per Table IWL-2521-1, Note 1.

For the 35th Year Inspection period, Callaway has selected tendons 31AC, 25CB, 45BA and 49CB for examination. Per IWL-2521(b), tendon 45BA has been designated the common tendon. Per IWL-2523.1, tendon 25CB has been selected for wire and strand sample examination and testing.

For the 40th Year Inspection period, Callaway has selected tendons 33BA, 42BA, 45BA, and 47AC for examination. Per IWL-2521(b), tendon 45BA has been designated the common tendon. Per IWL-2523.1, tendon 42BA has been selected for wire and strand sample examination and testing.

In accordance with IWL-2521(a), tendons to be examined during an inspection shall be selected on a random basis except as noted in IWL-2521(b), (c), and (d), and IWL-2521.2. In order to assure the selection of tendons was random, the Callaway tendons which have not been previously examined (or are not scheduled for examination) were assigned random numbers.

The tendons were then sorted in ascending order, and the tendons with the lowest random numbers were selected for the 35th and 40th Year Inspections.

This selection methodology will also be used in the event that additional random selections are required per Table IWL-2521-1, Note 2. In addition, future tendon selections (Year 45 and later)

Enclosure Evaluation of the Proposed Change 71 will utilize the same selection methodology if supported by applicable 10 CFR 50.55a and Section XI requirements General Notes On Tendon Selection:

1)

Selection of substitute tendons per IWL-2521.1(b) is performed by the Callaway Staff and approved by the Responsible Engineer.

2)

Tendons that have been selected randomly or selected as substitutes per IWL-2521 shall be removed from the tendon population from which future (35th Year Inspections and later) random selections are made per IWL-2521(a).

Test or Examination Method The column lists the test or examination method(s) required by ASME Section XI, Table IWL-2500-1.

Inspection Period This column indicates the inspection period, (applicable Inspection Year) for the components to be examined

Enclosure Evaluation of the Proposed Change 72 Table 3.5.3-1, Subsection IWL Inservice Inspection Summary Table Examination Category Item Number Description of Components Examined Component Test or Examination Method Inspection Period (Note 2)

L-A Concrete Surface L1.10 Concrete surface, including coated areas, shall be visually examined for evidence of conditions which may be indicative of damage or degradation. Note 1 1

General Visual, IWL-2510 35th and 40th Years L-B Unbonded Post-Tensioning System L2.10 Tendon (Includes Force and Elongation Measurement) 25CB, 31AC, 45BA, 49CB, V39, V65, V81 IWL-2522 35th Year L2.20 Wire 25CB, V39 IWL-2523.2 35th Year L2.30 Anchorage Hardware and Surrounding Concrete All 7 Detailed Visual, IWL-2524 35th Year L2.40 Corrosion Protection Medium All 7 IWL-2525.2(a) 35th Year L2.50 Free Water All 7 IWL-2525.2(b) 35th Year L2.10 Tendon (Includes Force and Elongation Measurement) 33BA, 42BA, 45BA, 47AC, V24, V59, V65 IWL-2522 40th Year L2.20 Wire 42BA, V24 IWL-2523.2 40th Year L2.30 Anchorage Hardware and Surrounding Concrete All 7 Detailed Visual, IWL-2524 40th Year L2.40 Corrosion Protection Medium All 7 IWL-2525.2(a) 40th Year L2.50 Free Water All 7 IWL-2525.2(b) 40th Year

Enclosure Evaluation of the Proposed Change 73 Notes:

1) Includes concrete surfaces at tendon anchorage areas not selected by IWL-2521 or exempt per IWL-1220(a).

2) The inspection plan for the 45th and 50th years has not been developed at this time.

Enclosure Evaluation of the Proposed Change 74 Alternative Requirements to Section XI, 2007 Edition with the 2008 Addenda This section lists the alternative requirements to ASME Section XI, 2007 Edition, 2008 Addenda being adopted for the Third Containment Inservice Inspection Interval Program at Callaway Plant. The alternative requirements presented are in accordance with ASME Section XI and 10 CFR 50.55a, as applicable.

Adoption of Code Cases This Section addresses the adoption of Code Cases during the Third Containment Inservice Inspection Interval at Callaway. Code Cases adopted for ISI use during this interval will be listed in Table 3.5.3-2. Code Cases for Repair/Replacement activities are not addressed in this Inspection Program. In all cases, the use and adoption of Code Cases will be in accordance with ASME Section XI, IWA-2440, 10 CFR 50.55a and RG 1.147.

Table 3.5.3-2, List of Adopted Code Cases Code Case No.

Title Reg Guide 1.147 Revision Date Adopted N-532-5 Alternative Requirements to Repair and Replacement Documentation Requirements and Inservice Summary Report Preparation and Submission as Required by IWA-4000 and IWA-6000,Section XI, Division 1 18 02/16/2018 Inservice Inspection Technical Approach and Positions This section provides guidance and clarification as to how ASME Section XI Subsection IWL will be applied at Callaway where Relief Requests are not required.

Table 3.5.3-3, List of Technical Approach and Positions Position Rev.

Summary TAP-IWL-1 0

10 CFR 50.55a(b)(2)(viii) includes one mandatory (1) condition to ASME Section XI, Subsection IWL, the 2007 Edition with 2008 Addenda. This Technical Approach and Position provides Callaway implementation details for this condition.

Enclosure Evaluation of the Proposed Change 75 Technical Approach and Position Number: TAP-IWL-1 Revision 0 Component Identification Code Class:

Class CC Components

Reference:

10 CFR 50.55a, Codes and Standards, dated January 1, 2018 Code or Regulatory Requirement The January 1, 2018, version of 10 CFR 50.55a includes one mandatory (1) condition applicable to ASME Section XI, Subsection IWL, the 2007 Edition with 2008 Addenda. This mandatory condition is included in 10 CFR 50.55a(b)(2)(viii) and will be implemented at Callaway as discussed in the following Position statement:

Callaway Technical Approach and Position (1) 10 CFR 50.55a(b)(2)(viii)(E): For Class CC applications, the licensee shall evaluate the acceptability of inaccessible areas when conditions exist in accessible areas that could indicate the presence of or result in degradation to such inaccessible areas. For each inaccessible area identified, the licensee shall provide the following in the ISI Summary Report required by IWA-6000:

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

An evaluation of each area, and the result of the evaluation, and; x

A description of necessary corrective actions.

Callaway Technical Approach and Position: The ISI Summary Report required by IWA-6000 shall be submitted to the Authorized Nuclear Inservice Inspector (ANII) and regulatory authorities as required. This report will be in the form of an Owners Activity Report (OAR-1),

prepared in accordance with Code Case N 532-5. Attached to the OAR-1 will be the Final Containment Inspection Report which contains all applicable inspection data, results, acceptance criteria, and evaluations for the degraded area(s) identified.

Use of Subsequent Editions of ASME Section XI In accordance with the January 1, 2018, Edition of the Code of Federal Regulations, 10 CFR 50.55a(g)(3)(v), components (including supports) may meet the requirements set forth in subsequent editions of Codes and Addenda, or portions thereof, which are incorporated by reference in 10 CFR 50.55a(b), subject to the conditions listed therein. This Section of the Inservice Inspection Plan provides for alternative requirements from approved subsequent Code editions that may be adopted during the Third Containment Inservice Inspection Interval. This Inservice Inspection Plan will be amended for adoption of subsequent Code rules.

Enclosure Evaluation of the Proposed Change 76 Containment Inspection Relief Requests This section contains Relief Requests written in accordance with 10 CFR 50.55a(g)(5) when specific ASME Section XI requirements for inservice inspection are considered impractical. The Relief Requests are listed in Table 3.5.3-4 and are subject to change throughout the inspection interval. If examination requirements are determined to be impractical during the course of the interval, additional or modified relief requests shall be submitted in accordance with 10 CFR 50.55a(g)(5).

Exceptions to Code required examinations may also be authorized by the NRC, Director Office of Nuclear Reactor Regulation (NRR), as allowed by 10 CFR 50.55a(z), provided that design, fabrication, installation, testing and inspection performed in compliance with Codes and Section XI requirements would result in hardship without a compensating increase in the level of quality and safety, or provided that the proposed alternative examination will assure an acceptable level of quality and safety. Specific exceptions may also be documented in the form of Relief Requests and included in this Section, as applicable. These Relief Requests are also listed in Table 3.5.3-4 below and are subject to change throughout the inspection interval.

Relief Requests for incomplete examinations shall be submitted in accordance with 10 CFR 50.55a(g)(5)(iv) throughout the interval as limitations are identified. Due to ongoing changes in nondestructive examination procedures, techniques and requirements, Ameren Missouri considers that submitting Relief Requests for incomplete examinations when they are evaluated will provide a more accurate representation of the limitations.

Table 3.5.3-4, List of Relief Requests Relief Request Summary Rev.

Status No Interval 3 Relief Requests Have Been Issued to Date Records and Reports General Record and Reports will be prepared and maintained in accordance with IWA-6000, IWL-3300, IWL-4000, and IWL-5000 of the ASME Section XI, 2007 Edition with the 2008 Addenda, as modified by Code Case N-532-5 and 10 CFR 50.55a.

Enclosure Evaluation of the Proposed Change 77 3.5.4 Results of Recent Inspections RF24 IWE Activities Inspection Zones 1 through 28 and 31 through 34 were inspected with no defects noted.

IWE General inspection of the Incore Pit.

Several areas of coating distress were noted. Areas were cleaned and scraped to expose the liner for inspection. The liner is free of defects and accepted. QC Initiated CR 202005276 and Job 20003148 to evaluate and repair the degraded coatings.

RF25 IWE Activities P022 - Reactor Coolant Pump 'B' Seal Water Supply Containment Penetration VT-1 Inspection of the Reactor Coolant Pump 'B' Seal Water Supply Containment Penetration (P022) was completed satisfactorily.

P058 - Accumulator Fill line From SI Pump P-EM-01A Containment Penetration VT-1 Inspection of the Accumulator Fill Line from SI Pump P-EM-01A Containment Penetration (P058) was completed satisfactorily.

P064 - Nuclear Sample LYS Loop 3 Hot Leg Sample Pressurizer Liquid Sample Containment Penetration VT-1 Inspection of the Nuclear Sample LYS Loop 3 Hot Leg Sample Pressurizer Liquid Sample Containment Penetration (P064) was completed satisfactorily.

P069 - Pressurizer Vapor Sample Containment Penetration VT-1 Inspection of the Pressurizer Vapor Sample Containment Penetration (P069) was completed satisfactorily.

P091 - Reactor Vessel Level Instrumentation System Containment Penetration VT-1 Inspection of the Reactor Vessel Level Instrumentation System Containment Penetration (P091) was completed satisfactorily.

P093 - Reactor Coolant Loop 1 Hot Leg Sample Containment Penetration VT-1 Inspection of the Reactor Coolant Loop 1 Hot Leg Sample Containment Penetration (P093) was completed satisfactorily.

P099 - Containment Atmosphere Monitor Post Accident H2 Analyzer Supply Containment Penetration VT-1 Inspection of the Containment Atmosphere Monitor Post Accident H2 Analyzer Supply Containment Penetration (P099) was completed satisfactorily.

Enclosure Evaluation of the Proposed Change 78 35th Year IWL Activities Examination Category and Item Number:

L-B, L2.20 Item

Description:

Tendon Wire Evaluation

Description:

CR 202103224 - Containment Post Tension system tendon V39 wire did not meet the ultimate tensile strength acceptance criteria during the lab testing of the removed tendon wire. The lab test average was found to be 235.3 thousand pounds per square inch (ksi) whereas the ultimate tensile strength acceptance criteria is a minimum of 240 ksi.

Reviews of the original containment design parameters, the containment tendon system design basis, and the containment response to a postulated severe accident has demonstrated there is significant margin in the original design. The original specification for the tendon wire required a yield stress value of 192 ksi which results in a containment pressure capacity of 152 psig. The average yield for the wire samples that failed the ultimate strength requirements were 217.97 ksi. Inryco designed the tendon system using an average wire stress of 156.9 ksi. Based on the final number of tendons installed, Engineering calculated an actual wire minimum prestress of 127.16 ksi at design pressure. Engineering also calculated a maximum containment pressure capacity of 152 psig vs. a design pressure of 60 psig. The minimum tested yield for all 6 samples was 213.2 ksi. This is 11% greater than the design basis yield stress of 192 ksi. The ultimate strength of the wire is not used as a design basis for the containment building tendons.

Maximum design stress of 192 ksi is significantly less than the ultimate strength results from the sample wire, indicating that the tendon will not fail prior to the maximum design stress. The lift-off forces for all surveyed tendons were within or above the projected acceptance band at the 35th year surveillance point. This verifies that each of the tendons is operable by meeting its overall design function.

3.5.5 Containment Leakage Rate Testing Program - Type B and Type C Testing Program Types B and C testing at Callaway ensures that containment penetrations such as air locks, flanges, sealing mechanisms, electrical penetrations and containment isolation valves are essentially leak tight. There are no pressure retaining bellows used on containment penetrations at Callaway.

The initial test frequency for performing a leak test on Type B and Type C components is a base interval of 30 months. For Type B components, the interval may be extended to up to 120 months based on acceptable performance. Type B components whose test intervals are extended to greater than 60 months are tested on a staggered basis to allow for early detection of any common mode failure mechanism. For Type C components, the interval may be extended up to 60 months based upon acceptable performance. Acceptable performance for extending the 30-month interval is established by passing two as-found (AF) LLRTs with

Enclosure Evaluation of the Proposed Change 79 leakage less than or equal to the established administrative limits and that are at least 24 months apart (or separated by a normal refueling interval).

In accordance with TS 5.5.16, Types B and C acceptance criteria prior to unit startup must be less than 0.60 La. Table 3.5.6-1 provides a summary of the as-left (AL) maximum pathway (MXPLR) and as-left minimum pathway (MNPLR) running totals for Types B and C components from Refuel (RF) 20 (Fall 2014) to the most recent refueling outage, RF25 (Spring 2022), which included the performance of the last ILRT in 2014.

In accordance with TS 5.5.16, the allowable maximum pathway total Types B and C leakage is 0.6 La (252,028 standard cubic centimeters per minute (sccm)) where La equals 420,046 sccm.

As discussed in NUREG-1493 (Reference 6), Type B and Type C tests can identify the vast majority of all potential containment leakage paths. Type B and Type C testing will continue to provide a high degree of assurance that containment integrity is maintained.

A review of the As-Left (AL) test values for Callaway can be summarized as follows:

x As-Left MNPLR leak rate shows an average of 9.94% of 1.0 La with a high of 13.0% of 1.0 La following the completion of RF21 repair activities.

x As-Left MXPLR leak rate shows an average of 14.9% of 1.0 La with a high of 18.6% of 1.0 La following the completion of RF21 repair activities.

Table 3.5.6-1 Callaway Types B and C LLRT Combined As-Left Trend Summary Outage &

Year RF19 2013 RF20 2014 RF21 2016 RF22 2017 RF23 2019 RF24 2020 RF25 2022 As-Left MXPLR (sccm) 105,979.08 170,116.74 53,139.63 77,988.19 67,624.8 56,091.93 57,724.19

%1.0 La 25.2 40.51 12.7 18.6 16.1 13.4 13.7 As-Left MNPLR (sccm) 90,310.17 149,231.5 40,832.79 54,694.91 35,493.52 44,411.35 33,681.04

%1.0La 21.5 35.5 9.7 13.0 8.4 10.6 8.0

1. During RF20, the overall Types B and C containment penetration leakage rate summary was more than the Maintenance Rule (MR) limit of 0.4 La (168,018.4 sccm). Reference Sections 3.6.6, "Containment Isolation System (SM) Exceeded the Maintenance Rule

Enclosure Evaluation of the Proposed Change 80 Performance Criteria - RF20," and 3.6.7, "Local Leak Rate Testing Program Effectiveness," of this LAR.

The As-Left minimum pathway summations following the completion of required component repair activities during RF21 represent the high quality of maintenance of Type B and Type C tested components while the As-Left maximum pathway summations following RF21 represent the effective management of the Containment Leakage Rate Testing Program by the program owner.

3.5.6 Type B and Type C Local Leak Rate Testing Program Implementation Review Object:

KCHV0253 Containment Fire Protection Penetration P-67 outer containment isolation valve.

Deviation:

During RF24, could not attain LLRT pressure to evaluate seat leakage. With not being able to quantify the seat leakage rate, the overall type B and C Containment leakage rate is exceeded. This required a Mode 4 Statement hold to be placed on the surveillance and KCHV0253 to be placed on the Equipment Out of Service list.

The reason for the failure was debris build up between the seats and disc of KCHV0253 due to the replacement activities associated with KCV0478, inner CIV for containment penetration P-067. KCHV0253 is a parallel double disc gate valve. Debris between the seats and disc holds the disc away from the seats which results in seat leakage.

KCHV0253 seat leakage was corrected in RF24 via JOB 20003382. In addition, the fire protection line was flushed. The cause of the failure, debris build up on the seats and disc of KCHV0253, has been identified and the internals of the valve have been cleaned. A post maintenance leak rate test has been performed and the leakage rate was acceptable at 28.5 sccm.

The reason for the failure was debris build up between the seats and disc of KCHV0253 due to the replacement activities associated with KCV0478. The debris was cleared and the LLRT of KCHVO253 was reran with acceptable results. Closure: No additional actions are required.

The results of testing during RF25 is as follows:

KCHV0253 - 1148.1sccm Object:

LFFV0096, CTMT NORM SMP PMPS DISCH HDR AUX BLD FCV Deviation:

LFFV0096, Containment Norm Sump Pumps Discharge Header Auxiliary Building Flow Control Valve, could not be pressed up during performance of

Enclosure Evaluation of the Proposed Change 81 LLRT testing, JOB 19505277.500. Not being able to attain test pressure resulted in LFFV0096 not being able to perform its design function as a Containment Pressure boundary and the leakage rate of containment exceeding the Code allowable leakage rate.

Task number 525 was added to JOB 19505277 to investigate and repair the reason LFFV0096 could not isolate correctly. The findings of the JOB identified the reason LFFV0096 could not isolate correctly, and support LLRT was a portion of a tie wrap was trapped in the valve body. The location of the tie wrap was between the seat and plug resulted in the plug being held off the seat. This is the reason a testing pressure could not be attained to obtain a seat leakage rate. After the tie wrap was removed, an internal inspection was performed with no other issues identified. LFFV0096 was later leak rate tested with satisfactory results. No additional actions were required to correct this condition as this failure is being classified as Foreign Material Exclusion Event, which occurred during RF24 prior to testing LFFV0096.

The results of testing during RF25 is as follows:

LFFV0096 - 111.17 sccm Object:

KCV0478 (inner containment isolation valve for containment penetration P-067)

Deviation:

KCV0478 is not seating properly and failing LLRT after fire water has been flowed through it during RF24. Mechanical agitation is required to attain seating and acceptable LLRT results.

With KCV0478 not seating correctly after flow, the potential exists for KCV0478 to no longer support isolation of Containment piping penetration P-067.

KCV0478, prior to RF24, was a 4 swing check valve. In RF22, the disc was found in the full open position. The reason the disc stuck open was body corrosion and an improperly configured disc. The disc was replaced in RF23, and the complete valve was replaced in RF24. The replacement valve is a differently configured valve. The replacement valve is an inline type check valve.

This design was selected because the disc seating surface and disc are centered in the valve body, which prevents corrosion products that precipitate after flow from being trapped between the disc and seat. This change was implemented via the corrective actions of CR201706233.

The issues with KCV0478, regardless of the design of the valve, has been the quality of the water and the amount of debris in the line. The debris is picked up by the water flow and transported to and in the CIVs for piping penetration P-067.

The valves are KCHV0253 (outer isolation valve, gate design motor operated) and KCV0478 (check design, spring to close). This CR questions if the new

Enclosure Evaluation of the Proposed Change 82 valve installed KCV0478 will function if it is exposed to fire water? The RF24 fire water flow tests have resulted in seating issues with KCV0478. The post flow LLRTs will not pass until the valve is agitated. Agitation more than likely centers the disc in the seat and allows the installed spring to press the disc/plunger into the seat, which then produces an acceptable LLRT.

Options do exist for this valve to improve seating. The options are an elastomer seat and a stiffer return spring. These options could decrease seating issues.

Remedial Action:

After KCV0478 was flowed with fire water and failed the LLRT, it was flushed with clean reactor makeup water (JOB 20003887.200). After flushing, the LLRT was repeated and the test failed again. The valve was mechanically agitated, which resulted in proper seating and returned a very low seat leakage rate.

Action:

Until such time as KCV0478 produces repetitive acceptable LLRT leakage rates after it has been flowed, the seating capability post flow will need to be verified.

The suggested action for verification is to keep the night order (as generated based upon finding KCV0478 in the open position in RF22) in place and generate an at power LLRT procedure to perform an LLRT after water has been flowed through KCV0478. CR 202007021 has been written to generate the at power LLRT procedure. JOB 20004381 has been written to figure out the corrective actions for the replacement valve installed in KCV0478. The goal of the Job will be to have it seat on its own and have good LLRT results and to be able to repeat seating after being exposed to the water.

Corrective Action:

At this time, KCV0478 is seated and supporting containment isolation to support the Limiting Condition of Operation for TS 3.6.3. No corrective actions are identified at this time. If KCV0478 will not pass an LLRT after being flowed, the effectiveness action of CR 201706233 (which details the performance criteria for the corrective actions implemented by CR 201706233) will be used to initiate corrective actions. The Effectiveness review was failed due to seating issues during RF24. JOB 20004381 has been written to initiate and perform corrective actions. Enhancement: Action (003) from CR 202005650 provided revised instructions for the draining of the fire protection lines associated with Penetration 67 prior to the performance Slave Relay Test procedure. The enhanced instructions increase the amount of piping being drained before the slave relay test is performed. The enhanced instruction increases the piping to be drained by draining all the piping between KCHV0253 and KCV0478. Prior to RF24, draining this piping required water to flow past KCV0478 to a drain valve down stream of KCV0478.

Enclosure Evaluation of the Proposed Change 83 During RF24 as part of the modification to install the new valve in location KCV0478 a new drain valve was added to the piping. The new drain valve (KCV0865) was added just upstream of KCV0478. This enhancement should drain all the water from the lines upstream of KCV0478 prior to performing the slave relay test, which requires KCHV0253 to be stroked open and closed.

The results of testing during RF25 is as follows:

KCV0478 - 1290.5 sccm KCHV0253 - 1148.1 sccm Performance Summary Of 65 Type B penetrations, 7% of the Type B penetrations are on extended intervals.

Of 109 Type C tested components, 59% of the Type C components are on extended intervals.

3.6 Operating Experience (OE)

During the conduct of the various examinations and tests conducted in support of the containment related programs previously mentioned, issues that do not meet established criteria or that provide indication of degradation, are identified, placed into the site's corrective action program, and corrective actions are planned and performed.

For the Callaway Primary Containment, the following site specific and industry events have been evaluated for impact:

x Information Notice (IN) 1992-20, Inadequate Local Leak Rate Testing x

IN 2004-09, Corrosion of Steel Containment and Containment Liner x

IN 2010-12, Containment Liner Corrosion x

IN 2014-07, Degradation of Leak-Chase Channel Systems for Floor Welds of Metal Containment Shell and Concrete Containment Metallic Liner x

Regulatory Issue Summary (RIS) 2016-07, Containment Shell or Liner Moisture Barrier Inspection x

Local Leak Rate Testing Program Effectiveness

Enclosure Evaluation of the Proposed Change 84 Each of these areas is discussed in detail in Sections 3.6.1 through 3.6.3, respectively.

3.6.1 IN 1992-20, Inadequate Local Leak Rate Testing The NRC issued IN 92-20 to alert licensees of problems with local leak rate testing of two-ply stainless-steel bellows used on piping penetrations at four different plants: Quad Cities, Dresden Nuclear Station, Perry Nuclear Plant, and the Clinton Station. Specifically, LLRTs could not be relied upon to accurately measure the leakage rate that would occur under accident conditions, because, during testing, the two plies in the bellows were in contact with each other, restricting the flow of the test medium to the crack locations. Any two-ply bellows of similar construction may be susceptible to the problem. The common issue in the four events was the failure to adequately perform local leak rate testing on different penetration configurations leading to problems that were discovered during ILRT tests in the first three cases.

Discussion:

There are no pressure retaining bellows used on containment penetrations at Callaway.

3.6.2 IN 2004-09, Corrosion of Steel Containment and Containment Liner The NRC issued IN 2004-09 to alert addressees to recent occurrences of corrosion in freestanding metallic containments and in liner plates of reinforced and pre-stressed concrete containments. It was expected that recipients will review this information for applicability to their facilities and consider actions, as appropriate. However, the suggestions in this information notice were not NRC requirements; therefore, no specific action or written response was required.

As discussed in Information Notice 97-10, Liner Plate Corrosion in Concrete Containments, the containment liners have safety factors well above the theoretically calculated strains. Any corrosion (metal thinning) of the liner plate or freestanding metallic containment could change the failure threshold of the containment under a challenging environmental or accident condition. Thinning changes the geometry of the containment shell or liner plate, which may reduce the design margin of safety against postulated accident and environmental loads.

Recent experience has shown that the integrity of the moisture barrier seal at the floor-to-liner or floor-to-containment junction is important in avoiding conditions favorable to corrosion and thinning of the containment liner plate material.

Inspections of containment at the floor level, as well as at higher elevations, have identified various degrees of corrosion and containment plate thinning. IN 2004-09 provides a partial listing of such occurrences at various nuclear power plants.

Discussion:

The Callaway containment liner plate is inspected in accordance with the requirements set forth in the ASME BPV Code,Section XI, Subsection IWE. As discussed in this IN, the use of the

Enclosure Evaluation of the Proposed Change 85 ASME BPV Code for the inspections of containment liner plates became effective September 9, 1996.

The IWE inspection is a general visual inspection of 100% of all accessible areas of the pressure boundary liner plate. The inspection is performed every other refueling outage.

The initial IWE inspection at Callaway took place during Refuel 10 (October 1999).

The second IWE inspection took place during Refuel 12 (October 2002).

The next IWE inspection was scheduled for Refuel 14 (October 2005).

During the initial IWE inspection in Refuel 10, a number of areas on the liner plate were identified as requiring further evaluation.

1.

Corrosion was identified at the concrete floor to liner plate interface in the incore tunnel area. Corrective Action Report (CAR) 199902304 documents this condition. RFR 19682A was completed to evaluate the affected liner plate material condition. W646485 was initiated to clean and paint the affected sections of liner plate to inhibit further corrosion. The W646485 work was performed in Refuel 12. W223703 was subsequently initiated to perform additional painting of the liner plate in the incore tunnel.

The W223703 work was scheduled to be performed in Refuel 13 (April 2004).

2.

Corrosion was identified in both the 'A' and 'B' containment normal sumps. CAR 199902740 documents this condition. RFR 20321A was completed to evaluate the affected liner plate material condition. Modification package 00-1008A was generated to install stainless steel plate over the carbon steel plates in the containment normal sumps. The stainless-steel plate would then become the containment pressure boundary in this area. The stainless-steel plate installation was completed in the 'B' sump during Refuel 11 (C653403) and in the 'A' sump during Refuel 12 (C653401).

During the second IWE inspection, in Refuel 12, a number of areas on the liner plate were identified as requiring further evaluation.

1.

Light rust was observed on the liner plate in Zone 10. Upon closer examination, the dark discoloration was actually the result of oil and dust accumulating on a roughened section of the liner plate. No corrosion appeared to exist and no material loss of the liner plate was observed.

2.

Paint damage was observed at the concrete floor to liner plate interface on the 2000' elevation of the Reactor Building. The liner plate, located at the floor to wall interface, was painted during Refuel 9 (April 1998) under C596672, as a result of corrosion identified during an inspection performed in Refuel 8 (October 1996). An area one foot up from the 2000' elevation floor and reaching around the entire perimeter of the Reactor Building was painted under the C596672 work activity. There were no indications

Enclosure Evaluation of the Proposed Change 86 identified during the Refuel 12 inspection to suggest any corrosion or damage to the liner plate existed, but the observation was noted (CAR 200207431) and W223992 was initiated to correct the paint damage. The W223992 work was completed during Refuel 13 (April 2004).

The Callaway Plant has no outstanding items related to the condition of the containment pressure boundary liner plate. All areas of concern have been addressed and the appropriate corrective actions taken.

The Callaway Plant does not have a moisture barrier seal that adjoins the concrete floor to the liner plate. Rather, a one-foot-thick fill slab is installed over the liner plate at both the 2000' elevation floor of containment and the 1970' elevation floor of the incore tunnel area in containment. The fill slab is installed directly over the horizontal liner plate on the floor, and directly against the vertical liner plate on the wall. Areas of corrosion that have been identified at the floor to wall interfaces have been adequately addressed though the plant's corrective action program and work control process.

A walkdown of the fill slab to liner plate interface at the 2000' elevation was performed by engineering personnel on 05/06/2004, while the plant was in Refuel 13. The walkdown confirmed that the fill slab to liner plate interface appears to be in good condition. No actions were taken as a result of the walkdown.

Callaway will continue to perform containment pressure boundary inspections of the liner plate in accordance with the requirements set forth in the ASME BPV Code,Section XI, Subsection IWE.

3.6.3 IN 2010-12, Containment Liner Corrosion The NRC issued IN 2010-12 to alert plant operators to three events that occurred where the steel liner of the containment building was degraded and corroded. Concrete reactor containments are typically lined with a carbon steel liner to ensure a high degree of leak tightness during operating and accident conditions. The reactor containment is required to be operable as specified in plant technical specifications to limit the leakage of fission product radioactivity from the containment to the environment. The regulations at 10 CFR 50.55a, Codes and Standards, require the use of Subsection IWE of ASME Section XI to perform inservice inspections of containment components. The required inservice inspections include periodic visual examinations and limited volumetric examinations using ultrasonic thickness measurements. The containment components include the steel containment liner and integral attachments for the concrete containment, containment personnel airlock and equipment hatch, penetration sleeves, moisture barriers, and pressure-retaining bolting. The NRC also requires licensees to perform leak rate testing of the containment pressure-retaining components and isolation valves according to 10 CFR Part 50, Appendix J, Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors, as specified in plant technical specifications. This operating experience highlights the importance of good quality assurance, housekeeping and high-quality construction practices during construction operations in

Enclosure Evaluation of the Proposed Change 87 accordance with 10 CFR Part 50, Appendix B, Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants.

Operating experience shows that containment liner corrosion is often the result of liner plates being in contact with objects and materials that are lodged between or embedded in the containment concrete (FME). Liner locations that are in contact with objects made of an organic material are susceptible to accelerated corrosion because organic materials can trap water that combined with oxygen will promote carbon steel corrosion. Organic materials can also cause a localized low pH area when they decompose. Organic materials located inside containment can come in contact with the containment liner and cause accelerated corrosion. However, corrosion that originates between the liner plate and concrete is a greater concern because visual examinations typically identify the corrosion only after it has significantly degraded the liner. In some cases, licensees identified such corroded areas by performing ultrasonic examination of suspect areas (e.g., areas of obvious bulging, hollow sound).

The objects and materials that caused liner corrosion that licensees have found lodged between or embedded in the containment concrete include both foreign material (e.g., wooden pieces, workers gloves, wire brush handles) and material that was deliberately installed as part of the design such as the felt material described in the above example at Brunswick Steam Electric Plant, Unit 1. Although there is no regulatory requirement to do so, one or more licensees have chosen to review design documents to identify locations where organic material was intentionally installed between the liner or penetration sleeve and schedule additional examinations of these areas to monitor for liner material loss.

Discussion:

Callaway currently has processes in place to examine the liner plate for such corrosion. The ASME Section XI, Subsection IWE inspections are conducted every three years during refueling outages. The entirety of the containment pressure boundary including the liner plate and all containment penetrations and hatches are inspected during each three-year period. The Containment Pressure Boundary Inservice Inspection Program plan details Callaway's commitments and processes for meeting ASME Section XI, Subsection IWE and 10 CFR 50.55a limitations and modifications. Callaway Procedures outline the actual inspection procedures and visual examination acceptance criteria.

Additionally, Callaway engineering conducts a Containment coatings survey during every Refuel to inspect for loose and damaged coatings in containment. This inspection would also identify any areas of corrosion on the containment liner plate.

3.6.4 IN 2014-07, Degradation of Leak-Chase Channel Systems for Floor Welds of Metal Containment Shell and Concrete Containment Metallic Liner The NRC issued this IN to inform addressees of issues identified by the NRC staff concerning degradation of floor weld leak-chase channel systems of steel containment shell and concrete containment metallic liner that could affect leak-tightness and aging management of

Enclosure Evaluation of the Proposed Change 88 containment structures. The NRC expected that recipients will review the information for applicability to their facilities and consider actions, as appropriate, to avoid similar problems.

The containment floor weld leak-chase channel system forms a metal-to-metal interface with the containment shell or liner, the test connection end of which is at the containment floor level.

Therefore, the leak-chase system provides a pathway for potential intrusion of moisture that could cause corrosion degradation of inaccessible embedded areas of the pressure-retaining boundary of the basemat containment shell or liner within it. In addition to protecting the test connection, the cover plates and plugs and accessible components of the leak-chase system within the access box are also intended to prevent intrusion of moisture into the access box and into the inaccessible areas of the shell/liner within the leak-chase channels, thereby protecting the shell and liner from potential corrosion degradation that could affect leak-tightness.

The containment ISI program required by 10 CFR 50.55a to be implemented in accordance with Subsection IWE, of the ASME Code,Section XI, subject to regulatory conditions, requires special consideration of areas susceptible to accelerated corrosion degradation and aging, and barriers intended to prevent intrusion of moisture and water accumulation against inaccessible areas of the containment pressure-retaining metallic shell or liner. The containment floor weld leak-chase channel system is one such area subject to accelerated degradation and aging if moisture intrusion and water accumulation is allowed on the embedded shell and liner within it.

Therefore, the leak-chase channel system is subject to ISI requirements of 10 CFR 50.55a(g)(4) and aging management requirements of 10 CFR 54.29(a)(1).

This IN provided examples of OE at some plants of water accumulation and corrosion degradation in the leak-chase channel system that has the potential to affect the leak-tight integrity of the containment shell or liner plate. In each of the examples, the licensee had no provisions in its ISI plan to inspect any portion of the leak-chase channel system for evidence of moisture intrusion and degradation of the containment metallic shell or liner within it. Therefore, these cases involved the licensees failure to perform required visual examinations of the containment shell or liner plate leak-chase systems in accordance with the ASME Code Section XI, Subsection IWE, as required by 10 CFR 50.55a(g)(4). The moisture intrusion and associated degradation found within leak-chase channels, if left uncorrected, could have resulted in more significant corrosion degradation of the containment shell or liner and associated seam welds. These examples and other similar previous industry operating experiences highlight the importance of licensees recognizing the existence of leak-chase channel systems in their containment floor. These experiences also highlight the importance of understanding the system configuration and how the leak-chase system components interact with the containment pressure-retaining metallic shell or liner plate within it to ensure that these systems are appropriately included for required examinations in the containment ISI program and the Subsection IWE aging management program.

For containments in which basemat shell/liner leak-chase channel systems exist with accessible interface at the containment floor level, licensees are required to comply with the containment ISI requirements of 10 CFR 50.55a(g)(4).

Enclosure Evaluation of the Proposed Change 89 Discussion:

Information Notice 2014-07 addressed the potential for corrosion of containment liners (containment pressure boundary) due to moisture intrusion through leak-chase channel access points. The systems addressed in this IN include access boxes with cover plates that could provide a pathway for moisture to reach inaccessible portions of the liner. The leak-chase channels at Callaway are seal-welded to the liner and extend to a point above the concrete floor, providing direct access to the leak-chase system and eliminating the need for access boxes. IN 2014-07 addresses leak-chase systems of a different configuration than is present at Callaway. Callaway's leak-chase system does not employ the use of access points at the concrete floor level.

IN 2014-07 also noted that the access boxes and leak-chase systems were not included in the ISI programs. At Callaway, the accessible portions of the leak-chase system (the seal welded segments above the concrete) are inspected under the IWE program as integral attachments to the containment liner. If any conditions are identified that could indicate the presence of or result in degradation to inaccessible areas, an evaluation is performed to determine appropriate corrective actions. No such conditions have been identified.

It should be noted that the leak chase system is not a system for conveying moisture as the name may imply, but rather a system that was installed during initial construction to facilitate pressure testing of the liner plate seam welds under the 2000' elevation concrete floor.

Also, any moisture introduced to the containment concrete floor to steel liner interface at the 2000' elevation, a possible location for water intrusion, would be conveyed away from the liner as the floor is sloped at all points around containment away from the steel liner. Also, no cracks or openings between the concrete floor and steel liner have been identified.

The steel liner to concrete floor interface is also inspected every 18 months (every refuel) under the Coatings Aging Management Program and every 4.5 years (every 3 refuels) under the Structural Monitoring Aging Management Program. The inspection under the Structural Monitoring Aging Management Program was completed in RF20 with no identified issues related to IN 2014-07.

The leak chase system at Callaway is comprised of 32 separate zones with each zone having its own access point above the floor level. There are 14 access points located along the circumference of the Reactor Building cylindrical steel liner along with an additional 18 at interior walls, all near and accessed from the 2000' elevation.

3.6.5 RIS 2016-07 Containment Shell or Liner Moisture Barrier Inspection The NRC issued this RIS to reiterate the NRC staffs position in regard to ISI requirements for moisture barrier materials, as discussed in the ASME Boiler and Pressure Vessel Code (hereinafter the ASME Code),Section XI, Rules for Inservice Inspection of Nuclear Power Plant Components, Subsection IWE. No specific action or written response is required.

Enclosure Evaluation of the Proposed Change 90 Background Information Requirements in 10 CFR 50.55a, Codes and standards, paragraph (g), Inservice inspection requirements, require, in part, that licensees implement an inservice inspection program for metal containments and metallic liners of concrete containments. The program shall be in accordance with the latest edition and addenda of Subsection IWE of Section XI of the ASME Code that has been incorporated by reference in 10 CFR 50.55a(b) 12 months before the start of the 120-month inspection interval and is subject to the applicable conditions in 10 CFR 50.55a(b)(2)(ix).

Section XI of ASME Code, Item E1.11, in Table IWE-2500-1 (E-A), requires general visual examination of 100 percent of accessible surface areas during each inspection period, while Item E1.30 in the same table requires general visual examination of 100 percent of accessible moisture barriers during each inspection period. Note 4 (Note 3 in editions before 2013) for Item E1.30 under the Parts Examined column states, Examination shall include moisture barrier materials intended to prevent intrusion of moisture against inaccessible areas of the pressure retaining metal containment shell or liner at concrete-to-metal interfaces and at metal-to-metal interfaces which are not seal-welded. Containment moisture barrier materials include caulking, flashing, and other sealants used for this application.

Note 4 also defines the moisture barriers that are required to be examined in terms of intended function. Table IWE-2500-1, third column, Examination Requirements/Fig. No., for Item E1.30, references Figure IWE-2500-1, Examination Areas for Moisture Barriers (shown below). Figure IWE-2500-1 depicts typical moisture barrier examination areas for concrete-to-metal interface moisture barriers. Figure IWE-2500-1 does not include an illustration of metal-to-metal interface moisture barriers, which are included as part of the definition of moisture barriers and are areas to be examined in accordance with Note 4. Thus, the NRC understands the figure to be an example, not a definitive depiction, of the parts required to be examined under the scope of Item E1.30. There may be other configurations in which a material has been applied to prevent moisture from making contact with inaccessible areas of the metal containment shell or liner. These materials should be inspected as a moisture barrier under item E1.30.

Discussion:

The Callaway containment liner plate is inspected in accordance with the requirements set forth in the ASME Code,Section XI, Subsection IWE. The use of the ASME Code for the inspections of containment liner plates became effective September 9, 1996. The IWE inspection is a general visual inspection of 100% of all accessible areas of the pressure boundary liner plate and this inspection is performed every other refueling outage. The initial IWE inspection at Callaway took place during Refuel 10 (October 1999).

Enclosure Evaluation of the Proposed Change 91 Callaway does not have a moisture barrier seal that adjoins the concrete floor to the liner plate.

Rather, a one-foot-thick fill slab is installed over the liner plate at both the 2000' elevation floor of containment and the 1970' elevation floor of the incore tunnel area in containment. The fill slab is installed directly over the horizontal liner plate on the floor, and directly against the vertical liner plate on the wall. Areas of corrosion that have been identified at the floor to wall interfaces have been adequately addressed through the plant's corrective action program and work control process. A walkdown of the fill slab to liner plate interface at the 2000' elevation was performed by engineering personnel on May 6, 2004, while the plant was in Refuel 13. The pressure boundary inspections are performed every other refuel in accordance with the requirements set forth in ASME Section XI, Subsection IWE.

Callaway has no outstanding items related to the condition of the containment pressure boundary liner plate. All areas of concern have been addressed and the appropriate corrective actions taken.

3.6.6 Containment Isolation System (SM) Exceeded the Maintenance Rule (MR)

Performance Criteria - RF20 (2014):

Performance Criteria for the SM System are as follows:

x No more than ten valves or other local leak rate tested penetrations (e.g., electrical penetrations, hatches, or flanged penetrations) that have leakage greater than their administrative limit at any given time.

x No simultaneous failure to activate on a valid Containment Isolation Signal of both the inside and outside valves in a single penetration.

x No more than two MPFFs per 18 months for failure to activate on a valid Containment Isolation Signal.

x Containment Leakage for Types B and C tests shall be less than or equal to 0.4 La which equates to 168,018.4 SCCM.

x Zero MPFFs for the Equipment Hatch Containment closure function when the function is required.

All of these performance requirements are currently satisfied with one exception: the Containment Leakage for Types B and C tests shall be less than or equal to 0.4 La which equates to 168,018.4 SCCM.

Enclosure Evaluation of the Proposed Change 92 3.6.7 Local Leak Rate Testing Program Effectiveness

==

Description:==

Multiple problems are impacting the effectiveness of the Local Leak Rate Testing (LLRT) program, consequently contributing to a lack of teamwork to drive resolution of long-standing test and valve performance issues. Test and valve issues have contributed to the containment isolation system exceeding MR program performance criteria, have resulted in outage scheduling delays, and can, if left uncorrected, further challenge meeting regulatory testing criteria as more discovery items are identified.

Below are examples impacting effectiveness of the LLRT program:

1. In RF20, CAR 201407586 documents that EMHV8964 exceeded the administrative leak rate. In RF21, CAR 201603231 documents the same issue.

Insight: The job to correct EMHV8964 is not scheduled until RF30. A number of LLRT valves were scheduled to be worked in RF21; however, the remaining valves are not scheduled yet or not scheduled until RF30 (CAR 201408400 action #1). Specifically, valves in penetrations 29, 71, and 73 had the most significant leaks and were therefore scheduled to be fixed in RF21 to reduce the overall Types B and C containment leakage to below MR (a)(1) limits. While this is a good short-term action, there is no long-term program plan to continue to reduce and remain below administrative leakage limits for other containment penetrations. Failure to properly schedule jobs to clean and fix degraded CIVs will challenge programmatic requirements as well as the reliability of these valves. In addition, when new discovery items are identified - see CAR 201602912 and 201603549 from RF21 - leakage rate margins will again be challenged.

2. In RF20, CAR 201407303 documents leak by past boundary valve EGV0125. This delayed testing of penetration 75 and caused additional hours on the job. In RF21, CARs 201603015 and 201602983 again document leak by past this boundary valve affecting outage scheduling and resources. Leak by past this valve also contributed to exceeding the planned duration of a yellow safety risk condition (CAR 201603409).

Insights: Based on discussions with the LLRT program owner, the RF22 schedule date to fix this valve was a surprise. Also, scheduling notes in Job 14005015 state that EGV0125 needs to be fixed in RF21 and was recoded back to RF21. Because it was assumed this valve would be fixed in RF21, no plans were made to implement test procedure changes, resulting in unnecessary scheduling and resource issues.

3. In RF20, CAR 201407622 documents Essential Service Water (ESW) boundary valves (EFV0343, 344, 345, and 346) would not isolate to support LLRT testing of penetrations 28, 29, 71, and 73. This CAR also documents this as a repeat issue, noting condition reports since RF17. In RF21, CARs 201603116 and 201603117 document the same issue.

Enclosure Evaluation of the Proposed Change 93 Insights: Health Issue (HI) 2014019 was generated in December 2014 to resolve this testing issue. Health issue documentation states that the Unit Reliability Team (URT) did not approve a solution until September 2015. Based on discussions with the LLRT program owner, an informal and undocumented URT met earlier in 2015 to discuss this HI so that valve procurement could begin. However, delays occurred in the procurement process, preventing procurement for RF21. It is documented in the HI that by September 2015, the URT knew that procurement of these valves would not occur in time to support RF21. Procedurally this requires implementing bridging actions.

However, the URT-approved bridging actions to use blind flanges in place of the boundary valves were not implemented for RF21. Based on discussion with the LLRT program owner, Outages did not accept adding the approved URT bridging actions. In addition, no action was taken to revise the bridging actions. Therefore, test jobs /

procedures were not changed before RF21, resulting in unnecessary scheduling and resource issues. At the quarterly URT meeting in March 2016, attendees were still unsure if these boundary valves were on site or had been ordered.

Recommended Actions to Resolve this Issue:

It is recommended that a multi-discipline team approach be utilized to help drive resolution of these long-standing test and valve performance issues. This approach can also help improve station alignment and understanding of program health and reliability margin for the containment isolation system. In addition, additional oversight and attention is warranted to help improve the effectiveness of the program.

The subject of the CAR is administrative or programmatic in nature and does not affect the ability or qualification of any structure, system, or component to perform its specified safety function. Thus, no Immediate Operability, Immediate Functionality, or Reportability Determination is required. The subject of the CAR is administrative or programmatic in nature and does not affect the ability or qualification of any structure, system, or component to perform its specified safety function.

Deviation.

The lack of a long-term plan to improve the health of the Appendix J program and a lack of teamwork.

Impact.

Leakage rates for Type C penetrations above administrative rates, operations work around to perform testing, inability to obtain individual valve leakage rates for the valves in penetrations 28, 29, 71 and 73, and in RF20 the overall type B and C summary leakage rate in excess of the MR performance criteria.

Enclosure Evaluation of the Proposed Change 94 Remedial Actions Taken:

Based on risk and consequence, document the Remedial Actions to address deficient conditions temporarily until permanent Corrective Actions can be implemented.

There are no remedial actions, which could be performed that would address the current conditions.

The leakage rate of EMHV8964 can only be improved by repair / replacement.

The inability to obtain independent leakage rates for the valves associated with penetrations 28, 29, 71 and 73 can only be corrected by replacing EFV0343, EFV0344, EFV0345 and EFV0346.

The testing alignment for penetration 75 is currently a work around for operations which works, but the preference is to repair / replace EGV0125.

Though these conditions are adverse, the lack of remedial actions does not increase the risk or consequences of these conditions.

Extent of Condition:

Accurately identify the Extent of the Condition as it relates to impacts on other plant systems, components, structures, programs, procedures, processes, or organizations. If Extent of Condition goes beyond departmental boundaries, contact your department head for organizational support.

Does the potential exist for this problem to cause further impact to this SSC or Process?

Yes. If the Appendix J program is not correctly monitored and if timely corrective actions are not implemented, there is the risk that it may not only stay MR (a)(1) the overall leakage could exceed the limits established in TS 3.6.3.

Does the potential exist for this problem to impact other SSCs or Processes?

No. The Extent of Condition is limited to the Appendix J program. Other SSCs, programs, procedures, processes, and organizations are not impacted or affected by the condition of the Appendix J program.

Component conditions which impact the Appendix J program:

Replacement of EGV0125:

Currently, the leakage of this valve is impacting the procedure associated with testing

Enclosure Evaluation of the Proposed Change 95 penetration P-075. The procedure (OSP-EG-LL075) contains line up instructions for the performance of the LLRT for valves EGHV0059, 60, 130, and 131, which requires EGV0125 to close. With EGV0125 leaking by the leakage rate for the EGHV valves results in a false high rate. Establishing an alternate boundary requires a separate JOB to provide the instructions to develop the boundary. Without this JOB, operations would be testing outside of procedure instructions.

Repair of BGHV8149A:

This valve is one of three testing boundary valves used to perform the LLRTs on valves BGHV8149 and BGHV8160. The seat leakage of BGHV8149A is resulting in a false high (approximately 13,000 SCCM) LLRT for BGHV8149 and BGHV 8160.

These actions are the remaining currently documented actions to further improve the performance of the Appendix J program. Additional actions will be identified based upon testing in future refuels.

Discussion or Explanation of Event:

Since RF19 there have been LLRT testing issues. The most significant testing issues occurred in RF20 where the overall Types B and C component leakage rate summary exceeded the MR criteria of 0.4La. To address this issue, CAR 201408400 was generated. This CAR evaluated the MR status change from (a)(2), normal monitoring, to (a)(1) increased monitoring and developed the corrective JOBs to lower the overall summary leakage rates.

During RF21, a portion of the corrective JOBs listed in 201408400 were worked to decrease the overall summary leakage rate. The reason the entire corrective action JOB list was not worked in RF21 is the scope was so extensive that station resources could not support the full scope.

During Cycle 21, the decision was made to support those JOBs which attained the highest gains and work the other corrective actions in future refueling outages. The list of RF22, RF23 and RF24 corrective action JOBs needs to be developed and prioritized.

Prior to RF21, the overall summary leakage rate was at 0.407 La. After RF21, the overall summary leakage rate is at 0.127 La. This reduction supports the MR criteria of no greater than 0.4La but it does not restore the Appendix J MR status to (a)(2). Restoration of the status comes from a combination of corrective action and a positive trend (in this case further leakage rate decrease) and a stable trend. To accomplish this goal, the remaining corrective actions of 201408400 and any other identified corrective actions during the evaluation period should be completed.

The issue with EFV0343, 344, 345 and 346 started in RF17. During RF17, it was identified that the valves seats in these valves were so worn that isolation was not occurring to support an LLRT of the valves in penetrations 28, 29, 71, and 73. At this time, a CAR was written to obtain replacement valves and testing of penetrations 28, 29, 71 and 73 changed from an independent valve test to a global test of the valves. The CAR requested an equivalent replacement valve

Enclosure Evaluation of the Proposed Change 96 via a Replacement Item Equivalence (RIE). This request was changed to a configuration change package modification in 2007. Later, the modification request was cancelled and changed back to an 74in 2009. This RIE request was not assigned resources based on the low priority.

During RF18, the leakage rates of penetrations 29, 71 and 73 started to increase with the leakage rates of penetration 71 and 73 over their administrative rates. Request were made to work the 2009 RIE but due to station resources and low priority no resources were assigned. In RF19 and RF20, the leakage rates continued to increase and eventually reached the point of being the major contributors to exceeding the MR criteria. With testing converted to global testing and lack of being able to separate out the valves to perform an independent seat leakage test specific, corrective action JOBs could not be generated.

After RF20, HI 2014019 was written to attain replacement valves. This combined with the Appendix J program being MR (a)(1) produced the necessary priority to attain replacement valves. Currently, BOM 403172 is in process to obtain vendor quotes to attain replacement valves. The work of replacement is floating between RF22 and RF23.

EGV0125 was written up in CAR 201407303. This SIG 5 CAR was processed per procedure and JOB 14005015 was written to repair/replace EGV0125 during RF21. Due to the LLRT could still be performed (albeit outside of procedure instructions), the priority assigned was low, as such resources were not assigned.

Results of Lower Tier Cause Evaluation (LTCE):

The result of the Binning:

Managing the Overall Types B and C summary based upon Margin. Lack of resources assigned to corrective maintenance. Lack of resources assigned to obtaining replacement valves. Corrective actions prioritized low based upon overall summary Margin.

Seat leakage rates do not increase in a linear manner once an adverse trend starts.

Typical increases follow an exponential rate once the seat leakage starts to increase.

This is why managing the Appendix J program with margin is not a best practice.

Describe the results of the evaluation in the form of a short executive summary. Include error precursors and flawed defenses considered.

The primary Common Cause is incorrect prioritization of corrective actions initiated by and related to the Appendix J program. Not until the program exceeded the MR performance criteria were corrective actions supported by station management.

Corrective actions and support were based upon Overall Types B and C summary margin instead of keeping the margin as high as possible. This is the secondary Common Cause.

Enclosure Evaluation of the Proposed Change 97 Corrective Actions:

Align station management to properly support and prioritize Corrective Actions.

Engineering to meet with outage manager, outage scheduling and OLT to understand the requirement to support Appendix J generated Corrective Actions and how managing the program via Margin is not the correct practice.

Scope remaining JOBs of the MR recovery plan and other JOBs to decrease leakage and support testing and have them assigned to the appropriate Refueling outages.

Assigned to Engineering and outage scheduling.

==

Conclusion:==

Table 3.5.6-1, Callaway Types B and C LLRT Combined As-Left Trend Summary, shows the effectiveness of the corrective actions taken to address LLRT Program Effectiveness.

3.7 License Renewal Aging Management In compliance with Callaway license condition 2.C.(17), of the Callaway renewed operating license, Chapter 19 of the FSAR contains the information required by 10 CFR 54.21(d) that was contained in the Callaway license renewal application, Appendix A, Final Safety Analysis Report Supplement. By letter dated December 15, 2011, Union Electric Company, doing business as Ameren Missouri (Ameren Missouri or the applicant), submitted the license renewal application (LRA) in accordance with 10 CFR Part 54, Requirements for Renewal of Operating Licenses for Nuclear Power Plants. Ameren Missouri requested renewal of the Callaway operating license (Operating License No. NPF-30) for a period of 20 years beyond the current expiration at midnight October 18, 2024.

The following Callaway FSAR programs/activities are credited with the aging management of the Primary Containment:

x 19.1.26 ASME Section XI, Subsection IWE The ASME Section XI, Subsection IWE program manages cracking, loss of material, loss of sealing, loss of preload, and loss of leak tightness by providing aging management of the steel liner of the concrete containment building, including the containment liner plate and its integral attachments, containment hatches and airlocks, and pressure-retaining bolting. IWE inspections are performed in order to identify and manage any containment liner aging effects that could result in loss of intended function.

Acceptance criteria for components subject to Subsection IWE examination requirements are specified in Article IWE 3000. The Callaway containment ISI program is consistent with the requirements of 2007 Edition of ASME Section XI, Subsection IWE (through the 2008 addenda), supplemented with the applicable requirements of 10 CFR

Enclosure Evaluation of the Proposed Change 98 50.55a(b)(2)(ix). In conformance with 10 CFR 50.55a(g)(4)(ii), the Callaway containment inservice inspections program will be updated during each successive 120-month inspection interval to comply with the requirements of the latest edition and addenda of the Code specified 12 months before the start of the inspection interval.

x 19.1.27 ASME Section XI, Subsection IWL The ASME Section XI, Subsection IWL program manages the following aging effects of the concrete containment building and post tensioned system:

x Cracking x

Cracking, loss of bond, and loss of material (spalling, scaling) x Increase in porosity and permeability, cracking, loss of material (spalling, scaling) x Increase in porosity and permeability, loss of strength x

Loss of material x

Loss of material (spalling, scaling) and cracking Inspections will be performed to identify and manage any aging effects of the containment concrete, post-tensioning tendons, tendon anchorages, and concrete surface around the anchorage that could result in loss of intended function. In conformance with 10 CFR 50.55a(g)(4)(ii), the ASME Section XI, Subsection IWL program will be updated during each successive 120-month inspection interval to comply with the requirements of the latest edition and addenda of the Code specified 12 months before the start of the inspection interval.

x 19.1.29 10 CFR Part 50, Appendix J The 10 CFR Part 50, Appendix J program manages cracking, loss of material, loss of leak tightness, loss of sealing, and loss of preload. The program monitors leakage rates through the containment pressure boundary, including the penetrations and access openings, in order to detect degradation of containment pressure boundary.

Containment leak rate tests are performed in accordance with 10 CFR 50 Appendix J, Primary Reactor Containment Leakage Testing for Water Cooled Power Reactors (Option B); NRC RG 1.163, Performance-Based Containment Leak-Test Program; NEI 94-01, Industry Guideline for Implementing Performance Based Option of 10 CFR Part 50 Appendix J; and ANSI/ANS 56.8, Containment System Leakage Testing Requirements.

Enclosure Evaluation of the Proposed Change 99 Containment leak rate tests are performed to assure that leakage through the primary containment and systems and components penetrating primary containment does not exceed allowable leakage limits specified in the TS. Corrective actions are taken if leakage rates exceed established administrative limits for individual penetrations or the overall containment pressure boundary.

x 19.1.33 Protective Coating and Monitoring and Maintenance Program The Protective Coating Monitoring and Maintenance Program manages loss of coating integrity for Service Level 1 coatings inside containment so that the intended functions of post-accident safety systems that rely on water recycled through the containment sump/drain system are maintained consistent with the current licensing basis. The program includes a visual examination of all accessible Service Level 1 coatings inside containment, including those applied to the steel containment liner, structural steel, supports, penetrations, and concrete walls and floors. The program is consistent with the ASTM requirements, but Callaway is not committing to all the requirements noted in RG 1.54, Service Level I, II, and III Protective Coatings Applied to Nuclear Power Plants, Revision 2 (Reference 72).

3.8 NRC SER Limitations and Conditions 3.8.1 Limitations and Conditions Applicable to NEI 94-01, Revision 2-A The NRC staff found that the use of NEI TR 94-01, Revision 2, was acceptable for referencing by licensees proposing to amend their TS to permanently extend the ILRT surveillance interval to 15 years, provided the following conditions as listed in Table 3.8.1-1 were satisfied:

Table 3.8.1-1 NEI 94-01 Revision 2-A Limitations and Conditions Limitation/Condition (From Section 4.0 of SE)

Callaway Response For calculating the Type A leakage rate, the licensee should use the definition in the NEI TR 94-01, Revision 2, in lieu of that in ANSI/ANS-56.8-2002. (Refer to SE Section 3.1.1.1.)

Callaway will utilize the definition in NEI 94-01 Revision 3-A, Section 5.0. This definition has remained unchanged from Revision 2-A to Revision 3-A of NEI 94-01.

The licensee submits a schedule of containment inspections to be performed prior to and between Type A tests. (Refer to SE Section 3.1.1.3.)

Reference Section 3.5.2, Table 3.5.2-1 and Section 3.5.3, Tables 3.5.3-1, 3.5.3-2, and 3.5.3-3, of this submittal.

Enclosure Evaluation of the Proposed Change 100 Table 3.8.1-1 NEI 94-01 Revision 2-A Limitations and Conditions Limitation/Condition (From Section 4.0 of SE)

Callaway Response The licensee addresses the areas of the containment structure potentially subjected to degradation. (Refer to SE Section 3.1.3.)

Reference Sections 3.5.2 and 3.5.3 of this submittal.

The licensee addresses any tests and inspections performed following major modifications to the containment structure, as applicable. (Refer to SE Section 3.1.4.)

Steam Generator replacements were performed using the installed equipment hatch.

The normal Type A test interval should be less than 15 years. If a licensee has to utilize the provision of Section 9.1 of NEI TR 94-01, Revision 2, related to extending the ILRT interval beyond 15 years, the licensee must demonstrate to the NRC staff that it is an unforeseen emergent condition. (Refer to SE Section 3.1.1.2.)

Callaway will follow the requirements of NEI 94-01 Revision 3-A, Section 9.1. This requirement has remained unchanged from Revision 2-A to Revision 3-A of NEI 94-01.

In accordance with the requirements of NEI 94-01, Revision 2-A, SER Section 3.1.1.2, Callaway will also demonstrate to the NRC staff that an unforeseen emergent condition exists in the event an extension beyond the 15-year interval is required.

For plants licensed under 10 CFR Part 52, applications requesting a permanent extension of the ILRT surveillance interval to 15 years should be deferred until after the construction and testing of containments for that design have been completed and applicants have confirmed the applicability of NEI 94-01, Revision 2, and EPRI Report No. 1009325, Revision 2, including the use of past containment ILRT data.

Not applicable. Callaway was not licensed under 10 CFR Part 52.

3.8.2 Limitations and Conditions Applicable to NEI 94-01, Revision 3-A The NRC staff found that the guidance in NEI TR 94-01, Revision 3, was acceptable for referencing by licensees in the implementation of the optional performance-based requirements of Option B to 10 CFR 50, Appendix J. However, the NRC staff identified two conditions on the use of NEI TR 94-01, Revision 3 (Reference NEI 94-01, Revision 3-A, NRC SER 4.0, Limitations and Conditions):

Enclosure Evaluation of the Proposed Change 101 Topical Report Condition 1 NEI TR 94-01, Revision 3, is requesting that the allowable extended interval for Type C LLRTs be increased to 75 months, with a permissible extension (for non-routine emergent conditions) of nine months (84 months total). The staff is allowing the extended interval for Type C LLRTs be increased to 75 months with the requirement that a licensee's post-outage report include the margin between the Type B and Type C leakage rate summation and its regulatory limit. In addition, a corrective action plan shall be developed to restore the margin to an acceptable level. The staff is also allowing the non-routine emergent extension out to 84-months as applied to Type C valves at a site, with some exceptions that must be detailed in NEI TR 94-01, Revision 3. At no time shall an extension be allowed for Type C valves that are restricted categorically (e.g., BWR MSIVs), and those valves with a history of leakage, or any valves held to either a less than maximum interval or to the base refueling cycle interval. Only non-routine emergent conditions allow an extension to 84 months.

Response to Condition 1 Condition 1 presents the following three (3) separate issues that are required to be addressed:

x ISSUE 1 - The allowance of an extended interval for Type C LLRTs of 75 months carries the requirement that a licensee's post-outage report include the margin between the Type B and Type C leakage rate summation and its regulatory limit.

x ISSUE 2 - In addition, a corrective action plan shall be developed to restore the margin to an acceptable level.

x ISSUE 3 - Use of the allowed 9-month extension for eligible Type C valves is only authorized for non-routine emergent conditions with exceptions as detailed in NEI 94-01, Revision 3-A, Section 10.1.

Response to Condition 1, ISSUE 1 The post-outage report shall include the margin between the Type B and Type C MNPLR summation value, as adjusted to include the estimate of applicable Type C leakage understatement, and its regulatory limit of 0.6 La.

Response to Condition 1, ISSUE 2 When the potential leakage understatement adjusted Types B and C MNPLR total is greater than the Callaway administrative leakage summation limit of 0.5 La, but less than the regulatory limit of 0.6 La, then an analysis and determination of a corrective action plan shall be prepared to restore the leakage summation margin to less than the Callaway leakage limit. The corrective action plan shall focus on those components which have contributed the most to the increase in the leakage summation value and what manner of timely corrective action, as

Enclosure Evaluation of the Proposed Change 102 deemed appropriate, best focuses on the prevention of future component leakage performance issues so as to maintain an acceptable level of margin.

Response to Condition 1, ISSUE 3 Callaway will apply the 9-month allowable interval extension period only to eligible Type C components and only for non-routine emergent conditions. Such occurrences will be documented in the record of tests.

Topical Report Condition 2 The basis for acceptability of extending the ILRT interval out to once per 15 years was the enhanced and robust primary containment inspection program and the local leakage rate testing of penetrations. Most of the primary containment leakage experienced has been attributed to penetration leakage and penetrations are thought to be the most likely location of most containment leakage at any time. The containment leakage condition monitoring regime involves a portion of the penetrations being tested each refueling outage, nearly all LLRTs being performed during plant outages. For the purposes of assessing and monitoring or trending overall containment leakage potential, the as-found minimum pathway leakage rates for the just tested penetrations are summed with the as-left minimum pathway leakage rates for penetrations tested during the previous 1 or 2 or even 3 refueling outages. Type C tests involve valves, which in the aggregate, will show increasing leakage potential due to normal wear and tear, some predictable and some not so predictable. Routine and appropriate maintenance may extend this increasing leakage potential. Allowing for longer intervals between LLRTs means that more leakage rate test results from farther back in time are summed with fewer just tested penetrations and that total is used to assess the current containment leakage potential. This leads to the possibility that the LLRT totals calculated understate the actual leakage potential of the penetrations. Given the required margin included with the performance criterion and the considerable extra margin most plants consistently show with their testing, any understatement of the LLRT total using a 5-year test frequency is thought to be conservatively accounted for.

Extending the LLRT intervals beyond 5 years to a 75-month interval should be similarly conservative provided an estimate is made of the potential understatement and its acceptability determined as part of the trending specified in NEI TR 94-01, Revision 3, Section 12.1.

When routinely scheduling any LLRT valve interval beyond 60-months and up to 75-months, the primary containment leakage rate testing program trending or monitoring must include an estimate of the amount of understatement in the Types B and C total leakage and must be included in a licensee's post-outage report. The report must include the reasoning and determination of the acceptability of the extension, demonstrating that the LLRT totals calculated represent the actual leakage potential of the penetrations.

Enclosure Evaluation of the Proposed Change 103 Response to Condition 2 Condition 2 presents the following two (2) separate issues that are required to be addressed:

x ISSUE 1 - Extending the LLRT intervals beyond 5 years to a 75-month interval should be similarly conservative provided an estimate is made of the potential understatement and its acceptability determined as part of the trending specified in NEI TR 94-01, Revision 3, Section 12.1.

x ISSUE 2 - When routinely scheduling any LLRT valve interval beyond 60 months and up to 75 months, the primary containment leakage rate testing program trending or monitoring must include an estimate of the amount of understatement in the Types B and C total and must be included in a licensee's post-outage report. The report must include the reasoning and determination of the acceptability of the extension, demonstrating that the LLRT totals calculated represent the actual leakage potential of the penetrations.

Response to Condition 2, ISSUE 1 The change in going from a 60-month extended test interval for Type C tested components to a 75-month interval, as authorized under NEI 94-01, Revision 3-A, represents an increase of 25%

in the LLRT periodicity. As such, Callaway will conservatively apply a potential leakage understatement adjustment factor of 1.25 to the actual As-Left leak rate, which will increase the As-Left leakage total for each Type C component currently on greater than a 60-month test interval up to the 75-month extended test interval. This will result in a combined conservative Type C total for all 75-month LLRTs being "carried forward" and will be included whenever the total leakage summation is required to be updated (either while on-line or following an outage).

When the potential leakage understatement adjusted leak rate total for those Type C components being tested on greater than a 60-month test interval up to the 75-month extended test interval is summed with the non-adjusted total of those Type C components being tested at less than or equal to a 60-month test interval, and the total of the Type B tested components, results in the MNPLR being greater than the Callaway administrative leakage summation limit of 0.50 La, but less than the regulatory limit of 0.6 La, then an analysis and corrective action plan shall be prepared to restore the leakage summation value to less than the Callaway leakage limit. The corrective action plan should focus on those components which have contributed the most to the increase in the leakage summation value and what manner of timely corrective action, as deemed appropriate, best focuses on the prevention of future component leakage performance issues.

Response to Condition 2, ISSUE 2 If the potential leakage understatement adjusted leak rate MNPLR is less than the Callaway administrative leakage summation limit of 0.50 La, then the acceptability of the greater than a 60-month test interval up to the 75-month LLRT extension for all affected Type C components

Enclosure Evaluation of the Proposed Change 104 has been adequately demonstrated and the calculated local leak rate total represents the actual leakage potential of the penetrations.

In addition to Condition 1, ISSUES 1 and 2, which deal with the MNPLR Types B and C summation margin, NEI 94-01, Revision 3-A, also has a margin-related requirement as contained in Section 12.1, Report Requirements.

A post-outage report shall be prepared presenting results of the previous cycles Type B and Type C tests, and Type A, Type B and Type C tests, if performed during that outage. The technical contents of the report are generally described in ANSI/ANS-56.8-2002 and shall be available on-site for NRC review. The report shall show that the applicable performance criteria are met and serve as a record that continuing performance is acceptable. The report shall also include the combined Type B and Type C leakage summation, and the margin between the Type B and Type C leakage rate summation and its regulatory limit. Adverse trends in the Type B and Type C leakage rate summation shall be identified in the report and a corrective action plan developed to restore the margin to an acceptable level.

At Callaway, in the event an adverse trend in the aforementioned potential leakage understatement adjusted Types B and C summation is identified, then an analysis and determination of a corrective action plan shall be prepared to restore the trend and associated margin to an acceptable level. The corrective action plan shall focus on those components, which have contributed the most to the adverse trend in the leakage summation value and what manner of timely corrective action, as deemed appropriate, best focuses on the prevention of future component leakage performance issues.

At Callaway, an adverse trend is defined as three (3) consecutive increases in the final pre-mode change Types B and C MNPLR leakage summation values, as adjusted to include the estimate of applicable Type C leakage understatement, as expressed in terms of La.

3.9 Conclusion NEI 94-01, Revision 3-A, dated July 2012, and the limitations and conditions specified in NEI 94-01, Revision 2-A, dated October 2008, describe an NRC-accepted approach for implementing the performance-based requirements of 10 CFR 50, Appendix J, Option B. It incorporated the regulatory positions stated in RG 1.163 and includes provisions for extending Type A intervals to 15 years and Type C test intervals to 75 months. NEI 94-01, Revision 3-A, delineates a performance-based approach for determining Type A, Type B, and Type C containment leakage rate surveillance test frequencies. Callaway is adopting the guidance of NEI 94-01, Revision 3-A, and the limitations and conditions specified in NEI 94-01, Revision 2-A, for the Callaway, Unit 1, 10 CFR 50, Appendix J testing program plan.

Based on the previous ILRTs conducted at Callaway, Union Electric Company (d.b.a. Ameren Missouri) concludes that the permanent extension of the containment ILRT interval from 10 to 15 years represents minimal risk to increased leakage. The risk is minimized by continued Type

Enclosure Evaluation of the Proposed Change 105 B and Type C testing performed in accordance with Option B of 10 CFR 50, Appendix J, and the overlapping inspection activities performed as part of the following Callaway inspection programs:

x Containment Inservice Inspection Program, Subsection IWE x

Containment Inservice Inspection Program, Subsection IWL x

Callaway Coatings Program This experience is supplemented by risk analysis studies, including the Callaway risk analysis provided in Attachment 1. The risk assessment concludes that increasing the ILRT interval on a permanent basis to a one-in-fifteen-year frequency is not considered to be significant because it represents only a small change in the Callaway risk profile.

4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory Requirements/Criteria The proposed change has been evaluated to determine whether applicable regulations and requirements continue to be met.

10 CFR 50.54(o) requires primary reactor containments for water-cooled power reactors to be subject to the requirements of Appendix J to 10 CFR 50, Leakage Rate Testing of Containment of Water Cooled Nuclear Power Plants. Appendix J specifies containment leakage testing requirements, including the types required to ensure the leak-tight integrity of the primary reactor containment and systems and components which penetrate the containment. In addition, Appendix J discusses leakage rate acceptance criteria, test methodology, frequency of testing and reporting requirements for each type of test.

The adoption of the Option B performance-based containment leakage rate testing for Type A, Type B and Type C testing did not alter the basic method by which Appendix J leakage rate testing is performed; however, it did alter the frequency at which Type A, Type B, and Type C containment leakage tests must be performed. Under the performance-based option of 10 CFR 50, Appendix J, the test frequency is based upon an evaluation that reviewed "as-found" leakage history to determine the frequency for leakage testing which provides assurance that leakage limits will be maintained. The change to the Type A test frequency did not directly result in an increase in containment leakage. Similarly, the proposed change to the Type C test frequencies will not directly result in an increase in containment leakage.

EPRI TR-1009325, Revision 2-A (Reference 11), provided a risk impact assessment for optimized ILRT intervals up to 15 years, utilizing current industry performance data and risk informed guidance. NEI 94-01, Revision 3-A, Section 9.2.3.1 (Reference 2), states that Type A

Enclosure Evaluation of the Proposed Change 106 ILRT intervals of up to 15 years are allowed by this guideline. The Risk Impact Assessment of Extended Integrated Leak Rate Testing Intervals, EPRI Report 1018243 (formerly TR-1009325, Revision 2-A), indicates that, in general, the risk impact associated with ILRT interval extensions for intervals up to 15 years is small. However, plant-specific confirmatory analyses are required.

The NRC staff reviewed NEI TR 94-01, Revision 2, and EPRI Report No. 1009325, Revision 2.

For NEI TR 94-01, Revision 2, the NRC staff determined that it described an acceptable approach for implementing the optional performance-based requirements of Option B to 10 CFR 50, Appendix J. This guidance includes provisions for extending Type A ILRT intervals up to 15 years and incorporates the regulatory positions stated in RG 1.163. The NRC staff finds that the Type A testing methodology, as described in ANSI/ANS-56.8-2002 (Reference 37), and the modified testing frequencies recommended by NEI TR 94-01, Revision 2, serve to ensure continued leakage integrity of the containment structure. Type B and Type C testing ensures that individual penetrations are essentially leak tight. In addition, aggregate Type B and Type C leakage rates support the leakage tightness of primary containment by minimizing potential leakage paths.

For EPRI Report No. 1009325, Revision 2, a risk-informed methodology using plant-specific risk insights and industry ILRT performance data to revise ILRT surveillance frequencies, the NRC staff finds that the proposed methodology satisfies the key principles of risk-informed decision making applied to changes to TS as delineated in RG 1.174 (Reference 3) and RG 1.177, An Approach for Plant-Specific, Risk-Informed Decisionmaking: Technical Specifications (Reference 42). The NRC staff, therefore, found that this guidance was acceptable for referencing by licensees proposing to amend their TS in regard to containment leakage rate testing, subject to the limitations and conditions noted in Section 4.2 of the SE.

The NRC staff reviewed NEI TR 94-01, Revision 3, and determined that it described an acceptable approach for implementing the optional performance-based requirements of Option B to 10 CFR 50, Appendix J, as modified by the limitations and conditions summarized in Section 4.0 of the associated SE. This guidance included provisions for extending Type C LLRT intervals up to 75 months. Type C testing ensures that individual CIVs are essentially leak tight. In addition, aggregate Type C leakage rates support the leakage tightness of primary containment by minimizing potential leakage paths. The NRC staff, therefore, found that this guidance, as modified to include two limitations and conditions, was acceptable for referencing by licensees proposing to amend their TS in regard to containment leakage rate testing. Any applicant may reference NEI TR 94-01, Revision 3, as modified by the associated SER and approved by the NRC, and the limitations and conditions specified in NEI 94-01, Revision 2-A, dated October 2008, in a licensing action to satisfy the requirements of Option B to 10 CFR 50, Appendix J.

Enclosure Evaluation of the Proposed Change 107 4.2 Precedent This LAR is similar in nature to the following license amendments to extend the Type A Test Frequency to 15 years and the Type C test frequency to 75 months as previously authorized by the NRC in the associated referenced SERs:

x Braidwood Station, Units 1 and 2, and Byron Station, Units 1 and 2, issued September 10, 2020 (Reference 31 - ML20149K698) x McGuire Nuclear Station, Units 1 and 2, issued January 31, 2018 (Reference 24 - ML18009A842) x Vogtle Electric Generating Plant, Units 1 and 2, issued October 29, 2018 (Reference 25 - ML18263A039) 4.3 No Significant Hazards Consideration Union Electric Company (d.b.a. Ameren Missouri) proposes to amend the Technical Specifications (TS) for Callaway Plant, Unit No. 1 (Callaway) to allow extension of the Type A and Type C leakage test intervals. The extension is based on the adoption of the Nuclear Energy Institute (NEI) 94-01, Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J, Revision 3-A, and the conditions and limitations set forth in Revision 2-A.

Specifically, the proposed change revises Callaway TS 5.5.16, Containment Leakage Rate Testing Program, paragraph a., by replacing the reference to Regulatory Guide (RG) 1.163, Performance-Based Containment Leak-Test Program, with a reference to NEI 94-01, Revision 3-A and the conditions and limitations specified in NEI 01, Revision 2-A.

Ameren Missouri has evaluated whether a significant hazards consideration is involved with the proposed amendment by focusing on the three standards set forth in 10 CFR 50.92, Issuance of amendment, as discussed below:

1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No.

The proposed activity involves the revision of Callaway Plant, Unit No. 1 (Callaway),

Technical Specification (TS) Section 5.5.16, Containment Leakage Rate Testing Program, to allow the extension of the Type A integrated leakage rate test (ILRT) containment test interval to 15 years, and the extension of the Type C local leakage rate test (LLRT) interval to 75 months. The current Type A test interval of 120 months (10 years) would be extended on a permanent basis to no longer than 15 years from the last Type A test. The current

Enclosure Evaluation of the Proposed Change 108 Type C test interval of 60 months for selected components would be extended on a performance basis to no longer than 75 months. Extensions of up to nine months (total maximum interval of 84 months for Type C tests) are permissible only for non-routine emergent conditions.

The proposed test interval extensions do not involve either a physical change to the plant or a change in the manner in which the plant is operated or controlled. The containment is designed to provide an essentially leak tight barrier against the uncontrolled release of radioactivity to the environment for postulated accidents. As such, the containment and the testing requirements invoked to periodically demonstrate the integrity of the containment exist to ensure the plant's ability to mitigate the consequences of an accident, and do not involve the prevention or identification of any precursors of an accident.

The change in Type A test frequency to once-per-fifteen years, measured as an increase to the total integrated plant risk for those accident sequences influenced by Type A testing, based on the internal events (IE) probabilistic risk analysis (PRA) is 0.025 person-rem/year for Callaway. Electric Power Research Institute (EPRI) Report No. 1009325, Revision 2-A states that DYHU\\VPDOOSRSXODWLRQLVGHILQHGDVDQLQFUHDVHRISHUVRQ-rem per year or

RIWKHWRWDOSRSXODWLRQGRVHZKLFKHYHULVOHVVUHVWULFWLYHIRUWKHULVNLPSDFW

assessment of the extended ILRT intervals. This is consistent with the Nuclear Regulatory Commission (NRC) Final Safety Evaluation for Nuclear Energy Institute (NEI) 94-01 and EPRI Report No. 1009325. Moreover, the risk impact when compared to other severe accident risks is negligible. Therefore, this proposed extension does not involve a significant increase in the probability of an accident previously evaluated.

In addition, as documented in NUREG-1493, Performance-Based Containment Leak-Test Program, dated September 1995, Types B and C tests have identified a very large percentage of containment leakage paths, and the percentage of containment leakage paths that are detected only by Type A testing is very small. The Callaway Type A test history supports this conclusion.

The integrity of the containment is subject to two types of failure mechanisms that can be categorized as: (1) activity based, and (2) time based. Activity-based failure mechanisms are defined as degradation due to system and/or component modifications or maintenance.

The LLRT requirements and administrative controls such as configuration management and procedural requirements for system restoration ensure that containment integrity is not degraded by plant modifications or maintenance activities. The design and construction requirements of the containment combined with the containment inspections performed in accordance with American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (BPV) Code,Section XI, Rules for Inservice Inspection of Nuclear Power Plant Components; Containment Coatings Program; and TS requirements, serve to provide a high degree of assurance that the containment would not degrade in a manner that is detectable only by a Type A test. Based on the above, the proposed test interval extensions do not significantly increase the consequences of an accident previously evaluated.

Enclosure Evaluation of the Proposed Change 109 The proposed amendment also deletes TS 5.5.16.a. exceptions 3 and 4. The exception from post-modification ILRT associated with the Steam Generator Replacement (SGR) was previously approved by the NRC in TS Amendment 168. The performance of the Type A test no later than October 25, 2014, was previously approved by the NRC in TS Amendment No. 195. These exceptions were for activities that have already taken place; therefore, their deletion is solely an administrative action that has no effect on any component and no impact on how the unit is operated.

Therefore, the proposed change does not result in a significant increase in the probability or consequences of an accident previously evaluated.

2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No.

The proposed amendment to the TS 5.5.16, "Containment Leakage Rate Testing Program,"

paragraph a., involves the extension of the Callaway Type A containment test interval to 15 years and the extension of the Type C test interval to 75 months. The containment and the testing requirements to periodically demonstrate the integrity of the containment exist to ensure the plants ability to mitigate the consequences of an accident and do not involve any accident precursors or initiators. The proposed change does not involve a physical change to the plant (i.e., no new or different type of equipment will be installed), nor does it alter the design, configuration, or change the manner in which the plant is operated or controlled beyond the standard functional capabilities of the equipment.

The proposed amendment also deletes TS 5.5.16.a. exceptions 3 and 4. The exception from post-modification ILRT associated with the SGR was previously approved by the NRC in TS Amendment 168. The performance of the Type A test no later than October 25, 2014, was previously approved by the NRC in TS Amendment No. 195. These exceptions were for activities that have already taken place; therefore, their deletion is solely an administrative action that does not result in any change in how the unit is operated or controlled.

Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated.

3. Does the proposed change involve a significant reduction in a margin of safety?

Response: No.

The proposed amendment to the Callaway TS 5.5.16.a. involves the extension of the Callaway Type A containment test interval to 15 years and the extension of the Type C test interval to 75 months for selected components. This amendment does not alter the manner in which safety limits, limiting safety system set points, or limiting conditions for operation

Enclosure Evaluation of the Proposed Change 110 are determined. The specific requirements and conditions of the TS Containment Leak Rate Testing Program exist to ensure that the degree of containment structural integrity and leaktightness that is considered in the plant safety analysis is maintained. The overall containment leak rate limit specified by TS is maintained.

The proposed change involves only the extension of the interval between Type A containment leak rate tests and Type C tests for Callaway. The proposed surveillance interval extension is bounded by the 15-year ILRT interval and the 75-month Type C test interval currently authorized within NEI 94-01, Revision 3-A. Industry experience supports the conclusions that Types B and C testing detects a large percentage of containment leakage paths and that the percentage of containment leakage paths that are detected only by Type A testing is small. The containment inspections performed in accordance with ASME Section Xl, Containment Coatings Program; and TS serve to provide a high degree of assurance that the containment would not degrade in a manner that is detectable only by Type A testing. The combination of these factors ensures that the margin of safety in the plant safety analysis is maintained. The design, operation, testing methods and acceptance criteria for Types A, B, and C containment leakage tests specified in applicable codes and standards would continue to be met, with the acceptance of this proposed change, since these are not affected by changes to the Type A and Type C test intervals.

The proposed amendment also deletes TS 5.5.16.a. exceptions 3 and 4. The exception from post-modification ILRT associated with the SGR was previously approved by the NRC in TS Amendment 168. The performance of the Type A test no later than October 25, 2014, was previously approved by the NRC in TS Amendment No. 195. These exceptions were for activities that have already taken place; therefore, the deletion is solely an administrative action and does not change how the unit is operated and maintained. Thus, there is no reduction in any margin of safety as a result of this administrative change.

Therefore, the proposed change does not involve a significant reduction in a margin of safety.

Based on the above, Ameren Missouri concludes that the proposed amendment does not involve a significant hazards consideration under the standards set forth in 10 CFR 50.92(c),

and, accordingly, a finding of no significant hazards consideration is justified.

4.4 Conclusion In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commissions regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

Enclosure Evaluation of the Proposed Change 111

5.0 ENVIRONMENTAL CONSIDERATION

A review has determined that the proposed amendment would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would change an inspection or surveillance requirement. However, the proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluent that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure.

Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.

6.0 REFERENCES

1.

Regulatory Guide 1.163, "Performance-Based Containment Leak-Test Program,"

September 1995 (ML003740058)

2.

NEI 94-01, Revision 3-A, "Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J," July 2012 (ML12221A202)

3.

RG 1.174, Revision 3, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis," January 2018 (ML17317A256)

4.

RG 1.200, Revision 2, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities," March 2009 (ML090410014)

5.

NEI 94-01, Revision 0, "Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J," dated July 21, 1995 (ML11327A025)

6.

NUREG-1493, "Performance-Based Containment Leak-Test Program - Final Report,"

September 1995 (ML9510200161)

7.

Electric Power Research Institute (EPRI) Topical Report No. 104285, "Risk Impact Assessment of Revised Containment Leak Rate Testing Intervals, Palo Alto, California,"

August 1994

8.

NEI 94-01, Revision 2-A, "Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J," October 2008 (ML100620847)

Enclosure Evaluation of the Proposed Change 112

9.

Letter from NRC (M. J. Maxin) to NEI (J. C. Butler), "Final Safety Evaluation for Nuclear Energy Institute (NEI) Topical Report (TR) 94-01, Revision 2, 'Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J' and Electric Power Research Institute (EPRI) Report No. 1009325, Revision 2, August 2007, 'Risk Impact Assessment of Extended Integrated Leak Rate Testing Intervals' (TAC No.

MC9663)," dated June 25, 2008 (ML081140105)

10. Letter from NRC (S. Bahadur) to NEI (B. Bradley), "Final Safety Evaluation of Nuclear Energy Institute (NEI) Report, 94-01, Revision 3, 'Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J' (TAC No. ME2164)," dated June 8, 2012 (ML121030286)

11.

EPRI TR-1018243, "Risk Impact Assessment of Extended Integrated Leak Rate Testing Intervals: Revision 2-A of 1009325," October 2008

12.

Letter from NRC (K. M. Thomas) to Union Electric Company (D. Schnell), "Issuance of Amendment No. 111, Revise TS to directly reference Regulatory Guide 1.163 as required by 10 CFR 50, Appendix J, Option B for the Type A containment integrated leak rate tests and the Type B and C local leak rate tests, for Callaway Unit 1 (TAC No. M94801)," dated May 28, 1996 (ML960531185)

13. Letter from NRC (L. R. Wharton) to Union Electric Company (D. Schnell), "Issuance of Amendment No. 98, deferred the requirement to perform the Type A Containment Integrated Leak Rate Test until Refuel 8 (October 1996), for Callaway Unit 1 (TAC No.

M91529)," dated April 5, 1995 (ML9504130255 & ML9504130259)

14. Letter from NRC (M. Khanna) to Union Electric Company (G. L. Randolph), "Callaway Plant, Unit 1 - Issuance of Amendment [No. 160] Re: Containment Tendon Surveillance Program and Containment Leakage Rate Testing Program (TAC No. MC1497)," dated March 17, 2004 (ML040820660)

15. Letter from NRC (J. Donohew) to Union Electric Company (C. D. Naslund), "Callaway Plant, Unit 1 - Issuance of Amendment [No. 168] Regarding the Steam Generator Replacement Project (TAC No. MC4437)," dated September 29, 2005 (ML052570054)

16. Letter from NRC (M. C. Thadani) to Union Electric Company (A. C. Heflin), "Callaway Plant, Unit 1 - Issuance of Amendment [No. 195] Re: Revision to Technical Specification 5.5.16, 'Containment Leakage Rate Testing Program' (TAC No. ME0986)," dated March 17, 2010 (ML100601323)

17. RG 1.200, Revision 3, "Acceptability of Probabilistic Risk Assessment Results for Risk-Informed Activities," December 2020.

Enclosure Evaluation of the Proposed Change 113

18. Nuclear Energy Institute (NEI) Topical Report (TR) NEI 06-09, "Risk-Informed Technical Specifications Initiative 4b, Risk-Managed Technical Specifications (RMTS) Guidelines,"

Revision 0-A, October 12, 2012 (ML12286A322).

19. NUREG-2122, Glossary of Risk-Related Terms in Support of Risk-Informed Decision Making, dated November 2013 (ML13311A353)

20. Interim Guidance for Performing Risk Impact Assessments In Support of One-Time Extensions for Containment Integrated Leakage Rate Test Surveillance Intervals, Rev. 4, Developed for NEI by EPRI and Data Systems and Solutions, November 2001.

21. RG 1.200, Revision 0, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities," February 2004 (ML040630078)

22. RG 1.147, "Inservice Inspection Code Case Acceptability, ASME Section XI, Division 1, Revision 17," August 2014 (ML13339A689)

23. Risk Impact Assessment of Revised Containment Leak Rate Testing Intervals, EPRI, Palo Alto, CA, EPRI TR-104285, August 1994.

24. Letter from NRC (M. Mahoney) to Duke Energy (T. D. Ray), "McGuire Nuclear Station, Units 1 and 2 - Issuance of Amendments to Extend the Containment Type A Leak Rate Test Frequency to 15 Years and Type C Leak Rate Test Frequency to 75 Months (CAC Nos. MF9020 and MF9021; EPID L-2016-LLA-0032)," dated January 31, 2018 (ML18009A842)

25. Letter from NRC (M. Orenak) to Southern Nuclear Operating Company, Inc. (C. A.

Gayheart), "Vogtle Electric Generating Plant, Units 1 and 2, Issuance of Amendments to Extend the Containment Type A Leak Rate Test Frequency to 15 Years and Type C Leak Rate Test Frequency to 75 Months (CAC Nos. MG0240 and MG0241; EPID L-2017-LLA-0295)," dated October 29, 2018 (ML18263A039)

26. Regulatory Guide 1.11, "Instrument Lines Penetrating the Primary Reactor Containment"

27. ANSI N18.2-1973, "Nuclear Safety Criteria for the Design of Stationary Pressurized Water Reactor Plants"

28. Regulatory Guide 1.29, "Seismic Design Classification for Nuclear Power Plants"

29. NEI 00-02, "Probabilistic Risk Assessment (PRA) Peer Review Process Guidance Rev.

A3, PSA Peer Review Enclosures," dated March 20, 2000 (ML003728023)

Enclosure Evaluation of the Proposed Change 114

30. ASME/ANS, "Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications, ASME/ANS RA-Sa-2009," dated March 2009. Addendum A to RA-S-2008.

31. Letter from NRC (J. S. Wiebe) to Exelon Generation Co. (B. C. Hanson), "Braidwood Station, Units 1 and 2, and Byron Station, Unit Nos. 1 and 2 - Issuance of Amendment Nos. 215, 215, and 219 Re: Permanent Extension of Type A and Type C Containment Leak Rate Test Frequencies," (EPID L-2019-LLA-0208) (ML20149K698)

32. Letter from Constellation Nuclear (C. H. Cruse) to NRC (Document Control Desk), "Calvert Cliffs Nuclear Power Plant, Unit No. 1; Docket No. 50-317 - Response to Request for Additional Information Concerning the License Amendment Request for a One-Time Integrated Leakage Rate Test Extension," dated March 27, 2002 (ML020920100)

33. ASME/ANS, "Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications, ASME/ANS RA-Sb-2009," dated March 2009. Addendum B to RA-S-2008

34. Letter ULNRC-06526, "Request for License Amendment and Regulatory Exemptions for a Risk-Informed Approach to Address GSI-191 and Respond to GL 2004-02 (LDCN 19-0014)," dated March 31, 2021 (ML21090A184)

35. Letter to F. Diya (Ameren Missouri) from M. Chawla (NRC), "Callaway Plant, Unit No. 1 -

Issuance of Amendment No. 228 Re: Revise Technical Specifications to Address Generic Safety Issue-191 And Respond To Generic Letter 2004-02 Using A Risk-informed Approach (EPID L-2021-LLA-0059)," Dated October 21, 2022 (ML22220A132)

36. Letter from Entergy Operations, Inc. (K. Mulligan) to NRC (Document Control Desk),

"Grand Gulf Nuclear Station Response to Request for Additional Information Regarding License Amendment Request to Revise Technical Specifications for Containment Leak Rate Testing, Grand Gulf Nuclear Station, Unit 1, Docket No. 50-416, License No. NPF-29, (GNRO-2015/00063)," dated October 28, 2015 (ML15302A042)

37. American Nuclear Society, ANSI/ANS 56.8-2002, Containment System Leakage Testing Requirements, LaGrange Park, Illinois, November 2002

38. ASME Boiler & Pressure Vessel Code,Section XI, Subsection IWE, "Requirements for Class MC and Metallic Liners of Class CC Components of Light-Water Cooled Plants"

39. ASME B&PV Code,Section XI, Subsection IWL, "Requirements for Class CC Concrete Components of Light-Water Cooled Plant"

40. NRC's letter, "Callaway Plant, Unit No. 1-Audit Summary for License Amendment Request and Regulatory Exemptions for a Risk-Informed Approach to Address Generic

Enclosure Evaluation of the Proposed Change 115 Safety Issue-191 and Respond to Generic Letter 2004-02 (EPID L-2021 LLA 0059 and EPID L-2021-LLE-0021)," dated September 14, 2021 (ML21238A138).

41. Letter from T. Witt (Ameren Missouri) to NRC, "Response to Request for Additional Information Regarding Request for License Amendment and Regulatory Exemptions for risk-Informed Approach to Address GSI-191 and Respond to Generic Letter 2004-02 (LDCN 19-0014)," (EPID L-2021-LLA-0059 and EPID L-2021-LLE-0021. (ULNRC-06735)

42. RG 1.177, "An Approach for Plant-Specific, Risk-Informed Decisionmaking: Technical Specifications, Revision 1," May 2011 (ML100910008)

43. Letter from Jennifer M. Golder (NRC) to Biff Bradley (NEI), "Final Safety Evaluation for Nuclear Energy Institute (NEI) Topical Report (TR) NEI 06-09, "Risk-Informed Technical Specifications Initiative 4b, Risk-Managed Technical Specifications (RMTS) Guidelines,"

May 17, 2007 (ML071200238).

44. NEI Letter to NRC, "Final Revision of Appendix X to NEI 05-04/07-12/12-16, Close-Out of Facts and Observations (F&Os)," dated February 21, 2017 (ML17086A431).

45. NRC Letter to Mr. Greg Krueger (NEI), "U.S. Nuclear Regulatory Commission Acceptance on Nuclear Energy Institute Appendix X to Guidance 05-04, 7-12, and 12-13, Close-Out of Facts and Observations (F&Os)," dated May 3, 2017 (ML17079A427).

46. NEI 17-07, Revision 2, "Performance of PRA Peer Reviews Using the ASME/ANS PRA Standard," July 2019 (ML19228A242).

47. PRA-OEH-ANALYSIS, Other External Hazards: Screening Assessment Notebook, Revision 0.

48. Enclosure 4, License Amendment Request: Revise Technical Specifications to Adopt Risk-Informed Completion Times TSTF-505, Revision 2, Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b, Information Supporting Justification of Excluding Sources of Risk Not Addressed by the PRA Models.

49. ASME/ANS RA-S-2009, Addenda to ASME/ANS RA-S-2008, "Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications," February 2009.

50. AMN#PES00031-REPT-001, "Callaway Energy Center Probabilistic Risk Assessment Focused Scope Peer Review," July 2020.

51. AMN#PES00031-REPT-002, "Callaway Energy Center Probabilistic Risk Assessment Peer Review F&Os Closure," July 2020.

Enclosure Evaluation of the Proposed Change 116

52. PWROG-19022-P, "Peer Review of the Callaway External Hazard Screening Assessment and High Winds Probabilistic Risk Assessment," April 2019.

53. PWROG-19034-P, "Independent Assessment of Facts & Observations Closure and Focused Scope Peer Review of the Callaway Probabilistic Risk Assessments," November 2019.

54. PWROG-18027-NP Revision 0, "Loss of Room Cooling in PRA Modeling," April 2020.

55. PWROG-19012-P, "Peer Review of the Callaway Internal Events and Internal Flood Probabilistic Risk Assessment Model," April 2019.

56. PWROG-19020-NP Revision 1, "Newly Developed Method Peer Review Pilot - General Screening Criteria for Loss of Room Cooling in PRA Modeling Risk Management Committee," PA-RMSC-1647, Revision 1, April 2020.

57. ASME/ANS RA-S CASE 1, Case for ASME/ANS RA-Sb-2013, "Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications," ASME and ANS, November 2017.

58. PWROG-18044-P, "Peer Review of the Callaway Seismic Probabilistic Risk Assessment,"

June 2018.

59. NRC Letter, "U.S. Nuclear Regulatory Commission Acceptance of ASME/ANS RA-S Case 1," March 12, 2018 (ML18017A964 and ML18017A966).

60. PWROG-19011-P, "Independent Assessment of Facts & Observations Closure and Focused Scope Peer Review of the Callaway Seismic Probabilistic Risk Assessment,"

March 2019.

61. NUREG/CR-6850 (also EPRI 1011989), "Fire PRA Methodology for Nuclear Power Facilities," September 2005, with Supplement 1 (EPRI 1019259), September 2010.

62. LTR-RAM-II-10-019, "Fire PRA Peer Review Against the Fire PRA Standard SRs From Section 4 of the ASME/ANS Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessments for Nuclear Power Plant Applications for The Callaway Nuclear Plant Fire PRA," October 2009.

63. AMN#PES00021-REPT-001, "Callaway Energy Center Fire Probabilistic Risk Assessment Peer Review F&Os Closure," June 2019.

64. AMN#PES00042-REPT-002, "Callaway Energy Center Fire Probabilistic Risk Assessment Peer Review F&Os Closure Review," February 2021.

Enclosure Evaluation of the Proposed Change 117

65. AMN#PES00031-REPT-003, "Callaway Energy Center Probabilistic Risk Assessment Focused Scope Peer Review," November 2020.

66. ASTM D5163-08, "Standard Guide for Establishing a Program for Condition Assessment of Coating Service Level I Coating Systems in Nuclear Power Plants."

67. Callaway Plant, Unit No. 1 - Audit Plan and Setup of Online Reference Portal for License Amendment Request Regarding Risk-Informed Approach for Closure of Generic Safety Issue-191 (EPID L-2021-LLA-0059)

68. RG 1.35, "Inservice Inspection of Ungrouted Tendons in Prestressed Concrete Containments, Proposed Revision 3," dated April 1979

69. Proposed RG 1.35.1, "Determining Prestressing Forces for Inspection of Prestressed Concrete Containments," dated April 1979.

70. ULNRC-3934, "Request for Relief from Certain American Society of Mechanical Engineers Code Requirements for Inservice Inspection of the Callaway Plant Unit 1 Containment Building," December 15, 1998

71. APC 17-13,

Subject:

NRC Acceptance of Industry Guidance on Closure of PRA Peer Review Findings, dated May 8, 2017

72. RG 1.54, Service Level I, II, and III Protective Coatings Applied to Nuclear Power Plants, Revision 2.

73. RIS-04-12, "Clarification on use of Later Editions and Addenda to the ASME OM Code and Section XI," dated July 2004