Text

License Amendment Request for Risk-Informed Approach to Resolution of Generic Letter 2004-02
	ML21280A381
Person / Time
Site:	Callaway
Issue date:	10/07/2021
From:	; Ameren Missouri
To:	; Office of Nuclear Reactor Regulation
Shared Package
ML21280A378	List: ML21280A379; ML21280A380; ML21280A381; ULNRC-06692, (Replacement for Enclosure 3, Callaway Methodology for a Risk-Informed Approach to Address Generic Letter 2004-02, of ULNRC-06526);
References
	ULNRC-06692
	Download: ML21280A381 (107)
	v • d • e

ULNRC-06692 Enclosure 2 to ULNRC-06692 (Replacement for Enclosure 2, "License Amendment Request,"

of ULNRC-06526)

(106 pages)

ULNRC-06526 Enclosure 2 License Amendment Request for Callaway Risk-Informed Approach to Resolution of Generic Letter 2004-02 ATTACHMENTS:

2-1 List of Regulatory Commitments 2-2 Technical Specification Page Markups 2-3 Technical Specification Bases Page Markups (for information only) 2-4 Re-typed Technical Specification Pages 2-5 Final Safety Analysis Report Page Markups (for information only)

ULNRC-06526 Enclosure 2 Page 1 of 36 License Amendment Request for Callaway Risk-Informed Approach to Resolution of Generic Letter 2004-02

Subject:

Pursuant to 10 CFR 50.90, Union Electric Company (d.b.a. Ameren Missouri) requests amendment of Renewed Operating License NPF-30 for Callaway Plant, Unit 1. The proposed amendment would revise the Final Safety Analysis Report and Technical Specifications for Callaway Plant, Unit 1 using a risk-informed approach to address safety issues discussed in Generic Safety Issue 191, Assessment of Debris Accumulation on Pressurized-Water Reactor Sump Performance [1], and NRC Generic Letter 2004-02, "Potential Impact of Debris Blockage on Emergency Recirculation during Design Basis Accidents at Pressurized-Water Reactors" [18].

1.

SUMMARY

DESCRIPTION

2. DETAILED DESCRIPTION 2.1 Current System Descriptions 2.2 Current Technical Specification Requirements 2.3 Reason for the Proposed Changes 2.4 Description of the Proposed Change

3. TECHNICAL EVALUATION

3.1 Background

3.2 Evaluation

4. REGULATORY EVALUATION 4.1 Applicable Regulatory Requirements/Criteria 4.2 Precedent 4.3 No Significant Hazards Consideration Determination 4.4 Conclusion

5. ENVIRONMENTAL CONSIDERATION

6. REFERENCES

ULNRC-06526 Enclosure 2 Page 2 of 36

1.

SUMMARY

DESCRIPTION In accordance with 10 CFR 50.59(c)(1)(i) and (c)(2)(viii), Ameren Missouri requests an amendment to Operating License NPF-30 for Callaway Nuclear Plant Unit 1 (Callaway) pursuant to 10 CFR 50.90. The proposed amendment would:

1. Revise the licensing basis as described in the Callaway Final Safety Analysis Report (FSAR) to allow the use of a risk-informed approach to address safety issues discussed in Generic Letter (GL) 2004-02, Potential Impact of Debris Blockage on Emergency Recirculation during Design Basis Accidents at Pressurized-Water Reactors" [2]. The risk-informed approach is consistent with the guidance of NRC Regulatory Guide (RG) 1.174, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis" [3].

2. Revise the Technical Specification (TS) for the Emergency Core Cooling System (ECCS) by deleting Surveillance Requirement (SR) 3.5.2.8 in TS 3.5.2, "ECCS -

Operating," and deleting its mention from SR 3.5.3.1 in TS 3.5.3, "ECCS -

Shutdown."

3. Add new TS 3.6.8, Containment Sumps, with appropriate Conditions, Required Actions and Completion Times, including new SR 3.6.8.1 for visual inspection of the containment sumps.

4. Revise TS 5.5.15, Safety Function Determination Program, to clarify its application when a supported system is made inoperable by a single TS support system.

In accordance with 10 CFR 50.59(c)(2)(viii), a license amendment shall be obtained prior to implementing a proposed change if the change would result in a departure from a method of evaluation described in the FSAR (as updated) used in establishing the design bases or in the safety analysis. Ameren Missouri proposes to amend the Callaway Operating License in order to add risk-informed methodology to FSAR Chapter 6, "Engineered Safety Features." (Chapters 3, "Design of Structures, Components, Equipment, and Systems," and 15, "Accident Analysis," will also refer to it.) Ameren Missouri also proposes to revise the Technical Specifications for the ECCS and Containment Spray System (CSS) in accordance with 10 CFR 50.59(c)(1)(i) and the NRC Safety Evaluation of Technical Specifications Task Force (TSTF) Improved Standard Technical Specifications Change Traveler TSTF-567-A, Rev. 1 "Add Containment Sump TS to Address GSI-191 Issues" [4]. The proposed TS changes would align the Callaway Technical Specifications with the risk-informed methodology change, and the proposed changes would apply only for the effects of debris as described in GSI-191 and GL 2004-02.

ULNRC-06526 Enclosure 2 Page 3 of 36

2. DETAILED DESCRIPTION The proposed change associated with the change in methodology is to use a risk-informed approach to determine the design requirements needed to address the effects of loss-of-coolant accident (LOCA) debris instead of a traditional deterministic-only approach. The details of the approach are provided in Enclosure 3 of this submittal.

The debris analysis covers a full spectrum of postulated LOCAs, including a range of partial breaks and double-ended guillotine breaks (DEGBs), for all pipe sizes up to and including the design-basis accident (DBA) LOCA in order to provide assurance that the most severe postulated loss-of-coolant accidents are evaluated. The deterministic licensing basis will continue to apply to LOCA break sizes that generate fine fiber debris that is bounded by Callaway plant-specific testing. The proposed methodology change would apply for LOCAs that can generate and transport fine fiber debris that is not bounded by the plant-specific testing. In the risk-informed approach, Callaway conservatively relegates to failure the LOCA break sizes that can generate and transport fine fiber debris that is not bounded by the Callaway plant-specific testing.

The proposed risk-informed methodology change would apply NUREG-1829 [5] to determine the break frequency for the smallest of those breaks to obtain the highest frequency and uses that frequency as the incremental (delta) core damage frequency (CDF) for comparison to the criteria in RG 1.174. The results of the evaluation show that the risk from the proposed change is "very small" in that it is in Region III of RG 1.174. The methodology includes conservatisms in the plant-specific testing and in the assumption that all the unbounded breaks are relegated to failure.

The proposed TS changes associated with the change in methodology would include a new TS, i.e., TS 3.6.8, "Containment Sumps," that contains Conditions and Required Actions with Completion Times that apply when the Limiting Condition for Operation (LCO) is not met. The ECCS and CSS are the only TS systems that depend on the containment sumps as a support system, and are therefore the only systems that are directly subject to the effects of debris. The purpose of the new TS is to have a TS dedicated to the containment sumps in lieu of having the sumps addressed only as a support system behind the ECCS and CSS Technical Specifications, and to establish Conditions and Actions that address the concerns of potential debris effects, including a required action time (Completion Time) for restoring compliance with the LCO that is commensurate with the very low risk associated with debris effects. The proposed action would be required if the limit for analyzed LOCA-generated and transported debris is determined to be exceeded, which is based on the amount of debris used in the Callaway plant-specific testing. The proposed completion time for this TS action is 90 days. In addition, a periodic surveillance to verify by visual inspection that the containment sumps do not show structural damage, abnormal corrosion, or debris blockage would be established per proposed SR 3.6.8.1. A more complete description of the proposed TS changes is provided in Section 2.4.

ULNRC-06526 Enclosure 2 Page 4 of 36 The proposed change to TS 5.5.15 "Safety Function Determination Program," (SFDP) clarifies its application when a supported system is made inoperable by the inoperability of a single TS support system.

The proposed change to the licensing basis implements a risk-informed approach in lieu of an entirely deterministic method to demonstrate acceptable system responses to a LOCA. In conjunction with the proposed license amendment request (LAR), Callaway is also requesting exemptions from 10 CFR 50.46(a)(1), General Design Criterion (GDC) 35, GDC 38 and GDC 41, as provided in Attachments 1-1 through 1-4 of of this submittal. Upon approval of the licensing basis changes, Callaway will make conforming updates to the FSAR. The FSAR markups are attached for the staffs information.

2.1 System Design and Operation The methodology change affects the analysis of systems and functions that are susceptible to the effects of LOCA debris. The affected systems are those that are supported by the strainers and sumps during the recirculation phase of LOCA mitigation, which are the ECCS (i.e., residual heat removal (RHR) pumps, safety injection (SI) pumps, and centrifugal charging pumps (CCPs)) and the CSS. The associated functions and associated regulations are:

Emergency Core Cooling: 10 CFR 50.46(a)(1) and GDC 35 Containment Heat Removal: GDC 38 Containment Atmosphere Cleanup: GDC 41 Emergency Core Cooling System The ECCS is designed to cool the reactor core and provide shutdown capability subsequent to the following accident conditions:

1. LOCA, including a pipe break or a spurious relief or safety valve opening in the reactor coolant system (RCS) which would result in a discharge larger than that which could be made up by the normal makeup system.

2. Rupture of a control rod drive mechanism, causing a rod cluster control assembly ejection accident.

3. Steam or feedwater system break accident, including a pipe break or a spurious relief or safety valve opening in the secondary steam system which would result in an uncontrolled steam release or a loss of feedwater.

4. A steam generator tube failure.

The primary function of the ECCS is to provide emergency core cooling in the event of a LOCA resulting from a break in the primary RCS or to provide emergency boration in

ULNRC-06526 Enclosure 2 Page 5 of 36 the event of a steam/or feedwater break accident resulting from a break in the secondary steam system.

The ECCS is safety related and is required to function following a DBA and to achieve and maintain the plant in a safe shutdown condition.

The ECCS meets the following design bases:

1. Except for the refueling water storage tank (RWST), the ECCS is protected from the effects of natural phenomena, such as earthquakes, tornadoes, hurricanes, floods, and external missiles (GDC 2). The RWST is designed to seismic Category I requirements only.

2. The ECCS is designed to remain functional after a safe shutdown earthquake (SSE) and to perform its intended function following the postulated hazards of fire, internal missiles, or pipe break (GDC 3 and 4).

3. Safety functions can be performed, assuming a single active component failure coincident with the loss of offsite power (GDC 35).

4. The active components are capable of being tested during plant operation.

Provisions are made to allow for in-service inspection of components at appropriate times specified in the ASME Boiler and Pressure Vessel Code,Section XI (GDC 36 and 37).

5. The ECCS is designed and fabricated to codes consistent with the quality group classification assigned by 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.

6. The capability to isolate components or piping is provided so that the ECCS safety function will not be compromised. This includes isolation of components to deal with leakage or malfunctions and to isolate nonsafety-related portions of the system (GDC 35).

7. The containment isolation valves in the system are selected, tested, and located in accordance with the requirements of GDC 54 and 55 and 10 CFR 50, Appendix J, Type A testing.

8. ECCS equipment design ensures acceptable performance for all environments anticipated under normal, testing, and design-basis accident conditions.

9. The functional requirements of the ECCS are derived from 10 CFR 50, Appendix K limits for fuel cladding temperature, etc., following any of the above accidents, as delineated in 10 CFR 50.46. The subsystem functional parameters are

ULNRC-06526 Enclosure 2 Page 6 of 36 integrated so that the Appendix K requirements are met over the range of anticipated accidents and single failure assumptions.

There are no power generation design bases for the ECCS function. Portions of the ECCS are also portions of the residual heat removal system (RHRS) and chemical and volume control system (CVCS), and are used during normal power operation.

10 CFR 50.46(b) provides the following criteria to judge the adequacy of the ECCS.

1. Peak clad temperature calculated shall not exceed 2,200°F.

2. The calculated total oxidation of the clad shall nowhere exceed 0.17 times the total clad thickness before oxidation.

3. The calculated total amount of hydrogen generated from the chemical reaction of the clad with water or steam shall not exceed 0.01 times the hypothetical amount that would be generated if all of the metal in the clad cylinders surrounding the fuel, excluding the clad around the plenum volume, were to react.

4. Calculated changes in core geometry shall be such that the core remains amenable to cooling.

5. After any calculated successful initial operation of the ECCS, the calculated core temperature shall be maintained at an acceptably low value and decay heat shall be removed for the extended period of time required by long-lived radioactivity remaining in the core.

In addition to, and as an extension of the 10 CFR 50.46(b) Final Acceptance Criteria, two accidents have more specific criteria, as described below.

In the case of the inadvertent opening of a steam generator relief or safety valve, an additional criterion for adequacy of the ECCS is: Assuming a stuck rod cluster control assembly (RCCA), offsite power available, and a single failure in the engineered safety features, there will be no return to criticality after reactor trip for a steam release equivalent to the spurious opening (with failure to close), of the larger of a single steam dump, relief, or safety valve.

For a steam system piping failure, the added criterion is: Assuming a stuck RCCA with or without offsite power, and assuming a single failure in the engineered safety features, the core remains in place and intact.

These two accidents and others listed in the ECCS description (e.g., main steam line break, reactor coolant pump seal leak) do not require the use of the containment sump strainers to mitigate accident generated and transported debris. Enclosure 3, the technical analysis performed for this LAR, provides further details on how these secondary risk contributors are assessed.

ULNRC-06526 Enclosure 2 Page 7 of 36 Containment Heat Removal System (CHRS)

The functional performance objective of the containment heat removal system, as an engineered safety features system, is to reduce the containment temperature and pressure following a LOCA or main steam line break (MSLB) accident by removing thermal energy from the containment atmosphere. These cooling systems also serve to limit offsite radiation levels by reducing the pressure differential between the containment atmosphere and the external environment, thereby diminishing the driving force for the leakage of fission products from the containment to the environment. The containment heat removal systems include the residual heat removal system (see ECCS system description), the CSS, and the containment cooling system (CtCS).

The CSS:

The CSS consists of two separate trains of equal capacity, each independently capable of meeting the design bases. Each train includes a containment spray pump, spray header and nozzles, spray recirculation path, valves, and the necessary piping, instrumentation, flushing connections, and controls.

The RWST supplies borated injection water to the CSS. Each train takes suction from separate containment sumps during the recirculation phase.

The CSS provides a spray of cold or subcooled borated water from the upper regions of the containment to reduce the containment pressure and temperature during either a LOCA or MSLB inside the containment.

Each CSS pump discharges into the containment atmosphere through an independent spray header. The spray headers are located in the upper part of the reactor building to allow maximum time for the falling spray droplets to reach thermal equilibrium with the steam-air atmosphere. The condensation of the steam by the falling spray results in a reduction in containment pressure and temperature. Each spray train provides adequate coverage to meet the design requirements with respect to both containment heat removal and iodine removal.

In the CSS, only the containment sumps, the trisodium phosphate baskets, the spray headers, nozzles, and associated piping and valves are located within the containment.

The remainder of the system is located within the auxiliary building, separated from that portion in the containment by motor-operated isolation valves.

Following a large break LOCA, the containment spray during the injection phase will be a boric acid solution having a pH of about 4.5. The desired pH level is greater than 7.0 to assure iodine retention in the sumps, to limit corrosion and the associated production of hydrogen, and to limit chloride induced stress-corrosion cracking of austenitic stainless steels. To adjust the sump solution pH into the desired range, a minimum of 9,000 pounds of trisodium phosphate dodecahydrate (NA3PO4

12H2O

1/4 NaOH) is stored in two baskets, one adjacent to each containment sump, at an elevation to

ULNRC-06526 Enclosure 2 Page 8 of 36 assure dissolution after a LOCA. This amount of trisodium phosphate is sufficient to assure that the equilibrium sump solution pH will be greater than or equal to 7.1.

The CSS meets the following design bases:

1. The CSS is protected from the effects of natural phenomena, such as earthquakes, tornadoes, hurricanes, floods, or external missiles (GDC 2).

2. The CSS is designed to remain functional after a SSE or to perform its intended function following the postulated hazard of a pipe break (GDC 3 and 4).

3. Safety functions can be performed, assuming a single active component failure coincident with the loss of offsite power (GDC 38).

4. The active components are capable of being tested during plant operation.

Provisions are made to allow for in-service inspection of components at appropriate times specified in the ASME Boiler and Pressure Vessel Code,Section XI (GDC 39 and 40).

5. The CSS is designed and fabricated to codes consistent with the quality group classification assigned by Regulatory Guide 1.26 and the seismic category assigned by Regulatory Guide 1.29. The power supply and control functions are in accordance with Regulatory Guide 1.32.

6. The capability of isolating components or piping is provided so that the CSS safety function will not be compromised. This includes isolation of components to deal with leakage or malfunctions (GDC 38).

7. The containment isolation valves in the system are selected, tested, and located in accordance with the requirements of GDC 54 and 56 and 10 CFR 50, Appendix J, Type A testing.

8. The CSS, in conjunction with the containment fan cooler system and the emergency core cooling system, is designed to be capable of removing sufficient heat and subsequent decay heat from the containment atmosphere following the hypothesized LOCA or MSLB to maintain the containment pressure below the containment design pressure.

9. The CSS remains operable in the accident environment.

10. The containment spray water does not contain substances which would be unstable in the thermal or radiolytic environment of the LOCA or cause extensive corrosive attack on equipment.

ULNRC-06526 Enclosure 2 Page 9 of 36

11. The CSS is designed so that adequate net positive suction head (NPSH) exists at the suction of the containment spray pumps during all operating phases, in accordance with Regulatory Guide 1.1.

12. The CSS is designed to prevent debris which could impair the performance of the containment spray pumps, valves, eductors, or spray nozzles from entering the recirculation piping. Design is in accordance with Regulatory Guide 1.82.

The CtCS:

The CtCS, in conjunction with the containment HVAC systems, functions during normal plant operation to maintain a suitable atmosphere for equipment located within the containment. Subsequent to a DBA within the containment, the containment cooling system provides a means of cooling the containment atmosphere to reduce pressure and thus reduce the potential for containment leakage of airborne and gaseous radioactivity to the environment.

The CtCS provides cooling by recirculation of the containment air across air-to-water heat exchangers. The bulk of this cooled air is supplied to the lower regions of the steam generator compartments. The remaining air is supplied to the instrument tunnel and at each level (operating floor and below) of the containment outside the secondary shield wall. The air supplied to each steam generator compartment is drawn upwards through the compartments by the hydrogen mixing fans and discharged into the upper elevations of the containment.

Combustible Gas Control in Containment The hydrogen control system (HCS) is an engineered safety feature which serves to control combustible gas concentrations in the containment. The HCS consists of redundant hydrogen recombiners, a redundant hydrogen mixing system, redundant hydrogen monitoring subsystem, and a backup hydrogen purge subsystem.

Sump Design To address debris-related concerns associated with GSl-191 and in response to the debris issues identified in GL 2004-02, new containment sump strainers were installed in April 2007 to replace the previously installed screens. The wetted surface area of the strainers was increased from approximately 400 square feet to approximately 6,600 square feet. The screen-hole size of the strainers was reduced from 1/8 inches to 0.045 inches. Small particles in water entering the suction pipe from the sump cannot clog the containment spray nozzles (which have a minimum constriction size of 7/16).

Previously, the sump screens had extended above the maximum, post-accident containment water level and would not be submerged. The replacement sump strainers are now inside the containment sump pit ensuring maximum strainer surface area is available during post-accident recirculation mode. Installation of the new strainers did

ULNRC-06526 Enclosure 2 Page 10 of 36 not affect the independence and redundancy of the sumps; each one of the two sumps has sufficient capacity to serve one of the redundant halves of the ECCS and CSS.

The sump strainer design implemented by these modifications meets the current design-basis requirements with respect to NPSH and ECCS performance. The sumps are designed according to RG 1.82 Revision 0, dated June 1974, which recommends a calculation of sump screen head loss due to 50% debris blockage of the wetted surface.

Utilizing the current licensing basis methodology, the pump NPSH is sufficient to accommodate this head loss. The Callaway sumps meet the function to preclude passage of debris particles large enough to damage downstream components in the ECCS and CSS.

The sump strainer design has been evaluated to meet the current licensing basis assumptions for analyzing the effects of post-accident debris blockage and for compliance with 10 CFR 50.46 for long term cooling, GDC 35 for emergency core cooling, GDC 36 for inspection of ECCS, GDC 38 for containment heat removal, GDC 39 for inspection of containment heat removal system, and GDC 41 for containment atmosphere cleanup.

The containment sumps meet each position of Regulatory Guide 1.82, Revision 0:

1. Two sumps are provided, and each has sufficient capacity to serve one of the redundant trains of the ECCS and CS systems.

2. The redundant sumps are physically separated from each other and from high energy piping.

3. The sumps are located at the 2000' elevation, which is the lowest floor elevation in the reactor building, exclusive of the reactor cavity. The strainers are installed in the recirculation sump pit and extend approximately one foot above the 2000 elevation of the Reactor Building. The intent is met.

4. The floor is level in the vicinity of the sump. However, a 6-inch concrete curb is provided which prevents high density particles from entering the sump.

5. All drains in the upper regions of the reactor building are terminated in such a manner that direct streams of water which may contain entrained debris will not impinge on the filter assemblies.

6. The containment sump strainers are fabricated from stainless steel perforated plate, including structural reinforcement, and are sufficiently rigid to preclude the use of a trash rack. The structural evaluation for the strainers concludes that the strainers meet the acceptance criteria for all applicable loadings during the recirculation phase of an event. The sumps and strainers are outside the secondary shield wall which provides protection from missiles and large debris.

The intent is met.

ULNRC-06526 Enclosure 2 Page 11 of 36

7. The containment sump strainers are composed of stainless steel perforated plate with 0.045-inch diameter holes. The approach velocity of the recirculation coolant flow at the sump strainer face is less than 0.2 ft/sec.

8. A concrete slab over the containment sump strainers is provided. The containment recirculation sump strainers will be fully submerged following a large break LOCA.

9. The containment sump strainers are designed as seismic Category I and have been evaluated acceptably for all applicable loadings.

10. The containment sump strainers have a nominal 0.045-inch hole size, which precludes particles larger than 0.045 inches from passing through the strainers.

The containment spray pump is designed to pass particles less than 1/4 inch in size, and the minimum restriction in the spray system is the 7/16-inch orifice in the spray nozzle.

11. The pump intake location in the sump is horizontal to limit any degrading effects due to vortexing.

12. The containment sump strainers are fabricated from stainless steel. Stainless steel has a low sensitivity to corrosion during power operation and after an event.

13. The containment sump strainers are provided with provisions to allow inspection of the strainer structure and areas downstream of the strainer.

14. In-service inspection requirements consist of visual examination during every scheduled refueling downtime.

The locations of the strainers provide significant protection from dynamic effects such as pipe whip, jet impingement, and missile impacts associated with a high-energy line break. The containment sump strainers are outside the secondary shield wall and are located inside a pit where approximately 1 foot of the strainers is above the reactor containment building floor. A concrete structure is also approximately 7 feet above the strainers, and a concrete wall divides the strainer trains. Structural steel that stabilizes the top of the strainer stacks also provides protection.

The replacement strainer design is a safety improvement that contributed to meeting the RG 1.174 criteria for Region III, "Very Small Changes," for the results obtained from the risk-informed methodology.

2.2 Current Technical Specification Requirements Under the current TS, the operability of the ECCS Subsystems is assured by the capability of the containment sump strainers to limit entry of debris into the sumps and recirculating lines and also allow adequate flow into the system. This capability ensures that the flow and net positive suction head requirements of ECCS are satisfied.

ULNRC-06526 Enclosure 2 Page 12 of 36 Assurance that containment debris will not block the sump strainers and render the ECCS Subsystem inoperable on emergency recirculation during design-basis accidents is provided by inspection and engineering evaluation.

2.2.1 TS 3.5.2 ECCS - Operating TS 3.5.2 requires two ECCS trains to be OPERABLE during Modes 1, 2, and 3. (Each train must include an ECCS centrifugal charging pump (high-head pump), a safety injection pump (intermediate-head pump), and a residual heat removal pump (low-head pump). The TS provides an Action (i.e., Required Action A.1) that allows 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months to restore one or more inoperable trains to meet the LCO requirement of having two ECCS trains OPERABLE, provided at least 100% of the ECCS flow equivalent to a single operable ECCS train is available. (The latter is met by having a combination of one high-head, one intermediate-head, and one low-head pump available from either train.)

Reliability analysis has shown that the impact of having one full ECCS train inoperable is sufficiently small to justify continued operation for 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . If the inoperable train is not restored within the 72-hour Completion Time, Required Action B.1 allows 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months in to be MODE 3, and in parallel, Required Action B.2 allows 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months to be in MODE 4.

2.2.2 TS 3.5.3 ECCS - Shutdown TS 3.5.3 requires one ECCS train to be OPERABLE during Mode 4. (For Mode 4, the required OPERABLE train must include an ECCS centrifugal charging pump and a residual heat removal pump.) The TS provides an Action (i.e., Required Action A.1) to initiate action to restore the required ECCS RHR subsystem to OPERABLE status immediately, and an Action (i.e., Required Action B.1) to restore the required ECCS Centrifugal Charging Pump subsystem to OPERABLE status within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . If at least one ECCS Centrifugal Charging Pump subsystem is not restored to OPERABLE status within the 1-hour Completion Time, Required Action C.1 allows 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to be MODE 5.

2.2.3 TS 3.6.6 Containment Spray and Cooling Systems The TS provides an Action (i.e., Required Action A.1) that allows 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months to restore an inoperable containment spray train to OPERABLE status (and 10 days from discovery of failure to meet the LCO) in order to meet the requirement of having two containment spray trains and two containment cooling trains OPERABLE. The 72-hour Completion Time is based on the ability of the remaining operable train to provide adequate coverage to meet design requirements with respect to both containment heat removal and iodine removal. The 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months takes into account reasonable time for repairs and the low probability of a DBA occurring during this period. If the inoperable train is not restored within the 72-hour Completion Time, Required Action B.1 allows 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months to be in MODE 3, and in parallel, Required Action B.2 allows 84 hours9.722222e-4 days 0.0233 hours 1.388889e-4 weeks 3.1962e-5 months to be in MODE 5. If one containment cooling train is inoperable, the TS provides an Action (i.e., Required Action C.1) that allows 7 days to restore containment spray train to OPERABLE status (and 10 days from discovery of failure to meet the LCO). If the inoperable containment cooling train is not restored within the 7-day Completion Time, Required Action D.1

ULNRC-06526 Enclosure 2 Page 13 of 36 allows 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months to be in MODE 3, and in parallel, Required Action D.2 allows 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months to be in MODE 5. If two spray trains or two cooling trains are inoperable, Required Action E.1 allows 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months to be in Mode 3, and in parallel, Required Action E.2 allows 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months to be in Mode 5.

2.2.4 TS 3.6.8 Containment Sumps There is currently no Containment Sumps TS.

2.2.5 TS 5.5.15 Safety Function Determination Program (SFDP)

The current SFDP ensures loss of safety function is detected and appropriate actions are taken. Additionally, other actions may be taken as a result of support system inoperability and corresponding exception to entering supported system Conditions and Required Actions.

2.3 Reason for the Proposed Changes Callaway is proposing to change the Technical Specifications related to the ECCS and CSS in order to address the effects of HELB generated and transported debris. The ECCS and CSS are the only systems potentially affected by debris effects because they are the only systems that are supported by the containment sumps and their strainers.

As noted previously, the purpose of the proposed, new Technical Specification for the containment sumps, i.e., TS 3.6.8, is to have a TS dedicated to the containment sumps in lieu of having the sumps addressed only as a support system behind the ECCS and CSS Technical Specifications, and to establish Conditions and Actions that address the concerns of potential debris effects.

2.4 Description of Proposed Changes Revising the plant's licensing basis to reflect resolution of the concerns identified per GSI-191 and GL 2004-02 requires changes to be made to the Technical Specifications and their associated TS Bases, as well as the FSAR.

2.4.1 Proposed TS Changes The changes proposed for the Callaway Technical Specifications are based on the changes proposed and described in NRC-approved TSTF-567-A, Rev. 1.

Of the various options pursued by licensees for resolving the GSI-191 and GL 2004-02 concerns, it is noted in the approved TSTF-567 package that the TS changes proposed per the TSTF may be considered applicable to all plants, regardless of the GSI-191 closure option selected. In addition, a model license amendment application for adopting the TSTF is included in the TSTF-567 package. In regard to that, however, it is noted in the TSTF package that the model application is intended for use by Option 1

ULNRC-06526 Enclosure 2 Page 14 of 36 plants and Option 2a plants that have closed, or will close, GSI-191 utilizing an NRC-approved deterministic approach. Option 2b or Option 3 plants using a risk-informed approach are required to submit a plant-specific license amendment request to close GSI-191, and will not use the model application. The proposed TS changes and associated justification can be used by a risk-informed option plant and included as part of the license amendment request with the necessary technical justification.

For the risk-informed Option 2b approach taken for Callaway, Ameren Missouri has determined that the TS changes proposed per TSTF-567 are applicable to Callaway and may be proposed for Callaway as-is. That is, although Callaway's approach to resolving the GSI-191 and GL 2004-02 concerns involves deterministic and risk-informed aspects, and although both of those aspects contributed to establishing the "containment accident generated and transported debris" limits referred to in the new containment sump TS proposed for Callaway (described further below), the combined approach does not require any deviations from the TS changes prescribed by TSTF-567-A (Rev. 1).

Consistent with TSTF-567, the primary change to be made to the Callaway Technical Specifications is a new Technical Specification dedicated to the containment sumps. As previously noted, the sumps are a support system for the ECCS and CSS. In the current Technical Specifications, there is a Surveillance Requirement (SR) for the sumps within the ECCS Technical Specifications, but there is no dedicated Limiting Condition for Operation (LCO) for the sumps. Per the proposed TS changes, the sumps will have their own LCO and SRs under a new TS dedicated to the sumps. This change and some ancillary TS changes, including removal of the sump-related SRs under the ECCS Technical Specifications, as well as a change to the Administrative Controls TS that describes the Safety Function Determination Program, are described in greater detail in the following subsections.

2.4.1.1 Proposed Change to TS 3.5.2, "ECCS - Operating" Per the current Callaway Technical Specifications, TS 3.5.2, "ECCS - Operating,"

contains an SR, i.e., SR 3.5.2.8, which requires periodic verification by visual inspection that "each ECCS train containment sump suction inlet is not restricted by debris and the suction inlet strainers show no evidence of structural distress or abnormal corrosion."

With the creation of a new TS dedicated to the containment sumps (i.e., new TS 3.6.8, as further described in subsection 2.4.1.3 below), SR 3.5.2.8 is no longer needed in its current place since the SRs pertinent to the containment sumps will be specified under the new containment sump TS. As explained in Section 2.4.1.3, the new SR under TS 3.6.8 is broader in scope such that SR 3.5.2.8 is fully included in the new SR.

A corresponding change to the TS Bases for TS 3.5.2 will be made to reflect the removal of SR 3.5.2.8. In accordance with 10 CFR 50.36, changes to the TS Bases will be made in accordance with the Technical Specifications Bases Control Program

ULNRC-06526 Enclosure 2 Page 15 of 36 following approval of the requested amendment. The TS Bases changes are provided for information only and approval of the Bases is not requested.

2.4.1.2 Proposed Change to TS 3.5.3, "ECCS - Shutdown" TS 3.5.3, "ECCS - Shutdown," contains a Surveillance Requirements section in which one SR is identified, i.e., SR 3.5.3.1. However, this SR identifies a number of applicable surveillances by identifying those SRs in TS 3.5.2 that also apply to SR 3.5.3.1. Thus, by its list of referenced SRs, SR 3.5.3.1 is satisfied by the satisfactory performance of SRs 3.5.2.1, 3.5.2.3, 3.5.2.4, 3.5.2.7, and 3.5.2.8. Since the list of referenced SRs includes SR 3.5.2.8, and since SR 3.5.2.8 is being eliminated from its present location, the reference to SR 3.5.2.8 should be eliminated from SR 3.5.3.1. Further, there is no need for SR 3.5.3 to refer the new/relocated SR under the new containment sump since that SR is now associated with that TS and is no longer associated with the ECCS TS.

(That is, for the same reason that SR 3.5.2.8 is being eliminated from TS 3.5.2 as explained above, it should be eliminated from TS 3.5.3 as well.) Summarily, the change to TS 3.5.3 is simply the elimination of "SR 3.5.2.8" from list of SRs contained in SR 3.5.3.1. (A corresponding change to the Bases for TS 3.5.3 is also proposed, as described in Attachment 2-3 to this Enclosure.)

2.4.1.3 Proposed New TS 3.6.8, "Containment Sumps" The principal change proposed for the Callaway Technical Specifications is the addition of an entirely new TS dedicated to the containment sumps, i.e., TS 3.6.8, "Containment Sumps," complete with an LCO and Applicability, Conditions and Required Actions, including specified Completion Times, as well as an applicable SR. (A new Bases section dedicated to this TS is also proposed, as described in Attachment 2-3 to this Enclosure.)

Consistent with the redundant sumps designed for Callaway, LCO 3.6.8 (as proposed) states that "two containment sumps shall be OPERABLE." In regard to its Applicability, it is proposed that Modes 1, 2, 3 and 4 be identified as the applicable Modes. This is consistent the TS requirements for the ECCS and CSS to be Operable in Modes 1, 2, 3 and 4 and the requirement for the containment sumps to be Operable to support those functions/systems. (In Modes 1, 2, and 3, the plant is at normal operating pressure and temperature where generation of design-basis quantities of debris can reasonably be postulated. For Mode 4, there is less energy in the reactor coolant system (RCS) and reduced capability to generate the zones of influence associated with pipe breaks, but Mode 4 is included in the Applicability of TS 3.6.8 nonetheless.) In Modes 5 and 6, the probability and consequences of events are reduced due to the pressure and temperature limitations of these Modes. Thus, the containment sumps are not required to be Operable in Modes 5 or 6.

For the proposed Actions section of TS 3.6.8, it is necessary to address the condition of having one or more containment sumps inoperable. Consistent with how the Conditions are to be specified per TSTF-567, however, Condition A would address the condition of

ULNRC-06526 Enclosure 2 Page 16 of 36 having one or more sumps inoperable due solely to a debris issue, and Condition B would address the condition of having one or more sumps inoperable for reasons other than just a debris issue.

Accordingly, proposed Condition A specifies that if one or more containment sumps are "inoperable due to containment accident generated and transported debris exceeding the analyzed limits," then Required Actions A.1, A.2 and A.3 are to be entered.

Respectively, these Required Actions require station personnel to: (1) immediately initiate action to reduce containment accident generated and transported debris, (2) perform SR 3.4.13.1 once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , and (3) restore the containment sumps to OPERABLE status within 90 days. (Surveillance Requirement 3.4.13.1 requires verification that RCS operational leakage is within limits by performance of an RCS water inventory balance.)

The phrase "containment accident generated and transported debris" is taken directly from TSTF-567 (Rev. 1) and is further defined in the proposed Bases for TS 3.6.8. (See -3 to this Enclosure.) On the basis of that wording, Condition A is meant to apply only for the effects of LOCA generated and transported debris that exceeds the amount that has been analyzed. It does not apply to non-conforming or degraded conditions that are not associated with the LOCA-generated and transported debris, such as a strainer obstructed by a tarp (a condition that makes the strainer non-functional regardless of debris) or a strainer with large gaps that would pass debris fragments which would make non-applicable the risk evaluation based on fine fiber debris. The TS Bases address the analyzed debris amounts/limits applicable to Condition A and Required Action A.1.

The Bases for proposed new TS 3.6.8 (as further described in Attachment 2-3 to this Enclosure) give examples of actions that may be taken to satisfy Required Action A.1, i.e., to "reduce containment accident generated and transported debris." These include the following:

Removing the debris source from containment or preventing the debris from being transported to the containment sumps; Evaluating the debris source against the assumptions in the analysis; Deferring maintenance that would affect availability of the affected systems and other LOCA-mitigating equipment; Deferring maintenance that would affect availability of primary defense-in-depth systems, such as containment coolers; Briefing operators on LOCA debris management actions; or Applying an alternative method to establish new limits.

The 90-day Completion Time proposed for Required Action A.3, i.e., for restoring the affected containment sump to Operable status, is the same as what is specified in TSTF-567 (Rev. 1) for this Required Action, as approved by the NRC (and reflected in Consolidated Line Item Improvement Process for the TSTF). The 90-day Completion

ULNRC-06526 Enclosure 2 Page 17 of 36 Time is reasonable for emergent conditions that involve debris which could be generated and transported under LOCA conditions. The likelihood of an initiating event in the 90-day Completion Time is very small (~1/4 of the LOCA annual frequency).

There are margins in the debris generation and transport analyses and in the downstream and in-core effects analyses. 90 days is a reasonable time to restore the affected subsystems to Operable status by completing the identification and implementation of mitigating or compensatory actions, such as removing the debris, securing or containing the debris so that it is not transportable, performing additional analysis to demonstrate Operability, or to obtain regulatory relief (e.g., Enforcement Discretion and/or an Emergency or Exigent TS change). Additionally, the 90-day Completion Time provides clarity for the operators with regard to application of the TS for degraded or nonconforming conditions associated with the effects of LOCA debris.

Proposed Condition B addresses having one or more containment sumps inoperable for reasons other than Condition A. For this Condition, Required Action B.1 must be entered, which requires restoring the containment sump(s) to operable status within the specified Completion Time of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . As proposed, Required Action B.1 is modified by two notes. Note 1 directs entering the "applicable Conditions and Required Actions of LCO 3.5.2, ECCS - Operating, and LCO 3.5.3, ECCS - Shutdown, for emergency core cooling trains made inoperable by the containment sump(s)." Note 2 directs entering the "applicable Conditions and Required Actions of LCO 3.6.6, Containment Spray and Cooling Systems, for containment spray trains made inoperable by the containment sump(s)."

Proposed Condition C is to be entered if the Required Actions and associated Completion Times for Condition A or Condition B are not met. Per Required Actions C.1 and C.2, as proposed, the plant must be in Mode 3 in 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and Mode 5 in 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months , respectively.

One SR is proposed for new TS 3.6.8. Specifically, new SR 3.6.8.1 would require verifying, "by visual inspection, [that] the containment sumps do not show structural damage, abnormal degradation, or debris blockage" (based on the wording prescribed in TSTF-567). The frequency for performance of this SR would be as specified per Callaway's Surveillance Frequency Control Program, which was established per License Amendment 202 of the Callaway Operating License and is described in Administrative Control TS 5.5.18.

It was noted previously that SR 3.5.2.8 is being eliminated in light of the fact that new SR 3.6.8.1 provides the appropriate surveillance for the containment sumps and encompasses the scope of SR 3.5.2.8. This can be discerned by comparing the wording of the former, "Verify, by visual inspection, each ECCS train containment sump suction inlet is not restricted by debris and the suction inlet strainers show no evidence of structural distress or abnormal corrosion," which focuses on the suction inlets and inlet strainers, to the latter, "Verify, by visual inspection, the containment sumps do not

ULNRC-06526 Enclosure 2 Page 18 of 36 show structural damage, abnormal corrosion, or debris blockage," which focuses on the containment sumps overall.

In regard to this change, therefore, the NRC concluded in their evaluation of TSTF-567, Rev. 1, that "the proposed change is acceptable since the existing requirements are either unchanged or expanded and continue to ensure the containment sump is unrestricted (i.e., unobstructed) and stays in proper operating condition. The proposed SR meets the requirements of 10 CFR 50.36(c)(3) because it provides an SR to assure the necessary quality of systems and components are maintained, that facility operation will be within safety limits, and that the LCOs will be met."

2.4.1.4 Proposed Change to TS 5.5.15, "Safety Function Determination Program (SFDP)"

Ameren Missouri proposes to add the following sentence at the end of TS 5.5.15, "Safety Function Determination Program (SFDP)," to clarify the SFDP, consistent with TSTF-567, Revision 1, and the associated NRC SE:

When a loss of safety function is caused by the inoperability of a single TS support system, the appropriate Conditions and Required Actions to enter are those of the support system.

The SFDP described in TS 5.5.15 is necessary for the proper implementation of LCO 3.0.6, in regard to addressing certain inoperable equipment conditions when the LCOs of systems involving support-supported system relationships are involved. Specifically, per LCO 3.0.6, when a supported system LCO is not met solely due to a support system LCO not being met, the Conditions and Required Actions associated with the supported system are not required to be entered. Only the support system LCO ACTIONS are required to be entered. This provision is an exception to LCO 3.0.2 for the supported system. In such cases, LCO 3.0.6 requires an evaluation to be performed in accordance with TS 5.5.15 (the SFDP) to determine whether a loss of safety function exists. If a loss of safety function is determined to exist by the SFDP, the appropriate Conditions and Required Actions of the LCO in which the loss of safety function exists are required to be entered.

Thus, for example, if one train of a support system having redundant trains is declared inoperable, cross-train checks are performed to identify whether a loss of safety function exists for the associated, supported system(s) that might also have redundant trains.

The cross train check verifies that the supported systems of the redundant support system are Operable, thereby ensuring safety function is retained. If this evaluation determines that a loss of safety function exists, the appropriate Conditions and Required Actions of the LCO in which the loss of safety function exists are required to be entered.

Callaways containment sump design includes intended redundancy such that there are two containment sumps. However, in light of the concerns raised by GSI-191 and GL

ULNRC-06526 Enclosure 2 Page 19 of 36 2004-02, the two sumps must be considered part of a single support system because containment accident generated and transported debris that could render one sump inoperable could render both sumps inoperable. Declaring one or both sumps inoperable per Condition A of proposed TS 3.6.8 (for one or more containment sumps "inoperable due to containment accident generated and transported debris exceeding the analyzed limits"), and then applying the provisions of the SFDP as currently described in TS 5.5.15, could (unnecessarily) result in declaring both ECCS trains and both CSS trains inoperable such that the Conditions and Required Actions for both trains inoperable would have to be entered under TS 3.5.2 and TS 3.6.6.

By considering the containment sumps to be a single TS support system, the words to be added to TS 5.5.15 would make it clear that when a loss of safety function is caused by the inoperability of a single TS support system, the appropriate Conditions and Required Actions to be entered are those of the support system (in lieu of the supported system).

As noted in the NRC's Safety Evaluation (SE) for approval of TSTF-567 (Rev. 1), the proposed change to the TS-described SFDP is applicable to plants that have more than one containment sump. Further, it is stated in the NRC's SE that the proposed addition to TS 5.5.15 clarifies the intent of the allowance (not to enter the Conditions and Required Actions of the supported systems) provided by LCO 3.0.6 and the SFDP for single-train support systems. It is noted that the proposed change is acceptable since the actions for the support system LCO adequately address the inoperability of that system. Therefore, as required by 10 CFR 50.36(c)(5), the TS-described SFDP (as revised by the proposed change) continues to provide adequate administrative controls to assure safe operation.

2.4.2 Proposed FSAR Changes Upon approval of the licensing basis changes, Ameren Missouri will make the following changes to the Callaway FSAR:

Add Appendix 6.3A, "Risk-Informed Approach to Potential Impact of Debris Blockage on Emergency Recirculation During Design-Basis Accidents." This appendix describes the evaluations performed using a risk-informed approach to address GSl-191 concerns including the effects on long-term cooling due to debris accumulation on containment sump strainers for ECCS and CSS in recirculation mode, as well as core flow blockage due to in-vessel effects, following LOCAs This section summarizes the methodology change(s) for which NRC approval is sought.

o Callaway requests NRC approval of Section 6.3A.2, "Change Control and Reporting," of Appendix 6.3A since it includes criteria for identifying changes that would require prior NRC approval.

Make conforming changes to FSAR Table 1.6-2

ULNRC-06526 Enclosure 2 Page 20 of 36 Make conforming changes to FSAR Chapter 3 descriptions of evaluations against GDC 35, GDC 38 and GDC 41 Make conforming changes to FSAR Chapter 6 Make conforming changes to FSAR Chapter 15 The proposed FSAR changes are provided as mark-ups in Attachment 2.5 to this Enclosure.

3. TECHNICAL EVALUATION

3.1 Background

GSI-191 concerns the possibility that debris generated during a LOCA could clog the containment sump strainers in pressurized-water reactors (PWRs) and result in loss of NPSH for the ECCS and CSS pumps, thus impeding the flow of water from the sumps.

GL 2004-02 requested licensees to address GSI-191 issues, with a focus on demonstrating compliance with the ECCS acceptance criteria in 10 CFR 50.46.

GL 2004-02 also requested licensees to perform new, more realistic analyses using an NRC-approved methodology and to confirm the functionality of the ECCS and CSS during design-basis accidents that require containment sump recirculation.

The Callaway risk-informed approach maintains the defense-in-depth measures in place to mitigate the residual risk of strainer or in-vessel issues to address GL 2004-02.

These measures include those implemented in response to NRC Bulletin 2003-01 [6]

and GL 2004-02 to address the potential for sump strainer clogging and other concerns associated with GSI-191. Additional measures, such as operating procedures and instrumentation to monitor core level and temperature, and actions taken by operators if core blockage is indicated, have been implemented. These actions are not the subject of this license amendment request. Detailed discussion regarding defense-in-depth is provided in Enclosure 3, Attachment 3-4 to this letter. These measures are part of the defense-in-depth for Callaway and remain in place.

The Commission issued Staff Requirements Memorandum (SRM)-SECY-10-0113, "Closure Options for Generic Safety Issue 191, Assessment of Debris Accumulation on Pressurized Water Reactor Sump Performance," [7] directing the staff to consider alternative options for resolving GSI-191 and GL 2004-02 that are innovative and creative, as well as risk-informed and safety conscious. Subsequently, through interactions with the staff, South Texas Project Nuclear Operating Company (STPNOC) developed a risk-informed approach to address GSI-191 based on the guidance in RG 1.174, which serves as a pilot for other licensees to adopt. Ameren Missouri has chosen to use this approach as described in the following two sections.

ULNRC-06526 Enclosure 2 Page 21 of 36 Plant-Specific Testing Callaway conducted successful plant-specific testing in July 2016 using approved prototypical debris, conservative chemical effects, prototypical simulation of strainer approach flow conditions, and a Callaway strainer module. This plant-specific test is described in more detail in Enclosure 3 and forms the basis for the deterministic scope of the proposed methodology change.

Use of a Risk-Informed Approach to Address GL 2004-02 The risk associated with GL 2004-02 issues has been quantified as described in and is very small as defined by Region III in RG 1.174. The proposed FSAR Appendix 6.3A describes the risk-informed approach used to confirm that there is high probability that the ECCS and CSS will perform their design basis functions following a LOCA when considering the impacts and effects described by GL 2004-02.

Therefore, no further physical modifications to Callaway are proposed as part of this license amendment request to implement the risk-informed approach. -1 to this enclosure provides the Licensee Commitment to implement the proposed amendment following approval and to revise affected sections of the plant TS identified in Attachment 2-2 and of the FSAR identified in Attachment 2-5 to this enclosure. Upon approval of the proposed amendment, applicable FSAR safety system and design bases descriptions that take credit for the evaluation described above will be revised. In addition, conforming changes to the TS Bases are provided in Attachment 2-3 to this enclosure for information only, to be implemented following NRC approval of the LAR.

System redundancy, independence, and diversity features are not changed for those safety systems credited in the accident analyses. No new programmatic compensatory activities or reliance on manual operator actions are required to implement this change.

3.2 Evaluation The proposed change meets the current regulations unless it is explicitly related to a requested exemption.

3.2.1 Engineering Analysis Overview The design and licensing basis descriptions of accidents requiring ECCS and CSS operation, including analysis methods, assumptions, and results provided in FSAR Chapters 6 and 15 remain unchanged. This is based on the functionality of the ECCS and CSS during design-basis accidents being confirmed by demonstrating that the calculated risk associated with GL 2004-02 for Callaway is "Very Small" and less than the Region III acceptance guidelines defined by RG 1.174.

ULNRC-06526 Enclosure 2 Page 22 of 36 In addition, as described in RG 1.174, Section 2.5.2, "Comparisons with Acceptance Guidelines," "if there is an indication that the CDF or LERF could considerably exceed 10-4 and 10-5, respectively, in order for the change to be considered, the licensee may need to show why steps should not be taken to reduce CDF or LERF." As shown in the following Table, the current Callaway PRA model of record baseline total aggregate risk CDF is 7.94E-05, and the corresponding LERF is 3.18E-06.

Baseline Baseline Model CDF LERF Internal Events (Excluding Internal Flooding) PRA 3.46E-06 5.93E-08 Internal Flooding PRA 7.07E-06 1.64E-08 Fire PRA 7.40E-06 2.79E-08 Seismic PRA 5.59E-05 2.82E-06 High Winds PRA 5.60E-06 2.6E-07 Total Aggregate Risk 7.94E-05(1) 3.18E-06(1)

(1)

Note: The PRA model risk metrics provided herein have not been adjusted to account for an open modeling issue recently identified in the Callaway PRA. The open issue is currently being addressed in accordance with the PRA model maintenance process. Preliminary evaluation of the impact indicates an increase to the aggregate baseline CDF value, but the change is not expected to be significant or to challenge the 1E-4 aggregate baseline risk threshold used to establish the "very low change in risk" determined for application of the RoverD methodology (per the Reg. Guide 1.174 criteria).

From the Section 7, "Baseline Results," in Enclosure 3, Attachment 3-3, it can be seen that the calculated RoverD mean CDF is 5.37E-07, and the corresponding mean LERF is 5.37E-08. Applying the changes in CDF and LERF associated with implementation of RoverD to the baseline CDF and LERF, respectively, the resulting final values are 7.99E-5 for CDF and 3.23E-6 for LERF, which are well within the RG 1.174, Section 2.5.2 acceptance guidelines.

The performance evaluations for accidents requiring ECCS operation described in Chapters 6 and 15, based on Callaway 10 CFR 50, Appendix K large-break LOCA analysis, demonstrate that for partial breaks and complete breaks up to and including the DEGB of a reactor coolant pipe, the ECCS will limit the clad temperature to below the limit specified in 10 CFR 50.46, and assure that the core will remain in place and substantially intact with its essential heat transfer geometry preserved.

ULNRC-06526 Enclosure 2 Page 23 of 36 The LAR is requested for the scope of breaks that can generate fiber debris on the containment sump strainers that exceeds the amount of fiber debris bounded by the plant-specific testing. In Attachment 3-2 and 3-3, Callaway determined that only large breaks on some RCS and RHR pipes were in this scope and has identified 60 weld break locations (listed in Attachment 3-3). Ameren Missouri is requesting an amendment to the license for this scope of breaks to allow evaluation of the debris effects using a risk-informed methodology because there is no practical deterministic methodology currently available. The amendment is requested to apply to the evaluation methodology and not to the specific set of break locations.

The WCAP-17788 methodology [8] that was used to evaluate the down-stream effects for the deterministic scope of breaks to assure long-term core cooling does not replace the ECCS evaluation methodology described in Callaway FSAR Chapter 15.6. The current Chapter 15.6 LOCA thermal-hydraulic analysis applies only through the LOCA re-flood phase and is not used for the assessment of long-term cooling required by the risk-informed assessment of debris effects.

3.2.2 Evaluation of Defense-In-Depth and Safety Margin Defense-in-Depth Analysis The proposed change is consistent with the defense-in-depth (DID) philosophy in that the following aspects of the facility design and operation are unaffected:

Functional requirements and the design configuration of systems Existing plant barriers to the release of fission products Design provisions for redundancy, diversity, and independence Plant's response to transients or other initiating events Preventive and mitigative capabilities of plant design features The proposed amendment does not involve a change in any functional requirements or the configuration of plant SSCs. , Attachment 3-4 provides a more detailed description of the defense-in-depth measures that address potential sump blockage and in-core effects, including the means available to operators for detecting and mitigating inadequate recirculation flow and inadequate core cooling flow. The proposed change does not involve a change in any functional requirements, the configuration, or method of performing functions of plant SSCs. The effects from a full spectrum of LOCAs, including a range of partial breaks and DEGBs for all piping sizes up to and including the largest pipe in the reactor coolant system, are analyzed. Appropriate redundancy and consideration of loss of

ULNRC-06526 Enclosure 2 Page 24 of 36 offsite power and worst case single failure are retained. This approach ensures that DID is maintained.

Safety Margin Analysis , Attachment 3-4 provides a more detailed discussion on how sufficient safety margins associated with the design are maintained by the proposed change.

Approval of the proposed change would add the results of a risk-informed evaluation to the FSAR that concludes that there is high probability that the containment sumps will perform their design basis functions in support of ECCS and CSS recirculation modes following a LOCA when considering the impacts and effects of debris on sump strainers.

Core flow blockage due to in-vessel effects is addressed via an analysis using WCAP-17788 methodology. The proposed change does not result in any changes to the safety analyses demonstrating safety margin for the barriers to the release of radioactivity as described in the FSAR, and does not involve a change in the functional requirements, configuration, or method of performing functions of plant SSCs.

3.2.3 Description of the PRA The Callaway PRA is an all-hazards, at-power Level I with large early release frequency (LERF) model. The Callaway PRA is representative of the as-built, as-operated plant and includes the internal events that are within the focus of the GL 2004-02 concerns related to LOCA. The Callaway PRA is reviewed for consistency with the as-built, as-operated plant at a nominal frequency of every two refueling cycles.

The PRA was not changed to address GL 2004-02 concerns. Instead, a detailed engineering analysis was performed in an uncertainty quantification framework that evaluates the required failure modes of ECCS and core cooling (in-vessel effects).

Significant detail is included in the engineering analysis, including physical models and mechanisms known to lead to failure, and the analyses account for experimental evidence used to support particular areas of concern.

Callaway's PRA is compliant with RG 1.200, Revision 2 for internal events and is therefore acceptable to support the assessment of the risk of internal events associated with GL 2004-02. , Attachment 3-3 to this letter provides a more detailed description of how the Callaway PRA results are used in conjunction with the detailed engineering analysis to estimate LERF and address secondary line break initiating events.

3.2.4 Implementation and Monitoring Program Design modifications addressing GL 2004-02 concerns, including installation of new sump strainers and replacement of some problematic insulation, have been previously implemented using the Callaway design change process.

ULNRC-06526 Enclosure 2 Page 25 of 36 Callaway has implemented procedures and programs for monitoring, controlling and assessing changes to the plant that have a potential impact on plant performance related to GL 2004-02 concerns. These provide the capability to monitor the performance of the sump strainers and the ability to assess impacts to the inputs and assumptions used in the PRA and the associated engineering analysis that support the proposed change. Programmatic requirements ensure that the potential for debris loading on the sumps does not materially increase. These include:

Programs and procedures have been implemented to evaluate and control sources of debris in containment:

o Technical Requirements implemented by Callaway procedures require visual inspections of all accessible areas of the containment to check for loose debris, and each containment sump to check for debris.

o The Callaway Engineering Change Control Procedure includes provisions for managing debris sources such as insulation, qualified coatings, addition of aluminum or zinc. The procedure has been augmented as applicable to require changes that involve any work or activity inside the containment be evaluated for the potential to affect the following:

Reactor coolant pressure boundary integrity Accident or post-accident equipment inside containment Quantity of metal inside containment Quantity or type of coatings inside containment Thermal insulation changed or added Addition or deletion of cable o The 10 CFR 50.59 change process will be entered in accordance with Callaway procedures for all Design Changes. This process ensures that new insulation material that may differ from the initial design is evaluated for GL 2004-02 concerns.

o Programs to ensure that Service Level 1 protective coatings used inside containment are procured, applied, and maintained in compliance with applicable regulatory requirements. Additional details are discussed in the Callaway coatings program developed in response to Generic Letter 98-04

[9].

o Procedures have been implemented to govern the use of signs and labels inside containment.

ULNRC-06526 Enclosure 2 Page 26 of 36 As a necessary and required support function for ECCS and CSS, the sump strainers are within the Callaway 10 CFR 50.65 Maintenance Rule program:

o As part of the Callaway Corrective Action Program, condition reports written due to adverse conditions identified during the containment inspections or containment sumps and strainers surveillances are reviewed for impact on Maintenance Rule scoped systems, as appropriate.

o The Callaway Maintenance Rule program includes performance monitoring of functions associated with ECCS and CSS, including sump recirculation. The inclusion of the ECCS and CSS into the Maintenance Rule program and the assessment of acceptable system performance provide continued assurance of the availability for performance of the required functions.

PRA Updates: For the purpose of monitoring future facility changes or other conditions that may impact the PRA results associated with GL 2004-02, appropriate changes to the as-built, as-operated plant are reflected in updates to the Callaway at-power PRA reference model. The Callaway PRA Program is a living program and, as such, is subject to periodic review and updates. These PRA model periodic updates are performed in accordance with Callaway procedures. The effect of changes incorporated into the at-power PRA model of record are assessed to ensure the results of the analysis used to close GL 2004-02 remain within the aggregate baseline acceptance criteria in RG 1.174.

Licensed Operator Training: Licensed Operators are trained on indications of and actions in response to sump blockage issues related to GL 2004-02, and performance is evaluated during training scenarios designed to simulate plant responses.

Operator actions required to respond to sump clogging are currently described in ECA-1.3, and operators are trained on implementing defense-in-depth actions (e.g., alternate flow paths) as a part of the Licensed Operator program.

Indications of sump blockage are included as part of the Licensed Operator training administered for Emergency Operating Procedure (EOP) performance of switchover activities in addition to general familiarization with the indications of loss of pump suction. Licensed Operator Training includes the monitoring of operating ECCS and CSS pumps during the evolution for transfer to cold-leg recirculation. Operator training also includes actions required on a total loss of sump recirculation capability.

Quality Assurance (QA): The Callaway Operating Quality Assurance Program (OQAP) is implemented and controlled in accordance with policies, manuals, procedures and the Operating Quality Assurance Manual (OQAM). This program is applicable to safety related structures, systems and components to an extent consistent with their importance to safety. The OQAP complies with the

ULNRC-06526 Enclosure 2 Page 27 of 36 requirements of 10 CFR 50, Appendix B and other program commitments [OQAP Introduction].

The QA Program is implemented with documented instructions, procedures, and drawings which include appropriate quantitative and qualitative acceptance criteria for determining that prescribed activities have been satisfactorily accomplished [OQAM 5.1]. Procedures control the sequence of required inspections, tests, and other operations when important to quality [OQAM 14.4].

To change these controls, the individual procedure must be changed and a similar level of review and approval given to the original procedure is required

[OQAM 5.2.1, 5.2.2, 6.3]. Such instructions, procedures, and drawings are reviewed and approved for compliance with requirements appropriate to their safety significance [OQAM 5.1, 5.2 - 5.9].

QA program controls are applied to safety-related SSCs to provide a high degree of confidence that they perform safely and activities are performed as expected

[OQAM 2.1, 2.4, 2.5, and 2.6]. The rigorous controls imposed by the QA program provide adequate quality control elements to ensure system component reliability for the required functions.

Callaway has adopted other programs that help provide early detection and mitigation of leakage in other applications. The proposed change does not involve any changes to ASME Section XI inspection programs or mitigation strategies that have been shown effective in early detection and mitigation of weld and material degradation in Class 1 piping.

3.2.5 Technical Evaluation Conclusion The technical evaluation results demonstrate that the calculated risk associated with GL 2004-02 concerns for Callaway is very small and lie in Region III defined by RG 1.174.

Acceptable containment sump design in support of ECCS and CSS during design-basis accidents is confirmed by demonstrating with high confidence that sufficient safety margin and defense-in-depth are maintained. Additional details can be found in of this submittal.

4. REGULATORY EVALUATION 4.1 Applicable Regulatory Requirements/Criteria The following regulations apply to the proposed amendment. Approval of the proposed amendment is contingent upon approval of the requests for exemptions from these regulations as provided and justified in Enclosure 1.

Regulatory Requirement 10 CFR 50.46(a)(1)

The regulation 10 CFR 50.46(a)(1), states:

ULNRC-06526 Enclosure 2 Page 28 of 36 Each boiling or pressurized light-water nuclear power reactor fueled with uranium oxide pellets within cylindrical zircaloy or ZIRLO cladding must be provided with an emergency core cooling system (ECCS) that must be designed so that its calculated cooling performance following postulated loss-of-coolant accidents conforms to the criteria set forth in paragraph (b) of this section. ECCS cooling performance must be calculated in accordance with an acceptable evaluation model and must be calculated for a number of postulated loss-of-coolant accidents of different sizes, locations, and other properties sufficient to provide assurance that the most severe postulated loss-of-coolant accidents are calculated. Except as provided in paragraph (a)(1)(ii) of this section, the evaluation model must include sufficient supporting justification to show that the analytical technique realistically describes the behavior of the reactor system during a loss-of-coolant accident. Comparisons to applicable experimental data must be made and uncertainties in the analysis method and inputs must be identified and assessed so that the uncertainty in the calculated results can be estimated. This uncertainty must be accounted for, so that, when the calculated ECCS cooling performance is compared to the criteria set forth in paragraph (b) of this section, there is a high level of probability that the criteria would not be exceeded. Appendix K, Part II Required Documentation, sets forth the documentation requirements for each evaluation model. This section does not apply to a nuclear power reactor facility for which the certifications required under §50.82(a)(1) have been submitted.

(ii) Alternatively, an ECCS evaluation model may be developed in conformance with the required and acceptable features of appendix K ECCS Evaluation Models.

The regulatory requirements of 10 CFR 50.46(a)(1) remain applicable to the model of record that meets the required features of 10 CFR 50, Appendix K. This evaluation model remains the licensing basis for demonstrating that the ECCS calculated cooling performance following postulated LOCAs does not exceed the acceptance criteria.

The proposed changes do not result in any physical changes to the facility or changes to the operation of the plant, and does not change any of the programmatic requirements. Based on demonstrating acceptable LOCA debris mitigation and containment sump and ECCS design for amending the current licensing basis for 10 CFR 50.46(a)(1) as described above, compliance with other regulatory requirements that rely on acceptable design for these systems and components continue to be met in the current licensing basis.

Regulatory Requirement 10 CFR 50 Appendix A

ULNRC-06526 Enclosure 2 Page 29 of 36 GDC 35, "Emergency core cooling," states that a system to provide abundant emergency core cooling shall be provided. The system safety function shall be to transfer heat from the reactor core following any loss of reactor coolant at a rate such that (1) fuel and clad damage that could interfere with continued effective core cooling is prevented and (2) clad metal-water reaction is limited to negligible amounts.

Suitable redundancy in components and features, and suitable interconnections, leak detection, isolation, and containment capabilities shall be provided to assure that for onsite electric power system operation (assuming offsite power is not available) and for offsite electric power system operation (assuming onsite power is not available) the system safety function can be accomplished, assuming a single failure.

GDC 38, "Containment heat removal," states that a system to remove heat from the reactor containment shall be provided. The system safety function shall be to reduce rapidly, consistent with the functioning of other associated systems, the containment pressure and temperature following any LOCA and maintain them at acceptably low levels.

Suitable redundancy in components and features, and suitable interconnections, leak detection, isolation, and containment capabilities shall be provided to assure that for onsite electric power system operation (assuming offsite power is not available) and for offsite electric power system operation (assuming onsite power is not available) the system safety function can be accomplished, assuming a single failure.

GDC 41, "Containment atmosphere cleanup," states that systems to control fission products, hydrogen, oxygen, and other substances which may be released into the reactor containment shall be provided as necessary to reduce, consistent with the functioning of other associated systems, the concentration and quality of fission products released to the environment following postulated accidents, and to control the concentration of hydrogen or oxygen and other substances in the containment atmosphere following postulated accidents to assure that containment integrity is maintained.

The proposed changes do not affect compliance with these regulations or guidance and will ensure that the lowest functional capabilities or performance levels of equipment required for safe operation are met.

NRC Regulatory Guide 1.174, "An Approach for using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis," provides the NRC staff's recommendations for using risk information in support of licensee-initiated Licensing Basis changes to a nuclear power plant that require NRC review and approval. This regulatory guide describes an acceptable approach for assessing the

ULNRC-06526 Enclosure 2 Page 30 of 36 nature and impact of proposed Licensing Basis changes by considering engineering issues and applying risk insights.

In implementing risk-informed decision making, Licensing Basis changes are expected to meet a set of key principles. These principles include the following:

1. The proposed change meets the current regulations unless it is explicitly related to a requested exemption (i.e., a specific exemption under 10 CFR 50.12, "Specific Exemptions').

The exemption request in Enclosure 1 to this letter implements this principle.

2. The proposed change is consistent with a defense-in-depth philosophy.

The proposed change is consistent with the DID philosophy in that the following aspects of the facility design and operation are unaffected:

Functional requirements and the design configuration of systems Existing plant barriers to the release of fission products Design provisions for redundancy, diversity, and independence Plant's response to transients or other initiating events Preventive and mitigative capabilities of plant design features The Callaway risk-informed approach analyzes a full spectrum of LOCAs, including a range of partial breaks and DEGBs for all piping sizes up to and including the largest pipe in the RCS. By requiring that mitigative capability be maintained in a realistic and risk-informed evaluation of GL 2004-02 for a full spectrum of LOCAs, the approach ensures that defense-in-depth is maintained.

3. The proposed change maintains sufficient safety margins.

The proposed change does not involve a change in any functional requirements or the configuration of plant SSCs. The safety analyses in the FSAR are unchanged.

Therefore, sufficient safety margins associated with the design will be maintained by the proposed change.

4. When proposed changes result in an increase in CDF or risk, the increases should be small and consistent with the intent of the Commission's Safety Goal Policy Statement.

ULNRC-06526 Enclosure 2 Page 31 of 36 The proposed change is defined as the risk associated with GL 2004-02 concerns.

Using engineering analysis and the PRA this risk has been calculated and shown to be less than the threshold for Region III, "Very Small Changes," and is therefore consistent with the Commission's Safety Goal Policy Statement.

5. The impact of the proposed change should be monitored using performance measurement strategies.

Section 3.2.4 of this Enclosure describes the programmatic requirements that ensure the potential for debris loading on the sump does not materially increase.

As noted in section 3.2.4, the effect of changes incorporated into the at-power PRA model of record are periodically assessed to ensure the results of the analysis used to close GL 2004-02 remain within the aggregate baseline acceptance criteria in RG 1.174.

NRC Regulatory Guide 1.200, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities," describes one acceptable approach for determining whether the quality of the PRA, in total or the parts that are used to support an application, is sufficient to provide confidence in the results, such that the PRA can be used in regulatory decision-making for light-water reactors.

The Callaway PRA model used for the risk-informed approach for addressing GL 2004-02 concerns is in compliance with Revision 2 of RG 1.200.

The proposed changes do not affect compliance with these regulatory guides and will ensure that the lowest functional capabilities or performance levels of equipment required for safe operation are met.

4.2 Precedent The NRC plans to use South Texas Plant (STP) Units 1 and 2 as a pilot for other licensees choosing a risk-informed approach for closure of GSI-191 (SECY-12-0093

[10] and ACRS letter to NRC "Safety Evaluation of License Amendment Request by South Texas Project Nuclear Operating Company to Adopt a Risk-informed Resolution of Generic Safety Issue-191" [11]). The STP-piloted risk-informed approach resulted in substantial benefit to both the NRC and industry in support of the development and implementation of risk-informed resolution of GSI-191.

The proposed amendment and accompanying exemption requests provided an approach for other licensees to revise their Licensing Basis in order to close GSI-191 and GL 2004-02.

By Reference 11, the ACRS approved the NRCs safety evaluation of the STPNOCs RoverD analysis results showing that the risks, CDF and large early release frequency (LERF) associated with GSI-191 concerns are less than the threshold for Region III,

ULNRC-06526 Enclosure 2 Page 32 of 36 "Very Small Changes," of RG 1.174, and the NRC accepted the associated license amendment request.

It is Ameren Missouris intent to follow that pilot program with the following exceptions, which deviate from STPNOCs submittal:

References to GSI-191 were replaced with GL 2004-02 in most places due to the administrative closing of GSI-191 in July 2019.

Ex-vessel downstream effects are not addressed in the Callaway risk-informed evaluation of GL 2004-02. Instead Callaway performed an analysis based on WCAP-16406 [12].

In-vessel downstream effects are not addressed in the Callaway risk-informed evaluation of GL 2004-02. Instead, Callaway performed an analysis based on WCAP-17788.

No extensive thermal-hydraulic analysis of the cores response to postulated blockage was conducted.

Completion times for proposed new TS 3.6.8 were based on TSTF-567, Revision 1, which applies a 90-day window to bring the ECCS strainer back to operable condition.

4.3 No Significant Hazards Consideration Determination Callaway has evaluated whether or not 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 change involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No.

The proposed changes involve a methodology change for assessment of debris effects that adds the results of a risk-informed evaluation to the Callaway licensing basis. Included are proposed changes to the Callaway Technical Specifications, for which the Emergency Core Cooling System (ECCS) Technical Specifications would be revised and a new Technical Specification (TS) dedicated to the

ULNRC-06526 Enclosure 2 Page 33 of 36 containment sumps would be established. The new TS would include Conditions and Required Actions for addressing containment conditions involving potential LOCA debris-related effects. The Required Action related to potential LOCA-generated debris effects includes a Completion Time that is longer than the Completion Time currently specified under the ECCS and Containment Spray System (CSS) Technical Specifications, but is justified on the basis of very low risk. Associated administrative TS changes are proposed as well.

The methodology change (to be reflected in the FSAR) concludes that the ECCS and Containment Spray System (CSS) will have sufficient defense-in-depth and safety margin and that there is high confidence that these systems will perform their design basis functions following a loss-of-coolant accident (LOCA) when considering the impacts and effects of debris accumulation on containment sump strainers in recirculation mode, as well as core flow blockage due to in-vessel effects, following loss-of-coolant accidents. The methodology change also supports the proposed TS changes.

There is no significant increase in the probability of an accident previously evaluated. The proposed changes address mitigation of loss-of-coolant accidents and have no effect on the probability of the occurrence of a LOCA. The proposed methodology and TS changes do not implement any physical changes to the facility or any SSCs and do not implement any changes in plant operation that could lead to a different kind of accident.

The proposed changes do not involve a significant increase in the consequences of an accident previously evaluated. The methodology change confirms that required SSCs supported by the containment sumps will perform their safety functions with a high probability, as required, and does not alter or prevent the ability of SSCs to perform their intended function to mitigate the consequences of an accident previously evaluated within the acceptance limits. The safety analysis acceptance criteria in the FSAR continue to be met for the proposed methodology change. The evaluation of the changes determined that containment integrity will be maintained. The dose consequences were considered in the assessment, and quantitative evaluation of the effects on dose using input from the risk-informed approach shows the increase in dose consequences is very small.

Therefore, the proposed change does not involve a significant increase in the probability or consequences of any the accident previously evaluated in the FSAR.

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 changes are a methodology change for assessment of debris effects from LOCAs that are already evaluated in the Callaway FSAR, establishment of a

ULNRC-06526 Enclosure 2 Page 34 of 36 new TS dedicated to the containment sumps that addresses potential LOCA-generated debris effects on the ECCS and CSS, and associated administrative changes to the TS. No new or different kind of accident is being evaluated. None of the changes install or remove any plant equipment, or alter the design, physical configuration, or mode of operation of any plant structure, system or component.

The proposed changes do not introduce any new failure mechanisms or malfunctions that can initiate an accident. The proposed changes do not introduce failure modes, accident initiators, or equipment malfunctions that would cause a new or different kind of accident.

Therefore, the proposed changes do not create the possibility for 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 changes are a methodology change for assessment of debris effects from LOCAs that are already evaluated in the Callaway FSAR, establishment of a new TS dedicated to the containment sumps that addresses potential LOCA-generated debris effects on the ECCS and CSS, and associated administrative changes to the TS. The effects from a full spectrum of LOCAs, including a range of partial breaks and DEGBs for all piping sizes up to and including the largest pipe in the reactor coolant system, are analyzed. Appropriate redundancy and consideration of loss of offsite power and worst case single failure are retained, such that defense-in-depth is maintained.

Application of the risk-informed methodology showed that the increase in risk from the contribution of debris effects is very small as defined by RG 1.174 and that there is adequate defense-in-depth and safety margin. Consequently, Callaway determined that the risk-informed method demonstrates the containment sumps will continue to support the ability of safety-related components to perform their design functions when the effects of debris are considered. The proposed change does not alter the manner in which safety limits are determined or acceptance criteria associated with a safety limit. The proposed change does not implement any changes to plant operation and does not significantly affect SSCs that respond to safely shutdown the plant and to maintain the plant in a safe shutdown condition. The proposed change does not significantly affect the existing safety margins in the barriers for the release of radioactivity. There are no changes to any of the safety analyses in the FSAR.

Therefore, the proposed change does not involve a significant reduction in a margin of safety.

ULNRC-06526 Enclosure 2 Page 35 of 36 Based on the above, Callaway concludes that the proposed changes do 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 Conclusions 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 Commission's regulations contingent upon approval of the exemption requested in Enclosure 1 to this letter, 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.

5. 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 a significant increase in the amounts of any effluents 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. REFERENCES

1. Generic Safety Issue 191, "Assessment of Debris Accumulation on Pressurized Water Reactor Sump Performance."

2. Nuclear Regulatory Commission Generic Letter 2004-02: Potential Impact of Debris Blockage on Emergency Recirculation during Design Basis Accidents at Pressurized-Water Reactors, September 13, 2004 (ML042360586).

3. Regulatory Guide 1.174, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis,"

Revision 2. U.S. Nuclear Regulatory Commission, May 2011 (ML100910006).

4. Letter from NRC to TSTF, Final Safety Evaluations of Technical Specifications Task Force Traveler TSTF-567, Revision 1, Add Containment Sump TS to Address GSI-191 Issues (EPID: L-2017-PMP-0005, July 3, 2018 (ML18109A077).

5. NUREG-1829 Vol. 1, Estimating Loss-of-Coolant Accident (LOCA) Frequencies Through the Elicitation Process, April 2008 (ML082250436).

ULNRC-06526 Enclosure 2 Page 36 of 36

6. NRC Bulletin 2003-01, Potential Impact of Debris Blockage on Emergency Sump Recirculation at Pressurized-Water Reactors. Nuclear Regulatory Commission, June 9, 2003.

7. SECY-10-0113, Closure Options for Generic Safety Issue - 191, Assessment of Debris Accumulation on Pressurized Water Reactor Sump Performance, August 26, 2010.

8. WCAP-17788-NP, Comprehensive Analysis and Test Program for GSI-191 Closure (PA-SEE-1090), Volume 1, Revision 0. Westinghouse, July 2015 (ML15210A669).

9. Generic Letter No. 98-04, Potential for Degradation of the Emergency Core Cooling System and the Containment Spray System after a Loss-of-Coolant Accident Because of Construction and Protective Coating Deficiencies and Foreign Material in Containment. Nuclear Regulatory Commission, July 14, 1998.

10. SECY-12-0093, Closure Options for Generic Safety Issue - 191, Assessment of Debris Accumulation on Pressurized-Water Reactor Sump Performance, July 9, 2012.

11. ACRS letter to NRC, "Safety Evaluation of License Amendment Request by South Texas Project Nuclear Operating Company to Adopt a Risk-informed Resolution of Generic Safety Issue-191," May 17, 2017 (ML17137A325).

12. WCAP-16406-P, Evaluation of Downstream Sump Debris Effects in Support of GSI-191 (PA-SEE-0195), Revision 0. Westinghouse, August 2007.

ULNRC-06526 Enclosure 2, Attachment 2-1 Page 1 of 2 Attachment 2-1 List of Regulatory Commitments

ULNRC-06526 Enclosure 2, Attachment 2-1 Page 2 of 2 List of Commitments The following table identifies the actions to which the licensee, Ameren Missouri, has committed. Statements in the submittal with the exception of those in the table below are provided for information purposes and are not considered regulatory commitments.

TYPE Scheduled Continuing Completion Date Commitment One-Time Compliance (if applicable)

[1.] Ameren Missouri shall complete all work required for final resolution of GL 2004-02 (i.e., all licensing 120 days actions such as FSAR, TS following and TS Bases changes) per X issuance of NRC the schedule agreed upon Safety Evaluation between the industry (NEI) of Callaway LAR and NRC.

Commitment number 50262

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ULNRC-06526 Enclosure 2, Attachment 2-5 Page 2 of 41 Callaway FSAR Page Markups Upon approval of the licensing basis changes, Ameren Missouri will add a new Appendix 6.3A, Risk-Informed Approach to Potential Impact of Debris Blockage on Emergency Recirculation during Design-Basis Accidents, to the FSAR. Appendix 6.3A will describe the evaluations performed using a risk-informed approach to address GSl-191 concerns including the effects on long-term cooling due to debris accumulation on containment sump strainers for ECCS and CSS in recirculation mode, as well as core flow blockage due to in-vessel effects, following LOCAs.

Note that FSAR Appendix 6.3A.2 describes change control requirements, monitoring requirements and reporting requirements which require prior NRC approval for changes, and NRC approval of the requirements described in FSAR Appendix 6.3A.2 is requested as part of this application.

FSAR Table 1.6-2 (WCAPs Incorporated by Reference and Sections 3.1 (GDC, 6.2 (Containment Systems, 6.3 (ECCS, and 15.6 (LOCA will be updated to reference Appendix 6.3A as needed.

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