RS-12-006, Units 1 & 2 - License Amendment Request for the Use of an Auxiliary Feedwater Cross-tie Between Units

Units 1 & 2 - License Amendment Request for the Use of an Auxiliary Feedwater Cross-tie Between Units

ML12033A023
Person / Time
Site:	Byron, Braidwood
Issue date:	01/31/2012
From:	Gullott D Exelon Generation Co, Exelon Nuclear
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
RS-12-006
Download: ML12033A023 (35)

Text

Exelon Generation Company, LLC www.exeloncorp.com 4300 Winfield Road r% I cl aT Warrenville, IL 60555 RS-12-006 10 CFR 50.90 January 31, 2012 10 CFR 50.59(c)(2)(ii)

U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001 Braidwood Station, Units 1 and 2 Facility Operating License Nos. NPF -72 and NPF-77 NRC Docket Nos. STN 50-456 and STN-50-457 Byron Station, Units 1 and 2 Facility Operating License Nos. NPF-37 and NPF-66 NRC Docket Nos. STN 50-454 and STN 50-455

Subject:

License Amendment Request for the use of an Auxiliary Feedwater Cross-tie Between Units In accordance with 10 CFR 50.90, "Application for amendment of license, construction permit or early site permit," Exelon Generation Company, LLC (EGC) is requesting an amendment to Facility Operating License Nos. NPF-72 and NPF-77 for Braidwood Station, Units 1 and 2, and Facility Operating License Nos. NPF-37 and NPF-66 for Byron Station, Units 1 and 2. The proposed change would revise the Updated Final Safety Analysis Report (UFSAR) to describe the use of an Auxiliary Feedwater (AF) cross-tie. Associated Technical Specification (TS)

Bases changes are included for information. Specifically, this change adds information to the UFSAR and the TS 3.7.5, "AF System" Bases describing the design and shared operation of cross-tie piping between the discharges of the Unit 1 and Unit 2 Train A motor-driven AF pumps.

EGC is requesting this amendment in accordance with the provisions of 10 CFR 50.90 and 10 CFR 50.59, "Changes, tests and experiments," paragraph (c)(2)(ii). The operation and use of the AF Train A unit cross-tie results in more than a minimal increase in the likelihood of occurrence of a malfunction of a Structure, System or Component (SSC) that is important to safety, previously evaluated in the UFSAR. NRC approval is requested for the use of the cross-tie since this function for the AF system has not previously been licensed to meet 10 CFR 50, Appendix A, General Design Criterion 5, "Sharing of structures, systems, and components" as a shared system. This proposed use of the AF Train A unit cross-tie would improve safety and the emergency response procedures for a beyond design basis total loss of secondary heat sink by providing greater assurance of achieving Reactor Coolant System conditions that allow the Residual Heat Removal System to be used for shutdown cooling.

The attached request is subdivided as follows:

- Attachment 1 provides an evaluation of the proposed change

- Attachment 2 includes the marked-up UFSAR consistent with the proposed change

- Attachment 3 includes example actions for use of the cross-tie that are applicable to an emergency procedure for a total loss of secondary heat sink

U. S. Nuclear Regulatory Commission January 31,2012 Page 2

- Attachment 4 includes the marked-up TS Bases changes for both Braidwood and Byron Stations, respectively, for information only. These TS Bases changes will be made in accordance with the TS Bases Control Program following NRC approval.

- Attachment 5 includes a figure of the AF system Approval of this amendment application is requested by August 1, 2012. The compressed schedule for approval is requested based upon the improvements to safety and reduced baseline probabilistic risk for station operation provided by the AF Train A unit cross-tie. NRC approval will allow the use of the cross-tie to be included in plant procedures and incorporated into operator training programs. Approval eliminates the need to exit procedures and plant licensing basis under the provisions of 10 CFR 50.54(x) in order to utilize the benefits of the cross-tie. Once approved, the amendment will be implemented within 30 days.

The proposed amendment has been reviewed by the Braidwood Station and Byron Station Plant Operations Review Committees and approved by the Nuclear Safety Review Board in accordance with the requirements of the EGC Quality Assurance Program.

EGC is notifying the State of Illinois of this application for a change to the UFSAR by sending a copy of this letter and its attachments to the designated State Official in accordance with 10 CFR 50.91, "Notice for public comment; State consultation," paragraph (b).

There are no regulatory commitments contained within this letter. Should you have any questions concerning this letter, please contact Richard Mcintosh at (630) 657-2816.

I declare under penalty of perjury that the foregoing is true and correct. Executed on the 31 th day of January 2012.

~41#

David M. Gullott Manager - licenSing Attachments:

1. Evaluation of Proposed Change

2. Markup / Annotated Pages of UFSAR

3. Actions on the Use of the Cross-tie in Response to a Total Loss of Secondary Heat Sink 4 . TS Bases Changes (For Information)

5. Figure: Auxiliary Feedwater System

U. S. Nuclear Regulatory Commission January 31,2012 Page 3 cc: NRC Regional Administrator, Region III NRC Senior Resident Inspector, Braidwood Station NRC Senior Resident Inspector, Byron Station NRC Project Manager. NRR - Braidwood and Byron Stations Illinois Emergency Management Agency - Division of Nuclear Safety

ATTACHMENT 1 Evaluation of Proposed Change 1.0

SUMMARY

DESCRIPTION 2.0 DETAILED DESCRIPTION

2.1 PROPOSED CHANGE

S TO THE UFSAR 2.2 ASSOCIATED CHANGES TO THE TS BASES SECTION 3.7.5, AF SYSTEM (FOR INFORMATION) 2.3 SYSTEM INFORMATION 2.4 CROSS-TIE INSTALLATION

3.0 TECHNICAL EVALUATION

3.1 PROCEDURES 3.2 RISK ASSESSMENT

4.0 REGULATORY EVALUATION

4.1 APPLICABLE REGULATORY REQUIREMENTS/CRITERIA 4.2 NO SIGNIFICANT HAZARDS CONSIDERATION DETERMINATION

6.0 REFERENCES

Page 1 of 14

ATTACHMENT 1 Evaluation of Proposed Change 1.0

SUMMARY

DESCRIPTION This evaluation supports a request to amend Facility Operating License Nos. NPF-72 and NPF-77 for Braidwood Station, Units 1 and 2, and Facility Operating License Nos. NPF-37 and NPF-66 for Byron Station, Units 1 and 2.

The proposed change would revise the Updated Final Safety Analysis Report (UFSAR) to describe the use of an Auxiliary Feedwater (AF) cross-tie. Associated Technical Specification (TS) Bases changes are included for information. Specifically, this change adds information to the UFSAR and the TS 3.7.5, "AF System" Bases describing the design and shared operation of cross-tie piping between the discharges of the Unit 1 and Unit 2 Train A motor-driven AF pumps. Exelon Generation Company, LLC (EGC) is requesting this amendment in accordance with the provisions of 10 CFR 50.90 and 10 CFR 50.59(c)(2)(ii). NRC approval is required for the use of the AF Train A unit cross-tie since this function for the AF system has not previously been licensed to meet 10 CFR 50, Appendix A, General Design Criterion (GDC) 5, "Sharing of structures, systems, and components," as a shared system.

Approval of this amendment application is requested by August 1, 2012. The compressed schedule for approval is requested based upon the improvements to safety and reduced baseline probabilistic risk for station operation provided by the AF Train A unit cross-tie. NRC approval will allow the use of the cross-tie to be included in plant procedures and incorporated into operator training programs. Approval eliminates the need to exit procedures and plant licensing basis under the provisions of 10 CFR 50.54(x) in order to utilize the benefits of the cross-tie. Once approved, the amendment will be implemented within 30 days.

2.0 DETAILED DESCRIPTION The AF Train A unit cross-tie piping and components were installed to improve safety and the emergency response procedures, by providing additional emergency beyond design basis operating flexibility that can help to maintain natural circulation and water level in the Steam Generators (SGs). The AF Train A unit cross-tie is normally isolated with two manual locked closed isolation valves. During the response to a beyond design basis loss of secondary heat sink, if the minimum AF flow to the SGs cannot be restored using the accident unit's dedicated AF trains, the AF Train A unit cross-tie may be placed into service. This requires the non-accident unit's AF Train A pump be started and the AF Train A unit cross-tie isolation valves to be locally unlocked and opened. This proposed use of the AF Train A unit cross-tie provides greater assurance of achieving Reactor Coolant System (RCS) conditions that allow the Residual Heat Removal (RH) system to be used for shutdown cooling.

Example actions that use the AF Train A unit cross-tie at Braidwood Station, Unit 1, are provided in Attachment 3 and are typical for both units at each station. A figure showing the AF system with the AF Train A unit cross-tie piping and components is provided in Attachment 5.

2.1 PROPOSED CHANGE

S TO THE UFSAR Pages from the UFSAR showing the proposed markups are provided in Attachment 2.

1. UFSAR Section 3.1.2.1.5, Evaluation Against Criterion 5 - Sharing of structures, systems, and components, currently states:

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ATTACHMENT 1 Evaluation of Proposed Change "Structures, systems, and components important to safety shall not be shared between nuclear power units unless it is shown that their ability to perform their functions, including, in the event of an accident in one unit, an orderly shutdown and cooldown of the remaining units, is not significantly impaired by the sharing."

An evaluation against GDC 5 is provided in UFSAR Section 3.1.2.1.5 that describes SSCs important to safety that are shared by the two units. The proposed change adds the following discussion to UFSAR Section 3.1.2.1.5 to reflect the capability to share the AF system between the two units:

In the event of a beyond design basis loss of all Auxiliary Feedwater (AF) on one unit, AF may be provided by the other unit via the Train A unit cross-tie connection. When the Train A unit cross-tie is operated in support of the accident unit, the motor-driven AF pump would not be available to perform its UFSAR described design basis function for the non-accident unit. If necessary, an orderly shutdown and cooldown of the non-accident unit could be accomplished using the main feedwater (FW) system. The diesel driven AF pump on the non-accident unit must be confirmed operable prior to use of the AF Train A unit cross-tie. If required, the redundant diesel-driven AF pump could also support shutdown and cooldown of the non-accident unit if the non-safety related FW system is unavailable.

2. UFSAR Section 3.1.2.4.5, Evaluation Against Criterion 34 - Residual Heat Removal, currently states:

"A system to remove residual heat shall be provided. The system safety function shall be to transfer fission product decay heat and other residual heat from the reactor core at a rate such that specified acceptable fuel design limits and the design conditions of the reactor coolant pressure boundary are not exceeded.

Suitable redundancy in components and features, and suitable interconnections, leak detection, and isolation 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."

An evaluation against GDC 34 is provided in UFSAR Section 3.1.2.4.5. The proposed change adds the following discussion to UFSAR Section 3.1.2.4.5:

The AF system automatically supplies feedwater to the steam generators (SGs) to remove decay heat from the Reactor Coolant System upon loss of normal feedwater supply. The AF system consists of a motor driven AF pump and diesel driven AF pump configured into two trains. Each pump provides 100% capacity to the SGs, as assumed in the accident analysis. One pump at full flow conditions is sufficient to remove decay heat and cool the unit to RH entry conditions. The AF system is capable of supplying, but does not normally supply, feedwater to the SGs during normal unit startup, shutdown, and hot standby conditions. The AF system is designed with suitable redundancy to offset the consequences of any single failure, with one exception during AF Page 3 of 16

ATTACHMENT 1 Evaluation of Proposed Change Train A unit cross-tie use. A normally isolated cross-tie between the discharges of both units' AF Train A pumps is available for emergency response to a beyond design basis total loss of secondary heat sink on one unit. With the Train A unit cross-tie in use, the AF Train A is not available to the non-accident unit. The diesel driven AF pump on the non-accident unit must be confirmed operable prior to use of the AF Train A unit cross-tie. The Technical Specifications limit operation with one train of AF inoperable. Use of the Train A unit cross-tie results in a temporary relaxation of the single failure criterion for the non-accident unit, which, consistent with overall system reliability considerations, provides a limited time to support the accident unit emergency response, and return the AF Train A to an operable status. Otherwise, a plant shutdown is required.

2.2 ASSOCIATED CHANGES TO THE TS BASES SECTION 3.7.5, AF SYSTEM (FOR INFORMATION)

The associated TS Bases changes for Braidwood and Byron Stations, respectively, are provided in Attachment 4. These TS Bases changes are provided for information and the changes to the TS Bases will be made to implement the proposed UFSAR changes following NRC approval in accordance with the Braidwood and Byron Station TS Bases Control Programs. The markups of TS Bases pages in Attachment 4 show what the TS Bases currently state.

1. The proposed change adds the following additional information to the BACKGROUND of TS Bases Section 3.7.5:

There is also an AF Train A unit cross-tie downstream of the motor driven AF pump at each unit that is normally isolated. Use of the AF Train A unit cross-tie, however, is incompatible with an OPERABLE motor driven AF pump on either unit and its use is limited to an emergency response for a beyond design basis event that involves a total loss of secondary heat sink.

2. The proposed change adds the following additional information to the LCO of TS Bases Section 3.7.5:

The motor-driven AF pump is not OPERABLE if the AF Train A unit cross-tie is unisolated, (i.e., both isolation valves open). The use of the AF Train A unit cross-tie is for an emergency response to a total loss of secondary heat sink on the accident unit. Use of the AF Train A unit cross-tie results in a temporary relaxation of the single failure criterion for the non-accident unit, which, consistent with overall system reliability considerations, provides a limited time to support the emergency response on the accident unit, and return the AF Train A to an OPERABLE status. Otherwise, a plant shutdown is required. The diesel driven AF pump on the non-accident unit must be confirmed OPERABLE prior to use of the AF Train A unit cross-tie.

2.3 SYSTEM INFORMATION The design basis safety function of the AF system is to remove decay heat and prevent core damage in the event of loss of normal feedwater or secondary side piping or component failure. During such transients, AF maintains SG water levels to ensure Page 4 of 16

ATTACHMENT 1 Evaluation of Proposed Change sufficient heat transfer area for removal of decay heat. AF also serves as a backup system for supplying feedwater to the secondary side of the SGs when the normal feedwater system is unavailable thus maintaining the heat sink capabilities of the SGs.

In Modes 1, 2, and 3, AF features are required to be operable to ensure that the SGs can remain the heat sink for the reactor. In Modes 4, 5, and 6, the SGs are not normally used for heat removal, and the AF system is not required.

The AF system is not used for normal startup and shutdown of the unit, or for normal operation. A simplified figure of the AF system is provided in Attachment 5.

Major components include two AF pumps (one electric motor-driven and one diesel driven); associated pump suction and discharge supply valves; and pump run-out protection orifices.

The AF system (for both Units 1 and 2) consists of a motor-driven AF pump and a diesel driven AF pump configured in two trains. Each pump provides 100 percent of the required AF capacity to the four SGs necessary to remove a conservative, core residual heat load, based on long-term operation at power. The pumps are equipped with recirculation lines to prevent pump operation against a closed system. The flow limiting orifices in each AF supply line to the SGs provide pump run out protection and minimum required flow to the unfaulted SGs in the event of a line break or faulted SG.

The AF system employs no common valving or other common components capable of spurious mechanical motion.

The Train A motor-driven AF pump is powered from an independent Class 1E power supply. The Train B diesel driven AF pump is supported by a diesel engine, an independent battery system, an essential service water booster pump, and a fuel oil day tank. Thus, the requirement for diversity in motive power sources for the AF system are met.

The AF system pumps are started on either low-low SG level, a safety injection signal, or a loss of power to the reactor coolant pumps. The AF Train A also automatically starts on an undervoltage signal from the Division 1 ESF bus. The Anticipated Transient Without Scram (ATWS) Mitigation System trips the main turbine and initiates the AF system whenever three-out-of-four SG levels are greater than 3% below the Reactor Protection System (RPS) low-low-setpoint, and the Turbine Impulse pressure is greater than 30% of nominal full power.

There are two sources of cooling water for the AF system. The AF pumps normally take suction from the Condensate Storage Tank (CST) and pump flow to the SG secondary side via separate and independent connections to the feedwater piping outside containment. If the CST is not available, the Essential Service Water System (SX) serves as the credited safety related supply to the AF pumps.

The condensate storage facility consists of two tanks, with associated pumps and valves, to serve both Units 1 and 2. The tanks at Byron have a capacity of 500,000 gallons each.

The tanks at Braidwood have a capacity of 650,000 gallons each. For both Byron and Braidwood, the CSTs have a TS minimum requirement of 212,000 gallons (useable volume), and are required to be operable in Modes 1, 2, and 3. This volume is sufficient to Page 5 of 16

ATTACHMENT 1 Evaluation of Proposed Change maintain the RCS in MODE 3 at normal operating pressure and temperature for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, followed by a cooldown to RH entry conditions at 50°F/hour, followed by a period not longer than one-hour to allow warmup of the RH pumps prior to placing the RH system into service in shutdown cooling mode.

When pressure in the supply line to the AF pumps is low and there is an actuation demand for AF, water supply to the pumps is automatically switched over to SX. At Byron, SX inventory for AF is obtained from the SX cooling tower basins. Normal makeup to the SX cooling tower basins is from the circulating water makeup pumps, while safety-related makeup is provided from two independent SX makeup pumps. A backup makeup water source is provided by two deep well pumps. The stored inventory of the SX cooling tower basins plus one of the sources of makeup water provides a sufficient supply of cooling water to support AF cooling. At Braidwood, the SX is provided by a cooling lake which provides an essentially unlimited backup AF supply. Thus, a backup supply of cooling water is available for both stations and a reasonable assurance that sufficient cooling water inventory exists.

2.4 CROSS-TIE INSTALLATION The addition of the AF Train A unit cross-tie between the 1A and 2A AF trains provides a measurable benefit in improving nuclear safety at Byron and Braidwood stations. The ability to use the AF Train A unit cross-tie is an improvement in safety for the accident unit that experiences a beyond design basis event involving a total loss of secondary heat sink.

This modification provides additional capability to supply feedwater to the SGs, and it may avoid relying heavily upon use of pressurizer power-operated relief valves (PORVs) and reactor head vents during a bleed and feed attempt to lower the pressure during such an emergency. This ensures the integrity of the RCS and reduces the impact of radiological consequences on the accident unit.

EGC installed the AF Train A unit cross-tie piping modifications in 2009 and 2010. The new piping modification provides the capability to tie the AF Train A pump discharge from one unit to the other unit, between two isolation valves. The AF Train A unit cross-tie piping is 6-inch diameter piping that is equivalent to the existing AF Train A pump discharge piping diameter. A simplified figure of the AF system is provided in Attachment

5. The AF Train A unit cross-tie line is located at the discharge of both unit's AF motor-driven pumps, 1AF01PA (Unit 1) and 2AF01PA (Unit 2), in the Auxiliary Building.

The AF Train A unit cross-tie is normally isolated with the two manual closed and locked isolation valves 1AF036 (Unit 1) and 2AF036 (Unit 2). AF Train A unit cross-tie piping is also maintained filled and vented between these two manual isolation valves.

The new piping for the cross-tie was tested for leakage (VT-2) when the modifications were installed.

The AF Train A unit cross-tie can deliver water to the accident unit's SGs, and recirculation flow returns to the non-accident unit's CST through the non-accident unit's AF022A valve, or to the SX return header through the non-accident unit's AF024 valve, depending on the water source. The accident unit's AF Train A pump recirculation is isolated for AF Train A cross-tie operation. This arrangement ensures a recirculation flow path for any running AF pump on the non-accident unit.

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ATTACHMENT 1 Evaluation of Proposed Change The operation of the AF Train A cross-tie on the discharge of the AF Train A pump is not impacted from where the suction is coming, (i.e., either CST or SX). However, when the CST is used, the Emergency Operating Procedure (EOP) involving the AF Train A cross-tie will locally unlock and close the accident unit's AF pump recirculation isolation valve (AF009A) that flows back to the CST. The EOP will then also locally open CST crosstie valve 0CD117. The CST crosstie is an existing design feature that is consistent with current EOPs. Refer to Attachment 3, example steps 6.g and 6.h. These steps perform two important functions. Closing the accident unit AF009A valve will prevent an additional recirculation flow path of non-accident unit AF flow from returning to the accident unit CST.

The non-accident AF pump is still protected with its own recirculation flowpath. Second, opening valve 0CD117 allows sharing of both the Unit 1 and Unit 2 CSTs. Opening the CST cross-tie isolation valve ensures the minimum required volume will remain available for the non-accident unit should it be required. The existing TS 3.7.6, "Condensate Storage Tank," will be applicable if CST level drops below a level that ensures the required usable volume of approximately 212,000 gallons in Modes 1, 2, and 3. If entry into TS LCO 3.7.6 is required, it will occur simultaneously for both units. The required action is to verify the operability of the backup water supply (SX) within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> and restore the CST level within 7 days. If the CST is unavailable and SX must be relied upon as the suction source, the requirements of TS 3.7.9, "Ultimate Heat Sink," ensure a sufficient volume of water is available. In addition, the resultant increase in SX flow on the non-accident unit's SX pump due to the additional load of supplying the accident unit's AF pump (approximately 990 gpm) is within the capacity of an SX pump.

EGC evaluated this piping modification under 10 CFR 50.59 and installed the change without prior NRC approval. In June 2011, the operation of the AF Train A unit cross-tie was reviewed during an NRC inspection. The review concluded that the operation of the AF Train A unit cross-tie resulted in more than a minimal increase in the likelihood of occurrence of a malfunction of a structure, system or component important to safety previously evaluated in the UFSAR. 10 CFR 50.59 requires, in part that for this type of change a license amendment is required. NRC Inspection Reports document a non-cited violation for the Braidwood and Byron Stations related to not receiving prior NRC approval, (References 2 and 3).

3.0 TECHNICAL EVALUATION

The availability of the AF Train A unit cross-tie is intended to help improve safety by restoring the normal secondary heat sink with a supply of water to the SGs. Without the cross-tie available, the EOPs direct operators to initiate bleed and feed to cope with the beyond design basis event involving the loss of secondary heat sink and a loss of heat transfer causing RCS temperature and pressure to rise. That bleed and feed strategy requires operators to initiate Safety Injection to inject into the RCS and then bleed reactor coolant out of the RCS by opening pressurizer Power-Operated Relief Valves (PORVs) or reactor head vents. The bleed and feed strategy essentially uses a controlled LOCA that introduce fission products into the Pressurizer Relief Tank or the containment atmosphere.

The AF Train A unit cross-tie can be initiated before more severe conditions develop in the RCS that necessitate a bleed and feed strategy, thereby reducing the potential for fuel damage. By utilizing the SGs per their design to transfer heat from the RCS to the Page 7 of 16

ATTACHMENT 1 Evaluation of Proposed Change secondary, non-contaminated system, the AF Train A unit cross-tie can re-fill the SGs, which will help maintain the integrity of SG tubing and the RCS, helping to keep fission products contained in RCS. Therefore, use of the AF Train A unit cross-tie allows for injection into the SGs, which protects the SG tubes from creep rupture, and scrubs fission products entering the SGs via tube leakage, while providing a heat sink for the RCS.

Considering these safety improvements, the AF Train A unit cross-tie is evaluated to reduce the overall unit baseline core damage frequency.

Either of the two AF trains, supplying the four SGs, provides sufficient feedwater to cool the unit down safely to the temperature at which the RH system can be utilized for shutdown cooling. The events which impose safety-related performance requirements on the design of the AF system include the following: loss of main feedwater transient, secondary system pipe breaks, loss of all a-c power, loss-of-coolant accident (LOCA),

and cooldown (after expected transients, accidents, and other scenarios).

Operation of the AF Train A unit cross-tie affects only the A Train associated with the motor-driven AF pumps. The AF Train A piping is safety related and seismically qualified.

The redundant diesel-driven B Train flow path is not affected because none of the added piping and components of the AF Train A unit cross-tie are connected to the B Train of AF and appropriate physical separation is maintained to preclude any adverse interactions with the redundant AF train. In addition, added cross-tie piping and valves will not create any adverse interactions with any AF control devices nor are any new AF failure modes introduced. Failure of the AF Train A unit cross-tie piping during design basis accidents and transients does not impact the ability of the AF system to perform its function because the piping is isolated from the normal system flowpath.

The criteria for AF system design basis conditions (UFSAR Table 10.4-5) and the summary of assumptions used in AF system design verification analyses (UFSAR Table 10.4-6) are not changed.

The AF system is designed to provide adequate feedwater to the unfaulted SGs in response to accidents/transients described in the UFSAR, considering a worse case single failure. The limiting accident considered in the AF failure modes-effects analysis is a break of a main feedwater line in close proximity to a SG, as documented in UFSAR Table 10.4-4. Failure of AF pumps, control valves, stop valves, feedwater lines, power supplies, and gate valves are evaluated in the failure analysis.

The AF Train A unit cross-tie does not add any active components subject to failure that would impact the failure modes-effects analysis or radiological consequences.

The redundant diesel driven AF pump flow path will not be affected because none of the added piping is connected to the B Train of AF and appropriate separation is maintained to preclude any adverse interactions with the redundant AF train.

Therefore, the previously evaluated radiological consequences will not be affected.

During normal operation and design basis events, the AF Train A unit crosstie remains isolated with at least one locked and closed valve. These valves are normally in a locked closed position and administratively controlled by procedures to ensure safety functions are fulfilled. Therefore, these added system interfaces will not result in an increase in the frequency of an accident previously evaluated in the UFSAR, because an accident is not initiated nor are any new failure modes introduced.

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ATTACHMENT 1 Evaluation of Proposed Change With the AF Train A unit cross-tie isolation valves in the normally locked closed position, the proposed piping configuration changes will have a negligible impact on the hydraulics of the AF system. Therefore, the required design basis flow to the SGs following an accident, as documented in UFSAR Table 10.4-8, is not impacted by this change.

During AF Train A unit cross-tie operation, the flow of AF through the additional short section of piping and valves was evaluated and determined to have a minor impact on the hydraulics of the AF system. The EOP will direct isolation of the accident unit recirculation flow path, however, if the recirculation flow path to the CST remains open on the unit with the accident, there may be a minor reduction in AF flow to the SGs.

However, since this is a condition that exceeds postulated design basis events, the design basis flow to the SGs defined in Table 10.4-8 are not applicable. In addition, AF pump runout is not an issue based on the use of existing system flow restriction orifices and based on existing operating procedures. Consequently, the motor-driven AF pump will not be adversely affected by the AF Train A unit cross-tie use.

Since the motor-driven AF pump will be operated hydraulically consistent with its current design, the imposed electrical loads on the pumps will be within current design limits.

With a loss of offsite power on a unit, the emergency diesel generators continue to provide electrical power to the motor-driven AF pump during design basis accidents and natural phenomena scenarios.

Although manual operator actions will be needed to align the motor-driven AF pump for cross-tie flow, this is not a permanent substitution of a manual action for an automatic function, since the AF Train A unit cross-tie is only used during an event that is more severe than postulated design basis conditions.

The inventory of the CST will be depleted when the AF Train A unit cross-tie is in operation. The CST crosstie isolation valve will be opened, which allows sharing of both the Unit 1 and Unit 2 CSTs. Opening the CST cross-tie isolation valve ensures the minimum required volume will remain available for the non-accident unit should it be required. The existing TS 3.7.6 for CST level will be applicable if CST level drops below a level that ensures the required useable volume of approximately 212,000 gallons in Modes 1, 2, and 3. The required action to verify the backup water supply (SX) is within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, with restoration of the CST level within 7 days. There are no expected conditions brought about by use of the AF Train A cross-tie that would prevent these actions from being met. If the CST is unavailable and SX must be relied upon as the suction source, the requirements of TS 3.7.9, "Ultimate Heat Sink," ensure a sufficient volume of water is available. In addition, the resultant increase in SX flow on the non-accident unit's SX pump due to the additional load of supplying the accident unit's AF pump (approximately 990 gpm) is within the capacity of an SX pump.

The AF Train A unit cross-tie line is not credited to mitigate the consequences of design basis accidents. There is no change to the radiological consequences because the non-accident unit will still be operating within the existing LCO.

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ATTACHMENT 1 Evaluation of Proposed Change The use of the AF Train A unit cross-tie allows for injection into the SGs, which protects the SG tubes from creep rupture, and scrubs fission products entering the SGs via tube leakage, while providing a heat sink for the RCS. Because the AF Train A unit cross-tie operation can be initiated to fill SGs before more severe conditions develop in the RCS that necessitate a bleed and feed strategy, this proposal reduces the potential for fuel damage. With the ability to operate the AF Train A unit cross-tie the operators may be able to avoid initiation of bleed and feed evolutions.

The ability to use the AF Train A unit cross-tie results in a reduction in operator actions required for safe shutdown of a unit experiencing a beyond design basis total loss of secondary heat sink, if the bleed and feed evolutions can be avoided. The reduction in operator actions is a reduced potential for human error. Therefore, the potential for operator error is not increased by the proposed change. In addition, there is no impact to operator actions required for the ability to isolate a faulted SG.

The AF Train A unit cross-tie piping and components are classified and designed in accordance with the UFSAR requirements and applicable referenced ASME code requirements for the design of the AF system piping. The piping is evaluated using moderate energy pipe break location methodologies as described in the UFSAR. The affected pipe supports were analyzed in accordance with the structural design requirements as described in the UFSAR. The new valves are seismically qualified in accordance with the existing UFSAR described methodology.

3.1 PROCEDURES The AF Train A unit cross-tie piping and components have been installed, and the manual isolation valves are normally in a locked closed position and administratively controlled by procedures. There is one anticipated evolution where both the AF Train A unit cross-tie isolation valves will be opened, that is to support a beyond design basis accident on one unit.

The approval of this request will permit implementation of the proposed UFSAR changes.

The associated TS Bases changes clarify existing TS LCO requirements and the applicability of the Condition for an inoperable train of AF on one unit when using the cross-tie. The approval of this request also allows EOPs for the Braidwood and Byron stations for a loss of secondary heat sink to be changed to reflect the example that has been provided in Attachment 3, for use of the AF Train A cross-tie piping and components.

Although specific EOP content or the changes that will be implemented upon approval of the amendment have been provided as an example in Attachment 3, the specific steps of the EOP provided with this proposal are for information only, and may be modified upon final development and implementation. The implementation of approval for this proposal will require EOP changes to use the AF Train A unit cross-tie.

Braidwood and Byron Stations utilize procedures that track the actions that are taken in the event any TS Limiting Condition for Operation (LCO) is not met. These procedures assist in complying with the TS by providing documentation that the unit is maintained in the required mode of operation for safe operation until the LCO is restored. If the AF Train A unit cross-tie must be placed into use, the AF motor-driven pump, Train A, is Page 10 of 16

ATTACHMENT 1 Evaluation of Proposed Change considered inoperable on the non-accident unit, and the associated action of the TS would be entered.

3.2 RISK ASSESSMENT Although this is not a risk informed license amendment request, the risk benefit was quantified by evaluating the current internal events Probabilistic Risk Assessment (PRA) model for Byron and Braidwood, with the AF Train A unit cross-tie available and with it disabled.

The benefit from implementation of the AF Train A unit cross-tie capability was modeled under this configuration and the results compared to the base Core Damage Frequency (CDF) and Large Early Release Frequency (LERF) results for the current model. This information is summarized in the table provided below.

CDF and LERF Benefits from AF Train A Unit Cross-Tie Unit 1 Unit 2 Station LERF LERF CDF Reduction CDF Reduction Reduction Reduction Byron 4% 1% 6% 12%

Braidwood 4% 2% 6% 12%

As shown, there is a significant reduction in risk for all four units.

4.0 REGULATORY EVALUATION

4.1 APPLICABLE REGULATORY REQUIREMENTS/CRITERIA Apart from the UFSAR changes proposed, there is no impact to the applicable regulatory requirements for the AF system. The following 10 CFR 50, Appendix A, GDC are applicable to the Byron Units 1 and 2 and Braidwood Units 1 and 2 AF systems:

 GDC 2 - Design Bases for Protection Against Natural Phenomena

 GDC 4 - Environmental and Missile Design Bases

 GDC 5 - Sharing of structures, systems, and components

 GDC 19 - Control Room

 GDC 34 - Residual Heat Removal

 GDC 44 - Cooling Water

 GDC 45 - Inspection of Cooling Water System

 GDC 46 - Testing of Cooling Water System The AF system was designed to meet the requirements of these GDC with respect to protection against natural phenomena, missiles, environmental effects, shared systems, operational capability from the control room, decay heat and cooling water capability, inservice inspection and functional testing. The AF system is designed consistent with guidelines of Regulatory Guide 1.29, "Seismic Design Criteria," and the Branch Technical Position (BTP) ASB 10-1, "Design Guidelines for Auxiliary Page 11 of 16

ATTACHMENT 1 Evaluation of Proposed Change Feedwater System Pump Drive and Power Supply Diversity for Pressurized Water Reactors," and RSB 5-1, "Design Requirements of the Residual Heat Removal System," concerning seismic classification, power diversity, and design of decay heat removal systems. The AF system also meets recommendations of NUREG-0611, "Generic Evaluation of Feedwater Transients and Small Break Loss-of-Coolant Accidents in Westinghouse-Designed Operating Plants," concerning generic improvements to the AF design, procedures, and TS and AF system reliability.

This change proposes additional information necessary to define AF system compliance with GDC 5, and 34, which reflects the use of the shared AF system function of the AF Train A unit cross-tie piping. The AF system is designed with suitable redundancy to offset the consequences of any single failure, with the exception to allow AF Train A unit cross-tie use during a beyond design bases event.

Use of the AF Train A unit cross-tie results in a temporary relaxation of the single failure criterion for the non-accident unit, which, consistent with overall system reliability considerations, provides a limited time to support the accident unit emergency response during a beyond design basis event, and return the AF Train A to an Operable status. Otherwise, a plant shutdown is required.

The proposed change continues to meet GDC 5 because in the event of an accident on one unit, an orderly shutdown and cooldown of the other unit is not impaired by the sharing. This is due to the fact that the AF system is not used to perform orderly/controlled shutdown of a unit. Therefore, GDC 5 conditions are met with the sharing of the AF Train A between units. The use of the AF Train A unit cross-tie is predicated upon the operability of the diesel driven AF pump.

The diesel driven AF pump functions independently of any onsite or offsite a-c power, and is thus not affected by a loss of offsite power. The exception to single failure criterion necessary for AF Train A unit cross-tie use is recognized in the proposed change to the UFSAR in response to GDC 34. However, existing TS are not changed and continue to ensure that the non-accident unit enters TS 3.7.5 Condition A for one inoperable AF Train. This TS condition allows the unit to operate with relaxed single failure criterion for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. This approach is consistent with GL 80-30 and NRC Inspection Manual Part 9900 which conclude that TS allow continued operation with only one train of a two train safety system operable. In these cases, the GDC continues to be met because the system design provides the necessary redundancy and the TS permit continued operation for a specified time.

EGC has evaluated the impact of the AF cross-tie against GDC 44, Cooling Water, and concluded that the SX system functions are maintained in accordance with this GDC. The SX supply to the AF system has been analyzed and does not impact the safety function of the SX system or the Ultimate Heat Sink. For the non-accident unit, the AF Train B will remain operable during cross-tie operation. In this condition, the SX supply to the AF Train B remains capable of withstanding a single failure since SX remains capable of supplying cooling water to AF via two independent SX trains.

Therefore, the AF Train B is able to perform its safety function assuming a single failure on the SX system. Furthermore, the resultant increase in SX flow on the non-accident unit's SX pump due to the additional load of supplying the accident unit's AF pump (approximately 990 gpm) is within the capacity of an SX pump. The impact of Page 12 of 16

ATTACHMENT 1 Evaluation of Proposed Change the cross-tie on the AF systems ability to withstand a single failure is addressed in the GDC 34 discussion above.

10 CFR 50.36 (c)(2)(ii), stipulates that a TS limiting condition for operation must be established for each item meeting one or more of the following criteria.

1. Installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary.

2. A process variable, design feature, or operating restriction that is an initial condition of a design basis accident or transient analysis that either assumes the failure of, or presents a challenge to the integrity of a fission product barrier.

3. A structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a design basis accident or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

4. A structure, system, or component which operating experience or probabilistic risk assessment has shown to be significant to public health and safety.

The AF system satisfies the Criterion 3 of 10 CFR 50.36(c)(2)(ii). However, no change to TS are proposed that would modify the manner in which the AF system continues to meet this regulation.

4.2 NO SIGNIFICANT HAZARDS CONSIDERATION DETERMINATION Exelon Generation Company, LLC (EGC) has evaluated the proposed changes to the Technical Specifications (TS) using the criteria in 10 CFR 50.92 and has determined that the proposed changes do not involve a significant hazards consideration.

Description of Amendment Request:

The proposed changes request an amendment to Facility Operating License Nos. NPF-72 and NPF-77 for Braidwood Station, Units 1 and 2, and Facility Operating License Nos.

NPF-37 and NPF-66 for Byron Station, Units 1 and 2.

The proposed change would revise the Operating License(s) to add information to the Updated Final Safety Analysis Report (UFSAR) and Technical Specification (TS) Bases describing the use of an Auxiliary Feedwater (AF) Train A unit cross-tie between units during emergency response to a beyond design basis total loss of secondary heat sink.

Specifically, this change adds information to the UFSAR and the TS 3.7.5, "AF System" Bases describing the design and shared operation of AF Train A unit cross-tie piping between the discharges of the Unit 1 and Unit 2 Train A motor-driven AF pumps. EGC is requesting this amendment in accordance with the provisions of 10 CFR 50.90 and 10 CFR 50.59(c)(2)(ii). NRC approval is required for the use of the AF Train A unit cross-tie since this shared function between units for the AF system has not previously been licensed to meet 10 CFR 50, Appendix A, General Design Criteria (GDC) 5, "Sharing of structures, systems, and components," as a shared system.

Page 13 of 16

ATTACHMENT 1 Evaluation of Proposed Change Basis for proposed no significant hazards determination:

As required by 10 CFR 50.91(a), the EGC analysis of the issue of no significant hazards consideration is presented below:

1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No The AF system is normally in standby and a failure of the AF system during normal operations or emergency operations cannot initiate any of the accidents previously evaluated. The use of the AF Train A unit cross-tie does not interface with the reactor coolant system, containment, or engineered safeguards features in such a way as to be a precursor or initiator for an accident previously evaluated. The AF system is capable of performing the safety-related functions required to mitigate the effects of design basis accidents. Conditions which impose safety-related performance requirements on the design of the AF system include the following:

loss of main feedwater transient, secondary system pipe breaks, loss of all a-c power, loss-of-coolant accident (LOCA), and cooldown (after expected transients, accidents, and other scenarios). For the non-accident unit, controls ensure compliance with existing TS conditions that ensure one train remains operable and the condition exists for a limited time. The AF system will continue to be used in compliance with the existing conditions in the TS. Since the AF system is assured of performing its intended design function in mitigating the effects of design basis accidents, the consequences of accidents previously evaluated in the UFSAR will not be increased. Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No Failures of the AF system cannot initiate an accident. The proposed use of an AF Train A unit cross-tie will not interface with the reactor coolant system, containment, or engineered safeguards features. Failure modes and effects described in the UFSAR are not impacted. The electrical power supplies and AF system pumps will be maintained in design basis train alignments. Use of an AF Train A unit cross-tie will have no impact on the range of initiating events previously assessed. Thus, the accident analysis presented in the UFSAR is not impacted. The change is consistent with the safety analysis assumptions.

Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

Page 14 of 16

ATTACHMENT 1 Evaluation of Proposed Change

3. Does the proposed change involve a significant reduction in a margin of safety?

Response: No The margin of safety is not reduced. Results of the existing UFSAR accident analysis are not impacted, and therefore the safety margins are not impacted. The proposed change will not reduce a margin of safety because the non-accident unit will be operated within existing TS conditions. For the non-accident unit, controls ensure compliance with existing TS conditions that ensure one train remains operable and the condition exists for a limited time. The AF Train A unit cross-tie is not a credited flow path in design basis or needed to meet a safety function. The AF Train A unit cross-tie is an additional strategy made available if a total loss of secondary heat sink should occur. The AF Train A unit cross-tie would be initiated if the feed flow to at least one SG cannot be verified during the event, and an appropriate SG level cannot be maintained to regain secondary heat sink. As such, the AF Train A unit cross-tie is an improvement in emergency procedures for a total loss of heat sink, and this improves probabilistic risk assessment. The proposed change, therefore, does not involve a reduction in a margin of safety.

Based on the above evaluation, EGC concludes that the proposed amendment presents no significant hazards consideration under the standards set forth in 10 CFR 50.92, paragraph (c), and accordingly, a finding of no significant hazards consideration is justified.

4.3 CONCLUSION

In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by the operation of Byron and Braidwood Units 1 and 2 in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or the health and safety of the public.

5.0 ENVIRONMENTAL CONSIDERATION

EGC has evaluated the proposed amendment for environmental considerations, consistent with 10 CFR 51.21, "Criteria for and identification of licensing and regulatory actions requiring environmental assessments." The review results in the determination that the proposed change will modify a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would change an inspection or surveillance requirement. However, the proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any 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 paragraph (c)(9) of 10 CFR 51.22, "Criterion for categorical exclusion; identification of licensing and regulatory actions eligible for categorical exclusion or otherwise not requiring environmental review." Therefore, in accordance with 10 CFR 51.22(b), no environmental impact statement or environmental assessment needs to be prepared in connection with the proposed amendment.

Page 15 of 16

ATTACHMENT 1 Evaluation of Proposed Change

6.0 REFERENCES

1. Braidwood Station and Byron Station Updated Final Safety Analysis Report (UFSAR),

Revision 13.

2. Letter from NRC to M. J. Pacilio, "Byron Station Units 1 and 2 NRC Integrated Inspection Report 05000454/2011004; 05000455/2011004," dated November 3, 2011

3. Letter from NRC to M. J. Pacilio, "Braidwood Station, Units 1 and 2, NRC Integrated Inspection Report 05000456/2011004; 05000457/2011004," dated November 9, 2011 Page 16 of 16

ATTACHMENT 2 Braidwood Station, Units 1 and 2 Byron Station, Units 1 and 2 Facility Operating License Nos.: NPF-72, NPF-77, NPF-37 and NPF-66 Docket Nos.: 50-456, 50-457, 50-454 and 50-455 Markup / Annotated Pages of UFSAR

ATTACHMENT 2 Markup / Annotated Pages of UFSAR 3.1..2'.1.4 Evaluation A g ainstt Criterion 4 - Environmental and ssi a Design bases USruuctures, systems and cctvononts inportaxit to safety shall be designed ter ac comt date the effects of and. be coups ibi a with the vironm^ntal conditions associated with normal operation, maintenance, testing and postulated accidents,' including loss of coolant accidents. Thesestructures, systems, and components shall be a ropriately pratected..against c .effects, including he affects of uu.ssiles, pipe whipping, . and, discharging fluids, that may result f rota ecltxip nt failures and from events and conditions outside tine nuclear power unit."

RES PONSE ety-related cyst , couponents, and structures in this plant are designed to ac ate all normal or routine environmental conditions as. well as those. associated with post hated accidents (where ,Mate) . The design includes provisions to tact (by physical. separation, barriers, or appropriate restra ts) safety-relate i t from dynamic facts rul t' f rom component fail ures , and specific. credible. external events, and conditions_

The design criteria for these systems, conponants, are discussed in the remainder of this chapter.

3.12.1. 5 Evaluation Against Criteri on S - Sharing of cures, Sys Eaffs, and cour Wants systems, and components important to safety shall not be shared between nuclear power units unless it is shown that their ability to perform their functions, including, in the event of an .cadent in one unit, an orderly shutdown ar cooldown of the remaining units, is. not significantly :impaired- by the sharing-RES P4NSE use system, structures, and ccsnconet s important: to safety shared by tone two units are the ultimate heat sinks and the associated Byron " water system; various" heating..

ventilating, and air condition ing sys tem within the shared auxiliary and fuel handling buildin 1 and a c ent cooling heat exchanger which, can be valved to serve one unit or the other. These shared systems, structures, an d co nents are mare fully described elsewhere in this report.. No safety -related system, structures, or componants are shared unless such sharing has been evaluated to ensure that there will be no 'significant adverse impact on safety functions.

in the event of a beyond design basis loss of all Auxiliary Feedwater (AF) on one unit, AF may be provided by the other unit via the Train A unit cross-tie connection. When the Train A unit cross-tie is operated in support of the accident unit, the motor-driven AF pump would not be available to perform its UFSAR described design basis function for the non-accident unit. If necessary, an orderly shutdown and cooldown of the non-accident unit could be accomplished using the main feedwater (FW) system. The diesel driven AF pump on the non-accident unit must be con d operable prior to use of the AF Train A unit cross-tie. If required, the redundant d' -driven AF pump could also suppcxt shutdown and coo n of the non-accident unit if the non-safety related FW system is unavailable.

ATTACHMENT 2 Markup I Annotated Pages of UFSAR No changes this page For Information Only The material properties surveillance program includes not only the conventional, tensile and impact tests, but also fracture mechanics specimens. The observed shifts in RTC of the core region materials with irradiation will be used to confirm the allowable limits calculated for all operational transients.

See the appropriate sections in Chapter 5.0 for further details on inspection and surveillance requirements.

3.1.2.4.4 Evaluation Against Criterion 33' - Reactor Coolant eu "A system to supply reactor coolant makeup for protection against small breaks in the reactor coolant pressure boundary shall be provided. The system safety function shall be to assure that specified acceptable fuel design limits are not exceeded as a result of reactor coolant loss due to leakage from the reactor coolant pressure boundary and rupture of small piping or other small components which are part of the boundary. The system shall be designed 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 using the piping, pumps,. and valves used to maintain coolant inventory during normal reactor operation."

RESPONSE

The chemical and volume control system provides a. means of reactor coolant makeup to ensure appropriate makeup supply for small breaks as described in Subsection 9.3.4 and adjustment of the boric acid concentration. Makeup is added automatically if the level in the volume control tank falls below the preset level. The high-pressure centrifugal charging pumps provided are capable of supplying the required makeup and reactor coolant seal injection flow when power is available from either onsite or offsite electric power systems. These pumps also serve as high-head safety injection pumps. Functional reliability is assured by provision of standby components assuring a safe response to probable modes of failure. Details of system design are included in Section 6.3, with details of the electric power system included in Chapter 8.0.

3.1.2.4.5 Evaluation A ainst Criterion 34 - Residual Heat emova "A system to remove residual heat shall be provided. The system safety function shall be to transfer fission product decay heat and other residual heat from the reactor core at a rate such that specified acceptable fuel design limits and the design conditions of the reactor coolant pressure boundary are not exceeded.

ATTACHMENT 2 Markup / Annotated Pages of UFSAR B/B-UFSAR

• Suitable redundancy in components and features, and suitable interconnections, leak detection, and isolation capabilities shall be provided to assure that for onsite electric power system operation (assuming offeite 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."

RESPC tfSE The residual heat removal (RHR) system in conjunction with the steam and power conversion system, is designed to transfer the fission production decay heat and other residual heat from the reactor core within acceptable limits. The crossover from the steam and power conversion system to the residual beat removal system occurs at approximately 350GF and. 360 psig.

Suitable redundancy at tempe ratures below approximately 3550*F is accomplished with the two residual beat removal pumps (located in rate c t wags a it le for drab itoring of leakag y, the two heat exchangers; and the associated piping; cabling, and. electric power sources. The residual heat removal system is capable of operating en either onsite or offsite electrical power, "ents Suitable redundancy at temperatures above approximately 354°R is provided by the four steam generators and attendant piping, Details of the system designs are given in Sections 5.4 and 9.2 and Chapter 10.0.

3.1.2.4.6 Evaluation Against Criterion 35 - fteEgency Core ooI ng

• A s stem The AF system automatically supplies feedwater to the steam generators provided.

from the r (SGs) to remove decay heat from the Reactor Coolant System upon loss of rate such normal feedwater supply. The AF system consists of a motor driven AF continued pump and diesel driven AF pump configured into two trains. Each pump metal-ware provides 100% capacity to the SGs, as assumed in the accident analysis.

"Suitable One pump at full flow conditions is sufficient to remove decay heat and cool interconn the unit to RH entry conditions. The AF system is capable of supplying, but capabiliti power syst does not normally supply, feedwater to the SGs during normal unit startup, and for of shutdown, and hot standby conditions. The AF system is designed with power is z accomplish suitable redundancy to offset the consequences of any single failure, with one exception during AF Train A unit cross-tie use. A normally isolated RESPONSE cross-tie between the discharges of both units` AF Train A pumps is available An a rgen{ for emergency response to a beyond design basis total loss of secondary lose-of-co heat sink on one unit. With the Train A unit cross-tie in use, the AF Train A water is a is not available to the non-accident unit. The diesel driven AF pump on the at a rate non-accident unit must be confirmed operable prior to use of the AF Train A unit cross-tie. The Technical Specifications limit operation with one train of AF inoperable. Use of the Train A unit cross-fie results in a temporary relaxation of the single failure criterion for the non-accident unit, which, consistent with overall system reliability considerations, provides a limited time to support the accident unit emergency response, and return the AF Train A to an operable status. Otherwise, a plant shutdown is required.

ATTACHMENT 22 ATTACHMENT Markup I/ Annotated Pages Pages of ofUFSAR UFSAR No changes this page page --

or Information Only FFor Cool Cooldowndown The coolcooldown down function performed by the auxiliary feedwater feedwater system is is aa partial one since the the reactor reactor coolantcoolant system system isis reduced reduced from from normal normal zero load temperature to a a hot leg temperature of approximately approximately 350°F. 350°F. The latter is the the maximum maximum temperature temperature recommended for placing the the residual residual heat heat removal removal system system (RHRS) (RHRS) into service. The RHR RHR system completes the cooldown to cold shutdown conditions.

Cooldown may be be required following expected transients, transients, following an accident such as a main main feedline feedline break, break or l or during during aa normal normal cooldown prior to refueling or or performing performing reactor reactor plant plant maintenance. If If the the reactor is tripped tripped following following extended extended operation at rated power level, level the t the AFWS A.FWS is is capable capable ofof delivering delivering sufficient AFW to remove decay decay heat heat andand reactor reactor coolant coolant pump pump (RCP)

(RCP) heat heat following following reactor trip while maintaining the the steam steam generator (SG) (SG) water water level.level. Following transients transients or or accidents, accidents I the the recommended cooldown rate rate is is consistent consistent with with expected expected needs needs and at the same time time does not not imposeimpose additional additional requirements requirements on on the the capacities of the auxiliary auxiliary feedwaterfeedwater pumps, considering aa pumpsl considering single single failure.

failure. In In any any event, event the l the process process consistsconsists of of being being able able to dissipate plant sensible heat heat in in addition addition to to the the decay decay heat heat produced by by thethe reactor reactor core. core.

Descriptions of the the analyses analyses of of thethe design design and and supporting supporting assumptions are provided provided in in Subsection Subsection 10.4.9.3.2.

10.4.9.3.2.

10.4 10.4.9.2 . 9 .2 System Description Description The auxiliary feedwater feedwater system system consistsconsists of of two two subsystems.

subsystems. One One subsystem utilizes an electric-motor-drivenelectric-motor-driven pump, pump, which which is is powered from powered from one one of of the the emergency emergency onsite onsite power power systems systems supplied supplied from aa diesel generator; from generator; the other other utilizes utilizes aa pump pump that that is is directly powered by by a dieseldiesel engineengine throughthrough aa gear gear increaser.

increaser.

Each of the the twotwo subsystems subsystems can can deliver deliver feedwater feedwater to to all all four four steam generators. The system system has been been designed designed to to provide provide adequate feedwater to to thethe unfaulted unfaulted steam steam generators generators in in thethe event event of a a main feedwater feedwater or or steamline steamline break break coupled coupled with with aa single single active or passive failure failure in in the the auxiliary auxiliary feedwater feedwater system, system, as as shown in Table Table 10.4-4. Equipment redundancy, 10.4-4. Equipment redundancy, flow flow paths, paths safety l safety class and and quality quality groupgroup boundaries, boundaries, major major line sizes, sizes and t andsystem system~________ _~

operation operation are areillustrated illustrated on the on system the system diagram,diagram, Drawing Drawing MMM- 3 7 For this For this page, page,

~

shows the this shows the The auxiliary feedwater systems systems are not utilized utilized for for n'rmal n rmal UFSAR as UFSAR it as it startup and startup and shutdown shutdown of of the the units.

units. They They are, therefo ee l are, therefo currently exists.

currently exists.

classified as classified as moderate-energy moderate-energy systems. systems. There are There are no no A--c-r-o-s-s-t-l.-*

crosstie e---l-i-n-e-o-f-f--t-h-e-d-l.-'

line off the discharge s-c-h-a-r-g-e-p-l.-*

p-i-n-g-o-f-t-h-e--m-o-t-o-r---d-r-i-v-e-n----1 piping of the motor-driven new new changes changes toto this page that I auxiliary feedwater auxiliary feedwater pumps, pumps 1 fromfrom one one unitunit to to thethe other, other 1 provides provides this page that addi tional beyond-design-basis additional beyond-design-basis emergency emergency operating flexibility are operating flexibility proposed so are proposed so and risk and risk enhancement.

enhancement. The The AF AF crosstie crosstie line line between between thethe units uni ts this this page page is is for for I information contains two contains two manual manual isolation isolation valves valves that that are are locked locked closed closed information during normal normal plant plant operation.

operation. AF crosstie crosstie operationoperation is is notnot only.

only.

c__r_e_d_i_t_e_d___

credited in in ___ __y__design an any d_e_s__i_g_n__

b_a_s_e__

bases s_e_v_e_n_t event__ r_e__sp

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response.

10.4-33 10.4-33 REVISION 13 DECEMBER 2010 13 -- DECEMBER 2010

ATTACHMENT 3 Braidwood Station, Units 1 and 2 Byron Station, Units 1 and 2 Facility Operating License Nos.: NPF-72, NPF-77, NPF-37 and NPF-66 Docket Nos. 50-456, 50-457, 50-454 and 50-455 Actions on the Use of the Cross-tie in Response to a Total Loss of Secondary Heat Sink The steps shown in this attachment are an example of content applicable to Braidwood Station, Unit 1, and are typical for both units at each station.

ATTACHMENT 3 Actions on the Use of the Cross-tie in Response to a Total Loss of Secondary Heat Sink A. PURPOSE Th is procedu r e provid es actions to r espo nd to a l oss of second a ry he aL sI nk In al l s Leam gcneraLors .

B. SYMPTOMS OR ENTRY CONDITIONS Th is procedur e is e nt ered fro m:

o 1Bwl';P-O , RI';AC'I' OR 'l'R I P OR SAlo' l';'I' Y I NJ I'; C'I' ION , Step 1 5 , wh e n mi ni mum AF flow i s not verified AND narrow rang e l e v el in a ll 8Gs i s l ess Lha n 10% (3 1 % ADVERS E CNMT) .

o 1 BwST-J , HEAT SINK, Critical Safety Functio n Status Tre e o n a RED condition .

ATTACHMENT 3 Actions on the Use of the Cross-tie in Response to a Total Loss of Secondary Heat Sink ACTIONJEXPECTED RESPONSE RESPONSE NOT OBTAINED J.

~

The fo lowing step wi 1 .ake the 2A AF pump and the nit 2 CST J.

J. inoperable.

1t 1t 1t *1t ./ti 1t*1t ." ." 1t ." ." ." ." ." 1t ."

6 CROSSTI E A-TRA I N Af FROM OPPOSITE

.Il1il.I:

a. C~eck Unit 2 - I N MOPE 1-3 a . .II. Unit 2A AF pump is available, lB£R perfo~ the following:

1) Locally energize UBit 2A AF pump Awe Oil pump b :ea ke : :

• MeC 23lX3 Al

2) GO TO Step 6c .

.II. Unit 2A AF pump is llill::.

available, IB£a GO TO Ste. 7 (Page I ) .

b. C~eck Unit 2 AF pumps - BQ!H b. GO TO Step 7 (Page 11) .

OPERABLE

c. Close th.e fol ow:ing Unit 2 c. Locally iso ate AF flow:

Train A AF Iso l va ves:

1) Dispatch an operator to

• 2AF013A f ail Unit 2 Train A AF

• 2AF 13B flow control va.lves by

• 2AF013 iso ating air t o t h e

• 2AF 13:> valve operators.

2) Locally close 2AFOOSA thru D (364' PI O).

d. Locally unlock and open Train A d . .II. both crosstie valves AF unit crosstie isol v alv es: cannot be open ed, lB£R GO TO Ste 7

• AF036 (383 ' IUa ) ( F2 ey) (Page 1 ) -

• 2AF036 (383 ' IUa ) ( F2 ey)

Step continued on next page

ATTACHMENT 3 Actions on the Use of the Cross-tie in Response to a Total Loss of Secondary Heat Sink ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED Step 6 (continued) e . Start Unit 2A AF pump f . Check total feed to flow to f. If any feed flow to at Unit 1 SGs - GREATER THEN 500 least one SG is verified, GPM THEN perform the following:

1) Maintain feed flow to restore narrow range level to greater than 10% (31% ADVERSE CNMT).

2) WHEN narrow range level is greater than 10% (31%

ADVERSE CNMT),

THEN RETURN TO procedure and step in effect .

3) GO TO Step 7 (Next Page) .

~ feed flow is NOT verified ,

THEN GO TO 7 (Ne xt Page) .

g . Locally unlock and close 1A AF pump recirc isol valve:

• 1AF009A (383' M18) (lU4 key) h . Locally open CST crosstie valve:

• OCDl17 (U - 2 CST valve pit)

(G - O key) i . RETURN TO procedure and step in effect

ATTACHMENT 4 Braidwood and Byron Stations, Units 1 and 2 Facility Operating License Nos.: NPF-72, NPF-77, NPF-37, and NPF-66 Docket Nos.: 50-456, 50-457, 50-454, and 50-455 Technical Specifications Bases Changes All TS Bases Pages are Included for Information Only

ATTACHMENT 4 Technical Specifications Bases Changes AF System B 3.7.5 B 3.7 PLANT SYSTEMS B 3.7.5 Auxiliary Feedwater CAF) System BASES BACKGROUND The AF System automatically supplies feedwater to the Steam Generators (SGs) to remove decay heat from the Reactor Coolant System upon the loss of normal feedwater supply.

The AF pumps normally take suction from the condensate storage tank (CST) (LCO 3.7.6) and pump to the steam generator secondary side via separate and independent connections to the feedwater piping outside containment. If the CST is not available, AF can be supplied by the Essential Service Water System. The steam generators function as a heat sink for core decay heat . The heat load is dissipated by releasing steam to the atmosphere from the steam generators via the Main Steam Safety Va lves CMSSVs)

(LCO 3.7.1 ) or SG Power Operated Relief Valves (PORVs)

(LCO 3.7.4). If the main condenser is available, steam may be released via the steam dump valves and recirculated to the CST.

The AF System consists of a motor driven AF pump and a diesel driven pump configured into two trains. Each pump provides 100% of the required AF capacity to the steam generators, as assumed in the accident ana lys is. The pumps are equipped with independent recirculation lines to prevent pump operation against a closed system. The motor driven AF pump is powered from an independent Class IE power supply and feeds four steam generators. The diesel driven AF pump is powered from an independent diesel and also feeds four steam generators. The diesel driven AF pump is supported by a diesel engine, an independent battery system, an essential service water booster pump, and a fuel oil day tank. Thus, the requirement for diversity in motive power sources for the AF System is met.

The AF System is capable of supplying, but does not normally supply, feedwater to the steam generators during normal unit startup, shutdown, and hot standby conditions.

One pump at full flow is sufficient to remove decay heat and cool the unit to Residual Heat Removal CRHR) entry conditions.

There is also an AF Train A unit cross-tie downstream of the motor driven AF pump at each unit that is normally isolated. Use of the AF Train A unit cross-tie, however, is incompatible with an OPERABLE motor driven AF pump on either unit and its use is limited to an emergency response for a beyond design basis event that involves a total loss of secondary heat sink.

BRAIDWOOD - UNITS 1 &2 B 3.7.5 - 1 Revision 0

ATTACHMENT 4 Technical Specifications Bases Changes AF System No changes on this page B 3.7.5

- For Information Only -

BASES BACKGROUND (continued)

The AF System is designed to supply sufficient water to the steam generator(s ) to remove decay heat with steam generator pressure at the setpoint of the MSSVs. Subsequently, the AF System suppl ies sufficient water to cool the unit to RHR entry conditions, with steam released through the SG PORVs.

The AF System actuates automatical ly on low-2 steam generator water level, Safety Injection and Undervoltage (UV) on the Reactor Coolant Pump buses. The motor driven AF pump also actuates on an UV on bus 141(241).

The AF System is discussed in the UFSAR, Section 10.4.9 (Ref. 1).

APPLICABLE The AF System mitigates the consequences of any event with SAFETY ANALYSES loss of normal feedwater.

The design basis of the AF System is to supply water to the steam generator to remove decay heat and other residual heat by delivering at least the minimum required flow rate to the steam generators at pressures corresponding to the maximum steam pressure inside an intact steam generator during the long term cooling portion of the design basis accident (i .e., after steam line isolation occurs). This maximum steam pressure is 1250 psia (Ref. 2).

In addition, the AF System must supply enough makeup water to replace steam generator secondary inventory lost as the unit cools to MODE 4 conditions. Sufficient AF flow must also be available to account for flow losses such as pump recirculation and line breaks.

The limiting Design Basis Accidents (DBAs) and transients for the AF System are as follows:

a. Feedwater Line Break (FWLB); and

b. Loss of normal feedwater.

BRAIDWOOD - UNITS 1 &2 B3.7.5-2 Revision 0

ATTACHMENT 4 Technical Specifications Bases Changes AF System B 3.7.5 No changes on this page

- For Information Only -

BASES APPLICABLE SAFETY ANALYSES (continued)

In addition, the minimum available AF flow and system characteristics are serious considerations in the analysis of a small break Loss Of Coolant Accident (LOCA) and loss of offsite power (Ref. 3).

The AF System design is such that it can perform its function following an R~LB between the main feedwater isolation valves and containment, combined with a loss of offsite power following turbine trip, and a single active failure of one AF pump. The AF lines to the SGs are orificed such that sufficient flow is delivered to the non faulted SGs. Reactor tri pis assumed to occur I-Jhen the faulted SG reaches the lOW-low level setpoint. Sufficient flow would be delivered to the intact steam generators by the other AF pump.

During the loss of all AC power events, the Engineered Safety Feature Actuation System (ESFAS) automatically actuates the AF diesel driven pump and associated controls to ensure an adequate supply to the steam generators during loss of power. Valves which can be manually controlled are provided for each AF line to control the AF flow to each steam generator during loss of all AC pmver events.

The AF System satisfies the requirements of Criterion 3 of 10 CFR 50.36(c)(2)(ii).

BRAIDWOOD - UNITS 1 &2 B 3.7.5 - 3 Revision 82

ATTACHMENT 4 Technical Specifications Bases Changes AF System B 3.7.5 BASES LCO This LCD provides assurance that the AF System will perform its design safety function to mitigate the consequences of accidents that could result in overpressurization of the reactor coolant pressure boundary. Two independent AF pwips in two diverse trains are required to be OPERABLE to ensure the availability of RHR capability for all events accompanied by a loss of offsite power and a single failure.

This is accomplished by powering one of the pumps from the emergency buses. The second AF pump is powered by a different means, a diesel engine.

The AF System is configured into two trains. The AF System is considered OPERABLE when the c is and flow paths required to provide redundant AF flow to the steam generators OPERABLE. This requires that motor driven AF pump and the diesel driven AF pump be OPERABLE and capable of supplying AF to each steam generator. The associated piping, valves, instrumentation, and controls in the required flow paths to perform the safety related function are also required to be OPERABLE.

APPLICABILITY In MODES 1, 2, and 3, the AF System is required to be OPERABLE in the event that it is called upon to function when feedwater is lost, In MODE 4, 5, or 6, the steam generators are not normally used for heat removal, and the AF System Is not required.

The motor-driven AF pump is not OPERABLE if the AF Train A unit cross-tie is unisolated, (i.e., both isolation valves open). The use of the AF Train A unit cross-tie is for an emergency response to a total loss of secondary heat sink on the accident unit. Use of the AF Train A unit cross-tie results in a temporary relaxation of the single failure criterion for the non-accident unit, which, consistent with overall system reliability considerations, provides a limited time to support the emergency response on the accident unit, and return the AF Train A to an OPERABLE status. Otherwise, a plant shutdown is required. The diesel driven AF pump on the non -accident unit must be confirmed OPERABLE prior to use of the AF Train A unit cross-tie.

BRAIOW00p - UNITS 1 & 2 B 3.7.5 - 4 Revision 0

ATTACHMENT 5 FIGURE:

AUXILIARY FEEDWATER SYSTEM

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start , Fails Fail as-is a -i Limit break nO\now so0 DCBu~112 loeD I Balt~ty

& loeDI totalI 10 on a IOta loss of air to the valve that other S/Gs thai IGs will actuator. has own receiver to auto get sufficient ufficient now to 1B SX open on low air press remove decay heat AUXILIARY FEEDWATER