ML12342A388

Report of Changes, Tests and Experiments

ML12342A388
Person / Time
Site:	Byron
Issue date:	12/17/2012
From:	Tulon T Exelon Generation Co, Exelon Nuclear
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
1.10.0101, 2012-0132
Download: ML12342A388 (21)

Text

December 17, 2012 10 CFR 5059(d) (2)

LTR: Byron 2012-0132 FILE: 1.10.0101 United States Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Byron Station, Unit 1 and 2 Facility Operating License Nos. NPF -37 and NPF-66 NRC Docket Nos. STN 50- 454 and STN 50-455

Subject:

Report of Changes, Tests and Experiments Pursuant to the requirements of 10 CFR 50. 59, "Changes, tests and experiments,"

paragraph (d)(2), Byron Station is providing the required report for Facility Operating License Nos. NPF-37 and NPF-66. This report is provided for the 2010, 2011 and 2012 calendar years and consists of 50.59 Review Coversheets for changes to the facility or procedures as described in the Updated Final Analysis Report (UFSAR) and test or experiments not described in the UFSAR.

Please direct any questions regarding this submittal to David Gudger, Regulatory Assurance Manager, at (815) 406-2800.

Respectfully, Tiiothy J. Tulon Sib Vice President Byron Station TJH/KM/cy , Byron Station 10 CFR 50. 50 Report

Attachment 1 Byron Station 10 CFR 50.59 Report 10 CFR 50.59 Review Coversheets for Calendar Years 2010, 2011, and 2012

Design Change packages (DCP), Drawings Change Requests (DCR) Engineering Changes (EC) and Temporary Modifications (TMOD) 6G-12-002 Revisionl

50.59 REVIEW COVERSHEET FO LS-AA-104-1001 Revision 3 Page I of 3 Station /Unit(s): Byron Unit 1 Activity/Document Number: EC 380012 Revision Number: 2

Title:

STEAM GENERATOR MARGIN TO OVERFILL (SG MTO) PORV TRIM REPLACE MENT AND VALVE BLOCK INSTALLATION NOTE: For 50. 59 Evaluations, information on this form will provide the basis for preparing the biennial summary report submitted to the NRC in accordance with the requirements of 10 CFR 50.59(d)(2).

Description of Activity:

(Provide a brief, concise description of what the proposed activity involves.)

The proposed activity will install a new valve internal trim in the Steam Generator (SG) Power Operated Relief Valve (PORV) valves (1MS018A- D) and a valve block which will limit the flow to the current design basis flows consistent with the Steam Generator Tube Rupture (SGTR) Margin to Overfill (MTO) and Dose Case analyses.

This change will also increase the actuator setpoints which is required to support opening and closing the valve with the increase trim size. This change will also replace the snubber installed on the I D MS line downstream of the 1MS018D with a larger size to improve stress margins for the higher valve thrust loads with the modified actuator setpoints and increased trim size.

UFSAR Draft Revision Package 14-053, included with EC 380012 Revision 1, updates the maximum stresses listed in UFSAR Section 5.2.2.5.3 for the SG PORV piping associated with the Unit 1 valves.

Revision 1 to 50.59 Evaluation 6G-12-002 removes references to performing the modification work at power and modifies the response to Question 6 due to a physical difference in the expected valve configuration as described in IR 1417983 which necessitated a change in the design to limit valve stem travel.

Reason for Activity:

(Discuss why the proposed activity is being performed.)

Currently the MTO analysis has low margin. Recent valve capacity testing has determined that the existing SG PORV valves have a capacity lower than previously assumed thus further challenging the available margin.

To increase the margin to overfill, the station has elected to increase the capacity of the SG PORV's on Unit 1. The maximum flow rate of the modified valves will not exceed the flow assumed in the SGTR Dose Case analysis. This change will install a new trim in the SG PORV's which, in conjunction with a block to limit the stroke of the valve, will increase the valve capacity to be consistent with the current design basis flow. The setpoint change for the actuator is required to increase the actuator capability with the new trim installed. The support change on the ID MS line is required due the increased loads on the support as a result of the higher valve thrust loads with the modified actuator setpoints and increased trim size.

Effect of Activity:

(Discuss how the activity impacts plant operations, design bases, or safety analyses described in the UFSAR.)

During normal operation the SG PORV's are closed. In the event the pressure increases, the valves automatically begin opening in response to a pressure controller to maintain SG pressures below the safety valve setpoints if possible. The modified SG PORV's will essentially perform the same but the flow capacity will be limited by the valve block. The stroke time to open and close will decrease with the reduced stroke. The piping has been conservati vely analyzed for the maximum flow through the valve assuming the valve block is removed. This condition is bounding for the configuration with the valve block installed. During a SG tube rupture event, operator actions require the operators isolate the ruptured SG by closing the feedwater, auxiliary feedwater and main steam isolation valves and the affected SG PORV.

The intact SG PORV's are then used to cool down the RCS below the saturation temperature of the ruptured steam generator.

Once the cooldown is complete, the RCS is then depressurized to equalize pressure with the ruptured steam generator and terminate the primary to secondary leakage. Given the capacity will be consistent with the current design basis, there is no adverse impact on the SGTR Dose Case and margins will improve for the MTO Case. The operator action times associated with the SGTR event are not changed.

50.59 REVIEW COVERSHEET FORM LS-AA-104-1001 Revision 3 Page 2 of 3 Station/Unit(s): Byron Unit I Activity/Document Number: EC 380012 Revision Number:

Title:

STEAM GENERATOR MARGIN TO OVERFILL (SG MTO) PORV TRIM REPLACEMENT AND VALVE BLOCK INSTALLATION Summary of Conclusion for the Activity's 50.59 Review:

(Provide justification for the conclusion, including sufficient detail to recognize and understand the essential arguments leading to the conclusion. Provide more than a simple statement that a 50.59 Screening, 50.59 Evaluation, or a License Amendment Request, as applicable, is not required.)

1. The SG PORV (often described as a SG Relief Valve) valve control system is not changed, so the likelihood of a failure of the control system which could cause an inadvertent opening of the SG relief valve remains the same. The installation of the block does not contribute to failing the valve open since it does not directly operate the valve. The valve block will limit the travel of the valve. The valve is a flow over the seat. Therefore, pressure helps maintain this valve in the closed position. This would tend to decrease the frequency of occurrence of a SG PORV valve failing open. Section 15.6.3 discusses the Steam Generator Tube Rupture Event. The SG PORV's are credited in this analysis but are not the accident initiator. Therefore, the installation of the new trim, actuator setpoint change, support replacement, and valve block does not increase the frequency of occurrence of an accident previously evaluated in the UFSAR.

2. UFSAR Section 15.1.4 discusses the possible malfunction of the SG PORV (SG Relief) to open inadvertently. As discussed above the likelihood of this malfunction has not increased. UFSAR Sections 15.6.3 and 15.3.3 discuss the possibility of the valve to fail to reclose once opened in response to a transient. The valve is a flow over the seat, therefore flow will tend to close the valve. As a result, the required force from the actuator to close the valve will be decreased and decrease the likelihood of the malfunction. Various sections in Chapter 15 discuss that the valve opens in response to transients. The modified valve will require increased thrust to open the valve due to the increase in the trim size. The valve operator has been evaluated by the manufacturer and is capable of operating the valve with sufficient margin with the hydraulic actuator setpoint change described above. Furthermore, flow testing has demonstrated that the modified valve is capable of opening under maximum differential pressure conditions.

Therefore, the likelihood of a failure to open malfunction is not increased for the modified valve. In addition, the UFSAR sections often do not credit the SG PORV function but instead rely on the main steam safety valves to open to provide the heat removal function. Therefore, the installation of new trim, actuator setpoint change, snubber replacement, and valve block will not result in more than a minimal increase in the likelihood of occurrence of a malfunction of an SSC important to safety previously evaluated in the UFSAR.

3. Various accidents discuss the SG PORV as a potential release path that is evaluated as part of the dose calculations.

These accidents will not be affected since the new flow capacity of the valve is bounded by the flow assumed in the safety analyses. Also many of the accidents are not based on the flow rate from the valve but the primary to secondary allowable tube leakage which will not be affected by this modification. Failure of the valve block is not considered credible such that the valve would travel full open since the valve block design has significant margin to failure compared to the actuator stem. Therefore, if the actuator thrust were increased the valve stem would fail prior to the valve block failing which would cause the valve to go closed. Therefore, it is concluded that the installation of the new trim, actuator setpoint change, snubber replacement and valve block does not result in more than a minimal increase in the consequences of an accident previously evaluated in the UFSAR.

4. The SG tube rupture accident and Reactor Coolant Pump Shaft Seizure dose analysis assumes a single SG PORV sticks open. This is still considered a credible failure mode. The consequences of this accident are not affected by the new trim and valve block installation since the maximum flow rate from the valve will not be affected and failure of the blocking device is not considered credible. The actions to mitigate this malfunction are unchanged as an operator will be dispatched to close the SG PORV upstream block valve. Therefore, it is concluded that the installation of the new trim, actuator setpoint change, snubber replacement and valve block does not result in more than a minimal increase in the consequences of a malfunction of an SSC important to safety previously evaluated in the UFSAR.

5. The new valve trim and valve block do not affect accident initiation sequences or response scenarios as modeled in the safety analyses. No new operating configuration is being imposed by the proposed change. The actuator setpoint

50.59 REVIEW COVERSHEET FORM LS-AA- 104-1001 Revision 3 Page 3 of 3 Station/Unit(s): Byron Unit 1 Activity/Document Number: EC 380012 Revision Number: 2 I

Title:

STEAM GENERATOR MARGIN TO OVERFILL (SG MTO) PORV TRIM REPLACEMENT AND VALVE BLOCK INSTALLATION change will allow the valve to continue to operate the valve similar to the existing operation but with increased capability and within the design limits of the actuator. Therefore, it is concluded that the installation of the new trim, actuator setpoint change, snubber replacement, and valve block does not create a possibility for an accident of a different type than any previously evaluated in the UFSAR.

6. The current malfunctions evaluated are failure of the SG PORV to open and a failure of the SG PORV to close when required. The valve failing closed will have the same result. The valve block will be set such that the valve capacity will not exceed the current design basis flow rate. When the valve fails open, the result would be the same because the flow rate would not be affected. Failure of the valve block is not considered credible. This is because the device is passive, it is safety related, and the blocking device is a robust design and was demonstrated to be effective during testing. Therefore, the installation of the new trim, actuator setpoint change, snubber replacement, and valve block does not create a possibility for a malfunction of an SSC important to safety with a different result than any previously evaluated in UFSAR.

7. Installation of the new trim, actuator setpoint change, support modification and valve block installation does not affect the integrity of the fission product barriers utilized for mitigation of radiological dose consequences as a result of an accident. Plant response as modeled in the safety analyses is unaffected and no parameter which impacts a fission product barrier is changed. Hence, the mass and radioactivity releases used as input to the dose calculations are unchanged from those previously assumed. Therefore, it is concluded that the new trim, actuator setpoint change, snubber replacement, and valve block installation does not result in a design basis limit for a fission product barrier as described in the UFSAR being exceeded or altered.

8. The assumptions and methods used in the plant accident analyses are not affected by the new trim, actuator setpoint change, support modification and valve block installation. The current SG Tube Rupture accident analysis margin to overfill case assumes two SG PORV's are available for cooldown and depressurization of the RCS. This is not affected by this change. The dose case assumes one SG PORV fails open. This input is not changed nor is the evaluation methodology. The piping analysis performed for this design change uses the methodologies and criteria described in UFSAR Chapter 3.9. Therefore, it is concluded that the new trim, actuator setpoint change, snubber replacement and valve block installation does not result in a departure from a method of evaluation described in the UFSAR used in establishing the design bases or in the safety analyses.

Based on the above, the replacement of the SG PORV valve trim, actuator setpoint change and support modification does not result in an unreviewed safety question and maybe implemented without NRC approval.

Attachments:

Attach all 50.59 Review forms completed, as appropriate.

Forms Attached: (Check all that apply.)

q Applicability Review q 50.59 Screening 50.59 Screening No. Rev.

50.59 Evaluation 50.59 Evaluation No. 6G.12-002 Rev. 1

Procedure Revisions and Special Processes Procedures (SPP)

None listed

Technical Requirement Manual (TRM), Changes and Draft Revision Packages (DRP), Operability Evaluations 6G-10-003, Revision 0 6G-11-004 Revision 0 6G-12-005 Revision 0 6G-12-009 Revision 0

50.59 REVIEW COVERSHEET FORM LS-AA- 104-1001 Revision 3 Station/Unit(s): Byron/Unit 2 Page 1 of 4 Activity/Document Number: Technical Requirements Manual (TRM) Change 10-008, 1/2BOSR 9. a.1-1, 1/2BVSR 9.a.1-2, 1/2BOL 9. a Revision Numbers: NA, 5/7, 010, 517

Title:

TRM Change 10-008, TRM Section 3.9.a, "Decay Time" NOTE: For 50.59 Evaluations, information on this form will provide the basis for preparing the biennial summary report submitted to the NRC in accordance with the requirements of 10 CFR 50.59(d)

(2).

Description of Activity:

(Provide a brief, concise description of what the proposed activity involves.)

The following Technical Requirements Manual (TRM) changes will be made to revise the Incore Decay Time (ICDT) for future outages:

Byron TRM Section 3.9.a, "Decay Time," TLCO states, "The reactor shall be subcritical for? the last 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> (For B2R15, the reactor shall be subcritical for? the last 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, AND the point described by the number of assemblies offloaded and the time after shutdown shall be within the Acceptable Regions of Figure 3.9.a-1)."

This shall be changed to read:

"The reactor shall be subcritical for >_ the last 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br />; OR the reactor shall be subcritical for > the last 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> AND the point described by the number of assemblies offloaded and the time after shutdown shall be within the Acceptable Regions of the appropriate figure (Figure 3.9.a-1 through 3.9.a-9, as applicab le), based on cycle-specific SFP Background Heat Load Margin."

Byron TRM Section 3.9.a Condition A states, "Reactor subcritical for < 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> (For B2R 15: Reactor subcritical < 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, OR the point described by the number of assemblies offloaded and the time after shutdown not within the Acceptab le Regions of Figure 3.9.a-1.)"

This shall be changed to read:

"Reactor subcritical for < 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br />; OR Reactor subcritical < 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> OR the point described by the number of assemblies offloaded and the time after shutdown not within the Acceptable Regions of the appropriate figure (Figure 3.9.a-1 through 3.9.a-9)."

Byron TRM Surveillance TSR 3.9.a.1 states, "Verify the reactor subcritical >_ 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> by confirming the date and time of subcriticality. (For B2R15: Verify the reactor subcritical > 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> by confirming the date and time of subcriticality AND verify the point described by the number of assemblies offloaded and the time after shutdown is within the Acceptab le Regions of Figure 3.9.a-1.)"

This shall be changed to read:

"Verify the reactor subcritical ? 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> by confirming the date and time of subcriticality; OR Verify the reactor subcriticai > 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> by confirming the date and time of subcriticality AND verify the point described by the number of assemblies offloaded and the time after shutdown is within the Acceptable Regions of the appropriate figure (Figure 3.9.a-l through 3.9.a-9, as applicable), based on cycle-specific SFP Backgrou nd Heat Load Margin determined by TSR 3.9.a.2."

Byron TRM Surveillance TSR 3.9.1 .a Frequency states, "Prior to initial movement of irradiated fuel in the reactor vessel each outage (For B2R15: Verify subcriticality prior to initial movement of irradiated fuel in the reactor vessel AND verify the point described by the number of assemblies offloaded and the time after shutdown is within the Acceptable Regions of Figure 3.9.a-1 prior to transporting each fuel assembly through the transfer tube out of containment.)"

This shall be changed to read:

"Prior to initial movement of irradiated fuel in the reactor vessel each outage.

If cycle-specific SFP Background Heat Load Margin is used, verify subcriticality prior to initial movement of irradiated fuel in the reactor vessel AND verify the point described by the number of assemblies offloaded and the time after shutdow n is within the Acceptable Regions of

50.59 REVIEW COVERSHEET FORM LS-AA-104-1001 Revision 3 Station/Unit(s): Byron/Unit 2 Page 2 of 4 Activity/Document Number: Technical Requirements Manual (TRM) Change 10-008, 1/2BOSR 9.a.1-1, 1/2BVSR 9.a. 1-2, 1/2BOL 9.a Revision Numbers: NA, 5/7, 0/0, 5/7

Title:

TRM Change 10-008, TRM Section 3.9.a, "Deca y Time" the appropriate figure (Figure 3.9.a-1 through 3.9.a-9, as applicable), based on cycle-specific SFP Background Load Margin prior to transporting each fuel assembly Heat through the transfer tube out of containment."

Figure 3.9.a-1, "B2R15 Incore Decay Time Requireme nt" will be replaced with 9 new figures (Figures 3.9.a:

for SFP Background Heat Load Margins of 41 MBTU/hr 1 through 3.9.a-9),

through 49 MBTU/hr respectively.

New TSR 3.9.a.2 will be added that will read:

"Determine the Spent Fuel Pool Background Heat Load Margin for the specific cycle for use in TSR 3.9.a. 1.

margin down to the next MBTU/hr value to determine Round the the applicable SFP Background Heat Load Margin Figur 3.9.a-1 through 3.9.a-9)." e (Figure This TSR will calculate the cycle-specific SFP Backg round Heat Load Margin and thereby determine which Figures is to be used for the specific outage. of the 9 new A Note will proceed this TSR stating:

"Surveillance only required to be performed if cycle-specific SFP Background Heat Load figures are used 3.9.a.1." in TSR If it is desired to simply use the 100 hour0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> ICDT value

, then the surveillance to determine the SFP Background is not necessary. Heat Load Margin The frequency of the new TSR 3.9.a.2 will state:

"Prior to movement of irradiated fuel in the reactor vessel each outage."

Current Technical Requirements Manual Surveillanc e procedures 1/2BOSR 9.a.1-1, and LCOAR procedures revised accordingly. New TRM Surveillance procedures 1/2BOL 9.a will be 1/2BVSR 9.a. 1-2 will be written to calculate the cycle Background Heat Load Margin. -specific SFP Reason for Activity:

(Discuss why the proposed activity is being performed.

)

It is anticipated that during futures outages, work activi ties will be completed and the required plant configuratio established to support commencing movement of irradi n will be ated fuel from the reactor vessel to the Spent Fuel Pool current requirement of 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> after reactor shutdown. (SFP) prior to the The current SFP cooling analysis assumes that fuel transf after 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> decay time in the reactor core. This evalu er begins ation is being performed to determine if the proposed made under the provisions of 10 CFR 50.59. changes can be The proposed activity does not address the radiologica l consequences of a Fuel Handling Accident (FHA).

consequences of a FHA were revised under the Powe The radiological r Uprate (PUR) program using an ICDT of 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.

applicable to the FHA did not require review under 50.59 The changes

, as it was reviewed and approved by the NRC in NRC May 4, 2001 to Oliver D. Kingsley (Exelon),

Subject:

Letter dated Issuance of Amendments; Increase in Reactor Power, 1 and 2, and Braidwood Station, Units 1 and 2. Subsequent Byron Station, Units ly, the radiological consequences of a FHA were revise the implementation of the Alternate Source Term (AST d again under

) project. The FHA is now evaluated for an ICDT of as incorporated into the UFSAR under DRP 11-075. less than 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, However, this activity will still prevent fuel moves prior other controls and evaluations would be required to be to 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, since implemented if moves prior to 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> were desired.

Effect of Activity:

(Discuss how the activity impacts plant operations, design bases, or safety analyses described in the UFSAR.)

The proposed activity will allow starting fuel offloading activities as early as 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />. If offload begins this early, allowed maximum transfer rate of 8 assemblies per hour the UFSAR is acceptable, however the decay heat load margin limit will ultimately

50.59 REVIEW COVERSHEET FORM LS-AA- 104-1001 Revision 3 Station/Unit(s): Byron/Unit 2 Page 3 of 4 Activity/Document Number: Technical Requirements Manual (TRM)

Change 10-008, 1/2BOSR 9.a.1-1, 1/2BVSR 9.a.1-2, l/2BOL 9. a Revision Numbers: NA, 5/7,0/0,5/7

Title:

TRM Change 10-008, TRM Section 3.9.a, "Decay Time" restrict the offload rate as the number of assemblies offloaded increas es. 'Occupational dose on the refueling machine may increase slightly.

Summary of Conclusion for the Activity' s 50.59 Review:

(Provide justification for the conclusion, including sufficient detail to recognize and understand the essential arguments leading to the conclusion. Provide more than a simple statement that a 50.59 Screening, 50.59 Evaluation, or a License Amendment Request, as applicable, is not required.)

The proposed activity may be implemented without prior NRC review and approval based upon the following:

Changing the ICDT from 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> to 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, with the total numbe r of fuel assemblies offloaded based on total heat load and time after shutdown, does not change the frequency of an accident because the proposed change does not increase the failure rate of refueling equipment or increase the risk of a FHA due to human error. Spent fuel handling tools will not change, nor will the method/procedures used for handling spent fuel assemblies. The total number of fuel assemblies to be transferred remains the same. There is no effect on the failure probabilities of the SFP cooling system.

Calculation BYRI0-136 has been completed to evaluate the impact of changing the ICDT. This calculation determines the number of fuel assemblies that may be placed in the SFP based on the margin in the background decay heat load, as a function time after shutdown, which are provided as figures in the TRM. of The results of the calculation verified the calculated SFP Bulk Water Temperature, Maximum Local Water Temperature, Maxim um Cladding Temperature and Maximum Clad Heat Flux (DNB) remain acceptable. I/2BVSR 9.a.1-2 calculates the actual cycle-specific background heat load margin using either a CaskLoader analysis or a manual calculation, both of which are based on NRC Branch Technical Position ASB 9-2, Rev. 2, Residual Decay Energy For Light-Water Reactors For Long-Term Cooling, which is the currently licensed Byron methodology.

The design basis SFP criticality analysis (for the SFP re-rack project

) assumes a bulk pool water temperature of 4°C (39°F). The proposed change would potentially increase the temper ature of the water in the SFP, thus adding negative reactivity. The SFP criticality analysis is therefore not adversely affected.

There are no offsite dose consequences impacted by this change

. The ICDT associated with radiological concerns (dropped fuel assembly) has been reduced to 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> under the PUR program which has been reviewed and approved by the NRC and subsequently reduced to < 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> (provided the FHB ventilation system is operable) per the AST project as incorporated into the UFSAR per DRP 11 -075. However, this activity will still preven t fuel moves prior to 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, since other controls and evaluations would be required to be implemented if moves prior to 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> were desired.

Beginning core alteration and fuel transfer operation as early as 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, with the total number of fuel assemblies offloaded Lased on total heat load and time after shutdown, is not expected to significantly increase the occupational dose. UFSAR Tables 12.3-1 and 12.3-2 divide areas in the plant into radiation zones. The design dose rate for each zone is selected to ensure that the exposure limit of 10 CFR 20 is not exceeded. Shielding is established based on ALARA to minimize the dose rate for the selected areas. The areas affected by the defueling operation are designated as High Radiation areas (Zone III) with a design dose rate of >I 00 mrem/hr. Access to these areas is controlled in accordance with station procedures and radiation work permits. Electronic dosimeters are required to continuously monitor the dose rate in the areas in order to limit personnel exposure to below 10 CFR 20 limits.

These existing controls are not affected.

The fission product barriers potentially affected by this change are the fuel clad and the reactor containment. This change does not result in a change to the internal containment pressu re that would represent a challenge to the containment design basis limit of 50 psig. Calculation BYRIO-136 determined reducing the ICDT to >_ 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, with the total number of fuel assemblies offloaded based on total heat load and time after shutdown, would not exceed the currently assumed SFP temperature limits, nor will it increase the maximum calculated local water temperature, the maximum calculated cladding temperature, and the actual maxim um clad heat flux (DNB) beyond acceptable values.

Therefore, no design basis limit for a fission product barrier is being exceeded or altered.

50.59 REVIEW COVERSHEET FORM LS-AA-104-1001 Revision 3 Station /Unit(s): Byron/Unit 2 Page 4 of 4 Activity/Document Number. Technical Requirements Manua l (TRM) Change 10-008, 1/2BOSR 9.a.1-1, 1/2BVSR 9.a.1-2, 1/2BOL 9. a Revision Numbers: NA, 5/7, 0/0, 5/7

Title:

TRM Change 10-008, TRM Section 3.9.a, "Decay Time" The changes made by this activity represent changes in input parame ters to the design basis analysis. Decay heat input to the SFP was calculated for the earlier ICDT using the method describ ed in NRC Branch Technical Position ASB 9-2. This is the same method used in the existing analysis. Therefore, this activity does not change the method of evaluation described in the UFSAR or in the Safety Evaluation for the PUR project.

Since all the Evaluation questions have been answered "No," this activity does not require NRC review and approval. This activity may be implemented per the applicable governing proced ures.

Attachments:

Attach all 50.59 Review forms completed, as appropriate.

Forms Attached: (Check all that apply.)

Applicability Review 50.59 Screening 50.59 Screening No. Rev.

50.59 Evaluation 50.59 Evaluation No. 6G-10-003 Rev. 0

50.59 REVIEW COVERSHEET FORM LS-AA-104-1001 Revision 3 Page I of 3 Station/Unit(s): Boron/ Unit 1 and 2 Activity/Document Number: EC 385829 and DRP #14-040 Revision Number: 0/0

Title:

Tornado Missile Design Basis for the Essential Service Water Cooling Towers NOTE: For 50 .59 Evaluations, information on this form will provide the basis for preparing the biennial summary report submitted to the NRC in accordance with the requirements of 10 CFR 50.59(d)(2).

Description of Activity:

(Provide a brief, concise description of what the proposed activity involves.)

EC 385829 revises the design basis associated with the UHS capability of mitigating the effects of tornado missiles. The proposed activity also revises the UFSAR described tornado design bases for the Essential Service Water Cooling Towers (SXCTs). UFSAR Sections 3.5.2, 3.5.4.1, 9.2.5.3.1, and 9.2.5.3.2 are revised to provide a description of the modified design bases and a summary of the analysis of SXCT performance with the loss of SXCT fans due to postulated tornado missiles.

Reason for Activity:

(Discuss why the proposed activity is being performed.)

New analysis has been performed to evaluate Ultimate Heat Sink (UHS) performance associated with a tornado event including loss of SXCT fans due to a postulated tornado missile. In an 11/14/2007 inspection report the NRC had identified a Non-Cited Violation involving the UHS capability of mitigating the effects of tornado missiles.

Effect of Activity:

(Discuss how the activity impacts plant operations, design bases, or safety analyses described in the UFSAR.)

UFSAR Sections 3.5.4.1 and 9.2.5.3.2 describe an analysis of SXCT capacity for a design basis tornado event assuming no fans are available. The UFSAR described analysis was based on the following inputs:

n Wet Bulb Temperature = 78° F n Number of SXCT cells with water flow = 8 n Fan Operation = All SXCT fans are assumed to be inoperable due to impact of tornado missiles n Plant Cooling Load (normal shutdown cooling loads for both units plus diesel cooling loads) = 150 x 106 Btu/hr n Initial Service Water Temperature entering the plant = 91° F As described in IR 295141, the analysis was not updated to reflect the changes to heat load made during the 1991design basis reconstitution. The heat load was also not updated for the replacement of the Unit 1 steam generators or for power uprate.

Additionally the analysis did not consider that SXCT cells may be out of service with water flow isolated to the out of service cells.

The new analysis is based on a revised design bases with a minimum of two SXCT fans available for cooling. Two fans are assumed to be initially out of service, two fans are assumed to be inoperable due to the impact of a tornado missile, and a single failure (passive electrical failure) is assumed to result in the loss of power to two additional fans. The analysis assumes operator action will be taken to delay RH initiation until an adequate number of SXCT fans are available for shutdown cooling. RH initiation for the two units may need to be offset and operator action may be needed to adjust the cool down rates of each unit to match the heat input to the capacity of the SXCT.

Summary of Conclusion for the Activity's 50.59 Review:

(Provide justification for the conclusion, including sufficient detail to recognize and understand the essential arguments leading to the conclusion. Provide more than a simple statement that a 50.59 Screening, 50.59 Evaluation, or a License Amendment Request, as applicable, is not required.)

The proposed revision to the design bases associated with the UHS capability of mitigating the effects of tornado missiles does not introduce a change in the frequency of a tornado event or the frequency of tornado generated missiles impacting the Essential Service Water Cooling Towers (SXCTs). The UFSAR Section 3.3.2 description of tornado design parameters and the UFSAR Section 3.5.1.4 description of tornado generated missiles are unchanged. The proposed change does not result in more than a

50.59 REVIEW COVERSHEET FORM LS-AA-104-1001 Revision 3 Page 2 of 3 Station/Unit(s): Byron! Unit 1 and 2 Activity/Document Number: EC 385829 and DRP #14-040 Revision Number: 0/0

Title:

Tornado Missile Design Basis for the Essential Service Water Cooling Towers minimal increase in the likelihood of occurrence of a malfunction of an SSC important to safety because; 1) The change meets the applicable regulatory requirements and design bases requirements, 2) The calculated peak SX supply water temperature is below the UFSAR described maximum allowed temperature and 3) The required operator actions to support the analysis are reflected in plant procedures and operator training program. There is no specific time requirement to perform the action and recovery can be made from credible errors.

The proposed revision to the design bases associated with the UHS capability of mitigating the effects of tornado missiles does not adversely impact the capability to safely shutdown both units in the event of a design basis tornado. The revised transient temperature analysis for the UHS indicates that with a minimum of two SXCT fans available that the maximum SX supply temperature is 105.9 ° F. As indicated in UFSAR Section 9 .21.2.1, all safety-related heat transfer equipment is designed for a 100°F essential service water inlet temperature. As currently indicated in UFSAR Sections 3.5.4.1 and 9.2.5.3.2 in the event of tornado missiles SX supply temperature may be as high as 110° F and although this exceeds the normal maximum temperature of 100° F, no adverse impact on safety will result. This is supported by the inherent design margin in the SX cooled components required for cool down operation.

The proposed revision to the design bases associated with the UHS capability of mitigating the effects of tornado missiles does not adversely impact the capability to safely shutdown both units in the event of a design basis tornado. The analysis conservatively assumes that two SXCT fan motors could be disabled and inoperable from a tornado missile. Additionally a single failure (passive electrical failure) is assumed to result in the loss of power to two additional fans. The revised transient temperature analysis for the UHS indicates that with a minimum of two SXCT fans available that the maximum SX supply temperature remains below the current UFSAR described allowed SX temperature for a tornado event.

The proposed revision to the design bases associated with the UHS capability of mitigating the effects of tornado missiles does not introduce the possibility of a new accident because the design bases change is not the initiator of any accident and no new failure modes are introduced.

UFSAR Sections 3.5.4.1 and 9.2.5.3.2 currently describe that an analysis of the cooling tower capacity for postulated multiple tornado missiles was made conservatively assuming no operating fans (i.e. all SXCT fans inoperable due to tornado missiles).

Based upon a review of Regulatory Guide (RG) 1.27, RG 1.76, Standard Review Plan (SRP) 3.3.2, SRP 3.5.1.4, and SRP 3.5.3 multiple missile impacts on an SSC are never discussed, and since RG 1.27 states "the complex (but not necessarily its individual features) must be capable of withstanding each of the most severe natural phenomena expected," it is concluded that the SXCT need only withstand the impact of a single tornado missile. The proposed revision to the design bases associated with the UHS capability of mitigating the effects of tornado missiles considers two separate malfunctions. The first malfunction considered is a failure of two SXCT fans due to a tornado generated missile strike on the UHS cooling tower complex. The second malfunction considered is the loss of power to two SXCT fans do to a postulated passive electrical failure. In both the current UFSAR described analysis and the revised analysis the result of the missile strike is the loss of one or two of the forced draft cooling fan(s) providing air flow thru a tower cell (a single tornado generated missile impact that comes in tangent to the fan stacks in line with the motors could potentially impact the power cables for two fan motors). Thus the result of the malfunction due to a postulated missile strike is unchanged. For the limiting condition with 2 fans initially out of service and the outside air wet bulb temperature of 78° F, the 6 OPERABLE fan riser valves would be open and a postulated passive electrical failure would result in the impacted cells only providing natural draft cooling. For less limiting conditions (lower outside air wet bulb temperatures or at least 7 SXCT fans initially OPERABLE) a passive electrical failure could result in a failed closed riser valve. Having an additional operable SXCT cell or improved tower performance with a lower outside air wet bulb temperature would more than offset the loss of natural draft cooling from a cell with a closed riser valve. Thus the proposed change does not create a possibility for a malfunction of a SSC important to safety with a different result than any previously evaluated in the UFSAR.

The proposed revision to the design bases associated with the UHS capability of mitigating the effects of tornado missiles does not adversely impact the capability to safely shutdown both units in the event of a design basis tornado. The revised transient temperature analysis for the UHS indicates that the maximum SX supply temperature remains below the UFSAR described allowed SX temperature for a tornado event. Since the UHS temperature remains acceptable there is no indirect impact on any fission product barrier. Therefore the change does not result in a DBLFB as described in the UFSAR being exceeded or altered.

50.59 REVIEW COVERSHEET FORM LS-AA-104-1001 Revision 3 Page 3 of 3 Station/Unit(s): Boron/ Unit 1 and 2 Activity/Document Number: EC 385829 and DRP #14-040 Revision Number: 0/0

Title:

Tornado Missile Design Basis for the Essential Service Water Cooling Towers The proposed revision to the design bases associated with the UHS capability of mitigating the effects of tornado missiles does not result in a departure from a method of evaluation described in the UFSAR. UFSAR Sections 3.5.4.1 and 9.2.5.3.2 describe the analysis and inputs used in the analysis but the specific method of analysis is not described. The revised transient temperature analysis for the UHS was performed using service water cooling tower performance curves generated using the method described in UFSAR Section 9.2.5.3.1.1.2 and the time dependent two cooling tower model described in UFSAR Section 9.2.5.3.1.1.3. These methods have been previously accepted by the NRC for calculating UHS temperatures at Byron. For calculating the maximum SX supply water temperature for a postulated tornado event the method of evaluation was unchanged; only the input parameters were changed. Thus the proposed activity does not result in a departure from a method of evaluation described in the UFSAR used in establishing the design bases.

Attachments:

Attach all 50.59 Review forms completed, as appropriate.

Forms Attached: (Check all that apply.)

q Applicability Review q 50.59 Screening 50.59 Screening No. Rev.

0 50.59 Evaluation 50.59 Evaluation No. 6G-11-004 Rev. 0

50.59 REVIEW COVERSHEET FORM LS-AA-104-1001 Revision 3 Page 1 of 3 Station/Unit(s): Byron / 1&2 Activity/Document Number: EC 387444, DRP 14-079 Revision Number: 000, N/A

Title:

Replace SFPBC Bridge Drive Gearboxes NOTE: For 50.59 Evaluations, information on this form will provide the basis for preparing the biennial summary report submitted to the NRC in accordance with the requirements of 10 CFR 50.59(d)(2).

Description of Activity:

(Provide a brief, concise description of what the proposed activity involves.)

This activity will replace the Spent Fuel Bridge Crane (SFPBC) gearboxes to allow for an increase in the SFPBC bridge speed to a maximum value of 40 fpm. In conjunction with the replacement of the gearboxes, if required, the SFPBC motor controls will be adjusted to limit the speed of the SFPBC bridge speed to a maximum of 40 fpm.

DRP 14-079 will modify UFSAR Section 9.1.4.2.2 to specify the Byron SFPBC bridge speed is ranged from approximately O fpm to 40 fpm (in lieu of the current 30 fpm).

Reason for Activity:

(Discuss why the proposed activity is being performed.)

Recently, the SFPBC control system was upgraded (Reference EC 375823) inclusive of installation of a new programmable logic controller (PLC) and replacement of the bridge drive motor. Standard CMAA-70, referenced in the UFSAR, indicates that a maximum bridge speed of 50 fpm would be acceptable for a crane design consistent with the SFPBC. This activity will take advantage of the aforementioned SFPBC control system upgrades and UFSAR-referenced design standard limitations on bridge speed to safely enhance fuel handling performance by reducing the transit time for various fuel handling activities.

Effect of Activity:

(Discuss how the activity impacts plant operations, design bases, or safety analyses described in the UFSAR.)

As indicated above, the proposed change will enhance fuel handling performance by reducing the transit time for various fuel handling activities. The structural integrity of the fuel assemblies will not be compromised by increasing the SFPBC bridge speed to 40 fpm. Note also, as discussed below, the fuel handling accident scenario conservatively assumes total failure of the fuel assembly. Therefore, increasing the SFPBC bridge speed during fuel handling activities cannot result in additional structural damage to the fuel assemblies - than what is already assumed in the UFSAR. Since the SFPBC gearbox and resulting maximum bridge speed increase from 30 fpm to 40 fpm has no affect on the function of the systems supporting postulated fuel handling accidents, since bridge speed is not utilized as an input to the fuel handling accident analysis, and since increasing the bridge speed cannot not result in any additional damage to the current total failure of the ficel assembly already assumed in the accident analyses, the increase in bridge speed to 40 fpm will not affect the results of a fuel handling accident in the Fuel Handling Building.

As documented in EC 387444, PaR Nuclear performed an assessment documenting that the SFPBC electrical systems were capable of supporting a bridge speed of 40 fpm. Therefore, increasing the bridge speed to 40 fpm will not adversely impact the UFSAR described functions of the variable speed controllers and installed electronics. Braidwood calculation 060312(CMED)

(which was determined to be applicable to Byron via calculation 8.1.9-BYROI -098) evaluated the structural integrity of the bridge crane inclusive of an evaluation for overturning potential for a bridge crane speed of 40 fpm. The results of this assessment concluded that the bridge crane was structurally adequate and that overturning would not occur at a 40 fpm bridge speed. The proposed SFPBC bridge speed of 40 fpm is less that the suggested "slow operation" speed of 50 fpm, suggested in standard CMAA Specification #70 criteria defined as applicable in the UFSAR. This activity which replaces the SFPBC bridge gear case (and potential adjustment of the SFPBC motor control speed setting) resulting in an increase in bridge speed does not change any of the current SFPBC safetyfeatures, control features or interlocks identified in UFSAR Section 9.1.4.3.1 to ensure safe fuel handling.

50.59 REVIEW COVERSHEET FORM LS-AA-104-1001 Revision 3 Page 2 of 3 Station/Unit(s): Byron / 1&2 Activity/Document Number: EC 387444, DRP 14-079 Revision Number: 000, N/A

Title:

Replace SFPBC Bridge Drive Gearboxes Summary of Conclusion for the Activity's 50.59 Review:

(Provide justification for the conclusion, including sufficient detail to recognize and understand the essential arguments leading to the conclusion. Provide more than a simple statement that a 50.59 Screening, 50.59 Evaluation, or a License Amendment Request, as applicable, is not required.)

The UFSAR does not explicitly identify any design limitations on the SFPBC relating to the maximum speed of the bridge movement. Through UFSAR reference to CMAA Specification #70, the suggested speed of the SFPBC bridge shall be maximized at 50 fpm, which is greater than the proposed speed of 40 fpm. The higher proposed bridge speed will not result in any additional fuel assembly damage than is already assumed in fuel handling accident analyses and will not adversely affect any supporting systems credited in the fuel handling accident scenarios. Fuel accident analyses do not utilize bridge speed as a design input and calculations have determined that the increase in bridge speed will not adversely impact the structural integrity of the SFPBC. Therefore, the proposed activity does not involve a change to an SSC that adversely affects an UFSAR described design function.

This activity does not change any of the current SFPBC safety features, control features or interlocks identified in the UFSAR to ensure safe fuel handling. The existing SFPBC controls will continue to provide interlocks and control function to ensure that the bridge speed is restricted when specified criteria are met. Design standards (CMAA Specification #70) referenced in the UFSAR Section 9.1.4.3.1 suggest a maximum operating bridge speed of 50 fpm which is greater that the proposed bridge speed of 40 fpm. Therefore, replacement of the SFPBC gearbox and resulting increase in the maximum bridge speed from 30 fpm to 40 fpm will not adversely affect how UFSAR described design functions are performed or controlled.

UFSAR Section 4.2.1.5 describes how the integrity of the fuel assemblies was assured by assessment of stresses and deformation due to various non-operational, operational and accident loads. However, none of the methods described for the evaluation of the fuel assemblies utilizes the SFPBC bridge speed as an input to the method of evaluation. The higher proposed bridge speed will not result in any additional fuel assembly damage than is already assumed in fuel handling accident analyses. Therefore, increasing the SFPBC speed will not adversely impact the aforementioned UFSAR described methodology to assess the structural integrity of the fuel assemblies. Methods of analysis to assess the structural integrity of the SFPBC are not explicitly described in the UFSAR, therefore, the Braidwood calculation 060312(CMED) which evaluated the structural integrity of the bridge crane inclusive of an evaluation for overturning potential for a bridge crane speed of 40 fpm [which is conservative, since this speed exceeds the bridge bumper evaluation speed recommended in UFSAR-referenced OSHA Standard 1910.1 79(e)(2)(i)(a) requirements] does not involve an adverse change or alternate evaluation methodology to that described in the UFSAR. In addition, methods of analysis to assess the SFPBC electrical systems are not explicitly described in the UFSAR, therefore, analyses performed by PAR documented in EC 387444 which document that the SFPBC electrical systems are capable of supporting a 40 fpm bridge speed do not involve an adverse change or alternate evaluation methodology to that described in the UFSAR. As such, this activity does not involve an adverse change to an element of a UFSAR described evaluation methodology, or use of an alternative evaluation methodology, that is used in establishing the design bases or used in the safety analyses.

The existing SFPBC interlocks and controls will continue to monitor the motor for overload or abnormal operating conditions.

The increased bridge speed is within the electrical and mechanical design limitations of the equipment as documented in EC 387444, therefore, the equipment will not be operated beyond the design bounds for the equipment design as described in the UFSAR. Therefore, this activity does not involve a test or experiment not described in the UFSAR, where an SSC is utilized or controlled in a manner that is outside the reference bounds of the design for that SSC.

Based on a review of the Technical Specifications and Technical Requirements Manual (TRM) Sections related to Refueling, Spent Fuel, Fuel Handling Building, Spent Fuel Pool and supporting systems, it was determined that equipment protective devices and setpoints related to SFPBC bridge speed are not explicitly or implicitly addressed in the Technical Specifications, TRM or in the Operating License. Therefore, the Technical Specifications and Operating License requirements have not been affected by the proposed change.

50.59 REVIEW COVERSHEET FORM LS-AA-104-1001 Revision 3 Page 3 of 3 Station/Unit(s): Byron / 1&2 Activity/Document Number: EC 387444, DRP 14-079 Revision Number: 000. N/A

Title:

Replace SFPBC Bridge Drive Gearboxes Attachments:

Attach all 50.59 Review forms completed, as appropriate.

Forms Attached: (Check all that apply.)

Applicability Review M 50.59 Screening 50.59 Screening No. 6E-12-117 Rev. 0 50.59 Evaluation 50.59 Evaluation No. Rev.

50.59 REVIEW COVERSHEET FORM LS-AA-1 04 -1001 Revision 3 Pace I of 3 Station /Unit(s): Braidwood, Byron /Unit I and 2 Activity/Document Number: Abnormal Component Position Sheet for Closing the Isolation Valves to the MSIV Hydraulic Accumulator Manifolds' Relief Valves Revision Number: N/A

Title:

Close the Isolation Valves for the Relief Valves on each of the MSIV Hydraulic Accumulator Manifolds NOTE: For 50.59 Evaluations, information on this form will provide the basis for preparing the biennial summary report submitted to the NRC in accordance with the requirements of 10 CFR 50.59(d)(2).

Description of Activity:

(Provide a brief, concise description of what the proposed activity involves.)

Change the normal position of the relief valve isolation valves (1/2MSOOIAJB/C/D-HJHI) on the hydraulic accumulator manifolds assembly for the Main Steam Isolation Valves (MSIVs). The normal position of the relief valves isolation valves is changed from "Open" to "Closed". The normal position of the l/2MSO01A/B/C/D-H/H1 valves is indicated on drawing #D-10239 for Braidwood, M-152 Sh. 57 for Byron and on the PASSPORT D031 panels for the valves.

Reason for Activity:

(Discuss why the proposed activity is being performed.)

The screening criteria in procedure OP-AA-108- 101, "Control of Equipment and System Status",

Attachment 2, "Abnormal Component Position Sheet (ACPS)", is applied to activities that result in abnormal positioning of plant components. Review in accordance with this procedure applies to the activity to close the 1

/2MSO01AJB/C/D-HJHI relief valves isolation valves. A review of the Attachment 2 responses indicates that a IOCFR50.59 review is required for this activity.

A condition was recently noted whereby, during a steamline break accident, the MSIV accumula tors could experience a higher temperature than previously considered. Changes in ambient temperature will cause the accumulators' temperature to rise which causes the accumulators' hydraulic and pneumatic pressures to rise above the pressure range of 4900 to 5100 psig normally maintained in the accumulator. At approximately 5400 psig in the accumulator, the.

hydraulic manifold relief valve actuates and bleeds hydraulic oil back to the reservoir. If an excessive amount of hydraulic oil is relieved to the reservoir as MSIV room temperatures rise, there may not be a sufficient inventory of hydraulic oil left to fully close the MSIV regardless of the accumulator pressure when demande d to respond to a Main Steam Line Break event (Reference IR #01409899 for Byron and IR #01409900 for Braidwoo d).

Effect of Activity:

(Discuss how the activity impacts plant operations. design bases, or safety analyses described in the UFSAR.)

The relief valves were added to relieve increased accumulator pressure caused by fluctuatio ns in ambient temperatures during normal plant operation. Prior to the installation of the relief valves (and relief valves bypass valves) it was necessary to reduce accumulator pressure below administrative limits by performing a partial stroke of the MSIVs. The frequency of the partial stroke evolutions challenged actuator elastomers and potentially accelerate d wear of the equipment thus reducing component reliability. The isolation of the relief device will result in the need for manual action to reduce accumulator pressure below the administrative limit of 5400 psi, but a partial stroke of the MSIVs will not be needed. Steps to take this action are included in procedure BwOP MS-S Rev. 28 for Braidwoo d and BOP MS-S Rev. 17 for Byron. The supporting documentation for the addition of the hydraulic relief valves (Braidwo od EC # 41550. Byron ECs 77503 & 77504) shows that the potential for relief valves failure was recognized; thus, the modification included a manual isolation valve to maintain the MSIV safety function while including a relief valve bypass valve to provide a manual method to control /limit actuator pressure.

Isolating the hydraulic manifold relief valves will not impact the assumptions made in the safety analyses. The action to isolate the hydraulic relief valves will ensure that the hydraulic accumulators retain the oil inventory that is required to close the MSIVs when demanded. The MSIVs will thus be able to close as demanded on a Main Steamline Isolation signal generated automatically or manually.

As discussed in the Byron SER (Supplement 5) and the Braidwood SER (Supplement 2), the MSIV's accumulator was tested to a atinimum internal pressure of 7500 psi. The NRC was concerned about the valve accumulators exceeding their proof pressure under high temperature ambient accident conditions. The MSIV actuator was successfully tested, under simulated accident conditions, as part of the Environmental Qualification activities without exceeding the

50.59 REVIEW COVERSHEET FORM Ls-AA- 104-1001 Revision 3 tation/Unit(s): Braidwood, Byron/Unit 1 and 2 Page2of3 Activity/Document Number: Abnormal Component Position Sheet for Closing the Isolation Valves to the MSIV H draulic Accumulator Manifolds' Relief Valves Revision Number: N/A

Title:

Close the Isolation Valves for the Relief Valves on each of the MSIV Hydraulic Accumulator Manifolds accumulator internal pressure limit (The maximum accumulator pressure reported in EQ Binder EQ-BB-024 is 5,800 psig). The tested configuration did not include the hydraulic relief valves.

The hydraulic relief valves were not added to provide overpressure protection. The relief valves were added by Byron/Braidwood Station modifications to relieve increased accumu lator pressure caused by fluctuations in ambient temperatures during normal plant operation. Prior to the installa tion of the relief valves (and relief valve bypass valves) it was necessary to reduce accumulator pressure by performing a partial stroke surveillance of the MSIVs. The relief valves were not installed to limit the accumulators below their maxim um allowable working pressure or to satisfy a specific ASME code design requirement. The relief valves provide a pressure controlling or limiting function to preserve the long term reliability of elastomers in the actuator assembly. The design includes a 0.013 inch orifice to limit the hydraulic volume relieved from the system such that it is within the capacity of the system hydraulic pump to maintain pressure in the event of relief valve leakage or failing open. Theref ore, the design of the relief valve was intended as an operational convenience to preserve the long term reliability of the MSIVs without creating a potential loss of function due to relief valve failure.

Summary of Conclusion for the Activity' s 50.59 Review:

(Provide justification for the conclusion, including sufficient detail to recognize and understand the essential arguments leading to the conclusion. Provide more than a simple statement that a 50.59 Screeni ng, 50.59 Evaluation, or a License Amendment Request, as applicable, is not required.)

This activity does not increase the frequency of occurrence of a Main Steam Line Break Accident or an Inadvertent MSIV closure event, or increase the likelihood of occurrence of a malfunction of an SSC important to safety. The change does not impact the pressure boundary integrity of the Main Steam System. The activity does not affect the seismic and environmental qualification of the MSIVs, nor does it impact the Main Steam piping loading conditions.

This activity does not result in an increase in the consequences of an accident or in the consequences of a malfunction of an SSC important to safety. The Large SLB outside containment upstrea m of MSIV is limiting for Offsite Dose consequences. This activity ensures that the MSIVs close as assume d in the accident analyses to isolate the main steam lines on a MSLB. This verifies that the consequences of a MSLB are not impacted.

In a Steam Generator Tube Rupture accident, the MSIVs are closed by operator action to isolate the rupture Steam Generator. The SGTR event does not affect the environmental conditi ons in the MSIV room, thus, isolating the hydraulic relief for the MSIV hydraulic manifold does not have any impact on the assumed MSIV closure in the SGTR event.

The MSIVs are assumed to close on a main steamline isolation signal in a feedwater system pipe break event. As discussed in the UFSAR, the feedwater line break with the most significant consequences would be the one that occurred inside containment between a steam generator and the feedwater check valve. In this case, the contents of the steam generator would be released to containment. For this scenario, environ mental conditions in the MSIV rooms will not be affected and the 1'YISIVs' function to close will be maintained. The environmental consequences of less severe feedline breaks in the MSIV rooms are bounded by main steamline breaks.

This activity does not create a possibility for an accident of a differe nt type than any previously evaluated in the UFSAR.

The MSIVs are maintained in the open position during normal plant operation. A consequence of a malfunction of the hydraulic circuit could only be closure of the MSIVs. Inadvertent MSIV closure is evaluated in UFSAR section 15.2.4.

The proposed change does not change the results of postulated events where an MSIV is assumed to fail open.

This activity does not create a possibility for a malfunction of an SSC important to safety with a different result than any previously evaluated in UFSAR. Inadvertent closure of a MSIV has been evaluated in the UFSAR. The isolation of the hydraulic relief valves does not change the results evaluated in the UFSAR for this failure. The MSIV inadvertent closure is bounded by a Turbine Trip and closure of the MSIVs is not assumed for this event.

50.59 REVIEW COVERSHEET FORM I-S-AA- 104-1001 Revision 3 Page 3of3 Station/Unit(s): Braidwood, Byron/Unit 1 and 2 Activity/Document Number: Abnormal Component Position Sheet for Closinc the Isolation Valves to the MSIV Hydraulic Accumulator Manifolds' Relief Valves Revision Number: NIA

Title:

Close the Isolation Valves for the Relief Valves on each of the MSIV Hydraulic Accumula tor Manifolds This activity does not result in a change that would cause any system parameter to change.

Therefore, the activity does not result in a Design Basis Limit Fission Product Barrier as described in the UFSAR being exceeded or altered.

The results of the limiting steamline break accidents are not affected by this activity.

This activity does not impact the calculated containment temperature and pressure upon a MSLB. Therefore, this activity does not exceed or alter design basis limits related to fuel cladding, RCS pressure boundary and containment.

This activity does not result in changing the method that has been used to evaluate the MSLB or other accidents that are analyzed in the UFSAR.

Based on the results of the 50.59 Evaluation, the activity may be implemented without a license amendment.

Attachments:

Attach all 50.59 Review forms completed, as appropriate.

Forms Attached: (Check all that apply.)

q Applicability Review q 50.59 Screening 50.59 Screening No. Rev.

50.59 Evaluation 50.59 Evaluation No. 6G-12-009 Rev. 0 BRW-E-2012-186 0