NOC-AE-14003180, Annual Update to the License Renewal Application (TAC Nos. ME4936 and ME4937)

Annual Update to the License Renewal Application (TAC Nos. ME4936 and ME4937)

ML14308A073
Person / Time
Site:	South Texas
Issue date:	10/22/2014
From:	Gerry Powell South Texas
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
NOC-AE-14003180, TAC ME4936, TAC ME4937
Download: ML14308A073 (18)

Text

Nuclear Operating Company NEN South Texas ProJectElectricGeneratingStation PO.Box 289 Wadsworth, Texas 77483 .

October 22, 2014 NOC-AE-14003180 10 CFR 54 File: G25 U. S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, DC 20555-0001 South Texas Project Units 1 and 2 Docket Nos. STN 50-498, STN 50-499 2014 Annual Update to the South Texas Project License Renewal Application (TAC NOS. ME4936 and ME4937)

Reference:

STPNOC Letter from G. T. Powell to NRC Document Control Desk, "License Renewal Application", dated October 25, 2010 (NOC-AE-1 0002607)

(ML103010257)

By the referenced letter, STP Nuclear Operating Company (STPNOC) submitted an application to the Nuclear Regulatory Commission (NRC) for the renewal of Facility Operating Licenses NPF-76 and NPF-80, for South Texas Project (STP) Units 1 and 2, respectively. The application included the License Renewal Application (LRA), and the Applicant's Environmental Report - Operating License Renewal Stage. As required by 10 CFR 54.21(b), each year following submittal of the LRA, an amendment to the LRA must be submitted that identifies any change to the current licensing basis (CLB) that materially affects the contents of the LRA, including the Updated Final Safety Analysis Report (UFSAR) supplement.

This LRA update covers the period from September 1, 2013 through August 31, 2014. identifies STP LRA changes that are being made to: (1) reflect the CLB that materially affect the LRA; and (2) reflect completed enhancements and commitments. Changes to LRA pages described in Enclosure 1 are depicted as line-in/line-out pages provided in .

License Renewal Application revised regulatory commitments are provided in Enclosure 2.

There are no other regulatory commitments in this letter.

Should you have any questions regarding this letter, please contact Arden Aldridge, STP License Renewal Project Lead, at (361) 972-8243, or Rafael Gonzales, STP License Renewal Project regulatory point-of-contact, at (361) 972-4779.

STI: 33950846

NOC-AE-14003180 Page 2 of 3 I declare under penalty of perjury that the foregoing is true and correct.

Executed on ge,.

Date G. T. Powell Site Vice President rjg

Enclosures:

1. STPNOC License Renewal Application (LRA) Changes Reflected in 2014 Annual LRA Update

2. STP LRA Changes with Line-in/Line-out Annotations

NOC-AE-14003180 Page 3 of 3 cc:

(paper copy) (electronic copy)

Regional Administrator, Region IV Steve Frantz U. S. Nuclear Regulatory Commission Morgan, Lewis & Bockius, LLP 1600 East Lamar Boulevard Arlington, Texas 76011-4511 John Ragan Balwant K. Singal Chris O'Hara Senior Project Manager Jim von Suskil U.S. Nuclear Regulatory Commission NRG South Texas LP One White Flint North (MS 8B1) 11555 Rockville Pike Rockville, MD 20852 NRC Resident Inspector Kevin Polio U. S. Nuclear Regulatory Commission Cris Eugster P. O. Box 289, Mail Code: MN116 L.D. Blaylock Wadsworth, TX 77483 City Public Service Jim Collins Peter Nemeth City of Austin Crain Caton & James, P.C.

Electric Utility Department 721 Barton Springs Road C. Mele Austin, TX 78704 John Wester City of Austin John W. Daily Richard A. Ratliff License Renewal Project Manager (Safety) Robert Free U.S. Nuclear Regulatory Commission Texas Department of State Health Services One White Flint North (MS 011-Fl)

Washington, DC 20555-0001 Tam Tran Balwant K. Singal License Renewal Project Manager John W. Daily (Environmental) Tam Tran U. S. Nuclear Regulatory Commission U. S. Nuclear Regulatory Commission One White Flint North (MS 011 F01)

Washington, DC 20555-0001

Enclosure 1 NOC-AE-14003180 Enclosure 1 STPNOC License Renewal Application (LRA)

Changes Reflected in 2014 Annual LRA Update

Enclosure 1 NOC-AE-14003180 Page 1 of 1 STPNOC License Renewal Application (LRA)

Changes Reflected in 2013 Annual LRA Update Following Changes Materially Affect the LRA Reason for Change Affected LRA Sections or Tables Revised Table 3.3.2-4, to reflect current information Table 3.3.2-4 regarding items related to Auxiliary Systems - Summary of Aging Management Evaluation - Essential Cooling Water and ECW Screen Wash System.

Revised Table 3.3.2-4, to reflect current information Table 3.3.2-19 regarding items related to Auxiliary Systems - Summary of Aging Management Evaluation - Chemical and Volume Control System, i.e. the notes section.

Revised Table 3.3.2-4, to reflect current information Table 3.5.2-11 regarding items related to Containments, Structures, and Component Supports - Summary of Aging Management Evaluation - Supports Section 3.5.2.1.11 was revised to include Copper Alloy as 3.5.2.1.11 a material used for the construction of support components.

This section was revised to capture minor administrative 4.3.2.11 corrections, e.g. line out of "Primary Coolant System",

from NOC-AE-14003078 Enclosure 2 pages 4 and 5.

This change lined out the following, "the pressurizer surge A3.2.1.11 line, and the accumulator line", from the first paragraph of A3.2.1.11 from NOC-AE-14003078 Enclosure 2 page 6.

This change updated Item 33, on Table A4-1, Table A4-1 implementation schedule, to, "Continued into the period of extended operation", and Item 43 removal of seal cap enclosures from Unit 2 Safety Injection System Check Valve SI0010A. Items 44 and 46 have been updated to reflect completion of the respective commitments.

Enclosure 2 NOC-AE-1 4003180 Enclosure 2 STP LRA Changes with Line-in/Line-out Annotations

Enclosure 2 NOC-AE-14003180 Page 1 of 12 Table 3.3.2-4 Auxiliary Systems - Summary of Aging Management Evaluation - Essential Cooling Water and ECW Screen Wash System CompoentIntened

• Type

...

,,

Function

..

oMateria,"

... ._ .. ___

Environment

..

T Aging Effect .

Requiring' Management.

Management AgingProgram .

[

NUREG-11801 2Item Vol. , Table

___"___

I.mI1"j Notes Closure Bolting 1LBS, PB, Copper Alloy Plant Indoor Air Loss of preload Bolting Integrity (B2.1.7) None None ;F, 1

'SIA _(Ext) _ _ __ _ _ _ _

Closure Boltingq PB CopperAlloy 'Raw Water (Ext) Loss of preload 'Bolting Inteqrity (B2.1.7) None None

,Closure Bolting PB CCopper Alloy Raw Water (Ext) Loss of material Open-Cycle Cooling 1-9 3.3.1.81 1_B

_ _ _ ____! Water System (B2.1.9)) _

IClosure Bolting LBS, PB, Stainless Plant Indoor Air ;Loss of preload :Bolting Integrity (B2.1.7) None None IH' 1 iSIA Steel _ (Ext) ......

...

Piping :LBS, PB, Stainless Raw Water (Int) Loss of material !Open-Cycle Cooling VI1.C1-15 3.3.1.79 B SIA Steel _Water  ! System (B2.1.9) __ _

ýPum p P13B Coppier Alloy iPlant Indoor Air None 'None IVIII.I-2 3.4.1.41 IA i ~~Ext) I IPumI 1PB Copper Alloy Raw Water (Ext) Loss of material ?Open-Cycle Cooling VII.C1-9 3.3.1.81 B i

Pump 4

---

IPB i

Copper Alloy Plant Indoor Air (Aluminum > .(Ext) jNone vater W

None System jD2. 1.9) 1 tVIII.1-2 13.4.1.41 <A Copper Alloy Raw Water (Ext) Loss of material Open-Cycle Cooling Vll.Cl-9 3.3.1.81 B iPump ~ PB - (Aluminum > ;Water System (B2.1.9)

. ... ..

.8 . ....... . .. .......... ..........

........ .. .

Pump PB Copper Alloy Raw Water (Ext) Loss of material Selective Leaching of VII.C1-10 I3.3.1.84 E3 (Aluminum > !Aluminum Bronze PB_ 0 (B2.1.3ý7 Copper Alloy IRaw Water (Int) Loss of material Open-Cycle Cooling VI.C1-9 3.3.1.81 (Aluminum > Water System (B2.1.9) 8% _ ...

SIPB Copper Alloy Raw Water (Int) Loss of material !Selective Leaching of IVII.C-l1 3.3.1.84 E,3 (Aluminum > Aluminum Bronze 8°/o%, _____B2.1.37)

Enclosure 2 NOC-AE-14003180 Page 2 of 12 Table 3.3.2-19 Auxiliarv Svstems - Summary of Aging Manaaement Evaluation - Chemical and Volume Control System Component Type Intended Material Environment Aging Effect Aging Management NUREG- Table I Notes Fun~ction Requiring y Program 1801 Vol. Item

.... ___,Management 2item_

Insulation INS Aluminum Plant Indoor None None V.F-2 3.2.1.50 C Air (Ext)

Insulation INS Insulation Plant Indoor None None H, 5 Calcium Silicate Air (Ext) Reduced thermal External Surfaces insulation Monitoring Program resistance due to (B2.1.20) moisture intrusion Insulation INS Insulation Plant Indoor None None None None H, 5 Fiberglass Air (Ext)

Orifice PB, TH Stainless Steel Borated Water None None VII.J-16 3.3.1.99 A Leakage (Ext)

Enclosure 2 NOC-AE-14003180 Page 3 of 12 Table 3.5.2-11 Contanmen.z VVIIL*IIIIIIVI fL*t VLItrucrtiurcq

• VL*I V*I *l andJ Conmpnennt w* VV*II* V Simpnnds~- Siimmarv of Anyina Mananement Fvalua~t ion - Smot Component f Intended

• Type Function Material 1 Environment. Aging Effect Requiring Management.

Aging Management Program.

NUREG- Table 1 Ite, " Notes 1801 Vol.

2 Item .......... .__

Supports Mech ISS Concrete Plant Indoor Air Reduction in ;Structures Monitoring 111.B4-1 3.5.1.40 A Equip Class 2 (Structural) (Ext) concrete anchor Program (B2.1.32)

.capact and 3 _

Supports Mech SS Concrete  :(Submerqed Increase in Regulatory Guide 1.127, III.A6-6 3.5.1.37 B Equip Class 2 ,(Structural) (Ext) porosity and Inspection of Water-and 3 permeability, loss Control Structures of strength Associated with Nuclear

• Power Plants (B2.1.33)

Supports Mech ISS Concrete Submergqed Loss of material Regulatory Guide 1.127.,- III.A6-7 3.5.1.45 Equip Class 2 i(Structural) (Ext) Inspection of Water-and 3 Control Structures Associated with Nuclear

-1 Power Plants (B2._1.33) rSupoorts Mech _SS Copper Alloy Submerged Loss of material iASME Section XI, Non1e None _3 Equip Class 2 (Structural) (Ext) ISubsection IWF

- ...

- 1 .................. .......... . . . . .. .

. - _ _ 1 1 2.1.2 9 )

.....

........

...

iSupports Mech 1SS Stainless Borated Water None 'None Ill.B1.2-8 3.5.1.59 A IEquip Class 2 1 Steel Leakage (Ext)

Land 3  !

Notes for Table 3.5.2-1 1:

Standard Notes:

A *Consistentwith NUREG-1801 item for component, material, environment, and aging effect. AMP is consistent with NUREG-1801 AMP.

B Consistent with NUREG-1 801 item for component, material, environment, and aging effect. AMP takes some exceptions to NUREG-1801 AMP.

E Consistent with NUREG-1 801 for material, environment, and aging effect, but a different aging management program is credited or NUREG-1801 identifies a plant-specific aging management program.

H Aging effect not in NUREG-1801 for this component, material, and environment combination.

J Neither the component nor the material and environment combination is evaluated in NUREG-1801.

Enclosure 2 NOC-AE-1 4003180 Page 4 of 12 Plant Specific Notes:

1 NUREG-1 801 does not provide a line to evaluate stainless steel components outdoors under the ASME Section XI, Subsection IWF program (B2.1.29).

2 GALL Rev 1 does not identify Loss of Preload as an AERM for structural bolting. This line is consistent with GALL Rev 2, Ill.B1.1 .TP-229.

3 NUREG-1801 does not provide a line to evaluate copper component support in a submerged environment. The ASME Section XI, Subsection IWF (B2.1.29) AMP managqes the aging of the ECW pump column copper alloy supports when the pump is removed for maintenance.

Enclosure 2 NOC-AE-14003180 Page 5 of 12 3.5.2.1.11 Supports Materials The materials of construction for the supports component types are:

0 Aluminum

• Carbon Steel

• Copper Alloy

• Concrete

• High Strength Low Alloy Steel (Bolting)

• Lubrite

• Stainless Steel Environment The supports component types are exposed to the following environments:

• Atmosphere/ Weather (Structural)

• Borated Water Leakage

• Plant Indoor Air (Structural)

• Submerged (Structural)

Aging Effects Requiring Management The following supports aging effects require management:

• Cracking

• Increase in porosity and permeability, loss of strength

• Loss of material

• Loss of mechanical function

• Loss of preload

• Reduction in concrete anchor capacity Aging Management Programs The following aging management programs manage the aging effects for the supports component types:

• ASME Section XI, Subsection IWF (B2.1.29)

• Bolting Integrity (B2.1.7)

• Boric Acid Corrosion (B2.1.4)

• Regulatory Guide 1.127, Inspection of Water-Control Structures Associated with Nuclear Power Plants (B2.1.33)

• Structures Monitoring Program (B2.1.32)

Enclosure 2 NOC-AE-14003180 Page 6 of 12 4.3.2.11 Fatigue Crack Growth Assessments and Fracture Mechanics Stability Analyses for Leak-Before-Break (LBB) Elimination of Dynamic Effects of Primary Loop Piping Failures Summary Description A leak-before-break analysis eliminated the need to postulate longitudinal and circumferential breaks in the reactor coolant system primary loop piping, under a 10 CFR 50.12 exemption.

Elimination of these breaks omitted the need to install pipe whip restraints in the primary loop and eliminated the requirement to design for dynamic (jet and whip) effects of primary loop breaks.

The containment pressurization, emergency core cooling system, and environmental qualification large-break design bases were not affected.

NRC approval of the use of leak-before-break in the reactor coolant system primary loop piping was granted with STP SER, NUREG-0781, Supplement No. 2.

Analysis The STP LBB analysis demonstrates that reactor coolant system primary loop pipe breaks are highly unlikely and need not be included in the design basis because flaws in reactor coolant system piping would have significant leaks for extended periods before developing into a large break. Such flaws would be detected by the reactor coolant pressure boundary leak detection system long before they become full size breaks.

Fatigue Crack Growth Analyses Primary Coolant System The final LBB submittal for STP included a fatigue crack growth assessment for a range of materials at a high stress location bounding the primary coolant system. The submittal concluded that the effects of low and high cycle fatigue on the integrity of primary piping are negligible.

FractureMechanics Evaluation The STP leak-before-break analysis for the primary loop, includes a fracture mechanics evaluation which depends on the crack initiation energy integral, JIN. The primary coolant loops at STP are SA 351 Grade CF8A cast stainless steel, which at PWR operating temperatures is subject to time-dependent thermal embrittlement reducing the JIN integral.

Thermal embrittlement effects depend logarithmically on time (more rapid initially, approaching a saturation value over time). The Westinghouse LBB analysis for the primary loop cites a study which determined the effects of thermal aging on piping integrity for a material at thermal embrittlement saturation. The fracture mechanics evaluation considers the thermal embrittlement aging mechanism and is defined by the current operating term. Therefore the fracture mechanics evaluation is a TLAA.

Effects of Power Uprate and Steam Generator Replacement on the LBB Analysis The Westinghouse power uprate report determined that power uprate had no effects on the LBB analysis for the primary loop piping, the pressurizer surge line, or the accumulator lines. (The pressurizer surge line and the accumulator lines are addressed in Section 4.3.2.10 in the discussion on the increase in the CUF for break consideration.) Westinghouse determined that the conclusions of the previous LBB analysis for the reactor coolant piping, pressurizer surge line, and accumulator lines remain valid after steam generator replacement.

Enclosure 2 NOC-AE-14003180 Page 7 of 12 Disposition: Validation, 10 CFR 54.21(c)(1)(i) and Aging Management, 10 CFR 54.21(c)(1)(iii)

Aging Management of the FatiqueCrack Growth Analysis The LBB analysis found that fatigue crack growth effects will be negligible. The basis for evaluation of fatigue crack growth effects in the LBB analysis will remain unchanged so long as the number of transient occurrences remains below the number assumed for the analysis of fatigue crack growth effects.

The Metal Fatigue of the Reactor Coolant Pressure Boundary program described in Section 4.3.1 and B3.1 ensures that the numbers of transients remain below the number actually experienced during the period of extended operation remain below the assumed number; or that appropriate corrective actions maintain the design and licensing basis by other acceptable means. The effects of fatigue will therefore be managed for the period of extended operation. This TLAA is dispositioned in accordance with 10 CFR 54.21(c)(1)(iii). Continuation of the 10 CFR 50.12 LBB exemption is therefore justified for the period of extended operation.

Validation of the FractureMechanics Evaluation The material fracture toughness properties selected for use in the LBB analysis are sufficiently embrittled that they bound the amount of thermal embrittlement that will occur in 60 years.

Therefore this TLAA is valid for the period of extended operation and is dispositioned in accordance with 10 CFR 54.21(c)(1 )(i).

Enclosure 2 NOC-AE-14003180 Page 8 of 12 A3.2.1.11 Fatigue Crack Growth Assessments and Fracture Mechanics Stability Analyses for Leak-Before-Break (LBB) Elimination of Dynamic Effects of Primary Loop Piping Failures A leak-before-break analysis eliminated the need to postulate longitudinal and circumferential breaks in the reactor coolant system primary loop piping, the pressurizer surge line, and the accumul6Ator line. Elimination of these breaks omitted the need to install pipe whip restraints in the primary loop and eliminated the requirement to design for dynamic (jet and whip) effects of primary loop breaks. The containment pressurization, emergency core cooling system, and environmental qualification large-break design bases were not affected.

A-ging Management of the Fatique Crack Growth Analysis The final LBB submittal for STP included a fatigue crack growth assessment for a range of materials at a high stress locations bounding the primary coolant system._The LBB analysis found that fatigue crack growth effects will be negligible. The basis for evaluation of fatigue crack growth effects in the LBB analysis will remain unchanged so long as the number of transient occurrences remains below the number assumed for the analysis of fatigue crack growth effects.

The Metal Fatigue of Reactor Coolant Pressure Boundary program, described in Section A2.1, ensures that the numbers of transients actually experienced during the period of extended operation remain below the assumed number; or that appropriate corrective actions maintain the design and licensing basis by other acceptable means. The effects of fatigue will therefore be managed for the period of extended operation. This TLAA is dispositioned in accordance with 10 CFR 54.21(c)(1)(iii).

Validation of the FractureMechanics Evaluation The STP leak-before-break analysis for the primary loop, includes a fracture mechanics evaluation which depends on the crack initiation energy integral, JIN. The material fracture toughness properties selected for use in the LBB analysis are sufficiently embrittled that they bound the amount of thermal embrittlement that will occur in 60 years. Therefore this TLAA is valid for the period of extended operation and is dispositioned in accordance with 10 CFR 54.21(c)(1)(i).

Enclosure 2 NOC-AE-14003180 Page 9 of 12 A4 LICENSE RENEWAL COMMITMENTS Table A4-1 identifies proposed actions committed to by STPNOC for STP Units 1 and 2 in its License Renewal Application. These and other actions are proposed regulatory commitments. This list will be revised, as necessary, in subsequent amendments to reflect changes resulting from NRC questions and STPNOC responses. STPNOC will utilize the STP commitment tracking system to track regulatory commitments. The Condition Report (CR) number in the Implementation Schedule column of the table is for STPNOC tracking purposes and is not part of the amended LRA.

Enclosure 2 NOC-AE-14003180 Page 10 of 12 LICENSE RENEWAL COMMITMENTS Table A4-1 identifies proposed actions committed to by STPNOC for STP Units 1 and 2 in its License Renewal Application. These and other actions are proposed regulatory commitments. This list will be revised, as necessary, in subsequent amendments to reflect changes resulting from NRC questions and STPNOC responses. STPNOC will utilize the STP commitment tracking system to track regulatory commitments.

Table A4-1 License Renewal Commitments Item#... .,.:!... ,..:'Commitment "* *..."*i LRA: Implementation Item. Commitmet " : Section Schedule 33 Periodic inspection of a sample of transmission conductor connections for loose connections 3.6.2.2.3 Complete no later using thermography is currently performed as part of the preventive maintenance activities. than SiX mntEh,' prio The periodic thermography will continue into the period of extended operation. -. .... ,9"f.-.

'

extended operation Replacoe~net to be complete no later th-an six months prior to the PE* or tl.h en.

of the last refueling outage priorF to the PEO, whichcver GGGUFS lateF.

Continued into the period of extended operation CR 10-23608 43 The seal cap enclosures from Unit 2 Safety Injection System Check Valve S10010A B2.1.7 Unit 1 completed and from Unit 1 and Unit 2 Chemical Volume Control System Check Valves CV0001, CV0002, CV0004, and CV0005 will be permanently removed. After removal of the seal 201-3Refueli44 cap enclosures, the component bolting will be replaced or inspected for intergranular tage (*. Rn*.-*)

stress corrosion cracking. Unit 2 completed CR 12-21155

Enclosure 2 NOC-AE-14003180 Page 11 of 12 44 Enhance the Selective Leaching of Aluminum Bronze procedure to update the structural B2.1.37 july 31, 2014 integrity analyses, confirm load carrying capacity, and determine degree of dealloying as (revised -pe follows: NOC AE 14003000)

• Perform volumetric examinations of leaking aluminum bronze components where the configuration supports this type of examination to conclude with reasonable assurance Completed that cracks are not approaching a critical size.

" Perform Profile Examinations (PE) on 100% of leaking components. The PE consists NOC-AE-43003135 of non-destructive examination of the leaking component for the presence of any visual crack identifications (Inside/outside surfaces) and distractive examinations for CR 12-22150 microstructure, degree of dealloying, percent of dealloying through wall thickness and chemical composition (including aluminum content). When sufficient material is available for preparation of a test coupon, mechanical properties (ultimate strength, yield strength, and/or fracture toughness) will be obtained.

" Perform pressure and bending tests (Analysis Confirmatory Tests (ACTs) on leaking components to obtain pressure and bending moment.

• Require ACTs be performed on 100% of the leaking components until 3 confirmatory ACTs from 3 different component sizes have been tested. Following the 9 confirmatory ACTs then 20% of all removed leaking aluminum bronze components will have ACTs performed until the end of the Period of Extended Operation.

Require at least two components be tested (PEs and ACTs) during the each 10-year interval.

If at least two leaking components are not identified two years prior to the end of each 10 year testing interval, a risk-ranked approach based on those components most susceptible to degradation will be used to identify candidate components for removal testing.

• Perform an engineering evaluation at the end of each PEs and ACTs testing interval to confirm the analytical methodology used to calculate the load carrying capacity and structural integrity of the leak components is conservative.

• Update the analytical methodology used to demonstrate structural integrity used to demonstrate structural integrity as required confirming that the load carrying capacity of the aluminum bronze material remains adequate to support the intended function of the ECW system through the period of extended operation.

• Trend the degree of dealloying and cracking by comparing examination results with previous examination results. Trend ultimate strength, yield strength, and/or fracture toughness results from the PE testing.

• Upon completion of each test, incorporate new test data updating existing trend to evaluate impact on the acceptance criteria.

• Specify the ASME Code Section XI structural factors for the normal/upset conditions (2.77) as well as the emergency and faulted conditions (1.39).

Enclosure 2 NOC-AE-14003180 Page 12 of 12

" Specify the acceptance criteria criterion for ultimate tensile strength and yield strength values of dealloyed aluminum bronze material is greater than or equal to 30 ksi.

Specify the acceptance criterion for fracture toughness is 65 ksi inl/2 for nondealloyed aluminum bronze castings and at welded joints in the heat affected zones.

• Initiate a corrective action document when the acceptance the criterion is not met.

• Specify that upon discovery of visual evidence of through-wall dealloying, components are scheduled for replacement by the next outage.

• Specify that when the ACTs does not confirm the structural integrity analyses, o The corrective action program as defined in 10 CFR Part 50 Appendix B will be followed to address emergent conditions to assure continued safe operation of the units.

o That a Operational Decision-Making Issue (ODMI) detailing specific steps based on identified conditions will be developed. These steps include notifying the control room of the condition, initiating a condition report and performing field walkdowns to determine compensatory action.

Leak rates that could occur upstream of any individual component supplied by the ECW N/A '3,, 2014 j"ly system will be determined to validate the maximum size flaw for which piping can still perform by4his-LetteF its intended function. NOC AE 14003135

• A summary of the results of these leak rates will be provided to the NRC for review.

Completed Results submitted in NOC-AE-43003135 CR 12-27257