Text

Tech Spec Pages for Amendments 150, 150, 144 and 144 Regarding Steam Generator Tube Surveillance Program
	ML071000245
Person / Time
Site:	Byron, Braidwood
Issue date:	03/30/2007
From:	; NRC/NRR/ADRO/DORL/LPLIII-2
To:	;
References
	TAC MC8966, TAC MC8967, TAC MC8968, TAC MC8969
	Download: ML071000245 (49)
	v • d • e

(3)

Exelon Generation Company, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess, and use at any time any byproduct, source and special nuclear material as sealed neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required; (4)

Exelon Generation Company, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess, and use in amounts as required any byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; and (5)

Exelon Generation Company, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility.

C.

The license shall be deemed to contain and is subject to the conditions specified in the Commission's regulations set forth in 10 CFR Chapter I and is subject to all applicable provisions of the Act and to the rules, regulations and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:

(1)

Maximum Power Level The licensee is authorized to operate the facility at reactor core power levels is not in excess of 3586.6 megawatts thermal (100 percent rated power) in accordance with the conditions specified herein and other items identified in Attachment 1 to this license. The items identified in to this license shall be completed as specified. is hereby incorporated into this license.

(2)

Technical Specifications The Technical Specifications contained in Appendix A as revised through Amendment No. 150, and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, are hereby incorporated into this license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

(3)

Emergency Planning In the event that the NRC finds that the lack of progress in completion of the procedures in the Federal Emergency Management Agency's final rule, 44 CFR Part 350, is an indication that a major substantive problem exists in achieving or maintaining an adequate state of emergency preparedness, the provisions of 10 CFR Section 50.54(s)(2) will apply.

Amendment No. 150 (3)

Pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess, and use at any time any byproduct, source and special nuclear material as sealed neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required; (4)

Pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive,,possess, and use in amounts as required any byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; and (5)

Pursuant to the Act and 10 CFR Parts 30, 40 and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility.

C.

This license shall be deemed to contain and is subject to the conditions specified in the Commission's regulation set forth in 10 CFR Chapter I and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:

(1)

Maximum Power Level The licensee is authorized to operate the facility at reactor core power levels not in excess of 3586.6 megawatts thermal (100 percent rated power) in accordance with the conditions specified herein.

(2)

Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A (NUREG-1 113), as revised through Amendment No.150, and the Environmental Protection Plan contained in Appendix B, both of which were attached to License No. NPF-37, dated February 14, 1985, are hereby incorporated into this license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

(3)

Deleted.

(4)

Deleted.

(5)

Deleted.

Amendment No. 150

TABLE OF CONTENTS - TECHNICAL SPECIFICATIONS 3.4 3.4.1 3.4.2 3.4.3 3.4.4 3.4.5 3.4.6 3.4.7 3.4.8 3.4.9 3.4.10 3.4.11 3.4.12 3.4.13 3.4.14 3.4.15 3.4.16 3.4.17 3.4.18 3.4.19 3.5 3.5.1 3.5.2 3.5.3 3.5.4 3.5.5 3.6 3.6.1 3.6.2 3.6.3 3.6.4 3.6.5 3.6.6 3.6.7 3.6.8 3.7 3.7.1 3.7.2 3.7.3 3.7.4 3.7.5 3.7.6 REACTOR COOLANT SYSTEM (RCS)

RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB)

Limits...............

RCS Minimum Temperature for Criticality..............

RCS Pressure and Temperature (P/T) Limits............

RCS Loops-MODES 1 and 2..............................

RCS.Loops-MODE 3.....................................

RCS Loops-MODE 4.....................................

RCS Loops-MODE 5, Loops Filled.......................

RCS Loops-MODE 5, Loops Not Filled...................

Pressurizer..........................................

Pressurizer Safety Valves.......

Pressurizer Power Operated Relief Valves (PORVs).....

Low Temperature Overpressure Protection (LTOP)

System...........................................

RCS Operational LEAKAGE..............................

RCS Pressure Isolation Valve (PIV)

Leakage...........

RCS Leakage Detection Instrumentation...............

RCS Specific Activity................................

RCS Loop Isolation Valves............................

RCS Loops-Isolated................................

Steam Generator (SG)

Tube Integrity..............

EMERGENCY CORE COOLING SYSTEMS (ECCS)

Accumulators.........................................

ECCS-Operating.......................................

ECCS-S hutdown........................................

Refueling Water Storage Tank (RWST)..................

Seal Injection Flow..................................

CONTAINMENT SYSTEMS Containment..........................................

Containment Air Locks................................

Containment Isolation Valves.........................

Containment Pressure.................................

Containment Air Temperature..........................

Containment Spray and Cooling Systems................

Spray Additive System................................

(Deleted)............................................

PLANT SYSTEMS Main Steam Safety Valves (MSSVs).....................

Main Steam Isolation Valves (MSIVs)..................

Secondary Specific Activity..........................

Steam Generator (SG)

Power Operated Relief Valves (PORVs)

Auxiliary Feedwater(AF) System.....................

Condensate Storage Tank (CST)........................

3.4.1-1 3.4.2-1 3.4.3-1 3.4.4-1 3.4.5-1 3.4.6-1 3.4.7-1 3.4.8-1 3.4.9-1 3.4.10-1 3.4.11-1 3.4.12-1 3.4.13-1 3.4.14-1 3.4.15-1 3.4.16-1 3.4.17-1 3.4.18-1 3.4.19-1 3.5.1-1 3.5.2-1 3.5.3-1 3.5.4-1 3.5.5-1 3.6.1-1 3.6.2-1 3.6.3-1 3.6.4-1 3.6.5-1 3.6.6-1 3.6.7-1 3.6.8-1 3.7.1-1 3.7.2-1 3.7.3-1 3.7.4-1 3.7.5-1 3.7.6-1 BYRON - UNITS 1 & 2 ii Amendment 150

Defi ni ti ons 1.1 1.1 Definitions LEAKAGE LEAKAGE shall be:

a.

Identified LEAKAGE

1.

LEAKAGE, such as that from pump seals or valve packing (except Reactor Coolant pump (RCP) seal water injection or leakoff),

that is captured and conducted to collection systems or a sump or collecting tank;

2.

LEAKAGE into the containment atmosphere from sources that are both specifically located and known either not to interfere with the operation of leakage detection systems or not to be pressure boundary LEAKAGE; or

3. Reactor Coolant through a Steam System (primary System (RCS)

LEAKAGE Generator to the Secondary to secondary LEAKAGE);

I

b. Unidentified LEAKAGE All LEAKAGE or leakoff)

(except RCP seal water injection that is not identified LEAKAGE;

c.

Pressure Boundary LEAKAGE MASTER RELAY TEST LEAKAGE (except primary to secondary LEAKAGE) through a nonisolable fault in an RCS component body, pipe wall, or vessel wall.

A MASTER RELAY TEST shall consist of energizing each master relay and verifying the OPERABILITY of each relay.

The MASTER RELAY TEST shall include a continuity check of each associated slave relay.

I BYRON - UNITS 1 & 2 1.1 -

4 Amendment150

RCS Operational LEAKAGE 3.4.13 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.13 RCS Operational LEAKAGE LCO 3.4.13 RCS operational LEAKAGE shall be limited to:

a.

No pressure boundary LEAKAGE;

b.

1 gpm unidentified LEAKAGE;

c.

10 gpm identified LEAKAGE; and I

d.

150 gallons per day primary to secondary LEAKAGE through any one steam generator (SG).

APPLICABILITY:

MODES 1, 2, 3, and 4.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

RCS operational A.1 Reduce LEAKAGE to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months LEAKAGE not within within limits.

limits for reasons other than pressure boundary LEAKAGE or primary to secondary LEAKAGE.

B. Required Action and B.1 Be in MODE 3.

6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time of Condition A AND not met.

B.2 Be in MODE 5, 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months OR Pressure boundary LEAKAGE exists.

OR Primary to secondary LEAKAGE not within limit.

I BYRON - UNITS I & 2 3.4.13 -

1 Amendment 150

RCS Operational LEAKAGE 3.4.13 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.13.1 -------------------

NOTES-------------------

1. Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after establishment of steady state operation.

2.

Not applicable to primary to secondary LEAKAGE.

Verify RCS operational LEAKAGE is within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months limits by performance of RCS water inventory balance.

SR 3.4.13.2 ------------------- NOTE --------------------

Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after establishment of steady state operation.

Verify primary to secondary LEAKAGE is 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months

< 150 gallons per day through any one SG.

BYRON -

UNITS 1 & 2 3.4.13 - 2 Amendment150

SG Tube. Integrity 3.4.19 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.19 Steam Generator (SG)

Tube Integrity LCO 3.4.19 SG tube integrity shall be maintained.

AND All SG tubes satisfying the tube repair criteria shall be plugged or repaired in accordance with the Steam Generator Program.

APPLICABILITY:

MODES 1, 2, 3, and 4.

ACTIONS

NOTE -----------------------------------------

Separate Condition entry is allowed for each SG tube.

CONDITION REQUIRED ACTION COMPLETION TIME A. One or more SG tubes A.1 Verify tube integrity 7 days satisfying the tube of the affected repair criteria and tube(s) is maintained not plugged or until the next repaired in accordance refueling outage or with the Steam SG tube inspection.

Generator Program.

AND A.2 Plug or repair the Prior to affected tube(s) in entering MODE 4 accordance with the following the Steam Generator next refueling Program.

outage or SG tube inspection B. Required Action and B.1 Be in MODE 3.

6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time of Condition A AND not met.

B.2 Be in MODE 5.

36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months OR SG tube integrity not maintained.

BYRON - UNITS 1 & 2 3.4.19 - 1 Amendment 150

SG Tube Integrity 3.4.19 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.19.1 Verify SG tube integrity in accordance with In accordance the Steam Generator Program.

with the Steam Generator Program SR 3.4.19.2 Verify that each inspected SG tube that Prior to satisfies the tube repair criteria is entering MODE 4 plugged or repaired in accordance with the following a SG Steam Generator Program.

tube inspection BYRON -

UNITS 1 & 2 3.4.19 - 2 Amendment 150

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.9 Steam Generator (SG)

ProQram A Steam Generator Program shall be established and implemented to ensure that SG tube integrity is maintained.

In addition, the Steam Generator Program shall include the following provisions:

a.

Provisions for condition monitoring assessments.

Condition monitoring assessment means an evaluation of the "as found" condition of the tubing with respect to the performance criteria for structural integrity and accident induced leakage.

The "as found" condition refers to the condition of the tubing during an SG inspection outage, as determined from the inservice inspection results or by other means, prior to the plugging or repair of tubes.

Condition monitoring assessments shall be conducted during each outage during which the SG tubes are inspected, plugged, or repaired to confirm that the performance criteria are being met.

b.

Performance criteria for SG tube integrity.

SG tube integrity shall be maintained by meeting the performance criteria for tube structural integrity, accident induced leakage, and operational LEAKAGE.

1.

Structural integrity performance criterion:

All in-service steam generator tubes shall retain structural integrity over the full range of normal operating conditions (including startup, operation in the power range, hot standby, and cool down and all anticipated transients included in the design specification) and design basis accidents.

This includes retaining a safety factor of 3.0 against burst under normal steady state full power operation primary-to-secondary pressure differential and a safety factor of 1.4 against burst applied to the design basis accident primary-to-secondary pressure differentials.

Apart from the above requirements, additional loading conditions associated with the design basis accidents, or combination of accidents in accordance with the design and licensing basis, shall also be evaluated to determine if the associated loads contribute significantly to burst or collapse.

In the assessment of tube integrity, those loads that do significantly affect burst or collapse shall be determined and assessed in combination with the loads due to pressure with a safety factor of 1.2 on the combined primary loads and 1.0 on axial secondary loads.

BYRON - UNITS I & 2 5.5-7 Amendment 150

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.9 Steam Generator (SG)

Proqram (continued)

2.

Accident induced leakage performance criterion:

The primary to secondary accident induced leakage rate for any design basis accident, other than a SG tube rupture, shall not exceed the leakage rate assumed in the accident analysis in terms of total leakage rate for all SGs and leakage rate for an individual SG.

Leakage is not to exceed a total of 1 gpm for all SGs.

3.

The operational LEAKAGE performance criteria is specified in LCO 3.4.13, "RCS Operational LEAKAGE."

c.

Provisions for SG tube repair criteria.

1.

Tubes found by inservice inspection to contain flaws in a non-sleeved region with a depth equal to or exceeding 40% of the nominal wall thickness shall be plugged or repaired except if permitted to remain in service through application of the alternate repair criteria discussed in TS 5.5.9.c.4.

For Unit 2 only, during Refueling Outage 13 and the subsequent operating cycle, flaws identified in the portion of the tube from the top of the hot leg tubesheet to 17 inches below the top of the tubesheet shall be plugged or repaired upon detection.

2.

Sleeves found by inservice inspection to contain flaws with a depth equal to. or exceeding the following percentages of the nominal sleeve wall thickness shall e plugged:

i.

TIG welded sleeves (per TS 5.5.9.f.2.i): 32%

3.

Tubes with a flaw in a sleeve to tube joint that occurs in the sleeve or in the original tube wall of the joint shall be plugged.

4.

The following tube repair criteria may be applied as an alternate to the 40% depth-based criteria of Technical Specification 5.5.9.c.1:

i.

For Unit 2 only, during Refueling Outage 13 and the subsequent operating cycle, flaws found in the portion of the tube below 17 inches from the top of the hot leg tubesheet do not require plugging or repair.

BYRON -

UNITS 1 & 2 5.5-8 Amendment 150

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.9 Steam Generator (SG)

Proqram (continued)

d.

Provisions for SG tube inspections.

Periodic SG tube inspections shall be performed.

The number and portions of the tubes inspected and methods of inspection shall be performed with the objective of detecting flaws of any type (e.g., volumetric flaws, axial and circumferential cracks) that may be present along the length of the tube, from the tube-to-tubesheet weld at the tube inlet to.the tube-to-tubesheet weld at the tube outlet, and that may satisfy the applicable tube repair criteria.

For Unit 2 only, during Refueling Outage 13 and the subsequent operating cycle, the portion of the tube below 17 inches from the top of the hot leg tubesheet is excluded.

The tube-to-tubesheet weld is not part of the tube.

In addition to meeting the requirements of d.1, d.2, and d.3 below, the inspection scope, inspection methods, and inspection intervals shall be such as to ensure that SG tube integrity is maintained until the next SG inspection.

An assessment of degradation shall be performed to determine the type and location of flaws to which the tubes may be susceptible and, based on this assessment, to determine which inspection methods need to be employed and at what locations.

1.

Inspect 100% of the tubes in each SG during the first refueling outage following SG replacement.

2.

Inspect 100% of the Unit 1 tubes at sequential periods of 144, 108, 72, and, thereafter, 60 effective full power months.

The first sequential period shall be considered to begin after the first inservice inspection of the SGs.

In addition, inspect 50% of the tubes by the refueling outage nearest the midpoint of the period and the remaining 50% by the refueling outage nearest the end of the period.

No SG shall operate for more than 72 effective full power months or three refueling outages (whichever is less) without being inspected.

Inspect 100% of the Unit 2 tubes at sequential periods of 120, 90, and, thereafter, 60 effective full power months.

The first sequential period shall be considered to begin after the first inservice inspection of the SGs.

In addition, inspect 50% of the tubes by the refueling outage nearest the midpoint of the period and the remaining 50% by the refueling outage nearest the end of the period.

No SG shall operate for more than 48 effective full power months or two refueling outages (whichever is less) without being inspected.

BYRON -

UNITS 1 & 2 5.5 - 9 Amendment 150

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.9 Steam Generator (SG)

Program (continued)

3.

If crack indications are found in any SG tube, then the next inspection for each SG for the degradation mechanism that caused the crack indication shall not exceed 24 effective full power months or one refueling outage (whichever is less).

If definitive information, such as from examination of a pulled tube, diagnostic non-destructive testing, or engineering evaluation indicates that a crack-like indication is not associated with a crack(s), then the indication need not be treated as a crack.

e.

Provisions for monitoring operational primary to secondary LEAKAGE.

f.

Provisions for SG tube repair methods.

Steam generator tube repair methods shall provide the means to reestablish the RCS pressure boundary integrity of SG tubes without removing the tube from service.

For the purposes of these Specifications, tube plugging is not a repair.

1.

There are no approved tube repair methods for the Unit i SGs.

2.

All acceptable repair methods for the Unit 2 SGs are listed below.

i.

TIG welded sleeving as described in ABB Combustion Engineering Inc., Technical Reports:

Licensing Report CEN-621-P, Revision 00, "Commonwealth Edison Byron and Braidwood Unit 1 and 2 Steam Generators Tube Repair Using Leak Tight Sleeves, FINAL REPORT," April 1995; and Licensing Report CEN-627-P, "Operating Performance of the ABB CENO Steam Generator Tube Sleeve for Use at Cormmonwealth Edison Byron and Braidwood Units 1 and 2," January 1996; subject to the limitations and restrictions as noted by the NRC Staff.

BYRON - UNITS 1 & 2 5.5 - 10 Amendment 150

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.10 Secondary Water Chemistry Program This program provides controls for monitoring secondary water chemistry to inhibit SG tube degradation.

The program shall incl ude:

a.

Identification of a sampling schedule for the critical variables and control points for these variables;

b.

Identification of the procedures used to measure the values of the critical variables;

c.

Identification of process sampling points, which shall include monitoring the discharge of the condensate pumps for evidence of condenser inleakage;

d.

Procedures for the recording and management of data;

e.

Procedures defining corrective actions for all off control point chemistry conditions; and

f.

A procedure identifying the authority responsible for the interpretation of the data and the sequence and timing of administrative events, which is required to initiate corrective action.

BYRON - UNITS 1 & 2 5.5 - 11 Amendment 150

Programs and Manuals 5.5 5.5 Programs and Manuals

5.11 Venti:lation Filter Testing Program (VFTP)

A program shall be established to implement the following required testing of Engineered Safety Feature (ESF) filter ventilation systems at the frequencies specified in conformance with Regulatory Guide 1.52, Revision 2, and ANSI-N1510-1980, with any exceptions noted in Appendix A of the UFSAR.

a.

Demonstrate for each of the ESF filter systems that an inplace test of the High Efficiency Particulate Air (HEPA) filters shows a penetration specified below when tested in conformance with Regulatory Guide 1.52, Revision 2. and ANSI N510-1980, with any exceptions noted in Appendix A of the UFSAR, at the system flow rate specified below.

Verification of the specified flow rates may be accomplished during the performance of SRs 3.7.10.4, 3.7.12.4, and 3.7.13.5, as applicable:

ESF Ventilation System Control Room Ventilation (VC)

Filtration System (makeup)

Nonaccessible Area Exhaust Filter Plenum Ventilation System (after structural maintenance of the HEPA filter housings)

Nonaccessible Area Exhaust Filter Plenum Ventilation System (for reasons other than structural maintenance of the HEPA filter housings)

Fuel Handling.

Building Exhaust Filter Plenum (FHB)

Venti l ation System Flow Rate Penetration

> 5400 cfm and K 6600 cfm

> 55,669 cfm and K 68,200 cfm per train, and

> 18,556 cfm and K 22,733 cfm per bank

< 0.05%

< 1%

<1%

_> 55,669

_< 68,200 train cfm and cfm per

> 18,900 cfm and K 23,100 cfm BYRON -

UNITS 1 & 2 5.5 - 12 Amendment 150

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.11 vuent laton Filter Testing Program ('JFTP)

(continued)

b.

Demonstrate for each of the ESF filter systems that an inplace test of the charcoal adsorber shows a bypass specified below when tested in conformance with Regulatory Guide 1.52, Revision 2, and ANSI N510-1980, with any exceptions noted in Appendix A of the UFSAR, at the system flow rate specified below.

Verification of the specified flow rates may be accomplished during the performance of SRs 3.7.10.4, 3.7.12.4, and 3.7.13.5, as applicable:

ESF Ventilation System Flow Rate VC Filtration System (makeup)

VC Filtration System (recirculation, charcoal bed after complete or partial replacement)

VC Filtration System (recirculation for reasons other than complete or partial charcoal bed replacement)

Nonaccessible Area Exhaust Filter Plenum Ventilation System (after structural maintenance of the charcoal adsorber housings)

Nonaccessible Area Exhaust Filter Plenum Ventilation System (for reasons other than structural "mainenance of the charcoal adsorber housings)

FHB Ventilation System

Ž 5400 cfm and 6600 cfm 44,550 cfm and 54,450 cfm

> 44,550 cfm and

54,450 cfm

55,669 cfm and

68,200 cfm per train, and

Ž 18,556 cfm and

< 22,733 cfm per bank Bypass

< 1%

< 0.1%

< 2%

< 1%

> 55,669

68,200 train cfm and cfm per

Ž 18,900 cfm and

23,100 cfm per train BYRON -

UNITS I & 2 5.5 -

13 Amendment 150

Programs and Manuals 5.5 5.5 Programs and Manuals 55il Ventilation Filter Testing Program (VFFP)

(continued)

c.

Demonstrate for each of the [SF filter systems that a laboratory test of a sample of the charcoal adsorber, when obtained as described in Regulatory Guide 1.52, Revision 2, shows the methyl iodide penetration less than the value specified below when tested in conformance with Regulatory Guide 1.52, Revision 2, ANSi N510-1980, and ASTM D3803-1989, with any exceptions noted in Appendix A of the UFSAR, at a temperature of 300C and a Relative Humidity (RH) specified below:

ESF Ventilation System Penetration RH VC Filtration System (makeup)

VC Filtration System (reci rcul ati on)

Nonaccessible Area Exhaust Filter Plenum Ventilation System FHB Ventilation System 2.0%

4%

4.5%

10%

70%

9 5%/

d.

Demonstrate for each of the ESF filter systems that the pressure drop across the combined HEPA filters and the charcoal adsorbers is < 6 inches of water gauge when tested in conformance with Regulatory Guide 1.52, Revision 2, and ANSI N510-1980, with any exceptions noted in Appendix A of the UFSAR, at the system flow rate specified below.

Verification of the specified flow rates may be accomplished during the performance of SRs 3.7.10.4, 3.7.12.4, and 3.7.13.5, as applicable:

ESF Ventilation System VC Filtration System (makeup)

Nonaccessible Area Exhaust Filter Plenum Ventilation System FHB Ventilation System Flow Rate 5400 cfm and

6600 cfm 55,669 cfm and 68,200 cfm per train

.18,900 cfm and

< 23,100 cfm BYRON -

UNITS 1 & 2 5.5 -

14 Amendment 150

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.11 Ventilation Filter. Testing Program (VFTP)

(continued)

e.

Demonstrate for each of the ESF filter systems that a bypass test of the combined HEPA filters and damper leakage shows a total bypass specified below at the system flow rate specified below.

Verification of the specified flow rates may be accomplished during the performance of SRs 3.7.12.4 and 3.7.13.5, as applicable:

ESF Ventil ation System Flow Rate Bypass Nonaccessible Area Exhaust Filter Plenum Ventilation System FHB Ventilation System

Ž 55,669 cfm and

68,200 cfm per train

Ž 18,900 cfm and

< 23,100 cfm

< 1%

f.

Demonstrate that the heaters for each of the ESF filter systems dissipate the value specified below when tested in conformance with ANSI N510-1980, with any exceptions noted in Appendix A of the UFSAR.

[SF Ventiilation System Wattage VC Filtration System The provisions of SR 3.0.2 and test frequencies.

_> 24.0 kW SR 3.0.3 are applicable to the VFTP BYRON -

UNITS 1 & 2 5.5 -

15 Amendment 150 I

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.12 ExDlosive Gas and Storaae Tank Radioactivity Monitorina Proaram This program provides controls for potentially explosive gas mixtures contained in the waste gas system, the quantity of radioactivity contained in gas decay tanks or fed into the off gas treatment system, and the quantity of radioactivity contained in unprotected outdoor liquid storage tanks.

The gaseous radioactivity quantities shall be determined following the methodology in Branch Technical Position (BTP)

ETSB 11-5, "Postulated Radioactive Release due to Waste Gas System Leak or Failure."

The liquid radwaste quantities shall be determined in accordance with the ODCM.

The program shall include:

a.

The limits for concentrations of hydrogen and oxygen in the waste gas system and a surveillance program to ensure the limits are maintained.

Such limits shall be appropriate to the system's design criteria (i.e., whether or not the system is designed to withstand a hydrogen explosion);

b.

A surveillance program to ensure that the quantity of radioactivity contained in each gas decay tank and fed into the offgas treatment system is less than the amount that would result in a whole body exposure of Ž 0.5 rem to any individual in an unrestricted area, in the event of an uncontrolled release of the tanks' contents; and

c.

A surveillance program to ensure that the quantity of radioactivity contained in all outdoor liquid radwaste tanks that are not surrounded by liners, dikes, or walls, capable of holding the tanks' contents and that do not have tank overflows and surrounding area drains connected to the liquid radwaste treatment system is less than the amount that would result in concentrations less than the limits of 10 CFR 20, Appendix B, Table 2, Column 2, at the nearest potable water supply and the nearest surface water supply in an unrestricted area, in the event of an uncontrolled release of the tanks' contents.

The provisions of SR 3.0.2 and Explosive Gas and Storage Tank surveillance frequencies.

SR 3.0.3 are applicable to the Radioactivity Monitoring Program BYRON -

UNITS 1 & 2 5.5 -

16 Amendment 150

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.13 Diesel Fuel Oil Testing Program A diesel fuel oil testing program to implement required testing of both new fuel oil and stored fuel oil shall be established.

The program shall include sampling and testing requirements, and acceptance criteria, all in accordance with applicable ASTM Standards.

The purpose of the program is to establish the fol l owing:

a.

Acceptability of new fuel oil for use prior to addition to storage tanks by determining that the fuel oil has:

1.

an API gravity or an absolute specific gravity within

limits,

2.

a flash point and kinematic viscosity within limits, and

3.

a clear and bright appearance with proper color or a water and sediment content within limits;

b.

Other properties of new fuel oil are within limits within 30 days following sampling and addition to storage tanks; and

c.

Total particulate concentration of the fuel oil is < 10 mg/l when tested every 31 days.

The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Diesel Fuel Oil Testing Program test frequencies.

BYRON -

UNITS 1 & 2 5.5 -

17 Amendment 150 I

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.14 Technical SDecifications (TS)

Bases Control Program This program provides a means for processing changes to the Bases of these Technical Specifications.

a.

Changes to the Bases of the TS shall be made under appropriate administrative controls and reviews.

b.

Licensees may make changes to Bases without prior NRC approval provided the changes do not require either of the following:

1.

a change in the TS incorporated in the license; or

2.

a change to the UFSAR or Bases that requires NRC approval pursuant to 10 CFR 50.59.

c.

The Bases Control Program shall contain provisions to ensure that the Bases are maintained consistent with the UFSAR.

d.

Proposed changes that meet the criteria of Specification 5.5.14.b above shall be reviewed and approved by the NRC prior to implementation.

Changes to the Bases implemented without prior NRC approval shall be provided to the NRC on a frequency consistent with 10 CFR 50.71(e) as modified by approved exemptions.

BYRON -

UNITS 1 & 2 5.5 - 18 Amendment 150

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.15 Safety Function Determination Program (SFDP)

This program ensures loss of safety function is detected and appropriate actions taken.

Upon entry into LCO 3.0.6, an evaluation shall be made to determine if loss of safety function exists.

Additionally, other appropriate actions may be taken as a result of the support system inoperability and corresponding exception to entering supported system Condition and Required Actions.

This program implements the requirements of LCO 3.0.6.

The SFDP shall contain the following:

a.

Provisions for cross train checks to ensure a loss of the capability to perform the safety function assumed in the accident analysis does not go undetected;

b.

Provisions for ensuring the plant is maintained in a safe condition if a loss of function condition exists;

c.

Provisions to ensure that an inoperable supported system's Completion Time is not inappropriately extended as a result of multiple support system inoperabilities; and

d.

Other appropriate limitations and remedial or compensatory actions.

A loss of safety function exists when, assuming no concurrent single failure, a safety function assumed in the accident analysis cannot be performed.

For the purpose of this program, a loss of safety function may exist when a support system is inoperable, and:

a.

A required system redundant to the system(s) supported by the inoperable support system is also inoperable; or

b.

A required system redundant to the system(s) in turn supported by the inoperable supported system is also inoperable; or

c.

A required system redundant to the support system(s) for the supported systems (a) and (b) above is also inoperable.

BYRON - UNITS I & 2 5.5 -

19 Amendment 150

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.15 Safety Function Determination Program (SFDP)

(continued)

The SFDP identifies where a loss of safety function exists.

If a loss of safety function is determined to exist by this program, the appropriate Conditions and Required Actions of the LCO in which the loss of safety function exists are required to be entered.

5.5.16 Containment Leakage Rate Testing Program A program shall be established to implement the leakage rate testing of the containment as required by 10 CFR 50.54(o) and 10 CFR 50, Appendix J, Option B, as modified by approved exemptions.

This program shall be in accordance with the guidelines contained in Regulatory Guide 1.163, September 1995 and NEI 94-01, Revision 0.

The peak calculated containment internal pressure for the design basis loss of coolant accident, P., is 42.8 psig for Unit 1 and 38.4 psig for Unit 2 The maximum allowable containment leakage rate, L..,

at P',, shall be 0.20% of containment air weightper day.

Leakage Rate acceptance criteria are:

a.

Containment leakage rate acceptance criterion is

1.0 La.

During the first unit startup following testing in accordance with this program, the leakage rate acceptance criteria are

< 0.60 L, for the Type B and C tests and < 0.75 L, for Type A tests; and BYRON - UNITS I & 2 5.5 - 20 Amendment 150

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.16 Containment Leakage Rate Testing Program (continued)

b.

Air lock testing acceptance criteria are:

1.

Overall air lock leakage rate is _ 0.05 La when tested at P,a; and

2.

For each door, seal leakage rate is:

i.

< 0.0024 Lat. when pressurized to > 3 psig, and ii.

< 0.01 La, when pressurized to > 10 psig.

The provisions of SR 3.0.2 do not apply to the test frequencies specified in the Containment Leakage Rate Testing Program.

The provisions of SR 3.0.3 are applicable to the Containment Leakage Rate Testing Program.

Battery Monitoring and Maintenance Program This program provides for restoration and maintenance, based on the recommendations of IEEE Standard 450, "IEEE Recommended Practice for Maintenance, Testing, and Replacement of Vented Lead - Acid Batteries For Stationary Applications," or of the battery manufacturer of the following:

A.

Actions to restore battery cells with float voltage

< 2.13 V, and B.

Actions to equalize and test battery cells that had been discovered with electrolyte level below the minimum established design limit.

5.5.17 BYRON -

UNITS 1 & 2 5.5 - 21 Amendment I

150

Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.8 Tendon Surveillance Report Any abnormal degradation of the containment structure detected during the tests required by the Pre-Stressed Concrete Containment Tendon Surveillance Program shall be reported in the Inservice Inspection Summrary Report in accordance with 10 CFR 50.55a and ASME Section XI, 1992 Edition with the 1992 Addenda.

5.6.9 Steam Generator (SG) Tube Inspection Report A report shall be submitted within 180 days after the initial entry into MODE 4 following completion of an inspection performed in accordance with Specification 5.5.9, Steam Generator (SG)

Program.

The report shall include:

a.

The scope of inspections performed on each SG,

b.

Active degradation mechanisms found, c*.

Nondestructive examination techniques utilized for each degradati on mechanism,

d.

Location, orientation (if linear), and measured sizes (if available) of service induced indications,

e.

Number of tubes plugged or repaired during the inspection outage for each active degradation mechanism,

f.

Total number and percentage of tubes plugged or repaired to.

date,

g.

The results of condition monitoring, including the results of tube pulls and in-situ testing,

h.

The effective plugging percentage for all plugging and tube repairs in each SG, and

i.

Repair method utilized and the number of tubes repaired by each repair method.

j.

For Unit 2, following completion of an inspection performed in Refueling Outage 13 (and any inspections performed in the subsequent operating cycle), the number of indications and location, size, orientation, and whether initiated on primary or secondary side for each indication detected in the upper 17-inches of the tubesheet thickness.

BYRON - UNITS 1 & 2 5.6-6 Amendment150

Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.9 Steam Generator (SG) Tube Inspection Report (continued)

k.

For Unit 2, following completion of an inspection performed in Refueling Outage 13 (and any inspections performed in the subsequent operating cycle), the operational primary to secondary leakage rate observed (greater than three gallons per day) in each steam generator (if it is not practical to assign the leakage to an individual steam generator, the entire primary to secondary leakage should be conservatively assumed to be from one steam generator) during the cycle preceding the inspection which is the subject of the report.

1.

For Unit 2, following completion of an inspection performed in Refueling Outage 13 (and any inspections performed in the subsequent operating cycle), the calculated accident leakage rate from the lowermost 4-inches of tubing for the most limiting accident in the most limiting steam generator.

In addition, if the calculated accident leakage rate from the most limiting accident is less than 2 times the maximum operational primary to secondary leakage rate, the report should describe how it was determined.

BYRON -

UNITS 1 & 2 5.6-7 Amendment150 (3)

Exelon Generation Company, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess, and use at any time any byproduct, source and special nuclear material as sealed neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required; (4)

Exelon Generation Company, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess, and use in amounts as required any byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; and (5)

Exelon Generation Company, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility.

C.

The license shall be deemed to contain and is subject to the conditions specified in the Commission's regulations set forth in 10 CFR Chapter I and is subject to all applicable provisions of the Act and to the rules, regulations and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:

(1)

Maximum Power Level The licensee is authorized to operate the facility at reactor core power levels is not in excess of 3586.6 megawatts thermal (100 percent rated power) in accordance with the conditions specified herein and other items identified in Attachment 1 to this license. The items identified in to this license shall be completed as specified. is hereby incorporated into this license.

(2)

Technical Specifications The Technical Specifications contained in Appendix A as revised through Amendment No.144, and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, are hereby incorporated into this license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

(3)

Emergency Planning In the event that the NRC finds that the lack of progress in completion of the procedures in the Federal Emergency Management Agency's final rule, 44 CFR Part 350, is an indication that a major substantive problem exists in achieving or maintaining an adequate state of emergency preparedness, the provisions of 10 CFR Section 50.54(s)(2) will apply.

Amendment No. 144 material as sealed neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required; (4)

Exelon Generation Company, LLC pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess, and use in amounts are required any byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; and (5)

Exelon Generation Company, LLC, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility.

C.

The license shall be deemed to contain and is subject to the conditions specified in the Commission's regulations set forth in 10 CFR Chapter I and is subject to all applicable provisions of the Act and to the rules, regulations and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:

(1)

Maximum Power Level The licensee is authorized to operate the facility at reactor core power levels is not in excess of 3586.6 megawatts thermal (100 percent rated power) in accordance with the conditions specified herein and other items identified in Attachment 1 to this license. The items identified in to this license shall be completed as specified. is hereby incorporated into this license.

(2)

Technical Specifications The Technical Specifications contained in Appendix A as revised through Amendment No. 144, and the Environmental Protection Plan contained in Appendix B, both of which are attached to License No. NPF-72, dated July 2, 1987, are hereby incorporated into this license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

(3)

Emergency Planning In the event that the NRC finds that the lack of progress in completion of the procedures in the Federal Emergency Management Agency's final rule, 44 CFR Part 350, is an indication that a major substantive problem exists in achieving or maintaining an adequate state of emergency preparedness, the provisions of 10 CFR Section 50.54(s)(2) will apply.

Amendment No. 144

TABLE OF CONTENTS - TECHNICAL SPECIFICATIONS 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.1 RCS Pressure, Temperature, and Flow Departure

. from Nucleate Boiling (DNB)

Limits...............

3.4.1-1 3.4.2 RCS Minimum Temperature for Criticality..............

3.4.2-1 3.4.3 RCS Pressure and Temperature (P/T) Limits............

3.4.3-1 3.4.4 RCS Loops-MODES 1 and 2..............................

3.4.4-1 3.4.5 RCS Loops-MODE 3......

3.4.5-1 3.4.6 RCS Loops-MODE 4......................................

3.4.6-1 3.4.7 RCS Loops-MODE 5, Loops Filled.......................

3.4.7-1 3.4.8 RCS Loops-MODE 5, Loops Not Filled...................

3.4.8-1 3.4.9 Pressurizer..............................

3.4.9-1 3.4.10 Pressurizer Safety Valves............................

3.4.10-1 3.4.11 Pressurizer Power Operated Relief Valves (PORVs)..... 3.4.11-1 3.4.12 Low Temperature Overpressure Protection (LTOP)

System..............

3.4.12-1 3.4.13 RCS Operational LEAKAGE..............................

3.4.13-1 3.4.14 RCS Pressure Isolation Valve (PIV)

Leakage........... 3.4.14-1 3.4.15 RCS Leakage Detection Instrumentation...............

3.4.15-1 3.4.16 RCS Specific Activity........

3.4.16-1 3.4.17 RCS Loop Isolation Valves............................

3.4.17-1 3.4.18 RCS Loops-Isolated

........... 3.4.18-1 3.4.19 Steam Generator (SG) Tube Integrity.................

3.4.19-1 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 3.5.1 Accumulators.........................................

3.5.1-1 3.5.2 ECCS-Operating.......................................

3.5.2-1 3.5.3 ECCS-Shutdown.................

3.5.3-1 3.5.4 Refueling Water Storage Tank (RWST).............

3.5.5 Seal Injection Flow..................................

3.5.5-1 3.6 CONTAINMENT SYSTEMS 3.6.1 Containment..........................................

3.6.1-1 3.6.2 Containment Air Locks....

3.6.2-1 3.6.3 Containment Isolation Valves....

3.6.3-1 3.6.4 Containment Pressure.................................

3.6.4-1 3.6.5 Containment Air Temperature..........................

3.6.5-1 3.6.6 Containment Spray and Cooling Systems................

3.6.6-1 3.6.7 Spray Additive System................................

3.6.7-1 3.6.8 (Deleted)............................................

3.6.8-1 3.7 PLANT SYSTEMS 3.7.1 Main Steam Safety Valves (MSSVs).....................

3.7.1-1 3.7.2 Main Steam Isolation Valves (MSIVs)..................

3.7.2-1 3.7.3 Secondary Specific Activity..........................

3.7.3-1 3.7.4 Steam Generator (SG)

Power Operated Relief Valves (PORVs)..............

3.7.4-1 3.7.5 Auxiliary Feedwater (AF) System......................

3.7.5-1 3.7.6 Condensate Storage Tank (CST)........................

3.7.6-1 BRAIDWOOD - UNITS 1 & 2 ii Amendment 144

Definitions 1.1 1.1 Definitions LEAKAGE LEAKAGE shall be:

a.

Identified LEAKAGE

1.

LEAKAGE, such as that from pump seals or valve packing (except Reactor Coolant pump (RCP) seal water injection or leakoff),

that is captured and conducted to collection systems or a sump or collecting tank;

2.

LEAKAGE into the containment atmosphere from sources that are both specifically located and known either not to interfere with the operation of leakage detection systems or not to be pressure boundary LEAKAGE; or

3. Reactor Coolant System (RCS)

LEAKAGE through a steam generator to the Secondary System (primary to secondary LEAKAGE);

b.

Unidentified LEAKAGE All LEAKAGE (except RCP seal water injection or leakoff) that is not identified LEAKAGE;

c.

Pressure Boundary LEAKAGE LEAKAGE (except primary to secondary LEAKAGE) through a nonisolable fault in an RCS component body, pipe wall, or vessel wall.

MASTER RELAY TEST A MASTER RELAY TEST shall consist of energizing each master relay and verifying the OPERABILITY of each relay.

The MASTER RELAY TEST shall include a continuity check of each associated slave relay.

BRAIDWOOD - UNITS 1 & 2 1.1-4 Amendment 144

RCS Operational LEAKAGE 3.4.13 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.13 RCS Operational LEAKAGE LCO 3.4.13 RCS operational LEAKAGE shall be limited to:

a.

No pressure boundary LEAKAGE;

b.

1 gpm unidentified LEAKAGE;

c.

10 gpm identified LEAKAGE; and

d.

150 gallons per day primary to secondary LEAKAGE through any one steam generator (SG).

I APPLICABILITY:

MODES 1, 2, 3, and 4.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

RCS operational A.1 Reduce LEAKAGE to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months LEAKAGE not within within limits.

limits for reasons other than pressure boundary LEAKAGE or primary to secondary LEAKAGE.

B. Required Action and B.1 Be in MODE 3.

6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time of Condition A AND not met.

B.2 Be in MODE 5.

36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months OR Pressure boundary LEAKAGE exists.

OR Primary to secondary LEAKAGE not within limit.

BRAIDWOOD - UNITS 1 & 2 3.4.13 -

1 Amendment144

RCS Operational LEAKAGE 3.4.13 SURVEILLANCE REQUIREMENTS SURVEI LLANCE FREQUENCY SR 3.4.13.1 -------------------

NOTES -------------------

1.

Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after establishment of steady state operation.

2.

Not applicable to primary to secondary LEAKAGE.

I Verify RCS operational LEAKAGE is within limits by performance of RCS water inventory balance.

72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months SR 3.4.13.2

NOTE---------------

Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after establishment of steady state operation.

Verify primary to secondary LEAKAGE is 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months 150 gallons per day through any one SG.

BRAIDWOOD -

UNITS I & 2 3.4. 13 - 2 Amendment 144

SG Tube Integrity 3.4.19 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.19 Steam Generator (SG)

Tube Integrity LCO 3.4.19 SG tube integrity shall be maintained.

AND All SG tubes satisfying the tube repair criteria shall be plugged or repaired in accordance with the Steam Generator Program.

APPLICABILITY:

MODES 1, 2, 3, and 4.

ACTIONS

NOTE -----------------------------------------

Separate Condition entry is allowed for each SG tube.

CONDITION REQUIRED ACTION COMPLETION TIME A. One or more SG tubes A.1 Verify tube integrity 7 days satisfying the tube.

of the affected repair criteria and tube(s) is maintained not plugged or until the next repaired in accordance refueling outage or with the Steam SG tube inspection.

Generator Program.

AND A.2 Plug or repair the Prior to affected tube(s) in entering MODE 4 accordance with the following the Steam Generator next refueling Program.

outage or SG tube inspection B. Required Action and B.1 Be in MODE 3.

6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time of Condition A AND not met.

B.2 Be in MODE 5.

36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months OR SG tube integrity not maintained.

BRAIDWOOD - UNITS 1 & 2 3.4.19 - I Amendment 144

SG Tube Integrity 3.4.19 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.19.1 Verify SG tube integrity in accordance with In accordance the Steam Generator Program.

with the Steam Generator Program SR 3.4.19.2 Verify that each inspected SG tube that Prior to satisfies the tube repair criteria is entering MODE 4 plugged or repaired in accordance with the following a SG Steam Generator Program.

tube inspection BRAIDWOOD - UNITS 1 & 2 3.4.19 - 2 Amendment 144

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.9 Steam Generator (SG)

Program A Steam Generator Program shall be established and implemented to ensure that SG tube integrity is maintained.

In addition, the Steam Generator Program shall include the following provisions:

a.

Provisions for condition monitoring assessments.

Condition monitoring assessment means an evaluation of the "as found" condition of the tubing with respect to the performance criteria for structural integrity and accident induced leakage.

The "as found" condition refers to the condition of the tubing during an SG inspection outage, as determined from the inservice inspection results or by other means, prior to the plugging or repair of tubes.

Condition monitoring assessments shall be conducted during each outage during which the SG tubes are inspected, plugged, or repaired to confirm that the performance criteria are being met.

b.

Performance criteria for SG tube integrity.

SG tube integrity shall be maintained by meeting the performance criteria for tube structural integrity, accident induced leakage, and operational LEAKAGE.

1.

Structural integrity performance criterion:

All in-service steam generator tubes shall retain structural integrity over the full range of normal operating conditions (including startup, operation in the power range, hot standby, and cool down and all anticipated transients included in the design specification) and design basis accidents.

This includes retaining a safety factor of 3.0 against burst under normal steady state full power operation primary-to-secondary pressure differential and a safety factor of 1.4 against burst applied to the design basis accident primary-to-secondary pressure differentials.

Apart from the above requirements, additional loading conditionsassociated with the design basis accidents, or combination of accidents in accordance with the design and licensing basis, shall also be evaluated to determine if the associated loads contribute significantly to burst or collapse.

In the assessment of tube integrity, those loads that do significantly affect burst or collapse shall be determined and assessed in combination with the loads due to pressure with a safety factor of 1.2 on the combined primary loads and 1.0 on axial secondary loads.

BRAIDWOOD - UNITS 1 & 2 5.5 - 7 Amendment 14zI*

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.9 Steam Generator (SG)

Program (continued)

2.

Accident induced leakage performance criterion:

The primary to secondary accident induced leakage rate for any design basis accident, other than a SG tube rupture, shall not exceed the leakage rate assumed in the accident analysis in terms of total leakage rate for all SGs and leakage rate for an individual SG.

Leakage is not to exceed a total of 1lgpm for all SGs.

3.

The operational LEAKAGE performance criteria is specified in LCO 3.4.13, "RCS Operational LEAKAGE."

c.

Provisions for SG tube repair criteria.

1.

Tubes found by inservice inspection to contain flaws in a non-sleeved region with a depth equal to or exceeding 40% of the nominal wall thickness shall be plugged or repaired except if permitted to remain in service through application of the alternate repair criteria discussed in TS 5.5.9.c.4.

For Unit 2 only, during Refueling Outage 12 and the subsequent operating cycle, flaws identified in the portion of the tube from the top of the hot leg tubesheet to 17 inches below the top of the tubesheet shall be plugged or repaired upon detection.

2.

Sleeves found by inservice inspection to contain flaws with a depth equal to or exceeding the following ercent ages of the nominal sleeve wall thickness shall e plugged:

i.

TIG welded sleeves (per TS 5.5.9.f.2.i): 32%

3.

Tubes with a flaw in a sleeve to tube joint that occurs in the sleeve or in the original tube wall of the joint shall be plugged.

4.

The following tube repair criteria may be applied as an alternate to the 40% depth-based criteria of Technical Specification 5.5.9.c.1:

i.

For Unit 2 only, during Refueling Outage 12 and the subsequent operatingcycle, flaws found in the portion of the tube below 17 inches from the top of the hot leg tubesheet do not require plugging or repair.

BRAIDWOOD -

UNITS I & 2 5.5 -

8 Amendment 14*7 T4

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.9 Steam Generator (SG)

Program (continued)

d.

Provisions for SG tube. inspections.

Periodic SG tube inspections shall be performed.

The number and portions of the tubes inspected and methods of inspection shall be performed with the objective of detecting flaws of any type (e.g., volumetric flaws, axial and circumferential cracks) that may be present along the length of the tube, from the tube-to-tubesheet weld at the tube inlet to the tube-to-tubesheet weld at the tube outlet, and that may satisfy the applicable tube repair criteria.

For Unit 2 only, during Refueling Outage 12 and the subsequent operating cycle, the portion of the tube below 17 inches from the top of the hot leg tubesheet is excluded.

The tube-to-tubesheet weld is not part of the tube.

In addition to meeting the requirements of d.1, d.2, and d.3 below, the inspection scope, inspection methods, and inspection intervals shall be such as to ensure that SG tube integrity is maintained until the next SG inspection.

An assessment of degradation shall be performed to determine the type and location of flaws to which the tubes may be susceptible and, based on this assessment, to determine which inspection methods need to be employed and at what locations.

1.

Inspect 100% of the tubes in each SG during the first refueling outage following SG replacement.

2.

Inspect 100% of the Unit 1 tubes at sequential periods of 144,.108, 72, and, thereafter, 60 effective full power months.

The first sequential period shall be considered to begin after the first inservice inspection of the SGs.

In addition, inspect 50% of the tubes by the refueling outage nearest the midpoint of the period and the remaining 50% by the refueling outage nearest the end of the period.

No SG shall operate for more than 72 effective full power months or three refueling outages (whichever is less) without being inspected.

Inspect 100% of the Unit 2 tubes at sequential periods of 120, 90, and, thereafter, 60 effective full power months.

The first sequential period shall be considered to begin after the first inservice inspection of the SGs.

In addition, inspect 50% of the tubes by.the refueling outage nearest the midpoint of the period and the remaining 50% by the refueling outage nearest the end of the period.

No SG shall operate for more than 48 effective full power months or two refueling outages (whichever is less) without being inspected.

BRAIDWOOD -

UNITS I & 2 5.5 -

9 Amendment 1444

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.9 Steam Generator (SG)

Program (continued)

3.

If crack indications are found in any SG tube, then the next inspection for each SG for the degradation mechanism that caused the crack indication shall not exceed 24 effective full power months or one refueling outage (whichever is less).

If definitive information, such as from examination of a pulled tube, diagnostic non-destructive testing, or engineering evaluation indicates that a crack-like indication is not associated with a crack(s), then the indication need not be treated as a crack.

e.

Provisions for monitoring operational primary to secondary LEAKAGE.

f.

Provisions for SG tube repair methods.

Steam generator tube repair methods shall provide the means to reestablish the RCS pressure boundary integrity of SG tubes without removing the tube from service.

For the purposes of these Specifications, tube plugging is not a repair.

1.

There are no approved tube repair methods for the Unit 1 SGs.

2.

All acceptable repair methods for the Unit 2 SGs are listed below..

i.

TIG welded sleeving as described in ABB Combustion Engineering Inc., Technical Reports:

Licensing Report CEN-621-P, Revision 00, "Commonwealth Edison Byron and Braidwood Unit 1 and 2 Steam Generators Tube Repair Using Leak Tight Sleeves, FINAL REPORT," April 1995; and Licensing Report CEN-627-P, "Operating Performance of the ABB CENO Steam Generator Tube Sleeve for Use at Colmmonwealth Edison Byron and Braidwood Units 1 and 2," January 1996; subject to the limitations and restrictions as noted by the NRC Staff.

BRAIDWOOD -

UNITS I & 2 5.5 - 10 Amendment 144

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.10 Secondary Water Chemistry Program This program provides controls for monitoring secondary water chemistry to inhibit SG tube degradation.

The program shall include:

a.

Identification of a sampling schedule for the critical variables and control points for these variables;

b.

Identification of the procedures used to measure the values of the critical variabl es;

c.

Identification of process sampling points, which shall include monitoring the discharge of the condensate pumps for evidence of condenser inleakage;

d.

Procedures for the recording and management of data;

e.

Procedures defining corrective actions for all off control point chemistry conditions; and

f.

A procedure identifying the authority responsible for the interpretation of the data and the sequence and timing of administrative events, which is required to initiate corrective action.

BRAIDWOOD -

UNITS 1 & 2 5.5 -

11 Amendment-,

..+/-

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.11 Ventilation Filter Testing Program (VFTP)

A program shall be established to implement the following required testing of Engineered Safety Feature (ESF) filter ventilation systems at the frequencies specified in conformance with Regulatory Guide 1.52, Revision 2, and ANSI N510-1980, with any exceptions noted in Appendix A of the UFSAR.

a.

Demonstrate for each of the ESF filter systems that an inplace test of the High Efficiency Particulate Air (HEPA) filters shows a penetration specified below when tested in conformance with Regulatory Guide 1.52, Revision 2, and ANSI N510-1980, with any exceptions noted in Appendix A of the UFSAR, at the system flow rate specified below.

Verification of the specified flow rates may be accomplished during the performance of SRs 3.7.10.4, 3.7.12.4, and 3.7.13.5, as applicable:

ESF Ventilation System Control Room Ventilation (VC)

Filtration System (makeup)

Nonaccessibl e Area Exhaust Filter Plenum Ventilation System (after structural maintenance of the HEPA filter housings)

Nonaccessibl e Area Exhaust Filter Plenum Ventil ation System (for reasons other than structural maintenance of the HEPA filter housings)

Fuel Handling Building Exhaust Filter Plenum (FHB)

Ventilation System Flow Rate Penetration 5400 cfm and 6600 cfm 60,210 cfm and 73,590 cfm per train, and 20,070 cfm and 24,530 cfm per bank

< 0.05%

< 1%

> 60,210

z 73,590 train cfm and cfm per 18,900 cfm and 23,100 cfm BRAIDWOOD - UNITS 1 & 2 5.5 - 12 Amendment I

144

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.11 Ventilation Filter Testing Program (VFTP)

(continued)

b.

Demonstrate for each of the ESF filter systems that an inplace test of the charcoal adsorber shows a bypass specified below when tested in conformance with Regulatory Guide 1.52, Revision 2, and ANSI N510-1980, with any exceptions noted in Appendix A of the UFSAR, at the system flow rate specified below.

Verification of the specified flow rates may be accomplished during the performance of SRs 3.7.10.4, 3.7.12.4, and 3.7.13.5, as applicable:

ESF Ventilation System VC Filtration System (makeup)

VC Filtration System (reci rcul ati on, charcoal bed after complete or partial replacement)

VC Filtration System (reci rcul ati on for reasons other than complete or partial charcoal bed replacement)

Nonaccessible Area Exhaust Filter Plenum Ventilation System (after structural maintenance of the charcoal adsorber housings)

Nonaccessible Area Exhaust Filter Plenum Ventilation System (for reasons other than structural maintenance of the charcoal adsorber housings)

FHB Ventilation System Flow Rate

Ž 5400 cfm and

< 6600 cfm 44,550 cfm and

<54,450 cfm

> 44,550 cfm and

54,450 cfm

60,210 cfm and

73,590 cfm per train, and 20,070 cfm and 24,530 cfm per bank

> 60,210 cfm and

< 73,590 cfm per train 13,900 cfm and

< 23,100 cfm per train Bypass

< 1%

< 0.1%

< 2%

< 1%

BRAIDWOOD -

UNITS 1 & 2 5.5 -

23 Amendment 144

Programs and Manuals 5.5 5.5 Programs and Manuals 6.5.11 Ventilation Filter Testing Program (VFTP)

(continued)

c.

Demonstrate for each of the ESF filter systems that a laboratory test of a sample of the charcoal adsorber, when obtained as described in Regulatory Guide 1.52, Revision 2, shows the methyl iodide penetration less than the value specified below when tested in conformance with Regulatory Guide 1.52, Revision 2, ANSI N510-1980, and ASTM D3803-1989, with any exceptions noted in Appendix. A of the UFSAR, at a temperature of 301C and a Relative Humidity (RH) specified below:

ESF Ventilation System Penetration RH VC Filtration System 2.0%

70%

(makeup)

VC Filtration System 4%

70%

(reci rcul ati on)

Nonaccessible Area 4.5%

70%

Exhaust Filter Plenum Ventilation System FHB Ventilation System 10%

95%

d.

Demonstrate for each of the ESF filter systems that the pressure drop across the combined HEPA filters and the charcoal adsorbers is < 6 inches of water gauge when tested in conformance with Regulatory Guide 1.52, Revision 2, and ANSI N510-1980., with any exceptions noted in Appendix A of the UFSAR, at the system flow rate specified below.

Verification of the specified flow rates may be accomplished during the performance of SRs 3.7.10.4, 3.7.12.4, and 3.7.13.5, as applicable:

ESF Ventilation System Flow Rate VC Filtration System

_ 5400 cfm and _< 6600 cfm (makeup)

Nonaccessible Area

_ 60,210 cfm and Exhaust Filter Plenum

< 73,590 cfm per train Ventilation System FHB Ventilation System

_ 18,900 cfm and

< 23,100 cfm BPAIDWOOD -

UNITS I & 2 5.5 -

14 Amendment I

144

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.11 Ventilation Filter TestinO Program (VFTP)

(continued)

e.

Demonstrate for each of the ESF filter systems that a bypass test of the combined HEPA filters and damper leakage shows a total by pass specified below at the system flow rate specified below.

Verification of the specified flow rates may be accomplished during the performance of SRs 3.7.12.4 and 3.7.13.5, as applicable:

ESF Ventilation System Flow Rate Bypass Nonaccessible Area Exhaust Filter Plenum Ventilation System FHB Ventilation System

Ž 60,210 cfm and

73,590 cfm per train

Ž 18,900 cfm and

< 23,100 cfm

< 1%

_ 1%

f.

Demonstrate that the heaters for each of the ESF filter systems dissipate the value specified below when tested in conformance with ANSI N510-1980, with any exceptions noted in Appendix A of the UFSAR.

ESF Ventilation System Wattaqe VC Filtration System The provisions of SR 3.0.2 test frequencies.

29.9 kW and > 24.5 kW and SR 3.0.3 are applicable to the VFTP BRAIDWOOD -

UNITS I & 2 5.5 - 15 Amendment 144 I

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.12 Explosive Gas and Storage Tank Radioactivity Monitoring Program This program provides controls for potentially explosive gas mixtures contained in the waste gas system, the quantity of radioactivity contained in gas decay tanks or fed into the off gas treatment system, and the quantity of radioactivity contained in unprotected outdoor liquid storage tanks.

The gaseous radioactivity quantities shall be determined following the methodology in Branch Technical Position (BTP)

ETSB 11-5, "Postulated Radioactive Release due to Waste'Gas System Leak or Failure."

The liquid radwaste quantities shall be determined in accordance with the ODCM.

The program shall include:

a.

The limits for concentrations of hydrogen and oxygen in the waste gas system and a surveillance program to ensure the limits are maintained.

Such limits shall be appropriate to the system's design criteria (i.e., whether or not the system is designed to withstand a hydrogen explosion);

b.

A surveillance program to ensure that the quantity of radioactivity contained in each gas decay tank and fed into the offgas treatment system is less than the amount that would result in a whole body exposure of Ž 0.5 rem to any individual in an unrestricted area, in the event of an uncontrolled release of the tanks' contents; and

c.

A surveillance program to ensure that the quantity of radioactivity contained in all outdoor liquid radwaste tanks that are not surrounded by liners, dikes, or walls, capable of holding the tanks' contents and that do not have tank overflows and surrounding area drains connected to the liquid radwaste treatment system is less than the amount that would result in concentrations less than the limits of 10 CFR 20, Appendix B, Table 2, Column 2, at the nearest potable water supply and the nearest surface water supply in an unrestricted area, in the event of an uncontrolled release of the tanks' contents.

The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Explosive Gas and Storage Tank Radioactivity Monitoring Program surveillance frequencies.

BRAIDWOOD -

UNITS I & 2 5.5 -

16 Amendment 144

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.13 Diesel Fuel Oil Testing Program A diesel fuel oil testing program to implement required testing of both new fuel oil and stored fuel oil shall be established.

The program shall include sampling and testing requirements, and acceptance criteria, all in accordance with applicable ASTM Standards.

The purpose of the program is to establish the following:

a.

Acceptability of new fuel oil for use prior to addition to storage tanks by determining that the fuel oil has:

1.

an API gravity or an absolute specific gravity within

limits,

2.

a flash point and kinematic viscosity within limits, and

3.

a clear and bright appearance with proper color or a water and sediment content within limits;

b.

Other properties of new fuel oil are within limits within 30 days following sampling and addition to storage tanks; and

c.

Total particulate concentration of the fuel oil is when tested every 31 days.

10 mg/l The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Diesel Fuel Oil Testing Program test frequencies.

BRAIDWOOD -

UNITS 1 & 2 5.5 - 17 Amendment 144

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.14 Technical Soecifications (TS) Bases Control Proaram This program provides a means for processing changes to the Bases of these Technical Specifications.

a.

Changes to the Bases of the TS shall be made under appropriate administrative controls and reviews.

b.

Licensees may make changes to Bases without prior NRC approval provided the changes do not require either of the fo1lowing:

1.

a change in the TS incorporated in the license; or

2.

a change to the UFSAR or Bases that requires NRC approval pursuant to 10 CFR 50.59.

c.

The Bases Control Program shall contain provisions to ensure that the Bases are maintained consistent with the UFSAR.

d.

Proposed changes that meet the criteria of Specification 5.5.14.b above shall be reviewed and approved by the NRC prior to implementation.

Changes to the Bases implemented without prior NRC approval shall be provided to the NRC on a frequency consistent with 10 CFR 50.71(e) as modified by approved exemptions.

BRAIDWOOD -

UNITS 1 & 2 5.5 -

18 Amendment 144

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.15 Safety Function Determination Program (SFDP)

This program ensures loss of safety function is detected and appropriate actions taken.

Upon entry into LCO 3.0.6, an evaluation shall be made to determine if loss of safety function exists.

Additionally, other appropriate actions may be taken as a result of the support system inoperability and corresponding exception to entering supported system Condition and Required Actions.

This program implements the requirements of LCO 3.0.6.

The SFDP shall contain the following:

a.

Provisions for cross train checks to ensure a loss of the capability to perform the safety function assumed in the accident analysis does not go undetected;

b.

Provisions for ensuring the plant is maintained in a safe condition if a loss of function condition exists;

c.

Provisions to ensure that an inoperable supported system's Completion Time is not inappropriately extended as a result of multiple support system inoperabilities; and

d.

Other appropriate limitations and remedial or compensatory actions.

A loss of safety function exists when, assuming no concurrent single failure, a safety function assumed in the accident analysis cannot be performed.

For the purpose of this program, a loss of safety function may exist when a support system is inoperable, and:

a.

A required system redundant to the system(s) supported by the inoperable support system is also inoperable; or

b.

A required system by the inoperable

c.

A required system supported systems redundant to the system(s) in turn supported supported system is also inoperable; or redundant to the support system(s) for the (a) and (b) above is also inoperable.

BRAIDWOOD -

UNITS 1 & 2 5.5 -

19 Amendment 144

Programs and Manuals 5.5 5.5 Programs and Manuals

.5.15 Sa.ety Function Determination Program (SFDP)

(continued)

The SFDP identifies where a. loss of safety function exists.

If a loss of safety function is determined to exist by this program, the appropriate Conditions and Required Actions of the LCO in which the loss of safety function exists are required to be entered.

5.5.16 Containment Leakaqe Rate Testinq Proqram A program shall be established to implement the leakage rate testing of the containment as required by 10 CFR 50.54(o) and 10 CFR 50, Appendix J, Option B, as modified by approved exemptions.

This program shall be in accordance with the guidelines contained in Regulatory Guide 1.163, September 1995 and NEi 94-01, Revision 0.

The peak calculated containment internal pressure for the design basis loss of coolant accident, P., is 42.8 psig for Unit 1 and 38.4 psig for Unit 2 The maximum allowable containment leakage rate, L., at P., shall be 0.201% of containment air weight per day.

Leakage Rate acceptance criteria are:

a.

Containment leakage rate acceptance criterion is

1.0 L.

During the first unit startup following testing in accordance with this program, the leakage rate acceptance criteria are

< 0.60 L, for the Type B and C tests and < 0.75 La for Type A tests; and BP,.AIDWOOD - UNITS I & 2 5.5 - 20 Amendment 144

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.16 Containment Leakage Rate Testing Program (continued)

b.

Air lock testing acceptance criteria are:

1.

Overall at > Pa; air lock leakage rate is _< 0.05 La when tested and

2.

For each door, seal leakage rate is:

i.

< 0.0024 La, when pressurized to Ž 3 psig; and ii.

< 0.01 La, when pressurized to Ž 10 psig.

The provisions of SR 3.0.2 do not apply to the test frequencies specified in the Containment Leakage Rate Testing Program.

The provisions of SR 3.0.3 are applicable to the Containment Leakage Rate Testing Program.

Battery Monitorina and Maintenance Proaram 5.5.17 This program provides for restoration and maintenance, based on the recommendations of IEEE Standard 450, "IEEE Recommended Practice for Maintenance, Testing, and Replacement of Vented Lead-Acid Batteries For Stationary Applications," or of the battery manufacturer of the following:

a.

Actions to restore battery cells with float voltage

< 2.13 V, and

b.

Actions to equalize and test battery cells that had been discovered with electrolyte level below the minimum established design limit.

BRAIDWOOD -

UNITS 1 & 2 5.5 - 21 Amendment 144

Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.8 Tendon Surveillance Report Any abnormal degradation of the containment structure detected during the tests required-by the Pre-Stressed Concrete Containment Tendon Surveillance Program shall be reported in the Inservice Inspection Summary Report in accordance with 10 CFR 50.55a and ASME Section XI, 1992 Edition with the 1992 Addenda.

5.6.9 Steam Generator.(SG) Tube Inspection Report I

A report shall be submitted within 180 days after the initial entry into MODE 4 following completion of an inspection performed in accordance with Specification 5.5.9, Steam Generator (SG)

Program.

The report shall include:

a.

The scope of inspections performed on each SG,

b.

Active degradation mechanisms found,

c.

Nondestructive examination techniques utilized for each degradation mechanism,

d.

Location, orientation (if linear), and measured sizes (if available) of service induced indications,

e.

Number of tubes plugged or repaired during the inspection outage for each active degradation mechanism,

f.

Total number and percentage of tubes plugged or repaired to

date,

g.

The results of condition monitoring, including the results of tube pulls and in-situ testing,

h.

The effective plugging percentage for

.repairs in each SG, and

i.

Repair method utilized and the number each repair method.

all plugging and tube of~tubes repaired by BRAIDWOOD - UNITS 1 & 2 5.6-6 Amendment 144

v t e Letter Sequence Other
TAC:MC8966, Steam Generator Tube Integrity (Open) TAC:MC8967, Steam Generator Tube Integrity (Open) TAC:MC8968, Steam Generator Tube Integrity (Open) TAC:MC8969, Steam Generator Tube Integrity (Open)
Initiation Request Acceptance... Supplement Administration Withholding Request Acceptance, Withholding Request Acceptance Questions 4 August 2006 13 December 2006 28 March 2007 Answers 15 February 2007 Results Approval... Other: ML062780287, ML062780503, ML062990266, ML070810337, ML071000245, ML071210555
MONTHYEAR2007-02-15 [Table View]

ML071000245

Text

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