Text

Submittal of Technical Specification Bases Revisions
	ML061240288
Person / Time
Site:	Brunswick
Issue date:	04/13/2006
From:	Ivey R; Progress Energy Carolinas
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	BSEP 06-0044
	Download: ML061240288 (72)
	v • d • e

C4J Progress Energy APR 1 3 2006 SERIAL: BSEP 06-0044 U. S. Nuclear Regulatory Commission AITN: Document Control Desk W'ashington, DC 20555-0001

Subject:

Brunswick Steam Electric Plant, Unit Nos. 1 and 2 Docket Nos. 50-325 and 50-324/License Nos. DPR-71 and DPR-62 Submittal of Technical Specification Bases Revisions Ladies and Gentlemen:

In accordance with Technical Specification (TS) 5.5.10 for the Brunswick Steam Electric Plant (BSEP), Unit Nos. 1 and 2, Carolina Power & Light Company (CP&L), now doing business as Progress Energy Carolinas, Inc., is submitting Revisions 44, 45, and 46 to the B SEP, Unit 1 TS Bases and Revisions 42, 43, and 44 to the BSEP, Unit 2 TS Bases.

In addition, the TS Table of Contents are controlled by CP&L. For convenience, updated Table of Contents for Units 1 and 2 are included in Enclosure 5. Changes to the Table of Contents resulted from issuance of Amendment Nos. 239 and 267 for Units 1 and 2, respectively.

Please refer any questions regarding this submittal to Mr. Leonard R. Beller, Supervisor -

Licensing/Regulatory Programs, at (910) 457-2073.

Sincerely, aa'j' C O7 Randy C. Ivey Manager - Support Services Brunswick Steam Electric Plant Pr gress Energy Carolinas. Inc.

Brinswick Nuclear Plant PC.O Box 10429 So ithport, NC 28461

Document Control Desk BSEP 06-0044 / Page 2 WRM/wrm

Enclosures:

1. Summary of Revisions to Technical Specification Bases

2. Page Replacement Instructions

3. Unit 1 Technical Specification Bases Replacement Pages

4. Unit 2 Technical Specification Bases Replacement Pages

5. Unit 1 and 2 Technical Specification Table of Contents Replacement Pages

Document Control Desk B SEP 06-0044 / Page 3 cc: (with enclosures):

U. S. Nuclear Regulatory Commission, Region II A1TTN: Dr. William D. Travers, Regional Administrator Sam Nunn Atlanta Federal Center 61 Forsyth Street, SW, Suite 23T85 Atlanta, GA 30303-8931 U. S. Nuclear Regulatory Commission ATTN: Mr. Eugene M. DiPaolo, NRC Senior Resident Inspector 8470 River Road Southport, NC 28461-8869 U. S. Nuclear Regulatory Commission (Electronic Copy Only)

ATTN: Ms. Brenda L. Mozafari (Mail Stop OWFN 8G9) 11555 Rockville Pike Rockville, MD 20852-2738 Ms. Jo A. Sanford Chair - North Carolina Utilities Commission P.O. Box 29510 Raleigh, NC 27626-0510 Ms. Beverly 0. Hall, Section Chief Radiation Protection Section, Division of Environmental Health North Carolina Department of Environment and Natural Resources 3825 Barrett Drive Raleigh, NC 27609-7221

BSEP 06-0044 Page 1 Df 2 Summary of Revisions to Technical Specification (TS) Bases Affected Revisi Di1 Unit Date Implemented Title/Description 44 1

February 16,2006

Title:

HPCI[RCIC Surveillance Testing 42 2

==

Description:==

Revision 44 for Unit 1 and 42 for Unit 2 incorporated changes to Sections 3.3.5.1, 3.3.5.2, 3.3.6.:,

and 3.5.1 of the TS Bases to facilitate online testing of the High Pressure Coolant injection and Reactor Core Isolation Cooling Systems.

45 1

March 3, 2006

Title:

Main Steam Isolation Valve Leakage 43 2

Limit

==

Description:==

Revision 45 for Unit 1 and 43 for Unit 2 incorporated changes to Sections 3.3.7.1, 3.6.1.3, and 3.7.3 of the TS Bases associated with Amendment 239 for Unit I and Amendment 267 for Unit 2, issued on March 2, 2006. These amendments increase the allowable main steam isolation valve leakage rate and revise Control room Emergency Ventilation initiation logic.

46 1

March 24, 2006

Title:

Instrument Accuracy for Primary 44 2

Containment Leakage Rate Testing

==

Description:==

Revision 46 for Unit 1 and 44 for Unit 2 incorporated changes to Sections 3.6.1.1, 3.6.1.2, and 3.6.1.3 of the TS Bases associated with Amendment 238 for Unit 1 and Amendment 266 for I

Revision 44 for Unit 1 and Revision 42 for Unit 2 incorporated change package TSB-2006-01.

Revision 45 for Unit 1 and Revision 43 for Unit 2 incorporated change packages TSC-2005-05 and TRMI-2005-07.

Revision 46 for Unit I and Revision 44 for Unit 2 incorporated change packages TSC-2005-04 and TSB-2005-04.

BSEP 06-0044 Enclosure I Page 2 of 2 Summary of Revisions to Technical Specification (TS) Bases Affected Revision1 Unit Date Implemented Title/Description Unit 2, issued on February 8, 2006. These amendments revised TS 5.5.12, "Primary Containment Leakage Rate Testing Program," by removing an exception that allowed for compensation of flow meter instrument inaccuracies in accordance with ANSI/ANS-56.8-1987.

BSEP 06-0044 Page 1 of 2 Page Replacement Instructions - Unit 1 Remove Insert Unit 1 - Bases Book 1 Title Page, Revision 43 Title Page, Revision 46 LOEP-1, Revision 43 LOEP-1, Revision 46 LOEP-3, Revision 41 LOEP-3, Revision 44 LOEP-4, Revision 36 LOEP-4, Revision 45 B 3.3.5.1-31, Revision 36 B 3.3.5.1-31, Revision 44 B 3.3.5.2-11, Revision 31 B 3.3.5.2-11, Revision 44 B 3.3.6.1-29, Revision 31 B 3.3.6.1-29, Revision 44 B 3.3.7.1-1, Revision 31 B 3.3.7.1-1, Revision 45 B 3.3.7.1-2, Revision 31 B 3.3.7.1-2, Revision 45 B 3.3.7.1-3, Revision 31 B 3.3.7.1-3, Revision 45 B 3.3.7.1-4, Revision 31 B 3.3.7.1-4, Revision 45 Unit I - Bases Book 2 LOEP-1, Revision 43 LOEP-1, Revision 46 LOEP-2, Revision 41 LOEP-2, Revision 46 LOEP-3, Revision 41 LOEP-3, Revision 45 B 3.5.1-14, Revision 31 B 3.5.1-14, Revision 44 B 3.5.1-15, Revision 31 B 3.5.1-15, Revision 44 B 3.6.1.1-4, Revision 31 B 3.6.1.1-4, Revision 46 B 3.6.1.1-5, Revision 31 B 3.6.1.1-5, Revision 46 B 3.6.1.1-6, Revision 31 B 3.6.1.1-6, Revision 46 B 3.6.1.2-7, Revision 31 B 3.6.1.2-7, Revision 46 B 3.6.1.2-8, Revision 31 B 3.6.1.2-8, Revision 46 B 3.6.1.2-9, Revision 31 B 3.6.1.3-12, Revision 31 B 3.6.1.3-12, Revision 46 B 3.6.1.3-13, Revision 31 B 3.6.1.3-13, Revision 46 B 3.6.1.3-14, Revision 31 B 3.6.1.3-14, Revision 46 B 3.6.1.3-15, Revision 31

=

B 3.7.3-2, Revision 36 B 3.7.3-2, Revision 45

BSEP 06-0044 EnclosuLre 2 Page 2 of 2 Page Replacement Instructions - Unit 2 Remove Insert Unit 2 - Bases Book 1 Title Page, Revision 41 Title Page, Revision 44 LOEP-1, Revision 41 LOEP-1, Revision 44 LOEP-3, Revision 39 LOEP-3, Revision 42 LOEP-4, Revision 35 LOEP-4, Revision 43 B 3.3.5.1-31, Revision 33 B 3.3.5.1-31, Revision 42 B 3.3.5.2-11, Revision 30 B 3.3.5.2-11, Revision 42 B 3.3.6.1-30, Revision 30 B 3.3.6.1-30, Revision 42 B 3.3.7.1-1, Revision 30 B 3.3.7.1-1, Revision 43 B 3.3.7.1-2, Revision 30 B 3.3.7.1-2, Revision 43 B 3.3.7.1-3, Revision 30 B 3.3.7.1-3, Revision 43 B 3.3.7.1-4, Revision 30 B 3.3.7.1-4, Revision 43 Unit 2 - Bases Book 2 LOEP-1, Revision 41 LOEP-1, Revision 44 LOEP-2, Revision 39 LOEP-2, Revision 44 LOEP-3, Revision 39 LOEP-3, Revision 43 B 3.5.1-14, Revision 30 B 3.5.1-14, Revision 42 B 3.5.1-15, Revision 30 B 3.5.1-15, Revision 42 B 3.6.1.1-4, Revision 30 B 3.6.1.1-4, Revision 44 B 3.6.1.1-5, Revision 30 B 3.6.1.1-5, Revision 44 B 3.6.1.1-6, Revision 30 B 3.6.1.1-6, Revision 44 B 3.6.1.2-7, Revision 30 B 3.6.1.2-7, Revision 44 B 3.6.1.2-8, Revision 30 B 3.6.1.2-8, Revision 44 B 3.6.1.2-9, Revision 30 B 3.6.1.3-12, Revision 30 B 3.6.1.3-12, Revision 44 B 3.6.1.3-13, Revision 30 B 3.6.1.3-13, Revision 44 B 3.6.1.3-14, Revision 30 B 3.6.1.3-14, Revision 44 B 3.6.1.3-15, Revision 30 B 3.7.3-2, Revision 33 B 3.7.3-2, Revision 43

BSEP 06-0044 Unit 1 Technical Specification Bases Replacement Pages

Unit 1 - Bases Book 1 Replacement Pages

BASES TO THE FACILITY OPERATING LICENSE DPR-71 TECHNICAL SPECIFICATIONS FOR BRUNSWICK STEAM ELECTRIC PLANT UNIT I CAROLINA POWER & LIGHT COMPANY REVISION 46

LIST OF EFFECTIVE PAGES - BASES Page No.

Title Page Revision No.

46 List of Effective Pages - Book 1 LOEP-1 LOEP-2 LOEF'-3 LOEF'-4 46 36 44 45 i.

Hl 42 31 B 2.1.1-1 B 2.1.1-2 B 2.1.1-3 B 2.1.1-4 B 2.1.1-5 B 2.1.2-1 B 2.1.2-2 B 2.1.2-3 B 3.0-1 B 3.0-2 B 3.0-3 B 3.0-4 B 3.0-5 B 3.0-6 B 3.0-7 B 3.0-8 B 3.0-9 B 3.0-10 B 3.0-11 B 3.0-12 B 3.0-13 B 3.0-14 B 3.0-15 B 3.0-16 B 3.0-17 B 3.1.1-1 B 3.1.1-2 B 3.1.1-3 B 3.1.1-4 B 3.1.1-5 31 31 31 31 31 31 31 31 31 31 31 31 41 41 41 41 41 41 41 41 41 41 41 41 41 31 31 31 31 31 Page No.

B 3.1.1-6 B 3.1.2-1 B 3.1.2-2 B 3.1.2-3 B 3.1.2-4 B 3.1.2-5 B 3.1.3-1 B 3.1.3-2 B 3.1.3-3 B 3.1.3-4 B 3.1.3-5 B 3.1.3-6 B 3.1.3-7 B 3.1.3-8 B 3.1.3-9 B 3.1.4-1 B 3.1.4-2 B 3.1.4-3 B 3.1.4-4 B 3.1.4-5 B 3.1.4-6 B 3.1.4-7 B 3.1.5-1 B 3.1.5-2 B 3.1.5-3 B 3.1.5-4 B 3.1.5-5 B 3.1.6-1 B 3.1.6-2 B 3.1.6-3 B 3.1.6-4 B 3.1.6-5 B 3.1.7-1 B 3.1.7-2 B 3.1.7-3 B 3.1.7-4 B 3.1.7-5 B 3.1.7-6 B 3.1.8-1 B 3.1.8-2 B 3.1.8-3 B 3.1.8-4 B 3.1.8-5 Revision No.

31 31 31 31 31 31 31 31 31 31 31 31 31 31 31 31 31 31 31 42 31 31 31 31 31 31 31 31 31 31 31 31 34 31 31 31 31 34 31 37 37 31 31 I

I (continued)

Brunswick Unit 1 LOEP-1 FRevision 46 l

LIST OF EFFECTIVE PAGES - BASES (continued)

Page No.

Revision No.

Page No.

Revision No.

B 3.3.3.1-4 B 3.3.3.1-5 B 3.3.3.1-6 B 3.3.3.1-7 B 3.3.3.1-8 B 3.3.3.1-9 B 3.3.3.1-10 B 3.3.3.2-1 B 3.3.3.2-2 B 3.3.3.2-3 B 3.3.3.2-4 B 3.3.3.2-5 B 3.3.4.1-1 B 3.3.4.1-2 B 3.3.4.1-3 B 3.3.4.1-4 B 3.3.4.1-5 B 3.3.4.1-6 B 3.3.4.1-7 B 3.3.4.1-8 B 3.3,.4.1-9 B 3.',.5.1-1 B 3.3.5.1-2 B 3.3.5.1-3 B 3.3.5.1-4 B 3.3.5.1-5 B 3.3.5.1-6 B 3.',.5.1-7 B 3.3,.5.1-8 B 3.3.5.1-9 B 3.3.5.1-10 B 3.3',.5.1-11 B 3.3.5.1-12 B 3.3.5.1-13 B 3.3.5.1-14 B 3.3.5.1-15 B 3.3.5.1-16 31 31 40 41 40 40 40 41 41 41 41 41 31 31 31 31 31 31 31 31 31 31 31 31 36 31 31 31 31 31 31 31 31 36 36 36 36 B 3.3.5.1-17 B 3.3.5.1-18 B 3.3.5.1-19 B 3.3.5.1-20 B 3.3.5.1-21 B 3.3.5.1-22 B 3.3.5.1-23 B 3.3.5.1-24 B 3.3.5.1-25 B 3.3.5.1-26 B 3.3.5.1-27 B 3.3.5.1-28 B 3.3.5.1-29 B 3.3.5.1-30 B 3.3.5.1-31 B 3.3.5.2-1 B 3.3.5.2-2 B 3.3.5.2-3 B 3.3.5.2-4 B 3.3.5.2-5 B 3.3.5.2-6 B 3.3.5.2-7 B 3.3.5.2-8 B 3.3.5.2-9 B 3.3.5.2-10 B 3.3.5.2-11 B 3.3.6.1-1 B 3.3.6.1-2 B 3.3.6.1-3 B 3.3.6.1-4 B 3.3.6.1-5 B 3.3.6.1-6 B 3.3.6.1-7 B 3.3.6.1-8 B 3.3.6.1-9 B 3.3.6.1-10 B 3.3.6.1-11 B 3.3.6.1-12 B 3.3.6.1-13 B 3.3.6.1-14 31 36 36 31 31 31 31 31 31 31 31 31 31 31 44 31 31 31 31 31 31 31 31 31 31 44 31 31 31 31 31 31 31 32 31 31 31 31 31 31 I

I (continued)

Brunswick Unit I LOEP-3 Revision 44 l

LIST OF EFFECTIVE PAGES - BASES (continued)

Page No.

Revision No.

Page No.

Revision No.

B 3.3.6.1-15 B 3.3.6.1-16 B 3.3.6.1-17 B 3.3.6.1-18 B 3.3.6.1-19 B 3.3.6.1-20 B 3.3.6.1-21 B 3.3.6.1-22 B 3.3.6.1-23 B 3.3.6.1-24 B 3.3.6.1-25 B 3.3.6.1-26 B 3.3.6.1-27 B 3.3.6.1-28 B 3.3.6.1-29 B 3.3.6.1-30 B 3.3.6.1-31 B 3.3.6.2-1 B 3.3.6.2-2 B 3.3.6.2-3 B 3.3.6.2-4 B 3.3.6.2-5 B 3.3.6.2-6 B 3.3.6.2-7 B 3.3.6.2-8 B 3.3.6.2-9 B 3.3.6.2-10 B 3.3.6.2-11 B 3.3.7.1-1 B 3.3.7.1-2 B 3.3.7.1-3 B 3.3.7.1-4 B 3.3.7.1-5 B 3.3.7.1-6 B 3.3.7.1-7 B 3.3.7.2-1 B 3.3.7.2-2 B 3.3.7.2-3 B 3.3.7.2-4 B 3.3.7.2-5 B 3.3.7.2-6 31 31 31 31 31 31 31 31 31 31 31 31 31 31 44 31 31 31 31 31 31 31 31 31 31 31 31 31 45 45 45 45 31 31 36 31 31 31 31 31 31 B 3.3.7.2-7 B 3.3.8.1-1 B 3.3.8.1-2 B 3.3.8.1-3 B 3.3.8.1-4 B 3.3.8.1-5 B 3.3.8.1-6 B 3.3.8.1-7 B 3.3.8.2-1 B 3.3.8.2-2 B 3.3.8.2-3 B 3.3.8.2-4 B 3.3.8.2-5 B 3.3.8.2-6 B 3.3.8.2-7 31 31 31 31 31 31 31 31 31 31 31 31 31 31 31 Brunswick Unit 1 LOEP-4 Revision 45 l

ECCS Instrumenlation B 3.3.5.1 BASES SURVEILLANCE REQUIREMENTS (continued)

SR 3.3.5.1.5 The LOGIC SYSTEM FUNCTIONAL TEST demonstrates the OPERABILITY of the required initiation logic and simulated automatic operation for a specific channel. The system functional testing performed in LCO 3.5.1, LCO 3.5.2, LCO 3.8.1, and LCO 3.8.2 overlaps this Surveillance to complete testing of the assumed safety function.

Based on minimal assumed risk in performing this Surveillance with the reactor at power, the surveillance is not required to be performed during a refueling outage. Operating experience has demonstrated that these components will usually pass the SR when performed at the 24 month Frequency (originally based on the refueling cycle). Therefore, the Frequency is concluded to be acceptable from a reliability standpoint.

REFERENCES

1.

UFSAR, Section 5.2.

2.

3.

4.

5.

UFSAR, Section 6.3.

UFSAR, Chapter 15.

10 CFR 50.36(c)(2)(ii).

NEDC-31624P, Brunswick Steam Electric Plant Units 1 and 2 SAFER/GESTR-LOCA Loss-of-Coolant Accident Analysis (Revision 2), July 1990.

UFSAR, Section 9.2.6.2.

NEDC-30936-P-A, BWR Owners' Group Technical SpecificatiDn Improvement Methodology (With Demonstration for BWR ECC(S Actuation Instrumentation), Parts 1 and 2, December 1988.

6.

7.

Brunswick Unit 1 B 3.3.5.1-31 Revision No. 44 l

RCIC System Instrumentation B 3.3.5.2 BASES SURVEILLANCE REQUIREMENTS (continued)

SR 3.3.5.2.5 The LOGIC SYSTEM FUNCTIONAL TEST demonstrates the OPERABILITY of the required initiation logic for a specific channel and includes simulated automatic actuation of the channel. The system functional testing performed in LCO 3.5.3 overlaps this Surveillance to provide complete testing of the safety function.

Based on minimal assumed risk in performing this Surveillance with the reactor at power, the surveillance is not required to be performed during a refueling outage. Operating experience has demonstrated that these components will usually pass the SR when performed at the 24 month Frequency (originally based on the refueling cycle). Therefore, the Frequency is concluded to be acceptable from a reliability standpoint.

REFERENCES

1.

10 CFR 50.36(c)(2)(ii).

2.

GENE-770-06-2P-A, Bases for Changes to Surveillance Test Intervals and Allowed Out-of-Service Times for Selected Instrumentation Technical Specifications, December 1992.

Brunswick Unit 1 B 3.3.5.2-1 1 Revision No. 44 l

Primary Containment Isolation Instrumentation B 3.3.6.1 BASES SUR\\/EILLANCE SR 3.3.6.1.4 and SR 3.3.6.1.6 (continued)

REQUIREMENTS CALIBRATION leaves the channel adjusted to account for instrument drifts between successive calibrations consistent with the plant specific setpoint methodology.

The Frequency of SR 3.3.6.1.4 is based on the assumption of a 92 day calibration interval in the determination of the magnitude of equipment drift in the setpoint analysis. The Frequency of SR 3.3.6.1.6 is based on the assumption of a 24 month calibration interval in the determination of the magnitude of equipment drift in the setpoint analysis.

SR 3.3.6.1.7 The LOGIC SYSTEM FUNCTIONAL TEST demonstrates the OPERABILITY of the required isolation logic for a specific channel and includes simulated automatic operation of the channel. The system functional testing performed on PCIVs in LCO 3.6.1.3 overlaps this Surveillance to provide complete testing of the assumed safety function.

The 24 month Frequency was developed considering it is prudent that this Surveillance be performed during a unit outage. Operating experience has demonstrated that these components will pass this Surveillance when performed at the 24 month Frequency. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint.

SR 3.3.6.1.8 This SR ensures that the individual channel response times are less than or equal to the maximum values assumed in the accident analysis.

Testing is performed only on channels where the assumed response time does not correspond to the diesel generator (DG) start time. For channels assumed to respond within the DG start time, sufficient margin exists in the 10 second start time when compared to the typical channel response time (milliseconds) so as to assure adequate response without a specific measurement test (Ref. 9).

(continued)

Brunswick Unit I B 3.3.6.1-29 Revision No. 44 l

CREV System Instrumentation B 3.3.7.1 B 3.3 INSTRUMENTATION B 3.3.7.1 Control Room Emergency Ventilation (CREV) System Instrumentation BASES BACKGROUND The CREV System is designed to provide a radiologically controlled environment to ensure the habitability of the control room for the safety of control room operators under all plant conditions. Two independent CREV subsystems are each capable of fulfilling the stated safety function.

The instrumentation and controls for the CREV System automatically initiate action to pressurize the main control room (MCR) to minimize the consequences of radioactive material in the control room environment:.

In the event of a loss of coolant accident (LOCA), the Unit 1 Secondary Containment Isolation-CREV Auto-Start signal will automatically start the CREV System in the radiation/smoke protection mode. The automatic CREV actuation is generated from the portion of the secondary containment isolation logic associated with Reactor Vessel Water Level-Low Level 2 and the Drywell Pressure-High. CREV automatic actuation will not occur as a result of secondary containment isolations due to Reactor Building Exhaust Radiation-High signals. The Control Building Air Intake-Radiation High function provides protection for non-LOCA events. In the event of a Control Building Air Intake Radiation-High signal, the CREV System is automatically started in the radiation/smoke protection mode. Air is then recirculated through the charcoal filter, and sufficient outside air is drawn in through the normal intake to maintain the MCR slightly pressurized with respect to outside atmosphere.

The CREV System instrumentation has two trip systems, either of wh'ch can initiate the CREV System. The Reactor Vessel Water Level-Low Level 2 and the Drywell Pressure-High signals are arranged in such a manner that opening of either A and B, or C and D, relay contacts of either the Reactor Vessel Water Level-Low Level 2 or the Drywell Pressure-High will provide the CREV start signal. Thus, the automatic CREV initiation, using signals from the secondary containment isolation logic, provides redundant/diverse protection for control room operators in the event of a LOCA. The Control Building Air Intake Radiation-High Function is arranged in a one-out-of-two logic for each trip system. The channels include electronic equipment (e.g., trip units) that compares measured input signals with pre-established setpoints. When the selpoint is exceeded, the channel output relay actuates, which then outputs a CREV System initiation signal to the initiation logic.

APPLICABLE The ability of the CREV System to maintain the habitability of the MCR is SAFETY ANALYSES, explicitly assumed for the design basis accident as discussed in the LCO, and UFSAR safety analyses (Refs. 1 and 2). CREV System operation APPLICABILITY ensures that the radiation exposure of control room personnel, through (continued)

Brunswick Unit 1 B 3.3.7.1-1 Revision No. 45 1

CREV System Instrumentation B 3.3.7.1 BASE:S APPLICABLE the duration of any one of the postulated accidents, does not exceed the SAFETY ANALYSES, limits set by GDC 19 of 10 CFR 50, Appendix A.

LCO, and APPLICABILITY CREV System instrumentation satisfies Criterion 3 of (continued) 10 CFR 50.36(c)(2)(ii) (Ref. 3).

The OPERABILITY of the CREV System instrumentation is dependent upon the OPERABILITY of the individual instrumentation Channel Functions specified in Table 3.3.7.1-1. The Functions must have a required number of OPERABLE channels, with their setpoints within tie specified Allowable Values provided in Tables 3.3.7.1-1 and 3.3.6.2-1.

The actual setpoints are calibrated consistent with applicable setpoint methodology assumptions.

Allowable Values are specified for each Function specified in Tables 3.3.7.1-1 and 3.3.6.2-1. Trip setpoints are specified in the setpoint calculations. The setpoints are selected to ensure that the trip settings do not exceed the Allowable Value between successive CHANNEL CALIBRATIONS. Operation with a trip setting less conservative than the trip setpoint, but within its Allowable Value, is acceptable. A channel is inoperable if its actual trip setting is not within its required Allowable Value. Trip setpoints are those predetermined values of output at wh ch an action should take place. The setpoints are compared to the actual process parameter (e.g., control building air intake radiation), and when the measured output value of the process parameter exceeds the setpoint, the associated device (e.g., trip unit) changes state. The analytic limits are derived from the limiting values of the process parameters obtained from the safety analysis. The trip setpoints are determined from the analytic limits, corrected for defined process, calibration, and instrument errors. The Allowable Values are then determined, based on the trip setpoint values, by accounting for calibration based errors. These calibration based instrument errors are limited to instrument drift, errors associated with measurement and test equipment, and calibration tolerance of loop components. The trip setpoints and Allowable Values determined in this manner provide adequate protection because instrumentation uncertainties, process effects, calibration tolerances, instrument drift, and severe environment errors (for channels that must function in harsh environments as defined by 10 CFR 50.49) are accounted for and appropriately applied for the instrumentation.

The specific Applicable Safety Analyses, LCO, and Applicability discussions are listed below on a Function by Function basis.

1. Control Building Air Intake Radiation-Hiqh The control building air intake radiation monitors measure radiation levels in the control building air intake plenum. A high radiation level may pose a threat to MCR personnel; thus, automatically initiating the CREV Syslem.

(continued)

Brunswick Unit 1 B 3.3.7.1-2 Revision No. 45 1

CREV System Instrumentation B 3.:3.7.1 BASES APPLICABLE The Control Building Air Intake Radiation-High Function consists of iwo SAFETY ANALYSES, independent monitors. Two channels per trip system of Control Building LCO, and Air Intake Radiation-High are available and are required to be APPLICABILITY OPERABLE to ensure that no single instrument failure can preclude (continued)

CREV System initiation. The Allowable Value was selected to ensure protection of the control room personnel.

The Control Building Air Intake Radiation-High Function is required to be OPERABLE in MODES 1, 2, and 3 and during OPDRVs and movement of recently irradiated fuel assemblies in the secondary containment, to ensure that control room personnel are protected during a LOCA, fuel handling event, or vessel draindown event. During MODES 4 and 5, when these specified conditions are not in progress (e.g., OPDRVs), the probability of a LOCA, main steam line break accident, or control rod drop accident is low; thus, the Function is not required. Also due to radioactive decay, this Function is only required to initiate the CREV System during fuel handling accidents involving handling recently irradiated fuel (i.e., fuel that has occupied part of a critical reactor core within the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ).

2. Unit 1 Secondary Containment Isolation-CREV Auto-Start The Unit 1 Secondary Containment Isolation-CREV Auto-Start Function provides post-LOCA operator protection. Since Reactor Vessel Water Level-Low Level 2 and Drywell Pressure-High provide primary indication of a LOCA, only secondary containment isolations resulting from these signals provide CREV Auto-Start. The Reactor Vessel Water Level-Low Level 2 and the Drywell Pressure-High signals are arranged in such a manner that opening of either A and B, or C and D, relay contacts of either the Reactor Vessel Water Level-Low Level 2 or the Drywell Pressure-High will provide the CREV start signal. Thus, automatic CREV initiation, using signals from the secondary containment isolation logic, provides redundant/diverse protection for control room operators in the event of a LOCA.

The Allowable Values for the Secondary Containment Isolation Instrumentation are provided in Table 3.3.6.2-1.

The Unit 1 Secondary Containment Isolation-CREV Auto-Start Function is required to be OPERABLE in MODES 1,2, and 3 to ensure that control room personnel are protected in the event of a LOCA.

ACTIONS A Note has been provided to modify the ACTIONS related to CREV System instrumentation channels. Section 1.3, Completion Times, specifies that once a Condition has been entered, subsequent divisions, subsystems, components, or variables expressed in the Condition, discovered to be inoperable or not within limits, will not result in separate entry into the Condition. Section 1.3 also specifies that Required Actions (continued)

Brunswick Unit I B 3.3.7.1-3 Revision No. 45 l

CREV System Instrumenlation B 3.3.7.1 BASES ACTIONS of the Condition continue to apply for each additional failure, with (continued)

Completion Times based on initial entry into the Condition. However, the Required Actions for inoperable CREV System instrumentation channels provide appropriate compensatory measures for separate inoperable channels. As such, a Note has been provided that allows separate Condition entry for each inoperable CREV System instrumentation channel.

A.1 Because of the redundancy of sensors available to provide initiation signals and the redundancy of the CREV System design, an allowable out of service time of 7 days is provided to permit restoration of any inoperable channel to OPERABLE status. This out of service time is Dnly acceptable provided the affected Function is still maintaining CREV System initiation capability (refer to Required Action B.1 Bases). If the Function is not maintaining CREV System initiation capability, Condition B must be entered.

If the inoperable channel cannot be restored to OPERABLE status within the 7 day allowable out of service time, one CREV subsystem must be placed in the radiation/smoke protection mode of operation per Required Action A.1. The method used to place the CREV subsystem in operation must provide for automatically re-initiating the subsystem upon restoration of power following a loss of power to the CREV subsystem. Placing one CREV subsystem in the radiation/smoke protection mode of operation provides a suitable compensatory action to ensure that the automatic radiation protection function of the CREV System is not lost.

B.1 Required Action B.1 is intended to ensure that appropriate action is taken if multiple, inoperable, untripped channels result in the affected Function not maintaining CREV System initiation capability. The Function is considered to be maintaining CREV System initiation capability when sufficient channels are OPERABLE or in trip such that one trip system will generate an initiation signal for one CREV subsystem from the FunctiDn on a valid signal. For the Control Building Air Intake Radiation-High Function, this would require one trip system to have one channel OPERABLE or in trip. For the Unit I Secondary Containment Isolation-CREV Auto Start Reactor Vessel Water Level-Low Level 2 or Drywe II Pressure-High, this would require all channels to be operable or the inoperable channels in the tripped condition. With CREV System initiation capability not maintained, one CREV subsystem must be placed in the radiation/smoke protection mode of operation per Required Action B. l to ensure that control room personnel will be protected in the event of a Design Basis Accident. The method used to place the CREV subsystem in operation must provide for automatically re-initiating the subsystem upon restoration of power following a loss of power to the CREV subsystem.

(continued)

Brunswick Unit 1 B 3.3.7.1-4 Revision No. 45 l

Unit 1 - Bases Book 2 Replacement Pages

LIST OF EFFECTIVE PAGES - BASES Page No.

Revision No.

List of Effective Pages - Book 2 LOEP-1 LOEF'-2 LOEF'-3 LOEF'-4 LOEF'-5 46 46 45 31 31 Paqe No.

B 3.4.7-5 B 3.4.8-1 B 3.4.8-2 B 3.4.8-3 B 3.4.8-4 B 3.4.8-5 B 3.4.9-1 B 3.4.9-2 B 3.4.9-3 B 3.4.9-4 B 3.4.9-5 B 3.4.9-6 B 3.4.9-7 B 3.4.9-8 B 3.4.9-9 B 3.4.10-1 B 3.4.10-2 i

31 31 B 3.4.1-1 B 3.4.1-2 B 3.4.1-3 B 3.4.1-4 B 3.4.1-5 B 3.4.1-6 B 3.4.2-1 B 3.4.2-2 B 3.4.2-3 B 3.4.2-4 B 3.4.3-1 B 3.4.3-2 B 3.4.3-3 B 3.4.3-4 B 3.4.4-1 B 3.4.4-2 B 3.4.4-3 B 3.4.4-4 B 3.4.4-5 B 3.4.5-1 B 3.4.5-2 B 3.4.5-3 B 3.4.5-4 B 3.4.6-1 B 3.4.6-2 B 3.4.6-3 B 3.4.7-1 B 3.4.7-2 B 3.4.7-3 B 3.4.7-4 31 31 31 31 31 31 31 31 31 31 31 31 31 31 31 31 31 31 31 31 31 43 41 41 41 41 31 31 41 41 B 3.5.1-1 B 3.5.1-2 B 3.5.1-3 B 3.5.1-4 B 3.5.1-5 B 3.5.1-6 B 3.5.1-7 B 3.5.1-8 B 3.5.1-9 B 3.5.1-10 B 3.5.1-11 B 3.5.1-12 B 3.5.1-13 B 3.5.1-14 B 3.5.1-15 B 3.5.1-16 B 3.5.1-17 B 3.5.2-1 B 3.5.2-2 B 3.5.2-3 B 3.5.2-4 B 3.5.2-5 B 3.5.2-6 B 3.5.3-1 B 3.5.3-2 Revision No.

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Brunswick Unit 1 LOEP-1 Revisimn 46 l

LIST OF EFFECTIVE PAGES - BASES (continued)

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B 3.5.3-3 B 3.5.3-4 B 3.5.3-5 B 3.5.3-6 B 3.5.3-7 B 3.6.1.1-1 B 3.6.1.1-2 B 3.6.1.1-3 B 3.6.1.1-4 B 3.6.1.1-5 B 3.6.1.1-6 B 3.6.1.2-1 B 3.6.1.2-2 B 3.6.1.2-3 B 3.6.1.2-4 B 3.6.1.2-5 B 3.6.1.2-6 B 3.6.1.2-7 B 3.6.1.2-8 B 3.6.1.3-1 B 3.6.1.3-2 B 3.6.1.3-3 B 3.6.1.3-4 B 3.6.1.3-5 B 3.6.1.3-6 B 3.6.1.3-7 B 3.6.1.3-8 B 3.6.1.3-9 B 3.6.1.3-10 B 3.6.1.3-11 B 3.6.1.3-12 B 3.6.1.3-13 B 3.6.1.3-14 B 3.6.1.4-1 B 3.6.1.4-2 B 3.6.1.4-3 B 3.6.1.5-1 B 3.6.1.5-2 B 3.6.1.5-3 B 3.6.1.5-4 B 3.6.1.5-5 B 3.6.1.5-6 31 31 31 31 31 31 31 31 46 46 46 31 31 31 31 31 31 46 46 31 31 31 31 31 31 31 31 31 31 31 46 46 46 31 31 31 31 31 31 31 31 31 B 3.6.1.5-7 B 3.6.1.5-8 B 3.6.1.5-9 B 3.6.1.6-1 B 3.6.1.6-2 B 3.6.1.6-3 B 3.6.1.6-4 B 3.6.1.6-5 B 3.6.1.6-6 B 3.6.2.1-1 B 3.6.2.1-2 B 3.6.2.1-3 B 3.6.2.1-4 B 3.6.2.1-5 B 3.6.2.2-1 B 3.6.2.2-2 B 3.6.2.2-3 B 3.6.2.3-1 B 3.6.2.3-2 B 3.6.2.3-3 B 3.6.2.3-4 B 3.6.3.1-1 B 3.6.3.1-2 B 3.6.3.1-3 B 3.6.3.2-1 B 3.6.3.2-2 B 3.6.3.2-3 B 3.6.3.2-4 B 3.6.3.2-5 B 3.6.4.1-1 B 3.6.4.1-2 B 3.6.4.1-3 B 3.6.4.1-4 B 3.6.4.1-5 B 3.6.4.2-1 B 3.6.4.2-2 B 3.6.4.2-3 B 3.6.4.2-4 B 3.6.4.2-5 B 3.6.4.2-6 B 3.6.4.3-1 B 3.6.4.3-2 B 3.6.4.3-3 31 36 31 31 31 31 31 31 31 31 31 31 31 31 31 31 31 31 41 41 41 31 31 31 31 31 41 41 41 31 31 31 31 31 31 31 31 31 31 31 31 31 31 I

1 (continued)

Brunswick Unit 1 LOEP-2 Revision 46 l

LIST OF EFFECTIVE PAGES - BASES (continued)

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B 3.6.4.3-4 B 3.6.4.3-5 B 3.6.4.3-6 Revision No.

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31 31 31 B 3.7.7-1 B 3.7.7-2 B 3.7.7-3 Revision No.

31 31 31 B 3.7.1-1 B 3.7.1-2 B 3.7.1-3 B 3.7.1-4 B 3.7.1-5 B 3.7.2-1 B 3.7.2-2 B 3.7.2-3 B 3.7.2-4 B 3.7.2-5 B 3.7.2-6 B 3.7.2-7 B 3.7.2-8 B 3.7.2-9 B 3.7.2-10 B 3.7.2-11 B 3.7.2-12 B 3.7.2-13 B 3.7.2-14 B 3.7.3-1 B 3.7.3-2 B 3.7.3-3 B 3.7.3-4 B 3.7.3-5 B 3.7.3-6 B 3.7.3-7 B 3.7.4-1 B 3.7.4-2 B 3.7.4-3 B 3.7.4-4 B 3.7.4-5 B 3.7.5-1 B 3.7.5-2 B 3. 7.5-3 B 3.7.6-1 B 3.7.6-2 B 3.7.6-3 B 3.7.6-4 31 31 41 41 41 31 31 31 31 31 31 31 31 31 31 31 31 31 31 31 45 31 31 31 31 36 31 31 31 31 31 31 31 31 31 31 31 31 B 3.8.1-1 B 3.8.1-2 B 3.8.1-3 B 3.8.1-4 B 3.8.1-5 B 3.8.1-6 B 3.8.1-7 B 3.8.1-8 B 3.8.1-9 B 3.8.1-10 B 3.8.1-11 B 3.8.1-12 B 3.8.1-13 B 3.8.1-14 B 3.8.1-15 B 3.8.1-16 B 3.8.1-17 B 3.8.1-18 B 3.8.1-19 B 3.8.1-20 B 3.8.1-21 B 3.8.1-22 B 3.8.1-23 B 3.8.1-24 B 3.8.1-25 B 3.8.1-26 B 3.8.1-27 B 3.8.1-28 B 3.8.1-29 B 3.8.1-30 B 3.8.1-31 B 3.8.1-32 B 3.8.1-33 B 3.8.1-34 B 3.8.2-1 B 3.8.2-2 B 3.8.2-3 B 3.8.2-4 B 3.8.2-5 31 31 36 41 41 31 31 31 31 31 31 31 31 31 31 31 31 31 31 31 31 31 31 31 31 31 31 38 31 31 31 31 36 31 31 31 31 31 31 I

(continued)

Brunswick Unit 1 LOEP-3 Revision 45 l

ECCS-Operating B 3.5.1 BASES SURVEILLANCE SR 3.5.1.6, SR 3.5.1.7, and SR 3.5.1.8 (continued)

REQUIREMENTS Therefore, SR 3.5.1.7 and SR 3.5.1.8 are modified by Notes that state the Surveillances are not required to be performed until 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months after the reactor steam pressure is adequate to perform the test.

The 92 day Frequency for SR 3.5.1.6 and SR 3.5.1.7 is consistent with the Inservice Testing Program requirements. The 24 month Frequency for SR 3.5.1.8 is based on the need to perform the Surveillance under the conditions that apply just prior to or during a startup from a plant outage.

Operating experience has demonstrated that these components will usually pass the SR when performed at the 24 month Frequency, which is based on the refueling cycle. Therefore, the Frequency is considered to be acceptable from a reliability standpoint.

SR 3.5.1.9 The ECCS subsystems are required to actuate automatically to perform their design functions. This Surveillance verifies that, with a required system initiation signal (actual or simulated), the automatic initiation logic of HPCI, CS, and LPCI will cause the systems or subsystems to operate as designed, including actuation of the system throughout its emergency operating sequence, automatic pump startup and actuation of all automatic valves to their required positions. This SR also ensures that the HPCI System will automatically restart on an RPV low water level signal received subsequent to an RPV high water level trip and that the suction is automatically transferred from the CST to the suppression pool on a CST low level signal or a suppression pool high water level signal.

The LOGIC SYSTEM FUNCTIONAL TEST performed in LCO 3.3.5.1, "ECCS Instrumentation," overlaps this Surveillance to provide complete testing of the assumed safety function.

Based on minimal assumed risk in performing this Surveillance with the reactor at power, the surveillance is not required to be performed during a refueling outage. Operating experience has demonstrated that these components will usually pass the SR when performed at the 24 month Frequency (originally based on the refueling cycle). Therefore, the Frequency is concluded to be acceptable from a reliability standpoint.

(continued)

Brunswick Unit 1 B 3.5.1-14 Revision No. 44 l

ECCS-Operating B 3.5.1 BASES SURVEILLANCE SR 3.5.1.9 (continued)

REQUIREMENTS This SR is modified by a Note that excludes vessel injection/spray during the Surveillance. Since all active components are testable and full flow can be demonstrated by recirculation through the test line, coolant injection into the RPV is not required during the Surveillance.

SR 3.5.1.10 The ADS designated SRVs are required to actuate automatically upon receipt of specific initiation signals. A system functional test is performed to demonstrate that the mechanical portions of the ADS function (i.e.,

solenoids) operate as designed when initiated either by an actual or simulated initiation signal, causing proper actuation of all the required components. SR 3.5.1.11 and the LOGIC SYSTEM FUNCTIONAL TEST performed in LCO 3.3.5.1 overlap this Surveillance to provide complete testing of the assumed safety function.

The 24 month Frequency is based on the need to perform the Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power. Operating experience has demonstrated that these components will usually pass the SR when performed at the 24 month Frequency, which is based on the refueling cycle. Therefore, the Frequency is considered to be acceptable from a reliability standpoint.

This SR is modified by a Note that excludes valve actuation since the valves are individually tested in accordance with SR 3.5.1.1 1. This also prevents an RPV pressure blowdown.

(continued)

Brunswick Unit 1 B 3.5.1-15 Revision No. 44 l

Primary Containment B 3.5.1.1 BASES (continued)

SURVEILLANCE SR 3.6.1.1.1 REQIJIREMENTS Maintaining the primary containment OPERABLE requires compliance with the visual examinations and leakage rate test requirements of the Primary Containment Leakage Rate Testing Program. The Primary Containment Leakage Rate Testing Program has been established in accordance with 10 CFR 50.54(o) to implement the requirements of 10 CFR Part 50, Appendix J, Option B (Ref. 3). The Primary Containment Leakage Rate Testing Program also conforms with Regulatory Guide 1.163 (Ref. 6) and Nuclear Energy Institute (NEI) 94-01 (Ref. 7) except for the following:

a.

BNP may use the criteria and Total Time method specified in Bechtel Topical Report BN-TOP-1 (Ref. 8) for calculating the l

primary containment leakage during reduced duration Type A testing. This is an exemption from the requirements of 10 CFR 50 Appendix J (Ref. 3) which, in accordance with NEI 94-01 (Ref. 7),

requires the methods for calculating primary containment leakage described in ANSIJANS 56.8-1994 (Ref. 9). The basis for this exemption is described in References 10 and 11.

b.

Type C testing is not required for the hydrogen and oxygen monitor isolation valves. This is an exemption from the requirements of 10 CFR 50 Appendix J (Ref. 3). The basis for this exemption is described in Reference 12.

Failure to meet air lock leakage limits (SR 3.6.1.2.1) or main steam isolation valve leakage (SR 3.6.1.3.9) does not necessarily result in a failure of this SR. The impact of the failure to meet SR 3.6.1.2.1 must be evaluated against the Type A, B, and C acceptance criteria of the Primary Containment Leakage Rate Testing Program, and failure to meet SR 3.6.1.3.9 must be evaluated against Type A acceptance criteria of the Primary Containment Leakage Rate Testing Program.

As left leakage prior to the first startup after performing required leakage testing is required to be < 0.6 La for combined Type B and C leakage, and

< 0.75 La for overall Type A leakage. At all other times between required leakage rate tests, the acceptance criteria is based on an overall Type A leakage limit of

1.0 La. At < 1.0 La the offsite dose consequences are bounded by the assumptions of the safety analysis. The Frequency is required by the Primary Containment Leakage Rate Testing Program.

(continued)

Brunswick Unit 1 B 3.6.1.1-4 Revision 1\\;o. 46 l

Primary Containment B 3.1.1.1 BASES SURVEILLANCE REQUIREMENTS (continued)

SR 3.6.1.1.2 Maintaining the pressure suppression function of primary containment.

requires limiting the leakage from the drywell to the suppression chamber.

Thus, if an event were to occur that pressurized the drywell, the steam would be directed through the downcomers into the suppression pool.

This SR measures drywell to suppression chamber differential pressure during a 10 minute period to ensure that the leakage paths that would bypass the suppression pool (downcomers) are within allowable limits.

Satisfactory performance of this SR can be achieved by establishing a known differential pressure between the drywell and the suppression chamber and verifying that the differential pressure between the suppression chamber and the drywell does not decrease by more than 0.25 inch of water per minute over a 10 minute period. The leakage test is performed every 24 months. The 24 month Frequency was developed considering it is prudent that this Surveillance be performed during a unit outage and also in view of the fact that component failures that might have affected this test are identified by other primary containment SR3.

REFERENCES

1.

UFSAR, Section 6.2.

2.

UFSAR, Section 15.6.

3.

10 CFR 50, Appendix J, Option B.

4.

NEDC-33039P, Safety Analysis Report for Brunswick Steam Electric Plant Units 1 and 2, Extended Power Uprate, August 2001.

5.

10 CFR 50.36(c)(2)(ii).

6.

NRC Regulatory Guide 1.163, Performance-Based Containment Leak-Rate Testing Program, September 1995.

7.

Nuclear Energy Institute (NEI) 94-01, Industry Guideline for Implementing Performance-Based Option of 10 CFR 50 Appendix J, July 26, 1995.

8.

Bechtel Topical Report BN-TOP-1, Revision 1, November 1, 1972.

9.

ANSI/ANS 56.8-1994.

(continued)

Brunswick Unit 1 B 3.6.1.1-5 Revision Nlo. 46 l

Primary Containment B 3.6.1.1 BASES REFERENCES (continued)

10.

NRC SER; Issuance of Amendment No. 181 to Facility Operating License No. DPR-71 and Amendment No. 213 to Facility Operating License No. DPR-62 Regarding 10 CFR 50 Appendix J, Option B - Brunswick Steam Electric Plant, Units 1 and 2 (BSEP 95-0316) (TAC Nos. M93679 and M93680); dated February 1, 1996.

11.

NRC SER, Exemption from the Requirements of Appendix J for Brunswick Steam Electric Plant, Units 1 and 2, dated February 17, 1988.

12.

NRC SER, Technical Exemption from the Requirements of Appendix J, dated May 12, 1987.

I Brunswick Unit 1 B 3.6.1.1-6 Revision No. 46 l

Primary Containment Air Lock B 3.13.1.2 BASES (continued)

SURVEILLANCE REQUIREMENTS SR 3.6.1.2.1 Maintaining the primary containment air lock OPERABLE requires compliance with the leakage rate test requirements of the Primary Containment Leakage Rate Testing Program. The Primary Containmant Leakage Rate Testing Program has been established in accordance with 10 CFR 50.54(o) to implement the requirements of 10 CFR Part 50, Appendix J, Option B (Ref. 4), and conforms with Regulatory Guide 1.163 (Ref. 5) and Nuclear Energy Institute (NEI) 94-01 (Ref. 6) except for the following:

a.

The local leak rate testing requirements of the primary containment air lock doors may be modified to perform the tests at a pressure less than Pa following replacement of the air lock door seals. This is an exception from the requirements of NEI 94-01 (Ref. 6). The basis for this exception is described in Reference 7.

I I

This SR reflects the leakage rate testing requirements with respect to air lock leakage (Type B leakage tests). The acceptance criteria were established as a small fraction of the total allowable primary containment leakage. The periodic testing requirements verify that the air lock leakage does not exceed the allowed fraction of the overall primary containment leakage rate. The Frequency is required by the Primary Containment Leakage Rate Testing Program.

The SR has been modified by two Notes. Note 1 states that an inoperable air lock door does not invalidate the previous successful performance of the overall air lock leakage test. This is considered reasonable since either air lock door is capable of providing a fission product barrier in the event of a DBA. Note 2 has been added to this SR, requiring results to be evaluated against the acceptance criteria which are applicable to SR 3.6.1.1.1. This ensures that air lock leakage is properly accounted for in determining the combined Type B and C primary containment leakage rate.

(continued)

Brunswick Unit 1 B 3.6.1.2-7 Revision No. 46 l

Primary Containment Air Lock B 3.6.1.2 BASES SURVEILLANCE REQIJIREMENTS (continued)

SR 3.6.1.2.2 The air lock interlock mechanism is designed to prevent simultaneous.

opening of both doors in the air lock. Since both the inner and outer doors of the air lock are designed to withstand the maximum expected post accident primary containment pressure, closure of either door will support primary containment OPERABILITY. Thus, the interlock feature supports primary containment OPERABILITY while the air lock is being used fcr personnel transit in and out of the containment. Periodic testing of this interlock demonstrates that the interlock will function as designed and that simultaneous inner and outer door opening will not inadvertently occur.

Due to the purely mechanical nature of this interlock, and given that the interlock mechanism is not normally challenged when the primary containment airlock door is used for entry and exit (procedures require strict adherence to single door opening), this test is only required to be performed every 24 months. The 24 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage, and the potential for loss of primary containment OPERABILITY if the Surveillance were performed with the reactor at power. The 24 month Frequency for the interlock is justified based on generic operating experience. The Frequency is based on engineering judgment and is considered adequate given that the interlock is not challenged during use of the air lock.

REFERENCES

1.

UFSAR, Section 3.8.2.4.3.2.

2.

NEDC-33039P, Safety Analysis Report for Brunswick Units 1 and 2 Extended Power Uprate, August 2001.

3.

4.

10 CFR 50.36(c)(2)(ii).

10 CFR 50, Appendix J, Option B.

5.

NRC Regulatory Guide 1.163, Performance-Based Containment Leak-Rate Testing Program, September 1995.

6.

Nuclear Energy Institute (NEI) 94-01, Industry Guideline for Implementing Performance-Based Option of 10 CFR 50 Appendix J, July 26, 1995.

7.

NRC SER, Brunswick 1 & 2 - Amendments No. 10 and 36 to Operating Licenses Revising Technical Specifications to Grart Exemptions from Specific Requirements of 10 CFR 50 Appendix J, dated November 8,1977.

I Brunswick Unit 1 B 3.6.1.2-8 Revision No. 46 l

PCIIVs B 3.13.1.3 BASES SURVEILLANCE SR 3.6.1.3.6 (continued)

REQUIREMENTS complete testing of the safety function. The 24 month Frequency was developed considering it is prudent that this Surveillance be performed only during a unit outage since isolation of penetrations would eliminate cooling water flow and disrupt the normal operation of many critical components. Operating experience has demonstrated that these components will pass this Surveillance when performed at the 24 month Frequency. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint.

SR 3.6.1.3.7 This SR requires a demonstration that a representative sample of reactor instrumentation line excess flow check valves (EFCVs) is OPERABLE by verifying that the valves actuate to the isolation position on an actual or simulated instrument line break signal. This may be accomplished by cycling the EFCVs through one complete cycle of full travel. The representative sample consists of an approximately equal number of EFCVs, such that each EFCV is tested at least once every 10 years (nominal). In addition, the EFCVs in the samples are representative of the various plant configurations, models, sizes, and operating environments.

This ensures that any potentially common problem with a specific type or application of EFCV is detected at the earliest possible time. This SR provides assurance that the instrumentation line EFCVs will perform so that predicted radiological consequences will not be exceeded during a postulated instrument line break event. The 24 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power. Operating experience has demonstrated that these components will pass this Surveillance when performed at the 24 month Frequency. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint.

The nominal 10-year interval is based on performance testing as discussed in NEDO-32977-A (Ref. 5). Furthermore, any EFCV failures will be evaluated to determine if additional testing in that test interval is warranted to ensure overall reliability is maintained. Operating experience has demonstrated that these components are highly reliable and that failures to isolate are very infrequent. Therefore, testing of a representative sample was concluded to be acceptable from a reliability standpoint.

(continued)

Brunswick Unit 1 B 3.6.1.3-12 Revision No. 46 I

PCIVs B 3.6.1.3 BASES SURVEILLANCE REQUIREMENTS (continued)

SR 3.6.1.3.8 The TIP shear isolation valves are actuated by explosive charges. Art in place functional test is not possible with this design. The explosive squib is removed and tested to provide assurance that the valves will actuale when required. The replacement charge for the explosive squib shall be from the same manufactured batch as the one fired or from another batch that has been certified by having one of the batch successfully fired. The Frequency of this SR is in accordance with the requirements of the Inservice Testing Program.

SR 3.6.1.3.9 The analyses in References 2, 6, 7, and 8 are based on leakage that is less than the specified leakage rate. Leakage through each main steam line must be < 100 scfh when tested at 2 Pt (25 psig). The combined leakage rate for all four mains steam lines must be < 150 scfh when tested at > 25 psig in accordance with the Primary Containment Leakage Rate Testing Program. The Primary Containment Leakage Rate Tesling Program has been established in accordance with 10 CFR 50.54(o) to implement the requirements of 10 CFR Part 50, Appendix J, Option B (Ref. 9), and conforms with Regulatory Guide 1.163 (Ref. 10) and Nuclear Energy Institute (NEI) 94-01 (Ref. 11) except for the following:

a.

Local leak rate testing of the MSIVs may be performed at a pressure less than Pa. This is an exemption from the requirements of 10 CFR 50 Appendix J (Ref. 9). The basis for this exemption is described in Reference 12.

The Frequency is required by the Primary Containment Leakage Rate Testing Program.

I I

(continued)

Brunswick Unit 1 B 3.6.1.3-13 Revision No. 46 1

PCIVs B 3.6.1.3 BASES REFERENCES

1.

UFSAR, Chapter 15.

2.

NEDC-32466P, Power Uprate Safety Analysis Report for Brunswick Steam Electric Plant Units 1 and 2, September 1995.

3.

10 CFR 50.36(c)(2)(ii).

4.

Technical Requirements Manual.

5.

NEDO-32977-A, "Excess Flow Check Valve Testing Relaxation,"

June 2000.

6.

UFSAR, Section 15.2.3.

7.

NRC letter, Brunswick Steam Electric Plant, Units 1 and 2 -

Issuance of Amendment Re: Alternative Source Term, May 30, 2002.

8.

BNP Calculation No. BNP-RAD-007, Rev. 2, DBA-LOCA Radiological Dose With Alternate Source Term.

9.

10 CFR 50, Appendix J, Option B.

10.

NRC Regulatory Guide 1.163, Performance-Based Containment Leak-Rate Testing Program, September 1995.

11.

Nuclear Energy Institute (NEI) 94-01, Industry Guideline for Implementing Performance-Based Option of 10 CFR 50 Appendix J, July 26, 1995.

12.

NRC SER, Brunswick 1 & 2 -Amendments No. 10 and 36 to Operating Licenses Revising Technical Specifications to Grant Exemptions from Specific Requirements of 10 CFR 50 Appendix J, dated November 8, 1977.

Brunswick Unit 1 B 3.6.1.3-14 Revision No. 46 I

CREV System B 3.7.3 BASEES BACKGROUND The CREV System is designed to maintain the control room environrrent (continued) for a 30 day continuous occupancy after a DBA without exceeding 5 ram whole body dose or its equivalent to any part of the body. A single CREV subsystem will slightly pressurize the control room to prevent infiltration of air from surrounding buildings. CREV System operation in maintaining control room habitability is discussed in the UFSAR, Sections 6.4 and 9.4, (Refs. 1 and 2, respectively).

APPILICABLE The ability of the CREV System to maintain the habitability of the control SAFEETY ANALYSES room is an explicit assumption for the design basis accident presented in the UFSAR (Ref. 3). The radiation/smoke protection mode of the CREV System is assumed (explicitly or implicitly) to operate following a loss of coolant accident, fuel handling accident, main steam line break, and control rod drop accident. The radiological doses to control room personnel as a result of a DBA are summarized in Reference 3.

Postulated single active failures that may cause the loss of outside or recirculated air from the control room are bounded by BNP radiological dose calculations for control room personnel.

The CREV System satisfies Criterion 3 of 10 CFR 50.36(c)(2)(ii) (Ref. 4).

LCO Two redundant subsystems of the CREV System are required to be OPERABLE to ensure that at least one is available, assuming a single failure disables the other subsystem. Total system failure could resul: in exceeding a dose of 5 rem to the control room operators in the event of a DBA if unfiltered leakage into the control room is > 2,000 cfm.

The CREV System is considered OPERABLE when the individual components necessary to support the radiation protection mode are OPERABLE in both subsystems. A subsystem is considered OPERABLE when its associated:

a.

Emergency recirculation fan is OPERABLE;

b.

HEPA filter and charcoal adsorber bank are not excessively restricting flow and are capable of performing their filtration and adsorption functions; and (continued)

Brunswick Unit 1 B 3.7.3-2 Revision No. 45 l

BSEP 06-0044 Unit 2 Technical Specification Bases Replacement Pages

Unit 2 - Bases Book 1 Replacement Pages

BASES TO THE FACILITY OPERATING LICENSE DPR-62 TECHNICAL SPECIFICATIONS FOR BRUNSWICK STEAM ELECTRIC PLANT UNIT 2 CAROLINA POWER & LIGHT COMPANY REVISION 44

LIST OF EFFECTIVE PAGES - BASES Paqe No.

Title Page Revision No.

44 List cf Effective Pages - Book 1 LOEP-1 LOEP-2 LOEFP-3 LOEP-4 44 37 42 43 ii 40 30 B 2.1.1-1 B 2.1.1-2 B 2.1.1-3 B 2.1.1-4 B 2.1.1-5 B 2.1.2-1 B 2.1.2-2 B 2.1.2-3 B 3.0-1 B 3.0-2 B 3.0-3 B 3.0-4 B 3.0-5 B 3.0-6 B 3.0-7 B 3.0-8 B 3.0-9 B 3.C(-10 B 3.0-11 B 3.0-12 B 3.0-13 B 3.0-14 B 3.0-15 B 3.0-16 B 3.0-17 B 3.1.1-1 B 3.1.1-2 B 3.1.1-3 B 3.1.1-4 B 3.1.1-5 30 30 30 30 30 30 30 30 30 30 30 30 39 39 39 39 39 39 39 39 39 39 39 39 39 30 30 30 30 30 Page No.

B 3.1.1-6 B 3.1.2-1 B 3.1.2-2 B 3.1.2-3 B 3.1.2-4 B 3.1.2-5 B 3.1.3-1 B 3.1.3-2 B 3.1.3-3 B 3.1.3-4 B 3.1.3-5 B 3.1.3-6 B 3.1.3-7 B 3.1.3-8 B 3.1.3-9 B 3.1.4-1 B 3.1.4-2 B 3.1.4-3 B 3.1.4-4 B 3.1.4-5 B 3.1.4-6 B 3.1.4-7 B 3.1.5-1 B 3.1.5-2 B 3.1.5-3 B 3.1.5-4 B 3.1.5-5 B 3.1.6-1 B 3.1.6-2 B 3.1.6-3 B 3.1.64 B 3.1.6-5 B 3.1.7-1 B 3.1.7-2 B 3.1.7-3 B 3.1.7-4 B 3.1.7-5 B 3.1.7-6 B 3.1.8-1 B 3.1.8-2 B 3.1.8-3 B 3.1.8-4 B 3.1.8-5 Revision No.

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Brunswick Unit 2 LOEP-1 Revision 44 l

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B 3.3.3.1-4 B 3.3.3.1-5 B 3.3.3.1-6 B 3.3.3.1-7 B 3.3.3.1-8 B 3.3.3.1-9 B 3.3.3.1-10 B 3.3.3.2-1 B 3.3.3.2-2 B 3.3.3.2-3 B 3.3.3.2-4 B 3.3.3.2-5 B 3.3.4.1-1 B 3.3.4.1-2 B 3.3.4.1-3 B 3.3.4.1-4 B 3.3.4.1-5 B 3.3.4.1-6 B 3.3.4.1-7 B 3.3.4.1-8 B 3.3.4.1-9 B 3.3.5.1-1 B 3.3S.5.1-2 B 3.3.5.1-3 B 3.3.5.1-4 B 3.-.5.1-5 B 3.3.5.1-6 B 3.3.5.1-7 B 3.3.5.1-8 B 3.3.5.1-9 B 3.3.5.1-10 B 3.30.5.1-11 B 3.3.5.1-12 B 3.3.5.1-13 B 3.3.5.1-14 B 3.3.5.1-15 B 3.3.5.1-16 30 30 38 39 38 38 38 39 39 39 39 39 30 30 30 30 30 30 30 30 30 30 30 30 33 30 30 30 30 30 30 30 30 33 33 33 33 B 3.3.5.1-17 B 3.3.5.1-18 B 3.3.5.1-19 B 3.3.5.1-20 B 3.3.5.1-21 B 3.3.5.1-22 B 3.3.5.1-23 B 3.3.5.1-24 B 3.3.5.1-25 B 3.3.5.1-26 B 3.3.5.1-27 B 3.3.5.1-28 B 3.3.5.1-29 B 3.3.5.1-30 B 3.3.5.1-31 B 3.3.5.2-1 B 3.3.5.2-2 B 3.3.5.2-3 B 3.3.5.2-4 B 3.3.5.2-5 B 3.3.5.2-6 B 3.3.5.2-7 B 3.3.5.2-8 B 3.3.5.2-9 B 3.3.5.2-10 B 3.3.5.2-11 B 3.3.6.1-1 B 3.3.6.1-2 B 3.3.6.1-3 B 3.3.6.1-4 B 3.3.6.1-5 B 3.3.6.1-6 B 3.3.6.1-7 B 3.3.6.1-8 B 3.3.6.1-9 B 3.3.6.1-10 B 3.3.6.1-11 B 3.3.6.1-12 B 3.3.6.1-13 B 3.3.6.1-14 30 33 33 30 30 30 30 30 30 30 30 30 30 30 42 30 30 30 30 30 30 30 30 30 30 42 30 30 30 30 30 30 30 31 30 30 30 30 30 30 I

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Brunswick Unit 2 LOEP-3 Revision 42 l

LIST OF EFFECTIVE PAGES - BASES (continued)

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B 3.3.6.1-15 B 3.3.6.1-16 B 3.3.6.1-17 B 3.3.6.1-18 B 3.3.6.1-19 B 3.3.6.1-20 B 3.3.6.1-21 B 3.3.6.1-22 B 3.3.6.1-23 B 3.3.6.1-24 B 3.3.6.1-25 B 3.3.6.1-26 B 3.3.6.1-27 B 3.3.6.1-28 B 3.3.6.1-29 B 3.3.6.1-30 B 3.3.6.1-31 B 3.3.6.1-32 B 3.3.6.2-1 B 3.3.6.2-2 B 3.3.6.2-3 B 3.3.6.2-4 B 3.3.6.2-5 B 3.3.6.2-6 B 3.3.6.2-7 B 3.3.6.2-8 B 3.3.6.2-9 B 3.3.6.2-10 B 3.3.6.2-11 B 3.3.7.1-1 B 3.3.7.1-2 B 3.3.7.1-3 B 3.3.7.1-4 B 3.3.7.1-5 B 3.31.7.1-6 B 3.3.7.1-7 B 3.3.7.2-1 B 3.3.7.2-2 B 3.3.7.2-3 B 3.3.7.2-4 B 3.3.7.2-5 B 3.32.7.2-6 30 30 30 30 30 30 30 30 30 30 30 30 30 30 30 42 30 30 30 30 30 30 30 30 30 30 30 30 30 43 43 43 43 30 30 33 30 30 30 30 30 30 B 3.3.7.2-7 B 3.3.8.1-1 B 3.3.8.1-2 B 3.3.8.1-3 B 3.3.8.1-4 B 3.3.8.1-5 B 3.3.8.1-6 B 3.3.8.1-7 B 3.3.8.2-1 B 3.3.8.2-2 B 3.3.8.2-3 B 3.3.8.2-4 B 3.3.8.2-5 B 3.3.8.2-6 B 3.3.8.2-7 30 30 30 30 30 30 30 30 30 30 30 30 30 30 30 I

Brunswick Unit 2 LOEP-4 Revision 43 l

ECCS Instrumentation B 3.3.5.1 BASES SURVEILLANCE REQUIREMENTS (continued)

SR 3.3.5.1.5 The LOGIC SYSTEM FUNCTIONAL TEST demonstrates the OPERABILITY of the required initiation logic and simulated automatic operation for a specific channel. The system functional testing performed in LCO 3.5.1, LCO 3.5.2, LCO 3.8.1, and LCO 3.8.2 overlaps this Surveillance to complete testing of the assumed safety function.

Based on minimal assumed risk in performing this Surveillance with the reactor at power, the surveillance is not required to be performed during a refueling outage. Operating experience has demonstrated that these components will usually pass the SR when performed at the 24 month Frequency (originally based on the refueling cycle). Therefore, the Frequency is concluded to be acceptable from a reliability standpoint.

REFERENCES

1.

UFSAR, Section 5.2.

2.

3.

4.

5.

UFSAR, Section 6.3.

UFSAR, Chapter 15.

10 CFR 50.36(c)(2)(ii).

NEDC-31624P, Brunswick Steam Electric Plant Units 1 and 2 SAFER/GESTR-LOCA Loss-of-Coolant Accident Analysis (Revision 2), July 1990.

UFSAR, Section 9.2.6.2.

NEDC-30936-P-A, BWR Owners' Group Technical Specification Improvement Methodology (With Demonstration for BWR ECCS Actuation Instrumentation), Parts I and 2, December 1988.

6.

7.

Brunswick Unit 2 B 3.3.5.1-31 Revision No. 42 1

RCIC System Instrumentation B 3.3.5.2 BASES SURVEILLANCE REQUIREMENTS (continued)

SR 3.3.5.2.5 The LOGIC SYSTEM FUNCTIONAL TEST demonstrates the OPERABILITY of the required initiation logic for a specific channel and includes simulated automatic actuation of the channel. The system functional testing performed in LCO 3.5.3 overlaps this Surveillance to provide complete testing of the safety function.

Based on minimal assumed risk in performing this Surveillance with the reactor at power, the surveillance is not required to be performed during a refueling outage. Operating experience has demonstrated that these components will usually pass the SR when performed at the 24 month Frequency (originally based on the refueling cycle). Therefore, the Frequency is concluded to be acceptable from a reliability standpoint.

REFERENCES

1.

10 CFR 50.36(c)(2)(ii).

2.

GENE-770-06-2P-A, Bases for Changes to Surveillance Test Intervals and Allowed Out-of-Service Times for Selected Instrumentation Technical Specifications, December 1992.

Brunswick Unit 2 B 3.3.5.2-1 1 Revision No. 42 l

Primary Containment Isolation Instrumentation B 3.3.6.1 BASES SURVEILLANCE SR 3.3.6.1.4 and SR 3.3.6.1.6 (continued)

REQUIREMENTS CALIBRATION leaves the channel adjusted to account for instrument drifts between successive calibrations consistent with the plant specific setpoint methodology.

The Frequency of SR 3.3.6.1.4 is based on the assumption of a 92 day calibration interval in the determination of the magnitude of equipment drift in the setpoint analysis. The Frequency of SR 3.3.6.1.6 is based on t'le assumption of a 24 month calibration interval in the determination of the magnitude of equipment drift in the setpoint analysis.

SR 3.3.6.1.7 The LOGIC SYSTEM FUNCTIONAL TEST demonstrates the OPERABILITY of the required isolation logic for a specific channel and includes simulated automatic operation of the channel. The system functional testing performed on PCIVs in LCO 3.6.1.3 overlaps this Surveillance to provide complete testing of the assumed safety function.

The 24 month Frequency was developed considering it is prudent that this Surveillance be performed during a unit outage. Operating experience has demonstrated that these components will pass this Surveillance when performed at the 24 month Frequency. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint.

SR 3.3.6.1.8 This SR ensures that the individual channel response times are less than or equal to the maximum values assumed in the accident analysis.

Testing is performed only on channels where the assumed response lime does not correspond to the diesel generator (DG) start time. For channels assumed to respond within the DG start time, sufficient margin exists in the 10 second start time when compared to the typical channel response time (milliseconds) so as to assure adequate response without a specific measurement test (Ref. 9).

(continued)

Brunswick Unit 2 B 3.3.6.1-30 Revision No. 42 l

CREV System Instrumentation B 3.:3.7.1 B 3.3 INSTRUMENTATION B 3.3.7.1 Control Room Emergency Ventilation (CREV) System Instrumentation BASES BACKGROUND The CREV System is designed to provide a radiologically controlled environment to ensure the habitability of the control room for the safety of control room operators under all plant conditions. Two independent CREV subsystems are each capable of fulfilling the stated safety function.

The instrumentation and controls for the CREV System automatically initiate action to pressurize the main control room (MCR) to minimize the consequences of radioactive material in the control room environment:.

In the event of a loss of coolant accident (LOCA), the Unit 2 Secondary Containment Isolation-CREV Auto-Start signal will automatically start the CREV System in the radiation smoke/protection mode. The automatic CREV actuation is generated from the portion of the secondary containment isolation logic associated with Reactor Vessel Water Level-Low Level 2 and the Drywell Pressure-High. CREV automatic actuation will not occur as a result of secondary containment isolations due to Reactor Building Exhaust Radiation-High signals. The Control Building Air Intake-Radiation High function provides protection for non-LOCA events. In the event of a Control Building Air Intake Radiation-High signal, the CREV System is automatically started in the radiation/smo'ce protection mode. Air is then recirculated through the charcoal filter, and sufficient outside air is drawn in through the normal intake to maintain the MCR slightly pressurized with respect to outside atmosphere.

The CREV System instrumentation has two trip systems, either of which can initiate the CREV System. The Reactor Vessel Water Level-Low Level 2 and the Drywell Pressure-High signals are arranged in such a manner that opening of either A and B, or C and D, relay contacts of either the Reactor Vessel Water Level-Low Level 2 or the Drywell Pressure-High will provide the CREV start signal. Thus, the automatic CREV initiation, using signals from the secondary containment isolation logic:,

provides redundant/diverse protection for control room operators in the event of a LOCA. The Control Building Air Intake Radiation-High Function is arranged in a one-out-of-two logic for each trip system. The channels include electronic equipment (e.g., trip units) that compares measured input signals with pre-established setpoints. When the setpoint is exceeded, the channel output relay actuates, which then outputs a CREV System initiation signal to the initiation logic.

APPLICABLE The ability of the CREV System to maintain the habitability of the MCR is SAFETY ANALYSES, explicitly assumed for the design basis accident as discussed in the LCO, and UFSAR safety analyses (Refs. 1 and 2). CREV System operation APPLICABILITY ensures that the radiation exposure of control room personnel, through (continued)

Brunswick Unit 2 B 3.3.7.1 -1 Revision No. 43 l

CREV System Instrumentation B 3.3.7.1 BASES APPLICABLE the duration of any one of the postulated accidents, does not exceed the SAFETY ANALYSES, limits set by GDC 19 of 10 CFR 50, Appendix A.

LCO, and APPLICABILITY CREV System instrumentation satisfies Criterion 3 of (continued) 10 CFR 50.36(c)(2)(ii) (Ref. 3).

The OPERABILITY of the CREV System instrumentation is dependent upon the OPERABILITY of the individual instrumentation Channel Functions specified in Table 3.3.7.1-1. The Functions must have a required number of OPERABLE channels, with their setpoints within the specified Allowable Values provided in Tables 3.3.7.1-1 and 3.3.6.2-1.

The actual setpoints are calibrated consistent with applicable setpoinl methodology assumptions.

Allowable Values are specified for each Function specified in Tables 3.3.7.1-1 and 3.3.6.2-1. Trip setpoints are specified in the setpoint calculations. The setpoints are selected to ensure that the trip settings do not exceed the Allowable Value between successive CHANNEL CALIBRATIONS. Operation with a trip setting less conservative than the trip setpoint, but within its Allowable Value, is acceptable. A channel is inoperable if its actual trip setting is not within its required Allowable Value. Trip setpoints are those predetermined values of output at which an action should take place. The setpoints are compared to the actual process parameter (e.g., control building air intake radiation), and when the measured output value of the process parameter exceeds the setpoint, the associated device (e.g., trip unit) changes state. The analytic limits are derived from the limiting values of the process parameters obtained from the safety analysis. The trip setpoints are determined from the analytic limits, corrected for defined process, calibration, and instrument errors. The Allowable Values are then determined, based on the trip setpoint values, by accounting for calibration based errors. These calibration based instrument errors are limited to instrument drift, errors associated with measurement and test equipment, and calibration tolerance of loop components. The trip setpoints and Allowable Values determined in this manner provide adequate protection because instrumentation uncertainties, process effects, calibration tolerances, instrument drift, and severe environment errors (for channels that must function in harsh environments as defined by 10 CFR 50.49) are accounted for and appropriately applied for the instrumentation.

The specific Applicable Safety Analyses, LCO, and Applicability discussions are listed below on a Function by Function basis.

1. Control Building Air Intake Radiation-High The control building air intake radiation monitors measure radiation levels in the control building air intake plenum. A high radiation level may pose a threat to MCR personnel; thus, automatically initiating the CREV System.

(continued)

Brunswick Unit 2 B 3.3.7.1-2 Revision No. 43

CREV System Instrumenlation B 3.3.7.1 BASES APPLICABLE The Control Building Air Intake Radiation-High Function consists of two SAFETY ANALYSES, independent monitors. Two channels per trip system of Control Building LCO, and Air Intake Radiation-High are available and are required to be APPLICABILITY OPERABLE to ensure that no single instrument failure can preclude (continued)

CREV System initiation. The Allowable Value was selected to ensure!

protection of the control room personnel.

The Control Building Air Intake Radiation-High Function is required lo be OPERABLE in MODES 1, 2, and 3 and during OPDRVs and movement of recently irradiated fuel assemblies in the secondary containment, to ensure that control room personnel are protected during a LOCA, fuel handling event, or vessel draindown event. During MODES 4 and 5, when these specified conditions are not in progress (e.g., OPDRVs), the probability of a LOCA, main steam line break accident, or control rod drop accident is low; thus, the Function is not required. Also due to radioactive decay, this Function is only required to initiate the CREV System during fuel handling accidents involving handling recently irradiated fuel (i.e., fuel that has occupied part of a critical reactor core within the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ).

2. Unit 2 Secondary Containment Isolation-CREV Auto-Start The Unit 2 Secondary Containment Isolation-CREV Auto-Start Function provides post-LOCA operator protection. Since Reactor Vessel Water Level-Low Level 2 and Drywell Pressure-High provide primary indication of a LOCA, only secondary containment isolations resulting from these signals provide CREV Auto-Start. The Reactor Vessel Water Level-Low Level 2 and the Drywell Pressure-High signals are arranged in such a manner that opening of either A and B, or C and D, relay contacts of either the Reactor Vessel Water Level-Low Level 2 or the Drywell Pressure-High will provide the CREV start signal. Thus, automatic CREV initiation, using signals from the secondary containment isolation logic, provides redundant/diverse protection for control room operators in the event of a LOCA.

The Allowable Values for the Secondary Containment Isolation Instrumentation are provided in Table 3.3.6.2-1.

The Unit 2 Secondary Containment Isolation-CREV Auto-Start Function is required to be OPERABLE in MODES 1,2, and 3 to ensure that control room personnel are protected in the event of a LOCA.

ACTIONS A Note has been provided to modify the ACTIONS related to CREV System instrumentation channels. Section 1.3, Completion Times, specifies that once a Condition has been entered, subsequent divisions, subsystems, components, or variables expressed in the Condition, discovered to be inoperable or not within limits, will not result in separate entry into the Condition. Section 1.3 also specifies that Required Actions (continued)

Brunswick Unit 2 B 3.3.7.1-3 Revision Nho. 43 l

CREV System Instrumentation B 3.:3.7.1 BASES ACTIONS of the Condition continue to apply for each additional failure, with (continued)

Completion Times based on initial entry into the Condition. However, the Required Actions for inoperable CREV System instrumentation channels provide appropriate compensatory measures for separate inoperable channels. As such, a Note has been provided that allows separate Condition entry for each inoperable CREV System instrumentation channel.

A.1 Because of the redundancy of sensors available to provide initiation signals and the redundancy of the CREV System design, an allowable out of service time of 7 days is provided to permit restoration of any inoperable channel to OPERABLE status. This out of service time is only acceptable provided the affected Function is still maintaining CREV System initiation capability (refer to Required Action B.1 Bases). If the Function is not maintaining CREV System initiation capability, Condition B must be entered.

If the inoperable channel cannot be restored to OPERABLE status within the 7 day allowable out of service time, one CREV subsystem must be placed in the radiation/smoke protection mode of operation per Required Action A.1..The method used to place the CREV subsystem in operation must provide for automatically re-initiating the subsystem upon restoration of power following a loss of power to the CREV subsystem. Placing one CREV subsystem in the radiation/smoke protection mode of operation provides a suitable compensatory action to ensure that the automatic radiation protection function of the CREV System is not lost.

B.1 Required Action B.1 is intended to ensure that appropriate action is taken if multiple, inoperable, untripped channels result in the affected Function not maintaining CREV System initiation capability. The Function is considered to be maintaining CREV System initiation capability when sufficient channels are OPERABLE or in trip such that one trip system will generate an initiation signal for one CREV subsystem from the Function on a valid signal. For the Control Building Air Intake Radiation-High Function, this would require one trip system to have one channel OPERABLE or in trip. For the Unit 2 Secondary Containment Isolation-CREV Auto Start Reactor Vessel Water Level-Low Level 2 or Drywell Pressure-High, this would require all channels to be operable or the inoperable channels in the tripped condition. With CREV System init ation capability not maintained, one CREV subsystem must be placed in the radiation/smoke protection mode of operation per Required Action B.1 to ensure that control room personnel will be protected in the event of a Design Basis Accident. The method used to place the CREV subsystem in operation must provide for automatically re-initiating the subsystem upon restoration of power following a loss of power to the CREV subsystem.

(contiiued)

Brunswick Unit 2 B 3.3.7.1-4 Revision No. 43

Unit 2 - Bases Book 2 Replacement Pages

LIST OF EFFECTIVE PAGES - BASES Paqe No.

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Paqe No.

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Title Page N/A B 3.4.7-4 39 B 3.4.7-5 39 List of Effective Pages - Book 2 B 3.4.8-1 30 B 3.4.8-2 30 LOEF'-1 44 B 3.4.8-3 30 LOEF'-2 44 B 3.4.8-4 30 LOEF'-3 43 B 3.4.8-5 30 LOEF-4 30 B 3.4.9-1 30 LOEF'-5 30 B 3.4.9-2 35 B 3.4.9-3 35 i

30 B 3.4.9-4 30 ii 30 B 3.4.9-5 35 B 3.4.9-6 35 B 3.4.1-1 30 B 3.4.9-7 30 B 3.4.1-2 30 B 3.4.9-8 30 B 3.4.1-3 30 B 3.4.9-9 35 B 3.4.1-4 30 B 3.4.10-1 30 B 3.4.1-5 30 B 3.4.10-2 30 B 3.4.1-6 30 B 3.4.2-1 30 B 3.5.1-1 30 B 3.4.2-2 30 B 3.5.1-2 30 B 3.4.2-3 30 B 3.5.1-3 30 B 3.4.2-4 30 B 3.5.1-4 33 B 3.4.3-1 30 B 3.5.1-5 33 B 3.4.3-2 30 B 3.5.1-6 39 B 3.4.3-3 30 B 3.5.1-7 30 B 3.4.3-4 30 B 3.5.1-8 30 B 3.4.4-1 30 B 3.5.1-9 30 B 3.4.4-2 30 B 3.5.1-10 30 B 3.4.4-3 30 B 3.5.1-11 30 B 3.4.4-4 30 B 3.5.1-12 30 B 3.4.4-5 30 B 3.5.1-13 30 B 3.4.5-1 30 B 3.5.1-14 30 B 3.4.5-2 30 B 3.5.1-15 42 B 3.4.5-3 41 B 3.5.1-16 42 B 3.4.5-4 39 B 3.5.1-17 30 B 3.4.6-1 39 B 3.5.2-1 30 B 3.4.6-2 39 B 3.5.2-2 30 B 3.4.6-3 39 B 3.5.2-3 30 B 3.4.7-1 30 B 3.5.2-4 30 B 3.4.7-2 30 B 3.5.2-5 30 B 3.4.7-3 39 B 3.5.2-6 30 B 3.5.3-1 30 B 3.5.3-2 39 (continued)

Brunswick Unit 2 LOEP-1 Revision 44 l

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B 3.5.3-3 B 3.5.3-4 B 3.5.3-5 B 3.5.3-6 B 3.5.3-7 B 3.6.1.1-1 B 3.6.1.1-2 B 3.6.1.1-3 B 3.6.1.1-4 B 3.6.1.1-5 B 3.6.1.1-6 B 3.6.1.2-1 B 3.6.1.2-2 B 3.6.1.2-3 B 3.6.1.2-4 B 3.6.1.2-5 B 3.6.1.2-6 B 3.6.1.2-7 B 3.6.1.2-8 B 3.6.1.3-1 B 3.6.1.3-2 B 3.6.1.3-3 B 3.6.1.3-4 B 3.6.1.3-5 B 3.6.1.3-6 B 3.6.1.3-7 B 3.6.1.3-8 B 3.6.1.3-9 B 3.6.1.3-10 B 3.6.1.3-11 B 3.6.1.3-12 B 3.6.1.3-13 B 3.6.1.3-14 B 3.6.1.4-1 B 3.6.1.4-2 B 3.6.1.4-3 B 3.6.1.5-1 B 3.6.1.5-2 B 3.6.1.5-3 B 3.6.1.5-4 B 3.6.1.5-5 B 3.6.1.5-6 30 30 30 30 30 30 30 30 44 44 44 30 30 30 30 30 30 44 44 30 30 30 30 30 30 30 30 30 30 30 44 44 44 30 30 30 30 30 30 30 30 30 B 3.6.1.5-7 B 3.6.1.5-8 B 3.6.1.5-9 B 3.6.1.6-1 B 3.6.1.6-2 B 3.6.1.6-3 B 3.6.1.6-4 B 3.6.1.6-5 B 3.6.1.6-6 B 3.6.2.1-1 B 3.6.2.1-2 B 3.6.2.1-3 B 3.6.2.1-4 B 3.6.2.1-5 B 3.6.2.2-1 B 3.6.2.2-2 B 3.6.2.2-3 B 3.6.2.3-1 B 3.6.2.3-2 B 3.6.2.3-3 B 3.6.2.3-4 B 3.6.3.1-1 B 3.6.3.1-2 B 3.6.3.1-3 B 3.6.3.2-1 B 3.6.3.2-2 B 3.6.3.2-3 B 3.6.3.2-4 B 3.6.3.2-5 B 3.6.4.1-1 B 3.6.4.1-2 B 3.6.4.1-3 B 3.6.4.1-4 B 3.6.4.1-5 B 3.6.4.2-1 B 3.6.4.2-2 B 3.6.4.2-3 B 3.6.4.2-4 B 3.6.4.2-5 B 3.6.4.2-6 B 3.6.4.3-1 B 3.6.4.3-2 B 3.6.4.3-3 30 33 30 30 30 30 30 30 30 30 30 30 30 30 30 30 30 30 39 39 39 30 30 30 30 30 39 39 39 30 30 30 30 30 30 30 30 30 30 30 30 30 30 I

(continued)

Brunswick Unit 2 LOEP-2 RevisiDn 44

LIST OF EFFECTIVE PAGES - BASES (continued)

Page No.

B 3.6.4.3-4 B 3.6.4.3-5 B 3.6.4.3-6 B 3.7.1-1 B 3.7.1-2 B 3.7.1-3 B 3.7.1-4 B 3.7.1-5 B 3.7.2-1 B 3.7.2-2 B 3.7.2-3 B 3.7.2-4 B 3.7.2-5 B 3.7.2-6 B 3.7.2-7 B 3.7.2-8 B 3.7.2-9 B 3.7.2-10 B 3.7.2-11 B 3.7.2-12 B 3.7.2-13 B 3.7.2-14 B 3.7.3-1 B 3.7.3-2 B 3.7.3-3 B 3.7.3-4 B 3.7.3-5 B 3.7.3-6 B 3.7.3-7 B 3.7.4-1 B 3.7.4-2 B 3.7.4-3 B 3.7.4-4 B 3.7.4-5 B 3.7.5-1 B 3.7.5-2 B 3.7.5-3 B 3.7.6-1 B 3.7.6-2 B 3.7.6-3 B 3.7.6-4 Revision No.

Page No.

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30 30 30 30 30 33 39 39 30 30 30 30 30 30 30 30 30 30 30 30 30 30 30 30 30 30 30 30 30 30 35 30 30 30 30 33 30 30 30 30 30 30 I

(continued)

Brunswick Unit 2 LOEP-3 Revision 43 l

ECCS-Operating B 3.5.1 BASES SURVEILLANCE SR 3.5.1.6, SR 3.5.1.7. and SR 3.5.1.8 (continued)

REQUIREMENTS Therefore, SR 3.5.1.7 and SR 3.5.1.8 are modified by Notes that state the Surveillances are not required to be performed until 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months after the reactor steam pressure is adequate to perform the test.

The 92 day Frequency for SR 3.5.1.6 and SR 3.5.1.7 is consistent with the Inservice Testing Program requirements. The 24 month Frequency for SR 3.5.1.8 is based on the need to perform the Surveillance under the conditions that apply just prior to or during a startup from a plant outage.

Operating experience has demonstrated that these components will usually pass the SR when performed at the 24 month Frequency, which is based on the refueling cycle. Therefore, the Frequency is considered to be acceptable from a reliability standpoint.

SR 3.5.1.9 The ECCS subsystems are required to actuate automatically to perform their design functions. This Surveillance verifies that, with a required system initiation signal (actual or simulated), the automatic initiation logic of HPCI, CS, and LPCI will cause the systems or subsystems to operate as designed, including actuation of the system throughout its emergency operating sequence, automatic pump startup and actuation of all automatic valves to their required positions. This SR also ensures that the HPCI System will automatically restart on an RPV low water level signal received subsequent to an RPV high water level trip and that the suction is automatically transferred from the CST to the suppression pool on a CST low level signal or a suppression pool high water level signal.

The LOGIC SYSTEM FUNCTIONAL TEST performed in LCO 3.3.5.1, "ECCS Instrumentation," overlaps this Surveillance to provide complete testing of the assumed safety function.

Based on minimal assumed risk in performing this Surveillance with the reactor at power, the surveillance is not required to be performed during a refueling outage. Operating experience has demonstrated that these components will usually pass the SR when performed at the 24 month Frequency (originally based on the refueling cycle). Therefore, the Frequency is concluded to be acceptable from a reliability standpoint.

(continued)

Brunswick Unit 2 B 3.5.1-14 Revision No. 42

ECCS-Operating B 3.5.1 BASES SURVEILLANCE SR 3.5.1.9 (continued)

REQUIREMENTS This SR is modified by a Note that excludes vessel injection/spray during the Surveillance. Since all active components are testable and full flow can be demonstrated by recirculation through the test line, coolant injection into the RPV is not required during the Surveillance.

SR 3.5.1.10 The ADS designated SRVs are required to actuate automatically upon receipt of specific initiation signals. A system functional test is performed to demonstrate that the mechanical portions of the ADS function (i.e.,

solenoids) operate as designed when initiated either by an actual or simulated initiation signal, causing proper actuation of all the required components. SR 3.5.1.11 and the LOGIC SYSTEM FUNCTIONAL TEST performed in LCO 3.3.5.1 overlap this Surveillance to provide complete testing of the assumed safety function.

The 24 month Frequency is based on the need to perform the Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power. Operating experience has demonstrated that these components will usually pass the SR when performed at the 24 month Frequency, which is based on the refueling cycle. Therefore, the Frequency is considered to be acceptable from a reliability standpoint.

This SR is modified by a Note that excludes valve actuation since the valves are individually tested in accordance with SR 3.5.1.1 1. This also prevents an RPV pressure blowdown.

(continued)

Brunswick Unit 2 B 3.5.1-15 Revision No. 42 l

Primary Containment B 3.6.1.1 BASES (continued)

SURVEILLANCE SR 3.6.1.1.1 REQUIREMENTS Maintaining the primary containment OPERABLE requires compliance with the visual examinations and leakage rate test requirements of the Primary Containment Leakage Rate Testing Program. The Primary Containment Leakage Rate Testing Program has been established in accordance with 10 CFR 50.54(o) to implement the requirements of 10 CFR Part 50, Appendix J, Option B (Ref. 3). The Primary Containment Leakage Rate Testing Program also conforms with Regulatory Guide 1.163 (Ref. 6) and Nuclear Energy Institute (NEI) 94-01 (Ref. 7) except for the following:

a.

BNP may use the criteria and Total Time method specified in Bechtel Topical Report BN-TOP-1 (Ref. 8) for calculating the primary containment leakage during reduced duration Type A testing. This is an exemption from the requirements of 10 CFR 50 Appendix J (Ref. 3) which, in accordance with NEI 94-01 (Ref. 7),

requires the methods for calculating primary containment leakage described in ANSI/ANS 56.8-1994 (Ref. 9). The basis for this exemption is described in References 10 and 11.

b.

Type C testing is not required for the hydrogen and oxygen monitor isolation valves. This is an exemption from the requirements of 10 CFR 50 Appendix J (Ref. 3). The basis fo -this exemption is described in Reference 12.

Failure to meet air lock leakage limits (SR 3.6.1.2.1) or main steam isolation valve leakage (SR 3.6.1.3.9) does not necessarily result in a failure of this SR. The impact of the failure to meet SR 3.6.1.2.1 must be evaluated against the Type A, B, and C acceptance criteria of the Primary Containment Leakage Rate Testing Program, and failure to meet SR 3.6.1.3.9 must be evaluated against Type A acceptance criteria of the Primary Containment Leakage Rate Testing Program.

As left leakage prior to the first startup after performing required leakage testing is required to be < 0.6 La for combined Type B and C leakage, and

< 0.75 La for overall Type A leakage. At all other times between required leakage rate tests, the acceptance criteria is based on an overall Type A leakage limit of < 1.0 L. At < 1.0 La the offsite dose consequences are bounded by the assumptions of the safety analysis. The Frequency is required by the Primary Containment Leakage Rate Testing Program.

(conti iued)

Brunswick Unit 2 B 3.6.1.1-4 Revision No. 44 l

Primary Containment B 3.6.1.1 BASES SURVEILLANCE REQUIREMENTS (continued)

SR 3.6.1.1.2 Maintaining the pressure suppression function of primary containment requires limiting the leakage from the drywell to the suppression chamber.

Thus, if an event were to occur that pressurized the drywell, the steam would be directed through the downcomers into the suppression pool.

This SR measures drywell to suppression chamber differential pressure during a 10 minute period to ensure that the leakage paths that woulc bypass the suppression pool (downcomers) are within allowable limits.

Satisfactory performance of this SR can be achieved by establishing a known differential pressure between the drywell and the suppression chamber and verifying that the differential pressure between the suppression chamber and the drywell does not decrease by more than 0.25 inch of water per minute over a 10 minute period. The leakage test is performed every 24 months. The 24 month Frequency was developed considering it is prudent that this Surveillance be performed during a unit outage and also in view of the fact that component failures that might have affected this test are identified by other primary containment SRs.

REFERENCES

1.

UFSAR, Section 6.2.

2.

UFSAR, Section 15.6.

3.

10 CFR 50, Appendix J, Option B.

4.

NEDC-33039P, Safety Analysis Report for Brunswick Steam Electric Plant Units 1 and 2, Extended Power Uprate, August 2001.

5.

10 CFR 50.36(c)(2)(ii).

6.

NRC Regulatory Guide 1.163, Performance-Based Containment Leak-Rate Testing Program, September 1995.

7.

Nuclear Energy Institute (NEI) 94-01, Industry Guideline for Implementing Performance-Based Option of 10 CFR 50 Appendix J, July 26, 1995.

8.

Bechtel Topical Report BN-TOP-1, Revision 1, November 1,1972.

9.

ANSI/ANS 56.8-1994.

(continued)

I Brunswick Unit 2 B 3.6.1.1-5 Revision No.44 l

Primary Containment B 3.6.1.1 BASES REFERENCES (continued)

10.

NRC SER; Issuance of Amendment No. 181 to Facility Operaling License No. DPR-71 and Amendment No. 213 to Facility Operating License No. DPR-62 Regarding 10 CFR 50 Appendix J, Option B - Brunswick Steam Electric Plant, Units 1 and 2 (BSEP 95-0316) (TAC Nos. M93679 and M93680); dated February 1, 1996.

11.

NRC SER, Exemption from the Requirements of Appendix J for Brunswick Steam Electric Plant, Units I and 2, dated Februart 17, 1988.

12.

NRC SER, Technical Exemption from the Requirements of Appendix J, dated May 12,1987.

I Brunswick Unit 2 B 3.6.1.1-6 Revision No. 44 l

Primary Containment Air Lock B 3.6.1.2 BASE:S (continued)

SUR\\VEILLANCE SR 3.6.1.2.1 REQUIREMENTS Maintaining the primary containment air lock OPERABLE requires compliance with the leakage rate test requirements of the Primary Containment Leakage Rate Testing Program. The Primary Containment Leakage Rate Testing Program has been established in accordance with 10 CFR 50.54(o) to implement the requirements of 10 CFR Part 50, Appendix J, Option B (Ref. 4), and conforms with Regulatory Guide 1.163 (Ref. 5) and Nuclear Energy Institute (NEI) 94-01 (Ref. 6) except for the following:

a.

The local leak rate testing requirements of the primary I

containment air lock doors may be modified to perform the tests at a pressure less than Pa following replacement of the air lock door seals. This is an exception from the requirements of NEI 94-01 (Ref. 6). The basis for this exception is described in Reference 7.

This SR reflects the leakage rate testing requirements with respect to air lock leakage (Type B leakage tests). The acceptance criteria were established as a small fraction of the total allowable primary containment leakage. The periodic testing requirements verify that the air lock leakage does not exceed the allowed fraction of the overall primary containment leakage rate. The Frequency is required by the Primary Containment Leakage Rate Testing Program.

The SR has been modified by two Notes. Note 1 states that an inoperable air lock door does not invalidate the previous successful performance of the overall air lock leakage test. This is considered reasonable since either air lock door is capable of providing a fission product barrier in the event of a DBA. Note 2 has been added to this SR, requiring results to be evaluated against the acceptance criteria which are applicable to SR 3.6.1.1.1. This ensures that air lock leakage is properly accounted for in determining the combined Type B and C primary containment leakage rate.

(continued)

Brunswick Unit 2 B 3.6.1.2-7 Revision t\\o. 44 l

Primary Containment Air Lock B 3.6.1.2 BASES SURVEILLANCE REQIJIREMENTS (continued)

SR 3.6.1.2.2 The air lock interlock mechanism is designed to prevent simultaneous opening of both doors in the air lock. Since both the inner and outer doors of the air lock are designed to withstand the maximum expected post accident primary containment pressure, closure of either door will support primary containment OPERABILITY. Thus, the interlock feature supports primary containment OPERABILITY while the air lock is being used fcr personnel transit in and out of the containment. Periodic testing of this interlock demonstrates that the interlock will function as designed and that simultaneous inner and outer door opening will not inadvertently occur.

Due to the purely mechanical nature of this interlock, and given that the interlock mechanism is not normally challenged when the primary containment airlock door is used for entry and exit (procedures require strict adherence to single door opening), this test is only required to be performed every 24 months. The 24 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage, and the potential for loss of primary containment OPERABILITY if the Surveillance were performed with the reactor at power. The 24 month Frequency for the interlock is justified based on generic operating experience. The Frequency is based on engineering judgment and is considered adequate given that the interlock is not challenged during use of the air lock.

REFERENCES

1.

UFSAR, Section 3.8.2.4.3.2.

2.

NEDC-33039P, Safety Analysis Report for Brunswick Units 1 and 2 Extended Power Uprate, August 2001.

3.

10 CFR 50.36(c)(2)(ii).

4.

10 CFR 50, Appendix J, Option B.

5.

NRC Regulatory Guide 1.163, Performance-Based Containment Leak-Rate Testing Program, September 1995.

6.

Nuclear Energy Institute (NEI) 94-01, Industry Guideline for Implementing Performance-Based Option of 10 CFR 50 Appendix J, July 26, 1995.

7.

NRC SER, Brunswick 1 & 2 - Amendments No. 10 and 36 to Operating Licenses Revising Technical Specifications to Grant Exemptions from Specific Requirements of 10 CFR 50 Appendix J, dated November 8, 1977.

I Brunswick Unit 2 B 3.6.1.2-8 Revision No. 44 l

PCIVs B 3.5.1.3 BASES SURVEILLANCE SR 3.6.1.3.6 (continued)

REQUJIREMENTS complete testing of the safety function. The 24 month Frequency was developed considering it is prudent that this Surveillance be performei only during a unit outage since isolation of penetrations would eliminate cooling water flow and disrupt the normal operation of many critical components. Operating experience has demonstrated that these components will pass this Surveillance when performed at the 24 month Frequency. Therefore, the Frequency was concluded to be acceptab e from a reliability standpoint.

SR 3.6.1.3.7 This SR requires a demonstration that a representative sample of reactor instrumentation line excess flow check valves (EFCVs) is OPERABLE: by verifying that the valves actuate to the isolation position on an actual or simulated instrument line break signal. This may be accomplished by cycling the EFCVs through one complete cycle of full travel. The representative sample consists of an approximately equal number of EFCVs, such that each EFCV is tested at least once every 10 years (nominal). In addition, the EFCVs in the samples are representative of the various plant configurations, models, sizes, and operating environments.

This ensures that any potentially common problem with a specific type or application of EFCV is detected at the earliest possible time. This SR provides assurance that the instrumentation line EFCVs will perform so that predicted radiological consequences will not be exceeded during a postulated instrument line break event. The 24 month Frequency is based on the need to perform this Surveillance under the conditions tVat apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power. Operating experience has demonstrated that these components will pass this Surveillance when performed at the 24 month Frequency. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint.

The nominal 10-year interval is based on performance testing as discussed in NEDO-32977-A (Ref. 5). Furthermore, any EFCV failures will be evaluated to determine if additional testing in that test interval is warranted to ensure overall reliability is maintained. Operating experience has demonstrated that these components are highly reliable and that failures to isolate are very infrequent. Therefore, testing of a representative sample was concluded to be acceptable from a reliability standpoint.

(continued)

Brunswick Unit 2 B 3.6.1.3-12 Revision No. 44 l

PCIVs B 3.6.1.3 BASES SURVEILLANCE REQUIREMENTS (continued)

SR 3.6.1.3.8 The TIP shear isolation valves are actuated by explosive charges. An in place functional test is not possible with this design. The explosive squib is removed and tested to provide assurance that the valves will actua4:e when required. The replacement charge for the explosive squib shall be from the same manufactured batch as the one fired or from another batch that has been certified by having one of the batch successfully fired. The Frequency of this SR is in accordance with the requirements of the Inservice Testing Program.

SR 3.6.1.3.9 The analyses in References 2, 6, 7, and 8 are based on leakage that is less than the specified leakage rate. Leakage through each main steam line must be s 100 scfh when tested at 2 Pt (25 psig). The combined leakage rate for all four mains steam lines must be

150 scfh when tested at 2 25 psig in accordance with the Primary Containment Leakage Rate Testing Program. The Primary Containment Leakage Rate Testing Program has been established in accordance with 10 CFR 50.54(o) to implement the requirements of 10 CFR Part 50, Appendix J, Option e (Ref. 9), and conforms with Regulatory Guide 1.163 (Ref. 10) and Nuclear Energy Institute (NEI) 94-01 (Ref. 11) except for the following:

a.

Local leak rate testing of the MSIVs may be performed at a pressure less than Pa. This is an exemption from the requirements of 10 CFR 50 Appendix J (Ref. 7). The basis for this exemption is described in Reference 10.

The Frequency is required by the Primary Containment Leakage Rate Testing Program.

II (continued)

Brunswick Unit 2 B 3.6.1.3-13 Revision No. 44 l

PCIVs B 3.6.1.3 BASES REFERENCES

1.

UFSAR, Chapter 15.

2.

NEDC-32466P, Power Uprate Safety Analysis Report for Brunswick Steam Electric Plant Units 1 and 2, September 1995.

3.

10 CFR 50.36(c)(2)(ii).

4.

Technical Requirements Manual.

5.

NEDO-32977-A, "Excess Flow Check Valve Testing Relaxaticn,"

June 2000.

6.

UFSAR, Section 15.2.3.

7.

NRC letter, Brunswick Steam Electric Plant, Units I and 2 -

Issuance of Amendment Re: Alternative Source Term, May 30, 2002.

8.

BNP Calculation No. BNP-RAD-007, Rev. 2, DBA-LOCA Radiological Dose With Alternate Source Term.

9.

10 CFR 50, Appendix J, Option B.

10.

NRC Regulatory Guide 1.163, Performance-Based Containment Leak-Rate Testing Program, September 1995.

11.

Nuclear Energy Institute (NEI) 94-01, Industry Guideline for Implementing Performance-Based Option of 10 CFR 50 Appendix J, July 26, 1995.

12.

NRC SER, Brunswick 1 & 2 -Amendments No. 10 and 36 to Operating Licenses Revising Technical Specifications to Grant Exemptions from Specific Requirements of 10 CFR 50 Appendix J, dated November 8, 1977.

Brunswick Unit 2 B 3.6.1.3-14 Revision No. 44 l

CREV System B 3.7.3 BASES BACKGROUND The CREV System is designed to maintain the control room environment (continued) for a 30 day continuous occupancy after a DBA without exceeding 5 rem whole body dose or its equivalent to any part of the body. A single CREV subsystem will slightly pressurize the control room to prevent infiltration of air from surrounding buildings. CREV System operation in maintaining control room habitability is discussed in the UFSAR, Sections 6.4 and 9.4, (Refs. 1 and 2, respectively).

APPLICABLE The ability of the CREV System to maintain the habitability of the control SAFETY ANALYSES room is an explicit assumption for the design basis accident presented in the UFSAR (Ref. 3). The radiation/smoke protection mode of the CREV System is assumed (explicitly or implicitly) to operate following a loss of coolant accident, fuel handling accident, main steam line break, and control rod drop accident. The radiological doses to control room personnel as a result of a DBA are summarized in Reference 3.

Postulated single active failures that may cause the loss of outside or recirculated air from the control room are bounded by BNP radiological dose calculations for control room personnel.

The CREV System satisfies Criterion 3 of 10 CFR 50.36(c)(2)(ii) (Ref. 4).

LCO Two redundant subsystems of the CREV System are required to be OPERABLE to ensure that at least one is available, assuming a single failure disables the other subsystem. Total system failure could result in exceeding a dose of 5 rem to the control room operators in the event of a DBA if unfiltered leakage into the control room is > 2,000 cfm.

The CREV System is considered OPERABLE when the individual components necessary to support the radiation protection mode are OPERABLE in both subsystems. A subsystem is considered OPERABLE when its associated:

a.

Emergency recirculation fan is OPERABLE;

b.

HEPA filter and charcoal adsorber bank are not excessively restricting flow and are capable of performing their filtration and adsorption functions; and (continued)

Brunswick Unit 2 B 3.7.3-2 Revision No. 43 l

BSEP 06-0044 Unit 1 and 2 Technical Specification Table of Contents Replacement Pages

TABLE OF CONTENTS 1.0 USE AND APPLICATION................................................

1.1-1 1.1 Definitions................................................

1.1*1 1.2 Logical Connectors................................................

1.2-1 1.3 Completion Times.....................................................................................

1.3-1 1.4 Frequency................................................

1.4-1 2.0 SAFETY LIMITS (SLs)................................................

2.0-1 2.1 SLs................................................

2.0-1 2.2 SL Violations................................................

2.0-1 3.0 LIMITING CONDITION FOR OPERATION (LCO) APPLICABILITY.........

.. 3.0-1 3.0 SURVEILLANCE REQUIREMENT (SR) APPLICABILITY..............

............ 3.0 4 3.1 REACTIVITY CONTROL SYSTEMS................................................

3.1-1 3.1.1 SHUTDOWN MARGIN (SDM)................................................

3.1-1 3.1.2 Reactivity Anomalies.................................................

3.1-5 3.1.3 Control Rod OPERABILITY................................................

3.1-7 3.1.4 Control Rod Scram Times.................................................

3.1-12 3.1.5 Control Rod Scram Accumulators........................................

........ 3.1-15 3.1.6 Rod Pattern Control................................................

3.1-18 3.1.7 Standby Liquid Control (SLC) System................................................

3.1-20 3.1.8 Scram Discharge Volume (SDV) Vent and Drain Valves.......................

3.1-25 3.2 POWER DISTRIBUTION LIMITS.......................

3.2-1 3.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR).....................................

3.2-1 3.2.2 MINIMUM CRITICAL POWER RATIO (MCPR)..................................... 3.2-2 3.3 INSTRUMENTATION...........................................

3.3-1 3.3.1.1 Reactor Protection System (RPS) Instrumentation...............................

3.3-1 3.3.1.2 Source Range Monitor (SRM) Instrumentation..................................... 3.3.12 3.3.2.1 Control Rod Block Instrumentation.......................................

3.3-18 3.3.2.2 Feedwater and Main Turbine High Water Level Trip Instrumentation................................................................................... 3.3-24 3.3.3.1 Post Accident Monitoring (PAM) Instrumentation.................................. 3.3-26 3.3.3.2 Remote Shutdown Monitoring Instrumentation................................... 3.3-30 3.3.4.1 Anticipated Transient Without Scram Recirculation Pump Trip (ATWS-RPT) Instrumentation..

3.3-32 3.3.5.1 Emergency Core Cooling System (ECCS)

Instrumentation................................................................................... 3.3-35 3.3.5.2 Reactor Core Isolation Cooling (RCIC) System Instrumentation................................................................................... 3.3-45 3.3.6.1 Primary Containment Isolation Instrumentation.................................

3.3-49 (cortinued)

Brunswick Unit 1 i

Revisiorn 3/8/02 l

TAB3LE OF CONTENTS 3.3 INSTRUMENTATION (continued) 3.3.6.2 Secondary Containment Isolation Instrumentation

............................... 3.3-59 3.3.7.1 Control Room Emergency Ventilation (CREV) System Instrumentation................................................................................. 3.3-63 3.3.7.2 Condenser Vacuum Pump Isolation Instrumentation............................ 3.3-66 3.3.8.1 Loss of Power (LOP) Instrumentation.....................................

3.3-69 3.3.8.2 Reactor Protection System (RPS) Electric Power Monitoring......................................................................................... 3.3-72 3.4 REACTOR COOLANT SYSTEM (RCS)..............................

3.4-1 3.4.1 Recirculation Loops Operating..............................

3.4-1 3.4.2 Jet Pumps..............................

3.4-3 3.4.3 Safety/Relief Valves (SRVs)..............................

3.4-5 3.4.4 RCS Operational LEAKAGE..............................

3.4-7 3.4.5 RCS Leakage Detection Instrumentation..............................

3.4-9 3.4.6 RCS Specific Activity..............................

3.4-12 3.4.7 Residual Heat Removal (RHR) Shutdown Cooling System-Hot Shutdown..

3.4-14 3.4.8 Residual Heat Removal (RHR) Shutdown Cooling System-Cold Shutdown................................

3.4-17 3.4.9 RCS Pressure and Temperature (P/T) Limits................................

3.4-19 3.4.10 Reactor Steam Dome Pressure................................

3.4-28 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) AND REACTOR CORE ISOLATION COOLING (RCIC) SYSTEM...................................

3.5-1 3.5.1 ECCS-Operating...................................

3.5-1 3.5.2 ECCS-Shutdown...................................

3.5-8 3.5.3 RCIC System...................................

3.5-12 3.6 CONTAINMENT SYSTEMS...................................

3.6-1 3.6.1.1 Primary Containment...................................

3.6-1 3.6.1.2 Primary Containment Air Lock...................................

3.6-3 3.6.1.3 Primary Containment Isolation Valves (PCIVs)................................... 3.6-7 3.6.1.4 Drywell Air Temperature...................................

3.6-14 3.6.1.5 Reactor Building-to-Suppression Chamber Vacuum Breakers......................................

3.6-15 3.6.1.6 Suppression Chamber-to-Drywell Vacuum Breakers........................... 3.6-18 3.6.2.1 Suppression Pool Average Temperature......................................

3.6-20 3.6.2.2 Suppression Pool Water Level......................................

3.6-23 3.6.2.3 Residual Heat Removal (RHR) Suppression Pool Cooling.....................................

3.6-24 3.6.3.1 Primary Containment Oxygen Concentration..................................... 3.6-26 3.6.3.2 Containment Atmosphere Dilution (CAD) System................................

3.6-27 3.6.4.1 Secondary Containment.....................................

3.6-29 3.6.4.2 Secondary Containment Isolation Dampers (SCIDs)............................ 3.6-31 3.6.4.3 Standby Gas Treatment (SGT) System.....................................

3.6-34 (continued)

Brunswick Unit 1 ii Revision 3/2/06 l

TABLE OF CONTENTS (continued) 3.7 PLANT SYSTEMS.............................................

3.7-1 3.7.1 Residual Heat Removal Service Water (RHRSW) System............

....... 3.7-1 3.7.2 Service Water (SW) System and Ultimate Heat Sink (UHS)..........................................

3.7-4 3.7.3 Control Room Emergency Ventilation (CREV) System................

......... 3.7-11 3.7.4 Control Room Air Conditioning (AC) System......................................... 3.7-15 3.7.5 Main Condenser Offgas..........................................

3.7-18 3.7.6 The Main Turbine Bypass System..........................................

3.7-20 3.7.7 Spent Fuel Storage Pool Water Level................................

.......... 3.7-22 3.8 ELECTRICAL POWER SYSTEMS..........................................

3.8-1 3.8.1 AC Sources-Operating.......................................... 3.8-1 3.8.2 AC Sources-Shutdown..........................................

3.8-16 3.8.3 Diesel Fuel Oil..........................................

3.8-20 3.8.4 DC Sources-Operating.......................................... 3.8-23 3.8.5 DC Sources-Shutdown..........................................

3.8-27 3.8.6 Battery Cell Parameters.......................................... 3.8-30 3.8.7 Distribution Systems-Operating...................

....................... 3.8-34 3.8.8 Distribution Systems-Shutdown...................

....................... 3.8-38 3.9 REFUELING OPERATIONS.......................................... 3.9-1 3.9.1 Refueling Equipment Interlocks..........................................

3.9-1 3.9.2 Refuel Position One-Rod-Out Interlock..........................................

3.9-3 3.9.3 Control Rod Position..........................................

3.9-5 3.9.4 Control Rod Position Indication..........................................

3.9-6 3.9.5 Control Rod OPERABILITY-Refueling..........................................

3.9-8 3.9.6 Reactor Pressure Vessel (RPV) Water Level........................................ 3.9-9 3.9.7 Residual Heat Removal (RHR)- High Water Level.............................. 3.9-10 3.9.8 Residual Heat Removal (RHR)- Low Water Level.....................

......... 3.9-13 3.10 SPECIAL OPERATIONS..........................................

3.10-1 3.10.1 Inservice Leak and Hydrostatic Testing Operation...........................

..... 3.10-1 3.10.2 Reactor Mode Switch Interlock Testing..........................................

3.10-4 3.10.3 Single Control Rod Withdrawal-Hot Shutdown...................................

3.10-6 3.10.4 Single Control Rod Withdrawal-Cold Shutdown.................................. 3.10-9 3.10.5 Single Control Rod Drive (CRD) Removal-Refueling.......................... 3.10-13 3.10.6 Multiple Control Rod Withdrawal-Refueling........................................ 3.10-16 3.10.7 Control Rod Testing-Operating.................

......................... 3.10-18 3.1 0.8 SHUTDOWN MARGIN (SDM) Test-Refueling.................................... 3.10-20 4.0 DESIGN FEATURES...........................................

4.C-1 4.1 Site Location..........................................

4.0-1 4.2 Reactor Core..........................................

4.C-1 4.3 Fuel Storage..........................................

4.0-2 (continued)

BrUnswick Unit 1 iii Revision 4/15/99 l

TAI3LE OF CONTENTS (continued) 5.0 ADMINISTRATIVE CONTROLS......................

5.0-1 5.1 Responsibility......................

5.0-1 5.2 Organization......................

5.0-2 5.3 Facility Staff Qualifications......................

5.0-4 5.4 Procedures......................

5.0-5 5.5 Programs and Manuals......................

5.0-6 5.6 Reporting Requirements......................

5.0-18 5.7 High Radiation Area......................

5.0-22 Brunswick Unit I iv Revision 3/8/02 l

TAI3LE OF CONTENTS 1.0 USE AND APPLICATION..................................................

1.1-1 1.1 Definitions..................................................

1.1-1 1.2 Logical Connectors..................................................

1.2-1 1.3 Completion Times..................................................

1.3-1 1.4 Frequency..................................................

1.4-1 2.0 SAFETY LIMITS (SLs)..................................................

2.0-1 2.1 SLs..................................................

2.0-1 2.2 SL Violations..................................................

2.0-1 3.0 LIMITING CONDITION FOR OPERATION (LCO) APPLICABILITY........... 3.0-1 3.0 SURVEILLANCE REQUIREMENT (SR) APPLICABILITY.............

............. 3.04 3.1 REACTIVITY CONTROL SYSTEMS..................................................

3.1-1 3.1.1 SHUTDOWN MARGIN (SDM)..................................................

3.1-1 3.1.2 Reactivity Anomalies..................................................

3.1-5 3.1.3 Control Rod OPERABILITY..................................................

3.1-7 3.1.4 Control Rod Scram Times..................

................................ 3.1-12 3.1.5 Control Rod Scram Accumulators..................................................

3.1-15 3.1.6 Rod Pattern Control..................................................

3.1-18 3.1.7 Standby Liquid Control (SLC) System.................................................. 3.1-20 3.1.8 Scram Discharge Volume (SDV) Vent and Drain Valves.......................

3.1-25 3.2 POWER DISTRIBUTION LIMITS.......................

3.2-1 3.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR)......................................

3.2-1 3.2.2 MINIMUM CRITICAL POWER RATIO (MCPR)..................................... 3.2-2 3.3 INSTRUMENTATION......................................

3.3-1 3.3.1.1 Reactor Protection System (RPS) Instrumentation...............................

3.3-1 3.3.1.2 Source Range Monitor (SRM) Instrumentation..................................... 3.3-12 3.3.2.1 Control Rod Block Instrumentation...............

....................... 3.3-18 3.3.2.2 Feedwater and Main Turbine High Water Level Trip Instrumentation................................................................................... 3.3-24 3.3.3.1 Post Accident Monitoring (PAM) Instrumentation.................................. 3.3-26 3.3.3.2 Remote Shutdown Monitoring Instrumentation.................................... 3.3-30 3.3.4.1 Anticipated Transient Without Scram Recirculation Pump Trip (ATWS-RPT) Instrumentation..

3.3-32 3.3.5.1 Emergency Core Cooling System (ECCS)

Instrumentation................................................................................... 3.3-35 3.3.5.2 Reactor Core Isolation Cooling (RCIC) System Instrumentation...................................................................................

3.345 3.3.6.1 Primary Containment Isolation Instrumentation..................................

3.349 (continued)

Brunswick Unit 2 i

Revision 3/8/03 l

TAB3LE OF CONTENTS 3.3 INSTRUMENTATION (continued) 3.3.6.2 Secondary Containment Isolation Instrumentation............................... 3.3--59 3.3.7.1 Control Room Emergency Ventilation (CREV) System Instrumentation.................................................................................

3.3-63 3.3.7.2 Condenser Vacuum Pump Isolation Instrumentation............................ 3.3*-66 3.3.8.1 Loss of Power (LOP) Instrumentation......................................

3.3*-69 3.3.8.2 Reactor Protection System (RPS) Electric Power Monitoring.........................................................................................

3.3-72 3.4 REACTOR COOLANT SYSTEM (RCS)..............................

3.4-1 3.4.1 Recirculation Loops Operating..............................

3.4*-1 3.4.2 Jet Pumps..............................

3.4-3 3.4 3 Safety/Relief Valves (SRVs)..............................

3.4-5 3.4.4 RCS Operational LEAKAGE..............................

3.4-7 3.4 5 RCS Leakage Detection Instrumentation..............................

3.4-9 3.4 6 RCS Specific Activity..............................

3.4-12 3.4.7 Residual Heat Removal (RHR) Shutdown Cooling System-Hot Shutdown..

3.4-14 3.4..8 Residual Heat Removal (RHR) Shutdown Cooling System-Cold Shutdown................................

3.4-17 3.4.9 RCS Pressure and Temperature (P/T) Limits................................

3.4-19 3.4.10 Reactor Steam Dome Pressure................................

3.4-28 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) AND REACTOR CORE ISOLATION COOLING (RCIC) SYSTEM....................................

3.5-1 3.5.1 ECCS-Operating....................................

3.5-1 3.5.2 ECCS-Shutdown...................................

3.5-8 3.5.3 RCIC System..................................

3.5-12 3.6 CONTAINMENT SYSTEMS...................................

3.6-1 3.6.1.1 Primary Containment..................................

3.6-1 3.6.1.2 Primary Containment Air Lock...................................

3.6-3 3.6.1.3 Primary Containment Isolation Valves (PCIVs)................................... 3.6-7 3.6.1.4 Drywell Air Temperature...................................

3.6-14 3.6.1.5 Reactor Building-to-Suppression Chamber Vacuum Breakers......................................

3.6-15 3.6.1.6 Suppression Chamber-to-Drywell Vacuum Breakers........................... 3.6-18 3.6.2.1 Suppression Pool Average Temperature......................................

3.6-20 3.6.2.2 Suppression Pool Water Level......................................

3.6-23 3.6.2.3 Residual Heat Removal (RHR) Suppression Pool Cooling.....................................

3.6-24 3.6 3.1 Primary Containment Oxygen Concentration..................................... 3.6-26 3.6.3.2 Containment Atmosphere Dilution (CAD) System................................

3.6-27 3.6.4.1 Secondary Containment.....................................

3.6-29 3.6.4.2 Secondary Containment Isolation Dampers (SCIDs)............................ 3.6-31 3.6.4.3 Standby Gas Treatment (SGT) System.....................................

3.6-34 (continued)

Brunswick Unit 2 ii Revision 3/2/06 1

TABLE OF CONTENTS (continued) 3.7 PLANT SYSTEMS............................................

3.7-1 3.7.1 Residual Heat Removal Service Water (RHRSW) System...........

........ 3.7-1 3.7.2 Service Water (SW) System and Ultimate Heat Sink (UHS)..........................................

3.7-4 3.7.3 Control Room Emergency Ventilation (CREV) System...............

.......... 3.7-11 3.7.4 Control Room Air Conditioning (AC) System......................................... 3.7-15 3.7.5 Main Condenser Offgas..........................................

3.7-18 3.7.6 The Main Turbine Bypass System..........................................

3.7-20 3.7.7 Spent Fuel Storage Pool Water Level..................................

........ 3.7-22 3.8 ELECTRICAL POWER SYSTEMS..........................................

3.8-1 3.8.1 AC Sources-Operating..........................................

3.8-1 3.8.2 AC Sources-Shutdown.......................................... 3.8-16 3.8.3 Diesel Fuel Oil..........................................

3.8-20 3.8.4 DC Sources-Operating..........................................

3.8.23 3.8.5 DC Sources-Shutdown..........................................

3.8..27 3.8.6 Battery Cell Parameters.......................................... 3.8-.30 3.8.7 Distribution Systems-Operating...................

....................... 3.8-.34 3.8.8 Distribution Systems-Shutdown...................

....................... 3.8-38 3.9 REFUELING OPERATIONS.......................................... 3.9-*1 3.9.1 Refueling Equipment Interlocks...............

........................... 3.9-1 3.9.2 Refuel Position One-Rod-Out Interlock..........................................

3.9-.3 3.9.3 Control Rod Position..........................................

3.9-.5 3.9.4 Control Rod Position Indication..........................................

3.9.6 3.9.5 Control Rod OPERABILITY-Refueling..........................................

3.9-8 3.9.6 Reactor Pressure Vessel (RPV) Water Level........................................ 3.9.9 3.9.7 Residual Heat Removal (RHR-High Water Level................................ 3.9.10 3.9.8 Residual Heat Removal (RHR-Low Water Level................................. 3.9.13 3.10 SPECIAL OPERATIONS..........................................

3.10-1 3.10.1 Inservice Leak and Hydrostatic Testing Operation..........................

...... 3.11)-1 3.10.2 Reactor Mode Switch Interlock Testing..........................................

3.10-4 3.10.3 Single Control Rod Withdrawal-Hot Shutdown...................................

3.10-6 3.10.4 Single Control Rod Withdrawal-Cold Shutdown.................................. 3.10-9 3.10.5 Single Control Rod Drive (CRD) Removal-Refueling.......................... 3.10)-13 3.10.6 Multiple Control Rod Withdrawal-Refueling........................................ 3.10-16 3.10.7 Control Rod Testing-Operating..................

........................ 3.10-18 3.10.8 SHUTDOWN MARGIN (SDM) Test-Refueling....................................

3.10-20 4.0 DESIGN FEATURES..........................................

4.0-1 4.1 Site Location..........................................

4.0-1 4.2 Reactor Core..........................................

4.0-1 4.3 Fuel Storage..........................................

4.0-2 (continued)

Brunswick Unit 2 iii Revision 4/15199 I

TABLE OF CONTENTS (continued) 5.0 ADMINISTRATIVE CONTROLS.......................

5.0-1 5.1 Responsibility.......................

5.0-1 5.2 Organization.......................

5.0-2 5.3 Facility Staff Qualifications.......................

5.0-4 5.4 Procedures.......................

5.0-5 5.5 Programs and Manuals.......................

5.0-6 5.6 Reporting Requirements.......................

5.0-18 5.7 High Radiation Area.......................

5.0-22 Brunswick Unit 2 iv Revision 11/21/02 l

BSEP 06-0044, Submittal of Technical Specification Bases Revisions

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