Text

Application to Allow Selective Implementation of Alternate Source Term to Analyze the Dose Consequences Associated with Fuel Handling Accidents (WBN-TS-11-19)
	ML12171A317
Person / Time
Site:	Watts Bar
Issue date:	06/13/2012
From:	James Shea; Tennessee Valley Authority
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	WBN-TS-11-19
	Download: ML12171A317 (132)
	v • d • e

Tennessee Valley Authority, 1101 Market Street, Chattanooga, Tennessee 37402 June 13, 2012 10 CFR 50.90 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, D.C. 20555-0001 Watts Bar Nuclear Plant, Unit 1 Facility Operating License No. NPF-90 NRC Docket No. 50-390

Subject:

Watts Bar Nuclear Plant Unit 1 - Application to Allow Selective Implementation of Alternate Source Term to Analyze the Dose Consequences Associated with Fuel Handling Accidents (WBN-TS-11-19)

In accordance with the provisions of 10 CFR 50.90, "Application for amendment of license, construction permit, or early site permit," and 10 CFR 50.67, "Accident source term," the Tennessee Valley Authority (TVA) requests a change (WBN-TS-1 1-19) to Watts Bar Nuclear Plant (WBN), Unit 1, Facility Operating License No. NPF-90.

The proposed amendment will:

1. Permit selective implementation of the Alternate Source Term (AST) methodology in accordance with 10 CFR 50.67 and Regulatory Guide (RG) 1.183, "Alternative Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors." Implementation of the AST methodology will be limited to the analysis of Fuel Handling Accidents (FHAs) for WBN, Unit 1;

2. Add WBN, Unit 1 Technical Specification (TS) 3.9.10, "Decay Time," to restrict movement of irradiated fuel assemblies until 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months after the reactor core has become sub-critical. TS 3.9.10 ensures that the irradiated fuel meets the minimum decay time established in the radiological analysis of the FHA;

3. Modify WBN, Unit 1 TS 3.3.6, "Containment Vent Isolation Instrumentation," TS 3.3.8, "Auxiliary Building Gas Treatment System (ABGTS) Actuation Instrumentation," and TS 3.7.12, "Auxiliary Building Gas Treatment System (ABGTS)," to eliminate the requirements associated with movement of irradiated fuel in the containment or the fuel handling area; printed on recycled paper

U.S. Nuclear Regulatory Commission Page 2 June 13, 2012

4. Eliminate TS 3.9.4, "Containment Penetrations," and TS 3.9.8, "Reactor Building Purge Air Cleanup Units;" and

5. Modify WBN, Unit 1 TS 5.7.2.20, "Control Room Envelope Habitability Program," to incorporate the Control Room dose limit defined in 10 CFR 50.67(b)(2)(iii). to this letter provides a description, technical evaluation, regulatory evaluation, and discussion of environmental considerations of the proposed changes.

Attachments 1 and 2 to the enclosure provide the existing TS and TS Bases pages marked-up to show the proposed changes. Attachments 3 and 4 to the enclosure provide the existing TS and TS Bases pages retyped to show the proposed changes.

The TS Bases pages are provided to the NRC for information only. to this letter provides the WBN, Unit 1 calculation utilized to determine the radiological consequences associated with a FHA. provides the data files for the meteorological data that supports calculation of the atmospheric dispersion factors.

TVA requests approval of the proposed license amendment by June 13, 2013, with implementation within 60 days of issuance.

The WBN Plant Operations Review Committee and the WBN Nuclear Safety Review Board have reviewed this proposed change and determined that operation of WBN in accordance with the proposed change will not endanger the health and safety of the public.

Additionally, in accordance with 10 CFR 50.91(b)(1), TVA is sending a copy of this letter and Enclosures to the Tennessee Department of Environment and Conservation.

This submittal does not contain any new regulatory commitments. Please address any questions regarding this request to Terry Cribbe, Corporate Licensing Manager, at 423-751-3850.

U.S. Nuclear Regulatory Commission Page 3 June 13, 2012 I declare under penalty of perjury that the foregoing is true and correct. Executed on this 13 day of June 2012.

Respe ly, SShea nager, Corporate Nuclear Licensing

Enclosures:

1. Evaluation of Proposed Change

2. Calculation WBNTSR-009, Control Room Operator and Offsite Doses from a Fuel Handling Accident, Revision 14

3. Compact Disc with Meteorological Data File for Calendar Years 1991 - 2010 and the ARCON96 Output File cc (Enclosures 1 and 2):

NRC Regional Administrator - Region II NRC Resident Inspector- Watts Bar Nuclear Plant, Unit 1 NRC Resident Inspector - Watts Bar Nuclear Plant, Unit 2 Director, Division of Radiological Health - Tennessee State Department of Environment and Conservation

ENCLOSURE I TENNESSEE VALLEY AUTHORITY WATTS BAR NUCLEAR PLANT, UNIT I EVALUATION OF PROPOSED CHANGE

Subject:

Application to Allow Selective Implementation of Alternate Source Term to Analyze the Dose Consequences Associated with Fuel Handling Accidents (WBN-TS-1 1-19)

1.

SUMMARY

DESCRIPTION

2. DETAILED DESCRIPTION 2.1 Proposed Changes 2.2 Need for Proposed Changes

3. TECHNICAL EVALUATION 3.1 Introduction 3.2 Computer Codes 3.3 Accident Source Term 3.4 Dose Calculation Methodology 3.5 Radiological Consequences - FHA Analysis

4. REGULATORY EVALUATION 4.1 Applicable Regulatory Requirements 4.2 Precedent 4.3 Significant Hazards Consideration 4.4 Conclusion

5. ENVIRONMENTAL CONSIDERATION ATTACHMENTS

1. Proposed TS Changes (Mark-Up) for WBN, Unit 1

2. Proposed TS Bases Changes (Mark-Up) for WBN, Unit 1

3. Proposed TS Changes (Final Typed) for WBN, Unit 1

4. Proposed TS Bases Changes (Final Typed) for WBN, Unit 1

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SUMMARY

DESCRIPTION The Tennessee Valley Authority (TVA) is proposing to amend Watts Bar Nuclear Plant (WBN), Unit 1, Facility Operating License No. NPF-90. The proposed change will selectively implement an Alternate Source Term (AST) methodology in accordance with Regulatory Position C. 1.2.2 of Regulatory Guide (RG) 1.183, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors," by modifying the WBN, Unit 1 licensing basis for determining offsite and Control Room doses due to a Fuel Handling Accident (FHA). A license amendment is required for AST implementation in accordance with 10 CFR 50.67(b)(1) which states:

"A licensee who seeks to revise its current accident source term in design basis radiological consequence analyses shall apply for a license amendment under Sec. 50.90. The application shall contain an evaluation of the consequences of applicable design basis accidents previously analyzed in the safety analysis report."

The proposed amendment will:

1. Add WBN, Unit 1 Technical Specification (TS) 3.9.10, "Decay Time," to restrict movement of irradiated fuel assemblies until 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months after the reactor core has become sub-critical. TS 3.9.10 ensures that the irradiated fuel meets the minimum decay time established in the radiological analysis of the FHA;

2. Modify WBN, Unit 1 TS 3.3.6, "Containment Vent Isolation Instrumentation," TS 3.3.8, "Auxiliary Building Gas Treatment System (ABGTS) Actuation Instrumentation," and TS 3.7.12, "Auxiliary Building Gas Treatment System (ABGTS)," to eliminate the requirements associated with movement of irradiated fuel in the containment or the fuel handling area;

3. Eliminate TS 3.9.4, "Containment Penetrations," and TS 3.9.8, "Reactor Building Purge Air Cleanup Units;" and

4. Modify WBN, Unit 1 TS 5.7.2.20, "Control Room Envelope Habitability Program," to incorporate the Control Room dose limit defined in 10 CFR 50.67(b)(2)(iii).

Mark-ups of the affected TS and TS Bases pages are included in Attachments 1 and 2.

Final typed versions of the affected TS and TS Bases pages are included in Attachments 3 and 4. The TS Bases pages are provided to the NRC for information only.

2.0 DETAILED DESCRIPTION 2.1 Proposed Changes The proposed change to the WBN, Unit 1 licensing basis involves the adoption of the AST methodology for calculating accident doses to Control Room personnel and offsite receptors following a FHA in accordance with 10 CFR 50.67 and RG 1.183. The following changes to the WBN, Unit 1 Technical Specifications are requested to support adoption of the new AST analysis of the FHA.

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TS 3.3.6. "Containment Vent Isolation Instrumentation" TVA proposes the following changes to TS 3.3.6:

1. Elimination of the specified condition "during movement of irradiated fuel assemblies within containment" from the Applicability Section;

2. Elimination of the Note from Condition B that states that it is only applicable in MODE 1, 2, 3, or 4;

3. Elimination of Condition C and the associated Required Actions;

4. Elimination of footnotes (a) and (b) from Table 3.3.6-1; and

5. Elimination of Allowable Values for the Containment Purge Exhaust Radiation Monitors that apply during movement of irradiated fuel assemblies within containment.

These changes reflect that: 1) the AST analysis of the FHA does not credit containment isolation, and 2) TS 3.9.10 is added to restrict movement of irradiated fuel assemblies until 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months after the reactor core has become sub-critical. TS 3.9.10 ensures that the irradiated fuel meets the minimum decay time established in the radiological analysis of the FHA.

TS 3.3.8, "Auxiliary Building Gas Treatment System (ABGTS) Actuation Instrumentation" TVA proposes the following changes to TS 3.3.8:

1. Elimination of Condition C and the associated Required Actions;

2. Elimination of the reference to MODE 1, 2, 3, or 4 in Condition D;

3. Renumber Condition D, Required Action D.1, and Required Action D.2 as Condition C, Required Action C.1, and Required Action C.2;

4. Elimination of Surveillance Requirements (SRs) 3.3.8.1, 3.3.8.2, and 3.3.8.4;

5. Renumbering SR 3.3.8.3 as SR 3.3.8.1 in the Surveillance Requirements Section and Table 3.3.8-1;

6. Elimination of footnote (a) from Table 3.3.8-1; and

7. Elimination of the requirements regarding the Fuel Pool Area Radiation Monitors from Table 3.3.8-1.

These changes reflect that: 1) the AST analysis of the FHA does not credit actuation of the Auxiliary Building Gas Treatment Gas (ABGTS), and 2) TS 3.9.10 is added to restrict movement of irradiated fuel assemblies until 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months after the reactor core has become sub-critical. TS 3.9.10 ensures that the irradiated fuel meets the minimum decay time established in the radiological analysis of the FHA.

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TS 3.7.12, "Auxiliary Building Gas Treatment System (ABGTS)"

TVA proposes the following changes to TS 3.7.12:

1. Elimination of the specified condition "during movement of irradiated fuel assemblies in the fuel handling area" from the Applicability Section;

2. Elimination of the references to MODE 1, 2, 3, or 4 in Condition B; and

3. Elimination of Conditions C and D and their associated Required Actions.

These changes reflect: 1) the AST analysis of the FHA does not credit actuation of the ABGTS, and 2) TS 3.9.10 is added to restrict movement of irradiated fuel assemblies until 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months after the reactor core has become sub-critical. TS 3.9.10 ensures that the irradiated fuel meets the minimum decay time established in the radiological analysis of the FHA.

TS 3.9.4, "Containment Penetrations" TVA proposes to eliminate TS 3.9.4 in its entirety is proposed. This change reflects that:

1) the AST analysis of the FHA does not credit containment isolation, and 2) TS 3.9.10, is added to restrict movement of irradiated fuel assemblies until 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months after the reactor core has become sub-critical. TS 3.9.10 ensures that the irradiated fuel meets the minimum decay time established in the radiological analysis of the FHA.

TS 3.9.8, "Reactor Building Purge Air Cleanup Units" TVA proposes to eliminate TS 3.9.8 in its entirety. This change reflects that: 1) the AST analysis of the FHA does not credit containment isolation, and 2) TS 3.9.10 is added to restrict movement of irradiated fuel assemblies until 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months after the reactor core has become sub-critical. TS 3.9.10 ensures that the irradiated fuel meets the minimum decay time established in the radiological analysis of the FHA.

TS 3.9.10, "Decay Time" TVA proposes to add TS 3.9.10. TS 3.9.10 is added to restrict movement of irradiated fuel assemblies until 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months after the reactor core has become sub-critical. TS 3.9.10 ensures that the irradiated fuel meets the minimum decay time established in the radiological analysis of the FHA.

TS 5.7.2.20, "Control Room Envelope Habitability Program TS 5.7.2.20 currently states:

"...The program shall ensure that adequate radiation protection is provided to permit access and occupancy of the CRE under design basis accident (DBA) conditions without personnel receiving radiation exposures in excess of 5 rem whole body or its equivalent to any part of the body for the duration of the accident..."

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TS 5.7.2.20 would be revised to state:

"...The program shall ensure that adequate radiation protection is provided to permit access and occupancy of the CRE under design basis accident (DBA) conditions without personnel receiving radiation exposures in excess of the applicable regulatory requirement (i.e., 5 rem Total Effective Dose Equivalent (TEDE) for a fuel handling accident or 5 rem whole body or its equivalent to any part of the body for other accidents) for the duration of the accident..."

Bases for Technical Specifications 3.3.6, 3.3.8. 3.7.12, 3.7.13, 3.9.4, 3.9.7, 3.9.8, and 3.9.10 Conforming changes to the Bases for TSs 3.3.6, 3.3.8, 3.7.12, 3.9.4, and 3.9.8 are made to address the selective implementation of the AST for the FHA analysis. In addition, Bases for TS 3.9.10 are added, and changes to the Bases for TSs 3.7.13 and 3.9.7 are made to denote that 10 CFR 50.67 and Regulatory Position C.4.4 of RG 1.183 define the regulatory dose limits for the fuel handling accident.

2.2 Need for Proposed Changes Section 15.5.6 of the WBN, Unit 1 Updated Final Safety Analysis Report (UFSAR) provides the current WBN, Unit 1 licensing basis for the radiological analyses of the FHA. The analysis is based on RG 1.25, "Assumptions Used for Evaluating the Potential Radiological Consequences of a Fuel Handling Accident in the Fuel Handling and Storage Facility for Boiling and Pressurized Water Reactors," and NUREG/CR-5009, "Assessment of the Use of Extended Burnup Fuel in Light Water Power Reactors."

The dose was determined by utilizing dose equations from TID-14844. Dose conversion factors in International Commission on Radiation Protection (ICRP) Publication 30 were used to determine thyroid doses in place of those found in TID-14844.

On September 14, 2010, TVA discovered a conflict between the dose calculation for the FHA analysis and the calculation that determines the response time for the radiation monitor system, including the time for closure of the isolation damper. The dose calculation assumes that automatic isolation of the Auxiliary Building Ventilation System occurs, such that no unfiltered releases occur subsequent to a FHA. The response time calculation determined that the total response time for isolation of the Auxiliary Building Ventilation System was 12.5 seconds and the gas travel time between the spent fuel pool monitor and the Auxiliary Building Secondary Containment Enclosure (ABSCE) isolation damper was 6.3 seconds. Therefore, the potential exists for an unfiltered release to occur for 6.2 seconds following a FHA in the Auxiliary Building. Problem Evaluation Report (PER) 252012 was initiated to document the issue.

A functional evaluation was performed in accordance with NRC Inspection Manual, Part 9900: "Technical Guidance." The functional evaluation utilized an alternative analytical method in accordance with Appendix C.4 of NRC Inspection Manual, Part 9900 to establish that the offsite and Control Room doses would be within the regulatory limits without isolation of the Auxiliary Building following a FHA. The alternative method utilized the AST methodology.

10 CFR 50.67 was issued to allow holders of operating licenses to voluntarily revise the traditional accident source terms used in the design basis accident (DBA) radiological E1-5

consequence analyses with ASTs, because of advances made in understanding the timing, magnitude, and chemical form of fission product releases from severe nuclear power plant accidents. RG 1.183 and NUREG-1465, "Accident Source Terms for Light-Water Nuclear Power Plants," provide guidance on selective application of the AST methodology in revising the accident source terms used in design basis radiological consequences analyses, as allowed by 10 CFR 50.67.

The proposed changes to the WBN, Unit 1 licensing basis and TS are required to resolve the existing plant condition identified in PER 252012. The analysis of the dose consequences resulting from a FHA demonstrates that the regulatory acceptance criteria are met. NRC approval of the requested changes will establish the acceptability of the use of the AST methodology for WBN, Unit 1 to analyze the dose consequences associated with a FHA.

The proposed changes to WBN, Unit 1 TS 3.3.6, TS 3.3.8, TS 3.7.12, TS 3.9.4, TS 3.9.8, and TS 5.7.2.20 and the addition of TS 3.9.10 are made to reflect the assumptions made in the revised FHA analyses. The following FHA scenarios are addressed:

1. The drop of a single fuel assembly in the spent fuel pool/Auxiliary Building with no Auxiliary Building Isolation (ABI) and with unfiltered releases through the Auxiliary Building vent.

2. The drop of a single fuel assembly in the containment. The containment is assumed to be open, and an unfiltered release occurs through the Shield Building vent for 12.7 seconds until the Reactor Building Purge Ventilating System (RBPVS) is isolated. After 12.7 seconds, the remaining release occurs through the Auxiliary Building vent with no ABI and no filtration.

Both analyses assume that the fuel decays for 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months prior to the FHA occurring, Control Room isolation occurs within 40 seconds, and the Control Room Emergency Ventilation System (CREVS) filters the air provided to the Control Room.

The Case # 1 analysis of the FHA in the Auxiliary Building is the bounding accident, because it results in a higher dose to the occupants of the Control Room. The doses to the receptors at the Exclusion Area Boundary (EAB) and Low Population Zone (LPZ) are the same for both FHA analyses.

3.0 TECHNICAL EVALUATION

3.1 Introduction 3.1.1 Control Room Area Ventilation System Description Section 9.4.1 of the WBN, Unit 1 UFSAR provides a description of the Control Room Area Ventilation System.

The Control Building air-conditioning systems are engineered safety features. Each pair of full-capacity (one redundant) water chillers and each redundant set of air handling units is served from a separate train of the emergency power system and from a coordinated separate loop of the Essential Raw Cooling Water System (ERCW).

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The Control Building outside air intakes are provided with radiation monitors, and smoke detectors. Indicators are provided with the radiation monitors. Control room common annunciation is provided. Isolation of the Main Control Room Habitability Zone (MCRHZ) occurs automatically upon the actuation of a safety injection signal or upon indication of high radiation, or smoke concentrations in the outside air supply stream to the building.

Upon receipt of a signal for Main Control Room Habitability System (MCRHS) area isolation, Control Room Isolation (CRI), the following conditions are automatically implemented:

1. The Control Building emergency air cleanup fans operate to recirculate a portion of the MCRHS area air-conditioning system return air through the cleanup trains composed of HEPA filters and charcoal adsorbers.

2. The Control Building emergency pressurizing air supply fan operates to supply a reduced stream of outside air to the Control Room air-conditioning system to maintain the MCRHZ pressurized relative to outside and the adjacent areas. This fresh air is routed through the emergency air cleanup trains.

3. The Control Room electrical board rooms air handling units continue to draw outside air to maintain the lower floor spaces at atmospheric pressure.

4. The exhaust fan in the toilet rooms is stopped, and double isolation dampers are closed.

5. The spreading room supply and exhaust fans are stopped and any operating battery room exhaust fan continues to run.

6. Double isolation dampers in the spreading room supply duct and isolation dampers in the exhaust duct close.

7. The Auxiliary Building Elevation 757 shutdown board rooms pressurizing air supply fans are automatically de-energized.

8. Double isolation valves close to isolate the normal pressurizing supply to the MCRHZ.

MCRHZ isolation may be accomplished manually at any time by the Control Room operators.

The following building air-conditioning and ventilating system components are each provided with two 100% capacity units. Each meets the single failure criterion, and automatic switchover is assured if one of the units fails. These systems include the:

1. Control room air-conditioning system, water chillers, air handling units, and piping.

2. Control Building emergency air cleanup supply fans and filter assemblies.

3. Control Building emergency pressurizing air supply fans.

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3.1.2 Fuel Handling Area Ventilation System Description Section 9.4.2 of the WBN, Unit 1 UFSAR describes the Fuel Handling Area Ventilation System. It is a subsystem of the Auxiliary Building Ventilation System.

A FHA in the Auxiliary Building is detected by two gamma radiation monitors, mounted above the spent fuel pool. The high radiation signals via redundant trains will shut off the fuel handling and Auxiliary Building general supply and exhaust fans and start the ABGTS. To accomplish its safety function following a FHA, the fuel handling area ventilation system must:

1. Isolate the normal ventilation pathways between the spent fuel pool and the environment.

2. Filter the contaminants out of the air by the ABGTS before exhausting it to the environment.

The two redundant radiation monitors (safety-related) located above the spent fuel pool assure that the accident is promptly detected and that a high radiation signal is provided to each ventilation train, even if one monitor fails. Also, during refueling operations when containment or the annulus is open to the ABSCE spaces, a Containment Vent Isolation (CVI) signal is procedurally configured to assure that a FHA in containment is promptly detected and the CIV signal is provided to each ventilation train.

3.1.3 Reactor Building Purge Ventilating System Section 9.4.6 of the WBN, Unit 1 UFSAR describes the Reactor Building Purge Ventilating System (RBPVS).

The RBPVS consists of two trains, each designed to provide 50% of the capacity required for normal operation. Each train contains an air supply fan, an air exhaust fan, a cleanup filter unit, containment isolation valves, system air flow control valves, and all necessary ductwork. The system also includes single air supply distribution and air exhaust collection subsystems as well as an instrument room supply fan and an instrument room exhaust fan.

The filtered air is discharged to the outdoors by means of the Shield Building exhaust vent located in the annular space of the Reactor Building and extending through the roof of the Reactor Building. The purge air filtration units and associated exhaust ductwork provide a safety-related filtration path following a FHA.

Three signals cause the system to change from the normal purge mode to the accident isolation mode. These signals (i.e., manual, safety injection system auto-initiate, and high purge exhaust radiation (automatic)) initiate a CVI signal. Additionally, during refueling operations whenever containment or the annulus is open to the ABSCE spaces, a high radiation signal from the spent fuel pool accident radiation monitors automatically cause the system to change from the purge mode to the accident isolation mode.

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3.1.4 Current Licensing Basis for the FHA The current WBN, Unit 1 radiological analyses of the FHA is based on RG 1.25 and NUREG/CR-5009. The dose was determined by utilizing dose equations from TID-14844. Dose conversion factors in ICRP-30 were used to determine thyroid doses in place of those found in TID-14844.

Major assumptions of the current analysis include: 1) the accident occurs 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months after plant shutdown; 2) all of the fuel rods in one fuel assembly are damaged; 3) the 24 Tritium Producing Burnable Absorber Rods (TPBARs) in a single fuel assembly are damaged and release their entire tritium content to the environment; and 4) the release is filtered by the RBPVS or ABGTS filters.

3.1.5 Proposed Changes to the WBN, Unit I Current Licensing Basis TVA proposes to revise the WBN, Unit 1 licensing basis to selectively implement the AST described in RG 1.183 through reanalysis of the radiological consequences of the FHA. As part of this selective implementation of AST, the following changes are assumed in the analysis:

The total effective dose equivalent (TEDE) acceptance criterion of 10 CFR 50.67(b)(2) and Regulatory Position C.4.4 of RG 1.183 replace the previous whole body and thyroid dose guidelines of 10 CFR 100.11 and 10 CFR 50, Appendix A, General Design Criterion (GDC) 19.

The gap activity is revised to be consistent with the guidance of RG 1.183.

An overall decontamination factor of 200 was applied to iodines in accordance with the guidance of RG. 1.183.

The release of radioactive materials, including tritium, is assumed to be linear over a two-hour time frame to be consistent with the guidance of RG 1.183.

Once or twice burned fuel assemblies may contain up to 24 TPBARs. All 24 TPBARs in a fuel assembly are assumed to break, with 25% of the tritium inventory being released to the environment.

No filtration of the release by the RBPVS or ABGTS to the environment is assumed.

No Auxiliary Building isolation is assumed.

The release path for the containment scenario is changed to include 12.7 seconds of unfiltered release through the Shield Building vent, with the remainder of the unfiltered release through the Auxiliary Building vent.

The time to isolate the Control Room is increased from 20.6 seconds to 40 seconds.

New onsite (Control Room) and offsite atmospheric dispersion factors based on more recent meteorological data (1991 through 2010) are used.

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3.2 Computer Codes Table 1 provides the computer codes used to perform the AST analysis of the FHA.

Table 1 Computer Codes Utilized in the AST Analysis of the FHA Computer Version Purpose Regulatory Precedence Code ARCON96 June 25, Onsite Safety Evaluation Report for License Amendment 1997 Atmospheric No. 59 to Facility Operating License No. NPF-90 Dispersion for WBN, Unit 1, dated January 6, 2006 Factors SCALE 4.3 Conventional NUREG/CR-0200, "SCALE: A Modular Code Core System for Performing Standardized Computer Fission Analyses for Licensing Evaluation," September Product 1998 Inventory ORIGEN - S 3.0 Safety Evaluation Report for License Amendment No. 293 to Renewed Facility Operating License No.

NPF-6 for Arkansas Nuclear One, Unit No. 2, dated April 26, 2011 ORIGEN 2.1 Tritium Safety Evaluation Report for License Amendment Production Nos. 269 and 273 to Renewed Facility Operating Core Fission License Nos. DPR-44 and DPR-56 for the Peach Product Bottom Atomic Power Station, Units 2 and 3, dated Inventory September 5, 2008 Referenced in RG 1.183 STP 7 Activity Safety Evaluation Report for License Amendment Released No. 59 to Facility Operating License No. NPF-90 after a FHA for WBN, Unit 1, dated January 6, 2006 COROD 7.1 Control Safety Evaluation Report for License Amendment Room Doses No. 59 to Facility Operating License No. NPF-90 for WBN, Unit 1, dated January 6, 2006 FENCDOSE 5 Offsite Safety Evaluation Report for License Amendment Doses No. 59 to Facility Operating License No. NPF-90 I _for WBN, Unit 1, dated January 6, 2006 The offsite X/Qs are determined utilizing a calculation methodology that conforms to RG 1.145, "Atmospheric Dispersion Models for Potential Accident Consequence Assessments at Nuclear Power Plants."

3.3 Accident Source Term 3.3.1 Fission Product Inventory Table 2 provides the specific parameters used in the core inventory calculations.

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Table 2 Parameters Used in Core Inventory Calculations Parameter Conventional Core Tritium Production Core Core Thermal Power 3459 Megawatts-thermal 3459 MWt (MWt)

Emergency Core Cooling 3.06% 0.6%

System Uncertainty Factor Number of Assemblies 193 193 (96 Once Burned, 96 Twice Burned, and 1 Thrice Burned)

Fuel Rods per Assembly 264 264 Burn-up 1500 Effective Full Power 510/1020/1530 EFPD Days (EFPD)

Enrichment 5 weight percent U-235 4.95 weight percent U-235 Core Average Assembly 18.47 MWt 18.03 MWt Power Radial Peak to Average 1.65 1.65 Ratio for Discharge Assembly The analysis assumes that all of the fuel rods in a fuel assembly rupture. Thus, the fission product inventory of the damaged fuel assembly was determined by dividing the total core inventory by the number of fuel assemblies in the core. The values assumed for individual fission product inventories are calculated assuming full power operation at the end of core life immediately preceding shutdown with a radial peaking factor of 1.65 for the standard core assembly and TPC assembly, except tritium (discussed in Section 3.3.2). The factor of 1.65 is the maximum peaking factor allowed by the Core Operating Limit Report (COLR).

Table 3 provides the source terms for the 1500 EFPD maximum burn-up of a standard core utilized in an 18 month fuel cycle. Table 4 provides the source terms for the once burned, twice burned, and three-times burned assemblies for the Tritium Production Core (TPC).

The analysis assumes a decay time of 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months prior to the movement of spent fuel.

The source terms presented in Tables 3 and 4 do not include this decay time, but it is accounted for in the STP model that is utilized to determine the activity released after a fuel handling accident.

The fission product inventory conforms to Regulatory Position C.3.1 of RG 1.183, except that an additional source of tritium is included.

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Table 3 Source Terms for 1500 EFPD Maximum Burn-up of a Standard Core for an 18-Month Fuel Cycle Nuclide Ci/Assembly Nuclide CllAssembly Nuclide CilAssembly 1 Kr-83m 5.20E+04 37 Rb-88 2.90E+05 76 Sb-127 4.69E+04 2 Kr-85m 1.04E+05 38 Rb-89 3.74E+05 77 Sb-129 1.65E+05 3 Kr-85 7.02E+03 39 Rb-90m 1.13E+05 78 Sb-130m 2.17E+05 4 Kr-87 2.06E+05 40 Rb-90 3.39E+05 79 Sb-130 5.45E+04 5 Kr-88 2.82E+05 41 Rb-91 4.65E+05 80 Sb-133 3.08E+05 6 Kr-89 3.44E+05 42 Se-84 9.25E+04 81 Te-125m 1.29E+03 7 Kr-90 3.64E+05 43 Sr-89 3.90E+05 82 Te-127m 7.91E+03 8 Xe-131m 5.64E+03 44 Sr-90 6.17E+04 83 Te-127 4.65E+04 9 Xe-133m 3.22E+04 45 Sr-91 5.06E+05 84 Te-129m 3.18E+04 10 Xe-133 9.63E+05 46 Sr-92 5.55E+05 85 Te-129 1.57E+05 11 Xe-135m 2.16E+05 47 Sr-93 6.46E+05 86 Te-131m 1.04E+05 12 Xe-135 2.90E+05 48 Sr-94 6.55E+05 87 Te-131 4.14E+05 13 Xe-137 9.15E+05 49 Y-90 6.56E+04 88 Te-132 7.06E+05 14 Xe-138 8.31E+05 50 Y-91 m 2.94E+05 89 Te-133m 4.37E+05 15Xe-139 5.99E+05 51 Y-91 5.24E+05 90 Te-133 5.35E+05 16 Xe-140 4.10E+05 52 Y-92 5.59E+05 91 Te-134 8.47E+05 53 Y-93 4.39E+05 92 Ba-1 37m 8.45E+04 171-130 1.79E+04 54 Y-94 7.11E+05 93 Ba-139 8.66E+05 18 1-131 4.94E+05 55 Y-95 7.51E+05 94 Ba-140 8.71E+05 19 1-132 7.21E+05 56 Y-96 7.35E+05 95 Ba-141 7.80E+05 20 1-133 1.OOE+06 57 Zr-95 8.05E+05 96 Ba-142 733E+05 21 1-134 1.10E+06 58 Zr-97 8.14E+05 97 La-140 9.43E+05 22 1-135 9.60E+05 59 Nb-95 8.11E+05 98 La-141 7.88E+05 23 1-136m 2.10E+05 60 Nb-97m 7.73E+05 99 La-142 7.64E+05 61 Nb-97 8.20E+05 100 La-143 7.12E+05 24 Br-83 5.20E+04 62 Mo-99 9.16E+05 101 Ce-141 7.94E+05 25 Br-84m 2.63E+03 63 Tc-99m 8.06E+05 102 Ce-143 7.20E+05 26 Br-84 9.51E+04 64 Tc-99 0.OOE+00 103 Ce-144 6.64E+05 27 Br-85 1.03E+05 65 Tc-101 8.48E÷05 104 Ce-145 5.94E+05 28 Br-87 1.61E+05 66 Ru-103 8.48E+05 105 Pr-143 6.96E+05 67 Ru-105 6.30E+05 106 Pr-144 6.69E+05 29 Cs-134 1.66E+05 68 Ru-106 3.85E+05 107 Pr-145 4.94E+05 30 Cs-135 0.OOE+00 69 Ru-107 3.88E+05 108 Np-239 1.11E+07 31 Cs-136 4.90E+04 70 Rh-103m 8.46E+05 32 Cs-137 8.90E+04 71 Rh-105m 1.79E+05 33 Cs-138 9.09E+05 72 Rh-105 5.95E+05 34 Cs-139 8.40E+05 73 Rh-106 4.08E+05 35 Cs-140 7.52E+05 74 Rh-107 3.88E+05 36 Cs-141 5.70E+05 75 Sn-130 1.62E+05 Others (corrosioniactivation products) CilAssembly Cr-51 1.65E+04 Fe-55 3.65E+03 Co-60m 9.82E+03 Mn-54 9.22E+02 Fe-59 2.74E+02 Ni-63 3.98E+02 Mn-56 2.41E+04 Co-58 6.07E+03 Ni-65 5.84E+02 Mn-57 3.21E+00 Co-60 6.60E+03 I E1-12

Table 4 (Page I of 3)

Sources Terms for the Once Burned, Twice Burned, and Three Times Burned Fuel Assemblies for the TPC Core WBN 96-Feed Equilibrium Core, End-of-cycle Operation at 3480 MWt for 510 days I Average Assembly Inventories (Ci)

Nuclide Total Core Inventory IX Burned 2X Burned 3X Burned Core Avg.

Kr-83m 1.23E+07 7.63E+04 5.15E+04 6.13E+04 6.39E+04 Kr-85m 2.69E+07 1.69E+05 1.iOE+05 1.25E+05 1.39E+05 Kr-85 8.81 E+05 3.56E+03 5.54E+03 6.84E+03 4.56E+03 Kr-87 5.23E+07 3.31 E+05 2.11 E+05 2.36E+05 2.71 E+05 Kr-88 7.38E+07 4.68E+05 2.97E+05 3.31 E+05 3.82E+05 Kr-89 9.1OE+07 5.81 E+05 3.63E+05 3.97E+05 4.72E+05 Kr-90 9.01 E+07 5.76E+05 3.59E+05 3.92E+05 4.67E+05 Xe-i31mn 9.54E+05 5.31 E+03 4.56E+03 6.18E+03 4.94E+03 Xe-133m 5.80E+06 3.41 E+04 2.60E+04 3.45E+04 3.01 E+04 Xe-133 1.88E+08 1.11 E+06 8.36E+05 1.09E+06 9.75E+05 Xe-1 35m 3.59E+07 2.08E+05 1.63E+05 2.19E+05 1.86E+05 Xe-i 35 4.96E+07 2.84E+05 2.30E+05 2.19E+05 2.57E+05 Xe-137 1.65E+08 9.75E+05 7.30E+05 9.51 E+05 8.53E+05 Xe-138 1.59E+08 9.55E+05 6.93E+05 8.79E+05 8.24E+05 Xe-139 1.26E+08 7.57E+05 5.43E+05 6.83E+05 6.50E+05 Xe- 140 8.32E+07 5.07E+05 3.54E+05 4.36E+05 4.31 E+05 1-130 2.34E+06 9.02E+03 1.51 E+04 3.14E+04 1.21 E+04 1-131 9.01 E+07 5.24E+05 4.09E+05 5.49E+05 4.67E+05 1-132 1.31 E+08 7.63E+05 5.89E+05 7.87E+05 6.77E+05 1-133 1.88E+08 1.11E+06 8.35E+05 1.09E+06 9.75E+05 1-134 2.08E+08 1.23E+06 9.18E+05 1.19E+06 1.08E+06 1-135 1.76E+08 1.04E+06 7.81 E+05 1.02E+06 9.09E+05 1-1 36m 5.05E+07 3.02E+05 2.21 E+05 2.83E+05 2.62E+05 Br-83 1.23E+07 7.63E+04 5.14E+04 6.11 E+04 6.38E+04 Br-84mn 6.68E+05 3.86E+03 3.06E+03 4.13E+03 3.46E+03 Br-84 2.18E+07 1.37E+05 8.95E+04 1.03E+05 1.13E+05 Br-85 2.65E+07 1.67E+05 1.08E+05 1.23E+05 1.38E+05 Br-87 4.40E+07 2.79E+05 1.77E+05 1.98E+05 2.28E+05 Cs-1 34 1.12E+07 3.13E+04 8.41 E+04 1.48E+05 5.81 E+04 Cs-1 35 3.60E+01 1.30E-01 2.42E-01 3.48E-01 1.87E-01 Cs-136 3.67E+06 1.53E+04 2.24E+04 4.57E+04 1.90E+04 Cs-137 8.81 E+06 3.34E+04 5.76E+04 7.66E+04 4.56E+04 Cs-138 1.75E+08 1.05E+06 7.67E+05 9.80E+05 9.09E+05 Cs-i 39 1.66E+08 9.95E+05 7.27E+05 9.28E+05 8.61 E+05 Cs-140 1.49E+08 8.95E+05 6.53E+05 8.32E+05 7.74E+05 Cs-141 1.12E+08 6.75E+05 4.86E+05 6.14E+05 5.81 E+05 Rb-88 7.48E+07 4.74E+05 3.02E+05 3.37E+05 3.88E+05 Rb-89 9.64E+07 6.13E+05 3.87E+05 4.29E+05 4.99E+05 Rb-90mn 2.13E+07 1.33E+05 8.74E+04 1.01E+05 1.1OE+05 Rb-9o 9.41 E+07 6.OOE+05 3.76E+05 4.13E+05 4.88E+05 Rb-91 1.15E+08 7.28E+05 4.66E+05 5.24E+05 5.96E+05 Se-84 2.12E+07 1.33E+05 8.66E+04 9.93E+04 1.10E+05 Sr-89 1.02E+08 6.47E+05 4.06E+05 4.54E+05 5.26E+05 Sr-90 6.94E+06 2.78E+04 4.40E+04 5.45E+04 3.60E+04 Sr-91 1.23E+08 7.72E+05 4.99E+05 5.68E+05 6.35E+05 Sr-92 1.31E+08 8.14E+05 5.39E+05 6.27E+05 6.76E+05 Sr-93 1.45E+08 8.97E+05 6.08E+05 7.26E+05 7.53E+05 Sr-94 1.36E+08 8.39E+05 5.75E+05 6.92E+05 7.07E+05 Y-90 7.21 E+06 2.89E+04 4.57E+04 5.78E+04 3.74E+04 Y-91 M 7.11 E+07 4.48E+05 2.90E+05 3.30E+05 3.68E+05 Y-91 1.29E+08 8.12E+05 5.23E+05 5.99E+05 6.67E+05 El-13

Table 4 (Page 2 of 3)

Sources Terms for the Once Burned, Twice Burned, and Three Times Burned Fuel Assemblies for the TPC Core WBN 96-Feed Equilibrium Core, End-of-cycle Operation at 3480 MWt for 510 days - continued Average Assembly Inventories (Ci)

Total Core Nuclide Inventory (Ci) 1X Burned 2X Burned 3X Burned Core Avg.

Y-92 1.31E+08 8.17E+05 5.41E+05 6.30E+05 6.79E+05 Y-93 1.49E+08 9.16E+05 6.23E+05 7.46E+05 7.70E+05 Y-94 1.48E+08 9.07E+05 6.28E+05 7.63E+05 7.67E+05 Y-95 1.57E+08 9.54E+05 6.72E+05 8.31E+05 8.13E+05 Y-96 1.48E+08 8.99E+05 6.38E+05 7.93E+05 7.69E+05 Zr-95 1.64E+08 9.97E+05 7.05E+05 8.71E+05 8.51E+05 Zr-97 1.57E+08 9.38E+05 6.89E+05 8.83E+05 8.13E+05 Nb-95 1.66E+08 1.00E+06 7.13E+05 8.78E+05 8.59E+05 Nb-97m 1.49E+08 8.89E+05 6.53E+05 8.38E+05 7.72E+05 Nb-97 1.58E+08 9.45E+05 6.95E+05 8.92E+05 8.20E+05 Mo-99 1.68E+08 9.93E+05 7.48E+05 9.80E+05 8.72E+05 Tc-99m 1.47E+08 8.70E+05 6.55E+05 8.58E+05 7.63E+05 Tc-99 1.12E+03 4.36E+00 7.23E+00 9.15E+00 5.81E+00 Tc-101 1.54E+08 8.96E+05 6.94E+05 9.29E+05 7.95E+05 Ru-103 1.31 E+08 7.25E+05 6.28E+05 8.97E+05 6.78E+05 Ru-105 8.13E+07 4.21E+05 4.19E+05 6.55E+05 4.21E+05 Ru-106 3.56E+07 1.37E+05 2.30E+05 3.49E+05 1.84E+05 Ru-1 07 4.33E+07 2.08E+05 2.38E+05 3.96E+05 2.24E+05 Rh-103m 1.18E+08 6.53E+05 5.66E+05 8.08E+05 6.1OE+05 Rh-105m 2.28E+07 1.18E+05 1.17E+05 1.83E+05 1.18E+05 Rh-1 05 7.59E+07 3.92E+05 3.92E+05 5.92E+05 3.93E+05 Rh-1 06 3.94E+07 1.58E+05 2.49E+05 3.93E+05 2.04E+05 Rh-1 07 4.35E+07 2.09E+05 2.39E+05 3.98E+05 2.25E+05 Sn-130 3.14E+07 1.81E+05 1.43E+05 1.95E+05 1.63E+05 Sb-127 8.87E+06 4.88E+04 4.30E+04 6.26E+04 4.59E+04 Sb-129 2.77E+07 1.56E+05 1.30E+05 1.82E+05 1.43E+05 Sb-130m 4.16E+07 2.40E+05 1.90E+05 2.59E+05 2.16E+05 Sb-130 8.97E+06 5.04E+04 4.23E+04 5.96E+04 4.65E+04 Sb-133 5.52E+07 3.34E+05 2.38E+05 2.99E+05 2.86E+05 Te-125m 2.17E+05 7.83E+02 1.45E+03 1.95E+03 1.12E+03 Te-127m 1.15E+06 6.11E+03 5.81E+03 8.21E+03 5.97E+03 Te-127 8.77E+06 4.80E+04 4.27E+04 6.20E+04 4.54E+04 Te-129m 4.1OE+06 2.32E+04 1.93E+04 2.68E+04 2.12E+04 Te-129 2.73E+07 1.54E+05 1.28E+05 1.79E+05 1.41E+05 Te-131m 1.27E+07 7.32E+04 5.87E+04 8.04E+04 6.60E+04 Te-131 8.OOE+07 4.66E+05 3.62E+05 4.85E+05 4.15E+05 Te-132 1.29E+08 7.54E+05 5.80E+05 7.73E+05 6.67E+05 Te-133m 7.15E+07 4.34E+05 3.08E+05 3.82E+05 3.71E+05 Te-133 1.10E+08 6.48E+05 4.90E+05 6.43E+05 5.69E+05 Te-134 1.62E+08 9.81E+05 6.97E+05 8.70E+05 8.39E+05 Ba-137m 8.35E+06 3.17E+04 5.45E+04 7.26E+04 4.33E+04 Ba-139 1.71E+08 1.02E+06 7.49E+05 9.61E+05 8.84E+05 Ba-140 1.65E+08 9.88E+05 7.24E+05 9.30E+05 8.57E+05 Ba-141 1.55E+08 9.29E+05 6.79E+05 8.67E+05 8.05E+05 Ba-142 1.48E+08 8.92E+05 6.45E+05 8.17E+05 7.69E+05 La-140 1.69E+08 1.00E+06 7.48E+05 9.78E+05 8.77E+05 La-141 1.56E+08 9.33E+05 6.83E+05 8.71E+05 8.08E+05 La-142 1.52E+08 9.11E+05 6.61E+05 8.37E+05 7.86E+05 La-143 1.46E+08 8.85E+05 6.31 E+05 7.89E+05 7.58E+05 Ce-141 1.59E+08 9.51 E+05 6.94E+05 8.85E+05 8.23E+05 Ce-143 1.47E+08 8.90E+05 6.36E+05 7.96E+05 7.63E+05 Ce-144 1.17E+08 6.10E+05 6.OOE+05 6.79E+05 6.06E+05 Ce-145 9.96E+07 6.00E+05 4.32E+05 5.45E+05 5.16E+05 Pr-143 1.46E+08 8.82E+05 6.30E+05 7.84E+05 7.56E+05 Pr-144 1.18E+08 6.15E+05 6.04E+05 6.85E+05 6.1OE+05 El-14

Table 4 (Page 3 of 3)

Sources Terms for the Once Burned, Twice Burned, and Three Times Burned Fuel Assemblies for the TPC Core WBN 96-Feed Equilibrium Core, End-of-cycle Operation at 3480 MWt for 510 days - continued Average Assembly Inventories (Ci)

Total Core Nuclide Inventory (Ci) 1X Burned 2X Burned 3X Burned Core Avg.

Pr-145 9.97E+07 6.OOE+05 4.32E+05 5.45E+05 5.16E+05 Np-239 1.53E+09 8.24E+06 7.61 E+06 1.16E+07 7.95E+06 H-3 2.68E+07 3.3.2 Tritium Inventory WBN, Unit 1 is licensed to permit the production of tritium. Thus, in the analysis of the FHA, TVA includes a release of tritium, even though the FHA does not involve a temperature excursion that would result in boiling of the water covering the fuel assemblies. Note: RG 1.183 does not include the release of tritium as part of the analysis of the FHA, because standard plants do not include TPBARs.

TPBARs are installed in once and twice burned fuel assemblies, but they are not installed in fuel assemblies that are burned three times.

Following a FHA in the spent fuel pool, all 24 TPBARs in a TPC once or twice burned fuel assembly are assumed to break and release their tritium contents. Each TPBAR has 1.2 grams of tritium.

25% of the tritium released is assumed to be released to the environment following the FHA through evaporation of water. Tritium was assumed to evaporate at a constant rate over 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

100% of the tritium released from the TPBARs following a FHA will not be released to the environment, because the event does not involve temperatures that would result in boiling of the water covering the fuel assemblies.

The water tritium concentration is conservatively assumed to be 60 jtCi/gm. At this concentration, the total tritium inventory would be 84,490 Ci. This is calculated as follows:

60giCi/gm

372,000 gal

3,785.4 cc/gal

1 gm/cc

1 E-6 Ci/jgCi = 84,490 Ci 25% of this value equals 21,123 Ci (84,490 Ci

0.25)

If the temperature of the water in the spent fuel pool is maintained below the boiling point, a large fraction of the inventory will not evaporate in 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months . Ifthe normal spent fuel pool cooling system is not in service, the spent fuel pool will not reach 212°F for at least 9 hours1.041667e-4 days 0.0025 hours 1.488095e-5 weeks 3.4245e-6 months .

In the unlikely event that the spent fuel pool does boil, the boil off rate is 24,496.7 lb/hr, which is approximately 3,000 gallons/hr. Over a period of 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , this would result in a total evaporation of 6,000 gallons of spent fuel pool water. This volume is less than 2%

El-15

of the total spent fuel pool water volume. Therefore, assuming that 25% of the spent fuel pool water evaporates in 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or less is conservative.

For the containment analysis, assuming that 25% of the tritium is released to the environment following a FHA is conservative, because fuel movement would be terminated if any disruption in decay heat removal was experienced. Therefore, no boiling is expected to occur.

3.3.3 Release Fractions The FHA analysis utilizes the following release fractions: 1-131 = 0.08, Kr-85 = 0.10, and other noble gases and iodines = 0.05. Even though Table 3 of RG 1.183 specifies a gap activity for alkali metals of 12%, the FHA analysis assumes that no alkali metals are released, because particulates have essentially an infinite partition factor, which is consistent with Regulatory Position 3 of Appendix B of RG 1.183.

TVA confirmed the applicability of these release fractions, by ensuring that all of the fuel assemblies complied with Footnote 11 of RG 1.183 which states:

"The release fractions listed here have been determined to be acceptable for use with currently approved LWR fuel with a peak burnup up to 62,000 MWD/MTU provided that the maximum linear heat generation rate does not exceed 6.3 kw/ft peak rod average power for burnups exceeding 54 GWD/MTU."

Based on the previous cycles, burnups did not exceed 54 GWD/MTU except for a very small number of assemblies. The linear heat generation rates for these assemblies were much less than 6.3 kW/ft. The WBN fuel design guide is being updated to ensure that these limits are not exceeded. Thus, the use of the release fractions is appropriate for use at WBN.

The core inventory release fractions utilized in the FHA analysis conform to Regulatory Position C.3.2 of RG 1.183, Table 3 of RG 1.183, and Regulatory Positions 1.2 and 3 of Appendix B to RG 1.183.

3.3.4 Timing of Release Phases For the FHA analysis, the release from the fuel gap and the fuel pellet are assumed to occur instantaneously with the onset of the projected damage. In addition, the releases to the environment are assumed to occur in a linear ramp manner over the duration of the event.

These assumptions conform to Regulatory Position C.3.3 of RG 1.183 and Regulatory Position 3 of Appendix B of RG 1.183.

3.3.5 Radionuclide Composition As established in Section 3.3.3 of this enclosure, the core inventory release fractions utilized in the FHA analysis conform to Regulatory Positions C.3.2 and C.3.4 of RG 1.183, Table 3 of RG 1.183, and Regulatory Position 1.2 of Appendix B to RG 1.183.

E1-16

3.3.6 Chemical Form An overall effective decontamination factor of 200 is applied to the iodine in accordance with Regulatory Position 2 in Appendix B of RG 1.183. Thus, the chemical form of iodine was not expressly considered in the analysis. This conforms to Regulatory Position C.3.5 of RG 1.183.

3.3.7 Fuel Damage in FHA All the fuel rods in a single fuel assembly (264 fuel rods) are assumed to be damaged in the FHA. This assumption is consistent with the current licensing basis established in Section 15.5.6 of the WBN, Unit 1 UFSAR. This conforms to Regulatory Position C.3.6 of RG 1.183 and Regulatory Position 1.1 of Appendix B to RG 1.183.

3.4 Dose Calculation Methodology 3.4.1 Offsite Dose Consequences The offsite dose analysis of the FHA:

Determines the TEDE for the most limiting person at the Exclusion Area Boundary (EAB) and the Low Population Zone (LPZ). The STP code is utilized to determine the activity following 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months of decay. The output from this code is used as input to computer code FENCDOSE to determine the offsite doses. The offsite atmospheric dispersion factors utilized in the dose analysis of the FHA are:

Exclusion Area Boundary X/Q = 6.382E-4 sec/m3 3

Low Population Zone X/Q = 1.784E-4 sec/m This analysis conforms to the guidance of Regulatory Positions C.4.1.1, C.4.1.5 and C.4.1.6 of RG 1.183.

Dose conversion factors (DCF) from Table 5-1 of EPA 400-R-92-001, Manual of Protective Action Guides and Protective Actions for Nuclear Incidents, were utilized to calculate the TEDE. This total exposure DCF consists of 3 contributors, external exposure, inhalation exposure and exposure from ground deposition. These DCFs can be found in Tables 5-3, 5-4, and 5-5 of EPA 400-R-92-001. The Deep Dose Equivalent (DDE) component of this dose conversion factor utilizes the value from DOE/EH-0070, External Dose-Rate Conversion Factors for Calculation of Dose to the Public. The exposure to the committed effective dose equivalent (CEDE) component of these dose conversion factors are based on data provided in ICRP Publication 30.

The DDE DCFs are an exception from the guidance of Regulatory Position C.4.1.4 of RG 1.183. The methodology utilized to calculate the offsite doses, including the exceptions to RG 1.183 noted above, is conservative (higher dose) when compared to using the RG 1.183 methodology. This is due to including the DCFs associated with ground deposition. (Appendix B of Enclosure 2)

E1-17

" A breathing rate of 3.33E-4 m 3/s is embedded in the values in Table 5-4 of EPA 400-R-92-001 which is used to determine the overall DCF given in Table 5-1 as part of the dose conversion factor. This is an exception from the guidance of Regulatory Position C.4.1.3 of RG 1.183. The methodology utilized to calculate the offsite doses, including the exceptions to RG 1.183 noted above, is conservative (higher dose) when compared to using the RG 1.183 methodology. This is due to including the DCFs associated with ground deposition. (Appendix B of Enclosure 2)

" The analysis did not make any corrections for depletion of the effluent plume by deposition on the ground, but the DCF used does take into account the dose received from ground deposition. This conforms to the guidance of Regulatory Position C.4.1.7 of RG 1.183.

3.4.2 Control Room Dose Consequences The Control Room dose analysis of the FHA:

Determines the TEDE dose to the Control Room occupants due to the radioactive release associated with the FHA. The computer code COROD determines dose due to: 1) time dependent concentration of airborne activity in the Control Room; and

2) shine through the Control Room roof, Control Room ends, Auxiliary Building, Turbine Building, and Cable Spreading Room.

The Control Room dose model includes a recirculation filter model along with filtered air intake, unfiltered air inleakage, and an exhaust path. Only one train of the CREVS is assumed to be in operation. Intake flow to the Control Room is assumed to be 3,200 cubic feet per minute (cfm) before isolation. Control room isolation occurs at 40 seconds. After isolation, the total recirculation flow rate into the Control Room is 3,600 cfm (711 cfm of pressurization flow and 2,889 cfm of recirculated flow) of filtered flow plus 51 cfm of unfiltered flow from various sources (e.g., open doors, leaky valves, etc).

This conforms to the guidance of Regulatory Positions C.4.2.1 and C.4.2.4 of RG 1.183.

Utilizes the same source term, transport, and release assumptions used for determining offsite doses to determine Control Room doses. This conforms to the guidance of Regulatory Position C.4.2.2 of RG 1.183.

" Utilizes the STP code to determine the activity released to the environment. The STP code output was utilized as input to the COROD code to determine the Control Room doses. This model is consistent with the current licensing basis described in Section 15.5.6 of the WBN, Unit 1 UFSAR. This conforms to the guidance of Regulatory Position C.4.2.3 of RG 1.183.

" Does not take any credit for the use of personal protective equipment or prophylactic drugs. This conforms to the guidance of Regulatory Position C.4.2.1 of RG 1.183.

" Assumes that the hypothetical maximum exposed individual was present in the Control Room for 100% of the time during the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after the event, 60% of E1-18

the time between 1 and 4 days, and 40% of the time from 4 days to 30 days. A breathing rate of 3.33E-4 m 3/sec was utilized to calculate the CEDE. COROD utilizes Table 5-4 from EPA 400-R-92-001, Manual of Protective Action Guides and Protective Actions for Nuclear Incidents, which has the breathing rate embedded in the DCF. With the exception of the breathing rate, these assumptions conform to the guidance of Regulatory Position C.4.2.6 of RG 1.183. The methodology utilized to calculate the control room doses is conservative (higher dose) when compared to using the RG 1.183 methodology. This is due to using the point kernal integration methodology as opposed to use of the DDE DCFs. (Appendix B to Enclosure 2)

COROD calculates TEDE by summing up 100% of the gamma dose + 1% of the beta dose + concentration*DCF (Table 5-4 of EPA 400-R-92-001). The gamma dose is calculated using a point kernal integration method and the beta dose is calculated multiplying the average beta energy per disintegration

total concentration.

The parameters, including the atmospheric dispersion factors, assumed in the Control Room dose analysis of the FHA are provided in Table 5.

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Table 5 Parameters Assumed in the Control Room Dose Analysis of the FHA Parameter Value Basis Control Room Volume 257,198 ft3 Consistent with WBN, Unit 1 UFSAR Section 15.5.6 Control Room Intake Flow 3,200 cfm Consistent with WBN, Unit 1 Prior to Isolation UFSAR Section 15.5.6 Control Room 711 cfm Consistent with WBN, Unit 1 Makeup/Pressurization UFSAR Section 15.5.6 Flow Control Room Recirculation 2,889 cfm Consistent with WBN, Unit 1 Flow UFSAR Section 15.5.6 Control Room Unfiltered 51 cfm Consistent with WBN, Unit 1 Intake UFSAR Section 15.5.6 Control Room Filter 95% first pass Consistent with WBN, Unit 1 Efficiency 70% second pass UFSAR Section 15.5.6 0% for noble gases, and Tritium Control Room Isolation 40 seconds Conservative Assumption -

Time Change to WBN, Unit 1 UFSAR Section 15.5.6 Control Room Occupation 0 - 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months - 100% Consistent with WBN, Unit 1 Factors 1 - 4 days - 60% UFSAR Section 15.5.6 4 - 30 days - 40%

Release Height 32.5 meters Consistent with current licensing basis Distance to Intake 41.4 meters Consistent with current licensing basis Intake Height 14.3 meters Consistent with current licensing basis Auxiliary Building Vent X/Q 2.56E-3 sec/m 3 Change to WBN, Unit 1 0 - 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months UFSAR Table 15.5.-14 Shield Building Vent X/Q 1.09E-3 sec/m 3 Change to WBN, Unit 1 0 - 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months UFSAR Table 15.5.-14 E1-20

3.4.3 Meteorology Assumptions 3.4.3.1 Meteorological Data Meteorological data over a 20-year period (1991 through 2010) were used in the development of the X/Qs used in the AST analysis of the FHA. The WBN, Unit 1 onsite meteorological measurements program is described in Section 2.3.3 of the WBN, Unit 1 UFSAR. It states:

"The meteorological program has been developed to be consistent with the guidance given in RG 1.23 (Revision 1)..."

ARCON96 analyzes the meteorological data file used and lists the total number of hours of data processed and the number of hours of missing data in the case output. A meteorological data recovery rate may be determined from this information. The ARCON96 files present the number of hours of data processed as 175,320, and the number of missing data hours as 3,846. This yields a meteorological data recovery rate of 97.8%. Regulatory Position C.5 of RG 1.23 requires a 90% data recovery threshold for measuring and capturing meteorological data. The 97.8% valid meteorological data rate exceeds the 90% data recovery limit set forth by RG 1.23. With a data recovery rate of 97.8% and a total of 20 years of data, the contents of the meteorological data files are representative of the long-term meteorological trends at the WBN site.

3.4.3.2 Atmospheric Dispersion Factors Section 2.3.4 of the WBN, Unit 1 UFSAR provides the current licensing basis regarding the derivation of the offsite X/Qs. The current offsite X/Qs are based on onsite meteorological data for the time period of 1974 through 1993 and a RG 1.145, "Atmospheric Dispersion Models for Potential Accident Consequence Assessments at Nuclear Power Plants," calculation methodology.

Table 15.5-14 of the WBN, Unit 1 UFSAR provides the current onsite X/Qs. They are based on onsite meteorological data for the time period of 1974 through 1993 and were determined utilizing the ARCON96 computer code.

New WBN, Unit 1 onsite and offsite X/Qs are utilized in the FHA analysis. The onsite and offsite X/Qs are calculated consistent with the current licensing basis methodology, except the meteorological data was updated to reflect a more recent 20-year time period (1991 through 2010). The onsite and offsite X/Qs utilized in the dose analysis of the FHA are presented in Sections 3.4.1 and 3.4.2 of this evaluation.

The possible release pathways were considered from a FHA in either the Auxiliary Building (i.e., spent fuel pool) or containment. The most conservative pathway to the Control Room was modeled. The bounding pathway is an unfiltered release from the Auxiliary Building vent, which has the largest calculated Control Room X/Q. The EAB and LPZ X/Qs encompass all possible release points; therefore, they are bounding.

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3.4.4 Acceptance Criteria Offsite and Control Room doses must meet the guidance of Regulatory Position C.4.4 of RG 1.183 and the requirements of 10 CFR 50.67. Regulatory Position C.4.4 of RG 1.183 states:

"The radiological criteria for the EAB, the outer boundary of the LPZ, and for the control room are in 10 CFR 50.67. These criteria are stated for evaluating reactor accidents of exceedingly low probability of occurrence and low risk of public exposure to radiation, e.g., a large-break LOCA. The control room criterion applies to all accidents. For events with a higher probability of occurrence, postulated EAB and LPZ doses should not exceed the criteria tabulated in Table 6."

10 CFR 50.67(b)(2)(iii) provides the acceptance criterion for the Control Room. It states:

"Adequate radiation protection is provided to permit access to and occupancy of the control room under accident conditions without personnel receiving radiation exposures in excess of 0.05 Sv (5 rem) total effective dose equivalent (TEDE) for the duration of the accident."

Table 6 of RG 1.1.83 defines the EAB and LPZ dose criteria for the FHA as 6.3 rem TEDE.

3.5 Radiological Consequences - FHA Analysis 3.5.1 FHA Scenario Description The following FHA scenarios are addressed:

1) The drop of a single fuel assembly in the spent fuel pool/Auxiliary Building with no Auxiliary Building Isolation (ABI) and with unfiltered releases through the Auxiliary Building vent.

2) The drop of a single fuel assembly in the containment. The containment is assumed to be open, and an unfiltered release occurs through the Shield Building vent for 12.7 seconds until the RBPVS is isolated. After 12.7 seconds, the remaining release occurs through the Auxiliary Building vent with no ABI and no filtration.

Case #1 above was determined to be the bounding accident, because the Atmospheric Dispersion Factors (X/Qs) for the Auxiliary Building are greater than the X/Qs for the Shield Building. As a result, no credit is taken for isolation of the RBPVS.

Sections 3.3 and 3.4 provide the assumptions regarding the accident source term and the dose calculation methodology. The assumptions are summarized in Table 6.

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Table 6 (Page 1 of 2)

Parameters Utilized in the Analysis of the FHA Parameter AST Analysis Basis Delay Before Fuel 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months Consistent with the WBN, Unit 1 Movement UFSAR Section 15.5.6 Consistent with proposed TS 3.9.10 Offsite and 3.33E-4 m 3/sec This is an exception.to Control Room for all time periods Regulatory Positions C.4.1.3 and Breathing Rate C.4.2.6 of RG 1.183. The breathing rate is embedded into the DCFs used for CEDE.

Dose Conversion Tables 5-1 (Offsite) and This is an exception to Factors 5-4 (Control Room) of Regulatory Positions C.4.1.2 and EPA 400-R-92-001 C.4.1.4 of RG 1.183 Damage to Fuel 264 fuel rods (all fuel rods in Consistent with WBN, Unit 1 Assembly one fuel assembly) are UFSAR Section 15.5.6 damaged Conforms to Regulatory Position 1.1 of Appendix B to RG 1.183 Fission Product 1-131 = 0.08 Conforms to Regulatory Position Gap Fractions Kr-85 = 0.10 C.3.2 of RG 1.183, Table 3 of RG Other noble gases and iodines 1.183, and Regulatory Position

= 0.05 1.2 of Appendix B to RG 1.183 Release Time for Instantaneous Conforms to Regulatory Position Gap Activity 1.2 of Appendix B to RG 1.183 Form of Iodine Specific form of radioiodine not While Regulatory Position 1.3 of Activity Released considered due to use of an Appendix B to RG 1.183 specifies to SFP overall DCF of 200 for iodines the chemical form of radioiodine to be released from the fuel, Regulatory Position 2 of Appendix B to RG 1.183 defines an overall DCF that should be utilized.

Conforms to Regulatory Position 2 of Appendix B to RG 1.183 Water Level Minimum water level of 23 feet Consistent with WBN, Unit 1 TS Above Damaged is maintained above the 3.7.13 Fuel damaged fuel for the duration of the event Conforms to Regulatory Position 2 of Appendix B to RG 1.183 El-23

Table 6 (Page 2 of 2)

Parameters Utilized in the Analysis of the FHA Parameter AST Analysis Basis Decontamination Overall Iodine - 200 Conforms to Regulatory Positions Factor in SFP Noble Gases - 1 2 and 3 of Appendix B to RG Particulates - Infinite 1.183 Release Time for All radioisotopes are linearly This conforms to Regulatory Radioisotopes released to the environment Positions 4.1 and 5.3 of Appendix within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months B to RG 1.183 Filter Efficiencies No filtration by RBPVS or Conservative assumption in RBPVS and ABGTS assumed ABGST Conforms to Regulatory Position 4.2 and 5.4 of Appendix B to RG 1.183 Amount of Mixing None Conforms to Regulatory Positions of Activity in 4.3 and 5.5 of Appendix B to RG Containment or 1.183 Auxiliary Building Containment The Containment is not Conforms to Regulatory Positions Isolation isolated during fuel movement. 5.1, 5.2 and 5.3 of Appendix B to The radiological consequences RG 1.183 associated with an FHA in containment were determined to be bounded by an FHA in the Auxiliary Building 3.5.2 Radiological Consequences The methodology utilized by TVA to calculate the offsite and control room doses, including the exceptions to RG 1.183 noted above, is conservative (higher dose) when compared to using the RG 1.183 methodology. For the offsite doses, this is due to including the DCFs associated with ground deposition. For the control room doses, this is due to using the point kernal integration methodology as opposed to use of the DDE DCFs.

The radiological consequences are shown in Tables 7 and 8. The results for Control Room, EAB, and LPZ doses are within the appropriate acceptance criteria of 10 CFR 50.67(b)(2) and Table 6 of RG 1.183.

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Table 7 Radiological Consequences for FHA in the Auxiliary Building Conventional TPC Once TPC Twice TPC Thrice TEDE Core (rem) Burned (rem) Burned Burned (rem) Limit (rem)

(rem)

Control 1.015E+00 2.869E+00 2.602E+00 1.1 36E+00 5 Room EAB 2.383E+00 2.834E+00 2.268E+00 2.650E+00 6.3 LPZ (0a 6.660E-01 7.923E-01 6.339E-01 7.407E-01 (30-day) 6.3 Table 8 Radiological Consequences for FHA in Containment Conventional TPC Once TPC Twice TPC Thrice TEDE Core (rem) Burned (rem) Burned Burned (rem) Limit (rem)

(rem)

Control 1.OOOE+00 2.277E+00 2.014E+00 1.119E+00 5 Room EAB 2.383E+00 2.834E+00 2.268E+00 2.650E+00 6.3 LPZ (0a 6.660E-01 7.923E-01 6.339E-01 7.407E-01 (30-day) ,11 6.3

4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory Requirements The proposed changes have been evaluated to determine whether applicable regulations and requirements continue to be met.

10 CFR 50, Appendix A, General Design Criterion (GDC) 19, "Control Room," states:

"...Adequate radiation protection shall be provided to permit access and occupancy of the control room under accident conditions without personnel receiving radiation exposures in excess of 5 rem whole body, or its equivalent to any part of the body, for the duration of the accident..."

The original WBN, Unit 1 licensing basis was established based on the whole body, thyroid, and skin dose limits of 10 CFR 100.11 as described in the WBN, Unit 1 UFSAR.

As stated in RG 1.183, the applicable acceptance criterion to establish compliance with GDC 19 for facilities licensed with an AST is the 5 rem TEDE criterion of 10 CFR 50.67(b)(2)(iii). The new FHA analysis demonstrates that the WBN, Unit 1 Control Room dose complies with this 5 rem TEDE requirement.

E1-25

4.2 Precedent TVA evaluated precedent license amendment requests in which the NRC had approved implementation of the AST methodology to address FHAs. TVA identified the following precedents that were applicable, in part, to the changes TVA is proposing in this license amendment request:

Arkansas Nuclear One, Unit No. 2 - Issuance of Amendment Re: Use of Alternate Source Term (TAC No. ME3678), dated April 26, 2011

Peach Bottom Atomic Power Station, Units 2 and 3 - Issuance of Amendments Re: Application of Alternative Source Term Methodology (TAC Nos. MD6806 and MD6807), dated September 5, 2008

" Nine Mile Point Nuclear Station, Unit No. 2 - Issuance of Amendment Re:

Implementation of Alternative Radiological Source Term (TAC No. MD5758),

dated May 29, 2008 The approach used in the above license amendments regarding the application of the AST methodology to analyze the dose consequences of the FHA is similar to the proposed change to the WBN, Unit 1 licensing basis to selectively implement the AST methodology to analyze the dose consequences associated with FHAs. Sections 3.3, 3.4, and 3.5 of this enclosure describe WBN, Unit l's compliance with the guidance of RG 1.183, including a description of any exceptions taken.

4.3 Significant Hazards Consideration TVA is proposing to:

1. Permit selective implementation of the Alternate Source Term (AST) methodology in accordance with 10 CFR 50.67 and Regulatory Guide (RG) 1.183, "Alternative Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors." Implementation of the AST methodology will be limited to the analysis of Fuel Handling Accidents (FHAs) for WBN, Unit 1;

2. Add WBN, Unit 1 Technical Specification (TS) 3.9.10 to restrict movement of irradiated fuel assemblies until 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months after the reactor core has become sub-critical. TS 3.9.10 ensures that the irradiated fuel meets the minimum decay time established in the radiological analysis of the FHA;

3. Modify WBN, Unit 1 TS 3.3.6, "Containment Vent Isolation Instrumentation," TS 3.3.8, "Auxiliary Building Gas Treatment System (ABGTS) Actuation Instrumentation," and TS 3.7.12, "Auxiliary Building Gas Treatment System (ABGTS)," to eliminate the requirements associated with movement of irradiated fuel assemblies in the containment or the fuel handling area;

4. Eliminate TS 3.9.4, "Containment Penetrations," and TS 3.9.8, "Reactor Building Purge Air Cleanup Units;" and

5. Modify WBN, Unit 1 TS 5.7.2.20 to incorporate the Control Room dose limit defined in 10 CFR 50.67(b)(2)(iii).

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TVA concludes that these changes do not involve a significant hazards consideration.

TVA's conclusion is based on its evaluation in accordance with 10 CFR 50.91 (a)(1) of the three standards set forth in 10 CFR 50.92, "Issuance of amendment," as discussed below:

1. Does the proposed amendment involve a significant increase in the probabilityor consequence of an accident previously evaluated?

Response: No.

The equipment affected by the proposed changes is mitigative in nature, and relied upon after an accident has been initiated. Application of the AST does not involve any physical changes to the plant design. While the operation of various systems will change as a result of these proposed changes, these systems are not accident initiators. Application of the AST is not an initiator of a design basis accident. The proposed changes to the TS, while they revise certain performance requirements, do not involve any physical modifications to the plant. As a result, the proposed changes do not affect any of the parameters or conditions that could contribute to the initiation of any accidents. As such, removal of operability requirements during the specified conditions will not significantly increase the probability of occurrence for an accident previously analyzed. Since design basis accident initiators are not being altered by adoption of the AST analysis of the FHA, the probability of an accident previously evaluated is not affected.

The dose consequences of a FHA have been re-evaluated utilizing the AST methodology recognized by 10 CFR 50.67 and the guidance contained within Regulatory Guide 1.183. Based upon the results of this analysis, TVA has demonstrated that, with the requested changes, the dose consequences of the FHA are within the appropriate acceptance criteria of 10 CFR 50.67(b)(2) and Table 6 of RG 1.183. The AST involves quantities, isotopic composition, chemical and physical characteristics, and release timing of radioactive material for use as inputs to the dose analysis of the FHA. Selective implementation of the AST does not create any conditions that could significantly increase the consequences of any of the events being evaluated.

Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. Does the proposed change create the possibility of a new or different kind of accidentfrom any accidentpreviously evaluated?

Response: No.

The proposed changes would not require any new or different accidents to be postulated, since no changes are being made to the plant that would introduce any new accident causal mechanisms. This license amendment request does not impact any plant systems that are potential accident initiators. The AST methodology involves quantities, isotopic composition, chemical and physical characteristics, and release timing of radioactive material for use as inputs to the dose analysis of the FHA.

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

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

Response: No.

TVA is proposing to modify the methodology for responding to a FHA. Selective implementation of the AST methodology is relevant only to the calculated dose consequences for the FHA. The radiological analysis of the FHA does not credit containment isolation, operation of the Auxiliary Building Gas Treatment System, or operation of the Reactor Building Purge Air Cleanup Units. The results of the revised dose consequences analysis demonstrate that the regulatory acceptance criteria regarding onsite and offsite doses are met for the FHA.

In addition, the selective implementation of the AST methodology does not affect the transient behavior of non-radiological parameters (e.g., RCS pressure, Containment pressure) that are pertinent to a margin of safety.

Therefore, the proposed change does not involve a significant reduction in the margin of safety.

Based on the above, TVA concludes that the proposed amendments do not involve a significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of "no significant hazards consideration" is justified.

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

5.0 ENVIRONMENTAL CONSIDERATION

TVA determined that the proposed amendment would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR Part 20, or would change an inspection or surveillance requirement.

However, the proposed amendment does not involve (i) a significant hazards consideration, (ii)a significant change in the types or a significant increase in the amounts of any effluents that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental assessment needs to be prepared in connection with the proposed amendment.

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ATTACHMENT I PROPOSED TS CHANGES (MARK-UPS) FOR WBN, UNIT I

Containment Vent Isolation Instrumentation 3.3.6 3.3 INSTRUMENTATION 3.3.6 Containment Vent Isolation Instrumentation LCO 3.3.6 The Containment Vent Isolation instrumentation for each Function in Table 3.3.6-1 shall be OPERABLE.

4 APPLICABILITY: MODES 1, 2, 3, and T F I uurin~i movomont OT rrnaiaioa Tuol 3ccomDIIoc witnin cont3lnmont.

ACTIONS Separate Condition entry is allowed for each Function.

CONDITION REQUIRED ACTION COMPLETION TIME A. One radiation monitoring A.1 Restore the affected channel 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months channel inoperable, to OPERABLE status.

(continued)

Watts Bar-Unit 1 3.3-52 Amendment 35

Containment Vent Isolation Instrumentation 3.3.6 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME B. NOTE NOTE ----------

OnlY appli.abl, in MODE 1-, One train of automatic actuation logic 2, 3, 8F 4.. may be bypassed and Required Action B.1 may be delayed for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for Surveillance testing provided the One or more Functions with other train is OPERABLE.

one or more manual or automatic actuation trains inoperable. B.1 Enter applicable Conditions Immediately and Required Actions of OR LCO 3.6.3, "Containment Isolation Valves," for Two radiation monitoring containment purge and channels inoperable, exhaust isolation valves made inoperable by isolation OR instrumentation.

Required Action and associated Completion Time of Condition A not met.

(continued)

Watts Bar-Unit 1 3.3-53

Containment Vent Isolation Instrumentation 3.3.6 ACTIONS (continued) _________

CONDITION REQUIRED ACTION COMPLETION TIME 0?NOTE G4 Paco and Fmaintain ImmFediately Onily appliooblo during containmon~t purgo and movcmont Of irradiatod fuel exhaust valvcs in clocod asscmbliesc within peeltleR.

eentaiwneet.

Ono or moro FunctioeR with4 G- Entor applicablo Conditione IMme1doately ono or moroF manual or and Roguirod Actionc of automatic actuation trains LCO 3.9.4, "Contaminan.RRt.

nepeFeble. Pcnctrationc," for containonteP pur~ge and exhaust isolatio OR valvcs m~ado incpsrablc by iccloticnistuctain Two radiaticn mcn~itcring cha~nnls inoporabic.

Roguirod Action and ossosiated Comnplotion Time for Condition A not maet.

Watts Bar-Unit 1 3.3-54 Amendment 35

Containment Vent Isolation Instrumentation 3.3.6 Table 3.3.6-1 (page 1 of 1)

Containment Vent Isolation Instrumentation FUNCTION REQUIRED SURVEILLANCE ALLOWABLE CHANNELS REQUIREMENTS VALUE

1. Manual Initiation 2 SR 3.3.6.6 NA

2. Automatic Actuation Logic 2 trains SR 3.3.6.2 NA and Actuation Relays SR 3.3.6.3 SR 3.3.6.5

3. Containment Purge Exhaust 2 SR 3.3.6.1 _ 02-O2-ee Radiation Monitors SR 3.3.6.4 (4 SR 3.3.6.7 < 2.8E-02 pCi/cci (2.8E+04 cpm)

4. Safety Injection Refer to LCO 3.3.2, "ESFAS Instrumentation," Function 1, for all initiation functions and requirements.

(8) During m'V.'.m,,t of i,Radiated fude assammbliot w.ithin AAonbinmo~t.

(b) Medcz 1, 2, 3, and 1.

Watts Bar-Unit 1 3.3-56 Amendment 74

ABGTS Actuation Instrumentation 3.3.8 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME B. (continued) B.2 Place both trains in Immediately emergency radiation protection mode.

G- Roquirod Action and G.4 SUcp98nd moevomont of IeAFediete4 acccicatod ComplotioR Timo irr-Aditotfu o.4accomblioc in for Ccditin .A Or nB t Mct thoeo handling aroa durliong emzvmct oef irda fuel assemblies in the fuol 19he ffRg aFe&-

.# Required Action and Q*.I Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time for Condition A or B not met in AND MODE 1, 2, 3, arF Q.2 Be in MODE 5. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months SURVEILLANCE REQUIREMENTS

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

Refer to Table 3.3.8-1 to determine which SRs apply for each ABGTS Actuation Function.

SURVEILLANCE FREQUENCY SR-&&.4 Po~form CHANNEL C;HECK. 1:2-hews SR-3.& PefeF*-,G*T, 92-days (continued)

Watts Bar-Unit 1 3.3-62

ABGTS Actuation Instrumentation 3.3.8 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.3.8 --------------------- NOTE-----------------

Verification of setpoint is not required.

Perform TADOT. 18 months SR3Pcrfrm CH,,.'NEL CALIGR,"BRATION. 18 me4ths Watts Bar-Unit 1 3.3-63

ABGTS Actuation Instrumentation 3.3.8 Table 3.3.8-1 (page 1 of 1)

ABGTS Actuation Instrumentation APPLICABLE MODES OR OTHER SPECIFIED REQUIRED SURVEILLANCE ALLOWABLE CONDITIONS REQUIREMENTS VALUE FUNCTION CHANNELS

1. Manual Initiation 1,2,3,4 2 SR 3.3.8.3 NA (a) 2 SR 3.3.8.3t NA

2. .'...P..Are...R.d...... (Ra) 2 2R-_44464.

S;o 2.2.84 Deleted

3. Containment Isolation - Refer to LCO 3.3.2, Function 3.a., for all Phase A initiating functions and requirements.

uunnzi mcvcmznr OT irro~iatc~ TUZ~I oc~cmoiic~ an tnc TUCI nzanaiana arca.

Watts Bar-Unit 1 3.3-64

ABGTS 3.7.12 3.7 PLANT SYSTEMS 3.7.12 Auxiliary Building Gas Treatment System (ABGTS)

LCO 3.7.12 Two ABGTS trains shall be OPERABLE.

APPLICABILITY: MODES 1, 2, 3, and 4, DUuOrng movcment Of IrrzScliated tuol asseRo3mbli *nthe fuel handlin.E argo.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One ABGTS train A.1 Restore ABGTS train to 7 days inoperable. OPERABLE status.

B. Required Action B.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and associated Completion Time of AND Condition A not met in MODE 1, 2,-37 B.2 Be in MODE 5. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months er-4.

OR Two ABGTS trains inoperable i, MOIDE 1, 2, 3, 9r 4.

4- Reguircd Aetion and G Place OPEiRABLE ABGT-S Immediaeley asseeiated CampletieA train in cperatiefn.

Timce ef Ccniditic A not met during moevement of (R iradfiatod fuol accombioc0 in the uol h-AndlIrg ara. G4 SuScprd mo'.mArt of i,.,1ate1, irrAdi**td fuoW Acco1b1io i-n the fuel handling arna.

(continued)

Watts Bar-Unit 1 3.7-27

ABGTS 3.7.12 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME

&. Twe ASCTS9 trakin &.4 SuspeR ndmvcmonit of IFRFediet*

incepcrable during rradiatcd fuel assemblioc i Mo'.'omcnt of irradiatod fuel the fuel handingaro arcsomblioc in the fuel SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.12.1 Operate each ABGTS train for > 10 continuous hours 31 days with the heaters operating.

SR 3.7.12.2 Perform required ABGTS filter testing in accordance In accordance with with the Ventilation Filter Testing Program (VFTP). the VFTP SR 3.7.12.3 Verify each ABGTS train actuates on an 18 months actual or simulated actuation signal.

SR 3.7.12.4 Verify one ABGTS train can maintain a pressure 18 months on a between -0.25 and -0.5 inches water gauge with STAGGERED TEST respect to atmospheric pressure during the post BASIS accident mode of operation at a flow rate >_9300 and

< 9900 cfm.

Watts Bar-Unit 1 3.7-28

.- oCntainmont PenUtAtion 3.9.4 nDeleted 3.9 REFUELING OPERATIONS 3.9.4 Q..t*!,nAPct P e-rtretecns LGO " D.4Teeleted afinmont penotrationS Shall be in the following statusi a- The oguipmcnt hatch cicced and held in place by a minimum of four One doorF in each air look elocoed; or eapablo of being closed providod ABGT-S is OPERABLE in accordance with T-S 3.7.12: anqd if-- -- L ..... L----L*

GrM racnpon"raio pr3oving air ct aceos tro

conainmon atmoshor toth otside atmos3hreF oithor:

1, oslosod by a manual or autom~atic isolation valvo, blind flango, Or equvient, ref capable of being c.loed by an OPERABE Containmet Ven irselatfeR System.

kIrrr!!!!

IVl V to the Futeid..atmosphere may bo unisolatod under administrativo controls provided ABGTS iGs OPERABLE in accordance with TS 3.7.122.

APPLICAB"IITY Duin.g movement of irradiatod fuel ascemblios within RARteUOReRt~

AQWI@N&

GO.DITION REQUIRED ACTION CO.MPLETIOTN TIME A- Ono or m9Fre containent A4 Suspend moVeent ,* M.M.ditely penetratfione not iiRradiated fuel assemblies Feq~iFe4-64a4Y6 within containm~ent.

Watts Bar-Unit 1 3.9-6 Amendment 26, 35

Contaimonet Pnt~in 3.9.4 nDeleted SWRVEILLANC~e REQUIREMENTS RI RVE- I A~N GE FREQU1IENICY SR-&.94 Verify each rcguiroed eentaimomnt pcnetrntion 7cdan the rcguirod sta~tus.

sR.-94~aVorify,oach roguiro~d contanmonRt Y'ont 0eolation V'aIV9 48-m 1 thS actuiatcc to the icolation poc~tionR OR an actual or simulatod actuatien signal-Watts Bar-Unit 1 3.9-7

Reast8r Building Purgo AiFr Cloanup Unit 3.9.8 E eleJe 3.9 REFUELING OPERATIONS 3.9,8 6katr#R Buildina PZUrac Air Cinanun1 Unitn S~eleted-iGc A.0~ +we ca*ct:or MuEMa.n rur;*gv fir ic-anup units snail bse.--'\-'-.

APPLICABILIT-Y: IDumrig movomant of irradiated fuel assemblioc within thc contaiFnmont.

GCODQTQO, REQUIRED ,A, CTION CO.MPLETIO.N TIME A- One RnartGr Building A4 Iclsate thiepoal air Immediately Purgc Air Cleanup Unit eeefwi RepeFable ANQ A-2 Vorib' tho OPERABLE air Mm~ediaeley cloanup unit ic in oporation.

a- TWO Reactor. Building 84 Suepond meo....ont* tf 4m1eiditely Purge AiFr Cicanup Unite irradiatd fuel acsemblies ieepeFabgle. ihncnanet Watts Bar-Unit 1 3.9-14 Amendment 35

K~atEcwr t5HIIENin Purgo :J8 MIF~fU UflIU31 tRK 3.9.8 Deleted SWRV9EILL=NGCE REQUIREMENTS 6URVEILLANC FREQUENCY SR 3.94 Pecf9Fm roguired RFilo tccting ini accordanc3 With tho In accrr.danco With Ventibation Filter Tootfing ProgrnmA (VF-TP). the VFT-P Watts Bar-Unit 1 3.9-15

Decay Time 3.9.10 3.9 REFUELING OPERATIONS 3.9.10 Decay Time LCO 3.9.10 The reactor shall be subcritical for > 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months .

APPLICABILITY: During movement of irradiated fuel assemblies within the containment.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Reactor subcritical for A.1 Suspend movement of Immediately

< 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months . irradiated fuel assemblies within containment.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.9.10.1 Verify the reactor has been subcritical for > 100 Prior to movement of hours. irradiated fuel within containment Watts Bar-Unit I 3.9-17

Procedures, Programs, and Manuals 5.7 5.7 Procedures, Programs, and Manuals 5.7.2.19 Containment Leakage Rate Testing Program (continued)

Leakage rate acceptance criteria are:

a. Containment overall leakage rate acceptance criterion is < 1.0 La. During the first unit startup following testing in accordance with this program, the leakage rate acceptance criteria are < 0.60 La for the combined Type B and Type C tests, and

< 0.75 La for Type A tests.

b. Air lock testing acceptance criteria are:

1) Overall air lock leakage rate is < 0.05 La when tested at > Pa.

2) For each door, leakage rate is < 0.01 La when pressurized to _>6 psig.

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

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

5.7.2.20 Control Room Envelope Habitability Program A Control Room Envelope (CRE) Habitability Program shall be established and implemented to ensure that CRE habitability is maintained such that, with an OPERABLE Control Room Emergency Ventilation System (CREVS), CRE occupants can control the reactor safely under normal conditions and maintain it in a safe condition following a radiological event, hazardous chemical release, or a smoke challenge. The program shall ensure that adequate radiation protection is provided to permit access and occupancy of the CRE under design basis accident (DBA) conditions without personnel receiving radiation exposures in excess of . . .. b. . it . ........... W ef the 199dy f9r the d'-rati9R, 9f,the The*,program shall include the following elements: Il~nsert 5.7.2.20-1]

a. The definition of the CRE and the CRE boundary.

b. Requirements for maintaining the CRE boundary in its design condition including configuration control and preventive maintenance.

c. Requirements for (i) determining the unfiltered air inleakage past the CRE boundary into the CRE in accordance with the testing methods and at the Frequencies specified in Sections C. 1 and C.2 of Regulatory Guide 1.197, "Demonstrating Control Room Envelope Integrity at Nuclear Power Reactors,"

Revision 0, May 2003, and (ii) assessing CRE habitability at the Frequencies specified in Sections C.1 and C.2 of Regulatory Guide 1.197, Revision 0.

(continued)

Wafts Bar-Unit 1 5.0-25 Amendment .5, 70, 78

Insert 5.7.2.20-1 the applicable regulatory requirement (i.e., 5 rem Total Effective Dose Equivalent (TEDE) for a fuel handling accident or 5 rem whole body or its equivalent to any part of the body for other accidents) for the duration of the accident.

ATTACHMENT 2 PROPOSED TS BASES CHANGES (MARK-UPS) FOR WBN, UNIT I

Containment Vent Isolation Instrumentation B 3.3.6 B 3.3 INSTRUMENTATION B 3.3.6 Containment Vent Isolation Instrumentation BASES BACKGROUND Containment Vent Isolation (CVI) Instrumentation closes the containment isolation valves in the Containment Purge System. This action isolates the containment atmosphere from the environment to minimize releases of radioactivity in the event of an accident. The Reactor Building Purge System may be in use during reactor operation and with the reactor shutdown.

Containment vent isolation is initiated by a safety injection (SI) signal or by manual actuation. The Bases for LCO 3.3.2, "Engineered Safety Feature Actuation System (ESFAS) Instrumentation," discuss initiation of SI signals.

Redundant and independent gaseous radioactivity monitors measure the radioactivity levels of the containment purge exhaust, each of which will initiate its associated train of automatic Containment Vent Isolation upon detection of high gaseous radioactivity.

The Reactor Building Purge System has inner and outer containment isolation valves in its supply and exhaust ducts. This system is described in the Bases for LCO 3.6.3, "Containment Isolation Valves."

The plaRt dcsign bazSi .... i. r, that when mo.ing irradiated fuel in the Amu liar; Building and r- C-nRt;;ainnt with1the oRtainm\nt open t the ; A xii*aF' Buildi*g ABG!! pae, ignal ulpe adaiAF9it8rc monteset0 RE 00 102 and 103 will initiato a Containment Vontilation Iselation (CVI) in additicn to thoir nriFmalW funcRtion. In addition, a cignal fromA the containmont purgo radiation moiterc I RE 00 130, and 13 1 or othcr CVI cignal ilintat that poo of the ABI nrmFally initiated by the Gpont fuol pol aiain oitore.

(continued)

Wafts Bar-Unit 1 B 3.3-154 Revision 43, 87, 110

Containment Vent Isolation Instrumentation B 3.3.6 BASES (continued)

APPLICABLE The containment isolation valves for the Reactor Building Purge System SAFETY ANALYSES close within six seconds following the DBA. The containment vent isolation radiation monitors act as backup to the SI signal to ensure closing of the purge air system supply and exhaust valves. Thcy aro also thc primary moans for n*

ehutdewR. Containment isolation in turn ensures meeting the containment leakage rate assumptions of the safety analyses, and ensures that the calculated accidental offsite radiological doses are below 10 CFR 100 (Ref. 1) limits.

The Containment Vent Isolation instrumentation satisfies Criterion 3 of the NRC Policy Statement.

The ABCT is Fr*cguied to be operablc d,.uring .. ocment of iradiatod fuel in the ilinr. B uilding any

.. . and ; .. - ... of irra. intrA fue in th s

^,

Reator Bu;ilding when the Reao-cr Building is established as po of the ABSCGE boundary (coo TS 3.3.8, 3.7.12, & 3.9.4). ~hon m~eovin irraqdiated fuoel fins~ide contafinment, at ceast onc train of thc containment purgc s'ystem must b3e opcralting crthe containment must bo isolated. WhoR nemoing irradiatcd fuol in thc Auxiliary Building durIing times whon thc containment is opcni to thc Auxiliary Building ABSCE! spasos, conta*minat purge raR boeoperated, bu6t oporation of the s'ystcmi s5not roguired.

Asstated previously, provisions arc providod to neco n c tho ABI R* and the CVI iRnstru* entation to fac..ilitate tho astuationf.. fa. n. A BI* whc C-a ,usinitiatod and tho astuation of a CVPwhen an ABI* is initiatod. Tho following table s*pocifies the radiation m onitors1 and the ass ociated acetation i*n*stnr m RtatiR that m~ust be availablo whon thc ABI* and CVI functions are intereonnlectod anid fuel-.

is being moevod either insido containmenRt orin tho Auxiliary Building orif fuol uelM dation eReoqemt aeid d A:R ctuntainmten tC Containment i Rmnatru 4 . m -m o nt ti o n:H atc h. . evn e tr ti. o; ns ;

Gle sedi C ontainmento Epont o r nr c o w8 e.o. iWith A u xilia ry B u "ild i ng Rn n side C on t a i nm e nt@

'-RE-90 34a 3.3.c cc a eWoith uxiliary B uildin E 0 0 2O 3 .3 ..6 . ..... ... c..c .r n i.th Q RE 90 41GO3.383 G.7. . 3.8a nd 12I....

I R -m _ 0 13Ina ccordancew With CI=

In sideC ontai m~ ent E Q O- 434 LL 0G0 .3 ._ _ _ _ _ _ _ _ _ _ _ 3 -"__

and inAux"ie2 a accor9122

)dancew WithA SC6E m aintainedin builaing 0 Rn0 0 =GQ 3.9.4a ccordancewA i.thL CO_-

_ _ __ __ _1_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ 3. 3.8 a n d 3 .7.12*

3__

That~ sdioncotthe ABIi nitiated b yt he~ Sp ntFuelP ooelKadiatienM oitorsW,U 0 RE UQ 402a nd;14

-RE-9910-3(

(continued)

Watts Bar-Unit 1 B 3.3-154A Revision 43, 87, 110

Containment Vent Isolation Instrumentation B 3.3.6 BASES LCO 3. Containment Radiation (continued)

The LCO specifies two required channels of radiation monitors to ensure that the radiation monitoring instrumentation necessary to initiate Containment Vent Isolation remains OPERABLE.

For sampling systems, channel OPERABILITY involves more than OPERABILITY of the channel electronics. OPERABILITY may also require correct valve lineups and sample pump operation, as well as detector OPERABILITY, if these supporting features are necessary for trip to occur under the conditions assumed by the safety analyses.

Table 3.3.6-1 specifies the-twe Allowable Value& (AV&) for the Containment Purge Exhaust Radiation Monitors. One AV ic applic~abl in IThis AV is ODES 1,2,3,and 4 and theoccnd AV is applicable during the movo-mont of irradiated f'ul accsomblieG"Oido contrainment Whon the potential f9r a fuol handling accidont (PHA) oxiste. Both of these AVe aro based on expected concentrations for a small break LOCA, which is more restrictive than the 10 CFR 100 limits. In addition, the courco term far -An FHA, ic cignificantly groatcr thaRn the courco torm for a ema*l brcak The specified . ocA which w'uld rcu'-t in the 3ontainmont puFrgo moneitrc rosponding more conservative than the analytical limit assumed in the safety analysis in order to account for instrument uncertainties appropriate to the trip function. The actual nominal Trip Setpoint is normally still more conservative than that required by the AVs. If the setpoint does not exceed the applicable AV, the radiation monitor is considered OPERABLE.

4. Safety Iniection (SI)

Refer to LCO 3.3.2, Function 1, for all initiating Functions and requirements.

APPLICABILITY The Manual Initiation, Automatic Actuation Logic and Actuation Relays, Safety Injection, and Containment Radiation Functions are required OPERABLE in MODES 1, 2, 3, and 4- and during movomont of irrdiatod fu-el acomblio -ithin ccnta!nmeRt. Under these conditions, the potential exists for an accident that could release significant fission product radioactivity into containment.

Therefore, the Containment Vent Isolation Instrumentation must be OPERABLE in these MODES. See additional discussion in the Background and Applicable Safety Analysis sections.

(continued)

Watts Bar-Unit 1 B 3.3-156 Revision 45, 87, 93 Amendment 35, 74

Containment Vent Isolation Instrumentation B 3.3.6 BASES APPLICABILITY While in MODES 5 and 6 without fuel haRdling in progrcc,, the Containment (continued) the Containmont Vent Isolation Instrumentation need not be OPERABLE since the potential for radioactive releases is minimized and operator action is sufficient to ensure post accident offsite doses are maintained within the limits of Reference 1.

ACTIONS The most common cause of channel inoperability is outright failure or drift sufficient to exceed the tolerance allowed by unit specific calibration procedures.

Typically, the drift is found to be small and results in a delay of actuation rather than a total loss of function. If the Trip Setpoint is less conservative than the tolerance specified by the calibration procedure, the channel must be declared inoperable immediately and the appropriate Condition entered.

A Note has been added to the ACTIONS to clarify the application of Completion Time rules. The Conditions of this Specification may be entered independently for each Function listed in Table 3.3.6-1. The Completion Time(s) of the inoperable channel(s)/train(s) of a Function will be tracked separately for each Function starting from the time the Condition was entered for that Function.

A. 1 Condition A applies to the failure of one containment purge isolation radiation monitor channel. Since the two containment radiation monitors are both gaseous detectors, failure of a single channel may result in loss of the redundancy.

Consequently, the failed channel must be restored to OPERABLE status. The 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months allowed to restore the affected channel is justified by the low likelihood of events occurring during this interval, and recognition that one or more of the remaining channels will respond to most events.

(continued)

Watts Bar-Unit 1 B 3.3-157

Containment Vent Isolation Instrumentation B 3.3.6 BASES ACTIONS B.1 (continued)

Condition B applies to all Containment Vent Isolation Functions and addresses the train orientation of the Solid State Protection System (SSPS) and the master and slave relays for these Functions. It also addresses the failure of multiple radiation monitoring channels, or the inability to restore a single failed channel to OPERABLE status in the time allowed for Required Action A.1.

If a train is inoperable, multiple channels are inoperable, or the Required Action and associated Completion Time of Condition A are not met, operation may continue as long as the Required Action for the applicable Conditions of LCO 3.6.3 is met for each valve made inoperable by failure of isolation instrumentation. A Note has been added above the Required Actions to allow one train of actuation logic to be placed in bypass and to delay entering the Required Actions for up to four hours to perform surveillance testing provided the other train is OPERABLE. The 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months allowance is consistent with the Required Actions for actuation logic trains in LCO 3.3.2, "Engineered Safety Features Actuation System Instrumentation" and allows periodic testing to be conducted while at power without causing an actual actuation. The delay for entering the Required Actions relieves the administrative burden of entering the Required Actions for isolation valves inoperable solely due to the performance of surveillance testing on the actuation logic and is acceptable based on the OPERABILITY of the opposite train.

Asaduditin B ORM , )opplicablo (continued)

Watts Bar-Unit 1 B 3.3-158

Containment Vent Isolation Instrumentation B 3.3.6 BASES ACTIONS (continued)

Condition C applies to all Containmon~t Vcnt lzolation Func~tionc and addrocecss the train orfiontation o~f the SS;RPS- -Andtho- mactor and clavo rclayc for thoco Functionc. It alco addroccoc the failur of multipic radiation moni~toring channoic, or the inabilit' to roctoro a cinglo failod chRnnol to OPERARBL-E statu-c in tho timo all.w.d fcr R1g..rcd Action A.I. If R tra;in* is inpoablo, Multiplo chanMeol aro dinopcrable, or the Reguirod Action and accociatod Complotion TimoF ot Condition A aro not mcet, cperatien m~ay rentin ue as long ac the ouio Actio to plare and maintafin containmon~t pugoe an~d oxq;hauc'Ft ioelation Yalyoc in thoem clocad pecition is met or the applicablo ConRditionc%of LCOG 3.9.4, "Containmon~t Pznctrationz,". arc mcet for oach Yalyc mnado inopcrable by failurc of isolation Dnctrumzntatien. The mpl.tioR

.. Time far thcse Ro.uircd A cqt-nir Immediately.

A Ntestaesthat Cedte is. appilcaule dW*0mym~ f Fa.tdfe asscmbliac Within eentainmcnt.

determines A Note has been added to the SR Table to clarify that Table 3.3.6-1 SURVEILLANCE A Note has been added to the SR Table to clarify that Table 3.3.6-1 determines REQUIREMENTS which SRs apply to which Containment Vent Isolation Functions.

SR 3.3.6.1 Performance of the CHANNEL CHECK once every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ensures that a gross failure of instrumentation has not occurred. A CHANNEL CHECK is normally a comparison of the parameter indicated on one channel to a similar parameter on other channels. It is based on the assumption that instrument channels monitoring the same parameter should read approximately the same value.

Significant deviations between the two instrument channels could be an indication of excessive instrument drift in one of the channels or of something even more serious. A CHANNEL CHECK will detect gross channel failure; thus, it is key to verifying the instrumentation continues to operate properly between each CHANNEL CALIBRATION.

(continued)

Watts Bar-Unit 1 B 3.3-159 Revision 45 Amendment 35

ABGTS Actuation Instrumentation B 3.3.8 B 3.3 INSTRUMENTATION B 3.3.8 Auxiliary Building Gas Treatment (ABGTS) Actuation Instrumentation BASES BACKGROUND The ABGTS ensures that radioactive materials in the fuel building atmosphere following 8 f'-ol handling accid,,t or a loss of coolant accident (LOCA) are filtered and adsorbed prior to exhausting to the environment. The system is described in the Bases for LCO 3.7.12, "Auxiliary Building Gas Treatment System." The system initiates filtered exhaust of air from the fuel handling area, ECCS pump rooms, and penetration rooms automatically following receipt of-&

fuol pool aroa high radiatieRn* gnal or a Containment Phase A Isolation signal.

Initiation may also be performed manually as needed from the main control room.

E 0fiaitetR,. Ench .A.BGTS9 traiR si",t3o by, ,,=hg OrNAtR detected by a GhARRI dcdmcated t8 t~hat tr3aR. There are a total wo channels, one for each train.

High, rFadi..!c* deteetd by a,' me*!ter ort Phase A isolation signal from the Engineered Safety Features Actuation System (ESFAS) initiates auxiliary building isolation and starts the ABGTS. These actions function to prevent exfiltration of contaminated air by initiating filtered ventilation, which imposes a negative pressure on the Auxiliary Building Secondary Containment Enclosure (ABSCE).

The plant design ba r ,ic u .. that

.. when

.. mov .ig .. .atod fu in the

.l Auxilia '

Building andior ContaiHnmcnt;wfith the ConRtaiwnmot and/or annulu6 opcn to the Auxilia. ' ,*, ,;, ^, GF!! a,- a G l from +ho cpont fuel poeol radiatio monimterc 0 RE! 00 102 and 103 YVll initiat a Containmont Vontilation Icolation (CVI) in addition te their nrmalR- funcRAtion. In addition, acignal 4fro thO containmnt purgo radiation m.nqitrc ! RE g0 130, and 131 or ether CVI cignal w.,ill initiato that po*ti*o ef the ABIrnormally initiat*d by the ept fuel pool4 radiation. m tR ....

(continued)

Watts Bar-Unit 1 B 3.3-171 Revision 87, 110

ABGTS Actuation Instrumentation B 3.3.8 BASES (continued)

APPLICABLE The ABGTS ensures that radioactive materials in the ABSCE atmosphere SAFETY ANALYSES following a fuol' handling accidont or a LOCA are filtered and adsorbed prior to being exhausted to the environment. This action reduces the radioactive content in the auxiliary building exhaust following a LOCA or fuol handling accidot so that offsite doses remain within the limits specified in 10 CFR 100 (Ref. 1).

The ABGTS Actuation Instrumentation satisfies Criterion 3 of the NRC Policy Statement.

Thc ABOTSg it roguircd to bo oporablo duFrin mo9Vomont Of irradiatod fuol i tho A nxilinr' Building dligany mnedn nn d mn'nmnt irninn fini thn R.a.tor Building. when tho Rcactor Building it establishcd as part of the ABSCE bcundar; (see T-8 3.3.8, 3.7.12, & 3.9.4). When moving irradiatcd fuci insido containmcnt, at ceast ono train of thc contaminFant purgc systcm musttb opcrating or the containmcnt m~ust bo icolatod. Whcn movying irradiatcd fuel in tho Aux..ilia Building duFrig times whon .potho co.taimon..t ito to theaAuxili.

Building ABSCE! cpacat, contaiminat purgo can bo8 oporatod, but oporation of the system it not rcguircd.

As stated frc"'-i-usly, nrovic, ted to inteFG9nncct the oBI* and tho CVI itrinttio t facilitato thc actuation of an AB3I* Whon a CVI i ntao anId thoeet*uation of a CVI whcni an ABI* it itat. Tho following table tpecifiet+tho ridi*atioei*no nitort and.. . tho *attociatod actuation) inr m that mut-Ft ho availablo whon tho AB'* an~dCV funcRAtiontR aroinRtorconnoc~tod and fuo'A it boing Fnovod either intidc containm~ent Or in the Auxiliary Building or if fuel movomon_:1_.t *'it takino DMIaco in both aroac:-

Reuire - __rd rb2an Ot~RRn Fuel Movement: Radiation IRequired Actuation' CoMaitc-R Cnatratinment Incio CctAinmont I RF Q42AL .. accord~anco withacodnowt

.... ...... . -0RE 90-102 4QQC 3.3. AB..CE maintainod i Building AuRiliary G~peR-eF-CGIGS accordacae with LCOQ 3.3.8 and 3.7.12 kI accordance with LCO R1 E - 9 0 -90

_- _4-01RE a -49 O

amnd-OR-Auwoe~ 2Q 2A;rordancoacc9)i with ABSCE mnaintOaGindi BU11Idig LGQ-3.94 accorFdancoR With LCGO

_________U _ 1_______RE 90 103 __________

2___3.23.98

_and 2_7412 ii__:¢___ r* i'-*ff /*/* Ar*/-I =_J That *orOT tnc

.. I lnn.iat.. By the tcRn, B ,-uci .-'ooi ,iaoiafuonR vonitort;, U .r. tl ... ano 0 RP 990-49.

(continued)

Watts Bar-Unit 1 B 3.3-172 Revision 87, 110

ABGTS Actuation Instrumentation B 3.3.8 BASES (continued)

LCO The LCO requirements ensure that instrumentation necessary to initiate the ABGTS is OPERABLE.

1. Manual Initiation The LCO requires two channels OPERABLE. The operator can initiate the ABGTS at any time by using either of two switches in the control room. This action will cause actuation of all components in the same manner as any of the automatic actuation signals.

The LCO for Manual Initiation ensures the proper amount of redundancy is maintained in the manual actuation circuitry to ensure the operator has manual initiation capability.

Each channel consists of one hand switch and the interconnecting wiring to the actuation logic relays.

2. Fuwel Pool Aroa Radiation

[Deleted! Thc LCO spocifioc two requirod Fuoel Pool Aroa Radiation Monitore to onsurs that the radiation moni.to;*rg intrumon.. tation nccacry to initiate the ABOTSg remains OPERABLE. Ono radiation moniRtor is deodiatodt sash train of ABOTS.

For sampling systems, hannel OPERABILITY involves mor.e thar OP12ER.ABILITY of channel olecteroics. OPERABILITY m~a" aleo roquir corret valve ,inoups, samplc pump oporation, and filter mot9or oparltion* ,

as well as doteGeOt OPERABILITY, if thoco supporting fpaturoc are ncccssa' for trip to occur under tho conditions assumod by the eafety Only the Allowable Value is speeifiod forF the Fuoel Pool Aroa Radliation, Monitors in the LCO. Tho Allowable Value specfified it moree consor.ative than thea~aý*a liissumeod in tho safoty analysis inR order to accoun..t for instrument uerta.intie app ropiate to the trip fuction. The actul nromirnal Trip Sotpoint is no.rmally still ,mrce o vthan that reoquFd by the Allwab Value. If themos Sotpoint doos not exceed the Alleowa-ble Value, tho radiation- monietori conIsiod O-PERABL6E.

(continued)

Watts Bar-Unit 1 B 3.3-173

ABGTS Actuation Instrumentation B 3.3.8 BASES LCO 3. Containment Phase A Isolation (continued)

Refer to LCO 3.3.2, Function 3.a, for all initiating Functions and requirements.

APPLICABILITY The manual ABGTS initiation must be OPERABLE in MODES 1, 2, 3, and 4-apd

,.,h mo.iRg irradiatod fuel a,..mblioc in the fuel haRd*in* aroa, to ensure the ABGTS operates to remove fission products associated with leakage after a LOCA or a fuol handling accidont. The Phase A ABGTS Actuation is also required in MODES 1, 2, 3, and 4 to remove fission products caused by post LOCA Emergency Core Cooling Systems leakage.

High radiaticn initiation of the ABCTS m~uct be OPERABLE9 inany MODS during moevement cf irraldiated fuel assemblios in the fuel handling arca te encuro automatic initiation ef the ABGTS when the petcntial far aafuel handling ascidont While in MODES 5 and 6 without fuol haRdling iR progr;ce, the ABGTS instrumentation need not be OPERABLE cinO: a fuA-W handlRig acdidont cGaRot eeeH. See additional discussion in the Background and Applicable Safety Analysis sections.

ACTIONS The most common cause of channel inoperability is outright failure or drift sufficient to exceed the tolerance allowed by unit specific calibration procedures.

Typically, the drift is found to be small and results in a delay of actuation rather than a total loss of function. If the Trip Setpoint is less conservative than the tolerance specified by the calibration procedure, the channel must be declared inoperable immediately and the appropriate Condition entered.

A Note has been added to the ACTIONS to clarify the application of Completion Time rules. The Conditions of this Specification may be entered independently for each Function listed in Table 3.3.8-1 in the accompanying LCO. The Completion Time(s) of the inoperable channel(s)/train(s) of a Function will be tracked separately for each Function starting from the time the Condition was entered for that Function.

(continued)

Watts Bar-Unit 1 B 3.3-174 Revision 87

ABGTS Actuation Instrumentation B 3.3.8 BASES ACTIONS A.1 (continued)

Condition A applies to the actuation logic train function from the Phase A Isolation, the radiaticn mc*nitcr f*u*...", and the manual function. Condition A applies to the failure of a single actuation logic train, radiation monitor cha.nol, or manual channel. If one channel or train is inoperable, a period of 7 days is allowed to restore it to OPERABLE status. If the train cannot be restored to OPERABLE status, one ABGTS train must be placed in operation. This accomplishes the actuation instrumentation function and places the unit in a conservative mode of operation. The 7 day Completion Time is the same as is allowed if one train of the mechanical portion of the system is inoperable. The basis for this time is the same as that provided in LCO 3.7.12.

B.1.1, B.1.2, B.2 Condition B applies to the failure of two ABGTS actuation logic signals from the Phase A Isolation, tMo radiation ,,monitrc, or two manual channels. The Required Action is to place one ABGTS train in operation immediately. This accomplishes the actuation instrumentation function that may have been lost and places the unit in a conservative mode of operation. The applicable Conditions and Required Actions of LCO 3.7.12 must also be entered for the ABGTS train made inoperable by the inoperable actuation instrumentation. This ensures appropriate limits are placed on train inoperability as discussed in the Bases for LCO 3.7.12.

Alternatively, both trains may be placed in the emergency radiation protection mode. This ensures the ABGTS Function is performed even in the presence of a single failure.

Conditfioni G applies when the Rcgu~irod Action and ascociatod Complotien Timo fer Condition .Aor BR hWVo Het boon met and- irrad-ia*Atod fuel a66omblioc aro boing mROVod- in tho-ful buildfina. Movomon~t of irradiAtod fulASccomb~lioc inRh (continued)

Watts Bar-Unit 1 B 3.3-175

ABGTS Actuation Instrumentation B 3.3.8 BASES ACTIONS ee GOtR4+e4)

!We' Obuiln v bo SUG198ndod wmmedmately to olwminato the petontmal tor ovontc_

muR9cR1t that could roguiro ABGT-9 actuation. PorlormFanco; o-f th8co actionc chall not c prccludo movinrg a compononrt to a cafo pocition".

.1land Condition : plies when the Required Action and associated Completion Time for Condition A or B have not been met and the plant is in MODE 1, 2, 3, or 4.

The plant must be brought to a MODE in which the LCO requirements are not applicable. To achieve this status, the plant must be brought to MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and MODE 5 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

SURVEILLANCE A Note has been added to the SR Table to clarify that Table 3.3.8-1 determines REQUIREMENTS which SRs apply to which ABGTS Actuation Functions.

&R-g4-3.

Prefermanco of the CHANNEL CH4ECK onco 8OvY 12 hourGeoncuroc thata grooc failuro of inctrumon~tation hac not occurrod. A CHANNEL1 CH-ECK ic n.rm.ally a compac..n of..tc aramotr nd*icated on onotchanrn, to a cimilar paramctor on othor Ghannolc. it is based on the accumFption that incRumonR:(t channolc mon~itoring the Game paraqmotor chould road approximately the comoe Yaluoe. Significant doviationc boetwoon the tAo incRumon Ahannels coulId he An indiccatin of xcocc'ivo drif in ono of the channolc or of Fcmothing 4nctr'-mnt ovon ,morc sc'iouc. A CHANNEL CHECK will dctcct grocs channol failuro; thus, it ic koy toe Y.ifying the inctrumntatien cOntinucc to oporato propcrly botwoon Pr~ih CHA!~NWR c2AJIRRATION' (continued)

Watts Bar-Unit 1 B 3.3-176

ABGTS Actuation Instrumentation B 3.3.8 BASES SURVEILLANCE SR 3.3.8.1 (conitinucd)

REQUIREMENTS Age R..t .. . are detcFrm...d

.itcria b" the Wit.ta. based o a scombination of the channel inotrument unccrtaintioc, including indirsation and roadability. if a ehanncl iseutside the criteria, it may; be an indication that tho concrRF or the s.ignal Prccocc in ogip t has drifted outside itc limit.

The Froquenoy ic based on operating cXper~inec that domOnctratoc channol failurc icraro. The CHANNEL= CHEC!K supplemontc 10cc foFrmal, but moroe frgquont, chocks of channoic during normal operational usc of the dicplay accociated- with the COre-quired channoic.

A COT is peufFFuu once v~ 92, days on ~ Iu e U~Flq1 nRi~i to oncure the GIRuu entiro ehannal will pcrf9Frm the intcnded function. T-hic toct;'oarifio th~e capability

e. the in;Gtru...ntation to p. . ide the ABG.TS actuation The ctpeointiR, .hall be left conciotent with the unit specific calibration procoduro tolcRRAn.Th Frogquoney ef 02 dayc isc based en the kn.wn eliability, ^f the manito;ing n^

equipmcnt and has boon ohGWn to be aeeeptablo through eperalting exporionce SR 3.3.8f SR 3.3.8.3 is the performance of a TADOT. This test is a check of the manual actuation functions and is performed every 18 months. Each manual actuation function is tested up to, and including, the relay coils. In some instances, the test includes actuation of the end device (e.g., pump starts, valve cycles, etc.). The Frequency is based on operating experience and is consistent with the typical industry refueling cycle.

The SR is modified by a Note that excludes verification of setpoints during the TADOT. The Functions tested have no setpoints associated with them.

(continued)

Watts Bar-Unit 1 B 3.3-177

ABGTS Actuation Instrumentation B 3.3.8 BASES SURVEILLANCE GR-344 REQUIREMENTS (continued) A CHANNEL CA*L^1-R^T,-ION is peoferm..d ovo 18 months, or approimato"ly at ovcr,' rcfucling. CHANNEL CLB TIN is a comploto chock of thoentrmn loop, includEing the scncor. The test Yerifioc, that the channol rocpendc to a mceacUrcd parameter Within the nococcar,' rango and accuracy. The Froguoncy 0s basod en eppomting expcrienco and icconcictont With the t'picaI nut, Fefue4Rg eyele-REFERENCES 1. Title 10, Code of Federal Regulations, Part 100.11, "Determination of Exclusion Area, Low Population Zone, and Population Center Distance."

Watts Bar-Unit 1 B 3.3-178

ABGTS B 3.7.12 B 3.7 PLANT SYSTEMS B 3.7.12 Auxiliary Building Gas Treatment System (ABGTS)

BASES BACKGROUND The ABGTS filters airborne radioactive particulates from the aro3a of the fuo, pool f"llowing a fuel handling aeidnt and from the area of active Unit 1 ECCS components and Unit 1 penetration rooms following a loss of coolant accident (LOCA).

The ABGTS consists of two independent and redundant trains. Each train consists of a heater, a prefilter, moisture separator, a high efficiency particulate air (HEPA) filter, two activated charcoal adsorber sections for removal of gaseous activity (principally iodines), and a fan. Ductwork, valves or dampers, and instrumentation also form part of the system. A second bank of HEPA filters follows the adsorber section to collect carbon fines and provide backup in case the main HEPA filter bank fails. The downstream HEPA filter is not credited in the analysis. The system initiates filtered ventilation of the Auxiliary Building Secondary Containment Enclosure (ABSCE) exhaust air following receipt of a Phase A containment isolation signal or a high radiaticn *ignal from the 'pont fule~e'aFea-The ABGTS is a standby system, not used during normal plant operations.

During emergency operations, the ABSCE dampers are realigned and ABGTS fans are started to begin filtration. Air is exhausted from the Unit 1 ECCS pump rooms, Unit 1 penetration rooms, and fuel handling area through the filter trains.

The prefilters or moisture separators remove any large particles in the air, and any entrained water droplets present, to prevent excessive loading of the HEPA filters and charcoal adsorbers.

The plant de.ig. baiG. requiroc that when m.Vi.g irradiated fuel in the Au-Xilia'y Building and/or ContainRmn* t with the Containment open to the Auxiliar,' Building A99GE spacoc, acignal fromB the epont fuel pool radiation monitore 0 RE 00 10Q2 andc 103 ll initiae a *:R nti -*Containment (CVI) in additioR to thAir lolatio.

norm.al function. In addition, a cignal from the contai*nmnt purge radiaton mPnit~r; 1 RE 00 130, and 131 or etheqr CVI signal will initiate that portion of the A^I norm.ally initiated by the epent fuel peel radiatioR monitor'. In aadditi** , the

........... .......... operable of......

. t....h....... pL RotrtiRG We 9p8R tar to e uAMMIOR'x Bu ilding du FOAringmvement of iraiae uanid otainment The ABGTS is discussed in the FSAR, Sections 6.5.1, 9.4.2, 15.0, and 6.2.3 (Refs. 1, 2, 3, and 4, respectively).

(continued)

Watts Bar-Unit 1 B 3.7-62 Revision 87, 110

ABGTS B 3.7.12 BASES (continued)

APPLICABLE The ABGTS design basis is establishey the consequences of the limiting SAFETY ANALYSES Design Basis Accident (DBA), which is a . dr det. The ar~aeysis 9f the fuel handling accidnt, gi in orno 32ac that all fUo rods in an assembly arc damaged. The analysis of the LOCA assumes that radioactive materials leaked from the Emergency Core Cooling System (ECCS) are filtered and adsorbed by the ABGTS. The DBA analysis of the fuel handliRng accidont assumes that only one train of the ABGTS is functional due to a single failure that disables the other train. The accident analysis accounts for the reduction in airborne radioactive material provided by the one remaining train of this filtration system. The amount of fission products available for release from the ABSCE is determined for a fuel handling a. . ,iden and for a LOCA. The accumptians and the analyciG for a fuel handling accident follow the guidanca provided in Regulatery Guide 1.25 (Ref. 6) and NUREG!CR 5009 (Ref. 11). The assumptions and analysis for a LOCA follow the guidance provided in Regulatory Guide 1.4 (Ref. 6).

The ABGTS satisfies Criterion 3 of the NRC Policy Statement.

The ABO*TS ic rcquired to be.p..abl, durng .... m.nt of i.radiatod fuol in tho uIxi,,ia. Building"HiRg- R .m;de d durin

,R of n fuel iR Reacter Building when the Rcacter Buiolding eicctablichod ac, part of the ABSCGE boundar; (see ' T-6

"^.* ^ux!ar ,,, 3.3.8, a4;*during

, andmcs,,*...,,".,

3.7.12, &3.9.4). Whon moVing irradiated contan~mt *4.

irro.o~n +k.Iincide fuoml Auiiar en~tanmcnet, At loac-1t ARA train Of the contafinmont purFgo cycto: mRuct 199 eperating or the contaminFAnt must be icolatod. I.Whon moVing irradiatod fuo in1 the ,A~uiliar; Building during times when the containmont ic opon to tho AuxilfiarI Building ABSCE Gpaecc, containment purgo can be operated, but oporation of the.y.t.M ic not required.

As stated previously, ProVicienc are provided to itronc the A[3I* and the; CVI instr'-mcntatOn to facilitato th actu-atiAn of an ABI* Whona CVI ic in.tiatod and the actuatRo-n f a C;VI when an ABI* .The folloWg tab pcifies the radiation monitorc and the accaciatod actuation incii on-to that must be available when the A9I* and GCI funcrtion6 Are interconnected and fiuell Gcbeing moved cither inside r-antainmontPR or in the Auxiliary' Building or iffuel mevement ic taking placa in bath areas:

-Tha RE portion ot thoAB: initiated by the Spent Fuel Pool P0 a-iation M itoe, 0 RE n010,2 and Q-403.

(continued)

Watts Bar-Unit 1 B 3.7-62A Revision 87, 110

ABGTS B 3.7.12 BASES (continued)

LCO Two independent and redundant trains of the ABGTS are required to be OPERABLE to ensure that at least one train is available, assuming a single failure that disables the other train, coincident with a loss of offsite power. Total system failure could result in the atmospheric release from the ABSCE exceeding the 10 CFR 100 (Ref. 7) limits in the event of a fuoel handling Accidont ef LOCA.

The ABGTS is considered OPERABLE when the individual components necessary to control exposure in the fuel handling building are OPERABLE in both trains. An ABGTS train is considered OPERABLE when its associated:

a. Fan is OPERABLE;

b. HEPA filter and charcoal adsorber are not excessively restricting flow, and are capable of performing their filtration function; and

c. Heater, moisture separator, ductwork, valves, and dampers are OPERABLE, and air circulation can be maintained.

APPLICABILITY In MODE 1, 2, 3, or 4, the ABGTS is required to be OPERABLE to provide fission product removal associated with ECCS leaks due to a LOCA and leakage from containment and annulus.

In MODE 5 or 6, the ABGTS is not required to be OPERABLE since the ECCS is not required to be OPERABLE.

IDurng moevement af irradiatod fuol in the fuol handlin~g aroa, the ABGTS;i roqlufirod to be OPERA.BLE to alleviate the concoquoncoc of afuel handlin a*e*d-nt. Sce addiVtnal dis.us.ion in the Background and Applicablo Safoty (continued)

Watts Bar-Unit 1 B 3.7-63 Revision 55, 87

ABGTS B 3.7.12 BASES (continued)

ACTIONS A.1 With one ABGTS train inoperable, action must be taken to restore OPERABLE status within 7 days. During this period, the remaining OPERABLE train is adequate to perform the ABGTS function. The 7 day Completion Time is based on the risk from an event occurring requiring the inoperable ABGTS train, and the remaining ABGTS train providing the required protection.

B.1 and B.2 -F.h[ n In MVOGnDE 1, 2, 3, or 4, whn Required Action A.1 cannot be completed within the associated Completion Time, or when both ABGTS trains are inoperable, the plant must be placed in a MODE in which the LCO does not apply. To achieve this status, the plant must be placed in MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and in MODE 5 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . The Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

"VWho RequirenAA.od c a.1cRnnot b complotod ...within the rcquirod p.otion Time, during MoV9omont of ful in tho fi fuoA handlin aroa, uo8Abioc tho

.... RA... ABGT. train. mu..t. be tart. d immodiately or fcl movemnt suspcndcd. This action oncuroc that the remainin~g train is GPERAR'\BLE, that no undotctodfailuroc freycnting 6yctcm: operation wall occur, and that any activo failuro will be readily doteeted.

ifthecsystem is not plaeed in epeffati8n, this actien roquirc of.fluell Rupna meycment, which prccluidcz a fuel handling accident. hidc ntpecluide the mavzmcn~t of fucel assemblies to a safe pocsitian.

(continued)

Watts Bar-Unit 1 B 3.7-64

ABGTS B 3.7.12 BASES W t .ho

tAnc 4 oA,fthe ABGT-l aFr inpeoablo durng9 mv omont of irradiated fuo -

accomblics in the fuel handling arca, actin musct be takon to plao tho uitot iA candition in which the LCO dooc not apply. Actio must be takonimditl to suspcnd meycment ef irradiated fuel assemblies in the fuel handling aroa. Thic, deeS nct prcclude tho meycment of fuol to a safe pocition SURVEILLANCE SR 3.7.12.1 REQUIREMENTS Standby systems should be checked periodically to ensure that they function properly. As the environmental and normal operating conditions on this system are not severe, testing each train once every month provides an adequate check on this system.

Monthly heater operation dries out any moisture accumulated in the charcoal from humidity in the ambient air. The system must be operated for > 10 continuous hours with the heaters energized. The 31 day Frequency is based on the known reliability of the equipment and the two train redundancy available.

SR 3.7.12.2 This SR verifies that the required ABGTS testing is performed in accordance with the Ventilation Filter Testing Program (VFTP). The ABGTS filter tests are in accordance with Regulatory Guide 1.52 (Ref. 8). The VFTP includes testing HEPA filter performance, charcoal adsorber efficiency, minimum system flow rate, and the physical properties of the activated charcoal (general use and following specific operations). Specific test frequencies and additional information are discussed in detail in the VFTP.

(continued)

Watts Bar-Unit 1 B 3.7-65

ABGTS B 3.7.12 BASES Deleted.

o ,

REFERENCES 5. .1w at9FYIcbIlue Into, MaFGH 1t9f-e-, "Ac.."uRD.O.,c Uced (continued) 9F yalatg te PteialRaIu819::; %=R68U8RGuu 9+ a Puei Handlinig Accident 6n the Fuel Handling and Storago Facility for Boiling@

and Prczzurizl-Pd Wvatcr ReActors."

6. Regulatory Guide 1.4, "Assumptions Used for Evaluating the Potential Radiological Consequences of a Loss of Coolant Accident for Pressurized Water Reactors."

7. Title 10, Code of Federal Regulations, Part 100.11, "Determination of Exclusion Area, Low Population Zone, and Population Center Distance."

8. Regulatory Guide 1.52 (Rev. 2), "Design, Testing and Maintenance Criteria for Post Accident Engineered-Safety-Feature Atmospheric Cleanup System Air Filtration and Adsorption Units of Light-Water Cooled Nuclear Power Plants."

9. NUREG-0800, Section 6.5.1, "Standard Review Plan," Rev. 2, "ESF Atmosphere Cleanup System," July 1981.

10. Watts Bar Drawing 1-47W605-242, "Electrical Tech Spec Compliance IDeleted. 1_ ,.* ,Tables-'"

11. 000, "Assessment of the Uco of RAExtondo 69E!C Bur.up Fuol in Light Wator Pe'vor Racstoe," W.S.

Nu.lcar Rcgulater Co i.....on, Fobruar,' 1088.

Watts Bar-Unit 1 B 3.7-67 Revision 29, 55

Insert B 3.7.13-1 1.183 (Ref. 6). The Total Effective Dose Equivalent (TEDE) for control room occupants, individuals at the exclusion area boundary, and individuals within the low population zone will remain within 10 CFR 50.67 (Ref. 7) and Regulatory Position C.4.4 of Regulatory Guide 1.183 (Ref. 6) for a fuel handling accident.

Insert B 3.7.13-2

6. Regulatory Guide 1.183, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors," July 2000.

7. Title 10, Code of Federal Regulations, 10 CFR 50.67, "Accident Source Term."

Fuel Storage Pool Water Level B 3.7.13 B 3.7 PLANT SYSTEMS B 3.7.13 Fuel Storage Pool Water Level BASES BACKGROUND The minimum water level in the fuel storage pool meets the assumptions of iodine decontamination factors following a fuel handling accident. The specified water level shields and minimizes the general area dose when the storage racks are filled to their maximum capacity. The water also provides shielding during the movement of spent fuel.

A general description of the fuel storage pool design is given in the FSAR, Section 9.1.2 (Ref. 1). A description of the Spent Fuel Pool Cooling and Cleanup System is given in the FSAR, Section 9.1.3 (Ref. 2). The assumptions of the fuel handling accident are given in the FSAR, Section 15.4.5 (Ref. 3).

I ýns e rtýB& 7.ý13 -1 APPLICABLE The minimum water level in the fuel storage pool me4 the assumptions of the SAFETY fuel handling accident described in Regulatory Guide 1.25 (Ref. 4). The ANALYSES .. ultant 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months thyroid do.. po. por..n at the.X.u.in aro*a boundary i6 a Fmall fr*ation oft . 0 th0FR 100 (Ref. 5) limit*.6 According to Referenc 4, there is 23 ft of ater between the top of the damaged fuel bundle and the fuel pool surface dur' a fuel handling accident. With 23 ft of water, the assumptions of Reference 4 can be used directly. In practice, this LCO preserves this assumption for the bulk of the fuel in the storage racks. In the case of a single bundle dropped and lying horizontally on top of the spent fuel racks, however, there may be < 23 ft of water above the top of the fuel bundle and the surface, indicated by the width of the bundle. To offset this small nonconservatism, the analysis assumes that all fuel rods fail, although analysis shows that only the first few rows fail from a hypothetical maximum drop.

The fuel storage pool water level satisfies Criterion 2 of the NRC Policy Statement.

(continued)

Watts Bar-Unit 1 B 3.7-68

Fuel Storage Pool Water Level B 3.7.13 BASES SURVEILLANCE SR 3.7.13.1 (continued)

REQUIREMENTS pool is normally stable. Water level changes are controlled by plant procedures and are acceptable based on operating experience.

During refueling operations, the level in the fuel storage pool is in equilibrium with the refueling canal, and the level in the refueling canal is checked daily in accordance with SR 3.9.7.1.

REFERENCES 1. Watts Bar FSAR, Section 9.1.2, "Spent Fuel Storage."

2. Watts Bar FSAR, Section 9.1.3, "Spent Fuel Pool Cooling and Cleanup System."

3. Watts Bar FSAR, Section 15.4.5, "Fuel Handling Accident."

4* * -" +/-EIF Guid 4.2 , MBF*I*' 4*9: 1 .* ,,-'.d, 1 fqr * * *..

thc Ptental RdiolGical onquoncoc of Fua Fa Handling Accidant i Deleted.e cuH Stoe Facilityfor Boig and PrczzuriZed Wate D. '" Titlo 10, Code of .,"DotFR-iatin Fcdoral Regulations,Part 100.11 of EcuinAroa, Low Population Zone, and Population Contor Distanc.

Watts Bar-Unit 1 B 3.7-70

Contaitnmct Ponotrations

-ýB Peletd 3.9.4 B 3.9 REFUELING OPERATIONS B 3.9.4 QARteiwe:1Rt-eHetFatfieAG DuFrig moevcmcnt of irradiatcd fuel assemblqios within contafinment, a rclcaso oe fission product rad fioactivity MWiti con_.tainment Will bo restricted fromR oscaping to thc onvironmon~t whcn thc LCO roquirments arc mct. In MODES!!! 4, 2, 3, and 4, T. EULNGOEAIN this isaccomplishcd by maintaining containmont OPERABLE as doscribed i LCOG 3.6.1, "Containmont." In MOIDE 6, tho potcntial for containm~ent pressurization as a recult of an accidcnt is not likely; therefoe*R rcquircmeRnG to iso.late thc cntainmcnt fro, thc outside atmo,,pher can bo losstrin4gcnt. Tho LCO roquircmcnts aro rcfcrred to as "containmcn~t slosuro" rathor tha

'containm~ent OPERABILITY." Con)tainmonAt clocro; F moans that all potential eceapc paths arc closed or capablo of being closed. Since thore-is no potcntial for containmont pressurization, the Appondix J leakage critoria and tests, are not FeqwiFed.

The containment serues t contain fission productr .adoa.ctivity that may be relcased fromR the reactor core following an accident, such that offsito radiation exposures are mnaintained well within the requirements of 10 CFR 100G.

..dditioally, tha containment prov,.ides radiation shielding fromR the fission products, that may be present in the containm~ent atmosrphere following accident eeedutueee 7 The containment equipment hatch, which iOpa. of theonentainment prccSur .

beundary, provides a mceanc for moeving large equipment and components into and out of contaminanet. DuFrig movement of irradiated fuel assemblfies within on~tainment, the equipment hatch mus~t bo hold in place by at least four bolts.

Good ongineering practice dicatates that the bolts required by this LCO b approx(imately equally spacod.

The containment air locks, which are also padt of the contaiminat prossuro boundaF', provido a m~eans for pcrsennel ascoss during MOIDE!S 1, 2, 3, and 4 unit operation in accordance with LCO 3.66.2, "Containment Air Locks." Emach air lock has a door at both ends. The doors are nrmFFally interlocked tc prevent Jinultaneous opening whon contaminFant OPERABILITY is required. IDurin periods of unit shuitdown when containment closurwe isent required, tho door interlock rnobhanismR may be disabled, allGowig both dooArs; of anR air*lock to8 reai ope for extended periods when frequont containmen~t ont'y isnOcosca'.

IDuFrig mo9vemen.t oef irradiated fuel assemblies within containmen8t, containm~ent EyoIriirnA t;V* rmqt3 ra ;

0% rnr orv ro,

+kn Ann.

v

~+ 1rin rnnnnrn Our ri or wo -t Moty I

r. nir n'n I "UM dfisabledl, but one air lock door must alwy rmicapable ot being closed.

This Page is Intentionally Left Blank.

Also, Bases Pages B3.9-13 through B3.9-16 Were Deleted.

(continued)

Watts Bar-Unit 1 B 3.9-12 Revision 37, 45 Amendment 26, 35

Contaim*eRt PoeRntrations B-3.94 BACIKGROUNID Tho rcquiromonts for containm~ent penotration closuro onsuro that a roloaso (6eeIVe~ed) of fission produc~t radioactivity within conRtainmon~t will bo rostrictod to withi Thc Rcatorte Building Purge Ventilation SystcmR oporatos to suipply ouitsido air inRto thc containm~ent for ventilation and cooling or hoating, to oqualizo intornal and external pressures.I and to reduce the concentration of noblo Gacos; within containmcnt prir to and uring porsonnie ass. ihe supply an. oxnau.... nos oach contain txo fisolation valvos. Becauso of their largoe sizc, tho 214nc containment loWOr empa tc aro physically Rt purgo valvass t rosttoritd~ tod my dogroes open. Tho Roactor Building Purgo andI Vontilation Systom valvos can be oponod in MOIDES 5 andI 6, but aro closod auitomatically by tho Enginoorod Safcty Foaturos Actuiatfion System (EiSFAS). In MODE 6, largo air oxchangos are noesesar,' to conduct rofuoling oporationc. Tho nrmFFal 21 inch purgo systemR is used for this purFposc. Tho ventilation systomn must bo oithor i~solatod or capablo of boing automatically isolatod upon dotoction of high radiation ovalsg

.Athin contaimonet.

The other containment penetrations that provide direct aeeecc fromR containment atmocphcre to outside atmosphers m~ust be isolatcd on at ceast onoe sid!

Isolation mnay be achieved by an OPERABLE! automatic isolation yak e, or b, a m~anual fisslation valve, blind flango, or oquivalent. Equivalont isolation m~ethods must be approved andI ma" incluido use of a m~aterial that can previde-a temnporar,, atmospheri prossuro, ventilation barrier for the otheFrecotainmen~t ponetrations during fuel movyemcnts (Ref. 1). Closroby~ o3 ther valves, or blind flanges may be used ifthey arc similar in capabilit' to thesa previdod-for containment isolation. Thcco may be constructed of standard matorials, and may be iustified on the basis of either nrAmal analysic methods or reasonbe

-ngin..r....judgmcnt (Ref. )..

APPFYIABLE IDuring moveymont of irradiated fuiel assemblies within containm~ent, tho mRost SAPFE-TY ANALYS9ES9 s.v.r rad*iological

. .cosoquwoceG result from a; fuel handling a*cident. The fuel handling accident is a pestulated oe'nt that involves damage to irradiated f-el (Ref. 2). Fuel handling accid8ets, analyzed in Reference 3, include drepping a single irradiatod fuel assembly and handling tool or a heavy objoet onto other iradaedfel assemblies. The requiremon8_tsF of LCO-3.0.7, "IRefueling Cavity Water Level," in conju-nction wth a minimum decay time of100 hours prior to ir.ra~dia~teAd f4uel moveement With Gontainmon~t closur~e eapabgilit' ensResF tha;t the relcass of fission product radioactivity, subsequent to a fuel handlinfg accidlent, rcsults indoses that arc wall within the guideline values specified in 10 CFR 100.

Stad*ard Reviow Plan, Sectin 16.7.4, Rey. 1 (Ref. 3), definRes "well withI "

10 CFR 100 to be 25% or less of the 10 CFR 100 values. The aEceptance limits for cffsite rodiotisn ex(posure will be 251%of 10 CFR 1 00 values Or the NRC staf approved licensing b~asis (e.g., a specified fraction of 10 CFIR 100 limnits).F (continued)

Wafts Bar-Unit 1 B 3.9-13 Revision 37, 45, 46, 73 Amendment 26, 35

Containmon~t Pcnotrationc R A44 RASPS; APPLICABL6E Contafinment ponetratioeR Satisf' CritoFrion :3 of the NRC Policy StatomonRt.

SAF~FlP ANALYSES L-GO This LCOG limits the concequences of a fuel handlinfg accidont in containmfent by limiting the potential cscapc paths for fission produst radioactivity rel cascd within containment. The LCOQ requires any penetration providing dircct access fromA the ccntainmcnt atmoespherc to ths ouitside atoshr to bsslosed cxecpt for tho OPERABLE Roastor Buwilding Purge and Vctain Sysem penotrations, and tho containment pcrsonncl airlocks. For tho OPERABLE Roastor Building Purgo and Ven.tilation Sy.tcm. ponetrations, thic LCO encuros. that tho.o ponotrationc are icolable by thc Containment VentilationIslaio Systom. The OPERABILITY roquiromon~tc for thic LCGO onsuro that tho automatic purge9 and cxhaust valvc closure times spccifiod in the FSEAR c~an bo ac-hieve.Ad_ -and, therefore, mnect the assumptions uced in thc sefot analysis to ensuro that releaecs thro)ugh tho valves aro termlinatcd, sush that radiological doses arc

'-thin thc acceptance limit.

The rcntaiwnmet pctrsonnl airlc'k dofors ma" be opcn duf'rig movement Af i Radiated fuel ir the containment prov.ided that one doer is capable of beirg closed-t- in the evenA-.t o-f -Afe handling accident and provid-ed- that ABGTSQ i OPERABLE in accordance with TS 3.7.12. Should a fuel handling accident occur inside containment, one personnel airleck door Will be closed following en cvaeuation ef -eontainment-.R* The LCOQ is modified by a No to allowing ponotration flow paths with direcAt ar-RPsses f.-Re;m the con~tainm~ent atmosphere to the outside atmosphcrc to be unisolated under adm~inistrative controls. Administraiv centrels ensure that 1) approprfiate personnel are aware of the open statuis of the penetration flew path duFrig movyement of irradiated fuel assem~blfies within on~tainmont, 2) specified individuals arc designated and readily avai able to isoelate the flow path in the event of a fuiel handling accident, 3) pen~etration flow paths, penetrating the Auxiliary Building Seconder,' Containment Enclosure (ABSCE) bounder;, arc limited to less than the ABSCE breach allowance, end 4) the ABOT-S is OPERABLE finaccordance with T-6 3.7.12. Operability of ABOQTS is required to alleviate the consequences of a FHA inside containment resulting in leakage of airborne radioactive m~aterial pact the open airlock or penetration flow paths prior to their closur~e.

(continued)

Watts Bar-Unit 1 B 3.9-14 Revision 37, 45, 73, 74 Amendment 26, 35

Containmont Ponotrations B-3.9.4 APPLICIABILI=TY Tho contaminanet pcnctratin roqirmets aro applicsablo dur~ing moGvemont oe iRadiatcd fuie assemblies wit-hin1R contRa.inmonl.t boc~auto thit s6whon thoro it; a potontial for tho limiting fuol handling accident. In MODE 1, 2, 3, an~d 1, contaminwant ponetratia roquiromcnts aro addrocsaed b"LCOQ 3.6.1. In MODESi 6 and 6, whan movcmcnt of irradiated fuael assombliosr within containment is not being canductcd, the potontial for a fuol handling accidont doees not exist. T-horcfore, undor theta conditions no reguiremon~te are placad on containmonet ponetration status.

AGTIGNS A.4 if tho coentainment equipment hatch, air lockc, Or any containmont ponotration that provwides direct access fromA the centainmcn~t atmRosphere to the outtido atinospharo ic not in thc rcguired status, including the Containmont Ventilation Isolation Systcm not capable cf automatic actuation when the purge and exhaust valvcs are opcn, thc unit must bc plaocd in a condition where thc isolato function is not nccdcd. Thic ic accomplished by immcdiatcly cucpcnding@

movemant of irradiatod fuel assemblios within contafinm~ent. Pe~fFotqrmancefa these acticns chall not precludc complctien of mavcmcn~t sf a compenant tc a safe pesiten7.

SURVE'L~L=ANCE &R-3-9A4.

REQUIREMENTS This Sur.'cillanao dcmRonctratcs that each Of the con;tainmenRt panotrations rcguircd to be in its slesed position is in that positionI. The Sur.'cillanca An the opon purge and cxhaust valves will demonstrate that the valves are not blocked fromR closing. Also the Sur.'eillance will damonstrate that each valve operator hat mo~tive power, which will entur~e that each valve is,capable of being clotad by an OPEF!RABLE!! automatfic. containm~ent ventilatfion isolation Signal.

The SurP.'oillance it- pe~frmo~d every' 7Edayt during movRFemet of irradiated fuel assemblies within containm~ent. The Survewillane finter.al is selected to be commensr8Rbata with the nrmFRal duration of time to complete fu el handling operations. Acu.'ilacebfore the stadI ofrfeln perations Will provide two sr.ailansc verifications durIing the applicable period for this CO As-or hrc usbh, this Sur.'cllansc ensuras that a postulatad fuel handling accident that releases fission product radioactivity within the containm~ent Will net result in a relcase of significant fission produc~t radioactivit' to the environenFet in excess of these Froemmfended by StanRdard ReviewA. PlanR Seton1.7.1 (Reference 3).

(continued)

Watts Bar-Unit 1 B 3.9-15 Revision 37, 45 Amendment 26, 35

Containmont Ponotrations SURVEILL=ANCE REQUIREM ENTS (e9Awe~Ed) T-his SurVoillanso domonsRtratos that oach containmonRt purge and oxhaus~t valvo aetuatos ta isslaonposition On mganual initiation Or On an actual or simu lated high rodiaticn signal. The 18 month Frg usncy maintainase nsistoncy with other mimlar ESFAS inGtrumcnetation and valvc tcsting reguirae~nets. LCO3..6 Centainmont Vcntilation Isolation Instru-mentation requircs a CHANNEL CHECK evor,' 12 hou-;ARScandaOT eyo~'O2 days to ensuro tho channel OPERABILITY duri.g refuoI.ng opcrations. Ev-er 18 mon.ths a CHANNEL CALIBRATION is pc~frmBEd. The system actuation responsc time Is dleemontrated ovary 18 months, dUurig refuoling, on a STAGGERED TEFS;T BASIS. SR 3.6..

dcmonstrates th.at the i otime of cs*h valve isWinaccordn with tho Insc*..ie Tasting PrFogram. equiFremeRnG. These Sueillan.es peF.ormd duFrin MODEF 65will ensure that the valves arc capabgle of cl9osin after a postu lated fuoel C- "Hq a"" Hn-M -- , 8* .

OR-..-.- - - - -.,,R... .,v eentaenent.

REFERENCES "s of Silione.,v Sealant to Maintain Containm.ent ,ntegr,. IS," CP Nuvelar Sof'"Evalu.ation SE 000200000, Rev. 0, May 20, 1088.

Bar FSAR, Section 15.4.5, "Design Rasis Fuel' Handlng

.atts Aeeimdents.

N, 0800, Standard Review Plan, Section 45.7.4, "Radiolegisal vUREG C-nsalu.Rnsa of F-wel Hanid.ling Accidents," Rev. 4, July 1081.

4- Generic.Lette 88 17, "Less of Dsay Heat Removal."

Watts Bar-Unit 1 B 3.9-16 Revision 37 Amendment 26

Insert B 3.9.7-1 the limits defined in 10 CFR 50.67 (Ref. 7) and Regulatory Position C.4.4 of Regulatory Guide 1.183 (Ref. 8)

Insert B 3.9.7-2 (Regulatory Position 2 of Appendix B to Regulatory Guide 1.183 (Ref. 8)) allows an overall iodine decontamination factor of 200 Insert B 3.9.7-3 8% of the 1-131, 10% of the Kr-85, and 5% of the other noble gases and iodides from the total fission product inventory in accordance with Regulatory Position 3.1 of Regulatory Guide 1.183 (Ref. 8)

Insert B 3.9.7-4 without containment closure or Auxiliary Building isolation Insert B 3.9.7-5

7. Title 10, Code of Federal Regulations, 10 CFR 50.67, "Accident Source Term."

8. Regulatory Guide 1.183, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors," July 2000.

Refueling Cavity Water Level B 3.9.7 B 3.9 REFUELING OPERATIONS B 3.9.7 Refueling Cavity Water Level BASES BACKGROUND The movement of irradiated fuel assepnblies within containment requires a minimum water level of 23 ft above th" top of the reactor vessel flange. During refueling, this maintains sufficient wal er level in the containment, refueling canal, fuel transfer canal, refueling cavity, abd spent fuel pool. Sufficient water is necessary to retain iodine fission priluct activity in the water in the event of a fuel handling accident (Refs. 4-eRd 2). Sufficient iodine activity would be retained to limit offsite doses from the accident t 4 25% of 10 CFR 100 .imite,as pro'vded by the guidanco of Rcferenco 3.

Insert B 3.9.7-1 --Iln tB3.9.7-2I APPLICABLE During movement of irradiated fuel assemblies, the watelevel in the refueling (Refs. 2 SAFETY ANALYSES canal and the refueling cavity is an initial condition desi n paramete .t - and 8) analysis of a fuel handling accident in containment, 'ttd b3,' R99--*t9."

Guide 1.25 (Ref. 1). A minimum water level of 23 ft kegulatory Position .4.ee Ref. 1)allews a deE;9ntamnination faotor of 100 (RcgUlatory Pocition C.1.g o Ref. 4) to be used in the accident analysis for iodine. This relates to the assumptionjtaa % of the total iodine released from the pellet to cladding gap of ropped fuel assembly rods is retained by the refueling cavity water.

99.5*The fuel pellet to cladding gap is assumed to contain 0-al f*ec I The fuel handling accident analysis inside containment is described nsert B 3.9.7-3 Reference 2. With a minimum water level of 23 ft 41 a minimum decay time of 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months prior to fuel handling the analysis and test ograms demonstrate that the iodine release due to a pos ated fuel handling acci**t is adequately captured by the water and offsite ses are maintained wit lallowable limits (Refs. .Ijpe ). Insert B 3.9.7-4 in conjunction with E7 and 8 ýRefueling cavity water level satisfies Criterion 2 of the NRC Policy Statement.

(Continued)

Watts Bar-Unit 1 B 3.9-25 Revision 45, 55 Amendment 35

Refueling Cavity Water Level B 3.9.7 BASES (continued)

SURVEILLANCE SR 3.9.7.1 REQUIREMENTS Verification of a minimum water level of 23 ft above the top of the reactor vessel flange ensures that the design basis for the analysis of the postulated fuel handling accident during refueling operations is met. Water at the required level above the top of the reactor vessel flange limits the consequences of damaged fuel rods that are postulated to result from a fuel handling accident inside containment (Ref. 2).

The Frequency of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is based on engineering judgment and is considered adequate in view of the large volume of water and the normal procedural controls of valve positions, which make significant unplanned level changes unlikely.

REFERENCES 1. rKeaw late." Guide 1.25, "As'-mottioGc Ucod for E'.'a-uatiRE thO Pctontial IDeleted.

Radiological Consequcnccc of a Fuel Handling Aeeidcnit in the Fuel Handling andE Stcrago Facility for Boiling and PReGcuriZed Water Reac~tGrc," U.S. NuEcloar Regulator; Commiscion, March 23, 107-2.

2. Watts Bar FSAR, Section 15.4.5, "Fuel Handling Accident."

3. NUREG-0800, "Standard Review Plan," Section 15.7.4, "Radiological Consequences of Fuel-Handling Accidents," U.S. Nuclear Regulatory Commission.

4. Title 10, Code of Federal Regulations, Part 20.1201(a), (a)(1), and(2)(2),

"Occupational Dose Limits for Adults."

5. Ma.RinwGki, ). P., Bell, M. j., DUhn, E., and Locanto, j., W.AP 7-828, Radiological ConeequoncocF_ of -AFuel- Handling AccGident,

{Deleted. Docembo~hr 1971-.

6. &,.NUREG/CR5099, "AcscGsmoqnt of tho Ico of Extonqdod 1":.F,-p Fuol8 in L!ght Water Power Roactrcr," U. S. N-cloar Regulatoer Commiscion, 7Deleted.ý-ý FebFia-, 4-99.

ilnsert B 3.9.7-5 Watts Bar-Unit 1 B 3.9-27 Revision 55

Rcactor Building- Puree Air Cloanue r B 3.9.8

>-*Deeed B 3.9 REFUELING OPERATIONS B 3.9.8 e twgt4r I-R AiedGeae4p YfNt BASES BACKGROWNID Tho Roactor Building Pu'rg Air Cloanup Unite aro an onginoorod cafot' foature of tho- Ro-actor Bulilding Purgo Vontilatio Syctom which ic a non caft foaturo

'cntilation s *"y*m. Tho a!rF cloaup unite* cotaiR profilte.re, HEPA Fto, 2i thick charcoal adsorborc, houcingc and ductwork. Anytime fuel handlingq oporation arc being carried on incido tho prim~arY contafinment, oithor tho containment ventilatioR Will be icolated or tho Reactor Building Pu'rge air Geoaup units will be OPE!!RAI3BLE (Ref. 1).

The Rooster Building Purge Ventilation Syctomf provides mnechanical ventfilation of the prim;*ay containment, tho inetrument room located within the contoinmont, and tho annuluc. The cyctom is decignod te cupply froch air for breathing and conRtamfination control to allow preonnolP acceAss for mnaintenanso and rofuoling oporatioe. Tho. oxhaut air i. filtered by tho Roactor Buiding Purgo Air Cleanup UniRte- to-limit tho rolooco-RA o-f radioaetivit' to tho- A-.onvronmont.

The contaminant upper and lower cempadtmonte aro purgod with froth afir by tho Reactor Building Purgo Vontilation SyctomA boforo occupancy. Tho annulucF cAnR be purged with frech air during rcactor chutdown Or at timoce whon tho annuluc

aeuuFm control system of the Emoregcncy Gas Troeatment System it shut dewp,.

The instrumsnet room is purged with froth air duIFrin oporation cf the Reactcr Buildinig Purgo Ventilation System cr it scparatcly purgod by the Instrument Room Purge Sub*,yctm. All pur,, vontilation funotioRn aFe nFn Safety rolat1d.

The Roastor Buildinig Purgo Vontilatio Syctemf it sized to prcvide adequato

entilation for porconnl to p~erForm work incid~e tho prim~arY containmont and tho annuuc dringall nrmFFal oporatione. In tho; oVont of Afuol handling accident, tho Reactor Building Pur~go Vontilatien Systtom it icolatod. Tho Reactor Building Purge Air Cleanup Unite a*r always availablo at pacsive inline com.pononte- to perform their

÷ function immediatoly aftFr a fuel handling aGcIdet to proce..

activity contained in exhaust air bofore it roachoc tho outsido enviGronment.

Bases Page B 3.9-29 Intentionally Left Blank Also, Bases Pages B 3.9-30 through B 3.9-32 Were Deleted (continued)

Watts Bar-Unit 1 B 3.9-29

Roactor Buwilding Purgo Air Cicanup Units 63.9.

BASES&

BACKGROUND Tho Prim.ar' containm.ent cxhaust is mon itorod by

.. a radiatioR dotoctr which (eee#RUG4d) providcs automatic containment purgc vontilation systemioainuo dctccting the sctpeint radioactivity in the exhaust air stroa~m.. Thoeotimn purgc ventilation isolation valvos will be automatically clesed upon the actuation ef4 -AConcc-Rn.

unFcn Vent Isolation (CVI) signal whorever thc pinmal' rcntainAment Is boein purgcd duHrin RnGoal o*poation Or up**manual n a*

c tVaio Ufro tho Main Control Room (Rot 2). Rcqirments for Containment Vent Iatio Instruontation arovoereofd by LC 0 3.3.6.

APPLIGABLE The Reaceto Buolding PuHge VentilatioR System air cleanu.p unite onsfu that tho SA÷ETY ANALYSlES roloase of radioaetivity to tho envirnmeVnt Is limited by leaRi* g UpIontainment cxhauct during a fuel ha dlingbefore tho containment purge exhaust n accddent

.alves are isolated. Reactor Building Pur~ge Ventilation System filter effioeisI one of the inputs for the analysis of the envieronmental consequences of a fuel handling accident. Containtnmet isolation can only result in smaller releases o radioactivity to the envronment (Ref. 1). Tho Contafinm~ent Vent Isolation System ensures that the containment vent and purgo penotrations Will be autematisally iselated upon doteetion of high radiation levols within the containmot (Ref. 2).

Containmient Vent Isolatin I*t*rtumentation iRI adressed by LCO 3.3.6.

The Reactor Building Purge Air Cleanup Units satiy Criterfion 3 of the NRC Pelicy Statoment.

In addition, durin moeot of irr-ad-iatedt fuelR in thEI Au-xilia; Building when on~tainm~ent isoe6ote uiir'Bilding spaces, a high radiation signal fromR the spent fuel pool accident radiation moni~tors Will initiato a CVI.

LOGThe safety function of the Reactor Bulilding Purge Air Cloanu1p Unit is related to the initial control of offsite radiation exposures resulting fromA a fuel handling acciodent inside containment. DuFrig a fuel handling accident inside contafinm~ent, the Rcaatsr Building Purge Air Cloanup Unit provides, a filtered path for cleaning Up any air leaving the contaiRnmont until the containm~ent ventilation is, isolated.

The plant design basis roquliros that when mRoving irradiated fuel in the Auxiliary Building and/or Containmcnt With the Containment epen to the Aux(iliar; Building paes km hesp~tfuiel radiation ABG i aGi~a 0R 0402 1FOief and 103 Will initiate- -A CVI in addition to their no9rmal- funcAtion. In Maddiio, a signal fFrom the contafinment purge radiation mon9itors; 1 RE 00 130, and 1,31 or-o-ther' (continued)

Watts Bar-Unit 1 B 3.9-30 Revision 87,

KoactorG4F Bu1ilingM Purgev Air Cleanup11 -Ueite 8404 CVI\GGsignal will initiatc that pod-ion of the ABI normally initiatod by the cpont

{eee9tinbled) fuel Pool raidiatfion monietors. in addition, thc ABOTS8 must remain* operable i these containmon~t penetrations are opcn to the Auxiliar' Building durin mevomont of irradiatod fuel in sido contamiennt.

APPLICABIAn AR initial asmption in the analysis ef a fuel handling aceident incido conainentisthat tho accidont occursFF whle rrdito fuol is being handled.

T-herefore, LCO 3.0.8 is applicablo only at this timo. See additional discussion6in the Applicable Safo Analysis and LCOQ sectiens.

AGCTONS A.-and-A.2 If one Reactr-,tAF Bulding Purge Air Cloanup Unit is,inperablc, that aiirleanup nit must *selatcd.

be This places thc systeminORthe requirod ac.ident sonfiguration, thus allowing refueling to continue..afte-r ver.ig . the reainRig airc*leanup Ruit i alignod and OPERABLEi.

The immediate Completeon Time is consistcnt with the required times fer actions to bc pcIfeFrrnd without delay anId in a Go*nrolled mann*er1.

8.4 With two Reactor Building Purgo Air Cleanup Units ineperablo, movement ef iRradiated fuel assemblies within containment must be suspended. Ti precludes tho possibility of a fuel handlinRg accident in contafinment with both ReactrfBuilding Purge Air Cleanup Unitstin PerFoeprm,;anne of this action shall not proclude moving a compoenet to a+aopsto.

The immediate Cempletien Time is con~sitcnt with the required times6 for actions-to be porfoFrmod without delay and in a controlled mannr.

SURV EILLANCE 8R34944 SEQU IREM ENTS The-Ventilatfion Filte Testing Program (VFmTP) 9enmpasses tho Reactor Building Pur~ge Air Cleanu1p UnIit filterF tests6 in accorFdance with Regulator,' Guide 1.52 (Ref. 3). The VFT-P includes testing the per9fonrmane of the HEPA filter, charcoal adeorbor efficiency, mini mum. flow rate, and the phySi*al Ipepies et.,

9 (continued)

Watts Bar-Unit 1 B 3.9-31 Revision 45, 87, 110 Amendment 35

Reactor Building Purgoe Air Cloanup BASES SURVEILL'ANGE SR 3.0.8.1 (eentinued)

REQUIREMENTS the aetivated Gharcoal. Spocfifie test FrFeguencsie and additional infoFrmation arc diEccusrcd in detail in the VFT:P-RFE=RF=NCES Fm~ ;ABrw FS; R,Soctio 415.5.6, "EnviFEronmental ConcoguoncRqGe f a Potula Itod Fuol8 Handling Accident'

.2 Vatt Bar FSAR, Section 0.4.6, "Reactor Building Pur~ge Ventilating SyeteFA.

3- Regulator; Guide 1.52 (Rev. 02), 'Dcoign, Toc~ting and Maintenance Critcria fer Rest Accident En~ginccrcd igaf* Featuro Atmocphfer Cleanup System Air Filtration and Adcorption Un~fits of Light Watcr Ocoled Nuclear Pewer PlantG."

Watts Bar-Unit 1 B 3.9-32

Decay Time B 3.9.10 B 3.9 REFUELING OPERATIONS B33.9. 10 Decay Time BASES BACKGROUND Section 15.5.6 of the WBN, Unit 1 UFSAR (Ref. 1) defines the assumptions of the fuel handling accident radiological analysis, including a minimum decay time for irradiated fuel assemblies prior to movement. This assumption ensures that the inventory of radioactive isotopes is at a level that supports the safety analysis assumptions.

To ensure that irradiated fuel assemblies have decayed for the appropriate period of time, a limitation is established to require the reactor core to be subcritical for a time period at least equivalent to the minimum decay time assumption in the fuel handling analysis prior to allowing irradiated fuel to be moved.

Given that no irradiated fuel assembly will be moved outside of the containment until the minimum decay time requirement is met, this requirement also ensures that any irradiated fuel assemblies that are moved outside of the containment meet the decay time assumption in the radiological analysis of the fuel handling accident.

APPLICABLE The radiological analysis of the fuel handling accident (Ref. 1) assumes a SAFETY ANALYSES minimum decay time prior to movement of irradiated fuel assemblies. The requirements of LCO 3.3.7, "Control Room Emergency Ventilation System (CREVS) Actuation Instrumentation," LCO 3.7.10, "Control Room Emergency Ventilation System (CREVS)," LCO 3.7.11, "Control Room Emergency Air Temperature Control System (CREATCS)," and LCO 3.9.7, "Refueling Cavity Water Level," in conjunction with a minimum decay time of 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months prior to irradiated fuel movement ensures that the release of fission product radioactivity, subsequent to a fuel handling accident, results in doses that are within the requirements of 10 CFR 50.67 (Ref. 2) and Regulatory Position C.4.4 of Regulatory Guide 1.183 (Ref. 3).

The decay time satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii).

Watts Bar-Unit 1 B 3.9-35 Amendment

Decay Time B 3.9.10 BASES LCO A minimum decay time of 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months is required prior to moving irradiated fuel assemblies within containment. This preserves an assumption in the fuel handling accident analysis (Ref. 1), and ensures that the radiological consequences of a postulated fuel handling accident inside containment are within acceptable limits.

APPLICABILITY This LCO applies during movement of irradiated fuel assemblies within the containment, since the potential for a release of fission products exist.

ACTIONS A.1 When the initial conditions for prevention of an accident cannot be met, steps should be taken to preclude the accident from occurring. When the reactor is subcritical for < 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months , movement of irradiated fuel assemblies within containment must be suspended. This action precludes the possibility of a fuel handling accident in containment. This action does not preclude moving a fuel assembly to a safe position.

The immediate Completion Time is consistent with the required times for actions to be performed without delay and in a controlled manner.

SURVEILLANCE SR 3.9.10.1 REQUIREMENTS This SR verifies that the reactor has been subcritical for at least 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months prior to moving irradiated fuel assemblies by confirming the date and time of subcriticality. This ensures that any irradiated fuel assemblies have decayed for at least 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months prior to movement. The Frequency of "Prior to movement of irradiated fuel in the containment" is appropriate, because it ensures that the decay time requirement has been met just prior to moving the irradiated fuel.

REFERENCES 1. Watts Bar UFSAR, Section 15.5.6, "Environmental Consequences of a Postulated Fuel Handling Accident."

2. Title 10, Code of Federal Regulations, 10 CFR 50.67, "Accident Source Term."

3. Regulatory Guide 1.183, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors," July 2000.

Watts Bar-Unit 1 B 3.9-36 Amendment

ATTACHMENT 3 PROPOSED TS CHANGES (FINAL TYPED) FOR WBN, UNIT I

Containment Vent Isolation Instrumentation 3.3.6 3.3 INSTRUMENTATION 3.3.6 Containment Vent Isolation Instrumentation LCO 3.3.6 The Containment Vent Isolation instrumentation for each Function in Table 3.3.6-1 shall be OPERABLE.

APPLICABILITY: MODES 1, 2, 3, and 4 ACTIONS

NOTIz ---------------------------------------------------------------

Separate Condition entry is allowed for each Function.

CONDITION REQUIRED ACTION COMPLETION TIME A. One radiation monitoring A.1 Restore the affected channel 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months channel inoperable, to OPERABLE status.

(continued)

Wafts Bar-Unit 1 3.3-52 Amendment 35,

Containment Vent Isolation Instrumentation 3.3.6 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME B. One or more Functions with --------------- NOTE ----------

one or more manual or One train of automatic actuation logic automatic actuation trains may be bypassed and Required Action inoperable. B.1 may be delayed for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for Surveillance testing provided the OR other train is OPERABLE.

Two radiation monitoring channels inoperable. B.1 Enter applicable Conditions Immediately and Required Actions of OR LCO 3.6.3, "Containment Isolation Valves," for Required Action and containment purge and associated Completion Time exhaust isolation valves made of Condition A not met. inoperable by isolation instrumentation.

Watts Bar-Unit 1 3.3-53 Amendment

Containment Vent Isolation Instrumentation 3.3.6 SURVEILLANCE REQUIREMENTS

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

Refer to Table 3.3.6-1 to determine which SRs apply for each Containment Vent Isolation Function.

SURVEILLANCE FREQUENCY SR 3.3.6.1 Perform CHANNEL CHECK. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months This surveillance is only applicable to the actuation logic of the ESFAS instrumentation.

SR 3.3.6.2 Perform ACTUATION LOGIC TEST. 92 days on a STAGGERED TEST BASIS This surveillance is only applicable to the master relays of the ESFAS instrumentation.

SR 3.3.6.3 Perform MASTER RELAY TEST. 92 days on a STAGGERED TEST BASIS SR 3.3.6.4 Perform COT. 92 days SR 3.3.6.5 Perform SLAVE RELAY TEST. 92 days OR 18 months for Westinghouse type AR relays SR 3.3.6.6 ------------------- NOTE -----------

Verification of setpoint is not required.

Perform TADOT. 18 months SR 3.3.6.7 Perform CHANNEL CALIBRATION. 18 months Wafts Bar-Unit 1 3.3-54 Amendment 17, 68

Containment Vent Isolation Instrumentation 3.3.6 Table 3.3.6-1 (page 1 of 1)

Containment Vent Isolation Instrumentation FUNCTION REQUIRED SURVEILLANCE ALLOWABLE CHANNELS REQUIREMENTS VALUE

1. Manual Initiation 2 SR 3.3.6.6 NA

2. Automatic Actuation Logic 2 trains SR 3.3.6.2 NA and Actuation Relays SR 3.3.6.3 SR 3.3.6.5

3. Containment Purge Exhaust 2 SR 3.3.6.1 < 2.8E-02 pCi/cc Radiation Monitors SR 3.3.6.4 (2.8E+04 cpm)

SR 3.3.6.7

4. Safety Injection Refer to LCO 3.3.2, "ESFAS Instrumentation," Function 1, for all initiation functions and requirements.

Watts Bar-Unit 1 3.3-55 Amendment 74,

Containment Vent Isolation Instrumentation 3.3.6 Page Intentionally Left Blank Watts Bar-Unit 1 3.3-56 Amendment

ABGTS Actuation Instrumentation 3.3.8 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME B. (continued) B.2 Place both trains in Immediately emergency radiation protection mode.

C. Required Action and C.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time for Condition A or B not met. AND C.2 Be in MODE 5. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months SURVEILLANCE REQUIREMENTS

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

Refer to Table 3.3.8-1 to determine which SRs apply for each ABGTS Actuation Function.

SURVEILLANCE FREQUENCY SR 3.3.8.1 -------------------- NOTE -----------------

Verification of setpoint is not required.

Perform TADOT. 18 months Watts Bar-Unit 1 3.3-62 Amendment

ABGTS Actuation Instrumentation 3.3.8 Table 3.3.8-1 (page 1 of 1)

ABGTS Actuation Instrumentation APPLICABLE MODES OR OTHER SPECIFIED REQUIRED SURVEILLANCE ALLOWABLE CHANNELS REQUIREMENTS VALUE FUNCTION

1. Manual Initiation 1,2,3,4 2 SR 3.3.8.1 NA (a) 2 SR 3.3.8.1 NA

2. Deleted

3. Containment Isolation - Refer to LCO 3.3.2, Function 3.a., for all Phase A initiating functions and requirements.

I Watts Bar-Unit 1 3.3-63 Amendment

ABGTS Actuation Instrumentation 3.3.8 Page Intentionally Left Blank Watts Bar-Unit 1 3.3-64 Amendment

ABGTS 3.7.12 3.7 PLANT SYSTEMS 3.7.12 Auxiliary Building Gas Treatment System (ABGTS)

LCO 3.7.12 Two ABGTS trains shall be OPERABLE.

APPLICABILITY: MODES 1, 2, 3, and 4 ACTIONS I

CONDITION REQUIRED ACTION COMPLETION TIME A. One ABGTS train A.1 Restore ABGTS train to 7 days inoperable. OPERABLE status.

B. Required Action B.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and associated Completion Time of AND Condition A not met.

B.2 Be in MODE 5. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months OR Two ABGTS trains inoperable.

Watts Bar-Unit 1 3.7-27 Amendment

ABGTS 3.7.12 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.12.1 Operate each ABGTS train for > 10 continuous hours 31 days with the heaters operating.

SR 3.7.12.2 Perform required ABGTS filter testing in accordance In accordance with with the Ventilation Filter Testing Program (VFTP). the VFTP SR 3.7.12.3 Verify each ABGTS train actuates on an 18 months actual or simulated actuation signal.

SR 3.7.12.4 Verify one ABGTS train can maintain a pressure 18 months on a between -0.25 and -0.5 inches water gauge with STAGGERED TEST respect to atmospheric pressure during the post BASIS accident mode of operation at a flow rate > 9300 and

< 9900 cfm.

Watts Bar-Unit 1 3.7-28 Amendment

Deleted 3.9.4 3.9 REFUELING OPERATIONS 3.9.4 Deleted Wafts Bar-Unit 1 3.9-6 Amendment 26, 35,

Deleted 3.9.4 Page Intentionally Left Blank Watts Bar-Unit 1 3.9-7 Amendment

Deleted 3.9.8 3.9 REFUELING OPERATIONS 3.9.8 Deleted Watts Bar-Unit 1 3.9-14 Amendment 35,

Deleted 3.9.8 Page Intentionally Left Blank Watts Bar-Unit 1 3.9-15 Amendment

Decay Time 3.9.10 3.9 REFUELING OPERATIONS 3.9.10 Decay Time LCO 3.9.10 The reactor shall be subcritical for > 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months .

APPLICABILITY: During movement of irradiated fuel assemblies within the containment.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Reactor subcritical for A.1 Suspend movement of Immediately

< 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months . irradiated fuel assemblies within containment.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.9.10.1 Verify the reactor has been subcritical for > 100 Prior to movement of hours. irradiated fuel within containment Watts Bar-Unit 1 3.9-17

Procedures, Programs, and Manuals 5.7 5.7 Procedures, Programs, and Manuals 5.7.2.19 Containment Leakage Rate Testing Program (continued)

Leakage rate acceptance criteria are:

a. Containment overall leakage rate acceptance criterion is < 1.0 La. During the first unit startup following testing in accordance with this program, the leakage rate acceptance criteria are < 0.60 La for the combined Type B and Type C tests, and

< 0.75 La for Type A tests.

b. Air lock testing acceptance criteria are:

1) Overall air lock leakage rate is < 0.05 La when tested at > Pa.

2) For each door, leakage rate is < 0.01 La when pressurized to > 6 psig.

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

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

5.7.2.20 Control Room Envelope Habitability Program A Control Room Envelope (CRE) Habitability Program shall be established and implemented to ensure that CRE habitability is maintained such that, with an OPERABLE Control Room Emergency Ventilation System (CREVS), CRE occupants can control the reactor safely under normal conditions and maintain it in a safe condition following a radiological event, hazardous chemical release, or a smoke challenge. The program shall ensure that adequate radiation protection is provided to permit access and occupancy of the CRE under design basis accident (DBA) conditions without personnel receiving radiation exposures in excess of the applicable regulatory requirement (i.e., 5 rem Total Effective Dose Equivalent (TEDE) for a fuel handling accident or 5 rem whole body or its equivalent to any part of the body for other accidents) for the duration of the accident. The program shall include the following elements:

a. The definition of the CRE and the CRE boundary.

b. Requirements for maintaining the CRE boundary in its design condition including configuration control and preventive maintenance.

c. Requirements for (i) determining the unfiltered air inleakage past the CRE boundary into the CRE in accordance with the testing methods and at the Frequencies specified in Sections C. 1 and C.2 of Regulatory Guide 1.197, "Demonstrating Control Room Envelope Integrity at Nuclear Power Reactors,"

Revision 0, May 2003, and (ii) assessing CRE habitability at the Frequencies specified in Sections C.1 and C.2 of Regulatory Guide 1.197, Revision 0.

(continued)

Wafts Bar-Unit 1 5.0-25 Amendment 5, 70, 78,

ATTACHMENT 4 PROPOSED TS BASES CHANGES (FINAL TYPED) FOR WBN, UNIT 1

Containment Vent Isolation Instrumentation B 3.3.6 B 3.3 INSTRUMENTATION B 3.3.6 Containment Vent Isolation Instrumentation BASES BACKGROUND Containment Vent Isolation (CVI) Instrumentation closes the containment isolation valves in the Containment Purge System. This action isolates the containment atmosphere from the environment to minimize releases of radioactivity in the event of an accident. The Reactor Building Purge System may be in use during reactor operation and with the reactor shutdown.

Containment vent isolation is initiated by a safety injection (SI) signal or by manual actuation. The Bases for LCO 3.3.2, "Engineered Safety Feature Actuation System (ESFAS) Instrumentation," discuss initiation of SI signals.

Redundant and independent gaseous radioactivity monitors measure the radioactivity levels of the containment purge exhaust, each of which will initiate its associated train of automatic Containment Vent Isolation upon detection of high gaseous radioactivity.

The Reactor Building Purge System has inner and outer containment isolation valves in its supply and exhaust ducts. This system is described in the Bases for LCO 3.6.3, "Containment Isolation Valves."

(continued)

Wafts Bar-Unit 1 B 3.3-154 Revision 43, 87, 110,

Containment Vent Isolation Instrumentation B 3.3.6 BASES (continued)

APPLICABLE The containment isolation valves for the Reactor Building Purge System SAFETY ANALYSES close within six seconds following the DBA. The containment vent isolation radiation monitors act as backup to the SI signal to ensure closing of the purge air system supply and exhaust valves. Containment isolation in turn ensures meeting the containment leakage rate assumptions of the safety analyses, and ensures that the calculated accidental offsite radiological doses are below 10 CFR 100 (Ref. 1) limits.

The Containment Vent Isolation instrumentation satisfies Criterion 3 of the NRC Policy Statement.

(continued)

Watts Bar-Unit 1 B 3.3-154A Revision 43, 87, 110,

Containment Vent Isolation Instrumentation B 3.3.6 BASES LCO 3. Containment Radiation (continued)

The LCO specifies two required channels of radiation monitors to ensure that the radiation monitoring instrumentation necessary to initiate Containment Vent Isolation remains OPERABLE.

For sampling systems, channel OPERABILITY involves more than OPERABILITY of the channel electronics. OPERABILITY may also require correct valve lineups and sample pump operation, as well as detector OPERABILITY, if these supporting features are necessary for trip to occur under the conditions assumed by the safety analyses.

Table 3.3.6-1 specifies the Allowable Value (AV) for the Containment Purge Exhaust Radiation Monitors. This AV is based on expected concentrations for a small break LOCA, which is more restrictive than the 10 CFR 100 limits. The specified AV is more conservative than the analytical limit assumed in the safety analysis in order to account for instrument uncertainties appropriate to the trip function. The actual nominal Trip Setpoint is normally still more conservative than that required by the AV. If the setpoint does not exceed the applicable AV, the radiation monitor is considered OPERABLE.

4. Safety Injection (SI)

Refer to LCO 3.3.2, Function 1, for all initiating Functions and requirements.

APPLICABILITY The Manual Initiation, Automatic Actuation Logic and Actuation Relays, Safety Injection, and Containment Radiation Functions are required OPERABLE in MODES 1, 2, 3, and 4. Under these conditions, the potential exists for an accident that could release significant fission product radioactivity into containment. Therefore, the Containment Vent Isolation Instrumentation must be OPERABLE in these MODES. See additional discussion in the Background and Applicable Safety Analysis sections.

(continued)

Watts Bar-Unit 1 B 3.3-156 Revision 45, 87, 93 Amendment 35, 74,

Containment Vent Isolation Instrumentation B 3.3.6 BASES APPLICABILITY While in MODES 5 and 6, the Containment Vent Isolation Instrumentation need (continued) not be OPERABLE since the potential for radioactive releases is minimized and operator action is sufficient to ensure post accident offsite doses are maintained within the limits of Reference 1.

ACTIONS The most common cause of channel inoperability is outright failure or drift sufficient to exceed the tolerance allowed by unit specific calibration procedures.

Typically, the drift is found to be small and results in a delay of actuation rather than a total loss of function. Ifthe Trip Setpoint is less conservative than the tolerance specified by the calibration procedure, the channel must be declared inoperable immediately and the appropriate Condition entered.

A Note has been added to the ACTIONS to clarify the application of Completion Time rules. The Conditions of this Specification may be entered independently for each Function listed in Table 3.3.6-1. The Completion Time(s) of the inoperable channel(s)/train(s) of a Function will be tracked separately for each Function starting from the time the Condition was entered for that Function.

A.1 Condition A applies to the failure of one containment purge isolation radiation monitor channel. Since the two containment radiation monitors are both gaseous detectors, failure of a single channel may result in loss of the redundancy.

Consequently, the failed channel must be restored to OPERABLE status. The 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months allowed to restore the affected channel is justified by the low likelihood of events occurring during this interval, and recognition that one or more of the remaining channels will respond to most events.

(continued)

Wafts Bar-Unit 1 B 3.3-157 Amendment

Containment Vent Isolation Instrumentation B 3.3.6 BASES ACTIONS B..1 (continued)

Condition B applies to all Containment Vent Isolation Functions and addresses the train orientation of the Solid State Protection System (SSPS) and the master and slave relays for these Functions. It also addresses the failure of multiple radiation monitoring channels, or the inability to restore a single failed channel to OPERABLE status in the time allowed for Required Action A.I.

If a train is inoperable, multiple channels are inoperable, or the Required Action and associated Completion Time of Condition A are not met, operation may continue as long as the Required Action for the applicable Conditions of LCO 3.6.3 is met for each valve made inoperable by failure of isolation instrumentation. A Note has been added above the Required Actions to allow one train of actuation logic to be placed in bypass and to delay entering the Required Actions for up to four hours to perform surveillance testing provided the other train is OPERABLE. The 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months allowance is consistent with the Required Actions for actuation logic trains in LCO 3.3.2, "Engineered Safety Features Actuation System Instrumentation" and allows periodic testing to be conducted while at power without causing an actual actuation. The delay for entering the Required Actions relieves the administrative burden of entering the Required Actions for isolation valves inoperable solely due to the performance of surveillance testing on the actuation logic and is acceptable based on the OPERABILITY of the opposite train.

Watts Bar-Unit 1 B 3.3-158 Amendment

Containment Vent Isolation Instrumentation B 3.3.6 BASES SURVEILLANCE A Note has been added to the SR Table to clarify that Table 3.3.6-1 determines REQUIREMENTS which SRs apply to which Containment Vent Isolation Functions.

SR 3.3.6.1 Performance of the CHANNEL CHECK once every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ensures that a gross failure of instrumentation has not occurred. A CHANNEL CHECK is normally a comparison of the parameter indicated on one channel to a similar parameter on other channels. It is based on the assumption that instrument channels monitoring the same parameter should read approximately the same value.

Significant deviations between the two instrument channels could be an indication of excessive instrument drift in one of the channels or of something even more serious. A CHANNEL CHECK will detect gross channel failure; thus, it is key to verifying the instrumentation continues to operate properly between each CHANNEL CALIBRATION.

(continued)

Watts Bar-Unit 1 B 3.3-159 Revision 45 Amendment 35,

ABGTS Actuation Instrumentation B 3.3.8 B 3.3 INSTRUMENTATION B 3.3.8 Auxiliary Building Gas Treatment (ABGTS) Actuation Instrumentation BASES BACKGROUND The ABGTS ensures that radioactive materials in the fuel building atmosphere following a loss of coolant accident (LOCA) are filtered and adsorbed prior to exhausting to the environment. The system is described in the Bases for LCO 3.7.12, "Auxiliary Building Gas Treatment System." The system initiates filtered exhaust of air from the fuel handling area, ECCS pump rooms, and penetration rooms automatically following receipt of a Containment Phase A Isolation signal. Initiation may also be performed manually as needed from the I main control room.

There are a total of two channels, one for each train. A Phase A isolation signal from the Engineered Safety Features Actuation System (ESFAS) initiates auxiliary building isolation and starts the ABGTS. These actions function to prevent exfiltration of contaminated air by initiating filtered ventilation, which imposes a negative pressure on the Auxiliary Building Secondary Containment Enclosure (ABSCE).

(continued)

Wafts Bar-Unit 1 B 3.3-171 Revision 87, 110,

ABGTS Actuation Instrumentation B 3.3.8 BASES (continued)

APPLICABLE The ABGTS ensures that radioactive materials in the ABSCE atmosphere SAFETY ANALYSES following a LOCA are filtered and adsorbed prior to being exhausted to the environment. This action reduces the radioactive content in the auxiliary building exhaust following a LOCA so that offsite doses remain within the limits specified in 10 CFR 100 (Ref. 1).

The ABGTS Actuation Instrumentation satisfies Criterion 3 of the NRC Policy Statement.

(continued)

Watts Bar-Unit 1 B 3.3-172 Revision 87, 110,

ABGTS Actuation Instrumentation B 3.3.8 BASES (continued)

LCO The LCO requirements ensure that instrumentation necessary to initiate the ABGTS is OPERABLE.

1. Manual Initiation The LCO requires two channels OPERABLE. The operator can initiate the ABGTS at any time by using either of two switches in the control room. This action will cause actuation of all, components in the same manner as any of the automatic actuation signals.

The LCO for Manual Initiation ensures the proper amount of redundancy is maintained in the manual actuation circuitry to ensure the operator has manual initiation capability.

Each channel consists of one hand switch and the interconnecting wiring to the actuation logic relays.

2. Deleted (continued)

Watts Bar-Unit 1 B 3.3-173 Amendment

ABGTS Actuation Instrumentation B 3.3.8 BASES LCO 3. Containment Phase A Isolation (continued)

Refer to LCO 3.3.2, Function 3.a, for all initiating Functions and requirements.

APPLICABILITY The manual ABGTS initiation must be OPERABLE in MODES 1, 2, 3, and 4 to ensure the ABGTS operates to remove fission products associated with leakage after a LOCA. The Phase A ABGTS Actuation is also required in MODES 1, 2, 3, and 4 to remove fission products caused by post LOCA Emergency Core Cooling Systems leakage.

While in MODES 5 and 6, the ABGTS instrumentation need not be OPERABLE.

See additional discussion in the Background and Applicable Safety Analysis sections.

ACTIONS The most common cause of channel inoperability is outright failure or drift sufficient to exceed the tolerance allowed by unit specific calibration procedures.

Typically, the drift is found to be small and results in a delay of actuation rather than a total loss of function. If the Trip Setpoint is less conservative than the tolerance specified by the calibration procedure, the channel must be declared inoperable immediately and the appropriate Condition entered.

A Note has been added to the ACTIONS to clarify the application of Completion Time rules. The Conditions of this Specification may be entered independently for each Function listed in Table 3.3.8-1 in the accompanying LCO. The Completion Time(s) of the inoperable channel(s)/train(s) of a Function will be tracked separately for each Function starting from the time the Condition was entered for that Function.

(continued)

Watts Bar-Unit 1 B 3.3-174 Revision 87

ABGTS Actuation Instrumentation B 3.3.8 BASES ACTIONS A.1 (continued)

Condition A applies to the actuation logic train function from the Phase A Isolation and the manual function. Condition A applies to the failure of a single actuation logic train or manual channel. If one channel or train is inoperable, a period of 7 days is allowed to restore it to OPERABLE status. If the train cannot be restored to OPERABLE status, one ABGTS train must be placed in operation.

This accomplishes the actuation instrumentation function and places the unit in a conservative mode of operation. The 7 day Completion Time is the same as is allowed if one train of the mechanical portion of the system is inoperable. The basis for this time is the same as that provided in LCO 3.7.12.

B.1.1, B.1.2, B.2 Condition B applies to the failure of two ABGTS actuation logic signals from the Phase A Isolation or two manual channels. The Required Action is to place one ABGTS train in operation immediately. This accomplishes the actuation instrumentation function that may have been lost and places the unit in a conservative mode of operation. The applicable Conditions and Required Actions of LCO 3.7.12 must also be entered for the ABGTS train made inoperable by the inoperable actuation instrumentation. This ensures appropriate limits are placed on train inoperability as discussed in the Bases for LCO 3.7.12.

Alternatively, both trains may be placed in the emergency radiation protection mode. This ensures the ABGTS Function is performed even in the presence of a single failure.

(continued)

Watts Bar-Unit 1 B 3.3-175 Amendment

ABGTS Actuation Instrumentation B 3.3.8 BASES ACTIONS C.land C.2 Condition C applies when the Required Action and associated Completion Time for Condition A or B have not been met. The plant must be brought to a MODE in which the LCO requirements are not applicable. To achieve this status, the plant must be brought to MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and MODE 5 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

SURVEILLANCE A Note has been added to the SR Table to clarify that Table 3.3.8-1 determines REQUIREMENTS which SRs apply to which ABGTS Actuation Functions.

SR 3.3.8.1 SR 3.3.8.1 is the performance of a TADOT. This test is a check of the manual actuation functions and is performed every 18 months. Each manual actuation function is tested up to, and including, the relay coils. In some instances, the test includes actuation of the end device (e.g., pump starts, valve cycles, etc.). The Frequency is based on operating experience and is consistent with the typical industry refueling cycle.

The SR is modified by a Note that excludes verification of setpoints during the TADOT. The Functions tested have no setpoints associated with them.

REFERENCES 1. Title 10, Code of Federal Regulations, Part 100.11, "Determination of Exclusion Area, Low Population Zone, and Population Center Distance."

Watts Bar-Unit 1 B 3.3-176 Amendment

Page Intentionally Left Blank Watts Bar-Unit 1 B 3.3-177 Amendment

Page Left Intentionally Blank Watts Bar-Unit 1 B 3.3-178 Amendment

ABGTS B 3.7.12 B 3.7 PLANT SYSTEMS B 3.7.12 Auxiliary Building Gas Treatment System (ABGTS)

BASES BACKGROUND The ABGTS filters airborne radioactive particulates from the area of active Unit 1 ECCS components and Unit 1 penetration rooms following a loss of coolant accident (LOCA).

The ABGTS consists of two independent and redundant trains. Each train consists of a heater, a prefilter, moisture separator, a high efficiency particulate air (HEPA) filter, two activated charcoal adsorber sections for removal of gaseous activity (principally iodines), and a fan. Ductwork, valves or dampers, and instrumentation also form part of the system. A second bank of HEPA filters follows the adsorber section to collect carbon fines and provide backup in case the main HEPA filter bank fails. The downstream HEPA filter is not credited in the analysis. The system initiates filtered ventilation of the Auxiliary Building Secondary Containment Enclosure (ABSCE) exhaust air following receipt of a Phase A containment isolation signal.

The ABGTS is a standby system, not used during normal plant operations.

During emergency operations, the ABSCE dampers are realigned and ABGTS fans are started to begin filtration. Air is exhausted from the Unit 1 ECCS pump rooms, Unit 1 penetration rooms, and fuel handling area through the filter trains.

The prefilters or moisture separators remove any large particles in the air, and any entrained water droplets present, to prevent excessive loading of the HEPA filters and charcoal adsorbers.

The ABGTS is discussed in the FSAR, Sections 6.5.1, 9.4.2, 15.0, and 6.2.3 (Refs. 1, 2, 3, and 4, respectively).

(continued)

Wafts Bar-Unit 1 B 3.7-62 Revision 87, 110,

ABGTS B 3.7.12 BASES (continued)

APPLICABLE The ABGTS design basis is established by the consequences of the limiting SAFETY ANALYSES Design Basis Accident (DBA), which is a LOCA. The analysis of the LOCA assumes that radioactive materials leaked from the Emergency Core Cooling System (ECCS) are filtered and adsorbed by the ABGTS. The DBA analysis assumes that only one train of the ABGTS is functional due to a single failure that disables the other train. The accident analysis accounts for the reduction in airborne radioactive material provided by the one remaining train of this filtration system. The amount of fission products available for release from the ABSCE is determined for a LOCA. The assumptions and analysis for a LOCA follow the guidance provided in Regulatory Guide 1.4 (Ref. 6).

The ABGTS satisfies Criterion 3 of the NRC Policy Statement.

(continued)

Wafts Bar-Unit 1 B 3.7-62A Revision 87, 110,

ABGTS B 3.7.12 BASES (continued)

LCO Two independent and redundant trains of the ABGTS are required to be OPERABLE to ensure that at least one train is available, assuming a single failure that disables the other train, coincident with a loss of offsite power. Total system failure could result in the atmospheric release from the ABSCE exceeding the 10 CFR 100 (Ref. 7) limits in the event of a LOCA.

The ABGTS is considered OPERABLE when the individual components necessary to control exposure in the fuel handling building are OPERABLE in both trains. An ABGTS train is considered OPERABLE when its associated:

a. Fan is OPERABLE;

b. HEPA filter and charcoal adsorber are not excessively restricting flow, and are capable of performing their filtration function; and

c. Heater, moisture separator, ductwork, valves, and dampers are OPERABLE, and air circulation can be maintained.

APPLICABILITY In MODE 1, 2, 3, or 4, the ABGTS is required to be OPERABLE to provide fission product removal associated with ECCS leaks due to a LOCA and leakage from containment and annulus.

In MODE 5 or 6, the ABGTS is not required to be OPERABLE since the ECCS is not required to be OPERABLE.

(continued)

Watts Bar-Unit 1 B 3.7-63 Revision 55, 87,

ABGTS B 3.7.12 BASES (continued)

ACTIONS A. 1 With one ABGTS train inoperable, action must be taken to restore OPERABLE status within 7 days. During this period, the remaining OPERABLE train is adequate to perform the ABGTS function. The 7 day Completion Time is based on the risk from an event occurring requiring the inoperable ABGTS train, and the remaining ABGTS train providing the required protection.

B.1 and B.2 When Required Action A.1 cannot be completed within the associated Completion Time, or when both ABGTS trains are inoperable, the plant must be placed in a MODE in which the LCO does not apply. To achieve this status, the plant must be placed in MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and in MODE 5 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

The Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

Watts Bar-Unit 1 B 3.7-64 Amendment

ABGTS B 3.7.12 BASES SURVEILLANCE SR 3.7.12.1 REQUIREMENTS Standby systems should be checked periodically to ensure that they function properly. As the environmental and normal operating conditions on this system are not severe, testing each train once every month provides an adequate check on this system.

Monthly heater operation dries out any moisture accumulated in the charcoal from humidity in the ambient air. The system must be operated for > 10 continuous hours with the heaters energized. The 31 day Frequency is based on the known reliability of the equipment and the two train redundancy available.

SR 3.7.12.2 This SR verifies that the required ABGTS testing is performed in accordance with the Ventilation Filter Testing Program (VFTP). The ABGTS filter tests are in accordance with Regulatory Guide 1.52 (Ref. 8). The VFTP includes testing HEPA filter performance, charcoal adsorber efficiency, minimum system flow rate, and the physical properties of the activated charcoal (general use and following specific operations). Specific test frequencies and additional information are discussed in detail in the VFTP.

(continued)

Watts Bar-Unit 1 B 3.7-65 Amendment

ABGTS B 3.7.12 BASES REFERENCES 5. Deleted.

(continued)

6. Regulatory Guide 1.4, "Assumptions Used for Evaluating the Potential Radiological Consequences of a Loss of Coolant Accident for Pressurized Water Reactors."

7. Title 10, Code of Federal Regulations, Part 100.11, "Determination of Exclusion Area, Low Population Zone, and Population Center Distance."

8. Regulatory Guide 1.52 (Rev. 2), "Design, Testing and Maintenance Criteria for Post Accident Engineered-Safety-Feature Atmospheric Cleanup System Air Filtration and Adsorption Units of Light-Water Cooled Nuclear Power Plants."

9. NUREG-0800, Section 6.5.1, "Standard Review Plan," Rev. 2, "ESF Atmosphere Cleanup System," July 1981.

10. Watts Bar Drawing 1-47W605-242, "Electrical Tech Spec Compliance Tables."

11. Deleted.

Watts Bar-Unit 1 B 3.7-67 Revision 29, 55

Fuel Storage Pool Water Level B 3.7.13 B 3.7 PLANT SYSTEMS B 3.7.13 Fuel Storage Pool Water Level BASES BACKGROUND The minimum water level in the fuel storage pool meets the assumptions of iodine decontamination factors following a fuel handling accident. The specified water level shields and minimizes the general area dose when the storage racks are filled to their maximum capacity. The water also provides shielding during the movement of spent fuel.

A general description of the fuel storage pool design is given in the FSAR, Section 9.1.2 (Ref. 1). A description of the Spent Fuel Pool Cooling and Cleanup System is given in the FSAR, Section 9.1.3 (Ref. 2). The assumptions of the fuel handling accident are given in the FSAR, Section 15.4.5 (Ref. 3).

APPLICABLE The minimum water level in the fuel storage pool meets the assumptions of the SAFETY fuel handling accident described in Regulatory Guide 1.183 (Ref. 6). The Total ANALYSES Effective Dose Equivalent (TEDE) for control room occupants, individuals at the exclusion area boundary, and individuals within the low population zone will remain within 10 CFR 50.67 (Ref. 7) and Regulatory Position C.4.4 of Regulatory Guide 1.183 (Ref. 6) for a fuel handling accident.

According to Reference 6, there is 23 ft of water between the top of the damaged fuel bundle and the fuel pool surface during a fuel handling accident. With 23 ft of water, the assumptions of Reference 6 can be used directly. In practice, this LCO preserves this assumption for the bulk of the fuel in the storage racks. In the case of a single bundle dropped and lying horizontally on top of the spent fuel racks, however, there may be < 23 ft of water above the top of the fuel bundle and the surface, indicated by the width of the bundle. To offset this small nonconservatism, the analysis assumes that all fuel rods fail, although analysis shows that only the first few rows fail from a hypothetical maximum drop.

The fuel storage pool water level satisfies Criterion 2 of the NRC Policy Statement.

(continued)

Watts Bar-Unit 1 B 3.7-68 Amendment

Fuel Storage Pool Water Level B 3.7.13 BASES SURVEILLANCE SR 3.7.13.1 (continued)

REQUIREMENTS pool is normally stable. Water level changes are controlled by plant procedures and are acceptable based on operating experience.

During refueling operations, the level in the fuel storage pool is in equilibrium with the refueling canal, and the level in the refueling canal is checked daily in accordance with SR 3.9.7.1.

REFERENCES 1. Watts Bar FSAR, Section 9.1.2, "Spent Fuel Storage."

2. Watts Bar FSAR, Section 9.1.3, "Spent Fuel Pool Cooling and Cleanup System."

3. Watts Bar FSAR, Section 15.4.5, "Fuel Handling Accident."

4. Deleted.

5. Deleted.

6. Regulatory Guide 1.183, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors," July 2000.

7. Title 10, Code of Federal Regulations, 10 CFR 50.67, "Accident Source Term."

Watts Bar-Unit 1 B 3.7-70 Amendment

Deleted B 3.9.4 B 3.9 REFUELING OPERATIONS B 3.9.4 Deleted This Page is Intentionally Left Blank.

Also, Bases Pages B 3.9-13 through B 3.9-16 Were Deleted.

Watts Bar-Unit 1 B 3.9-12 Revision 37 Amendment 26, 35,

Refueling Cavity Water Level B 3.9.7 B 3.9 REFUELING OPERATIONS B 3.9.7 Refueling Cavity Water Level BASES BACKGROUND The movement of irradiated fuel assemblies within containment requires a minimum water level of 23 ft above the top of the reactor vessel flange. During refueling, this maintains sufficient water level in the containment, refueling canal, fuel transfer canal, refueling cavity, and spent fuel pool. Sufficient water is necessary to retain iodine fission product activity in the water in the event of a fuel handling accident (Refs. 2 and 8). Sufficient iodine activity would be retained to limit offsite doses from the accident to the limits defined in 10 CFR 50.67 (Ref. 7) and Regulatory Position C.4.4 of Regulatory Guide 1.183 (Ref. 8).

APPLICABLE During movement of irradiated fuel assemblies, the water level in the refueling SAFETY ANALYSES canal and the refueling cavity is an initial condition design parameter in the analysis of a fuel handling accident in containment (Refs. 2 and 8). A minimum water level of 23 ft (Regulatory Position 2 of Appendix B to Regulatory Guide 1.183 (Ref. 8)) allows an overall iodine decontamination factor of 200 to be used in the accident analysis for iodine. This relates to the assumption that 99.5% of the total iodine released from the pellet to cladding gap of all the dropped fuel assembly rods is retained by the refueling cavity water. The fuel pellet to cladding gap is assumed to contain 8% of the 1-131, 10% of the Kr-85, and 5% of the other noble gases and iodides from the total fission product inventory in accordance with Regulatory Position 3.1 of Regulatory Guide 1.183 (Ref. 8).

The fuel handling accident analysis inside containment is described in Reference 2. With a minimum water level of 23 ft in conjunction with a minimum decay time of 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months prior to fuel handling without containment closure or Auxiliary Building isolation, the analysis and test programs demonstrate that the iodine release due to a postulated fuel handling accident is adequately captured by the water and offsite doses are maintained within allowable limits (Refs. 7 and 8).

Refueling cavity water level satisfies Criterion 2 of the NRC Policy Statement.

(Continued)

Watts Bar-Unit 1 B 3.9-25 Revision 45, 55 Amendment 35,

Refueling Cavity Water Level B 3.9.7 BASES (continued)

SURVEILLANCE SR 3.9.7.1 REQUIREMENTS Verification of a minimum water level of 23 ft above the top of the reactor vessel flange ensures that the design basis for the analysis of the postulated fuel handling accident during refueling operations is met. Water at the required level above the top of the reactor vessel flange limits the consequences of damaged fuel rods that are postulated to result from a fuel handling accident inside containment (Ref. 2).

The Frequency of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is based on engineering judgment and is considered adequate in view of the large volume of water and the normal procedural controls of valve positions, which make significant unplanned level changes unlikely.

REFERENCES 1. Deleted.

2. Watts Bar FSAR, Section 15.4.5, "Fuel Handling Accident."

3. NUREG-0800, "Standard Review Plan," Section 15.7.4, "Radiological Consequences of Fuel-Handling Accidents," U.S. Nuclear Regulatory Commission.

4. Title 10, Code of Federal Regulations, Part 20.1201(a), (a)(1), and(2)(2),

"Occupational Dose Limits for Adults."

5. Deleted.

6. Deleted.

7. Title 10, Code of Federal Regulations, 10 CFR 50.67, "Accident Source Term."

8. Regulatory Guide 1.183, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors," July 2000.

Watts Bar-Unit 1 B 3.9-27 Revision 55 Amendment

Deleted B 3.9.8 B 3.9 REFUELING OPERATIONS B 3.9.8 Deleted Bases Page B 3.9-29 Intentionally Left Blank Also, Bases Pages B 3.9-30 through B 3.9-32 Were Deleted Watts Bar-Unit 1 B 3.9-29 Amendment

Decay Time B 3.9.10 B 3.9 REFUELING OPERATIONS B 3.9. 10 Decay Time BASES BACKGROUND Section 15.5.6 of the WBN, Unit 1 UFSAR (Ref. 1) defines the assumptions of the fuel handling accident radiological analysis, including a minimum decay time for irradiated fuel assemblies prior to movement. This assumption ensures that the inventory of radioactive isotopes is at a level that supports the safety analysis assumptions.

To ensure that irradiated fuel assemblies have decayed for the appropriate period of time, a limitation is established to require the reactor core to be subcritical for a time period at least equivalent to the minimum decay time assumption in the fuel handling analysis prior to allowing irradiated fuel to be moved.

Given that no irradiated fuel assembly will be moved outside of the containment until the minimum decay time requirement is met, this requirement also ensures that any irradiated fuel assemblies that are moved outside of the containment meet the decay time assumption in the radiological analysis of the fuel handling accident.

APPLICABLE The radiological analysis of the fuel handling accident (Ref. 1) assumes a SAFETY ANALYSES minimum decay time prior to movement of irradiated fuel assemblies. The requirements of LCO 3.3.7, "Control Room Emergency Ventilation System (CREVS) Actuation Instrumentation," LCO 3.7.10, "Control Room Emergency Ventilation System (CREVS)," LCO 3.7.11, "Control Room Emergency Air Temperature Control System (CREATCS)," and LCO 3.9.7, "Refueling Cavity Water Level," in conjunction with a minimum decay time of 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months prior to irradiated fuel movement ensures that the release of fission product radioactivity, subsequent to a fuel handling accident, results in doses that are within the requirements of 10 CFR 50.67 (Ref. 2) and Regulatory Position C.4.4 of Regulatory Guide 1.183 (Ref. 3).

The decay time satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii).

Watts Bar-Unit 1 B 3.9-35 Amendment

Decay Time B 3.9.10 BASES LCO A minimum decay time of 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months is required prior to moving irradiated fuel assemblies within containment. This preserves an assumption in the fuel handling accident analysis (Ref. 1), and ensures that the radiological consequences of a postulated fuel handling accident inside containment are within acceptable limits.

APPLICABILITY This LCO applies during movement of irradiated fuel assemblies within the containment, since the potential for a release of fission products exist.

ACTIONS A..1 When the initial conditions for prevention of an accident cannot be met, steps should be taken to preclude the accident from occurring. When the reactor is subcritical for < 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months , movement of irradiated fuel assemblies within containment must be suspended. This action precludes the possibility of a fuel handling accident in containment. This action does not preclude moving a fuel assembly to a safe position.

The immediate Completion Time is consistent with the required times for actions to be performed without delay and in a controlled manner.

SURVEILLANCE SR 3.9.10.1 REQUIREMENTS This SR verifies that the reactor has been subcritical for at least 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months prior to moving irradiated fuel assemblies by confirming the date and time of subcriticality. This ensures that any irradiated fuel assemblies have decayed for at least 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months prior to movement. The Frequency of "Prior to movement of irradiated fuel in the containment" is appropriate, because it ensures that the decay time requirement has been met just prior to moving the irradiated fuel.

REFERENCES 1. Watts Bar UFSAR, Section 15.5.6, "Environmental Consequences of a Postulated Fuel Handling Accident."

2. Title 10, Code of Federal Regulations, 10 CFR 50.67, "Accident Source Term."

3. Regulatory Guide 1.183, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors," July 2000.

Watts Bar-Unit 1 B 3.9-36 Amendment

v t e Letter Sequence Request
TAC:ME8877, (Approved, Closed)
Initiation Request, Request Acceptance Supplement Administration Questions 20 December 2012 Answers 4 February 2013 Results Approval Other: ML12207A303
MONTHYEAR2012-08-02 [Table View]

ML12171A317

Text

Subject:

Enclosures:

Subject:

SUMMARY

SUMMARY

3.0 TECHNICAL EVALUATION

4.0 REGULATORY EVALUATION

5.0 ENVIRONMENTAL CONSIDERATION

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