ML13141A564

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Issuance of Amendment Regarding Application to Allow Selective Implementation of Alternative Source Term to Analyze the Dose Consequence Associating with Fuel-Handling Accidents
ML13141A564
Person / Time
Site: Watts Bar Tennessee Valley Authority icon.png
Issue date: 06/19/2013
From: Andrew Hon
Plant Licensing Branch II
To: James Shea
Tennessee Valley Authority
Hon A NRR/DORL/LPL2-2
References
TAC ME8877
Download: ML13141A564 (35)
v • d • e


Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 June 19, 2013 Mr. Joseph W. Shea Vice President, Nuclear Licensing Tennessee Valley Authority 3R Lookout Place 1101 Market Street, LP 3D-C Chattanooga, TN 37402-2801

SUBJECT:

WATTS BAR NUCLEAR PLANT, UNIT 1 - ISSUANCE OF AMENDMENT TO ALLOW SELECTIVE IMPLEMENTATION OF ALTERNATE SOURCE TERM TO ANALYZE THE DOSE CONSEQUENCES ASSOCIATED WITH FUEL-HANDLING ACCIDENTS (TAC NO. ME8877)

Dear Mr. Shea:

The U.S. Nuclear Regulatory Commission (NRC) has issued the enclosed Amendment No.92 to Facility Operating License No. NPF-90 for Watts Bar Nuclear Plant, Unit 1. This amendment consists of changes to the license and the Technical Specification (TS) in response to your application dated June 13, 2012, as supplemented by a letter dated February 4, 2013.

The proposed amendment will permit selective implementation of the Alternate Source Term methodology for the analysis of Fuel Handling Accidents (FHAs) and make TS changes consistent with the revised FHA dose analyses, in accordance with Title 10 of the Code of Federal Regulations (10 CFR) 50.67, "Accident source term," and Regulatory Guide 1.183, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors," Rev. 0, July 2000. The amendment includes the following changes to the TS:

  • TS 3.3.6, "Containment Vent Isolation Instrumentation"
  • TS 3.9.8, "Reactor Building Purge Air Cleanup Units"

J. Shea - 2 A copy of the related Safety Evaluation is also enclosed. The Notice of Issuance will be included in the Commission's biweekly Federal Register notice.

Sincerely, Andrew Hon, Project Manager Plant Licensing Branch 11-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-390

Enclosures:

1. Amendment No. 92 to NPF-90
2. Safety Evaluation cc w/encls: Distribution via Listserv

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 TENNESSEE VALLEY AUTHORITY DOCKET NO. 50-390 WATTS BAR NUCLEAR PLANT, UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 92 License No. NPF-90

1. The Nuclear Regulatory Commission (the Commission) has found that:

A. The application for amendment by the Tennessee Valley Authority (the licensee) dated June 13, 2012, as supplemented by letter dated February 4,2013, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in Title 10 of the Code of Federal Regulations (10 CFR) Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

- 2

2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. NPF-90 is hereby amended to read as follows:

(2) Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 92and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, are hereby incorporated into this license. TVA shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

3. This license amendment is effective as of the date of its issuance, and shall be implemented no later than 60 days from the date of its issuance.

FOR THE NUCLEAR REGULATORY COMMISSION

~ (j' ~ :fc{

Jessie F. Quichocho, Chief Plant Licensing Projects Branch 11-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Operating License No. NPF-90 and the Technical Specifications Date of Issuance: June 19, 2013

ATTACHMENT TO LICENSE AMENDMENT NO. 92 FACILITY OPERATING LICENSE NO. NPF-90 DOCKET NO. 50-390 Replace Page 3 of Operating License NPF-90 with the attached Page 3.

Replace the following pages of the Appendix A Technical Specifications with the attached revised pages. The revised pages are identified by amendment number and contain a marginal line indicating the area of change.

REMOVE INSERT 3.3-52 3.3-52 3.3-53 3.3-53 3.3-54 3.3-54 3.3-55 3.3-55 3.3-56 3.3-56 3.3-62 3.3-62 3.3-63 3.3-63 3.3-64 3.3-64 3.7-27 3.7-27 3.7-28 3.7-28 3.9-6 3.9-6 3.9-7 3.9-7 3.9-14 3.9-14 3.9-15 3.9-15 3.9-17 5.0-25 5.0-25

- 3 (4) TVA, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess, and use in amounts as required, any byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis, instrument calibration, or other activity associated with radioactive apparatus or components; and (5) TVA, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility.

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

(1) Maximum Power Level TVA is authorized to operate the facility at reactor core power levels not in excess of 3459 megawatts thermal.

(2) Technical Specifications and Environmental Protection Plan The Technical SpeCifications contained in Appendix A as revised through Amendment NO.92 and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, are hereby incorporated into this license. TVA shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

(3) Safety Parameter Display System (SPDS) (Section 18.2 of SER Supplements 5 and 15)

Prior to startup following the first refueling outage, TVA shall accomplish the necessary activities, provide acceptable responses, and implement all proposed corrective actions related to having the Watts Bar Unit 1 SPDS operational.

(4) Vehicle Bomb Control Program (Section 13.6.9 of SSER 20>

During the period of the exemption granted in paragraph 2.0.(3) of this license, in implementing the power ascension phase of the approved initial test program, TVA shall not exceed 50% power until the requirements of 10 CFR 73.55(c)(7) and (8) are fully implemented. TVA shall submit a letter under oath or affirmation when the requirements of 73.55(c)(7) and (8) have been fully implemented.

Amendment No.92

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

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

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 <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> channel inoperable. to OPERABLE status.

(continued)

Watts Bar-Unit 1 3.3-52 Amendment 35, 92

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 <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for Surveillance testing provided the other train is OPERABLE.

Two radiation monitoring channels inoperable. B.1 Enter applicable Conditions Immediately and Required Actions of 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 No. 92

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 <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />

--------~-------------. -----

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 instrunlt~lllduull.

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 Watts Bar-Unit 1 3.3-54 Amendment 17. 68 ,92

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 5 2.8E-02 !lCi/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, 92

Containment Vent Isolation Instrumentation 3.3.6 Page Intentionally Left Blank Watts Bar-Unit 1 3.3-56 Amendment No. 92

ABGTS Actuation Instrumentation 3.3.8 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME B. (continued) 8.2 Place both trains in Immediately emergency radiation protection mode.

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

Verification of setpoint is not required.

Perform TADOT. 18 months Watts Bar-Unit 1 3.3-62 Amendment No. 92

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 3.3.8.1 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.

Watts Bar-Unit 1 3.3-63 Amendment No. 92

ABGTS Actuation Instrumentation 3.3.8 Page Intentionally Left Blank Watts Bar-Unit 1 3.3-64 Amendment No. 92

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 CONDITION REQUIRED ACTION ON 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 <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and associated Completion Time of AND Condition A not met.

8.2 Be in MODE 5. 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> OR Two ABGTS trains inoperable.

Watts Bar-Unit 1 3.7-27 Amendment No. 92

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). theVFTP 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

5: 9900 cfm.

Watts Bar-Unit 1 3.7-28 Amendment No. 92

Deleted 3.9.4 3.9 REFUELING OPERATIONS 3.9.4 Deleted Watts Bar-Unit 1 3.9-6 Amendment 26, 35, 92

Deleted 3.9.4 Page Intentionally Left Blank Watts Bar-Unit 1 3.9-7 Amendment No. 92

Deleted 3.9.8 3.9 REFUELING OPERATIONS 3.9.8 Deleted Watts Bar-Unit 1 3.9-14 Amendment 35, 92

Deleted 3.9.8 Page Intentionally Left Blank Watts Bar-Unit 1 3.9-15 Amendment No. 92

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 <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br />.

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 <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br />. 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 Amendment No. 92

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 2. 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 O.

(continued)

Watts Bar-Unit 1 5.0-25 Amendment 5, 70, 78, 92

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 92 TO FACILITY OPERATING LICENSE NO. NPF-90 TENNESSEE VALLEY AUTHORITY WATTS BAR NUCLEAR PLANT, UNIT 1 DOCKET NO. 50-390

1.0 INTRODUCTION

By application dated June 13, 2012 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML12171A317), as supplemented by a letter dated February 4,2013 (ADAMS Accession No. ML13038A011), Tennessee Valley Authority (TVA) requested a license amendment request (LAR) for Watts Bar Nuclear Plant (WBN), Unit 1, Facility Operating License No. NPF-90. The proposed amendment will permit selective implementation of the Alternate Source Term (AST) methodology for the analysis of Fuel Handling Accidents (FHAs) and make Technical Specification (TS) changes consistent with the revised FHA dose analyses, in accordance with Title 10 of the Code of Federal Regulations (10 CFR) 50.67, "Accident source term," and Regulatory Guide (RG) 1.183, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors," Rev. 0, July 2000.

The supplemental letter dated February 4, 2013, provided additional information that clarified the application. It did not change the original proposed no significant hazard consideration determination as published in the Federal Register 77 FR 56882 on September 14, 2012.

2.0 REGULATORY EVALUATION

In the early 1970s, the Nuclear Regulatory Commission (NRC) staff issued regulatory guidance for evaluating the consequences of design basis accidents (DBAs) using the radiological source term described in Technical Information Document (TID)-14844, "Calculation of Distance Factors for Power and Test Reactor Sites." Since the publication of TID-14844, significant advances in understanding timing, magnitude, and chemical form of fission product releases from nuclear power plant accidents have occurred. In 1995, the NRC published NUREG-1465, "Accident Source Terms for Light-Water Nuclear Power Plants." NUREG-1465 used updated research to provide more realistic estimates of the accident source term that were physically based on, and could be applied to, the design of future light-water power reactors. In addition, the NRC determined that the analytical approach, based on the TID-14844 source term, would continue to be adequate to protect public health and safety for the current licensed power

-2 reactors. The NRC staff also determined that current licensees may wish to use the NUREG-1465 source term, referred to as the AST, in analyses to support cost-beneficial licensing actions. The NRC staff, therefore, initiated several actions to provide a regulatory basis for operating reactors to use an AST in design basis analyses. These initiatives resulted in the development and issuance of 10 CFR 50.67 and RG 1.183. The Issuance of RG 1.183 provided the first comprehensive guidance for analyzing DBAs for radiological consequences using the AST. A holder of an operating license issued prior to January 10, 1997, or a holder of a renewed license under 10 CFR Part 54, "Conditions of licenses," whose initial operating license was issued prior to January 10,1997, can, in accordance with 10 CFR 50.67, voluntarily revise the accident source term used in design basis radiological consequence analyses.

However, to ensure proper implementation of the AST, the NRC required, in 10 CFR 50.67(b),

that, "A licensee who seeks to revise its current accident source term in design basis radiological consequence analyses shall apply for a license amendment under 10 CFR 50.90.

The application shall contain an evaluation of the consequences of applicable DBAs previously analyzed in the safety analysis report."

In addition to developing the AST and providing regulatory guidance for its implementation, the NRC determined that new dose criteria for protection of public health and safety were appropriate and included these performance-based criteria in 10 CFR 50.67 as follows:

  • Paragraph (b)(2)(i) states, "An individual located at any point on the boundary of the exclusion area [EAB] for any 2-hour period following the onset of the postulated fission product release, would not receive a radiation dose in excess of 0.25 Sv [Sievert]

(25 rem [roentgen equivalent man]) total effective dose equivalent (TEDE)."

  • Paragraph (b)(2)(ii) states, ~An individual located at any point on the outer boundary of the low population zone [LPZ], who is exposed to the radioactive cloud resulting from the postulated fission product release (during the entire period of its passage), would not receive a radiation dose in excess of 0.25 Sv (25 rem) total effective dose equivalent (TEDE)."
  • Paragraph (b)(2)(iii) provides control room (CR) habitability criteria. 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."

The acceptance criteria in 10 CFR 50.67 is intended to apply to a design basis radiological consequence analysis based upon a major accident, or possible event, resulting in dose consequences not exceeded by those from any accident considered credible (maximum hypothetical accident). Unlike the design basis loss-of-coolant accident (LOCA), used to evaluate the emergency core cooling system (ECCS) requirements of 10 CFR 50.46, the general scenario used to postulate a maximum hypothetical dose consequence accident does not represent any specific accident sequence. Rather, the maximum hypothetical accident is intended to be a surrogate to enable a deterministic evaluation of the response of a facility's engineered safety features (ESFs) such as the primary containment system. Although the maximum hypothetical dose consequence LOCA is typically the maximum credible accident,

- 3 NRC staff experience in reviewing license applications has indicated the need to consider other accident sequences of possible occurrence including other dose consequence DBAs such as the FHA. These accident analyses are intentionally conservative in order to compensate for known uncertainties in accident progression, airborne activity product transport, and atmospheric dispersion.

RG 1.183 states in Regulatory Position 1.2 that a complete implementation of an AST would upgrade all existing radiological analyses. Although a complete re-assessment of all facility radiological analyses would be desirable, the NRC staff determined that recalculation of all design analyses for operating reactors would generally not be necessary. Full implementation is a modification of the facility design basis that addresses all characteristics of the AST:

composition and magnitude of the radioactive material, its chemical and physical form, and the timing of its release. Full implementation revises the facility's licensing basis to specify the AST in place of the previous accident source term and establishes new TEDE acceptance criteria.

This applies not only to the analyses performed in the application, which may only include a subset of the plant analyses, but also to all future design basis analyses. As a minimum for full implementation, the maximum credible dose consequence LOCA must be analyzed using the guidance in Appendix A of RG 1.183.

As stated in Regulatory Position 5.2 of RG 1.183, the DBAs addressed in the appendices of RG 1.183, other than Appendix A for LOCA dose consequence analysis where the source term is defined by regulation, were selected from accidents that may involve damage to irradiated fuel. The inclusion or exclusion of a particular dose consequence DBA in RG 1.183 should not be interpreted as indicating that an analysis of that DBA is required or not required. Licensees should analyze the DBAs that are affected by the specific proposed applications of an AST.

The licensee proposed to revise their current licensing basis (CLB) FHAs using the AST methodology described in RG 1.183. As such, this LAR is a selective implementation of the AST as described in Regulatory Position 1.2.2 of RG 1.183 which states that, "Selective implementation is a modification of the facility design basis that (1) is based on one or more of the characteristics of the AST or (2) entails re-evaluation of a limited subset of the design basis radiological analyses.

The regulatory requirements from which the NRC staff used in performing its review are the accident dose criteria in 10 CFR 50.67(b)(2), as supplemented in Regulatory Position 4.4 of RG 1.183 and Standard Review Plan (SRP) Section 15.0.1, "Radiological Consequence Analyses Using Alternative Source Terms." In addition, the NRC staffs evaluation is based upon the following regulatory codes, guides, and standards:

  • RG 1.183, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors," Rev. 0, July 2000.
  • RG 1.52, "Design, Inspection, and Testing Criteria for Air Filtration and Adsorption Units of Post-Accident Engineered-Safety-Feature Atmosphere Cleanup Systems in Light-Water-Cooled Nuclear Power Plants," Rev. 3, June 2001.

-4

  • RG 1.194, "Atmospheric Relative Concentrations for Control Room Radiological Habitability Assessments at Nuclear Power Plants," Rev. 0, June 2003.
  • NUREG-0800, "Standard Review Plan," Section 2.3.4, "Short-Term Diffusion Estimates for Accidental Atmospheric Releases," Rev. 3, March 2007.
  • NUREG/CR-6331, Revision 1, "Atmospheric Relative Concentrations in Building Wakes", May 1997.
  • 10 CFR 50, Appendix A, General Design Criterion (GDC) 19, "Control Room," which 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 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).)

3.0 TECHNICAL EVALUATION

In performing its technical and safety review, the NRC staff evaluated the licensee's FHA re-analysis for compliance with regulations, adherence to NRC acceptable accident consequence analysis assumptions and methods as described in the above applicable regulatory codes, guides, and standards, and approved precedents. The NRC staff also performed confirmatory accident dose calculations where appropriate. Finally, the NRC staff reviewed the licensee's current licensing and design basis, as described in its updated final safety analysis report (UFSAR) and TSs.

3.1 Radiological Evaluation 3.1.1 FHA Source Term Consistent with the current licensing basis (CLB), the licensee assumed a conservative estimate of 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> decay time before any movement of fuel. The core fission product inventory that constitutes the source term for this event is the gap activity in the 264 fuel rods assumed to be

- 5 damaged as a result of the postulated design basis FHA. Volatile constituents of the core fission product inventory migrate from the fuel pellets to the gap between the pellets and the fuel rod cladding during normal power operations.

Consistent with the CLB the proposed FHA analysis postulates that a spent fuel assembly is dropped during fuel handling operations damaging the cladding in all of the fuel rods and thereby releasing the radionuclide within the fuel rod gap to the spent fuel pool (SFP) or the reactor cavity water. The affected assembly is conservatively assumed to contain the maximum inventory of fission products. The fission product inventory in the fuel rod gap of the damaged fuel rods is assumed to be instantaneously released to the surrounding water as a result of the FHA. The licensee evaluated the fission product release from the breached fuel following the guidance of RG 1.183 Regulatory Position 3.2, Table 3, which specifies the fraction of fission product inventory assumed to be present in the fuel rod gap.

Fission products released from the damaged fuel are decontaminated by passage through the overlaying water in the reactor cavity or SFP, depending on their physical and chemical form.

Consistent with RG 1.183 guidance, the licensee assumed that the chemical form of radioiodine released from the fuel to the surrounding water would be 95 percent cesium iodide (Csl), 4.85 percent elemental iodine, and 0.15 percent organic iodide. The Csi released from the fuel is assumed to completely dissociate in the pool water. Because of the low pH of the pool water, the iodine re-evolves as elemental iodine. This process is assumed to occur instantaneously.

The NRC staff finds that the licensee followed the guidance in RG 1.183 in the evaluation of the fission product source term used in the FHA and therefore finds the evaluation to be acceptable.

Since WBN, Unit 1 is licensed to produce tritium; the licensee's FHA includes an analysis of the release of tritium. To conservatively bound the release of tritium in an FHA, the licensee assumed that 25 percent of the tritium released to the surrounding water is subsequently released to the environment during the two hour accident evaluation period. RG 1.183 does not include the release of tritium as part of the analysis of the FHA since standard plants are not licensed to produce tritium. The NRC staff notes that tritium released to either the cavity or SFP water from an FHA would be expected to largely be retained in the water. Releases of tritium would only be expected due to evaporation. Therefore, the NRC staff considers the licensee's evaluation of tritium releases during an FHA to be conservative and therefore acceptable for use in a bounding FHA dose analysis.

3.1.2 FHA Water Depth As corrected by item 8 of RIS 2006-04 (ADAMS Accession Number ML053460347), RG 1.183, Appendix B, Regulatory Position 2, should read as follows:

"If the depth of water above the damaged fuel is 23 feet or greater, the decontamination factors for the elemental and organic species are 285 and 1, respectively, giving an overall effective decontamination factor of 200 (Le., 99.5% of the total iodine released from the damaged rods is retained by the water). This difference in decontamination factors for elemental (99.85%) and organic iodine (0.15%) species results in the iodine above the water being composed of 70% elemental and 30% organic species."

- 6 Consistent with the CLB, the licensee assumed that a minimum water level of 23 feet above the damaged fuel assembly is maintained for release locations both inside the containment and the auxiliary building. This minimum water covering acts as a barrier to many of the radio nuclides released from the dropped assembly. Consistent with RG 1.183 guidance, the licensee assumed retention of all non-iodine particulates in the pool, while the iodine releases from the fuel gap into the pool are assumed to be decontaminated by an overall factor of 200. This OF results in 0.5% (Le., 99.5% of the iodine is retained in the pool) of the radioiodine escaping the overlying water with a composition of 70% elemental iodine and 30% organic iodide. In accordance with RG 1.183, Appendix B, Regulatory Position 3, the licensee assumed that 100%

of the noble gas activity is released from the water covering the fuel. The NRC staff finds that the licensee's assumptions regarding the retention of fission products in the water covering the fuel during an FHA follow the guidance in RG 1.183 and are therefore acceptable to the staff.

3.1.3 FHA within the Auxiliary Building The licensee evaluated the FHA within the auxiliary building without taking credit for filtration of the auxiliary building exhaust. Consistent with the guidance in RG 1.183, the licensee assumed that the release of all fission products to the environment 'from an FHA occurring within the auxiliary building occurs over a two hour period. Consistent with the guidance in RG 1.183, the licensee did not credit holdup or dilution of the released activity within the auxiliary building. The NRC staff finds that the licensee's assumptions regarding the release of fission products from an FHA within the auxiliary building to be consistent with the guidance in RG 1.183 and therefore acceptable.

3.1.4 FHA within the Containment For the FHA occurring inside containment, the licensee assumed that the equipment maintenance hatch is open at the time of the accident and that the release from the containment occurs with no credit taken for containment isolation and no credit for filtration of the released effluent. Consistent with the guidance in RG 1.183, the licensee assumed that the release of all fission products to the environment from an FHA occurring within the containment occurs over a two hour period. Consistent with the guidance in RG 1.183 the licensee did not credit holdup or dilution of the released activity within the containment. The NRC staff finds that the licensee's assumptions regarding the release of fission products from an FHA within the containment to be consistent with the guidance in RG 1.183 and therefore acceptable.

3.1.5 FHA CR Dose Analysis The licensee evaluated the dose to the control room (CR) occupants due to the radioactive release associated with the FHA considering the time dependent concentration of airborne activity in the CR and the direct shine dose accounting for CR shielding. The licensee only credited one train of the CR emergency ventilation system (CREVS). Prior to CR isolation, which is initiated by redundant CR intake monitors, the CR intake flow rate is assumed to be 3,200 cubic feet per minute (cfm). After CR isolation, the total flow rate into the CR is assumed to be 3,600 cfm. This consists of a filtered pressurization flow rate of 711 cfm, a filtered recirculation flow rate of 2,889 cfm and an additional assumed unfiltered inleakage of 51 cfm.

The modeling assumptions for the flow rates and filter efficiencies associated with CREVS operation are consistent with the CLB. The licensee conservatively increased the CR isolation

-7 time from the CLB value of 20.6 seconds to 40 seconds. The NRC staff finds that the assumptions used in the FHA CR dose analysis are conservative relative to the CLB assumptions and therefore acceptable.

3.2 Atmospheric Dispersion Estimates 3.2.1 Meteorological Data The licensee used 20 years of onsite meteorological data collected from calendar years 1991 through 2010 to develop the updated CR atmospheric dispersion factors (X/a values) used in the FHA AST dose assessment. These data, in the form of hourly meteorological data files in the ARCON96 atmospheric dispersion computer code input format (NUREG/CR-6331), were provided as part of the June 13, 2012, LAR. The data were also provided for a previous licensing action, were reviewed by the NRC, and are discussed in the safety evaluation (SE) associated with WBN, Unit 1, Amendment No. 91, dated December 5, 2012 (ADAMS Accession Number ML12279A115). For calculation of doses at the EAB and LPZ, the licensee used current licensing basis X/a values which were based on the 1991 through 2010 meteorological data, in joint wind direction, wind speed, and atmospheric stability distribution format, which are also discussed in the staffs SE for Amendment No. 91.

On the basis of the prior review and approval conducted in support of Amendment No. 91, the NRC staff has concluded that the 1991 through 2010 data files provided by the licensee give an adequate representation of the site conditions to facilitate calculation of the CR, EAB, and LPZ X/a values for the FHA AST dose assessment for WBN, Unit 1.

3.2.2 CR Atmospheric Dispersion Factors The licensee generated revised CR X/a values for the current LAR using the ARCON96 methodology and provided the output files as part of the June 13, 2012, LAR. As asserted by the licensee, the CR X/a values were calculated consistent with the CLB methodology, except the meteorological data were updated to reflect a more recent 20-year time period, 1991 through 2010, rather than utilizing the older period of data. The licensee postulated six source/receptor pairs and selected the X/a values associated with the limiting pairs for use in the CR FHA AST dose assessment. Releases were assumed to occur from the Unit 1 Shield Building vent, Unit 2 Shield Building vent, and Auxiliary Building vent to the normal and emergency CR air intakes.

RG 1.194 states that ARCON96 is an acceptable methodology for assessing CR X/a values for use in DBA radiological analyses. NRC staff evaluated the applicability of the ARCON96 model and concluded that there were no unusual siting, building arrangement, release characterization, source-receptor configuration, meteorological regime, or terrain conditions that precluded use of this model in support of the WBN, Unit 1 LAR. The NRC staff qualitatively reviewed the licensee's inputs to the ARCON96 computer runs and found them adequately consistent with site configuration drawings and staff practices. NRC staff noted that the licensee used the ARCON96 default surface roughness length and averaging sector width constant values presented in NUREG/CR-6331, Revision 1, rather than the default values listed in RG 1.194. NRC staff used ARCON96 and the RG 1.194 default values to generate X/a values to compare with the X/a values calculated by the licensee. Because the resultant values

- 8 were similar to those calculated by the licensee, NRC staff has concluded that the CR X/O values identified by the licensee as listed in Table 1 below are acceptable for use in the FHA AST dose assessment associated with the WBN, Unit 1, LAR.

3.2.3 Offsite Atmospheric Dispersion Factors The licensee used previously approved EAB and LPZ X/O values which were generated using the 1991 through 2010 meteorological data. On the basis of the previous review discussed in the SE associated with Amendment No. 91, the NRC staff has concluded that the EAB and LPZ X/O values identified by the licensee and presented in Table 1 are acceptable for use in the WBN, Unit 1, FHA AST dose assessment.

3.3 Evaluation of Proposed TS Changes The licensee has proposed several TS changes to ensure consistency with the revised AST FHA analysis for refueling operations only. The NRC staff has reviewed the following proposed TS changes and found that they accurately reflect the conditions assumed in the revised AST analysis:

  • The licensee proposed to modify 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. This change is appropriate since the revised FHA analysis does not credit the ABGTS.

  • The licensee proposed to eliminate TS 3.9.4, "Containment Penetrations," and TS 3.9.8, "Reactor Building Purge Air Cleanup Units." These TS are no longer necessary since the revised FHA does not require containment isolation or credit the reactor building air cleanup units.
  • The licensee proposed to eliminate TS 3.9.8, "Reactor Building Purge Air Cleanup Units." The reactor building air cleanup units are ESF passive components and are part of the non-safety related reactor building purge ventilation system (RBPVS). The RBPVS provides ventilation for personnel to perform work in the primary containment and the annulus during all normal operations. In the event of a FHA, the RBPVS is isolated. In the current licensing basis, immediately after a FHA, the reactor building air cleanup units are required to always be operable to perform their safety function by filtering the exhaust air to limit the offsite dose. The licensee proposes to delete TS 3.9.8, in its entirety, because no credit is taken for the operation of reactor building air cleanup units for the dose analysis during a FHA.
  • The licensee proposed to add TS 3.9.10, "Decay Time," to restrict movement of irradiated fuel assemblies until 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> after the reactor core has become sub-critical.

The added TS 3.9.10 will ensure that the irradiated fuel meets the minimum decay time established in the revised radiological analysis of the FHA.

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3.4 Summary The licensee evaluated an FHA in the SFP area and in an open containment with no credit taken for the ABGTS or containment purge system filters and concluded that the radiological consequences at the EAB, the LPZ and the CR comply with the reference values provided in 10 CFR 50.67 and the accident-specific dose guidelines specified in the SRP and RG 1.183.

The NRC staff's review has found that the licensee used analyses, assumptions, and inputs consistent with applicable regulatory guidance identified in Section 2.0 of this SE. The licensee's calculated dose results are given in Table 2 and Table 3 of this SE. The assumptions found acceptable to the NRC staff are presented in Table 4. The NRC staff finds, with reasonable assurance, that the licensee's estimates of the dose consequences of a design basis FHA will comply with the requirements of 10 CFR 50.67 and the guidance found in RG 1.183, and are therefore acceptable.

4.0 STATE CONSULTATION

In accordance with the Commission's regulations, the Tennessee State official was notified of the proposed issuance of the amendments. The State official had no comments.

5.0 ENVIRONMENTAL CONSIDERATION

The amendment changes a requirement with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20 and changes surveillance requirements. The NRC staff has determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendment involves no significant hazards consideration, and there has been no public comment on such finding (77 FR 56882). Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b) no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendment.

-10

6.0 CONCLUSION

The Commission has concluded, based on the considerations discussed above, that: (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 amendments will not be inimical to the common defense and security or to the health and safety of the public.

Contributors: L. Brown, NRR J. Parillo, NRR A.Saliman, NRR Date: June 19, 2013

- 11 Table 1 - Offsite and Onsite Atmospheric Dispersion Factors (X/Qs)

Receptor Location Time Period X/Q (sec/m3)

Exclusion Area Boundary (EAB) 0-2 hours 6.382E-4 i Outer Boundary of the Low Population Zone (LPZ) 0-2 hours 1.784E-4

! Unit 1 Control Room (CR)

Auxiliary Building Vent 0-2 hours 2.56E-03 Shield Building Vent 0-2 hours 1.09E-03 Table 2 - Radiological Consequences (rem) for FHA in the Auxiliary Building Conventional TPC Once TPC Twice TPC Thrice Acceptance Location Core Burned Burned Burned Criteria I

CR 1.015E+00 2.869E+00 2.602E+00 1.136E+00 5 EAB 2.383E+00 2.834E+00 2.268E+00 2.650E+00 6.3 LPZ 6.660E-01 7.923E-01 6.339E-01 7.407E-01 6.3 Table 3 - Radiological Consequences (rem) for FHA in the Containment Conventional TPC Once TPC Twice TPC Thrice Acceptance Location Core Burned Burned Burned Criteria CR 1.000E+00 2.277E+00 2.014E+00 1.119E+00 5 EAB 2.383E+00 2.834E+00 2.268E+00 2.650E+00 6.3 LPZ 6.660E-01 7.923E-01 6.339E-01 7.407E-01 6.3

- 12 Table 4 - Data and Assumptions for the FHA Core thermal power level 3459 MWt Radial peaking factor 1.65 Number of fuel assemblies in the core 193 Fuel rods per assembly 264 Core average assembly power 18.47 MWth Number of fuel assemblies damaged 1 (all rods ruptured)

Minimum post shutdown fuel handling time (decay time) 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> Minimum pool water depth 23 feet Fuel clad damage gap release fractions 1-131 8%

Remainder of halogens 5%

Kr-85 10%

Remainder of noble gases 5%

Pool OF Noble gases and organic iodine 1 Aerosols Infinite Elemental iodine (23 ft of water cover) 285 Overall iodine (23 ft of water cover) 200 (effective OF)

Chemical form of iodine released Elemental 99.85%

Organic 0.15%

Duration of release to the environment 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> release CR volume 257,198 ft3 CR intake flow prior to isolation 3,200 cfm CR pressurization intake flow 711 cfm CR unfiltered in leakage 51 cfm CR filter efficiency 95% first pass 70% second pass

J.Shea -2 A copy of the related Safety Evaluation is also enclosed. The Notice of Issuance will be included in the Commission's biweekly Federal Register notice.

Sincerely, IRA!

Andrew Hon, Project Manager Plant Licensing Branch 11-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-390

Enclosures:

1. Amendment No. 92 to NPF-90
2. Safety Evaluation cc wtencls: Distribution via Listserv DISTRIBUTION:

PUBLIC LPL2-2 rtf RidsN rrDssStsb RidsNrrDorlDpr RidsNrrDorlLpl2-2 RidsNrrPMWattsBar1 RidsNrrLABClayton RidsOgcMailCenter Resource RidsRgn2MailCenter RidsAcrsAcnw_MaiICTR RidsNrrDraAadb ASaliman, NRR

~IParillo, NRR LBrown, NRR ADAM SAccesslon N0.: ML13141A564 *b e-mal*1 OFFICE NRR/LPL2-2/PM NRR/LPL2-2/LA NRR/AADB/B NRRISCVB/B NRRlSTSB/B BClayton NAME AHon TTate* RDennig RElliotl*

(SFigueroa for)

DATE 5/23/13 5/23/13 6/6113 6/12/13 5/23/13 OFFICE NRR/AHPB/B OGC - NLO NRR/LPB2-/BC NRR/LP2-2/PM JQuichocho NAME UShoop* SUtlal AHon (SLingam for)

DATE 5/17/13 6/17/13 06/19/13 6119113 OFFICIAL AGENCY RECORD

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