ML12279A115

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Issuance of Amendment Regarding Technical Specifications Changes in Dose Equivalent 1-131 Spike Limit and Allowable Value for Control Room Air Intake Radiation Monitors
ML12279A115
Person / Time
Site: Watts Bar Tennessee Valley Authority icon.png
Issue date: 12/05/2012
From: Andrew Hon, Jessie Quichocho
Plant Licensing Branch II
To: James Shea
Tennessee Valley Authority
Hon, A
References
TAC ME8156
Download: ML12279A115 (19)
v • d • e


Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 December 5, 2012 Mr. Joseph W. Shea Corporate Manager, Nuclear Licensing Tennessee Valley Authority 3R Lookout Place 1101 Market Street Chattanooga, TN 37402-2801

SUBJECT:

WATIS BAR NUCLEAR PLANT, UNIT 1 - ISSUANCE OF AMENDMENT REGARDING TECHNICAL SPECIFICATION CHANGES IN DOSE EQUIVALENT 1-131 SPIKE LIMIT AND ALLOWABLE VALUE FOR CONTROL ROOM AIR INTAKE RADIATION MONITORS (TAC NO. ME8156)

Dear Mr. Shea:

The U.S. Nuclear Regulatory Commission (NRC) has issued the enclosed Amendment No. 91 to Facility Operating License No. NPF-90 for Watts Bar Nuclear Plant, Unit 1. This amendment consists of changes to the license in response to your application dated March 8, 2012, as supplemented by a letter dated July 18, 2012.

The amendment revises: (1) Technical Specification (TS) 3.3.7, "Control Room Emergency Ventilation System (CREVS) Actuation Instrumentation," by changing the Allowable Value for the main control room air intake radiation monitoring instrumentation in Table 3.3.7-1 from less than or equal to (S) 9.45E-05 micro-Curie per cubic centimeter (j.lCi/cc) (3,308 counts per minute (cpm>> to S 1.647E-04 j.lCi/cc (3,308 cpm); and (2) TS 3.4.16, "RCS [Reactor Coolant System] Specific Activity," by lowering the DOSE EQUIVALENT iodine 131 spike limit from 21 micro-Curie/gram (j.lCi/gm) to 14 j.lCi/gm in Required Action A.1 and Condition C.

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,. ,Ii J

~p'~ i-f---

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. 91 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. 91 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 March 8,2012, as supplemented July 18, 2012, 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 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 (0 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.

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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. 91 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. 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

Attachment:

Changes to the Operating License No. NPF-90 and the Technical Specifications Date of Issuance: December 5, 2012

ATTACHMENT TO LICENSE AMENDMENT NO. 91 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 3.3-60 3.3-60 3.4-39 3.4-39 3.4-40 3.4-40

- 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. 91 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. 91

CREVS Actuation Instrumentation 3.3.7 Table 3.3.7-1 (page 1 of 1)

CREVS Actuation Instrumentation FUNCTION REQUIRED SURVEILLANCE ALLOWABLE VALUE CHANNELS REQUIREMENTS

1. Manual Initiation 2 trains SR 3.3.7.3 NA
2. Control Room Radiation 2 SR 3.3.7.1 s1.647E-04 IJC/cc Control Room Air Intakes SR 3.3.7.2 (3,308 cpm)

SR 3.3.7.4

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

Watts Bar-Unit 1 3.3-60 Amendment 41, 91

RCS Specific Activity 3.4.16 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.16 RCS Specific Activity LCO 3.4.16 The specific activity of the reactor coolant shall be within limits.

APPLICABILITY: MODES 1 and 2, MODE 3 with RCS average temperature (Tavg) 2500°F.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. DOSE EQUIVALENT 1-131 ----------------------NOTE ---------------***-

> 0.265 IlCi/gm. LCO 3.0.4.c is applicable.

A.1 Verify DOSE EQUIVALENT 1-131 Once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />

~ 14 IlCilgm.

AND A.2 Restore DOSE EQUIVALENT 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> 1-131 to within limit.

B. Gross specific activity of the B.1 Perform SR 3.4.16.2. 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> reactor coolant not within limit.

AND B.2 Be in MODE 3 with Tavg < 500°F. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> (continued)

Watts Bar-Unit 1 3.4-39 Amendment 41, 55,91

RCS Specific Activity 3.4.16 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME C. Required Action and C.1 Be in MODE 3 with 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time Tav9 < 500°F.

of Condition A not met.

OR DOSE EQUIVALENT 1-131

> 14 IlCi/gm.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.16.1 Verify reactor coolant gross specific 7 days activity.:s 1001 E IlCi/gm.

SR 3.4.16.2 -----------------------------N0 TE---------------------------------

Only required to be performed in MODE 1.

Verify reactor coolant DOSE EQUIVALENT 1-131 specific 14 days activity:o; 0.265 IlCilgm.

Between 2 and 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> after a THERMAL POWER change of

~ 15% RTP within a 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> period (continued)

Watts Bar-Unit 1 3.4-40 Amendment 41, 91

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 91 TO FACILITY OPERATING LICENSE NO. NPF-90 TENNESSEE VALLEY AUTHORITY WATIS BAR NUCLEAR PLANT, UNIT 1 DOCKET NO. SO-390

1.0 INTRODUCTION

By application dated March 8, 2012 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML 12072A20S), as supplemented by a letter dated July 18, 2012 (ADAMS Accession No. ML 1220SA002), Tennessee Valley Authority (TVA) or (the licensee) requested a license amendment for Watts Bar Nuclear Plant (WBN) Unit 1, Facility Operating License No. NPF-90. The proposed amendment will revise:

1) Technical Specification (TS) 3.3.7, "Control Room Emergency Ventilation System (CREVS) Actuation Instrumentation," by changing the Allowable Value for the main control room air intake radiation monitoring instrumentation in Table 3.3.7-1 from less than or equal to (S) 9.4SE-OS micro-Curie per cubic centimeter (~Ci/cc) (3,308 counts per minute (cpm>> to S 1.647E-04 ~Ci/cc (3,308 cpm); and
2) TS 3.4.16, "RCS [Reactor Coolant System] Specific Activity," by lowering the DOSE EQUIVALENT iodine 131 (DEI) spike limit from 21 micro-Curie/gram (~Ci/gm) to 14

~Ci/gm in Required Action A.1 and Condition C.

TVA proposed to lower the above DEI spike limit because TVA determined the current value in TS 3.4.16 may not assure safety under certain postulated accident conditions. Until TVA implements this, Nuclear Regulatory Commission (NRC) approved License Amendment that lowers the DEI spike limit from 21 ~Ci/gm to 14 ~Ci/gm, the new numerical limit for DEI will be controlled administratively in accordance with NRC Administrative Letter 98-10, "Dispositioning of Technical Specifications That Are Insufficient to Assure Plant Safety."

The supplement dated July 18, 2012, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the NRC staff's original proposed no significant hazards consideration determination as published in the Federal Register on May 1S, 2012 (77 FR 28633).

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2.0 REGULATORY EVALUATION

2.1 Regulations and Guidance The NRC staff reviewed the licensee's evaluation of the radiological consequences of design basis accidents (DBAs) against the requirements specified in Title 10 of the Code of Federal Regulations (10 CFR), Section 100.11, "Determination of exclusion area, low population zone, and population center distance." Section 100.11 of 10 CFR, requires that the licensee determine:

  • An exclusion area of such size that an individual located at any point on its boundary for two hours immediately following onset of the postulated fission product release would not receive a total radiation dose to the whole body in excess of 25 roentgen equivalent man (rem) or a total radiation dose in excess of 300 rem to the thyroid from iodine exposure.
  • A low population zone of such size that an individual located at any point on its outer boundary 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 total radiation dose to the whole body in excess of 25 rem or a total radiation dose in excess of 300 rem to the thyroid from iodine exposure.

In addition the NRC staff reviewed the licensee's evaluation of the radiological consequences of DBAs against the requirements specified Appendix A to 10 CFR Part 50, "General Design Criteria [GDC] for Nuclear Power Plants," Criterion 19, "Control room," which states that:

A control room shall be provided from which actions can be taken to operate the nuclear power unit safely under normal conditions and to maintain it in a safe condition under accident conditions, including loss-of-coolant accidents.

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.

This safety evaluation (SE) addresses the impact of the proposed changes on previously analyzed DBA radiological consequences and the acceptability of the revised analysis results.

The regulatory requirements from which the NRC staff based its acceptance are the reference values in 10 CFR 100.11 and GDC 19. In addition, the NRC staff used the regulatory guidance provided in:

  • NUREG-0800, "Standard Review Plan (SRP) for the Review of Safety Analysis Reports for Nuclear Power Plants," Section 2.3.4, "Short-Term Atmospheric Dispersion Estimates for Accident Releases," Revision 3.
  • NUREG-0800, SRP Section 15.1.5, "Steam System Piping Failures Inside and Outside of Containment (PWR)," Appendix A, "Radiological Consequences of Main Steam Line Failures Outside Containment," Revision 2.

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  • NUREG-0917, "Nuclear Regulatory Commission Staff Computer Programs for Use with Meteorological Data," July 1982.
  • NUREG/CR-6331, Revision 1, "Atmospheric Relative Concentrations in Building Wakes," May 1997.
  • NUREG/CR-2858, "PAVAN: An Atmospheric Dispersion Program for Evaluating Design Basis Accidental Releases of Radiological Materials from Nuclear Power Stations,"

November 1982.

  • RG 1.23, Revision 1, "Meteorological Monitoring Programs for Nuclear Power Plants,"

March 2007.

  • RG 1.145, "Atmospheric Dispersion Models for Potential Accident Consequence Assessments at Nuclear Power Plants," February 1983.
  • RG 1.194, "Atmospheric Relative Concentrations for Control Room Radiological Habitability Assessments at Nuclear Power Plants," June 2003.
  • RG 1.195, "Methods and Assumptions for Evaluating Radiological Consequences of Design Basis Accidents at Light Water Nuclear Power Reactors," May 2003.

2.2 Precedents TVA requested a license amendment on January 14, 2002 (ADAMS No. ML020170416) to change WBN, Unit 1 TSs. It requested that the value of DEI-131 be lowered from 60 IJCi/gm to 21 IJCi/gm DEI-131, and it modified the value for the Allowable Value for the main control room air intake radiation monitors. The NRC approved the requested changes with the issuance of Amendment 41 to WBN, Unit 1 Operating License on November 18, 2002 (ADAMS No. ML023240483). The staff did not consider this precedent as it has no impact on the SE.

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3.0 TECHNICAL EVALUATION

3.1 Changes to TS 3.3.7, "CREVS Actuation Instrumentation" The actuation instrumentation for the CREVS consists of redundant radiation monitors. A high radiation signal from any detector will initiate its associated train of the CREVS. On initiation of the CREVS, the supply of unfiltered outside air to the control room will be terminated and the emergency filtration and pressurization of the control room will be initiated. These actions are necessary to ensure the control room is kept habitable for the operators stationed there during accident recovery and post-accident operations by minimizing the radiation exposure of control room personnel.

The licensee has proposed to change the Allowable Value for the main control room (CR) air intake radiation monitors in TS Table 3.3.7-1. The licensee states that this change is due to correcting an error in the sensitivity used to calculate the Allowable Value as measured in IJCi/cc. The original sensitivity value for Xenon 133 (3.SE+7 cpm/lJCi/cc) was based on a graphical extrapolation from a vendor test report. A more recent test report from the same vendor included a higher quality graph and a vendor established value for the same point of 2.27E+7 cpm/lJCi/cc. As a result of the correction in sensitivity the Allowable Value corresponding to 3,308 cpm will change from s 9.4SE-OS IJCi/cc to s 1.647E-04 IJCi/cc. The Allowable Value expressed in cpm presented in TS Table 3.3.7-1 remains unchanged at 3,308 cpm. The licensee states that the margin to the Analytical Limit for the previously implemented setpoint was 2.48E-06 IJCi/cc and that the margin to the Analytical Limit for the new setpoint is 2.S4E-06 IJCi/cc. The NRC staff finds that this change is acceptable from an accident dose consequence perspective since the margin to the Analytical Limit for the resultant setpoint remains essentially unchanged and therefore this change will have no impact on the dose consequence analyses.

3.2 Changes to TS 3.4.16, "RCS Specific Activity" 3.2.1 Applicable Accident Analyses The NRC staff evaluated the impact of the proposed changes as they relate to the radiological consequences of affected DBAs that use the RCS inventory as the source term. As stated in RG 1.19S, the source term assumed in radiological analyses should be based on the activity associated with the projected fuel damage or the maximum RCS TS values, whichever maximizes the radiological consequences. The limits on RCS specific activity ensure that the offsite doses are appropriately limited for accidents that are based on releases from the RCS with no significant amount of fuel damage.

The licensee's current licensing basis calculations demonstrate that no fuel damage is postulated for both the main steam line break (MSLB) accident and the steam generator tube rupture (SGTR) accident. Therefore the activity released is based on the maximum coolant activity allowed by TS. In performing the dose consequence analyses for a coolant release at maximum TS values two radioiodine spiking cases are considered. The first case is referred to as a pre-accident iodine spike and assumes that a reactor transient has occurred prior to the postulated accident that has raised the primary coolant iodine concentration to the maximum value permitted by the TS for a spiking condition. The second case assumes that the primary

-5 system transient associated with the accident causes an iodine spike in the primary system.

This case is referred to as an accident-induced spike or a concurrent iodine spike. Initially, the plant is assumed to be operating with the RCS iodine activity at the TS limit for normal operation. The increase in primary coolant iodine concentration for the concurrent iodine spike case is estimated using a spiking model that assumes that as a result of the accident, iodine is released from the fuel rods to the primary coolant at a rate that is much greater than the iodine equilibrium release rate corresponding to the iodine concentration at the TS limit for normal operation. The iodine release rate at equilibrium is equal to the rate at which iodine is lost due to radioactive decay, RCS purification, and RCS leakage. The iodine release rate is also referred to as the iodine appearance rate.

3.2.2 Impact of TS change on the Accident Analyses The accident analyses for a MSLB and a SGTR, including the assumed parameters, are presented in Sections 15.4.2 and 15.4.3 of the WBN Unit 1, Updated Final Safety Analysis Report (UFSAR), respectively. The licensee has not proposed any changes to the operator actions and the thermal and hydraulic analysis previously performed to determine plant response for a postulated MSLB and a postulated SGTR accident as described in the UFSAR.

Based on the findings of a recent Problem Evaluation Report, which documented a discrepancy in the use of dose conversion factors (DCFs), the licensee has proposed to lower the pre-accident DEI spike limit 'from 21 jJCi/gm to 14 jJCi/gm in order to ensure that calculated doses from postulated MSLB and SGTR accidents are within the acceptance criteria specified in SRP Sections 15.1.5 and 15.6.3. This discrepancy is a result of the fact that for a given mix of iodine isotopes (iodine 131 through iodine 135), the use of the DCFs from the International Commission on Radiological Protection (ICRP) Publication ICRP 2 will result in a higher DEI value than when updated DCFs from either RG 1.109 or ICRP 30 are used. Currently, the proposed pre-accident iodine DEI spike limit of 14 jJCi/gm is being controlled administratively.

This change will result in a more restrictive condition for plant operations. The NRC staff finds this change to be conservative and, therefore, acceptable.

3.3 Atmospheric Dispersion Estimates TVA used new WBN Unit 1, CR, exclusion area boundary (EAB), and low population zone (LPZ) atmospheric dispersion factors (X/O values) in the MSLB and SGTR dose assessments. In the March 8,2012, license amendment request (LAR), TVA affirmed that these X/O values were calculated consiste,nt with the current licensing basis methodology, except the meteorological data were updated'to reflect a more recent 20-yeartime period, 1991 through 2010. NRC staff notes that these calculations are also consistent with prior calculations for WBN Unit 2, as discussed in NUREG-0847, Supplement 25, "Safety Evaluation Report Related to the Operation of Watts Bar Nuclear Plant, Unit 2" (SSER 25), dated November 2011 (ADAMS Accession No. ML12011A024).

3.3.1 Meteorological Data TVA provided the 1991 through 2010 meteorological data as part of a prior licensing action, the Watts Bar, Unit 2, license application. The data were provided in hourly format for input into the ARCON96 atmospheric dispersion computer code (NUREG/CR-6331, Revision 1, "Atmospheric

- 6 Relative Concentrations in Building Wakes") by letter dated October 17, 2011 (ADAMS Accession No. ML11294A461). In addition, TVA provided the data in the form of a jOint (wind speed, wind direction, and atmospheric stability) frequency distribution (~IFD) suitable for input to the PAVAN atmospheric dispersion computer code (NUREG/CR-2858, "PAVAN: An Atmospheric Dispersion Program for Evaluating Design Basis Accidental Releases of Radiological Materials from Nuclear Power Stations") by letter dated November 7, 2011 (ADAMS Accession No. ML11314A116).

As discussed in SSER 25, NRC staff previously performed a review of the 1991 through 2010 onsite hourly meteorological data using the methodology described in NUREG-0917, "Nuclear Regulatory Commission Staff Computer Programs for Use with Meteorological Data." Further statistical review was performed using computer spreadsheets. Data recovery for the 20-year period for all parameters was consistently in the upper 90 percentiles, which exceeds the recommendation of RG 1.23 of at least 90 percent data recovery. With respect to atmospheric stability measurements, stable and neutral conditions were consistently reported to occur at night and unstable and neutral conditions during the day. The frequency, length, and time of occurrence of stable and unstable atmospheric conditions were very congruent with expected meteorological conditions. Wind speed and direction frequency distributions for each measurement channel were also very consistent from year to year and when comparing measurements between the measurement heights. A comparison of the ~IFD generated by NRC staff from the hourly data with the JFD provided by TVA showed reasonably good agreement.

On the basis of this review, the NRC staff concluded that the 1991 through 2010 data files provided by the TVA give an adequate representation of the site conditions to facilitate calculation of the CR, EAB, and LPZ X/a values for the SGTR and MSLB dose assessments for Watts Bar, Unit 1.

3.3.2 Control Room Atmospheric Dispersion Factors TVA used guidance in RG 1.194 and the ARCON96 methodology to calculate CR X/a values by incorporating inputs discussed in SSER 25 and in the WBN Unit 1, LAR, as confirmed to be applicable to WBN Unit 1, in the ~Iuly 18, 2012, supplemental letter.

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 Watts Bar Unit 1, LAR. The NRC staff qualitatively reviewed TVA's inputs to the ARCON96 computer runs and found them adequately consistent with site configuration drawings and staff practices. NRC staff noted that TVA 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 calculate X/a values to compare with the X/a values calculated by TVA and has concluded that the CR X/a values identified in Table 1 are acceptable for use in the SGTR and MSLB dose assessments associated with the WBN Unit 1, LAR.

-7 3.3.3 Offsite Atmospheric Dispersion Factors TVA updated the EAB and LPZ X/Q values for WBN Unit 1, using guidance in RG 1.145, as discussed in the WBN Unit 1, LAR and SSER 25.

The NRC staff qualitatively reviewed the inputs and assumptions used by TVA and found them reasonably consistent with NRC regulatory guidance and staff practices. In addition, NRC staff made comparison calculations using the PAVAN computer code and obtained results similar to the EAB and LPZ X/Q values generated by TVA On the basis of this review, the staff has concluded that the EAB and LPZ X/Q values presented in Table 1 are acceptable for use in the WBN Unit 1, MSLB and SGTR dose assessments.

3.4 Summary The licensee evaluated the radiological consequences resulting from the postulated SGTR and MSLB accidents and concluded that the radiological consequences at the EAB, the LPZ, and the CR comply with the reference values provided in 10 CFR 100.11 and the accident specific dose guidelines specified in RG 1.195. 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 and that the revised DEI spike limit of 14 IJCi/gm will impose a more restrictive condition on plant operations. The other assumptions previously found acceptable to the NRC staff are presented in the WBN Unit 1, UFSAR. The licensee's calculated dose results are given in Table 2 and Table 3 of this SE. The NRC staff finds, with reasonable assurance, that the licensee's estimates of the dose consequences of a design basis SGTR and an MSLB will comply with the requirements of 10 CFR 100.11 and the accident specific dose guidelines specified in RG 1.195, 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 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 28633). Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22{c)(1 0). 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.

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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 Date: December 5, 2012

Attachment:

Tables 1 - 3

Table 1 ~ WBN Unit 1 Revised Offsite and Onsite Atmospheric Dispersion Factors (X/Qs)

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

Exclusion Area Boundary (EAB) 0-2 hours 6.382 x 10-4 Outer Boundary of the Low Population Zone (LPZ) 0-2 hours 1.784 x 10-4 2 - 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 8.835 x 10-0 8 - 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 6.217 x 1O-l) 1 - 4 days 2.900 x 10-::'

4 - 30 days 9.811 x 10-6 Unit 1 Control Room (CR) 0-2 hours 3.85 x 10-;:5 2 - 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 3.22 x 1O-;J Table 2 - WBN Unit 1 MSLB Radiolo ical Conse Acceptance EAB (2) LPZ (3)

Acceptance Criteria Criteria Accident Initiated Iodine Spike Case (0.265 IJCi/gm steady state)

Gamma 1.25E-02 5 1.04E-01 1.23E-01 2.5 Thyroid 1.73E+01 30 3.20E+00 4.59E+00 30 Pre-Accident Iodine Spike Case (14 IJCi/gm maximum peak)

Gamma 7.07E-03 5 2.92E-02 1.16E-02 Thyroid 1.31 E+01 30 2.63E+00 1.27 LPZ (3)

Accident Initiated Iodine Spike Case (0.265 IJCi/gm steady state)

Gamma 8.11 E-02 5 5.03E-01 1.47E-01 2.5 Thyroid 3.37E+00 30 6.33E+00 1.86E+00 30 Pre-Accident Iodine Spike Case (14 IJCi/gm maximum peak)

Gamma 8.56E02 5 3.50E-01 1.03E-01 25 Thyroid 2.18E+01 30 1.33E+01 3.81E+00 300 (1) CR - Integrated 30 day dose (2) Exclusion area boundary hour dose (3) Low population zone - Integrated 30 day dose Attachment

December 5, 2012 Mr. Joseph W. Shea Corporate Manager, Nuclear Licensing Tennessee Valley Authority 3R Lookout Place 1101 Market Street Chattanooga, TN 37402-2B01

SUBJECT:

WATTS BAR NUCLEAR PLANT, UNIT 1 -ISSUANCE OF AMENDMENT REGARDING TECHNICAL SPECIFICATION CHANGES IN DOSE EQUIVALENT 1-131 SPIKE LIMIT AND ALLOWABLE VALUE FOR CONTROL ROOM AIR INTAKE RADIATION MONITORS (TAC NO. MEB156)

Dear Mr. Shea:

The U.S. Nuclear Regulatory Commission (NRC) has issued the enclosed Amendment No.

to Facility Operating License No. NPF-90 for Watts Bar Nuclear Plant, Unit 1. This amendment consists of changes to the license in response to your application dated March B, 2012, as supplemented by a letter dated July 1B, 2012.

The amendment revises: (1) Technical Specification (TS) 3.3.7, "Control Room Emergency Ventilation System (CREVS) Actuation Instrumentation," by changing the Allowable Value for the main control room air intake radiation monitoring instrumentation in Table 3.3.7-1 from less than or equal to (::;;) 9A5E-05 micro-Curie per cubic centimeter (j..JCi/cc) (3,30B counts per minute (cpm>> to::;; 1.647E-04 j..JCi/cc (3,30B cpm); and (2) TS 3.4.16, "RCS [Reactor Coolant System] Specific Activity," by lowering the DOSE EQUIVALENT iodine 131 spike Iirnitfrom 21 micro-Curie/gram (j..JCi/gm) to 14 j..JCi/gm in Required Action A.1 and Condition C.

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. 91to NPF-90
2. Safety Evaluation cc w/encls: Distribution via Listserv Distribution:

PUBLIC RidsNrrDorlDpr RidsNrrLABClayton RidsAcrsAcnw MailCTR LBrown. NRR LPL2-2 r/f RidsNrrDorlLpl2-2 RidsOgcRp Resource RidsNrrDraAadh RidsNrrDssStsb RidsNrrPMWattsBar1 RidsRgn2MailCenter JParilio. NRR ADAMS Accession No ML12279A115 OFFICE NRRlLP2-21PM NRRlLP2-21LA NRR/AADB/BC NRRlSTSB/BC OGC NLO NRR/LPB2-21BC 1NRRlLP2-21PM NAME AHon BClayton TTate RElliott LSubin JQuichocho I: AHon DATE 10/16/12 10/16/12 10/17112 10/16/12 10/26112 12/5/12 [12/5/12 OFFICIAL AGENCY RECORD

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