ML110980197

Issuance of Amendment No. 293, Technical Specification Changes and Analyses Regarding Use of Alternate Source Term

ML110980197
Person / Time
Site:	Arkansas Nuclear
Issue date:	04/26/2011
From:	Kalyanam N Plant Licensing Branch IV
To:	Entergy Operations
Kalyanam N, NRR/DORL/LPL4, 415-1480
References
TAC ME3678
Download: ML110980197 (51)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 April 26, 2011 Vice President, Operations Arkansas Nuclear One Entergy Operations, Inc.

1448 S.R. 333 Russellville, AR 72802

SUBJECT:

ARKANSAS NUCLEAR ONE, UNIT NO.2 - ISSUANCE OF AMENDMENT RE:

USE OF ALTERNATE SOURCE TERM (TAC NO. ME3678)

Dear Sir or Madam:

The Nuclear Regulatory Commission (NRC) has issued the enclosed Amendment No. 293 to Renewed Facility Operating License No. NPF-6 for Arkansas Nuclear One, Unit NO.2. The amendment consists of changes to the Technical Specifications (TSs) in response to your application dated March 31, 2010, as supplemented by letters dated June 23, June 24, August 9, and September 16, 2010.

The amendment modifies the requirements of the TS definitions, requirements, and terminology related to the use of an Alternate Source Term (AST) associated with accident offsite and control room dose consequences. In addition, implementation of the AST supports adoption of the control room envelope habitability controls in accordance with NRC-approved Technical Specification Task Force (TSTF)-448, Revision 3, "Control Room Habitability."

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

Sincerely, N. Kaly Kalyanam, Project Manager Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-368

Enclosures:

1. Amendment No. 293 to NPF-6

2. Safety Evaluation cc w/encls: Distribution via Listserv

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555*0001 ENTERGY OPERATIONS, INC.

DOCKET NO. 50~368 ARKANSAS NUCLEAR ONE, UNIT NO.2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 293 Renewed License No. NPF~6

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

A. The application for amendment by Entergy Operations, Inc. (the licensee), dated March 31, 2010, as supplemented by letters dated June 23, June 24, August 9, and September 16, 2010, 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 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 license 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.

Enclosure 1

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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 Renewed Facility Operating License No. NPF-6 is hereby amended to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 293, are hereby incorporated in the renewed license. The licensee shall operate the facility in accordance with the Technical Specifications

3. The license amendment is effective as of its date of issuance and shall be implemented within 90 days from the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION B o..\w A~t \c ~S.""i~ .~

Michael T. Markley, Chief Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Renewed Facility Operating License No. NPF-6 Technical Specifications Date of Issuance: Apr i 1 26, 2011

ATIACHMENT TO LICENSE AMENDMENT NO. 293 RENEWED FACILITY OPERATING LICENSE NO. NPF-6 DOCKET NO. 50-368 Replace the following pages of the Renewed Facility Operating License No. NPF-6 and Appendix A Technical Specifications with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

Operating License REMOVE INSERT

-3 -3 Technical Specifications REMOVE INSERT 1-4 1-4 3/44-18 3/44-18 6-16 6-16

3 (4) EOI, pursuant to the Act and 10 CFR Parts 30,40 and 70 to receive, possess and use at any time any byproduct, source and special nuclear material as sealed neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required; (5) EOI, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess, and use in amounts as required any byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; and (6) EOI, pursuant to the Act and 10 CFR Parts 30 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 renewed license shall be deemed to contain and is subject to conditions specified in the following Commission regulations in 10 CFR Chapter 1; Part 20, Section 30.34 of Part 30, Section 40.41 of Part 40, Sections 50.54 and 50.59 of Part 50, and Section 70.32 of Part 70; 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 EOI is authorized to operate the facility at steady state reactor core power levels not in excess of 3026 megawatts thermal. Prior to attaining this power level EOI shall comply with the conditions in Paragraph 2.C.(3).

(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 293 are hereby incorporated in the renewed license. The licensee shall operate the facility in accordance with the Technical Specifications.

Exemptive 2nd paragraph of 2.C.2 deleted per Amendment 20, 3/3/81.

(3) Additional Conditions The matters specified in the following conditions shall be completed to the satisfaction of the Commission within the stated time periods following issuance of the renewed license or within the operational restrictions indicated.

The removal of these conditions shall be made by an amendment to the renewed license supported by a favorable evaluation by the Commission.

2.C.(3)(a) Deleted per Amendment 24,6/19/81.

Renewed License No. NPF-6 Amendment No. 293 Revised by letter dated July 18, 2007

DEFINITIONS UNIDENTIFIED LEAKAGE 1.15 UNIDENTIFIED LEAKAGE shall be all leakage which is not IDENTIFIED LEAKAGE or controlled leakage.

PRESSURE BOUNDARY LEAKAGE 1.16 PRESSURE BOUNDARY LEAKAGE shall be leakage (except primary to secondary leakage) through a non~isolable fault in a Reactor Coolant System component body.

pipe wall or vessel wall.

AZIMUTHAL POWER TILT - Tg 1.17 AZIMUTHAL POWER TILT shall be the power asymmetry between azimuthally symmetric fuel assemblies.

DOSE EQUIVALENT 1~131 1.18 DOSE EQUIVALENT 1-131 shall be that concentration of 1-131 (microcuries per gram) that alone would produce the same committed effective dose equivalent (CEDE) as the quantity and isotopic mixture of 1-131.1-132.1-133.1-134. and 1-135 actually present.

The CEDE dose conversion factors used to determine the DOSE EQUIVALENT 1-131 shall be performed using Table 2.1 of EPA Federal Guidance Report No. 11. 1988.

"Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation. Submersion. and Ingestion."

DOSE EQUIVALENT XE-133 1.19 DOSE EQUIVALENT XE-133 shall be that concentration of Xe-133 (microcuries per gram) that alone would produce the same acute dose to the whole body as the combined activities of noble gas nuclides Kr-85m. Kr-85, Kr-87, Kr-88. Xe-131m, Xe-133m, Xe-133, Xe-135m. Xe-135, and Xe-138 actually present. If a specific noble gas nuclide is not detected. it should be assumed to be present at the minimum detectable activity. The determination of DOSE EQUIVALENT XE-133 shall be performed using effective dose conversion factors for air submersion listed in Table 111.1 of EPA Federal Guidance Report No. 12. 1993, "External Exposure to Radionuclides in Air, Water, and SoiL" STAGGERED TEST BASIS 1.20 A STAGGERED TEST BASIS shall consist of:

a. A test schedule for n systems, subsystems. trains or other designated components obtained by dividing the specified test interval into n equal subintervals, and

b. The testing of one system, subsystem, train or other designated component at the beginning of each subinterval.

FREQUENCY NOTATION 1.21 The FREQUENCY NOTATION specified for the performance of Surveillance Requirements shall correspond to the intervals defined in Table 1.2.

ARKANSAS - UNIT 2 1-4 Amendment No. 4a+.~.~,~. 293

REACTOR COOLANT SYSTEM SPECIFIC ACTIVITY LIMITING CONDITION FOR OPERATION 3.4.8 RCS DOSE EQUIVALENT 1-131 and DOSE EQUIVALENT XE-133 specific activity shall be within limits.

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

ACTION Note: The provisions of Specification 3.0.4.c are applicable to ACTION a and b.

a. With the DOSE EQUIVALENT 1-131 not within limit:

1. Verify DOSE EQUIVALENT 1-131 :S 60 IJCi/gm once every 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, and

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

b. With the DOSE EQUIVALENT XE-133 not within limit, restore DOSE EQUIVALENT XE-133 within limit within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.

c. With the requirements of ACTION a and/or b not met, or with DOSE EQUIVALENT 1-131

> 60 IJCi/gm, be in at least HOT STANDBY in 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

SURVEILLANCE REQUIREMENTS 4.4.8.1 Verify reactor coolant DOSE EQUIVALENT XE-133 specific activity :S 3100 IJCi/gm once every 7 days.*

4.4.8.2 Verify reactor coolant DOSE EQUIVALENT 1-131 specific activity:S 1.0 IJCi/gm:*

a. once every 14 days, and

b. between 2 and 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> after THERMAL POWER change of ~ 15% RATED THERMAL POWER within a 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> period.

• Only required to be performed in MODE 1.

ARKANSAS - UNIT 2 3/44-18 Amendment No. ~,~,~, 293 Next Page is 3/4 4-22

ADMINISTRATIVE CONTROLS 6,5,12 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 5 rem Total Effective Dose Equivalent (TEDE) 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.

d, Measurement, at designated locations, of the CRE pressure relative to all external areas adjacent to the CRE boundary during the pressurization mode of operation by one train of the CREVS, operating at the flow rate required by the VFTP, at a Frequency of one train every 18 months. The results shall be trended and used as part of the 18 month assessment of the CRE boundary, e, The quantitative limits on unfiltered air inleakage into the CRE, These limits shall be stated in a manner to allow direct comparison to the unfiltered air inleakage measured by the testing described in paragraph c. The unfiltered air inleakage limit for radiological challenges is the inleakage flow rate assumed in the licensing basis analyses of DBA consequences. Unfiltered air inleakage limits for hazardous chemicals must ensure that exposure of CRE occupants to these hazards will be within the assumptions in the licensing basis.

f. The provisions of Specification 4,0.2 are applicable to the Frequencies for assessing CRE habitability, determining CRE unfiltered inleakage, and measuring CRE pressure and assessing the CRE boundary as required by paragraphs c and d, respectively, ARKANSAS - UNIT 2 6-16 Amendment No, 200,288, 293

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 293 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-6 ENTERGY OPERATIONS, INC.

ARKANSAS NUCLEAR ONE, UNIT NO.2 DOCKET NO. 50-368

1.0 INTRODUCTION

By application dated March 31, 2010 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML100910241), as supplemented by letters dated June 23, June 24, August 9, and September 16,2010 (ADAMS Accession Nos. ML102000199, ML101750247, ML102250417, and ML102590567, respectively), Entergy Operations, Inc. (the licensee), requested changes to the Technical Specifications (TSs) for Arkansas Nuclear One, Unit No.2 (ANO-2). The supplemental letters June 23, June 24, August 9, and September 16, 2010, provided additional information that clari'fied the application, did not expand the scope of the application as originally noticed, and did not change the U.S. Nuclear Regulatory Commission (NRC) staffs original proposed no significant hazards consideration determination as published in the Federal Register on June 29,2010 (75 FR 37475).

The amendment would implement an alternative source term (AST) for ANO-2 and would modify the ANO-2 TS requirements related to the use of an AST associated with accident offsite and control room dose consequences. Implementation of the AST supports adoption of the control room envelope habitability controls in accordance with Technical Specification Task Force (TSTF)-448, Revision 3, "Control Room Habitability," TSTF-448 related changes were approved by the NRC for ANO-2 via Amendment No. 288 on October 29, 2009 (ADAMS Accession No. ML082520574).

With significant advances that have been made in understanding the timing, magnitude, and chemical form of fission product releases from severe nuclear power plant accidents, a holder of an operating license issued prior to January 10, 1997, is allowed by Section 50.67, "Accident source term," of Title 10 of the Code of Federal Regulations (10 CFR) to voluntarily revise the traditional accident source term used in design basis radiological consequence analyses.

Enclosure 2

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2.0 REGULATORY EVALUATION

The NRC staff evaluated the radiological consequences of the postulated design-basis accidents (DBAs) against the dose criteria specified in 10 CFR 50.67. The applicable criteria are 5 roentgen equivalent man (rem) Total Effective Dose Equivalent (TEDE) in the control room, 25 rem TEDE at the exclusion area boundary (EAB), and 25 rem TEDE at the outer boundary of the low population zone (LPZ).

The regulatory requirements upon which the NRC staff based its acceptance are NUREG-0800, "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants:

LWR [Light-Water Reactor] Edition," (SRP), Section 15.0.1, "Radiological Consequence Analyses Using Alternative Source Terms"; 10 CFR Part 50, Appendix A, "General Design Criteria [GDC] for Nuclear Power Plants," Criterion 19, "Control room";. and the accident dose criteria in 10 CFR 50.67, as supplemented in Regulatory Position 4.4 and Table 6 of NRC Regulatory Guide (RG) 1.183, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors," July 2000. The licensee has not proposed any deviation or departure from the guidance provided in RG 1.183. The NRC staff's evaluation is based upon the following regulatory codes, regulatory guides, and standards, in addition to relevant information in the ANO-2 Final Safety Analysis Report (FSAR) and TSs, as well as considerations for any applicable alternative documentation the licensee may have provided:

• 10 CFR 50.34, "Contents of applications; technical information," requires that each applicant for a construction permit or operating license provide an analysis and evaluation of the design and performance of structures, systems, and components of the facility with the objective of assessing the risk to public health and safety resulting from operation of the facility.

• 10 CFR 100.11, "Determination of exclusion area, low population zone, and population center distance," (under "Subpart A--Evaluation Factors for Stationary Power Reactor Site Applications Before January 10, 1997 and for Testing Reactor") provides criteria for evaluating the radiological aspects of the proposed site.

• 10 CFR 50.67, "Accident source term," provides an optional provision for licensees to revise the AST used in design basis radiological analyses.

• 10 CFR 50.49, "Environmental qualification of electric equipment important to safety for nuclear power plants," requires that the safety related electrical equipment which are relied upon to remain functional during and following design basis events be qualified for accident (harsh) environment.

• 10 CFR Part 50, Appendix A, GDC 17, "Electric power systems," requires, in part, that nuclear power plants have onsite and offsite electric power systems to permit the functioning of structures, systems, and components that are important to safety.

• 10 CFR Part 50, Appendix A, GDC 19, "Control room."

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• SRP Section 2.3.4, "Short-Term Atmospheric Dispersion Estimates for Accident Releases."

• SRP Section 6.4, "Control Room Habitability Systems."

• SRP Section 15.0.1, "Radiological Consequence Analyses Using Alternative Source Terms."

• NUREG/CR-6604, "RADTRAD: Simplified Model for RADionuclide Iransport and Removalgnd Dose Estimation," June 1999 (ADAMS Accession No. ML081850607, (non-public>>.

• NUREG/CR-6331, Revision 1, "Atmospheric Relative Concentrations in Building Wakes," May 1997 (ADAMS Accession No. ML003769825, (non-public)).

• NUREG/CR-2858, "PAVAN: An Atmospheric Dispersion Program for Evaluating Design Basis Accidental Releases of Radioactive Materials from Nuclear Power Stations," November 1982.

• NUREG-0917, "Nuclear Regulatory Commission Staff Computer Programs for Use with Meteorological Data," July 1982.

• RG 1.23, "Onsite Meteorological Programs," February 1972.

• RG 1.145, "Atmospheric Dispersion Models for Potential Accident Consequence Assessments at Nuclear Power Plants," November 1982 (ADAMS Accession No. ML003740205).

• RG 1. 183, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors," (ADAMS Accession No. ML003716792).

• RG 1.194, "Atmospheric Relative Concentrations for Control Room Radiological Habitability Assessments at Nuclear Power Plants," Revision 0, June 2003 (ADAMS Accession No. ML0314305050.)

• RG 1.197, "Demonstrating Control Room Envelope Integrity at Nuclear Power Reactors," May 2003 (ADAMS Accession No. ML031390664).

• NRC Regulatory Issue Summary (RIS) 2006-04, "Experience with Implementation of Alternative Source Terms," March 7, 2006 (ADAMS Accession No. ML053460347).

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3.0 TECHNICAL EVALUATION

3.1 Proposed TS Changes Current TS Section 1.0, "DEFINITIONS," TS 1.18, "DOSE EQUIVALENT 1-131," states:

1.18 DOSE EQUIVALENT 1-131 shall be that concentration of 1-131 (microcuries per gram) that alone would produce the same dose when inhaled as the combined activities of iodine isotopes 1-131,1-132,1-133, 1-134, and 1-135 actually present. The determination of DOSE EQUIVALENT 1-131 shall be performed using Table III of TID-14844, AEC, 1962, "Calculation of Distance Factors for Power and Test Reactor Sites."

Revised TS 1.18 would state:

1.18 DOSE EQUIVALENT 1-131 shall be that concentration of 1-131 (microcuries per gram) that alone would produce the same committed effective dose equivalent (CEDE) as the quantity and isotopic mixture of 1-131, 1-132, 1-133, 1-134, and 1-135 actually present. The CEDE dose conversion factors used to determine the DOSE EQUIVALENT 1-131 shall be performed using Table 2.1 of EPA Federal Guidance Report No. 11, 1988, "Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion."

Current TS 1.19, "DOSE EQUIVALENT XE-133," states:

1.19 DOSE EQUIVALENT XE-133 shall be that concentration of Xe-133 (microcuries per gram) that alone would produce the same acute dose to the whole body as the combined activities of noble gas nuclides Kr-85m, Kr-85, Kr-87 , Kr-88 , Xe-131m, Xe-133m, Xe-133, Xe-135m, Xe-135, and Xe-138 actually present. If a specific noble gas nuclide is not detected, it should be assumed to be present at the minimum detectable activity. The determination of DOSE EQUIVALENT XE-133 shall be performed using the average disintegration energies derived from the dose conversion factors provided in ICRP Publication 2, "Report of ICRP Committee II on Permissible Dose for Internal Radiation" for non-fuel damage events.

Revised TS 1.19 would state:

1.19 DOSE EQUIVALENT XE-133 shall be that concentration of Xe-133 (microcuries per gram) that alone would produce the same acute dose to the whole body as the combined activities of noble gas nuclides Kr-85m, Kr-85, Kr-87, Kr-88 , Xe-131m, Xe-133m, Xe-133, Xe-135m, Xe-135, and Xe-138 actually present. If a specific noble gas nuclide is not detected, it should be assumed to be present at the minimum detectable activity. The determination of DOSE EQUIVALENT XE-133 shall be performed using

- 5 effective dose conversion factors for air submersion listed in Table 111.1 of EPA Federal Guidance Report No. 12, 1993, "External Exposure to Radionuclides in Air, Water, and Soil."

Current TS 3.4.8, "Reactor Coolant System - Specific Activity," SR 4.4.8.1 states:

4.4.8.1 Verify reactor coolant DOSE EQUIVALENT XE-133 specific activity s 3200 IJCi/gm once every 7 days.'"

Revised SR 4.4.8.1 would state:

4.4.8.1 Verify reactor coolant DOSE EQUIVALENT XE-133 specific activity s 3100 IJCi/gm once every 7 days.'"

Current TS 6.5.12, "Control Room Envelope Habitability Program," states that:

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 5 rem whole body or its equivalent to any part of the body for the duration of the accident. The program shall include the following elements:

a.-f. [no change]

Revised TS 6.5.12 would state:

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 5 rem Total Effective Dose Equivalent (TEDE) for the duration of the accident. The program shall include the following elements:

a.-f. [no change]

-6 3.2 Atmospheric Dispersion Estimates 3.2.1 Control Room Atmospheric Dispersion Factors The licensee used hourly meteorological data previously provided in support of the ANO-2 power uprate license amendment request (LAR) to generate the control room atmospheric dispersion factors (X/Q values) used in the dose assessment for the current LAR. These X/Q values were generated using the site meteorological data collected from 1995 through 1999 and the ARCON96 atmospheric dispersion computer code (NUREG/CR-6331, Revision 1, "Atmospheric Relative Concentrations in Building Wakes"). RG 1.194, "Atmospheric Relative Concentrations for Control Room Radiological Habitability Assessments at Nuclear Power Plants," states that ARCON96 is an acceptable methodology for assessing control room X/Q values for use in DBA radiological analyses. The NRC staff evaluated the applicability of the ARCON96 model and concluded that there are no unusual siting, building arrangements, release characterization, source-receptor configuration, meteorological regimes, or terrain conditions that preclude use of this model in support of the current LAR for ANO-2.

The licensee stated that the accident scenarios and X/Q values model the limiting doses considering multiple release scenarios, including those due to loss of offsite power or other single failure. Releases were postulated to occur from six release locations and to be transported to both control room intakes. The licensee provided a table of those 12 sets of X/Q values from which the X/Q value associated with the limiting source-receptor pair for each time period was identified for each DBA:

• Containment building wall to intake VPH-1 -loss of coolant accident (LOCA),

and control rod ejection accident (CREA)

• Main steam safety valves (MSSV) to intake VPH steam generator tube rupture (SGTR) accident, main steam line break (MSLB) accident, locked rotor accident (LRA), and CREA

• Atmospheric dump valves (ADV) to intake VPH MSLB, SGTR, LRA, and CREA

• Penetration room ventilation system (PRVS) exhaust to intake VPH LOCA

• Main steam pipe to intake VPH MSLB

• Fuel handling area to intake VPH fuel handling accident (FHA)

All sources were modeled as ground-level releases, consistent with guidance in RG 1.194. The release from the containment was modeled as a diffuse source based upon RG 1.194 guidance and the other releases, as listed above, were modeled as point sources. For each potential release location, the licensee used the most limiting X/Q values.

-7 In summary, the NRC staff qualitatively reviewed the inputs to the ARCON96 computer runs for the control room X/a value assessment and found them generally consistent with site configuration drawings and staff practice. The staff also reviewed the licensee's dispersion modeling and performed confirmatory calculations, and concludes that the licensee's X/a values which are listed in Table 3.1-1 in this safety evaluation (SE) are acceptable for use in control room DBA dose assessment associated with the current LAR.

3.2.2 Offsite Atmospheric Dispersion Factors The licensee used the current ANO-2 licensing basis EAB and LPZ X/a values listed in Table 3.1-2 of this SE to assess the radiological consequences of the DBAs postulated in the current LAR. These X/a values are presented in Table 15.1.0-5 of Amendment 20 to the ANO-2 FSAR.

The NRC staff previously reviewed these EAB and LPZ X/a values in support of a prior licensing action, related to ANO-2 Amendment No. 244 dated April 24, 2002 (ADAMS Accession No. ML021130826) and noted the methodology used by the licensee resulted in EAB and LPZ X/a values that were generated prior to issuance of RG 1.145. RG 1.145 states that when using a direction dependent model, the higher of the 0.5 percentile X/a value by direction or the 5 percentile X/a value for the overall site should be selected. The NRC staff determined that it was not necessary to revise the EAB and LPZ X/a calculations proposed in support of Amendment No. 244 since 10 CFR Part 100, "Reactor site criteria," dose limits could still be met using the RG 1.145 guidance. NRC staff further noted that future license amendments involving dose assessments using the X/a values listed in Table 15.1.0-5 of Amendment 20 to the ANO-2 FSAR should be reviewed on a case-by-case basis.

In summary, consistent with the guidance provided in RG 1.183, the NRC staff concludes that it is not necessary to revise the licensing basis EAB and LPZ X/a values used in the dose assessment for the current LAR. However, NRC staff concludes that any future license amendments involving dose assessments using the X/a values in Table 15.1 .0-5 of Amendment 20 should be reviewed on a case-by-case basis.

3.3 Radiological Conseguences of Design Basis Accidents To support the proposed implementation of an AST, the licensee analyzed the radiological dose consequences and provided all major inputs and assumptions for the following DBAs:

• LOCA

• FHA

• MSLB

• SGTR

• LRA

• CREA As a minimum effort to revise the ANO-2 licensing basis to incorporate a full implementation of the AST, RG 1.183, Position 1.2.1, specifies that the DBA LOCA must be re-analyzed using the

-8 appropriate guidance therein. The licensee re-analyzed the six DBAs listed above, which includes the design basis LOCA at ANO-2.

The licensee's submittal reports the results of the radiological consequence analyses for the above DBAs to show compliance with 10 CFR 50.67 dose acceptance criteria, or fractions thereof, as defined in SRP Section 15.0.1, for doses offsite and in the control room. For full implementation of the AST DBA analysis methodology, the dose acceptance criteria specified in 10 CFR 50.67 provides an alternative to the previous whole body and thyroid dose guidelines stated in 10 CFR 100.11 and GDC 19. The subject LAR is considered a request for full implementation of the AST.

The following sections discuss the NRC staff's review of the DBA dose assessment performed by the licensee to support the LAR submittal dated March 31, 2010, as supplemented by the letter dated June 23,2010.

Core Inventory RG 1.183, Regulatory Position 3.1, "Fission Product Inventory," states, in part, the following:

The inventory of fission products in the reactor core and available for release to the containment should be based on the maximum full power operation of the core with, as a minimum, current licensed values for fuel enrichment, fuel burnup, and an assumed core power equal to the current licensed rated thermal power times the ECCS [Emergency Core Cooling System] evaluation uncertainty. The period of irradiation should be of sufficient duration to allow the activity of dose-significant radionuclides to reach equilibrium or to reach maximum values.

The core inventory should be determined using an appropriate isoto~e generation and depletion computer code such as ORIGEN 2 (Ref. [1 ) or ORIGEN-ARP (Ref. [2J).

For accident analyses postulating fuel damage or substantial core melt, and in accordance with the guidance of RG 1.183, the licensee calculated the core isotopic inventory available for release using the ORIGEN-S isotope generation and depletion computer code, and then multiplied the isotopic-specific activities by the relevant power level and release fractions. In addition, the licensee added an additional 4 percent margin to its calculated source terms in order to bound potential future increases in uranium fuel enrichment. The licensee stated that sensitivity analyses show that this 4 percent margin produces results that bound fuel enrichments up to 4.6 weight percent (W/o) U-235; the current fuel enrichment implemented at ANO-2 is 4.6 w/0 U-235. The NRC staff concludes that the licensee's use of the cited isotope generation and depletion computer code is acceptable for establishing the core inventory for AST accident analyses, because it is consistent with the guidance in RG 1.183.

A.G. Croff, "A User's Manual for the ORIGEN 2 Computer Code," ORNLfTM-7175, Oak Ridge National Laboratory, July 1980.

2 S.M. Bowman and L.C. Leal, "The ORIGNARP Input Processor for ORIGEN-ARP," Appendix F7.A in SCALE: A Modular Code System for Performing Standardized Analyses for Licensing Evaluation, NUREG/CR-0200, USNRC, March 1997.

-9 As stated in RG 1.183, the release fractions associated with the LWR core inventory released into containment for the design basis LOCA and non-LOCA events have been determined to be acceptable for use with currently approved LWR fuel with a peak burnup of 62,000 megawatt days per metric ton of uranium (MWd/MTU) provided that the maximum linear heat generation rate does not exceed 6.3 kilowatt per foot (kw/ft) peak rod average power for burnups exceeding 54,000 MWd/MTU.

Confirmatory Calculation To perform independent confirmatory dose calculations for the DBAs, the NRC staff used the NRC-sponsored radiological consequence computer code, "RADTRAD: Simplified Model for RADionuciide Iransport and Removal 6nd Dose Estimation," Version 3.03, as described in NUREG/CR-6604. The RADTRAD code, developed by the Sandia National Laboratories for the NRC, estimates transport and removal of radionuclides and the resulting radiological consequences at selected receptors.

3.3.1 Loss-of-Coolant Accident The current ANO-2 design basis LOCA analysis is based on the traditional accident source term described in Technical Information Document (TID)-14844, "Calculation of Distance Factors for Power and Test Reactor Sites." The current licensing basis radiological consequence analysis for the postulated LOCA is provided in the ANO-2 FSAR Section 15.1.13, "Major Rupture of Pipes Containing Reactor Coolant up to and Including Double-Ended Rupture of the Largest Pipe in the Reactor Coolant System (Loss of Coolant Accident)." To support implementation of the AST, as requested by the subject LAR, the licensee reanalyzed the offsite and control room radiological consequences of the postulated LOCA. This reanalysis was performed to demonstrate that the engineered safety features (ESFs) designed to mitigate radiological consequences at ANO-2 will remain adequate after implementation of the AST.

In addition to the LAR, the licensee submitted the AST -based reanalysis of the LOCA including the assumptions, parameters, and newly calculated offsite and control room doses associated with implementing the AST methodology. The licensee cites RG 1.183 as providing the primary radiological analysis assumptions for its reanalysis of the postulated design basis LOCA.

Specifically, the NRC staff's guidance for the assumptions is detailed in RG 1.183, Appendix A, "Assumptions for Evaluating the Radiological Consequences of a LWR Loss-of-Coolant Accident."

3.3.1.1 Activity Source For the LOCA analysis, the licensee assumed that the core isotopic inventory, that is available for release into the containment atmosphere, is based on maximum full power operation of the core at 3,086.52 megawatt thermal (MWth), or 1.02 times the current licensed thermal power level of 3,026 MWth, in order to account for the ECCS evaluation uncertainty. Additionally, the current licensed values for fuel enrichment and burnup are assumed when determining the core isotopic inventory.

- 10 The core inventory release fractions and release timing for the gap and early in-vessel release phases of the DBA-LOCA were taken from RG 1.183, Table 2, "PWR [Pressurized-Water Reactor] Core Inventory Fraction Released Into Containment," and Table 4, "LOCA Release Phases," respectively. Also consistent with RG 1.183 guidance, the licensee assumed that the radioactive iodine speciation released from failed fuel is 95.00 percent aerosol (particulate),

4.85 percent elemental, and 0.15 percent organic. Whereas, the radioactive iodine speciation released from steaming and flashed liquid is 97.00 percent elemental and 3.00 percent organic.

3.3.1.2 Transport Methodology and Assumptions The licensee calculated the onsite and offsite dose consequences of the design basis LOCA by modeling the transport of activity released from the core to the environment, while accounting for appropriate activity dilution, holdup, and removal mechanisms.

The NRC staff reviewed the licensee's assessment of the following potential post-LOCA activity release pathways:

• Containment Leakage

• ESF Recirculation Leakage Also, the NRC staff reviewed the licensee's assessment of the following potential post-LOCA shine dose pathways:

• External Plume Shine

• Containment Cloud Direct Shine The following sections detail the NRC staff's review of the licensee's analysis of these post accident contributors to both control room and offsite dose.

3.3.1.2.1 Containment Leakage The current ANO-2 design basis containment leak rate, La, is equal to 0.1 percent volume per day (% per day), at containment peak pressure, as expressed in ANO-2 FSAR Table 15.1.13-1, "Assumptions for Design Basis Loss-of-Coolant-Accident." The licensee assumed the design basis leak rate La for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of the accident, then a subsequent reduction to half of this value (0.05 % per day) for the remainder of the accident duration, as is consistent with the guidance of RG 1.183.

In its submitted LOCA analysis, the licensee does not assume any credit for mitigation of airborne activity releases from containment by natural processes, e.g., natural deposition, or filtration systems, e.g., Penetration Room Ventilation System (PRVS). However, as a design basis, the licensee has proposed to credit mitigation of airborne activity releases by the ANO-2 containment spray system (CSS).

The CSS is assumed to be operating at 3 minutes following initiation of the accident and will continue operating for the accident duration. The licensee stated that the sprays were actually determined to be operating at 1 minute; therefore, using 3 minutes is a conservative

- 11 assumption. To model removal of activity by the CSS, the licensee separated the containment volume into two individually well-mixed compartments, one "sprayed" and one "unsprayed," as sprays directly affect only 78 percent of the containment volume. In the licensee's model, mixing between the two volumes is accomplished by the containment fan coolers that provide 11,880 cubic feet per minute (cfm) of mixing flow between the sprayed and unsprayed regions of the ANO-2 containment. The guidance of RG 1.183 allows for an assumption of a mixing rate equal to two turnovers of the unsprayed volume per hour, due to natural convection. Because the volume of the ANO-2 unsprayed region determined by the licensee is 392,000 cubic feet (fe), the natural convection mixing rate could potentially have been calculated to be approximately 13,000 cfm, which is considerably more than the mixing rate assumed by the licensee. The licensee's model of airborne activity removal by the ANO-2 CSS is conservative and consistent with regulatory guidance and is, therefore, acceptable to the NRC staff.

3.3.1.2.2 ESF Recirculation Leakage The guidance in RG 1.183, Appendix A, "Assumptions for Evaluating the Radiological Consequences of a LWR Loss-of-Coolant Accident," Section 5.2, states that leakage from the ESF system should be taken as two times the sum of the simultaneous leakage from all components in the ESF recirculation system above which the TSs, or licensee commitments to item 111.0.1.1 of NUREG-0737, "Clarification of TMI Action Plan Requirements," November 1980, would require declaring such systems inoperable. For ANO-2, the licensee assumed that leakage of coolant from ESF systems is possible upon initiation of those systems. The ANO-2 TSs do not provide a specific limitation on the total amount of operational leakage that is allowed from ESF systems. However, as a current design basis, the ANO-2 FSAR Section 15.1.13.4.1, "Recirculation Leakage," assumes that the total leakage from ECCS is equal to 2,060 cubic centimeters per hour (cc/hr), assuming two containment spray and three high-pressure safety injection (HPSI) pumps in operation. In accordance with the guidance of RG 1.183, the licensee assumed twice this amount (4,120 cc/hr) in analyzing the dose consequence from the ESF leakage pathway.

The licensee's DBA analysis assumed leakage to start at 1,560 seconds, which the licensee determined to be the time that the recirculation phase begins, and is consistent with AST activity release assumptions. The leakage was assumed to continue for the duration of the accident.

The licensee conservatively took no credit for holdup or filtration of this leakage by assuming that the leaked activity is immediately available for release to the atmosphere.

The licensee assumed that, with the exception of iodine, all radioactive material in the recirculating coolant is retained in the liquid phase, and also in accordance with RG 1.183, the chemical form of the released iodine activity is assumed to be 97.00 percent elemental and 3.00 percent organic. The licensee determined the maximum ANO-2 sump temperature to be greater than 212 degrees Fahrenheit CF), and therefore, assuming constant enthalpy, applied the equation shown in RG 1.183, Appendix A, Section 5.4 to calculate the flashing fraction of leaked coolant. The licensee determined this value to be 2.29 percent; however, in accordance with RG 1.183 guidance, a conservative flashing fraction of 10 percent is used in the licensee's analysis.

- 12 The licensee's treatment of ESF leakage is consistent with its current design basis and the applicable regulatory guidance and is, therefore, acceptable to the NRC staff.

3.3.1.2.3 External Plume Shine For the calculation of external cloud shine dose to the control room, the licensee applied the source term that was calculated using the model of accident releases into containment and out to the environment, as described in Sections 3.3.1.2.1 and 3.3.1.2.2 in this SE. As modeled by the licensee in its submitted LOCA design analysis, airborne activity that leaks from containment results in a radioactive plume external to the ANO-2 control room. During the postulated accident, this cloud will in turn result in additional dose to the control room operator inside the control room. For its design basis model, the licensee calculated the dose contribution from this external plume by using the simulations that were used to calculate the control room doses due to airborne activity transport, but modifying them slightly. The licensee created new dose conversion factor (DCF) input files to apply to these existing models. The new DCF input files set inhalation DCFs to zero, and adjusted the external exposure DCFs by multiplying them by a calculated attenuation factor associated with the concrete of the ANO-2 control room wall. The NRC staff determined this method will lead the RADTRAD code to output dose calculations resulting from direct activity shine only; the ingestion and inhalation component of the dose has been ignored, as it was accounted for in the licensee's model which is evaluated in Sections 3.3.1.2.1 and 3.3.1.2.2. Also, the staff determined it is mathematically appropriate and analytically acceptable to apply the control room wall attenuation factor directly to the DCF.

Based on the above, the NRC staff concludes that the licensee's method for calculating the external plume shine to control room personnel is acceptable.

3.3.1.2.4 Containment Cloud Shine To determine the dose contribution of the activity that is postulated to accumulate and form a cloud internal to the ANO-2 containment building, the licensee calculated gamma ray dose rates interior to the control room using its SCAP-II computer code. The SCAP-II code implements the point-kernel method to calculate radiation dose. Based on engineering judgment, the NRC staff finds the point-kernel method to be acceptable method in low-scatter systems characterized by simple source-receiver paths and has previously approved this method for other AST amendments. The geometry that the licensee applied to its SCAP-II calculation included simulation of the control room walls and ceiling, as well as of the outer structure of the containment building. The licensee modeled the cylindrical portion of the containment building as a 116-foot-diameter cylindrical shell with a concrete thickness of 3.75 feet. The height of the cylinder was considered to be 179 feet. The licensee applied the gamma source strength determined from the ORIGEN-S computer code to the model of the ANO-2 containment. The control room was treated as a structure with concrete wall and ceiling thickness of 1.5 feet.

Therefore, the licensee's method for calculating the dose contribution from the activity cloud internal to the ANO-2 containment is acceptable to the NRC staff, because it accurately and conservatively models the post-accident ANO-2 source, strengths and building dimensions.

- 13 3.3.1.3 Summary The licensee concluded that the radiological consequences at the EAB, LPZ, and control room are within the dose criteria specified in 10 CFR 50.67 and the accident-specific dose acceptance criteria specified in SRP Section 15.0.1 and RG 1.183. These accident-specific dose acceptance criteria for the LOCA are a TEDE of25 rem at the EAB for the worst 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, 25 rem at the outer boundary of the LPZ, and 5 rem for access to and occupancy of the control room for the duration of the accident. Based on the above discussion of the licensee's LOCA analysis, the NRC staff concludes that the licensee used sufficiently conservative analysis assumptions and inputs consistent with applicable regulatory guidance identified in Section 2.0 of this SE and with those stated in the ANO-2 FSAR as design bases. The staff also performed independent calculations of the dose consequences of the postulated LOCA releases, using the licensee's assumptions, which are consistent with RG 1.183 guidelines, for input to the RADTRAD computer code. The staff's calculations confirmed the licensee's dose results. The major parameters and assumptions used by the licensee and found acceptable to the staff are presented in Table 3.2.1 in this SE. The results of the licensee's design basis radiological consequence calculation are provided in Table 3.2. The staff concludes that the EAB, LPZ, and control room doses estimated by the licensee for the LOCA meet the applicable accident dose criteria and are, therefore, acceptable.

3.3.2 Fuel Handling Accident (FHA)

The current ANO-2 design basis FHA analysis is based on the traditional accident source term described in TID-14844. The ANO-2 licensing basis analysis is presented in FSAR Section 15.1.23, "Fuel Handling Accident." To support implementation of the AST as requested by the subject LAR, the licensee reanalyzed the offsite and control room radiological consequences of the postulated FHA. This reanalysis was performed to demonstrate that the ESFs designed to mitigate the radiological consequences at ANO-2 will remain adequate after implementation of the AST.

The licensee submitted the AST-based reanalysis of the FHA as part of the LAR. Included in this reanalysis are the assumptions, parameters, and newly calculated offsite and control room doses associated with implementing the AST methodology. The licensee cites RG 1.183 as providing the primary radiological analysis assumptions for its reanalysis of the postulated design basis FHA. Specifically, the NRC staffs guidance is detailed in RG 1.183, Appendix B, "Assumptions for Evaluating the Radiological Consequences of a Fuel Handling Accident."

The new ANO-2 design basis consists of an FHA occurring outside containment in the ANO-2 Fuel Handling Building. This postulated FHA outside containment results in damage to the dropped and impacted fuel assemblies and the release of the associated volatile gaseous radioactivity.

3.3.2.1 Activity Source For the FHA analysis, the licensee assumed that the core isotopic inventory is based on maximum full power operation of the core at 3,026 MWth, or 1.02 times the current licensed thermal power level of 3,086.52 MWth, in order to account for the ECCS evaluation uncertainty.

- 14 The licensee assumed the design basis radial peaking factor of 1.70. Additionally, the licensee accounted for current licensed values for fuel enrichment and burnup when determining the core isotopic inventory. The fraction of core isotopic activity assumed to be available for release from the gap of failed fuel (Le., fuel experiencing cladding failure as a result of the drop) is provided in Table 3, "Non-lOCA Fraction of Fission Product Inventory in Gap," of RG 1.183. Table 1.7.4-1 of Attachment 3 to the letter dated March 31, 2010, shows the core inventory used by the licensee. Therefore, upon approval of this LAR, the ANO-2 design basis will be analyzed for fuel burnup and enrichment values specified in Section 3.3 of this SE. These values meet the RG 1.183 limits of applicability expressed in its footnote 11.

The licensee assumed that, as a design basis, spent fuel will have decayed for at least 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> after shutdown before being handled. The licensee assumed that 472 pins will be damaged (two full fuel assemblies), as the result of the postulated FHA, and thus release all of its available gap activity over a 2-hour period. This bounds the current ANO-2 licensing basis, as described in ANO-2 FSAR Section 15.1.23, and is consistent with the guidance expressed in RG 1.183.

3.3.2.2 Transport Methodology and Assumptions Per the ANO-2 LAR, the minimum fuel depth over the reactor core when handling fuel and over the spent fuel in the fuel handling building is 23 feet. Therefore, consistent with RG 1.183, Appendix B, Section 2, "Water Depth," an overall effective iodine decontamination factor (DF) of 200 was assumed, which resulted in an iodine chemical speciation released from the water of 70 percent elemental and 30 percent organic. All particulate radionuclides were assumed to be retained in the water. The 23-foot depth water coverage easily bounds the available water coverage for a drop over the reactor core. Therefore, the licensee's assumption of a drop over the spent fuel pool is conservative.

The licensee took no credit for holdup or filtration of the activity release in the fuel handling building, and assumed the limiting dispersion associated with a release from the containment equipment hatch for conservatism. The activity release is modeled to occur within a time period not to exceed 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, in accordance with the guidance of RG 1.183.

3.3.2.3 Summary The licensee concluded that the radiological consequences at the EAB, lPZ, and control room are within the dose criteria specified in 10 CFR 50.67 and the accident-specific dose acceptance criteria specified in SRP 15.0.1 and RG 1.183. These accident-specific dose acceptance criteria for the FHA are a TEDE of 6.3 rem at the EAB for the worst 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, 6.3 rem at the outer boundary of the LPZ, and 5 rem in the control room for the duration of the accident.

Based on the above discussion of the licensee's FHA analysis, the NRC staff concludes that the licensee used analysis assumptions and inputs consistent with applicable regulatory guidance identified in Section 2.0 of this SE and with those stated in the ANO-2 FSAR as design bases.

The NRC staff also performed independent calculations to verify the conservatism of certain parameters used by the licensee. The major parameters and assumptions used by the licensee and found acceptable to the staff are presented in Table 3.2.2 in this SE. The results of the licensee's design basis radiological consequence calculation are provided in Table 3.2 in this

- 15 SE. The staff concludes that the EAB, LPZ, and control room doses estimated by the licensee for the FHA accident meet the applicable accident dose criteria and are, therefore, acceptable.

3.3.3 Main Steam Line Break (MSLB) Accident The current ANO-2 design basis MSLB analysis is based on the traditional accident source term described in TID-14844. The ANO-2 licensing basis analysis is presented in FSAR Section 15.1.14.1, "Steam Line Break Accident." To support implementation of the AST, as requested by the subject LAR, the licensee reanalyzed the offsite and control room radiological consequences of the postulated MSLB. This reanalysis was performed to demonstrate that the ESFs designed to mitigate the radiological consequences at ANO-2 will remain adequate after implementation of the AST.

The licensee submitted the AST-based reanalysis of the MSLB accident as part of the LAR.

Included in this reanalysis are the assumptions, parameters, and newly calculated offsite and control room doses associated with implementing the AST methodology. The licensee cites RG 1.183 as providing the primary radiological analysis assumptions for its reanalysis of the postulated design basis MSLB accident. Specifically, the NRC staff's guidance is detailed in RG 1.183, Appendix E, "Assumptions for Evaluating the Radiological Consequences of a PWR Main Steam Line Break Accident."

3.3.3.1 Activity Source The licensee postulated no fuel failure to occur as a result of this design basis MSLB accident.

Therefore, activity from defective and leaking fuel was assumed to be introduced instantaneously and homogenously into the primary coolant. As a design basis, the licensee considered following two cases of activity spiking in order to determine the maximum offsite and control room dose:

• Pre-existing Iodine Spike (PIS) Case: For this case, the licensee assumed that a reactor transient has occurred prior to the postulated MSLB and has raised the primary reactor coolant system (RCS) iodine concentration to 60 times the new TS limit of 1.0 jJCi/gm (microcuries per gram) dose equivalent lodine-131 (DE 1-131), as is consistent with the guidance of RG 1.183, when no fuel failure is assumed.

• Accident-initiated Iodine Spike (AIS) Case: For this case, the licensee assumed that the RCS transient associated with the MSLB causes an iodine spike in the primary RCS. It is assumed that the iodine release rate from the fuel rods to the primary RCS increases to a value 500 times greater than the release rate corresponding to the iodine concentration at the equilibrium value, as is consistent with the guidance of RG 1.183, when no fuel failure is assumed. The licensee conservatively assumed a spike duration of 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />; that is, the iodine activity increases during the 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> of the transient as a result of release from the defective fuel at the spiked iodine equilibrium appearance rate. The 8-hour assumption is consistent with the guidance of RG 1.183.

-16 In addition, for the two cases considered, the licensee also assumed that the TS maximum secondary coolant iodine concentration is available for release, consistent with RG 1.183 guidance. As shown in the ANO-2 TS, and as input to the MSLB analysis, the maximum secondary coolant iodine concentration at ANO-2 is 0.1 IJCi/gm DE 1-131.

Consistent with RG 1.183 guidance, the licensee assumed that the radioactive iodine speciation released from failed fuel is 95.00 percent aerosol (particulate), 4.85 percent elemental, and 0.15 percent organic. Whereas, the radioactive iodine speciation released from the steam generators (SGs) is 97.00 percent elemental and 3.00 percent organic.

3.3.3.2 Transport Methodology and Assumptions (consistent with RG 1.183)

The design basis MSLB accident is typically defined as the pre-trip double-ended rupture of a main steam line outside containment. The licensee modeled unimpeded blowdown of steam from the broken steam line, whereby the affected SG rapidly depressurizes and releases its initial contents, and plant cooldown is then achieved via the remaining unaffected SG. The energy removal from the RCS causes a reduction of coolant temperature and pressure, and steam releases occur until shutdown cooling (SOC) is initiated at 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. Activity is transported from the primary RCS to the secondary side, and made available for release to the environment, through SG tube leakage, also referred to as primary-to-secondary leakage.

The licensee conservatively assumed that the entire maximum coolant inventory initially in the faulted SG, and all of its associated activity, is released to the environment within 1 minute. The licensee also used the maximum value for this pathway in order to maximize the total activity release. Then, consistent with the current ANO-2 design basis, a maximum primary-to secondary leakage rate of 0.5 gallons per minute (gpm) per SG was assumed. All primary-to secondary leakage into the faulted SG and all noble gas activity associated with this leakage were assumed to be flashed, or vaporized, and released directly to the environment. Primary to-secondary leakage was assumed to continue until the RCS temperature was reduced to 212 OF at 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. No credit was taken for holdup, reduction or mitigation through this pathway.

For the release of activity through the intact SG, the licensee assumed a typical moisture carryover fraction of 0.1 percent, which equates to a partitioning coefficient of 0.001. The licensee calculated moisture carryover to be 0.06 percent; therefore, the selection of this typical value is conservative. For the time-sensitive primary-to-secondary leakage into the intact SG, the licensee assumed the minimum liquid coolant volume was present, thereby increasing the activity concentration of the release. The licensee calculated that an assumption of 5 percent was appropriate for the amount of primary-to-secondary leakage into the intact SG that flashes to vapor. The licensee based this assumption on the calculation of the flashing fraction at the start of tube uncovery and re-initiation of auxiliary feedwater approximately 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> later. Using a constant enthalpy equation, the licensee calculated a flashing fraction of 0.0762 at time equal to zero. At time equal to 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, the licensee calculated that the RCS and SG are near an equilibrium pressure due to depressurization as a result of the steam line break, resulting in effectively no flashing. Assuming a linear relationship, and taking the average between these times, the licensee determined that the overall effective flashing fraction would be 0.0381. The licensee then assumed 5 percent flashing for the 8-hour release duration for conservatism. The licensee assumed the remaining 95 percent of the 0.5 gpm primary-to-secondary leakage to mix

- 17 with the minimized coolant mass of the intact SG and be released by steaming associated with the actions to reach SOC (in 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />).

The flashing fraction and release assumed by the licensee was based on a calculation using thermodynamic and physical principles, and adequately accounts for uncertainty, which is in accordance with RG 1.183. Also, by selecting limiting parameters and by assuming maximized releases, the licensee conservatively modeled MSLB accident activity transport. The NRC staff reviewed the methods, parameters, and assumptions that the licensee used in its radiological dose consequence analyses and concludes that they are consistent with the conservative guidance provided in RG 1.183, and, therefore, acceptable.

3.3.3.3 Summary The licensee concluded that the radiological consequences at the EAB, LPZ, and control room are within the dose criteria specified in 10 CFR 50.67 and the accident-specific dose acceptance criteria specified in SRP 15.0.1 and RG 1.183. These accident-specific dose acceptance criteria for the MSLB assuming a PIS are a TEOE of 25 rem at the EAB for the worst 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, 25 rem at the outer boundary of the LPZ, and 5 rem for the control room. For the MSLB assuming an AIS, the accident-specific dose acceptance criteria are a TEOE of 2.5 rem at the EAB for the worst 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, 2.5 rem at the outer boundary of the LPZ, and 5 rem for the control room. Based on the above discussion of the licensee's MSLB analysis, the NRC staff concludes that the licensee used analysis assumptions and inputs consistent with applicable regulatory guidance identified in Section 2.0 of this SE and with those stated in the ANO-2 FSAR as design bases. The staff also performed an independent calculation of the dose consequences of the MSLB accident using the licensee's assumptions for input to the RAOTRAO computer code. The staff's calculation confirmed the licensee's dose results. The major parameters and assumptions used by the licensee and found acceptable to the staff are presented in Table 3.2.3 in this SE. The results of the licensee's design basis radiological consequence calculation are provided in Table 3.2 in this SE. The NRC staff concludes that the EAB, LPZ, and control room doses estimated by the licensee for the MSLB accident meet the applicable accident dose criteria and are, therefore, acceptable.

3.3.4 Steam Generator Tube Rupture Accident The current ANO-2 design basis SGTR accident analysis is based on the traditional accident source term described in TIO-14844. The ANO-2 licensing basis analysis is presented in FSAR Section 15.1.18, "Steam Generator Tube Rupture With or Without a Concurrent Loss of AC

[Alternating Current] Power." To support implementation of the AST, as requested by the subject LAR, the licensee reanalyzed the offsite and control room radiological consequences of the postulated SGTR. This reanalysis was performed to demonstrate that the ESFs designed to mitigate the radiological consequences at ANO-2 will remain adequate after implementation of the AST.

The licensee submitted the AST-based reanalysis of the SGTR accident as part of the LAR.

Included in this reanalysis are the assumptions, parameters, and newly calculated offsite and control room doses associated with implementing the AST methodology. The licensee cites RG 1.183 as providing the primary radiological analysis assumptions for its reanalysis of the

- 18 postulated design basis SGTR accident. Specifically, the NRC staffs guidance is detailed in RG 1.183, Appendix F, "Assumptions for Evaluating the Radiological Consequences of a PWR Steam Generator Tube Rupture Accident."

3.3.4.1 Activity Source The licensee determined that there is no fuel failure associated with the design basis SGTR accident. As a result, the licensee considered the following two cases of postulated activity release following a design basis SGTR, in order to determine the maximum offsite and control room dose:

• Pre-existing Iodine Spike (PIS) Case: For this case, the licensee assumed that a reactor transient has occurred prior to the postulated MSLB and has raised the primary RCS iodine concentration to 60 times the TS limit of 1.0 jJCi/gm DE 1-131, as is consistent with the guidance of RG 1.183, when no fuel failure is assumed.

• Accident-initiated Iodine Spike (AIS) Case: For this case, the licensee assumed that the RCS transient associated with the MSLB causes an iodine spike in the primary RCS. It is assumed that the iodine release rate from the fuel rods to the primary RCS increases to a value 335 times greater than the release rate corresponding to the iodine concentration at the equilibrium value, as is consistent with the guidance of RG 1.183, when no fuel failure is assumed. Also consistent with RG 1.183 guidance, the spike duration persists for a period of 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

In addition, as part of this LAR, the licensee has proposed to modify its TS limit on the allowable Dose Equivalent X-133 (DEX) concentration in the primary RCS. Therefore, in its accident analysiS, the licensee assumed the newly proposed limit of 3100 jJCi/gm for the noble gas activity available for release.

Consistent with RG 1.183 guidance, the licensee assumed that the radioactive iodine speciation released from the SGs is 97.00 percent elemental and 3.00 percent organic.

3.3.4.2 Transport Methodology and Assumptions Typically, the limiting SGTR accident is analyzed as a complete double-ended tube break that is postulated to occur due to a complete failure of a tube-to-sheet weld or the rapid propagation of a circumferential crack. The SGTR allows primary coolant to leak into the secondary system via the SG. The primary coolant transfer causes the pressurizer level to decrease, provided that the tube leak rate exceeds the charging pump capacities and causes the level in the affected SG to increase. In the event of a coincident loss of offsite power, or failure of the condenser steam dump system, discharge of activity to the atmosphere takes place via the MSSVs and/or ADV.

For a double-ended SGTR, the leak rate typically exceeds the charging pump capacities and, consequently, the pressurizer level will decrease. The decrease in the pressurizer level and the inability of the heaters to maintain pressurizer pressure causes the RCS pressure to decrease.

- 19 The drop in the pressure will cause a reactor trip and ensure that the departure from nucleate boiling (ONB) fuel design limit is not exceeded. The licensee assumed no fuel failure resulting from a design basis SGTR accident at ANO-2.

For the SGTR at ANO-2, the licensee assumed that primary-to-secondary leakage continues until after the SOC system has been placed into service and both SGs have been isolated. The affected SG was conservatively assumed to be isolated by operator action after 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. This isolation would effectively terminate releases from the affected SG, while primary-to-secondary leakage would continue to provide activity for release from the unaffected SG via its AOV until the SOC system is placed in service and SOC (RCS temperature less than 212 OF) is achieved.

The licensee conservatively assumed this happens within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

As stated in its submittal, the licensee assumed that a portion of the primary-to-secondary leakage and break flow through the SG tubes flashes, or vaporizes, based upon the thermodynamic conditions in the primary and secondary coolant. The licensee calculated this flashing fraction to be a maximum of 0.05, or 5 percent, of the flow from primary-to-secondary leakage and ruptured tube flow. The leakage that immediately flashes to vapor was assumed to be immediately released to the environment with no mitigation (i.e., no reduction for scrubbing within the SG bulk water was credited.) Partitioning of this release to the condenser would typically be credible, but was not credited by the licensee. The licensee also calculates, but does not credit, that operator action to isolate the affected SG and open the AOV on the intact SG occurs at 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, and at the same time the SG level is recovered, thus covering the tubes and preventing further flashing. All leakage that does not immediately flash, which is the remaining 95 percent of the leakage and break flow, was assumed to mix with the secondary side bulk water. The radioactivity within this secondary side coolant was assumed to become vapor at a rate that is the function of the steaming rate and the partition coefficient. A partition coefficient for iodine of 100 was assumed for the leakage from the bulk water. This value is consistent with the guidance of RG 1.183. The licensee assumed that all noble gas activity is released directly to the environment.

The licensee's model is consistent with the guidance of RG 1.183, and in addition, by selecting limiting parameters and by assuming maximized releases, the licensee conservatively modeled SGTR accident activity transport. Based on the above, the NRC staff concludes this release methodology to be acceptable.

3.3.4.3 Summary The licensee concluded that the radiological consequences at the EAB, LPZ, and control room are within the dose criteria specified in 10 CFR 50.67 and the accident-specific dose acceptance criteria specified in SRP Section 15.0.1 and RG 1.183. These accident-specific dose acceptance criteria for the SGTR assuming a PIS are a TEOE of 25 rem at the EAB for the worst 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, 25 rem at the outer boundary of the LPZ, and 5 rem for the control room. For the SGTR assuming an AIS, the accident-specific dose acceptance criteria are a TEOE of 2.5 rem at the EAB for the worst 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, 2.5 rem at the outer boundary of the LPZ, and 5 rem for the control room. Based on the above discussion of the licensee's SGTR analysis, the NRC staff concludes that the licensee used analysis assumptions and inputs consistent with applicable regulatory guidance identified in Section 2.0 of this SE and with those stated in the ANO-2

- 20 FSAR as design bases. The staff also performed an independent calculation of the dose consequences of the SGTR accident using the licensee's assumptions for input to the RADTRAD computer code. The staff's calculation confirmed the licensee's dose results. The major parameters and assumptions used by the licensee and found acceptable to the staff are presented in Table 3.2.4 in this SE. The results of the licensee's design basis radiological consequence calculation are provided in Table 3.2 in this SE. The NRC staff concludes that the EAB, LPZ, and control doses estimated by the licensee for the SGTR accident meet the applicable accident dose criteria and are, therefore, acceptable.

3.3.5 Locked Rotor Accident (LRA)

The current ANO-2 design basis LRA analysis is based on the traditional accident source term described in TID-14844. The ANO-2 licensing basis analysis is presented in FSAR Section 15.1.5, "Total and Partial Loss of Reactor Coolant Forced Flow." To support implementation of the AST, as requested by the subject LAR, the licensee reanalyzed the offsite and control room radiological consequences of the postulated LRA. This reanalysis was performed to demonstrate that the ESFs designed to mitigate the radiological consequences at ANO-2 will remain adequate after implementation of the AST.

The licensee submitted the AST-based reanalysis of the LRA as part of the LAR. Included in this reanalysis are the assumptions, parameters, and newly calculated offsite and control room doses associated with implementing the AST methodology. The licensee cites RG 1.183 as providing the primary radiological analysis assumptions for its reanalysis of the postulated design basis LRA. Specifically, the NRC staff's guidance is detailed in RG 1.1.8, Appendix G, "Assumptions for Evaluating the Radiological Consequences of a PWR Locked Rotor Accident."

3.3.5.1 Activity Source To determine the maximum offsite and control room dose resulting from the postulated design basis LRA, the licensee assumed 14 percent fuel cladding failure in its reanalysis. This failed fuel fraction is consistent with the current ANO-2 design basis, as shown ANO-2 FSAR Section 15.1.5.2.3.4, "Cycle 15 Loss of Coolant Flow Resulting from a Pump Shaft Seizure Radiological Assessment." The licensee's current licensing basis states that, based on the transient analyses, less than 14.0 percent of the fuel rods will fail; therefore, the use of 14.0 percent in the design basis LRA dose consequence analysis is conservative. The licensee stated that the initial primary and secondary coolant activity was not included in the source term for the ANO-2 LRA analysis because such activity is of negligible significance due to activity associated with the failed fuel being much larger in comparison. The NRC staff compared the initial primary and secondary coolant activity to the failed fuel activity and found the failed fuel activity to be at least ten times higher than both the initial primary and secondary activity.

Based on the above, the NRC concludes that the licensee's assumption is acceptable.

For the postulated LRA, the initial thermal power was assumed to be 3,086.52 MWth, which is a factor of 1.02 times the current licensed thermal power of 3,026 MWth, in order to account for the ECCS evaluation uncertainty. The licensee assumed the design basis radial peaking factor of 1.65. Additionally, the licensee accounted for current licensed values for fuel enrichment and burnup when determining the core isotopic inventory. Table 1.7.4-1 of Attachment 3 to the letter

- 21 dated March 31, 2010, shows the core inventory used by the licensee. The fraction of core isotopic activity the licensee assumed to be available for release from the gap of failed fuel (i.e.,

fuel experiencing cladding failure) is provided in Table 3 of RG 1.183. In its LAR submittal, the licensee stated that it has processes in place to ensure the gap fraction applicability restrictions in RG 1.183, Footnote 11! are met. Therefore, upon approval of this LAR, the ANO-2 design basis will be analyzed for fuel burnup and enrichment values specified in Section 3.3 of this SE.

These values meet the RG 1.183 limits of applicability expressed in its footnote 11.

Consistent with RG 1.183 guidance, the licensee assumed that the radioactive iodine speciation released from failed fuel is 95.00 percent aerosol (particulate), 4.85 percent elemental, and 0.15 percent organic. Whereas, the radioactive iodine speciation released from the SGs is 97.00 percent elemental and 3.00 percent organic.

3.3.5.2 Transport Methodology and Assumptions The licensee has defined the design basis LRA as the instantaneous seizure of a primary reactor coolant pump (RCP) rotor. Typically, such a seizure is postulated to occur due to a mechanical failure or a loss of component cooling to the pump shaft seals. Reactor coolant flow would then be rapidly reduced to a three-pump value. Because a rapid reduction in coolant flow results in a rapid reduction in the margin to DNB, the licensee postulated that a low departure from nucleate boiling ratio (DNBR) trip would occur. The licensee also postulated that, at the beginning of the LRA (i.e., at the time the rotor in one of the RCPs is assumed to seize), the plant is assumed to be in operation under adverse steady-state operating conditions. The licensee determined these conditions to be maximum steady-state power level, maximum steady-state pressure, and maximum steady-state coolant average temperature.

Following initiation of the reactor trip, heat stored in the fuel rods continues to be transferred to the RCS causing the coolant to expand. At the same time, heat transfer to the shell side of the SGs is reduced, first because the reduced flow results in a decreased tube side film coefficient, and then because the reactor coolant in the tubes cools down while the shell side temperature increases (turbine steam flow is reduced to zero upon plant trip). The rapid expansion of the coolant in the reactor core, combined with reduced heat transfer in the SGs causes an insurgence into the pressurizer and a pressure increase throughout the RCS.

ANO-2 uses both SGs for plant cooldown. The licensee assumed a portion of the primary-to secondary leakage to flash to vapor based on the thermodynamic conditions in the primary and secondary coolants. To address iodine transport for release from the ANO-2 SGs, a flashing fraction model was developed, as was done for the MSLB accident analysis, and a conservative flashing fraction of 0.05 was selected to bound that which was calculated. The flashed portion of the primary-to-secondary leakage was assumed to be released directly to the environment with no credit for partitioning in the SG. The remaining 95 percent of the primary-to-secondary activity leakage that is discharged as liquid was assumed to be mixed with the SG secondary side coolant inventory and released to the environment with partitioning via steam releases from the bulk fluid in the SG; which is consistent with the guidance of RG 1.183. The SG tubes were assumed to be partially uncovered during the first hour. The licensee assumed operator action is taken at 30 minutes to open the ADVs thereby allowing SG level to be recovered within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> of event initiation. The licensee did not take credit for mitigation of released activity by

- 22 scrubbing due to the submergence of the tubes after the first hour; the licensee assumed the 5 percent flashing rate to persist for the duration of the cooldown. The radioactivity within the bulk water was assumed to become vapor at a rate that is a function of the steaming rate and the partition coefficient. The licensee assumed a partition coefficient of 100 for iodine. As is discussed in RG 1,183, the retention of particulate radionuclides in the SG is limited by the moisture carryover from the SG. The licensee conservatively assumed a partition coefficient of 0.001 for alkali metals. Noble gases were assumed to be released directly to the environment without credit for dilution, holdup, or mitigation of any kind.

The NRC staff determined that the licensee's treatment of the LRA activity release and transport through the SG pathways is conservative and consistent with its current design basis and the guidance shown in Appendix G of RG 1.183. Based on the above, the NRC staff concludes that the licensee's model is acceptable.

3.3.5.3 Summary The licensee concluded that the radiological consequences at the EAB, LPZ, and control room are within the dose criteria specified in 10 CFR 50.67 and the accident-specific dose acceptance criteria specified in SRP Section 15.0.1 and RG 1.183. The accident-specific dose acceptance criteria for the LRA at ANO-2 are a TEDE of 2.5 rem at the EAB for the worst 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, 2.5 rem at the outer boundary of the LPZ, and 5 rem for access to and occupancy of the control room. Based on the above discussion of the licensee's LRA analysis, the NRC staff concludes that the licensee used analysis assumptions and inputs consistent with applicable regulatory guidance identified in Section 2.0 of this SE and with those stated in the ANO-2 FSAR as design bases. The staff also performed an independent calculation of the dose consequences of the LRA using the licensee's assumptions for input to the RADTRAD computer code. The staff's calculation confirmed the licensee's dose results. The major parameters and assumptions used by the licensee and found acceptable to the staff are presented in Table 3.2.5 in this SE. The results of the licensee's design basis radiological consequence calculation are provided in Table 3.2 in this SE. The NRC staff concludes that the EAB, LPZ, and control room doses estimated by the licensee for the LRA meet the applicable accident dose criteria and are, therefore, acceptable.

3.3.6 Control Rod Ejection Accident (CREA)

The current ANO-2 design basis CREA analysis is based on the traditional accident source term described in TID-14844. The ANO-2 licensing basis analysis is presented in FSAR Section 15.1.20, "Control Element Assembly Ejection." To support implementation of the AST, as requested by the subject LAR, the licensee reanalyzed the offsite and control room radiological consequences of the postulated CREA. This reanalysis was performed to demonstrate that the ESFs designed to mitigate the radiological consequences at ANO-2 will remain adequate after implementation of the AST.

The licensee submitted the AST-based reanalysis of the CREA as part of the LAR. Included in this reanalysis are the assumptions, parameters, and newly calculated offsite and control room doses associated with implementing the AST methodology. The licensee cites RG 1.183 as providing the primary radiological analysis assumptions for its reanalysis of the postulated

- 23 design basis CREA. Specifically, the NRC staff's guidance is detailed in RG 1.183, Appendix H "Assumptions for Evaluating the Radiological Consequences of a PWR Rod Ejection Accident."

3.3.6.1 Activity Source To determine the maximum offsite and control room dose resulting from the postulated design basis CREA, the licensee determined that 14.0 percent of the fuel experiences DNB and is therefore assumed to fail; however, the licensee determined that no fuel melts. This assumption was applied to both accident scenarios. This failed fuel fraction is consistent with the current ANO-2 design basis shown in ANO-2 FSAR Section 15.1.20.4.7, "Cycle 16 CEA [Control Element Assembly] Ejection Analysis," as calculated for the cycle 16 control element assembly ejection analysis. For the postulated CREA, the initial thermal power was assumed to be 3,086.52 MWth, which is a factor of 1.02 times the current licensed thermal power of 3,026 MWth, in order to account for the ECCS evaluation uncertainty. The licensee assumed the design basis radial peaking factor of 1.65, consistent with its current design basis, and accounted for current licensed values for fuel enrichment and burnup when determining the core isotopic inventory. The licensee did not calculate the contribution from primary RCS or secondary coolant releases in addition to the release of core activity. The NRC staff compared the primary RCS and secondary coolant activity release to the failed fuel activity and found the 14 percent failed fuel activity release to be at least ten times higher than both the initial primary and secondary activity. Based on the above, the NRC concludes that the licensee's assumption is acceptable.

The licensee assumed that 10 percent of the noble gas, 10 percent of the iodine, and 12 percent of the alkali metals are released, representing all of the available gap activity, and consistent with the guidance of RG 1.183. In its LAR submittal, the licensee stated that it has processes in place to ensure the gap fraction applicability restrictions in RG 1.183, footnote 11, are met. Therefore, upon approval of this LAR, the ANO-2 design basis will be analyzed for fuel burnup and enrichment values specified in Section 3.2 of this SE. These values meet the RG 1.183 limits of applicability expressed in its footnote 11.

Consistent with RG 1.183 guidance, the licensee assumed that the radioactive iodine speciation released from failed fuel is 95.00 percent aerosol (particulate), 4.85 percent elemental, and 0.15 percent organic. Whereas, the radioactive iodine speciation released from the SGs is 97.00 percent elemental and 3.00 percent organic.

3.3.6.2 Transport Methodology and Assumptions In Section 2.6.1 of Attachment 3 to the letter dated March 31, 2010, the licensee has defined the ANO-2 design basis CREA as the uncontrolled withdrawal of a single control rod that results in a reactivity insertion that leads to a core power increase and subsequent reactor trip.

The licensee has identified two separate credible scenarios for activity release to the environment resulting from the postulated rod ejection accident. The two scenarios are as follows: (1) a 100 percent release of activity from the damaged fuel that instantaneolJsly and homogeneously mixes with the containment atmosphere, as can be attributed to a postulated breach of the reactor pressure vessel (RPV); or (2) the dissolution of 100 percent of the activity

- 24 from the damaged fuel in the primary coolant, which is then made available for release to the secondary side through primary-to-secondary leakage. Through analysis, the first CREA case, the containment leakage release, was determined to be bounding. This is consistent with Appendix H of RG 1.183.

3.3.6.2.1 Containment Release Case 3.3.6.2.1.1 Containment Leakage For the containment leakage release, the ejected control rod would effectively cause the equivalent of a small break LOCA by breaching the RPV. The licensee assumed that the activity leaks from the containment atmosphere to the environment at the design basis containment leak rate for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of the accident. The current ANO-2 design basis containment leak rate, La, is equal to 0.1 percent per day, at containment peak pressure, as expressed in ANO-2 FSAR Table 15.1.13-1. The licensee then assumed that, based on the containment pressure decreasing over time, the leak rate is reduced to 50 percent of the TS value (0.05 percent per day) at 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and for the duration of the accident. This is consistent with the guidance of RG 1.183. Though the licensee did not credit removal of activity by containment sprays, it still separated the containment volume into two individually well-mixed compartments, one "sprayed" and one "unsprayed." In the licensee's model, mixing between the two volumes is accomplished by the containment fan coolers that provide 11,880 cfm of mixing flow between the sprayed and unsprayed regions of the ANO-2 containment. The guidance of RG 1.183 allows for an assumption of a mixing rate equal to two turnovers of the unsprayed volume per hour, due to natural convection. Because the volume of the ANO-2 unsprayed region determined by the licensee is 392,000 ft3, the natural convection mixing rate could potentially have been calculated to be approximately 13,000 cfm, which is considerably more than the mixing rate assumed by the licensee. The licensee's model for achieving a well mixed containment atmosphere is conservative and consistent with the appropriate regulatory guidance and is, therefore, acceptable to the NRC staff.

The licensee took credit for sedimentation of aerosols. Conservatively, the licensee takes no credit for the removal of released activity by the containment sprays or the internal containment recirculation HEPA filters. These assumptions are consistent with the appropriate regulatory guidance and are, therefore, acceptable to the NRC staff.

3.3.6.2.1.2 ESF Recirculation Leakage For ANO-2, it was assumed that leakage of coolant from ESF systems is possible upon initiation of those systems. The ANO-2 TSs do not provide a specific limitation on the total amount of operational leakage that is allowed from ESF systems. However, as a current design basis, the licensee assumed, consistent with the analyses discussed in Section 0 of this evaluation, that the total leakage from ECCS is equal to 2,060 cc/hr, assuming two containment spray and three high pressure safety injection (HPSI) pumps in operation. In accordance with the guidance of RG 1.183, the licensee assumed twice this amount (4,120 cc/hr) in analyzing the dose consequence from the ESF leakage pathway.

- 25 The licensee's DBA analysis assumed leakage to start at 1,560 seconds, which the licensee determined to be the time that the recirculation phase begins, and is consistent with AST activity release assumptions. The leakage was assumed to continue for the duration of the accident.

The licensee conservatively took no credit for holdup or filtration of this leakage by assuming that the leaked activity is immediately available for release to the atmosphere.

The licensee assumed that, with the exception of iodine, all radioactive material in the recirculating coolant is retained in the liquid phase, and also in accordance with RG 1.183, the chemical form of the released iodine activity is assumed to be 97.00 percent elemental and 3.00 percent organic. The licensee determined the maximum ANO-2 sump temperature to be greater than 212 OF, and therefore, assuming constant enthalpy, applied the equation shown in RG 1.183, Appendix A, Section 5.4 to calculate the flashing fraction of leaked coolant. The licensee determined this value to be 2.29 percent; however, in accordance with RG 1.183 guidance, a conservative flashing fraction of 10 percent is used in the licensee's analysis.

The licensee's treatment of ESF leakage is consistent with its current design basis and the applicable regulatory guidance, and the NRC staff concludes it to be acceptable.

3.3.6.2.1.3 Shine Contribution For shine dose resulting from an activity cloud internal to the ANO-2 containment, the licensee stated that the contribution to the total CREA dose for this case would be negligible. This assumption was based upon the internal containment cloud shine dose calculated for the LOCA and a comparison between the LOCA and CREA source terms. The licensee stated, and the NRC staff agrees, that the activity released into the containment following a CREA is far less than that released to the containment during a LOCA. This is because the CREA assumes only 14 percent fuel damage and a percent fuel melt, while the LOCA assumes 100 percent fuel melt. All transport characteristics between the two events would be identical. Therefore, based on the above discussion, the NRC staff concludes that the shine contribution due to the CREA is bounded by that due to the LOCA. Therefore, it is acceptable for this DBA analysis to not include a dose contribution from internal containment cloud shine.

For the calculation of external cloud shine dose to the control room, the licensee applied the source term that was calculated using the model of accident releases into containment and out to the environment. As modeled by the licensee in its submitted CREA design analysis, airborne activity that leaks from containment results in a radioactive plume external to the ANO-2 control room. During the postulated accident, this cloud will in turn result in additional dose to the control room operator inside the control room due to shine. The licensee calculated the dose contribution from this external plume by using the simulations that were used to calculate the control room doses due to airborne activity transport with slight modifications. The licensee created new DCF input files to apply to these existing models. The new DCF input files set inhalation DCFs to zero, and adjusted the external exposure DCFs by multiplying them by a calculated attenuation factor associated with the concrete of the ANO-2 control room wall. This method will result in the RADTRAD code to output dose calculations resulting from direct activity shine only; the ingestion and inhalation component of the dose has been ignored. Also, it is mathematically appropriate and analytically acceptable to apply the control room wall attenuation factor directly to the DCF. Based on the above, the NRC staff concludes that the

- 26 licensee's method for calculating the external plume shine to control room personnel is acceptable.

3.3.6.2.2 Secondary RCS Release Case For the secondary RCS release pathway, the licensee did not assume that the RPV was breached following the rod ejection. In this case, the licensee assumed that RCS integrity was maintained and all activity from damaged fuel that has been mixed with the RCS leaks to the secondary RCS through the SG tubes at a rate of 150 gallons per day (gpd) per SG, for a total of 300 gpd, as limited by ANO-2 TS 3.4.6.2. To address iodine transport for release from the ANO-2 SGs, a flashing fraction model was developed, as was done for the MSLB accident analysis, and a conservative flashing fraction of 0.05 (which is double the calculated flashing fraction) was selected to bound that which was calculated. The flashed portion of the primary to-secondary leakage was assumed to be released directly to the environment with no credit for partitioning in the SG. The remaining 95 percent of the primary-to-secondary activity leakage that is discharged as liquid was assumed to be mixed with the SG secondary side coolant inventory and released to the environment with partitioning via steam releases from the bulk fluid in the SG; which is consistent with the guidance of RG 1.183. The SG tubes were assumed to be partially uncovered during the first hour. The licensee assumed operator action is taken at 30 minutes to open the ADVs thereby allowing SG level to be recovered within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> of event initiation. The licensee did not take credit for reduction in mitigation of released activity by scrubbing due to the submergence of the tubes after the first hour; the licensee assumed the 5 percent flashing rate to persist for the duration of the cooldown. The radioactivity within the bulk water was assumed to become vapor at a rate that is a function of the steaming rate and the partition coefficient. The licensee assumed a partition coefficient of 100 for iodine, which is consistent with RG 1.183. As is discussed in RG 1,183, the retention of particulate radionuclides in the SG is limited by the moisture carryover from the SG. The licensee conservatively and appropriately assumed a partition coefficient of 0.001 for alkali metals. Noble gases were assumed to be released directly to the environment without credit for dilution, holdup, or mitigation of any kind.

The licensee's treatment of this case of CREA activity release and transport through the SG pathway is conservative and consistent with its current design basis and the guidance shown in Appendix H of RG 1.183. Based on the above, the NRC staff concludes that the licensee's model is acceptable.

3.3.6.3 Summary The licensee concluded that the radiological consequences at the EAB, LPZ, and control room are within the dose criteria specified in 10 CFR 50.67 and the accident-specific dose acceptance criteria specified in SRP Section 15.0.1 and RG 1.183. The accident-specific dose acceptance criteria for the CREA at ANO-2 are a TEDE of 6.3 rem at the EAB for the worst 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, 6.3 rem at the outer boundary of the LPZ, and 5 rem for access to and occupancy of the control room. The NRC staff concludes that the licensee used analysis assumptions and inputs consistent with applicable regulatory guidance identified in Section 2.0 of this SE and with those stated in the ANO-2 FSAR as design bases. Based on the above discussion of the licensee's CREA analysis, the staff also performed an independent calculation of the dose

- 27 consequences of the CREA using the licensee's assumptions for input to the RADTRAD computer code. The staff's calculation confirmed the licensee's dose results. The major parameters and assumptions used by the licensee and found acceptable to the staff are presented in Table 3.2.6 in this SE. The results of the licensee's design basis radiological consequence calculation are provided in Table 3.2 in this SE. The staff concludes that the EAB, LPZ, and control room doses estimated by the licensee for the CREA meet the applicable accident dose criteria and are, therefore, acceptable.

3.3.7 Control Room Habitability and Modeling The current ANO-2 non-LOCA DBA analyses, as shown in FSAR Chapter 15, do not calculate control room dose; therefore, the control room dose model provided in the revised DBA accident analyses that support this AST-based LAR, represents a change in the ANO-2 licensing basis.

For its revised analyses, the licensee credits post-accident control room isolation with filtered recirculation and pressurization only after the emergency operation mode is initialized. The control room ventilation system is designed and modeled as providing 35,200 cfm of unfiltered makeup air in normal operation mode and 333 cfm of filtered makeup air in emergency operation mode. As stated by the licensee, this emergency operation mode filtration system provides 99 percent filtration of aerosol, elemental, and organic forms of iodine. In addition to makeup air flow, the ANO-2 design basis credits 1,667 cfm of filtered recirculation flow, again, only after the emergency operation mode is initialized. The filtration associated with the recirculation system is 99 percent removal of aerosol activity and 95 percent for elemental and organic forms of iodine. For its analyses, the licensee assumed that recirculation flow is initiated immediately upon isolation of the normal control room intake, which was assumed to occur at 10 seconds following the initiation of the accident. Also, upon isolation of the normal control room intake, the licensee took credit for the filters associated with the control room emergency operation mode for the duration of the DBA.

The total bounding unfiltered inleakage into the ANO-2 control room assumed by the licensee was 82 cfm. This value was also assumed for the accident duration. In a letter dated August 28, 2003 (ADAMS Accession No. ML032450205), in response to NRC Generic Letter 2003-01, "Control Room Habitability," the licensee indicated to the NRC staff that the maximum measured unfiltered inleakage into the ANO-2 control room was 61 cfm. Therefore, the modeled unfiltered inleakage value is conservative and provides margin for future measurements of control room inleakage.

3.3.8 Summary As described above, the NRC staff reviewed the assumptions, inputs, and methods used by the licensee to assess the radiological consequences of the postulated DBA analyses with the proposed TS changes. The staff concludes that the licensee used analysis methods and assumptions consistent with the conservative regulatory requirements and guidance identified in Section 2.0. The staff compared the doses estimated by the licensee to the applicable criteria identified in Section 2.0. The staff also concludes, with reasonable assurance, that the licensee's estimates of the EAB, LPZ, and control room doses will comply with these criteria.

The staff further concludes with reasonable assurance that ANO-2, as modified by this

- 28 approved license amendment, will continue to provide sufficient safety margins, with adequate defense-in-depth, to address unanticipated events and to compensate for uncertainties in accident progression, analysis assumptions, and input parameters. Therefore, the staff concludes that proposed license amendment is acceptable with respect to the radiological consequences of the DBAs.

3.4 Containment Sump pH and Iodine Spray Removal Coefficients 3.4.1 Technical Evaluation After a LOCA, a variety of different chemical species are released from the damaged core. One of them is radioactive iodine. This iodine, when released to the outside environment, will significantly contribute to radiation doses. It is, therefore, essential to keep it confined within the plant's containment. According to NUREG-1465, "Accident Source Terms for Light-Water Nuclear Power Plants," February 1995, iodine is released from the core in three different chemical forms; at least 95 percent is released in ionic form as cesium iodide (Csl) and the remaining 5 percent as elemental iodine (Iz) and hydriodic acid (HI); the release contains at least 1 percent of each 12 and HI. Csi and HI are ionized in water and are, therefore, soluble.

However, elemental iodine is scarcely soluble. To sequester the iodine in water, it is desirable to maintain as much as possible of the released iodine in ionic form. In radiation environments existing in containment, some of the ionic iodine dissolved in water is converted into elemental form. The degree of conversion varies significantly with the pH of water. At a higher pH, conversion to elemental form is lower and at pH >7 it becomes negligibly small. The relationship between the rate of conversion and pH is specified in Figure 3.1 of NUREG/CR-5950, "Iodine Evolution and pH ControL" ANO-2 uses sodium tetraborate to control post-LOCA sump pH. A supplemental letter dated June 24, 2010, provided clarifying information on the current application. In addition, the NRC staff has previously reviewed the licensee's post-LOCA sump pH analysis as part of a prior license amendment. Additional information from the prior license amendment is available in a request for additional information response from the licensee dated February 19, 2008 (ADAMS Accession No. ML080580200). The licensee's analysis considered boron concentrations and volumes for the refueling water tank (RWT), the Boric Acid Makeup tank (BAM), the Safety Injection Tank (SIT), and RCS. Additional inputs included the minimum quantity of buffering agent contained in the sodium tetra borate baskets, dissolution time of the sodium tetra borate and the impact of strong acids generated by radiation of cable insulation and sump water.

The mechanism for strong acid generation from cable insulation and sump fluids under accident radiation levels are detailed in NUREG/CR-5950, "Iodine Evolution and pH Control". The licensee calculated that the bounding strong acid quantities were 3,473.6 moles for hydrochloric acid and 147.1 moles for nitric acid based on a 3~-day integrated dose for a LOCA. The minimum pH evaluation used the maximum borated water source volumes and concentrations with the minimum quantity of sodium tetraborate available, as specified in ANO-2 TS of 308 ft3 (15,000 Ibs) at a density of 48.7 Ibmlft3. The licensee's calculation determined that under LOCA conditions, including strong acid generation, worse case boron concentrations and sump fluid volumes, an equilibrium sump pH of 7.0 or greater would be maintained. Additionally, the licensee administratively controls the sodium tetraborate basket level at a minimum level of

- 29 443 fe (21 ,561 Ibs) at a density of 48.7 Ibm/fe. This additional quantity of sodium tetraborate provides an additional margin to maintain pH above 7.0 during the time period following a LOCA.

The NRC staff reviewed the licensee's assumptions and analysis and concluded that these are in accordance with the guidance provided in RG 1.183, and conservative values were used for the key parameters of the calculation. In addition the staff verified that the current ANO-2 TS requirements for sodium tetra borate volume will ensure sufficient buffering of the sump pool such that the pH will not drop below 7.0.

3.4.2 Summary The NRC staff reviewed the licensee's assumptions to minimize iodine re-evolution as presented in the re-analysis of the radiological consequences for a LOCA. The methodology relies on using buffering action of sodium tetra borate. The assumptions are appropriate and consistent with the methods accepted by the staff for the calculation of post-accident containment sump pH. In addition, the staff verified that the post-accident containment sump pH will be maintained above 7.0 for 30 days following a LOCA. Based on the above, the NRC staff concludes that the assumptions used are in accordance with the guidance provided in RG 1.183, and the post-accident containment sump pH will be maintained above 7.0 for 30 days following a LOCA.

3.5 Environmental Qualifications of Electrical Equipment 3.5.1 Technical Evaluation In this section of the SE, the NRC staff has reviewed the impact of the use of an AST methodology, if any, on the electrical safety-related systems and environmental qualifications (EQ) of electrical equipment.

In Section 2.7 of Attachment 3 to the letter dated March 31,2010, the licensee stated that RG 1.183, Regulatory Position 6 allows the use of either the AST or TID-14844 assumptions for performing the required EQ analysis until such time that a generic issue related to the effect of increased cesium releases on EQ doses is resolved. The ANO-2 EQ analyses will continue to be based on TID-14844 assumptions. The NRC staff concludes this position is acceptable.

In its letter dated August 9,2010, the licensee confirmed that no components are added to the ANO-2 10 CFR 50.49 program as a result of the AST adoption. Since the licensee will continue to use the TID-14844 assumptions for performing EQ analysis, and no new equipment is added to its 10 CFR 50.49 program, the NRC staff concludes that the current EQ of electrical equipment will remain valid for the purpose of implementation of an AST.

Additionally, the licensee stated that no non-safety-related electrical systems or components are credited in the AST analyses. The non-safety-related ADVs credited in analysis requiring cooldown to shutdown cooling entry conditions are non-electrical and are designed to fail open, upon loss of air. The steam release can be controlled by the safety-related motor-operated isolation valves located upstream of each ADV. The safety-related Class 'I E circuits remain

- 30 electrically independent of non-Class 1E circuits. If the ADV path became inoperable, the control room operators would be alerted of the isolation of the pathway, via a rapid increase in the associated SG pressure and via an eventual opening of an MSSV which is monitored through an acoustic monitoring panel and no loads were added to the ANO-2 emergency diesel generators as a result of the AST adoption.

The NRC staff concludes that since the licensee will continue to use the TID-14844 assumptions for performing EQ analysis, and no new equipment is added to its 10 CFR 50.49 program, the current EQ of electrical equipment will remain unchanged for the purpose of implementation of an AST.

3.5.2 Summary Based on the above, the NRC staff concludes that the proposed TS changes will not impact the licensee's compliance with 10 CFR 50.49, 10 CFR 50.67, and GDC 17 requirements, are consistent with the guidance in RG 1.183, and there is no impact on the safety-related electrical systems as a result of adoption of AST. Therefore, the staff concludes that the proposed changes are acceptable.

4.0 STATE CONSULTATION

In accordance with the Commission's regulations, the Arkansas State official was notified of the proposed issuance of the amendment. 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 dated July 27,2010 (75 FR 44024). Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9) and 10 CFR 51.22(c)(10). 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.

- 31

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 amendment will not be inimical to the common defense and security or to the health and safety of the public.

Principal Contributors: D. Duvigneaud A. Boatright L. Brown J. White S. Basturescu C. Hunt Date: April 26, 2011

- 32 Table 3.1-1 ANO-2 Control Room Atmospheric Dispersion Factors (X/O Values, sec/m 3)

Release Location 0- 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> - . J!-- 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 1 -4 days 4- 30 days Containment 8.49 x 10-4 5.58 X 10-4 2.29 X 10-4 1.69 X 10-4 1.28 X 10-4 MSSV 8.05 X 10-4 4.64 X 10-4 2.08 X 10-4 1.42 x 10-4 1.06 X 10'4 ADV 6.31 x 10'4 3.65 X 10-4 1.64 X 10'4 1.09 x 10'4 8.26 X 10,5 PRVS 9.77 x 10'4 5.76 X 10-4 2.56 X 10-4 1.68 X 10'4 1~

Main Steam Pipe 5.48 x 10'4 3.23 X 10-4 1.43 X 10'4 1.01 X 10-4 7.45 X 10,5 Fuel Handling Area 1.20 x 10'3 6.72 X 10'4 3.12x10,4 2.00 X 10-4 1.54 X 10-4 Table 3.1-2 ANO-2 EAB and LPZ Atmospheric Dispersion Factors eX/o Values, sec/m 3)

EAB 0-2 hours 6.5 x 10'4 LPZ 0-8 hours 3.1 x 10,5 8-24 hours 3.6 x 10'6 1-4 days 2.3 x 10.6 4-30 days 1.4 x 10.6

- 33 Table 3.2 Licensee Calculated Radiological Consequences of Design Basis Accidents EAS*.OO$&)

- l.p,ZOOSe CR Dose Design Basis Accident (rem TEDE) (remTEDE) (rem TEDE)

LOCA 9.084 0.673 2.062 SGTR Pre-existing Iodine Spike 2.80 0.14 0.59 MSLB Pre-existing Iodine Spike 0.815 0.064 1.008 Acceptance Criteria 25.0 25.0 5.0 SGTR Accident-initiated Iodine 2.07 0.11 0.22 Spike MSLB Accident-initiated Iodine 2.252 0.225 1.143 Spike LRA 1.01 0.16 0.27 Acceptance Criteria 2.5 2.5 5.0 FHA 1OO-hr decay 4.01 0.19 1.19 CREA Containment Release 2.265 0.327 0.95 CREA Secondary Release 1.422 0.217 0.321 Acceptance Criteria 6.3 6.3 5.0

- 34 Table 3.2.1 Key Parameters Used in Radiological Consequence Analysis of Loss of Coolant Accident Parameter Value Reactor Core Power, MWth 3087 Containment Volume, ft3 1.78E6 Gap Release Phase 30 sec 0.5 hrs Early In-Vessel Release Phase 0.5 -1.8 hrs Gap Release Fraction 0.05 for noble gases, halogens, and alkali metals only Iodine species distribution (percent)

Particulate 95 Elemental 4.85 Organic 0.15 Containment Leak Rates 0.1 percent/day S 24 hrs 0.05 percent/day> 24 hrs Unsprayed Containment Volume, ft3 3.92E5 Sprayed Containment Volume, ft3 1.388E6 Sump Volume, ft3 62,898 Containment Sprayed Fractions 0.22 unsprayed 0.78 sprayed Containment Mixing Rates 11,880 cfm unsprayed to sprayed 11,880 cfm sprayed to unsprayed Containment Spray Removal Coefficients, h(1 Spray elemental iodine removal Injection 20 Recirculation 10 Spray particulate removal Injection 3.97 Recirculation 4.24 Spray Initiation Time (no termination), sec 180 Natural Deposition in Unsprayed Region No credit taken Penetration Room Ventilation System No credit taken Dose Conversion Factors Federal Guidance Report 11 CEDE Federal Guidance Report 12 EDE ESF Leakage Rate, cc/hr 4120 Fraction of Released Iodine in Sump Solution 1.0

- 35 Table 3.2.1 (continued)

Kev Parameters Used in Radiological Consequence Analysis of Loss of Coolant Accident Parameter Value Iodine Species Distribution in Sump 0.97 elemental 0.03 organic Time to Recirculation, sec 1560 (0.434 hr)

Flashing Fraction, percent 10 Control Room Isolation 10 seconds Control Room X/Q Table 3.1.1 Offsite X/Q Table 3.1.2

- 36 Table 3.2.2 Key Parameters Used in Radiological Consequence Analysis of Fuel Handling Accident Parameter Value Reactor Core Power, MWt 3087 Peaking Factor 1.7 Number of Failed Fuel Rods 472 Fuel Decay Time, hrs 100 Fission Product Gap Fractions Kr-85 0.10 1-131 0.08 Other Noble Gases 0.05 Other lodines 0.05 Alkali Metals 0.05 Minimum Water Depth Above Damaged Fuel, ft 23 Pool Decontamination Factors Overall 200 Elemental Iodine 286 Organic Iodine and Noble Gases 1 Iodine Form in Pool Elemental 99.85 percent Organic 0.15 percent Iodine Form Above Pool Elemental 70 percent Organic 30 percent Time of CR Isolation 36 seconds CR Unfiltered inleakage, cfm 250 Offsite and CR Breathing Rate (duration of event), m3/s 3.5E-04 Dose Conversion Factors Federal Guidance Report 11 CEDE Federal Guidance Report 12 EDE Control Room X/Q Table 3.1.1 Offsite X/Q Table 3.1.2

- 37 Table 3.2.3 Key Parameters Used in Radiological Consequence Analysis of Main Steam Line Break Accident Parameter V..lue Reactor Core Power, MWth 3086.52 MWt Initial Primary Coolant Activity 1.0 IJCi/gm DE 1-131 Activity with Pre-existing Iodine Spike 60 IJCi/g DE 1-131 Initial Secondary Coolant Activity 0.1 IJCi/g 1-131 Accident-Initiated Iodine Spike Factor 500 Accident-Initiated Iodine Spike Duration 8 hrs Primary-to-Secondary Leak Rate 0.5 gpm/SG Time to Begin Cooldown (operator action) 60 min Time to Isolation of Intact SG (initiation of SDC) 8 hrs and to Reach 212 of (terminate steam release)

Flashing Fraction in Intact SG 0.05 Partition Coefficient 1.0 (faulted SG via flashing and vaporization; intact SG via flashing)

Partition Coefficients (intact SG via steaming) lodines 0.01 Alkali metals 0.001 RCS Mass Maximum 4.54E5 Ibm Minimum 4.32E5 Ibm SG Secondary Mass Maximum 1.936E51bm Minimum 1.382E5 Ibm Iodine Form of Secondary Release Particulate o percent Elemental 97 percent Organic 3 percent Time of CR Isolation 10 seconds CR Unfiltered Inleakage pre-existing iodine spike case o cfm accident-initiated iodine spike case 250 cfm Dose Conversion Factors Federal Guidance Report 11 CEDE Federal Guidance Report 12 EDE Control Room XlQ Table 3.1.1 Offsite XlQ Table 3.1.2

- 38 Table 3.2.4 Key Parameters Used in Radiological Consequence Analysis of Steam Generator Tube Rupture Accident Parameter Value Reactor Core Power, MWth 3086.52 MWt Initial Primary Coolant Activity 1.0 IJCi/gm DE 1-131 Activity with Pre-existing Iodine Spike 60 IJCilg DE 1-131 Initial Secondary Coolant Activity 0.1 IJCi/g 1-131 Accident-Initiated Iodine Spike Factor 335 Accident-Initiated Iodine Spike Duration 8 hrs Initial Ruptured SG Tube Leak Rate 240 gpm Primary-to-Secondary Leak Rate 150 gpd per SG Time of Reactor Trip/LOOP o sec Time to Isolation of Faulted SG (operator action) 60 min Time to Isolation of Intact SG (initiation of SDC) 8 hrs Flashing Fraction in SGs 5 Partition Coefficient (via flashing) 1.0 Partition Coefficients (via steaming) lodines 0.01 Alkali metals 0.001 RCS Mass Maximum 4.54E5 Ibm Minimum 4.32E5 Ibm SG Secondary Mass 1.382E51bm Time of CR Isolation 36 seconds CR Unfiltered Inleakage 250 cfm Dose Conversion Factors Federal Guidance Report 11 CEDE Federal Guidance Report 12 EDE Control Room X/Q Table 3.1.1 Offsite X/Q Table 3.1.2

- 39 Table 3.2.5 Key Parameters Used in Radiological Consequence Analysis of Locked Rotor Accident Parameter Value Reactor Core Power, MWth 3086.52 Fuel Failure (rod in DNS) 14 percent Peaking Factor 1.65 Fission Product Gap Fractions 1-131 0.08 Kr-85 0.10 Other noble gases and iodines 0.05 Alkali metals 0.12 Secondary Release Iodine Species Distribution, percent Particulate 0 Elemental 97 Organic 3 Primary-to-Secondary (P-S) Leak Rate 150 gpd/SG Duration of Steam Release (switch to SOC system) 8 hrs Flashing Fraction of P-S Leakage during Cooldown 0.05 Partition Coefficients lodines 0.01 Alkali metals 0.001 SG Secondary Mass 1.253E8 Time of CR Isolation 36 seconds CR Unfiltered Inleakage 250 cfm Federal Guidance Report 11 CEDE Dose Conversion Factors Federal Guidance Report 12 EDE Control Room X/Q Table 3.1.1 Offsite X/Q Table 3.1.2

- 40 Table 3.2.6 Key Parameters Used in Radiological Consequence Analysis of Control Rod Ejection Accident Parameter Va.lue Reactor Core Power, MWth 3086.52 Fuel Failure (rod in DNB) 14 percent Peaking Factor 1.65 Fission Product Gap Fractions Noble gases and iodines 0.10 Alkali metals 0.12 Containment Release Iodine Species Distribution, percent 95 Particulate 4.85 Elemental 0.15 Organic Secondary Release Iodine Species Distribution, percent Particulate 0 Elemental 97 Organic 3 Primary-to-Secondary (P-S) Leak Rate 300 gpd (secondary release model)

Duration of Steam Release (switch to SDC system) 8 hrs Flashing Fraction of P-S Leakage during Cooldown 0.05 0.1 percent/day so 24 hrs Containment Leak Rates 0.05 percent/day> 24 hrs Sedimentation Coefficient (Particulates only) 0.1/hr until DF =1000. then 0 Containment Spray No credit taken Penetration Room Ventilation System No credit taken Partition Coefficients lodines 0.01 Alkali metals 0.001 RCS Mass

- 41 Table 3.2.6 (continued)

Key Parameters Used in Radiological Consequence Analysis of Control Rod Ejection Accident Parameter Value SG Secondary Mass 1.253E8 Time of CR Isolation 10 seconds CR Unfiltered Inleakage 250 cfm Federal Guidance Report 11 CEDE Dose Conversion Factors Federal Guidance Report 12 EDE Control Room X/Q Table 3.1.1 Offsite X/Q Table 3.1.2

- 42 Table 3.2.7 Key Parameters Used in Modeling the Control Room for Design Basis Radiological Consequence Analyses Parameter Value Control Room Volume, ft3 40,000 ft3 Recirculation Flow Rate, cfm 1667 cfm Emergency Filtration System Initiation Delay, hours 0 Emergency Filtered Intake Flow Rate, cfm 333 Recirculation Filter Efficiency, percent Elemental 95 percent Organic 95 percent Aerosol/Particulate 99 percent Emergency Intake Filter Efficiency, percent Elemental 99 percent Organic 99 percent Aerosol/Particulate 99 percent 250 Unfiltered Inleakage, cfm Occupancy Factors 0-24 hours 1.0 24 - 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> 0.6 96 - 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> 0.4 Breathing Rate, m3/sec 3.5E-04 Atmospheric Dispersion Factors Table 3.1.2

April 26, 2011 Vice President, Operations Arkansas Nuclear One Entergy Operations, Inc.

1448 S.R. 333 Russellville, AR 72802

SUBJECT:

ARKANSAS NUCLEAR ONE, UNIT NO.2 - ISSUANCE OF AMENDMENT RE:

USE OF ALTERNATE SOURCE TERM (TAC NO. ME3678)

Dear Sir or Madam:

The Nuclear Regulatory Commission (NRC) has issued the enclosed Amendment No. 293 to Renewed Facility Operating License No. NPF-6 for Arkansas Nuclear One, Unit No.2. The amendment consists of changes to the Technical Specifications (TSs) in response to your application dated March 31, 2010, as supplemented by letters dated June 23, June 24, August 9, and September 16, 2010.

The amendment modifies the requirements of the TS definitions, requirements, and terminology related to the use of an Alternate Source Term (AST) associated with accident offsite and control room dose consequences. In addition, implementation of the AST supports adoption of the control room envelope habitability controls in accordance with NRC-approved Technical Specification Task Force (TSTF)-448, Revision 3, "Control Room Habitability."

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

Sincerely, IRA!

N. Kaly Kalyanam, Project Manager Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket 1\10. 50-368

Enclosures:

1. Amendment No. 293 to NPF-6

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