ML090890627

Issuance of Amendments Regarding Adoption of the Alternative Source Term Radiological Analysis Methodology

ML090890627
Person / Time
Site:	McGuire, Mcguire
Issue date:	03/31/2009
From:	Stang J F Plant Licensing Branch II
To:	Hamilton B H Duke Energy Carolinas
Thompson Jon, NRR/DORL/LPL 2-1, 415-1119
References
TAC MD8400, TAC MD8401
Download: ML090890627 (45)

Text

UNITED NUCLEAR REGULATORY WASHINGTON, D.C. 20555-0001 March 31, 2009 Mr. Bruce H. Hamilton Vice President McGuire Nuclear Station Duke Energy Carolinas, LLC 12700 Hagers Ferry Road Huntersville, NC 28078 MCGUIRE NUCLEAR STATION, UNITS 1 AND 2, ISSUANCE OF AMENDMENTS REGARDING ADOPTION OF THE ALTERNATE SOURCE TERM RADIOLOGICAL AI\IALYSIS METHODOLOGY (TAC I\IOS. MD8400 and MD8401)

Dear Mr. Hamilton:

The Nuclear Regulatory Commission has issued the enclosed Amendment No. 251 to Renewed Facility Operating License NPF-9 and Amendment No. 231 to Renewed Facility Operating License NPF-17 for the McGuire Nuclear Station, Units 1 and 2. The amendments consist of changes implemented in order to adopt the alternate source term (AST) radiological analysis methodology in response to your application dated March 20, 2008, as supplemented by letters dated May 28,2008, October 6,2008, December 17, 2008, February 12, 2009, and March 25, 2009.. The amendments represent full scope implementation of the AST as allowed by Title 10 of the Code of Federal Regulations, Part 50, Section 50.67, "Accident source term," for the Loss of Coolant Accident described in NRC Regulatory Guide 1.183, "Alternate Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors, Revision 0." A copy of the related Safety Evaluation is also enclosed. A Notice of Issuance will be included in the Commission's biweekly Federal Register notice.

B. Hamilton -2If you have any questions, please call me at 301-415-1345.

Sincerely, ManagerPlant Licensing Branch Division of Operating Reactor Office of Nuclear Reactor Docket Nos. 50-369 and

Enclosures:

1. Amendment No. 251 to 2. Amendment No. 231 to 3. Safety Evaluation cc w/encls: Distribution via Listserv UNITED NUCLEAR REGULATORY WASHINGTON, D.C. 20555-0001 DUKE ENERGY CAROLINAS, LLC DOCKET NO. 50-369 MCGUIRE NUCLEAR STATION, UNIT 1 AMENDMENT TO RENEWED FACILITY OPERATING liCENSE Amendment No. 251 Renewed License No. NPF-9 The Nuclear Regulatory Commission (the Commission) has found that: The application for amendment to the McGuire Nuclear Station, Unit 1 (the facility), Renewed Facility Operating License No. NPF-9, filed by the Duke Energy Carolinas, LLC (licensee), dated March 20, 2008, as supplemented by letters dated May 28,2008, October 6,2008, December 17, 2008, February 12, 2009, and March 25, 2009, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations as set forth in 10 CFR Chapter I; The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; 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 set forth in 10 CFR Chapter I; The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and 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

-2Accordingly, Paragraph 2.C.(2) of Renewed Facility Operating License No. NPF-9 is hereby amended to read as follows: Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 251 ,are hereby incorporated into this license. The licensee shall operate the facility in accordance with the Technical Specifications. This license amendment is effective as of its date of issuance and shall be implemented within 60 days of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION Melanie C.Wong, Chief Plant Licensing Branch 11-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to License No. NPF-9 and the Technical Specifications Date of Issuance:

March 31) 2009 UNITED NUCLEAR REGULATORY WASHINGTON, D.C. 20555-0001 DUKE ENERGY CAROLINAS, LLC DOCKET NO. 50-370 MCGUIRE NUCLEAR STATION, UNIT 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 231 Renewed License No. NPF-17 The Nuclear Regulatory Commission (the Commission) has found that: The application for amendment to the McGuire Nuclear Station, Unit 2 (the facility), Renewed Facility Operating License No. NPF-17, filed by the Duke Energy Carolinas, LLC (the licensee), dated March 20, 2008, as supplemented by letters dated May 28,2008, October 6,2008, December 17, 2008, February 12,2009, and March 25, 2009, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations as set forth in 10 CFR Chapter I; The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; 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 set forth in 10 CFR Chapter I; The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and 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 2

-2Accordingly, Paragraph 2.C.(2) of Renewed Facility Operating License No. NPF-17 is hereby amended to read as follows: . Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 231 ,are hereby incorporated into this license. The licensee shall operate the facility in accordance with the Technical Specifications. This license amendment is effective as of its date of issuance and shall be implemented within 60 days of issuance. FOR THE NUCLEAR REGULATORY COMMISSION . Melanie C. Wong, Chief Plant Licensing Branch 11-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to License No. NPF-17 and the Technical Specifications Date of Issuance:

March 31, 2009 ATTACHMENT TO LICENSE AMENDMENT NO. RENEWED FACILITY OPERATING LICENSE NO. DOCKET NO. LICENSE AMENDMENT NO. RENEWED FACILITY OPERATING LICENSE NO.

DOCKET NO. Replace the following pages of the Renewed Facility Operating Licenses and the Appendix A Technical Specifications (TSs) with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

Remove License Page License Page NPF-9, page 3 NPF-9, page 3 NPF-17, page 3 NPF-17, page 3 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 materia' without restriction to chemical or physical form, for: sample ana'ysis or instrument calibration or associated with radioactive apparatus or components; Pursuant to the Act and 10 CFR Parts 30,40 and 70, to possess, but not separate, such byproducts and special nuclear materials as may be produced by the operation of McGuire Nuclear Stati,on, Units l' and 2',. and'; Pursuant to the Act and 10 CFR Parts 30 and 40, to receive, possess and process for release or transfer such byproduct matena' as may be produced by the Duke Training and Technology Center. This renewed operating license shall be deemed to contain and is SUbjectto,the conditions specified in the Commission's regulations set forth in 10 CFR Chapter , 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 The licensee is authorized to operate the facility at a reactor core full steady state power level of 3411 megawatts thermal (100%). Technical Specifications The Technical Soecifications contained in Appendix 'A, as revised through Amendment No. 251 "are hereby incorporated into this renewed operating license. The licensee shall operate the facility In accordance with the Technical Specifications. Updated Final Safety Analysis Report The Updated Final Safety Analysis Report supplement submitted pursuant to 10 CFR 54.21(d), as revised on December 16, 2002, descrlbes certain future activities to be completed before the period of extended operation. Duke shall complete these activities no later than June 12, 2021, and shall notify the NRC in writing when implementation of these activities is complete and can be verified by NRC inspection.

The Updated Final Safety Analysis'Report supplement as revised on December 16,2002, described above, shall be included in the next scheduled update to the Updated Final Safety Analysis Report required by 10 CFR 50.71(e)(4), following issuance of this renewed operating license. Until that update is complete, Duke may make changes to the programs described in such supplement without prior Commission approval, provided that Duke evaluates each such change pursuant to the criteria set forth in 10 CFR 50,59 and otherwee complies with the reculrememe in that section. Renewed License No, NPF-9 Amendment No. 251 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; Pursuant to the Act and 10 CFR Parts 30,40 and 70. to possess. but not separate, such byproducts and' special nuclear materials as may be produced by the operation of McGuire Nuclear Station, Units 1 and 2; and, Pursuant to the Act and 10 CFR'Parts 30 and 40, to receive, possess and process for release or transfer such byproduct material as may be produced by the Duke Training and Technology Center. This renewed operating license shall be deemed to contain and is subject to the conditions specified In the Commission's regUlations set forth in 10 CFR Chapter I and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below: (1 Maximum Power Level The licensee is authorized to operate the facility at a reactor core full steady state power level of 3411 megawatts thermal (100%). Technical Specifications . The Technical Soecifications contained in Appendix A, as revised through Amendment No. 231. hereby incorporated into this renewed operating license. The licensee shall operate the facility in accordance with the Technical Specifications. Updated Final Safety Analysis Report The Updated FInal Safety Analysis Report supplement submlUed pursuant to 10 eFR 54.21(d), as revised on December 16.2002, describes certain future activities to be completed before the period of extended operation. Duke shall complete these activities no later than March 3,2023, and shall notify the NRC in writin9 when implementation of these activities is complete and can be verified by NRC inspection.

The Updated Final Safety Analysis Report supplement as revised on December 16,2002, described above, shall be included in the next scheduled update to the Updated Final Safety Analysis Report required by 10 CFR 50.71(eX4), following issuance of this renewed operating license. Until that update Is complete, Duke may make changes to the programs described in such supplement without prior Commission approval, provided that Duke evaluates each such change pursuant to the criteria set forth in 10 CFR 50,59, and otherwise complies with the requirements in that section. Renewed license No. NPF-17 . Amendment No. 231 UNITED NUCLEAR REGULATORY WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 251 TO RENEWED FACILITY OPERATING LICENSE NPF-9 AND AMENDMENT NO. 231 TO RENEWED FACILITY OPERATING LICENSE NPF-17 DUKE ENERGY CAROLINAS, LLC MCGUIRE NUCLEAR STATION, UNITS 1 AND 2 DOCKET NOS. 50-369 AND 50-370

1.0 INTRODUCTION

By application dated March 20, 2008 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML080930505), as supplemented by letters dated May 28, 2008 (ADAMS Accession No. ML081560395), October 6,2008 (ADAMS Accession No. ML082830014), December 17,2008 (ADAMS Accession No. ML083590342)

February 12, 2009 (ADAMS Accession No. ML090540682), and March 25, 2009 (ADAMS Accession No. ML090860776)

Duke Energy Carolinas, LLC (Duke, the licensee), requested changes to the Renewed Facility Operating Licenses (FOLs) and Updated Final Safety Analysis Reports (UFSARs) for the McGuire Nuclear Station, Units 1 and 2 (McGuire 1 and 2). The supplements dated May 28,2008, October 6,2008, December 17, 2008, February 12, 2009, and March 25, 2009, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staff's original proposed no significant hazards consideration determination as published the Federal Register on February 27,2009 (74 FR 9009).

The amendments represent full scope implementation of the alternate source term (AST) radiological analysis methodology as allowed by Title 10 of the Code of Federal Regulations (10 CFR), Part 50, Section 50.67, "Accident source term," for the Loss of Coolant Accident [LOCA] described in NRC Regulatory Guide [RG] 1.183, "Alternate Radiological Source Terms for Evaluating Design Basis Accidents

[DBAs] at Nuclear Power Reactors, Revision 0."

2.0 REGULATORY EVALUATION

The Nuclear Regulatory Commission (NRC) staff reviewed the licensee's evaluation of the radiological consequences of the design basis LOCA for implementation of the AST Enclosure 3

-2 methodology against the requirements specified in 10 CFR 50.67(b)(2).

Section 50.67(b)(2) requires that the licensee's analysis demonstrates with reasonable assurance that: An individual located at any point on the boundary of the exclusion area for any 2-hour period following the onset of the postulated fission product release, would not receive a radiation dose in excess of 25 roentgen equivalent man (rem) total effective dose equivalent (TEDE). An individual located at any point on the outer boundary of the low population zone, who is exposed to the radioactive cloud resulting from the postulated fission product release during the entire period of its passage, would not receive a radiation dose in excess of 25 rem TEDE. Adequate radiation protection is provided to permit access to and occupancy of the control room (CR) under accident conditions without personnel receiving radiation exposures in excess of 5 rem TEDE for the duration of the accident.

This safety evaluation (SE) addresses the impact of the proposed changes on previously analyzed DBA's radiological consequences and the acceptability of the revised analysis results. The regulatory requirements from which the NRC staff based its acceptance are the reference values in 10 CFR 50.67. The licensee has not proposed any significant deviation or departure from the guidance provided in RG 1.183. The NRC staffs evaluation is based upon the following regulations, regulatory guides, and standards: 10 CFR Part 50.67, "Accident Source Term." 10 CFR Part 50, Appendix A, "General Design (GDC) Criterion for Nuclear Power Plants": GDC 19, "Control room." RG 1.23, "Onsite Meteorological Programs," Rev.O, February 1972. RG 1.52, "Design, Inspection, and Testing Criteria for Air Filtration and Adsorption Units of Post Accident Engineered Safety Feature Atmosphere Cleanup Systems in Light Water Cooled Nuclear Power Plants," Rev. 3, June 2001. RG 1.145, "Atmospheric Dispersion Models for Potential Accident Consequence Assessments at Nuclear Power Plants," Rev. 1, November 1982. RG 1.183, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors," Rev. 0, July 2000. RG 1.194, "Atmospheric Relative Concentrations for Control Room Radiological Habitability Assessments at Nuclear Power Plants," Rev. 0, June 2003. RG 1.196, "Control Room Habitability at Light-Water Nuclear Power Reactors," Rev. 0, May 2003.

-3 NUREG-0800, "Standard Review Plan," Section 2.3.4, "Short-Term Diffusion Estimates for Accidental Atmospheric Releases," Rev. 3, March 2007.

• NUREG-0800, "Standard Review Plan," Section 6.4, "Control Room Habitability Systems," Rev. 3, March 2007.

• NUREG-0800, "Standard Review Plan," Section 6.5.2, "Containment Spray as a Fission Product Cleanup System," Rev. 4, March 2007.

• NUREG-0800, "Standard Review Plan," Section 15.0.1, "Radiological Consequence Analyses Using Alternative Source Terms," Rev. 0, July 2000.

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

• NUREG/CR-5950, "Iodine Evolution and pH Control," December 1992. 3.0 TECHNICAL EVALUATION

3.1 Radiological

Consequences of Design Basis Accidents (DBAs) As stated in RG 1.183, Regulatory Position 5.2, the DBAs addressed in the appendices of RG 1.183 were selected from accidents that may involve damage to irradiated fuel. RG 1.183 does not address DBAs with radiological consequences based on technical specification (TS) reactor or secondary coolant specific activities only. The inclusion or exclusion of a particular DBA in RG 1.183 should not be interpreted as indicating that an analysis of that DBA is required or not required.

Licensees should analyze the DBAs that are affected by the specific proposed applications of an AST. In the application dated March 20, 2008, the licensee performed an analysis of the design basis LOCA for a full-scope implementation of the AST, in accordance with the guidance in RG 1.183, and SRP Section 15.0.1. As discussed in RG 1.183, Regulatory Position 1.2.1, full implementation is a modification of the facility design basis that addresses all characteristics of the AST, that is, composition and magnitude of the radioactive material, its chemical and physical form, and the timing of its release. Full implementation revises the plant licensing basis to specify the AST in place of the previous accident source term and establishes the TEDE dose as the new acceptance criteria. This applies not only to the analyses performed in this application, which only includes the LOCA, but also to all future design basis dose consequence analyses at McGuire 1 and 2. Since, upon acceptance of this licensing action request (LAR), the AST and TEDE criteria will become part of the design basis for McGuire 1 and 2, new applications of the AST would not require prior NRC approval unless stipulated by 10 CFR 50.59, "Changes, Tests, and Experiments," or unless the new application involved a change to a TS. However, a change from an approved AST to a different AST that is not approved for use at McGuire 1 and 2 would require a license amendment under 10 CFR 50.67.

The licensee states that no physical plant modifications are required by the new analysis or to implement the AST and that no TS changes are proposed in this LAR. The licensee further states

-4 that implementation of the AST will permit McGuire 1 and 2 to retire a non-conforming issue and associated current operability evaluation related to measured CR unfiltered in-leakage and its impact on the radiological consequences to CR operators from a design basis LOCA. The DBA radiological source term used in the AST analyses was developed based on a core power level of 3479 megawatts thermal (MWt). The core power level used in the AST analysis of 3479 MWt represents the licensed power level of 3411 MWt with a 2% increase to account for measurement uncertainties. The use of 3479 MWt for the AST DBA radiological source term analyses bounds the current licensed core thermal power level of 3411 MWt and is therefore acceptable to the NRC staff for use in the full implementation of the AST at McGuire 1 and 2.

Regulatory Position 6 of RG 1.183 states that the NRC staff is assessing the effect of increased cesium releases on equipment qualification (EQ) doses to determine whether licensee action is warranted and that until such time as this generic issue is resolved, licensees may use either the AST or the Technical Information Document (TID)-14844 assumptions for performing the required EQ analyses. In a letter dated October 6, 2008, the licensee provided additional information stating that the current TID-14844, Atomic Energy Commission (AEC), 1962, "Calculation of Distance Factors for Power and Test Reactors Sites," accident source term will remain the licensing basis for EQ until the generic issue regarding cesium releases is resolved. The issue of the effect of increased cesium releases on EQ doses has been resolved as documented in a memo dated April 30, 2001 (ADAMS Accession No. ML011210348) and in NUREG-0933, Supplement 25, June 2001 (ADAMS Accession No. ML012190402).

The conclusion to Generic Issue 187, "The Potential Impact of Postulated Cesium Concentration on Equipment Qualification in the Containment Sump," states the following: "The NRC staff concluded that there was no clear basis for back-fitting the requirement to modify the design basis for equipment qualification to adopt the AST. There would be no discernible risk reduction associated with such a requirement.

Licensees should be aware, however, that a more realistic source term would potentially involve a larger dose for equipment exposed to sump water for long periods of time. Longer term equipment operability issues associated with severe fuel damage accidents, (with which the AST is associated) could also be addressed under accident management or plant recovery actions as necessary." Therefore, in consideration of the cited references, the NRC staff finds that it is acceptable for the TID 14844 accident source term to remain the licensing basis for EQ at McGuire 1 and 2. RG 1.183, Regulatory Position 4.3, Other Dose Consequences, states that: "The guidance provided in Regulatory Positions 4.1 and 4.2 should be used, as applicable, in re-assessing the radiological analyses identified in Regulatory Position 1.3.1, such as those in NUREG-0737.

Design envelope source terms provided in NUREG-0737 should be updated for consistency with the AST. In general, radiation exposures to plant personnel identified in Regulatory Position 1.3.1 should be expressed in terms of TEDE." RG 1.183, Regulatory Position 1.3.4, "Updating Analyses Following Implementation," states that: "Full implementation of the AST replaces the previous accident source term with the approved AST and the TEDE criteria for all design basis radiological analyses. The implementation may have been supported in part by sensitivity or scoping analyses that concluded many of the design basis radiological analyses would remain bounding for the AST and the TEDE criteria and would not require updating. After the implementation is complete, there may be a subsequent need (e.g.,

-a planned facility modification) to revise these analyses or to perform new analyses. For these recalculations, the NRC staff expects that all characteristics of the AST and the TEDE criteria incorporated into the design basis will be addressed in all affected analyses on an individual as-needed basis. Re-evaluation using the previously approved source term may not be appropriate. Since the AST and the TEDE criteria are part of the approved design basis for the facility, use of the AST and TEDE criteria in new applications at the facility does not constitute a change in analysis methodology that would require NRC approval." In a letter dated October 6, 2008, the licensee provided additional information stating that McGuire 1 and 2 plans to incorporate the AST into the remaining NUREG-0737 analyses on an as-needed basis. The licensee also stated that since the full implementation of the AST at McGuire 1 and 2 does not require any physical changes to the plant, the implementation of the AST has no impact on the assumptions or inputs to the current TID-14844-based analyses. The NRC staff finds that since the results of TID-14844 analyses generally bound the results of AST analyses, it is acceptable for the licensee to incorporate the AST into the remaining NUREG-0737 analyses on an as-needed basis. The DBA LOCA dose consequence analyses evaluated the integrated TEDE dose at the exclusion area boundary (EAB) for the worst 2-hour period following the onset of the accident.

The integrated TEDE doses at the outer boundary of the low population zone (LPZ) and the integrated dose in the McGuire 1 and 2 CR were evaluated for the duration of the accident. The dose consequence analyses were performed by the licensee using the Bechtel proprietary LOCADOSE computer code. LOCADOSE is the primary radiological effluent analysis code used by the licensee to analyze design basis accidents. The licensee used the LOCADOSE code to compute the off-site and CR doses for the Catawba Nuclear Station, Units 1 and 2 (Catawba 1 and 2) AST LOCA analyses which were accepted by the NRC staff (ADAMS Accession No. ML052730312). The licensee used the LOCADOSE code to model activity transfer to simulate the progression of the accident and the transport of activity within the plant and to the environment and receptors.

LOCADOSE is similar in its application to the RADTRAD: "Simplified Model for RADionuclide Transport and Removal And Dose Estimation," computer code. The NRC sponsored the development of the RADTRAD radiological consequence computer code, as described in NUREG/CR-6604. The RADTRAD code was developed by Sandia National Laboratories for the NRC. The RADTRAD code estimates transport and removal of radionuclides and radiological consequence doses at selected receptors. The NRC staff uses the RADTRAD computer code to perform independent confirmatory dose evaluations as needed to ensure a thorough understanding of the licensee's methods. The results of the evaluations performed by the licensee, as well as the applicable dose acceptance criteria from RG 1.183, are shown in Table 1 of this SE. RG 1.183, Regulatory Position 3.1, "Fission Product Inventory," states that, "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 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 isotope generation and depletion computer code such as ORIGEN 2 or ORIGEN ARP."

-6 In accordance with RG 1.183, Regulatory Position 3.1, the licensee generated the core radionuclide inventories for use in determining source term inventories using the ORIGEN-S code.

ORIGEN-S is an updated version of the ORIGEN code and is acceptable to the NRC staff for generating the core inventories used in dose consequence analyses. The licensee investigated a large number of combinations of fuel enrichments, batch loadings, and fuel burn-up to bound current and potential future core designs. From these assessments the licensee chose conservative input values to generate the LOCA source term inventories for the AST LOCA analysis.

As discussed in RG 1.183, the release fractions associated with the light water reactor (LWR) core inventory released into containment for the DBA 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. The licensee varied fuel assembly burnups in the LOCA cases up to and including 62,000 MWD/MTU. Therefore, the results of the licensee's AST LOCA analysis are valid for individual fuel assembly burnup levels up to 62,000 MWD/MTU for fuel assemblies to be discharged at the end of the cycle. On a core-averaged basis, the LOCA cases bound core average burnups above 44,000 MWD/MTU, which is in excess of anticipated core designs which are expected to have a peak core averaged burnup of about 40,000 MWD/MTU. The licensee used committed effective dose equivalent (CEDE) and effective dose equivalent (EDE) dose conversion factors (DCFs) from Federal Guidance Reports (FGR) 11 and 12 to determine the TEDE dose as is required for AST evaluations. The use of ORIGEN-S and DCFs from FGR-11 and FGR-12 is in accordance with RG 1.183 guidance and is, therefore, acceptable to the NRC staff.

3.1.1 Loss of Coolant Accident (LOCA) The radiological consequence design basis LOCA analysis is a deterministic evaluation based on the assumption of a major rupture of the primary reactor coolant system (RCS) piping. The accident scenario assumes the deterministic failure of the emergency core cooling system (ECCS) to provide adequate core cooling which results in a significant amount of core damage as specified in RG 1.183. This general scenario does not represent any specific accident sequence, but is representative of a class of severe damage incidents that were evaluated in the development of the RG 1.183 source term characteristics. Such a scenario would be expected to require multiple failures of systems and equipment and lies beyond the severity of incidents evaluated for design basis transient analyses. The LOCA considered in this evaluation is a complete and instantaneous circumferential severance of the primary RCS piping, which would result in the maximum fuel temperature and primary containment pressure among the full range of LOCAs. Due to the postulated loss of core cooling, the fuel heats up, resulting in the release of fission products. The fission product release is assumed to occur in phases over a 2-hour period.

-7When using the AST for the evaluation of a design basis LOCA for a pressurized water reactor PWR, it is assumed that the initial fission product release to the containment will last for 30 seconds and will consist of the radioactive materials dissolved or suspended in the RCS liquid. After 30 seconds, fuel damage is assumed to begin and is characterized by clad damage that releases the fission product inventory assumed to reside in the fuel gap. The fuel gap release phase is assumed to continue until 30 minutes after the initial breach of the RCS. As core damage continues, the gap release phase ends and the early in-vessel release phase begins. The early in-vessel release phase continues for the next 1.3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />. The licensee used the LOCA source term release fractions, timing characteristics, and radionuclide grouping as specified in RG 1.183 for evaluation of the AST.

The licensee's bounding design basis LOCA scenario includes a limiting single failure. The licensee postulated single failures of individual pieces of equipment and of whole trains of systems, including ventilation component failures, and mitigating system component failures to determine the bounding scenario. For McGuire 1 and 2, this scenario is referred to as the "Minimum Safeguards" scenario. In this scenario the licensee postulates that a power failure results in the loss of one train of the following LOCA mitigation systems: Containment Spray System (NS)

Containment Air Return System (VX)

Annulus Ventilation System (VE) Control Room Ventilation System (VC) In the evaluation of the LOCA design basis radiological analysis, the licensee considered dose contributions from the following potential activity release pathways:

Containment leakage bypassing the annulus ventilation system, Containment leakage as processed by the annulus ventilation system, ECCS leakage into the auxiliary building, and ECCS back-leakage into the refueling water storage tank (FWST). The licensee considered the following potential DBA LOCA dose contributors to the control room habitability envelope (CRHE) analysis:

Contamination of the CR atmosphere by intake and infiltration of radioactive material from the containment leakage and ECCS system leakage; External radioactive plume shine contribution from the containment and ECCS leakage releases with credit for CR structural shielding; The direct shine dose contribution from the containment's contained accident activity with credit for both containment and CR structural shielding; and

-8* The direct shine dose contribution from the activity collected on the CR ventilation filters.

3.1.1.1 LOCA Source Term The licensee followed all aspects of the guidance outlined in RG 1.183, Regulatory Position 3, regarding the core inventory and the release fractions and timing for the evaluation of the LOCA. Typically, the LOCA analysis assumes that iodine will be removed from the containment atmosphere by containment sprays and natural diffusion to the containment walls. The licensee credits the removal of iodine by the containment spray system. However, for additional conservatism the licensee did not credit natural removal mechanisms in the LOCA analysis. As a result of the containment spray removal mechanisms, a large fraction of the released activity will be deposited in the containment sump. The sump water will retain soluble gaseous and soluble fission products, such as iodine and cesium, but not noble gases. The guidance from RG 1.183 specifies that the iodine deposited in the sump water can be assumed to remain in solution as long as the containment sump pH is maintained at or above 7.

According to NUREG-1465, "Accident Source Terms for Light-Water Nuclear Power Plants," iodine released from the damaged core to the containment after a LOCA is composed of 95% cesium iodide (Csl) which is a highly ionized salt, soluble in water. Iodine in this form does not present any radiological problems since it stays dissolved in the sump water and does not enter the containment atmosphere. However, in a radiation field existing in the containment, some of this iodine could be transformed from the ionic (1-) to the elemental form (1 2) which is scarcely soluble in water and can be, therefore, released to the containment atmosphere.

Conversion of iodine to the elemental form depends on several parameters, of which pH is one of the more important. Maintaining pH basic in the sump water 7) will ensure that this conversion will be minimized. The licensee used the method described in NUREG/CR-5950, "Iodine Evolution and pH Control," for calculating generation of this elemental iodine. The licensees calculations have indicated that at the higher sump water pH less of the iodides are converted into elemental form and at pH 7 or higher elemental iodine generated from this source become insignificant relative to the elemental iodine release directly to the containment from the damage core. In the McGuire plant, the sump water pH is controlled by the presence of sodium tetraborate released from the melting ice in the ice condenser. Sodium tetraborate is a salt derived from a strong base and a weak acid. It acts, therefore, as a buffer and helps to stabilize the sump water pH. After a LOCA several acids are either generated or are added to the containment.

Relative amounts of these acids and that of sodium tetraborate determine the value of pH reached by the containment sump water. After a LOCA, boric acid from the reactor coolant system, cold leg accumulators and borated water storage tank is discharged into the sump. The licensee assumed that in all these systems, the concentration of boron is 2875 ppm. This is a conservative assumption, because the concentration of boron in the reactor coolant varies during a fuel cycle and is much lower toward the end of cycle. Also, the value of pH will be continuously decreasing due to formation of hydrochloric and nitric acids in the containment.

Hydrochloric acid is formed from decomposition of chlorinated polymer cable insulation by radiation. The licensee used a generation rate of 4.6 x10-4 moles of HCI per pound of insulation per Mrad which is consistent with the value in NUREG-5950. Nitric acid is formed in the containment by irradiation of water and air. The amount of nitric acid produced is proportional to the time-integrated dose rate for gamma and beta

-radiation. Based on the information provided in NUREG-5950, the licensee calculated a generation rate of 7.3 x10-6 moles of HN0 3 liter per Mrad. Both acids are strong acids and will contribute to lowering sump pH. In order to neutralize the boric, hydrochloric and nitric acids, the licensee chose to buffer the sump pool water by using sodium tetraborate. Such buffering action could maintain basic pH in the sumr pool despite the presence of the acids. The licensee has calculated that by adding 1.89 x 10 Ibm of ice containing 1800 ppm of sodium tetraborate to the sump pool, it will maintain the pH in the sump water as basic.

The licensee also referenced a letter dated September 22, 2004 (ML042720440) in order to describe the methodology used to calculate the pH over time. The licensee stated in the RAI response dated March 25, 2008, that the McGuire and Catawba sump pH analyses use the same methodology and that all the assumptions and methods are the same for both sites except for a few key inputs: reactor coolant system volumes and boron and lithium refueling water storage tank volumes and boron

• cold leg accumulator volumes and boron concentrations

• the sump temperature profiles

• gamma radiation dose rates

• ice mass and sodium tetraborate concentration Based on the similarity of the plants, and since the differences are all inputs to the code, the NRC staff finds that using the previously-accepted Catawba methodology is acceptable for use in the McGuire calculation.

The methodology that the licensee used to calculate the sump water pH applied an EXCEL spreadsheet with the Visual Basic Program PHSC. The EXCEL spreadsheet was used to calculate the time-dependent concentration of the chemicals dissolved in the sump water and the PHSC program was used to determine the resulting time-dependent pH. The inventories of chemicals and water in the containment sump were calculated by solving separate time-dependent mass balance equations. These chemicals included boron in the form of boric acid and sodium tetraborate, sodium in the form of sodium tetraborate, lithium in the form of lithium hydroxide added to the sump water for chemistry control, chlorides in the form of hydrochloric acid, and nitrates in the form of nitric acid. The PHSC program used the mathematical model based on correlations and data for solutions of boric acid, sodium hydroxide, and other acids and bases from the EPRI reports NP-5561-CCML and TR-105714.

The licensee calculated values of pH that varied with time. The lowest value was 7.3. As recommended by NUREG-5950, Appendix C, all the calculated pH values were determined for the temperature of 77 of. The licensee also performed hand calculations. The RAI response sent by letter dated March 25, 2009, stated, "the calculation [at 30 days post LOCAl yielded a decrease of only 0.02 in the sump pH at 25C and 0.01 in the sump pH at solution temperature compared to the values calculated at 3000 minutes." Even with a 0.02 decrease in pH, the pH is still maintained above 7. However, in calculating conversion of ionic to molecular iodine in the containment spray, the licensee used a pH at the containment sump water temperature which was about one pH unit lower. This introduced significant conservatism in the licensee's calculation. Using this pH, the licensee calculated iodine conversion from ionic to elemental form. The calculated value was so low that its concentration in the containment atmosphere was

-10 negligible compared to the concentration of molecular iodine directly release to the containment atmosphere from the damaged core. The NRC staff reviewed the licensee's methodology for calculating sump water pH and evaluated its results and found it acceptable. The NRC staff reviewed the licensee's assumptions, methodology, and conclusions regarding the pH of sump water and the corresponding fraction of the dissolved iodine in the sump water that is converted into the elemental form. The calculations were made for the 30-day period following a LOCA. The NRC staff performed independent calculations to verify the licensee's calculation. From the results of these calculations the staff concluded that, although the value of pH varied with time, it never dropped below 7.3. Maintaining pH above 7 resulted in negligible fraction of the dissolved iodine converted into elemental form and low release of radioactive iodine to the environment. Therefore, in accordance with the applicable regulatory guidance, the licensee assumed that the chemical form of the radioiodine released to the containment is 95% cesium iodide (Csl), 4.85%

elemental iodine, and 0.15%

organic iodine. With the exception of elemental and organic iodine and noble gases, fission products are assumed to be in particulate form. Based on these considerations the NRC staff finds the licensee's proposal acceptable.

3.1.1.2 Assumptions on Transport in the Primary Containment 3.1.1.2.1 Containment Mixing, Natural Deposition and Leak Rate Section 6.2 of the McGuire 1 and 2 UFSAR describes the containment as consisting of a containment vessel and a separate reactor building enclosing an annulus air space. The containment vessel is a free standing welded steel structure with a vertical cylinder, hemispherical dome, and a flat circular base. The reactor building is a reinforced concrete structure similar in shape to the containment vessel. The two structures are separated by an annular air space. The containment vessel is designed to contain the radioactive material that could be released from a loss of integrity of the reactor coolant pressure boundary. The reactor building is a concrete structure which protects the containment vessel from external missiles, provides biological shielding, and provides a means of controlling radioactive fission products that could leak from the containment vessel if an accident should occur. The double enclosure concept affords minimal interaction between the containment vessel and the reactor building, a margin of conservatism in leakage rate from the use of two structures in addition to the VE, and a reduction of gaseous and particulate radioactive release due to annulus mixing and holdup prior to filtering and release. McGuire 1 and 2 incorporates an ice condenser containment design which can accept large amounts of energy and mass inputs and maintain low internal pressures and leakage rates.

Another advantage of the ice condenser design is that its actuation is passive and does not require an actuation system signal. The VX works to promote the exchange of the atmospheres of the upper and lower containment compartments. The VX returns air from the upper containment to the lower containment. The air in the lower containment is forced through the ice condenser and into the upper containment where the spray system can remove the activity deposited in the upper containment atmosphere.

Conservatively, the licensee did not credit activity removal by the ice condensers during this process. The licensee conservatively modeled the VX system to start 10 minutes after accident initiation, which is the latest possible time based upon worst case diesel generator loading.

-11 Although credit for the removal of activity in the containment by natural deposition has been routinely accepted by the NRC staff, the licensee has conservatively chosen not to credit activity removal by natural processes including natural deposition.

Not crediting this potential mitigation process increases the conservatism of the analysis since it results in an increase in the assumed quantity of radionuclides available for release. RG 1.183, Regulatory Position 3.7 states that, "The primary containment should be assumed to leak at the peak pressure technical specification 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 />. For PWRs, the leak rate may be reduced after the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to 50% of the technical specification leak rate." Accordingly, the licensee assumed a containment vessel leak rate of 0.3 percent per day for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, after which the containment leak rate is reduced to 0.15 percent per day for the duration of the accident.

3.1.1.2.2 Containment Spray Assumptions The NS is the primary activity mitigation system inside of containment. NS is used to remove heat from containment and to remove iodine from the containment atmosphere to reduce the radiological consequences of the LOCA. Iodine is removed by the NS system by "washing" it from the atmosphere through interactions between the airborne iodine and the sprayed water droplets. The spray headers are located high in the upper containment.

Because the operating deck separates the upper and lower containment, NS operation is only effective at removing iodine from the upper containment atmosphere.

The operating deck effectively prevents direct communication between upper and lower containment, so the release from the breech in the NC is made to the lower containment only. The source term remains in lower containment until the VX starts at 600 seconds after accident initiation. The VX fans force air flow from upper containment to lower containment which promotes the flow of the lower containment atmosphere through the ice condensers to the upper containment. Once the activity reaches upper containment it can be removed by the spray system which provides full spray flow coverage to upper containment by 120 seconds after accident initiation. Initially, during the injection phase, the spray system uses FWST water to provide suction to the spray pumps. When the FWST supply is exhausted, the spray system is realigned to recirculation mode to take suction from the containment sump. The licensee evaluated the removal of airborne activity from the NS operation assuming that the system starts and provides spray flow within 2 minutes after accident initiation. This initiation time bounds the worst case diesel generator sequencer loading. The licensee assumed that NS provides spray flow to the upper containment at a reduced rate accounting for the possibility of the diesel generator operating at a reduced frequency and voltage. Full spray flow coverage is provided to the upper containment, and the licensee accounted for the reduction in spray effectiveness as a result of the nozzles which point toward containment walls. The licensee assumed that initially, the NS provides no radiological benefit since the release is made to lower containment. The lower containment is separated from upper containment by the operating deck which limits effective airflow between the upper and lower containment without a motive force. Prior to VX operation, the released activity is confined in the lower containment and the spray into upper containment has no mitigation impact. With VX initiation, air flow is forced from the upper

-12 containment to the lower containment which promotes airflow in the opposite direction through the ice condensers. This results in the transfer of activity from the lower containment, through the ice condensers into the upper containment, where activity can be mitigated by NS operation.

The licensee did not credit the potential natural air flow through the ice condensers and into the upper containment prior to VX start as a result of the thermal conditions in lower containment.

Therefore, the licensee did not credit NS activity removal until VX initiation. The licensee assumed a bounding worst case diesel loading sequence to determine the VX start time.

Activity removal by the spray system is modeled using removal constants referred to as "spray lambdas". The lambdas model the ability of the NS system to remove elemental and particulate iodines from the upper containment atmosphere in a post-LOCA environment.

The licensee derived spray lambda values using inputs related to the characteristics of the containment, the flow characteristics of the NS system, and, during recirculation, the chemistry (e.g. pH) and temperature profiles of the containment sump fluid. Once the decontamination factor (OF) exceeds its limit, as prescribed in the regulatory guidance, the effectiveness of the NS in removing the iodine from the containment atmosphere is reduced.

In accordance with the applicable regulatory guidance, the licensee reduced the effectiveness of the NS for particulates by a factor of 10 when the OF for particulates reaches 50. The licensee determined that this reduction will occur at 7100 seconds after accident initiation.

In accordance with the applicable regulatory guidance, the licensee ended the credit for elemental iodine removal after attaining a OF of 200, which occurs at 46,000 seconds after accident initiation. The licensee did not credit the removal of organic iodines from NS operation. The licensee modeled the response to the LOCA using the minimum safeguards scenario, in which only one train of the NS is operational. Assuming the minimum safeguards scenario, full injection spray flow from the FWST is provided to the upper containment by 120 seconds after accident initiation. After 3000 seconds, sump recirculation begins. During spray system realignment for sump recirculation, spray flow is supplied from the auxiliary spray header via the Residual Heat Removal (NO) System pumps. At 3240 seconds, the NS system begins supplying recirculation spray flow from the containment sump. This continues until spray flow is assumed to be secured 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after accident initiation. Full spray coverage is achieved under minimum safeguards NS operation. The spray flow credited by the licensee reflects reductions for water flow which could impinge upon the containment walls, rather than falling through upper containment to the operating deck. The licensee reduced the spray pump flow rates to bound the effect of reduced diesel generator frequency and voltage. The NRC staff has reviewed the licensee's application of credit for iodine removal from the operation of the containment spay system and has found that the analysis follows the applicable regulatory guidance, is conservative, and is therefore acceptable.

3.1.1.3 Assumptions on Dual Containments Section 6.2 of the McGuire 1 and 2 UFSAR describes the containment as consisting of a containment vessel and a separate reactor building enclosing an annulus air space. The

-13 containment vessel is a free standing welded steel structure with a vertical cylinder, hemispherical dome, and a flat circular base. The reactor building is a reinforced concrete structure similar in shape to the containment vessel. The VE provides a vacuum in the annulus to promote air flow from adjacent higher pressure spaces into the annulus where it can be held and filtered prior to being released. Prior to the establishment of vacuum in the annulus, it is assumed that containment vessel leakage is released directly to the environment. After the establishment of annulus vacuum, the licensee assumes that seven percent of the containment vessel leakage bypasses the annulus and is released directly to the environment. In response to a LOCA, the VE initiates in the exhaust mode, discharging through the filters, to reduce pressure in the annulus. The VE is required to maintain the annulus between -0.5 and -3.5 inches of water gage (wg) by TS Surveillance Requirement (SR) 3.6.16.2. In order to assure that the model reflects this requirement or is conservative relative to it, the licensee modeled the worst case impact from a difference between the thermodynamic conditions in the annulus and those conditions external to it by including the effect on stagnant head pressure from the elevation difference between the top of the annulus and the pressure detector with an extreme outside air temperature. The licensee included the effects of instrument uncertainty in this evaluation. The licensee determined that these factors, when combined, could cause the pressure detector to indicate up to 0.7 inches wg lower than the conditions at the top of the annulus. To ensure that the modeled setpoints were inclusive of the pressure at the most remote location in the annulus, and to ensure that the pressure requirement was satisfied in the worst case at all locations, the licensee applied -0.7 inches wg to the setpoints to arrive at a modeled control band of -1.2 inches wg to

-4.2 inches wg.

Therefore, in the licensee's model, when the annulus pressure reaches

-4.2 inches wg, the VE system switches to recirculation mode. In the recirculation mode the annulus environment is recirculated by the VE fans through the VE filters. No release is modeled from the annulus air space to the environment during the recirculation mode. When the modeled annulus vacuum drops to

-1.2 inches wg as a result of inleakage from the containment vessel and the environment, the VE system realigns to the exhaust mode. Under the limiting minimum safeguards scenario, the performance of only one train of VE is modeled due to the postulated single failure which removes the other train from service. Annulus vacuum is established at slightly greater than one minute after accident initiation with only one train available. In the exhaust mode, annulus vacuum increases in the model until

-4.2 inches wg is reached. The system then realigns to recirculation mode and the annulus atmosphere is recirculated. In the recirculation mode, the discharge of the VE system is directed through the VE filters and back into the annulus. During recirculation, the pressure in the annulus will increase as it loses vacuum due to in-leakage from surrounding spaces and the environment to the annulus.

The licensee modeled the VE system in-leakage based on an exterior environment temperature of 18°F which satisfies the 95th percentile temperature data requirement of RG 1.183. When the indicated annulus pressure reaches

-1.2 inches wg, the system realigns to the exhaust mode and the atmosphere in the annulus is again discharged through the VE filters to the unit vent stack. The system continues to change modes between exhaust and recirculation as these set points are met for the duration of the accident. RG 1.183, Appendix A, Regulatory Position 4.4, states that, "Credit for dilution in the secondary containment may be allowed when adequate means to cause mixing can be demonstrated. Otherwise, the leakage from the primary containment should be assumed to be transported

-14 directly to exhaust systems without mixing. Credit for mixing, if found to be appropriate, should generally be limited to 50%. This evaluation should consider the magnitude of the containment leakage in relation to contiguous building volume or exhaust rate, the location of exhaust plenums relative to projected release locations, the recirculation ventilation systems, and internal walls and floors that impede stream flow between the release and the exhaust." The licensee credited 50 percent mixing in the annulus air space based on several physical characteristics of the annulus volume and VX operation. The VX takes suction at the upper annulus elevations, whereas the VX discharges at lower elevations. The majority of the leakage from the containment vessel into the annulus is expected to occur from the penetrations at the lower elevations. The lower containment will contain a large post-LOCA heat source relative to the upper containment. The licensee asserts, and the NRC staff agrees, that the upper portion of the annulus will be cooler than the lower portion just as upper containment is cooler than lower containment due to the position of the heat source as well as the operation of the spray system. As a result of the distribution of the penetrations around the annulus, the relatively small width of the annulus which is approximately 2 feet, and the large elevation difference between the VE discharges and main in-leakage locations and the VE return header, the licensee assumed a value of 50% mixing for the leakage into the annulus prior to discharge from the VE system. This assumption is conservative, consistent with the assumptions used in the previously-accepted AST evaluation for Catawba 1 and 2 and is therefore acceptable to the NRC staff for use at McGuire 1 &2. The Catawba 1 and 2 AST analysis was accepted by the NRC in Amendment No. 227 (ADAMS Accession No. ML052730312). The licensee considered the effect of high wind speed on VE performance. The assumption of high wind speeds can challenge the performance of the VE. However, the licensee asserts, and the NRC staff agrees, that the assumption of high wind speeds would greatly reduce the off-site and control room atmospheric dispersion factors (X/Os). The licensee has determined that using low wind speeds to calculate X/Os while not including high wind speeds in post-LOCA operation of the VE System results in higher (more conservative) calculated radiation doses than does modeling high wind speed in the calculation of X/Os and post-LOCA VE operation. Therefore, the NRC staff finds that the evaluation of VE performance and X/Os based on low wind speeds is conservative and, therefore, acceptable for the AST analysis. The NRC staff has reviewed the licensee's assumptions and methodology related to the duel containment design features of McGuire 1 and 2 and has found that the analysis follows the applicable regulatory guidance, is conservative, and is therefore acceptable.

3.1.1.4 Assumptions on ECCS System Leakage To evaluate the radiological consequences of ECCS leakage, the licensee used the deterministic approach as prescribed in RG 1.183. This approach assumes that except for the noble gases, all of the fission products released from the fuel mix instantaneously and homogeneously in the containment sump water. Except for iodine, all of the radioactive materials in the containment sump are assumed to be in aerosol form and retained in the liquid phase. As a result, the licensee assumed that the fission product inventory available for release from ECCS leakage consists of 40 percent of the core inventory of iodine. This amount is the combination of five percent released to the containment sump water during the gap release phase and 35 percent released to

-15 the containment sump water during the early in-vessel release phase. This source term assumption is conservative in that 100 percent of the radioiodines released from the fuel are assumed to reside in both the containment atmosphere and in the containment sump concurrently. ECCS leakage develops when ECCS systems circulate containment sump water outside containment and leaks develop through packing glands, pump shaft seals and flanged connections.

3.1.1.4.1 Assumptions on ECCS Leakage to the Auxiliary Building The licensee assumed that ECCS leakage starts at the time that realignment for sump recirculation and auxiliary spray begins. The licensee used a value of one gallon per minute (gpm), for the evaluation of ECCS leakage to the auxiliary building. This value is consistent with the current licensing basis value and represents two times the leakage rate permitted by the leakage monitoring program. This assumption is in accordance with the guidance specified in RG 1.183, Appendix A, Item 5.2. As stated above, actual ECCS leakage would not begin until after the recirculation phase of the accident begins. The licensee assumed that ECCS leakage will start at 50 minutes into the event and continues for the 30-day duration of the accident evaluation period. RG 1.183, Appendix A, Item 5.5, states that, "If the temperature of the leakage is less than 212°F or the calculated flash fraction is less than 10%, the amount of iodine that becomes airborne should be assumed to be 10% of the total iodine activity in the leaked fluid, unless a smaller amount can be justified based on the actual sump pH history and area ventilation rates." The licensee has determined that the sump temperature does not exceed 180 of during the time that the plant is in sump recirculation. Because the temperature of the leakage is less than 212°F, RG 1.183 Appendix A, Item 5.5, provides the option of a static model that releases 10 percent of the iodine activity during the accident, or it allows for the potential for the application of smaller partitioning factors based upon a technically justifiable plant-specific model. The same partition factors and model that was used in the Catawba 1 and 2 AST analysis were adopted by the licensee for the McGuire 1 and 2 analysis based on a comparison of the site-specific conditions affecting the determination of a partition factor. The Catawba 1 and 2 AST analysis was accepted by the NRC in Amendment No. 227 (ADAMS Accession No.

ML052730312). The iodine partition fraction for post-LOCA ECCS leakage depends on several variables such as pH, iodine concentration, auxiliary building ventilation airflow rates and air temperatures in the vicinity of the leakage, characteristics of the rooms in which the leakage occurs and the assumed system leak rate. The licensee determined that relative to the Catawba 1 and 2 analysis, the pH is slightly higher for McGuire 1 and 2, which would result in a lower iodine partition factor. Due to the differences on the sump water volumes, the iodine concentration is slightly higher for the McGuire 1 and 2 analysis. However, this difference is offset by the slightly lower sump water temperature in the McGuire 1 and 2 evaluation which would result in a lower iodine partition fraction. The airflow rate for McGuire 1 and 2 is approximately half of the rate at Catawba 1 and 2, which would result in lower partition factors. In addition, the licensee has determined that the post-accident air temperature in the affected rooms is slightly lower at McGuire 1 and 2, which would tend to result in a lower amount of iodine partitioning.

-16 Based on the comparison of the variables governing the partitioning factors, the licensee has determined that the results used in the Catawba 1 and 2 evaluation are conservative for use at McGuire 1 and 2.

Therefore, the licensee used the same time-dependent partitioning factors for ECCS leakage to the auxiliary building as was used in the Catawba 1 and 2 AST LOCA analysis. During the review of the Catawba 1 and 2 AST LOCA analysis several concerns were raised pertaining to the use of partitioning factors for ECCS leakage. In a letter dated July 8, 2005 (ADAMS Accession No. ML052030337), the licensee provided additional information which is pertinent to the current LAR for McGuire 1 and 2 since the same partitioning factors were applied. The licensee described several conservative attributes used in the determination of the dose consequence analysis from post-LOCA ECCS leakage. The licensee noted that the method used in NUREG/CR-5950 is based on the use of equilibrium and reaction rate constants that are referenced at 25'C (7TF). The licensee's calculations for the formation of volatile iodine were performed with the pH of the leakage based on the higher leakage temperature for ESF leakage in the auxiliary building and inventory temperature for ESF leakage to the FWST. This introduces additional conservative margin in the analysis of iodine partitioning for ESF leakage in the auxiliary building and to the FWST. The licensee noted that additional processes which may prevent the escape of iodine airborne in the auxiliary building were not credited. In particular, no credit was taken for plate-out of iodine in the auxiliary building. In addition, dilution of airborne iodine within the auxiliary building was not credited in the analysis. The licensee set a lower bound of one percent for the iodine partition fraction for ESF leakage in the auxiliary building following a design basis LOCA. This lower bound provides additional conservatism relative to the actual calculated results.

The licensee's model for convective transfer of iodine across a surface of a pool of ECCS leakage in the auxiliary building conservatively set the liquid film iodine mass transfer coefficient to zero. The licensee set the concentration of iodine in the airspace above the pool of ECCS leakage in the auxiliary building to zero to maximize the flux of iodine partitioning from the pool. The licensee did not credit any forced convection in the determination of the temperature of the ECCS leakage in the Auxiliary Building. However, the highest values for the ventilation airflow were assumed by the licensee in the calculation of the iodine partition fraction for ECCS leakage in the auxiliary building. The licensee used additional conservative assumptions in the calculation of the dose consequence "from ECCS leakage as documented in the July 8, 2005, letter responding to NRC staff questions concerning the modeling of the dose consequence from ECCS leakage for Catawba 1 and 2. The licensee assumed that the ECCS leakage is released directly into the auxiliary building and released instantaneously into the environment with credit for auxiliary building ECCS area filtration. The licensee credited ECCS area filtration efficiencies of 92 percent for elemental and organic iodine and 98 percent for particulates. As noted previously, the licensee assumed that 100 percent of the particulate activity is retained in the sump water. The licensee did not credit a reduction of activity released to the auxiliary building as a result of dilution or holdup.

In accordance with RG 1.183, for ECCS leakage into the auxiliary building, the licensee assumed that the chemical form of the released iodine is 97 percent elemental and 3 percent organic.

-17The NRC staff has reviewed the licensee's analysis of the dose consequence from ECCS leakage to the auxiliary building and has determined that the analysis follows the applicable regulatory guidance, is conservative, and is, therefore, acceptable.

3.1.1.4.2 Assumptions on ECCS Back-leakage to the FWST The licensee evaluated the dose consequence from ECCS back-leakage to the FWST by assuming a back-leakage rate of 20 gpm. The licensee assumed that this leakage starts at 50 minutes into the event, when recirculation begins, and continues throughout the 30-day analysis period. Based on the sump pH remaining at seven or above, the iodine in the sump solution is assumed to all be nonvolatile. However, when introduced into the acidic solution of the FWST inventory, there is a potential for the particulate iodine to convert into the elemental form. The fraction of the total iodine in the FWST which becomes elemental is both a function of the FWST pH and the total iodine concentration. The amount of elemental iodine in the FWST fluid which then enters the FWST air space is a function of the temperature-dependent iodine partition coefficient. The licensee included stable 1-127 and as well as the long-lived 1-129 in the calculation of the total iodine concentration in the containment sump. For conservatism the licensee did not model refill of the FWST during the accident analysis period. The licensee evaluated the release rate of iodine from the FWST vent by determining the air displacement from the back-leakage and as well as the effects of diurnal expansion of the tank contents. The licensee employed similar methods to calculate the portion of iodine entrained in the ECCS leakage to the FWST that is released to the environment following a design basis LOCA for the Catawba 1 and 2 AST amendment (ADAMS Accession No.

ML052730312). The licensee compared the significant input parameters to the model for both Catawba 1 and 2 and McGuire 1 and 2 to determine whether the Catawba 1 and 2 release fractions could conservatively be applied to the McGuire 1 and 2 LOCA. The results of this comparison indicated the following: The minimum sump pH for McGuire 1 and 2 is the same as that reported for Catawba 1 and 2 and the equilibrium pH is slightly higher for McGuire 1 and 2. The lower sump pH for Catawba 1 and 2 results in a more limiting sump chemistry relative to McGuire 1 and 2. The upper bound temperature of the water in the FWST is the same for both McGuire 1 and 2 and Catawba 1 and 2. The upper bound value of the initial boron concentration for Catawba 1 and 2 is higher than the initial boron concentration for McGuire 1 and 2 and, therefore, limiting relative to McGuire 1 and 2. Other important parameters in the evaluation such as; the tank dimensions, the back-leakage flow rate, and the meteorological parameters germane to the iodine release, were essentially the same for both plants. The licensee has determined that the significant parameters in the evaluation are the same or nearly the same for the McGuire 1 and 2 and Catawba 1 and 2 analyses. Where a slight difference exists, the effect of the Catawba 1 and 2 value was shown to be conservative relative to

-18McGuire 1 and 2.

Therefore, the licensee has concluded, and the NRC staff agrees, that the inputs associated with the Catawba 1 and 2 ECCS back-leakage iodine FWST release fractions bound those which would be associated with similar releases from the McGuire 1 and 2 FWST. The NRC staff has reviewed the licensee's analysis of the dose consequence from ECCS back-leakage into the FWST and has determined that the analysis follows the applicable regulatory guidance, is conservative and is therefore acceptable.

3.1.1.5 Assumptions on Containment Purging The licensee stated that the containment purge system is not used in response to a LOCA. During normal operation the purge supply and exhaust valves are sealed closed per TS SR 3.6.3.1. Additionally, the licensee evaluated the containment air and release system in the LOCA analysis for similar impact. The potential release from this system was very small, and completed well before the onset of the assumed gap release. As a result, the licensee has determined, and the NRC staff agrees, that the impact of potential releases from the containment purge system on the LOCA dose calculation is not significant.

3.1.1.6 CR Habitability 3.1.1.6.1 CR Ventilation Assumptions for the LOCA The licensee assumed that the very rapid pressure increase in containment will actuate the engineered safeguards actuation system nearly instantaneously with the initiation of the accident. The CR is automatically pressurized by the VC system in response to the engineered safeguards actuation system actuation. In the minimum safeguards scenario, only a single train of filtered pressurization intake responds. In addition, the licensee conservatively assumed that the fan provides airflow at the lower limit of the test acceptance band. Using these conservative assumptions, the licensee has determined that CR pressurization will begin 11 seconds after the accident and that effective pressurization is assumed to be complete 30 seconds after accident initiation. The CR is assumed to remain pressurized for the duration of the accident. The McGuire 1 and 2 VC originally included provisions for filtered recirculation of the CR atmosphere. Currently, the filtered recirculation mode of operation is not used, and therefore the licensee did not model or credit filtered recirculation in the CR dose consequence analysis. The licensee modeled the McGuire 1 and 2 CR ventilation system as a filtered intake pressurization ventilation system as is the current mode of operation.

Pressurization of the CR affects the modeled rates of unfiltered in-leakage, which the licensee based upon the results of CR tracer gas testing conducted in October 2003. The licensee adjusted the nominal test results to include the maximum quantified test error band, and an allowance of 10 cubic feet per minute (cfm) for CR ingress and egress. The licensee then increased this total for additional conservatism. The licensee selected a value of 625 cfm for in-leakage prior to pressurization as a bounding value based upon the nominal testing value of 505+/-15 cfm.

The licensee computed the unfiltered in-leakage value after pressurization in a similar fashion based upon the most conservative testing result. The pressurization unfiltered in-leakage test

-19 was performed in both two train and single train configurations. The licensee used the most conservative nominal test value as the basis for the computation of the post pressurization unfiltered in-leakage value. The licensee adjusted the nominal test results to include the maximum quantified test error band, and an allowance of 10 cfm for CR ingress and egress. The licensee then increased this total for additional conservatism. The licensee selected a value of 210 cfm as a bounding value based upon the limiting case test result of 131+/-36 cfm. The values assumed for unfiltered in-leakage are consistent with those used in the McGuire 1 and 2 AST Fuel Handling Accident submittal (ADAMS Accession No. ML063100406). These values bound the current testing results with allowances for uncertainties and CR ingress and egress. The licensee stated that results of future tracer gas tests could change the values for CR unfiltered in-leakage. The McGuire 1 and 2 CR incorporates a dual intake design with no manual or automatic selection controls. Each train of the VC has one intake location with two separate inlets for a total of four inlets. The intake locations are sufficiently separated to eliminate the possibility of both locations being exposed to a concentrated plume of activity. The VC outside air inlets are Seismic Category I. In addition, the outside air inlets are protected from turbine missiles and are designed to withstand tornado wind loading. The licensee has not identified any failure mode that will cause an outside air inlet or one of its isolation valves to close. The licensee assumed that at the beginning of the postulated accident, the VC was in the minimum normal plant operation configuration which provides for one of the four inlets to be out of service or closed. In this alignment there are two inlets open at one intake location and one inlet open at the other. The licensee states that the CR intake flow distribution between the two intake locations changes with the system alignment and that the system may not be perfectly balanced at all particular points in time. The licensee recently tested the VC to quantify the percentage of flow from each CR intake in nine potential alignments.

The alignment tests encompassed combinations of either fan running alone or both fans running. In addition, the tests considered a single inlet isolated at either location or all inlets open. The results indicated that a flow split of 65 percent from the contaminated intake and 35 percent from the non-contaminated intake would bound the test results from all system alignments. Therefore, for the AST LOCA dose consequence analysis, 65% of the VC intake flow is assumed to be from the contaminated stream and 35% is assumed to be from the non-contaminated stream for all normal VC alignments. The licensee applied this flow split to the CR atmospheric dispersion factors prior to their use in the radiological consequences model. In accordance with applicable regulatory guidance, the licensee incorporated a safety factor of two when performing the computation of filter efficiencies as margin against the degradation of filter performance between filter tests. The licensee has a monitoring/maintenance program in place for ensuring VC system in-leakage performance does not significantly degrade relative to the CR in-leakage test results. This program includes intrusion from toxic gas, unfiltered in-leakage, and smoke.

-20 3.1.1.6.2 CR Direct Shine Dose Assumptions RG 1.183, Regulatory Position 4.2.1 states that "The TEDE analysis should consider all sources of radiation that will cause exposure to control room personnel. The applicable sources will vary from facility to facility, but typically will include:

Contamination of the control room atmosphere by the intake or infiltration of the radioactive material contained in the radioactive plume released from the facility, Contamination of the control room atmosphere by the intake or infiltration of airborne radioactive material from areas and structures adjacent to the control room envelope, Radiation shine from the external radioactive plume released from the facility, Radiation shine from radioactive material in the reactor containment, and Radiation shines from radioactive material in systems and components inside or external to the control room envelope, e.g., radioactive material buildup in recirculation filters." In accordance with requlatory guidance, the licensee considered all applicable sources contributing to the CR TEDE. In a letter dated October 6, 2008, the licensee provided additional information describing the details of how each individual contribution to the CR direct shine dose was evaluated. The licensee conservatively maximized the individual contributions to CR dose for each of the sources evaluated. For instance, in the evaluation of the dose from filtration units in close proximity to the CR, the licensee used a filter efficiency of 100 percent to maximize the filter loading and consequently maximize the direct shine dose to the CR. The licensee applied this approach to each of the direct shine components to CR dose to ensure conservative results. The NRC staff reviewed the licensee's assumptions for each individual contribution to CR direct shine dose and finds that the analysis is conservative, consistent applicable regulatory quldance and is, therefore, acceptable. The 30-day direct shine dose to personnel in the CR, considering the appropriate occupancy factors, is provided in Table 4 of this SE.

3.2 Atmospheric

Dispersion Estimates

3.2.1 Meteorological

Data In response to comments in Section 3.3.1.2 of the McGuire 1 and 2 AST fuel-handling accident SE associated with Amendments 236 and 218, respectively, dated December 22, 2006 (ADAMS Accession No. ML063100406), Duke generated new CR atmospheric dispersion factors (X/Q values) using only data from the upgraded meteorological tower which became operational in 1998. The meteorological data used in the current LAR were collected from 2001 through 2005. Wind speed and wind direction data were measured at heights of 10 meters and 60 meters above the ground.

Temperature difference data, which were used to determine atmospheric stability class, were measured between the 60-meter and 1O-meter levels. The data was provided for NRC staff review in the form of hourly meteorological data files for input into the ARCON96 atmospheric dispersion computer code (NUREG/CR-6331, Revision 1, "Atmospheric Relative Concentrations in Building Wakes"). The resulting CR X/Q values represent a change from those

-21 presented in the current McGuire 1 and 2 UFSAR. The licensee used the current licensing basis X/O values listed in the McGuire 1 and 2 UFSAR to assess dose consequences for postulated releases to the EAB and LPZ. NRC staff performed a quality review of the ARCON96 hourly meteorological database using the methodology described in NUREG-0917, "Nuclear Regulatory Commission Staff Computer Programs for Use with Meteorological Data." Further review was performed using computer spreadsheets. Data files provided as part of the March 20, 2008 (ML080930505) LAR contained formatting errors which were subsequently amended. Files containing the final revisions were submitted by letter dated December 17,2008 (ML083590342).

Examination of these revised data files revealed that stable and neutral atmospheric conditions were generally reported to occur at night and unstable and neutral conditions during the day, as expected. Wind speed, wind direction, and stability class frequency distributions for each measurement channel were very similar from year to year. Winds were predominantly from the southwest and northeast at both 10 and 60 m. With respect to the reported atmospheric stability measurements, the daily duration of stable and unstable conditions were consistent with expected meteorological conditions as were the occurrences with respect to time of day. Stable and neutral conditions were generally reported to occur at night and unstable and neutral conditions during the day. The combined data recovery of the wind speed, wind direction, and stability data was in the mid-to upper 90 percentiles at both levels during each of the 5 years. This meets the data recovery recommendation of Regulatory Guide (RG) 1.23, Revision 0, "Onsite Meteorological Programs." In summary, the NRC staff has reviewed available information relative to the onsite meteorological measurements program, the 2001 through 2005 meteorological data measured at the McGuire 1 and 2 site, and the ARCON96 meteorological data input files provided by the licensee. On the basis of this review, the NRC staff has concluded that these data provide an acceptable basis for making estimates of atmospheric dispersion for design basis accident CR dose assessments.

3.2.2 Control

Room Atmospheric Dispersion Factors In the LAR dated March 20, 2008, the licensee postulated releases from the following McGuire 1 and 2 locations to the McGuire 1 and 2 CR air intakes.

Equipment hatch (EO) -point

& diffuse source Fuel building (FUEL) -diffuse source Inboard doghouse (NDOG) -diffuse source Inboard doghouse plus steam

& feedwater line penetrations (VNDOG) -diffuse source Outboard doghouse (ODOG) -diffuse source Outboard doghouse plus steam

&feedwater line penetrations (VODOG) -point & diffuse source Reactor building surface (RX) -diffuse source FWST -point source Unit vent (UV) -point source Containment purge supply intake vents (VP) -point source Steam & feedwater line penetrations of inboard doghouse (AGin) -point source Steam & feedwater line penetrations of outboard doghouse (AGout) -point source Steam generator power-operated relief valves on inboard doghouse (PORVin) -point source

-22 Steam generator power-operated relief valves on outboard doghouse (PORVout) -point source Main steam safety valves on inboard doghouse (MSSVin) -point source Main steam safety valves on outboard doghouse (MSSVout) -point source Turbine driven auxiliary feedwater pump exhaust vents (AFW) -point source Postulated releases from the EO, FWST, UV and VP were associated with the LOCA analysis in the current LAR. With regard to the other locations, in to the letter dated December 17, 2008, the licensee stated that the full population of CR X/a values associated with potential release locations which could impact any of the design basis accidents was provided in the current LAR to satisfy the condition imposed in the SE associated with McGuire 1 and 2, Amendments 236 and 218, dated December 22, 2006, concerning use of data collected only from the upgraded meteorological tower. The licensee noted that a review of the recomputed CR X/a values would support their use in future implementation of AST for the remaining design basis accidents at McGuire 1 and 2. The licensee initially generated CR X/a values using the ARCOl\l96 computer code for all release locations except four sets associated with the McGuire 1 RX and MSSVout to McGuire 1 CR intake and McGuire 2 RX and MSSVout to McGuire 2 CR intake. RG 1.194 states that if the distance to the receptor is less than about 10 meters, the ARCON96 code should not be used to assess X/a values and that such situations should be analyzed on a case-by-case basis. Because the distances between these source/receptor pairs were initially judged to be less than 10 meters, the licensee proposed use of the Murphy-Campe methodology discussed in RG 1.194 as an alternative to the ARCON96 methodology for generating X/a values. However, RG 1.194 also states that the Murphy-Campe methodology should not be used when the distance to the receptor is less than about 10 meters. As a result of NRC staff requests for additional information, including further justification for use of the Murphy-Campe methodology, diffuse source assumptions and source/receptor distances and direction inputs, the licensee reviewed the release locations originally identified in the LAR dated March 20, 2008, by performing an examination of detailed plant drawings and conducting walk-downs. As a result, the licensee determined that the list of release locations could be shortened and simplified and that certain inputs and resultant X/a values should be updated as discussed in letters dated May 28, 2008 (ADAMS Accession No. ML081560395), December 17, 2008 (ADAMS Accession No. ML083590342), and February 12, 2009 (ADAMS Accession No. ML090540682). The licensee identified the revised list of release points to be considered for the full scope AST LAR as the EO, FWST, UV and VP, all associated with the LOCA analysis, and the PORVin, PORVout, MSSVin, MSSVout, and AFW, as the principal secondary system release locations. The RX, FUEL, NDOG, ODOG, VNDOG, VODOG, AGin and AGout release locations were removed from consideration under the current LAR. All sources were modeled as ground level releases using the ARCON96 methodology, based upon guidance provided in RG 1.194. The review was simplified by concentrating on the limiting point release/receptor combination(s) for each release location.

-23 EQ was modeled as both a point and diffuse source, but was retained as a point source because that assessment resulted in more limiting X/Q values than those generated assuming a diffuse release.

Therefore, all of the diffuse sources were removed from further consideration. As part of the review of more detailed plant drawings and walk-downs, the licensee re-measured and determined that the separation distance between the most-limiting MSSV and the associated same unit CR intake was greater than 10 meters. With the revision to the release locations under consideration in the current LAR including deletion of RX, the distances between all remaining source/receptor pairs were greater than 10 meters. Thus, all X/Q values were generated using the ARCON96 computer code and the Murphy-Campe methodology was no longer applied in the current LAR. RG 1.194, states that ARCON96 is an acceptable methodology for assessing CR X/Q values for use in design basis accident radiological analyses. NRC staff evaluated the applicability of the ARCON96 model and concluded that there are no unusual siting conditions, 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 McGuire 1 and 2. For the AFW, the licensee generated X/Q values for both straight-line and arc distances and wind directions between the source and receptor locations and proposed use of the limiting case, which varied as a function of time. The CR air intake X/Q values for the analyses discussed above were used to analyze unfiltered in-leakage to the CR for the LOCA dose assessment. The licensee stated that the most likely and most bounding path for control room in-leakage is via the CR ventilation system, including its inlets, which provide the shortest and easiest flow path for unfiltered radioactivity to enter the CR. McGuire 1 and 2 has dual intakes. As part of the current LAR, the licensee applied an adjustment factor following guidance in RG 1.94 to take credit for an assumed CR intake flow split of 65/35 percent between the two intakes which resulted in a 35 reduction in the X/Q values used in the LOCA dose assessment.

These adjusted LOCA X/Q values are listed in Table 2.1. In summary, the NRC staff qualitatively reviewed the inputs to the ARCON96 computer runs for the CR X/Q value assessment and found them generally consistent with site configuration drawings and staff practice. The NRC staff also reviewed the licensee's dispersion modeling. The resulting CR X/Qvalues are presented in Table 2.2. On the basis of this review, the NRC staff has concluded that the X/Q values listed in Tables 2.1 and 2.2 are acceptable for use in CR design basis accident dose assessment. The NRC staff has not determined the acceptability of CR X/Q values for the RX, FUEL, NDOG, ODOG, VNDOG, VODOG, AGin and AGout release locations which were removed from consideration under the current LAR. 3.2.3 Offsite Atmospheric Dispersion Factors The licensee used the current McGuire 1 and 2 licensing basis EAB and LPZ X/Q values listed in Table 2.3 to assess the radiological consequences of the LOCA postulated in this LAR. These X/Q values are presented in the McGuire 1 and 2 UFSAR in Table 15-33 which lists parameters

-applicable to the LOCA. They were also previously approved in McGuire 1 and 2 Amendments 236 and 218 concerning selective implementation of the alternative source term radiological analysis methodology.

3.3 Technical

Specification Changes The AST LAR for McGuire 1 and 2 does not contain any Technical Specification changes. 3.4 Commitments The AST LAR for McGuire 1 and 2 does not contain any licensee commitments.

3.5 Summary

of Technical Evaluation As described above, the NRC staff reviewed the assumptions, inputs, and methods used by the licensee to assess the radiological consequences of a DBA LOCA for full implementation of an AST at McGuire 1 and 2. The NRC staff concludes that the licensee used methods of analysis and assumptions consistent with the conservative regulatory requirements and guidance described in Section 2.0 above. The NRC staff compared the doses estimated by the licensee to the applicable dose guidelines and criteria referenced in Section 2.0, "Regulatory Evaluation." Based on that comparison, the NRC staff concludes that there is a reasonable assurance that the licensee's estimates of the EAB, LPZ, and CR doses comply with the regulatory requirements. The NRC staff also concludes that there is reasonable assurance that McGuire 1 and 2, as modified by the requested license amendment, will continue to provide sufficient safety margins and adequate defense-in-depth, under conditions of unanticipated events, and in presence of the uncertainties in accident progression, assumptions, parameters, and analyses outlined above. Therefore, the proposed changes to the licensing basis are acceptable with respect to the radiological consequence of the DBA LOCA. This licensing action is considered a full implementation of the AST. With this approval, the previous accident source term in the McGuire 1 and 2 licensing basis is superseded by the revised licensing basis, incorporating the AST as proposed by the licensee. The previous offsite and CR accident dose criteria expressed in terms of whole body, thyroid, and skin doses are superseded by the TEDE guidelines and criteria of 10 CFR 50.67, or fractions thereof, as defined in RG 1.183. All future radiological accident analyses performed to show compliance with regulatory requirements shall address all characteristics of the AST and the TEDE criteria as defined the McGuire 1 and 2 design basis, and modified by the present amendment. FINAL NO SIGNIFICANT HAZARDS CONSIDERATION The Commission's regulations in 10 CFR 50.92(c), "Issuance of amendment," state that the Commission may make a final determination that a license amendment involves no significant hazards consideration if operation of the facility in accordance with the amendment would not Involve a significant increase in the probability or consequences of an accident previously evaluated; or

-25 (2) Create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) Involve a significant reduction in a margin of safety. The following analysis was provided by the licensee in its letter dated March 20, 2008: Does this LAR involve a significant increase in the probability or consequences of an accident previously evaluated? No. AST is an updated methodology used to evaluate the dose consequences of the Loss of Coolant Accident (LOCA). This type of change is analytical, thus, does not increase the probability of an accident previously evaluated. It has been demonstrated that the dose consequences of the re-analyzed accident remain within the dose limits of 10 CFR 50.67 and Regulatory Guide 1.183. This proposed change assumes an increase in the amount of unfiltered air in-leakage into the control room. The current Technical Information Document (TID) based McGuire dose consequence analysis for the LOCA assumed control room unfiltered in-leakage of 10 scfm. Tracer gas testing performed at McGuire revealed that unfiltered in-leakage into the control room exceeded this amount by as much as 167 scfm as discussed in McGuire's response to NRC GL 2003-01 dated February 19, 2004. Use of the AST methodology can accommodate a larger control room pressurization unfiltered in-leakage rate without exceeding any regulatory dose limits.

A comparison of the AST analysis results and the TID values (UFSAR Table 15-12) shows that the EAB and LPZ (off-site) doses decrease while the control room dose increases. The new AST based analysis not only implements changes which affect both off-site and control room doses, such as the change in source term methodology, it also includes changes to the LOCA model which only impact the control room dose, and are responsible for the increased result. These new attributes include a control room in-leakage model that reflects the control room tracer gas testing results and a recomputed control room shine component of the post LOCA control room dose. The dose consequences of the revised analysis. however, are below the 10 CFR 50.67 acceptance criteria for both off-site and control room doses and are not considered a significant increase.

AST radiological methodology does not adversely affect accident initiators or precursors. Nor will it alter or prevent the ability of structures, systems, and components from performing their intended function to mitigate the consequences of an accident.

Therefore, this LAR will not involve a significant increase in the probability or consequences of an accident previously evaluated. Does this LAR create the possibility of a new or different kind of accident from any accident previously evaluated? No. AST is an updated methodology that was used to re-evaluate the dose consequences of the McGuire UFSAR previously analyzed accidents. This new analysis does not cause any change in the post accident operation of any plant system, structure, or component.

-26 This LAR does not involve an addition or modification to any plant system, structure, or component. This change does not affect the post accident operation of any plant system, structure, or component as directed in plant procedures. New or modified equipment or personnel failure modes that might initiate a new or different type accident are not created as a result of the proposed change.

Therefore, no new or different accident is created by changing to the AST methodology prescribed in Regulatory Guide 1.183.

3. Does this LAR involve a significant reduction in a margin of safety? No. Margin of safety is related to the confidence in the ability of the fission product barriers to perform their design functions during and following accident conditions. These barriers include the fuel cladding, the reactor coolant system, and the containment system. The proposed re-analysis of the LOCA dose consequences using AST will have no affect on the performance of these barriers. This LAR does not involve an addition or modification to any plant system, structure, or component. This change will not affect the post accident operation of any plant system, structure, or component as directed in plant procedures. Therefore, the LAR will not involve a significant reduction in a margin of safety. The NRC staff has reviewed the licensee's analysis and, based on this review, has concluded that the three standards of 10 CFR 50.92(c) are satisfied.

Therefore, the NRC staff has made a final determination that the proposed amendment involves no significant hazards consideration.

5.0 STATE CONSULTATION

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

6.0 ENVIRONMENTAL

CONSIDERATION The amendments change a requirement with respect to the installation or use of facility components located within the restricted area as defined in 10 CFR Part 20. The NRC staff has determined that the amendments involve 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 amendments involve no significant hazards consideration, and there has been no public comment on such finding (74 FR 9009).

Accordingly, the amendments meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9).

Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendments.

-

7.0 CONCLUSION

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

Principal Contributors: John Parillo Leta Brown Date: March 31,2009

-28 Table McGuire Units 1 and 2 Radiological Consequences Expressed as TEDE Design Basis Loss of Coolant Accident (LOCA) EAB (2) LPZ (3) CR(4) Containment effluent 8.2 1.7 2.2 ECCS effluent 1.3 0.23 0.80 Total effluent dose 9.5 1.9 3.0 External CR shine dose N/A N/A 1.3 Total LOCA dose 9.5 1.9 4.3 Dose acceptance criteria 25 25 5 (1) Total effective dose equivalent (2) Exclusion area boundary -worst 2-hour dose (3) Low population zone -integrated 30 day dose (4) Unfiltered CR inleakage 625 cfm prior to pressurization; 210 cfm after pressurization Note: Licensee's results are expressed to a limit of two significant figures.

-29 Table 2.1 McGuire LOCA CR Atmospheric Dispersion Factors (X/Q, With Reduction Credit Due to 65/35 Flow Split Between Time interval EO FWST UV 0-2 hours 2.66 x 10-3 1.20 X 10-3 1.10 x 10-3 2 -8 hours 2.35 x 10-3 1.11 x 10-3 1.01 X 10-3 8 -24 hours 1.12 x 10-3 5.26 x 10-4 4.89 X 10-4 1 -4 days 8.32 x 10-4 3.88 x 10-4 3.69 X 10-4 4 -30 days 6.49 x 10-4 3.00 X 10-4 2.81 x 10-4 EO -equipment hatch FWST -refueling water storage tank UV -unit vent

-30 Table 2.2 McGuire LOCA CR Atmospheric Dispersion Factors (X/Q, sec/m 3Without 65/35 Flow Time interval EO FWST UV Vp* 0-2 hours 4.09 x 10-3 1.84 X 10-3 1.69 X 10-3 1.58 X 10-3 0-4 hours 3.88 x 10-3 1.80 X 10-3 1.65 X 10-3 1.50 X 10-3 0-8 hours 3.73 x 10-3 1.75 X 10-3 1.59 X 10-3 1.38 X 10-3 2 -8 hours 3.61 x 10-3 1.71 X 10-3 1.56 X 10-3 1.31 X 10-3 4 -8 hours 3.58 x 10-3 1.70 X 10-3 1.53 X 10-3 1.26 X 10-3 8 -24 hours 1.72 x 10-3 8.09 X 10-4 7.52 X 10-4 5.51 x 10-4 1 -4 days 1.28 x 10-3 5.97 X 10-4 5.68 X 10-4 4.20 X 10-4 4 -30 days 9.99 x 10.4 4.62 X 10-4 4.32 X 10-4 3.49 x 10.4 EO -equipment hatch FWST -refueling water storage tank UV -unit vent VP -containment purge supply intake vents

• VP X/O values were not used in LOCA dose assessment, but are provided to show that they are limited by UV X/O values.

-31 Table 2.2 McGuire Secondary Side CR Atmospheric Dispersion Factors (X/Q Values, Without 65/35 Flow Time Interval PORVin PORVout MSSVin MSSVout AFW 0-2 hours 3.03 x 10-3 6.71 X 10-3 3.56 x 10-3 9.70 X 10-3 5.22 x 10-3 0-4 hours 2.90 x 10-3 6.52 x 10-3 3.45 x 10-3 9.44 X 10-3 4.92 x 10-3 0-8 hours 2.81 x 10-3 6.23 x 10-3 3.32 x 10-3 9.04 X 10-3 4.42 x 10-3 2 -8 hours 2.73 x 10-3 6.07 x 10-3 3.24 x 10-3 8.82 X 10-3 4.16 X 10-3 4 -8 hours 2.72 x 10-3 5.94 x 10-3 3.19 x 10-3 8.64 X 10-3 3.97 x 10-3 8 -24 hours 1.30 x 10-3 2.83 X 10-3 1.44 x 10-3 4.19x10-3 1.92 X 10-3 1 -4 days 9.57 x 10-4 2.17 X 10-3 1.09 x 10-3 3.20 X 10-3 1.40 X 10-3 4 -30 days 7.40 x 10-4 1.68 X 10-3 8.16x10-4 2.46 X 10-3 1.70 X 10-3 PORVin -steam generator power powered relief valves on inboard doghouse PORVout -steam generator power powered relief valves on outboard doghouse MSSVin -main steam safety valves on inboard doghouse MSSVout -main steam safety valves on outboard doghouse AFW -turbine driven auxiliary feedwater pump exhaust vents Table 3 McGuire EAB and LPZ Atmospheric Dispersion Factors (XlQ Values, sec/m') EAB 0-2 hours 9.0x10-4 LPZ 0-8 hours 8.0x10-s 8-24 hours 5.2 x 10-6 1-4 days 1.7 x 10-6 4-30 days 3.7 x 10-7

-32 Table McGuire Units 1 and 2 LOCA Control Room Data and and Direct Shine Control room volume 1.07 x 10+5 ft3 Prior to pressurization

-normal operation Assumed unfiltered inleakage 625 cfm Nominal test value 505 15 cfm Emergency operation assuming a single VC fan and train

@ T =30 seconds Filtered Make-up Flow Rate 1800 cfm Filtered Recirculation Flow Rate Not modeled Assumed unfiltered inleakage 210 cfm Nominal test value 131 36 cfm Duel intake flow split Contaminated air stream 65% Non-contaminated air stream 35% Credited control room filter efficiencies Particulates 99% Elemental iodine 98.1% Organic iodine 98.1% CR operator breathing rate 3/sec 0-720 hours 3.5 x 10-4 m CR occupancy factors 0-24 hours 1.0 24 -96 hours 0.6 96 -720 hours 0.4 LOCA CR direct shine dose 1.26 rem

-33 Table 5 (Page 1 of 3)

McGuire Units 1 and 2 Data and Assumptions for the LOCA Core Power level 3479 MWt (includes 2% uncertainty) Core Average Fuel Burnup 44,000 MWO/MTU Fuel Enrichment 3.5 -5 weight percent (w/o) Upper Containment Volume 826,752 ft3 Lower Containment Volume 370,623 ft3 Total Containment Volume 1,197,375 ft3 Primary containment leak rate 0-24 hours 0.3% (by weight)/day 24 -720 hours 0.15% (by weight)/day Start Annulus exhaust 39 seconds Annulus vacuum established 71 seconds Annulus bypass fraction and bypass leak paths 7% Equipment hatch 5% of bypass flow Unit vent stack 95% of bypass flow Credited Spray Lambdas (As) for Minimum Safeguards Scenario Start time (seconds) End time (seconds)

Elemental AS Particulate AS o 600 1 0 0 600 3000 20 9.36 3000 3240 0.22 7.19 3240 3500 0.50 16.5 3500 4000 0.53 16.5 4000 4500 0.56 16.5 4500 5000 0.58 16.5 5000 7100 0.59 16.5 7100 24,600 0.59 1.65 2 24,600 30,000 0.58 1.65 30,000 40,000 0.56 1.65 40,000 46,000 0.53 1.65 46,000 86,400 0 (No credit)" 1.65 86,400 end 0 (No credit)' 0 (No credn)" 1 McGuire spray flow starts by 120 seconds but is not credited until VX starts at 600 2 Particulate AS are reduced by a factor of 10 due to reaching a OF of 3 Elemental OF reaches 200 and credit for elemental spray removal 4 No spray credit after 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> (86,400

-Table 5 (Page 2 of 3)

McGuire Units 1 and 2 Data and Assumptions for the LOCA Time steps for sump water volume model Sump Volume (sec) (hrs) fe o 0 o 45 0.0125 19,000 1560 0.4333 56,240 1800 0.5 59,600 1830 0.5083 60,020 3000 0.8333 72,140 3600 1.0 74,075 4800 1.333 76,900 6000 1.667 77,300 8700 2.417 77,400 10,200 2.833 77,600 ECCS Leakage to Auxiliary building (2 times limit) 1 gpm ECCS Iodine Partition Fractions and release rates for one gpm leakage to Auxiliary Building Start time (hr) End time (hr) Iodine partition factor Release rate (cfm) o 3 0.100 0.01337 3 72 0.028 0.00374 72 720 0.010 0.001337 Chemical form of released iodine from ECCS leakage Elemental 97% Organic 3% Particulate 0% Auxiliary building and Annulus ventilation filter efficiencies Elemental 92% Organic 92% Particulate 98%

-35 Table 5 (Page 3 of McGuire Units 1 and 2 Data and Assumptions for the FWST Release Model for 20 gpm ECCS Back-leakage Start time (sec) o End time (Release fraction o 9.197 x 10-1 2.894 X 10-3.443 X 10-9.799 x 10-1.772 X 10-3.486 x 10-4.228 x 10-4.128 X 10-3.916 X 10-3.376 x 10-3.284 X 10-1.873 X 10-3.444 X 10-6.388 X 10-Release rate (cfm) o 2.459 x 10-1 7.739 X 10-9.207 x 10-2.620 x 10-4.738 X 10-9.321 X 10-1.131 X 10-1.104 x 10-1.047 x 10-9.027 x 10-8.781 X 10-5.008 x 10-9.209 X 10-1.708 X 10-

ML 090890627 OFFICE NRRlLPL2-1/PM NRRlLPL2-1/lA DRAlAPLAlBC' DSS/SCVB/BC DCI/CSGBI ABC" OGC NRRlLPL2-1/BC NRRlLPL2-1/PM NAME JThompson MO'Brien RTaylor RDennig MYoder AJones MWong JStang DATE 03111/09 03111/09 02126/09 03117109 03126/09 03127109 03131/09 03130109