Text

Issuance of Amendments 269 and 273 Application of Alternative Source Term Methodology
	ML082320406
Person / Time
Site:	Peach Bottom ; (DPR-044, DPR-056)
Issue date:	09/05/2008
From:	John Hughey; NRC/NRR/ADRO/DORL/LPLI-2
To:	Pardee C; Exelon Generation Co
	Hughey J, NRR/DORL, 301-415-3204
Shared Package
ML082320257	List: ML082320406; ML082490831;
References
	TAC MD6806, TAC MD6807
	Download: ML082320406 (52)
	v • d • e

September 5, 2008 Mr. Charles G. Pardee Chief Nuclear Officer and Senior Vice President Exelon Generation Company, LLC 200 Exelon Way Kennett Square, PA 19348

SUBJECT:

PEACH BOTTOM ATOMIC POWER STATION, UNITS 2 AND 3 - ISSUANCE OF AMENDMENTS RE: APPLICATION OF ALTERNATIVE SOURCE TERM METHODOLOGY (TAC NOS. MD6806 AND MD6807)

Dear Mr. Pardee:

The Nuclear Regulatory Commission has issued the enclosed Amendment Nos. 269 and 273 to Renewed Facility Operating License Nos. DPR-44 and DPR-56 for the Peach Bottom Atomic Power Station (PBAPS), Units 2 and 3. These amendments consist of changes to the Technical Specifications (TSs) in response to your application dated July 13, 2007, (Agencywide Documents Access and Management System (ADAMS) Accession No. ML072570151), as supplemented by letters dated February 28, 2008, March 28, 2008, April 17, 2008, May 23, 2008, July 29, 2008, August 7, 2008, and August 21, 2008 (ADAMS Accession Nos.

ML080670277, ML080910502, ML081090447, ML081500373, ML082140237, ML082260396 and ML082340796, respectively).

The amendments revise the PBAPS Units 2 and 3 TSs to support application of an alternative source term methodology. A copy of our Safety Evaluation is also enclosed and a Notice of Issuance will be included in the Commissions biweekly Federal Register notice.

Sincerely,

/ra/

John D. Hughey, Project Manager Plant Licensing Branch I-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-277 and 50-278

Enclosures:

1. Amendment No. 269 to Renewed DPR-44

2. Amendment No. 273 to Renewed DPR-56

3. Safety Evaluation cc w/encls: See next page

Peach Bottom Atomic Power Station, Unit Nos. 2 and 3 cc:

Site Vice President Peach Bottom Atomic Power Station Exelon Generation Company, LLC 1848 Lay Road Delta, PA 17314 Plant Manager Peach Bottom Atomic Power Station Exelon Generation Company, LLC 1848 Lay Road Delta, PA 17314 Regulatory Assurance Manager Peach Bottom Atomic Power Station Exelon Generation Company, LLC 1848 Lay Road Delta, PA 17314 Chief Operating Officer (COO)

Exelon Nuclear Exelon Generation Company, LLC 4300 Winfield Road Warrenville, IL 60555 Resident Inspector U.S. Nuclear Regulatory Commission Peach Bottom Atomic Power Station P.O. Box 399 Delta, PA 17314 Regional Administrator, Region I U.S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406 Mr. Roland Fletcher Department of Environment Radiological Health Program 2400 Broening Highway Baltimore, MD 21224 Correspondence Control Desk Exelon Generation Company, LLC 200 Exelon Way, KSA 1N1 Kennett Square, PA 19348 Director, Bureau of Radiation Protection Pennsylvania Department of Environmental Protection Rachel Carson State Office Building P.O. Box 8469 Harrisburg, PA 17105-8469 Board of Supervisors Peach Bottom Township 545 Broad Street Ext.

Delta, PA 17314-9203 Mr. Richard McLean Power Plant and Environmental Review Division Department of Natural Resources B-3, Tawes State Office Building Annapolis, MD 21401 Dr. Judith Johnsrud National Energy Committee Sierra Club 433 Orlando Avenue State College, PA 16803 Manager-Financial Control & Co-Owner Affairs Public Service Electric and Gas Company P.O. Box 236 Hancocks Bridge, NJ 08038-0236 Manager Licensing-Peach Bottom Atomic Power Station Exelon Generation Company, LLC P.O. Box 160 Kennett Square, PA 19348 Vice President - Licensing and Regulatory Affairs Exelon Generation Company, LLC 4300 Winfield Road Warrenville, IL 60555

Peach Bottom Atomic Power Station, Unit Nos. 2 and 3 cc:

Senior Vice President-Mid-Atlantic Operations Exelon Generation Company, LLC 200 Exelon Way, KSA 3-N Kennett Square, PA 19348 Senior Vice President - Operations Support Exelon Generation Company, LLC 200 Exelon Way, KSA 3-N Kennett Square, PA 19348 Director - Licensing and Regulatory Affairs Exelon Generation Company LLC Correspondence Control P.O. Box 160 Kennett Square, PA 19348 Associate General Counsel Exelon Generation Company, LLC 4300 Winfield Road Warrenville, IL 60555

ML080670277, ML080910502, ML081090447, ML081500373, ML082140237, ML082260396 and ML082340796, respectively).

The amendments revise the PBAPS Units 2 and 3 TSs to support application of an alternative source term methodology. A copy of our Safety Evaluation is also enclosed and a Notice of Issuance will be included in the Commissions biweekly Federal Register notice.

Sincerely,

/ra/

John D. Hughey, Project Manager Plant Licensing Branch I-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-277 and 50-278

Enclosures:

1. Amendment No. 269 to Renewed DPR-44

2. Amendment No. 273 to Renewed DPR-56

3. Safety Evaluation cc w/encls: See next page DISTRIBUTION:

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by memo dated

EXELON GENERATION COMPANY, LLC PSEG NUCLEAR LLC DOCKET NO. 50-277 PEACH BOTTOM ATOMIC POWER STATION, UNIT 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 269 License No. DPR-44

1.

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

A.

The application for amendment by Exelon Generation Company, LLC (Exelon Generation Company), and PSEG Nuclear LLC (the licensees), dated July 13, 2007, as supplemented by letters dated February 28, 2008, March 28, 2008, April 17, 2008, May 23, 2008, July 29, 2008, August 7, 2008, and August 21, 2008, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.

There is reasonable assurance: (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.

The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C(2) of Renewed Facility Operating License No. DPR-44 is hereby amended to read as follows:

(2)

Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 269, are hereby incorporated in the license. Exelon Generation Company shall operate the facility in accordance with the Technical Specifications.

3.

Implementation Requirements:

This license amendment is effective as of the date of issuance, and shall be implemented within 90 days of issuance. Implementation of the amendment shall include updating the UFSAR in accordance with 10 CFR 50.71(e). This update shall include, but not be limited to, a discussion that reflects that the minimum containment pressure available (MCPA) is maintained greater than or equal to the containment overpressure required (COPR).

FOR THE NUCLEAR REGULATORY COMMISSION

/ra/

Harold K. Chernoff, Chief Plant Licensing Branch I-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the License and Technical Specifications Date of Issuance: September 5, 2008

ATTACHMENT TO LICENSE AMENDMENT NO. 269 RENEWED FACILITY OPERATING LICENSE NO. DPR-44 DOCKET NO. 50-277 Replace the following page of the Facility Operating License with the attached revised page.

The revised page is identified by amendment number and contains marginal lines indicating the areas of change.

Remove Insert Page 3 Page 3 Replace the following pages of the Appendix A, Technical Specifications, with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

Remove Insert 1.1 - 2 1.1 - 2 1.1 - 5 1.1 - 5 3.1 - 20 3.1 - 20 3.1 - 21 3.1 - 21 3.3 - 54 3.3 - 54 3.3 - 58 3.3 - 58 3.6 - 12 3.6 - 12 3.6 - 16 3.6 - 16 3.6 - 34 3.6 - 34 3.6 - 35 3.6 - 35 3.6 - 36 3.6 - 36 3.6 - 38 3.6 - 38 3.6 - 40 3.6 - 40 3.6 - 41 3.6 - 41 5.0 - 17 5.0 - 17 5.0 - 18 5.0 - 18

(5)

Exelon Generation Company, pursuant to the Act and 10 CFR Parts 30 and 70, to possess, but not to separate, such byproduct and special nuclear material as may be produced by operation of the facility.

C.

This renewed license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations in 10 CFR Chapter I: Part 20, Section 30.34 of Part 30, Section 40.41 of Part 40, Section 50.54 of Part 50, and Section 70.32 of Part 70; all applicable provisions of the Act and the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified below:

(1)

Maximum Power Level Exelon Generation Company is authorized to operate the Peach Bottom Atomic Power Station, Unit 2, at steady state reactor core power levels not in excess of 3514 megawatts thermal.

(2)

Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 269, are hereby incorporated in the license. Exelon Generation Company shall operate the facility in accordance with the Technical Specifications.

(3)

Physical Protection Exelon Generation Company shall fully implement and maintain in effect all provisions of the Commission-approved physical security, training and qualification, and safeguards contingency plans including amendments made pursuant to provisions of the Miscellaneous Amendments and Search Requirements revisions to 10 CFR 73.55 (51 FR 27817 and 27822), and the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The combined set of plans1, submitted by letter dated May 17, 2006, is entitled: "Peach Bottom Atomic Power Station Security Plan, Training and Qualification Plan, Safeguards Contingency Plan, and Independent Spent Fuel Storage Installation Security Program, Revision 3." The set contains Safeguards Information protected under 10 CFR 73.21.

(4)

Fire Protection The Exelon Generation Company shall implement and maintain in effect all provisions of the approved fire protection program as described in the Updated Final Safety Analysis Report for the facility, and as approved in the NRC Safety Evaluation Report (SER) dated May 23, 1979, and Supplements dated August 14, September 15, October 10 and November 24, 1980, and in the NRC SERs dated September 16, 1993, and August 24, 1994, subject to the following provision:

The Exelon Generation Company may make changes to the approved 1 The Training and Qualification Plan and Safeguards Contingency Plan are Appendices to the Security Plan.

Renewed License No. DPR-44 Revised by letter dated October 28, 2004 Revised by Letter dated May 29, 2007 Amendment No. 269 Page 3

EXELON GENERATION COMPANY, LLC PSEG NUCLEAR LLC DOCKET NO. 50-278 PEACH BOTTOM ATOMIC POWER STATION, UNIT 3 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 273 License No. DPR-56

1.

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

A.

The application for amendment by Exelon Generation Company, LLC (Exelon Generation Company), and PSEG Nuclear LLC (the licensees), dated July 13, 2007, as supplemented by letters dated February 28, 2008, March 28, 2008, April 17, 2008, May 23, 2008, July 29, 2008, August 7, 2008, and August 21, 2008, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.

There is reasonable assurance: (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.

The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C(2) of Renewed Facility Operating License No. DPR-56 is hereby amended to read as follows:

(2)

Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 273, are hereby incorporated in the license. Exelon Generation Company shall operate the facility in accordance with the Technical Specifications.

3.

Implementation Requirements:

This license amendment is effective as of the date of issuance, and shall be implemented within 90 days of issuance. Implementation of the amendment shall include updating the UFSAR in accordance with 10 CFR 50.71(e). This update shall include, but not be limited to, a discussion that reflects that the minimum containment pressure available (MCPA) is maintained greater than or equal to the containment overpressure required (COPR).

FOR THE NUCLEAR REGULATORY COMMISSION

/ra/

Harold K. Chernoff, Chief Plant Licensing Branch I-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the License and Technical Specifications Date of Issuance: September 5, 2008

ATTACHMENT TO LICENSE AMENDMENT NO. 273 RENEWED FACILITY OPERATING LICENSE NO. DPR-56 DOCKET NO. 50-278 Replace the following page of the Facility Operating License with the attached revised page.

The revised page is identified by amendment number and contains marginal lines indicating the areas of change.

Remove Insert Page 3 Page 3 Replace the following pages of the Appendix A, Technical Specifications, with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

Remove Insert 1.1 - 2 1.1 - 2 1.1 - 5 1.1 - 5 3.1 - 20 3.1 - 20 3.1 - 21 3.1 - 21 3.3 - 54 3.3 - 54 3.3 - 58 3.3 - 58 3.6 - 12 3.6 - 12 3.6 - 16 3.6 - 16 3.6 - 34 3.6 - 34 3.6 - 35 3.6 - 35 3.6 - 36 3.6 - 36 3.6 - 38 3.6 - 38 3.6 - 40 3.6 - 40 3.6 - 41 3.6 - 41 5.0 - 17 5.0 - 17 5.0 - 18 5.0 - 18

(5)

Exelon Generation Company, pursuant to the Act and 10 CFR Parts 30 and 70, to possess, but not to separate, such byproduct and special nuclear material as may be produced by operation of the facility.

C.

This renewed license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations in 10 CFR Chapter I: Part 20, Section 30.34 of Part 30, Section 40.41 of Part 40, Section 50.54 of Part 50, and Section 70.32 of Part 70; all applicable provisions of the Act and the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified below:

(1)

Maximum Power Level Exelon Generation Company is authorized to operate the Peach Bottom Atomic Power Station, Unit No. 3, at steady state reactor core power levels not in excess of 3514 megawatts thermal.

(2)

Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 273, are hereby incorporated in the license. Exelon Generation Company shall operate the facility in accordance with the Technical Specifications.1 (3)

Physical Protection Exelon Generation Company shall fully implement and maintain in effect all provisions of the Commission-approved physical security, training and qualification, and safeguards contingency plans including amendments made pursuant to provisions of the Miscellaneous Amendments and Search Requirements revisions to 10 CFR 73.55 (51 FR 27817 and 27822), and the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The combined set of plans2, submitted by letter dated May 17, 2006, is entitled: Peach Bottom Atomic Power Station Security Plan, Training and Qualification Plan, Safeguards Contingency Plan, and Independent Spent Fuel Storage Installation Security Program, Revision 3." The set contains Safeguards Information protected under 10 CFR 73.21.

(4)

Fire Protection The Exelon Generation Company shall implement and maintain in effect all provisions of the approved fire protection program as described in the Updated Final Safety Analysis Report for the facility, and as approved in 1Licensed power level was revised by Amendment No. 250, dated November 22, 2002, and will be implemented following the 14th refueling outage currently scheduled for Fall 2003.

2The training and Qualification Plan and Safeguards Contingency Plan and Appendices to the Security Plan.

Renewed License No. DPR-56 Revised by letter dated October 28, 2004 Revised by letter dated November 5, 2004 Revised by letter dated May 29, 2007 Amendment No. 273 Page 3

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 269 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-44 AND AMENDMENT NO. 273 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-56 EXELON GENERATION COMPANY, LLC PSEG NUCLEAR LLC PEACH BOTTOM ATOMIC POWER STATION, UNITS 2 AND 3 DOCKET NOS. 50-277 AND 50-278

1.0 INTRODUCTION

By letter dated July 13, 2007, (Agencywide Documents Access and Management System (ADAMS) Accession No. ML072570151), Exelon Generation Company, LLC, the licensee for Peach Bottom Atomic Power Station (PBAPS or Peach Bottom), Units 2 and 3, requested to amend the Technical Specifications (TSs), Appendix A, of Renewed Facility Operating License numbers DPR-44 and DPR-56 for PBAPS Units 2 and 3. The amendments revise the PBAPS Units 2 and 3 TSs to support application of an alternative source term (AST) methodology. The application provides the TS changes and evaluations of the radiological consequences of dose consequence design basis accidents (DBAs) for implementation of a full-scope AST pursuant to Title 10 of the Code of Federal Regulations Section 50.67 (10 CFR 50.67), Accident source term, and using the methodology described in Regulatory Guide (RG) 1.183, Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors.

During its review, the Nuclear Regulatory Commission (NRC) staff determined that additional information was necessary. Exelon Generation Company, LLC (the licensee), responded to the NRCs requests for additional information (RAIs) by letters dated February 28, 2008, March 28, 2008, April 17, 2008, May 23, 2008, July 29, 2008, August 7, 2008, and August 21, 2008 (ADAMS Accession Nos. ML080670277, ML080910502, ML081090447, ML081500373, ML082140237, ML082260396 and ML082340796, respectively). The responses clarified the application, did not expand the scope of the application as originally noticed, and did not change the staffs original proposed no significant hazards consideration determination as published in the Federal Register on May 6, 2008 (73 FR 25040).

The NRC staff has completed its review and finds that that the requested modification is acceptable, as discussed in this safety evaluation (SE).

2.0 REGULATORY EVALUATION

In the early 1970s, the NRC staff issued regulatory guidance for evaluating the consequences of DBAs using the radiological source term described in Technical Information Document (TID)-

14844, Calculation of Distance Factors for Power and Test Reactor Sites. Since the publication of TID-14844, significant advances in understanding timing, magnitude, and chemical form of fission product releases from severe nuclear power plant accidents have occurred. In 1995, the NRC published NUREG-1465, Accident Source Terms for Light-Water Nuclear Power Plants. NUREG-1465 used updated research to provide more realistic estimates of the accident source term that were physically based on, and could be applied to, the design of future light-water power reactors. In addition, the NRC determined that the analytical approach based on the TID-14844 source term would continue to be adequate to protect public health and safety for the current licensed power reactors. The NRC staff also determined that current licensees may wish to use the NUREG-1465 source term referred to as the AST in analyses to support cost-beneficial licensing actions. The NRC staff, therefore, initiated several actions to provide a regulatory basis for operating reactors to use an AST in design basis analyses. These initiatives resulted in the development and issuance of 10 CFR 50.67 and RG 1.183 (July 2000). Issuance of RG 1.183 provided the first comprehensive guidance for analyzing design basis accidents for radiological consequences using the AST.

A holder of an operating license issued prior to January 10, 1997 (Operating Reactors), or a holder of a renewed license under 10 CFR Part 54 Conditions of licenses whose initial operating license was issued prior to January 10, 1997, can, in accordance with 10 CFR 50.67, voluntarily revise the AST used in design basis radiological consequence analyses. However, to ensure proper implementation of the AST, the NRC requires in 10 CFR 50.67(b) that A licensee who seeks to revise its current accident source term in design basis radiological consequence analyses shall apply for a license amendment under § 50.90. The application shall contain an evaluation of the consequences of applicable DBAs previously analyzed in the safety analysis report.

In addition to developing the AST and providing regulatory guidance for its implementation, the NRC determined that new dose criteria for protection of public health and safety were appropriate and included these performance-based criteria in 10 CFR 50.67. Paragraph (b)(2)(i) of 10 CFR 50.67 states, 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 0.25 sievert (Sv) or 25 roentgen equivalent in man (rem) total effective dose equivalent (TEDE). Paragraph (b)(2)(ii) of 10 CFR 50.67 states, 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 0.25 Sv (25 rem) TEDE. For those plants applying to use the AST, paragraph (b)(2)(iii) of 10 CFR 50.67 provides CR habitability criteria. It states, Adequate radiation protection is provided to permit access to and occupancy of the control room under accident conditions without personnel receiving radiation exposures in excess of 0.05 Sv (5 rem) total effective dose equivalent (TEDE) for the duration of the accident. An applicants analysis must demonstrate with reasonable assurance that the three criteria of 10 CFR 50.67(b)(2) are met.

A design basis radiological consequence analysis is intended to be based upon a major accident, or possible event, resulting in a dose consequence that is not exceeded by any accident considered credible (maximum hypothetical accident). Unlike the DBA loss-of-coolant accident (LOCA), used to evaluate the emergency core cooling system (ECCS) requirements of 10 CFR 50.46, the general scenario used to postulate a maximum hypothetical dose consequence AST does not represent any specific accident sequence. Rather, the maximum hypothetical accident is intended to be a surrogate to enable a deterministic evaluation of the response of a facilitys engineered safety features (ESFs) such as the primary containment system. Although the maximum hypothetical dose consequence LOCA is typically the maximum credible accident, NRC staff experience in reviewing license applications has indicated the need to consider other accident sequences of possible occurrence including other dose consequence DBAs such as the fuel handling accident (FHA). The accident analyses are intentionally conservative in order to compensate for known uncertainties in accident progression, airborne activity product transport, and atmospheric dispersion.

RG 1.183 states, in part, in Regulatory Position 1.2, that a complete implementation of an AST would upgrade all existing radiological analyses. Although a complete re-assessment of all facility radiological analyses would be desirable, the NRC staff determined that recalculation of all design analyses for operating reactors would generally not be necessary. Full implementation is a modification of the facility design basis that addresses all characteristics of the AST: composition and magnitude of the radioactive material, its chemical and physical form, and the timing of its release. Full implementation revises the facilitys licensing basis to specify the AST in place of the previous accident source term and establishes the new TEDE dose as the new acceptance criteria. This applies not only to the analyses performed in the application (which may only include a subset of the plant analyses), but also to all future design basis analyses. As a minimum for full implementation, the maximum credible dose consequence LOCA (maximum hypothetical accident) must be determined and then analyzed using the guidance in Appendix A of RG 1.183.

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

Licensees should analyze the DBAs that are affected by the specific proposed applications of an AST.

Regulatory Position 6 of RG 1.183 states, in part, that the NRC staff is assessing the effect of increased cesium releases on equipment environmental 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 TID-14844 assumptions for performing the required EQ analyses. This issue has been resolved as documented in a memo dated April 30, 2001, Initial Screening of Candidate Generic Issue 187, The Potential Impact of Postulated Cesium Concentration on Equipment Qualification in the Containment Sump (ADAMS Accession No. ML011210348) and in the June 2001 NUREG-0933, Supplement 25, A Prioritization of Generic Safety Issues (ADAMS Accession No. ML012190402). In NUREG-0933, Supplement 25, the conclusion to Generic Issue 187, The Potential Impact of Postulated Cesium Concentration on Equipment Qualification in the Containment Sump in Nuclear Power Plants, states the following:

The staff concluded that there was no clear basis for backfitting 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 considerations.

The regulatory requirements that the NRC staff used in performing its review are the accident dose criteria in 10 CFR 50.67(b)(2), as supplemented in Regulatory Position 4.4 of RG 1.183 and NUREG-0800 Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants (Standard Review Plan or SRP) Section 15.0.1. In addition, the NRC staff's evaluation is based upon the following regulatory codes, guides, and standards:

10 CFR 50.67, Accident source term.

10 CFR 50.44, Combustible gas control for nuclear power reactors.

10 CFR 100.11, Determination of exclusion area, low population zone, and population center distance.

RG 1.183, Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors, Rev. 0, July 2000.

RG 1.23, Onsite Meteorological Programs, Rev. 0, February 1972.

RG 1.23, Meteorological Monitoring Programs for Nuclear Power Plants, Rev. 1, March 2007.

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 1.194, 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.

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

NUREG-0800, Standard Review Plan, Section 6.4, Control Room Habitability System, 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-0800, "Standard Review Plan," Section 15.7.4, Radiological Consequences of Fuel Handling Accidents, (Rev, 1, July 1981) provides guidance to the NRC staff for the review and evaluation of system design features and plant procedures provided for the mitigation of the radiological consequences of postulated FHAs.

NUREG/CR-5950, Iodine Evolution and pH control, December 1992.

NUREG-1465, Accident Source Terms for Light-Water Nuclear Power Plants, February 1995.

NUREG-0737, Clarification of TMI Action Plan Requirements, November 1980, (including Item II.E.4.2, Containment Isolation Dependability.)

NUREG-1433, Rev. 3.0, Standard Technical Specifications General Electric Plants, BWR/4, June 2004 Technical Specification (TS) Task Force Traveler TSTF-51, Revision 2 Revise Containment Requirements During Handling Irradiated Fuel and Core Alterations (ADAMS Accession No. ML040400343) approved by the NRC on October 13, 1999, which provides for the relaxation of some TS requirements during refueling after a sufficient decay period has occurred.

Review Guideline Guidance on the Assessment of a BWR SLC System for pH Control, February 12, 2004 (ADAMS Accession No. ML040640364).

NRC Generic Letter 97-04, Assurance of Sufficient Net Positive Suction Head for Emergency Core Cooling and Containment Heat Removal Pumps, October 7, 1997.

The NRC requested that addressees submit information necessary to confirm the adequacy of the net positive suction head (NPSH) available for emergency core cooling (including core spray and decay heat removal) and containment heat removal pumps.

The NRC staffs review includes the seismic qualification of components such as main steam lines, electric equipment, and heating ventilation and air conditioning (HVAC) duct work that may be affected by the implementation of the AST at PBAPS. The NRC staffs review also considered the requirements of 10 CFR 50.55a as they relate to structures and components being designed, fabricated, erected, constructed, tested, and inspected to quality standards commensurate with the importance of the safety function to be performed.

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

3.0 TECHNICAL EVALUATION

The licensee proposed a full implementation of the AST, in accordance with the guidance in RG 1.183, and Section 15.0.1 of the SRP. The scope of the licensee AST analyses included the four boiling water reactor (BWR) DBAs identified in its UFSAR Chapter 14, Plant Safety Analysis, and described in RG 1.183 as BWR DBAs that could potentially result in significant CR and offsite doses. These DBAs include:

Loss-of-coolant accident (LOCA);

Fuel handing accident (FHA);

Control rod drop accident (CRDA); and

Main steam line break accident (MSLBA).

The licensee has determined that the current TID-14844, Atomic Energy Commission (AEC),

1962, Calculation of Distance Factors for Power and Test Reactors Sites, AST will remain the licensing basis for EQ. As stated in the regulatory analysis section of this NRC SE, continued use of the TID-14844 source term for EQ is acceptable to the NRC staff.

The licensee completed a review of the impact of proposed AST requirements applicability to Peach Bottom NUREG-0737 commitments and modifications. The licensee determined that the AST will not affect its current NUREG-0737 licensing basis. Therefore, the staff finds that the licensee fully addressed the issue of maintaining consistency with the NUREG-0737 evaluations while incorporating the AST into the plant licensing basis for dose consequence analyses.

The DBA 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 to a Peach Bottom Units 2 and 3 CR operator were evaluated for the duration of the accident. The dose consequence analyses were performed by the licensee using the RADTRAD: Simplified Model for RADionuclide Transport and Removal And Dose Estimation, Version 3.03, 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 code estimates transport and removal of radionuclides and radiological consequence doses at selected receptors. The staff performs independent confirmatory dose evaluations, as needed, using the RADTRAD computer code. 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 2 of this SE.

3.1 Proposed TS Changes

The licensees application of the AST methodology is being used to implement TS and licensing basis changes for Peach Bottom that would allow some additional operational flexibility. The proposed changes are as follows:

3.1.1 TS Section 1.1, "Definitions" The proposed change revises the definition of DOSE EQUIVALENT I-131 in TS Section 1.1 to remove the word "thyroid" and to add a reference to Federal Guidance Report 11, "Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion," Second Printing, 1989.

3.1.2 TS Section 1.1, "Recently Irradiated Fuel" The proposed change, as modified by the licensees supplement dated August 21, 2008, revises Section 1.1 to add a new definition for RECENTLY IRRADIATED FUEL. RECENTLY IRRADIATED FUEL is fuel that has occupied part of a critical reactor core within the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months with the stipulation that specific secondary containment ground-level hatches must be maintained closed during movement of irradiated fuel in secondary containment.

3.1.3 TS Section 3.1.7, "Standby Liquid Control (SLC) System" The proposed change revises the applicability of TS Section 3.1.7 to add the requirement for the Limiting Condition for Operation (LCO) to be met in Mode 3. This change implements AST assumptions regarding the use of the SLC System to buffer the suppression pool following a LOCA involving significant fission product release. The required actions for Condition D are being revised to add an additional requirement to be in Mode 4 with a completion time of 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

3.1.4 TS Section 3.3.6.1, "Primary Containment Isolation Instrumentation" TS Section 3.3.6.1, Table 3.3.6.1-1 lists the applicability requirements for Primary Containment Isolation Instrumentation. The proposed change revises the applicability of the SLC System Initiation Function of the Reactor Water Cleanup System Isolation instrumentation to add the requirement for this function to be operable in Mode 3. The revised applicability for this function is consistent with the revised applicability for the SLC System.

3.1.5 TS Section 3.3.6.2, "Secondary Containment Isolation Instrumentation" The proposed change revises Footnote (b) of TS Table 3.3.6.2-1 by deleting, "CORE ALTERATIONS, and during, which eliminates the requirement for Function 3 (i.e., RB Ventilation Exhaust Radiation - High) and Function 4 (i.e., Refueling Floor Ventilation Exhaust Radiation - High) of the Secondary Containment Isolation Instrumentation to be operable during core alterations. The proposed change also relaxes TS requirements to require these functions to be operable only when handling recently irradiated fuel. With the application of AST, secondary containment is not credited for the FHA after a 24-hour decay period.

3.1.6 TS Section 3.6.1.3, "Primary Containment Isolation Valves (PCIVs)"

A. TS 3.6.1.3 Purge or Vent valve change The proposed change involves revising TS Section 3.6.1.3 to include a new LCO requirement stipulating that the accumulated time a purge or vent flow path is open shall be limited to 90 hours0.00104 days 0.025 hours 1.488095e-4 weeks 3.4245e-5 months per calendar year, with the reactor in MODES 1 and 2, and reactor pressure greater than 100 psig. In the event that flow paths are open greater than 90 hours0.00104 days 0.025 hours 1.488095e-4 weeks 3.4245e-5 months in a calendar year, the penetration(s) must be isolated with 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or the reactor shall be in MODE 3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , and in MODE 4 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . This will limit the total time that a flow path exists through certain containment penetrations. Consequently, the impact on plant risks resulting from ECCS NPSH during a LOCA while purging contribute little to the likelihood of ECCS equipment malfunctions.

B. Surveillance Requirement (SR) 3.6.1.3.14 - Increase allowable Main Steam Isolation Valve (MSIV) leakage rate The proposed change revises SR 3.6.1.3.14 to increase the allowable limit for the combined leakage rate for all MSIV leakage paths. Currently, the allowable limit is less than or equal to 11.5 standard cubic feet per hour (scfh) for each MSIV when tested at greater than or equal to 25 psig. This limit will be increased to less than or equal to 204 scfh for all four main steam lines and less than or equal to 116 scfh for any one main steam line, when tested at greater than or equal to 25 psig. The proposed revised SR 3.6.1.3.14 reads:

Verify combined MSIV leakage rate for all four main steam lines is < 204 scfh, and <116 scfh for any one steam line, when tested at 25 psig.

3.1.7 TS Section 3.6.4.1, "Secondary Containment" A. Change in TS-3.6.4.1 applicability The proposed change deletes "During CORE ALTERATIONS" from the applicability statement for TS LCO 3.6.4.1 and relaxes TS requirements to require LCO 3.6.4.1 to be applicable only when handling recently irradiated fuel.

B. Change in TS3.6.4.1 conditions The proposed change revises Condition C, and associated required actions and completion times, to reflect the revision of the applicability requirements for LCO 3.6.4.1. With the application of AST, secondary containment is not credited for the FHA after a 24-hour decay period.

C. Change in TS SR 3.6.4.1.3 requirements The proposed change to TS 3.6.4.1 is to SR 3.6.4.1.3, increasing the secondary containment drawdown time from less than or equal to 120 seconds to less than or equal to 180 seconds.

3.1.8 TS Section 3.6.4.2, "Secondary Containment Isolation Valves (SCIVs)"

The proposed change deletes "During CORE ALTERATIONS" from the applicability statement for TS LCO 3.6.4.2 and relaxes TS requirements to require LCO 3.6.4.2 to be applicable only when handling recently irradiated fuel. The proposed change revises Condition D, and associated required actions and completion times, to reflect the revision of the applicability requirements for LCO 3.6.4.2. With the application of AST, closure of secondary containment isolation valves is not credited for the FHA after a 24-hour decay period.

3.1.9 TS Section 3.6.4.3, "Standby Gas Treatment (SGT) System" The proposed change deletes "During CORE ALTERATIONS" from the applicability statement for TS LCO 3.6.4.3 and relaxes TS requirements to require LCO 3.6.4.3 to be applicable only when handling recently irradiated fuel. The proposed change revises Condition C and Condition E, and associated required actions and completion times, to reflect the revision of the applicability requirements for LCO 3.6.4.3.

3.1.10 TS Section 5.5.12, "Primary Containment Leakage Rate Testing Program" The proposed change increases the maximum allowable primary containment leakage rate, La, at Pa, from 0.5 percent to 0.7 percent of primary containment air weight per day.

3.1.11 UFSAR Section 5.2.4.3.2, "Minimum Containment Pressure Available" The initial amendment request submittal dated July 13, 2007, proposed to increase the Containment Overpressure License (COPL) described in UFSAR Figure 5.2.16, while maintaining adequate margin between the revised Minimum Containment Pressure Available (MCPA) and the proposed COPL. However, the licensees supplement dated April 17, 2008, revised the submittal and proposes to eliminate the COPL curve from the UFSAR. Exelon states that they will continue to demonstrate that the containment pressure available during the postulated events meets or exceeds that required by the residual heat removal (RHR) and core spray (CS) systems.

3.2 AST Design Basis Accident Analyses 3.2.1 DBA Radiation Source Terms 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.

In accordance with RG 1.183, the licensee generated the core and worst case fuel assembly radionuclide inventories using the ORIGEN code version 2.1. The licensee assumed a period of irradiation that was sufficient to allow the activity of dose-significant radionuclides to reach equilibrium or to reach maximum values. Consideration was also given to fuel enrichment in order to ensure bounding conditions. The dose consequence maximum hypothetical accident analysis is based on a maximum power level of 3,528 MWt, which includes 0.4 percent margin for ECCS evaluation uncertainty. The source terms are based on a 2-year fuel cycle with a nominal 711 effective full power days (EFPD) per cycle. The Peach Bottom LOCA analysis uses a core average inventory of all fuel assemblies. Peach Bottom used peaking factors of 1.7 for DBA events that do not involve the entire core, with fission product inventories for damaged fuel rods determined by dividing the total core inventory by the number of fuel rods in the core.

As stated, in part, in RG 1.183, Section 3.3, 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 has stated that its dose consequence analysis conforms to this specified guidance.

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 and DCFs from FGR-11 and FGR-12 is in accordance with RG 1.183, Section 3.1, guidance and is therefore acceptable to the NRC staff.

3.2.2 Loss of Coolant Accident (LOCA) 3.2.2.1 AST LOCA Description The radiological consequence design basis LOCA (AST LOCA) is a surrogate for a maximum hypothetical accident (MHA), required by regulation and postulated from considerations of possible accidental events that would result in potential hazards (dose consequences) not exceeded by those from any accident considered credible. The AST based on NUREG-1465, like its predecessor the TID-14844 source term, is used to evaluate the ESFs used to protect public health and safety in the unlikely event of a nuclear accident that results in substantial meltdown of the core with subsequent release of appreciable quantities of fission products. The light water nuclear MHA source term described by TID-14844 or NUREG-1465 is derived from a deterministic evaluation based on the assumption of a major rupture of the primary reactor coolant system (RCS) piping referred to as a LOCA. The AST LOCA accident scenario assumes the deterministic failure of the ECCS to provide adequate core cooling which results in a significant amount of core damage as specified in RG 1.183. Unlike the DBA LOCA used to evaluate the ECCS requirements of 10 CFR 50.46, 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 ECCS or design basis transient analyses.

The MHA considered in an AST evaluation is the complete and instantaneous circumferential severance of un-isolable primary RCS piping, which would result in the maximum dose consequence among the full range of LOCAs considered. 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.

In response to an RAI regarding the licensees AST LOCA licensing basis description, the licensee stated, in its supplement dated July 29, 2008, The AST LOCA analysis performed for the [Peach Bottom] PBAPS, Units 2 and 3, considered a spectrum of line breaks and applicable source term guidance and has determined that the AST LOCA is representative of the maximum hypothetical accident (MHA). For the purpose of consideration of the MHA the LOCA analysis assumes a line break in a main steam line upstream of one inboard main steam isolation valve (MSIV) within the primary containment (drywell). The licensee further explained that this postulated accident sequence for AST purposes maximizes the dose since one steam line that would otherwise be available is postulated to be broken and cannot be credited for iodine deposition. The faulted steam line was chosen such that the line with the most deposition credit possible is not available for deposition, The NRC staff finds that this AST LOCA description for design basis purposes, that was postulated from considerations of possible accidental events that would result in potential hazards (dose consequences) not exceeded by those from any accident considered credible, is a conservative representation for implementation of an AST at Peach Bottom.

When evaluating an AST LOCA for a BWR, it is assumed that the initial fission product release to the containment will last for 2 minutes and will consist of the radioactive materials dissolved or suspended in the RCS liquid. After 2 minutes, 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 1.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months . The licensee used the AST release fractions, timing characteristics, and radionuclide grouping as specified in RG 1.183 Table 1.

In its evaluation of the proposed licensing basis AST LOCA, the licensee considered dose contributions from the following potential radioactive material release pathways:

$ Primary containment bypass leakage directly to the environment;

$ Engineered safety feature (ESF) system leakage; and

$ Main Steam Isolation Valve (MSIV) Leakage to the environment via the turbine stop valves (TSVs).

The licensee considered the following DBA LOCA dose contributors to the CR habitability envelope (CRHE) analysis:

$ Post-LOCA airborne activity inside the CR;

$ Post-LOCA airborne cloud external to CR;

$ Post-LOCA containment shine to CR; and

$ Post-LOCA Main Control Room Emergency Ventilation (MCREV) filter shine.

3.2.2.2 AST LOCA Source Term The licensee followed all aspects of the guidance outlined in RG 1.183, Regulatory Position 3, regarding the core inventory, release fractions and timing for the evaluation of its dose consequence LOCA. The licensees description of the AST LOCA analysis in Section 4.3.2 of the license amendment request states the following:

The radioactivity released from the fuel is assumed to mix instantaneously and homogeneously throughout the free air volume of the primary containment (drywell). The radioactivity release into the drywell is assumed to terminate at the end of the early in-vessel phase, which occurs at 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months after the onset of a LOCA. The reduction in drywell leakage activity by dilution in the reactor building (RB) and removal by the standby gas treatment (SGT) system filtration is not credited. The analysis dilutes the radioactivity released from the core into the drywell air volume during the first 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months of the LOCA, and then into the combined drywell plus suppression chamber air volume after 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , at which time the containment volume becomes well mixed following the restoration of core cooling. The thermal-hydraulic conditions in the primary containment are expected to be quite active at this time due to a very high flow established between the drywell and wetwell as a result of steaming and condensing phenomenon.

Peach Bottom proposes to credit control of the pH in the suppression pool following a LOCA by means of injecting sodium pentaborate into the reactor core with the SLC system. To demonstrate that the SLC system is capable of performing the safety function assumed in the AST LOCA dose analysis, the licensee has completed an analysis that demonstrates that the suppression pool pH will remain greater than 7 for the duration of the accident. The NRC staff review of the Peach Bottom SLC system applicability to pH control for AST purposes is provided in Section 3.4.3 of this SE.

Since the post-LOCA minimum suppression chamber water pH has been demonstrated by the licensee to be greater than 7.0 for the 30-day duration of the accident, the chemical form of radioiodine released into the containment is assumed to be 95 percent cesium iodide (Csl), 4.85 percent elemental iodine, and 0.15 percent organic iodide in accordance with Appendix A of RG 1.183 Regulatory Position 2, and is acceptable to the NRC staff.

3.2.2.3 AST LOCA Radioactive Material Transport in the Primary Containment The Peach Bottom primary containment is an essentially leak tight enclosure for the reactor vessel, the reactor coolant recirculation system, and other branch connections of the reactor coolant system. The primary containment includes a drywell and a pressure suppression chamber connected by vents, isolation valves, vacuum breakers, containment cooling systems, and other service equipment. The drywell is a steel pressure vessel in the shape of an upside down light bulb, and the pressure suppression chamber is a torus-shaped steel pressure vessel, located below and encircling the drywell, which is partially filled with water for pressure suppression purposes in the unlikely event of a LOCA. The primary containment is a seismic Class I structure and is designed to withstand the jet forces resulting from a rupture of a reactor coolant system pipe.

In accordance with Appendix A of RG 1.183 Regulatory Position 3.1, the licensee assumed that the activity released from the fuel is mixed instantaneously and homogeneously throughout the free air volume of the drywell. The licensee used the core release fractions and timing as specified in RG 1.183 with the termination of the release into containment drywell set at the end of the early in-vessel phase at 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months following the initial release. The licensees analysis dilutes the radioactivity released from the core into the drywell air volume during the first 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months of the LOCA, and then dilutes the source term further into the combined drywell plus suppression chamber (wet-well) air volume after 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , at which time the containment volume becomes well mixed following the restoration of core cooling.

The NRC staff finds that the mass and energy (steaming and steam condensation) created by reflooding (arresting reactor pressure vessel failure) and core quenching will provide sufficient energy to mix the drywell and wet-well air when vacuum breaker cycling occurs during this pressure transient. The licensee assumed that the radioactivity release is diluted into the larger volume of the wet-well plus drywell air spaces after 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months . Before this time, the radioactivity is only assumed to be released into the drywell net free volume.

3.2.2.4 AST LOCA Radioactive Material Deposition in the Primary Containment The licensee credited gravitational deposition of aerosols from the containment atmosphere by using the RADTRAD "Powers Model" with a 10th percentile uncertainty distribution resulting in the lowest removal rate of aerosols from the containment.

Although containment sprays are not credited, the licensee credited removal of the elemental iodine by wall deposition on wetted surfaces inside containment in a similar way as containment spray iodine removal would be credited, providing a removal rate of 3.36 hr -1 for the first 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months of the accident and a removal rate of 1.86 hr -1 after 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months until a decontamination factor (DF) of elemental iodine of 200 is reached at 3.85 hours9.837963e-4 days 0.0236 hours 1.405423e-4 weeks 3.23425e-5 months . In response to an RAI on this assumption, the licensee stated, in its supplement dated May 23, 2008, that the moisture released during and following a LOCA is continuously condensed on the relatively colder surfaces including but not limited to the drywell wall, floor, and ceiling concrete surfaces, various piping surfaces, duct and equipment surfaces, and torus water surface. Additional wetted surface area due to the drywell spray operation is not accounted for in the wetted surface estimate in the licensee calculations. The estimated wetted surface area is conservatively reduced by 25 percent and compared with the corresponding wetted surface areas of the licensees other BWR Mark I plants, Dresden and Quad Cities, to establish a conservative basis for the selection of the DF value for Peach Bottom. Iodine removal by suppression pool scrubbing is not credited since the bulk of core activity is released to containment well after the initial mass and energy release.

The NRC staff agrees with the licensees assessment of elemental iodine deposition on wetted surfaces inside containment. The wetted drywell wall deposition is equivalent to credit for containment sprays in that the accident would cause all exposed drywell surfaces to be wetted in the course of the accident. Conservatively, Peach Bottom did not credit reduction in fission products transferred to the suppression pool air space from the drywell by suppression pool scrubbing. Instead, Peach Bottom assumed a well mixed suppression pool air space and drywell after 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

Peach Bottom credited the reduction in airborne aerosol radioactivity in the containment by natural deposition consistent with the guidelines provided in RG 1.183, Appendix A, Section 3.2.

Acceptable models for removal of iodine and aerosols are described in NUREG-0800, SRP Chapter 6.5.2, Containment Spray as a Fission Product Cleanup System," Rev. 2, December 1988 and in NUREG/CR-6189, A Simplified Model of Aerosol Removal by Natural Processes in Reactor Containments. The NUREG/CR-6189 model is incorporated into RADTRAD. This simplified model in NUREG/CR-6189 was derived by correlation of the results of Monte Carlo uncertainty analyses of detailed models of aerosol behavior in the containment under accident conditions. Peach Bottom conservatively used the 10th percentile Power's Aerosol Decontamination Model in RADTRAD to account for the reduction in airborne radioactivity in the containment by natural deposition. Because Peach Bottoms approach is consistent with RG 1.183, the NRC staff finds this approach acceptable.

3.2.2.5 AST LOCA Containment Leakage Pathway Appendix A of RG 1.183, Regulatory Position 3.7 states that, The primary containment (i.e.,

drywell for Mark I and II containment designs) should be assumed to leak at the peak pressure technical specification leak rate [La] for the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . For BWRs, leakage may be reduced after the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , if supported by plant configuration and analyses, to a value not less than 50 percent of the technical specification leak rate. The Peach Bottom analysis credits this reduction after 38 hours4.398148e-4 days 0.0106 hours 6.283069e-5 weeks 1.4459e-5 months as determined in its plant design analysis described in Attachment 1, reference 7.36 of the licensee submittal dated July 13, 2007.

3.2.2.6 AST LOCA Containment Leakage to Secondary Containment Pathway The Peach Bottom RB (secondary containment), in conjunction with other engineered safeguards, is designed to limit the ground level release of airborne radioactive materials and to provide means for controlled elevated release of the secondary containment atmosphere by the safety-related standby gas treatment system (SGTS). The Peach Bottom SGTS limits the ground level release from the RB, and will release primary and secondary containment air at an elevated release point via the Peach Bottom main stack following initiation of the SGTS and establishment of negative building pressure or drawdown. For the licensees AST LOCA analysis, the containment leakage during the RB drawdown is assumed to be directly released to the atmosphere via the RB stack. The licensee has assumed a secondary containment (RB) drawdown time of 3 minutes which was increased from the 2 minutes assumed in its current licensing basis (CLB).

The licensees analysis did not credit filtration through the SGTS charcoal and high efficiency particulate air (HEPA) filters. Leakage from the primary containment is postulated to be released directly to the environment without mixing in the RB free air volume initially as a ground level release, and as an elevated release via the Peach Bottom main stack following initiation of the SGTS and establishment of negative RB pressure or drawdown.

Appendix A of RG 1.183, Regulatory Position 4.2 states that, Leakage from the primary containment is assumed to be released directly to the environment as a ground-level release during any period in which the secondary containment does not have a negative pressure as defined in technical specifications. Since the licensees AST LOCA analysis assumptions for dual containments are in conformance with regulatory guidance, the NRC staff finds them acceptable.

In the licensees RAI response dated February 28, 2008, to an NRC staff RAI regarding the seismic qualification of the SGTS, the licensee confirmed that the systems associated ductwork and components credited in the proposed AST are capable of withstanding and performing their intended safety functions following a safe-shutdown earthquake (SSE). The licensee stated that during 1988, Peach Bottom upgraded the ductwork above the main control room under Modification 1729, "Control Room Upgrade." Ductwork, its supports and support anchorage were seismically designed to withstand the SSE loads. The licensee, in its RAI response, also provided calculations demonstrating the seismic qualification of the ductwork and its supports associated with the SGTS at PBAPS. Review of these calculations shows that the ductwork and its supports are capable of withstanding a design basis earthquake and performing their intended design function. Therefore, the NRC staff finds that the ductwork and components of the SGTS will continue to operate safely upon implementation of the proposed AST.

3.2.2.7 Assumptions on Engineered Safety Feature (ESF) System Leakage Peach Bottom ESF system leakage develops when ECCS systems located in the RB circulate suppression pool water outside primary containment and leaks develop through packing glands, pump shaft seals and flanged connections. To evaluate the radiological consequences of ESF leakage, Peach Bottom used the deterministic approach as described 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 suppression pool water. Except for iodine, all of the radioactive materials in the suppression pool are assumed to be retained in the liquid 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 suppression pool concurrently. Since the post-LOCA temperature of torus water recirculated through the ESF systems is less than 212 °F, 10 percent of the iodine activity in the leaked ESF liquid is assumed to become airborne in accordance with Appendix A of RG 1.183, Regulatory Position 5.5. Reduction in ESF leakage activity by dilution in the RB and radioiodine removal by SGT filtration are not credited.

The licensee assumed a value of 10.0 gallons per minute (gpm) for its ESF leakage, which is 2 times the expected leakage of 5.0 gpm for the Peach Bottom AST LOCA analysis, as specified in Appendix A of RG 1.183, Regulatory Position 5.2. The licensee assumed that ESF leakage will start at the beginning of the LOCA and continue for the 30 day duration of the accident evaluation. The radiological consequences from the postulated leakage are analyzed and combined with the consequences from other fission product release paths to determine the total AST LOCA calculated radiological consequences. The NRC staff finds the assumptions and methodology used by the licensee is in accordance with regulatory guidance and is therefore acceptable.

3.2.2.8 Assumptions on Main Steam Isolation Valve Leakage The main steam lines (MSL) in BWR plants, including Peach Bottom, contain main steam isolation valves (MSIVs). The MSIVs are designed to limit the release of radioactive material and loss of reactor cooling water in case of a major steam leak outside the primary containment and to limit the release of radioactive material from the reactor coolant system and the primary containment in the event of a design basis LOCA. Two MSIVs are welded in a horizontal run of each of the four main steam lines at Peach Bottom, with one valve as close as possible to the primary containment barrier inside, and the other just outside the primary containment barrier.

Each valve is a "Y"-shaped, 26-inch globe valve and is installed in a matching 26-inch MSL pipe. Although the MSIVs are designed to provide a leak-tight barrier, it is recognized that some leakage through the valves will occur. Since the MSIVs are functionally part of the primary containment boundary, leakage through these valves provides a path for fission products that bypass the secondary containment and enter the environment as a ground-level release.

Appendix A of RG 1.183, Section 6, provides guidance for the evaluation of the radiological consequences from MSIV leakage which should be combined with other fission product pathways to determine the total calculated radiological consequences from an AST LOCA.

Following the guidance in RG 1.183, Peach Bottom assumed that the activity available for release via MSIV leakage is that activity determined to be in the drywell for evaluating containment leakage. Peach Bottom did not credit activity reduction by the steam separators and steam dryers or by iodine partitioning in the reactor vessel. The wet-well free air volume was included with the drywell free air volume after 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months as previously discussed in Section 3.2.2.3 of this SE. In the Peach Bottom AST LOCA analysis, MSIV leakage is evaluated at the maximum leak rate above which the proposed TS would require the MSIVs to be declared inoperable (less than 205 standard cubic feet per hour (scfh) from any one valve or less than 360 scfh total from four valves) calculated for the drywell peak pressure of 49.1 psig. These leakage values correspond to the licensee proposed increased TS leakage values of less than or equal to 204 scfh for all four MSLs and less than or equal to 116 scfh for any one MSL, when tested at greater than or equal to 25 psig (SR 3.6.1.3.14).

The Peach Bottom-described MHA analysis for the AST LOCA assumes a steam line break inside containment on the shortest steam line with a faulted inside containment (inboard) MSIV.

The licensee did not credit deposition of aerosol and removal of elemental iodine in this steam line between the reactor pressure vessel (RPV) nozzle and the outboard MSIV. In addition, the licensee assumed the maximum TS leakage flow of 205 scfh at 49.1 psig in this broken line and assumed that the remaining leakage flow of 155 scfh (360 scfh minus 205 scfh) would come from the second shortest intact MSL providing a conservative maximum dose consequence analysis for the licensee described MHA AST LOCA. The leakage is assumed to continue for the duration of the accident and is reduced to 50 percent after 38 hours4.398148e-4 days 0.0106 hours 6.283069e-5 weeks 1.4459e-5 months in accordance with Peach Bottom specific design analysis as described in Attachment 1, reference 7.36 of the licensee submittal dated July 13, 2007.

Appendix A of RG 1.183, Section 6.5 allows radioactive material dose reduction from MSIV leakage due to holdup and deposition in MSL piping downstream of the MSIVs and in the main condenser, including the treatment of air ejector effluent by off-gas systems, if the components and piping systems used in the release path are capable of performing their safety function during and following an SSE. Peach Bottom has conservatively credited only its seismic Class I piping up to the TSVs in its MSIV leakage calculation for the AST LOCA analysis. The deposition of aerosols and removal of elemental iodine are conservatively credited in only the horizontal pipe between the outboard MSIV and TSV for 0-96 hrs for the broken MSL and the horizontal pipe segments between the RPV nozzle and TSV for 0-96 hrs in the second shortest intact MSL.

In the licensees RAI response dated February 28, 2008, the licensee provided a summary of the analysis performed on the main-steam line piping system from the outboard MSIV to the TSVs to demonstrate seismic qualification. The Code of Record for the main-steam line piping is the American National Standards Institute (ANSI) B31.1 Code for Power Piping, 1973 Edition, including Summer 1973 Addenda. The licensee indicated that, for the portion of piping under review, a finite element piping analysis was performed to demonstrate seismic qualification of the main-steam piping system from the outboard MSIVs to the TSVs. The analyses provided by the licensee confirm that the main-steam piping stresses and associated pipe support stresses are within code-allowable values and are seismically qualified to withstand an SSE. Therefore, the NRC staff finds that the portion of the main-steam lines downstream of the MSIVs will continue to operate safely upon implementation of the proposed AST.

The Peach Bottom model for aerosol settling is based on the methodology used by the NRC staff in its review of the implementation of an AST at the Perry Nuclear Power Plant and Clinton Nuclear Station. The aerosol settling model is described in the report, AEB-98-03, Assessment of Radiological Consequences for the Perry Pilot Plant Application Using the Revised (NUREG-1465) Source Term, (ADAMS Accession No. ML011230531) which was prepared by the NRC Office of Nuclear Regulatory Research. AEB-98-03 gives a distribution of aerosol settling velocities that are estimated to apply in the MSL piping. The NRC-approved model used in the Perry and Clinton assessments assumed aerosol settling may occur in the MSL upstream of the outboard MSIV at the median (50th percentile) settling velocity given by the Monte Carlo analysis described in the AEB-98-03 report. The NRC staff is concerned however, with how much deposition (i.e., what settling velocity value) is appropriate for other licensees that use this model in their MSL MSIV leakage analysis.

In response to these NRC staff concerns, Peach Bottom used a conservative rate constant (s) for its different piping segments in the MSIV leakage release paths calculation (Calculation No.

PM-1077, Rev 0, reference 7.37 of Attachment 1 to the license amendment request (LAR) submittal dated July 13, 2007). This design analysis calculation used a 40th percentile aerosol settling velocity rather than the medium 50th percentile aerosol settling velocity recommended in AEB 98-03. The aerosol removal efficiencies due to gravitational depositions on the horizontal pipe surfaces were calculated using the mass balance equation for well-mixed volumes. The natural removal efficiency for elemental iodine in each steam line volume was assumed to be 50 percent which also is recommended in AEB 98-03.

In response to several RAIs, the licensee in its supplements dated May 23, 2008, and July 29, 2008, further explained the details of their conservative assumptions and described a parametric study done to confirm that the assumptions used were conservative in regards to the NRC-approved AEB 98-03 MSIV leakage deposition model. In the parametric study the AEB 98-03 50th percentile settling velocity and resulting aerosol removal efficiencies was used in the first node of each line similar to the acceptable method used by Perry Nuclear Power Plant and Clinton Nuclear Station. The settling velocities in the main steam lines beyond the outboard MSIVs in both Peach Bottom release paths (failed line and intact line) were varied from the 10th percentile through the 50th percentile. The licensee then compared the CR doses for the different settling velocities in the MSL piping beyond the outboard MSIV in the parametric study with its proposed design basis MSIV leakage pathway CR dose based on its design analysis calculation. The results of the licensee parametric study are contained in the revised design analysis calculation (No. PM-1077, Rev 1) that was submitted to the NRC in a supplement dated August 7, 2008, which further demonstrated that the methods used in the licensee MSL MSIV design analysis are conservative in regards to the use of the AEB 98-03 model for Peach Bottom.

The licensee concluded in its parametric study that:

The aerosol gravitational deposition and elemental iodine removal in the post-LOCA MSIV leakage release paths are conservatively modeled for Peach Bottom and maintain the conservative characteristics of the method described in AEB 98-03.

The MSIV leakage model is structured to provide a conservative bound for a large range of settling velocities including very fine aerosol particles represented by a lower settling velocity, lesser deposition, and higher dose as well as coarser aerosol particles represented by a higher settling velocity, larger deposition, and lower dose.

The conservative model used in the licensee design analysis provides an appropriate and prudent safety margin against unpredicted events in the course of an accident and compensates for large uncertainties in facility parameters, accident progression, radioactive material transport, and atmospheric dispersion.

The NRC staff acknowledges that aerosol settling is expected to occur in the MSL piping, however, because of NRC staff concerns regarding the AEB-98-03 report and the lack of additional confirmatory information, it is not clear how much deposition (i.e., which settling velocity value) is appropriate for BWR MSL MSIV leakage analysis. The licensee has used a model based on the methodology of AEB-98-03, but included some additional conservatism to address NRC staff questions on the applicability of the AEB-98-03 methodology to Peach Bottom. The licensee assumed in its analysis that the TS allowed leakage at accident peak pressure is only from the two shortest MSLs. The licensee assumed a break of the shortest MSL with inboard MSIV failure leaking at the maximum TS leakage. Additionally, the licensee assumed that the second shortest intact MSL leaks at the maximum remaining TS allowed flow leakage at peak accident pressure. The other two MSLs were assumed to have no leakage for the proposed AST LOCA analysis. In addition, the licensee used the AEB-98-03 model with a 40th percentile settling velocity which is more conservative than use of the median settling velocity noted as reasonable in AEB-98-03. The licensee also showed in its parametric study that use of the 40th percentile settling velocity used for the credited MSL piping in its design analysis was bounding for the range of settling velocities described in AEB-98-03 for the piping between the outboard MSIVs to the TSVs. Given this information, the NRC staff finds the Peach Bottom MSL settling model for aerosol deposition used in its AST LOCA analysis to be reasonably conservative.

The licensee assumed deposition of elemental iodine in the MSL piping was modeled in accordance with guidance provided in Appendix A of RG 1.183, Regulatory Position 6.5. The Peach Bottom elemental iodine deposition model used is described in a letter report dated March 26, 1991, by J. E. Cline, MSIV Leakage Iodine Transport Analysis, which the NRC staff has found acceptable for estimating elemental iodine deposition. The Cline Report provides elemental iodine deposition velocities, re-suspension rates and fixation rates. The deposition velocities were used in a well-mixed model formulation for use with AEB-98-03. Because elemental deposition is not gravity dependent, the licensee, as in AEB-98-03, assumed elemental iodine deposition occurs on the entire surface area of the horizontal and vertical piping of the modeled MSLs. The licensee determined that the elemental iodine removal efficiency for Peach Bottom based on the Cline methodology is 56.4 percent, which is greater than the 50.0 percent used in the analysis in AEB 98-03. This elemental iodine removal efficiency is negligibly reduced due to re-suspension of the elemental iodine, which further gets adsorbed on the piping surface during its migration through the main steam piping. For conservatism the licensee used 50 percent elemental iodine deposition in the MSLs for its AST LOCA analysis. In addition, the licensee calculated the elemental iodine removal efficiency using the maximum post-LOCA gas temperature which is a conservative assumption based on an increase in elemental iodine adsorption rate on cooling piping surfaces during the course of the postulated event sequence. Based on the discussion above, the NRC staff also finds the Peach Bottom MSL elemental iodine removal efficiency used in its AST LOCA analysis to be reasonably conservative.

3.2.2.9 Control Room Habitability AST LOCA The Peach Bottom main CR heating, ventilating and air conditioning (HVAC) system consists of the safety-related CR fresh air supply (CRFAS) system and the safety-related main CR emergency ventilation (MCREV) filter system which is common to both Peach Bottom Units 2 and 3. The Peach Bottom CR HVAC system is designed to provide a suitable environment for continuous personnel occupancy as well as operability of CR equipment and instrumentation under normal and accident conditions. The CR HVAC system provides isolation, filtration, ventilation, cooling, heating, and humidification. The primary safety function of the CR HVAC system is to provide isolation, filtration, and ventilation during normal plant operation as well as during accident conditions.

Upon receiving a high radiation or low flow signal, as sensed by the associated instruments, the CRFAS system and associated components are automatically isolated and the MCREV filter system initiates to provide the control room with filtered, fresh outside air. The CR envelope (CRE) is maintained under habitable conditions following an accident. The envelope consists of the main CR, the Shift Supervisors office, Shift Managers office, auxiliary offices, lunchroom, and toilet room (all on elevation 165-0"). CRE construction joints and penetrations for cable, pipe, HVAC equipment, dampers, and doors have been specifically designed for leak-tightness.

The CRE zone served by the HVAC system in the emergency mode is approximately 176,000 cubic feet.

In the licensees RAI response dated February 28, 2008, to an NRC staff RAI regarding the seismic qualification of the MCREV system, the licensee confirmed that the systems associated ductwork and components credited in the proposed AST are capable of withstanding and performing their intended safety functions following an SSE. The licensee stated that during 1988, Peach Bottom upgraded the ductwork above the main control room under Modification 1729, "Control Room Upgrade. Ductwork, its supports and support anchorage were seismically designed to withstand the SSE loads. The licensee, in its RAI response, also provided calculations demonstrating the seismic qualification of the ductwork and its supports associated with the MCREV system at PBAPS. Review of these calculations shows that the ductwork and its supports are capable of withstanding a design basis earthquake and performing their intended design function. Therefore, the NRC staff finds that the ductwork and components of the MCREV system will continue to operate safely upon implementation of the proposed AST.

3.2.2.10 CR Ventilation Assumptions for AST LOCA During emergency modes of operation, the CR HVAC supplies 3,000 standard cubic feet per minute (scfm) of filtered, outdoor air to maintain the control room at 0.1-inch water column positive pressure with respect to adjacent areas. Infiltration following isolation is assumed to be 500 scfm of unfiltered inleakage, which includes CR ingress and egress.

The licensees assumption of 500 scfm unfiltered CR inleakage was validated by inleakage testing at Peach Bottom conducted in October 2004, as described in a letter dated January 21, 2005 (ADAMS Accession No. ML040890545). The testing was conducted using the tracer gas method described in American Society for Testing and Materials (ASTM) E741-00, Standard Test Method for Determining Air Change in a Single Zone by Means of a Tracer Gas Dilution, and was in response to NRC Generic Letter 2003-01, "Control Room Habitability. The test results were 369 scfm unfiltered inleakage for MCREV Train A and 21 scfm for MCREV Train B based on a mean of six measurements.

The post-LOCA MCREV system initiation time is conservatively delayed for 30 minutes in the licensee AST LOCA analysis. The use of a large CR unfiltered intake during this 30 minutes delay produces a conservative CR dose during the 3-minute RB drawdown time.

In response to an RAI on the licensees unfiltered inleakage assumptions, the licensee stated, in its supplement dated May 23, 2008, that CR inleakage infiltration paths were not specifically identified, because the CR was verified to be at a positive pressure with respect to adjacent areas. Therefore the licensee assumed that the CR inleakage is through the ventilation system.

The ventilation system, including the air supply ducts and fans, is located in the radwaste building adjacent to the CR and is the major source for unfiltered inleakage to the CR. The licensee also conducted a parametric study to determine the CR intake flow rate that maximizes the CR dose during the 30 minutes in which the MCREV initiation is delayed; the CR AST LOCA dose analysis includes the infiltration of 18,500 scfm during the MCREV initiation delay. Based on the information above, the NRC staff concludes that the unfiltered inleakage assumption of 18,500 cfm prior to MCREV initiation and the 500 scfm inleakage during MCREV initiation conservatively bounds the CR unfiltered inleakage assumptions used for the Peach Bottom AST LOCA analysis.

3.2.2.11 CR Ventilation Dose Evaluations for AST LOCA The radioactivity from ground level sources including the containment prior to drawdown, ESF and MSIVs leakages are assumed to be released into the atmosphere and transported to the CR air intake, where it may leak into the CR envelope or be filtered by the CR intake filtration system prior to being distributed in the CR envelope.

The CR intake filtration by the Peach Bottom MCREV is credited in the licensees LOCA dose analysis upon automatic initiation of the system following a 30-minute delay on a high radiation signal in the CR intake. The analysis used intake filter efficiencies of 98 percent for aerosols and 89 percent for elemental and organic iodine. The bounding total inlet ventilation flow used in the licensees AST LOCA analysis through the Peach Bottom MCREV filters is 3000 cfm minus 10 percent which equals 2700 cfm.

3.2.2.12 CR Direct Shine Dose Evaluations for AST LOCA The total Peach Bottom CR LOCA dose includes direct shine contributions from the effluent plume outside the CR, the direct shine from the buildup of activity on the CR filters, the direct shine from piping containing contaminated fluids as well as the direct shine from radioactive material in buildings adjacent to the control structure including the containment, RB and the turbine building ( TB.) The post-AST LOCA radioactive plume contains the radioactive sources from the containment, ESF, and MSIV leakages. The Peach Bottom common CR is located at the center of the plant in the TB. The gamma radiation due to the external radioactive plume shine to the CR personnel is attenuated by the 2 feet (ft) - 6 inches (in) minimum concrete ceiling thickness. Since the containment and ESF leakages contribute insignificant direct shine whole body dose to the CR operator, they are not considered important sources for the external cloud dose. Therefore, only the MSIV leakage path is evaluated to determine the external cloud dose to the CR operator. The resulting gamma dose from the external cloud shine is added to the dose contributions from other post-LOCA sources for CR dose.

The licensee assumed that the post-AST LOCA containment and ESF leakage is uniformly distributed inside the RB. The airborne activity confined in the space above the operating floor of the RB contributes direct shine dose to the CR operators. The concrete block wall/steel shielding on the RB operating floor and multiple concrete floor shielding is conservatively not credited in the analysis. The containment shine dose is calculated for the Peach Bottom Unit 2 CR, which is equally applicable to Unit 3 due to the symmetrical shielding geometry. The resulting time dependent CR containment shine dose is calculated based on the dose rates and integrated dose.

The MCREV charcoal filter is located in the radwaste building and the concrete wall between the MCREV charcoal filter and CR panel is 3 ft thick. The charcoal bed is conservatively modeled by the licensee as a rectangular source of 2 ft x 2 ft x 4 ft at approximately 24 ft from the concrete wall. The separation distance is approximately 28 ft - 6 in. The post-AST LOCA CR doses indicate that MSIV leakage contributes the maximum dose to CR operators.

Therefore, the MSIV leakage path is used to assess iodine and aerosol activity buildup on the CR charcoal filter. The resulting CR filter shine dose is adjusted for the containment and ESF leakage dose contributions. The licensee has determined that the accumulation of iodine due to MSIV leakage is insignificant. This is due to the fact that most of the elemental iodine is removed by elemental deposition in the main steam piping as discussed in Section 3.2.2.8 of this SE, before it is released to the environment and it is further reduced by air dilution before it migrates to the CR air intake.

The total aerosol mass deposited on the MCREV HEPA filter due to the MSIV leakage is calculated based on the HEPA filter efficiency and MCREV intake flow rate. The licensee then determined the aerosol mass to curie relationship, and the total mass of aerosols on the HEPA filter. The isotopic aerosol activities deposited on the MCREV filter due to the MSIV leakage was then calculated. The resulting isotopic aerosol activity is insignificant due to the fact that most of the aerosols deposit out in the main steam piping horizontal surfaces before being released to the environment. The total post-LOCA iodine and aerosol activity accumulated on the MCREV charcoal and HEPA filters were then determined.

The licensee used the MicroShield computer code (Reference 7.26 of Attachment 1 to the LAR submittal dated July 13, 2007) to calculate the post-AST LOCA MCREV charcoal filter shine dose to CR operators. The AST LOCA iodine and aerosol activity is uniformly distributed on the MCREV charcoal bed with a dose point located 1 ft from the concrete wall at the center of the source to the CR. This MicroShield direct dose model is conservative with respect to the locations of the MCREV charcoal filter, filter dimensions, and CR operator normal occupancy in the vicinity of the CR panels.

The NRC staff finds that the licensees evaluation of the potential direct shine dose contributions to the CR AST LOCA dose analysis used conservative assumptions and sound engineering principles and is therefore acceptable.

3.2.2.13 AST LOCA Analysis Conclusions The licensee evaluated the radiological consequences resulting from the postulated AST LOCA and concluded that the radiological consequences at the EAB, LPZ, and CR are within the dose requirements provided in 10 CFR 50.67 and accident dose criteria specified in SRP Section 15.0.1. The NRC staff=s review found that the licensee used analysis assumptions and inputs consistent with applicable regulatory guidance identified in Section 2.0 of this SE. The assumptions found acceptable to the staff are presented in Table 5 and the licensee=s calculated dose results are given in Table 2 of this SE. The NRC staff performed independent confirmatory dose evaluations as necessary to ensure a thorough understanding of the licensee=s methods and results. The NRC staff finds that the EAB, LPZ, and CR doses calculated by the licensee for the Peach Bottom AST LOCA analysis meet the applicable accident dose criteria and are therefore acceptable.

3.2.3 AST Fuel Handling Accident (FHA) 3.2.3.1 Analysis Summary The Peach Bottom FHA involves the unlikely event of a fuel assembly drop on top of other fuel assemblies during refueling operations. Exelon, in its CLB for Peach Bottom, concluded that the FHA is bounding for accidents postulated to occur when the primary containment (drywell) is open which would result in the release of radioactive materials directly to the secondary containment. The licensee states in its UFSAR for Peach Bottom, Section 14.6, Accidents that result in radioactive material releases (fuel assembly drops) directly to the secondary containment with the primary containment not intact [drywell and reactor vessel head removed]

is a limiting accident category for Peach Bottom. This DBA is described in the Peach Bottom final safety analysis report (FSAR) as, Refueling accident (fuel assembly drops on core during refueling). In response to an RAI, the licensee provided, in a supplemental letter dated May 23, 2008, a summary table of the proposed Peach Bottom licensing basis parameters that were used as the basis for their revised FHA analysis utilizing the AST methodology.

The licensee evaluated the dose consequences of an FHA following a 24-hour decay time since reactor shutdown (subcritical for greater than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ). This evaluation was based on the AST guidelines outlined in NUREG-1465 and RG 1.183. Computations were made using the NRC dose consequence computer code RADTRAD, Version 3.03. To support the proposed Peach Bottom TS changes, the licensees evaluation demonstrated that the radiological dose consequences at the EAB, LPZ, and in the CR are within regulatory limits. These limits are met without crediting the operability of safety-related systems including secondary containment, secondary containment isolation valves, or the SGTS, after a 24-hour decay period following reactor shutdown. Likewise, the licensees evaluation demonstrated that CR radiological dose consequences are within regulatory limits without credit for the Peach Bottom MCREV system operation.

The licensees CLB limiting postulated FHA event assumed a fuel assembly is dropped on to the reactor core during refueling operations from a height of 34 feet, which is a maximum bounding height allowed by the fuel handling equipment while over the reactor core during refuel mode in accordance with Peach Bottom UFSAR Reference 14.0.6., General Electric Boiling Water Reactor Generic Reload Fuel Application," (GESTAR II), NEDE-24011-P-A-11, November, 1995. This event could also occur over the spent fuel pool (SFP). However, in response to an RAI, the licensee states in a supplemental letter dated May 23, 2008, that the FHA calculation specifically determined that a drop over the reactor core is more limiting than accidents in the SFP.

The licensees proposed LAR would allow the Peach Bottom secondary containment to be open after sufficient fuel decay time (24 hrs) following reactor shutdown. Therefore, in its AST FHA analysis during refueling operations, the licensee assumed the radionuclides released, in the unlikely event of a FHA, would pass through the water in the reactor cavity and enter the RB refueling floor atmosphere instantaneously. The water pool above the core serves as a barrier to the release of radionuclides and is significantly greater than the 23 feet above the fuel to allow for the decontamination factors cited in RG 1.183, Appendix B, Fuel Handling Accident, Regulatory Position 2.0. The remaining radionuclides that become airborne in the RB were released conservatively to the outside environment over a 2-hour period as provided in RG 1.183, Appendix B, Regulatory Position 5.3.

3.2.3.2 AST Fuel Handling Accident Source Term The fission product inventory that constitutes the source term for the postulated Peach Bottom FHA event analysis utilizing the AST methodology is the gap activity in the fuel rods assumed to be damaged as a result of the fuel drop in the reactor core. Volatile constituents of the core fission product inventory migrate from the fuel pellets to the gap between the pellets and the fuel rod cladding during normal power operations. The fission product inventory in the fuel rod gap of the damaged fuel rods is assumed to be instantaneously released to the surrounding water as a result of the accident in conformance with RG 1.183, Appendix B, Regulatory Position 1.3.

The licensee performed a detailed analysis to ensure that the most restrictive case would be considered for the proposed AST FHA dose consequence analysis in accordance with RG 1.183, Appendix B, Regulatory Position 1.1.

For its revised AST FHA event, the licensee based its fraction of core fuel damage on the GESTAR II, NEDE-24011-P-A-14, September 2005 (Reference 7.41 of the licensee submittal dated July 13, 2008) limiting case of damaging 172 fuel pins (based on a "Heavy Mast" design) from GE12 or GE14 10x10 fuel bundle arrays with the analyzed equivalent of 87.33 pins per bundle, and with all of the damaged fuel assumed to have a limiting radial peaking factor (PF) of 1.7. This represents a postulated core damage fraction of approximately 0.44 percent of the Peach Bottom core. This analysis is for an assembly and mast drop from the maximum height allowed by the refueling platform 34 ft above the top of the core and bounds all locations in terms of fuel damage potential. Peach Bottom core loads consist of all GE14 10x10 fuel assemblies, therefore, the results of the GESTAR II limiting case are consistent with current Peach Bottom core loading.

In accordance with RG 1.183, Regulatory Position 3.1, the licensee determined the fission product inventory of each of the damaged fuel rods in its AST FHA analysis by dividing the total core fission product inventory determined for the AST LOCA analysis by the number of fuel rods in the core. To account for differences in power level across the core, the maximum core radial PF of 1.7 was applied in determining the inventory of the worst-case damaged rods.

The licensee calculated the activity in the gap using the non-LOCA gap inventory fractions presented in Table 3 of RG 1.183. These release fractions for non-LOCA events are acceptable for LWR fuel with a peak fuel exposure up to 62,000 MWD/MTU. The fuel peak exposure is acceptable provided that the fuel operating linear heat generation rate does not exceed 6.3 kw/ft peak rod average power for fuel burnup or exposure exceeding 54,000 MWD/MTU in accordance with Table 3 of RG 1.183, Footnote 11. The licensee stated that the limits above will not be exceeded at Peach Bottom and that these fuel limits will be evaluated before each refuel cycle.

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

Following the guidance in RG 1.183, Regulatory Position 3.5 and Appendix B, Section 1.3, the licensee assumed that the chemical form of radioiodine released from the fuel to the SFP consists of 95 percent cesium iodide (CsI), 4.85 percent elemental iodine, and 0.15 percent organic iodide. The CsI released from the fuel is assumed to completely dissociate in the pool water, and because of the low pH of the pool water, the iodine as part of the CsI re-evolves as elemental iodine. This results in a final iodine distribution of 99.85 percent elemental iodine and 0.15 percent organic iodine. The licensee assumed that the release to the pool water and the chemical redistribution of the iodine species occurs instantaneously.

In accordance with RG 1.183, Appendix B, Regulatory Position 2, the licensee credited an overall iodine DF of 200 for a water cover depth of 23 feet. Consistent with RG 1.183, the licensee credited an infinite DF for the remaining particulate forms of the radionuclides contained in the gap activity. In accordance with RG 1.183, the licensee did not credit decontamination from water scrubbing for the noble gas constituents of the gap activity.

The NRC staff finds the licensees conclusion that the consequences of an FHA over the reactor core bounds those for an FHA over the SFP acceptable, and further finds that the use of an effective decontamination factor of 200 for the FHA AST event analysis is acceptable as well.

The NRC staff also finds that the licensees source term used for the proposed Peach Bottom AST FHA event analysis conforms to regulatory guidance, is conservative, and is therefore acceptable.

3.2.3.3 AST Fuel Handling Accident Transport As discussed previously, the licensees proposed LAR would allow the Peach Bottom secondary containment to be open after sufficient fuel decay time (24 hrs) following reactor shutdown to allow more flexibility and potential time savings during the plants refueling operations. Since the accident release through the plant stack via the SGTS is not assumed due to a potentially open secondary containment, the postulated AST FHA release was treated as a ground level (rather than elevated) release. No credit is taken for mixing or dilution inside containment and no credit is taken for filtration by the Peach Bottom SGTS. Release points for the potential openings in the secondary containment (Table 4.4-2 of the licensee submittal dated July 13, 2008) were considered and evaluated by the licensee and by the NRC staff. The licensee then applied its limiting atmospheric dispersion values associated with these potential release points to determine the most limiting dose consequence results for the postulated FHA using the AST methodology. Radioactive material from the reactor core cavity is released to the environment over a 2-hour time period as specified in RG 1.183, Appendix B, Section 5.3. In support of the proposed secondary containment TS changes, the licensee found its licensing basis limiting release point to be the RB Roof Scuttle as described in the proposed LAR submittal dated July 13, 2008.

3.2.3.4 CR Habitability for the AST Fuel Handling Accident The normal operation of the Peach Bottom main CR HVAC system supplies outside air that is drawn through a filter by a fresh air supply fan and is discharged to the inlet of the air conditioning supply fan suction, and is then discharged to duct work leading to the main CR and adjacent offices. This air is conditioned to maintain a controlled temperature environment using heating and cooling coils. The licensee, in its analysis for the AST FHA, did not credit the MCREV system and the associated filters. The normal air supply was assumed with a design basis intake of 20,600 scfm plus an allowance of 1600 scfm for additional in leakage. The dose consequence to CR operators for the AST FHA is calculated over a total duration of 30 days and the dose receptor for this analysis conforms to RG 1.183, Regulatory Position 4.2.6.

3.2.3.5 AST Fuel Handling Accident Analysis Conclusions The licensee evaluated the radiological consequences resulting from the postulated AST FHA and concluded that the radiological consequences at the EAB, LPZ, and CR are within the dose guidelines provided in 10 CFR 50.67, RG 1.183 guidelines, and the SRP 15.0.1 radiological dose acceptance criteria for an FHA using an AST methodology. These criteria are 6.3 rem TEDE at the EAB for the worst 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , 6.3 rem TEDE at the LPZ for the duration of the accident, and 5 rem TEDE in the CR for the duration of the accident. The NRC staff=s review found that the licensee used analysis assumptions and inputs consistent with applicable regulatory guidance identified in Section 2.0 of this SE. The assumptions found acceptable to the staff are presented in Table 6 and the licensee=s calculated dose results and are given in Table 2 of this SE. The NRC staff performed independent confirmatory dose evaluations as necessary to ensure a thorough understanding of the licensee=s methods and results. The NRC staff finds that the EAB, LPZ, and CR doses estimated by the licensee for the AST FHA meet the applicable accident dose criteria and are therefore acceptable.

The licensee included a bounding calculation for a postulated release through grade level torus room hatch openings (i.e., torus room concrete plug penetrations), for the time requested of 84 hrs based on its CR MCREV assumptions in the initial submittal dated July 13, 2007. However, the supplement dated August 21, 2008, revised the original submittal to define RECENTLY IRRADIATED FUEL as fuel that has occupied part of a critical reactor core within the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months provided that all outside secondary containment ground-level hatches (hatches H15 through H24 and the Unit 2 and 3 Torus room access hatches) remain closed. Therefore, the NRC staff review of the licensing basis and TS changes proposed in the LAR for the AST FHA did not include an evaluation or approval of this bounding analysis for use in licensing basis assumptions for Peach Bottom dose consequence AST FHA.

3.2.4 AST Control Rod Drop Accident (CRDA) 3.2.4.1 Analysis Summary The postulated Peach Bottom control rod drop accident (CRDA) involves the rapid removal (i.e.,

drop) of a highest worth control rod resulting in a reactivity excursion. The postulated DBA event evaluated assumes that a control rod has been fully inserted and becomes stuck in this position. The control rod drive is assumed to be uncoupled and withdrawn. The rod subsequently becomes free and rapidly falls out of the core onto the withdrawn drive coupling.

The amount of positive reactivity introduced into the reactor core is at a rate consistent with the maximum control rod drop velocity, resulting in the insertion of a large positive reactivity and a localized power excursion. Peach Bottom assumed for its evaluation of the AST CRDA, that as a result of the accident, fuel damage would occur consisting of localized damage to fuel cladding with a limited amount of fuel melt occurring in the damaged rods.

3.2.4.2 AST Control Rod Drop Accident Source Term The licensee determined the CRDA core source terms are those associated with a DBA power level of 3528 MWth as discussed in Section 3.3.1 of this SE. In the CLB for Peach Bottom, fuel damage from a CRDA is based on GE 10x10 fuels in an 87.33 equivalent fuel pin array. There are 1,200 fuel rods breached with melting in 0.77 percent of the fuel contained in the breached rods. A conservative radial PF of 1.7 is used and of the 0.77 percent of the fuel that melts during the CRDA, 100 percent of noble gases and 50 percent of the iodines contained in the melted fuel fraction are assumed to be released to the reactor coolant in accordance with Appendix C of RG 1.183, Regulatory Position 1. The source term used for the CRDA analysis is the combination of the release of the gap activity from the fuel rods postulated to be damaged and the 0.77 percent fuel melt that occurs in the same number of damaged fuel rods.

Consistent with the guidance in RG 1.183, Appendix C, Sections 3.1 and 3.2, the licensee assumed that the gap activity and the activity from fuel pellet melting mixes instantaneously in the reactor coolant within the reactor pressure vessel with no credit for partitioning or removal by the steam separators.

3.2.4.3 AST Control Rod Drop Accident Transport The following release scenarios were considered by the licensee in its analysis of the AST CRDA:

The main condenser is assumed to leak activity into the TB at a rate of 1 percent per day. This activity is then released, unfiltered, to the environment by way of the RB/TB Exhaust Ventilation Stacks, taking no credit for holdup in the TB.

During power operation, the steam jet air ejector's (SJAE) discharge to the augmented off-gas system (AOG).

Flow from the sealing steam exhauster Low Power operation of the mechanical vacuum pump (MVP).

As described in the Peach Bottom UFSAR Section 7.12, the MSLs include MSL radiation monitors (MSLRM) designed to give prompt indication of a gross release of fission products from the fuel. When a significant increase in the main steam line radiation level is detected, trip signals are transmitted to the reactor protection system, the primary containment and reactor vessel isolation control system and to the off-gas system. Upon receipt of the high radiation trip signals, the reactor protection system initiates a scram, a primary containment and reactor isolation which initiates closure of all MSIVs, trips the MVP, if running, and closes the MVP suction valve. The MSL radiation trip setting is selected so that a high radiation trip results from the fission products released in the design basis CRDA. The radiation monitor trip setting is set above the background radiation level in the vicinity of the MSLs so that spurious trips are avoided at rated power. The setting is low enough that the monitor can respond to the fission products released during a postulated design basis CRDA, which occurs at a low steam flow condition.

Since the MVP, SJAE, and off gas systems are all immediately shutdown due to the automatic MSIV isolation function of the MSLRM caused by the high radiation levels following a CRDA, all forced flow paths are automatically disabled, all leakage from the main steam turbine condenser leaks to the atmosphere from the RB/TB ventilation exhaust stack. This becomes the only release of concern for the analysis of the AST CRDA. In accordance with Appendix C of RG 1.183, the licensee assumed that of the activity that reaches the turbine and condenser, 100 percent of the noble gases, 10 percent of the iodine, and 1 percent of the particulate radionuclides are available for release to the environment at a leak rate of 1 percent per day, for a period of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

3.2.4.4 CR Habitability for the AST Control Rod Drop Accident The licensee in its analysis for the AST CRDA did not credit the MCREV system and the associated filters. The normal air supply was assumed with a design basis intake of 20,600 scfm plus an allowance of 1600 scfm for additional in leakage. The dose consequence to CR operators for the AST CRDA is calculated over a total duration of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and the control room occupancy factor is 1 because the duration of this AST CRDA analysis is 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , which conforms to RG 1.183, Regulatory Position 4.2.6.

3.2.4.5 AST Control Rod Drop Accident Analysis Conclusions Peach Bottom evaluated the radiological consequences resulting from the postulated AST CRDA and concluded that the radiological consequences at the EAB, LPZ, and CR are within the dose guidelines provided in 10 CFR 50.67 and accident specific dose criteria specified in SRP 15.0.1. The NRC staffs review found that Peach Bottom used analysis assumptions and inputs consistent with applicable regulatory guidance identified in Section 2.0 of this SE. The assumptions found acceptable to the NRC staff are presented in Table 7 and the licensees calculated dose results are given in Table 2 of this SE. The NRC staff performed independent confirmatory dose evaluations as needed to ensure a complete understanding of the licensees methods. The EAB, LPZ, and CR doses estimated by the licensee for the AST CRDA were found to meet the applicable accident dose criteria and are therefore acceptable.

3.2.5 AST Main Steam Line Break Accident (MSLBA) 3.2.5.1 Analysis Summary Accidents that result in the release of radioactive materials outside the secondary containment are the result of postulated breaches in the nuclear system process barrier. The Peach Bottom CLB DBA is a complete severance of one main steam line outside the secondary containment.

Like the CLB main steam line break accident (MSLBA), the postulated Peach Bottom AST MSLBA assumes a double-ended break of one main steam line outside the secondary containment. The Peach Bottom MSLB accident is described in the FSAR Section 15.6.5, Main Steam Line Break Accident. Appendix D of RG 1.183 identifies acceptable radiological analysis assumptions for an MSLB. The licensee conservatively assumes an MSIV closure time of 10.5 seconds which maximizes the mass of coolant released in the TB. The break mass released includes the amount of steam in the steam line and connecting lines at the time of the break, plus the amount of steam that passes through the valves prior to closure in accordance with regulatory guidance. Because the engineered safeguard MSL flow restrictors are sized for the MSLBA, reactor vessel water level remains above the top of active fuel throughout the accident sequence. Therefore, adequate core cooling is maintained and no fuel failures are assumed in the CLB as well as for the postulated AST MSLBA.

3.2.5.2 AST Main Steam Line Break Accident Source Term Appendix D of RG 1.183, Section 2, states that if no or minimal fuel damage is postulated for the limiting event, the released activity should be the maximum coolant activity allowed by TS for the equilibrium case and for the pre-accident iodine spike case. Therefore, consistent with the Peach Bottom CLB and RG 1.183, the licensee evaluated the MSLBA based on the maximum equilibrium reactor coolant dose equivalent I-131 (DEI) concentration of 0.2 Ci/gm and, in a separate analysis, evaluated the MSLB assuming a pre-accident iodine spike DEI concentration of 4.0 Ci/gm as specified in TS 3.4.6 and its bases.

3.2.5.3 AST Main Steam Line Break Accident Transport The licensees analysis assumes the MSLBA to be an instantaneous ground level release. Two models are considered for assessment of the radiological consequences CR dose and offsite dose consequences. In both models the licensee based its assumptions on Appendix D of RG 1.183. The licensee states the following in Section 4.6.2 of the license amendment request, In the CR dose consequence model, the released reactor coolant and steam at operating temperature and pressure is conservatively assumed to expand to a hemispheric volume at atmospheric pressure and temperature. Dilution of the steam cloud by the air into which the steam is ejected is not credited in the licensee analysis. Neither the TB structure nor its ventilation system is assumed to have an effect on the cloud resulting from the MSLB accident. This hemisphere is then assumed to move at a speed of 1 meter per second downwind past the CR intake. No credit is taken for buoyant rise of the steam cloud or for decay in transit. Dilution (i.e., dispersion) of the activity in the plume in transit was also conservatively ignored.

For offsite dose consequence locations, the buoyant rise of the steam cloud is similarly ignored, and the ground level dispersion is based on an instantaneous release converted to an equivalent curie release. No credit is taken for decay over this release time, or in transit.

The radiological consequences resulting from a design basis MSLBA to a person at the EAB, to a person at the LPZ, and to an operator in the CR following the accident, was performed using a spreadsheet calculation.

3.2.5.4 CR Habitability for the AST Main Steam Line Break Accident CR operator doses are determined somewhat differently than the other AST DBAs, because steam cloud concentrations are used, rather than the meteorological dispersion factor (/Q) times a curie release rate. No CR filter credit is taken and, therefore, for inhalation, a dose for a location outside of the CR is used. For cloud submersion, a geometry factor is used to credit the reduced plume size seen in the control room.

3.2.5.5 AST Main Steam Line Break Accident Analysis Conclusions Peach Bottom evaluated the radiological consequences resulting from the postulated AST MSLBA and concluded that the radiological consequences at the EAB, LPZ, and CR are within the dose guidelines provided in 10 CFR 50.67 and accident specific dose criteria specified in SRP 15.0.1. The NRC staffs review found that Peach Bottom used analysis assumptions and inputs consistent with applicable regulatory guidance identified in Section 2.0 of this SE. The assumptions found acceptable to the NRC staff are presented in Table 8 and the licensees calculated dose results are given in Table 2 of this SE. The NRC staff reviewed the licensee dose evaluations to ensure a complete understanding of the licensees methods. The EAB, LPZ, and CR doses estimated by the licensee for the AST MSLBA were found to meet the applicable accident dose criteria and are therefore acceptable.

3.3 Atmospheric Dispersion Estimates 3.3.1 Meteorological Data The licensee used 5 years of onsite hourly meteorological data collected during calendar years 1984 through 1988 to generate new atmospheric dispersion factors (/Q values) for use in the proposed LAR. These data were 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 May 1997) and joint frequency distributions (JFDs) for input to the PAVAN atmospheric dispersion computer code (NUREG/CR-2858, November 1982, PAVAN: An Atmospheric Dispersion Program for Evaluating Design Basis Accidental Releases of Radiological Materials from Nuclear Power Stations). The data were used to generate CR, EAB, and outer boundary LPZ /Q values for the LOCA, CRDA, and FHA events evaluated in the proposed LAR.

Atmospheric dispersion estimates for MSLB accident releases were determined assuming a set of default meteorological conditions as discussed in the following subsections. The resulting

/Q values represent a change from those used in the current Peach Bottom UFSAR analyses for Units 2 and 3.

The onsite meteorological data were used to model atmospheric dispersion for six sets of release pathways: the RB stacks, RB personnel access doors, RB roof scuttles, railroad bay doors, ground level hatches and the off-gas stack. The FHA releases were assumed to be discharged to the environment via the personnel access doors, railroad bay doors, RB roof scuttle, and ground level hatches. The CRDA releases were assumed to be discharged via the RB/TB stacks. The LOCA primary containment releases are also assumed to occur through the RB stacks for a 3 minute period during and immediately following the secondary containment drawdown period. Once this initial period is completed, subsequent LOCA primary containment leakage releases are assumed to occur via the off-gas stack.

The set of meteorological data used in the proposed LAR atmospheric dispersion analyses was selected by the licensee from 35 years of available historical onsite data records. The licensee examined release locations and configurations in conjunction with the sharply varying topography (both in the vicinity of the release and at the desired receptor locations) in order to select the data sets for use in their dispersion analyses as shown in Table 1:

TABLE 1 Peach Bottom Onsite Meteorological Data Period of Record: 1984-1988 Measurement Elevations(a)

Tower Name Tower Location Wind Speed &

Direction Delta-T(b)

Data Applications(c)

Weather Tower 1A On the river bank SE of the plant at approximately 1 meter above plant grade 10 and 28 meters 27-10 meters

RB stack releases, RB personnel access doors, RB roof scuttles, railway bay doors, and ground hatches to the CR (ARCON96) and to the EAB and LPZ (PAVAN)

Delta-T with River Tower wind data for RB stack releases to the EAB 10 meters 46-10 meters

RB stack releases to the CR (ARCON96),

EAB and LPZ (PAVAN) 23 meters 46-10 meters Off-gas stack and RB stack releases to the CR (ARCON96)

Microwave Tower (Tower 2)

On a hill NW of the plant at approximately 76 meters above plant grade 98 meters 96-10 meters

Off-gas stack releases to the CR (ARCON96 and PAVAN)

Off-gas stack releases to the EAB/LPZ (PAVAN)

River Tower In the Susquehanna River NNE of the plant 14 meters (d)

Wind data for RB stack releases to the EAB (PAVAN)

(a) Height of sensors above tower grade.

(b) Stability class was based on delta-temperature (delta-T) measurements.

(c) Whenever a release was modeled with more than one meteorological data set, the resulting bounding (highest) /Q values were utilized in the subsequent dose analyses.

(d) Delta-T is not measured on the River Tower; subsequently, the 27-10 meter delta-T measurements from Weather Tower 1A were used in conjunction with the River Tower wind data.

The licensee stated that the 1984 through 1988 period was selected because continuous 5-year periods of data meeting system accuracy and data recovery specifications as defined in Regulatory Guide (RG) 1.23 were available from these tower locations during this period.

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. Examination of the data revealed that stable and neutral atmospheric conditions were generally reported to occur at night and unstable and neutral conditions during the day, as expected. NRC staff noted a very high occurrence of stable measurements on Tower 1A.

While Tower 1A is located in the river valley where stable conditions are expected to occur more frequently, delta-T measurements between 27 and 10 meters is an interval of 17 meters.

Measurements using such an interval would typically indicate a higher occurrence of stable and unstable conditions than would be measured using the 50 meter measurement interval recommended in RG 1.23, Rev. 1. However, this is expected to result in conservative estimates of the resultant /Q values as these data were used in this specific AST analysis.

Wind speed, wind direction, and stability class frequency distributions for each measurement channel were reasonably similar from year to year, except that an unusually high occurrence of low wind speeds (less than 0.5 miles per hour) was recorded for Tower 1A during 1984 as compared to 1985 through 1988. In addition, wind speed measurements in 1984 on all towers were recorded to the nearest mile per hour and wind direction measurements were recorded to the nearest 5° interval. This was due to the measurement recording methodology which the licensee explained was an artifact of the characteristics of the wind speed sensors and strip chart recorders used to collect and record the data during 1984. To assess the effects of the 1984 low wind speed data and recording methodology, NRC staff performed a random set of ARCON96 calculations using only the 1985 through 1988 meteorological data and compared these results to the licensees calculations using the complete 1984 through 1988 meteorological data set. NRC staff found no significant differences between the two sets of /Q values. Otherwise, any observed differences in wind and stability frequencies among the different measurement channels can be explained by flow patterns (e.g., channeling and slope flows) associated with a river valley site.

The licensee provided a set of joint wind speed, wind direction, and atmospheric stability distributions (JFDs) in Attachment J to the LAR dated July 13, 2007, which used seven wind speed categories of 0.5, 3.5, 7.5, 12.5, 18.5, 24.0, and 55.0 miles per hour based on RG 1.23, Revision 0. NRC staff requested that the licensee provide alternate JFDs that had a large number of wind speed categories at the lower wind speeds in order to reduce data clustering and skewing of wind speed classes and produce the best results using PAVAN. As a result, the licensee provided a revised set of JFDs using eleven wind speed categories of 0.5, 2.35, 3.47, 4.59, 6.82, 9.06, 11.30, 13.53, 18.01, 22.37, and 55.0 miles per hour. The licensee used these JFDS in the PAVAN analyses of offsite /Q values that were subsequently used in the EAB and LPZ dose assessments.

In summary, the NRC staff has reviewed the available information relative to the onsite meteorological measurements program and the ARCON96 and PAVAN meteorological data input files provided by the licensee. On the basis of this review, the NRC staff finds that these data provide an acceptable basis for making estimates of atmospheric dispersion for the proposed AST DBA assessments.

3.3.2 CR Atmospheric Dispersion Factors The licensee calculated CR air intake /Q values for the LOCA, CRDA, and FHA events using 1984 through 1988 onsite meteorological data and guidance provided in RG 1.194, Atmospheric Relative Concentrations for Control Room Radiological Habitability Assessment at Nuclear Power Plants. RG 1.194 states that ARCON96 is an acceptable methodology for assessing CR /Q values for use in DBA radiological analyses. NRC staff evaluated the applicability of the ARCON96 model and concluded that there are no unusual siting, building arrangements, release characterization, source-receptor configuration, meteorological regimes, or terrain conditions that preclude use of this model in support of the proposed LAR for Peach Bottom. When generating both ground level and elevated release /Q values, the license used the ARCON96 default values for surface roughness length and averaging sector width constant rather than the revised values listed in RG 1.194. However, results of comparison calculations performed by the NRC staff showed that any differences in the resultant /Q values were not significant.

The wind direction values provided as input to ARCON96 for 1985, 1986, and 1987 ranged from 0° to 360° instead of 1° to 360°, as specified in the ARCON96 Users Guide (Section 4.6.5 of NUREG/CR-6331). As such, wind direction values of 0° were interpreted as missing data instead of being treated as valid wind direction values. Since wind direction values of 0° occurred infrequently in the ARCON96 meteorological database, this did not have a significant impact on the licensees resulting /Q values.

RB stack, RB personnel access door, RB roof scuttle, railroad bay door and ground level hatch release CR /Q values were modeled as ground level releases based upon guidance provided in RG 1.194. The licensee also used the taut string methodology described in RG 1.194 to model the distance between the source and receptor locations. Because the top of the RB stacks are at approximately the same elevation as the lower level wind sensors on Tower 2, both Tower 1A and Tower 2 data were used comparatively to determine the most limiting /Q values for RB stack releases to the CR air intake for both Units 2 and 3 RB stacks and the resulting limiting /Q values were utilized in the subsequent dose analyses. Postulated releases from the RB personnel access doors, RB roof scuttles, railroad bay doors and ground level hatches were determined from Tower 1A data only as these data were measured in the river valley adjacent to the Peach Bottom facility.

Off-gas stack release CR /Q values were calculated using ARCON96 as well as the PAVAN atmospheric dispersion computer code and were modeled as an elevated release based upon guidance in RG 1.194. The resulting /Q values from these two models were then analyzed in accordance with RG 1.194, Section 3.2.2 guidance to derive the applicable CR /Q values. For the 0-2 hour time period, the licensee compared the maximum PAVAN /Q value with the ARCON96 value and determined that the PAVAN derivation resulted in the most conservative

/Q. The licensee noted that the higher of the maximum sector and site limit /Q was used in the assessment. For both the 1-4 days and 4-30 days time period, the licensee used the effective /Q value as allowed for stack releases per RG 1.194. This deterministic approach assumes that the stack plume reverses direction 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months daily throughout the event. Both the maximum PAVAN value and ARCON96 /Q value were used in this analysis as outlined in Equations [1] and [2] of RG 1.194 dated June 2003. The 2-8 hour and 8-24 hour time periods only used results from the ARCON96 calculations. Use of these methods is found acceptable by NRC staff.

The CR air intake /Q values for the analyses discussed above were used to analyze both pre-isolation outside air intake and post-isolation unfiltered inleakage to the CR. That is, the CR air intake /Q values were considered to be bounding for all potential inleakage pathways. This is discussed further in Section 3.2.2.1.

The licensee modeled the MSLB event to the CR using steam cloud concentrations rather than

/Q values. This is discussed further in Section 3.2.5.4.

In summary, the NRC staff qualitatively reviewed the inputs to the ARCON96 and PAVAN computer runs for the CR /Q value assessment and found them generally consistent with site configuration drawings and staff practice. The NRC staff also reviewed the licensees assessments of CR post-accident dispersion conditions generated from the licensees meteorological data and atmospheric dispersion modeling. The resulting CR /Q values are presented in Table 3-1. On the basis of this review, the NRC staff finds that these /Q values are acceptable for use in DBA CR dose assessments.

3.3.3 EAB/LPZ Atmospheric Dispersion Factors The licensee calculated EAB and LPZ /Q values for the LOCA, CRDA, and FHA events using 1984 through 1988 onsite meteorological data and guidance provided in RG 1.145, Atmospheric Dispersion Models for Potential Accident Consequence Assessments at Nuclear Power Plants. Atmospheric dispersion factors for the RB stacks, personnel access doors, railway bay doors, RB roof scuttle, and ground level hatches and off-gas stack release pathways were calculated using the PAVAN computer code. The licensee recalculated the EAB and LPZ /Q values using a finer wind speed breakdown as described in Section 3.3.1 above.

RB stacks, personnel access doors, railway bay doors, RB roof scuttles, and ground level hatches were treated as ground-level releases. EAB and LPZ /Q values were calculated using 46-10 meter delta-T data from Tower 2 and wind data from Tower 2 (10 meter level), Tower 1A (28 meter level wind data and 27-10 meter delta-T), and the River Tower (14 meter level wind data and 27-10 meter delta-T). The licensee stated that meteorological data measured on Tower 1A and the River Tower would not be generally representative of conditions at the LPZ, but for purposes of conservatism, the licensee also calculated /Q values using the Tower 1A data. The resulting highest /Q values from each wind data set were then selected for use in the subsequent dose analyses.

Off-gas stack releases were treated as elevated releases. Both EAB and LPZ /Q values were calculated using the 98 meter wind data and the 96-10 meter delta-T data from Tower 2. In accordance with the guidance provided in RG 1.145 for an inland site, fumigation conditions were assumed to occur during the first half-hour time period.

The licensee derived both EAB and LPZ atmospheric dispersion factors for the MSLB event using the /Q algorithm presented in RG 1.5, Assumptions Used for Evaluating the Potential Radiological Consequences of a Steam Line Break Accident for Boiling Water Reactors. This algorithm assumes the resulting steam cloud travels downwind at a height of 30 meters and is uniformly distributed in the vertical between the ground and 30 meters (e.g., fumigation conditions). Using EAB and LPZ distances of 823 and 7300 meters, respectively, default meteorological conditions of F stability and 1 meter per second wind speed were assumed.

In summary, the NRC staff qualitatively reviewed the inputs to the PAVAN computer runs and found them generally consistent with site configuration drawings and NRC staff practice. In addition, NRC staff reviewed the licensees assessments of EAB and LPZ post-accident dispersion conditions generated from the licensees meteorological data and atmospheric dispersion modeling. The resulting EAB and LPZ /Q values are presented in Table 3-2 and 3-3, respectively. On the basis of this review, the NRC staff finds that these /Q values are acceptable for use in DBA EAB and LPZ dose assessments.

3.3.4 Secondary Containment Drawdown - Meteorology RG 1.183, Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors, states that the effect of high winds on the ability of the secondary containment to maintain a negative pressure should be evaluated on an individual case basis.

The wind speed to be assumed is the 1-hour average value that is exceeded only 5 percent of the total number of hours in the data set. The licensee estimated the relevant Peach Bottom 95 percent wind speed as about 10 miles per hour. NRC staff finds this estimate acceptable from the 1984 through 1988 wind data measured on Tower 1A at the 10 meter level.

Design Basis Accidents EAB (2)

LPZ (3)

CR Loss of Coolant Accident 1.10E+01 9.00E+00 4.70E+00 Dose Criteria 2.50E+01 2.50E+01 5.00E+00 Fuel Handling Accident 2.50E+00 3.8E01 3.80E+00 Dose Criteria 6.30E+00 6.30E+00 5.00E+00 Control Rod Drop Accident (4) 8.6E02 3.0E02 3.0E01 Dose Criteria 6.30E+00 6.30E+00 5.00E+00 Main steamline break accident (5) 8.0E02 1.4E02 1.6E01 Dose Criteria 2.50E+00 2.50E+00 5.00E+00 Main steamline break accident (6) 1.60E+00 2.7E01 3.20E+00 Dose Criteria 2.50E+01 2.50E+01 5.00E+00 (1) Total effective dose equivalent (2) Exclusion area boundary (3) Low population zone (4) 1200 failed rods at full power (5) Maximum RCS equilibrium iodine activity (6) Pre-accident iodine spike Note: Licensee results are expressed to a limit of two significant figures Table 2 Peach Bottom Radiological Consequences Expressed as TEDE (1)

(rem)

Table 3-1 Peach Bottom CR Atmospheric Dispersion Factors Release Point: RB Stacks Accidents: CRDA and LOCA (during drawdown)

Time Interval

/Q Value (sec/m3) 0-2 hrs 1.18 x 10-3 2-8 hrs 9.08 x 10-4 8-24 hrs 4.14 x 10-4 1-4 days 2.90 x 10-4 4-30 days 2.26 x 10-4 Accident: FHA Time Interval: 0-2 hrs Release Locations(1)

/Q Value (sec/m3)

Personnel Access Doors 1.04 x 10-3 Railroad Bay Doors 4.54 x 10-4 RB Roof Scuttle 1.90 x 10-3 Ground Level Hatches 1.28 x 10-2 (1) The limiting /Q value was used for each release category (i.e., two personnel access doors, two railroad bay doors, two roof scuffles, and four ground level hatches).

Release Point: Off-Gas Stack Accident: LOCA (post drawdown)

Time Interval

/Q Value (sec/m3) 0-2 hrs 3.31 x 10-6 2-8 hrs 1.00 x 10-15 8-24 hrs 1.00 x 10-15 1-4 days 1.64 x 10-8 4-30 days 4.54 x 10-9 Table 3-2 Peach Bottom EAB Atmospheric Dispersion Factors Release Point: RB Exhaust Stacks, RB Personnel Access Doors, RB Roof, Railway Bay Doors, and Ground Level Hatches Accidents: FHA CRDA and LOCA (during drawdown)

Time Interval

/Q Value (sec/m3) 0-2 hrs 9.11 x 10-4 Release Point: Off-Gas Stack Accident: LOCA (post drawdown)

Time Interval

/Q Value (sec/m3) 0-0.5 hrs 5.30 x 10-5 0.5-2 hrs 9.17 x 10-6 Release Point: Main Steam Line Accident: MSLB Time Interval

/Q Value (sec/m3)

Puff Release 4.29 x 10-4 Table 3-3 Peach Bottom LPZ Atmospheric Dispersion Factors Release Point: RB Exhaust Stacks, RB Personnel Access Doors, RB Roof Scuttles, Railway Bay Doors, and Ground Level Hatches Accidents: FHA, CRDA, and LOCA (during drawdown)

Time Interval

/Q Value (sec/m3) 0-2 hrs 1.38 x 10-4 2-8 hrs 5.81 x 10-5 8-24 hrs 3.77 x 10-5 1-4 days 1.48 x 10-5 4-30 days 4.15 x 10-6 Release Point: Off-Gas Stack Accident: LOCA (post drawdown)

Time Interval

/Q Value (sec/m3) 0-0.5 hrs 1.75 x 10-5 0.5-2 hrs 9.05 x 10-6 2-8 hrs 4.01 x 10-6 8-24 hrs 2.67 x 10-6 1-4 days 1.10 x 10-6 4-30 days 3.10 x 10-7 Release Point: Main Steam Line Accident: MSLB Time Interval

/Q Value (sec/m3)

Puff Release 5.97 x 10-5 Control structure habitability envelope total volume 176,000 ft3 Control Room Normal Intake Flow 20,600 scfm Assumed Unfiltered Inleakage 1,600 scfm MCREV pressurization flow rate 3,000 scfm CR unfiltered inleakage 500 scfm CR isolation and MCREV initiation LOCA Automatic MSLB Not credited CRDA Not credited FHA Not credited Credited manual actions relative to the MCREV None MCREV HEPA filter efficiency credited in the analysis 98%

MCREV Charcoal filter efficiency credited in the analysis 89%

CRHE operator breathing rate 0 - 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months 3.5E04 m3/sec CR occupancy factors 0 - 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 1

24 - 96 hours0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months 0.6 96 - 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months 0.4 Table 4 Peach Bottom AST Control Room Data and Assumptions Core thermal power level 3528 MWt Drywell free volume (ft3) 159,000 Wetwell free volume (ft3) 127,700 Total free volume (ft3) 286,700 Primary containment leak rate 0 - 38 hours4.398148e-4 days 0.0106 hours 6.283069e-5 weeks 1.4459e-5 months 0.7%/day 38 - 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months 0.35%/day Primary containment aerosol deposition 10th percentile Powers Model Iodine chemical form in containment atmosphere cesium iodide 95%

elemental iodine 4.85%

organic iodine 0.15%

Containment sump pH 7

MSIV leak rate total (4 lines) 360 scfh for < 38 hrs @ 49.1 psig Broken line leakage 205 scfh for < 38 hrs @ 49.1 psig 102.5 scfh for >38 hrs @ 49.1 psig First intact line 155 scfh for < 38 hrs @ 49.1 psig 77.5 scfh for > 38 hrs @ 49.1 psig Second and Third intact line 0 scfh for < 30 days RB free air volume used (ft3) 2,500,000 RB Post-LOCA drawdown time assumed 3 minutes RB mixing efficiency Not Credited SGTS filter efficiency Not Credited Minimum post-LOCA suppression pool volume 122,900 ft3 Maximum post-LOCA suppression pool temperature

< 212 F Chemical Form of Iodine in ESF Leakage elemental 97%

organic 3%

ESF leakage assumption 10 gpm ESF flash fraction 10%

MCREV Initiation 30 minutes after LOCA MCREV air intake flow 2,700 cfm MCREV unfiltered inleakage 500 cfm Table 5 Peach Bottom AST Data and Assumptions for the LOCA Power level 3528 MWt Minimum post shutdown fuel handling time (decay time) 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Number of fuel assemblies in core 764 Number of equivalent fuel rods per assembly - GE12&GE14 10 x 10 87.33 Number of failed pins for fuel handling accident 172 Core radial peaking factor 1.7 Limiting Damaged Core Fraction with Power Factor (PF) 0.004382 Reactor Well minimum water level

> 23 feet Fuel bundle peak burnup will not exceed 62 GWD/MTU Fuel clad damage gap fractions I-131 8%

Remainder of halogens 5%

Kr-85 10%

Remainder of noble gases 5%

Alkali metals 12%

Pool DF Noble gases 1

Aerosols Infinite lodines - limiting event is over the reactor well 200 Duration of release 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months No credit is taken for filtration by the SGTS or Stack CR isolation and MCREV initiation Not Credited Normal flow intake rates, plus an unfiltered inleakage flow, is used Table 6 Peach Bottom AST Data and Assumptions for the FHA Core thermal power level 3528 MWt Radial peaking factor 1.7 Number of fuel assemblies in core 764 Number of equivalent fuel rods per assembly - GE12&GE14 10 x 10 87.33 Number of fuel rods damaged in full power CRDA 1200 Fraction of fission product inventory in gap Noble gases 0.1 Iodines 0.1 Alkali metals (Cs and Rb) 0.12 Fraction of damaged rods experiencing fuel melt 0.77%

Fraction of activity in melted regions released to RCS Noble gas 100.00%

Iodines 50.00%

Others as specified by Table 1 of Reg. Guide 1.183 Fraction of activity release in RCS reaching condenser Noble gas 100%

Iodines 10%

Others 1%

Fraction of activity from condenser available for release to environment Noble gas 100%

Iodines 10%

Others 1%

Release rate from condenser:

To turbine building 1% per day for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months CR isolation and MCREV initiation Not Credited Normal flow intake rates, plus an unfiltered inleakage flow, is used Control Room Occupancy Factor for 24hr duration of accident 1

Table 7 Peach Bottom AST Data and Assumptions for the CRDA Core thermal power level 3528 MWt Noble gas design source term 100,000 Ci/sec after 30 min Design offgas release rate 403,000 Ci/sec after 30 min Maximum equilibrium iodine 0.2 Ci/gm DE I-131(Case 1)

Pre-accident iodine spike 4.0 Ci/gm DE I-131(Case 2)

Iodine Isotope Activity Ci (Case 1)

Ci (Case 2) 1-131 6.07E+00 1.21E+02 1-132 3.64E+01 7.28E+02 1-133 3.71E+01 7.42E+02 1-134 5.25E+01 1.05E+03 1-135 4.64E+01 9.28E+02 Noble Gas Activity Kr-83M 2.25E-02 2.25E-02 Kr-85M 4.03E-02 4.03E-02 Kr-85 1.32E-04 1.32E-04 Kr-87 1.32E-01 1.32E-01 Kr-88 1.32E-01 1.32E-01 Kr-89 8.58E-01 8.58E-01 Xe-1 31M 9.90E-05 9.90E-05 Xe-133M 1.91E-03 1.91E-03 Xe-133 5.41E-02 5.41E-02 Xe-1 35M 1.72E-01 1.72E-01 Xe-135 1.45E-01 1.45E-01 Xe-137 9.90E-01 9.90E-01 Xe-138 5.87E-01 5.87E-01 MSIV isolation time assumed 10 seconds Liquid release 165,120 Ibm Steam release 25,800 Ibm CR isolation and MCREV initiation Not Credited Control Room Envelope 176,000 ft3 Table 8 Peach Bottom AST Data and Assumptions for the MSLBA 3.4 Evaluation of Technical Specification Changes The licensee has proposed the following TS changes for Peach Bottom Units 2 and 3. The proposed changes apply to both units and the NRC staffs review and acceptance of these changes apply to both units unless otherwise noted.

3.4.1 TS Section 1.1, "Definitions" The proposed change revises the definition of DOSE EQUIVALENT 1-131 in TS Section 1.1 to remove the word "thyroid" and to add a reference to Federal Guidance Report 11. This change is consistent with the TEDE basis of the radiological consequence analyses and provides an improved correlation between the TS specific activity LCO (where this definition is used) and the projected offsite and control room doses. The NRC staff, therefore, finds this proposed change acceptable.

3.4.2 TS Section 1.1, "Recently Irradiated Fuel" The proposed change revises Section 1.1 to add a new definition for RECENTLY IRRADIATED FUEL. This change adds the definition of RECENTLY IRRADIATED FUEL as TS definition 1.35. The initial submittal included a definition of RECENTLY IRRADIATED FUEL that included fuel that has occupied part of a critical reactor core within the previous 84 hours9.722222e-4 days 0.0233 hours 1.388889e-4 weeks 3.1962e-5 months with out any requirements that secondary containment ground-level hatches remain closed. However, the licensees supplement dated August 21, 2008, revised the original submittal to define RECENTLY IRRADIATED FUEL as fuel that has occupied part of a critical reactor core within the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months with the stipulation that specific secondary containment ground-level hatches must be maintained closed during movement of irradiated fuel in secondary containment. These hatches are H15, H16, H17, H18, H19, and H33 for Unit 2 and H20, H21, H22, H23, H24, and H34 for Unit 3.

The NRC staff has reviewed the proposed change and determined that the associated dose consequence analysis results are within acceptable limits. Therefore, the NRC staff finds that the insertion of this definition and its use for reference to allowed containment operability conditions is acceptable.

3.4.3 TS Section 3.1.7, "Standby Liquid Control (SLC) System" The proposed change revises the Applicability of TS Section 3.1.7 to add the requirement for the LCO to be met in Mode 3. This change implements AST assumptions regarding the use of the SLC System to buffer the suppression pool following a LOCA involving significant fission product release. The required actions for Condition D are being revised to add an additional requirement to be in Mode 4 with a completion time of 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

The SLC system is used in the AST analysis to ensure a torus water pH of greater than 7 for the duration of the postulated AST LOCA to provide the chemical form of radioiodine released into the containment assumed in the licensee dose consequence evaluation. After a LOCA, a variety of different chemical species are released from the damaged core. One of these species is radioactive iodine. This iodine, when released to the outside environment will significantly contribute to radiation doses. It is, therefore, essential to keep it confined within the plants containment. According to NUREG-1465, Accident Source Terms for Light-Water Nuclear Power Plant, iodine is released from the core in three different chemical forms; at least 95 percent is released in ionic form as cesium iodide (CsI) and the remaining 5 percent as elemental iodine (I2) and hydriodic acid (HI), with at least 1 percent of each of them. CsI and HI are ionized in water environment and are, therefore, soluble. However, elemental iodine is scarcely soluble. It is of interest, therefore, to maintain as much as possible of the released iodine in ionic form. Unfortunately, in radiation environments existing in containment, some of the ionic iodine dissolved in water is converted into elemental form. The degree of conversion varies significantly with the pH of water. At a higher pH, conversion to elemental form is lower and at pH >7 it becomes negligibly small. The relationship between the degree of conversion and pH is specified in Figure 3.1 of NUREG/CR-5950, Iodine Evolution and pH Control.

In Peach Bottom Units 2 and 3, most of the iodine is released from the core to the suppression pool. Therefore, in order to keep it dissolved, the suppression pool water should be kept at pH7 throughout the 30-day post-LOCA period. The licensee has demonstrated that because of strong acid formation in the containment, this is not achievable, without adding buffering chemicals to control the water pH.

The licensee calculated that after a LOCA, the pH value will be continuously decreasing due to formation of hydrochloric and nitric acid in containment. Hydrochloric acid is formed from the decomposition of Hypalon cable insulation by radiation. About 3.21E-4 mols/liter of hydrochloric acid is formed during the 30-day period. Nitric acid is formed by irradiation of air and water and about 8.74E-5 mols/liter of nitric acid is formed during the same period. Both are strong acids and will significantly contribute to lowering suppression pool pH. In order to neutralize their effect, the licensee credited the buffering effect of sodium pentaborate from the SLC system.

The main purpose of the SLC system is to control reactivity in the case of control rod failure.

However, since sodium pentaborate is derived from a strong base and a weak acid it can also act as a buffer. Such buffering action could maintain basic pH in the suppression pool despite the presence of strong acids. The licensee has calculated that adding 162.7 lbm of sodium pentaborate from the SLC system (Technical Specification SR 3.1.7.7) will maintain a basic pH in the suppression pool for 30 days. In the licensees analysis the addition is accomplished within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months after a LOCA.

In order to evaluate beneficial effect of sodium pentaborate, the licensee calculated suppression pool pH for unbuffered and buffered cases. As expected, without addition of sodium pentaborate but taking only credit for the presence of Cs(OH), the value of pH during the 30 day period was below 7, reaching a minimum pH value of 3.5. With the addition of sodium pentaborate, the pH will increase rapidly above 7 and remains above a pH of 8 for the 30 days post-LOCA. The NRC staff has independently verified the licensees calculations and has determined that by using sodium pentaborate as a buffer, the pH of the suppression pool will remain above a pH of 7 for 30 days post-LOCA.

To demonstrate that the SLC system was capable of performing its intended safety function during a LOCA following AST implementation, the licensee utilized the guidance provided by the NRC in the review guideline document titled GUIDANCE ON THE ASSESSMENT OF A BWR SLC SYSTEM FOR pH CONTROL (ADAMS Accession No. ML040640364). The fourth guideline in this document concerns the lack of redundancy of a plants SLC system with respect to its active components. A plant can offset this lack of redundancy by showing acceptable quality and reliability of non-redundant active components and/or compensatory actions in the event of failure of the non-redundant active components by meeting the six criteria outlined in the document. The second criterion of these six requires the licensee to provide the design-basis conditions for the components and the environmental and seismic conditions under which the components may be required to operating during a DBA.

One of the non-redundant, active components described by the licensee is REMOTE SWITCH RMS-2(3)-11A-S001 which turns on either the A or B SLC pump in the event that SLC injection is necessary. In its July 13, 2007 submittal, the licensee indicated that this switch is located in the main control room and is subject to a mild environment. In its February 28, 2008 response to the NRC staffs RAI, the licensee provided additional information regarding the switch and indicated that the subject switches are dynamically qualified and are designed to function during a DBA and following an SSE. The licensee also provided detailed analyses demonstrating the seismic qualification calculations related to the consoles and instruments associated with the switch under review. It is noted in these calculations that for this particular switch, seismic qualification was based on testing which is documented in GE document NEFO-10678, Seismic Qualification of Class I Electric Equipment. The NRC staff has reviewed the analyses provided by the licensee and finds that the aforementioned electric equipment will continue to operate safely upon implementation of the proposed AST.

In summary, the licensee described its methodology for controlling the post-LOCA pH in the suppression pool to remain above a pH value of 7. The methodology relies on using buffering action of sodium pentaborate, introduced into the suppression pool from the SLC system and provided analyses to support the ability of the SLCS to operate safely upon implementation of the proposed AST. The licensee also provided analyses justifying that using 162.7 lbm of sodium pentaborate will ensure that the pH in the suppression pool will stay above 7 for 30 days after a LOCA. The NRC staff has reviewed the calculations and justifications provided by the licensee and, based in this review, finds that the analysis presented in the licensees submittal support the conclusion that the SLC system will operate as required and that the suppression pool pH will stay basic for the period of 30 days after a LOCA. Therefore, the NRC staff finds the licensees analysis acceptable.

3.4.4 TS Section 3.3.6.1, "Primary Containment Isolation Instrumentation" TS Section 3.3.6.1, Table 3.3.6.1-1 lists the applicability requirements for Primary Containment Isolation Instrumentation. The proposed change adds the requirement that the SLC System Initiation Function of the Reactor Water Cleanup System Isolation Instrumentation be operable in Mode 3. The revised applicability for the SLC System Initiation function is consistent with the revised applicability for the SLC system as discussed in Section 3.4.3 of this SE. In addition, this change is consistent with the secondary containment operability changes related to TSTF-51 which is related to the acceptable dose consequences for the licensee AST FHA analysis.

Therefore, the NRC staff finds this change to be acceptable.

3.4.5 TS Section 3.3.6.2, "Secondary Containment Isolation Instrumentation" The proposed change revises footnote (b) of TS Table 3.3.6.2-1 by deleting, "CORE ALTERATIONS, and during," and replaces irradiated fuel with recently irradiated fuel. This eliminates the requirement for Function 3 (i.e., RB Ventilation Exhaust Radiation - High) and Function 4 (i.e., Refueling Floor Ventilation Exhaust Radiation - High) of the Secondary Containment Isolation Instrumentation to be operable during core alterations after the fuel assemblies no longer meet the definition of recently irradiated. The licensee proposed change is consistent with the secondary containment operability changes related to TSTF-51 which is related to the acceptable dose consequences for the licensee AST FHA analysis. Therefore, the NRC staff finds this change to be acceptable.

3.4.6 TS Section 3.6.1.3, "Primary Containment Isolation Valves (PCIVs)"

A. TS 3.6.1.3 Purge or Vent valve change The proposed change involves revising TS Section 3.6.1.3 to renumber existing CONDITIONS E and F to F and G respectively. A new CONDITION E will add an LCO requirement stipulating that the accumulated time a purge or vent flow path shall be open shall be limited to 90 hours0.00104 days 0.025 hours 1.488095e-4 weeks 3.4245e-5 months per calendar year, with the reactor in MODES 1 and 2, and reactor pressure greater than 100 psig. In the event that flow paths are open greater than 90 hours0.00104 days 0.025 hours 1.488095e-4 weeks 3.4245e-5 months in a calendar year, the penetration(s) must be isolated within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or the reactor shall be in MODE 3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in MODE 4 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . This will limit the total time that a flow path exists through certain containment penetrations.

The licensee stated that containment purge as a combustible gas or pressure control measure is not required within 30 days following a DBA LOCA. Therefore the release from containment purge is not analyzed per RG 1.183, Appendix A, Section 7. This is consistent with the regulatory guide and is acceptable. In 1989 TS Amendment Nos. 144 and 146 originally granted the incorporation into the PBAPS TS of the 90-hour limit for purge and vent valve operation and the requirement that both trains of the standby gas treatment system (SGTS) be operable and that only one train of SGTS be in operation. However, as a result of the PBAPS conversion to the Improved Standard Technical Specifications (ITS), many of the custom TS requirements were relocated out of TS and into various licensee controlled documents. When PBAPS converted to ITS, the 90-hour limit for purge and vent valve operation and the SGTS operational requirements were relocated to plant procedures per Amendment Nos. 210 and 214. The NRC staff finds it appropriate and acceptable to restore the TS requirement to limit the purge/vent flowpath accumulated open time to less than or equal to 90 hours0.00104 days 0.025 hours 1.488095e-4 weeks 3.4245e-5 months for the calendar year while in MODE 1 or 2 with the reactor pressure greater than 100 psig.

B. SR 3.6.1.3.14-Increase allowable MSIV leakage rate The licensee proposed to revise the Peach Bottom TS SR 3.6.1.3.14 to increase the allowable limit for the combined leakage rate for all MSIV leakage paths to less than or equal to 204 scfh for all four main steam lines and less than or equal to 116 scfh for any one main steam line, when tested at greater than or equal to 25 psig. The licensee concluded and the NRC staff accepts that these increased allowed MSIV leakages, combined with other leakage pathways in its AST LOCA analysis, showed that the radiological consequences at the EAB, LPZ, and CR are within the dose guidelines provided in 10 CFR 50.67 and accident dose criteria specified in SRP Section 15.0.1. Therefore, the NRC staff finds this change to be acceptable.

3.4.7 TS Section 3.6.4.1, "Secondary Containment" The proposed change deletes "During CORE ALTERATIONS" from the applicability statement for TS LCO 3.6.4.1 and relaxes TS requirements to require LCO 3.6.4.1 to be applicable only when handling recently irradiated fuel. The proposed change revises Condition C, and associated required actions and completion times, to reflect the revision of the applicability requirements for LCO 3.6.4.1. With the application of AST, secondary containment is not credited for the FHA after a 24-hour decay period when all outside secondary containment ground-level hatches (hatches H15 through H24 and Units 2 and 3 Torus room access hatches) are closed. The proposed change to increase the secondary containment drawdown time from less than or equal to 120 seconds to less than or equal to 180 seconds was used in the licensee AST LOCA analysis which showed that the radiological consequences at the EAB, LPZ, and CR are within the dose guidelines provided in 10 CFR 50.67 and accident dose criteria specified in SRP Section 15.0.1. These changes are consistent with the secondary containment operability changes related to TSTF-51 which is related to the acceptable dose consequences for the licensee AST FHA analysis. Therefore, the NRC staff finds this change to be acceptable.

3.4.8 TS Section 3.6.4.2, "Secondary Containment Isolation Valves (SCIVs)"

The proposed change deletes "During CORE ALTERATIONS" from the applicability statement for TS LCO 3.6.4.2 and relaxes TS requirements to require LCO 3.6.4.2 to be applicable only when handling recently irradiated fuel. The proposed change revises Condition D, and associated required actions by replacing the phrase irradiated fuel with recently irradiated fuel to reflect the revision of the applicability requirements for LCO 3.6.4.2. The REQUIRED ACTION D.2, Suspend CORE ALTERATION and its completion time are deleted since the applicability applies to recently irradiated fuel which is addressed by REQUIRED ACTION D.1.

REQUIRED ACTION D.3 is renumbered as D.2. With the application of AST, closure of secondary containment isolation valves is not credited for the FHA after a 24-hour decay period when all outside secondary containment ground-level hatches (hatches H15 through H24 and Units 2 and 3 Torus room access hatches) are closed. The proposed changes to TS 3.6.4.2 are consistent with the secondary containment operability changes related to TSTF-51, which is related to the acceptable dose consequences for the licensee AST FHA analysis. Therefore, the NRC staff finds this change to be acceptable.

3.4.9 TS Section 3.6.4.3, "Standby Gas Treatment (SGT) System" The proposed change deletes "During CORE ALTERATIONS" from the applicability statement for TS LCO 3.6.4.3 and relaxes TS requirements to require LCO 3.6.4.3 to be applicable only when handling recently irradiated fuel by replacing the phrase irradiated fuel with recently irradiated fuel. The proposed change revises Condition C and Condition E, and associated required actions and completion times, to reflect the revision of the applicability requirements for LCO 3.6.4.3. The REQUIRED ACTION E.2, Suspend CORE ALTERATION and its completion time are deleted since the applicability applies to recently irradiated fuel which is addressed by REQUIRED ACTION E.1. REQUIRED ACTION E.3 is renumbered as E.2.

These changes are being made to reflect that, with application of AST, the SGT System is no longer required to be operable during movement of irradiated fuel assemblies, which have decayed at least 24-hours when all outside secondary containment ground-level hatches are maintained closed (hatches H15 through H24 and Units 2 and 3 Torus room access hatches), in the secondary containment, or during core alterations, since this system is not credited for the FHA after a 24-hour decay period. The proposed changes to TS 3.6.4.3 are consistent with the secondary containment operability changes related to TSTF-51, which is related to the acceptable dose consequences for the licensee AST FHA analysis. Therefore, the NRC staff finds this change to be acceptable.

3.4.10 TS Section 5.5.12, "Primary Containment Leakage Rate Testing Program" The proposed change to increase the primary containment leakage to 0.7 percent of primary containment air weight per day was used in the licensee AST LOCA analysis which showed that the radiological consequences at the EAB, LPZ, and CR are within the dose guidelines provided in 10 CFR 50.67 and accident dose criteria specified in SRP Section 15.0.1. Therefore, the NRC staff finds this change to be acceptable.

3.4.11 UFSAR Section 5.2.4.3.2, "Minimum Containment Pressure Available" In a request for a license amendment dated August 11, 1999, the licensee proposed revising the PBAPS UFSAR to clarify the licensing basis with regard to the allowable containment overpressure (accident pressure greater than atmospheric) for ensuring adequate NPSH for the ECCS pumps. The NRC granted this change in a letter from the U.S. NRC to Mr. J. A. Hutton, PECO Energy Company, Peach Bottom Atomic Power Station, Units Nos. 2 and 3 - Issuance of Amendment Regarding Crediting of Containment Overpressure for Net Positive Suction Head Calculations for Emergency Core Cooling Pumps, dated August 14, 2000.

The UFSAR discussion contains two curves. These are denoted the MCPA and the COPL.

The MCPA is the minimum containment pressure available for a given accident. This is the calculated containment accident pressure as a function of time, conservatively minimized. The COPL is the containment accident pressure required to ensure an acceptable NPSH margin for the most limiting ECCS pump. The COPL curve is less than the MCPA curve.

As part of the proposed change to AST the licensee is changing the TS limits for maximum allowable primary containment leakage (see Section 3.4.10 above). The changes in the AST proposed containment leakage values for the MSIVs and general containment leakage, required the MCPA to be revised.

Per Section 5.2 of the PBAPS UFSAR, emergency pumps that take suction from the suppression pool rely on some amount of containment pressure to provide adequate NPSH at elevated suppression pool temperatures. The bounding event for containment overpressure required (COPR) is the design basis LOCA. The MCPA following a LOCA must be greater than the COPR to assure that the NPSH-available is greater than or equal to the NPSH-required to prevent cavitation in the ECCS pumps. The licensees supplement dated April 17, 2008, proposes to eliminate the COPL curve from the UFSAR and design basis calculations. The NRC staff finds this acceptable since maintaining MCPA greater than or equal to COPR assures acceptable operation of the RHR and CS pumps. In addition, any future changes that could impact MCPA or COPR for these systems will be evaluated by the licensee in accordance with the requirements of 10 CFR 50.59.

3.5 Summary The NRC staff reviewed the assumptions, inputs, and methods used by the licensee to assess the radiological consequences of DBAs with full implementation of an AST at Peach Bottom Units 2 and 3. The NRC staff finds that the licensee used analysis methods and assumptions consistent with the conservative regulatory requirements and guidance identified in Section 2.0 above. The NRC staff also finds, with reasonable assurance that the estimates of the EAB, LPZ, and CR doses comply with the criteria specified in Section 2.0. The NRC staff further finds reasonable assurance that Peach Bottom Units 2 and 3, as modified by these license amendments, will continue to provide sufficient safety margins with adequate defense-in-depth to address unanticipated events and to compensate for uncertainties in accident progression and analysis assumptions and parameters. Therefore, the NRC staff finds that the proposed license amendments are acceptable with respect to the radiological consequences of DBAs.

In addition, the NRC staff has reviewed the licensees assessment of the impact of the proposed amendments on portions of the main-steam system, SLC system, and HVAC ductwork and components with regard to the seismic qualification involved with these items as they relate to the AST implementation at PBAPS. On the basis of this review, the NRC staff finds that the proposed AST implementation will not have an adverse impact on the ability of these systems to withstand and perform their intended safety functions following an SSE.

The NRC staff finds that there is reasonable assurance that there will be adequate protection of public health and safety and the environment if the requested amendments are implemented.

This licensing action is considered a full implementation of the AST. With this approval, the previous AST in the Peach Bottom design basis is superseded by the AST 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 criteria of 10 CFR Part 50.67, or fractions thereof, as defined in RG 1.183.

3.0 STATE CONSULTATION

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

4.0 ENVIRONMENTAL CONSIDERATION

The amendments change a requirement with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20 and changes SRs.

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 (73 FR 25040). 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.

5.0 CONCLUSION

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

Principal Contributor: James Shea Alexander Tsirigotis Leta Brown William Jessup Emma L. Wong Date: September 5, 2008

v t e Letter Sequence Approval
TAC:MD6806, (Approved, Closed) TAC:MD6807, (Open)
Initiation Acceptance... Supplement, Supplement Administration Questions Answers 28 March 2008 23 May 2008 29 July 2008 Results Approval Other: ML082490831
MONTHYEAR2008-05-23 [Table View]

Person / Time
ML082320406
Site:	Peach Bottom (DPR-044, DPR-056)
Issue date:	09/05/2008
From:	John Hughey NRC/NRR/ADRO/DORL/LPLI-2
To:	Pardee C Exelon Generation Co
Hughey J, NRR/DORL, 301-415-3204
Shared Package
ML082320257	List: ML082320406 ML082490831
References
TAC MD6806, TAC MD6807
Download: ML082320406 (52)
v • d • e

ML082320406

Text

SUBJECT:

Dear Mr. Pardee:

Enclosures:

Enclosures:

Attachment:

Attachment:

1.0 INTRODUCTION

2.0 REGULATORY EVALUATION

3.0 TECHNICAL EVALUATION

3.1 Proposed TS Changes

3.0 STATE CONSULTATION

4.0 ENVIRONMENTAL CONSIDERATION

5.0 CONCLUSION

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