ML14211A019

License Amendment Request Revise Technical Specifications Definition for Recently Irradiated Fuel

ML14211A019
Person / Time
Site:	Peach Bottom
Issue date:	07/25/2014
From:	Jim Barstow Exelon Generation Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
Download: ML14211A019 (22)

Text

o200 Exelon Way EKennett Square. PA 19348 Exe~on Generation.

www.exeloncorp.com 10 CFR 50.90 July 25, 2014 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Peach Bottom Atomic Power Station, Units 2 and 3 Renewed Facility Operating License Nos. DPR-44 and DPR-56 NRC Docket Nos. 50-277 and 50-278

Subject:

License Amendment Request Revise Technical Specifications Definition for RECENTLY IRRADIATED FUEL In accordance with 10 CFR 50.90, "Application for amendment of license, construction permit, or early site permit, " Exelon Generation Company, LLC (Exelon) requests amendments to Appendix A, Technical Specifications (TS) of Renewed Facility Operating License Nos. DPR-44 and DPR-56 for Peach Bottom Atomic Power Station (PBAPS), Units 2 and 3, respectively.

This submittal requests changes to the definitions described in the PBAPS, Unit 2 and 3, TS for the term RECENTLY IRRADIATED FUEL. Currently, the definitions include limitations requiring that certain ground-level hatches remain closed during movement of any irradiated fuel in Secondary Containment. The proposed changes will modify the definitions for RECENTLY IRRADIATED FUEL to: 1) revise the specific restriction identifying the Secondary Containment hatches listed, and 2) address a discrepancy in the designation for identifying the Secondary Containment hatch numbers. provides the evaluation of the proposed changes. Attachment 2 provides the marked-up TS pages indicating the proposed changes. Attachments 3 and 4 provide calculations supporting the requested changes.

Exelon has concluded that the proposed changes present no significant hazards consideration under the standards set forth in 10 CFR 50.92.

Exelon requests approval of the proposed amendment by July 25, 2015. This schedule is being requested in order to support outage activities for Unit 3 in September 2015. Upon NRC approval, the amendment shall be implemented within 60 days of issuance.

These proposed changes have been reviewed and approved by the station's Plant Operations Review Committee and by the Nuclear Safety Review Board.

U.S. Nuclear Regulatory Commission Revise Definition for Recently Irradiated Fuel July 25, 2014 Page 2 In accordance with 10 CFR 50.91, "Notice forpublic comment; State consultation," paragraph (b),

Exelon is notifying the Commonwealth of Pennsylvania of this application for license amendment by transmitting a copy of this letter and its attachments to the designated State Official.

There are no regulatory commitments contained within this submittal.

If you have any questions or require additional information, please contact Richard Gropp at (610) 765-5557.

I declare under penalty of perjury that the foregoing is true and correct. Executed on the 25th day of July 2014.

Respectfully, James Barstow Director, Licensing and Regulatory Affairs Exelon Generation Company, LLC Attachments:

1. Evaluation of Proposed Changes

2. Proposed Technical Specifications Pages

3. Calculation PM-1059, Revision 5, "EAB, LPZ, and CR Doses Due to Fuel Handling Accident (FHA)"

4. Calculation PM-1 170, Revision 0, "PBAPS Atmospheric Dispersion Factors (X/Qs) for post-FHA Ground Hatch Releases" cc:

Regional Administrator - NRC Region I NRC Senior Resident Inspector - Peach Bottom Atomic Power Station NRC Project Manager, NRR - Peach Bottom Atomic Power Station S. T. Gray, State of Maryland R. R. Janati, Commonwealth of Pennsylvania w/ Attachments

ATTACHMENT 1 License Amendment Request Peach Bottom Atomic Power Station, Units 2 and 3 Docket Nos. 50-277 and 50-278 EVALUATION OF PROPOSED CHANGES Revise Technical Specifications Definitions for RECENTLY IRRADIATED FUEL

2.0 PROPOSED CHANGE

S

5.0 REGULATORY ANALYSIS

5.1 No Significant Hazards Consideration 5.2 Applicable Regulatory Requirements/Criteria

6.0 ENVIRONMENTAL CONSIDERATION

7.0 PRECEDENT

8.0 REFERENCES

Page 1 of 16 ATTACHMENT 1 EVALUATION OF PROPOSED CHANGES

1.0 DESCRIPTION

In accordance with 10 CFR 50.90, "Application for amendment of license, construction permit, or early site permit," Exelon Generation Company, LLC (Exelon) requests amendments to Appendix A, Technical Specifications (TS) of Renewed Facility Operating License Nos. DPR-44 and DPR-56 for Peach Bottom Atomic Power Station (PBAPS), Units 2 and 3, respectively.

This submittal requests changes to the definition described in the PBAPS, Units 2 and 3, TS for the term RECENTLY IRRADIATED FUEL. Currently, the definitions include limitations requiring that certain ground-level hatches (located on the west-side of the Reactor Buildings - Figure 1) remain closed during movement of any irradiated fuel in Secondary Containment (SC). The proposed changes will modify the definition for RECENTLY IRRADIATED FUEL to: 1) revise the specific restriction identifying the SC hatches listed, and 2) address a discrepancy in the designation for identifying the SC hatch numbers.

The evaluation of the proposed changes provided in this attachment includes a detailed discussion and description of the proposed TS changes, a safety assessment of the proposed TS changes, information supporting a finding of No Significant Hazards Consideration, and information supporting an Environmental Assessment. Attachment 2 provides the marked-up TS pages for the proposed changes. Attachments 3 and 4 include copies of the supporting calculations (i.e., PM-1059, Revision 5 and PM-1 170, Revision 0, respectively).

2.0 PROPOSED CHANGE

S TS Section 1.1, Definitions - RECENTLY IRRADIATED FUEL Currently, the TS definition of RECENTLY IRRADIATED FUEL for PBAPS, Units 2 and 3 limits the opening of several ground-level hatches when using the definition of RECENTLY IRRADIATED FUEL to suspend the applicability of Limiting Condition for Operations (LCOs) associated with the SC, SC Isolation Valves, and Standby Gas Treatment (SGT).

Exelon is proposing to modify the definition for PBAPS, Units 2 and 3, as follows:

Unit 2 Current 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 <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. When using this definition to suspend the Applicability of LCOs, secondary containment ground-level hatches H15, H16, H1 7, H18, H19, and H33 shall be closed during the movement of any irradiated fuel in Secondary Containment.

Page 2 of 16 Proposed Definition for RECENTLY IRRADIATED FUEL RECENTLY IRRADIA TED FUEL is fuel that has occupied part of a critical reactor core within the previous 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

Unit 3 Current Definition for RECENTLY IRRADIATED FUEL RECENTLY IRRADIA TED FUEL is fuel that has occupied part of a critical reactor core within the previous 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. When using this definition to suspend the Applicability of LCOs, secondary containment ground-level hatches H20, H21, H22, H23, H24, and H34 shall be closed during the movement of any irradiated fuel in Secondary Containment.

Proposed Definition for RECENTLY IRRADIATED FUEL RECENTLY IRRADIATED FUEL is fuel that has occupied part of a critical reactor core within the previous 312 hours0.00361 days <br />0.0867 hours <br />5.15873e-4 weeks <br />1.18716e-4 months <br />. This 312-hour time period may be reduced to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> if secondary containment hatches H2, H21, H22 and H34 are closed.

The proposed changes are supported by the analysis provided in Calculation PM-1059, Revision 5 (Attachment 3) which demonstrates that the resulting post-Fuel Handling Accident (FHA) radiological dose consequences due to the releases from the opened SC hatches are within allowable regulatory limits. The proposed changes are also supported by Calculation PM-1 170, Revision 0 (Attachment 4), which provides revised X/Q values for the ground-level releases.

3.0 BACKGROUND

Exelon is proposing to change the Unit 2 and Unit 3 definitions of RECENTLY IRRADIATED FUEL to allow for the performance of more efficient outages. The current PBAPS TS definitions require SC ground-level hatches, specifically, those above the Residual Heat Removal (RHR)

Rooms to be in place during movement of recently irradiated fuel assemblies. As a result of Extended Power Uprate (EPU) modifications that will be installed in Unit 3 in the Fall of 2015, changes to the definition of RECENTLY IRRADIATED FUEL are needed specifically for installation of the PBAPS Unit 3 RHR cross-tie modification in the RHR Rooms. The installation will be facilitated by removal of the RHR Room hatches, which are currently listed in the definition of RECENTLY IRRADIATED FUEL in the PBAPS TS for Units 2 and 3. Without an NRC-approved change to the PBAPS TS, movement of irradiated fuel is not permitted without the RHR Room hatches being in place. This situation potentially complicates the PBAPS Unit 3 cross-tie modification in the RHR Rooms with possible impact on the outage duration and cost.

In addition, allowing opening of the west-side hatches currently listed in the TS definitions of RECENTLY IRRADIATED FUEL will facilitate access through these hatches to efficiently perform maintenance, along with modification and replacement of equipment and components located in the High Pressure Coolant Injection (HPCI) Room, Torus Room, and RHR Rooms without impacting the refueling outage.

Page 3 of 16 The current definitions in the Unit 2 and Unit 3 TS are as follows:

Unit 2 RECENTLY RECENTLY IRRADIATED FUEL is fuel that has occupied IRRADIA TED FUEL part of a critical reactor core within the previous 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

When using this definition to suspend the Applicability of LCOs, secondary containment ground-level hatches H15, H16, H17, H18, H19, and H33 shall be closed during the movement of any irradiated fuel in Secondary Containment.

Unit 3 RECENTLY RECENTLY IRRADIA TED FUEL is fuel that has occupied IRRADIA TED FUEL part of a critical reactor core within the previous 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

When using this definition to suspend the Applicability of LCOs, secondary containment ground-level hatches H20, H21, H22, H23, H24, and H34 shall be closed during the movement of any irradiated fuel in Secondary Containment.

Based on calculations and analyses performed by Exelon, the definitions described below are being proposed:

Unit 2 RECENTLY RECENTLY IRRADIATED FUEL is fuel that has occupied I IRRADIATED FUEL part of a critical reactor core within the previous 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

Unit 3 RECENTLY RECENTLY IRRADIATED FUEL is fuel that has occupied IRRADIATED FUEL part of a critical reactor core within the previous 312 hours0.00361 days <br />0.0867 hours <br />5.15873e-4 weeks <br />1.18716e-4 months <br />.

This 312-hour time period may be reduced to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> if secondary containment hatches H2, H21, H22 and H34 are closed.

An administrative change is also proposed to address a discrepancy in the listing of the SC hatches. Hatches H1 (Unit 2) and H2 (Unit 3) establish the SC pressure boundary for the Unit 2 and Unit 3 HPCI Rooms, respectively. These hatches are located in the ceiling of the HPCI Rooms. The ceiling of the HPCI Room is the floor for the Reactor Building Closed Cooling Water (RBCCW) Room that is located on the 116' elevation. The RBCCW Room is not part of the SC. Hatches H19 and H20 are in the ceiling of the RBCCW Rooms for Units 2 and 3, respectively. Hatches H19 and H20 open to the outside at grade-level (135' elevation), to the west-side of the Reactor Buildings (RBs). For HPCI maintenance that necessitates removing large components, hatches H1 and H19 (Unit 2) or hatches H2 and H20 (Unit 3) would need to be opened to remove these large components from the Unit 2 or Unit 3 HPCI Rooms. This existing discrepancy in the TS discussed above was inadvertently introduced in connection with Page 4 of 16 an August 21, 2008, supplemental response (Reference 1) when the ground-level H19/H20 hatches were designated in lieu of hatches H1/H2. The ground-level hatches H19 and H20 establish the release paths to the environment for releases through hatches H1 and H2, respectively, during a Fuel Handling Accident (FHA). Therefore, the discussions in the following section and the analysis in PM-1059, Revision 5 (Attachment 3) about the post-FHA releases through ground-level hatches H19 and H20 are directly applied to the releases from hatches H1 and H2, respectively.

For Unit 2, recently revised calculation PM-1059, Revision 5 (Attachment 3) demonstrates that the resulting radiological dose consequences from post-FHA release paths, including the ground-level hatches, are within allowable regulatory dose limits after a fuel decay time of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

For Unit 3, recently revised calculation PM-1059, Revision 5 (Attachment 3), demonstrates that the resulting radiological dose consequences from post-FHA release paths are within allowable regulatory dose limits. Hatch H2 can be opened after a fuel decay time of 288 hours0.00333 days <br />0.08 hours <br />4.761905e-4 weeks <br />1.09584e-4 months <br />, and hatches H21, H22, and H34 can be opened after a fuel decay time of 312 hours0.00361 days <br />0.0867 hours <br />5.15873e-4 weeks <br />1.18716e-4 months <br /> and still allow Main Control Room (MCR) dose to be maintained below regulatory limits. Hatches H23 and H24 can be opened during irradiated fuel movement after a minimum fuel decay time of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

When hatches H19 and H20 were designated in the August 21, 2008 (Reference 1) submittal to the NRC, the focus was on the ground-level release at the 135' elevation, and therefore, hatches H19 and H20 were chosen. However, this is not technically correct since the area between hatches H1 and H19 for Unit 2, and between hatches H2 and H20 for Unit 3, is not part of the SC. Therefore, hatches HI and H2 should have been chosen as the boundary to be identified in the TS definitions of RECENTLY IRRADIATED FUEL and H19 and H20 should have been indicated as the release locations for corresponding releases from hatches H1 and H2.

Additionally, the wording "ground-level" is proposed to be removed from the TS definition since hatches H1 and H2 reside at the 116' elevation. Hatches H19 and H20 reside at the 135' elevation, which is at ground-level on the west-side of the RBs. Because hatches H1 and H2 are the appropriate SC hatches to reference, the term "ground-level" is not appropriate. The term is not required to be in the TS since it adds no other important information. Hatch designations and locations are controlled in accordance with the plant design documents and procedures.

4.0 TECHNICAL ANALYSIS

Of the four design basis radiological accidents contained in the PBAPS licensing basis for Alternate Source Term (AST), the only accident that is affected by the opening of the subject hatches is the FHA. The FHA design calculation PM-1059, Revision 5, "Re-Analysis of Fuel Handling Accident Using Alternative Source Terms" (Attachment 3) and supporting Calculation PM-1170, Revision 0, "PBAPS Atmospheric Dispersion Factors (X'Qs) for post-FHA Ground Hatch Releases" (Attachment 4) are included in this submittal.

Page 5 of 16 Methodoloqy PBAPS Amendments 269 (Unit 2) and 273 (Unit 3) (Reference 2) approved AST methodology for the FHA event. With its approval, the previous accident source term in the PBAPS design basis FHA was superseded by the AST. The previous offsite and MCR accident dose criteria expressed in terms of whole body, thyroid and skin doses were superseded by a fraction of the TEDE criteria of 10 CFR 50.67 (Reference 4), as defined in Regulatory Guide (RG) 1.183, Table 6 (Reference 3).

Consistent with the regulatory guidance in RG 1.183, Section 3.1 (Reference 3) for the Design Basis Accident (DBA) events that do not involve the degradation of the entire core, like the FHA, the fission product inventory of each of the damaged fuel rods is determined by dividing the total core inventory by the number of fuel rods in the core. To account for differences in power level across the core, a radial peaking factor is applied in determining the inventory of the worst-case damaged rods.

Because of radioactivity decay, the worst-case FHA is that associated with handling fuel that has recently been part of a critical core operating at full power immediately prior to the reactor shutdown.

The FHA release pathway through hatches H19 and H20 is conservatively used for the FHA dose evaluation for the corresponding releases from the HPCI Room hatches H1 and H2, respectively. The post-FHA doses due to releases through the ground-level hatches are calculated using newly developed sets of MCR X/Q values (Calculation PM-1 170, Revision 0 - ) for the as-built location of the MCR air intake and source term information (i.e.,

activity release rates to the environment) for PBAPS power levels 102% of 3,951 MWt. The post-FHA doses for some of these hatches bound the releases from the other surrounding hatches as follows:

Post-FHA doses from Unit 2 hatch H18 release bound the Unit 2 releases from hatches H17 and H33 due its shorter distance from MCR air intake.

Post-FHA doses from Unit 3 hatch H21 release bound the Unit 3 releases from hatches H22 and H34 due its shorter distance from MCR air intake.

Post-FHA doses from Unit 3 hatch H23 release bound the releases from Unit 3 hatch H24 and Unit 2 hatches H15 and H16 because:

1. Unit 3 hatch H23 is located at a shorter distance to the MCR intake than Unit 3 hatch H24.

2.

Unit 3 hatches H23 and H24 are mirror images of Unit 2 hatches H15 and H16.

3.

Unit 3 ground hatches are located in unfavorable wind sectors. Therefore, the set of X/Q values for hatch H23 conservatively bound the releases from Unit 2 hatches H15 and H16.

Page 6 of 16 Fuel Source Term To maximize the activity release during an FHA, it is conservative to model a fuel bundle type that maximizes both the number of damaged fuel bundles (i.e., maximizes the fraction of the core inventory released to the environment) and the radial Peaking Factor (PF).

FHA in the Reactor Vessel A survey of the various conditions that could exist when the drywell is open reveals that the greatest potential for the release of radioactive material occurs when the drywell head and reactor vessel head have been removed. In this case, radioactive material released as a result of fuel failure is available for transport directly to the secondary containment. It is concluded that the only accident that could result in the release of significant quantities of fission products to the secondary containment during this mode of operation is one resulting from the accidental dropping of a fuel bundle onto the top of the core.

A Decontamination Factor (DF) of 200 is assumed based on the guidance of RG 1.183.

For the FHA occurring in the reactor well, the source could exit the SC directly to the atmosphere through building openings or penetrations without mixing and diluting in the SC unless the SC integrity is maintained.

FHA in Refuelinq Pool When a FHA occurs in the Spent Fuel Pool (SFP), it is assumed that the dropped bundle will rest on the tops of the bail handles of the fuel in the storage racks. Therefore, for a drop over the SFP, coverage over a dropped assembly is slightly less than 23 feet, assuming that the dropped bundle is lying across the tops of fuel bundles within the spent fuel racks. There is greater than 23 feet of water above top of active fuel for bundles within the racks.

For the FHA occurring in the SFP, source term could exit the SC directly to the atmosphere through building openings or penetrations unless SC integrity is maintained.

Maximum Linear Heat Generation Rate (MLHGR)

Note 11 to Table 3 of RG 1.183 requires that the MLHGR does not exceed 6.3 kW/ft peak rod average power for burnups exceeding 54 GWD/MTU. The PBAPS fuel management program has determined that there will be no fuel assemblies being exposed to a maximum LHGR that exceeds 6.3 kW/ft at fuel burnups between 54 and 62 GWD/MTU.

RADTRAD Model The RADTRAD3.03 Code is used in this analysis. The same RADTRAD release models are used to model the FHA occurring in either the SFP or in the refueling cavity, both being inside the SC.

The RADTRAD model considers a source term volume of 100 cubic feet, which initially contains all of the activity that is released from the damaged spent fuel assemblies to the SC air space.

Any value of the source term volume can be used, which will result in the same Ci/sec release rate. This source term considers 2.009 damaged fuel assemblies (172 fuel pins) that have Page 7 of 16 decayed for a minimum of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, with a radial PF of 1.7, fuel rod gap release fractions per RG 1.183, Table 3, and pool (i.e., SFP or reactor cavity) water iodine, noble gas, and particulate DFs per RG 1.183.

A building release rate was calculated that will exhaust at least 99.9999% of this 100 cubic foot volume of the radioactive material to the environment over a 2-hour time period.

The FHA events analyzed use atmospheric dispersion factors modeled in previous revisions for the Exclusion Area Boundary (EAB), and Low Population Zone (LPZ). Since the SC integrity is not assumed to be maintained in the supporting calculation during the refueling outage, the post-FHA activity could release to the atmosphere via open doors and various containment hatches including the ground-hatches. The X/Q information for the hatches has been revised and replaced by the new sets of X/Q values established in PM-1 170, Revision 0 (Attachment 4) because of the change in the modeling of the actual locations of the releases and as-built location of the MCR air intake.

The assumption of inoperable SC integrity during the refueling outage results in the post-FHA releases through various hatches shown in Calculation PM-1059 (Attachment 3). The hatches are grouped together in a conservative manner to calculate the limiting sets of X/Q values for the as-built location of the MCR air intake. For the given fuel source term and MCR response, the MCR dose is proportional to X!Q values providing the dilution of post-FHA activity release from the containment. The different fuel decay times are used for the hatch releases due to varying severity of new set of X/Q values based on their locations from the MCR air intake and wind sector.

Key Assumptions RG 1.183, Appendix B (Reference 3) provides guidance on modeling assumptions that are acceptable to the NRC for the evaluation of the radiological consequences of a FHA. The following modeling assumptions were made for the FHA analysis.

Source Term Assumptions

1. Consistent with RG 1.183, Section 3.2 (Reference 3), the fractions of the core inventory assumed to be in the gap for the various radionuclides are as given in Table 3 of RG 1.183.

The release fractions from Table 3 are incorporated in conjunction with the core fission product inventory with the maximum core radial PF of 1.7 and with a proposed core thermal power level of 4,030 MWt. Fuel bundle peak burnup will not exceed the limits of RG 1.183, Footnote 11 (Reference 3).

2. Consistent with RG 1.183, Appendix B, Section 1.1 (Reference 3), the number of fuel rods damaged during the FHA is based on a conservative analysis that considers the most limiting case. All of the fuel rods in 2.009 spent fuel assemblies are assumed to be damaged. Additionally, it is assumed consistent with RG 1.183, Appendix B, Section 1.2 (Reference 3) that the fission product release from the breached fuel is based on the fission product inventory in the fuel rod gap (RG 1.183, Table 3 - Reference 3) and the estimated number of fuel rods breached (RG 1.183, Tables 1 and 2 - Reference 3).

Page 8 of 16

3. It is assumed that all the gap activity in the damaged rods is instantaneously released to the pool water (i.e., SFP or reactor cavity). The radionuclides included are xenons, kryptons, and iodines. Irradiated fuel shall not be removed from the reactor without SC integrity until the unit has been shutdown for at least 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Movement of recently irradiated fuel will not occur sooner than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after the fuel has occupied a critical reactor core. This value continues to be a very conservative assumption for BWRs, given the extensive operations necessary associated with reactor disassembly before commencing fuel movement. Longer fuel decay times are used with the post-FHA releases from two of the hatches due to their higher associated XJQ values and to bring the resulting doses within the allowable regulatory dose limits. The curie per megawatt-thermal inventory of fission products in the reactor core and available for gap release from damaged fuel is based on the core thermal power level of 4,030 MWt (102% of 3,951 MWt). The fission product inventory is based on the current fuel enrichment of 3.8 wt % (weight percent) and 4.2 wt % U-235, and a core average burnup of 36,471 MWD/MTU.

4. Consistent with RG 1.183, Section 3.3 (Reference 3), for non-Loss of Coolant Accident (LOCA) Design Basis Accidents (DBAs) in which fuel damage is projected, the release from the fuel gap is assumed to occur instantaneously with the onset of the projected damage.

5. Consistent with RG 1.183, Appendix B, Section 1.3 (Reference 3), the chemical form of radioiodine released from the fuel to the surrounding water should be assumed to be 95%

cesium iodide (Csl), 4.85 percent elemental iodine, and 0.15 percent organic iodine. The Csl released from the fuel is assumed to completely dissociate in the pool water (i.e., SFP of reactor cavity). Because of the low pH of the pool water, the iodine re-evolves as elemental iodine. The release to the pool water is assumed to occur instantaneously.

6. If the depth of water above the damaged fuel is 23 feet or greater, the overall effective DF for iodine is 200 (i.e., 99.5% of the total iodine released from the damaged rods is retained by the water) (RG 1.183, Appendix B, Section 2 - Reference 3). The DF of 200 is also applicable for water depths of as little as approximately 21 feet above damaged fuel in the PBAPS spent fuel pool. This iodine above the water is composed of 57% elemental and 43% organic species (RG 1.183, Appendix B, Section 2).

7. The retention of noble gases in the pool water is negligible (i.e., DF of 1). Particulate radionuclides are assumed to be retained by the pool water (i.e., infinite DF) (RG 1.183, Appendix B, Section 3 - Reference 3).

MCR Dose Consequences Assumptions RG 1.183, Section 4.2 (Reference 3) provides guidance to be used in determining the Total Effective Dose Equivalent (TEDE) for persons located in the MCR. The following assumptions were made for the PBAPS FHA analysis.

1. Consistent with RG 1.183, Section 4.2.1 (Reference 3), the MCR TEDE analysis should consider the following sources of radiation that will cause exposure to MCR personnel:

Contamination of the MCR atmosphere by the intake or infiltration of the radioactive material contained in the post-accident radioactive plume released from the facility (via MCR air intake).

Page 9 of 16 Contamination of the MCR atmosphere by the intake or infiltration of airborne radioactive material from areas and structures adjacent to the MCR envelope (via MCR unfiltered inleakage).

Radiation shine from the external radioactive plume released from the facility (external airborne cloud).

Radiation shine from radioactive material in the reactor containment (containment shine dose).

Radiation shine from radioactive material in systems and components inside or external to the MCR envelope (e.g., radioactive material buildup in recirculation filters) (MCR filter shine dose).

2. Consistent with RG 1.183, Section 4.2.3 (Reference 3), the models used to transport radioactive material into and through the MCR, and the shielding models used to determine radiation dose rates from external sources, should be structured to provide suitably conservative estimates of the exposure to MCR personnel. The radioactive material releases and radiation levels used in the MCR dose analysis are determined using the same source term, transport, and release assumptions used for determining the EAB and the LPZ TEDE values. These parameters do not result in non-conservative results for the MCR.

3. Consistent with RG 1.183, Section 4.2.6 (Reference 3), the MCR dose receptor is the hypothetical maximum exposed individual who is present in the MCR for 100% of the time during the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after the event, 60% of the time between one (1) and four (4) days, and 40% of the time from four (4) days to 30 days. For the duration of the event, the breathing rate of this individual should be assumed to be 3.5 x 10.4 cubic meters per second.

4. Consistent with RG 1.183, Section 4.4 and RG 1.183, Table 6 (Reference 3), the postulated MCR doses should not exceed the 5 Rem TEDE criterion established in 10 CFR 50.67 (Reference 4).

Offsite Dose Conseauences RG 1.183, Section 4.1 (Reference 3) provides guidance to be used in determining the TEDE for persons located at the EAB and at the outer boundary of the LPZ. The following assumptions were made for the PBAPS FHA analysis:

1. Consistent with RG 1.183, Section 4.1.1 (Reference 3), the dose calculation determines the TEDE, which is the sum of the Committed Effective Dose Equivalent (CEDE) from inhalation and the Effective Dose Equivalent (EDE) from external exposure; and these two components of the TEDE consider all radionuclides, including progeny from the decay of parent radionuclides that are significant with regard to dose consequences and the released radioactivity.

2. Consistent with RG 1.183, Section 4.1.2 (Reference 3), the exposure-to-CEDE factors for inhalation of radioactive material are derived from the data provided in International Commission on Radiological Protection (ICRP) Publication 30, "Limits for Intakes of Radionuclides by Workers." The supporting calculation models the CEDE Dose Conversion Factors (DCFs) in the column headed "effective" yield doses in Table 2.1 of Federal Page 10 of 16 Guidance Report 11, "Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion."

3. Consistent with RG 1.183, Section 4.1.4 (Reference 3), Table 111.1 of Federal Guidance Report 12, "External Exposure to Radionuclides in Air, Water, and Soil," provides external EDE conversion factors acceptable to the NRC.

4. Consistent with RG 1.183, Section 4.1.3 (Reference 3), for the first eight (8) hours, the breathing rate of person offsite is assumed to be 3.5 x 104 cubic meters per second. From eight (8) to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following the FHA, the breathing rate is assumed to be 1.8 x 104 cubic meters per second. After that and until the end of the FHA, the rate is assumed to be 2.3 x 10-4 cubic meters per second. The PBAPS supporting analysis conservatively models the EAB with only the conservative initial breathing rate of 3.5 x 10-4 cubic meters per second.

5. Consistent with RG 1.183, Section 4.1.7 (Reference 3), no correction is made for depletion of the effluent plume by deposition on the ground.

Acceptance Criteria The following NRC regulatory requirement and guidance documents are applicable to the Alternative Source Term FHA Calculation:

Regulatory Guide 1.183, Table 6 (Reference 3) 10 CFR 50.67 (Reference 4)

Standard Review Plan Section 15.0.1 (Reference 5)

Dose Acceptance Criteria are:

Regulatory Dose Limits Dose Type Control Room EAB and LPZ (rem TEDE)

(rem TEDE)

TEDE Dose 5

6.3 Results Summary:

The postulated post-FHA EAB, LPZ, and MCR doses are summarized in the following table:

Fuel Decay Post-FHA Dose Post-FHA Release Point Time EAB LPZ MCR (hrs)

(Rem TEDE)

(Rem TEDE)

(Rem TEDE)

Unit 2 Roof Scuttle Limiting Release Case Applicable to 24 2.99E+00 4.53E-01 4.30E+00 All Openings Except Hatches Listed Below Page 11 of 16 Fuel Decay Post-FHA Dose Post-FHA Release Point Time EAB LPZ MCR (hrs)

(Rem TEDE)

(Rem TEDE)

(Rem TEDE)

Hatches H17, H18 and H33 24 2.99E+00 4.53E-01 3.17E+00 Hatch Hi 24 2.99E+00 4.53E-01 3.75E+00 Hatch H2 288 8.04E-01 1.22E-01 4.52E+00 Hatches H21, H22, and H34 312 7.35E-01 1.11E-01 4.59E+00 Hatches H23, H24, H15, H16) 24 2.99E+00 4.53E-01 3.58E+00 Allowable Dose Limit (Rem TEDE) 6.3 6.3 5.0 The results of the analysis indicate that the EAB, LPZ, and MCR doses are within allowable regulatory dose limits for an FHA occurring either in the reactor vessel or the SFP without SC integrity with the limited hatches H15, H16, H17, H18, H1, H2, H21, H22, H23, H24, H33, and H34 remaining open during irradiated fuel movement after given fuel decay times.

5.0 REGULATORY ANALYSIS

5.1 No Significant Hazards Consideration Exelon has concluded that the proposed changes to the Peach Bottom Atomic Power Station, Units 2 and 3, Technical Specifications (TS) definition for RECENTLY IRRADIATED FUEL, as described in TS Section 1.1, "Definitions,"do not involve a Significant Hazards Consideration.

In support of this determination, an evaluation of each of the three (3) standards, set forth in 10 CFR 50.92, "Issuance of amendment," is provided below.

1.

Will operation of the facility in accordance with the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No.

The proposed changes to revise the PBAPS, Units 2 and 3, TS definition for RECENTLY IRRADIATED FUEL do not introduce new equipment or new equipment operating modes, nor do the proposed changes alter existing system relationships. The proposed changes do not affect plant operation, design function, or any analysis that verifies the capability of a Structure, System, or Component (SSC) to perform a design function. There are no changes or modifications to plant SSC. The plant Engineered Safety Features (ESFs) will continue to Page 12 of 16 function as designed in all modes of operation. There are no significant changes to procedures or training being introduced by the proposed changes to the TS definition.

Based upon the results of the FHA analysis, it has been demonstrated that, with the requested changes, the dose consequences remain within the regulatory guidance provided by the NRC as specified in 10 CFR 50.67 and associated Regulatory Guide (RG) 1.183. The calculations used to evaluate the consequences of the FHA accident in support of the proposed changes do not by themselves affect the plant response, but better represent the physical characteristics of the release, so that appropriate mitigation techniques may be applied. Therefore, the consequences of an accident previously evaluated are not significantly increased.

There is no adverse impact on systems designed to mitigate the consequences of accidents.

The proposed changes do not adversely affect system or component pressures, temperatures, or flowrates for systems designed to prevent accidents or mitigate the consequences of an accident. Since these conditions are not adversely affected, the likelihood of failure of SSC is not increased.

The proposed changes do not increase the likelihood of the malfunction of any SSC or impact any analyzed accident. Consequently, the probability or consequences of an accident previously evaluated are not affected.

Based on the above, Exelon concludes that the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.

2.

Will operation of the facility in accordance with the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No.

The proposed changes to revise the PBAPS, Units 2 and 3, TS definition for RECENTLY IRRADIATED FUEL do not alter the design function or operation of any SSC. There are no changes or modifications to plant SSC. The plant ESFs will continue to function as designed.

There is no new system component being installed, no new construction, and no performance of a new test or maintenance function. The proposed TS changes do not create the possibility of a new credible failure mechanism or malfunction. The proposed changes do not introduce new accident initiators or precursors of a new or different kind of accident. New equipment or personnel failure modes that might initiate a new type of accident are not created as a result of the proposed changes. SC integrity is not adversely impacted and radiological consequences from the analyzed FHA remain within specified regulatory limits. The proposed changes do not adversely impact system or component pressures, temperatures, or flowrates for systems designed to prevent accidents or mitigate the consequences of an accident. Since these conditions are not adversely impacted, the likelihood of failure of SSC is not increased.

Consequently, the proposed changes cannot create the possibility of a new or different kind of accident from any accident previously evaluated.

Based on the above, Exelon concludes that the proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.

Page 13 of 16

3.

Will operation of the facility in accordance with the proposed amendment involve a significant reduction in a margin of safety?

Response: No.

The proposed changes to revise the PBAPS, Units 2 and 3, TS definition for RECENTLY IRRADIATED FUEL do not alter the design function or operation of any SSC. There are no changes or modifications to plant SSC. The plant ESFs will continue to function as designed.

The proposed changes do not increase system or component pressures, temperatures, or flowrates for systems designed to prevent accidents or mitigate the consequences of an accident.

Safety margins and analytical conservatisms have been evaluated and have been found acceptable. The analyzed event has been evaluated and margin has been retained to ensure that the analysis adequately bounds the postulated FHA event. The dose consequences resulting from analyzing the FHA design basis accident comply with the requirements of 10 CFR 50.67 and the guidance of RG 1.183.

The proposed changes continue to ensure that the doses at the Exclusion Area Boundary (EAB) and Low Population Zone (LPZ) boundary, as well as the Main Control Room (MCR), remain within corresponding regulatory limits.

Based on the above, Exelon concludes that the proposed changes do not involve a significant reduction in a margin of safety.

Conclusion Based on the above evaluation of the three criteria, Exelon concludes that the proposed amendment presents no significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of "No Significant Hazards Consideration" is justified.

5.2 Applicable Regulatory Requirements/Criteria 10 CFR 50.67, 'Accident source term" (a) Applicability. The requirements of this section apply to all holders of operating licenses issued prior to January 10, 1997, and holders of renewed licenses under part 54 of this chapter whose initial operating license was issued prior to January 10, 1997, who seek to revise the current accident source term used in their design basis radiological analyses.

(b) Requirements. (1) 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 design basis accidents previously analyzed in the safety analysis report.

(2) The NRC may issue the amendment only if the applicant's analysis demonstrates with reasonable assurance that:

(i) 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 Page 14 of 16 receive a radiation dose in excess of 0.25 Sv (25 rem) total effective dose equivalent (TEDE).

(ii) 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) total effective dose equivalent (TEDE).

(iii) 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.

RG 1.183 provides assumptions and methods that are acceptable to the NRC for performing design basis radiological analyses using an AST. This guidance supersedes corresponding radiological analysis assumptions provided in the older regulatory guides and SRP chapters when used in conjunction with an approved AST and the TEDE criteria provided in 10 CFR 50.67.

Also, the NRC published a new SRP section to address AST. It is SRP Section 15.0.1, Revision 0, "Radiological Consequence Analyses Using Alternative Source Terms." It provides guidance on which NRC will review various aspects of an AST-related license amendment request, but otherwise is consistent with the guidance found in RG 1.183. The plant-specific information provided in this license amendment request adequately addresses the applicable guidance.

In conclusion, based on the considerations discussed above, (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 NRC'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.

6.0 ENVIRONMENTAL CONSIDERATION

The proposed amendments do not change a requirement with respect to the installation or use of a facility component located within the restricted area as defined in 10 CFR 20 and do not change surveillance requirements. The proposed amendments involve changes that will modify the definitions for RECENTLY IRRADIATED FUEL to: 1) revise the specific restriction identifying the SC hatches, and 2) address a discrepancy in the designation for identifying the SC hatch numbers. The proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluent that may be released offsite, or (iii) a significant increase in the individual or cumulative occupational radiation exposure. Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, in accordance with 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.

Page 15 of 16 7.0 PRECEDENT AST has been implemented at PBAPS as approved by the NRC (Reference 2). This approval included FHA design assumptions. The proposed changes in this submittal also correct a discrepancy in identifying specific hatch numbers and refine the FHA AST dose calculation.

8.0 REFERENCES

1. Letter from Pamela B. Cowan, Exelon Generation Company, LLC, to U.S. Nuclear Regulatory Commission - "Supplemental Response for License Amendment Request for Alternative Source Term," dated April 21, 2008 (ML082340796).

2. Letter from John D. Hughey, U.S. Nuclear Regulatory Commission to Charles G. Pardee, Exelon Generation Company, LLC - "Peach Bottom Atomic Power Station, Units 2 and 3 -

Issuance of Amendments RE: Application of Alternative Source Term Methodology," dated September 5, 2008 (ML082320406).

3. Regulatory Guide (RG) 1.183, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors," July 2000.

4. 10 CFR 50.67, Accident Source Term.

5. Standard Review Plan Section 15.0.1, Revision 0, "Radiological Consequence Analyses Using Alternative Source Terms."

6. Letter from Pamela B. Cowan, Exelon Generation Company, LLC, to U. S. Nuclear Regulatory Commission, "License Amendment Request - Application of Alternative Source Term," dated July 13, 2007 (ML072570156).

Peach Bottom Atomic Power Station, Units 2 and 3 NRC Docket Nos. 50-277 and 50-278 Revise Technical Specifications Definition for RECENTLY IRRADIATED FUEL Proposed Technical Specifications Unit 2 Unit 3 1.1-5 1.1-5

Definitions 1.1 1.1 Definitions PHYSICS TESTS (continued)

PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)

RATED THERMAL POWER (RTP)

b.

Authorized under the provisions of 10 CFR 50.59; or

c.

Otherwise approved by the Nuclear Regulatory Commission.

The PTLR is the unit-specific document that provides the reactor vessel pressure and temperature limits, including heatup and cooldown rates, for the current reactor vessel fluence period.

These pressure and temperature limits shall be determined for each fluence period in accordance with Specification 5.6.7.

RTP shall be a total reactor core heat transfer rate to the reactor coolant of 3514 MWt.

REACTOR PROTECTION SYSTEM (RPS)

RESPONSE TIME RECENTLY IRRADIATED FUEL The RPS RESPONSE TIME shall be that time interval from the opening of the sensor contact up to and including the opening of the trip actuator contacts.

RECENTLY IRRADIATED FUEL is fuel that has occupied part of a critical reactor core within the previous 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

When using this definition to suspend the ApplicabitY Of LGOS, secondary continmen.t aroAund level hatchcs HI5.

H16.

H17.

H18.

H!9.

and I I I

P1 -1

~~i; H A 4

[tHl (I

>L UU I

1 H -

Flu6 4 L I 'i i ijV '

lU i i i i '[Fi -.

I jI 1 1' 4 12Rr'Il I-I I'

Till' I 4A iAl'lIlrI4jFr' 6AA11 -I IrllII'L--

SHUTDOWN MARGIN (SDM)

SDM shall be the amount of reactivity by which the reactor is subcritical or would be subcritical assuming that:

a.

The reactor is xenon free;

b.

The moderator temperature is 68°F; and

c.

All control rods are fully inserted except for the single control rod of highest reactivity worth, which is assumed to be fully withdrawn.

With control rods not capable of being fully inserted, the reactivity worth of these control rods must be accounted for in the determination of SDM.

(continued)

PBAPS UNIT 2 1.1-5 Amendment No.

286

Definitions 1.1 1.1 Definitions PHYSICS TESTS (continued)

b.

Authorized under the provisions of 10 CFR 50.59; or PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)

RATED THERMAL POWER (RTP)

REACTOR PROTECTION SYSTEM (RPS)

RESPONSE TIME

c.

Otherwise approved by the Nuclear Regulatory Commission.

The PTLR is the unit-specific document that provides the reactor vessel pressure and temperature limits, including heatup and cooldown rates, for the current reactor vessel fluence period.

These pressure and temperature limits shall be determined for each fluence period in accordance with Specification 5.6.7.

RTP shall be a total reactor core heat transfer rate to the reactor coolant of 3514 MWt.

The RPS RESPONSE TIME shall be that time interval from the opening of the sensor contact up to and including the opening of the trip actuator contacts.

RECENTLY IRRADIATED FUEL SHUTDOWN MARGIN (SDM)

RECENTLY IRRADIATED FUEL is fuel that has occupied part of a critical reactor core within the previous 312-94 hours.

This 312-hour time period may be reduced to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> if secondary containment hatches H2,

H21, H22 and H34 are closed. When usiRg-thiS d..finitio to. SUpcnd the App!icability of

LCOs, SecoRdar..

co*nt.in.ent ground le...el hAtchS H29, H21, H22, H23, H24, an d H4 shall bh closed during the movement Of anR i4rradi4-ated fuel in SeconRdary Con-

.t-A4AnMcn-t-.

SDM shall be the amount of reactivity by which the reactor is subcritical or would be subcritical assuming that:

a.

The reactor is xenon free;

b.

The moderator temperature is 68 0 F; and

c.

All control rods are fully inserted except for the single control rod of highest reactivity worth, which is assumed to be fully withdrawn.

With control rods not capable of being fully inserted, the reactivity worth of these control rods must be accounted for in the determination of SDM.

(continued)

PBAPS UNIT 3 1.1-5 Amendment No.

289