Text

License Amendment Request to Administratively Clarify the Operating Licenses and Technical Specifications
	ML082120328
Person / Time
Site:	Byron, Braidwood
Issue date:	07/29/2008
From:	Simpson P; Exelon Generation Co, Exelon Nuclear
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	RS-08-087
	Download: ML082120328 (86)
	v • d • e

RS-08-087 July 29, 2008 www_exelcncorp.corn U. S. Nuclear Regulatory Commission ATTN : Document Control Desk Washington, DC 20555-0001 Braidwood Station, Units 1 and 2 Facility Operating License Nos. NPF-72 and NPF-77 NRC Docket Nos. STN 50-456 and STN 50-457 Byron Station, Units 1 and 2 Facility Operating License Nos. NPF-37 and NPF-66 NRC Docket Nos. STN 50-454 and STN 50-455 This request is subdivided as follows :

Exellon The proposed amendment has been reviewed by the Braidwood and Byron Station Plant Operations Review Committees and approved by the Nuclear Safety Review Board in accordance with the requirements of the EGC Quality Assurance Program.

Nuclear 10 CFR 50.90 Subject :

License Amendment Request to Administratively Clarify the Operating Licenses and Technical Specifications In accordance with 10 CFR 50.90, "Application for amendment of license or construction permit," Exelon Generation Company, LLC (EGC) is requesting an amendment to Facility Operating License Nos. NPF-72 and NPF-77 for Braidwood Station, Units 1 and 2, respectively, and NPF-37 and NPF-66 for Byron Station, Units 1 and 2, respectively. The proposed amendment removes time, cycle, or modification-related items from the Operating Licenses and Technical Specifications (TS) at both stations. Additionally, the proposed amendment proposes the correction of typographical errors introduced into the TS at both stations in a previous amendment. The time, cycle, or modification-related items have been implemented or superseded, are no longer applicable, and no longer need to be maintained in their associated Operating Licenses or Technical Specifications.

" provides an evaluation of the proposed changes.

" Attachments 2 and 3 include marked-up copies of the operating license, TS, and TS Bases pages with the proposed changes indicated for Braidwood and Byron Stations, respectively.

The TS Bases pages are provided for information only, and do not require NRC approval.

EGC requests approval of the proposed change by July 29, 2009, with the amendment being implemented within 60 days of issuance.

U. S. Nuclear Regulatory Commission July 29, 2008 Page 2 In accordance with 10 CFR 50.91, "Notice for public comment; State consultation," EGC is notifying the State of Illinois of this application for a change to the TS by sending a copy of this letter and its attachments to the designated State Official.

There are no regulatory commitments associated with the changes proposed by this request.

Should you have any questions about this letter, please contact Mr. Mitchel A. Mathews at (630) 657-2819.

1 declare under penalty of perjury that the foregoing is true and correct. Executed on the 29`h day of July 2008.

Patrick R. Simpson Manager - Licensing : Evaluation of Proposed Changes : Markup of Braidwood Station Proposed Operating License, Technical Specifications, and Technical Specifications Bases Changes Markup of Byron Station Proposed Operating License, Technical Specifications, and Technical Specifications Bases Changes

ATTACHMENT 1 Evaluation of Proposed Changes CONTENTS 1.0 DESCRIPTION

2.0 PROPOSED CHANGE

S

3.0 BACKGROUND

4.0 TECHNICAL ANALYSIS

5.0 REGULATORY ANALYSIS

5.1 No Significant Hazards Consideration 5.2 Applicable Regulatory Requirements/Criteria

6.0 ENVIRONMENTAL CONSIDERATION

7.0 REFERENCES Page 1 of 11

1.0 DESCRIPTION

2.0 PROPOSED CHANGE

S ATTACHMENT 1 Evaluation of Proposed Changes In accordance with 10 CFR 50.90, "Application for amendment of license or construction permit," Exelon Generation Company, LLC (EGC) is requesting an amendment to Facility Operating License Nos. NPF-72 and NPF-77 for Braidwood Station, Units 1 and 2, respectively, and NPF-37 and NPF-66 for Byron Station, Units 1 and 2, respectively. The proposed amendment removes time, cycle, or modification-related items and typographical errors from the Operating Licenses and Technical Specifications at both stations. The time, cycle, and modification-related items have been implemented or superseded, are no longer applicable, and no longer need to be maintained in their associated Operating Licenses or Technical Specifications.

Braidwood Station Unit 1 Facilitv Operatinq License Delete Unit 1 Facility Operating License condition (OLC) 2.C.4, "Initial Startup Test Program,"

since the license condition has been satisfied. This OLC required EGC to report any changes to the Initial Test Program described in Chapter 14 of the Final Safety Analysis Report (FSAR) within one month of such change. The initial startup test program is complete and no further changes can be made.

Delete Unit 1 Facility Operating License OLC 2.C.5, "Regulatory Guide 1.97 Revision 2 Compliance." Operating License OLC 2.C.5 required EGC to submit a report and schedule for implementation within six months of NRC approval of the Detailed Control Room Design Review (DCRDR). EGC proposes to delete this OLC since the action has been completed. EGC's Final Report for Regulatory Guide 1.97 compliance was transmitted via a letter from S. C.

Hunsader (Commonwealth Edison Company) to T. E. Murley (NRC) dated September 1, 1987 (Reference 1).

Delete an exemption from the requirements of Appendix J to 10 CFR 50 concerning containment air locks from section 2.D. The need for this exemption was eliminated following the issuance of Amendment 73 (10 CFR 50, Appendix J, Option B), which was issued on April 4, 1996 (Reference 2). This specific exemption no longer needs to be included in the Operating License.

Delete Attachment 1 to the Unit 1 Facility Operating License, "Work Items to be completed." refers to letters to the NRC from Commonwealth Edison Company (ComEd) detailing the operation and testing of the Auxiliary Building Ventilation (VA) System during Unit 1 startup and operation, and Unit 2 construction. Construction is complete on both units at Braidwood Station and the VA System is fully operational. Therefore, Attachment 1 no longer needs to be included in the Braidwood Station Unit 1 Facility Operating License.

Page 2 of 11

Braidwood Station Unit 2 Operating License ATTACHMENT 1 Evaluation of Proposed Changes Delete Unit 2 Facility Operating License OLC 2.C.4, "Initial Startup Test Program," since the OLC has been satisfied. This OLC required EGC to report any changes to the Initial Test Program described in Chapter 14 of the FSAR within one month of such change. The initial startup test program is complete and no further changes can be made.

Delete an exemption from the requirements of Appendix J to 10 CFR 50 concerning containment air locks from section 2.D. The need for this exemption was eliminated following the issuance of Amendment 73 (10 CFR 50, Appendix J, Option B), which was issued on April 4, 1996 (Reference 2). This specific exemption no longer needs to be included in this section of the Operating License.

Delete a time-related specific exemption regarding the requirements of 10 CFR 50.49(f) and 10 CFR 50.49(j) from section 2.D. This exemption was required until the startup following the Braidwood Station, Unit 2 surveillance outage that was scheduled in January 1989. This outage was completed as planned; therefore, this exemption is no longer required and can be removed from this section of the Operating License.

Delete the Unit 2 Facility Operating License Attachment 1, "Work Items to be completed." is listed as an attachment in the Unit 2 Facility Operating License; however, does not physically exist for Unit 2. Therefore, the reference to Attachment 1 can be deleted from the list of attachments in the Braidwood Station Unit 2 Facility Operating License.

Braidwood Station TS Table of Contents Revise the Braidwood Station TS Table of Contents. The proposed change revises the Table of Contents by deleting the listings of TS Tables and TS Figures from the Table of Contents. The Tables and Figures are not required to be listed separately in the Table of Contents according to NUREG-1431, Revision 3.0, "Standard Technical Specifications Westinghouse Plants," dated June 2004.

Braidwood Station TS 3.3.1. "FITS Instrumentation" Revise TS 3.3.1, "FITS Instrumentation," Condition N, Required Action Note. The proposed change corrects an editorial oversight in the nomenclature and structure of the Condition N Required Action Note. The change deletes the number of the Note (i.e., "1."), changes "Notes" to "Note", and removes an extra blank line in the Note. This editorial oversight was inadvertently introduced by Amendment 148 (Reference 3).

Page 3 of 11

ATTACHMENT 1 Evaluation of Proposed Changes Braidwood Station TS Surveillance Requirements (SR) 3.7.2.1 and SR 3.7.2.2 Revise TS 3.7.2, "Main Steam Isolation Valves," for Braidwood Station. EGC is requesting deletion of Note 2 from both SR 3.7.2.1 and SR 3.7.2.2, as the requirements imposed by these Notes which were added in Amendment 119 (Reference 4) have been implemented, the SRs are currently being met, and no longer need to be modified by these Notes.

Braidwood Station TS 3.7.8, "Essential Service Water (SX) System" Revise TS 3.7.8 for Braidwood Station. EGC is requesting the deletion of a Note from Condition A, and the deletion of Condition B in its entirety from TS 3.7.8. Maintenance activities (i.e.,

replacement of the SX pump suction isolation valves), as discussed in TS Amendment 130 (Reference 5), necessitated the addition of the Note in Condition A, and the insertion of Condition B. These maintenance activities were completed during Unit 2 Refueling Outage 11 as scheduled ; therefore, the Condition A Note, and Condition B in its entirety, no longer need to be maintained in TS 3.7.8. This revision will result in various editorial and formatting changes.

Braidwood Station Unit 1 and Unit 2 Facility Operating License Appendix C "Additional License Conditions" Revise the Unit 1 and Unit 2 Facility Operating License Appendix C to delete items that have been satisfied and are no longer required. Specifically, EGC is proposing the deletion of all Unit 1 and Unit 2 Operating License Appendix C items associated with License Amendments 98 (Reference 6) and 113 (Reference 7). The Conditions associated with Amendment 98 were implemented within 180 days of the issuance of Amendment 98 which was issued on December 22, 1998. The Conditions associated with Amendment 113 are related to power uprate activities and have been completed as documented in References 8, 9, 10, and 11. Therefore, these additional conditions have been satisfied and no longer need to be tracked in the Facility Operating License.

Page 4 of 11

Boron Station TS Table of Contents Bvron Station TS 3.3.1 "FITS Instrumentation" ATTACHMENT 1 Evaluation of Proposed Changes Revise the Byron Station TS Table of Contents. The proposed change revises the Table of Contents by deleting the listings of TS Tables and TS Figures from the Table of Contents. The Tables and Figures are not required to be listed separately in the Table of Contents according to NUREG-1431, Revision 3.0, "Standard Technical Specifications Westinghouse Plants," dated June 2004.

Revise TS 3.3.1, "FITS Instrumentation," Condition N, Required Action Note. The proposed change corrects an editorial oversight in the nomenclature and structure of the Condition N Required Action Note. The change deletes the number of the Note (i.e., "1."), changes "Notes" to "Note", and removes an extra blank line in the Note. This editorial oversight was inadvertently introduced by Amendment 153 (Reference 3).

Boron Station TS Surveillance Requirements (SR)3.7.2.1 and SR 3.7.2.2 Revise TS 3.7.2, "Main Steam Isolation Valves," for Byron Station. EGC is requesting deletion of Note 2 from both SR 3.7.2.1 and SR 3.7.2.2, as the requirements imposed by these Notes which were added in TS amendment 124 (Reference 4) have been implemented, the SRs are currently being met, and no longer need to be modified by these Notes.

Bvron Station TS 3.7.8, "Essential Service Water (SXLSvstem" Revise TS 3.7.8 for Byron Station. EGC is requesting the deletion of a Note from Condition A, and the deletion of Condition B in its entirety from TS 3.7.8. Maintenance activities (i.e.,

replacement of the SX pump suction isolation valves), as discussed in TS Amendment 136 (Reference 5), necessitated the addition of the Note in Condition A, and the insertion of Condition B. These maintenance activities were completed during Unit 1 Refueling Outage 13 as scheduled ; therefore, the Condition A Note, and Condition B in its entirety, no longer need to be maintained in TS 3.7.8. This revision will result in various editorial and formatting changes.

Page 5 of 11

ATTACHMENT 1 Evaluation of Proposed Changes Byron Station TS 3.7.15, TS 3.7.16 and TS 4.3.1 Revise TS 3.7.15, TS 3.7.16, and TS 4.3.1 at Byron Station. The proposed revision removes all references to Joseph Oat spent fuel pool storage racks in these TS and their associated Bases.

The Joseph Oat spent fuel storage racks have been replaced with Holtec spent fuel storage racks as of January 1, 2001. Since the Joseph Oat spent fuel storage racks are no longer installed at Byron Station, the removal of these references is editorial only and will delete wording that is no longer required. The revisions will result in various editorial and formatting requirements, and as a result of the proposed revision to TS 3.7.16, SR 3.7.16.3, and existing Figures 3.7.16-1, 3.7.16-2, and 3.7.16-3 will be deleted in their entirety. Figure 3.7.16-4 will be renumbered to Figure 3.7.16-1.

This change is not proposed for Braidwood Station, as the TS for Braidwood Station were previously amended to remove reference to Joseph Oat fuel storage racks under amendment 145 (Reference 12).

Byron Station Unit 1 and Unit 2 Facility Operating License Appendix C, "Additional License Conditions" Revise the Unit 1 and Unit 2 Facility Operating License Appendix C to delete items that have been satisfied and are no longer required. Specifically, EGC is proposing the deletion of all Unit 1 and Unit 2 Operating License Appendix C items associated with License Amendments 106 (Reference 6) and 119 (Reference 7). The Conditions associated with Amendment 106 were implemented within 180 days of the issuance of Amendment 106 which was issued on December 22, 1998. The Conditions associated with Amendment 119 are related to power uprate activities and have been completed as documented in References 11, 13, and 14.

Therefore, these additional conditions have been satisfied and no longer need to be tracked in the Facility Operating License.

3.0 BACKGROUND

Historically, conditions, exceptions, or exemptions that are date, cycle, or modification-related have been tracked in the Facility Operating License. Over time, the actions or requirements that these items contain are implemented, and the issues no longer need to be tracked in the Facility Operating License. In an effort to clarify, and avoid any confusion regarding the current requirements contained in the Operating License, EGC is proposing the administrative removal of these items from the Operating License and TS for Braidwood and Byron Stations.

Page 6 of 11

4.0 TECHNICAL ANALYSIS

All proposed changes are administrative in nature and require no technical analysis.

5.0 REGULATORY ANALYSIS

ATTACHMENT 1 Evaluation of Proposed Changes 5.1 No Significant Hazards Consideration In accordance with 10 CFR 50.90, "Application for amendment of license or construction permit," Exelon Generation Company, LLC (EGC) is requesting an amendment to Facility Operating License Nos. NPF-72 and NPF-77 for Braidwood Station, Units 1 and 2, respectively, and NPF-37 and NPF-66 for Byron Station, Units 1 and 2, respectively. The proposed amendment provides for the administrative removal of time, cycle, or modification-related items from the Operating Licenses at both stations. These items have been implemented or superseded, are no longer applicable, and therefore, no longer need to be maintained in their associated Operating License.

According to 10 CFR 50.92, "Issuance of amendment," paragraph (c), a proposed amendment to an operating license involves no significant hazards consideration if operation of the facility in accordance with the proposed amendment would not:

Involve a significant increase in the probability or consequences of an accident previously evaluated ; or Create the possibility of a new or different kind of accident from any accident previously evaluated ; or Involve a significant reduction in a margin of safety.

(1)

(2)

(3)

In support of this determination, an evaluation of each of the three criteria set forth in 10 CFR 50.92 is provided below regarding the proposed license amendment.

Page 7 of 11

ATTACHMENT 1 Evaluation of Proposed Changes 1.

Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?

Response : No The initial conditions and methodologies used in the accident analyses remain unchanged. The proposed changes do not change or alter the design assumptions for the systems or components used to mitigate the consequences of an accident. Therefore, accident analyses results are not impacted.

All changes proposed by EGC in this amendment request are administrative in nature, and are removing one-time requirements that have been satisfied or items that are no longer applicable. There are no physical changes to the facilities, nor any changes to the station operating procedures, limiting conditions for operation, or limiting safety system settings.

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

2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated?

Response : No None of the proposed changes affect the design or operation of any system, structure, or component in the plant. The safety functions of the related structures, systems, or components are not changed in any manner, nor is the reliability of any structure, system, or component reduced by the revised surveillance or testing requirements. The changes do not affect the manner by which the facility is operated and do not change any facility design feature, structure, system, or component. No new or different type of equipment will be installed. Since there is no change to the facility or operating procedures, and the safety functions and reliability of structures, systems, or components are not affected, the proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.

Based on this evaluation, the proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.

Page 8 of 11

Response : No ATTACHMENT 1 Evaluation of Proposed Changes

3. Does the proposed amendment involve a significant reduction in a margin of safety?

The proposed changes to the Facility Operating Licenses and TS are administrative in nature and have no impact on the margin of safety of any of the TS. There is no impact on safety limits or limiting safety system settings. The changes do not affect any plant safety parameters or setpoints. The Operating License Conditions have been satisfied as required. There are no changes to the conditions themselves.

Based on this evaluation, the proposed change does not involve a significant reduction in a margin of safety.

Therefore, EGC concludes that the proposed changes do not involve a significant hazards consideration under the criteria set forth in 10 CFR 50.92(c).

5.2 Applicable Regulatory Requirements/Criteria 10 CFR 50.36 details the information that must be included in each station's TS. The proposed changes modify or delete time, cycle, or modification-related items that have been implemented or superseded, and are no longer applicable. The proposed changes have no impact on current Safety Limits, Limiting Safety System Settings, Limiting Control Settings, Limiting Conditions for Operation, Surveillance Requirements, Design Features, or Administrative Controls. Therefore, EGC concludes that the methods used to comply with 10 CFR 50.36 are not modified by the proposed changes, and the requirements continue to be met.

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 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.

6.0 ENVIRONMENTAL CONSIDERATION

Portions of the proposed amendment change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, "Standards for protection against radiation," or would change an inspection or surveillance requirement.

However, 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 individual or cumulative occupational radiation exposure. Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in paragraph (c)(9) of 10 CFR 51.22, "Criterion for categorical exclusion; identification of licensing and regulatory actions eligible for categorical exclusion or otherwise not requiring environmental review." Therefore, in accordance with 10 CFR 51.22(b),

Page 9 of 11

ATTACHMENT 1 Evaluation of Proposed Changes no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.

Portions of the proposed amendment are confined to (i) changes to surety, insurance, and/or indemnity requirements, or (ii) changes to recordkeeping, reporting, or administrative procedures or requirements. Accordingly, these portions of the proposed amendment meet the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(10). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with these portions of the proposed amendment.

7.0 REFERENCES

1.

Letter from S. C. Hunsader (Commonwealth Edison Company) to U.S. NRC, "Byron Station Units 1 and 2 Braidwood Station Units 1 and 2 Regulatory Guide 1.97 Compliance Final Report," dated September 1, 1987 2.

Letter from R. R. Assa (U. S. NRC) to D. L. Farrar (Commonwealth Edison Company),

"Issuance of Amendments (TAC Nos. M94212, M94213, M94214, and M94215)," dated April 4, 1996

3. Letter from M. M. Thorpe-Kavanaugh (U. S. NRC) to Charles Pardee (EGC), "Byron Station, Units 1 and 2, and Braidwood Station, Units 1 and 2 Issuance of Amendments Re: Technical Specification Request to Extend Reactor Trip System and Engineered Safety Features Actuation System Completion Times, Bypass Test Times, and Surveillance Test Intervals (TAC Nos. MD4009, MD4010, MD4011, and MD4012),"

dated January 29, 2008

4.

Letter from Mahesh Chawla (U.S. NRC) to O. D. Kingsley (EGC), "Issuance of Amendments - Request for License Amendment for Technical Specification 3.7.2, "Main Steam Isolation Valves (MSIVs)(TAC Nos. MB3075, M133076, M133088, and MB3089),"

dated November 1, 2001 5.

Letter from G. F. Dick (U.S. NRC) to C. M. Crane (EGC), "Issuance of Amendments Re:

One-Time Change to the Completion Time for Restoration of a Unit Specific Essential Service Water Train (TAC Nos. MB9547, MB9548, MB9545, and MB9546)," dated March 18, 2004

6.

Letter from R. R. Assa (U. S. NRC) to O. D. Kingsley (Commonwealth Edison Company), "Issuance of Amendments (TAC Nos. M97546, M97547, M97548, and M97549)," dated December 22, 1998

7.

Letter from G. F. Dick (U. S. NRC) to O. D. Kingsley (EGC), "Issuance of Amendments ;

Increase in Reactor Power, Byron Station, Units 1 and 2, and Braidwood Station, Units 1 and 2 (TAC Nos. MA9428, MA9429, MA 9426, and MA9427)," dated May 4, 2001

8. Letter from Letter from J. D. von Suskil (EGC) to U. S. NRC, "Startup Report for Braidwood Station, Units 1 and 2 - Mid-Cycle Power Uprate," dated August 15, 2001 Page 10 of 11

ATTACHMENT 1 Evaluation of Proposed Changes

9.

Letter from Letter from J. D. von Suskil (EGC) to U.S. NRC, "Supplemental Startup Report for Braidwood Station, Unit 1 - Full Power Uprate Ascension," dated January 14, 2002 10. Letter from Letter from J. D. von Suskil (EGC) to U. S. NRC, "Supplemental Startup Report for Braidwood Station, Unit 2 - Full Power Uprate Power Ascension," dated August 13, 2002 11. Letter from G. F. Dick, Jr. (U. S. NRC) to J. L. Skolds (EGC), "Hot Leg Switchover Confirmatory Analysis - Braidwood Station, Units 1 and 2, and Byron Station, Units 1 and 2 (TAC Nos. MB5237, MB5238, MB5239, and MB5240)," dated September 27, 2002 12. Letter from R. F. Kuntz (U. S. NRC) to C. M. Crane (Exelon Generation Company, LLC (EGC)), "Braidwood Station, Units 1 and 2 - Issuance of Amendments re : Use of Areva NP Inc. Advanced Mark-BW(A) Fuel Assemblies (TAC Nos. MD3079 and MD3080),"

October 4, 2007

13. Letter from R. P. Lopriore (EGC) to U. S. NRC, "Startup Report for Byron Station, Units 1 and 2 - Mid-Cycle Power Uprate," dated August 8, 2001 14. Letter from R. P. Lopriore (EGC) to U. S. NRC, "Supplemental Startup Report for Byron Station, Units 1 and 2 - Mid-Cycle Power Uprate," dated November 5, 2001

15. Letter from G. F. Dick (U. S. NRC) to O. D. Kingsley (Commonwealth Edison Company),

"Byron and Braidwood - Issuance of Amendments on Spent Fuel Storage Racks," dated March 1, 2000 16. Letter from R. M. Krich (Commonwealth Edison Company) to U. S. NRC, "Request for an Amendment to Technical Specifications to Support Installation of New Spent Fuel Pool Storage Racks at Byron and Braidwood Stations," dated March 23, 1999 17. Letter from R. M. Krich (Commonwealth Edison Company) to U. S. NRC, "Response to Request for Additional and Clarifying Information Regarding Holtec International Report, HI-982083, 'Licensing Report for Spent Fuel Rack Installation at Byron and Braidwood Nuclear Stations,"' dated October 21, 1999 18. Letter from R. M. Krich (Commonwealth Edison Company) to U. S. NRC, "Editorial Correction to Technical Specification Amendment Request to Support Installation of New Spent Fuel Pool Storage Racks at Byron and Braidwood Stations," dated December 15, 1999 Page 1 1 of 11

ATTACHMENT 2 Markup of Braidwood Station Proposed Operating License, Technical Specifications, and Technical Specifications Bases Changes Braidwood Station, Units 1 and 2 Facility Operating License Nos. NPF-72 and NPF-77 License Amendment Request to Administratively Clarify the Operating Licenses and Technical Specifications for Braidwood Station and Byron Station

Any FSA report hanges t ade In In accod VI" the Initi

ordai ance Test Pr with th 5Q.59(b ram de rovision within gn ed In of 10 C 1 month ection 1 R 50.59 h ch of the all be ng~.

~N' if The li Implem isee sr ntation submit In six e final r the of rt and IC app schedu al of th for DCRDR.

(6)

Deleted.

The Additional Conefitions contained fn Appendix C, as,revised through Amendment Noj are hereby incorporated into this license. The' licensee shall operate the facility in accordance with the Additional Conditions.

(8)

Exelon Generation Company shall provide to the Director of the Office of Nuclear Reactor Regulation a copy of any application, at the time It Is filed, to transfer (excluding grants of security Interests or liens) from Exelon Generation Company to Its direct or indirect parent, or to any other affiliated company, facilities for the production, transmission, or distribution of electric energy having a depredated book value exceeding ten percent (10%) of Exelon Generation Company's consolidated net utility plant, as recorded on Exelon Generation Company's books of account.

Braidwood Unit 1

$154,273,345 (9)

Exelon Generation Company shall have decommissioning trust funds for Braldwood, Unit 1, in the following minimum amount, when Braldwood, Unit 1, Is transferred to Exelon Generation Company:

(10)

The decommissioning trust agreement for Braidwood, Unit 1, at the time the transfer of the unit to Exelon Generation Company Is effected and thereafter, Is subject to the following:

(a)

The decommissioning. trust agreement must be in a form acceptable to the NRC.

AMENDMENT NO.

Deleted

~

Deleted

-4b-4.

Procedures for implementing integrated fire response strategy

5.

Identification of readily-available pre-staged equipment G.

Training on integrated fire response strategy

7.

Spent fuel pool mitigation measures Actions to minimize release to include consideration of:

1.

Water spray scrubbing

2.

Dose to onsite responders is by low, will not present an w M* to the public health-and safes, end is-wneistne An exemption was previously granted pursuant to 10 CFR 70.24. The exemption was granted with NRC materials license No. SNM-1938, issued October 8, 1985, and relieved the licensee from the requirement of having a criticality alarm system.

Therefore, the licensee is exempted from the criticality alarm system provision of 10 CFR 70.24 so far as this section applies to the storage of fuel assemblies held under this license.

E.

The licensee shall implement and maintain in effect all provisions of the approved fire protection program as described in the Final Safety Analysis Report, as supplemented and amended, and as approved in the SER dated November 1983 and its supplements, subject to the following provision:

The licensee may make changes to the approved fire protection program without prior approval of the Commission, only if those changes would not adversely affect the ability to achieve and maintain safe shutdown in the event of a fire.

Amendment No.(j1%

Revised by letter dated June 27, 2007

F.

Exelon Generation Company shall fully implement and maintain in effect all provisions of the Commission-approved physical security, training and qualifications, 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 50354(p). The combined set of plans',

which contain Safeguards Information protected under 10 CRF 73.21, is entitled:

"Braidwood Station Security Plan, Training and Oualdication Plan, and Safeguards Contingency Plan, Revision 3," submitted by letter dated May 17,

2006, G.

Deleted H.

The licensee shall have and maintain financial protection of such type and in such amounts as the Commission shall require in accordance with Section 170 of the Atomic Energy Act of 1954, as amended, to cover public liability claims.

I.

This license is effective as of the date of issuance and shah expire at midnight on October 17.2026.

Date of Issuance : July 2, 1987 FOR THE NUCLEAR REGULATORY COMMSSION original signed by :

Thomas E. Mudey, Director Office of Nuclear Reactor RegWation Attachments:

Appendix A - Technical Specifications (NUREG-1278)

Appendix 8 - Environmental Protection Plan Appendix C - Additional Conditions

'The training and Qualification Plan and Safeguards Contingency Plan are Appendices to the Security Plan Amendment No Revised by letter dated February 1,

Thi sati The i conduc 1987. 51 E. Murlt 1986.

atta faction i erim ope in acc C. Huns r dated identif accorda tion pla dance wi to Th 11.

1 (es speci e with for the

h 1 etter s E. Mur

. and A.

ATT c item operati xi N ary

,s. C dated J D. most ich must a ) modes iiding V ader to 23. 1 o Harold couplet s identi tilation

. Bert 7. 5. C.

. Denton to the ied beli 1(VA) Syst is dated sader t dated omission wi 11 be une 26.

Thaws st 26.

In Any' FSA report est P hanges made In d In acc the In ccordari cdance v ra i Test Pr e with th 50.59(

ram de provisl within o ribed in sof 10C month ection 1 R 50.59 f such 4 of the hall nge.

Deleted (5)

Deleted.

(6)

Additional Conditions The Additional Co Amendment No licensee shall opera Conditions.

ditions contained In Appendix C, as revised through are hereby Incorporated into this license. The e the facility in accordance with the Additional Exelon Generation Company, LLC, shall piovide the Director of the Office of Nuclear Reactor Regulation, a copy of any application, at the time It is filed, to transfer (excluduutig granti of, security interests or liens) from Exelon Generation Company, LLC to Its direct or Indirect parent, or to any other affiliated company, facaRitjes for the production, transmission, or distribution of electric energy having a depreciated took value exceeding ten percent (10°/a) of Exelon Generation Company, LLC's consolidated net utility plant, as recorded on Exelon Generation Company, LLCs books of account.

(8)

Exelon Generation Company, LLC, shall have decommissioning trust funds for Braldwood, Unit 2, in the following minimum amount, when Braldwood, Unit 2, Is transferred to Exelon Generation Company, LLC :

Braidwood Unit 2

$154,448,967 The decommissioning trust agreement for Srafdwood, Unit 2, at the time the transfer of the unit to Exelon Generation Company, LLC is effected and thereafter, is subject to the following:

The decommissioning trust agreement must be in a form acceptable to the NRC.

(b)

With respect to the decommissioning trust fund, investments in the securities or other obligations of Exelon Corporation or affiliates thereof, or their successors or assigns are prohibited.

Except for Investments tied to market indexes or other non-nuclear sector mutual funds, investments in any entity owning one or more nuclear power plants are prohibited.

AMENDMENT NO.

(c)

Actions to minimize release to include consideration of.

1.

Water spray scrubbing 2.

Dose to onsite responders 4b-by law; w41 net presvAt an-w*due-4;ish to-tho public t}ealtla- -A -afetyL, aad- ~

An exemption was previously granted pursuant to 10 CFR 70.24. The exemption was granted with NRC materials license No

. SNM-1938, issued October 8, 1985, and relieved the licensee from the requirement of having a criticality alarm system. Therefore, the licensee is exempted from the criticality alarm system provision of 10 CFR 70.24 so far as this section applies to the storage of fuel assemblies held under this license.

50.

Title Each shall e aragra action item following Testin similar to be qu vesting a t the eq uires a

.49(}).

f facility 9(f) and 0 of the der of ablish a (b) in

.49(f) of electric ethods :

an identi nditions ified is a imilar its ipment to empor exempti license i the e to safety underi alysis to support able.

r u.c i+wr-.i!ra each ap' rogram (

is sftl CFR uipme I item o ith a su

eptable.

of equi e qualift scant f qualifyi states:

import ding a ant with is a from th 10 CFR) o operat c equip ust be ntical how that 9 analy require art 50.4 a nudea ant de ualifted b ditions o equip to sho ents of 1 (a) state power d in one of under ant CFR nt Amendment No.

Revised by letter dated June 27, 1

90'J 70101 4a ~~d 3l ArfOJ.VlfUlr Ot Irb-~1 E.

Tied licensee, shall iopleront nod eaintals to effect all PM1stens of the app194 ~ fire protectloo program as dasembed 1s the Final Safety Anal ysis Aeport. as supplawooted mod aromded. sad as approved is the SO fated fvaaber IOU and Its supplsneaf, subject to the follorfM pfovistons wlD "N 16-9F9~~

P~~~~~q Sac life.

tioo l A dis 3.11 a This is pu"1,

01:41 nod of the*

detest

.1t. T (ls7 This e llcanse is Seat (SSER 6?

undue r s do

crlpti, uest ai

,10CFR9 it.

b as suppo foe S0.4 horeby slam and the Safi is t health kslrcl that staffe s 1111%).

s the ant

,(j) of 10 snt fd an ova matt Evt l as amthori 4d Safe PA use Nat cl 1

yttcal a CFII :0 s tnclu of ten Ion by

. and is relatt to 1 as suaptt test as part exsoptt~

1 11 sued the to tltf do axis Was

~ ape cof f this llo.6 Preseat With 1team' aseelarti Is reglii publt uslole.

road ti d

sign I 4i t

~~t the licensee may arks changes to the approved fire pretmietido program without prier $""val of the Cawaiselso, oel 1f ties" changes tumid W tdversriy affect the ability to achieve sed maintain safe shutdoon is the event of 4 fire.

A pare ash f or+e instal ad of l i f lcati

, tncl doc nutioo (a)

M this s

t1on, tbeW low tlio ire rt d dwrir~

ich the covered to tiK w1' ~

p la or is ored to an for Mod A audi ble is fours u f to permit erificat on that Item f e1 c equ" tepo to safe cow E by this toe:

(1) is ltfled or its a 1 icstt am

2) Neat its spa fled ps onssma qul reawi 1

11 sub) tad to t conditt $ predi tad to present It must rfa~ t safety unction to the rod of i qualifI 3.

Es Hence w h idantl 1 or 0 far equi t un cc" ctoes wi a SVppo iM one sis to J that aqui Fit to gw1M is &cct table.

4 Analy s to lnation th porn 1 type t dots T

eaaiptl to requ until 1tdw1og no Brat lint Z suet Van" ou age sch sum Mary 191 War whi ties 1n unqua l l I1 Rasp I

talelsat

.4 1 ass scion have qwi fled or 1"af d

ones Filch ha been pro tousl ties h

ceasing torte a Mta te the fa lily.

Exelon Generation Company shall fully implement and maintain in effect all provisions of the Commission-approved physical security, training and qualifications, 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 50354(p). The combined set of plans',

which contain Safeguards Information protected under 10 CRF 73.21, is entitled:

"Braidwood Station Security Plan, Training and Oualitication Plan, and Safeguards Contingency Plan, Revision 3," submitted by letter dated May 17, 2006.

G.

Deleted H.

The licensee shall have and maintain financial protection of such type and in such amounts as the Commission shah require in accordance with Section 170 of the Atomic Energy Act of 1954, as amended, to cover public liability claims.

This license is effective as of the date of issuance and shall expire at midnight on December 18, 2027.

Attachments:

a<.

Wa*4tems to be-sonpfeted Appendix A - Technical Specifications (NUREG-1278)

Appendix 13-Environmental Protection Plan Appendix C - Additional Conditions Date of Issuance : May 20, 1988 FOR THE NUCLEAR REGULATORY COMA41SSION orOnal signed by:

James H. Sniezek, Deputy Director Office of Nuclear Reactor Regulation

'The training and Qualification Plan and Safeguards Contingency Plan are Appendices to the Security Plan Amendment No _

Revised by letter dated February 1, 2007

MODES.....

.. 1.1-9 Reactor Trip System Instrumentation

... 3.3 Engineered Safety Feature Actuation System Instrumentation....

Post Accident Monitoring Instrumentation.....

Remote Shutdown Monitoring Instrumentation....

Containment Ventilation Isolation Instrumentati VC Filtration System Actuation Instrumentatio FHB Ventilation System Actuation Instrument OPERABLE Main Steam Safety Valves versus Applicable Power in Percent of RATED POWER.

~n Steam Safety Valve Lift Settings ry Cell Parameters Requirements

Reactor Core Safety Limits...

. 2.0 Moderator Temperature Coefficient vs. Power Leve3.

Reactor Coolant DOSE EQUIVALENT I-131 Specific Activity Limit Versus Percent of RATED THERMAL POWER....

Seal Injection Flow Limits Region 2 Fuel Assembly BurnupRequirements...

ACTIONS (continued)

REQUIRED ACTION

NOTEX-____--____-

One train may be bypassed for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for surveillance testing, provided the other train is OPERABLE.

CONDITION N.

One RTB train inoperable.

0.

One or more channels inoperable.

BRAIDWOOD - UNITS 1 & 2 3.3.1 - 6 RTS Instrumentation 3.3.1 COMPLETION TIME (continued)

Amendment N.1 Restore train to OPERABLE status.

24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months OR N.2 Be in MODE 3.

30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months 0.1 Verify interlock is 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months in required state for existing unit conditions.

OR 0.2 Be in MODE 3.

7 hours8.101852e-5 days 0.00194 hours 1.157407e-5 weeks 2.6635e-6 months

SURVEILLANCE REQUIRBEVIS SR 3.7.2.2

NQIE5( ------------------

'1C required to be performed in MS 1 and 2.

Verify each MSIV actuates to the isolation position on an actual or simulated actuation signal MSIVs 3.7.2 In accordance with the Inservice Testing Program BRAIDWOOD - UNITS 1 & 2 3.7.2 - 2 Amendment

BASES SURVEILLANCE SR 3.7.2.1 REQUIREMENTS This SR verifies that MSIV closure time is <_ 5 seconds.

The MSIV closure time is assumed in the accident and containment analyses.

This Surveillance is normally performed upon returning the unit to operation following a refueling outage.

Based on ASME Code Section XI (Ref. 5), the MSIVs are not closure time tested at power.

The Frequency is in accordance with the Inservice Testing Program.

This test is conducted in MODE 3 with the unit at ure and pressure.

This SR is modified by allows entry into and operation in MODE 3 1 prior o per orming the SR.

This allows a delay of testing until MODE 3, to establish conditions consistent with_ those under which the acceptance criterion was qenerated.

r SR 3.7.2.2 This SR verifies that each MSIV can close on an actual or simulated actuation signal.

This Surveillance is normally performed upon returning the unit to operation following a refueling outage.

The frequency of MSIV testing is every 18 months.

The 18 month Frequency for testing is based on the refueling cycle.

Operating experience has shown that these components usually pass the Surveillance when performed at the 18 month Frequency.

Therefore, this a., ,tn T~ ;n R,,.+n Frequency i s acceptable from a reliability standpoint.

Ea Note a him avvtc This SR is modified b'y t e

allows entry into and operation in MODE 3 prior to performing the SR.

This allows a delay of testing until MODE 3, to establish conditions consistent with those under which the acceptance 1

11 was enerated.

r REFERENCES 1.

.)F AR, Secti,~

F, X0_3.

lIFSAR, 'Secti L5 _ _. 5 iF AR, Sect o 4.

O CFR 50.07.

5 -

AYE, Bcil&- and P' BRAIDWOOD - UNITS 1 & 2 B 3.7.2 - 6 insert ssel 'C'ode, Se-t.t 4; on X I.

Revision MSIVs B 3.7.2

3.7 PLANT SYSTEMS 3.7.8 Essential Service Water (SX) System APPLICABILITY :

MODES 1, 2, 3, and 4.

ACTIONS A.

CONDITION One unit-specific SX train inoperable.

BRAIDWOOD - UNITS 1 & 2 3.7.8 - 1

NOTES --------

1.

Enter applicable Conditions and Required Actions of LCO 3.8.1, "AC Sources-Operating," for Emergency Diesel Generator made inoperable by SX.

2.

Enter applicable Conditions and Required Actions of LCO 3.4.6, "RCS Loops-MODE 4," for Residual Heat Removal loops made inoperable by SX.

Restore unit-specific 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months SX train to OPERABLE status.

Amendment SX System 3.7.8 LCO 3.7.8 The following SX trains shall be OPERABLE :

a.

Two unit-specific SX trains ; and b.

One opposite-unit SX train for unit-specific support.

REQUIRED ACTION I COMPLETION TIME (continued)

B.

ACTIONS (continued)

SX System 3.7.8 BRAIDWOOD - UNITS 1 & 2 3.7.8 - 2 Amendment CONDITION REQUIRED ACTION COMPLETION TIME B.

NOT ------ -

B.1

NOTES- ------

Only pplica le to 1

Ente appli able Unit 1 durin Cond tions d

replac ment o the S Requi ed Act'ons uction isoIa. on of LC 3.8.1, "AC alves uring nit 2 Source -

fuelin 11 wh'le perati g," f U it 2 i in MO E 5, mergen y Dies 1 6, or def elect.

nerato made i operab e by S.

One nit-s cific X 2.

En er app icable trai inope able.

Con itions and Req fired A tions of L 0 3.4.

"RCS Loops-MODE

," for Resid 1 Heat Remova loops ade i perabl SX.

Restor unit-pecif.

144 ours X trai to 0 RABLE tatus.

Opposite-unit SX train Restore opposite-unit 7 days inoperable.

SX train to OPERABLE B.1 status.

Required Action and

,~

Be in MODE 3.

6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion i

C. i Time of ~_on i Vii_

!AND

!i.~

e ___

__ _ 1 C.2

.~,

A or B not met.

~- we i

MODE 5.

36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months

BASES APPLICABILITY In MODES 1, 2, 3, and 4, the unit-specific SX System is a normally operating system that is required to support the OPERABILITY of the equipment serviced by the SX System and required to be OPERABLE in these MODES.

ACTIONS A.1 While a specific unit is in MODES l, 2, 3, or 4, the opposite-unit SX System must be available (independent of the opposite unit's MODE or condition) for unit-specific support.

This minimizes the risk associated with loss of all unit-specific SX.

In MODES 5 and 6 the OPERABILITY requirements of the unit-specific SX System are determined by the systems it supports and there are no opposite-unit SX System requirements.

BRAIDWOOD - UNITS 1 & 2 B 3.7.8 - 4 SX System B 3.7.8 If one unit-specific SX train is inoperable, action must be taken to restore OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

In this Condition, the remaining OPERABLE SX train is adequate to perform the heat removal function.

However, the overall reliability is reduced because a single failure in the OPERABLE SX train could result in loss of the SX System function in the short term.

The 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Completion Time is based on the redundant capabilities afforded by the OPERABLE train, and the low probability of a DBA occurring during this time period.

C Co th whi ndit dit SX eU ion A ni ucti at 2 is not ni is i odif appl' o] at MOD ed b icab on v 5,

a e to Ives 6, o to Unit dur de hat ld ng U ele indi ing it tes' rep]

Ref this cem Iin to 11 Required Action A.1 is modified by two Notes.

The first Note indicates that the applicable Conditions and Required Actions of LCO 3.8.1, "AC Sources-Operating," should be entered if an inoperable SX train results in an inoperable emergency diesel generator.

The second Note indicates that the applicable Conditions and Required Actions of LCO 3.4.6, "RCS Loops-MODE 4," should be entered if an inoperable SX train results in an inoperable decay heat removal train.

These are exceptions to LCO 3.0.6 and ensure the proper actions are taken for these components.

Revision

BASES ACTIONS (continued)

C.1 and C.2 SX System B 3.7.8 B'

Du Uni inop with risk-risk ccept nditi C nditio th SX s wh le Uni RegAired A Actin n A.1 ng re 2 Ref

rable, n 144 informe ith th ble.

nBis' is on tion i 2 is tion B :

bove.

laceme eling action urs.

asses units odifi y app]

olatid n MODE is m of th 1, if must b This C ment t in the by a cable valve 5, 6, ified SX su e uni taken pleti at con pecifi ote tri o Unit durin r defu y two tion i c speci to res Time luded d conf t indi 1 duri Unit led.

otes a olatio jc SX ore OP is base hat th (gurati ates t repl Refue descr valve ain is RABLE upon associ n is is ement ing 11 ibed in duri n

,tatus ted f

If the opposite-unit SX train is not OPERABLE for unit-specific support, action must be taken to restore OPERABLE status within 7 days.

In this Condition, if a complete loss of unit-specific SX were to occur, the SX System function would be lost.

The 7 day Completion Time is based on the capabilities of the unit-specific SX System and the low probability of a DBA with a loss of all unit-specific SX occurring during this time period.

If the unit-specific SX train or the opposite-unit SX train cannot be restored to OPERABLE status within the associated Completion Time, the unit must be placed in a MODE in which the LCO does not apply.

To achieve this status, the unit must be placed in at least MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in MODE 5 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging plant systems.

BRAIDWOOD - UNITS 1 & 2 B 3.7.8 - 5 Revision

APPENDIX C ADDITIONAL CONDITIONS FACILITY OPERATING_ LICENSE NO. NPF-72 The licensee shall comply with the following conditions on the schedules noted below :

Amendment implementation 8

9 ai F

A Jul Se Oct Nov Dece taffs men 9

Th req,

con ame thes,

s de eve cto b d Jai brua il 16,'

31, emb er 1!

ber ber afe ent.

a b

th be pe 98.

whos reduc

egin, rfor lice irem

'oiled dme requi scribe ber 1 10,,

uary 13,

' une ugust 25, Oct 9, N 19c Eval For Ope is du hat b

=or S' cludi id SR ing e end ins o rme

-or S inter d, th

upon ed a ee is nts in ocu t shal emen in th 199 cto

,7, Ja

~brua Jun 11, A ctob er 2

+embi

,an tion u

Rs frog at th gins

,s that Ig SR who ende the the d prior s that also first r ompl sr im autho lude ents.

inclu lstot liven Feb 28, ary 24, 2, Ju gust 1, O Nov

~r 23, eval ssoci at ar icen end t imp

,exist with e inte the st su to th imp

!xist perfo aduce tion emen ized t in Ap mple e the Lapp ee's I uary d De Fe

ebru, 2, J

., Se tober mber ove ted i ited 1 new i NP f the men prio od if als rst pe eillai SUN ment 1pho mane surv f the ation relo endix ientat utial opriat tters 4, Se emb uary ry 26, ly 8, J temb Oct ber 3 the th thi Ame 72, th rst su bon to A d ace, perf form

e int illanc tion to A are illanc

t su f Am ate ce Atoli n of 1ocat doc ated temb 8, 1

,pril ly 30 ;

r21, ber 5!

and C

dme first feilla Ame sndm Iptanc nanc ice is rval t was I Ame

sndm, ing inte eilla ndme in ense is n of ent r 2, 7,

3, 98 to erfo e int, dme

,nt 98,1 crite are

'ue a t

st dme Int 98 ;

I e

it is A

Facili ance rval 98.

hall ith i n uanc lend wit iss Am impl 0 da of gent 98 II be in 18 nce dme ment afte mpler~

days f

t 98 Amendment No

ADDITIONAL CONDITIONS FACILITY OP ERATING LICENSE NO. NPF-72 The licensee shall comply with the following conditions on the schedules noted below:

Amendment Implementation Number Additional Condition Date 145 The safety limit equation specified in TS 2.1.1.3 regarding fuel centerline melt temperature (i.e., less than 5080 °F, decreasing by 58 °F per 10,000 MWD/MTU burnup as described in WCAP-12610-P-A, "VANTAGE+ Fuel Assembly Reference Core Report,"

April 1995) is valid for uranium oxide fuel without the presence of poisons mixed homogeneously into the fuel pellets.

If fuel pellets incorporating homogeneous poisons are used, the topical report documenting the fuel centerline melt temperature basis must be reviewed and approved by the NRC and referenced in this license condition. TS 2.1.1.3 must be modified to also include the fuel centerline melt temperature limit for the fuel with homogeneous poison. During operation in Cycles 15, 16, and 17, up to eight (8) AREVA NP Advanced Mark-BW(A) fuel assemblies containing fuel pellets incorporating homogeneous poisons may be placed in nonlimiting Unit 1 core locations provided the fuel cycle designs are developed such that the TS 2.1.1.3 Safety Limit equation for Westinghouse fuel is bounding. The design basis for the AREVA NP fuel rod centerline melt follows that given in BAW-10162P-A, "TAC03 - Fuel Pin Thermal Analysis Computer Code," October 1989, and BAW-10184P-A, "GDTACO - Urania Gadolinia Fuel Pin Thermal Analysis Code," February 1995.

With imple-mentation of the amend-ment AMENDMENT NO.

APPENDIX C ADDITIONAL CONDITIONS FACILITY OPERATING LICENSE NO. NPF-77 The licensee shall comply with the following conditions on the schedules noted below :

Amendment Implementation Number Additional Condition Date a

F Af Jul Sep' Octd Nove Dece taff s mend ecei ctobe d Jan' rua il 16, 31, A embe' er 1 ;

ber ber 1 afe ent.

er 1 10, ary 13, tune T gust 25, Octo 9, No 191 Eval 199 ctobe Jan'

~brua Jun 1, A

!ctobe er 2 eml an tion Feb 28, a jary 2 24, 2, Ju ust 1, O Nov~

Er 23, eval ssoci ary id De Fe ebru<

2, J

., Se tober mber ove ted i ited 1

4, Sel emb ary ry 26, ly 8, J temb Oct ber 3 the ith thi temb 8, 1 pril ly 30 ;

r21, ber 5 and C

r 2, 7,

Amendment No.

9 For, Rs th t are, ew i Ame dmen' 98 to 'acili S

11 be' mple ente Ope ting icens NPF,77, th first erfor ance wit in 18 days fter is du at th end the st su eilla e int ~rval iss nce f hat b gins impl ;ment tion Ame dmen 98.

Am idme t 98 or SR that ~xiste prior to A ndm nt 98, it cludii SRs' with odif acc tanc chte a a d SR who inte als o perffi anc are be ng ex nde the t pe forma ce is 'ue at the nd o the -fl st su eillan e inb al t

begl is on, he d e the' ;urve Ilanc was I st perf, rmed rior t impl ment tion Amei dmen 98.

or S that xiste prio ~to A ;ndm nt 98, whos inte gals o erfo ianc are ing educ d, the first r duce surv illanc inte I

egins pon ompl tion `the st sui reillan e p rfor

~d aft ~r imp men tion f Am dme t 9

8 Th licen ee is' iutho ized t relo to ce ain hall b impl ment req irem is in iuded in Ap, endix Atoli sense ithin 1' 0 da afte con tolled ocu ents.

pie ienta n of is is uanc of ame dme shall' inclu the iitial docat n of A endi :nt 98 these requi ;men to th app ppriat doc ents s de cribe in th licen ee's I tters ated ADDITIONAL CONDITIONS FACILITY OPERATING LICENSE NO. NPF-77 The licensee shall comply with the following conditions on the schedules noted below:

Amendment Implementation Number Additional Condition Date T

as T

ant the leg i Eval ime u e lice descr h,

dis to m<

rid i acku lice' lysis alue iecti tion

ing t see ed i lice rise intaie cludi time see sing f 8.5 foll ecti e cur hall Sec ee s in S the ar setti hall mo hour?

wing 3.1' entty ake ion 4.

all i tion abilit ducti

s.

bmit el ac for t loss

) ; or' cce e in 5.2 lem

.11.

of th

~n in t to th epta e tim of-co' ecal eted trum the nt m of th Brai e exi NRC e to of s%

lant late

tho tatio

afet, ifica Safe woo ting co eN itcho ccid he s logy' chaff Eval ions Ev tran cal b irma C jus er to' t (S itcho ges tion' s

uatio miss reake ry ifyin of ety er m

ful rat n

S J

for t ntati ow con Pr me full rate bmit

ie 1, imple n of up-itions rtoi tatio

,ower onditi y

002 ple-'

of p-ns 122 The safety limit equation specified in TS 2.1.1.3 With imple-regarding fuel centerline melt temperature (i.e., less mentation of than 5080 °F, decreasing by 58 °F per 10,000 the amend MWD/MTU burnup as described in WCAP-12610-P-A, ment "VANTAGE+ Fuel Assembly Reference Core Report,"

April 1995) is valid for uranium oxide fuel without the presence of poisons mixed homogeneously into the fuel pellets. If fuel pellets incorporating homogeneous poisons are used, the topical report documenting the fuel centerline melt temperature basis must be reviewed and approved by the NRC and referenced in this license condition. TS 2.1.1.3 must be modified to also include the fuel centerline melt temperature limit for the fuel with homogeneous poison.

AMENDMENT NO.~k

ATTACHMENT 3 Markup of Byron Station Proposed Operating License, Technical Specifications, and Technical Specifications Bases Changes Byron Station, Units 1 and 2 Facility Operating License Nos. NPF-37 and NPF-66 License Amendment Request to Administratively Clarify the Operating Licenses and Technical Specifications for Braidwood Station and Byron Station

ABLE OF CONTENTS - TECHNICAL SPECIFICATION TABLES YRON - UNITS 1 & 2 v

Amendment 150 4

Tab]

1.1-1 Table

.3.1-1 MODES..

Reactor Trip System Instrumentation 1.1-9 3.3.1 Tab] e 3> 3.2-1 Engineered Safety Feature Actuation System Instrumentation...

... 3.

.2-9 Table 3.3.

-1 Post Accident Monitoring Instrumentation.

'3.3-4 Table 3.3.4`

Remote Shutdown Monitoring Instrumentation..

.3.4-3 Table 3.3.6-1 :

Containment Ventilation Isolation Instrumentati 3.3.6-4 Table 3.3.7-1 VC Filtration System Actuation Instrumentatio 3.3.7-4 Table 3.3.8-1 FHB Ventilation System Actuation Instrumenta ion 3.3.8-4 Table 3.7.1-1 OPERABLE Main Steam Safety Valves versus Applicable Power in Percent of RATED ERMAL POWER.

.. 3.7.1-3 Table 3.7.1-2 Ma Steam Safety Valve Lift Settings

... 3.7.1-4 Table 3.8.6-1 Bat ry Cell Parameters Requirements, '....

... 3.8.6-4

Reactor Core Safety Limits 2.0-Moderator Temperature Coefficient vs. Power Level 3.

Reactor Coolant DOSE EQUIVALENT I-131 Specific Activity Limit Versus Percent of RATED THERMAL POWER.

Seal Injection Flow Limits.

Region 2 All Cell Configuration Burnup Credit, Requirements..

Region 2 3-out-of-4Checkerboard Configura Burnup Credit Requirements.

Region 2 2-out-of-4 Checkerboard Config Burnup Credit Requirements.

ACTIONS (continued)

REQUIRED ACTION

NOTE------------

(

One train may be bypassed for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for surveillance testing, provided the other train is OPERABLE.

CONDITION N.

One RTB train inoperable.

0.

RTS Instrumentation 3.3.1 COMPLETION TIME (continued)

BYRON - UNITS 1 & 2 3.3.1 - 6 Amendment 153 N.1 Restore train to 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months OPERABLE status.

N.2 Be in MODE 3.

30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months One or more channels 0.1 Verify interlock is 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months inoperable.

in required state for existing unit conditions.

OR 0.2 Be in MODE 3.

7 hours8.101852e-5 days 0.00194 hours 1.157407e-5 weeks 2.6635e-6 months

SURVEILLANCE REQUIREMENTS

_________________NOD___________________

required to be performed in 1 and 2.

Onl MOD SURVEILLANCE rify closure time of each MSIV is 5 seconds.

NUDES __________________-

-}-.

Only~~required to be performed in MODES 1 and 2.

Verify each MSIV actuates to the isolation position on an actual or simulated actuation signal.

13YRON - UNITS 1 & 2 3.7.2 - 2 FREQUENCY MSIVs 3.7.2 In accordance with the Inservice Testing Program 18 months Amendment SR 3.7.2.1 Ve<_

SR 3.7.2.2

BASES SURVEILLANCE SR 3.7.2.1 REQUIREMENTS a Note. This Note a Note. This Note SR 3.7.2.2 MSIVs B 3.7.2 This SR verifies that MSIV closure time is <_ 5 seconds.

The MSIV closure time is assumed in the accident and containment analyses.

This Surveillance is normally performed upon returning the unit to operation following a refueling outage.

Based on ASME Code Section XI (Ref.

5), the MSIVs are not closure time tested at power.

The Frequency is in accordance with the Inservice Testing Program.

This test is conducted in MODE 3 with the unit at operating temperature and pressure.

This SR is modified by allows entry into and operation in MODE 3 prior to performing the SR.

This allows a delay of testing until MODE 3, to establish conditions consistent with those under which the acceptance criterion was generated.

Nete 2 This SR verifies that each MSIV can close on an actual or simulated actuation signal This Surveillance is normally performed upon returning the unit to operation following a refueling outage.

The frequency of MSIV testing is every 18 months.

The 18 month Frequency for testing is based on the refueling cycle.

Operating experience has shown that these components usually pass the Surveillance when performed at the 18 month Frequency.

Therefore, this Frequency is acceptable from a reliability standpoint.

This SR is modified by e

allows entry into and operation in MODE 3 prior to performing the SR.

This allows a delay of testing until MODE 3, to establish conditions consistent with those under which the acceptance criterion was generated.

E REFERENCE'S I.

UESAR, Section 10.3.

2.

U SAR. Section 15.1.5.

3 UE

.0 CSR 50.07.

5.

AS1E Boiler and Pressure Vessel Code, Section ection XI BYRON - UNITS 1 & 2 B 3.7.2 - 6 Revision

3.7 PLANT SYSTEMS 3.7.8 Essential Service Water (SX) System LCO 3.7.8 The following SX trains shall be OPERABLE :

a.

Two unit-specific SX trains ; and b.

One opposite-unit SX train for unit-specific support.

APPLICABILITY :

MODES l, 2, 3, and 4.

ACTIONS A.

CONDITION I

REQUIRED ACTION NOTES

~U,=ri- ~1--t g

" Unit 2 QLTT"Yng SX sH^C G 'e't+

Refueling 13 _h ;l w,-rrre i~-

c Unit I is i19 MODE 5 6-,-er d

Restore unit-specific One unit-specific SX SX train to OPERABLE train inoperable.

status.

BYRON - UNITS 1 & 2 3.7.8 - 1 A.1

NOTES --------

1.

Enter applicable Conditions and Required Actions of LCO 3.8.1, "AC Sources-Operating," for Emergency Diesel Generator made inoperable by SX.

2.

Enter applicable Conditions and Required Actions of LCO 3.4.6, "RCS Loops-MODE 4," for Residual Heat Removal loops made inoperable by SX.

72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Amendment SX System 3.7.8 COMPLETION TIME (continued)

ACTIONS (continued)

SX System 3.7.8 BYRON - UNITS 1 & 2 3.7.8 - 2 Amendment CONDITION REQUIRED ACTION COMPLETION TIME Cn

~-r enffleftt *4 i7T° T F..sa.~.ri.G SX 5Eieti-6t++

11 111 r dud-t 2

:rsu:

"-¬$

J n

J i

L,.i1Y

~

7ir

-~i~.~

Mn--

.r"0 :i S

IIRGS i

~T~I

~ n

_ r _

MODE n 1

e\\mu~r-i-rvprm'i

`

A49D E 5,

6- ~

by SX.

.-1e H el ed.

VIRW

' F' SX

7N~7i~Tir~

t vn -, ; in-4r

.7 Opposite-unit SX train

,'i

~

Restore opposite-unit 7 days inoperable.

g.l SX train to OPERABLE status.

Required Action and

~~~w-Be in MODE 3.

6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated ComUl eti

'Zl Be i n MODE 5.

36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months A or B not met. -

BASES APPLICABILITY In MODES l, 2, 3, and 4, the unit-specific SX System is a normally operating system that is required to support the OPERABILITY of the equipment serviced by the SX System and required to be OPERABLE in these MODES.

ACTIONS A. l While a specific unit is in MODES l, 2, 3, or 4, the opposite-unit SX System must be available (independent of the opposite unit's MODE or condition) for unit-specific support.

This minimizes the risk associated with loss of all unit-specific SX.

In MODES 5 and 6 the OPERABILITY requirements of the unit-specific SX System are determined by the systems it supports and there are no opposite-unit SX System requirements.

If one unit-specific SX train is inoperable, action must be taken to restore OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

In this Condition, the remaining OPERABLE SX train is adequate to perform the heat removal function.

However, the overall reliability is reduced because a single failure in the OPERABLE SX train could result in loss of the SX System function in the short term.

The 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Completion Time is based on the redundant capabilities afforded by the OPERABLE train, and the low probability of a DBA occurring during this time period.

BYRON - UNITS 1 & 2 B 3.7.8 - 4 SX System B 3.7.8 t f i e i U~ + i i5 4H MODE 5 u ur iirk H e'f LTCTrTT~-- 13

~'--while 6 8F Required Action A.1 is modified by two Notes.

The first Note indicates that the applicable Conditions and Required Actions of LCO 3.8.1, "AC Sources-Operating," should be entered if an inoperable SX train results in an inoperable emergency diesel generator.

The second Note indicates that the applicable Conditions and Required Actions of LCO 3.4.6, "RCS Loops-MODE 4," should be entered if an inoperable SX train results in an inoperable decay heat removal train.

These are exceptions to LCO 3.0.6 and ensure the proper actions are taken for these components.

Revision

BASES ACTIONS (continued)

BYRON - UNITS 1 & 2 B 3.7.8 - 5 SX System B 3.7.8 If the opposite-unit SX train is not OPERABLE for unit-specific support, action must be taken to restore OPERABLE status within 7 days.

In this Condition, if a complete loss of unit-specific SX were to occur, the SX System function would be lost.

The 7 day Completion Time is based on the capabilities of the unit-specific SX System and the low probability of a DBA with a loss of all unit-specific SX occurring during this time period.

Revision

BASES ACTIONS (continued)

SURVEILLANCE SR 3.7.8.1 REQUIREMENTS SX System B 3.7.8 If the unit-specific SX train or the opposite-unit SX train cannot be restored to OPERABLE status within the associated Completion Time, the unit must be placed in a MODE in which the LCO does not apply.

To achieve this status, the unit must be placed in at least MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in MODE 5 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging plant systems.

Verifying the correct alignment for manual, power operated, and automatic valves in the unit-specific SX flow path provides assurance that the proper flow paths exist for unit-specific SX operation.

This SR does not apply to valves that are locked, sealed, or otherwise secured in position, since they are verified to be in the correct position prior to being locked, sealed, or secured.

This SR does not require any testing or valve manipulation ;

rather, it involves verification that those valves capable of being mispositioned are in the correct position.

This SR does not apply to valves that cannot be inadvertently misaligned, such as check valves.

The 31 day Frequency is based on engineering judgment, is consistent with the procedural controls governing valve operation, and ensures correct valve positions.

This SR is modified by a Note indicating isolation of the SX components does not affect the OPERABILITY of the SX System.

Isolation of components may render those components inoperable.

BYRON - UNITS 1 & 2 B 3.7.8 - 6 Revision

3.7 PLANT SYSTEMS 3.7.15 Spent Fuel Pool Boron Concentration LCO 3.7.15 The spent fuel pool boron concentration shall be; as ACTIONS Spent Fuel Pool Boron Concentration 3.7.15 APPLICABILITY :

Whenever fuel assemblies are stored in the spent fuel pool

_____________________________________NOTE_____________________________________

LCO 3.0.3 is not applicable.

AND Suspend movement of fuel assembl ies in the spent fuel pool.

A.2 Initiate action to restore spent fuel pool bores concentration to within limit.

COITION A.

Spent fuel pool boron concentration not within limit.

BYRON - UNITS 1 & 2 3.7.15 - 1 REQUIRED ACTION COMPLETION TIME Immediately Immediately Amendment

B 3.7 PLANT SYSTEMS B 3.7.15 Spent Fuel Pool Boron Concentration BASES BACKGROUND The spent fuel pool provides for storage of various Westinghouse Optimized Fuel Assembly (OFA) types of different initial fuel enrichments and exposure histories in two distinct regions.

(For this discussion, the term OFA is intended to refer to the specific reduced fuel rodlet diameter, and includes all analyzed fuel types with this diameter,. such as Vantage 5.) ^

In 4*14

, du-iiflg die The m#h 4 wee spent fuel pool storage racks, hi preside placement locations for a total of 2~4 new or used fuel assemblies.

Of the 24 He4ee spent fuel pool stora~ge racks, four are designated "Region 1" with the remaining 20 racks designated as "Region 2." The analytical methodology used for the criticality analyses is in accordance with established NRC guidelines (Ref. 2)

BYRON - UNITS 1 & 2 B 3.7.15 - 1 Spent Fuel Pool Boron Concentration B 3.7.15 Revision

Spent Fuel Pool Boron Concentrati B 3.7.

BASES BACK Abso enri,

4).

boron load 1.5 mg B'°/i material 1 Region 2 racks for storing Westi configurations

1) "All Cells" 2 "3-out-of-4,

3) "2-out-of; For the "All Ce assemblies may

<- 1.14 weigh bui nup or initial when fue are c For sto (continued) aph Oat Spent Fuel Pool Storage Racks Region 1 racks contain 392 cells which are area, storing Westinghouse OFAs in an "All Cells" (that is, the criticality analysis assumes i~lies reside in all available cell 1 t ion of the boundary requirements).

'lies may contain an initial norm weight percent U-235 (without In Hers (I

) installed) up to an/

't of < 5.0 weight percent

~nt for a minimum number o.

IFBAs are required of 1.0X, equal to This is the mi offered by retain 247,

.w ka s" sto contain an iri percent U-235 ioactive decay of inal enrichment of <_

buinup and radioactive item.

"3-out-of-4 Checkerboard" sto rea reel assemblies ma contain an i ichment of < 1.64 weig~it percent U-2 it for fuel lxunup or radioactive constituents) to an initial nominal enri weight percent-235, when fuel burnup and of dl constituents are credited.

In this s there can be no more than three stored assembli matrix of cell lattices.

for t

t spent fuel tions, with the The stored fuel enrichment of egral Fuel Burnable initial nominal

'-235, provided that the 16 IFBAs is met (Refs. 3 have, as a minimum, a amount of imum standard poison tinghouse for 17X17 OFAs.

cells which are also analyzed FAs in a combination of storage tterns are :

ov erboard" Storage ; and erboard" Storage.

configuration, the stored fuel itial nominal enrichment of thout taking credit for fuel el constituents) up to an

.0 weight rcent U-235, 3ecay of rl constituents configuration, the itial nominal (without taking of fuel nt of < 5.0 ioactive decay rage pattern, in any 2X2 5

,r YRON - UNITS 1 & 2 B 3.7.15 - 2 Revision 9

BASES BACKGROUND (continued)

APPLICABLE SAFETY ANALYSES ion 1 racks contain 396 cells which are analyzed for storing Westinghouse OFAs in an "All Cells" arrangement (that is, the criticality analysis assumes that t fuel assemblies reside in all available cell locations)

The stored fuel assemblies may contain an initial nominal enrichment of 5 5.0 wei t percent U-235 (with or without IFBAs installed) (Ref. P)

Spent Fuel Pool Boron Concentration B 3.7.15 Region 2 racks contain 2588 cells which are also analyzed for storing Westinghouse OFAs in an "All Cells" arrangement (that is, the criticality analysis assumes that spent fuel assemblies reside in all available cell locations)

For the "All Cells" storage configuration, the stored fuel assemblies may certain an initial nominal enrichment of

<- 5.0 weight percent U-235 with credit for ixunup.

The water in the spent fuel pool normally contains soluble boron which results in large subcriticality margins under actual operating conditions.

approved methodologies mere used to develop the criticality analyses (Ref. 2) for the lee spent fuel pool storage racks.

The fuel handling accident analyses are described in Reference 6.

Additional evaluations performed (Refs. 4

--awl 8) to support placement of the Byres lead test assemblies with higher density pellets in the spent fuel pool.

BYRON - UNITS 1 & 2 B 3.7.15 Revision

BASES APPLICABLE SAFETY ANALYSES (continued)

Spent Fuel Pool Boron Concentration B 3.7.15 The criticality analyses for the spent fuel assembly storage racks confirm that kef remains

<_ 0.95 for the-lea tee spent fuel pool storage racks (including uncertainties and tolerances) at a 95fo probability with a 95% confidence level (95/95 basis), based on the accident condition of the pool being flooded with unborated water.

Thus, the deliM of both regions assumes the use of unborated water while maintaining stored fuel in a subcritical condition.

However, the presence of soluble boron has been credited to provide adequate safe t~argin to maintain spent fuel assembly storage rack

<_ 0.95 (also on a 95/95 basis) for all postulated accident scenarios involving dropped or misloaded fuel assemblies Crediting the presence of soluble boron for mitigation of these scenarios is acceptable based on applying the "double contingency principle" which states that there is no requirement to assume two unlikely, independent, concurrent events to ensure protection against a criticality accident (Refs. 9 and 10).

The accident analyses address the following five postulated scenarios :

1) fuel assembly drop on top of rack ;

2) fuel assembly drop between rack modules ;

3) fuel assembly drop between rack modules and spent fuel pool wall ;

4 change in t fuel pool water temperature ; and 5~ fuel assemb loaded contrary to placement restrictions.

only scenarios 2, 3, and have the capacity to increase reactivity for the wee spent fuel pool storage racks.

BYRON - UNITS 1 & 2 B 3.7.15 -

Revision

BASES APPLICABLE SAFETY ANALYSES (continued)

Spent Fuel Pool Boron Concentration B 3.7.15 Calculations were use performed, for the wee spent fuel pool storage racks, for a spent fuel pool temperature of 4°C (39°F) which is well below the lowest normal operating temperature (50°F).

Because the temperature coefficient of reactivity in the spent fuel pool is negative, temperatures greater than 4°C will result in a decrease in reactivity.

Calculations were also performed to show the largest reactivity increase caused by a Westinghouse 17X17 OFA fuel assembly misplaced into a lal-tee Region 2 storage cell for which the restrictions on enrichment or burnup are not satisfied.

The assembly misload accident can only occur during fuel handling operations in the spent fuel pool BYRON - UNITS 1 & 2 B 3.7.15 - 5 Revision

BASES APPLICABLE SAFETY ANALYSES (continued)

Should For the above vostulated accident conditions, the double contingency principle can be lied.

S ifically, the presence of soluble boron in spent pool water can assumed as a realistic initial condition since not assuming its presence wild be a second unlikely event.

Fop Spent Fuel Pool Boron Concentration B 3.7.15 For the H9 1 tee spent fuel pool storage racks, spent fuel pool soluble boron has been credited in the criticality safety analysis to offset the reactivity caused by postulated accident conditions.

Because theRegion 1 racks are designed for the storage of fresh fuel assemblies, a fuel assembly misload accident has no consequences from a criticality standpoint (i.e., the acceptance criteria for storage are satisfied by all assemblies in the spent fuel pool).

pool. 44i-it-P.

a fuel assembly misload accident occur in the Region gage cells, k ff will be maintained <- 0.95 due to the presence of at least 300 ppm of soluble boron in the spent fuel pool water.

BYRON - UNITS 1 & 2 B 3.7.15 - 6 Revision

BASES APPLICABLE SAFETY ANALYSES (continued)

Spent Fuel Pool Boron Concentration B 3.7.15 The concentration of dissolved boron in the spent fuel pool satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii).

The spent fuel pool boron concentration. is required to be

~! 300_ppm s-awe.

The specified concentration of dissolved boron in the spent fuel pool preserves the assumptions used in the analyses of the potential critical accident scenarios as described in References 3, 5, 6, and 7.

The di

-rae1ts-The dissolved boron concentration of 300 ppm botuxls the minimum required concentration for accidents occurring duri fuel assembly movement within the spent fuel poi BYRON - UNITS 1 & 2 B 3.7.15 - 7 Revision

BASES APPLICABILITY This LCO applies whenever fuel assemblies are stored in the spent fuel pool.

ACTIONS The ACTIONS have been modified by a Note indicating that LCO 3.0.3 does not apply.

A.1 and A.2 SURVEILLANCE SR 3.7.15.1 REQUIREMVTS Spent Fuel Pool Boron Concentration B 3.7.15 When the concentration of boron in the spent fuel pool is less than required, immediate action must be taken to preclude the occurrence of an accident or to mitigate the consequences of an accident in progress.

This is most efficiently achieved by immediately suspending the movement of fuel assemblies.

This does not preclude movement of a fuel assembly to a safe position.

Immediate actions are also taken to restore spent fuel pool boron concentration.

If moving fuel assemblies while in MODE 5 or 6, LCO 3.0.3 would not specify any action.

If moving fuel assemblies while in ES 1, 2, 3, and 4, the fuel movement is independent of reactor operations.

Therefore, inability to suspend movement of fuel assemblies is not sufficient reason to require a reactor shutdown.

This SR verifies that the concentration of boron in the spent fuel pool is within the required limit.

As long as this SR is met, the analyzed accidents are fully addressed.

BYRON - UNITS 1 & 2 B 3.7.15 - 8 Revision

BASES SURVEILLANCE REQUIREMFNTS (continued)

REFERENCES The 7 day frequency is appropriate based on operating experience and takes into consideration that no maJ or replenishment of spent fuel pool water is expected to occur over such a short period of time.

3.

4.

5.

Spent Fuel Pool Boron Concentration B 3.7.15 2.

NRC Memorandum from L. Kopp to T. Collins, "Guidance on the Regulatory Requirements for Criticality Analysis of Fuel StoraRe at Light Water Reactor Povuer Plants," dated August I9, 1998.

Holtec International Report, HI-982094, "Criticality Analysis for the n/Braidwood Rack Installation Project," Project 80944, 1998.

6.

UFSAR, Section 15.7.4.

7.

"Byron/Braidwood Spent Fuel Pool Dilution Analysis,"

Rev. 3, dated June 17, 1997.

8 CN-CRIT-141 "Analysis Suorting the LTA Assemblies for Byron/Braidwood SFP,' dated February 4, 1999.

9.

Double contingency principle of ANSI N16.1 - 1975, as ified in the April 14, 1978 NRC letter (Section 1.2) and implied in the pop revision to Regulatory Guide 1.13 (Section 1.4, Appendix A).

10.

ANSI/ANS 8.1 - 1983 "American National Standard for Nuclear Criticality Safety in Operations with Fissionable Materials Outside Reactors."

BYRON - UNITS 1 & 2 B 3.7.15 - 9 Revision 9 Deleted

3.7 PLANT SYSTEMS 3.7.16 Spent Fuel Assembly Storage LCO 3.7.16 Each spent fuel assembly stored in the spent fuel pool shall, as applicable :

BYRON - UNITS 1 & 2 3.7.16-1 Spent Fuel Assembly Storage 3.7.16 Region 1 of Holtee spent fuel pool storage racks Have an initial nominal enrichment of -< 5.0 weight percent U-235 to permit storage in any cell location.

Region 2 of Haltee spent fuel pool storage racks Have a combination of initial enrichment and within the Acceptable Burnup Domain of Figure APPLICABILITY :

Whenever fuel assemblies are stored in the spent fuel pool.

Amendment

ACTIONS LCO 3.0.3 is not applicable.

CONDITION I

REQUIRED ACTION irements of the I A.1 not met.

SURVEILLANCE REQUIREMENTS a-- -Z_Initial nominal enrichment of the fuel assembly is <- 5.0 weight percent U-235.

BYRON - UNITS 1 & 2 3.7.16 - 2 Spent Fuel Assembly Storage 3.7.16 Initiate action to move the noncomplying fuel assembly into a location which restores compliance.

COMPLETION TIME Immediately Prior to storing the fuel assembly in Region 1 Amendment

Spent Fuel Assembly Storage 3.7.16 BYRON - UNITS 1 & 2 3.7.16 - 3 Amerxinent SURVEILLANCE REQUIRIIl1EVIS SURVEILLANCE FREQUENCY b,

initial

..1 eh fwnt :,

y^r.T..Tw -- :-

is 4-v T~_

SR 3.7.16.2 F4-

.a~

~~.4.46

'-7

.~.-

cait.

... _. ". _Y!

N M e 3 4

" 1 TT7T!1 Verify by administrative means t lx Prior to combination of initial enrichment/ burnup, storing the

~-time, as livable, of the fuel fuel assembly assembly is within Acceptable Burnup in Region 2 Domain of Figure 3.7.16-1, ?.7.' 1-2, SR 3.7.16.3 B

n" "Jrw~" r"iw

~I7W storage t-adts.

..t ;

Pp i 01-te inte A "SR~~l"`a""" l!11.\\1" lLw~ll1"S! "!~" 1~,"

TI-I-1

Figure 3.7.16-2 (page 1 of 1) 2 3-out-of-4 Checkerboard Configuration Burnup Credit Require'Npnts (Joseph Oat Spent Fuel Pool Storage Racks)

S' T 113 LE 01IAIX

Figure 3_7.16-3 (page 1 of 1)

RegXn 2 2-out-of-4 Checkerboard Configuration Burnup Credit Require (Joseph Oat Spent Fuel Pool Storage Racks)

-4."?

A.I.

-1.f>

INITIAL I-

`' :i :) E\\ HWII-NIENT (w/o)

MTEPTAI3LE

InUP DOMAIN 1(."("EPTAI3LE

~I.'I' I)0m 1

BYRON - UNITS 1 & 2 Figuret~(page 1 of 1)

Region 2 Fuel Assembly Burnup Requirements (weltee speni=Fz;ej --P-GA-,-

~P AgP QA L~-;

3.7.16 -

Spent Fuel Assembly Storage 3.7.16 Amendment

B 3.7 PLANT SYSTEMS B 3.7.16 Spent Fuel Assembly Storage BASES BACKGRO(M The spent fuel pool provides for storage of various Westinghouse Optimized Fuel Assembly (OFA) types of different initial fuel enrichments and exposure histories in tm distinct regions.

(For this discussion, the term OFA is intended to refer to the specific reduced fuel rodlet diameter, and includes all analyzed fuel types with this diameter, such as Vantage 5.)

addition, dEwitig the

". " Ir!_. a !N om_~

a..71%am_!_ ".

141-MM WI L" 1--M1 r, 0 ate:

"i.:fi b+"Mi WA_%

+

v is WtiWOii :inal~w wwr ~wra:rwar

-IL-0 a ~i

_iMw=w ". ~m AT No1% MR=

+ri~

t S"

~ "

BYRON - UNITS 1 & 2 B 3.7.16 - 1 Spent Fuel Assembly Storage B 3.7.16 Revision

BASES BACKGROM (continued)

BYRON - UNITS 1 & 2 Spent Fuel Assembly Storage B 3.7.16 "TP

.ais: .U'::wi:A-10i:

7-C W.

0a lwtkq ss"iWa+iiy`riiN i"iir+"rWV_Wisyi:iWrryi+~i

~6!~i~fF.

~

~!7~SThr les/'

Q"I9lv " i ;,;l "

1 MW R!!!JlE~~ f !7lf.iTC "F ~!~!l~T D-T!fY""fY~~lNlr B 3.7.16 vision

BASES BACKGROUND (continued)

For the -"2-out-of-4 Checkerboard" storage configuration, APPLICABI.E SAFETY ANALYSES Spent Fuel Assembly Storage B 3.7.16 Region 1 racks contain 396 cells which are analyzed for storing Westinghouse OFAs in an "All Cells" arrangement (that is, the criticality analysis assumes that t fuel assemblies reside in all available cell locations)

The stored fuel assemblies may contain an initial nominal enrichment of <- 5.0 wei t percent U-235 (with or without IFBAs installed) (Ref. ~)

Region 2 racks contain 2588 cells which are also analyzed for storing Westinghouse OFAs in an "All Cells" arrangement (that is, the criticality analysis assumes that spent fuel assemblies reside in all available cell locations For the "All Cells" storage configuration, the stored fuel assemblies may contain an initial nominal enrichment of

_< 5.0 weight percent U-2,35 with credit for burnup.

The water in the spent fuel pool normally contains soluble boron which results in large subcriticality margins under.

actual operating conditions.

approved methodologies-were used to develop the criticality analyses (Ref. 2) for the Holtec spent fuel pool storage racks.

The fuel handling accident analyses are described in Reference 6.

Additional evaluations werepe performed (Refs. 4; aM 8) to su~o support placement of the Byyron lead test assemblies with higher density pellets in the spent fuel pool.

BYRON - UNITS 1 & 2 B 3.7.16 430 Revisi

BASES APPLICABLE SAFETY ANALYSES (continued)

Spent Fuel Assembly Storage B 3.7.16 The criticality analyses for the spent fuel assembly storage racks confirm that Kff remains

_< 0.95 for the Ho+tee spent fuel pool storage racks (including uncertainties and tolerances) at a 95% probability with a 95% confidence level (95/95 basis), based on the accident condition of the pool being flooded with unborated water.

Thus, the desi~n of both regions assumes the use of unborated water while maintaining stored fuel in a subcritical condition.

However, the presence of soluble boron has been credited to provide adequate safe mar~in to maintain spent fuel assembly storage rackIff

.95 (also on a 95/95 basis) for all postulated accident scenarios involving dropped or misloaded fuel assemblies Crediting the preseiCe of soluble boron for mitigation of these scenarios is acceptable based on applying the "double contingency principle" which states that there is no requirement to assume two unlikely, independent, concurrent events to ensure protection against a criticality accident (Refs. 9 and 10).

The accident analyses address the following five postulated scenarios :

1) fuel assembly drop on top of rack ;

2) fuel assembly drop between rack modules ;

3) fuel assembly drop between rack modules and spent fuel pool wall ;

4 change in spent fuel pool water temperature ; and 5~ fuel assembly loaded contrary to placement restrictions.

only scenarios 2, 3, and 5 have the `capacity to increase reactivity for the Herl-tee spent fuel pool storage racks.

BYRON - UNITS 1 & 2 B 3.7.16 - @

Revisionm

BASES APPLICABLE SAFM ANALYSES (continued)

Spent Fuel Assembly Storage B 3.7.16 Calculations were also perfo for a spent fuel pool temperature of 4°C (39°F) which is well below the lowest normal operating temperature (50°F).

Because the temperature coefficient of reactivity in the spent fuel pool is negative, temperatures greater than 4°C will result in a decrease in reactivity.

For the fuel assembly misload accident, calculations were performed to show the largest reactivity increase caused by a Westinghouse 17X17 OFA fuel assembly misplaced into a leltee Region 2 storage cell for which the restrictions on enrichment or burnup are not satisfied.

The assembly misload accident can only occur during fuel handling operations in the spent fuel pool.

BYRON - UNITS 1 & 2 B 3.7.16 -

Revision

BASES APPLICABLE SAFETY ANALYSES (continued)

Spent Fuel Assembly Storage B 3.7.16 For the above postulated accident conditions, the double contingency principle can be lied.

Specifically, the presence of soluble boron in spent fuel pool water can assumed as a realistic initial condition since not assuming its presence would be a second unlikely event. FE spent fuel pool storage racks, sp~

t fuel pool soluble boron has been credited in the criticality safety analysis to offset the reactivity caused by postulated accident conditions.

Because theRegion 1 racks are designed for the storage of fresh fuel assemblies, a fuel assembly misload accident has no consequences from a criticality standpoint (i.e., the acceptance criteria for storage are satisfied by all assemblies in the spent fuel pool).

tae--eel-ls'; k-Shcau ld __.."

should a fuel assembly misload accident occur in the Region 2 storage cells, k, will be maintained <_ 0.95 due to the presence of at least 300 ppm of soluble boron in the spent fuel pool water.

BYRON - UNITS 1 & 2 For the Heltee B 3.7.16 -YGd

Revision,

BASES APPLICABLE SUM ANALYSES (continued) 3.7. 16-1 Spent Fuel Assembly Storage B 3.7.16 The configuration of fuel assemblies in the spent fuel pool satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii)

The restrictions on the placement of fuel assemblies within the spent fuel pool in accordance with the requirements in the accompanying LCO ensure that the off of the spent fuel PC 0.95 assuming the pool is rated water for the wee spent fuel pool For-the_ He l ee spent fuel pool storage r'ack,-, in IM Fi

-u`f ~' the Acceptable Burn up Domain lies on, above, and to the left of the line.

BYRON - UNITS 1 & 2 B 3.7.16 - 7 Revision

BASES LCO (continued)

The APPLICABILITY This LCO applies whenever fuel assemblies are stored in the spent fuel pool.

ACTIONS The ACTIONS have been modified by a Note indicating that LCO 3.0.3 does not apply.

SURVEILLANCE SR 3.7.16.1 REQUIRBENTS BYRON - UNITS 1 & 2 Spent Fuel Assembly Storage B 3.7.16

`

use of linear interpolation between minimum burnups is acceptable.

When the configuration of fuel assemblies stored in the spent fuel pool is not in accordance with the requirements of the LCO, immediate action mast be taken to make the necessary fuel assembly movement(s) to bring the configuration into compliance.

If moving fuel assemblies while in MODE 5 or 6, WO 3.0.3 would not specify arty action.

If moving fuel assemblies while in MODES 1, 2, 3, and 4, the fuel movement is independent of reactor operations.

Therefore, inability to suspend movement of fuel assemblies is not sufficient reason to require a reactor shutdown.

Item a and item b areperformed, as applicable, is performed prior to storing the fuel assembly in the intended spent fuel pool storage location.

The frequency is appropriate because compliance with the SR ensures that the relationship between the fuel assembl and its storage location will meet

~

analyses.

of the and preserve the assumptions of yses.

This SR verifies by administrative means that the initial nominal enrichment of the fuel assembly

-er

}is met to ensure that the assumptions of the safety analyses are preserved.

B 3.7.16 -W Revision

BASES SURVEILLANCE REQUIREMENIS (continued)

Spent Fuel Assembly Storage B 3.7.16 SR 3.7.16.2 SR 3.7.16.2 is performed prior to storir~ the fuel assembly in the intended spent fuel pool storage location.

The frequency is appropriate because compliance with the SR ensures that the relationship between the fuel assembly and its storage location will meet the requirements of the L.CO and preserve the assumptions of the analyses.

This SR verifies by administrative means that the combination of initial enrichment, lx~rnuQ, and decay time, as applicable, of the fuel assembly is within the Acceptable &urnyp Domain of Figure 3.7.16-1, 3.

'--r

'-x66 2 ;-

for the intended storage configuration to ensure that the assumptions of the safety analyses are preserved.

BYRON - UNITS 1 & 2 B 3.7.16 - N Revision

(continued )

REFERENCES BYRON - UNITS 1 & 2 2.

NRC Memoranchn from L. Kopp to T. Collins, "Guidance on the Regulatory Requirements for Criticality Analysis of Fuel Storage at Light Water Reactor Power Plants," dated August 19, 1998.

3.

4.

5.

Spent Fuel Assembly Storage B 3.7.16 Holtec International Report, HI-982094, "Criticality Analysis for the Syrorm/Braidwood Rack Installation Project," Project No. 80944, 1998.

6.

UFSAR, Section 15.7.4.

7.

"Byron/Braidwood Spent Fuel Pool Dilution Analysis,"

Rev. 3, dated June 17, 1997.

8 CN-GRIT-141 "Analysis Supporting the LTA Assemblies for Byron/Braidwood SFP,' dated February 4, 1999.

9.

Double contingency principle of ANSI N16.1 - 1975, as specified in the April 14, 1978 NRC letter (Section 1.2) and implied in the proposed revision to Regulatory Guide 1.13 (Section 1.4, Appendix A) 10.

ANSI/ANS 8.1 - 1983 "American National Standard for Nuclear Criticality Safety in Operations with Fissionable Materials Outside Reactors."

B 3.7.16 -

Revision a Deleted

DESIGN FEATURES (continued) 4.3 Fuel Storage 4.3.1 Criticality C.

d.

The spent fuel storage racks are designed and shall be maintained, as applicable, with :

a.

Fuel assemblies having a maximum U-235 enrichment of 5.0 weight percent ;

4-i

~ 9 ~2c

ti irrc-rti ea=~-L-

+

w~.r~

+

ceF~s~'t"' to xv-~-,z-r-rrr6-t~-b-rs6t~' - a-rd ra-19.42 r---w Design Features 4.0

,. k~F <_ 0.95 i f ful 1 flooded with unborated water, which inclo es an allowance for uncertainties as described in Holtec International Report HI-982094, "Criticality Analysis for Byron/Braidwood Rack Installation Project," Project No. 80944, 1998 ;

a - ae-I(s.

a mi na l 10.888 A inch-north-south and 10.574 inch east-west c n er to cen e E distance between fuel assemblies placed in Region 1 racks ;

and nominal 8.97 inch center to center distance between fue assemblies placed in Region 2 racks.

A BYRON - UNITS 1 & 2 4.0 - 2 Amendment

APPENDIX C ADDITIONAL CONDITIONS FACILITY OPERATING LICENSE NO. NPF-37 The licensee shall comply with the following conditions on the schedules noted below:

Amendment Implementation 06 int first'p surveilla Surveillaric implementati existed prior to of performance ar reduced surveillance completion of the first s after implementation of A The licensee is authorized t requirements included in controlled documents.

amendment shall in these requiremen as described in December 1 October 1 and Ja Febr Apfi For SRs that are new in Amendment 106 to Shall be implemen Facility Operating License NPF-37, the first within 180 days Oer performance is due at the end of the first issuance of surveillance interval that begins at Amendmen006 implementation of Amendment 106

. For SRS that existed prior to Amendment 106, including SRs

~h modified acceptance criteria and SRs whose als of performance are being extended, the ormance is due at the end of the first e interval that begins on the date the was last performed prior to of Amendment 106. For SR hat endment 106, whose i rvals eing reduced, the st terval begins on

~tveillance erformed nd nt 106.

cate certain Shall be implemented A to licensee-within 180 days after of this issuance of ton of Amendment 106 ments r

ppen plementati de the initial rel to the appropriate dd e licensee's letters dated

!1996, February 24, SeptembeF

!October 28, and December 8, 1997; ary 27, January 29, February 6, ry 13, February 24, February 26, April 13, 16, June 1, June 2, July 2, July 8, July 30, ly 31, August 11, August 12. September 21, September 25, October 1, October 2, October 5, October 15, October 23, November 6, November 19, November 23, November 30 and December 14, 1998, and evaluated in the NRC staffs Safety Evaluation associated with this amendment.

Ad Amendment No ADDITIONAL CONDITIONS FACILITY OPERATING LICENSE NO. NPF-37 The licensee shall comply with the following conditions on the schedules noted below:

Amendment Implementation Number Additional Condition Date 1 T

.9...,_

The licensee shall implement modifications as discussed in Section 5.11.9 of the Safety Evaluation to maintain the stability of the Byron transmission grid.

modifications include a reduction in the existing loc brea ackup time settings and a revision of the-UM trip schem 119 The licensee shall s it to the N analysis using a model ac t

the value of 8.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months for leg injection followin Evaluation Sect time usin Prior to i me e-on of power up-rate conditions confirmatory Submit by to the NRC justifying June 1, 2002 of switchover to hot accident (Safety switchover Prior to imple-ntation of er up-n ns ti oss-of-coo 3.1.3) ; or recalculate currently accepted methodology:

licensee shall make the instrumentation changes a described in Section 4.15.2 of the Safety Evaluation.

full rate co 127 The safety limit equation specified in TS 2.1.1.3 regarding fuel centerline melt temperature (i.e., less than 5080 °F, decreasing by 58 °F per 10,000 MWD/MTU burnup as described in WCAP-12610-P-A, "VANTAGE+ Fuel Assembly Reference Core Report,"

April 1995) is valid for uranium oxide fuel without the presence of poisons mixed homogeneously into the fuel pellets.

If fuel pellets incorporating homogeneous poisons are used, the topical report documenting the fuel centerline melt temperature basis must be reviewed and approved by the NRC and referenced in this license condition. TS 2.1.1.3 must be modified to also include the fuel centerline melt temperature limit for the fuel with homogeneous poison.

With imple-mentation of the amend-ment AMENDMENT NO.

APPENDIX C ADDITIONAL CONDITIONS FACILITY OPERATING LICENSE NO. NPF-66 The licensee shall comply with the following conditions on the schedules noted below:

Amendment Implementation Number Additional Condition Date 106 int first -0 surveill Surveillan implementati existed prior to of performance ar reduced surveillance completion of the first s after implementation of A The licensee is authorized t requirements included in controlled documents.

amendment shall in these requiremen as described in December 1 October 1 and Jan Febr Apyi For SRs that are new in Amendment 1d6 to Facility Operating License NPF-66, the first performance is due at the end of the first surveillance interval that begins at implementation of Amendment 106. For SRs that existed prior to Amendment 106, including SRs

[th modified acceptance criteria and SRs whose als of performance are being extended, the ormarice is due at the end of the first ce interval that begins on the date the,

was last performed prior to of Amendment 106

. For SIR endment 106, eing reduce reduced, the interval begins eillance nd or at Shall be implemen within 180 days issuance of Amendmen r

ppen pleme de the initial to the appropriate d e licensee's letters dated X1996, February 24, Septembe

!October 28, and December 8, 1997; ry 27, January 29, February 6, "ry 13, February 24, February 26, April 13, 16, June 1, June 2, July 2, July 8, July 30, ly 31, August 11, August 12, September 21, September 25, October 1, October 2, October 5, October 15, October 23, November 6, November 19, November 23, November 30 and December 14, 1998, and evaluated in the NRC staffs Safety Evaluation associated with this mendment.

whose i rvals ns on erformed nt 106.

06 er ovate certain Shall be implemented

' A to licensee-within 180 days after ntat n of this issuance of tial re tion of Amendment 106 hate d uments Amendment No

Amendment Number 127 Additional Condition ADDITIONAL CONDITIONS FACILITY OPERATING LICENSE NO. NPF-66 The licensee shall comply with the following conditions on the schedules noted below:

The safety limit equation specified in TS 2.1.1.3 regarding fuel centerline melt temperature (i.e., less than 5080 °F, decreasing by 58 °F per 10,000 MWD/MTU burnup as described in WCAP-12610-P-A, "VANTAGE+ Fuel Assembly Reference Core Report,"

April 1995) is valid for uranium oxide fuel without the presence of poisons mixed homogeneously into the fuel pellets.

If fuel pellets incorporating homogeneous poisons are used, the topical report documenting the fuel centerline melt temperature basis must be reviewed and approved by the NRC and referenced in this license condition. TS 2.1.1.3 must be modified to also include the fuel centerline melt temperature limit for the fuel with homogeneous poison.

Implementation Date The licensee shall implement modifications as discussed in Section 5.11.9 of the Safety Evaluation to maintain the stability of the Byron transmission grid.

he modifications include a reduction in the existing log er backup time settings, a revision of the uni _VA nd the installation of a power syst if to th Prior to i me bP scherri stabilizer.

The licensee shall su analysis using a model a the value of 8.5 hour5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months leg injection foll Evaluation time usi C a confirmatory ble to the NRC justifying f switchover to hot ccident (Safety witchover e-ion of power up-rate conditions Submit by June 1, 2002 Prior to imple-ntation of r up-Nitiens "r the ti fig a loss-of-cools fiction 3.1.3) ; or recalculate t g the currently accepted methodology The licensee shall make the instrumentation changes as described in Section 4.15.2 of the Safety Evaluation.

full p c

rate con With imple-mentation of the amend-ment AMENDMENT NO.=

RS-08-087, License Amendment Request to Administratively Clarify the Operating Licenses and Technical Specifications

Contents

Text

2.0 PROPOSED CHANGE

3.0 BACKGROUND

4.0 TECHNICAL ANALYSIS

5.0 REGULATORY ANALYSIS

6.0 ENVIRONMENTAL CONSIDERATION

2.0 PROPOSED CHANGE

3.0 BACKGROUND

4.0 TECHNICAL ANALYSIS

5.0 REGULATORY ANALYSIS

6.0 ENVIRONMENTAL CONSIDERATION

7.0 REFERENCES

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