ML093080158

Technical Specification Change TS-467 - Utilization of Areva Fuel and Associated Analysis Methodologies

ML093080158
Person / Time
Site:	Browns Ferry
Issue date:	10/23/2009
From:	Krich R Tennessee Valley Authority
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
TS-467
Download: ML093080158 (149)

Text

Tennessee Valley Authority, 1101 Market Street, Chattanooga, Tennessee 37402-2801 October 23, 2009 10 CFR 50.90 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555-0001 Browns Ferry Nuclear Plant, Unit 1 Facility Operating License No. DPR-33 NRC Docket No. 50-259

Subject:

Technical Specification Change TS-467- Utilization of AREVA Fuel and Associated Analysis Methodologies

References:

1. Letter from NRC to TVA, "Summary of January 28, 2009, Meeting with the Tennessee Valley Authority Regarding Proposed Fuel Transition Amendment (TAC No. ME0438)," dated March 23, 2009

2. Letter from NRC to TVA, "Summary of March 16, 2009, Meeting with the Tennessee Valley Authority Regarding Proposed Fuel Transition Amendment (TAC No. ME0438)," dated June 3, 2009 In accordance with the provisions of 10 CFR 50.90, "Application for amendment of license, construction permit, or early site permit," the Tennessee Valley Authority (TVA) is submitting a request for amendments to the Technical Specifications (TS) for Browns Ferry Nuclear Plant, Unit 1.

The proposed amendment adds the AREVA NP analysis methodologies to the list of approved methods to be used in determining the core operating limits in the Core Operating Limits Report (COLR). Additional Technical Specification changes are requested to reflect the AREVA NP specific methods for monitoring and enforcing of the thermal limits.

The enclosure provides a description of the proposed changes, technical evaluation of the proposed changes, regulatory evaluation, and a discussion of environmental considerations. Attachment 1 identifies regulatory commitments. Attachment 2 provides the existing Unit 1 TS pages marked-up to show the proposed changes. shows the existing Unit 1 TS pages retyped to show the proposed changes. Attachment 4 provides Unit 1 TS Bases pages marked up to show the associated proposed changes. Attachment 5 provides Unit 1 TS Bases pages retyped to show the associated proposed changes.

Pnnted on recycled paper

U. S. Nuclear Regulatory Commission Page 2 October 23, 2009 In support of the proposed TS changes, certain technical information related to the transition core design and licensing analyses, as well as information related to the AREVA analysis methodologies, has been provided in Attachments 6 through 23 of this submittal. These attachments also provide the information requested during meetings, summarized in References 1 and 2, between TVA and NRC representatives. As discussed between TVA and NRC representatives during these meetings, the information attached to this submittal is based on Extended Power Uprate (EPU) conditions. However, due to recent delays in the EPU review for the Browns Ferry Nuclear Plant, TVA will provide supplemental information based on the Unit 1 current licensed thermal power by the end of 2009. The submittal of this supplemental information was previously agreed to between NRC and TVA representatives during a teleconference on June 2, 2009.

Attachments 6, 8, 10, 12, 14, 18, 20 and 22 contain information that AREVA NP considers to be proprietary in nature and subsequently, pursuant to 10 CFR 2.390, "Public inspections, exemptions, requests for withholding," paragraph (a)(4), it is requested that such information be withheld from public disclosure. Attachment 24 provides the affidavits supporting this request. Attachments 7, 9, 11, 13, 15, 19, 21 and 23 contain the redacted versions of the proprietary attachments with the proprietary material removed, which are suitable for public disclosure.

TVA has determined that there are no significant hazards considerations associated with the proposed changes and that the TS changes qualify for a categorical exclusion from environmental review pursuant to the provisions of 10 CFR 51.22(c)(9). Additionally, in accordance with 10 CFR 50.91(b)(1), TVA is sending a copy of this letter and the enclosure to the Alabama State Department of Public Health.

TVA requests approval of these TS changes by October 22, 2010, and that the implementation of the revised TS be made prior to the startup of Unit 1 for Cycle 9.

There are no regulatory commitments in this submittal as reflected in Attachment 1.

Please direct any questions concerning this matter to Dan Green at (423) 751-8423

• I declare under penalty of perjury that the foregoing is true and correct.

Executed on the 23rd day of October 2009.

Respectfully, 224 R. M. Krich Vice President Nuclear Licensing

U. S. Nuclear Regulatory Commission Page 3 October 23, 2009

Enclosure:

Revision of Technical Specifications to Allow Utilization of AREVA Fuel and Associated Analytical Methodologies Attachments:

1 List of Regulatory Commitments 2 Proposed Technical Specifications Changes (Mark-up) 3 Retyped Proposed Technical Specifications Pages 4 Proposed Technical Specification Bases Changes (Mark-up) 5 Retyped Proposed Technical Specification Bases Pages 6 Mechanical Design Report (proprietary)

ANP-2833(P), Revision 0, Mechanical Design Report for Browns Ferry Unit 1 Reload BFE1-9 ATRIUM-10 Fuel Assemblies, AREVA NP Inc., September 2009.

7 Mechanical Design Report (non-proprietary)

ANP-2833(NP), Revision 0, Mechanical Design Report for Browns Ferry Unit 1 Reload BFE1-9 ATRIUM-10 Fuel Assemblies, AREVA NP Inc., September 2009.

8 Thermal Hydraulic Design Report (proprietary)

ANP-2807(P), Revision 0, Browns Ferry Unit 1 Thermal-Hydraulic Design Report for ATRIUM-10 Fuel Assemblies, AREVA NP Inc.,

June 2009.

9 Thermal Hydraulic Design Report (non-proprietary)

ANP-2807(NP), Revision 0, Browns Ferry Unit 1 Thermal-Hydraulic Design Report for ATRIUM-10 Fuel Assemblies, AREVA NP Inc., June 2009.

10 Fuel Cycle Design Report (proprietary)

ANP-2850(P), Revision 0, Browns Ferry Unit 1 Fuel Cycle Design, AREVA NP Inc., July 2009.

11 Fuel Cycle Design Report (non-proprietary)

ANP-2850(NP), Revision 0, Browns Ferry Unit 1 Fuel Cycle Design, AREVA NP Inc., August 2009.

12 Reload Safety Analysis Report (proprietary)

ANP-2864(P), Revision 2, Browns Ferry Unit 1 Cycle Reload Safety Analysis, AREVA NP Inc., October 2009.

13 Reload Safety Analysis Report (non-proprietary)

ANP-2864(NP), Revision 2, Browns Ferry Unit 1 Cycle Reload Safety Analysis, AREVA NP Inc., October 2009.

14 LOCA Break Spectrum Analysis Report (proprietary)

EMF-2950(P), Revision 2, Browns Ferry Units 1, 2, and 3 Extended Power Uprate LOCA Break Spectrum Analysis, AREVA NP Inc., August 2009.

15 LOCA Break Spectrum Analysis Report (non-proprietary)

EMF-2950(NP), Revision 0, Browns Ferry Units 1, 2, and 3 Extended Power Uprate LOCA Break Spectrum Analysis, AREVA NP Inc., August 2009.

U. S. Nuclear Regulatory Commission Page 4 October 23, 2009 16 Response to NRC Comments Regarding Browns Ferry Unit 1 Proposed Fuel Transition Amendment (non-proprietary) 51-9121503-002, Response to NRC Comments Regarding Browns Ferry Unit 1 Proposed Fuel Transition Amendment, AREVA NP Inc., October 2009.

17 Boiling Water Reactor Licensing Methodology Compendium (non-proprietary)

ANP-2637, Revision 2, Boiling Water Reactor Licensing Methodology Compendium, AREVA NP Inc., December 2007.

18 Applicability of AREVA NP BWR Methods to Extended Power Uprate Conditions (proprietary)

ANP-2638(P), Revision 2, Applicability of AREVA NP BWR Methods to Extended Power Uprate Conditions, AREVA NP Inc.,

October 2009.

19 Applicability of AREVA NP BWR Methods to Extended Power Uprate Conditions (non-proprietary)

ANP-2638(NP), Revision 2, Applicability of AREVA NP BWR Methods to Extended Power Uprate Conditions, AREVA NP Inc.,

October 2009.

20 Part 1: Previous NRC Requests for Additional Information Matrix and Text Part 2: Browns Ferry Unit 1 - Summary of Response to Requests for Additional Information (proprietary)

ANP-2860(P), Revision 2, Browns Ferry Unit 1 - Summary of Responses to Requests for Additional Information, AREVA NP Inc., October 2009.

21 Part 1: Previous NRC Requests for Additional Information Matrix and Text Part 2: Browns Ferry Unit 1 - Summary of Response to Requests for Additional Information (non-proprietary)

ANP-2860(NP), Revision 2, Browns Ferry Unit 1 - Summary of Responses to Requests for Additional Information, AREVA NP Inc., October 2009.

22 Safety Limit Minimum Critical Power Ratio (proprietary) 51-9119738-000, Browns Ferry Unit 1 Cycle 9 MCPR Safety Limit Analysis (120% OLTP), AREVA NP Inc., September 2009.

23 Safety Limit Minimum Critical Power Ratio (non-proprietary) 51-9121246-000, Browns Ferry Unit 1 Cycle 9 MCPR Safety Limit Analysis (120% OLTP), AREVA NP Inc., September 2009.

24 Affidavits cc (Enclosure):

NRC Regional Administrator - Region II NRC Senior Resident Inspector - Browns Ferry Nuclear Plant State Health Officer, Alabama State Department of Public Health

ATTACHMENT 24 Browns Ferry Nuclear Plant (BFN)

Unit 1 Technical Specifications (TS) Change 467 Revision of Technical Specifications to allow utilization of AREVA NP fuel and associated analysis methodologies AFFIDAVITS Attached are the AREVA NP AFFIDAVITS supporting the request to withhold proprietary information from the public.

AFFIDAVIT STATE OF WASHINGTON )

) ss.

COUNTY OF BENTON )

1. My name is Alan B. Meginnis. I am Manager, Product Licensing, for AREVA NP Inc. and as such I am authorized to execute this Affidavit.

2. I am familiar with the criteria applied by AREVA NP to determine whether certain AREVA NP information is proprietary. I am familiar with the policies established by AREVA NP to ensure the proper application of these criteria.

3. I am familiar with the AREVA NP information contained in the report ANP-2833(P), Revision 0, entitled, "Mechanical Design Report for Browns Ferry Unit 1 Reload BFE1-9 ATRIUM-10 Fuel Assemblies," dated September 2009 and referred to herein as "Document." Information contained in this Document has been classified by AREVA NP as proprietary in accordance with the policies established by AREVA NP for the control and protection of proprietary and confidential information.

4. This Document contains information of a proprietary and confidential nature and is of the type customarily held in confidence by AREVA NP and not made available to the public. Based on my experience, I am aware that other companies regard information of the kind contained in this Document as proprietary and confidential.

5. This Document has been made available to the U.S. Nuclear Regulatory Commission in confidence with the request that the information contained in this Document be withheld from public disclosure. The request for withholding of proprietary information is made in accordance with 10 CFR 2.390. The information for which withholding from disclosure is

requested qualifies under 10 CFR 2.390(a)(4) "Trade secrets and commercial or financial information."

6. The following criteria are customarily applied by AREVA NP to determine whether information should be classified as proprietary:

(a) The information reveals details of AREVA NP's research and development plans and programs or their results.

(b) Use of the information by a competitor would permit the competitor to significantly reduce its expenditures, in time or resources, to design, produce, or market a similar product or service.

(c) The information includes test data or analytical techniques concerning a process, methodology, or component, the application of which results in a competitive advantage for AREVA NP.

(d) The information reveals certain distinguishing aspects of a process, methodology, or component, the exclusive use of which provides a competitive advantage for AREVA NP in product optimization or marketability.

(e) The information is vital to a competitive advantage held by AREVA NP, would be helpful to competitors to AREVA NP, and would likely cause substantial harm to the competitive position of AREVA NP.

The information in the Document is considered proprietary for the reasons set forth in paragraphs 6(b), 6(d) and 6(e) above.

7. In accordance with AREVA NP's policies governing the protection and control of information, proprietary information contained in this Document have been made available, on a limited basis, to others outside AREVA NP only as required and under suitable agreement providing for nondisclosure and limited use of the information.

8. AREVA NP policy requires that proprietary information be kept in a secured file or area and distributed on a need-to-know basis.

9. The foregoing statements are true and correct to the best of my knowledge, information, and belief.

SUBSCRIBED before me this /0 day 2009.-

~,  ; ~c~NOTARy

, -..*"" *.,,,;,,,,,,c, 0PUBLIC Ak.*?,,,- .

1,,0-,*

Susan K. McCoy NOTARY PUBLIC, STATE OF WASHINGTON MY COMMISSION EXPIRES. 1/10/12

AFFIDAVIT STATE OF WASHINGTON )

) ss.

COUNTY OF BENTON

1. My name is Alan B. Meginnis. I am Manager, Product Licensing, for AREVA NP Inc. and as such I am authorized to execute this Affidavit.

2. I am familiar with the criteria applied by AREVA NP to determine whether certain AREVA NP information is proprietary. I am familiar with the policies established by AREVA NP to ensure the proper application of these criteria.

3. I am familiar with the AREVA NP information contained in the report ANP-2807P, Revision 0, entitled "Browns Ferry Unit 1 Thermal-Hydraulic Design Report for ATRIUM-10 Fuel Assemblies," dated June 2009 and referred to herein as "Document."

Information contained in this Document has been classified by AREVA NP as proprietary in accordance with the policies established by AREVA NP for the control and protection of proprietary and confidential information.

4. This Document contains information of a proprietary and confidential nature and is of the type customarily held in confidence by AREVA NP and not made available to the public. Based on my experience, I am aware that other companies regard information of the kind contained in this Document as proprietary and confidential.

5. This Document has been made available to the U.S. Nuclear Regulatory Commission in confidence with the request that the information contained in this Document be withheld from public disclosure. The request for withholding of proprietary information is made in accordance with 10 CFR 2.390. The information for which withholding from disclosure is

requested qualifies under 10 CFR 2.390(a)(4) "Trade secrets and commercial or financial information."

6. The following criteria are customarily applied by AREVA NP to determine whether information should be classified as proprietary:

(a) The information reveals details of AREVA NP's research and development plans and programs or their results.

(b) Use of the information by a competitor would permit the competitor to significantly reduce its expenditures, in time or resources, to design, produce, or market a similar product or service.

(c) The information includes test data or analytical techniques concerning a process, methodology, or component, the application of which results in a competitive advantage for AREVA NP.

(d) The information reveals certain distinguishing aspects of a process, methodology, or component, the exclusive use of which provides a competitive advantage for AREVA NP in product optimization or marketability.

(e) The information is vital to a competitive advantage held by AREVA NP, would be helpful to competitors to AREVA NP, and would likely cause substantial harm to the competitive position of AREVA NP.

The information in the Document is considered proprietary for the reasons set forth in paragraphs 6(b), 6(d) and 6(e) above.

7. In accordance with AREVA NP's policies governing the protection and control of information, proprietary information contained in this Document have been made available, on a limited basis, to others outside AREVA NP only as required and under suitable agreement providing for nondisclosure and limited use of the information.

8. AREVA NP policy requires that proprietary information be kept in a secured file or area and distributed on a need-to-know basis.

9. The foregoing statements are true and correct to the best of my knowledge, information, and belief.

SUBSCRIBED before me this ___

day of .J ,2009. ,

ý4(1

- P*UBLIC ,L Susan K. McCoyWA  %,.,,* ,,,WJ "

NOTARY PUBLIC, STATE OF INGTON MY COMMISSION EXPIRES: 1/10/12

AFFIDAVIT COMMONWEALTH OF VIRGINIA )

) Ss.

CITY OF LYNCHBURG )

1. My name is Gayle F. Elliott. I am Manager, Product Licensing, for AREVA NP Inc. and as such I am authorized to execute this Affidavit.

2. I am familiar with the criteria applied by AREVA NP to determine whether certain AREVA NP information is proprietary. I am familiar with the policies established by AREVA NP to ensure the proper application of these criteria.

3. I am familiar with the AREVA NP information contained in the report ANP-2850(P), Revision 0, entitled "Browns Ferry Unit I Cycle 9 Fuel Cycle Design," dated July 2009 and referred to herein as "Document." Information contained in this Document has been classified by AREVA NP as proprietary in accordance with the policies established by AREVA NP for the control and protection of proprietary and confidential information.

4. This Document contains information of a proprietary and confidential nature and is of the type customarily held in confidence by AREVA NP and not made available to the public. Based on my experience, I am aware that other companies regard information of the kind contained in this Document as proprietary and confidential.

5. This Document has been made available to the U.S. Nuclear Regulatory Commission in confidence with the request that the information contained in this Document be withheld from public disclosure. The request for withholding of proprietary information is made in

accordance with 10 CFR 2.390. The information for which withholding from disclosure is requested qualifies under 10 CFR 2.390(a)(4) "Trade secrets and commercial or financial information."

6. The following criteria are customarily applied by AREVA NP to determine whether information should be classified as proprietary:

(a) The information reveals details of AREVA NP's research and development plans and programs or their results.

(b) Use of the information by a competitor would permit the competitor to significantly reduce its expenditures, in time or resources, to design, produce, or market a similar product or service.

(c) The information includes test data or analytical techniques concerning a process, methodology, or component, the application of which results in a competitive advantage for AREVA NP.

(d) The information reveals certain distinguishing aspects of a process, methodology, or component, the exclusive use of which provides a competitive advantage for AREVA NP in product optimization or marketability.

(e) The information is vital to a competitive advantage held by AREVA NP, would be helpful to competitors to AREVA NP, and would likely cause substantial harm to the competitive position of AREVA NP.

The information in the Document is considered proprietary for the reasons set forth in paragraphs 6(b) and 6(c) above.

7. In accordance with AREVA NP's policies governing the protection and control of information, proprietary information contained in this Document have been made available, on a limited basis, to others outside AREVA NP only as required and under suitable agreement providing for nondisclosure and limited use of the information.

8. AREVA NP policy requires that proprietary information be kept in a secured file or area and distributed on a need-to-know basis.

9. The foregoing statements are true and correct to the best of my knowledge, information, and belief.

SUBSCRIBED before me this____

day of August 2009.

Sherry L. McFaden NOTARY PUBLIC, COMMONWEALTH OF VIRGINIA MY COMMISSION EXPIRES: 10/31/10 Reg. # 7079129 SHERRY L.MC*PAON Notary Pfbile Commonweafth of Virginlo 7079129 My Commisalon Expires Oct 31, 2010 1 V- - - - - ,,i w _. _ ,

AFFIDAVIT STATE OF WASHINGTON )

) ss.

COUNTY OF BENTON )

1. My name is Alan B. Meginnis. I am Manager, Product Licensing, for AREVA NP Inc. and as such I am authorized to execute this Affidavit.

2. I am familiar with the criteria applied by AREVA NP to determine whether certain AREVA NP information is proprietary. I am familiar with the policies established by AREVA NP to ensure the proper application of these criteria.

3. I am familiar with the AREVA NP information contained in the report EMF-2950(P), Revision 2, entitled, "Browns Ferry Units 1, 2, and 3 Extended Power Uprate LOCA Break Spectrum Analysis," dated August 2009 and referred to herein as "Document."

Information contained in this Document has been classified by AREVA NP as proprietary in accordance with the policies established by AREVA NP for the control and protection of proprietary and confidential information.

4. This Document contains information of a proprietary and confidential nature and is of the type customarily held in confidence by AREVA NP and not made available to the public. Based on my experience, I am aware that other companies regard information of the kind contained in this Document as proprietary and confidential.

5. This Document has been made available to the U.S. Nuclear Regulatory Commission in confidence with the request that the information contained in this Document be withheld from public disclosure. The request for withholding of proprietary information is made in accordance with 10 CFR 2.390. The information for which withholding from disclosure is

requested qualifies under 10 CFR 2.390(a)(4) "Trade secrets and commercial or financial information."

6. The following criteria are customarily applied by AREVA NP to determine whether information should be classified as proprietary:

(a) The information reveals details of AREVA NP's research and development plans and programs or their results.

(b) Use of the information by a competitor would permit the competitor to significantly reduce its expenditures, in time or resources, to design, produce, or market a similar product or service.

(c) The information includes test data or analytical techniques concerning a process, methodology, or component, the application of which results in a competitive advantage for AREVA NP.

(d) The information reveals certain distinguishing aspects of a process, methodology, or component, the exclusive use of which provides a competitive advantage for AREVA NP in product optimization or marketability.

(e) The information is vital to a competitive advantage held by AREVA NP, would be helpful to competitors to AREVA NP, and would likely cause substantial harm to the competitive position of AREVA NP.

The information in the Document is considered proprietary for the reasons set forth in paragraphs 6(b), 6(d) and 6(e) above.

7. In accordance with AREVA NP's policies governing the protection and control of information, proprietary information contained in this Document have been made available, on a limited basis, to others outside AREVA NP only as required and under suitable agreement providing for nondisclosure and limited use of the information.

8. AREVA NP policy requires that proprietary information be kept in a secured file or area and distributed on a need-to-know basis.

9. The foregoing statements are true and correct to the best of my knowledge, information, and belief.

SUBSCRIBED before me this I ,)

day o ,2009. '.

" NOTARy.

- .* A.(.. ,:.' ,,:

SsnK. McCoy ("\",,p *w*

NOTARY PUBLIC, STATE OF WASHI *eON I""I*I*

MY COMMISSION EXPIRES: 1/10/12

AFFIDAVIT STATE OF WASHINGTON )

) ss.

COUNTY OF BENTON )

1. My name is Alan B. Meginnis. I am Manager, Product Licensing, for AREVA NP Inc. and as such I am authorized to execute this Affidavit.

2. I am familiar with the criteria applied by AREVA NP to determine whether certain AREVA NP information is proprietary. I am familiar with the policies established by AREVA NP to ensure the proper application of these criteria.

3. I am familiar with the AREVA NP information contained in the report 51-9119738-000, entitled, "Browns Ferry Unit 1 Cycle 9 MCPR Safety Limit Analysis (120%

OLTP)," dated September 2009 and referred to herein as "Document." Information contained in this Document has been classified by AREVA NP as proprietary in accordance with the policies established by AREVA NP for the control and protection of proprietary and confidential information.

4. This Document contains information of a proprietary and confidential nature and is of the type customarily held in confidence by AREVA NP and not made available to the public. Based on my experience, I am aware that other companies regard information of the kind contained in this Document as proprietary and confidential.

5. This Document has been made available to the U.S. Nuclear Regulatory Commission in confidence with the request that the information contained in this Document be withheld from public disclosure. The request for withholding of proprietary information is made in accordance with 10 CFR 2.390. The information for which withholding from disclosure is

requested qualifies under 10 CFR 2.390(a)(4) "Trade secrets and commercial or financial information."

6. The following criteria are customarily applied by AREVA NP to determine whether information should be classified as proprietary:

(a) The information reveals details of AREVA NP's research and development plans and programs or their results.

(b) Use of the information by a competitor would permit the competitor to significantly reduce its expenditures, in time or resources, to design, produce, or market a similar product or service.

(c) The information includes test data or analytical techniques concerning a process, methodology, or component, the application of which results in a competitive advantage for AREVA NP.

(d) The information reveals certain distinguishing aspects of a process, methodology, or component, the exclusive use of which provides a competitive advantage for AREVA NP in product optimization or marketability.

(e) The information is vital to a competitive advantage held by AREVA NP, would be helpful to competitors to AREVA NP, and would likely cause substantial harm to the competitive position of AREVA NP.

The information in the Document is considered proprietary for the reasons set forth in paragraphs 6(b), 6(d) and 6(e) above.

7. In accordance with AREVA NP's policies governing the protection and control of information, proprietary information contained in this Document have been made available, on a limited basis, to others outside AREVA NP only as required and under suitable agreement providing for nondisclosure and limited use of the information.

8. AREVA NP policy requires that proprietary information be kept in a secured file or area and distributed on a need-to-know basis.

9. The foregoing statements are true and correct to the best of my knowledge, information, and belief.

% N SUBSCRIBED before me this 3 day of' 2009.

Susan K. McCoy NOTARY PUBLIC, STATE OF WASHINGTON MY COMMISSION EXPIRES: 1/10/12

AFFIDAVIT STATE OF WASHINGTON )

) ss.

COUNTY OF BENTON )

1. My name is Alan B. Meginnis. I am Manager, Product Licensing, for AREVA NP Inc. and as such I am authorized to execute this Affidavit.

2. I am familiar with the criteria applied by AREVA NP to determine whether certain AREVA NP information is proprietary. I am familiar with the policies established by AREVA NP to ensure the proper application of these criteria.

3. I am familiar with the AREVA NP information contained in the report ANP-2638P, Revision 2, entitled, "Applicability of AREVA NP BWR Methods to Extended Power Uprate Conditions," dated October 2009 and referred to herein as "Document." Information contained in this Document has been classified by AREVA NP as proprietary in accordance with the policies established by AREVA NP for the control and protection of proprietary and confidential information.

4. This Document contains information of a proprietary and confidential nature and is of the type customarily held in confidence by AREVA NP and not made available to the public. Based on my experience, I am aware that other companies regard information of the kind contained in this Document as proprietary and confidential.

5. This Document has been made available to the U.S. Nuclear Regulatory Commission in confidence with the request that the information contained in this Document be withheld from public disclosure. The request for withholding of proprietary information is made in accordance with 10 CFR 2.390. The information for which withholding from disclosure is

requested qualifies under 10 CFR 2.390(a)(4) "Trade secrets and commercial or financial information."

6. The following criteria are customarily applied by AREVA NP to determine whether information should be classified as proprietary:

(a) The information reveals details of AREVA NP's research and development plans and programs or their results.

(b) Use of the information by a competitor would permit the competitor to significantly reduce its expenditures, in time or resources, to design, produce, or market a similar product or service.

(c) The information includes test data or analytical techniques concerning a process, methodology, or component, the application of which results in a competitive advantage for AREVA NP.

(d) The information reveals certain distinguishing aspects of a process, methodology, or component, the exclusive use of which provides a competitive advantage for AREVA NP in product optimization or marketability.

(e) The information is vital to a competitive advantage held by AREVA NP, would be helpful to competitors to AREVA NP, and would likely cause substantial harm to the competitive position of AREVA NP.

The information in the Document is considered proprietary for the reasons set forth in paragraphs 6(b), 6(d) and 6(e) above.

7. In accordance with AREVA NP's policies governing the protection and control of information, proprietary information contained in this Document have been made available, on a limited basis, to others outside AREVA NP only as required and under suitable agreement providing for nondisclosure and limited use of the information.

8. AREVA NP policy requires that proprietary information be kept in a secured file or area and distributed on a need-to-know basis.

9. The foregoing statements are true and correct to the best of my knowledge, information, and belief.

i4-~ y SUBSCRIBED before me this __

day of C) a o " ,2009. , K

" "* NOTAR

- - PUBLIC -

Susan K. McCoy NOTARY PUBLIC, STATE OF WASHINGTON MY COMMISSION EXPIRES: 1/10/12

A FFID A VI T STATE OF WASHINGTON )

) Ss.

COUNTY OF BENTON )

1. My name is Alan B. Meginnis. I am Manager, Product Licensing, for AREVA NP Inc. and as such I am authorized to execute this Affidavit.

2. I am familiar with the criteria applied by AREVA NP to determine whether certain AREVA NP information is proprietary. I am familiar with the policies established by AREVA NP to ensure the proper application of these criteria.

3. I am familiar with the AREVA NP information contained in the report ANP-2864(P), Revision 2, entitled, "Browns Ferry Unit 1 Cycle 9 Reload Safety Analysis," dated October 2009 and referred to herein as "Document." Information contained in this Document has been classified by AREVA NP as proprietary in accordance with the policies established by AREVA NP for the control and protection of proprietary and confidential information.

4. This Document contains information of a proprietary and confidential nature and is of the type customarily held in confidence by AREVA NP and not made available to the public. Based on my experience, I am aware that other companies regard information of the kind contained in this Document as proprietary and confidential.

5. This Document has been made available to the U.S. Nuclear Regulatory Commission in confidence with the request that the information contained in this Document be withheld from public disclosure. The request for withholding of proprietary information is made in accordance with 10 CFR 2.390. The information for which withholding from disclosure is

requested qualifies under 10 CFR 2.390(a)(4) "Trade secrets and commercial or financial information."

6. The following criteria are customarily applied by AREVA NP to determine whether information should be classified as proprietary:

(a) The information reveals details of AREVA NP's research and development plans and programs or their results.

(b) Use of the information by a competitor would permit the competitor to significantly reduce its expenditures, in time or resources, to design, produce, or market a similar product or service.

(c) The information includes test data or analytical techniques concerning a process, methodology, or component, the application of which results in a competitive advantage for AREVA NP.

(d) The information reveals certain distinguishing aspects of a process,,

methodology, or component, the exclusive use of which provides a competitive advantage for AREVA NP in product optimization or marketability.

(e) The information is vital to a competitive advantage held by AREVA NP, would be helpful to competitors to AREVA NP, and would likely cause substantial harm to the competitive position of AREVA NP.

The information in the Document is considered proprietary for the reasons set forth in paragraphs 6(b), 6(d) and 6(e) above.

7. In accordance with AREVA NP's policies governing the protection and control of information, proprietary information contained in this Document have been made available, on a limited basis, to others outside AREVA NP only as required and under suitable agreement providing for nondisclosure and limited use of the information.

8. AREVA NP policy requires that proprietary information be kept in a secured file or area and distributed on a need-to-know basis.

9. The foregoing statements are true and correct to the best of my knowledge, information, and belief.

A SUBSCRIBED before me this 9 day of , . ,2009.

Susan K. McCoy

• NOTARY PUBLIC, STATE OF VC-INGTON MY COMMISSION EXPIRES: 1/10/12

/

AFFIDAVIT STATE OF WASHINGTON )

) ss.,

COUNTY OF BENTON )

1. My name is Alan B. Meginnis. I am Manager, Product Licensing, for AREVA NP Inc. and as such I am authorized to execute this Affidavit.

2. I am familiar with the criteria applied by AREVA NP to determine whether certain AREVA NP information is proprietary. I am familiar with the policies established by AREVA NP to ensure the proper application of these criteria.

3. I am familiar with the AREVA NP information contained in Attachment 20 of the TVA Letter to NRC entitled, "Technical Specification Change TS-467 - Utilization of AREVA Fuel and Associated Analysis Methodologies," dated October 23, 2009 and referred to herein as "Document." Information contained in this Document has been classified by AREVA NP as proprietary in accordance with the policies established by AREVA NP for the control and protection of proprietary and confidential information.

4. This Document contains information of a proprietary and confidential nature and is of the type customarily held in confidence by AREVA NP and not made available to the public. Based on my experience, I am aware that other companies regard information of the kind contained in this Document as proprietary and confidential.

5. This Document has been made available to the U.S. Nuclear Regulatory Commission in confidence with the request that the information contained in this Document be withheld from public disclosure. The request for withholding of proprietary information is made in accordance with 10 CFR 2.390. The information for which withholding from disclosure is

requested qualifies under 10 CFR 2.390(a)(4) "Trade secrets and commercial or financial information."

6. The following criteria are customarily applied by AREVA NP to determine whether information should be classified as proprietary:

(a) The information reveals details of AREVA NP's research and development plans and programs or their results.

(b) Use of the information by a competitor would permit the competitor to significantly reduce its expenditures, in time or resources, to design, produce, or market a similar product or service.

(c) The information includes test data or analytical techniques concerning a process, methodology, or component, the application of which results in a competitive advantage for AREVA NP.

(d) The information reveals certain distinguishing aspects of a process, methodology, or component, the exclusive use of which provides a competitive advantage for AREVA NP in product optimization or marketability.

(e) The information is vital to a competitive advantage held by AREVA NP, would be helpful to competitors to AREVA NP, and would likely cause substantial harm to the competitive position of AREVA NP.

The information in the Document is considered proprietary for the reasons set forth in paragraphs 6(b), 6(d) and 6(e) above.

7. In accordance with AREVA NP's policies governing the protection and control of information, proprietary information contained in this Document have been made available, on a limited basis, to others outside AREVA NP only as required and under suitable agreement providing for nondisclosure and limited use of the information.

8. AREVA NP policy requires that proprietary information be kept in a secured file or area and distributed on a need-to-know basis.

9. The foregoing statements are true and correct to the best of my knowledge, information, and belief.

A=

SUBSCRIBED efore me this 3 day of() o-- 12009.

Susan K. McCoy-NOTARY PUBLIC, STATE OF WAS GTON MY COMMISSION EXPIRES: 1/10/12

Enclosure Browns Ferry Nuclear Plant (BFN)

Unit I Technical Specifications (TS) Change 467 Revision of Technical Specifications to Allow Utilization of AREVA Fuel and Associated Analytical Methodologies 1.0

SUMMARY

DESCRIPTION This evaluation supports a request to amend Operating License DPR-33 for BFN Unit 1. The proposed changes would revise the Operating License to allow the use of AREVA fuel and analytical methodologies for BFN Unit 1. Unit 1 will transition from using Global Nuclear Fuel's (GNF) GE14 design, to using the AREVA ATRIUM-10 fuel design commencing with the reload batch delivered in the fall of 2010.

2.0 DETAILED DESCRIPTION The Tennessee Valley Authority (TVA) intends to begin utilizing the ATRIUM-10 design in BFN Unit 1 Cycle 9. The first reload of ATRIUM-10 targeted for insertion into the core is the fall 2010 outage. The ATRIUM-10 product is an industry proven fuel design in use at BFN 2 and BFN 3 since 2005 and 2004, respectively. The initial Unit 1 reload, and at least one follow on reload, will utilize Blended Low Enriched Uranium (BLEU) provided to TVA under a joint project with the Department of Energy. However, TVA may also elect to utilize ATRIUM-10 fuel in Unit 1 with standard commercial grade uranium in future reloads.

In order to extend the use of this fuel design to BFN Unit 1, several changes to the Technical' Specifications (TS) are required. TS 5.6.5.b address the analytical methods which may be used to determine input to the Core Operating Limits Report (COLR). Currently, the BFN Unit 1 specification only includes GNF analytical methods. Unit 1 TS 5'6.5.b will be revised to add appropriate NRC approved AREVA analytical methodologies.

Also, TS 3.2.3 (Linear Heat Generation Rate (LHGR)) requires an administrative correction.

Word processing of a previous change caused the header to incorrectly state "APRM Gain and Setpoints," (instead of LHGR). The header and section number are corrected. The change is:

editorial in nature and has no impact on public health and safety and no impact on the environment.

In addition, two other TS changes will be made to reflect the manner by which AREVA methodologies monitor and enforce thermal limits. The affected TS sections are 3.3.4.1 (End of Cycle Recirculation Pump Trip (EOC-RPT) Instrumentation), and 3.7.5 (Main Turbine Bypass System); both are modified to require a linear heat generation rate limit adjustment when operating with EOC-RPT out of service, operating with a single recirculation loop, and operating with turbine bypass out of service, respectively.

1

Finally, TS 2.1.1.2 (Minimum Critical Power Ratio Safety Limit) includes revised values (dual loop and single loop) resulting from cycle specific analysis of the transition cycle design at 120%

original licensed thermal power (OLTP). Note the TS 2.1.1.2 page markup in Attachment 2 is based on the current Technical Specification as of the date of this submittal.

The submittal also addresses the required changes to the Technical Specification Bases.

Changes are related to adding information pertaining to AREVA analytical methodologies (including Reference documents), and information related to AREVA specific monitoring and enforcement of fuel thermal limits.

The previous AREVA fuel transition submittal for BFN (Reference 1) addressed Unit 1 to the extent of providing a description of the AREVA fuel (TS 4.2.1, Reactor Core - Fuel Assemblies),

and to modify the fuel storage criticality requirement to a k-effective basis (TS 4.3.1, Fuel Storage - Criticality). Unit 1 was included in this prior change to allow for the possibility of storing AREVA fuel bundles in the BFN Unit 1 spent fuel pool. Consequently, these two TSs do not require alteration, and are not included in the current change request.

In a meeting with the NRC staff on January 28, 2009, the overall approach for the BFN Unit 1 fuel transition submittal was discussed. In addition to providing guidance on submittal timing, the NRC provided a list of eleven technical items to be addressed in the submittal, per Reference 2. A follow-up meeting was held on March 16, 2009 in which the specific contents of the transition submittal were agreed upon per Reference 3. In addition to the eleven items mentioned above, the NRC requested certain AREVA reload documents pertaining to the design and licensing analyses of the transition cycle, as well as selected generic reports related to methodologies, be included in the submittal. The NRC also provided a specific list of prior Requests for Additional Information (RAIs), which should be answered for Unit 1 (addressing the co-resident GNF fuel impacts as appropriate). Responses to prior RAIs, are addressed in three of the attachments to this submittal (Attachments 12, 18, and 20). Attachment 16 provides responses to technical items identified in Reference 2, along with information on BLEU material.

Specific information requested by the NRC is included in the following attachments.

Attachment Description 1 List of Regulatory Commitments 2 Proposed Technical Specifications Changes (Mark-up) 3 Retyped Proposed Technical Specifications Pages 4 Proposed Technical Specification Bases Changes (Mark-up) 5 Retyped Proposed Technical Specification Bases Pages 6 Mechanical Design Report (proprietary)

ANP-2833(P), Revision 0, Mechanical Design Report for Browns Ferry Unit 1 Reload BFE1-9 ATRIUM-10 Fuel Assemblies, AREVA NP Inc., September 2009.

7 Mechanical Design Report (non-proprietary)

ANP-2833(NP), Revision 0, Mechanical Design Report for Browns Ferry Unit 1 Reload BFE1-9 ATRIUM-1 0 Fuel Assemblies, AREVA NP Inc., September 2009.

8 Thermal Hydraulic Design Report (proprietary)

ANP-2807(P), Revision 0, Browns Ferry Unit 1 Thermal-Hydraulic Design Report for ATRIUM-1 0 Fuel Assemblies, AREVA NP Inc., June 2009.

2

Description Attachment Description 9 Thermal Hydraulic Design Report (non-proprietary)

ANP-2807(NP), Revision 0, Browns Ferry Unit 1 Thermal-Hydraulic Design Report for ATRIUM-10 Fuel Assemblies, AREVA NP Inc., June 2009.

10 Fuel Cycle Design Report (proprietary)

ANP-2850(P), Revision 0, Browns Ferry Unit 1 Fuel Cycle Design, AREVA NP Inc., July 2009.

11 Fuel Cycle Design Report (non-proprietary)

ANP-2850(NP), Revision 0, Browns Ferry Unit 1 Fuel Cycle Design, AREVA NP Inc., August 2009.

12 Reload Safety Analysis Report (proprietary)

ANP-2864(P), Revision 2, Browns Ferry Unit 1 Cycle Reload Safety Analysis, AREVA NP Inc.,

October 2009.

13 Reload Safety Analysis Report (non-proprietary)

ANP-2864(NP), Revision 2, Browns Ferry Unit 1 Cycle Reload Safety Analysis, AREVA NP Inc.,

October 2009.

14 LOCA Break Spectrum Analysis Report (proprietary)

EMF-2950(P), Revision 2, Browns Ferry Units 1, 2, and 3 Extended Power Uprate LOCA Break Spectrum Analysis, AREVA NP Inc., August 2009.

15 LOCA Break Spectrum Analysis Report (non-proprietary)

EMF-2950(NP), Revision 0, Browns Ferry Units 1, 2, and 3 Extended Power Uprate LOCA Break Spectrum Analysis, AREVA NP Inc., August 2009.

16 Response to NRC Comments Regarding Browns Ferry Unit 1 Proposed Fuel Transition Amendment (non-proprietary) 51-9121503-002, Response to NRC Comments Regarding Browns Ferry Unit 1 Proposed Fuel Transition Amendment, AREVA NP Inc., October 2009.

17 Boiling Water ReactorLicensing Methodology Compendium (non-proprietary)

ANP-2637, Revision 2, Boiling Water Reactor Licensing Methodology Compendium, AREVA NP Inc., December 2007.

18 Applicability of AREVA NP BWR Methods to Extended Power Uprate Conditions (proprietary)

ANP-2638(P), Revision 2, Applicability of AREVA NP BWR Methods to Extended Power Uprate Conditions, AREVA NP Inc., October 2009.

19 Applicability of AREVA NP BWR Methods to Extended Power Uprate Conditions (non-proprietary)

ANP-2638(NP), Revision 2, Applicability of AREVA NP BWR Methods to Extended Power Uprate Conditions, AREVA NP Inc., October 2009.

20 Part 1: Previous NRC Requests for Additional Information Matrix and Text Part 2: Browns Ferry Unit 1 - Summary of Response to Requests for Additional Information (proprietary)

ANP-2860(P), Revision 2, Browns Ferry Unit 1 - Summary of Responses to Requests for Additional Information, AREVA NP Inc., October 2009.

21 Part 1: Previous NRC Requests for Additional Information Matrix and Text Part 2: Browns Ferry Unit 1 - Summary of Response to Requests for Additional Ihformation (non-proprietary)

ANP-2860(NP.), Revision 2, Browns Ferry Unit 1 - Summary of Responses to Requests for Additional Information, AREVA NP Inc., October 2009.

22 Safety Limit Minimum Critical Power Ratio (proprietary) 51-9119738-000, Browns Ferry Unit 1 Cycle 9 MCPR Safety Limit Analysis (120% OLTP),

AREVA NP Inc., September 2009.

23 Safety Limit Minimum Critical Power Ratio (non-proprietary) 51-9121246-000, Browns Ferry Unit 1 Cycle 9 MCPR Safety Limit Analysis (120% OLTP),

AREVA NP Inc., September 2009.

24 Affidavits 3

Attachment information is based on Extended Power Uprate (EPU, 120% OLTP operations),

consistent with previous meetings between TVA and NRC representatives, documented in :

References 2 and 3. As part of the meeting documented in Reference 3, representatives also discussed the potential to provide supplemental information supporting a 105% OLTP core design option. At the time of the Reference 3 meeting, TVA proposed to have all submittal material to NRC by the end of December. TVA and AREVA took actions to determine if an earlier date could be supported. A subsequent proposal to NRC indicated TVA would provide the EPU portion of the submittal first, followed up shortly thereafter with additional material supporting a 105% OLTP core design option.

A related public meeting was held between TVA and NRC representatives on August 11, 2009 to discuss the status of the TVA's pending EPU submittal. During the meeting, representatives discussed the potential impact of delaying the EPU submittal approval, in the context of the BFN Unit 1 fuel transition. To address the possibility of a 105% OLTP based fuel transition, TVA stated that a 120% OLTP based submittal would be made initially, with a submittal supplement to be provided by the end of December 2009 supporting 105% OLTP. NRC concurred with TVA's position to provide supplemental information based on the current licensed thermal power (105% OLTP).

Much of the information in the current submittal applies to both 120% and 105% OLTP core designs. Consequently, it is of value for the NRC to receive it now, such that a technical review can commence. Future supplemental information (specific to 105% OLTP) will include as a minimum: Fuel Cycle Design, Reload Safety Analysis, and Thermal Hydraulic Design reports.

In accordance with discussions between TVA and NRC representatives, the fuel designs used for the 105% OLTP core design are a subset of the fresh bundle types developed for the 120%

OLTP core design. Bundle design information being submitted as part of this license amendment request package will also be applicable to the supplemental 105% OLTP core design, helping to minimize any impact on NRC review time due to the submittal supplement.

All critical power results provided in the submittal (Attachments 8, 9, 10, 11, 12, 13, 22, & 23) are based on corrected additive constants approved by NRC in the addendum discussed above.

Upon issuance of a revision to EMF-2209(P)(A), TVA will provide a revision to ANP-2637(Attachment 17), referencing the new version. ANP-2637 (Attachment 17) identifies Reference 8 as an approved methodology report. Reference 8 is the approved version at the time of this submittal. There was an outstanding 10 CFR 21, "Reporting of Defects and

• Noncompliance," issue related to Reference 8. The issue had to do with reported ATRIUM-t:0 additive constants values having been found non-conservative. AREVA submitted the Reference 9 addendum, to Reference 8, correcting ATRIUM-10 additive constants values. The NRC has approved the Reference 9 correction per Reference 13.

A second 10 CFR 21 issue was recently identified in Reference 10. The issue of operating limit error is related to the fact that LaSalle operates with Zinc levels well beyond the industry standard set by Electric Power Research Institute (EPRI) guidance in References 11 and 12.

LaSalle measured unusually high liftoff levels, which were attributed to operating water chemistry with high levels of Zinc. All BFN units operate within the EPRI water chemistry guidance, and measured BFN liftoff levels remain consistent with AREVA methodology assumptions. Therefore, this particular 10 CFR 21 issue is not applicable to any BFN unit.

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3.0 TECHNICAL EVALUATION

The fuel design to be introduced into Unit 1 in 2010 is the AREVA ATRIUM-1 0 product. This design utilizes a 10x10 array of fuel rods, with eighty-three full length fuel rods and eight partial length fuel rods. The partial length fuel rods are approximately two-thirds the length of the full length fuel rods. The use of partial length rods improves fuel utilization in the high void upper region of the bundle, and also enhances cold shutdown margin, stability, and pressure drop performance.

The ATRIUM-10 design does not utilize tie rods as the structural tie between the upper and lower tie plates. Instead, the design uses a central water channel, having a mechanical connection to the two tie plates. The central water channel carries the mechanical loads during fuel handling. It displaces a 3x3 array of fuel rods within the bundle, and serves to improve fuel economy by improving internal neutron moderation. The lower ends of the fuel rods rest on top of the lower tie plate, with their lower ends laterally restrained by a spacer grid located just above the lower tie plate. No expansion springs are required on each fuel rod, because a single, large reaction spring is used on the central water channel to hold the upper tie plate in the latched position. The ATRIUM-10 design uses a total of eight fuel rod spacers to provide lateral support for the fuel rods, and to enhance thermal hydraulic performance. The ATRIUM-10 design to be employed at Unit 1 utilizes a debris resistant lower tie plate to limit introduction of foreign material into the assembly from below.

The ATRIUM-10 design was developed using the thermal mechanical design bases and limits outlined in Reference 4. Compliance with Reference 4 ensures the fuel design meets the fuel system damage, fuel failure, and fuel coolability criteria identified in the Reference 5 Standard Review Plan. The NRC reviewed and approved (per Reference 6) the use of Reference 4 for making changes and improvements to fuel designs; specifically stating such changes and improvements do not require specific NRC review and approval, provided the criteria are satisfied. The ATRIUM-10 design fully complies with the criteria of Reference 4, and therefore meets all of the required fuel licensing criteria in the Reference 5 Standard Review Plan.

Changes to the Updated Final Safety Analysis Report (UFSAR), required as a result of implementing AREVA ATRIUM-10 fuel, were previously addressed during the BFN Units 2 & 3 AREVA fuel transition. Changes to the following UFSAR sections were made during the initial implementation for BFN Units 2 and 3:

• Section 3.2 Fuel Mechanical Design

• Section 3.3 Reactor Vessel Internals Mechanical Design

• Section 3.6 Nuclear Design

• Section 3.7 Thermal and Hydraulic Design

• Section 6.5 Safety Evaluation

• Section 13.10 Refueling Test Program

• Section 14.4 Approach to Safety Analysis

• Section 14.5 Analyses of Abnormal Operational Transients

• Section 14.6 Analysis of Design Basis Accidents Given the UFSAR applies to all three units, and the AREVA fuel product for Unit 1 is the same design used in Units 2 and'3, introduction of AREVA ATRIUM-10 fuel into Unit I does not require any changes to the UFSAR.

5

The AREVA analytical methods and topical reports to be added to Technical Specification 5.6.5.b are those utilized to evaluate the fuel mechanical design, along with both cycle dependent and independent safety analyses, used to establish limits identified in the COLR.

Additionally, Reference 4 is also being added to the Technical Specifications as the basis for acceptance of the ATRIUM-10 fuel design.

Each analytical methodology being added to Technical Specification 5.6.5.b has been previously reviewed and approved by the NRC. In August 2008, the NRC staff performed additional reviews of the AREVA analytical methods, specifically with EPU application in mind.

The review concluded AREVA methodologies are adequate for application to EPU conditions, with two exceptions. These exceptions are related to the impact on calculated vessel overpressure arising from potential void quality correlation uncertainties at higher void conditions. Information on how this concern is addressed is contained in Attachment 18.

The impact of the ATRIUM-10 design on the UFSAR accident analyses will be accounted for by cycle specific reload and accident analyses. Limiting transients from UFSAR Chapter 14 categories of pressure increase events, vessel water temperature decrease events, control rod withdrawal error events, core flow increase events, and increase in vessel inventory events are evaluated each cycle. Limiting analyses results, for the transition cycle, are presented in 2./

Introduction of the ATRIUM-10 design fuel will not adversely impact UFSAR accident analyses.

AREVA evaluates the control rod drop accident (UFSAR section 14.6.2) on a cycle specific basis. Attachment 12 includes the cycle specific evaluation of the control rod drop accident for the transition cycle. The evaluation shows the number of rods calculated to fail in this event remains well below the value of 850 assumed in the UFSAR radiological evaluation of this event. The doses, from the control rod drop accident, remain within limits required by 10 CFR 50.67, "Accident Source Term," and Regulatory Guide 1.183 (Reference 7).

Regarding the LOCA analysis (UFSAR section 14.6.3), a baseline LOCA break spectrum analysis of ATRIUM-10 fuel was previously performed, covering all three BFN units; it is included as Attachment 14. Cycle specific fuel design MAPLHGR limits are analyzed consistent with assumptions used in the baseline LOCA analysis. Peak cladding temperature, cladding oxidation, and hydrogen generation analyses results of record are included in Attachment 12, The introduction of ATRIUM-10 fuel will not challenge the peak clad temperature, cladding oxidation, or hydrogen generation limits specified in 10 CFR 50.46 "Acceptance criteria for -

emergency core cooling systems for light-water nuclear power reactors," paragraph (b).

The ATRIUM-10 design will also not challenge the UFSAR basis of the refueling accident (UFSAR section 14.6.4). The BFN UFSAR accident is based on a bounding event using a 7x7 fuel design. While the number of rods calculated to fail for an ATRIUM-10 bundle (154) is higher than the number calculated to fail in a 7x7 bundle (111), the activity is allocated over a greater number of rods. The ATRIUM-10 bundle has the equivalent of 88.33 fuel length rods (83 full length plus 8 partial length rods with approximately two thirds the full length), while the 7x7 bundle has 49 full length rods. Therefore, the accident release with ATRIUM-1 0 fuel would be approximately (154/111) x (49/88.33), or 77% of the release from the design basis 7x7 fuel.

The UFSAR radiological analysis of this event accounts for the isotopic inventory from 120%

OLTP operation. Consequently, the fuel handling accident described in the UFSAR remains 7 bounding for ATRIUM-10 fuel. The doses resulting from this event will remain within the limits specified in 10 CFR 50.67.

6

The main steam line break accident (UFSAR section 14.6.5) is not affected by a change in fuel design. As stated in the UFSAR, no fuel failures are expected to occur as a result of this accident. The radionuclide inventory, released from the primary coolant system, is present in the coolant prior to the event; UFSAR section 14.6.5.2.1 provides details regarding the assumed accident inventory. Therefore, the fuel design change does not alter the consequences of a main steam line break accident.

The NRC has previously reviewed and approved transitions from GE14 to ATRIUM-10 (see section 4.1 below). Previous reviews confirmed the acceptability of transitioning from GE14 to ATRIUM-10. The scope of the technical analyses provided in support of the Unit 1 submittal is consistent with, and surpasses, the technical analyses provided with the precedent submittals.

In summary, the ATRIUM-10 fuel design fully complies with applicable fuel licensing criteria provided in Reference 5, as documented in Reference 4. The analytical methodologies to be used for design and licensing of ATRIUM-10 reloads are NRC approved, and acceptable for establishing COLR limits. Application of these methods to EPU conditions will be in compliance with the restrictions identified by the NRC staff during the August 2008 review of the AREVA analytical methods. The proposed changes to Technical Specifications 2.1.1.2, 5.6.5.b, 3.3.4.1, and 3.7.5, are necessary and appropriate to implement the AREVA fuel design, and associated analytical methodologies. Given the prior transition to ATRIUM-10 fuel on BFN Units 2 & 3, the required changes to the UFSAR have already been completed.

4.0 REGULATORY EVALUATION

4.1 PRECEDENT A search of NRC actions on Technical Specification changes revealed the Nuclear Regulatory Commission has previously approved similar changes for the following plants:

• "Browns Ferry Nuclear Plant, Units 2 and 3 - Issuance of Amendments Regarding Core Operating Limits (TAC Nos. MB8433 and MB8434)," December 30, 2003.

(ML033650142) 0 "Safety Evaluation by the Office of Nuclear Reactor Regulation Related to Amendment Nos. 246 and 274 to Renewed Facility Operating Licenses Nos. DPR-71 and DPR-62, Carolina Power & Light Company, Brunswick Steam Electric Plant, Units 1 and 2, Docket Nos. 50-325 and 50-324," March 27, 2008. (ML080870546) 4.2 SIGNIFICANT HAZARDS CONSIDERATIONS This analysis addresses the proposed change to amend Operating License DPR-33 for BFN Unit 1 to allow the use of AREVA fuel and analytical methodologies.

TVA has evaluated whether or not a significant hazards consideration is involved with the proposed amendment(s) by focusing on the three standards set forth in 1 OCFR 50.92, "Issuance of amendment," as discussed below:

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

7

Response: No Changing fuel designs and making an editorial change to TS will not increase the probability of a loss of coolant accident. The fuel cannot increase the probability of a primary coolant system breach or rupture, as there is no interaction between the fuel and the system piping.

The fuel will continue to meet the 10 CFR 50.46 limits for peak clad temperature, oxidation fraction, and hydrogen generation. Therefore, the consequences of a LOCA will not be increased.

Similarly, changing the fuel design and making an editorial change to TS cannot increase the probability of an abnormal operating occurrence (AOO). As a passive component, the fuel does not interact with plant operating or control systems. Therefore, the fuel change cannot affect the initiators of the previously evaluated AOO transient events. Thermal limits for the new fuel will be determined on a reload specific basis, ensuring the specified acceptable fuel design limits continue to be met. Therefore, the consequences of a previously evaluated AOO will not increase.

The refueling accident is potentially affected by a change in fuel design, due td the mechanical interaction between the fuel and the refueling equipment. However, the probability of the refueling accident with ATRIUM-10 fuel is not increased because the upper bail handle is designed to be mechanically compatible with existing fuel handling equipment. The design weight of the ATRIUM-10 design is similar to other designs in use at Browns Ferry, and is well within the design capability of the refueling equipment. The consequences of the refueling accident are similar to the current GEl4 fuel, remaining well within the design basis (7x7 Fuel) evaluation in the UFSAR.

,The probability of a control rod drop accident does not increase because the ATRIUM-10 fuel channel is mechanically compatible with the co-resident fuel, and existing control blade designs. The mechanical interaction and friction forces between the ATRIUM-10 channel,..and control blades, would not be higher than previous designs. In addition, routine plant testing includes confirmation of adequate control blade to control rod drive coupling. The probability of a rod drop accident is not increased with the use of ATRIUM-10 fuel. Control rod drop accident consequences are evaluated on a cycle specific basis, confirming the number of calculated rod failures remains with the UFSAR design basis.

The dose consequences of all the previously evaluated UFSAR accidents remain with the limits of 10 CFR 50.67.

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

Response: No The ATRIUM-10 fuel product has been designed to maintain neutronic, thermal-hydraulic, and mechanical compatibility with the NSSS vendor fuel designs. The ATRIUM-i 0 fuel has been designed to meet fuel licensing criteria specified in NUREG-08000, "Standard Review Plan for Review of Safety Analysis Reports for Nuclear Power Plants." Compliance with these criteria ensures the fuel will not fail in an unexpected manner.

8

A change in fuel design and an editorial change to TS cannot create any new accident initiators because the fuel is a passive component, having no direct influence on the performance of operating plant systems and equipment. Hence, a fuel design change cannot create a new type of malfunction leading to a new or different kind of transient or accident.

Consequently, the proposed fuel design change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

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

Response: No The ATRIUM-10 fuel is designed to comply with the fuel licensing criteria specified in NUREG-0800. Reload specific and cycle independent safety analyses are performed ensuring no fuel failures will occur as the result of abnormal operational transients, and dose consequences for accidents remain with the bounds of 10CFR50.67. All regulatory margins and requirements are maintained.

Therefore, the proposed change does not involve a significant reduction in a margin of safety.

Based on the above, TVA concludes the proposed amendment does not involve a significant hazards consideration under the standards set forth in 10 CFR 50.92 (c), and, accordingly, a finding of "no significant hazards consideration" is justified.

4.3 CONCLUSION

The proposed use of ATRI UM-10 fuel (using BLEU, or commercial grade uranium), and the adoption of AREVA analytical methodologies for BFN Unit 1, is acceptable based on the following:

ATRIUM-10 fuel has been designed so as to comply with the fuel related licensing criteria specified in the Standard Review Plan (Reference 5).

) Analytical methodologies being added to the Technical Specifications have been previously reviewed and approved by NRC.

> Analytical methodologies have been reviewed for EPU application, and found to be . .

acceptable, with the caveat of two restrictions related to vessel overpressure margins.

These two restrictions have been incorporated into Unit 1 transition analyses.

> Transition core design analyses demonstrate acceptability of using ATRIUM-10 in Unit 1, including mixed core compatibility with co-resident GE14 fuel.

> The impacts of BLEU material do not adversely impact the neutronic, thermal-hydraulic, or mechanical performance of the fuel, including analytical methods used to perform these evaluations.

In conclusion, based on the considerations discussed above, (1) there is reasonable assurance the health and safety of the public will not be endangered by operation in the proposed manner, 9

(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.

5.0 ENVIRONMENTAL CONSIDERATION

A review has determined the proposed amendment would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, 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 effluents 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 10 CFR 51.22(c)(1 0). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.

6.0 REFERENCES

1. Letter from R. G. Jones (TVA) to NRC, "Browns Ferry Nuclear Plant (BFN) - Units 1, 2, and 3 - Technical Specifications (TS) Change 421 - Framatome Fuel Design and Storage," dated February 13, 2003.

2. Letter from Ms. Eva A. Brown (NRC) to TVA, "Summary of January 28, 2009, Meeting with the Tennessee Valley Authority Regarding Proposed Fuel Transition Amendment (TAC No.,

ME0438)," dated March 23, 2009.

3. Letter from Ms. Eva A. Brown (NRC) to TVA, "Summary of March 16, 2009, Meeting with the Tennessee Valley Authority Regarding Proposed Fuel Transition Amendment (TAC No.

ME0438)," dated June 3, 2009.

4. ANF-89-98(P)(A) Revision 1 and'Supplement 1, "Generic Mechanical Design Criteria for BWR Designs," Advanced Nuclear Fuels Corporation, dated May 1995.

5. NUREG-0800, "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants," Section 4.2, 'Fuel System Design,' Revision 3, dated March 2007.

6. Letter from R.C. Jones (NRC) to R. Copeland (Siemens Power Corporation), "Acceptance for Referencing of Topical Report ANF-89-98(P), Revision 1, "Generic Mechanical Design Criteria for BWR Fuel Designs," (TAC No. M81070)," dated April 20, 1995.

7. Regulatory Guide 1.183, "Alternative Radiological Source Terms for Evaluating Design.

Basis Accidents at Nuclear Power Reactors," U.S. Nuclear Regulatory Commission, dated July 2000.

8. EMF-2209(P)(A), Revision 2, "SPCB Critical Power Correlation," Framatome ANP, September 2003.

10

9. EMF-2209(P)(A), Revision 2, Addendum 1 Revision 0, "SPCB Additive Constants for ATRIUM-10 Fuel," AREVA NP, April 2008.

10. Letter from R. L. Gardener (AREVA NP, Inc.) to Document Control Desk (NRC),

NRC:09:092, "10 CFR Part 21 Notification of an Error in LaSalle Units 1 & 2 Power Dependent MCPR and LHGR Operating Limits Calculation Due to High Measured Liftoff,"

AREVA NP Inc., dated August 27, 2009.

11. 1008192, "BWRVIP-130: BWR Vessel and Internals Project, BWR Water Chemistry Guidelines - 2004," EPRI, October 2004.

12. 1016579, "BWRVIP-190: BWR Vessel and Internals Project, BWR Water Chemistry Guidelines - 2008," EPRI, October 2008.

13. Letter from Thomas B. Blount (NRC) to R. L. Gardener (AREVA NP, Inc.), Final Safety Evaluation for AREVA NP, Inc. (AREVA) Topical Reports (TR) EMF-2209P, Revision 2, Addendum 1, "SPCB Additive Constants For ATRIUM-10 Fuel," and ANP-10249 (P),

Revision 0, Supplement 1, "ACE Additive Constants For ATRIUM -10 Fuel," (TAC Nos.

MD8754 AND ME0162), dated September 23, 2009.

11

ATTACHMENT I Browns Ferry Nuclear Plant (BFN)

Unit 1 Technical Specifications (TS) Change 467 Revision of Technical Specifications to allow utilization of AREVA NP fuel and associated analysis methodologies Regulatory Commitments None

ATTACHMENT 2 Browns Ferry Nuclear Plant (BFN)

Unit 1 Technical Specifications (TS) Change 467 Revision of Technical Specifications to allow utilization of AREVA NP fuel and associated analysis methodologies Proposed Technical Specifications Changes (Mark-up)

The following pages have been revised to reflect the proposed changes. On the affected pages a line has been drawn through the deleted text and new or revised text is shaded.

SLs 2.0 2.0 SAFETY LIMITS (SLs) 2.1 SLs 2.1.1 Reactor Core SLs 2.1.1.1 With the reactor steam dome pressure < 785 psig or core flow

< 10% rated core flow:

THERMAL POWER shall be *25% RTP.

2.1.1.2 With the reactor steam dome pressure _>785 psig and core flow

_>10% rated core flow:

MCPR shall be > 1.07 for two recirculation loop operation or > 1.10 for single loop operation.

2.1.1.3 Reactor vessel water level shall be greater than the top of active irradiated fuel.

2.1.2 Reactor Coolant System Pressure SL Reactor steam dome pressure shall be _ 1325 psig.

2.2 SL Violations With any SL violation, the following actions shall be completed within 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />s:

2.2.1 Restore compliance with all SLs; and 2.2.2 Insert all insertable control rods.

BFN-UNIT 1 2.0-1 Amendment No. 236, 267 Februr*y 06, 2007

LHGRAPRM -Gainand Setpoin, 3.2.43 3.2 POWER DISTRIBUTION LIMITS 3.2.3 LINEAR HEAT GENERATION RATE (LHGR)

LCO 3.2.3 All LHGRs shall be less than or equal to the limits specified in the COLR.

APPLICABILITY: THERMAL POWER Ž 25% RTP.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Any LHGR not within A.1 Restore LHGR(s) to within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> limits, limits.

B. Required Action and B.1 Reduce THERMAL 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> associated Completion POWER to < 25% RTP.

Time not met.

BFN-UNIT 1 3.2-5 Amendment No. 2-34

LHGRAPRM Gain and Setpoit, 3.2.43 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.2.3.1 Verify all LHGRs are less than or equal to the Once within limits specified in the COLR. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after

Ž> 25% RTP AND 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> thereafter BFN-UNIT 1 3.2-6 Amendment No. 234

EOC-RPT Instrumentation 3.3.4.1 3.3 INSTRUMENTATION 3.3.4.1 End of Cycle Recirculation Pump Trip (EOC-RPT) Instrumentation LCO 3.3.4.1 a. Two channels per trip system for each EOC-RPT instrumentation Function listed below shall be OPERABLE:

1. Turbine Stop Valve (TSV) - Closure; and

2. Turbine Control Valve (TCV) Fast Closure, Trip Oil Pressure

- Low.

OR

b. LCO 3.2.2, "MINIMUM CRITICAL POWER RATIO (MCPR),"

limits for inoperable EOC-RPT as specified in the COLR are made applicable; and

c. LCO 3.2.3, "LINEAR HEAT GENERATION RATE (LHGR),"

limits for an inoperable EOC-RPT, as specified in the COLR, are made applicable.

APPLICABILITY: THERMAL POWER > 30% RTP.

BFN-UNIT 1 3.3-29 Amendment No. 234

EOC-RPT Instrumentation 3.3.4.1 ACTIONS

---

NOTE-Separate Condition entry is allowed for each channel.

CONDITION REQUIRED ACTION COMPLETION TIME A. One or more channels A.1 Restore channel to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> inoperable. OPERABLE status.

OR A.2 -------- NOTE------

Not applicable if inoperable channel is the result of an inoperable breaker.

Place channel in trip. 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> B. One or more Functions B.1 Restore EOC-RPT trip 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> with EOC-RPT trip capability.

capability not maintained.

OR AND B.2 Apply t-he MCPR and 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> MCPR and LHGR limits LHGR limits for for inoperable EOC-RPT inoperable EOC-RPT as not made applicable, specified in the COLR.

C. Required Action and C.1 Reduce THERMAL 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> associated Completion POWER to < 30% RTP.

Time not met.

BFN-UNIT 1 3.3-30 Amendment No. 2U4

Main Turbine Bypass System 3.7.5 3.7 PLANT SYSTEMS 3.7.5 Main Turbine Bypass System LCO 3.7.5 The Main Turbine Bypass System shall be OPERABLE.

OR The following limits are made applicable:

a. LCO 3.2.1, "AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR)," limits for an inoperable Main Turbine Bypass System, as specified in the COLR; and

b. LCO 3.2.2, "MINIMUM CRITICAL POWER RATIO (MCPR),"

limits for an inoperable Main Turbine Bypass System, as specified in the COLR; and

c. LCO 3.2.3, "LINEAR HEAT GENERATION RATE (LHGR),"

limits for an inoperable Main Turbine Bypass System, as specified in the COLR.

APPLICABILITY: THERMAL POWER >_25% RTP.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Requirements of the LCO A.1 Satisfy the requirements 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> not met. of the LCO.

B. Required Action and B.1 Reduce THERMAL 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> associated Completion POWER to < 25% RTP.

Time not met.

BFN-UNIT 1 3.7-16 Amendment No. 2-4

Reporting Requirements 5.6 5.6.4 (Deleted).

5.6.5 CORE OPERATING LIMITS REPORT (COLR)

a. Core operating limits shall be established prior to each reload cycle, or prior to any remaining portion of a reload cycle, and shall be documented in the COLR for the following:

(1) The APLHGRs for Specification 3.2.1; (2) The LHGR for Specification 3.2.3; (3) The MCPR Operating Limits for Specification 3.2.2; and (4) The RBM setpoints and applicable reactor thermal power ranges for each of the setpoints for Specification 3.3.2.1, Table 3.3.2.1-1.

b. The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents (latest approved versions applicable to BFN): NEDE 24011 P A, "General Eectric Standard App'icatoen for Reactor Fuel," (latest appro'.ed version for BF-N)-.

1. NEDE-2401 1-P-A, General Electric Standard Application for Reactor Fuel.

2. XN-NF-81-58(P)(A), RODEX2 Fuel Rod Thermal-Mechanical Response Evaluation Model.

3. XN-NF-85-67(P)(A), Generic Mechanical Design for Exxon Nuclear Jet Pump BWR Reload Fuel.

4. EMF-85-74(P)(A), RODEX2A (BWR) Fuel Rod Thermal-Mechanical Evaluation Model.

5. ANF-89-98(P)(A), Generic Mechanical Design Criteria for BWR Fuel Designs.

(continued)

BFN-UNIT 1 5.0-24 Amendment No. 234,-239, 2.2 January 25, 2005

Reporting Requirements 5.6 5.6 Reporting Requirements (continued)

6. XN-NF-80-19(P)(A) Volume 1, Exxon Nuclear Methodology for Boiling Water Reactors - Neutronic Methods for Design and Analysis.

7. XN-NF-80-19(P)(A) Volume 4, Exxon Nuclear Methodology for Boiling Water Reactors: Application of the ENC Methodology to BWR Reloads.

8. EMF-2158(P)(A), Siemens Power Corporation Methodology for Boiling Water Reactors: Evaluation and Validation of CASMO-4/MICROBURN-B2.

9. XN-NF-80-19(P)(A) Volume 3, Exxon Nuclear Methodology for Boiling Water Reactors, THERMEX: Thermal Limits Methodology Summary Description.

10. XN-NF-84-105(P)(A) Volume 1, XCOBRA-T: A Computer Code for BWR Transient Thermal-Hydraulic Core Analysis.

11 .ANF-524(P)(A), ANF Critical Power Methodology for Boiling Water Reactors.

12.ANF-913(P)(A) Volume 1, COTRANSA2: A Computer Program for Boiling Water Reactor Transient Analyses.

13.ANF-1358(P)(A), The Loss of Feedwater Heating Transient in Boiling Water Reactors.

14. EMF-2209(P)(A), SPCB Critical Power Correlation.

15. EMF-2245(P)(A), Application of Siemens Power Corporation's Critical Power Correlations to Co-Resident Fuel.

16. EMF-2361 (P)(A), EXEM BWR-2000 ECCS Evaluation Model.

17. EMF-2292(P)(A), ATRIUMTM-10: Appendix K Spray Heat Transfer Coefficients.

The COLR will contain the complete identification for each of the Technical Specification referenced topical reports used to prepare the COLR (i.e., report number, title, revision, date and any supplements)

(continued)

BFN-UNIT 1 5.0-24a Amendment No. 234,239,-2 January 25, 2005

ATTACHMENT 3 Browns Ferry Nuclear Plant (BFN)

Unit I Technical Specifications (TS) Change 467 Revision of Technical Specifications to allow utilization of AREVA NP fuel and associated analysis methodologies Retyped Proposed Technical Specifications Pages The following pages have been revised to reflect the proposed changes. These are the retyped pages relative to the markups found in Attachment 2.

SLs 2.0 2.0 SAFETY LIMITS (SLs) 2.1 SLs 2.1.1 Reactor Core SLs 2.1.1.1 With the reactor steam dome pressure < 785 psig or core flow

< 10% rated core flow:

THERMAL POWER shall be *<25% RTP.

2.1.1.2 With the reactor steam dome pressure > 785 psig and core flow

_>10% rated core flow:

MCPR shall be Ž>1.07 for two recirculation loop operation or Ž_1.10 for single loop operation.

2.1.1.3 Reactor vessel water level shall be greater than the top of active irradiated fuel.

2.1.2 Reactor Coolant System Pressure SL Reactor steam dome pressure shall be < 1325 psig.

2.2 SL Violations With any SL violation, the following actions shall be completed within 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />s:

2.2.1 Restore compliance with all SLs; and 2.2.2 Insert all insertable control rods.

BFN-UNIT 1 2.0-1 Amendment No. 236, 267 February 06, 2007

LHGR 3.2.3 3.2 POWER DISTRIBUTION LIMITS 3.2.3 LINEAR HEAT GENERATION RATE (LHGR)

LCO 3.2.3 All LHGRs shall be less than or equal to the limits specified in the COLR.

APPLICABILITY: THERMAL POWER Ž 25% RTP.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Any LHGR not within A.1 Restore LHGR(s) to within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> limits, limits.

B. Required Action and B.1 Reduce THERMAL 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> associated Completion POWER to < 25% RTP.

(

Time not met.

BFN-UNIT 1 3.2-5 Amendment No. 2-4

LHGR 3.2.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.2.3.1 Verify all LHGRs are less than or equal to the Once within limits specified in the COLR. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after

Ž 25% RTP AND 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> thereafter BFN-UNIT 1 3.2-6 Amendment No. 234

EOC-RPT Instrumentation 3.3.4.1 3.3 INSTRUMENTATION 3.3.4.1 End of Cycle Recirculation Pump Trip (EOC-RPT) Instrumentation LCO 3.3.4.1 a. Two channels per trip system for each EOC-RPT instrumentation Function listed below shall be OPERABLE:

1. Turbine Stop Valve (TSV) - Closure; and.

2. Turbine Control Valve (TCV) Fast Closure, Trip Oil Pressure

- Low.

OR

b. LCO 3.2.2, "MINIMUM CRITICAL POWER RATIO (MCPR),"

limits for inoperable EOC-RPT as specified in the COLR are made applicable; and 2

c. LCO 3.2.3, "LINEAR HEAT GENERATION RATE (LHGR),"

limits for an inoperable EOC-RPT, as specified in the COLR, are made applicable.

APPLICABILITY: THERMAL POWER > 30% RTP.

BFN-UNIT 1 3.3-29 Amendment No. 2-34 1

EOC-RPT Instrumentation 3.3.4.1 ACTIONS

,1I,f~T I

---

lvi r-I-----------------------------------

Separate Condition entry is allowed for each channel.

CONDITION REQUIRED ACTION COMPLETION TIME A. One or more channels A.1 Restore channel to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> inoperable. OPERABLE status.

OR A.2 -------- NOTE------

Not applicable if inoperable channel is the result of an inoperable breaker.

Place channel in trip. 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> B. One or more Functions B.1 Restore EOC-RPT trip 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> with EOC-RPT trip capability.

capability not maintained.

OR AND B.2 Apply MCPR and LHGR 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> MCPR and LHGR limits limits for inoperable for inoperable EOC-RPT EOC-RPT as specified in not made applicable, the COLR.

C. Required Action and C.1 Reduce THERMAL 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> associated Completion POWER to < 30% RTP.

Time not met.

BFN-UNIT 1 3.3-30 Amendment No. 2-,4

Main Turbine Bypass System 3.7.5 3.7 PLANT SYSTEMS 3.7.5 Main Turbine Bypass System LCO 3.7.5 The Main Turbine Bypass System shall be OPERABLE.

OR The following limits are made applicable:

a. LCO 3.2.1, "AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR)," limits for an inoperable Main Turbine Bypass System, as specified in the COLR; and,

b. LCO 3.2.2, "MINIMUM CRITICAL POWER RATIO (MCPR),"

limits for an inoperable Main Turbine Bypass System, as specified in the COLR; and

c. LCO 3.2.3, "LINEAR HEAT GENERATION RATE (LHGR),"

limits for an inoperable Main Turbine Bypass System, as specified in the COLR.

APPLICABILITY: THERMAL POWER _Ž 25% RTP.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Requirements of the LCO A.1 Satisfy the requirements 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> not met. of the LCO.

B. Required Action and B.1 Reduce THERMAL 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> associated Completion POWER to < 25% RTP.

Time not met.

BFN-UNIT 1 3.7-16 Amendment No. 2-34

Reporting Requirements 5.6 5.6.4 (Deleted).

5.6.5 CORE OPERATING LIMITS REPORT (COLR)

a. Core operating limits shall be established prior to each reload cycle, or prior to any remaining portion of a reload cycle, and shall be documented in the COLR for the following:

(1) The APLHGRs for Specification 3.2.1; (2) The LHGR for Specification 3.2.3; (3) The MCPR Operating Limits for Specification 3.2.2; and (4) The RBM setpoints and applicable reactor thermal power ranges for each of the setpoints for Specification 3.3.2.1, Table 3.3.2.1-1.

b. The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents (latest approved versions applicable to BFN):

1. NEDE-24011-P-A, General Electric Standard Application for Reactor Fuel.

2. XN-NF-81-58(P)(A), RODEX2 Fuel Rod Thermal-Mechanical Response Evaluation Model.

3. XN-NF-85-67(P)(A), Generic Mechanical Design for Exxon Nuclear Jet Pump BWR Reload Fuel.

4. EMF-85-74(P)(A), RODEX2A (BWR) Fuel Rod Thermal-Mechanical Evaluation Model.

5. ANF-89-98(P)(A), Generic Mechanical Design Criteria for BWR Fuel Designs.

6. XN-NF-80-19(P)(A) Volume 1, Exxon Nuclear Methodology for Boiling Water Reactors - Neutronic Methods for Design and Analysis.

(continued)

BFN-UNIT 1 5.0-24 Amendment No. 234, 239, 252 January 26, 2005

Reporting Requirements 5.6 5.6 Reporting Requirements (continued)

7. XN-NF-80-19(P)(A) Volume 4, Exxon Nuclear Methodology for Boiling Water Reactors: Application of the ENC Methodology to BWR Reloads.

8. EMF-2158(P)(A), Siemens Power Corporation Methodology for Boiling Water Reactors: Evaluation and Validation of CASMO-4/MICROBURN-B2.

9. XN-NF-80-19(P)(A) Volume 3, Exxon Nuclear Methodology for Boiling Water Reactors, THERMEX: Thermal Limits Methodology Summary Description.

10.XN-NF-84-105(P)(A) Volume 1, XCOBRA-T: A Computer Code for BWR Transient Thermal-Hydraulic Core Analysis.

11 .ANF-524(P)(A), ANF Critical Power Methodology for Boiling Water Reactors.

12.ANF-913(P)(A) Volume 1, COTRANSA2: A Computer Program for Boiling Water Reactor Transient Analyses.

13.ANF-1358(P)(A), The Loss of Feedwater Heating Transient in Boiling Water Reactors.

14. EMF-2209(P)(A), SPCB Critical Power Correlation.

15. EMF-2245(P)(A), Application of Siemens Power Corporation's Critical Power Correlations to Co-Resident Fuel.

16. EMF-2361 (P)(A), EXEM BWR-2000 ECCS Evaluation Model.

17. EMF-2292(P)(A), ATRIUMTM-1 0: Appendix K Spray Heat Transfer Coefficients.

The COLR will contain the complete identification for each of the Technical Specification referenced topical reports used to prepare the COLR (i.e., report number, title, revision, date and any supplements)

(continued)

BFN-UNIT 1 5.0-24a Amendment No. 234, 239, 252 Janay-2,5, 2005

ATTACHMENT 4 Browns Ferry Nuclear Plant (BFN)

Unit 1 Technical Specifications (TS) Change 467 Revision of Technical Specifications to allow utilization of AREVA NP fuel and associated analysis methodologies Proposed Technical Specification Bases Changes (Mark-up)

/

The following pages have been revised to reflect the proposed changes. On the affected pages a line has been drawn through the deleted text and new or revised text is shaded.

Reactor Core SLs B 2.1.1 B 2.0 SAFETY LIMITS (SLs)

B 2.1.1 Reactor Core SLs BASES BACKGROUND GDC 10 (Ref. 1) requires, and SLs ensure, that specified acceptable fuel design limits are not exceeded during steady state operation, normal operational transients, and abnormal operational transients.

The fuel cladding integrity SL is set such that no fuel damage is calculated to occur if the limit is not violated. Because fuel damage is not directly observable, a stepback approach is used to establish an SL, such that the MCPR is not less than the limit specified in Specification 2.1.1.2 for Gencrl El.e.tric

• Company (G-)-fuel. MCPR greater than the specified limit represents a conservative margin relative to the conditions required to maintain fuel cladding integrity.

The fuel cladding is one of the physical barriers that separate the radioactive materials from the environs. The integrity of this cladding barrier is related to its relative freedom from perforations or cracking. Although some corrosion or use related cracking may occur during the life of the cladding, fission product migration from this source is incrementally cumulative and continuously measurable. Fuel cladding perforations, however, can result from thermal stresses, which occur from reactor operation significantly above design conditions.

While fission product migration from cladding perforation is just as measurable as that from use related cracking, the thermally caused cladding perforations signal a threshold beyond which still greater thermal stresses may cause gross, rather than (continued)

BFN-UNIT 1 B 2.0-1 Revision 0

Reactor Core SLs B 2.1.1 BASES APPLICABLE 2.1.1.1 Fuel Cladding Integrity SAFETY ANALYSES (continued) GE critiGa*l po, r correlations arc applicable for all critical power ca4cQtin atpessures Ž!785 psig and coreflws; IQ1% of rated-flow.- The SPCB critical power correlation is used for both AREVA and coresident fuel and is valid at pressures >700 psia, and bundle mass fluxes >0.1 X106 Ibm/hr-ft 2 (2:12,000 lbm/hr, i.e.,

>10% core flow, on a per bundle basis) for ATRIUM-10 and GE14 fuel types. For thermal margin monitoring at 25% power and higher, the hot channel flow rate will be >28,000 Ibm/hr (core flow not less than natural circulation, i.e., -25%-30% core flow for 25% power); therefore, the fuel cladding integrity SL is conservative relative to the applicable range of the SPCB critical power correlation. For operation at low pressures or low flows, another basis is used, as follows:

The static head across the fuel bundles due only to elevation effects from liquid only in the channel, core bypass region, and annulus at zero power, zero flow is approximately 4.5 psi. At all operating conditions, this pressure differential is maintained by the bypass region of the core and the annulus region of the vessel. The elevation head provided by the annulus produces natural circulation flow conditions which have balancing pressure head and loss terms inside the core shroud. This natural circulation principle maintains a core plenum to plenum pressure drop of about 4.5 to 5 psid along the natural circulation flow line of the P/F operating map. In the range of power levels of interest, approaching 25% of rated power below which thermal margin monitoring is not required, the pressure drop and density head terms tradeoff for power changes such that natural circulation flow is nearly independent of reactor power.

This characteristic is represented by the nearly vertical portion of the natural circulation line on the P/F operating map.

Analysis has shown that the hot channel flow rate is >28,000 Ibm/hr (>0.23 x 106 Ibm/hr-ft2 ) in the region of operation with power -25% and core pressure drop of about 4.5 to 5 psid. Full scale ATLAS test data taken at pressures from 14.7 psia to 800 psia indicate that the fuel assembly critical power at 28,000 (continued)

BFN-UNIT 1 B 2.0-3 Revision G

Reactor Core SLs B 2.1.1 BASES APPLICABLE 2.1.1.1 Fuel Cladding Integrity (continued)

SAFETY ANALYSES Ibm/hr is approximately 3 MWt. With the design peaking factors, this corresponds to a core thermal power of more than 50%.

Thus operation up to 25% of rated power with normal natural circulation available is conservatively acceptable even if reactor pressure is equal to or below 800 psia-(the-limit of the range of appliability of GETAIB/G=EXL for GE fueI). If reactor power is significantly less than 25% of rated (e.g., below 10% of rated), the core flow and the channel flow supported by the available driving head may be less than 28,000 Ibm/hr (along the lower portion of the natural circulation flow characteristic on the P/F map).

However, the critical power that can be supported by the core and hot channel flow with normal natural circulation paths available remains well above the actual power conditions. The inherent characteristics of BWR natural circulation make power and core flow follow the natural circulation line as long as normal water level is maintained.

Thus, operation with core thermal power below 25% of rated without thermal margin surveillance is conservatively acceptable even for reactor operations at natural circulation. Adequate fuel thermal margins are also maintained without further surveillance for the low power conditions that would be present if core natural circulation is below 10% of rated flow (the-limit-of appl,"ablity of the G*ETABI*IGEXI ,GOl-ation_ forGE- fuel).

(continued)

BFN-UNIT 1 B 2.0-4 Revision 0

Reactor Core SLs B 2.1.1 BASES APPLICABLE 2.1.1.2 MCPR SAFETY ANALYSES (continued) The fuel cladding integrity SL is set such that no fuel damage is calculated to occur if the limit is not violated. Since the parameters that result in fuel damage are not directly observable during reactor operation, the thermal and hydraulic conditions that result in the onset of transition boiling have been used to mark the beginning of the region in which fuel damage could occur. Although it is recognized that the onset of transition boiling would not result in damage to BWR fuel rods, the critical power at which boiling transition is calculated to occur has been adopted as a convenient limit. However, the uncertainties in monitoring the core operating state and in the procedures used to calculate the critical power result in an uncertainty in the value of the critical power. Therefore, the fuel cladding integrity SL is defined as the critical power ratio in the limiting fuel assembly for which more than 99.9% of the fuel rods in the core are expected to avoid boiling transition, considering the power distribution within the core and all uncertainties.

The MCPR SL is determined using a statistical model that-combinesing all the uncertainties in operating parameters and the procedures used to calculate critical power. The probability of the occurrence of boiling transition is determined using the approved GenerFa E... tc*. Cr.t.al PoWer correlations. Detail. of the fuel cladding integrity SL calculation are given in Reference 2. Referen-e 2 also incO*ld*es a tabhlation of the i m iI I I iN r 4i i mA* A I i F uncrtintesused ;n thle deeriato orh hA"c R and r lm i

• iI

• iJ*i, Ii6,I*

-;(Ad-o I¢..*I inh nominal values of mte Pa*rameters used in the M'#IG P m statistical analysis, AREVA critical power correlations. One specific uncertainty included in the SL is the uncertainty inherent in the SPCB critical power correlation. References 2, 3, and 4 describe the uncertainties and methodologies used in determining the MCPR SL.

(continued)

BFN-UNIT 1 B 2.0-5 Revision 0

Reactor Core SLs B 2.1.1 BASES (continued)

SAFETY LIMIT Exceeding an SL may cause fuel damage and create a potential VIOLATIONS for radioactive releases in excess of 10 CFR 50.67, "Accident Source Term," limits (Ref. 35). Therefore, it is required to insert all insertable control rods and restore compliance with the SLs within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. The 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> Completion Time ensures that the operators take prompt remedial action and also ensures that the probability of an accident occurring during this period is minimal.

REFERENCES 1. 10 CFR 50, Appendix A, GDC 10.

Z. "t N16 NO. 010 SupplementiA, -anuar; 19, 1uu~

2. EMF-2209(P)(A), "SPCB Critical Power Correlation,"

(as identified in the COLR).

3. EMF-2245(P)(A), "Application of Siemens Power Corporation's Critical Power Correlations to Co-Resident Fuel," (as identified in the COLR).

4. ANF-524(P)(A), "ANF Critical Power Methodology for Boiling Water Reactors," (as identified in the COLR).

35. 10 CFR 50.67.

BFN-UNIT 1 B 2.0-7 Revision OT2-Ja*nuar' 25, 2005

SDM B 3.1.1 B 3.1 REACTIVITY CONTROL SYSTEMS B 3.1.1 SHUTDOWN MARGIN (SDM)

BASES BACKGROUND SDM requirements are specified to ensure:

a. The reactor can be made subcritical from all operating conditions and transients and Design Basis Events;

b. The reactivity transients associated with postulated accident conditions are controllable within acceptable limits; and

c. The reactor will be maintained sufficiently subcritical to preclude inadvertent criticality in the shutdown condition.

These requirements are satisfied by the control rods, as described in GDC 26 (Ref. 1), which can compensate for the reactivity effects of the fuel and water temperature changes experienced during all operating conditions.

APPLICABLE The control rod drop accident (CRDA) analysis (Refs. 2, afid 3, SAFETY ANALYSES 9, and 10) assumes the core is subcritical with the highest worth control rod withdrawn. Typically, the first control rod withdrawn has a very high reactivity worth and, should the core be critical during the withdrawal of the first control rod, the consequences of a CRDA could exceed the fuel damage limits for a CRDA (see Bases for LCO 3.1.6, "Rod Pattern Control").

Also, SDM is assumed as an initial condition for the control rod removal error during refueling (Ref. 4) and fuel assembly insertion error during refueling (Ref. 5) accidents. The analysis of these reactivity insertion events assumes the refueling interlocks are OPERABLE when the reactor is in the refueling mode of operation. These interlocks prevent the withdrawal of more (continued)

BFN-UNIT 1 B 3.1 -1 Revision 0

SDM B 3.1.1 BASES (continued)

REFERENCES 1. 10 CFR 50, Appendix A, GDC 26.

2. FSAR, Section 14.6.2.

3. NEDE-2401 1-P-A-1 3, "General Electric Standard Application for Reactor Fuel," Section S.2.2.3.1, August 1996.

4. FSAR, Section 14.5.3.3.

5. FSAR, Section 14.5.3.4.

6. FSAR, Section 3.6.5.2.

7. NEDE-2401 1-P-A-1 3, "General Electric Standard Application for Reactor Fuel," Section 3.2.4.1, August 1996.

8. NRC 93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.

9. XN-NF-80-19(P)(A), Volume 1 "Exxon Nuclear Methodology for Boiling Water Reactors - Neutronic Methods for Design and Analysis," (as identified in the COLR).

10. EMF-2158(P)(A), "Siemens Power Corporation Methodology for Boiling Water Reactors: Evaluation and Validation of CASMO-4/MICROBURN-B2," (as identified in the COLR).

BFN-UNIT 1 B 3.1-8 Revision 0

Control Rod Scram Times B 3.1.4 BASES (continued)

APPLICABLE The analytical methods and assumptions used in evaluating the SAFETY ANALYSES control rod scram function are presented in References 2, 3, and 4. The Design Basis Accident (DBA) and transient analyses assume that all of the control rods scram at a specified insertion rate. The resulting negative scram reactivity forms the basis for the determination of plant thermal limits (e.g., the MCPR). Other distributions of scram times (e.g., several control rods scramming slower than the average time with several control rods scramming faster than the average time) can also provide sufficient scram reactivity. Surveillance of each individual control rod's scram time ensures the scram reactivity assumed in the DBA and transient analyses can be met.

The scram function of the CRD System protects the MCPR Safety Limit (SL) (see Bases for SL 2.1.1, "Reactor Core SLs,"

and LCO 3.2.2, "MINIMUM CRITICAL POWER RATIO (MCPR)") and the 1% cladding plastic strain fuel design limit (see Bases for LCO 3.2.1, "AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR)", and LCO 3.2.3, "LINEAR HEAT GENERATION RATE (LHGR)"), which ensure that no fuel damage will occur if these limits are not exceeded. Above 800 psig, the scram function is designed to insert negative reactivity at a rate fast enough to prevent the actual MCPR from becoming less than the MCPR SL, during the analyzed limiting power transient. Below 800 psig, the scram function is assumed to perform during the control rod drop accident (Refs. 5, 8, and 9) and, therefore, also provides protection against violating fuel damage limits during reactivity insertion accidents (see Bases for LCO 3.1.6, "Rod Pattern Control").

For the reactor vessel overpressure protection analysis, the scram function, along with the safety/relief valves, ensure that the peak vessel pressure is maintained within the applicable ASME Code limits.

Control rod scram times satisfy Criterion 3 of the NRC Policy Statement (Ref. 7).

(continued)

BFN-UNIT 1 B 3.1-27 Revision 0

Control Rod Scram Times B 3.1.4 BASES (continued)

REFERENCES 1. 10 CFR 50, Appendix A, GDC 10.

2. FSAR, Section 3.4.6.

3. FSAR, Section 14.5.

4. FSAR, Section 14.6.

5. NEDE-24011-P-A-13, "General Electric Standard Application for Reactor Fuel," Section 3.2.4.1, August 1996.

6. Letter from R. F. Janecek (BWROG) to R. W. Starostecki (NRC), "BWR Owners Group Revised Reactivity Control System Technical Specifications," BWROG-8754, September 17, 1987.

7. NRC No.93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.

8. XN-NF-80-19(P)(A), Volume 1 "Exxon Nuclear Methodology for Boiling Water Reactors - Neutronic Methods for Design and Analysis," (as identified in the COLR).

9. EMF-2158(P)(A), "Siemens Power Corporation Methodology for Boiling Water Reactors: Evaluation and Validation of CASMO-4/MICROBURN-B2," (as identified in the COLR).

BFN-UNIT 1 B 3.1-34 Revision 0

Rod Pattern Control B 3.1.6 B 3.1 REACTIVITY CONTROL SYSTEMS B 3.1.6 Rod Pattern Control BASES BACKGROUND Control rod patterns during startup conditions are controlled by the operator and the rod worth minimizer (RWM) (LCO 3.3.2.1, "Control Rod Block Instrumentation"), so that only specified control rod sequences and relative positions are allowed over the operating range of all control rods inserted to 10% RTP.

The sequences limit the potential amount of reactivity addition that could occur in the event of a Control Rod Drop Accident (CRDA).

This Specification assures that the control rod patterns are consistent with the assumptions of the CRDA analyses of References 1, and 2, 10, and 11.

APPLICABLE The analytical methods and assumptions used in evaluating SAFETY ANALYSES the CRDA are summarized in References 1, a-Rd 2, 10, and 11.

CRDA analyses assume that the reactor operator follows prescribed withdrawal sequences. These sequences define the potential initial conditions for the CRDA analysis. The RWM (LCO 3.3.2.1) provides backup to operator control of the withdrawal sequences to ensure that the initial conditions of the CRDA analysis are not violated.

Prevention or mitigation of positive reactivity insertion events is necessary to limit the energy deposition in the fuel, thereby preventing significant fuel damage which could result in the undue release of radioactivity. Since the failure consequences for U0 2 have been shown to be insignificant below fuel energy (continued)

BFN-UNIT 1 B 3.1-41 Revision O

Rod Pattern Control B 3.1.6 BASES APPLICABLE depositions of 300 cal/gm (Ref. 3), the fuel damage limit of SAFETY ANALYSES 280 cal/gm provides a margin of safety from significant core (continued) damage which would result in release of radioactivity (Refs. 4 and 5). Generic evaluations (Refs. 1, and 6, and 10) of a design basis CRDA (i.e., a CRDA resulting in a peak fuel energy deposition of 280 cal/gm) have shown that if the peak fuel enthalpy remains below 280 cal/gm, then the maximum reactor pressure will be less than the required ASME Code limits (Ref. 7) and the calculated offsite doses will be well within the required limits (Ref. 5).

Control rod patterns analyzed in References 1, 10, and 11, follow the banked position withdrawal sequence (BPWS). The BPWS is applicable from the condition of all control rods fully inserted to 10% RTP (Ref. 2). For the BPWS, the control rods are required to be moved in groups, with all control rods assigned to a specific group required to be within specified banked positions (e.g., between notches 08 and 12). The banked positions are established to minimize the maximum incremental control rod worth without being overly restrictive during normal plant operation. Generic

. analysis of the B PIA1DS (Ref.-8)-4has-Analyses are performed using the Reference 10 methodology demonstratinge4-that the 280 cal/gm fuel damage limit will not be violated during a CRDA while following the BPWS mode of operation. The evaluation provided by the generic BPWS analysis (Ref. 8) allows a limited number (i.e.,

eight) and corresponding distribution of fully inserted, inoperable control rods, that are not in compliance with the sequence.

Rod pattern control satisfies Criterion 3 of the NRC Policy Statement (Ref. 9).

(continued)

BFN-UNIT 1 B 3.1-42 Revision 0

Rod Pattern Control B 3.1.6 BASES (continued)

REFERENCES 1. NEDE-2401 1-P-A-1 3, "General Electric Standard Application for Reactor Fuel," Section 2.2.3.1, August 1996.

2. Letter from T. Pickens (BWROG) to G. C. Lainas (NRC),

Amendment 17 to General Electric Licensing Topical Report, NEDE-24011-P-A, August 15, 1986.

3. NUREG-0979, Section 4.2.1.3.2, April 1983.

4. NUREG-0800, Section 15.0.1.

5. 10 CFR 50.67.

6. NEDO-21778-A, "Transient Pressure Rises Affected Fracture Toughness Requirements for Boiling Water Reactors," December 1978.

7. ASME, Boiler and Pressure Vessel Code.

8. NEDO-21231, "Banked Position Withdrawal Sequence,"

January 1977.

9. NRC No.93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.

10. XN-NF-80-19(P)(A), Volume 1 "Exxon Nuclear Methodology for Boiling Water Reactors - Neutronic Methods for Design and Analysis," (as identified in the COLR).

11. EMF-2158(P)(A), "Siemens Power Corporation Methodology for Boiling Water Reactors: Evaluation and Validation of CASMO-4/MICROBURN-B2," (as identified in the COLR).

BFN-UNIT 1 B 3.1-46 Revision O0-2-9

APLHGR B 3.2.1 B 3.2 POWER DISTRIBUTION LIMITS B 3.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR)

BASES BACKGROUND The APLHGR is a measure of the average LHGR of all the fuel rods in a fuel assembly at any axial location. Limits on the APLHGR are specified to ensure that the fuel design limits identified in Reference 1 are not exceeded during abnormal operational transients and that the peak cladding temperature (PCT) during the postulated design basis loss of coolant accident (LOCA) does not exceed the limits specified in 10 CFR 50.46.

APPLICABLE The analytical methods and assumptions used in evaluating SAFETY ANALYSES the fuel design limits are presented in References 1, aRd 2, and

11. The analytical methods and assumptions used in evaluating Design Basis Accidents (DBAs), abnormal operational transients, and normal operation that determine the APLHGR limits are presented in References 1, 2, 3, 4, and 7 for GE fuel; References 11, 12, 13, 14, and 15 for AREVA fuel.

4 C I1 ,d t ;, n aA. i t+. n ' , t v .a r i A + a A. , +n +*a ,

+&r, +, +&%

1%iMit On the fuel cladding p~astic str~n :And other fuel' de g liisdescribed inReference 1 are not eXceeded during abnormal operation~al tranients fo operation With ILHGRs up to the operating limit LHGR. APLHGR limits are equivalent to the L=HGR limit forF each fuel rod divided by the local peaking facstor of the fuel assembly.

APLHGR limits arc developed as a functio of exposure and the

. I, ne

. . eF6ta

. . . e6 . te.u..-..-

. en. .. U~

Arka a +- - . - -A.;- -. '.+- - . .. k. .A- .. - . . .

I

+ (D f 7\ M~.L A ra rA + AD! Wf(-D 1immt i r rcerib w" 5 ý w - 7- x7irr wpwn wN W 5 cc rw deteRnined using the thFee dimensienal BVVR simulatGF rae (Ref. 9) to aRalyze slew flOW FURGUt twsieRts- The ftlAf depeRdent multipliw, MAPFAGf, is dependent eR the maximum n6ore flew r-61RGUt Gapability. The m i A- -t fl(Wf iR (continued)

BFN-UNIT 1 B 3.2-1 Revision Q-,49 GetebeF 26, 2096

APLHGR B 3.2.1 BASES APPLICABLE dependent on the-e-Xistt-ngsetting Of t-he coeflowA limiter in the-SAFETY ANALYSES RecGircul-1at-ion FloGW Conrol System.

(continued)

Based on aFnalyrcs of limiting plant transicnts (other than core floW inrGeases) over a range of poWer and flow conditions, power dependent mnultipliers, MAPFACO, are also generated Due to the sensitivity of the transient repnec~ to initial core flow levels at power levels below thosAe at.w.hich turbin~e stop valve closures an~d turbine control valve fast closure Scram: trips are bypassed, both high and IOW cor.e flew MAPFACG "I*mit6a&e proVided for operation at power levels between 25% RTP a~nd-the previously m "entioned bypass power level. The exposu depeRdent APLHGR limits a.e reduced by MAPF.A..,and FueI MAPFGAC at various operating conditions to ensure that all design criteria are mnet for normnal operation and abRFnoral operatfional transients. A complete discussion of the analy.

coede is provided in Refer-ence 9.

GE Fuel LOCA analyses are then-performed to ensure that the above ddGeermi4e4-APLHGR limits are adequate to meet the PCT and maximum oxidation limits of 10 CFR 50.46. The analysis is performed using calculational models that are consistent with the requirements of 10 CFR 50, Appendix K. A complete discussion of the analysis code is provided in Reference 5. The PCT following a postulated LOCA is a function of the average heat generation rate of all the rods of a fuel assembly at any axial location and is not strongly influenced by the rod to rod power distribution within an assembly. The APLHGR limits specified are equivalent to the LHGR of the highest powered fuel rod assumed in the LOCA analysis divided by its local peaking factor. A conservative multiplier is applied to the LHGR assumed in the LOCA analysis to account for the uncertainty associated with the measurement of the APLHGR.

(continued)

BFN-UNIT 1 B 3.2-2 Revision 40

,M.mendm'ntNo. 236 Oc~tober 26, 2006

APLHGR B 3.2.1 BASES APPLICABLE For single recirculation loop operation, an APLHGR multiplier is SAFETY ANALYSES applied to the APLHGR limit (Ref. 5 and Ref. 10). The (continued) multiplier is documented in the COLR. This multiplier is due to the conservative analysis assumption of an earlier departure from nucleate boiling with one recirculation loop available, resulting in a more severe heatup during a LOCA.

AREVA Fuel For AREVA fuel, the APLHGR limits are developed as a function of exposure and, along with the LHGR limits, ensure adherence to fuel design limits during abnormal operational transients. No power- or flow-dependent corrections are applied to the APLHGR (referred to as the maximum APLHGR or MAPLHGR).

AREVA APLHGR limits are intended to be bound by the LHGR limits.

The calculational procedure used to establish the AREVA fuel MAPLHGR limits is based on LOCA analyses as defined in 10 CFR 50.46, Appendix K. MAPLHGR limits are created to assure that the peak cladding temperature of AREVA fuel following a postulated design basis LOCA will not exceed the PCT and maximum oxidation limits specified in 10 CFR 50.46, Appendix K. The calculational models and methodology are described in References 11 and 12.

The AREVA fuel MAPLHGR limits for two-loop operation are specified in the COLR. For single-loop operation, a MAPLHGR multiplier is applied to the MAPLHGR limit (Ref. 11). The multiplier is documented in the COLR.

The APLHGR satisfies Criterion 2 of the NRC Policy Statement (Ref. 6).

(continued)

BFN-UNIT 1 B 3.2-3 Revision 4G Ame*dMent Wo. 236-October 26, 2006

APLHGR B 3.2.1 BASES (continued)

LCO The APLHGR limits specified in the COLR are the result of the fuel design, DBA, and transient analyses. For operation at other than 100% power and 100%04 rec-irc-ulation flow coenditions,th APL=HGR operating limit is determnined by multiplying the smnaller of the IMAPFACG and MAPF=A,* fa*c*tS times the exposure dependent APLHGR limit. With only one recirculation loop in operation, in conformance with the requirements of LCO 3.4.1, "Recirculation Loops Operating," the limit is determined by multiplying the exposure dependent limit by an APLHGR correction factor (Ref. 5 and Ref. 10). Cycle specific APLHGR correction factors for single recirculation loop operation are documented in the COLR. APLHGR limits are selected such that no power or flow dependent corrections are required.

Additional APLHGR operating limit adjustments may be provided in the COLR supporting other analyzed equipment out-of-service conditions.

APPLICABILITY The APLHGR limits are primarily derived from fuel design evaluations and LOCA and transient analyses that are assumed to occur at high power levels. Design calculations (Ref. 4) and operating experience have shown that as power is reduced, the margin to the required APLHGR limits increases. This trend continues down to the power range of 5% to 15% RTP when entry into MODE 2 occurs. When in MODE 2, the intermediate range monitor scram function provides prompt scram initiation during any significant transient, thereby effectively removing any APLHGR limit compliance concern in MODE 2. Therefore, at THERMAL POWER levels < 25% RTP, the reactor is (continued)

BFN-UNIT 1 B 3.2-3a Revision 40 Amendment No. 236 October 26, 2006

APLHGR B 3.2.1 BASES SURVEILLANCE SR 3.2.1.1 (continued)

REQUIREMENTS operation. The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> allowance after THERMAL POWER

> 25% RTP is achieved is acceptable given the large inherent margin to operating limits at low power levels.

REFERENCES 1. NEDE-24011-P-A-13 "General Electric Standard Application for Reactor Fuel," August 1996.

2. FSAR, Chapter 3.

3. FSAR, Chapter 14.

4. FSAR, Appendix N.

5. NEDC-32484P, "Browns Ferry Nuclear Plant Units 1, 2, and 3, SAFER/GESTR-LOCA Loss-of-Coolant Accident Analysis," Revision 2, December 1997.

6. NRC No.93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.

7. NEDC-32433P, "Maximum Extended Load Line Limit and ARTS Improvement Program Analyses for Browns Ferry Nuclear Plant Units 1,2, and 3," April 1995.

8. NEDO-30130-A, "Steady State Nuclear Methods,"

May 1985.

9. NEDO-24154, "Qualification of the One-Dimensional Core Transient Model for Boiling Water Reactors," October 1978.

10. NEDO-24236, "Browns Ferry Nuclear Plant Units 1,2, and 3, Single-Loop Operation," May 1981.

11. EMF-2361 (P)(A), "EXEM BWR-2000 ECCS Evaluation Model," (as identified in the COLR).

(continued)

BFN-UNIT 1 B 3.2-5 Revision 40 OAmendment No. 23 October 26, 2006

APLHGR B 3.2.1 BASES REFERENCES 12. EMF-2292(P)(A), "ATRIUMTM-10: Appendix K Spray Heat (continued) Transfer Coefficients," (as identified in the COLR).

13. XN-NF-81-58(P)(A), "RODEX2 Fuel Rod Thermal-Mechanical Response Evaluation Model," (as identified in the COLR).

14. XN-NF-80-19(P)(A), Volume 1 "Exxon Nuclear Methodology for Boiling Water Reactors - Neutronic Methods for Design and Analysis," (as identified in the COLR).

15. XN-NF-80-19(P)(A) Volume 4, "Exxon Nuclear Methodology for Boiling Water Reactors: Application of the ENC Methodology to BWR Reloads," (as identified in the COLR).

BFN-UNIT 1 B 3.2-5a Revision 40 Avmendmwent No. 23 October 26, 2006

MCPR B 3.2.2 BASES (continued)

APPLICABLE The analytical methods and assumptions used in evaluating SAFETY ANALYSES the abnormal operational transients to establish the operating limit MCPR are presented in References 2, 3, 4, 5, 8, ad-l10, 11, 12, 13, 14, and 15. To ensure that the MCPR SL is not exceeded during any transient event that occurs with moderate frequency, limiting transients have been analyzed to determine the largest reduction in critical power ratio (CPR). The types of transients evaluated are loss of flow, increase in pressure and power, positive reactivity insertion and coolant temperature decrease.

The limiting transient yields the largest change in CPR (ACPR).

When the largest ACPR is added to the MCPR SL, the required operating limit MCPR is obtained.

The MCPR operating limits derived from the transient analysis are dependent on the operating core flow and power state (M.PR, and MCPRe, ,peGtiyely) to ensure adherence to fuel design limits during the worst transient that occurs with moderate frequency (Reference 8). Flow dependent MCPR (MCPRf) limits are determined by steady state thermal hydraulic methods wih-key physics esosinputs benchmnarked using the thre dmnsion-al BWlýR simulator code (Reference 6) to analyze slow-flow runout trani;ents using the three dimensional BWR simulator code (Ref. 12) and the multichannel thermal hydraulics code (Ref. 13). The operating limit is dependent on the maximum core flow limiter setting in the Recirculation Flow Control System.

Power dependent MCPR limits (MCPRp) are determined by the One di*men..ional tan.ent code (Refercnce 0) three-dimensional BWR simulator code (Ref. 12) and the one-dimensional transient codes (Refs. 14 and 15). Due to the sensitivity of the transient response to initial core flow levels at power levels below those at which the turbine control valve fast closure scrams are bypassed, high and low flow MCPRp operating limits are provided for operating between 25% RTP and the previously mentioned bypass power level.

The MCPR satisfies Criterion 2 of the NRC Policy Statement (Ref. 7).

(continued)

BFN-UNIT 1 B 3.2-7 Revision 49 A~mendment hio. 23 October 26, 2006

MCPR B 3.2.2 BASES (continued)

LCO The MCPR operating limits specified in the COLR are the result of the Design Basis Accident (DBA) and transient analysis.

Additional MCPR operating limits supporting analyzed equipment out-of-service conditions are provided in the COLR.

The operating limit MCPR is determined by the larger of the MCPRf and MCPRp limits.

APPLICABILITY The MCPR operating limits are primarily derived from transient analyses that are assumed to occur at high power levels. Below 25% RTP, the reactor is operating at a minimum recirculation pump speed and the moderator void ratio is small. Surveillance of thermal limits below 25% RTP is unnecessary due to the large inherent margin that ensures that the MCPR SL is not exceeded even if a limiting transient occurs. Statistical analyses indicate that the nominal value of the initial MCPR expected at 25% RTP is > 3.5. Studies of the variation of limiting transient behavior have been performed over the range of power and flow conditions. These studies encompass the range of key actual plant parameter values important to typically limiting transients.

The results of these studies demonstrate that a margin is expected between performance and the MCPR requirements, and that margins increase as power is reduced to 25% RTP.

This trend is expected to continue to the 5% to 15% power range when entry into MODE 2 occurs. When in MODE 2, the intermediate range monitor provides rapid scram initiation for any significant power increase transient, which effectively eliminates any MCPR compliance concern. Therefore, at THERMAL POWER levels < 25% RTP, the reactor is operating with substantial margin to the MCPR limits and this LCO is not required.

ACTIONS A._1 If any MCPR is outside the required limits, an assumption regarding an initial condition of the design basis transient analyses may not be met. Therefore, prompt action should be taken to restore the MCPR(s) to within the required limits such (continued)

BFN-UNIT 1 B 3.2-8 Revision Q-40 October 26, 2006

MCPR B 3.2.2 BASES SURVEILLANCE SR 3.2.2.2 REQUIREMENTS (continued) Because the transient analysis takes credit for conservatism in the scram speed performance, it must be demonstrated that the specific scram speed distribution is consistent with that used in the transient analysis. SR 3.2.2.2 determines the value of--

Which is- .amoasgure of the acatual scramn speed distribution copared with the assumned d-istri-butionF. The MCPR epcrating-limit is the~n deterrmnend barsed on anF interpolatioan bet-woen the "Co-ntrol Ro~d Scoram T-imors") and Option mi""hptwdirnr t~ rrI BTh(realisticw 7 hi ir scr-r;amA times) analyses. The parameter Tmust be determnined once

,-..- .- .~---.I within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> after eacah set of scramn time tests requiredlby-SR 3.1A.4. and- SR 3-1.4.2 bcuethe effective scramn 6peed-I..WarnrIaImnn m;ray . .,nnan .,dAO the ,a~l *k2 The houF

,.k..... v.i.... Fi b d, V,,cm I.. .. w'v ta sv ra spee.

changes in T-expected during the fuel cycle actual scram speed distribution and compares it with the assumed distribution. The MCPR operating limit is determined based either on the applicable limit associated with scram times of LCO 3.1.4, "Control Rod Scram Times," or the nominal scram times. The scram speed dependent MCPR limits are contained in the COLR. This determination must be performed within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> after each set of control rod scram time tests required by SR 3.1.4.1 and SR 3.1.4.2 because the effective scram speed distribution may change during the cycle. The 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Completion Time is acceptable due to the relatively minor changes in the actual control rod scram speed distribution expected during the fuel cycle.

REFERENCES 1. NUREG-0562, "Fuel Rod Failure As a Consequence of Departure from Nucleate Boiling or Dryout," June 1979.

2. NEDE-24011-P-A-13, "General Electric Standard Application for Reactor Fuel," August 1996.

3. FSAR, Chapter 3.

(continued)

BFN-UNIT 1 B 3.2-10 Revision 40

,Armendrnent No. 23 October 26, 2006

MCPR B 3.2.2 BASES REFERENCES 4. FSAR, Chapter 14.

(continued)

5. FSAR, Appendix N.

6. NEDO-30130-A, "Steady State Nuclear Methods,"

May 1985.

7. NRC No.93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.

8. NEDC-32433P, "Maximum Extended Load Line Limit and ARTS Improvement Program Analyses for Browns Ferry Nuclear Plant Units 1, 2, and 3," April 1995.

9. NEDO-24154, "Qualification of the One-Dimensional Core Transient Model for Boiling Water Reactors," October 1978.

10. NEDO-24236, "Browns Ferry Nuclear Plant Units 1, 2, and 3, Single-Loop Operation," May 1981.

11. ANF-524(P)(A), "ANF Critical Power Methodology for Boiling Water Reactors," (as identified in the COLR).

12. EMF-2158(P)(A), "Siemens Power Corporation Methodology for Boiling Water Reactors: Evaluation and Validation of CASMO-4/MICROBURN-B2," (as identified in the COLR).

13. XN-NF-80-19(P)(A) Volume 3, "Exxon Nuclear Methodology for Boiling Water Reactors, THERMEX: Thermal Limits Methodology Summary Description," (as identified in the COLR).

14. ANF-913(P)(A) Volume 1, "COTRANSA2: A Computer Program for Boiling Water Reactor Transient Analyses," (as identified in the COLR).

15. XN-NF-84-105(P)(A) Volume 1, "XCOBRA-T: A Computer Code for BWR Transient Thermal-Hydraulic Core Analysis,"

(as identified in the COLR).

BFN-UNIT 1 B 3.2-10a Revision 40 Oc÷tober 26, 2006

LHGR B 3.2.3 BASES APPLICABLE A value of 1% plastic strain of the fuel cladding has been SAFETY ANALYSES defined as the limit below which fuel damage caused by (continued) overstraining of the fuel cladding is not expected to occur (Ref. 3).

Fuel design evaluations have been performed and demonstrate that the 1% fuel cladding plastic strain design limit is not exceeded during continuous operation with LHGRs up to the operating limit specified in the COLR. The analysis also includes allowances for short term transient operation above the operating limit to account for abnormal operational transients, plus an allowance for densification power spiking.

LHGR limits are multiplied by the smaller of either the flow-dependent LHGR factor (LHGRFACf) or the power-dependent LHGR factor (LHGRFACp) corresponding to the existing core flow and power state to ensure adherence to the fuel mechanical design bases during the limiting transient.

LHGRFACf is generated to protect the core from slow flow runout transients. A curve is provided based on the maximum credible flow runout transient. LHGRFACp is generated to protect the core from plant transients other than core flow increases. LHGRFAC multipliers are provided in the COLR.

The LHGR satisfies Criterion 2 of the NRC Policy Statement (Ref. 4).

LCO The LHGR is a basic assumption in the fuel design analysis.

The fuel has been designed to operate at rated core power with sufficient design margin to the LHGR calculated to cause a 1%

fuel cladding plastic strain. The operating limit to accomplish this objective is specified in the COLR.

Additional LHGR operating limits adjustments may be provided in the COLR to support analyzed equipment out-of-service operation.

(continued)

BFN-UNIT 1 B 3.2-12 Revision Q

LHGR B 3.2.3 BASES (continued)

APPLICABILITY The LHGR limits are derived from fuel design analysis that is limiting at high power level conditions. At core thermal power levels < 25% RTP, the reactor is operating with a substantial margin to the LHGR limits and, therefore, the Specification is only required when the reactor is operating at >_25% RTP.

(continued)

BFN-UNIT 1 B 3.2-12a Revision 0

Control Rod Block Instrumentation B 3.3.2.1 BASES (continued)

APPLICABLE 1. Rod Block Monitor SAFETY ANALYSES, LCO, and The RBM is designed to prevent violation of the MCPR SL and APPLICABILITY the cladding 1% plastic strain fuel design limit that may result from a single control rod withdrawal error (RWE) event. The analytical methods and assumptions used in evaluating the RWE event are summarized in References 3, 12, and 13. A statistical analysis of RWE events was performed to determine the RBM response for both channels for each event. From these responses, the fuel thermal performance as a function of RBM Allowable Value was determined. The Allowable Values are chosen as a function of power level. Based on the specified Allowable Values, operating limits are established.

The RBM Function satisfies Criterion 3 of the NRC Policy Statement (Ref. 10).

Two channels of the RBM are required to be OPERABLE, with their setpoints within the appropriate Allowable Value for the associated power range to ensure that no single instrument failure can preclude a rod block from this Function. The setpoints are calibrated consistent with applicable setpoint methodology (nominal trip setpoint).

Nominal trip setpoints are specified in the setpoint calculations.

The nominal setpoints are selected to ensure that the setpoints do not exceed the Allowable Values between successive CHANNEL CALIBRATIONS. Operation with a trip setpoint less conservative than the nominal trip setpoint, but within its Allowable Value, is acceptable. Trip setpoints are those (continued)

BFN-UNIT 1 B 3.3-59 Revision 0,40 October 26, 2006

Control Rod Block Instrumentation B 3.3.2.1 BASES APPLICABLE 2. Rod Worth Minimizer SAFETY ANALYSES, LCO, and The RWM enforces the banked position withdrawal sequence APPLICABILITY (BPWS) to ensure that the initial conditions of the CRDA (continued) analysis are not violated. The analytical methods and assumptions used in evaluating the CRDA are summarized in References 4, 5, 6,-and 7, 12, and 13. The BPWS requires that control rods be moved in groups, with all control rods assigned to a specific group required to be within specified banked positions. Requirements that the control rod sequence is in compliance with the BPWS are specified in LCO 3.1.6, "Rod Pattern Control."

The RWM Function satisfies Criterion 3 of the NRC Policy Statement (Ref. 10).

Since the RWM is designed to act as a backup to operator control of the rod sequences, only one channel of the RWM is available and required to be OPERABLE (Ref. 7). Special circumstances provided for in the Required Action of LCO 3.1.3, "Control Rod OPERABILITY," and LCO 3.1.6 may necessitate bypassing the RWM to allow continued operation with inoperable control rods, or to allow correction of a control rod pattern not in compliance with the BPWS. The RWM may be bypassed as required by these conditions, but then it must be considered inoperable and the Required Actions of this LCO followed.

Compliance with the BPWS, and therefore OPERABILITY of the RWM, is required in MODES 1 and 2 when THERMAL POWER is < 10% RTP. When THERMAL POWER is > 10% RTP, there is no possible control rod configuration that results in a control (continued)

BFN-UNIT 1 B 3.3-61 Revision 0

Control Rod Block Instrumentation B 3.3.2.1 BASES (continued)

REFERENCES 1. FSAR, Section 7.5.8.2.3.

2. FSAR, Section 7.16.5.3.1.k.

3. NEDC-32433P, "Maximum Extended Load Line Limit and ARTS Improvement Program Analyses for Browns Ferry Nuclear Plant Unit 1, 2 and 3," April 1995.

4. NEDE-2401 1-P-A-US, "General Electrical Standard Application for Reload Fuel," Supplement for United States, (revision specified in the COLR).

5. "Modifications to the Requirements for Control Rod Drop Accident Mitigating Systems," BWR Owners' Group, July 1986.

6. NEDO-21231, "Banked Position Withdrawal Sequence,"

January 1977.

7. NRC SER, "Acceptance of Referencing of Licensing Topical Report NEDE-2401 1-P-A," "General Electric Standard Application for Reactor Fuel, Revision 8, Amendment 17," December 27, 1987.

8. NEDC-30851-P-A, Supplement 1, "Technical Specification Improvement Analysis for BWR Control Rod Block Instrumentation," October 1988.

9. GENE-770-06-1, "Addendum to Bases for Changes to Surveillance Test Intervals and Allowed Out-of-Service Times for Selected Instrumentation Technical Specifications," February 1991.

10. NRC No.93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.

11. NEDC-32410P-A, "Nuclear Measurement Analysis and Control Power Range Neutron Monitor (NUMAC PRNM)

Retrofit Plus Option III Stability Trip Function,"

October 1995.

(continued)

BFN-UNIT 1 B 3.3-71 Revision 0-40 Octobcr 26, 2006

Control Rod Block Instrumentation B 3.3.2.1 BASES REFERENCES 12. XN-NF-80-19(P)(A), Volume 1 "Exxon Nuclear (continued) Methodology for Boiling Water Reactors - Neutronic Methods for Design and Analysis," (as identified in the COLR).

13. EMF-2158(P)(A), "Siemens Power Corporation Methodology for Boiling Water Reactors: Evaluation and Validation of CASMO-4/MICROBURN-B2," (as identified in the COLR).

BFN-UNIT 1 B 3.3-71a Revision 05-40 October 26, 2006

EOC-RPT Instrumentation B 3.3.4.1 B 3.3 INSTRUMENTATION B 3.3.4.1 End of Cycle Recirculation Pump Trip (EOC-RPT) Instrumentation BASES BACKGROUND The EOC-RPT instrumentation initiates a recirculation pump trip (RPT) to reduce the peak reactor pressure and power resulting from turbine trip or generator load rejection transients to provide additional margin to core thermal MCPR Safety Limits (SLs),

and LHGR limits.

The need for the additional negative reactivity in excess of that normally inserted on a scram reflects end of cycle reactivity considerations. Flux shapes at the end of cycle are such that the control rods may not be able to ensure that thermal limits are maintained by inserting sufficient negative reactivity during the first few feet of rod travel upon a scram caused by Turbine Control Valve (TCV) Fast Closure, Trip Oil Pressure - Low or Turbine Stop Valve (TSV) - Closure. The physical phenomenon involved is that the void reactivity feedback due to a pressurization transient can add positive reactivity at a faster rate than the control rods can add negative reactivity.

The EOC-RPT instrumentation, as shown in Reference 1, is composed of sensors that detect initiation of closure of the TSVs or fast closure of the TCVs, combined with relays, logic circuits, and fast acting circuit breakers that interrupt power from the recirculation pump variable frequency drives (VFD) to each of the recirculation pump motors. When the channels pre-established setpoint is exceeded, the channel output relay actuates, which then outputs an EOC-RPT signal to the trip logic. When the RPT breakers trip open, the recirculation pumps coast down under their own inertia. The EOC-RPT has two identical trip systems, either of which can actuate an RPT.

(continued)

BFN-UNIT 1 B 3.3-105 Revision Q-49 Ap4[l30, 2007

EOC-RPT Instrumentation B 3.3.4.1 BASES BACKGROUND Each EOC-RPT trip system is a two-out-of-two logic for each (continued) Function; thus, either two TSV - Closure or two TCV Fast Closure, Trip Oil Pressure - Low signals are required for a trip system to actuate. If either trip system actuates, both recirculation pumps will trip. There are two EOC-RPT breakers in series per recirculation pump. One trip system trips one of the two EOC-RPT breakers for each recirculation pump, and the second trip system trips the other EOC-RPT breaker for each recirculation pump.

APPLICABLE The TSV - Closure and the TCV Fast Closure, Trip Oil SAFETY ANALYSES, Pressure - Low Functions are designed to trip the recirculation LCO, and pumps in the event of a turbine trip or generator load rejection APPLICABILITY to mitigate the increase in neutron flux, heat flux, and reactor pressure, and to increase the margin to the MCPR SL, and LHGR limits. The analytical methods and assumptions used in evaluating the turbine trip and generator load rejection are summarized in References 2, 3, and 4.

To mitigate pressurization transient effects, the EOC-RPT must trip the recirculation pumps after initiation of closure movement of either the TSVs or the TCVs. The combined effects of this trip and a scram reduce fuel bundle power more rapidly than a scram alone, resulting in an increased margin to the MCPR SL, and LHGR limits. Alternatively, MCPR limits for an inoperable EOC-RPT, as specified in the COLR, are sufficient to prevent violation of the MCPR Safety Limit, and fuel mechanical limits.

The EOC-RPT function is automatically disabled when turbine first stage pressure is < 30% RTP.

EOC-RPT instrumentation satisfies Criterion 3 of the NRC Policy Statement (Ref. 6).

(continued)

BFN-UNIT 1 B 3.3-106 Revision 0

EOC-RPT Instrumentation B 3.3.4.1 BASES APPLICABLE for calibration, process, and some of the instrument errors. The SAFETY ANALYSES, trip setpoints are then determined accounting for the remaining LCO, and instrument errors (e.g., drift). The trip setpoints derived in this APPLICABILITY manner provide adequate protection because instrumentation (continued) uncertainties, process effects, calibration tolerances, instrument drift, and severe environmental effects (for channels that must function in harsh environments as defined by 10 CFR 50.49) are accounted for.

The specific Applicable Safety Analysis, LCO, and Applicability discussions are listed below on a Function by Function basis.

Alternatively, since this instrumentation protects against a MCPR SL violation, with the instrumentation inoperable, modifications to the MCPR limits (LCO 3.2.2) may be applied to allow this LCO to be met. The MCPR and LHGR penaltiesy for the EOC-RPT inoperable condition isare specified in the COLR.

Turbine StoD Valve - Closure Closure of the TSVs and a main turbine trip result in the loss of a heat sink that produces reactor pressure, neutron flux, and heat flux transients that must be limited. Therefore, an RPT is initiated on TSV - Closure in anticipation of the transients that would result from closure of these valves. EOC-RPT decreases reactor power and aids the reactor scram in ensuring that the MCPR SL and LHGR limits are-is not exceeded during the worst case transient.

(continued)

BFN-UNIT 1 B 3.3-108 Revision G

EOC-RPT Instrumentation B 3.3.4.1 BASES APPLICABLE Turbine Stop Valve - Closure (continued)

SAFETY ANALYSES, LCO, and Closure of the TSVs is determined by measuring the position of APPLICABILITY each valve. There are two separate position signals associated with each stop valve, the signal from each switch being assigned to a separate trip channel. The logic for the TSV -

Closure Function is such that two or more TSVs must be closed to produce an EOC-RPT. This Function must be enabled at THERMAL POWER > 30% RTP. This is normally accomplished automatically by pressure transmitters sensing turbine first stage pressure; therefore, opening the turbine bypass valves may affect this function. To consider this function OPERABLE, bypass of the function must not occur when bypass valves are open. Four channels of TSV - Closure, with two channels in each trip system, are available and required to be OPERABLE to ensure that no single instrument failure will preclude an EOC-RPT from this Function on a valid signal. The TSV -

Closure Allowable Value is selected to detect imminent TSV closure.

This protection is required, consistent with the safety analysis assumptions, whenever THERMAL POWER is > 30% RTP.

Below 30% RTP, the Reactor Vessel Steam Dome Pressure -

High and the Average Power Range Monitor (APRM) Fixed Neutron Flux - High Functions of the Reactor Protection System (RPS) are adequate to maintain the necessary margin to the MCPR Safety-Li#it, and LHGR limits.

(continued)

BFN-UNIT 1 B 3.3-109 Revision 0

EOC-RPT Instrumentation B 3.3.4.1 BASES APPLICABLE Turbine Control Valve Fast Closure. TriD Oil Pressure - Low SAFETY ANALYSES, (PS-47-142, PS-47-144, PS-47-146, and PS-47-148)

LCO, and APPLICABILITY Fast closure of the TCVs during a generator load rejection (continued) results in the loss of a heat sink that produces reactor pressure, neutron flux, and heat flux transients that must be limited.

Therefore, an RPT is initiated on TCV Fast Closure, Trip Oil Pressure - Low in anticipation of the transients that would result from the closure of these valves. The EOC-RPT decreases reactor power and aids the reactor scram in ensuring that the MCPR SL, and LHGR limits are-is not exceeded during the worst case transient.

Fast closure of the TCVs is determined by measuring the electrohydraulic control fluid pressure at each control valve.

There is one pressure switch associated with each control valve, and the signal from each switch is assigned to a separate trip channel. The logic for the TCV Fast Closure, Trip Oil Pressure - Low Function is such that two or more TCVs must be closed (pressure switch trips) to produce an EOC-RPT. This Function must be enabled at THERMAL POWER > 30% RTP.

This is normally accomplished automatically by pressure transmitters sensing turbine first stage pressure; therefore, opening the turbine bypass valves may affect this function. To consider this function OPERABLE, bypass of the function must not occur when bypass valves are open. Four channels of TCV Fast Closure, Trip Oil Pressure - Low, with two channels in each trip system, are available and required to be OPERABLE to ensure that no single instrument failure will preclude an EOC-RPT from this Function on a valid signal. The TCV Fast Closure, Trip Oil Pressure - Low Allowable Value is selected high enough to detect imminent TCV fast closure.

(continued)

BFN-UNIT 1 B 3.3-110 Revision 0

EOC-RPT Instrumentation B 3.3.4.1 BASES ACTIONS A. 1 (continued)

With one or more channels inoperable, but with EOC-RPT trip capability maintained (refer to Required Actions B.1 and B.2 Bases), the EOC-RPT System is capable of performing the intended function. However, the reliability and redundancy of the EOC-RPT instrumentation is reduced such that a single failure in the remaining trip system could result in the inability of the EOC-RPT System to perform the intended function.

Therefore, only a limited time is allowed to restore compliance with the LCO. Because of the diversity of sensors available to provide trip signals, the low probability of extensive numbers of inoperabilities affecting all diverse Functions, and the low probability of an event requiring the initiation of an EOC-RPT, 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> is provided to restore the inoperable channels (Required Action A.1) or apply the EOC-RPT inoperable MCPR and LHGR limits. Alternately, the inoperable channels may be placed in trip (Required Action A.2) since this would conservatively compensate for the inoperability, restore capability to accommodate a single failure, and allow operation to continue. As noted, placing the channel in trip with no further restrictions is not allowed if the inoperable channel is the result of an inoperable breaker, since this may not adequately compensate for the inoperable breaker (e.g., the breaker may be inoperable such that it will not open). If it is not desired to place the channel in trip (e.g., as in the case where placing the inoperable channel in trip would result in an RPT, or if the inoperable channel is the result of an inoperable breaker),

Condition C must be entered and its Required Actions taken.

(continued)

BFN-UNIT 1 B 3.3-112 Revision 0

EOC-RPT Instrumentation B 3.3.4.1 BASES ACTIONS B.1 and B.2 (continued)

Required Actions B.1 and B.2 are intended to ensure that appropriate actions are taken if multiple, inoperable, untripped channels within the same Function result in the Function not maintaining EOC-RPT trip capability. A Function is considered to be maintaining EOC-RPT trip capability when sufficient channels are OPERABLE or in trip, such that the EOC-RPT System will generate a trip signal from the given Function on a valid signal and both recirculation pumps can be tripped.

Alternately, Required Action B.2 requires the MCPR and LHGR limits for inoperable EOC-RPT, as specified in the COLR, to be applied. This also restores the margin to MCPR and LHGR limits assumed in the safety analysis.

The 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> Completion Time is sufficient time for the operator to take corrective action, and takes into account the likelihood of an event requiring actuation of the EOC-RPT instrumentation during this period. It is also consistent with the 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> Completion Time provided in LCO 3.2.2 for Required Action A.1, since this instrumentation's purpose is to preclude a MCPR or LHGR violation.

C.1 With any Required Action and associated Completion Time not met, THERMAL POWER must be reduced to < 30% RTP within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. The allowed Completion Time of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is reasonable, based on operating experience, to reduce THERMAL POWER to < 30% RTP from full power conditions in an orderly manner and without challenging plant systems.

(continued)

BFN-UNIT 1 B 3.3-113 Revision 0

Main Turbine Bypass System B 3.7.5 BASES (continued)

APPLICABLE The Main Turbine Bypass System is assumed to function during SAFETY ANALYSES abnormal operational transients (e.g., the feedwater controller failure-maximum demand event), as discussed in the FSAR, Section 14.5.1.1 (Ref. 2). Opening the bypass valves during the event mitigates the increase in reactor vessel pressure, which affects the MCPR during the event. An inoperable Main Turbine Bypass System may result in APLHGR,-aod MCPR, and LHGR penalties.

The Main Turbine Bypass System satisfies Criterion 3 of the NRC Policy Statement (Ref. 3).

LCO The Main Turbine Bypass System is required to be OPERABLE to limit peak pressure in the main steam lines and maintain reactor pressure within acceptable limits during events that cause rapid pressurization, so that the Safety Limit MCPR is no-t exceeded. With the Main Turbine Bypass System inoperable, modlifircations to the APLHGR limits (LC 3.2.1, "AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR))--an t.he M*rCPR lim4t-. (L1C" 3.2.2,"MINIMUM11,4 CRITICAL PDIWAIER RATIO0 (MCP2R)") may be applied to allow this LCO to be mnet.

The APLHGR and MCPR APLHGR limits, MCPR Safety Limit, and LHGR limits are not exceeded. With the Main Turbine Bypass System inoperable, modifications to the APLHGR limits (LCO 3.2.1, "AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR)"), the MCPR limits (LCO 3.2.2, "MINIMUM CRITICAL POWER RATIO (MCPR)"), and LHGR limits (LCO 3.2.3, "LINEAR HEAT GENERATION RATE (LHGR)") may be applied to allow this LCO to be met. The APLHGR, MCPR, and LHGR limits for the inoperable Main Turbine Bypass System are specified in the COLR. An OPERABLE Main Turbine Bypass System requires the bypass valves to open in response to increasing main steam line pressure. This response is within the assumptions of the applicable analysis (Ref. 2).

(continued)

BFN-UNIT 1 B 3.7-33 Revision 0

Main Turbine Bypass System B 3.7.5 BASES (continued)

APPLICABILITY The Main Turbine Bypass System is required to be OPERABLE at _Ž 25% RTP to ensure that the fuel cladding integrity Safety Limit is not violated during abnormal operational transients. As discussed in the Bases for LCO 3.2.1 and LCO 3.2.2, sufficient margin to these limits exists at < 25% RTP. Therefore, these requirements are only necessary when operating at or above this power level.

ACTIONS A.1 If the Main Turbine Bypass System is inoperable (one or more bypass valves inoperable), or the APLHGR,-apA MCPR, and LHGR limits for an inoperable Main Turbine Bypass System, as specified in the COLR, are not applied, the assumptions of the design basis transient analysis may not be met. Under such circumstances, prompt action should be taken to restore the Main Turbine Bypass System to OPERABLE status or adjust the APLHGR,-ao4 MCPR, and LHGR limits accordingly. The 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> Completion Time is reasonable, based on the time to complete the Required Action and the low probability of an event occurring during this period requiring the Main Turbine Bypass System.

B.1 If the Main Turbine Bypass System cannot be restored to OPERABLE status or the APLHGR,-an4 MCPR, and LHGR limits for an inoperable Main Turbine Bypass System are not applied, THERMAL POWER must be reduced to < 25% RTP.

As discussed in the Applicability section, operation at

< 25% RTP results in sufficient margin to the required limits, and the Main (continued)

BFN-UNIT 1 B 3.7-34 Revision 0

Control Rod Testing - Operating B 3.10.7 B 3.10 SPECIAL OPERATIONS B 3.10.7 Control Rod Testing - Operating BASES BACKGROUND The purpose of this Special Operations LCO is to permit control rod testing, while in MODES 1 and 2, by imposing certain administrative controls. Control rod patterns during startup conditions are controlled by the operator and the rod worth minimizer (RWM) (LCO 3.3.2.1, "Control Rod Block Instrumentation"), such that only the specified control rod sequences and relative positions required by LCO 3.1.6, "Rod Pattern Control," are allowed over the operating range from all control rods inserted to the low power setpoint (LPSP) of the RWM. The sequences effectively limit the potential amount and rate of reactivity increase that could occur during a control rod drop accident (CRDA). During these conditions, control rod testing is sometimes required that may result in control rod patterns not in compliance with the prescribed sequences of LCO 3.1.6. These tests include SDM demonstrations, control rod scram time testing, and control rod friction testing. This Special Operations LCO provides the necessary exemption to the requirements of LCO 3.1.6 and provides additional administrative controls to allow the deviations in such tests from the prescribed sequences in LCO 3.1.6.

APPLICABLE The analytical methods and assumptions used in evaluating SAFETY ANALYSES the CRDA are summarized in References 1 ,-a.i4 2, 3, and 4.

CRDA analyses assume the reactor operator follows prescribed withdrawal sequences. These sequences define the potential initial conditions for the CRDA analyses. The RWM provides backup to operator control of the withdrawal sequences to ensure the initial conditions of the CRDA analyses are not (continued)

BFN-UNIT 1 B 3.10-36 Revision O

Control Rod Testing - Operating B 3.10.7 BASES APPLICABLE violated. For special sequences developed for control rod SAFETY ANALYSES testing, the initial control rod patterns assumed in the safety (continued) analysis of References 1 ,-and 2, 3, and 4 may not be preserved.

Therefore, special CRDA analyses may be required to demonstrate that these special sequences will not result in unacceptable consequences, should a CRDA occur during the testing. These analyses, performed in accordance with an NRC approved methodology, are dependent on the specific test being performed.

As described in LCO 3.0.7, compliance with Special Operations LCOs is optional, and therefore, no criteria of the NRC Policy Statement apply. Special Operations LCOs provide flexibility to perform certain operations by appropriately modifying requirements of other LCOs. A discussion of the criteria satisfied for the other LCOs is provided in their respective Bases.

LCO As described in LCO 3.0.7, compliance with this Special Operations LCO is optional. Control rod testing may be performed in compliance with the prescribed sequences of LCO 3.1.6, and during these tests, no exceptions to the requirements of LCO 3.1.6 are necessary. For testing performed with a sequence not in compliance with LCO 3.1.6, the requirements of LCO 3.1.6 may be suspended, provided additional administrative controls are placed on the test to ensure that the assumptions of the special safety analysis for the test sequence are satisfied. Assurances that the test sequence is followed can be provided by either programming the test sequence into the RWM, with conformance verified as specified in SR 3.3.2.1.8 and allowing the RWM to monitor (continued)

BFN-UNIT 1 B 3.10-37 Revision G

Control Rod Testing - Operating B 3.10.7 BASES LCO control rod withdrawal and provide appropriate control rod (continued) blocks if necessary, or by verifying conformance to the approved test sequence by a second licensed operator or other qualified member of the technical staff (i.e., personnel trained in accordance with an approved training program for this test).

These controls are consistent with those normally applied to operation in the startup range as defined in the SRs and ACTIONS of LCO 3.3.2.1, "Control Rod Block Instrumentation."

APPLICABILITY Control rod testing, while in MODES 1 and 2, with THERMAL POWER greater than 10% RTP, is adequately controlled by the existing LCOs on power distribution limits and control rod block instrumentation. Control rod movement during these conditions is not restricted to prescribed sequences and can be performed within the constraints of LCO 3.2.1, "AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR)," LCO 3.2.2, "MINIMUM CRITICAL POWER RATIO (MCPR)," LCO 3.2.3, "LINEAR HEAT GENERATION RATE (LHGR)," and LCO 3.3.2.1. With THERMAL POWER less than or equal to 10% RTP, the provisions of this Special Operations LCO are necessary to perform special tests that are not in conformance with the prescribed sequences of LCO 3.1.6.

While in MODES 3 and 4, control rod withdrawal is only allowed if performed in accordance with Special Operations LCO 3.10.3, "Single Control Rod Withdrawal - Hot Shutdown," or Special Operations LCO 3.10.4, "Single Control Rod Withdrawal - Cold Shutdown," which provide adequate controls to ensure that the assumptions of the safety analyses of References 1,-and 2, 3, and 4 are (continued)

BFN-UNIT 1 B 3.10-38 Revision O

Control Rod Testing - Operating B 3.10.7 BASES SURVEILLANCE SR 3.10.7.2 REQUIREMENTS (continued) When the RWM provides conformance to the special test sequence, the test sequence must be verified to be correctly loaded into the RWM prior to control rod movement. This Surveillance demonstrates compliance with SR 3.3.2.1.8, thereby demonstrating that the RWM is OPERABLE. A Note has been added to indicate that this Surveillance does not need to be performed if SR 3.10.7.1 is satisfied.

REFERENCES 1. NEDE-24011-P-A-13, "General Electric Standard Application for Reactor Fuel," August 1996.

2. Letter from T. Pickens (BWROG) to G. C. Lainas (NRC)

"Amendment 17 to General Electric Licensing Topical Report NEDE-24011-P-A," August 15, 1986.

3. XN-NF-80-19(P)(A), Volume 1 "Exxon Nuclear Methodology for Boiling Water Reactors - Neutronic Methods for Design and Analysis," (as identified in the COLR).

4. EMF-2158(P)(A), "Siemens Power Corporation Methodology for Boiling Water Reactors: Evaluation and Validation of CASMO-4/MICROBURN-B2," (as identified in the COLR).

BFN-UNIT 1 B 3.10-40 Revision 0

SDM Test -Refueling B 3.10.8 BASES (continued)

APPLICABLE Prevention and mitigation of unacceptable reactivity excursions SAFETY ANALYSES during control rod withdrawal, with the reactor mode switch in the startup/hot standby position while in MODE 5, is provided by the intermediate range monitor (IRM) neutron flux scram (LCO 3.3.1.1, "Reactor Protection System (RPS)

Instrumentation"), and control rod block instrumentation (LCO 3.3.2.1, "Control Rod Block Instrumentation"). The limiting reactivity excursion during startup conditions while in MODE 5 is the control rod drop accident (CRDA).

CRDA analyses assume that the reactor operator follows prescribed withdrawal sequences. For SDM tests performed within these defined sequences, the analyses of References 1,-

a~d 2, 3, and 4 are applicable. However, for some sequences developed for the SDM testing, the control rod patterns assumed in the safety analyses of References 1 ,-aAd 2, 3, and 4 may not be met. Therefore, special CRDA analyses, performed in accordance with an NRC approved methodology, may be required to demonstrate the SDM test sequence will not result in unacceptable consequences should a CRDA occur during the testing. For the purpose of this test, the protection provided by the normally required MODE 5 applicable LCOs, in addition to the requirements of this LCO, will maintain normal test operations as well as postulated accidents within the bounds of the appropriate safety analyses (Refs. 1,-ad 2, 3, and 4). In addition to the added requirements for the RWM, APRM, and control rod coupling, the notch out mode is specified for out of sequence withdrawals. Requiring the notch out mode limits withdrawal steps to a single notch, which limits inserted reactivity, and allows adequate monitoring of changes in neutron flux, which may occur during the test.

(continued)

BFN-UNIT 1 B 3.10-42 Revision 0

SDM Test -Refueling B 3.10.8 BASES SURVEILLANCE SR 3.10.8.5 REQUIREMENTS (continued) Coupling verification is performed to ensure the control rod is connected to the control rod drive mechanism and will perform its intended function when necessary. The verification is required to be performed any time a control rod is withdrawn to the "full out" notch position, or prior to declaring the control rod OPERABLE after work on the control rod or CRD System that could affect coupling. This Frequency is acceptable, considering the low probability that a control rod will become uncoupled when it is not being moved as well as operating experience related to uncoupling events.

SR 3.10.8.6 CRD charging water header pressure verification is performed to ensure the motive force is available to scram the control rods in the event of a scram signal. Since the reactor is depressurized in MODE 5, there is insufficient reactor pressure to scram the control rods. Verification of charging water pressure ensures that if a scram is required, capability for rapid control rod insertion would exist. The minimum pressure of 940 psig, which is well below the expected pressure of approximately 1100 psig, ensures sufficient pressure for rapid control rod insertion. The 7 day Frequency has been shown to be acceptable through operating experience and takes into account indications available in the control room.

REFERENCES 1. NEDE-2401 1-P-A-1 3, "General Electric Standard Application for Reactor Fuel," August 1996.

2. Letter from T. Pickens (BWROG) to G. C. Lainas, NRC, "Amendment 17 to General Electric Licensing Topical Report NEDE-24011 -P-A," August 15, 1986.

(continued)

BFN-UNIT 1 B 3.10-48 Revision 0

SDM Test -Refueling B 3.10.8 BASES REFERENCES 3. XN-NF-80-19(P)(A), Volume 1 "Exxon Nuclear (continued) Methodology for Boiling Water Reactors - Neutronic Methods for Design and Analysis," (as identified in the COLR).

4. EMF-2158(P)(A), "Siemens Power Corporation Methodology for Boiling Water Reactors: Evaluation and Validation of CASMO-4/MICROBURN-B2," (as identified in the COLR).

BFN-UNIT 1 B 3.10-48a Revision 8

ATTACHMENT 5 Browns Ferry Nuclear Plant (BFN)

Unit 1 Technical Specifications (TS) Change 467 Revision of Technical Specifications to allow utilization of AREVA NP fuel and associated analysis methodologies Retyped Proposed Technical Specification Bases Pages The following pages have been revised to reflect the proposed changes. These are the retyped pages relative to the markups found in Attachment 4.

Reactor Core SLs B 2.1.1 B 2.0 SAFETY LIMITS (SLs)

B 2.1.1 Reactor Core SLs BASES BACKGROUND GDC 10 (Ref. 1) requires, and SLs ensure, that specified acceptable fuel design limits are not exceeded during steady state operation, normal operational transients, and abnormal operational transients.

The fuel cladding integrity SL is set such that no fuel damage is calculated to occur if the limit is not violated. Because fuel damage is not directly observable, a stepback approach is used to establish an SL, such that the MCPR is not less than the limit specified in Specification 2.1.1.2. MCPR greater than the specified limit represents a conservative margin relative to the conditions required to maintain fuel cladding integrity.

The fuel cladding is one of the physical barriers that separate the radioactive materials from the environs. The integrity of this cladding barrier is related to its relative freedom from perforations or cracking. Although some corrosion or use related cracking may occur during the life of the cladding, fission product migration from this source is incrementally cumulative and continuously measurable. Fuel cladding perforations, however, can result from thermal stresses, which occur from reactor operation significantly above design conditions.

While fission product migration from cladding perforation is just as measurable as that from use related cracking, the thermally caused cladding perforations signal a threshold beyond which still greater thermal stresses may cause gross, rather than (continued)

BFN-UNIT 1 B 2.0-1 Revision 0

Reactor Core SLs B 2.1.1 BASES APPLICABLE 2.1.1.1 Fuel Claddingq Integrity SAFETY ANALYSES (continued) The SPCB critical power correlation is used for both AREVA and coresident fuel and is valid at pressures >700 psia, and bundle mass fluxes >0.1 X 106 Ibm/hr-ft 2 (>12,000 Ibm/hr, i.e., Ž10% core flow on a per bundle basis) for ATRIUM-10 and GE14 fuel types. For thermal margin monitoring at 25% power and higher, the hot channel flow rate will be >28,000 Ibm/hr (core flow not less than natural circulation, i.e., -25%-30% core flow for 25%

power); therefore, the fuel cladding integrity SL is conservative relative to the applicable range of the SPCB critical power correlation. For operation at low pressures or low flows, another basis is used, as follows:

The static head across the fuel bundles due only to elevation effects from liquid only in the channel, core bypass region, and annulus at zero power, zero flow is approximately 4.5 psi. At all operating conditions, this pressure differential is maintained by the bypass region of the core and the annulus region of the vessel. The elevation head provided by the annulus produces natural circulation flow conditions which have balancing pressure head and loss terms inside the core shroud. This natural circulation principle maintains a core plenum to plenum pressure drop of about 4.5 to 5 psid along the natural circulation flow line of the P/F operating map. In the range of power levels of interest, approaching 25% of rated power below which thermal margin monitoring is not required, the pressure drop and density head terms tradeoff for power changes such that natural circulation flow is nearly independent of reactor power.

This characteristic is represented by the nearly vertical portion of the natural circulation line on the P/F operating map.

Analysis has shown that the hot channel flow rate is >28,000 Ibm/hr (>0.23 x 106 Ibm/hr-ft2) in the region of operation with power -25% and core pressure drop of about 4.5 to 5 psid. Full scale ATLAS test data taken at pressures from 14.7 psia to 800 psia indicate that the fuel assembly critical power at 28,000 Ibm/hr is approximately 3 MWt. With the design peaking factors, this corresponds to a core thermal power of more than 50%.

(continued)

BFN-UNIT 1 B 2.0-3 Revision 0 1

Reactor Core SLs B 2.1.1 BASES APPLICABLE 2.1.1.1 Fuel Cladding Integrity (continued)

SAFETY ANALYSES Thus operation up to 25% of rated power with normal natural circulation available is conservatively acceptable even if reactor pressure is equal to or below 800 psia. If reactor power is significantly less than 25% of rated (e.g.,

below 10% of rated), the core flow and the channel flow supported by the available driving head may be less than 28,000 Ibm/hr (along the lower portion of the natural circulation flow characteristic on the P/F map). However, the critical power that can be supported by the core and hot channel flow with normal natural circulation paths available remains well above the actual power conditions. The inherent characteristics of BWR natural circulation make power and core flow follow the natural circulation line as long as normal water level is maintained.

Thus, operation with core thermal power below 25% of rated without thermal margin surveillance is conservatively acceptable even for reactor operations at natural circulation. Adequate fuel thermal margins are also maintained without further surveillance for the low power conditions that would be present if core natural circulation is below 10% of rated flow.

(continued)

BFN-UNIT 1 & 2.0-4 Revision 1

Reactor Core SLs B 2.1.1 BASES APPLICABLE 2.1.1.2 MCPR SAFETY ANALYSES (continued) The fuel cladding integrity SL is set such that no fuel damage is calculated to occur if the limit is not violated. Since the parameters that result in fuel damage are not directly observable during reactor operation, the thermal and hydraulic conditions that result in the onset of transition boiling have been used to mark the beginning of the region in which fuel damage could occur. Although it is recognized that the onset of transition boiling would not result in damage to BWR fuel rods, the critical power at which boiling transition is calculated to occur has been adopted as a convenient limit. However, the uncertainties in monitoring the core operating state and in the procedures used to calculate the critical power result in an uncertainty in the value of the critical power. Therefore, the fuel cladding integrity SL is defined as the critical power ratio in the limiting fuel assembly for which more than 99.9% of the fuel rods in the core are expected to avoid boiling transition, considering the power distribution within the core and all uncertainties.

The MCPR SL is determined using a statistical model combining all the uncertainties in operating parameters and the procedures used to calculate critical power. The probability of the occurrence of boiling transition is determined using the approved AREVA critical power correlations. One specific uncertainty included in the SL is the uncertainty inherent in the SPCB critical power correlation. References 2, 3, and 4 describe the uncertainties and methodologies used in determining the MCPR SL.

(continued)

BFN-UNIT 1 B 2.0-5 Revision 0 1

Reactor Core SLs B 2.1.1 BASES (continued)

SAFETY LIMIT Exceeding an SL may cause fuel damage and create a potential VIOLATIONS for radioactive releases in excess of 10 CFR 50.67, "Accident Source Term," limits (Ref. 5). Therefore, it is required to insert all insertable control rods and restore compliance with the SLs within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. The 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> Completion Time ensures that the operators take prompt remedial action and also ensures that the probability of an accident occurring during this period is minimal.

REFERENCES 1. 10 CFR 50, Appendix A, GDC 10.

2. EMF-2209(P)(A), "SPCB Critical Power Correlation,"

(as identified in the COLR).

3. EMF-2245(P)(A), "Application of Siemens Power Corporation's Critical Power Correlations to Co-Resident Fuel," (as identified in the COLR).

4. ANF-524(P)(A), "ANF Critical Power Methodology for Boiling Water Reactors," (as identified in the COLR).

5. 10 CFR 50.67.

BFN-UNIT 1 B 2.0-7 Revision OQ.F2 January 25, 2005

SDM B 3.1.1 B 3.1 REACTIVITY CONTROL SYSTEMS B 3.1.1 SHUTDOWN MARGIN (SDM)

BASES BACKGROUND SDM requirements are specified to ensure:

a. The reactor can be made subcritical from all operating conditions and transients and Design Basis Events;

b. The reactivity transients associated with postulated accident conditions are controllable within acceptable limits; and

c. The reactor will be maintained sufficiently subcritical to preclude inadvertent criticality in the shutdown condition.

These requirements are satisfied by the control rods, as described in GDC 26 (Ref. 1), which can compensate for the reactivity effects of the fuel and water temperature changes experienced during all operating conditions.

APPLICABLE The control rod drop accident (CRDA) analysis (Refs. 2, 3, 9, SAFETY ANALYSES and 10) assumes the core is subcritical with the highest worth control rod withdrawn. Typically, the first control rod withdrawn has a very high reactivity worth and, should the core be critical during the withdrawal of the first control rod, the consequences of a CRDA could exceed the fuel damage limits for a CRDA (see Bases for LCO 3.1.6, "Rod Pattern Control"). Also, SDM is assumed as an initial condition for the control rod removal error during refueling (Ref. 4) and fuel assembly insertion error during refueling (Ref. 5) accidents. The analysis of these reactivity insertion events assumes the refueling interlocks are OPERABLE when the reactor is in the refueling mode of operation. These interlocks prevent the withdrawal of more (continued)

BFN-UNIT 1 B 3.1 -1 Revision 0 1

SDM B 3.1.1 BASES (continued)

REFERENCES 1. 10 CFR 50, Appendix A, GDC 26.

2. FSAR, Section 14.6.2.

3. NEDE-2401 1-P-A-13, "General Electric Standard Application for Reactor Fuel," Section S.2.2.3.1, August 1996.

4. FSAR, Section 14.5.3.3.

5. FSAR, Section 14.5.3.4.

6. FSAR, Section 3.6.5.2.

7. NEDE-2401 1-P-A-1 3, "General Electric Standard Application for Reactor Fuel," Section 3.2.4.1, August 1996.

8. NRC 93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.

9. XN-NF-80-19(P)(A), Volume 1 "Exxon Nuclear Methodology for Boiling Water Reactors - Neutronic Methods for Design and Analysis," (as identified in the COLR).

10. EMF-2158(P)(A), "Siemens Power Corporation Methodology for Boiling Water Reactors: Evaluation and Validation of CASMO-4/MICROBURN-B2," (as identified in the COLR).

BFN-UNIT 1 B 3.1-8 Revision G I

Control Rod Scram Times B 3.1.4 BASES (continued)

APPLICABLE The analytical methods and assumptions used in evaluating the SAFETY ANALYSES control rod scram function are presented in References 2, 3, and 4. The Design Basis Accident (DBA) and transient analyses assume that all of the control rods scram at a specified insertion rate. The resulting negative scram reactivity forms the basis for the determination of plant thermal limits (e.g., the MCPR). Other distributions of scram times (e.g., several control rods scramming slower than the average time with several control rods scramming faster than the average time) can also provide sufficient scram reactivity. Surveillance of each individual control rod's scram time ensures the scram reactivity assumed in the DBA and transient analyses can be met.

The scram function of the CRD System protects the MCPR Safety Limit (SL) (see Bases for SL 2.1.1, "Reactor Core SLs,"

and LCO 3.2.2, "MINIMUM CRITICAL POWER RATIO (MCPR)") and the 1% cladding plastic strain fuel design limit (see Bases for LCO 3.2.1, "AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR)", and LCO 3.2.3, "LINEAR HEAT GENERATION RATE (LHGR)"), which ensure that no fuel damage will occur if these limits are not exceeded. Above 800 psig, the scram function is designed to insert negative reactivity at a rate fast enough to prevent the actual MCPR from

-becoming less than the MCPR SL, during the analyzed limiting power transient. Below 800 psig, the scram function is assumed to perform during the control rod drop accident (Refs. 5, 8, and 9) and, therefore, also provides protection against violating fuel damage limits during reactivity insertion accidents (see Bases for LCO 3.1.6, "Rod Pattern Control").

For the reactor vessel overpressure protection analysis, the scram function, along with the safety/relief valves, ensure that the peak vessel pressure is maintained within the applicable ASME Code limits.

Control rod scram times satisfy Criterion 3 of the NRC Policy Statement (Ref. 7).

(continued)

BFN-UNIT 1 B 3.1-27 Revision 0 1

Control Rod Scram Times B 3.1.4 BASES (continued)

REFERENCES 1. 10 CFR 50, Appendix A, GDC 10.

2. FSAR, Section 3.4.6.

3. FSAR, Section 14.5.

4. FSAR, Section 14.6.

5. NEDE-2401 1-P-A-1 3, "General Electric Standard Application for Reactor Fuel," Section 3.2.4.1, August 1996.

6. Letter from R. F. Janecek (BWROG) to R. W. Starostecki (NRC), "BWR Owners Group Revised Reactivity Control System Technical Specifications," BWROG-8754, September 17, 1987.

7. NRC No.93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.

8. XN-NF-80-19(P)(A), Volume 1 "Exxon Nuclear Methodology for Boiling Water Reactors - Neutronic Methods for Design and Analysis," (as identified in the COLR).

9. EMF-2158(P)(A), "Siemens Power Corporation Methodology for Boiling Water Reactors: Evaluation and Validation of CASMO-4/MICROBURN-B2," (as identified in the COLR).

BFN-UNIT 1 B 3.1-34 Revision 0 1

Rod Pattern Control B 3.1.6 B 3.1 REACTIVITY CONTROL SYSTEMS B 3.1.6 Rod Pattern Control BASES BACKGROUND Control rod patterns during startup conditions are controlled by the operator and the rod worth minimizer (RWM) (LCO 3.3.2.1, "Control Rod Block Instrumentation"), so that only specified control rod sequences and relative positions are allowed over the operating range of all control rods inserted to 10% RTP.

The sequences limit the potential amount of reactivity addition that could occur in the event of a Control Rod Drop Accident (CRDA).

This Specification assures that the control rod patterns are consistent with the assumptions of the CRDA analyses of References 1,2, 10, and 11.

APPLICABLE The analytical methods and assumptions used in evaluating SAFETY ANALYSES the CRDA are summarized in References 1, 2, 10, and 11.

CRDA analyses assume that the reactor operator follows prescribed withdrawal sequences. These sequences define the potential initial conditions for the CRDA analysis. The RWM (LCO 3.3.2.1) provides backup to operator control of the withdrawal sequences to ensure that the initial conditions of the CRDA analysis are not violated.

Prevention or mitigation of positive reactivity insertion events is necessary to limit the energy deposition in the fuel, thereby preventing significant fuel damage which could result in the undue release of radioactivity. Since the failure consequences for U0 2 have been shown to be insignificant below fuel energy (continued)

BFN-UNIT 1 B 3.1-41 Revision 0

Rod Pattern Control B 3.1.6 BASES APPLICABLE depositions of 300 cal/gm (Ref. 3), the fuel damage limit of SAFETY ANALYSES 280 cal/gm provides a margin of safety from significant core (continued) damage which would result in release of radioactivity (Refs. 4 and 5). Generic evaluations (Refs. 1, 6, and 10) of a design basis CRDA (i.e., a CRDA resulting in a peak fuel energy deposition of 280 cal/gm) have shown that if the peak fuel enthalpy remains below 280 cal/gm, then the maximum reactor pressure will be less than the required ASME Code limits (Ref. 7) and the calculated offsite doses will be well within the required limits (Ref. 5).

Control rod patterns analyzed in References 1, 10, and 11, follow the banked position withdrawal sequence (BPWS). The BPWS is applicable from the condition of all control rods fully inserted to 10% RTP (Ref. 2). For the BPWS, the control rods are required to be moved in groups, with all control rods assigned to a specific group required to be within specified banked positions (e.g., between notches 08 and 12). The banked positions are established to minimize the maximum incremental control rod worth without being overly restrictive during normal plant operation. Analyses are performed using the Reference 10 methodology demonstrating the 280 cal/gm fuel damage limit will not be violated during a CRDA while following the BPWS mode of operation. The evaluation provided by the generic BPWS analysis (Ref.,8) allows a limited number (i.e., eight) and corresponding distribution of fully inserted, inoperable control rods, that are not in compliance with the sequence.

Rod pattern control satisfies Criterion 3 of the NRC Policy Statement (Ref. 9).

(continued)

BFN-UNIT 1 B 3.1-42 Revision 0

Rod Pattern Control B 3.1.6 BASES (continued)

REFERENCES 1. NEDE-2401 1-P-A-1 3, "General Electric Standard Application for Reactor Fuel," Section 2.2.3.1, August 1996.

2. Letter from T. Pickens (BWROG) to G. C. Lainas (NRC),

Amendment 17 to General Electric Licensing Topical Report, NEDE-24011-P-A, August 15, 1986.

3. NUREG-0979, Section 4.2.1.3.2, April 1983.

4. NUREG-0800, Section 15.0.1.

5. 10 CFR 50.67.

6. NEDO-21778-A, "Transient Pressure Rises Affected Fracture Toughness Requirements for Boiling Water Reactors," December 1978.

7. ASME, Boiler and Pressure Vessel Code.

8. NEDO-21231, "Banked Position Withdrawal Sequence,"

January 1977.

9. NRC No.93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.

10. XN-NF-80-19(P)(A), Volume 1 "Exxon Nuclear Methodology for Boiling Water Reactors - Neutronic Methods for Design and Analysis," (as identified in the COLR).

11. EMF-2158(P)(A), "Siemens Power Corporation Methodology for Boiling Water Reactors: Evaluation and Validation of CASMO-4/MICROBURN-B2," (as identified in the COLR).

BFN-UNIT 1 B 3.1-46 Revision 0-2-9 januaFy2,,,2005

(

APLHGR B 3.2.1 B 3.2 POWER DISTRIBUTION LIMITS B 3.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR)

BASES

(

BACKGROUND The APLHGR is a measure of the average LHGR of all the fuel rods in a fuel assembly at any axial location. Limits on the APLHGR are specified to ensure that the fuel design limits identified in Reference 1 are not exceeded during abnormal operational transients and that the peak cladding temperature (PCT) during the postulated design basis loss of coolant accident (LOCA) does not exceed the limits specified in 10 CFR 50.46.

APPLICABLE The analytical methods and assumptions used in evaluating SAFETY ANALYSES the fuel design limits are presented in References 1, 2, and 11.

The analytical methods and assumptions used in evaluating Design Basis Accidents (DBAs), abnormal operational transients, and normal operation that determine the APLHGR limits are presented in References 1, 2, 3, 4, and 7 for GE fuel; References 11, 12, 13, 14, and 15 for AREVA fuel.

(continued)

BFN-UNIT 1 B 3.2-1 Revision 07-4, Octobc~r 26,'20065

APLHGR B 3.2.1 BASES APPLICABLE GE Fuel SAFETY ANALYSES (continued) LOCA analyses are performed to ensure APLHGR limits are adequate to meet the PCT and maximum oxidation limits of 10 CFR 50.46. The analysis is performed using calculational models that are consistent with the requirements of 10 CFR 50, Appendix K. A complete discussion of the analysis code is provided in Reference 5. The PCT following a postulated LOCA is a function of the average heat generation rate of all the rods of a fuel assembly at any axial location and is not strongly influenced by the rod to rod power distribution within an

-assembly. The APLHGR limits specified are equivalent to the LHGR of the highest powered fuel rod assumed in the LOCA analysis divided by its local peaking factor. A conservative multiplier is applied to the LHGR assumed in the LOCA analysis to account for the uncertainty associated with the measurement of the APLHGR.

For single recirculation loop operation, an APLHGR multiplier is applied to the APLHGR limit (Ref. 5 and Ref. 10). The multiplier is documented in the COLR. This multiplier is due to the conservative analysis assumption of an earlier departure from nucleate boiling with one recirculation loop available, resulting in a more severe heatup during a LOCA.

(continued)

Revision 40 BFN-UNIT 1 B 3.2-2

,A.mendment hi. 236 October 26, 2006

APLHGR B 3.2.1 BASES APPLICABLE AREVA Fuel SAFETY ANALYSES (continued) For AREVA fuel, the APLHGR limits are developed as a function of exposure and, along with the LHGR limits, ensure adherence to fuel design limits during abnormal operational transients. No power- or flow-dependent corrections are applied to the APLHGR (referred to as the maximum APLHGR or MAPLHGR).

AREVA APLHGR limits are intended to be bound by the LHGR limits.

The calculational procedure used to establish the AREVA fuel MAPLHGR limits is based on LOCA analyses as defined in 10 CFR 50.46, Appendix K. MAPLHGR limits are created to assure that the peak cladding temperature of AREVA fuel following a postulated design basis LOCA will not exceed the PCT and maximum oxidation limits specified in 10 CFR 50.46, Appendix K. The calculational models and methodology are described in References 11 and 12.

The AREVA fuel MAPLHGR limits for two-loop operation are specified in the COLR. For single-loop operation, a MAPLHGR multiplier is applied to the MAPLHGR limit (Reference 11). The multiplier is documented in the COLR.

The APLHGR satisfies Criterion 2 of the NRC Policy Statement (Ref. 6).

(continued)

BFN-UNIT 1 B 3.2-3 Revision 40 A~m.e ndmenmt hNo . 236 October 26, 2006

APLHGR B 3.2.1 BASES (continued)

LCO The APLHGR limits specified in the COLR are the result of the fuel design, DBA, and transient analyses. With only one recirculation loop in operation, in conformance with the requirements of LCO 3.4.1, "Recirculation Loops Operating," the limit is determined by multiplying the exposure dependent limit by an APLHGR correction factor (Ref. 5 and Ref. 10). Cycle specific APLHGR correction factors for single recirculation loop operation are documented in the COLR. APLHGR limits are selected such that no power or flow dependent corrections are required. Additional APLHGR operating limit adjustments may be provided in the COLR supporting other analyzed equipment out-of-service conditions.

APPLICABILITY' The APLHGR limits are primarily derived from fuel design evaluations and LOCA and transient analyses that are assumed to occur at high power levels. Design calculations (Ref. 4) and operating experience have shown that as power is reduced, the margin to the required APLHGR limits increases. This trend continues down to the power range of 5% to 15% RTP when entry into MODE 2 occurs. When in MODE 2, the intermediate range monitor scram function provides prompt scram initiation during any significant transient, thereby effectively removing any APLHGR limit compliance concern in MODE 2. Therefore, at THERMAL POWER levels < 25% RTP, the reactor is (continued)

BFN-UNIT 1 B 3.2-3a Revision 40 Amendment No. 23 October 26, 2006

APLHGR B 3.2.1 BASES SURVEILLANCE SR 3.2.1.1 (continued)

REQUIREMENTS operation. The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> allowance after THERMAL POWER

> 25% RTP is achieved is acceptable given the large inherent margin to operating limits at low power levels.

REFERENCES 1. NEDE-2401 1-P-A-1 3 "General Electric Standard Application for Reactor Fuel," August 1996.

2. FSAR, Chapter 3.

3. FSAR, Chapter 14.

4. FSAR, Appendix N.

5. NEDC-32484P, "Browns Ferry Nuclear Plant Units 1, 2, and 3, SAFER/GESTR-LOCA Loss-of-Coolant Accident Analysis," Revision 2, December 1997.

6. NRC No.93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.

7. NEDC-32433P, "Maximum Extended Load Line Limit and ARTS Improvement Program Analyses for Browns Ferry Nuclear Plant Units 1, 2, and 3," April 1995.

8. NEDO-30130-A, "Steady State Nuclear Methods,"

May 1985.

9. NEDO-24154, "Qualification of the One-Dimensional Core Transient Model for Boiling Water Reactors," October 1978.

10. NEDO-24236, "Browns Ferry Nuclear Plant Units 1, 2, and 3, Single-Loop Operation," May 1981.

11. EMF-2361(P)(A), "EXEM BWR-2000 ECCS Evaluation Model," (as identified in the COLR).

(continued)

BFN-UNIT 1 B 3.2-5 Revision 40 Amend~ment No. 236-Octobcr 26, 2006

APLHGR B 3.2.1 BASES REFERENCES 12. EMF-2292(P)(A), "ATRIUM TM-10: Appendix K Spray Heat (continued) Transfer Coefficients." (as identified in the COLR).

13. XN-NF-81-58(P)(A), "RODEX2 Fuel Rod Thermal-Mechanical Response Evaluation Model." (as identified in the COLR).

14. XN-NF-80-19(P)(A), Volume 1 "Exxon Nuclear Methodology for Boiling Water Reactors - Neutronic Methods for Design and Analysis," (as identified in the COLR).

15. XN-NF-80-19(P)(A) Volume 4, "Exxon Nuclear Methodology for Boiling Water Reactors: Application of the ENC Methodology to BWR Reloads." (as identified in the COLR).

BFN-UNIT 1 B 3.2-5a Revision 40

.A.mendme~nt No. 23 October 26, 2006

MCPR B 3.2.2 BASES (continued)

APPLICABLE The analytical methods and assumptions used in evaluating SAFETY ANALYSES the abnormal operational transients to establish the operating limit MCPR are presented in References 2, 3, 4, 5, 8, 10, 11, 12, 13, 14, and 15. To ensure that the MCPR SL is not exceeded during any transient event that occurs with moderate frequency, limiting transients have been analyzed to determine the largest reduction in critical power ratio (CPR). The types of transients evaluated are loss of flow, increase in pressure and power, positive reactivity insertion and coolant temperature decrease.

The limiting transient yields the largest change in CPR (ACPR).

When the largest ACPR is added to the MCPR SL, the required operating limit MCPR is obtained.

The MCPR operating limits derived from the transient analysis are dependent on the operating core flow and power to ensure adherence to fuel design limits during the worst transient that occurs with moderate frequency (Reference 8). Flow dependent MCPR (MCPRf) limits are determined by steady state thermal hydraulic methods using the three dimensional BWR simulator code (Ref. 12) and the multichannel thermal hydraulics code (Ref. 13). The operating limit is dependent on the maximum core flow limiter setting in the Recirculation Flow Control System.

Power dependent MCPR limits (MCPRp) are determined by the three-dimensional BWR simulator code (Ref. 12) and the one-dimensional transient codes (Refs. 14 and 15). Due to the sensitivity of the transient response to initial core flow levels at power levels below those at which the turbine control valve fast closure scrams are bypassed, high and low flow MCPRp operating limits are provided for operating between 25% RTP and the previously mentioned bypass power level.

The MCPR satisfies Criterion 2 of the NRC Policy Statement (Ref. 7).

(continued)

BFN-UNIT 1 B 3.2-7 Revision 40 Amendmnt No.23 October 26, 2006

MCPR B 3.2.2 BASES (continued)

LCO The MCPR operating limits specified in the COLR are the result of the Design Basis Accident (DBA) and transient analysis.

Additional MCPR operating limits supporting analyzed equipment out-of-service conditions are provided in the COLR.

The operating limit MCPR is determined by the larger of the MCPRf and MCPRp limits.

APPLICABILITY The MCPR operating limits are primarily derived from transient analyses that are assumed to occur at high power levels. Below 25% RTP, the reactor is operating at a minimum recirculation pump speed and the moderator void ratio is small. Surveillance of thermal limits below 25% RTP is unnecessary due to the large inherent margin that ensures that the MCPR SL is not exceeded even if a limiting transient occurs. Statistical analyses indicate that the nominal value-of the initial MCPR expected at 25% RTP is > 3.5. Studies of the variation of limiting transient behavior have been performed over the range of power and flow conditions. These studies encompass the range of key actual plant parameter values important to typically limiting transients.

The results of these studies demonstrate that a margin is expected between performance and the MCPR requirements, and that margins increase as power is reduced to 25% RTP.

This trend is expected to continue to the 5% to 15% power range when entry into MODE 2 occurs. When in MODE 2, the intermediate range monitor provides rapid scram initiation for any significant power increase transient, which effectively eliminates any MCPR compliance concern. Therefore, at THERMAL POWER levels < 25% RTP, the reactor is operating with substantial margin to the MCPR limits and this LCO is not required.

ACTIONS A.1 If any MCPR is outside the required limits, an assumption regarding an initial condition of the design basis transient analyses may not be met. Therefore, prompt action should be taken to restore the MCPR(s) to within the required limits such (continued)

BFN-UNIT 1 B 3.2-8 Revision 0-40 Octobcr 26, 2006

MCPR B 3.2.2 BASES SURVEILLANCE SR 3.2.2.2 REQUIREMENTS (continued) Because the transient analysis takes credit for conservatism in the scram speed performance, it must be demonstrated that the specific scram speed distribution is consistent with that used in the transient analysis. SR 3.2.2.2 determines the actual scram speed distribution and compares it with the assumed distribution. The MCPR operating limit is determined based either on the applicable limit associated with scram times of LCO 3.1.4, "Control Rod Scram Times," or the nominal scram times. The scram speed dependent MCPR limits are contained in the COLR. This determination must be performed within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> after each set of control rod scram time tests required by SR 3.1.4.1 and SR 3.1.4.2 because the effective scram speed distribution may change during the cycle. The 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Completion Time is acceptable due to the relatively minor changes in the actual control rod scram speed distribution expected during the fuel cycle.

REFERENCES 1. NUREG-0562, "Fuel Rod Failure As a Consequence of Departure from Nucleate Boiling or Dryout," June 1979.

2. NEDE-2401 1-P-A-1 3, "General Electric Standard Application for Reactor Fuel," August 1996.

3. FSAR, Chapter 3.

4. FSAR, Chapter 14.

5. FSAR, Appendix N.

6. NEDO-30130-A, "Steady State Nuclear Methods,"

May 1985.

7. NRC No.93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.

(continued)

BFN-UNIT 1 B 3.2-10 Revision 40 AmeGnedeF 2 2306

MCPR B 3.2.2 BASES REFERENCES 8. NEDC-32433P, "Maximum Extended Load Line Limit and (continued) ARTS Improvement Program Analyses for Browns Ferry Nuclear Plant Units 1, 2, and 3," April 1995.

9. NEDO-24154, "Qualification of the One-Dimensional Core Transient Model for Boiling Water Reactors," October 1978.

10. NEDO-24236, "Browns Ferry Nuclear Plant Units 1,2, and 3, Single-Loop Operation," May 1981.

11. ANF-524(P)(A), "ANF Critical Power Methodology for Boiling Water Reactors," (as identified in the COLR).

12. EMF-2158(P)(A,), "Siemens Power Corporation Methodology for Boiling Water Reactors: Evaluation and Validation of CASMO-4/MICROBURN-B2," (as identified in the COLR).

13. XN-NF-80-19(P)(A) Volume 3, "Exxon Nuclear Methodology for Boiling Water Reactors, THERMEX: Thermal Limits Methodology Summary Description," (as identified in the COLR).

14. ANF-913(P)(A) Volume 1, "COTRANSA2: A Computer Program for Boiling Water Reactor Transient Analyses," (as identified in the COLR).

15. XN-NF-84-105(P)(A) Volume 1, "XCOBRA-T: A Computer Code for BWR Transient Thermal-Hydraulic Core Analysis,"

(as identified in the COLR).

BFN-UNIT 1 B 3.2-10a Revision 40 OctObcl 26, 2096

LHGR B 3.2.3 BASES APPLICABLE A value of 1% plastic strain of the fuel cladding has been SAFETY ANALYSES defined as the limit below which fuel damage caused by (continued) overstraining of the fuel cladding is not expected to occur (Ref. 3).

Fuel design evaluations have been performed and demonstrate that the 1% fuel cladding plastic strain design limit is not exceeded during continuous operation with LHGRs up to the operating limit specified in the COLR. The analysis also includes allowances for short term transient operation above the operating limit to account for abnormal operational transients, plus an allowance for densification power spiking.

LHGR limits are multiplied by the smaller of either the flow-dependent LHGR factor (LHGRFACf) or the power-dependent LHGR factor (LHGRFACp) corresponding to the existing core flow and power state to ensure adherence to the fuel mechanical design bases during the limiting transient.

LHGRFACf is generated to protect the core from slow flow runout transients. A curve is provided based on the maximum credible flow runout transient. LHGRFACp is generated to protect the core from plant transients other than core flow increases. LHGRFAC multipliers are provided in the COLR.

The LHGR satisfies Criterion 2 of the NRC Policy Statement (Ref. 4).

LCO The LHGR is a basic assumption in the fuel design analysis.

The fuel has been designed to operate at rated core power with sufficient design margin to the LHGR calculated to cause a 1%

fuel cladding plastic strain. The operating limit to accomplish this objective is specified in the COLR.

Additional LHGR operating limits adjustments may be provided in the COLR to support analyzed equipment out-of-service operation.

(continued)

BFN-UNIT 1 B 3.2-12 Revision 0 1

LHGR B 3.2.3 BASES (continued) r APPLICABILITY The LHGR limits are derived from fuel design analysis that is limiting at high power level conditions. At core thermal power levels < 25% RTP, the reactor is operating with a substantial margin to the LHGR limits and, therefore, the Specification is only required when the reactor is operating at _>25% RTP.

r (continued)

BFN-UNIT 1 B 3.2-12a Revision 0

Control Rod Block Instrumentation B 3.3.2.1 BASES (continued)

APPLICABLE 1. Rod Block Monitor SAFETY ANALYSES, LCO, and The RBM is designed to prevent violation of the MCPR SL and APPLICABILITY the cladding 1% plastic strain fuel design limit that may result from a single control rod withdrawal error (RWE) event. The analytical methods and assumptions used in evaluating the RWE event are summarized in References 3, 12, and 13. A statistical analysis of RWE events was performed to determine the RBM response for both channels for each event. From these responses, the fuel thermal performance as a function of RBM Allowable Value was determined. The Allowable Values are chosen as a function of power level. Based on the specified Allowable Values, operating limits are established.

The RBM Function satisfies Criterion 3 of the NRC Policy Statement (Ref. 10).

Two channels of the RBM are required to be OPERABLE, with their setpoints within the appropriate Allowable Value for the associated power range to ensure that no single instrument failure can preclude a rod block from this Function. The setpoints are calibrated consistent with applicable setpoint methodology (nominal trip setpoint).

Nominal trip setpoints are specified in the setpoint calculations.

The nominal setpoints are selected to ensure that the setpoints do not exceed the Allowable Values between successive CHANNEL CALIBRATIONS. Operation with a trip setpoint less conservative than the nominal trip setpoint, but within its Allowable Value, is acceptable. Trip setpoints are those (continued)

BFN-UNIT 1 B 3.3-59 Revision 0-49 October 26, 2006

Control Rod Block Instrumentation B 3.3.2.1 BASES APPLICABLE 2. Rod Worth Minimizer SAFETY ANALYSES, LCO, and The RWM enforces the banked position withdrawal sequence APPLICABILITY (BPWS) to ensure that the initial conditions of the CRDA (continued) analysis are not violated. The analytical methods and assumptions used in evaluating the CRDA are summarized in References 4, 5, 6, 7, 12, and 13. The BPWS requires that control rods be moved in groups, with all control rods assigned to a specific group required to be within specified banked positions. Requirements that the control rod sequence is in compliance with the BPWS are specified in LCO 3.1.6, "Rod Pattern Control."

The RWM Function satisfies Criterion 3 of the NRC Policy Statement (Ref. 10).

Since the RWM is designed to act as a backup to operator control of the rod sequences, only one channel of the RWM is available and required to be OPERABLE (Ref. 7). Special circumstances provided for in the Required Action of LCO 3.1.3, "Control Rod OPERABILITY," and LCO 3.1.6 may necessitate bypassing the RWM to allow continued operation with inoperable control rods, or to allow correction of a control rod pattern not in compliance with the BPWS. The RWM may be bypassed as required by these conditions, but then it must be considered inoperable and the Required Actions of this LCO followed.

Compliance with the BPWS, and therefore OPERABILITY of the RWM, is required in MODES 1 and 2 when THERMAL POWER is < 10% RTP. When THERMAL POWER is > 10% RTP, there is no possible control rod configuration that results in a control (continued)

BFN-UNIT 1 B 3.3-61 Revision 0 1

Control Rod Block Instrumentation B 3.3.2.1 BASES (continued)

REFERENCES 1. FSAR, Section 7.5.8.2.3.

2. FSAR, Section 7.16.5.3.1.k.

3. NEDC-32433P, "Maximum Extended Load Line Limit and ARTS Improvement Program Analyses for Browns Ferry Nuclear Plant Unit 1,2 and 3," April 1995.

4. NEDE-2401 1-P-A-US, "General Electrical Standard Application for Reload Fuel," Supplement for United States, (revision specified in the COLR).

5. "Modifications to the Requirements for Control Rod Drop Accident Mitigating Systems," BWR Owners' Group, July 1986.

6. NEDO-21231, "Banked Position Withdrawal Sequence,"

January 1977.

7. NRC SER, "Acceptance of Referencing of Licensing Topical Report NEDE-2401 1-P-A," "General Electric Standard Application for Reactor Fuel, Revision 8, Amendment 17," December 27, 1987.

8. NEDC-30851-P-A, Supplement 1, "Technical Specification Improvement Analysis for BWR Control Rod Block Instrumentation," October 1988.

9. GENE-770-06-1, "Addendum to Bases for Changes to Surveillance Test Intervals and Allowed Out-of-Service Times for Selected Instrumentation Technical Specifications," February 1991.

10. NRC No.93-102, "Final Policy Statement on Technical Specification Improvements," July 23, 1993.

11. NEDC-3241 OP-A, "Nuclear Measurement Analysis and Control Power Range Neutron Monitor (NUMAC PRNM)

Retrofit Plus Option III Stability Trip Function,"

October 1995.

(continued)

BFN-UNIT 1 B 3.3-71 Revision 0,40 October 26, 2006

Control Rod Block Instrumentation B 3.3.2.1 BASES REFERENCES 12. XN-NF-80-19(P)(A), Volume 1 "Exxon Nuclear (continued) Methodology for Boiling Water Reactors - Neutronic Methods for Design and Analysis," (as identified in the COLR).

13. EMF-2158(P)(A), "Siemens Power Corporation Methodology for Boiling Water Reactors: Evaluation and Validation of CASMO-4/MICROBURN-B2," (as identified in the COLR).

BFN-UNIT 1 B 3.3-71a Revision OQ-4.

October 26-1 2006

EOC-RPT Instrumentation B 3.3.4.1 B 3.3 INSTRUMENTATION B 3.3.4.1 End of Cycle Recirculation Pump Trip (EOC-RPT) Instrumentation BASES BACKGROUND The EOC-RPT instrumentation initiates a recirculation pump trip (RPT) to reduce the peak reactor pressure and power resulting from turbine trip or generator load rejection transients to provide additional margin to core thermal MCPR Safety Limits (SLs),

and LHGR limits.

The need for the additional negative reactivity in excess of that normally inserted on a scram reflects end of cycle reactivity considerations. Flux shapes at the end of cycle are such that the control rods may not be able to ensure that thermal limits are maintained by inserting sufficient negative reactivity during the first few feet of rod travel upon a scram caused by Turbine Control Valve (TCV) Fast Closure, Trip Oil Pressure - Low or Turbine Stop Valve (TSV) - Closure. The physical phenomenon involved is that the void reactivity feedback due to a pressurization transient can add positive reactivity at a faster rate than the control rods can add negative reactivity.

The EOC-RPT instrumentation, as shown in Reference 1, is composed of sensors that detect initiation of closure of the TSVs or fast closure of the TCVs, combined with relays, logic circuits, and fast acting circuit breakers that interrupt power from the recirculation pump variable frequency drives (VFD) to each of the recirculation pump motors. When the channels pre-established setpoint is exceeded, the channel output'relay actuates, which then outputs an EOC-RPT signal to the trip logic. When the RPT breakers trip open, the recirculation pumps coast down under their own inertia. The EOC-RPT has two identical trip systems, either of which can actuate an RPT.

(continued)

BFN-UNIT 1 B 3.3-105 Revision 0,49 AP.p30, 2*.7

EOC-RPT Instrumentation B 3.3.4.1 BASES BACKGROUND Each EOC-RPT trip system is a two-out-of-two logic for each (continued) Function; thus, either two TSV - Closure or two TCV Fast Closure, Trip Oil Pressure - Low signals are required for a trip system to actuate. If either trip system actuates, both recirculation pumps will trip. There are two EOC-RPT breakers in series per recirculation pump. One trip system trips one of the two EOC-RPT breakers for each recirculation pump, and the second trip system trips the other EOC-RPT breaker for each recirculation pump.

APPLICABLE The TSV - Closure and the TCV Fast Closure, Trip Oil SAFETY ANALYSES, Pressure - Low Functions are designed to trip the recirculation LCO, and pumps in the event of a turbine trip or generator load rejection APPLICABILITY to mitigate the increase in neutron flux, heat flux, and reactor pressure, and to increase the margin to the MCPR SL, and LHGR limits. The analytical methods and assumptions used in evaluating the turbine trip and generator load rejection are summarized in References 2, 3, and 4.

To mitigate pressurization transient effects, the EOC-RPT must trip the recirculation pumps after initiation of closure movement of either the TSVs or the TCVs. The combined effects of this trip and a scram reduce fuel bundle power more rapidly than a scram alone, resulting in an increased margin to the MCPR SL, and LHGR limits. Alternatively, MCPR limits for an inoperable EOC-RPT, as specified in the COLR, are sufficient to prevent violation of the MCPR Safety Limit, and fuel mechanical limits.

The EOC-RPT function is automatically disabled when turbine first stage pressure is < 30% RTP.

EOC-RPT instrumentation satisfies Criterion 3 of the NRC Policy Statement (Ref. 6).

(continued)

BFN-UNIT 1 B 3.3-106 Revision 0 1

EOC-RPT Instrumentation B 3.3.4.1 BASES APPLICABLE for calibration, process, and some of the instrument errors. The SAFETY ANALYSES, trip setpoints are then determined accounting for the remaining LCO, and instrument errors (e.g., drift). The trip setpoints derived in this APPLICABILITY manner provide adequate protection because instrumentation (continued) uncertainties, process effects, calibration tolerances, instrument drift, and severe environmental effects (for channels that must function in harsh environments as defined by 10 CFR 50.49) are accounted for.

The specific Applicable Safety Analysis, LCO, and Applicability discussions are listed below on a Function by Function basis.

Alternatively, since this instrumentation protects against a MCPR SL violation, with the instrumentation inoperable, modifications to the MCPR limits (LCO 3.2.2) may be applied to allow this LCO to be met. The MCPR and LHGR penalties for the EOC-RPT inoperable condition are specified in the COLR.

Turbine Stop Valve - Closure Closure of the TSVs and a main turbine trip result in the loss of a heat sink that produces reactor pressure, neutron flux, and heat flux transients that must be limited. Therefore, an RPT is initiated on TSV - Closure in anticipation of the transients that would result from closure of these valves. EOC-RPT decreases reactor power and aids the reactor scram in ensuring that the MCPR SL and LHGR limits are not exceeded during the worst case transient.

(continued)

BFN-UNIT 1 B 3.3-108 Revision 0 1

EOC-RPT Instrumentation B 3.3.4.1 BASES APPLICABLE Turbine Stop Valve - Closure (continued)

SAFETY ANALYSES, LCO, and Closure of the TSVs is determined by measuring the position of APPLICABILITY each valve. There are two separate position signals associated with each stop valve, the signal from each switch being assigned to a separate trip channel. The logic for the TSV -

Closure Function is such that two or more TSVs must be closed to produce an ,EOC-RPT. This Function must be enabled at THERMAL POWER > 30% RTP. This is normally accomplished automatically by pressure transmitters sensing turbine first stage pressure; therefore, opening the turbine bypass valves may affect this function. To consider this function OPERABLE, bypass of the function must not occur when bypass valves are open. Four channels of TSV - Closure, with two channels in each trip system, are available and required to be OPERABLE to ensure that no single instrument failure will preclude an EOC-RPT from this Function on a valid signal. The TSV -

Closure Allowable Value is selected to detect imminent TSV closure.

This protection is required, consistent with the safety analysis assumptions, whenever THERMAL POWER is > 30% RTP.

Below 30% RTP, the Reactor Vessel Steam Dome Pressure -

High and the Average Power Range Monitor (APRM) Fixed Neutron Flux - High Functions of the Reactor Protection System (RPS) are adequate to maintain the necessary margin to the MCPR SL, and LHGR limits.

(continued)

BFN-UNIT 1 B 3.3-109 Revision 0 1

EOC-RPT Instrumentation B 3.3.4.1 BASES APPLICABLE Turbine Control Valve Fast Closure, Trip Oil Pressure - Low SAFETY ANALYSES, (PS-47-142, PS-47-144, PS-47-146, and PS-47-148)

LCO, and APPLICABILITY Fast closure of the TCVs during a generator load rejection (continued) results in the loss of a heat sink that produces reactor pressure, neutron flux, and heat flux transients that must be limited.

Therefore, an RPT is initiated on TCV Fast Closure, Trip Oil Pressure - Low in anticipation of the transients that would result from the closure of these valves. The EOC-RPT decreases reactor power and aids the reactor scram in ensuring that the MCPR SL, and LHGR limits are not exceeded during the worst case transient..

Fast closure of the TCVs is determined by measuring the electrohydraulic control fluid pressure at each control valve.

There is one pressure switch associated with each control valve, and the signal from each switch is assigned to a separate trip channel. The logic for the TCV Fast Closure, Trip Oil Pressure - Low Function is such that two or more TCVs must be closed (pressure switch trips) to produce an EOC-RPT. This Function must be enabled at THERMAL POWER >_30% RTP.

This is normally accomplished automatically by pressure transmitters sensing turbine first stage pressure; therefore, opening the turbine bypass valves may affect this function. To consider this function OPERABLE, bypass of the function must not occur when bypass valves are open. Four channels of TCV Fast Closure, Trip Oil Pressure - Low, with two channels in each trip system, are available and required to be OPERABLE to ensure that no single instrument failure will preclude an EOC-RPT from this Function on a valid signal. The TCV Fast Closure, Trip Oil Pressure - Low Allowable Value is selected high enough to detect imminent TCV fast closure.

(continued)

BFN-UNIT 1 B 3.3-110 Revision@ 1

EOC-RPT Instrumentation B 3.3.4.1 BASES ACTIONS A.1 (continued)

With one or more channels inoperable, but with EOC-RPT trip capability maintained (refer to Required Actions B.1 and B.2 Bases), the EOC-RPT System is capable of performing the intended function. However, the reliability and redundancy of the EOC-RPT instrumentation is reduced such that a single failure in the remaining trip system could result in the inability of the EOC-RPT System to perform the intended function.

Therefore, only a limited time is allowed to restore compliance with the LCO. Because of the diversity of sensors available to provide trip signals, the low probability of extensive numbers of inoperabilities affecting all diverse Functions, and the low probability of an event requiring the initiation of an EOC-RPT, 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> is provided to restore the inoperable channels (Required Action A.1) or apply the EOC-RPT inoperable MCPR and LHGR limits. Alternately, the inoperable channels may be placed in trip (Required. Action A.2) since this would conservatively compensate for the inoperability, restore capability to accommodate a single failure, and allow operation to continue. As noted, placing the channel in trip with no further restrictions is not allowed if the inoperable channel is the result of an inoperable breaker, since this may not adequately compensate for the inoperable breaker (e.g., the breaker may be inoperable such that it will not open). If it is not desired to place the channel in trip (e.g., as in the case where placing the inoperable channel in trip would result in an RPT, or if the inoperable channel is the result of an inoperable breaker),

Condition C must be entered and its Required Actions taken.

(continued)

BFN-UNIT 1 B 3.3-112 Revision 0

EOC-RPT Instrumentation B 3.3.4.1 BASES ACTIONS B.1 and B.2 (continued)

Required Actions B.1 and B.2 are intended to ensure that appropriate actions are taken if multiple, inoperable, untripped channels within the same Function result in the Function not maintaining EOC-RPT trip capability. A Function is considered to be maintaining EOC-RPT trip capability when sufficient channels are OPERABLE or in trip, such that the EOC-RPT System will generate a trip signal from the given Function on a valid signal and both recirculation pumps can be tripped.

Alternately, Required Action B.2 requires the MCPR and LHGR limits for inoperable EOC-RPT, as specified in the COLR, to be applied. This also restores the margin to MCPR and LHGR limits assumed in the safety analysis.

The 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> Completion Time is sufficient time for the operator to take corrective action, and takes into account the likelihood of an event requiring actuation of the EOC-RPT instrumentation during this period. It is also consistent with the 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> Completion Time provided in LCO 3.2.2 for Required Action A.1, since this instrumentation's purpose is to preclude a MCPR or LHGR violation.

C.1 With any Required Action and associated Completion Time not met, THERMAL POWER must be reduced to < 30% RTP within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. The allowed Completion Time of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is reasonable, based on operating experience, to reduce THERMAL POWER to < 30% RTP from full power conditions in an orderly manner and without challenging plant systems.

(continued)

BFN-UNIT 1 B 3.3-113 Revision 0

Main Turbine Bypass System B 3.7.5 BASES (continued)

APPLICABLE The Main Turbine Bypass System is assumed to function during SAFETY ANALYSES abnormal operational transients (e.g., the feedwater controller failure-maximum demand event), as discussed in the FSAR, Section 14.5.1.1 (Ref. 2). Opening the bypass valves during the event mitigates the increase in reactor vessel pressure, which affects the MCPR during the event. An inoperable Main Turbine Bypass System may result in APLHGR, MCPR, and LHGR penalties.

The Main Turbine Bypass System satisfies Criterion 3 of the NRC Policy Statement (Ref. 3).

LCO The Main Turbine Bypass System is required to be OPERABLE to limit peak pressure in the main steam lines and maintain reactor pressure within acceptable limits during events that cause rapid pressurization, so that the APLHGR limits, MCPR Safety Limit, and LHGR limits are not exceeded. With the Main Turbine Bypass System inoperable, modifications to the APLHGR limits (LCO 3.2.1, "AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR)"), the MCPR limits (LCO 3.2.2, "MINIMUM CRITICAL POWER RATIO (MCPR)"),

and LHGR limits (LCO 3.2.3, "LINEAR HEAT GENERATION RATE (LHGR)") may be applied to allow this LCO to be met.

The APLHGR, MCPR, and LHGR limits for the inoperable Main Turbine Bypass System are specified in the COLR. An OPERABLE Main Turbine Bypass System requires the bypass valves to open in response to increasing main steam line pressure. This response is within the assumptions of the applicable analysis (Ref. 2).

(continued)

BFN-UNIT 1 B 3.7-33 Revision 0

Main Turbine Bypass System B 3.7.5 BASES (continued)

APPLICABILITY The Main Turbine Bypass System is required to be OPERABLE at > 25% RTP to ensure that the fuel cladding integrity Safety Limit is not violated during abnormal operational transients. As discussed in the Bases for LCO 3.2.1 and LCO 3.2.2, sufficient margin to these limits exists at < 25% RTP. Therefore, these requirements are only necessary when operating at or above this power level.

ACTIONS A.1 Ifthe Main Turbine Bypass System is inoperable (one or more bypass valves inoperable), or the APLHGR, MCPR, and LHGR limits for an inoperable Main Turbine Bypass System, as specified in the COLR, are not applied, the assumptions of the design basis transient analysis may not be met. Under such

-J circumstances, prompt action should be taken to restore the Main Turbine Bypass System to OPERABLE status or adjust the APLHGR, MCPR, and LHGR limits accordingly. The 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> Completion Time is reasonable, based on the time to complete the Required Action and the low probability of an event occurring during this period requiring the Main Turbine Bypass System.

B.1 If the Main Turbine Bypass System cannot be restored to OPERABLE status or the APLHGR, MCPR, and LHGR limits for an inoperable Main Turbine Bypass System are not applied, THERMAL POWER must be reduced to < 25% RTP. As discussed in the Applicability section, operation at < 25% RTP results in sufficient margin to the required limits, and the Main (continued)

BFN-UNIT 1 B 3.7-34 Revision 0I

Control Rod Testing - Operating B 3.10.7 B 3.10 SPECIAL OPERATIONS B 3.10.7 Control Rod Testing - Operating BASES BACKGROUND The purpose of this Special Operations LCO is to permit control rod testing, while in MODES 1 and 2, by imposing certain administrative controls. Control rod patterns during startup conditions are controlled by the operator and the rod worth minimizer (RWM) (LCO 3.3.2.1, "Control Rod Block Instrumentation"), such that only the specified control rod sequences and relative positions required by LCO 3.1.6, "Rod Pattern Control," are allowed over the operating range from all control rods inserted to the low power setpoint (LPSP) of the RWM. The sequences effectively limit the potential amount and rate of reactivity increase that could occur during a control rod drop accident (CRDA). During these conditions, control rod testing is sometimes required that may result in control rod patterns not in compliance with the prescribed sequences of LCO 3.1.6. These tests include SDM demonstrations, control rod scram time testing, and control rod friction testing. This Special Operations LCO provides the necessary exemption to the requirements of LCO 3.1.6 and provides additional administrative controls to allow the deviations in such tests from the prescribed sequences in LCO 3.1.6.

APPLICABLE The analytical methods and assumptions used in evaluating SAFETY ANALYSES the CRDA are summarized in References 1, 2, 3, and 4. CRDA analyses assume the reactor operator follows prescribed withdrawal sequences. These sequences define the potential initial conditions for the CRDA analyses. The RWM provides backup to operator control of the withdrawal sequences to ensure the initial conditions of the CRDA analyses are not (continued)

BFN-UNIT 1 B 3.10-36 Revision 0 1

Control Rod Testing - Operating B 3.10.7 BASES APPLICABLE violated. For special sequences developed for control rod SAFETY ANALYSES testing, the initial control rod patterns assumed in the safety (continued) analysis of References 1, 2, 3, and 4 may not be preserved.

Therefore, special CRDA analyses may be required to demonstrate that these special sequences will not result in unacceptable consequences, should a CRDA occur during the testing. These analyses, performed in accordance with an NRC approved methodology, are dependent on the specific test being performed.

As described in LCO 3.0.7, compliance with Special Operations LCOs is optional, and therefore, no criteria of the NRC Policy Statement apply. Special Operations LCOs provide flexibility to perform certain operations by appropriately modifying requirements of other LCOs. A discussion of the criteria satisfied for the other LCOs is provided in their respective Bases.

LCO As described in LCO 3.0.7, compliance with this Special Operations LCO is optional. Control rod testing may be performed in compliance with the prescribed sequences of LCO 3.1.6, and during these tests, no exceptions to the requirements of LCO 3.1.6 are necessary. For testing performed with a sequence not in compliance with LCO 3.1.6, the requirements of LCO 3.1.6 may be suspended, provided additional administrative controls are placed on the test to ensure that the assumptions of the special safety analysis for the test sequence are satisfied. Assurances that the test sequence is followed can be provided by either programming the test sequence into the RWM, with conformance verified as specified in SR 3.3.2.1.8 and allowing the RWM to monitor (continued)

BFN-UNIT 1 B 3.10-37 Revision 0 1

Control Rod Testing - Operating B 3.10.7 BASES LCO control rod withdrawal and provide appropriate control rod (continued) blocks if necessary, or by verifying conformance to the approved test sequence, by a second licensed operator or other qualified member of the technical staff (i.e., personnel trained in accordance with an approved training program for this test).

These controls are consistent with those normally applied to operation in the startup range as defined in the SRs and ACTIONS of LCO 3.3.2.1, "Control Rod Block Instrumentation."

APPLICABILITY Control rod testing, while in MODES 1 and 2, with THERMAL POWER greater than 10% RTP, is adequately controlled by the existing LCOs on power distribution limits and control rod block instrumentation. Control rod movement during these conditions is not restricted to prescribed sequences and can be performed within the constraints of LCO 3.2.1, "AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR)," LCO 3.2.2, "MINIMUM CRITICAL POWER RATIO (MCPR)," LCO 3.2.3, "LINEAR HEAT GENERATION RATE (LHGR)," and LCO 3.3.2.1. With THERMAL POWER less than or equal to 10% RTP, the provisions of this Special Operations LCO are necessary to perform special tests that are not in conformance with the prescribed sequences of LCO 3.1.6.

While in MODES 3 and 4, control rod withdrawal is only allowed if performed in accordance with Special Operations LCO 3.10.3,

."Single Control Rod Withdrawal - Hot Shutdown," or Special Operations LCO 3.10.4, "Single Control Rod Withdrawal - Cold Shutdown," which provide adequate controls to ensure that the assumptions of the safety analyses of References 1, 2, 3, and 4 are (continued)

BFN-UNIT 1 B 3.10-38 Revision 0 1

Control Rod Testing - Operating B 3.10.7 BASES SURVEILLANCE SR 3.10.7.2 REQUIREMENTS (continued) When the RWM provides conformance to the special test sequence, the test sequence must be verified to be correctly loaded into the RWM prior to control rod movement. This Surveillance demonstrates compliance with SR 3.3.2.1.8, thereby demonstrating that the RWM is OPERABLE. A Note has been added to indicate that this Surveillance does not need to be performed if SR 3.10.7.1 is satisfied.

REFERENCES 1. NEDE-2401 1-P-A-1 3, "General Electric Standard Application for Reactor Fuel," August 1996.

2. Letter from T. Pickens (BWROG) to G. C. Lainas (NRC)

"Amendment 17 to General Electric Licensing Topical Report NEDE-24011-P-A," August 15, 1986.

3. XN-NF-80-19(P)(A), Volume 1 "Exxon Nuclear Methodology for Boiling Water Reactors - Neutronic Methods for Design and Analysis," (as identified in the COLR).

4. EMF-2158(P)(A), "Siemens Power Corporation Methodology for Boiling Water Reactors: Evaluation and Validation of CASMO-4/MICROBURN-B2," (as identified in the COLR).

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SDM Test -Refueling B 3.10.8 BASES (continued)

APPLICABLE Prevention and mitigation of unacceptable reactivity excursions SAFETY ANALYSES during control rod withdrawal, with the reactor mode switch in the startup/hot standby position while in MODE 5, is provided by the intermediate range monitor (IRM) neutron flux scram (LCO 3.3.1.1, "Reactor Protection System (RPS)

Instrumentation"), and control rod block instrumentation (LCO 3.3.2.1, "Control Rod Block Instrumentation"). The limiting reactivity excursion during startup conditions while in MODE 5 is the control rod drop accident (CRDA).

CRDA analyses assume that the reactor operator follows prescribed withdrawal sequences. For SDM tests performed within these defined sequences, the analyses of References 1, 2, 3, and 4 are applicable. However, for some sequences developed for the SDM testing, the control rod patterns assumed in the safety analyses of References 1, 2, 3, and 4 may not be met. Therefore, special CRDA analyses, performed in accordance with an NRC approved methodology, may be required to demonstrate the SDM test sequence will not result in unacceptable consequences should a CRDA occur during the testing. For the purpose of this test, the protection provided by the normally required MODE 5 applicable LCOs, in addition to the requirements of this LCO, will maintain normal test operations as well as postulated accidents within the bounds of the appropriate safety analyses (Refs. 1, 2, 3, and 4). In addition to the added requirements for the RWM, APRM, and control rod coupling, the notch out mode is specified for out of sequence withdrawals. Requiring the notch out mode limits withdrawal steps to a single notch, which limits inserted reactivity, and allows adequate monitoring of changes in neutron flux, which may occur during the test.

(continued)

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SDM Test -Refueling B 3.10.8 BASES SURVEILLANCE SR 3.10.8.5 REQUIREMENTS (continued) Coupling verification is performed to ensure the control rod is connected to the control rod drive mechanism and will perform its intended function when necessary. The verification is required to be performed any time a control rod is withdrawn to the "full out" notch position, or prior to declaring the control rod OPERABLE after work on the control rod or CRD System that could affect coupling. This Frequency is acceptable, considering the low probability that a control rod will become uncoupled when it is not being moved as well as operating experience related to uncoupling events.

SR 3.10.8.6 CRD charging water header pressure verification is performed to ensure the motive force is available to scram the control rods in the event of a scram signal. Since the reactor is depressurized in MODE 5, there is insufficient reactor pressure to scram the control rods. Verification of charging water pressure ensures that if a scram is required, capability for rapid control rod insertion would exist. The minimum pressure of 940 psig, which is well below the expected pressure of approximately 1100 psig, ensures sufficient pressure for rapid control rod insertion. The 7 day Frequency has been shown to be acceptable through operating experience and takes into account indications available in the control room.

REFERENCES 1. NEDE-2401 1-P-A-1 3, "General Electric Standard Application for Reactor Fuel," August 1996.

2. Letter from T. Pickens (BWROG) to G. C. Lainas, NRC, "Amendment 17 to General Electric Licensing Topical Report NEDE-24011-P-A," August 15, 1986.

(continued)

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SDM Test -Refueling B 3.10.8 BASES REFERENCES 3. XN-NF-80-19(P)(A), Volume 1 "Exxon Nuclear (continued) Methodology for Boiling Water Reactors - Neutronic Methods for Design and Analysis," (as identified in the COLR).

4. EMF-2158(P)(A), "Siemens Power Corporation Methodology for Boiling Water Reactors: Evaluation and Validation of CASMO-4/MICROBURN-B2," (as identified in the COLR).

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