Text

Issuance of Amendment Adoption of TSTF-475, Revision 1, Control Rod Notch Testing Frequency and SRM Insert Control Rod Action Using Consolidated Line Item Improvement Process
	ML083520395
Person / Time
Site:	Susquehanna
Issue date:	01/02/2009
From:	Bhalchandra Vaidya; Plant Licensing Branch 1
To:	Spence W; PPL Corp
	vaidya b k
References
	TAC MD9303, TAC MD9304 TSTF-475, Rev 1
	Download: ML083520395 (27)
	v • d • e

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 January 2, 2009 Mr. William H. Spence Executive Vice President Chief Operating Officer/Chief Nuclear Officer PPL Corporation Two North Ninth Street, GENTW16 Allentown, PA 18101-1179

SUBJECT:

SUSQUEHANNA STEAM ELECTRIC STATION, UNITS 1 AND 2 ISSUANCE OF AMENDMENT RE: ADOPTION OF TSTF-475, REVISION 1, "CONTROL ROD NOTCH TESTING FREQUENCY AND SRM INSERT CONTROL ROD ACTION" USING CONSOLIDATED LINE ITEM IMPROVEMENT PROCESS (TAC NOS. MD9303 AND MD9304)

Dear Mr. Spence:

The Commission has issued the enclosed Amendment No. 25eto Facility Operating License No. NPF-14 and Amendment No229 to Facility Operating License No. NPF-22 for the Susquehanna Steam Electric Station (SSES), Units 1 and 2. These amendments consist of changes to the Technical Specifications (TSs) in response to your application dated July 7, 2008.

The proposed amendment adopts Technical Specification Task Force (TSTF)-475, Revision 1 "Control Rod Notch Testing Frequency and SRM Insert Control Rod Action." The proposed amendment would: (1) delete TS Surveillance Requirement (SR) 3.1.3.2 (specifies a 7-day test frequency for fully withdrawn rods) in TS 3.1.3, "Control Rod OPERABILITY," (2) revise SR 3.1.3.3 to make its 31-day test frequency applicable not only to partially withdrawn rods but also to fully withdrawn rods, (3) revise Example 1.4-3 in Section 1.4 "Frequency" to clarify the applicability of the 1.25 surveillance test interval extension, and (4) some editorial changes pertaining to renumbering of the TS SRs. This operating license improvement was made available by the U.S. Nuclear Regulatory Commission on November 13,2007 (72 FR 63935) as part of the consolidated line item improvement process (CLlIP).

W. H. Spence

-2 A copy of our safety evaluation is also enclosed. The Notice of Issuance will be included in the Commission's next regular Biweekly Federal Register Notice.

Sincerely, Bhalchandra K. Vaidya, Project Manager Plant Licensing Branch 1-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-387 and 50-388

Enclosures:

1. Amendment NO.250 to License No. NPF-14

2. Amendment No.229 to License No. NPF-22

3. Safety Evaluation cc w/encls: Distribution via ListServe

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 PPL SUSQUEHANNA, LLC ALLEGHENY ELECTRIC COOPERATIVE, INC.

DOCKET NO. 50-387 SUSQUEHANNA STEAM ELECTRIC STATION, UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment NcQ50 License No. NPF-14

1.

The Nuclear Regulatory Commission (the Commission or the NRC) having found that:

A.

The application for the amendment filed by PPL Susquehanna, LLC, dated July 7,2008, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's regulations set forth in 10 CFR Chapter I; B.

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

There is reasonable assurance: (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations set forth in 10 CFR Chapter I; D.

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

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

- 2

2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment and paragraph 2.C.(2) of the Facility Operating License No. NPF-14 is hereby amended to read as follows:

(2)

Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 250 and the Environmental Protection Plan contained in Appendix B, are hereby incorporated in the license. PPL Susquehanna, LLC shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

3. This license amendment is effective as of its date of issuance and shall be implemented within 60 days.

FOR THE NUCLEAR REGULATORY COMMISSION

~vr~

~

Mark G. Kowal, Chief Plant Licensing Branch 1-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the License and Technical Specifications Date of Issuance: January 2, 2009

ATTACHMENT TO LICENSE AMENDMENT NO. 250 FACILITY OPERATING LICENSE NO. NPF-14 DOCKET NO. 50-387 Replace the following page of the Facility Operating License with the attached revised page.

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

REMOVE INSERT Page 3 Page 3 Replace the following pages of the Appendix A Technlcal Specifications with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

REMOVE INSERT 1.4-4 1.4-4 1.4-5 1.4-5 3.1-8 3.1-8 3.1-10 3.1-10

-3 (4)

PPL Susquehanna, LLC, pursuant to the Act and 10 CFR Parts 30, 40, and 70, to receive, possess, and use in amounts as required any byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; and (5)

PPL Susquehanna, LLC, pursuant to the Act and 10 CFR Parts 30, 40, and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility.

C. This license shall be deemed to contain and is subject to the conditions specified in the Commission's regulations set forth in 10 CFR Chapter I and is subject to all applicable provisions of the Act and to the rules, regulations and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:

(1)

Maximum Power Level PPL Susquehanna, LLC is authorized to operate the facility at reactor core power levels not in excess of 3952 megawatts thermal in accordance with the conditions specified herein. The preoperational tests, startup tests and other items identified in License Conditions 2.C.(36), 2.C.(37), 2.C.(38), and 2.C.(39) to this license shall be completed as specified.

(2)

Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 250 and the Environmental Protection Plan contained in Appendix B are hereby incorporated in the license. PPL Susquehanna, LLC shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

For Surveillance Requirements (SRs) that are new in Amendment 178 to Facility Operating License No. NPF-14, the first performance is due at the end of the first surveillance interval that begins at implementation of Amendment 178. For SRs that existed prior to Amendment 178, including SRs with modified acceptance criteria and SRs whose frequency of performance is being extended, the first performance is due at the end of the first surveillance interval that begins on the date the Surveillance was last performed prior to implementation of Amendment 178.

(3)

Conduct of Work Activities During Fuel Load and Initial Startup The operating licensee shall review by committee all facility construction, Preoperational Testing, and System Demonstration activities performed concurrently with facility initial fuel loading or with the facility Startup Test Amendment No. a, 44J, +78, ~, +88, -UM, ~~, ~, 2J8, ~, ~, ~, ~, 24J, ~, ~, ~,

U7,~,249,250

1.4 PPL Rev.

Frequency 1.4 Frequency EXAMPLES EXAMPLE 1.4-2 (continued)

"Thereafter" indicates future performances must be established per SR 3.0.2, but only after a specified condition is first met (i.e., the "once" performance in this example). If reactor power decreases to <25% RTP, the measurement of both intervals stops. New intervals start upon reactor power reaching 25% RTP.

EXAMPLE 1.4-3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY

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

Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after >25% RTP.

Perform channel adjustment.

7 days The interval continues whether or not the unit operation is <25% RTP between performances.

As the Note modifies the required performance of the Surveillance, it is construed to be part of the "specified Frequency." Should the 7 day interval be exceeded while operation is <25% RTP, this Note allows 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after power reaches :::: 25% RTP to perform the Surveillance. The Surveillance is still considered to be within the "specified Frequency." Therefore, if the Surveillance were not performed within the 7 day interval (plus the extension allowed by SR 3.0.2), but operation was <25% RTP, it would not constitute a failure of the SR or failure to meet the LCO. Also, no violation of SR 3.0.4 occurs when changing MODES, even with the 7 day Frequency not met, provided operation does not exceed 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months (plus the extension allowed by SR 3.0.2) with power >25% RTP.

(continued)

SUSQUEHANNA - UNIT 1 TS /1.4-4 Amendment 4+8,250

1.4 PPL Rev.

Frequency 1.4 Frequency EXAMPLES EXAMPLE 1.4-3 (continued)

Once the unit reaches 25% RTP, 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months would be allowed for completing the Surveillance. If the Surveillance were not performed within this 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months interval (plus the extension allowed by SR 3.0.2), there would then be a failure to perform a Surveillance within the specified Frequency, and the provisions of SR 3.0.3 would apply.

EXAMPLE 1.4-4 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY

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

Only required to be met in MODE 1.

Verify leakage rates are within limits.

24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Example 1.4-4 specifies that the requirements of this Surveillance do not have to be met until the unit is in MODE 1. The interval measurement for the Frequency of this Surveillance continues at all times, as described in Example 1.4-1. However, the Note constitutes an "otherwise stated" exception to the Applicability of this Surveillance. Therefore, if the Surveillance were not performed within the 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months interval (plus the extension allowed by SR 3.0.2), but the unit was not in MODE 1, there would be no failure of the SR nor failure to meet the LCO. Therefore, no violation of SR 3.0.4 occurs when changing MODES, even with the 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Frequency exceeded, provided the MODE change was not made into MODE 1. Prior to entering MODE 1 (assuming again that the 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Frequency were not met), SR 3.0.4 would require satisfying the SR.

SUSQUEHANNA - UNIT 1 TS /1.4-5 Amendment ~, 250

3.1.3 PPL Rev.

Control Rod OPERABILITY ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A,

(continued)

A,3 Perform SR 3.1.3.3 for each withdrawn OPERABLE control rod.

AND A,4 Perform SR 3.1.1.1.

24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months from discovery of Condition A concurrent with THERMAL POWER greater than the low power setpoint (LPSP) of the RWM.

72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months B.

Two or more withdrawn control rods stuck.

B.1 Be in MODE 3.

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months C.

One or more control rods inoperable for reasons other than Condition A or B.

C.1 ----------------NO-rE---------------

RWM may be bypassed as allowed by LCO 3.3.2.1, if required, to allow insertion of inoperable control rod and continued operation.

FUlly insert inoperable control rod.

AND 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months (continued)

SUSQUEHANNA - UNIT 1 TS / 3.1-8 Amendment ++8250

3.1.3 PPL Rev.

Control Rod OPERABILITY SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.3.1 Determine the position of each control rod.

24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months SR 3.1.3.2 NOT USED SR 3.1.3.3

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

Not required to be performed until 31 days after the control rod is withdrawn and THERMAL POWER is greater than the LPSP of the RWM.

Insert each withdrawn control rod at least one notch.

31 days SR 3.1.3.4 Verify each control rod scram time from fully withdrawn to notch position 05 is < 7 seconds.

In accordance with SR 3.1.4.1, SR 3.1.4.2, SR 3.1.4.3, and SR 3.1.4.4 (continued)

SUSQUEHANNA - UNIT 1 TS/3.1-10 Amendment -++S, 250

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 PPLSUSQUEHANNA,LLC ALLEGHENY ELECTRIC COOPERATIVE, INC.

DOCKET NO. 50-388 SUSQUEHANNA STEAM ELECTRIC STATION, UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No229 License No. NPF-22

1.

The Nuclear Regulatory Commission (the Commission or the NRC) having found that:

A.

The application for the amendment filed by PPL Susquehanna, LLC, dated July 7, 2008, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's regulations set forth in 10 CFR Chapter I; B.

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

There is reasonable assurance: (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations set forth in 10 CFR Chapter I; D.

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

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

- 2

2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment and paragraph 2.C.(2) of the Facility Operating License No. NPF-14 is hereby amended to read as follows:

(2)

Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 229 and the Environmental Protection Plan contained in Appendix B, are hereby incorporated in the license. PPL Susquehanna, LLC shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

3. This license amendment is effective as of its date of issuance and shall be implemented within 60 days.

FOR THE NUCLEAR REGULATORY COMMISSION Mark G. Kowal, Chief Plant Licensing Branch 1-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the License and Technical Specifications Date of Issuance: January 2, 2009

ATTACHMENT TO LICENSE AMENDMENT NO.229 FACILITY OPERATING LICENSE NO. NPF-22 DOCKET NO. 50-388 Replace the following page of the Facility Operating License with the attached revised page.

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

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

REMOVE INSERT 1.4-4 1.4-4 1.4-5 1.4-5 3.1-8 3.1-8 3.1-10 3.1-10

-3 (4)

PPL Susquehanna, LLC, pursuant to the Act and 10 CFR Parts 30, 40, and 70, to receive, possess, and use in amounts as required any byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; and (5)

PPL Susquehanna, LLC, pursuant to the Act and 10 CFR Parts 30, 40, and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility.

C. This license shall be deemed to contain and is subject to the conditions specified in the Commission's regulations set forth in 10 CFR Chapter I and is subject to all applicable provisions of the Act and to the rules, regulations and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:

(1)

Maximum Power Level PPL Susquehanna, LLC is authorized to operate the facility at reactor core power levels not in excess of 3952 megawatts thermal in accordance with the conditions specified herein. The preoperational test, startup tests and other items identified in License Conditions 2.C.(20), 2.C.(21), 2.C.(22), and 2.C.(23) to this license shall be completed as specified.

(2)

Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No.229 and the Environmental Protection Plan contained in Appendix S, are hereby incorporated in the license. PPL Susquehanna, LLC shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

For Surveillance Requirements (SRs) that are new in Amendment 151 to Facility Operating License No. NPF-22, the first performance is due at the end of the first surveillance interval that begins at implementation of Amendment 151. For SRs that existed prior to Amendment 151, including SRs with modified acceptance criteria and SRs whose frequency of performance is being extended, the first performance is due at the end of the first surveillance interval that begins on the date the Surveillance was last performed prior to implementation of Amendment 151..

2.C.(3)

PPL Susquehanna, LLC shall implement and maintain in effect all provisions of the approved fire protection program as described in the Fire Protection Review Report for the facility and as approved in Fire Protection Program, Section 9.5, SER, SSER#1, SSER#2, SSER#3, SSER#4, SSER#6, Safety Evaluation of Fire Protection dated August 9, 1989, Safety Evaluation Amendment No. +,.a, +00, +W, +e-+, ~, ~, +W, ~, ~,.a+4, zrs, ~,.2-+7, ~,.a+9, sao, ~,~, ~, ~, aas, ~, 2Zl-, ~, 229

1.4 PPL Rev.

Frequency 1.4 Frequency EXAMPLES EXAMPLE 1.4-2 (continued)

"Thereafter" indicates future performances must be established per SR 3.0.2, but only after a specified condition is first met (Le., the "once" performance in this example). If reactor power decreases to < 25% RTP, the measurement of both intervals stops. New intervals start upon reactor power reaching 25% RTP.

EXAMPLE 1.4-3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY

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

Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after ~ 25% RTP.

7 days Perform channel adjustment.

The interval continues whether or not the unit operation is < 25% RTP between performances.

As the Note modifies the required performance of the Surveillance, it is construed to be part of the "specified Frequency." Should the 7 day interval be exceeded while operation is < 25% RTP, this Note allows 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after power reaches ~ 25% RTP to perform the Surveillance. The Surveillance is still considered to be within the "specified Frequency." Therefore, if the Surveillance were not performed within the 7 day interval (plus the extension allowed by SR 3.0.2), but operation was < 25% RTP, it would not constitute a failure of the SR or failure to meet the LCO. Also, no violation of SR 3.0.4 occurs when changing MODES, even with the 7 day Frequency not met, provided operation does not exceed 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months (plus the extension allowed by SR 3.0.2) with power ~ 25% RTP.

(continued)

SUSQUEHANNA - UNIT 2 TS / 1.4-4 Amendment -+M,229

1.4 PPL Rev.

Frequency 1.4 Frequency EXAMPLES EXAMPLE 1.4-3 (continued)

Once the unit reaches 25% RTP, 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months would be allowed for completing the Surveillance. If the Surveillance were not performed within this 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months interval (plus the extension allowed by SR 3.0.2), there would then be a failure to perform a Surveillance within the specified Frequency, and the provisions of SR 3.0.3 would apply.

EXAMPLE 1.4-4 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY

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

Only required to be met in MODE 1.

Verify leakage rates are within limits.

24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Example 1.4-4 specifies that the requirements of this Surveillance do not have to be met until the unit is in MODE 1. The interval measurement for the Frequency of this Surveillance continues at all times, as described in Example 1.4-1. However, the Note constitutes an "otherwise stated" exception to the Applicability of this Surveillance. Therefore, if the Surveillance were not performed within the 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months interval (plus the extension allowed by SR 3.0.2), but the unit was not in MODE 1, there would be no failure of the SR nor failure to meet the LCO. Therefore, no violation of SR 3.0.4 occurs when changing MODES, even with the 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Frequency exceeded, provided the MODE change was not made into MODE 1. Prior to entering MODE 1 (assuming again that the 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Frequency were not met), SR 3.0.4 would require satisfying the SR.

SUSQUEHANNA - UNIT 2 TS / 1.4-5 Amendment 4-a+,229

3.1.3 PPL Rev.

Control Rod OPERABILITY ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

(continued)

A.3 Perform SR 3.1.3.3 for each withdrawn OPERABLE control rod.

AND A.4 Perform SR 3.1.1.1.

24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months from discovery of Condition A concurrent with THERMAL POWER greater than the low power setpoint (LPSP) of the RWM.

72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months B.

Two or more withdrawn control rods stuck.

B.1 Be in MODE 3.

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months C.

One or more control rods inoperable for reasons other than Condition A or B.

C.1

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

RWM may be bypassed as allowed by LCO 3.3.2.1, if required, to allow insertion of inoperable control rod and continued operation.

Fully insert inoperable control rod.

AND 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months (continued)

SUSQUEHANNA - UNIT 2 TS / 3.1-8 Amendment +a-+,229

3.1.3 PPL Rev.

Control Rod OPERABILITY SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.3.1 Determine the position of each control rod.

24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months SR 3.1.3.2 NOT USED SR 3.1.3.3

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

Not required to be performed until 31 days after the control rod is withdrawn and THERMAL POWER is greater than the LPSP of the RWM.

Insert each withdrawn control rod at least one notch.

31 days SR 3.1.3.4 Verify each control rod scram time from fully withdrawn to notch position 05 is < 7 seconds.

In accordance with SR 3.1.4.1, SR 3.1.4.2, SR 3.1.4.3, and SR 3.1.4.4 (continued)

SUSQUEHANNA - UNIT 2 TS/3.1-10 Amendment -ta-+, 229

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 250 TO FACILITY OPERATING LICENSE NO. NPF-14 AND AMENDMENT NO. 229 TO FACILITY OPERATING LICENSE NO. NPF-22 PPL SUSQUEHANNA. LLC ALLEGHENY ELECTRIC COOPERATIVE, INC.

SUSQUEHANNA STEAM ELECTRIC STATION, UNITS 1 AND 2 DOCKET NOS. 50-387 AND 388

1.0 INTRODUCTION

By application dated July 7,2008, (Agencywide Documents Access and Management system (ADAMS) Accession No. ML082040624), PPL Susquehanna, LLC (the licensee),

requested changes to the Technical Specifications (TSs) for Susquehanna Steam Electric Station, Units 1 and 2 (SSES-1 and 2).

The proposed amendment adopts Technical Specification Task Force (TSTF)-475, Revision 1 "Control Rod Notch Testing Frequency and SRM Insert Control Rod Action." The proposed amendment would: (1) delete TS Surveillance Requirement (SR) 3.1.3.2 (specifies a 7-day test frequency for fully withdrawn rods) in TS 3.1.3, "Control Rod OPERABILITY," (2) revising the frequency of notch testing of fully withdrawn control rods, in SR 3.1.3.2 from "7 days after the control rod is withdrawn and THERMAL POWER is greater than the Low Power Set Point (LPSP) of the Rod Worth Minimizer (RWM)" to "31 days after the control rod is withdrawn and THERMAL POWER is greater than the LPSP of the RWM," by changing the scope of SR 3.1.3.3 to include all withdrawn rods (i.e., both partially and fully withdrawn rods) and (3) revise Example 1.4-3 in Section 1.4 "Frequency" to clarify that the 1.25 surveillance test interval extension in SR 3.0.2 is applicable to time periods discussed in NOTES in the "SURVEILLANCE" column in addition to the time periods in the "FREQUENCY" column.

The licensee's application further states:

PPL is not proposing variations or deviations from the TS changes described in TSTF-475, Revision 1 and the NRC staff's model safety evaluation dated November 13, 2007 (72 FR 63935) as part of the CLlIP [Consolidated Line Item Improvement process] Notice of Availability." However, some editorial changes are proposed and are described below:

The renumbering of the TS SRs has not been incorporated as these editorial changes would create conflicts with other documented references, if adopted.

The existing SR numbering is maintained by identifying SR 3.1.3.2 as "not used."

- 2 The U.S. Nuclear Regulatory Commission (NRC) staff issued a notice of opportunity for comment in the Federal Register on August 16, 2007 (72 FR 46103), on possible amendments to revise the plant specific TS, to allow: (1) revising the frequency of SR 3.1.3.2, notch testing of fully withdrawn control rod, from "7 days after the control rod is withdrawn and THERMAL POWER is greater than the LPSP of RWM" to "31 days after the control rod is withdrawn and THERMAL POWER is greater than the LPSP of the RWM", (2) adding the word "fully" to limiting condition for operation (LCO) 3.3.1.2 Required Action E.2 to clarify the requirement to fully insert all insertable control rods in core cells containing one or more fuel assemblies when the associated Source Range Monitor (SRM) instrument is inoperable, and (3) revising Example 1.4-3 in Section 1.4 "Frequency" to clarify that the 1.25 surveillance test interval extension in SR 3.0.2 is applicable to time periods discussed in NOTES in the "SURVEILLANCE" column in addition to the time periods in the "FREQUENCY" column., including a model safety evaluation and model No Significant Hazards Consideration Determination (NSHCD), using the CLlIP. The NRC staff subsequently issued a notice of availability of the models for referencing in license amendment applications in the Federal Register on November 13, 2007, (72 FR 63935).

The licensee affirmed the applicability of the model NSHCD in its application dated July 7, 2008.

1.1 Background

By letter dated August 30, 2004, the TSTF submitted a request (Reference 1) for changes to the Standard Technical Specifications (STS): NUREG-1430 STS B&W Plants (Reference 2);

NUREG-1431 STS Westinghouse Plants (Reference 3); NUREG-1432 STS Combustion Engineering Plants (Reference 4); NUREG-1433, STS General Electric Plants, BWR/4 (Reference 5); and NUREG-1434, STS General Electric Plants, BWR/6 (Reference 6). The proposed changes would: (1) revise the TS control rod notch surveillance frequency in TS 3.1.3, "Control Rod OPERABILITY," (NUREG-1433 and NUREG-1434), (2) clarify the TS requirement for inserting control rods for one or more inoperable SRMs in MODE 5 (NUREG 1434 only), and (3) revise one Example in Section 1.4 "Frequency" to clarify the applicability of the 1.25 surveillance test interval extension (NUREG-1430 through NUREG-1434).

These changes are based on the NRC approved TSTF change traveler TSTF-475, Revision 1, that revised the reference STS by: (1) revising the frequency of SR 3.1.3.2, notch testing of each fully withdrawn control rod, from 7 days after the control rod is withdrawn and THERMAL POWER is greater than the Low Power Setpoint (LPSP) of the Rod Worth Minimizer (RWM) to "31 days after the control rod is withdrawn and THERMAL POWER is greater than the LPSP of the RWM" (NUREG-1433 and NUREG-1434) and (2) revising Example 1.4-3 in Section 1.4 "Frequency" to clarify that the 1.25 surveillance test interval extension in SR 3.0.2 is applicable to time periods discussed in NOTES in the "SURVEILLANCE" column in addition to the time periods in the "FREQUENCY" column (NUREG-1430 through NUREG-1434).

The purpose of the surveillances is to confirm control rod insertion capability which is demonstrated by inserting each partially or fully withdrawn control rod at least one notch and observing that the control rod moves. Control rods and the control rod drive (CRD)

Mechanism (CRDM), by which the control rods are moved, are components of the CRD System (CRDS), which is the primary reactivity control system for the reactor. By design, the

- 3 CRDM is highly reliable with a tapered design of the index tube which is conducive to control rod insertion.

A stuck control rod is an extremely rare event and industry review of plant operating experience did not identify any incidents of stuck control rods while performing a rod notch surveillance test.

The purpose of these revisions is to reduce the number of control rod manipulations and, thereby, reduce the opportunity for reactivity control events.

The purpose of the change to Example 1.4-3 in Section 1.4 "Frequency" is to clarify the applicability of the 25% allowance of SR 3.0.2 to time periods discussed in NOTES in the "SURVEILLANCE" column as well as to time periods in the "FREQUENCY" column.

2.0 REGULATORY EVALUATION

Title 10 of the Code of Federal Regulations (CFR), Part 50, Appendix A, General Design Criterion (GDC) 29, Protection against anticipated occurrence, requires that the protection and reactivity control systems be designed to assure an extremely high probability of accomplishing their safety functions in an event of anticipated operational occurrences. The design relies on the CRDS to function in conjunction with the protection systems under anticipated operational occurrences, including loss of power to all recirculation pumps, tripping of the turbine generator, isolation of the main condenser, and loss of all offsite power. The CRDS provides an adequate means of inserting sufficient negative reactivity to shut down the reactor and prevent exceeding acceptable fuel design limits during anticipated operational occurrences. Meeting the requirements of GDC 29 for the CRDS prevents occurrence of mechanisms that could result in fuel cladding damage such as severe overheating, excessive cladding strain, or exceeding the thermal margin limits during anticipated operational occurrences. Preventing excessive cladding damage in the event of anticipated transients ensures maintenance of the integrity of the cladding as a fission product barrier.

3.0 TECHNICAL EVALUATION

The NRC staff previously reviewed the following information provided by the TSTF to support the staff's review and approval of TSTF-475, Revision 1. Specifically, the following documents were reviewed during the NRC staff's evaluation:

a TSTF letter TSTF-04-07 (Reference 1) - Provided a description of the proposed changes in TSTF-475 that changes the weekly rod notch frequency to monthly and clarify the applicability of the 25% allowance in Example 1.4-3.

a TSTF letter TSTF-06-13 (Reference 8) - Provided responses to NRC staff request for additional information (RAI) on (1) industry experience with identifying stuck rods, (2) tests that would identify stuck rods, (3) continue compliance with General Electric (GE) Services Information Letter (SIL) 139, (4) industry experience on collet failures, and (4) applying the 25% grace period to the 31 day control rod notch SR test frequency.

-4 a

Boiling Water Reactor Owners Group (BWROG) letter BWROG-06036 (Reference 9) - Provided the GE Nuclear Energy Report, "CRD Notching a Surveillance Testing for Limerick Generating Station," in which CRD notching frequency and CRD performance were evaluated.

a TSTF letter TSTF-07-19 (Reference 10) - Provided response to NRC staff RAI on CRD performance in Control Cell Core (CCC) designed plants, including TSTF-475, Revision 1.

The CRD System at Susquehanna Steam Electric Station Units 1 and 2, is the primary reactivity control system for the reactor. The CRD System, in conjunction with the Reactor Protection System, provides the means for the reliable control of reactivity changes to ensure under all conditions of normal operation, including anticipated operational occurrences that specified acceptable fuel design limits are not exceeded. Control rods are components of the CRD System that have the capability to hold the reactor core subcritical under all conditions and to limit the potential amount and rate of reactivity increase caused by a malfunction in the CRD System.

The CRD System consists of a CRDM, by which the control rods are moved, and a hydraulic control unit (HCU) for each control rod. The CRDM is a mechanical hydraulic latching cylinder that positions the control blades. The CRDM is a highly reliable mechanism for inserting a control rod to the full-in position. The collet piston mechanism design feature ensures that the control rod will not be inadvertently withdrawn. This is accomplished by engaging the collet fingers, mounted on the collet piston, in notches located on the index tube. Due to the tapered design of the index tube notches, the collet piston mechanism will not impede rod insertion under normal insertion or scram conditions.

The collet retainer tube (CRT) is a short tube welded to the upper end of the CRD which houses the collet mechanism which consist of the locking collet, collet piston, collet return spring and an unlocking cam. The collet mechanism provides the locking/unlocking mechanism that allows the insert/withdraw movement of the control rod. The CRT has three primary functions: (a) to carry the hydraulic unlocking pressure to the collet piston, (b) to provide an outer cylinder, with a suitable wear surface for the metal collet piston rings, and (c) to provide mechanical support for the guide cap, a component which incorporates the cam surface for holding the collet fingers open and also provides the upper rod guide or bushing.

The NRC staff approved TSTF-475 which revised the TS SR 3.1.3.2, "Control Rod OPERABILITY" in the STS (NUREG-1433 and NUREG-1434) from 7 days to monthly based on the following: (1) slow crack growth rate of the CRT; (2) the improved CRT design; (3) a higher reliable method (scram time testing) to monitor CRD scram system functionality; (4)

GE chemistry recommendations; and (5) no known CRD failures have been detected during the notch testing exercise, the NRC staff concluded that the changes would reduce the number of control rod manipulations thereby reducing the opportunity for potential reactivity events while having a very minimal impact on the extremely high reliability of the CRD system.

- 5 The following paragraphs describe the bases for the NRC staff's approval of TSTF-475:

According to the BWROG, at the time of the first CRT crack discovery in 1975, each partially or fully withdrawn operable control rod was required to be exercised one notch at least once each week. It was recognized that notch testing provided a method to demonstrate the integrity of the CRT. Control rod insertion capability was demonstrated by inserting each partially or fully withdrawn control rod at least one notch and observing that the control rod moves. The control rod may then be returned to its original position. This ensures the control rod is not stuck and is free to insert on a scram signal.

It was determined that during scrams, the CRT temperature distribution changes substantially at reactor operating conditions. Relatively cold water moves upward through the inside of the CRT and exits via the flow holes into the annulus on the outside. At the same time, hot water from the reactor vessel flows downward on the outside surface of the CRT. There is very little mixing of the cold water flowing from the three flow holes into the annulus and the hot water flowing downward. Thus, there are substantial through wall and circumferential temperature gradients during scrams which contribute to the observed CRT cracking.

Subsequently, many BWRs have reduced the frequency of notch testing for partially withdrawn control rods from weekly to monthly. The notch test frequency for fully withdrawn control rods are still performed weekly. The change for partially withdrawn control rods was made because of the potential power reduction required to allow control rod movement for partially withdrawn control rods, the desire to coordinate scheduling with other plant activities, and the fact that a large sample of control rods are still notch tested on a weekly basis. The operating experience related to the changes in CRD performance also provided additional justification to reduce the notch test frequency for the partially withdrawn control rods.

In response to NRC's requests for additional information (RAls) and to support their position to reduce the CRD notch testing frequency, the BWROG provided plant data and a GE Nuclear Energy report entitled, "CRD Notching Surveillance Testing for Limerick Generating Station" (CRDNST). The GE report provided a description of the cracks noted on the original design CRT surfaces. These cracks, which were later determined to be intergranular, were generally circumferential, and appeared with greatest frequency below and between the cooling water ports, in the area of the change in wall thickness. Subsequently, cracks associated with residual stresses were also observed in the vicinity of the attachment weld.

Continued circumferential cracking could lead to 360 degree severance of the CRT that would render the CRD inoperable which would prevent insertion, withdrawal or scram. Such failure would be detectable in any fully or partially withdrawn control rod during the surveillance notch testing required by the TSs. To a lesser degree, cracks have also been noted at the welded joint of the interim design CRT but no cracks have been observed in the final improved CRT design. Neither the BWROG nor the NRC staff were able to find evidence of a collet housing failure since 1975. To date, operating experience data shows no reports of a severed CRT at any BWR. No collet housing failures have been noted since 1975. On a numerical basis for instance, based on BWROG assumption that there are 137 control rods for a typical BWR/4 and 193 control rods for a typical BWR/6, the yearly performance would be 6590 rod notch tests for a BWRl4 plant and 9284 for a BWR/6 plant.

For example, if all BWRs operating in the U.S. are taken into consideration, the yearly

- 6 performances of rod notch data would translate into approximately 240,000 rod notch tests without detecting a failure.

In addition, the intergranular stress-corrosion cracking (IGSCC) crack growth rates were evaluated, at Limerick Generating Station, using GE's PLEDGE model with the assumption that the water chemistry condition is based on GE recommendations. The model is based on fundamental principles of stress-corrosion cracking which can evaluate crack growth rates as a function of water oxygen level, conductivity, material sensitization and applied loads. It was determined that the additional time of 24 days represented an additional 10 mils of growth in total crack length. The small difference in growth rate would have little effect on the behavior between one notch test and the next subsequent test. Therefore, from the materials perspective based on low crack growth rates, a decrease in the notch test frequency would not affect the reliability of detecting a CRDM failure due to crack growth.

Also, the BWR scram system has extremely high reliability. In addition to notch testing, scram time testing can identify failure of individual CRD operation resulting from IGSCC initiated cracks and mechanical binding. Unlike the CRD notch tests, these single rod scram tests cover the other mechanical components such as scram pilot solenoid operated valves, the scram inlet and outlet air operated valves, and the scram accumulator, as well as operation of the control rods. Thus, the primary assurance of scram system reliability is provided by the scram time testing since it monitors the system scram operation and the complete travel of the control rod.

Also, the HCUs, CRD drives, and control rods are tested during refueling outages, approximately every 18-24 months. Based on the data collected during the preceding cycle of operation, selected control rod drives, are inspected and, as required, their internal components are replaced. Therefore, increasing the CRD notch testing frequency to monthly would have very minimal impact on the reliability of the scram system.

The licensee stated in its application that they have reviewed the basis for the staff's acceptance of TSTF-475, Revision 1, and concluded that the basis is applicable to Susquehanna Steam Electric Station, Units 1 &2's Operating Licenses, and supports their adoption of the TSTF-475 changes into its both units TS. The NRC staff also reviewed the TSTF-475, Revision 1 basis, and similarly concluded that the basis for the TSTF is applicable to Susquehanna Steam Electric Station, Units 1 &2's Operating Licenses, and therefore, the TSTF is appropriate for adoption by the licensee. In addition, the NRC staff reviewed the licensee's proposed changes against the corresponding changes made to the STS by TSTF-475, revision 1, which the staff has found to satisfy applicable regulatory requirements, as described above. The proposed changes would: (1) revise the Technical Specifications (TS) control rod notch surveillance frequency in TS 3.1.3, "Control Rod OPERABILITY," and (2) revise one Example in Section 1.4 "Frequency" to clarify the applicability of the 1.25 surveillance test interval extension. The staff found that the proposed changes are consistent with the changes approved by the staff in TSTF-475, Revision 1. The NRC staff, therefore, finds these changes acceptable.

The !\\IRC staff has reviewed the licensee's proposal to amend existing Susquehanna Steam Electric Station, Units 1 & 2's TS Sections SR 3.1.3 "Control Rod OPERABILITY," and Example 1.4-3, "Frequency" applicable to SR 3.0.2. The NRC staff has concluded that the TS revisions will have a minimal effect on the high reliability of the CRD system while reducing the opportunity for potential reactivity events; thus, meeting the requirement of

- 7 10 CFR Part 50, Appendix A, GDC 29, and will clarify the applicability of the 1.25 provision in SR 3.0.2. Therefore, the staff concludes that the amendment request is acceptable.

4.0 STATE CONSULTATION

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

5.0 ENVIRONMENTAL CONSIDERATION

The amendment changes a requirement with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20 and changes to SRs. The NRC staff has determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendment involves no significant hazards consideration, and there has been no public comment on such finding (73 FR 58675). Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b) no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendment.

6.0 CONCLUSION

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

7.0 REFERENCES

1.

Letter TSTF-04-07 from the Technical Specifications Task Force to the NRC, TSTF 475 Revision 0, "Control Rod Notch Testing Frequency and SRM Insert Control Rod Action," August 30, 2004, Agencywide Documents and Management System (ADAMS) Accession No. ML042520035.

2.

NUREG-1430, "Standard Technical Specifications Babcock and Wilcox Plants, Revision," August 31,2003.

3.

NUREG-1431, "Standard Technical Specifications Westinghouse Plants, Revision 3,"

August 31, 2003.

4.

NUREG-1432, "Standard Technical Specifications Combustion Engineering Plants, Revision 3," August 31,2003.

5.

NUREG-1433, "Standard Technical Specifications General Electric Plants, BWR/4, Revision 3," August 31,2003.

- 8

6.

NUREG-1434, "Standard Technical Specifications General Electric Plants, BWR/6, Revision 3," August 31,2003.

7.

Letter TSTF-07-19, Response from the Technical Specifications Task Force to the NRC, "Request for Additional Information (RAI) Regarding TSTF-475 Revision 0, "Control Rod Notch Testing Frequency and SRM Insert Control Rod Action," dated February 28,2007, (TSTF-475 Revision 1 is an enclosure), ADAMS Accession 1\\10.

ML071420428.

8.

Letter TSTF-06-13 from the Technical Specifications Task Force to the NRC, "Response to NRC Request for Additional Information Regarding TSTF-475, Revision 0," dated July 3,2006, ADAMS Accession No. ML0618403421.

9.

Letter BWROG-06036 from the BWR Owners Group to the NRC, "Response to NRC Request for Additional Information Regarding TSTF-475, Revision 0," dated November 16, 2006, with Enclosure of the GE Nuclear Energy Report, "CRD Notching Surveillance Testing for Limerick Generating Station," dated November 2006, ADAMS Accession No. ML063250258.

10.

Letter TSTF-07 -19 from the Technical Specifications Task Force to the NRC, "Response to NRC Request for Additional Information Regarding TSTF-475, Revision 0," dated May 22,2007, ADAMS Accession No. ML071420428.

11.

PPL Susquehanna, LLC (the licensee) license amendment request, dated JUly 7, 2008, ADAMS Accession No. IVIL 082040624.

Principal Contributor: R. Grover Date: January 2, 2009

W. H. Spence

-2 A copy of our safety evaluation is also enclosed. Notice of Issuance will be included in the Commission's next regular Biweekly Federal Register Notice.

Sincerely,

/raJ Bhalchandra K. Vaidya, Project Manager Plant Licensing Branch 1-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-387 and 50-388

Enclosures:

1. Amendment No. 250 to License No. NPF-14

2. Amendment No. 229 to License No. NPF-22

3. Safety Evaluation cc w/encls: Distribution via ListServe DISTRIBUTION PUBLIC PDI-1 RF RidsNrrDorlLPL1-1 RidsNrrPMBVaidya(hard copy)

RidsNrrLASLittle (hard copy) RidsOGCRp RidsNrrltsb..... RidsNrrDorIDpr RidsAcrsAcnw&mMailCenter RidsRg 1MailCenter ADAMS Accession No.: ML083520395

(*) No substantial changes to SE Input Memo OFFICE LPLI-1/PM LPLI-1/LA NRR/ITSB/BC(*)

OGC LPLI-1/BC NAME BVaidya SLittie RElliot LSubin MKowal (DPickett for)

DATE 12/22/08 12/18/08 12/11/08 12/24/08 12/30/08 OFFICIAL RECORD COpy

v t e Letter Sequence Approval
TAC:MD9303, Control Rod Notch Testing Frequency and SRM Insert Control Rod Action (Approved, Closed) TAC:MD9304, Control Rod Notch Testing Frequency and SRM Insert Control Rod Action (Approved, Closed)
Initiation Request Acceptance Administration Questions Answers Results Approval Other:
MONTHYEAR2008-08-13 [Table View]

ML083520395

Text

SUBJECT:

Dear Mr. Spence:

Enclosures:

Attachment:

Attachment:

1.0 INTRODUCTION

1.1 Background

2.0 REGULATORY EVALUATION

3.0 TECHNICAL EVALUATION

4.0 STATE CONSULTATION

5.0 ENVIRONMENTAL CONSIDERATION

6.0 CONCLUSION

7.0 REFERENCES

Enclosures:

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